A Review of Radionuclide Release from HTGR Cores During Normal Operation WARNING: Please read the Export Control Agreement on the back cover.

Transcription

1 A Review of Radionuclide Release from HTGR Cores During Normal Operation WARNING: Please read the Export Control Agreement on the back cover. Technical Report

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3 A Review of Radionuclide Release From HTGR Cores During Normal Operation Final Report, February 2004 EPRI Project Manager L. Sandell EPRI 3412 Hillview Avenue, Palo Alto, California PO Box 10412, Palo Alto, California USA

4 DISCLAIMER OF WARRANTIES AND LIMITATION OF LIABILITIES THIS DOCUMENT WAS PREPARED BY THE ORGANIZATION(S) NAMED BELOW AS AN ACCOUNT OF WORK SPONSORED OR COSPONSORED BY THE ELECTRIC POWER RESEARCH INSTITUTE, INC. (EPRI). NEITHER EPRI, ANY MEMBER OF EPRI, ANY COSPONSOR, THE ORGANIZATION(S) BELOW, NOR ANY PERSON ACTING ON BEHALF OF ANY OF THEM: (A) MAKES ANY WARRANTY OR REPRESENTATION WHATSOEVER, EXPRESS OR IMPLIED, (I) WITH RESPECT TO THE USE OF ANY INFORMATION, APPARATUS, METHOD, PROCESS, OR SIMILAR ITEM DISCLOSED IN THIS DOCUMENT, INCLUDING MERCHANTABILITY AND FITNESS FOR A PARTICULAR PURPOSE, OR (II) THAT SUCH USE DOES NOT INFRINGE ON OR INTERFERE WITH PRIVATELY OWNED RIGHTS, INCLUDING ANY PARTY'S INTELLECTUAL PROPERTY, OR (III) THAT THIS DOCUMENT IS SUITABLE TO ANY PARTICULAR USER'S CIRCUMSTANCE; OR (B) ASSUMES RESPONSIBILITY FOR ANY DAMAGES OR OTHER LIABILITY WHATSOEVER (INCLUDING ANY CONSEQUENTIAL DAMAGES, EVEN IF EPRI OR ANY EPRI REPRESENTATIVE HAS BEEN ADVISED OF THE POSSIBILITY OF SUCH DAMAGES) RESULTING FROM YOUR SELECTION OR USE OF THIS DOCUMENT OR ANY INFORMATION, APPARATUS, METHOD, PROCESS, OR SIMILAR ITEM DISCLOSED IN THIS DOCUMENT. ORGANIZATION(S) THAT PREPARED THIS DOCUMENT General Atomics ORDERING INFORMATION Requests for copies of this report should be directed to EPRI Orders and Conferences, 1355 Willow Way, Suite 278, Concord, CA 94520, (800) , press 2 or internally x5379, (925) , (925) (fax). Electric Power Research Institute and EPRI are registered service marks of the Electric Power Research Institute, Inc. EPRI. ELECTRIFY THE WORLD is a service mark of the Electric Power Research Institute, Inc. Copyright 2004 Electric Power Research Institute, Inc. All rights reserved.

5 CITATIONS This report was prepared by General Atomics 3550 General Atomics Court San Diego, CA Principal Investigator D. Hanson This report describes research sponsored by EPRI. The report is a corporate document that should be cited in the literature in the following manner: A Review of Radionuclide Release From HTGR Cores During Normal Operation, EPRI, Palo Alto, CA: iii

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7 PRODUCT DESCRIPTION The release of radionuclides from the core of high-temperature gas-cooled reactors (HTGRs) especially direct-cycle HTGRs during normal plant operation has significant design, O&M, and safety implications. A hallmark philosophy of all modern HTGRs is to design the plant so that radionuclides are retained in the core during normal operation and postulated accidents. The key to achieving this safety goal is twofold: (1) a reliance on ceramic-coated fuel particles for primary fission product containment at their source and (2) passive cooling to assure that the integrity of the coated particles is maintained even if the normal cooling systems were permanently disrupted. This report reviews and evaluates the existing database of available international information on radionuclide release phenomena, specifically as it applies to the gas turbine-modular helium reactor (GT-MHR) and to the pebble bed modular reactor (PBMR). Results & Findings Written from the perspective of an HTGR core designer, the report describes design methods for predicting fuel performance and fission product release from the core and summarizes the international fuel performance and fission product release databases. It also discusses the comprehensive core performance analysis performed for the plutonium consumption-modular helium reactor (PC-MHR). Additional technology developments required to reduce the large uncertainties in core release predictions are summarized. Challenges & Objectives Design methods have been developed to model coated-particle fuel performance and radionuclide transport in core materials during normal plant operation and postulated accident conditions. These design methods have been used extensively for core design and safety analysis for both prismatic and pebble-bed HTGRs. However, these methods are not yet fully validated, in part, because of the remaining uncertainties in the phenomena that control coated-particle fuel performance and radionuclide transport in HTGR cores. Consequently, additional technology development will be necessary before the radionuclide source terms are validated. To establish what additional technology development is required, it is first necessary to evaluate the current state-of-the-art and the current database on which it is predicated. v

8 Applications, Values & Use There is a renewed international interest in advanced HTGR designs to help resolve key national and international energy and environmental issues. For example, the helium-cooled very high temperature reactor (VHTR) is the nearest-term system capable of producing nuclear hydrogen and/or high-efficiency electricity. A direct-cycle GT-MHR is being developed under a joint U.S./Russian program for the purpose of destroying surplus Russian weapons plutonium, and a South Africa-based PBMR project continues to progress toward construction of a commercial prototype. All these projects depend on the capability of predicting fuel performance and fission product release from HTGR cores during normal operation. EPRI Perspective A concerted effort to improve the capability of reactor designers to reliably predict fuel performance and fission product release from HTGR cores during normal operation will be required to support future plant design and deployment. Since the models and material property data used by the various international design and technology organizations differ significantly in certain regards, a series of benchmark problems leading to quantitative comparisons of results would be beneficial. In the course of defining these benchmark problems, the existing database would need to be reexamined to determine if further validation efforts are warranted. An International Atomic Energy Agency (IAEA) Coordinated Research Project on Advances in HTGR Fuel Technology is in process to address these needs. Approach Investigators began their research by describing the radionuclide release issue for advanced HTGRs. They reviewed design methods used for predicting coated-particle fuel performance and the attendant radionuclide release from HTGR cores. Next, they acquired, evaluated, and documented the existing experimental database to assess the validity of design methods and radionuclide source terms for normal operation. The next major step was to predict fuel performance and radionuclide release from the core of direct-cycle HTGRs. Finally, they identified further technology development appropriate to complete validation of design methods and plateout source terms. Keywords High-temperature gas-cooled reactors Coated fuel particles Radionuclide release O&M dose vi

9 ABSTRACT There is a renewed international interest in advanced high-temperature gas-cooled reactor (HTGR) designs to help resolve key national and international energy and environmental issues. For example, a direct-cycle gas turbine-modular helium reactor (GT-MHR) is being developed under a joint U.S./Russian program for the purpose of destroying surplus Russian weapons plutonium, and a South Africa-based pebble bed modular reactor (PBMR) project continues to progress toward construction of a commercial prototype. All these projects depend on the capability of predicting fuel performance and fission product release from HTGR cores during normal operation. A hallmark philosophy of all modern HTGRs is to design the plant so that radionuclides are retained in the core during normal operation and postulated accidents. Written from the perspective of an HTGR core designer, this report describes design methods for predicting fuel performance and fission product release from the core and summarizes the international fuel performance and fission product release databases. It also discusses the comprehensive core performance analysis performed for the plutonium consumption-modular helium reactor (PC-MHR). Additional technology developments required to reduce the large uncertainties in core release predictions are summarized. vii

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11 ACRONYMS AND ABBREVIATIONS Acronym AGR AVR BAF o BISO CEA CG/TG Ci CRP DB-MHR DDN DOE EAB EFPD EOL EPA EPZ EPRI FDDM/F Definition Advanced Gas Reactor [in this report, AGR refers to Advanced HTGRs and not to the British CO 2 -cooled Advanced Gas Reactor] Arbeitsgemeinschaft Versuchs Reactor [German pebble-bed HTR] Bacon Anisotropy Factor Bi-ISOtropic coated-fuel particle design with three materials in coating system (low-density PyC, high-density PyC) Commissariat à L'Energie Atomique [French Atomic Energy Commission] coating gas/total gas Curie(s) Coordinated Research Program [IAEA terminology] Deep-Burn Modular Helium Reactor Design Data Need [US] Department of Energy Exclusion Area Boundary effective full power days end-of-life [US] Environmental Protection Agency Emergency Planning Zone Electric Power Research Institute Fuel Design Data Manual, Issue F [GA Proprietary Information] ix

12 Acronym Definition FRG FSV FTE FZ Juelich GA GCR Gen IV GT-MHR HEU HFIR HHT HM HPS HTGR HTI HTTR HTR-10 IAEA IMGA INEEL INET Federal Republic of Germany Fort St. Vrain [US prismatic HTGR] fuel test element Forschungszentrum Juelich [German national laboratory; formerly KFA] General Atomics gas-cooled reactor Generation IV Gas Turbine-Modular Helium Reactor high enriched uranium [typically 93% U-235] High Flux Isotope Reactor Hochtemperaturreaktor mit Helium Turbine [German direct-cycle HTR] heavy-metal helium purification system High Temperature Gas-Cooled Reactor 1 [prismatic core] High Temperature Isotropic [PyC coatings] High Temperature Test Reactor [Japanese HTGR] High Temperature Reactor [Chinese HTR] International Atomic Energy Agency Irradiated Microsphere Gamma Analyzer Idaho National Engineering and Environmental Laboratory Institute of Nuclear Energy Technology [Tsinghua University, Beijing, China] 1 In this report, HTGR is used generically to represent all high temperature, helium-cooled reactors with coatedparticle fuel, regardless of core configuration, fuel-element type, or power conversion cycle; in contrast, HTR refers specifically to HTGRs with pebble-bed cores. x

13 Acronym Definition ISI KFA LEU LBE LTI LWR MHTGR MHR MINATOM MOX MTR MTS NFI NGNP NMAC NP-MHTGR NPP NQA NRC O&M OKBM ORNL PAG PB in-service inspection Kernforschungsanlage Juelich [former German national laboratory, now FZJ] low enriched uranium [<20% U-235] Licensing Basis Event Low Temperature Isotropic [PyC coatings] light water reactor Modular High Temperature Gas-Cooled Reactor [steam-cycle HTGR] Modular Helium Reactor [Russian Ministry of Atomic Energy] mixed oxide [(U,Pu)O 2 ±x] materials test reactor (MTR) methyl trichlorosilane Nuclear Fuel Industries [Japanese fuel manufacturer] Next Generation Nuclear Plant [EPRI] Nuclear Maintenance Application Center New Production-Modular High Temperature Gas-Cooled Reactor nuclear power plant Nuclear Quality Assurance [ASME/ANSI QA protocol for nuclear activities] [US] Nuclear Regulatory Commission operations & maintenance [Experimental Bureau of Mechanical Engineering, Russian] Oak Ridge National Laboratory [EPA] Protective Action Guides Peach Bottom Unit 1 [prototype HTGR] xi

14 Acronym PBMR PC-MHR PCRV PCS PIE PSID QA QC R/B RCCS RF RN TECDOC THTR TRISO V&V VHTR VLPC WPu Definition Pebble Bed Modular Reactor [South African HTR] Plutonium Consumption-Modular Helium Reactor prestressed concrete reactor vessel power conversion system postirradiation examination Preliminary Safety Information Document quality assurance quality control release rate-to-birth rate ratio reactor cavity cooling system Russian Federation radionuclide [IAEA technical report nomenclature] Thorium Hochtemperatur Reaktor [German pebble-bed HTR] TRi-ISOtropic coated-fuel particle design with three materials in coating system (low-density PyC, high-density PyC, and SiC) verification and validation [of design methods] Very High Temperature Reactor vented low-pressure containment weapons [grade] plutonium [>90% Pu-239] xii

15 EXECUTIVE SUMMARY The Electric Power Research Institute has contracted with General Atomics to acquire, evaluate and summarize the existing data base on the radionuclide release from the cores of High Temperature Gas-Cooled Reactors under normal operating conditions. This report should complement an earlier EPRI-sponsored review report on the data base on the deposition ( plateout ) of condensable radionuclides in the primary coolant circuits of HTGRs. The release of radionuclides, especially silver-110m and cesium-134/-137, from the core during normal operation is of particular importance to direct-cycle, gas-turbine HTGR designs, such as the Gas Turbine-Modular Helium Reactor and the Pebble Bed Modular Reactor, because of their impact on O&M dose rates. In addition, radionuclides that accumulate in the primary coolant circuit during normal operation, especially I-131, can become important source terms for postulated accidents. Moreover, there is a growing interest in using HTGRs for producing process heat for high-temperature energy-intensive processes, including hydrogen production by thermochemical water splitting or high-temperature electrolysis; for such applications, core outlet temperatures of 1000 o C and above are thermodynamically attractive. For such very high temperature designs, fission product release from the core may become a critical issue. The intention is that this review be supportive of both prismatic- and pebble-bed core designs. A hallmark philosophy of all modern HTGRs is to design the plant such that the radionuclides would be essentially retained in the core during normal operation and postulated accidents. The key to achieving this safety goal is the reliance on ceramic-coated fuel particles for primary fission product containment at their source, along with passive cooling to assure that the integrity of the coated particles is maintained even if the normal cooling systems were permanently disrupted. Consequently, these designs mandate the development and qualification of coatedparticle fuels that meet stringent requirements for as-manufactured quality and in-service coating integrity for both normal operation and postulated accident conditions. Coated-particle fuel for use in both prismatic and pebble-bed HTGRs has been under development for the past four decades. The various HTGR fuel development programs have performed hundreds of irradiation tests with a large variety of coated-particle designs and conducted reactor surveillance programs to assess fuel performance in operating HTGRs. Initial fuel development focused on pyrocarbon-coated carbide particles, and such particles were mass produced and used in the early prototype HTGRs (Dragon, AVR, and Peach Bottom 1) and in the THTR. However, the superior fission product retention capability of the so-called TRISO particle - because of the addition of a SiC layer - was recognized early (first serving as the driver fuel in Dragon reload cores and then mass produced for FSV and for AVR reloads), and its development and qualification became the emphasis by the late 1970s. xiii

16 TRISO particles consist of multiple ceramic coatings of silicon carbide and pyrolytic carbon surrounding dense microspheres ( kernels ) containing the fissile and fertile materials which constitute the nuclear fuel. The TRISO coating system functions as a multi-shell pressure vessel and as a diffusion barrier to retain fission products. The coatings are capable of maintaining their integrity and retaining fission products at temperatures higher than those imposed during normal operation and postulated accident conditions. The fuel particles are bonded together by a carbonaceous matrix and formed into cylindrical compacts that are contained in vertical holes in graphite fuel blocks for use in prismatic cores, or they are formed into a spherical compact encapsulated by a spherical shell of unfueled matrix material for use in pebble-bed cores. The ultimate figure-of-merit for judging fuel performance is the ability of the coated particle to retain fission products. Typically, the dominant sources of fission product release from the core are as-manufactured, heavy-metal contamination (i.e., leachable heavy metal outside of coated particles) and particles whose coatings fail in service. In addition, the volatile metals (e.g., Cs, Ag, Sr) can, at sufficiently high temperatures for sufficiently long times, diffuse through the SiC coating and be released from intact TRISO particles; however, diffusive release from intact particles is only significant compared to other sources for silver release. The important mechanisms that can cause coating failure under irradiation have been identified as a result of irradiation testing and postirradiation examination. Both structural/mechanical mechanisms and thermochemical mechanisms have been observed, and phenomenological models have been developed to predict them for reactor design and safety analysis. Analytical modeling, including structural analysis, in combination with experimental testing have identified practical methods for eliminating, or at least controlling to acceptable levels, these particle failure mechanisms. The TRISO particle is the critical part of a multiple-barrier radionuclide containment system for an HTGR, which reflects a defense-in-depth philosophy. While the particle coatings are the most effective release barrier in the containment system, credit is also taken for fission product retention in the fuel kernels, the fuel matrix, and the fuel-element graphite (in the case of a prismatic core). Consequently, fission product transport in each of these core materials must be characterized experimentally and modeled as part of the core design. An extensive international data base has been assembled from laboratory measurements, in-pile experiments, and reactor surveillance programs to provide the basis for deriving the material properties and models necessary to describe fission product transport in core materials. Obviously, the prediction of fuel performance (i.e., coating failure rates) is a prerequisite to predicting fission product release from the core, and the reactor designer and safety analyst must do both. Design methods have been developed internationally during the past four decades to model coated-particle fuel performance and radionuclide transport in core materials during normal plant operation and postulated accident conditions. These design methods have been used extensively for core design and safety analysis for both prismatic and pebble-bed HTGRs. An overview is given of these methods for normal operation, and an example of their application to predict the fuel performance and fission product release during normal operation from a prismatic core with an 850 o C outlet temperature is given herein. These results suggest that a properly optimized prismatic core design with a 850 o C outlet temperature should be capable of meeting the provisional radionuclide control criteria for normal operation also described herein. xiv

17 Producing a VHTR fuel and core design with a 1000 o C outlet temperature to meet these same criteria will be more challenging. Integral radionuclide release data are available from irradiation capsules, in-pile loops and operating HTGRs. The validity of the design methods for predicting fuel performance and fission product release during normal operation has been assessed by evaluating their ability to reproduce these integral release data. While there have been some notable successes, especially the prediction of fission product release from the Fort St. Vrain HTGR core, the discrepancies between the predictions and the measurements are often larger than the current design margins to accommodate such uncertainties in the design methods. A concerted effort to improve the capability of the reactor designer to reliably predict fuel performance and fission product release from HTGR cores during normal operation will be required to support future plant design and deployment. Since the models and material property data used by the various international design and technology organizations differ significantly in certain regards, a series of benchmark problems leading to quantitative comparisons of results would be beneficial. In the course of defining these benchmark problems, the existing data base would need to be reexamined to determine if further validation efforts are warranted. An IAEA Coordinated Research Project on Advances in HTGR Fuel Technology is in process to address these needs. While new insight would result from further examination of the existing data base, significant additional testing will be required to complete the development and qualification of reliable predictive methods for calculating fuel performance and fission product release, especially under very high-temperature conditions. The requisite Design Data Needs and the attendant testing programs to improve the predictive methods have been identified for the commercial GT-MHR, and are summarized herein. Developers of HTGRs for gas turbine power generation and high temperature process heat applications are encouraged to review these DDNs for relevance to specific designs. Since many of these issues are largely generic and the requisite test programs are technically challenging and expensive, there is considerable opportunity for international cooperation. Efforts are under way through the US DOE International Nuclear Energy Research Initiative, the US DOE initiated Generation IV International Forum, and the IAEA CRP noted above to foster international cooperation. xv

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19 CONTENTS 1 INTRODUCTION AND BACKGROUND Purpose and Scope of Review Current Status of Modular HTGR Development GT-MHR Program Programmatic Design Overview Manufacture of HTGR Fuel Manufacturing Technology Fuel Kernels Coated Particles Fuel Elements Prismatic Fuel Elements Spherical Fuel Elements Manufacturing Experience Report Organization RADIONUCLIDE RELEASE FROM HTGR CORES Radionuclide Control Philosophy Radionuclide Containment System Phenomena Controlling Release As-Manufactured Fuel Attributes Kernel Composition U-C-O Phase Equilibria Carbon Monoxide Generation Fuel Failure Mechanisms Structural/Mechanical Mechanisms Pyrocarbon Performance Pressure-Vessel Failure xvii

20 OPyC Irradiation-Induced Failure IPyC Irradiation-Induced Failure Thermochemical Mechanisms Kernel Migration Chemical Attack of SiC Thermal Decomposition of the SiC Layer Heavy-Metal Dispersion Radionuclide Transport Mechanisms Radionuclide Transport in Fuel Kernels Fission Gas Release from Fuel Kernels Fission Metal Release from Fuel Kernels Radionuclide Transport in Particle Coatings Radionuclide Transport in Fuel-Compact Matrix Radionuclide Transport in Fuel-Element Graphite Radionuclide Control Requirements Methodology Radionuclide Design Criteria Fuel Design Criteria for Advanced HTGRs Provisional Fuel Requirements Provisional Fuel Product Specifications DESIGN METHODS FOR PREDICTING RADIONUCLIDE RELEASE Phenomenological Models and Computer Codes U.S. Computer Codes Particle Analysis Codes Core Performance Codes Support Codes German Computer Codes Other Foreign Computer Codes Phenomenological Models and Material Property Data Particle Material Properties Fuel Performance Models Fission Product Transport Models/Correlations FUEL PERFORMANCE DATA BASE xviii

21 4.1 Fuel Irradiation Capsules Conventional TRISO-Coated Particles HEU Irradiation Tests LEU Irradiation Tests U.S. LEU Program German LEU Program Non-Conventional TRISO-Coated Particles TRISO-P Particles TRISO-Coated UO * 2 Particles Barrier Coatings Key Differences between U.S. and FRG TRISO Particles Fuel Test Elements Peach Bottom Fuel Test Elements HEU/Th Test Elements Pu Test Element Fort St. Vrain Fuel Test Elements HTGR Operating Experience Dragon HTGR Peach Bottom HTGR Fort St. Vrain HTGR AVR HTR THTR HTR High Temperature Test Reactor High Temperature Reactor Conclusions and Implications RADIONUCLIDE RELEASE DATA BASE Material Property Data Transport in Kernels Fission Gas Release Fission Metal Release Transport in Coatings Transport in Matrix/Graphite Integral Radionuclide Release Data Irradiation Capsules xix

22 R2 K TRISO-P Capsules In-Pile Loops CPL-2 Test Program COMEDIE BD-1 Test Program Operating HTGRs Peach Bottom Core Fort St. Vrain Fission Gas Release Fission Metal Release AVR HTR THTR HTR Conclusions and Implications PREDICTED RADIONUCLIDE RELEASE FROM DIRECT-CYCLE HTGR CORES PC-MHR Description Methodology and Assumptions Fuel Performance Models for Plutonium Oxide Fuel Pressure-Vessel Failure OPyC Irradiation-Induced Failure SiC Corrosion Failure SiC Thermal Decomposition Failure Kernel Migration Failure Heavy-Metal Dispersion Failure Fission Gas Release Fission Metal Release Core Performance Analysis Fuel and Graphite Temperature Predictions Fast Fluence and Burnup Distributions Fuel Particle Failure Fission Gas Release Fission Metal Release Cs-137 Release Ag-110m Release Sr-90 Release xx

23 6.4 Summary and Implications REQUIRED TECHNOLOGY DEVELOPMENT GT-MHR Design Data Needs Required Test Programs for the GT-MHR International Development Opportunities CONCLUSIONS Design Implications Design Methods for Radionuclide Release Experimental Data Base REFERENCES xxi

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25 LIST OF FIGURES Figure 1-1 VHTR Demonstration Module Figure 1-2 NGNP Master Schedule Figure 1-3 GT-MHR Module Figure 1-4 Annular Core Figure 1-5 Prismatic Fuel-Element Components Figure 1-6 Functions of the Individual Coating Layers Figure 1-7 GT-MHR Power Conversion System Figure 1-8 Spherical Fuel-Element Components Figure 1-9 Process Flow Diagram for Prismatic Fuel Elements Figure 1-10 Process Flow Diagram for Sphere Making Figure 2-1 MHR Radionuclide Containment System Figure 2-2 OPyC Failure versus Material Properties Figure 2-3 OPyC Failure vs. Process Variables Figure 2-4 SiC Material Properties vs. Process Variables Figure 2-5 Effect of Ar Diluent on SiC Microstructure Figure 2-6 Effect of Ar Diluent on SiC Material Properties Figure K Isotherm of the U-C-O System Figure 2-8 Oxygen Potential vs. Temperature for Various Oxide-Carbide Equilibria Figure 2-9 Phases Present below 1800 K for HEU UCO Kernel Figure 2-10 TRISO Particle Failure Mechanisms Figure 2-11 Pressure Vessel of UO 2 Particle in HRB-8 Irradiation Figure 2-12 OPyC Failure from Matrix-Coating Interaction Figure 2-13 Radial Crack in IPyC Layer of TRISO-P Particle Figure 2-14 Kernel Migration in High-Burnup Oxide Fuels Figure 2-15 SiC Corrosion by Fission Products Figure 2-16 CO Corrosion of SiC Co Figure 2-17 Effect of High Temperature Heating on TRISO Particle Figure 2-18 Principal Steps in Radionuclide Release from an HTGR Core Figure 2-19 Isotopic R/Bs vs. Radioactive Half Life Figure 2-20 Temperature Dependence of Fission Gas Release Figure 2-21 Effect of Hydrolysis on Fission Gas Release Figure 2-22 Fission Metal Release from Individual Fuel Particles xxiii

26 Figure 2-23 Cesium Transport in R2 K13 Capsule Figure 2-24 Logic for Derivation of Fuel Quality Requirements Figure 2-25 Radionuclide Design Criteria Figure 4-1 Fission Gas Release from FRG Capsule FRJ2-K Figure 4-2 Fission Gas Release versus Fast Fluence for TRISO-P Capsules Figure 4-3 HTTR Fuel Element Design Figure 5-1 Effect of Temperature on Half-Life Dependence for Gas Release Figure 5-2 Reduced Diffusion Coefficients for Cs Transport in ThO 2 Kernels Figure 5-3 Postirradiation Heating Data for UO 2 at 1700 C Figure 5-4 Cs-in-PyC Diffusion Coefficients Figure 5-5 Cs-in-SiC Diffusion Coefficients Figure 5-6 Ag-in-SiC Diffusion Coefficients Figure 5-7 Ag Diffusion Coefficients for German A3 Matrix Figure 5-8 Cs Diffusivities for Nuclear Graphites Figure 5-9 Cs Sorption on Irradiated H 451 Graphite Figure 5-10 Kr-85m R/B versus Time for Capsule r2-k Figure 5-11 Comparison of Measured and Predicted Gas Release for R2-K13/Cell Figure 5-12 Ag Release from Capsule R2-K Figure 5-13 Comparison of Measured and Predicted Gas Release in COMEDIE BD Figure 5-14 Peach Bottom Fuel Element Figure 5-15 Comparison of Measured and Calculated Gas Release from PB Core Figure 5-16 Cs Redistribution within Peach Bottom Core 2 Fuel Elements Figure 5-17 Cs-137 Distribution in Fuel Element E Figure 5-18 Predicted and Measured Kr 85m R/B in FSV Figure 5-19 AVR Coolant Temperature and Fission Gas Release from Figure 5-20 THTR Coolant Gas Activity Figure 6-1 PC-MHR Module Figure 6-2 PC-MHR Annular Core Figure 6-3 PC-MHR Fuel Components Figure 6-4 PuO 1.68 Particle Irradiated to 70% FIMA in FTE Figure 6-5 R/B Multiplier for High-Burnup Fuels Figure 6-6 Peak Fuel Temperature Volume Distribution (Segment 1) Figure 6-7 Fuel Temperature History for Peak Temperature Location Figure 6-8 Location with Persistently High Fuel Temperature Figure 6-9 Time-Average Fuel Temperature Volume Distribution (Segment 1) Figure 6-10 Peak Graphite Temperature Volume Distribution (Segment 1) Figure 6-11 Time-Average Graphite Temperature Volume Distribution (Segment 1) Figure 6-12 Burnup Volume Distribution (Segment 1) xxiv

27 Figure 6-13 Fast Fluence Volume Distribution (Segment 1) Figure 6-14 Exposed Kernel Fraction vs. Time Figure 6-15 SiC Failure Fraction vs. Time Figure 6-16 Core-Average R/B for Kr Figure 6-17 Core-Average R/B for I Figure 6-18 Cs-137 Fractional Release vs. Time Figure 6-19 Ag-110m Fractional Release vs. Time Figure 7-1 Process for Identifying DDNs xxv

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29 LIST OF TABLES Table 1-1 GT-MHR Key Design Parameters Table 1-2 Production of Coated-Particle Fuel Table 2-1 Key Features of Candidate Kernel Compositions Table 2-2 Classes of Radionuclides of Interest for HTGR Design Table 2-3 GT-MHR Provisional Fuel Requirements Table 2-4 GT-MHR Provisional Fission Metal Release Limits Table 2-5 Provisional Fuel Service Conditions for Advanced HTGRs Table 2-6 GT-MHR Coated Particle Nominal Design Parameters Table 2-7 GT-MHR Fuel Compact Nominal Design Parameters Table 4-1 HEU Irradiation Capsules Table 4-2 LEU Irradiation Capsules Table 4-3 German High Quality LEU UO 2 Irradiation Summary Table 4-4 TRISO-P Irradiation Capsules Table 4-5 Peach Bottom Fuel Test Elements Table 4-6 Fort St. Vrain Fuel Test Elements Table 4-7 Specifications for FSV Fuel Particles Table 4-8 Specifications for FSV Fuel Compacts Table 4-9 AVR Fuel Reloads in the Years 1966 to Table 5-1 Comparison of Measured and Calculated Metal Release in R2-K Table 5-2 Comparison of Measured and Calculated FSV Metal Release Table 5-3 Specific Activities in AVR Primary Coolant Table 6-1 PC-MHR Key Design Parameters Table 6-2 Comparison of PC-MHR and GT-MHR Particle Designs Table 6-3 Comparison of PC-MHR and GT-MHR Fuel Requirements Table 6-4 PC-MHR Performance Analysis Results Table 7-1 GT-MHR Fuel/Fission Product DDNs Table 7-2 Workscope for Release-Related DDNs Table 7-3 FP Release-related DDN/Test Program Matrix xxvii

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31 1 INTRODUCTION AND BACKGROUND This report is a review - and an evaluation - of the available international information on the radionuclide release from the cores of High Temperature Gas-Cooled Reactors (HTGRs) during normal plant operation with emphasis on direct-cycle, gas-turbine plant designs. 1.1 Purpose and Scope of Review The Electric Power Research Institute (EPRI) has contracted with General Atomics (GA) to acquire, evaluate and summarize the existing international data base on the radionuclide release from HTGR cores during normal operation. 2 The scope of work for this review follows: 1. Description of the radionuclide release issue for HTGRs. 2. Review of design methods used for predicting release. 3. Acquisition, evaluation, and documentation of the existing experimental database to validate design methods and predicted core release rates. 4. Predicted core release rates (with emphasis on the direct-cycle PC-MHR for which the most information is currently available). 5. Summary description of further technology development needed ( Design Data Needs, DDNs) to complete validation of the design methods and predicted core release rates. This radionuclide release review is a sequel to a review of plateout phenomena in HTGRs (Ref. 1-1) that was prepared by GA for EPRI in Because the topics of radionuclide release from the core and plateout in the primary circuit are intimately related, there is a certain amount of overlap between the two documents (i.e., the core release is the plateout source term). While both reviews are intended to be stand-alone documents, the interested reader is encouraged to consult the plateout review as well. The intention is that this review be supportive of the prismatic-core Gas Turbine-Modular Helium Reactor (GT-MHR), the Pebble Bed Modular Reactor (PBMR), and any other future HTGR projects. Hence, it emphasizes the generic aspects of the core release issue (e.g., release from fuel particles), but potential differences between designs with prismatic and pebble-bed cores are also briefly addressed. Since GA is an active participant in the USDOE/MINATOM International GT-MHR program and a long term developer of prismatic core designs, more design-specific information was available for prismatic designs than for pebble bed designs. 2 Research and Development Agreement Number EP-P58034/C4049 (GA 30138). 1-1

32 Introduction and Background 1.2 Current Status of Modular HTGR Development There is a renewed international interest in Advanced Gas Reactor (AGR) designs, 3 based upon HTGR technology, to contribute to the resolution of key national and international issues. For example, among the Generation IV (Gen-IV) concepts, the helium-cooled Very High Temperature Reactor (VHTR) is the nearest-term system (estimated to be deployable before 2020) capable of producing nuclear hydrogen and/or high-efficiency electricity (Ref. 1-2). The direct-cycle GT-MHR is already being developed under a joint USDOE/MINATOM program for the purpose of destroying surplus Russian weapons plutonium (Ref. 1-3). Moreover, the GT- MHR with a modified core design is also being evaluated as an efficient burner of transuranic materials. The primary benefit of the so-called Deep-Burn MHR (DB-MHR) would be to significantly reduce the long-term storage requirements for high-level waste generated by the currently operating light-water reactors around the world (Ref. 1-4). The South Africa-based PBMR project (e.g., Ref. 1-5) to construct a prototype of a commercial pebble-bed modular HTGR is also in progress. In June 2003, the PBMR project received a positive Record of Decision on the environmental impact assessment study by the South African Department of Environmental Affairs and Tourism which is a major step towards the completion of the detailed feasibility phase of the project. The PBMR is a modular, direct-cycle, pebble-bed High Temperature Reactor (HTR) based upon the pebble-bed reactor technology successfully developed and demonstrated in the Federal Republic of Germany (FRG). The intention is to build a 160 MW(e) demonstration module at Koeberg, near Cape Town. The USDOE Generation IV project has identified reactor system concepts for producing electricity, which excelled at meeting Generation IV goals related to safety, sustainability, proliferation resistance and physical security, and economics. One of these reactor system concepts, the VHTR, is also uniquely suited for producing hydrogen without the consumption of fossil fuels or the emission of greenhouse gases. As a result DOE has selected this system for the Next Generation Nuclear Power (NGNP) Project, a project to demonstrate emissions-free nuclear-assisted electricity and hydrogen production by 2015 (e.g., Ref. 1-6). The NGNP would be constructed at the Idaho National Engineering and Environmental Laboratory (INEEL) The NGNP will be a helium-cooled, graphite-moderated, thermal neutron spectrum reactor with an outlet temperature of 1000 o C or higher. The reactor core will be either a prismatic-block core or a pebble bed core; the final selection of a reference core design will be made during the first year of preconceptual design (e.g., Ref. 1-6). The NGNP will produce both electricity and hydrogen. The process heat for hydrogen production will be transferred to the hydrogen plant through an intermediate heat exchanger. The reactor thermal power and core configuration will be designed to assure passive decay heat removal without fuel damage during accidents. One or more processes will use the heat from the high temperature helium coolant to produce hydrogen. The first process of interest is the thermochemical splitting of water into hydrogen and oxygen; the leading candidate is the iodine-sulfur (IS) process. The second process of interest is thermally assisted electrolysis of water. A high efficiency gas turbine can be used to 3 In this report, AGR refers to He-cooled HTGRs and not to the British CO 2 -cooled Advanced Gas Reactors. 1-2

33 Introduction and Background generate the electricity to produce hydrogen from water by electrolysis. The efficiency of this process can be substantially improved by heating the water to high-temperature steam before electrolysis. A possible NGNP configuration with the above production capabilities is shown in Figure 1-1, and a possible deployment schedule for the NGNP is shown in Figure 1-2 (Ref. 1-7). An IAEA TECDOC (Ref. 1-8), published in 2001, provides an overview of the international Modular Helium Reactor (MHR) development with comprehensive descriptions of the USDOE/MINATOM GT-MHR program, the South African-based PBMR program, the Japanese High Temperature Test Reactor (HTTR) program and the Chinese HTR-10 program. This TECDOC is readily available on the IAEA web site, 4 along with links to the various international HTGR programs. The reader is referred to that site for background information and timely status reports as required. Figure 1-1 VHTR Demonstration Module 4 Available (in.pdf format) from the IAEA web site at 1-3

34 Introduction and Background Project Timeline Year: VHTR Demonstrator - Project Research & Development Pre-conceptual Design Conceptual Preliminary Final Construction Facility Startup Testing Hydrogen Production Demonstrator Research & Development Pilot Plant Design & Construction Facility Startup Testing Operations/Demonstration Testing Figure 1-2 NGNP Master Schedule As developed in this report, most of the fuel/fission product issues related to Modular HTGR design and licensing, including coated-particle fuel performance and fission product release from the core, are generic issues. While the issues are generic, it is constructive to consider them in the context of a specific reactor design. Of the current Modular HTGR designs, the International GT-MHR and the PBMR are the most mature with their preliminary design phases having been completed at this writing (in contrast, the NGNP is still in the pre-conceptual design phase). Since GA is an active participant in the GT-MHR program, more design-specific information is available to the writer for that reactor than for the PBMR; consequently, the focus here will be on the International GT-MHR program. A description of that program is given in the next section. 1.3 GT-MHR Program Programmatic The GT-MHR is a modular, helium-cooled, direct-cycle, nuclear power plant for high-efficiency electricity production intended to provide superior safety, economic, nonproliferation and environmental characteristics (Ref. 1-3). Currently, an international program, involving the United States and the Russian Federation, is developing the GT-MHR for the purpose of 1-4

35 Introduction and Background destroying surplus Russian weapons-grade plutonium (WPu, Ref. 1-9). This program includes extensive testing of the gas-turbine power conversion system and WPu coated-particle fuel. The International GT-MHR program workscope is to: Design, construct and operate a prototype GT-MHR module in Russia Design, construct and license a WPu fuel fabrication facility in Russia Operate the first 4-module GT-MHR plant for Pu disposition The GT-MHR is also an effective nuclear electric generation plant for commercial deployment when fueled with low-enriched uranium (LEU); consequently, a program has been proposed for commercial application in the United States and foreign markets of the technology being developed in Russia (Ref. 1-10). The commercial GT-MHR is expected to meet all Generation IV goals (Ref. 1-2) with significant margins. The engineering tasks necessary for adapting the technology being developed in Russia for commercial plant deployment in the U.S. consists of: GT-MHR design and technology transfer from Russia to the U.S.; Incremental design tasks which include: A LEU core design to replace the WPu core, A vented low-pressure containment (VLPC), An air-cooled, reactor cavity cooling system (RCCS); Uranium fuel fabrication development; Plant safety and licensing; Plant level design and analysis. For the present review, it is sufficient to note that the fission product release issue is generic (although the total core releases and speciation are influenced by the fuel cycle) and that the design and operating conditions for the plutonium consumption and commercial GT-MHR reactor systems are practically the same Design Overview The commercial GT-MHR plant is comprised of four 600 MW(t) modules which generate a total of 1145 MW(e). As shown in Figure 1-3, the module components are contained within two steel pressure vessels connected by a third cross vessel. All three vessels are sited underground in a concrete silo, which serves as part of an independent, vented low pressure containment structure 1-5

36 Introduction and Background Figure 1-3 GT-MHR Module The GT-MHR core, located inside the reactor pressure vessel, is designed to produce 600 MW(t) at a power density of 6.6 W/cm 3. The active core consists of an assembly of prismatic fuel elements in the form of hexagonal graphite blocks containing nuclear fuel compacts and coolant channels. The fuel elements are stacked 10 high in the core to form columns that rest on graphite support structures. The active core is composed of 102 fuel columns in an annular arrangement as shown in Figure 1-4. The annular core configuration was adopted to achieve the maximum power rating and still permit passive core heat removal while maintaining the peak fuel temperature below 1600 o C during the worst case accident condition of combined total loss of coolant and loss of flow, thereby assuring that fuel integrity is maintained. The design includes ceramic control rods in the outer side reflector for power control and in-core rods for reactor shutdown. Some key plant attributes are summarized Table

37 Introduction and Background REPLACEABLE CENTRAL & SIDE REFLECTORS 36 X OPERATING CONTROL RODS CORE BARREL BORATED PINS (TYP) REFUELING PENETRATIONS ACTIVE CORE 102 COLUMNS 10 BLOCKS HIGH 12 X START-UP CONTROL RODS PERMANENT SIDE REFLECTOR 18 X RESERVE SHUTDOWN CHANNELS Figure 1-4 Annular Core Table 1-1 GT-MHR Key Design Parameters Thermal Power Electrical Power Parameter Thermal efficiency 48% Core power density 6.6 W/cm 3 Core Inlet/outlet temperature Core Inlet/outlet pressure He mass flow rate Fuel form Feedstock Fuel cycle Fuel burnup Safety and licensing Proliferation resistance Final waste form Design Value 600 MW/module; 4 modules/standard plant 286 MW/module; 1145 MW/standard plant 491/850 o C 7.07/7.02 MPa 320 kg/s TRISO (ceramic) coated particles; prismatic graphite fuel elements LEU (<19.9% enriched) [options: LEU/Th, (Pu,U), (Pu,Th), etc.] Once-through; whole-element disposal without processing ~120,000 MWD/T with LEU fuel Meet EPA PAGs for all events with frequency of >5 X 10-7 /year Fresh fuel: LEU, enrichment to HEU required; dilute fuel form Spent fuel: unfavorable isotopics; difficult Pu recovery from fuel Radionuclides contained in robust TRISO particles encased in large monolithic graphite blocks 1-7

38 Introduction and Background The basic fuel element components are shown in Figure 1-5. The TRISO 5 particles consist of multiple ceramic coatings of silicon carbide and pyrolytic carbon surrounding the fuel kernels. They constitute tiny independent pressure vessels that retain fission products. The coatings are capable of maintaining their integrity and retaining fission products at temperatures higher than those imposed during normal operation and bounding accident conditions. The fuel particles are bonded together in fuel compacts that are contained in sealed vertical holes in the graphite fuel blocks. Figure 1-5 Prismatic Fuel-Element Components The primary barrier to fission product release from an HTGR is the fuel particle with its ceramic coatings. The fissile particles consist of 350 µm diameter spherical kernels containing 19.9% enriched uranium oxycarbide; the TRISO-coated particles have four coating layers with a combined thickness of 225 µm for an overall particle diameter of 800 µm. The fertile particles contain 500 µm diameter kernels of natural UCO with similar coatings for a total diameter of 880 µm. The fuel particles are bonded into fuel compacts by injecting a hot thermoplastic or thermosetting resin binder into a mold containing the fuel particles plus graphite shim particles. The shim particles are approximately the same size as the fuel particles and are used to allow a variation of the fuel loading while maintaining a constant, compact particle packing fraction. 5 TRISO = A coated-fuel particle design with three materials in coating system (low-density PyC, high-density PyC, and SiC). 1-8

39 Introduction and Background The four coating layers of a TRISO particle have specialized purposes (see Figure 1-6), but, in composite, they provide a high-integrity pressure vessel which is extremely retentive of fission products. The purpose of the low-density buffer layer is to provide a reservoir for fission gases released from the fuel kernel, to attenuate fission recoils, and to accommodate kernel swelling under irradiation. The main purposes of the high-density, inner pyrocarbon coating (IPyC) are to provide a smooth regular substrate for the deposition of a high-integrity SiC coating and to prevent Cl 2 and HCl from permeating the fuel kernel during the SiC deposition process; hence, a major benefit of IPyC coating is realized during fuel fabrication. The IPyC coating, which is intimately bonded to the SiC coating, also helps to maintain the SiC coating in compression as the former shrinks under irradiation whereas the latter is dimensionally stable. Figure 1-6 Functions of the Individual Coating Layers The most important coating in a TRISO particle is the SiC which provides most of the structural strength and the dimensional stability and which serves as the primary barrier to the release of fission products, particularly volatile metallic fission products such as cesium and silver. The high-density, outer pyrocarbon coating (OPyC), which shrinks under irradiation, also generates a compressive stress in the dimensionally stable SiC which partially compensates for the tensile stress component induced by the internal gas pressure. The PyC coatings have also been shown 1-9

40 Introduction and Background to effectively retain fission gases in fuel particles with defective or failed 6 SiC layers up to about 1800 o C. The fuel compacts are 12.4 mm (0.49 in) diameter by 50.1 mm (2.00 in) long right circular cylinders containing the particles embedded in a carbonaceous matrix. The matrix enhances heat transfer and prevents mechanical interaction of the particles with the fuel element graphite. Proportions of the fissile and fertile particle types within the fuel compacts of particular fuel elements are determined according to the fuel cycle reload prescription and the power zoning requirements with the total particle packing limited to <58% by volume. A standard prismatic fuel element is shown in Figure 1-5; certain other fuel elements contain an additional hole for the reserve shutdown control pellets which contain boron for reactivity control. The fuel compacts are stacked in 12.7 mm (0.5 in) diameter holes in graphite blocks to form the fuel elements. The graphite fuel elements are hexagonal right prisms with arrays of fuel and coolant holes. In the center of the top end is a pickup hole for remote fuel handling. Each fuel element is 360 mm (14.17 in) across the hexagonal flats and 793 mm (31.22 in) in height. The standard fuel element has a total of 174 fuel holes and 91 coolant holes. The 12.7 mm (0.5 in) diameter fuel holes are blind holes terminating about 12.7 mm (0.5 in) from the bottom of the element. After the fuel compacts are inserted, the fuel holes are capped with 12.7 mm (0.5 in) thick graphite plugs which are cemented in place. The power conversion system (PCS), also shown in Figure 1-3, operates on the Brayton cycle. The PCS contains the gas turbine, electric generator, and two compressor sections mounted on a single vertical shaft, suspended by electromagnetic bearings. The magnetic bearings control rotational stability and support thrust loads while eliminating the need for lubricants within the primary coolant system. Recuperator, precooler, and intercooler heat exchangers are enclosed along with the balance of the PCS within the single power conversion vessel. The 95% effective recuperator, which recovers a large fraction of the turbine exhaust heat, makes possible the high thermal efficiency. With the reference core outlet temperature of 850 o C, the net thermal efficiency of the plant is 48% which significantly improves the plant economics compared to steam cycle systems. Even higher plant efficiencies (>50%) are achievable with higher core outlet temperatures. The high efficiency of the GT-MHR also results in reduced high-level wastes and a 50% reduction in thermal discharge relative to water-cooled nuclear plants on a per unit electric generation basis. As shown in Figure 1-7, helium is heated in the reactor core and flows to the PCS where it expands through a gas turbine to run a generator to produce electricity. From the turbine exhaust, the helium flows through the hot side of the recuperator to transfer residual heat to the helium returning to the reactor system on the recuperator cold side. From the recuperator, the helium flows through a precooler where it is further cooled. The cooled helium passes through 6 In the convention adopted here (and in most of the coated-particle fuel literature), the terms defect or defective refer to the as-manufactured state of the fuel, and the terms failure and failed refer to in-service consequences.(both normal operation and postulated accidents). 1-10

41 Introduction and Background low and high-pressure compressors with intercooling. From the compressor section, the helium flows through the cold side of the recuperator, where it is heated for return to the reactor system. Figure 1-7 GT-MHR Power Conversion System A reactor cavity cooling system (RCCS) is located on the concrete structure external to the reactor vessel to remove core heat from the reactor vessel when forced core cooling is unavailable. The core heat is transferred by conduction, convection, and radiation to the aircooled RCCS. The RCCS transfers the heat by natural circulation to the atmosphere. This system is completely passive and has no controls, valves, circulating fans, or other active components. Moreover, even if the passive RCCS were assumed to be failed, passive radiation and conduction to the reactor cavity, walls, and environment would be sufficient to constrain peak core temperatures and to limit radionuclide release. The spent GT-MHR fuel elements are extremely stable and resistant to corrosion, and they can be placed directly in a repository without further processing. Long term corrosion models, developed from accelerated leach tests conducted at Oak Ridge National Laboratory (ORNL) and Pacific Northwest Laboratory, predict a coated particle failure fraction of only 10-4 after 10 6 years in a repository (Ref. 1-11). 1-11

42 Introduction and Background 1.4 Manufacture of HTGR Fuel The primary safety goal for all modern HTGRs is to design the plant such that the radionuclides are retained in the core during normal operation and postulated accidents. The key to achieving this goal is the reliance on ceramic-coated fuel particles for primary fission product containment at their source. As will be discussed in Section 2, the in-service performance of coated-particle fuel is strongly influenced by its as-manufactured attributes; consequently, it may be constructive to review briefly coated-particle manufacturing technology to provide a context. A comprehensive description of coated-particle fuel manufacturing technology is beyond the scope of this review. The intention here is simply to provide an introduction and to provide the interested reader with key references. The fabrication of coated-particle fuel essentially involves the serial production of three components: (1) fuel kernels, (2) coated particles, and (3) fuel elements. These fuel components are shown in Figure 1-5 for a prismatic fuel element and in Figure 1-8 for a spherical fuel element. The overall manufacturing process is illustrated in Figure 1-9 which shows the process used by General Atomics to make prismatic fuel elements (Ref. 1-12). The process for manufacturing spherical fuel elements (e.g., Ref. 1-13) is quite similar through the coated particle phase; the process for sphere making is shown in Figure Manufacturing Technology Fuel Kernels The small ( µm) spheres of fuel material at the center of the coated particles, consisting of ceramic compounds of fissile or fertile metals (e.g., U, Th, Pu) and oxygen or carbon, are referred to as kernels. Kernels for early coated particle fuel, including fuel for Peach Bottom and Fort St. Vrain in the US, as well as initial loadings of the AVR in the FRG, were of a carbide form (e.g., (U,Th)C 2 ). The Dragon Reactor in the United Kingdom was an early experimental reactor fueled with an oxide (UO 2 ). The majority of international coated particle fuel development since the early 1970s has been directed toward the oxide fuel form. A mixture of oxide and carbide fuel, labeled UCO, was selected in 1981 as the reference fuel in the US due to its projected superior performance at higher temperatures and burnups anticipated for economically competitive prismatic core reactors. A number of techniques have been used to fabricate fuel kernels for coated particles, including sphere melt, powder agglomeration, and gel formation; however, all modern kernel making processes are based upon gel formation where spherical kernels are formed from droplets of aqueous metal solutions (e.g., Ref. 1-12). In the simplest terms, the process commences with the formation of uniform size droplets by passing a solution containing heavy metal nitrates through vibrating needles, resulting in each drop containing a precise amount of metal. The droplets are then gelled by various chemical reactions. After washing, the gelled microspheres are dried and then undergo a controlled heat treatment with an appropriate atmosphere that sinters the microspheres into dense ceramic spheres with the required stoichiometry. Such gel formation processes have been used to manufacture kernels of various enrichments and compositions. 1-12

43 Figure 1-8 Spherical Fuel-Element Components Introduction and Background 1-13

44 Introduction and Background Figure 1-9 Process Flow Diagram for Prismatic Fuel Elements 1-14

45 Figure 1-10 Process Flow Diagram for Sphere Making Introduction and Background 1-15

46 Introduction and Background Ammonia-based precipitation processes developed more recently, referred to as either internal gelation (e.g., Ref. 1-14) or external gelation (e.g., Ref. 1-15), have rendered the traditional sol-gel processes obsolete. Both internal and external gelation processes use a concentrated heavy-metal nitrate feed that is acid-deficient, and therefore require a concentration/denitration solution preparation step. These newer gel-precipitation methods are simpler, more reliable, and more suitable for scale-up and adaptation to radiochemical processing than the traditional sol-gel techniques. Both internal and external gelation processes have undergone extensive development for the production of microspheres containing UO 2, UCO, ThO 2, and mixed systems of U-Th, U-Pu, and Th-Pu as well as pure carbides and nitrides. Only limited work has been done on pure Pu or Purich systems. An external gelation/precipitation process (which was invented in Italy) has been used to manufacture both UCO and UO 2 kernels at GA and ORNL in the USA, at Nuclear Fuel Industries (NFI) in Japan, and at HOBEG Company in Germany. The feasibility of producing 350 µm and 500 µm diameter UCO by this method has been demonstrated, although the quality requirements were not completely demonstrated for UCO material. INET in China has produced 500 µm UO 2 kernels using a variation of the external gelation process labeled total gelation. HEU UCO kernels of 195 µm diameter were fabricated in laboratory scale equipment by an internal gelation/precipitation process under the New Production-Modular High Temperature Gas-Cooled Reactor (NP-MHTGR) program by BWXT Corporation. That material showed more uniform structure and density than similar kernels from external gelation, and on that basis the internal gelation process was selected as the reference process for the DOE-sponsored AGR fuel development program (Ref. 1-16) which has the mission of developing and qualifying coated-particle fuel for the VHTR program introduced above Coated Particles The basic principles for applying TRISO coatings on dense microspheres by chemical vapor deposition are well established after nearly four decades of international development. Each coating layer is deposited on the fuel particle by chemical vapor deposition in a high-temperature fluidized bed (e.g., Ref. 1-12). The type and attributes of the coatings deposited are determined by complex chemical kinetics and fluidization dynamic mechanisms. The thermal decomposition and deposition of PyC coatings on fuel particles was first described by LeFevre (Ref. 1-17). The exhaustive study by Voice in the 1960s of the relationships between coating process parameters and the attendant physical properties of the SiC coating remains a seminal work (Ref. 1-18). More recent TRISO coating process experience in the United Kingdom and Germany has also been described (Refs and 1-19, respectively). Nevertheless, in certain regards, the production of high-quality coatings, especially PyC coatings, is as much art as science. 1-16

47 Introduction and Background The coating operation begins by preheating the empty coater to approximately the coating temperature with an inert purge gas. Once the particle loading temperature is reached, the fuel kernels or partially coated particles are loaded into the coater. The gas velocity in the coater is adjusted to completely fluidize the fuel particle bed; all introduced from the bottom of the coater. The gas flow into coater is switched from the purge gas to the appropriate coating gas by the control system and the coating process commences. Pyrocarbon coating (buffer, IPyC, and OPyC) is accomplished by thermal decomposition of acetylene, propylene or a mixture of the two gases. The hydrocarbon gas(es) decomposes in the coater hot zone and deposits PyC on the fluidized particles. Argon and hydrogen are used as the diluent and fluidizing gases. The lowdensity buffer layer is typically deposited at ~1250 o C, and the high-density IPyC and OPyC layers are typically deposited at ~1300 o C. 7 SiC coating is accomplished by the thermal decomposition of methyl trichlorosilane (MTS) which is introduced into the coater as a vapor with pure hydrogen as the carrier gas and diluent. Acceptable SiC coatings can be deposited in the 1500 to 1650 o C range, but the lower end of this range appears to be optimal. After the prescribed coating time to produce the required coating thickness has elapsed, the levitation gas (argon or nitrogen) in turned on and the coating gas turned off by the control system. If more than one coating layer is to be deposited in a single coating run (e.g., the buffer plus the IPyC layers), the coater is heated (or cooled) to the next required temperature. At the new coating temperature, the new coating gas is introduced, and the coating sequence resumes. After the coating run is complete, an inert gas is introduced and the coater is allowed to cool down. At about 1100 o C or lower, the coated particles are discharged from the coater. The coated particles are sampled at various stages during the coating process for quality control/quality assurance (QC/QA) measurements. Once all the coating layers have been applied and approved by quality control, the particles are tabled, screened and/or elutriated to remove over/under size ranges, misshapen particles, and debris before compositing for compact or sphere fabrication. The optimal process conditions for the production of high-quality, high-performance, TRISOcoated fuel particles has been a much discussed topic (e.g., Refs and 1-21); however, there is an emerging international consensus that the reference German coating process conditions are near optimal. Consequently, the AGR fuel development program has effectively adopted the German coating technology as the basis for its planned coating process development activities (Ref. 1-16). The key coating process variables (e.g., Ref. 1-20) are the bed fluidization conditions, coating temperature, the coating rate and the ratio of active coating gas to carrier gas (i.e., the concentration of coating gas). For the pyrocarbon coatings, key material properties are the permeability and the degree of anisotropy. Permeable IPyC coatings allow the formation of volatile heavy-metal chlorides during SiC coating, and anisotropic PyC coatings can fail excessively under neutron irradiation. The process variables that most influence these PyC 7 These coatings are so-called Low Temperature Isotropic (LTI) PyC coatings; earlier, High Temperature Isotropic (HTI) PyC coatings deposited from methane at ~2000 o C were also investigated, but they exhibited high failure rates at high fast fluences. 1-17

48 Introduction and Background properties are the coating rate and the active gas concentration; bed fluidization conditions and coating temperature are less important. For the SiC coating, key material properties are the coating strength and grain structure; weak SiC coatings can fail mechanically, and large columnar grains and/or free Si or carbon are thought to facilitate diffusive release of volatile fission metals and enhanced SiC corrosion rates. The process variables that most influence these SiC properties are the coating temperature and coating rate Fuel Elements As previously stated, the processes for fabricating fuel kernels and coated particles for prismatic and spherical fuel elements are essentially the same. The processes for fabricating fuel compacts for prismatic fuel elements and for fabricating fuel spheres are also similar but with some important differences as described in the following subsections Prismatic Fuel Elements With prismatic fuel elements, the coated fuel particles are incorporated into small solid cylinders called fuel compacts or, earlier, fuel rods - for placement within the graphite fuel blocks for handling and performance considerations (e.g., higher thermal conductivity, elimination of particle-block interactions, etc.). The particles are bonded together in each compact with a carbonaceous matrix binder; this binder can be derived from either a thermoplastic feedstock (e.g., petroleum pitch, coal-tar pitch, etc.) or a thermosetting resin. The fabrication process (e.g., Ref. 1-12) involves several steps for combining the fuel particles and other raw materials and forming them into finished compacts. With thermoplastic matrices, the matrix is first formulated by mixing heated petroleum (or coal tar) pitch, graphite filler and selected additives, and then granulating the cooled and hardened matrix cakes. Fuel particles and graphite shim particles (to maintain a constant particle volume in the compact) are weighed and blended, then loaded into steel mold cavities where the heated fluidized matrix is pressure injected into the particle bed to form green compacts. Since shrinkage of the matrix during heat treatment and irradiation may be a source of particle failures, the amount of graphite filler, which is more dimensionally stable than the pitch, is maximized within limits on viscosity of the mixture during the injection phase. A two-step inert gas heat treatment drives off matrix volatiles and carbonizes the compact binder. High-temperature leaching with gaseous HCl has been used extensively by GA to remove heavymetal contamination, exposed kernels, and metal impurities from the compacts prior to final heat treatment. This HCl cleaning step was especially important for the fabrication of the fuel compacts used in the Fort St. Vrain HTGR. A number of fuel compact fabrication processes based upon resin matrices have been developed over the years (e.g., Ref. 1-22). The admix process that was developed for the early fuel loads of the Dragon reactor is an early example. In the admix process the basic ingredients of the fuel compact matrix are a high char yield resin and a graphitized petroleum coke filler. Fuel particle volume fractions in this process were typically >25%. Some compacts with particle volume fractions of ~35% were made but there was an increased tendency for fuel particle coating 1-18

49 Introduction and Background damage to occur at these higher volume fractions. Consequently, the Dragon project developed an overcoating process to reduce fuel particle-to-particle interaction and allow for higher volume fraction compacts to be manufactured. Variants of this overcoating process have been used to manufacture the fuel compacts for the Japanese HTTR (Ref. 1-23) and to manufacture fuel spheres for pebble-bed reactors (see next subsection). A primary goal of compact (and pebble) fabrication process is to incorporate fuel particles into the compacts without damaging the particles or introducing contaminants into the fuel element. Until fairly recently, the greatest concern was mechanical interaction between the particles during compacting; consequently, key parameters, such as injection pressure, injection rates, temperature, etc., were carefully controlled. This concern was also the inspiration of the aforementioned overcoating process. Undoubtedly, particle interactions during compacting can lead to mechanical coating failures and must be minimized; however, a comprehensive investigation performed in 1996 revealed that chemical attack of the SiC coatings by transition metal impurities, principally iron, during high-temperature heat treatment of the green compacts may be an even greater cause of coating failure (Ref. 1-24). Such transition-metal impurities are present in relatively high concentrations in petroleum pitch; however, these impurities are removed by high-temperature HCl cleaning. The primary source of the metal impurities in the 1996 study was identified to be impurities volatilized from the graphite liners of the hightemperature furnace used for final heat treatment. (High temperature heat treatment in a vacuum furnace reduces the impurity content in the furnace atmosphere and is likely the preferred approach for future fuel manufacturing.) Based upon the available test results and fabrication experience, high-quality fuel compacts can be fabricated using either a thermoplastic or a thermosetting resin matrix. However, several other factors influence the selection of the binder, and thus selection of an appropriate binder must be a compromise. Pitch is a thermoplastic material (i.e., it will soften when heated during the pyrolysis step) requiring the compacts to be supported in a pack of alumina powder. Moreover, pitch, being derived from oil or coal, contains chemical impurities (particularly transition metals) that could attack the SiC coatings and contribute to particle failure. The use of a man-made thermosetting resin is therefore attractive since it cures (hardens) during the compacting stage and can be specified to contain low maximum levels of chemical impurities. Not having to restrain the compact in alumina during carbonization also eliminates a further opportunity for chemical attack of the particles and significantly reduces the volume of low-level process waste (by eliminating contaminated alumina). Consequently, the AGR and Russian fuel development programs have adopted thermosetting resin as the reference compact matrix for the VHTR (Refs and 1-22) Spherical Fuel Elements The spherical fuel elements developed by the Germans for use in pebble-bed HTRs were ultimately manufactured with a derivative overcoating process (Refs and 1-25). The spherical fuel element ( pebble ) consists of a spherical fuel particle compact contained in a spherical shell of unfueled matrix material. As shown in Figure 1-10, the graphite powder and phenolic resin (diluted with methanol) are mixed, dried, milled, and homogenized. Part of this matrix mixture is then fed into a rotating drum along with the fuel particles and methanol. The 1-19

50 Introduction and Background overcoated particles are then dried, sieved, and isostatically pressed in a rubber mold at ~30 MPa (with additional matrix to form the fuel compact. The remainder of the matrix is then molded around the fuel compact to form the fuel-free shell of the fuel pebble, and the entire pebble is subjected to high-pressure isostatic molding at ~300 MPa. The fuel pebble is then slowly heated to 800 o C in an inert atmosphere to carbonize the resin binder and in a subsequent step is heat treated at temperatures up to 1950 o C in vacuum to degas the fuel pebbles. The matrix mixture used in the German molding process was designated the A3 matrix and consisted of a mixture of 64 wt% natural flake graphite, 16 wt % graphitized petroleum coke, and 20 wt % phenolic resin, and in the carbonized condition it contained ~90 wt% graphite and only ~10 wt % of the less stable resin char (Ref. 1-22). While this method is very similar to the Dragon method, a thinner particle overcoating is used in pebble production since a fraction of the matrix is introduced into the fuel compact after the overcoating step. The fuel particle volume fraction, including the fuel free shell, in the German pebble process was typically rather low (~5% for the AVR and ~12% for the THTR). The Chinese essentially replicated the German fabrication processes to manufacture spherical fuel elements for their pebble-bed HTR-10 test reactor (Ref. 1-26). Similarly, the PBMR Project will replicate the German processes for fabrication of spherical fuel elements in South Africa for their pebble-bed, direct-cycle Modular HTR (Ref. 1-5) Manufacturing Experience As summarized in Table 1-2 (Ref. 1-27), coated-particle fuel has been produced in quantity for use in fuel development programs, experimental reactors and power reactors. The asmanufactured quality (e.g., heavy-metal contamination, coating defects, etc.) of these fuels varied considerably. General Atomics mass produced TRISO-coated HEU (Th,U)C 2 /ThC 2 fuel (33,000 kg Heavy Metal) for the Fort St. Vrain HTGR (Ref. 1-28), but the as-manufactured heavy-metal (HM) contamination and coating defect fractions were at least an order-ofmagnitude higher than that anticipated to be required for the VHTR. The German company HOBEG mass produced TRISO-coated LEU UO 2 reload fuel for the AVR (Ref. 1-25) which met these anticipated VHTR requirements. The fuel manufactured by NFI for the Japanese HTTR and by INET for the Chinese HTR-10 also approaches this quality level. 1-20

51 Introduction and Background Table 1-2 Production of Coated-Particle Fuel Reactor/Manufacturer Fuel Description Total Heavy Metal (kg) ROVER/GA LANL BISO 8 in compacts 1 Peach Bottom/ GA BISO in compacts 3,500 UHTREX/GA LANL TRIPLEX 9 in compacts 200 DRAGON TRISO, BISO compacts 300 FSV/ GA TRISO in compacts 33,400 THTR/ NUKEM BISO in spheres 11,000 AVR/ NUKEM HEU BISO, TRISO spheres 1,700 AVR/ NUKEM Modern LEU TRISO spheres 480 US Fuel Development (GA) Modern TRISO UCO 500 HTTR/ NFI Modern LEU TRISO compacts 900 HTR-10/ INET Modern LEU TRISO spheres Report Organization As indicated in Section 1.1, the goal of this review is to acquire, evaluate and summarize the existing international data base on the release of radionuclides from HTGR cores during normal operation. Fortunately, there are a number of useful resources that can serve as the point of departure for this review. Perhaps, the best single resource is IAEA TECDOC-978, entitled Fuel Performance and Fission Product Behavior in Gas Cooled Reactors, (henceforth, referred to this report as TECDOC-978, Ref. 1-29) which is essentially an encyclopedia of HTGR fuel performance and fission product transport data. This report is intended to be a stand-alone document (although the serious reader is strongly encouraged to acquire key references, especially those alluded to above). It is organized as follows. After the introductory and background information presented here in Section 1, the design methods for predicting fuel performance and fission product release from the core are described in Sections 2 and 3, respectively. The international fuel performance and fission product release data bases are summarized in Sections 4 and 5, respectively; for reader convenience, TECDOC-978 is excerpted extensively in Sections 4 and 5. The comprehensive core performance analysis performed for the Plutonium Consumption-Modular Helium Reactor (PC-MHR) is discussed in Section 6; much of the analytical results presented in this section are reproduced from Ref which may not be readily available outside of GA. The additional technology development required to reduce the large uncertainties in core release predictions is summarized in Section 7. Finally, a series of conclusions and recommendations are presented in Section 8. 8 BISO = A coated-fuel particle design with two materials in the coating system (low-density PyC, high-density PyC) 9 TRIPLEX = A coated-particle design consisting of buffer/sic/opyc. 1-21

52

53 2 RADIONUCLIDE RELEASE FROM HTGR CORES 2.1 Radionuclide Control Philosophy A fundamental requirement in the design of any nuclear power plant is the containment and control of the radionuclides produced by the nuclear reactions; in response, different radionuclide containment systems have been designed and employed for different reactor designs. For modular HTGR designs, a hallmark philosophy has been adopted since the early 1980s to design the plant such that the radionuclides would be retained in the core during normal operation and postulated accidents. The key to achieving this safety goal is the reliance upon ceramic-coated fuel particles for primary fission product containment at their source, along with passive cooling to assure that the integrity of the coated particles is maintained even if the normal cooling systems were permanently disrupted. This innovative design philosophy - radionuclide containment at the source for all credible plant conditions has been discussed in numerous publications, but it is perhaps best elaborated in a Preliminary Safety Information Document (PSID) for the 350 MW(t) steam-cycle Modular HTGR (MHTGR) that was submitted to the U.S. Nuclear Regulatory Commission (NRC) in 1987 (Ref. 2-1). This philosophy has been carried forward for all subsequent MHR designs. 2.2 Radionuclide Containment System The radionuclide containment system for an HTGR, which reflects a defense-in-depth philosophy, is comprised of multiple barriers to limit radionuclide release from the core to the environment to insignificant levels during normal operation and a spectrum of postulated accidents. As shown schematically in Figure 2-1, the five principal release barriers are: (1) the fuel kernel, (2) the particle coatings, particularly the SiC coating, (3) the fuel element structural graphite, (4) the primary coolant pressure boundary; and (5) the reactor building/containment structure. The effectiveness of these individual barriers in containing radionuclides depends upon a number of fundamental factors including the chemistry and half-lives of the various radionuclides, the service conditions, and irradiation effects. The effectiveness of these release barriers is also event specific. The first barrier to fission product release is the fuel kernel itself. Under normal operating conditions, the kernel retains >95% of the radiologically important, short-lived fission gases such as Kr-88 and I-131. However, the effectiveness of a UCO kernel for retaining gases can be reduced at elevated temperatures or if an exposed kernel is hydrolyzed by reaction with trace amounts of water vapor which may be present in the helium coolant (the UO 2 kernel used in PBMR fuel is somewhat less susceptible to hydrolysis effects than is UCO). The retentivity of 2-1

54 Radionuclide Release From HTGR Cores oxidic fuel kernels for long-lived, volatile fission metals such as Cs, Ag, and Sr is strongly dependent upon the temperature and the burnup. Figure 2-1 MHR Radionuclide Containment System The second - and most important - barrier to fission product release from the core is the silicon carbide and pyrocarbon coatings of each fuel particle. Both the SiC and PyC coatings provide a barrier to the release of fission gases. The SiC coating acts as the primary barrier to the release of metallic fission products because of the low solubilities and diffusion coefficients of fission metals in SiC; the PyC coatings are partially retentive of Cs at lower temperatures but provide little holdup of Ag and Sr. With a prismatic core, the fuel-compact matrix and the graphite fuel block collectively are the third release barrier (with a pebble-bed core, the analog is the pebble matrix, including the unfueled outer shell). The fuel-compact matrix is relatively porous and provides little holdup of the fission gases which are released from the fuel particles. However, the matrix is a composite material which has a high content of amorphous carbon, and this constituent of the matrix is highly sorptive of metallic fission products, especially Sr. While the matrix is highly sorptive of metals, it provides little diffusional resistance to the release of fission metals because of its high interconnected porosity. The fuel element graphite, which is denser and has a more ordered structure than the fuelcompact matrix, is somewhat less sorptive of the fission metals than the matrix, but it is much more effective as a diffusion barrier than the latter. The effectiveness of the graphite as a release 2-2

55 Radionuclide Release From HTGR Cores barrier decreases as the temperature increases. Under typical core conditions, the fuel element graphite attenuates the release of Cs from the core by an order of magnitude, and the Sr is essentially completely retained. The extent to which the graphite attenuates Ag release is not nearly as well characterized, and there is some evidence that the retention of Ag by graphite increases as the total system pressure increases (implying gas-phase transport through the interconnected pore structure of the graphite). Typically, the two dominant sources of fission product release from the core are asmanufactured, heavy metal contamination (i.e., heavy metal outside the coated particles) and particles whose coatings are defective or fail in service. In addition, the volatile metals (e.g., Cs, Ag, Sr) can, at sufficiently high temperatures for sufficiently long times, diffuse through the SiC coating and be released from intact TRISO particles; however, diffusive release from intact particles during normal operation is only significant compared to other sources for silver release. Fission products resulting from fissions in heavy-metal contamination outside of the particles are obviously not attenuated by the kernels or coatings, nor are the fission products produced in the kernels of failed particles appreciably attenuated by the failed coatings. In these cases, the fission products must be controlled by limiting the respective sources and by the fuel element graphite in the case of the fission metals and actinides. The fourth release barrier is the primary coolant pressure boundary. Once the fission products have been released from the core into the coolant, they are transported throughout the primary circuit by the helium coolant. The helium purification system (HPS) efficiently removes both gaseous and metallic fission products from the primary coolant at a rate determined by the gas flow rate through the purification system (the primary purpose of the HPS is to control chemical impurities in the primary coolant). However, for the condensable fission products, the dominant removal mechanism is deposition ( plateout ) on the various helium-wetted surfaces in the primary circuit (i.e., the deposition rate far exceeds the purification rate). The plateout rate is determined by the mass transfer rates from the coolant to the fixed surfaces and by the sorptivities of the various materials of construction for the volatile fission products and by their service temperatures. Condensable radionuclides may also be transported throughout the primary circuit sorbed on particulates ( dust ) which may be present in the primary coolant; the plateout distribution of these contaminated particulates may be considerably different than the distribution of radionuclides transported as atomic species. The circulating and plateout activities in the primary coolant circuit are potential sources of environmental release in the event of primary coolant leaks or as a result of the venting of primary coolant in response to overpressuring of the primary circuit (e.g., in response to significant water ingress in a steam-cycle plant). The fraction of the circulating activity lost during such events is essentially the same as the fraction of the primary coolant that is released, although the radionuclide release can be mitigated by pump down through the HPS if the leak rate is sufficiently slow. A small fraction of the plateout may also be reentrained, or lifted off, if the rate of depressurization is sufficiently rapid. The amount of fission product liftoff is expected to be strongly influenced by the amount of dust in the primary circuit as well as by the presence of friable surface films on primary circuit components which could possibly spall off during a rapid depressurization. 2-3

56 Radionuclide Release From HTGR Cores Other mechanisms which can potentially result in the removal and subsequent environmental release of primary circuit plateout activity are steam-induced vaporization and washoff. In both cases, the vehicle for radionuclide release from the primary circuit is water which has entered the primary circuit. In principle, both water vapor and liquid water could partially remove plateout activity. However, even if a fraction of the plateout activity were removed from the fixed surfaces, there would be environmental release only in the case of venting of helium/steam from the primary circuit. For all but the largest water ingress events the pressure relief valve does not lift. Moreover, the radiologically important nuclides such as iodine and cesium are expected to remain preferentially in the liquid water which remains inside the primary circuit. The probability of large water ingress with a direct-cycle plant is much lower than for a steam-cycle plant because with the former, the secondary water pressures are lower than the primary He pressures, and the heat-exchanger tubes experience much lower temperatures during normal operation. The reactor building/containment structure is the fifth barrier to the release of radionuclides to the environment. Its effectiveness as a release barrier is highly event-specific. The vented low-pressure containment may be of limited value during rapid depressurization transients; however, it is of major importance during longer term, higher risk core conduction cool-down transients during which forced cooling is unavailable. Under such conditions, the natural removal mechanisms occurring in the VLPC, including condensation, fallout and plateout, serve to attenuate the release of condensable radionuclides, including radiologically important iodines, by at least an order of magnitude. 2.3 Phenomena Controlling Release As-Manufactured Fuel Attributes The as-manufactured attributes of coated fuel particles strongly influence their capability to retain radionuclides during normal operation and postulated accidents. Since the particle coatings are the primary barriers to radionuclide release, the most obvious as-manufactured attributes affecting radionuclide release are the amount of heavy-metal contamination and the fraction of defective coatings. The required as-manufactured fuel attributes including the allowable HM contamination and coating defects are defined in detail in a Fuel Product Specification (e.g., Ref. 2-2); the types of defects controlled and the limits imposed for various HTGR designs, including those for the commercial GT-MHR, are described in Section 2.4. The in-service coating performance may also be profoundly influenced by the as-manufactured coating material properties (e.g., the performance of PyC coatings under irradiation is strongly influenced by the degree of anisotropy). Ideally, it would be desirable to rely exclusively on product specifications to control the as-manufactured attributes of coated-particle fuel (especially if there were multiple independent fuel suppliers). Unfortunately, the mechanistic understanding of particle performance and the available QC methods are not yet adequate to guarantee with sufficient confidence the required in-service performance based upon measured product attributes; consequently, it is necessary to supplement the product specifications with process specifications, at least for the foreseeable future. Pertinent examples are presented below for 2-4

57 Radionuclide Release From HTGR Cores PyC and SiC coatings to illustrate the point; the reader is referred to the specialist literature (e.g., Refs. 2-3 and 2-4) for a more comprehensive treatment of this complicated, and often confusing and contradictory, subject. As previously mentioned, the irradiation performance of PyC coatings is dependent upon the anisotropy because anisotropic PyC coatings shrink excessively under neutron irradiation, accumulate tensile stresses, and may fail as a consequence. For IPyC coatings the failure mode is typically radial cracks which may serve as stress risers and crack initiators for the SiC layer which may result in pressure-vessel failure of the entire coating system or may serve as pathways for fission product release and for corrosive agents, including CO in UO 2 particles, to attack the SiC coating. For OPyC coatings, excessive irradiation-induced shrinkage results in high failure rates, thereby eliminating a compressive stress imposed upon the SiC coating which serves to counteract a tensile stress component induced by the internal gas pressure. For OPyC coatings, there is a complicating factor as a result of possible matrix/coating interaction. If the matrix, which shrinks more than the OPyC coating under irradiation, is too tightly bonded to the coating, it will tear the OPyC coating off the SiC coating. Open surface porosity in the OPyC layer encourages an excessively strong bond with the matrix; hence, this porosity must be controlled. For coated-particle fuel, the anisotropy of PyC coatings has historically been determined by an optical technique that measures the property as a ratio of the intensities of polarized light in two perpendicular directions reflected by the coating cross-section (Ref. 2-5); this ratio is referred to as the optical Bacon Anisotropy Factor (BAF o ). Unfortunately, the OPyC failure and BAF o are not well correlated as is illustrated in Figure 2-2a, and there are few reliable data to determine the degree of correlation between IPyC cracking and BAF o measurements (Ref. 2-3). 10 As shown in Figure 2-2b, the OPyC failure rate correlates somewhat better with porosity (as measured by high-pressure mercury intrusion), but very high failure rates are occasionally observed at the lowest measured porosities. Consequently, these OPyC coating product specifications must be supplemented by process specifications. As shown in Figure 2-3, it was determined experimentally that the OPyC failure rate best correlated with active coating gas ratio (and coating rate, not shown), but it appeared to be independent of coating temperature (Ref. 2-3). The circumstances for SiC coatings are similar, but not identical, to those for OPyC coatings. For SiC coatings, material properties can in general be specified that are satisfactory indicators of good in-service performance. For example, the density, mechanical strength, etc., can be reliably specified and measured with existing QC methods. The exception is the relationship between the SiC microstructure and the coating s ability to retain volatile fission metals, including silver and cesium. A number of studies have been conducted, especially by the Germans (Refs. 2-6 and 2-7), to develop correlations between fission metal retention and SiC material properties, especially microstructure, with limited success. 10 The importance of the IPyC layer to the overall structural performance of a TRISO-coated particle was only recognized in the past 10 years. Previously, when much of the TRISO irradiation testing was done, it was thought that the primary purpose of the IPyC coating was to prevent the formation of volatile heavy-metal chlorides during SiC coating. Consequently, there is a paucity of anisotropy measurements for IPyC coatings in historical test fuel. 2-5

58 Radionuclide Release From HTGR Cores (a) (b) Figure 2-2 OPyC Failure versus Material Properties 2-6

59 Radionuclide Release From HTGR Cores Figure 2-3 OPyC Failure vs. Process Variables 2-7

60 Radionuclide Release From HTGR Cores While there is a general consensus that theoretically dense β-phase SiC is preferred and that less ordered α-phase SiC and free Si and carbon are to be avoided, reliable correlations between grain size and structure and fission metal retention have not been determined. SiC coatings with small columnar grains have been shown to be stronger, and intuitively they should provide better fission metal retention because the more discontinuous grain boundaries, which are characteristic of a small grain structure, should reduce the effective diffusivities; however, the studies to date (e.g., Ref. 2-7) have been inconclusive. Consequently, the QC approach with respect to the SiC coating has been historically to rely upon density measurements and process specifications to ensure good in-service performance of the SiC coating (Ref. 2-3). Although the structure and properties of pyrolytic SiC are somewhat less sensitive to process variables than that of pyrocarbon, a number of process variables, including the deposition temperature, coating rate and active gas coating ratio, are nevertheless important. While satisfactory SiC coatings can be apparently deposited over a fairly broad temperature range (say, from o C), coatings deposited at ~1500 o C, as shown in Figure 2-4, are somewhat stronger and have small columnar grains which is thought to be desirable (Ref. 2-3). The Germans have historically used a SiC deposition temperature of about 1500 o C, and GA had previously used 1650 o C; however, based upon the success of the German fuel development program and a review of the process data, GA also adopted a SiC deposition temperature of ~1500 o C in 1994 (Ref. 2-3). Consistent with this emerging consensus, the AGR fuel development program has also adopted ~1500 o C as the reference SiC deposition temperature (Ref. 2-8). Another intriguing aspect of the aforementioned German process studies is perhaps noteworthy here. All modern TRISO fuel, including the high-performance German fuel fabricated by HOBEG, has been produced using a mixture of MTS and pure hydrogen to deposit the SiC coating. However, process studies performed at the Kernforschungsanlage Juelich (KFA) in the early 1980s (Refs. 2-6 and 2-7) suggested that a superior, more uniform SiC coating could be produced if argon was used as a diluent (Ar is used as a diluent in PyC coating). As illustrated in Figures 2-5 and 2-6, the use of a mixture of H 2 /Ar instead of pure H 2 as the carrier gas appears to produce a superior SiC coating which is also much less sensitive to variations in deposition temperature. Given the inherent variability of the temperatures within a fluidized bed, this reduced sensitivity to deposition temperature should produce a more uniform product. Surprisingly, there is no published indication that any manufacturer of TRISO-coated fuel particles, including HOBEG, pursued the use of an Ar diluent instead of pure hydrogen. To reiterate, the purpose of this subsection was to demonstrate that the as-manufactured attributes of coated-particle fuel can have a dramatic impact on the in service performance of the particle. Consequently, potential changes in the particle design and/or in the process conditions which promise improvements in the as-manufactured fuel quality (e.g., lower coating defects) should not be adopted until irradiation data and postirradiation heating data are available to quantify the performance consequences of these apparent improvements. Moreover, experience teaches that it is particularly unwise to make multiple simultaneous changes in the particle design or process conditions. A dramatic illustration follows. 2-8

61 Radionuclide Release From HTGR Cores Figure 2-4 SiC Material Properties vs. Process Variables 2-9

62 Radionuclide Release From HTGR Cores Figure 2-5 Effect of Ar Diluent on SiC Microstructure 2-10

63 Radionuclide Release From HTGR Cores Figure 2-6 Effect of Ar Diluent on SiC Material Properties 2-11

64 Radionuclide Release From HTGR Cores The TRISO-P particle (Ref. 2-9) was adopted as the reference particle for gas-cooled New Production Reactor. The TRISO-P design featured both a significantly thicker and denser IPyC layer and an added porous protective (P-PyC) outer layer. Both design changes were made to solve perceived problems during fuel fabrication. The IPyC layer was thickened and made less porous (and more anisotropic) to improve the quality of SiC coating by reducing the potential for heavy metal dispersion. The outer P-PyC layer was added to reduce the potential for introducing SiC defects from particle-to-particle contact during compacting. The design changes resolved these process issues, and the as-manufactured quality of the fuel compacts was dramatically improved. However, under irradiation the thicker (and more anisotropic) IPyC developed radial cracks which served as stress risers in the SiC layer, and the porous P-PyC layers shrank excessively which caused a high fraction of the OPyC layers to fail. The combined result of these design improvements was an order-of-magnitude increase in the in-service failure rates compared to that of conventional U.S.-made TRISO particles even though the as-manufactured quality had been much improved Kernel Composition The fissile and fertile materials in an HTGR fuel element are present in the form of dense microspheres ( kernels ) which are the first barriers to fission product release in the multiplebarrier radionuclide containment system (Section 2.2). Coated-fuel particles with a variety of kernels compositions, including UO 2, UCO, UC 2, ThO 2, ThC 2, (U,Th)O 2, (U,Th)C 2, PuO 2-x and (Th,Pu)O 2-x, and enrichments have been fabricated and tested; but the major emphasis here will be placed upon LEU UCO which is the reference kernel for the commercial GT-MHR and on LEU UO 2 which is the reference kernel for the PBMR. The PuO 2-x kernel is also of interest since it is the reference kernel for the International GT-MHR program which uses excess WPu the feedstock. As illustrated in Table 2-1, selection of the kernel composition is a tradeoff since each type has distinct advantages and disadvantages. Carbide-based kernels, which were used in Peach Bottom, Fort St. Vrain and initially in AVR, permit high burnups and are thermally stable; however, they readily hydrolyze when exposed to water with increased fission gas release, and they do not retain fission metals well. Oxide-based fuel kernels, such as UO 2, are (relatively) easy to fabricate, best retain metallic fission products, and are most resistant to corrosion by water or air. However, with pure UO 2, CO is formed at higher burnups which contributes to the internal gas pressure, and CO may corrode the SiC layer at accident temperatures; UO 2 also has the greatest potential for kernel migration (a failure mechanism described in the next section). These disadvantages of UO 2 can be eliminated by including a fraction of UC 2 in the kernel with the optimal proportion of UC 2 being dependent the maximum design burnup. Given the importance of the kernel chemistry on the performance of coated-particle fuel, it is discussed in greater detail below. 2-12

65 Radionuclide Release From HTGR Cores Table 2-1 Key Features of Candidate Kernel Compositions Kernel Type Key Advantages Key Disadvantages UC 2 UO 2 UCO (UC 2 /UO 2 mixture) No CO formation (minimum internal gas pressure) Highest U loading Insignificant kernel migration Permits high burnup Highest retention of metallic fission products Most resistant to hydrolysis Easy fabrication by sol gel processes Large body of successful irradiation and postirradiation heatup data for high quality LEU fuel Negligible CO formation Good retention of metallic fission products Retention of La fission products reduces SiC corrosion Less susceptible to hydrolysis than UC 2 Permits high burnup Hydrolyzes when exposed to H 2 O with increased FP release Fabrication requires very high temperature Lowest retention of metallic fission products Release of lanthanide fission products enhances SiC corrosion CO formation (maximum internal gas pressure, possible SiC attack) Lowest U loading Highest kernel migration Apparent burnup limit Careful control of kernel stoichiometry required More reactive with HCl gas formed during SiC coating phase Eu isotopes released from UCO The initial composition of the fuel kernel influences the degree of fission product retention for a number of reasons: (1) it determines the extent to which carbon monoxide will be generated as function of burnup; (2) it determines the potential for kernel/coating interactions; and (3) it establishes the chemical forms of key fission products which determine their mobility in the kernel and coatings U-C-O Phase Equilibria The solid-state phase equilibria in the U-C-O system are a major controlling parameter in the production of the fuel kernel (especially for UCO kernels) and are important to its performance during normal operation and most accident conditions. The phase relationships in the U-C-O system have been studied (e.g., Refs through 2-15); Appendix A of Reference 2-16 provides an excellent summary of the available information; however, it is not generally available so it is excerpted here. 2-13

66 Radionuclide Release From HTGR Cores The 1573 K isothermal section of the U-C-O ternary phase diagram proposed by Javed (Ref. 2-11) is shown in Figure The fuels indicated by the numbered circles are fully dense LEU fuels with different C/O ratios irradiated to maximum burnups of <28% FIMA, and UO 2 is shown on the abscissa. The vectors drawn from the starting composition points represent the change in fuel composition from the loss of uranium from fissioning, without considering the effects of the fission product elements on phase composition. The phase designated UC 1-y O y in the diagram refers to uranium monoxycarbide, which is the oxycarbide analog of UC. The complexity of this phase diagram derives partly from the nature of the 1573 K isotherm of the two solid-state binary phase diagrams, U-C and U-O, upon which the ternary diagram is based. The diagram presented in the figure is incomplete because no phases are indicated for the high carbon and oxygen region. However, the only solid phases possible in this region are UO 2 and carbon; gaseous CO and CO 2 will also be found in this region. During high-temperature, fuel heat fabrication treatments, such as sintering, the CO partial pressure is used to control kernel C/O ratios in the production of uranium oxycarbide fuels (e.g., Ref. 2-16). The equilibrium CO partial pressure over the three-phase region (UO 2, UC 1.92 O 0.08, C) in which the fuel will start operating will be insensitive to the phase composition at a given temperature. Because the fuel kernel will be surrounded by carbon in a closed chemical environment within the coated-fuel particle, it is best represented as an equilibrium mixture of the phases UO 2, UC 1.92 O 0.08, and carbon. Postirradiation electron microprobe examination of oxycarbide fuels indicates that the UO 2 phase strongly retains the rare-earth fission products (cerium, lanthanum, neodymium, etc.) as oxides (e.g., Ref. 2-16). Zirconium, strontium, and barium are also retained as oxides in the UO 2 phase in fuels with starting composition >90 mol% UO 2, but, increasingly, these fission products are found in the carbide phase and the densified buffer carbon phase as carbides when the starting kernel composition shifts toward larger amounts of UC 2. Molybdenum, ruthenium, and technetium were present as metallic inclusions at the surface of fission gas bubbles within the UO 2 phase and at the surface of the UO 2 phase in fuels with high (>85 mol%) UO 2 starting composition. Relatively low concentrations of these fission products were found within the bulk of the oxide and carbide phases, although high concentrations were observed in the densified buffer. With increasing UC 2 fraction in the starting fuel composition, the concentration of these fission products tended to increase in the oxide and carbide phases, and the distribution tended to become more uniform within the phases. At the same time, the concentration of these fission products within the densified buffer decreased. For all compositions, xenon collected in the porosity in the oxide phase, and cesium was absent from the oxide phase and found to a limited extent in the carbide phase, most strongly in the densified buffer. Fission gas release data for UO 2 kernels that were fully dense at the beginning of life indicate that the porous structures that develop at higher burnups do not retain gaseous fission products very well. For all compositions, palladium was observed to migrate out of the kernel and accumulate at the SiC coating on the cool side of the particle (Ref. 2-16). 11 The U-C-O phase diagram is applicable to pure UO 2 kernels as well because the oxide kernel is surrounded by carbon in a closed chemical system; however, with dense UO 2 kernels the time to approach equilibrium will probably be longer than for UCO kernels wherein the oxide and carbide phases are intimately mixed. 2-14

67 Radionuclide Release From HTGR Cores Figure K Isotherm of the U-C-O System Carbon Monoxide Generation With stiochiometric oxide fuel particles, including UO 2, carbon monoxide is produced from excess oxygen liberated upon fissioning of the heavy metal because the fission products in the aggregate are thermochemically incapable of binding all of the oxygen. One effective way to control the CO pressure within uranium fuel particles (and therefore kernel migration) is to maintain carbides within the kernel that can be oxidized in preference to elemental carbon. Each U-235 fission within UO 2 leads to fission products that, at maximum (assuming the formation of oxides of yttrium, cerium, lanthanum, neodymium, praseodymium, 2-15

68 Radionuclide Release From HTGR Cores samarium, zirconium, strontium, europium, and barium), may combine with only ~1.62 of the two oxygen atoms released, leaving, at a minimum, 0.38 atoms available to oxidize other materials, such as carbon or carbides. The equilibria of uranium and fission product oxides with their carbides, plotted as a function of oxygen potential in Figure 2-8 (Ref. 2-17), show the stabilities of the oxides relative to one another and to CO. In this plot, the equilibria at the lowest oxygen potentials (the lowest lines) represent the most stable oxides (least stable carbides). That is, the rare earths yttrium, cerium, lanthanum, neodymium, and praseodymium are the most stable oxides, whereas barium is the least stable fission product oxide (the most stable fission product carbide) on the plot. Note that SiC is very stable. The equilibrium between C, CO 2 and CO has a negative slope with temperature, and therefore the oxidation of carbon to CO is more favored than the oxidation of BaC 2 to BaO above 1700 K and more favored than the oxidation of ZrC to ZrO 2 above 1900 K. Figure 2-8 Oxygen Potential vs. Temperature for Various Oxide-Carbide Equilibria Oxycarbide fuel is designed such that UC 2 is converted to UO 2 from the reaction with O 2 liberated by fissioning of UO 2. The oxygen potential is fixed by the UC 2 /UO 2 equilibrium, meaning that rare-earth fission products will form oxides and the fission products zirconium, strontium, europium, and barium will form carbides. After exhaustion of UC 2, the oxygen potential in the kernel will shift upwards to the ZrC/ZrO 2 equilibrium. Figure 2-8 indicates that CO can be formed at temperatures less than 1700 K only after total conversion of BaC 2 to BaO. Note that SiC oxidation to SiO 2 can occur at oxygen potentials close to the BaC 2 /BaO equilibrium. 2-16

69 Radionuclide Release From HTGR Cores The phases present in the kernel are a function of both the starting UC 2 /UO 2 ratio in the kernel and the burnup, as is demonstrated in Figure 2-9 (Ref. 2-17). 12 Here it can be seen that the oxygen potential in the kernel increases with burnup for a given initial kernel composition. For example, with an initial kernel composition of 90 mol% UO 2 and 10 mol% UC 2, the oxygen potential increases from the UC 2 /UO 2 equilibrium up to 18% FIMA, to SrC 2 /SrO from 18 to 20% FIMA, to ZrC/ZrO 2 from 20 to 45% FIMA, to BaC 2 /BaO from 45 to 55% FIMA, to CO production above 55% FIMA. With an initial kernel composition of 80 mol% UO 2 and 20 mol% UC 2, the oxygen potential at 75% FIMA is controlled by the ZrC/ZrO 2 equilibrium and significant quantities of CO are not predicted. The suppression of excessive CO formation during irradiation of oxide-based fuel particles is an extremely important particle design objective, especially for higher burnups. Consequently, the thermochemical calculations presented above need to be confirmed experimentally by measuring the CO contents of irradiated fuel particles over a range of burnups. Such measurements have been made on irradiated UO 2 and (Th,U)O 2 particle (Refs and 2-19, respectively), and they generally confirm the thermochemical calculations. Such CO measurements have not yet been made on high burnup LEU UCO fuel, but they are planned as part of the AGR fuel program (Ref. 2-8). Additionally, the Russians are planning to measure the CO inventories of highburnup, substoichiometric PuO 2-x particles to confirm the expected low values Fuel Failure Mechanisms During the past four decades of coated-particle fuel development and demonstration, a number of mechanisms have been identified - and quantified which can compromise the capability of the coated fuel particles to retain radionuclides (i.e., functional failure of the coated particle). A considerable number of documents have been prepared on the topic of coated particle failure mechanisms (e.g., Refs through 2-26). 13 TECDOC-978 (Ref. 2-26) provides a good summary along with an extensive bibliography. The following failure mechanisms have been identified as capable of causing partial 14 or total failure of the TRISO coating system under irradiation and during postulated accidents; these mechanisms are shown schematically in Figure Phenomenological performance models, typically inspired by first principles and correlated with experimental data, have been developed to model each of these mechanisms. Design methods incorporating these models have been developed to predict fuel performance and fission product release from the reactor core to the 12 The illustrations given here are for HEU fuel (U-235 fission yields, etc.), but the underlying thermochemistry and qualitative trends are the same for other fuels, including LEU UCO and substiochiometric PuO 2-x. 13 The September, 1977, issue of the journal Nuclear Technology was devoted to the topic of coated-particle fuel; it provides an excellent summary of the then available international information on the irradiation performance of coated fuel particles. Practically all of the information in this special issue is still relevant. 14 In a partially failed particle, one or more of the coating layers is defective or has failed in service, but the kernel is still encapsulated by at least one high-density structural coating (i.e. IPyC, SiC and/or OPyC). 2-17

70 Radionuclide Release From HTGR Cores primary coolant; these design methods are described in Section 3, and an example of their application is given in Section 6. Figure 2-9 Phases Present below 1800 K for HEU UCO Kernel 2-18

71 Radionuclide Release From HTGR Cores Figure 2-10 TRISO Particle Failure Mechanisms 1. Coating damage during fuel manufacture, resulting in heavy metal contamination on coating surfaces and in the fuel compact matrix. 2. Pressure vessel failure in standard particles (i.e., particles without manufacturing defects). 3. Pressure vessel failure in particles with defective or missing coatings. 4. Irradiation-induced failure of the OPyC coating; 5. Irradiation-induced failure of the IPyC coating and potential SiC cracking; 6. Failure of the SiC coating due to kernel migration in the presence of a temperature gradient. 7. Failure of the SiC coating caused by fission product/sic interactions. 8. Failure of the SiC coating by thermal decomposition. 9. Failure of the SiC coating due to heavy-metal dispersion in the IPyC coating. It should be noted that other failure mechanisms have been observed experimentally, and yet others have been proposed. As an example of the former, transition metal impurities, especially iron, which have been introduced during processing, have been shown to corrode the SiC coating. As an example of the latter, severely faceted ( flat spots ) particles are predicted to be more susceptible to pressure vessel failure than highly spherical particles. Nevertheless, it 2-19

72 Radionuclide Release From HTGR Cores appears that particles susceptible to these failure mechanisms can be reliably detected with existing QC methods and practically eliminated by appropriate processing; consequently, they are not included here. It is also possible that, if the failure mechanisms listed above were practically eliminated by design and/or process improvements, other second-order failure mechanisms might be revealed and become dominant, albeit at a lower frequency of occurrence. The first mechanism listed above as-manufactured heavy-metal contamination - is not an in-service failure mechanism per se but rather an extreme case of as-manufactured coating defects whereby trace amounts of heavy metal are not encapsulated by a single intact coating layer (analogous to tramp uranium in LWR fuel). Modern fuel product specifications only allow small fractions of HM contamination (~10-5 is typical); nevertheless, it is an important source of fission product release which will be addressed in Section As elaborated below, the next four failure mechanisms are structural/mechanical mechanisms, and the latter four are thermochemical mechanisms. Before addressing individual mechanisms, it should be noted that they can occur in isolation or in combination. As an example of the former, a particle whose SiC coating has thermally decomposed at very high temperature does not typically undergo pressure-vessel failure of the PyC coatings; apparently, the PyC becomes sufficiently porous that the internal gas pressure is relieved (an example of leak-before-break ). In contrast, a particle which experiences irradiation-induced failure of its IPyC layer and/or OPyC layer has a much greater probability of experiencing complete pressure-vessel failure of the remaining coatings (which is what happened with the TRISO-P particle, Section 2.3.1) Structural/Mechanical Mechanisms During irradiation, long-lived and stable fission gases are released from the kernel into the buffer, which increases the internal gas pressure. As described above, carbon monoxide can also be generated during irradiation for some particle designs, which further increases the gas pressure. Because the SiC layer has a much higher elastic modulus than the pyrocarbon layers, 15 it bears most of the internal pressure force, which produces a tensile stress. However, the pyrocarbon layers undergo shrinkage during irradiation, which produces compressive forces in the SiC layer. The compressive forces from pyrocarbon shrinkage more than compensate for the tensile stresses from internal pressure, such that the SiC remains in compression, provided at least one of the pyrocarbon layers remains intact. From a structural/mechanical perspective, the SiC layer will remain intact provided that: (1) it remains in compression, or (2) the tensile stress in the SiC layer does not exceed its strength. 15 In other words, SiC is much stiffer than pyrocarbon. Because of this property, it is reasonable to assume the IPyC and OPyC are isolated from each other when evaluating performance of these layers and overall performance of the TRISO coating system. 2-20

73 Radionuclide Release From HTGR Cores Pyrocarbon Performance Shrinkage of the pyrocarbon layers during irradiation is a favorable attribute, in terms of the compressive forces applied to the SiC layer, as long as the pyrocarbon layers remain intact. However, pyrocarbon shrinkage produces tensile stresses in the pyrocarbon layers themselves, which can lead to failure of these layers. The strains, stresses and creep generated in the pyrocarbon layers are complex functions of fast neutron fluence, irradiation temperature, and coating material properties. Cracking and differential shrinkage of the PyC layers can impose high local stresses on the SiC layer, depending on the local bond strength between the PyC and SiC layers, which can lead to through wall SiC cracks. As discussed above, a property that greatly affects pyrocarbon performance is anisotropy which is usually expressed in terms of the optical Bacon Anisotropy Factor (BAF o, see Section for definition). For a perfectly isotropic material, BAF o = 1, and for a perfectly oriented medium, BAF o =. Pyrocarbon layers are able to perform well out to high fast neutron fluences because the irradiation-induced strains and stresses are relaxed to some extent by irradiation-induced creep (Ref. 2-27). Unfortunately, the measured data for pyrocarbon creep coefficients are badly scattered. Reference 2-28 provides a useful summary of the available PyC material property data Pressure-Vessel Failure Pressure-vessel models with varying degrees of complexity have been proposed to predict the structural performance of coated-particle fuel (e.g., Refs and 2-30). In the absence of compressive forces from the pyrocarbon layers, the tensile stress σ SiC in the SiC layer may be calculated with reasonable accuracy using the spherical thin-shell approximation: where: P rsic σ SiC =, Equation 2-1 t 2 SiC P internal pressure inside the particle r SiC radius to the middle of the SiC layer t SiC thickness of the SiC layer Pressure vessel failure occurs when σ SiC exceeds the strength of the SiC layer. The fraction of particles with a failed SiC coating 16 (f SiC ) is calculated using Weibull statistical strength theory (which represents a distribution of SiC strengths within the particle population). Assuming volume flaws and a uniform stress distribution in the SiC layer, the quantity f SiC is determined from: 16 This fraction applies to the population of particles that have a failed IPyC layer and a failed OPyC layer. 2-21

74 Radionuclide Release From HTGR Cores where: m σ SiC f SiC 1 exp V σ o = SiC, Equation 2-2 σ o m Weibull characteristic strength (MPa) Weibull modulus V SiC volume of the SiC layer. Typical strength values of σ o = 96.3 MPa and of m = 6 are given in Ref Figure 2-11 is an example of a particle that has undergone pressure-vessel failure. Figure 2-11 Pressure Vessel of UO 2 Particle in HRB-8 Irradiation 2-22

75 Radionuclide Release From HTGR Cores In fact, multi-shell structural models of varying degrees of sophistication have been available for the past four decades to predict the mechanical performance of coated particle fuel (e.g., Ref. 2-29). These models, which are solved numerically, typically treat all of the coating layers and the interactions between them. These models are discussed in more detail in Section 3. These numerical models developed to predict pressure vessel failure typically use the thick-shell equation and a Weibull distribution to characterize the failure probability of the SiC layer. These models have the following functional form (which is somewhat different from the thin-shelled approximation given above): where the variables are as defined above. f SiC = 1 - exp [-ln2(σ o /σ SiC ) m ], Equation 2-3 Since an HTGR core typically contains ~10 10 coated particles and since 100% QC of finished fuel compacts or fuel spheres is not practical (at least not with the current QC methods), statistical sampling methods are used; consequently, fuel product specifications typically allow small fractions of various coating defects (e.g., missing buffer layers). Not surprisingly, these defective particles are more susceptible to in-service pressure-vessel failure. Consequently, for core design purposes, structural analysis models for predicting pressure-vessel failure probabilities are formulated not only for standard particles (particles with no defective coatings) but also for particles with various kinds of as-manufactured coating defects and in-service coating failures, including: (1) particles with missing buffer coatings, and (2) particles with failed OPyC but intact SiC coatings. Missing-buffer particles have the highest probability of pressure-vessel failure so there is great incentive to limit this type of defect OPyC Irradiation-Induced Failure The irradiation performance of PyC coatings is dependent upon the anisotropy because anisotropic PyC coatings shrink excessively under neutron irradiation and may fail as a consequence (e.g., Ref. 2-28). For OPyC coatings, excessive irradiation-induced shrinkage results in high failure rates, thereby eliminating a compressive stress imposed upon the SiC coating which serves to counteract a tensile stress component induced by the internal gas pressure. For OPyC coatings, there is a complicating factor as a result of possible matrix/coating interaction. If the matrix, which shrinks considerably under irradiation, is too tightly bonded to the OPyC coating, it will tear the OPyC coating off the SiC coating. Open surface porosity in the OPyC layer encourages an excessively strong bond with the matrix; hence, it must be controlled. Examples of OPyC failure as a result of matrix/coating interaction is shown in Figure

76 Radionuclide Release From HTGR Cores IPyC Irradiation-Induced Failure For IPyC coatings the irradiation-induced failure mode is typically radial cracks which serve as stress risers for the SiC layer which may result in cracking of the SiC layer or pressure-vessel failure of the entire coating system and may serve as pathways for fission product release and for corrosive agents, including CO in UO 2 particles, to attack the SiC coating. An example of an irradiation-induced radial crack in a TRISO-P particle is shown in Figure Figure 2-12 OPyC Failure from Matrix-Coating Interaction 2-24

77 Radionuclide Release From HTGR Cores Figure 2-13 Radial Crack in IPyC Layer of TRISO-P Particle Thermochemical Mechanisms Fuel failure caused by thermochemical mechanisms can be controlled in large measure through the nuclear and thermal-hydraulic design of the reactor core. In order for the fuel to satisfy performance criteria, peak fuel temperatures must be kept sufficiently low, and the fraction of fuel that experiences relatively high temperatures for long time periods must be kept sufficiently small. Thermochemical failure mechanisms that have been observed in coated-particle fuel are described below Kernel Migration Local fuel temperatures and temperature gradients across the fuel compact can be relatively high when the reactor is producing power. Under these conditions, oxide and carbide fuel kernels can migrate up the temperature gradient, as shown in Figure This phenomenon is often referred to as the amoeba effect and can lead to complete failure of the coating system, especially if the kernel migrates to the point of contacting the SiC coating. For carbide kernels, migration is caused by solid-state diffusion of carbon to the cooler side of the kernel (Ref. 2-31). For oxide kernels, migration may be caused by carbon diffusion or gas-phase diffusion of CO or other gaseous carbon compounds (Ref. 2-32). Empirically, it is observed that if CO formation is suppressed, kernel migration is also suppressed (Ref. 2-33) 2-25

78 Radionuclide Release From HTGR Cores Figure 2-14 Kernel Migration in High-Burnup Oxide Fuels Failure by this mechanism is correlated as a function of temperature, temperature gradient, and thicknesses of the buffer and IPyC layers. Failure is assumed to occur when the migrating kernel contacts the SiC layer. The particle-to-particle variations in the buffer and IPyC thicknesses (expressed as normal distributions with measured variances) are accounted for when calculating the failure probability. The kernel migration rate is calculated according to: where: Θ x& KM = KMC(T) Equation T x& KM kernel migration rate (m/s) KMC kernel migration coefficient (m 2 -K/s) Θ T temperature gradient across the particle (K/m) absolute temperature (K) The kernel migration coefficient (KMC) for each kernel type is typically correlated as an Arrhenius function of temperature. In some oxide fuels, including ThO 2, there is an incubation period prior to the onset of measurable kernel migration which may correspond to the onset of CO generation (Ref. 2-34). Kernel migration coefficients for a variety of kernel compositions can be found in Ref

79 Radionuclide Release From HTGR Cores Chemical Attack of SiC Noble metals (e.g., Ru, Rh, Pd, and Ag) are produced with relatively high yields during fission of uranium and plutonium fuels. During irradiation, the thermochemical conditions are not conducive for these elements to form stable oxides, and they can readily migrate out of the fuel kernel, regardless of its composition. Reactions of SiC with Pd have been observed during postirradiation examinations (PIEs) of TRISO fuel (e.g., Refs and 2-37). Although the quantity of Pd is small compared with the mass of the SiC layer, the reaction is highly localized, and complete penetration of the SiC layer can occur if high temperatures are maintained for long periods of time (see Figure 2-15, Ref. 2-38). The cumulative fission yield of long-lived and stable Pd isotopes in Pu-239 is more than an order of magnitude higher than that in U-235; consequently, this failure mechanism may be more of a concern for LEU and Pu fuels than for HEU fuels. Several correlations have been proposed for calculating the probability of SiC failure caused by fission-product attack (e.g., Refs and 2-40). The 1985 Goodin-Nabielek correlation (Ref. 39) is a fairly complex, empirical function of temperature, prior irradiation-temperature history, temperature gradient, burnup, and fast fluence. The basis for this correlation is a statistical analysis of data obtained from postirradiation heating of reference German LEU (10.6% enriched) TRISO UO 2 fuel. Figure 2-15 SiC Corrosion by Fission Products 2-27

80 Radionuclide Release From HTGR Cores Chemical attack of the SiC layer by CO has been observed in UO 2 particles irradiated at temperatures above approximately 1400 C (Ref. 2-41). As shown in Figure 2-16, degradation occurred near locations where the IPyC layer was cracked. The kernels of particles with degraded SiC layers were examined by electron microprobe, which showed the presence of silicon in the form of fission-product silicides. The authors performed thermochemical calculations which supported their hypothesis that silicon is transported to the kernel in the form of SiO gas, which then reacts with fission products Thermal Decomposition of the SiC Layer At very high temperatures, SiC will decompose into its constituent elements; the silicon vaporizes, leaving a porous carbon structure. As shown in Figure 2-17, the coating system remains ostensibly intact. The PyC coatings do not typically undergo pressure-vessel failure; apparently, the PyC becomes sufficiently porous that the internal gas pressure is relieved. Figure 2-16 CO Corrosion of SiC Co 2-28

81 Radionuclide Release From HTGR Cores Figure 2-17 Effect of High Temperature Heating on TRISO Particle Correlations for thermal decomposition are part of the 1985 Goodin-Nabielek model (Ref. 2-39); as with the SiC corrosion model, this correlation is an empirical function of temperature, prior irradiation-temperature history, temperature gradient, burnup, and fast fluence. The basis for this correlation is a statistical analysis of data obtained from postirradiation heating of reference German TRISO-coated LEU UO 2 fuel. Based on calculations performed for previous core designs, thermal decomposition is not an important contributor to fuel failure at normal operating temperatures. However, relatively high failure rates can occur if temperatures higher than 1700 C to 1800 C are maintained for extended periods of time, and thermal decomposition of SiC occurs rapidly at temperatures above 2000 C Heavy-Metal Dispersion Heavy metal dispersion results when a defective or porous IPyC layer allows HCl produced during the SiC coating deposition to react with heavy metal in the fuel kernel to form volatile heavy-metal chlorides which are in turn transported out of the kernel into the buffer and IPyC layers (Ref. 2-37). Particles with heavy metal dispersed in the buffer and IPyC are observed to exhibit enhanced SiC attack by fission products and SiC coating failure. Due a paucity of quantitative data, the failure probability due to heavy-metal dispersion is correlated as a function 2-29

82 Radionuclide Release From HTGR Cores that increases linearly with burnup, with the maximum failure probability assumed to be 0.5 at full design burnup. With proper product and process specifications for the IPyC layer, heavy-metal dispersion is reduced to insignificant levels (Ref. 2-42). It has been speculated that the potential for heavymetal dispersion is further reduced by the use of a lower SiC deposition temperature which would reduce HCl transport and reaction rates with the kernel as well as the vapor pressures of the resulting HM chlorides. The basis for this speculation is that HM dispersion is apparently not seen in German TRISO-coated particles for which the SiC is deposited at ~1500 o C but is occasionally seen in GA particles for which the SiC is deposited at 1650 o C even though the former s IPyC layers are more permeable than the latter s. The fact that the German UO 2 fuel kernels are typically denser with less surface porosity than the GA UCO kernels may also contribute to reduced transport rates Radionuclide Transport Mechanisms As with fuel particle failure, a number of mechanisms have been identified - and quantified which govern the transport of radionuclides in HTGR core materials, and a large number of documents have been prepared on the topic (e.g., Refs. 2-23, 2-24, 2-43 through 2-47). Especially notable is the 1974 Dragon Project Report DP-828, Part III, by Nabielek which provides a comprehensive set of transport models along with analytical solutions for many bounding cases (Ref. 2-43); this report remains as useful today as it was three decades ago despite the development of numerical methods for predicting fission product transport. Once again, TECDOC-978 (Ref. 2-26) provides a good summary of radionuclide transport phenomena in HTGR core materials along with an extensive bibliography. The transport of radionuclides from the location of their birth through the various material regions of the core to their release into the helium coolant is a relatively complicated process. The principal steps and pathways are shown schematically in Figure For a pebble-bed core, those steps related to transport across the gap between the fuel compact and the fuelelement and transport in the fuel-element graphite are eliminated. Also for certain classes of radionuclides, some steps are eliminated (e.g., noble gases are not diffusively released from intact TRISO particles, but noble gases are not significantly retarded by the compact matrix or fuel-element graphite). 2-30

83 Radionuclide Release From HTGR Cores Figure 2-18 Principal Steps in Radionuclide Release from an HTGR Core As discussed in Section 2.2, the two dominant sources of fission product release from the core are as-manufactured heavy metal contamination and particles whose coatings fail in service. The latter source can be subdivided into (1) coating failure during normal operation and (2) incremental coating failure during core heatup accidents. In addition, certain volatile fission metals (notably Ag) can, at sufficiently high temperatures and long times, diffuse through the SiC coatings of intact TRISO particles. Expressed in the simplest terms, the fractional release of a radionuclide from the core is given by the following relationship: (f.r.) core C(f.r.) + F(f.r.) + [1 C F](f.r.) c F D = Equation 2-5 AFgraphite where: (f.r.) core fractional release from core C heavy-metal contamination fraction (f.r.) C fractional release from contamination F failure fraction 2-31

84 Radionuclide Release From HTGR Cores (f.r.) F fractional release from failed particles (f.r.) D fractional diffusive release from intact particles AF graphite graphite attenuation factor 17 In reality, the problem of calculating the full-core fractional release is much more complicated than implied by Equation (2-5). For example, the fissile and fertile particle failure fractions are generally different and vary in space and time, the fractional releases from contamination and failed particles and graphite attenuation factors vary in space and time, and "partially" failed particles (i.e., particles with a failed SiC coating but with intact inner and/or outer pyrocarbon coatings) must also be considered. Full-core computer codes (Section 3) are needed to keep track of all these effects; nevertheless, the results given by Equation (2-5) are quite intuitive. As implied by Equation (2-5), radionuclide transport must be modeled in the fuel kernel, in the particle coatings, in fuel-compact matrix, and fuel-element graphite. While the actual radionuclide transport phenomena in an HTGR core are complex and remain incompletely characterized after four decades of modeling efforts, the basic approach remains unchanged; radionuclide transport is essentially treated as a transient solid-state diffusion problem with various modifications and/or additions to account for the effects of irradiation and heterogeneities in the core materials. The point of departure is Fick s first law of diffusion (Ref. 2-48) which in slab geometry simply postulates that the diffusive flux is proportional to the negative concentration gradient with the proportionality constant being the diffusion coefficient: c φ = D, Equation 2-6 x The mass balance equation in slab geometry, ignoring for the moment source and radioactive decay, relates the difference in of flux of mass transport across a differential volume to the change in concentration in the volume: c φ = t x, Equation 2-7 Substituting Eqn (2-7) into Eqn (2-6) gives the transient diffusion equation, which is usually called Fick s second law of diffusion: c = t c D x x, Equation Graphite attenuation factor = fission product release from fuel compact/release into coolant. 2-32

85 Radionuclide Release From HTGR Cores where the variables in Eqns. (2-6) through (2-9) are: φ diffusive flux of atoms (atoms/cm 2 -s) D diffusion coefficient (cm 2 -s) c concentration of atoms (atoms/cm 3 ) x spatial coordinate (cm) t time (s) For solid-state diffusion processes, the diffusion coefficient D is typically observed to be exponentially dependent upon the absolute temperature, and the familiar Arrhenius equation is generally used to fit the data: where: ( E / RT), D o preexponential factor (D for T ) (cm 2 /s) E activation energy (cal/mole) T absolute temperature (K) R gas constant (= cal-mole/k) D = Do exp Equation 2-9 For the present problem, radionuclides are produced by various nuclear reactions in the core, and the radiologically significant ones decay so that the mass balance equation (2-8) is generalized to include a source term and a decay term (e.g. Ref. 2-43): c c = D λc+ s t x x, Equation

86 Radionuclide Release From HTGR Cores where: λ decay constant (s -1 ) s volumetric source rate (atoms/cm 3 -s) To model the radionuclide transport in an HTGR core, the balance equations become more complicated by the addition of decay chains and the appropriate boundary and interface equations; nevertheless, the essence of these models, which are described in Section 3, is captured in Eqn (2-10). The transport of the various classes of radionuclides (Table 2-2) in the kernels, coatings, matrix and graphite are considered in the following subsections. Table 2-2 Classes of Radionuclides of Interest for HTGR Design Form in Fuel Key Nuclide Breathing DCF (Sv/Bq) Form in Fuel Special Behavior Noble Gases Xe-133 ~ 0 Element (Gas) Retained by PyC Halogens I x 10-7 Element (Gas) Retained by PyC Alkali Metals Cs x10-8 Oxide-Element Tellurium Group Te x10-8 Complex Mobile Alkaline Earths Sr x 10-7 Oxide-Carbide Noble Metals Ag-110m 1.2 x 10-7 Element Alloys/Permeates Lanthanides La x 10-9 Oxide Actinides Pu x 10-4 Oxide-Carbide Retained by Graphite Radionuclide Transport in Fuel Kernels In general, the release of fission products from the fuel kernel is treated as a transient diffusion process. In spherical geometry, the generalized continuity equation (2-10) takes the form: c 1 c = D λc + s 2 t r r r where r is the radius of the sphere., Equation

87 Radionuclide Release From HTGR Cores The radiologically significant fission gases (e.g., 2.8-hr Kr-88) are typically short-lived, the production rates and release rates reach equilibrium quickly, and the steady-state, core service conditions change relatively slowly by comparison; consequently, steady-state approximations are typically used in core design and analysis methods for fission gas release during normal operation (e.g., Ref. 2-44); these approximations are described below. In contrast, the key fission metals (e.g., 30-yr Cs-137) are typically long-lived, and transient analysis methods are necessary Fission Gas Release from Fuel Kernels The release of fission gases from heavy-metal contamination and from fuel kernels is typically expressed in terms of the release rate-to-birth rate ratio (R/B); at steady-state, the R/B ratio is numerically equal to the fractional release. At steady-state, the R/B for isotope i of element j is calculated from the following relationship (e.g., Ref. 2-44): where: 1 2 R Equation 2-12 B ji λ λ i ξ j i ξ j ( ) ξ coth λi j = 3, (R/B) ji fractional release for isotope i of element j from as-manufactured HM contamination or failed fuel particles λ i decay constant for isotope i of element j (s -1 ) ξ j diffusion parameter for chemical element j (s -1 ) Equation (2-12) has been derived and discussed by several authors, but normally it is attributed to Booth (Ref. 2-49). It is the steady-state solution of the diffusion equation (Equation 2-11) in spherical coordinates assuming (1) constant birth rate, and (2) zero concentration at the surface of the sphere. The diffusion parameter ξ j is defined as ξ j = D j /a 2 where D j is a diffusion coefficient characterizing release of a fission gas and where a is a characteristic distance through which the fission gases diffuse before release. The values of D j and a are never used directly in core analysis. Rather, the fractional release (R/B) is determined from in-pile measurements, and this value is substituted into Equation 2-12 from which ξ j is calculated. The fractional release of a radioisotope of a given element is dependent on its decay constant; for fractional releases (R/Bs) less than ~0.1 (i.e., small values of ξ/λ), Equation 2-12 can be approximated by: ξ j ( ) = 3, R Equation 2-13 B ji λi 2-35

88 Radionuclide Release From HTGR Cores The gas release models give the R/Bs from contamination and failed particles as a function of chemical element, isotope half-life, temperature, and burnup. These functional dependencies are determined experimentally. At high irradiation temperatures, experimental fission gas release measurements show that the R/B is proportional to the square root of isotope half life as required by Eqn (2-13); an example is given in Figure 2-19 which shows the measured fission gas release from Peach Bottom Core 2 (BISO-coated HEU (Th,U)C 2 fuel) and as-manufactured Fort St. Vrain fuel (TRISO-coated HEU (Th,U)C 2 fuel). Figure 2-19 Isotopic R/Bs vs. Radioactive Half Life While this observation is consistent with a diffusive release mechanism, the actual release mechanism is evidently much more complicated. For example, some fission fragments born near the exterior surface of the kernel are released by recoil, and high-energy neutrons produce irradiation-induced defects and other structural changes. In addition, gaseous fission products may become trapped at intrinsic or irradiation-induced defects to form bubbles which themselves migrate. In fact, at lower irradiation temperatures (say, <1000 o C) and at very high neutron fluxes, significant deviations from the square root-of-half life dependence are often observed. Moreover, at lower temperatures, the R/Bs tend to plateau and become almost athermal as indicated in Figure 2-20; this transition as rationalized as the diffusive release component becoming small compared to the recoil component (e.g., Ref. 2-23). 2-36

89 Radionuclide Release From HTGR Cores Figure 2-20 Temperature Dependence of Fission Gas Release The modeling expedient for such complexities has typically been to add additional terms to Equation (2-13). For example, the reference GA gas release model (Ref. 2-50) starts with ( R ξ j ) = 3 f ( T) f (Bu), B ji ( R ξo ) ( ) exp [, j E A + 1 A 1+ σ Bu n ], B ji where: λ i = i R T T Equation 2-14 λ o Bu burnup (% FIMA) A, T o, σ, n constants The values of the constants must be determined experimentally for each gaseous fission product element (Kr, Xe) and for each fuel kernel composition (Ref. 2-50). 2-37

90 Radionuclide Release From HTGR Cores The fractional releases of fission gases from exposed fuel kernels under irradiation can be enhanced if the kernel is hydrolyzed by reaction with trace amounts of water vapor which may be present in the helium coolant (e.g., Ref. 2-26). The magnitude of the effect depends upon the kernel composition with UC 2 showing the largest enhancement and UO 2 showing the least; the effect for UCO is intermediate. With kernels containing uranium carbide, the water reacts to convert the carbide phase completely to oxide along with kernel swelling and increased porosity. With UO 2 kernels, a hyperstoichiometric uranium-oxide phase may be formed; in any case, kernel porosity is increased with enhanced gas release. As illustrated in Figure 2-21, 18 the general response of the exposed kernel to the introduction of water vapor consists of three distinct phases: (1) a transient release of stored fission gas with a concomitant increase in the steady-state fractional release, (2) a period of constant steady-state release, and, upon removal of the water vapor, (3) a monotonic decline in the fractional release to prehydrolysis values (or nearly so). Relatively complicated empirical models have been derived from the experimental data which reproduce these transient results (Ref. 2-51); however, for design purposes, a simpler approach is taken in recognition that the transient release phase is short lived, and that a new steady-state R/B is established if the water is persistent, or the R/B returns to its original value if the water is removed. For fully hydrolyzed UCO fuel, a correlation similar to Equation (2-14) is used but with different fitting constants, and a hydrolysis increase factor of 1.7 is included (Ref. 2-50). Figure 2-21 Effect of Hydrolysis on Fission Gas Release 18 In the illustration the exposed kernels were originally LEU UCO (80% UO 2 and 20% UC 2 ), but the UC 2 phase had already been converted to UO 2 by prior water injections so the observed response is effectively that of UO 2. This conclusion is confirmed by other tests with as-fabricated UO 2 which yielded essentially identical results (Ref. 2-26). 2-38

91 Radionuclide Release From HTGR Cores For perspective, it is noteworthy that the technical specifications for HTGR normal plant operation will limit concentrations of oxidants in the primary coolant to very low levels (e.g., for the Fort St. Vrain HTGR the limit was <10 ppm total oxidants, Ref. 2-52). If the technical specification limits are reached, corrective action must be taken, including removal of the oxidants by action of the helium purification system. Moreover, Advanced HTGR designs, including the commercial GT-MHR and the PBMR, are direct-cycle systems rather than steamcycle systems. The probability of a large water ingress with a direct-cycle plant is much lower than for a steam-cycle plant because with the former designs, the secondary water pressures are lower than the primary He pressures, and the heat-exchanger tubes experience much lower temperatures during normal operation. The R/B values for the radioisotopes of krypton and xenon are conveniently measured experimentally; however, the effectivities of noble gases are generally low. In contrast, the effectivities of radioisotopes of the halogens (Br, I) and chalcogens (Se, Te) produced by fission tend to be significantly higher, hence more significantly to reactor design and safety analysis. Experimental measurement of the release rates of these volatile radionuclides is more difficult (e.g., they deposit in sample lines, etc.) but manageable with special techniques. Fortunately, the R/B values for iodine and tellurium have been shown to be slightly lower than that of xenon isotopes with the same half life (Ref. 2-53). Consequently, it is convenient and conservative to assume that iodine and tellurium have the same release characteristics as xenon; by analogy, it is further assumed that bromine and selenium have the same release characteristics as krypton (Ref. 2-44) Fission Metal Release from Fuel Kernels The transport of fission metals through the kernel is modeled as a transient diffusion process (e.g., Refs and 2-46). Reference 2-43 provides a number of analytical solutions for bounding conditions (e.g., constant power and constant temperature), but Equation 2-10 is typically solved numerically with appropriate boundary and interface conditions which vary depending upon whether a bare kernel or an encapsulated kernel in an intact coated particle is being modeled (e.g., Ref. 2-54). For fission metal transport in fuel kernels, the reference GA diffusivity is given by the following equations with D' = D o exp (-Q/RT), Equation 2-15 D o = C 1 [1 + (1 + n) C 2 F n ], Equation

92 Radionuclide Release From HTGR Cores where D' reduced diffusion coefficient (s -1 ), D o a constant (s -1 ), Q T R F activation energy (J/mol), temperature (K), gas constant (8.314 J/mol-K), burnup (% FIMA), C 1, C 2, n constants. The release of metallic fission products is calculated with computer codes (e.g., TRAFIC-FD, Ref. 2-54) which assume that the kernel material is homogeneous. Therefore, it is necessary to determine an effective homogeneous diffusion coefficient. Use of an effective diffusion coefficient in code calculations will result in the same fractional release of fission products from the fuel kernel as was observed in the measurements from which the coefficient was derived. The effective diffusion coefficient is calculated according to the equation where D effective diffusion coefficient (cm 2 /s), D = D'r 2, Equation 2-17 D' reduced diffusion coefficient (s -1 ) given in Equation 2-17, R kernel radius (cm). The reference GA correlations for kernel diffusivities include a very strong burnup dependence; e.g., for Cs transport in UCO and ThO 2 kernels, the diffusivity is a function of burnup to the 4 th power in Eqn (2-16) (Ref. 2-50); this conclusion is derived primarily from measured Cs releases from ThO 2 with burnups from 1 6 % FIMA (Ref. 2-55). Whether this strong burnup dependence really applies to UCO kernels at burnups >20 % FIMA needs further confirmation. As with the transport of fission gases in the kernel, the transport of mobile fission metals, including Cs, Ag, Sr and Eu isotopes, is undoubtedly much more complicated than classical Fickian diffusion. As discussed in Section , the fission product speciation in the kernel changes with burnup, especially with UCO kernels as the oxygen potential changes, and these changes in chemistry could affect the mobility of oxide-forming species, including Cs and Sr. The probable exception is silver which appears to remain in elemental form for all kernel compositions and burnups of interest. 2-40

93 Radionuclide Transport in Particle Coatings Radionuclide Release From HTGR Cores Under normal operating conditions, the fission gases, including iodines, are quantitatively retained by the coatings of an intact TRISO particle. The transport of the volatile fission metals, including Ag, Cs, Sr, and Eu, in the PyC and SiC coatings is also treated as a transient Fickian diffusion process (Eqn 2-10). In this case, the geometry is a spherical shell, and the interface conditions between the layers are assumed to be described by partition factors p (i) : p (i) c (i) = c (i+1), Equation 2-18 where c (i) is the concentration at the outer surface of the i th coating layer, and c (i+1) is the concentration at the inner surface of the adjacent exterior layer. Partition factors are introduced to account for possible discontinuities in the concentration at the interface of adjacent materials with different chemical affinities for the migrating species (e.g., Refs and 2-54). In fact, a constant partition factor is a gross simplification; nevertheless, there are currently no reliable data to even estimate these partition factors, and they are typically set to unity in most modeling work. As with transport in the fuel kernel, radionuclide transport in the PyC and SiC coatings is undoubtedly more complex than homogeneous Fickian diffusion. These apparent migration coefficients are generally structure sensitive which indicates that the migration process is not a simple diffusion process but likely a combination of lattice diffusion, grain boundary diffusion, pore diffusion, etc., complicated further by effects like irradiation-enhanced trapping and adsorption. Consequently, any quoted diffusion coefficients should be called effective diffusion coefficients which implies that the overall migration process can be approximately described by Fick s laws (to paraphrase Ref. 2-43). Perhaps, the most dramatic indication of the limitations of treating metal transport in the coatings as simple Fickian diffusion comes from the high-temperature annealing of individual irradiated fuel particles (e.g., Ref. 2-56). The fission product inventories in individual, Japanese, TRISOcoated UO 2 particles were measured before and after heating at 1700 and 1800 o C. These measurements were made at ORNL using the Irradiated Microsphere Gamma Analyzer (IMGA). They revealed that the fission product release behavior of the individual particles was not uniform (Figure 2-22), and large particle-to-particle variations in the release behavior of Ag- 110m, Cs-134, Cs-137 and Eu-154 were observed. These variations could not be consistently explained by the presence or absence of cracks in the SiC coating layers. To date, no completely satisfactory explanation has been proposed for the observed behavior. Similar particle-toparticle variability has also been observed in IMGA data for other fuel compositions. 2-41

94 Radionuclide Release From HTGR Cores Figure 2-22 Fission Metal Release from Individual Fuel Particles For all practical purposes, an intact SiC coating retains quantitatively all radionuclides, except silver (and tritium), under normal operating conditions (e.g., Ref. 2-26). It remains to be demonstrated whether the use of a Fickian model with effective diffusion coefficients derived from particle release data will give sufficient accuracy for predicting Ag release when applied to analysis of large populations of particles in irradiation capsules or in reactor fuel elements Radionuclide Transport in Fuel-Compact Matrix As previously stated, the fuel-compact matrix is relatively porous and provides little holdup of the fission gases which are released from the fuel particles, and the effect is generally neglected (e.g., Ref. 2-26). However, the matrix is a composite material which has a high content of amorphous carbon, and this constituent of the matrix is highly sorptive of metallic fission products, especially Sr. While the matrix is highly sorptive of metals, it provides little diffusional resistance to the release of fission metals because of its high interconnected porosity; the matrix of a spherical fuel element is denser and partially graphitized so it does provide more diffusional resistance as discussed in the next subsection. For prismatic designs, fission metal transport in the fuel compact matrix is again modeled as a transient Fickian diffusion process: the transient diffusion equation for cylindrical geometry is solved with an evaporative boundary condition. It is assumed that sorption equilibrium prevails in the gap between the fuel compact and the fuel hole surface of the fuel block. At equilibrium, the vapor pressure in the helium-filled gap and solid-phase concentration on the fuel-compact surface are uniquely related to one another by a sorption isotherm which is determined experimentally. 2-42

95 Radionuclide Release From HTGR Cores Conceptually, the partial pressure of the migrating species in the gas gap is calculated from its concentration on the surface of the fuel compact using a matrix sorption isotherm, and then the equilibrium surface concentration on the fuel-hole surface of the fuel block is calculated from the partial pressure in the gap using a graphite sorption isotherm. In general, the compact matrix material is more sorptive than the graphite so there is a discontinuity in surface concentrations across the gap (a partition factor using the terminology introduced above). The magnitude of the partition factor is less than the ratio of the intrinsic sorptivities of the matrix and graphite because there is also a temperature drop across the gap (say, ~30 o C), and the lower temperature increases the effective sorptivity of the graphite. Several different sorption isotherms have been derived by making various assumptions about the potential energy distributions of the sorption sites which lead to different functional dependencies between the gas-phase partial pressure and the surface concentration; however, for the sorption of fission products on core materials, the experimental data are generally correlated with a simple Henrian isotherm (linear dependence) for low sorbate concentrations and with a Freundlich isotherm (exponential dependence) at higher sorbate concentrations (Refs and Refs. 2-47). Functionally, the partial pressure in the gap is assumed to be the sum of the pressures calculated with the two isotherms: ln P F B E = A + + D + ln Cgr, Equation 2-19 T T B E ln P H = A + + D 1+ lnct + lncgr, Equation 2-20 T T ln P = P F + P H, Equation 2-21 Ct 1 2 = d d T, Equation 2-22 where: P F partial pressure from Freundlich isotherm (Pa) P H partial pressure from Henrian isotherm (Pa) T absolute temperature (K) C gr mass concentration of sorbate (mol/kg) A, D, d 1 constants (dimensionless) B, E constants (K) D 2 constant (K -1 ) 2-43

96 Radionuclide Release From HTGR Cores Sorption isotherms for Cs, Sr and Ag have been measured for a variety of nuclear graphites and matrix materials; the data are summarized in TECDOC-978 (Ref. 2-26). Measurements have been made on both unirradiated and irradiated materials. For matrix materials with a high content of amorphous carbon, irradiation has little effect; however, for highly graphitic materials, the Cs and Sr sorptivities are observed to increase with increasing neutron fluence. Apparently, neutron irradiation of the crystalline component causes damage which serves to create additional sorption sites (e.g., Refs and 2-47). Consequently, the sorption isotherms have been modified to include a fast fluence dependence which is fit to the experimental data (Ref. 2-50) Radionuclide Transport in Fuel-Element Graphite The fuel element graphite, which is denser and has a more ordered structure than the fuelcompact matrix, is somewhat less sorptive of the fission metals than the matrix, but it is much more effective as a diffusion barrier than the latter. Traditionally, the transport of fission metals in the fuel-element graphite and in the outer unfueled shell of a fuel sphere has been modeled as transient Fickian diffusion with an evaporative boundary condition at the coolant interface; the graphite web of a prismatic fuel element is treated as an equivalent slab, and the pebble is modeled in spherical coordinates. As already discussed for transport in other core materials, the transport of fission metals, including Cs, Sr, and Ag, in nuclear graphite is more complex than simple homogeneous Fickian diffusion (e.g., Ref. 2-43). Consequently, other more complicated models have been proposed including a two-phase diffusion model (Ref. 2-57), a coupled fast-slow diffusion model (Ref. 2-58), a diffusion-trapping model (Ref. 2-59), and others. Unfortunately, in all cases, there are insufficient experimental data available to derive the material property data necessary for the models or even to develop reliable empirical fits. At the coolant boundary, the mass flux φ from the surface into the flowing coolant is given by the product of a convective mass transfer coefficient and a concentration driving force which is the difference between the desorption pressure (expressed as a volumetric concentration) and the free stream or mixed mean concentration in the coolant: 2-44 where: φ diffusive flux of atoms (atoms/cm 2 -s) h mass transfer coefficient (cm/s) φ = h(c v - C ), Equation 2-23 C v equilibrium desorption concentration in the boundary layer (atoms/cm 3 ) C mixed mean coolant concentration (atoms/cm 3 ) The equilibrium desorption pressure in the boundary layer is calculated with a sorption isotherm as described in the previous subsection and converted to a volumetric concentration C v using the ideal gas law. The mixed mean coolant concentration C is often conservatively set to zero;

97 Radionuclide Release From HTGR Cores alternatively, a two-dimensional model can be used that integrates the total flux into the coolant as it passes through the fuel-element or core, thereby providing the coolant concentration at each local point. For prismatic fuel elements, the mass transfer coefficient is calculated from an empirical correlation for the Sherwood number. The reference GA correlations for predicting convective mass transfer coefficients for forced convection and free convection are cataloged in the PADLOC code (Ref. 2-60). 19 In general, the Sherwood number is given as functions of the Reynolds, Schmidt, and Grashof numbers. The general forms of these relations are given below. a. forced convection: α3 D d α4 α5 h = α1 + α2 Re Sc, χ L Equation 2-24 and b. free convection: β 5 D β d 2 β3 β 4 h = β1 Re Sc Gr. Equation 2-25 χ L where, h mass transfer coefficient (cm/s) D diffusion coefficient of fission product n helium (cm 2 /s) Re Sc Gr d H Reynolds Number Schmidt Number Grashof Number hydraulic diameter of conduit (cm) L hydrodynamic entry length (cm), and, χ L for flat plates χ d H for other geometries. The coefficients α i and β i for the different flow regimes and geometries are given in the PADLOC user s manual (Ref. 2-60). 19 Convective heat and mass transfer correlations can be found in various editions of standard engineering handbooks (e.g., Perry s Chemical Engineers Handbook, Marks Standard Handbook for Mechanical Engineers, etc.). 2-45

98 Radionuclide Release From HTGR Cores The definitions of the dimensionless groups are as follows: 1. vρd Re = µ H, where Equation v average coolant velocity, (cm/s) ρ helium density, (g/cm 3 ) µ dynamic viscosity of helium, (g/(cm-s) d H 4 A/P, (A = cross section area, P = wetted perimeter). (cm) µ Sc =, ρd Equation ρm 3. Gr = g L ρm - ( ρm ). Equation µ m In Equation (2-28) the index m stands for mixture of fission product and helium coolant, (ρ m ) 4 is the average density in the free stream, and g is the acceleration of gravity. The formula for the diffusion coefficients is D = Tc P 1 M , Equation 2-29 where M is the molecular mass of the fission produce species, and the dynamic viscosity is computed as 20 where: µ dynamic viscosity (g/cm-s) m P helium mass flow rate (g/s) helium pressure (atm) T c helium temperature ( o C) T W wall temperature ( o C) = 5.31x Tc µ 10. Equation Wilson, M. P. Jr., "Thermodynamic and Transport Properties of Helium," GA-1355, General Atomic,

99 With this input, the density is calculated from the perfect gas law, Radionuclide Release From HTGR Cores MP ρ =, Equation 2-31 RT where R = atm.cm 3 /(mol-k), the universal gas constant. The coolant velocity follows from m& v = = ρa R M m& T Aρ. Equation 2-32 The qualitative features of these models for fission metal transport in the fuel compact and graphite web are observed experimentally. For example, consider the transport of cesium in the two graphite fuel bodies of irradiation capsule R2-K13 (Ref. 2-61). 21 R2 K13 was an irradiation of GA-fabricated, TRISO-coated LEU UCO/ThO 2 in fuel compacts in H-451 graphite fuel bodies; the irradiation was performed in the R2 materials test reactor at Studsvik, Sweden. Cell 2 operated at time-average fuel temperature of 1190 o C, and Cell 3 operated at an average temperature of 980 o C. As shown in Figure 2-23, the measured Cs-137 profiles in the fuel-compact matrix and graphite web are qualitatively as predicted (solid lines): (1) the concentration profile in the matrix is flat (high diffusion coefficient); (2) a partition factor is observed across the gap, (3) the concentration profile in the graphite decreases exponentially with penetration distance, and the profile is flatter at the higher temperature. However, the absolute concentrations in the matrix and graphite are significantly underpredicted for Cell 2, suggesting that the release from the fuel particles into the compact matrix was underpredicted. Surprisingly, the total Cs-137 release from the graphite fuel body for Cell 2 (1190 o C) was overpredicted by a factor of 1.2. In contrast, the concentration in the graphite for Cell 3 (980 o C) was slightly overpredicted, but the release from the body was underpredicted by a factor of 37. Such discrepancies and apparent inconsistencies are common when attempting to predict metallic release from irradiation capsules (and reactor cores). Explanations are confounded by the difficulty of isolating potential errors in the prediction of the fuel particle failure rates from potential errors in the prediction of fission product transport. 21 The R2-K13 experiment will be discussed in more detail in Section

100 Radionuclide Release From HTGR Cores Figure 2-23 Cesium Transport in R2 K13 Capsule 2.4 Radionuclide Control Requirements As described in Section 2.2, the radionuclide containment system for the commercial GT-MHR (and, presumably, for the VHTR as well although the design is still in the pre-conceptual phase) is comprised of multiple barriers to limit radionuclide release from the core to the environment to insignificant levels during normal operation and a spectrum of postulated accidents. To reiterate, the five principal release barriers are: (1) the fuel kernel, (2) the particle coatings, particularly the SiC coating, (3) the fuel-element structural graphite, (4) the primary coolant pressure boundary; and (5) the Vented Low-Pressure Confinement building. As part of the design process, performance requirements must be derived for each of these release barriers. Of these multiple release barriers, the particle coatings are the most important. Moreover, the inreactor performance characteristics of the coated-particle fuel are strongly influenced by its asmanufactured attributes. Consequently, the fuel performance requirements and fuel quality requirements (allowable, as-manufactured HM contamination and coating defects) must be systematically defined and controlled Methodology The logic for deriving these fuel quality specifications is illustrated in Figure 2-24 (Ref. 2-62). Top-level requirements for an advanced HTGR design will be defined by both the regulators and the users. Lower-level requirements will then be systematically derived using a top-down functional analysis methodology. With this approach, the radionuclide control requirements for 2-48

101 Radionuclide Release From HTGR Cores each of the release barriers can be defined. For example, starting with the allowable doses at the site boundary, limits on Curie releases from the reactor building, from the reactor vessel, and from the reactor core will be successively derived. Fuel failure criteria are in turn derived from the allowable core release limits. Finally, the required as-manufactured fuel attributes will be derived from the in-reactor fuel failure criteria, with consideration of achievable values based on existing fuel experience, thereby providing a logical basis for the fuel quality specifications. Figure 2-24 Logic for Derivation of Fuel Quality Requirements Radionuclide Design Criteria Standard GA design practice is to define a two-tier set of radionuclide design criteria, - referred to as Maximum Expected and Design criteria, - (or allowable core releases for normal operation and Anticipated Operational Occurrences); this practice has been followed since the design of the Peach Bottom 1 prototype U.S. HTGR up through the current commercial GT-MHR. The Design criteria are derived from externally imposed requirements, such as the site-boundary dose limits, occupational exposure limits, etc.; in principle, any of these radionuclide control requirements could be the most constraining for a given reactor design. 2-49

102 Radionuclide Release From HTGR Cores Once the Design criteria have been derived from the radionuclide control requirements, the corresponding Maximum Expected, criteria are derived by dividing the Design criteria by an uncertainty factor, or design margin, to account for uncertainties in the design methods. This uncertainty factor is typically a factor of four for the release of fission gases from the core and a factor of 10 for the release of fission metals. The fuel and core are to be designed such that there is at least a 50% probability that the fission product release will be less than the Maximum Expected criteria and at least a 95% probability that the release will be less than the Design criteria. This GA approach to implementing such radionuclide design criteria is illustrated in Figure (No particular scale is implied in this figure; it is simply a conceptual illustration of the approach.) In the example given in the figure, the Preliminary Design predictions (solid lines) slightly exceed the criteria (triple lines) at the 50% confidence level: i.e., the nominal (50% confident) prediction is slightly higher than the Maximum Expected criterion, but the 95% confident prediction meets the Design criterion, primarily because a large design margin was chosen to accommodate the considerable uncertainties in the current design methods at the Preliminary Design stage. This example was chosen because it is anticipated to roughly reflect the current prediction of Ag-110m release from a commercial GT-MHR core, based upon previous GA analysis of a WPu core operating with an 850 o C core outlet temperature. Silver release is of concern because it can be diffusively released from intact TRISO particles at high temperatures and preferentially deposit on the turbine, where it is predicted to be a dominant contributor to O&M dose rates (it is only a minor contributor to offsite dose rates because of its low effectivity). 22 There are several candidate options for resolving this design issue. The first option is simply to relax the Maximum Expected criterion and to design the plant to accommodate the currently predicted levels of Ag release and the large uncertainties in the predictive methods; however, this option implies high O&M dose rates and the attendant requirements for fully remote turbine maintenance, etc. Another option is to develop and qualify efficient decontamination protocols to reduce the dose rates from the turbine prior to refurbishment to levels permitting hands-on maintenance. A third option (dashed lines) is to reduce the predicted Ag release and the uncertainties therein by a combination of design optimization (primarily to reduce the nominal prediction) and technology development (primarily to reduce the uncertainty in the prediction). 22 As discussed in Section 7, silver release and plateout will be somewhat less of an issue in a commercial GT-MHR because less Ag-110m is produced with LEU fuel compared to WPu fuel, but even with LEU fuel it is still predicted to a major contributor to O&M dose rates, along with Cs-134 and Cs

103 Radionuclide Release From HTGR Cores RADIONUCLIDE DESIGN CRITERIA PRELIMINARY DESIGN PREDICTIONS FINAL DESIGN PREDICTIONS Design Margin Design (P>95%) Upper Bound (P=95%) Uncertainty Analysis Technology Development Upper Bound (P=95%) Maximum Expected (P>50%) Nominal Prediction (P=50%) Design Optimization Nominal Prediction (P=50%) Figure 2-25 Radionuclide Design Criteria Since diffusive release from intact particles is the dominant source of Ag release, the most effective design changes to reduce Ag release are those that reduce the peak fuel temperatures in the core. Some reduction in peak temperatures can be achieved by improved fuel zoning to optimize the core power distribution for minimum Ag release, and further reductions are possible with various fuel shuffling schemes. Larger fuel temperature reductions require more dramatic changes in the fuel-block design and/or in core operating conditions (e.g., power density); such changes have broad implications for the overall plant design and fuel cycle costs. A comprehensive trade study would be required to identify the optimal combination of the above options to resolve the Ag plateout issue. In any case, it would be prudent to design a first-of-akind, direct-cycle HTGR to permit fully remote turbine maintenance should the actual gamma dose rates prove to be higher than predicted. Regarding the maintainability of direct-cycle HTGRs, the EPRI Nuclear Maintenance Application Center (NMAC) completed an independent review of the International GT-MHR design and the underlying O&M philosophy (Ref. 2-63). The NMAC team concluded that the GT-MHR was a viable Generation IV nuclear plant concept. They noted that the success of a first-of-a-kind design depends strongly on the identification of potential problems and that, once identified, these potential problems can be avoided or mitigated by providing margin in the plant 2-51

104 Radionuclide Release From HTGR Cores and/or incorporating contingencies in the project plan. This perspective is consistent with the GT-MHR design approach described above which emphasizes the inclusion of sufficient design margin to accommodate the current uncertainties in the design methods, combined with contingency planning for fully remote ISI and maintenance of the PCS, including the turbine, should the actual gamma dose rates prove to be significantly higher than predicted. The subject radionuclide design criteria, including the uncertainty factors, are design specific and depend strongly on the top-level radionuclide control requirements. For example, the Fort St. Vrain (FSV) HTGR (Ref. 2-52) was designed to meet 10CFR100 accident dose limits whereas the steam-cycle MHTGR (Ref. 2-64) was designed to meet the much more restrictive EPA Protective Action Guides; consequently, the core release criteria and the subordinate in-service fuel failure limits and as-manufactured fuel quality requirements for the MHTGR were much more stringent than those for FSV. Another example, because of the paucity of data for the irradiation performance of high-burnup TRISO-coated plutonium fuel, the design margins for the PC-MHR with WPu fuel were increased by a factor of 10 compared to the margins for the steamcycle MHTGR with LEU fuel. In practice, as illustrated in Figure 2-24, an iterative procedure is required to develop optimized radionuclide design criteria. With each iteration, the lower-level requirements are refined as a result of analyses performed for the reactor systems and components. The goal is to produce an optimum design which meets all the top-level requirements with sufficient, but not excessive, margin. The elements involved in optimizing the core release criteria for a particular plant design can be conveniently categorized as follows: (1) development and optimization of design requirements and subordinate criteria, (2) core and plant performance assessments to determine if the design meets the applicable requirements, and (3) technology development and demonstration to validate the design and the underlying design methods and codes 2.5 Fuel Design Criteria for Advanced HTGRs A comprehensive top-down functional analysis to define fuel requirements has not yet been performed for either the commercial GT-MHR or the VHTR. 23 However, a preliminary fuel product specification for the GT-MHR has been prepared (Ref. 2-2), based upon extrapolation of the functional analysis performed for the steam-cycle MHTGR (Ref. 2-62). In addition, a generic fuel product specification for Advanced HTGRs (Ref. 2-65) has been prepared as input to the AGR fuel development program, which is developing a fuel product and process specification to be responsive to the evolving needs of the VHTR. In-service fuel performance requirements and as-manufactured fuel quality requirements have not yet been defined for a generic VHTR or for the VHTR Demonstration Module. The fuel performance and quality requirements adopted for a given HTGR design along with the fuel service conditions will determine the amount of technology development that will be necessary to support the design and license the plant. Consequently, it is critically important that a comprehensive set of fuel requirements be derived for the VHTR early in the design process. 23 Since the VHTR program is in a very early phase, and since representative information on fuel design criteria is available for the GT-MHR, information in this section is predominantly associated with the GT-MHR to illustrate the requirements anticipated for advanced HTGRs. 2-52

105 Radionuclide Release From HTGR Cores Overall, the most constraining radionuclide control requirement for the steam-cycle MHTGR was to comply with the dose limits specified in the EPA Protective Action Guides (PAGs) at the 425-m Exclusion Area Boundary (EAB) so that the Emergency Planning Zone (EPZ) could be located at the EAB to preclude the need for public evacuation plans (Ref. 2-62). The PAGs limit both whole body and thyroid doses; these dose limits were used to derive allowable environmental releases of noble gases and iodines, respectively, during Licensing Basis Events (LBEs). The limit on iodine-131 (the dominant iodine isotope) release from the plant was used to derive the limit on I-131 release from the core which, in turn, was used to set the limit on inservice fuel failure. Finally, this limit on in-service coating failure was used to derive the limits on certain as-manufactured defects, including the missing-buffer layer fraction. The second, most constraining, top-level radionuclide control requirement for the steam-cycle MHTGR was to limit the occupational exposure to 10% of 10CFR20 (i.e., a factor of 10 ALARA margin was imposed on the design). A detailed occupational exposure assessment was not performed for the MHTGR. Hence, in deriving limits on plateout activity consistent with the subject goal, it was necessary to rely heavily upon previous occupational exposure assessments for earlier steam-cycle HTGR designs and upon engineering judgment. On that basis, it was projected that the 10% of 10CFR20 goal would be met if the gamma radiation fields around the primary circuit due to fission product plateout were limited to 10 mr/hr for scheduled maintenance activities (e.g., circulator ISI) and to 100 mr/hr for unscheduled maintenance activities (e.g., steam-generator tube plugging). These limits on gamma dose rates were in turn used to set limits on the primary circuit plateout inventories, in particular, limits on the releases of metallic fission products from the core, including Ag-110m, Cs-134, and Cs-137. Finally, the limits on Cs release from the core were used to derive limits on as-manufactured SiC defects. It is anticipated that off-site dose limits and occupational exposure limits will again be the most constraining, top-level radionuclide control requirements for both the International and commercial GT-MHR designs. It remains to be determined which of these requirements will be the more constraining with respect to in-service fuel failure and as-manufactured fuel quality for the GT-MHR (or for the VHTR) Provisional Fuel Requirements The provisional fuel performance and quality requirements for the commercial GT-MHR are summarized in Table 2-3, and the provisional metal release limits are shown in Table 2-4. For perspective, the allowable metal release limits for the US steam-cycle MHTGR plant and for the German direct-cycle HHT plant are also shown in the latter table (Ref. 2-66). The limits on volatile metal release are particularly speculative at this writing, and considerable plant design and fuel development will likely be required to optimize them. 2-53

106 Radionuclide Release From HTGR Cores Table 2-3 GT-MHR Provisional Fuel Requirements Commercial GT-MHR Parameter >50% Confidence >95% Confidence Missing or defective buffer <1.0 x 10-5 <2.0 x 10-5 Defective SiC <5.0 x 10-5 <1.0 x 10-4 HM contamination <1.0 x 10-5 <2.0 x 10-5 HM fraction outside intact SiC <6.0 x 10-5 <1.2 x 10-4 Normal operation <5.0 x 10-5 <2.0 x 10-4 Core heatup accidents [<1.5 x 10-4 ] 24 [<6.0 x 10-4 ] Table 2-4 GT-MHR Provisional Fission Metal Release Limits Allowable Core Fractional Release Reactor Plant Type Cs-137 Ag-110m COT ( o C) Expected Design Expected Design MHTGR Steam-cycle x x x x 10-3 HHT Direct-cycle x x x x 10-4 GT-MHR Direct-cycle x x x x 10-3 As a point of departure, the fuel requirements for the VHTR with a 1000 o C core outlet temperature may be assumed to be the same as those for the direct-cycle GT-MHR with an 850 o C core outlet temperature (Ref. 2-2). This assumption would be supported by efforts to reduce the difference between core outlet temperature and maximum fuel temperature but may prove to be too ambitious. It is reasonable to expect that these as-manufactured fuel quality limits can be met since the Germans met or exceeded comparable limits in the late 1970s (e.g., Refs and 2-67). However, the in-service fuel performance limits could prove problematic; in particular, the allowable core metal release limits (Ag, Cs, etc.) may have to be increased even if the failure limits are maintained because of the higher core temperatures which will result in less retention by the fuel kernels of failed particles and by the fuel-element graphite. 24 Values in [square brackets] are provisional and subject to revision as the design and safety analysis evolve. 2-54

107 Radionuclide Release From HTGR Cores Peak service conditions for prismatic VHTR fuel are assumed here that are consistent with previous core designs with outlet temperatures of 850 o C and higher. They are subject to revision if and when the conceptual and preliminary core designs are completed for a prismaticcore VHTR. These fuel service conditions are intended to enable the prismatic VHTR to achieve its goals of nuclear hydrogen production and high-efficiency electricity generation. These assumed prismatic VHTR service conditions are compared with the conditions for the 850 o C GT-MHR in Table 2-5 (Ref. 2-68). Table 2-5 Provisional Fuel Service Conditions for Advanced HTGRs Performance Parameters GT-MHR Prismatic VHTR Core outlet temperature Core power density Fuel element design 10-row block 10-row block Core Residence Time (EFPD) 425 Determined by core design Burnup - Fissile (% FIMA) Burnup - Fertile (% FIMA) 7 7 Maximum Fast Neutron Fluence (E>29 fj), (n/m 2 ) 5 x x Maximum Fuel Temperature ( o C): normal operation 1250 [1400] accident conditions <1600 [<1800] These provisional service conditions, especially the fuel temperature limits, serve only as an initial guide to the fuel development and the reactor core design. Such bounding conditions are needed to perform fuel-particle design analyses, to prepare provisional fuel product specifications, and to plan the details of the fuel irradiation and testing programs. Core designers need this information to guide them in the trade studies required to optimize the core design. Since certain coating failure mechanisms depend on the exact history of time, temperature, burnup, and fast neutron fluence, it may be necessary to define more detailed limits for combinations of these core and fuel cycle parameters as part of the overall VHTR design and development effort. Undoubtedly, these service conditions will be refined as the designs of the commercial GT-MHR and the VHTR evolve Provisional Fuel Product Specifications The conventional, TRISO-coated, fissile and fertile particle designs specified for the GT-MHR (Ref. 2-2) are summarized in Tables 2-6 and 2-7. The extensive international experience with a large variety of TRISO-coated fuel particles strongly indicates that SiC-based coating systems should prove adequate for a broad range of AGR applications with core outlet temperatures of at least 850 o C and, perhaps, up to 950 o C (with certain core design changes to minimize fuel temperatures). However, as core outlet temperatures are increased to 1000 o C and higher, the ultimate performance limits of SiC-based, conventional TRISO coatings will be reached, and advanced fuels will have to be developed and qualified (e.g., Ref. 2-69). 2-55

108 Radionuclide Release From HTGR Cores The fuel design described in Tables 2-6 and 2-7 is for a prismatic core. However, the fuel requirements summarized in Section are believed to applicable (certainly to within an order of magnitude and probably within a factor of several) to advanced HTGR designs with pebblebed cores as well. The PBMR Project does not appear to have published their fuel performance and quality requirements in the open literature at this writing; however, their fuel design is essentially the German fuel design for the steam-cycle HTR Modul, and the fuel requirements for the Modul (Ref. 2-70) were comparable to those given here. Moreover, although modern pebble-bed HTR designs have consistently used 500 µm LEU UO 2 particles (typically ~10%- enriched), pebble-bed core designs could also use 350 µm LEU UCO particles (19.9% enriched) with the following benefits: (1) higher burnups (~25% versus ~10% FIMA), (2) suppression of CO formation, and (3) elimination of potential kernel migration. In reality, the fuel product specifications control many as-manufactured fuel attributes in addition to the ones addressed herein, including heavy-metal loading requirements, kernel properties, coating properties, compact properties, impurity limits, etc. Moreover, for each step in the fabrication process, they impose acceptance criteria and often mandate the QC techniques to be used to demonstrate compliance. 2-56

109 Radionuclide Release From HTGR Cores Table 2-6 GT-MHR Coated Particle Nominal Design Parameters Parameter Fissile Particle Fertile Particle Composition UC 0.5 O 1.5 UC 0.5 O 1.5 Uranium enrichment, % (Natural Uranium) Design burnup (% FIMA) 26 7 Dimensions (µm Kernel Diameter Buffer thickness IPyC thickness SiC thickness OPyC thickness Particle diameter Material Densities (g/cm 3 ) Kernel Buffer IPyC SiC OPyC Elemental Content Per Particle (µg) Carbon Oxygen Silicon Uranium Total particle mass (µg)

110 Radionuclide Release From HTGR Cores Table 2-7 GT-MHR Fuel Compact Nominal Design Parameters Parameter Value Diameter, mm Length, mm 49.3 Volume, cm Shim particle composition Shim particle size H-451 or TS-1240 graphite 99 wt % < 1.19 mm; 95 wt % < 0.59 mm Shim particle density (g/cm 3 ) 1.74 Binder type Filler Thermosetting resin Petroleum derived graphite flour Matrix density (g/cm 3 ) 0.8 to 1.2 Volume fraction occupied by matrix 0.39 Volume fraction occupied by shim particles in an average compact Volume fraction occupied by fissile particles in an average compact Volume fraction occupied by fertile particles in an average compact Number of fissile particles in an average compact 4310 Number of fertile particles in an average compact

111 3 DESIGN METHODS FOR PREDICTING RADIONUCLIDE RELEASE Design methods have been developed during the past four decades to model coated-particle fuel performance and radionuclide transport in core materials during normal plant operation. These design methods have been used extensively for core design and safety analysis for both prismatic and pebble-bed HTGRs. The major phenomena controlling fuel performance and radionuclide release from HTGR cores were introduced in Section 2.2. The codes that are available to model these phenomena are described in this section. The validation status of these codes is discussed in Sections 4 and 5, and the additional technology development needed to complete their validation is addressed in Section Phenomenological Models and Computer Codes A number of different phenomenological models and associated computer codes have been developed internationally to predict fuel performance and radionuclide release from HTGR cores. Typically, the utility of the more sophisticated models has been limited by unavailability and/or unreliability of the material property data required as input to these codes. The models, codes and requisite material property data are summarized below by country of origin U.S. Computer Codes The computer codes currently available to predict fuel performance and fission product transport in prismatic cores during normal operation are listed below where they are categorized by particle analysis codes, core performance codes, and support codes Particle Analysis Codes SOLGASMIX-PV (Ref. 3-2): A thermochemical code that calculates equilibrium relationships in complex chemical systems by minimizing the free energy while preserving the masses of each element present for either constant pressure or volume. The code can calculate equilibria in systems containing a gaseous phase, condensed phase solutions, and condensed phases of invariant and variable stoichiometry. It has been used extensively to model kernel chemistry (Section 2.3.2) 3-1

112 Design Methods for Predicting Radionuclide Release FUEL (Ref. 3-3): A code that performs Monte Carlo calculations of fuel particle "pressure vessel" performance for fuel particle design and product specification development. FUEL uses a simplified, spherically symmetric, thick-walled shell stress analysis model to determine the failure probability of a statistical sample of fuel particles under constant irradiation conditions. ABAQUS (Ref. 3-4): A suite of industry-standard. general purpose, finite-element, structural analysis codes which can be used to perform full, deterministic, non-linear stress analysis. ABAQUS has been used to develop 2-D and 3-D pressure-vessel models for TRISO-coated fuel particles and to model various types of flaws, defects and structural abnormalities in the coating system. PISA (Ref. 3-5): A one-dimensional, spherically symmetric, coupled, thermal-stress finite element code used for fuel particle design, specification development, and capsule analysis. PISA performs deterministic, non-linear stress analysis of fuel particle "pressure vessel" performance for arbitrary irradiation histories. PISA can also be used to perform Monte Carlo calculations. MIT Failure Model (Ref. 3-6) 25 : An integrated fuel performance model for coated particle fuel based on a probabilistic fracture mechanics approach. The mechanical analysis includes effects of anisotropic creep and swelling and the possibility of variable physical properties as a function of fluence. Monte Carlo sampling of particles is employed in the fuel failure prediction to capture the statistical features of dimensions, material properties, and in the case of the pebble bed concept, the statistical nature of the refueling process. PARFUME (Ref. 3-7): An advanced, stress analysis code for TRISO fuel being developed by Petti et al. at INEEL. PARFUME considers structural failures due to overpressure, cracking of layers, debonding between layers, and asphericity. The code is still under development, and additional features, such as kernel thermochemistry, SiC/fission product corrosion reactions, etc., are to be added Core Performance Codes SURVEY (Ref. 3-8): An analytical/finite difference, core survey code that calculates the steady state, full core, fuel particle coating failure and the full core fission gas releases rates. An automatic interface with the core physics codes provides burnup, fast fluence and temperature distributions; likewise, the temperature and fuel failure distributions calculated by SURVEY are passed on to the metallic release code TRAFIC. SURVEY contains component models for each of the fuel failure mechanisms (Section 2.3.3) and fission gas release models for failed particles and HM contamination (Section ) 25 This MIT structural analysis model/code is unnamed in Ref

113 Design Methods for Predicting Radionuclide Release SURVEY/HYDROBURN (Ref. 3-8): An optional subroutine in SURVEY which calculates the corrosion of fuel element graphite and the hydrolysis of failed fuel particles by coolant impurities, particularly water vapor. Transport of water vapor through the graphite web of the fuel element is modeled as a combination of diffusion and convection due to cross block pressure gradients. The effects of catalysts and burnoff on the graphite corrosion kinetics are modeled. TRAFIC-FD (Ref. 3-9): A core survey code for calculating the full core release of metallic fission products and actinides. TRAFIC-FD is a finite difference solution to the transient diffusion equation for multi-hole fuel element geometry with a convective boundary condition at the coolant hole surface. The effect of fluence on graphite sorptivity is modeled explicitly. The temperature and failure distributions required as input are supplied by an automatic interface with the SURVEY code. TRAFIC-FD contains component models for fission metal transport in kernels, coatings, compact matrix and fuel-element graphite (Section 2.3.4) COPAR-FD (Ref. 3-10): A stand alone code as well as a subroutine in the TRAFIC-FD code which calculates the transient fission product release from failed and intact coated particles with burnup dependent kernel diffusivities (Sections and ). COPAR-FD is a finitedifference solution to the transient diffusion equation for multi-region spherical geometry and arbitrary temperature and failure histories Support Codes CAPPER (Ref. 3-11): A code (Capsule Performance) which calculates coated particle failure and fission gas release for irradiation test capsules. CAPPER also models fuel performance for out-of-reactor tests that simulate HTGR accident conditions. It has the capability of modeling test conditions (temperature, burnup, fluence, and dimensions) that vary arbitrarily with time and position. Analogous to the SURVEY code, CAPPER contains component models for each of the fuel failure mechanisms (Section 2.3.3) and fission gas release models for failed particles and HM contamination (Section ) TRAMP (Ref. 3-12): A one-dimensional, finite-difference local-point code which provides a generalized transient solution for the coupled diffusive transport of radionuclides along multiple, coupled parallel paths. Multiple, coupled transport models with arbitrary boundary and interface conditions can be conveniently programmed for solution by the code. TRAMP is used for hot spot analysis for the core and for analysis of irradiation experiments. The code contains component models for fission metal transport in kernels, coatings, compact matrix and fuelelement graphite (Section 2.3.4); the standard version of TRAMP has the same component models as TRAFIC-FD. RADC (Ref. 3-13): A zero-dimensional, steady-steady inventory code for calculating an overall plant mass balance for radionuclides in the core, primary circuit (both coolant and plateout inventories), and helium purification system during normal plant operation. 3-3

114 Design Methods for Predicting Radionuclide Release RANDI (Ref. 3-14): An advanced inventory code with capabilities beyond those of RADC, including explicit treatment of transient effects, a more detailed determination of core inventories (e.g., inventories within fissile and fertile particles), and a compartment model of the primary circuit German Computer Codes The primary German computer codes used for particle design and for predicting fuel performance during normal operation are listed below. 26 CONVOL (Ref. 3-16): A multi-shell, pressure-vessel code to calculate the particle failure fraction under isothermal irradiation conditions. It determines the tangential stresses within the SiC layer taking into account the internal fission gas pressure, the tensile stress imposed by kernel swelling, and the compressive stress in the SiC layer created by fast fluence-induced shrinkage of the pyrocarbon layers. STADIF-II (Ref. 3-17): A KFA computer code for calculating fission gas release during normal operation. It describes the steady-state noble gas and iodine release from defective and failed particles, including a recoil contribution, and from HM contamination using an uncoupled twopath (grain, pore or grain boundary) diffusion model, but it neglects sorption on graphite surfaces. The code also calculates the coolant activities, including the effects of plateout and the helium purification system. FRESCO-II (Ref. 3-18): A KFA diffusion code to calculate the metallic fission product release from a spherical fuel element under both irradiation and core heatup conditions. The model includes irradiation effects such as recoil and the buildup of fission product inventories. It is based on effective diffusion coefficients for the fission product species in the different fuel-element materials. PANAMA (Ref. 3-19): A KFA fuel performance code used primarily for predicting coating failure during core heatup accidents but occasionally used for normal operation as well. The code considers two failure mechanisms: (1) pressure vessel failure using a simplified model wherein the particle is predicted to fail when the tensile stress induced in the SiC layer by the internal gas pressure exceeds the tensile strength, and (2) thermal decomposition of the SiC layer as described by Weibull statistics. The strength of the SiC layer is also assumed to decrease due to fission product corrosion as described by a uniform thinning rate. 26 The PBMR project is using the German fuel performance and fission product transport codes (Ref. 3-15). 3-4

115 Design Methods for Predicting Radionuclide Release Other Foreign Computer Codes FORNAX (Ref. 3-20): A JAERI computer code which is similar to the German FRESCO-II code for describing metallic fission product release from the particle kernel by diffusion and recoil and the diffusive transport through the various materials of the fuel element. Three different types of fuel particles are considered: standard (intact) particles, failed particles (i.e., exposed kernels) and particles with a degraded SiC layer (simulated by a larger diffusion coefficient). The code is also applied to accident conditions. GOLT_v1 (Ref. 3-21): A Russian, multi-shell, pressure-vessel code for modeling the structural performance of high-burnup TRISO-coated particles, especially PuO 2-x particles. The model, which is similar to the US PISA code, uses Weibull statistics to describe the SiC layer and treats the effects of fluence-dependent shrinkage and creep on the performance of the PyC layers. The code is still under development, and additional features, such as CO formation, statistical variabilities in particle properties etc., are to be added. 3.2 Phenomenological Models and Material Property Data The phenomenological models and attendant material property data used in calculating fuel performance and radionuclide release from HTGR cores were introduced in Sections and 2.3.4, respectively. These models and material property data have been catalogued by several different authors as described in the following subsections. The most recent and the most comprehensive compilation is IAEA TECDOC-978 (Ref. 3-22). The primary strength of TECDOC-978 is its comprehensiveness under one cover, including an impressive bibliography; its principal weakness is that it is fundamentally a data compilation report with little evaluation and few recommendations; 27 for example, it contains disparate U.S., German, Japanese, and Russian correlations for Ag and Cs transport in PyC and SiC coatings without recommending any particular correlations for use in reactor design. Despite four decades of R&D, the uncertainties in many of the fuel performance and fission product transport models are large. Consequently, as discussed in Section 2.4, significant margin has to be included in the design to accommodate these uncertainties, and a number of Design Data Needs have been identified which call for additional technology development to refine the component models and to validate the design methods. These DDNs are presented in Section 7. Because of these circumstances, it is important that design manuals that provide models and correlations for predicting fuel performance and fission product transport include estimates of the uncertainties in these models and correlations. 27 In fairness, the original intent was to include a critical data evaluation phase in Coordinated Research Programme-2 that produced TECDOC-978, but the funding was terminated prematurely. 3-5

116 Design Methods for Predicting Radionuclide Release The reference GA component models and material property correlations are contained in the Fuel Design Data Manual, Issue F (FDDM/F, Ref. 3-23); 28 it is an invaluable document to control and standardize the models and correlations for reactor design and licensing. The FDDM/F has several notable limitations: (1) it was released in 1987 and needs to be updated (the process was begun in 1992 but not completed) to reflect more recent data; (2) it presents models and correlations along with extensive references, but it does not include the experimental data from which they were derived; and (3) it contains uncertainty estimates (estimated standard deviations), but it does not include the statistical analyses upon which they are based. The reference German models and material property correlations were documented in 1987 (Ref. 3-24) and updated in 1993 (Ref. 3-25); the latter summarized the reference models from other international HTGR programs as well. These German compilations suffer the same limitations as the FDDM/F: viz., they also do not include the experimental data from which the models and correlations were derived. In recognition of the above limitations, Martin of ORNL prepared a compilation in 1993 which collected the US and German models and the supporting data base under a single cover (Ref. 3-26). It was a highly ambitious and worthwhile exercise. However, it was originally planned to be a multi-year task, but it was not funded for the full duration. Consequently, it does have its limitations which it acknowledges. First, there is little or no discussion of the computer codes in which the component models are embedded and for which material property correlations serve as input; it is by way of these computer codes that these models and data are actually applied for core design and safety analysis. Secondly, there is no definition of the core environmental conditions during normal operation and postulated accidents to allow the reader to judge the extent to which the current data base encompasses the ranges of conditions over which the models and data are applied by the core designer and safety analysts. Finally, the statistical bases for the uncertainty estimates in the design manuals are not addressed. Presumably, had the task received full funding, these limitations would have been eliminated. At about same time that this ORNL review was being performed, a cooperative effort between Professor Kasten and two graduate students at the University of Tennessee and coated-particle fuel performance specialists at KFA was conducted to provide an improved fuel performance and fission product release model for pebble-bed reactors under core heatup conditions (Ref. 3-27). While their emphasis was on accident conditions, they did review irradiation performance data as well, and their model is applicable to both normal operation and accidents. Their model does appear to provide better agreement with the extensive postirradiation heating data for TRISO-coated LEU UO 2 fuel. They also identify additional R&D that will need to be performed before such models can be completely validated for reactor design and safety analysis. 28 FDDM/F (Ref. 3-21) is GA Proprietary Information and is not generally available; however, much of the information contained therein is included in TECDOC-978 (Ref. 3-20), especially in Appendix A. 3-6

117 Design Methods for Predicting Radionuclide Release Particle Material Properties Several compilations have been prepared of the material property data for kernels, PyC coatings and SiC coatings that are required for the various particle structural analysis models described above (e.g., Refs through 3-31). The summary of material models for PyC and SiC coatings by Ho (Ref. 3-28) is probably the best available compilation of coating property data at this writing. A standard linear viscoelastic material model and Weibull statistical strength theory (assuming volume flaws) are recommended to model the material behavior and failure strength, respectively, for both PyC and SiC coatings. Material constants for these models are given as a function of neutron fluence, temperature, degree of anisotropy, density and crystallite size. Specific models and correlations are recommended for particle design and analysis. With regard to kernel chemistry, especially CO formation as a function of burnup, the journal articles by Homan (Ref. 3-32) and Proksch (Ref. 3-33) are considered the most useful to the core designer. The most significant source of uncertainty in the structural performance modeling appears to derive from the uncertainty in the correlation for irradiation-induced pyrocarbon creep (Refs. 3-28, 3-34, and 3-35). A DDN has been prepared to address this uncertainty (Section 7) Fuel Performance Models For LEU TRISO-coated UCO particles in fuel compacts, the GA fuel performance models are available for use despite their considerable uncertainties. They are readily available in the ORNL review report prepared by Martin (Ref. 3-26). For LEU TRISO-coated UO 2 particles in spherical fuel elements, the KFA fuel performance models are available for use. They are presented in Section 2 and Appendix A of TECDOC-978 and through the citations given therein. The most significant source of uncertainty in modeling fuel performance under irradiation is the uncertainty in the model for predicting fission product/sic coating reactions at high irradiation temperatures (Ref. 3-35). Ironically, as programmatic interest gravitates to advanced HTGRs with higher core outlet temperatures, the predicted fuel performance becomes more sensitive to this corrosion model (Section ). In particular, the time-dependence of these corrosion reactions is not well established. This uncertainty is particularly of concern when extrapolating the results of accelerated (i.e., short-term) irradiation tests to predict real-time, in-reactor SiC corrosion rates. A DDN has been prepared to address this issue (Section 7) Fission Product Transport Models/Correlations For fission product transport in the materials in US prismatic-core designs (i.e., LEU TRISOcoated UCO particles in fuel compacts), the GA fission product transport models and correlations are also available for use despite their considerable uncertainties. They are readily available in the ORNL review report prepared by Martin (Ref. 3-25). For LEU TRISO UO 2 particles in spherical fuel elements, the KFA fission product transport models and correlations are also available for use. They are presented in Ref and in Appendix A of TECDOC

118 Design Methods for Predicting Radionuclide Release The most significant source of uncertainty in modeling in-core radionuclide transport is the uncertainty in the correlations for predicting Ag transport in SiC coatings and, to a lesser extent, in fuel-element graphite. As previously mentioned, Ag release and plateout are of particular concern for direct-cycle HTGRs because the impact on turbine maintenance, and that concern is exacerbated as core outlet temperatures are increased because silver is diffusively released from intact TRISO particles at sufficiently high temperatures (Section ). There is some indication that Ag transport in the SiC layer is dependent upon coating process conditions, but the dependence is not well characterized. There are essentially no reliable data to characterize the transport of silver in irradiated fuel-element graphite at reactor system pressures. DDNs have been prepared to address both these issues (Section 7). 3-8

119 4 FUEL PERFORMANCE DATA BASE The most important consideration, of course, for coated particle fuel is its performance under irradiation and under postulated accident conditions (the emphasis here being on irradiation performance). The ultimate figure-of-merit for judging fuel performance is the ability of the coated particle to retain fission products; consequently, the decision to discuss certain data sets (e.g., reactor surveillance data, etc.) in this section on fuel performance or in the next section on fission product release is somewhat arbitrary. Obviously, the prediction of fuel performance (i.e., coating failure rates) is a prerequisite to predicting fission product release from the core, and the reactor designer and safety analyst must do both. In this review, those integral data sets which have been used to assess the validity of the design methods for predicting fission product release are assigned to Section 5. Coated-particle fuel for use in both prismatic and pebble-bed HTGRs has been under development for the past four decades. The various international HTGR fuel development programs have performed hundreds of irradiation tests with a large variety of coated-particle designs and conducted reactor surveillance programs to assess fuel performance in operating HTGRs. Initial fuel development focused on BISO-coated carbide particles, and BISO particles were mass produced and used in the first prototype HTGRs (Dragon, AVR, and Peach Bottom 1) and in the THTR. However, the superior fission product retention capabilities of TRISO-coated particles, because of the addition of the SiC layer, were recognized early, and its development and qualification became the emphasis by the late 1970s. Other coating systems (e.g., ZrC, Si-BISO alloys, exotic coatings like NbC, etc.) have also been investigated, but they are beyond the scope of this review. Because certain coatings, such as ZrC, are more refractory than SiC, there is renewed interest in their development for very high temperature applications (e.g., Ref. 4-1). The international data base for the irradiation performance of TRISO-coated fuel particles is robust. TRISO-coated particles with a variety of fuel kernel compositions have been fabricated, using a range of process conditions, and irradiated in materials test reactors (MTRs) and in operating HTGRs. In the latter case, TRISO particles have been used as the driver fuel (FSV, AVR reloads, HTTR, and HTR-10), and they have been irradiated as fuel test elements in operating HTGRs as well (especially Dragon). In a typical fuel development program, the test fuel is first designed to meet certain performance goals; the fuel is then fabricated to product and process specifications; the as-manufactured attributes are determined by a broad spectrum of QC test techniques; the fuel is irradiated in a reactor with the coating performance monitored by on-line fission gas release measurements; and finally, the end-of-life condition of the fuel is characterized by postirradiation examination using both nondestructive and destructive techniques, including the measurement of the distribution and cumulative release of fission metals. If any of these steps is omitted, the interpretation of the test results becomes much more difficult. 4-1

120 Fuel Performance Data Base The international irradiation data base for TRISO fuel particles is summarized in the following subsections where it is categorized by data from fuel irradiation capsules, from fuel test elements, and from operating HTGRs. This categorization is appropriate because each has its own advantages and disadvantages. The fuel capsule irradiations have been in materials test reactors which are either light-water or heavy-water moderated (usually the former). These capsule irradiations are typically accelerated by factors of two to 10 compared to the irradiation times to full burnup in operating HTGRs (i.e., the neutron flux levels are higher in MTRs), and the fast fluence to burnup ratios are often substantially different than would be encountered in operating HTGRs. The main advantages of irradiation capsule tests are: (1) full burnup can be achieved relatively quickly, (2) the capsules can be highly instrumented, (3) the test conditions can be well controlled and maintained relatively constant; and (4) the on-line performance can be monitored by fission gas release measurements; the main disadvantages are: (1) the neutron spectrum is not prototypical, (2) high acceleration may introduce artifacts, and (3) the results must be extrapolated to real times for reactor application. The main advantages and disadvantages for testing in an operating HTGR are essentially the inverse of those for irradiation capsules. Stated differently, a well designed irradiation capsule in certain MTRs can produce test data that are almost differential in nature; irradiation data from operating HTGRs will always be integral in nature. Thus, in principle, capsule irradiations are best suited for producing data for model development, and operating HTGR data are best suited for model validation. In reality, irradiation capsules in certain MTRs (e.g., the HFIR at ORNL) experience a wide range of temperatures during an irradiation cycle and yield data that are also essentially integral in nature. There have been previous attempts to summarize the international coated-particle irradiation data base. Once again, TECDOC-978 is probably the best single source (Ref. 4-2); in Section 2.4 of the TECDOC, the U.S., German, Japanese, Russian, and Chinese coated-particle irradiation data bases are discussed. Dragon Project Report DP-1000 (Ref. 4-3) provides an excellent summary of the Dragon fuel development program and the many irradiation tests in the Dragon HTGR; these results are older but still quite relevant. Appendix A of the NPR Fuel Development Plan (Ref. 4-4) has a good summary of the HEU TRISO irradiations, but the reference is not generally available. Nabielek has published a number of summaries of the German data base with emphasis on their high-quality LEU UO 2 fuel (e.g., Ref. 4-5). Petti has also prepared a comparative summary of the U.S. and German irradiation data and attempted to identify the key reasons why the German fuel has performed so much better than the U.S. fuel to date (Ref. 4-6). Each of these references is excerpted (and credited) in this section. In this report, emphasis will be placed on the U. S. and German TRISO irradiation data bases since the fuel performance and fission product release models for normal reactor operation that are presented herein (Section 3) are largely derived from these data along with some selected, more recent Japanese data. 4.1 Fuel Irradiation Capsules Most TRISO-coated fuel particle designs (Section 1.3.2) consist of the standard sequence of coatings: buffer, IPyC, SiC and OPyC (a so-called 4-layer TRISO coating system); in some particle designs, a thin (~5 µm) PyC seal coat is added between the buffer and IPyC to facilitate 4-2

121 Fuel Performance Data Base certain QC measurements during fabrication (a 5-layer particle). Most German TRISO particles are 4-layer particles, and most GA TRISO particles are 5-layer particles. Since the seal coat appears to have no performance implications, these two particle designs can be considered together and are referred to herein as conventional TRISO particles. The on-going international fuel development efforts, including the U.S. AGR program, the PBMR program, the INET program in China, and the EU program, are all focused on conventional TRISO coating systems. Other non-conventional TRISO coating designs have also been fabricated and irradiated. The TRISO-P particle (Section 2.3.1) with its added porous PyC layer external to the dense OPyC is a prime example. These non-conventional TRISO designs will be briefly addressed in a separate subsection. None of these non-conventional designs has been used as the driver fuel in operating HTGRs to date Conventional TRISO-Coated Particles As noted earlier, the discussion here is focused on irradiations in the US and Germany, recognizing that a substantial number of additional irradiation tests of conventional TRISO particles were conducted in Japan, Russia and the United Kingdom, and irradiation testing of fuel made in China has been recently completed in Russia HEU Irradiation Tests A large number of irradiation capsule results have been obtained for HEU UC 2, UCO, UO 2, (Th,U)O 2, and (Th,U)C 2 particles with both BISO and TRISO coatings as shown in Table 4-1 (principally from Ref. 4-4, embellished by Refs. 4-7 and 4-8). The fuel in these HEU irradiations was either a single particle containing Th as well as HEU or a two-particle system with an HEU fissile particle and a separate fertile particle (usually containing Th but also natural U). In many of the early irradiations, the fertile particle was BISO-coated. The reference fuel cycle of the day included recycle of the bred U-233 from the fertile particle, and the use of a BISO-coated fertile particle simplified the recovery of the bred U-233 while the TRISO coating on the fissile particle minimized the crossover of the bred U-236 (a neutron poison) from the fissile particle during reprocessing (Ref. 4-9). The disadvantage of a BISO fertile particle was the excessive diffusive release of Cs isotopes from intact BISO particles at high fuel temperatures. By the late 1970s (after nuclear fuel reprocessing had been banned in the U.S.), BISO fertile particles were eliminated from further consideration. Diffusive Cs release from intact BISO fertile particles also complicates the interpretation of irradiation data from capsules with mixed TRISO fissile and BISO fertile fuel. In the U.S., 61 capsules with HEU fuel were irradiated, and an additional 22 were irradiated as part of the German development effort. In 17 of the U.S. capsules the burnup exceeded 70% FIMA, the temperature was in excess of 1000 C, and the fast neutron exposure was greater than 5 X n/m 2 (E>29 fj) HTGR. A list of the capsules and exposure conditions of burnup, fast fluence and temperature experienced in specific HEU fuel irradiation capsules and reactors is provided in Table 4-1 (Refs. 4-4, 4-7 and 4-8). 4-3

122 Fuel Performance Data Base Table 4-1 HEU Irradiation Capsules Capsule Fissile Particle Type Temperature (K) U.S. Capsules Burnup (% FIMA) P7 UC 2 -T&B P8 UC 2 -T&B P9 P10 P11 (Th, U)O 2 -B, (Th,U)C 2 -B&T (Th, U)O 2 -B, (Th,U)C 2 -B&T UC 2 -T&B, (Th, U)O 2 -T, (Th,U)C 2 -T&B P12 (Th, U)O 2 -B, (Th,U)C 2 -B P13E UC 2 -B, (Th, U)O 2 -B&T, (Th,U)C 2 -B P13F (Th,U)C 2 -B&T P13G (Th, U)C 2 -B, (Th, U)O 2 -T P13H (Th, U)C 2 -T&B P13J (Th, U)C 2 -T&B P13K UC 2 -B, (Th, U)C 2 -B P13L UC 2 -T&B, (Th, U)C 2 -T, UO 2 - T&B P13M UC 2 -T, (Th, U)C 2 -T&B P13N UC 2 -T, UO 2 -T (Th, U)O 2 -T P13P UC 2 -T, UO 2 -T, (Th, U)O 2 -T P13Q UC 2 -T, (Th, U)O 2 -T P13R UC 2 -T P13S UC 2 -T P13T UC 2 -T P14 UC 2 -B, (Th, U)C 2 -B P14B (Th, U)C 2 -B Fast Fluence (x10-25 n/m 2, E>0.18 Mev) 4-4

123 Fuel Performance Data Base Table 4-1 (Continued) HEU Irradiation Capsules Capsule Fissile Particle Type Temperature (K) U.S. Capsules Burnup (% FIMA) P15 (Th, U)C 2 -B P16 (Th, U)C 2 -B P17 (Th, U)C 2 -B P18 (Th, U)C 2 -B P19 (Th, U)C 2 -B P20 (Th, U)C 2 -T P21 (Th, U)C 2 -T P22 (Th, U)C 2 -T P23 (Th, U)C 2 -T&B F25 (Th, U)C 2 -T F26 (Th, U)C 2 -T F27 (Th, U)C 2 -T&B P28 (Th, U)C 2 -T F29 (Th, U)C 2 -T F30 (Th, U)C 2 -T SSL-2 UC 2 -T HRB-6 (Th, U)C 2 -T HRB-7 UCO WAR 1 (?) HRB-8 UCO WAR (?) HRB-9 UCO WAR (?) HRB-10 UCO WAR (?) HRB-l1 UC 2 -T < HRB-12 UC 2 -T HRB-13 UCO WAR (?) HRB-17 UCO-T, UC 2 -T HBR-18 UCO-T, UC 2 -T OF-1 UC 2 (?) NPR-1 UCO-TP NPR-1A UCO-TP NPR-2 UCO-TP Fast Fluence (x10-25 n/m 2, E>0.18 Mev) 4-5

124 Fuel Performance Data Base Table 4-1 (Continued) HEU Irradiation Capsules Capsule Fissile Particle Type Temperature (K) French Capsules Burnup (% FIMA) FR-1 (Th, U)C 2 -T FR-2 (Th, U)C 2 -T FR-3 (Th, U)C 2 -T GF-1 UC 2 -T, (Th, U)O 2 -T GF-2 UC 2 -T, (Th, U)O 2 -T GF-3 UC 2 -T, (Th, U)O 2 -T GF-4 UC 2 -T, UCO(WAR)-T GF-5 UCO(WAR)-T GF-6 UCO(WAR)-T German Capsules BR2-P23 UCO-T BR2-P24 (Th,U)O 2 -B BR2-P25 (Th,U)O 2 -T BR2-P9 (Th,U)O 2 -B BR2-P10 (Th,U)O 2 -B HFR-K1 (Th,U)O 2 -B R2-K3 (Th,U)O 2 -B, (Th,U)C 2 -B R2-K12 (Th,U)O 2 -T UC 2 -T Fast Fluence (x10-25 n/m 2, E>0.18 Mev) R2-K13 3 (Th,U)O 2 -T, Cells 1& FRJ2-P11 (Th,U)O 2 -B FRJ2-P12 (Th,U)O 2 -B FRJ2-P14 UC 2 -T, UO 2 -T FRJ2-P16 (Th,U)O 2 -B, (Th,U)C 2 -B FRJ2-P17 (Th,U)O 2 -B FRJ2-P18 (Th,U)O 2 -B, UO 2 -B FRJ2-P19 (Th,U)O 2 -B

125 Fuel Performance Data Base Table 4-1 (Continued) HEU Irradiation Capsules Capsule Fissile Particle Type Temperature (K) German Capsules Burnup (% FIMA) FRJ2-P20 (Th,U)O 2 -B FRJ2-P22 (Th,U)O 2 -B FRJ2-P23 (Th,U)O 2 -T FRJ2-P25 UCO-T, (Th,U)O 2 -B, (Th,U)O 2 - T, UC 2 -T FRJ2-K10 UCO-T FRJ2-K11 (ThU)O 2 -T Note: T = TRISO; TP = TRISO-P, B = BISO; WAR = Weak Acid Resin (a kernel fabrication process) LEU Irradiation Tests Fast Fluence (x10-25 n/m 2, E>0.18 Mev) In the late 1970s, HEU fuel was replaced by LEU fuel (< 20 % U-235) as the reference fuel for the HTGR in accord with the general goal of preventing the proliferation of nuclear weapons. All commercial HTGR designs since then have employed LEU fuel cycles. Irradiation performance data, mainly for LEU TRISO UC 2, UCO, and UO 2, have been obtained from nine U.S. and 15 German irradiation capsules (Refs. 4-8 through 4-18) as shown in Table 4-2, as well as other capsules in Japan and Russia which are not included here. In addition to the characterization of fissile coated fuel particles, greater than 60 of the irradiation capsule tests, including most of those HEU tests described above, contained TRISO ThO 2 fertile particles (Ref. 4-16). As discussed above, the generic nature of the irradiation data permits the coating material performance results from the complete spectrum of fuel types to be used in the design of HTGR fuel. The earlier HEU data are of substantial benefit in establishing a general understanding of basic phenomena affecting LEU TRISO performance; but because of major differences in fuel form, irradiation conditions, and quality requirements, predicting the performance of high quality LEU TRISO fuel with confidence must rely primarily on irradiation testing of high quality LEU TRISO fuel U.S. LEU Program In the period 1978 to 1981 a U.S. program was conducted to test and select suitable LEU fuel materials (Ref. 4-19). The fissile, two-phase, oxycarbide system of UC 2 and UO 2, represented commonly by the abbreviation UCO, and the fertile ThO 2 system were selected. The selection process involved the consideration of heavy metal loadings, radionuclide release limits, fabrication, irradiation response, performance margins, manufacturing and schedule risks, and transitions from LEU to HEU cycles. As presented in Ref. 4-20, the UCO particle promised: (1) a lower CO pressure and consequent reduction in kernel migration, and (2) kernel retention of rare earths and the consequent reduction in SiC attack by the rare earths. Also noted was that a UC 0.2 O 1.8 particle irradiated to 78% FIMA did not show evidence of kernel migration or chemical attack on the SiC. 4-7

126 Fuel Performance Data Base Table 4-2 LEU Irradiation Capsules Capsule Fissile Particle type Temperature (K) Burnup (% FIMA) Fast fluence (x10-25 n/m 2, E>0.18 Mev) U.S. Capsules HRB-14 UCO, UC 2, UO 2 - all T HRB-15A UCO, UC 2, UO 2, UO 2 * 29 -all T HRB-15B UCO, UC 2, UO 2 *-all T HRB-16 UCO, UC 2, UO 2, UO 2 *-all T HRB-17 UCO, (HEU UCO) 30 -all T HRB-18 UCO-T, (HEU UCO-T) R2-K13 31 UCO-T HFR-B1 UCO-T HRB-21 UCO-TP German Capsules BR2-P12 UO 2 -B BR2-P15 UO 2 -T HFR-P4 UO 2 -T HFR-P5 UCO-T, UO 2 -T HFR-M5 UO 2 -T SL-P1 UO 2 -T FRJ2-P15 UO 2 -B FRJ2-P24 UCO-T FRJ2-P27 UO 2 -T FRJ2-K13 UO 2 -T FRJ2-K15 UO 2 -T FRJ2-P28 UO 2 -T UO 2 * represents a either a particle in which ZrC is deposited within the buffer layer or a particle in which thin PyC and ZrC layers are deposited adjacent to the UO 2 kernel. 30 HEU UCO loose particles included in encapsulated holders ( piggy-back samples). 31 U.S. /German cooperative irradiation experiments. 4-8

127 Fuel Performance Data Base Table 4-2 (Continued) LEU Irradiation Capsules Capsule Fissile Particle type Temperature (K) German Capsules Burnup (% FIMA) HFR-K3 UO 2 -T HFR-K5 UO 2 -T Steady state: Cycled: HFR-K6 UO 2 -T Steady state Cycled: Fast fluence (x10-25 n/m 2, E>0.18 Mev) Major tests of fuel compacts containing UCO (75% UO 2 and 25% UC 2 ) were conducted in irradiation experiments HRB-15A in the High Flux Isotope Reactor (HFIR) at ORNL and R2- K13 in the R2 Reactor at Studsvik, Sweden. The data from the two experiments were consistent in terms of the in-service failure fraction as a function of the accumulated fast fluence, E >29 fj) HTGR, the failed fraction being based on the on-line Kr-85m R/B measurements. For these experiments, the peak tolerable fast fluence appears to be of the order 4 x n/m 2 ; higher fluences led to an increasing failure fractions. The data from the experiments HRB-15A and R2-K13, however, were not consistent in terms of fission product attack on the SiC coating. The attack was extensive in HRB-15A (Ref. 4-11) and essentially absent in R2-K13 (Ref. 4-15). This implies that the accelerated irradiation in HRB-15A was the cause of the attack whereas in R2-K13, under essentially normal conditions, no attack was discernible (Refs and 4-21). The SiC attack was also extensive in other HRB experiments with related batches of particles Under HRB capsule irradiation conditions in HFIR, coated fuel particles experience significantly larger temperature gradients than an operating HTGR or certain other MTRs, such as R2 and HFR Petten. The gradients promote fission product transport to the cool side of the particle, accumulation of Ag, Pd, Ru, and rare earths at the inner SiC surface, penetration of the SiC and reaction of the named fission products with SiC. The latter leads to thinning of the SiC coating and an enhancement of fission product release from the particle. The HRB-21 capsule, which contained LEU UCO with a TRISO-P coating, was also part of the US LEU program. TRISO-P fuel is discussed in Section German LEU Program The in-reactor performance of German LEU fuels, together with follow-up postirradiation examinations, fission product release analyses, and postirradiation accident simulation tests, have demonstrated a high level of fuel performance of the reference UO 2 TRISO particle. The inservice failed fraction was derived from on-line R/B measurements in five representative irradiation experiments with fuel elements with LEU TRISO particles. These experiments 4-9

128 Fuel Performance Data Base include a total of 19 fuel elements and 276,680 TRISO particles (capsules FRJ2-K13, FRJ2-K15, HFR-K3, HFR-K5, and HFR-K6); the burnups and neutron fluences achieved, as well as the R/Bs for Kr-85m and Cs-137 fractional releases at the end of irradiation, are compiled in Table 4-3 (Ref. 4-22), together with other high quality TRISO fuels of different geometries, typically cylindrical compacts (capsules FRJ2-P27, HFR-P4, and SL-P1). Table 4-3 German High Quality LEU UO 2 Irradiation Summary Test Specimens/ Capsule No. of Particles Irrad. Time, (EFPD) Maximum Temperature ( o C) Max. Burnup (%FIMA) Max. Fast Fluence (x10 25 n/m 2, E>16fJ) EOL R/B Kr-85m Cs-137 Fractional Release HFR-P4 spheres x x 10-5 SL-P1 spheres x x 10-6 HFR-K3 1FE x x 10-5 FRJ2-K13 1 FE x x 10-5 FRJ2-K15 1 FE x x 10-6 FRJ2-P27 3 compacts x x 10-4 HFR-K5 1 FE Cycled x 10-7 ND HFR-K6 1 FE Cycled x 10-6 ND The analysis of the measured noble gas releases shows that in no single case was a failed particle generated during irradiation. Tests HFR-K3, FRJ2-K13 and -K15 also had no manufacturing defects. HFR-K5/6 had one manufacturing defect, but no additional in-pile failure; the other spheres had zero defects and zero failures. Shown in Figure 4-1, the R/B measurements from capsule FRJ2-K15 in the Juelich DIDO reactor with three fuel elements indicate that none of the three elements, which were irradiated in individually monitored capsules, contained failed particles. The level of uranium contamination is extremely low and can be derived to be ~4 ppb natural uranium contamination in the fuel element matrix and graphite capsule parts. The slow increase in the R/B over time is attributed to the release of noble gases from fissioning in the Pu-239 bred in this natural U contamination (Ref. 4-22). In summary, the German program produced a substantial number of successful irradiations of high quality LEU UO 2 fuel and demonstrated the feasibility of producing fuel that could perform at quality levels required by the modular HTR concepts. 4-10

129 Fuel Performance Data Base Figure 4-1 Fission Gas Release from FRG Capsule FRJ2-K

130 Fuel Performance Data Base The TRISO-coated UCO fuel irradiated in the German capsule FRJ2-P24 (Ref. 4-23) closely resembles the UCO now recommended by the U.S. In this irradiation test, 300 µm LEU (~20% enriched) UCO kernels were coated with the standard German TRISO coating system. The capsule contained about 50,000 particles in 12 compacts bonded with resin matrix. The 298 EFPD irradiation in the DIDO reactor, Juelich, FRG, reached a maximum burnup of 22% FIMA, a maximum fast fluence of 2.8 X n/m 2 E>29 fj) at a maximum temperature of 1350 o C. The on-line R/B data indicated no failure and the postirradiation examination showed no apparent inservice coating failure. The limited metallography found no exposed kernels. No postirradiation heating tests were performed on the test samples. The German HTR program concluded that, while this was a successful test, they had already adopted the 10% enriched UO 2 particle as their reference fuel, and UCO development was not pursued further Non-Conventional TRISO-Coated Particles TRISO-P Particles As introduced in Section 2.3.1, the TRISO-P particle (Ref. 4-24) was adopted as the reference particle for gas-cooled New Production Reactor and for the concurrent, commercial steam-cycle MHTGR program. The TRISO-P design featured both a significantly thicker and denser IPyC layer and an added porous protective (P-PyC) outer layer. Both design changes were made to solve perceived problems during fuel fabrication. The design changes resolved the process issues, and the as-manufactured quality of the fuel compacts was dramatically improved. The low-density P-PyC layer was included in the design to separate the particles during compact formation. The intent of the design was that the formation forces of compaction and matrix injection would cause the protective layer to be crushed without affecting the structural layers, therefore prevent coating mechanical damage. This layer likely also acted as a barrier to transfer of impurities from the furnace parts to the particle coatings during fabrication in the high temperature firing of the compacts. (At the firing temperature, 1600 o C to 1700 o C, the impurities can react with the SiC layer and cause coating defects.) This protective layer system worked extremely well in preventing SiC layer damage during compact fabrication. As-manufactured SiC defects were reduced to the range of 5 x 10-6 to 2.4 x 10-5 for the commercial LEU fuel and to 3 x 10-6 for the NPR capsule fuel. The heavy-metal contamination was <1 x 10-5 for HRB-21 and 5 x 10-7 for the NPR capsules. The initial R/B measurements for 85m Kr in capsules HRB-21, and NPR-1, 1A, and 2 were of the order 10-8 to The coating process for the inner pyrocarbon layer IPyC) was also changed from the normal TRISO procedure to produce a less permeable layer with the intent to reduce reactions between the chlorine liberated during SiC coating and the kernel. These reactions generate volatile products that cause SiC layer defects during coating. As a further measure to reduce SiC defects, the IPyC layer thickness was increased from 35 µm to 50 µm because GA data indicate that SiC defects are reduced as the IPyC layer thickness is increased. The IPyC coating rate was decreased to produce a more impermeable but unfortunately also more anisotropic coating layer. The IPyC anisotropy factor was higher than for pyrocarbon layers made at the normal coating rate but it still was within the specification. The decrease in permeability decreased the accessibility of the kernel to chlorine but produced a more anisotropic coating that meant more azimuthal dimensional change of the layer would occur under fast neutron irradiation. 4-12

131 Fuel Performance Data Base TRISO-P particles with HEU UCO kernels were irradiated in capsules NPR-1, NPR-2 and NPR-1A (Ref. 4-24), and TRISO-P particles with LEU UCO and ThO 2 kernels were irradiated in capsule HRB-21 (Ref. 4-25) in MTRs at nearly the same time for an initial test of the TRISO-P design; Table 4-4 describes the irradiation capsules. Table 4-4 TRISO-P Irradiation Capsules Capsule Test Reactor Uranium Enrichment (%) Kernel Diameter m) IPyC BAF o Fuel Temperature ( o C) Fluence at Initial Failure (n/m 2 ) Burnup at Initial Failure (%) HRB-21 HFIR Fissile x Fertile NPR-1 HFIR Fissile x NPR-2 HFIR Fissile x NPR-1A ATR Fissile x As shown in Figure 4-2, premature coating failure was detected in all four capsules by observing increases in the noble gas R/Bs and by spikes on the ion chambers monitoring the capsule sweep gas. As can be seen in Table 4-4, the initial failures were correlated with the fast neutron fluence with, perhaps, a lesser dependence on the capsule operating temperature. The failures were not correlated with burnup. Correlation with fast neutron fluence implied that the coating failures were related to the behavior of the pyrocarbon coatings, and the subsequent post-irradiation examinations confirmed this. In nearly all of the HRB-21 particles examined, the P-PyC and OPyC coating cracks are associated; failed IPyC layers are also present. From study of a large number of photomicrographs of irradiated TRISO-P particles, it was observed that SiC cracks always occurred in particles with broken IPyC layers. Data from the post-irradiation examinations and from supporting analyses of the irradiation behavior of the particle indicate that the failure of the SiC in the four capsules with TRISO-P particles was directly caused by the poor performance of the pyrocarbon layers (Ref. 4-26). The poor performance of the pyrocarbon layers was caused by deviations from the design and manufacturing techniques used for the conventional TRISO particle which have for many years generally performed well in fuel produced prior to the application of high quality levels required by the modular HTR concepts. 4-13

132 Fuel Performance Data Base Figure 4-2 Fission Gas Release versus Fast Fluence for TRISO-P Capsules TRISO-Coated UO 2 * Particles As described in Section 2.3.2, the rationale for the UCO fuel particle is add sufficient UC 2 to the UO 2 kernel to getter the excess oxygen liberated upon fissioning of the UO 2 in order to suppress CO formation and the attendant kernel migration. An alternative means of accomplishing the same objective is to add zirconium to the particle to act as the oxygen getter. One such design, of which there are two variants, is the so-called UO 2 * particle which is essentially a modification of the standard German TRISO UO 2 particle. The only design changes are use of a µm kernel and the addition of ZrC to the particle: either as a thin ZrC coating applied directly on the UO 2 kernel or codeposited with the porous PyC buffer layer. Such particles have been tested, and the results are intriguing; the limited experience with UO 2 * has been summarized in several documents, including in Section 6 of TECDOC-978; References 4-27 and 4-28 are the most cogent and complete. As indicated in Table 4-3, TRISO-coated UO 2 * particles were included in several HRB irradiation capsules in HFIR. No detrimental irradiation effects were ceramographically found, except for carbon precipitates near the outer edge of the kernels, which may result from the gettering action of the ZrC. Kernel migration was not observed in the UO 2 * particles but was 4-14

133 Fuel Performance Data Base observed in the conventional TRISO-coated UO 2 particles in the same capsules (HRB-15A and HRB-16). The absence of kernel migration in the UO 2 particles was good evidence that this potentially detrimental phenomenon in oxide fuels can be controlled by the addition of oxygen getters. There also appeared to be a reduction in fission product corrosion of the SiC layers even in cases where ZrC layer eventually failed. Improved silver retention of the SiC-TRISO coated UO 2 particles during irradiation was found by IMGA measurements of individual particles. The best evidence was that the particles irradiated in the HRB-16 capsule retained silver at or near 100 % during irradiation. Apparently, it was not necessary to maintain the ZrC layer intact throughout irradiation to achieve improved the silver retention with UO 2 * particles. The key to the improved irradiation performance seemed to be the better quality SiC layer obtained in these particles which could be made possible by: (1) less heavy metal contamination of the SiC layer as a result of better retention of kernel material during coating process because of the presence of an intact ZrC layer during SiC deposition, and/or (2) reduced corrosion of the SiC layer from chemical reactions with fission products during irradiation because of the shortened reaction time remaining subsequent to ZrC failure. The most intriguing data for UO 2 * particles are from a series of comparative postirradiation tests (Ref. 4-28). The tests included irradiated TRISO-coated particles with five different types of kernels: UCO, UC 2, UO 2, and the two variants of UO 2 * (ZrC seal coat and ZrC dispersed in the buffer layer). The particles were annealed at 1500 C for 10,000 h, and the UO 2 * particle with ZrC seal coat was the only one to retain 100% of its highly diffusive silver and europium inventories. To better appreciate how superior the retention of the UO 2 * particle was during heating, it should be mentioned that all 30 test particles of the other three fuel types without ZrC released Eu-154 at levels between 15 and 100% and that 22 of 30 particles released Ag-110m at levels between 10 and 100%, while none of the 10 UO 2 * particles with the ZrC seal coat released either of these metals. Complete Ag retention at 1500 C for 10,000 h has never been observed for any other TRISO fuel particle. This particle design was not pursued further because of the need to concentrate scarce development resources on the UCO reference design, which had been selected prior to the availability of the postirradiation heating data. However, it remains a possible future option if metallic fission product release needs to be better controlled for higher temperature applications Barrier Coatings As described above, certain fission products, as well as carbon monoxide, can corrode the SiC coating at sufficiently high temperatures (Section ). Two basic approaches have been investigated to prevent such reactions: (1) getters to retain the corrosive agents in the fuel kernel, and (2) barrier coatings to prevent the corrosive agents from reaching the SiC coating. The UCO kernel is an example of the use of a getter (UC 2 ) to prevent the generation of CO. The greatest challenge is to getter Pd (and other noble metal) isotopes since they do not form oxides. The Japanese have investigated the addition of sacrificial SiC to the particle to protect the main structural SiC layer (Ref. 4-29). Presumably, the Pd isotopes released from the kernel will react with this sacrificial Si to form silicides, thereby preventing them from attacking the structural SiC layer. 4-15

134 Fuel Performance Data Base Three different variants to a conventional TRISO-coated UO 2 particle were fabricated and tested by JAERI: (1) an additional layer of SiC+PyC adjacent to the interior of the structural SiC layer; (2) a layer SiC+PyC separated from the structural SiC layer by a dense PyC layer; and (3) a pure sacrificial SiC layer separated from the structural SiC layer by a dense PyC layer. Test particles were fabricated using conventional coating techniques and irradiated to up to 7% FIMA at temperatures up to 1330 o C. Other test particles were prepared for out-of-pile tests wherein they were heated with Pd powder at 1500 o C for 1 hr; since the Pd source for these tests was external, the sacrificial layers were deposited outside of the structural SiC layer and in reverse order. In all cases, the advanced coating systems had good irradiation performance, and the additional layers of SiC and SiC+PyC trapped palladium effectively to prevent the corrosion of the structural SiC layer. In addition, the intermediate PyC layer between the sacrificial layer and the structural SiC layer was found to interrupt the radial extension of the corrosion zone from the sacrificial layer to the SiC layer. The use of SiC (and ZrC) as getters for Pd appears to merit further investigation Key Differences between U.S. and FRG TRISO Particles From the information presented above, it is evident that German-made, conventional TRISOcoated particles have performed superbly under irradiation and that the performance to date of U.S.-made TRISO particles has been inferior. Petti of INEEL has prepared a comparative summary of the U.S. and German irradiation data and attempted to identify the key reasons for the performance differences (Ref. 4-6); his conclusions are reproduced below without editorial comment or rebuttal. This review has concluded that there has historically been a difference in the quality of U.S. and German fuel. This difference has been traced to technical differences in the fabrication processes used in Germany and the U.S. as well as different philosophies used to implement the irradiation and testing programs in the two countries. A review of the fabrication processes used in Germany and the U.S. to make coated particle fuel indicates that the scale of fuel fabrication and development efforts in the last 25 years were quite different. German fabrication of modern TRISO fuel was industrial/production scale incorporating improvements from fuel manufactured for the German AVR and THTR reactors. Only 100 defects were measured in 3.3 million particles produced. The post Fort St. Vrain U.S. Program was a mixture of lab scale and larger scale fabrication. The initial defect levels varied greatly and were generally much greater than those produced in Germany. A comparison of the fabrication processes has revealed many differences in the overall process. Three specific technical differences in the nature of the TRISO coating that can be attributed to differences in the fabrication processes are: pyrocarbon microstructure and density, the nature of the IPyC/ SiC interface, and SiC microstructure. 4-16

135 Fuel Performance Data Base A review of the U.S. and German irradiation programs over the last 25 years indicates that the irradiation programs were implemented quite differently with vastly different results. The German program s focus was on UO 2 -TRISO fuel for AVR and all future designs such as HTR Modul. The U.S. program produced and tested many different variants (different coatings, different kernels) using different coaters and different coating conditions, with apparently few lessons learned from one irradiation to the next, and insufficient feedback to the fabrication process. The on-line gas release data indicate that German fuel exhibits about a factor of 1000 less fission gas release under irradiation than U.S. fuel under a broad range of conditions (i.e., temperature, burnup, fluence). Furthermore, the postirradiation examination confirms the more extensive gas release data. German fuel is excellent. Out of ~380,000 LEU UO 2 and 80,000 HEU (Th,U)O 2 particles tested there were no in-pile failures and only a few damaged particles due to experimental anomalies. Gas release was attributed only to as-manufactured defects and heavy metal contamination. U.S. fuel did not perform very well. Percent level failures of fuel, and in many cases very high levels of failures of individual layers of the TRISO coating were observed following irradiation in most experiments. A variety of failure mechanisms were noted which were related to effects of accelerated irradiation and attributes of the fabrication process. Extensive testing has been done on German TRISO-coated fuel to characterize the behavior under long term depressurized conduction cooldown. Much less work has been done on U.S. UCO fuel. The German data show excellent behavior for fuel irradiated to burnups of less than 9% FIMA and fast fluences less than 4 x n/m 2 annealed at 1600 C. Greater releases were observed at higher temperature or 1600 C in fuel irradiated to 14% FIMA and fluences above 4.6 x nim 2. The work has resulted in better understanding of the mechanisms that challenge the integrity of SiC with respect to retention of fission products of the expected source term from the fuel for such events. 4.2 Fuel Test Elements Peach Bottom Fuel Test Elements HEU/Th Test Elements Peach Bottom 1 (PB) was the first HTGR in the United States. The plant operated for seven years until it was shut down for decommissioning in October In addition to producing commercial power (over 1.2 million MW(e)-hr for the Philadelphia Electric Company grid), this prototype reactor was used as a test facility for HTGR fuels and materials. In the Peach Bottom Fuel Test Element (FTE) program 32 (Ref. 4-30), 33 full-sized fuel test elements with various HEU TRISO and BISO fuels were irradiated at time-averaged temperatures as high as 1640 o C, fast neutron exposures up to 4.2 x 1O 25 n/m 2 and maximum burnup of 60 % FIMA; details are provided in Table 4-5 (some FTEs also contained separate fertile particles). Extensive postirradiation examinations and evaluations of 21 of these 32 Participants in the PB fuel test element program included GA, ORNL, USDOE, EPRI, Kernforschungsanlage Jülich GmbH (KFA), Hochtemperatur Brennelement Gesellschaft (HOBEG, formerly NUKEM), and the United Kingdom Atomic Energy Authority (UKAEA). 4-17

136 Fuel Performance Data Base irradiation experiments were performed and demonstrated the physical integrity of the coatings to be consistent with results obtained from accelerated fuel tests. Table 4-5 Peach Bottom Fuel Test Elements Test Element No. FTE-1 FTE-2 FTE-3 FTE-4 FTE-5 FTE 6 Fissile Particle (Th,U)C 2 BISO (Th,U)C 2 TRISO UC 2 BISO (Th,U)C 2 BISO (Th,U)C 2 TRISO UC 2 BISO (Th,U)C 2 TRISO UC 2 TRISO UO 2 TRISO (Th,U)C 2 TRISO UC 2 TRISO UO 2 TRISO (Th,U)C 2 TRISO UC 2 BISO UC 2 TRISO (Th,U)C 2 BISO (Th,U)C 2 TRISO UC 2 TRISO UO 2 TRISO Time (EFPD) Max. Temp. ( o C) Max. Burnup (%FIMA) FTE-7 UO 2 TRISO FTE-8 UC 2 TRISO FTE-9 (Th,U)C 2 TRISO FTE-10 (Th,U)C 2 TRISO FTE-12 (Th,U)O 2 BISO FTE-13 FTE-14 FTE-15 PuO 2 TRISO (Th,Pu)O 2 TRISO (Th,U)C 2 BISO (Th,U)O 2 TRISO UC 2 TRISO UO 2 TRISO (Th,U)O 2 TRISO UC 2 TRISO UO 2 TRISO Fast Fluence (x10 25 n/m 2 E>29 fj) Comment With EPRI FTE-16 (Th,U)C 2 TRISO FTE-17 (Th,U)C 2 TRISO FTE-18 (Th,U)O 2 BISO With HOBEG, Molded fue1 block irradiation FBTE-1 (Th,U)C 2 BISO Unbonded particle bed irradiation FBTE-2 UC 2 BISO Unbonded particle bed irradiation FBTE-3 UC 2 TRISO Unbonded particle bed irradiation 4-18

137 Fuel Performance Data Base Table 4-5 (Continued) Peach Bottom Fuel Test Elements Test Element No. Fissile Particle Time (EFPD) Max. Temp. ( o C) Max. Burnup (%FIMA) Fast Fluence (x10 25 n/m 2 E>29 fj) Comment FBTE-4 (Th,U)C 2 TRISO Unbonded particle bed irradiation FBTE-5 (Th,U)C 2 BISO (Th,U)C 2 TRISO UC 2 BISO 897 ND Unbonded particle bed irradiation FBTE-6 (Th,U)C 2 BISO 252 ND Unbonded particle bed irradiation RTE-1 UC 2 BISO UC 2 TRISO UO 2 TRISO UO 2 BISO (Th,U)O 2 BISO RTE-2 UC 2 BISO UC 2 TRISO (Th, U)O 2 BISO RTE-4 UC 2 BISO UC 2 TRISO (Th,U)O 2 BISO RTE-5 UC 2 BISO UC 2 TRISO UO 2 BISO (Th,U)O 2 BISO RTE-6 UC 2 BISO UC 2 TRISO UO 2 BISO (Th,U)O 2 BISO RTE-7 UC 2 BISO UC 2 TRISO UO 2 BISO (Th,U)O 2 BISO RTE-8 UC 2 BISO UC 2 TRISO Th, U)O 2 BISO Recycle test irradiation Recycle test irradiation Recycle test irradiation Recycle test irradiation Recycle test irradiation Recycle test irradiation Recycle test irradiation PTE-2 (Th,U)C 2 TRISO FSV Proof Test # 2 FPTE-1 U-238) O 2 TRISO With UKAEA, Fuel Pin Irradiation FPTE-2 U-238) O 2 TRISO With UKAEA, Fuel Pin irradiation Notes: FTE = Fuel Test Element FBTE = Fuel Bed Test Element RTE = Recycle Test Element PTE = Proof Test Element FPTE = Fuel Pin Test Element 4-19

138 Fuel Performance Data Base The FTE irradiations were simulated using HTGR design codes and data. Calculated burnups, power profiles, fast neutron fluences, and temperatures were verified via destructive burnup measurements, gamma scanning, and in-pile thermocouple readings (corrected for decalibration effects). Analytical techniques were developed to improve the quality of temperature predictions through feedback of nuclear measurements into thermal calculations. Extensive data on irradiationinduced strain in fuel compacts and H-327 graphite structures, residual stress and strength distributions in irradiated H-327 graphite structures, fuel performance, and the transport of metallic fission products in H-327 graphite were obtained from postirradiation examinations of the fuel test elements. The fuel test elements irradiated in Peach Bottom contained at least a fraction of TRISO fuel, especially the 15 test elements in which all of the fissile particles were TRISO coated. A large number of different kernel compositions were tested as TRISO-coated particles, including HEU (Th,U)C 2 /ThC 2 (the reference FSV fuel), HEU UC 2, UO 2, PuO x and (Th,Pu)O 2. A comprehensive summary report describing the Peach Bottom FTE Program is available (Ref. 4-30), and it includes a large number of citations of detailed reports on individual test elements. The Peach Bottom core was an excellent test bed for HTGR fuel development because many of the test element positions in the core were individually purged so that the R/Bs for test elements could be measured directly without contributions from the rest of the core. In addition, most of the fuel test elements contained additional test samples irradiated as piggy back samples in the central graphite spine; these samples included various kinds of loose particles and fission product diffusion samples. Upon discharge from the Peach Bottom core, many of these test elements were subjected to detailed PIEs, typically at ORNL or GA. The results of the various fuel test element irradiations in Peach Bottom are best understood by reviewing the operational and PIE reports for the individual PIEs. Nevertheless, several common characteristics of these irradiations are noteworthy: (1) large numbers of particles (>10 5 ) could be irradiated in a single test element; (2) the irradiation was in a true HTGR neutron flux spectrum; and (3) the irradiations were real time so all of the problems, real and imagined, associated with accelerated tests were avoided (e.g., excessive temperature gradients) Pu Test Element Most of the PB FTEs contained HEU/Th fuels, but a plutonium fuel test element was also irradiated. Funded by EPRI, FTE-13 (Refs and 4-32) consisted of 56 standard size fuel compacts containing TRISO-coated Pu fuel particles irradiated in a full-size fuel element. The fuel was irradiated at peak temperatures of up to 1440 o C and a peak burnup of 657,000 MW d /MT. The fast neutron fluence was 2.3 x n/m 2 (E > 0.18 MeV). The irradiation was for 512 EFPD; essentially all of the 239 Pu and ~75% of the total plutonium were fissioned. Two different, highly enriched (88% fissile), PuO 2-x kernel compositions and two different (Th,Pu)O 2-x kernels were tested. Post irradiation examination of the fuel revealed good performance for the fuel with a lower oxygen to plutonium (O/Pu) ratio (1.68:1) and both the mixed-oxide particles. However, poor fuel performance was observed for the fuel with a higher O/Pu ratio (1.81:1). This poor fuel performance resulted from excessive kernel migration. In this case the kernel migration was likely the result of the higher CO formation in the fuel particle during irradiation. Kernel migration was not observed in the fuel with the lower O/Pu ratio. 4-20

139 Fuel Performance Data Base Even though the palladium yield is much higher for Pu fission than for U fission, significant palladium attack was not seen in the irradiations of plutonium oxide kernels containing a mixture of PuO 2 and Pu 2 O 3 (overall O/Pu = PuO 1.68 ) Fort St. Vrain Fuel Test Elements The Fort St. Vrain Nuclear Generating Station was a 842 MW(t) HTGR that was operated by the Public Service Company of Colorado from 1974 to The FSV core used prismatic fuel elements with hexagonal cross-sections (e.g., Ref. 4-33). (The FSV fuel element was essentially identical to that specified for the commercial GT MHR, and it will be the point of departure for optimization of the fuel element for the prismatic NGNP.) Fuel for the reactor was based on the 93% enriched uranium/thorium cycle (HEU/Th). Separate TRISO-coated (Th,U)C 2 fissile and ThC 2 fertile particles were used. Fuel lifetime in the core was six years; about one-sixth of the fuel elements were removed and replaced at each refueling. Test elements were also designed, built and irradiated in the FSV reactor. Eight test elements were irradiated in FSV as shown in Table 4-6 (Ref. 4-34). Table 4-6 Fort St. Vrain Fuel Test Elements Test Element Planned Residence Time (Yr) Max. Temp. ( o C) Burnup, (%FIMA) Fast Fluence (x n/m 2 E>0.18MeV) HEU Fissile Fuel Type (see notes) FTE UC 2 TRISO CIP FTE LEU UC 2 TRISO (Th, U)C 2 TRISO (Th, U)O 2 TRISO UCO WAR TRISO FTE UC 2 TRISO CIP FTE LEU UC 2 TRISO (Th, U)C 2 TRISO (Th, U)O 2 TRISO UCO WAR TRISO FTE UC 2 TRISO CIP FTE LEU UC 2 TRISO (Th, U)C 2 TRISO (Th, U)O 2 TRISO UCO WAR TRISO FTE (Th, U)C 2 TRISO CIB FTE (Th, U)C 2 TRISO CIB Notes: FTE = Fuel Test Element LEU = Low Enriched Uranium, ~19.5%; all other fuel is HEU, High Enriched Uranium WAR = Weak Acid Resin kernels CIP = Cure in Place fuel compact carbonization within the FSV fuel block hole CIT = Cure in Tube within graphite crucibles, simulating conditions as experienced in CIP CIB = Cure in alumina Bed, reference FSV process Comment With ORNL/CEA, CIP & CIT With ORNL/CEA, CIP & CIT With ORNL/CEA, CIP & CIT 4-21

140 Fuel Performance Data Base Because of the limited nuclear operation of the FSV reactor after the FTEs were inserted in the core, no destructive PIEs of any of the FSV test elements was conducted. However, limitedscope destructive PIEs were carried out on two driver fuel elements. There was less fuel compact dimensional change than predicted based on the accelerated irradiation test data. However, the amount of SiC corrosion, although small, appeared to be greater than expected (Refs through 4-38). The information available from these two PIEs is so limited that it is not much value for design methods validation. The full-core, noble gas R/B measurements that were made throughout the plant operating life and the two plateout probe data sets (Section ) are of more value for that purpose. 4.3 HTGR Operating Experience Dragon HTGR The OECD-sponsored, High-Temperature Reactor Project Dragon did an extensive amount of pioneering work in the field of HTGR fuel and fission products. The central achievement of the Dragon Project was the construction and successful operation of the 20 MW(t) Dragon HTGR with a prismatic core at Winfrith in the UK. The helium inlet/outlet gas temperatures were 350/750 o C and the plant operated with a power density of 14 MW/m 3 (quite high by HTGR standards; e.g., the HTTR core power density is 2.2 MW/m 3 ). These investigations spanned many areas, including: TRISO fuel particle design, fabrication, and irradiation testing, fuel performance modeling, in-reactor radiochemical and coolant chemistry surveillance, fission product release and transport measurement and modeling, and reactor component decontamination. The Dragon Project began in 1959 and was terminated in 1976 because of unfavorable political circumstances. During the Project, a remarkable variety of coated particle fuels were fabricated and irradiated in materials test reactors and in-pile loops throughout Europe and, especially, in the Dragon HTGR. Their fuel development efforts were organized into four phases: 1. Initial survey of a broad spectrum of particle designs with various kernel and coating designs; 2. HEU/Th fuel cycles, including both one- and two-particle systems; 3. LEU fuel cycles, which emphasized the optimization and testing of a TRISO-coated UO 2 particle; 4. Advanced fuel systems, including a number of different TRISO coated PuO x and PuO x /C particles, were also successfully fabricated and irradiated. The work was systematically documented in a long series of Dragon Project DP reports and was summarized in a final, comprehensive overview report DP-1000 (Ref. 4-3), which contains a large number of references to their fuel and fission product research and development. Despite the fact that the research was done two or three decades ago, a number of these Dragon studies remain seminal works to this day: the exhaustive study by Voice (Ref. 4-39) of the relationships 4-22

141 Fuel Performance Data Base between coating process parameters and the attendant physical properties of SiC coatings is a classic example. All current and future researchers in the field of HTGR fuel and fission products would benefit from an in-depth study of the Dragon Project work, and DP-l000 is probably the best point of departure Peach Bottom HTGR The Peach Bottom Atomic Power Station Unit 1 was a 40 MW(e) HTGR prototype plant. The heart of the PB nuclear steam supply system was a helium-cooled, graphite-moderated, 115 MWth) reactor operating with a 700 C gas outlet temperature on a thorium-uranium fuel cycle. Peach Bottom operated successfully for seven years until it was shut down for decommissioning in late 1974 because it had completed its demonstration mission. An extensive and highly successful End-of-Life (EOL) R&D Program, jointly sponsored by USDOE and EPRI, was conducted with the primary goal of generating real-time integral data to validate HTGR design methods with emphasis on reactor physics, core thermal/fluid dynamics, fission product transport, and materials performance, especially performance of the Incoloy 800 used for the steam-generator superheaters (Ref. 4-40). The reactor core consisted of 804 graphite fuel elements oriented vertically in a close packed array within the steel reactor vessel. Each fuel element, which was 3.5 inches in diameter and 144 inches long, contained 30 annular fuel compacts comprised of coated fuel particles in a carbonaceous matrix. The fuel kernels were HEU (Th,U)C 2. The Core 1 fuel particles were coated with a single PyC layer solely to prevent hydrolysis of the carbide kernels during manufacture. However, these so-called LAMINAR particles were dimensionally unstable under fast neutron irradiation causing the fuel compacts to swell which in turn caused mechanical interaction between the compacts and outer sleeve, resulting in the cracking of up to 10 % of the sleeves. As a consequence, the plant was refueled early, and Core 2 contained BISO-coated particles which eliminated the compact swelling problem and provided enhanced fission product retention. Peach Bottom Core 2 performed exceptionally well. The utilization of the PB operating data and EOL program to assess the validity of core design methods is described in Section Fort St. Vrain HTGR The FSV HTGR was introduced above in the subsection on fuel test elements. The plant was designed to produce 842 MW(t)/330 MW(e) and had many design features common to prismatic-core MHRs, e.g., graphite moderator, helium coolant, and similar designs for fuel particles, fuel elements, and control rods (e.g., Ref. 4-41). Unlike the MHR designs with their steel pressure vessels, the FSV primary coolant circuit was wholly contained within a prestressed concrete reactor vessel (PCRV) with the core and reflectors located in the upper part of the cavity, and the steam generators and circulators located in the lower part. The helium coolant flowed downward through the reactor core and was then directed into the reheater, superheater, evaporator, and the economizer sections of the 12 steam generators. From the steam generators, the helium entered the four circulators and was pumped up, around the outside of the core support floor and the core barrel before entering the plenum above the core to complete the 4-23

142 Fuel Performance Data Base circuit. The superheated and reheated steam was converted to electricity in a conventional, steam-cycle power conversion turbine-generator system with a thermodynamic efficiency of ~40%. For FSV, 2448 fuel elements, 7.1 million fuel compacts containing 26,600 kg of fissile and fertile material in TRISO-coated fuel particles were produced. The fissile particle kernels contained fully-enriched uranium carbide and thorium carbide in a ratio of 1 to 3.6. The fertile particle kernels were 100% thorium carbide. Fissile and fertile particles were bonded together to form fuel compacts 1.26 cm in diameter and 5 cm long. Two different diameter fissile particles ( fissile A and fissile B ) and two different diameter fertile particles ( fertile A and fertile B ) were utilized in order to permit different heavy-metal loadings in a constant-volume fuel compact; the particle specifications are summarized in Table 4-7 and the fuel compact specifications in Table 4-8 (Ref. 4-35). In the later reload segments, graphite shim particles were used as well to allow fabrication of compacts with various fissile and fertile loadings. Table 4-7 Specifications for FSV Fuel Particles Kernel Particle Fissile Fertile Smaller (A) Larger (B) Smaller (A) Larger (B) Material (3.6Th*, U)C 2 (3.6Th*, U)C 2 ThC 2 ThC 2 Diameter Enrichment (% 235 U) Coating Buffer Seal <5 <5 <5 <5 IPyC SiC OPyC >25 >35 >30 >40 Defective Coatings Heavy Metal Migration + Missing Buffer + Missing IPyC 1 x x 10-3 Defective SiC 3 x x 10-3 Missing OPyC 1 x x

143 Fuel Performance Data Base Table 4-8 Specifications for FSV Fuel Compacts Diameter (ring gauge) Specification Length (mechanical measurement) ~1.26 cm ~5 cm Coke Content (% (coke + filler)) <0.36 Iron Sulfur Chlorine Vapor 1600 o C <20 µg/compact < 1200 ppm 5 X 10-8 atmospheres Value Contamination (average for all rods in segment) Uranium Thorium Fission Gas Release ( 85m o C) 3 x 10-5 Heavy Metal Contamination - Burn Leach (contamination + totally exposed kernels) Defective SiC 3 x x 10-2 Not Specified Thorium Contamination (hydrolysis) 1x10-4 Fuel Dispersion 1x10-3 1x10-3 Burn/leach (HM Contamination + SiC Defects) 3x10-3 1x AVR HTR The Arbeitsgemeinschaft Versuchs Reactor (AVR) was a 46 MW(t)/15 MW(e) prototype pebble-bed HTR which began operation in 1968 on a site adjacent to the KFA Juelich national laboratory (now called FZ Juelich). The AVR was used extensively as a test bed for the development and qualification of spherical fuel elements. A broad spectrum of coated-particle types was tested (Table 4-9), ranging from the initial core of HEU BISO (Th,U)C 2 fuel to highquality, LEU TRISO UO 2 reload fuel, beginning in Initially, the AVR operated with a core outlet temperature of 850 o C; in 1974, the outlet temperature was raised to 950 o C (Ref. 4-42); this temperature increase caused no serious operational problems, but the release rates of Sr and Cs from the older BISO-coated carbide fuel did increase significantly producing rather high plateout inventories in the primary circuit at EOL in In summary, the AVR proved to be a superb vehicle for the successful development and demonstration of pebble-bed reactor technology, especially for spherical fuel elements (e.g., Ref. 4-43). 4-25

144 Fuel Performance Data Base Table 4-9 AVR Fuel Reloads in the Years 1966 to 1987 Reload Number Insertion Date Fuel Type Number of Elements CP-Kernel Coating 0 7/66 UCC 30,155 Th,U)C 2 HTI BISO /68 T 7,510 Th,U)C 2 HTI BISO GK 17,770 Th,U)C 2 HTI BISO /70 GK 6,210 Th,U)C 2 HTI BISO /70 GK 25,970 Th,U)C 2 HTI BISO /71 GO-1 20,825 Th,U)O 2 HTI BISO /73 GO-1 7,840 Th,U)O 2 HTI BISO /73 GO-1 11,000 Th,U)O 2 HTI BISO /73 GLE-1 2,446 UO 2 LTI BISO /74 GFB-1 1,440 UO 2 Th /74 GFB-2 1,610 UO 2 ThO 2 LTI BISO 93 LTI LTI BISO TRISO 9 9/74 THTR-l 5,145 Th,U)0 2 HTI BISO /74 THTR-2 10,000 Th,U)O 2 HTI BISO /74 THTR-2 5,000 Th,U)O 2 HTI BISO /76 GO-1 11,325 Th,U)O 2 HTI BISO /76 GO-1 9,930 Th,U)O 2 HTI BISO /77 GFB-3 6,077 UC 2 ThO /77 GFB-5 5,354 UCO Th GFB-4 5,861 UC 2 ThO 2 LTI LTI BISO TRISO LTI TRISO 92 LTI LTI BISO TRISO 15 2/81 GO-2 6,087 Th,U)O 2 LTI TRISO /81 GO-3 11,547 Th,U)O 2 LTI BISO /82 GLE-3 24,615 UO 2 LTI TRISO /84 GLE-4 20,250 UO 2 LTI TRISO /85 GO-2 1 1,854 Th,U)O 2 LTI TRISO /86 THTR 15,228 Th,U)O 2 HTI BISO /87 GLE-4 8,740 UO 2 LTI TRISO Enrichment (% U-235) 4-26

145 Fuel Performance Data Base A large supply of irradiated fuel elements could be drawn from AVR. They were used for extensive characterization tests, for accident simulation tests, and for the demonstration of longterm storage behavior (including leaching tests in brine). It was demonstrated that the strength of the fuel elements did not deteriorate. A fundamental difficulty in evaluating irradiated fuel elements from the AVR is the exact irradiation history during their multiple passes through the bed core can not be determined with great certainty due to the stochastic nature of a ball flow in a pebble-bed core. The EOL burnup can be accurately determined by gamma scanning and radiochemistry. However, the accumulated fast neutron fluence and the irradiation temperatures have to be estimated from codes simulating sphere flow, thermohydraulics and nucleonics in the AVR core. Particularly difficult is the assessment of irradiation temperatures in the AVR reactor. This was only made possible with the development of 3D codes to simulate sphere flow, and the distribution of neutrons and temperatures in the core. In addition, a sophisticated experiment ( HTA-8 ) was conducted in which 200 unfueled spheres containing meltable wires were inserted into the AVR core. A variation of alloys were included that span the temperature range from 898 to 1280 C. These spheres were detected after one passage through the core and removed. Their radial position in the core could be derived from the exit time, and X-rays made the wires visible (melted or not), giving the maximum temperature during the core passage. Unfortunately, 1280 C was chosen as the highest melting point. Approximately 20% of all spheres had all wires melted and had therefore seen surface temperatures >1280 C. For an average passage of GLE-3 elements, a maximum surface temperature of 1150 C had been predicted during 950 C operation THTR HTR The Thorium-High-Temperature Reactor (THTR) was a 756 MW(t)/300 MW(e) pebble-bed power plant. The core consisted of 600,000 spherical fuel elements with HTI BISO-coated HEU (Th,U)0 2 particles. The plant operated for 423 EFPD, beginning in 1983, until final shutdown in 1988 (Ref. 4-44) because of a combination of factors, including anti-nuclear politics in Germany. Fuel performance during THTR operation was monitored by on-line measurements of circulating noble gas activity. The circulating activity increased by a factor of three between 70 and 150 EFPD and then remained relatively constant until 300 EFPD, after which the activity trended downward. It was concluded that initially the dominant source of gas release was asmanufactured uranium contamination in the fuel-element matrix and OPyC coatings 33. An additional source of fission product release developed during reactor operation as a result of mechanical damage to a relatively large number of fuel spheres from the insertion of control rods into the pebble bed. A strong correlation was found between the number of damaged fuel element in the core and the circulating gas activity. The upper limits on the fraction of exposed fuel kernels were estimated to be 8 x 10-5 for the entire core and 5 x 10-3 for damaged fuel elements. The THTR fission product release data are discussed in Section Since the HTI BISO coatings used in THTR are deposited at very high temperature (>1600 o C), such particles typically have relatively high heavy-metal contamination in the OPyC coating. 4-27

146 Fuel Performance Data Base High Temperature Test Reactor The High Temperature Test Reactor (HTTR) is a 30 MW(t) test reactor constructed at Oarai, Japan. The reactor is designed primarily to investigate nuclear process heat applications. A portion of the nuclear heat (10 MW) is transported from the prismatic core to the secondary cooling loop through an intermediate heat exchanger (IHX); the remainder of the heat is dumped to a pressurized water cooler. The plant is designed to operate initially with a core outlet temperature of 850 C and then at 950 C for process heat applications. Hydrogen production by both steam reforming of natural gas and thermochemical water splitting (sulfur-iodine cycle) will be investigated. The reactor achieved initial criticality in 1998 and full power with a core outlet temperature of 850 C in The HTTR fuel element is a pin-in-block design that was developed by the British HTGR program in the 1970s. The core consists of 30 fuel columns, and each fuel column has five blocks for a total of 150 blocks. The initial core has design life of 660 EFPD. As shown in Figure 4-3, the coated fuel particles are contained in annular fuel compacts which are encased in a graphite sleeve. These graphite fuel rods are then placed in holes in the hexagonal graphite block. The advantages of this design are that the graphite block can be used for multiple fuel loads (until its fast fluence limit is reached) and that fuel compact recovery for reprocessing (the reference Japanese fuel cycle) is simplified. The disadvantage is that the heat transfer from the fuel compact to the coolant is less efficient than with a multi-hole FSV-type prismatic fuel element. Consequently, HTTR fuel will run very hot (~1400 o C). Figure 4-3 HTTR Fuel Element Design 4-28

147 Fuel Performance Data Base The reference particle design for the first HTTR core is TRISO-coated LEU UO 2 fuel with a variable enrichment ( %, average 6 %) to permit power shaping in the core. The TRISO particle has a kernel diameter of 600 µm and layer thicknesses of 60 µm for buffer, 30 µm for the IPyC, µm for SiC, and 45 µm for the OPyC. The main differences compared to the FRG and U.S. designs are a larger particle kernel and a thinner buffer layer. The burnup design limit for the HTTR reference fuel is as low as 3.6 %FIMA. For the HTTR second core, the use of an advanced fuel is contemplated; options include: (1) reference particle with improved fabrication processes, (2) optimized particle design (e.g., a smaller kernel and thicker buffer and SiC layers to allow for a higher burnup) and (3) the development of a ZrC-coated particle. The fabrication of the initial core has been described, including a series of process improvements that significantly reduced the number of SiC defects as the manufacturing campaign proceeded (Ref. 4-45). The fission gas release has been monitored since the initial rise-to-power and will be compared to fuel performance predictions. Fuel specialists from Japan Atomic Energy Research Institute (JAERI) and Research Center Juelich (FZJ), Germany, have published a paper (Ref. 4-46) comparing their predictions of the fuel performance and fission product behavior in the HTTR under normal operating conditions. In the prediction of the core average failure fraction, the JAERI model gave a much earlier failure and about two times larger than the FZJ model. In the fission gas release prediction, the FZJ model showed a later increase of fractional release than JAERI model and rapidly increased towards the end of the fuel life time. As noted above, the HTTR fuel will run very hot because of the fuel-element mechanical design; however, the temperature gradients across the fuel compacts are low because of the low power density. Once the initial core has reached its design life time of 660 EFPD and destructive PIEs have been performed, it will be interesting to determine the degree of SiC/fission product interactions and to compare the results to irradiation capsule results which are characterized by much higher power densities and temperature gradients but much shorter fuel residence times. The other very interesting result from the initial HTTR core will be the amount of Ag-110m that is released to the primary coolant as well as the amount and distribution of Ag-110m that is found in the graphite sleeves of the cooler running elements. Since the primary coolant pressure in HTTR is 4 MPa, the previous suggestion (Ref. 4-47) that Ag is quantitatively retained by nuclear graphite at temperatures up of ~1050 o C at system pressures can be evaluated High Temperature Reactor-10 The 10 MW High Temperature Gas-cooled Reactor-Test Module (termed as HTR-10) has been constructed at the Institute of Nuclear Energy Technology (INET) of Tsinghua University, Beijing, China, to allow the Chinese to develop an expertise in HTGR technology for potential future applications, including electricity production in direct-cycle plants and process heat applications (e.g., Ref. 4-48). The HTR-10 is a pebble-bed HTR, based upon German plant and fuel technology. The reactor core and the steam generator are housed in two steel pressure vessels, which are arranged side by side. The reactor thermal power is 10MW and the mean power density of the core is 2 MW/m 3. The helium coolant in the primary system is designed at the pressure of 3.0 MPa and mean core inlet and outlet temperatures of 250 o C and 700 o C, respectively. 4-29

148 Fuel Performance Data Base The core contains 27,000 spherical fuel elements (6 cm in diameter) with TRISO-coated, 17% enriched 500 µm UO 2 kernels. The average discharge burnup after multiple passes through the core is 80,000 MWD/t. The TRISO coating design is the same as the reference German design. The as-manufactured quality of the first core has been reported (Ref. 4-49). The average burnleach value (sum of HM contamination and SiC defects) for the initial core was 5.0 x 10-5 (which can be compared to the German and U.S. specification of 6.0 x 10-5 ). The design and operation of the HTR-10 has proceeded more or less in parallel with the fuel development program; stated differently, Chinese production fuel elements were not irradiated to full-burnup before the HTR-10 received a license for initial power operation. Instead, four spherical fuel elements were sampled randomly from the first and second production runs and have been under irradiation in the Russian IVV-2M reactor since July The HTR-10 achieved initial criticality in December The burnup in the irradiation tests must precede the burnup in the HTR-10 core to provide assurance that the fuel will perform satisfactorily. Through October, 2002, the maximum burnup and fast neutron fluence of the test fuel elements had reached 90,000 MWd/t(U) and 1.1 x n/cm 2, respectively; the fuel performance results, as inferred from the on-line R/B measurements, have been quite satisfactory to date (Ref. 4-50). 4.4 Conclusions and Implications The international data base for the irradiation performance of TRISO-coated fuel particles is robust. TRISO-coated particles with a variety of fuel kernel compositions have been fabricated, using a range of process conditions, and irradiated in MTRs and in operating HTGRs. In the latter case, TRISO particles have been used as the driver fuel (Dragon, FSV, AVR reloads, HTTR, and HTR-10), and they have also been irradiated as fuel test elements in operating HTGRs as well (especially Dragon). The main sources of fission product release during normal operation are as-manufactured heavymetal contamination and particles with defective or failed coatings. The important mechanisms that can cause coating failure under irradiation have been identified as a result of irradiation testing and postirradiation examination (Section 2.3.3). Both structural/mechanical failure mechanisms and thermochemical mechanisms have been observed, and phenomenological models have been developed to predict them for reactor design and safety analysis. The existing data, principally the German data for LEU UO 2, indicate that a properly designed and fabricated TRISO fuel particle, which is free of as-manufactured defects, will not experience significant pressure-vessel failure to well beyond the current design goals for the on-going HTGR design programs. However, particles with certain as-manufactured coating defects, especially missing buffer layers, will experience high in-service failure rates. These defects can be controlled and minimized during fuel manufacture by proper product- and process specifications and the attendant QC/QA protocols. The properties of the PyC coatings are especially important for acceptable mechanical performance; however, the PyC material properties and related QC techniques are not currently adequate, and additional technology development will be required. 4-30

149 Fuel Performance Data Base Thermochemical mechanisms have been also been identified, including kernel migration, fission product/sic corrosion interactions, and thermal decomposition. Under normal operating conditions, fission product/sic corrosion reactions appear to be the ultimate performance limiting failure mechanism. With oxidic fuels, kernel migration can apparently be controlled by designing the particle to suppress CO formation from the reaction of excess oxygen liberated during fission with the carbon buffer layer; this can be accomplished by adding carbon to the kernel (i.e., using a UCO kernel) or by adding an oxygen getter, such as zirconium. The control of fission product/ SiC corrosion reactions appears to be more difficult. Lanthanide attack which is prominent with carbide kernels can be practically eliminated by the use of an oxide or oxycarbide kernel. However, controlling SiC attack by noble-metal fission products, especially Pd isotopes which are produced in quantity in LEU and Pu fuels, is problematic because the noble metals do not form oxides. For conventional TRISO particles, FP/SiC interactions will have to be controlled primarily by employing core designs that limit the time at high temperature (<1400 o C). In that regard, it will be interesting to note how the HTTR fuel performs after extended times at high temperatures. The German LEU UO 2 fuel has performed superbly under irradiation and can be considered fully qualified to at least 10% FIMA, fast fluences of 4 x n/rn 2, and temperatures of at least 1200 o C. A small amount of data at higher burnups suggests that there are no fundamental burnup limitations on TRISO-coated UO 2 for normal operation up to burnups approaching 20%. However, with ungettered UO 2, very significant CO partial pressures will be generated at these higher burnups which may affect pressure-vessel performance. Perhaps more importantly, there is some evidence of SiC corrosion by CO or fission product metals in particles with UO 2 kernels during post-irradiation heating of high burnup particles. Planned European Union-sponsored tests will address this uncertainty over the next several years. If CO generation at high burnups proves to be performance limiting, the problem can be solved by use of a UCO kernel with the proper stoichiometry for the design burnup. If the problem is caused by fission product metal corrosion of SiC, other options will be needed. In that regard, LEU UO 2 fuel may be acceptable for pebble-bed HTRs even with core outlet temperatures >850 o C (as noted earlier, the AVR operated for years with a core outlet temperature of 950ºC) because their typically low core power densities and low temperature gradients during normal operation should preclude significant kernel migration. However, with prismatic designs with their higher temperature gradients and higher burnups, the prudent choice is UCO with its CO suppression to eliminate kernel migration during normal operation (and possible CO corrosion of the SiC during core heatup accidents). In fact, this strategy has been adopted by U.S. AGR fuel development program with its mission of qualifying TRISO fuel for the NGNP and VHTR. One other fundamental lesson learned after four decades of TRISO fuel development deserves mentioned: viz., it is extremely unwise to make multiple simultaneous changes in the coated particle design. Experience also indicates that is essential to get early irradiation and postirradiation heating data before the effects of particle design changes can be reliably determined. This lesson was dramatically demonstrated by the TRISO-P experience. For this particle design, a significantly thicker and denser inner IPyC layer and an added porous outer layer yielded large improvements in the as-manufactured quality; however, the combined result of these improvements was an order-of-magnitude increase in the in-service failure rates compared to that of conventional US-made TRISO particles. 4-31

150

151 5 RADIONUCLIDE RELEASE DATA BASE The codes used to predict radionuclide transport in an HTGR core during normal operation were described in Section 3. These codes contain phenomenological component models and material property correlations to predict radionuclide transport in each material region in the core. Almost without exception these models and correlations have been derived from experimental data produced from various international HTGR development programs conducted over the past four decades and from operating HTGRs. This data base is summarized in two subsections: Section 5.1 summarizes the material property data base from which the input correlations for these codes were derived, and Section 5.2 describes the previous efforts to validate these codes by comparing code predictions with the observed radionuclide transport behavior in operating reactors and test facilities. Some of the more important data will be reproduced here, but in other cases, only the essential results will be provided along with the relevant citations; in all cases, the interested reader is encouraged to obtain the primary references. 5.1 Material Property Data The material property database provides quantitative values for attributes that provide the input to the codes listed in Section 3. For all such data, the seminal reference is TECDOC-978 (Ref. 5-1). The reference GA correlations and component models are documented and controlled in the Fuel Design Data Manual (Ref. 5-2) Transport in Kernels Fission Gas Release There is an extensive international database on the release of fission gases from HTGR core materials. Fission gases are completely retained by intact TRISO fuel particles. For those particles that have defective or failed SiC coatings, the radiologically important short-lived gases, including I-131, are retained by the outer pyrocarbon coating, even at the peak fuel temperatures that occur during depressurized core conduction cooldown events. Consequently, the dominant sources of fission gas release during normal operation and postulated accidents are (1) heavy-metal contamination outside the particle coatings and (2) exposed fuel kernels which occur from in-service coating failure. 34 The 1993 ORNL review by Martin (Ref. 5-3) is widely available. 5-1

152 Radionuclide Release Data Base The present data base for fission gas release from heavy-metal contamination and from failed particles under irradiation is derived primarily from TRIGA measurements on fuel-compact matrix doped with uranium and on laser-failed, irradiated fuel particles from capsules (e.g., Refs. 5-3 through 5-11). Examples of fission gas R/B data are given in Figures 2-19 and 2-20 which give the R/Bs for exposed fuel kernels as a function of temperature and isotope half life, respectively (Ref. 5-4). Additional data for gas release from failed UCO particles are given in Figure 5-1 which demonstrates that the half life dependence falls below the square root dependence associated with Fickian diffusion as the temperature is reduced; evidently, at lower temperatures, the diffusive component no longer dominates the athermal processes, such as recoil and knockout (Ref. 5-11). Isothermal, in-pile hydrolysis tests for the reaction of exposed kernels with water for LEU UCO fuel (HRB 17/18) were investigated at ORNL, and the temperature dependence of gas release from both unhydrolyzed and hydrolyzed LEU UCO fuel was addressed in the HFR B1 test which was conducted in HFR Petten in the Netherlands (e.g., Ref. 5-12). As illustrated in Figure 2-21, the general response of exposed UCO and UO 2 kernels to the introduction of water vapor is complex, involving a transient release of stored fission gases with a concomitant increase in the steady-state fractional release and, upon removal of the water vapor, a monotonic decline in the fractional release to prehydrolysis values (Section ). Irradiation tests indicate no strong burnup dependence for fission gas release from LEU UCO kernels up to a burnup of ~18% FIMA (e.g., Refs and 5-12). Circumstantial evidence from the NPR-1/-2 irradiations of HEU UCO fuel to ~75% FIMA suggest a large burnup dependence (5-10x increases) at the higher burnup (e.g., Ref. 5-13). Limited German data for intermediateto-high burnup UO 2 also suggest a burnup dependence for gas release (e.g., Ref. 5-1). While the existing international data base on gas release from U and Th fuels is extensive, there are relatively few measurements on LEU UCO fuel particles, and there are no direct measurements of the fission gas release characteristics of TRISO coated Pu fuels Fission Metal Release There is an extensive international database on the release of fission metals from oxide-based U and Th fuel kernels (e.g., 5-1, 5-3, 5-7 and 5-14). Only silver, cesium and palladium (and perhaps other noble metals) are diffusely released to a significant degree from oxide fuel kernels at normal operating temperatures; the other fission metals, including radiologically important Sr-90, are only released by fission recoil. Additional fission metals, including Sr, Ba, and the lanthanides are released from carbide kernels. There are German data for Cs, Sr and Ag retention in exposed oxide particles that were irradiated under near real-time conditions, as well as limited laboratory data on Cs release from ThO 2 kernels. There is a considerable amount of German data for diffusion of Cs, Sr and, to a lesser extent, Ag diffusion in exposed oxide-based fuels (Ref. 5-1). The present data base for fission metal diffusivities in fuel kernels is derived primarily from measurement on particles irradiated in accelerated capsules (Ref. 5-3). 5-2

153 Radionuclide Release Data Base Figure 5-1 Effect of Temperature on Half-Life Dependence for Gas Release 5-3

154 Radionuclide Release Data Base The reference GA correlations for kernel diffusivities include a very strong burnup dependence (Section ). This burnup dependence is derived primarily from measured Cs releases from ThO 2 with burnups from 1 6 % FIMA which is shown in Figure 5-2 (Ref. 5-15). The strong burnup dependence was derived primarily from the set of measurements at 1400 o C (plotted at 6 x 10-4 K -1 ). Whether this strong burnup dependence really applies to UCO kernels at burnups >20 % FIMA needs further confirmation. Figure 5-2 Reduced Diffusion Coefficients for Cs Transport in ThO 2 Kernels 5-4

155 Radionuclide Release Data Base The release of plutonium, americium and curium from fuel kernels of various compositions [(Th,U)O 2, UO 2, UC 2 and UCO] under irradiation at high temperature ( C) has been investigated (Refs through 5-18). The fractional releases from the kernels to the IPyC layer in intact particles were strongly dependent upon kernel composition. The actinides appear to be completely retained by the UO 2 kernel, but some release was observed from kernels containing as little as 3% UC 2. The apparent diffusion coefficient for Am in UC 2 was an order of magnitude higher than that for Pu at 1350 C. The diffusivity of Pu in irradiated MOX pellets [(U,Pu)O 2±x ] has been measured in the C range and shown to be a function of temperature, burnup and kernel stoichiometry (Ref. 5-19). However, the actinides in MOX fuel are typically fully saturated with oxygen and the actinides in substoichiometric PuO 2-x kernels may behave differently. There are no available data on the release of actinides from failed TRISO-coated PuO x particles Transport in Coatings There are considerable international data on the transport of fission products in the SiC and PyC coatings of TRISO fuel particles (e.g., Refs. 5-1 and 5-3). Most of the data were obtained by the heating of irradiated particles, and effective diffusion coefficients were deduced from the observed time histories of fission product release. However, much of the early SiC data were compromised because the test particle batches contained unknown amounts of defective or failed SiC coatings. This circumstance would imply an excessively large apparent diffusion coefficient. More recent international data suggest that volatile fission metals, like fission gases, do not significantly diffuse through intact SiC coatings under normal operating conditions - with the exception of Ag isotopes (and tritium) at the higher temperatures. Such postirradiation heating results are illustrated in Figure 5-3 which shows the data obtained by ORNL for Japanese LEU UO 2 TRISO particles from Capsule HRB-22 (Ref. 5-20). A batch of 25 particles was recovered from a fuel compact and heated at 1700 o C for 270 hours. No particle experienced complete coating failure; i.e., at least one PyC coating remained intact (complete failure of a single particle would have resulted in ~4% fractional release of Kr-85). The release profiles indicate that, at 1700 o C, Ag is rapidly diffusively released from intact TRISO particles, Kr is retained by PyC coatings, and Cs and Eu are slowly released as the SiC coating degrades. 5-5

156 Radionuclide Release Data Base Figure 5-3 Postirradiation Heating Data for UO 2 at 1700 C The data for fission metal diffusion in pyrocarbon coatings also come mainly from postirradiation heating tests, and some in-pile measurements, with earlier BISO-coated fuel particles with various kernel compositions, including UC 2, (Th,U)C 2, UO 2, (Th,U)O 2 and ThO 2. The data from carbide particles is perhaps easier to interpret because there is less hold up of the volatile metals in the kernels to complicate the release time history. An example of such data is given in Figure 5-4 for Cs diffusion in pyrocarbon coatings (Ref. 5-15). 5-6

157 Radionuclide Release Data Base Figure 5-4 Cs-in-PyC Diffusion Coefficients There is an unresolved controversy (e.g., see Section 4.6 of Ref. 5-3) regarding the diffusive transport of Cs isotopes through good intact SiC coatings (where good implies SiC coatings free of defects, no free Si or C phases, fine-grained, etc.). The earlier literature viewed postirradiation data such as that in Figure 5-3 and concluded that Cs was diffusively released from intact SiC at high temperatures (e.g., Ref. 5-15). Later, other researchers, prominently Goodin and Nabielek (Ref. 5-21), concluded that cesium was not diffusively released through good SiC until the coating began to degrade at high temperatures due to fission product corrosion and/or thermal decomposition. In fact, in their modeling work, the Cs release history 5-7

158 Radionuclide Release Data Base was interpreted as the primary indicator of the rate of degradation of the SiC coating. Contemporaneously with the Goodin/Nabielek work, a diffusion model was again proposed for Cs release at temperatures below which the SiC coating should be uncompromised (Ref. 5-22). In any case, apparent diffusion coefficients for Cs in SiC from a number of researchers are shown in Figure 5-5 (Ref. 5-3). Figure 5-5 Cs-in-SiC Diffusion Coefficients 5-8

159 Radionuclide Release Data Base Until this issue is definitively resolved, standard design practice, for completeness, is to calculate a diffusive release of Cs from intact particles during normal operation using a Fickian diffusion model and apparent diffusion coefficients; the contribution to the total release is unimportant compared to the Cs release from HM contamination, exposed kernels and partially failed particles (defective and/or failed SiC with at least one intact PyC coating). The key issue is the diffusive release of silver through the SiC of intact TRISO fuel particles at high temperatures. There are data from irradiation capsules, operating reactors and postirradiation heating tests that clearly demonstrate that a significant fraction (>1%) of the silver can be released from intact TRISO particles; however, there is considerable scatter in the Ag-in-SiC diffusion coefficients derived from these data (e.g., Refs. 5-1 and 5-3). Apparent diffusion coefficients for Ag in SiC from a number of researchers are shown in Figure 5-6 (Ref. 5-3). Figure 5-6 Ag-in-SiC Diffusion Coefficients 5-9

160 Radionuclide Release Data Base Standard design practice is to calculate a diffusive release of Ag isotopes (250-day Ag-110m is most important) from intact particles during normal operation using a Fickian diffusion model and apparent diffusion coefficients; typically, this contribution from intact particles dominates the total predicted Ag release from the core, especially for designs with core outlet temperatures 850 o C (e.g., see Section 6). However, as discussed in Section , data from the hightemperature annealing of individual irradiated fuel particles revealed that the fission product release behavior of the individual particles was not uniform (Figure 2-22), and large particle-toparticle variations in the release behavior of Ag-110m, Cs-134, Cs-137 and Eu-154 were observed. Consequently, it remains to be demonstrated whether the use of a Fickian model with effective diffusion coefficients derived from particle release data will give sufficient accuracy for predicting Ag release when applied to analysis of large populations of particles in irradiation capsules or in reactor fuel elements. The present data base for fission product transport in particle coatings resulted largely from diffusivity measurements for various fission products in laboratory tests (e.g., 5-3). These data are supported by limited in-pile data for Cs and Sr inferred from the results of irradiation experiments. These data imply that the effective diffusivities in SiC increase with increasing neutron fluence, presumably as a result of irradiation damage; this irradiation damage may anneal out during core conduction cooldown transients, but this effect has not been investigated. There are also data on the diffusive release of fission gases from BISO particles (e.g., Ref. 5-23), but the relevance of these data to the transport of gases in the OPyC coatings of TRISO particles has to be confirmed. In any case, this issue is not relevant at the fuel temperatures characteristic of normal operation; under this conditions, the short-lived fission gases, including radiologically important 8-day I-131, are quantitatively retained by dense PyC coatings. The issue becomes important for temperatures >~1700 o C for long times which leads to significant degradation of the SiC coating. There are limited data regarding the diffusion of Pu, Am and Cm in PyC and SiC coatings on uranium coated particles (Refs. 5-18, 5-20 and 5-24). The diffusion of these actinides in HTI and LTI PyC appeared to be relatively rapid at high temperature. The actinides were quantitatively retained by SiC to at least 1400 C; measurable releases were reported at 1600 C, but the condition of the SiC coatings in these experiments is unknown. There are no data available for Pu release from failed PuO x particles or transport in PyC and SiC coatings on coated Pu particles Transport in Matrix/Graphite The international data base for radionuclide transport in matrix and graphite has been reviewed and summarized previously (e.g., Refs. 5-1, 5-3 and 5-25). The data base for nuclear graphite is large in recognition of its effectiveness as a release barrier in HTGR cores. As described above for oxide-based fuel kernels, only cesium and silver nuclides effectively migrate through the fuel element graphite at normal operating temperatures. The other fission metals, including radiologically important Sr-90 and the actinides, are completely retained by the graphite during normal operation. However, the existing data are scattered, and there are considerable differences in the reported transport properties for different grades of nuclear graphite. 5-10

161 Radionuclide Release Data Base The fuel-compact matrix used in prismatic cores is relatively porous and provides little holdup of the fission gases which are released from the fuel particles (e.g., Ref. 5-3). However, the matrix is a composite material which has a high content of amorphous carbon, and this constituent of the matrix is highly sorptive of metallic fission products, especially Sr. While the matrix is highly sorptive of metals, it provides little diffusional resistance to the release of fission metals because of its high interconnected porosity. The matrix material used in spherical fuel elements is denser and somewhat more graphitized than the fuel-compact matrix; consequently, it does provide some diffusive resistance to the release of metallic fission products. The fuel-element graphite used in prismatic cores is denser and highly graphitized. It is less sorptive of the fission metals than matrix materials, but it is more effective as a diffusion barrier than the latter. Diffusion data are available for cesium, strontium and silver for German A3-3 matrix as well as the influence of irradiation and corrosion on the diffusion process; the Ag data are shown in Figure 5-7. The A3-3 matrix was used for the THTR fuel elements. Modern German fuel elements in the LEU UO 2 development program used A3-27 matrix. Comparative measurements by Hoinkis obtained diffusion coefficients for cesium and silver in as-received A3-27 at 1000 C that are lower by factors of 20 and 7, respectively, compared to as-received A3-3 (Ref. 5-26). Figure 5-7 Ag Diffusion Coefficients for German A3 Matrix 5-11

162 Radionuclide Release Data Base The present correlations for fission metal diffusivities in core graphite are derived largely from laboratory measurements on unirradiated nuclear graphites and from profile measurements in various irradiated graphites (e.g., Refs. 5-9 and 5-25; Appendix A of Ref. 5-1 provides a good summary). The data for Cs diffusion in various nuclear graphites are shown in Figure 5-8 (Ref. 5-15). The current correlation for Ag diffusivity in irradiated H-451 graphite (Ref. 5-2) was inferred from the measured Ag diffusivity in German A3 matrix (Figure 5-7). Figure 5-8 Cs Diffusivities for Nuclear Graphites The sorptivities of unirradiated and irradiated nuclear graphites and matrix materials for various fission metals and iodine have been investigated by the various international HTGR programs (Appendix A of Ref. 5-1 again provides a good summary). The sorptivities of Cs and Sr on H-451 and H-327 graphites and over petroleum pitch matrix materials (all materials used in Fort St. Vrain) have been measured in the laboratory at partial pressures >10-10 atm (several orders of magnitude higher than typical in-reactor partial pressures during normal operation). The data for Cs sorption on irradiated H-451 graphite is shown in Figure 5-9 (Ref. 5-15). The sorptivities of Cs and Sr on nuclear graphites have been shown to increase with increasing fast fluence, but the effect may anneal out at high temperature in the absence of a neutron flux. The sorptivity of 5-12

163 Radionuclide Release Data Base pitch matrix is independent of fast fluence. The sorptivities Cs, Sr and Ag on German thermosetting-resin matrix, including A3 matrix, have been measured and may apply to candidate U.S. resin matrix materials. There are limited laboratory data that indicate the vapor pressure of Cs over graphite increases in the presence of coolant impurities and as a consequence of partial graphite oxidation (Ref. 5-1). Figure 5-9 Cs Sorption on Irradiated H 451 Graphite There are no measurements of Ag transport in irradiated H-451 graphite. Early Dragon Project data suggest that Ag transport through graphite may be reduced dramatically by elevated system pressures at temperatures below ~1050 C (Ref. 5-27). The implication of this apparent pressure dependence is that Ag may migrate in the vapor phase through the interconnected porosity of the graphite. If so, in-reactor Ag releases might be significantly lower than those observed in irradiation capsules which are typically operated at or slightly above atmospheric pressure. In-pile loop tests at elevated pressures are needed to provide a definitive answer. 5-13

164 Radionuclide Release Data Base The diffusivity of plutonium in unirradiated H-451 graphite has been measured up to 1350 C (Ref. 5-28), and the desorption pressure of Pu sorbed on H-451 has been measured up to 1350 C (Ref. 5-29). Based upon these measured transport properties, the release of Pu from the graphite into the primary coolant was predicted to be negligible for both normal operation and depressurized core conduction cooldown conditions. In addition, the sorption of PuC on H-451 graphite (Ref. 5-30) and uranium diffusion in H-451 graphite have been measured (Ref. 5-31). The transport properties of actinides in compact-matrix material have not been measured. 5.2 Integral Radionuclide Release Data Integral radionuclide release data are available from irradiation capsules, in-pile loops and operating HTGRs. The validity of the design methods for predicting fission product release during normal operation has been assessed by evaluating their ability to reproduce these integral release data. The results of such comparisons are often ambiguous because the measurements are quite integral. The measured radionuclide releases from a fuel irradiation capsule or, even more so, from a reactor core represent the integral of multiple sources (HM contamination, failed particles, etc.), the fractional releases of radionuclides from each of the sources, and, in case of metallic fission products, the degree of attenuation by the compact matrix and fuel-element graphite. Moreover, large variations in fuel burnup, fast fluence, and especially temperature are common in irradiation capsules and inevitable in a reactor core. To model such systems, the sources (e.g., coating failure rates, etc.) and the radionuclide transport through each release barrier must be predicted. Consequently, when there is good apparent agreement between predictions and measurements, one can not in general be certain that there were not compensating errors. Likewise, when there are significant discrepancies, the cause(s) may be difficult to isolate. For some experimental programs, this ambiguity has been removed by including a known fission product source. For example, laser-failed (LF) particles, bare kernels, and designed-to-fail (DTF) particles (standard fuel kernels with a thin PyC seal coat) have been seeded in fuel compacts to provide a known source. Such seeded compacts have been irradiated in capsules and in-pile loops. The fission gas release from irradiation capsules containing LEU UCO/ThO 2 fuel is generally predicted to within a factor of about five (Ref. 5-32). However, these capsules operated dry, so the hydrolysis model was not tested independent of the data from which it was derived. Moreover, there is the aforementioned inherent ambiguity in these data since the fuel failure fraction is not known with high accuracy independent of the gas release data (certain PIE measurements do provide some independent indication of the particle failure fraction). The validity of the methods for predicting fission metal release from the core during normal operation have been assessed by applying the reference design methods to predict the observed metal release in irradiation capsules: e.g., SSL-1 (Ref. 5-33), SSL2 (Ref. 5-34), Idylle 03 (Refs and 5-36), and R2 K13 (Ref. 5-37); in-pile loops: e.g., four CPL2 loops (Ref. 5-38), and COMEDIE BD-1 (Ref. 5-39); and in operating HTGRs: e.g., Peach Bottom Core 2 (Ref. 5-14

165 Radionuclide Release Data Base 5-40) and FSV (Ref. 5-41). Most of the available data are for the Cs isotopes with a small amount of Ag and Sr data. The releases of fission metals were often underpredicted by large factors, and in some cases, by more than a factor of 10. The cause of the underpredictions is ambiguous because the SiC defect fractions and the particle failure fractions are typically not well known. However, there is circumstantial evidence that the transport across the fuel compact/fuel element gap and the transport in the graphite web are not properly modeled. The most significant of these methods validation efforts are described in greater detail in the following subsections. The emphasis here is placed on data for LEU UCO fuel and on data from operating HTGRs Irradiation Capsules R2 K13 Irradiation capsule R2-K13 (Ref. 5-42) was introduced in Section as an example of anomalous behavior in irradiation capsules. R2 K13 was an irradiation of GA-fabricated, TRISO-coated LEU UCO/ThO 2 in fuel compacts in H-451 graphite fuel bodies; the irradiation was performed in the R2 materials test reactor at Studsvik, Sweden. Cell 2 operated at timeaverage fuel temperature of 1190 o C, and Cell 3 operated at an average temperature of 980 o C. The measured Kr-85m R/B as a function of time is shown in Figure The R/B history for Cell 2 has features that are typical for irradiation capsules and can be easily rationalized in terms of additional particle failures taking place during the irradiation because it is a continuously increasing function of time. In contrast, the Kr-85m R/B for Cell 3 stays approximately constant (~2.0 x 10-7 ) for most of the irradiation and then takes a step increase by a factor of about 125 (to 2.5 x 10-5 ) at about 400 EFPD into the irradiation. Nondestructive fission gas release measurements were made on the as-built Cells 2 and 3 fuel bodies prior to irradiation, and the release rates were almost the same. Therefore, the difference in the initial in-pile R/Bs should only have reflected the 200 o C lower temperature in Cell 3; consequently, the reported early Cell 3 R/Bs are at least an order of magnitude lower than expected. Despite considerable investigation, no satisfactory explanation was ever provided for this unusual experimental observation; it is strongly suspected that the data prior to 400 EFPD are wrong These Cell 3 results are included here as an indication of anomalous experimental results that are not uncommon in the fuel performance and fission product transport data base. It is also noteworthy that the R2 K13 experiment was planned and conducted by experienced researchers and in compliance with QA requirements and protocols. 5-15

166 Radionuclide Release Data Base Figure 5-10 Kr-85m R/B versus Time for Capsule r2-k13 The measured and calculated R/Bs for Kr-85m for Cell 2 are compared in Figure 5-11; the agreement is quite good, considering that the accuracy goal for fission gas release predictions is typically a factor of four. No attempt was made to model the gas release from Cell 3 because of the anomalous behavior described above. 5-16

167 Radionuclide Release Data Base Figure 5-11 Comparison of Measured and Predicted Gas Release for R2-K13/Cell 2 The measured and calculated fractional releases of Cs-137 and Ag-110m are compared in Table The measured and predicted Cs-137 and Ag-110m profiles in the fuel-compact matrix and the fuel-body graphite are shown in Figures 2-23 and 6-12, respectively. The measured Cs-137 profiles in the fuel-compact matrix and graphite web are qualitatively as predicted (solid lines): (1) the concentration profile in the matrix is flat (high diffusion coefficient); (2) a partition factor is observed across the gap, (3) the concentration profile in the graphite decreases exponentially with penetration distance, and the profile is flatter at the higher temperature. However, the absolute concentrations in the matrix and graphite are significantly underpredicted for Cell 2, suggesting that the release from the fuel particles into the compact matrix was underpredicted. Surprisingly, the total Cs-137 release from the graphite fuel body for Cell 2 (1190 o C) was overpredicted by a factor of 1.2. In contrast, the concentration in the graphite for Cell 3 (980 o C) was slightly overpredicted, but the release from the body was underpredicted by a factor of The reported fractional releases of Ag-110m are based upon calculated total production of the radionuclide by neutron activation of stable Ag-109 which is produced primarily by fissions in bred Pu in the LEU fissile particle; hence, there is a degree of uncertainty (Ag-110m can not be easily detected by gamma spectroscopy of spent fuel.) 5-17

168 Radionuclide Release Data Base Table 5-1 Comparison of Measured and Calculated Metal Release in R2-K13 Fractional Release Nuclide Cell Measured Calculated Measured/ Calculated Cs x x Cs x x Ag-110m x x Ag-110m x x Figure 5-12 Ag Release from Capsule R2-K1 The measured Ag-110m profile for the Cell 2 graphite goes through a maximum which is not consistent with Fickian diffusion. It is perhaps a loading profile, reflecting the temperature drop (hence, increased sorptivity) across the web, rather than a diffusion profile; this possibility would be consistent with the earlier suggestion that Ag migrates in the gas phase through the interconnected porosity of the graphite. No Ag-110m profile was reported for Cell 3, and no explicit explanation was given in the final report (Ref. 5-42); most likely, the concentration of Ag-110m was low compared to that of Cs-137/Cs-134 and other gamma-emitting radionuclides in the graphite. The predicted Ag-110m fractional release for Cell 2 exactly matches the reported fractional release (recall the uncertainty in this value); the predicted release is dominated by diffusive release from intact particles. The reported fractional Ag-110m release 5-18

169 Radionuclide Release Data Base for Cell 3 is underpredicted by a factor of 8. The predicted release for this lower temperature cell (~200 o C lower) is dominated by release from failed particles; the predicted release from intact particles is negligible at these temperatures. Such discrepancies and apparent inconsistencies are common when attempting to predict metallic release from irradiation capsules. Explanations are confounded by the difficulty of isolating potential errors in the prediction of the fuel particle failure rates from potential errors in the prediction of fission product transport. The value of conducting fission product release experiments with a well defined fission product source is evident TRISO-P Capsules As introduced in Section 2.3.1, the TRISO-P particle (Ref. 5-43) was adopted as the reference particle for gas-cooled New Production Reactor and for the concurrent, commercial steam-cycle MHTGR program. The TRISO-P design featured both a significantly thicker and denser IPyC layer and an added porous protective (P-PyC) outer layer. Both design changes were made to solve perceived problems during fuel fabrication. The design changes resolved the process issues, and the as-manufactured quality of the fuel compacts was dramatically improved. TRISO-P particles with HEU UCO kernels were irradiated in capsules NPR-1, NPR-2 and NPR-1A (Ref. 5-44), and TRISO-P particles with LEU UCO and ThO 2 kernels were irradiated in capsule HRB-21 (Ref. 5-45). Under irradiation the thicker (and more anisotropic) IPyC developed radial cracks which served as stress risers and crack initiators in the SiC layer, and the porous P-PyC layers shrank excessively which caused a high fraction of the OPyC layers to fail. The fission gas release predictions from the TRISO-P particles in the three NPR capsules and the HRB-21 capsule were originally grossly under predicted with the reference design methods. The reasons were that (1) the FDDM failure models, which were based upon conventional 5-layer TRISO fuel, did not properly account for the coating failure mechanisms introduced with TRISO-P fuel; and (2) the FDDM fission gas release model did not account for the large burnup dependence observed for gas release from failed HEU UCO particles at burnups >30% FIMA. Empirical failure models were developed for TRISO-P particles based upon PIE measurements, and the fission gas release rates were determined by postirradiation R/B measurements on individual irradiated compacts from these capsules. From these R/B measurements, a burnup multiplier was derived and added to the reference gas release model (see Figure 6-5). With these revisions to the reference design methods, the in-pile fission gas release from the four TRISO-P capsules was reasonably well predicted (Fig. 4-2, Ref. 5-13). 5-19

170 Radionuclide Release Data Base In-Pile Loops CPL-2 Test Program The CPL-2 test program was a series of four in-pile loop tests in the PEGASE materials test reactor in Cadarache, France. 37 The essential features of the CPL-2 loop were a fuel element (representative of a prismatic block), reflector element, and a counter-current, shell-and-tube heat exchanger-recuperator. Selected fuel compacts contained 1-4% bare HEU UO 2 kernels in addition to TRISO-coated driver particles to provide a well defined fission product source. The fuel particles, fuel compacts, and graphite fuel blocks were fabricated by the CEA using French materials of construction (e.g., the fuel and reflector blocks were fabricated from Pechiney P 3 JHAN graphite). Four tests were performed in the CPL-2 program. Tests CPL-2/1 (Ref. 5-46), CPL2/1-Bis (Ref. 5-47), and CPL-2/3 (Ref. 5-48) were characterized by variable coolant chemistries during the two-month irradiations with the latter test operating with up to 100 ppm total oxidants for most of the irradiation. CPL-2/4 (Ref. 5-49) was an in situ depressurization test: the loop was operated under nominal conditions for 60 days and then rapidly depressurized with the effluent passing through a series of traps to recover the radionuclides released from the loop. The CPL-2 test program generated a broad spectrum of fission product transport data, including fission gas and metal release data, plateout distribution data, and reentrainment data. Attempts were made by GA in the 1970s to predict the fission gas release for comparison with the on-line R/B measurements and to predict fission metal transport in the various fuel materials for comparison with the PIE measurements. These predictions were made with the reference methods of that time period. The computer codes used were generally precursors of the current reference codes described in Section 3, but they were based upon similar transport models (e.g., Fickian diffusion in the fuel-element graphite with an evaporative boundary condition). The material properties (e.g., graphite diffusion coefficients) used were the reference GA correlations of the day, supplemented in some cases with measurements by the CEA for the actual French materials. A series of reports were prepared in the late 1970s, and a comprehensive summary report was prepared in 1985 (Ref. 5-38). In general, the radionuclide release data obtained in the CPL-2 tests was qualitatively consistent with the transport models of the day, but the data were much more complex. The bare UO 2 kernels used to provide a known fission product source were initially highly porous, and they densified under irradiation; consequently, the R/Bs decreased from initially high values by almost two orders of magnitude during the two-month irradiation, apparently reaching steadystate during the last week of the test. The reference gas release models are based upon measurements with initially dense fuel kernels, and they predicted an essentially constant R/B with time over the short irradiation period. The measured Cs profiles in the graphite fuel block were indicative of a two-path Cs transport process (Reference 5-38 suggests bulk diffusion coupled with faster grain-boundary diffusion). The Sr-90 release to the coolant appeared to be 37 The program was conducted under a private GA/CEA Joint Accord in the mid-1970s; later, evaluation of the test data at GA was funded by USERDA. 5-20

171 Radionuclide Release Data Base exclusively from the release and subsequent decay of its Kr-90 precursor as predicted by the reference models. Since HEU fuel was used in the four CPL-2 tests, no Ag release data was generated. It would be possible to re-analyze the CPL-2 fission product release data with the current design methods or with the improved methods that the AGR program should produce in the next five or so years. In fact, such a re-analysis of the Cs and I plateout distribution data was performed in the mid-1980s with a considerable improvement in the agreement between predictions and measurements. However, the issue of how representative these release data might be for future prismatic HTGR cores would have to be addressed before such an undertaking. Probably, the most attractive future application would be to use the data for Cs transport in graphite to test any new model (e.g., diffusion/trapping, coupled two path, etc.) COMEDIE BD-1 Test Program The CEA COMEDIE loop was an in-pile test facility in the SILOI materials test reactor in Grenoble, France (Ref. 5-50). This loop was designed with the specific goal of characterizing the release, transport, deposition and liftoff of fission products in HTGRs during normal operation and during rapid depressurization transients. The loop was capable of providing engineering scale, integral test data under realistic reactor operating conditions to validate the methodology used to predict HTGR source terms. The COMEDIE loop consisted of an in-pile section and an out-of-pile section. The in-pile section included a fuel element which was the source of fission products and also produced nuclear heating to operate the loop components at the desired temperatures. The fuel element (representative of a prismatic fuel element) contained fuel compacts seeded with a known fission product source (e.g., designed-to-fail particles, fuel kernels with a thin PyC seal coat) and coated-particle driver fuel. Immediately downstream of the fuel element was a graphite block, simulating a core reflector element, to determine the deposition of condensable fission products on core structural graphite. The reflector block was followed by a plateout section where condensable fission products were deposited. The plateout section was a straight tube, counterflow, gas-to-gas heat exchanger/recuperator simulating the steam generator and the other metallic components in the primary circuit of an HTGR. The BD-1 test in COMEDIE in 1992 was conducted to obtain integral test data to validate the methods and transport models used to predict fission product release from the core and plateout in the primary coolant circuit of a steam-cycle HTGR during normal operation and reentrainment ( liftoff ) during rapid depressurization transients. GA-made LEU UCO fuel and US materials of construction were used for the fuel and reflector elements and for the loop heat exchanger. The primary objective in the BD-1 test was to perform a series of successively more energetic in situ blowdown tests to characterize the liftoff of plateout activity as a result of large primary coolant leaks; that objective was met (Ref. 5-39). The COMEDIE BD-1 test also generated considerable data on fission product release from LEU UCO fuel. The fission gas release was predicted with the CAPPER code (Section 3) using the FDDM/F release models. The Kr-85m R/B at end-of-life was predicted to within 2x, but the 5-21

172 Radionuclide Release Data Base Xe-133 R/B was underpredicted by 5x, compared to the accuracy goal of 4x; these comparisons are shown in Figure In general, the dependence of R/B on isotope half-life was greater than predicted before the test. Data on fission metal release from LEU UCO fuel are also available from the BD-1 test. However, the value of the BD-1 metal release data was seriously compromised by the failure to perform the planned destructive PIE of the fuel element and reflector element. Figure 5-13 Comparison of Measured and Predicted Gas Release in COMEDIE BD-1 Release of key metallic radionuclides Cs-137, Cs-134, Ag-110m, Sr-90 and Sr-89 from fuel particles was modeled using the COPAR-FD code to predict the diffusive release from the fuel kernels to the fuel matrix material. Metallic transport from the fuel compact through the graphite web to the coolant was calculated by the TRAMP code using COPAR-FD results as input (Section 3). The cesium release was underpredicted by nearly a factor of 30, using the FDDM/F kernel diffusion correlation (this discrepancy is consistent with prior anecdotal evidence that this correlation with its strong burnup dependence underpredicts release of cesium for low burnups). However, Cs release was predicted to within a factor of 2 using the German UO 2 correlation (Ref. 5-51), well within the specified accuracy goal of 10x. The total Ag-110m production appears to have been significantly underestimated by CEA (the plateout inventory of Ag-110m on the loop heat exchanger far exceeded the calculated total production in the fuel); thus, detailed comparisons of predicted and measured releases are not meaningful. The modest amount of Sr-90 plateout measured on the loop heat exchanger was consistent with the prediction that the Sr-90 release was dominated by the release and subsequent decay of its gaseous Kr-90 precursor. 5-22

173 Radionuclide Release Data Base With the failure to perform a destructive PIE on the fuel and reflector element, the plateout and liftoff data became the most useful test data from the BD-1 test. The evaluation of the plateout distribution data, including comparisons with PADLOC code predictions, was described in the EPRI plateout review report (Ref. 5-52) Operating HTGRs Peach Bottom Core 2 The 40 MW(e) Peach Bottom Unit 1 reactor was the first operational gas-cooled reactor in the United States. It was owned and operated by Philadelphia Electric. The reactor was operated with two batch-loaded cores, each consisting of 804 fuel elements. When Core 2 achieved full design burnup in 1974, the plant was shutdown permanently and placed in SAFSTOR 38 mode for future decommissioning. This circumstance provided a unique opportunity for validation of HTGR design methods and codes in a prototypical HTGR environment over significant operating times. In response, the Peach Bottom End-of-Life Program, jointly funded by USDOE and EPRI, was conducted with the prime objective of validating core design codes, including fission product transport codes, and high-temperature materials performance (Ref. 5-40). The basic PB fuel element consisted of a graphite cylinder 3.5 in. in diameter and 12 ft long as shown in Figure Within each fuel element were 30 annular fuel compacts, each 3 in. long, consisting of coated fuel particles bonded in a carbonaceous matrix; Core 2 was fueled with BISO-coated HEU (Th,U)C 2 particles. Peach Bottom was designed with a fuel element purge system to control fission product release from the core. A purge stream of main coolant helium was drawn into the top of the fuel element (core exit) and flowed over the fuel compacts in the radial gaps between the stack of compacts and the external sleeve and central spine; this sweep gas passed through an internal trap and was then routed to the helium purification system for removal of chemical and radioactive impurities. The system worked quite well (circulating activity during Core 2 was <1 Ci compared to the "Design" activity of 4225 Ci. The particle coatings were originally included in the design to prevent hydrolysis of the carbide kernels during manufacture, not to retain fission products in core. The full-core release of fission gases from the fuel compacts into the purge stream was calculated with the PERFOR code, a module of the SURVEY code (Section 3). Performance models were added to predict failure of the (Th,U)C 2 particles from manufacturing defects, pressure vessel, and kernel migration; the code also modified to account for the single particle fuel, cylindrical fuel element geometry, and a batch-loaded core. The calculated and measured R/Bs (release rate into purge gas divided by birth rate in the fuel) for the reference nuclides Kr- 85m and Xe-138 are compared in Figure A decommissioning protocol characterized by those activities required to place and maintain a radioactive facility in such condition that the risk to safety is within acceptable bounds and that the facility can be safely stored (for up to ~100 yr) and subsequently decontaminated to levels which permit release of the facility for unrestricted use. 5-23

174 Radionuclide Release Data Base Core 2 fission gas release was accurately predicted, especially considering the uncertainties in the input data and the limited resolution of Nal detectors used to measure the isotopics (Ref. 5-40). The largest deviation for Kr-85m was a 2.5x underprediction near end-of-life which is within 4x accuracy goal. The dominant source of gas release was heavy-metal contamination. The observed dependence of R/B on half life (0.66 power for krypton and 0.60 for xenon) was greater than the expected 0.5 power, but measurement errors were suspected. Even so, the release of all observed isotopes was predicted to well within design margins. Figure 5-14 Peach Bottom Fuel Element 5-24

175 Radionuclide Release Data Base Figure 5-15 Comparison of Measured and Calculated Gas Release from PB Core 2 The fuel temperatures in PB Core 2 were very high ( 1400 o C) because the fuel compacts shrank so much that large radial gaps (40-50 mils) opened up between the external sleeve and the fuel compacts. As a result of these high temperatures, there was a massive diffusive release of Cs- 137/Cs-134 from the BISO carbide fuel in the hottest fuel elements. Fortunately, the Cs released from the hottest fuel compacts was carried by the purge flow toward the colder end of the fuel elements where the cesium resorbed on the colder fuel compacts and on the graphite sleeve and spine. The exact value is unknown, but something like 10% of the Cs produced in Core 2 was released from the fuel particles. However, because the action of the purge flow and retention by the graphite sleeve, the core-average fractional release into the primary circuit was only 2 x There was also a significant, but less well characterized, release of Sr-90 from the particles. This redistribution of Cs isotopes within the hottest PB Core 2 elements is clearly evident from the axial gamma scans which were performed during the postirradiation examination of several driver fuel elements at ORNL as a part of an extensive end-of-life R&D program. As shown in Figure 5-16a, the axial distributions of Cs-137/Cs-134 in fuel elements that ran at peak fuel temperatures of ~1200 o C, such as element E1401 (Ref. 5-53), tracked the axial power distribution and exhibited an approximately cosine shape. However, as shown in Figure 5-16b, the axial Cs distributions in fuel elements that ran significantly hotter, such as element E0301 (Ref. 5-54) which operated with a peak temperature of ~1400 o C, exhibit a dramatic depletion of Cs isotopes from the hottest compacts and their resorption on the colder compacts. Other more refractory gamma-emitting radionuclides, such as Zr-95 and Ce-144, showed the same characteristic cosine distribution even for the hottest elements so the Cs redistribution was confirmed. 5-25

176 Radionuclide Release Data Base (a) Cs134 and Cs137 Inventories in PB Element E1401 (Tmax = ~1200 C (b) Cs134 and Cs137 Inventories in PB Element F0301 (Tmax = ~1400 C) Figure 5-16 Cs Redistribution within Peach Bottom Core 2 Fuel Elements 5-26

177 Radionuclide Release Data Base This diffusive release and redistribution of Cs isotopes within PB Core 2 fuel elements was modeled with some success (Refs and 5-56). The FIPER Q code (Ref. 57), a precursor to the TRAFIC and TRAMP codes (Section 3), was modified to simulate axial transport in the helium purge flow. Analyses and comparisons were made for six fuel elements that underwent destructive PIE at ORNL. Comparison areas included axial and radial profiles for cesium and strontium in the element spine, fuel compacts, and sleeve; metallic inventories in these locations; and Cs and Sr release into the purge stream and primary circuit. Sensitivity studies were performed to assess effects of diffusion coefficients, fuel thermal conductivity, operating temperature, and power. An example of the predicted and measured Cs profiles in Fuel Element E0101 are given in Figure 5-17 (the radial profile shown is at the axial elevation of compact 28 near the top of the core which is a high-temperature location). The V shaped, measured radial profile in the sleeve is typical of the measured profiles in the PB element sleeves. The PB sleeves were triple impregnated with char to reduce the permeability (to improve the efficiency of the purge system). It is reasonable to expect that the char was preferentially deposited near the inner and outer surfaces of the sleeve. Consequently, the measured profiles may well be loading profiles reflecting the sorption of Cs on the non-uniformly distributed char in the sleeve. The total core releases were estimated in 1981 from the six PIE elements and three additional elements. The sensitivity studies demonstrated that the calculated Cs releases from the individual fuel elements was remarkably sensitive to the choice of material properties; in particular, the assumed cesium diffusivity for the Peach Bottom sleeve graphite (impregnated HLM-85) dramatically influenced calculated cesium release. The limited data available indicate that in-pile diffusivities (after significant irradiation) of most graphites are similar, although the diffusivities in the unirradiated materials may vary considerably. It was assumed, therefore, that the then reference (FDDM/C) diffusivity data (based on the in-pile behavior of H-451, H-327, and Speer graphites) applied to the Peach Bottom graphite. With this assumption, the estimated total core Cs release was a factor of three less than the measured fractional release 2 x 10-4 (Ref. 5-56) Fort St. Vrain The Fort St. Vrain Nuclear Generating Station was a 842 MW(t) HTGR that was operated by the Public Service Company of Colorado from 1974 to The FSV core used prismatic fuel elements with hexagonal cross-sections (e.g., Ref. 5-58). (The FSV fuel element was essentially identical to that specified for the commercial GT-MHR, and it will be the point of departure for optimization of the fuel element for the prismatic NGNP concept.) Fuel for the reactor was based on the 93% enriched uranium/thorium cycle (HEU/Th). Separate, TRISO-coated (Th,U)C 2 fissile and ThC 2 fertile particles were used. Fuel lifetime in the core was six years; about one-sixth of the fuel elements were removed and replaced at each refueling. The FSV fuel particle and compact designs were described in Section

178 Radionuclide Release Data Base Figure 5-17 Cs-137 Distribution in Fuel Element E

179 Radionuclide Release Data Base The standard GA fuel and reactor core design codes were used to predict the fuel performance and fission product release during operation. As-operated core power levels, corresponding 3-D power distributions, and measured and calculated coolant flow rates and temperatures were used to calculate the fuel-element temperature histories at a large number of locations in the reactor core. Three-dimensional burnup and fast neutron fluence distributions were calculated as a function of time based on actual reactor power and control rod positions. These temperatures, burnups, fast neutron fluences, and flow rates were used in the core design codes to predict fuel performance and fission product transport. The as-manufactured condition of the fuel as it was loaded into the core was represented in the core design codes by fissile and fertile material contamination in the fuel compacts, the number of particles with defective buffers and other coating layers, all measured as part of the fuel quality control program during the fabrication of the fuel. The reference fuel performance computer codes SURVEY and the metallic fission product release code TRAFIC (Section 3) were used. The FDDM/F fuel performance and fission product transport models were used except for certain revisions in the fuel failure and fission gas release models for application to FSV fuel. The most significant changes are in the performance model for irradiation-induced OPyC failure for FSV fuel, and in the diffusion parameter for the release of xenon from heavy-metal contamination. The revisions were based upon earlier FSV surveillance data and other irradiation capsule data for FSV TRISO fuel particles which showed higher OPyC failure rates than predicted with the FDDM/F model Fission Gas Release Radionuclides in the circulating coolant and plated out on the surfaces of the primary circuit provide information on the performance of the FSV fuel. Grab samples were periodically removed from the primary coolant system of FSV and the radionuclide content of the sample determined (e.g., Ref. 5-59). The SURVEY code calculated the fuel failure distribution and the full-core fission gas release. The predicted and measured fission gas release histories for Kr-85m are compared in Figure In the figure, the measured release rates are shown as points; the calculated R/Bs from asmanufacture heavy metal contamination plus release from particles whose coatings failed in service are shown as a solid line, and the R/Bs from the HM contamination alone are shown as a dashed line. The dominant source of coating failure in FSV was predicted to be the fast fluenceinduced failure of OPyC coatings on particles with as-manufactured SiC defects, resulting in an exposed kernel. However, the measured R/Bs indicate that for most of the operation of FSV there was less coating failure than predicted. The Kr-85m measurements indicate that the release contribution from failed particles became significant in cycle 4 after about 1400 EFPD of operation. 5-29

180 Radionuclide Release Data Base Figure 5-18 Predicted and Measured Kr 85m R/B in FSV The noble gas release from FSV at end-of-life was overpredicted by about 2x, well within the 4x accuracy goal; the cause of this overprediction is ambiguous. Fuel coating failure may have been overpredicted, the long-term effect of hydrolysis may be less severe than observed in lab tests, or a combination of both these effects may be the cause (Ref. 5-58). No destructive PIEs were performed on FSV fuel elements so no independent estimate of the particle failure fraction is available Fission Metal Release Data were obtained on the metallic fission product release from the FSV core into the primary coolant circuit from two plateout probes (Refs through 5-61). These probes were designed to sample vapor and aerosol-borne condensable fission products by drawing a small portion of the coolant into the probe. Each probe drew samples from the inlet to the steam generator and a second sample from the coolant in the lower plenum prior to return of the coolant to the core inlet. Plated out metallic fission products were also measured on one of the circulators located downstream from the steam generators (Ref. 5-62). The TRAFIC computer code was used to predict the release of fission product metals from the core during FSV operation. Analyses were performed for the key nuclides including Sr-90, Cs-134, and Cs-137. The calculated cumulative releases are compared with the actual total releases derived from the probe data in Table 5-2. The Sr-90 plateout based on the plateout probe data was overpredicted by ~40% primarily due to overprediction of the amount of its gaseous precursor Kr-90 (Sr-90 plateout comes almost exclusively from this source). 5-30

181 Radionuclide Release Data Base Table 5-2 Comparison of Measured and Calculated FSV Metal Release Cumulative Release (Ci) Nuclide Measured Calculated Measured/Calculated FSAR 30-yr Expected Sr Cs Cs with the predictions, well within the design goals of 4x and 10x for predicting gas and metal release, respectively. Based upon the plateout probe data and plateout data from the C2105 circulator, the fission product inventories are significantly lower than the Expected values given in the FSAR. Since the fuel in FSV was HEU/Th, the total production of Ag isotopes was low compared to the Cs isotopes. While low concentrations of Ag 110m were occasionally detected by gamma scanning of the aforementioned circulators and plateout probes, these data were inadequate to define the total Ag plateout inventory or distribution in the primary circuit AVR HTR A brief overview of AVR operation and the various spherical fuel elements tested therein was provided in Section Continuous measurements of noble-gas activity were conducted in the AVR (Ref. 5-63). Particle failures could then be detected by a considerable increase of the activity during the period Also with the increase of the average gas outlet temperature up to 950 C, a prompt increase of the noble gas activity by 25% was observed which is in the expected range of statistical deviation during normal operation (Ref. 5-64). Elevated values found since 1979 may be partly attributed to the water ingress accident in The relatively low fission gas values since 1983 are due to the increased fraction of low releasing LEU fuel (see Figure 5-19). A typical distribution of specific activities in the hot coolant as measured in the period 1984 to 1987 at full power and 950 C gas outlet temperature is presented in Table 5-3 (Ref. 5-64) The activities of condensable fission products are several orders of magnitude lower compared to the gaseous species. The measured activity in the coolant mostly originated from the heavy metal contamination in the fuel element matrix. Cesium profiles in fuel elements with normal particle performance show an increase of concentration near the surface due to re-adsorption of cesium from the gaseous phase in cooler parts of the core (cross-contamination). The cesium level in the fuel free zone is much higher compared to profiles in an MTR-irradiated fuel element of the same type. It originates from the high levels of fission product metals in the primary circuit with old fuel releasing Cs, Ag, and Sr from contamination and particle failure. However, the reduction in this adsorbed activity observed over the years reflects the gradual replacement of old BISO fuel by high-quality TRISO fuel in the AVR core. 5-31

182 Radionuclide Release Data Base Figure 5-19 AVR Coolant Temperature and Fission Gas Release from Table 5-3 Specific Activities in AVR Primary Coolant Nuclide Specific Activity (Bq/m 3 ) Total noble gases 4.6 x 10 8 H x 10 7 C x 10 7 Co x 10 1 Sr x 10 2 Ag-110 m 4.9 x 10 1 I x 10 2 Cs x

183 Radionuclide Release Data Base THTR HTR A brief overview of the THTR was provided in Section Fuel performance during THTR operation was monitored by on-line measurements of circulating noble gas activity (Figure 5-20, Ref. 5-65). The circulating activity increased by a factor of three between 70 and 150 EFPD and then remained relatively constant until 300 EFPD after which the activity trended downward. It was concluded that, initially, the dominant source of gas release was as-manufactured uranium contamination in the fuel-element matrix and OPyC coatings 39. An additional source of fission product release developed during reactor operation as a result of mechanical damage to a relatively large number of fuel spheres from the insertion of control rods into the pebble bed. A strong correlation was found between the number of damaged fuel elements in the core and the circulating gas activity. The upper limits on the fraction of exposed fuel kernels were estimated to be 8 x 10-5 for the entire core and 5 x 10-3 for damaged fuel elements. Despite this additional source of fission gas release, the total circulating activity remained well below the licensed allowable activity (<4% of the Design activity). Figure 5-20 THTR Coolant Gas Activity 39 Since the HTI BISO coatings used in THTR are deposited at very high temperature, such particles typically have relatively high heavy-metal contamination in the OPyC coating. 5-33

184 Radionuclide Release Data Base In addition to the noble gases, condensable fission products were undoubtedly released from the THTR core as well, especially considering the use of BISO coatings and the mechanical damage to fuel elements from control rod insertion. One would certainly expect significant quantities of Cs-134/-137 in the primary circuit (and of I-131 which has long since decayed). Since the THTR fuel was also HEU/Th, the total production of Ag isotopes was low compared to the Cs isotopes, but its fractional release from the BISO fuel would have been high. Nevertheless, no data on measured plateout inventories in THTR are reported in TECDOC-978 or in the primary THTR references cited therein (e.g., Ref. 5-65). The only exception is the measurement of the radionuclide contamination levels on samples of dust recovered from THTR which are discussed below. A 1997 paper (Ref. 5-66) which describes the decommissioning of THTR (it was placed in the German equivalent of the U.S. SAFSTOR decommissioning option) makes no mention of the estimated plateout inventories in the primary circuit at EOL. One likely reason that plateout in the THTR primary circuit has received so little apparent attention is that THTR, like FSV, had a PCRV which serves as a massive biological shield. When the primary circuit components are embedded in a PCRV, in situ gamma scanning is practically precluded. Under these circumstances, the plateout inventories can only be estimated if the plant instrumentation included plateout probes and whenever primary circuit components are removed from the PCRV. The radionuclide loadings on graphite dust collected in the THTR from the moisture sensors were measured after 300 EFPD as part of an effort to estimate the total dust production (Ref. 5-66). The estimated range of dust production was kg with an expected value of 25 kg. The distribution of dust within the primary circuit depended upon on particle size, coolant velocity, and geometry. The specific activity was relatively high, 2 x 10 8 Bq/g maximum, mainly caused by the radionuclides Co-60, Sr-95 40, Hf-181, and Pa-235 (Ref. 5-67) due to the enhanced particle failure from fuel element fracture. The measurements for the radiologically relevant cesium and silver isotopes were 1.1 x 10 6 Bq/g for Cs-137, 5.7 x 10 5 Bq/g for Cs-134, and 5.0 x 105 Bq/g for Ag-110 m (Ref. 5-68). If the Cs plateout were primarily on dust, if these loadings were characteristic, and if 25 kg is a good estimate of the total dust in the primary circuit, then the total Cs-137 plateout inventory was <1 Ci which corresponds to a fractional release from the core of <1 x 10-6 which seems very unlikely with BISO fuel and mechanically failed fuel spheres. 5.3 Conclusions and Implications The international data base for fission product release from HTGR cores during normal operation is also robust. Data for fission product transport in core materials are available from a variety of sources: out-of-pile laboratory experiments, in-pile experiments in research reactors (especially the GA TRIGA reactor), in-pile irradiation capsules and loops in MTRs, and operating HTGRs. 40 The presence of 25-s Sr-95 on THTR dust samples, as reported in TECDOC-978, is not credible; presumably, this is a typographical error, and the detected nuclide was 64-day Zr-95. The reported presence of 24.4-m Pa-235 is also not credible and should be taken as 27-d Pa-233. With HEU/Th fuel, the presence of nuclides such as Zr-95 and Pa-233 is a clear indication of fuel debris in the primary circuit from broken fuel spheres since these nuclides are not released from intact HTR fuel elements even at extreme temperatures. 5-34

185 Radionuclide Release Data Base While the data are extensive, some of the data are ambiguous because the fission product source (e.g., particle failure fraction) is not well defined. The data base is more complete for German LEU UO 2 fuel than for U.S. LEU UCO fuel. In general, the uncertainties in the fission product transport data and correlations are large; the estimated uncertainties often exceed an order of magnitude. The main sources of fission product release during normal operation are as-manufactured heavymetal contamination and particles with defective or failed coatings. While the SiC coatings are the most important barrier to radionuclide release from an HTGR core, the kernels, compact matrix, and fuel-element graphite are also important barriers. Credit is taken for all of these barriers when predicting radionuclide release from the core. Hence, radionuclide transport in each of these materials has been investigated and partially characterized. Fission gases, including radiologically important I-131, are completely retained by intact TRISO particles. Even in the case of coating failure, >95% of the short-lived fission gases are retained by the fuel kernel. The fractional release of fission gases can be increased if the exposed kernels are hydrolyzed by reaction with trace amounts of water that may be introduced into the primary coolant. The volatile metals (e.g., Cs, Ag, and Sr) can, at sufficiently high temperatures for sufficiently long times, diffuse through the SiC coating and be released from intact TRISO particles; however, diffusive release from intact particles is only significant compared to other sources for silver release. The fuel-compact matrix and fuel-element graphite do not significantly retain fission gases; however, they are important barriers to the release of fission metals and actinides. Radiologically important Sr-90 and the actinides are quantitatively retained by the matrix and graphite during normal operation. The retention of Cs isotopes is strongly dependent upon the temperature. The matrix used in spherical fuel elements provides limited retention of Ag-110m at high temperatures; Ag transport in prismatic core graphites has not been characterized. Early Dragon Project data suggest that Ag transport through graphite may be reduced dramatically by elevated system pressures at temperatures below ~1050 C, but this possibility needs to be confirmed. Design methods have been developed to predict fuel performance and fission product release from the core, and these methods have been applied extensively by international design organizations to support the design and safety analysis for various HTGRs, including prismaticand pebble-bed HTGRs. Despite numerous attempts to benchmark code predictions against experimental data, the design methods have not been fully verified and validated either by pragmatic standards or by formal V&V protocols. The models for fission product transport in core materials generally assume that solid-state Fickian diffusion is an adequate representation of phenomena which are obviously more complex. The adequacy of these models for core design and safety analysis has yet to be convincingly demonstrated Radionuclide transport in core materials is typically exponentially temperature dependent; for fission gases, the activation energies are modest, but for fission metals, the activation energies are often large. Consequently, radionuclide release from the core during normal operation will likely become an increasing important design issue as core outlet temperatures are increased for 5-35

186 Radionuclide Release Data Base more efficient electricity production and for process heat applications. In particular, diffusive release of Ag-110m from intact TRISO particles and reduced Cs-134/-137 retention by the fuel element graphite will need to be addressed. Technology development will be required to complete the development and validation of reliable design methods to predict radionuclide release from HTGR cores. Single-effects tests are required to refine the transport models and to reduce the uncertainties in the material property data correlations which serve as input to the models. Independent integral tests will be necessary to confirm and convincingly demonstrate the validity of these improved design methods for predicting transport. 5-36

187 6 PREDICTED RADIONUCLIDE RELEASE FROM DIRECT-CYCLE HTGR CORES A full-core, fuel performance and fission product release analysis has not yet been conducted for a commercial GT-MHR with LEU fuel; however, such an analysis has been performed for the direct-cycle PC-MHR with WPu fuel. 41 The results of the PC-MHR analysis (Ref. 6-1), which are summarized in this section, should provide a conservative upper bound on the performance of an LEU-fueled commercial GT-MHR for the reasons discussed below. First, the burnups in the PC-MHR are much higher than in the GT-MHR (87% versus 26%, respectively). Secondly, the fuel temperatures are higher in the PC-MHR core because it employs a single fissile particle; the GT-MHR design includes both 19.8%-enriched LEU fissile and natural U fertile particles which permits better fuel zoning and power shaping. From a fission product transport perspective, the LEU-fueled GT-MHR and WPu-fueled PC-MHR designs are largely equivalent with several important exceptions. The inventories of individual radionuclides in the core are different because the fission yields are different for different fissile nuclides. Generally, the differences are small (compared to the uncertainties in the core release rates), but there is a notable exception: the total production of Ag-110m in the LEU core is about four times lower than in the WPu core because the yields of its Ag-109 precursor are much higher for Pu fissions than for U fissions (this difference in yields is mitigated by the fact that with a high-burnup LEU core about 25% of the fissions take place in bred Pu-239). Finally, the uncertainties in the fuel performance and fission product release rates are smaller for LEU fuel than for WPu fuel since the experimental data base for the former is much more extensive; consequently, the design margins ( Design / Maximum Expected, see Section 2.5) adopted by GA for WPu cores are factor of 10 larger than those for LEU cores. 6.1 PC-MHR Description The Plutonium Consumption-Modular Helium Reactor (PC-MHR) program was conducted at General Atomics during under USDOE funding (Refs. 6-1 and 6-2). The PC-MHR program demonstrated the superior potential of a graphite-moderated HTGR to efficiently destroy surplus weapons-grade plutonium in a single pass through the reactor. A 600 MW(t) 41 The PC-MHR analysis was performed by V. Jovanovic of General Atomics as part of the USDOE-funded MHTGR Plutonium Consumption Study (Refs. 6-1 and 6-2). Reference 6-1 is not generally available (it is not subject to any Applied Technology or UCNI restrictions) so the section on fuel performance is excepted here. 6-1

188 Predicted Radionuclide Release From Direct-Cycle HTGR Cores direct-cycle PC-MHR was shown to be capable of destroying almost 90% of the Pu-239 (and >60% of the total Pu) in a single pass while producing electricity with a 48% thermal efficiency. Moreover, the combination of a graphite core, Pu fuel encapsulated in TRISO-coated particles, inert He coolant, and the large heat capacity of the reactor core results in a plant design which has significant safety advantages. A standard PC-MHR plant unit would include four reactor modules, the supporting fuel fabrication facilities, and the spent fuel storage facilities. The arrangement of a typical PC-MHR module is shown in Figure 6-1, and the key design parameters are summarized in Table 6-1. A 600 MW(t) module produces 286 MW(e) of power by the transfer of energy from the reactor directly to a helium turbine which is coupled to an electrical generator. Each module is located within a steel lined, reinforced concrete, high pressure, low leakage reactor containment structure which is located underground. 6-2 Figure 6-1 PC-MHR Module

189 Predicted Radionuclide Release From Direct-Cycle HTGR Cores Table 6-1 PC-MHR Key Design Parameters Parameter Design Value Plant design Thermal power Outlet Temperature Direct-cycle HTGR 600 MW/module; 4 module/standard plant 850 o C Thermal efficiency 48% Feedstock Fuel form Fuel cycle Fuel burnup Surplus weapons Pu (94% Pu-239) TRISO PuO 1.68 particles in prismatic fuel element Once-through <85% FIMA Fast Fluence <5 x n/m 2 Safety & licensing Final waste form EPA PAGs (e.g., 5 rem thyroid, 1 rem WB) Whole-element disposal without processing 6-3

190 Predicted Radionuclide Release From Direct-Cycle HTGR Cores The active core contains columns of hexagonal fuel elements which are arranged in an annular array with columns of inner and outer graphite reflector elements. A plan view of the arrangement is shown in Figure 6-2, which is almost identical to configuration for the commercial GT-MHR shown in Figure 2-4. The active core contains 1020 fuel elements in 102 columns, arranged in three concentric rings of fuel elements. Of these, 72 columns contain standard elements having only fuel and burnable poison materials. The remaining 30 columns contain control rod elements in 12 column locations and reserve shutdown channels in 18 column locations. The fuel element components are shown in Figure 6-3, which again are almost identical to those for the commercial GT-MHR shown in Figure 1-5. The primary differences are in the dimensions of the coated particles which are adjusted to reflect the use of highly enriched WPu feedstock instead of the LEU feedstock in the commercial designs. The particle designs are compared in Table 6-2. Figure 6-2 PC-MHR Annular Core 6-4

191 Predicted Radionuclide Release From Direct-Cycle HTGR Cores Figure 6-3 PC-MHR Fuel Components 6-5

192 Predicted Radionuclide Release From Direct-Cycle HTGR Cores Table 6-2 Comparison of PC-MHR and GT-MHR Particle Designs Parameter Commercial GT-MHR Fissile Particle Fertile Particle PC-MHR Composition UC 0.5 O 1.5 UC 0.5 O 1.5 PuO 1.60 Enrichment, % 19.8 (U-235) 0.7 (natural U) 94 (Pu-239) Design burnup (% FIMA) Dimensions (µm) Kernel Diameter Buffer thickness IPyC thickness SiC thickness OPyC thickness Particle diameter Material Densities (g/cm 3 ) Kernel Buffer IPyC SiC OPyC

193 Predicted Radionuclide Release From Direct-Cycle HTGR Cores The specification for the substoichiometric PuO 1.68 kernel is the same as the particle successfully tested in Peach Bottom Fuel Test Element (FTE)-13 to peak temperatures of 1440 C, peak burnup of 70% FIMA, and peak fast neutron fluence of 2.2 x n!m 2 (Refs. 6-3 and 6-4). Particles with PuO 1.68 kernels did not exhibit kernel migration whereas particles with PuO 1.81 kernels migrated excessively; an irradiated PuO 1.68 particle is shown in Figure 6-4 (Ref. 6-4). Mixtures of plutonium oxide and plutonium carbide have also been successfully irradiated to high burnup in TRISO coated particles by the Dragon Project during the early 1970s (Ref. 6-5). The TRISO coating system specification is essentially same as the system developed and qualified by the German program and adopted by the AGR fuel development program. Figure 6-4 PuO 1.68 Particle Irradiated to 70% FIMA in FTE

194 Predicted Radionuclide Release From Direct-Cycle HTGR Cores The key fuel performance and fuel quality requirements for the PC-MHR are compared to those for the commercial GT-MHR in Table 6-3 (also given in Table 2-3). Several differences are noteworthy. First, the specification on missing-buffer layers is a factor of five higher for the PC-MHR; this will prove important because the failure of particles with missing-buffer layers is typically the dominant cause of exposed kernels. Secondly, the margins (ratio of 95%/50% values) are an order-of-magnitude higher for the PC-MHR. These very large margins were adopted for two reasons: (1) the paucity of data on the performance of high-quality PuO x fuels; and (2) a programmatic scheduling requirement which required the plant design to proceed almost in parallel with the Pu fuel development program (Ref. 6-1). Such large design margins would not be expected for a commercial GT-MHR or VHTR. Table 6-3 Comparison of PC-MHR and GT-MHR Fuel Requirements Parameter As-Manufactured Fuel Quality Commercial GT-MHR >50% Confidence >95% Confidence >50% Confidence PC-MHR >95% Confidence Missing or defective buffer <1.0 x 10-5 <2.0 x 10-5 <1.0 x 10-5 <2.0 x 10-4 Defective SiC <5.0 x 10-5 <1.0 x 10-4 <5.0 x 10-5 <3.0 x 10-3 HM contamination <1.0 x 10-5 <2.0 x 10-5 <5.0 x 10-5 <1.0 x 10-3 HM fraction outside intact SiC <6.0 x 10-5 <1.2 x -4 <4.0 x 10-5 <8.0 x 10-4 In-Service Fuel Performance Normal operation <5.0 x 10-5 <2.0 x 10-4 <1.0 x 10-4 <4.0 x 10-3 Core heatup accidents [<1.5 x 10-4 ] 42 [<6.0 x 10-4 ] [<3.9 x 10-3 ] [<0.16] 42 Values in [square brackets] are provisional and subject to revision as the design and safety analysis evolve. 6-8

195 Predicted Radionuclide Release From Direct-Cycle HTGR Cores 6.2 Methodology and Assumptions The reference GA design codes and FDDM/F models/correlations (Section 3) were modified as described below for analysis of PC-MHR core with TRISO-coated PuO 1.68 fuel Fuel Performance Models for Plutonium Oxide Fuel Prior to the analysis, fuel performance models were formulated for the TRISO-coated PuO 1.68 fuel particles. In general, the fuel performance models were developed based upon the available data for irradiated PuO 2-x particles, including data from FTE-13 (Ref. 6-4). However, since the data base for the plutonium oxide fuel is limited, numerous assumptions regarding the applicability of the fuel performance models for other fuels had to be made. The assumptions used in developing the model for each fuel failure mechanism and for fission product release from the kernel are discussed in the following subsections Pressure-Vessel Failure Pressure vessel failure is defined as structural failure of the SiC coating. As introduced in Section , the model uses the thick-shell equation and a Weibull distribution to characterize the failure probability of the SiC layer. The models for pressure vessel failure were developed for: (1) standard particles (particles with no defective coatings), (2) particles with failed OPyC but intact SiC coatings, and (3) particles with missing buffer coatings. Using the PC-MHR fuel particle dimensions and the appropriate void volume, the internal gas pressure is calculated as function of burnup and fuel temperature. Then, using this gas pressure in the thickshell equation, the SiC stress is calculated. Pressure vessel failure is assumed to occur when the stress in the SiC coating exceeds the fracture strength OPyC Irradiation-Induced Failure Irradiation-induced failure of the OPyC coating is caused by the interaction of fast neutrons with the microstructure of the outer pyrocarbon coating (Section ). Fast neutrons induce dimensional changes in pyrocarbon which cause the OPyC layer to shrink in the tangential direction. If there is sufficient shrinkage, the OPyC layer can separate or crack and eventually fail. The OPyC irradiation-induced failure model from FDDM/F was used; in this model, the failure probability is correlated as a function that increases linearly with fast neutron fluence to a maximum value and then remains constant. The maximum value of the failure probability is 0.03, and it is assumed to occur at a fast fluence of 2 x n/m 2 (E >0.18 MeV) 6-9

196 Predicted Radionuclide Release From Direct-Cycle HTGR Cores SiC Corrosion Failure SiC coating failure from SiC-fission product corrosion is caused by chemical reaction between certain fission products and SiC (Section ). This reaction reduces the thickness of the SiC layer in fuel particles and can lead to release of fission products from particles. Fissioning of plutonium produces fission products such as palladium, lanthanides, and rare earths which can attack and corrode the SiC layer during the reactor operation. Since SiC corrosion has been shown to be a highly temperature-dependent process (e.g., Ref. 6-6), various rate data from outof-pile experiments and in-pile Peach Bottom fuel tests for UC 2 and (Th,U)C 2 TRISO particles were placed in the simple form below representing the processes of generation of reactants, their transport to the SiC, and their reaction with the SiC (Ref. 6-7): where: dx/dt = x 10 5 exp(-3.03 x 10 4 /T), Equation 6-1 dx/dt = rate of change in SiC thickness (µm/h) due to SiC corrosion, T = fuel temperature (K). As significantly higher palladium production results from plutonium fissions than from uranium fissions, higher Pd accumulation was observed in PuO 1.68 fuel at the SiC/IPyC interface (Ref. 6-4) than in UC 2 fuel. In order to account for higher Pd inventory in Pu fuel than in UC 2 fuel, the above equation was conservatively adjusted by multiplying the pre-exponential factor by the ratio of cumulative palladium fission yields (24.5x). Failure of the SiC coating is assumed to occur when the SiC thickness is reduced by 50% or more. The distribution of SiC thickness is described by a normal distribution with a mean of 35 µm and a standard deviation of 5.15 µm SiC Thermal Decomposition Failure SiC coatings can fail by way of the thermal decomposition of SiC (Section ). If temperatures are sufficiently high, typically greater than 1600 o C for sufficient numbers of hours, SiC will thermally decompose into its constituent elements silicon and carbon. If sufficient decomposition occurs, the structural integrity of the SiC coating can be compromised to fail this coating and allow release of fission products from particles. For this analysis, the model from FDDM/F was used; this model assumes a Weibull distribution for the SiC decomposition rate. The failure probability is correlated as a function of temperature, prior irradiation temperature, and fast neutron fluence. (In fact, the fuel temperatures during normal operation are not sufficiently high for this mechanism to be significant; it is included primarily for completeness.) 6-10

197 Predicted Radionuclide Release From Direct-Cycle HTGR Cores Kernel Migration Failure Kernel migration is the movement of a fuel kernel toward the hot side of the coated particle, and it occurs at sufficiently high temperatures and in presence of a large temperature gradient across the particle (Section ). The fraction of particles that have failed by kernel migration at any given time is determined by the extent of the migration of the kernel and the distribution function for the buffer plus IPyC coating thickness. The distribution for the combined thickness of the buffer and IPyC layers is assumed to be normal. Kernel migration in coated-fuel particles can be detected by post-irradiation examination. The FTE-13 PIE indicated that no kernel migration occurred for O/Pu = However, excessive kernel migration was observed for O/Pu = The migration of PuO x could be influenced by gas-phase transport involving CO, produced by reaction of excess oxygen from fissioning PuO x with carbon from the buffer layer. Even though kernel migration should be precluded in the PuO x particles by controlling the stoichiometry, the kernel migration failure was conservatively calculated in this analysis. Kernel migration data for plutonium oxide fuel given in Ref. 6-4 have not been correlated. Since these data are lower than the data for UO 2, the most recent Japanese, UO 2 kernel migration coefficient (Ref. 6-8) was conservatively used in conjunction with Equation (2-4): Heavy-Metal Dispersion Failure KMC(T) = 2 x 10-6 exp(-14,800/t). Equation 6-2 Heavy-metal dispersion results from transport of heavy metal from the fuel kernel into buffer layer during SiC deposition; the dispersion is thought to be caused by intrusion of Cl (through a porous or defective IPyC) into the kernel with the formation of volatile HM chlorides (Section ). Particles with heavy metal dispersed in the buffer are observed to exhibit enhanced SiC attack by fission products and SiC coating failure. In this analysis, the correlation from FDDM/F was used. The failure probability is correlated as a function that increases linearly with burnup, with the maximum failure probability assumed to be 0.5 at full design burnup Fission Gas Release The fission gas release model from FDDM/F was used; this model accounts for fission gas release from (1) failed fuel (exposed kernels) as a function of fuel temperature, burnup, hydrolysis condition, and the decay constant, diffusion parameter and release characteristics of the particular fission gas; and (2) release from heavy-metal contamination in the fuel compact matrix (Section ). Since the model is for the LEU UCO fuel with a maximum burnup of 26% FIMA, a burnup multiplier was used to account for the burnup dependence of R/B at higher burnups. The burnup multiplier, developed in Ref. 6-9 for a HEU-UCO fission gas release model, is also a function of temperature. As shown in Figure 6-5, the multiplier is essentially independent of burnup and temperature at low burnups (FIMA < 0.2); at higher burnups (FIMA > 0.2), the multiplier increases rapidly with burnup and, for a given burnup, decreases as the temperature increases. 6-11

198 Predicted Radionuclide Release From Direct-Cycle HTGR Cores Figure 6-5 R/B Multiplier for High-Burnup Fuels Fission Metal Release The kernel release model for fission metals from FDDM/F was used; this model includes kernel diffusion coefficients, which are given as a function of temperature and burnup (Section ). For diffusion in fuel kernels, the diffusivity is given by Eqns. (2-15) and (2-16). Diffusion in the buffer is assumed to be so rapid for metallic fission products that all quantities released from the kernel are instantly homogeneously distributed throughout the buffer. Diffusion coefficients for pyrocarbon and SiC coatings and for H-451 fuel element graphite are given as a function of temperature. Sorption equations for fuel compact matrix and graphite are given as a function of temperature and fast fluence. FDDM/F correlations were used for these properties. 6-12

199 6.3 Core Performance Analysis Predicted Radionuclide Release From Direct-Cycle HTGR Cores The fuel performance models for plutonium fuel described in the preceding section were incorporated into the fuel performance code SURVEY (Ref. 6-10) and the metallic fission product release code TRAFIC-FD (Ref. 6-11). The analysis was performed with these codes using the results of the three dimensional depletion of the PC MHR core with the DIF3D code for six fuel cycles. The analysis was performed at the bottom points of all 10 axial layers of fuel elements, with seven radial calculation points per fuel element. Taking advantage of the 120 degree symmetry of the core, only one-third of the core was analyzed at each of the 10 axial locations. Using the calculated fuel temperature, burnup and fast fluence histories, the fuel particle failures and fission product releases were calculated as a function of time. The gaseous fission product releases were calculated for the two key isotopes Kr-88 and I-131. The releases for other isotopes can be obtained by assuming that the R/Bs vary as the square root of isotope half-life. Moreover, it is assumed that bromine and selenium isotopes have the same release characteristics as krypton and that iodine and tellurium isotopes have the same release characteristics as xenon. The analysis of fission metal release considered diffusion in the fuel kernel and coatings, diffusion in the fuel element graphite and sorption at the gap between the fuel compact and fuel hole, and at the coolant hole surface. The results of the fuel performance analysis are summarized in Table 6-4 and discussed in the following subsections. Included are fuel and graphite temperature predictions, fast fluence and burnup fractions, and fuel particle failure and fission product release predictions. A comparison is made between the predicted fission product releases and the radionuclide design criteria (Section 2.1 of Ref. 6-1). 6-13

200 Predicted Radionuclide Release From Direct-Cycle HTGR Cores Table 6-4 PC-MHR Performance Analysis Results Parameter Value Nuclear Maximum Burnup (FIMA) 0.87 Maximum Fast Fluence (n/m 2 ) 4.2 x Temperatures Peak Fuel Temperature ( o C) 1350 Max. Time-Averaged Fuel Temperature ( o C) 1240 Peak Graphite Temperature ( o C) 1268 Max. Time-Averaged Graphite Temperature ( o C) 1155 Fuel Failure Maximum Expected Criterion 1.0 x 10-4 Maximum Predicted Failure (50% Confident) 4.8 x 10-5 Fission Gas Release Kr-88 Maximum Expected Criterion 1.5 x 10-6 Predicted R/B 1.6 x 10-6 I-131 Maximum Expected Criterion 3.9 x 10-6 Predicted R/B 3.9 x 10-6 Fission Metal Release Cs-137 Maximum Expected Criterion 2.0 x 10-5 Predicted Release 1.8 x 10-5 Ag-110m Maximum Expected Criterion 1.5 x 10-3 Predicted Release Fraction 4.1 x

201 6.3.1 Fuel and Graphite Temperature Predictions Predicted Radionuclide Release From Direct-Cycle HTGR Cores Volume distributions of peak fuel centerline temperatures were calculated for each of the three refueling segments considered in the analysis; illustrative results for Segment 1 are given in Figure 6-6. A peak fuel temperature of 1350 o C was predicted to occur in Segment 3 at the corereflector interface of an inner ring fuel element at the bottom of the core during the first year of operation of the initial core. This temperature and other high temperatures are typically maintained for relatively short periods of time because they occur in buffered sections with fresh fuel which burns out quickly. The peak temperature is maintained for 28 days and then drops to 1213 o C, as can be seen from the temperature history for this point, which is shown in Figure 6-7. Figure 6-6 Peak Fuel Temperature Volume Distribution (Segment 1) 6-15

202 Predicted Radionuclide Release From Direct-Cycle HTGR Cores Figure 6-7 Fuel Temperature History for Peak Temperature Location Detailed examination of the temperature predictions for the full core indicated that there are some locations in the PC-MHR core where relatively high fuel and graphite temperatures are maintained throughout an entire cycle (260 EFPD). The temperature history for one such location is shown in Figure 6-8. These locations are typically in the unbuffered middle ring of fuel elements, and they are quite different in nature from the location of the peak temperature, which occurs in the buffered section of a fuel element at the reflector interface. Since fuel failure and fission product release are time-at-temperature phenomena, some of these locations with cycle-average fuel temperatures of 1283 o C could be more limiting with respect to fuel performance and release than the location of the peak fuel temperature of 1350 o C which is maintained for 28 days. 6-16

203 Predicted Radionuclide Release From Direct-Cycle HTGR Cores Figure 6-8 Location with Persistently High Fuel Temperature Volume distributions of time-averaged fuel temperatures were also calculated for each of the three refueling segments considered in the analysis; illustrative results for Segment 1 are given in Figure 6-9. The maximum fuel temperature on time average basis was predicted to be 1240 o C. Consequently, the core thermal design satisfies the MHR design guideline that the time-average fuel temperature should be 1250 o C to preclude significant temperature-induced fuel failure. Only a small fraction of the fuel volume is predicted to have relatively high temperatures, as can be seen in the figure. 6-17

204 Predicted Radionuclide Release From Direct-Cycle HTGR Cores Figure 6-9 Time-Average Fuel Temperature Volume Distribution (Segment 1) 6-18

205 Predicted Radionuclide Release From Direct-Cycle HTGR Cores Volume distributions of the peak graphite temperatures for Segment 1 are shown in Figure 6-10 and on a time averaged basis in Figure The peak predicted graphite temperature was 1268 o C, and the maximum time-averaged graphite temperature was 1135 o C. A summary of the predicted fuel and graphite temperatures is given in Table 6-4. Figure 6-10 Peak Graphite Temperature Volume Distribution (Segment 1) 6-19

206 Predicted Radionuclide Release From Direct-Cycle HTGR Cores Figure 6-11 Time-Average Graphite Temperature Volume Distribution (Segment 1) 6-20

207 Predicted Radionuclide Release From Direct-Cycle HTGR Cores Fast Fluence and Burnup Distributions Volume distributions of particle burnup and fast fluence were calculated for each of the three refueling segments considered in the analysis; illustrative results for Segment 1 are given in Figure 6-12 and 6-13, respectively. The maximum fissile burnup fraction was 0.87 FIMA, which was predicted to occur in Segment 2 at the end of Cycle 5, and the peak fast neutron fluence of 4.2 x n/m 2 was predicted to occur in Segment 3 at the end of Cycle 6. Figure 6-12 Burnup Volume Distribution (Segment 1) 6-21

208 Predicted Radionuclide Release From Direct-Cycle HTGR Cores Figure 6-13 Fast Fluence Volume Distribution (Segment 1) Fuel Particle Failure The fuel performance models account for the effect of partially failed fuel particles on fission gas and fission metal release; the particles with failed SiC but intact OPyC coatings retain fission gases but not metals. The maximum predicted, core averaged fuel particle failures from all 6-22

209 Predicted Radionuclide Release From Direct-Cycle HTGR Cores mechanisms are given in Table 6-4 which shows the failure fractions used to predict both the fission gas and fission metal releases. Time histories of particle failure fractions are shown in Figure 6-14 and 6-15 for the fission gas (exposed kernel fraction) and fission metal (SiC failure fraction) releases, respectively. At each reload, the failure fraction in the core decreases due to the replacement of oldest one third of the fuel in the core with fresh fuel. The maximum particle coating failure fraction from all mechanisms for fission gas release (complete coating failure resulting in an exposed kernel) was predicted to be 4.8 x This predicted failure fraction satisfies the "Maximum Expected" criterion of 1.0 x 10 4 (Section 2.1 of Ref. 6-1). Figure 6-14 Exposed Kernel Fraction vs. Time 6-23