PROPOSAL OF A GUIDE TO PERFORMANCE ASSESSMENT OF FUEL RODS FOR NUCLEAR POWER PLANTS

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1 2013 International Nuclear Atlantic Conference - INAC 2013 Recife, PE, Brazil, November 24-29, 2013 ASSOCIAÇÃO BRASILEIRA DE ENERGIA NUCLEAR - ABEN ISBN: PROPOSAL OF A GUIDE TO PERFORMANCE ASSESSMENT OF FUEL RODS FOR NUCLEAR POWER PLANTS Nilo G. Silva, Daniel A. P Palma, Auro C. Pontedeiro Comissão Nacional de Energia Nuclear CNEN/Sede Av. General Severiano, Rio de Janeiro Brazil dapalma@cnen.gov.br ABSTRACT The purpose of this paper is to present a proposal for a procedure to be adopted by the Brazilian Nuclear Energy Commission (CNEN) to evaluate the safety of fuel rods being used in nuclear power reactors in operation in the country. It should also guide the licensing process of new fuel rods designs clearly delimiting safety criteria related to its thermo-mechanical behavior. This activity is under technical collaboration of multidisciplinary design INSC BR3.01/09-BR/RA/01 Project signed between Brazil and the European Union (EU). This paper presents a first step towards establishing a CNEN standard on specific safety requirements to be met by designs of fuel elements of NPP reactors (PWR) that are operating in Brazil. 1. INTRODUCTION Currently the nuclear sector in Brazil is regulated by CNEN following different standards. These standards have criteria related to safety limits of fuel rods used in nuclear power reactors in operation in the country in a very dispersed way. These standards are of the eighties, a time where the claddings were different from the current ones and the computer codes were inferior than those used nowadays. The operational experience worldwide has increased considerably in recent years and there are technical documents where the accepted safety limits used by regulators in different countries are compiled [1]. This document will present the Brazilian criteria, according to Brazilian standards regarding the safety of nuclear fuel during operation in stationary and transient regimes. This study does not apply to storage and transportation of fuel, which has its own rules and it is beyond the scope of the task 6 technical collaboration project INSC BR3.01/09-BR/RA/01. It also does not apply to situations of accidents beyond the design basis. 2. BRAZILIAN STANDARDS RELATED TO THE PERFORMANCE OF NUCLEAR FUEL The set of standards published by CNEN is divided into eight groups, and all regulations related to the performance of nuclear fuel are members of the group 1 - Nuclear Facilities. This section presents the existing criteria in each one of them Emergency Core Cooling System Acceptance Criteria for Light Water Reactors (NE ) [2]

2 The objective of this standard is to establish acceptance criteria for emergency core cooling systems (ECCS) of light-water reactors (LWR) of nuclear power plants. This standard applies only to power reactors, cooled by light-water and loaded with fuel rods with pellets of uranium oxide (UO 2 ) contained in cylindrical claddings from zircaloy. The NPPs Angra1 and and Angra 2 are pressurized water reactors (PWR) and they are still loaded with zircaloy cladding rods but also with Zyrlo and M5 alloys. This trend of gradual replacement of zircaloy by most modern cladding alloys suggests that the standard CNEN-NE 1.20 requires updates. In section 4.2 of this standard are the following criteria on the matter in question: Maximum Cladding Temperature: The maximum temperature of the zircaloy cladding calculated during a loss of coolant accident (LOCA) should not exceed 1200 C. Maximum Cladding Oxidation: b) The total oxidation of the cladding during LOCA calculated with correlation of Baker&Just [3] must not exceed at any position in the core 17% ECR (ECR: equivalent cladding reacted). In case of fuel rod burst, the inner surface of the fuel rod cladding has to be included in the calculation of oxidation. Maximum Hydrogen Generation: The total calculated amount of hydrogen generated by the chemical reaction of zircaloy cladding with water or steam should not exceed 1% of the amount that would be generated if all of the metal surrounding the cylinder liner fuel pellets, excluding surrounding coating of full, reacted stoichiometrically. 2.2 Quality Assurance in Acquisition, Design and Manufacturing of Fuel Elements (NE ) [4] The standard NE 1:27 is not directly related to the performance and safety criteria of fuel elements, but comes to the quality of its manufacturing. Quality assurance ensures mastery and documentation of all stages of manufacturing the element. This includes the uncertainties associated with each parameter of interest to the present study, whether geometric or chemical composition. The item 1 of Annex I of the standard NE 1.27 states that the following items related to the manufacture of nuclear fuel must have a guaranteed quality: Enrichment; Amounts of impurities; Geometrical dimensions. The item 3 of Annex I of the standard NE 1:27 states that the following items related to the nuclear fuel cladding manufacturing process must have a guaranteed quality. These cladding parameters are: Chemical composition; Impurities; Resistance to corrosion; Mechanical properties;

3 Integrity; Dimensions: (e.g. cladding diameter, cladding thickness) All parameters listed above have to comply with defined design tolerances in order to guarantee the desired fuel rod performance for normal operation. In item 5 of Annex I of the standard CNEN-NE-1.27 [4] is highlighted the quality of predicting the fuel rod thermodynamic parameters for both all operational states and postulated accidents. These thermodynamic parameters are: Heat transfer; Fuel temperature; Fission gas release; Internal pressure. As the criteria related to fuel performance are located in different standards would be desirable to establish a unique document containing the safety criteria which would address the fuel in use. 3. NUCLEAR FUEL SAFETY CRITERIA IN GERMANY AND FRANCE There are three types of criteria relating to the safety of nuclear installations [1]: Safety criteria - Criteria imposed by the regulator. If preserved, safety criteria ensure that the impact of the DBA (Design Based Accident) on the environment is acceptable. Design criteria - Design criteria ensure that fuel rod failure will not occur systematically as a consequence of an insufficient fuel rod design. Operational criteria - Operational criteria, specific to the particular fuel element under consideration, are partly provided by the vendor as part of the overall licensing basis. Operational criteria are preserved during the regular operation and anticipated transients. Unfortunately, numerous studies related to fuel damage criteria are used in fuel design and safety analyses and these fuel criteria may differs from country to country, as can visualized in the reference [1]. In this paper, only the safety criterias will be considered. In subsequent tables (Tab. 1 and Tab. 2) fuel rod safety criteria from the OECD report on nuclear fuel safety criteria [1] have been summarized: Table 1: Selected safety criteria for fuel rods under normal operation. Parameters Criteria Cladding Oxide Thickness <100 µm Hydride concentration in the cladding Fuel Rod Internal Pressure <500 wppm < Primary coolant pressure

4 Cladding Hoop Strain for fast transient < 1% Cladding Equivalent Strain for long term transient < 3.5% Centerline Temperature < Melting point Cladding Circumferential Stress Stay below Stress of collapse Stay below Euler buckling Cladding Axial Stress force for bending Cyclic Bending Forces on the Cladding Stay below 50 MPa Power load ramp rate and Pellet Cladding Interaction (PCI) total power step are limited to prevent PCI Table 2: Selected safety criteria for fuel rods in design basis accident. Parameters Criteria Peak cladding temperature during LOCA 1200 C Maximum local equivalent cladding reacted in 17% core during LOCA To limit the injected energy Reactivity Initiate Accident (RIA) in the fuel rod depending on fuel rod corrosion status Based on the EU experience the Brazilian regulatory body intends to standardize the assessment of fuel performance safety in NPPs in the country. In the next section a set of new requirements will be reported. 4. NEW REQUIREMENTS IN THE CORE RELOAD REPORT This section will present a proposal for a guideline for submission of report related to the specific performance of nuclear fuel rods from the operator to the Brazilian regulatory body (CNEN). This report, which would be an addendum to the reload report named Nuclear and Thermo-hydraulic Project Report (NTPR), has the role of enhancing the safety of NPPs Angra I and Angra II and allows CNEN s independent calculations to check that no safety criterion are violated. Safety limits must be observed both in operation at full power as under postulated accident conditions in the project. CNEN uses the codes FRAPCON/FRAPTRAN [5] to conduct their independent calculations. The operator shall provide a table in NTPR with all geometry and composition parameters for each type of fuel rod used in the core loading with their respective manufacturing uncertainties as described in section 2.2 of this paper.

5 Safety parameters should contain but not be limited to the Table 3 below: Table 3: Safety parameters. 1. Average Power 20. Oxide Thickness 2. Fuel Stack Axial Extension 21. Surface Heat Flux 3.Cladding Axial Extension 22. Coolant Density 4. Plenum Pressure 23. Coolant Mass Flux 5. Total Void Volume 24. Coolant Pressure 6. Average Fuel Temperature 25. Centerline Temperature 7. Fuelled Region Stored Energy 26. Fuel Pellet Surface Temperature 8. Rod Average Burnup 27. Cladding Inside Temperature 9. Fission Gas Release 28. Cladding Average Temperature 10. Cladding Axial Strain 29. Cladding Outside Temperature 11. Cladding Hoop Strain 30. Bulk Coolant Temperature 12. Cladding Radial Strain 31. Fuel Surface Displacement 13. Cladding Perm Axial Strain 32. Gap Gas Pressure 14. Cladding Perm Hoop Strain 33. Axial Power 15. Cladding Perm Radial Strain 34. Gap Interface Pressure 16. Cladding Axial Stress 35. Fuel Surface Axial Strain 17. Cladding Hoop Stress 36. Cladding Elastic Hoop Strain 18. Effective Cladding Stress 37. Cladding Elastic Axial Strain 19. Structural Radial Gap 38. Cladding Elastic Radial Strain The operator must provide what will be the power history for each type of fuel rod in the case of mixed core, as complete as possible and with an appropriate and justified safety factor. In exceptional cases, a history of average power can be used in the simulations. The rods with higher burnup must be taken into consideration never violating the maximum burnup reported in FSAR [6]. In the next section an example of independent calculation will be performed. 5. AN EXAMPLE OF SIMULATION USING THE FRAPCON CODE This section will present the results obtained with the FRAPCON code to the fuel rod exposed to the most severe historical power calculated with the CNFR code [7]. For simplicity, it was chosen to simulate a cycle of 363 days of operation at full rated thermal input of 1876 MW fully charged with fresh fuel elements (8 0 cycle of Angra I). There are three distinct regions of enrichment of the isotope 235 U. All burnable poison elements simulated are also fresh. The code is able to indicate the different regions of enrichment, location of control rods banks and burnable poison. An example of displaying a map of the core is shown in Figure 1, which was generated from the CNFR code that will be used by the operator. The fuel rod analyzed is in the G12 element and the maximum linear heat generation rate (q max ) is considered q max = x W/m = W/m. For comparison, we also consider the rod located at the position J12, which presents the minimum linear heat generation rate (q min ) of q min = x W/m = W/m. The power density distribution can be visualized in Figure 2 and is standard information provides from operator in each NTPR submited to CNEN.

6 Figure 1. Visualization of the reactor core in the CNFR code. Figure 2. Power density distribution in the core using CNFR code.

7 Figure 3 has a summary of some important safety parameters that are provided directly by the output file FRAPCON code. Figure 3: Regulatory Summary of important safety parameters. From the figure 3 is possible to conclude that fuel safety criteria were not violated in this cycle. There is a difference of 10.0% between the higher and lower maximum fuel rod internal pressure. A difference of 64.8% between the higher and lower maximum strain increment was visualized which shows a higher sensitivity of this parameter. 6. CONCLUSIONS This paper proposes initial ideas to build a guide for submission of reports related to the performance of nuclear fuels used in NPP in Brazil. With the recent acquisition of the new version of the code FRAPCON/FRAPTRAN is possible to perform independent calculations, which is desirable to improve the licensing of nuclear installations in operation in the country. ACKNOWLEDGMENTS We acknowledge the valuable support from technical collaboration project INSC BR3.01/09- BR/RA/01 in the course of the development of this work. REFERENCES 1. Nuclear Energy Agency - OECD, Nuclear Fuel Safety Criteria Technical Review, NEA No. 7072, (2012). 2. Comissão Nacional de Energia Nuclear - CNEN, Emergency Core Cooling System Acceptance Criteria for Light Water Reactors, CNEN-NE-1.20, (1985). 3. Baker, L., Just, L. C., Studies of metal-water reactions at high temperatures III. experimental and theoretical studies of the zirconium-water reaction. ANL 6548, (1962). 4. Comissão Nacional de Energia Nuclear CNEN, Quality Assurance in Acquisition, Design and Manufacturing of Fuel Elements, CNEN-NE-1.27, (1999).

8 5. Geelhood, K.J., Luscher, W.G., Beyer, C.E., FRAPCON-3.4: A computer code for the calculation of steady-state, thermal-mechanical behavior of oxide fuel rods for highburnup, NUREG/CR-7022, Vol. 1., (2011). 6. Comissão Nacional de Energia Nuclear CNEN, Final Safety Analysis Report - Central Nuclear Almirante Álvaro Alberto, Unit 1, Rev. 36, September (2011). 7. Silva, A. C.; Silva, F.C., Martinez, A. S., Calculation of the power factor using the neutron diffusion hybrid equation, Annals of Nuclear Energy, v. 51, pp (2013).