2017 Water Reactor Fuel Performance Meeting September 10 (Sun) ~ 14 (Thu), 2017 Ramada Plaza Jeju Jeju Island, Korea

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1 NUCLEAR DESIGN AND SAFETY ANALYSIS OF ACCIDENT TOLERANT FUEL CANDIDATES IN OPR1000 Wang-Kee In 1, Ser-Gi Hong 2, Tae-Wan Kim 3, Tae-Hyun Chun 1, Chang-Hwan Shin 1 1 Korea Atomic Energy Research Institute: Daedeok-daero Yuseong, Daejeon, 34039, and wkin@kaeri.re.kr 2 Kyunghee University: 1732 Deokyoung-daero, Yongin, Geyongki-Do, 17104, and sergihong@khu.ac,kr 3 Incheon University: 119 Academiro, Yeonsu, Incheon, 22012, and taewan.kim@inu.ac.kr ABSTRACT: The Korea Atomic Energy Research Institute (KAERI) has developed the accident tolerant fuel (ATF) candidates which are micro-cell fuel pellet and coated cladding. The micro-cell U pellet with metal additions will decrease the fuel temperature by increasing thermal conductivity and also reduce the release of fission product in accident condition. The KAERI has also developed a Cr-alloy coated cladding to significantly reduce the hydrogen generation by lowering the oxidation rate in high temperature particularly in beyond design basis accident (BDBA). This paper presents the nuclear design and safety analysis of the KAERI ATF candidates against the reference fuel for optimum power reactor, OPR1000 in Korea. A neutronic analysis was performed to compare the nuclear physics of OPR1000 core with the ATF candidates (metallic micro-cell U-Cr or U-Mo pellets and Cr-alloy coated cladding) and the reference fuel (U pellet and Zircaloy cladding). The neutronic calculation shows that the cycle length is decreased by 11 days for the Cr-alloy coated cladding and by more than 60 days for the metallic micro-cell pellets because of reduced heavy metal inventories and large thermal absorption cross sections of the Mo and Cr isotopes. The safety analysis was also performed to evaluate the ATF enhancement of reactor safety in the events of large break LOCA (LBLOCA), small break LOCA (SBLOCA) with partial safety injection (SI), and station black out (SBO). The system transient thermal-hydraulic codes, RELAP5/Mod3.3 and SCDAP/RELAP5 were used in this study to perform the safety analysis. The peak cladding temperature (PCT) and hydrogen generation were compared for the ATF and reference fuel in BDBA conditions. The ATF with metallic micro-cell pellets appear to decrease the PCT in LBLOCA by 100 K due to high thermal conductivity. The safety analysis for the SBLOCA with partial SI and SBO accident showed the delay of PCT increase and less generation of hydrogen with the ATF candidates due to low metal(clad)-water reaction heat generation. KEYWORDS: Accident tolerant fuel (ATF), Nuclear design, Safety analysis, Cycle length, Peak cladding temperature (PCT), Hydrogen, Metal-water reaction. I. INTRODUCTION Accident tolerant fuel (ATF) concepts have been developed in worldwide to enhance the safety of light water reactors in accident condition as well as to improve the fuel performance in normal operation, especially after the Fukushima nuclear disaster (Refs. 1-3). The advanced fuel campaign (AFC) in USA organized a workshop on accident tolerant cladding at Oak Ridge National Laboratory (ORNL) in 2013 (Refs. 4-5). The purpose of this workshop was to discuss the technology development for advanced cladding and coated cladding. The Korea Atomic Energy Research Institute (KAERI) has developed the accident tolerant fuel (ATF) candidates for the optimum power reactor (OPR1000) in Korea, which are microcell fuel pellet and coated cladding. The new UO 2 pellet with metal or ceramic additions (Ref. 6) will decrease the fuel temperature by increasing thermal conductivity and also reduce the release of fission product in accident conditions. The KAERI has also developed a Cr-alloy coated cladding (Ref. 7) to significantly reduce the hydrogen generation by lowering the oxidation rate at high temperature, particularly during beyond design basis accidents (BDBA). 1

2 Figure 1 shows the comparison of thermal properties of reference fuel (U) and ATF (U+Cr, U+Mo). It is noted that the thermal conductivity of ATF is twice higher than the reference conductivity. This is because a metal (Cr or Mo) is included in the ATF candidates. The reference and ATF shows almost identical heat capacity. Fig. 1. Comparison of thermal conductivity and heat capacity for reference fuel and ATF candidates Figure 2 shows the oxidation characteristics of ATF clad (Cr-alloy coating) in comparison to reference clad (Zry-4). The ATF clad indicates the decrease of oxidation rate (k p ) and weight gain by 1/1000 compared to the reference clad. Hence, the value of proportional coefficient in clad oxidation model is decreased by 1/1000 in the safety analysis for ATF. Fig. 2. Comparison of oxidation rate and weight gain for reference clad (Zry-4) and ATF clad (Cr-alloy). It is important to investigate core physics and safety enhancement for the application of KAERI ATF to OPR1000. This paper presents the nuclear design and safety analysis of the KAERI ATF candidates against the reference fuel for the OPR1000. A neutronic analysis was performed to compare the nuclear physics of OPR1000 core with the ATF candidates (metallic micro-cell U-Cr or U-Mo pellets and Cr-alloy coated cladding) and the reference fuel (U pellet and Zircaloy cladding). The safety analysis was also performed to evaluate the enhancement in reactor safety resulting from ATF adoption in the event of large break LOCA (LBLOCA), small break LOCA (SBLOCA) with partial safety injection (SI), and station black out (SBO) accident. The system transient thermal-hydraulic codes, RELAP5/Mod3.3 (Ref. 8) and 2

3 SCDAP/RELAP5 (Ref. 9) were used in this study to perform the safety analysis. It is noted that the SCDAP/RELAP5 code can be used to assess the reactor safety in the event of BDBA as well as DBA. II. NUCLEAR DESIGN OF ATF CORE FOR OPR1000 The ATF concepts investigated in this work use metallic micro-cell U pellets (containing Cr or Mo) as fuel and Zircaloy coated with Cr-based alloy as cladding, which have been suggested as the ATF concepts in KAERI. The DeCART2D (Deterministic Core Analysis based on Ray Tracing for 2-Dimensional Core) code was used for fuel assembly calculations. This code has been developed in KAERI to generate few group homogenized neutron cross section data for nodal diffusion core analysis code (Ref. 10). Then, the table sets which include functionalized group constants are produced by using the PROLOG program and HGC file prepared with DeCART2D. The calculations for core analysis are performed by using the MASTER (Multi-purpose Analyzer for Static and Transient Effects of Reactors) code, also developed in KAERI. This latter code is a nuclear analysis and design code capable to simulate PWR or BWR cores in 1-, 2-, or 3-dimensional Cartesian or Hexagonal geometry with the advanced nodal diffusion methods (Ref. 11). In this study, we compared four different cases having the same uranium enrichments. CASE 1 is the reference case which uses conventional UO 2 pellets with enrichments of 4.60/4.10 (normal/zoning) wt% and conventional (uncoated) Zircaloy cladding. The selected uranium enrichments were determined to satisfy at cycle length of 480 EFPDs. CASE 2 is similar to the reference case except for a 0.005cm thick, Cr-based alloy coating on the cladding outer surface, and it is selected in order to evaluate the effect of the coating only. CASE 3 and 4 use metallic micro-cell UO 2 pellets containing small amounts of Cr and Mo, respectively, with the weight percentages of each of these additives determined to have same amount of initial heavy metal between these two designs. Both CASE 3 and CASE 4 use Cr-based alloy coated Zircaloy as fuel rod cladding. All the fuel assemblies adopt an enrichment zoning to reduce the pin power peaking, which places low uranium enrichment fuel rods around the instrument thimbles. Burnable absorber (BA) rods are used to reduce initial excess reactivity. The BA rods included the cutbacks on the top and bottom of the rods in order to flatten axial power distribution. Unlike the addition of Cr and Mo to the fuel pellets, which decreases the heavy metal inventory by ~5%, the adoption of Cr coating does not lead to any variation in initial heavy metal because such coating only slightly reduces the thickness of the cladding base metal, without any effect on the pellet diameter. The equilibrium core design used for this study is shown in Figure 3. This design was obtained by modeling fuel burnup following the 7 th cycle of the Hanbit-3 nuclear power plant, which was chosen as the detailed design data from the 1 st to 7 th cycle were available for this plant. The loading pattern of the 7 th cycle was simplified to find equilibrium cycle as soon as possible. The low-leakage loading pattern coupled with the concept of three batch refueling scheme was considered and there were no fresh fuel assemblies in the center of the core in order to mitigate power peaking. Core burnup characteristics were found not to change after Cycle 12, which was therefore chosen as reference equilibrium cycle for this study. Figure 4 compares the critical boron concentrations (CBC) and 3-dimensional peaking factors of the equilibrium cycle (12 th cycle) for all the cases. In CASE 2, which uses conventional UO 2 pellets and Cr-based alloy coating, the cycle length is shortened by 11.4 EFPDs in comparison with the reference case because of the higher capture cross section of Cr-52 (contained in the coating) compared to Zr-90. The cycle lengths of CASE 3 and 4 are further reduced by 60 and 106 EFPDs, respectively, because of the lower amount of initial heavy metal characterizing these cases and, although to a lower extent, because of the high capture cross sections of the Mo and Cr isotopes. As a consequence of the shorter cycle length, the average discharge burnup of the ATF-fueled cores is reduced to 45.0 and 40.1 MWD/kgU, respectively for CASE 3 and 4, relative to the 48.9 MWD/kgU average discharge burnup of the reference case. The maximum local peaking factors for all cases are typical, around 2.5. The axial offset for all cases are within -4.1% to 7.2 %. It is notable that the axial power distribution of CASE 4 using the micro-cell UO 2 -Mo pellet and Cr-based alloy coating moves towards the core bottom at beginning of the cycle (BOC), which results from the strong negative MTC of this core. 3

4 Critical Boron Concentration (ppm) 3-dimensional peaking factor, Fq 2017 Water Reactor Fuel Performance Meeting A B C D E F G H J K L M N P R 180 O0 O0 1 2 O0 O0 3 4 O1 O1 O1 O1 5 O0 O1 O1 O0 6 O0 O1 O0 O1 O0 O1 O O1 O O0 O1 O0 O1 O0 O1 O0 9 O0 O1 O1 O0 10 O1 O1 O1 O O0 O O0 O Fig. 3. Low-leakage loading pattern at equilibrium cycle (12 th cycle) CASE 1 CASE 2 CASE 3 CASE CASE 1 CASE 2 CASE 3 CASE Time (EFPD) Time (EFPD) Fig. 4. Comparison of critical boron concentrations and 3-D peaking factor at equilibrium cycle. Figure 5 shows the moderator temperature coefficients (MTC) and the fuel temperature coefficients (FTC) for the four cases analyzed. It shows that CASE 4 (UO 2 -Mo pellets and Cr-based alloy coating) has the most negative MTC, which is due to the strong resonance absorption cross section of Mo. In particular, the significant more negative MTC of CASE 4 is also due to its lowest CBC. Core-level simulation showed that CASE 3, which uses UO 2 -Cr pellets and Cr-based alloy coating, has more negative MTC than CASE 1 and 2, which results from its lower CBC. On the other hand, assembly-level simulations showed that CASE 3 has less negative MTC than CASE 1 and 2. The assembly-level calculation may be resulted from the fact that Cr isotopes have smaller resonance cross sections than U-238. Figure 5 also compares the fuel temperature coefficient (FTC) for the four cases. It should be noted, however, that this parameter was obtained from assembly-level simulations as fuel temperatures could not be accurately estimated in core-level calculations with the MASTER code. Moreover, no differences in thermal conductivity of the different fuels was accounted for, as the main focus of this study on neutronic differences between the cores analyzed. The micro-cell UO 2 -Mo and UO 2 -Cr pellets resulted to have slightly more negative FTCs. It should be noted that these differences in FTC are partially due to the 4

5 Moderator Temperature Coefficient (pcm/ o C) Fuel Temperature Coefficient (pcm/ o C) 2017 Water Reactor Fuel Performance Meeting burnup and so these differences will be reduced if the FTCs are compared at the same burnup CASE 1 CASE 2 CASE 3 CASE CASE 1 CASE 2 CASE 3 CASE Time (EFPD) Time (EFPD) Fig. 5. Comparison of moderator and fuel temperature coefficients at equilibrium cycle. III. SAFETY ANALYSIS OF ATF-FUELED OPR1000 A safety analysis was performed to assess the effect of the KAERI ATF candidates on the safety performance of the OPR1000 plant. The postulated accidents for this study are loss of coolant accident (LOCA), and station-black-out (SBO) accident. The system transient codes used in this analysis are the RELAP5/MOD3.3(P 4) and SCDAP/RELAP5/MOD3.3. The RELAP5/MOD3.3 code uses a Cathcart-Pawel correlation for the metal(cladding)-water reaction model and the SCDAP code employs a physical oxidation model based on integral diffusion theory. The reactivity coefficients and other nuclear design data, needed for the safety analysis discussed in this section, were generated from the core neutronic analysis as described in section II. CASE 4 (U+Mo pellet) shows a somewhat difference in MTC and slight difference in axial power distribution from the reference core (CASE 1). Overall the differences in core neutronic characteristics between the reference fuel and ATF are found not to affect reactor performance during the transients considered in this work. The physical properties required for the safety analysis are thermal conductivity and heat capacity of fuel. Figure 6 illustrates the system nodalization of OPR1000 for use in this safety analysis. The fuel rod consists of hot pin and average pin in cylindrical heat structure. The number of mesh in the core (fuel) is 12 and 8 in axial and radial directions, respectively. The safety analysis for large-break LOCA (LBLOCA) was performed using the RELAP5/MOD3.3 code. This analysis assumed the double-ended guillotine break (DEGB) in cold leg and a loss of one LPSI pump. At 8 seconds after initiation of LBLOCA, the safety injection signal and turbine trip occur. The safety injection tank (SIT) starts the operation at 13 seconds and the LPSI pump injects the coolant at 38 seconds. After approximately 72 seconds, the SIT becomes empty. 5

6 Fig. 6. System nodalization of OPR1000 for safety analysis. The LBLOCA analysis for reference and ATF cores shows almost same predictions of primary pressure, core water level, break flow rate and SI flow rate. Figure 7 compares the peak cladding temperature (PCT) for reference fuel and ATF candidates. The ATF cases C1, D1 and G1 correspond to the CASE 2, 3 and 4 in section II. It shows the decrease of PCT for ATF (Case D1, G1) by 100 K because the thermal conductivity of ATF pellets is higher than the reference case. After 50 seconds of LBLOCA, the reflood PCT appears to show somewhat difference between reference fuel and ATF. This is because the oxidation heat generated by clad-water reaction will decrease for the ATF case. Fig. 7. Peak cladding temperature in LBLOCA for reference fuel and ATF candidates. 6

7 H2 generation at top of fuel(kg/s) Peak cladding temperature, PCT(K) 2017 Water Reactor Fuel Performance Meeting A special LOCA scenario was developed to consider the decrease of SI flow rate due to debris in SI pipe, which is defined as small-break LOCA with partial SI flow, SBLOCA+. The SBLOCA+ assumes 12% of SI flow rate with the 1.5 inch break of cold leg. The RELAP5/MOD3.3 code was used to perform safety analysis in the event of SBLOCA+. The pressurizer pressure rapidly decreases to 10 MPa and continually decreases to 6 MPa after 4000 seconds. The core water level decreases to 3 m after 700 seconds and the top portion of fuel is exposed to steam environment. There is no significant difference in pressure and core level between reference fuel and ATF. Figure 8 compares the hydrogen generation and PCT for reference fuel and ATF D1 (CASE 3). The reference fuel A1 (CASE 1) shows rapid increase of hydrogen generation due to high clad oxidation as the fuel is exposed to steam. However, the ATF case D1 shows almost negligible level of hydrogen generation because the Cr-alloy coated cladding resulted in very low rate of oxidation in high temperature. The PCT for reference fuel starts to increase after 4500 seconds and increases to cladding melting temperature (2100 K) at 8000 seconds due to high oxidation heat as well as decay heat. The ATF D1 shows the increase of PCT up to 1800 K lower than the cladding melting temperature. This is because the Cr-alloy coated cladding for ATF D1 reduces the oxidation heat by 1/1000 in comparison with the reference Zircaloy cladding. The SCDAP/RELAP5 code was also used to simulate the SBLOCA+ for reference fuel and ATF. It should be noted that the cladding-water(steam) reaction is neglected in this simulation for the ATF case. The system thermal-hydraulic parameters such as pressurizer pressure, core water level, break flow rate and SI flow rate are almost same for reference fuel and ATF. Figure 9 compares the decay heat and clad oxidation heat for reference fuel. The oxidation heat rapidly increases to 53 MW near 9000 seconds which is larger than the decay heat of 33 MW. This is because the Zircaloy clad oxidation increases significantly as the cladding temperature increases above approximately 1200 K. The hydrogen generation rate and PCT in SBLOCA+ are compared in Fig. 10. The hydrogen generation rate for reference fuel starts to increase at 8000 seconds and shows a sudden jump near 9000 seconds. However, the ATF (case D1) did not show any hydrogen generation because the ATF clad oxidation is neglected in this simulation. The PCT for reference fuel begins to increase at 6000 seconds and reaches at 1300 K at 8500 seconds. The reference PCT increases exponentially and reaches the clad melting temperature at approximately 9000 seconds. This is because the oxidation heat by the reference cladding (Zircaloy) is larger than the decay heat near 9000 seconds. The ATF PCT also increases to 1300 K at 8500 seconds and shows a peak value (approx K) which is lower than the cladding melting temperature. The SCDAP/RELAP5 results indicate that the ATF may not melt in the event of SBLOCA with partial SI. It should be noted that the reference fuel(cladding) will melt in this postulated accident Ref. A1 ATF D Clad melting Ref. A1 ATF D Time(sec) Time(sec) Fig. 8. Hydrogen generation and PCT in SBLOCA with partial SI for reference fuel and ATF D1 (RELAP5/MOD3.3). 7

8 Fig. 9. Decay heat and clad oxidation heat in SBLOCA with partial SI for reference fuel (SCDAP/RELAP5). Fig. 10. Hydrogen generation and PCT in SBLOCA with partial SI for reference fuel and ATF D1 (SCDAP/RELAP5). The station-black-out (SBO) accident was also simulated in this safety analysis since it represents one of beyond design basis accidents (BDBA). The total loss of feedwater (TLOFW) was assumed in this SBO scenario by neglecting operation of auxiliary feedwater pump. The SBO accident was analyzed using the RELAP5/MOD3.3 and SCDAP/RELAP5 codes. The system transient codes predicted the oscillation of pressurizer pressure after 3000 seconds owing to opening or closure of safety valve. Figure 11 shows the core water level and hydrogen generation rate for reference fuel and ATF. The core water level drops by 0.5 m at 5000 seconds by opening safety valve and water evaporation, and then rapidly decreases at 6500 seconds. After 7200 seconds, the water level in active core is lower than 1 m. The reference fuel and ATF shows almost same core water level. The hydrogen generation for reference fuel increases to 10 kg/s within 500 seconds from 6700 seconds. However, the ATF case shows the hydrogen generation rate lower than 1 kg/s even after 1000 seconds from 6700 seconds. The PCT in Fig. 12 rapidly increases starting at approximately 6500 second, i.e. when the core water level decreases to about 2 m. The PCT without the oxidation heat appears to increase more slowly than for the reference fuel, reaching melting with a delay of about 11 minutes. However, the previous SBO analysis for GE BWR by Oak Ridge National Lab. (Ref.12) presented the time delay of more than 2 hour to clad melting for the reference fuel without clad oxidation and the FeCrAl clad. The ORNL analysis using the MELCOR code showed that rapid cladding oxidation with the reference fuel occurs after the core is mostly uncovered. For more accurate estimation of time delay to clad melting for the ATF, it is necessary to include the effect of cladding oxidation at high temperature and fuel deformation in the future safety analysis. 8

9 Peak cladding temperature, PCT(K) 2017 Water Reactor Fuel Performance Meeting Fig. 11. Core water level and hydrogen generation rate in SBO accident for reference fuel and ATF D1 (RELAP5/MOD3.3) SBO(TLOFW) Ref. A1 No clad oxidation heat Time (sec) Fig. 12. Effect of Oxiadtion Heat on PCT in SBO accident for reference fuel (RELAP5/MOD3.3). IV. CONCLUSIONS An analysis was performed to assess the effects, on neutronic and safety performance, of replacing conventional fuel/cladding materials with the ATF concepts considered by KAERI, i.e. metallic micro-cell UO 2 -Cr and, UO 2 -Mo pellets and Cr-based alloy coated cladding. The OPR1000 plant was used as reference. In this study, the same initial uranium enrichment was used for all core designs. The neutronic calculation shows that adoption of the Cr-based coating would result in shortening the cycle length by 11 days. More significant reduction cycle length, of about 60 and 108 days, would instead result from the adoption of the metallic micro-cell pellets containing Cr and Mo, respectively, in comparison with the reference UO 2 fueled core. The analysis also showed that other neutronic parameters, such as reactivity coefficients, do not differ significantly between the ATF and the reference core. There is an on-going effort to increase the cycle length of ATF core by optimizing the ATF pellet design. The safety analysis considered large break loss-of-coolant accident (LBLOCA), small break LOCA with partial SI (SBLOCA+) and a station-black-out (SBO) accident. Neutronic performance parameters, e.g., reactivity coefficients, and thermo-physical properties specific to the ATF materials were used in this analysis, together with a modified cladding-water reaction model to capture the lower oxidation characterizing Cr-coated cladding. The LBLOCA analysis showed that the ATF with metallic micro-cell pellets decreases the PCT by about 100 K. In the event of SBLOCA+, the ATF PCT reaches a 9

10 peak value of approximately 1500 K, which is lower than the cladding melting temperature. It should be noted that, in the same scenario, the reference fuel(cladding) core is anticipated to reach cladding melting. For the SBO accident, the ATF decreases hydrogen generation by a factor of 10. Future improvements in the analysis method will need to include the effect of cladding oxidation at high temperature and fuel deformation in the future safety analysis. ACKNOWLEDGMENTS This work has been carried out under the Nuclear R&D Program supported by the Ministry of Science, ICT&Future Planning. (NRF-2017M2A8A ) REFERENCES 1. OECD/NEA, OECD/NEA Workshop on Accident Tolerant Fuels of LWRs, NEA Headquarters, December 10-12, (2012). 2. MIT, Canes-MIT Symposium on Advanced LWR Fuels, MIT Student Center, March 20, (2012). 3. OECD/NEA, IAEA TWG on Fuel Performance & Technology, IAEA Headquarters, April 24-26, (2013). 4. L. Braase, S. Bragg-Sitton, Advanced Fuel Campaign Cladding & Coatings Meeting Summary, INL/EXT , March (2013). 5. L. Braase, Enhanced Accident Tolerant LWR Fuels National Metrics Workshop Report, INL/EXT , January (2013). 6. D. J. Kim, Y. W. Rhee, J. H. Kim, K. S. Kim, J. S. Oh, J. H. Yang, Y. H. Koo, K. W. Song, Fabrication of Micro-cell U-Mo Pellet with Enhanced Thermal Conductivity, Journal of Nuclear Materials, 462, (2015). 7 H. G. Kim, I. H. Kim, Y. I. Jung, D. J. Park, J. Y. Park, Y. H. Koo, Adhesion Property and High-Temperature Oxidation Behavior of Cr-coated Zircaloy-4 Cladding Tube Prepared by 3D Laser Coating, Journal of Nuclear Materials, 465, (2015). 8. Information Systems Laboratories, Inc., Idaho Falls, USA, RELAP5/MOD3.3 Code Manual Vol. I VIII, NUREG/CR- 5535/Rev P4, October (2010). 9. Idaho National Engineering and Environmental Laboratory, Idaho Falls, USA, SCDAP/RELAP5/MOD3.3 Code Manual Vol. 1-5, NUREG/CR-6150, INEL-96/0422, (2001). 10. J. Y. Cho et al., DeCART2D v1.0 User s Manual, KAERI/TR-5116/ J. Y. Cho et al., MASTER 3.0 USER S MANUAL, KAERI/UM-8/ L.J. Ott, K.R. Robb, D. Wang, Preliminary Assessment of Accident-Tolerant Fuels on LWR Performance During Normal Operation and Under DB and BDB accident conditions, Journal of Nuclear Materials, 448, (2014). 10