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

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1 Behavior of High-burnup Advanced LWR Fuels under Design-basis Accident Conditions Masaki Amaya 1, Yutaka Udagawa 1, Takafumi Narukawa 1, Takeshi Mihara 1, Yoshinori Taniguchi 1 1 Nuclear Safety Research Center, Japan Atomic Energy Agency: Tokai-mura, Naka-gun, Ibaraki, Japan amaya.masaki@jaea.go.jp ABSTRACT: Advanced fuels for light water reactors (LWRs) which consist of cladding materials with high corrosion resistance and pellets with lower fission gas release have been developed by utilities and fuel vendors to improve the fuel performance even in the high burnup region and also raise the safety level of current nuclear power plants to a higher one. In order to evaluate adequacy of present safety criteria and safety margins in terms of such advanced fuels and provide a database for future regulation on them, Japan Atomic Energy Agency (JAEA) has conducted a research program called ALPS-II program, which has been sponsored by Nuclear Regulation Authority (NRA), Japan. In this program, tests simulating a reactivity-initiated accident (RIA) and a loss-of-coolant accident (LOCA) have been performed on high burnup advanced fuels irradiated in commercial PWRs or BWRs in Europe. The failure limits of the high-burnup advanced fuels under RIA conditions have been obtained by pulse irradiation tests at the Nuclear Safety Research Reactor (NSRR) in JAEA. Post-test examinations are being performed on the fuel rods after pulse irradiation, and the information about fission gas release, pellet fragmentation, etc. during the pulse irradiations was obtained. In terms of the simulated LOCA test, integral thermal shock tests and high-temperature oxidation tests have been performed at the Reactor Fuel Examination Facility in JAEA, and the fracture limits under LOCA and post-loca conditions, oxidation rates in high-temperature steam, etc. of the high-burnup advanced fuel cladding have been investigated. From the results of these tests, it was found that in terms of these materials the fracture boundaries do not decrease and the oxidation does not significantly accelerate in the burnup level examined. KEYWORDS: Advanced fuels, High-burnup, Reactivity-initiated accident, Loss-of-coolant accident, Failure, Fracture, Oxidation. I. INTRODUCTION Japan Atomic Energy Agency (JAEA) performs the Advanced Light Water Reactor (LWR) Fuel Performance and Safety Research (ALPS) program to better understand behavior of high burnup fuels under accident conditions and to provide a database for regulatory judgment. The program was launched in the Japanese fiscal year (JFY) 2002 and extensive examinations including reactivity-initiated accident (RIA) studies and loss-of-coolant accident (LOCA) studies have been performed by using some unique facilities such as Nuclear Safety Research reactor (NSRR) and Reactor Fuel Examination Facility (RFEF) in JAEA, Tokai. The ALPS program is separated into two phases. In the first phase of the program (ALPS-I) which had been conducted from JFYs 2002 to 2007, high burnup UO 2 and MOX fuels with conventional fuel materials were subjected to pulseirradiation experiments in the NSRR and LOCA-simulated tests in the RFEF. Based on these results, the threshold for fuel failure, data on fuel behavior, etc. under RIA and LOCA have been successfully obtained in the high burnup range from about 45 to 78 GWd/t. Following the ALPS-I program, the second phase of the program (ALPS-II) started in JFY The main objectives of the ALPS-II program are to obtain technical information and create a database which would be required for regulatory evaluation and judgement on the fuels with advanced cladding materials with high corrosion resistance and/or pellets with lower fission gas release, and JAEA has conducted experiments simulating design-basis accident conditions for 1

2 the fuels with advanced fuel materials by using the NSRR and RFEF. Up to now, five RIA-simulated tests, eight LOCAsimulated integral thermal shock tests, post-test examinations related to these tests, etc. have been completed 1-3. The present paper describes some recent experimental results with respect to RIA and LOCA which have been obtained in this program. II. TEST FUEL FOR ALPS-II PROGRAM In the ALPS-II program, UO 2 and MOX fuels which had been irradiated to GWd/t (local burnup) in six European commercial reactors and UO 2 disk specimens which had been irradiated to about 130 GWd/t at the Halden reactor in Norway were gathered into a site in Europe. The cladding materials for these fuels are M-MDA TM, low-tin ZIRLO TM, M5 TM, Zircaloy-2/LK3, and the pellet materials are undoped UO 2, UO 2 with additives and MOX. These fuels with advanced materials were successfully transported from Europe to JAEA, Tokai in Japan at the beginning of 2011, and are being subjected to a series of RIA and LOCA experiments. Some information about the advanced fuels for the ALPS-II program is listed in TABLE I. III. RIA STUDY Reactor type PWR BWR TABLE I. List of advanced fuels for ALPS-II program. Fuel type Nuclear power plant Cladding material Burnup (GWd/t) Low-tin ZIRLO TM 80 M-MDA TM x17 UO 2 Vandellos Gravelines M5 TM * x15 UO 2 Ringhals M5 TM 68 17x17 MOX Chinon M5 TM 64 10x10 UO 2 Zry-2/LK Leibstadt Zry x10 Doped-UO 2 Oskarshamn Zry-2 63 HBWR Disk UO 2 Halden * M5 is a trademark of AREVA NP registered in the USA and in other countries. III.A. Outline of the RIA test conducted at the NSRR The NSRR is a modified TRIGA (Training, Research, Isotopes, General Atomics) annular core pulse reactor which can be used for simulating a rapid power pulse anticipated in a RIA. The details of the NSRR and NSRR experiments were described elsewhere 4. As a typical power history during the pulse irradiation, the full width at half maximum of power pulse and peak power is about 4 ms and 23 GW at the maximum reactivity insertion of $4.6, respectively. In the case of tests on the fuel irradiated in commercial reactors, a short test rod is prepared from the irradiated fuel rod at the RFEF in JAEA. Detailed examinations, e.g. visual and dimensional inspections, axial gamma scanning, X-ray radiography, etc., are carried out for the test rod before the pulse irradiation test. After the test rod has been transported from the RFEF to the NSRR, the test rod and test capsule are instrumented and assembled in hot cells in the NSRR. In the case of RIA tests on irradiated fuels at the NSRR, a test capsule with double walls which is made of stainless steel is used in order to ensure air tightness for preventing the release of radioactive materials from the capsule. Figures 1(a) and (b) show schematic diagrams of the test capsules used for NSRR experiments 5. Pulse irradiation tests with stagnant coolant water at room temperature (~20ºC) and atmospheric pressure (~0.1 MPa) are performed by using the capsule illustrated in Figure 1(a), and the capsule illustrated in Figure 1(b) is used for pulse irradiation tests under high-temperature-and-high-pressure conditions: the inner capsule is equipped with an electric heater to raise the coolant temperature and pressure up to ~280ºC and ~6.4 MPa (saturation pressure at 280ºC), respectively. 2

3 At the center of the NSRR core, there is a large cavity (220 mm in diameter) with loading tubes; and after the test capsule is loaded into the reactor core through the loading tube, the reactor is operated under the test condition desired. After a pulse irradiation test, the test capsule is dismantled; only the inner capsule with the test rod is transported to the RFEF and the test rod is retrieved from the inner capsule in the RFEF hot cells. The retrieved test rod is subjected to detailed post-testexaminations which are similar to those conducted before the pulse irradiation test. Fig. 1. Schematics of the test capsule used for NSRR test 5. III.B. Recent results of RIA tests and post-test examinations for advanced fuels Advanced cladding materials with high corrosion resistance and pellets with lower fission gas release have been developed by utilities and fuel vendors to improve the fuel performance even in the high burnup region and also to increase the safety of current nuclear power plants to a higher level. JAEA has started RIA tests for fuels with such advanced materials in the ALPS-II program in order to obtain data which are needed for regulatory judgement concerning RIA. In this paper, some results of post-test examinations which were newly obtained on the test fuel rods of tests VA-5 and - 6 are described. Tests VA-5 and -6 were conducted in JFYs 2014 and In terms of test VA-5, the test fuel rod was segmented from a PWR fuel rod irradiated in Vandellos-2 Nuclear power plant in Spain. Here, the test fuel rod for test VA-5 had M-MDA TM cladding with a final fabrication heat treatment under stress-relieved (SR) conditions, and had pellets with a conventional grain size. The average burnup of the test segment was ~80 GWd/t. The test rod with a 117 mm fuel stack length was prepared by cutting the segmented fuel rod in the RFEF in JAEA. The test rod was subjected to a pulse irradiation at room temperature in the NSRR with a peak fuel enthalpy of 580 J/g and a pulse width of 4.4 ms, respectively. Test VA-5 resulted in fuel failure at an enthalpy of 298 J/g. Regarding test VA-6, the test fuel rod had M-MDA TM cladding with a final fabrication heat treatment under recrystallized (RX) conditions and had pellets with a conventional grain size. The test rod was prepared in a way similar to that of VA-5. The average burnup of the test segment was ~80 GWd/t. The rod was subjected to a pulse irradiation at room temperature in the NSRR with a peak fuel enthalpy of 601 J/g and a pulse width of 4.4 ms, respectively. The test resulted in fuel failure at an enthalpy of 142 J/g. Specifications of test rods for tests VA-5 and -6, NSRR test conditions and some test results are summarized in TABLE II. Concerning test VA-5, the amount of fission gas release (FGR) during the test was estimated based on the analysis results 3

4 of capsule inner gas and the amount of fission gas generated during normal operation which was calculated by using the ORIGEN-2 code 6, 7. The FGR value was reported as ~30.6% 3. TABLE II. NSRR test conditions and results. Test ID VA-5 VA-6 Fuel enrichment (%) 4.9 Fuel pellet diameter (mm) 8.19 Cladding inner/outer diameter (mm) 8.36/9.5 Cladding material M-MDA TM (SR) M-MDA TM (RX) Rod burnup (GWd/t) Averaged cladding oxide thickness (μm) Fuel stack length (mm) 117 Fill gas in the rod (MPa) 0.1 (He) Coolant condition 0.1 MPa, stagnant Water ~ K Max. increase in fuel enthalpy (J/g) Fuel enthalpy increase at failure (J/g) Estimated FGR during the test (%) ~30.6 (will be estimated) Commercial use of M-MDA TM (RX) as cladding is not planned. The fuel enthalpies at failure are plotted as a function of fuel burnup in Figure 2 3. The current Japanese PCMI failure criteria shown in the figure were defined up to 75 GWd/t in 1998, on the basis of the data obtained up to ~64 GWd/t at that time. It is found that the result from test VA-5 is above the PCMI failure threshold with an adequate margin. On the other hand, the result of test VA-6 is slightly lower than the PCMI failure threshold. Based on optical microscope observation results on a transverse cross section of each cladding before the pulse irradiation 2, 3 test and SEM (scanning electron microscope) observation results on the fracture surfaces of the cladding failed during the pulse irradiation 3, it is considered that orientations of the hydrides precipitated in the cladding affects the PCMI failure limit of the cladding during the pulse irradiation. BWR NSRR PWR ATR JMTR SPERT CABRI &PBF MOX UO2 failure non failure Fuel enthalpy increase [J/g] VA 5 PCMI failure criteria VA Burnup [GWd/tU] Fig. 2 Fuel enthalpy increases at failure as a function of fuel burnup 3. Fuel enthalpy increase [cal/g] 4

5 The size distribution of the pellet fragments collected from the test capsule of test VA-6 are compared with that of test VA-5 in Figure 3. Here, it should be noted that not all amounts of pellet fragments released from the fuel rods during these tests were included because the test rod transportation from NSRR to RFEF was conducted after the test. Regarding the fraction of pellet fragments below 32 m in diameter, it is found that the fraction for test VA-6 tends to be larger than that of test VA-5 although the test rods used in tests VA-5 and -6 have the same design specification, power history and pulse irradiation condition. Therefore, it is considered that the difference in the size distributions of pellet fragments between tests VA-5 and -6 is mainly due to the difference between the test rods for tests VA-5 and -6 with respect to cladding conditions such as final heat treatment at fabrication, oxide thickness and the amount of absorbed hydrogen. Since these cladding conditions could influence the PCMI failure limit of cladding, the difference in the size distribution of pellet fragments may reflect that in the PCMI failure limit of cladding: pellet fragmentation behavior would change with temperature (i.e. the amount of expansion) of gas bubbles in the pellet, the temperature at which pellets in the test rod contact coolant in the test capsule, etc. during the pulse irradiation tests. (a) Test VA-5 (b) Test VA-6 IV. LOCA STUDY Fig. 3. Size distribution of the pellet fragments collected from the test capsules after tests VA5 and -6. In a safety analysis for a postulated LOCA (a design-basis-accident LOCA, a DBA-LOCA), it is estimated that fuel cladding is exposed to steam at high temperatures for several minutes until the emergency core cooling water quenches the fuel bundles in the reactor core. During this process, it is possible that the fuel cladding is severely oxidized and embrittled. In order to maintain the coolable geometry of the reactor core during a LOCA, the Japanese LOCA criteria require that the oxidation of the cladding, calculated by using the oxidation rate equation of Baker-Just, shall not exceed 15% of the cladding thickness (ECR: Equivalent Cladding Reacted) in order to prevent massive failure of fuel rods and/or fuel bundles by quenching. The limit is mainly based on the thermal shock resistance (fracture/no-fracture boundary) of oxidized cladding which was experimentally determined under simulated DBA-LOCA conditions. With increasing fuel burnup, the effects of corrosion, hydrogen absorption and neutron irradiation on the mechanical properties of cladding become considerable. Since the current Japanese LOCA criteria are based mostly on the data obtained with unirradiated cladding, the thermal shock resistance of high burnup fuel rods has been a concern for the safety of LWRs in Japan. In addition, fuel with advanced cladding material has been developed and is considered to be introduced in near future. From the viewpoint of the safety review of nuclear power plants, investigations on the behavior of fuel with advanced cladding material under LOCA conditions is needed. JAEA has conducted studies on fuel behavior under a DBA-LOCA condition by carrying out experiments such as integral thermal shock tests and cladding oxidation rate tests. The outline of the integral thermal shock test is as follows: a segment of about 190 mm in length was cut from the fuel rod, and the fuel pellets inside the segment were mechanically removed by drilling. Afterwards dummy Al 2 O 3 pellets were inserted into the cladding specimen in order to simulate the thermal condition of a fuel rod during a LOCA and end plugs 5

6 were welded on both ends of the cladding specimen. The inside of the test rod was pressurized with Ar gas of about 5MPa at room temperature and sealed. Four R-type thermocouples were spot-welded onto the outer surface of the test rod at different elevations to control and measure the cladding temperature during the test. The test rod was set inside the quartz tube of a test apparatus, and heated up at a rate of 3 to 10 K/s to a desired target oxidation temperature (usually 1473 K (1200 o C)) by using an infrared furnace under steam flow conditions. The test rod was ballooned and ruptured during the heat-up stage, and was isothermally oxidized after the rupture, that is, from both the inner and outer surfaces of the ruptured cladding for a predetermined oxidation period at the desired target oxidation temperature. After the isothermal oxidation, the rod was cooled in a steam flow to about 970 K and finally quenched by flooding water from the bottom. To achieve a constrained condition during the quench, which is expected in the bundle geometry, both ends of the test rod were fixed just before the cooling stage initiates. The tensile load increases as the rod was cooled and quenched, because cladding shrinkage was restrained. Since it is considered that a fully constrained condition is too severe, the restraint load was controlled to not exceed 540 N: this load was determined based on previous studies 8, 9. The outline of the cladding oxidation rate test is as follows: test specimens (~10 mm in length) were prepared from the fuel cladding after the fuel pellets were mechanically removed. One of the test specimens was set on the holder of an apparatus for the cladding oxidation rate test, and the holder with the specimen was inserted into the furnace of the apparatus kept at a desired temperature with a steam flow. After isothermal oxidation of the specimen for a desired period, the holder was quickly withdrawn from the furnace and the specimen was quickly cooled down to room temperature. The weight of the specimen after oxidation was measured, and the rate constant of oxidation was evaluated based on the weight and surface area of the specimen before oxidation. In the ALPS-II program, the integral thermal shock tests and oxidation rate tests have been conducted from JFY Eight integral thermal shock tests and two oxidation rate tests for several advanced cladding materials have been completed and post-test-examinations are now being conducted. Figure 4 shows the fracture map relevant to ECR and oxidation temperature for advanced fuel cladding tubes 3. In this figure, the black and gray symbols show the data obtained in the ALPS-II program and previous tests, respectively, and the notations show the specimen IDs which include cladding materials (MMDAs: M-MDA, LMFIs: M5 TM, LZIRs: low-tin ZIRLO TM, and LZRTs: Zircaloy-2(LK3)), and the amount of oxidation which is shown as ECR was calculated by using the Baker-Just equation 10 with oxidation temperature and time, taking account of the metal layer thickness in the cladding before the test and the decrease in cladding-wall thickness due to the ballooning that occurred during the test. As shown in this figure, it was confirmed that the fracture limits of the fuel cladding tubes with the advanced materials are located above the current Emergency Core Cooling System (ECCS) safety criteria in Japan in the burnup range up to about 85 GWd/t. Fig. 4. The fracture map relevant to ECR and oxidation temperature for advanced fuel cladding. Open symbol: non-fractured during the test. Solid symbol: fractured during the test. 6

7 In order to investigate the mechanical strength of a cladding tube under e.g. seismic conditions after the cladding tube experienced a LOCA, a four-point bend test was conducted on a low-tin ZIRLO TM cladding specimen after the integral thermal shock test. Characteristics of the specimen are summarized in TABLE III. Figure 5 compares the maximum bending moment obtained in this study with the literature data obtained at 410 K for unirradiated Zry-4 cladding 11. From this figure, it is found that the maximum bending moment of the low-tin ZIRLO TM cladding is quite similar to those of unirradiated Zry-4 cladding. This suggests that the burnup and the difference in cladding material hardly affect the maximum bending moment of cladding. Test rod ID Burnup (GWd/t) LZIR4 80 TABLE III. Characteristics of the specimen subjected to a four-point bend test. Cladding material Specimen conditions before the integral thermal shock test Oxide layer thickness ( m) The amount of absorbed hydrogen (ppm) Results of the integral thermal shock test Isothermal oxidation temperature (K) ECR* (%) Fracture/ non-fracture Axial restraint load at quench (N) Low-tin ZIRLO TM Non-fracture 526 * Calculated by using the Baker-Just equation with oxidation temperature and time. In this calculation, the metal layer thickness in the cladding before the test and the decrease in cladding-wall thickness due to ballooning during the test were taken into account. Maximum bending moment (N m) The amount of oxidation (ECR, %) Fig. 5. The relationship between the maximum bending moment and the amount of oxidation for the low-tin ZIRLO TM specimen with a burnup of ~80GWd/t after the integral thermal shock test. The open symbols show the literature data 11 and the solid symbol shows the data of the low-tin ZIRLO specimen obtained in this study. Table IV summarizes the characteristics of the irradiated fuel cladding with advanced materials which were subjected to the oxidation test. The weight gains of irradiated low-tin ZIRLO TM cladding specimens after being oxidized at 1473 K for up to 900 s are plotted in Figure 6, and Figure 7 shows the temperature dependence of the parabolic rate constants for weight gain. The data for unirradiated Zry-4 and low-tin ZIRLO TM materials are also shown for comparison in both figures. In terms of the oxidation rate at 1473 K, it is found that low-tin ZIRLO TM has a value similar to those of unirradiated Zry-4 and low-tin ZIRLO TM materials. This tendency suggests that the oxidation rate of low-tin ZIRLO TM does not significantly accelerate in the burnup and temperature levels examined. It is also found that the values estimated by using Baker-Just equation 10 give 7

8 conservative values in the burnup range up to about 80 GWd/t in the case of low-tin ZIRLO TM material. These tendencies are similar to those observed in other materials 3, 12. TABLE IV. Characteristics of the irradiated fuel cladding with advanced materials subjected to the oxidation tests. Specimen ID Burnup (GWd/t) OZIR2-1~5 80 Cladding Isothermal oxidation condition material Temperature (K) Duration (s) Low-tin ZIRLO TM ~900 15~40 The amount of oxidation (ECR, %)* * Calculated by using the Baker-Just equation 10 with oxidation temperature and time, based on the metal layer thickness in the cladding before the oxidation tests. Weight gain (g/m 2 ) 重量増加量 (g/m 2 ) 酸化温度 :1473 K 50 OZIR2 Unirradiated 未照射低スズlow-tin ZIRLO(H22 試験 TM ) Time 時間 (s) Fig. 6. Weight gains of irradiated low-tin ZIRLO TM as a function of oxidation time at 1473 K. Parabolic rate constant, kw (g 2 /m 4 /s) OZIR2 Temperature (K) Baker-Just eq. 10 Unirradiated low-tin ZIRLO TM 1200 Unirradiated Zry Reciprocal temperature ( 10-4 /K) Fig. 7. Temperature dependence of the parabolic rate constants for weight gain of low-tin ZIRLO TM. 8

9 V. SUMMARY Advanced fuels for light water reactors (LWRs) which consist of cladding materials with high corrosion resistance and pellets with lower fission gas release have been developed by utilities and fuel vendors to improve the fuel performance even in the high burnup region and also raise the safety level of current nuclear power plants to a higher one. In order to evaluate the adequacy of present safety criteria and safety margins in terms of such advanced fuels and provide a database for their future regulation, Japan Atomic Energy Agency (JAEA) has conducted a research program called ALPS-II program, which has been sponsored by the Nuclear Regulation Authority (NRA), Japan. In this program, tests simulating a reactivity-initiated accident (RIA) and a loss-of-coolant accident (LOCA) have been performed on the high burnup advanced fuels irradiated in commercial PWRs or BWRs in Europe. The failure limits of the high-burnup advanced fuels under RIA conditions have been obtained by pulse irradiation tests at the Nuclear Safety Research Reactor (NSRR) in JAEA. Post-test examinations are being performed on the fuel rods after pulse irradiation. In terms of the simulated LOCA test, integral thermal shock tests and high-temperature oxidation tests have been performed at the Reactor Fuel Examination Facility in JAEA, and the fracture limits under LOCA and post-loca conditions, oxidation rates in high-temperature steam, etc. of the high-burnup advanced fuel cladding have been investigated. Some post-test examinations were performed on the fuel rods after the pulse irradiation tests, and information on fission gas release, pellet fragmentation, etc. during the pulse irradiations was obtained. A burnup effect on fission gas release during pulse irradiation was suggested, and a difference in the extent of fragmentation of the pellets released into the experimental capsules was observed. The latter seems to be due to the difference in the contact surface area between fuel pellet and coolant water just after the fuel failure in these pulse irradiation tests. In terms of the simulated LOCA test, integral thermal shock tests and high temperature oxidation tests have been conducted on high-burnup PWR and BWR fuel cladding with advanced materials, and their fracture limits, high temperature oxidation rates, etc. were investigated. In the integral thermal shock tests for these high-burnup advanced cladding tubes, no fractures were observed. Regarding the oxidation rate tests, the hightemperature oxidation rates of these high-burnup advanced cladding tubes were similar to those of unirradiated Zry-4 and lower than the values estimated by the Baker-Just equation. From the results of these tests, it was found that in terms of these materials the fracture boundaries do not decrease and the oxidation does not significantly accelerate in the burnup level examined. ACKNOWLEDGMENTS The ALPS program has been sponsored by Nuclear Regulation Authority (NRA), Japan. REFERENCES 1. M. AMAYA, F. NAGASE, T. SUGIYAMA, Y. UDAGAWA, T. NARUKAWA and A. SAWADA, Current Studies at JAEA on Fuel Behaviors under Accident Conditions, Proceedings of WRFPM 2014, Sendai, Japan, September 14-17, 2014, Paper (2014). 2. M. AMAYA, Y. UDAGAWA, T. NARUKAWA, T. MIHARA and T. SUGIYAMA, Behavior of high burnup advanced fuels for LWR during design-basis accidents, Proceedings of TopFuel 2015, Zurich, Switzerland, September 13-17, 2015, Paper A0032 (2015). 3. M. AMAYA, Y. UDAGAWA, T. NARUKAWA, T. MIHARA and Y. TANIGUCHI, Behavior of High-burnup Advanced LWR Fuels under Accident Conditions, Proceedings of Top Fuel 2016, Boise, Idaho, USA, September 11-15, 2016, pp (2016) 4. T. FUKETA, H. SASAJIMA and T. SUGIYAMA, Behavior of High-Burnup PWR Fuels with Low-Tin Zircaloy-4 Cladding under Reactivity-Initiated-Accident Conditions, Nucl. Technol., 133, 50 (2001). 5. T. FUKETA, F. NAGASE, T. SUGIYAMA and M. AMAYA, Behavior of High Burnup LWR Fuels during Design-basis Accidents; Key Observations and an Outline of the Coming Program, Proceedings of 2010 LWR Fuel Performance/TopFuel/WRFPM, Orlando, Florida, USA, September 26-29, 2010, Paper 0033 (2010). [CD-ROM]. 6. S. B. LUDWIG and J. P. RENIER, Standard and Extended Burnup PWR and BWR Reactor Models for the ORIGEN2 Computer Code, ORNL/TM-11018, Oak Ridge National Laboratory (1989). 9

10 7. J. KATAKURA, M. KATAOKA, K. SUYAMA, T. JIN and S. Ohki, A Set of ORIGEN2 Cross Section Libraries Based on JENDL-3.3 Library: ORLIBJ33, JAERI-Data/Code , Japan Atomic Energy Agency (2004) [in Japanese]. 8. T. MURATA, Y. TANIGUCHI, S. URATA, et al., LOCA Simulation Test of the Cladding for High-Burnup Fuel, Proceedings of ANS Annual Meeting, Milwaukee, Wisconsin, USA, 2001 (2001). 9. T. HONMA, S. DOI, M. OZAWA, et al., Thermal-Shock Behavior of PWR High-Burnup Fuel Cladding Under Simulated LOCA Conditions, Proceedings of ANS Annual Meeting, Milwaukee, Wisconsin, USA, 2001 (2001). 10. L. BAKER and L. C. JUST, Studies of Metal-Water Reaction at High Temperatures III. Experimental and Theoretical Studies of the Zirconium-Water Reaction, ANL-6548, Argonne National Laboratory (1962). 11. M. YAMATO, F. NAGASE and M. AMAYA, Evaluation of fracture resistance of ruptured, oxidized, and quenched Zircaloy cladding by four-point-bend tests, J. Nucl. Sci. Technol., 51, 1125 (2014). 12. T. CHUTO, F. NAGASE and T. FUKETA, High Temperature Oxidation of Nb-containing Zr Alloy Cladding in LOCA Conditions, Nucl. Eng. Technol., 41, 163 (2009). 10