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JAEA Reports

Improvement of intragranular fission gas behavior model for fuel performance code FEMAXI-8

Udagawa, Yutaka; Tasaki, Yudai

JAEA-Data/Code 2021-007, 56 Pages, 2021/07

JAEA-Data-Code-2021-007.pdf:5.05MB

Japan Atomic Energy Agency (JAEA) has released FEMAXI-8 in 2019 as the latest version of the fuel performance code FEMAXI, which has been developed to analyze thermal and mechanical behaviors of a single fuel rod in mainly normal operation conditions and anticipated transient conditions. This report summarizes a newly developed model to analyze intragranular fission gas behaviors considering size distribution of gas bubbles and their dynamics. While the intragranular fission gas behavior models implemented in the previous FEMAXI versions have supported only treating single bubble size for a given fuel element, the new model supports multiple gas groups according to their size and treats their dynamic behaviors, making the code more versatile. The model was first implemented as a general module that takes arbitrary number of bubble groups and spatial mesh division for a given fuel grain system. An interface module to connect the model to FEMAXI-8 was then developed so that it works as a 2 bubble groups model, which is the minimum configuration of the multi-grouped model to be operated with FEMAXI-8 at the minimum calculation cost. FEMAXI-8 with the new intragranular model was subjected to a systematic validation calculation against 144 irradiation test cases and recommended values for model parameters were determined so that the code makes reasonable predictions in terms of fuel center temperature, fission gas release, etc. under steady-state and power ramp conditions.

JAEA Reports

Technical basis of ECCS acceptance criteria for light-water reactors and applicability to high burnup fuel

Nagase, Fumihisa; Narukawa, Takafumi; Amaya, Masaki

JAEA-Review 2020-076, 129 Pages, 2021/03

JAEA-Review-2020-076.pdf:3.9MB

Each light-water reactor (LWR) is equipped with the Emergency Core Cooling System (ECCS) to maintain the coolability of the reactor core and to suppress the release of radioactive fission products to the environment even in a loss-of-coolant accident (LOCA) caused by breaks in the reactor coolant pressure boundary. The acceptance criteria for ECCS have been established in order to evaluate the ECCS performance and confirm the sufficient safety margin in the evaluation. The limits defined in the criteria were determined in 1975 and reviewed based on state-of-the-art knowledge in 1981. Though the fuel burnup extension and necessary improvements of cladding materials and fuel design have been conducted, the criteria have not been reviewed since then. Meanwhile, much technical knowledge has been accumulated regarding the behavior of high-burnup fuel during LOCAs and the applicability of the criteria to the high-burnup fuel. This report provides a comprehensive review of the history and technical bases of the current criteria and summarizes state-of-the-art technical findings regarding the fuel behavior during LOCAs. The applicability of the current criteria to the high-burnup fuel is also discussed.

Journal Articles

Thresholds for failure of high-burnup LWR fuels by pellet cladding mechanical interaction under reactivity-initiated accident conditions

Udagawa, Yutaka; Sugiyama, Tomoyuki; Amaya, Masaki

Journal of Nuclear Science and Technology, 56(12), p.1063 - 1072, 2019/12

 Times Cited Count:3 Percentile:57.07(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Model updates and performance evaluations on fuel performance code FEMAXI-8 for light water reactor fuel analysis

Udagawa, Yutaka; Amaya, Masaki

Journal of Nuclear Science and Technology, 56(6), p.461 - 470, 2019/06

 Times Cited Count:3 Percentile:57.07(Nuclear Science & Technology)

no abstracts in English

JAEA Reports

Development of fuel performance code FEMAXI-8; Model improvements for light water reactor fuel analysis and systematic validation

Udagawa, Yutaka; Yamauchi, Akihiro*; Kitano, Koji*; Amaya, Masaki

JAEA-Data/Code 2018-016, 79 Pages, 2019/01

JAEA-Data-Code-2018-016.pdf:2.75MB

FEMAXI-8 is the latest version of the fuel performance code FEMAXI developed by JAEA. A systematic validation work has been achieved against 144 irradiation test cases, after many efforts have been made, in development of new models, improvements in existing models and the code structure, bug-fixes, construction of irradiation-tests database and other infrastructures.

Journal Articles

Technical basis of accident tolerant fuel updated under a Japanese R&D project

Yamashita, Shinichiro; Nagase, Fumihisa; Kurata, Masaki; Nozawa, Takashi; Watanabe, Seiichi*; Kirimura, Kazuki*; Kakiuchi, Kazuo*; Kondo, Takao*; Sakamoto, Kan*; Kusagaya, Kazuyuki*; et al.

Proceedings of 2017 Water Reactor Fuel Performance Meeting (WRFPM 2017) (USB Flash Drive), 10 Pages, 2017/09

In Japan, the research and development (R&D) project on accident tolerant fuel and other components (ATFs) of light water reactors (LWRs) has been initiated in 2015 for establishing technical basis of ATFs. The Japan Atomic Energy Agency (JAEA) has coordinated and carried out this ATF R&D project in cooperation with power plant providers, fuel venders and universities for making the best use of the experiences, knowledges in commercial uses of zirconium-base alloys (Zircaloy) in LWRs. ATF candidate materials under consideration in the project are FeCrAl steel strengthened by dispersion of fine oxide particles(FeCrAl-ODS) and silicon carbide (SiC) composite, and are expecting to endure severe accident conditions in the reactor core for a longer period of time than the Zircaloy while maintaining or improving fuel performance during normal operations. In this paper, the progresses of the R&D project are reported.

Journal Articles

Results from studies on high burn-up fuel behavior under LOCA conditions

Nagase, Fumihisa; Fuketa, Toyoshi

NUREG/CP-0192, p.197 - 230, 2005/10

The Japanese regulatory criterion for a loss-of-coolant-accident (LOCA) is based on a threshold of fuel rod fracture during quenching, which was experimentally determined under simulated LOCA conditions. In order to evaluate the fracture threshold of high burn-up fuel rods, JAERI performs integral thermal shock tests simulating LOCA conditions. The tests have been performed with pre-hydrided, unirradiated claddings and high burn-up fuel claddings irradiated to 39 and 44 GWd/t at a PWR. It was shown that fracture/no-fracture threshold primarily depends on the oxidation amount and that the threshold decreases with increases in hydrogen concentration and axial restraint during the quench. It was also shown that fracture conditions of the tested high burn-up fuel claddings are consistent with the fracture threshold derived from unirradiated claddings with similar hydrogen concentrations.

Journal Articles

Investigation on Innovative Water Reactor for Flexible Fuel Cycle (FLWR), 1; Conceptual design

Uchikawa, Sadao; Okubo, Tsutomu; Kugo, Teruhiko; Akie, Hiroshi; Nakano, Yoshihiro; Onuki, Akira; Iwamura, Takamichi

Proceedings of International Conference on Nuclear Energy System for Future Generation and Global Sustainability (GLOBAL 2005) (CD-ROM), 6 Pages, 2005/10

no abstracts in English

JAEA Reports

Report on the 8th Workshop on the Innovative Water Reactor for Flexible Fuel Cycle; February 10, 2005, Koku-kaikan, Minato-ku, Tokyo

Kobayashi, Noboru; Okubo, Tsutomu; Uchikawa, Sadao

JAERI-Review 2005-029, 119 Pages, 2005/09

JAERI-Review-2005-029.pdf:11.01MB

The research on Innovative Water Reactor for Flexible fuel cycle (FLWR) has been performed in JAERI for the development of future innovative reactors. The workshop on the FLWRs has been held every year since 1998 aiming at information exchange between JAERI and other organizations. The 8th workshop was held on Feb. 10, 2005 under the joint auspices of JAERI and North Kanto and Kanto-Koetsu branches of Atomic Energy Society of Japan with 75 participants. The workshop began with 3 presentations on FLWRs entitled "Framework and Status of Research and Development on FLWRs", "Long-Term Fuel Cycle Scenarios for Advanced Utilization of Plutonium from LWRs", and "Experiments on Characteristics on Hydrodynamics in Tight-Lattice Core". Then 3 lectures followed: "Development of Evaluation Method for Accuracy in Predicting Neutronics Characteristics of Tight-Lattice Core" by Osaka University, "Development of Cost-Reduced Low-Moderation Spectrum Boiling Water Reactor" by Toshiba Corporation and "Design and Analysis on Super-Critical Water Cooled Power Reactors" by Tokyo University.

Journal Articles

RIA- and LOCA-simulating experiments on high burnup LWR fuels

Fuketa, Toyoshi; Nagase, Fumihisa; Sugiyama, Tomoyuki

Proceedings of IAEA Technical Meeting on Fuel Behaviour Modelling under Normal, Transient and Accident Conditions, and High Burnups (CD-ROM), 15 Pages, 2005/09

To provide a data base for the regulatory guide of light water reactors, behaviors of reactor fuels during off-normal and postulated accident conditions such as reactivity-initiated accident (RIA) and loss-of-coolant accident (LOCA) are being studied at the Japan Atomic Energy Research Institute (JAERI). A series of RIA-simulating experiments with high burnup fuel rods is being performed by using pulse-irradiation capability of the Nuclear Safety Research Reactor (NSRR). Fuel behaviors during a LOCA are also examined in an extensive program comprising of integral thermal shock tests and separate tests for oxidation rate and mechanical properties of fuel cladding.

Journal Articles

Rationalization of the fuel integrity and transient criteria for the super LWR

Yamaji, Akifumi*; Oka, Yoshiaki*; Ishiwatari, Yuki*; Liu, J.*; Koshizuka, Seiichi*; Suzuki, Motoe

Proceedings of 2005 International Congress on Advances in Nuclear Power Plants (ICAPP '05) (CD-ROM), 7 Pages, 2005/05

Ensuring the fuel integrities is one of the most fundamental parts in the High Temperature Supercritical-Pressure Light Water Reactor. Most abnormal transient events of SCLWR-H last for a short period of time and the fuel rods are replaced after being irradiated in the core. In this study, the fuel integrity criteria are rationalized based on the fact that the fuel rod mechanical failures can be represented by the strain of the fuel rod cladding. A new fuel rod is designed with a Stainless Steel cladding. It is internally pressurized to reduce the stress on the cladding and also to increase the gap conductance between the pellet and the cladding. The fuel integrities both at normal operation and abnormal transient conditions are evaluated using the fuel analysis code FEMAXI-6 of JAERI.

Journal Articles

Introduction to nuclear fuel engineering, 9; LWR fuel behavior

Fuketa, Toyoshi; Nagase, Fumihisa; Sasahara, Akihiro*

Nihon Genshiryoku Gakkai-Shi, 47(2), p.112 - 119, 2005/02

Behavior of LWR fuel during reactivity-initiated accident (RIA) and loss-of-coolant accident (LOCA) is described.

Journal Articles

Analysis on lift-off experiment in Halden reactor by FEMAXI-6 code

Suzuki, Motoe; Kusagaya, Kazuyuki*; Saito, Hiroaki*; Fuketa, Toyoshi

Journal of Nuclear Materials, 335(3), p.417 - 424, 2004/12

 Times Cited Count:5 Percentile:36.97(Materials Science, Multidisciplinary)

Experimental analysis was conducted on the Lift-Off experiment IFA-610.1 in Halden reactor by the FEMAXI-6 code using the detailed measured conditions of test-irradiation. Calculated fuel center temperatures on the two assumptions, i.e., (1) an enhanced thermal conductance across the pellet-clad bonding layer is maintained during the cladding creep-out by over-pressurization, and (2) the bonding layer is broken by the cladding creep-out, were compared with the measured data to analyze the effect of the creep-out by over-pressure inside the test pin. The measured center temperature rise was higher by a few tens of K than the prediction performed on the assumption (1), though this difference was much smaller than the predicted rise on the assumption (2). Therefore, it is appropriate to attribute the measured center temperature rise to the decrease of effective thermal conductance by irregular re-location of pellet fragments, etc. which was caused by cladding creep-out.

Journal Articles

Influence of hydride re-orientation on BWR cladding rupture under accidental conditions

Nagase, Fumihisa; Fuketa, Toyoshi

Journal of Nuclear Science and Technology, 41(12), p.1211 - 1217, 2004/12

 Times Cited Count:13 Percentile:66.18(Nuclear Science & Technology)

Hydride precipitation along the radial-axial plane increases in high burn-up BWR fuel claddings. The radial hydrides may have an important role during fuel behavior in a RIA and may reduce ductility of the cladding under PCMI conditions. In order to promote a better understanding of the influence of the radial hydrides on cladding failure behavior under the PCMI conditions, tube burst tests were conducted for unirradiated BWR claddings charged with 200 to 650 ppm of hydrogen. About 20 to 30% of hydrides were re-oriented and precipitated along the radial-axial plane. The claddings exhibited large burst openings with an axial crack at room temperature and 373 K. However, the influence of the radial hydrides on both burst pressure and residual hoop strain was very small. It is accordingly expected that ductility of high burn-up BWR cladding is significantly reduced not only by precipitation of radial hydrides as far as hydrogen concentration and radial hydride fraction range in the present study.

JAEA Reports

A Set of ORIGEN2 cross section libraries based on JENDL-3.3 library; ORLIBJ33

Katakura, Junichi; Kataoka, Masaharu*; Suyama, Kenya; Jin, Tomoyuki*; Oki, Shigeo*

JAERI-Data/Code 2004-015, 115 Pages, 2004/11

JAERI-Data-Code-2004-015.pdf:16.52MB

A set of cross section libraries for ORIGEN2 code, ORLIBJ33, has been produced based on the latest Japanese Evaluated Nuclear Data Library JENDL-3.3. The produced libraries are for LWR's which include PWR, BWR and their MOX fuels. The libraries for FBR's are also produced. Using the libraries for LWR, comparisons with old libraries based on JENDL-3.2 were performed. The comparisons with measured PIE data were also carried out. For the libraries for FBR, the comparisons with the calculations using the old libraries were performed and the effects using different libraries were discussed.

Journal Articles

Results from simulated LOCA experiments with high burnup PWR fuel claddings

Nagase, Fumihisa; Fuketa, Toyoshi

Proceedings of 2004 International Meeting on LWR Fuel Performance, p.500 - 506, 2004/09

A systematic research program is being conducted at the Japan Atomic Energy Research Institute (JAERI), which aims at a wide range database for evaluating the influence of further burnup extension on fuel behavior under LOCA conditions. As a part of the program, integral thermal shock tests simulating the whole LOCA sequence have been conducted with Zircaloy-4 fuel claddings, irradiated to 39 and 44GWd/t at a PWR. One cladding, oxidized to about 30% ECR, fractured during the quench. The fracture condition agrees with the fracture criteria for non-irradiated claddings that have similar hydrogen concentrations (about 25% ECR). Two claddings, oxidized to about 16 and 18% ECR, survived the quench, indicating that fracture/non-fracture boundary is not reduced so significantly by irradiation for the examined burnup range. The present paper describes information obtained from the tests including oxidation kinetics and rupture behavior.

Journal Articles

Effect of pre-hydriding on thermal shock resistance of Zircaloy-4 cladding under simulated loss-of-coolant accident conditions

Nagase, Fumihisa; Fuketa, Toyoshi

Journal of Nuclear Science and Technology, 41(7), p.723 - 730, 2004/07

 Times Cited Count:41 Percentile:92.64(Nuclear Science & Technology)

Experiments simulating loss-of-coolant accident (LOCA) conditions were performed to evaluate effect of pre-hydriding on thermal-shock resistance of oxidized Zircaloy-4 cladding. Artificially hydrided (400 to 600 ppm) and non-hydrided claddings were subjected to the tests. Since cladding fracture on quenching primarily depends on the oxidation amount, fracture threshold was evaluated in terms of "Equivalent Cladding Reacted (ECR)". Under axially non-restrained condition, the fracture threshold is 56% ECR and the influence of pre-hydriding is not seen. The fracture threshold is decreased by restraining the test rods on quenching, and it is more remarkable in pre-hydrided claddings than in non-hydrided claddings. Consequently, the fracture threshold was 20% ECR and 10% ECR for non-hydrided and pre-hydrided claddings, respectively, under the fully restrained condition. These results indicate possible decrease of fracture threshold of high burnup fuel claddings under LOCA conditions.

Journal Articles

Analysis of benchmark results for reactor physics of LWR next generation fuels

Kitada, Takanori*; Okumura, Keisuke; Unesaki, Hironobu*; Saji, Etsuro*

Proceedings of International Conference on Physics of Fuel Cycles and Advanced Nuclear Systems; Global Developments (PHYSOR 2004) (CD-ROM), 8 Pages, 2004/04

Burnup calculation benchmark has been carried out for the LWR next generation fuels aiming at high burnup up to 70 GWd/t with UO$$_{2}$$ and MOX. Based on the submitted results by many benchmark participants, the present status of calculation accuracy has been confirmed for reactor physics parameters of the LWR next generation fuels, and the factors causing the calculation differences were analyzed in detail. Moreover, the future experiments and research subjects necessary to reduce the calculation differences were discussed and proposed.

JAEA Reports

Development of pellet melting temperature measuring technique; Melting temperature measuring technique for small sample

Harada, Katsuya; Nakata, Masahito; Harada, Akio; Nihei, Yasuo; Yasuda, Ryo; Nishino, Yasuharu

JAERI-Tech 2004-034, 13 Pages, 2004/03

JAERI-Tech-2004-034.pdf:0.69MB

The Department of Hot Laboratories has been aiming the establishment of the melting temperature measuring technique for small samples obtained from the micro-region of irradiated fuel pellet. Due to the modification of the shape of tungsten capsule contained sample and the improvement of the detection method for melting temperature from indistinct thermal arrest point owing to small sample, it is possible to determine the melting temperature of small sample and to utilize effectively for the irradiated fuel pellet by using the existing apparatus. This paper describes the technique of the melting temperature measurement for small sample and the experimental results by using tantalum, molybdenum, hafnium oxide and un-irradiated UO$$_{2}$$ pellet.

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