Udagawa, Yutaka; Tasaki, Yudai
JAEA-Data/Code 2021-007, 56 Pages, 2021/07
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.
Kakiuchi, Kazuo; Udagawa, Yutaka; Amaya, Masaki
Annals of Nuclear Energy, 155, p.108171_1 - 108171_11, 2021/06
Khatib-Rahbar, M.*; Barrachin, M.*; Denning, R.*; Gabor, J.*; Gauntt, R.*; Herranz, L. E.*; Hobbins, R.*; Jacquemain, D.*; Maruyama, Yu; Metcalf, J.*; et al.
NUREG/CR-7282, ERI/NRC 21-204 (Internet), 160 Pages, 2021/04
Kitamura, Akira; Akahori, Kuniaki; Nagata, Masanobu*
Genshiryoku Bakkuendo Kenkyu (CD-ROM), 27(2), p.83 - 93, 2020/12
Direct disposal of spent nuclear fuel (SNF) in deep underground repositories (hereafter "direct disposal") is a concept that disposal canisters stored fuel assemblies dispose without reprocessing. Behavior of radionuclide release from SNF must be different from that from vitrified glass. The present study established a methodology on determination of instant release fraction (IRF) of radionuclides from SNF, which is the one of the parameters on radionuclide release based on the latest safety assessment reports in other countries, especially for IRF values proportional to a fission gas release ratio (FGR). Recommended and maximum values of FGR have been estimated using the fuel performance code FEMAXI-7 after collecting FGR values on Japanese SNFs. Furthermore, recommended and maximum values of IRF for Japanese SNFs used in a pressurized water reactor (PWR) have been estimated using the presently obtained FGR values and experimentally obtained IRF values on foreign SNFs. The recommended and maximum IRF values obtained in the present study have been compared with those of the latest safety assessment reports in other countries.
Udagawa, Yutaka; Fuketa, Toyoshi*
Comprehensive Nuclear Materials, 2nd Edition, Vol.2, p.322 - 338, 2020/08
Udagawa, Yutaka; Amaya, Masaki
Journal of Nuclear Science and Technology, 56(6), p.461 - 470, 2019/06
no abstracts in English
Udagawa, Yutaka; Yamauchi, Akihiro*; Kitano, Koji*; Amaya, Masaki
JAEA-Data/Code 2018-016, 79 Pages, 2019/01
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.
Aihara, Jun; Ueta, Shohei; Goto, Minoru; Inaba, Yoshitomo; Shibata, Taiju; Ohashi, Hirofumi
JAEA-Technology 2018-002, 70 Pages, 2018/06
HTFP code is code for calculation of additional release amount of fission product (FP) from fuel rod in high temperature gas-cooled reactor (HTGR) after stop of fission. Minory changed Fornax-A code also can calculate that. Therefore, release behavior of Cs calculated with HTFP code was compared with that calculated with minory modified FORNAX-A code in this report. Release constants of Cs evaluated with minory modified FORNAX-A code are rather different from default values for HTFP code.
Udagawa, Yutaka; Sugiyama, Tomoyuki; Amaya, Masaki
Proceedings of 2016 International Congress on Advances in Nuclear Power Plants (ICAPP 2016) (CD-ROM), p.1183 - 1189, 2016/04
Aihara, Jun; Ueta, Shohei; Nishihara, Tetsuo
JAEA-Technology 2015-040, 32 Pages, 2016/02
Original FORNAX-A is a calculation code for amount of fission product (FP) released from fuel rods of pin-in-type high temperature gas-cooled reactors (HTGRs). This report is for explanation what calculations become possible with minor changed FORNAX-A.
Nomoto, Yasunobu; Aihara, Jun; Nakagawa, Shigeaki; Isaka, Kazuyoshi; Ohashi, Hirofumi
JAEA-Data/Code 2015-008, 39 Pages, 2015/06
HTFP is a calculation code for amount of additionally released fission product (FP) from fuel rods of pin-in-type according to transient of core temperature at the accident of high temperature gas-cooled reactors (HTGRs). This code analyzes FP release inventory from core according to the transient of core temperature at the accident as an input data and considering FP release rate from a fuel compact and a graphite sleeve and radioactive decay of FP. This report describes the outline of HTFP code and its input data. The computed solutions using the HTFP code were compared to those of HTCORE code, which was used for the design of the High Temperature Engineering Test Reactor (HTTR) to validate the analysis models of the HTFP code. The comparison of HTFP code results with HTCORE code results showed the good agreement.
Ueta, Shohei; Shaimerdenov, A.*; Gizatulin, S.*; Chekushina, L.*; Honda, Masaki*; Takahashi, Masashi*; Kitagawa, Kenichi*; Chakrov, P.*; Sakaba, Nariaki
Proceedings of 7th International Topical Meeting on High Temperature Reactor Technology (HTR 2014) (USB Flash Drive), 7 Pages, 2014/10
A capsule irradiation test with the high temperature gas-cooled reactor (HTGR) fuel is being carried out using WWR-K research reactor in the Institute of Nuclear Physics of the Republic of Kazakhstan (INP) to attain 100 GWd/t-U of burnup under normal operating condition of a practical small-sized HTGR. This is the first HTGR fuel irradiation test for INP in Kazakhstan collaborated with Japan Atomic Energy Agency (JAEA) in frame of International Science and Technology Center (ISTC) project. In the test, TRISO coated fuel particle with low-enriched UO (less than 10% of U) is used, which was newly designed by JAEA to extend burnup up to 100 GWd/t-U comparing with that of the HTTR (33 GWd/t-U). Both TRISO and fuel compact as the irradiation test specimen were fabricated in basis of the HTTR fuel technology by Nuclear Fuel Industries, Ltd. in Japan. A helium-gas-swept capsule and a swept-gas sampling device installed in WWR-K were designed and constructed by INP. The irradiation test has been started in October 2012 and will be completed up to the end of February 2015. The irradiation test is in the progress up to 69 GWd/t of burnup, and integrity of new TRISO fuel has been confirmed. In addition, as predicted by the fuel design, fission gas release was observed due to additional failure of as-fabricated SiC-defective fuel.
Amaya, Masaki; Sugiyama, Tomoyuki; Fuketa, Toyoshi
Journal of Nuclear Science and Technology, 41(10), p.966 - 972, 2004/10
Pulse irradiation simulating RIA condition was carried out for test rod prepared from fuel irradiated in a commercial reactor. After the pulse irradiation, optical microscopy (OM) and scanning electron microscopy (SEM) observations and electron probe micro analysis (EPMA) were conducted for the test rod as a part of destructive tests. Fission gas release behavior during pulse irradiation was investigated by EPMA and puncture test. Xeon depression was observed in the fuel pellet after pulse irradiation at periphery and center region. It is considered that fission gas was mainly released from the pellet center region during pulse irradiation. The amount of xenon release during pulse irradiation was estimated to be 10-12% from the EPMA results and this estimated value was comparable with the puncture test result. Comparing the estimated value with other results of out-of-pile annealing tests, it was concluded that most fission gas, which was accumulated at grain boundary during base irradiation, was released from the center region of test fuel pellet during pulse irradiation.
Hatakeyama, Yuichi; Sudo, Kenji; Kanazawa, Hiroyuki
JAERI-Tech 2004-033, 29 Pages, 2004/03
The amount of fission gas (Kr, Xe) in irradiated fuel pellet increases with extending the burn up and that exerts a serious influence upon thermal and mechanical properties of light water reactor fuel. Therefore, the accumulation of the data on the release behavior of fission gas is important in the investigation program of safety and reliability for extended burn up fuel. In the post irradiation examination at the Reactor Fuel Examination Facility in JAERI,the fission gas which released into the plenum region from UO pellet during irradiation has been measured by puncturing test of irradiated fuel rod. The results of puncturing test show the most of fission gas remained in the pellet. It can be seen that the additional release of fission gas might occur under higher burn up and accident conditions. To know the fission gas release behavior from irradiated fuel, the Out Gas analyzer(OGA)which has the performance to heat up the UO pellet stepwise up to 2300C and to measure the released fission gas instantly from the pellet has been developed and installed at RFEF.
Sugiyama, Tomoyuki; Nakamura, Takehiko; Kusagaya, Kazuyuki*; Sasajima, Hideo; Nagase, Fumihisa; Fuketa, Toyoshi
JAERI-Research 2003-033, 76 Pages, 2004/01
Boiling water reactor (BWR) fuels with burnups of 41 to 45 GWd/tU were pulse-irradiated in the Nuclear Safety Research Reactor (NSRR) to investigate fuel behavior under cold startup reactivity-initiated-accident (RIA) conditions. BWR fuel segment rods of 88BJ (STEP I) type from Fukushima-Daiichi Unit 3 nuclear power plant were refabricated into short test rods, and they were subjected to prompt enthalpy insertion from 293 to 607 J/g (70 to 145 cal/g) within about 20 ms. The fuel cladding had enough ductility against the prompt deformation due to pellet cladding mechanical interaction. The plastic hoop strain reached 1.5% at the peak location. The cladding surface temperature locally reached about 600 deg C. Recovery of irradiation defects in the cladding due to high temperature during the pulse irradiation was indicated via X-ray diffractometry. Fission gas release during the pulse irradiation was from 3.1% to 8.2%, depending on the peak fuel enthalpy and the normal operation conditions.
Suzuki, Motoe; Saito, Hiroaki*
JAERI-Data/Code 2003-019, 423 Pages, 2003/12
A light water reactor fuel analysis code FEMAXI-6 is an advanced version which has been produced by integrating the former version with a number of improvements. In particular, the FEMAXI-6 code has attained a complete coupled solution of thermal analysis and mechanical analysis, permitting an accurate prediction of pellet-clad gap size and PCMI in high burnup fuel rods. Also, such new models have been implemented as pellet-clad bonding and fission gas bubble swelling, and the coupling with burning analysis code has been enhanced. Furthermore, a number of new materials properties and parameters have been introduced. With these advancements, the FEMAXI-6 code is a versatile tool not only in the normal operation but also in transient conditions. This report describes the design, basic theory, models and numerical method, improvements, and model modification. In order to facilitate effective and wide-ranging application of the code, formats and methods of input/output, and a sample output in an actual form are included.
Kosugiyama, Shinichi; Takei, Masanobu; Takizuka, Takakazu; Takada, Shoji; Yan, X.; Kunitomi, Kazuhiko
Nihon Genshiryoku Gakkai Wabun Rombunshi, 2(4), p.532 - 545, 2003/12
no abstracts in English
Ueta, Shohei; Emori, Koichi; Tobita, Tsutomu*; Takahashi, Masashi*; Kuroha, Misao; Ishii, Taro*; Sawa, Kazuhiro
JAERI-Research 2003-025, 59 Pages, 2003/11
In the safety design requirements for the High Temperature Engineering Test Reactor (HTTR) fuel, it is determined that "the as-fabricated failure fraction shall be less than 0.2%" and "the additional failure fraction shall be small through the full service period". Therefore the failure fraction should be quantitatively evaluated during the HTTR operation. In order to measure the primary coolant activity, primary coolant radioactivity signals the in safety protection system, the fuel failure detection (FFD) system and the primary coolant sampling system are provided in the HTTR. The fuel and fission product behavior was evaluated based on measured data in the rise-to-power tests (1) to (4). The measured fractional releases are constant at 210 up to 60% of the reactor power, and then increase to 710 at full power operation. The prediction shows good agreement with the measured value. These results showed that the release mechanism varied from recoil to diffusion of the generated fission gas from the contaminated uranium in the fuel compact matrix.
Ueta, Shohei; Tobita, Tsutomu*; Takahashi, Masashi*; Sawa, Kazuhiro
JAERI-Tech 2002-055, 24 Pages, 2002/07
no abstracts in English
Sawa, Kazuhiro; Sumita, Junya; Ueta, Shohei; Takahashi, Masashi; Tobita, Tsutomu*; Hayashi, Kimio; Saito, Takashi; Suzuki, Shuichi*; Yoshimuta, Shigeharu*; Kato, Shigeru*
JAERI-Research 2002-012, 39 Pages, 2002/06
no abstracts in English