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Mohamad, A. B.; Udagawa, Yutaka
Nuclear Technology, 210(2), p.245 - 260, 2024/02
Times Cited Count：1 Percentile：68.31(Nuclear Science & Technology)Takino, Kazuo; Oki, Shigeo
JAEA-Data/Code 2023-003, 26 Pages, 2023/05
JAEA-Data-Code-2023-003.pdf:1.66MB
Since next-generation fast reactors aim to achieve a higher core discharge burn-up than conventional reactors do, core neutronics design methods must be refined. Therefore, a suitable analysis condition is required for the analysis of burn-up nuclear characteristics to accomplish sufficient estimation accuracy while maintaining a low computational cost. We investigated the effect of the analysis conditions on the accuracy of estimation of the burn-up nuclear characteristics of next-generation fast reactors in terms of neutron energy groups, neutron transport theory, and spatial mesh. This study treated the following burn-up nuclear characteristics: criticality, burn-up reactivity, control rod worth, breeding ratio, assembly-wise power distribution, maximum linear heat rate, sodium void reactivity, and Doppler coefficient for the equilibrium operation cycle. As a result, it was found that the following conditions were the most suitable: 18-energy-group structure, 6 spatial meshes per assembly with diffusion approximation. Additionally, these conditions should apply to correction factors for energy group structure, spatial mesh and transport effects.
Goto, Minoru; Aihara, Jun; Inaba, Yoshitomo; Ueta, Shohei; Fukaya, Yuji; Okamoto, Koji*
Proceedings of 9th International Topical Meeting on High Temperature Reactor Technology (HTR 2018) (USB Flash Drive), 6 Pages, 2018/10
JAEA has conducted design studies of a Pu-burner HTGR. The Pu-burner HTGR incinerates Pu by fission, and hence a high burn-up is required for the efficient incineration. In the fuel design, a thin ZrC layer, which acts as an oxygen getter and suppresses the internal pressure, was coated on the fuel kernel to prevent the CFP failure at the high burn-up. A stress analysis of the SiC layer, which acts as a pressure vessel for the CFP, was performed for with consideration of the depression effect due to the ZrC layer. As a result, the CFP failure fraction at high burn-up of 500 GWd/t satisfied the target value. In the reactor core design, an axial fuel shuffling was employed to attain the high burn-up, and the nuclear burn-up calculations with the whole core model and the fuel temperature calculations were performed. As a result, the nuclear characteristics, which are the shutdown margin and the temperature coefficient of reactivity, and the fuel temperature satisfied their target values.
Kora, Kazuki*; Nakaya, Hiroyuki*; Matsuura, Hideaki*; Goto, Minoru; Nakagawa, Shigeaki; Shimakawa, Satoshi*
Nuclear Engineering and Design, 300, p.330 - 338, 2016/04
Times Cited Count：6 Percentile：49.05(Nuclear Science & Technology)In order to investigate the potential of high temperature gas-cooled reactors (HTGRs) for transmutation of long-lived fission products (LLFPs), numerical simulation of four types of HTGRs were carried out. In addition to the gas-turbine high temperature reactor system "GTHTR300", a small modular HTGR plant "HTR50S" and two types of plutonium burner HTGRs "Clean Burn with MA" and "Clean Burn without MA" were considered. The simulation results show that an early realization of LLFP transmutation using a compact HTGR may be possible since the HTR50S can transmute fair amount of LLFPs for its thermal output. The Clean Burn with MA can transmute a limited amount of LLFPs. However, an efficient LLFP transmutation using the Clean Burn without MA seems to be convincing as it is able to achieve very high burn-ups and produce LLFP transmutation more than GTHTR300. Based on these results, we propose utilization of variety of HTGRs for LLFP transmutation and storage.
Okumura, Keisuke
JAEA-Data/Code 2015-015, 162 Pages, 2015/10
JAEA-Data-Code-2015-015.pdf:3.99MB
JAEA-Data-Code-2015-015-appendix(CD-ROM).zip:3.38MB
MOSRA-SRAC is a lattice calculation module of the Modular code System for nuclear Reactor Analyses (MOSRA). This module performs the neutron transport calculation for various types of fuel elements including existing light water reactors, research reactors, etc. based on the collision probability method with a set of the 200-group cross-sections generated from the Japanese Evaluated Nuclear Data Library JENDL-4.0. It has also a function of the isotope generation and depletion calculation for up to 234 nuclides in each fuel material in the lattice. In these ways, MOSRA-SRAC prepares the burn-up dependent effective microscopic and macroscopic cross-section data to be used in core calculations.
Harada, Masahide; Watanabe, Noboru; Teshigawara, Makoto; Kai, Tetsuya; Maekawa, Fujio; Kato, Takashi; Ikeda, Yujiro
LA-UR-06-3904, Vol.2, p.700 - 709, 2006/06
Pulse characteristics data for every neutron beam line are indispensable in designing devices for neutron scattering experiments of JSNS. A detailed model was built and pulse characteristics of each beam line were estimated using the PHITS code and the MCNP-4C code. These results have been disclosed on the J-PARC homepage since September 2004. Due to changes of moderator shapes in a progress of manufacture design, we observed from the calculation that pulse structures of decoupled moderators were deteriorated, especially, those of pulse tail. We found that this deterioration was caused by leakage neutron from gaps between decouplers and absorbing liners of the reflector. For a final stage of the manufacture design, we carefully tried to find other factors which deteriorated the pulse characteristics. Furthermore, pulse structures of poisoned and unpoisoned decoupled moderators were evaluated with the consideration of heterogeneous burn-up and leakage neutron spectra including high-energy region up to GeV were estimated for each neutron beam hole.
Okumura, Keisuke; Oki, Shigeo*; Yamamoto, Munenari*; Matsumoto, Hideki*; Ando, Yoshihira*; Tsujimoto, Kazufumi; Sasahara, Akihiro*; Katakura, Junichi; Matsumura, Tetsuo*; Aoyama, Takafumi*; et al.
JAERI-Research 2004-025, 154 Pages, 2005/01
JAERI-Research-2004-025.pdf:19.46MB
This report summarizes the activity (FY2000-2003) of Working Group (WG) on Evaluation of Nuclide Generation and Depletion under Subcommittee on Nuclear Fuel Cycle of Japanese Nuclear Data Committee. In the WG, analyses of Post Irradiation Examinations have been carried out for UO
and MOX fuels irradiated in PWRs, BWRs and FBRs, and for actinide samples irradiated in fast reactors, by using ORIGEN or more detailed calculation codes with their libraries based on JENDL-3.2, JENDL-3.3 and other foreign nuclear data files. From these results, current prediction accuracy and problems for evaluation of nuclide generation and depletion are discussed. Furthermore, this report covers other products of our activity; development of the ORIGEN libraries for PWR, BWR and FBR based on JENDL-3.3, study on introduction of neutron spectrum index to ORIGEN calculations, and results of questionnaire survey on desirable accuracy of ORIGEN calculations.
Nishitani, Takeo; Osakabe, Masaki*; Shinohara, Koji; Ishikawa, Masao
Purazuma, Kaku Yugo Gakkai-Shi, 80(10), p.860 - 869, 2004/10
no abstracts in English
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 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.
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 pellet.
Research Committee on Reactor Physics
JAERI-Research 2004-004, 409 Pages, 2004/03
JAERI-Research-2004-004.pdf:28.53MB
This report summarizes the second phase (FY2001-2002) activity of "the Working Party (WP) on Reactor Physics for LWR Next Generation Fuels". The next generation fuels mean the ones aiming at further extended burn-up such as 70GWd/t over the current design. In the WP, the benchmark activity has been conducted to investigate and improve the calculation accuracy of the nuclear characteristics of the next generation fuels. In the second phase activity, all benchmark results were compiled and compared. Based on the comparison, the present status of calculation accuracy for the next generation fuels has been confirmed, and the factors causing the calculation differences were analyzed in detail. Moreover, analyses of the post irradiation and critical experiments with the codes used in the benchmark were reviewed, and future experiments and research subjects necessary to reduce the calculation differences were discussed and proposed.
Suzuki, Motoe; Saito, Hiroaki*
JAERI-Data/Code 2003-019, 423 Pages, 2003/12
JAERI-Data-Code-2003-019.pdf:17.7MB
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.
Shelley, A.; Shimada, Shoichiro*; Kugo, Teruhiko; Okubo, Tsutomu; Iwamura, Takamichi
Nuclear Engineering and Design, 224(3), p.265 - 278, 2003/10
Times Cited Count：15 Percentile：66.87(Nuclear Science & Technology)Parametric studies have been done for a PWR-type reduced-moderation water reactor (RMWR) with seed-blanket fuel assembles to achieve a high conversion ratio, negative void reactivity coefficient and a high burnup. It was found that 50 to 60% of seed in a seed-blanket assembly has higher conversion ratio. The number of seed-blanket layers is 20, in which the number of seed layers is 15 and blanket layers is 5. The fuel assembly with the height of seed of 1000mm
2, internal blanket of 150 mm and axial blanket of 400mm2 is recommended. The conversion ratio is 1.0 and the average burnup in core region is 38.2 GWd/t. The enrichment of fissile Pu is 14.6 wt%. The void coefficient is +21.8 pcm/% void, however, it is expected that the void coefficient will be negative if the radial neutron leakage is taken into account. It is also possible to use this fuel assembly for a high core averaged burnup of 45GWd/t, however, the height of seed must be 500mm2 to improve the void coefficient. The conversion ratio is 0.97 and void coefficient is +20.8 pcm/%void.
Iwamura, Takamichi; Okubo, Tsutomu; Kureta, Masatoshi; Nakatsuka, Toru; Takeda, Renzo*; Yamamoto, Kazuhiko*
Proceedings of 13th Pacific Basin Nuclear Conference (PBNC 2002) (CD-ROM), 7 Pages, 2002/10
In order to ensure sustainable energy supply in Japan, the reduced-moderation water reactor (RMWR) has been developed by JAERI since 1998. MOX fuel assemblies with tight lattice arrangement are used to increase the conversion ratio. In order to establish negative void reactivity coefficient, the core should be short and flat to increase neutron leakage from the core. The core designs were accomplished to a large core with 1,356MWe and a small core with 330MWe. For both cores, negative void coefficient and natural circulation cooling of the core were realized. To confirm thermal-hydraulic feasibility, critical heat flux experiments were performed using 7-rod bundles with the gap width of 1mm and 1.3mm. The results indicated that enough cooling was assured for the tight lattice core. Further R&D studies, including large scale thermal-hydraulic experiments, reactor physics experiments, development of high burn-up fuel cladding material and simplified reprocessing technology, are necessary to realize commercial introduction of RMWR by 2020's for the replacement of current generation LWRs.
pelletHarada, Katsuya; Nakata, Masahito; Yasuda, Ryo; Nishino, Yasuharu; Amano, Hidetoshi
HPR-356, 11 Pages, 2001/00
no abstracts in English
Okumura, Keisuke; Mori, Takamasa; Nakakawa, Masayuki; Kaneko, Kunio*
Journal of Nuclear Science and Technology, 37(2), p.128 - 138, 2000/02
no abstracts in English
Mori, Takamasa; Okumura, Keisuke; Nagaya, Yasunobu; Ando, Hiroei
Transactions of the American Nuclear Society, 83, p.283 - 284, 2000/00
no abstracts in English
Yanagisawa, Kazuaki; Yamashita, Toshiyuki; Kanazawa, Hiroyuki; Amano, Hidetoshi; Muromura, Tadasumi
Journal of Nuclear Science and Technology, 36(12), p.1153 - 1159, 1999/12
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)no abstracts in English
Naka, Michihiro; Nagao, Yoshiharu; Komukai, Bunsaku; Tabata, Toshio
Proceedings of 7th Meeting of the International Group on Research Reactors (IGORR-7) (CD-ROM), 7 Pages, 1999/10
no abstracts in English
Mori, Takamasa; Okumura, Keisuke; Nagaya, Yasunobu; Nakakawa, Masayuki
Mathematics and Computation, Reactor Physics and Environmental Analysis in Nuclear Applications, 2, p.987 - 996, 1999/09
no abstracts in English
