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Sono, Hiroki; Izawa, Kazuhiko; Yoritsune, Tsutomu; Suyama, Kenya; Tonoike, Kotaro
Proceedings of 12th International Conference on Nuclear Criticality Safety (ICNC2023) (Internet), 7 Pages, 2023/10
Japan Atomic Energy Agency (JAEA) has constructed and operated nine critical assemblies. Of these nine facilities as of 2023, four have already been dismantled, four are under decommissioning, and only STACY is active but under temporally shutdown. STACY is scheduled to restart in 2024 after core modification from a "critical assembly using uranium nitrate solution fuel" to a "general-purpose critical assembly using uranium fuel rods and light-water moderator." The immediate objective of new STACY is to acquire criticality data for fuel debris removal from the damaged reactors in Fukushima-Daiichi Nuclear Power Plant. After the critical experiment program regarding fuel debris, the new STACY is expected to be used for various R&D on next-generation power reactors and others. In addition, the new STACY will serve as an educational and training reactor. These activities are useful not only for Japan but also for international collaborative research and joint use.
Fukushima, Masahiro; Goda, J.*; Oizumi, Akito; Bounds, J.*; Cutler, T.*; Grove, T.*; Hayes, D.*; Hutchinson, J.*; McKenzie, G.*; McSpaden, A.*; et al.
Nuclear Science and Engineering, 194(2), p.138 - 153, 2020/02
Times Cited Count:7 Percentile:55.95(Nuclear Science & Technology)To validate lead (Pb) nuclear cross sections, a series of integral experiments to measure lead void reactivity worth was conducted systematically in three fast spectra with different fuel compositions on the Comet critical assembly of the National Criticality Experiments Research Center. Previous experiments in a high-enriched uranium (HEU)/Pb and a low-enriched uranium (LEU)/Pb systems had been performed in 2016 and 2017, respectively. A follow-on experiment in a plutonium (Pu)/Pb system has been completed. The Pu/Pb system was constructed using lead plates and weapons grade plutonium plates that had been used in the Zero Power Physics Reactor (ZPPR) of Argonne National Laboratory until the 1990s. Furthermore, the HEU/Pb system was re-examined on the Comet critical assembly installed newly with a device that can guarantee the gap reproducibility with a higher accuracy and precision, and then the experimental data was re evaluated. Using the lead void reactivity worth measured in these three cores with different fuel compositions, the latest nuclear data libraries, JENDL 4.0 and ENDF/B VIII.0, were tested with the Monte Carlo calculation code MCNP version 6.1. As a result, the calculations by ENDF/B-VIII.0 were confirmed to agree with lead void reactivity worth measured in all the cores. It was furthermore found that the calculations by JENDL 4.0 overestimate by more than 20% for the Pu/Pb core while being in good agreements for the HEU/Pb and LEU/Pb cores.
Ono, Ayako; Yamashita, Susumu; Suzuki, Takayuki*; Yoshida, Hiroyuki
Proceedings of 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-18) (USB Flash Drive), p.666 - 677, 2019/08
An evaluation methodology of critical heat fluxes (CHFs) based on a mechanism for fuel assemblies in light water reactors (LWRs) is needed in order to design and evaluate the safety for the fuel assemblies in LWRs. In our study, the numerical simulation with surface-tracking will be applied for the two-phase flow in fuel assemblies in order to obtain the detail data relating to the size and velocity of bubbles in the subchannel, which is needed to predict the CHF based on the mechanism. In this study, the numerical simulation of two-phase flow in 44 bundle was implemented by using JUPITER in order to establish the evaluation method of the size and velocity of bubbles by the numerical simulation, which is the multi-physics simulation code and enable to track the gas-liquid surface. The simulation results are validated by the curve of flow regime for air-water under the adiabatic condition. The bubble and velocity of bubbles obtained by simulation results are analyzed.
Partitioning and Transmutation Technology Division, Nuclear Science and Engineering Center
JAEA-Technology 2017-033, 383 Pages, 2018/02
JAEA is pursuing research and development (R&D) on volume reduction and mitigation of degree of harmfulness of high-level radioactive waste. Construction of Transmutation Experimental Facility (TEF) is under planning as one of the second phase facilities in the Japan Proton Accelerator Complex (J-PARC) program to promote R&D on the transmutation technology with using accelerator driven systems (ADS). The TEF consists of two facilities: ADS Target Test Facility (TEF-T) and Transmutation Physics Experimental Facility (TEF-P). Development of spallation target technology and study on target materials are to be conducted in TEF-T with impinging a high intensity proton beam on a liquid lead-bismuth eutectic target. Whereas in TEF-P, by introducing a proton beam to minor actinide loaded cores, reactor physical properties of the cores are to be studied, and operation experiences of ADS are to be acquired. This report summarizes results of safety design for establishment permit of one of two TEF facilities, TEF-P.
Fukushima, Masahiro; Goda, J.*; Bounds, J.*; Cutler, T.*; Grove, T.*; Hutchinson, J.*; James, M.*; McKenzie, G.*; Sanchez, R.*; Oizumi, Akito; et al.
Nuclear Science and Engineering, 189, p.93 - 99, 2018/01
Times Cited Count:9 Percentile:62.92(Nuclear Science & Technology)To validate lead (Pb) nuclear cross sections, a series of integral experiments to measure lead void reactivity worths was conducted in a high-enriched uranium (HEU)/Pb system and a low enriched uranium (LEU)/Pb system using the Comet Critical Assembly at NCERC. The critical experiments were designed to provide complementary data sets having different sensitivities to scattering cross sections of lead. The larger amount of the U present in the LEU/Pb core increases the neutron importance above 1 MeV compared with the HEU/Pb core. Since removal of lead from the core shifts the neutron spectrum to the higher energy region, positive lead void reactivity worths were observed in the LEU/Pb core while negative values were observed in the HEU/Pb core. Experimental analyses for the lead void reactivity worths were performed with the Monte Carlo calculation code MCNP6.1 together with nuclear data libraries, JENDL 4.0 and ENDF/B VII.1. The calculation values were found to overestimate the experimental ones for the HEU/Pb core while being consistent for the LEU/Pb core.
Iwamoto, Hiroki; Nishihara, Kenji; Yagi, Takahiro*; Pyeon, C.-H.*
Journal of Nuclear Science and Technology, 54(4), p.432 - 443, 2017/04
Times Cited Count:20 Percentile:86.43(Nuclear Science & Technology)Morita, Kenji; Morimoto, Makoto; Hisada, Masaki; Fukui, Yasutaka
JAEA-Technology 2015-037, 28 Pages, 2016/01
Deuterium Critical Assembly (DCA) achieved first critically in 1969 and used for research and development program of Advanced Thermal Reactor. To achieved the aim of facility, DCA decommissioning work started in 2002. Decommissioning schedule consists of 4 stages. The third stage, which is the main work (To dismantle and remove reactor vessel and main equipment), was started in 2008 and will be finished at 2023. This report describes DCA decommissioning work and data (Ability of cutting tools and Man-hours) in 2013.
Oigawa, Hiroyuki; Sasa, Toshinobu; Kikuchi, Kenji; Nishihara, Kenji; Kurata, Yuji; Umeno, Makoto*; Tsujimoto, Kazufumi; Saito, Shigeru; Futakawa, Masatoshi; Mizumoto, Motoharu; et al.
Proceedings of 4th International Workshop on the Utilisation and Reliability of High Power Proton Accelerators, p.507 - 517, 2005/11
Under the framework of J-PARC, the Japan Atomic Energy Research Institute (JAERI) plans to construct the Transmutation Experimental Facility (TEF). The TEF consists of two facilities: the Transmutation Physics Experimental Facility (TEF-P) and the ADS Target Test Facility (TEF-T). The TEF-P is a critical facility which can accept a 600 MeV - 10 W proton beam. The TEF-T is a material irradiation facility using a 600 MeV - 200 kW proton beam, where a Pb-Bi target is installed, but neutron multiplication by nuclear fuel will not be attempted. This report describes the purposes of the facility, the present status of the conceptual design, and the expected experiments to be performed.
Akimoto, Hajime; Tamai, Hidesada; Onuki, Akira; Takase, Kazuyuki
Nihon Konsoryu Gakkai Nenkai Koenkai 2005 Koen Rombunshu, p.229 - 230, 2005/08
A thermal-hydraulic research program for Reduced-Moderation Water Reactor (RMWR) has been performed since 2002. The RMWR has a tight-lattice core to attain the breeding of nuclear fuel for the effective use of Plutonium in a light-water reactor system. In this R&D program, large-scale thermal-hydraulic tests, several model experiments and development of advanced numerical analysis codes are being carried out to confirm the cooling performance in tight-lattice fuel assembly of the RMWR. In this paper, outline of the research program is described as well as the latest results of critical power measurement in the large-scale thermal-hydraulic tests and model experiments, which simulates the tight-lattice core of the RMWR.
Shiroya, Seiji*; Misawa, Tsuyoshi*; Unesaki, Hironobu*; Ichihara, Chihiro*; Kobayashi, Keiji*; Nakamura, Hiroshi*; Shin, Kazuo*; Imanishi, Nobutsugu*; Kanazawa, Satoshi*; Mori, Takamasa
JAERI-Tech 2004-025, 93 Pages, 2004/03
In view of the future plan of Research Reactor Institute, Kyoto University, the present study consisted of (1) the transmission experiments of high energy neutrons through materials, (2) experimental simulation of ADSR using the Kyoto University Critical Assembly(KUCA), and (3) conceptual neutronics design study on KUR type ADSR using the MCNP-X code. Through the present study, valuable knowledge on the basic nuclear characteristics of ADSR, which is indispensable to promote the study on ADSR, was obtained both theoretically and experimentally. For the realization of ADSR, it is considered to be necessary to accumulate results of research steadily. For this purpose, it is inevitable (1) to compile the more precise nuclear data for the wide energy range, (2) to establish experimental techniques for reactor physics study on ADSR including subcriticality measurement and absolute neutron flux measurement, and (3) to develop neutronics calculation tools which take into account the neutron generation process by the spallation reaction and the delayed neutron behavior.
Ando, Masaki; Nakano, Yoshihiro; Okajima, Shigeaki; Kawasaki, Kenji
JAERI-Research 2003-029, 72 Pages, 2003/12
The objectives of this study is to clarify calculation accuracy for the Doppler effect of the resonance materials; erbium (Er), tungsten (W) and thorium (ThO). Doppler effect measurements were carried out in a fast neutron spectrum (XX-2 core) and in an intermediate neutron spectrum (XXI-1D2 core) by the sample-heated and reactivity worth measurement method up to 800
C using FCA. The experiment was analyzed with the standard analysis method for fast reactor cores at FCA with the use of the JENDL-3.2. The SRAC system was also used to investigate the calculation accuracy of the system and to compare it with that of the FCA standard analysis method. The standard analysis method underestimated for the XX-2 core and agreed the experiments within the experimental errors for the XXI-1D2 core. The analysis with the SRAC system gave smaller values by 3%
10% for the Er sample and bigger values by 2%
5% for the W sample than the standard analysis method.
Okuno, Hiroshi
Journal of Nuclear Science and Technology, 40(7), p.544 - 551, 2003/07
Times Cited Count:0 Percentile:0.00(Nuclear Science & Technology)A method for classifying benchmark results of criticality calculations according to similarity was proposed in this paper. After formulation of the method utilizing correlation coefficients, it was applied to burnup credit criticality benchmarks Phase III-A and II-A, which were conducted by the Expert Group on Burnup Credit Criticality Safety under auspices of the OECD/NEA. Phase III-A benchmark was a series of criticality calculations for irradiated BWR fuel assemblies, whereas Phase II-A benchmark was a suite of criticality calculations for irradiated PWR fuel pins. These benchmark problems and their results were summarized. The correlation coefficients were calculated and sets of benchmark results were classified according to the criterion that the correlation coefficients were no less than 0.15 for Phase III-A and 0.10 for Phase II-A benchmarks. When a couple of results were in a same group, one result was found predictable from the other. An example was shown for each of the Benchmarks. The evaluated nuclear data seemed the main factor of errors.
Ando, Masaki; Nakano, Yoshihiro; Okajima, Shigeaki; Kawasaki, Kenji
Journal of Nuclear Materials, 319, p.126 - 130, 2003/06
Times Cited Count:0 Percentile:0.00(Materials Science, Multidisciplinary)Doppler effect experiments on resonance materials for ROX fuels were carried out to examine the calculation accuracy in the intermediate neutron spectrum using Fast Critical Assembly (FCA) at Japan Atomic Energy Research Institute. This study is the second phase of a series of the Doppler effect experiments on the resonance materials, which is following the measurements in the fast neutron spectrum. The Doppler effect was measured as the sample reactivity change between the heated and unheated samples. The cylindrical samples of the resonance materials such as erbium (Er), tungsten (W) and thorium (ThO) were used. The sample was heated up to 800
C at the center of the FCA core. The Doppler effect measurements were analyzed using the SRAC 95 code system with the use of JENDL 3.2. The calculated values agreed with the experiment within the experimental error for the W and ThO
samples, while the calculation overestimated the experiment for the Er sample about 10 %.
Okuno, Hiroshi; Naito, Yoshitaka*; Suyama, Kenya
JAERI-Research 2002-001, 181 Pages, 2002/02
The report describes the final results of the Phase IIIB Benchmark conducted by the Expert Group on Burnup Credit Criticality Safety under the auspices of the OECD/NEA. The Benchmark was intended to compare the predictability of current computer code and data library combinations for the atomic number densities of an irradiated BWR fuel assembly model, which was irradiated under specific power of 25.6 MW/tHM up to 40 GWd/tHM and cooled for five years. The void fraction was assumed to be uniform and constant, at 0, 40 and 70%, during burnup. In total, 16 results were submitted from 13 institutes of 7 countries. The calculated densities of 12 actinides and 20 fission product nuclides were found mostly within a range of +- 10% relative to the average, although some results, esp. 155Eu and gadolinium isotopes, exceeded the band. Pin-wise burnup results agreed well among the participants. The results in the multiplication factor also accorded well with each other for void fractions of 0 and 40%; however some results deviated from the average noticeably for the void fraction of 70%.
Ando, Masaki; Okajima, Shigeaki; Oigawa, Hiroyuki
JAERI-Research 2001-017, 20 Pages, 2001/03
U Doppler effect measurements were carried out using 6 uranium samples with various composition or diameter varying self-shielding factors of
U. Calculation accuracy of the standard analysis method for fast reactor cores at FCA was investigated.The Doppler effect were measured by the sample-heated and reactivity worth measurement method up to 800 degree in the FCA XVII-1 assembly, which simulated Na-cooled MOX fueled FBR. In the analysis, the Doppler effect was calculated using the effective cross sections obtained by the PEACO-X code. The calculated values based on the diffusion theory agreed well with the experiment for the U-metal and UO
samples. The calculation underestimated the experiment from 10% to 30% for the UO
samples containing BeO. This underestimation was improved by the use of the transport calculation. The calculation, however, underestimated about 20% for the sample whose background cross section is greater than 300 barns.
Ogawa, Hironobu; Mukaiyama, Takehiko
JAERI-Tech 99-041, 188 Pages, 1999/05
no abstracts in English
Akino, Fujiyoshi; Yamane, Tsuyoshi
Nihon Genshiryoku Gakkai-Shi, 40(4), p.262 - 264, 1998/00
no abstracts in English
Okuno, Hiroshi;
Criticality Safety Challenges in the Next Decade, 0, p.150 - 155, 1997/00
no abstracts in English
Yamane, Tsuyoshi; Akino, Fujiyoshi;
PHYSOR 96: Int. Conf. on the Physics of Reactors, 2, p.E290 - E299, 1996/00
no abstracts in English
Ogawa, Hironobu; ; Mukaiyama, Takehiko
JAERI-Research 94-026, 199 Pages, 1994/11
no abstracts in English