Journal of Nuclear Science and Technology, 14 Pages, 2022/04
The reactivity was estimated from a time profile of neutron count rate or a simulated data in a quasi-steady state after sudden change of reactivity or external neutron source strength. The estimation was based on the equation of power in subcritical quasi-steady state. The purpose of the study is to develop the method of timely reactivity estimation from complicated time profile of neutron count rate. The developed method was applied to the data simulating neutron count rate created by using one-point kinetics code, AGNES, and Poisson-distributed random noise and to the transient subcritical experiment data measured by using TRACY. The result shows that the difference of the estimated and reference value was within about 5% or less for (simeq) -1.4 and -3.1 for the experimental data. It was also shown that the possibility of the reactivity estimation several ten seconds after the status change.
Collaborative Laboratories for Advanced Decommissioning Science; Tokyo Institute of Technology*
JAEA-Review 2021-037, 61 Pages, 2022/01
The Collaborative Laboratories for Advanced Decommissioning Science (CLADS), Japan Atomic Energy Agency (JAEA), had been conducting the Nuclear Energy Science & Technology and Human Resource Development Project (hereafter referred to "the Project") in FY2020. The Project aims to contribute to solving problems in the nuclear energy field represented by the decommissioning of the Fukushima Daiichi Nuclear Power Station (1F), Tokyo Electric Power Company Holdings, Inc. (TEPCO). For this purpose, intelligence was collected from all over the world, and basic research and human resource development were promoted by closely integrating/collaborating knowledge and experiences in various fields beyond the barrier of conventional organizations and research fields. The sponsor of the Project was moved from the Ministry of Education, Culture, Sports, Science and Technology to JAEA since the newly adopted proposals in FY2018. On this occasion, JAEA constructed a new research system where JAEA-academia collaboration is reinforced and medium-to-long term research/development and human resource development contributing to the decommissioning are stably and consecutively implemented. Among the adopted proposals in FY2019, this report summarizes the research results of the "Improvement of critical safety technology in fuel debris retrieval" conducted in FY2019 and FY2020. Since the final year of this proposal was FY2020, the results for two fiscal years were summarized. The purpose of research was to improve the criticality safety analysis methods in the case of fuel debris removal with the collaboration with Russian university, which has a lot of experiences in the criticality analysis. This research has been performed as two fiscal years project in FY 2019 and FY 2020 by Tokyo Institute of Technology (Tokyo Tech) and Tokyo City University (TCU) as the Japanese side, and National Research Nuclear University MEPhI as the Russian side. In FY2019, Tokyo Tech introduced a GPU server
JAEA-Technology 2021-023, 190 Pages, 2021/11
Computational analyses on nuclear criticality characteristics were carried out for heterogeneous lattice systems composed of water moderator and fuel rods utilized in low-power research and test reactors, in which the depletion of fuel due to burnup is relatively small, by using the continuous-energy Monte Carlo code MVP Version 2 with the evaluated nuclear data library JENDL-4.0. In the analyses, the minimum critical number of fuel rods was evaluated using calculated neutron multiplication factors for the heterogeneous systems of the uranium dioxide fuel rod in the Static Experiments Critical Facility (STACY) and the Tank-type Critical Assembly (TCA), and the uranium-zirconium hydride fuel rod in the Nuclear Safety Research Reactor (NSRR). In addition, six sorts of the ratio of reaction rates, which are components of neutron multiplication factors, were calculated in the analyses to explain the variation of neutron multiplication factors with the ratio of water moderator to fuel volume in a unit fuel rod cell. Those results of analyses are considered to be useful for the confirmation of reasonableness and validity of criticality safety measures as data showing criticality characteristics for water-moderated heterogeneous lattice systems composed of the existing fuel rods in research and test reactors, of which criticality data are not sufficiently provided by the Criticality Safety Handbook.
Collaborative Laboratories for Advanced Decommissioning Science; Tokyo Institute of Technology*
JAEA-Review 2020-041, 30 Pages, 2020/12
JAEA/CLADS had been conducting the Nuclear Energy Science & Technology and Human Resource Development Project in FY2019. Among the adopted proposals in FY2019, this report summarizes the research results of the "Improvement of Critical Safety Technology in Fuel Debris Retrieval" conducted in FY2019.
Journal of Nuclear Science and Technology, 57(8), p.926 - 931, 2020/08
An equation of power in subcritical quasi-steady state has been derived based on one-point kinetics equations for the purpose of utilizing it for the development of timely reactivity estimation from complicated time profile of neutron count rate. It linearly relates power, , to a new variable , which is a function of time differential of the power. It has been confirmed by using one-point kinetics code, AGNES, that the calculated points () are perfectly in a line described by the new equation and that points () calculated from transient subcritical experiments by using TRACY made a line with a slope indicated by the new equation.
Okuno, Hiroshi; Suyama, Kenya; Ryufuku, Susumu*
JAEA-Review 2017-010, 93 Pages, 2017/06
There is an ongoing discussion on the application of burnup credit to the criticality safety controls of facilities that treat spent fuels. With regard to such application of burnup credit in Japan, this document summarizes the current technical status of the prediction of the isotopic composition and criticality of spent fuels, as well as safety evaluation concerns and the current status of legal affairs. This report is an English translation of A Guide to Introducing Burnup Credit, Preliminary Version, originally published in Japanese as JAERI-Tech 2001-055 by the Nuclear Fuel Cycle Facility Safety Research Committee.
Nauchi, Yasushi*; Takezawa, Hiroki*; Tonoike, Kotaro
Nihon Genshiryoku Gakkai-Shi ATOMO, 58(4), p.247 - 252, 2016/04
no abstracts in English
Yamamoto, Kento*; Akie, Hiroshi; Suyama, Kenya; Hosoyamada, Ryuji*
JAEA-Technology 2015-019, 110 Pages, 2015/10
In the direct disposal of used nuclear fuel (UNF), criticality safety evaluation is one of the important issues since UNF contains some amount of fissile material. The recent development of higher-enrichment fuel has enhanced the benefit of the application of Burnup Credit. In the present study, the effects of the several parameters on the reactivity of disposal canister model were evaluated for used PWR fuel. The parameters are relevant to the uncertainties of depletion calculation code, irradiation history, and axial and horizontal burnup distribution, which are known to be important effect in the criticality safety evaluation adopting burnup credit. The latest data or methodology was adopted in this evaluation, based on the various latest studies. The appropriate margin of neutron multiplication factor in the criticality safety evaluation for UNF can be determined by adopting the methodology described in the present study.
Nakajima, Ken*; Itahara, Kuniyuki*; Okuno, Hiroshi
Proceedings of International Conference on Nuclear Criticality Safety (ICNC 2015) (DVD-ROM), p.496 - 502, 2015/09
An outline of the standard "Procedures for Applying Burnup Credit to Criticality Safety Control of a Reprocessing Facility: 2014" (AESJ-SC-F025: 2014) published in April 2015 by the Atomic Energy Society of Japan (AESJ) is presented. The AESJ published more than 60 Standards. However, many of them were in the field of nuclear power reactors or radioactive wastes. Ten years ago the AESJ published "Basic Items of Criticality Safety Control: 2004" (AESJ-SC-F004:2004), which prescribed basic ideas, requirements and methods on nuclear criticality safety controls of facilities handling with nuclear fuel materials in general for preventing a nuclear criticality accident. However, it did not include any specific procedures for adopting burnup credit. Therefore, a new standard was envisaged as the first Standard for fuel reprocessing plants, which clarified the specific procedures to apply burnup credit to designers, operators, maintenance persons and administrators.
Tonoike, Kotaro; Yamane, Yuichi; Umeda, Miki; Izawa, Kazuhiko; Sono, Hiroki
Proceedings of International Conference on Nuclear Criticality Safety (ICNC 2015) (DVD-ROM), p.20 - 27, 2015/09
From the viewpoint of safety regulation, criticality control of the fuel debris in the Fukushima Daiichi Nuclear Power Station would be a risk-informed control to mitigate consequences of criticality events, instead of a deterministic control to prevent such events. The Nuclear Regulation Authority of Japan has set up a research and development program to tackle this challenge. The Nuclear Safety Research Center of Japan Atomic Energy Agency, commissioned by the authority, has launched activities such as computations of criticality characteristics of the fuel debris, development of criticality risk assessment method, and preparation of criticality experiments to support them.
International Symposium NUCEF 2005 Working Group
JAERI-Conf 2005-007, 359 Pages, 2005/08
Japan Atomic Energy Research Institute (JAERI) held the international symposium NUCEF2005 at Techno Community Square RICOTTI in Tokai-mura on February 9 and 10, 2005. This symposium was co-organized by Japan Nuclear Cycle Development Institute (JNC), and Nuclear Fuel Cycle Safety Research Committee authorized the program. Two hundred thirty-nine participants from 11 countries presented fifty-nine papers, and discussed recent research activities and its outputs on waste disposal safety, fuel cycle facility safety including criticality safety, and separation process development. The presented papers are compiled in the proceedings.
Abe, Hitoshi; Tashiro, Shinsuke; Morita, Yasuji
JAERI-Conf 2005-007, p.199 - 204, 2005/08
Source term data for estimating release behavior of radioactive nuclides is necessary to evaluate synthetic safety of nuclear fuel cycle facility under accident conditions, such as fire and criticality. In JAERI, the data has been obtained by performing some demonstration tests. In this paper, the data for the criticality accident in fuel solution obtained from the TRACY experiment, will be mainly reviewed. At 4.5 h after the transient criticality, the release ratio of the iodine were about 0.2% for re-insertion of transient rod at just after transient criticality and about 0.9% for not re-insertion. Similarly the release coefficient and release ratio for Xe were estimated. It was proved that the release ratio of Xe-141 from the solution was over 90% in case that the inverse period was over about 100 (1/s). Furthermore, outline of the study on the fire accident as future plan will be also mentioned.
Yamamoto, Toshihiro; Miyoshi, Yoshinori
Transactions of the American Nuclear Society, 91, p.583 - 584, 2004/11
MOX powder and additives are mixed in the process of MOX fuel fabrication. A non-uniform mixing state of MOX powder and additives occurs during the homogenization mixing process. However, ordinary criticalit safety evaluations for mixtures assume that the mixtures have a uniform distribution of the mixing state. A non-uniform distribution of the mixing state in a sphere, which maximizes the effective neutron multiplication factor, was obtained using a concept of the fuel importance. As a result, the central portion of the sphere is composed of an optimal moderation region, and the surrounding region is composed of pure MOX powder. While keff is 0.545 for the uniform distribution, keff for the optimal non-uniform distribution is 0.590. That is, keff increases by 0.045.
Abe, Hitoshi; Tashiro, Shinsuke; Morita, Yasuji
JAERI-Research 2004-014, 19 Pages, 2004/09
no abstracts in English
Okuno, Hiroshi; Takada, Tomoyuki
Journal of Nuclear Science and Technology, 41(4), p.481 - 492, 2004/04
Nuclear characteristic parameters were calculated and subcriticality judgement graphs were drawn for revision purposes of the Data Collection for the Nuclear Criticality Safety Handbook. The nuclear characteristic parameters were the neutron multiplication factor in infinite media, migration area and diffusion constants for 11 kinds of typical fuels encountered in criticality safety evaluation of nuclear fuel cycle facilities. These fuels included ADU-HO, UF6-HF and Pu(NO)-UO(NO) solution, of which data were not cited in the Data Collection. The calculation was made with the Japanese evaluated nuclear data library JENDL-3.2 and a sequence of criticality calculation codes, SRAC, POST and SIMCRI. The subcriticality judgement graphs that depict the region satisfying the inequality relation of the neutron multiplication factor less than 0.98 between the two variables (a) uranium enrichment, 239Pu/Pu ratio or plutonium enrichment and (b) H/(Pu+U) ratio were drawn for the same kinds of fuels except UF6-HF in infinite media.
Okuno, Hiroshi; Ryufuku, Susumu*; Suyama, Kenya; Nomura, Yasushi; Tonoike, Kotaro; Miyoshi, Yoshinori
JAERI-Conf 2003-019, p.116 - 121, 2003/10
This paper outlines the data prepared for the 2nd version of Data Collection of the Nuclear Criticality Safety Handbook. These data are discussed in the order of its preliminary table of contents. The nuclear characteristic parameters (k, M, D) were derived, and subcriticality judgment graphs were drawn for eleven kinds of fuels which were often encountered in criticality safety evaluation of fuel cycle facilities. For calculation of criticality data, benchmark calculations using the combination of the continuous energy Monte Carlo criticality code MVP and the Japanese Evaluated Nuclear Data Library JENDL-3.2 were made. The calculation errors were evaluated for this combination. The implementation of the experimental results obtained by using NUCEF facilities into the 2nd version of the Data Collection is under discussion. Therefore, related data were just mentioned. A database is being prepared to retrieve revised data easily.
Cao, X.; Suzaki, Takenori; Kugo, Teruhiko; Mori, Takamasa
JAERI-Tech 2003-069, 36 Pages, 2003/08
From the viewpoint of nuclear criticality safety of fuel rod storage and transport, a series of critical experiments concerning effects of water hole size, water gap width, water-to-fuel volume ratio and non-uniform arrangement of water moderator have been performed at the Tank-type Critical Assembly (TCA) of Japan Atomic Energy Research Institute. In the present study, the effects of volume fraction and non-uniform arrangement of water moderator on reactivity are evaluated by the water level worth method and analyzed by the SRAC code. Error sources of experiments and calculations are discussed, especially for an energy group model. The calculation results of diffusion model with 17-group model show good agreement with the experiment results within a few dozen cents.
Takada, Tomoyuki; Miyoshi, Yoshinori; Katakura, Junichi
JAERI-Tech 2003-036, 80 Pages, 2003/03
In order to perform accuracy evaluation of the critical calculation by the combination of multi-group constant library MGCL and 3-dimensional Monte Carlo code KENO-IV among critical safety evaluation code system JACS, benchmark calculation was carried out from 1980 in 1982. Some cases where the neutron multiplication factor calculated in the heterogeneous system in it was less than 0.95 were seen. In this report, it re-calculated by considering the cause about the heterogeneous system of the U+Pu nitric acid solution systems containing the neutron poison shown in JAERI-M 9859. The present study has shown that the keff value less than 0.95 given in JAERI-M 9859 is caused by the fact that the water reflector below a cylindrical container was not taken into consideration in the KENO-IV calculation model. By taking into the water reflector, the KENO-IV calculation gives a keff value greater than 0.95 and a good agreement with the experiment.
Umeda, Miki; Nakazaki, Masato; Kida, Takashi; Sato, Kenji; Kato, Tadahito; Kihara, Takehiro; Sugikawa, Susumu
JAERI-Tech 2003-024, 23 Pages, 2003/03
MOX dissolution with silver mediated electrolytic oxidation method is planned for the preparation of plutonium nitrate solution to be used for criticality safety experiments at Nuclear Fuel Cycle Safety Engineering Research Facility (NUCEF). Silver mediated electrolytic oxidation method uses the strong oxidisation ability of Ag(II) ion. This method is thought to be effective for the dissolution of MOX, which is difficult to be dissolved with nitric acid.In this paper, the results of experiments on dissolution with 100 g of MOX are described. It was confirmed by the results that the MOX powder to be used at NUCEF was completely dissolved by silver mediated electrolytic oxidation method and that Pu(VI) ion in the obtained solution was reduced to tetravalent by means of NO purging.
Kuroishi, Takeshi; Hoang, A.; Nomura, Yasushi; Okuno, Hiroshi
JAERI-Tech 2003-021, 60 Pages, 2003/03
The reactivity effect of the asymmetry of axial burnup profile is studied for PWR spent fuel transport cask proposed in OECD/NEA Phase II-C benchmark. The axial burnup profiles are based on in-core flux measurements. Criticality calculations are performed with the continuous energy Monte Carlo code MCNP-4B2 and the nuclear data library JENDL-3.2. Calculations are carried out not only for cases in the benchmark but also for symmetric burnup cases. Both actinide-only approach and actinide plus fission product approach is considered. The end effect is more sensitive to higher burnup asymmetry. The axial fission distribution becomes strongly asymmetric as its peak shifts toward the fuel top end. The peak of fission distribution gets higher with the increase of either the burnup asymmetry or the assembly-averaged burnup. The conservatism of uniform axial burnup assumption for the actinide-only approach is estimated quantitatively in comparison with the keff result calculated with experiment-based strongest asymmetric axial burnup profile for the actinide plus fission product approach.