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Fukuda, Kodai; Yamane, Yuichi
Journal of Nuclear Science and Technology, 60(12), p.1514 - 1525, 2023/12
Times Cited Count:1 Percentile:72.91(Nuclear Science & Technology)This study aims to clarify the effect of fuel particle radius on the criticality transient behavior and the total number of fissions in water-moderated solid fuel dispersion systems. Neutronics/thermal hydraulics-coupled kinetics analysis was performed in a hypothetical fuel debris system, where small fuel particles aggregate in water and become supercritical. Results showed that the number of fissions is 10 times larger when the fuel particle radius is reduced by one order of magnitude under conditions where heat transfer, i.e. from fuel to water, is emphasized. Moreover, there is a possibility that lower reactivity could give a larger number of fissions when the fuel particle size is very small. In addition, the number of fissions may be overestimated or underestimated to an unexpected extent unless appropriate fuel particle size is set on the analysis.
Yamane, Yuichi
Journal of Nuclear Science and Technology, 59(11), p.1331 - 1344, 2022/11
Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)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 ( -1) for simulated data and within about 7% or less for -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.
Nakamura, Hideo; Bentaib, A.*; Herranz, L. E.*; Ruyer, P.*; Mascari, F.*; Jacquemain, D.*; Adorni, M.*
Proceedings of International Conference on Topical Issues in Nuclear Installation Safety; Strengthening Safety of Evolutionary and Innovative Reactor Designs (TIC 2022) (Internet), 10 Pages, 2022/10
Yokoyama, Kenji; Maruyama, Shuhei; Taninaka, Hiroshi; Oki, Shigeo
JAEA-Data/Code 2021-019, 115 Pages, 2022/03
In JAEA, several versions of unified cross-section set for fast reactors have been developed so far; we have developed a new unified cross-section set ADJ2017R, which is an improved version of the unified cross-section setADJ2017 for fast reactors. The unified cross-section set is used for reflecting information of C/E values (analysis / experiment values) obtained by integral experiment analyses in reactor core design via the cross-section adjustment methodology; the values are stored in the standard database for FBR core design. In the methodology, the cross-section set is adjusted by integrating the information such as uncertainty (covariance) of nuclear data, uncertainty of integral experiment / analysis, sensitivity of integral experiment with respect to nuclear data. ADJ2017R basically has the same performance as ADJ2017, but we conducted an additional investigation on ADJ2017 and revised the following two points. The first is to unify the evaluation method of the correlation coefficient of uncertainty caused by experiments (hereinafter referred to as the experimental correlation coefficient). Because it was found that the common uncertainty used in the evaluation of the experimental correlation coefficient was evaluated by two different methods, the experimental correlation coefficients were revised for all experimental data, and the evaluation method was unified. The second is the review of the integral experiment data used for the cross-section adjustment calculation. It was found that one of the experimental values of composition ratio after irradiation of the Am-243 sample has a problem in uncertainty evaluation because its experimental uncertainty is extremely small compared to the others. The cross-section adjustment calculation was, therefore, redone by excluding the experimental value. In the creation of ADJ2017, a total of 719 data sets were analyzed and evaluated, and eventually adopted 620 integral experimental data sets. In contrast, a total of 61
Takino, Kazuo; Sugino, Kazuteru; Oki, Shigeo
Annals of Nuclear Energy, 162, p.108454_1 - 108454_7, 2021/11
Times Cited Count:1 Percentile:16.35(Nuclear Science & Technology)Sato, Ikken
Nuclear Engineering and Design, 383, p.111426_1 - 111426_19, 2021/11
Times Cited Count:5 Percentile:65.59(Nuclear Science & Technology)Takeda, Tetsuaki*; Inagaki, Yoshiyuki; Aihara, Jun; Aoki, Takeshi; Fujiwara, Yusuke; Fukaya, Yuji; Goto, Minoru; Ho, H. Q.; Iigaki, Kazuhiko; Imai, Yoshiyuki; et al.
High Temperature Gas-Cooled Reactors; JSME Series in Thermal and Nuclear Power Generation, Vol.5, 464 Pages, 2021/02
As a general overview of the research and development of a High Temperature Gas-cooled Reactor (HTGR) in JAEA, this book describes the achievements by the High Temperature Engineering Test Reactor (HTTR) on the designs, key component technologies such as fuel, reactor internals, high temperature components, etc., and operational experience such as rise-to-power tests, high temperature operation at 950C, safety demonstration tests, etc. In addition, based on the knowledge of the HTTR, the development of designs and component technologies such as high performance fuel, helium gas turbine and hydrogen production by IS process for commercial HTGRs are described. These results are very useful for the future development of HTGRs. This book is published as one of a series of technical books on fossil fuel and nuclear energy systems by the Power Energy Systems Division of the Japan Society of Mechanical Engineers.
Okita, Shoichiro; Fukaya, Yuji; Goto, Minoru
Journal of Nuclear Science and Technology, 58(1), p.9 - 16, 2021/01
Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)Suppressing the kernel migration rates, which depend on both the fuel temperature and the fuel temperature gradient, under normal operation condition is quite important from the viewpoint of the fuel integrity for High Temperature Gas-cooled Reactors. The presence of the ideal axial power distribution to minimize the maximum kernel migration rate allows us to improve efficiency of design work. Therefore, we propose a new method based on Lagrange multiplier method in consideration of thermohydraulic design in order to obtain the ideal axial power distribution to minimize the maximum kernel migration rate. For one of the existing conceptual designs performed by JAEA, the maximum kernel migration rate for the power distribution to minimize the maximum kernel migration rate proposed in this study is lower by approximately 10% than that for the power distribution as a conventional design target to minimize the maximum fuel temperature.
Yamane, Yuichi
Journal of Nuclear Science and Technology, 57(8), p.926 - 931, 2020/08
Times Cited Count:1 Percentile:12.16(Nuclear Science & Technology)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.
Miwa, Shuhei; Takase, Gaku; Imoto, Jumpei; Nishioka, Shunichiro; Miyahara, Naoya; Osaka, Masahiko
Journal of Nuclear Science and Technology, 57(3), p.291 - 300, 2020/03
Times Cited Count:7 Percentile:55.67(Nuclear Science & Technology)For the evaluation of transport behavior of control material boron in a severe accident of BWR from the viewpoint of chemical effects on cesium and iodine behavior, boron chemistry during transportation in the high temperature region above 400 K was experimentally investigated. The heating tests of boron oxide specimen were conducted using the dedicated experimental apparatus reproducing fission product release and transport in steam atmosphere. Released boron oxide vapor was deposited above 1,000 K by the condensation onto stainless steel. The boron deposits and/or vapors significantly reacted with stainless steel above 1,000 K and formed the stable iron-boron mixed oxide (FeO)BO. These results indicate that released boron from degraded BWR control blade in a severe accident could remain in the high temperature region such as a Reactor Pressure Vessel. Based on these results, it can be said that the existence of boron deposits in the high temperature region would decrease the amount of transported cesium vapors from a Reactor Pressure Vessel due to possible formation of low volatile cesium borate compounds by the reaction of boron deposits with cesium vapors.
Hidaka, Akihide
Journal of Nuclear Science and Technology, 56(9-10), p.831 - 841, 2019/09
Times Cited Count:12 Percentile:80.27(Nuclear Science & Technology)The insoluble Cs particles (Type A) were firstly observed in Tsukuba-city on the morning of March 15. The particles have been considered to be generated in RPV of Unit 2 by evaporation/condensation based on the measured Cs/Cs ratio and the core temperatures of each unit. However, the Type A particles with smaller diameter than the Type B particles of Unit 1 origin, are covered by almost pure silicate glass and have a trace of the quenching. This indicates that the particles could have been generated due to the melting of the HEPA filter in SGTS by the fire of H detonation at Unit 3, and atomization followed by quenching of the molten materials by air blast of the explosion. Although the particles were mostly dispersed to the sea because of the wind direction, some of them deposited onto the lower elevation of R/B at Unit 3, could have been subsequently re-suspended and released into the environment, by the steam flow in the R/B caused by restart of the Unit 3 core cooling water injection at 2:30 of March 15.
Okumura, Taiga*; Yamaguchi, Noriko*; Dohi, Terumi; Iijima, Kazuki; Kogure, Toshihiro*
Microscopy, 68(3), p.234 - 242, 2019/06
Times Cited Count:10 Percentile:68.41(Microscopy)Radiocesium-bearing microparticles (CsMPs), consisting substantially of silicate glass, were released to the environment during the Fukushima nuclear accident in March 2011. We investigated a total of nine CsMPs using transmission electron microscopy (TEM) and inferred the atmosphere in the reactors during the accident. From elemental mapping using energy-dispersive X-ray spectrometry, Fe and Zn showing radial inhomogeneities were found in the CsMPs, in addition to the Cs that had been previously reported. Four of the CsMPs included submicron crystals, which were identified as chromite, franklinite, acanthite, molybdenite, and hessite. The chromium-containing spinels, chromite and franklinite, indicated the presence of ferrous iron (Fe), suggesting that the inside of the reactors was reductive to some extent. Electron energy-loss spectroscopy also confirmed that the CsMPs did not contain boron, and therefore the atmosphere in which they were formed might be boron-free.
Takahatake, Yoko; Koma, Yoshikazu
Proceedings of International Topical Workshop on Fukushima Decommissioning Research (FDR 2019) (Internet), 4 Pages, 2019/05
Miyahara, Naoya; Miwa, Shuhei; Horiguchi, Naoki; Sato, Isamu*; Osaka, Masahiko
Journal of Nuclear Science and Technology, 56(2), p.228 - 240, 2019/02
Times Cited Count:7 Percentile:61.94(Nuclear Science & Technology)In order to improve LWR source term under severe accident conditions, the first version of a fission product (FP) chemistry database named "ECUME" was developed. The ECUME is intended to include major chemical reactions and their effective kinetic constants for representative SA sequences. It is expected that the ECUME can serve as a fundamental basis from which FP chemical models in the SA analysis codes can be elaborated. The implemented chemical reactions in the first version were those for representative gas species in Cs-I-B-Mo-O-H system. The chemical reaction kinetic constants were evaluated from either literature data or calculated values using ab-initio calculations. The sample chemical reaction calculation using the presently constructed dataset showed meaningful kinetics effects at 1000 K. Comparison of the chemical equilibrium compositions by using the dataset with those by chemical equilibrium calculations has shown rather good consistency for the representative Cs-I-B-Mo-O-H species. From these results, it was concluded that the present dataset should be useful to evaluate FP chemistry in Cs-I-B-Mo-O-H system under LWA SA conditions.
Ito, Chikara; Naito, Hiroyuki; Ishikawa, Takashi; Ito, Keisuke; Wakaida, Ikuo
JPS Conference Proceedings (Internet), 24, p.011038_1 - 011038_6, 2019/01
A high-radiation resistant optical fiber has been developed in order to investigate the interiors of the reactor pressure vessels and the primary containment vessels at the Fukushima Daiichi Nuclear Power Station. The tentative dose rate in the reactor pressure vessels is assumed to be up to 1 kGy/h. We developed a radiation resistant optical fiber consisting of a 1000 ppm hydroxyl doped pure silica core and 4 % fluorine doped pure silica cladding. We attempted to apply the optical fiber to remote imaging technique by means of fiberscope. The number of core image fibers was increased from 2000 to 22000 for practical use. The transmissive rate of infrared images was not affected after irradiation of 1 MGy. No change in the spatial resolution of the view scope by means of image fiber was noted between pre- and post-irradiation. We confirmed the applicability of the probing system, which consists of a view scope using radiation-resistant optical fibers.
Nishimura, Akihiko; Furusawa, Akinori; Takenaka, Yusuke*
AIP Conference Proceedings 2033, p.080002_1 - 080002_5, 2018/11
Times Cited Count:0 Percentile:0.06(Green & Sustainable Science & Technology)We developed a cpmpact laser maintenance device in order to access a 23 mm diameter for heat exchanger tubes of nuclear power plants. A laser instrumentation device was desighned and assembled to measure the corrosion depth at the inlet of heat exchanger tubes. This device can be applied for heat exchanger tubes in CSP where erosion or cracking might be caused by repetitive thermal induced stress.
Seki, Masakazu; Maekawa, Tomoyuki; Izawa, Kazuhiko; Sono, Hiroki
JAEA-Technology 2017-038, 52 Pages, 2018/03
The Japan Atomic Energy Agency is conducting a reactor modification project of the Static Experiment Critical Facility (STACY). In the modification, STACY is to be converted from a thermal reactor using solution fuel into that using fuel rods and light water moderator. Reactivity of the modified STACY core is controlled by the water level fed in the core tank as well as the present STACY. In order to verify the basic design of the water feed and drain system of the modified STACY, we constructed a mockup test apparatus with almost the same structure and specifications as the modified STACY. In the mockup test, performance checks were pursued regarding limitation of maximum flow of water feeding, adjustment of the flow rate of water feeding, stop of water feeding and others. This report describes the outline and results of the mock-up test of the water feed and drain system of the modified STACY.
Ohgama, Kazuya; Nakano, Yoshihiro; Oki, Shigeo
Journal of Nuclear Science and Technology, 53(8), p.1155 - 1163, 2016/08
Times Cited Count:1 Percentile:10.71(Nuclear Science & Technology)The power distribution and core characteristics in various configurations of fuel subassemblies with an innerduct structure in the Japan Sodium-cooled Fast Reactor were evaluated using a Monte Carlo code for neutron transport and burnup calculation. The correlation between the fraction of fuel subassemblies facing outward and the degree of power increase at the core center was observed regardless of the compositions. This indicated that the spatial fissile distribution caused by innerduct configurations was the major factor of the difference in the power distribution. A power increase was also found in an off-center region, and it tended to be greater than that at the core center because of the steep gradient of neutron flux intensity. The differences in the worth of control rods caused by innerduct configurations were confirmed.
Chino, Masamichi; Terada, Hiroaki; Nagai, Haruyasu; Katata, Genki; Mikami, Satoshi; Torii, Tatsuo; Saito, Kimiaki; Nishizawa, Yukiyasu
Scientific Reports (Internet), 6, p.31376_1 - 31376_14, 2016/08
Times Cited Count:56 Percentile:98.67(Multidisciplinary Sciences)Takamatsu, Kuniyoshi
Hokeikyo Nyusu, (56), p.2 - 4, 2015/10
JP, 2010-166333 Licensable Patent Information Database Patent publication (In Japanese)In our study, we focused on a nondestructive inspection method by cosmic-ray muons which could be used to observe the internal reactor from outside the RPV and the CV. We conducted an observation test on the HTTR to evaluate the applicability of the method to the internal visualization of a reactor. We also analytically evaluated the resolution of existing muon telescopes to assess their suitability for the HTTR observation, and were able to detect the major structures of the HTTR based on the distribution of the surface densities calculated from the coincidences measured by the telescopes. Our findings suggested that existing muon telescopes could be used for muon observation of the internal reactor from outside the RPV and CV.