Sato, Shunsuke*; Nauchi, Yasushi*; Hayakawa, Takehito*; Kimura, Yasuhiko; Kashima, Takao*; Futakami, Kazuhiro*; Suyama, Kenya
Journal of Nuclear Science and Technology, 9 Pages, 2022/10
A new non-destructive method for evaluating Cs activity in spent nuclear fuels was proposed and experimentally demonstrated for physical measurements in burnup credit implementation. Cs activities were quantified using gamma ray measurements and numerical detector response simulations without reference fuels, in which 137Cs activities are well known. Fuel samples were obtained from a lead use assembly (LUA) irradiated in a commercial pressurized water reactor (PWR) up to 53 GWd/t. Gamma rays emitted from the samples were measured using a bismuth germinate (BGO) scintillation detector through a collimator attached to a hot cell. The detection efficiency of gamma rays with the detector was calculated using the PHITS particle transport calculation code considering the measurement geometry. The relative activities of Cs, Cs, and Eu in the sample were measured with a high-purity germanium (HPGe) detector for more accurate simulations of the detector response for the samples. The absolute efficiency of the detector was calibrated by measuring a standard gamma ray source in another geometry. Cs activity in the fuel samples was quantified using the measured count rate and detection efficiency. The quantified Cs activities agreed well with those estimated using the MVP-BURN depletion calculation code.
Higa, Nonoka*; Ito, Takashi; Yogi, Mamoru*; Hattori, Taisuke; Sakai, Hironori; Kambe, Shinsaku; Guguchia, Z.*; Higemoto, Wataru; Nakashima, Miho*; Homma, Yoshiya*; et al.
Physical Review B, 104(4), p.045145_1 - 045145_7, 2021/07
Morita, Yasuji; Fukushima, Masahiro; Kashima, Takao*; Tsubata, Yasuhiro
JAEA-Data/Code 2020-013, 38 Pages, 2020/09
Critical Masses of Cm, Am and the mixture were calculated in metal-water mixtures with water reflector as a basic data for criticality safety assessment of minor actinide separation process. In the mixture of Cm-244 and Cm-245, higher ratio of Cm-245 gives smaller critical mass, but the amount of Cm-245 in the critical mass can be obtained by concentration of Cm-245 in the Cm mixture without depending on the Cm-245 ratio. Critical mass of Cm isotope mixture with 30% Cm-245 was smaller than that of Pu isotope mixture in the practical reprocessing (71% Pu-239 + 17% Pu-240 + 12% Pu-241). When Cm is separated from other element including Am and the solution is concentrated, measure for the critical accident has to be taken. Critical mass of Am-242m is smaller than that of Cm-245, but the ratio of Am-242m in the Am contained in practical spent fuel is small enough, about several percent, and therefore the critical accident by Am does not have to be considered. That by the mixture of Am and Cm does not either.
Matsuda, Shinya*; Ota, Joji*; Nakaima, Kenri*; Iha, Wataru*; Gochi, Jun*; Uwatoko, Yoshiya*; Nakashima, Miho*; Amako, Yasushi*; Honda, Fuminori*; Aoki, Dai*; et al.
Philosophical Magazine, 100(10), p.1244 - 1257, 2020/04
Takeuchi, Tetsuya*; Haga, Yoshinori; Taniguchi, Toshifumi*; Iha, Wataru*; Ashitomi, Yosuke*; Yara, Tomoyuki*; Kida, Takanori*; Tahara, Taimu*; Hagiwara, Masayuki*; Nakashima, Miho*; et al.
Journal of the Physical Society of Japan, 89(3), p.034705_1 - 034705_15, 2020/03
Onuki, Yoshichika*; Kakihana, Masashi*; Iha, Wataru*; Nakaima, Kenri*; Aoki, Dai*; Nakamura, Ai*; Honda, Fuminori*; Nakashima, Miho*; Amako, Yasushi*; Gochi, Jun*; et al.
JPS Conference Proceedings (Internet), 29, p.012001_1 - 012001_9, 2020/02
Iha, Wataru*; Kakihana, Masashi*; Matsuda, Shinya*; Honda, Fuminori*; Haga, Yoshinori; Takeuchi, Tetsuya*; Nakashima, Miho*; Amako, Yasushi*; Gochi, Jun*; Uwatoko, Yoshiya*; et al.
Journal of Alloys and Compounds, 788, p.361 - 366, 2019/06
Takeuchi, Tetsuya*; Yara, Tomoyuki*; Ashitomi, Yosuke*; Iha, Wataru*; Kakihana, Masashi*; Nakashima, Miho*; Amako, Yasushi*; Honda, Fuminori*; Homma, Yoshiya*; Aoki, Dai*; et al.
Journal of the Physical Society of Japan, 87(7), p.074709_1 - 074709_14, 2018/07
Suyama, Kenya; Uchida, Yuriko*; Kashima, Takao; Ito, Takuya*; Miyaji, Takamasa*
NEA/NSC/R(2015)6 (Internet), 253 Pages, 2016/03
The Expert Group on Burnup Credit Criticality Safety (EGBUC) of Working Party of Nuclear Criticality Safety (WPNCS) under the Nuclear Science Committee (NSC) of OECD/NEA has been assessing the accuracy of the burnup calculation code systems by organizing several international benchmarks. This Phase IIIC benchmark specification for BWR 9 by 9 type fuel assembly infinite two-dimensional model was proposed and approved in the meeting of the OECD/NEA/NSC/WPNCS Expert group on burnup credit criticality safety in September 2012 and distributed in October 2012 to the members of the WPNCS. We have set of thirty-five calculation results from sixteen institutes of nine countries. This report presents the results of the benchmark phase IIIC. By this benchmark results, we can confirm the certain progress of the burnup calculation capability than the time of Phase IIIB benchmark. The difference of the neutron multiplication factor generated by the difference of the burnup calculation results by the latest code systems is less than 3%.
Suyama, Kenya; Kashima, Takao
Proceedings of International Conference on Nuclear Criticality Safety (ICNC 2015) (DVD-ROM), p.273 - 282, 2015/09
In the technical development of the criticality safety control of the fuel debris of Fukushima accident in Japan, there have been a discussion on a possibility of adopting BUC with FP. The Expert Group on Burnup Credit Criticality Safety (EGBUC) under the Working Party on Nuclear Criticality Safety (WPNCS) in OECD/NEA Nuclear Science Committee had carried out an international burnup calculation benchmark "Phase-IIIB" and "Phase-IIIC" for BWR fuel assemblies. In these benchmarks the difference of the calculation results of Gd among the participants obtained keen interests because it showed rather larger difference among the participants. Authors has been carried out additional analyses on the accumulation of the gadolinium isotopes in the used nuclear fuel during the burnup. Without cooling time, the assembly-averaged amount of Gd against the burnup value depends on the burnout property of gadolinium in the burnable poison rods. However, after few year cooling time, Gd increase drastically by the decay of Eu. In this case, the amount of gadolinium isotopes in the burnable poison rods has less importance. It means that the adopted parameters and data concerning the Eu generation have much more importance than the burnup treatment of the burnable poison rods for better prediction of Gd.
Kashima, Takao; Suyama, Kenya; Mochizuki, Hiroki*
Energy Procedia, 71, p.159 - 167, 2015/05
The nuclear fuel cycle program of Japan would be delayed because of the impact of the Fukushima Daiichi NPP accident in 2011. Excessive plutonium, however, has to be utilized as mixed-oxide (MOX) fuel to reduce the quantity of plutonium possessed by Japan. Calculation codes and libraries adopted in the fuel cycle analyses of MOX fuel should be benchmarked based on comparison between calculation results and experimental data. From another viewpoint, nuclide inventory analyses of MOX fuel is important for evaluations of the Fukushima accident because MOX fuel has been loaded in the Unit 3 reactor. ARIANE is a PIE program which includes measurements of nuclide compositions of spent MOX fuels discharged from both of pressurized and boiling water reactors. In this study, the PIE data of MOX fuels irradiated in a pressurized water reactor were analyzed by the integrated burnup code system SWAT4 that combines the point burnup system ORIGEN2 and neutron transport calculation solvers, the continuous energy Monte Carlo code MVP or MCNP, and the deterministic neutronics calculation code SRAC. The calculation results of SWAT4 have generally same trends with the case of UO fuel analyses. For major uranium and plutonium isotopes, deviations less than 5% were obtained. This means that SWAT4 has the same accuracy to predict isotopic compositions of irradiated MOX fuel with the case of UO fuel. The radial distribution of isotopes in a pellet was also analyzed, whose results were compared with that measured by SIMS. SWAT4 predicted well the isotope and burnup distributions in an irradiated MOX pellet.
Kashima, Takao; Suyama, Kenya; Takada, Tomoyuki*
JAEA-Data/Code 2014-028, 152 Pages, 2015/03
There have been two versions of SWAT depending on details of its development history: the revised SWAT that uses the deterministic calculation code SRAC as a neutron transportation solver, and the SWAT3.1 that uses the continuous energy Monte Carlo code MVP or MCNP5 for the same purpose. It takes several hours, however, to execute one calculation by the continuous energy Monte Carlo code even on the super computer of the Japan Atomic Energy Agency. Moreover, two-dimensional burnup calculation is not practical using the revised SWAT because it has problems on production of effective cross section data and applying them to arbitrary fuel geometry when a calculation model has multiple burnup zones. Therefore, SWAT4.0 has been developed by adding, to SWAT3.1, a function to utilize the deterministic code SARC2006, which has shorter calculation time, as an outer module of neutron transportation solver for burnup calculation. SWAT4.0 has been enabled to execute two-dimensional burnup calculation by providing an input data template of SRAC2006 to SWAT4.0 input data, and updating atomic number densities of burnup zones in each burnup step. This report describes outline, input data instruction, and examples of calculations of SWAT4.0.
Abe, Hitoshi; Kashima, Takao; Uchiyama, Gunzo
JAEA-Research 2011-015, 27 Pages, 2011/06
To contribute on confirmation of safety of fuel cycle facilities, evaluation method for soundness of confinement capability of the facilities under fire has been investigated. Panel materials of glove-box and cable sheath materials were considered to be an examination object as the representative organic materials in the facilities. Combustion property data, such as mass loss rate and soot generation ratio of the materials, and clogging property data of HEPA filter with combustion of the materials were measured as a parameter with radiation heat given to the materials, supply flow rate to the materials and oxygen concentration in the supply flow. Furthermore, progress of rising differential pressure of HEPA filter under hypothetical scenario of fire accident was evaluated quantitatively by applying these data mutually.
Sugo, Takanobu; Tamada, Masao; Seguchi, Tadao; Shimizu, Takao*; Uotani, Masaki*; Kashima, Ryoichi*
Nihon Genshiryoku Gakkai-Shi, 43(10), p.1010 - 1016, 2001/10
The cost of uranium recovered from seawater was estimated by using the amidoxime adsorbent of polymer fibers synthesized by radiation modification, and the technical problems in the recovery system were extracted. The cost of adsorbent materials, storage in seawater for uranium absorption, and the uranium detachment from the adsorbent was estimated respectively in three different systems of the storage in seawater as a buoy, floating body, and chain binding system. The recovery cost of uranium from seawater was estimated to be 810 times of that from mine uranium. More than 80% of the total cost was occupied by the cost for storage in seawater, which is owing to a weight of metal cage for the holding of adsorbents. The cost can be attained to half by the reduction of the weight to 1/4. One of facing research subject is the improvement of adsorbent ability, since the cost directly depends on the adsorbent performance.
Nishimori, Nobuyuki; Sagara, K.*; Fujita, T.*; Wakamatsu, Fumihiko*; Bussaki, Toru*; Maeda, Kazuhide*; Akiyoshi, H.*; Tsuruta, Kaoru*; Nakamura, Hiroyuki*; Nakashima, Takao*
Nuclear Fusion, 631, p.697C - 700C, 1998/03
no abstracts in English
Eda, Itsumu*; Omine, Mayumi*; Nemoto, Norimasa*; Shimizu, Tomoko*; Tanaka, Sachiko*; Kashima, Takao*; Ito, Yukari*; Taniyama, Hiroshi*; Kamei, Mitsuru*; Yonezawa, Rika; et al.
no journal, ,
no abstracts in English
Abe, Hitoshi; Kashima, Takao; Tashiro, Shinsuke; Uchiyama, Gunzo; Tsuchino, Susumu*; Ishibashi, Takashi*
no journal, ,
In Japan Atomic Energy Agency, to contribute on confirmation of safety of fuel cycle facilities, evaluation method for soundness of confinement capability of the facilities under fire accident has been investigated. Panel materials of Glove-box and cable sheath materials were considered to be an examination object as the representative organic materials in the facilities. Combustion property data, such as mass loss rate of the materials and soot generation ratio, and clogging property data of HEPA filter with combustion of the materials were measured as a parameter with radiation heat given to the materials, supply flow rate to the materials and oxygen concentration in the supply flow. Furthermore, progress of rising differential pressure of HEPA filter under concrete scenario of fire accident was evaluated by connecting these data mutually.
Abe, Hitoshi; Kashima, Takao; Uchiyama, Gunzo
no journal, ,
To contribute on confirmation of safety of fuel cycle facilities, evaluation method for soundness of confinement capability of the facilities under fire accident has been investigated. Panel materials of glove-box and cable sheath materials were considered to be an examination object as the representative organic materials in the facilities. Combustion property data, such as mass loss rate of the materials and soot generation ratio, and clogging property data of HEPA filter with combustion of the materials were measured as a parameter with radiation heat given to the materials, supply flow rate to the materials and oxygen concentration in the supply flow. Furthermore, progress of rising differential pressure of HEPA filter under a scenario of fire accident was evaluated on the basis of these data.
Okubo, Kiyoshi; Suyama, Kenya; Kashima, Takao; Tonoike, Kotaro; Takada, Tomoyuki*
no journal, ,
Criticality safety analysis is necessary for the damaged-fuel handling in the Fukushima Daiichi NPP decommissioning. This presentation show influence of structural materials such as Zry-2, Fe, concrete expected to be present in the damaged fuel. Multiplication factor (kinf) decreases most by replacing moisture, in the damaged fuel, with iron. Replacement of all moisture with Zry-2 gives the same influence as iron, although decrease rate of kinf is lower because of the smaller absorb cross section of Zry-2. Concrete has much less influence due to the neutron moderation by hydrogen contained in concrete, which calls attention on handling of the concrete-fuel mixture. Effect as reflector of the materials is also evaluated.
Uchida, Yuriko; Suyama, Kenya; Kashima, Takao; Tonoike, Kotaro
no journal, ,
It is necessary to evaluate isotopic composition of spent fuel for the criticality safety evaluation of the fuel damaged in the Fukushima Daiichi NPS accident, which demands to evaluate uncertainty of the burnup calculation code system. Analysis with the latest nuclear data library and code was conducted on the burnup calculation benchmark "Phase-IIIB" established by OECD/NEA in late '90s to grasp difference between the new and old calculation results. This presentation shows the re-analysis result of the Phase-IIIB with the Integrated Burnup Code System SWAT3.1 which drives the continuous energy Monte Carlo code MVP and the combined point burnup calculation code ORIGEN2.