Absolute quantification of Cs activity in spent nuclear fuel with calculated detector response function

Sato, Shunsuke*; Nauchi, Yasushi*; Hayakawa, Takehito*; Kimura, Yasuhiko; Kashima, Takao*; Futakami, Kazuhiro*; Suyama, Kenya

Journal of Nuclear Science and Technology, 9 Pages, 2022/10

https://doi.org/10.1080/00223131.2022.2130462

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.

Critical mass evaluation of minor actinides in aqueous solution; Data for criticality safety assessment of separation process

Morita, Yasuji; Fukushima, Masahiro; Kashima, Takao*; Tsubata, Yasuhiro

JAEA-Data/Code 2020-013, 38 Pages, 2020/09

JAEA-Data-Code-2020-013.pdf:1.94MB

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.

Spin glass behavior in EuCuSi single crystal grown by the flux method

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

https://doi.org/10.7566/JPSJ.89.034705

Anomalous ferromagnetic ordering in EuCuP

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

https://doi.org/10.1016/j.jallcom.2019.02.213

Burn-up credit criticality safety benchmark phase III-C; Nuclide composition and neutron multiplication factor of a boiling water reactor spent fuel assembly for burn-up credit and criticality control of damaged nuclear fuel

Suyama, Kenya; Uchida, Yuriko*; Kashima, Takao; Ito, Takuya*; Miyaji, Takamasa*

NEA/NSC/R(2015)6 (Internet), 253 Pages, 2016/03

https://www.oecd-nea.org/science/docs/2015/nsc-r2015-6.pdf

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%.

Validation of burnup calculation code SWAT4 by evaluation of isotopic composition data of mixed oxide fuel irradiated in pressurized water reactor

Kashima, Takao; Suyama, Kenya; Mochizuki, Hiroki*

Energy Procedia, 71, p.159 - 167, 2015/05

https://doi.org/10.1016/j.egypro.2014.11.866

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.

Experiment on evaluation of confinement capability of fuel cycle facility under combustion of glove-box and cable sheath materials (Contract research)

Abe, Hitoshi; Kashima, Takao; Uchiyama, Gunzo

JAEA-Research 2011-015, 27 Pages, 2011/06

JAEA-Research-2011-015.pdf:1.81MB

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.

Precise experiment on (anti-n)+d scattering at 12MeV

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

Study on evaluation of confinement capability of fuel cycle facility

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.

Criticality safety evaluation of damaged burned nuclear fuel; Effect of structural materials

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.