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Journal Articles

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

Journal Articles

Effects of magnetic field and pressure on the valence-fluctuating antiferromagnetic compound EuPt$$_2$$Si$$_2$$

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

 Times Cited Count:2 Percentile:29.91(Physics, Multidisciplinary)

Journal Articles

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

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%.

Journal Articles

Accumulation of gadolinium isotopes in used nuclear fuel

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 $$^{155}$$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 $$^{155}$$Gd against the burnup value depends on the burnout property of gadolinium in the burnable poison rods. However, after few year cooling time, $$^{155}$$Gd increase drastically by the decay of $$^{155}$$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 $$^{155}$$Eu generation have much more importance than the burnup treatment of the burnable poison rods for better prediction of $$^{155}$$Gd.

Journal Articles

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

 Percentile:100

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$$_{2}$$ 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$$_{2}$$ 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.

JAEA Reports

SWAT4.0; The Integrated burnup code system driving continuous energy Monte Carlo codes MVP, MCNP and deterministic calculation code SRAC

Kashima, Takao; Suyama, Kenya; Takada, Tomoyuki*

JAEA-Data/Code 2014-028, 152 Pages, 2015/03

JAEA-Data-Code-2014-028.pdf:13.39MB

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.

JAEA Reports

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.

Journal Articles

Recovery system for uranium from seawater with fibrous adsorbent and its preliminary cost estimation

Sugo, Takanobu; Tamada, Masao; Seguchi, Tadao; Shimizu, Takao*; Uotani, Masaki*; Kashima, Ryoichi*

Nippon Genshiryoku Gakkai-Shi, 43(10), p.1010 - 1016, 2001/10

 Times Cited Count:12 Percentile:30.25(Nuclear Science & Technology)

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 8$$sim$$10 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.

Journal Articles

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

Oral presentation

Environmental monitoring from resident's point of view

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

Oral presentation

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.

Oral presentation

Study on evaluation of confinement capability of fuel cycle facility under fire accident

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.

Oral presentation

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.

Oral presentation

Re-analyses of the OECD/NEA burnup calculation benchmark Phase-IIIB for BWR fuel assembly

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.

Oral presentation

Development and validation of the integrated burnup analysis code system SWAT4

Kashima, Takao; Suyama, Kenya; Uchida, Yuriko; Tonoike, Kotaro; Takada, Tomoyuki*

no journal, , 

no abstracts in English

Oral presentation

The Outline and a draft report of the OECD/NEA burnup calculation benchmark "Phase-IIIC" for BWR 9$$times$$9 fuel assembly

Uchida, Yuriko; Suyama, Kenya; Kashima, Takao; Tonoike, Kotaro

no journal, , 

It is necessary to evaluate the isotopic composition of the spent fuels for the criticality safety evaluation of the fuel damaged during the Fukushima Daiichi (1F) accident, which demands to evaluate the uncertainty of the burnup calculation code system. The Expert Group on Burnup Credit Criticality Safety of Working Party of Nuclear Criticality Safety under the NSC of OECD/NEA carried out the burnup calculation benchmark "Phase-IIIB" for BWR STEP-2 fuel, although it has been done more than 10 years ago. Because the latest BWR STEP-3 fuel were adopted in 1F and the calculation code and the nuclear data libraries have been revised, the Expert Group carries out the burnup calculation benchmark "Phase-IIIC" for BWR STEP-3 fuel since September 2012 by the proposal of JAEA. This presentation shows the outline and a draft report of this benchmark.

Oral presentation

XAFS study on effect of molybdenum addition on photo-induced particle formation of palladium

Saeki, Morihisa; Matsumura, Daiju; Yomogida, Takumi; Taguchi, Tomitsugu; Tsuji, Takuya; Kusano, Shogo*; Miyazaki, Tatsuya*; Takao, Koichiro*; Oba, Hironori; Nakashima, Nobuaki*

no journal, , 

Recently, we found that photo-induced particle formation of Pd$$^{2+}$$ was promoted by coexistence of molybdenum anion in aqueous solution. In this study, we performed in-situ time-resolved XAFS measurement in the photoreduction of Pd$$^{2+}$$ in presence of various concentration of molybdenum anion. The effect of molybdenum addition on photo-induced particle formation of Pd will be discussed based on the relationship between the photoreduction rate of Pd$$^{2+}$$ and the concentration of molybdenum anion.

Oral presentation

XAFS study on effect of molybdenum addition on photo-induced particle formation of palladium

Saeki, Morihisa; Matsumura, Daiju; Yomogida, Takumi; Taguchi, Tomitsugu; Tsuji, Takuya; Kusano, Shogo*; Miyazaki, Tatsuya*; Takao, Koichiro*; Oba, Hironori; Nakashima, Nobuaki*

no journal, , 

no abstracts in English

Oral presentation

Temperature dependence of temperature coefficient and void coefficient of reactivity for dilute plutonium solutions

Watanabe, Tomoaki; Kashima, Takao; Yamane, Yuichi

no journal, , 

In order to develop evaluation techniques of the number of nuclear fission for criticality accidents with boiling of solution fuel, temperature dependence of temperature coefficients and void coefficients of reactivity for dilute plutonium solutions is analytically investigated. The results show that evaluation accuracy of the power when solution fuel is boiling is expected to be improved by about 10% by considering the temperature dependence.

19 (Records 1-19 displayed on this page)
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