Refine your search:     
Report No.
 - 
Search Results: Records 1-20 displayed on this page of 23

Presentation/Publication Type

Initialising ...

Refine

Journal/Book Title

Initialising ...

Meeting title

Initialising ...

First Author

Initialising ...

Keyword

Initialising ...

Language

Initialising ...

Publication Year

Initialising ...

Held year of conference

Initialising ...

Save select records

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

 Times Cited Count:2 Percentile:83.55

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

SWAT3.1; The Integrated burnup code system driving continuous energy Monte Carlo codes MVP and MCNP

Suyama, Kenya; Mochizuki, Hiroki*; Takada, Tomoyuki*; Ryufuku, Susumu*; Okuno, Hiroshi; Murazaki, Minoru; Okubo, Kiyoshi

JAEA-Data/Code 2009-002, 124 Pages, 2009/05

JAEA-Data-Code-2009-002.pdf:14.09MB

Integrated burnup calculation code system SWAT is a system that combines neutronics calculation code SRAC widely used in Japan and point burnup calculation code ORIGEN2. It has been used to evaluate the composition of the uranium, plutonium, minor actinide and the fission products in the spent nuclear fuel. Because of the ability to treat the arbitrary fuel geometry and no requirement of generating the effective cross section data, there is a great advantage to introduce continuous energy Monte Carlo Code into the burnup calculation code. Based on this idea, the integrated burnup calculation code system SWAT3.1 was developed by combining the continuous energy Monte Carlo code MVP and MCNP and ORIGEN2. This report describes the outline, input data instruction and several example of the calculation.

Journal Articles

Corrections to the $$^{148}$$Nd method of evaluation of burnup for the PIE samples from Mihama-3 and Genkai-1 reactors

Suyama, Kenya; Mochizuki, Hiroki*

Annals of Nuclear Energy, 33(4), p.335 - 342, 2006/03

 Times Cited Count:8 Percentile:49.7(Nuclear Science & Technology)

The value of the burnup is one of the most important parameters of samples taken by post irradiation examination (PIE). In this study, concerning the PIE data from Mihama-3 and Genkai-1 PWRs, which were taken at the Japan Atomic Energy Research Institute, the burnup values of the PIE samples were re-evaluated and the PIE data are re-analyzed using SWAT and SWAT2 code systems with JENDL-3.3 library. This analysis concludes that the burnup values of samples from Mihama-3 and Genkai-1 PWRs should be corrected of 2-3%. The effect of re-evaluation of the burnup value on the neutron multiplication factor is approximately 1% for PIE samples having the burnup of larger than 30 GWd/t. Comparison between calculation results using a single pin cell model and an assembly model is carried out. Because the both results agreed within a few percents, we concluded that the single pin cell model is suitable for the analysis of PIE samples and the underestimation of plutonium isotopes does not result from the geometry model.

Journal Articles

Effect of neutron induced reactions of neodymium-147 and 148 on burnup evaluation

Suyama, Kenya; Mochizuki, Hiroki*

Journal of Nuclear Science and Technology, 42(7), p.661 - 669, 2005/07

 Times Cited Count:15 Percentile:69.97(Nuclear Science & Technology)

Burnup is important value for criticality safety evaluation of spent nuclear fuel. Nd-148 method is one of most important method to evaluate the burnup of post irradiation examination (PIE) samples, and well known that it has good accuracy. However, the evaluated burnup values could be perturbed by the neutron capture reaction of Nd-147 and Nd-148. And in the analysis of PIE data from PWR, the calculation results of Nd-148 have approximately more than 1% deviation from experiment. In this study, the contribution of neutron capture reaction of Nd-147 and Nd-148 to Nd-148 amount are discussed. Especially for Nd-147 contribution, it is shown that the current evaluated cross section of Nd-147 is not supported and the new evaluation is consistent with the analysis of PIE data. Possible perturbed amount of Nd-148 by both reactions is less than 0.7% for normal reactor operation condition, and it is approximately 0.1% for 30 GWd/t (BWR) and 40 GWd/t (PWR). Finally, we confirm again that Nd-148 method is good evaluation method.

Journal Articles

Validation of integrated burnup code system SWAT2 by the analyses of isotopic composition of spent nuclear fuel

Suyama, Kenya; Mochizuki, Hiroki*; Okuno, Hiroshi; Miyoshi, Yoshinori

Proceedings of International Conference on Physics of Fuel Cycles and Advanced Nuclear Systems; Global Developments (PHYSOR 2004) (CD-ROM), 10 Pages, 2004/04

This paper provides validation results of SWAT2, the revised version of SWAT, which is a code system combining point burnup code ORIGEN2 and continuous energy Monte Carlo code MVP, by the analysis of post irradiation examinations (PIEs). Some isotopes show differences of calculation results between SWAT and SWAT2. However, generally, the differences are smaller than the error of PIE analysis that was reported in previous SWAT validation activity, and improved results are obtained for several important fission product nuclides. This study also includes comparison between an assembly and a single pin cell geometry models.

Journal Articles

Improvements to SFCOMPO; A Database on isotopic composition of spent nuclear fuel

Suyama, Kenya; Nouri, A.*; Mochizuki, Hiroki*; Nomura, Yasushi*

JAERI-Conf 2003-019, p.890 - 892, 2003/10

Isotopic composition is one of the most relevant data to be used in the calculation of burnup of irradiated nuclear fuel. Since autumn 2002, the Organisation for Economic Co-operation and Development/Nuclear Energy Agency OECD/NEA) has operated a database of isotopic composition; SFCOMPO, initially developed in Japan Atomic Energy research Institute. This paper describes latest version of SFCOMPO and the future development plan in OECD/NEA.

Journal Articles

Deuterium effect on the subcritical limit for fissile-to-hydrogen ratio

Okuno, Hiroshi; Akiyama, Hideo*; Mochizuki, Hiroki*

Journal of Nuclear Science and Technology, 40(1), p.57 - 60, 2003/01

 Times Cited Count:1 Percentile:10.89(Nuclear Science & Technology)

Low-level waste (LLW) drums are required to transport as fissile material if the current IAEA's Regulations for the Safe Transport of Radioactive Material are rigorously applied. This problem is a consequence that water contents of concrete in LLW drums contained deuterium (D) in quantities more than 0.1% of fissile material mass, therefore they are not excepted from packages containing fissile material. Consideration of differences in the absorption cross sections of light hydrogen and D shows that the relative increase in the neutron multiplication factor by a presence of D in natural water for hydrogen (H)-moderated systems is not larger than 0.015%. A numerical calculation confirms that the infinite multiplication factor of a mixture of $$^{235}$$U-metal and water in a $$^{235}$$U/H mass ratio of 5% increases proportionally to the D/H atomic ratio, and that its relative increase is less than 0.03% for the D/H atomic ratio of 0.015%. The limiting fissile-to-H mass ratio of 5% in the exception rule is concluded to be applicable to H-moderated systems including D in natural water.

JAEA Reports

Derivation of correction factor to be applied for calculated results of BWR fuel isotopic composition by ORIGEN2.1 code

Nomura, Yasushi; Mochizuki, Hiroki*

JAERI-Tech 2002-068, 131 Pages, 2002/11

JAERI-Tech-2002-068.pdf:5.59MB

no abstracts in English

Journal Articles

Revised burnup code system SWAT; Description and validation using postirradiation examination data

Suyama, Kenya; Mochizuki, Hiroki*; Kiyosumi, Takehide*

Nuclear Technology, 138(2), p.97 - 110, 2002/05

 Times Cited Count:23 Percentile:80.69(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Nuclide composition benchmark data set for verifying burnup codes on spent light water reactor fuels

Nakahara, Yoshinori; Suyama, Kenya; Inagawa, Jun; Nagaishi, Ryuji; Kurosawa, Setsumi; Kono, Nobuaki; Onuki, Mamoru; Mochizuki, Hiroki*

Nuclear Technology, 137(2), p.1 - 16, 2002/02

no abstracts in English

JAEA Reports

Derivation of correction factor to be applied for calculated results of PWR fuel isotopic composition by ORIGEN2 code

Suyama, Kenya; Murazaki, Minoru*; Mochizuki, Hiroki*; Nomura, Yasushi

JAERI-Tech 2001-074, 119 Pages, 2001/11

JAERI-Tech-2001-074.pdf:4.21MB

no abstracts in English

JAEA Reports

Spent fuel composition data base system on WWW; SFCOMPO on WWW Ver.2

Mochizuki, Hiroki*; Suyama, Kenya; Nomura, Yasushi; Okuno, Hiroshi

JAERI-Data/Code 2001-020, 394 Pages, 2001/08

JAERI-Data-Code-2001-020.pdf:17.99MB

no abstracts in English

JAEA Reports

Reactivity effect of spent fuel depending on burn-up history

Hayashi, Takafumi*; Suyama, Kenya; Mochizuki, Hiroki*; Nomura, Yasushi

JAERI-Tech 2001-041, 158 Pages, 2001/06

JAERI-Tech-2001-041.pdf:5.15MB

no abstracts in English

JAEA Reports

Analyses of PWR spent fuel composition using SCALE and SWAT code systems to find correction factors for criticality safety applications adopting burnup credit

Hee, S. S.*; Suyama, Kenya; Mochizuki, Hiroki*; Okuno, Hiroshi; Nomura, Yasushi

JAERI-Research 2000-066, 131 Pages, 2001/01

JAERI-Research-2000-066.pdf:6.36MB

no abstracts in English

JAEA Reports

Revised SWAT; The Integrated burnup calculation code system

Suyama, Kenya; Kiyosumi, Takehide*; Mochizuki, Hiroki*

JAERI-Data/Code 2000-027, 88 Pages, 2000/07

JAERI-Data-Code-2000-027.pdf:4.08MB

no abstracts in English

Oral presentation

Increase in the intensity and brightness enhancement of the slow positron beam and its application at the Slow Positron Facility, KEK

Wada, Ken*; Hyodo, Toshio*; Kosuge, Takashi*; Saito, Yuki*; Yagishita, Akira*; Ikeda, Mitsuo*; Osawa, Satoshi*; Suwada, Tsuyoshi*; Furukawa, Kazuro*; Shirakawa, Akihiro*; et al.

no journal, , 

no abstracts in English

Oral presentation

Beam-line improvement and new experiment stations of KEK-IMSS slow-positron facility

Wada, Ken*; Mochizuki, Izumi*; Hyodo, Toshio*; Kosuge, Takashi*; Saito, Yuki*; Shidara, Tetsuo*; Osawa, Satoshi*; Ikeda, Mitsuo*; Shirakawa, Akihiro*; Furukawa, Kazuro*; et al.

no journal, , 

no abstracts in English

Oral presentation

Beam-line improvement and recent results of KEK-IMSS slow-positron facility

Wada, Ken*; Mochizuki, Izumi*; Hyodo, Toshio*; Kosuge, Takashi*; Saito, Yuki*; Shidara, Tetsuo*; Osawa, Satoshi*; Ikeda, Mitsuo*; Shirakawa, Akihiro*; Furukawa, Kazuro*; et al.

no journal, , 

no abstracts in English

Oral presentation

Present status of the KEK-IMSS Slow Positron Facility; A New beam-line branch and a rearrangement of the experiment stations

Wada, Ken*; Mochizuki, Izumi*; Hyodo, Toshio*; Kosuge, Takashi*; Saito, Yuki*; Nigorikawa, Kazuyuki*; Shidara, Tetsuo*; Osawa, Satoshi*; Ikeda, Mitsuo*; Shirakawa, Akihiro*; et al.

no journal, , 

no abstracts in English

Oral presentation

Recent developments and results of the KEK slow positron facility

Wada, Ken*; Mochizuki, Izumi*; Hyodo, Toshio*; Kosuge, Takashi*; Saito, Yuki*; Nigorikawa, Kazuyuki*; Shidara, Tetsuo*; Osawa, Satoshi*; Ikeda, Mitsuo*; Shirakawa, Akihiro*; et al.

no journal, , 

no abstracts in English

23 (Records 1-20 displayed on this page)