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Journal Articles

Criticality characteristics of MCCI products possibly produced in reactors of Fukushima Daiichi Nuclear Power Station

Tonoike, Kotaro; Okubo, Kiyoshi; Takada, Tomoyuki*

Proceedings of International Conference on Nuclear Criticality Safety (ICNC 2015) (DVD-ROM), p.292 - 300, 2015/09

The damaged Unit 1-3 reactors of the Fukushima Daiichi Nuclear Power Station may contain fuel debris of a significant amount that is in a form of molten-core-concrete-interaction (MCCI) product with porous structure. Such low density MCCI product including fissile material is a great concern for its criticality control, especially under submerged condition, due to its fairly good neutron moderation. This report shows computation results of basic criticality characteristics of the MCCI product, which will facilitate criticality risk assessments during decommissioning of the reactors. The results imply that water bound in concrete may raise the risk from the viewpoints of possibility of criticality events and of effectiveness of mitigation measures such as neutron poison injection into coolant water.

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

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.

JAEA Reports

Examination on small-sized cogeneration HTGR for developing countries

Sakaba, Nariaki; Tachibana, Yukio; Shimakawa, Satoshi; Ohashi, Hirofumi; Sato, Hiroyuki; Yan, X.; Murakami, Tomoyuki; Ohashi, Kazutaka; Nakagawa, Shigeaki; Goto, Minoru; et al.

JAEA-Technology 2008-019, 57 Pages, 2008/03

JAEA-Technology-2008-019.pdf:8.59MB

The small-sized and safe cogeneration High Temperature Gas-cooled Reactor (HTGR) that can be used not only for electric power generation but also for hydrogen production and district heating is considered one of the most promising nuclear reactors for developing countries where sufficient infrastructure such as power grids is not provided. Thus, the small-sized cogeneration HTGR, named High Temperature Reactor 50-Cogeneration (HTR50C), was studied assuming that it should be constructed in developing countries. Specification, equipment configuration, etc. of the HTR50C were determined, and economical evaluation was made. As a result, it was shown that the HTR50C is economically competitive with small-sized light water reactors.

JAEA Reports

Benchmark analyses of criticality calculation codes based on the evaluated dissolver-type criticality experiment systems

Okuno, Hiroshi; Takada, Tomoyuki; Yoshiyama, Hiroshi; Miyoshi, Yoshinori

JAEA-Data/Code 2005-001, 117 Pages, 2005/11

JAEA-Data-Code-2005-001.pdf:9.37MB

Criticality calculation codes/code systems MCNP, MVP, SCALE and JACS, which are currently typically used in Japan for nuclear criticality safety evaluation, were benchmarked for so called dissolver-typed systems, i.e., fuel rod arrays immersed in fuel solution. The benchmark analyses were made for the evaluated critical experiments published in the International Criticality Safety Benchmark Evaluation Project (ICSBEP) Handbook: one evaluation representing five critical configurations from heterogeneous core of low-enriched uranium dioxides at the Japan Atomic Energy Research Institute and two evaluations representing 16 critical configurations from heterogeneous core of mixed uranium and plutonium dioxides (MOXs) at the Battelle Pacific Northwest Laboratories of the U.S.A.. The results of the analyses showed that the minimum values of the neutron multiplication factor obtained with MCNP, MVP, SCALE and JACS were 0.993, 0.990, 0.993, 0.972, respectively, which values are from 2% to 4% larger than the maximum permissible multiplication factor of 0.95.

Journal Articles

Calculation of nuclear characteristic parameters and drawing subcriticality judgment graphs of infinite fuel systems for typical nuclear fuels

Okuno, Hiroshi; Takada, Tomoyuki

Journal of Nuclear Science and Technology, 41(4), p.481 - 492, 2004/04

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

Nuclear characteristic parameters were calculated and subcriticality judgement graphs were drawn for revision purposes of the Data Collection for the Nuclear Criticality Safety Handbook. The nuclear characteristic parameters were the neutron multiplication factor in infinite media, migration area and diffusion constants for 11 kinds of typical fuels encountered in criticality safety evaluation of nuclear fuel cycle facilities. These fuels included ADU-H$$_{2}$$O, UF6-HF and Pu(NO$$_{3}$$)$$_{4}$$-UO$$_{2}$$(NO$$_{3}$$)$$_{2}$$ solution, of which data were not cited in the Data Collection. The calculation was made with the Japanese evaluated nuclear data library JENDL-3.2 and a sequence of criticality calculation codes, SRAC, POST and SIMCRI. The subcriticality judgement graphs that depict the region satisfying the inequality relation of the neutron multiplication factor less than 0.98 between the two variables (a) uranium enrichment, 239Pu/Pu ratio or plutonium enrichment and (b) H/(Pu+U) ratio were drawn for the same kinds of fuels except UF6-HF in infinite media.

JAEA Reports

Revaluation of JACS code system benchmark analyses of the heterogeneous system; Fuel rods in U+Pu nitric acid solution system

Takada, Tomoyuki; Miyoshi, Yoshinori; Katakura, Junichi

JAERI-Tech 2003-036, 80 Pages, 2003/03

JAERI-Tech-2003-036.pdf:3.67MB

In order to perform accuracy evaluation of the critical calculation by the combination of multi-group constant library MGCL and 3-dimensional Monte Carlo code KENO-IV among critical safety evaluation code system JACS, benchmark calculation was carried out from 1980 in 1982. Some cases where the neutron multiplication factor calculated in the heterogeneous system in it was less than 0.95 were seen. In this report, it re-calculated by considering the cause about the heterogeneous system of the U+Pu nitric acid solution systems containing the neutron poison shown in JAERI-M 9859. The present study has shown that the keff value less than 0.95 given in JAERI-M 9859 is caused by the fact that the water reflector below a cylindrical container was not taken into consideration in the KENO-IV calculation model. By taking into the water reflector, the KENO-IV calculation gives a keff value greater than 0.95 and a good agreement with the experiment.

JAEA Reports

Acceleration of criticality analysis solution convergence by matrix eigenvector for a system with weak neutron interaction

Nomura, Yasushi; Takada, Tomoyuki; Kadotani, Hiroyuki*; Kuroishi, Takeshi

JAERI-Tech 2003-020, 88 Pages, 2003/03

JAERI-Tech-2003-020.pdf:4.31MB

no abstracts in English

Journal Articles

Analysis of the (N,xN') reactions by quantum molecular dynamics plus statistical decay model

Niita, Koji*; Chiba, Satoshi; Maruyama, Toshiki; Maruyama, Tomoyuki*; Takada, Hiroshi; Fukahori, Tokio; Nakahara, Yasuaki; Iwamoto, Akira

Physical Review C, 52(5), p.2620 - 2635, 1995/11

 Times Cited Count:315 Percentile:99.49(Physics, Nuclear)

no abstracts in English

Oral presentation

Criticality safety evaluation of damaged burned nuclear fuel; Basic parameters

Suyama, Kenya; Totsuka, Masayoshi; Uchiyama, Gunzo; Takada, Tomoyuki*

no journal, , 

Decommission of the Fukushima Daiichi NPP is under discussion. It is not possible for us to assure the fuel assemblies keep the original geometry, and the nuclide composition of the material of the damaged fuel and their positions in the reactor are also unknown now. So that, in this stage, it is difficult for us to judge whether the parameters adopted in the criticality safety evaluation is reasonable or on the contrary over conservative. Based on this view, for the further study on the criticality safety evaluation of the damaged nuclear fuel, we have stared evaluating the basic criticality parameters of such fuel materials.

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

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

Critical Mass estimation of MCCI products

Tonoike, Kotaro; Okubo, Kiyoshi; Takada, Tomoyuki*

no journal, , 

no abstracts in English

Oral presentation

A Micro pixel chamber based neutron imaging detector ($$mu$$ NID) with boron converter for energy-resolved neutron imaging at J-PARC

Parker, J. D.*; Shinohara, Takenao; Harada, Masahide; Hayashida, Hirotoshi*; Hiroi, Kosuke; Kai, Tetsuya; Matsumoto, Yoshihiro*; Oikawa, Kenichi; Segawa, Mariko; Su, Y.; et al.

no journal, , 

14 (Records 1-14 displayed on this page)
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