Izawa, Kazuhiko; Tonoike, Kotaro; Sono, Hiroki; Miyoshi, Yoshinori
JAEA-Conf 2014-003, Appendix (CD-ROM), 13 Pages, 2015/03
Critical assemblies of thermal neutron system are decreasing in number in spite of their important roles in the reactor physics research. On the other hand, the extension of the utilization term of the LWRs brings new research themes requiring critical experiments of thermal neutron system. JAEA is modifying the Static Critical Experiment Facility (STACY) to revive the critical experiments. The modified STACY will be an infrastructure for the experimental research of reactor physics on thermal neutron system. The primary mission of the modified STACY at present is the critical experiments for fuel debris to contribute to the criticality safety control of the fuel debris generated by the severe accident of the Fukushima Daiichi Nuclear Power Station. This report introduces the plan of criticality safety research in Japan Atomic Energy Agency following the accident, and describes the role of the modified STACY in the retrieval work of fuel debris from the damaged reactor.
Yamashita, Takuya*; Shimoyama, Yusuke*; Haga, Yoshinori; Matsuda, Tatsuma*; Yamamoto, Etsuji; Onuki, Yoshichika; Sumiyoshi, Hiroaki*; Fujimoto, Satoshi*; Levchenko, A.*; Shibauchi, Takasada*; et al.
Nature Physics, 11(1), p.17 - 20, 2015/01
Izawa, Kazuhiko; Aoyama, Yasuo; Sono, Hiroki; Ogawa, Kazuhiko; Yanagisawa, Hiroshi; Miyoshi, Yoshinori
Proceedings of 9th International Conference on Nuclear Criticality (ICNC 2011) (CD-ROM), 11 Pages, 2012/02
For reactor physics and criticality safety researches, the Static Experiment Critical Facility (STACY) will be modified. In the modification, the present STACY, solution-fuel-type homogeneous cores, will be converted to fuel-pin-type heterogeneous cores moderated by light water. For nuclear safety design of the modified STACY, computational analyses have been carried out by using a Monte Carlo code MVP and a transport code system DANTSYS with cross-section data based on the JENDL-3.3. In the analyses, basic nuclear characteristics have been evaluated, such as criticality, water-level worth and reactor shutdown margin. By the results of these analyses, the feasibility of reactivity control mechanism and the sufficiency of reactor shutdown margin of the modified STACY were confirmed. In addition, temperature and void coefficients of reactivity and kinetic parameters were obtained to comprehend nuclear characteristics of the modified STACY.
Tonoike, Kotaro; Miyoshi, Yoshinori; Uchiyama, Gunzo
Proceedings of 9th International Conference on Nuclear Criticality (ICNC 2011) (CD-ROM), 11 Pages, 2012/02
Several series of critical experiments have been conducted in the Static Critical Experiment Facility (STACY) to measure critical volume of low-enriched uranyl nitrate solution. Its purpose is to obtain critical benchmark data for validation of computation methods used in the criticality safety analysis of nuclear facilities, especially, reprocessing plants. The experiment series are: homogeneous single-unit system, homogeneous multi-unit system, and heterogeneous system. Experimental conditions such as core tank geometry and fuel solution composition, and measurement results of critical volume were carefully and precisely evaluated to produce critical benchmark data. Sensitivity analysis on uncertainties of such experimental conditions was conducted to estimate overall uncertainties of the benchmarks. In this report, features of each experimental system will be highlighted by describing results of the experiments and the sensitivity analysis. Also presented will be lessons leaned from the experimental and evaluation experience which might be valuable for design of future critical experiments.
Tonoike, Kotaro; Suyama, Kenya; Okuno, Hiroshi; Miyoshi, Yoshinori; Uchiyama, Gunzo
Proceedings of 9th International Conference on Nuclear Criticality (ICNC 2011) (CD-ROM), 8 Pages, 2012/02
The 1st version of criticality safety handbook of Japan was published in 1988. A criticality safety analysis code system JACS was validated, and minimum critical mass and safety limit mass of various fissile materials were calculated. During more than two decades since then, new critical experimental data were taken in the Static Critical Experiment Facility (STACY), and more precise benchmark data of wider range of fissile materials were accumulated by the International Criticality Safety Benchmark Evaluation Project (ICSBEP). Computational capability has greatly grown, and new codes and nuclear data have been developed. The 2nd version of the handbook utilizes the results of validation of the criticality analysis method with a continuous energy Monte-Carlo code MVP and a nuclear data library JENDL-3.2 using the benchmark data chosen from the ICSBEP handbook. Results of the benchmark calculation were statistically studied, from which the safety limit value of multiplication factor was derived as 0.98. Based on the conclusion, minimum critical mass and safety limit mass were calculated. Future plan of research activities on the criticality safety in JAEA will be also overviewed.
Tonoike, Kotaro; Miyoshi, Yoshinori; Uchiyama, Gunzo
Journal of Nuclear Science and Technology, 48(7), p.1118 - 1128, 2011/07
A series of critical experiments were performed using heterogeneous cores at STACY in JAEA in order to obtain systematic benchmark criticality data concerning the dissolving process in reprocessing plants. Focusing on the use of gadolinium as a soluble poison, critical mass was measured for a combination of uranium dioxide fuel rods and uranyl nitrate solution poisoned with gadolinium (Gd). The Gd concentrations were varied up to 0.1 gGd/L. The other series of experiments were also conducted, as reference cases, varying uranium concentration in the fuel solution without Gd. The results provided benchmark criticality data for validation of neutron multiplication factor calculation on heterogeneous systems such as a dissolver. Validation calculation of JACS based on the newly obtained benchmarks supports the justification of its utilization for the criticality safety analysis.
Okuno, Hiroshi; Suyama, Kenya; Tonoike, Kotaro; Yamane, Yuichi; Yamamoto, Toshihiro; Miyoshi, Yoshinori; Uchiyama, Gunzo
JAEA-Data/Code 2009-010, 175 Pages, 2009/08
The report revised the Data Collection part of Nuclear Criticality Safety Handbook, which was published in 1988. This second version provided criticality data on homogeneous U-HO and UF-HF, which were not cited in the previous version, and increased those data on the medium-enriched uranium fuels. Calculations were performed mainly with the Continuous-Energy Monte Carlo Criticality Calculation Code, MVP, and the Japanese Evaluated Nuclear Data Library, JENDL-3 Revision 2, JENDL-3.2, both of which were developed at the late Japan Atomic Energy Research Institute (JAERI). Data on atomic number densities of actinide metal and oxide were additionally supplied, and nuclide compositions of irradiated fuels were improved from the first version. One million histories of neutrons were followed in benchmark calculations of critical experiments and in calculations of single-unit criticality data, i.e., critical mass, volume, dimensions, etc., to attain almost ten times higher precision than the first version.
Tonoike, Kotaro; Yamamoto, Toshihiro; Miyoshi, Yoshinori; Uchiyama, Gunzo; Watanabe, Shoichi*
Journal of Nuclear Science and Technology, 46(4), p.354 - 365, 2009/04
A series of critical experiments were performed using heterogeneous cores at the Static Experiment Critical Facility in Japan Atomic Energy Agency in order to obtain systematic benchmark data concerning dissolving process in a reprocessing plant. Focusing on the introduction of the burn-up credit, critical mass measurement was conducted for a combination of uranium dioxide fuel rods (5wt% U) and uranyl nitrate solution (6wt% U) poisoned with pseudo fission product (FP) elements - samarium, cesium, rhodium, and europium. Fuel rods were arrayed with an 1.5-cm lattice interval in the poisoned fuel solution in a 60-cm diameter cylindrical tank. The uranium concentrations of the solution was roughly kept at about 320gU/L, and the FP element concentrations were adjusted to be equivalent to a burn-up of about 30GWd/t. The result provides basic experimental data for validation of computational methods to evaluate a reactivity effect of each FP element, as well as benchmark criticality data for validation of neutron multiplication factor calculation of heterogeneous systems of spent fuel. In the report, detail of the experiments and its benchmark models will be presented as well as the procedure and the result of separate reactivity worth evaluation for each FP element. The experimental result and the computational evaluation will also be compared.
Rugama, Y.*; Blomquist, R.*; Brady Raap, M.*; Briggs, B.*; Gulliford, J.*; Miyoshi, Yoshinori; Suyama, Kenya; Ivanova, T.*
Proceedings of International Conference on the Physics of Reactors, Nuclear Power; A Sustainable Resource (PHYSOR 2008) (CD-ROM), 5 Pages, 2008/09
Over the years, substantial progress has been made in developing nuclear data and computer codes to evaluate criticality safety for nuclear fuel handling. This state-of-the-art knowledge also has an economic impact. Increased understanding of uncertainties in safety margins allow rational and more economical designs for manipulation, storage and transportation of fissile materials. In the mid-nineties, several activities related to criticality-safety were grouped together into the Working Party on Nuclear Criticality Safety. Six expert groups co-ordinate various activities that run the gamut from experimental evaluations to code and data inter-comparisons, for the study of static and transient criticality behaviors. The various reports produced by the expert groups attempt to establish practical rules and identify applicable tools when appropriate.
Rugama, Y.*; Blomquist, R.*; Brady Raap, M.*; Briggs, B.*; Gulliford, J.*; Miyoshi, Yoshinori; Suyama, Kenya
Proceedings of 2008 International Congress on Advances in Nuclear Power Plants (ICAPP '08) (CD-ROM), p.1391 - 1393, 2008/06
The OECD Nuclear Energy Agency (NEA) started dealing with criticality-safety related subjects during the nineteen-seventies. In the mid-nineties, several activities related to criticality-safety were grouped together into the Working Party on Nuclear Criticality Safety. This working party has since been operating and reporting to the Nuclear Science Committee. Six expert groups coordinate various activities ranging from experimental evaluations to code and data intercomparisons for the study of static and transient criticality behaviours. The paper describes current activities performed in this framework and the achievements of the various expert groups.
Okuno, Hiroshi; Suyama, Kenya; Okuda, Yasuhisa*; Yoshiyama, Hiroshi*; Miyoshi, Yoshinori
Proceedings of 8th International Conference on Nuclear Criticality Safety (ICNC 2007), p.140 - 143, 2007/05
In this research, a preliminary critical safe evaluation of a canister was performed, which stored either (1) four UO fuel assemblies (initial uranium enrichment of 4.1 wt%) or (2) four mixed uranium and plutonium oxide (MOX) fuel assemblies (initial plutonium enrichment of 10 wt%) for pressurized-water reactors (PWRs) in the earth for 1000 years without a crash of a fuel bundle. Ten actinide nuclides were chosen, most of which based on "A Guide Introducing Burnup Credit, Preliminary Version," and their compositions were computed with the SWAT code system. Criticality calculations were carried out with the MVP code adopting the computed composition, and the neutron multiplication factor was calculated to be less than 0.9. Issues for consideration were finally summarized.
Tonoike, Kotaro; Miyoshi, Yoshinori; Uchiyama, Gunzo; Watanabe, Shoichi*; Yamamoto, Toshihiro*
Proceedings of 8th International Conference on Nuclear Criticality Safety (ICNC 2007), p.222 - 227, 2007/05
In order to obtain systematic benchmark criticality data concerning dissolving process in a reprocessing plant for LWR spent fuel, a series of critical experiments were performed using heterogeneous cores at the Static Experiment Critical Facility (STACY) in Japan Atomic Energy Agency (JAEA). Focusing on the introduction of the burn-up credit to the process, critical mass measurement was conducted for a combination of uranium fuel rods and uranium solution where pseudo fission product (FP) materials were doped. In this report, the "pseudo FP materials" means elements such as Sm, Cs, Rh and Eu whose isotopic composition is natural but which contains some FP nuclide(s). The result is going to provide basic experimental data for validation of computational methods to evaluate a reactivity effect of each FP material, as well as benchmark criticality data for validation of neutron multiplication factor calculation of heterogeneous systems of spent fuel. In the report, detail of the experiments including a differential reactivity worth curve over the solution level variation is going to be provided as well as the procedure and the result of separate reactivity worth evaluation of each pseudo FP material. Comparison of the experimental result and the computational evaluation will also be presented.
Abe, Hitoshi; Tashiro, Shinsuke; Miyoshi, Yoshinori
Nippon Genshiryoku Gakkai Wabun Rombunshi, 6(1), p.10 - 21, 2007/03
In MOX fuel fabrication facility, zinc stearate will be added into the MOX powder as addition material. If the material is added in large excess by miss operation, criticality characteristics of the MOX fuel would be influenced because it has neutron moderation effect. If criticality condition should be induced by the excess addition, physical variations, such as melting and pyrolysis of the material, must be caused by the fission energy and dynamic characteristics of the MOX fuel must be affected. To contribute quantitative evaluation of the dynamic characteristics, thermal properties data such as exo/endothermic calorific values, reaction rates, etc. with the respective physical variations and release behavior of pyrolysis gas were measured. It was found the exo/endothermic behavior with rinsing temperature of the material could be divided into six regions and rapid pressure rise caused by the pyrolysis reaction over about 400 C. Furthermore, on the basis of the results, evaluation model for the thermal properties under the criticality condition was also investigated.
Okuno, Hiroshi; Yoshiyama, Hiroshi; Miyoshi, Yoshinori
Journal of Nuclear Science and Technology, 43(11), p.1406 - 1413, 2006/11
Single-unit criticality condition data were calculated for homogeneous uranium materials in six chemical forms for revision of the Data Collection section that was attached as the appendix to the Nuclear Criticality Safety Handbook. The calculated criticality condition data were the estimated critical and the estimated lower-limit critical masses and volumes of spheres, diameters of infinitely-long cylinders and infinite slab thicknesses for uranium materials in six chemical forms encountered in criticality safety evaluation of nuclear fuel cycle facilities. The calculation was made with a continuous-energy Monte Carlo criticality calculation code MVP and the Japanese Evaluated Nuclear Data Library JENDL-3.2. The values and precision of the present calculations were discussed in comparison with the literature and the previous results.
Okuno, Hiroshi; Suyama, Kenya; Takahashi, Satoshi*; Watanabe, Shoichi*; Tonoike, Kotaro; Miyoshi, Yoshinori
Transactions of the American Nuclear Society, 95(1), p.283 - 284, 2006/11
no abstracts in English
Yamane, Yuichi; Sakai, Mikio*; Abe, Hitoshi; Yamamoto, Toshihiro*; Okuno, Hiroshi; Miyoshi, Yoshinori
JAEA-Data/Code 2006-021, 75 Pages, 2006/10
Propety data of MOX, Zinc Stearate, etc. were investigated and examined as part of the development for criticality accident evaluation method for MOX fuel fabrication facility. Property data gathered for the powders of MOX, UO, Zinc Stearate, Tungsten and their mixture were density, specific heat, thermal conductivity and etc. as well as the data concerning fluidization or degree of mixing.
Yamamoto, Toshihiro; Miyoshi, Yoshinori; Takeda, Toshikazu*
Journal of Nuclear Science and Technology, 43(1), p.77 - 87, 2006/01
no abstracts in English
Okuno, Hiroshi; Takada, Tomoyuki; Yoshiyama, Hiroshi; Miyoshi, Yoshinori
JAEA-Data/Code 2005-001, 117 Pages, 2005/11
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.
Takahashi, Satoshi*; Okuno, Hiroshi; Miyoshi, Yoshinori
JAERI-Tech 2005-056, 51 Pages, 2005/09
In the heterogeneous system of the mixed oxide fuel of uranium and plutonium, hereafter, MOX fuel, it was investigated whether the system could be modeled as a homogeneous system on the conditions which dealt with the MOX fuel of particle diameter 0.02mm or less in MOX fuel fabrication facilities in Japan. The infinite multiplication factor of the homogeneous system of the MOX fuel was first calculated, and the optimum moderation condition over the each ratio of PuO was determined. It was verified that carried out critical calculation for the heterogeneous system of the MOX fuel in which the spherical fuel diameter in a cube unit cell increased, and an atomic number ratio of hydrogen to heavy metal fixed conditions, and the probability for neutrons to escape resonance by a spherical fuel diameter no less than 0.1mm, and analyzed critical conditions etc. using a contiguous energy Monte Carlo code MVPII and JENDL3.3. The details of these calculations are reported. These results are expected to be quoted in a revised edition of "Nuclear Criticality Safety Handbook."
Sakai, Mikio; Yamamoto, Toshihiro; Murazaki, Minoru; Miyoshi, Yoshinori
Nuclear Technology, 149(2), p.141 - 149, 2005/02
In the conventional criticality evaluation of the nuclear powder system, the effects of particulate behavior have not been considered. In other words, it is difficult to reflect the particle behavior into the conventional criticality evaluation. We have developed a novel criticality evaluation code to resolve this issue. The criticality evaluation code, coupling a Discrete Element Method simulation code with a continuous-energy Monte Carlo transport code, makes it possible to study the effect of the particulate behavior on a criticality evaluation. The criticality evaluation code has been applied to the powder system of the MOX fuel powder agitation process. The criticality evaluations have been performed under mixing the MOX fuel powder in a stirred vessel to investigate the effects of the powder boundary deformation and particulate mixture conditions on the criticality evaluation. The evaluation results revealed that the powder uniformity mixture condition and the boundary deformation could make the neutron effective multiplication factor decrease.