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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.
Takemoto, Noriyuki; Imaizumi, Tomomi; Kimura, Nobuaki; Tsuchiya, Kunihiko; Hori, Junichi*; Sano, Tadafumi*; Nakajima, Ken*
Proceedings of International Conference on the Physics of Reactors; The Role of Reactor Physics toward a Sustainable Future (PHYSOR 2014) (CD-ROM), 11 Pages, 2014/09
Neutronic evaluations in JMTR have been performed for irradiation tests by Monte Carlo method with thermal neutron scattering law, S(
, 
), data for beryllium metal, etc. The calculation accuracy of fast and thermal neutron fluxes are 
10% and 30%, respectively. Analytical and experimental investigations to achieve higher calculation accuracy, especially for the thermal neutron flux up to the fast neutron flux level, have been therefore performed to offer higher value data technically to the JMTR users. In order to investigate an effect of fabrication method of beryllium material on the calculation accuracy, total cross-sections of beryllium specimens were measured using KURRI-LINAC, and it was found that the total cross-section was different from the theoretical one, and depended on the crystal texture, etc. The S(, ) was adjusted based on the measured data, and the applicability to the neutronic evaluation in the JMTR was verified.
Kugo, Teruhiko; Sugino, Kazuteru; Uematsu, Mari Mariannu; Numata, Kazuyuki*
Proceedings of International Conference on the Physics of Reactors; The Role of Reactor Physics toward a Sustainable Future (PHYSOR 2014) (CD-ROM), 12 Pages, 2014/09
The present paper summarizes calculation results for an international benchmark proposed under the framework of the Working Party on scientific issues of Reactor Systems (WPRS) of the Nuclear Energy Agency of the OECD. It focuses on the large size oxide-fueled SFR. Library effect for core performance characteristics and reactivity feedback coefficients is analyzed using sensitivity analysis. The effect of ultra-fine energy group calculation in effective cross section generation is also analyzed. The discrepancy is about 0.4% for a neutron multiplication factor by changing JENDL-4.0 with JEFF-3.1. That is about -0.1% by changing JENDL-4.0 with ENDF/B-VII.1. The main contributions to the discrepancy between JENDL-4.0 and ENDF/B-VII.1 are 
Pu capture, 
U inelastic scattering and 
Pu fission. Those to the discrepancy between JENDL-4.0 and JEFF-3.1 are 
Na inelastic scattering, 
Fe inelastic scattering, Pu fission, Pu capture, Pu fission, U inelastic scattering, Pu fission and Pu nu-value. As for the sodium void reactivity, JEFF-3.1 and ENDF/B-VII.1 underestimate by about 8% compared with JENDL-4.0. The main contributions to the discrepancy between JENDL-4.0 and ENDF/B-VII.1 are Na elastic scattering, Na inelastic scattering and Pu fission. That to the discrepancy between JENDL-4.0 and JEFF-3.1 is Na inelastic scattering. The ultra-fine energy group calculation increases the sodium void reactivity by 2%.
Yokoyama, Kenji
Proceedings of International Conference on the Physics of Reactors; The Role of Reactor Physics toward a Sustainable Future (PHYSOR 2014) (CD-ROM), 11 Pages, 2014/09
A burnup sensitivity analysis solver based on generalized perturbation theory has been developed in a multi-purpose analysis framework MARBLE. The new solver has capability to calculate sensitivity coefficients based on the diffusion theory for transmuted nuclear fuel composition after burnup, i.e. atomic number density, with respect to nuclear data not only in 2-dimensional R-Z model but also in 3-dimensional Hexagonal-Z model. In the present paper, a new numerical verification method named component-wise direct calculation is proposed to check the results of each term. A numerical experiment of the new verification method was performed, and it shows that the new verification method is valid. In addition, a sample application result is shown in order to evaluate calculation model effects due to difference between 2- and 3-dimensional models.
Sugawara, Takanori; Iwamoto, Hiroki; Nishihara, Kenji; Tsujimoto, Kazufumi; Sasa, Toshinobu; Oigawa, Hiroyuki
Proceedings of International Conference on the Physics of Reactors; The Role of Reactor Physics toward a Sustainable Future (PHYSOR 2014) (CD-ROM), 8 Pages, 2014/09
The Japan Atomic Energy Agency (JAEA) has the plan to construct Transmutation Physic Experimental Facility (TEF-P) under a framework of J-PARC (Japan Proton Accelerator Research Complex) project. TEF-P is a critical assembly which can load Minor Actinide (MA) fuels to perform reactor physics experiments for transmutation systems such as Accelerator-Driven System (ADS) or Fast Reactor (FR). The facility can also use proton beam from the J-PARC accelerator to investigate the controllability of ADS. Cur-rent status and activities for TEF-P are described.
Takano, Kazuya; Mori, Tetsuya; Kishimoto, Yasufumi; Hazama, Taira
Proceedings of International Conference on the Physics of Reactors; The Role of Reactor Physics toward a Sustainable Future (PHYSOR 2014) (CD-ROM), 13 Pages, 2014/09
This paper describes details of the IAEA/CRP benchmark calculation by JAEA on the control rod withdrawal test in the Phenix End-of-Life Experiments. The power distribution deviation by the control rod insertion/withdrawal, which is the major target of the benchmark, is well simulated by calculation. In addition to the CRP activities, neutron and photon transport effect is evaluated in the nuclear heating calculation of the benchmark analysis. It is confirmed that the neutron and photon transport effect contributes to the improvement of the absolute power calculation results in the breeder blanket region.
Kora, Kazuki*; Nakaya, Hiroyuki*; Kubo, Kotaro*; Matsuura, Hideaki*; Shimakawa, Satoshi; Goto, Minoru; Nakagawa, Shigeaki
Proceedings of International Conference on the Physics of Reactors; The Role of Reactor Physics toward a Sustainable Future (PHYSOR 2014) (CD-ROM), 12 Pages, 2014/09
In this study, the capability of HTGR as LLFP transmuter was evaluated in terms of neutron economy. Considering gas turbine high-temperature reactor with 300 MWe nominal capacity (GTHTR300) as HTGR, transmutations of four types of LLFP nuclide were estimated using Monte Carlo transport code MVP and ORIGEN. In addition, burn-up simulations for whole-core region were carried out using MVP-BURN. It was numerically shown that the neutron fluxes change significantly depending on the arrangement of LLFP in the core. When 15 t of LLFP is placed in an ideal manner, the GTHTR300 can sustain sufficient reactivity for one year while transmuting up to 30 kg per year. Additionally, there are more space available for storing larger amount of LLFP without affecting the reactivity. These results suggest that there is a possibility of using GTHTR300 as both LLFP storage and transmuter.
Takeda, Toshikazu*; Hazama, Taira; Fujimura, Koji*; Sawada, Shusaku*
Proceedings of International Conference on the Physics of Reactors; The Role of Reactor Physics toward a Sustainable Future (PHYSOR 2014) (CD-ROM), 15 Pages, 2014/09
A national project started in 2013 in Japan entitled "technology development for the environmental burden reduction". The present study is one of the studies adopted as the national project. We are aiming to develop MA transmutation core concepts harmonizing MA transmutation performance with core safety and to improve design accuracy related to MA transmutation performance. To validate and improve design accuracy of the high safety and high MA transmutation performance of SFR cores, we develop methods for calculating the transmutation rate of individual MA nuclides and estimating uncertainty of MA transmutation by using burnup sensitivity. Also we develop reliable reactor physics database to reduce the uncertainty of MA transmutation calculations. The overall consistency of the measured data is investigated by evaluating the usefulness of conventional static data as well as those related to MA transmutation obtained from various facilities like Monju, Joyo, FCA, BFS and PFR.
Pascal, V.*; Prulhi
re, G.*; Vanier, M.*; Fontaine, B.*; Devan, K.*; Chellapandi, P.*; Kriventsev, V.*; Monti, S.*; Mikityuk, K.*; Chenu, A.*; et al.
Proceedings of International Conference on the Physics of Reactors; The Role of Reactor Physics toward a Sustainable Future (PHYSOR 2014) (CD-ROM), 16 Pages, 2014/09
no abstracts in English
Kojima, Kensuke; Okumura, Keisuke
Proceedings of International Conference on the Physics of Reactors; The Role of Reactor Physics toward a Sustainable Future (PHYSOR 2014) (CD-ROM), 9 Pages, 2014/09
The MOSRA system has been developing to improve the applicability of the neutronic characteristic analyses. The cell calculation module MOSRA-SRAC is a core module of MOSRA, and applicability tests for realistic problems are required. As a test, we joined the benchmark "Burnup Credit Criticality Benchmark Phase IIIC." The benchmark requested the neutronic characteristics for a BWR fuel assembly with gadolinium, which had been used in the TEPCO's Fukushima Daiichi Nuclear Power Station. Because of a restriction of MOSRA-SRAC, the geometry was partially homogenized. To verify the module's applicability including the homogenization effects, the multiplication factor and the nuclide compositions were compared with the well-validated code MVP-BURN. As the results, the applicability of MSORA-SRAC for the assembly was verified. Additionally, it was also shown that the homogenization effects were smaller than the difference due to the calculation methods.
pin cell geometryNagaya, Yasunobu
Proceedings of International Conference on the Physics of Reactors; The Role of Reactor Physics toward a Sustainable Future (PHYSOR 2014) (CD-ROM), 13 Pages, 2014/09
Monte Carlo analysis has been performed to investigate the impact of the exact resonance elastic scattering model on the Doppler reactivity coefficient for the UO pin cell geometry with the parabolic temperature profile. As a result, the exact scattering model affects the coefficient similarly for both the flat and parabolic temperature profiles; it increases the contribution of uranium-238 (U) resonance capture in the energy region from 
16 eV to 150 eV and does uniformly in the radial direction. Then the following conclusions hold for both the exact and asymptotic resonance scattering models. The Doppler reactivity coefficient is well reproduced with the definition of the effective fuel temperature (equivalent flat temperature) proposed by Grandi et al.
Uematsu, Mari Mariannu; Kugo, Teruhiko; Numata, Kazuyuki*
Proceedings of International Conference on the Physics of Reactors; The Role of Reactor Physics toward a Sustainable Future (PHYSOR 2014) (CD-ROM), 15 Pages, 2014/09
In the frame work of the working party on reactor and system (WPRS) of the OECD/NEA, the benchmark on SFR was conducted. Within the OECD/NEA/WPRS benchmark, study on medium size metallic fuel core was performed using a code system for fast reactor core calculation with deterministic method MARBLE and with a Monte Carlo method MVP. The latest nuclear library JENDL-4.0 is used for evaluation of eigenvalues (k
) and reactivity (sodium void, Doppler and control rod worth) calculations. Depletion calculations are conducted using MARBLE/BURNUP with deterministic method for flux calculation and MVP-BURN with Monte Carlo method. The analysis results and discrepancies between different analysis methods are summarized in this paper. Sensibility studies of eigenvalue and sodium void reactivity of the medium size metallic fuel benchmark core are also conducted to determine the main reactions contributing to the difference between JENDL-4.0 and other libraries JEFF-3.1 and ENDF/B-VII.
Buiron, L.*; Rimpault, G*; Fontaine, B.*; Kim, T. K.*; Stauff, N. E.*; Taiwo, T. A.*; Yamaji, Akifumi*; Gulliford, J.*; Fridmann, E.*; Pataki, I.*; et al.
Proceedings of International Conference on the Physics of Reactors; The Role of Reactor Physics toward a Sustainable Future (PHYSOR 2014) (CD-ROM), 16 Pages, 2014/09
Within the activities of the Working Party on Scientific Issues of Reactor Systems (WPRS) of the OECD, an international collaboration is ongoing on the neutronic analyses of several Generation-IV Sodium-cooled Fast Reactor (SFR) concepts. This paper summarizes the results obtained by participants from institutions of different countries (ANL, CEA, ENEA, HZDR, JAEA, CER, KIT, UIUC) for the large core numerical benchmarks. These results have been obtained using different calculation methods and analysis tools to estimate the core reactivity and isotopic composition evolution, neutronic feedbacks and power distribution. For the different core concepts analyzed, a satisfactory agreement was obtained between participants despite the different calculation schemes used. A good agreement was generally obtained when comparing compositions after burnup, the delayed neutron fraction, the Doppler coefficient, and the sodium void worth. However, some noticeable discrepancies between the k-effective values were observed and are explained in this paper. These are mostly due to the different neutronic libraries employed (JEFF3.1, ENDFB7.0 or JENDL-4.0) and to a lesser extent the calculations methods.
Iwamoto, Hiroki; Mathias, M.*; Nishihara, Kenji
no journal, ,
