Investigation of the core neutronics analysis conditions for evaluation of burn-up nuclear characteristics of the next-generation fast reactors

Takino, Kazuo; Oki, Shigeo

JAEA-Data/Code 2023-003, 26 Pages, 2023/05

JAEA-Data-Code-2023-003.pdf:1.66MB

Since next-generation fast reactors aim to achieve a higher core discharge burn-up than conventional reactors do, core neutronics design methods must be refined. Therefore, a suitable analysis condition is required for the analysis of burn-up nuclear characteristics to accomplish sufficient estimation accuracy while maintaining a low computational cost. We investigated the effect of the analysis conditions on the accuracy of estimation of the burn-up nuclear characteristics of next-generation fast reactors in terms of neutron energy groups, neutron transport theory, and spatial mesh. This study treated the following burn-up nuclear characteristics: criticality, burn-up reactivity, control rod worth, breeding ratio, assembly-wise power distribution, maximum linear heat rate, sodium void reactivity, and Doppler coefficient for the equilibrium operation cycle. As a result, it was found that the following conditions were the most suitable: 18-energy-group structure, 6 spatial meshes per assembly with diffusion approximation. Additionally, these conditions should apply to correction factors for energy group structure, spatial mesh and transport effects.

Contribution to risk reduction in decommissioning works by the elucidation of basic property of radioactive microparticles (Contract research); FY2018 Center of World Intelligence Project for Nuclear Science/Technology and Human Resource Development

Collaborative Laboratories for Advanced Decommissioning Science; Ibaraki University*

JAEA-Review 2019-041, 71 Pages, 2020/03

JAEA-Review-2019-041.pdf:3.38MB

JAEA/CLADS, has been conducting the Center of World Intelligence Project for Nuclear Science/Technology and Human Resource Development (hereafter referred to "the Project") since FY2018. The Project aims at solving problems in nuclear energy field represented by the decommissioning of the Fukushima Daiichi Nuclear Power Station, Tokyo Electric Power Company Holdings, Inc. For this purpose, intelligence has been collected from all over the world, and basic research and human resource development have been promoted by closely integrating/collaborating knowledge and experiences in various fields beyond the barrier of conventional organizations and research fields. Among the adopted proposals in FY2018, this report summarizes the research results of the "Contribution to Risk Reduction in Decommissioning Works by the Elucidation of Basic Property of Radioactive Microparticles". In order to establish the decommissioning procedures (recovery of the melted fuels, decontamination inside the reactors, ensuring the safety of the workers, etc.) of the Fukushima Daiichi Nuclear Power Station, radioactive microparticles released by the accident are an important information source for clarifying what had happened inside the reactors in the course of the accident. The purpose of the present study is to obtain detailed knowledge on the basic properties (particle size, composition, electrical/optical properties, etc.) of the radioactive microparticles, as well as to further elucidate the various properties of the radioactive microparticles including the quantitative evaluation of alpha-ray-emitters, through the Japan-UK synergetic research. Thus, we are conducting research and development that will contribute to the comprehensive works towards the risk reduction in the "decommissioning" plan.

Mechanical and rheological characteristics of the siliceous mudstone at the Horonobe Underground Research Laboratory site

Hashiba, Kimihiro*; Fukui, Katsunori*; Sugita, Yutaka; Aoyagi, Kazuhei

Proceedings of ITA-AITES World Tunnel Congress 2017 (WTC 2017) (USB Flash Drive), 8 Pages, 2017/06

It is essential to understand the mechanical and rheological characteristics of diatomaceous and siliceous mudstones for the construction of underground structures and for the assessment of their long-term stability. In this study, the siliceous mudstone of the Wakkanai Formation was applied to various laboratory tests: compression test, creep test, relaxation test, drying shrinkage test, and slaking test. The test results showed that water has a major impact on the mechanical and rheological properties of the siliceous mudstone. In addition, water content at a tunnel wall was measured in the Horonobe URL. Comparing the results of the laboratory tests and the in situ measurement, the effect of water on the tunnel stability was discussed.

Analytical results of coolant flow reduction test in the HTTR

Takamatsu, Kuniyoshi; Nakagawa, Shigeaki; Iyoku, Tatsuo

Proceedings of 11th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-11) (CD-ROM), 12 Pages, 2005/10

Safety demonstration tests using the HTTR are in progress to verify the inherent safety features, to improve the safety design and the technologies for High Temperature Gas-cooled Reactors (HTGRs). The coolant flow reduction test by tripping one or two out of three gas circulators is one of the safety demonstration tests. The reactor power safely becomes a stable level without a reactor scram and the temperature transient of the reactor-core is very slow. The SIRIUS code was developed to analyze reactor transient during the tests with reactor dynamics. This paper describes the validation of the SIRIUS code with the measured values of one and two gas circulators tripping test at 30% (9 MW). It was confirmed that the SIRIUS code was able to analyze the reactor transient within 10% during the tests. The result of this study and the way of resolving problems can be applied to development for not only the commercial HTGRs but also the Very High Temperature Reactor (VHTR) as one of the Generation IV reactors.

Heat transfer characteristics evaluation of heat exchangers of mock-up test facility with full-scale reaction tube for HTTR hydrogen production system (Contract research)

Shimizu, Akira; Ohashi, Hirofumi; Kato, Michio; Hayashi, Koji; Aita, Hideki; Nishihara, Tetsuo; Inaba, Yoshitomo; Takada, Shoji; Morisaki, Norihiro; Sakaki, Akihiro*; et al.

JAERI-Tech 2005-031, 174 Pages, 2005/06

JAERI-Tech-2005-031.pdf:20.71MB

Connection of hydrogen production system by steam reforming of methane to the High Temperature Engineering Test Reactor (HTTR) of the Japan Atomic Energy Research Institute (JAERI) has been surveyed until now. Mock-up test facility of this steam reforming system with full-scale reaction tube was constructed in FY 2001, and a lot of operational test data on heat exchanges were obtained in these tests.In this report specifications, structures and heat transfer formulae of steam reformer, steam superheater, steam generator, condenser, helium gas cooler, feed gas heater and feed gas superheater were described. Evaluation codes were newly made to evaluate heat transfer characteristics from measured test data. Overall heat-transfer coefficient obtained from the experimental data were compared and evaluated with the prospective value calculated with heat transfer formulae. As a result, heat transfer performance and thermal efficiency of these heat exchangers were confirmed to be appropriate.

Proposal for evaluation methods of reactor outlet coolant temperature in HTGRs

Takamatsu, Kuniyoshi; Nakagawa, Shigeaki

JAERI-Tech 2005-030, 21 Pages, 2005/05

JAERI-Tech-2005-030.pdf:1.06MB

The High Temperature engineering Test Reactor (HTTR) is a graphite moderated and gas cooled reactor with the thermal power of 30MW and the reactor outlet coolant temperature of 850C/950C. Rise-to-power test in the HTTR was performed from March 31th to May 1st in 2004 as phase 5 test up to 30MW in the high temperature test operation mode. It was confirmed that the thermal reactor power and the reactor outlet coolant temperature reached to 30MW and 950C respectively on April 19th. Achievement of the reactor outlet coolant temperature of 950C is the first time in Japan as well as the world. This report describes proposal for evaluation methods of reactor outlet coolant temperature in the HTGRs through the HTTR operation experiments. The equation is derived from relationships among PRM reading values, reactor outlet coolant temperature, reactor thermal power and heat removal by VCS. The deliberation processes in this study will be applicable to the research and developments of HTGRs in the future.

Rise-to-power test in high temperature engineering test reactor in the high temperature test operation mode; Test progress and summary of test results up to 30MW of reactor thermal power

Takamatsu, Kuniyoshi; Nakagawa, Shigeaki; Sakaba, Nariaki; Takada, Eiji*; Tochio, Daisuke; Shimakawa, Satoshi; Nojiri, Naoki; Goto, Minoru; Shibata, Taiju; Ueta, Shohei; et al.

JAERI-Tech 2004-063, 61 Pages, 2004/10

JAERI-Tech-2004-063.pdf:3.14MB

The High Temperature engineering Test Reactor (HTTR) is a graphite moderated and gas cooled reactor with the thermal power of 30MW and the reactor outlet coolant temperature of 850C/950C. Rise-to-power test in the HTTR was performed from March 31th to May 1st in 2004 as phase 5 test up to 30MW in the high temperature test operation mode. It was confirmed that the thermal reactor power and the reactor outlet coolant temperature reached to 30MW and 950C respectively on April 19th in the single operation mode using only the primary pressurized water cooler. The parallel loaded operation mode using the intermediate heat exchanger and the primary pressurized water cooler was performed from June 2nd and JAERI (Japan Atomic Energy Research Institute) obtained the certificate of the pre-operation test on June 24th from MEXT (Ministry of Education Culture Sports Science and Technology) after all the pre-operation tests were passed successfully in the high temperature test operation mode. Achievement of the reactor-outlet coolant temperature of 950C is the first time in the world. It is possible to extend highly effective power generation with a high-temperature gas turbine and produce hydrogen from water with a high-temperature. This report describes the results of the high-temperature test operation of the HTTR.

Evaluation of neutronic characteristics of TRACY water-reflected core system

Sono, Hiroki; Yanagisawa, Hiroshi*; Miyoshi, Yoshinori

JAERI-Tech 2003-096, 84 Pages, 2004/01

JAERI-Tech-2003-096.pdf:3.6MB

Prior to the supercritical experiments using a water-reflected core of the TRACY Facility, neutronic characteristics regarding criticality and reactivity of the core system were evaluated. In the analyses, a continuous energy Monte Carlo code, MVP, and a two-dimensional transport code, TWOTRAN, were used together with a nuclear data library, JENDL-3.3. By comparison to the characteristics in the former-used bare core system of TRACY, the water reflector was estimated not to change the kinetic parameter and to reduce the critical solution level by 20 %, the temperature coefficient of reactivity by 610 %, and the void coefficient of reactivity by 18 %, respectively. According to the Nordheim-Fuchs model, the first peak power during a power excursion was evaluated to be 15 % smaller than that in the bare system under the same conditions of fuel and reactivity insertion. The influence of the void feedback effect of reactivity, which is left out of consideration in the model, on the power characteristics will be evaluated from the results of the experiments.

Experimental results obtained with the simulated cold moderator system; System characteristics and technical issues

Aso, Tomokazu; Kaminaga, Masanori; Haga, Katsuhiro; Kinoshita, Hidetaka; Takahashi, Toshio*; Hino, Ryutaro

JAERI-Tech 2002-096, 76 Pages, 2002/12

JAERI-Tech-2002-096.pdf:7.2MB

The JAERI and the High Energy Accelerator Research Organization have been developing a Mega-Watt scale spallation target system. In the system, neutrons generated in a target are sorted out their energy to the proper values in liquid-hydrogen moderators. Then, the liquid-hydrogen is forced to circulate in order to suppress hydrogen temperature increase. In the operation of moderators, it is very important to establish a safety protection system against emergency shutdown of the accelerator or accidents of the cold moderator system. In order to obtain a technical data for design and safety review of the liquid-hydrogen system, we have fabricated an experimental apparatus simulated the cold moderator system using liquid nitrogen (max. 1.5MPa, mini. 77K) instead of liquid hydrogen. The experiments on a controllability of the system were carried out to investigate dynamic characteristics of the system. This report presents the experimental results and technical issues for the construction of a practical liquid-hydrogen moderator system of the Mega-Watt scale target system.

Rise-to-power test in High Temperature Engineering Test Reactor; Test progress and summary of test results up to 30MW of reactor thermal power

Nakagawa, Shigeaki; Fujimoto, Nozomu; Shimakawa, Satoshi; Nojiri, Naoki; Takeda, Takeshi; Saikusa, Akio; Ueta, Shohei; Kojima, Takao; Takada, Eiji*; Saito, Kenji; et al.

JAERI-Tech 2002-069, 87 Pages, 2002/08

JAERI-Tech-2002-069.pdf:10.12MB

Rise-to-power test in the HTTR has been performed from April 23rd to June 6th in 2000 as phase 1 test up to 10MW, from January 29th to March 1st in 2001 as phase 2 test up to 20MW in the rated operation mode and from April 14th to June 8th in 2001 as phase 3 test up to 20MW in the high temperature test operation mode. Phase 4 test to achieve the thermal reactor power of 30MW started from October 23rd in 2001. On December 7th it was confirmed that the thermal reactor power reached to 30MW and the reactor outlet coolant temperature reached to 850C. JAERI obtained the certificate of pre-operation test from MEXT because all the pre-operation tests by MEXT were passed successfully. From the test results of rise-up-power test up to 30MW, the performance of reactor and cooling system were confirmed, and it was confirmed that an operation of reactor facility could be performed safely. Some problems to be solved were found through tests. By means of solving them, the reactor operation with the reactor outlet coolant temperature of 950C will be achievable.