Evaluation of heat removal during the failure of the core cooling for new critical assembly

Eguchi, Yuta; Sugawara, Takanori; Nishihara, Kenji; Tazawa, Yujiro; Tsujimoto, Kazufumi

Proceedings of 26th International Conference on Nuclear Engineering (ICONE-26) (Internet), 8 Pages, 2018/07

https://doi.org/10.1115/ICONE26-82012

In order to investigate the basic neutronics characteristics of the accelerator-driven subcritical system (ADS), JAEA has a plan to construct a new critical assembly in the J-PARC project, Transmutation Physics Experimental Facility (TEF-P). This study aims to evaluate the natural cooling characteristics of TEF-P core which has large decay heat by minor actinide (MA) fuel, and to achieve a design that does not damage the core and the fuels during the failure of the core cooling system. In the evaluation of the TEF-P core temperature, empty rectangular lattice tube outer of the core has a significant effect on the heat transfer characteristics. The experiments by using the mockup device were performed to validate the heat transfer coefficient and experimental results were obtained. By using the obtained experimental results, the three-dimensional heat transfer analysis of TEF-P core were performed, and the maximum core temperature was obtained, 294C. This result shows TEF-P core temperature would be less than 327C that the design criterion of temperature.

Fabrication and test results of testing equipment for remote-handling of MA fuel, 2; Evaluation of heat transfer parameter crossing rectangular lattice matrix

Eguchi, Yuta; Sugawara, Takanori; Nishihara, Kenji; Tazawa, Yujiro; Inoue, Akira; Tsujimoto, Kazufumi

JAEA-Technology 2015-052, 34 Pages, 2016/03

JAEA-Technology-2015-052.pdf:5.02MB

Transmutation Physics Experimental Facility (TEF-P) planned in the J-PARC project uses minor actinide (MA) fuel which has large decay heat. So it is necessary to consider the increase of the core temperature when the core cooling system is stopped. This change of the core temperature was evaluated by thermal conduction analysis. It was found that the impact of thermal insulation in the empty rectangular lattice matrix area was large. So it is necessary to verify reliability and accuracy of heat transfer effect used in this area. Testing equipment was fabricated to verify the accuracy of calculation model for the empty lattice matrix which was the free convection model of sealed fluid. By using this equipment, thermal distribution and one dimensional heat flow through the lattice were measured. It was observed that the actual equivalent thermal conductivity in the lattice was larger than the free convection model. It was also confirmed that the insertion of the aluminum block into the empty lattice could achieve the higher equivalent thermal conductivity. These results could be the useful data for the thermal conduction analysis.

Conceptual study of transmutation experimental facility, 5; Investigation of MA fuel handling

Sugawara, Takanori; Nishihara, Kenji; Sasa, Toshinobu; Tsujimoto, Kazufumi; Tazawa, Yujiro; Oigawa, Hiroyuki

JAEA-Technology 2014-044, 59 Pages, 2015/03

JAEA-Technology-2014-044.pdf:14.46MB

Transmutation Physics Experimental Facility (TEF-P) planned in the J-PARC is a critical assembly with low thermal output and it will treat large amount of highly-radioactive minor actinide (MA) fuels in the experiments. Handling of the MA fuels in each stage of storage, transport and loading/unloading to the core was conceptually investigated, then, criticality, dose and cooling performance were assessed. For the criticality, it was shown that the effective multiplication factors in each step, storage, transport and loading, were sufficiently lower than 1.0. For the dose, the dose for workers will be reduced by installing remote handling devices to treat the MA fuels. For the cooling performance, it was found that the temperature of the core would be kept low in the normal operation. On the other hand, in the case which the air conditioning or the blower for the core stopped for long period, it was shown that there would be a possibility of the MA fuel failure.

A Small-sized HTGR system design for multiple heat applications for developing countries

Ohashi, Hirofumi; Sato, Hiroyuki; Goto, Minoru; Yan, X.; Sumita, Junya; Tazawa, Yujiro*; Nomoto, Yasunobu; Aihara, Jun; Inaba, Yoshitomo; Fukaya, Yuji; et al.

International Journal of Nuclear Energy, 2013, p.918567_1 - 918567_18, 2013/00

https://doi.org/10.1155/2013/918567

Japan Atomic Energy Agency (JAEA) has conducted a conceptual design of a 50 MWt small-sized high temperature gas cooled reactor (HTGR) for multiple heat applications, named HTR50S, with the reactor outlet coolant temperature of 750 C and 900 C. It is first-of-a-kind of the commercial plant or a demonstration plant of a small-sized HTGR system to deploy it in developing countries in the 2020s. The design concept of HTR50S is to satisfy the user requirements for multipurpose heat application, to upgrade its performance compared to that of HTTR without significant R&D utilizing the knowledge obtained by the HTTR design and operation, and to fulfill the high level of safety by utilizing the inherent features of HTGR and a passive decay heat removal system.

Core design and safety analyses of 600 MWt, 950C high temperature gas-cooled reactor

Nakano, Masaaki*; Takada, Eiji*; Tsuji, Nobumasa*; Tokuhara, Kazumi*; Ohashi, Kazutaka*; Okamoto, Futoshi*; Tazawa, Yujiro; Tachibana, Yukio

Proceedings of 6th International Topical Meeting on High Temperature Reactor Technology (HTR 2012) (USB Flash Drive), 6 Pages, 2012/10

The conceptual core design study of High Temperature Gas-cooled Reactor (HTGR) is performed. The major specifications are 600 MW thermal output, 950C outlet coolant temperature, prismatic core type, enriched uranium fuel. The decay heat in the core can be removed with only passive measures, for example, natural convection reactor cavity cooling system (RCCS), even if any electricity is not supplied (station blackout). The transient thermal analysis of the depressurization accident in the case the primary coolant decreases to the atmosphere pressure shows that the fuels and the reactor pressure vessel temperatures are kept under their safety limit criteria. The fission product release, Ag and Cs from the fuels under the normal operation is small as to make maintenance of devices in the primary cooling system, such as a gas turbine, without remote maintenance. The HTGRs can achieve the advanced safety features based on their inherent passive safety characteristics.

Estimation of acceptable beam trip frequencies of accelerators for ADS and comparison with experimental data of accelerators

Takei, Hayanori; Nishihara, Kenji; Tsujimoto, Kazufumi; Furukawa, Kazuro*; Yano, Yoshiharu*; Ogawa, Yujiro*; Oigawa, Hiroyuki

Proceedings of International Topical Meeting on Nuclear Research Applications and Utilization of Accelerators (CD-ROM), 9 Pages, 2009/05

Frequent beam trips as experienced in existing high power proton accelerators may cause thermal fatigue problems in ADS components which may lead to degradation of their structural integrity and reduction of their lifetime. Thermal transient analyses were performed to investigate the effects of beam trips on the reactor components, with the objective of formulating ADS design that had higher engineering possibilities and determining the requirements for accelerator reliability. These analyses were made on the thermal responses of four parts of the reactor components; the beam window, the cladding tube, the inner barrel and the reactor vessel. Our results indicated that the acceptable frequency of beam trips ranged from 43 to times per year depending on the beam trip duration to keep the plant availability 70%. In order to consider measures to reduce the frequency of beam trips on the high power accelerator for ADS, we compared the acceptable frequency of beam trips with the operation data of existing accelerators. The result of this comparison showed that the beam trip frequency for durations of 10 seconds or less was within the acceptable level, while that exceeding five minutes should be reduced to about 1/35 to satisfy the plant availability conditions.

Basic principles on the safety evaluation of the HTGR hydrogen production system

Ohashi, Kazutaka*; Nishihara, Tetsuo; Tazawa, Yujiro; Tachibana, Yukio; Kunitomi, Kazuhiko

JAEA-Technology 2008-093, 56 Pages, 2009/03

JAEA-Technology-2008-093.pdf:2.31MB

As HTGR hydrogen production systems, such as HTTR-IS system or GTHTR300C currently being developed by Japan Atomic Energy Agency, consists of nuclear reactor and chemical plant, which are without a precedent in the world, safety design philosophy and regulatory framework should be newly developed. In this report, phenomena to be considered and events to be postulated in the safety evaluation of the HTGR hydrogen production systems were investigated and basic principles to establish acceptance criteria for the explosion and toxic gas release accidents were provided. Especially for the explosion accident, quantitative criteria to the reactor building are proposed with relating sample calculation results. It is necessary to treat abnormal events occurred in the hydrogen production system as external events to the nuclear plant in order to classify the hydrogen production system as no-nuclear facility and basic policy to meet such requirement was also provided.

Conceptual core design study of the Very High Temperature gas-cooled Reactor (VHTR); Upgrading the core performance by using multihole-type fuel

Ohashi, Kazutaka; Nishihara, Tetsuo; Kunitomi, Kazuhiko; Nakano, Masaaki*; Tazawa, Yujiro*; Okamoto, Futoshi*

Nihon Genshiryoku Gakkai Wabun Rombunshi, 7(1), p.32 - 43, 2008/03

https://doi.org/10.3327/taesj.J07.005

Interests on the development of the Very High-Temperature Gas-Cooled Reactor (VHTR), of which the reactor outlet temperature is 950C or much higher, are recently increasing world-widely and it was selected as one of the candidate reactor types of the GIF. Japan Atomic Energy Agency has already initiated R&D efforts on the electricity and hydrogen co-generation plant with VHTR system, GTHTR300C. Although technical feasibility of its VHTR reactor using Pin-in-block fuel, which has experience to be already used in the HTTR, has been shown fundamentally, more improvements of the core performances, such as decrease of the occupational exposure doses during the plant maintenance, are desired. This report presents the results of the conceptual core design study using Multi-hole type fuel and the study on the occupational exposure doses. The latter results shows much better plant maintainability compared to the previous results of the GTHTR-300.

Research and development program on accelerator driven subcritical system in JAERI

Tsujimoto, Kazufumi; Oigawa, Hiroyuki; Ouchi, Nobuo; Kikuchi, Kenji; Kurata, Yuji; Mizumoto, Motoharu; Sasa, Toshinobu; Nishihara, Kenji; Saito, Shigeru; Umeno, Makoto*; et al.

Proceedings of International Conference on Nuclear Energy System for Future Generation and Global Sustainability (GLOBAL 2005) (CD-ROM), 6 Pages, 2005/10

The Japan Atomic Energy Research Institute (JAERI) has been proceeding with the research and development (R&D) on accelerator-driven subcritical system (ADS). The ADS proposed by JAERI is a lead-bismuth (Pb-Bi) eutectic cooled fast subcritical core with 800 MWth. To realize such an ADS, some technical issues should be studied, developed and demonstrated. JAERI has started a comprehensive R&D program since the fiscal year of 2002 to acquire knowledge and elemental technology that are necessary for the validation of engineering feasibility of the ADS. The first stage of the program had been continued for three years. The program is conducted by JAERI, and many institutes, universities and private companies were involved. Items of R&D are concentrated on three technical areas peculiar to the ADS: (1) superconducting linear accelerator (SC-LINAC), (2) Pb-Bi eutectic as spallation target and core coolant, and (3) subcritical core design and technology. In the present work, the outline and the results in the first stage of the program are reported.

Conceptual study of transmutation experimental facility, 3; Study on transmutation physics experiment facility

Oigawa, Hiroyuki; Maekawa, Fujio; Meigo, Shinichiro; Kasugai, Yoshimi; Tsujimoto, Kazufumi; Nishihara, Kenji; Sasa, Toshinobu; Konno, Chikara; Kai, Tetsuya; Ikeda, Yujiro

JAERI-Tech 2002-037, 220 Pages, 2002/03

JAERI-Tech-2002-037.pdf:11.09MB

no abstracts in English