Development of simple method to incorporate out-of-core cooling effect on thorium conversion in multi-pass fueled reactor and investigation on characteristics of the effect

Fukaya, Yuji

Annals of Nuclear Energy, 81, p.301 - 305, 2015/07

https://doi.org/10.1016/j.anucene.2015.03.005

Development of a simple method to incorporate the out-of-core cooling effect on the thorium conversion in multi-pass fueled reactors and investigation on characteristics of the effect have been performed. For multi-pass fueled reactors, such as Molten Salt Breeder Reactor (MSBR) and Pebble-Bed Modular Reactor (PBMR), fuel moves in the core and exits from the core. The nuclides decay also out of the core, and it should be also considered if it is important for core characteristics. In the present study, Pa is considered to evaluate the thorium conversion accurately. To take the effect into account, in the present study, an effective decay constant is proposed to make equilibrium concentration of Pa without out-of-core cooling equal to that of out-of-core cooling. With the effective decay constant, the out-of-core cooling effect can be incorporated even with the code system using macroscopic cross sections generated by cell burn-up calculations without any code modification. In addition, the characteristic of out-of-core cooling effect for the thorium conversion is evaluated for thorium fueled reactors of MSBR and PBMR. It is concluded that the out-of-core cooling effect is suitable for MSBR to enhance thorium conversion because of the fast flow rate of fuel salt. On the other hand, the effect is not important and not realistic to employ for PBMR because the in-core residence time of approximately 100 days is longer than the half-life of Pa of 27.0 days, and the effect cannot improve the conversion ratio drastically.

Advances of study on thermal/hydraulic performance in tight-lattice rod bundles for reduced-moderation water reactors

Onuki, Akira; Takase, Kazuyuki; Kureta, Masatoshi*; Yoshida, Hiroyuki; Tamai, Hidesada; Liu, W.; Nakatsuka, Toru; Akimoto, Hajime

Proceedings of 13th International Conference on Nuclear Engineering (ICONE-13) (CD-ROM), 8 Pages, 2005/05

R&D project to investigate thermal-hydraulic performance in tight-lattice rod bundles for Reduced-Moderation Water Reactor (RMWR) is started at Japan Atomic Energy Research Institute in collaboration with power company, reactor vendors, universities since 2002. The RMWR can attain the favorable characteristics such as effective utilization of uranium resources, multiple recycling of plutonium, high burn-up and long operation cycle, based on matured LWR technologies. The confirmation of thermal-hydraulic feasibility is one of the most important R&D items for the RMWR because of the tight-lattice configuration. In this paper, we will show the R&D plan and describe some advances on experimental and analytical studies. The experimental study is performed mainly using large-scale (37-rod bundle) test facility and the analytical one aims to develop a predictable technology for geometry effects such as gap between rods etc. using advanced 3-D two-phase flow simulation methods. Steady-state and transient critical power experiments are conducted with the test facility (Gap width between rods: 1.3mm and 1.0mm) and the experimental data reveal the feasibility of RMWR.

Designing of reduced-moderation water reactor and related development issues

Okubo, Tsutomu

Konsoryu, 17(3), p.228 - 235, 2003/09

The Reduced-Moderation Water Reactor is supposed to realize plutonium multiple recycling, and furthermore, plutonium breeding cycle, based on the well-established Light Water Reactor technologies. In the present paper, the overview of the design study is presented and the related R&D issues are introduced, especially focusing on the thermal hydraulic activities.

Development of Reduced-Moderation Water Reactor (RMWR) for sustainable energy supply

Iwamura, Takamichi; Okubo, Tsutomu; Kureta, Masatoshi; Nakatsuka, Toru; Takeda, Renzo*; Yamamoto, Kazuhiko*

Proceedings of 13th Pacific Basin Nuclear Conference (PBNC 2002) (CD-ROM), 7 Pages, 2002/10

In order to ensure sustainable energy supply in Japan, the reduced-moderation water reactor (RMWR) has been developed by JAERI since 1998. MOX fuel assemblies with tight lattice arrangement are used to increase the conversion ratio. In order to establish negative void reactivity coefficient, the core should be short and flat to increase neutron leakage from the core. The core designs were accomplished to a large core with 1,356MWe and a small core with 330MWe. For both cores, negative void coefficient and natural circulation cooling of the core were realized. To confirm thermal-hydraulic feasibility, critical heat flux experiments were performed using 7-rod bundles with the gap width of 1mm and 1.3mm. The results indicated that enough cooling was assured for the tight lattice core. Further R&D studies, including large scale thermal-hydraulic experiments, reactor physics experiments, development of high burn-up fuel cladding material and simplified reprocessing technology, are necessary to realize commercial introduction of RMWR by 2020's for the replacement of current generation LWRs.

Conversion of JRR-4 fuel to LEU

Watanabe, Shukichi; Nakajima, Teruo; Kaieda, Keisuke

JAERI-Conf 99-006, p.119 - 124, 1999/08

https://jopss.jaea.go.jp/pdfdata/JAERI-Conf-99-006.pdf

no abstracts in English

Measurement of flux tilt and eigenvalue separation in axially decoupled core

Ando, Masaki; *; Nishina, Kojiro*; *

Journal of Nuclear Science and Technology, 34(5), p.445 - 453, 1997/05

https://doi.org/10.3327/jnst.34.445

no abstracts in English

Performance evaluations of a high conversion boiling water reactor core with an axially heterogeneous core concept

*; Okumura, Keisuke;

JAERI-M 92-067, 35 Pages, 1992/05

JAERI-M-92-067.pdf:1.06MB

no abstracts in English

Conceptual design study of high conversion light water reactor

Okumura, Keisuke; ; Mori, Takamasa; Nakakawa, Masayuki;

JAERI-M 90-096, 169 Pages, 1990/06

JAERI-M-90-096.pdf:5.02MB

no abstracts in English

Thermal-Hydrauluc Analysis of Loss-of-Coolant Accident in the JMTR

;

JAERI-M 85-001, 33 Pages, 1985/02

JAERI-M-85-001.pdf:0.74MB

no abstracts in English