Summary of the 5th Workshop on the Reduced Moderation Water Reactor; March 8, 2002, JAERI, Toaki

Nakano, Yoshihiro; Ishikawa, Nobuyuki; Nakatsuka, Toru; Iwamura, Takamichi

JAERI-Conf 2002-012, 219 Pages, 2002/12

JAERI-Conf-2002-012.pdf:17.4MB

no abstracts in English

Neutronic study of SCR core for under-sea scientific research vessel

Odano, Naoteru; Ishida, Toshihisa; Wada, Koji*; Imai, Hiroshi*

JAERI-Research 2001-039, 59 Pages, 2001/07

JAERI-Research-2001-039.pdf:4.47MB

A very small reactor, SCR (Submersible Compact Reactor), whose thermal output is 1250 kW, is an integral-pressurized type reactor to be used as a power source for a scientific research vessel in medium depth region of the Arctic Ocean. Neutronic study has been carried out for design of the SCR core of which could achieve continuous long-term operation without refueling for 10 years considering 50 % of load factor of the core. In the present study, arrangement of fuel rods, U enrichment of UO fuel rods and reflector materials were surveyed. The U enrichment has been determined to be 9.5 wt% to satisfy design criteria. In the present study Be metal was adopted as a reflector material. Reactor physics parameters including reactivity coefficients and power distributions were evaluated for the determined core specifications. Reactor physics parameters related to core safety were also confirmed and the evaluated parameters indicated that the determined core specifications in this study satisfied design conditions.

Study on reduced-moderation water reactor (RMWR) core design; Joint research report, FY1998-1999 (Joint research)

Research Group for Advanced Reactor System; Research Group for Reactor Physics; Research Group for Thermal and Fluid Engineering

JAERI-Research 2000-035, 316 Pages, 2000/09

JAERI-Research-2000-035.pdf:19.81MB

no abstracts in English

Pu Vector Sensitivity Study for a Pu Burning Fast Reactor Part II:Rod Worth Assessment and Design Optimization

Hunter

PNC-TN9410 97-057, 106 Pages, 1997/05

PNC-TN9410-97-057.pdf:2.99MB

This study was based on a 'pancake' type fast reactor core design of 600 MW(e), which had been optimized for Pu burning with a feed Pu vector appropriate to once-through irradiation of MOX fuel in a PWR. The purpose of the study was to investigate the effects of varying the Pu vector, examining various methods of offsetting the effects of such a change, and finally to produce fuel cycles optimized for the different qualities of Pu vector within the same basic design. In addition to the reference (once-through) Pu vector, two extreme Pu vectors were examined: high quality Pu from military stockpiles; low quality Pu corresponding to the equilibrium point of multiple recycling in a Pu burning fast reactor. Variations in Pu quality were overcome by changing the fuel inventory - replacing some of the fuel by diluent material, and altering the fuel pin size. Using absorber material (BC) as diluent improves the rod worth shutdown margin but degrades the Na void and Doppler safety parameters, a non-absorber diluent has the opposite effects, so a mix of the 2 material types was used to optimize the core characteristics. Of the non-absorber diluent materials examined, ZrH gave significantly better performance than all others; BC was the second choice for non-absorber diluent, because of its compatibility with BC absorber. It was not possible to accommodate the lower quality (multi-recycled) Pu vector without a significant increase in the fuel pin volume. It was not generally possible, especially with the increased fuel pin size, to achieve positive rod worth shutdown margins - this was overcome by increasing the number of control rods. For the higher quality Pu vectors to maintain ratings within limits, it was necessary to adopt hollow fuel pellets, or else to use the diluent material as an inert matrix in the fuel pellets. It proved possible to accommodate both extremes of Pu vector within a single basic design, maintaining ...

None

PNC-TN8020 91-003, 49 Pages, 1990/12

PNC-TN8020-91-003.pdf:1.18MB

no abstracts in English

Evaluation of Symbiotic Energy System Between Gas-Cooled Fast Breeder Reactor and Multipurpose Very High Temperature Reactor(III); Reference GCFR Core Configuration and It's Thermo-Hydraulic Characteristics

; ; *; *; *

JAERI-M 82-183, 68 Pages, 1982/12

JAERI-M-82-183.pdf:1.92MB

no abstracts in English

Consideration of directional effects of inner-ducts in fuel sub-assemblies to power distribution in JSFR core design

Moriwaki, Hiroyuki*; Kan, Taro*; Oki, Shigeo

no journal, , 

no abstracts in English

Core design for the next generation sodium-cooled fast reactor, 1; Core performance requirements and design conditions

Oki, Shigeo; Chikazawa, Yoshitaka; Kubo, Shigenobu; Hibi, Koki*; Kan, Taro*

no journal, , 

no abstracts in English

Current status and future prospects of Monte Carlo calculation methods in reactor physics

Nagaya, Yasunobu

no journal, , 

We review Monte Carlo capabilities developed recently and Monte Carlo applications to reactor analysis and reactor design. Finally, we describe future prospects of the Monte Carlo methodological development and the applicability of the Monte Carlo methods to reactor design.

HTGR licensing process, HTGR technology and design, materials for HTRGs, hydrogen production technology for HTGR cogeneration

Nishihara, Tetsuo; Tachibana, Yukio; Takada, Shoji; Ohashi, Hirofumi; Goto, Minoru; Noguchi, Hiroki

no journal, , 

Lectures on HTGR safety requirements and experience with HTTR licensing are conducted as HTGR licensing process. Lectures on core design, components design and safety analysis are conducted as HTGR technology and design. Lectures on graphite metallic material and fuel are conducted as materials for HTGRs. Additionally, lecture on hydrogen production technology for HTGR cogeneration is conducted.