JOPSS - Search Results

JAEA Reports

Analytical study of two-region TCA critical experiments with PWR-type MOX fuel by using Monte Carlo code MVP

Rahman, M.*; Suzaki, Takenori; Mori, Takamasa

JAERI-Research 2003-007, 68 Pages, 2003/03

JAERI-Research-2003-007.pdf:2.48MB

no abstracts in English

JAEA Reports

Neutronic study of a very small reactor MR-1G core for exclusive use of heat supply

Odano, Naoteru; Ishida, Toshihisa; Ochiai, Masaaki

JAERI-Research 2001-044, 53 Pages, 2001/10

JAERI-Research-2001-044.pdf:2.35MB

A very small reactor, MR-1G with a thermal output of 1 MW, is the integral-pressurized type reactor to be used for heat supply to an office building in a city. Neutronic study has been carried out for design of the MR-1G, the core of which could achieve continuous long-term operation without refueling for 10 years assuming a load factor of the core of 44 %. In the present study, arrangement of fuel rods and $^{235}$ U enrichment of UO $_{2}$ fuel rods were surveyed as design parameters. The $^{235}$ U enrichment was determined to be 8.5 wt% to satisfy design requirement. 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 analyzed. Results of the safety analysis indicated that the determined core specifications satisfied design conditions. Reactor shutdown performance by dropping the reflector, which was adopted as a passive reactor shutdown system, was analyzed and confirmed it's availability from a viewpoint of reactor physics.

JAEA Reports

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, $^{235}$ U enrichment of UO $_{2}$ fuel rods and reflector materials were surveyed. The $^{235}$ 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.

JAEA Reports

Neutronic study of DRX (Deep Sea Reactor X) core for deep sea research vessel

Odano, Naoteru; Ishida, Toshihisa

JAERI-Research 2001-004, 36 Pages, 2001/03

JAERI-Research-2001-004.pdf:1.51MB

no abstracts in English

JAEA Reports

Study on thermal-hydraulics during a PWR reflood phase

Iguchi, Tadashi

JAERI-Research 98-054, 216 Pages, 1998/10

JAERI-Research-98-054.pdf:7.94MB

no abstracts in English

JAEA Reports

None

Yamaguchi, Takashi

PNC TN1410 97-026, 16 Pages, 1997/07

PNC-TN1410-97-026.pdf:0.29MB

no abstracts in English

JAEA Reports

Educational reactor-physics experiments with the critical assembly TCA

*; *; *; Suzaki, Takenori; Horiki, Oichiro*

JAERI-Review 96-010, 40 Pages, 1996/08

JAERI-Review-96-010.pdf:1.14MB

no abstracts in English

JAEA Reports

None

; Seki, Masayuki; ; ; ; Tobita, Noriyuki

PNC TN8410 95-023, 130 Pages, 1995/01

PNC-TN8410-95-023.pdf:15.78MB

None

Journal Articles

Geometric buckling expression for regular polygons, I; Measurements in low-enriched UO $_{2}$ -H $_{2}$ O lattices

Miyoshi, Yoshinori; ; ; Hirose, Hideyuki;

Nuclear Technology, 103, p.380 - 391, 1993/09

https://doi.org/10.13182/NT93-A34858

Times Cited Count：3 Percentile：38.1(Nuclear Science & Technology)

no abstracts in English

JAEA Reports

RODBURN: A Code for calculating power distribution in fuel rods

Uchida, Masaaki; Saito, Hioraki*

JAERI-M 93-108, 36 Pages, 1993/05

JAERI-M-93-108.pdf:0.96MB

no abstracts in English

JAEA Reports

Computer code (DIRAD) for fuel temperature evaluation of LMFBR at early burnup

; *; ; ; ; Matsumoto, Mitsuo

PNC TN8410 92-187, 21 Pages, 1992/05

PNC-TN8410-92-187.pdf:0.22MB

Computer code for fuel temperature evaluation of LMFBR at early burnup has been developed. On fule temperature evaluation, especially at early burnup, the fuel restructuring model and the gap conductance medel are important. These models which are installed in the temperature evaluation code, were verified based on the results of irradiation tests using the foreign fast reactor and "JOYO". This paper describes the essential parts of the models and the functions of the code (DIRAD) which is used for the fuel temperature evaluation at early burnup of LMFBR such as the prototype fast breeder reactor "MONJU".

JAEA Reports