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JAEA Reports

Prototype fast breeder reactor Monju; Its history and achievements

Tsuruga Comprehensive Research and Development Center

JAEA-Technology 2019-007, 159 Pages, 2019/07

JAEA-Technology-2019-007.pdf:19.09MB
JAEA-Technology-2019-007-high-resolution1.pdf:42.36MB
JAEA-Technology-2019-007-high-resolution2.pdf:33.56MB
JAEA-Technology-2019-007-high-resolution3.pdf:38.14MB
JAEA-Technology-2019-007-high-resolution4.pdf:48.82MB
JAEA-Technology-2019-007-high-resolution5.pdf:37.61MB

This report summarizes the history and achievements of the prototype fast breeder reactor Monju. The development of Monju started in 1968 as a prototype reactor following the experimental fast reactor Joyo. The development covers all the activity related to the fast reactor; plant design, mockup tests, construction, operation, and plant management. This report summarizes the history and achievements for 11 technical areas: history and principal achievements, design and construction, operation test, plant safety, core physics, fuel, plant system, sodium technology, materials and mechanical design, plant management, and trouble management.

JAEA Reports

Investigation of utilizing plutonium as mixed oxide fuel (5); BWR for next generation

*; *; *; *

JNC-TJ9440 2000-007, 43 Pages, 2000/03

JNC-TJ9440-2000-007.pdf:1.73MB

Planning of the plutonium utihzation in the Light water thermal reactor has been investigated to evaluate scenario for FBR development. Plans for MOX fuel utilization in the ABWR including Ooma plant are studied, and information of high burnup fuels for a future BWR is summarized based on public documents. Nuclear compositions of the present burnup fuel (45,000MWd/t) and a high burnup fue (60,000MWd/t) have been evaluated using an open code: SRAC. Results of the study are follows; (1)Surveying the status of MOX fuel utilization. The status of MOX and UO$$_{2}$$ fuel utilization in the present BWR and future BWR have been summarized based on public documents. (2)Evaluation of spent MOX and UO$$_{2}$$ fuel composition. Nuclear compositions of spent MOX and UO$$_{2}$$ fuels at 45,000MWd/t and 60,000MWd/t burnup have been evaluated and summarized for recycle scenarios by FBR.

JAEA Reports

Measurement of neutron capture cross sections of Tc-99

*

JNC-TJ9400 2000-008, 61 Pages, 2000/02

JNC-TJ9400-2000-008.pdf:2.5MB

For studies on nuclear transmutation of long-lived fission products (LLFPs) in a fast reactor, detailed characteristics of reactor core such as transmutation performance have to be investigated, so accurate neutron cross section data of LLFPs become necessary. Therefore, the keV-neutron capture cross sections of Tc-99, which is one of important LLFPs, were measured in the present study to obtain the accurate data. The measurement was relative to the standard capture cross sections of Au-197. A neutron time-of-flight method was adopted with a ns-pulsed neutron source by a Pelletron accelerator and a large anti-Compton NaI(TI) gamma-ray detector. As a result, the capture cross sections of Tc-99 were obtained with the error of about 5 % in the incident neutlon energy region of 10 to 600 keV. The present data were compared with other experimental data and the evaluated values of JENDL-3.2, and it was found that the evaluations of JENDL-3.2 were 15-20 % smaller than the present measurements.

JAEA Reports

Measurement of neutron capture cross sections of Tc-99

*

JNC-TJ9400 99-001, 78 Pages, 1999/03

JNC-TJ9400-99-001.pdf:2.07MB

For studies on incineration of long-lived fission products (LLFPs) in a fast reactor, detailed characteristics of reactor core such as incineration performance have to be investigated. Therefore, accurate neutron cross section data of LLFPs become necessary. In the present study, in order to perform the precise measurements of keV-neutron capture cross sections of Tc-99, which is one of most important LLFPs, the details of the Tc-99 sample and the measurements with our experimental facilities were investigated.

JAEA Reports

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 ($$^{10}$$B$$_{4}$$C) 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; $$^{11}$$B$$_{4}$$C was the second choice for non-absorber diluent, because of its compatibility with $$^{10}$$B$$_{4}$$C 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 ...

JAEA Reports

None

PNC-TJ1678 95-003, 97 Pages, 1995/02

PNC-TJ1678-95-003.pdf:2.59MB

None

JAEA Reports

None

*; *; *; *; *

PNC-TJ1678 95-002, 121 Pages, 1995/02

PNC-TJ1678-95-002.pdf:4.83MB

None

JAEA Reports

None

PNC-TJ1678 95-006, 181 Pages, 1994/11

PNC-TJ1678-95-006.pdf:5.25MB

None

JAEA Reports

Outline of reactor physics tests conducted during the power-up tests in the nuclear ship Mutsu

; Miyoshi, Yoshinori; *; *

JAERI-M 92-172, 62 Pages, 1992/11

JAERI-M-92-172.pdf:2.01MB

no abstracts in English

Oral presentation

Core design for the next generation sodium-cooled fast reactor, 2; Reference core design

Tsuboi, Toru*; Moriwaki, Hiroyuki*; Ogura, Masashi*; Hibi, Koki*; Maeda, Seiichiro; Ohgama, Kazuya; Chikazawa, Yoshitaka; Oki, Shigeo

no journal, , 

no abstracts in English

Oral presentation

R&D of advanced stainless steels for BWR fuel claddings, 3-2; Applicability of MOX fuel core design

Takano, Sho*; Goto, Daisuke*; Kusagaya, Kazuyuki*; Sakamoto, Kan*; Yamashita, Shinichiro

no journal, , 

For practical use of FeCrAl-ODS stainless steel, it is necessary to confirm that a design request in a reactor core is achieved even if the loss of neutron economy is caused by the big neutron absorption cross section. In previous study, the core design was established in the case of cladding thickness of 0.3 mm, water rod thickness of 0.3 mm and channel box thickness of 1.0 mm. In this report, the core property was evaluated when FeCrAl-ODS stainless steel and MOX fuel were applied.

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