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Graphite for High Temperature Gas-Cooled Nuclear Reactors (STP-NU-009 - 2008)
| 분야 | STP : ASME Standards Technology |
|---|---|
| 적용범위 | This technical report presents the basic information relative to bulk graphite production, structure, chemical properties, physical properties and neutron irradiation behavior. Bulk graphite characteristics, its manufacture, properties and irradiation behavior as well as a new generation of nuclear grades are briefly reviewed. An overview of graphite moderated gas-cooled reactor designs is also presented. The report serves as a summary of the training seminar on Nuclear Graphite conducted during the ASME Boiler and Pressure Vessel Code week, October 30?November 3, 2006, in Louisville, KY. There is no universally accepted code for the design of graphite moderator structures. The history of graphite moderated reactors is traced from the beginnings in 1942 to the most recent utility start-up in 1989. Developments have continued over the intervening years especially in the area of helium cooled High Temperature Reactors. Prismatic 30MWth, and pebble-bed 10MWth, test reactors were brought into operation in Japan and China, respectively. |
| 국제분류(ICS)코드 | |
| 페이지수 | 40 |
| Edition |
| No. | 표준번호 | 표준명 | 발행일 | 상태 |
|---|---|---|---|---|
| 1 | ASME STP-NU-009-2008 | Graphite for High Temperature Gas-Cooled Nuclear Reactors (STP-NU-009 - 2008) | 2008-01-01 | 표준 |
ASME STP-PT-005-2006 - Design Factor Guidelines for High-Pressure Composite Hydrogen Tanks (STP-PT-005 - 2006) 상세보기
ASME STP-PT-007-2006 - Comparison of Pressure Vessel Codes: ASME Section VIII & EN13445 (Technical, Commercial, & Usage Comparison Design Fatigue Life Comparison) (STP-PT-007 - 2006) 상세보기
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