ASME STP-NU-009-2008
高温气体冷却核反应堆用石墨

Graphite for High Temperature Gas-Cooled Nuclear Reactors

ASME STP-NU-009-2008 发布历史

This document 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.

ASME STP-NU-009-2008由美国机械工程师协会 US-ASME 发布于 2008。

ASME STP-NU-009-2008 在中国标准分类中归属于: Q51 石墨材料，在国际标准分类中归属于: 27.120.10 反应堆工程。

ASME STP-NU-009-2008的历代版本如下：

• 2008年 ASME STP-NU-009-2008 高温气体冷却核反应堆用石墨

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标准号

ASME STP-NU-009-2008

发布日期

2008年

实施日期

废止日期

无

中国标准分类号

Q51

国际标准分类号

27.120.10

发布单位

US-ASME

适用范围

This document 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.