F40 核材料、核燃料综合 标准查询与下载

ASTM C1430-00 用大气均衡法测定烧结的二氧化铀和氧化钆-二氧化铀颗粒中铀、氧铀比(O/U)和氧金属比(O/M)的标准试验方法

1.1 This test method applies to the determination of uranium, the oxygen to uranium (O/U) ratio in sintered uranium dioxide pellets, and the oxygen to metal (O/M) ratio in sintered gadolinium oxide-uranium dioxide pellets with a Gd2O3 concentration of up to 12 weight %. The O/M calculations assume that the gadolinium and uranium oxides are present in a metal dioxide solid solution.This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use. For specific hazards statements, see Section .

Standard Test Method for Determination of Uranium, Oxygen to Uranium (O/U), and Oxygen to Metal (O/M) in Sintered Uranium Dioxide and Gadolinia-Uranium Dioxide Pellets by Atmospheric Equilibration

BS ISO 13463:1999 制造轻水反应堆MOX燃料用核级二氧化钚粉末.帮助定义产品规范的指南

This International Standard applies to nuclear-grade plutonium dioxide powder (PuO2), for fabrication of light water reactor MOX fuel. It defines minimal requirements for the product, quality assurance and quality control. It can be modified or amended once the powder supplier and the MOX fuel fabricator agree. This International Standard sets, in that respect, guidelines for the definition of a product specification.

Nuclear-grade plutonium dioxide powder for fabrication of light water reactor MOX fuel - Guidelines to help in the definition of a product specification

ISO 13463:1999 制造轻水反应堆MOX燃料用的核级二氧化钚粉末 帮助定义产品规范的指南

This International Standard applies to nuclear-grade plutonium dioxide powder (PuO2), for fabrication of light water reactor MOX fuel. It defines minimal requirements for the product, quality assurance and quality control. It can be modified or amended once the powder supplier and the MOX fuel fabricator agree. This International Standard sets, in that respect, guidelines for the definition of a product specification.

Nuclear-grade plutonium dioxide powder for fabrication of light water reactor MOX fuel - Guidelines to help in the definition of a product specification

ANSI/ANS 15.2-1999 片状铝铀核燃料元件的质量控制

Sets forth general requirements for the establishment and execution of a program designed to verify that the quality of plate-type uranium-aluminum fuel elements being purchased for research reactors conforms to the requirements of the contract and applicable technical documents, including specifications, standards, and drawings.

Quality Control for Plate-Type Uranium-Aluminum Fuel Elements

EJ/T 1099-1999 贫化四氟化铀技术条件

Specification for depleted uranium tetrafluoride

ASTM C1431-99(2005) 维护容器清理用铝基废核燃料腐蚀检验的标准导则

1.1 This guide covers corrosion testing of aluminum-based spent nuclear fuel in support of geologic repository disposal (per the requirements in 10 CFR 60 and 40 CFR 191). The testing described in this document is designed to provide data for analysis of the chemical stability and radionuclide release behavior of aluminum-based waste forms produced from aluminum-based spent nuclear fuels. The data and analyses from the corrosion testing will support the technical basis for inclusion of aluminum-based spent nuclear fuels in the repository source term. Interim storage and transportation of the spent fuel will precede geologic disposal; therefore, reference is also made to the requirements for interim storage (per 10 CFR 72) and transportation (per 10 CFR 71). The analyses that will be based on the data developed are also necessary to support the safety analyses reports (SARs) and performance assessments (PAs) for disposal systems. 1.2 Spent nuclear fuel that is not reprocessed must be safely managed prior to transportation to, and disposal in, a geologic repository. Placement is an interim storage facility may include direct placement of the irradiated fuel or treatment of the fuel prior to placement, or both. The aluminum-based waste forms may be required to be ready for geologic disposal, or road ready, prior to placement in extended interim storage. Interim storage facilities, in the United States, handle fuel from civilian commercial power reactors, defense nuclear materials production reactors, and research reactors. The research reactors include both foreign and domestic reactors. The aluminum-based fuels in the spent fuel inventory in the United States are primarily from defense reactors and from foreign and domestic research reactors. The aluminum-based spent fuel inventory includes several different fuel forms and levels of 235U enrichment. Highly enriched fuels (235U enrichment leves 062 20%) are part of this inventory. 1.3 Knowledge of the corrosion behavior of aluminum-based spent nuclear fuels is required to ensure safety and to support licensing or other approval activities, or both, necessary for disposal in a geologic repository. The response fo the aluminum-based spent nuclear fuel waste form(s) to disposal environments must be established for configuration-safety analyses, criticality analyses, PAs, and other analyses required to assess storage, treatment, transportation, and disposal of spent nuclear fuels. This is particularly important for the highly enriched, aluminum-based spent nuclear fuels. The test protocols described in this guide are designed to establish material response under the repository relevant conditions. 1.4 The majority of the aluminum-based spent nuclear fuels are aluminum clad, aluminum-uranium alloys. The aluminum-uranium alloy typically consists of uranium aluminide particles dispersed in an aluminum matrix. Other aluminum-based fuels include dispersions of uranium oxide, uranium silicide, or uranium carbide particles in an aluminum matrix. These particles, including the aluminides, are generally cathodic to the aluminum matrix. Selective leaching of the aluminum in the exposure environment may provide a mechanism for redistribution and relocation of the uranium-rich particles. Particle redistribution tendencies will depend on the nature of the aluminum corrosion processes and the size, shape, distribution and relative reactivity of the uranium-rich particles. Interpretation of test data will require an understanding of the material behavior. This understanding will enable evaluation of the design and configuration of the waste package to ensure that unfilled regions in the waste package do not provide sites for the relocation of the uranium-rich particles into nuclear critical configurations. Test samples must be evaluated, prior to testing, to ensure that the size and shape of the uranium-rich particles in the test samples are......

Standard Guide for Corrosion Testing of Aluminum-Based Spent Nuclear Fuel in Support of Repository Disposal

ASTM C1431-99 维护容器清理用铝基废核燃料腐蚀检验的标准导则

1.1 This guide covers corrosion testing of aluminum-based spent nuclear fuel in support of geologic repository disposal (per the requirements in 10 CFR 60 and 40 CFR 191). The testing described in this document is designed to provide data for analysis of the chemical stability and radionuclide release behavior of aluminum-based waste forms produced from aluminum-based spent nuclear fuels. The data and analyses from the corrosion testing will support the technical basis for inclusion of aluminum-based spent nuclear fuels in the repository source term. Interim storage and transportation of the spent fuel will precede geologic disposal; therefore, reference is also made to the requirements for interim storage (per 10 CFR 72) and transportation (per 10 CFR 71). The analyses that will be based on the data developed are also necessary to support the safety analyses reports (SARs) and performance assessments (PAs) for disposal systems. 1.2 Spent nuclear fuel that is not reprocessed must be safely managed prior to transportation to, and disposal in, a geologic repository. Placement is an interim storage facility may include direct placement of the irradiated fuel or treatment of the fuel prior to placement, or both. The aluminum-based waste forms may be required to be ready for geologic disposal, or road ready, prior to placement in extended interim storage. Interim storage facilities, in the United States, handle fuel from civilian commercial power reactors, defense nuclear materials production reactors, and research reactors. The research reactors include both foreign and domestic reactors. The aluminum-based fuels in the spent fuel inventory in the United States are primarily from defense reactors and from foreign and domestic research reactors. The aluminum-based spent fuel inventory includes several different fuel forms and levels of 235U enrichment. Highly enriched fuels (235U enrichment leves 062 20%) are part of this inventory. 1.3 Knowledge of the corrosion behavior of aluminum-based spent nuclear fuels is required to ensure safety and to support licensing or other approval activities, or both, necessary for disposal in a geologic repository. The response fo the aluminum-based spent nuclear fuel waste form(s) to disposal environments must be established for configuration-safety analyses, criticality analyses, PAs, and other analyses required to assess storage, treatment, transportation, and disposal of spent nuclear fuels. This is particularly important for the highly enriched, aluminum-based spent nuclear fuels. The test protocols described in this guide are designed to establish material response under the repository relevant conditions. 1.4 The majority of the aluminum-based spent nuclear fuels are aluminum clad, aluminum-uranium alloys. The aluminum-uranium alloy typically consists of uranium aluminide particles dispersed in an aluminum matrix. Other aluminum-based fuels include dispersions of uranium oxide, uranium silicide, or uranium carbide particles in an aluminum matrix. These particles, including the aluminides, are generally cathodic to the aluminum matrix. Selective leaching of the aluminum in the exposure environment may provide a mechanism for redistribution and relocation of the uranium-rich particles. Particle redistribution tendencies will depend on the nature of the aluminum corrosion processes and the size, shape, distribution and relative reactivity of the uranium-rich particles. Interpretation of test data will require an understanding of the material behavior. This understanding will enable evaluation of the design and configuration of the waste package to ensure that unfilled regions in the waste package do not provide sites for the relocation of the uranium-rich particles into nuclear critical configurations. Test samples must be evaluated, prior to testing, to ensure that the size and shape of the uranium-rich particles in the test samples are......

Standard Guide for Corrosion Testing of Aluminum-Based Spent Nuclear Fuel in Support of Repository Disposal

EJ/T 1095-1999 核材料许可证申请中衡算与控制文件的编写要求

Compiling requirements for the nuclear material control and accounting documents in nuclear material license application

EJ/T 1089-1998 铀离心浓缩厂核材料衡算与控制视察程序

Inspection procedure of nuclear material control and accounting for uranium centrifuge enrichment plants

EJ/T 1087-1998 压水堆核电厂用涂料漆膜耐化学介质的测定

Determination of chemical resistance of coatings used in pressurized water reactor nuclear power plants

EJ/T 1086-1998 压水堆核电厂用涂料.漆膜在模拟设计基准事故条件下的评价试验方法

Test method for evaluating coatings used in pressurized water reactor nuclear power plants at simulated design basis accident (DBA) conditions

DIN EN 1911-3:1998 固定源放射.HCl的手动法测定.第3部分:吸收溶液分析和结果计算

Stationary source emissions - Manual method of determination of HCl - Part 3: Absorption solutions analysis and calculation; German version EN 1911-3:1998

DIN EN 1911-1:1998 固定源放射.HCl的手动法测定.第1部分:气体的取样

The document specify a manual measuring method for the determination of volatile inorganic chlorides, expressed as hydrogen chloride (HCl), in emissions of stationary sources. It describes the design of the sampling system, measurement planning and the procedure, how a representative sample of the waste gas stream is sucked in under defined conditions by means of a sampling system for isokinetic or non-isokinetic sampling.

Stationary source emissions - Manual method of determination of HCl - Part 1: Sampling of gases; German version EN 1911-1:1998

DIN EN 1911-2:1998 固定源放射.HCl的手动法测定.第2部分:气态化合物的吸收

The document specifies a method for the absorption of a waste gas sample which hase been taken according to DIN EN 1911-1 in an aqueous solution. The method applies to waste gases, in which the chloride concentration is in the range between 1 mg/m and 5000 mg/m under normal pressure and temperature conditions.#,,#

Stationary source emissions - Manual method of determination of HCl - Part 2: Gaseous compounds absorption; German version EN 1911-2:1998

ANSI/ANS 8.7-1998 裂变材料储存的核临界安全指南

This standard is applicable to the storage of fissile materials. Mass and spacing limits are tabulated for uranium containing greater than 30 wt-% 235U, for 233U, and for plutonium, as metals and oxides. Criteria for the range of application of these limits are provided.

Nuclear Criticality Safety in the Storage of Fissile Materials, Guide for

EJ/T 1054-1997 核材料实物保护导则

Guide for physical protection of nuclear material

EJ 1055-1997 核电厂核材料衡算与控制视察程序

Nuclear material control and accounting inspection procedure of nuclear power plants

EJ 1053-1997 核燃料元件制造厂核材料衡算与控制视察程序

Material control and accounting inspection procedure for fuel fabrication plants

DIN 25421:1997 裂变物质溶液硼硅玻璃拉希格圈中子吸收剂的使用

The document specifies requirements relating to the use of borosilicate glass raschig rings as a neutron in solutions of fissile material being the only media to guarantee criticality safety.

Use of borosilicate glass raschig rings as a neutron absorber in solutions of fissile material

ARMY MIL-C-48665 NOTICE 1-1997 环形核

This notice should be filed in front of MIL-C-48665(AR) dated 28 April 1986

CORE; TOROID