F48 核燃料元件及其分析试验方法 标准查询与下载

EJ/T 541-2005 铀三硅二-铝板型燃料组件

本标准规定了低浓铀(LEU)铀三硅二-铝(USiAl)板型燃料组件的技术要求、试验方法、验收规则与标志、包装、运输、贮存等要求。 本标准适用于低浓铀(LEU)USiAl板型燃料组件的制造和检验。 本标准不包括防止临界事故和对卫生、安全和环境保护方面的要求。

Specifications for USi-Al plate-type fuel assembly

EJ/T 1167-2004 压水堆燃料组件导向管与格架电阻点焊的金相检验

本标准规定了压水堆燃料组件导向管与格架电阻点焊的金相检验方法。 本标准适用于压水堆燃料組件不銹鋼及鑽合金導向管與格架电阻點焊的金相檢驗。

PWR fuel assembly metallographic examination for resistance spot weld of guide thimble to grid

EJ/T 802-2004 压水堆燃料组件管座焊缝检验X射线照相法

Test method for nozzle weld of PWR fuel assembly X-ray photo method

EJ/T 1155-2002 洗衣房用固定式放射性污染监测仪

本标准规定了洗衣房用放射性污染监测仪的最低性能要求和试验方法。 本标准适用于工作服放射性污染监测仪。该监测仪用于工作服清洗后，决定复用前其放射性污 染水平在可接受水平。 本标准不适用于厂区外使用的工作服监测。

Installed monitors for the detection of radioactive contamination of laundry

ASTM C781-02 高温气冷核反应堆用石墨及硼酸化石墨元部件试验的标准实施规程

Property data obtained with the recommended test methods identified herein may be used for research and development, design, manufacturing control, specifications, performance evaluation, and regulatory statutes pertaining to high temperature gas-cooled reactors. The test methods are applicable primarily to specimens in the unirradiated and unoxidized state. Many are also applicable to specimens in the irradiated or oxidized state, or both, provided the specimens meet all requirements of the test method. The user is cautioned to consider the instructions given in the test methods. Additional test methods are in preparation and will be incorporated. The user is cautioned to employ the latest revision. 1.1 This practice covers the test methods for measuring those properties of graphite and boronated graphite materials that may be used for the design and evaluation of high-temperature gas-cooled reactors. 1.2 The test methods referenced herein are applicable to materials used for replaceable and permanent components as defined in Section 7 and include fuel and removable reflector elements; target elements and insulators; permanent side reflector elements; core support pedestals and elements; control rod, reserve shutdown, and burnable poison compacts, and neutron shield material. 1.3 This practice includes test methods that have been selected from existing ASTM standards, ASTM standards that have been modified, and new ASTM standards that are specific to the testing of materials listed in 1.2. Comments on individual test methods for graphite and for boronated graphite components are given in Sections 8 and 10, respectively. The test methods are summarized in Tables 1 and 2. 1.4 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.

Standard Practice for Testing Graphite and Boronated Graphite Components for High-Temperature Gas-Cooled Nuclear Reactors

EJ/T 653-1998 高通量工程试验反应堆燃料组件技术条件

Specification for fuel assembly of high flux engineering test reactor

ASTM C1207-97 无源中子复合计数法测定废弃物中钚的无损检验试验方法

1.1 This test method describes the nondestructive assay of scrap or waste for plutonium content using passive thermal-neutron coincidence counting. This test method provides rapid results and can be applied to a variety of carefully sorted materials in containers as large as 208-L drums. It has been used to assay items whose plutonium content ranges from 0.01 to 6000 g. 1.2 This test method requires knowledge of the relative abundances of the plutonium isotopes to determine the total plutonium mass. 1.3 This test method may not be applicable to the assay of scrap or waste containing other spontaneously fissioning nuclides. 1.4 This test method assumes the use of shift-register-based coincidence electronics (1). 1.5 This standard does not purport to address all of the safety problems, 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.

Standard Test Method for Nondestructive Assay of Plutonium in Scrap and Waste by Passive Neutron Coincidence Counting

ASTM C1205-97 α光谱法对土壤中镅-241进行放射化学测定的试验方法

1.1 This test method covers the determination of americium-241 in soil by means of chemical separations and alpha spectrometry. It is designed to analyze up to ten grams of soil or other sample matrices that contain up to 30 mg of combined rare earths. 1.2 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 a specific precaution statement, see note in Section 8.

Standard Test Method for The Radiochemical Determination of Americium-241 in Soil by Alpha Spectrometry

ASTM C1205-97(2002) α光谱法对土壤中镅-241进行放射化学测定的试验方法

1.1 This method covers the determination of americium-241 in soil by means of chemical separations and alpha spectrometry. It is designed to analyze up to ten grams of soil or other sample matrices that contain up to 30 mg of combined rare earths. This method allows the determination of americium-241 concentrations from ambient levels to applicable standards. The values stated in SI units are to be regarded as standard.1.2 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 precaution statements, see Section 9.

Standard Test Method for The Radiochemical Determination of Americium-241 in Soil by Alpha Spectrometry

DIN 25711:1996 α光谱分析测定含铀核燃料溶液中铀235同位素含量

Determination of the <(hoch)232>U isotopic content in uranium containing nuclear fuel solutions by spectrometry

ASTM C1133-96 分段无源γ射线扫描法对低密度废品及残渣中特殊核材料进行无损检验的标准试验方法

1.1 This test method covers the nondestructive assay of gamma-ray emitting special nuclear materials (SNMs) in low-density scrap or waste, packaged in cylindrical containers (with measured transmission values greater than 1%). High resolution gamma-ray spectroscopy is used to detect and measure the nuclides of interest and to measure and correct for average gamma-ray attenuation in discrete vertical segments of the container. Corrections are also made for counting losses occasioned by signal processing characteristics (1-3). 1.2 Two conditions must be met to optimize assay results as follows: 1.2.1 The particles containing the nuclides of interest must be small to minimize self absorption of emitted gamma radiation. 1.2.2 The mixture of material within a package segment must be reasonably uniform in order to apply an attenuation correction factor, computed from a single measurement of item transmission through the segment. 1.3 The assay technique may be applicable to loadings of from one to several hundred grams of nuclide with more restricted ranges to be applicable, depending on specific packaging and counting equipment considerations. Measured transmission values must be greater than 1% to permit valid attenuation corrections. 1.4 The values stated in SI units are to be regarded as the standard. The values given in parentheses are for information only. 1.5 This standard may involve hazardous materials, operations, and equipment. 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. Specific precautionary statements are given in Section 9.

Standard Test Method for Nondestructive Assay of Special Nuclear Material in Low Density Scrap and Waste by Segmented Passive Gamma-Ray Scanning

ASTM C781-96 高温气冷核反应堆石墨及硼酸化石墨元部件试验的标准实施规程

1.1 This practice covers the test methods for measuring those properties of graphite and boronated graphite materials that may be used for the design and evaluation of high-temperature gas-cooled reactors. 1.2 The test methods referenced herein are applicable to materials used for replaceable and permanent components as defined in Section 7 and include fuel and removable reflector elements; target elements and insulators; permanent side reflector elements; core support pedestals and elements; control rod, reserve shutdown, and burnable poison compacts, and neutron shield material. 1.3 This practice includes test methods that have been selected from existing ASTM standards, ASTM standards that have been modified, and new ASTM standards that are specific to the testing of materials listed in 1.2. Comments on individual test methods for graphite and for boronated graphite components are given in Sections 8 and 10, respectively. The test methods are summarized in Tables 1 and 2. 1.4 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.

Standard Practice for Testing Graphite and Boronated Graphite Components for High-Temperature Gas-Cooled Nuclear Reactors

DIN 25488-2:1995 热室用元件.第2部分:抗辐射、耐腐蚀和去污性的要求

Components for hot cells - Part 2: Requirements on radiation resistance, corrosion resistance and decontaminability

DIN 25488-1:1995 热室用元件.第1部分:遥控设计的要求

The document is valid for components, i.e. machines, apparatus, devices, remote handling devices and tools, which are used or remote handled in hot cells with lead or concrete shieldings. The document can also be applied to components in other areas of nuclear installations with the necessity for remote handling, such as reactors, fusion reactor power plants and high energy particle accelerators.#,,#

Components for hot cells - Part 1: Requirements on design for remote handling

ASTM E402-95 用氧化镓载体技术作氧化铀(U308)光谱分析的标准试验方法

1.1 This test method covers the semiquantitative spectrographic analysis of high-purity U O for the 32 elements in the ranges indicated in Table 1. (Quantitative analyses of boron, chromium, iron, magnesium, manganese, nickel, and other impurities can be performed using densitometric methods.) 1.2 The test method can be applied to those samples of uranium and uranium compounds, or both, which can be converted to the black oxide (U O ) and which are of approximately 99.5% purity or better. 1.3 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.

Standard Test Method for Spectrographic Analysis of Uranium Oxide (U₃O₈)by Gallium Oxide-Carrier Technique

ISO 9891:1994 二氧化铀粉末和烧结芯块中碳含量的测定 高频感应炉燃烧 滴定法/库仑法/红外吸收法

The methods specified are based on heating a portion of the test sample at a temperature of at least 1100 鳦 to 1200 鳦 in an oxygen atmosphere, passing the evolved oxidation products over a purification trap filled with manganese dioxide catalyst and silver permanganate catalyst (that oxidises CO to CO), trapping the CO2, restoring the initial pH continuously by addition of hydroxyl ions either by potentiostatic titrimetry or by coulometry, or alternatively determining the CO by IR absorption and integration of the signal obtained. Is applicable to the determination of 5 鎔 to 500 鎔 of carbon.

Determination of carbon content in uranium dioxide powder and sintered pellets - High-frequency induction furnace combustion - Titrimetric/coulometric/infrared absorption methods

EJ/T 910-1994 U₃Si₂-Al燃料板芯体定位方法

U₃Si₂-Al fuel plate core positioning method

EJ/T 868-1994 核燃料元件制造中的质量保证

本标准规定了核燃料元件制造中的质量保证要求、内容和程序等项目。 本标准适用于各种堆型燃料元件制造中的质量保证。 使用本标准时允许根据核燃料元件的特点及复条程序进行裁剪。

Quality assurance in the manufacture of nuclear fuel elements

ISO 10979:1994 核动力反应堆用燃料组件标识

Specifies requirements for the unique identification. Was developed primarily for commercial light-water reactor fuel, but may be used for any reactor fuel contained in discrete fuel assemblies that can be identified with an identification code. Defines the characters and proposed sequence to be used in assigning identification to the fuel assemblies. The identification is intended to be borne by the fuel assembly throughout its lifetime. Aims at providing an organizing principle for fuel assembly identification systems in order to guarantee unequivocal identification at any time and any place in the world.

Identification of fuel assemblies for nuclear power reactors

EJ/T 817-1994 压水堆核电厂新燃料组件包装、运输、装卸和贮存规定

Regulations on packaging, transportation, handling and storage of new fuel assemblies in pressurized water reactor nuclear power plants