27.120.01 - Nuclear energy in general
ICS 27.120.01 Details
Nuclear energy in general
Kerntechnik im allgemeinen
Energie nucléaire en général
Jedrska energija na splošno
General Information
Frequently Asked Questions
ICS 27.120.01 is a classification code in the International Classification for Standards (ICS) system. It covers "Nuclear energy in general". The ICS is a hierarchical classification system used to organize international, regional, and national standards, facilitating the search and identification of standards across different fields.
There are 110 standards classified under ICS 27.120.01 (Nuclear energy in general). These standards are published by international and regional standardization bodies including ISO, IEC, CEN, CENELEC, and ETSI.
The International Classification for Standards (ICS) is a hierarchical classification system maintained by ISO to organize standards and related documents. It uses a three-level structure with field (2 digits), group (3 digits), and sub-group (2 digits) codes. The ICS helps users find standards by subject area and enables statistical analysis of standards development activities.
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This document specifies methods for the measurement of the absorbed-dose rate in a tissue-equivalent slab phantom in the ISO 6980 reference beta-particle radiation fields. The energy range of the beta-particle-emitting isotopes covered by these reference radiations is 0,22 MeV to 3,6 MeV maximum beta energy corresponding to 0,07 MeV to 1,2 MeV mean beta energy. Radiation energies outside this range are beyond the scope of this document. While measurements in a reference geometry (depth of 0,07 mm or 3 mm at perpendicular incidence in a tissue‑equivalent slab phantom) with an extrapolation chamber used as primary standard are dealt with in detail, the use of other measurement systems and measurements in other geometries are also described, although in less detail. However, as noted in ICRU 56, the ambient dose equivalent, H*(10), used for area monitoring, and the personal dose equivalent, Hp(10), as used for individual monitoring, of strongly penetrating radiation, are not appropriate quantities for any beta radiation, even that which penetrates 10 mm of tissue (Emax > 2 MeV).
This document is intended for those organizations wishing to establish primary dosimetry capabilities for beta particles and serves as a guide to the performance of dosimetry with an extrapolation chamber used as primary standard for beta‑particle dosimetry in other fields. Guidance is also provided on the statement of measurement uncertainties.
- Standard50 pagesEnglish languagee-Library read for1 day
This document specifies the requirements for reference beta radiation fields produced by radioactive sources to be used for the calibration of personal and area dosemeters and dose-rate meters to be used for the determination of the quantities Hp(0,07), H'(0,07;Ω), Hp(3) and H'(3;Ω), and for the determination of their response as a function of beta particle energy and angle of incidence. The basic quantity in beta dosimetry is the absorbed-dose rate in a tissue-equivalent slab phantom. This document gives the characteristics of radionuclides that have been used to produce reference beta radiation fields, gives examples of suitable source constructions and describes methods for the measurement of the residual maximum beta particle energy and the dose equivalent rate at a depth of 0,07 mm in the International Commission on Radiation Units and Measurements (ICRU) sphere. The energy range involved lies between 0,22 MeV and 3,6 MeV maximum beta energy corresponding to 0,07 MeV to 1,2 MeV mean beta energy and the dose equivalent rates are in the range from about 10 µSv·h-1 to at least 10 Sv·h-1.. In addition, for some sources, variations of the dose equivalent rate as a function of the angle of incidence are given. However, as noted in ICRU 56[5], the ambient dose equivalent, H*(10), used for area monitoring, and the personal dose equivalent, Hp(10), as used for individual monitoring, of strongly penetrating radiation, are not appropriate quantities for any beta radiation, even that which penetrates 10 mm of tissue (Emax > 2 MeV).
This document is applicable to two series of reference beta radiation fields, from which the radiation necessary for determining the characteristics (calibration and energy and angular dependence of response) of an instrument can be selected.
Series 1 reference radiation fields are produced by radioactive sources used with beam-flattening filters designed to give uniform dose equivalent rates over a large area at a specified distance. The proposed sources of 106Ru/106Rh, 90Sr/90Y, 85Kr, 204Tl and 147Pm produce maximum dose equivalent rates of approximately 200 mSv·h–1.
Series 2 reference radiation fields are produced without the use of beam-flattening filters, which allows large area planar sources and a range of source-to-calibration plane distances to be used. Close to the sources, only relatively small areas of uniform dose rate are produced, but this series has the advantage of extending the energy and dose rate ranges beyond those of series 1. The series also include radiation fields using polymethylmethacrylate (PMMA) absorbers to reduce the maximum beta particle energy. The radionuclides used are those of series 1; these sources produce dose equivalent rates of up to 10 Sv·h–1.
- Standard29 pagesEnglish languagee-Library read for1 day
This document provides information and guidelines on the decommissioning of a medical cyclotron facility, with a focus on activated or contaminated parts. Useful information and guidelines are given on decommissioning strategy and plan, safety assessment, and various decommissioning activities. This document also provides the guideline on the estimation of activation level using Monte Carlo simulation and the methodology for the measurement of activated radionuclides in the main structure, system components, and shielding walls, ceilings and floors during operation and decommissioning. Financial provisions and radioactive waste management aspects are also included. This document can be used by organizations responsible for operation and decommissioning of a medical cyclotron facility. In addition, it is expected that organizations that design a medical cyclotron or manage radioactive waste generated by cyclotron can utilize or refer to this document in whole or in part.
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This document describes procedures for calibrating and determining the response of dosemeters and dose-rate meters in terms of the operational quantities for radiation protection purposes defined by the International Commission on Radiation Units and Measurements (ICRU). However, as noted in ICRU 56, the ambient dose equivalent, H*(10), used for area monitoring, and the personal dose equivalent, Hp(10), as used for individual monitoring, of strongly penetrating radiation, are not appropriate quantities for any beta radiation, even that which penetrates 10 mm of tissue (Emax > 2 MeV).
This document is a guide for those who calibrate protection-level dosemeters and dose-rate meters with beta-reference radiation and determine their response as a function of beta-particle energy and angle of incidence. Such measurements can represent part of a type test during the course of which the effect of other influence quantities on the response is examined. This document does not cover the in-situ calibration of fixed, installed area dosemeters. The term “dosemeter” is used as a generic term denoting any dose or dose-rate meter for individual or area monitoring. In addition to the description of calibration procedures, this document includes recommendations for appropriate phantoms and the way to determine appropriate conversion coefficients. Guidance is provided on the statement of measurement uncertainties and the preparation of calibration records and certificates.
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This document deals with the terminological data used in the standards regarding the standardization and promotion of good practices associated with the planning, design, construction, operation and decommissioning of installations, processes and technologies involving radioactive materials. The vocabulary of nuclear installations, processes and technologies includes fuel cycle, ex-reactor nuclear criticality safety, analytical methodologies, transport of radioactive materials, materials characterization, radioactive waste management and decommissioning. NOTE See Annex A for the methodology used to develop the vocabulary.
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This document complements the existing requirements of ISO/IEC 17021-1 for bodies providing audit and certification of quality management systems against ISO 19443.
NOTE This document can be used as a criteria document for accreditation, peer assessment or other audit processes.
- Technical specification26 pagesEnglish languagee-Library read for1 day
This document complements the existing requirements of ISO/IEC 17021-1 for bodies providing audit and certification of quality management systems against ISO 19443.
NOTE This document can be used as a criteria document for accreditation, peer assessment or other audit processes.
- Technical specification26 pagesEnglish languagee-Library read for1 day
This document complements the existing requirements of ISO/IEC 17021-1 for bodies providing audit and certification of quality management systems against ISO 19443. NOTE This document can be used as a criteria document for accreditation, peer assessment or other audit processes.
- Technical specification17 pagesEnglish languagesale 15% off
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This document is applicable to housed scintillators for registration and spectrometry of alpha-, beta-, gamma-, X-ray and neutron radiation. Their basic parameters such as a light output and intrinsic resolution are established. The document does not apply to gas or liquid scintillators and scintillators for counting or
current measurement.
- Standard21 pagesEnglish languagee-Library read for1 day
This document is applicable to housed scintillators for registration and spectrometry of alpha-, beta-, gamma-, X-ray and neutron radiation. Their basic parameters such as a light output and intrinsic resolution are established. The document does not apply to gas or liquid scintillators and scintillators for counting or current measurement.
- Standard21 pagesEnglish languagee-Library read for1 day
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This document provides guidance on the implementation of the ISO 19443 requirements, with examples of possible steps an organization can take to meet the requirements.
It does not add to, subtract from, or in any way modify those requirements.
This document does not prescribe mandatory approaches to implementation, or provide any preferred method of interpretation.
- Technical report65 pagesEnglish languagee-Library read for1 day
This document provides guidance on the implementation of the ISO 19443 requirements, with examples of possible steps an organization can take to meet the requirements.
It does not add to, subtract from, or in any way modify those requirements.
This document does not prescribe mandatory approaches to implementation, or provide any preferred method of interpretation.
- Technical report65 pagesEnglish languagee-Library read for1 day
This International Standard specifies requirements for a quality management system when an organization:
a) needs to demonstrate its ability to consistently provide products and services that meet customer and applicable statutory and regulatory requirements, and
b) aims to enhance customer satisfaction through the effective application of the system, including processes for improvement of the system and the assurance of conformity to customer and applicable statutory and regulatory requirements.
All the requirements of this International Standard are generic and are intended to be applicable to any organization, regardless of its type or size, or the products and services it provides.
NOTE 1 In this International Standard, the terms "product" or "service" only apply to products and services intended for, or required by, a customer.
NOTE 2 Statutory and regulatory requirements can be expressed as legal requirements.
This International Standard applies to organizations supplying ITNS products or services.
Application of this standard to organizations performing activities on a licensed nuclear site is subject to prior agreement by the Licensee.
Requirements specified in this International Standard are complementary (not alternative) to customer and applicable statutory and regulatory requirements.
- Standard59 pagesEnglish languagee-Library read for1 day
This International Standard specifies requirements for a quality management system when an organization:
a) needs to demonstrate its ability to consistently provide products and services that meet customer and applicable statutory and regulatory requirements, and
b) aims to enhance customer satisfaction through the effective application of the system, including processes for improvement of the system and the assurance of conformity to customer and applicable statutory and regulatory requirements.
All the requirements of this International Standard are generic and are intended to be applicable to any organization, regardless of its type or size, or the products and services it provides.
NOTE 1 In this International Standard, the terms "product" or "service" only apply to products and services intended for, or required by, a customer.
NOTE 2 Statutory and regulatory requirements can be expressed as legal requirements.
This International Standard applies to organizations supplying ITNS products or services.
Application of this standard to organizations performing activities on a licensed nuclear site is subject to prior agreement by the Licensee.
Requirements specified in this International Standard are complementary (not alternative) to customer and applicable statutory and regulatory requirements.
- Standard59 pagesEnglish languagee-Library read for1 day
This document gives the rules of naming, technical requirements, test methods, inspection, marking, packaging, transportation, storage and accompanying documents for electron linear accelerator equipment for Non-Destructive Testing (NDT). This document applies to NDT electron linear accelerator equipment in the X-ray energy range of 1 MeV to 15 MeV, including the accelerator equipment for radiographic film,computed radiography with imaging plates, real-time imaging, digital detector array and industrial computerized tomography.
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IEC 63175:2021 is applicable to hydrogen ion H- acceleration proton cyclotrons with one or more fixed energies within the range of 10 MeV to less than 30 MeV and a beam intensity equal to or greater than 300 μA. This document specifies the performance and safety requirements, structure, technical requirements, test methods, identification, packing, transportation, storage and accompanying documents for such cyclotrons. This document is intended for manufacturers of high intensity proton cyclotron within the energy range of 10 MeV to less than 30 MeV, and responsible organizations where such cyclotrons are installed.
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IEC 62976:2017 gives the rules of naming, technical requirements, test methods, inspection, marking, packaging, transportation, storage and accompanying documents for electron linear accelerator equipment for Non-Destructive Testing (NDT). This document applies to NDT electron linear accelerator equipment in the X-ray energy range of 1 MeV to 15 MeV, including the accelerator equipment for radiographic film, computed radiography with imaging plates, real-time imaging, digital detector array and industrial computerized tomography.
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IEC 63148:2021 specifies the requirements of tracking systems for radioactive materials. Such systems identify and locate the position of the radioactive materials transported using global navigation satellite systems (GNSS) and radio frequency identification (RFID).
The system provides a set of safety controls of the radioactive material, by which the transporter can improve safety during transportation. This document may also be used as supplementary guidance to the regulatory body.
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This document specifies the format of binary list-mode data at the output of digital data acquisition devices used for the detection and measurement of radiation. Such data acquisition devices may employ digital signal processors (DSP) and field-programmable gate arrays (FPGA) in combination with memory and a communication interface with a computer.
- Standard113 pagesEnglish languagee-Library read for1 day
This document specifies the format of binary list-mode data at the output of digital data acquisition devices used for the detection and measurement of radiation. Such data acquisition devices may employ digital signal processors (DSP) and field-programmable gate arrays (FPGA) in combination with memory and a communication interface with a computer.
- Standard113 pagesEnglish languagee-Library read for1 day
IEC 62372:2021 is applicable to housed scintillators for registration and spectrometry of alpha-, beta-, gamma-, X-ray and neutron radiation. This document specifies the requirements for the testing equipment and test methods of the basic parameters, of housed scintillators, such as:
- the direct method is applicable to measure the light output of housed scintillators based on scintillation material. The housed scintillators certified by this method can be used as working standard of housed scintillators (hereinafter: working standard) when performing measurements by a relative method of comparison.
- the relative method of comparison with the working standard is applicable to housed scintillators based on the same scintillation material as the working standard.
This second edition cancels and replaces the first edition published in 2006. This edition includes the following significant technical changes with respect to the previous edition:
- Title has been modified.
- To review the existing requirements and to update the terminology, definitions and normative references.
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SIGNIFICANCE AND USE
4.1 The mission of an analytical laboratory is to provide quality analyses on nuclear fuel cycle materials. An analytical laboratory QA program is comprised of planned and systematic actions needed to provide confidence that this mission is conducted in an acceptable and consistent manner.
4.2 The analytical laboratories involved in the analysis of nuclear fuel cycle materials are required to implement a documented QA program. Regulatory agencies may mandate some form of control requirements for all or a part of a laboratory's operation. A documented QA program is also necessary for those laboratory operations required to comply with ASME NQA-1 or ISO/IEC 17025, or the requirements of many accreditation bodies. Even when not mandated, laboratory QA programs should be established as a sound and scientific technical practice. This guide provides guidance for establishing and maintaining a QA program to control those analytical operations vital to ensuring the quality of chemical analyses.
4.3 Quality assurance programs are designed and implemented by organizations to assure that the quality requirements for a process, product or service will be fulfilled. The quality system is complementary to technical requirements that may be specific to a process or analytical method. Each laboratory should identify applicable program requirements and use standards to implement a quality program that meets the appropriate requirement. This guide may be used to develop and implement an analytical laboratory QA program. Other useful implementation standards and documents are listed in Section 2 and Appendix X1.
4.4 The guides for QA in the analytical laboratory within the nuclear fuel cycle have been written to provide guidance for each of the major activities in the laboratory and are displayed in Fig. 1. The applicable standard for each subject is noted in the following sections.
FIG. 1 Essential Elements of Analytical Laboratory Quality Assurance System
4.5 Althoug...
SCOPE
1.1 This guide covers the establishment and maintenance of a quality assurance (QA) program for analytical laboratories within the nuclear industry. References to key elements of ASME NQA-1 and ISO/IEC 17025 provide guidance to the functional aspects of analytical laboratory operations. When implemented as recommended, the practices presented in this guide will provide a comprehensive QA program for the laboratory. The practices are grouped by functions, which constitute the basic elements of a laboratory QA program.
1.2 The essential, basic elements of a laboratory QA program appear in the following order:
Section
Organization
5
Quality Assurance Program
6
Training and Qualification
7
Procedures
8
Laboratory Records
9
Control of Records
10
Management of Customer Requests and Commitments to Customers
11
Control of Procurement
12
Control of Measuring Equipment and Materials
13
Control of Measurements
14
Control of Nonconforming Work
15
Candidate Actions
16
Preventative Actions
17
1.3 Collection of samples and associated sampling procedures are outside the scope of this guide. The user may refer to sampling practices developed by Subcommittee C26.02.
1.4 Nuclear laboratories are required to handle a variety of hazardous materials, including but not limited to radioactive samples and materials. The need for proper handling of these materials is discussed in 13.2.4. While this guide focuses on the nuclear laboratory QA program, proper handling of nuclear materials is essential for proper function of the QA program.
1.5 This international standard was developed in accordance with internationally recognized principles on standardization established in the Decision on Principles for the Development of International Standards, Guides and Recommendations issued by the World Trade Organization Technical Barriers to Tra...
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- Guide12 pagesEnglish languagesale 15% off
SIGNIFICANCE AND USE
4.1 A laboratory quality assurance program is an essential program for laboratories within the nuclear industry. Guide C1009 provides guidance for establishing a quality assurance program for an analytical laboratory within the nuclear industry. This guide deals with the control of measurements aspect of the laboratory quality assurance program. Fig. 1 shows the relationship of measurement control with other essential aspects of a laboratory quality assurance program.
FIG. 1 Quality Assurance of Analytical Laboratory Data
4.2 The fundamental purposes of a measurement control program are to provide the with-use assurance (real-time control) that a measurement system is performing satisfactorily and to provide the data necessary to quantify measurement system performance. The with-use assurance is usually provided through the satisfactory analysis of quality control samples (reference value either known or unknown to the analyst). The data necessary to quantify measurement system performance is usually provided through the analysis of quality control samples or the duplicate analysis of process samples, or both. In addition to the analyses of quality control samples, the laboratory quality control program should address (1) the preparation and verification of standards and reagents, (2) data analysis procedures and documentation, (3) calibration and calibration procedures, (4) measurement method qualification, (5) analyst qualification, and (6) other general program considerations. Other elements of laboratory quality assurance also impact the laboratory quality control program. These elements or requirements include (1) chemical analysis procedures and procedure control, (2) records storage and retrieval requirements, (3) internal audit requirements, (4) organizational considerations, and (5) training/qualification requirements. To the extent possible, this standard will deal primarily with quality control requirements rather than overall quality assurance re...
SCOPE
1.1 This guide provides guidance for establishing and maintaining a measurement system quality control program. Guidance is provided for general program considerations, preparation of quality control samples, analysis of quality control samples, quality control data analysis, analyst qualification, measurement system calibration, measurement method qualification, and measurement system maintenance.
1.2 This guidance is provided in the following sections:
Section
General Quality Control Program Considerations
5
Quality Control Samples
6
Analysis of Quality Control Samples
7
Quality Control Data Analysis
8
Analyst Qualification
9
Measurement System Calibration
10
Qualification of Measurement Methods and Systems
11
Measurement System Maintenance
12
1.3 This international standard was developed in accordance with internationally recognized principles on standardization established in the Decision on Principles for the Development of International Standards, Guides and Recommendations issued by the World Trade Organization Technical Barriers to Trade (TBT) Committee.
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- Guide7 pagesEnglish languagesale 15% off
SIGNIFICANCE AND USE
4.1 Quality assurance, as covered by this practice, comprises all those planned and systematic actions necessary to provide adequate confidence that safety-related coating work in nuclear facilities as defined in Guide D5144, will perform satisfactorily in service.
4.2 It is not practical to impose all the requirements of this practice on certain specific items that require only a small quantity of coating material. The licensee, consistent with his formal Quality Assurance Program, may accept affidavits of compliance or certification attesting to the quality of a shop or field coating for such items. If required by licensing commitment; safety-related coatings that are not qualified or for which the quantification basis is indeterminate as defined in Guide D5144, shall be identified, quantified, and documented.
4.3 This practice may be incorporated in a project specification by direct reference or may be used to provide guidelines for the quality assurance program for coatings, on the basis of the licensee’s requirements. Effective use of this practice may also require the incorporation of applicable sections in project specifications for coatings on concrete, steel, equipment, and other related items.
SCOPE
1.1 This standard replaces ANSI N101.4 and provides a common basis for, and specifically comprises quality assurance requirements applicable to, safety-related protective coating work in Coating Service Level I areas of nuclear facilities as defined in Guide D5144.
1.2 This standard meets the requirements of ANSI N101.4 while also recognizing advancements in technology and industry practices since transfer to ASTM responsibility for updating, rewriting, and issuing replacement standards to ANSI N101.4.
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, health, and environmental practices and determine the applicability of regulatory limitations prior to use.
1.4 This international standard was developed in accordance with internationally recognized principles on standardization established in the Decision on Principles for the Development of International Standards, Guides and Recommendations issued by the World Trade Organization Technical Barriers to Trade (TBT) Committee.
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IEC 63048:2020 defines the general requirements for Mobile Remotely Controlled Systems (MRCSs) for nuclear and radiological applications such as integrity inspections, repair of components, handling of radioactive materials, and monitoring of physical conditions and radiation dose intensity in specific areas.
This document applies to MRCSs that are used to support nuclear and radiological facilities. These general requirements encompass high-level performance requirements regarding sensors, monitoring devices, control devices, interfacing mechanisms, simulation methods, and verification methods thereof in a normal environment or extreme environmental conditions, such as high radiation, high temperature, and high humidity environments.
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This document provides guidance on the implementation of the ISO 19443 requirements, with examples of possible steps an organization can take to meet the requirements. It does not add to, subtract from, or in any way modify those requirements. This document does not prescribe mandatory approaches to implementation, or provide any preferred method of interpretation.
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ABSTRACT
This specification defines essential criteria for all material combinations in boron-based neutron-absorbing material systems used for nuclear spent fuel storage racks in nuclear light water reactors, spent-fuel assemblies, or disassembled components. The boron-based neutron absorbing materials normally consist of metallic boron or a boron-containing boron compound supported by a matrix of aluminum, steel, or other materials. Material systems covered in this specification should always be capable of maintaining a B10 areal density that can support the required subcriticality depending on the design specification for service life.
SCOPE
1.1 This specification defines criteria for boron-based neutron absorbing material systems used in racks in a pool environment for storage of nuclear light water reactor (LWR) spent-fuel assemblies or disassembled components to maintain sub-criticality in the storage rack system.
1.2 Boron-based neutron absorbing material systems normally consist of metallic boron or a chemical compound containing boron (for example, boron carbide, B4C) supported by a matrix of aluminum, steel, or other materials.
1.3 In a boron-based absorber, neutron absorption occurs primarily by the boron-10 isotope that is present in natural boron to the extent of 18.3 ± 0.2 % by weight (depending upon the geological origin of the boron). Boron enriched in boron-10 could also be used.
1.4 The materials systems described herein shall be functional (that is, always be capable to maintain a boron-10 areal density such that subcriticality is maintained depending on the design specification for the service life in the operating environment of a nuclear spent fuel pool).
1.5 Observance of this specification does not relieve the user of the obligation to conform to all applicable international, national, and local regulations.
1.6 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, health, and environmental practices and determine the applicability of regulatory limitations prior to use.
1.7 This international standard was developed in accordance with internationally recognized principles on standardization established in the Decision on Principles for the Development of International Standards, Guides and Recommendations issued by the World Trade Organization Technical Barriers to Trade (TBT) Committee.
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ABSTRACT
This specification covers the required content and provides retention requirements for records needed for in-processing of nuclear facility transient workers. This specification is not intended to cover specific skills records (such as equipment operating licenses, ASME inspection qualifications, or welding certifications), nor does it reduce any regulatory requirement for records retention at a licensed nuclear facility. Rather, the records shall contain data elements for the following information: worker identification; occupational external radiation exposure; occupational internal radiation exposure; lifetime occupational radiation exposure history; security screening; medical approval; and radiation worker training.
SIGNIFICANCE AND USE
4.1 The standardization of nuclear facility transient worker records will provide the individual transient worker with a greater assurance that the radiation doses that may be received are within regulatory limits.
4.2 This specification establishes a fixed content for nuclear facility transient worker records. Standardizing the content of these records will facilitate interfacing with industry-wide record keeping systems, such as the Nuclear Energy Institute (NEI) Personnel Data System (PADS).
4.3 The standardization of nuclear facility transient worker records will reduce the time required for in-processing of these workers.
SCOPE
1.1 This specification covers the required content and provides retention requirements for records needed for in-processing of nuclear facility transient workers.
1.2 This specification applies to records to be used for in-processing only.
1.3 This specification is not intended to cover specific skills records (such as equipment operating licenses, ASME inspection qualifications, or welding certifications).
1.4 This specification doesPresident Barack Obama eulogy at John Lewis funeral not reduce any regulatory requirement for records retention at a licensed nuclear facility.
Note 1: Nuclear facilities operated by the U.S. Department of Energy (DOE) are not licensed by the U.S. Nuclear Regulatory Commission (NRC), nor are other nuclear facilities that may come under the control of the U.S. Department of Defense (DOD) or individual agreement states. The references in this specification to licensee, the U.S. NRC Regulatory Guides, and Title 10 of the U.S. Code of Federal Regulations are to imply appropriate alternative nomenclature with respect to DOE, DOD, or agreement state nuclear facilities. This distinction does not alter the required content of records needed for in-processing of nuclear facility transient workers.
Note 2: This specification does not define the form of the required worker records (such as a passport or central computerized record keeping system).
1.5 This international standard was developed in accordance with internationally recognized principles on standardization established in the Decision on Principles for the Development of International Standards, Guides and Recommendations issued by the World Trade Organization Technical Barriers to Trade (TBT) Committee.
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SIGNIFICANCE AND USE
4.1 This standard guide applies to concrete that is still in place with a defined geometry and known, documented history.
4.2 It is not intended for use on concrete that has already been rubbelized where it is difficult to measure the radiation levels and not easy to remove surface contamination to reduce radiation levels after concrete has been rubbelized.
4.3 This standard guide applies to surface or volumetrically contaminated concrete, where the depth of contamination can be measured or estimated based on the history of the concrete.
4.4 This standard guide does not apply to the reinforcement bar (rebar) found in concrete. Although most concrete contains rebar, it is generally removed before the concrete is dispositioned. In addition, rebar may be activated, and is covered under procedures for reuse of scrap metal.
4.5 General unit-dose and unit-cost data to support the calculations is provided in the appendices of this standard guide. However, if site-specific data is available, it should be used instead of the general information provided here.
4.6 This standard guide helps determine estimated doses to the public during disposal of concrete and to future residents of disposal areas. It does not include dose to radiation workers already involved in a radiation control program. It is assumed that the dose to radiation workers is already tracked and kept within acceptable levels through a radiation control program. The cost and dose to radiation workers could be added in to find an overall cost and dose for each option.
SCOPE
1.1 This standard guide defines the process for developing a strategy for dispositioning concrete from nuclear facility decommissioning. It outlines a 10-step method to evaluate disposal options for radioactively contaminated concrete. One of the steps is to complete a detailed analysis of the cost and dose to nonradiation workers (the public); the methodology and supporting data to perform this analysis are detailed in the appendices. The resulting data can be used to balance dose and cost and select the best disposal option. These data, which establish a technical basis to apply to release the concrete, can be used in several ways: (1) to show that the release meets existing release criteria, (2) to establish a basis to request release of the concrete on a case-by-case basis, (3) to develop a basis for establishing release criteria where none exists.
1.2 This standard guide is based on the “Protocol for Development of Authorized Release Limits for Concrete at U.S. Department of Energy Sites,” (1)2 from which the analysis methodology and supporting data are taken.
1.3 Guide E1760 provides a general process for release of materials containing residual amounts of radioactivity. In addition, Guide E1278 provides a general process for analyzing radioactive pathways. This standard guide is intended for use in conjunction with Guides E1760 and E1278, and provides a more detailed approach for the release of concrete.
1.4 This international standard was developed in accordance with internationally recognized principles on standardization established in the Decision on Principles for the Development of International Standards, Guides and Recommendations issued by the World Trade Organization Technical Barriers to Trade (TBT) Committee.
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This document contains the terms, definitions, notes to entry and examples corresponding to the basic concepts of the nuclear energy, nuclear technologies, and radiological protection subject fields. It provides the minimum essential information for each cross-cutting concept represented by a single term. NOTE A full understanding of concepts goes with a background knowledge of nuclear energy, nuclear technologies, and radiological protection. It is intended to facilitate communication and promote common understanding.
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This document gives the rules of naming, technical requirements, test methods, inspection, marking, packaging, transportation, storage and accompanying documents for electron linear accelerator equipment for Non-Destructive Testing (NDT).
This document applies to NDT electron linear accelerator equipment in the X-ray energy range of 1 MeV to 15 MeV, including the accelerator equipment for radiographic film,computed radiography with imaging plates, real-time imaging, digital detector array and industrial computerized tomography.
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SIGNIFICANCE AND USE
4.1 This guide presents concise guidance and approach to developing a test document for qualifying a coating for CSLI service, whether a new or existing coating. Guidance for evaluating existing qualification test data for applicability is presented in Guide D8104.
4.2 The requirements for qualification testing can be found in Quality Assurance Criteria III (Design Control), IX (Control of Special Processes), and XI (Test Control) of 10 CFR 50, Appendix B, as implemented, respectively, by Requirements III, IX, and XI of NQA-1. A test document developed per this guide is intended to be compliant with these requirements.
4.3 This guide implements the guidance provided in Guide D5144 for qualification of coatings for use in CSLI service. Additional guidance is provided in Regulatory Guide 1.54, Revisions 0 through 3, as may be invoked by the licensee.
4.4 For plants with a license basis that predates the requirements of ANSI N5.12 and N101.2, this guide also is applicable. For these plants, the coatings or coating systems may be designated as acceptable, rather than qualified.
4.5 All qualification testing shall comply with the licensee’s approved quality assurance program.
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1.1 This guide provides an approach to identifying the need for and development of a test document to qualify coatings for Coating Service Level I (CSLI) service in nuclear power plants.
1.2 It is the intent of this guide to provide a recommended basis for establishing a coatings qualification test document, not to mandate a singular basis for all test documents. Variations or simplifications of the process described in this guide may be appropriate for any given operating or new construction nuclear power plant depending on its licensing commitments.
1.3 The values stated in inch-pound units are to be regarded as standard. The values given in parentheses are mathematical conversions to SI units that are provided for information only and are not considered standard.
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, health, and environmental practices and determine the applicability of regulatory limitations prior to use.
1.5 This international standard was developed in accordance with internationally recognized principles on standardization established in the Decision on Principles for the Development of International Standards, Guides and Recommendations issued by the World Trade Organization Technical Barriers to Trade (TBT) Committee.
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SIGNIFICANCE AND USE
5.1 Information is provided in this document and other referenced documents to assist the licensee and the licensor in analyzing the materials aspects of performance of SNF and DCSS components during extended storage. The effects of the service conditions of the first licensing period are reviewed in the license renewal process. These service conditions are highlighted and discussed in Annex A1 as factors that affect materials performance in an ISFSI. Emphasis is on the effects of time, temperature, radiation, and the environment on the condition of the SNF and the performance of components of ISFSI storage systems.
5.2 The storage of SNF that is irradiated under the regulations of 10 CFR Part 50 is governed by regulations in 10 CFR Part 72. Regulatory requirements for the subsequent geologic disposal of this SNF are presently given in 10 CFR Part 60, with specific requirements for the use of Yucca Mountain as a repository being given in the regulatory requirements of 10 CFR Part 63. Between the life-cycle phases of storage and disposal, SNF may be transported under the requirements of 10 CFR Part 71. Therefore, in storage, it is important to acknowledge the transport and disposal phases of the life cycle. In doing this, the materials properties that are important to these subsequent phases are to be considered in order to promote successful completion of these subsequent phases in the life cycle of SNF. Retrievability of SNF (or high-level radioactive waste) is set as a requirement in 10 CFR Part 72.122(g)(5) and 10 CFR Part 72.122(l). Care should be taken in operations conducted prior to disposal, for example, storage, transfer, and transport, to ensure that the SNF is not abused and that SNF assemblies will be retrievable, the protective value of the cladding is not degraded and remains capable of serving as an active barrier to radionuclide release during transfer and transport operations. It is possible that cladding could be altered during dry storage. The ...
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1.1 Part of the total inventory of commercial spent nuclear fuel (SNF) is stored in dry cask storage systems (DCSS) under licenses granted by the U.S. Nuclear Regulatory Commission (NRC). The purpose of this guide is to provide information to assist in supporting the renewal of these licenses, safely and without removal of the SNF from its licensed confinement, for periods beyond those governed by the term of the original license. This guide provides information on materials behavior under conditions that may be important to safety evaluations for the extended service of the renewal period. This guide is written for DCSS containing light water reactor (LWR) fuel that is clad in zirconium alloy material and stored in accordance with the Code of Federal Regulations (CFR), at an independent spent-fuel storage installation (ISFSI).2 The components of an ISFSI, addressed in this document, include the commercial SNF, canister, cask, and all parts of the storage installation including the ISFSI pad. The language of this guide is based, in part, on the requirements for a dry SNF storage license that is granted, by the U.S. Nuclear Regulatory Commission (NRC), for up to 20 years. Although government regulations may differ for various nations, the guidance on materials properties and behavior given here is expected to have broad applicability.
1.2 This guide addresses many of the factors affecting the time-dependent behavior of materials under ISFSI service [10 CFR Part 72.42]. These factors are those regarded to be important to performance, in license extension, beyond the currently licensed 20-year period. Examples of these factors are given in this guide and they include materials alterations or environmental conditions for components of an ISFSI system that, over time, could have significance related to safety. For purposes of this guide, a license period of an additional 20 to 80 years is assumed.
1.3 This guide address...
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SIGNIFICANCE AND USE
4.1 To describe the uncertainties of a standard test method, precision and bias statements are required.3 The formulation of these statements has been addressed from time to time, and at least two standards practices (Practices E177 and E691) have been issued. The 1986 Compilation of ASTM Standard Definitions (6)4 devotes several pages to these terms. This guide should not be used in cases where small numbers of test results do not support statistical normality.
4.2 The intent of this guide is to help analysts prepare and interpret precision and bias statements. It is essential that, when the terms are used, their meaning should be clear and easily understood.
4.3 Appendix X1 provides the theoretical foundation for precision and bias concepts and Practice E691 addresses the problem of sources of variation. To illustrate the interplay between sources of variation and formulation of precision and bias statements, a hypothetical data set is analyzed in Appendix X2. This example shows that depending on how the data was collected, different precision and bias statements are possible. Reference to this example will be found throughout this guide.
4.4 There has been much debate inside and outside the statistical community on the exact meaning of some statistical terms. Thus, following a number of the terms in Section 3 is a list of several ways in which that term has been used. This listing is not meant to indicate that these meanings are equivalent or equally acceptable. The purpose here is more to encourage clear definition of terms used than to take sides. For example, use of the term systematic error is discouraged by some. If it is to be used, the reader should be told exactly what is meant in the particular circumstance.
4.5 This guide is intended as an aid to understanding the statistical concepts used in precision and bias statements. There is no intention that this be a self-contained introduction to statistics. Since many analysts have no formal sta...
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1.1 This guide covers terminology useful for the preparation and interpretation of precision and bias statements. This guide does not recommend a specific error model or statistical method. It provides awareness of terminology and approaches and options to use for precision and bias statements.
1.2 In formulating precision and bias statements, it is important to understand the statistical concepts involved and to identify the major sources of variation that affect results. Appendix X1 provides a brief summary of these concepts.
1.3 To illustrate the statistical concepts and to demonstrate some sources of variation, a hypothetical data set has been analyzed in Appendix X2. Reference to this example is made throughout this guide.
1.4 It is difficult and at times impossible to ship nuclear materials for interlaboratory testing. Thus, precision statements for test methods relating to nuclear materials will ordinarily reflect only within-laboratory variation.
1.5 No units are used in this statistical analysis.
1.6 This guide does not involve the use of materials, operations, or equipment and does not address any risk associated.
1.7 This international standard was developed in accordance with internationally recognized principles on standardization established in the Decision on Principles for the Development of International Standards, Guides and Recommendations issued by the World Trade Organization Technical Barriers to Trade (TBT) Committee.
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SIGNIFICANCE AND USE
5.1 Refer to Practice E261 for a general discussion of the determination of decay rates, reaction rates, and neutron fluence rates with threshold detectors (1-29).3 Refer to Practice E1006, Practice E185 and Guide E1018 for the use and application of results obtained by this test method.(30-32)
5.2 The half-life of 93mNb is 16.1 (2)4 years5(34) and has a K X-ray emission probability of 0.11442 ± 3.356 % per decay (35). The Kα and Kβ X-rays of niobium are at 16.521–16.615 and 18.607–18.9852 keV, respectively (35). The recommended 93Nb(n,n′)93mNb cross section comes from the International Reactor Dosimetry and Fusion File (IRDFF version 1.05, cross section compendium (36), and is shown in Fig. 1. This nuclear data evaluation is part of the Russian Reactor Dosimetry File (RRDF), cross section evaluations (37). The nuclear decay data referenced here are not taken from the latest dosimetry recommended database (33) but are selected to be consistent with the nuclear data used in the recommended IRDFF evaluation.
FIG. 1 RRDF/IRDFF-1.05 Cross Section Versus Energy for the 93Nb(n,n′) 93mNb Reaction
5.3 Chemical dissolution of the irradiated niobium to produce very low mass-per-unit area sources is an effective way to obtain consistent results. The direct counting of foils or wires can produce satisfactory results provided appropriate methods and interpretations are employed. It is possible to use liquid scintillation methods to measure the niobium activity provided the radioactive material can be kept uniformly in solution and appropriate corrections can be made for interfering activities.
5.4 The measured reaction rates can be used to correlate neutron exposures, provide comparison with calculated reaction rates, and determine neutron fluences. Reaction rates can be determined with greater accuracy than fluence rates because of the current uncertainty in the cross section versus energy shape.
5.5 The 93Nb(n,n′)93mNb reaction has the desirable properties o...
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1.1 This test method describes procedures for measuring reaction rates by the activation reaction 93Nb(n,n′) 93mNb.
1.2 This activation reaction is useful for monitoring neutrons with energies above approximately 0.5 MeV and for irradiation times up to about 48 years (three half-lives), provided that the analysis methods described in Practice E261 are followed.
1.3 With suitable techniques, fast-neutron reaction rates for neutrons with energy distribution similar to fission neutrons can be determined in fast-neutron fluences above about 1016 cm−2. In the presence of high thermal-neutron fluence rates (>1012cm−2·s−1), the transmutation of 93mNb due to neutron capture should be investigated. In the presence of high-energy neutron spectra such as are associated with fusion and spallation sources, the transmutation of 93mNb by reactions such as (n,2n) may occur and should be investigated.
1.4 Procedures for other fast-neutron monitors are referenced in Practice E261.
1.5 Fast-neutron fluence rates can be determined from the reaction rates provided that the appropriate cross section information is available to meet the accuracy requirements.
1.6 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard.
1.7 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, health, and environmental practices and determine the applicability of regulatory limitations prior to use.
1.8 This international standard was developed in accordance with internationally recognized principles on standardization established in the Decision on Principles for the Development of International Standards, Guides and Recommendations issued by the World Trade Organization Technical Barriers to Trade (TBT) Committee.
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This International Standard specifies requirements for a quality management system when an organization: a) needs to demonstrate its ability to consistently provide products and services that meet customer and applicable statutory and regulatory requirements, and b) aims to enhance customer satisfaction through the effective application of the system, including processes for improvement of the system and the assurance of conformity to customer and applicable statutory and regulatory requirements. All the requirements of this International Standard are generic and are intended to be applicable to any organization, regardless of its type or size, or the products and services it provides. NOTE 1 In this International Standard, the terms "product" or "service" only apply to products and services intended for, or required by, a customer. NOTE 2 Statutory and regulatory requirements can be expressed as legal requirements. This International Standard applies to organizations supplying ITNS products or services. Application of this standard to organizations performing activities on a licensed nuclear site is subject to prior agreement by the Licensee. Requirements specified in this International Standard are complementary (not alternative) to customer and applicable statutory and regulatory requirements.
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- nomenclature of scintillation detectors was expanded by phoswich detector and single-line multi-channel detector;
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- review the existing requirements;
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