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ASTM E944-96

(Guide)

Standard Guide for Application of Neutron Spectrum Adjustment Methods in Reactor Surveillance, (IIA)

Standard Guide for Application of Neutron Spectrum Adjustment Methods in Reactor Surveillance, (IIA)

SCOPE
1.1 This guide covers the analysis and interpretation of the physics dosimetry for Light Water Reactor (LWR) surveillance programs. The main purpose is the application of adjustment methods to determine best estimates of neutron damage exposure parameters and their uncertainties.  
1.2 This guide is also applicable to irradiation damage studies in research reactors.
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.

General Information

Status
Historical
Publication Date
31-Dec-1995
ICS
27.120.20 - Nuclear power plants. Safety
Technical Committee
E10 - Nuclear Technology and Applications
Drafting Committee
E10.05 - Nuclear Radiation Metrology
Current Stage
Ref Project

Relations

Replaced By
ASTM E944-02 - Standard Guide for Application of Neutron Spectrum Adjustment Methods in Reactor Surveillance, E706 (IIA)
Effective Date
10-Jun-2002

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ASTM E944-96 - Standard Guide for Application of Neutron Spectrum Adjustment Methods in Reactor Surveillance, (IIA)ASTM E944-96 - Standard Guide for Application of Neutron Spectrum Adjustment Methods in Reactor Surveillance, (IIA)
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ASTM E944-96 - Standard Guide for Application of Neutron Spectrum Adjustment Methods in Reactor Surveillance, (IIA)
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Standards Content (Sample)

Designation: E 944 – 96
Standard Guide for
Application of Neutron Spectrum Adjustment Methods in
1
Reactor Surveillance, (IIA)
This standard is issued under the fixed designation E 944; the number immediately following the designation indicates the year of
original adoption or, in the case of revision, the year of last revision. A number in parentheses indicates the year of last reapproval. A
superscript epsilon (e) indicates an editorial change since the last revision or reapproval.
2,3
1. Scope for Reactor Vessel Surveillance, E 706(IID)
E 523 Test Method for Measuring Fast-Neutron Reaction
1.1 This guide covers the analysis and interpretation of the
2
Rates by Radioactivation of Copper
physics dosimetry for Light Water Reactor (LWR) surveillance
E 526 Test Method for Measuring Fast-Neutron Reaction
programs. The main purpose is the application of adjustment
2
Rates by Radioactivation of Titanium
methods to determine best estimates of neutron damage expo-
E 693 Practice for Characterizing Neutron Exposures in
sure parameters and their uncertainties.
Iron and Low Alloy Steels in Terms of Displacements Per
1.2 This guide is also applicable to irradiation damage
,
2 3
Atom (DPA), E 706(ID)
studies in research reactors.
E 704 Test Method for Measuring Reaction Rates by Ra-
1.3 This standard does not purport to address all of the
2
dioactivation of Uranium-238
safety concerns, if any, associated with its use. It is the
E 705 Test Method for Measuring Reaction Rates by Ra-
responsibility of the user of this standard to establish appro-
2
dioactivation of Neptunium-237
priate safety and health practices and determine the applica-
E 706 Master Matrix for Light-Water Reactor Pressure
bility of regulatory limitations prior to use.
2
Vessel Surveillance Standards
4
2. Referenced Documents
(IIE) Benchmark Testing of Reactor Vessel Dosimetry
(IIID) Application and Analysis of Damaged Monitors for
2.1 ASTM Standards:
4
Reactor Vessel Surveillance
E 170 Terminology Relating to Radiation Measurements
2
E 844 Guide for Sensor Set Design and Irradiation for
and Dosimetry
,
2 3
Reactor Surveillance, E 706(IIC)
E 262 Test Method for Determining Thermal Neutron Re-
2
E 853 Practice for Analysis and Interpretation of Light-
action and Fluence Rates by Radioactivation Techniques
,
2 3
Water Reactor Surveillance Results, E 706(IA)
E 263 Test Method for Measuring Fast-Neutron Reaction
2
E 854 Test Method for Application and Analysis of Solid
Rates by Radioactivation of Iron
State Track Recorder (SSTR) Monitors for Reactor Sur-
E 264 Test Method for Measuring Fast-Neutron Reaction
2,3
2
veillance, E 706(IIIB)
Rates by Radioactivation of Nickel
E 910 Test Method for Application and Analysis of Helium
E 265 Test Method for Measuring Reaction Rates and
2
Accumulation Fluence Monitors for Reactor Vessel Sur-
Fast-Neutron Fluences by Radioactivation of Sulfur-32
,
2 3
veillance, E706(IIIC)
E 266 Test Method for Measuring Fast-Neutron Reaction
2
E 1005 Test Method for Application and Analysis of Radio-
Rates by Radioactivation of Aluminum
metric Monitors for Reactor Vessel Surveillance,
E 343 Test Method for Measuring Reaction Rates by Analy-
2
E 706(IIIA)
sis of Molybdenum-99 Radioactivity from Fission Dosim-
2
E 1018 Guide for Application of ASTM Evaluated Cross
eters
2,3
Section Data File, E706(IIB)
E 393 Test Method for Measuring Reaction Rates by Analy-
5
2
2.2 Nuclear Regulatory Commission Documents:
sis of Barium-140 from Fission Dosimeters
NUREG/CR-1861 PCA Experiments and Blind Test
E 481 Test Method for Measuring Neutron Fluence Rate by
2
NUREG/CR-2222 Theory and Practice of General Adjust-
Radioactivation of Cobalt and Silver
ment and Model Fitting Procedures
E 482 Guide for Application of Neutron Transport Methods
NUREG/CR-3318 LWR Pressure Vessel Surveillance Do-
simetry Improvement Program: PCA Experiments, Blind
1
This guide is under the jurisdiction of ASTM Committee E-10 on Nuclear
3
Technology and Applicationsand is the direct responsibility of Subcommittee The unnumbered references in parentheses refer to Section 5 as well as Figs. 1
E10.05on Nuclear Radiation Metrology. A brief overview of Guide E 944 appears in and 2 of Matrix E 706.
4
Master Matrix E 706 in 5.3.1(IIA). For standards that are in the draft stage and have not received an ASTM
Current edition approved July 10, 1996. Published August 1996. Originally designation, see Section 5 as well as Figs. 1 and 2 of Matrix E 706.
5
published as E 944 – 83. Last previous edition E944 – 89. Available from Superintendents of Documents, U. S. Government Printing
2
Annual Book of ASTM Standards, Vol 12.02. Office, Washington, DC 20402.
Copyright © ASTM, 100 Barr Harbor Drive, West Conshohocken, PA 19428-2959, United States.
1

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E 944
E
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Test, and Physics-Dosimetry Support for the PSF Experi-
s 5 F E s E dE/F
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ij * i j
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NUREG/CR-3319 LWR Power Reactor Surveillance
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...

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ASTM E853-23(Practice)
Standard Practice for Analysis and Interpretation of Light-Water Reactor Surveillance Neutron Exposure Results

SIGNIFICANCE AND USE
3.1 The objectives of a reactor vessel surveillance program are twofold. The first requirement of the program is to monitor changes in the fracture toughness properties of ferritic materials in the reactor vessel beltline region resulting from exposure to neutron irradiation and the thermal environment. The second requirement is to make use of the data obtained from the surveillance program to determine the conditions under which the vessel can be operated throughout its service life.  
3.1.1 To satisfy the first requirement of 3.1, the tasks to be carried out are straightforward. Each of the irradiation capsules that comprise the surveillance program may be treated as a separate experiment. The goal is to define and carry to completion a dosimetry program that will, a posteriori, describe the neutron field to which the material test specimens were exposed. The resultant information will then become part of a database applicable in a stricter sense to the specific plant from which the capsule was removed, but also in a broader sense to the industry as a whole.  
3.1.2 To satisfy the second requirement of 3.1, the tasks to be carried out are somewhat complex. The objective is to describe accurately the neutron field to which the pressure vessel itself will be exposed over its service life. This description of the neutron field must include spatial gradients within the vessel wall. Therefore, heavy emphasis must be placed on the use of neutron transport techniques as well as on the choice of a design basis for the computations. Since a given surveillance capsule measurement, particularly one obtained early in plant life, is not necessarily representative of long-term reactor operation, a simple normalization of neutron transport calculations to dosimetry data from a given capsule may not be appropriate (1-67).  
3.2 The objectives and requirements of a reactor vessel's support structure's surveillance program are much less stringent, and at present, are limited to ...
SCOPE
1.1 This practice covers the methodology, summarized in Annex A1, to be used in the analysis and interpretation of neutron exposure data obtained from LWR pressure vessel surveillance programs and, based on the results of that analysis, establishes a formalism to be used to evaluate present and future condition of the pressure vessel and its support structures2 (1-74).3  
1.2 This practice relies on, and ties together, the application of several supporting ASTM standard practices, guides, and methods (see Master Matrix E706) (1, 5, 13, 48, 49).2 In order to make this practice at least partially self-contained, a moderate amount of discussion is provided in areas relating to ASTM and other documents. Support subject areas that are discussed include reactor physics calculations, dosimeter selection and analysis, and exposure units.  
1.3 This practice is restricted to direct applications related to surveillance programs that are established in support of the operation, licensing, and regulation of LWR nuclear power plants. Procedures and data related to the analysis, interpretation, and application of test reactor results are addressed in Practice E1006, Guide E900, and Practice E1035.  
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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ASTM E3104-17(2023)(Specification)
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This specification prescribes the performance criteria for strippable/removable coatings used in immobilizing radioactive contamination, minimizing worker exposure, and facilitating subsequent decontamination or protecting uncontaminated areas against the spread of radioactive contamination. It covers the minimum performance requirements (shelf life, tensile strength, adhesion, abrasion resistance, dry/cure time, decontamination factor, airborne release fraction) as well as the mechanical and chemical properties for strippable/removable coatings. The strippable/removable coating is intended to reduce: migration of the radioactive contamination into or along buildings, equipment, and other surfaces; resuspension of contamination into the air and the airborne intake hazards of the contamination; and the spread of contamination as a result of external forces such as pedestrian traffic. The strippable/removable coating shall: be applicable to both vertical and horizontal surfaces; work within a range of environmental and radiological conditions; and be readily applied to both porous and nonporous materials such as concrete, wood, metal, ceramics, and plastics. Furthermore, the strippable/removable coating may include constituents that will physically or chemically bind and immobilize radioactive contamination.
SCOPE
1.1 This specification is intended to provide a basis for identification of strippable/removable materials used to immobilize radioactive contamination, minimize worker exposure, and facilitate subsequent decontamination or to protect uncontaminated areas against the spread of radioactive contamination.  
1.2 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.  
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ASTM E3105-17(2023)(Specification)
Standard Specification for Permanent Coatings Used to Mitigate Spread of Radioactive Contamination

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This specification prescribes the performance criteria for non-removable permanent coatings and fixatives as a long-term measure used to immobilize radioactive contamination, minimize worker exposure, and protect uncontaminated areas against the spread of radioactive contamination. It covers the minimum performance requirements (shelf life, adhesion, abrasion resistance, dry/cure time, decontamination factor, airborne release fraction, respirable fraction, radiation resistance) as well as the mechanical and chemical properties for permanent coatings that are intended to immobilize dispersible radioactive contamination deposited on buildings and equipment as might result from anticipated to unanticipated events to include normal operating conditions, decommissioning, and radiological release. The coating is intended to reduce: migration of the contamination into or along buildings, equipment, and other surfaces; resuspension of contamination into the air; and the spread of contamination as a result of external forces such as pedestrian traffic. It shall: be applicable to both vertical and horizontal surfaces; work within a range of environmental and radiological conditions; and be applicable to both porous and nonporous materials such as concrete, wood, metal, ceramics, and plastics. Furthermore, the coating may include constituents that will physically or chemically bind and hold radioactive contamination.
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ASTM E1035-18(2023)(Practice)
Standard Practice for Determining Neutron Exposures for Nuclear Reactor Vessel Support Structures

SIGNIFICANCE AND USE
3.1 Prediction of neutron radiation effects to pressure vessel steels has long been a part of the design and operation of light water reactor power plants. Both the federal regulatory agencies (see 2.3) and national standards groups (see 2.1 and 2.2) have promulgated regulations and standards to ensure safe operation of these vessels. The support structures for pressurized water reactor vessels may also be subject to similar neutron radiation effects (1, 3-6).2 The objective of this practice is to provide guidelines for determining the neutron radiation exposures experienced by individual vessel supports.  
3.2 It is known that high-energy photons can also produce displacement damage effects that may be similar to those produced by neutrons. These effects are known to be much less at the belt line of a light water reactor pressure vessel than those induced by neutrons. The same has not been proven for all locations within vessel support structures. Therefore, it may be prudent to apply coupled neutron-photon transport methods and photon-induced displacement cross sections to determine whether gamma-induced dpa exceeds the screening level of 3.0 × 10–4 used in this practice for neutron exposures. (See 1.3.)
SCOPE
1.1 This practice covers procedures for monitoring the neutron radiation exposures experienced by ferritic materials in nuclear reactor vessel support structures located in the vicinity of the active core. This practice includes guidelines for:  
1.1.1 Selecting appropriate dosimetric sensor sets and their proper installation in reactor cavities.  
1.1.2 Making appropriate neutronics calculations to predict neutron radiation exposures.  
1.2 The values stated in SI units are to be regarded as standard; units that are not SI can be found in Terminology E170 and are to be regarded as standard. Any values in parentheses are for information only.  
1.3 This practice is applicable to all pressurized water reactors whose vessel supports will experience a lifetime neutron fluence (E > 1 MeV) that exceeds 1 × 1017 neutrons/cm2 or exceeds 3.0 × 10−4 dpa (1).2 (See Terminology E170.)  
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Standard Guide for Use of Melt Wire Temperature Monitors for Reactor Vessel Surveillance

SIGNIFICANCE AND USE
3.1 Temperature monitors are used in surveillance capsules in accordance with Practice E2215 to estimate the maximum value of the surveillance specimen irradiation temperature. Temperature monitors are needed to give evidence of overheating of surveillance specimens beyond the expected temperature. Because overheating causes a reduction in the amount of neutron radiation damage to the surveillance specimens, this overheating could result in a change in the measured properties of the surveillance specimens that would lead to an unconservative prediction of damage to the reactor vessel material.  
3.2 The magnitude of the reduction of radiation damage with overheating depends on the composition of the material and time at temperature. Guide E900 provides an accepted method for quantifying the temperature effect. Because the evidence from melt wire monitors gives no indication of the duration of overheating above the expected temperature as indicated by melting of the monitor, the significance of overheating events cannot be quantified on the basis of temperature monitors alone. Indication of overheating does serve to alert the user of the data to further evaluate the irradiation temperature exposure history of the surveillance capsule.  
3.3 This guide is included in Master Matrix E706 that relates several standards used for irradiation surveillance of light-water reactor vessel materials. It is intended primarily to amplify the requirements of Practice E185 in the design of temperature monitors for the surveillance program. It may also be used in conjunction with Practice E2215 to evaluate the post-irradiation test measurements.
SCOPE
1.1 This guide describes the application of melt wire temperature monitors and their use for reactor vessel surveillance of light-water power reactors as called for in Practices E185 and E2215.  
1.2 The purpose of this guide is to recommend the selection and use of the common melt wire technique where the correspondence between melting temperature and composition of different alloys is used as a passive temperature monitor. Guidelines are provided for the selection and calibration of monitor materials; design, fabrication, and assembly of monitor and container; post-irradiation examinations; interpretation of the results; and estimation of uncertainties.  
1.3 The values stated in SI units are to be regarded as standard. The values given in parentheses are mathematical conversions to inch-pound units that are provided for information only and are not considered standard.  
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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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Standard Guide for Application of Neutron Transport Methods for Reactor Vessel Surveillance

SIGNIFICANCE AND USE
3.1 General:  
3.1.1 The methodology recommended in this guide specifies criteria for validating computational methods and outlines procedures applicable to pressure vessel related neutronics calculations for test and power reactors. The material presented herein is useful for validating computational methodology and for performing neutronics calculations that accompany reactor vessel surveillance dosimetry measurements (see Master Matrix E706 and Practice E853). Briefly, the overall methodology involves: (1) methods-validation calculations based on at least one well-documented benchmark problem, and (2) neutronics calculations for the facility of interest. The neutronics calculations of the facility of interest and of the benchmark problem should be performed consistently, with important modeling parameters kept the same or as similar as is feasible. In particular, the same energy group structure and common broad-group microscopic cross sections should be used for both problems. Further, the benchmark problem should be characteristically similar to the facility of interest. For example, a power reactor benchmark should be utilized for power reactor calculations. Non-power reactors may have special features that may affect pressure vessel fluence and require consideration when developing a benchmark, such as beam tubes, irradiation facilities, and non-core neutron sources. The neutronics calculations involve two tasks: (1) determination of the neutron source distribution in the reactor core by utilizing diffusion theory (or transport theory) calculations in conjunction with reactor power distribution measurements, and (2) performance of a fixed fission rate neutron source (fixed-source) transport theory calculation to determine the neutron fluence rate distribution in the reactor core, through the internals and in the pressure vessel. Some neutronics modeling details for the benchmark, test reactor, or the power reactor calculation will differ; therefore, the proc...
SCOPE
1.1 Need for Neutronics Calculations—An accurate calculation of the neutron fluence and fluence rate at several locations is essential for the analysis of integral dosimetry measurements and for predicting irradiation damage exposure parameter values in the pressure vessel. Exposure parameter values may be obtained directly from calculations or indirectly from calculations that are adjusted with dosimetry measurements; Guide E944 and Practice E853 define appropriate computational procedures.  
1.2 Methodology—Neutronics calculations for application to reactor vessel surveillance encompass three essential areas: (1) validation of methods by comparison of calculations with dosimetry measurements in a benchmark experiment, (2) determination of the neutron source distribution in the reactor core, and (3) calculation of neutron fluence rate at the surveillance position and in the pressure vessel.  
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Standard Guide for Environmental Monitoring Plans for Decommissioning of Nuclear Facilities

SIGNIFICANCE AND USE
5.1 Use of this guide will ensure that the potential impact on the surrounding environment from planned decommissioning activities has been properly assessed.  
5.2 Use of this guide will ensure that the adequacy of environmental sampling has been assessed for location, frequency, analytical techniques, and media type to monitor the environment and to detect site-related releases and their impact.
SCOPE
1.1 This guide covers the development or assessment of environmental monitoring plans for decommissioning nuclear facilities. This guide addresses: (1) development of an environmental baseline prior to commencement of decommissioning activities; (2) determination of release paths from site activities and their associated exposure pathways in the environment; and (3) selection of appropriate sampling locations and media to ensure that all exposure pathways in the environment are monitored appropriately. This guide also addresses the interfaces between the environmental monitoring plan and other planning documents for site decommissioning, such as radiation protection, site characterization, and waste management plans, and federal, state, and local environmental protection laws and guidance. This guide is applicable up to the point of completing D&D activities and the reuse of the facility or area for other purposes.  
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Standard Guide for Post-Deactivation Surveillance and Maintenance of Radiologically Contaminated Facilities

SIGNIFICANCE AND USE
4.1 The purpose of this guide is to provide the user information and guidance for preparation of a plan for the surveillance and maintenance of nuclear facilities that have been deactivated and are awaiting D&D.  
4.1.1 This document provides guidance for performing S&M in a way that will ensure worker and public safety, while also addressing stakeholder requirements.  
4.1.2 Use of this guide helps standardize the basic requirements for S&M of nuclear facilities.  
4.2 Use of this guide helps ensure that the S&M plan addresses the significant activities and actions necessary to maintain these facilities in a safe and stable condition until they can be decommissioned.
SCOPE
1.1 This guide outlines a method for developing a Surveillance and Maintenance (S&M) plan for inactive nuclear facilities. It describes the steps and activities necessary to prevent loss or release of radioactive or hazardous materials, and to minimize physical risks between the deactivation phase and the start of facility decontamination and decommissioning (D&D).  
1.2 The primary concerns for S&M are related to (1) animal intrusion, (2) structural integrity degradation,  (3) water in-leakage, (4) contamination migration, (5) unauthorized personnel entry, and (6) theft/intrusion. This document is intended to serve as a guide only, and is not intended to modify existing regulations.  
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ASTM E1892-15(2021)(Guide)
Standard Guide for Preparing Characterization Plans for Decommissioning Nuclear Facilities

SIGNIFICANCE AND USE
5.1 Knowledge of the nature and extent of contamination in a nuclear facility to be decommissioned is crucial to choosing the optimum methods for decontamination and decommissioning, and estimating the resulting waste volumes and personnel exposures. Implementing a characterization plan, developed in accordance with this standard, will result in obtaining or deriving the above information.  
5.2 Information on the proposed decommissioning methods, waste volumes, and estimated personnel radiation exposures can be used to define the overall work scope, costs, schedules, and manpower needs for the decommissioning project. This information may be included in the Decommissioning Plan. The extent of over- or under-estimating these project parameters will be a function of the sampling plan and statistical designs, described in Sections 6.1.4 and 6.1.5.
SCOPE
1.1 This standard guide applies to developing nuclear facility characterization plans to define the type, magnitude, location, and extent of radiological and chemical contamination within the facility to allow decommissioning planning. This guide amplifies guidance regarding facility characterization indicated in ASTM Standard E1281 on Nuclear Facility Decommissioning Plans. This guide does not address the methodology necessary to release a facility or site for unconditional use. This guide specifically addresses:  
1.1.1 the data quality objective for characterization as an initial step in decommissioning planning.  
1.1.2 sampling methods,  
1.1.3 the logic involved (statistical design) to ensure adequate characterization for decommissioning purposes; and  
1.1.4 essential documentation of the characterization information.  
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, health, and environmental practices and determine the applicability of regulatory limitations prior to use.  
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ASTM E1281-15(2021)(Guide)
Standard Guide for Nuclear Facility Decommissioning Plans

SIGNIFICANCE AND USE
4.1 The standardization of decommissioning plans will provide the nuclear facility owner with a greater assurance that all basic planning elements and requirements have been identified, examined, and addressed.  
4.2 In applying the guidance contained in this standard, the nuclear facility owner will address the significant subject areas necessary to describe a comprehensive decommissioning plan. Additional guidance on the planning of decommissioning projects, and the preparation of decommissioning plans can be found in such references as NUREG-1757 on decommissioning standard review plans, and Regulatory Guide 1.179 on the format and content of license termination plans. Recent new guidance on all aspects of decommissioning is contained in an ASME publication titled The Decommissioning Handbook.6  
4.3 This decommissioning plan will be developed to serve as the executive document that describes the objectives of the decommissioning program and identifies and defines the elements necessary to accomplish the program.  
4.4 A detailed implementation plan describing how the objectives of the decommissioning plan will be met should be prepared. Some of the documents or implementation plans that may be required to support the overall decommissioning program include an engineering plan; a cost, schedule, and financing plan (10 CFR 140 and 170); a field implementation plan; a health and safety plan (29 CFR 1910.120, Guide E1167); a quality assurance plan (10 CFR 50.59 and 10 CFR 830.120); an emergency plan; an environmental monitoring plan (Guide E1819); a radiological protection plan (10 CFR 20, 10 CFR 835, Guide E1167); and a physical security plan (10 CFR 73). These implementation plans shall be separate from and consistent with the decommissioning plan.
SCOPE
1.1 This guide applies to decommissioning plans for any nuclear facility whose operation was (is) governed by Nuclear Regulatory Commission (NRC), Agreement State license, under Department of Energy (DOE) orders, or whose operation was overseen by another federal, state, or local agency.  
1.2 The guide applies to the preparation and content of the decommissioning plan document itself.  
1.3 The detailed description and development of implementation plans identified in Section 4 is outside the scope of this guide.  
Note 1: Nuclear facilities operated by the U.S. DOE are not licensed by the U.S. NRC, nor are other nuclear facilities which may come under the control of the U.S. Department of Defense or individual agreement states. The references in this guide to licensee, 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 should not alter the content of decommissioning plans for nuclear facilities.  
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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