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ASTM E3104-17

(Specification)

Standard Specification for Strippable & Removable Coatings to Mitigate Spread of Radioactive Contamination

Standard Specification for Strippable & Removable Coatings to Mitigate Spread of Radioactive Contamination

ABSTRACT
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.  
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.  
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.

General Information

Status
Historical
Publication Date
30-Jun-2017
ICS
27.120.20 - Nuclear power plants. Safety
Technical Committee
E10 - Nuclear Technology and Applications
Drafting Committee
E10.03 - Radiological Protection for Decontamination and Decommissioning of Nuclear Facilities and Components
Current Stage
Ref Project

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ASTM E3104-17 - Standard Specification for Strippable & Removable Coatings to Mitigate Spread of Radioactive ContaminationASTM E3104-17 - Standard Specification for Strippable & Removable Coatings to Mitigate Spread of Radioactive Contamination
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ASTM E3104-17 - Standard Specification for Strippable & Removable Coatings to Mitigate Spread of Radioactive Contamination
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NOTICE: This standard has either been superseded and replaced by a new version or withdrawn.
Contact ASTM International (www.astm.org) for the latest information
Designation: E3104 − 17
Standard Specification for
Strippable & Removable Coatings to Mitigate Spread of
1
Radioactive Contamination
This standard is issued under the fixed designation E3104; 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 (´) indicates an editorial change since the last revision or reapproval.
1. Scope Radioactive Contamination after a Radiological Disper-
sion Event
1.1 This specification is intended to provide a basis for
identification of strippable/removable materials used to immo-
3. Terminology
bilize radioactive contamination, minimize worker exposure,
3.1 Definitions:
and facilitate subsequent decontamination or to protect uncon-
3.1.1 contamination, n—radioactive material in an un-
taminated areas against the spread of radioactive contamina-
wanted location.
tion.
3.1.2 decontamination, n—removal of contamination from
1.2 The values stated in inch-pound units are to be regarded
an unwanted location.
as standard. The values given in parentheses are mathematical
3.1.3 decontamination system, n—one or more products or
conversions to SI units that are provided for information only
procedures, or both, that, when used together, form a stabiliz-
and are not considered standard.
ing material to hold or bind radioactive contamination for
1.3 This standard does not purport to address all of the
subsequent removal from the substrate.
safety concerns, if any, associated with its use. It is the
3.1.4 environmental conditions, n—external factors that
responsibility of the user of this standard to establish appro-
may contribute to the performance of the coating including, but
priate safety and health practices and determine the applica-
not limited to, temperature, humidity, and ventilation.
bility of regulatory limitations prior to use.
3.1.5 immobilize, v—to fix in place; to prevent movement or
1.4 This international standard was developed in accor-
re-aerosolization of particulates due to mechanical or environ-
dance with internationally recognized principles on standard-
mental forces such as by tracking, precipitation, or wind.
ization established in the Decision on Principles for the
(E2731–09)
Development of International Standards, Guides and Recom-
mendations issued by the World Trade Organization Technical
3.1.6 lower flammability limit (LFL), n—the lower end of
Barriers to Trade (TBT) Committee.
the concentration range over which a flammable mixture of gas
or vapor in air can be ignited at a given temperature and
2. Referenced Documents
pressure.
2
2.1 ASTM Standards:
3.1.7 removable coating, n—products that fall outside the
D412 Test Methods for Vulcanized Rubber and Thermoplas-
purview of strippable coatings, that is, products that are
tic Elastomers—Tension
temporary coatings and removed through methods other than
D4541 Test Method for Pull-Off Strength of Coatings Using
stripping, that are used to physically or chemically hold or bind
Portable Adhesion Testers
radioactive particulate that may also be capable of subsequent
D4060 Test Method for Abrasion Resistance of Organic
removal of radioactive contamination.
Coatings by the Taber Abraser
3.1.8 strippable coating, n—durable film-forming product
E2731 Specification for Materials to Mitigate the Spread of
used to physically or chemically hold or bind radioactive
particulate that may also be capable of subsequent removal of
radioactive contamination.
1
This specification is under the jurisdiction of ASTM Committee E10 on
Nuclear Technology and Applications and is the direct responsibility of Subcom- 3.1.9 waste acceptance criteria (WAC), n—the criteria that a
mittee E10.03 on Radiological Protection for Decontamination and Decommission-
material must meet for acceptance in a waste disposal site;
ing of Nuclear Facilities and Components.
these criteria may vary per disposal site.
Current edition approved July 1, 2017. Published July 2017. DOI: 10.1520/
E3104–17
3.1.10 working time, n—the time period between the open-
2
For referenced ASTM standards, visit the ASTM website, www.astm.org, or
ing of the material storage container or mixing of components
contact ASTM Customer Service at service@astm.org. For Annual Book of ASTM
until the prepared material can no longer be successfully
Standards volume information, refer to the standard’s Document Summary page on
the ASTM website. applied to a surface.
Copyright © ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959. United States
1

---------------------- Page: 1 ----------------------
E3104 − 17
4. Significance and Use 5.4 The strippable/removable coating should be applicable
to the desired surface without
...

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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.  
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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).  
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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.  
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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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ABSTRACT
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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1.1 This specification is intended to provide a basis for identification of non-removable permanent coatings and fixatives as a long-term measure used to immobilize radioactive contamination, minimize worker exposure, and to protect uncontaminated areas against the spread of radioactive contamination.  
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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
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Standard Specification for Strippable and Removable Coatings to Mitigate Spread of Radioactive Contamination

ABSTRACT
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.
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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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4.1 A waste management plan based on the contents of this guide will provide for the successful identification of potential waste streams anticipated from decommissioning activities, and provide a clear and concise methodology for the handling of identified waste from generation to final disposition.  
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SCOPE
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1.3 This guide provides a description of the key elements of waste management plans that if followed will successfully allow for the characterization, packaging, transportation, and off-site treatment or disposal, or both, of conventional, hazardous, and radioactive waste streams.  
1.4 This guide does not address the on-site treatment, long term storage, or on-site disposal of these potential waste streams.  
1.5 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.  
1.6 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 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.  
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Standard Guide for Nuclear Facility Decommissioning Plans

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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  
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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.  
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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.  
1.2 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.

  • Guide
    9 pages
    English language
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