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ASTM C781-96

(Practice)

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

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

SCOPE
1.1 This practice covers the test methods for measuring those properties of graphite and boronated graphite materials that may be used for the design and evaluation of high-temperature gas-cooled reactors.
1.2 The test methods referenced herein are applicable to materials used for replaceable and permanent components as defined in Section 7 and include fuel and removable reflector elements; target elements and insulators; permanent side reflector elements; core support pedestals and elements; control rod, reserve shutdown, and burnable poison compacts, and neutron shield material.
1.3 This practice includes test methods that have been selected from existing ASTM standards, ASTM standards that have been modified, and new ASTM standards that are specific to the testing of materials listed in 1.2. Comments on individual test methods for graphite and for boronated graphite components are given in Sections 8 and 10, respectively. The test methods are summarized in Tables 1 and 2.
1.4 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

General Information

Status
Historical
Publication Date
31-Dec-1995
ICS
27.120.10 - Reactor engineering
Technical Committee
D02 - Petroleum Products, Liquid Fuels, and Lubricants
Drafting Committee
D02.F0 - Manufactured Carbon and Graphite Products
Current Stage
Ref Project

Relations

Replaced By
ASTM C781-02 - Standard Practice for Testing Graphite and Boronated Graphite Components for High-Temperature Gas-Cooled Nuclear Reactors
Effective Date
10-Dec-2002
Referred By
ASTM D7219-19 - Standard Specification for Isotropic and Near-isotropic Nuclear Graphites
Effective Date
01-Jan-1996
Referred By
ASTM D7301-21 - Standard Specification for Nuclear Graphite Suitable for Components Subjected to Low Neutron Irradiation Dose
Effective Date
01-Jan-1996
Referred By
ASTM C714-17 - Standard Test Method for Thermal Diffusivity of Carbon and Graphite by Thermal Pulse Method
Effective Date
01-Jan-1996
Referred By
ASTM D7846-21 - Standard Practice for Reporting Uniaxial Strength Data and Estimating Weibull Distribution Parameters for Advanced Graphites
Effective Date
01-Jan-1996

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ASTM C781-96 - Standard Practice for Testing Graphite and Boronated Graphite Components for High-Temperature Gas-Cooled Nuclear ReactorsASTM C781-96 - Standard Practice for Testing Graphite and Boronated Graphite Components for High-Temperature Gas-Cooled Nuclear Reactors
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ASTM C781-96 - Standard Practice for Testing Graphite and Boronated Graphite Components for High-Temperature Gas-Cooled Nuclear Reactors
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Standards Content (Sample)

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.
An American National Standard
Designation: C 781 – 96
Standard Practice for
Testing Graphite and Boronated Graphite Components for
1
High-Temperature Gas-Cooled Nuclear Reactors
This standard is issued under the fixed designation C 781; 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.
4
1. Scope sorption Cross Section of Nuclear Graphite
C 651 Test Method for Flexural Strength of Manufactured
1.1 This practice covers the test methods for measuring
Carbon and GraphiteArticles Using Four-Point Loading at
those properties of graphite and boronated graphite materials
2
Room Temperature
that may be used for the design and evaluation of high-
C 695 Test Method for Compressive Strength of Carbon
temperature gas-cooled reactors.
2
and Graphite
1.2 The test methods referenced herein are applicable to
C 709 Terminology Relating to Manufactured Carbon and
materials used for replaceable and permanent components as
2
Graphite
defined in Section 7 and include fuel and removable reflector
C 747 Test Method for Moduli of Elasticity and Fundamen-
elements; target elements and insulators; permanent side re-
tal Frequencies of Carbon and Graphite Materials by Sonic
flector elements; core support pedestals and elements; control
2
Resonance
rod, reserve shutdown, and burnable poison compacts; and
C 749 Test Method for Tensile Stress-Strain of Carbon and
neutron shield material.
2
Graphite
1.3 This practice includes test methods that have been
C 816 Test Method for Sulfur in Graphite by Combustion-
selected from existing ASTM standards, ASTM standards that
2
Iodometric Titration Method
have been modified, and newASTM standards that are specific
C 838 Test Method for Bulk Density of As-Manufactured
tothetestingofmaterialslistedin1.2.Commentsonindividual
2
Carbon and Graphite Shapes
test methods for graphite and boronated graphite components
C 1179 Test Method for Oxidative Mass Loss of Manufac-
are given in Sections 8 and 10, respectively. The test methods
2
turered Carbon and Graphite Materials in Air
are summarized in Table 1 and Table 2.
C 1251 Test Method for Determination of Specific Surface
1.4 This standard does not purport to address all of the
2
Areas ofAdvanced Ceramic Materials by GasAdsorption
safety concerns, if any, associated with its use. It is the
D 2854 Test Method for Apparent Density of Activated
responsibility of the user of this standard to establish appro-
2
Carbon
priate safety and health practices and determine the applica-
D 2862 Test Method for Particle Size Distribution of
bility of regulatory limitations prior to use.
2
Granular Activated Carbon
2. Referenced Documents
D 4292 Test Method for Vibrated Bulk Density of Calcined
5
Petroleum Coke
2.1 ASTM Standards:
E 132 Test Method for Poisson’s Ratio at Room Tempera-
C 559 Test Method for Bulk Density by Physical Measure-
6
2
ture
ments of Manufactured Carbon and Graphite Articles
2
E 228 Test Method for Linear Thermal Expansion of Solid
C 560 Test Methods for Chemical Analysis of Graphite
7
2
Materials with a Vitreous Silica Dilatometer
C 561 Test Method for Ash in a Graphite Sample
3
E 261 Practice for Determining Neutron Fluence Rate, Flu-
C 577 Test Method for Permeability of Refractories
8
ence, and Spectra by Radioactivation Techniques
C 625 Practice for Reporting Irradiation Results on Graph-
2
E 1461 Test Method for Thermal Diffusivity of Solids by
ite
7
the Flash Method
C 626 Methods for Estimating the Thermal Neutron Ab-
3. Terminology
3.1 Definition:
1
This practice is under the jurisdiction of ASTM Committee D02 on Petroleum
Products and Lubricants and is the direct responsibility of Subcommittee D02.F on
4
Manufactured Carbon and Graphite Products . Discontinued; see 1985 Annual Book of ASTM Standards, Vol 15.01.
5
Current edition approved Nov. 10, 1996. Published December 1996. Originally Annual Book of ASTM Standards, Vol 05.02.
6
published as C 781 – 77. Last previous edition C 781 – 93. Annual Book of ASTM Standards, Vol 03.01.
2 7
Annual Book of ASTM Standards, Vol 05.05. Annual Book of ASTM Standards, Vol 14.02.
3 8
Annual Book of ASTM Standards, Vol 15.01. Annual Book of ASTM Standards, Vol 12.02.
Copyright © ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959, United States.
1

---------------------- Page: 1 ----------------------
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.
C 781
TABLE 1 Summary of Test Methods for Graphite Components
NOTE 1—Designations under preparation
...

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5.1 The Molten Salt Reactor is a nuclear reactor which uses graphite as reflector and structural material, and molten salt as coolant. The graphite components will be submerged in the molten salt during the lifetime of the reactor. The porous structure of graphite may lead to molten salt permeation, which can affect the thermal and mechanical properties of graphite. Consequently, it may be necessary to measure the various strengths of the manufactured graphite materials after impregnation with molten salt and before exposure to the reactor environment in a range of test configurations in order for designers or operators to assess their performance.
Note 1: Depending upon the salt selected for the reactor, there may be some chemical reaction between the salt and the graphite that could affect properties. The user should establish, prior to following this guide, that any interactions between the molten salt and graphite are understood and any implications for the validity of the strength tests have been assessed.  
5.2 For gas-cooled reactors, the strength of a graphite specimen is usually measured at room temperature. However, for molten salt reactors, the operating temperature of the reactor must be higher than the melting temperature of the salt, and so the salt will be in solid state at room temperature. Consequently, room temperature measurements may not be representative of the performance of the material at its true operating conditions. It is therefore necessary to measure the strength at an elevated temperature where the salt is in liquid form.
Note 2: Users should be aware that a small increase in graphite strength is expected with increasing temperature. Testing at the plant operating temperature will eliminate this small uncertainty.  
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SCOPE
1.1 This guide covers the best practice for strength measurements at elevated temperature of graphite impregnated with molten salt.  
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ASTM C781-20(Practice)
Standard Practice for Testing Graphite Materials for Gas-Cooled Nuclear Reactor Components

SIGNIFICANCE AND USE
4.1 Property data obtained with the recommended test methods identified herein may be used for research and development, design, manufacturing control, specifications, performance evaluation, and regulatory statutes pertaining to nuclear reactors that utilize graphite.  
4.2 The referenced test methods are applicable primarily to specimens in the non-irradiated and non-oxidized state. Testing irradiated specimens often requires specimen geometries that do not meet the requirements of the standard. Specific instructions or recommendations with respect to testing non-conforming geometries can be found in STP 15784 and/or Guide D7775. When testing irradiated specimens at elevated temperatures, the effects of annealing should be considered (see Note 1).
Note 1: Exposure to fast neutron radiation will result in atomic and microstructural changes to graphite. This radiation damage occurs when energetic particles, such as fast neutrons, impinge on the crystal lattice and displace carbon atoms from their equilibrium positions, creating a lattice vacancy and an interstitial carbon atom. The lattice strain that results from displacement damage causes significant structural and property changes in the graphite and is a function of the irradiation temperature and dose. When the temperature of the graphite is brought above the temperature at which it was irradiated, enough energy is provided that the structure of the graphite will anneal back to its original condition. Therefore, measurement techniques that bring the specimen temperature above the irradiation temperature can result in property values that change during the measurement process. For this reason, measurements made on irradiated test specimens below the irradiation temperature will produce results that are representative of the irradiation damage. However, measurements made at temperatures above the irradiation temperature could include the effects of annealing.  
4.3 Additional test methods are in preparation a...
SCOPE
1.1 This practice covers the application and limitations of test methods for measuring the properties of graphite materials. These properties may be used for the design and evaluation of gas-cooled reactor components.  
1.2 The test methods referenced herein are applicable to materials used for replaceable and permanent components as defined in Section 7 and includes fuel elements; removable reflector elements and blocks; permanent side reflector elements and blocks; core support pedestals and elements; control rod, reserve shutdown, and burnable poison compacts; and neutron shield material. Specific aspects with respect to testing of irradiated materials are addressed.  
1.3 This practice includes test methods that have been selected from ASTM standards and guides that are specific to the testing of materials listed in 1.2. Comments on individual test methods for graphite components are given in Section 8. The test methods are summarized in Table 1.  
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Standard Guide for Monitoring the Neutron Exposure of LWR Reactor Pressure Vessels

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4.1 Regulatory Requirements—The USA Code of Federal Regulations (10CFR Part 50, Appendix H) requires the implementation of a reactor vessel materials surveillance program for all operating LWRs. Other countries have similar regulations. The purpose of the program is to (1) monitor changes in the fracture toughness properties of ferritic materials in the reactor vessel beltline6 resulting from exposure to neutron irradiation and the thermal environment, and (2) make use of the data obtained from surveillance programs to determine the conditions under which the vessel can be operated with adequate margins of safety throughout its service life. Practice E185, derived mechanical property data, and (r, θ, z) physics-dosimetry data (derived from the calculations and reactor cavity and surveillance capsule measurements (1) using physics-dosimetry standards) can be used together with information in Guide E900 and Refs. 4, 11-18 to provide a relation between property degradation and neutron exposure, commonly called a “trend curve.” To obtain this trend curve at all points in the pressure vessel wall requires that the selected trend curve be used together with the appropriate (r, θ, z) neutron field information derived by use of this guide to accomplish the necessary interpolations and extrapolations in space and time.  
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SCOPE
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Standard Practice for Investigating the Effects of Neutron Radiation Damage Using Charged-Particle Irradiation

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4.1 A characteristic advantage of charged-particle irradiation experiments is the precise, individual control over most of the important irradiation conditions such as dose, dose rate, temperature, and quantity of gases present. Additional attributes are the lack of induced radioactivation of specimens and, in general, a substantial compression of irradiation time, from years to hours, to achieve comparable damage as measured in displacements per atom (dpa). An important application of such experiments is the investigation of radiation effects that may occur in materials exposed to environments which do not currently exist, such as in first wall materials used in fusion reactors.  
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SIGNIFICANCE AND USE
5.1 To establish a proper calibration area for nuclear surface gauges.  
5.2 To reduce the chance of improper calibration.
Note 1: The quality of the results produced by this standard is dependent on the competence of the personnel performing it, and the suitability of the equipment and facilities used. Agencies that meet the criteria of practice D3740 are generally considered capable of competent and objective testing/inspection/etc. Users of this standard are cautioned that compliance with practice D3740 does not in itself assure a means of evaluating some of those factors.
SCOPE
1.1 This guide outlines procedures for setup of a nuclear gauge calibration facility in either a shielded bay or an unshielded area—Guide A and Guide B, respectively.  
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4.1 The purpose of this guide is to provide general guidelines for the design and operation of hot cell equipment to ensure longevity and reliability throughout the period of service.  
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4.3 This guide is intended as a supplement to other standards, and to federal and state regulations, codes, and criteria applicable to the design of equipment intended for hot cell use.  
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SCOPE
1.1 Intent:  
1.1.1 The intent of this guide is to provide general design and operating considerations for the safe and dependable operation of remotely operated hot cell equipment. Hot cell equipment is hardware used to handle, process, or analyze nuclear or radioactive material in a shielded room. The equipment is placed behind radiation shield walls and cannot be directly accessed by the operators or by maintenance personnel because of the radiation exposure hazards. Therefore, the equipment is operated remotely, either with or without the aid of viewing.  
1.1.2 This guide may apply to equipment in other radioactive remotely operated facilities such as suited entry repair areas, canyons or caves, but does not apply to equipment used in commercial power reactors.  
1.1.3 This guide does not apply to equipment used in gloveboxes.  
1.2 Applicability:  
1.2.1 This guide is intended for persons who are tasked with the planning, design, procurement, fabrication, installation, or testing of equipment used in remote hot cell environments.  
1.2.2 The equipment will generally be used over a long-term life cycle (for example, in excess of two years), but equipment intended for use over a shorter life cycle is not excluded.  
1.2.3 The system of units employed in this standard is the metric unit, also known as SI Units, which are commonly used for International Systems, and defined by IEEE/ASTM SI 10: American National Standard for Use of the International System of Units (SI): The Modern Metric System.  
1.3 Caveats:  
1.3.1 This guide does not address considerations relating to the design, construction, operation, or safety of hot cells, caves, canyons, or other similar remote facilities. This guide deals only with equipment intended for use in hot cells.  
1.3.2 Specific design and operating considerations are found in other ASTM documents.  
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ASTM C1615/C1615M-17(2022)(Guide)
Standard Guide for Mechanical Drive Systems for Remote Operation in Hot Cell Facilities

SIGNIFICANCE AND USE
4.1 Mechanical drive systems operability and long-term integrity are concerns that should be addressed primarily during the design phase; however, problems identified during fabrication and testing should be resolved and the changes in the design documented. Equipment operability and integrity can be compromised during handling and installation sequences. For this reason, the subject equipment should be handled and installed under closely controlled and supervised conditions.  
4.2 This standard is intended as a supplement to other standards, and to federal and state regulations, codes, and criteria applicable to the design of equipment intended for this use.  
4.3 This standard is intended to be generic and to apply to a wide range of types and configurations of mechanical drive systems.
SCOPE
1.1 Intent:  
1.1.1 The intent of this standard is to provide general guidelines for the design, selection, quality assurance, installation, operation, and maintenance of mechanical drive systems used in remote hot cell environments. The term mechanical drive systems used herein, encompasses all individual components used for imparting motion to equipment systems, subsystems, assemblies, and other components. It also includes complete positioning systems and individual units that provide motive power and any position indicators necessary to monitor the motion.  
1.2 Applicability:  
1.2.1 This standard is intended to be applicable to equipment used under one or more of the following conditions:  
1.2.1.1 The materials handled or processed constitute a significant radiation hazard to man or to the environment.
1.2.1.2 The equipment will generally be used over a long-term life cycle (for example, in excess of two years), but equipment intended for use over a shorter life cycle is not excluded.
1.2.1.3 The equipment can neither be accessed directly for purposes of operation or maintenance, nor can the equipment be viewed directly, for example, without radiation shielding windows, periscopes, or a video monitoring system (Guides C1572 and C1661).  
1.2.2 The values stated in either SI units or inch-pound units are to be regarded separately as standard. The values stated in each system may not be exact equivalents; therefore, each system shall be used independently of the other. Combining values from the two systems may result in non-conformance with the standard.  
1.3 User Caveats:  
1.3.1 This standard is not a substitute for applied engineering skills, proven practices and experience. Its purpose is to provide guidance.
1.3.1.1 The guidance set forth in this standard relating to design of equipment is intended only to alert designers and engineers to those features, conditions, and procedures that have been found necessary or highly desirable to the design, selection, operation and maintenance of mechanical drive systems for the subject service conditions.
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Standard Guide for Hot Cell Specialized Support Equipment and Tools

SIGNIFICANCE AND USE
4.1 This guide is relevant to the design of specialized support equipment and tools that are remotely operated, maintained, or viewed through shielding windows, or combinations thereof, or by other remote viewing systems.  
4.2 Hot cells contain substances and processes that may be extremely hazardous to personnel or the external environment, or both. Process safety and reliability are improved with successful design, installation, and operation of specialized mechanical and support equipment.  
4.3 Use of this guide in the design of specialized mechanical and support equipment can reduce costs, improve productivity, reduce failed hardware replacement time, and provide a standardized design approach.
SCOPE
1.1 Intent:  
1.1.1 This guide presents practices and guidelines for the design and implementation of equipment and tools to assist assembly, disassembly, alignment, fastening, maintenance, or general handling of equipment in a hot cell. Operating in a remote hot cell environment significantly increases the difficulty and time required to perform a task compared to completing a similar task directly by hand. Successful specialized support equipment and tools minimize the required effort, reduce risks, and increase operating efficiencies.  
1.2 Applicability:  
1.2.1 This guide may apply to the design of specialized support equipment and tools anywhere it is remotely operated, maintained, and viewed through shielding windows or by other remote viewing systems.  
1.2.2 Consideration should be given to the need for specialized support equipment and tools early in the design process.  
1.2.3 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.  
1.3 Caveats:  
1.3.1 This guide is generic in nature and addresses a wide range of remote working configurations. Other acceptable and proven international configurations exist and provide options for engineer and designer consideration. Specific designs are not a substitute for applied engineering skills, proven practices, or experience gained in any specific situation.  
1.3.2 This guide does not supersede federal or state regulations, or both, or codes applicable to equipment under any conditions.  
1.3.3 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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ASTM
ASTM E2006-22(Guide)
Standard Guide for Benchmark Testing of Light Water Reactor Calculations

SIGNIFICANCE AND USE
4.1 This guide deals with the difficult problem of benchmarking neutron transport calculations carried out to determine fluences for plant specific reactor geometries. The calculations are necessary for fluence determination in locations important for material radiation damage estimation and which are not accessible to measurement. Typically, the most important application of such calculations is the estimation of fluence within the reactor vessel of operating light water reactors (LWR) to provide accurate estimates of the irradiation embrittlement of the base and weld metal in the vessel. The benchmark procedure must not only prove that calculations give reasonable results but that their uncertainties are propagated with due regard to the sensitivities of the different input parameters used in the transport calculations. Benchmarking is achieved by building up data bases of benchmark experiments that have different influences on uncertainty propagation. For example, in simple vessel wall mockups where measurements are made within a simulated reactor vessel wall, the integral effect of uncertainties in iron cross sections (absorption and elastic and inelastic scattering) are dominant and have been bounded by the agreement between calculation and measurement. For more complicated integral benchmarks, other factors such as: uncertainties in the distribution of fission sources, geometry, the energy-dependent cross sections, and the angular scattering distribution for elemental components of major materials in the neutron field (such as water and iron) may all be important uncertainty contributors. This guide describes general procedures for using neutron fields with known characteristics to corroborate the calculational methodology and nuclear data used to derive neutron field information from measurements of neutron sensor response.  
4.2 The bases for benchmark field referencing are usually irradiations performed in standard neutron fields with well-known energy spe...
SCOPE
1.1 This guide covers general approaches for benchmarking neutron transport calculations for pressure vessel surveillance programs in light water reactor systems. A companion guide (Guide E2005) covers use of benchmark fields for testing neutron transport calculations and cross sections in well controlled environments. This guide covers experimental benchmarking of neutron fluence calculations (or calculations of other exposure parameters such as dpa) in more complex geometries relevant to reactor pressure vessel surveillance. Particular sections of the guide discuss: the use of well-characterized benchmark neutron fields to provide an indication of the accuracy of the calculational methods and nuclear data when applied to typical cases; and the use of plant specific measurements to indicate bias in individual plant calculations. Use of these two benchmark techniques will serve to limit plant-specific calculational uncertainty, and, when combined with analytical uncertainty estimates for the calculations, will provide uncertainty estimates for reactor fluences with a higher degree of confidence.  
1.2 Although this guide and the companion guide, Guide E2005, are focused on power reactors, the principle of this guide is also applicable to non-power light water reactor pressure vessel surveillance programs.  
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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ASTM
ASTM D7727-21(Practice)
Standard Practice for Calculation of Dose Equivalent Xenon (DEX) for Radioactive Xenon Fission Products in Reactor Coolant

SIGNIFICANCE AND USE
5.1 Each power reactor has a specific DEX value that is their technical requirement limit. These values may vary from about 200 to about 900 μCi/g based upon the height of their plant vent, the location of the site boundary, the calculated reactor coolant activity for a condition of 1 % fuel defects, and general atmospheric modeling that is ascribed to that particular plant site. Should the DEX measured activity exceed the technical requirement limit, the plant enters an LCO requiring action on plant operation by the operators.  
5.2 The determination of DEX is performed in a similar manner to that used in determining DEI, except that the calculation of DEX is based on the acute dose to the whole body and considers the noble gases 85mKr, 85Kr, 87Kr, 88Kr, 131mXe, 133mXe, 133Xe, 135mXe, 135Xe, and 138Xe which are significant in terms of contribution to whole body dose.  
5.3 It is important to note that only fission gases are included in this calculation, and only the ones noted in Table 1. For example 83mKr is not included even though its half-life is 1.86 hours. The reason for this is that this radionuclide cannot be easily determined by gamma spectrometry (low energy X-rays at 32 and 9 keV) and its dose consequence is vanishingly small compared to the other, more prevalent krypton radionuclides.  
5.4 Activity from 41Ar, 19F, 16N, and 11C, all of which predominantly will be in gaseous forms in the RCS, are not included in this calculation.  
5.5 If a specific noble-gas radionuclide is not detected, it should be assumed to be present at the minimum-detectable activity. The determination of dose-equivalent Xe-133 shall be performed using effective dose-conversion factors for air submersion listed in Table III.1 of EPA Federal Guidance Report No. 12,3 or the average gamma-disintegration energies as provided in ICRP Publication 38 (“Radionuclide Transformations”) or similar source.
SCOPE
1.1 This practice applies to the calculation of the dose equivalent to 133Xe in the reactor coolant of nuclear power reactors resulting from the radioactivity of all noble gas fission products.  
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, 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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