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ASTM D7727-21

(Practice)

Standard Practice for Calculation of Dose Equivalent Xenon (DEX) for Radioactive Xenon Fission Products in Reactor Coolant

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.

General Information

Status
Published
Publication Date
14-Dec-2021
ICS
27.120.10 - Reactor engineering
Technical Committee
D19 - Water
Drafting Committee
D19.04 - Methods of Radiochemical Analysis
Current Stage
Ref Project

Relations

Refers
ASTM D7902-20 - Standard Terminology for Radiochemical Analyses
Effective Date
01-May-2020

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ASTM D7727-21 - Standard Practice for Calculation of Dose Equivalent Xenon (DEX) for Radioactive Xenon Fission Products in Reactor CoolantASTM D7727-21 - Standard Practice for Calculation of Dose Equivalent Xenon (DEX) for Radioactive Xenon Fission Products in Reactor Coolant
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REDLINE ASTM D7727-21 - Standard Practice for Calculation of Dose Equivalent Xenon (DEX) for Radioactive Xenon Fission Products in Reactor CoolantREDLINE ASTM D7727-21 - Standard Practice for Calculation of Dose Equivalent Xenon (DEX) for Radioactive Xenon Fission Products in Reactor Coolant
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Standards Content (Sample)

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.
Designation: D7727 − 21
Standard Practice for
Calculation of Dose Equivalent Xenon (DEX) for Radioactive
1
Xenon Fission Products in Reactor Coolant
This standard is issued under the fixed designation D7727; 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 the same acute dose to the whole body as the combined
85m 85 87 88 131m
activitiesofnoble-gasnuclides Kr, Kr, Kr, Kr, Xe,
1.1 This practice applies to the calculation of the dose
133m 133 135m 135 138
133 Xe, Xe, Xe, Xe, and Xe actually present.
equivalent to Xe in the reactor coolant of nuclear power
reactors resulting from the radioactivity of all noble gas fission
3.1.1.1 Discussion—This is the general definition of DEX.
products.
Each utility may have adopted modifications to this definition
through agreement with the U.S. Nuclear Regulatory Commis-
1.2 The values stated in inch-pound units are to be regarded
sion (U.S. NRC).The definition as approved for each utility by
as standard. The values given in parentheses are mathematical
the U.S. NRC is the one that should be applied to the
conversions to SI units that are provided for information only
calculations in this practice.
and are not considered standard.
1.3 This standard does not purport to address all of the
4. Summary of Practice
safety concerns, if any, associated with its use. It is the
4.1 A sample of fresh reactor coolant is analyzed for noble
responsibility of the user of this standard to establish appro-
gas activities using gamma-ray spectrometry. The individual
priate safety, health, and environmental practices and deter-
activity of each detectable radioactive fission gas is divided by
mine the applicability of regulatory limitations prior to use.
133
a factor that normalizes its dose to that of Xe. This practice
1.4 This international standard was developed in accor-
replaces the previous practice of calculating the reactor coolant
dance with internationally recognized principles on standard-
%
E calculation when allowed by the plant’s revised technical
ization established in the Decision on Principles for the
specifications. The quantity DEX is acceptable from a radio-
Development of International Standards, Guides and Recom-
logical dose perspective since it will result in a limiting
mendations issued by the World Trade Organization Technical
condition of operation (LCO) that more closely relates the
Barriers to Trade (TBT) Committee.
non-iodine RCS activity limits to the dose consequence analy-
ses which form their bases.
2. Referenced Documents
NOTE 1—It is incumbent on the licensee to ensure that the dose
2
2.1 ASTM Standards:
conversion factors (DCFs) used in the determination of DEX are consis-
D3648 Practices for the Measurement of Radioactivity tent with the DCFs used in the applicable dose consequence analysis used
by the plant in their dose calculation manual for radioactive releases.
D7282 Practice for Set-up, Calibration, and Quality Control
of Instruments Used for Radioactivity Measurements
5. Significance and Use
D7902 Terminology for Radiochemical Analyses
5.1 Each power reactor has a specific DEX value that is
their technical requirement limit. These values may vary from
3. Terminology
about 200 to about 900 µCi/g based upon the height of their
3.1 Definitions:
plant vent, the location of the site boundary, the calculated
133
3.1.1 dose-equivalent Xe-133 (DEX), n—shall be that Xe
reactor coolant activity for a condition of 1 % fuel defects, and
concentration(microcuriespergram)thatalonewouldproduce
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
1
This practice is under the jurisdiction ofASTM Committee D19 on Water and
is the direct responsibility of Subcommittee D19.04 on Methods of Radiochemical action on plant operation by the operators.
Analysis.
5.2 The determination of DEX is performed in a similar
Current edition approved Dec. 15, 2021. Published January 2022. Originally
manner to that used in determining DEI, except that the
approved in 2011. Last previous edition approved in 2016 as D7727 – 11 (2016).
DOI: 10.1520/D7727-21.
calculation of DEX is based on the acute dose to the whole
2
For referenced ASTM standards, visit the ASTM website, www.astm.org, or 85m 85 87 88
body and considers the noble gases Kr, Kr, Kr, Kr,
contact ASTM Customer Service at service@astm.
...

This document is not an ASTM standard and is intended only to provide the user of an ASTM standard an indication of what changes have been made to the previous version. Because
it may not be technically possible to adequately depict all changes accurately, ASTM recommends that users consult prior editions as appropriate. In all cases only the current version
of the standard as published by ASTM is to be considered the official document.
Designation: D7727 − 11 (Reapproved 2016) D7727 − 21
Standard Practice for
Calculation of Dose Equivalent Xenon (DEX) for Radioactive
1
Xenon Fission Products in Reactor Coolant
This standard is issued under the fixed designation D7727; 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
133
1.1 This practice applies to the calculation of the dose equivalent to Xe 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.
2. Referenced Documents
2
2.1 ASTM Standards:
D3648 Practices for the Measurement of Radioactivity
D7282 Practice for Set-up, Calibration, and Quality Control of Instruments Used for Radioactivity Measurements
D7902 Terminology for Radiochemical Analyses
3. Terminology
3.1 Definitions:
133
3.1.1 dose-equivalent Xe-133 (DEX), n—shall be that Xe concentration (microcuries per gram) that alone would produce the
85m 85 87 88 131m 133m 133
same acute dose to the whole body as the combined activities of noble-gas nuclides Kr, Kr, Kr, Kr, Xe, Xe, Xe,
135m 135 138
Xe, Xe, and Xe actually present.
3.1.1.1 Discussion—
This is the general definition of DEX. Each utility may have adopted modifications to this definition through agreement with the
U.S. Nuclear Regulatory Commission (U.S. NRC). The definition as approved for each utility by the U.S. NRC is the one that
should be applied to the calculations in this practice.
1
This practice is under the jurisdiction of ASTM Committee D19 on Water and is the direct responsibility of Subcommittee D19.04 on Methods of Radiochemical Analysis.
Current edition approved Nov. 1, 2016Dec. 15, 2021. Published November 2016January 2022. Originally approved in 2011. Last previous edition approved in 20112016
ɛ1
as D7727 – 11D7727 . DOI: 10.1520/D7727-11R16. – 11 (2016). DOI: 10.1520/D7727-21.
2
For referenced ASTM standards, visit the ASTM website, www.astm.org, or contact ASTM Customer Service at service@astm.org. For Annual Book of ASTM Standards
volume information, refer to the standard’s Document Summary page on the ASTM website.
Copyright © ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959. United States
1

---------------------- Page: 1 ----------------------
D7727 − 21
4. Summary of Practice
4.1 A sample of fresh reactor coolant is analyzed for noble gas activities using gamma ray gamma-ray spectrometry. The
133
individual activity of each detectable radioactive fission gas is divided by a factor that normalizes its dose to that of Xe. This
%
practice is to replace replaces the previous practice of calculating the reactor coolant EĒ calculation when allowed by the plant’s
revised technical specifications. The quantity DEX is acceptable from a radiological dose perspective since it will result in a
limiting condition of operation (LCO) that more closely relates the non-iodine RCS activity limits to the dose consequence
analyses which form their bases.
NOTE 1—It is incumbent on the licensee to ensure that the dose conversion factors (DCFs) used in the determination of DEX are consistent with the DCFs
used in the applicable dose consequence analysis used by the plant in their dose calculation manual for radioactive releases.
5. 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
...

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SCOPE
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1.3.1.2 The guidance set forth results from discoveries of conditions, practices, features, or lack of features that were found to be sources of operational or maintenance problems, or causes of failure.  
1.3.2 This standard does not supersede federal or state regulations, or both, and 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 E185-21(Practice)
Standard Practice for Design of Surveillance Programs for Light-Water Moderated Nuclear Power Reactor Vessels

SIGNIFICANCE AND USE
4.1 Predictions of neutron radiation effects on pressure vessel steels are considered in the design of light-water moderated nuclear power reactors. Changes in system operating parameters often are made throughout the service life of the reactor vessel to account for radiation effects. Due to the variability in the behavior of reactor vessel steels, a surveillance program is warranted to monitor changes in the properties of actual vessel materials caused by long-term exposure to the neutron radiation and temperature environment of the reactor vessel. This practice describes the criteria that should be considered in planning and implementing surveillance test programs and points out precautions that should be taken to ensure that: (1) capsule exposures can be related to beltline exposures, (2) materials selected for the surveillance program are samples of those materials most likely to limit the operation of the reactor vessel, and (3) the test specimen types are appropriate for the evaluation of radiation effects on the reactor vessel.  
4.2 Guides E482 and E853 describe a methodology for estimation of neutron exposure obtained for reactor vessel surveillance programs. Regulators or other sources may describe different methods.  
4.3 The design of a surveillance program for a given reactor vessel must consider the existing body of data on similar materials in addition to the specific materials used for that reactor vessel. The amount of such data and the similarity of exposure conditions and material characteristics will determine their applicability for predicting radiation effects.
SCOPE
1.1 This practice covers procedures for designing a surveillance program for monitoring the radiation-induced changes in the mechanical properties of ferritic materials in light-water moderated nuclear power reactor vessels. New advanced light-water small modular reactor designs with a nominal design output of 300 MWe or less have not been specifically considered in this practice. This practice includes the minimum requirements for the design of a surveillance program, selection of vessel material to be included, and the initial schedule for evaluation of materials.  
1.2 This practice was developed for all light-water moderated nuclear power reactor vessels for which the predicted maximum fast neutron fluence (E > 1 MeV) exceeds 1 × 1021 neutrons/m 2  (1 × 1017 n/cm2) at the inside surface of the ferritic steel reactor vessel.  
1.3 This practice does not provide specific procedures for monitoring the radiation induced changes in properties beyond the design life. Practice E2215 addresses changes to the withdrawal schedule during and beyond the design life.  
1.4 The values stated in SI units are to be regarded as the standard. The values given in parentheses are for information only.
Note 1: The increased complexity of the requirements for a light-water moderated nuclear power reactor vessel surveillance program has necessitated the separation of the requirements into three related standards. Practice E185 describes the minimum requirements for design of a surveillance program. Practice E2215 describes the procedures for testing and evaluation of surveillance capsules removed from a reactor vessel. Guide E636 provides guidance for conducting additional mechanical tests. A summary of the many major revisions to Practice E185 since its original issuance is contained in Appendix X2.
Note 2: This practice applies only to the planning and design of surveillance programs for reactor vessels designed and built after the effective date of this practice. Previous versions of Practice E185 apply to earlier reactor vessels. See Appendix X2.  
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 Or...

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ASTM
ASTM E900-21(Guide)
Standard Guide for Predicting Radiation-Induced Transition Temperature Shift in Reactor Vessel Materials

SIGNIFICANCE AND USE
4.1 Operation of commercial power reactors must conform to pressure-temperature limits during heatup and cooldown to prevent over-pressurization at temperatures that might cause non-ductile behavior in the presence of a flaw. Radiation damage to the reactor vessel is compensated for by adjusting the pressure-temperature limits to higher temperatures as the neutron damage accumulates. The present practice is to base that adjustment on the TTS  produced by neutron irradiation as measured at the Charpy V-notch 41-J (30-ft·lbf) energy level. To establish pressure temperature operating limits during the operating life of the plant, a prediction of TTS  must be made.  
4.1.1 In the absence of surveillance data for a given reactor material (see Practice E185 and E2215), the use of calculative procedures are necessary to make the prediction. Even when credible surveillance data are available, it will usually be necessary to interpolate or extrapolate the data to obtain a TTS  for a specific time in the plant operating life. The embrittlement correlation presented herein has been developed for those purposes.  
4.2 Research has established that certain elements, notably copper (Cu), nickel (Ni), phosphorus (P), and manganese (Mn), cause a variation in radiation sensitivity of reactor pressure vessel steels. The importance of other elements, such as silicon (Si), and carbon (C), remains a subject of additional research. Copper, nickel, phosphorus, and manganese are the key chemistry parameters used in developing the calculative procedures described here.  
4.3 Only power reactor (PWR and BWR) surveillance data were used in the derivation of these procedures. The measure of fast neutron fluence used in the procedure is n/m2  (E > 1 MeV). Differences in fluence rate and neutron energy spectra experienced in power reactors and test reactors have not been accounted for in these procedures.
SCOPE
1.1 This guide presents a method for predicting values of reference transition temperature shift (TTS) for irradiated pressure vessel materials. The method is based on the TTS exhibited by Charpy V-notch data at 41-J (30-ft·lbf) obtained from surveillance programs conducted in several countries for commercial pressurized (PWR) and boiling (BWR) light-water cooled (LWR) power reactors. An embrittlement correlation has been developed from a statistical analysis of the large surveillance database consisting of radiation-induced TTS and related information compiled and analyzed by Subcommittee E10.02. The details of the database and analysis are described in a separate report (ADJE090015-EA).2,3 This embrittlement correlation was developed using the variables copper, nickel, phosphorus, manganese, irradiation temperature, neutron fluence, and product form. Data ranges and conditions for these variables are listed in 1.1.1. Section 1.1.2 lists the materials included in the database and the domains of exposure variables that may influence TTS but are not used in the embrittlement correlation.  
1.1.1 The range of material and irradiation conditions in the database for variables used in the embrittlement correlation:  
1.1.1.1 Copper content up to 0.4 %.
1.1.1.2 Nickel content up to 1.7 %.
1.1.1.3 Phosphorus content up to 0.03 %.
1.1.1.4 Manganese content within the range from 0.55 to 2.0 %.
1.1.1.5 Irradiation temperature within the range from 255 to 300°C (491 to 572°F).
1.1.1.6 Neutron fluence within the range from 1 × 1021 n/m2 to 2 × 1024 n/m2 (E> 1 MeV).
1.1.1.7 A categorical variable describing the product form (that is, weld, plate, forging).  
1.1.2 The range of material and irradiation conditions in the database for variables not included in the embrittlement correlation:  
1.1.2.1 A533 Type B Class 1 and 2, A302 Grade B, A302 Grade B (modified), and A508 Class 2 and 3. Also, European and Japanese steel grades that are equivalent to these ASTM Grades.
1.1.2.2 Submerged arc welds, shielded a...

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ASTM
ASTM D8377-21a(Guide)
Standard Guide for High Temperature Strength Measurements of Graphite Impregnated with Molten Salt

SIGNIFICANCE AND USE
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.  
5.3 The purpose of this guide is to provide considerations, which should be included in testing graphite specimens impregnated with molten salt at elevated temperature.  
5.4 For the test results to be meaningful, the...
SCOPE
1.1 This guide covers the best practice for strength measurements at elevated temperature of graphite impregnated with molten salt.  
1.2 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this 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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