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ASTM D7727-11

(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
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.
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.
It is important to note that only fission gases are included in this calculation, and only the ones noted in 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.
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.
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, 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 given in parentheses are mathematical conversions to SI units, which are provided for information only and are not considered standard.
1.3 The values stated in inch-pound units are to be regarded as standard. The values given in parentheses are mathematical conversions to SI units that are provided for information only and are not considered standard.

General Information

Status
Historical
Publication Date
14-May-2011
ICS
27.120.10 - Reactor engineering
Technical Committee
D19 - Water
Drafting Committee
D19.04 - Methods of Radiochemical Analysis
Current Stage
Ref Project

Relations

Replaced By
ASTM D7727-11e1 - Standard Practice for Calculation of Dose Equivalent Xenon (DEX) for Radioactive Xenon Fission Products in Reactor Coolant
Effective Date
15-May-2011

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ASTM D7727-11 - Standard Practice for Calculation of Dose Equivalent Xenon (DEX) for Radioactive Xenon Fission Products in Reactor CoolantASTM D7727-11 - Standard Practice for Calculation of Dose Equivalent Xenon (DEX) for Radioactive Xenon Fission Products in Reactor Coolant
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ASTM D7727-11 - Standard Practice for Calculation of Dose Equivalent Xenon (DEX) for Radioactive Xenon Fission Products in Reactor Coolant
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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: D7727 − 11
StandardPractice for
Calculation of Dose Equivalent Xenon (DEX) for Radioactive
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 a factor that normalizes its dose to that of Xe. This practice
is to replace the previous practice of calculating the reactor
1.1 This practice applies to the calculation of the dose
133 coolant Ē calculation when allowed by the plant’s revised
equivalent to Xe in the reactor coolant of nuclear power
technical specifications. The quantity DEX is acceptable from
reactors resulting from the radioactivity of all noble gas fission
a radiological dose perspective since it will result in a limiting
products.
condition of operation (LCO) that more closely relates the
1.2 The values given in parentheses are mathematical con-
non-iodine RCS activity limits to the dose consequence analy-
versions to SI units, which are provided for information only
ses which form their bases.
and are not considered standard.
NOTE 2—It is incumbent on the licensee to ensure that the dose
1.3 Thevaluesstatedininch-poundunitsaretoberegarded
conversion factors (DCFs) used in the determination of DEX are consis-
as standard. The values given in parentheses are mathematical
tent with the DCFs used in the applicable dose consequence analysis used
by the plant in their dose calculation manual for radioactive releases.
conversions to SI units that are provided for information only
and are not considered standard.
5. Significance and Use
2. Referenced Documents
5.1 Each power reactor has a specific DEX value that is
2.1 ASTM Standards:
their technical requirement limit. These values may vary from
D3648 Practices for the Measurement of Radioactivity
about 200 to about 900 µCi/g based upon the height of their
D7282 Practice for Set-up, Calibration, and Quality Control
plant vent the location of the site boundary, the calculated
of Instruments Used for Radioactivity Measurements
reactor coolant activity for a condition of 1% fuel defects, and
general atmospheric modeling that is ascribed to that particular
3. Terminology
plant site. Should the DEX measured activity exceed the
3.1 DOSE-EQUIVALENT XE-133 (DEX),n—shall be that technical requirement limit the plant enters an LCO requiring
action on plant operation by the operators.
Xe concentration (microcuries per gram) that alone would
produce the same acute dose to the whole body as the
5.2 The determination of DEX is performed in a similar
85 85 87
combined activities of noble-gas nuclides mKr, Kr, Kr,
manner to that used in determining DEI, except that the
88 131m 133m 133 135m
Kr, Xe, Xe, Xe, Xe,
calculation of DEX is based on the acute dose to the whole
135 138
Xe, and Xe actually present. 85 85 87 88
body and considers the noble gases mKr, Kr, Kr, Kr,
131m 133m 133 135m 135 138
Xe, Xe, Xe, Xe, Xe, and Xe which are
NOTE 1—This is the general definition of DEX. Each utility may have
adopted modifications to this definition through agreement with the US
significant in terms of contribution to whole body dose.
Nuclear Regulatory Commission (USNRC). The definition as approved
5.3 It is important to note that only fission gases are
for each utility by the USNRC, is the one that should be applied to the
calculations in this method. included in this calculation, and only the ones noted in . 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
4. Summary of Practice
easily determined by gamma spectrometry (low energy x-rays
4.1 A sample of fresh reactor coolant is analyzed for noble
at 32 and 9 keV) and its dose consequence is vanishingly small
gas activities using gamma ray spectrometry. The individual
compared to the other, more prevalent krypton radionuclides.
activity of each detectable radioactive fission gas is divided by
41 19 16 11
5.4 Activity from Ar, F, N, and C, all of which
predominantly will be in gaseous forms in the RCS, are not
1 included in this calculation.
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
5.5 If a specific noble-gas radionuclide is not detected, it
Analysis.
should be assumed to be present at the minimum-detectable
Current edition approved May 15, 2011. Published June 2011. DOI: 10.1520/
D7727–11. activity. The determination of DOSE-EQUIVALENT XE-133
Copyright © ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959. United States
D7727 − 11
shall be performed using effective dose-conversion factors for 9.2 An appropriate aliquant of the sample is counted as a
air submersion listed in Table III.1 of EPA Federal Guidance pressurized liquid or degasified and the removed gas counted
Report No. 12, or the average gamma-disintegration energies
on a gamma ray spectrometer immediately after degasification
as provided in ICRP Publication 38 (''Radionuclide Transfor-
occurs.
mations’’) or similar source.
9.3 If a separated gas sample is counted, the method used
6. Interferences
should ensure that noble gas radionuclides are no retained by
the liquid phase. If they are, then the concentration from the
6.1 The analytical determination of the radionuclides used
liquid phase should be included in the calculation in 10.1.
for this calculation is made by gamma ray
...

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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.  
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SCOPE
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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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Irradiation Techniques (including Helium Injection)  
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SCOPE
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SCOPE
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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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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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