ASTM E1018-09(2013)e1
(Guide)Standard Guide for Application of ASTM Evaluated Cross Section Data File
Standard Guide for Application of ASTM Evaluated Cross Section Data File
SIGNIFICANCE AND USE
4.1 The ENDF/B library in the United States and similar libraries elsewhere, such as JEF (22), JENDL (19), and BROND (20), provide a compilation of neutron cross section and other nuclear data for use by the nuclear community. The availability of these excellent and consistent evaluations makes possible standardized usage, thereby allowing easy referencing and intercomparisons of calculations. However, as the first ENDF/B files were developed it became apparent that they were not adequate for all applications. This need resulted in the development of the ENDF/B Dosimetry File (17, 23), consisting of activation cross sections important for dosimetry applications. This file was made available worldwide. Later, other“ Special Purpose” files were introduced (24). In the ENDF/B-VI compilation (25), dosimetry files were identified, but they no longer appeared as separate evaluation files. The ENDF/V-VII compilation (26) removed most of the covariance files used by the dosimetry community. It kept the covariance files for the “standard cross sections” in a special sub-library, but the covariance data in this sub-library is only provided over the energy range in which each reaction is considered to be a “standard”, and does not include the full energy range required for LWR PVS dosimetry applications.
4.2 Another file of evaluated neutron cross section data has been established by the International Atomic Energy Agency (IAEA) for reactor dosimetry applications. This file, the International Reactor Dosimetry File (IRDF-2002) (18) , draws upon the ENDF/B files and supplements these evaluations with a set of reactions evaluated by groups often outside of the United States. Some of the IRDF-2002 supplemental reactions represent material evaluations that are currently being examined by the CSEWG. The supplemental IRDF-2002 evaluations only include the specific reactions of interest to the dosimetry community and not a full material evaluation. The ENDF community requires a co...
SCOPE
1.1 This guide covers the establishment and use of an ASTM evaluated nuclear data cross section and uncertainty file for analysis of single or multiple sensor measurements in neutron fields related to light water reactor LWR-Pressure Vessel Surveillance (PVS). These fields include in- and ex-vessel surveillance positions in operating power reactors, benchmark fields, and reactor test regions.
1.2 Requirements for establishment of ASTM-approved cross section files address data format, evaluation requirements, validation in benchmark fields, evaluation of error estimates (covariance file), and documentation. A further requirement for components of the ASTM-approved cross section file is their internal consistency when combined with sensor measurements and used to determine a neutron spectrum.
1.3 Specifications for use include energy region of applicability, data processing requirements, and application of uncertainties.
1.4 This guide is directly related to and should be used primarily in conjunction with Guides E482 and E944, and Practices E560, E185, and E693.
1.5 The ASTM cross section and uncertainty file represents a generally available data set for use in sensor set analysis. However, the availability of this data set does not preclude the use of other validated data, either proprietary or nonproprietary. When alternate cross section files are used that deviate from the requirements laid out in this standard, the deviations should be noted to the customer ofr the dosimetry application.
1.6 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.
1.7 This international standard was developed in accordance with internationally recognized principles on standardization established ...
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´1
Designation: E1018 − 09 (Reapproved 2013)
Standard Guide for
Application of ASTM Evaluated Cross Section Data File
This standard is issued under the fixed designation E1018; the number immediately following the designation indicates the year of
original adoption or, in the case of revision, the year of last revision.Anumber in parentheses indicates the year of last reapproval.A
superscript epsilon (´) indicates an editorial change since the last revision or reapproval.
ε NOTE—The title of this guide and the Referenced Documents were updated in May 2017.
1. Scope 1.7 This international standard was developed in accor-
dance with internationally recognized principles on standard-
1.1 This guide covers the establishment and use of an
ization established in the Decision on Principles for the
ASTMevaluatednucleardatacrosssectionanduncertaintyfile
Development of International Standards, Guides and Recom-
for analysis of single or multiple sensor measurements in
mendations issued by the World Trade Organization Technical
neutron fields related to light water reactor LWR-Pressure
Barriers to Trade (TBT) Committee.
Vessel Surveillance (PVS). These fields include in- and ex-
vessel surveillance positions in operating power reactors,
2. Referenced Documents
benchmark fields, and reactor test regions.
2.1 ASTM Standards:
1.2 Requirements for establishment of ASTM-approved
E170Terminology Relating to Radiation Measurements and
cross section files address data format, evaluation
Dosimetry
requirements, validation in benchmark fields, evaluation of
E185Practice for Design of Surveillance Programs for
error estimates (covariance file), and documentation.Afurther
Light-Water Moderated Nuclear Power Reactor Vessels
requirement for components of the ASTM-approved cross
E482Guide for Application of Neutron Transport Methods
section file is their internal consistency when combined with
for Reactor Vessel Surveillance
sensor measurements and used to determine a neutron spec-
E560Practice for Extrapolating Reactor Vessel Surveillance
trum.
Dosimetry Results, E 706(IC) (Withdrawn 2009)
1.3 Specifications for use include energy region of E693Practice for Characterizing Neutron Exposures in Iron
applicability, data processing requirements, and application of
and Low Alloy Steels in Terms of Displacements Per
uncertainties. Atom
E844Guide for Sensor Set Design and Irradiation for
1.4 This guide is directly related to and should be used
Reactor Surveiillance
primarily in conjunction with Guides E482 and E944, and
E853PracticeforAnalysisandInterpretationofLight-Water
Practices E560, E185, and E693.
Reactor Surveillance Results
1.5 TheASTM cross section and uncertainty file represents
E854Test Method for Application and Analysis of Solid
a generally available data set for use in sensor set analysis.
State Track Recorder (SSTR) Monitors for Reactor Sur-
However, the availability of this data set does not preclude the
veillance
use of other validated data, either proprietary or nonpropri-
E910Test Method for Application and Analysis of Helium
etary. When alternate cross section files are used that deviate
Accumulation Fluence Monitors for Reactor Vessel Sur-
from the requirements laid out in this standard, the deviations
veillance
should be noted to the customer ofr the dosimetry application.
E944Guide for Application of Neutron Spectrum Adjust-
1.6 This standard does not purport to address all of the
ment Methods in Reactor Surveillance
safety concerns, if any, associated with its use. It is the
E1005Test Method for Application and Analysis of Radio-
responsibility of the user of this standard to establish appro-
metric Monitors for Reactor Vessel Surveillance
priate safety and health practices and determine the applica-
E2005Guide for Benchmark Testing of Reactor Dosimetry
bility of regulatory limitations prior to use.
in Standard and Reference Neutron Fields
1 2
This guide is under the jurisdiction of ASTM Committee E10 on Nuclear For referenced ASTM standards, visit the ASTM website, www.astm.org, or
Technology and Applicationsand is the direct responsibility of Subcommittee contact ASTM Customer Service at service@astm.org. For Annual Book of ASTM
E10.05 on Nuclear Radiation Metrology. Standards volume information, refer to the standard’s Document Summary page on
Current edition approved June 1, 2013. Published July 2013. Originally the ASTM website.
publishedasE1018–84.Lastpreviouseditionapprovedin2009asE1018-09.DOI: The last approved version of this historical standard is referenced on
10.1520/E1018-09R13E01. www.astm.org.
Copyright © ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959. United States
´1
E1018 − 09 (2013)
3. Terminology 3.1.1.3 controlled environment—these environments are
well-defined neutron fields with some spectral definitions,
3.1 Definitions of Terms Specific to This Standard:
employed for a restricted set of validation experiments over a
3.1.1 benchmark field—a limited number of neutron fields
range of energies.
have been identified as benchmark fields for the purpose of
3.1.2 dosimetry cross sections—cross sections used for do-
dosimetry sensor calibration and dosimetry cross section data
simetry application and which provide the total cross section
developmentandtesting (1, 2). SeeTerminologyE170.These
for production of particular (measurable) reaction products.
fields are permanent facilities in which experiments can be
These include fission cross sections for production of fission
repeated. In addition, differential neutron spectrum measure-
products, activation cross sections for the production of radio-
ments have been performed in many of the fields to provide,
active nuclei, and cross sections for production of measurable
togetherwithtransportcalculationsandintegralmeasurements,
stable products, such as helium.
the best state-of-the-art neutron spectrum evaluation. To
supplement the data available from benchmark fields, most of
3.1.3 evaluated data—values of physical quantities repre-
which are limited in fluence rate intensity, reactor test regions
senting a current best estimate. Such estimates are developed
for dosimetry method validation have also been defined,
by experts considering measurements or calculations of the
including both in-reactor and ex-vessel dosimetry positions.
quantity of interest, or both. Cross section evaluations, for
Table 1 lists some of the neutron fields that have been used for
example, are conducted by teams of scientists such as the
data development, testing, and evaluation. Other benchmark
ENDF/B Cross Section Evaluation Working Group (CSEWG)
fields used for testing LWR calculations are described in
(see also section 3.1.5.2).
E2005.
3.1.4 Evaluated Nuclear Data File (ENDF)—consists of
3.1.1.1 standard field—these fields are produced by facili-
neutron cross sections and other nuclear data evaluated from
ties and apparatus that are stable, permanent, and whose fields
available experimental measurements and calculations. Two
are reproducible with neutron fluence rate intensity, energy
types of ENDF files exist.
spectra, and angular fluence rate distributions characterized to
3.1.4.1 ENDF/B files—evaluatedfilesofficiallyapprovedby
state-of-the-art accuracy. Important standard field quantities
CSEWG [see ENDF documents 102 (15), 201 (16), and 216
mustbeverifiedbyinterlaboratorymeasurements.Thesefields
(17)] after suitable review and testing.
exist at the National Institute of Standards and Technology
3.1.4.2 ENDF/A files—evaluated files including outdated
(NIST) and other laboratories.
versions of ENDF/B, the International Reactor Dosimetry File
3.1.1.2 reference field—these fields are produced by facili-
(IRDF-2002) (18), the Japanese Evaluated Nuclear Data Li-
ties and apparatus that are permanent and whose fields are
brary(JENDL) (19),BROND(USSR) (20)andotherevaluated
reproducible, less well characterized than a standard field, but
cross section libraries. These files include partial as well as
acceptable as a measurement reference by the community of
complete evaluations.
users.
3.1.5 integral data/differential data—integral data are data
points that represent an integrated sensor’s response over a
range of energy. Examples are measurements of reaction rates
or fission rates in a fission neutron spectrum. Differential data
The boldfaced numbers in parentheses refer to the list of references at the end
of this guide.
TABLE 1 Partial List of Neutron Fields for Validating Dosimetry Cross Sections
Energy
Sample Facility Useful Energy Range Reference
Neutron Field
A
Location for Data Testing Documentation
Median Average
Standard Fields
Thermal Maxwellian NIST . . <0.51 eV
Cf Fission NIST (3) 1.68 MeV 2.13 MeV 100 keV–8 MeV Ref 3
Designation XCF-5-N1
U Thermal Fission NIST (3) 1.57 MeV 1.97 MeV 250 keV–3 MeV Ref 3
Mol-χ (4, 5) Designation XU5-5-N1
ISNF NIST (6) 0.56 MeV ;1.0 MeV 10 keV–3.5 MeV Ref 3
NISUS (7) Designation ISNF(5)-1-L1
Mol-^^ (8)
Reference Fields
BIG TEN LANL (9, 10) 0.33 MeV 0.58 MeV 10 keV–3 MeV Ref 9
Fast Reactor Benchmark
CFRMF EGG-Idaho (9, 11) 0.375 MeV 0.76 MeV 4 keV–2.5 MeV Ref 9
Dosimetry Benchmark 1
Controlled Environments
PCA-PV ORNL (12) . . 100 keV–10 MeV Ref 12
EBR-II ANL-West (13) . . 1 keV–10 MeV Ref 13
FFTF HEDL (14) . . 1 keV–10 MeV Ref 14
A
The requirements for the data testing energy range are much more strict for reference and standard fields than for controlled fields. These testing energy ranges reflect
comparison with calculations based on published spectra for reference and standard fields, but only address data reproducibility for controlled environments.
´1
E1018 − 09 (2013)
are measurements at single energy points or over a relatively support of radiometric, solid state track recorder, helium
smallenergyrange.Examplesaretime-of-flightmeasurements, accumulation dosimetry methods (see Test Methods E853,
proton recoil spectrometry, etc. (21). E854, E910, and E1005).
4.3.2 Other cross sections or sensor response functions
3.1.6 uncertainty file—the uncertainty in cross section data
useful for active or passive dosimetry measurements, for
hasbeenincludedwithevaluatedcrosssectionlibrariesthatare
example, the use of neutron absorption cross sections to
used for dosimetry applications. Because of the correlations
represent attenuation corrections due to covers or self-
between the data points or cross section parameters, these
shielding.
uncertainties, in general, cannot be expressed as variances, but
4.3.3 Cross sections for damage evaluation, such as dis-
rather a covariance matrix must be specified. Through the use
placements per atom (dpa) in iron.
of the covariance matrix, uncertainties in derived quantities,
4.3.4 Related nuclear data needed for dosimetry, such as
such as average cross sections, can be calculated more accu-
branching ratios, fission yields, and atomic abundances.
rately.
4.4 TheASTM-recommended cross sections and uncertain-
4. Significance and Use
ties are based mostly on the ENDF/B-VI and IRDF-2002
4.1 The ENDF/B library in the United States and similar
dosimetry files. Damage cross sections for materials such as
libraries elsewhere, such as JEF (22), JENDL (19), and
iron have been added in order to promote standardization of
BROND (20), provide a compilation of neutron cross section
reported dpa measurements within the dosimetry community.
and other nuclear data for use by the nuclear community. The
Integral measurements from benchmark fields and reactor test
availabilityoftheseexcellentandconsistentevaluationsmakes
regions shall be used to ensure self-consistency and establish
possiblestandardizedusage,therebyallowingeasyreferencing
correlationsbetweencrosssections.Thetotalfileisintendedto
and intercomparisons of calculations. However, as the first
beasself-consistentaspossiblewithrespecttobothdifferential
ENDF/B files were developed it became apparent that they
and integral measurements as applied in LWR environments.
werenotadequateforallapplications.Thisneedresultedinthe
This self-consistency of the data file is mandatory for LWR-
development of the ENDF/B Dosimetry File (17, 23), consist-
pressure vessel surveillance applications, where only very
ing of activation cross sections important for dosimetry appli-
limiteddosimetrydataareavailable.Wheremodificationstoan
cations. This file was made available worldwide. Later, other“
existing evaluated cross section have been made to obtain this
SpecialPurpose”fileswereintroduced (24).IntheENDF/B-VI
self-consistence in LWR environments, the modifications shall
compilation (25), dosimetry files were identified, but they no
be detailed in the associated documentation (see 5.6).
longer appeared as separate evaluation files. The ENDF/V-VII
compilation (26) removed most of the covariance files used by
5. Establishment of Cross Section File
the dosimetry community. It kept the covariance files for the
5.1 Committee—The cross section and uncertainty file shall
“standard cross sections” in a special sub-library, but the
be established and maintained under a responsible task group
covariance data in this sub-library is only provided over the
appointed by Subcommittee E10.05 on Nuclear Radiation
energy range in which each reaction is considered to be a
Metrology. The task group shall review, and approve all data
“standard”,anddoesnotincludethefullenergyrangerequired
before insertion of the file and ensure the adequate testing has
for LWR PVS dosimetry applications.
been performed on the file contents. The task group shall
4.2 Another file of evaluated neutron cross section data has
establish requirements, data formats, etc.
been established by the International Atomic Energy Agency
5.2 Formats—Formats shall generally conform to one of
(IAEA) for reactor dosimetry applications. This file, the
two types. The first format type is that referred to as the
International Reactor Dosimetry File (IRDF-2002) (18), draws
ENDF-6 format and is specified in ENDF-201 (16). The
upontheENDF/Bfilesandsupplementstheseevaluationswith
second format type consists of multigroup data in the 640
a set of reactions evaluated by groups often outside of the
groupSAND-II (27,28)energystructure(seePracticeE693for
United States. Some of the IRDF-2002 supplemental reactions
SAND-II energy group structure). The multigroup data format
represent material evaluations that are currently being exam-
isthepreferredformsinceitismorecompatiblewiththeforms
ined by the CSEWG. The supplemental IRDF-2002 evalua-
tions only include the specific reactions of interest to the typically used to represent facility neutron spectra. The spec-
trum weighting function used to collapse the point cross
dosimetry community and not a full material evaluation. T
...
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.
´1
Designation: E1018 − 09 (Reapproved 2013) E1018 − 09 (Reapproved 2013)
Standard Guide for
Application of ASTM Evaluated Cross Section Data File,
Matrix E706 (IIB)File
This standard is issued under the fixed designation E1018; 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.
ε NOTE—The title of this guide and the Referenced Documents were updated in May 2017.
1. Scope
1.1 This guide covers the establishment and use of an ASTM evaluated nuclear data cross section and uncertainty file for
analysis of single or multiple sensor measurements in neutron fields related to light water reactor LWR-Pressure Vessel
Surveillance (PVS). These fields include in- and ex-vessel surveillance positions in operating power reactors, benchmark fields,
and reactor test regions.
1.2 Requirements for establishment of ASTM-approved cross section files address data format, evaluation requirements,
validation in benchmark fields, evaluation of error estimates (covariance file), and documentation. A further requirement for
components of the ASTM-approved cross section file is their internal consistency when combined with sensor measurements and
used to determine a neutron spectrum.
1.3 Specifications for use include energy region of applicability, data processing requirements, and application of uncertainties.
1.4 This guide is directly related to and should be used primarily in conjunction with Guides E482 and E944, and Practices
E560, E185, and E693.
1.5 The ASTM cross section and uncertainty file represents a generally available data set for use in sensor set analysis. However,
the availability of this data set does not preclude the use of other validated data, either proprietary or nonproprietary. When
alternate cross section files are used that deviate from the requirements laid out in this standard, the deviations should be noted
to the customer ofr the dosimetry application.
1.6 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.
1.7 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
2.1 ASTM Standards:
E170 Terminology Relating to Radiation Measurements and Dosimetry
E185 Practice for Design of Surveillance Programs for Light-Water Moderated Nuclear Power Reactor Vessels
E482 Guide for Application of Neutron Transport Methods for Reactor Vessel Surveillance
E560 Practice for Extrapolating Reactor Vessel Surveillance Dosimetry Results, E 706(IC) (Withdrawn 2009)
E693 Practice for Characterizing Neutron Exposures in Iron and Low Alloy Steels in Terms of Displacements Per Atom (DPA)
E706 Master Matrix for Light-Water Reactor Pressure Vessel Surveillance Standards
E844 Guide for Sensor Set Design and Irradiation for Reactor SurveillanceSurveiillance
E853 Practice for Analysis and Interpretation of Light-Water Reactor Surveillance Results
This guide is under the jurisdiction of ASTM Committee E10 on Nuclear Technology and Applicationsand is the direct responsibility of Subcommittee E10.05 on Nuclear
Radiation Metrology.
Current edition approved June 1, 2013. Published July 2013. Originally published as E1018 – 84. Last previous edition approved in 2009 as E1018-09. DOI:
10.1520/E1018-09R13.10.1520/E1018-09R13E01.
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.
The last approved version of this historical standard is referenced on www.astm.org.
Copyright © ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959. United States
´1
E1018 − 09 (2013)
E854 Test Method for Application and Analysis of Solid State Track Recorder (SSTR) Monitors for Reactor Surveillance
E910 Test Method for Application and Analysis of Helium Accumulation Fluence Monitors for Reactor Vessel Surveillance,
E706 (IIIC)Surveillance
E944 Guide for Application of Neutron Spectrum Adjustment Methods in Reactor Surveillance
E1005 Test Method for Application and Analysis of Radiometric Monitors for Reactor Vessel Surveillance
E2005 Guide for Benchmark Testing of Reactor Dosimetry in Standard and Reference Neutron Fields
3. Terminology
3.1 Definitions of Terms Specific to This Standard:
3.1.1 benchmark field—a limited number of neutron fields have been identified as benchmark fields for the purpose of dosimetry
sensor calibration and dosimetry cross section data development and testing (1, 2). See Terminology E170. These fields are
permanent facilities in which experiments can be repeated. In addition, differential neutron spectrum measurements have been
performed in many of the fields to provide, together with transport calculations and integral measurements, the best state-of-the-art
neutron spectrum evaluation. To supplement the data available from benchmark fields, most of which are limited in fluence rate
intensity, reactor test regions for dosimetry method validation have also been defined, including both in-reactor and ex-vessel
dosimetry positions. Table 1 lists some of the neutron fields that have been used for data development, testing, and evaluation.
Other benchmark fields used for testing LWR calculations are described in E2005 Guide for the Benchmark Testing of Reactor
Dosimetry in Standard and Reference Neutron Fields, E706 (IIE-1).
3.1.1.1 standard field—these fields are produced by facilities and apparatus that are stable, permanent, and whose fields are
reproducible with neutron fluence rate intensity, energy spectra, and angular fluence rate distributions characterized to
state-of-the-art accuracy. Important standard field quantities must be verified by interlaboratory measurements. These fields exist
at the National Institute of Standards and Technology (NIST) and other laboratories.
3.1.1.2 reference field—these fields are produced by facilities and apparatus that are permanent and whose fields are
reproducible, less well characterized than a standard field, but acceptable as a measurement reference by the community of users.
3.1.1.3 controlled environment—these environments are well-defined neutron fields with some spectral definitions, employed
for a restricted set of validation experiments over a range of energies.
3.1.2 dosimetry cross sections—cross sections used for dosimetry application and which provide the total cross section for
production of particular (measurable) reaction products. These include fission cross sections for production of fission products,
activation cross sections for the production of radioactive nuclei, and cross sections for production of measurable stable products,
such as helium.
3.1.3 evaluated data—values of physical quantities representing a current best estimate. Such estimates are developed by
experts considering measurements or calculations of the quantity of interest, or both. Cross section evaluations, for example, are
conducted by teams of scientists such as the ENDF/B Cross Section Evaluation Working Group (CSEWG) (see also section
3.1.5.2).
The boldfaced numbers in parentheses refer to the list of references at the end of this guide.
TABLE 1 Partial List of Neutron Fields for Validating Dosimetry Cross Sections
Energy
Sample Facility Useful Energy Range Reference
Neutron Field
A
Location for Data Testing Documentation
Median Average
Standard Fields
Thermal Maxwellian NIST . . <0.51 eV
Cf Fission NIST (3) 1.68 MeV 2.13 MeV 100 keV–8 MeV Ref 3
Designation XCF-5-N1
U Thermal Fission NIST (3) 1.57 MeV 1.97 MeV 250 keV–3 MeV Ref 3
Mol-χ (4,5) Designation XU5-5-N1
ISNF NIST (6) 0.56 MeV ;1.0 MeV 10 keV–3.5 MeV Ref 3
NISUS (7) Designation ISNF(5)-1-L1
Mol-^^ (8)
Reference Fields
BIG TEN LANL (9,10) 0.33 MeV 0.58 MeV 10 keV–3 MeV Ref 9
Fast Reactor Benchmark
CFRMF EGG-Idaho (9, 11) 0.375 MeV 0.76 MeV 4 keV–2.5 MeV Ref 9
Dosimetry Benchmark 1
Controlled Environments
PCA-PV ORNL (12) . . 100 keV–10 MeV Ref 12
EBR-II ANL-West (13) . . 1 keV–10 MeV Ref 13
FFTF HEDL (14) . . 1 keV–10 MeV Ref 14
A
The requirements for the data testing energy range are much more strict for reference and standard fields than for controlled fields. These testing energy ranges reflect
comparison with calculations based on published spectra for reference and standard fields, but only address data reproducibility for controlled environments.
´1
E1018 − 09 (2013)
3.1.4 Evaluated Nuclear Data File (ENDF)—consists of neutron cross sections and other nuclear data evaluated from available
experimental measurements and calculations. Two types of ENDF files exist.
3.1.4.1 ENDF/B files—evaluated files officially approved by CSEWG [see ENDF documents 102 (15), 201 (16), and 216 (17)]
after suitable review and testing.
3.1.4.2 ENDF/A files—evaluated files including outdated versions of ENDF/B, the International Reactor Dosimetry File
(IRDF-2002) (18), the Japanese Evaluated Nuclear Data Library (JENDL) (19), BROND (USSR) (20) and other evaluated cross
section libraries. These files include partial as well as complete evaluations.
3.1.5 integral data/differential data—integral data are data points that represent an integrated sensor’s response over a range of
energy. Examples are measurements of reaction rates or fission rates in a fission neutron spectrum. Differential data are
measurements at single energy points or over a relatively small energy range. Examples are time-of-flight measurements, proton
recoil spectrometry, etc. (21).
3.1.6 uncertainty file—the uncertainty in cross section data has been included with evaluated cross section libraries that are used
for dosimetry applications. Because of the correlations between the data points or cross section parameters, these uncertainties, in
general, cannot be expressed as variances, but rather a covariance matrix must be specified. Through the use of the covariance
matrix, uncertainties in derived quantities, such as average cross sections, can be calculated more accurately.
4. Significance and Use
4.1 The ENDF/B library in the United States and similar libraries elsewhere, such as JEF (22), JENDL (19), and BROND (20),
provide a compilation of neutron cross section and other nuclear data for use by the nuclear community. The availability of these
excellent and consistent evaluations makes possible standardized usage, thereby allowing easy referencing and intercomparisons
of calculations. However, as the first ENDF/B files were developed it became apparent that they were not adequate for all
applications. This need resulted in the development of the ENDF/B Dosimetry File (17, 23), consisting of activation cross sections
important for dosimetry applications. This file was made available worldwide. Later, other“ Special Purpose” files were introduced
(24). In the ENDF/B-VI compilation (25), dosimetry files were identified, but they no longer appeared as separate evaluation files.
The ENDF/V-VII compilation (26) removed most of the covariance files used by the dosimetry community. It kept the covariance
files for the “standard cross sections” in a special sub-library, but the covariance data in this sub-library is only provided over the
energy range in which each reaction is considered to be a “standard”, and does not include the full energy range required for LWR
PVS dosimetry applications.
4.2 Another file of evaluated neutron cross section data has been established by the International Atomic Energy Agency (IAEA)
for reactor dosimetry applications. This file, the International Reactor Dosimetry File (IRDF-2002) (18), draws upon the ENDF/B
files and supplements these evaluations with a set of reactions evaluated by groups often outside of the United States. Some of the
IRDF-2002 supplemental reactions represent material evaluations that are currently being examined by the CSEWG. The
supplemental IRDF-2002 evaluations only include the specific reactions of interest to the dosimetry community and not a full
material evaluation. The ENDF community requires a complete evaluation before including it in the main ENDF/B evaluated
library.
4.3 The application to LWR surveillance dosimetry may introduce new data needs that can best be satisfied by the creation of
a dedicated cross section file. This file shall be in a form designed for easy application by users (minimal processing). The file shall
consist of the following types of information or indicate the sources of the following type of data that should be used to supplement
the file contents:
4.3.1 Dosimetry cross sections for fission, activation, helium production sensor reactions in LWR environments in support of
radiometric, solid state track recorder, helium accumulation dosimetry methods (see Test Methods E853, E854, E910, and E1005).
4.3.2 Other cross sections or sensor response functions useful for active or passive dosimetry measurements, for example, the
use of neutron absorption cross sections to represent attenuation corrections due to covers or self-shielding.
4.3.3 Cross sections for damage evaluation, such as displacements per atom (dpa) in iron.
4.3.4 Related nuclear data needed for dosimetry, such as branching ratios, fission yields, and atomic abundances.
4.4 The ASTM-recommended cross sections and uncertainties are based mostly on the ENDF/B-VI and IRDF-2002 dosimetry
files. Damage cross sections for materials such as iron have been added in order to promote standardization of reported dpa
measurements within the dosimetry community. Integral measurements from benchmark fields and reactor test regions shall be
used to ensure self-consistency and establish correlations between cross sections. The total file is intended to be as self-consistent
as possible with respect to both differential and integral measurements as applied in LWR en
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