ASTM E1018-09(2013)
(Guide)Standard Guide for Application of ASTM Evaluated Cross Section Data File, Matrix E706 (IIB)
Standard Guide for Application of ASTM Evaluated Cross Section Data File, Matrix E706 (IIB)
SIGNIFICANCE AND USE
4.1 The ENDF/B library in the United States and similar libraries elsewhere, such as JEF (9), JENDL (7), and BROND (8), 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 (5, 10), consisting of activation cross sections important for dosimetry applications. This file was made available worldwide. Later, other“ Special Purpose” files were introduced (23). In the ENDF/B-VI compilation (41), dosimetry files were identified, but they no longer appeared as separate evaluation files. The ENDF/V-VII compilation (37) 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) (6) , 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 complet...
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.
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Designation: E1018 − 09 (Reapproved 2013)
Standard Guide for
Application of ASTM Evaluated Cross Section Data File,
Matrix E706 (IIB)
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.
1. Scope 2. Referenced Documents
1.1 This guide covers the establishment and use of an 2.1 ASTM Standards:
ASTMevaluatednucleardatacrosssectionanduncertaintyfile E170Terminology Relating to Radiation Measurements and
for analysis of single or multiple sensor measurements in Dosimetry
neutron fields related to light water reactor LWR-Pressure E185Practice for Design of Surveillance Programs for
Vessel Surveillance (PVS). These fields include in- and ex- Light-Water Moderated Nuclear Power Reactor Vessels
vessel surveillance positions in operating power reactors, E482Guide for Application of Neutron Transport Methods
benchmark fields, and reactor test regions. for Reactor Vessel Surveillance, E706 (IID)
E560Practice for Extrapolating Reactor Vessel Surveillance
1.2 Requirements for establishment of ASTM-approved
Dosimetry Results, E 706(IC) (Withdrawn 2009)
cross section files address data format, evaluation
E693Practice for Characterizing Neutron Exposures in Iron
requirements, validation in benchmark fields, evaluation of
and Low Alloy Steels in Terms of Displacements Per
error estimates (covariance file), and documentation.Afurther
Atom (DPA), E 706(ID)
requirement for components of the ASTM-approved cross
E706MasterMatrixforLight-WaterReactorPressureVessel
section file is their internal consistency when combined with
Surveillance Standards, E 706(0) (Withdrawn 2011)
sensor measurements and used to determine a neutron spec-
E844Guide for Sensor Set Design and Irradiation for
trum.
Reactor Surveillance, E 706 (IIC)
1.3 Specifications for use include energy region of
E853PracticeforAnalysisandInterpretationofLight-Water
applicability, data processing requirements, and application of
Reactor Surveillance Results, E706(IA)
uncertainties.
E854Test Method for Application and Analysis of Solid
State Track Recorder (SSTR) Monitors for Reactor
1.4 This guide is directly related to and should be used
primarily in conjunction with Guides E482 and E944, and Surveillance, E706(IIIB)
E910Test Method for Application and Analysis of Helium
Practices E560, E185, and E693.
Accumulation Fluence Monitors for Reactor Vessel
1.5 TheASTM cross section and uncertainty file represents
Surveillance, E706 (IIIC)
a generally available data set for use in sensor set analysis.
E944Guide for Application of Neutron Spectrum Adjust-
However, the availability of this data set does not preclude the
ment Methods in Reactor Surveillance, E 706 (IIA)
use of other validated data, either proprietary or nonpropri-
E1005Test Method for Application and Analysis of Radio-
etary. When alternate cross section files are used that deviate
metric Monitors for Reactor Vessel Surveillance, E 706
from the requirements laid out in this standard, the deviations
(IIIA)
should be noted to the customer ofr the dosimetry application.
E2005Guide for Benchmark Testing of Reactor Dosimetry
1.6 This standard does not purport to address all of the
in Standard and Reference Neutron Fields
safety concerns, if any, associated with its use. It is the
responsibility of the user of this standard to establish appro-
3. Terminology
priate safety and health practices and determine the applica-
3.1 Definitions of Terms Specific to This Standard:
bility of regulatory limitations prior to use.
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/E1018-09R13. www.astm.org.
Copyright © ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959. United States
E1018 − 09 (2013)
3.1.1 benchmark field—a limited number of neutron fields 3.1.1.3 controlled environment—these environments are
have been identified as benchmark fields for the purpose of well-defined neutron fields with some spectral definitions,
dosimetry sensor calibration and dosimetry cross section data employed for a restricted set of validation experiments over a
developmentandtesting (1, 2). SeeTerminologyE170.These range of energies.
fields are permanent facilities in which experiments can be
3.1.2 dosimetry cross sections—cross sections used for do-
repeated. In addition, differential neutron spectrum measure-
simetry application and which provide the total cross section
ments have been performed in many of the fields to provide,
for production of particular (measurable) reaction products.
togetherwithtransportcalculationsandintegralmeasurements,
These include fission cross sections for production of fission
the best state-of-the-art neutron spectrum evaluation. To
products, activation cross sections for the production of radio-
supplement the data available from benchmark fields, most of
active nuclei, and cross sections for production of measurable
which are limited in fluence rate intensity, reactor test regions
stable products, such as helium.
for dosimetry method validation have also been defined,
3.1.3 evaluated data—values of physical quantities repre-
including both in-reactor and ex-vessel dosimetry positions.
senting a current best estimate. Such estimates are developed
Table 1 lists some of the neutron fields that have been used for
by experts considering measurements or calculations of the
data development, testing, and evaluation. Other benchmark
quantity of interest, or both. Cross section evaluations, for
fieldsusedfortestingLWRcalculationsaredescribedinE2005
example, are conducted by teams of scientists such as the
Guide for the Benchmark Testing of Reactor Dosimetry in
ENDF/B Cross Section Evaluation Working Group (CSEWG)
Standard and Reference Neutron Fields, E706 (IIE-1).
(see also section 3.1.5.2).
3.1.1.1 standard field—these fields are produced by facili-
3.1.4 Evaluated Nuclear Data File (ENDF)—consists of
ties and apparatus that are stable, permanent, and whose fields
neutron cross sections and other nuclear data evaluated from
are reproducible with neutron fluence rate intensity, energy
available experimental measurements and calculations. Two
spectra, and angular fluence rate distributions characterized to
types of ENDF files exist.
state-of-the-art accuracy. Important standard field quantities
3.1.4.1 ENDF/B files—evaluatedfilesofficiallyapprovedby
mustbeverifiedbyinterlaboratorymeasurements.Thesefields
CSEWG [see ENDF documents 102 (3), 201 (4), and 216 (5)]
exist at the National Institute of Standards and Technology
after suitable review and testing.
(NIST) and other laboratories.
3.1.4.2 ENDF/A files—evaluated files including outdated
3.1.1.2 reference field—these fields are produced by facili-
versions of ENDF/B, the International Reactor Dosimetry File
ties and apparatus that are permanent and whose fields are
(IRDF-2002) (6),theJapaneseEvaluatedNuclearDataLibrary
reproducible, less well characterized than a standard field, but
(JENDL) (7), BROND (USSR) (8) and other evaluated cross
acceptable as a measurement reference by the community of
sectionlibraries.Thesefilesincludepartialaswellascomplete
users.
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
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 (24) 1.68 MeV 2.13 MeV 100 keV–8 MeV Ref 24
Designation XCF-5-N1
U Thermal Fission NIST (24) 1.57 MeV 1.97 MeV 250 keV–3 MeV Ref 24
Mol-χ (25, 26) Designation XU5-5-N1
ISNF NIST (27) 0.56 MeV ;1.0 MeV 10 keV–3.5 MeV Ref 24
NISUS (28) Designation ISNF(5)-1-L1
Mol-^^ (29)
Reference Fields
BIG TEN LANL (30, 31) 0.33 MeV 0.58 MeV 10 keV–3 MeV Ref 30
Fast Reactor Benchmark
CFRMF EGG-Idaho (30, 32) 0.375 MeV 0.76 MeV 4 keV–2.5 MeV Ref 30
Dosimetry Benchmark 1
Controlled Environments
PCA-PV ORNL (33) . . 100 keV–10 MeV Ref 33
EBR-II ANL-West (34) . . 1 keV–10 MeV Ref 34
FFTF HEDL (35) . . 1 keV–10 MeV Ref 35
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.
E1018 − 09 (2013)
or fission rates in a fission neutron spectrum. Differential data 4.3.1 Dosimetry cross sections for fission, activation, he-
are measurements at single energy points or over a relatively lium production sensor reactions in LWR environments in
smallenergyrange.Examplesaretime-of-flightmeasurements, support of radiometric, solid state track recorder, helium
proton recoil spectrometry, etc. (38). accumulation dosimetry methods (see Test Methods E853,
E854, E910, and E1005).
3.1.6 uncertainty file—the uncertainty in cross section data
4.3.2 Other cross sections or sensor response functions
hasbeenincludedwithevaluatedcrosssectionlibrariesthatare
useful for active or passive dosimetry measurements, for
used for dosimetry applications. Because of the correlations
example, the use of neutron absorption cross sections to
between the data points or cross section parameters, these
represent attenuation corrections due to covers or self-
uncertainties, in general, cannot be expressed as variances, but
shielding.
rather a covariance matrix must be specified. Through the use
4.3.3 Cross sections for damage evaluation, such as dis-
of the covariance matrix, uncertainties in derived quantities,
placements per atom (dpa) in iron.
such as average cross sections, can be calculated more accu-
4.3.4 Related nuclear data needed for dosimetry, such as
rately.
branching ratios, fission yields, and atomic abundances.
4. Significance and Use
4.4 TheASTM-recommended cross sections and uncertain-
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 (9), JENDL (7), and BROND
iron have been added in order to promote standardization of
(8), provide a compilation of neutron cross section and other
reported dpa measurements within the dosimetry community.
nucleardataforusebythenuclearcommunity.Theavailability
Integral measurements from benchmark fields and reactor test
of these excellent and consistent evaluations makes possible
regions shall be used to ensure self-consistency and establish
standardized usage, thereby allowing easy referencing and
correlationsbetweencrosssections.Thetotalfileisintendedto
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-
developmentoftheENDF/BDosimetryFile (5, 10),consisting
pressure vessel surveillance applications, where only very
of activation cross sections important for dosimetry applica-
limiteddosimetrydataareavailable.Wheremodificationstoan
tions. This file was made available worldwide. Later, other“
existing evaluated cross section have been made to obtain this
SpecialPurpose”fileswereintroduced (23).IntheENDF/B-VI
self-consistence in LWR environments, the modifications shall
compilation (41), 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 (37) removed most of the covariance files used by
5. Establishment of Cross Section File
the dosimetry community. It kept the covariance files for the
“standard cross sections” in a special sub-library, but the
5.1 Committee—The cross section and uncertainty file shall
covariance data in this sub-library is only provided over the
be established and maintained under a responsible task group
energy range in which each reaction is considered to be a
appointed by Subcommittee E10.05 on Nuclear Radiation
“standard”,anddoesnotincludethefullenergyrangerequired
Metrology. The task group shall review, and approve all data
for LWR PVS dosimetry applications.
before insertion of the file and ensure the adequate testing has
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
(IAEA) for reactor dosimetry applications. This file, the 5.2 Formats—Formats shall generally conform to one of
International Reactor Dosimetry File (IRDF-2002) (6), draws
two types. The first format type is that referred to as the
upontheENDF/Bfilesandsupplementstheseevaluationswith ENDF-6 format and is specified in ENDF-201 (4).The second
a set of reactions evaluated by groups often outside of the
format type consists of multigroup data in the 640 group
United States. Some of the IRDF-2002 supplemental reactions
SAND-II (11,12) energy structure (see Practice E693 for
represent material evaluations that are currently being exam-
SAND-II energy group structure). The multigroup data format
ined by the CSEWG. The supplemental IRDF-2002 evalua-
isthepreferredformsinceitismorecompatiblewiththeforms
tions only include the specific reactions of interest to the
typically used to represent facility neutron spectra. The spec-
dosimetry community and not a full material evaluation. The
trum weighting function used to collapse the point cro
...
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