ASTM E1018-95
(Guide)Standard Guide for Application of ASTM Evaluated Cross Section Data File, Matrix E 706 (IIB)
Standard Guide for Application of ASTM Evaluated Cross Section Data File, Matrix E 706 (IIB)
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 r
elated to 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 E 482 and E 944, and Practices E 560, E 185, and E 693.
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.
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: E 1018 – 95
Standard Guide for
Application of ASTM Evaluated Cross Section Data File,
Matrix E 706 (IIB)
This standard is issued under the fixed designation E 1018; the number immediately following the designation indicates the year of
original adoption or, in the case of revision, the year of last revision. A number in parentheses indicates the year of last reapproval. A
superscript epsilon (e) indicates an editorial change since the last revision or reapproval.
2,3
1. Scope Water-Cooled Nuclear Power Reactor Vessels, E706 (IF)
E 482 Guide for Application of Neutron Transport Methods
1.1 This guide covers the establishment and use of an
2,3
for Reactor Vessel Surveillance, E706 (IID)
ASTM evaluated nuclear data cross section and uncertainty file
E 560 Practice for Extrapolating Reactor Vessel Surveil-
for analysis of single or multiple sensor measurements in
2,3
lance Dosimetry Results, E706 (IC)
neutron fields r
E 693 Practice for Characterizing Neutron Exposures in
elated to LWR-Pressure Vessel Surveillance (PVS). These
Iron and Low Alloy Steels in Terms of Displacements Per
fields include in- and ex-vessel surveillance positions in
,
2 3
Atom (DPA), E706 (ID)
operating power reactors, benchmark fields, and reactor test
E 706 Master Matrix for Light-Water Reactor Pressure
regions.
,
2 3
Vessel Surveillance Standards, E706 (O)
1.2 Requirements for establishment of ASTM-approved
E 844 Guide for Sensor Set Design and Irradiation for
cross section files address data format, evaluation require-
2,3
Reactor Surveillance, E706 (IIC)
ments, validation in benchmark fields, evaluation of error
E 853 Practice for Analysis and Interpretation of Light-
estimates (covariance file), and documentation. A further
2,3
Water Reactor Surveillance Results, E706 (IA)
requirement for components of the ASTM-approved cross
E 854 Test Method for Application and Analysis of Solid
section file is their internal consistency when combined with
State Track Recorder (SSTR) Monitors for Reactor Sur-
sensor measurements and used to determine a neutron spec-
,
2 3
veillance, E706 (IIIB)
trum.
E 910 Test Method for Application and Analysis of Helium
1.3 Specifications for use include energy region of applica-
Accumulation Fluence Monitors for Reactor Vessel Sur-
bility, data processing requirements, and application of uncer-
2,3
veillance, E706 (IIIC)
tainties.
E 944 Guide for Application of Neutron Spectrum Adjust-
1.4 This guide is directly related to and should be used
2,3
ment Methods in Reactor Surveillance, (IIA)
primarily in conjunction with Guides E 482 and E 944, and
E 1005 Test Method for Application and Analysis of Radio-
Practices E 560, E 185, and E 693.
metric Monitors for Reactor Vessel Surveillance, E706
1.5 The ASTM cross section and uncertainty file represents
,
2 3
(IIIA)
a generally available data set for use in sensor set analysis.
However, the availability of this data set does not preclude the
3. Terminology
use of other validated data, either proprietary or nonpropri-
3.1 Definitions of Terms Specific to This Standard:
etary.
3.1.1 benchmark field—a limited number of neutron fields
1.6 This standard does not purport to address all of the
have been identified as benchmark fields for the purpose of
safety concerns, if any, associated with its use. It is the
dosimetry sensor calibration and dosimetry cross section data
responsibility of the user of this standard to establish appro-
development and testing (1, 2). These fields are permanent or
priate safety and health practices and determine the applica-
semi-permanent facilities in which experiments can be re-
bility of regulatory limitations prior to use.
peated. In addition, differential neutron spectrum measure-
2. Referenced Documents ments have been performed in many of the fields to provide,
together with transport calculations and integral measurements,
2.1 ASTM Standards:
the best state-of-the-art neutron spectrum evaluation. To
E 185 Practice for Conducting Surveillance Tests for Light
supplement the data available from benchmark fields, most of
1 2
This guide is under the jurisdiction of ASTM Committee E-10 on Nuclear The reference in parentheses refers to Section 5 as well as Figs. 1 and 2 of
Technology and Applicationsand is the direct responsibility of Subcommittee Matrix E 706.
E10.05on Nuclear Radiation Metrology. Annual Book of ASTM Standards, Vol. 12.02.
Current edition approved April 15, 1995. Published June 1995. Originally The boldfaced numbers in parentheses refer to the list of references at the end
published as E 1018 – 84. Last previous edition E 1018 – 84 (1991). of this guide.
Copyright © ASTM, 100 Barr Harbor Drive, West Conshohocken, PA 19428-2959, United States.
E 1018
which are limited in flux intensity, reactor test regions for 3.1.4.1 ENDF/B files—evaluated files officially approved by
dosimetry method validation have also been defined, including CSEWG [see ENDF documents 102 (3), 201 (4), and 216 (5)]
both in-reactor and ex-vessel dosimetry positions. Table 1 lists
after suitable review and testing. The current recommended set
the neutron fields that are being used for data development, of ENDF/B files is ENDF/B-VI, revision 2, July 1993.
testing, and evaluation.
3.1.4.2 ENDF/A files—evaluated files including outdated
3.1.1.1 standard 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-90) (6), the Japanese Evaluated Nuclear Data Library
reproducible with neutron flux intensity, energy spectra, and
(JENDL) (7), and BROND (USSR) (8) evaluated cross section
angular flux distributions characterized to state-of-the-art ac-
libraries.
curacy. These fields exist at the National Institute of Standards
3.1.5 integral data/differential data—integral data are data
and Technology (NIST) and other laboratories.
points that represent an integrated sensor’s response over a
3.1.1.2 reference field—these fields are produced by facili-
range of energy. Examples are measurements of reaction rates
ties and apparatus that are permanent and whose fields are
or fission rates in a fission neutron spectrum. Differential data
reproducible, less well characterized than a standard field, but
are measurements at single energy points or over a relatively
acceptable as a measurement reference by the community of
small energy range. Examples are time-of-flight measurements,
users.
proton recoil spectrometry, etc.
3.1.1.3 controlled environment—these environments are
3.1.6 uncertainty file—the uncertainty in cross section data
well-defined neutron fields with some spectral definitions,
has been included with evaluated cross section libraries that are
employed for a restricted set of validation experiments over a
used for dosimetry applications. Because of the correlations
range of energies.
between the data points or cross section parameters, these
3.1.2 dosimetry cross sections—cross sections used for
uncertainties, in general, cannot be expressed as variances, but
dosimetry application and which provide the total cross section
rather a covariance matrix must be specified. Through the use
for production of particular (measurable) reaction products.
of the covariance matrix, uncertainties in derived quantities,
These include fission cross sections for production of fission
such as average cross sections, can be calculated more accu-
products, activation cross sections for the production of radio-
rately.
active nuclei, and cross sections for production of measurable
stable products, such as helium.
4. Significance and Use
3.1.3 evaluated data—values of physical quantities repre-
senting a current best estimate. Such estimates are developed 4.1 The ENDF/B library in the United States and similar
by experts considering measurements or calculations of the libraries elsewhere, such as JEF (9), JENDL (7), and BROND
quantity of interest, or both. Cross section evaluations, for (8), provide a compilation of neutron cross section and other
example, are conducted by teams of scientists such as the nuclear data for use by the nuclear community. The availability
ENDF/B Cross Section Evaluation Working Group (CSEWG) of these excellent and consistent evaluations makes possible
(see also section 3.1.5.2). standardized usage, thereby allowing easy referencing and
3.1.4 Evaluated Nuclear Data File (ENDF)—consists of intercomparisons of calculations. However, as the first
neutron cross sections and other nuclear data evaluated from ENDF/B files were developed it became apparent that they
available experimental measurements and calculations. Two were not adequate for all applications. This need resulted in the
types of ENDF files exist. development of the ENDF/B Dosimetry File (5, 10), consisting
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-x (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.
E 1018
of activation cross sections important for dosimetry applica- Metrology. The task group shall review, test, and approve all
tions. This file was made available worldwide. Later, other“ data before insertion of the file. The task group shall establish
Special Purpose” files were introduced. In the latest ENDF/ requirements, data formats, etc.
B-VI compilation, dosimetry files are identified, but do not
5.2 Formats—Formats shall generally conform to one of
typically appear as separate evaluation files.
two types. The first format type is that referred to as the
4.2 Another file of evaluated neutron cross section data has
ENDF-6 format and is specified in ENDF-201 (4). The second
been established by the International Atomic Energy Agency
format type consists of multigroup data in the 640 group
(IAEA) for reactor dosimetry applications. This file, the
SAND-II (11, 12) energy structure (see Practice E 693 for
International Reactor Dosimetry File (IRDF-90) (6), draws
SAND-II energy group structure). The multigroup data format
upon the ENDF/B-VI files and supplements these evaluations
is the preferred form since it is more compatible with the forms
with a set of reactions evaluated by groups often outside of the
typically used to represent facility neutron spectrum. The
United States. Some of the IRDF-90 supplemental reactions
spectrum weighting function used to collapse the point cross
represent material evaluations that are currently being exam-
section data onto the multigroup energy grid should be generic
ined by the CSEWG. The supplemental IRDF-90 evaluations
in nature and shall be completely specified in the associated
only include the specific reactions of interest to the dosimetry
documentation.
community and not a full material evaluation. The ENDF
5.3 Cross Section Evaluation—Most evaluations generally
community requires a complete evaluation before including it
shall be based on the IRDF-90 Dosimetry File. Cross sections
in the ENDF/B evaluated library.
shall be consistent within error bounds for selected benchmark
4.3 The application to LWR surveillance dosimetry intro-
fields (see 5.4 and Table 1). Dosimetry cross sections presently
duces new data needs that can best be satisfied by the creation
not in ENDF/B or IRDF-90 shall be obtained from other
of a special cross section file. This file shall be in a form
sources or new evaluations. Other cross sections may be
designed for easy application by users (minimal processing).
obtained from other sources, for example, the dpa cross section
The file shall consist of the following:
for iron may be obtained from Practice E 693.
4.3.1 Dosimetry cross sections for fission, activation, he-
5.4 Cross Section Validation—The cross section file will be
lium production, and damage sensor reactions in LWR envi-
validated for LWR applications using dosimetry measurements
ronments in support of radiometric, solid state track recorder,
made in benchmark fields. Such validation may result in
helium accumulation, and damage monitor dosimetry methods
necessary modifications to cross sections to eliminate signifi-
(see Test Methods E 853, E 854, E 910, and E 1005 and Matrix
cant biases. Modification of ENDF/B and IRDF-90 files shall
E 706-IIID).
be done in a manner consistent with the uncertainties specified
4.3.2 Other cross sections or sensor response functions
for the differential data, using a least squares methodology.
useful for active or passive dosimetry measurements, for
5.5 Related Nuclear Data for Dosimetry Application—All
example, the use of neutron absorption cross sections to
necessary related data shall be specified. These data include
represent attenuation corrections due to covers or self-
isotopic abundances, gamma branching ratios, fission yields,
shielding.
half-lives, etc., as appropriate. Updates of these data shall
4.3.3 Cross sections for damage evaluation, such as dis-
require, in general, a revalidation of the cross section (see 5.4).
placements per atom (dpa) in iron.
In the ENDF-6 format this data can be specified as comment
4.3.4 Related nuclear data needed for dosimetry, such as
cards in the File 1 General Information section. The evaluation
branching ratios, fission yields, and atomic abundances.
file or associated documentation may cite a comprehensive
4.4 The ASTM-recommended cross sections and uncertain-
dosimetry-quality source, such as the Nuclear Data Guide for
ties are based mostly on the ENDF/B and IRDF dosimetry
Reactor Neutron Metrology (13), for the related nuclear data.
files. Damage cross sections for materials such as iron may be
5.5.1 If the related data is not explicitly provided in the
added in order to promote standardization of reported dpa
cross section evalu
...
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