ASTM E482-11e1
(Guide)Standard Guide for Application of Neutron Transport Methods for Reactor Vessel Surveillance, E706 (IID)
Standard Guide for Application of Neutron Transport Methods for Reactor Vessel Surveillance, E706 (IID)
SIGNIFICANCE AND USE
3.1 General:
3.1.1 The methodology recommended in this guide specifies criteria for validating computational methods and outlines procedures applicable to pressure vessel related neutronics calculations for test and power reactors. The material presented herein is useful for validating computational methodology and for performing neutronics calculations that accompany reactor vessel surveillance dosimetry measurements (see Master Matrix E706 and Practice E853). Briefly, the overall methodology involves: (1) methods-validation calculations based on at least one well-documented benchmark problem, and (2) neutronics calculations for the facility of interest. The neutronics calculations of the facility of interest and of the benchmark problem should be performed consistently, with important modeling parameters kept the same or as similar as is feasible. In particular, the same energy group structure and common broad-group microscopic cross sections should be used for both problems. The neutronics calculations involve two tasks: (1) determination of the neutron source distribution in the reactor core by utilizing diffusion theory (or transport theory) calculations in conjunction with reactor power distribution measurements, and (2) performance of a fixed fission rate neutron source (fixed-source) transport theory calculation to determine the neutron fluence rate distribution in the reactor core, through the internals and in the pressure vessel. Some neutronics modeling details for the benchmark, test reactor, or the power reactor calculation will differ; therefore, the procedures described herein are general and apply to each case. (See NUREG/CR–5049, NUREG/CR–1861, NUREG/CR–3318, and NUREG/CR–3319.)
3.1.2 It is expected that transport calculations will be performed whenever pressure vessel surveillance dosimetry data become available and that quantitative comparisons will be performed as prescribed by 3.2.2. All dosimetry data accumulated that are applicable to a ...
SCOPE
1.1 Need for Neutronics Calculations—An accurate calculation of the neutron fluence and fluence rate at several locations is essential for the analysis of integral dosimetry measurements and for predicting irradiation damage exposure parameter values in the pressure vessel. Exposure parameter values may be obtained directly from calculations or indirectly from calculations that are adjusted with dosimetry measurements; Guide E944 and Practice E853 define appropriate computational procedures.
1.2 Methodology—Neutronics calculations for application to reactor vessel surveillance encompass three essential areas: (1) validation of methods by comparison of calculations with dosimetry measurements in a benchmark experiment, (2) determination of the neutron source distribution in the reactor core, and (3) calculation of neutron fluence rate at the surveillance position and in the pressure vessel.
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 and health practices and determine the applicability of regulatory requirements prior to use.
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Standards Content (Sample)
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Designation: E482 − 11
StandardGuide for
Application of Neutron Transport Methods for Reactor
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Vessel Surveillance, E706 (IID)
This standard is issued under the fixed designation E482; 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.
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ε NOTE—Missing references to Practice E693 were added to 3.2.1.6, 3.2.1.7 and 3.4.8.3 editorially in November 2012.
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1. Scope Surveillance Standards, E 706(0) (Withdrawn 2011)
E844 Guide for Sensor Set Design and Irradiation for
1.1 Need for Neutronics Calculations—An accurate calcu-
Reactor Surveillance, E 706 (IIC)
lation of the neutron fluence and fluence rate at several
E853 Practice forAnalysis and Interpretation of Light-Water
locations is essential for the analysis of integral dosimetry
Reactor Surveillance Results, E706(IA)
measurements and for predicting irradiation damage exposure
E944 Guide for Application of Neutron Spectrum Adjust-
parameter values in the pressure vessel. Exposure parameter
ment Methods in Reactor Surveillance, E 706 (IIA)
values may be obtained directly from calculations or indirectly
E1018 Guide for Application of ASTM Evaluated Cross
from calculations that are adjusted with dosimetry measure-
Section Data File, Matrix E706 (IIB)
ments; Guide E944 and Practice E853 define appropriate
E2006 Guide for Benchmark Testing of Light Water Reactor
computational procedures.
Calculations
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1.2 Methodology—Neutronics calculations for application
2.2 Nuclear Regulatory Documents:
to reactor vessel surveillance encompass three essential areas:
NUREG/CR-1861 LWR Pressure Vessel Surveillance Do-
(1) validation of methods by comparison of calculations with
simetry Improvement Program: PCA Experiments and
dosimetry measurements in a benchmark experiment, (2)
Blind Test
determination of the neutron source distribution in the reactor
NUREG/CR-3318 LWR Pressure Vessel Surveillance Do-
core, and (3) calculation of neutron fluence rate at the surveil-
simetry Improvement Program: PCA Experiments, Blind
lance position and in the pressure vessel.
Test, and Physics-Dosimetry Support for the PSF Experi-
ments
1.3 This standard does not purport to address all of the
NUREG/CR-3319 LWR Pressure Vessel Surveillance Do-
safety concerns, if any, associated with its use. It is the
simetry Improvement Program: LWR Power Reactor Sur-
responsibility of the user of this standard to establish appro-
veillance Physics-Dosimetry Data Base Compendium
priate safety and health practices and determine the applica-
NUREG/CR-5049 Pressure Vessel Fluence Analysis and
bility of regulatory requirements prior to use.
Neutron Dosimetry
2. Referenced Documents
3. Significance and Use
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2.1 ASTM Standards:
3.1 General:
E693 Practice for Characterizing Neutron Exposures in Iron
3.1.1 Themethodologyrecommendedinthisguidespecifies
and Low Alloy Steels in Terms of Displacements Per
criteria for validating computational methods and outlines
Atom (DPA), E 706(ID)
procedures applicable to pressure vessel related neutronics
E706 MasterMatrixforLight-WaterReactorPressureVessel
calculationsfortestandpowerreactors.Thematerialpresented
herein is useful for validating computational methodology and
for performing neutronics calculations that accompany reactor
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This guide is under the jurisdiction of ASTM Committee E10 on Nuclear
vessel surveillance dosimetry measurements (see Master Ma-
Technology and Applications and is the direct responsibility of Subcommittee
trix E706 and Practice E853). Briefly, the overall methodology
E10.05 on Nuclear Radiation Metrology.
involves: (1) methods-validation calculations based on at least
Current edition approved June 1, 2011. Published June 2011. Originally
approved in 1976. Last previous edition approved in 2007 as E482 – 07 DOI:
10.1520/E0482-11E01.
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For referenced ASTM standards, visit the ASTM website, www.astm.org, or The last approved version of this historical standard is referenced on
contact ASTM Customer Service at service@astm.org. For Annual Book of ASTM www.astm.org.
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Standards volume information, refer to the standard’s Document Summary page on Available from Superintendent of Documents, U.S. Government Printing
the ASTM website. Office, Washington, DC 20402.
Copyright © ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959. United States
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E482 − 11
one well-documented benchmark problem, and (2) neutronics 3.2.1.7 Reaction rates (preferably established relative to
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calculations for the facility of interest. The neutronics calcula- neutron fluence stan
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