ASTM E482-89(1996)
(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)
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 E 944 and Practice E 853 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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Designation: E 482 – 89 (Reapproved 1996)
Standard Guide for
Application of Neutron Transport Methods for Reactor
Vessel Surveillance, E706 (IID)
This standard is issued under the fixed designation E 482; 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.
1. Scope E 853 Practice for Analysis and Interpretation of Light-
Water Reactor Surveillance Results, E706 (IA)
1.1 Need for Neutronics Calculations—An accurate calcu-
E 944 Guide for Application of Neutron Spectrum Adjust-
lation of the neutron fluence and fluence rate at several
ment Methods in Reactor Surveillance, (IIA)
locations is essential for the analysis of integral dosimetry
E 1018 Guide for Application of ASTM Evaluated Cross
measurements and for predicting irradiation damage exposure
Section Data File (ENDF/A), E706(IIB)
parameter values in the pressure vessel. Exposure parameter
E 706(IE) Damage Correlation for Reactor Vessel Surveil-
values may be obtained directly from calculations or indirectly
lance
from calculations that are adjusted with dosimetry measure-
E 706(IIE) Benchmark Testing of Reactor Vessel Dosim-
ments; Guide E 944 and Practice E 853 define appropriate
etry
computational procedures.
2.2 Nuclear Regulatory Documents:
1.2 Methodology—Neutronics calculations for application
NUREG/CR-1861 LWR Pressure Vessel Surveillance Do-
to reactor vessel surveillance encompass three essential areas:
simetry Improvement Program: PCA Experiments and
(1) validation of methods by comparison of calculations with
Blind Test
dosimetry measurements in a benchmark experiment, (2)
NUREG/CR-3318 LWR Pressure Vessel Surveillance Do-
determination of the neutron source distribution in the reactor
simetry Improvement Program: PCA Experiments, Blind
core, and (3) calculation of neutron fluence rate at the surveil-
Test, and Physics-Dosimetry Support for the PSF Experi-
lance position and in the pressure vessel.
ments
1.3 This standard may involve hazardous materials, opera-
NUREG/CR-3319 LWR Pressure Vessel Surveillance Do-
tions, and equipment. This standard does not purport to
simetry Improvement Program: LWR Power Reactor Sur-
address all of the safety concerns, if any, associated with its
veillance Physics-Dosimetry Data Base Compendium
use. It is the responsibility of the user of this standard to
NUREG/CR-5049 Pressure Vessel Fluence Analysis and
establish appropriate safety and health practices and deter-
Neutron Dosimetry
mine the applicability of regulatory limitations prior to use.
3. Significance and Use
2. Referenced Documents
3.1 General:
2.1 ASTM Standards:
3.1.1 The methodology recommended in this guide specifies
E 170 Terminology Relating to Radiation Measurements
2 criteria for validating computational methods and outlines
and Dosimetry
procedures applicable to pressure vessel related neutronics
E 560 Practice for Extrapolating Reactor Vessel Surveil-
2 calculations for test and power reactors. The material presented
lance Dosimetry Results, E706(IC)
herein is useful for validating computational methodology and
E 693 Practice for Characterizing Neutron Exposures in
for performing neutronics calculations that accompany reactor
Iron and Low Alloy Steels in Terms of Displacements Per
vessel surveillance dosimetry measurements (see Master Ma-
Atom (DPA), E706(ID)
trix E 706 and Practice E 853). Briefly, the overall methodol-
E 706 Master Matrix for Light-Water Reactor Pressure
ogy involves: (1) methods-validation calculations based on at
Vessel Surveillance Standards, E706(0)
least one well documented benchmark problem, and (2) neu-
E 844 Guide for Sensor Set Design and Irradiation for
tronics calculations for the facility of interest. The neutronics
Reactor Surveillance, E706(IIC)
calculations on the facility of interest and on the benchmark
This guide is under the jurisdiction of ASTM Committee E-10 on Nuclear
Technology and Applicationsand is the direct responsibility of Subcommittee
E10.05on Nuclear Radiation Metrology. For standards that are in the draft stage and have not received an ASTM
Current edition approved Oct. 27, 1989. Published December 1989. Originally designation, see Section 5, as well as, Figures 1 and 2 of Matrix E 706.
1 4
published as E 482 – 76. Last previous edition E 482 – 82e . Available from Superintendent of Documents, U.S. Government Printing
Annual Book of ASTM Standards, Vol 12.02. Office, Washington, DC 20402.
Copyright © ASTM, 100 Barr Harbor Drive, West Conshohocken, PA 19428-2959, United States.
E 482
238 58 54
problem should be as nearly the same as is feasible; in or U(n,f), and Ni(n,p) or Fe(n,p); additional reactions that
particular, the group structure and common broad-group mi- aid in spectral characterization, such as provided by Cu, Ti, and
croscopic cross sections should be preserved for both prob- Co-A1, should also be included in the benchmark measure-
lems. The neutronics calculations involve two tasks: (1) ments. The Np(n,f) reaction is an important reaction since it
determination of the neutron source distribution in the reactor gives information similar to dpa. Practices E 693 and E 853
core by utilizing diffusion theory (or transport theory) calcu- and Guides E 844 and E 944 discuss this criterion.
lations in conjunction with reactor power distribution measure- 3.2.2 Methodology Validation—It is essential that the neu-
ments, and (2) performance of a fixed fission rate neutron tronics methodology employed for predicting neutron fluence
source (fixed-source) transport theory calculation to determine in a power reactor pressure vessel be validated by accurately
the neutron fluence rate distribution in the reactor core, through predicting appropriate benchmark dosimetry results. In addi-
the internals and in the pressure vessel. Some neutronics tion, the following documentation should be submitted: (1)
modeling details for the benchmark, test reactor, or the power convergence study results, and (2) estimates of variances and
reactor calculation will differ; therefore, the procedures de- covariances for fluences and reaction rates arising from uncer-
scribed herein are general and apply to each case. (See tainties in both the source and geometric modeling.
NUREG/CR–5049, NUREG/CR–1861, NUREG/CR–3318, 3.2.2.1 For example, model specifications for S methods on
n
and NUREG/CR–3319.) which convergence studies should be performed include: (1)
3.1.2 It is expected that transport calculations will be group structure, (2) spatial mesh, and (3) angular quadrature.
performed whenever pressure vessel surveillance dosimetry One-dimensional calculations may be performed to check the
data become available and that quantitative comparisons will adequacy of group structure and spatial mesh. Two-
be performed as prescribed by 3.2.2. All dosimetry data dimensional calculations should be employed to check the
accumulated that are applicable to a particular facility should adequacy of the angular quadrature. A P cross section expan-
be included in the comparisons. sion is recommended along with an S minimum quadrature.
3.2 Validation—Prior to performing transport calculations 3.2.2.2 Uncertainties that are propagated from known un-
for a particular facility, the computational methods must be certainties in nuclear data are recommended, but they are not
validated by comparing results with measurements made on a required. Appropriate computer programs and covariance data
benchmark experiment. Criteria for establishing a benchmark are available, however, and sensitivity data may be obtained as
experiment for the purpose of validating neutronics methodol- an intermediate step in determining uncertainty estimates.
ogy should include those set forth in Guide E 944 as well as 3.2.2.3 Effects of known uncertainties in geometry and
those prescribed in 3.2.1. A discussion of the limiting accuracy source distribution should be evaluated based on the following
of benchmark validation procedures for the LWR surveillance test cases: (1) reference calculation with a time-averaged
program is given in Footnote 5. source distribution and with best estimates of the core, thermal
3.2.1 Requirements for Benchmarks—In order for a particu- shield and pressure vessel locations, (2) reference case geom-
lar experiment to qualify as a calculational benchmark, the etry with maximum and minimum expected deviations in the
following criteria are recommended: source distribution, and (3) reference case source distribution
3.2.1.1 Sufficient information must be available to accu- with maximum expected spatial perturbations of the core,
rately determine the neutron source distribution in the reactor thermal shield, and pressure vessel.
core, 3.2.2.4 Measured and calculated integral parameters should
3.2.1.2 Measurements must be reported in at least two be compared for all test cases. It is expected that larger
ex-core locations, well separated by steel or coolant, uncertainties are associated with geometry and neutron source
3.2.1.3 Uncertainty estimates should be reported for dosim- specifications than with parameters included in the conver-
etry measurements and calculated fluences including calculated gence study. Problems associated with space, energy, and angle
exposure parameters and calculated dosimetry activities, discretizations can be identified and corrected. Uncertainties
3.2.1.4 Quantitative criteria, consistent with those specified associated with geometry specifications are inherent in the
in the methods validation 3.2.2, must be published and dem- structure tolerances. Calculations based on the expected ex-
onstrated to be achievable, tremes provide a measure of the sensitivity of integral param-
3.2.1.5 Differences between measurements and calculations eters to the selected variables. Variations in the proposed
should be consistent with the uncertainty estimates in 3.2.1.3, convergence and uncertainty evaluations are appropriate when
3.2.1.6 Results for exposure parameter values of neutron the above procedures are inconsistent with the methodology to
fluence greater than 1 MeV and 0.1 MeV [f(E > 1 MeV and be validated. As-built data could be used to reduce the
0.1 MeV)] and of displacements per atom (dpa) should be uncertainty in geometrical dimensions.
reported consistent with Practices E 693 and E 853, and
3.2.1.7 Reaction rates (preferably established relative to
neutron fluence standards) must be reported for Np(n,f) 6
Weisbin, C. R., et al, Application of FORSS Sensitivity and Uncertainty
Methodology to Fast Reactor Benchmark Analysis, ORNL/TM-5563, December
1976.
5 7
Carlson, B. J., and Lathrop, K. O., “Transport Theory—The Method of Discrete Much of the nuclear covariance and sensitivity
...
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