ASTM E2006-99

NOTICE: This standard has either been superseded and replaced by a new version or withdrawn.
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Designation:E2006–99
Standard Guide for
Benchmark Testing of Light Water Reactor Calculations
This standard is issued under the fixed designation E 2006; 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 706 Master Matrix for Light Water Reactor Pressure
Vessel Surveillance Standards, E 706 (O)
1.1 This guide covers general approaches for benchmarking
E 844 Guide for Sensor Set Design and Irritation for Reac-
neutron transport calculations in light water reactor systems.A
tor Surveillance
companionguide(GuideE 706–IIE1)coversuseofbenchmark
E 853 Practice for Analysis and Interpretation of Light-
fields for testing neutron transport calculations and cross
Water Reactor Surveillance Results, E 706 (IA)
sections in well controlled environments. This guide covers
E 854 Test Method for Application and Analysis of Solid
experimental benchmarking of neutron fluence calculations (or
State Track Recorder (SSTR) Monitors for Reactor Sur-
calculations of other exposure parameters such as dpa) in more
veillance, E 706 (IIB)
complex geometries relevant to reactor surveillance. Particular
E 910 Test Method forApplication andAnalysis of Helium
sections of the guide discuss: the use of well-characterized
Accumulation Fluence Monitors for Reactor Vessel Sur-
benchmark neutron fields to provide an indication of the
veillance, E 706 (IIIC)
accuracy of the calculational methods and nuclear data when
E 944 Guide for Application of Neutron Spectrum Adjust-
applied to typical cases; and the use of plant specific measure-
ment Methods in Reactor Surveillance, E 706 (IIA)
ments to indicate bias in individual plant calculations. Use of
E 1006 Practice for Analysis and Interpretation of Physics
these two benchmark techniques will serve to limit plant-
Dosimetry Results for Test Reactors, E 706 (II)
specific calculational uncertainty, and, when combined with
E 1018 Guide for Application of ASTM Evaluated Cross
analytical uncertainty estimates for the calculations, will pro-
Section Data File, E 706 (IIB)
vide uncertainty estimates for reactor fluences with a higher
degree of confidence.
3. Significance and Use
1.2 This standard does not purport to address all of the
3.1 This guide deals with the difficult problem of bench-
safety concerns, if any, associated with its use. It is the
markingneutrontransportcalculationscarriedouttodetermine
responsibility of the user of this standard to establish appro-
fluences for plant specific reactor geometries. The calculations
priate safety and health practices and determine the applica-
are necessary for fluence determination in locations important
bility of regulatory limitations prior to use.
for material radiation damage estimation and which are not
2. Referenced Documents accessible to measurement. The most important application of
such calculations is the estimation of fluence within the reactor
2.1 ASTM Standards:
vessel of operating power plants to provide accurate estimates
E 170 Terminology Relating to Radiation Measurements
of the irradiation embrittlement of the base and weld metal in
and Dosimetry
the vessel. The benchmark procedure must not only prove that
E 261 Practice for Determining Neutron Fluence Rate, Flu-
calculations give reasonable results but that their uncertainties
ence, and Spectra by Radioactivation Techniques
are propagated with due regard to the sensitivities of the
E 262 Test Method for Determining Thermal Neutron Re-
different input parameters used in the transport calculations.
action and Fluence Rates by Radioactivation Techniques
Benchmarking is achieved by building up data bases of
E 482 Guide forApplication of Neutron Transport Methods
benchmark experiments which have different influences on
for Reactor Vessel Surveillance, E 706 (IID)
uncertainty propagation. For example, fission spectra are the
E 560 Practice for Extrapolating Reactor Vessel Surveil-
fundamental data bases which control propagation of cross
lance Dosimetry Results, E 706 (IC)
section uncertainties, while such physics-dosimetry experi-
ments as vessel wall mockups, where measurements are made
ThistestmethodisunderthejurisdictionofASTMCommitteeE-10onNuclear within a simulated reactor vessel wall, control error propaga-
Technology and Applications and is the direct responsibility of Subcommittee
tion associated with geometrical and methods approximations
E10.05 on Nuclear Radiation and Metrology.
in the transport calculations. This guide describes general
Current edition approved Feb. 10, 1999. Published April 1999.
procedures for using neutron fields with known characteristics
Annual Book of ASTM Standards, Vol 12.02.
Copyright © ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959, United States.
E2006
to corroborate the calculational methodology and nuclear data well-characterized facilities that each mock-up part of an
used to derive neutron field information from measurements of LWR-type system. These facilities have the advantage over
neutron sensor response. operating plants that the dimensions and material compositions
3.2 The bases for benchmark field referencing are usually can be more accurately defined, the neutron source can be well
irradiations performed in standard neutron fields with well characterized, and measurements can be made in a large
known energy spectra and intensities. There are, however, less number of locations that would not be accessible in actual
well known neutron fields that have been designed to mockup power systems. In power reactors, one is interested in the
special environments, such as pressure vessel mockups in transport of neutrons from the distributed source in the fuel,
which it is possible to make dosimetry measurements inside of through the reactor internals and water to the vessel, and
the steel volume of the “vessel”. When such mockups are through the vessel to the reactor cavity. Three mockups that
suitably characterized they are also referred to as benchmark together encompass this entire transport problem are described
fields. A benchmark is that against which other things are in 5.1. Modeling and calculating of neutron transport in these
referenced,hencetheterminology“tobenchmarkreference”or various geometries can be expected to identify any bias in
“benchmark referencing”. A variety of benchmark neutron specific parts of the calculations. Biases that can be detected
fields, other than standard neutron fields, have been developed, include those due to modeling the irregular fuel geometry and
or pressed into service, to improve the accuracy of neutron distributed neutron source, those due to errors in the cross-
dosimetry measurement techniques. Some of these special sections or neutron spectra, and those due to calculational
benchmark experiments are discussed in this standard because approximations.
they have identified needs for additional benchmarking or
4.1.3 The benchmarking described above does not provide
because they have been sufficiently documented to serve as
checks on geometries identical to actual plants and does not
benchmarks.
include bias that may exist in the definition of a specific plant
3.3 One dedicated effort to provide benchmarks whose
model. Identification of these types of bias can only be
radiation environments closely resemble those found outside
accomplished using actual plant measurements. Benchmarking
the core of an operating reactor was the Nuclear Regulatory
using these measurements is described in 5.2 and 5.3.
Commission’s Light Water Reactor Pressure Vessel Surveil-
4.1.4 The final aspect of benchmarking is the benchmarking
lance Dosimetry Improvement Program (LWR-PV-SDIP) (1) .
of the dosimetry results. This aspect is treated in Matrix
This program promoted better monitoring of the radiation
E 706(IIE1). It is assumed that the measurements in the
exposure of reactor vessels and, thereby, provided for better
benchmarked facilities and in the actual operating plants are
assessmentofvesselend-of-lifeconditions.Anobjectiveofthe
carried out using benchmarked reactions and dosimeters. This
LWR-PV-SDIP was to develop improved procedures for reac-
involvesusingreactionswhosecrosssectionshavebeenshown
tor surveillance and document them in a series of ASTM
to be consistent with results in these types of neutron environ-
standards (see Matrix E 706). The primary means chosen for
ments.Also,thedosimetersandmeasurementfacilitiesmustbe
validating LWR-PV-SDIP procedures was by benchmarking a
ofadequatequalityandhavemeasurementaccuraciesthathave
series of experimental and analytical studies in a variety of
been verified (such as through round-robin testing). Periodic
fields (see Matrix E 706 IIE1).
recalibration of laboratory measurement devices is also re-
quired using appropriate reference standards.
4. Particulars of Benchmarking Transport Calculations
4.1.4.1 Selection and use of dosimetry should be according
4.1 Benchmarking of neutron transport calculations in-
to Guide E 844, and evaluation of the dosimetry results should
volves several distinct steps that are detailed below.
be in accordance with Practice E 261 and Test Method E 262.
4.1.1 Nuclear data used for transport calculations are evalu-
In particular, to compare measured dosimetry results with
ated using differential data or a combination of integral and
calculated reaction rates or fluences, the following effects must
differential data. This process results in a library of cross
be accounted for: effects of dosimetry perturbations, position
sections and other needed nuclear data (including fission
or gradient corrections, gamma attenuation in counted foils,
spectra) that, in the opinion of the evaluator, gives the best fit
differences in counting geometry from that of calibration
to the available experimental and theoretical results. Some of
standards, dosimeter or reaction product burnup, effects of
information used in evaluating the cross sections may be the
competing reactions in impurities and photofission or photoin-
same as that used directly for benchmarking transport calcula-
duced reactions, and proper treatment of the irradiation history.
tions for LWR systems (see 4.1.2). The cross section bench-
4.1.4.2 The benchmarking of the dosimetry results will also
marking itself is not addressed in this standard. It is assumed
have indicated any bias that exists in the dosimetry cross
that the cross-section set is derived in this fashion to be
sections. These cross sections are essentially independent of
applicable to a variety of calculational geometries and may not
the transport cross sections discussed in 4.1.1. Recommended
give the most accurate answer for LWR geometries. Thus
dosimetry cross sections are given in Guide E 1018.
further benchmarking in LWR geometries is required.
4.1.5 The use of the benchmark data to determine bias in
4.1.2 Transport calculations in LWR geometries may be
calculations and to determine best values for fluence in
benchmarked using measurements made in well-defined and
complex geometries is not straightforward. It often is not clear
how to eight the impact of the different types of information
when inconsistencies exist. Although, most calculations pro-
The boldface numbers given in parentheses refer to a list of references at the
end of the text. duce results that agree with measurements within acceptable
E2006
tolerance, the cause of discrepancies within the tolerance may and by the core-boundary fuel power distributions could not be
not be apparent from the available information. In this case, ignored, otherwise the calculations could contain undetected
there is not universal agreement on the “best” answer, and the biases. Such biases could be further exacerbated by the use of
various approaches to use of the benchmark data can be low-leakage fuel-management schemes.
adopted. Some of these approaches are described in Section 6.
5.1.2.3 A second configuration, VENUS-2, contained a
Caution should be used if it is necessary to extrapolate beyond
plutonium-fueled zone at the periphery of the core (to simulate
the limits of the benchmarks.
burned fuel), and its objective was to investigate how much the
fast neutron fluence is affected by such a core loading, and if
5. Summary of Reference Benchmarks for Reactor
changes in calculational modeling are necessary to account for
Pressure Vessel Surveillance Dosimetry
any effects. The VENUS facility can also provide data to be
5.1 Special Benchmark Irradiation Fields:
used in validation of other sources asymmetries, such as those
5.1.1 One dedicated effort to provide benchmarks whose
due to loading of absorber pins or dummy fuel rods in external
radiation environments closely resemble those found outside
assemblies to limit neutron leakage.
the core of an operating reactor was the Nuclear Regulatory
5.1.3 The PCA/PSF Benchmark:
Commission’s LWR-PV-SDIP (1). This program promoted
5.1.3.1 The task of developing benchmark fields to meet
better monitoring of the radiation exposure of reactor vessels
surveillance dosimetry needs began with the construction,
and, thereby, provided for better assessment of vessel end-of-
adjacent to the Oak Ridge National Laboratory (ORNL) Pool
life conditions. In cooperation with other organizations nation-
Critical Assembly (PCA), of a full-scale-section mockup of a
ally and internationally this program resulted in three bench-
pressure vessel wall in passive and active dosimetry measure-
mark configurations, VENUS (2, 3, 4, 5, 6, 7, 8), PCA/PSF (9,
ments (including neutron spectroscopy) could be made both
10, 11, 12, 13, 14, 15), and NESDIP (16, 17, 18, 19).
outside and within the steel mockup (9, 10, 20). Measurement
5.1.1.1 To serve as benchmarks, these special neutron envi-
14 12 34
positions corresponding to the ⁄, ⁄, and ⁄ thicknesses of the
ronments had to be well characterized both experimentally and
pressure vessel were provided. A simulated surveillance cap-
theoretically. This came to mean that difference between
sule was added to the mockup also. Extensive measurements
measurements and calculations were reconciled and that un-
and calculations provided sufficient characterization of the
certainty bounds for exposure parameters were well defined.
PCAbenchmark experiment so that it was used for a blind test
Target uncertainties were 5 % to 10 % (1s). To achieve these
of neutron transport calculations (9).
objectives, benchmarked dosimetry measurements were com-
5.1.3.2 The PCA benchm
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