ASTM E2006-22

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Designation: E2006 − 16 E2006 − 22
Standard Guide for
Benchmark Testing of Light Water Reactor Calculations
This standard is issued under the fixed designation E2006; 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.
1. Scope
1.1 This guide covers general approaches for benchmarking neutron transport calculations for pressure vessel surveillance
programs in light water reactor systems. A companion guide (Guide E2005) covers use of benchmark fields for testing neutron
transport calculations and cross sections in well controlled environments. This guide covers experimental benchmarking of neutron
fluence calculations (or calculations of other exposure parameters such as dpa) in more complex geometries relevant to reactor
pressure vessel surveillance. Particular sections of the guide discuss: the use of well-characterized benchmark neutron fields to
provide an indication of the accuracy of the calculational methods and nuclear data when applied to typical cases; and the use of
plant specific measurements to indicate bias in individual plant calculations. Use of these two benchmark techniques will serve to
limit plant-specific calculational uncertainty, and, when combined with analytical uncertainty estimates for the calculations, will
provide uncertainty estimates for reactor fluences with a higher degree of confidence.
1.2 Although this guide and the companion guide, Guide E2005, are focused on power reactors, the principle of this guide is also
applicable to non-power light water reactor pressure vessel surveillance programs.
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 safety, health, and healthenvironmental practices and determine the
applicability of regulatory limitations prior to use.
1.4 This international standard was developed in accordance with internationally recognized principles on standardization
established in the Decision on Principles for the Development of International Standards, Guides and Recommendations issued
by the World Trade Organization Technical Barriers to Trade (TBT) Committee.
2. Referenced Documents
2.1 ASTM Standards:
E170 Terminology Relating to Radiation Measurements and Dosimetry
E261 Practice for Determining Neutron Fluence, Fluence Rate, and Spectra by Radioactivation Techniques
E262 Test Method for Determining Thermal Neutron Reaction Rates and Thermal Neutron Fluence Rates by Radioactivation
Techniques
E706 Master Matrix for Light-Water Reactor Pressure Vessel Surveillance Standards
E844 Guide for Sensor Set Design and Irradiation for Reactor Surveillance
E944 Guide for Application of Neutron Spectrum Adjustment Methods in Reactor Surveillance
E1006 Practice for Analysis and Interpretation of Physics Dosimetry Results from Test Reactor Experiments
E1018 Guide for Application of ASTM Evaluated Cross Section Data File
This test method guide is under the jurisdiction of ASTM Committee E10 on Nuclear Technology and Applications and is the direct responsibility of Subcommittee
E10.05 on Nuclear Radiation Metrology.
Current edition approved June 1, 2016Feb. 1, 2022. Published July 2016March 2022. Originally approved in 1999. Last previous edition approved in 20102016 as
E2006 – 10.E2006 – 16. DOI: 10.1520/E2006-16.
For referenced ASTM standards, visit the ASTM website, www.astm.org, or contact ASTM Customer Service at service@astm.org. For Annual Book of ASTM Standards
volume information, refer to the standard’s Document Summary page on the ASTM website.
Copyright © ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959. United States
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E2005 Guide for Benchmark Testing of Reactor Dosimetry in Standard and Reference Neutron Fields
3. Terminology
3.1 Definitions—definitions of terms used in this guide may be found in Terminology E170.
4. Significance and Use
4.1 This guide deals with the difficult problem of benchmarking neutron transport calculations carried out to determine fluences
for plant specific reactor geometries. The calculations are necessary for fluence determination in locations important for material
radiation damage estimation and which are not accessible to measurement. The Typically, the most important application of such
calculations is the estimation of fluence within the reactor vessel of operating power plants light water reactors (LWR) to provide
accurate estimates of the irradiation embrittlement of the base and weld metal in the vessel. The benchmark procedure must not
only prove that calculations give reasonable results but that their uncertainties are propagated with due regard to the sensitivities
of the different input parameters used in the transport calculations. Benchmarking is achieved by building up data bases of
benchmark experiments that have different influences on uncertainty propagation. For example, fission spectra are the fundamental
data bases which control propagation of cross section uncertainties, while such physics-dosimetry experiments as vessel wall
mockups,in simple vessel wall mockups where measurements are made within a simulated reactor vessel wall, control error
propagation associated with geometrical and methods approximations in the transport calculations. the integral effect of
uncertainties in iron cross sections (absorption and elastic and inelastic scattering) are dominant and have been bounded by the
agreement between calculation and measurement. For more complicated integral benchmarks, other factors such as: uncertainties
in the distribution of fission sources, geometry, the energy-dependent cross sections, and the angular scattering distribution for
elemental components of major materials in the neutron field (such as water and iron) may all be important uncertainty
contributors. This guide describes general procedures for using neutron fields with known characteristics to corroborate the
calculational methodology and nuclear data used to derive neutron field information from measurements of neutron sensor
response.
4.2 The bases for benchmark field referencing are usually irradiations performed in standard neutron fields with well-known
energy spectra and intensities. There are, however, less well known neutron fields that have been designed to mockup special
environments, such as pressure vessel mockups in which it is possible to make dosimetry measurements inside of the steel volume
of the “vessel”. When such mockups are suitably characterized, they are also referred to as benchmark fields. A benchmark is that
against which other things are referenced, hence the terminology “to benchmark reference” or “benchmark referencing”. A variety
of benchmark neutron fields, other than standard neutron fields, have been developed, or pressed into service, to improve the
accuracy of neutron dosimetry measurement techniques. Some of these special benchmark experiments are discussed in this
standard because they have identified needs for additional benchmarking or because they have been sufficiently documented to
serve as benchmarks.
4.3 One dedicated effort to provide benchmarks whose radiation environments closely resemble those found outside the core of
an operating reactor was the Nuclear Regulatory Commission’s Light Water Reactor Pressure Vessel Surveillance Dosimetry
Improvement Program (LWR-PV-SDIP) (1) . This program promoted better monitoring of the radiation exposure of reactor vessels
and, thereby, provided for better assessment of vessel end-of-life conditions. An objective of the LWR-PV-SDIP was to develop
improved procedures for reactor surveillance and document them in a series of ASTM standards (see Matrix E706). The primary
means chosen for validating LWR-PV-SDIP procedures was by benchmarking a series of experimental and analytical studies in
a variety of fields (see Guide E2005).
5. Particulars of Benchmarking Transport Calculations
5.1 Benchmarking of neutron transport calculations involves several distinct steps that are detailed below.
5.1.1 Nuclear data used for transport calculations are evaluated using differential data or a combination of integral and differential
data. This process results in a library of cross sections and other needed nuclear data (including fission spectra) that, in the opinion
of the evaluator, gives the best fit to the available experimental and theoretical results. Some of the information used in evaluating
the cross sections may be the same as that used directly for benchmarking transport calculations for LWR systems (see 4.1.25.1.2).
The cross section benchmarking itself is not addressed in this standard. It is assumed that the cross-section set is derived in this
The boldface numbers given in parentheses refer to a list of references at the end of the text.
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fashion to be applicable to a variety of calculational geometries and may not give the most accurate answer for LWR geometries.
Thus further benchmarking in LWR geometries is required.
5.1.2 Transport calculations in LWR geometries may be benchmarked using measurements made in well-defined and
well-characterized facilities that each mock-up part of an LWR-type system. These facilities have the advantage over operating
plants that the dimensions and material compositions can be more accurately defined, the neutron source can be well characterized,
and measurements can be made in a large number of locations that would not be accessible in actual power systems. In power
reactors, one is interested in the transport of neutrons from the distributed source in the fuel, through the reactor internals and water
to the vessel, and through the vessel to the reactor cavity. Three mockups that together encompass this entire transport problem
are described in systems. 5.1. Modeling and calculating of neutron transport in these various geometries can be expected to identify
any bias in specific parts of the calculations. Biases that can be detected include those due to modeling the irregular fuel geometry
and distributed neutron source, those due to errors in the cross-sections or neutron spectra, and those due to calculational
approximations.
5.1.2.1 In power reactors, one is interested in the transport of neutrons from the distributed source in the fuel, through the reactor
internals and water to the vessel, and through the vessel to the reactor cavity. Three mockups that together encompass this entire
transport problem are described in 6.1. Modeling and calculating of neutron transport in these various geometries can be expected
to identify any bias in specific parts of the calculations. Biases that can be detected include those due to modeling the irregular
fuel geometry and distributed neutron source, those due to errors in the cross-sections or neutron spectra, and those due to
calculational approximations.
5.1.2.2 In non-power reactors, the objective is the same in that the purpose is to characterize the transport of neutrons from the
distributed source in the fuel to and through the pressure vessel. However, in many non-power reactors, the geometries between
the reactor core and the pressure vessel are significantly different from those represented by the mockups described in Section 6.
In this case the evaluator must justify the validity of using the benchmarks discussed in Section 6. If these benchmarks cannot be
justified, other benchmarks must be identified and their use justified.
5.1.3 The benchmarking described above does not provide checks on geometries identical to actual plants and does not include
bias that may exist in the definition of a specific plant model. Identification of these types of bias can only be accomplished using
actual plant measurements. Benchmarking using these measurements is described in 5.26.2 and 5.36.3.
5.1.4 The final aspect of benchmarking is the benchmarking of the dosimetry results. This aspect is treated in Guide E2005. It is
assumed that the measurements in the benchmarked facilities and in the actual operating plants are carried out using benchmarked
reactions and dosimeters. This involves using reactions whose cross sections have been shown to be consistent with results in these
types of neutron environments. Also, the dosimeters and measurement facilities must be of adequate quality and have measurement
accuracies that have been verified (such as through round-robin testing). Periodic recalibration of laboratory measurement devices
is also required using appropriate reference standards.
5.1.4.1 The selection and use of dosimeters should be according to Guide E844, and evaluation of the dosimetry results should
be in accordance with Practice E261 and Test Method E262. In particular, to compare measured dosimetry results with calculated
reaction rates or fluences, the following effects must be accounted for: effects of dosimetry perturbations, position or gradient
corrections, gamma attenuation in counted foils, differences in counting geometry from that of calibration standards, dosimeter or
reaction product burnup, effects of competing reactions in impurities and photofission or photoinduced reactions, and proper
treatment of the irradiation history.
5.1.4.2 The benchmarking of the dosimetry results will also have indicated any bias that exists in the dosimetry cross sections.
These cross sections are essentiallymostly independent of the transport cross sections discussed in 4.1.15.1.1. , although some
hidden correlations may be present due to evaluations taking into account integral results. Recommended dosimetry cross sections
are given in Guide E1018.
5.1.5 The use of the benchmark data to determine bias in calculations and to determine best values for fluence in complex
geometries is not straightforward. It often is not clear how to weight the impact of the different types of information when
inconsistencies exist. Although, most calculations produce results that agree with measurements within acceptable tolerance, the
cause of discrepancies within the tolerance may not be apparent from the available information. In this case, there is not universal
agreement on the “best” answer, and the various approaches to use of the benchmark data can be adopted. Some of these
approaches are described in Section 67. Caution should be used if it is necessary to extrapolate beyond the limits of the
benchmarks.
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6. Summary of Reference Benchmarks for Transport Calculations for Reactor Pressure Vessel Surveillance Programs
6.1 Special Benchmark Irradiation Fields:
6.1.1 One dedicated effort to provide benchmarks whose radiation environments closely resemble those found outside the core of
an operating reactor was the Nuclear Regulatory Commission’s LWR-PV-SDIP (1). This program promoted better monitoring of
the radiation exposure of reactor vessels and, thereby, provided for better assessment of vessel end-of-life conditions. In
cooperation with other organizations nationally and internationally this program resulted in three benchmark configurations,
VENUS (22-8, 3, 4, 5, 6, 7, 8), PCA/PSF (99-15, 10, 11, 12, 13, 14, 15), and NESDIP (1616-19, 17, 18, 19).
6.1.1.1 To serve as benchmarks, these special neutron environments had to be well characterized both experimentally and
theoretically. This came to mean that differences between measurements and calculations were reconciled and that uncertainty
bounds for exposure parameters were well defined. Target uncertainties were 5 % to 10 % (1σ). To achieve these objectives,
benchmarked dosimetry measurements were combined with neutron transport calculations, and statistical uncertainty analysis and
spectral adjustment techniques were used to establish the uncertainty bounds.
6.1.1.2 Taken together, the three benchmarks provide coverage from the fuel region to the vessel cavity. The VENUS facility was
set up to measure spatial fluence distributions and neutron spectra near the fuel region and core barrel/thermal shield region. The
PCA/PSF measurements looked at surveillance capsule effects and the fluence variation within the vessel itself. The NESDIP
measurements overlap the PCA/PSF measurements and extend into the cavity behind the vessel. Investigations of axial streaming
in the cavity were also conducted in NESDIP.
6.1.2 The VENUS Benchmark:
6.1.2.1 The special benchmark field was developed at the VENUS Critical Facility CEN/SCK Laboratories, Belgium (22-8, 3, 4,
5, 6, 7, 8). The facility could mock up PWR fuel geometries to investigate the fluence rate distributions in regions affected by the
deviations from cylindrical symmetry. In addition, measurements on the VENUS fuel investigated the edge effects on power
produced by individual pins at the outside of the fuel region and thus better established the neutron source. These data provided
verification of both the flux magnitude and the azimuthal flux shape. The mock up included a simulated core barrel and thermal
shield.
6.1.2.2 There were several phases to the VENUS program. The first PV mockup configuration studies (VENUS-I) provided a link
between the PCA and PSF tests and the actual environments of LWR power plants. Indeed for actual power plants, the azimuthal
variation of the power distribution determined largely by complex stair-step-shaped core peripheries and by the core-boundary fuel
power distributions could not be ignored, otherwise the calculations could contain undetected biases. Such biases could be further
exacerbated by the use of low-leakage fuel-management schemes.
6.1.2.3 A second configuration, VENUS-2, contained a plutonium-fueled zone at the periphery of the core (to simulate burned
fuel), and its objective was to investigate how much the fast neutron fluence is affected by such a core loading, and if changes in
calculational modeling are necessary to account for any effects. The VENUS facility could also provide data to be used in
validation of other sources asymmetries, such as those due to loading of absorber pins or dummy fuel rods in external assemblies
to limit neutron leakage.
6.1.3 The PCA/PSF Benchmark:
6.1.3.1 The task of developing benchmark fields to meet surveillance dosimetry needs began with the construction, adjacent to the
Oak Ridge National Laboratory (ORNL) Pool Critical Assembly (PCA), of a full-scale-section mockup of a pressure vessel wall
in which passive and active dosimetry measurements (including neutron spectroscopy) could be made both outside and within the
1 1 3
steel mockup (9, 10, 20). Measurement positions corresponding to the ⁄4, ⁄2 , and ⁄4 thicknesses of the pressure vessel were
provided. A simulated surveillance capsule was added to the mockup also. Extensive measurements and calculations provided
sufficient characterization of the PCA benchmark experiment so that it was used for a blind test of neutron transport calculations
(9).
6.1.3.2 The PCA benchmark also served as the critical facility for a higher fluence model of the PCA built at the Pool Side Facility
(PSF) of the 30 MW Oak Ridge Research Reactor (ORR). The PSF made it possible to perform simultaneous dosimetry and
metallurgical irradiations at the simulated surveillance capsule position and positions within the vessel wall. Such measurements
within the vessel wall are not possible in an operating power reactor. The PSF measurements consisted of a startup experiment to
confirm similarity with the PCA results, a long-term vessel wall irradiation with extensive dosimetry contained in capsules with
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dosimetry specimens, and three additional experiments to investigate surveillance capsule effects. The PSF irradiation facility
consisting of the pressure vessel simulator was identified as the Simulated Dosimetry Measurement Facility (SDMF). The SDMF
irradiations were carried out at high-flux with the Oak Ridge Reactor at 30 MW in a series of seven experiments; refer to Appendix
A of reference 13 for the identification of each of these experiments and reference 15 for additional summary commentary on the
SDMF Experiments 1, 2, 3 and 4.
6.1.3.3 The SDMF-1 Startup Experiment, with dosimetry in dummy surveillance capsules in place of the instrumented ones, was
performed prior to the metallurgical irradiation to determine accurately the irradiation times needed to reach the target fluence. A
set of calculations was performed to account for 52 different core loadings and their associated irradiation histories. Calculations
were performed for each of three exposures: two surveillance capsules (SSC-1 and SSC-2) and a pressure vessel capsule.
Comparisons of the ORNL-calculated end-of-life dosimeter activities with measurements indicated agreement, generally within
1 1 3
15 % for the first surveillance capsule, 5 % for the second capsule, and 10 % for the three locations ( ⁄4 T, ⁄2 T, and ⁄4 T) in the
pressure vessel capsule (20).
6.1.3.4 NUREG/CR-3320, Vol.Vol 2 (12) provides documentation of the SDMF-1 Experiment and the results of dosimetry
measurements and studies by the LWR-PV-SDIP participants. The following laboratories participated in radiometric analyses of
the dosimeters: HEDL; ORNL; CEN/SCK (Mol); KFA (Julich); Harwell (England - counting for Rolls Royce Assoc. Ltd.); PTB
(Federal Republic of Germany); and Petten (Netherlands). NBS (presently known as NIST) Certified Fluence Standards were
supplied.
6.1.3.5 The results of the SDMF-1, SDMF-2, and SDMF-3 experiments are primarily based on radiometric sensor measurements.
The SDMF-4 experiment provided benchmark referencing data for the full complement of dosimetry sensors (radiometric, solid
state track recorders, helium accumulation fluence monitors, and damage monitors) which were under development and testing for
PWR and BWR surveillance program applications (15). Therefore, the SDMF-4 measured results are particularly appropriate for
benchmarking the methodology, nuclear data, and accuracy of derived neutron exposure parameter for surveillance applications.
6.1.3.6 The later SDMF experiments were specialized geometry experiments to study the effects on dosimeter response caused by
placement of the surveillance capsules in the water environment of the reactor downcomer region.
6.1.4 The NESDIP Benchmark—The NESTOR Shielding and Dosimetry Improvement Program (NESDIP) was started in 1982
(1616-18, 17, 18). NESDIP experiments have been divided into three phases, the third of which is simulation of actual commercial
LWR cavity configurations in accord with cooperative interests of the NRC and US utilities and reactor vendors (19). The emphasis
was on an internal study of the accuracy of transport theory methods, S and Monte Carlo methods, for predicting neutron
N
penetration and attenuation for the radial shield and cavity region of LWRs.
6.1.5 Other Benchmarks—Other benchmarks exist which may be used for comparisons for special geometries or for other reactor
types. These benchmarks include those described in the benchmark referencing standard (Guide E2005). Additional benchmarks
that may be applicable include the DOMPAC benchmark (21, 22), the OSIRIS benchmark (23, 24), the LR-0/VVER440
benchmark (25, 26), the TAPIRO source reactor benchmark (27), the KORPUS benchmark (28), the concrete benchmark (29), and
the KUCA/KUR/UTR-KINKI benchmarks (30, 31)).
6.2 Benchmarks at Power Reactor Facilities:
6.2.1 In parallel with the PV mockup experiments were efforts in the Arkansas Power and Light Reactor ANO-1 to initiate
ex-vessel cavity dosimetry as a supplement or replacement for vessel monitoring dosimetry in the surveillance capsule (32). This
led to benchmarking, by LWR-PV-SDIP of cavity dosimetry in special experiments in the H.B. Robinson nuclear power reactor
(33, 34) as well as a number of others (35).
6.2.2 The H.B. Robinson measurements have the advantage that simultaneous dosimetry results were obtained from a dummy
surveillance capsule and from ex-vessel capsules irradiated during a single reactor cycle. Thus direct comparisons may be made
with calculations on both sides of the reactor vessel.
6.3 Specific Plant Measurements:
6.3.1 The use of actual plant measurements to obtain fluence results is covered in Practice E1006. However, these results are seen
in the benchmark context as part of the overall benchmarking process to obtain the evaluated plant specific fluence.
6.3.1.1 A large body of data, including both surveillance capsule and ex-vessel dosimetry measurements, has been obtained.
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