ASTM E2956-23

Standard Guide for Monitoring the Neutron Exposure of LWR Reactor Pressure Vessels

SIGNIFICANCE AND USE
4.1 Regulatory Requirements—The USA Code of Federal Regulations (10CFR Part 50, Appendix H) requires the implementation of a reactor vessel materials surveillance program for all operating LWRs. Other countries have similar regulations. The purpose of the program is to (1) monitor changes in the fracture toughness properties of ferritic materials in the reactor vessel beltline6 resulting from exposure to neutron irradiation and the thermal environment, and (2) make use of the data obtained from surveillance programs to determine the conditions under which the vessel can be operated with adequate margins of safety throughout its service life. Practice E185, derived mechanical property data, and (r, θ, z) physics-dosimetry data (derived from the calculations and reactor cavity and surveillance capsule measurements (1) using physics-dosimetry standards) can be used together with information in Guide E900 and Refs. 4, 11-18 to provide a relation between property degradation and neutron exposure, commonly called a “trend curve.” To obtain this trend curve at all points in the pressure vessel wall requires that the selected trend curve be used together with the appropriate (r, θ, z) neutron field information derived by use of this guide to accomplish the necessary interpolations and extrapolations in space and time.  
4.2 Neutron Field Characterization—The tasks required to satisfy the second part of the objective of 4.1 are complex and are summarized in Practice E853. In doing this, it is necessary to describe the neutron field at selected (r, θ, z) points within the pressure vessel wall. The description can be either time dependent or time averaged over the reactor service period of interest. This description can best be obtained by combining neutron transport calculations with plant measurements such as reactor cavity (ex-vessel) and surveillance capsule or RPV cladding (in-vessel) measurements, benchmark irradiations of dosimeter sensor materials, and knowledge of th...
SCOPE
1.1 This guide establishes the means and frequency of monitoring the neutron exposure of the LWR reactor pressure vessel throughout its operating life.  
1.2 The physics-dosimetry relationships determined from this guide may be used to estimate reactor pressure vessel damage through the application of Practice E693 and Guide E900, using fast neutron fluence (E > 1.0 MeV and E > 0.1 MeV), displacements per atom (dpa), or damage-function-correlated exposure parameters as independent exposure variables. Supporting the application of these standards are the E853, E944, E1005, and E1018 standards, identified in 2.1.  
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, health, and environmental 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.

General Information

• Status: Published
• Publication Date: 31-Aug-2023
• ICS: 27.120.10 - Reactor engineering
• Technical Committee: E10 - Nuclear Technology and Applications
• Drafting Committee: E10.05 - Nuclear Radiation Metrology

Relations

• Refers: ASTM E1018-20e1 - Standard Guide for Application of ASTM Evaluated Cross Section Data File
• Effective Date: 01-Mar-2020
• Refers: ASTM E1018-20 - Standard Guide for Application of ASTM Evaluated Cross Section Data File
• Effective Date: 01-Mar-2020
• Refers: ASTM E944-19 - Standard Guide for Application of Neutron Spectrum Adjustment Methods in Reactor Surveillance
• Effective Date: 01-Oct-2019
• Refers: ASTM E2215-19 - Standard Practice for Evaluation of Surveillance Capsules from Light-Water Moderated Nuclear Power Reactor Vessels
• Effective Date: 01-Jun-2019
• Refers: ASTM E2215-18 - Standard Practice for Evaluation of Surveillance Capsules from Light-Water Moderated Nuclear Power Reactor Vessels
• Effective Date: 01-Aug-2018
• Refers: ASTM E844-18 - Standard Guide for Sensor Set Design and Irradiation for Reactor Surveillance
• Effective Date: 01-Jun-2018
• Refers: ASTM E170-17 - Standard Terminology Relating to Radiation Measurements and Dosimetry
• Effective Date: 01-Jun-2017
• Refers: ASTM E2215-16 - Standard Practice for Evaluation of Surveillance Capsules from Light-Water Moderated Nuclear Power Reactor Vessels
• Effective Date: 01-Dec-2016
• Refers: ASTM E170-16a - Standard Terminology Relating to Radiation Measurements and Dosimetry
• Effective Date: 01-Oct-2016
• Refers: ASTM E170-16 - Standard Terminology Relating to Radiation Measurements and Dosimetry
• Effective Date: 15-Feb-2016
• Refers: ASTM E170-15a - Standard Terminology Relating to Radiation Measurements and Dosimetry
• Effective Date: 01-Sep-2015
• Refers: ASTM E1005-15 - Standard Test Method for Application and Analysis of Radiometric Monitors for Reactor Vessel Surveillance
• Effective Date: 01-Jul-2015
• Refers: ASTM E185-15e1 - Standard Practice for Design of Surveillance Programs for Light-Water Moderated Nuclear Power Reactor Vessels
• Effective Date: 01-Jun-2015
• Refers: ASTM E185-15 - Standard Practice for Design of Surveillance Programs for Light-Water Moderated Nuclear Power Reactor Vessels
• Effective Date: 01-Jun-2015
• Refers: ASTM E2215-15 - Standard Practice for Evaluation of Surveillance Capsules from Light-Water Moderated Nuclear Power Reactor Vessels
• Effective Date: 01-Jun-2015

Overview

ASTM E2956-23: Standard Guide for Monitoring the Neutron Exposure of LWR Reactor Pressure Vessels provides essential guidance for nuclear power plant operators and engineers responsible for the integrity of Light Water Reactor (LWR) pressure vessels. As regulatory frameworks worldwide mandate reactor vessel materials surveillance programs, this ASTM standard offers uniform methods to monitor neutron exposure, evaluate pressure vessel material degradation, and ensure safe long-term reactor operation.

Neutron irradiation and thermal effects gradually reduce the fracture toughness of reactor vessel materials, impacting both condition monitoring and predictive maintenance. Effective monitoring of neutron exposure throughout a reactor's operating life, including periods extending to 60 or 80 years, is critical for regulatory compliance and plant safety.

Key Topics

• Regulatory Requirements: Ensures alignment with US Code of Federal Regulations (10CFR Part 50, Appendix H) and similar international directives, providing a framework for meeting mandatory surveillance needs.
• Neutron Exposure Monitoring Techniques: Offers guidance on in-vessel and ex-vessel neutron dosimetry, use of surveillance capsules, and sampling of reactor pressure vessel (RPV) cladding to track neutron flux and accumulated fluence.
• Data Analysis and Interpretation: Recommends combined use of mechanical property data, neutron field characterization, and physics-dosimetry standards to develop "trend curves" correlating material degradation to neutron exposure.
• Frequency of Assessment: Advises plant-specific approaches to determine monitoring frequency, based on operational changes, material exposure margins, and dosimetry half-lives.
• Analytical Procedures: Covers validated computational methods, including neutron transport and Monte Carlo simulations, benchmarking requirements, and the adjustment of calculated fields with physical measurements.
• Uncertainty Management: Addresses estimation and minimization of uncertainty in neutron exposure projections through comprehensive data integration and spatial analysis.

Applications

ASTM E2956-23 is crucial for:

• Nuclear Power Plant Operators: Supporting extended operating licenses by demonstrating ongoing reactor vessel integrity and compliance with safety criteria.
• Regulatory Compliance: Meeting national and international standards for reactor pressure vessel surveillance and material property trending over time.
• Maintenance Planning: Informing aging management programs and preventive maintenance strategies based on accurate monitoring of neutron-induced degradation.
• Plant Modifications and Upgrades: Providing baseline and periodic assessments before and after operational changes, such as fuel cycle adjustments or power uprates, ensuring neutron exposure remains within safe limits.
• Data-Driven Decisions: Enabling the use of surveillance data to refine neutron field models, improve operational safety margins, and extend the safe use of expensive reactor components.
• Design and Construction: Informing material selection and dosimetry provisions in new reactors for prospective exposure tracking and retrospective analysis.

Related Standards

This guide references and works in conjunction with several key industry standards, ensuring a comprehensive approach to neutron exposure monitoring:

• ASTM E185 – Practice for Design of Surveillance Programs for LWR Nuclear Reactor Vessels
• ASTM E853 – Practice for Analysis and Interpretation of LWR Surveillance Neutron Exposure Results
• ASTM E693 – Practice for Characterizing Neutron Exposures in Iron and Low Alloy Steels
• ASTM E900 – Guide for Predicting Radiation-Induced Transition Temperature Shift in Reactor Vessel Materials
• ASTM E482, E844, E944, E1005, E1018 – Addressing aspects from neutron transport calculations to sensor design and dosimetry data analysis
• ASME Boiler and Pressure Vessel Code – Sections III and XI, for fracture mechanics and vessel integrity analysis
• Code of Federal Regulations, 10CFR50, Appendix H – Reactor Vessel Material Surveillance Program Requirements

Summary

ASTM E2956-23 delivers a robust, practical guide for monitoring the neutron exposure of LWR reactor pressure vessels. Its application ensures the reliability, safety, and regulatory compliance of nuclear power plants, underpinning both long-term operational confidence and public protection against radiation risks.

Standards Content (Sample)

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.
Designation: E2956 − 23
Standard Guide for
Monitoring the Neutron Exposure of LWR Reactor Pressure
Vessels
This standard is issued under the fixed designation E2956; 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.
INTRODUCTION
Light Water Reactor (LWR) power plant safety analysis reports and subsequent neutron exposure
parameter calculations for the reactor pressure vessel (RPV) wall and critical welds need to be verified
using modern codes and information from surveillance dosimetry. The location of critical welds
relative to the axial and azimuthal fluence rate map should be taken into account, as well as changes
in fuel loading during periods when surveillance capsules are exposed and beyond to the end of the
reactor’s operating license. For many reactors today this interval is 60 years. In the nuclear industry,
there is active consideration and evaluation of operating intervals of 80 years. Most reactor
surveillance programs were designed based on the guidance of Practice E185 with an operating life
of 40 years. The Practice E185 surveillance programs are designed to select and irradiate the RPV
material test specimens. The dosimetry in the surveillance capsule is there primarily to measure the
neutron fluence to which the capsule’s material specimens have been exposed.
In addition, those programs were based on the operating assumptions in place at the time; typically
annual out-in core loading patterns and base load operation at a fixed reactor power level. Reactor
operations have evolved so that low-leakage core loading patterns (L P) are the norm as are 18 month
and 24 month fuel cycles and reactor power up-ratings of up to 20 %. Many reactors have now
installed flux suppression features such as natural uranium fuel rods, full or part-length hafnium or
B C rods, or stainless steel rods to minimize the neutron exposure of critical areas of the RPV. Such
developments increase the need to comprehensively monitor the RPV accrued fluence through the
extended operation period.
This guide is intended to be used together with other Standards to provide best estimates of the
neutron exposure and exposure rate (together with uncertainties) at positions at the inner diameter and
within the pressure vessel wall of a light water reactor. Also provided will be estimates of gamma-ray
exposure and exposure rates to interpret dosimetry sensor photo-reaction and other gamma-ray
induced effects. Information used to make these estimates is obtained from coupled neutron-gamma
ray transport calculations and from neutron and gamma-ray sensors located in surveillance positions
on the core side of the vessel and in the reactor cavity outside the vessel wall (1). Benchmark field
irradiations of similar monitors also provide valuable information used in the verification of the
accuracy of the calculations (1).
Knowledge of the time-dependent relationship between exposure parameters at surveillance
locations and selected (r, θ, z) locations within the pressure vessel wall is required to allow
determination of the time-dependent radiation damage to the RPV. The time dependency must be
known to allow proper accounting for complications due to burn-up, as well as changes in core loading
configurations (2-5). An estimate of the uncertainty in the neutron exposure parameter values at
selected (r, θ, z) points in the vessel wall (1) is also needed to place an upper bound on the allowable
operating lifetime of the reactor vessel without remedial action (6-9). (See Guide E509.)
Copyright © ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959. United States
E2956 − 23
1. Scope metric Monitors for Reactor Vessel Surveillance
E1018 Guide for Application of ASTM Evaluated Cross
1.1 This guide establishes the means and frequency of
Section Data File
monitoring the neutron exposure of the LWR reactor pressure
E2005 Guide for Benchmark Testing of Reactor Dosimetry
vessel throughout its operating life.
in Standard and Reference Neutron Fields
1.2 The physics-dosimetry relationships determined from
E2006 Guide for Benchmark Testing of Light Water Reactor
this guide may be used to estimate reactor pressure vessel
Calculations
damage through the application of Practice E693 and Guide
E2215 Practice for Evaluation of Surveillance Capsules
E900, using fast neutron fluence (E > 1.0 MeV and E > 0.1
from Light-Water Moderated Nuclear Power Reactor Ves-
MeV), displacements per atom (dpa), or damage-function-
sels
correlated exposure parameters as independent exposure vari-
2.2 American Society of Mechanical Engineers Standard:
ables. Supporting the application of these standards are the
Boiler and Pressure Vessel Code, Sections III and XI
E853, E944, E1005, and E1018 standards, identified in 2.1.
2.3 Nuclear Regulatory Document:
1.3 This standard does not purport to address all of the
Code of Federal Regulations, Chapter 10, Part 50, Appendix
safety concerns, if any, associated with its use. It is the
A – “General Design Criteria for Nuclear Power Plants,”
responsibility of the user of this standard to establish appro- Appendix G – “Fracture Toughness Requirements,” and
priate safety, health, and environmental practices and deter-
Appendix H – Reactor Vessel Material Surveillance Pro-
mine the applicability of regulatory limitations prior to use. gram Requirements”
1.4 This international standard was developed in accor-
3. Terminology
dance with internationally recognized principles on standard-
ization established in the Decision on Principles for the
3.1 Definitions for terms used in this guide are found in
Development of International Standards, Guides and Recom-
Terminology E170.
mendations issued by the World Trade Organization Technical
4. Significance and Use
Barriers to Trade (TBT) Committee.
4.1 Regulatory Requirements—The USA Code of Federal
2. Referenced Documents
Regulations (10CFR Part 50, Appendix H) requires the imple-
mentation of a reactor vessel materials surveillance program
2.1 ASTM Standards:
for all operating LWRs. Other countries have similar regula-
E170 Terminology Relating to Radiation Measurements and
tions. The purpose of the program is to (1) monitor changes in
Dosimetry
the fracture toughness properties of ferritic materials in the
E185 Practice for Design of Surveillance Programs for
reactor vessel beltline resulting from exposure to neutron
Light-Water Moderated Nuclear Power Reactor Vessels
irradiation and the thermal environment, and (2) make use of
E482 Guide for Application of Neutron Transport Methods
the data obtained from surveillance programs to determine the
for Reactor Vessel Surveillance
conditions under which the vessel can be operated with
E509 Guide for In-Service Annealing of Light-Water Mod-
adequate margins of safety throughout its service life. Practice
erated Nuclear Reactor Vessels
E185, derived mechanical property data, and (r, θ, z) physics-
E693 Practice for Characterizing Neutron Exposures in Iron
dosimetry data (derived from the calculations and reactor
and Low Alloy Steels in Terms of Displacements Per
cavity and surveillance capsule measurements (1) using
Atom (DPA)
physics-dosimetry standards) can be used together with infor-
E844 Guide for Sensor Set Design and Irradiation for
mation in Guide E900 and Refs. 4, 11-18 to provide a relation
Reactor Surveillance
between property degradation and neutron exposure, com-
E853 Practice for Analysis and Interpretation of Light-Water
monly called a “trend curve.” To obtain this trend curve at all
Reactor Surveillance Neutron Exposure Results
points in the pressure vessel wall requires that the selected
E900 Guide for Predicting Radiation-Induced Transition
trend curve be used together with the appropriate (r, θ, z)
Temperature Shift in Reactor Vessel Materials
neutron field information derived by use of this guide to
E944 Guide for Application of Neutron Spectrum Adjust-
accomplish the necessary interpolations and extrapolations in
ment Methods in Reactor Surveillance
space and time.
E1005 Test Method for Application and Analysis of Radio-
Available from American Society of Mechanical Engineers (ASME), ASME
This guide is under the jurisdiction of ASTM Committee E10 on Nuclear International Headquarters, Two Park Ave., New York, NY 10016-5990, http://
Technology and Applications and is the direct responsibility of Subcommittee www.asme.org.
E10.05 on Nuclear Radiation Metrology. Available from U.S. Government Printing Office Superintendent of Documents,
Current edition approved Sept. 1, 2023. Published September 2023. Originally 732 N. Capitol St., NW, Mail Stop: SDE, Washington, DC 20401, http://
approved in 2014. Last previous edition approved in 2021 as E2956 – 21. DOI: www.access.gpo.gov.
10.1520/E2956-23. Per USNRC Regulatory Issue Summary 2014-11 (10), the reactor vessel
The boldface numbers in parentheses refer to the list of references appended to beltline is defined as those portions of the RPV where the accumulated neutron
17 –2
this guide. fluence (E > 1.0 MeV) at the end of reactor operation will exceed 10 cm . The
For referenced ASTM standards, visit the ASTM website, www.astm.org, or reactor vessel extended beltline is a term commonly used to refer to materials
contact ASTM Customer Service at [e-mail]. For Annual Book of ASTM located outside of the region opposite the active core height that are also expected
Standards volume information, refer to the standard’s Document Summary page on to accumulate neutron fluence (E > 1.0 MeV) at the end of reactor operation
17 –2
the ASTM website. exceeding 10 cm .
E2956 − 23
4.2 Neutron Field Characterization—The tasks required to within the reactor, and near the vessel wall; that is, at the
satisfy the second part of the objective of 4.1 are complex and surveillance capsule locations (1). In actual practice, the
are summarized in Practice E853. In doing this, it is necessary surveillance capsules may be located within the reactor at an
to describe the neutron field at selected (r, θ, z) points within azimuthal position that differs from that associated with the
the pressure vessel wall. The description can be either time
maximum neutron exposure (or that differs from the azimuthal
dependent or time averaged over the reactor service period of and axial location of the assumed flaw); and at a radial position
interest. This description can best be obtained by combining
a few centimeters or more from the flaw and the pressure vessel
neutron transport calculations with plant measurements such as wall (4, 5). Although the surveillance capsule dosimetry does
reactor cavity (ex-vessel) and surveillance capsule or RPV
provide points for normalization of the neutron physics trans-
cladding (in-vessel) measurements, benchmark irradiations of
port calculations, it is still necessary to use analytical methods
dosimeter sensor materials, and knowledge of the spatial core
that provide an accurate representation of the spatial variation
power distribution, including the time dependence. Because
(axial, radial and azimuthal) of the neutron fluence (refer to
core power distributions change with time, reactor cavity or
Guide E482). It is also necessary to use other measurements to
surveillance capsule measurements obtained early in plant life
confirm the spatial distribution of RPV neutron exposure.
may not be representative of long-term reactor operation.
4.5.2 Given that surveillance capsules are located radially
Therefore, a simple normalization of neutron transport calcu-
closer to the core than the surface of the RPV, they may be
lations to dosimetry data from a given capsule is unlikely to
shifted azimuthally away from the peak exposure location in
give a satisfactory solution to the problem over the full reactor
order to limit the magnitude of the surveillance capsule lead
lifetime. Guide E482 and Guide E944 provide detailed infor-
factor. The lead factor is defined as the ratio of the fast neutron
mation related to the characterization of the neutron field for
fluence at the center of the surveillance capsule to the peak fast
BWR and PWR power plants.
neutron fluence at the clad–base metal interface of the RPV.
4.3 Fracture Mechanics Analysis—Currently, operating One adverse effect of this azimuthal shift away from the peak
limitations for normal heat up and cool down transients
is that the surveillance capsule dosimetry does not “see” the
imposed on the reactor pressure vessel are based on the fracture
part of the core that produces the peak exposure of the reactor
mechanics techniques outlined in the ASME Boiler and Pres-
vessel. As a result, the surveillance capsule is unable to monitor
sure Vessel Code. This code requires the assumption of the
the effect of changes in the core power distribution that are
presence of a surface flaw of depth equal to one fourth of the
made to reduce the peak RPV neutron exposure. Another
pressure vessel thickness. In addition, the fracture mechanics
adverse effect is that with larger lead factors, the capsules are
analysis of accident-induced transients (Pressurized Thermal
rapidly exposed to a high neutron fluence. For example, with a
Shock, (PTS)) may involve evaluating the effect of flaws of
lead factor of five, a surveillance capsule will receive an
varying depth within the vessel wall (4). Thus, information is
exposure in as little as twelve years that is equivalent to what
required regarding the distribution of neutron exposure and the
the reactor pressure vessel peak may see in 60 years of
corresponding radiation damage within the pressure vessel,
operation. Practices E185 and E2215 suggest not exceeding
both in space and time (4). In this regard, Practice E185
twice the maximum design fluence (MDF) or twice the
provides guidelines for designing a minimum surveillance
end-of-license fluence (EOLF). In this example, this would
program, selecting materials, and evaluating metallurgical
require withdrawing any remaining surveillance capsules after
specimen test results for BWR and PWR power plants. Practice
24 years of operation. Thus, without taking other steps, the
E2215 covers the evaluation of test specimens and dosimetry
reactor would be operated for the remaining 36 years (of a 60
from LWR surveillance capsules.
year life) with no dosimetry present.
4.4 Neutron Spectral Effects and DPA—Analysis of the
4.5.3 New or replacement surveillance capsules should
neutron fields of operating power reactors has shown that the
recognize and correct operating deficiencies by using improved
neutron spectral shape changes with radial depth into the
capsule dosimetry. For example, for one class of PWR, the
pressure vessel wall (2, 3). The ratio of dpa/ϕt (where ϕ is the
copper wire is cadmium shielded to minimize interference
fast (E > 1.0 MeV) neutron fluence rate and t is the time that
from trace amounts of cobalt. In about one third of the
the material was exposed to an average fluence rate) changes
measurements the copper has become incorporated into the
by factors of the order of 2.0/1.0 in traversing from the inner to
cadmium preventing separation and further processing. A
the outer radius. Although dpa, since it includes a more detailed
simple solution to this problem is to use stainless steel
modeling of the displacement phenomenon, should theoreti-
hypodermic tubing to contain and separate the radiometric
cally provide a better correlation with property degradation
monitor wire inside the cadmium tubing. Example dimensions
than fluence (E > 1.0 MeV) (1, 19), this topic is still
include: Typical radiometric monitor wire outside diameter =
controversial and the available experimental data does not
0.020 in. (0.5 mm). Typical 19 gauge stainless steel tubing is
provide clear guidance (19, 20). Thus it is recommended to
0.042 in. outside diameter by 0.027 in. inside diameter, 0.008
calculate and report both quantities; see Practice E853 and
in. wall thickness. Typical cadmium tubing is 0.090 in. outside
Practice E693.
diameter by 0.050 in. inside diameter, 0.020 in. wall thickness.
4.5 In-Vessel Surveillance Programs: 4.5.4 Guide E844 states that radionuclides with half-lives
4.5.1 The neutron dosimetry monitors used in reactor vessel that are short compared to the irradiation duration should not
surveillance capsules provide measurements of the neutron be used. For one class of BWR reactor, the surveillance capsule
fluence and fluence rate at single points on the core midplane dosimetry is minimal; consisting of an iron wire and a copper
E2956 − 23
wire (sometimes also a nickel wire). This dosimetry is not generally limited to PWRs due to the extensive structure (jet
suitable for longer irradiations as the “memory” of the activa- pumps, etc.) blocking general access to the RPV cladding of
many BWRs. It may be possible to take a more limited set of
tion products is too short to measure the accumulated fluence.
For example, for the iron (n,p) activation product, Mn, the samples from the cladding of a BWR RPV.
half-life is 312 d. For the copper (n,α) activation product, Co,
4.5.6 The design and manufacture of new reactor pressure
the half-life is 5.27 a. After three half-lives the remaining
vessels should consider using one of the stainless steels or
activity is on the same order as the counting statistics. The
Inconel alloys that contains niobium for the purpose of
result is that the iron wire has “forgotten” everything that has
cladding the inner surface of the vessel. This would result in a
happened more than two cycles ago and the copper wire has
designed-in retrospective dosimetry system that would capture
forgotten everything that has happened more than eight cycles
neutron exposure data from reactor startup.
ago. This assumes 24-month-long fuel cycles. The copper (n,α)
4.6 Ex-Vessel Surveillance Program:
reaction is induced by high energy neutrons and that at a BWR
4.6.1 Ex-vessel neutron dosimetry (EVND) has also been in
surveillance capsule position only 1 % to 3 % of the fast (E >
wide scale application in nuclear reactors for over 30 years (28,
1.0 MeV) neutrons are of high enough energy. This limits the
29, 31, 33, 35, 37-97). The main advantages of EVND are the
value of the copper wire as a neutron fluence monitor. In order
relative simplicity and the relatively low cost of the dosimetry
to monitor the neutron exposure of the RPV other dosimetry is
system. Removal and replacement of irradiated dosimetry
needed. Installation of ex-vessel neutron dosimetry is the most
takes little time. Typical installations have dosimetry that spans
reasonable and cost-effective option.
the active core height and continues to cover the extended
4.5.5 The neutron fluence calculation on the RPV inner
beltline region of the RPV. Installation of dosimetry at multiple
surface can be further verified by means of analyzing small
angles allows full octant coverage (for octant symmetric
samples of the irradiated stainless steel RPV cladding. Analyz-
cores). Some EVND installations include multiple measure-
ing RPV cladding samples has been a well-established practice
ments at symmetric azimuthal angles to confirm symmetry in
for over 30 years (21-36). During the reactor shut down
the azimuthal fluence rate distributions. Asymmetries may
periods, small samples (50 mg to 100 mg) can be machined
result from such things as non-symmetric core power
from the RPV cladding. For retrospective dosimetry purposes
distributions, differences in water temperatures from one loop
54 58 93m
the measured Mn, Co, and Nb activities are used.
to another, or ovality in the as-built dimensions for the reactor
93m
Because of its long half-life, Nb is especially useful for
internals or RPV. Dosimetry capsules typically contain a full
integrating fluence over time periods where accurate neutron
complement of radiometric monitors (refer to Guide E844) to
transport calculations are not available. With sample locations
ensure good spectral coverage and fluence integration.
properly selected, the fast neutron fluence distribution and its
Typically, capsules are connected and supported by stainless
maximum on the RPV inner surface can be determined. By
steel wires or chains, which are, in turn, segmented and
comparison of these data to the dosimetry data of the surveil-
counted to provide axial gradient information.
lance capsules, the lead factor at the time of measurement can
4.6.2 In order to minimize measurement field perturbation,
also be obtained. This technique works best if the cladding
the dosimeter capsules should be made of a neutron-transparent
material is one of the niobium-stabilized stainless steels. Type
material such as aluminum. This also serves to reduce the
347 with 0.7 % niobium is one example. Retrospective
radiation dose rates encountered when removing and replacing
dosimetry has been successfully demonstrated for ordinary
dosimetry. The gradient chains or wires should be a low mass
Type 304 stainless steel cladding with only a trace (~50 ppm)
per linear foot material, again to reduce the dose rates
of niobium (35). It is important that the cladding surface is first
encountered during handling of irradiated dosimetry.
polished to remove radioactive corrosion products before the
4.6.3 An ex-vessel neutron dosimetry system needs to be
sample is machined otherwise competing activity may com-
accurately located with respect to well-known and easily
promise the sample. The tooling used to take these samples
verified reactor features. A reasonable accuracy target is 625
needs to be accurately located relative to reactor landmarks in
mm axially and azimuthally. The effect of the dosimetry
order to know the actual axial and azimuthal locations of the
position error can be estimated by examining the spatial fast
samples. A reasonable accuracy target is 625 mm axially and
neutron fluence rate gradient in the vicinity of the measurement
azimuthally. The effect of the sampling position error can be
point. In general, in the areas where the fast neutron fluence is
estimated by examining the spatial fast neutron fluence rate
the greatest, the gradient tends to be very small; approaching
gradient in the vicinity of the sample point. In general, in the
flat in the case of the axial distribution opposite the middle of
areas where the fast neutron fluence is the greatest, the gradient
the core. At extreme axial positions, well beyond the ends of
tends to be very small; approaching flat in the case of the axial
the core, the gradient is steep. There the positioning error could
distribution opposite the middle of the core. At extreme axial
lead to an estimated fluence error of 620 %. A similar
positions, well beyond the ends of the core, the gradient is
discussion applies to the azimuthal fluence rate gradients.
steep. There the positioning error could lead to an estimated
fluence error of 620 %. A similar discussion applies to the 4.6.4 Ideally, the ex-vessel neutron dosimetry is installed
azimuthal fluence rate gradients. The tooling also needs to be
before reactor startup so that it can provide data over the
designed to completely retain all machined cladding chips and operating lifetime of the reactor. It is recommended that the
to prevent cross-contamination from one sample to another. ex-vessel neutron dosimetry be analyzed before and after
Access to the full extent of azimuthal and axial clad samples is significant plant modifications that would alter the neutron
E2956 − 23
exposure of the reactor vessel. Some examples include switch- cant changes in exposure rates are calculated, new dosimetry
ing from low-leakage core loading patterns back to out-in measurements may be required to ensure exposure estimates
loading patterns (or vice versa), performing a significant are within required accuracy limits. Accurate analysis to relate
(>10 %) uprating of the plant power, adding (or removing) core dosimetry measurements to exposures at critical locations
flux suppression absorbers or dummy fuel rods, or modifying requires fluence calculations for each fuel cycle that the
the reactor internals geometry. The typical dosimetry replace- dosimetry is irradiated and, if shorter half-life dosimeter
ment interval is between one and five 18-month-long fuel reactions are used, may require calculations for several time
cycles (or equivalent intervals for other fuel cycle lengths). intervals within a fuel cycle.
4.6.5 Periodic measurements (either RPV cladding samples 5.2.3 Guide E482 provides detailed guidance related to the
or EVND) serve to confirm neutron fluence projections and calculational determination of neutron exposure for BWR and
help to avoid problems that result from errors in reactor- PWR power plants, and the benchmarking of those calcula-
specific calculational models (98). tions. Test Method E1005 describes procedures for measuring
4.6.6 Calculations of neutron fields in commercial reactors the specific activities of radioactive nuclides produced in
show that the neutron exposure (dpa) at the inner diameter of radiometric monitors by nuclear reactions induced during
the pressure vessel can vary by a factor of three or more as a surveillance exposures for reactor vessels and support struc-
function of azimuthal position (2, 3). Dosimetry monitors in tures.
the reactor cavity outside the reactor pressure vessel are a
5.3 Frequency of Monitoring:
useful tool, therefore, in determining the accuracy of the
5.3.1 The frequency with which neutron exposure monitor-
neutron field calculations at points inside the pressure vessel
ing activities should be performed is dependent upon circum-
wall. Practice E853 recommends the use of ex-vessel reactor
stances unique to each reactor. To determine an appropriate
cavity neutron dosimetry measurements for verification of the
time interval for the re-assessment of neutron exposure
physics transport calculations. The status of benchmark field
projections, consideration should be given to the degree of
and power reactor applications as well as studies of this
consistency of actual power operation with the assumptions
approach are discussed in Refs. 1, 18, 19, 37-40, 99-112.
used in developing the neutron exposure projections, the
anticipated margin remaining between current and projected
5. Neutron Exposure Monitoring
neutron exposure levels, the physical constraints on the half-
5.1 Initial Conditions: lives of the sensor material used in the dosimeters, and
5.1.1 This guide assumes the existence of an analysis of potential ancillary uses for the results of the neutron exposure
record that provides projections of future neutron exposure for calculations (for example, equipment qualification or aging
materials in the reactor vessel that are predicted to experience management of the reactor vessel internal structures). Non-
sufficient neutron radiation damage to be considered in the technical considerations may also be important. Over long
selection of the most limiting material with regard to radiation periods of time, staff turnover may lead to challenges recov-
damage. Projections of future values of neutron exposure are ering the necessary input data, or a loss of organizational focus
then used in subsequent reactor vessel integrity evaluations to may occur on important issues relating to radiation damage and
demonstrate the reactor can be operated safely during normal aging management. It is important that plans be in place to
and off-normal conditions. ensure that all nuclear quality assurance requirements are met,
5.1.2 The operational parameters used to generate projec- including documentation of all inputs to exposure estimates,
and all calculations be carried out and reviewed by qualified
tions of future neutron exposure are frequently subject to
change. A program to periodically re-assess the neutron expo- personnel.
sure should be instituted to confirm that the neutron exposure 5.3.2 For example, consider a reactor that has accrued
projections used in the reactor vessel integrity evaluations sufficient neutron exposure to place it near regulatory screening
remain valid. For highest accuracy, calculations of exposure criteria limits. If continued operation is desired, such a plant
should be made for all past fuel cycles and projected to the may consider implementing fuel changes for the purpose of
future using best estimates of future fuel management. Signifi- reducing reactor vessel neutron exposure. For such a reactor, a
cant changes in the neutron exposure projections may neces- shorter monitoring interval may be appropriate to ensure safe
sitate revisions of the reactor vessel integrity evaluations. operability in-line with analyzed conditions. By contrast, a
When changes in calculated exposure rates are observed, the plant with wide margins between current and projected neutron
differences should be investigated, and the basis of such exposure, that operates with core loading patterns and opera-
differences understood. tional parameters that are highly consistent with the projection
assumptions, may be justified in using a longer monitoring
5.2 Means of Monitoring:
interval.
5.2.1 Neutron exposure monitoring can be achieved by
periodically performing or updating calculations to reflect
6. Supplementary Analytical Procedures
actual plant operating conditions, by collection and analysis of
additional reactor dosimetry measurements to validate calcu- 6.1 Basic Approach—ASTM Practice E853 covers various
lated exposure projections, or both.
aspects of the extrapolation problem. The basic approach is
5.2.2 Long periods of operation without any reactor dosim- that a transport calculation (benchmarked per Guide E482) is to
etry measurements can leave undetected errors in the inputs to be used to supply the neutron field information at the (r, θ, z)
the neutron exposure calculation methodology. When signifi- points in the pressure vessel wall where property deterioration
E2956 − 23
information will be calculated using Guide E900, or other trend surement results are used, the modeling in the reactor cavity
curves (4, 11-18). The dosimetry information obtained from and external shield should be adequate to provide usable
reactor cavity and surveillance capsule measurements and
calculations for the neutron field in the reactor cavity region.
retrospective dosimetry measurements from reactor internals
This requires an attention to mesh size in the ex-vessel region
structures and RPV cladding is to be used to ensure that the
and an accurate representation of the geometry and chemical
transport calculation is valid and to adjust the transport results
makeup of the external shield. Regardless of the method
if needed. The adjustments are to be accomplished using the
chosen, the effects of relevant input parameter variations on the
guidelines presented in Guide E944. Dosimetry from monitors
calculated results should be well understood. Reference param-
in the reactor cavity and surveillance capsules can provide
eter variation studies focused on the reactor cavity and ex-
limits on uncertainties for the calculated neutron field at
tended beltline region are available in Ref (118).
selected (r, θ, z) positions in the reactor pressure vessel wall.
6.2.1.2 Benchmarking—It is not the purpose of this guide to
Time dependence of the core power distribution (due to burnup
dictate the type of transport calculation to be used in the region
within a given cycle, or due to variations in cycle to cycle fuel
between the core and the outer radius of the pressure vessel (or
loading), surveillance capsule perturbation effects, and dosim-
into the biological shield) or the adjustment procedures, but
etry monitor experimental effects must be recognized as
any such calculations or adjustment procedures should be
complications, and these effects must be accounted for in the
adequately benchmarked by calculations of well defined prob-
calculation and adjustment methods chosen (1-6, 11).
lems (for example, PCA Blind Test (100), VENUS (107),
6.2 Spatial Extrapolations:
NESDIP (108), BWR (104, 105), and PWR (1, 37-40, 99). For
6.2.1 Transport Codes—In general, three-dimensional re-
further details on benchmarking refer to Guide E2006 and
sults need to be obtained for the neutron and gamma ray fields
Guide E944. USNRC Regulatory Guide 1.190 (119) also
in the region from the core to the interior of the biological
addresses benchmarking of neutron transport calculations for
shield beyond the pressure vessel. As a minimum, a three-
RPV surveillance in some detail.
dimensional synthesis analysis should be performed using a
6.2.2 Power Distribution—As discussed in Practice E853,
two-dimensional transport code. The transport calculations in
obtain a valid, adequately time dependent, core power distri-
that case are carried out using the following three-dimensional
bution using a diffusion calculation, or a transport calculation
synthesis technique:
(99, 100, 107). Experimental verification of the accuracy of the
ϕ r , z
~ !
results is desirable, but may be difficult to obtain. This is
ϕ~r , θ , z! 5 ϕ~r , θ! (1)
ϕ~r!
especially important for the pin-by-pin power distributions at
the core periphery and the axial power distributions at the ends
where ϕ (r, θ, z) is the synthesized three-dimensional flux
of the core. The uncertainties in the core power distribution
distribution, ϕ (r,θ) is the transport solution in r,θ geometry, ϕ
tend to be the largest in these areas. Fuel assembly geometric
(r,z) is the two-dimensional solution for a cylindrical reactor
model, and ϕ (r) is the one-dimensional solution for a cylin- features also need to be considered in the development and
drical reactor model using the same source per unit height as modeling of the core power distribution. For example, some
that used in the two-dimensional r,θ calculation.
PWR fuel assemblies use low-enrichment axial blankets and
some BWR fuel bundles use several different fuel rod lengths
6.2.1.1 However, for problems and regions of interest where
the transport solution, that is, radiation fields, have a nonsepa- within the bundle.
rable three-dimensional nature (due to the core power
6.2.2.1 Typically, calculations are performed on a fuel
distribution, or reactor internals structures, or away from the
cycle-by-fuel cycle basis rather than using a single power
core midplane), the synthesis technique may reduce the accu-
distribution that is averaged over many fuel cycles. A well-
racy of the results, thus dictating that a full three-dimensional
documented basis should be used for extrapolating core power
method be used. Analysis of the extended beltline, which often
distributions into the future. Extrapolations should be based on
includes RPV nozzles, also dictates a full three-dimensional
best estimate projections of future fuel cycles. One common
approach. An efficient way to carry out large 3D discrete
approach is to average the three most recent core power
ordinates S transport calculations is the use of multiple
n
distributions and to use that for extrapolation. The assumption
processors running in parallel (112-116). Monte Carlo methods
being that a similar core loading strategy will continue to be
are also used and these are traditionally run in parallel
used. This assumption should be revisited whenever new
processing computing environments. Guide E482 should be
measurements or core designs become available.
followed for the calculations and Guide E944 for measured
6.2.2.2 The power distribution should include the assembly-
dosimetry adjustments. If a discrete ordinates method is used,
wise and axial variation of power as well as the finer,
the spatial mesh should be fine enough and the order of angular
pin-by-pin distribution in the peripheral assemblies adjacent to
quadrature should be high enough to ensure a sufficiently
the reactor internals. Details of the initial U enrichment and
accurate solution in all regions of importance. The impact of
the cycle changes in assembly burnup should also be deter-
the selected discretization settings on the discrete ordinates
mined as this is needed in order to define the mix of fissioning
results should be assessed (117). Methods of ensuring that the
235 238 239 240 241
isotopes (for example U, U, Pu, Pu, Pu, and
mesh is sufficiently fine are the province of Guide E482.
Pu) in each fuel assembly. Some BWR fuel bundles use
Similar considerations apply to tallying techniques in Monte
Carlo calculations. If ex-vessel reactor cavity dosimeter mea- multiple U enrichments axially within a given fuel rod.
E2956 − 23
6.2.3 Ex-Core Regions—Perform a transport calculation for limits, a physics-dosimetry adjustment code analysis should be
the neutron field in all ex-core regions, using adequate mod- performed as outlined in 6.2.6.
6.2.6 Physics-Dosimetry Adjustment Code Analysis—Guide
eling of the reactor geometry, and adequate modeling of the
E944 should be used to combine the transport calculation with
ex-vessel region. The biological shield is to be accurately
the dosimeter results. The Guide E944 adjustment procedure
modeled both in terms of geometry (ex-core detector wells,
should be used to indicate whether the dosimeter measure-
support columns, and the presence or absence of a liner plate),
ments and associated uncertainties are consistent with the
and materials including the biological shield composition
transport calculation and with uncertainties implied from
(cement, aggregate, water content, and distribution of reinforc-
benchmark tests of the transport code (PCA, VENUS,
ing steel). The water content in the biological shield will vary
NESDIP, and an appropriate Commercial BWR or PWR; see
over time (120). The energy, angle, and space discretization as
Refs 1, 37-40, 99, 100, 104, 105). Having established the
well as neutron balance should be checked in all regions to
required consistency, the adjusted results of the transport
make sure the calculation has converged, watching in particu-
calculation may be used to calculate the best estimate neutron
lar for spatial oscillations or ray effects in ex-vessel regions.
field at all points in the pressure vessel wall with the uncer-
Monte Carlo calculations should be checked to confirm that
tainty estimates derived from the application of the adjustment
acceptable tally statistics have been achieved.
codes.
6.2.4 Power Plant Dimensions—In all calculations, as-built
6.2.7 Measurement Results—If the calculated neutron field
dimensions should be used. If they are unavailable, docu-
at the measurement location is inconsistent with the experi-
mented logic should be presented to defend the dimensions
mental dosimetry results, an attempt should be made to
used, and the uncertainty in the final results should reflect the
uncover and correct errors in order to obtain consistency.
added uncertainty. The thickness of the reactor pressure vessel
Particular attention should be paid to sensor monitor correction
(RPV) is a key dimension in the analysis of ex-vessel neutron
factors such as capsule perturbation, photo-reactions,
dosimetry. There are two ways in which the accuracy of the
impurities, burn-in / burn-out, and other effects. Discussions of
assumed RPV thickness may be assessed.
how to proceed when calculations and measurements do not
237 54
6.2.4.1 The ratio of the Np fission rate to the Fe(n,p)
agree may be found in Practice E853, especially Section 7.3.
reaction rate in the reactor cavity may be used as a spectral
6.3 Time Extrapolations—In the case where a time averaged
index. This ratio is very sensitive to the thickness of the RPV.
core loading has been used to define the future neutron source
For example, over an RPV thickness range from 100 mm to
term, the fluence or dpa in future years is estimated by
200 mm, the reaction rate ratio increases by nearly a factor of
multiplying by the expected integrated time at full power.
two. Therefore, when the calculated spectral index from a
calculation with an assumed RPV thickness agrees with the
7. Report and Bias of Results
measured spectral index, one can have a high degree of
7.1 As a minimum, the documentation of results should
confidence that the assumed thickness is correct. A difference
include the following information:
in the spectral index can also indicate how much the assumed
7.1.1 A description of the analytical technique used, includ-
RPV thickness is off and in which direction. The calculated
ing a listing of pertinent input parameters that may affect the
spectral index needs to be determined at the same azimuthal
bias of the calculation. For example, if the discrete ordinates
angle as the measurement being compared.
approach is used, specify or reference the source of the
6.2.4.2 RPV pre-service or in-service inspections are usu-
cross-section data, cross-section preparation procedures, en-
ally performed using ultrasonic testing (UT) looking for flaws
ergy group structure, spatial mesh, S order, and P order.
N L
in the material. Usually these are multi-angle scans. However,
Dimensions and material compositions of key structures in-
sometimes a zero degree (normal incidence) scan is performed.
cluded in the model need to be included. Some of this
This UT scan can provide a direct measured thickness for the
information may be proprietary. In that case, the source of the
RPV. With sufficient advance notice, a zero degree scan can be
data used and a general description should be provided.
added to a future ISI program if the spectral index assessment
7.1.2 Information indicating the bias of the analytical ap-
indicates that the design basis RPV thickness is incorrect.
proach in steel-water systems, including the details of bench-
6.2.5 Dosimetry Sensor Analysis—For analysis of any given
mark calculations used to validate the procedures, and data and
set of reactor cavity or surveillance capsule dosimetry sensors,
the bias attained in the benchmark tests.
the integral reactions or reaction rates of the individual sensors,
7.1.3 The calculated total, thermal, epi-thermal (also known
or both, should be calculated, using the results of the transport
as epi-cadmium fluence rate), E > 0.1 MeV, and E > 1.0 MeV
calculation. The measurement and analysis procedures for
neutron fluence rate-fluence values, and energy spectrum at the
individual sensors should be benchmarked for each sensor
surveillance capsule, and any ex-vessel dosimetry locations.
type; refer to Guide E2005. If the calculated and experimental
Also report calculated values of dpa/s and dpa at the same
integral results (C/E ratios) agree to within the required locations.
accuracy (~5 % to 15 %, 1σ being the best attainable, see Ref
7.1.3.1 The location of peak fluence rate-fluence points on
100) expected from the benchmark calibration of the transport
the surface and in the interior of the vessel wall are calculated
code, the transport calculation may be used directly to calculate values that are required for all the above exposure and
the neutron field at all (r, θ, z) points in the pressure vessel
exposure rate parameters, except for the thermal and epither-
wall. If the C/E ratios do not agree within acceptable accuracy mal fluence rates, which generally can be best determined by
E2956 − 23
dosimetry measurements. For some damage analysis studies, results and uncertainties. Methods of extrapolation and inter-
all of the above information is needed (111, 121-125). polation must specifically be delineated. If the transport calcu-
7.1.3.2 At dosimetry measurement locations, gamma ray
lation spatial mesh or tally size is sufficiently fine, interpolation
fluence rate should be estimated to the precision required to
does not introduce significant error.
make necessary photo-reaction corrections. Similarly, gamma
7.1.5 Details must be given relative to the methods used to
ray field parameters (for example, heat generation rates) should
assign uncertainties for calculated values of neutron fluence
be estimated to whatever precision is needed to allow tempera-
rate, fluence, dpa/s, and dpa. Uncertainties for calculated
tur
...

This document is not an ASTM standard and is intended only to provide the user of an ASTM standard an indication of what changes have been made to the previous version. Because
it may not be technically possible to adequately depict all changes accurately, ASTM recommends that users consult prior editions as appropriate. In all cases only the current version
of the standard as published by ASTM is to be considered the official document.
Designation: E2956 − 21 E2956 − 23
Standard Guide for
Monitoring the Neutron Exposure of LWR Reactor Pressure
Vessels
This standard is issued under the fixed designation E2956; 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.
INTRODUCTION
Light Water Reactor (LWR) power plant safety analysis reports and subsequent neutron exposure
parameter calculations for the reactor pressure vessel (RPV) wall and critical welds need to be verified
using modern codes and information from surveillance dosimetry. The location of critical welds
relative to the axial and azimuthal fluence rate map should be taken into account, as well as changes
in fuel loading during periods when surveillance capsules are exposed and beyond to the end of the
reactor’s operating license. For many reactors today this is a 60-year-long interval. interval is 60 years.
In the nuclear industry, there is active consideration and evaluation of an 80-year-long operating
interval. operating intervals of 80 years. Most reactor surveillance programs were designed based on
the guidance of Practice E185 with a 40-year operating life in mind. an operating life of 40 years. The
Practice E185 surveillance programs are designed to select and irradiate the RPV material test
specimens. The dosimetry in the surveillance capsule is there primarily to measure the neutron fluence
to which the capsule’s material specimens have been exposed.
In addition, those programs were based on the operating assumptions in place at the time; typically
annual out-in core loading patterns and base load operation at a fixed reactor power level. Reactor
operations have evolved so that low-leakage core loading patterns (L P) are the norm as are 18- and
24-month-long 18 month and 24 month fuel cycles and reactor power up-ratings of up to 20 %. Many
reactors have now installed flux suppression features such as natural uranium fuel rods, full or
part-length hafnium or B C rods, or stainless steel rods to minimize the neutron exposure of critical
areas of the RPV. Such developments increase the need to comprehensively monitor the RPV accrued
fluence through the extended operation period.
This guide is intended to be used together with other Standards to provide best estimates of the
neutron exposure and exposure rate (together with uncertainties) at positions at the inner diameter and
within the pressure vessel wall of a light water reactor. Also provided will be estimates of gamma-ray
exposure and exposure rates to interpret dosimetry sensor photo-reaction and other gamma-ray
induced effects. Information used to make these estimates is obtained from coupled neutron-gamma
ray transport calculations and from neutron and gamma-ray sensors located in surveillance positions
on the core side of the vessel and in the reactor cavity outside the vessel wall (1). Benchmark field
irradiations of similar monitors also provide valuable information used in the verification of the
accuracy of the calculations (1).
Knowledge of the time-dependent relationship between exposure parameters at surveillance
locations and selected (r, θ, z) locations within the pressure vessel wall is required to allow
determination of the time-dependent radiation damage to the RPV. The time dependency must be
known to allow proper accounting for complications due to burn-up, as well as changes in core loading
This 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 Feb. 1, 2021Sept. 1, 2023. Published March 2021September 2023. Originally approved in 2014. Last previous edition approved in 20142021
as E2956-14.E2956 – 21. DOI: 10.1520/E2956-21.10.1520/E2956-23.
The boldface numbers in parentheses refer to the list of references appended to this guide.
Copyright © ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959. United States
E2956 − 23
configurations (2-5). An estimate of the uncertainty in the neutron exposure parameter values at
selected (r, θ, z) points in the vessel wall (1) is also needed to place an upper bound on the allowable
operating lifetime of the reactor vessel without remedial action (6-9). (See Guide E509.)
1. Scope
1.1 This guide establishes the means and frequency of monitoring the neutron exposure of the LWR reactor pressure vessel
throughout its operating life.
1.2 The physics-dosimetry relationships determined from this guide may be used to estimate reactor pressure vessel damage
through the application of Practice E693 and Guide E900, using fast neutron fluence (E > 1.0 MeV and E > 0.1 MeV),
displacements per atom – dpa, (dpa), or damage-function-correlated exposure parameters as independent exposure variables.
Supporting the application of these standards are the E853, E944, E1005, and E1018 standards, identified in 2.1.
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, health, and environmental 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
E185 Practice for Design of Surveillance Programs for Light-Water Moderated Nuclear Power Reactor Vessels
E482 Guide for Application of Neutron Transport Methods for Reactor Vessel Surveillance
E509 Guide for In-Service Annealing of Light-Water Moderated Nuclear Reactor Vessels
E693 Practice for Characterizing Neutron Exposures in Iron and Low Alloy Steels in Terms of Displacements Per Atom (DPA)
E844 Guide for Sensor Set Design and Irradiation for Reactor Surveillance
E853 Practice for Analysis and Interpretation of Light-Water Reactor Surveillance Neutron Exposure Results
E900 Guide for Predicting Radiation-Induced Transition Temperature Shift in Reactor Vessel Materials
E944 Guide for Application of Neutron Spectrum Adjustment Methods in Reactor Surveillance
E1005 Test Method for Application and Analysis of Radiometric Monitors for Reactor Vessel Surveillance
E1018 Guide for Application of ASTM Evaluated Cross Section Data File
E2005 Guide for Benchmark Testing of Reactor Dosimetry in Standard and Reference Neutron Fields
E2006 Guide for Benchmark Testing of Light Water Reactor Calculations
E2215 Practice for Evaluation of Surveillance Capsules from Light-Water Moderated Nuclear Power Reactor Vessels
2.2 American Society of Mechanical Engineers Standard:
Boiler and Pressure Vessel Code, Sections III and XI
2.3 Nuclear Regulatory Document:
Code of Federal Regulations, Chapter 10, Part 50, Appendix A – “General Design Criteria for Nuclear Power Plants,” Appendix
G – “Fracture Toughness Requirements,” and Appendix H – Reactor Vessel Material Surveillance Program Requirements”
3. Terminology
3.1 Definitions for terms used in this guide are found in Terminology E170.
4. Significance and Use
4.1 Regulatory Requirements—The USA Code of Federal Regulations (10CFR Part 50, Appendix H) requires the implementation
For referenced ASTM standards, visit the ASTM website, www.astm.org, or contact ASTM Customer Service at [e-mail]. For Annual Book of ASTM Standards
volume information, refer to the standard’s Document Summary page on the ASTM website.
Available from American Society of Mechanical Engineers (ASME), ASME International Headquarters, Two Park Ave., New York, NY 10016-5990, http://
www.asme.org.
Available from U.S. Government Printing Office Superintendent of Documents, 732 N. Capitol St., NW, Mail Stop: SDE, Washington, DC 20401, http://
www.access.gpo.gov.
E2956 − 23
of a reactor vessel materials surveillance program for all operating LWRs. Other countries have similar regulations. The purpose
of the program is to ((1)1) monitor changes in the fracture toughness properties of ferritic materials in the reactor vessel beltline
resulting from exposure to neutron irradiation and the thermal environment, and ((2)2) make use of the data obtained from
surveillance programs to determine the conditions under which the vessel can be operated with adequate margins of safety
throughout its service life. Practice E185, derived mechanical property data, and (r, θ, z) physics-dosimetry data (derived from the
calculations and reactor cavity and surveillance capsule measurements (1) using physics-dosimetry standards) can be used together
with information in Guide E900 and Refs. 4, 11-18 to provide a relation between property degradation and neutron exposure,
commonly called a “trend curve.” To obtain this trend curve at all points in the pressure vessel wall requires that the selected trend
curve be used together with the appropriate (r, θ, z) neutron field information derived by use of this guide to accomplish the
necessary interpolations and extrapolations in space and time.
4.2 Neutron Field Characterization—The tasks required to satisfy the second part of the objective of 4.1 are complex and are
summarized in Practice E853. In doing this, it is necessary to describe the neutron field at selected (r, θ, z) points within the
pressure vessel wall. The description can be either time dependent or time averaged over the reactor service period of interest. This
description can best be obtained by combining neutron transport calculations with plant measurements such as reactor cavity
(ex-vessel) and surveillance capsule or RPV cladding (in-vessel) measurements, benchmark irradiations of dosimeter sensor
materials, and knowledge of the spatial core power distribution, including the time dependence. Because core power distributions
change with time, reactor cavity or surveillance capsule measurements obtained early in plant life may not be representative of
long-term reactor operation. Therefore, a simple normalization of neutron transport calculations to dosimetry data from a given
capsule is unlikely to give a satisfactory solution to the problem over the full reactor lifetime. Guide E482 and Guide E944 provide
detailed information related to the characterization of the neutron field for BWR and PWR power plants.
4.3 Fracture Mechanics Analysis—Currently, operating limitations for normal heat up and cool down transients imposed on the
reactor pressure vessel are based on the fracture mechanics techniques outlined in the ASME Boiler and Pressure Vessel Code. This
code requires the assumption of the presence of a surface flaw of depth equal to one fourth of the pressure vessel thickness. In
addition, the fracture mechanics analysis of accident-induced transients (Pressurized Thermal Shock, (PTS)) may involve
evaluating the effect of flaws of varying depth within the vessel wall (4). Thus, information is required regarding the distribution
of neutron exposure and the corresponding radiation damage within the pressure vessel, both in space and time (4). In this regard,
Practice E185 provides guidelines for designing a minimum surveillance program, selecting materials, and evaluating metallurgical
specimen test results for BWR and PWR power plants. Practice E2215 covers the evaluation of test specimens and dosimetry from
LWR surveillance capsules.
4.4 Neutron Spectral Effects and DPA—Analysis of the neutron fields of operating power reactors has shown that the neutron
spectral shape changes with radial depth into the pressure vessel wall (2, 3). The ratio of dpa/ϕt (where ϕ is the fast (E > 1.0 MeV)
neutron fluence rate and t is the time that the material was exposed to an average fluence rate) changes by factors of the order of
2.0/1.0 in traversing from the inner to the outer radius. Although dpa, since it includes a more detailed modeling of the
displacement phenomenon, should theoretically provide a better correlation with property degradation than fluence (E > 1.0 MeV)
(1, 19), this topic is still controversial and the available experimental data does not provide clear guidance (19, 20). Thus it is
recommended to calculate and report both quantities; see Practice E853 and Practice E693.
4.5 In-Vessel Surveillance Programs:
4.5.1 The neutron dosimetry monitors used in reactor vessel surveillance capsules provide measurements of the neutron fluence
and fluence rate at single points on the core midplane within the reactor, and near the vessel wall; that is, at the surveillance capsule
locations (1). In actual practice, the surveillance capsules may be located within the reactor at an azimuthal position that differs
from that associated with the maximum neutron exposure (or that differs from the azimuthal and axial location of the assumed
flaw); and at a radial position a few centimeters or more from the flaw and the pressure vessel wall (4, 5). Although the surveillance
capsule dosimetry does provide points for normalization of the neutron physics transport calculations, it is still necessary to use
analytical methods that provide an accurate representation of the spatial variation (axial, radial and azimuthal) of the neutron
fluence (refer to Guide E482). It is also necessary to use other measurements to confirm the spatial distribution of RPV neutron
exposure.
4.5.2 Given that surveillance capsules are located radially closer to the core than the surface of the RPV, they may be shifted
Per USNRC Regulatory Issue Summary 2014-11 (10), the reactor vessel beltline is defined as those portions of the RPV where the accumulated neutron fluence (E >
17 -2–2
1.0 MeV) at the end of reactor operation will exceed 10 cm . The reactor vessel extended beltline is a term commonly used to refer to materials located outside of the
17 -2–2
region opposite the active core height that are also expected to accumulate neutron fluence (E > 1.0 MeV) at the end of reactor operation exceeding 10 cm .
E2956 − 23
azimuthally away from the peak exposure location in order to limit the magnitude of the surveillance capsule lead factor. The lead
factor is defined as the ratio of the fast neutron fluence at the center of the surveillance capsule to the peak fast neutron fluence
at the clad – base clad–base metal interface of the RPV. One adverse effect of this azimuthal shift away from the peak is that the
surveillance capsule dosimetry does not “see” the part of the core that produces the peak exposure of the reactor vessel. As a result,
the surveillance capsule is unable to monitor the effect of changes in the core power distribution that are made to reduce the peak
RPV neutron exposure. Another adverse effect is that with larger lead factors, the capsules are rapidly exposed to a high neutron
fluence. For example, with a lead factor of five, a surveillance capsule will receive an exposure in as little as 12twelve years that
is equivalent to what the reactor pressure vessel peak may see in 60 years of operation. Practices E185 and E2215 suggest not
exceeding twice the maximum design fluence (MDF) or twice the end-of-license fluence (EOLF). In this example, this would
require withdrawing any remaining surveillance capsules after 24 years of operation. Thus, without taking other steps, the reactor
would be operated for the remaining 36 years (of a 60-year 60 year life) with no dosimetry present.
4.5.3 New or replacement surveillance capsules should recognize and correct operating deficiencies by using improved capsule
dosimetry. For example, for one class of PWR, the copper wire is cadmium shielded to minimize interference from trace amounts
of cobalt. In about one third of the measurements the copper has become incorporated into the cadmium preventing separation and
further processing. A simple solution to this problem is to use stainless steel hypodermic tubing to contain and separate the
radiometric monitor wire inside the cadmium tubing. Example dimensions include: Typical radiometric monitor wire outside
diameter = 0.020 in. (0.5 mm). Typical 19 gauge stainless steel tubing is 0.042 in. outside diameter by 0.027 in. inside diameter,
0.008 in. wall thickness. Typical cadmium tubing is 0.090 in. outside diameter by 0.050 in. inside diameter, 0.020 in. wall
thickness.
4.5.4 PracticeGuide E844 states that radionuclides with half-lives that are short compared to the irradiation duration should not
be used. For one class of BWR reactor, the surveillance capsule dosimetry is minimal; consisting of an iron wire and a copper wire
(sometimes also a nickel wire). This dosimetry is not suitable for longer irradiations as the “memory” of the activation products
is too short to measure the accumulated fluence. For example, for the iron (n,p) activation product, Mn, the half-life is 312 d.
For the copper (n,α) activation product, Co, the half-life is 5.27 y.a. After three half-lives the remaining activity is on the same
order as the counting statistics. The result is that the iron wire has “forgotten” everything that has happened more than two cycles
ago and the copper wire has forgotten everything that has happened more than eight cycles ago. This assumes 24-month-long fuel
cycles. The copper (n,α) reaction is induced by high energy neutrons and that at a BWR surveillance capsule position only 1 %
to 3 % of the fast (E > 1.0 MeV) neutrons are of high enough energy. This limits the value of the copper wire as a neutron fluence
monitor. In order to monitor the neutron exposure of the RPV other dosimetry is needed. Installation of ex-vessel neutron dosimetry
is the most reasonable and cost-effective option.
4.5.5 The neutron fluence calculation on the RPV inner surface can be further verified by means of analyzing small samples of
the irradiated stainless steel RPV cladding. Analyzing RPV cladding samples has been a well-established practice for over 30 years
(21-36). During the reactor shut down periods, small samples (50–100 (50 mg to 100 mg) can be machined from the RPV cladding.
54 58 93m 93m
For retrospective dosimetry purposes the measured Mn, Co, and Nb activities are used. Because of its long half-life, Nb
is especially useful for integrating fluence over time periods where accurate neutron transport calculations are not available. With
sample locations properly selected, the fast neutron fluence distribution and its maximum on the RPV inner surface can be
determined. By comparison of these data to the dosimetry data of the surveillance capsules, the lead factor at the time of
measurement can also be obtained. This technique works best if the cladding material is one of the niobium-stabilized stainless
steels. Type 347 with 0.7 % niobium is one example. Retrospective dosimetry has been successfully demonstrated for ordinary
Type 304 stainless steel cladding with only a trace (~ 50 (~50 ppm) of niobium (35). It is important that the cladding surface is
first polished to remove radioactive corrosion products before the sample is machined otherwise competing activity may
compromise the sample. The tooling used to take these samples needs to be accurately located relative to reactor landmarks in order
to know the actual axial and azimuthal locations of the samples. A reasonable accuracy target is 625 mm axially and azimuthally.
The effect of the sampling position error can be estimated by examining the spatial fast neutron fluence rate gradient in the vicinity
of the sample point. In general, in the areas where the fast neutron fluence is the greatest, the gradient tends to be very small;
approaching flat in the case of the axial distribution opposite the middle of the core. At extreme axial positions, well beyond the
ends of the core, the gradient is steep. There the positioning error could lead to an estimated fluence error of 620 %. A similar
discussion applies to the azimuthal fluence rate gradients. The tooling also needs to be designed to completely retain all machined
cladding chips and to prevent cross-contamination from one sample to another. Access to the full extent of azimuthal and axial clad
samples is generally limited to PWRs due to the extensive structure (jet pumps, etc.) blocking general access to the RPV cladding
of many BWRs. It may be possible to take a more limited set of samples from the cladding of a BWR RPV.
4.5.6 The design and manufacture of new reactor pressure vessels should consider using one of the stainless steels or Inconel
alloys that contains niobium for the purpose of cladding the inner surface of the vessel. This would result in a designed-in
retrospective dosimetry system that would capture neutron exposure data from reactor startup.
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4.6 Ex-Vessel Surveillance Program:
4.6.1 Ex-vessel neutron dosimetry (EVND) has also been in wide scale application in nuclear reactors for over 30 years (28, 29,
31, 33, 35, 37-97). The main advantages of EVND are the relative simplicity and the relatively low cost of the dosimetry system.
Removal and replacement of irradiated dosimetry takes little time. Typical installations have dosimetry that spans the active core
height and continues to cover the extended beltline region of the RPV. Installation of dosimetry at multiple angles allows full octant
coverage (for octant symmetric cores). Some EVND installations include multiple measurements at symmetric azimuthal angles
to confirm symmetry in the azimuthal fluence rate distributions. Asymmetries may result from such things as non-symmetric core
power distributions, differences in water temperatures from one loop to another, or ovality in the as-built dimensions for the reactor
internals or RPV. Dosimetry capsules typically contain a full complement of radiometric monitors (refer to PracticeGuide E844)
to ensure good spectral coverage and fluence integration. Typically, capsules are connected and supported by stainless steel wires
or chains, which are, in turn, segmented and counted to provide axial gradient information.
4.6.2 In order to minimize measurement field perturbation, the dosimeter capsules should be made of a neutron-transparent
material such as aluminum. This also serves to reduce the radiation dose rates encountered when removing and replacing
dosimetry. The gradient chains or wires should be a low mass per linear foot material, again to reduce the dose rates encountered
during handling of irradiated dosimetry.
4.6.3 An ex-vessel neutron dosimetry system needs to be accurately located with respect to well-known and easily verified reactor
features. A reasonable accuracy target is 625 mm axially and azimuthally. The effect of the dosimetry position error can be
estimated by examining the spatial fast neutron fluence rate gradient in the vicinity of the measurement point. In general, in the
areas where the fast neutron fluence is the greatest, the gradient tends to be very small; approaching flat in the case of the axial
distribution opposite the middle of the core. At extreme axial positions, well beyond the ends of the core, the gradient is steep.
There the positioning error could lead to an estimated fluence error of 620 %. A similar discussion applies to the azimuthal fluence
rate gradients.
4.6.4 Ideally, the ex-vessel neutron dosimetry is installed before reactor startup so that it can provide data over the operating
lifetime of the reactor. It is recommended that the ex-vessel neutron dosimetry be analyzed before and after significant plant
modifications that would alter the neutron exposure of the reactor vessel. Some examples include switching from low-leakage core
loading patterns back to out-in loading patterns (or vice versa), performing a significant (>10 %) uprating of the plant power,
adding (or removing) core flux suppression absorbers or dummy fuel rods, or modifying the reactor internals geometry. The typical
dosimetry replacement interval is between one and five 18-month-long fuel cycles (or equivalent intervals for other fuel cycle
lengths).
4.6.5 Periodic measurements (either RPV cladding samples or EVND) serve to confirm neutron fluence projections and help to
avoid problems that result from errors in reactor-specific calculational models (98).
4.6.6 Calculations of neutron fields in commercial reactors show that the neutron exposure (dpa) at the inner diameter of the
pressure vessel can vary by a factor of three or more as a function of azimuthal position (2, 3). Dosimetry monitors in the reactor
cavity outside the reactor pressure vessel are a useful tool, therefore, in determining the accuracy of the neutron field calculations
at points inside the pressure vessel wall. Practice E853 recommends the use of ex-vessel reactor cavity neutron dosimetry
measurements for verification of the physics transport calculations. The status of benchmark field and power reactor applications
as well as studies of this approach are discussed in Refs. 1, 18, 19, 37-40, 99-112.
5. Neutron Exposure Monitoring
5.1 Initial Conditions:
5.1.1 This guide assumes the existence of an analysis of record that provides projections of future neutron exposure for materials
in the reactor vessel that are predicted to experience sufficient neutron radiation damage to be considered in the selection of the
most limiting material with regard to radiation damage. Projections of future values of neutron exposure are then used in
subsequent reactor vessel integrity evaluations to demonstrate the reactor can be operated safely during normal and off-normal
conditions.
5.1.2 The operational parameters used to generate projections of future neutron exposure are frequently subject to change. A
program to periodically re-assess the neutron exposure should be instituted to confirm that the neutron exposure projections used
in the reactor vessel integrity evaluations remain valid. For highest accuracy, calculations of exposure should be made for all past
E2956 − 23
fuel cycles and projected to the future using best estimates of future fuel management. Significant changes in the neutron exposure
projections may necessitate revisions of the reactor vessel integrity evaluations. When changes in calculated exposure rates are
observed, the differences should be investigated, and the basis of such differences understood.
5.2 Means of Monitoring:
5.2.1 Neutron exposure monitoring can be achieved by periodically performing or updating calculations to reflect actual plant
operating conditions, by collection and analysis of additional reactor dosimetry measurements to validate calculated exposure
projections, or both.
5.2.2 Long periods of operation without any reactor dosimetry measurements can leave undetected errors in the inputs to the
neutron exposure calculation methodology. When significant changes in exposure rates are calculated, new dosimetry
measurements may be required to ensure exposure estimates are within required accuracy limits. Accurate analysis to relate
dosimetry measurements to exposures at critical locations requires fluence calculations for each fuel cycle that the dosimetry is
irradiated and, if shorter half-life dosimeter reactions are used, may require calculations for several time intervals within a fuel
cycle.
5.2.3 Guide E482 provides detailed guidance related to the calculational determination of neutron exposure for BWR and PWR
power plants, and the benchmarking of those calculations. Test Method E1005 describes procedures for measuring the specific
activities of radioactive nuclides produced in radiometric monitors by nuclear reactions induced during surveillance exposures for
reactor vessels and support structures.
5.3 Frequency of Monitoring:
5.3.1 The frequency with which neutron exposure monitoring activities should be performed is dependent upon circumstances
unique to each reactor. To determine an appropriate time interval for the re-assessment of neutron exposure projections,
consideration should be given to the degree of consistency of actual power operation with the assumptions used in developing the
neutron exposure projections, the anticipated margin remaining between current and projected neutron exposure levels, the
physical constraints on the half-lives of the sensor material used in the dosimeters, and potential ancillary uses for the results of
the neutron exposure calculations (for example, equipment qualification or aging management of the reactor vessel internal
structures). Non-technical considerations may also be important. Over long periods of time, staff turnover may lead to challenges
recovering the necessary input data, or a loss of organizational focus may occur on important issues relating to radiation damage
and aging management. It is important that plans be in place to ensure that all nuclear quality assurance requirements are met,
including documentation of all inputs to exposure estimates, and all calculations be carried out and reviewed by qualified
personnel.
5.3.2 For example, consider a reactor that has accrued sufficient neutron exposure to place it near regulatory screening criteria
limits. If continued operation is desired, such a plant may consider implementing fuel changes for the purpose of reducing reactor
vessel neutron exposure. For such a reactor, a shorter monitoring interval may be appropriate to ensure safe operability in-line with
analyzed conditions. By contrast, a plant with wide margins between current and projected neutron exposure, that operates with
core loading patterns and operational parameters that are highly consistent with the projection assumptions, may be justified in
using a longer monitoring interval.
6. Supplementary Analytical Procedures
6.1 Basic Approach—ASTM Practice E853 covers various aspects of the extrapolation problem. The basic approach is that a
transport calculation (benchmarked per Guide E482) is to be used to supply the neutron field information at the (r, θ, z) points in
the pressure vessel wall where property deterioration information will be calculated using Guide E900, or other trend curves (4,
11-18). The dosimetry information obtained from reactor cavity and surveillance capsule measurements and retrospective
dosimetry measurements from reactor internals structures and RPV cladding is to be used to ensure that the transport calculation
is valid and to adjust the transport results if needed. The adjustments are to be accomplished using the guidelines presented in
Guide E944. Dosimetry from monitors in the reactor cavity and surveillance capsules can provide limits on uncertainties for the
calculated neutron field at selected (r, θ, z) positions in the reactor pressure vessel wall. Time dependence of the core power
distribution (due to burnup within a given cycle, or due to variations in cycle to cycle fuel loading), surveillance capsule
perturbation effects, and dosimetry monitor experimental effects must be recognized as complications, and these effects must be
accounted for in the calculation and adjustment methods chosen (1-6, 11).
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6.2 Spatial Extrapolations:
6.2.1 Transport Codes—In general, the minimum analysis for the calculation ofthree-dimensional results need to be obtained for
the neutron and gamma ray fields in the region from the core to the interior of the biological shield beyond the pressure vessel
would be vessel. As a minimum, a three-dimensional synthesis analysis should be performed using a two-dimensional transport
code. The transport calculations would be in that case are carried out using the following three-dimensional synthesis technique:
ϕ r , z
~ !
ϕ~r , θ , z! 5 ϕ~r , θ! (1)
ϕ~r!
where ϕ (r, θ, z) is the synthesized three-dimensional flux distribution, ϕ (r,θ) is the transport solution in r,θ geometry, ϕ (r,z)
is the two-dimensional solution for a cylindrical reactor model, and ϕ (r) is the one-dimensional solution for a cylindrical reactor
model using the same source per unit height as that used in the two-dimensional r,θ calculation.
6.2.1.1 However, other complexities in defining the three-dimensional nature of for problems and regions of interest where the
transport solution, that is, radiation fields, have a nonseparable three-dimensional nature (due to the core power distribution and
reactor internals structures will usually dictate distribution, or reactor internals structures, or away from the core midplane), the
synthesis technique may reduce the accuracy of the results, thus dictating that a full three-dimensional method be used. Analysis
of the extended beltline, which often includes RPV nozzles, also dictates a full three-dimensional approach. An efficient way to
carry out large 3D discrete ordinates S transport calculations is the use of multiple processors running in parallel (112-116). Monte
n
Carlo methods are also used and these are traditionally run in parallel processing computing environments. Guide E482 should be
followed for the calculations and Guide E944 for measured dosimetry adjustments. If a discrete ordinates method is used, the
spatial mesh should be fine enough and the order of angular quadrature should be high enough to ensure a sufficiently accurate
solution in all regions of importance. The impact of the selected discretization settings on the discrete ordinates results should be
assessed (117). Methods of ensuring that the mesh is sufficiently fine are the province of Guide E482. Similar considerations apply
to tallying techniques in Monte Carlo calculations. If ex-vessel reactor cavity dosimeter measurement results are used, the
modeling in the reactor cavity and external shield should be adequate to provide usable calculations for the neutron field in the
reactor cavity region. This requires an attention to mesh size in the ex-vessel region and an accurate representation of the geometry
and chemical makeup of the external shield. Regardless of the method chosen, the effects of relevant input parameter variations
on the calculated results should be well understood. Reference parameter variation studies focused on the reactor cavity and
extended beltline region are available in Ref (118).
6.2.1.2 Benchmarking—It is not the purpose of this guide to dictate the type of transport calculation to be used in the region
between the core and the outer radius of the pressure vessel (or into the biological shield) or the adjustment procedures, but any
such calculations or adjustment procedures should be adequately benchmarked by calculations of well defined problems (for
example, PCA Blind Test (100), VENUS (107), NESDIP (108), BWR (104, 105), and PWR (1, 37-40, 99). For further details on
benchmarking refer to Guide E2006 and Guide E944. USNRC Regulatory Guide 1.190 (118119) also addresses benchmarking of
neutron transport calculations for RPV surveillance in some detail.
6.2.2 Power Distribution—As discussed in Practice E853, obtain a valid, adequately time dependent, core power distribution using
a diffusion calculation, or a transport calculation (99, 100, 107). Experimental verification of the accuracy of the results is desirable,
but may be difficult to obtain. This is especially important for the pin-by-pin power distributions at the core periphery and the axial
power distributions at the ends of the core. The uncertainties in the core power distribution tend to be the largest in these areas.
Fuel assembly geometric features also need to be considered in the development and modeling of the core power distribution. For
example, some PWR fuel assemblies use low-enrichment axial blankets and some BWR fuel bundles use several different fuel rod
lengths within the bundle.
6.2.2.1 Typically, calculations are performed on a fuel cycle-by-fuel cycle basis rather than using a single power distribution that
is averaged over many fuel cycles. A well-documented basis should be used for extrapolating core power distributions into the
future. Extrapolations should be based on best estimate projections of future fuel cycles. One common approach is to average the
three most recent core power distributions and to use that for extrapolation. The assumption being that a similar core loading
strategy will continue to be used. This assumption should be revisited whenever new measurements or core designs become
available.
6.2.2.2 The power distribution should include the assembly-wise and axial variation of power as well as the finer, pin-by-pin
distribution in the peripheral assemblies adjacent to the reactor internals. Details of the initial U enrichment and the cycle
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changes in assembly burnup should also be determined as this is needed in order to define the mix of fissioning isotopes (for
235 238 239 240 241 242 235
example U, U, Pu, Pu, Pu, and Pu) in each fuel assembly. Some BWR fuel bundles use multiple U enrichments
axially within a given fuel rod.
6.2.3 Ex-Core Regions—Perform a transport calculation for the neutron field in all ex-core regions, using adequate modeling of
the reactor geometry, and adequate modeling of the ex-vessel region. The biological shield is to be accurately modeled both in
terms of geometry (ex-core detector wells, support columns, and the presence or absence of a liner plate), and materials including
the biological shield composition (cement, aggregate, water content, and distribution of reinforcing steel). The water content in the
biological shield will vary over time (119120). The energy, angle, and space discretization as well as neutron balance should be
checked in all regions to make sure the calculation has converged, watching in particular for spatial oscillations or ray effects in
ex-vessel regions. Monte Carlo calculations should be checked to confirm that acceptable tally statistics have been achieved.
6.2.4 Power Plant Dimensions—In all calculations, as-built dimensions should be used. If they are unavailable, documented logic
should be presented to defend the dimensions used, and the uncertainty in the final results should reflect the added uncertainty. The
thickness of the reactor pressure vessel (RPV) is a key dimension in the analysis of ex-vessel neutron dosimetry. There are two
ways in which the accuracy of the assumed RPV thickness may be assessed.
237 54
6.2.4.1 The ratio of the Np fission rate to the Fe(n,p) reaction rate in the reactor cavity may be used as a spectral index. This
ratio is very sensitive to the thickness of the RPV. For example, over an RPV thickness range from 100 mm to 200 mm, the reaction
rate ratio increases by nearly a factor of two. Therefore, when the calculated spectral index from a calculation with an assumed
RPV thickness agrees with the measured spectral index, one can have a high degree of confidence that the assumed thickness is
correct. A difference in the spectral index can also indicate how much the assumed RPV thickness is off and in which direction.
The calculated spectral index needs to be determined at the same azimuthal angle as the measurement being compared.
6.2.4.2 RPV pre-service or in-service inspections are usually performed using ultrasonic testing (UT) looking for flaws in the
material. Usually these are multi-angle scans. However, sometimes a zero degree (normal incidence) scan is performed. This UT
scan can provide a direct measured thickness for the RPV. With sufficient advance notice, a zero degree scan can be added to a
future ISI program if the spectral index assessment indicates that the design basis RPV thickness is incorrect.
6.2.5 Dosimetry Sensor Analysis—For analysis of any given set of reactor cavity or surveillance capsule dosimetry sensors, the
integral reactions or reaction rates of the individual sensors, or both, should be calculated, using the results of the transport
calculation. The measurement and analysis procedures for individual sensors should be benchmarked for each sensor type; refer
to Guide E2005. If the calculated and experimental integral results (C/E ratios) agree to within the required accuracy (~ 5 (~5 %
to 15 %, 1σ being the best attainable, see Ref 100) expected from the benchmark calibration of the transport code, the transport
calculation may be used directly to calculate the neutron field at all (r, θ, z) points in the pressure vessel wall. If the C/E ratios
do not agree within acceptable accuracy limits, a physics-dosimetry adjustment code analysis should be performed as outlined in
6.2.6.
6.2.6 Physics-Dosimetry Adjustment Code Analysis—Guide E944 should be used to combine the transport calculation with the
dosimeter results. The Guide E944 adjustment procedure should be used to indicate whether the dosimeter measurements and
associated uncertainties are consistent with the transport calculation and with uncertainties implied from benchmark tests of the
transport code (PCA, VENUS, NESDIP, and an appropriate Commercial BWR or PWR; see Refs 1, 37-40, 99, 100, 104, 105).
Having established the required consistency, the adjusted results of the transport calculation may be used to calculate the best
estimate neutron field at all points in the pressure vessel wall with the uncertainty estimates derived from the application of the
adjustment codes.
6.2.7 Measurement Results—If the calculated neutron field at the measurement location is inconsistent with the experimental
dosimetry results, an attempt should be made to uncover and correct errors in order to obtain consistency. Particular attention
should be paid to sensor monitor correction factors such as capsule perturbation, photo-reactions, impurities, burn-in / burn-out,
and other effects. Discussions of how to proceed when calculations and measurements do not agree may be found in GuidePractice
E853, especially Section 7.3.
6.3 Time Extrapolations—In the case where a time averaged core loading has been used to define the future neutron source term,
the fluence or dpa in future years is estimated by multiplying by the expected integrated time at full power.
7. Report and Bias of Results
7.1 As a minimum, the documentation of results should include the following information:
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7.1.1 A description of the analytical technique used, including a listing of pertinent input parameters that may affect the bias of
the calculation. For example, if the discrete ordinates approach is used, specify or reference the source of the cross-section data,
cross-section preparation procedures, energy group structure, spatial mesh, S order, and P order. Dimensions and material
N L
compositions of key structures included in the model need to be included. Some of this information may be proprietary. In that
case, the source of the data used and a general description should be provided.
7.1.2 Information indicating the bias of the analytical approach in steel-water systems, including the details of benchmark
calculations used to validate the procedures, and data and the bias attained in the benchmark tests.
7.1.3 The calculated total, thermal, epi-thermal (also known as epi-cadmium fluence rate), E > 0.1 MeV, and E > 1.0 MeV neutron
fluence rate-fluence values, and energy spectrum at the surveillance capsule, and any ex-vessel dosimetry locations. Also report
calculated values of dpa/s and dpa at the same locations.
7.1.3.1 The location of peak fluence rate-fluence points on the surface and in the interior of the vessel wall are calculated values
that are required for all the above exposure and exposure rate parameters, except for the thermal and epithermal fluence rates,
which generally can be best determined by dosimetry measurements. For some damage analysis studies, all of the above
information is needed (111, 120-121-124125).
7.1.3.2 At dosimetry measurement locations, gamma ray fluence rate should be estimated to the precision required to make
necessary photo-reaction corrections. Similarly, gamma ray field parame
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