ASTM E853-23

Standard Practice for Analysis and Interpretation of Light-Water Reactor Surveillance Neutron Exposure Results

SIGNIFICANCE AND USE
3.1 The objectives of a reactor vessel surveillance program are twofold. The first requirement of the program is to monitor changes in the fracture toughness properties of ferritic materials in the reactor vessel beltline region resulting from exposure to neutron irradiation and the thermal environment. The second requirement is to make use of the data obtained from the surveillance program to determine the conditions under which the vessel can be operated throughout its service life.  
3.1.1 To satisfy the first requirement of 3.1, the tasks to be carried out are straightforward. Each of the irradiation capsules that comprise the surveillance program may be treated as a separate experiment. The goal is to define and carry to completion a dosimetry program that will, a posteriori, describe the neutron field to which the material test specimens were exposed. The resultant information will then become part of a database applicable in a stricter sense to the specific plant from which the capsule was removed, but also in a broader sense to the industry as a whole.  
3.1.2 To satisfy the second requirement of 3.1, the tasks to be carried out are somewhat complex. The objective is to describe accurately the neutron field to which the pressure vessel itself will be exposed over its service life. This description of the neutron field must include spatial gradients within the vessel wall. Therefore, heavy emphasis must be placed on the use of neutron transport techniques as well as on the choice of a design basis for the computations. Since a given surveillance capsule measurement, particularly one obtained early in plant life, is not necessarily representative of long-term reactor operation, a simple normalization of neutron transport calculations to dosimetry data from a given capsule may not be appropriate (1-67).  
3.2 The objectives and requirements of a reactor vessel's support structure's surveillance program are much less stringent, and at present, are limited to ...
SCOPE
1.1 This practice covers the methodology, summarized in Annex A1, to be used in the analysis and interpretation of neutron exposure data obtained from LWR pressure vessel surveillance programs and, based on the results of that analysis, establishes a formalism to be used to evaluate present and future condition of the pressure vessel and its support structures2 (1-74).3  
1.2 This practice relies on, and ties together, the application of several supporting ASTM standard practices, guides, and methods (see Master Matrix E706) (1, 5, 13, 48, 49).2 In order to make this practice at least partially self-contained, a moderate amount of discussion is provided in areas relating to ASTM and other documents. Support subject areas that are discussed include reactor physics calculations, dosimeter selection and analysis, and exposure units.  
1.3 This practice is restricted to direct applications related to surveillance programs that are established in support of the operation, licensing, and regulation of LWR nuclear power plants. Procedures and data related to the analysis, interpretation, and application of test reactor results are addressed in Practice E1006, Guide E900, and Practice E1035.  
1.4 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.  
1.5 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: 27.120.20 - Nuclear power plants. Safety
• Technical Committee: E10 - Nuclear Technology and Applications
• Drafting Committee: E10.05 - Nuclear Radiation Metrology

Relations

• Refers: ASTM E1018-20 - Standard Guide for Application of ASTM Evaluated Cross Section Data File
• Effective Date: 01-Mar-2020
• Refers: ASTM E1018-20e1 - Standard Guide for Application of ASTM Evaluated Cross Section Data File
• Effective Date: 01-Mar-2020
• Refers: ASTM E854-19 - Standard Test Method for Application and Analysis of Solid State Track Recorder (SSTR) Monitors for Reactor Surveillance
• Effective Date: 01-Nov-2019
• 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 E910-18 - Standard Test Method for Application and Analysis of Helium Accumulation Fluence Monitors for Reactor Vessel Surveillance
• Effective Date: 01-Feb-2018
• 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 E1005-15 - Standard Test Method for Application and Analysis of Radiometric Monitors for Reactor Vessel Surveillance
• Effective Date: 01-Jul-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 E185-15e1 - 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
• Refers: ASTM E900-15e1 - Standard Guide for Predicting Radiation-Induced Transition Temperature Shift in Reactor Vessel Materials
• Effective Date: 01-Feb-2015
• Refers: ASTM E900-15 - Standard Guide for Predicting Radiation-Induced Transition Temperature Shift in Reactor Vessel Materials
• Effective Date: 01-Feb-2015

Overview

ASTM E853-23: Standard Practice for Analysis and Interpretation of Light-Water Reactor Surveillance Neutron Exposure Results provides a comprehensive methodology for evaluating neutron exposure data gathered from surveillance programs in light-water reactor (LWR) pressure vessels and their support structures. This standard practice, developed by ASTM International, ensures consistent and reliable analysis of neutron irradiation effects on reactor vessels, helping to assess integrity, safety, and operational longevity in compliance with regulatory and industry requirements. The document also ties together a suite of referenced ASTM standards and related guides, offering a self-contained resource for reactor operators, analysts, and regulators.

Key Topics

• Neutron Exposure Analysis: Outlining step-by-step procedures for interpreting dosimetry data from surveillance capsules, including how to apply reactor physics calculations, select and analyze dosimeters, and document neutron field characterization.
• Condition Assessment: Formalism for evaluating the present and future state of reactor pressure vessels by analyzing embrittlement and property changes in ferritic materials exposed to neutron irradiation.
• Benchmarking and Qualification: Emphasis on the importance of validated neutron transport methods and benchmarking against established experimental data for accurate spatial and temporal projections of neutron exposure.
• Uncertainty Management: Guidance on estimating, documenting, and addressing uncertainties in measurements and analysis-crucial for safe operation and regulatory compliance.
• Surveillance Program Implementation: Recommendations for designing and maintaining surveillance programs that meet operational and regulatory criteria, including proper placement of surveillance capsules and ongoing monitoring protocols.

Applications

ASTM E853-23 is essential for:

• Nuclear Plant Operators: Ensuring continued safe operation by systematically monitoring material degradation in reactor vessel pressure boundaries and support structures.
• Regulatory Compliance: Providing the methodology required for licensing, re-licensing, and operational extensions of LWR nuclear power plants, as referenced by domestic and international nuclear regulatory authorities.
• Condition-Based Maintenance: Supporting informed decisions regarding pressure vessel and support structure maintenance, annealing, or replacement based on quantified embrittlement and neutron exposure.
• Reactor Life Assessment: Enabling long-term projections and risk assessments that underpin asset management and life extension programs for reactors.
• Technical Evaluations: Facilitating cross-industry data sharing and analysis by establishing standardized protocols, thereby benefiting the broader nuclear industry through comparability and statistical validation.

Related Standards

ASTM E853-23 references multiple supporting standards and guides, providing a robust framework for analysis:

• ASTM E185: Design of Surveillance Programs for Light-Water Moderated Nuclear Power Reactor Vessels
• ASTM E482: Guide for Application of Neutron Transport Methods for Reactor Vessel Surveillance
• ASTM E706: Master Matrix for LWR Pressure Vessel Surveillance Standards
• ASTM E900: Guide for Predicting Radiation-Induced Transition Temperature Shift in Reactor Vessel Materials
• ASTM E2215: Practice for Evaluation of Surveillance Capsules from LWR Vessels
• ASTM E2956: Guide for Monitoring the Neutron Exposure of LWR Reactor Pressure Vessels
• ASTM E1035: Practice for Determining Neutron Exposures for Nuclear Reactor Vessel Support Structures

Regulatory References:

• ASME Boiler and Pressure Vessel Code (Sections III and IX)
• U.S. Code of Federal Regulations, Title 10, Part 50, Appendices G and H

Practical Value

By following ASTM E853-23, organizations achieve high standards in reactor surveillance program implementation and analysis, reducing uncertainty in material behavior assessments and supporting the safe and economical operation of light-water nuclear power plants. The standard’s harmonization with globally recognized principles also makes it a trusted reference for international nuclear safety and reliability programs.

Keywords: light-water reactor, neutron exposure analysis, reactor surveillance, pressure vessel embrittlement, ASTM E853-23, dosimetry, nuclear safety, reactor vessel monitoring, regulatory compliance, benchmarking, reactor operation longevity

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: E853 − 23
Standard Practice for
Analysis and Interpretation of Light-Water Reactor
Surveillance Neutron Exposure Results
This standard is issued under the fixed designation E853; the number immediately following the designation indicates the year of
original adoption or, in the case of revision, the year of last revision. A number in parentheses indicates the year of last reapproval. A
superscript epsilon (´) indicates an editorial change since the last revision or reapproval.
1. Scope mendations issued by the World Trade Organization Technical
Barriers to Trade (TBT) Committee.
1.1 This practice covers the methodology, summarized in
Annex A1, to be used in the analysis and interpretation of
2. Referenced Documents
neutron exposure data obtained from LWR pressure vessel
surveillance programs and, based on the results of that analysis, 2.1 ASTM Standards:
establishes a formalism to be used to evaluate present and E185 Practice for Design of Surveillance Programs for
future condition of the pressure vessel and its support struc- Light-Water Moderated Nuclear Power Reactor Vessels
2 3
tures (1-74). E482 Guide for Application of Neutron Transport Methods
for Reactor Vessel Surveillance
1.2 This practice relies on, and ties together, the application
E509 Guide for In-Service Annealing of Light-Water Mod-
of several supporting ASTM standard practices, guides, and
2 erated Nuclear Reactor Vessels
methods (see Master Matrix E706) (1, 5, 13, 48, 49). In order
E706 Master Matrix for Light-Water Reactor Pressure Vessel
to make this practice at least partially self-contained, a mod-
Surveillance Standards
erate amount of discussion is provided in areas relating to
E844 Guide for Sensor Set Design and Irradiation for
ASTM and other documents. Support subject areas that are
Reactor Surveillance
discussed include reactor physics calculations, dosimeter se-
E854 Test Method for Application and Analysis of Solid
lection and analysis, and exposure units.
State Track Recorder (SSTR) Monitors for Reactor Sur-
1.3 This practice is restricted to direct applications related to
veillance
surveillance programs that are established in support of the
E900 Guide for Predicting Radiation-Induced Transition
operation, licensing, and regulation of LWR nuclear power
Temperature Shift in Reactor Vessel Materials
plants. Procedures and data related to the analysis,
E910 Test Method for Application and Analysis of Helium
interpretation, and application of test reactor results are ad-
Accumulation Fluence Monitors for Reactor Vessel Sur-
dressed in Practice E1006, Guide E900, and Practice E1035.
veillance
1.4 This standard does not purport to address all of the
E944 Guide for Application of Neutron Spectrum Adjust-
safety concerns, if any, associated with its use. It is the ment Methods in Reactor Surveillance
responsibility of the user of this standard to establish appro-
E1005 Test Method for Application and Analysis of Radio-
priate safety, health, and environmental practices and deter- metric Monitors for Reactor Vessel Surveillance
mine the applicability of regulatory limitations prior to use.
E1006 Practice for Analysis and Interpretation of Physics
1.5 This international standard was developed in accor- Dosimetry Results from Test Reactor Experiments
dance with internationally recognized principles on standard-
E1018 Guide for Application of ASTM Evaluated Cross
ization established in the Decision on Principles for the Section Data File
Development of International Standards, Guides and Recom-
E1035 Practice for Determining Neutron Exposures for
Nuclear Reactor Vessel Support Structures
E1214 Guide for Use of Melt Wire Temperature Monitors
This practice is under the jurisdiction of ASTM Committee E10 on Nuclear
for Reactor Vessel Surveillance
Technology and Applications and is the direct responsibility of Subcommittee
E2006 Guide for Benchmark Testing of Light Water Reactor
E10.05 on Nuclear Radiation Metrology.
Calculations
Current edition approved Sept. 1, 2023. Published September 2023. Originally
approved in 1981. Last previous edition approved in 2018 as E853 – 18. DOI:
10.1520/E0853-23.
ASTM Practice E185 gives reference to other standards and references that
address the variables and uncertainties associated with property change measure- For referenced ASTM standards, visit the ASTM website, www.astm.org, or
ments. The referenced standards are A370, E8, E21, E23, and E208. contact ASTM Customer Service at [e-mail]. For Annual Book of ASTM
The boldface numbers in parentheses refer to the list of references appended to Standards volume information, refer to the standard’s Document Summary page on
this practice. For an updated set of references, see Master Matrix E706. the ASTM website.
Copyright © ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959. United States
E853 − 23
E2215 Practice for Evaluation of Surveillance Capsules determine the condition of any support structure steels that
from Light-Water Moderated Nuclear Power Reactor Ves- might be subject to neutron induced property changes (1, 29,
sels 44-58, 65-70).
E2956 Guide for Monitoring the Neutron Exposure of LWR
Reactor Pressure Vessels
4. Establishment of the Surveillance Program
2.2 Other Documents:
4.1 Practice E185 describes the criteria that should be
NUREG/CR-1861 HEDL-TME 80-87 LWR Pressure Vessel
considered in planning and implementing surveillance test
Surveillance Dosimetry Improvement Program: PCA Ex-
programs and points out precautions that should be taken to
periments and Blind Test
ensure that: (1) capsule exposures can be related to beltline
ASME Boiler and Pressure Vessel Code, Sections III and
exposures, (2) materials selected for the surveillance program
IX
are samples of those materials most likely to limit the operation
Code of Federal Regulations, Title 10, Part 50, Appendixes
of the reactor vessel, and (3) the tests yield results useful for
G and H
the evaluation of radiation effects on the reactor vessel.
4.1.1 From the viewpoint of the radiation analyst, the
3. Significance and Use
criteria explicated in Practice E185 are met by the completion
3.1 The objectives of a reactor vessel surveillance program
of the following tasks: (1) Determine the locations within the
are twofold. The first requirement of the program is to monitor
reactor that provide suitable lead factors (see Practice E185)
changes in the fracture toughness properties of ferritic materi-
for each irradiation capsule relative to the pressure vessel; (2)
als in the reactor vessel beltline region resulting from exposure
Select neutron sensor sets that provide adequate coverage over
to neutron irradiation and the thermal environment. The second
the energy range and fluence range of interest; (3) Specify
requirement is to make use of the data obtained from the
sensor set locations within each irradiation capsule to define
surveillance program to determine the conditions under which
neutron field gradients within the metallurgical specimen array.
the vessel can be operated throughout its service life.
For reactors in which the end of life shift in RT of the
NDT
3.1.1 To satisfy the first requirement of 3.1, the tasks to be
pressure vessel beltline material is predicted to be less than
carried out are straightforward. Each of the irradiation capsules
100 °F, gradient measurements are not required. In that case
that comprise the surveillance program may be treated as a
sensor set locations may be chosen to provide a representative
separate experiment. The goal is to define and carry to
measurement for the entire surveillance capsule; and (4)
completion a dosimetry program that will, a posteriori, de-
Establish and adequately benchmark neutron transport meth-
scribe the neutron field to which the material test specimens
odology to be used both in the analysis of individual sensor sets
were exposed. The resultant information will then become part
and in the projection of materials properties changes to the
of a database applicable in a stricter sense to the specific plant
vessel itself.
from which the capsule was removed, but also in a broader
4.1.2 The first three items listed in the preceding paragraph
sense to the industry as a whole.
are carried out during the design of the surveillance program.
3.1.2 To satisfy the second requirement of 3.1, the tasks to
However, the fourth item, which directly addresses the analysis
be carried out are somewhat complex. The objective is to
and interpretation of surveillance results, is performed follow-
describe accurately the neutron field to which the pressure
ing withdrawal of the surveillance capsules from the reactor. To
vessel itself will be exposed over its service life. This descrip-
provide continuity between the designer and the analyst, it is
tion of the neutron field must include spatial gradients within
recommended that the documentation describing the surveil-
the vessel wall. Therefore, heavy emphasis must be placed on
lance programs of individual reactors provide details of irra-
the use of neutron transport techniques as well as on the choice
diation capsule construction, locations of the capsules relative
of a design basis for the computations. Since a given surveil-
to the reactor core and internals, and sensor set design that are
lance capsule measurement, particularly one obtained early in
adequate to allow accurate evaluations of the surveillance
plant life, is not necessarily representative of long-term reactor
measurement by the analyst. Well-documented (1) metallurgi-
operation, a simple normalization of neutron transport calcu-
cal and (2) physics-dosimetry databases now exist for use by
lations to dosimetry data from a given capsule may not be
the analyst based on both power reactor surveillance capsule
appropriate (1-67).
and test reactor results (1, 12, 19-38, 58-64).
3.2 The objectives and requirements of a reactor vessel’s 4.1.3 Information regarding the choice of neutron sensor
support structure’s surveillance program are much less sets for LWR surveillance applications is provided in Master
stringent, and at present, are limited to physics-dosimetry Matrix E706: Guide E844, Sensor Set Design; Test Method
measurements through ex-vessel cavity monitoring coupled E1005, Radiometric Monitors; Test Method E854, Solid State
with the use of available test reactor metallurgical data to Track Recorder Monitors; Test Method E910, Helium Accu-
mulation Fluence Monitors; and Damage Monitors. Dosimeter
materials currently in common usage and acceptable for use in
238 237
surveillance programs include Cu, Ti, Fe, Ni, Nb, U , Np ,
Available from NRC Public Document Room, 1717 H St., NW, Washington,
DC 20555.
U , and Co-Al. All radionuclide analysis of dosimeters
Available from American Society of Mechanical Engineers, Three Park Ave.,
should be calibrated to known sources such as those supplied
New York, NY 10016-5990.
by the National Institute of Standards and Technology (NIST)
Available from Superintendent of Documents, U. S. Government Printing
Office, Washington, DC 20402. or the International Atomic Energy Agency (IAEA). All quality
E853 − 23
assurance information pertinent to the sensor sets must be 4.2 As stated in 3.2, the requirements for the establishment
documented with the description of the surveillance program of a surveillance program for reactor vessel support structures
are much less stringent than for the reactor vessel, and the
(1, 40-43, 48, 51-58).
analyst is referred to Practice E1035 for more information.
4.1.4 As indicated in 4.1.1, neutron transport methods are
used both in the design of the surveillance program and in the
5. Analysis of Individual Surveillance Capsules
analysis and interpretation of capsule measurements. During
the design phase, neutron transport calculations are used to
5.1 For surveillance programs designed according to Prac-
define the neutron field within the pressure vessel wall and, in
tice E185, individual surveillance capsules are periodically
conjunction with damage trend curves, to predict the degree of
removed for analysis throughout plant life. Practice E2215
embrittlement of the reactor vessel over its service life.
provides guidance on testing and evaluating irradiated surveil-
Embrittlement gradients are in turn used to determine pressure-
lance capsules, as well as guidance on updating withdrawal
temperature limitations for normal plant operation as well as to
schedules in circumstances where reactor operation beyond the
evaluate the effect of various heat-up/cool-down transients on
original design life is planned.
vessel condition.
5.2 It is recognized that for many operating power reactors,
4.1.5 The neutron transport methodology used for these
the documentation of baseline neutron transport calculations
computations must be well benchmarked and qualified for
and sensor set design information may not be available. In that
application to LWR configurations. The PCA (Experiment and
event, to whatever extent possible the required information
Blind Test) data documented in Ref 47 provide one configu-
should be provided by the service laboratory in the respective
ration for benchmarking basic transport methodology as well
surveillance report (1, 29, 58).
as some of the input data used in power reactor calculations.
5.3 Radiometric analysis of capsule sensor sets should
Other suitably defined and documented benchmark
follow procedures outlined in Test Method E1005. For sensors
experiments, such as those for VENUS (1, 43, 45) and for
such as the fission monitors which may be gamma-ray
NESDIP (1, 46, 50), may also be used to provide method
sensitive, photo reaction corrections should be derived from
verification. However, further analytical/experimental com-
the results of gamma-ray transport calculations performed for
parisons are required to qualify a method for application to
the explicit capsule configuration under examination. Photo
LWRs that have a more complex geometry and that require a
reaction corrections in LWR environments have been shown to
more complex treatment of some input parameters, particularly
be extremely configuration dependent (1, 29, 58). Gamma-ray
of reactor core power distributions (1, 65-67). This additional
calculations should be well benchmarked. One such suitable
qualification may be achieved by comparison with measure-
reactor geometry benchmark is VENUS-1 (75, 76).
ments taken in the reactor cavity external to the pressure vessel
of selected operating reactors (1, 51-57).
5.4 In calculating spectrum-averaged reaction cross sections
from neutron transport calculations, care should be taken to
4.1.6 All experimental/analytical comparisons that com-
model the explicit capsule configuration and location under
prise the qualification program for a neutron transport meth-
examination (see Guide E482). It will be necessary to deter-
odology must be documented. At a minimum, this documen-
mine uncertainties associated with the determination of dam-
tation should provide an assessment of the uncertainty or error
age exposure parameters. The procedures outlined in Guide
inherent in applying the methodology to the evaluation of
E944, IIA can, in many cases, be useful for accomplishing this.
surveillance capsule dosimetry and to the determination of
To achieve satisfactory uncertainty bounds for the damage
damage gradients within the beltline region of the pressure
parameters, a sufficiently large set of foils should be used as
vessel (1, 12, 19-21, 23-29, 36, 38, 43-48, 50-57).
stipulated in 4.1.3 (1, 29, 36).
4.1.7 In the application of neutron transport methodology to
the evaluation of surveillance dosimetry as well as to the
5.5 The report of the capsule analysis should contain the
prediction of damage within the pressure vessel, several
following information. Uncertainties should be included in all
options are available regarding the choice of design basis data (1, 29, 36).
power distributions, the necessary detail in the geometric
5.5.1 Damage exposure parameters at the position of the
mockup, and the normalization of the analytical results. The
metallurgical specimens. These values will be used for corre-
methodology chosen by any analyst should be documented
lation with metallurgical data to develop damage trend curves.
with sufficient detail to permit a critical evaluation of the
Neutron fluence (E > 1.0 MeV) is presently required. However,
overall approach. Further discussions of the application of
iron dpa (displacements per atom) and neutron fluence (E > 0.1
neutron transport methods to LWRs are provided in Guide
MeV) should also be included for future reference. These
E482.
exposure values are derived from a combination of measure-
ments and calculations and must include estimates of uncer-
4.1.8 To ensure that metallurgical results obtained from
tainty bounds.
surveillance capsule measurements may be applied to the
determination of the pressure vessel fracture toughness, the 5.5.2 The neutron spectra, reaction rates, reaction cross
irradiation temperature of the surveillance test specimens must sections, and all other nuclear constants used in the derivation
be documented (see Guide E1214). of exposure values for the capsule.
E853 − 23
5.5.3 The gamma-ray energy spectra and reaction cross 6.1.5 If the measurements differ from the calculations by
sections used to make photoreaction corrections for the neutron more than the margins indicated by the benchmark
sensor sets. documentation, further investigation of the measurement ap-
5.5.4 The power-time history of the reactor during the proach and the mode of operation of the reactor in question
irradiation period of the subject capsule. should be undertaken. Any adjustments made to vessel em-
5.5.5 Spatial gradients of neutron fluence rate (E > 1.0 brittlement projections based on the results of these investiga-
MeV), neutron fluence (E > 1.0 MeV), and dpa throughout the tions should be justified and fully documented in the surveil-
metallurgical specimen array. lance report.
5.5.6 In addition, the documentation supporting the
6.2 Reactor Vessel Support Structures—The analyst is re-
benchmarking/qualification of sensor sets and reactor physics
ferred to Practice E1035.
methodology should be either referenced or included as an
6.3 Monitoring—Operational parameters used to generate
appendix to the dosimetry report.
projections of future neutron exposure are frequently subject to
6. Projection of Vessel and Support Structure Condition
change. A program to perform periodic monitoring of the
for Future Plant Operation
neutron exposure should be instituted. Neutron exposure moni-
6.1 Reactor Vessel: toring can be performed by way of updating calculations to
6.1.1 This practice requires the use of a fully benchmarked
reflect actual operating conditions, by collection and analysis
and qualified neutron transport methodology in both the design of additional in-vessel or ex-vessel reactor dosimetry measure-
of the surveillance program and the analysis of individual
ments to confirm exposure projections, or both. Guide E2956
surveillance capsules. The neutron field information obtained provides guidance on neutron exposure monitoring for LWR
from these computations should also be used to project damage
pressure vessels.
gradients within the pressure vessel wall. Currently, such
projections are based on effective vessel wall neutron fluence
7. Extrapolating Reactor Vessel Surveillance Dosimetry
(E > 1.0 MeV). It is common practice to quantify neutron
Results
exposure for all pressure vessel materials and locations ex-
7.1 Knowledge of the time-dependent relationship between
pected to accrue a neutron fluence (E > 1.0 MeV) of 1.0E + 17
exposure parameters at surveillance locations and selected (r,
n/cm at the end of facility operation (77). Guide E900 and
θ, z) locations within the pressure vessel wall is required to
NRC Regulatory Guide 1.99 (74) specify methods for deter-
allow determination of the time-dependent radiation damage to
mining the attenuation of the effective vessel wall fluence, up
the pressure vessel. The time dependency must be known to
to a specified depth from the inner surface. However, it is
allow proper accounting for complications due to burn-up, as
recommended that supplementary projections based on dpa
well as change in core loading configurations (20, 65-67). An
maps throughout the pressure vessel beltline region/
estimate of the uncertainty in the neutron exposure parameter
surveillance capsule geometry be included in the surveillance
values at selected (r, θ, z) points in the vessel wall (1) is also
report (1, 12, 19-21, 23-29, 33, 36, 38-48, 51-67).
needed to place an upper bound on the allowable operating
6.1.2 It is recommended that all surveillance results for a
lifetime of the reactor vessel without remedial action (21, 22,
generic reactor type (similar reactor geometry and fuel loading)
71). (See Guide E509.)
be used as a database to qualify the reactor physics methodol-
ogy as to its applicability to a particular reactor system. This 7.2 Several other ASTM practices cover various aspects of
approach should, in the long term, provide a statistically the extrapolation problem (see 2.1). The basic approach is that
significant validation of the calculations. a benchmarked Guide E482 transport calculation is to be used
6.1.3 Capsules removed from symmetric positions in ge- to supply the neutron field information at the (r, θ, z) points in
neric reactor geometries represent a series of repeat measure- the pressure vessel wall where property deterioration informa-
ments. As such, the measured data will reflect the variability in tion will be calculated using Guide E900, or other trend curves
important parameters such as water temperature, reactor (3, 12, 20, 24-27, 72-74). The dosimetry information obtained
from reactor cavity (ex-vessel) and surveillance capsule (in-
dimensions, fuel loading, sensor set design, sensor set analysis,
and reactor operating characteristics. By taking advantage of a vessel) measurements is to be used to adjust the transport
large database obtained from these repeat measurements, the results and ensure that the transport calculation is valid. The
uncertainties introduced by these various parameters may be adjustments are to be accomplished using the guidelines
better understood and possibly reduced. presented in Guide E944. Dosimetry from monitors in the
6.1.4 When evaluating the results of a given surveillance reactor cavity and surveillance capsules will be used in
capsule analysis, the measured capsule exposure should be establishing uncertainties for the calculated neutron field at
compared directly with neutron transport analysis and with all selected (r, θ, z) positions in the pressure vessel wall. Time
available experimental data obtained from similar capsules dependence of the core power distribution (due to burnup
removed from reactors having the same design. If the agree- within a given cycle, or due to variations in cycle-to-cycle
ment between measurement and calculation is within the range loading), surveillance capsule perturbation effects, and dosim-
indicated by the benchmark documentation for the specified etry monitor experimental effects must be recognized as
methodology, the analytically derived neutron field parameters complications, and these effects must be accounted for in the
should be used for all damage determinations for the pressure calculation and adjustment methods chosen (1, 3, 20, 21,
vessel (29). 65-67).
E853 − 23
7.3 Spatial Extrapolations: account of the uncertainties in both the dosimetry and transport
results) and if the discrepancy cannot be resolved, then the
7.3.1 Transport Codes—a three-dimensional or two-
dimensional [(r, θ), (x, y)] transport code is needed for the transport results should be scaled up proportionally to obtain
calculation of the neutron and gamma fields in the region from agreement, following which the transport results are to be used
the core to the interior of the biological shield beyond the for extrapolation purposes. In this case, appropriate increases
pressure vessel. Guide E482 should be followed for the should be made in the stated uncertainties of the final result,
calculations and Guide E944 for measured dosimetry adjust-
and documented logic should be provided to defend the
ments.
assigned uncertainties.
7.3.2 Dosimetry Sensor Analysis—For analysis of any given
7.3.4 Ex-Vessel Surveillance Results—Ex-vessel reactor
set of reactor cavity or surveillance capsule dosimetry sensors,
cavity dosimetry is to be treated in the same manner as
the integral reactions or reaction rates of the individual sensors,
surveillance capsule dosimetry, but care must be exercised to
or both, should be calculated using the results of the transport
ensure that the physics calculation modeling is adequate and
calculation; see Guides E844, E1018, E2006, Test Methods
includes the proper modeling of the reactor cavity surveillance
E1005, E854, and E910 (see 2.1).
capsule and any covers, as well as any nearby vessel support
7.3.2.1 If the calculated and experimental integral results
members.
(C/E ratios) agree to within the required accuracy (~5 % to
7.3.4.1 The biological shield is accurately modeled.
15 %, 1σ being the best attainable, see (47)) expected from the
7.3.4.2 In the final calculation of the neutron and gamma
benchmark calibration of the transport code, the transport
field at any point in the vessel wall, proper statistical weight
calculation may be used directly to calculate the neutron field
should be given to ex-vessel dosimetry, taking account of
at all (r, θ, z) points in the pressure vessel wall.
modeling problems as well as the possibility that a larger
7.3.2.2 If the C/E ratios do not agree within acceptable
logarithmic extrapolation or interpolation in absolute fluence
accuracy limits, a physics-dosimetry adjustment code analysis
value exists from ex-vessel positions to a ⁄4 T location when
should be performed as described in Guide E944. Having
compared to the extrapolation or interpolation from an internal
established the required consistency, the adjusted transport
surveillance capsule position to a ⁄4 T location.
code results may be used to calculate the neutron field at all
7.3.5 Power Plant Dimensions—In all calculations, as-built
points in the pressure vessel wall with the uncertainty estimates
dimensions should be used. If they are unavailable, docu-
derived from the application of the adjustment codes.
mented logic should be presented to defend the dimensions
7.3.2.3 Direct use of the transport code results with appro-
used, and the uncertainty in the final results should reflect the
priate bias factors and uncertainties is another acceptable
added uncertainty. It should be noted that dpa declines ~10
approach.
%/cm of radial trave
...

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: E853 − 18 E853 − 23
Standard Practice for
Analysis and Interpretation of Light-Water Reactor
Surveillance Neutron Exposure Results
This standard is issued under the fixed designation E853; the number immediately following the designation indicates the year of
original adoption or, in the case of revision, the year of last revision. A number in parentheses indicates the year of last reapproval. A
superscript epsilon (´) indicates an editorial change since the last revision or reapproval.
1. Scope
1.1 This practice covers the methodology, summarized in Annex A1, to be used in the analysis and interpretation of neutron
exposure data obtained from LWR pressure vessel surveillance programs;programs and, based on the results of that analysis,
2 3
establishes a formalism to be used to evaluate present and future condition of the pressure vessel and its support structures (1-74).
1.2 This practice relies on, and ties together, the application of several supporting ASTM standard practices, guides, and methods
(see Master Matrix E706) (1, 5, 13, 48, 49). In order to make this practice at least partially self-contained, a moderate amount
of discussion is provided in areas relating to ASTM and other documents. Support subject areas that are discussed include reactor
physics calculations, dosimeter selection and analysis, and exposure units.
1.3 This practice is restricted to direct applications related to surveillance programs that are established in support of the operation,
licensing, and regulation of LWR nuclear power plants. Procedures and data related to the analysis, interpretation, and application
of test reactor results are addressed in Practice E1006, Guide E900, and Practice E1035.
1.4 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.
1.5 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:
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
E706 Master Matrix for Light-Water Reactor Pressure Vessel Surveillance Standards
This practice 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 Dec. 1, 2018Sept. 1, 2023. Published December 2018September 2023. Originally approved in 1981. Last previous edition approved in 20132018
as E853 – 13.E853 – 18. DOI: 10.1520/E0853-18.10.1520/E0853-23.
ASTM Practice E185 gives reference to other standards and references that address the variables and uncertainties associated with property change measurements. The
referencereferenced standards are A370, E8, E21, E23, and E208.
The boldface numbers in parentheses refer to the list of references appended to this practice. For an updated set of references, see the E706 Master Matrix.Master Matrix
E706.
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.
Copyright © ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959. United States
E853 − 23
E844 Guide for Sensor Set Design and Irradiation for Reactor Surveillance
E854 Test Method for Application and Analysis of Solid State Track Recorder (SSTR) Monitors for Reactor Surveillance
E900 Guide for Predicting Radiation-Induced Transition Temperature Shift in Reactor Vessel Materials
E910 Test Method for Application and Analysis of Helium Accumulation Fluence Monitors for Reactor Vessel Surveillance
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
E1006 Practice for Analysis and Interpretation of Physics Dosimetry Results from Test Reactor Experiments
E1018 Guide for Application of ASTM Evaluated Cross Section Data File
E1035 Practice for Determining Neutron Exposures for Nuclear Reactor Vessel Support Structures
E1214 Guide for Use of Melt Wire Temperature Monitors for Reactor Vessel Surveillance
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
E2956 Guide for Monitoring the Neutron Exposure of LWR Reactor Pressure Vessels
2.2 Other Documents:
NUREG/CR-1861 HEDL-TME 80-87 LWR Pressure Vessel Surveillance Dosimetry Improvement Program: PCA Experiments
and Blind Test
ASME Boiler and Pressure Vessel Code, Sections III and IX
Code of Federal Regulations, Title 10, Part 50, Appendixes G and H
3. Significance and Use
3.1 The objectives of a reactor vessel surveillance program are twofold. The first requirement of the program is to monitor changes
in the fracture toughness properties of ferritic materials in the reactor vessel beltline region resulting from exposure to neutron
irradiation and the thermal environment. The second requirement is to make use of the data obtained from the surveillance program
to determine the conditions under which the vessel can be operated throughout its service life.
3.1.1 To satisfy the first requirement of 3.1, the tasks to be carried out are straightforward. Each of the irradiation capsules that
comprise the surveillance program may be treated as a separate experiment. The goal is to define and carry to completion a
dosimetry program that will, a posteriori, describe the neutron field to which the materialsmaterial test specimens were exposed.
The resultant information will then become part of a data base database applicable in a stricter sense to the specific plant from
which the capsule was removed, but also in a broader sense to the industry as a whole.
3.1.2 To satisfy the second requirement of 3.1, the tasks to be carried out are somewhat complex. The objective is to describe
accurately the neutron field to which the pressure vessel itself will be exposed over its service life. This description of the neutron
field must include spatial gradients within the vessel wall. Therefore, heavy emphasis must be placed on the use of neutron
transport techniques as well as on the choice of a design basis for the computations. Since a given surveillance capsule
measurement, particularly one obtained early in plant life, is not necessarily representative of long-term reactor operation, a simple
normalization of neutron transport calculations to dosimetry data from a given capsule may not be appropriate (1-67).
3.2 The objectives and requirements of a reactor vessel’s support structure’s surveillance program are much less stringent, and at
present, are limited to physics-dosimetry measurements through ex-vessel cavity monitoring coupled with the use of available test
reactor metallurgical data to determine the condition of any support structure steels that might be subject to neutron induced
property changes (1, 29, 44-58, 65-70).
4. Establishment of the Surveillance Program
4.1 Practice E185 describes the criteria that should be considered in planning and implementing surveillance test programs and
points out precautions that should be taken to ensure that: (1) capsule exposures can be related to beltline exposures, (2) materials
selected for the surveillance program are samples of those materials most likely to limit the operation of the reactor vessel, and
(3) the tests yield results useful for the evaluation of radiation effects on the reactor vessel.
4.1.1 From the viewpoint of the radiation analyst, the criteria explicated in Practice E185 are met by the completion of the
following tasks: (1) Determine the locations within the reactor that provide suitable lead factors (see Practice E185) for each
irradiation capsule relative to the pressure vessel; (2) Select neutron sensor sets that provide adequate coverage over the energy
Available from NRC Public Document Room, 1717 H St., NW, Washington, DC 20555.
Available from American Society of Mechanical Engineers, Three Park Ave., New York, NY 10016-5990.
Available from Superintendent of Documents, U. S. Government Printing Office, Washington, DC 20402.
E853 − 23
range and fluence range of interest; (3) Specify sensor set locations within each irradiation capsule to define neutron field gradients
within the metallurgical specimen array. For reactors in which the end of life shift in RT of the pressure vessel beltline material
NDT
is predicted to be less than 100 °F, gradient measurements are not required. In that case sensor set locations may be chosen to
provide a representative measurement for the entire surveillance capsule; and (4) Establish and adequately benchmark neutron
transport methodology to be used both in the analysis of individual sensor sets and in the projection of materials properties changes
to the vessel itself.
4.1.2 The first three items listed in the preceding paragraph are carried out during the design of the surveillance program. However,
the fourth item, which directly addresses the analysis and interpretation of surveillance results, is performed following withdrawal
of the surveillance capsules from the reactor. To provide continuity between the designer and the analyst, it is recommended that
the documentation describing the surveillance programs of individual reactors provide details of irradiation capsule construction,
locations of the capsules relative to the reactor core and internals, and sensor set design that are adequate to allow accurate
evaluations of the surveillance measurement by the analyst. Well documentedWell-documented (1 (1) ) metallurgical and (2)(2
physics-dosimetry data bases ) physics-dosimetry databases now exist for use by the analyst based on both power reactor
surveillance capsule and test reactor results (1, 12, 19-38, 58-64).
4.1.3 Information regarding the choice of neutron sensor sets for LWR surveillance applications is provided in Master Matrix
E706: Guide E844, Sensor Set Design; Test Method E1005, Radiometric Monitors; Test Method E854, Solid State Track Recorder
Monitors; Specification Test Method E910, Helium Accumulation Fluence Monitors; and Damage Monitors. Dosimeter materials
238 237 235
currently in common usage and acceptable for use in surveillance programs include Cu, Ti, Fe, Ni, Nb, U , Np , U , and
Co-Al. All radionuclide analysis of dosimeters should be calibrated to known sources such as those supplied by the National
Institute of Standards and TerchnologyTechnology (NIST) or Thethe International Atomic Energy Agency (IAEA). All quality
assurance information pertinent to the sensor sets must be documented with the description of the surveillance program (1, 40-43,
48, 51-58).
4.1.4 As indicated in 4.1.1, neutron transport methods are used both in the design of the surveillance program and in the analysis
and interpretation of capsule measurements. During the design phase, neutron transport calculations are used to define the neutron
field within the pressure vessel wall and, in conjunction with damage trend curves, to predict the degree of embrittlement of the
reactor vessel over its service life. Embrittlement gradients are in turn used to determine pressure-temperature limitations for
normal plant operation as well as to evaluate the effect of various heat-up/cool-down transients on vessel condition.
4.1.5 The neutron transport methodology used for these computations must be well benchmarked and qualified for application to
LWR configurations. The PCA (Experiment and Blind Test) data documented in Ref 47 provide one configuration for
benchmarking basic transport methodology as well as some of the input data used in power reactor calculations. Other suitably
defined and documented benchmark experiments, such as those for VENUS (1, 43, 45) and for NESDIP (1, 46, 50), may also be
used to provide method verification. However, further analytical/experimental comparisons are required to qualify a method for
application to LWRs that have a more complex geometry and that require a more complex treatment of some input parameters,
particularly of reactor core power distributions (1, 65-67). This additional qualification may be achieved by comparison with
measurements taken in the reactor cavity external to the pressure vessel of selected operating reactors (1, 51-57).
4.1.6 All experimental/analytical comparisons that comprise the qualification program for a neutron transport methodology must
be documented. At a minimum, this documentation should provide an assessment of the uncertainty or error inherent in applying
the methodology to the evaluation of surveillance capsule dosimetry and to the determination of damage gradients within the
beltline region of the pressure vessel (1, 12, 19-21, 23-29, 36, 38, 43-48, 50-57).
4.1.7 In the application of neutron transport methodology to the evaluation of surveillance dosimetry as well as to the prediction
of damage within the pressure vessel, several options are available regarding the choice of design basis power distributions, the
necessary detail in the geometric mockup, and the normalization of the analytical results. The methodology chosen by any analyst
should be documented with sufficient detail to permit a critical evaluation of the overall approach. Further discussions of the
application of neutron transport methods to LWRs are provided in Guide E482.
4.1.8 To ensure that metallurgical results obtained from surveillance capsule measurements may be applied to the determination
of the pressure vessel fracture toughness, the irradiation temperature of the surveillance test specimens must be documented (see
Guide E1214).
4.2 As stated in 3.2, the requirements for the establishment of a surveillance program for reactor vessel support structures are much
less stringent than for the reactor vessel, and the analyst is referred to Practice E1035, for more information.
E853 − 23
5. Analysis of Individual Surveillance Capsules
5.1 For surveillance programs designed according to Practice E185, individual surveillance capsules are periodically removed for
analysis throughout plant life. Practice E2215 provides guidance on testing and evaluating irradiated surveillance capsules, as well
as guidance on updating withdrawal schedules in circumstances where reactor operation beyond the original design life is planned.
5.2 It is recognized that for many operating power reactors, the documentation of baseline neutron transport calculations and
sensor set design information may not be available. In that event, to whatever extent possible the required information should be
provided by the service laboratory in the respective surveillance report (1, 29, 58).
5.3 Radiometric analysis of capsule sensor sets should follow procedures outlined in Test Method E1005. For sensors such as the
fission monitors which may be gamma-ray-sensitive, gamma-ray sensitive, photo reaction corrections should be derived from the
results of gamma-ray transport calculations performed for the explicit capsule configuration under examination. Photo reaction
corrections in LWR environments have been shown to be extremely configuration dependent (1, 29, 58). Gamma-ray calculations
should be well benchmarked. One such suitable reactor geometry benchmark is VENUS-1 (75, 76).
5.4 In calculating spectrum averaged spectrum-averaged reaction cross sections from neutron transport calculations, care should
be taken to model the explicit capsule configuration and location under examination (see Guide E482.)). It will be necessary to
determine uncertainties associated with the determination of damage exposure parameters. The procedures outlined in Guide E944,
IIA can, in many cases, be useful for accomplishing this. To achieve satisfactory uncertainty bounds for the damage parameters,
a sufficiently large set of foils should be used as stipulated in 4.1.3 (1, 29, 36).
5.5 The report of the capsule analysis should contain the following information. Uncertainties should be included in all data (1,
29, 36).
5.5.1 Damage exposure parameters at the position of the metallurgical specimens. These values will be used for correlation with
metallurgical data to develop damage trend curves. Neutron fluence (E > 1.0 MeV) is presently required. However, iron dpa
(displacements per atom) and neutron fluence (E > 0.1 MeV) should also be included for future reference. These exposure values
are derived from a combination of measurements and calculations and must include estimates of uncertainty bounds,bounds.
5.5.2 The neutron spectra, reaction rates, reaction cross sections, and all other nuclear constants used in the derivation of exposure
values for the capsule,capsule.
5.5.3 The gamma-ray energy spectra and reaction cross sections used to make photoreaction corrections for the neutron sensor
sets, sets.
5.5.4 The power-time history of the reactor during the irradiation period of the subject capsule, andcapsule.
5.5.5 Spatial gradients of neutron fluence rate (E > 1.0 MeV), neutron fluence (E > 1.0 MeV), and dpa throughout the metallurgical
specimen array.
5.5.6 In addition, the documentation supporting the benchmarking/qualification of sensor sets and reactor physics methodology
should be either referenced or included as an appendix to the dosimetry report.
6. Projection of Vessel and Support Structure Condition for Future Plant Operation
6.1 Reactor Vessel:
6.1.1 This practice requires the use of a fully benchmarked and qualified neutron transport methodology in both the design of the
surveillance program and the analysis of individual surveillance capsules. The neutron field information obtained from these
computations should also be used to project damage gradients within the pressure vessel wall. Currently, such projections are based
on effective vessel wall neutron fluence (E > 1.0 MeV). It is common practice to quantify neutron exposure for all pressure vessel
materials and locations expected to accrue a neutron fluence (E > 1.0 MeV) of 1.0E + 17 n/cm at the end of facility operation
(77). Guide E900 and NRC Regulatory Guide 1.99 (74) specify methods for determining the attenuation of the effective vessel wall
fluence, up to a specified depth from the inner surface. However, it is recommended that supplementary projections based on dpa
maps throughout the pressure vessel beltline region/surveillance capsule geometry be included in the surveillance report (1, 12,
1919-21, 20, 21, 23-29, 33, 36, 38-48, 51-67).
E853 − 23
6.1.2 It is recommended that all surveillance results for a generic reactor type (similar reactor geometry and fuel loading) be used
as a data base database to qualify the reactor physics methodology as to its applicability to a particular reactor system. This
approach should, in the long term, provide a statistically significant validation of the calculations.
6.1.3 Capsules removed from symmetric positions in generic reactor geometries represent a series of repeat measurements. As
such, the measured data will reflect the variability in important parameters such as water temperature, reactor dimensions, fuel
loading, sensor set design, sensor set analysis, and reactor operating characteristics. By taking advantage of a large data base
database obtained from these repeat measurements, the uncertainties introduced by these various parameters may be better
understood and possibly reduced.
6.1.4 When evaluating the results of a given surveillance capsule analysis, the measured capsule exposure should be compared
directly with neutron transport analysis and with all available experimental data obtained from similar capsules removed from
reactors having the same design. If the agreement between measurement and calculation is within the range indicated by the
benchmark documentation for the specified methodology, the analytically derived neutron field parameters should be used for all
damage determinations for the pressure vessel (29).
6.1.5 If the measurements differ from the calculations by more than the margins indicated by the benchmark documentation,
further investigation of the measurement approach and the mode of operation of the reactor in question should be undertaken. Any
adjustments made to vessel embrittlement projections based on the results of these investigations should be justified and fully
documented in the surveillance report.
6.2 Reactor Vessel Support Structures—The analyst is referred to Practice E1035.
6.3 Monitoring—Operational parameters used to generate projections of future neutron exposure are frequently subject to change.
A program to perform periodic monitoring of the neutron exposure should be instituted. Neutron exposure monitoring can be
performed by way of updating calculations to reflect actual operating conditions, by collection and analysis of additional in-vessel
or ex-vessel reactor dosimetry measurements to confirm exposure projections, or both. PracticeGuide E2956 provides guidance on
neutron exposure monitoring for LWR pressure vessels.
7. Extrapolating Reactor Vessel Surveillance Dosimetry Results
7.1 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 pressure
vessel. The time dependency must be known to allow proper accounting for compicationscomplications due to burn-up, as well
as,as change in core loading configurations (20, 65-67).). 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 (21, 22, 71). (See Guide E509).)
7.2 Several other ASTM practices cover various aspects of the extrapolation problem (see 2.1). The basic approach is that a
benchmarked Guide E482, transport calculation 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 (3, 12,
20, 24-27, 72-74). The dosimetry information obtained from reactor cavity (ex-vessel) and surveillance capsule (invessel)(in-
vessel) measurements is to be used to adjust the transport results and ensure that the transport calculation is valid. The adjustments
are to be accomplished using the guidelines presented in Guide E944. Dosimetry from monitors in the reactor cavity and
surveillance capsules will be used in establishing uncertainties for the calculated neutron field at selected (r, θ, z) positions in the
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 cycle-to-cycle 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, 3,
20, 21, 65-67).
7.3 Spatial Extrapolations:
7.3.1 Transport Codes—a three dimensional or two dimensional three-dimensional or two-dimensional [(r, θ), (x,y)] transport code
is needed for the calculation of the neutron and gamma fields in the region from the core to the interior of the biological shield
beyond the pressure vessel. Guide E482 should be followed for the calculations and Guide E944 for measured dosimetry
adjustments.
E853 − 23
7.3.2 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,calculated using the results of the
transport calculation,calculation; see Guides E844, E1018, E2006, Test Methods E1005, E854, and E910 (See(see 2.1).
7.3.2.1 If the calculated and experimental integral results (C/E ratios) agree to within the required accuracy (~ 5 to 15 %, (~5 %
to 15 %, 1σ being the best attainable, see (47)) 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.
7.3.2.2 If the C/E ratios do not agree within acceptable accuracy limits, a physics-dosimetry adjustment code analysis should be
performed as described in Guide E944. Having established the required consistency, the adjusted transport code results may be
used to calculate the neutron field at all points in the pressure vessel wall with the uncertainty estimates derived from the
application of the adjustment codes.
7.3.2.3 Direct use of the transport code results with appropriate bias factors and uncertainties is another acceptable approach.
7.3.2.4 Guide E900, Section 6, provides spatial extrapolation formulae that can optionally be used for relating the neutron
exposure at the inside surface of a LWR pressure vessel to the neutron exposure at an arbitrary depth inside the pressure vessel.
However, these extrapolation formulae constitute simplified approximations of complicated physical phenomena. Neutron
transport calculations able to perform detailed space-, direction-, and energy-dependent accounting of changes in the neutron
population through the pressure vessel material should be used to according to Equation 7 in Guide E900 (or equivalently Equation
8, if applicable) to perform spatial extrapolations when available. The extrapolation formula in Equation 9 of Guide E900
constitutes a simplified approximation of complicated physical phenomena, and users are advised to be cognizant of its potential
limitations, particularly for regions that are not adjacent to the active core.
7.3.3 Surveillance Capsule Results—If the calculated neutron field at the surveillance capsule is inconsistent with the experimental
dosimetry results, an attempt should be made to uncover and correct errors in order to obtain consistency. Particular attention will
be required to sensor monitor correction factors for perturbation, photo-reaction, impurity, burn-in, and other effects.
7.3.3.1 If the transport result indicates a higher fluence than that indicated by the dosimetry, the transport result can be used for
extrapolation purposes, but with an appropriate increase in the stated uncertainty for the results.
7.3.3.2 If the transport calculation indicates a lower fluence than that which would be consistent with the dosimetry (taking
account of the uncertainties in both the dosimetry and transport results) and if the discrepancy cannot be resolved, then the transport
results should be scaled up proportionally to obtain agreement, following which the transport results are to be used for
extrapolation purposes. In this case, appropriate increases should be made in the stated uncertainties of the final result, and
documented logic should be provided to defend the assigned uncertainties.
7.3.4 Ex-Vessel Surveillance Results—Ex-vessel reactor cavity dosimetry is to be treated in the same manner as surveillance
capsule dosimetry, but care must be exercised to ensure that the physics calculation modeling is adequate and includes the proper
modeling of the reactor cavity surveillance capsule and any covers, as well as any nearby vessel support members.
7.3.4.1 The biological shield is accurately modeled.
7.3.4.2 In the final calculation of the neutron and gamma field at any point in the vessel wall, proper statistical weight should be
given to ex-vessel dosimetry, taking account of modeling problems as well as the possibility that a larger logarithmic extrapolation
or interpolation in absolute fluence value exists from ex-vessel positions to a ⁄4 T location when compared to the extrapolation
or interpolation from an internal surveillance capsule position to a ⁄4 T location.
7.3.5 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. It
should be noted that dpa declines ~10 %/cm of radial travel, in water, and deviations of ~3 cm between design dimensions and
as-built dimensions have been observed in commercial reactors.
7.4 Time Extrapolations—In the case where a time averaged time-averaged core loading has been used to define the neutron source
term, the fluence or dpa in future years is estimated by multiplying by the expected integrated time at full power. Existing problems
E853 − 23
associated with time extrapolations (for example, saturation effects and differences in the slope of trend curves for different ferritic
steels) are addressed elsewhere. The reader is referred to Refs (1, 3, 12, 21, 23-27, 72-74),) and Guide E900 for more information
on these subjects.
8. Uncertainties
8.1 Analysis and measurement accuracies (uncertainties and errors) in the areas of concern for this practice may be difficult to
determine. However, they should be properly addressed (1, 12, 19-22, 23-29, 36, 38, 39, 43, 44, 47, 48, 51, 58-64). When
uncertainties and errors are well defined, as in integral reaction rate measurements, they should be estimated and summarized in
an accuracy table. For more difficult uncertainty situations, such as for damage exposure parameters, the procedure for determining
uncertainties must be well documented. A statement must be included that indicates what the uncertainty estimates do and
...