ASTM E853-01(2008)

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.
´1
Designation:E853–87(Reapproved 1995) Designation: E 853 – 01 (Reapproved 2008)
Standard Practice for
Analysis and Interpretation of Light-Water Reactor
Surveillance Results, E706(IA)
This standard is issued under the fixed designation E 853; 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.
´ NOTE—Keywords were added editorially in April 1996.
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; and, based on the results of that analysis, establishes a
2 3
formalism to be used to evaluate present and future condition of the pressure vessel and its support structures (1-70).
1.2This practice relies on, and ties together, the application of several supportingASTM standard practices, guides, and methods
that are in various stages of completion (see Fig. 1 and Master Matrix E706
1.2 This practice relies on, and ties together, the application of several supporting ASTM standard practices, guides, and
methods (see Master Matrix E 706) (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 toASTM and other documents. Support subject areas that are discussed include
reactor physics calculations, dosimeter selection and analysis, and exposure units.
1.3Since several of the standards shown in Fig. 1 are not currently in place, some of the requirements listed inAnnexA1 should,
at this time, be treated as recommendations.Appropriate caution should be exercised until each of the standards has been put into
use.
1.4This 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 Matrix E706 (IE), Practice E560, Matrix E706 (IC), E706 (II), Guide E900, and
E706(IG).
1.5
NOTE 1—(Figure 1 is deleted in the latest update. The user is refered to Master Matrix E 706 for the latest figure of the standards interconnectivity).
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 E 560, Practice E 1006, Guide E 900, and Practice E 1035.
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 and health practices and determine the applicability of regulatory
limitations prior to use.
2. Referenced Documents
2.1 ASTM Standards:
E170Terminology Relating to Radiation Measurements and Dosimetry
E184Practice for Effects of High-Energy Neutron Radiation on the Mechanical Properties of Metallic Materials, E706 (IB)
E 185Practice for Conducting Surveillance Tests for Light Water Cooled Nuclear Power Reactor Vessels, E706 (IF) Practice
for Design of Surveillance Programs for Light-Water Moderated Nuclear Power Reactor Vessels
E 482 Guide for Application of Neutron Transport Methods for Reactor Vessel Surveillance, E706 (IID)
E 560Practice for Extrapolating Reactor Vessel Surveillance Dosimetry Results, E706 (IC)
This practice is under the jurisdiction of ASTM Committee E-10E10 on Nuclear Technology and Applications and is the direct responsibility of Subcommittee E10.05
on Nuclear Radiation Metrology.
Current edition approved Oct. 30, 1987.Nov. 1, 2008. Published December 1987.November 2008 Originally published as E853–81.approved in 1981. Last previous edition
´1
E853–84 . approved in 2001 E 853 – 01.
ASTM Practice E 185 gives reference to other standards and references that address the variables and uncertainties associated with property change measurements. The
reference 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.
Annual Book of ASTM Standards, Vol 12.02.
Copyright © ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959, United States.
E 853 – 01 (2008)
E636Guide for Conducting Supplemental Surveillance Tests for Nuclear Power Reactor Vessels, E706 (IH)
E693Practice for Characterizing Neutron Exposures in Iron and LowAlloy Steels in Terms of Displacements PerAtom (DPA),
E706 (ID) Practice for Extrapolating Reactor Vessel Surveillance Dosimetry Results, E 706(IC)
E 706Master Matrix for Light Water Reactor Pressure Vessel Surveillance Standards Master Matrix for Light-Water Reactor
Pressure Vessel Surveillance Standards, E 706(0)
IEDamage Correlation for Reactor Vessel Surveillance
IIEBenchmark Testing of Reactor Vessel Dosimetry
IIID Application and Analysis of Damage Monitors for Reactor Vessel Surveillance
IIIEApplication and Analysis of Temperature Monitors
E844Guide for Sensor Set Design and Irradiation for Reactor Surveillance, E706 (IIC)
E854Test Method forApplication andAnalysis of Solid State Track Recorder (SSTR) Monitors for Reactor Surveillance, E706
(IIIB)
E 844 Guide for Sensor Set Design and Irradiation for Reactor Surveillance, E 706(IIC)
E 854 Test Method for Application and Analysis of Solid State Track Recorder (SSTR) Monitors for Reactor Surveillance,
E706(IIIB)
E 900Guide for Predicting Neutron Radiation Damage to Reactor Vessel Materials, E706 (IIF)
E910Specification for Application and Analysis of Helium Accumulation Fluence Monitors for Reactor Vessel Surveillance,
E706 (IIIC)
Guide for Predicting Radiation-Induced Transition Temperature Shift in Reactor Vessel Materials, E706 (IIF)
E 910 Test Method for Application and Analysis of Helium Accumulation Fluence Monitors for Reactor Vessel Surveillance,
E706 (IIIC)
E 944 Guide for Application of Neutron Spectrum Adjustment Methods in Reactor Surveillance, E 706 (IIA)
E 1005Test Method for Application and Analysis of Radiometric Monitors for Reactor Vessel Surveillance, E706 (IIIA)
E1006Practice for Analysis and Interpretation of Physics Dosimetry Results for Test Reactors, E706 (II)
E1018GuideforApplicationofASTMEvaluatedCrossSectionDataFile(ENDF/A)—CrossSectionandUncertaintyFile,E706
(IIB)
Test Method for Application and Analysis of Radiometric Monitors for Reactor Vessel Surveillance, E 706(IIIA)
E 1006 Practice for Analysis and Interpretation of Physics Dosimetry Results for Test Reactors, E 706(II)
E 1035Practice for Determining Radiation Exposures for Nuclear Reactor Vessel Support Structures, (IG) Practice for
Determining NeutronExposures for Nuclear Reactor Vessel Support Structures
E 1214 Guide for Use of Melt Wire Temperature Monitors for Reactor Vessel Surveillance, E 706 (IIIE)
E 2006 Guide for Benchmark Testing of Light Water Reactor Calculations
2.2 Other Documents:
NUREG/CR-1861 HEDL-TME 80-87 LWR Pressure Vessel Surveillance Dosimetry Improvement Program: PCAExperiments
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 materials test specimens were exposed. The
resultant information will then become part of a data base 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
Annual Book of ASTM Standards, Vol 12.02. The reference in parentheses refers to Section 5 and Figs. 1 and 2 of Matrix E 706.
Available from NRC Public Document Room, 1717 H St., NW, Washington, DC 20555.
For standards that are in the draft stage and have not received an ASTM designation, see Section 5 and Figs. 1 and 2 of Matrix E706.
Available from American Society of Mechanical Engineers, Three Park Ave., New York, NY 10016-5990.
Available from NRC Public Document Room, 1717 H St. NW, Washington, DC 20555.
Available from Superintendent of Documents, U. S. Government Printing Office, Washington, DC 20402.
E 853 – 01 (2008)
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 E 185 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 opera-
tion 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 E 185 are met by the completion of the
following tasks: (1) Determine the locations within the reactor that provide suitable lead factors (see Practice E 185) for each
irradiation capsule relative to the pressure vessel; (2) Select neutron sensor sets that provide adequate coverage over the energy
range and fluence range of interest; (3) Specify sensor set locations within each irradiation capsule to define neutron field gradients
withinthemetallurgicalspecimenarray.ForreactorsinwhichtheendoflifeshiftinRT ofthepressurevesselbeltlinematerial
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
transportmethodologytobeusedbothintheanalysisofindividualsensorsetsandintheprojectionofmaterialspropertieschanges
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 documented (1) metallurgical and (2)
physics-dosimetry data bases 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 Matrix E 706:
IIC,GuideE 844,SensorSetDesign;IIIA,TestMethodE 1005,RadiometricMonitors;IIIB,TestMethodE 854,SolidStateTrack
Recorder Monitors; IIIC,Specification E 910, Helium Accumulation Fluence Monitors; and Damage Monitors. Dosimeter
238 237 235
materials currently in common usage and acceptable for use in surveillance programs include Cu, Ti, Fe, Ni, U ,Np ,U ,
and Co-Al.All radionuclide analysis of dosimeters should be calibrated to known sources such as those supplied by the National
BureauInstitute of Standards (NBS) and Terchnology (NIST) or The 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 Asindicatedin4.1.1,neutrontransportmethodsareusedbothinthedesignofthesurveillanceprogramandintheanalysis
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 f
...

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:E853–01 Designation: E 853 – 01 (Reapproved 2008)
Standard Practice for
Analysis and Interpretation of Light-Water Reactor
Surveillance Results, E706(IA)
This standard is issued under the fixed designation E 853; 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; and, based on the results of that analysis, establishes a
2 3
formalism to be used to evaluate present and future condition of the pressure vessel and its support structures (1-70).
1.2 This practice relies on, and ties together, the application of several supporting ASTM standard practices, guides, and
methods (see Master Matrix E 706) (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 toASTM and other documents. Support subject areas that are discussed include
reactor physics calculations, dosimeter selection and analysis, and exposure units.
NOTE 1—(Figure 1 is deleted in the latest update. The user is refered to Master Matrix E 706 for the latest figure of the standards interconnectivity).
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 E 560, Practice E 1006, Guide E 900, and Practice E 1035.
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 and health practices and determine the applicability of regulatory
limitations prior to use.
2. Referenced Documents
2.1 ASTM Standards:
E170Terminology Relating to Radiation Measurements and Dosimetry
4 4
E184Practice for Effects of High-Energy Neutron Radiation on the Mechanical Properties of Metallic Materials, E706 (IB)
E 185 Practice for Conducting Surveillance Tests for Light- Water Cooled Nuclear Power Reactor Vessels, E706 (IF) Practice
for Design of Surveillance Programs for Light-Water Moderated Nuclear Power Reactor Vessels
E 482 Guide for Application of Neutron Transport Methods for Reactor Vessel Surveillance, E706 (IID)
E 560 Practice for Extrapolating Reactor Vessel Surveillance Dosimetry Results, E706 (IC)
E636Guide for Conducting Supplemental Surveillance Tests for Nuclear Power Reactor Vessels, E706 (IH)
E693Practice for Characterizing Neutron Exposures in Iron and LowAlloy Steels in Terms of Displacements PerAtom (DPA),
E706 (ID) Practice for Extrapolating Reactor Vessel Surveillance Dosimetry Results, E 706(IC)
E 706 Master Matrix for Light-Water Reactor Pressure Vessel Surveillance Standards, E 706(0)
E 844 Guide for Sensor Set Design and Irradiation for Reactor Surveillance, E 706(IIC)
E 854 Test Method for Application and Analysis of Solid State Track Recorder (SSTR) Monitors for Reactor Surveillance,
E706(IIIB)
E 900 Guide for Predicting Neutron Radiation Damage to Reactor Vessel Materials, E706 (IIF) Guide for Predicting
Radiation-Induced Transition Temperature Shift in Reactor Vessel Materials, E706 (IIF)
E 910 Specification Test Method for Application and Analysis of Helium Accumulation Fluence Monitors for Reactor Vessel
Surveillance, E706 (IIIC)
E 944 Guide for Application of Neutron Spectrum Adjustment Methods in Reactor Surveillance, E 706 (IIA)
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.
´1
Current edition approved June 10, 2001. Published September 2001 Originally published as E853–81. Last previous edition E853–95 .
Current edition approved Nov. 1, 2008. Published November 2008 Originally approved in 1981. Last previous edition approved in 2001 E 853 – 01.
ASTM Practice E 185 gives reference to other standards and references that address the variables and uncertainties associated with property change measurements. The
reference 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.
Annual Book of ASTM Standards, Vol 12.02.
Copyright © ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959, United States.
E 853 – 01 (2008)
E 1005 Test Method for Application and Analysis of Radiometric Monitors for Reactor Vessel Surveillance, E 706(IIIA)
E 1006 Practice for Analysis and Interpretation of Physics Dosimetry Results for Test Reactors, E706 (II)
E1018Guide forApplication ofASTM Evaluated Cross Section Data File, E706 (IIB) Practice forAnalysis and Interpretation
of Physics Dosimetry Results for Test Reactors, E 706(II)
E 1035 Practice for Determining RadiationNeutron Exposures for Nuclear Reactor Vessel Support Structures, E706
(IG) Structures
E 1214 Guide for Use of Melt Wire Temperature Monitors for Reactor Vessel Surveillance, E706 (IIIE)
E2005Guide for the Benchmark Testing of Reactor Dosimetry in Standard and Reference Neutron Fields, E706 (IIE-1) Guide
for Use of Melt Wire Temperature Monitors for Reactor Vessel Surveillance, E 706 (IIIE)
E 2006 Guide for the Benchmark Testing of Light Water Reactor Calculations
2.2 Other Documents:
NUREG/CR-1861 HEDL-TME 80-87 LWR Pressure Vessel Surveillance Dosimetry Improvement Program: PCAExperiments
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 materials test specimens were exposed. The
resultant information will then become part of a data base 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 E 185 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 opera-
tion 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 E 185 are met by the completion of the
following tasks: (1) Determine the locations within the reactor that provide suitable lead factors (see Practice E 185) for each
irradiation capsule relative to the pressure vessel; (2) Select neutron sensor sets that provide adequate coverage over the energy
range and fluence range of interest; (3) Specify sensor set locations within each irradiation capsule to define neutron field gradients
withinthemetallurgicalspecimenarray.ForreactorsinwhichtheendoflifeshiftinRT ofthepressurevesselbeltlinematerial
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
transportmethodologytobeusedbothintheanalysisofindividualsensorsetsandintheprojectionofmaterialspropertieschanges
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
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.
E 853 – 01 (2008)
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 documented (1) metallurgical and (2)
physics-dosimetry data bases 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 Matrix E 706:
Guide E 844, Sensor Set Design; Test Method E 1005, Radiometric Monitors; Test Method E 854, Solid State Track Recorder
Monitors; Specification E 910, HeliumAccumulation Fluence Monitors; and Damage Monitors. Dosimeter materials currently in
238 237 235
common usage and acceptable for use in surveillance programs include Cu, Ti, Fe, Ni, 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 Terchnology (NIST) or The 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 Asindicatedin4.1.1,neutrontransportmethodsareusedbothinthedesignofthesurveillanceprogramandintheanalysis
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 Intheapplicationofneutrontransportmethodologytotheevaluationofsurveillancedosimetryaswellastotheprediction
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.
...