ASTM E853-87(1995)e1

NOTICE: This standard has either been superseded and replaced by a new version or discontinued.
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e1
Designation: E 853 – 87 (Reapproved 1995)
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 (e) indicates an editorial change since the last revision or reapproval.
e NOTE—Keywords were added editorially in April 1996.
1. Scope bility of regulatory limitations prior to use.
1.1 This practice covers the methodology, summarized in
2. Referenced Documents
Annex A1, to be used in the analysis and interpretation of
2.1 ASTM Standards:
neutron exposure data obtained from LWR pressure vessel
E 170 Terminology Relating to Radiation Measurements
surveillance programs; and, based on the results of that
and Dosimetry
analysis, establishes a formalism to be used to evaluate present
E 184 Practice for Effects of High-Energy Neutron Radia-
and future condition of the pressure vessel and its support
2 3 tion on the Mechanical Properties of Metallic Materials,
structures (1-70).
E706 (IB)
1.2 This practice relies on, and ties together, the application
E 185 Practice for Conducting Surveillance Tests for Light
of several supporting ASTM standard practices, guides, and
Water Cooled Nuclear Power Reactor Vessels, E706 (IF)
methods that are in various stages of completion (see Fig. 1 and
2 E 482 Guide for Application of Neutron Transport Methods
Master Matrix E 706) (1, 5, 13, 48, 49). In order to make this
for Reactor Vessel Surveillance, E706 (IID)
practice at least partially self-contained, a moderate amount of
E 560 Practice for Extrapolating Reactor Vessel Surveil-
discussion is provided in areas relating to ASTM and other
lance Dosimetry Results, E706 (IC)
documents. Support subject areas that are discussed include
E 636 Guide for Conducting Supplemental Surveillance
reactor physics calculations, dosimeter selection and analysis,
Tests for Nuclear Power Reactor Vessels, E706 (IH)
and exposure units.
E 693 Practice for Characterizing Neutron Exposures in
1.3 Since several of the standards shown in Fig. 1 are not
Iron and Low Alloy Steels in Terms of Displacements Per
currently in place, some of the requirements listed in Annex A1
Atom (DPA), E706 (ID)
should, at this time, be treated as recommendations. Appropri-
E 706 Master Matrix for Light Water Reactor Pressure
ate caution should be exercised until each of the standards has
Vessel Surveillance Standards
been put into use.
IE Damage Correlation for Reactor Vessel Surveillance
1.4 This practice is restricted to direct applications related to
IIE Benchmark Testing of Reactor Vessel Dosimetry
surveillance programs that are established in support of the
IIID Application and Analysis of Damage Monitors for
operation, licensing, and regulation of LWR nuclear power
Reactor Vessel Surveillance
plants. Procedures and data related to the analysis, interpreta-
IIIE Application and Analysis of Temperature Monitors
tion, and application of test reactor results are addressed in
E 844 Guide for Sensor Set Design and Irradiation for
Matrix E 706 (IE), Practice E 560, Matrix E 706 (IC), E706
Reactor Surveillance, E706 (IIC)
(II), Guide E 900, and E 706(IG).
E 854 Test Method for Application and Analysis of Solid
1.5 This standard does not purport to address all of the
State Track Recorder (SSTR) Monitors for Reactor Sur-
safety concerns, if any, associated with its use. It is the
veillance, E706 (IIIB)
responsibility of the user of this standard to establish appro-
E 900 Guide for Predicting Neutron Radiation Damage to
priate safety and health practices and determine the applica-
Reactor Vessel Materials, E706 (IIF)
E 910 Specification for Application and Analysis of Helium
This practice is under the jurisdiction of ASTM Committee E-10 on Nuclear
Accumulation Fluence Monitors for Reactor Vessel Sur-
Technology and Applicationsand is the direct responsibility of Subcommittee
veillance, E706 (IIIC)
E10.05 on Nuclear Radiation Metrology.
Current edition approved Oct. 30, 1987. Published December 1987. Originally
e1
published as E 853 – 81. Last previous edition E 853 – 84 .
2 4
ASTM Practice E 185 gives reference to other standards and references that Annual Book of ASTM Standards, Vol 12.02.
address the variables and uncertainties associated with property change measure- Annual Book of ASTM Standards, Vol 12.02. The reference in parentheses refers
ments. The reference standards are A370, E8, E21, E23, and E208. to Section 5 and Figs. 1 and 2 of Matrix E 706.
3 6
The boldface numbers in parentheses refer to the list of references appended to For standards that are in the draft stage and have not received an ASTM
this practice. For an updated set of references, see the E706 Master Matrix. designation, see Section 5 and Figs. 1 and 2 of Matrix E 706.
Copyright © ASTM, 100 Barr Harbor Drive, West Conshohocken, PA 19428-2959, United States.
E 853
FIG. 1 ASTM Standards for Surveillance of LWR Nuclear Reactor Pressure Vessels and Support Structures
E 944 Guide for Application of Neutron Spectrum Adjust- to neutron irradiation and the thermal environment. The second
ment Methods in Reactor Surveillance, (IIA) requirement is to make use of the data obtained from the
E 1005 Test Method for Application and Analysis of Radio- surveillance program to determine the conditions under which
metric Monitors for Reactor Vessel Surveillance, E706 the vessel can be operated throughout its service life.
(IIIA)
3.1.1 To satisfy the first requirement of 3.1, the tasks to be
E 1006 Practice for Analysis and Interpretation of Physics carried out are straightforward. Each of the irradiation capsules
Dosimetry Results for Test Reactors, E706 (II)
that comprise the surveillance program may be treated as a
E 1018 Guide for Application of ASTM Evaluated Cross separate experiment. The goal is to define and carry to
Section Data File (ENDF/A)—Cross Section and Uncer-
completion a dosimetry program that will, a posteriori, de-
tainty File, E706 (IIB) scribe the neutron field to which the materials test specimens
E 1035 Practice for Determining Radiation Exposures for
were exposed. The resultant information will then become part
Nuclear Reactor Vessel Support Structures, (IG) of a data base applicable in a stricter sense to the specific plant
2.2 Other Documents:
from which the capsule was removed, but also in a broader
NUREG/CR-1861 HEDL-TME 80-87 LWR Pressure Ves- sense to the industry as a whole.
sel Surveillance Dosimetry Improvement Program: PCA
3.1.2 To satisfy the second requirement of 3.1, the tasks to
Experiments and Blind Test
be carried out are somewhat complex. The objective is to
ASME Boiler and Pressure Vessel Code, Sections III and
describe accurately the neutron field to which the pressure
IX
vessel itself will be exposed over its service life. This descrip-
Code of Federal Regulations, Title 10, Part 50, Appendixes
tion of the neutron field must include spatial gradients within
G and H the vessel wall. Therefore, heavy emphasis must be placed on
the use of neutron transport techniques as well as on the choice
3. Significance and Use
of a design basis for the computations. Since a given surveil-
3.1 The objectives of a reactor vessel surveillance program lance capsule measurement, particularly one obtained early in
are twofold. The first requirement of the program is to monitor plant life, is not necessarily representative of long-term reactor
changes in the fracture toughness properties of ferritic materi- operation, a simple normalization of neutron transport calcu-
als in the reactor vessel beltline region resulting from exposure lations 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 strin-
Available from NRC Public Document Room, 1717 H St. NW, Washington, DC
20555.
gent, and at present, are limited to physics-dosimetry measure-
Available from American Society of Mechanical Engineers, 345 E. 47th St.,
ments through ex-vessel cavity monitoring coupled with the
New York, NY 10017.
use of available test reactor metallurgical data to determine the
Available from Superintendent of Documents, U. S. Government Printing
Office, Washington, DC 20402. condition of any support structure steels that might be subject
E 853
to neutron induced property changes (1, 29, 44-58, 65-70). used both in the design of the surveillance program and in the
analysis and interpretation of capsule measurements. During
4. Establishment of the Surveillance Program
the design phase, neutron transport calculations are used to
4.1 Practice E 185 describes the criteria that should be
define the neutron field within the pressure vessel wall and, in
considered in planning and implementing surveillance test
conjunction with damage trend curves, to predict the degree of
programs and points out precautions that should be taken to
embrittlement of the reactor vessel over its service life.
ensure that: (1) capsule exposures can be related to beltline
Embrittlement gradients are in turn used to determine pressure-
exposures, (2) materials selected for the surveillance program
temperature limitations for normal plant operation as well as to
are samples of those materials most likely to limit the opera-
evaluate the effect of various heat-up/cool-down transients on
tion of the reactor vessel, and (3) the tests yield results useful
vessel condition.
for the evaluation of radiation effects on the reactor vessel.
4.1.5 The neutron transport methodology used for these
4.1.1 From the viewpoint of the radiation analyst, the
computations must be well benchmarked and qualified for
criteria explicated in Practice E 185 are met by the completion
application to LWR configurations. The PCA (Experiment and
of the following tasks: (1) Determine the locations within the
Blind Test) data documented in Ref 47 provide one configu-
reactor that provide suitable lead factors (see Practice E 185)
ration for benchmarking basic transport methodology as well
for each irradiation capsule relative to the pressure vessel; (2)
as some of the input data used in power reactor calculations.
Select neutron sensor sets that provide adequate coverage over
Other suitably defined and documented benchmark experi-
the energy range and fluence range of interest; (3) Specify
ments, such as those for VENUS (1, 43, 45) and for NESDIP
sensor set locations within each irradiation capsule to define
(1, 46, 50), may also be used to provide method verification.
neutron field gradients within the metallurgical specimen array.
However, further analytical/experimental comparisons are re-
For reactors in which the end of life shift in RT of the
quired to qualify a method for application to LWRs that have
NDT
pressure vessel beltline material is predicted to be less than
a more complex geometry and that require a more complex
100°F, gradient measurements are not required. In that case
treatment of some input parameters, particularly of reactor core
sensor set locations may be chosen to provide a representative
power distributions (1, 65-67). This additional qualification
measurement for the entire surveillance capsule; and (4)
may be achieved by comparison with measurements taken in
Establish and adequately benchmark neutron transport meth-
the reactor cavity external to the pressure vessel of selected
odology to be used both in the analysis of individual sensor sets
operating reactors (1, 51-57).
and in the projection of materials properties changes to the
4.1.6 All experimental/analytical comparisons that com-
vessel itself.
prise the qualification program for a neutron transport meth-
4.1.2 The first three items listed in the preceding paragraph
odology must be documented. At a minimum, this documen-
are carried out during the design of the surveillance program.
tation should provide an assessment of the uncertainty or error
However, the fourth item, which directly addresses the analysis
inherent in applying the methodology to the evaluation of
and interpretation of surveillance results, is performed follow-
surveillance capsule dosimetry and to the determination of
ing withdrawal of the surveillance capsules from the reactor. To
damage gradients within the beltline region of the pressure
provide continuity between the designer and the analyst, it is
vessel (1, 12, 19-21, 23-29, 36, 38, 43-48, 50-57).
recommended that the documentation describing the surveil-
4.1.7 In the application of neutron transport methodology to
lance programs of individual reactors provide details of irra-
the evaluation of surveillance dosimetry as well as to the
diation capsule construction, locations of the capsules relative
prediction of damage within the pressure vessel, several
to the reactor core and internals, and sensor set design that are
options are available regarding the choice of design basis
adequate to allow accurate evaluations of the surveillance
power distributions, the necessary detail in the geometric
measurement by the analyst. Well documented (1) metallurgi-
mockup, and the normalization of the analytical results. The
cal and (2) physics-dosimetry data bases now exist for use by
methodology chosen by any analyst should be documented
the analyst based on both power reactor surveillance capsule
with sufficient detail to permit a critical evaluation of the
and test reactor results (1, 12, 19-38, 58-64).
overall approach. Further discussions of the application of
4.1.3 Information regarding the choice of neutron sensor
neutron transport methods to LWRs are provided in Practice
sets for LWR surveillance applications is provided in Matrix
E 560, IC, and Guide E 482, IID.
E 706: IIC, Sensor Set Design; IIIA, Radiometric Monitors;
4.1.8 To ensure that metallurgical results obtained from
IIIB, Solid State Track Recorder Monitors; IIIC, Helium
surveillance capsule measurements may be applied to the
Accumulation Fluence Monitors; and Damage Monitors. Do-
determination of the pressure vessel fracture toughness, the
simeter materials currently in common usage and acceptable
irradiation temperature of the surveillance test specimens must
for use in surveillance programs include Cu, Ti, Fe, Ni, U ,
be documented (see Matrix E 706 (IIIE)).
237 235
Np ,U , and Co-Al. All radionuclide analysis of dosim-
4.2 As stated in 3.2, the requirements for the establishment
eters should be calibrated to known sources such as those
o
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