ASTM E2215-19

Standard Practice for Evaluation of Surveillance Capsules from Light-Water Moderated Nuclear Power Reactor Vessels

SIGNIFICANCE AND USE
4.1 Neutron radiation effects are considered in the design of light-water moderated nuclear power reactors. Changes in system operating parameters may be made throughout the service life of the reactor to account for these effects. A surveillance program is used to measure changes in the properties of actual vessel materials due to the irradiation environment. This practice describes the criteria that should be considered in evaluating surveillance program test capsules.  
4.2 Prior to the first issue date of this standard, the design of surveillance programs and the testing of surveillance capsules were both covered in a single standard, Practice E185. Between its provisional adoption in 1961 and its replacement linked to this standard, Practice E185 was revised many times (1966, 1970, 1973, 1979, 1982, 1993 and 1998). Therefore, capsules from surveillance programs that were designed and implemented under early versions of the standard were often tested after substantial changes to the standard had been adopted. For clarity, the standard practice for surveillance programs has been divided into the new Practice E185 that covers the design of new surveillance programs and this standard practice that covers the testing and evaluation of surveillance capsules. Modifications to the standard test program and supplemental tests are described in Guide E636.  
4.3 This practice is intended to cover testing and evaluation of all light-water moderated reactor pressure vessel surveillance capsules. The practice is applicable to testing of capsules from surveillance programs designed and implemented under all previous versions of Practice E185.  
4.4 The radiation-induced changes in the properties of the reactor pressure vessel are generally monitored by measuring the index temperatures, the upper-shelf energy and the tensile properties of specimens from the surveillance program capsules. The significance of these radiation-induced changes is described in Practice E185.  
4...
SCOPE
1.1 This practice covers the evaluation of test specimens and dosimetry from light water moderated nuclear power reactor pressure vessel surveillance capsules.  
1.2 Additionally, this practice provides guidance on reassessing withdrawal schedule for design life and operation beyond design life.  
1.3 This practice is one of a series of standard practices that outline the surveillance program required for nuclear reactor pressure vessels. The surveillance program monitors the irradiation-induced changes in the ferritic steels that comprise the beltline of a light-water moderated nuclear reactor pressure vessel.  
1.4 This practice along with its companion surveillance program practice, Practice E185, is intended for application in monitoring the properties of beltline materials in any light-water moderated nuclear reactor.2  
1.5 Modifications to the standard test program and supplemental tests are described in Guide E636.  
1.6 The values stated in SI units are to be regarded as the standard. The values given in parentheses are for information only.  
1.7 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-May-2019
• ICS: 27.120.10 - Reactor engineering
• Technical Committee: E10 - Nuclear Technology and Applications
• Drafting Committee: E10.02 - Behavior and Use of Nuclear Structural Materials

Relations

• Replaces: ASTM E2215-18 - Standard Practice for Evaluation of Surveillance Capsules from Light-Water Moderated Nuclear Power Reactor Vessels
• Effective Date: 01-Jun-2019
• Refers: ASTM E23-24 - Standard Test Methods for Notched Bar Impact Testing of Metallic Materials
• Effective Date: 01-Apr-2024
• Refers: ASTM A370-24 - Standard Test Methods and Definitions for Mechanical Testing of Steel Products
• Effective Date: 01-Mar-2024
• Refers: ASTM E8/E8M-24 - Standard Test Methods for Tension Testing of Metallic Materials
• Effective Date: 01-Jan-2024
• Refers: ASTM E1921-23b - Standard Test Method for Determination of Reference Temperature, <emph type="bdit">T<inf >0</inf></emph>, for Ferritic Steels in the Transition Range
• Effective Date: 15-Dec-2023
• Refers: ASTM E1921-23a - Standard Test Method for Determination of Reference Temperature, <emph type="bdit">T<inf >0</inf></emph>, for Ferritic Steels in the Transition Range
• Effective Date: 01-Nov-2023
• Refers: ASTM E1921-23 - Standard Test Method for Determination of Reference Temperature, <emph type="bdit">T<inf >0</inf></emph>, for Ferritic Steels in the Transition Range
• Effective Date: 01-Jun-2023
• Refers: ASTM E1820-20e1 - Standard Test Method for Measurement of Fracture Toughness
• Effective Date: 01-Jan-2020
• Refers: ASTM E1820-20 - Standard Test Method for Measurement of Fracture Toughness
• Effective Date: 01-Jan-2020
• Refers: ASTM E1921-19b - Standard Test Method for Determination of Reference Temperature, <emph type="bdit">T<inf >o</inf></emph>, for Ferritic Steels in the Transition Range
• Effective Date: 15-Jul-2019
• Refers: ASTM E1921-19be1 - Standard Test Method for Determination of Reference Temperature, <emph type="bdit">T<inf >o</inf></emph>, for Ferritic Steels in the Transition Range
• Effective Date: 15-Jul-2019
• Refers: ASTM A370-19 - Standard Test Methods and Definitions for Mechanical Testing of Steel Products
• Effective Date: 01-Jul-2019
• Refers: ASTM E1921-19a - Standard Test Method for Determination of Reference Temperature, <emph type="bdit">T<inf >o</inf></emph>, for Ferritic Steels in the Transition Range
• Effective Date: 01-May-2019
• Refers: ASTM E1921-19 - Standard Test Method for Determination of Reference Temperature, <emph type="bdit">T<inf >o</inf></emph>, for Ferritic Steels in the Transition Range
• Effective Date: 01-Feb-2019
• Refers: ASTM E1820-18a - Standard Test Method for Measurement of Fracture Toughness
• Effective Date: 01-Nov-2018

Overview

ASTM E2215-19 is a critical standard practice developed by ASTM International for the evaluation of surveillance capsules removed from light-water moderated nuclear power reactor vessels. This document outlines comprehensive guidelines and methods for assessing the effects of neutron irradiation on reactor pressure vessel materials, particularly ferritic steels in the reactor beltline. By providing evaluation criteria for surveillance capsule test specimens and dosimetry, ASTM E2215-19 plays a vital role in ensuring the long-term safety, integrity, and effective operation of nuclear reactors. The insights gained from such surveillance programs support regulatory compliance and plant life extension decisions.

Key Topics

• Evaluation of Surveillance Capsules:
  Defines procedures for testing, analyzing, and documenting the condition and performance of surveillance capsules exposed to irradiation inside reactor vessels.

• Mechanical Testing:
  Specifies tension and Charpy impact tests, fracture toughness, and optional hardness testing for irradiated materials to monitor changes in yield strength, ductility, and energy absorption properties.

• Dosimetry and Irradiation Exposure:
  Outlines methods for measuring and recording neutron exposure, including power history, neutron fluence, and the use of dosimeters. This ensures accurate tracking of irradiation-induced changes over time.

• Assessment of Material Property Changes:
  Guides the evaluation of radiation-induced changes by comparing irradiated and unirradiated test data, especially focusing on index temperatures, upper-shelf energy, and transition temperature shifts.

• Withdrawal Schedule Management:
  Provides best practices for reviewing and updating capsule withdrawal schedules, especially for reactors planning operation beyond their original design life. This includes guidance for supplemental irradiation monitoring.

• Reporting Requirements:
  Details mandatory documentation, including material identification, test procedures, results, and deviations, as well as recommendations for electronic data storage and reporting formats.

Applications

• Nuclear Power Plant Operation and Safety:
  Used by nuclear plant operators, engineers, and regulators to monitor the condition of reactor pressure vessel materials, ensuring ongoing safety and compliance with industry requirements.

• Life Extension and License Renewal:
  Enables accurate projections of material performance, supporting decisions on reactor license renewal and operation beyond initial design life by identifying property degradation trends.

• Surveillance Program Design and Assessment:
  Integral to the broader reactor vessel surveillance program, ASTM E2215-19 ensures uniformity and reliability in testing and assessment across different plants and reactor designs.

• Material Property Analysis for Structural Integrity:
  Assists in the determination of the structural integrity of reactor vessels by allowing assessment of irradiation effects on mechanical properties, such as fracture toughness and ductility.

Related Standards

• ASTM E185: Practice for Design of Surveillance Programs for Light-Water Moderated Nuclear Power Reactor Vessels (companion document for surveillance program design).
• ASTM E636: Guide for Conducting Supplemental Surveillance Tests for Nuclear Power Reactor Vessels.
• ASTM E8/E8M: Test Methods for Tension Testing of Metallic Materials.
• ASTM E23 & A370: Methods for Notched Bar Impact and Mechanical Testing of Steel Products.
• ASTM E1921 & E1820: Test Methods for Determining Reference Temperature and Fracture Toughness.
• ASTM E853 & E900: Analysis and prediction of neutron exposure results and radiation-induced property shifts.
• ASME Boiler and Pressure Vessel Code (Section III and XI): Governing construction rules and analysis criteria for reactor components.

---

Keywords: ASTM E2215, surveillance capsule evaluation, nuclear reactor vessels, light-water reactor, irradiation effects, dosimetry, mechanical testing, material property change, reactor vessel integrity, standard practice, nuclear safety, pressure vessel monitoring

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: E2215 − 19
Standard Practice for
Evaluation of Surveillance Capsules from Light-Water
Moderated Nuclear Power Reactor Vessels
This standard is issued under the fixed designation E2215; 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 2. Referenced Documents
2.1 ASTM Standards:
1.1 This practice covers the evaluation of test specimens
A370 Test Methods and Definitions for Mechanical Testing
and dosimetry from light water moderated nuclear power
of Steel Products
reactor pressure vessel surveillance capsules.
E8/E8M Test Methods for Tension Testing of Metallic Ma-
1.2 Additionally, this practice provides guidance on reas-
terials
sessing withdrawal schedule for design life and operation
E21 TestMethodsforElevatedTemperatureTensionTestsof
beyond design life.
Metallic Materials
1.3 This practice is one of a series of standard practices that E23 Test Methods for Notched Bar Impact Testing of Me-
tallic Materials
outline the surveillance program required for nuclear reactor
pressure vessels. The surveillance program monitors the E170 Terminology Relating to Radiation Measurements and
Dosimetry
irradiation-induced changes in the ferritic steels that comprise
the beltline of a light-water moderated nuclear reactor pressure E185 Practice for Design of Surveillance Programs for
Light-Water Moderated Nuclear Power Reactor Vessels
vessel.
E208 Test Method for Conducting Drop-Weight Test to
1.4 This practice along with its companion surveillance
Determine Nil-Ductility Transition Temperature of Fer-
program practice, Practice E185, is intended for application in
ritic Steels
monitoring the properties of beltline materials in any light-
E509 Guide for In-Service Annealing of Light-Water Mod-
water moderated nuclear reactor.
erated Nuclear Reactor Vessels
1.5 Modifications to the standard test program and supple- E636 Guide for Conducting Supplemental Surveillance
mental tests are described in Guide E636. Tests for Nuclear Power Reactor Vessels
E693 Practice for Characterizing Neutron Exposures in Iron
1.6 The values stated in SI units are to be regarded as the
and Low Alloy Steels in Terms of Displacements Per
standard. The values given in parentheses are for information
Atom (DPA)
only.
E844 Guide for Sensor Set Design and Irradiation for
1.7 This international standard was developed in accor-
Reactor Surveillance
dance with internationally recognized principles on standard-
E853 Practice forAnalysis and Interpretation of Light-Water
ization established in the Decision on Principles for the
Reactor Surveillance Neutron Exposure Results
Development of International Standards, Guides and Recom-
E900 Guide for Predicting Radiation-Induced Transition
mendations issued by the World Trade Organization Technical
Temperature Shift in Reactor Vessel Materials
Barriers to Trade (TBT) Committee.
E1214 Guide for Use of Melt Wire Temperature Monitors
for Reactor Vessel Surveillance
E1253 Guide for Reconstitution of Irradiated Charpy-Sized
Specimens
This practice is under the jurisdiction of ASTM Committee E10 on Nuclear
Technology and Applications and is the direct responsibility of Subcommittee E1820 Test Method for Measurement of Fracture Toughness
E10.02 on Behavior and Use of Nuclear Structural Materials.
CurrenteditionapprovedJune1,2019.PublishedJuly2019.Originallyapproved
in 2002. Last previous edition approved in 2018 as E2215–18. DOI: 10.1520/ For referenced ASTM standards, visit the ASTM website, www.astm.org, or
E2215-19. contact ASTM Customer Service at [e-mail]. For Annual Book of ASTM
Prior to the adoption of these standard practices, surveillance capsule testing Standards volume information, refer to the standard’s Document Summary page on
requirements were only contained in Practice E185. the ASTM website.
Copyright © ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959. United States
E2215 − 19
E1921 Test Method for Determination of Reference fracture appearance obtained from the best-fit (average)
Temperature, T , for Ferritic Steels in the Transition Charpy transition curve.
o
Range
3.1.10 lead factor—the ratio of the average neutron fluence
2.2 ASME Standards:
(E > 1 MeV) of the specimens in a surveillance capsule to the
Boiler and Pressure Vessel Code, Section III Subarticle
peak neutron fluence (E > 1 MeV) of the corresponding
NB-2000, Rules for Construction of Nuclear Facility
material at the ferritic steel reactor pressure vessel inside
Components, Class 1 Components, Materials
surface calculated over the same time period.
Boiler and Pressure Vessel Code, Section XI Nonmandatory
3.1.10.1 Discussion—Changes in the reactor operating pa-
Appendix A, Analysis of Flaws, and Nonmandatory Ap-
rameters and fuel management may cause the lead factor to
pendix G, Fracture Toughness Criteria for Protection
change.
against Failure
3.1.11 limiting materials—typically, the weld and base ma-
3. Terminology terial with the highest predicted transition temperature using
the projected fluence at the end of design life of each material,
3.1 Definitions:
determined by adding the appropriate transition temperature
3.1.1 base metal—as-fabricated plate material or forging
shift (TTS) to the unirradiated RT . Guide E900 describes a
NDT
material other than a weld or its corresponding heat-affected-
method for predicting the TTS. Regulators or other sources
zone (HAZ).
may describe different methods for predicting TTS.
3.1.2 beltline—the irradiated region of the reactor vessel
3.1.12 maximum design fluence (MDF)—the maximum pro-
(shell material including weld seams and plates or forgings)
jected fluence at the inside surface of the ferritic pressure
that directly surrounds the effective height of the active core.
vessel at the end of design life (if clad, MDF is defined at the
Note that materials in regions adjacent to the beltline may
interface of the cladding to the ferritic steel).
sustain sufficient neutron damage to warrant consideration in
3.1.13 reference material—any steel that has been charac-
the selection of surveillance materials.
terized as to the sensitivity of its tensile, impact and fracture
3.1.3 Charpy transition temperature curve—a graphic or
toughness properties to neutron radiation-induced embrittle-
curve-fitted presentation, or both, of absorbed energy, lateral
ment and is included in the Practice E185 surveillance pro-
expansion, or fracture appearance as a function of test
gram.
temperature, extending over a range including the lower shelf
(5 % or less shear fracture appearance), transition region, and 3.1.14 reference temperature (RT )—see subarticle NB-
NDT
2300 of the ASME Boiler and Pressure Vessel Code, Section
the upper shelf (95 % or greater shear fracture appearance).
III, for the definition of RT for unirradiated material based
NDT
3.1.4 Charpy transition temperature shift—the difference in
on Charpy (Test Methods A370) and drop weight tests (Test
the 41 J (30 ft-lbf) index temperatures for the best fit (average)
Method E208). ASME Code Section XI, Appendices A and G
Charpy absorbed energy curve measured before and after
provide an alternative definition for the reference temperature
irradiation. Similar measures of temperature shift can be
(RT ) based on fracture toughness properties (Test Method
To
defined based on other indices in 3.1.3, but the current U.S.
E1921).
industry practice is to use 41 J (30 ft-lbf) and is consistent with
Guide E900. 3.1.15 standby capsule—a surveillance capsule meeting the
recommendations of this practice that is or has been in the
3.1.5 Charpy upper-shelf energy level—the average energy
reactor vessel irradiation location as defined by Practice E185,
value for all Charpy specimen tests (preferably three or more)
but the testing of which is not required by this practice during
whose test temperature is at or above the Charpy upper-shelf
the applicable operating license period.
onset.
3.2 Neutron Exposure Terminology:
3.1.6 Charpy upper-shelf onset—the temperature at which
3.2.1 Definitions of terms related to neutron dosimetry and
the fracture appearance of all Charpy specimens tested is at or
exposure are provided in Terminology E170.
above 95 % shear.
3.1.7 end-of-license (EOL) fluence—the maximum pre-
4. Significance and Use
dicted fluence at the inside surface of the ferritic pressure
vessel(ifclad,theinterfacebetweencladdingandferriticsteel) 4.1 Neutron radiation effects are considered in the design of
corresponding to the end of the applicable operating license light-water moderated nuclear power reactors. Changes in
period. system operating parameters may be made throughout the
service life of the reactor to account for these effects. A
3.1.8 heat-affected-zone (HAZ)—plate material or forging
surveillance program is used to measure changes in the
material extending outward from, but not including, the weld
properties of actual vessel materials due to the irradiation
fusion line in which the microstructure of the base metal has
environment. This practice describes the criteria that should be
been altered by the heat of the welding process.
considered in evaluating surveillance program test capsules.
3.1.9 index temperature—the temperature corresponding to
4.2 Prior to the first issue date of this standard, the design of
a predetermined level of absorbed energy, lateral expansion, or
surveillance programs and the testing of surveillance capsules
werebothcoveredinasinglestandard,PracticeE185.Between
AvailablefromAmericanSocietyofMechanicalEngineers,ThirdParkAvenue,
New York, NY 10016. its provisional adoption in 1961 and its replacement linked to
E2215 − 19
this standard, Practice E185 was revised many times (1966, 5.3 Irradiation Temperature History—The average capsule
1970, 1973, 1979, 1982, 1993 and 1998). Therefore, capsules temperature during full power operation shall be estimated for
from surveillance programs that were designed and imple- each reactor fuel cycle experienced by the capsule. The local
mented under early versions of the standard were often tested reactor coolant temperature may be used as a reasonable
after substantial changes to the standard had been adopted. For approximation, although gamma-ray heating should be consid-
clarity, the standard practice for surveillance programs has ered if it leads to a significant temperature difference. In a
been divided into the new Practice E185 that covers the design typical pressurized water reactor, the coolant inlet temperature
of new surveillance programs and this standard practice that may be used as an estimate of the capsule irradiation tempera-
covers the testing and evaluation of surveillance capsules. ture using a time-weighted average (see Guide E900). In a
Modifications to the standard test program and supplemental typicalboilingwaterreactor,therecirculationtemperaturemay
tests are described in Guide E636. be used as an estimate of the capsule irradiation temperature.
5.4 Peak Temperature—Temperature monitors shall be ex-
4.3 This practice is intended to cover testing and evaluation
amined and any evidence of melting shall be recorded in
of all light-water moderated reactor pressure vessel surveil-
accordance with Guide E1214.
lance capsules.The practice is applicable to testing of capsules
from surveillance programs designed and implemented under
6. Measurement of Irradiation Exposure
all previous versions of Practice E185.
6.1 The monthly power history of the reactor for all cycles
4.4 The radiation-induced changes in the properties of the
prior to capsule removal shall be recorded. Other data that are
reactor pressure vessel are generally monitored by measuring
needed on a fuel-cycle-specific basis include: assembly-wise
the index temperatures, the upper-shelf energy and the tensile
core power distributions, including enrichments and burnups,
properties of specimens from the surveillance program cap-
axial core power distributions, axial core void distributions
sules. The significance of these radiation-induced changes is
(BWRs only), and core and downcomer water temperatures.
described in Practice E185.
Otherkeychangesthatneedtoberecordedincludetheaddition
4.5 Alternative methods exist for testing surveillance cap-
or removal of flux suppression rods or shield rods, uprates or
sule materials. Some supplemental and alternative testing
derates of reactor power, and other reactor modifications such
methods are available as indicated in Guide E636. Direct
as adding neutron shielding or the removal or addition of
measurement of the fracture toughness is also feasible using
structures such as a thermal shield. Fuel assembly, reactor
the T Reference Temperature method defined in Test Method
o internals, and reactor pressure vessel dimensional information
E1921 or J-integral techniques defined in Test Method E1820.
also need to be recorded. Surveillance capsule locations and
Additionally, hardness testing can be used to supplement
movements: including storage periods outside the reactor, shall
standard methods as a means of monitoring the irradiation
be provided for the evaluation of irradiation exposure.
response of the materials.
6.2 Practice E853 describes practices for determining the
4.6 Practice E853 describes a methodology that may be
neutron fluence rate, neutron energy spectrum and neutron
used in the analysis and interpretation of neutron dosimetry
fluence of the surveillance specimens and the corresponding
data and the determination of neutron fluence. Regulators or
maximum values for the reactor vessel. Regulators or other
other sources may describe different methods.
sources may describe different methods.
4.7 Guide E900 describes a method for predicting the TTS. 6.3 Neutron fluence rate and fluence values (E > 1 MeV)
and dpa rate and dpa values per Practice E693 (or alternatives
Regulators or other sources may describe different methods for
predicting TTS. in regulatory guidance or prescribed by regulations) shall be
determined and recorded using a calculated spectrum adjusted
4.8 Guide E509 provides direction for development of a
or validated by dosimetry measurements.
procedure for conducting an in-service thermal anneal of a
light-watercoolednuclearreactorvesselanddemonstratingthe
7. Measurement of Mechanical Properties
effectiveness of the procedure including a post-annealing
7.1 Generally, all the materials contained in the capsule
vessel radiation surveillance program.
except the HAZ specimens (if included) should be tested.
Testing of the HAZ specimens is optional.(1)
5. Determination of Capsule Condition
7.2 Tension Tests:
5.1 Visual Examination—A complete visual exam of the
7.2.1 Method—Tension testing shall be conducted in accor-
capsuleconditionshouldbecompleteduponreceiptandduring
dance with Test Methods E8/E8M and E21.
disassembly at the testing laboratory. External identification
7.2.2 Test Temperature—In general, the test temperatures
marks on the capsule shall be verified. Signs of damage or
for each material shall include room temperature and reactor
degradation of the capsule exterior shall be recorded.
vesselservicetemperature.Otherspecimensshouldberetained
5.2 Capsule Content—The specimen loading pattern should
for tension testing at possible future fracture toughness test
be compared to the capsule fabrication records and any
deviations shall be noted. Any evidence of corrosion or other
damage to the specimens shall also be noted. The condition of
The boldface numbers in parentheses refer to a list of references at the end of
any temperature monitors shall be noted and recorded. this standard.
E2215 − 19
temperature(s). Specific consideration should be given to the be performed after a thorough evaluation of the potential
specific temperatures at which unirradiated specimens have usefulness of the specimen materials.
been tested.
7.2.3 Measurements—Determine yield strength, tensile 8. Evaluation of Test Data
strength, total and uniform elongation and reduction of area.
8.1 Tension Tests:
7.3 Charpy Tests: 8.1.1 Determine the amount of radiation-induced strength-
ening and loss of ductility by comparing irradiated test results
7.3.1 Method—Charpy tests shall be conducted in accor-
with unirradiated data from the surveillance capsule documen-
dance with Test Methods and Definitions A370 and Test
tation package.
Method E23. Instrumented tests are recommended and should
be performed in accordance with Guide E636. Broken Charpy
8.2 Charpy Tests:
specimens may be reconstituted for supplemental testing in
8.2.1 CurveFitting—Averagecurvesshallbedrawnthrough
accordance with Guide E1253.
the Charpy data to display the Charpy impact energy, lateral
7.3.2 Test Temperature—Specimens for each material shall
expansion and percent shear fracture appearance as a function
be tested at temperatures selected to define the full Charpy
of the test temperature. A similar analysis of unirradiated
energy transition curve. Particular emphasis should be placed
Charpy data from the surveillance capsule documentation
on defining the 41 J (30 ft-lbf) index temperature and the
should also be performed. The preferred method for determin-
upper-shelf energy level. It is recommended that upper-shelf
ing the average curves is fitting to a hyperbolic tangent
Charpy tests be conducted using at least three specimens tested
function as discussed in Appendix X1.
and evaluated in accordance with 3.1.5 of this practice.
8.2.2 Occasionally a single data point will unduly influence
Instrumented tests are recommended and should be performed
the average curve. In this case, the test record and specimen
in accordance with Guide E636.
should be examined for possible causes of discrepancy and its
7.3.3 Measurements—For each test specimen, measure the
disposition documented.
impact energy, lateral expansion, and percent shear fracture
8.2.3 Index Temperatures—Charpy index temperatures shall
appearance.
bedeterminedforthe41J(30ft-lbf)energyleveland0.89mm
(35 mils) lateral expansion level. Optionally, the fracture
7.4 Hardness Tests (Optional)—Hardness tests may be per-
appearance transition temperature corresponding to 50 % shear
formed on irradiated Charpy specimens. The measurements
fracture can be determined. Radiation-induced shifts in the
shall be taken (prior to Charpy testing, if possible, to avoid
index temperatures shall be determined by subtracting the
samplingmaterialdeformedbythetest)inareasawayfromthe
measured unirradiated index temperatures from the irradiated
fracture zone or the edges of the specimens. The tests shall be
index temperatures. If the differences among these three shift
conducted in accordance with Test Methods and Definitions
measurements exceed 15°C, then the test records and speci-
A370.
mens should be examined for possible causes of discrepancy
7.5 Fracture Toughness Tests (Optional):
and the outcome of the examination documented.
7.5.1 Specimens—Fracture toughness tests may be con-
8.2.4 Upper-Shelf Energy—The Charpy upper-shelf energy
ducted following Guide E636 using either fracture mechanics
shouldbedeterminedaccordingtothedefinitiongivenin3.1.5.
specimens from the surveillance capsule or broken Charpy
The radiation-induced change in the upper-shelf energy shall
specimens that have been reconstituted and precracked. Proce-
bedeterminedbycomparingthisdatatounirradiateddatafrom
dures for reconstitution of Charpy specimens are given in
the surveillance capsule documentation.
Guide E1253.
8.3 ReferenceMaterial—Ifreferencematerialspecimensare
7.5.2 Upper-Shelf Fracture Toughness—Testing to charac-
included in the surveillance capsule, they shall be tested and
terizeupper-shelftoughnessusingtheJ-integralmethodshould
evaluated. The measured irradiation response of the reference
be conducted in accordance with Test Method E1820.
material specimens should fall within the scatter band of the
7.5.3 Transition Fracture Toughness—The reference tem-
pre-existing database. In cases where the reference material
perature for ferritic steels in the transition range, T , can be
o
test results exhibit excessive scatter relative to the available
established using the methodology provided in Test Method
data, the source of the scatter should be investigated. Potential
E1921.
reasons that can be investigated include deviations from the
7.6 Retention of Test Specimens—It is recommended that all
expected surveillance capsule exposure conditions, a lack of
broken and unbroken test specimens be maintained in good
uniformityofpropertiesinthereferencematerialitself,orboth.
condition and retained. These test specimens may be useful in
8.4 Hardness Tests (Optional)—The hardness data may be
the event that additional analysis is required to explain anoma-
correlated to the yield or tensile strength of the material, or
lous results. Identification of all test specimens shall be
other parameters. Justification for any correlation used shall be
maintained. After it is determined that additional testing or
provided with the report.
analysis to explain anomalous results is not required, then it is
8.5 Fracture Toughness Tests (Optional):
recommended that specimens be either retained or used for
appropriate research to increase understanding of embrittle-
ment or for direct use or potential reconstitution to support
reactor vessel material surveillance programs during extended
See for example: ASTM Data Series DS54(2); NUREG/CR-4947(3) on HSST
operating periods. Final disposition of specimens should only plates; and IAEA-TECDOC-1230(4), on the JRQ plate.
E2215 − 19
8.5.1 Upper-Shelf Fracture Toughness—The resistance to 9.4 Anticipated Operation Beyond Design Life:
crack initiation and extension on the upper shelf may be
9.4.1 When operation beyond design life is anticipated, a
expressed in terms of the J-integral as described in Test
plan for reactor vessel surveillance should be developed to
Method E1820.
ensure that the vessel beltline is appropriately monitored
8.5.2 Transition Fracture Toughness—An appropriate refer-
throughout the period of operation. This may include fabrica-
ence temperature for fracture toughness in the transition region
tionandinsertionofanewcapsule,movinganexistingcapsule
can be determined using the procedure in Test Method E1921.
to a higher lead factor location, a change in status of a standby
This reference transition temperature can be used to define an
capsule to one scheduled for withdrawal and testing, or
alternate reference temperature (RT in place of RT )as
participation in an integrated surveillance program.
To NDT
defined in ASME Code Section XI, Appendices A and G.
9.4.2 Update EOL fluence, TTS, EOL reactor vessel mate-
rial property projections and limiting material for the operating
9. Withdrawal Schedule Review
period beyond design life.
9.1 The primary consideration in the review of the with-
9.4.3 The goal is to have limiting beltline material (or a
drawal schedule shall be ensuring that the vessel is appropri-
surrogate material, if this is not practical) index temperature
ately monitored throughout its projected design life. This
measurements at a fluence greater than the projected EOL
should include
...

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: E2215 − 18 E2215 − 19
Standard Practice for
Evaluation of Surveillance Capsules from Light-Water
Moderated Nuclear Power Reactor Vessels
This standard is issued under the fixed designation E2215; 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 evaluation of test specimens and dosimetry from light water moderated nuclear power reactor
pressure vessel surveillance capsules.
1.2 Additionally, this practice provides guidance on reassessing withdrawal schedule for design life and operation beyond
design life.
1.3 This practice is one of a series of standard practices that outline the surveillance program required for nuclear reactor
pressure vessels. The surveillance program monitors the irradiation-induced changes in the ferritic steels that comprise the beltline
of a light-water moderated nuclear reactor pressure vessel.
1.4 This practice along with its companion surveillance program practice, Practice E185, is intended for application in
monitoring the properties of beltline materials in any light-water moderated nuclear reactor.
1.5 Modifications to the standard test program and supplemental tests are described in Guide E636.
1.6 The values stated in SI units are to be regarded as the standard. The values given in parentheses are for information only.
1.7 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:
A370 Test Methods and Definitions for Mechanical Testing of Steel Products
E8/E8M Test Methods for Tension Testing of Metallic Materials
E21 Test Methods for Elevated Temperature Tension Tests of Metallic Materials
E23 Test Methods for Notched Bar Impact Testing of Metallic Materials
E170 Terminology Relating to Radiation Measurements and Dosimetry
E185 Practice for Design of Surveillance Programs for Light-Water Moderated Nuclear Power Reactor Vessels
E208 Test Method for Conducting Drop-Weight Test to Determine Nil-Ductility Transition Temperature of Ferritic Steels
E509 Guide for In-Service Annealing of Light-Water Moderated Nuclear Reactor Vessels
E636 Guide for Conducting Supplemental Surveillance Tests for Nuclear Power Reactor Vessels
E693 Practice for Characterizing Neutron Exposures in Iron and Low Alloy Steels in Terms of Displacements Per Atom (DPA)
E844 Guide for Sensor Set Design and Irradiation for Reactor Surveillance
E853 Practice for Analysis and Interpretation of Light-Water Reactor Surveillance Neutron Exposure Results
E900 Guide for Predicting Radiation-Induced Transition Temperature Shift in Reactor Vessel Materials
E1214 Guide for Use of Melt Wire Temperature Monitors for Reactor Vessel Surveillance
E1253 Guide for Reconstitution of Irradiated Charpy-Sized Specimens
E1820 Test Method for Measurement of Fracture Toughness
This practice is under the jurisdiction of ASTM Committee E10 on Nuclear Technology and Applications and is the direct responsibility of Subcommittee E10.02 on
Behavior and Use of Nuclear Structural Materials.
Current edition approved Aug. 1, 2018June 1, 2019. Published August 2018July 2019. Originally approved in 2002. Last previous edition approved in 20162018 as
E2215–16.–18. DOI: 10.1520/E2215-18.10.1520/E2215-19.
Prior to the adoption of these standard practices, surveillance capsule testing requirements were only contained in Practice E185.
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
E2215 − 19
E1921 Test Method for Determination of Reference Temperature, T , for Ferritic Steels in the Transition Range
o
2.2 ASME Standards:
Boiler and Pressure Vessel Code, Section III Subarticle NB-2000, Rules for Construction of Nuclear Facility Components, Class
1 Components, Materials
Boiler and Pressure Vessel Code, Section XI Nonmandatory Appendix A, Analysis of Flaws, and Nonmandatory Appendix G,
Fracture Toughness Criteria for Protection against Failure
3. Terminology
3.1 Definitions:
3.1.1 base metal—as-fabricated plate material or forging material other than a weld or its corresponding heat-affected-zone
(HAZ).
3.1.2 beltline—the irradiated region of the reactor vessel (shell material including weld seams and plates or forgings) that
directly surrounds the effective height of the active core. Note that materials in regions adjacent to the beltline may sustain
sufficient neutron damage to warrant consideration in the selection of surveillance materials.
3.1.3 Charpy transition temperature curve—a graphic or curve-fitted presentation, or both, of absorbed energy, lateral
expansion, or fracture appearance as a function of test temperature, extending over a range including the lower shelf (5 % or less
shear fracture appearance), transition region, and the upper shelf (95 % or greater shear fracture appearance).
3.1.4 Charpy transition temperature shift—the difference in the 41 J (30 ft-lbf) index temperatures for the best fit (average)
Charpy absorbed energy curve measured before and after irradiation. Similar measures of temperature shift can be defined based
on other indices in 3.1.3, but the current U.S. industry practice is to use 41 J (30 ft-lbf) and is consistent with Guide E900.
3.1.5 Charpy upper-shelf energy level—the average energy value for all Charpy specimen tests (preferably three or more) whose
test temperature is at or above the Charpy upper-shelf onset; specimens tested at temperatures greater than 83°C (150°F) above
the Charpy upper-shelf onset shall not be included, unless no data are available between the onset temperature and onset +83°C
(+150°F).onset.
3.1.6 Charpy upper-shelf onset—the temperature at which the fracture appearance of all Charpy specimens tested is at or above
95 % shear.
3.1.7 end-of-license (EOL) fluence—the maximum predicted fluence at the inside surface of the ferritic pressure vessel (if clad,
the interface between cladding and ferritic steel) corresponding to the end of the applicable operating license period.
3.1.8 heat-affected-zone (HAZ)—plate material or forging material extending outward from, but not including, the weld fusion
line in which the microstructure of the base metal has been altered by the heat of the welding process.
3.1.9 index temperature—the temperature corresponding to a predetermined level of absorbed energy, lateral expansion, or
fracture appearance obtained from the best-fit (average) Charpy transition curve.
3.1.10 lead factor—the ratio of the average neutron fluence (E > 1 MeV) of the specimens in a surveillance capsule to the peak
neutron fluence (E > 1 MeV) of the corresponding material at the ferritic steel reactor pressure vessel inside surface calculated over
the same time period.
3.1.10.1 Discussion—
Changes in the reactor operating parameters and fuel management may cause the lead factor to change.
3.1.11 limiting materials—typically, the weld and base material with the highest predicted transition temperature using the
projected fluence at the end of design life of each material, determined by adding the appropriate transition temperature shift (TTS)
to the unirradiated RT . Guide E900 describes a method for predicting the TTS. Regulators or other sources may describe
NDT
different methods for predicting TTS.
3.1.12 maximum design fluence (MDF)—the maximum projected fluence at the inside surface of the ferritic pressure vessel at
the end of design life (if clad, MDF is defined at the interface of the cladding to the ferritic steel).
3.1.13 reference material—any steel that has been characterized as to the sensitivity of its tensile, impact and fracture toughness
properties to neutron radiation-induced embrittlement and is included in the Practice E185 surveillance program.
3.1.14 reference temperature (RT ) —see subarticle NB-2300 of the ASME Boiler and Pressure Vessel Code, Section III, for
NDT
the definition of RT for unirradiated material based on Charpy (Test Methods A370) and drop weight tests (Test Method E208).
NDT
ASME Code Section XI, Appendices A and G provide an alternative definition for the reference temperature (RT ) based on
To
fracture toughness properties (Test Method E1921).
Available from American Society of Mechanical Engineers, Third Park Avenue, New York, NY 10016.
E2215 − 19
3.1.15 standby capsule—a surveillance capsule meeting the recommendations of this practice that is or has been in the reactor
vessel irradiation location as defined by Practice E185, but the testing of which is not required by this practice during the applicable
operating license period.
3.2 Neutron Exposure Terminology:
3.2.1 Definitions of terms related to neutron dosimetry and exposure are provided in Terminology E170.
4. Significance and Use
4.1 Neutron radiation effects are considered in the design of light-water moderated nuclear power reactors. Changes in system
operating parameters may be made throughout the service life of the reactor to account for these effects. A surveillance program
is used to measure changes in the properties of actual vessel materials due to the irradiation environment. This practice describes
the criteria that should be considered in evaluating surveillance program test capsules.
4.2 Prior to the first issue date of this standard, the design of surveillance programs and the testing of surveillance capsules were
both covered in a single standard, Practice E185. Between its provisional adoption in 1961 and its replacement linked to this
standard, Practice E185 was revised many times (1966, 1970, 1973, 1979, 1982, 1993 and 1998). Therefore, capsules from
surveillance programs that were designed and implemented under early versions of the standard were often tested after substantial
changes to the standard had been adopted. For clarity, the standard practice for surveillance programs has been divided into the
new Practice E185 that covers the design of new surveillance programs and this standard practice that covers the testing and
evaluation of surveillance capsules. Modifications to the standard test program and supplemental tests are described in Guide E636.
4.3 This practice is intended to cover testing and evaluation of all light-water moderated reactor pressure vessel surveillance
capsules. The practice is applicable to testing of capsules from surveillance programs designed and implemented under all previous
versions of Practice E185.
4.4 The radiation-induced changes in the properties of the reactor pressure vessel are generally monitored by measuring the
index temperatures, the upper-shelf energy and the tensile properties of specimens from the surveillance program capsules. The
significance of these radiation-induced changes is described in Practice E185.
4.5 Alternative methods exist for testing surveillance capsule materials. Some supplemental and alternative testing methods are
available as indicated in Guide E636. Direct measurement of the fracture toughness is also feasible using the T Reference
o
Temperature method defined in Test Method E1921 or J-integral techniques defined in Test Method E1820. Additionally, hardness
testing can be used to supplement standard methods as a means of monitoring the irradiation response of the materials.
4.6 Practice E853 describes a methodology that may be used in the analysis and interpretation of neutron dosimetry data and
the determination of neutron fluence. Regulators or other sources may describe different methods.
4.7 Guide E900 describes a method for predicting the TTS. Regulators or other sources may describe different methods for
predicting TTS.
4.8 Guide E509 provides direction for development of a procedure for conducting an in-service thermal anneal of a light-water
cooled nuclear reactor vessel and demonstrating the effectiveness of the procedure including a post-annealing vessel radiation
surveillance program.
5. Determination of Capsule Condition
5.1 Visual Examination—A complete visual exam of the capsule condition should be completed upon receipt and during
disassembly at the testing laboratory. External identification marks on the capsule shall be verified. Signs of damage or degradation
of the capsule exterior shall be recorded.
5.2 Capsule Content—The specimen loading pattern should be compared to the capsule fabrication records and any deviations
shall be noted. Any evidence of corrosion or other damage to the specimens shall also be noted. The condition of any temperature
monitors shall be noted and recorded.
5.3 Irradiation Temperature History—The average capsule temperature during full power operation shall be estimated for each
reactor fuel cycle experienced by the capsule. The local reactor coolant temperature may be used as a reasonable approximation,
although gamma-ray heating should be considered if it leads to a significant temperature difference. In a typical pressurized water
reactor, the coolant inlet temperature may be used as an estimate of the capsule irradiation temperature using a time-weighted
average (see Guide E900). In a typical boiling water reactor, the recirculation temperature may be used as an estimate of the
capsule irradiation temperature.
5.4 Peak Temperature—Temperature monitors shall be examined and any evidence of melting shall be recorded in accordance
with Guide E1214.
6. Measurement of Irradiation Exposure
6.1 The monthly power history of the reactor for all cycles prior to capsule removal shall be recorded. Other data that are needed
on a fuel-cycle-specific basis include: assembly-wise core power distributions, including enrichments and burnups, axial core
E2215 − 19
power distributions, axial core void distributions (BWRs only), and core and downcomer water temperatures. Other key changes
that need to be recorded include the addition or removal of flux suppression rods or shield rods, uprates or derates of reactor power,
and other reactor modifications such as adding neutron shielding or the removal or addition of structures such as a thermal shield.
Fuel assembly, reactor internals, and reactor pressure vessel dimensional information also need to be recorded. Surveillance
capsule locations and movements: including storage periods outside the reactor, shall be provided for the evaluation of irradiation
exposure.
6.2 Practice E853 describes practices for determining the neutron fluence rate, neutron energy spectrum and neutron fluence of
the surveillance specimens and the corresponding maximum values for the reactor vessel. Regulators or other sources may describe
different methods.
6.3 Neutron fluence rate and fluence values (E > 1 MeV) and dpa rate and dpa values per Practice E693 (or alternatives in
regulatory guidance or prescribed by regulations) shall be determined and recorded using a calculated spectrum adjusted or
validated by dosimetry measurements.
7. Measurement of Mechanical Properties
7.1 Generally, all the materials contained in the capsule except the HAZ specimens (if included) should be tested. Testing of
the HAZ specimens is optional.(1)
7.2 Tension Tests:
7.2.1 Method—Tension testing shall be conducted in accordance with Test Methods E8/E8M and E21.
7.2.2 Test Temperature—In general, the test temperatures for each material shall include room temperature and reactor vessel
service temperature. Other specimens should be retained for tension testing at possible future fracture toughness test
temperature(s). Specific consideration should be given to the specific temperatures at which unirradiated specimens have been
tested.
7.2.3 Measurements—Determine yield strength, tensile strength, total and uniform elongation and reduction of area.
7.3 Charpy Tests:
7.3.1 Method—Charpy tests shall be conducted in accordance with Test Methods and Definitions A370 and Test Method E23.
Instrumented tests are recommended and should be performed in accordance with Guide E636. Broken Charpy specimens may be
reconstituted for supplemental testing in accordance with Guide E1253.
7.3.2 Test Temperature—Specimens for each material shall be tested at temperatures selected to define the full Charpy energy
transition curve. Particular emphasis should be placed on defining the 41 J (30 ft-lbf) index temperature and the upper-shelf energy
level. It is recommended that upper-shelf Charpy tests be conducted using at least three specimens tested and evaluated in
accordance with 3.1.5 of this practice. Instrumented tests are recommended and should be performed in accordance with Guide
E636.
7.3.3 Measurements—For each test specimen, measure the impact energy, lateral expansion, and percent shear fracture
appearance.
7.4 Hardness Tests (Optional)—Hardness tests may be performed on irradiated Charpy specimens. The measurements shall be
taken (prior to Charpy testing, if possible, to avoid sampling material deformed by the test) in areas away from the fracture zone
or the edges of the specimens. The tests shall be conducted in accordance with Test Methods and Definitions A370.
7.5 Fracture Toughness Tests (Optional):
7.5.1 Specimens—Fracture toughness tests may be conducted following Guide E636 using either fracture mechanics specimens
from the surveillance capsule or broken Charpy specimens that have been reconstituted and precracked. Procedures for
reconstitution of Charpy specimens are given in Guide E1253.
7.5.2 Upper-Shelf Fracture Toughness—Testing to characterize upper-shelf toughness using the J-integral method should be
conducted in accordance with Test Method E1820.
7.5.3 Transition Fracture Toughness—The reference temperature for ferritic steels in the transition range, T , can be established
o
using the methodology provided in Test Method E1921.
7.6 Retention of Test Specimens—It is recommended that all broken and unbroken test specimens be maintained in good
condition and retained. These test specimens may be useful in the event that additional analysis is required to explain anomalous
results. Identification of all test specimens shall be maintained. After it is determined that additional testing or analysis to explain
anomalous results is not required, then it is recommended that specimens be either retained or used for appropriate research to
increase understanding of embrittlement or for direct use or potential reconstitution to support reactor vessel material surveillance
programs during extended operating periods. Final disposition of specimens should only be performed after a thorough evaluation
of the potential usefulness of the specimen materials.
The boldface numbers in parentheses refer to a list of references at the end of this standard.
E2215 − 19
8. Evaluation of Test Data
8.1 Tension Tests:
8.1.1 Determine the amount of radiation-induced strengthening and loss of ductility by comparing irradiated test results with
unirradiated data from the surveillance capsule documentation package.
8.2 Charpy Tests:
8.2.1 Curve Fitting—Average curves shall be drawn through the Charpy data to display the Charpy impact energy, lateral
expansion and percent shear fracture appearance as a function of the test temperature. A similar analysis of unirradiated Charpy
data from the surveillance capsule documentation should also be performed. The preferred method for determining the average
curves is fitting to a hyperbolic tangent function as discussed in Appendix X1.
8.2.2 Occasionally a single data point will unduly influence the average curve. In this case, the test record and specimen should
be examined for possible causes of discrepancy and its disposition documented.
8.2.3 Index Temperatures—Charpy index temperatures shall be determined for the 41 J (30 ft-lbf) energy level and 0.89 mm (35
mils) lateral expansion level. Optionally, the fracture appearance transition temperature corresponding to 50 % shear fracture can
be determined. Radiation-induced shifts in the index temperatures shall be determined by subtracting the measured unirradiated
index temperatures from the irradiated index temperatures. If the differences among these three shift measurements exceed 15°C,
then the test records and specimens should be examined for possible causes of discrepancy and the outcome of the examination
documented.
8.2.4 Upper-Shelf Energy—The Charpy upper-shelf energy should be determined according to the definition given in 3.1.5. The
radiation-induced change in the upper-shelf energy shall be determined by comparing this data to unirradiated data from the
surveillance capsule documentation.
8.3 Reference Material—If reference material specimens are included in the surveillance capsule, they shall be tested and
evaluated. The measured irradiation response of the reference material specimens should fall within the scatter band of the
pre-existing database. In cases where the reference material test results exhibit excessive scatter relative to the available data, the
source of the scatter should be investigated. Potential reasons that can be investigated include deviations from the expected
surveillance capsule exposure conditions, a lack of uniformity of properties in the reference material itself, or both.
8.4 Hardness Tests (Optional)—The hardness data may be correlated to the yield or tensile strength of the material, or other
parameters. Justification for any correlation used shall be provided with the report.
8.5 Fracture Toughness Tests (Optional):
8.5.1 Upper-Shelf Fracture Toughness—The resistance to crack initiation and extension on the upper shelf may be expressed
in terms of the J-integral as described in Test Method E1820.
8.5.2 Transition Fracture Toughness—An appropriate reference temperature for fracture toughness in the transition region can
be determined using the procedure in Test Method E1921. This reference transition temperature can be used to define an alternate
reference temperature (RT in place of RT ) as defined in ASME Code Section XI, Appendices A and G.
To NDT
9. Withdrawal Schedule Review
9.1 The primary consideration in the review of the withdrawal schedule shall be ensuring that the vessel is appropriately
monitored throughout its projected design life. This should include a review of the original objectives of the surveillance program
and the adequacy of the program to meet future needs. This shall also include monitoring the neutron exposure of the reactor vessel
throughout its projected design life using a combination of neutron fluence tracking analysis methods and fluence measurements.
The fluence measurements may consist of both in-vessel and ex-vessel neutron dosimetry. This practice provides guidelines to aid
in that analysis. The circumstances of any particular reactor surveillance program may require considerations of factors beyond
these guidelines.
9.2 The withdrawal schedule shall be reviewed upon completion of testing of each of the surveillance capsules. Proposed
adaptations must accommodate the restrictions imposed by the design of the original surveillance program. The number and
contents of the ca
...