Appendix H to Part 50 - Reactor Vessel Material Surveillance Program Requirements
10:1.0.1.1.30.0.117.88.30 : Appendix H
Appendix H to Part 50 - Reactor Vessel Material Surveillance
Program Requirements I. Introduction II. Definitions III.
Surveillance Program Criteria IV. Report of Test Results I.
Introduction
The purpose of the material surveillance program required by
this appendix is to monitor changes in the fracture toughness
properties of ferritic materials in the reactor vessel beltline
region of light water nuclear power reactors which result from
exposure of these materials to neutron irradiation and the thermal
environment. Under the program, fracture toughness test data are
obtained from material specimens exposed in surveillance capsules,
which are withdrawn periodically from the reactor vessel. These
data will be used as described in section IV of appendix G to part
50.
ASTM E 185-73, “Standard Recommended Practice for Surveillance
Tests for Nuclear Reactor Vessels”; ASTM E 185-79, “Standard
Practice for Conducting Surveillance Tests for Light-Water Cooled
Nuclear Power Reactor Vessels”; and ASTM E 185-82, “Standard
Practice for Conducting Surveillance Tests for Light-Water Cooled
Nuclear Power Reactor Vessels”; which are referenced in the
following paragraphs, have been approved for incorporation by
reference by the Director of the Federal Register. Copies of ASTM E
185-73, -79, and -82, may be purchased from the American Society
for Testing and Materials, 1916 Race Street, Philadelphia, PA 19103
and are available for inspection at the NRC Library, 11545
Rockville Pike, Two White Flint North, Rockville, MD
20852-2738.
II. Definitions
All terms used in this appendix have the same meaning as in
appendix G.
III. Surveillance Program Criteria
A. No material surveillance program is required for reactor
vessels for which it can be conservatively demonstrated by
analytical methods applied to experimental data and tests performed
on comparable vessels, making appropriate allowances for all
uncertainties in the measurements, that the peak neutron fluence at
the end of the design life of the vessel will not exceed 10 17 n/cm
2 (E >1 MeV).
B. Reactor vessels that do not meet the conditions of paragraph
III.A of this appendix must have their beltline materials monitored
by a surveillance program complying with ASTM E 185, as modified by
this appendix.
1. The design of the surveillance program and the withdrawal
schedule must meet the requirements of the edition of the ASTM E
185 that is current on the issue date of the ASME code to which the
reactor vessel was purchased; for reactor vessels purchased after
1982, the design of the surveillance program and the withdrawal
schedule must meet the requirements of ASTM E 185-82. For reactor
vessels purchased in or before 1982, later editions of ASTM E 185
may be used, but including only those editions through 1982. For
each capsule withdrawal, the test procedures and reporting
requirements must meet the requirements of the ASTM E 185 to the
extent practicable for the configuration of the specimens in the
capsule. If any of the optional provisions in paragraphs III.B.4(a)
through (d) of this section are implemented in lieu of ASTM E 185,
the number of specimens included or tested in the surveillance
program shall be adjusted as specified in paragraphs III.B.4(a)
through (d) of this section.
2. Surveillance specimen capsules must be located near the
inside vessel wall in the beltline region so that the specimen
irradiation history duplicates, to the extent practicable within
the physical constraints of the system, the neutron spectrum,
temperature history, and maximum neutron fluence experienced by the
reactor vessel inner surface. If the capsule holders are attached
to the vessel wall or to the vessel cladding, construction and
inservice inspection of the attachments and attachment welds must
be done according to the requirements for permanent structural
attachments to reactor vessels given in Sections III and XI of the
American Society of Mechanical Engineers Boiler and Pressure Vessel
Code (ASME Code). The design and location of the capsule holders
must permit insertion of replacement capsules. Accelerated
irradiation capsules may be used in addition to the required number
of surveillance capsules.
3. A proposed withdrawal schedule must be submitted with a
technical justification as specified in § 50.4. The proposed
schedule must be approved prior to implementation.
4. Optional provisions. As used in this section, references to
ASTM E 185 include the edition of ASTM E 185 that is current on the
issue date of the ASME Code to which the reactor vessel was
purchased through the 1982 edition.
(a) First Provision: Heat-Affected Zone Specimens - The
inclusion or testing of weld heat-affected zone Charpy impact
specimens within the surveillance program as specified in ASTM E
185 is optional.
(b) Second Provision: Tension Specimens - If this
provision is implemented, the minimum number of tension specimens
to be included and tested in the surveillance program shall be as
specified in paragraphs III.B.4(b)(i) and (ii) of this section.
(i) Unirradiated Tension Specimens - Two tension specimens from
each base and weld material required by ASTM E 185 shall be tested,
with one specimen tested at room temperature and the other specimen
tested at the service temperature; and
(ii) Irradiated Tension Specimens - Two tension specimens from
each base and weld material required by ASTM E 185 shall be
included in each surveillance capsule and tested, with one specimen
tested at room temperature and the other specimen tested at the
service temperature.
(c) Third Provision: Correlation Monitor Materials - The
testing of correlation monitor material specimens within the
surveillance program as specified in ASTM E 185 is optional.
(d) Fourth Provision: Thermal Monitor - The inclusion or
examination of thermal monitors within the surveillance program as
specified in ASTM E 185 is optional.
C. Requirements for an Integrated Surveillance Program.
1. In an integrated surveillance program, the representative
materials chosen for surveillance for a reactor are irradiated in
one or more other reactors that have similar design and operating
features. Integrated surveillance programs must be approved by the
Director, Office of Nuclear Reactor Regulation, on a case-by-case
basis. Criteria for approval include the following:
a. The reactor in which the materials will be irradiated and the
reactor for which the materials are being irradiated must have
sufficiently similar design and operating features to permit
accurate comparisons of the predicted amount of radiation
damage.
b. Each reactor must have an adequate dosimetry program.
c. There must be adequate arrangement for data sharing between
plants.
d. There must be a contingency plan to assure that the
surveillance program for each reactor will not be jeopardized by
operation at reduced power level or by an extended outage of
another reactor from which data are expected.
e. There must be substantial advantages to be gained, such as
reduced power outages or reduced personnel exposure to radiation,
as a direct result of not requiring surveillance capsules in all
reactors in the set.
2. No reduction in the requirements for number of materials to
be irradiated, specimen types, or number of specimens per reactor
is permitted.
3. After (the effective date of this section), no reduction in
the amount of testing is permitted unless previously authorized by
the Director, Office of Nuclear Reactor Regulation.
IV. Report of Test Results
A. Each capsule withdrawal and the test results must be the
subject of a summary technical report to be submitted, as specified
in § 50.4, within eighteen months of the date of capsule
withdrawal, unless an extension is granted by the Director, Office
of Nuclear Reactor Regulation.
B. The report must include the data required by ASTM E 185, as
specified in paragraph III.B.1 of this appendix, and the results of
all fracture toughness tests conducted on the beltline materials in
the irradiated and unirradiated conditions.
C. If a change in the Technical Specifications is required,
either in the pressure-temperature limits or in the operating
procedures required to meet the limits, the expected date for
submittal of the revised Technical Specifications must be provided
with the report.
[60 FR 65476, Dec. 19, 1995, as amended at 68 FR 75390, Dec. 31,
2003; 73 FR 5723, Jan. 31, 2008; 84 FR 65644, Nov. 29, 2019; 85 FR
62207, Oct. 2, 2020]