Home > Electronic Reading Room > Document Collections > NRC Regulations (10 CFR) >
Part Index >
Appendix H to Part 50—Reactor Vessel Material Surveillance Program
Requirements
Appendix H to Part 50—Reactor Vessel Material Surveillance Program
Requirements
- Introduction
- Definitions
- Surveillance Program Criteria
- 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capsules, which are withdrawn periodically from the reactor vessel.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 1017 n/cm2 (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 ASTM E 185 that is current
on the issue date of the ASME Code to which the reactor vessel was purchased.
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 ASTM E 185-82 to the extent
practicable for the configuration of the specimens in the capsule.
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.
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 or the Director, Office of New
Reactors, as appropriate, 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 or the Director, Office of New
Reactors, as appropriate.
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
one year 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; 68 FR 75390, Dec. 31, 2003; 73 FR 5723, Jan. 31, 2008]
|