[Federal Register Volume 60, Number 135 (Friday, July 14, 1995)]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 95-17295]
NUCLEAR REGULATORY COMMISSION
[Docket No. 50-280]
In the Matter of: Virginia Electric Power Company (Surry Power
Station Unit No. 1); Exemption
Virginia Electric and Power Company (the licensee) is the holder of
Facility Operating License No. DPR-37, which authorizes operation of
Surry Power Station, Unit 1 (the facility), at a steady-state reactor
power level not in excess of 2441 megawatts thermal. The facility is a
pressurized water reactor located at the licensee's site in Surry
County, Virginia. The license provide among other things, that it is
subject to all rules, regulations, and Orders of the U.S. Nuclear
Regulatory Commission (the Commission or NRC) now or hereafter in
Section III.D.1.(a) of Appendix J to 10 CFR Part 50 requires the
three Type A containment integrated leakage rate tests (ILRTs) of the
primary containment, at approximately equal intervals during each 10-
year service period. The third test of each set shall be conducted when
the plant is shut down for the 10-year inservice inspection program.
By letter dated April 28, 1995, the licensee requested temporary
relief from the requirement to perform a set of three Type A tests at
approximately equal intervals during each 10-year service period of the
primary containment. The requested exemption would permit a one-time
interval extension of the third Type A test by approximately 18 months
(from the October 1995 refueling outage, to the February 1997 refueling
outage) and would permit the third Type A test of the second 10-year
inservice inspection period to not correspond with the end of the
current American Society of Mechanical Engineers Boiler and Pressure
Vessel Code (ASME Code) inservice inspection interval.
The licensee's request cites the special circumstances of 10 CFR
50.12, paragraph (a)(2)(ii), as the basis for the exemption. The
licensee points out that the existing Type B and C testing programs are
not being modified by this request and will continue to effectively
detect containment leakage caused by the degradation of active
containment isolation components as well as containment penetrations.
It has been the experience at Surry Unit 1 during the Type A tests
conducted from 1986 to date, that the Type A tests have not identified
any significant sources of leakage in addition to those found by the
Type B and C tests.
During operation, the Surry Unit 1 containment is maintained at a
subatmospheric pressure (approximately 10.0 psia) which provides a good
indication of the containment integrity. Technical Specifications
require the containment to be subatmospheric whenever Reactor Coolant
System temperature and pressure exceeds 350 deg.F and 450 psig,
respectively. Containment air partial pressure is monitored in the
control room to ensure Technical Specification compliance. If the
containment air partial pressure increases above the established
Technical Specification limit, the unit is required to shut down.
In the licensee's April 28, 1995, exemption request, the licensee
stated that special circumstance 50.12(a)(2)(ii) is applicable to this
situation, i.e., that application of the regulation is not necessary to
achieve the underlying purpose of the rule.
Appendix J states that the leakage test requirements provide for
periodic verification by tests of the leak tight integrity of the
primary reactor containment. Appendix J further states that the purpose
of the tests ``is to assure that leakage through the primary reactor
containment shall not exceed the allowable leakage rate values as
specified in the Technical Specifications or associated bases''. Thus,
the underlying purpose of the requirement to perform type A containment
leak rate tests at intervals during the 10-year service period is to
ensure that any potential leakage pathways through the containment
boundary are identified within a time span that prevents significant
degradation from continuing or becoming unknown.
The NRC staff has reviewed the basis and supporting information
provided by the licensee in the exemption request. The NRC staff has
noted that the licensee's record of ensuring a leak-tight containment
has improved markedly since 1986. All ``as-found'' Type A tests since
1986 have passed and the results of the Type A testing have been
confirmatory of the Type B and C tests which will continue to be
performed. The licensee will perform the general containment inspection
although it is only required by Appendix J (Section V.A.) to be
performed in conjunction with Type A tests. The NRC staff considers
that these inspections, though limited in scope, provide an important
added level of confidence in the continued integrity of the containment
The Surry Unit 1 containment is of the subatmospheric design.
During operation,the containment is maintained at a subatmospheric
pressure (approximately 10 psia) which provides for constant monitoring
of the containment integrity and further obviates the need for Type A
testing at this time. If the containment air partial pressure exceeds
the established Technical Specification limit, the unit must be shut
The NRC staff has also made use of a draft staff report, NUREG-
1493, which provides the technical justification for the present
Appendix J rulemaking effort which also includes a 10-year test
interval for Type A tests. The integrated leakage rate test, or Type A
test, measures overall containment leakage. However, operating
experience with all types of containments used in this country
demonstrates that essentially all containment leakage can be detected
by local leakage rate tests (Type B and C). According to results given
in NUREG-1493, out of 180 ILRT reports covering 110 individual reactors
and approximately 770 years of operating history, only 5 ILRT failures
were found which local leakage rate testing could not detect. This is
3% of all failures. This study agrees well with previous NRC staff
studies which show that Type B and C testing can detect a very large
percentage of containment leaks.
The Nuclear Management and Resources Council (NUMARC), now the
Nuclear Energy Institute (NEI), collected and provided the NRC staff
with summaries of data to assist in the Appendix J rulemaking effort.
NUMARC collected results of 144 ILRTs from 33 units; 23 ILRTs exceeded
1.0La. Of these, only nine were not due to Type B or C leakage
penalties. The NEI data show that in about one-third of the cases
exceeding allowable leakage, the as-found leakage was less than
2La; in one case the leakage was found to be approximately
2La; in one case the as-found leakage was less than 3La; one
case approached 10La; and in one case the leakage was found to be
approximately 21La. For about half of the failed ILRTs the as-
found leakage was not quantified. These data show that, for those ILRTs
for which the leakage was quantified, the leakage values are small in
comparison to the leakage value at which the risk to the public starts
to increase over the value of risk corresponding to La
(approximately 200La, as discussed in NUREG-1493). Therefore,
based on those considerations, it is unlikely that an extension of one
cycle for the performance of the Appendix J, Type A test at Surry, Unit
1, would result in significant degradation of the overall containment
integrity. As a result, the application of the regulation in these
particular circumstances is not needed to achieve the underlying
purpose of the rule.
Based on generic and plant specific data, the NRC staff finds the
basis for the licensee's proposed exemption to allow a one-time
exemption to permit a schedular extension of one cycle for the
performance of the Appendix Type A test, provided that the general
containment inspection is performed, to be acceptable.
Pursuant to 10 CFR 51.32, the Commission has determined that
granting this Exemption will not have a significant impact on the
environment (60 FR 35439).
This Exemption is effective upon issuance and shall expire at the
completion of the 1997 refueling outage.
Dated at Rockville, Maryland, this 7th day of July 1995.
For the Nuclear Regulatory Commission.
Steven A. Varga,
Director of Reactor Projects--I/II Office of Nuclear Reactor
[FR Doc. 95-17295 Filed 7-13-95; 8:45 am]
BILLING CODE 7590-01-M