[Federal Register Volume 60, Number 68 (Monday, April 10, 1995)]
[Notices]
[Pages 18152-18153]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 95-8707]



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NUCLEAR REGULATORY COMMISSION
[Docket No. 50-318]


Exemption

    In the matter of Baltimore Gas and Electric Comp. (Calvert 
Cliffs Nuclear Power Plant Unit No. 2).

I

    Baltimore Gas and Electric Company (BG&E or the licensee) is the 
holder of Facility Operating License No. DPR-69, which authorizes 
operation of Calvert Cliffs Nuclear Power Plant Unit No. 2 (the 
facility/CC-2), at a steady-state reactor power level not in excess of 
2700 megawatts thermal. The facility is a pressurized water reactor 
located at the licensee's site in Calvert County, Maryland. The license 
provides 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 effect.

II

    Section III.D.1.(a) of appendix J to 10 CFR part 50 requires the 
performance of three Type A containment integrated leakage rate tests 
(ILRTs), at approximately equal intervals during each 10-year service 
period of the primary containment. The third test of each set shall be 
conducted when the plant is shutdown for the 10-year inservice 
inspection of the primary containment.

III

    By letter dated February 24, 1995, BG&E requested temporary relief 
for
CC-2 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 second Type A test by approximately 24 months 
(from the 1995 refueling outage, currently scheduled to begin in March 
1995, to the spring 1997 refueling outage) and would permit the third 
Type A test to be performed during the spring 1999 refueling outage, 
coincident with the end of the current American Society of Mechanical 
Engineers Boiler and Pressure Vessel Code (ASME Code) inservice 
inspection interval. This would extend the CC-2 second 10-year service 
period to 12 years.
    The licensee's request cites the special circumstance of 10 CFR 
50.12, paragraph (a)(2)(ii), as the basis for the exemption. 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. The licensee has analyzed the results 
of the previous Type A tests performed at  CC-2. Four Type A tests have 
been conducted from 1979 to date. The initial Type A test failed; 
however, prompt corrective actions were taken and the subsequent tests 
were successful as detailed in Section IV of this Exemption. It is also 
noted that the licensee, as a condition of the proposed exemption, will 
perform the visual containment inspection although it is only required 
by Appendix J to be conducted 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 boundary. Therefore, application of the 
regulation in this particular circumstance is not necessary to achieve 
the underlying purpose of the rule.

IV

    Section III.D.1.(a) of appendix J to 10 CFR part 50 states that a 
set of three Type A leakage rate tests shall be performed at 
approximately equal intervals during each 10-year service period.
    The licensee proposes an exemption to this section which would 
provide a one-time interval extension for the second Type A test by 
approximately 24 months. This would permit the test to be performed 
during the spring 1997 refueling outage, as noted above, and would 
extend the second 10-year service period to 12 years. The Commission 
has determined, for the reasons discussed below, that pursuant to 10 
CFR 50.12(a)(1) this exemption is authorized by law, will not present 
an undue risk to the public health and safety, and is consistent with 
the common defense and security. The Commission further determines that 
special circumstances, as provided in 10 CFR 50.12(a)(2)(ii), are 
present justifying the exemption; namely, that application of the 
regulation in the particular circumstances is not necessary to 
[[Page 18153]] achieve the underlying purpose of the rule. 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.
    As previously noted, the initial Type A test failed. This failure 
was due to three sources: (1) The containment recirculation sump 
isolation valve, MOV-4145; (2) the temporary level indicators on the 
steam generators; and (3) the packing gland of a main steam line 
inboard vent valve. The first leakage source was identified as a 
problem with the limit switch setting on MOV-4145 that prevented full 
closure. Resetting the switches and closing the valve electrically 
corrected the source of leakage. This valve is now tested periodically 
to ensure the limit switch settings allow full closure, and the value 
has not demonstrated excessive leakage in any subsequent Type A test. 
The temporary level indicators, are components which are only in place 
while the plant is shutdown. Upon identification of the leakage path, 
the temporary configuration was isolated and has not resulted in any 
further leakage. The third component condition which led to an 
excessive leakage rate during this test was attributed to a packing 
failure in the main steam inboard vent valves. This condition was 
corrected by backseating the vent valves to eliminate leakage. In a 
subsequent refueling outage, the vent valves were removed and the 
connection was sealed with blind flanges. Following the licensee's 
prompt identification and corrective actions, three additional Type A 
tests have been successful and have demonstrated a good containment 
performance. Thus, the Type A test results only confirm the results of 
the Type B and C test results. The NRC staff has noted that the 
licensee has a good record of ensuring a leak-tight containment. Since 
the first failure, all Type A tests have passed with significant margin 
and the licensee has noted that the results of the Type A testing have 
been confirmatory of the Type B and C tests which will continue to be 
performed.
    The NRC staff has also made use of the information in 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 test (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 
percent 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 CC-2 experience has also been 
consistent with these results as previously noted.
    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 also added another perspective. The NEI data 
show that in about one-third of the cases exceeding allowage 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 these considerations, it is unlikely 
that an extension of one cycle for the performance of the appendix J, 
Type A test at CC-2 would result in significant degradation of the 
overall containment integrity. As a result, the application of the 
regulation of these particular circumstances is not necessary 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 for CC-2 of one cycle (24 
months) for the performance of the appendix J, Type A test, and to 
permit the third Type A test to be performed during the spring 1999 
refueling which extends the second 10-year service period to 12 years 
to be acceptable. As a condition for granting this exemption, the 
licensee will perform visual containment inspections.
    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 14979).
    This Exemption is effective upon issuance and shall expire at the 
completion of the 1997 refueling outage.


    Dated at Rockville, Maryland, this 3rd day of April 1995.

    For the Nuclear Regulatory Commission.
Steven A. Varga,
Director, Division of Reactor Projects--I/II, Office of Nuclear Reactor 
Regulation.
[FR Doc. 95-8707 Filed 4-7-95; 8:45 am]
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