[Federal Register Volume 59, Number 224 (Tuesday, November 22, 1994)]
[Unknown Section]
[Page 0]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 94-28761]


[[Page Unknown]]

[Federal Register: November 22, 1994]


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NUCLEAR REGULATORY COMMISSION
[Docket No. 50-237]

 

Commonwealth Edison Co. (Dresden Nuclear Power Station, Unit 2)

Exemption

I

    Commonwealth Edison Company (ComEd, the licensee) is the holder of 
Facility Operating License No. DPR-19, which authorizes operation of 
the Dresden Nuclear Power Station, Unit 2 (the facility), at a steady-
state power level not in excess of 2527 megawatts thermal. The facility 
is a boiling water reactor located at the licensee's site in Grundy 
County, Illinois. This license provides, among other things, that the 
facility is subject to all rules, regulations, and Orders of the 
Nuclear Regulatory Commission (the Commission) now or hereafter in 
effect.

II

    By letter dated October 28, 1994, pursuant to 10 CFR 50.12(a), 
ComEd requested a schedular exemption for Dresden, Unit 2, from the 18-
month test interval for a Type A Integrated Leak Rate Test (ILRT) as 
required by 10 CFR Part 50, Appendix J, Section III.A.6.(b). The 
exemption is requested to avoid a potential reactor shut down to 
perform the Type A test.
    Due to two forced outages, ComEd has had to reschedule the Dresden, 
Unit 2, refueling outage from September 1994 to July 1995. 
Subsequently, ComEd requested a maximum extension of up to an 
additional 242 days for the 18-month Appendix J test interval for a 
Type A test. The Type A test can not be performed during power 
operation.

III

    In its letter dated October 28, 1994, ComEd requested a one-time 
exemption from the 18-month Type A test interval requirement of 
Appendix J. ComEd has provided leakage test results and maintenance 
information for the past two outage testing programs. The two 
consecutive Type A ILRT failures that placed Dresden, Unit 2, on the 
accelerated test schedule were because of the addition of the Type B 
and C test leakage results to the Type A leakage rate--not problems 
with the Type A boundaries. The minimum pathway data from the last two 
Unit 2 refueling outages (D2R12 and D2R13) indicate that, on a minimum 
pathway basis, the quality of primary containment does not degrade 
excessively through the course of the fuel cycle. The total containment 
leakage rate minus the Type B and C leakages for the last two Type A 
test failures during D2R12 and D2R13 are 285.5 standard cubic feet per 
hour (scfh) and 302.62 scfh, respectively. These values are 47 percent 
and 50 percent of the Type A ILRT acceptance criteria of 610.56 scfh 
(0.75La). The acceptance criteria are from the Dresden, Unit 2, 
Technical Specification (TS) Section 3.7.A.2.b.(1). During the D2R13 
refueling outage, the Type A ILRT failed due to a leak of the inboard 
flange of the reactor building to suppression chamber vacuum breaker 
valve (2-1601-20A). This leakage was quantified to be 12720.05 scfh. If 
this did not occur, the Type A ILRT results would have been 543.40 
scfh, which is less than the TS limit of 610.56 scfh (0.75La). 
During the D2R13 refueling outage, the Type A test failed due to the 
as-found minimum pathway leakage of primary containment isolation 
valves found during Type B and C testing. The volumes that were the 
major contributors to this failure are as follows: The ``B'' feedwater 
line isolation check valves (2-220-58B and 2-220-62B), shutdown cooling 
isolation valves (2-1001-1A, 2-1001-1B, 2-1001-2A, 2-1001-2B and 2-
1001-2C), reactor water clean-up isolation valves (2-1201, 2-1201-1A, 
2-1201-3, 2-1299-004 and 2-1299-005), low pressure coolant injection 
containment spray isolation valves (2-1501-27B and 2-1501-28B), high 
pressure coolant injection drain pot to suppression chamber valves (2-
2301-34 and 2-2301-71), traversing incore probe purge check valve (2-
4799-514), electrical penetration X-202W, and drywell bellow X-113. The 
total minimum pathway leakage for these volumes was 685.75 scfh. Each 
of the above valves and volumes were repaired during the refueling 
outage and subsequently passed post-maintenance Type B and C tests 
prior to restart of Unit 2 for the current operating cycle.
    At the time the pertinent requirements of Appendix J were 
established (1973), the typical nuclear power plant fuel cycle lasted 
12 months. Section III.A.6.(b) may have been written to require a Type 
A test at every refueling outage--specifying a numerical cap on the 
interval to prevent extreme cases of very long intervals (caused by 
extended shutdowns or operations at reduced power) between tests. An 
18-month cap was reasonable for 12-month fuel cycles, but with the 
current 18- and 24-month fuel cycles, it clearly is not sufficient. The 
intent of Section III.A.6.(b) is to increase the testing frequency for 
a containment that exhibits leakage problems, but not to increase it so 
much that special shutdowns are required. Refueling outages are the 
only reasonable time to perform Type A test, since most plants do not 
have extended shutdowns at any other time.
    In order to add an additional margin of safety and to account for 
the possible increase in the leakage of a containment penetration, the 
licensee has imposed an administrative limit for Dresden, Unit 2, on 
the total maximum allowable containment leakage rate until the 
completion of the current operating cycle. This limit will be 519.0 
scfh, which is 85 percent of the TS limit of 610.56 scfh (0.75La). 
All additional minimum pathway leakage will be added to the current 
total as operational Type B and C leak rate tests are performed during 
the current operating cycle. These corrective measures taken should 
reduce the chances of D2R14 failing the ``As Found'' Type A ILRT test.

IV

    The staff has reviewed ComEd's submittal regarding the Appendix J 
test interval exemption request. Based on the above, the staff 
concludes that the licensee's corrective action and administrative 
leakage limit have reduced the likelihood of excessive leakage during 
the proposed extension of the Type A test interval. Further, 
considering the intent of the 18-month interval cap and its relation to 
longer fuel cycles, the staff finds that the safety benefit to be 
derived from performing a Type A test at 18 months rather than 26 
months does not justify the hardship of a forced plant shutdown. 
Therefore, the staff finds that the requested Type A interval extension 
should be granted.
    This is a one-time exemption from the 18-month Type A test interval 
requirements as prescribed in 10 CFR Part 50, Appendix J, and is 
intended to be in effect until July 14, 1995.
    In its October 28, 1994, letter, ComEd also identified special 
circumstances. As discussed above, the exemption request is for a short 
duration relative to the 18-month requirement. This meets a criterion 
for a special circumstance per 10 CFR 50.12(a)(2)(v), i.e., ``The 
exemption would provide only temporary relief from the applicable 
regulation and the licensee or applicant has made good faith efforts to 
comply with the regulation.''

V

    Accordingly, the Commission has determined that, pursuant to 10 CFR 
50.12(a), that (1) this exemption is authorized by law and will not 
endanger life or property or the common defense and security, and is 
otherwise in the public interest, and (2) the Exemption would provide 
only temporary relief from the applicable regulation and the licensee 
has made good faith efforts to comply with the regulation. Therefore, 
the Commission hereby grants an exemption as described in Section III 
above from 10 CFR Part 50, Appendix J, Section III.A.6.(b), to the 
extent that the 18-month interval for performing the Type A test may be 
extended for 242 days until July 14, 1995, on a one-time only basis, 
for Dresden, Unit 2.
    Pursuant to 10 CFR 51.32 the Commission has determined that the 
granting of this Exemption will have no significant impact on the 
environment (59 FR 56095).
    This exemption is effective upon issuance.

    Dated at Rockville, Maryland this 10th day of November 1994.

    For the Nuclear Regulatory Commission.
Jack W. Roe,
Director, Division of Reactor Projects--III/IV, Office of Nuclear 
Reactor Regulation.
[FR Doc. 94-28761 Filed 11-21-94; 8:45 am]
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