[Federal Register Volume 60, Number 237 (Monday, December 11, 1995)]
[Notices]
[Pages 63546-63548]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 95-30048]



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NUCLEAR REGULATORY COMMISSION

[Docket No. 50-440]


Cleveland Electric Illuminating Company, Centerior Service 
Company, Duquesne Light Company, Ohio Edison Company, Pennsylvania 
Power Company, Toledo Edison Company (Perry Nuclear Power Plant, Unit 
1); Exemption

I

    Cleveland Electric Illuminating Company, (the licensee) is the 
holder of Facility Operating License No. NPF-58, which authorizes 
operation of the Perry Nuclear Power Plant, Unit 1 (PNPP). The 
operating license provides, among other things, that the licensee is 
subject to all rules, regulations, and orders of the Commission now and 
hereafter in effect.
    The facility consists of a single boiling water reactor located at 
the licensee's site in Lake County, Ohio.

II

    Containment leak rate testing is necessary to demonstrate that the 
measured leak rate is within the acceptance criteria cited in the 
licensing design basis. Periodic testing of the overall containment 
structure along with separate leak testing of the penetrations provides 
assurance that post-accident radiological consequences will be within 
the limits of 10 CFR Part 100. The Commission's requirements regarding 
leak rate testing are found in Appendix J to 10 CFR Part 50.
    In its letter dated October 21, 1994, the licensee applied for 
partial exemptions from the Commission's regulations. The subject 
exemptions, which are from the requirements in Appendix J, Option A, to 
10 CFR Part 50, include:
    <bullet> Section III.A.5(b)(2) states that the measured leakage 
from the containment integrated leak rate (Type A) test (Lam) shall be 
less than 75% of the maximum allowable leakage rate (0.75 La).
    <bullet> Sections III.B.3 and III.C.3 require that the combined 
leakage of valves and penetrations subject to Type B and C local leak 
rate testing be less than 0.6 times the maximum allowable leakage rate 
(0.6 La).
    <bullet> Section III.A.1(d) requires that all fluid systems that 
would be open to containment following post-accident conditions, be 
vented and drained prior to conducting the containment integrated leak 
rate test.
    <bullet> Section III.D.1(a) states that the third Type A test of 
each 10-year interval be conducted when the plant is shut down for the 
10-year plant inservice inspection.
    <bullet> Section III.D.3 states that Type C tests shall be 
performed during each reactor shutdown for refueling but in no case at 
intervals greater than 2 years. Type C tests are tests intended to 
measure containment isolation valve leakage rates.

III

    Section III.A.5(b)(2) states that the measured leakage from the 
containment integrated leak rate (Type A) test (L<INF>am) shall be less 
than 75% of the maximum allowable leakage rate (0.75 L<INF>a). The 
licensee proposes to exempt main steam line isolation valve leakage 
from Type A test results and consider leakage from the main steam lines 
separately. Sections III.B.3 and III.C.3 require that the combined 
leakage of valves and penetrations subject to Type B and C local leak 
rate testing be less than 0.6 

[[Page 63547]]
times the maximum allowable leakage rate (0.6 L<INF>a). The licensee 
proposes to exempt main steam line isolation valve leakage from the 
combined leakage from Type B and C local leak rate testing and consider 
leakage from the main steam lines separately. Section III.A.1(d) 
requires that all fluid systems that would be open to containment 
following post-accident conditions, be vented and drained prior to 
conducting Type A tests. The licensee proposes that the piping between 
the inboard and outboard main steam line isolation valves be flooded 
with water when Type A tests are conducted.
    During the original staff review of the PNPP, the licensee proposed 
separate treatment of measured leakage past the main steam isolation 
valves. This approach is consistent with the staff's Standard Review 
Plan (SRP) 15.6.5, Appendix D, ``Radiological Consequences of a Design 
Basis Loss-of-Coolant Accident: Leakage from Main Steam Isolation Valve 
Leakage Control System.'' In this SRP, the radiological consequences 
associated with leakage from the main steam lines is calculated 
separately and subsequently combined with the consequences from other 
fission product release paths.
    As described in the Final Safety Analysis Report, the licensee 
calculates off-site dose consequences by assuming separate 
contributions from the containment integrated leak rate and the main 
steam line isolation valve leak rate. These assumptions are supported 
by the staff's Safety Evaluation Report (NUREG-0887) and the PNPP 
Technical Specifications. Both the FSAR and Specification 3.6.1.2.a 
state that the overall containment integrated leak rate shall be less 
than 0.20 percent per day. NUREG-0887 lists this same value for the 
containment integrated leak rate and a separate contribution from main 
steam line leakage. Finally, Specification 3.6.1.2.b specifically 
states that main steam line leakage will not be considered part of the 
combined leak rate for penetrations and valves. Specification 3.6.1.2.c 
limits the maximum allowable leakage from each main steam line to 25 
standard cubic feet per hour.
    As described above, the licensee does not include leakage from the 
main steam line isolation valves in either the Type A test results or 
the combined Type B and C test results. Since the licensee measures 
main steam line leakage separately from other Appendix J related 
testing, the licensee does not want leakage from the main steam lines 
to inadvertently influence the Type A test results. Therefore, in lieu 
of venting and draining the piping between containment isolation valves 
as required by Appendix J, the licensee proposes filling this section 
of piping with water when Type A tests are performed. Filling these 
sections of pipe with water would ensure that air would not pass 
through these lines and thereby contribute to the Type A test results.
    The licensee has proposed alternative methods to the leak testing 
requirements of Appendix J. While the licensee is treating main steam 
line leakage separately from both Type A test results and the combined 
Type B and C test results, the licensee still meets the intent of 
Appendix J by demonstrating that the overall leakage is within design 
limits. Therefore, the staff concludes that special circumstances are 
present as required by 10 CFR 50.12(a)(2)(ii), in that application of 
the regulation is not needed to meet the underlying purpose of the 
rule. Furthermore, the staff finds that permitting the alternative 
methods of leak testing will not present an undue risk to the public 
health and safety.
    Section III.D.1(a) requires, in part, that ``* * * a set of three 
Type A tests shall be performed, at approximately equal intervals 
during each 10-year service period. The third test of each set shall be 
conducted when the plant is shutdown for the 10-year plant inservice 
inspections.'' The licensee proposes to perform the three Type A tests 
at approximately equal intervals within each 10-year period, with the 
third test of each set conducted as close as practical to the end of 
the 10-year period. However, there would be no required connection 
between the Appendix J 10-year interval and the inservice inspection 
10-year interval.
    The 10-year plant inservice inspection (ISI) is the series of 
inspections performed every 10-years in accordance with Section XI of 
the ASME Boiler and Pressure Vessel Code and Addenda as required by 10 
CFR 50.55a. The licensee performs the ISI volumetric, surface, and 
visual examinations of components and system pressure tests in 
accordance with 10 CFR 50.55a(g)(4) throughout the 10-year inspection 
interval. The major portion of this effort is presently being performed 
during the refueling outages. As a result, there is no extended outage 
in which the 10-year ISI examinations are performed.
    There is no benefit to be gained by the coupling requirement cited 
above in that elements of the PNPP ISI program are conducted throughout 
each 10-year cycle rather than during a refueling outage at the end of 
the 10-year cycle. Consequently, the subject coupling requirement 
offers no benefit either to safety or to the economical operation of 
the facility.
    Moreover, each of these two surveillance tests (i.e., the Type A 
tests and the 10-year ISI program) is independent of the other and 
provides assurances of different plant characteristics. The Type A test 
assures the required leak-tightness to demonstrate compliance with the 
guidelines of 10 CFR Part 100. The 10-year ISI program provides 
assurance of the integrity of the structures, systems and components as 
well as verifying operational readiness of pumps and valves in 
compliance with 10 CFR 50.55a. There is no safety-related concern 
necessitating their coupling in the same refueling outage. Accordingly, 
the staff finds that application of the regulation is not necessary to 
achieve the underlying purpose of the rule.
    On this basis, the staff finds that the licensee has demonstrated 
that there are special circumstances present as required by 10 CFR 
50.12(a)(2)(ii). Further, the staff also finds that the uncoupling of 
the Type A tests from the 10-year ISI program will not present an undue 
risk to the public health and safety.
    Section III.D.3 of Appendix J states that Type C tests shall be 
performed during each reactor shutdown for refueling but in no case at 
intervals greater than 2 years. The licensee requested relief from the 
requirement to perform Type C tests during each reactor shutdown for 
refueling. The licensee proposes to perform the required Type C tests 
while the plant is at power.
    Section II.D.3 of Appendix J requires that ``Type C tests shall be 
performed during each reactor shutdown for refueling but in no case at 
intervals greater than 2 years.'' Paragraph III.D.2 discusses the 
scheduling of Type B tests and contains the same wording but also 
includes an additional provision that allows Type B tests to be 
performed at ``other convenient intervals'' in lieu of during reactor 
shutdown for refueling. The licensee has requested that this same 
flexibility be applied to Type C local leak rate testing.
    The underlying purpose of the rule is to ensure that adequate 
testing is done to demonstrate containment integrity. From the 
standpoint of testing adequacy, when the testing is performed is not 
significant because the conditions of testing are the same regardless 
of when it is performed. As indicated by the licensee, the BWR/6 Mark 
III containment/suppression pool design is such that Type C local leak 
rate testing can be performed during power operation on certain 
systems. In addition, the Drywell and Containment Purge System 
containment isolation 

[[Page 63548]]
valves have surveillance requirements imposed on them to demonstrate 
leak tightness during power operation. These surveillance tests are the 
same exact leak rate tests as the Type C local leak rate tests 
performed during refueling outages.
    Taking credit for testing performed during power operation provides 
the same degree of assurance of containment integrity as taking credit 
for testing performed during shutdown. In addition, testing while at 
power may be preferable when considering ALARA and operability 
requirements. Therefore, the special circumstances of 10 CFR 
50.12(a)(2)(ii) are present in that application of the regulation in 
this particular circumstance is not necessary to achieve the underlying 
purpose of the rule.

IV

    The Commission has determined that pursuant to 10 CFR 50.12(a)(1) 
that 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 
this particular circumstance is not necessary to achieve the underlying 
purpose of the rule.
    Pursuant to 10 CFR 51.32, the Commission has determined that the 
granting of this Exemption will not have a significant impact on the 
quality of the human environment (60 FR 51821). This exemption is 
effective upon issuance.

    Dated at Rockville, Maryland, this 4th day of December 1995.

    For the Nuclear Regulatory Commission
Jack W. Roe,
Director, Division of Reactor Projects III/IV,Office of Nuclear Reactor 
Regulation
[FR Doc. 95-30048 Filed 12-8-95; 8:45 am]
BILLING CODE 7590-01-P