[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:
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).
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).
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.
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.
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 (Lam) shall be less
than 75% of the maximum allowable leakage rate (0.75 La). 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 La). 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
