[Federal Register Volume 60, Number 194 (Friday, October 6, 1995)]
[Notices]
[Pages 52431-52432]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 95-24895]
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NUCLEAR REGULATORY COMMISSION
[Docket No. 50-302]
Exemption; Florida Power Corporation, Crystal River Nuclear
Generating Plant Unit 3
I
Florida Power Corporation (the licensee) is the holder of Facility
Operating License No. DPR-72, which authorizes operation of the Crystal
River Nuclear Generating Plant Unit 3 (CR-3). The license provides,
among other things, that the licensee is subject to all rules,
regulations, and orders of the Commission now or hereafter in effect.
The facility is of a pressurized water reactor type and is located
in Citrus County, Florida.
II
Pursuant to Title 10 Code of Federal Regulations Part 50 (10 CFR
50), Appendix A, ``General Design Criteria for Nuclear Power Plants,''
Criterion 16, ``Containment design,'' ``Reactor containment and
associated systems shall be provided to establish an essentially leak-
tight barrier against the uncontrolled release of radioactivity to the
environment and to assure that the containment design conditions
important to safety are not exceeded for as long as postulated accident
conditions require.'' 10 CFR 50.54(o) states that ``Primary reactor
containments for water cooled power reactors shall be subject to the
requirements set forth in Appendix J to this part.'' 10 CFR 50,
Appendix J, sets forth requirements for periodic verification by tests
of the leak-tight integrity of the primary reactor containment and
establish the acceptance criteria for such tests to satisfy general
design criterion 16 of the Commission's regulations. 10 CFR 50,
Appendix J, Paragraph III.D.1, specifies a set of three integrated leak
rate tests (ILRT or Type A test) to be performed at approximately equal
intervals during each 10-year service period. Such tests are to be
limited to periods when the plant is non-operational and secured in the
shutdown condition under an administrative control and in accordance
with the safety procedures defined in the license.
For CR-3, the next available opportunity for performing the ILRT
would be in spring 1996. The licensee requested a one-time interval
extension for the ILRT by approximately 24 months from the spring 1996
refueling outage to the spring 1998 refueling outage. The licensee
indicated that approval of its request would save over two million
dollars and reduce personnel radiation exposure. An exemption from 10
CFR 50, Appendix J, Paragraph III.D.1, is needed to permit the licensee
to defer the ILRT.
By letter dated May 19, 1995, as supplemented August 8, 1995, the
licensee submitted its exemption request for this purpose.
III
Pursuant to 10 CFR 50.12, the Commission may, upon application by
any interested person or upon its own initiative, grant exemptions from
the requirements of 10 CFR Part 50 when (1) the exemptions are
authorized by law, will not present an undue risk to public health and
safety, and are consistent with the common defense and security; and
(2) when special circumstances are present. Special circumstances are
present whenever, according to 10 CFR 50.12(a)(2)(ii), ``Application of
the regulation in the particular circumstances would not serve the
underlying purpose of the rule or is not necessary to achieve the
underlying purpose of the rule * * *'' The underlying purpose of 10 CFR
50, Appendix J, Paragraph III.D.1., is to assure that periodic
surveillance of reactor containment penetrations is performed so that
proper maintenance and repairs are made during the service life of the
containment, and leakage through the primary reactor containment shall
not exceed allowable leakage rate values as specified in the technical
specifications (TS) or associated bases.
IV
In support of its exemption request, the licensee submitted
information pertaining to Type A, and local leak rate (LLRT or Types B
and C) testing history, structural capability, and risk assessment to
demonstrate that the proposed exemption would not present an undue risk
to the public health and safety and would be consistent with the common
defense and security, and would be authorized by law. The licensee
indicates that the Type A testing frequency of Appendix J is not
necessary to achieve the underlying purpose of the regulation and thus
[[Page 52432]]
special circumstances required by 10 CFR 50.12(a)(2)(ii) apply to this
situation.
The CR-3 containment is a reinforced concrete structure with a
cylindrical wall, a flat foundation mat, and a shallow dome roof. The
cylinder wall is prestressed with a post-tensioning system in the
vertical and horizontal directions. The dome roof is prestressed using
a three-way post-tensioning system. The inside surface of the
containment has a carbon steel liner to ensure a high degree of leak-
tightness during operating and accident conditions. The liner is
anchored to the concrete to ensure composite action with the concrete
shell. Piping penetrations have been designed to ensure that the liner
would not be breached due to rupture of any process pipe. The
containment is designed with an allowable leakage rate of 0.25% of
containment air weight per day (La) at the calculated maximum
allowable containment pressure (Pa) of 54.2 psig resulting from
the limiting design basis accidents.
The historical Type A test results as set forth in the exemption
request demonstrate that CR-3 has a low-leakage containment. The
current 10-year inservice inspection and inservice testing service
period is the second service period and started in March 1987 and ends
in March 1997. During this service period, the licensee performed one
ILRT in November 7, 1991. A prior ILRT conducted in November 1987 was
counted as the third test of the first 10-year interval and therefore,
the licensee did not take credit for the November 1991 test for the
current interval. These two ILRTs which have been performed during the
last seven years have shown acceptable containment leakage rates. There
have been no permanent or temporary modifications to the containment
structure, liner or penetrations since the last two Type A tests, and
no future modifications are planned prior to the 1998 refueling outage
that could adversely affect the Type A test results.
The licensee will continue to be required to conduct the Type B and
C local leak rate tests, which are in general the principal means of
detecting containment leakage paths, with the Type A tests confirming
the Type B and C test results. Types B and C testing history at CR-3
shows that the overall combined as-found leakage has been less than the
allowed combined leakage rate of 0.6 La (266,431 SCCM) at the
calculated maximum peak containment pressure as specified in Appendix
J. Successful performance of Types B and C testing demonstrates the
leak-tightness of the penetrations and associated components and
provides a high degree of assurance that the overall Type A leakage
rate would remain satisfactory while this exemption is in effect. The
licensee has stated that it will perform the general containment
inspection, although it is required by Appendix J (Section V.A.) to be
performed only 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.
The purpose of containment leak testing is to detect containment
leakage which could be the result of failures (active or passive)
before an accident occurs. Containment leakage caused by degradation of
sealing material within containment penetrations and containment
isolation components will continue to be effectively measured by the
Type B and C testing programs. The Type A tests are only confirmatory
of the results of the Type B and C test results. The only potential
failures not covered by Types B and C testing are failures of the
containment due to structural deterioration because of parameters such
as pressure or temperature. However, structural deterioration would
require longer than the proposed period for the exemption.
There are no mechanisms that would adversely affect the structural
capability of the containment, which is the only leakage mode not
captured by the Type B and C testing that will be performed. Absent
actual accident conditions, structural deterioration of containment due
to temperature, radiation, chemical, or other such effects is a gradual
phenomenon requiring periods of time well in excess of the proposed
interval extension and is subject to detection by periodic visual
inspections. At CR-3, there has been no evidence of structural
deterioration that would impact structural integrity or leak tightness.
Other than postulated accident conditions, the only over-pressure
challenge to containment is the integrated leak rate test itself. Thus,
there is significant assurance that the extended interval between Type
A tests in concert with Type B and C testing will continue to provide
adequate verification of the leak tight integrity of the containment.
The proposed one-time change in Type A leakage test frequency only
affects the length of time that the containment could be in an
undetected failed state as a result of a failure. As part of the CR-3
Individual Plant Examination (IPE) program, the risk of losing
containment integrity is considered negligible compared to other risks
such as those resulting from small break loss of coolant accidents or
station blackout.
Draft NUREG-1493, which provides the technical justification for
the ongoing Appendix J rulemaking effort (including a 10-year test
frequency), has shown that essentially all containment leakage can be
detected by LLRTs (Type B and C). According to results given in NUREG-
1493, only 5 ILRT failures out of 180 ILRT reports that covered 110
individual reactors and approximately 770 years of operating history,
were found that local leak rate testing could not have detected.
Therefore, it is unlikely that this one-time exemption for the
performance of Type A testing at CR-3 would result in significant
degradation of the overall containment integrity.
In summary, the testing history, structural capability of the
containment, and the risk assessment discussed previously establish
that (1) CR-3 has had acceptable containment leakage rate test results,
(2) the structural integrity of containment is assured, and (3) there
is negligible risk impact in changing the Type A test schedule on a
one-time basis.
Therefore, application of the regulation in this particular
circumstance would not serve, nor is it necessary to achieve, the
underlying purpose of the rule, and the exemption request meets the
requirements of 10 CFR 50.12.
Accordingly, the Commission has determined that, pursuant to 10 CFR
50.12(a), an exemption is authorized by law, will not endanger life or
property or common defense and security, and is otherwise in the public
interest. Therefore, the Commission hereby grants Florida Power
Corporation a one-time exemption from those requirements of 10 CFR 50,
Appendix J, relating to containment overall leak rate test and allows
deferring the performance of a Type A test from the spring 1996 to the
spring 1998 refueling outage, provided that the general containment
inspection is performed during the spring 1996 outage. Pursuant to 10
CFR 51.32, the Commission has determined that the granting of this
exemption will not result in any significant adverse environmental
impact (60 FR 46320).
Dated at Rockville, Maryland, this 29th day of September 1995.
For the Nuclear Regulatory Commission.
Steven A. Varga,
Director, Division of Reactor Projects--I/II, Office of Nuclear Reactor
Regulation.
[FR Doc. 95-24895 Filed 10-5-95; 8:45 am]
BILLING CODE 7590-01-P
