[Federal Register Volume 60, Number 64 (Tuesday, April 4, 1995)]
[Notices]
[Pages 17090-17092]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 95-8166]
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NUCLEAR REGULATORY COMMISSION
[Docket No. 50-339]
Virginia Electric and Power Co. (North Anna Power Station Unit
No. 2); Exemptions
I
Virginia Electric and Power Company (the licensee) is the holder of
Facility Operating License No. NPF-7, which authorizes operation of
North Anna Power Station, Unit 2 (the facility or NA-2), at a steady-
state reactor power level not in excess of 2893 megawatts thermal. The
facility is a pressurized water reactor located at the licensee's site
in Louisa County, Virginia. 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) of the primary containment, at approximately equal intervals
during each 10-year service period. The third test of each set shall be
conducted when the plant is shut down for the 10-year inservice
inspection program.
Section IV.A of appendix J to 10 CFR part 50 requires that any
modification, replacement of a component which is part of the primary
reactor containment boundary, or resealing a seal-welded door,
performed after the preoperational leakage rate test shall be followed
by either a Type A, Type B, or Type C test, as applicable for the area
affected by the modification.
III
By letter dated March 2, 1995, the licensee requested temporary
relief 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 third Type A test by approximately 16 months
(from the March 1995 steam generator replacement outage, to the October
1996 refueling outage).
The licensee's March 2, 1995, letter also requested temporary
relief from the requirements to perform a type A test following a major
modification or replacement of a component which is part of the primary
reactor coolant boundary. Specifically, the post-modification exemption
is requested from performing a Type A test due to the activities
associated with the upcoming NA-2 steam generator replacement. The
basis for the post-modification exemption request is that, in this
case, the ASME Section XI inspection and testing requirements more than
fulfill the intent of the requirements of Section IV.A of Appendix J.
The licensee's request cites the special circumstances of 10 CFR
50.12, paragraph (a)(2)(ii), as the basis for the exemption to Section
III.D.1.a of appendix J to 10 CFR part 50. The licensee points out that
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. It has been the
experience at NA-2 during the Type A tests conducted during the first
10-year inservice inspection interval (1984, 1989, and 1990), that
considerable margin exists between the Type A tests and the Technical
Specifications (TS) allowable leakage rate limit.
During operation, the NA-2 containment is maintained at a
subatmospheric pressure (approximately 10.0 psia) which provides a good
indication of the containment integrity. TS require the containment to
be subatmospheric when in Modes 4, 3, 2, and 1. Containment air partial
pressure is monitored in the control room to ensure TS compliance. If
the containment air partial pressure increases above the established TS
limit, the unit is required to shut down.
The licensee's request also cites the special circumstances of 10
CFR 50.12, paragraph (a)(2)(ii), as the basis for the exemption to
Section IV.A of appendix J to 10 CFR part 50.
The NA-2 plant design incorporates a ``closed system'' for
transferring steam from the steam generators inside of the primary
containment to the main turbine-generators in the turbine building. The
inside containment portion of this closed system consists of the main
steam lines, the feedwater lines, and the secondary side of the steam
generators. This closed system inside of containment forms a part of
the primary reactor containment boundary.
The planned replacement of the NA-2 steam generators includes the
following activities:
--Cutting and removing the mainsteam and feedwater lines from the steam
generators.
--Cutting and removing the upper assemblies of the steam generators
(steam domes).
--Cutting the reactor coolant piping and removing the steam generator
lower assemblies (tube bundles).
--Installing the new steam generator lower assemblies and re-welding
the reactor coolant piping.
--Re-installing the steam generator upper assemblies on the new lower
assemblies.
--Re-installing and re-welding the main steam and feedwater lines.
The planned replacement of the NA-2 steam generators affects only
this [[Page 17091]] closed piping system inside containment. The steam
generator replacement activities do not affect the containment
structure or the actual containment liner.
Section IV.A to Appendix J, Special Testing Requirements for
Containment Modifications, requires that any major modification or
replacement of a component which is part of the primary reactor
containment boundary shall be followed by either a Type A, Type B, or
Type C test, as applicable for the area affected by the modification.
The Type C testing requirements of Appendix J apply to leakage testing
of containment isolation valves. The planned replacement does not
affect any containment isolation valves and, therefore, the Type C
testing requirements are not applicable. The Type B testing
requirements of appendix J apply to leakage testing of gasketed or
sealed containment penetrations (e.g., electrical penetrations), air
lock door seals, and other doors with resilient seals or gaskets.
Although the secondary side of the steam generators have access manways
with gaskets, the Type B testing requirements do not address the other
areas of the containment boundary affected by the planned replacement,
i.e., weld seams in the steam generator and in the main steam and
feedwater piping. Hence, because the affected areas cannot be tested by
Type B or Type C testing, Section IV.A of Appendix J would require that
a Type A test be performed prior to startup following the planned steam
generator replacement.
However, the affected area of the primary containment boundary is
also part of the pressure boundary of an ASME Class 2 component/piping
system and, as such, the planned replacement of the steam generators is
subject to the repair and replacement requirements of ASME Section XI.
The ASME Section XI surface examination, volumetric examination, and
system pressure test requirements are more stringent than the Type A
testing requirements of Appendix J. The acceptance criteria for ASME
Section XI system pressure testing of welded joints is ``zero
leakage.'' In addition, the test pressure for the system pressure test
will be in excess of 20 times that of a Type A test (1356 psig vs. 44.1
spig).
Therefore, the ASME Section XI inspection and testing requirements
more than fulfill the intent of the requirements of Section IV.A of
appendix J.
IV
In the licensee's March 2, 1995, exemption request, the licensee
stated that special circumstance 50.12(a)(2)(ii) is applicable to this
situation, i.e., that application of the regulation is not necessary to
achieve the underlying purpose of the rule.
Appendix J states that the leakage test requirements provide for
periodic verification by tests of the leak tight integrity of the
primary reactor containment. Appendix J further states that the purpose
of the tests ``is to assure that leakage through the primary reactor
containment shall not exceed the allowable leakage rate values as
specified in the Technical Specifications or associated bases.'' Thus,
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 from the requirements
of Section III.D.1(a) of appendix J. The NRC staff has noted that the
licensee's record of ensuring a leak-tight containment has verified
containment integrity and, as noted previously, considerable margin
exists between the Type A test results and the TS allowable leakage
rate. The Type A tests performed in 1984, 1989, and 1990 have all
successfully verified containment integrity. All ``as-found'' Type A
test results since 1984 have been confirmatory of the Type B and C
tests which will continue to be performed. The licensee will perform
the general containment inspection although it is only required by
appendix J (Section V.A.) to be performed 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 NA-2 containment is of the subatmospheric design. During
operation, the containment is maintained at a subatmospheric pressure
(approximately 10 psia) which provides for constant monitoring of the
containment integrity and further obviates the need for Type A testing
at this time. If the containment air partial pressure exceeds the
established TS limit, the unit must be shut down.
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 tests (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% 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 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 allowable 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 those considerations, it is unlikely
that an extension of one cycle for the performance of the appendix J,
Type A test at NA-2 would result in significant degradation of the
overall containment integrity. As a result, the special circumstances
of 10 CFR 50.12(a)(2)(ii) are present in that the application of the
regulation in these particular circumstances is not needed 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 [[Page 17092]] allow a
one-time exemption to permit a schedular extension of one cycle for the
performance of the appendix Type A test, provided that the general
containment inspection is performed, to be acceptable.
Section IV.A of appendix J would normally require that a Type A
test be performed prior to startup following a containment modification
such as the planned steam generator replacement. However, in this case,
the affected area of the primary containment boundary is also part of
the pressure boundary of a ASME Class 2 component/piping system and, as
such, the planned replacement of the steam generators is subject to the
repair and replacement requirements of ASME Section XI. The ASME
Section XI surface examination, volumetric examination, and system
pressure testing requirements are more stringent than the Type A
testing requirements of appendix J. The objective of the Type A test
required by Section IV.A is to assure the leak-tight integrity of the
containment area affected by the modification. The ASME Section XI
inspection and testing requirements more than fulfill the intent of the
requirements of Section IV.A of appendix J. As a result, the special
circumstances of 10 CFR 50.12(a)(2)(ii) are present in that the
application of the regulation in these particular circumstances is not
needed to achieve the underlying purpose of the rule. Therefore, the
NRC staff finds the basis for the licensee's proposed exemption to
allow a one-time exemption from Type A testing for modification of the
primary containment boundary due to the forthcoming NA steam generator
replacement to be acceptable.
Pursuant to 10 CFR 51.32, the Commission has determined that
granting these Exemptions will not have a significant impact on the
environment (60 FR 15945).
The exemption from Section III.D.1.(a) of appendix J to 10 CFR part
50 is effective upon issuance and shall expire at the completion of the
NA-2 1996 refueling outage.
The exemption from Section IV.A of appendix J to 10 CFR part 50 is
effective upon issuance and shall expire at the completion of the NA-2
1995 steam generator replacement refueling outage.
For the Nuclear Regulatory Commission.
Dated at Rockville, MD, this 29th day of March 1995.
Steven A. Varga,
Director, Division of Reactor Projects--I/II, Office of Nuclear Reactor
Regulation.
[FR Doc. 95-8166 Filed 4-3-95; 8:45 am]
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