[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 
1.0L<INF>a. 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 2L<INF>a; in one case the leakage 
was found to be approximately 2L<INF>a; in one case the as-found 
leakage was less than 3L<INF>a; one case approached 10L<INF>a; and in 
one case the leakage was found to be approximately 21L<INF>a. 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 L<INF>a (approximately 200L<INF>a, 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]
BILLING CODE 7590-01-M