[Federal Register Volume 59, Number 58 (Friday, March 25, 1994)]
[Unknown Section]
[Page 0]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 94-7063]


[[Page Unknown]]

[Federal Register: March 25, 1994]


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NUCLEAR REGULATORY COMMISSION
[Docket No. 50-333]

 

Power Authority of the State of New York; James A. FitzPatrick 
Nuclear Power Plant

Exemption

I

    The Power Authority of the State of New York (PASNY or the 
licensee) is the holder of Facility Operating License No. DPR-59, which 
authorizes operation of the James A. FitzPatrick Nuclear Power Plant 
(the facility or FitzPatrick). The license provides, among other 
things, that the facility is subject to all the rules, regulations, and 
Orders of the Nuclear Regulatory Commission (the Commission) now or 
hereafter in effect.
    The facility is a boiling water reactor located at the licensee's 
site in Oswego County, New York.

II

    Section III of Appendix J to 10 CFR part 50 requires the 
development of a program to conduct periodic leak testing of the 
primary reactor containment and related systems and components, and 
components penetrating the primary containment pressure boundary. The 
interval between local leak rate tests for containment isolation valves 
(Type C tests) is specified by section III.D.3 to be no greater than 2 
years.

III

    By letter dated January 11, 1994, the licensee requested a 
schedular exemption pursuant to 10 CFR 50.12(a) from the requirements 
of 10 CFR part 50, Appendix J, section III.D.3. Specifically, the 
licensee requested one-time relief from the requirement to perform Type 
C tests (local leak rate tests) at intervals of no greater than 2 years 
for the shutdown cooling isolation valves (10MOV-17 and 10MOV-18). This 
one-time only delay, until the next refueling outage currently 
scheduled to begin in November 1994, was requested for the performance 
of these leakage tests. The licensee's request was necessitated by the 
extended 1991-1993 refueling outage and the length of the current 
operating cycle.
    The shutdown cooling valves were previously tested during the last 
refueling outage (Reload 10/Cycle 11). This was an extended outage that 
began in November 1991 and ended in January 1993. The Type C tests on 
the subject valves were performed on May 30, 1992, for the outboard 
isolation valve 10MOV-17, and June 5, 1992, for the inboard isolation 
valve 10MOV-18. Subsequent delays in the outage resulted in these tests 
being performed significantly in advance of the start of the operating 
cycle (more than 7 months prior to the end of the outage). As a result, 
the 2 year test interval will be reached for these valves (May 30, 
1994/June 5, 1994) 6 to 7 months prior to the next scheduled refueling 
outage. The exemption would permit a deferral in the performance of the 
Type C test of the shutdown cooling isolation valves beyond the 2-year 
limiting interval to the next refueling outage.
    The only effective means of removing reactor core decay heat is 
with the shutdown cooling mode of the RHR system. This requires both of 
the stated isolation valves to be in the open position. The shutdown 
cooling mode of the RHR system must be removed from service for 
approximately 24 hours to perform a local leak rate test (Type C) of 
its isolation valves. This is the time required to tag-out the system, 
drain the line, perform the test, refill the line, and return the 
system to service. To avoid overheating the reactor coolant system with 
the shutdown cooling mode inoperable, one of the following two 
conditions must exist:
    1. The reactor needs to be shutdown for several months to permit 
sufficient reduction in decay heat levels for use of an alternate 
shutdown cooling method without placing the plant in the refueling 
condition. The alternate cooling method with the highest heat removal 
capacity is the Reactor Water Cleanup system in the blowdown mode. 
However, the reactor must be shutdown for more than 3 months before 
this method can handle the decay heat load.
    2. The plant needs to be in the refueling condition; i.e., reactor 
head removed, reactor cavity flooded up and connected to the spent fuel 
pool. This permits the removal of the normal shutdown cooling system 
from operation and testing of these valves.
    A three week surveillance/maintenance outage is planned for spring 
1994. However, the decay heat levels present during any outage less 
than several months precludes the use of the alternate cooling method 
without placing the plant in the refueling configuration. The exemption 
would preclude the need to place the plant in the refueling 
configuration prior to the next scheduled refueling outage. Without the 
exemption, the licensee would be required to remove the drywell and 
reactor heads and connect the reactor cavity to the spent fuel pool 
solely for the purpose of testing the shutdown cooling isolation 
valves. Placing the plant in the refueling configuration would 
significantly lengthen the spring 1994 outage and would require 
significant resources. Furthermore, placing the plant in the refueling 
configuration to accommodate testing of the isolation valves would 
significantly increase occupational radiation exposures. For these 
reasons, the licensee has determined that compliance with the 
regulation would result in undue hardship and costs.

IV

    Section III.D.3 of Appendix J to 10 CFR part 50 states that Type C 
tests shall be performed during reactor shutdowns for refueling, at an 
interval not to exceed 2 years. The licensee has requested a one-time 
exemption from the regulations.
    The operating configuration of the shutdown cooling isolation 
valves and the RHR system when the reactor coolant system is 
pressurized (greater than 75 psig) substantially minimizes the 
possibility of gross leakage through these valves. A high reactor 
pressure interlock, as well as plant operating procedures, assures that 
these isolation valves are closed whenever reactor pressure is above 75 
psig. This protects the low pressure RHR system from 
overpressurization. The RHR system suction piping is designed for 450 
psig. Gross leakage while the reactor is pressurized would be detected 
by high pressure on the RHR suction piping or an increase in 
suppression pool inventory. Consequently, the maintenance of normal 
operating status of the RHR system assures the absence of gross leakage 
through these valves.
    These valves also receive an isolation signal in the event of a 
plant accident (reactor vessel low water level or high drywell 
pressure). This assures isolation of a potential leakage path from the 
reactor coolant system to the reactor building. For this path to exist, 
leakage through both isolation valves, and a breach of the RHR system 
piping would need to occur simultaneously. Since the isolation valves 
are maintained closed with the reactor pressurized, it is improbable 
the leakage through the valves will increase while the plant is 
operating. The redundant isolation valves provide two leakage barriers 
which limit the pathway leakage rate to that experienced by the valve 
with smallest leakage rate. For these reasons, the potential for 
significant leakage to the reactor building by way of the shutdown 
cooling line is minimal.
    The penetration included in the licensee's schedular exemption 
request represents only 6.4 percent of the total ``as left'' leakage at 
the beginning of the current operating cycle. The total ``as left'' 
minimum path leakage for all penetrations was only 0.073 La and the 
total ``as left'' minimum path leakage for the penetration addressed in 
the proposed exemption was only 0.0046 La. The replacement of both 
isolation valves with valves of improved design provides added 
confidence that excessive leakage will not be experienced. The inboard 
valve 10MOV-18 was replaced during the 1985 refueling outage and has 
successfully passed three out of four Type C tests performed during 
refueling outages since its replacement. The outboard isolation valve 
10MOV-17 was replaced with a similarly designed new valve during the 
last refueling outage (1992). The limited number of valve strokes these 
valves are subject to over any one operating cycle minimizes valve 
degradation due to wear. This provides reasonable assurance that the 
requested surveillance interval expansion will not result in the Types 
B and C leakage rate total exceeding the 0.6 La limit of 10 CFR part 
50, Appendix J. Therefore, the Commission concludes that there are no 
significant radiological environmental impacts associated with the 
proposed schedular exemption.
    The 2-year interval requirement for Type C testing is intended to 
be often enough to preclude significant deterioration between tests and 
long enough to permit the tests to be performed during routine plant 
outages. Leak rate testing of containment isolation valves during plant 
shutdown is preferable because of the lower radiation exposures to 
plant personnel. Furthermore, some containment isolation valves cannot 
be tested at power. For those valves that cannot be tested during power 
operation, or for which testing at power would yield unnecessary 
radiation exposure of personnel, the NRC staff believes the increase in 
confidence of containment integrity following a successful test is not 
significant enough to justify the hardships and costs associated with 
performing the tests within the 2-year time period.

V

    The Commission has determined that, pursuant to 10 CFR 50.12(a)(1), 
this exemption is authorized by law, will not present 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 the particular 
circumstance is not necessary to achieve the underlying purpose of the 
rule. The underlying purpose of Section III.D.3 of Appendix J to 10 CFR 
part 50 is to provide an interval short enough to prevent serious 
deterioration from occurring between tests and long enough to permit 
testing to be performed during regular plant outages. For containment 
isolation valves that cannot be tested at power, or for containment 
isolation valves where testing involves unreasonable risk to personnel 
and equipment, the increased confidence in containment integrity 
following successful testing is not significant enough to justify the 
hardships associated with performing the test within the 2-year 
interval. Specifically, any potential incremental benefit of performing 
the tests within the 2-year requirement would not be sufficient to 
offset the increased occupational radiation exposure associated with 
testing, the risk to plant safety associated with removing the primary 
method of decay heat removal from service, and the undue financial 
burden of placing the plant in the refueling configuration and 
significantly extending the length of the spring 1994 maintenance/
surveillance outage. The licensee has presented information accepted by 
the Commission, which gives a high degree of confidence that the 
components affected by this exemption will not degrade to an 
unacceptable extent. Acceptable leakage limits are defined in sections 
III.B.3(a) and III.C.3 of Appendix J to 10 CFR part 50.
    Pursuant to 10 CFR 51.32, the Commission has determined that 
granting the above exemption will have no significant impact on the 
quality of the human environment (March 16, 1994, 59 FR 12382).
    This Exemption is effective upon issuance and shall expire prior to 
restart following the next FitzPatrick refueling outage which is 
currently scheduled to commence in November 1994.

    Dated at Rockville, Maryland, this 18th day of March 1994.

    For the Nuclear Regulatory Commission.
Frederick J. Hebdon,
Acting Director, Division of Reactor Projects--I/II, Office of Nuclear 
Reactor Regulation.
[FR Doc. 94-7063 Filed 3-24-94; 8:45 am]
BILLING CODE 7590-01-M