ML17056C346

Safety Evaluation Supporting Amend 140 to License DPR-63

ML17056C346
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Site:	Nine Mile Point
Issue date:	04/12/1993
From:	Office of Nuclear Reactor Regulation
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ML17056C347	List: ML17056C346 ML18038A776
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Download: ML17056C346 (30)

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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR R ACTOR REGULATION RELATED TO AMENDMENT NO. 140 TO FACILITY OPERATING LICENSE NO.

DPR-63 NIAGARA MOHAWK POWER CORPORATION NINE MILE POINT NUCLEAR STATION UNIT NO.

1 DOCKE NO. 50-220

1. 0 INTRODUCTION By letter dated November 20,

1990, as superseded February 7,

1992, as supplemented June 22,

1992, January 29,

1993, February 18,

1993, and March 29,

1993, Niagara Mohawk Power Corporation (the licensee) submitted a request for changes to the Nine Mile Point Nuclear Station Unit No.. 1, Technical Specifications (TS).

The requested changes would revise the following sections of the TS:

(1) Table 3.2.7, Reactor Coolant System Isolation Valves; (2) Table 3.2.7. 1, Primary Coolant System Pressure Isolation Valves; (3) 3.3.3/4.3.3, Leakage Rate; (4) 3.3.4, Primary Containment Isolation Valves; (5) Table 3.3.4, Primary Containment Isolation Valves Lines Entering Free Space of the Containment; and (6) associated Bases.

The requested changes would update the reactor coolant system and the primary containment isolation valve tables and the containment leakage rate testing requirements to reflect the NRC staff's conclusions as described in the NRC staff's safety evaluation (SE) dated Hay 6,

1988, as supplemented by the licensee's letter dated July 28,

1988, and the NRC staff's letter dated November 9,

1988, regarding compliance with the requirements of 10 CFR Part 50, Appendix J, at Nine Mile Point Unit 1 (NHP-1).

The NRC staff reviewed the requested changes and additional information requested by letter dated November 30, 1992.

The licensee responded by letters dated January 29, 1993; February 18, 1993; and March 29, 1993.

The discussions below provide the NRC staff's evaluation of the requested

changes, including responses to the request for additional information.

The licensee's February 7,

1992, application superseded an application from the licensee dated November 20, 1990.

The June 22,

1992, January 29,

1993, February 18,

1993, and March 29, 1993, letters provided clarifying information that did not change the initial proposed no significant hazards consideration determination.

2.0 EVALUATION TS Table 3.2.7, Reactor Coolant System Isolation Valves TS Table 3.2.7 regarding reactor coolant system isolation valves has been revised to incorporate the following:

PDR ADQCK 05000220 P

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t The initiating signals for the main steam line and the emergency cooling high point vent to main steam line isolation valves have been identified as, reactor water level low-low or low reactor pressure (with mode switch in run) or main steam line high radiation, or main steam line high flow, or low-low-low condenser vacuum, or high temperature in the steam tunnel.

(2)

The initiating signals for the isolation valves of the emergency cooling steam line drain to main steam line have been identified as high system flow or reactor water level low-low or main steam line isolation.

(3)

Two emergency cooling high point vent lines which will be isolated on reactor water level low-low or main steam line isolation have been added.

(4)

The initial signals for the cleanup system have been revised to show isolation valve closure on reactor water level low-low or high area temperature or liquid poison initiation.

The TS changes proposed in items (I), (2), (3),

and (4) above will provide diversity in the parameters sensed for the initiation of containment isolation in accordance with the requirements as described in Item II.E.4.2 of NUREG-0737.

Therefore, the NRC staff finds them acceptable.

A new footnote (I) indicating the isolation valves which do not have to be vented during the Type A test has been added to the main steam, feedwater, reactor cleanup, liquid poison, shutdown cooling, scram discharge volume system vent, scram discharge volume drain, post accident sampling, reactor recirculation system sampling, emergency cooling steam leaving reactor, and emergency cooling condensate return to reactor lines.

This footnote is consistent with the NRC staff's conclusion described in its SE dated Nay 6, 1988.

Therefore, the NRC staff finds it acceptable.

(6)

Remote manual capability for initiating the AC motor operated isolation valve closure has been added to the feedwater lines.

The NRC staff finds that for each feedwater line, there are two containment isolation valves (one self-acting check valve and one AC motor-operated gate valve).

During a loss-of-coolant accident (LOCA), it is desirable to maintain reactor coolant make-up from all sources of supply.

This AC motor operated valve will not automatically actuate for isolation upon receipt of the signal from the protection system.

However, if there is a need for system or containment isolation, this AC motor-operated valve will be actuated remote-manually.

Thus, the NRC staff finds this proposed remote manual capability for initiating the AC motor-operated isolation 'valve closure acceptable.

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An automatic initiating signal for opening the emergency condenser condensate return valves following a LOCA has been added.

However, TS Table 3.2.7 also indicates that these valves will close upon receipt of the initiating signal from the reactor protection system.

The closure signal is initiated by high flow in the emergency cooling system (emergency condenser),

and that closure signal takes precedence if both the open and close signals should occur simultaneously.

It is appropriate for the closure signal to take precedence over the opening signal.

The parameter initiating closure, high emergency condenser system flow, would indicate a breach in the condenser integrity outside containment.

If the emergency condenser continued to operate, reactor coolant could continue to be released outside containment.

Thus even if an opening signal also occurred in this situation (e.g.,

on low-low reactor water level), the containment isolation valve should close.

The NRC staff finds the proposed TS changes for this item to be acceptable.

The reactor head spray line has been deleted.

The licensee indicated that this line was cut and capped and would be Type B tested.

The NRC staff finds this proposed TS change acceptable.

Footnote (2) has been added to indicate that ther e will be flow (whi ch forms a water seal) through the control rod drive hydraulic line isolation valves.

Therefore, these isolation valves will receive a

water leak rate test in accordance with the inservice test (IST) program.

This proposed footnote is consistent with the NRC staff's conclusion described in its SE dated May 6, 1988, therefore; the NRC staff finds it acceptable.

The penetrations, with footnote (5), for core spray condensate supply and core spray system valves have been added to the table.

Footnote (5) requires these valves be tested in accordance with TS 4.2.7.1a.

The NRC staff finds the addition of these penetrations acceptable.

Footnote (3) for the penetrations of core spray injection isolation valves indicates that the inside core spray injection isolation valves are water sealed during and after an accident and are leak rate tested with water in accordance with the IST program.

This part of the proposed footnote is consistent with the NRC staff's conclusion described in its SE dated May 6, 1988; therefore, the NRC staff finds it acceptable.

The footnote further indicates that the outside core spray injection isolation valves are electrically locked open with their breakers locked in the off position, and, therefore, these isolation valves do not have to be tested under the IST or Appendix J leakage test programs.

The NRC staff finds that the locked open, outside core spray injection valves need not be Type C tested

because, in effect, they are not containment isolation valves.

The second valve isolation function is provided by further out system valves (40-03 or 40-13) which are included in TS Table 3.2.7.

(12)

Footnote (4) has been added to indicate that the isolation valves for the core spray high point vent and core spray pump discharge lines which are water sealed shall be tested during each refueling outage not to exceed two years.

Leakage rates shall be limited to 0.5 gpm per nominal inch of valve diameter up to a maximum of 5 gpm.

Based on its review, the NRC staff finds this footnote in compliance with the Appendix J requirements and, therefore, acceptable.

(13)

A common footnote (1) for the post-accident reactor sampling and reactor recirculation system sampling penetrations indicates that the isolation valves do not have to be vented during Type A test.

However, Type C

leakage from these valves will be added to the Type A test results.

This common footnote for the above penetrations is consistent with the NRC staff's conclusion described in its SE dated May 6, 1988.

Therefore, the NRC staff finds it acceptable.

An additional footnote (6) was provided for the post-accident reactor sampling penetration to indicate that the first isolation valve which is a self-actuating check valve will be tested in accordance with Section 4.3.4(c) of the TS.

This TS section indicates that the above isolation valve will only be tested for operability at least once per operating cycle.

It is difficult, if not impossible, to perform a standard Type C

test on an excess-flow check valve with typical Type C testing equipment.

The problem is making the valve go closed, which only happens when flow through it exceeds a certain amount.

Of course, if a valve is not closed, one cannot measure its leak rate.

Therefore, excess-flow check valves are typically not Type C tested, but rather given a functional test of the type that would be required by the proposed footnote.

Additionally, the redundant containment isolation valve in this line is included in TS Table 3.2.7, will be Type C tested, and its leak rate added to the Type A test result, in accordance with the NRC staff's SE dated May 6, 1988.

Based on the above considerations, the NRC staff finds the testing of this penetration to be appropriate and the proposed TS to be acceptable.

TS Table 3.2.7. 1, Primary Coolant System Pressure Isolation Valves These valves also serve as containment isolation valves and are also listed in TS Table 3.2.7.

The footnotes on TS Table 3.2.7. 1 have been revised to specify leakage rates which are in compliance with Appendix J requirements; therefore, the proposed changes are acceptable.

TSs 3.3.3 and 4.3.3, Leakage Rate The intent of the proposed changes in these TS sections is to make the surveillance requirements for Types A, B, and C leak tests consistent with the

NRC staff's conclusion described in its SE dated Hay 6,

1988, and Appendix J requirements.

The NRC staff's evaluation of the significant proposed changes is provided below:

TS 3.3.3, Leakage Rate - Objectives The following paragraph has been added to this section as part of the objective for primary. containment leakage rate tests:

"To assure that periodic surveillance of reactor containment penetrations and isolation valves are performed so that proper maintenance and repairs are made during the service life of the containment, and systems and components penetrating primary containment."

The NRC staff finds this proposed addition to the objectives of leakage rate testing acceptable.

TS 4.3.3, Leakage Rate - Specifications TS 4.3.3.a, Integrated Primary Containment Leakage Rate - Type A Test The following specifications have been proposed to replace the previous specifications:

(I)

Integrated leak rate tests shall be performed at the test pressure (P,) of 22 psig.

Containment pressure shall not be permitted to decrease more than I psi below P,.

(2)

Type B and C tests should be completed prior to each Type A test.

Type B and C leakages (penalties) not accounted for in the Type A test shall be incorporated as minimum pathway additions to Upper Confidence Limit (UCL) to determine the overall as left integrated leakage rate.

The above proposed TS changes (Items (I) and (2)) are in compliance with the requirements of Appendix J.

Therefore, the NRC staff finds them acceptable.

(3) If the leakage rate exceeds the acceptance criterion, corrective action shall be required.

If, during the performance of a Type A test, excessive leakage occurs through locally testable penetrations or isolation valves to the extent that it would interfere with satisfactory completion of the test, these leakage paths may be isolated and the Type A re-test continued until completion.

The Type A test shall be considered

'a failed test.

A local leakage test shall be performed at P, before and after the repair of each isolated leakage path.

The sum of the post

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(4) repaired local leakage rates and the UCL shall be less than 75 percent of the maximum allowable leakage rates, L, (22).

Local leakage rates shall not be subtracted from the Type A test results to determine the acceptability of a test.

The as found and as left leakage data values of excessive leakage areas beyond acceptance criteria shall be provided to NRC.

Based on its review, the NRC staff finds the above proposed TS acceptable since it is consistent with the NRC staff's conclusion described in its SE dated Nay 6,

1988, and the Appendix J requirements.

A Type A test shall last a minimum of eight (8) hours with leakage rates calculated based on "Total Time" method.

If a twenty-four (24) hour test is performed the "Mass Point" method will be used to calculate leakage rates.

A verification test shall be performed following each Type A

test.

The verification test provides a method for assuring that systematic error or bias is given adequate consideration.

During the verification test, containment pressure may not decrease more than one (I) psi below P,.

Based on its review, the NRC staff finds this proposed TS and its Bases acceptable.

TS 4.3.3.b, Acceptance Criteria - Type A Test The the (I)

(2)

(3) following acceptance criteria have been proposed to replace previous acceptance criteria for Type A tests:

The maximum allowable leakage rate L, (22) shall not exceed

1. 19 weight percent of the contained air per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at the test pressure of 22 psig (P,).

The maximum allowable operational leakage rate L which shall be met prior to power operation following a Type A test (either as measured or following repairs and retest) shall not exceed 0.75 L, (0.892 weight percent per day).

Mhen adding the leakage rate measured during a Type C test to the results of a Type A test, the leakage rate shall be determined using minimum pathway analysis.

Based on its review, the NRC staff finds the above proposed TS changes (Items (I), (2) and (3)) consistent with the NRC staff's conclusion described in its SE dated Nay 6, 1988 and, therefore, acceptable.

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TS 4.3.3.c, Frequency The following frequency has been proposed to replace the previous frequency for Type A tests:

(I)

Three Type A tests shall be conducted during each ten year service interval at approximately equal intervals.

The third test will be conducted when the plant is shutdown for the 10 year inservice inspections.

Based on its review, the NRC staff finds that this proposed TS is consistent with the requirements of Appendix J and, therefore, acceptable.

(2)

Retesting (a) If a Type A test fails to meet the acceptance criteria of 4.3.3.b.(l),

a Corrective Action Plan that focuses attention on the cause of the problem shall be developed and implemented.

A Type A test that meets the requirements of 4.3.3.a.(3) and 4.3.3.b.(2) is required prior to plant start-up.

A report of the corrective action following the failed Type A shall be submitted to the NRC for review and approval with the Containment Leak Test Report.

Based on its review, the NRC staff finds the proposed TS consistent with the requirements of Appendix J and, therefore, acceptable.

(b) If any periodic Type A test fails to meet the acceptance criteria of 4.3.3.b.(l),

the test schedule for subsequent Type A tests will be reviewed and approved by the NRC.'ased on its review, the NRC staff finds this proposed TS in compliance with the requirements of the Appendix J and, therefore, acceptable.

(c) If two consecutive periodic Type A tests (not including an immediate retest under (a)) fail to meet the acceptance criteria of 4.3.3.a.(3),

4.3.3.b.(1) or 4.3.3.b.(2),

not withstanding the periodic retest schedule of 4.3.3.c.(l),

a Type A test must be performed at each refueling outage or every 18 months, whichever occurs first, unless alternative leak test requirements are accepted by the NRC by means of specific exemption from Appendix J per 10 CFR 50. 12.

This testing shall be

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performed until two consecutive periodic Type A tests (not including an immediate retest under (a)) meet the acceptance criteria of 4.3.3.a.(3), 4.3.3.b.(l) and 4.3.3.b.(2),

then the retest schedule specified in 4.3.3.c.(l) should be resumed.

Based on its review, the NRC staff finds the proposed TS consistent with the requirements of Appendix 0 and, therefore, acceptable.

4.3.3.d, Local Leak Rate - Type B and Type C Tests The following TS have been proposed to replace the current TS related to Types B and C tests:

(1)

Primary containment testable penetrations and isolation valves required to be Type B or Type C tested by regulatory requirements, shall be tested at a pressure of 35.0 psig (P )

each major refueling outage, not to exceed two years, except as provided below:

Bolted double gasket seals which shall be tested whenever the seal is closed after being opened and at least at each refueling outage not to exceed a two year interval.

Type B tests for primary containment penetrations employing a

continuous leakage monitoring system shall be conducted at intervals not to exceed three years.

Based on its review, the NRC staff finds the above proposed TS in accordance with the requirements of Appendix J and, therefore, acceptable.

(2)

When system pressure (P,

) on the opposite side of the isolation valve under test cannot be reduced to atmospheric

pressure, then the test pressure shall not be less than P, +

Psys The NRC staff finds this proposed test pressure in compliance with the requirement of Appendix J and, therefore, acceptable.

(3)

Personnel airlocks shall be leak tested in accordance with the following:

(a)

The airlocks shall be tested at a test pressure of 35 psig following a refueling outage or maintenance outage requiring drywell access prior to primary containment integrity being required.

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Appendix J states, in part, that airlocks opened during periods when containment is not required by the plant's TS shall be tested at the end of such periods at not less than P,.

Therefore, assuming refueling and maintenance outages are the only periods when containment integrity is not required by the TS, the NRC staff finds this proposed TS in compliance with the requirement of Appendix J and acceptable.

(b)

Airlocks opened during periods when primary containment integrity is required shall be tested within three days after being opened.

For airlock doors opened more frequently than once every three days, the airlocks shall be tested at least once every three days.

Based on its review, the NRC staff finds this proposed TS in accordance with the requirements of Appendix J and, therefore, acceptable.

(c)

The airlocks shall be tested every six months at a test pressure of 35 psig.

Based on its review, the NRC staff finds this proposed TS in accordance with the requirements of Appendix J and, therefore, acceptable.

(d)

Leakage rate for airlocks shall not exceed 0.05 L, at 35 psig.

The proposed acceptance criterion of 0.05 L

at 35 psig is an appropriate number, consistent with standard TS and the TS of many operating plants.

Therefore, the NRC staff finds the proposed TS to be acceptable.

(4)

Primary containment penetrations and isolation valves that are not defined as Type B or Type C test components (e.g.,

seal welded cold instrument lines, CRD lines, drywell to wetwell connections, etc.) shall not be individually tested.

The penetrations will be considered as integral parts of the Type A test.

Based on its review, the NRC staff finds this proposed TS consistent with the NRC staff's conclusion described in its SE dated May 6,

1988, and the requirements of Appendix J; therefore, it is acceptable.

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aa 4.3.3.e, Acceptance Criteria (for Type B and C tests)

The licensee has proposed the following acceptance criteria for Types B

and C tests:

The combined leakage r ate for penetrations and valves subject to Type B and C tests determined by maximum pathway analysis shall be less than 0.60 L,. If this value is exceeded, repairs and retests shall be performed to correct the condition.

The NRC staff finds these proposed acceptance criteria for Types B and C tests in compliance with the requirements described in Appendix J and, therefore, acceptable.

4.3.3.f, Continuous Leak Rate Monitor (1)

The licensee proposes to change the phrase,

"...the containment shall be continuously monitored for gross leakage by review of the inerting system makeup requirements...."

to "...the containment shall be monitored for gross leakage by a weekly review of the inerting system makeup requirements...."

Based on its review, the NRC staff finds this proposed TS change acceptable.

(2)

The licensee proposes to change the sentence, "This monitoring system may be taken out of service for the purpose of maintenance or testing but shall be returned to service as soon as practical." to "This monitoring system may be taken out of service for the purpose of maintenance or testing but shall be returned to service as these activities are completed."

Based on its review, the NRC staff finds this proposed TS change acceptable.

4.3.3.g, Inspection The licensee proposes to change the sentence, "The accessible interior surface of the drywell shall be visually inspected each operating cycle for evidence of deterioration." to "The accessible interior surfaces of the primary containment shall be visually inspected each, operating cycle for evidence of deterioration."

Based on its review, the NRC staff finds this proposed TS change acceptable.

In addition to the above proposed TS changes, the licensee has updated the Bases associated with the above TS Sections 3.3.3 and 4.3.3 to reflect the NRC staff's conclusion described in its SE dated Hay 6,

1988, and Appendix J

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Based on its review, the NRC staff offers no objection to the proposed changes.

3.3.4, Primary Containment Isolation Valves A.

TS 3.3.4b.

has been revised to allow 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to isolate a line when an isolation valve becomes inoperable.

The TS previously required an isolation valve to be closed if it become inoperable without specifying a time for completing this action.

Based on its review, the NRC staff finds that this 4-hour time limit is consistent with the time allowed by the standard TS.

Therefore, the NRC staff finds this proposed TS change acceptable.

B.

TS Table 3.3.4 regarding primary containment isolation valves has been revised to incorporate the followings:

(1)

The maximum operating time for the drywell and suppression chamber vent and purge valves has been changed from 60 seconds to 15 and 30 seconds for the Pn/DC solenoid (air) and motor-operated valves, respectively.

The licensee stated that these closure times are consistent with the NRC staff's requirements to resolve Hulti-Plant Action (HPA) Item B-24.

As a part of HPA Item B-24, the NRC staff evaluated the radiological consequences of a LOCA during containment purging at NHP-1.

Based on the reduction of the purge isolation valve closure time from the previous value of 60 seconds to 15 seconds, the NRC staff in a SE, dated December 8,

1983, concluded that the radiological consequences of a LOCA during purging at NHP-1 to be within the 10 CFR Part 100 dose guideline values.

The NRC staff further stated that, "The staff will, therefore, require NHP-1 to reduce the technical specification limit on purge/vent valves isolation system response time to 15 seconds or less."

In a memorandum for M. R. Butler, thru J. Kudrick, from H. fields, subject:

Hotor Operated Containment Isolation Valves in Purge and Vent Lines, dated January 30, 1985, the NRC staff considered the fact that plants with motor-operated purge/vent valves might not achieve full closure of those valves within the 15-second criterion set forth in HPA Item B-24, during a design-basis accident (DBA) with simultaneous loss of offsite power.

The NRC staff surveyed approximately 50 operating plants, including NHP-l, and discovered that 11 had motor-operated valves (HOVs) in their purge/vent systems.

Several, including NHP-l, were

identified as having an air-operated valve in series with each HOV, and the NRC staff concluded that the HPA Item B-24 criterion of complete valve closure within 15 seconds following onset of a DBA LOCA was satisfied for these plants.

The implied basis for this conclusion was that the air-operated valve, with its rapid (15 second) closure time, is sufficient by itself to satisfy the 15-second criterion, despite the fact that the redundant HOV may take longer than 15 seconds to close.

Plants were built with one air-operated valve and one HOV in each purge/vent line presumably in an effort to provide diversity and avoid potential common-mode failures that could prevent both valves in a line from isolating at the same time.

Further, the NRC staff recognizes that, in fact, large HOVs that close in 15 seconds or less are not available.

Considering the above, the NRC staff finds that the 15-second and 30-second closure times for, respectively, the air-operated and motor-operated purge/vent valves are acceptable.

Therefore, the NRC staff finds the proposed TS to be acceptable.

(2)

Footnote (1) has been added to the drywell equipment drain line, drywell floor drain line, 012 Hz/Oz sampling lines, the service water connection line, the Nz Purge-Tip indexers, and traversing incore probe lines.

This footnote indicates that the isolation valves for these lines do not have to be vented during Type A test.

However, Type C test leakage from these valves will be added to the Type A test results.

This footnote is consistent with the NRC staff's conclusion described in its SE, dated Hay 6, 1988; therefore, the NRC staff finds it acceptable.

(3)

The suppression chamber water makeup line has been deleted.

The licensee indicated that this line had been flanged and designated as a penetration subjected to Type B only.

The NRC staff finds this proposed TS change acceptable.

(4)

Footnote (2) has been added to indicate that the isolation valves for the reactor cleanup system relief valve discharge, containment spray drywell and suppression chamber common supply, drywell branch and suppression chamber

branch, containment spray test line to torus, and emergency cooling vent to torus lines are provided with a water seal capability.

No Appendix J testing is required.

pg I4 I

J This footnote is consistent with the NRC staff's conclusion described in its SE, dated May 6, 1988.

Therefore, the NRC staff finds it acceptable.

Footnote (3) has been added to indicate that the penetrations for core spray pump suction and containment spray pump suction from suppression chamber lines are water leak rate tested in accordance with the IST program.

The proposed footnote is consistent with the NRC staff's conclusion described in its SE, dated May 6, 1988; therefore, the NRC staff finds it acceptable.

Footnote (4) has been added to indicate that the isolation valves for the core spray high point vent, core spray pump discharge, and condensate supply lines shall be tested during each refuel outage not to exceed 2 years to be consistent with Appendix J water seal testing requirements.

Leakage rates shall be limited to 0.5 gpm per nominal inch of valve diameter up to a maximum of 5 gpm.

Based on its review, the NRC staff finds this footnote consistent with the NRC staff's conclusion described in its SE, dated May 6,

1988, and therefore, acceptable.

The maximum operating time for the core spray pump discharge test line to the suppression chamber has been reduced from 90 to 27 seconds.

The licensee indicated that this proposed TS change is consistent with the Appendix K reload analysis for core spray initiation and flow requirement.

Based on its review, the NRC staff finds this proposed TS change acceptable.

The initiating signals for isolating the core spray discharge test line to the suppression chamber has been revised to include the high drywell pressure signal.

The NRC staff finds this proposed TS change acceptable.

Footnote (5) has been added to indicate that the valves in closed loops inside containment do not meet the requirements of Appendix J,Section II.H.

The NRC staff finds this footnote consistent with the NRC staff's conclusion described in its SE, dated May 6,

1988, and therefore, acceptable.

The maximum operating time for the recirculation pump cooling and drywell cooler water return line DC motor isolation valves has been increased from 30 to 60 seconds.

This proposed TS change is within the containment isolation valve closure time limit as described in Section 6.2.4, "Containment Isolation System", of the Standard Review Plan.

Therefore, the NRC staff finds it acceptable.

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The NRC staff required an editorial modification to Note (2) of Table 3.3.4 of the proposed TS to clarify that no Appendix J or IST leakage rate testing of the affected valves is required.

However, the affected valves do require periodic full stroke exercising and stroke time testing in accordance with the requirements of the NMP-1 IST program.

This modification to Note (2), which changed "IST testing" to "IST leakage rate testing,"

was discussed with and agreed to by Hr. Nick Spagnoletti of the licensee's staff during a

telephone discussion on March 30, 1993.

3. 0 STATE CONSULTATION In accordance with the Commission's regulations, the New York State official was notified of the proposed issuance of the amendment.

The State official had no comments.

4.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes surveillance requirements.

The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released

offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure.

The Commission has previously issued a

proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding (56 FR 4866 and 57 FR lllll). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).

Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

5.0 CONCLUSION

The Commission has concluded, based on the considerations discussed

above, that:

(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed

manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributors:

D.

Shum J. Pulsipher D. Brinkman Date:

April 12, 1993

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Docket No. 50-220 Hr. B. Ralph Sylvia Executive Vice President, Nuclear Niagara Mohawk Power Corporation 301 Plainfield Road

Syracuse, New York 13212 April 12, 1993

SUBJECT:

ISSUANCE OF AMENDMENT FOR NINE MILE POINT NUCLEAR STATION UNIT NO.

1 (TAC NO. H79135)

The Commission has issued the enclosed Amendment No. 140 to Facility Operating License No.

DPR-63 for the Nine Mile Point Nuclear Station Unit No.

1 (NMP-1).

The amendment consists of changes to the Technical Specifications (TSs) in response to your application transmitted by letter dated November 20,

1990, which was superseded by an application dated February 7, 1992.

The February 7,

1992, application was supplemented by letters dated June 22,

1992, and January 29, 1993.

The TSs and supporting information in Attachment 2 of the February 7,

1992, application and supplemental letters dated June 22,

1992, and January 29,

1993, were replaced by information provided in a letter dated February 18,

1993, and supplemented by a letter dated March 29, 1993.

The amendment revises TSs 3.2.7. 1, 3.3.3, 4.3.3, and 3.3.4 and associated Bases to update these TSs to conform to the requirements of 10 CFR Part 50, Appendix J, and NRC Safety Evaluations (SEs) dated May 6,

1988, and November 9, 1988.

A copy of the related SE is enclosed.

A Notice of Issuance will be included in the Commission's next regular biweekly Federal Receister notice.

Sincerely, n

o

Enclosures:

l.

Amendment No. 140 to DPR-63 2.

Safety Evaluation, cc w/enclosures:

See next page Original signed by:

Donald S.

Brinkman, Senior Project Manager Project Directorate I-1 Division of Reactor Projects I/II Office of Nuclear Reactor Regulation LA:PDI-1 CVo an PM'DI-1 B

DBrinkman:smm R

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93

//7 93 g /8/93 Z

93 OFFICIAL RECORD COPY FILENAME: NM179135.AMD

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