ML17056C274

Proposed TS Table 3.2.7 Re RCS Isolation Valves,Table 3.2.7.1 Re Primary Coolant Sys Pressure Isolation Valves & Table 3.3.4 Re Primary Containment Isolation Valves,Lines Entering Free Space of Containment

ML17056C274
Person / Time
Site:	Nine Mile Point
Issue date:	02/18/1993
From:	NIAGARA MOHAWK POWER CORP.
To:
Shared Package
ML17056C275	List: ML17056C274 ML18038A734
References
NUDOCS 9303020492
Download: ML17056C274 (72)

Text

ATTACHMENT1 NIAGARAMOHAWKPOWER CORPORATION LICENSE NO. DPR-63 DOCKET NO. $ 0-220 PROPOSED CHANGES TO TECHNICAL SPECIFICATIONS Replace existing pages with the attached revised and additional pages. These pages have been retyped in their entirety with marginal markings to indicate changes to the text.

Exi tin Pa e R~id P 118 118 118a 118a 119 119 add 119a add 119b 120 120 120b 120b 135 135 136 136 137 137 138 138 139 139 140 140 add 140a add 140b add 140c 141 141 142 142 143 143 add 143a 144 144 146 146 147 147 148 148 add 148a add 148b 149 149

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LIMITING CONDITIONS FOR OPERATION Table 3.2.7 REACTOR COOLANT SYSTEM ISOLATION VALVES Location Relative Maximum Initiating Signal (All No. of Valves to Primary Normal Oper. Time Action on Valves have Remote Une or System (Each Une) Containment Position Motive Poweri (Sec) Initiating Signal Manual Backup)

Reactor water level low-low or low reactor pressure, (with mode switch in run) or main Main Ste'am Inside Open AC Motor 10 Close steam line high (Two Lines) Outside Open Pn/DC Solenoid 10 Close radiation, or main steam line high flow, or low-low-low condenser vacuum, or high temperature in the steam tunnel Feedwater Outside Open AC Motor 60 Remote ij/lanual (Two Lines) Outside Open Self Act. Ck.

Emer enc Cooiin Steam Leavin (Two Lines)

Reactor(1) Outside Open AC Motor 38 Close / High emergency Outside Open DC Motor 38 Close )i cooling system flow Condensate Return to Reactor( ) Inside Closed Self Act. Ck.

(Two Lines) Outside Closed Pn/DC Solenoid 60 Close High emergency cooling system flo 60 Open Reactor water level low-low or high reactor pressure AMENDMENT NO., 9, 1 118

C= p LIMITING CONDITIONS FOR OPERATION Table 3.2.7 (Continued)

REACTOR COOLANT SYSTEM ISOLATION VALVES Location Relative Maximum Initiating Signal (AII No. of Valves to Primary Normal Oper. Time Action on Valves Have Remote Line or System (Each Une) Containment Position Motive Power (Sec) Initiating Signal Manual Backup) I Reactor Cleanu Water Leavin Reactor Inside Open AC Motor 18 Close Reactor water level (One Line) Outside Open DC Motor 18 Close low-low or high area temperature or liquid Water Return to Reactor Inside Open AC Motor 18 Close poison initiation (One Line) Outside Open Self Act. Ck.

Shutdown Coolin Water Leavin Reactor( ) Inside Closed AC Motor 40 Close (One Line) Outside Closed DC Motor 40 Close Reactor water level low-.low, or high area temperature Water Return to Reactor( ) Inside Closed AC Motor 40 Close (One Line) Outside Closed Self Act. Ck.

AMENDMENT NO. 1 118a

( P LIMITINGCONDITIONS FOR OPERATION Table 3.2.7 (Continued)

REACTOR COOLANT SYSTEM ISOLATION VALVES Location Relative Maximum Initiating Signal (All No. of Valves To Primary Normal Oper. Time Action on Valves Have Remote Une or System (Each Une) Containment Position Motive Power (Sec) Initiating Signal Manual Backup)

Lktnid Poisoni i Inside Closed Self Act. Ck.

(One Line) Outside Closed Self Act. Ck.

Control Rod Drive H draulic{ I Inside Open Self Act. Ck.

(One Line) Outside Open Self Act. Ck.

Scram Dischar e Volume{ ) Outside Open Pn/AC Solenoid 10 Close

~deters Vent" Automatic or manual (One Line) reactor scram Scram Discher e Volume{ ) Outside Open Pn/AC Solenoid 10 Close

~deters Drain'

{One Line)

~Cole S ts Reactor water level Core S ra In'ection{ ) Inside Closed AC Motor 22.5 Open low-low or high (Two Lines) Outside Open AC Motor 22.5 Open drywell pressure coincident with reactor vessel pressure less than 365 psig Core S ra Hi h Point Vent{

(Two Lines)

) Inside Outside Closed Closed AC Motor Pn/DC Solenoid 27 27 Close Close Reactor water level low-low or high

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drywell pressure Core S ra Condensate Su I Outside Open Self Act. Ck.

(Keep Fill)

(Two Lines)

Core S ra S stem Valves Outside Closed Self Act. Ck.

(Two Lines)

Core S ra Pum Dischar e Outside Closed AC Motor 27 Close Reactor water level (Two Test Lines to Suppression Chamber) low-low or high drywell pressure AMENDMENT NO... 1 119

LIMITING CONDITIONS FOR OPERATION Table 3.2.7 (Continued)

REACTOR COOLANT SYSTEM ISOLATION VALVES Location Relative Maximum Initiaung Signal (All No. of Valves To Primary Normal Oper. Time Action on Valves Have Remote Une or System (Each Line) Containment Position Motive Powers (Sec) Initiating Signal Manual Backup)

Post Accident Reactor Sam lin ( Outside Open Self Act. Flow (One Line) Fuse Outside Closed Pn/DC Solenoid 30 Close Reactor water level low-low or main steam line high radiation or low-low-low condenser vacuum or reactor low pressure, (with mode Reactor Recirculation S stem Sam lin ( ) Inside AC Motor Closed 20 Close switch in run) or high (One Line) Outside Closed DC Motor 20 Close temperature in the steam tunnel or main steam line high flow AMENDMENT NO. 119al

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Notes:

" Pn - Pneumatically Operated

"" Section 3.1.1e for LCO Requirements (1) These valves do not have to be vented during the Type A test. However, Type C leakage from these valves is added to the Type A test results, if not vented.

(2) These valves have flow through them during and following an accident (a water seal) and receive a water leak rate test in accordance with the IST Program.

(3) The inside core spray injection isolation valves are water sealed during and after an accident. These valves are leak rate tested with water in accordance with the IST Program. The outside core spray injection isolation valves are open with their breakers locked in the off position. Therefore, the outside core spray injection isolation valves do not have to be tested under the IST or Appendix J Leakage Program.

(4) These valves are provided with a water seal. Valves shall be tested during each refuel outage not to exceed two years 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.

(5) These valves are tested in accordance with Section 4.2.7.1a.

(6) The self actuating flow fuse is tested in accordance with Section 4.3.4c.

AMENDMENT NO. 119b I

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BASES FOR 3.2.7 AND 4.2.7 REACTOR COOLANT SySTEM ISOLATION VALVES Double isolation valves are provided in lines which connect to the reactor coolant system to assure isolation and minimize reactor coolant loss in the event of a line rupture. The specified valve requirements assure that isolation is already accomplished with one valve shut or provide redundancy in an open line with two operative valves. Except where check valves are used as one or both of a set of double isolation valves, the isolation valves shall be capable of automatic initiation and the closure times presented in Table 3.2.7. These closure times were selected to minimize coolant losses in the event of the specific line rupturing. Using the longest closure time on the main-steam;line valves following a main-steam-line break (Section XV C.1.0) ", the core is still covered by the time the valves close. Following a specific system line break, the cleanup and shutdown cooling closing times will upon initiation from a low-low level signal limit coolant loss such that the core is not uncovered. Feedwater flow would quickly restore coolant levels to prevent clad damage. Closure times are discussed in Section VI-D.1.0 " .

The valve operability test intervals are based on periods not likely to significantly affect operations, and are consistent with testing of other systems. Results obtained during closure testing are not expected to differ appreciably from closure times under accident conditions as in most cases, flow helps to seal the valve.

The test interval of once per operating cycle for automatic initiation results in a failure probability of 1 1 x 10 (Fifth Supplement,

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p. 115)( ) that a line will not isolate. More frequent testing for valve operability results in a more reliable system.

(1) UFSAR (2) FSAR AMENDMENT NO. 120

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TABLE 3.2.7.1 PRIIVIARY COOLANT SYSTEM PRESSURE ISOLATION VALVES Maximum(a)

~Setem Valve No. Allowable Leaka e

1. Core Spray System 40-03 a5.0 gpm 40-13 <5.0 gpm

2. Condensate Supply to Core Spray (Keep Fill 40-20 <5.0 gpm System) 40-21 <5.0 gpm 40-22 <5.0 gpm 40-23 <5.0 gpm Footnote:

(a) 1;- Leakage rates shall be limited to 0.5 gpm per nominal inch of valve diameter up to,a maximum of 5 gpm.

2. Test differential pressure shall not be less than 150 psid.

3. The observed leakage at test differential pressure shall be adjusted to the functional maximum pressure differential.

AMENDMENT NO. 120b

~,p LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.3.3 LEAKAGE RATE 4.3.3 LEAKAGE RATE A licabili A Iicabilit Applies to the allowable leakage rate of the primary Applies to the primary containment system leakage containment system. rate.

~Ob'ec ive: ~Ob ective:

To assure the capability of the containment in limiting To verify that the leakage from the primary radiation exposure to the public from exceeding containment system is maintained within specified values specified in 10CFR100 in the event of a loss- values.

of-coolant accident accompanied by significant fuel cladding failure and hydrogen generation from a metal-water reaction.

To assure that periodic surveillances 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.

Whenever the reactor coolant system temperature is a. Inte rated Primar Containmen Leaka e Rate-above 215F the primary containment leakage rate I shall be within the limits of 4.3.3.b.

(1) Integrated leak rate tests shall be performed at the test pressure (Pt) of 22 psig. I AMENDMENT NO. 135

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LIIVIITINGCONDITION FOR OPERATION SURVEILLANCE REQUIREMENT Containment pressure shall not be permitted to decrease more than one (1) psi below Pt.

(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 the Upper Confidence Limit (UCL) to determine the overall as left integrated leakage rate.

(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 the 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 Pt before and after the repair of each isolated leakage path. The sum of the post repaired local leakage rates and the UCL shall be less than 75 percent of the maximum allowable leakage rate, 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 the NRC.

AMENDMENT NO. 136

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LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREIVIENT (4) Closure of the containment isolation valves for the purpose of the test shall be accomplished by the means provided for normal operation of the valves.

(5) 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 (1) psi below Pt.

b., Acce tance Criteria - T e A Tes it i Tha maximum allowable leakage rate Lt (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 (Pt). I (2) The maximum allowable operational leakage, )

rate L,o (22) which shall be met prior to power operation following a Type A test AMENDMENT NO. 137

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LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT (either as measured or following repairs and retest) shall not exceed 0.75 4 (22) (0.892 weight percent per day).

(3) When adding the leakage rate measured

'uring a Type C test to the results of a Type A test, the leakage rate shall be determined using minimum pathway analysis.

c. ~Fre uenc (1) 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.

(2) Retesting (a) If a Type A test fails to meet the acceptance criteria of 4.3.3.b.(1), 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 AMENDMENT NO. 138

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LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREIVIENT 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.

(b) If any periodic Type A test fails to meet the acceptance criteria of 4.3.3.b.(1), the test schedule for subsequent Type A tests will be reviewed and approved by the NRC.

(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) and 4.3.3.b.(2), not-withstanding the periodic retest schedule of (b), 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 10CFR50.12. This testing shall be performed until two AMENDMENT NO. 139

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LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT consecutive periodic Type A tests (not including an immediate retest under (a)) meet the acceptance criteria of 4.3.3.a.(3) or 4.3.3.b. (1), then the retest schedule specified in 4.3.3.c.(1) should be resumed.

d. Local Leak Rate-T e 8 and T e 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 (Pa) each major refueling outage, not to exceed two years, except as provided in (a) and (b) below.

(a) Bolted double gasketed 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 ~

(b) Type 8 tests for primary containment penetrations employing a continuous leakage monitoring system shall be conducted at intervals not to exceed three years.

AMENDMENT NO. 140

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LIMITINGCONDITION FOR OPERATION SURVEILLANCE REQUIREMENT (2) When system pressure (Psys) 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 Pa + Psys.

(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.

(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.

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

(d) Leakage rate for airlocks shall not exceed 0.05La at 35 psig.

AMENDMENT NO. 140a I

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LIMITINGCONDITION FOR OPERATION SURVEILLANCE REQUIREMENT (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.

e. Acce tance Criteria - T e B and T e C Tests The combined leakage rate for penetrations and valves subject to Type B and C tests determined by maximum pathway analysis shall be less than 0.60 La. If this value is exceeded, repairs and retests shall be performed to correct the condition.

f. Con inuous Leak Rate Moni or (1) When the primary containment is inerted, the containment shall be monitored for gross leakage by a weekly review of the inerting system makeup requirements.

(2) 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.

AMENDMENT NO. 140b I

I LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT

g. ~lne ection The accessible interior surfaces of the primary containment shall be visually inspected each operating cycle for evidence of deterioration.

AMENDMENT NO. 140c )

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BASES FOR 3.3.3 AND 4.3.3 LEAKAGE RATE The primary containment preoperational test pressures are based upon the calculated primary containment pressure response in the event of a loss-of-coolant accident. The peak drywell pressure would be 35 psig which would rapidly reduce to 22 psig within 100 seconds following the pipe break. The total time the drywell pressure would be above 22 psig is calculated to be about 10 seconds. Following the pipe break, the suppression chamber pressure rises to 22 psig within 10 seconds, equalizes with drywell pressure and thereafter rapidly decays with the drywell pressure decay. l l The design pressures of the drywell and suppression chamber are 62 psig and 35 psig, respectively.l l As pointed out above, the pressure response of the drywell and suppression chamber following an accident would be the same after about 10 seconds. Based on the calculated primary containment pressure response discussed above and the suppression chamber design pressure; primary containment preoperational test pressures were chosen. Also, based on the primary containment pressure response and the fact that the drywell and suppression chamber function as a unit, the primary containment will be tested as a unit rather than testing the individual components separately.

The design basis loss-of-coolant accident was evaluated at the primary containment maximum allowable accident leak rate of 1.9%/day at 35 psig. The analysis showed that with this leak rate and a standby gas treatment system filter efficiency of 90 percent for halogens, 95 percent for particulates, and assuming the fission product release fractions stated in TID-14844, the maximum total whole body passing cloud dose is about 6.0 rem and the maximum total thyroid dose is about 150 rem at the site boundary considering fumigation conditions over an exposure duration of two hours. The resultant doses that would occur for the duration of the accident at the low population distance of 4 miles are lower than those stated due to the variability of meteorological conditions that would be expected to occur over a 30-day period. Thus, the doses reported are the maximum that would be expected in the unlikely event of a design basis loss-of-coolant accident. - These doses are also based on the assumption of no holdup in the secondary containment resulting in a direct release of fission products from the primary containment through the filters and stack to the environs. Therefore, the specified primary containment leak rate and filter efficiency (Specification 4.4.4) are. conservative and provide margin between expected off-site doses and 10 CFR 100 guideline limits.

AMENDMENT NO. 141

0 0 J BASES FOR 3.3.3 AND 4.3.3 LEAKAGE RATE The maximum allowable leakage rate (La) is 1.5%/day at a pressure of 35 psig (Pa). This value for the test condition was derived from the [

maximum allowable accident leak rate of about 1.9%/day when corrected for the effects of containment environment under accident and test conditions. In the accident case, the containment atmosphere initially would be composed of steam and hot air depleted of oxygen whereas under test conditions the test medium would be air or nitrogen at ambient conditions. Considering the differences in mixture composition and temperatures, the appropriate correction factor applied was 0.8 and determined from the guide on containment testing. (3)

Although the dose calculations suggest that the allowable test leak rate could be allowed to increase to about 3.0%/day before the guideline thyroid dose limit given in 10 CFR 100 would be exceeded, establishing the limit at 1.5%/day provides an adequate margin of safety to assur~

the health and safety of the general public. It is further considered that the allowable leak rate should not deviate significantly from th~

containment design value to take advantage of the design leak-tightness capability of the structure over its service lifetime. Additional margin to maintain the containment in the "as built" condition is achieved by establishing the allowable operational leak rate. The operational limit is derived by multiplying the allowable test leak rate by 0.75 thereby providing a 25% margin to allow for leakage deterioration which may occur during the period between leak rate tests.

A reduced pressure test program is used for the integrated test. The test pressures are based on loss-of-coolant accident conditions. The peak primary containment pressure following a loss-of-coolant accident would be 35 psig. This would rapidly reduce to 22 psig. The total time drywell pressure would be above 22 psig would be about 10 seconds. Preoperational integrated leak tests were performed at test pressures at 35 psig and 22 psig. Subsequent integrated tests are performed at a test pressure of 22 psig.

Closure of the containment isolation valves for the purpose of the test is accomplished by the means provided for normal operation of the valves.

The reactor is vented to the containment atmosphere during testing.

The acceptance criteria states that the maximum allowable leakage rate (4) shall not exceed 1.19 weight percent of the contained air in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at 22 psig (Pt). This corresponds to the maximum allowable leakage rate (La) of 1.5 weight percent at 35 psig (Pa). The maximu allowable test leak rate 4 (at 22 psig) shall not exceed the 1.5%/day times the square root of the ratio of the pressures Pt (at 22 psig) and Pa (at 35 psig), respectively since the ratio of measured leakages for Nine Mile Point Unit 1 is 0.735. The allowable operational leakage rate, 4o (at 22 psig) shall not exceed 75 percent of Lt (at 22 psig) and shall be met prior to resumption of power operation following a test.

r AMENDMENT NO. 142

r BASES FOR 3.3.3 AND 4.3.3 LEAKAGE RATE The primary containment leak rate test frequency is based on maintaining adequate assurance that the leak rate remains within the specification.

The leak rate test frequency is based on 10 CFR 50 Appendix J. I The penetration and air purge piping leakage test frequency, along with the containment leak rate tests, is adequate to allow detection of leakage trends. Whenever a double-gasketed penetration (primary containment head equipment hatches and the suppression chamber access hatch) is broken and remade, the space between the gaskets is pressurized to determine that the seals are performing properly. The test pressure of 35 psig is consistent with the accident analyses and the maximum preoperational leak rate test pressure. It is expected that the majority of the leakage from valves, penetrations and seals would be into the reactor building. However, it is possible that leakage into other parts of the~

facility could occur. Such leakage paths that may affect significantly the consequences of accidents are to be minimized. If the leakage rates~

of the double-gasketed seal penetrations, testable penetration isolation valves, containment air purge inlets and outlets and the vacuum relief valves are at the maximum specified, they will total 90 percent of the allowed leak rate.( ) Hence, 10 percent margin is left for leakage through walls and untested components.

Leakage from airlocks is measured under accident pressures in accordance with 10 CFR 50 Appendix J.

Monitoring the-nitrogen make-up requirements of the inerting system provides a method of observing leak rate trends. This instrumentation ]

equipment must be periodically removed from service for test and maintenance, but this out-of-service time will be kept to a practical minimum.

The test program follows the guidelines stated in the Bechtel Topical Report. I4I This program provides adequate assurance that the test results realistically estimates the degree of containment leakage following a loss-of-coolant accident. The containment leakage rate is calculated using the Absolute Methodology.l I Containment leakage results are presented in the test report as calculated using the Total Time and Mass Point techniques. The results of local leak rate tests, including, "as-found" and "as-left" leakages, are also included in the containment leak test report.

t AMENDMENT NO. 143

~ t BASES FOR 3.3.3 AND 4.3.3 LEAKAGE RATE The specific treatment of selective valve arrangements including the acceptability of the interpretations of 10 CFR 50 Appendix J requirements are given in References 5, 6, and 7. They serve as the bases for alternative test configurations (e.g., reverse accident, multi-valve, water leakage flow tests) as well as relaxations from previous leakage limits or constraints.

References:

(1) FSAR, Volume II, Appendix E (2) UFSAR, Section Vl B.2.1 (3) TID-20583, Leakage Characteristics of Steel Containment Vessels and the Analysis of Leakage Determinations (4) BN-TOP-1 "Testing Criteria for Integrated Leakage Rate Testing of Primary Containment Structures for Nuclear Power Plants," Revision 1, Bechtel Corporation, November 1, 1972 (5) NRC Safety Evaluation Report dated May 6, 1988, "Regarding Proposed Technical Specifications and Exemption Requests Related to Appendix J."

(6) Niagara Mohawk Letter dated July 28, 1988, "Clarifications, Justifications 5 Conformance with 10 CFR 50 Appendix J SER."

(7) NRC Letter dated November 9, 1988, "Review of the July 28, 1988 Letter on Appendix J Containment Leakage Rate Testing at Nine Mile Point Unit 1."

(8) ANSI/ANS - 56.8 - 1987, "Containment System Leakage Testing Requirements."

AMENDMENT NO. 143a I

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LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.3.4 PRIMARY CONTAINMENT ISOLATION VALVES 4.3.4 PRIMARY CONTAINMENTISOLATION VALVES Applies to the operating status of the system of Applies to the periodic testing requirements of the isolation valves on lines open to the free space of the primary containment isolation valve system.

primary containment.

~Ob ective:

~ob ective:

To assure the operability of the primary containment To assure that potential leakage paths from the isolation valves to limit potential leakage paths from primary containment in the event of a loss-of-coolant the containment in the event of a loss-of-coolant accident are minimized. accident.

The primary containment isolation valves surveillance shall be performed as indicated (See Table 3.3 4)

a. Whenever the reactor coolant system temperature is greater than 215F, all a. At least once per operating cycle the operable containment isolation valves on lines open to the isolation valves that are power operated and free space of the primary containment shall be automatically initiated shall be tested for operable except as specified in 3.3.4b below. automatic initiation and closure times.

b. In the event any isolation valve becomes b. At least once per quarter all normally open power inoperable the system shall be considered operated isolation valves shall be fully closed and operable provided that within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> at least reopened.

one valve in each line having an inoperable valve is in the mode corresponding to the isolated condition.

AMENDMENT NO. 144

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LIMITING CONDITION FOR OPERATION Table 3.3e4 PRIMARY CONTAINMENT ISOLATION VALVES LINES ENTERING FREE SPACE OF THE CONTAINMENT Location Relative Max(mum Initiating Signal (All No. of Valves To Primary Normal Oper. Time Action on Valves Have Remote Une or System (Each Une) Containment Position Motive Power {Sec) Initiating Signal Manual Backup)

D well Vent gi Pur e Ng Connection Outside Closed Pn/DC Solenoid 15 Close Reactor water level (One Line) Outside Closed AC Motor 30 Close low-low or high dryweil pressure or Air Connection Outside Closed Pn/DC Solenoid 15 Close high radiation at stack (One Line) Outside Closed AC Motor 30 Close monitoring Su ression Chamber Vent 5 Pur e N2 Connection Outside Closed Pn/DC Solenoid 15 Close Reactor water level

{One Line) Outside Closed AC Motor 30 Close low-low or high drywell pressure or Air Connection Outside Closed Pn/DC Solenoid 15 Close high radiation at stack (One Line) Outside Closed AC Motor 30 Close monitoring

~ore ii N k~Makeu Reactor water level (One Line) Outside Closed Pn/DC Solenoid 60 Close low-low or drywell high pressure Su ressiaa Chamber Nk~Mskeu Reactor water level (One Line) Outside Closed Pn/DC Solenoid 60 Close low-low or drywell t high pressure D well E ui ment Drain Line(1) Inside Open AC Motor 60 Close (One Line) Outside Open Pn/DC Solenoid 60 Close Reactor water level low-low or drywell D well Floor Drain Line(1 I Inside Open AC Motor 60 Close high pressure (One Line) Outside Open Pn/DC Solenoid 60 Close  !

Vacuum Relief Atmosphere to Pressure Suppression System Outside Closed Pn/DC Solenoid Open Negative pressure (Three Lines) relative to atmosphere Outside Closed Self Act. Ck. I Reactor Cleanu S stem Relief Valve I

~oisaher e (One Line to Suppression Chamber) Outside Closed Self Act. Ck. I AMENDMENT NO. 1 4 146

LIMITINGCONDITIONS FOR OPERATION Table 3.3e4 (Continued)

PRIMARY CONTAINMENT ISOLATION VALVES LINES ENTERING FREE SPACE OF THE CONTAINMENT Location Relative Maximum Initiating Signal (All No. of Valves To Primary Normal Oper. Time Action on Valves Have Remote Une or System (Each Une) Containment Position Motive Power D (Sec) =

Initiating Signal Manual Backup) I Outside Open Pn/DC Solenoid 60 Close (Two Lines)

Su ression Chamber Su I Outside Open Pn/DC Solenoid 60 Close Reactor Water level (One Line) low-low or high drywall pressure

~Rett Return Outside Open Pn/DC Solenoid 60 Close (One Line)

Su ression Chamber Return Outside Open Pn/DC Solenoid 60 Close (One Line)

H202~012 Sam lln Outside Open Pn/DC Solenoid 60 Close Reactor water level (Three Lines) low-low or high drywell pressure Su ression Chamber Su I Outside Open Pn/DC Solenoid 60 Close (One Line)

~Dwelt Returnl 1 Outside Open Self Act. Ck.

(One Line)

Su ression Chamber Return( ) Outside Self Act. Ck.

Open (One Line)

AMENDMENT NO. 1 6 147

UNITING CONDITION FOR OPERATION Table 3.3e4 (continued)

PRIMARY CONTAINING>>ENT ISOLATION VALVES LINES ENTERING FREE SPACE OF THE CONTAINMENT Location Relative Nlaxlmum Inltfat>>ng Signal (All No. of Valves To Primary Normal Oper. Time Aotlon on Valves Have Remote Une or System (Eaoh Une) Containment Position Motive Power~ (Seo) Initiating Signal Manual Baokup)

~Cores ro

~om Sucrion"'Four Outside Open AC Motor 90 Remote Manual Lines From Suppression Chamber)

Outside Closed AC Motor 27 Close Reactor water level (Two Test Lines to Suppression Chamber) low-Iow or high drywell pressure Condensate Su I Outside Open Self Act.

(Keep Fill)

(Two Lines) Ck.'losed Core S ra Hi h Point Vent>> )

Outside Pn/DC Solenoid 27 Close Reactor water level (Two Lines) Inside Closed AC Motor 27 Close low-low or high h

drywall pressure Containment S ra 0 well & Su ression Chamber>> )

Outside Pn/DC Solenoid Common Su I Open eo Remote Manual (Four Lines)

Outside Closed Self Act. Ck.

(Four Lines)

Su ression Chamber Branch>> ) 20 ~

Outside Closed Self Act. Ck.

(One Branch for Each System)

Pum Suction From Su ression Chamber>> I Outside Open AC Motor 70 Remote Manual (Four Lines)

Containment S ra Test Line to Torus( I Outside Closed AC Motor eo Remote Manual (One Line) mer enc Coolin Vent to Torus( Outside Closed AC Motor Remote Manual (One Line)

AMENDMENT NO. 1 2 148

LIMITING CONDITIONS FOR OPERATION Table 3.3.4 (Continued)

PRIMARY CONTAINMENT ISOLATION VALVES LINES ENTERING FREE SPACE OF THE CONTAINMENT Location Relative Maximum Initiating Signal (All No. of Valves To Primary Normal Oper. Time Action on Valves Have Remote Une or System (Each Une) Containment Position Motive Powers (Sec) Initiating Signal Manual Backup)

Containment Atmos here Monitorin Su I Outside Open Pn/DC Solenoid 60 Close Reactor water level Line low-low or high (One Line) drywell pressure Containment Post LOCA Vent Outside Closed Pn/DC Solenoid 60 Close Reactor water level (Two Lines) low-low or high drywell pressure N2 Pur e - TIP Indexers( I Outside Closed Self Act. Ck.

(One Line)

Traversin Incore Probe( I Outside Closed AC Motor 60 Close Reactor water level (Four Lines) low-low or high drywell pressure Breathin Air Connection Inside Closed (One Line) Outside Closed Local Manual Service Water Connection( ) Inside Closed (One Line) Outside Closed LINES WITH A CLOSED LOOP INSIDE CONTAINMENTVESSELS Recirculation Pum Coolin Water( )

Supply Line Outside Open Self Act. Ck.

Return Line Outside Open DC Motor 60 Remote Manual I D ell Cooler Water(

Supply Line Outside Open Self Act. Ck.

Return Line Outside Open DC Motor 60 Remote Manual AMENDMENT NO. 148a i

1 Notes:

Pn - Pneumatically Operated One valve in each separate line and one valve in each common line.

(1) These valves do not have to be vented during the Type A test. However, Type C leakage from these valves is added to the Type A test results, if not vented.

(2) These valves are provided with a water seal capability. No Appendix J or IST testing is required.

(3) These valves are water leak rate tested and acceptance criteria are established in accordance with the IST Program.

(4) These valves are provided with a water seal. Valves shall be tested during each refuel outage not to exceed two years 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.

(5) These valves do not meet the requirements of 10CFR50 Appendix J, Section II-H. No testing required.

AMENDMENT NO. 148b I

I I BASES FOR 3.3.3 AND 4.3.4 PRIIVIARY CONTAINMENT ISOLATION VALVES Double isolation valves are provided on lines penetrating the primary containment and open to the free space of the containment. Closure of one of the valves in each line would be sufficient to maintain the integrity of the pressure suppression system. Except where check valves are used as one or both of a set of double isolation valves, the isolation closure times are presented in Table 3.3.4. Automatic initiation is required to minimize the potential leakage paths from the containment in the event of a loss-of-coolant accident. Details of the isolation valves are discussed in Section VI-D.(1) For allowable leakage rate specification, see Section 3.3.3/4.3.3. I For the design basis loss-of-coolant accident fuel rod perforation would not occur until the fuel temperature reached 1700F which occurs in approximately 100 seconds. A required closing time of 60 seconds for all primary containment isolation valves will be adequate to prevent fission product release through lines connecting to the primary containment.

For reactor coolant system temperatures less than 215F, the containment could not become pressurized due to a loss-of-coolant accident.

The 215F limit is based on preventing pressurization of the reactor building and rupture of the blowout panels.

The test interval of once per operating cycle for automatic initiation results in a failure probability of 1.1 x 10 7 that a line will not isolate (Fifth Supplement, p. 115). More frequent testing for valve operability results in a more reliable system. I In addition to routine surveillance as outlined in Section VI-D.1.0 " each instrument-line flow check valve will be tested for operability. All instruments on a given line will be isolated at each instrument. The line will be purged by isolating the flow check valve, opening the bypass valves, and opening the drain valve to the equipment drain tank. When purging is sufficient to clear the line of non-condensibles and crud, the flow-check valve will be cut into service and the bypass valve closed. The main valve will again be opened and the flow-check valve allowed to close. The flow-check valve will be reset by closing the drain valve and opening the bypass valve depressurizing part of the system. Instruments will be cut into service after closing the bypass valve. Repressurizing of the individual instruments assures that flow-check valves have reset to the open position.

An in-depth review of the NMP-1 design and operation relative to Appendix J requirements has evaluated the various system/valving configurations. The results of the evaluation and subsequent clarifications are reflected in this specification and its bases.

(1) UFSAR (2) Nine Mile Point Nuclear Generation Station Unit 1 Safer/Corecool/GESTR-LOCA Loss of Coolant Accident Analysis, NEDC-31446P, Supplement 3, September, 1990.

(3) FSAR (4) NRC Safety Evaluation Report, dated May 6, 1988, "Regarding Proposed Technical Specifications and Exemption Requests Related to Appendix J."

(5) Niagara Mohawk Letter dated July 28, 1988, "Clarifications, Justifications 5. Conformance with 10CFR50 Appendix J SER."

AMENDMENT NO. 149

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ATTACHMENT2 REVISED SUPPORTING II%FORMATION:

PERTAINING TO FEBRUARY 7, 1992 LICENSE AMENDMENT

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This Technical Specification change is to update the Reactor Coolant, Primary Coolant System Pressure Isolation Valves, and Primary Containment Isolation Valve Tables 3.2.7, 3.2.7.1, and 3.3.4 and to make changes required to comply with the NRC Safety Evaluation Report dated May 6, 1988 regarding proposed Technical Specifications and Exemption Requests related to Appendix J. In addition, Technical Specification Sections 3.3.3 and 4.3.3, "Leakage Rate" are being revised consistent with 10CFR50 Appendix J. When testing is performed in accordance with the IST Program, the frequency of test is that specified in Appendix J.

Each change is identified below by page.

~Pa e ii

1) The word "Maximum" has been added to the column heading Qger Time to be more descriptive.

2) A 'k))h following the table.

2 AAR<<A ~i I h d)g I I

3) The initiating signal for Main Steam Line Isolation Valves have been correctly identified as low-low-low condenser vacuum and not just low condenser vacuum. In addition, isolation on low reactor pressure with the mode switch in run has been added.

4) A new footnote (1) has been added to Main Steam, Feedwater, and Emergency Cooling Steam Leaving Reactor and Condensate Return to Reactor. This footnote states that these valves do not have to be vented during the Type A test. However, Type C leakage from these valves is added to the Type A test results. This footnote is consistent with the NRC SER dated May 6, 1988.

5) Emergency Cooling Steam Line Drain to Main Steam and Emergency Cooling High Point Vent to Main Steam have been removed from Table 3.2.7, since they are outside the Reactor Coolant Isolation boundary.

6) Remote Manual has been added to the initiating signal column for the AC motor operated feedwater valve and the Normal Position of "Open" for the Feedwater Self Act. Ck. has been added for consistency.

7) The initiating signal for Emergency Cooling System Isolation has been identified as high emergency cooling system flow.

8) The Emergency Cooling System return line to the reactor vessel has been correctly identified as the Condensate Return to Reactor instead of Condenser Return to Reactor and the Normal Position of "Closed" for Condensate Return Self Act. Ck. valve has been added for consistency.

9) The automatic initiating signals'which open the Emergency Cooling Condensate Return to Reactor have been added, since the valves would open post LOCA.

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10) Notes have been deleted from bottom of Page 118. They appear at the end of the table. The note originally designated as (1) is now designated as (*).

Pa e 11 The word "Maximum" haa been added to the column heading Queer Time to be more descriptive.

2) An asterisk (g) has been added to the Motive Power column heading to reference a footnote following the table.

3) Footnote (1) has been added to the Reactor Cleanup System and Shutdown Cooling System Lines. This footnote is consistent with the NRC SER dated May 6, 1988.

4) The initiating signals for the Reactor Cleanup System have been revised to show closure on reactor water level low-low or high area temperature or liquid poison initiation. Isolation on high system pressure, low system flow or high system temperature has been deleted,

5) The Normal Position "Open" for the Reactor Cleanup Water Return to Reactor Self Act. Ck, has been added for consistency,

6) The Normal Position "Closed" for the Shutdown Cooling Water Return to Reactor Self Act.

Ck. has been added for consistency.

Pa e 11 The Reactor Head Spray Line has been deleted. This line has been cut and flanged. The line is now a Type B tested penetration instead of a Type C tested Reactor Coolant System Isolation Valve.

2) Footnote (1) has been added to the Liquid Poison and Scram Discharge Volume System Vent and Drain. This footnote is consistent with the NRC SER dated May 6, 1988. The Normal Position for the Liquid Poison Self Act. Ck. has been changed to "Closed" for consistency.

3) Footnote (2) has been added to the Control Rod Drive Hydraulic Line. This footnote is self-explanatory. The Normal Position "Open" for the Control Rod Drive Hydraulic Line Self Act. Ck. has been added for consistency.

4) The motive power for the Scram Discharge Volume System Vent and Drain has been designated Pn/AC Solenoid and system titles have been corrected.

5) The Core Spray Injection, Core Spray Condensate Supply (Keep Fill), Core Spray System Valves, and Core Spray Pump Discharge have been added to Table 3.2.7.

6) Footnote (3), which applies to the Core Spray injection (inside and outside) has been added.

This footnote is self-explanatory.

V

7) Footnote (4), which applies to the Core Spray High Point Vent and Core Spray Pump Discharge has been added. This footnote is self-explanatory.

8) The closure time for the Core Spray High Point Vent Line Isolation Valves has been changed from 30 to 27 seconds. This change is consistent with the'Appendix K reload analysis for Core Spray initiation and slow requirements.

9) Footnote (5), which applies to Core Spray Condensate Supply (Keep Fill), and Core Spray System Valves has been added. This footnote is self-explanatory.

10) The notes have been moved to the end of the table. The note associated with Motive Power to define R.M.P.O and A.I.A.O has been deleted as it is no longer required.

~Pa e~11 L (New Page)

1) The Post Accident Reactor Sampling and Reactor Recirculation System Sampling Lines have been added to Table 3.2.7.

2) Footnote (1) has been added to Post Accident Reactor Sampling and Reactor Recirculation System Sampling Lines. This footnote is consistent with NRC SER dated May 6, 1988.

3) Footnote (6) has been added to the Post Accident Reactor Sampling line. This footnote is self-explanatory.

P~ae 11 b (New Page)

New page 119b has been added to contain the notes from Table 3.2.7.

P~ae 120b

1) Original Footnotes (a) 1, 2 and 3 have been deleted. Footnotes 4 and 5 have been renumbered and new footnote 3 has been added. The deletion of original footnotes 1, 2 and 3 is supported by the following:

'I a) NRC Generic Letter 89-04, position 10 identifies the leakage trending requirements as ineffective and therefore compliance with such requirements (ref: ASME Section XI, Article IWV-3427(b)) for Appendix J air leakage testing is not warranted.

b) Valves 40-20, 21, 22 and 23 are exempted from ASME Section XI leakage trending requirements since they are less than 6".

c) Leakage trending requirements for Primary Coolant System Isolation Valves are exempted per IST Program Relief Request VG-2 as approved by NRC SER dated March 7, 1991 (TAC No. 79447).

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The addition of new footnote 3 clarifies the testing requirements regarding adjustment of leakage rates based on the difference of test pressures vs. maximum design functional pressure. This is in conformance with ASME Section XI, Article IWV-3423(e) and the new ASME O&M 10 Standard.

Pae 1 -14c These pages have been revised to incorporate the Surveillance Requirements for Type A, B and C Appendix J leak tests. This includes adding new pages 140a, b, and c. These changes are consistent with the requirements of Appendix J.

Pa e 141 142 14 and 143 The Bases for 3.3.3 and 4.3.3 Leakage Rate has been updated to reflect the requirement of Appendix J and the NRC SER dated May 6, 1988. The design leak rate of 0.5%/day has been deleted, as this figure has been identified as a design objective, not a design limit.

~Pa e 144 Specification 3.3.4.b was 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 specification previously required an isolation valve to be closed if it became inoperable without specifying a time for completing this action. This is considered a clarification to the specification since it is understood that a finite amount of time is required to complete the action.

The four hour time limit is reasonable and consistent with the time allowed by Standard Technical Specifications.

Pa e 14 I) Tl h<<()h b ddt h

2) Footnote (a) has been deleted for consistency. Opening of these valves is controlled by operating procedures.

3) The Air/DC Solenoid Isolation Valves have had their motive power designation changed to Pn/DC Solenoid to be more descriptive.

4) 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. This is consistent with a telecon dated April 9, 1986 and NMPC letter dated April 11, 1990.

5) Footnote (b) has been deleted from Drywell and Suppression Chamber N, Makeup for consistency. Opening of the valves is controlled by operating procedures.

6) Footnote (1) has been added to the Drywell Equipment Drain and the Drywell Floor Drain lines. This footnote is consistent with the NRC SER dated May 6, 1988.'

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7) The Floor Drain line is now identified as the Drywell Floor Drain Line to be more descriptive.

8) The Suppression Chamber Water Makeup line has been deleted, as this line has been flanged and has been designated a Type B penetration.

9) The motive power of the Vacuum Relief Valves has been changed from AC motor to Pn/DC Solenoid, which is the correct designation of the motive power for this valve. The Normal Position of "Closed" for the Vacuum Relief Self Act. Ck. has been added for consistency.

10) Footnote (2) has been added to the Reactor Cleanup System Relief Valve Discharge Line.

This is consistent with NRC SER dated May 6, 1988; The Normal Position of "Closed" for the Reactor Cleanup System Relief Valve Discharge Self Act. Ck. has been added for consistency.

~Pa e 147 The footnote (*) has been added to the Motive Power column.

2)

• The 0, sampling lines have been identified as H,O, ¹11 and ¹12 sampling lines for greater clarity. All of the H~O, sample lines have been added to Table 3.3.4 for completeness.

3) Footnote (b) and notes on the bottom of the page have been deleted from 0, sampling for consistency. Opening of these valves are controlled by operating procedures.

4) Footnote (1) has been added to H,O, ¹12 sampling lines. This footnote is consistent with the NRC SER dated May 6, 1988.

~Pa e 148 The footnote (*) has been added to the Motive Power column.

2) Note (c) was removed from this page to be consistent with NRC SER dated May 6, 1988.

3) Footnote (3), which is self-explanatory, has been added to the Core Spray Pump Suction and Containment Spray Pump Suction from Suppression Chamber. This footnote is consistent with NRC SER dated May 6, 1988.

4) The maximum operating time for the Core Spray Pump Discharge line to the suppression chamber has been changed from 90 to 27 seconds. This change is consistent with the Appendix K Reload Analysis for Core Spray initiation and flow requirements.

5) The initiating signals for isolating the Core Spray Pump Discharge test line to the suppression chamber has been revised to include the high drywell pressure signal. These valves isolate on containment isolation signals of reactor water level low-low or high drywell pressure.

However, the high drywell pressure initiating signal was not included in the original Technical Specifications.

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6) Core Spray Condensate Supply (Keep Fill) and Core Spray High Point Vent valves have been added to Table 3.3.4 as Primary Containment Isolation Valves.

7) Footnote (4) has been added to the Core Spray Pump Discharge, This is consistent with NRC SER dated May 6, 1988.

8) Recirculation Pump Cooling Water Supply and Drywell Cooler Water Supply Valves titled LINES WITH A CLOSED LOOP INSIDE CONTAINMENT VESSELS were moved to new page 148a.

9) Deleted heading LINES WITH A CLOSED LOOP OUTSIDE CONTAINMENTVESSELS to be consistent with NRC SER dated May 6, 1988.

10) Footnote (2), which is self-explanatory, has been added to the Containment Spray Drywell and Suppression Chamber Common Supply, Drywell Branch, Suppression Chamber Branch Lines, Containment Spray Test Line to Torus and Emergency Cooling Vent to the Torus.

Those Containment Spray Isolation Valves designated Air/DC Solenoid have had their Motive Power designation changed to Pn/DC Solenoid.

12) The Normal Position of "Closed" for the Drywell and Suppression Branch Self Act. Ck. has been added for consistency.

13) Suppression Chamber Branch No. of Valves single asterisk has been changed to double asterisk and notes have been deleted from the bottom of page 148 and appear on next page 148b.

n

14) Containment Spray Test Line to Torus has been added to Table 3.3.4. This is consistent with the NRC SER dated May 6, 1988.

15) Emergency Cooling Vent to the Torus has been added to Table 3.3.4 to update the table to represent a historical design change.

P~ae 148 (New Page)

Containment Atmosphere Monitoring Supply Line, Containment Post LOCA Vent, N, Purge-TIP Indexers, Traversing Incore Probe, Breathing Air Connection, and Service Water Connection have been added to Table 3.3.4 to update the table to represent historical design changes and for consistency,

2) Footnote (1) has been added to N, Purge-TIP Indexers, Traversing Incore Probe, and Service Water Connection. This is consistent with NRC SER dated May 6, 1988, and for consistency.

3) The Recirc. Pump Cooling Water Supply and Drywell Cooler Water Supply Valves were retitled Recirculation Pump Cooling Water and Drywell Cooler Water for clarity.

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4) Footnote (5) has been added to the Recirculation Pump Cooling Water and Drywell Cooler Water Lines. These valves do not meet the requirements of 10CFR50 Appendix J, Section II-H. Additionally, a Probability Risk Analysis (PRA) study has been performed on these lines in response to NRC Information Notice 89-55. Assuming these valves fail to close in a worst-case High Energy Line Break (HELB) scenario, the probability of a radiological release is acceptably low.

5) The maximum operating time for the Recirculation Pump Cooling and Drywell Cooler water return line DC motor isolation valves has been changed from 30 to 60 seconds. This is consistent with the bases for Primary Containment Isolation Valves which states that a closure time of 60 seconds for'rimary Containment Isolation Valves will be adequate to prevent fission product release through lines connected to the Primary Containment.

~Pa e 14 b (New Page)

New page 148b has been added to contain the notes from Table 3.3.4.

~Pa e 14 The bases for Section 3.3.4 and 4.3.4 Primary Containment Isolation Valves have been updated to reference the requirements of the NRC SER dated May 6, 1988 and Niagara Mohawk's clarification letter of July 28, 1988 (NMP1L 0291).

All the changes previously identified above can be categorized into one of the following groups:

1) Administrative changes to clarify the table or to update the table to include isolation valves in lines which were not previously identified in the Technical Specification Isolation Valve tables.

2) Changes made consistent with the requirements of 10CFR50 Appendix J and the NRC SER dated May 6, 1988 regarding Proposed Technical Specifications and Exemption Requests Related to Appendix J.

3) Changes in operating time of certain isolation valves consistent with the original Bases and/or current analyses for operation of systems.

4) Changes requiring closure of containment vent and purge valves on high radiation and reduction of closure time of the containment vent and purge valves to limit releases of radioactivity.

5) Changes made to add or delete valves which define the reactor coolant and primary containment isolation boundary.

6) Changes made to remove footnotes and add check valve position to be consistent with industry practice.

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