ML20045C476

Proposed Tech Specs Re Bench Checking & Disassembling of Safety/Relief Valves,Functionally Testing Ten Percent of Each Snubber Type & Testing Instrument Line Excess Flow Check Valves

ML20045C476
Person / Time
Site:	FitzPatrick
Issue date:	06/16/1993
From:	POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:
Shared Package
ML20045C464	List: ML20045C463 ML20045C467 ML20045C476
References
JPTS-93-008, JPTS-93-8, NUDOCS 9306230170
Download: ML20045C476 (14)

Text

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ATTACHMENT I to JPN-93-041 PROPOSED TECHNICAL SPECIFICATION CHANGES ONE-TIME EXTENSION TO VARIOUS SURVElLLANCE INTERVALS (JPTS-93-008) 4 New York Power Authority JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333 DPR-59 i

9306230170 930616 IS, PDR ADOCK 05000333 P PDR {%

JAFNPP 3.6 (cont'd) 4.6 (cont'd) 1 E. Safety and Safetv/ Relief Valves E. Safety and Safetv/ Relief Valves

1. During reactor power operating conditions and prior to 1. At least one half of all safety / relief valves shall be bench startup from a cold condition, or whenever reactor coolant checked or replaced with bench checked valves once each pressure is greater than atmosphere and temperature operating cycle. The safety / relief valve settings shall be set greater than 212 F, the safety mode of all safety / relief as required in Specification 2.2.B. All valves shall be tested valves shall be operable, except as specified by every two operating cycles.*

l Specification 3.6.E.2. The Automatic Depressurization System valves shall be operable as required by specification 3.5.D.

The current surveillance interval for bench checking safety / relief valves is extended until the end of R11/C12 refueling outage scheduled for January,1995. This is a one-time extension, effective only for this surveillance interval. The next surveillance interval will_begin after the completion of the bench check testing and after the safety / relief valves are declared operable.

Amendment No. H1,MM M, TSQ, TS4, 142a

JAFNPP .

^

3.6 (cont'd) 4.6 (cont'd)

2. 2. At least one safety / relief valve shall be disassembled #

a. From and after the date that the safety valve function of one safety / relief valve is made or found to be inoperable, continued operation is permissible only during the succeeding 30 days unless such valve is made operable sooner.

b. From and after the time that the safety valve function on two safety / relief vt.lves is made or found to be noperabie, continued reactor operation is permissible only during the succeeding 7 days unless such valves are sooner made operable.

3. If Specifications 3.6.E.1 and 3.6.E.2 are not met, the 3. The integrity of the nitrogen system and components which reactor shall be placed in a cold condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. provide manual and ADS actuation of the safety / relief valves shall be demonstrated at least once every 3 months.

4. Low power physics testing and reactor operator training 4. An annual report of safety /refief vave failures and challenges shall be permitted with inoperable components as specified will be sent to the NRC in accordance with Section 6.9.A.2.b.

in item B.2 above, provided that reactor coolant temperature is s 212 F and the reactor vessel is vented or the reactor vessel head is removed.

The current surveillance interval for disassembling and

5. The safety and safety / relief valves are not required to be inspecting at least one safety / relief valve is extended until operable during hydrostatic pressure and leakage testing the end of R11/C12 refueling outage scheduled for January, with reactor coolant temperatures between 212 #F and 1995. This is a one-time extension, effective only for this 300 *F and irradiated fuel in the reactor vessel provided all surveillance interval.' The next surveillance interval will begin control rods are inserted. upon completion of this surveillance.

Amendment No. 49,7Q,130, tS4,190, 143

JAFNPP .

3.6 (cont'd) 4.6 (cont'd)

2. With one or more snubbers inoperable, within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 2. Visual inspection shall verify (1) that there are no visible ~

during normal operation, or within 7 days during Cold indications of damage or impaired OPERABILITY, (2)

Shutdown or Refue!ing mode of operation for systems attachments to the foundation or supporting structure are which are required to be operable ir ' modes, secure, and (3) in those locations where snubber complete Que of the following: movements can be manually induced without disconnecting the snubber, that the snubber has freedom of movement and

a. replace or restore the inoperm.a snubber (s) to is not frozen up. Snubbers which appear inoperable as a operable status or, result of visualinspections may be determined OPERABLE for the purpose of establishing the next visual inspection

b. declare the supported system inoperable and follow interval, providing that (1) the cause of the rejection is clearly the appropriate limiting condition for operation established and remedied for that particular snubber and for statement for that system or, other snubbers that may be generically susceptible; and (2) the affected snubber is functionally tested in the as found

c. perform an engineering evaluation to show the condition and determined OPERABLE per Specifications inoperable snubber is unnecessary to assure 4.E .7 or 4.6.l.8, as applicable. Hydraulic snubbers which cperability of the system or to meet the design have lost sufficient fluid to potentially cause uncovering of criteria of the system, and remove the snubber from the fluid reservoir-to-snubber valve assembly port or the system. bottoming of the fluid reservoir piston with the snubber in the fully extended position shall be functionally tested to determina operability.

3. With one or more snubbers found inoperable, within 72 3. Once each operating cycle,10% of each type of snubbers hours perform a visual inspection of the supported shall be functionally tested for operability, either in place or component (s) associated with the inoperable snubber (s) in a bench test. For each unit and subsequent unit that and document the results. For all modes of operation does not meet the requirements of 4.6.l.7 or 4.6.l.8, an except Cold Shutdown and Refueling, within 14 days additional 10% of that type of snubber shall be functionally complete an engineering evaluation as per Specification tested until no more failures are found, or all units have been 4.6.l.6 to ennure that the inoperable snubber (s) has not tested.*

l adversely affected the supported component (s). For Cold Shutdown or Refueling mode, this evaluation shall be The current surveillance irtterval for functionally testing completed within 30 days. snubbers is extended until the end of R11/C12 refueling outage scheduled for January,1995. This is a one-time extension, effective only for this surveillance interval. The next surveillance inter /al will begin upon completion of this surveillance.

l Amendment No. 20,92,98,148,180, 145c

JAFNPP 3.7 (cont'd) 4.7 (cont'd)

c. Secondary containment capability to maintain a 1/4 in. of water vacuum under calm wind conditions with a filter train flow rate of not more than 6,000 cfm, shall be demonstrated at each refueling outage prior to refueling.

D. Primary Containment Isolation Valves D. Primary Containment isolation Valves

1. Whenever primary containment integrity is required per 1. The primary containment isolation valves surveillance 3.7.A.2, containment isolation valves and all instrument shall be performed as follows:

line excess flow check valves shall be operable, except as specified in 3.7.D.2. The containment vent a. At least once per operating cycle, the operable and purge valves shall be limited to opening angles isolation valves that are power operated and less than or equal to that specified below: automatically initiated shall be tested for simulated automatic initiation and for closure Valve Number Maximum Ooenino Ano;e time.

27AOV-111 40" 27AOV-112 40' b. At least once per operating cycle, the instrument 27AOV-113 40' line excess flow clieck valvos shall be tested for 27AOV-114 50 proper operation.* l 27AOV-115 50 27AOV-116 50' c. At least once per quarter:

27AOV-117 50' 27AOV-118 50 (1.) All normally open power-operated isolation valves (except for the main stream line and Reactor Building Closed Loop Cooling Water System (RBCLCWS) power-operated isolation valves) shall be fully closed and reopened.

The current surveillance interval for testing instrument line excess flow check valves is extended until the end of the R11/C12 refueling outage scheduled for January,1995. This is a one-time extension, effective only for this surveillance interval. The next surveillance interval will begin upon completion of this survei!!ance.

Amendment No. %4. WQ, 185

9 4

p Attachment 11 to JPN-93-041 SAFETY EVALUATION FOR . ,

PROPOSED TECHNICAL SPECIFICATION CHANGES ONE-TIME EXTENSION TO VARIOUS SURVEILLANCE INTERVALS (JPTS-93-008) I page 1 of 9

1. DESCRIPTION OF THE PROPOSED CHANGES The proposed changes to the James A. FitzPatrick Technical Specifications are as follows:

Page 142a. Soecification 0.6.E.1 Add an asterisk to the end of Specification 4.6.E.1.

Add at the lower right hand corner of the page the following footnote:

"The current surveillance interval for bench checking safety / relief valves is extended until the end of R11/C12 refueling outage scheduled for January,  ;

1995. This is a one-time extension, effective only for this surveillance i

interval. The next surveillance interval will begin alter the completion of the bench check testing and after the safety / relief valves are declared-operable." l Pace 143. Soecification 4.6.E.2 ,

Add an asterisk to the end of Specification 4.6.E.2.

Add at the lower right hand corner of the page the following footnote: t "The current surveillance interval for disassembling and inspecting at least one i

safety / relief valve is extended until the end of R11/C12 refueling outage scheduled for January,1995. This is a one-time extension effective only for this surveillance interval. The next surveillance interval will begin upon completion of this surveillance."  ;

Page 145c. Soecification 4.6.l.3 Add an asterisk to the end of Specification 4.6.1.3. ,

Add at the lower right hand corner of the page the following footnote:

"The current surveillance interval f_or functionally testing snubbers is extended until the end of R11/C12 refueling outage scheduled for January,,1995. This is a one-time extension, effective only for this surve!! lance interval. The next surveillance interval will begin upon completion of this surveillance." l

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a q

Attachment 11 to JPN-93-041 SAFETY EVALUATION Page 2 of 9 Pace 185. Soecification 4.7.D.1.b Add an asterisk to the end of Specification 4.7.D.1.b.

Add at the lower right hand corner of the following footnote:

"The current interval for testing instrument line excess flow check valves is extended until the end of R11/C12 refueling outage scheduled for January,1995. This is a one-time extension, effective only for this inspection interval. The next surveillance interval will begin upon completion of this surveillance." 4

11. PURPOSE OF THE PQOPOSED CHANGES The proposed changes to the James A. FitzPatrick Technical Specifications involve a one-time extension to various surveillance intervals as described below. The purpose of these extensions is to provide operational flexibility to accommodate the length of the current operating cycle. Typically, an eighteen month surveillance interval, with the TS 4.0.B.1 provision to extend it by 25% is sufficient to accommodate normal variations in the length of an operating cycle. However, since FitzPatrick had an extended fourteen month outage, the continuity of the normal refueling outage schedule was interrupted. In order for these surveillance requirements to be met using the existing Technical Specifications, a lengthy outage would be required after only eight months of operation following the startup from the last refueling outage. A three week mid-cycle maintenance and surveillance outage is currently scheduled for September,1993. However, there is insufficient time to complete these surveillances during this outage. Granting of this one-time extension will prevent an untimely extended mid-cycle outage of FitzPatrick. The Authority requests that the application for this amendment be processed by August 1,1993 since the first surveillance with the 25% extension will become overdue shortly thereafter.

The safety / relief valves provide pressure relief to prevent a reactor vessel over pressure situation. Technical Specification 4.6.E.1 requires that at least half of all safety / relief valves be bench checked or replaced with benched checked valves once each operating cycle. The intent of Technical Specification 4.6.E.1 is to verify half of the valves each cycle and all of them within two cycles (~3 years). The purpose of this requirement is to verify the valve lift setpoint is within prescribed tolerance for each valve. A commitment in LER 92-016 specifies all safety / relief valves, rather than half as specified in Technical Specifications, be subject to bench ' check testing once each operating cycle. The '

scheduled due dates for the eleven safety / relief valves are between April 3,1993 and July 18,1993. Applying the 25% interval extension extends the due dates of the current interval from August 18,1993 to December 2,1993. This proposed change requests an additional extension until the end of R11/C12 refueling outage scheduled to begin January,1995.

Technical Specification 4.6.E.2 requires that at least one safety / relief valve be disassembled and inspected once per operating cycle. The purpose of this requirement  ;

i

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Attachment Il to JPN-93-041 SAFETY EVALUATION Page 3 of 9 is to verify physical condition and identify any degradation to the valve. The plant has only been operating since January 1993. Therefore any degradation to the valves would be minimal. This surveillance interval expired on April 15,1993. Applying the 25%

interval extension extends the completion date to August 30,1993. This proposed change requests an additional extension until the end of R11/C12 refueling outage scheduled to begin January,1995. If any valve is removed during the fall,1993 outage or the spring 1994 maintenance outage, it will be disassembled and inspected to satisfy the requirements of Technical Specification 4.6.E.2.

1 Snubbers are designed to prevent unrestrained pipe motion under dynamic loads which might occur during an earthquake or a severe transient, while allowing normal motion (eg. thermal growth) during startup and shutdown. These snubbers, mechanical and hydraulic, provide protection to the primary coolant system piping and other safety related components. Technical Specification 4.6.l.3 requires functional testing of 10% of each snubber type. The scheduled due date is May 21,1993. Applying the 25%

interval extension, extends the completion date to October 5,1993. This proposed change requests an additional extension until the end of the R11/C12 refueling outage scheduled to begin in January,1995.

Piping comprising a portion of the Reactor Coolant System, whose failure could result in uncovering the reactor core, are supplied with isolation capability, instrument piping connected to the Reactor Coolant System which leaves the primary containment is dead-ended at instruments located in the Reactor Building. These instrument lines are provided with excess flow check valves that close on forward flow indicative of a break in the line. In the event of an instrument line break, the excess flow check valve on that line closes to prevent leakage of reactor coolant outside the containment.. Technical Specification 4.7.D.1.b requires that all instrument line excess flow check valves be tested once per operating cycle. This surveillance requirement will expire on July 12, 1993. Applying the 25% extension extends the completion date to December 3,1993.

This proposed change requests an addtiional extension until the end of the R11/C12 refueling outage scheduled to begin in January,1995.

111. SAFETY IMPLICATIONS OF THE PROPOSED CHANGES A. Safety / Relief Valves The purpose of testing safety / relief valves every refueling outage is to detect failures and deterioration. The tolerance value specified in Section lli of the ASME Boiler and Pressure Vessel Code and TS section 2.2.B is plus or minus one percent of the specified setpoint pressure. The safety / relief valves have two functions; automatic pressure relief ana self-actuation on high reactor pressure. Pressure relief is a solenoid actuated function (Automatic Depressurization System) in which external instrumentation logic signals the valves to open. In addition, these valves can be operated manually.

The bench checking of all the safety / relief valves and disassembling of one valve can be

Attachment il to JPN-93-041 SRI ETY EVALUATION Page 4 of 9 safely extended until the R11/C12 refueling outage for the following reasons:

In accordance with Technical Section 2.2.1.B and 4.6.E, the safety / relief valves are required to open with 1% of the nominal setpoints. In the past, safety / relief valve setpoint drif t has caused setpoint tolerances to be exceeded. During the scheduled refueling outage on August 27,1988, seven safety / relief valves were tested in accordance with Technical Specification 2.2.1.B. Of the seven valves tested, five had values outside the allowable tolerance. Three of these five valves had setpoint drifts within two percent of the nameplate set pressure. Deviations of this magnitude from the nameplate set pressure are not unusual for this valve design. The other two valves experienced setpoint drifts of 5.5 and 9.0 percent of the nameplate set pressure. During the scheduled refueling outage beginning on January 11,1992, all eleven safety / relief va!ves were tested. Of the eleven tested, eight had values outside the allowable tolerance. Four of the eight valves had  !

setpoints within 2 percent of the nameplate. The remaining four of the eight valves experienced setpoint drifts between 6.0 and 8.6 percent of the nameplate set pressure.

Due to the setpoint drift problem, an analysis to determine the effects of safety / relief valve setpoint drift has been performed (Reference 1). This analysis considered plant operation with two safety / relief valves inoperable and established an upper bound for the remaining nine valves. The analysis showed that continuous operation of the plant would be acceptable with nine safety / relief valves actuating at 1195 psig.

The acceptance criteria for this analysis was 50 psi margin to the ASME code upset reactor vessel pressure limit of 1375 psig during the design basis overpressure event. Additionally, the analysis confirmed that setpoint drift of 9 safety / relief valves to the 1195 psig limit would not adversely affect the following:

-High Pressure Coolant Injection System

-Reactor Core Isolation Cooling System

-Primary Containment Integrity

-Fuel Thermal Limits

-Emergency Core Cooling System / Loss of Coolant Accident performance This analysis bounds the safety / relief valve setpoints identified by testing since only one safety / relief valve exhibited setpoint in excess of 1195 psig during the 1988 i l

refueling outage and only two safety / relief valves exhibited setpoints in excess of 1195 psig during the 1992 refueling outage. Based on this evaluation, plant operation with two safety / relief valves out of service and the remaining nine having a setpoint drift up to the upper limit of 1195 psig, will have no significant safety impact upon plant safety.

The ability of the safety / relief valves to limit any overpressure situation which may occur during abnormal operational transients will not be significantly degraded as a result of this extension. This extension involves no hardware modifications, changes to system operating procedures or affects the ability of the s

Attachment ll to JPN-93-041 SAFETY EVALUATION Page 5 of 9 system to perform its intended safety function.

No significant safety / relief valve degradation is expected during this extension period. Prior to startup from the last outage, pilot assemblies for each valve were replaced with refurbished and recertified assemblies. Pilot disc to seat corrosion induced bonding during normal operation is the most likely cause of safety / relief j valve setpoint drift.

1 i The safety / relief valves were not subject to an operating environment during j the fourteen month time period that FitzPatrick was idle. No degradation is expected to have occurred while they were idle. Corrosion induced bonding takes place during an operating environment. Degradation normally occurs due to the long term effects of moisture, heat, radiation, physical wear and dirt which occurs while the plant is in operation.

This is a one-time extension. This extension applies only to this surveillance

, interval. All intervals after this interval are expected to be the regular Technical Specification specified durations.

During the startup process in January 1993, each safety / relief valve was successfully cycled twice to ensure operability. Successful operation of these valves provide a high degree of confidence that these valves will perform their intended safety function if required.

During the current laservice inspection Interval (ISI), which began in 1985, five safety / relief valves disassembled as a part of this program have shown no serious degradation. The safety / relief valves are believed to be in good physical condition as substantiated by those inspections conducted in accordance with TS 4.6.E.2 which requires that at least one safety / relief valve be disassembled and inspected per refueling outage.

Each safety / relief valve has tall pipe temperature and acoustical detectors. The temperature detector identifies a high temperature situation indicative of safety / relief valve pilot or main disk leakage. The acoustical monitor detect flow noise through the safety / relief valves. These detection systems identify leakage through the valves.

During the current operating cycle, one safety / relief valve (02RV-71K) indicates a high tail pipe temperature reading due to an actual leakage condition. This leak is determined to be a main disc leak and not a pilot valve leak. This type of leakage does not affect the actuation setpoint for the safety / relief valve.

B. Snubbers  !

Snubbers are designed to prevent unrestrained pipe motion under dynamic loads which might occur during an earthquake or severe transient, while allowing normal thermal motion during startup and shutdown. Snubbers provide protection to the primary coolant system piping and other safety related components. To ensure that snubbers properly perform their intended t

s .

Attachment 11 to JPN-93-041 SAFETY EVALUATION Page 6 of 9 function, they are subjected to periodic functional and visual inspections. Functional tests verify that snubbers can operate within specific parameter limits. Visual inspections observe the condition of installed snubbers to verify that none are damaged or degraded. Visual examinations complement the functional testing program and provide additional assurance of snubber operability. To provide assurance of snubber functional reliability, a representative sample of the installed snubbers is functionally tested once each operating cycle. Selection of a representative sample of 10% of each group of safety related snubbers provides a confidence level within acceptable limits that these supports will be in an operable condition.

Failures of these sample snubbers require functional testing of additional units.

The 10% functional testing of each snubber group can be safely extended until the end of R11/C12 refueling outage for the following reasons:

Snubber protection is only required during low probability events. The likelihood of a seismic event or severe transient requiring snubbers to perform their safety function during the requested interval extension is remote.

This is a one-time extension. This extension applies only to this surveillance interval. All intervals after this interval are expected to be within the regular Technical Specification specified durations.

During the last refueling outage,10% of all mechanical and hydraulic snubbers were functionally tested. No mechanical snubber failures were identified. Two hydraulic snubber failures were identified which resulted in a total of 69 snubbers being tested. In total, there are approximately 250 hydraulic snubbers. Due to the low number of observed failures during the last functional test, confidence of snubber functional reliability is assured.

A significant amount of resources would be required during the upcoming mid-cycle maintenance and surveillance outage to meet these requirements.

Since this outage is of short duration, there would be insufficient time complete these requirements.

All visual snubber inspections will be performed as scheduled. Visual inspections will detect any obvious damage or degradation to the snubbers. This will provide additional assurance of snubber operability.

C. Instrument line excess flow check valves Instrument line excess flow check valves are tested once each operating cycle to verify proper operation. The Primary Containment is penetrated by numerous small diameter instrument lines connected to the Reactor Coolant System. Each instrument line contains a 0.25 inch '

restricting orifice inside the primary containment and an excess flow check valve outside the Primary Containment. In addition to the excess flow check valves, each instrument line is ,

provided with a manual isolation valve upstream of the excess flow check valves. This valve l can be manually open or closed when needed. In the event of a broken instrument line, the

Attachment 11 to JPN-93-041 SAFETY EVALUATION Page 7 of 9 excess flow check valves close to limit leakage of reactor coolant into the Reactor Building thereby limiting contamination.

The operability testing of all instrument line excess flow check valves can be safely extended until the R11/C12 refueling outage for the following reasons:

No significant excess flow check valve degradation is expected during this extension period. These components have a good history and few failures. An average of one percent failure per operating cycle is identified for these valves. Due to the low number of failures during previous inspections, confidence of excess flow check valve reliability is assured.

Degradation to the excess flow valve seats normally occurs during normal power operation. Since FitzPatrick was idle for fourteen months, the check valves did not operate under service conditions. No degradation is expected to have occurred while they were idle during the outage.

This is a one-time extension. This extension applies only to this surveillance interval. All intervals after this interval are expected to be the regular Technical Specification specified duration.

IV. EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATION Operation of the FitzPatrick plant in accordance with the proposed amendment would not involve a significant hazards consideration as defined in 10 CFR 50.92, since it would not:

1. involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed change revises the testing frequency of safety / relief valves for one surveillance interval. This change will not require a modification to any plant structures, systems, or components. This is a change only to the safety / relief valve testing frequency. The actual surveillance test effectiveness and operability as determined by this testing will remain the same. The potential small change in reliability will not initiate an accident and will not effect the consequences of an accident since the safety / relief valves are still available to perform their intended safety function.

The one-time surveillance interval extension to the functional testing frequency of snubbers involves no hardware changes, no changes to the operation of the snubbers nor does it change the ability of the snubbers to perform their intended functions. Increasing the snubber functional test interval produces a small increase in the probability that an inoperable snubber would not be detected. However, this small risk does not involvo a significant increase in the probability of an accident previously evaluated.

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Attachment ll to JPN-93-041 SAFETY EVALUATION >

Page 8 of 9 The proposed one-time surveillance interval extension of instrument line excess flow check valve testing involves no hardware modifications, changes to system operating procedures or affects the ability of any system to perform its intended function. The probability of excess flow check valve leakage is not increased significantly and the ability of plant personnel and equipment to respond to an accident is not affected. ,

2. create the possibility of a new or different kind of accident from those previously evaluated.

The proposed change to the safety / relief valve surveillance test interval does not change design, operation or the testing process. The nature of this change precludes the possibility of a new or different kind of accident.

The proposed change to the snubber functional test interval involves no hardware -

changes, no changes to the operation of the snubbers nor change the ability of the snubbers to perform their intended functions. Performing the testing on an extended schedule cannot initiate any type of accident.

The proposed change to the excess flow check valve surveillance test interval does i not change the design, operation, or the testing process. The change to the testing interval does not affect any condition that could result in a new or different type of accident.

3. involve a significant reduction in the margin of safety.

Safety / relief valves are required to open within 1% of the nominal setpoints as prescribed in the Technical Specifications. The purpose of bench checking all safety / relief valves is to verify these setpoints. In the past, safety / relief valve setpoint drift has caused the setpoint tolerances to be exceeded. As part of the analysis in Reference 1, continuous operation of the plant would be acceptable with the safety / relief valves actuating at 1195 psig. The value of the upper limit of 1195 psig, determined from the overpressure protection analysis, satisfies all safety concerns and does not reduce the ability of the system to pedorm its intended safety function.

The proposed one-time change to the snubber functional test interval does not change the operation of snubbers nor change the ability of the snubbers to perform their intended function. This one-time interval extension does not change the level of confidence in snubber operability developed from visual inspections which will be performed as scheduled.

The proposed one-time siension to the excess flow check valve functional test interval does not reduce the ability of these valves to perform their intended safety function. The small amount of degradation the valves may undergo during this extension period is insignificant and will not significantly reduce the ability of the valves to close to limit leakage of reactor coolant.

1 Attachment 11 to JPN-93-041 SAFETY EVALUATION Page 9 of 9 V. IMPLEMENTATION OF THE PROPOSED CHANGES Implementation of the proposed changes will not adversely affect the ALARA or Fire Protection Program at the FitzPatrick plant, nor will the changes impact the environment.

These changes will not result in any new releases to the environment since there are no hardware, structural, or operational changes. For these same ceasons, the changes pose no radiological or fire hazards.

VI. CONCLUSION The changes, as proposed, do not constitute an unreviewed safety question as defined in 10 CFR 50.59. That is, they:

1. will not change the probability nor the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the Safety Analysis Report; 2 will not increase the possibility of an accident or malfunction of a type different from any previously evaluated in the Safety Analysis Report; and

3. will not reduce the margin of safety as defined in the basis for any technical specification.

The changes involve no significant hazards consideration, as defined in 10 CFR 50.92.

Vll. REFERENCES

1. NYPA letter, R. E. Beedle to U.S. NRC, dated May 5,1993, (JPN-93-033),

regarding updated Safety Relief Valve Performance Requirements.

2. James A. FitzPatrick Nuclear Power Plant Updated Final Safety Analysis Report, Sections 12.5. 4.4 and 5.2.

3. James A. FitzPatrick Nuclear Power Plant Safety Evaluation Report (SER), dated November 20,1972, and Supplements.