ML20063A989

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Proposed Tech Spec,Changing Reactor Operation to 30 Days in Event That Safety/Relief Valve Becomes Inoperable & Limiting Conditions for Operation of HPCI Sys
ML20063A989
Person / Time
Site: FitzPatrick Constellation icon.png
Issue date: 08/12/1982
From:
POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:
Shared Package
ML20063A987 List:
References
NUDOCS 8208250144
Download: ML20063A989 (9)
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~_/ ATTACHMENT I PROPOSED TECHNICAL SPECIFICATION CHANGES CONCERNING ,

SAFETY / RELIEF, VALVE OPERABILITY

.y POUER AUTHORITY OF THE STATE OF NEW YORK JAMES A. FITZPATRICK NUCLEAR POWER PLANT DOCKET NO. 50-333 AUGUST 12, 1982 i;

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NOTES FOR TABLE 3.2-6 (CONTINUED)

2. In the event that all indications of this parameter is disabled and such indication cannot be restored in six (6) hours, an orderly shutdown shall be initiated and the reactor stiall be in

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a flot shutdown condition in six (6) hours and a Cold Shutdown condition in the following eighteen (18) hours. .

3. Three (3) indicators from level instrument channel A, B, & C. Channel A or B are utilir.ed for feedwater control, reactor water high and low level alarms, recircula tion pump runback. Iligh level trip of main turbine and feedwater pump turbine utilizes channel A, B, & C.
4. One (1) recorder utilized the same level instrument channel as selected for feedwater control.
5. Three (3) indicators from reactor pressure instrument channel A, n, & C. Channel A or n are utilized for feedwater control and reactor pressure high alarm.
6. One (1) recorder. U.tilizes the same reactor pressure instrument channel as selected for feedwater control.
7. The position of each of the 137 control rods is monitored by the Rod Position Information Systen. For control rods in which the position is unknown, refer to Paragraph 3.3.A.
8. Neutron monitoring operability requirements are specified by Table 3.1-1 and Paragraph 3.3.n.4.

9.- A mis! mum of 3 IRM or,2 APRM channels respectively must be operable (or tripped) in each safety system.

10. Each Safety Relief Valve is equipped with two acoustical detectors of which one is in service and a backup thermocouple detector. In the event that a thermocouple is inoperable SRV l p erformance shall be monitored daily with the associated acoustical detector. I
11. From and after the date that none of the acoustical detectors is operable but the thermocouple is operable, continued operation'is permissibic until the next outage in which a primary -

' containment. entry is made. Both acoustical detectors shall be made operable prior to restart.

12. In the event that both primary and secondary indications of this parameter for any one valve' are disabled and neither indication can be restored in forty-eight (48) hours, an orderly
shutdown shall be initiated and the reactor shall be in a llot Shutdown condition in twelve (12) hours and in a Cold Shutdown within the next twenty-four (24) hours.
13. From and after the date that the minimum number of operable instrument channels is one less than the minimum number specified for each parameter, continued operation is permissable during the suc'ceeding 7 days unless the minimum number specified is made' operable sooner.

Amendment No. 40 , y1, / 76 c

l 3.6 (cont'd) JAFNPP 4.6 (cont'd)

E. Safety and Safdty/ Relief ~ Valves E. Safety and Safety / Relief . Valves-f

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1. During reactor power operating 1. At'least one half of all conditions and prior to startup safety / relief valves shall from a cold condition, or whenever be' bench checked or re-reactor coolant pressure is placed with bench checked greater than atmosphere and valves once each operating 0 cycle. The safety / relief temperature greater than,212 F, the safety mode of all safety / relief valve settings shall be set valves shall be operable, except as required in Specification as specified by Specification 2.2.B. All valves shall be 3.6.E.2. The Automatic Depressurization tested every two operating -

System valves shall be operable as required cycles.

by Specification 3.5.D.

2. Reactor operation may continue with one safety / relief valve inoperable.

From and after the date that two safety / relief valves are made or found inoperable, continued reactor operation is permissible only during the succeeding 7 days,unless one valve is made operable. -

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( This page is effective until Amendment No. 142b the October 1982 outage.

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JAFNPP 4.6 (cont'd) 3.6:(cont'd)'

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3. L If Specification ' 3. 6. E. l amd 3. 6. E. 2 2. 'At least one safety / relief ,.

are not met the reactor shall be valve shall be disassembled placed in a cold condition within and inspected once/ operating cycle.

24 hr.

4. Low power physics testing and reactor 3. Deleted operator training shall be permitted The integrity of the nitrogen with inoperable. components as specified 4.

system and components which-in 3.6.E.2, and provided that reactor coolant. temperature.isa(_2120F and provide manual and' ADS actuation of the safety /reli~ef the-reactor vessel is v'ented or the valves shall be demonstrated vessel head is removed. at least once every 3 months.

Amendment No. 143a This page is effective until the October 1982 outage.

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bAFNPP 3.6 (cont'd) 5.. If, for a period of longer than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the temperature of any safety / relief discharge -

pipe is more than 400F above its steady state valu e, or the acoustical monitor reading of any safety / relief-valve discharge pipe is more than 3 times greater than its steady state value, the following actioas shall be taken:

a. a report shall be issued in accordance with 6.9.A.4.1 which addresses the actions that have been taken or a schedule of actions to be taken.
b. an engineering evaluation shall be per- .

formed justifying continued operation for the corresponding increase in tem-perature or acoustical monitor reading.

c. the_affected safety / relief valve shall be removed. at the next cold shutdown of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or more, tested in the as-found condition, and recalibrated as necessary prior to reinstallation.
d. NRC approval of the engineering evaluation specified in 3.6.E.6.b above ,

shall be obtained prior to continuing power operation for more than 90 days after the -

initial discovery of the 40 F increase -

in temperature or the factor of 3 increase in acoustical monitor reading.

The steady state values of temperature and acoustical monitor readings shall-be as measured after 5 days of steady state power operation.

Amendment No. 143b This page is effective until the October 1982 outage.

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2 ATTACHMENT II SAFETY EVALUATION CONCERNING .

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SAFETY / RELIEF VALVE-OPERABILITY I

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POUER AUTHORITY OF THE_ STATE OF NEW YORK JAMES A. . FITZPATRICK NUCLEAR POWER PLANT ,

DOCKET NO. 50-333 L AUGUST 12,.1982  ;

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Ssetion-I - Dencription of the Amandment This proposed amendment to Appendix A'of the James A. FitzPatrick Nuclear Power Plant Facility Operating License changes the safety / relief valve (SRV) operability requirements contained in

' Specifications 3.5.C, 3.5.D and 3.6.E. Currently, when an SRV becomes s '

or is found to be inoperable, continued reactor operation is permissible for only thirty days unless the valve can be repaired and made operable. If more than one valve becomes inoperable, continued operation is permissible for only seven days. If the affected valve (s) is part of the Automatic Depressurization System (ADS), continued reactor operation is permissible for: only 7 days with one inoperable; and, only 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> if more than one is inoperable. In addition, if one or more ADS valves are inoperable, HPCI System operability is also required.

The proposed changes would affect the limiting conditions for operation only until the October 1982 refueling outage. With one non-ADS SRV inoperable, continued operation would be permissible for an indefinite period with no power level or other restrictions. With two or more SRV's inoperable, the existing limiting conditions for operation are still applicable.

Additional changes delete the requirement for continuous monitoring of the integrity of safety / relief valve bellows; change specification numbers and references as required; and, correct typographical errors.

The changes in this proposed amendment are contained in new pages 76c, 142b, 143a, and 143b.

Section II - Purpose of the Amendment The proposed changes will delete conservative limiting conditions for operation which are not necessary to assure the safe operation of the plant or to protect the health and safety of the public.

The proposed amendment requires only the operability of sufficient SRV's to assure plant safety.

Section III - Impact of the Amendment The proposed amendment provides revised limiting conditions for operation when an SRV is inoperable for any reason. It is also intended to address concerns of the NRC staff that SRV leakage, as indicated by elevated SRV tailpipe temperatures, renders the affected valve inoperable. These concerns have been expressed as a result of a recent increase in one SRV tailpipe temperature.

SRV leakage is not monitored directly but is indicated by increased

. tailpipe thermocouple and acoustical monitor readings. The possibility of proposing SRV tailpipe temperatures at which the valve would be declared inoperable, because of leakage, has been investigated.

General Electric (the NSSS vendor) and Target Rock (the SRV vendor) have both stated that they have not been able to identify or develop a definitive correlation between leakage and tailpipe temperature, by calculation'or by using empirical data. However, "as found" testing conducted by Target Rock for one utility showed that a 400 to 600F increase in tailpipe temperature indicates a leakage rate from neglible up to 200 pounds per hour. Further testing by Target Rock has

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indicated that a leakage rate of 200 pounds per hour sh'ould not affect SRV cetpoint or response time (Reference C). Therefore, the existing elevated tailpipe temperature does not conclusively indicate that the valve is inoperable. However, a new specification, 3.6.E.5, is also proposed to specify actions which will be taken if significant increases occur in safety / relief valve discharge pipe temperatures or acoustical mon _itor readings.

There are several possible physical mechanisms by which leakage may affect SRV per formance. The possible physical mechanisms allow one to postulate either an increase or decrease in the setpoint due to leakage. Therefore, both possibilities have been reviewed for their impact on plant safety.

If the setpoint decreases, there is a potential for increased loads on the torus during a transient which results in multiple SRV actuation. This issue was reviewed in Reference D, which granted an amendment to the Technical Specifications to allow changing the setpoint of one SRV from 1140 psig to 1090 psig. This decrease in setpoint.was found to be acceptable based on the Mark I containment improvements completed to date.

The effect of an SRV spontaneously opening, whether due to leakage or not, has been considered in the design basis of the plant. This design basis event was reviewed and approved in References A and B.

If the setpoint of a valve were to increase, the worst case would involve failure of the valve to open at any pressure, or in other words, total inoperability. This has been analyzed and the results reported in Reference E. The analyses as reported in this reference show that the FitzPatrick plant may be safely operated with any one of the SRV's totally inoperable, whether it is part of the ADS or not.

The effects of valve inoperability on the torus have been reviewed and found acceptable.

Technical Sepcification Basis 3.5.D also states that only six of '

the seven ADS valves are required to operate and that loss of one of the valves does not significantly affect the pressure relieving capability. A similiar conclusion was reached in Amendment No. 43 to '

the Facility Operating License (Reference F).

Section IV - Conclusion The incorporation of these modifications: a) 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; b) will not increase the possibility of an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report; and c) will not reduce the margin of safety as defined in the basis for any Technical Specification; and d) does not constitute an unreviewed safety question.

Section V - References A. JAF FSAR B. JAF SER ,

C. Target Rock Report No. 1834, dated 12/2/76, entitled "Model 7567F l Design Verification Test Summary".

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.m D. Amendment No. 54 to'Facil'ity Operating License No, DPR-59 dated

' April 13, 1981. .

E ;- Letter J.M. Morris-(GE) to R.A. Burns (PASNY) dated August 11, 1982 transmitting FitzPatrick SRV Out.of Service Analysis.

F. Amendment No. 43 to the Facility Operating License dated November 22,-1978.

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