Text

Safety Evaluation Supporting Amend 62 to License DPR-59
	ML20040G310
Person / Time
Site:	FitzPatrick
Issue date:	01/29/1982
From:	; Office of Nuclear Reactor Regulation
To:	;
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ML20040G309	List: ML20040G308; ML20040G310; ML20040G313;
References
	NUDOCS 8202120149
	Download: ML20040G310 (2)
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NUCLEAR REGULATORY COMMISSION O

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 62 TO FACILITY OPERATING LICENSE N0. DPR-59 POWER AUTHORITY OF THE STATE OF NEW YORK JAMES A. FITZPATRICK NUCLEAR POWER PLANT DOCKET NO. 50-333 1.0 Introduction

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As a result of events invo1ving common cause failures of scram discharge volume (SDV) limit switches and SDV drain valve operability, the NRC staff issued IE Bulletin 80-14 on June 12, 1980.

In addition, the staff sent a letter dated July 7,1980 to all operating BWR licensees requesting that they propose Technical Specification changes to provide surveillance requirements for SDV vent and drain valves and LC0/ surveillance requirements on SDV limit switches.

Model Technical Specifications were enclosed with this letter to provide guidance.to licensees for preparation of the requested submittals.

By letter -dated January 6,1981 the Power. Authority of the State of New York (licensee) requested changes to the Technical Specifications for the James A. FitzPatrick Nuclear Power Plant relating to SDV.

2.0 Evaluation The enclosed report (TER-C5506-74) was prepared by the Franklin Research Center (FRC) as part of a technical assistance contract program. This FRC report provides the technical evaluation of the compliance of t*c licensee's submittal with NRC provided criteria.

FRC has concluded that the licensee's response does not meet the explicit requirements of paragraph 3.3-6 and Table 3.3.6-1 of the NRC staff's Model Technical Specifications (TSs). However, the FRC report concludes that technical bases are defined on p. 50 of the staff's " Generic Safety Evaluation Report BWR Scram Discharge System," December 1,1980 for this departure from the explicit requirements of the Model Technical Specifications.

We conclude that these technical bases justify a deviation from the explicit requirements of the Model Technical Specifications.

FRC has concluded that the licensee's proposed Technical Specifications

' revisions meet our criteria without the need for further revision.

Based upon our review of the contractor's report of his evaluations and discussions with the reviewer, we conclude that the licensee's proposed Technical Specifications satisfy our requirements for surveillance of SDV vent and drain valves and for LCOs and surveillance requirements for SDV limit switches.

Consequently, we find the licensee's proposed Technical Specifications acceptable.

8202120149 820129 PDR ADOCK 05000333 P

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3.0 Environmental Consideration We have determined that the amendment does not authorize a change in effluent types or total amounts nor' an increase in power level and, will not result-in any significant ~ environmental impact.

Having made this determination, we have further concluded that the amendment involves an action which is insignifica.nt from the standpoint of environmental impact' and, pursuant to 10 CFR 551.5(d)(4), that an impact statement or negative declaration and environmental impact appraisal need not be prepared in connection with the issuance of this amendment.

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4.0 ~ Conclusion

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We have concluded, based on the considerations discussed above, that:

(1) because the amendment does not involve a significant increase in the probability or consequences of accidents previously considered and does not involve a significant decrease in a safety margin; the amendment does not involve a significant hazards consideratior., (2) there is reasonable assurance that the health and safety.of the public will not beendangeredbyoperation.intheproposedmanner,and(3)such activities will be conducted'in compli.ance with the Commission's regulations and the issuance of this amendment will not be inimical to.

the common defense and security or to the health and safety of the public.

Datedi January 29, 1982 Author:

Phillip J. Polk Kenneth T. Eccleston

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g TECHNICAL EVALUATION REPORT BWR SCRAM DISCHARGE VOLUME LONG-TERM MODIFICATIONS POWER AUTHORITY OF THE STATE OF NEW YORK JAMES A. FITZPATRICK NUCLEAR POWER PLANT NRC DOCKET NO. 50-333 FRC PROJECT C5506 NRC TAC NO. 42220 FRCASSIGNMENT 2 NRC CONTRACT NO. NRC43-81 130 FRC TASK 74 Preparedby Franklin Research Center Autnor:

E. Mucha l

The Parkway at Twentieth Street Philadelphia, PA 19103 FRC Group Leader:

E. Mucha Prepared for Nuclear Regulatory Commission Washington, D.C. 20555 Lead NRC Engineer:

K. Eccleston December 3, 1981 This report was prepared as an account of work sponsored by an agency of the United States Govemment. Neither the United States Government nor any agency thereof, or any of their employees, makes any warranty, ex-pressed or implied, or assumes any legal liability or responsibility for any third party's use, or the results of such use, of any information, apparatus.

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product or process disclosed in this report, or represents that its use by such third party would not infringe privately owned rights.

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Section Title Page

SUMMARY

1-l' INTRODUCTION 2

1.1 Purpose of the Technical. Evaluation 2

2 1.2 Generic Issue Background

-1.3 Plant-Specif'ic Background.

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REVIEW CRITERIA.

6 2.1 Surveillance Requirements for SDV Drain and vent valves' 6

2.2 ICO/ Surveillance Requirements for Reactor Protection System SDV Limit Switches 7

2.3 LCO/ Surveillance Requirements for Control Rod Withdrawal Block SDV Limit Switches 9

3 METHOD OF EVALUATION 12 4

TECHNICAL EVALUATION 13 4.1 Surveillance Requirements for SDV Drain 13 and vent valves 4.2 ICO/ Surveillance Requirements for Reactor Protection System SDV Limit Switches 14 4.3 ICO/ Surveillance Requirements for Control Rod Withdrawal Block SDV Limit Switches 18 5

CONCLUSIONS.

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REFERENCES'.

24 APPENDIX A - NRC STAFF'S MODEL TECHNICAL SPECIFICATIONS APPENDIX B - POWE2 AUTHORITY OF 'IHE STATE OF NEW YORK LETTER 0F JANUARY 6, 1981 AND SUBMITTAL WITH PROPOSED TECHNICAL SPECIFICATIONS CHANGES FOR FITZPATRICK NUCLEAR POWER PLANT l

i lii NJ Franklin Research Center A Dhaean of The Franseninmunas

TER-C5506-74

SUMMARY

This technical evaluation report reviews and evaluates Phase 1 proposed changes in the FitzPatrick Nuclear Power Plant Technical Specifications for scram discharge volume (SW) long-term modifications'regarding surveillance requirements for SW vent and drain valves and the limiting condition for operation (LCO)/ surveillance requirements for reactor protection system and control rod withdrawal block SW limit switches. Conclusions were based on the degree of compliance of the Licensee's submittal with criteria from the Nuclear Regulatory Commission (NRC) staff's Model Technical Specifications.

The revised page 89 of the FitzPatrick Technical Specifications, and the Licensee's agreement to add to the proposed specifications changes a require-ment to cycle each valve a minimum of one full cycle at least quarterly, comply with the NBC staff's Model Technical Specifications, paragraphs 4.1.3.1.la and 4.1.3.1.1b.

Proposed revisions of pages 43, 44, 45, 45a, 46, 47, 81, 99, 89a, and 96 and unrevised pages 41a and 103 meet the remaining surveillance-requirements. Table 5-1 on pages 22 and 23 of this report summarizes the 1

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TER-C5506-74

1. INT 30D0CTION j

1.1 PURPOSE OF THE TECHNICAL EVALUATION he purpose of this technical evaluation report (TER) is to review and evaluate the proposed changes in the Technical Specifications of the FitzPatrick Nuclear Power Plant boiling water reactor (BWR) in regard to "BWR Scram Discharge Volume Long Term Modification," specifically o surveillance requirements for scram discharge volume (SDV) vent and drain valves o limiting condition for operation (LCO)/ surveillance requirements for the reactor protection system SDV limit switches o IcO/ surveillance requirements for the control rod withdrawal block SDV limit switches.

'Ihe evaluation used criteria proposed by the NRC staff in Model Technical Specifications (see Appendix A of this report). This effort is directed toward the NRC's objective of increasing the reliability of installed BWR scram dis-charge volume systems, the need for which was made apparent by, events described below.

1.2 GENERIC ISSUE BACKGROUND On June 13, 1979, while the reactor at Hatch Unit 1 was in the refuel mode, two SUV high level switches had been modified, tested, and found inoper-able. We remaining switches were operable.

Inspection of each inoperable level switch revealed a bent float rod binding against the side of the float chamber.

On October 19, 1979, Brunswick Unit I reported that water hammer due to slow closure of the SUV drain valve during a reactor scram damaged several pipe supports on the SUV drain line. Drain valve closure time was approximately 5 minutes because of a faulty solenoid controlling the air supply to the valve.

l Af ter repair, to avoid probable damage from a scram, the unit was started with the SDV vent and drain valves closed except for periodic draining. During this mode of operation, the reactor scrammed due to a high water level in the i

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TER-C5506-74s SDV system without prior actuation of either the high level alarm or rod block switch. Inspection revealed that the float ball on the rod block switch was bent,. making the switches inoperable. The water hammer was reported to be the l

cause of these level switch failures.

As a result of these events involving common-cause failures of SDV limit

'switches and SDV drain valve operability, the NRC ' issued IE Bulletin 80-14,

" Degradation of BWR Scram Discharge Volume Capability," on June ^12, 1980 (1].

In addition,. to strengthen the provisions of this bulletin and to ensure that.

the scram system would continue to work during reactor operation,' the NPC sent

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I a letter dated July 7, 1980 (2) to all operating BWR licensees requesting that they propose Technical Specifications changes to provide surveillance-re':1uire-ments for reactor protection system and control rod block SDV limit switches.

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The letter also contained the NRC staff's Model Technical Specifications to be used as a guide by licensees in preparing their submittals.

Meanwhile, during a routine shutdown of the Browns Ferry Unit 3 reactor on June 28,1980, 76 of 185 control rods failed to insert fully. Full irner-l tion required two additional manual scrams and an automatic scrant for a total l

elapsed time of approximately 15 minutes between the first scramsiniti.ntion and the complete insertion of all the rods. On July 3,1980, in response to both this event and the previous events at Hatch Unit 1 and Brunswick' Unit 1, the NRC issued (in addition to the earlier IE Bulletin 80-14) IE Bulletin 80-17 (3] followed by five supplements [4, 5, 6, 7, 8).

These initiated l

short-term and long-term programs described in " Generic Safety Evaluation Report BWR Scram Discharge System," NRC sta.'f, December 1, 1980 (9) and " Staff Report and Evaluation of Supplement 4 to IE Bulletin 80-17 (Continuous

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Monitoring Systems)" (10}.

N Analysis and evaluation of the Browns Ferry Unit 3 and other SDV system y

events convinced the NRC staff that SDV systems in all BWRs should be modified to assure long-term SUV reliability. Improvements were needed in three major areas: SDV-IV hydraulic coupling, level instrumentation, and system isolation.

To achieve these objectives, an Office of Nuclear Reactor Regulation (NRR) t'a sk force and a subgroup of the BWR Owners Group developed Revised Scram Discharge 3

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System Design and Safety Criteria.for use in establishing acceptable SW systems dodifications (9]. Also, an NRC letter dated October 1,1980 requested all operating BWR -licensees to reevaluate installed SW systems and modify them as"necessary to comply with the revised criteria.

<w In'Ife'feren'ce 9, the SW-IV hydraulic coupling at the Big Rock Point,

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Brunswickj1 & 2, Duane Arnold, and Hatch.1 & 2 BWRs was judged acceptable.

The remaining BWas will require modification to mest the revised SW-IV i

hydraulic coupling criteria, and all operating BWRs may require modification

.3 to meet' the revised, instrumentation and isolation criteria. The changes in Technical S'pecifications associated with this effort will be carried out in

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two phases:

y Phase 1. Improvements in surveillance for vent and drain i valves and instrument volume level switches.

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' Phas'e 2 - Technical Specifications improvements required as a,

ic result of long-term modifir

' - s made to comply with revised design and per criteria.

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This TER is a\\ review and evaluation of Technical Specifications, changes s

proposed for'Pr.ase 1.

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1.3 PLANT-SPFAIFIC., EACKGROUND f

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The July 7,1980 NBC letter [2] not only requested all BWR licensees to amend etheir -facilieles' Technical Specifications with respect to control rod drive 'SDV capabilit'i, but enclosed the NBC staff's proposed Model Technical Specifil:hti~ ens (see Appendix A of this TER) as a guide for the lice:Isees in

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preparing \\th'e ' requested submittals and as a source of criteria for an FBC technical evaluation of the submittals. In this TER, FRC has reviewed and evaluated Technical Specifications changes for the FitzPatrick Nuclear Power Plant proposedtin a January 6,1981 letter (see Appendix B) by the Licensee, the Power Authcrity of the State of New York (PASNY), in regard to "BWR Scram Discharge Volume (SW) Long-Term Modifications" and, specifically, the sur-veillance requirements for SW vent and drain valves and the limiting condition fqr operation (Iro)/ surveillance requirements for the reactor protection system x,

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and control rod withdrawal. block SDV limit switches. FRC assessed the adequacy i

with which the PASNY infor3ation documented compliance of the proposed Techni-cal Specifications changes with the NRC staff's Model Technical Specifications.

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2. REVIEW CRITERIA The criteria established by the NRC staff's Model Technical Specifica-tions involving surveillance requirements of the main SDV components and instrumentation cover three areas of concerns o surv illance requirements for SDV vent and drain valves o LCO/ surveillance requirements for reactor protection system SDV limit switches o LCO/ surveillance requirements for control rod block SDV limit switches.

2.1 SURVEILLANCE REQUIREMENTS FOR SDV DRAIN AND VENT VALVES The surveillance criteria of the NRC staff's Model Technical Specifica-tions for SDV drain valves are:

"4.1.3.1.1 - The scram discharge volume drain and vent valves shall be demonstrated OPERABLE bys a.

Verifying each valve to be open* at least once per 31 days and 7

b.

Cycling each valve at least one complete cycle of full travel at least once per 92 days.

These valves may be closed intermittently for testing under 1dministrative controls."

The Model Technical Specifications require testing the drain and vent valves, cbecking at least once every 31 days dbat each valve is fully open during normal operation, and cycling each valve at least one complete cycle of full travel under administrative controls at least once per 92 days.

l Pull opening of each valve during normal operation indicates that there is no degradation in the control air system and its components that control the air pressure to the pneumatic actuators of the drain and vent valves.

Cycling each valve checks whether the valve opens fully and whether its movement is smooth, jerky, or oscillatory.

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o TER-C5506-74 During normal operation, the drain and vent valves stay in the open position for very long periods. A silt of particulates such as metal chips and flakes, various fibers, lint, sand, and weld slag from the water or air may accumulate at moving parts of the valves and temporarily " freeze" them. A strong breakout force may be needed to overcome this temporary freeze, producing a violent jerk which may induce a severe water hatuner if it occurs during a scram or a scram resetting. Periodic cycling of the drain and vent valves is the best method to clear tne effects of particulate silting, thus promoting smooth opening and closing and more reliable valve operation. Also, in case of improper valve operation,-cycling can indicate whether excessive pressure transients may be generated during and after a reactor scram which might damage the SW piping system and cause a loss of system integrity or function.

2.2 Iro/SUWEILLANCE REQUIREMENTS FOR REACTOR PROTECTION SYSTEM SW LIMIT SWITCHES The paragraphs of the NIC staff's Model Technical Specifications pertinent to 140/ surveillance requirements for reactor protection' system SW limit switches are:

"3.3.1 - As a minimum, the reactor protection system instrumentation channels shown in Table 3.3.1-1 shall be OPERABLE with the REA0TCR PR0rECTION SYSTEM RESPONSE TIME as shown in Table 3.3.1-2.

Table 3.3.1-1.

Reactor Protection System Instrumentation Applicable Minimum Operable Functional Operational Channels Per Trip Unit Conditions System (a)

Action 8.

Scram Discharge Volume Water Level-nigh 1,2,5 (h) 2 4

Table 3.3.1-2.

Reactor Protection System Response Times Functional Response Time Unit (Seconds) 8.

Scram Discharge Volume Water Level-High NA" ~

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TER-C5 50 6-74 "4.3.1.1 - Each reactor protection system instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST, and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.1.1-1.

l Table 4.3.1.1-1.

Reactor Protection Syste:a Instrumentation Surveillance Requirements Operational conditions Channel in Which l

Functional Channel Functional Channel Surveillance Unit Check Test Calibration Required 6.

Scram Discharge Volume Water Level-High NA M

R 1,2,5 Notation (a) A channel may be placed in an inoperable status up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for required surveillance without placing the trip system in the tripped condition provided at least one OPERABLE channel in the same trip system is monitoring that parameter.

(h) With any control rod withdrawn. Not appl'icable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.

Action 4: In OPERATIONAL CONDITION 1 or 2, be in at least HCff SHITfDOWN within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

In OPERATIONAL CONDITION 5, suspend all operations involving CORE ALTERATIONS

and fully insert all insertable control rods within one hour.

Except movement of IRM, SRM or special movable detectors, or replacement of LPRM strings provided SRM instrumentation is OPERABLE per Specification 3.9.2."

Paragraph 3.3.1 and Table 3.3.1-1 of the Model Technical Specifications require the functional unit of SDV water level-high to have at least 2 operable channels containing 2 limit switches per trip system, a total of 4 operable channels containing 4 limit switches per two trip systems for the reactor protection system which automatically initiates a scram. The technical objective of these requirements is to provide 1-out-of-2-taken-twice logic for nklin Research Center A Dammen of The Fm insumme

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a TER-C5506-74 the reactor protection system. The response time of the reactor protection i

system for the functional unit of SW water level-high should be measured and kept available (it is not given in Table 3.3.1-2).

Paragraph 4.3.1.1 and Table 4.3.1.1-1 give reactor protection system instrumentation surveillance requirements for the functional unit M SW water level-high. Each reactor protection system instrumentation channel containing a limit switch should be shown to be operable by the Channel Functional Test

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monthly and Channel Calibration at each refueling outage.

2.3 Iro/ SURVEILLANCE REQUIRDGNTS FOR CONTROL ROD WITHDRAWAL BLOCK SDV LIMIT SWITCHES The NRC staff's Model Technical Specifications specify the following ICO/

surveillance requirements for control rod withdrawal block SUV limit switches:

"3.3.6 - The control rod withdrawal block instrumentation channel shown in Table 3.3.6-1 shall be OPERABLE with trip se* points set consistent -

with the values shown in the Trip Setpoint column of Table 3.3.6-2.

Table 3.3.6-1. Control Rod withdrawal Block Instrument 5 tion Minimum Operable Applicable Channels Per Trip Operational Trip Function Function Conditions Action 5.

Scram Discharge Volume a.

Water level-high 2

1, 2, 5**

62 b.

Scram trip bypassed 1

'(1, 2, 5 * * )

62 ACTION 62: With the number of OPERABLE channels less than required by the minimum OPERABLE channels por Trip Function requirement, place the inoperable channel in the tripped condition within one hour.

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- With more than one control rod withdrawn. Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.

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Table 3.3.6-2 Control Pod Withdrawal Block Instrumentation Setpoints Trip Function Trip Setpoint Allowable Value 5.

Scram Discharge Volume a.

Water level-high To be specified NA b.

Scram trip bypassed NA NA "4.3.6 - Each of the above control rod withdrawal block trip systems and instrumentation channels shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.6-1.

Table 4.3.6-1. Control Rod Withdrawal BlSck Instrumentation Surveillance Requirements Operational Conditions Channel in Which Trip Channel Functional Channel Surveillance Function Check Test Calibration Required 1

5.

Scram Discharge Volume 5

a.

Water Level-NA Q

R 1, 2, 5**

High b.

Scram Trip NA M

NA (1, 2, 5**)

Bypassed

- With more than one control rod withdrawn. Not applicable to control l

rods removed per Specification 3.9.10.1 or 3.9.10.2."

l Paragraph 3.3.6 and Table 3.3.6-1 of the Model Technical Specifications require the control rod withdrawal block instrumentation to have at least 2 t

operable channels containing 2 limit switches for SDV water level-high and 1 operable channel containing i limit switch for SDV scram trip bypassed. The technical objective of these requirements is to have at least one channel

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containing one limit switch available to monitor the SDV water level when the other channel-with a limit switch is being tested or undergoing maintenance.

The trip setpoint for control rod withdrawal block instrumentation monitoring SDV water level-high should be specified as indicated la Table 3.3.6-2.

The trip function prevents further withdrawal of any control rod when the control rod block SDV limit switches indicate water level-high.

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r TER-C5506-74 Paragraph 4.3.6 and Table 4.3.6-1 require that each control rod withdrawal block instrumentation channel containing a limit switch be shown to be operable by the Channel Fur stional Test once per 3 months for SDV water level-high and once per month for SDV scram trip bypassed, and by Channel Calibration at each refueling outage for SDV water level-high.

i The Surveillance Criteria of the BWR Owners Subgroup given in Appendir. A, "Long-Ters. Evaluation of Scram Discharge System," of " Generic Safety Evaluation Report BWR Scram Discharge System" (9] written by the NRC staff and issued on December 1, 1980, are:

1.

Vent and drain valves shall be periodically tested.

2.

Verifying and level detection instrumentation shall be periodically tested in place.

3.

The operability of the entire system as an integrated whole shall be demonstrated periodically and during each operating cycle, by demonstrating scram instrument response and valve function at pressure and tegerature at approximately 50% control rod density.

Analysis of the above criteria indicates that the NRC staff's Model Technical Specifications requirements, the acceptance criteria fo'r the present TER, fully cover the BWR Owners Subgroup Surveillance Criteria 1 and 2 and partially cover Criterion 3.

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TER-C5506-74 3.

METHOD OF EVALUATION The PASNY submittal for the FitzPatrick Nuclear Power Plant was evaluated in two stages, initial and final.

During the initial evaluation, only the NRC staff's Model Technical Specifications requirements were used to determine if o

the Licensee's submittal was responsive to the July 7, 1980 NRC request for proposed Technical Specifications changes involving the surveillance requirements of the SDV vent and drain valves, 140/ surveillance requirements for reactor protection system SDV limit.

switches, and LCO/ surveillance requirements for control rod block SDV limit switches o

the submitted information was sufficient to permit a detailed technical evaluation.

During the final evaluation, in addition to the NRC staff's Model Technical Specifications requirements, background material in References 1 through 10, pertinent sections of FitzPatrick Nuclear Power Plant Final Safety Analysis Report, and FitzPatrick Technical Specifications were studied to determine the technical bases for the design of SDV main composents and instrumentation. Subsequently, the Licensee's response was compared directly to the requirements of the NRC staff's Model Technical Specifications. The findings of the final evaluation are presented in Section 4 of this report.

'Ihe initial evaluation concluded that the Licensee's submittal was responsive to the NRC request of July 7,1980, and the submittal contained sufficient information to permit preparatier. 7f a TER without a Request for Additional Information.

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TECHNICAL EVALUATION 4.1 SURVEILLANCE REQUIREMENTS FOR SDV DRAIN AND VENT VALVES NRC STAFF'S MODEL TECHNICAL SPECIFICATIONS Paragraph 4.1.3.1.1 requires demonstrating that the SDV drain and vent valves are operable by:

a.

verifying each valve to be open at le.ast once per 31 days (valves may be closed intermittently for testing under administrative controls) b.

cycling each valve at least one complete cycle of full travel at least once per 92 days.

LICENSEE RESPONSE The Licensee proposed to revise pages 89 and 96 of the FitzPatrick Technical Specifications as follows (see Appendix B):

"4.3 (Cont'ci) b.

The scram discharge volume drain and vent valves shall be verified open at least once per 31 days (these valves may be I

closed intermittently for testing under administrative control)."

"3.

All control rods shall be determined operable once each operating

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cycle by demonstrating the scram discharge volume drain and vent j

valves operable when the scram test initiated by placing the mode l

switch in the SHUTDOWI position is performed as required by Table j

4.1-1 and by verifying that the drain and vent' valves:

i a.

Close within 80 seconds af ter receipt of a signal for control rods to scram, and b.

Open when the scram signal is reset or the scram discharge instrument volume trip is bypassed."

In addition, the Licensee agreed to revise proposed specifications changes to require cycling each valve at least one complete cycle of full travel at least quarterly.

FPC EVALUATION The added paragraph 4.3b on the revised page 89 of the FitzPatrick

Nuclear Power Plant Technical Specifications and the above agreed-upon additional nk!!n Research Center A DMuon of The Fraseen humane t

4 TER45506-74 revision comply with the requirements of paragraphs 4.1.3.1.la and 4.1.3.1.lb of the NRC staff's Mddel Technical Specifications.

4.2 ICO/ SURVEILLANCE REQUIREMENTS FOR REACTOR PROTECTION SYSTEM SDV LIMIT SWITCHES NRC STAFF'S MODEL TECHNICAL SPECIFICATIONS Paragraph 3.3.1 and Table 3.3.1-1 require the functional unit of SDV water level-high to have at least 2 operable channels containing 2 limit switches per trip system, a total of 4 operable channels containing 4 limit switches per two trip systems for the reactor protection system which automatically initiates scram.

Paragraph 3.3.1 and Table 3.3.1-2 concern the response time of the reactor protection system for the functional unit of SDV water level-high which should be specified for each BWR (it is not specified in the table). Paragraph 4.3.1.1 and Tab), 4.3.1.1-1 require that each reactor protection system instrumentation channel containing a limit switch be shown to be operable for the functional unit of SDV water level-high by the Channel Functional Test monthly and Channel Calibration at each refueling outage. The applicable operational ecnditions for these requirements are startup, run, and refuel.

LICENSEE RESPONSE The NRC staff's Model Technical Specifications requirements of paragraph 3.3.1 and Table 3.3.1-1 are addressed in the PitzPatrick Technical Specifica-tiens original page 41a, Table 3.1-1 (cont'd), Reactor Protection System (Scram) Instrumentation Requirement, which provides the following information for Trip Function High Water Level in Scram Discharge Volume "1.

Minimum No. of Operable Instrument Channels per Trip System (1) : 2 2.

Trip I4 vel Setting: < 36 gal 3.

Modes in Which Function Must be Operable: Refuel (2) (6), Startup, Run 4.

Total Number of Instrument Channels Provided by Design for Both Trip Systems: 4 Instrument Channels gigyghgter

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Action (1) :

A" Notes:

"(1)

There shall be two operable or tripped trip systems for each function, except as specified in 4.1.D.

From and af ter the time that the minimum number of operable instrument channels for a trip system cannot be set, the affected trip system shall be placed in the safe (tripped) condition, or the appropriate actions listed below shall be taken.

A.

. Initiate insertion of operable rods and complete insertion of all operable rods within four hours.

(2) Permissible to bypass in Refuel and Shutdown positions of the Reactor Mode Switch.

(3) When the reactor is subcritical and the reactor water temperature is less than 212"F, only the following trip functions need to be operable:

A.

Mode Switch in Shutdown B.

Manual Scram C.

High Flux IRM D.

Scram Discharge Instrument Volume High Level when any control rod in a control cell containing fuel is not fully inserted.

E.

APRM 15% Power Trip."

Note (3)D is taken from the revised page 43 (see Appendix B).

l Page 103 of the FitzPatrick Technical Specifications gives the reactor protection system response time as follows:

i "In the analytical treatment of the transients, 290 milliseconds are allowed between a neutron sensor reaching the scram point and the start of motion of the control rods. This is adequate and conservative when compared to the typical delay of about 210 milliseconds estimated from scram test results."

This fulfills the requirements of paragraph 3.3.1 and Table 3.3.1-2.

l The Licensee's response to the requirements of paragraph 4.3.1.1 and Table 4.3.1.1-1 was the sutaittal of the revised pages 44 through 47.

The revised page 44 contains Table 4.1-1, Reactor Protection System (Scram)

Instrument Functional Tests, Minimum Functional Test Frequencies for Safety nk!!n Research Center A Chemen of The Frarsen bushee m

,-i d

TER-C5506-74 Instrument and Control Circuits, with the following informatien in regard to High Water Level in Scram Discharge Instrument Volume:

"1.

Group (2): A 2.

Functional Test: Trip Channel and Alarm 3.

Minimum Frequency (3): Once/ month and before each startup (6), (7).

Notes:

(2)

A description of the three groups is included in the Bases of this Specification.

(3)

Functional tests are not required on the part of the system that is.

not required to be operable or are tripped.

(6)

Functional test of the instruments before each startup is required only if a scram has occurred since the last functional test or calibration.

(7)

The functional test shall be performed utilizing a water column or similar device to provide assurance that damage to a float or other portions of the float assembly will be detected.

A.

on-off sensors that provide a scram trip function."

Notes (2) and (3) are taken from the revised page 45.

Notes (6) and (7) were revised and are taken from the revised page 45a.

The revised pages 46 and 47 provide the following information in Table A 2, Reactor Protection System (Scram) Instrument Calibration, Minimum

...bration Frequencies for Reactor Protection Instrument Channels, for I...crument Channel High Water Level in Scram Discharge Instrument Volume:

"1.

Group (1): A 2.

Calibration (4):

W ter Column, Note (6) a 3.

Minimum Frequency (2)*: Once/ Operating Cycle, Note (6) l Notes:

l l

(1) A description of the three groups is included in the Bases of this Specification.

"The title of column 4 on the revised page 46, Table 4.1-2, should be " Minimum Frequency (2)," instead of " Minimum Frequency Once/ Week",

l Ub Frank!In Raearch Center t

A Osamen of The Fransen Insensee

O

(

TER-C5 506-74 (2) Calibration test is not required on the part of the system that is not required to be operable, or is tripped, but is required prior to return to service.

(4) Response time is not a part of the routine instrument channel test but will be checked once per operating cycle.

(6) Calibration shall be performed utilizing a water column or similar device to provide assurance that damage to a float or other portions -

of the float assembly will be detected."

FRC EVALUATION The original page 41a, Table 3.1-1 of the FitzPatrick Technical Specifica-tions meets the NRC staff's Model Technical Specifications requirements of paragraph 3.3.1 and Table 3.3.1-1.

The FitzPatrick reactor protection system SDV water level-high instrumentation consists of 2 operable channels contain-ing 2 limit switches per trip system, for a total of 4 operable channels con-taining 4 limit switches per two trip systems, making 1-out-of-2-taken-twice logic. The original page 41a, Table 3.1-1 also specifies < 36 gal as a trip setting for' scram initiation and applicable operating conditions of refuel, starttp, and run, which are acceptable.

The reactor protection system response time of 290 milliseconds specified l

on page 103 of the FitzPatrick Technical Specifications is acceptable and meets the requirements of paragraph 3.3.1 and Table 3.3.1-2.

The revised pages 44 through 47 with Table 4,1-1 and Table 4.1-2 of the FitzPatrick Technical Specifications comply fully with the NRC staff's Model Technical Specifications requirements of paragraph 4.3.1.1 and Table 4.3.1.1-1.

(

They prescribe the Channel Functional Test to be performed monthly and the Channel Calibration to be performed once per operating cycle, which is equiva-lent to refueling outage, as specified.

I l

OlS ' Franklin Research Center A DMmen of The Frename busewas

.~

- -.,..... ~, -. _, _,. -,.. _. _.,

. - l 6

TER-C5506-74 4

i 4.3 I40/ SURVEILLANCE REQUIREMENTS FOR CONTROL RCD WITHDRAWAL BLOCK SDV LIMIT SWITCHES NRC STAFF'S MODEL TECHNICAL SPECIFICATIONS Paragraph 3.3.6 and Table 3.3.6-1 require the control rod withdrawal block instrumentation to have at least 2 operable channels containing 2 limit j

switches for SDV water level-high, and 1 operable channel containing.1 limit switch for SDV trip bypassed. Paragraph 3.3.6 also requires specifying the trip setpoint for control rod withdrawal block instrumentation monitoring SDV-water level-high as indicated la Table 3.3.6-2.

Paragraph 4.3.6 a'nd Table 4.3.6-1 require each control rod withdrawal block instrumentation channel containing a limit switch to be shown to be operable by the Channel Functional Test once per 3 months for SDV water level-high and once per month for SDV scram trip bypassed, and by Channel Calibration at each refueling outage for SDV water level-high.

LICENSEE RESPONSE In response to the Model Technical Specifications paragraph 3.3.6, Table 3.3.6-1 and Table 3.3.6-2 requirements, the Licensee proposed revising pages 72 and 73 of the FitzPatrick Technical Specifications, which contain Table 3.2-3, Instrumentation That Initiates Control Rod Blocks, with the following information for Scram Discharge Instrument volume High Water Level "1.

Minimum No. of Operable Instrument Channels Per Trip Systems l'

2.

Trip Level Setting:,5 18 gallons 3.

Total Number of Instrument Channels Provided by Design for Both i

Channels: 1 Inst. Channel i

l-4.

Action (9), (10)

Notes t i

(9) When the reactor is subcritical and the reactor water temperature is l

less than 212*F, the control rod block is required to be operable only if any control rod in a control cell containing fuel is not i

fully inserted.

l.

4 -

I OUUU Frankun Research Center j

4 osamen er ne renneen m i

e i

TER-C5506-74 M

(10) When the control rod block function associated with scram discharge instrument volume high water level is not operable when required to be operable, the trip system shall be tripped."

The Licensee responded to the NBC staff's Model Technical Specifications requirements of paragraph 4.3.6 and Table 4.3.6-1 with proposed revision of page 81, Table 4.2-3, Minimum Test Calibration Frequency for Control Rod Blocks Actuation, which contains the following information for Instrument Channel Scram Discharge Instrument Volume-High Water Level:

"1.

Instrument Functional Test: Once/Menth (2) 2.

Calibration: Once/ Operating Cycle (2) 3.

Instrument Check: N/A Notes (2)

Functional tests, calibrations and instrument checks are not required when these instruments are not required to be operable or are tripped. Functional tests shall be performed before each startup with a required frequency not to exceed once per week.

Calibrations shall be performed prior to each startup or prior to preplanned shutdowns with a required frequency not to exceed once per week.

Instrument checks shall be performed at least once per day during these periods when the instruments are required to be operable."

FRC EVALUATION The existing FitzPatrick Nuclear Plant scram discharge system has six level switches on the scram discharge volume (see FSAR, page 3.5-11) set at three different water levels to guard against operation of the reactor without sufficient free volume present in the scram discharge headers to receive the scram discharge water in the event of a scram. At the first (lowest) level, one level switch initiates an alarm for operator action. At the second level, with the setpoint cf j[ 18 gallons (see the revised pag'e 72, Table 3.2-3), one l

level switch initiates a rod withdrawal block 'to prevent furthat withdrawal l

l of any control rod. At the third (highest) level, with the setpoint of j[ 36 gallons (see page 41a, Table 3.1-1 of the FitzPatrick Technical Specifications),

the four level switches (two for each reactor protection system trip system) initiate a scram to shut down the reactor while sufficient free volume is -

i Ud' Franklin Research Center 4 cza.aa as N rraman w an,.

TER-C5506-74 available to receive the scram discharge water. Reference 9, page 50, defines Design Criterion 9

(" Instrumentation shall be provided to aid the operator in the detection of water accumulation in the instrumented volume (s) prior to scram initiation"), gives the technical basis for "Long-Term Evaluation of Scram Discharge System," and defines acceptable compliance ("The present alarm and rod block instrumentation meets this criterion given adequate hydraul'ic coupling with the SDV headers"). Thus, if the FitzPatrick Nuclear Power Plant scram discharge system is modified (long term) so that the hydraulic coupling between scram discharge headers and instrumented volume is adequate and acceptable, then the present alarm and rod block instrumentation consisting of one operable instrument channel with one limit switch for control rod withdrawal block as specified on the revised page 72 is also acceptable.

In the FitzPatrick Nuclear Power Station, " Scram Discharge Volume Scram Trips" cannot be bypassed while the reactor is in operational conditiens of startup and run (see FSAR page 7.2-12) and operational condition " refuel with more than one control rod wir5 drawn" is not applicable, since interlocks are provided which prevent the withdrawal of more than one control rod with the mode switch in the refuel posit' ion. Thus, the NBC staf f's Model Technical Specifications requirements of paragraph 3.3.6 with Table 3.3.6-1 and paragraph 4.3.6 with Table 4.3.6-1 are not applicable to the FitzParrick Nuclear Power Station for " Trip Function 5. Scram Discharge Volume b. Scram Trip Bypassed" and were not addressed in the proposed revision of pages 72, 73, and 81.

This is acceptable.

I The 18-gallon trip level setting for control rod withdrawal block instrumentation is acceptable (see revised page 72 of the FitzPatrick Technical Specifications). The Licensee's proposed revision of page 81, Table 4.2-3 to meet the requirements of paragraph 4.3.6 and Table 4.3.6-1 is also acceptable since it prescribes the Channel Functional Test of each control rod withdrawal block instrumentation channel containing a limit switch once per month (required once per 3 months) and Channel Calibration once per operating cycle for SDV water level-high.

nklin Research Center A OMemn of The Frannen insamme

4 TER-C5 506-74 5.

CONCLUSIONS Table 5-1 summarizes results of the final review and evaluation of the FitzPatrick Phase 1 proposed Technical Specifications changes for SDV long-term modification in regard to surveillance requirements for SDV vent and drain valves and LCO/ surveillance requirements for reactor protection system and control rod block SDV limit switches. The following conclusions were mades o

The revised page 89 of the FitzPatrick Technical Specifications and the Licensee's agreement to add to the proposed specifications changes a requirement to cycle each valve a minunum of one full cycle at least quarterly comply with the NRC staf f's Model Technical Specifications, paragraphs 4.1.2.1.la and 4.1.3.1.lb.

o The remaining surveillance requirements are met by revised pages 43, 44, 45, 45a, 46, 47, 72, 73, 81, 99, 89a, and 96 of the FitzPatrick Technical Specifications and by pages 41a and 103 without revision.

l l

i i

nklin Research Center A Omerm et The Feween enmanae

4 i

S,E mf Table 5-1.

Evaluation of Phase 1 Proposed Technical Specifications Changes for Scram Discharge Volume Long-Term Modifications y;

gg FitzPatrick Nuclear Power Plant to 1d" 4

Technical Specifications

r NRC Staff Model Proposed by IQ Surveillance Requirements (Paragraph)

Licensee Evaluation to k

SUV DRAIN AIO VENT VALVES Verify each valve open Once per 31 days Once per 31 days Acceptable (4.1. 3.1. la)

(p. 89, revised)

Cycle each valve one Once per 92 days Quar terly Acceptable 4

complete cycle (4.1.3.1.lb)

(p. 89, revised)

Y REAC'IOR PR0rECTION SYSTEM SDV LIMIT SWI'IDIES Minimum operable channels 2

2 Acceptable per trip system (3.3.1, Table 3.3.1-1)

(p. 41a, Table 3.1-1)

SDV water level-high NA 290 millisec. max.

Acceptable response time (3.3.1, Table 3.3.1-2) 210 millisec test (p. 103)

SDV water level-high Channel functional test m nthly Once per month Acceptable (4.3.1.1, Table 4.3.1.1-1)

(p. 44, Table 4.1-1, revised)

Channel calibration Each refueling Once per operating Acceptable j

(4.3.1.1, Table 4.3.1.1-1) cycle (p. 46, 47, Table 4.1-2, revised) j l

t

Table 5-1 (Cont.)

Technical Specifications gg NRC Staff Model Proposed by

,}

Surveillance Requirements (Parag raph)

Licensee Evaluation CONTROL ROD BIDCK SDV LIMIT SWITHES

,9 g

Minimum operable channels E

per trip fur.ction Q

g SDV water level-high 2

1 Acceptable *

(3.3.6, Table 3.3.6-1)

(p. 72, 73, Table 3.2-3, revised)

SDV scram trip bypassed 1

NA Acceptable *

(3.3.6, Table 3.3.6-1)

(p. 72, 73, Table 3.2-3, revised)

SDV water level-high l

Trip setpoint NA

< 18 gallons Acceptable (3.3.6, Table 3.3.6-2)

(p. 72, 73 Table 3.2-3, revised) i 1

l Channel functional test Quarterly Once per month Acceptable l

(4.3.6, Table 4.3.6-1)

(p. 81, Table 4.2-3, revised) 1 Channel calibration Each refueling Once per operating cycle Acceptable' l

(4.3.6, Table 4.3.6-1)

(p. 81, Table 4.2-3, I

revised) 1 l

l SDV scram trip bypassed

?

Channel functional test Monthly NA Acceptable *

(4.3.6, Table 4.3.6-1)

See Reference 9, p. 50, and pp.19 and 20 of this TER.

t

o TER-C5506-74 6.

REFERENCES 1.

IE Bulletin 80-14, " Degradation of BWR Scram Discharge. Volume Capacity" NRC, Office of Inspection and Enforcement, June 12, 1980 2.

D. G. Eisenhut (NRR), letter "To All Operating Dolling Water Reactors (BWRs)" with enclosure, "Model Technical Specifications,"

July 7, 1980 3.

IE Bulletin 80-17, " Failure of 76 of 185 Control Rods to Fully Insert During a Scram at a BWR" NRC, Office of Inspection and Enforcement, July 3,1990 4.

IE Bulletin 80-17, Supplement 1, "Failute of 76 of 185 Control Rods to Fully Insert During a Scram at a BWR" NRC, Of fice of Inspection and Enforcement, July 18, 1980 5.

IE Bulletin 80-17, Supplement 2, " Failures Revealed by Testing Subsequent to Failure of Control Rods to Insert During a Scram at a DWR" NRC, Of fice of Inspection and Enforcement, July 22, 1980 6.

IE Bulletin 80-17, Supplement 3, " Failure of Control Rods to Insert During a Scram at a BWR" NRC, Office of Inspection and Enforcement, August 22, 1980 7.

IE Bulletin 80-17, Supplement 4, " Failure of Control Rods to Insert During a Scram at a BWR" NRC, Office of Inspection and Enforcement, December 18, 1980 8.

IE Bulletin 80-17, Supplement 5, " Failure of Control Rods to Insert During a Scram at a BWR" NRC, Office of Inspection and Enforcement, Februa ry 13, 1981 9.

P. S. Check (NRR), memorandum with enclosure, " Generic Safety Evaluation Report BWR Scram Discharge System" December 1, 1980 10.

P. S. Check (NRR), memorandum with enclosure, " Staff Report and Evaluation of Supplement 4 to IE Bulletin 80-17" June 10, 1981 nk!!n Research Center A Dnamon of The Frarwen ownme

a i

f APPENDIX A

.l i

1 NBC STAFF'S MODEL TECHNICAL SPECIFICATIONS

f i

i I

l I

l l

i

Note: Applicable changes are marked by vertical lines in the margins.

d. Franklin Re. search Center a on a w n em m 4

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DER-4:5506-74 REACTIVITY CONTROL SYSTEMS i

LIMITING CCNDITICN FOR CPERATION (Continued)

A,CTION (Continued) 2.

If the inoperable control rod (s) is inserted, within one hour disarm the associated directional control velves either:

a)

Electrically, or b)

Hydraulically by closing the drive water and exhaust water isolation valves.

3.

Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

c.

With more than 8 control rods inoperable, he in' at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

!URVEILLANCE #ECU1REMENTS 4.1.3.1.1 The scram discharge volume drain and vent valves shall be demonstrated OPERABLE by:

a.

Verifying each valve to be open* at least once per 31 days and b.

Cycling each valve through at least one complete cycle of full.

travel at least once per 92 days.

4.1.3.1.2 When above the preset power level of the RWM and RSC5, all withdrawn control rods not required to have their directional control valves disareed electrically or hycrawlically shall be demonstrated CPERABLE by moving eacn control rod at least one notch:

(

a.

At least once per 7 days, and b.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when any control rod is immovable as a result of excessive friction or mechanical interference.

4.1.3.1.3 All control rods shall be demonstrated CPERABLE by performance of Surveillance Retuirements 4.1.3.2, 4.1.3.4,.4.1.3.5, 4.1.3.6 and 4.1.3.7.

These valves may be closed intermittently for testing under administrative controls.

ra l

GE-STS 3/4 1-'

i l

A-1

. ng' Franidin Research Center duu A oheemn of The Fransen insusuas

.=

e TER-C5506-74 REACTIVITI CONTROL SYSTEWS CONTR:t, r.00 mar! MUM SCRAM IRf!RT CN TIMES LIMITTwG CCwCITION FCR CPE UTION 2.1. 2. 2 The etximu: scram insertion tize of each c:ntrol red from the fully withdrawn position to notch position (6), based on de-energitation of the scram pilot valve solenoids as time zero, shall not exceed (7.0) seconds.

APPLICASILITY: OPERATIONAL'CCNDITICNS 1 and 2.

ACTICH:

Vith the maximum scram insertion time of one or oore control rods exceeding (7.0) seconds:

a.

Declare the control rod (s) with the slow insertion time inoperable, and

~

b.

Perform the Surveillance Requirements of Specification 4.1.3.2.c at least once per 50 days when operation is cantinued with three or more control rods with maximum scram insertion times in excess of (7.0) seconds, or c.

Se in at least HOT SHUTDCVN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

2 SURVEILLANCE RECU1 RESENTS 4.1.3.2 T'ha r.aximum scram insertion time of the control rods shall be demon-strated througn measure:ent with reactor coolant pressure greater than or

' equal to 950 psig and, durin; single control rod scram ti=e tasts, the control rod drive pumps iso'ated from the accumulators:

\\

a.

Fer all control rods prior to THE7?AL POVER exceeding AC% of RATED THEKr.AL PCVER following CCRE ALTE7ATICNS or aftar a reactor shutdown j

that is greatar than 120 days, b.

For specifically affected individual cont ci rods following r.aintenance on or =ccification to the control red or control r:4 drive system which could affect the scran insertion time of those specific cont c) rods, and c.

For 1C% of the control rods, on a rotating tasis, at least once per 120 days of c;eration.

i CE-STS 2/4 1-5 l

t c-A-2

. n! Frank!!n Research Center

$1 4 osa,.an as N r,.n.. m

s 1

TER-C5506-74 3 /a. 3 INS 71'ENTATICN gaj.1 REACTOR PROTECTTCN SYST E INSTtt:.ENTATICN t.3viTIy2 CCNDITICH FOR CFERATION 3.3.1 As a cini=u=, the react:r protetti:n systas instru:zntation channels sa,n in Tule 3.3.1-1 shall be CPETAELE with the REACTOR ??.3TECTICN SYSTIC4 723PCNSE TIME as shown in Table 3.3.1-2.

17 8'.* :A!!LITY: As shown in Tabit 3.3.1-1.

/CICN:

Vith the numbe' of CPERAILE channels less than required by the Minimus a.

r CPE?ABLE Channels ;er Trip Systes requitecent for one trip system, place at least one inoperable channel in the tripped condition within one hour.

Vith the number of CPERAILE channels less than required by the Minimus CPE?AILE Channels per Trip Systes requirement for both trip systems, place at leut one inoperable channel in at least sne aria systes" in the tri;;ed c:ndition wiuin one hour and *-ake the ACTICN required by Ta:la 3.3.1-1.

The ; revisions of !;ecification 3.0.3 are not applicable in CPEP.ATICNAL CCNDITICM !.

3 Sv!!LLikCI *!CUIRD'ENTS 4.3.1.1 Ea:M reactse pr:tection system instrumentation channel shall be cas:r.strated CPE?A8LE by ue performance of the CF.ANNEL CHECX, CMANNEL TW,CTI*NAL TEST and CF.ANNEL CALI5 RATION :;erations for the C?i?ATIONAL C:NDITICNS and at the frequencies shown in Tcle 4.3.1.1-1.

a.3.1.2 LOGIC SYSTEM FL'NCTIONAL TESTS and simulated aute atic operation of all cannels shall be perfor:ed at least once per 18 monus.

4.3.1.3 The P.E.aCTCR PROTi*7*:N SY3 TEM RESPCMSE TIME of each reacter trip fun::t:n sh=wn in Taale 3.3.1-2 shall be demonstrated to be wiuin its limit at ieut :nca ;er 15 sanus. Each test shall include at least one logic train su:h that esta leaic t ains are tested at least :n:s per 35 senus and :ne enannel ;er funcdon such cat all channels are tes.ed at least once eve y N ti=ss 13 ::nus =here N is ce total nur.:er of reduncant channels in a,.

spe:f fi: rtac'.:P trip function.

of :c.3 : anneis are ine;eracle in one tri; rystam, select at least one

^

in::trable enannel in uat trip system to ;iace in the tripoed c:ndition, c:::. when uis culd cause na Trip Func-ion ta oc:ur.'

E-!"3 3/4 3-1

!]00hrenidin Research Center A Deussen of The Feuroen insubme

i om a 5' TABLE 3.3.1-1 (Continued)

RfAC10R PROT [CTION SYSTEM INSTRUMENTATION l0 m

.A APPLICAnlE MINIMUH

. OPERATI0ilAL OPEllAOLE CllANNELS fullCiloilAL UltlT CJHDIll0NS Pla 1AIP SYS1[M (a)

ACTION 8.

Scrane alscliarge Volume W ter II 3

l Level - l1196

1. 2. 5 'I 2

4 3

III III 4

7 9.

Turbine Stop Valve - Closure I

I*

10. Turbine Control Valve Fast Closure, II}

III 2

7 1 rip 011 Pressure - tow I

11. Reactor Hode Switch in 51ustdown Position 1,2,3,4,5 1

8

12. Nnual Scram 1,2,3,4,5 1

9 s.,

4 N

3

=

A 9

i l

a TER-C5506-74 T!!'.E 3.3.1-1 (Continued) fia: f FIOTECTION SYSTEh INSTRUMENTATICM ACTICN In 07ERATIw*WA:. CONDITICH 2, be in at least MDT 51 CIT 0*'aM within SOTIN1 6 ho:rs.

In 0?EEATICN L CONDITICN 5 suspend all operations involving CORE ALTE%TI:XS* and fully insert all insertable control rods within are ho:r.

Lock the reec:or mode switch in the Shutdown sc4ftion within 1CTION 2 one tour.

Se is at laast STARTUP within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

A". I:N3 In 0:E?.A*IONA:. CONDITICN 1 or 2, be in at least HtT SWTDChN ACTIN 4 within 6 neurs.

In 07E?A IONA*. CONDITION 5, suspend all operations involving CORE ALTruTI N5* and fully insert all insertable control rods with!n are hose.

Se 1: at least HOT SHUT 3C%W within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

1:TI:N5 Se f: START:J.P with the main staas ifne isolation valves closed A:"'!*.N 5 within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or in at least HOT SHUTD;"N within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

Ir.itfate a rsfuction in THEK."AL PCWEP. within 15 minutes. and A;TI;N 7 reda:e t:-::ine first stage pressure to < (250) psig, equivalent to T:iE?$L ?C'TR 1ess than (303% of PATED THE.ML PCWER, within 2 ho:rs..

In C:E?rIONA:. C hDITICH 1 or 2, he in at lacs?. ICT SM' TOO"nN J

- I NS within 6 heun.

In C7E?AMONE CONDITION 3 or 4, verify all insertable c:r. trol rods t.o te fully inserted wittin one hour.

In 07E?rIONE CONDITICN 5, suspend all operations involving CCRI ALTI.%TI;NS* and fully insert all insertable control rods within ora hoar.

A TION 9 In C:E?r!OMA*. CONDITICN 1 or 2, be in at least H31 SHUTLC"nN within 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months .

In C7E?A77CNE CONDITION 3 or 4, lock the reactor mode switch in the I.*u do.n pos ' tion within one hour.

In C?E:ATIONA*. CONDITICN 5, suspend all c;erations involvfag C RI A:TI.uTI:NS" and fully ir. sert, all insartable c:ntrol rods within ore noor.

Ex:::- covecent of I?#.. S??. or special c:vable detectors, or replacement of

_??.4 s.-ings pr:viceC !??. i:stru ea.tati:n is CPE?A3LE per Specificati:n 3.9.2.

II.75 3/43-4 Nbranklin Research Center h or m r,a.sn m

TER-C5506-74 TA3LE 3.3.1-1 (C ntinued)

A! ACTOR 787,TECTION SYSTEM INSTRUMENTAT1 N TA!LE NOTATICNS (a) A channel may be placed in an in=;erable status for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for required surveillar.:e without placing the trip system in the tripped concition provided at, least one CPERABLE channel in the same trip systaa is sonitoring thai. ;arar.eter, b)

The " shorting ifnks' shall be removed from the RP5 circuitry prior to and during the time any control rod is withdrawn" and shutdown margin demonstrations perforced per Specification 3.10.3.

(c) An APRM channel is incperable if there are less ttan 2 LPRM inputs per level or less thari (11) LPRM inputs to an APRM channel.

(d) These functions are act required to be CPERA3LE when the reacter pressure vessel head is ur.boited or removed per Specification 3.10.1.

(e) This function shall be automatically bypassee when the reactor sede switch is not in the Run position.

(f) This function is not required to be CP!FAILE when PRIMARY CONTAINWSVT INTEGRITY is not required.

(g) Also actuates the standby gas treatment system.

l (h) Vith any centrol red withdrawn. Not applicable to c=ntrol rods removed

er Specificati: n 2.9.10.1 or 3.9.10.2.

(i) These functicas are aut:r.atically bypassed wnen turbine first stage pressure is < (253) ;sig, equivalent to THEP."AL PCE"R 1ess t.,an (13)%

of RATED THEX,wAL M.'ER.

(j) Also actuates the E C-RPT system.

"N:

requirec 1:r c:nt ol r:ds re==ved per specification 3.9.10.1 or 3.9.10.2.'

n GE-575 3/4 3-5 4

f A-6 dj0 Frankun Resear.ch Center

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V37tUw!.NTATION J 'a. 3. 6 CONTROL RCD VI FDRAVAL ELOCK INSTRLYEWATICN L:w: TING CCNDITION FOR CPERATION

3. 3. 5.

The enntest red withdrawal block instrumentation channels shown in Tah e 3.3.5-1 shall be CPERA!LE wi.h their trip set;sints set consistant with t.

values sh: n in the Trip setpoint column of Ta21e 3.3.5-2.

A* St!CA8!LITY: As shcwn iis'Tatic 3.3.5-1.

A 7:CN:

a.

With a control rod withdrawal block instru=entation channel trip setpoint less conservative than the value shown in the Allowabie Values eslumn of Tanle 3.3.5-2, declare the channel inoperable until the channel is restored to CPERABLE status with its trip setpoint adjusted consistent with the Trip 5etpoint value.

b.

With the number of CPERA8LE channel.1 less than required by the Minimum OPEMBLE Channels per Trip Function, requirement, take the ACT*CN required by Tanle 3.3.5-tr.

c.

The ;rovisions of Specification 3.C.3 are :t a;plicabia in CPEM-TICNAL CCNDITICN 5.

!"RVE!LLANCE RECUIRESENTS 4.3.5 Each of the abcve required control r:d withdrawal hiock trip systems ard instru:entation channels shall be descastrated :PEU3LE ty the perfor=ance

f t.e CMANNEL CMECK, CHANNEL FUNCTICNAL 7537 anc CMANNEL CALIERATICN cpera-ti:es for the CPERATICNAL CCNDITICNS and at the frequencies shewn in Tacle 4.3.5-1.

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TER-C5506-74 TAtt! 3.3.5-1 (Continued)

CCh'TRCL R00 VITH0RAVal ttcCIC INSTR.?'ENTATICN ACTION Taka the ACTICM required by Specification 3.1.4.3.

AZ:N 53 With the aucher of CPERABLE Channels:

A:T :N 61 a.

Cne less than required by the Minime: OPERAELE Channels per Trip function requirement, restore the inopera.ble channel to CPE?A8LE s*.atus within 7 days er place the in=perable channel in the tripped candition within the next hour.

b.

Two or more less than required by the Minimus CPEM3LE Channels per Trip Function recuirement, place at least one frieperable channel in the tripped condition within one hour.

Vith the number of CPERA!L! channels less than required by the AZ:N 52 Mini =u:n CPE?.ABLE Channels ;er Tri; Function requirement, place tne ineparable enannel in the tripped c:ndition within ene hour.

NOTES Vitn THETWAL PCVER 3, (20)% of RATED THE??AL PCh!R.

With :: ore tnan ene :sntrol rod withdrawn. Not a;:lica31e ta c:ntrcl rods ree:ved per Specification 3.9.10.1 er 3.9.10.2.

t.

The REM shall be av.cmatically bypassed wnen a ;eripheral c:ntrol red is,

selected.

h.

This functica shall be aut:matical'.y byjassed if detector count ra*.a is

> 100 c;s or the 1T01 channels are on range (2) or higher.

c.

This function shall be automatically bpassed wnen the associated IF.M enar.nels are on range 8 or higher.

2.

This function shall be autsmatically bypassed when t.9e IRM char.nels are

n -ange 3 or higner.

This function shall be automatically typassed when the IPJi channe'Is ire a.

n -ange 1.

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_TA?.'.E a.3. 5-1 (Continued)

CONTROL 200 $!I d RAVAL f t00X INSTRUMEh'TATION SURVE?tUNCE REQUIREMENTS NOTES:

a.

Neut.en detact:rs may be excluded fr:m CFANNEL CALIERATICN.

b.

Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to starttep, if not performed within the previous 7 days.

c.

When making an unscheduled change fr:a OPERATIONAL CONDITION 1 to CPERATICKAL CON 3IT! N 2, perfor= the required s:.rveillanca within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after entering :PERATICNAL CONDITICM 2.

Vits THEVAL PCVER 1 (20)% of RATED THERPAL POVER.

Vith any control rod withdrawn. Not a;plicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.

0 2

SI-I'3 2/4 3-!!

r, A-14 du.d,- -Franidin Research Center 4 % w w r nica w

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TER-C5506-74 1

4 i

i APPENDIX B POWER AUTHORITY OF THE STATE OF NEW YORK LETTER OF JANUARY 6, 1981

.x AND SUBMITTAL WITH PROPOSED TECHNICAL SPECIFICATIONS CHANGES FOR FITZPATRICK NUCLEAR POWER PLANT l

t nklin Rese

.-~_ arch Ce.nter

-~

a TER-C5506-74 REGULATypt INFORMATION DISTR!auTION SYSTEM (RIDS 1 ACCESSIUM o R : 8101 t 20 212 00C.DATE: 81/01/06 NOTAR1 ZED: ho' 00Cxh7 a FACIL:So-333 James A. Fit 2PatPick nuclear. Power Plant, Power Autno 05000333 AdT4.NAeE AUTHJR AFFILIATION BAYNE,J.P.

Power Autnority at the State of New York RECIP.NAHL AECIPIsNT AFFILIATION 1PPOLIT0,T.A.

Coerating Reactors Scanch 2

SUBJECT:

Forwards acolication for amend to License DPR-$9 revising Ano A Tech specs 6 safety _ evaluation re scram cischarge vol.

Class II amend tee enc 1 A0013-COPIES RECE!vED LTR $1 ENCL d. SIZE:_J! ? [5I DISTRIBUTION CODE:

TITLE: General Distribution for after Issuance of Coerating License NOTES:

V Ctf6d.g :

N000,oo RECIPIENT COPIES RECIPIENT CCPTES 10 CODE /NAME LTTR ENCL' 10 CODE /NAbE LTTR ENCL ACTION:

IPPOLITO,T.

04 13 13 INTERNAL: U/QIR,NUM FAC06 1

1 OIR,DIV 0F LIC 1

1 ItC 00 4

2 NRC POR 02 I

I OELD 11 1

0 OR ASSESS RR 10 1

0 dEG FILE 01 1

1 EXTERNAL: ACMS 09 to 16 LPOR 03 1

1 NSIC 03 1

1 a

e t

]

TOTAL Nuw8ER OF COPIES dEwu1RLD: LTTR 39 ENCL 3T ank!!n Research Center A Ohomson ad The Fransen m

1 TER-C5506-74 POWER AUTHORITY OF THE STATE OF NEW YORK 10 Cot.uhsaus Cincs.a New Yonx. N. Y. too19 (212) 347.4200

$*$$M

. = _

y

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January 6, 1981

**=

e......es

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JPN-80-2 ua*',",'c g'ac4^'.a m:. =

Director of Nuclear Reactor Regulation

.y C. S. Nuclear Regulatory Ccanaission Washington, D. C.

20555 Attention:

Mr. Thomas A. Ippolito, Chief g

Operating Reactors Branch No. 2

~.

o Division of Licensing

Subject:

James A. FitsPatrick Nuclear Power Plant Docket No. 50-333 Proposed change to the Technical Specification Related to Scram Discharge volume Dear Sir Enclosed for filing are three (3) signed originais and nine-teen (19) copies.of a document entitled, " Application for Amendment to operating License", together with forty (40) copies of Attachment I and II thereto, comprising a statament of the proposed changes to the Technical Specifications and the associated Safety Evaluation.

This application seeks to amend Appendix A of the operating License in accordance with the Commission's 2 t 7, 1990 letter which requested Technical Specification changes to provide surveil-lance for SDV vent and drain valves and LCO/ surveillance require-ments for RPS and control red block SDIV limit switches.

A list of the preposed changes to the Tec-hnical Specifications is given below 1.

The proposed change on page 43 (Table 3.1-1) defines specifically the condition when the SDIV Eigh Level trip function needs to be operable during cold shutdown.

2.

The first change on page 44 (Table 4.1-1) corrects a typographical error.

S 3.

The second change on page 44 increases the minimum frequency of the Scram Discharge Instrument Volume ggy.

Water Level trip channel and alarm functional test.

f '{ C00.00 SIS f

B-2 2Ju Franidin Research Center A om.,ea as rh. re.ama w

e TER-C5506-74 U. S. Nuclear Ragulatory Consnission 4.

The change on page 45 would replace AEC with NRC.

The change on page 45a is the proposed revised Note (6) and new Note (7). The present form of Note (6) was applicable during the 1977 refueling outage only.

The proposed Notes (6) and (7) are based upon Item A.5 of I. E.Bulletin 80-14.

5.

The proposed changes to page 46 and 47 (Table 4.1-21 result frcan I & E aulletin 80-14 (Item A.5) and the NRC letter dated July 7, 1980.

The proposed addition of Table 3.2-2 (page 72),

6.

together with the new Notes (9) and (10) on page 73 is written to be consistent with the proposed change to Table 3.1-1 (page 43).

7.

" Changes to Items 5 and 6 of Table 4.2-3 (page 81) are proposed to achieve consistency with Note 3 of Table 4.2-6.

The remaining changes in Table 4.2-3 are proposed as a result of I. E.Bulletin 80-14.

8.

The addition proposed to page 89 results from the I. E.Bulletin 80-14.

Other changes to page 89 and to page 89a are merely a renumbering of paragraphs l

following the proposed addition.

9.

The proposed addition to page 96 results from I. E.

Bulletin 80-14.

The Authority has classified this application for amendment to the operating licerse as Class III, resulting from the NRC

(

I. E.Bulletin 80-14 on Degradation of the BWR Se: am Discharge volume Capability and from the reference NRC letter. Enclosed is a check in the amount of $4,000 as the filing fee per 10CFR 170.22, which the Authority pays under protest pending a final determination of the legality of the fee schedule.

, Aery truly yours, 3.j Senior Vice ' resident

.uclear Generation nklin Research Center A Dnaemn of The Frannan innende

s TER-C5506-74 BEFORE THE UNITED STATES NUCLI.AR REGULATORY COMMISSION In the Matter of

)

POWER AUTHORITY OF THE STATE OF NEW YORK

)

Docket No. 50-333

)

James A. FitzPatrick Nuclear Power Plant

)

APPLICATION FOR AMENDMENT TO OPERATING LICEMSE Pursuant to Section 50.90 of the regulations of the Nuclear Regulatory Commission, the Power Authority of the State of New York, as holder of Facility Operating License No. DPR-59, hereby applies for an Amendment to the Technical Specifications con-tained in Appendix A of this license.

The proposed changes to the James A. FitzPatrick Technical Specifications occur in Sections 3.1, 3.2, 4.1, 4.2, and 4.3.

These proposed changes result from the NRC letter dated July 7, 1980, and are therefore relatsd to the control rod drive scram discharge volume capability.

The proposed changes to the Technical Specifications are presented in Attachment I to this application. The Safety Evaluation corresponding to this change is included in Attachment II.

POWER AUTHORITY OF THE l

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l USe ior Vice President ruclear Generation Subscribed.and sworn to before me this day of....%J.

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Notary Puclm-Rtm4 c. ZA89 Notary wg State of New York p

No. W

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i ATTACHMENT I i

PROPOSED OPERATING LICENSE ADDITION i

RELATED TO 00NI w L ROD DRIVE i

i SCRAM DISCHARGE VOLUME CAPABILITT t

i l'

POWER AUTHORITY OF THE STATE OF NEW YORK JAMES A. FITIPATRICK NUCLEAR POWER PLANT DOCKET NO. 50-333 JANUARY 6,

1981 51 B-5 s

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0 JAFNPP TAB 12 3.1-1 (cont'd) h REACTOR PROTECTION SYSTDI (SCRAM) INSrRUMENTATION REQUIREMENT

mf h

hares OF TABt2 3.1-1 (cont'd)

SE

.High Flum IRM IL 3 c.

2 <"e y

D.

Scram Discharge' Instrument Volume High Imvel when any control rod in a control cell cont'aining fuel i

_g is not fully inserted f:r i

Q E.

APRM 15g Power Trip

E'.

4 7.

Not required to be operable when primary containment integrity is not required.

8 Hot required to be operable when the reactor pressure vessel head is not bolted to the vessel.

9.

%e APRM downscale trip la automatically bypassed when the Iles Instrumentation is operable and not high.

10.

An APRM will be considered operable if there are at least 2 LPRMinpute per level and at least 11 LPRM inputs of the normal complement, s

on 11.

See section 2.1.A.1.

12

%1s equation w!!! be used in the event of operation with a maalmum fraction of limiting power density (MFLPD) greater than the fraction of rated power (FRP).

wheres FRP

- Fraction of rated thermal power (2436 MWL) i HFLPD - Maximum fraction of limiting power density where the limiting power density is 18.5 EW/ft for 7x7 fuel and 13.4 MW/f t for ex8, 8x8R and P8uSR fuel.

We ratio of FRP to MFLPD shall be set equal to 1.0 unless the actual operating value is less than the design value of 1.0, in which case 4,he actual operating value will be used W

- Icop;decirculation flow in percent of rated (rated is 34.2 x 106 l

lb/hr)

Y 4

- scram setting in percent of initial un 13 We Average Power Range Monitor scram function is varied (Figure 1.1-1) as a function of recirculation loop ui fis = (N). The trip setting of this function must be maintained in accordance with Specification 2.1.A.1.c.

S AmendmontNo./

43 e

t

s s

e-JAFNPP Table 4,1-1 l'

g, REACTOR PROFECTION SYSTDI (SCRAN) INSTRUNENT FUNCTIONAL TESTS a

E'*

MINTNUM FUNCTIOHAL TEST FREQUENCIES FOR SAFETY INSTRUNENT AND CONTROL CIRCUITS i

f2C s

i EL Group (2) runctional Test Minimum Frequency (3) 3h Mode Switch in Shutdown A

Place Mode tivitch in Shutdown Eich refueling outage.

I[a Manual Scram A

Trip Channel and Alarm Every 3 months.

1 gh 4

h RPS Channel Test switch A

Trip Channel and Alarm Every refueling outage or 1

af ter channel maintenance.

IRN High Flux C

Trip Channel and Alarm (4)

Once per week during re-fueling or startup as.d before each startup.

as Inopetative C

Trip Channel and Alerm(4)

Once per week during re-l e

fueling or startup and 4

before each startup.

APRM High Flux a*

Trip output Relaya(4)

Once/ week.

Inoperativa 3

Trip output Relays (4)

Once/ week.

Downscale a

Trip output Relays (4)

Once/ week.

Flow Blaa B

Calibrate Flow Bias Signal (4)

Once/ month.(1)

High Flux in Startup or Refisel C

Trip Output Relays (4)

Once per week during refueling or startup and before each startup.

High Reactor Pressure 2

Trip channel and Alarm (4)

Once/ month. (1) (Instrument check once per day) d High Drywell Pressure A

Trip Diannel and Alarm Once/ month (1) 1 Reactor tat water 14 vel (5)

A Trip Channel add Alarm Once/ month (1)

N

o i

4 High Water Level in Scram A

Trip Channel and Alarm Once/ month and before each Q

Discharge Instrument Volume startup(6), (7) g e

Main Steam Line High Radiation R

Trip Cleannel and Alarm (4)

Once/ week.

5 Amends.ont No. (2 44 t

y JAFNPP c=>

Table 4.1-1 (cont'd)

]q DEACTOR PROFEETION SYSTEN(SCRAN) INSTRtN4ENT FUNCTIONAL TESTS

!; E MINImmt FUNCTIONAL TEST FREQUENCIES EVR SAFETY INSTRUNENT AND CONTMOL CIRCUITS

a. o g.

g:%

Group (2)

Functional Test Ninimum Frequency (3) r fQ uain steam I.ine Isolation valve A

Trip Channel and Alarm Once/ month. (1) 2,

?-

Closure Turbine Control Valve EllC 011 Trip Channel and Alarm Once/ month.

1 Pressure Turbine First EMge Pressure A

Trip Channel and Alaine Every 3 months. (1)

Permissive Turbine Stop Valve Closure A

Trip Channel and Alarm

' Once/ month. (1) 4 Seactor Pressure Permissive A

Trip Channel and Alann Every 3 months.

I NOTES FOR TABLJE 4.1-1 1.

Initially once every month until acceptable failure rate data are availables thereafter,a request may be made to the NBC to change the test frequency. The cewilation of instrument failure rate data may include data l

obtained from other boiling water reactors for which the same design instrument operates in an environment similar to that of JAFNPP.

2 A description of the three groupe is included in the Bases of this Specification.

3.

runctiohal tests are not required on the part of the system tnat is not required to be operable or are tripped.

If tests are missed on parts not required to be operab'.e or are tripped, then they shall be performed prior to returning the system to an operable status.

4.

was instrumentation is enesyted from the instrument channel test definition. S ie instrimment channel N

i functional test will consist of injecting a simulated electrical signal into the instrument channels.

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Amendment No. Af 45 d

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JAFHPF BE Table 4.1-1 (cont'd)

e. 3 REACTOR PMorECTION SYSTEM (SCRAML INSTIENE!?rr FUNCTIONAL TESTS g$

MINIMUN FUNCTIONAL TEST FREQUENCIES FOR SAFET1f 3NSTRUMENT AND CONTROL CIRCUITS jn

r hQ no E

.HoTEs Poa TAaI2 4.1-1 (cont'd) 5.

no water level in the reactor vessel will be perturbed and the corresponding level indicator changee will be monitored. This, perturbation test will be performed every month at'ter completion of the fgnctional test program.

e*

6.

Functional test of the instrumente before each startup le required only if a scram has occurred eince the last functional test or calibration.

7.

We functional test shall be performed utilla;., a water column or similar device to provide assurance that damage to a float or other portions of the float assembly will be detected.

G 9

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Amendment No. (9 45a I

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JAPHPP Table 4.1-2 g,

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REACTOR PROTECTION SYSTEM (SCRAM) INSTALMENT CALIBRATION y\\q MININUM CALIBRATION FREQUENCIES FOR REACTOR PROTECTION INSTRUNENT CHANNEI.S G

a' s N

IrStrument Channel Group (1)

Calibration (4)

Ninimun Frequency once/ week

'I IRN High Flux C

Comparison to APRM on Maximie frequency once/ week 7

Controlled Shutdowns APRM High Plus j

Output Signal B

Heat Balance Daily Plow Blas Signal B

Internal Power and E mry refueling outage Flow Test with Stan-dard Pressure Snurce 1.PRM signal B

TIP System Traverse Every 1000 effective full l

power hours y

High Reactor Pressure B

Standard Pressure Once/ operating cycle 4

Source H

i High Drywell Pressure A

Standard Pressure Every 3 monthe Source A

Pressure Standard Every 3 months Reactor Iow Water 14 vel High Water level in Scram Dis-A Water Column, Hote(6)

Once/ operating cycle, Note (6) f charge Instrument Volume Main Steam Line Isolation Valve A

Hote (5)

Note (5)

Closurs j

Maln Steam Line High Pad 14 tion 8

Standard Current Ryery 3 ponths Source (3)

Turbine Plant Stage Pressure A

Btandard Pressure Every 6 montha Permissive.

Source 30 Turbina control Valve Past Closure A

Standard Pressure onc1/ operating cycle g

011 Pressure Trip Source m

a Amendment No. 42, (S 46 1

s

JAFNPP p

Table 4.1-2 (cont'd) y REACTOR PROITCTION SYSTEM (SCRAM) INSTRIMENT CALIBRATION g

MINIMUM CAI.IBRATION FREQUENCIES FOR REACTO9 PRO 1TCTION INSTRUMENT CHANNELS 6

fx k'

Instrument diannel Group (1)

Calibration (4)

Minimum Frequency (2) 1 s

S-Turbine Stop Valve Closure A

Note (5)

Note (5) gn 5S Reactor Pressure Permissive A

Standard Pressure Every 6 monthe 4

Source 4

NOTES FOR TABIE 4.1-2 1.

A description of three groups is included in the Bases of this Specification.

2.

Calibration test is not required on the part of the system that is not required to be operable, or T

is tripped, but is required prior to return to service.

w" 3.

We current source provides an instrtunent channel alignment. Calibration using a radiation source shall be made each refueling outage.

4.

Response time is not a part of the routine instrument channel test but will be checked once per operating cycle.

5 Actuation of these switches by normal means will be performed during the refueling outages.

6 Calibration shall be performed utilizing a water column or similar device to provide assurance that damage to a float or other portions of the float assembly will be detected.

N N

A-us E

Amendment No. 43 47 osg a

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j JAFNPP j

TABLF. 3.2-3 INSTRUNENTATIOtt THAT INITIATES ColfrROL EOD BIDCKS g1

- pE; Minimms No.

1 83 of Operable Total pueber of 2"

Instrument Instrument Trip Level setting Instrument Channels Action Channels Per Provided by Design 3

rrip System for noth Channel-

r 2

APRM Upscale (Flow Blased)

S 1 (0.66W+42 tim FRP 6 Inst. Channels (1) i.'

Wl2D

~

4 2

APRM Upscale (Start-up 1 12%

6 Inst. Channels (1)

Mode) 1 2

APM Downscale 12.5 indicated on scale 6 Inst. Channels (1) 1 (6) l%1 Block Monitor S 1 0.66WeK (8) 2 Inst. Channels (1) as irlow Blased) e U

1 (6)

Rod alock Monitor 12.5 indicated on scale 2 Inst. Emannels (1)

I Downscale 3

IRN Downscale (2) 12% of full scale S Inst. Channels (1)-

3 IRN Detector not in (7) 8 Inst. Channels (1)

. Start-up Position 3

IRN Upecale 186.4% of full scale 8 Inst. Channels (1) 2 (4)

SRM Detector not in (3) 4 Inst. Channels (1)

Start-up Position 5

2 (4)(5) SRM Upscale 110 counts /soo 4 Inst. Channels (1) 1 Scram Discharge Instrument i le gallone 1 Inst. Channel (9) (10) l Volume High Water level _ _

Q lO h.

HOIT.S FOR TABLE 3.2-3 1

For the Start-up and Run pcsitions of the Reactor 95)de Selector Switch, there shall be two operable or S

tripped trip systems for each function. The SRM and IRM blocks rNed not be operable in run mode, and 4

s anwonshman t No. 49 72 i

I JAf'?.>

M E

TABLE 3.2-3 (Cont'd)

INSTRUNENTATION 'IMAT INITIATES CONTROL ROD SIICKS i

HO17.S FOR TAsts 3.2-3 p 23 Frca and after the time it >.

d the APRM and RDN rod blocks need not be operable in start-up mode.

y found that the first column cannot be set for one of the two trip systems, this condition may exist for up to seven days provided that d* sting that time the operable system is functionally tested h@

immediately and daily thereafter, if this condition lasts longer than seven days, the system shall 3g be tripped. From and after the time it is found that the first column cannot be met for both trip y

systems, the systems shall be tripped.

2.

I m downscale is bypassed when it is on its lowest range.

3.

This f.anction is bypassed when the count rate is > 100 cpe.

4.

One of the four SM inputs may h) bypassed.

to 5.

His SM Function is bypassed when the IRM range switches are on range 8 or above, eV 6.

We trip is bypassed when the reactor power is 1304 7.

His function is bypassed when the Mode Switch is placed in Run, s.

a = mod alock Honitor Setting in percent of initial.

lb/hr),

f W = Imop becirculation flow in percent of rated,(rated torp recirculation flow is 34.2 x 106 K = Intercept values of 394, 40%, dit and 42% can be wed with appropriate MCPR limits from Section 3.1.B.

9.

When the reactor is subcritical and the reactor water temperature is less than 2120F, the conttol rod block is required to be operable only if any control rod in a enntrol cell containing fuel is not fully inserted.

10. When the control rod block function sesociated with scram discharge instrument volume high water levet is not operable when required to be operable, the trip system shall be tripped.

A e.n un S

AmendmentNo.jd 73 b

3 t

6 P

E JAFNPP TABIE 4.2-3 MINIMUN TEST Am CALIBRATION FREQUENCY FOR COtfrROL HOD BIDCKS AC1UATION

e. 3 Instrument (1:annel Instrument Functional Test Calibration Instrument check (9)

1) APRM - Downscale (1) (3)

Once/3 months Once/ day (O

5$

2) APHN - Upscale (1) (3)

Once/3 months once/ day 4

(2) (3)

(2)

3) IBM

-Upscale

4) IRH - Downscale (2) -(3)

(2)

5) kBM - Upscale (1) (3)

Once/3 months Once/ day l

6) RBH - Downscale (1)

(3).

Once/3 months Once/ day l

1

[

7) SRM - Upscale (2) (3)

(2)

8) SRM - Detector Not in Stattup Position III I3I I2I

9) IRN - Detector Not in Startu(* Position III I3I III

10) Scram Discharge Instrument Volume - Higli Once/ month (2)

Once/ operating Cycle (2)

N/A l

water level I

]

togle system Functional Test (4 ) (6)

Frequency

1) System Iogic Check Once/6 months NOTES See listing of notes following Table 4.2-6 for the notes referred to herein.

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5 Amendment No.7 81 I

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3.3 (cout'd)

JAFNPP 4.3 (cont'd)

BE a.

Control rode which a.

Each partiallyor fully

(

cannot be moved with withdrawn operable contrcl 2e control rod drive rod shall be exercised one g3 pressure shall be notch at least once each g3 considered inoperable.

week when operating above 30 If a partially or fully percent power.

In the event g

withdrawn control rod power operation is continuing g

drive cannot be moved with three or more inoperable with drive or scram control rods, this test shall pressure, the reactor be perfonned at least once each shall be brought to day, when operating above 30 the Cold Shutdown con-percent power.

dition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and shall not be started b.

The scram discharge volume drain unless (1) investigation and vent valves shall be verified m

has demonstrated that the open at least once per 31 days cause of the failure is (these valves may be closed inter-a not a failed control rod mittently for testing under admin-drive mechanism collet intrative control).

housing, and (2) adequate shutdown margin has been c.

A second licensed operator l

demonstrated as required shall verify the conformance by Specification 4.3.A.

to Specification 3.3.A.2.d before a rod may be bypassed If investigation demonstrates in the Rod sequence control that the cause of control

System, i

rod failure is a cracked j

collet housing, or if this d.

Once per week check status l

possibility cannot be tuled of pressure and level alarme out, the reactor shall not for each accumulator.

be started until the affected coutrol rod drive has been replaced or repaired.

Ynu sn o

Amendment No. )I 89 m

=J A

a *

+

O I

-.m s

6 s

-1

=

m D

[E b.

n e control rod e.

men it is initially determined that l

a5 directional control a control rod is incapable of normal fiU valves for inoperable insertion, an attempt to fully insert

'I control rods shall be the control rod shall be made. If disarmed electrically.

the control rod cannot be fully inserted

r fQ c.

Control rods with E ?.

scram times grsater shutdown margin test shall be made 4

than those permitted to demonstrate under this condition by that the core can be made subcritical specificatius 3.3.C.3 for any reactivity condition during are inoperable, but the remainder of the operating cycle if they can be with the analytically determined, inserted with control highest worth control rod capable of rod drive pressure withdrawal, fully withdrawn, and all they need not be other control rods capable of inser-a disarmed tion fully inserted. If Specification

[

electrically.

3.3. A.1 and 4.3. A.1 are omst, reactor startup may proceed.

d.

Control rode with a failed "rull-in" or i

"rull-out" position switch may be bypassed in the mod Sequence Control System and considered operable if the actual rod position is known. nese rode must be mved in i

9 M

A vi Amendment No.,15 89a as l

l t

JAFNFP 3.3 (cont'd) 4.3 (cont'd) 2.

We average of the scram insertion 2.

At 8-week intervals,15 percent of i5 times for the three fastest operable the operable control rod drives shall f{

control rods of all groups of four

.be scram timed above 950 psig. When-control rods in a two-by-two array ever such scram time measurements are a

y shall be no greater thans made, an evaluation shall be made to

r provide reasonable assurance that Q

Control Dod

~ Average Scram proper control rod drive performance

?.

Notch Position Insertion Time is being maintained.

4 Observed (sec) 3.

All control rods shall be determined 46 0.361 operabie once each operating cycle 38 0.977 by demonstrating the scram discharge 24 2.112 volume drain and vent valves operable 04 3.764 when the scram test initiated by placing the mode switch in the SHU1DOWN position is performed as required by T

Table 4.1-1 and by verifying that the y

drain and vent valves:

a.

Close within 80 seconds after receipt of a signal for control rods to scram, and b.

Open when the scram signal is reset or the scram discharge instrument volume trip is bypassed.

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r TER-C5506-74 ATTACHMENT II SAFETY EVALUATICN RELATED TO SCRAM DISCHARGE VOLUME POWER AUTHORITY OF THE STATE OF NEW YORK JAMES A. FITZPATRICK NUCLEAR POWER PLANT DOCKET NO. 50-333

~ JANUARY 6,

1981' nklin Research Center A Cheson of The Franda inseede

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l TER-C5506-74 4

Section I - Description of Modification The modification provides surveillance requirements for SDV vent and drain valves and I.CO/ surveillance requirements for the RPS and control rod block SDIV limit switches, in accordance with the NRC letter dated July 7,1980 to all BWR I.icensees.

Section II - Purpose of the Modification The purpose of the modification is to ensure that the SDv is operable and that the control rod drive system is operable during reactor operation.

Section III - Impact of the Change These modifications will not alter the conclusion reached in the FSAR and SER accident analysis.

Section IV - Implementation of the Modification The modification as proposed will not impact the Fire Protection Program at JAF.

Section v - Conclusion The incorporation of these modifications: a) will not increase j

the probability nor the consequences of an accident as previously evaluated in the Safety Analysis Report; b) will not increase ths possibility for an accident or malfunction of a different type r

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 VI - References c

(a) JAF FSAR (b)

JAF SER 1

$f

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ML20040G310

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