ML20064L071

Technical Evaluation Rept for BWR Scram Discharge Vol Long-Term Mods Oyster Creek Nuclear Generating Station

ML20064L071
Person / Time
Site:	Oyster Creek
Issue date:	01/27/1982
From:	Mucha E FRANKLIN INSTITUTE
To:	Eccleston K NRC
Shared Package
ML20064L075	List: ML20064L071
References
CON-NRC-03-81-130, CON-NRC-3-81-130 TER-C5506-58, NUDOCS 8202010167
Download: ML20064L071 (55)

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TECHNICAL EVALUATION REPORT BWR SCRAM DISCHARGE VOLUME LONG-TERM MODIFICATIONS i

I JERSEY CENTRAL POWER & LIGHT COMPMY OYSTER CREEK NUCLEAR GENERATING STATION NRC DOCKET NO. 50-219 FRC PROJECT C5506 NRC TAC NO.

42215 FRC ASSIGNMENT 2 NRC CONTRACT NO. NRC43-81-130 FRC TASK 58 Prepared by Franklin Research Center Author:

E. Mucha 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 January 27, 1982 This report was prepared as an account of work sponsored by an agency of the United States Government. 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, product or process disclosed in this report, or represents that its use by such third party would not infringe private!y owned rights.

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TER-C5506-58 CONTENTS Section Title Page

SUMMARY

1 1

INTRODUCTION 2

1.1 Purpose of the Technical Evaluation 2

1.2 Generic Issue Background 2

1.3 Plant-Specific Background.

4 2

REVIEW CRITERIA.

5 2.1 Surveillance Requirements for SDV Drain and Vent valves 5

2.2 LCO/ Surveillance Requirements for Reactor Protection System SDV Limit Switches 6

2.3 LCO/ Surveillance Requirements for Control Rod Withdrawal Block SDV Limit Switches 8

3 METHOD OF EVALUATION 11 4

TECHNICAL EVALUATION 12 4.1 Surveillance Requirements for SDV Drain and Vent Valves 12 4.2 I40/ Surveillance Requirements for Reactor Protection System SDV Limit Switches 13 j

4.3 ICO/ Surveillance Requirements for Control Rod Withdrawal Block SDV Limit Switches 15 5

CONCLUSIONS.

19 6

REFERENCES.

22 APPENDIX A - NRC STAFF'S MODEL TECHNICAL SPECIFICATIONS APPENDIX B - JERSEY CENTRAL POWER AND LIGHT COMPANY LETTER OF MARCH 4,1981 AND SUBMITTAL WITH PROPOSED TECHNICAL SPECIFICATIONS CHANGES FOR OYSTER l

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TER-C5506-58 FOREWORD This Technical Evaluation Report was prepared by Franklin Research Center under a contract with the U.S. Nuclear Regulatory Connaission (Office of Nuclear Reactor Regulation, Division of Operating Reactors) for technical assistance in support of NRC operating reactor licensing actions. The technical evaluation was conducted in accordance with criteria established by the NRC.

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SUMMARY

This technical evaluation report reviews and evaluates proposed Phase 1 changes in the Oyster Creek Nuclear Station Technical Specifications for scram discharge volume (SDV) long-term modifications regarding surveillance requirements for SDV vent and drain valves and the limiting condition ~for operation (LCO)/ surveillance requirements for reactor protection system and control rod withdrawal block SDV limit switches. Conclusions are 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 4.2-2, with the Licensee's agreement to incorporate a revision in the proposed specifiaations changes that requires cycling each valve at least one complete cycle of full travel at least quarterly, complies with the NRC staff's Model Technical Specifications requirements of paragraphs 4.1.3.1.la and 4.1.3.1.lb.

The proposed revisions of pages 3.1-7, 3.1-11, 3.1-12a, 4.1-6a (af ter deleting " Instrument Channel 27b"), and 4.2.2, and unrevised pages 3.2-5 and 4.1-5 meet the remaining surveillance requirements. Table 5-1 on pages 21 and 22 summarizes the evaluation results.

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1. INTRODUCTION 1.1 PURPOSE OF THE TECHNICAL EVALUATION The purpose of this technical evaluation report (TER) is to review and evaluate the proposed changes in the Technical Specifications of the Oyster Creek Nuclear Generating Station boiling water reactor (BWR) in regard to "BWR Scram Discharge Volume Long Term Modification," gecificallys o surveillance requirements for scram discharge volume (SDV) vent and drain valves limiting condition for operation (LCO)/ surveillance requirements o

for the reactor protection system ICO/ surveillance requirements for the control rod withdrawal o

block SDV limit switches The evaluation uses criteria proposed by the NRC staff in Model Technical Specifications (see Appendix A of this report). This effort is directed toward the NRC objective of increasing the reliability of installed BWR scram discharge 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 SDV high level switches had been modified, tested, and found inoperable. The 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 SDV drain valve during a reactor scram damaged several pipe supports on the SDV drain line. Drain valve closure time was approximately 5 minutes because of a faulty solenoid controlling the air supply to the valve.

Af ter repair, to avoid probable damage from a scram, the unit was started with the SDV vent and draf n valves closed except for periodic draining. During this mode of operation, the reactor scrassed due to a high water level in the SDV system without prior actuation of either the high level alarm or rod block l nklin Rese

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t TER-C5506-58 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 r

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 W R 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 NRC sent a letter dated July 7, 1980 (2] to all operating BWR licensees request.ing that they propose Technical Specifications changes to provide surveillance. nuire-ments for reactor protection system and control rod block SDV limit switt.nes.

The letter also contained the NRC staff's Model Technical Specifications to be I

used as a guide by licensees in preparing their submittals, t

Meanwhile, during a routine shutdown of the Browns Ferry Unit 3 reactor l

l on June 28,1980, 76 of 185 control rods failed to insert fully. Full inser-tion required two additional manual scrams and an automatic scram for a total elapsed time of approximately 15 minutes between the first scram. initiation ano 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 followeo by five supplements. These initiated short-term and long-term programs described in " Generic Safety Evaluation Report BWR 3 cram Discharge System," NRC Staff, December 1,1980 [9] and " Staff Report and Evaluation of Supplement 4 to IE Bulletin 80-17 (Continuous Monitoring Systems)" (10].

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

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

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'Ib achieve these objectives, an Office of Nuclear Reactor Regulation (NRR) task force and a subgroup of the BWR Owners Group developed revised scram discharge I

system design and safety criteria for use in establishing acceptable SDV systems modifications (9]. Also, an NRC letter dated October 1, 1980 requested I

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TER-C5506-58 all operating BWR licensees to reevaluate installed SDV systems and modify them as necessary to comply with the revised criteria.

In Reference 9, the SDV-IV hydraulic coupling at the Big Rock Point, Brunswick Units 1 and 2, Duane Arnold, and Hatch Units 1 and 2 BWRs was judged acceptable. The remaining BWRs will require modification to meet the revised SDV-IV hydraulic coupling criteria, and all operating BWRs may require modification to meet the revised instrumentation and isolation criteria. The changes in Technical Specifications associated with this effort will be carried out in two phases:

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

Phase 2 - Improvements required as a result of long-term modifications made to comply with revised design and performance criteria.

This TER is a review and evaluation of Technical Specifications changes proposed for Phase I.

1.3 PLANT-SPECIFIC BACKGROUND The July 7,1980 NRC letter (2] not only requested all BWR licensees to amend their facilities' Technical Specifications with respect to control rod drive SDV capability, but enclosed the NRC staff's proposed Model Technical Specifications (see Appendix A of this TER) as a guide for the licensees in preparing the requested submittals and as a source of critsria for an FRC l

technical evaluation nf the submittals. In this TER, FRC has reviewed and l

evaluated Technical Specifications changes for the Oyster Creek Nuclear i

Generating Station proposed in a by the j

Licensee, the Jersey Central Power & Light Company (JCP&L), in regard to "BWR l

Scram Discharge Volume (SDV) Long-T2rm Modifications" and, specifically, the surveillance requirements for SDV vent and drain valves and the limiting condition for operation (LCO)/ surveillance requirements for the reactor l

l protection system and control rod withdrawal block SDV limit switches. FRC assessed the adequacy with which the JCP&L information documented compliance of the proposed Technical Specifications changes witn the NRC staff's Model Technical Specifications.

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2. REVIEN CRITERIA The criteria established by the NBC staff's Model Technical Specifica-tions involving surveillance requirements of the main SDV components and instrumentation cover three areas of concern:

4 surveillance requirements for SDV drain and vent valves o

ICO/ surveillance requirements for reactor protection system SDV o

limit switches ICO/ surveillance requirements for control rod block SDV limit switches.

o 2.1 SURVEILLANCE REQUIREMENTS FOR SDV DRAIN AND VENT VALVES The surveillance criteria of the NRC staff's Model Technical Specifications for SDV drain and vent valves are:

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

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

b.

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

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

The Model Technical Specifications require testing the drain and vent valves, checking at least once in every 31 days that 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.

Full opening of each valve during normal operation indicates 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 l

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

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'i TER-C5506-58 During normal operation, the drain and vent valves stay in the open position for very long periods. A silt of particulates such as retal 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 "freeza" them. A strong breakout force may be needed to overcome this temporary freeze, producing a violent jerk which may induce a severe water hammer if it occurs during a scram or a scram resetting. Periodic cycling of the drain and vent valves is the best method to clear the 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 SDV piping system and cause a loss of system integrity or function.

2.2 LCO/ SURVEILLANCE REQUIREMENTS FOR REACTOR PROTECTION SYSTEM SDV LIMIT SWI'ICHES The paragraphs of the NBC staff's Model Technical Specifications pertinent to I4O/ surveillance requirements for reactor protection system SDV 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 REACTOR PROTECTION SYSTEM RESPONSE TIME as shown in Table 3.3.1-2.

4 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-High 1,2,5 (h) 2 4

Table 3.3.1-2.

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

Scram Discharge Volume Water Level-High NA"

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i TER-C5506-58 "4.3.1.1 - Each reactor protection system instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECE, CHANNEL FUNCTIONAL TEST, and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and at the frequencie; shown in Table 4.3.1.1-1.

Table 4.3.1.1-1.

Reactor Protection System Instrumentation Surveillance Requirements Operational Conditions Channel in Which Functional Channel Functional Channel Surveillance Unit Check Test Calibration Reouired 8.

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 <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 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 applicable 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 ROT SHUTDOWN within 6 he 1s.

In OPERATIv.3AL 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 movanle 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 two operaple channels containing two limit switches per trip system, for a total of four operable channels containing four limit switches per two trip systems for the reactor protection system which automatically initiates a scraa. The technical objective of these requirements is to provide 1-out-of-2-taken-twice

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TER-C5506-58 logic for the reactor protection system. The respense time of the reactor protection system for the functional unit of SDV water level-high should be aasured and kept available (it ia not given in Table 3.3.1-2).

I Paragraph 4.3.1.1 ar.d Table 4.3.1.1-1 give reactor protection system instrune.r.tation surveillan:e requirements for the functional unit of SDV wr.ter level-high. Each reactor prottction system instrumentation channel containing a limit switch should. be shown to be operable by the Channel Functional Test monthly and Channel Calibration at each refueling outage.

2.3 I40/ SURVEILLANCE REQUIREMENTS FOR CONTROL RCD WITHDRAWAL BLOCK SCRAM DISCHARGE VOLUME LIMIT SWITCHES The NRC staff's Model Technical Specifications specifj the following LCO/

surveillance requirements for centrol rod withdrawal block SUV limit switches:

"3.3.5 - The control rod withdrawal block instrumentation channel shown in Table 3.3.6-1 shall be OF3RABLE with trip setpoints wt consistent with the values shown in the Trip Setpoint column of Table 3.3.6-2.

Table 3.3.6-1. Control Red Withdrawal Block Instrumentation Minimum Operable Applicable Channels Per Trip Operational Trip Function Function Conditions Action 5.

Scram Discharge Vol'4'te_

a.

Water level-high 2

1, 2, 5**

62 b.

Scram trip bypassed 1

(1, 2, 5**)

62 ACTION 62: With the number of OPF.RABLE channels less than required by the ministw OPERA 8LE channels per Trip Function requirement, place the inoperable channel in the tripped condition within one hop;.

• • With more than one control rod withdrawn. Not applicauc 1ontrol rods removed per Specification 3.9,10.1 or 3.9.10.2.

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t TER-C5506-58 Table 3.3.6-2.

Control Rod Withdrawal Block Instrumentation Setpoints Trio Function Trip Setpoint Allowable Value 5.

Scram Discharge Volume a.

Nater 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 CM3ABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIci4AL TEST and CHANNEL CALIBRATION oporations for the OPERATIONAL CONDITIONS and at the frequencies shown in Tacle 4.3.6-1.

Table 4.3.6-1. Control Rod Withdrawal Block Instrumentation Surveillance Requirementt Operational Conditions Channel in Which Trip Channel Functional Channel Surveillance Function Check Test Calibration Required 5.

Scram Discharge Volume 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 rods removed per Specification 3.9.10.1 or 3.9.10.2."

Paragraph 3.3.6 and Table 3.3.6-1 of the Model Technical Specifications require the control rod whhdrawal block instrumentation to have at lease two operable channels containing two limit switches for SDV water level-high and one operable channel containing one limit switch for SDV scram trip bypassed.

The technical objective of these requirements is to have at least one channel containing ons limit switch available to monitor the SUV 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 in Table 3.3.6-2.

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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 Functional Test once per 3 months for SDV water level-high, by the Channel Functional Test 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 BNR Owners Subgroup given in Appendix A, "Iong-Tern Evaluation of Scram Discharge System," of " Generic Safety Evaluation Report MfR Scram Discharge System," 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 temperature at approximately 50% control rod density.

Analysis of the above esiteria indicates that the-NRC staff's Model Technical Specifications req' irements, the acceptance criteria for the present u

TER, fully cover the BWR Owners Subgroup Surveillance Criteria 1 and 2 and partially cover criterion 3.

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METHOD OF EVALUATION The JCP&L submittal for the Oyster Creek Nuclear Generating Station 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 ift 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 SUV drain and vent valves, 140/ surveillance requirements for reactor protection system SDV limit switches, and ICO/ surveillance requirements for control rod block SDV limit switches o the suositted 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 " Facility Description and Safety Analysis Report, Oyster Creek Power Plant Unit 1," Vols. I and II, and Oyster Creek Technical Specifications were studied to determine the technical bases for the design of SDV main components and instrumentation. Subsequently, the Licensee's response was contpared 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.

The initial evaluation concluded that the Licensee's submittal was responsive to the NBC's request of July 7, 1980 and that the submittal contained sufficient information to permit preparation of a TER without a request for additional information.

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TECHNICAL EVALUATION i

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 ara operable by:

verifying each valve to be open (valves may be closed intermitte y

a.

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 page 4.2-2 of the Oyster Creek Technical Specifications as follows (see Appendix B):

"H.

The scram discharge volume drain and vent valves shall be verified open at least once per 31 days, except in shutdown mode.*

I.

All withdrawn control rods shall be determined OPERABLE by demonstrating the scram discharge volume drain and vent valves OPERABLE. This will be done at least once per refueling cycle by placing the mode switch in shutdown and by verifying that:

a. The drain and vent valves close within 60 seconds after receipt of a signal for control rods to scram, and

b. The scram signal can be reset and the drain and vent valves open when the scram discharge volune trip is bypassed.

• These valves may be closed intermittently for testing under administrative control.

Basis: The core reactivity limitations (Specification 3.2. A) requires that core reactivity be limited such that the core could be made subcritical at any time during the operating cycle, with the strongest operable control rod fully withdrawn and all other operable rods fully inserted. Compliance with this requirement can be demonstrated conveniently only at the time of refueling."

In addition, the Licensee agreed to revise proposed specifications changes on page 4.2-2 to require cycling each valve at least one complete cycle of full travel at least quarterly.

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TER-C5506-58 FRC EVALUATION The revised page 4.2-2, with the Licensee's agreement to incorporate a revision in the proposed specifications changes that requires cycling each valve at least one complete cycle of full travel at least quarterly, complies with the NRC staff's Model Technical Specifications requirements of paragraph 4 4.1.3.la and 4.1.3.1.lb.

4.2 LCO/ SURVEILLANCE REQUIREMENTS FOR REACTOR FROTECTION 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 two operable channels containing two limit switches per trip system, for a total of four operable channels containing four limit switches per two trip systems for the reactor protection system j

which automatically initiates scram.

Parangraph 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 BNR (it is not specified in the table). Paragraph 4.3.1.1 and Table 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 by Channel Calibration at each refueling outage. The applicable operational conditions for these requirements are startup, run, and refuel.

l LICENSEE RESPONSE In response to the NRC staff's Model Technical Specifications requirements of paragraph 3.3.1 and Table 3.1.1-1, the Licensee proposed revising pages 3.1-7 and 3.1-12a of the Oyster Creek Technical Specifications. The revised page 3.1-7 contains Table 3.1.1, " Protective Instrumentation Requirements,"

with the following information for function - scram on SDV high water levels "1.

Trip setting

< 37 gal.

2.

Reactor modes in which function must be operable: _enklin Rese_ arch _ Center 1

1 TER-C5506-58 Refuel (a), Startup (z), Run (z) 3.

Min, No. of Operable or Operating (Tripped) Trip systems: 2 4.

Min. No. of Operable Instrument channels per Operable Trip Systems: 2 NOTES:

a. Permissible to bypass, with control rod block, for reactor protection system reset in refuel mode.

z. 'Ihe bypass function to permit scram reset in the shutdown or refuel mode with control rod block must be operable in this mode."

(Note z i.s taken from the revised page 3.1-12a.)

Page 3.2-5 of the Oyster Creek Technical Specifications gives the reactor protection system response time as follows:

"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 7ompared to the typical delay of about 210 mill.iseconds estimated from scram test results."

This acoresses tne requirements of paragraph 3.3.1 and Table 3.3.1-2.

In response to the requirements of paragrapn 4.3.1.1 and Table 4.3.1.1-1 the Licensee submitted the original page 4.1-5 of the Oyster Creek Technical Specifications without revision. This contained Table 4.1.1, " Minimum Check, Calibration and Test Frequency for Protective Instrumentation," with the following information regarding instrument channel SDV high water level:

"1.

Check: N/A 2.

Calibrate: 1/3 mo.

3.

Test: Note 1 4.

Remarks (Applies to Test Calibration): By varying level in switch Columns.

NOTE la Initially once/mo., thereafter according to Fig. 4.1.1, with an interval no less than one sonth nor more than three raonths."

anklin Research Center

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TER-C5506-58 FRC EVALUATION The Licensee's response to the NRC staff's Model Technical Specifications requirements of paragraph 3.3.1 is acceptable. The Oyster Creek reactor protection system SDV water level-high instrumentation consists of two operable channels containing two limit switches per trip system, for a total of four operable channels containing four limit switches per two trip systems, making 1-out-of-2-taken-twice logic. The revised page 3.1-7 with Table 3.1.1 also specifies < 37 gal as a trip setting for scram initiation and applicable operating conditions of refuel, startup, and run, which are acceptable.

The reactor protection system response time of 290 milliseconds specified on page 3.2-5 of the Oyster Creek Tect.s 'al Specifications addresses the requirements of paragraph 3.3.1 and Table 3.3.1-2 and is acceptable.

The original provisions of the Oyster Creek Technical Specifications given in Table 4.1.1, page 4.1-5 (see Appendix B), in regard to reactor protection system SDV water level-high calibration av$ test frequency for protective instrumentation are. acceptable although they differ from paragraph 4.3.1.1 and Table 4.3.1.1-1 of the NRC staff's Model Technical Specifications, which require Channel Calibration each refueling outage (provided by Oyster Creek once per 3 months) and a Channel Functional Test monthly (provided by Oyster Creek initially once per month and thereaf ter at intervals no shorter than 1 month or longer than 3 months).

4.3 LCO/ SURVEILLANCE REQUIREMENTS FOR CONTROL ROD 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 t

I block instrumentation to have at least two operable channels containing two limit switches for SDV water level-high, and one operable channel containing one 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 in Table 3.3.cs*2. ranklin Resear A cm a e n. rr.n ch Center

.m l

)

TER-C5506-58 Paragrapn 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 Functional Test once per 3 months for SDV water level-high, by the Channel Functional Test 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 and Table 3.3.6-1 requirements, the Licensee proposed revising page 3.1-11 of the Oyster Creek Technical Specifications. The revised page 3.1-11 contains Table 3.1.1, " Protective Instrumentation (Contd)" with the following information for function - rod block SDV water level-high:

"1.

Trip setting: 18 gallons 2.

Reactor Modes in Which Function Must be Operable:

Refuel (z), Startup (z), Run (z).

3.

Min. No. _of Operable or Operating (Tripped) Trip Systems 1 4.

Min. No. of Operable Instrument Channels per Operable Trip Systems 1."

[ NOTE 2: Same as in LICENSEE RESPONSE, Section 4.2 of this report.]

The Licensee responded to the requirements of paragraph 4.'3.6 and Table 4.3.6-1 with a proposed revision of page 4.1-6a of the Oyster Creek Technical Specifications which contains Table 4.1.1, " Minimum Check, Calibration and Test Frequency for Protective Instrumentation," with the following information in regard to instrument channel-SDV (rod block).

"(a)

Water level hight 1.

Calibrate: Each refueling outage 2.

Test: Every 3 months 3.

Remarks (Applies to test and calibration): By varying level in owitch column nklin Rmarch Center A Osnmen of The Fransen m

1 TER-C5506-58 (b)

Scram trip bypass:

1.

Calibrate NA 2.

Test: Each refueling outage" FRC EVALUATION The existing Oyster Creek Nuclear Generating Station scram discharge system has six level switches on the scram discharge volume (see " Facility Description and Safety Analysis Report, Oyster Creek Power Plant Unit No.1,"

Appendix B, Section 2) 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 setraint of 18 gallons (see revised page 3.1-11, Table 3.1.1), one level switch initiates a rod withdrawal block to prevent further withdrawal of any control rod. At the third (highest) level, with the setpoint of < 37 gallons (see page 3.1-7, Table 3.1-1 of the Oyster Creek 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 available to receive the scram discharge water.

Reference 9, page 50, defines Design Criterion 9 (" Instrumentation shall be i

provided to aid the operator in the detection of water accumulation in the instrumented voluse(s) prior to scram initiation"), gives the technical basis for "Long-Tern Evaluation of Scram Discharge System," and defines acceptable conpliance ("The present alarm and rod block instrumentation meets this criterion given adequate hydraulic coupling with the SDV headers"). Thus, if the Oyster Creek Nuclear Generating Station scram discharge system is modified (long term) so that the hydraulic coupling between scram discharge headers and instrumented vclume 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 revised page 3.1-11 is also acceptable. nklin Research Center A Cheesen of The Fransen m

TER-C5506-58 In the Oyster Creek Nuclear Generating Station, " Scram Discharge Volume Scram Trips" cannot be bypassed while the reactor is in operational conditions cf startup and run (see FSAR Section 7), and operational condition " refuel with more than one control rod withdrawn" is not applicabM since interlocks are provided which prevent the withdrawal of more than one control rod with the mode switch in the refuel position. Thus, the NRC staff's Model Technical Specifications requirements of paragraph 3.3.6, Table 3.3.6-1, paragraph 4.3.6, and Taole 4.3.6-1 are not applicable to the Oyster Creek Nuclear Generating Station for " Trip Function 5, Scram Discharge Volume Scram Trip Bypassed," and

" Instrumentation Channel 27b, Scram Discharge Volume (Rod Block) Scram Trip Bypass" should be deleted from revised page 4.1-6a, Table 4.1.1.

Otherwise, the proposed revision of page 4.1-6a is acceptable.

The 18-gallon trip setpoint for control rod withdrawal block instrumenta-tion is acceptable (see revised page 3.1-11 of the Oyster Creek Technical Specifications). The Licensee's proposed revision of page 4.1-6a to meet the requirements of paragraph 4.3.6 and Table 4.3.6-1 is also acceptable after deletion of." Instrument Channel 27b" since it prescribes the Channel Functional Test of each control rod withdrawal block instrumentation channel containing a limit switch once per 3 months and Channel Calibration each refueling outage for SDV water level-high. nkiin Research Center A Onesson of The Fm m

TER-C5506-58 5.

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

The revised page 4.2-2, with the Licensee's agreement to irrorporate o

a revision in the proposed specifications changes that requires cycling each valve at least one complete cycle of full travel at least quarterly, complies with the NBC staff's Model Technical Specifications requirements of paragraphs 4.1.3.1.la and 4.1.3.1.lb.

j o

" Instrument Channel 27b, SDV (Rod Block) Scram Trip Bypass" should be deleted from revised page 4.1-6a.

It is not applicable to the Oyster Creek Nuclear Generating Station.

The remaining surveillance requirements are met by revised pages o

3.1-7, 3.1-11, 3.1-12a, 4.1-6a, and 4.2-2 of the Oyster Creek Technical Specifications, and by pages 3.2-5 and 4.1-5 without revision.

l I

l

_nklin Resea_rch._ Center

EL gg[

,[

Table 5-1.

Evaluation of Phase 1 Proposed Technical Specifications Changes for Scram Discharge Volume Long'-Term Modifications gh Oyster Creek Nuclear Generating Station as2" Technical Specifications h

NRC Staff-Model Proposed by Q

Surveillance Requirements (Paragraph)

Licensee Evaluation R

I SDV DRAIN AIO VENT VALVES Verify each valve open Once per 31 days Once per 31 days Acceptable (4.1.3.1.la)

(p. 4.2-2, revised)

Cycle each valve one Once per 92 days Once per 92 days Acceptable complete cycle (4.1. 3.1. lb)

(p. 4.2-2, revised)

Sl REACTOR PROTECTION SYSTEM SDV LIMIT SWITCHES Minimum operable channels 2,

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

(pp. 3.1-7 and 3.1-12a.

revised)

SDV water level-high NA 290 ma maximum Acceptable response time (3.3.1, Table 3.3.1-2) 210 ma test (p. 3.2-5)

SDV water level-high Channel functional test Monthly First monthly, Acceptable (4.3.1.1, Table 4.3.1.1-1) then at 1-3 month l

intervals (p. 4.1-5)

\\

Channel calibration Each refueling Once per 3 months Acceptable (4.3.1.1, Table 4.3.1.1-1)

(p. 4.1-5)

.- a..... _.... _...

i<

Table 5-1 (Cont.)

.=g i

Dh Technical Specifications a 5*

NRC Staff Model Proposed by 22 Surveillance Requirements (Paragraph)

Licensee Evaluation i.

i CONTHOL ROD BIDCK SDV LIMIT SWITCHES (g

n R$

Minimum operable channels 2

per trip function SDV water level-high 2

1 Acceptable *

(3. 3.6, Table 3. 3.6-1)

(p. 3.1-11, revised)

SW scram trip bypassed 1

NA Acceptable *=

.(3.3.6, Table 3.3.6-1)

(p. 3.1-11, revised)

U 8

SOV water level-high Trip set point NA 18 gal Acceptable (3. 3.6, Table 3. 3. 6-2)

(p. 3.1-11, revised)

Channel function.al test Quarterly Quarterly Acceptable (4.3.6, Table 4.3.6-1)

(p. 4.1-6a, ** revised)

Channel calibration Each refueling Each refueling Acceptable (4. 3.6, Table 4. 3. 6-1)

(p. 4.1-6a, ** revised)

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.18 and 19 of this 'IER.

• • " Instrument channel 27b" should be deleted.

i TER-C5506-58 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 Boiling 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, 1980 4.

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

IE Bulletin 80-17, Supplement 2, " Failures Revealed by '14 sting Subsequent to Failure of Control Rods to Insert During a Scram at a BWR" NRC, Office 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, February 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 @ Nil' Franklin Resear.ch Center J

4 osa an ne Th. r,

. m

~i I

TER-C5506-58 APPENDIX A NRC STAFF'S MODEL TECHNICAL SPECIFICATIONS *

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

I 0007tankun Research Center A Om g g rm %

'l TEstH05506-58 i

REACTIVITY CONTROL SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)

ACTION (Continued) 2.

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

a)

Electrically, or b)

Hydraulically by closing the drive water and exhaust water isolation valves, i

j 3.

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

c.

With more than 8 control rods inoperable, be in' at least NOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

SURVEILLANCE REQUIREMENTS 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 RSCS, all withdrawn control rods not required to have their directional control valves disarmed electrically or hydraulically shall be demonstrated OPERABLE by naving each control rod at least one notch:

a.

At least once per 7 days, and b.

At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 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 Requirements 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.

l i

't GE-STs 3/41-4 nklin Research Center A opussen of The Frereen rename

TER-C5506-58 REACTIVIT( COCOL SYSTEM.5 CONTROL r.00.uAXIMUM SCRAM INSERTION TIMES LIM! TING CONDITION FOR CPERATION T'he etximum scras insertion time of each c:ntrol red from the fully 3.1. 3. 2 withdrawn position to notch position (6), based on de-energiration of the pilot valve solencids as time :ero, shall not exceed (7.0) seconds.

scrt:

AFPLICASILITY: OPERATIONAL'tDNDITIONS 1 and 2.

ACTION:

With the maximum scram insertion time of one or more centrol rods exceeding (7.0) seconds:

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

and b.

Perform the Surveillance Requirements of Specification 4.1.3.2.c at least once per 60 days when operation is cdntinued with three or more control ' reds with maximum scram insertion times in excess of (7.0) seconds, or c.

Se in at least HGT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

SURVEILLANCE REOUIREWENTS 4.1.3.2 T'ha maximum scram insertion time of the control rods shall be demon-strated through measurement with reactor coolant pressure greater than or

' equal to 850 ptig and, during single control rod scram time tests, the control rod drive pumps isolated from the accumulators:

For all control rods prior to THEP.".AL POWER exceeding 40% of RA[ED a.

THEW'. POWER following CORE ALTE?ATIONS or aftar a reactor shutdown that is greater than 120 days, b.

For specifically affected ind!vidual control. rods following r.aintenance on or modification to the control rod or control r::d drive system which could affect the scram insertiert '.ime of those specific control rods, and For 1C% of the control rods, on a rotating basis, at least once per c.

120 days of cperation.

CE-575 2/4 1-5 p

A-2 0

ranidin Research Center

.o

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L TER-C5506-58 1/4.3 INS ILHENTATION 2 /4. 3.1 REAC10R PROTECTION SYSTE INSTRtHENTATION L3rITIN3 C:KDITICH FOR OPERATION

2. 3.1 As a cinicu=, the react:r protection rysta= instru= ntatien chtnnels sa:<n in Tule 3.3.1-1 shall be OPERABLE with the REACTOR PE37ECTION SYSTEM P.E3P:NSE TIME as sho.n in Table 3.3.1-2.

A::LICA!ILITY: As shown in Tabit 3.3.1-1.

i C ICN:

Vith the numbe' of OPERABLE channels less than required by the Minimum a.

r SPE?ABLE Channels per Trip System requirteent for one trip systes, place at least one inoperable channel in the tripped condition within one hour.

b.

Vith the number of SPERABLE channels less than required by the Minimus SPERA!LE Channels per Trip System requirement for both trip systems, place at least ene inoperable channel in n least :ne via systam* in the ri;:ed c:nditica within one hour and take one ACTION required by Tacle 3.3.1-1.

c.

The previsions of Specification 3.0.3 are not appitcable in GPE?.ATIONAL CONDITION 5.

3 EVE!LLINCE REQUIRE"ENTS 4.3.1.1 Each reactor pr:tection system instrumentation channel shall be

as:r.s. rated CFERABLE by the perfomance of the CHANNEL CHECK, CHANNEL TUNCTICNAL TEST and CMANNEL CALIBRATION :perations for the OPE *ATIONAL l

CNDITIONS and at the frequencies shown in Tatie 4.3.1.1-1.

}

4.3.1.2 LOGIC SYST3! FUNCTIONAL TESTS and sioulated automatic operation of l

ali cnannels shall be perfor=ed at least once per 18 months.

1 4.3.1.3 The P.EACTOR PROTECT!CN SYSTEM RESPONSE TIME of each reacter trip fu..: tion sh:wn in Table 3.3.1-2 shall be demonstrated to be within its limit at least once per 1S unths. Each test shall include at least one logic train su:h that :th logic trains are tested at leest :nce per 35 ::nths and one l

chtanel per function such that all channels are tested at least once eve y N tires 13 m:nths where N is the total nu=5er of ra:'undant channels in a,.

spe:ific reect:r trip funct.fon.

. 3:.n :nanneis are in=perable in one trir systam, select at least one

^

insperable enannel in that trip system to place in the tripped c:ndition, c.:ct:t vten this w:ule cause the Trip Function to occur.'

6 IE-i I 3/A 3-1 A-3 UN Franklin Research C. enter i

a cm a w The rr ana

1 g:=

TAulE 3.3.1-1 (Continued)

",i

,i; RfAC10R l'ROTECil0N SYST[H INSTRI#1 ENTAIL 0N ah b

y=

Al'PLICAnt.E

, OPERA 8LE CilANNELS

'I MINIMUM OPEllAII0tlAL f3

{UllCTIOilAL Utili CDiful T Infl5 PER TRIP SYS1[M (a)

ACTI0li fh 8.

Scram Discharge Volume Water E

level - liigh 1,2,5(h) 2 4

3 III 0) 9.

Turbine Stop Valve - Closure I

4 7

10. Turbino Control Valve Fast Closure.

II) 0)

Trip 011 Pressure - Low I

2 7

11. Reactor Mode Switch in Shutdown 1

Position 1, 2. J 4, 5 1

8

12. llanual Scras 1,2,3,4,5 1

9 u,

l 64 N

A

.l 1

1 i

e

- -. - ~.

j a

TER-C5506-58 T1SLE 3.3.1-1 (Continued)

?.E1:~04 770TE: TION SYSTEM INSTitudNTATION ACTICN 2:TI N 1 In C?EUUC^4A:. CONDITION 2, be in at least HOT SHUTDOW within 6 he:rs.

In 0?!KA1TCNAL CONDITION 5, suspend all operations involving CORE ALTE?ATIONS" and fully insert all inserta.ble control rods within ore hour.

ACTION 2 Lock the tw.ctor mode switch in the Shutdown position within one :aur.

ACTION 3 Be I: at le.as: STARTUP within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

.JCTIIN 4 In 0: ERA IGNR CONDITION 1 or 2, be in at least HDT SHUT 00W within 6 heurs.

In 0? ERA"IONE CONDITION 5, suspend all operations involving CORE ALTIMTIONS" and fully insert all insertable control rods wi*.hin ore hoJr.

2:TI:N5 Be it, at least HOT SHUTDOW within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

A; TION 6 Se 1: STARTU? vith the rafn staan ifne isolation valves closed within 2 hcurs or in at.least NOT SHUTDA'N within 6. hours.

ATION 7-Initiate a re:uction in THER".AL r.T=TR within 25 minutes and reda:e ::-bine first stage pressure to < (250) asig, equivalent to TiEML PC.TR 1ess than (30)% of RATID THERV.AL POWER, within 2 he:rs..

4:UCN S In GPEM IONE CONDITION 1 or 2, he in at least MDT SHUTCOW 1.

within 6 hcurs.

In 0?EM IONR CONDITION 3 or 4, verify all insertable centrol rods to te fully inserted witgin ene hour.

In 0?E:.A"7DNR CONDITION 5, suspend all eperations involving CORE ALTERATIONS" and fully insert all insertabia control reds within ere haar.

A CTION 9 In GPERA IONE CONDITION 1 or 2, he in at least HOT SHUTDOW within 5 haun.

In 0?E?ATIGRE CONDITION 3 or 4, lock the reactor mode switch in tie Stu_.do.n position within one hour.

In 0?I:.CIONC CCNDITION 5, suspend all cperations involving CORE A;.TI?ATIONS* and fully insert all insartable control rods within c e n :r.

4 "i.x:::t venent of I.18.. S?.v. er special c::vable detectors, or replacement of

'.??.M s rings providec !?X i:struser.tation is CPEFAELE per Specificati:n 3.9.2.

II.C5 3/4 I-4 A-5 N

ranklin Resea

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,_rch Center

~

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i l

TER-C5506-58 T!2LE 3.3.1-1 (Continued)

REAtT:R 77.0TECT10N SYSTEM INSTRUMENTATION TABLE NOTATIONS (a) A channel may be placed in an inoperable status for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for required surveillan:e without placing the trip systes in the tripped condition provided at, least one OPERABLE channel in the same trip system is =onitoring that paraseter.

b)

The " shorting links' shall be remnved from the RPS 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 ineparable if there are less than 2 LPRM inputs per level or less tha.n (H) LPRM inputs to an APRM channel.

(d) These functions are not required to be OPERA 5LE den the reactor pressure vessel head is ur. bolted or removed per Specification 3.10.1.

(e) This function shall be automatict.11y bypassed en the reactor ecde switch is not in the Run position.

(f) This function is not required to be OPEPABLE when PRIPARY C0hTAINW.ENT INTEGRITY is not requirac.

(g) Also actuates the standby gas treatment system.

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

(f) These funct.icas r-e automatically bypassed when turbine first stage pressure is < (253) pc';, equivalent to THEFJ'AL PC'r?ER less than (30)%

of FATED THEPJdAL p7.'ER.

(j) Also actuates the EOC-RPT system.

"Not requitec for cont o1 rods removed per Specification 3.9.10.1 or 3.S.10.2.

GE-STS 3/4 3-5 A-6 h

UbJ ranklin Research Center A Opmeson of The Frarwen mesame

i IA8ll 3.3 1-/

g

.it.E..A._C. ION P.a.o_lE.CTI0tt SYS.11H RESPONSE TIMES m

>Q

i 7s RESPollSE TillE l'h Fl#fCTI0ttil IlllIT

a. :

(5ccends) fy 1.

Intermediate Range Honitors:

Jg a.

ficutron Flux - Upscale NA tg.

b.

Inoperative

• NA i

In IS 2.

Average Power Range Honitor*:

if a.

Heutron Flux - Upscale, (15)%

HA b.

Flow Diased Simulated ther. sal Power - Upscale 5 (0.09)""~

c.

Flxed Neutron Flux - Upscale, (110)%

$ (O 09) d.

Inoperative s

NA e.

LPl4H NA 3.

Reactor Yessel Steam Done Pressure - liigh

$ (0.55)

R 4.

Reactor Vessel Fater Level - Low, Level 3

$ (1.05) f, 5.

Hain steam Line Isolation Valve - Closure

$ (0.06) y' 6

Hain Steam Line Radiation - liigh NA

<a 7.

Prisary Containment s'ressure - High MA 8

Scram Discharge Volume Water Level - High NA 9.

Turbine Stop Valve - Closure 1 (0.06)

10. Turbine Control Valve Fast Closure, -

Trip 011 Pressure - Low

< (0.08)#

11.

lleactor Hode Switch in Sh.atdown Position liA

12. Hanual Scram NA

~"lleutron detectors tre exempt from response time testing. Response time shall be measured f rom the nietector culput or from the input, of the first electronic component in the channel.

(This provision is not applicalile to Construction Permits docketed af ter January 1,1970.

See Regulatory Guide 1.18, November 1977.)

3

• • Hot including simulated thermal power time constant.

!il A

fHeasured from start uf turbine control valve fast closure.

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ranklin Research Center A

A Dhenson et The Fransen insuouse

TER-C5506-58

:3TRLwiNTATION 3 't. 3. 6 CONTP.0L ROC VITWORAVAL ELOCK INSTRLHENTATION L*w.:UN3 CONDITION TOR OPEPATION

3. 3. 5.

The cente:1 r:d withdreval block instrumentation channels shom in Tcle 3.3.5-1 shall be OPEPAELE whh their trip set;cints set consistent with

.e values sh: a in the Trip 5etpoint coluca of Tcle 3.3.6-2.

A: PLICA!ILITY: As shown in' Table 3.3.5-1.

A*T* 0N:

Vith a control rod withdrawal block instrumentation channel trip a.

se peint less conservative than the value showri in the A11ewable Values column of Table 3.3.5-2, declare the channel ineparable until the channel is restored to CPERAELE status with its trip setpoint adjusted consistent with the Trip 5etpoint value.

b.

Vith the number of OPEilAELE channel.dess than required by the Minie= CPERA5LE Channels per Trip Fun:tien, requirement, take the l

ACT*CH requir.ed by Table 3.3.5-k The ;rovisions of Specificatica 3.C.3 are 3:t z;;1icable in CPERA-c.

TIONAL CONDITION 5.

i

!"RVEILLANCE REOUIREENTS

4. 3. 5 Each of the above required cor.trel r:d withdrawal block trip systems ar.: instr =entation channels shall be ce::nstratec CPE?A3LE by tne perf:r an:e

f tne CMANNEL CHECX, CHANNEL FUNCTIONAL TEST and CMANNEL CALIEPATION :pera-ti:r.s f:r the ~PERATIONAL CONDITIONS and at the frecuencias shown in Tacie 4.3.5-1.

s.,

l IE-ETs 3/4 3-50 000 Frankun Research Center A ceassen of The Fransen inseeuse

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C0lliROL 1I00 WillulRMIAL llLDCK IH51RINi[hTAIIDit on/

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I

TER-C5506-58 TAELE 3.3.5-1 (Continued)

CCh' TROL R00 VITHORAVAL BLOCX INSTT WENTATION ACTION A7::N 60 Take the ACTICN requirt1 by Specificatica 3.1.4.3.

A:T ON 61 With the nucher of CPERABLE Channels:

One less than required by the Mini =u:n OPERABLE Channels a.

per Trip function requirement, restore the inoperable channel to CPERABLE status within 7 days er place the inoperable channel in the tripped c:ndition within the next hour.

b.

Two or more less than required by the Minimum CPERA3LE Channels per Trip Fun: fon re uirement, place at least one inoperable channel in the tripped condition within one hour.

A*T::N 52 Vith the number of CPERA!LE channels itss than required by the Minicu= CPERAdLE Channels ;~er Trip Fun: tion requirement, place the in=perable channel in the tripped ::nditicn within ene hour.

NOTES

~

Vita THEPy.AL POVER 3, (20)% of RATED THE??.AL POWER.

Vith rare than ene control rod with:tawn. Not ap:licable to control rods re :ved per Specification 3.9.10.1 er 3.9.10.2.

The RIM shall b's automatically bypassed wnen a peripheral control r:d in a.

selected.

t.

This function shall be automatically byfassed if detector count rate is

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

"his fun:tica shall be automatically bp assed wnin the assoc ated IFJi

nannels are on range 8 or higher, d.

This function shall be automatically bypassed when the ITJi :hannels are

n ange 3 or higher.

This function shall be rutematically bypassed when the IPJi channeIs ire e.

in range 1.

1

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.w TER-C5506-58 TAT,'_E 4.3. 5-1 (Continued)

CONTROL ACD k'ITM3RAVAL ELOCK INSTRUMENTATION !URVE!LU.NCE REQUIREMEhis

,N*>TES:

a.

Neutr n detect:rs may be excluded fro: CHANNEL CALIERATION.

b.

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

c.

When making an unscheduled change fres OPERATIONAL CONDITION 1 to CPERATICKAL CONDITION 2, perfor= the recuired surveillance within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after entering CPERATIONAL CDNDITION 2.

Vit.i THERMAL POWER > (20)% of RATED THERMAL POWER.

Vith any control rod withdrawn. Not a;plicable to control rods removed per Specifi:ation 2.9.10.1 or 3.5.10.2.

A 01-IT3 3/4 1-55 A-14 UOO Fianklin Research Center A Dhusen af The Fransen m

O P

TER-C5506-58 APPENDIX B JERSEY CENTRAL POWER AND LIGHT COMPANY LETTER OF MARCH 4, 1981 AND SUBMITTAL WITH PROPOSED TECHNICAL SPECIFICATIONS CHANGES FOR OYSTER CREEK NUCLEAR GENERATING STATION nklin Res

.h. Cen

.,m,ea.c ter r

l

..i 1

TER-C5306-58 Jersey Central Power & Ught Company Macrson Avenue at Punchoowl Road a

Momstown New Jersey 07960 201 539-6111 b.

p#

4,.

March 4, 1981 h

\\

Director Nuclear Reactor Regulation

,A United States Nuclear Regulatory Ccanission Washington, D. C.

20SSS e

1. 'i Dear Sir.

Subject:

Oyster Creek Nuclear Generating Station Docket No. 50-219 Technical Specification Change Request No.92 In accordance with 10CFRSO.59 and 10CFRSO.90, Jersey Central Power 1 !.ight Company, owner and operator of the Oyster Creek Nuclear Generating Station, Provisional Operating I.icense No. DPR-16, requests changes to Appendix A of that license.

Pursuant to your correspordence of July 7,1980 concerning the control rod drive scram discharge volume capability, sections 3.1 4.1 and 4.2 of the Oyster Creek Teclutical Specifications shall be revised.

n o Technical Specification Change Request has been reviewed and approved by the Station Superintendent, the Plant Operations Review Committee, and an Independent Safety Review Group in accordance with Sections 6.5 of the Oyster Creek Technical Specifications.

In accordance with your correspondence of July 22, 1980 which determined that the submittal is Class III per 10 CFR 170.22, a check for 34,000 is enclosed.

Very truly yours, w'.A - [

g

' Ivan R. Fi k, p.

Vice President col la j

Enclosure

/

w/ckch:

/Yoco.oo Glos 1107ag

[

Jersey centrai power a u;nt ccmoany e a uomoer et rne cen nu puac ummes Sycem nklin Research Center A esuman of The Frarwen kwamme

.m TER-C5506-58 JERSEY CENTRAL POWER & LIGfr COMPANY OYS11R CREEK NUCLEAR GENERATING STATION Provisional Operating License No. DPR-16 Technical Specification Change Request No. 92 Docket No. 50-219 Applicant submits by this Technical Specifi trion Change Request No. 92 to the Oyster Creek Nuclear Generating Station Technical Specifications, changes to Specifications 3.1, 4.1 and 4.2.

JERSEY CE?frRAL PCWER & LIGHT COMPANY BY M

(/

vice P s1pt STATE OF NEW JERSEY

)

)

COUNTY OF MORRIS

)

day of W K P' 1981.

I Sworn and subscribed to before me this 30ws a) -

's 3

s t o a cme _t Notary Public l

l B-2 UJI' Franklin Research Center A Dhemen of The Fransen kasame

'=w

TER-C5506-58 UNITED STATES OF AMERICA NUCLEAR REGULATORY CO MISSION IN THE MATTER OF

)

)

DOCKET NO. 50-219 JERSEY CEVfRAL POWER 4 LIGff COMPANY

)

CERTIFICATE OF SERVICE This is to certify that a copy of Technical Specification Change Request No. 92 for the Oyster Creek Nuclear Generating Station Technical Specifications, filed with che United States Nuclear Regulatory Comission on March 4, 1981, has this 4th day of March,1981 been served on the Mayor of Lacey Township, Ocean County, New Jersey by deposit in the United States mail, addressed as follows:

The Honorable Henry Von Spreckelsen Mayor of Lacey Township P. O. Box 475 Forked River, New Jersey 08731 JERSEY CENTRAL POWER $ LIGir COMPANY BY M

v VicePrpideg DATED: March 4, 1981 l

l nk!!n Research Center A chamon of The Frannen inessme l

l i

TER-C5506-58 Jersey Central Power & Ught Cornoany

. I ( ':, '=~);, L r3J re = = n m e := m =..no.c

?. tor-stown New Ja",ey 079G 201 '3H111 March 4, 1981 The Honorable Henry Von Spreckelsen Mayor of Lacey Township P. O. Box 475 Forked River, New Jersey 08731

Dear Mayor Von Spreckelsen:

Enclosed herewith is one copy of Technical Specification Change Request No. 92 for the Oyster Creek Nuclear Generating Station Operating License.

This document was filed with the United States Nuclesr Regulatory Commission on March 4, 1981.

Very truly yours.

'l l}

Ivan R. Fi k

Vice Presid et la Enclosure l

810s11a 70T

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• 71 1.

The interlock is not required during the start-up test program and demonstration of plant elactrical

}

h output but shall be provided following these actions.

i_

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Not required below 40% of turbine rated steam flow.

{

k.

All four (4) drywell pressure instrument channels may be made inoperable during the integrated primary 3R containment leakage rate test (See Specification 4.5), provided that primary containment integrity is not required and that no work is performed on the reactor or its connected systems which could result in g

lowering the reactor water level to less than 4'8" above the top of the active fuel.

R I.

Bypassed in IRN Ranges 8. 9, 4 10.

m.

There is one time delay relay associated with each of two pumps.

n.

One time delay relay per pump must be operable, j

o.

Here are two time delay relays associated with each of two pumps.

tn a

p.

Two time delay relays per pump must be operable, q.

Manual initiation of affected component can be accomplished after the automatic load sequencing is completed.

Time delay starts after closing of containment spray pump circuit breaker, r.

s.

These functions not required to be operable with the reactor temperature less than 212*F and the vessel head removed or vented.

t.

These functions may be lic;urable or bypassed when corresponding portions in the same core spray system logic train are inoperable por Specification 3.4.A.

u.

These functions are not required to be operable when primary containment integrity is not required to be maintained, v.

1hese functions not required to be operable winen the AlWi is not required to be operable, 1hese functions must he operable only when irradiated finel is in the fuel pool or reactor ressel b

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ami second.ary containment integrity is required por specification 3.5.B.

IS o

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The number of operahic channels m.ny be reduced to 2 per Specification 3.9-li aml 12

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% e hygwiss function to permit scram reset in the shutdown or re[uel mi.Je k.th control rod hlock must he operahic in this male.

Amendment No. 44 4

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EEL TABl.E 4.1.1 D 71 NINIMM OIECK, CALIBRATION AND TliST FREQtaENCY FOR PRGIELTIVE INSTRINiiNTATION as px Instrument Channel Clacek Calibrate Test Remarks (Applies to Test and Calibration) k

!t 1.

liigh Reactor Pressure NA I/3 mo.

Now !

By application of test pressure Srln 2.

liigh Drywell Pressure 3@3 (scram)

NA t/3 mo.

Noto !

i 3.

Iow Reactor Mater I.evel I/d I/3 me.

Noto !

4.

Iow-law Water Level I/d 1/3 mo.

Note i 5.

liigh Water Level in Scram Discharge" Volume (Scram) NA I/3 mo.

Note I By varying level in switch columns 6.

Inw-law-law Water Level NA I/3 mo.

Note 1 By application of test pressure 7.

liigh Flow in Main Steamline I/d 1/3 mo.

Note i 8.

Inw Pressure in niin NA I/3 mo.

Note !

Steamlino 9.

liigh drywell Pressure I/d Note !

(Core Cooling) 10.

Main Steam Isolation NA NA I/3 mo.

By exercising valve Valve (Scram)

II.

APRM Level NA I/3d NA Output adjustment using operational type heat balance during power operation Milli I: Initially once/mo., thereafter according to Fig. 4.8.I, with an interval not less than one month nor more than three months.

mrtli 2: At least daily during reactor power operation, the reactor neutron flux pealing factor shall be estimated 3,

h and the flow-s eferenced APHM scram aski rami block settings shall lee adjusted, if necessary, as specified un in Section 2.3, Specificat ions (I) (a) anal (J) (a).

E un.

4.1-6a.

a=,

iEl as Instrument Channel Check Calibrate Test Remarks (Applies to Test 4 Calibration) 4 l

Th 19.

Manual Scram Buttons NA NA I/3 mo l$

n 20.

liigh Temperature Main NA Each refuel-Each refuel-Using heat source box

[5 Steamline Tunnel ing outage ing outage 21.

.SRN Using built-in calibration equipment 22.

Isolation Condenser High NA I/3 no 1/3 no By application of test pressure Flow P (Steam and Water) 23.

Turbine Trip Scram NA Every a

3 months o

24.

Generator Load Rejection NA Every Every Scram 3 months 3 months 25.

Recirculation loop Flow NA Each Refuel-NA By application of test pressure ing Outage 26.

Iow Reactor Pressure NA Every Every By application of test pressure l

Core Spray Valve 3 months 3 months Permissive 27.

Scram Discharge Volume (Raj Block) a) Water level high NA Each Refuel-Every 3 By varying level in switch column.

Ing Dutogo months H

N hl Scram t rip bypass NA NA Each refuel-4 Ing outage ui mo i

• Calibrate prior to starteip and normal shutdown and thereafter check I/s and test 1/wk until no longer respaired.

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All :.1:ndr:un ::nte:1 r:ds shall ce de:cr.ir.ed "T RAC"_F by

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7he scraa signai can be reset and the drain and vent valves open when the scram discharge volume trip is bypassed.

'The:: v:.1vas ::y be closed intar ittantly for cstin; ur. der ad-ini:tr-tivc ::ntr:'.

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• .e cer ec::1vi y licitation C";:cific: tion

.2.A) requires

..:.t : ort retic' ivity be 11.it at su:!. Ont.t the core could be :.:;e ru:.:riti:r*. a:. ny ti ic durir.g tt.e 0;:crittin. Cycis, wi.h the stro::;es:

2;sr.le c:strol r=

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the specified limits, provide the require /*

[f*i rotection. In the analytical creatment of the transients milliseconos are allowed between a neutron sensor reachfag The scram point and

?d the start of motion of the control rods.

This is adequate and conservative when gare4 co tae. typical time delay of about 210 mil.

seconds estimated from scram test results.

Approximately the first 90 millisecc:

of eac f-these time intervals result from the s'ensor and circuit de' lays; then the pil scram solenoid de-energises.

i Approximately 120 milliseconds later, the controi rod motion la estimat to actually begin. However, 200 milliseconds is conservatiyaly.a g

assumed for this e me interval in the transient _=a_=1yses and this is also included in c.

allowable scram i to. specirled limics provide ~suftscient scram capao111cy to N-insertion times of 3 accommodate failure to scram of any one operable rod.

f This specification of,3-2-3J ailure is in addition to any inoperable rods that exist in the core, provided that those inoperable rods met the core reactivity Tpecification 3.2.A.

Control rods (8) which cannot be moved with control rod drive pressure are clearly indicative of an abr.ormal operating condition on the affected rods and are, therefore, considered to be inoperable.

Inoperable rods are valved out of service to fix their position in the core and assure predictable behavior. If the. rod is fully inserted and then valved out of service, it is in a safe position of unwimum contribution to shutdown reactivity. If it is valved out of service in a non-fully inserted position, that position is required to be consistent with the shutdown reactivity limitation stated in Specification 3.2.A. which assures the core can be shu.tdown at all times with control rods.

Although there are many possible patterns of inoperable control rods which would mast this specification, the operator will be p'rovided with,only a limited aumber of predetermined patterns which allow him to continue operation with inoperable rods.

The availability of allowable pactarris to the operator assures that l

l information for determining comp'11ance with the specification is immediately available to him at the time a control rod becomes inoperable and does not require reliance on calculations at that time before compliance can'be determined.

The allowable ' inoperable rod patterns will be determined using information obtained in the startup tesc program supplemented by calculations. During initial startup, the reactivity condition

~

of the as-built core will be determined. Also, sub-critical patterns of widely separated with' drawn control rods will be observed in the control rod sequences being used.

The 'bserva-o tions, together with calculated strengths of the strongest control rods in these patterns will comprise a set of allow-able separations of malfunctioning rods. During the fuel cycle,

,similar observations mado during any cold shutdown can be used to update and/or increase the allowable patterns.

The number of rods permitted to be valved out of serv,1(e' could

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