Text

BWR Scram Discharge Vol Long-Term Mods, Technical Evaluation Rept
	ML20244B663
Person / Time
Site:	Hatch
Issue date:	01/13/1982
From:	Mucha E; Franklin Research Ctr, Franklin Institute
To:	Eccleston K; Office of Nuclear Reactor Regulation
Shared Package
ML20079L375	List: ML20079L394; ML20244B663; ML20244B666;
References
	CON-NRC-03-81-130, CON-NRC-3-81-130, TAC 42218, TAC 42219 NUDOCS 8201180418, TER-C5506-78
	Download: ML20244B663 (55)
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TECHNICAL EVALUATION REPORT BWR SCRAM DISCHARGE VOLUME LONG-TERM MODIFICATIONS GEORGIA POWER COMPANY

WIN I. HATCH NUCLEAR PLANT UNIT 2 NRC DOCXETNO. 50-366 FRC PROJECT C5506 NRCTACNO. 42218 FRCASSIGNMENT 2 NRC CONTRACT NO. NRC-03-81 130 FRCTASK' 78 Preparedby Franklin Research Center Author:

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

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

K. Ecclescon January 13, 1982 i

This report was prepared as an account of work sponsored by an agency of the United States Gownment. Neither ths United States Governrnent 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 4

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 privately owned rights.

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TER-C5506-78 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

REvird CRITERIA.

5 2.1 Surveillance Requirements for SDV Drain and vent Valves 5

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

1 2.3 LOO / Surveillance Requirements for Centrol Rod l

Withdrawal Bleck SDV Limit Switche.

8 3

METHOD OF r/ALUATION 11 4

TIX:HNICAL EVALCATION 12 4.1 Surveillance Requirements for SDV Drain and Vent Valves 12 l

4.2 LCO/ Surveillance Requirements for Reactor Protection System SUV Limit Switches 14 4.3 ICO/ Surveillance Requirements for Control Rod Withdrawal Block SDV Limit Switches 16 5

CONCLUSIONS.

20 6

RE7ERENCES.

23 APPENDIX A - NPC STAFF'S MODEL TECHNICAL SPECIFICATIONS APPENDIX B - GEOPGIA POWER COMPANY LETTER OF FEBRUARY 26, 1981 AND SUBMITTAL WITH PRCPOSED TECHNICAL SPECIFICATIONS CHANGES FOR EDHIN I.

HATCH NUCLEAR PLANT UNITS 1 AND 2 APPENDIX C - GEO!GIA POWER COMPANY LETfER OF OCTOBER 1,1981 WITH ANSWER TO RFI FOR EDWIN I. HATCH NDCLEAR PLANT UNITS 1 AND 2

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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 i

I changes in the Hatch Nuclear Plant Unit 2 Technical Specifications for scram discharge volume (SDV) long-term modifications regarding surveillance

-requirements for SDV vent and drain valves and the limiting condition for I

operation (LCO)/ surveillance requirements for reactor protection system and control rod withdrawal block SDV 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 proposed placement of the SDV drain and vent valves in the tables of power-operated isolation valves (see revised pages 3/4 6-23 and 3/4 6-32 of

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the Hatch Nuclear Plant Unit 2 Technical Specifications) in order to apply isolation valve surveillance requirements to them is not acceptable. However, the Licensee's agreement to revise proposed specifications changes to require verifying each valve to be open at least once per 31 days and cycling each valve at least one complete cycle of full travel at least once per 92 days meets the NRC staff's Model Technical Specifications requirements of paragraphs 4.1.3.la and 4.1.3.lb, and is acceptable.

The remaining surveillance requirements are met by pages 3/4 1-5, 3/4 1-6, 3/4 1-7, 3/4 3-4, 3/4 3-7, 3/4 3-38, and 3/4 3-41 of the Hatch Nuclear Plant Unit 2 Technical Specifications without any revision. Table 5-1 on pages 21 and 22 of this report 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 I

evaluate the proposed changes in the Technical Specifications of the Hatch Nuclear Plant Unit 2 boiling water reactor (BWR) in regard to "BWR Scram Discharge volume Long Term Modification," specifically surveillance requirements for scram discharge volume (SDV) o vent and drain valves limiting condition for operation (LCO)/ surveillance requirements o

for the reactor protection system limit switches Iro/survelliance requirements for the control rod withdrawal o

block SDV limit switches.

The evaluation used criteria proposed by the NRC staff in Model Technical I

Specifications (see Appendix A of this report). Tr.is 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 i

the float chamber.

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On October 19, 1979, Brunswick Unit i reported that water hammer due to f

slow closure of the SUV drain valve during a reactor scram damaged several pipe J

Mpports on the SDV drain line. Drain valve closure time was approximately 5 minutes because of a faulty scienoid controlling the air supply to the valve.

Af ter repair, to avoid probable damage from a scram, the unit was started with 1

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 nklin Rese

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k' 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 cause of these level switch failures.

As a result of these events involving common-cause failures of S!nt limit switches and SDV drain valve operability, the NRC issued IE Bulletin 80-14, j

" 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 l

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 requesting that they propose Technical Specifications changes to provide surveillance require-ments for reactor protection system and control rod block SDV limit switches.

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 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 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 followed by five supplements. These initiated short-term and long-term programs described in " Generic Safety Evaluation Report BWR Scram 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 events convinced the NRC staff that SDV systems in all BWRs should be modified to assure long-tern SDV reliability. Improvements were needed in three major"-

areas: SDV-IV hydraulic ccupling, level instrumentation, and system isolation.

To achieve these objectives, in Office of Nuclear Reactor Regulation (NRR) task force and a subgroup of the BvR Owners Group developed Revised Scram Discharge System Design and Safety Critoria for use in establishing acceptable SDV systems modifications (9].

AJ so, an NRC letter dated October 1,1980 requested

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TER-C5506-78 all operating BWR licensees to reevaluate installed SDV systems and modify them j

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 isolatiion 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.

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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 1.

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 dri~s 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 criteria for an FRC technical evaluation of the submittals. In this TER, FRC has reviewed and evaluated Technical Specifications changes for the Hatch Nuclear Plant Unit 2 as proposed in letters dated February 26, 1981 and October 1,1981 (see Appendices B and C, respectively) by the Licensee, the Georgia Power Company (GPC), in regard to "BWR Scram Discharge Volume (SDV) Long-Term Modifications" and, specifically, in regard to the surveillance requirements for SDV vent and drain valves and the limiting c,ondition for operation (LCO)/ surveillance requirements for the reactor protection system and control rod withdrawal block SDV limit switches. FRC assessed the adequacy with which the GPC information documented compliance of the proposed Technical Specifications changes with the NRC staff's Model Technical Specifications.

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TER-C5506-78

2. REVIEW CRITERIA The criteria established by the NRC staff's Model Technical Specifications involving surveillance requirements of the main SDV components and instruments-tion cover three areas of concerns o surveillance 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 Specification for SDV drain and vent valves are:

"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 at least 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 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.

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.

During normal operation, the drain and vent valves stay in the open position for very long periods. A silt of particulate such as metal chips and flakes, various fibers, lint, sand, and weld slag from the water or air nklin Resea;ch Center A Oheeman of The Frerven w a

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may accumulate at moving parts of the valvas and temporarily " freeze" them. A strong breakout force may be needed to overcome this temnorary "f reeze,"

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 af ter a reactor scram which might damage the SDV piping system and cause a loss of system integrity or i

func tion.

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2.2 LCO/ SURVEILLANCE REQUIREMENTS FOR REAC'ICR PROTECTION SYSTEM SDV LIMIT l

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The paragraphs of the NRC staff's Model Technical Specifications l

pertinent to ICO/ surveillance requirements for reactor protection system SDV '

limit switches are:

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l "3.3.1 - As a minimum, the reactor protection system instrumentation channels shown in Table 3.3.1-1 shall be OPERABLE with the REAC'IOR PROTECTION SYSTEM RESPONSE TIME as shown in Table 3.3.1-2.

Table 3.3.1-1.

Reactor Protection System Instrumentation i

Applicable Minimum Operable Functional Operational Channels Per Trip Unit conditions System (a)

Action j

8.

Scram Discharge volume Water Level-High 1,2,5 (h) 2 4

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

Scram Discharge Volume Water Leve1-High NA" 1

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'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.

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 Required 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 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. Net 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 HOT SHUTDOWN 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 SUV water level-hign 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 2 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 r

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TER-C5506-78 logic for' the reactor protection system. The response time of.the reactor protection system for the functional unit of SDV 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 of SUV water level-high. Each reactor protection system instrumentation channel containing a limit switch should be shown;to be operable by the Channel Functional Test j

monthly and Channel Calibration at each refueling outage.

2.3 I4O/ SURVEILLANCE REQUIREMENTS FOR CONTROL ROD WITHDRAWAL BLOCK SDV LIMIT SWITCHES The NRC staf f's Model Technical Specifications specify the following I40/

surveillance requirements for control rod withdrawal block 'SDV limit switches:

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

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Table 3.3.6-1. Control Rod Withdrawal Block Instrumentation -

Minimum Operable

Applicable Channels Per Trip Operational Trip Function Function

Conditions Action 5.

Scram Discharge Volume a.

Water level-high 2

1, 1, 5**

62 b.

Scram trip bypassed 1

(1, 2, 5 * *)

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

- 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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TER-C5506-78 Table 3.3.6-2 control Pod Withdrawal Block Instrumentation Setpoints Trip Function Trio Setpoint Allowable Value 5.

Scram Discharge Volume a.

Water level-high To be specified NA b.

Scram trip bypassed NA NA" j

"4.3.6 - Each of the above control rod withdrawal block trip systems and instrumentation channels shall be demonstrated OPERABLE by the performance i

of the CHANNEL CHIEK, 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 Block Instrumentation Surveillance Requirements Operational Conditions Channel in Which Trip Channel Func tional Channel Surveillance Function Check Test Calibration Required 5.

Scram Discharge Volume a.

Water Level-NA Q

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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 withdrawal block instrumentation to have at least 2 operable channels containing 2 limit switches for SDV water level-high and 1 operable channel containing 1 limit switch for SUV scram trip bypassed. The technical objective of these requirements is to have at least one channel containing one limit switch available to monitor the SUV water level when the othse channel with a limit switch is being tested or undergoing maintenance.

The trip setpoint for control rod withdrawal block instrumentation monitoring i

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TER-C5506-78 SDV water level-high should be specified as indicated in 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.

Paragraph 4.3.6 and Table 4.3.6-1 require that each control rod withdrawal block inst;~ ~ mentation 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 SUV scram trip bypassed, and by Channel Calibration at each refueling outage for SDV water level-high.

The Surveillance Criteria of the BWR Owners Subgroup given in Appendix A, "Iong-Term Evaluation of Scram Discharge System," of " Generic Safety Evaluation Report BWR Scram Discharge System," writtea 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 criteria indicates that the NRC staff's Model Technical Specifications requirements, the acceptance criteria for the present TER, fully cover the BWR Owners Subgroup Surveillance Criteria 1 and 2 and partially cover Criterion 3.

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3.

METHOD OF EVALUATION The GPC submittal for the Hatch Nuclear Plant Unit 2 was evaluated in two stages, initini and final.

During the initial evaluation, only the NRC staff's Model Technical Specific.tions requirements were used to determine if t 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, LCO/ 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 Ref erences 1 through 10, pertinent sections of " Georgia Power Company Hatch Nuclear Plant Unit 2 Safety Analysis Report" and Hatch Unit 2 Technical Specifications were studied to determine the technical bases for the design of the SDV main compo-nents 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.

The initial evaluation concluded that the Licensee's submittal was responsive to the NRC request of July 7,1980, but certain information was not available. A request for additional information (RFI) was sent to GPC by the NRC on September 1,1981. Thus, this TER is based on the initial submittal and the Licensee's response to the RFI, dated October 1,1981 (see Appendix C).

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TECHNICAL EVALUATION 4.1 SURVEILLANCE REQUIREMENTS FOR SDV DRAIN AND VENT VALVES NRC STAFF'S MCDEL 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 least 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 the Hatch Nuclear Plant Unit 2 Technical Specifications pages 3/4 6-23 and 3/4 6-32 which contain Table 3.6.3-1 (Continued) with information given below.

" Table 3.6.3-1 (Continued)

Primary Containment Isolation Valves Isolation Time Valve Function and Number Valve Group (Seconds)

A.

Automatic Isolation Valves 27.

Scram Discharge Volume Vert Valves 2Cll-F010A (c) 60 2Cll-F010B (c) 60 28.

Scram Discharge Volume Drain Valve 2Cll-F011 (c) 60 C.

Other Isolation Valves 29.

Scram Discharge Volume Relief Valve 2C11-F012 (1)"

Notes:

"(c) Isolates on receipt of any scram signal (i) Pressure relief valve"

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a TER-C5506-78 In response to the RFI, the Licensee provided the following statements j

" Item 1 The mocal Technical Specifications contained in your July 7, 1980, letter placed the scram discharge volume vent and drain valves in section i

3/4.1.3.1 of the model Technical Specifications; ' Control Rod Operability.'

Item 1 of the FRC request asked for a reference to the section of the Technical Specifications where the request change is incorporated.

Our February 26, 1981, letter proposed that these valves be placed in the tables of power operated containment isolation valves instead of the

' Control Rod Operability' section. These valves do not affect control rod operability at Plant Hatch. The plant unique geometry of this system at Plant Hatch allows free communication between the scram level switches and the scram discharge volume (SDV). Thus, the level switches, not the vent and drain valves, protect the scram function, and in a sense control rod operability, by providing assurance that the SDV is empty. The vent and drain valves are important, however, insof ar as they provide a containment pressure boundary during the time that a scram is sealed-in.

For this reason we have chosen to place the valves in the tables of containment isolation valves. The surveillance requirements are therefore different than those proposed by the model Technical Specifications in order to be censistent with the requirements for other comparable containment isolation valves."

The Licensee agreed to revise proposed specifications changes to requires a.

verifying each valve to be open at least once per 31 days (valves may be closed intermittently for testing under administrative controls),

and b.

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

FRC EVALUATION The proposed placement of the SDV drain and vent valves in the tables of power-operated isolation valves (see revised pages 3/4 6-23 and 3/4 6-32 of the Hatch Nuclear Plant Unit 2 Technical Specifications) in order to apply isola-

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tion valve surveillance requirements t'o them is not acceptable. However, the Licensee's agreement to revise proposed specifications changes to require verifying each valve to be open at least once per 31 days and cycling each valve at least one complete cycle of full travel at least once per 92 days nidin Research Center A Osneson of The Frennan m

TER-C5506-78 meets the NRC staf f's Model Technical Specifications requirements of paragraphs 4.1.3.1.la and 4.1.3.1.lb, and is acceptable.

4.2 LCO/ 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 2 trip systems for the reactor protection system which automat-ically 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 Table 4.3.1.1-1 require that each reactor protection system instru-mentation 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 conditions for these requirements are Startup, Run, and Refuel.

LICENSEE RESPONSE The Licensee provided the following information in answer to the RFI:

" As indicated in our October 10, 1980, letter the scram level switches are currently covered by Technical Specifications on each unit. For Unit 1, please refer to Specifications 3.1 and 4.1, Tables 3.1-1 and 4.1-1, item 7.

For Unit 2, the appropriate reference is Specification 3/4.3.1, tables 3.3.1-1 and 4.3.1-1, item 8.

The instrument functional test frequency for Unit 1 is once every three months as initially approved by the Commission on issuance of the Unit 1 Operating License. We have not proposed to modify this specification. "

Page 3/4 3-3 of the Hatch Nuclear Plant Unit 2 Technical Specifications addresses the NRC staff's Model Technical Specifications requirements of paragraph 3.3.1 and Table 3.3.1-1, giving the following information in Table 3.3.1-1 (continued), Reactor Protection System Instrumentation, for

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TER-C5506-78 "1.

Applicable Operational conditions:

1, 2, 5 (h) 2.

Minimum Number Operable Channels Per Trip System (a): 2 3.

Action: 4 Action 4 - In operational condition 1 or 2, be in at least HOT SHUTDOWN within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

TABLE NOTATIONS:

A channel may be placed in an inoperable status for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for a.

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 3.9.11.1 or 3.9.11.2" The requirements of the NRC staff's Model Technical Specifications of paragraph 3.3.1 and Table 3.3.1-2 are covered in the Hatch Nuclear Plant Unit 2 Technical Specifications, Sections 3/4.1.3.2, 3/4.1.3.3 and 3/4.1.3.4 which specify control rod maximum scram insertion times, control rod average scram insertion times, and four control rod group scram insertion times, respectively.

Page 3/4 3-7 of the Hatch Nuclear Plant Unit 2 Technical Specifications addresses the NRC staff's Model Technical Specifications requirements of paragraph 4.3.1.1 and Table 4.3.1.1-1, providing Table 4.3.1-1, Reactor Protection System Instrumentation Surveillance Requirements, with the following information for " Functional Unit 8 Scram Discharge Volume Water Level-High":

"1.

Channel Check: NA 2.

Channel Functional Test: M (monthly) 3.

Channel Calibration (a): R(h) (each reheling) 4.

Operational Conditions in Which Surveillance Required:

1, 2, 5" Notes (f rom page 3/4 3-8) :

"a.

Neutron detectors may be excluded from channel calibration h.

Physical inspection and actuation of switches."

nklin Research Center A Opnman of The Frannen kusame

i l

l TER-C5506-78 FRC EVALUATION s

The Licensee's response to the NRC staff's Model Technical Specifications requirements of paragraph 3.3.1 and Table 3.3.1-1 is acceptable. The Hatch

. Nuclear Plant Unit 2 reactor protection system SDV water level-high instrumentation consists of 2 operable channels containing 2 limit switches s

per trip system, for a total of 4 operable channels containing 4 limit switches per 2 trip systems, making 1-out-of-2-taken-twice logic.

Although the Hatch Nuclear Plant Unit 2 Technical Specifications do not specify directly the reactor protection system SDV water level-high response time as required in the NRC staf f's Model Technical Specifications, paragraph 3.3.1 and Table 3.3.1-2, they have requirements for scram tirae tests, which include the required response time (see Sections 3/4.1. 3. 2, 3/4.1. 3. 3, and 3/4.1.3.4).

This approach is acceptable, since the reactor protection system SDV water level-high response time can be deduced from the scram time test.

The original provisions of the Hatch Nuclear Plant Unit 2 Technical Specifications given in page 3/4 3-7, Table 4.3.1-1, in regard to reactor-protection system SDV water level-high Channel Functionsi Test and Channel Calibration are acceptable. They meet the NRC staff's Model Technical Specifi-cations requirements of paragraph 4.3.1.1 and Table 4.3.1.1-1, which require t

the Channel Functional Test monthly and Channel Calibration each refueling outage.

4.3 LCO/ 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 switches for SDV water level-high, and 1 operable channel containing i 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 3.3.6-2.

! hyu FranWin Reseamh Center i

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3 TER-C5506-78 Paragraph 4.3.6 and Table 4.3.6-1 require that each control rod p

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 f

bypassed, and by Channel Calibration at each refueling outage for SDV water level-high.

LICENSEE RESPONSE The Licensee responded as follows:

"The SDV rod. block setpoint and surveillance requirements are specified in Unit 2 Technical Specification Section 3.3.5 and Tablas 3.3.5-1, 3.3.5-2 and 4.3.5-1.

In reviewing the Technical Specifications for our February 26, 1981 submitual, the absence of a comparable specification in the Unit 1 Technical Specifications was not noted. We agree that it is appropriate to specify the limits and surveillance requirements for the SDV cod block alarm switch and will propose an amendent to the Unit i license to incorporate requirements similar to those contained in our Unit 2 Specifications referenced above."

The information provided in Table 3.3.5-1, Control Rod Withdrawal Block Instrumentation, is as follows for " Trip Function 5.

Scram Discharge Volume

a. Water Level-High":

"1.-

Minimum Number of Operable Channels per Trip Function: 1 2.

Applicable Operational Conditions:

1, 2, 5 (f) "

Note "f. With any control rod withdrawn. Not applicable to control rods removed per Specification 3.9.11.1 or 3.9.11.2."

Table 3.3.5-2, control Rod Withdrawal Block Instrumentation Setpoints, contains the following information for " Trip Function 5.

Scram Discharge volume a. Water Level-High":

"1.

Trip Setpoints < 36.2 gallons 2.

Allowable Values < 36.2 gallons" The contents of Table 3.3.5-1 and 3.3.5-2 address the NRC staf f's Model Technical Specifications requirements of paragraph 3.3.6 and Table 3.3.6-1.

Table 4.3.5-1, Control Rod Withdrawal Block Instrumentation Surveillance nidin Research Center A Dhulen of The Fm busams

TER-C5506-78 Requirements, addresses the NRC staff's Model Technical Specifications requirements of paragraph 4.3.6 and Table 4.3.6-1, providing the following information for " Trip Function 5.

Scram Discharge Volume a. Water Level-High":

"1.

Channel Check: NA 2.

Channel Functional Test: Q (quarterly) 3.

Channel Calibration (a) : R (each refueling) 4.

Operational Conditions in Which Surveillance Required:

1, 2, 5(e)"

NOTES:

"a.

Neutron detectors may be excluded f rom CHANNEL CALIBRATION.

e.

With any control rod withdrawn. Not applicable to control rods i

removed per Specification 3.9.11.1 or 3.9.11.2. "

l l

FRC EVALUATION The existing Hatch Nuclear Plant Unit 2 discharge system has six level switches on the scram discharge volume (see FSAR, page 4.2-48) set at three different water levels to guard against operation of the reactor without sufficient f ree 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 of < 36.2 gallons (see page 3/43-40, Table 3.3.5-2 of the Hatch Unit 2 Technical Specifications), one level switch initiates a rod e

withdrawal block to prevent further withdrawal of any control rod. At the third (highest) level, with the setpoint of 50+6 or 50-1 gall.ons (see the Hatch Nuclear Plant Unit 2 FSAR, Table 7.2-1), 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 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 acceptacle compliance ("The present alarm and rod block instrumentation meets this criterion given adequate hydraulic coupling with

_nklin Rese_ arch._Cen.ter

_j

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TER-C5506-78 the SDV headers"). The Hatch Nuclear Plant Unit 2 has adequate hydraulic coupling between scram discharge headers and instrumented volume. Thus, the present alarm and rod block instrumentation is also acceptable.

In tho' Hatch Nuclear Plant Unit 2, " Scram Discharge Volume Scram Tripc" cannot be bypassed while the reactor is in operational conditions of startup and run (see FSAR page 7.2-10) and operational condition " refuel with more than one control rod withdrawn"'is not applicable, 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 with Table 4.3.6-1 and paragraph 4.3.6 with Table 4.3.6-1 are not applicable to the Hatch Nuclear Plant Unit 2 for " Trip' Function 5.b, SDV Scram Trip Bypassed.'

The proposed trip setpoint of < 36.2 gallons for control rod withdrawal block instrumentation channel is acceptable.

The provision of Table 4.3.5-1 for control rod withdrawal block instrumentation surveillance requirements of the Hatch Nuclear Plant Unit 2 Technical Specifications meets the NRC staff's Model Technical Specifications requirements of paragraph 4.3.6 and Table 4.3.6-1.

It prescribes the Channel Functional Test of each control rod withdrawal block instrumentation channel containing a limit switch quarterly and channel calibration each refueling for SDV water level-high.

/

. b Franklin Research Center A Chimen af The Fw w

l TER-C5506-78 5.

CONCLUSIONS

'Table 5-1 summarizes the results of the final review and evaluation of the Hatch Nuclear Plant Unit 2 proposed Phase 1 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 made:

The proposed placement of the SDV ' drain and vent valves in the tables o

of power-operated isolation valves (see revised pages 3/4 6-23 and 3/4 6-32 of the Hatch Nuclear Plant Unit 2 Technical Specifications) in order to apply isolation valve surveillance requirements to them is not acceptable. However, the Licensee's agreement to revise proposed specifications changes to require verifying each valve to be open at least once per 31 days and cycling each valve at least one complete cycle of full travel at least once per 92 days meets the NRC staff's Model Tecnnical Specifications requirements of paragraphs 4.1.3.1.la and 4.1.3.1.lb, and is acceptable.

o The remaining surveillance requirements are met by pages 3/4 1-5, 3/4 1-6, 3/4 1-7, 3/4 3-4, 3/4 3-7, 3/4 3-38, and 3/4 3-41 of the Hatch Nuclear Plant Unit 2 Technical Specifications without any revision.

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REFERENCES 1.

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

D. G. Eisenhut (NRR), letter "To All Operating Boiling Water Peactors (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" NBC, 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" NIC, Office 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 BWR" NBC, 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" NBC, 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" NIC, 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" NBC, 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 i

! 0 Franklin Research Center A Dmmen eiThe he mesmas

s.s u*. o..

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I l

APPENDIX A NRC STAFF'S MODEL TECHNICAL SPECIFICATIONS

I I

e

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

Ubh Franklin Research Center A DMeson of The Frennan m

x,

.. a TER-C5506-78 REACTIVITY CONTROL SYSTEMS LIMf71NO CONDIT70N FOR OPERATION (Cor.tinued) l ACTION (Continued) 2.

If the inocerable 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.

~

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, be in' at least H'07 SHUT 00WN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

(

SURVEILLANCE RE0UIREMENTS 4.1.3.1.1 The scram discharge volume drain and vent valves shall be demonstrated OPERABLE by:

a.

Verifying eacn 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 RWH and RSCS, all withdrawn esntrol rods not recuired to have their directional control valves disarmed electrically or hydraulically shall be demor.strated OPERABLE by moving each control rod at least one notch:

a.

At least once per 7 days, and b.

At least once Der 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 titerference.

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 os closed intermittently for testing under administrative controls.

CE-STS 3)41-4 A~ 1

_nklin Res,e_ arch _ Center l

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2 i

I TER-C5506-78 l

+

REACTIVITY CONTROL SYSTEVS CNTR*L ECD n'AXIMUM SCRAM INSEMTION TIMES i

)

~

LIMITTWG CON 0! TION FOR CPERATION 3.1. 3. 2 The etximum scra.m insertion time of each c:ntrol rod free the fully withdrawn position *.c rotch positiott (6), based on de-energization of the I

o scram pilot valve solenoids as time zero, shall not exceed (7.0) seconds.

1 APPLICASitTTY: OPERATIONAL'CONDITICHS.1 and 2.

s I

ACTION:

Vith the me.ximum scrata insertion time of one or core control rods exceeding (7.0) secones:

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

and b.

Perform the Surveillance Requitecents of Specification 4.1.3.2.e a'

{

1 east onca per 50 days vnen coeration is edntinued with three or i

more control rods witn maximum scram insertion times in excess of 1

(7.0) seconds, or i

c.

Se in at least HOT SHUT 00VN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

I i

)

SURVEILLANCE REQUIREWEHis J

4.1. 3. 2 fne maximum scram insertion time of the control rods shall be demon-strated th cugh measurement with reactor coolant ;ressure greater than or

]

equal to 850 psig and, during single control rod scram tice tes s, the control red drive pumps isolated from the ac:umulators-For all control rods prior to THERMAL POWER exceeding 4C% of RAhED 4.

THE.V.AL POWER following CORE ALTERATIONS or aftar a reactor shutdown i

that is greater than 120 days, L

b.

For specifically affected individual control rods following maintenances l

en or modification to the control rod or cent.ol rod crive system i

which could affect the scram insertion time of those specific control rods, and For 10% of the control roes, on a rotating basis, at least once per c.

120 days of operation.

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C"E-575 2/4 1-5 A

A-2 bEnklin Research Center

- A Osmeson of The Frerven kunnae

TER-C5506-78 1

2/4.3 INSRUMENTATICN 2/4.3.1 REAC~CR PCTICTION SYSTI* INSTEMENTATION l

tFI~~N3 CCNDITICH FOR OPERATION l

2.3.1 As a cini =, the react:r protection syste instru catation cheneis sa rn in Tele 3.3.1-1 shall be OPERASLI with the RIACTOR P?.2TECTION SYSTEM 7.EIPCN!" TIME u shown in Ta.ble 3.3.1-2.

A78t**A!ILITY: As shown in Table 3.3.1-1.

A_~~iC N:

Vith the nu=ce'r of OPERABLE channels less than required by the Minimu:n a.

CPE?ABLE Channels per Trip Systas requirement for one trip system, place at least one inoperable channel in the tripped c=ndition within one hour.

Vith the nu=cer of OPERA!LE channels less than required by the Minicus 0?E?AELE Channels per Trip System requirement for b:2 trip systems, place a: least one inopartsle channel in at least sne 'tric systa=" in tae tM::ed ::ndition wi*.nin one hour and take tne ACTICH required by Ta:1e 3.3.1-1.

The :r: visions of 5; edification 3.0.3 art not 1;plicable in OPERATIONAL CCNCITICH 5.

.~.7VE*LLANCI RECUIREuENTS

~ -

4.2.1.1 Each reacter pruection system instrursentation channel shall be.-..

t :nstratac CFE?.ASLI by the pa-formanca of the CF.ANNEL CH2CX, CHANNIL FUNC', CNAL 7737 tad CFANHEL CALIBRATION :;erations for the CPE?ATICNAL CMDITIGR5 and at the frequencies shown in Tule 4.3.1.1-1.

4.3.1.2 LOGIC SYSTIM FUNCTICHAL TESTS and simulated autc=atic operttTon of 2.11 enannels sna11 be perfor=ed at least once per 18 eenus.

4.3.1.3 The P.!ACTCR PROTICT*CN SY3 TEM RE37CNSE TIME ed each reacter trip fun: tion sh=wn in Table 3.3.1-2 shall be de=enstrated to be within its limit at letst nce :er 13 =cn u s.

Ea:3 test snell incluce at least one logic train su:h tas =cth logic. rains tra tested at least :ncit per 35 ::ntas tne one enannel ;er function such that all :hannels are tastec at least once every N ti=ts 'S m:ntas where N is the total n=ter of racuncant channels in a,.

4pe:ifi: sect:r Hp function.

^.7 ::.n :nanne ss are incperacle in one trip systs:2, select at letst one in ttrule enannel in that trip system to ciace in the tripsed c:ndition, t.:: :t when this wculd cause the Trip Function to =ccur.'

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T13L! 3.3.1-1 (Continued)

EEr.*~07 MOTECTION SYSTEM INSTRUMENTATION 1

ACTICN A:TI:N 1 In 07ERATICHA:. CON TION 2, be in at least NOT SHUTDOW within i

6 ho::rs.

In 0?ERAUCNA) CONDITION 5 suspend all operations involving CDP.E ALTE%TIONS" and fully insert all insertable control rods within are hoor.

ACTION 2 kek the reactor mode switch in the Shutdown position within ne tour.

A:n:N 3 8e is at le.as: STAATUP within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , acT1:N4 In 0: ERA"IC-NA:. CONDITION 1 or 2, be in at least HDT SNUTCCW within 6 neurs.

In 0?!?A ICHA' CONDITION 5, suspend all operations involving l

CORE ALTI.MTI NS" and fully insert aq inser*able control rods vitaf n ore nosr.

1;T*:N5 5e it, at least. HOT SNUT00W within 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months .

A*TI N 5 3e 1: ST/ATL'? with the main staam line isolation valves closed witAin 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or in at least NOT SHUTDOW within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

A;TI:N7 Initf a.a a rseuction in TEER."AL POWER within 15 minutes and redu:e :: =ine first stage pressure to < (250) psig, equiva' tant te T"iEM. L P::.TR less than (30)% of FATID THERMAL POWER, within 2 he:Ps.

4: TION S In 0?!?A30NA' CONDITICH 1 or 2, he in at least HDT SHU1JW within i h Urs.

1 In CFEM70NA* CONDITION 2 or 4, verify all insertable ::ntrol rods to te fully inserted wi,ttin one hour.

In 0?E?A~IOt!A*. CONDITION 5, suspend all eperations involving CORE ALTIAATIONS" and fully insert all inser*.able control rods within are hear.

A: TION 9 In CFE?AUC%'. CONDITION 1 or 2, he in at least ICT SHUTDOW within 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months .

In 07~?AUCNs? CONDITION 3 or 4, lock the reactor mode switch in t2e 5.'ut.com position within one hour. '

In C?E:ATICNR, CONDITICH 5, suspend all c;erations involving C0F.E A.i!.uTI RSS and fully insert all insart.able control rods within m e n:::r.

'ia::: revement of I?#.. S??. or special e:vable detectors, or replacement of

??." strings proviged $7/. i:stru:entation is OPE?ABLE per Specificatica 3.9.2.

!!.~5 3/4 2 4 A-5

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TER-CS506-78 TA!LE 3.3.1-1 (Continued)

FiA*' TOR PF.CTECTION SYSTEM INSTRUMEVTAUCN TABLE NOTATIONS (a)~ A channel may be placed in an inoperable status for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for requirad surveillance without placing the trip systas in the tripped i

condition provided at, least one OPERA 8LE channel in the same trip system is monitoring tha. parasater, b)

The " shorting links' shall be removed from the RPS circuitry prior to and during the time any control rod is withdrawn" and shutdown margin demonstrations perforimed 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 (11) LPRM inputs to an APRM channel.

(d) inese functions are not required to be OPERA 3LI sten the reactor pressure vessel head is un. belted or removed per Specification 3.10.1.

(e) This fur.c*. ion shall be auteratically bypassen when the reactor n ce switch is not in the Run position.

(f) This funct. ion is r.ct required to be OPERAHLI when PRIP'ARY CONTAINMENT i

INTIGRITY is no. re uired.

(g) Also actuates the standby gas treatment systas.

(h) Wi.h any centrol rod withdrawn. Not applicable to centrol rocs removed per Specification 3.S.10.1 or 3.9.10.2.

)

i i

(i) These functicas are aut:...atically bypassed vnen tv-bine first stage pressure is < (250) psig, equivalent to THEPy.AL PC'r:S less than (30)%

of FATED THETyAL PCTiER.

(j) Also actuates the EOC-APT system.

"Nc recatrec f or c:ntrol rods removed per Specification 3.9.10.1 or 3.9.10.2.

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V3 R'.NENTAT70N i

! 'a. 3. 6 C0HTP.0L R00 VITEDRAVAL BLOCX INSTRLHENTATICH

.l L*M*~ING CONOTTION FOR OPE:lATION

3. 3. 5.

The control red withdrawal block instrumentation channels shown in l

Tc'e 3.3.6-1 shall be OPER,a3LE whh their trip set;cints set consistent with

-.a values sh:wn in the Trip setpoint column of Table 3.3.5-2.

AS $t!CA8ILITY: As shown in' Table 3.3.6-1.

A T N:

With a control red withdrawal block instrumentation channel trip a.

sat;oint less conservative than the value shown in the Allowable Values column of Tasle 3.3.5-2, ceclare the channel inoperable until tne channel is restored to CPERABLE status with its trip satpoint

- adjustad consistent with the Trip 5etpoint value.

b.

Vith the nutnber of CPERABLE channett jess than requirsd by the Minimu= CPERAELI "hannels ;er Trip Fun:-icn esquirteent, take the ACT!CN requirec by Tamie 3.3.5-b The previsions of $ specification 2.0.3 are act a;;11cabla in OPERA-c.

IIONAL COHOITICN 5.

)

r.'?VE!LLANQRSOUIR9ENTS 4.3.5 fich of ne aoove required control red withdrawal block trip systams t..d instru entation enannels shall be cemenstratec :PERA3LI by the perf:r=ance

f tne CXANNEL CHECX, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION c; era-ti:ns f:r the OPERATICHAL CONDITICNS and at the frequencies snown in Ta:1e 4.3.!-1.

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.s TER-C5506-78 a

i TABLE 3.3.5-1 (Continued)

C:h*T20L P00 VITHDRAVal !LOCX !NSTDMENTATION I

ACTICJ T:.ke the ACTION required by Specification 3.1.4.3.

A:7 :N 60 A:T N 61 With the number of CPERABLE Channels a.

One less than required by the Minim:..a OPEFASLE Channels per Trip function requirement, restore the inoperable channel to CPEPABLE status within 7 days or place the in:perable channel in the tripped condition within the next hour, b.

Two or more less than required by the Minimum CPERA3LE Channels per Trip Function requirement, place at least i

one inoperable channel in the tripped condition within one hour.

C **N 52 Vith the num:er of CPERA!LE :sannels Itss than required by the Mini =um :PERA3LE Channels ;~er Trip Fun: tion requirement, place i

tne ineparable enannel in the trippec ::nditien within cae hour.

NOT*3 Wita TEP#AL PC'a'ER 1 (20)% of RATED THE??AL PCWER.

r With : ore than ene control rod with:rawn. Not a;:hcatie to control rods rem:ved par Specification 3.9.10.1 or 3.9.10.2.

a.

The REM shall be automatically bypassed wnen a ;sripheral control red is selec.ed.

3.

This function shall be automatically bydarsed if detector count rate is

> 100 :ps or the IFJi channels are on range (2) or higher.

This fun: tion shall be automatically byytssed wnen the associated IRM

nannels are on range 8 or higner.

This fun: tion shall be automatically sypassed when tne IRM channels are I

en range 3 or higner.

e.

This function shall be automatically bypassed when the IRM channels ire

n -an;e 1.

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CONTROL 203 YU*d3RWAl. !LOCX INSTILHENTATTON TURVE!LU.NCE RECUfREMEh*TS NOTES:

a.

Neutron detect:rs may be excluded fr:s CXANNEi. CALIBRATION.

h.

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

c.

Vhen making an unscheduled change fr:m OPERATIONAL CDHDITION 7 to l

OPERATIONAL CONDITICN 2, perfor= the required surveillance vithin 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after antaring OPERATIONAL CONDITICH 2.

l Wit.i Tji!RMAL POWER > (20)% of RATED THERFAL POWER.

Vith any control rod withdrawn. Not a:plicable to control rods ree. ved per Specifica;' ion 3.9.10,1 or 3.9.10.2.

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::ss i: ir. ::sse tha ::::stility of isvi:usly sesh:ed ac:!:F:s 0; malfunctions. Margins of safety are incressed by the so-ition of n;*.eM:

surveills.i:a en tr.ase vaivas *11cn ba ::.e ccettitaant iso;sti:n valves,:. sri ;

I tr. tirs pa: Loc i:110.in; a 3::53. and before tra se:Em is :sset.

Tae Slant 2sview 5:ard and the Safety.;eview Ost:. ave : view.?: te pic:: set :.inges to t'.e Tecnnical 50ecificati:ns 2nd tra :sfis, stated 4'*cv.s.

l fe tre ::::::s.:d 0..an.;es, 1nd have concludra that they do not involve an unrevisae: safatv. cuasticn.

0::::ir;1y,

.<a tht::f:rc :3.:2st ycur :: view

a.. ' 1:::: val *of ths -

p: ::se: chan;es to the r chnical be:ifi:sti:rs is shewn in tae e

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The p::cosed anin.::an: cess not involve a :: o'er issue, an envi::nmental issue or acts than one s.a f e t y issue;

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The p ro ce s ed s ? t od ?.e n t dces involve a single strety 1

issue, nteely, the a dition of scram :is:nt:ge v alu ie vent valves, drain volves, :nd p::ssure reiis f valves to the existing :tbles of centsinment isois;ien valves.

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staa. o c 2o m mmm

.. e "==.L*:=* O W C CCCXETS 50-321, 50-366 j

OPERATI?C LICDtSES CPR-57, NF'-5 Q

N..j/b. g M EDWIN I. HATCH NLCLEAR PLANT UNITS 1. 2 i

j RESPONSE TO FRA W LIN RESEARCH CENTER RE T EST CCFCERNING SCRAM OISCHARGE SYSTEM TECHNICAL SPECIFICATIONS

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CENTLEMm:

Your letter dated Sectemoe 1, D81, ccnveyed to Cecrgia Power Cemcany a request fc: acditicnal intonation from Franklin Researen Center (FRC) concerning cur Fecruary 26, 1981, sucmittal of p;ccesed mccifications to the Tecnnical Specifications regarcing the scram discnarge volume and associated instruments. The following infctmation is suppliac in response to the FCC request:

Item l_

The mcdel Tecnnical Specifications contained in you July 7,

D80, letter placed the scram disenar;e volume vent and drain valves in section 3/4.1.3.1 of tno mocel Tecnnical Specifications; " Cent:ci Roo Operability". Itam 1 of the FRC neuest asked for a refennce to tne section of the TechMc31 Specifications wnere the recuested enange is incorporated.

Our Fecruary 26, D81,1stter p cocsed that these valves be placed in the tables of power ocarated containment isciation valves instead of the

" Control Rod Oceracility" section. These valves cc not affect cent ci rod cerrability at Plant Haten. The plant unicu geometry of this systm at Plant Hatch allows frw OTmJnicat'en between the scram level switn es and the scram disena::ye volune (50V).

Thus, the level switenes, not the vent and crain valves, protect the scram function, and in a sense cont:cl red coeracility, by previcing assurance that the SDV is encty. Tre vent and crain valres are 1.w:crtant, however, insofar as they provide a cont.aiment, p.sst.re bouncary curing the time t.* rat a scraa is sealed-JA. For this ruason we have cnosen to place the valves in the tables of contairnent isolation valves.

The surveillance recuixecer are therefon diffennt than these procesed by the mecel Tecnnical.

fications in creer to be consistent with the requirements for other ca scle contalment isolation valves.

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PDA nkun Reeeerch Center A Dhenian af the Frannan enessas a

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TER-C5506-78 I

Georgia Power A 1

Director of Macinar Reactor Regulation l

U. S. M clear Regulatory Conrnission Octocer 1, 1981 Pege Two Items 2 and 3 l

As indicated in cur Octocer 10, 1980, letter the scram level switches are currently covered by Technical Soecificatiens on eacn unit. For Lnit 1, please refer to Specifications 3.1 and 4.1 Tacles 3.1 1 and 4.1-1, item 7.

For Unit *2, the accrc::riate reference is Specification 3/4 3.1. tables 3.3.1-1 and 4.3.1-1, item 8.

The instrument functional test frecuency for Unit 1 is ence every three men:ns as initially q

apprnved oy the Ccmissicn en issuance of the Unit 1 C;erating License.

1 j

We have not procesed to modify this specification, j

- Items a, 5 and 6 f

1 The SCV red clock set:cint and surveillance recuire.t.ents are scecified in Unit 2 Tecnnical Scacificatien Section 3.3.3 and Tables 3.3.5-1, 3.3.5-2 and 4.3.5-1.

In reviewing the Tecnnical Scecificatiens for cur i

Fecruary 2J,1981 sucmittal, the acsence of a ecmcaracle scecificatien in tne Unit 1 Tecnnical Specifications was net ncted, we acree tnat it is accrocriate to scecify the limits ano surveillance recuirstrents fer the cDV red bicek alarm switen and will propose an amencment to the Unit 4

1 licanse to incorporate requirements similar to asse contained in our i

Lnit 2 Specifications referenced above.

Very truly yours,

)Id. ?CNf + -w J. T. Beckham, Jr.

RDC/mc xc:

M. Manry R. F. Rogers, III b

$O $

A chinen of The Frmutn tumme

-

v t e Letter Sequence Other
TAC:42218, (Approved, Closed) TAC:42219, (Approved, Closed)
Initiation Request, Request, Request Acceptance... Administration Questions 1 September 1981 Answers Results Approval... Other: ML19327A527, ML19327A533, ML19330B521, ML19330B523, ML19331B934, ML19336A759, ML19340B179, ML19344B508, ML19347C984, ML19347C997, ML19350A630, ML20053A188, ML20080J137, ML20080P638, ML20085B165, ML20244B663, ML20244B666
MONTHYEAR1983-10-03 [Table View]

ML20244B663

Text

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