Text

BWR Scram Discharge Vol Long-Term Mods,Ga Power Co,Hatch Nuclear Plant,Unit 1, Technical Evaluation Rept
	ML20244B666
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 IEB-80-14, NUDOCS 8201180455, TER-C5506-73
	Download: ML20244B666 (55)
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e TECHNICAL EVALUATION REPORT BWR SCRAM DISCHARGE VOLUME LONG-TERM MODIFICATIONS GEORGIA POWER COMPANY EDWIN I. HATCH NUCLEAR PLANT UNIT 1 NRC DOCKET NO. 50-321 FRC PROJECT N N RC TAC NO. 42219 FRC ASSIGNMENT 2 NRC CONTRACT NO. N RC43-81 130 FRC TASK 73 Prepared by Franklin Research Center Author E. Mucha

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

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Prepared for Nuclear Regulatory Commission Washington, D.C. 20555 Lead NRC Engineer: K. Eccleston I

January 13, 1982 This report was prepared as an account of work soonsored by an agency of tne United States Govemment. Neither the United States Government not any agency thereof, or any of their employees, maxes any warranty, ex-pressed or imolied, or assumes any legal liacility or responsibility for any third party's use, or the results of such use, of any information, apparatus, product or process disclosed in this reocrt, or represents that its use by such third party would not infringe privately owned rights.

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TER#.,5506-73 CONTENTS Section Title Page

SUMMARY

1 1

INTRODUCTION 2

1.1 Purpose of the Technical Evaluation 2

1.2 Ceneric Issue Background 2

1.3 Plant-Specific Background.

4 2

REVIEW CRITERIA.

6 2.1 Surveillance Requirements for SDV Drain and. Vent Valves 6

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

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

3 METHOD OF EVALUATION 12 4

TECHNICAL EVALUATION 13 4.1 Surveillance Req irements for SDV Drain and Vent Valves 13 4.2 ICO/ Surveillance Requirements for Reactor Protection System SDV Limit Switches 15 4.3 LCC/ Surveillance Requirements for Control Rod Withdrawal Block SDV Limit Switches 18 5

CONCLUSIONS.

22 6.

REFERENCES.

25 APPENDIX A - NRC STAFF'S MCDEL TECHNICAL SPECIFICATIONS APPENDIX B - GEORGIA POWER COMPANY LETTER OF FEBRUARY 26, 1981,AND SUBMITTAL WITH PROPOSED TECHNICAL SPECIFICATIONS CHANGES FOR EDWIN I. HAICH NUCLEAR PLANT UNITS 1 AND 2 l

APPENDIX C - GEORGIA POWER COMPANY LETTER OF OCTOBER 1, 1981 WITH RESPONSE IO RF2 FOR EDWIN I. HATCD NUCLEAR PLANT UNITS 1 AND 2 I.

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

the NRC.

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SUMMARY

This technical evaluation report reviews and evaluates proposed Phase 1 changes in the Edwin I. Hatch Nuclear Plant Unit 1 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 were based on the degree of compliance of the Licensee's submitta'l 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 3.7-18a and 3.7-20 of the Hatch Nuclear Plant Unit 1 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 tne NRC staff's Model Technical Specifications requirements of paragraphs 4.1.3.1.la and 4.1.3.1.lb and is acceptable.

The Licensee's agreement to revise the original provisions of the Hatch Nuclear Plant Unit 1 Technical Specifications given on page 3.1-8, Table 4.1-1 in regard to performing the reactor protection system SDV water level-high Channel Functional Test monthly instead once per 3 months is acceptable. It meets the NRC staff's Model Technical Specifications requirements of paragraph

4. 3.1.1 and Table 4. 3.1.1-1.

The proposed amendment of the Hatch Nuclear Plant Unit 1 Technical Specifications to incorporate surveillance requirements for control rod withdrawal block SUV limit switches similar to those contained in the Unit 2 Technical Specifications Tables 3.3.5-1, 3.3.5-2, and 4.3.5-1 is acceptable.

I Table 5-1 on pages 22 and 23 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 Hatch Nuclear Plant Unit 1 boiling water reactor (BWR) in regard to "BWR Scram Discharge Volume long Term Modification," specifically:

surveillance requirements for scram dis' charge volume (SDV) o vent and drain valves o limiting condition for operation (LCO)/ surveillance seguirements for the reactor protection system limit switches i

o LCO/ surveillance requirements for the control rod withdrawal block SDV limit switches.

The evaluation used criteria propoJed by the NRC staff in Model Technical Specifications (see Appendix A of this report). This effort is directed toward the NRC cojective 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 tne float chamber.

On October 19, 1979, Brunswick Unit i reported that water haanner 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 f aulty 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 drain valves closed except for periodic draining. During this mode of operation, the reactor scrammed due to a high water level in the I b1 kuu Franklin Research Center i

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TER-C5506-73 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 SDV limit switches and SDV drain valve operability, the NRC issued IE Bulletin 80-14,

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

In addition, to strengthen the provisions of this'hulletin 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 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 f ailed 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 j

and the complete insertion of all the rods. On July 3, 1980, in response to both this event and the previous events at Eatch 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 Systemst " (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-term SDV reliability. Improvements were needed in three major areas: SUV-IV hydraulic coupling, level instrumentation, and system isolation.

To achieve these objectives, an Office of Nuclear Reactor Regulation (NRR) task I

force and a subgroup of the BWR Owners Group developed Revised Scram Discharge i

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I TER-C5506-73 System Design and Safety Criteria for use in establishing acceptable SDV systems modifications [9]. Also, an NRC letter dated October 1, 1980 requested all operating BWR licensees to reevaluate installed SDV systems and modify them i

j as necescary 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 modifi-cation to meet the revised instrumentation and isolation criteria. The changes in Tecnnical 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.

1 Phase 2 - Improvements required as a result of long-term 1

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 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 criteria for an FRC technical evaluation of the submittals. In this TER, FRC has reviewed and evaluated the Technical Specifications changes for the Hatch Nuclear Plant Unit 1 as proposed in letters dated February 26 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, the surveillance requirements for SDV vent and drain valves and the limiting condition for operation (LCO)/ surveillance requirements for

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TER-C5506-73 the reactor protection system and control rod withdrawal block SDV limit switches. FRC assessed the adequacy with which the GPC information documented

,contpliance of the proposed Technical-Specifications changes with the NRC staff's Model Technical Specifications.

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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 concern o surveillance requirements for SDV vent and drain valves o LCO/ surveillance requirements for reactor protection system SDV limit switches i

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

l 2.1 SURVEILLANCE REQUIREMENTS POR SDV DRAIN AND VENT VALVES The surveillance criteria of the NRC staff's Model Technical Specifica-tions for SDV drain and vent valves ares l

"4.1.3.1.1 - The scram discharge volume drain and vent valves shall be demonstrated CPERABLE 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.

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

The Model Technical Specifications ' equire testing the drain and vent r

valves, checking at least once in every 31 days that each valve is fully open curing normal oper'ation, and cycling each valve at least one complete cycle of i

1 full travel under administrative controls at least once per 92 days.

Full 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 4he 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 i

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l TER-C5506-73 and flakes, various fibers, lint, sand, and weld slag from the water or air may accumulate at moving parts of the valves and temporarily " freeze" them. A I

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 i

p'romoting smooth opening and closing and more reliable valve operation. Also, l

in case of improper valve operation, cycling can indicate whether excessive pressure transients may be generated during and aft'er a reactor scram which l

might damage the SDV piping system and cause a loss of system integrity or function.

2.2 LCO/ SURVEILLANCE REQUIREMENTS EVR REACTOR PROTECTION SYSTEM SDV LIMIT SWITCHES The paragraphs of the NRC staff's Model Technical Specifications pertinent to LCO/ 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.

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 Time l

Unit (Seconds) i 8.

Scram Discharge Volume Water Level-High NA"

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TER-C5506-73 "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 frequencies shown in Table 4.3.1.1-1.

Table 4.3.1.1-1.

Reactor Protection System Instrumentation Surveillance Requirements Operational Conditions in Which j

Channel Functional Chaniel Functional Channel Surveillance Unit Check Test Calibration Required B.

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 red 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 HOT SHUTDOWN within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

l In OPERATIONAL CONDITION 5, suspend all operations involving CORE ALTERATIONS

and fully insert all insertable control rods within one hour.

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

Paragraph 3.3.1 and Table 3.3.1-1 of the Model Technical Specifications require the functional unit of SDV water level-high to have at least 2 operable channels containing 2 limit switches per trip system, for 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 I

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i 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 l

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 SDV water level-high. Eacn :eactor protection system instrumentation channel containing a limit switch shed.d be shown to be operable by the Channel Functional Test monthly and Channel ca.ibration at each refueling'qutage.

2.3 LCO/ SURVEILLANCE REQUIRDiENTS FOR CCNGOL ROD WITHDRAWAL BICCK SDV LIMIT SWITCHES The NRC staff's Model Technical Specifications specify the following LCO/

surveillance requirements for control rod withdrawal block SDV limit switches:

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

Taole 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

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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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Control Rod Withdrawal Block Instrumentation Setpoints l

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Scram Discharce Volume a.

Water level-high To be specified NA b.

Scram trip bypassed NA NA" l

"4.3.6 - Each of the above control rod withdrawal block trip systems and l

instrumentation channels shall be demonstrated) OPERABLE by the l

performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.6-1.

Table 4.3.6-1. Control Rod Withdrawal Block Instrumentation Surveillance Requirements Operational Conditions Channel in Which Trip Channel Functional Channel Surveillance Function Check Test Calibration Reauired 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 i limit switch for SDV scram trip bypassed. The technical objective of these requirements is to have at least one channel containing one limit switch available to monitor the SDV water level when the other channel with a limit switch is being tested or undergoing maintenance.

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The

. trip function prevents further withdrawal of any control ro8 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 instrumentation channel containing a limit switch be shown to be operable by the Channel Functional Test once per 3 months for SUV 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 SD7 water level-high.

The Surveillance Criteria of the BWR Owners Subgroup given in Appendix A, "Long-Term Evaluation of Scram Discharge System," of " Generic Saf ety Evaluation

. Report BWR Scram Discharge System," written by the NitC sta'f f and issued on December 1, 1980, are:

1.

Vent and drain valves shall be periodically tested.

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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 N14C 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 I

partially cover Criterion 3.

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METHOD OF EVALUATION The GPC submittal for the Hatch Suclear Plant Unit 1 was evaluated in two stages, initial and final.

During the initial evaluation, only the NRC staff's Model Technical Specifications requirements were used to determine if t o the Licensee's submittal was responsive to the July 7, 1980 NRC request for proposed Techt.ical Specifications changes involving the surveillance requirements of the SDV vent and drain valves, g

LCO/ surveillance requirements for reactor protection system SDV limit 1,

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

limit switches I

o the submitted information was sufficient to permit a detailed l

technical evaluation.

During the finel evaluation, in addition to the NRC staff's Model Technical Specifications requirements, background material in References 1 through 10, pertinent sections of " Georgia Power Company Edwin I. Hatch Nuclear Plant Unit 1 Safety Analysis Report," and Hatch Nuclear Plant Unit 1 Technical Specifications were studied to determine the technical bases for the design of SDV main components 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's request of July 7,1980, but did not contain sufficient information to permit preparation of a TER.

A request for additional information (RFI) was sent to GPC by the NRC on September 1,1981.

Thus, this TER is based on the Licenses's initial submittal (see Appendix B) 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 9 e..

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),

and b.

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

LICENSEE RESPONSE The Licensee proposed to revise pages 3.7-18a and 3.7-20 of the Hatch Nuclear Plant Unit 1 Technical Specifications. The revised page 3.7-18a cenrains Table 3.7-1 with the information given below.

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" Table 3.7-1.

Primary Containment Isolation Valves l

I Isolation Number of Power Maximum Normal Action On Group Valve Oeerated Valves Operating Position Initiating (q)

Identification Inside outside Time (sec)

(a)

Signal (a)_

(f)

Scram Discharge 2

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GC Vent Valves (Cil-F010 A, Cll-F010B)

(f)

Scram Discharge 1

60 0

GC Drain Valve (Cll-F0ll)

(g)

Valve Number of Valves (g)

Normal (g)

Identification Inside outside Position Scram Discharge 1

C" Volume Relief Valve (Cll-F012)

From revised page 3.7-20:

" Notes to Table 3.7-1 (Concluded)

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TER-C5506-73 In response to the RFI, the Licensee provided the following statements j

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1 The model Technical Specifications contained in your July 7,1980, letter placed the scram discharge volume vent and drain valves in section 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 requested 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 Eatch. The plant unique geometry of this system at Plant Eatch 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, insofar as they provide a con-tainment pressure boundary during the time that a scram is sealed-in.

Forthis reason we have enosen to place the valves in the tables of con-tainment isolation valves. The surveillance requirements are therefore different tnan those proposed by the model Technical Specifications in order to be consistent with the requirements for other comparable con-tainment isolation valves."

The Licensee agreed to revise the 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 3.7-18a and 3.7-20 of the Eatch Nuclear Plant Unit 1 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 rankHn Research Center A Csamen of The hasuen sumsee

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meets the NRC staff's Model Technical Specifications requirements of paragraphs 4.1.3.1.la and 4.1.3.1.lb and is acceptable.

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4.2 LCO/ SURVEILLANCE REQUIREMENTS FOR REAC'[OR PROTECTION SYSTEM SDV LIMIT l

SWITCHES l

NRC STAFF'S MODEL TECHNICAL SPECIFICATIONS Paragraph 3.3.1 and Table 3.3.1-1 require the, functional unit of SDV l

water level-high to have at least 2 operable channels containing 2 limit switches per trip system, for a total of 4 operable channels containing 4 limit switches per 2 trip systems for the reactor protection system which automatically initiates scram.

Paragraph 3.3.1 and Table 3.3.1-2 concern the response time of the reactor protection system for the functional unit of SDV water level-high which snould be specified~for each EWR (it is not specified in tne table). Paragraph

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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, taoles 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 3.1-4 of the Hatch Nuclear Plant Unit 1 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 nklin Research Center A Onnman of The huusn summe

j TIR-C5506-73 3.1-1 (Cont.), Reactor Protection System (RPS) Instrumentation Requirements, for " Source of Scram Trip Signal Scram Discharge Volume High High Level":

"1.

Scram Number (a): 7 2.

Operable Channels Required Per Trip System (b): 2 3.

Operacle Trip Setting:

< 71 gallons 4.

Source of Scram Signal is Required to be operable Except as Indicated Below: Permissible to bypass (initiates control rod block) in order to reset RPS when the Mode Switch is in the Refuel or Shutdown position."

Notes for Table 3.1-1 "a.

The column entitled ' Scram Number' is for convenience so that a one-to-one relationship can be established between items in Table 3.1-1 and items in Table 4.1-1.

b.

There shall be two operable or tripped trip systems for each potential scram signal. If the number of operable channels cannot be met for one of the trip systems, that trip system shall be tripped. However, one trip signal channel of a trip system may Le inoperable for up to two (2) hours during periods of required surveillance testing without tripping the associated trip system, provided that the other remaining channel (s) monitoring that parameter within that trip system is (are) operable."

Page 3.3-12 of the Hatch Nuclear Plant Unit 1 Technical Specifications gives the reactor protection system response time as follows:

"In the analytical treatment of the transients, 390 milliseconds are allowed between a neutron sensor reaching the scram point and start of negative reactivity insertion. This is adequate and conservative when compared to the typically observe time delay of about 270 milliseconds.

Approximately 70 mi111 seconds af ter neutron flux reaches the trip point, the pilot scram valve solenoid power supply voltage goes to zer'o and approximately 200 milliseconds later control rod motion begins. The 200 milliseconds are included in the allowable scram insertion times specified in specification 3.3.C."

This covers the requirements of paragraph 3.3 1,and Table 3.3.1-2 of the NRC sta'ff 's Model Technical Specification.s.

Pages 3.1-7, 3.1-8, and 3.1-9 of the Hatch Nuclear Power Plant Unit 1 Technical Specifications address the NRC staff's Model Technical Specifications requirements of paragraph 4.3.1.1 and Table 4.3.1.1-1, providing Table 4.1-1, Raactor Protection System (RPS) Instrumentation Functional Test, Functional M1 Frankun Resear 4 cm n at n. r c.h Ce.nter

-.. ~

s TER-C5506-73 Test Minimum Frequency, and Calibration Minimum Frequency, with the following information for " Source of Scram Trip signal Scram Discharge Volume High High Level":

"1.

Scram Number (a): 7 2.

Group (b): A 3.' Instrument Functional Test Minimum Frequency (c): Every 3 months 4.

Instrument Calibration Minimum Frequencyt, (h)"

Notes for Table 4.1-1:

"a.

See Note a for Table 3.1-1.

b.

The definition of... Group As On-off sensors ttat provide a scram trip signal.

c.

Functional tests are not required when the systems are not required to be operaole or are tripped. However, if functional tests are missed they shall be performed prior to returning the system to an 1

operaole status.

h.

Physical inspection and actuation of these position swicches will be performed once per operating cycle. "

i The Licensee agreed to revise the origin'ai provisions of the Hatch Nuclear Plant Unit 1 Technical Specifications given on page 3.1-8, Table 4.1-1, in regard to performing the reactor protection system SDV water level-nigh Channel Functional Test monthly instead once per 3 months.

FRC EVALUATION The Licensee's response to the NRC staff's Model Technical Specifications l

requirements of paragraph 3.3.1 and Table 3.3.1-1 is acceptable. The Hatch

./

Nuclear riant Unit 1 stractor protection system SDV water level-high instrumen-tation consists of 2 operable channels containing 2 limit switches 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. The original page 3.1-4 with Table 3.1-1 also specifies < 71 gal as a trip setting for scram initia-tion, which is acceptable.

. NO Franklin Resear.ch Ce.nter 4 a or n. n.

1

TER-C5506-73 The reactor protection system response time of 390 milliseconds specified j

on original page 3.3-12 of the Hatch Nuclear, Plant Unit i Technical Specifica-tions addresses the requirements of paragraph 3.3.1 and Table 3.3.1-2 and is t

acceptable.

The Licensee's agreement to revise the original provisions of the Hatch Nuclear Plant Unit 1 Technical Specifications given in page 3.1-8, Table 4.1-1, in regard to performing the reactor protection system SDV water 4

level-high. Channel Functional Test monthly instead.of once per 3 months meets the NRC staff's Model Technical Specifications requirements of paragraph 4.3.1.1 and Table 4.3.1.1-1 and is acceptable.

9 4.3 LCD/ SURVEILLANCE REQUIREMENTS IUR CONTROL RCD WITHDRAWAL BIDCK SDV LIMIT SWITCHES NRC STAFF'S MCDEL TECHNICAL SPECIFICATIONS Paragraph 3.3.6 and Table 3.3.6-1 require the control rod withdrawal block instrumen:Ttion to have at least 2 operable channels containing 2 limit switches for SD7 water level-high, and 1 operable channel containing 1 limit switch for SDV trip bypassed. Paragraph 3.3.6 also requires specifying the trip setpoint for control rod withdrawal block instrumentation monitoring SDV water level-high as indicated in Table 3.3.6-2.

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.

LICENSEE RESPONSE The Licensee responded to the RFI as follows:

"The SDV red block setpoint and surveillance requirements are specified i

in Unit 2 Technical Specification Section 3.3.5 and Tables 3.3.5-1, 4

3.3.5-2 and 4.3.5-1.

In reviewing the Technical Specifications for our February 26, 1981 submittal, the absence of a comparable specification in the Unit i Technical Specifications was not noted. We agreee that it is appropriate to specify the limits and surveillance requirements for the l

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TER-C550 6-73 SDV rod block alarm switch and will propose an amendment to the Unit 1 license to incorporate requirements similar to those contained in our Unit 2 Specifications refesenced above."

As seen from the above statement, the present Batch Nuclear Plant Unit 1 Technical Specifications do not contain any LCD/ surveillance requirements for l

control rod withdrawal block SDV limit switches. 1he Licensee preposes to I

amend 2 e Hatch Nuclear Plant Unit 1 Technical Specificatic.m to incorporate surveillance requirements similar to those contained in the Unit 2 Technical Specifications given in Section 3.3.5 and Tables 3.3.5-1, 3.3.5-2, and 4.3.5-1.

The information contained in these tables will be evaluated in this TER for the Hatch Nuc' ear Plant Unit 1.

)

l The information provided in Table 3.3.5-1, control Ibd Withdrawal Block Instrumentation, is as follows for " Trip Function Scram Discharge Volume Water j

Level-Hig h * :

"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 Bod Withdrawal Block Instrumentation 'Setpoints, contains the following information for " Trip Function Scram Discharge Volume Water Level-Bigh*:

"1.

Trip Setpoints 1 36.2 gallons 2.

Allowable Values 1 36.2 gallons" The contents of Table 3.3.5-1 and 3.3.5-2 address the NRC staff's Model 1

Technical Specifications requirements of paragraph 3.3.6 and Table 3.3.6-1.

l Table 4.3.5-1, control Rod Withdrawal Block Instrumentation Surveillance 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 Scram Discharge volume Water Level-Bigh*:

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Channel Check: NA

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TER-C5506-73 2.

Channel Functional Test Q (quarterly) 3.

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

Operational Corditions in Which Surveillance Required:

1, 2, 5 (e) "

Notes:

"a.

Neutron detectors may be excluded from CHANNEL CALIBRATION.

e.

With any control rod withdrawn.

Not applicable to control rods removed per Specification 3.9.11.1 or 3.9.11.2. "

FRC EVALUATION The existing Eatch Unit 1 scram discharge system has six level switches t

on the scram discharge volume (see FSAR, page 3-43) set at three different

)

water levels to guard against operation of' the reactor without sufficient free volume present in the scram discharge headers to receive the scram discharge water in the event of a scram. At the first (lowest) level, one level switch initiates an alarm for operator action. At the second level, with the setpoint of 136.2 gallons (see page 3/43-40, Table 3.3.5-2, of the Hatch Unit 2 Tecnnical Specifications), 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 171 gallons (see page 3.1-4, Table 3.1-1, of the Hatch Unit 1 Technical Specification'ns), 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 acceptable compliance ("The present alarm and rod block instrumentation meets this criterion given adequate hydraulic coupling with the SDV headers"). The Batch scram discharge system has adequate hydraulic j

coupling between scram discharge headers and instrumented volume. Thus, the present alarm and rod block instrumentation is also acceptable.

I

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t TER-C5506-73 In Hatch Unit 1, " Scram Discharge Volume Scram Trips" cannot be bypassed while the reactor is in operational conditions of startup and run (see FSAR j

page 7-17), and operational condition " refuel with morn than one control rod withdrawn" is not applicable, since interlocks are provided which prevent the l

l 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 3.3.6-1 and paragraph 4.3.6 with Table 4.3.6-1 I

are not applicable to Hatch Unit i 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 Licensee's proposed amendment l

of the Hatch Nuclear Plant Unit 1 Tec hnical Specifications to incorporate surveillance requirements similar to those contained in the Unit 2 Technical Specifications of Table 4.3.5-1 is acceptable. 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.

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TER-C5506-73 5.

CONCLUSIONS Table 5-1 summarizes the results of the final review and evaluation of the Hatch Nuclear Plant Unit 1 Phase 1 proposed Technical Specifications changes for SDV long-term modification in regard to surveillance requirements for SDV vent and drain valves and LCD/ 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 a$d vent valves in the tables

.o of power-operated isolation valves (see revised pages 3.7-18a and 3.7-20 of the Hatch Nuclear Plant Unit 1 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 4.1.3.1.la anc 4.1.3.1.lb and is acceptable.

o The Licensee's agreement to revise the original provisions of the Hatch Nuclear Plant Unit 1 Technical Specifications given on page 3.1-8, Table 4.1-1, in regard to performing the reactor protection system SDV water level-high Channel Functional Test

monthly instead once per 3 months meets the NRC staff's Model Technical Specifica-tions requirements of paragraph 4.3.1.1 and Table 4.3.1.1-1 and is acceptable.

o The proposed amendment of the Hatch Nuclear Plant Unit 1 Technical Specifications to incorporate surveillance requirements for control l

rod withdrawal block SDV limit switches similar to those contained in the Unit 2 Technical Specifications Tables 3.3.5-1, 3.3.5-2, and 4.3.5-1 is acceptable.

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REFERENCES t

1.

IE Bulletin 80-14, " Degradation of BWR Scram Discharge Volume j

l Capacity" NRC, Office of Inspection and Enforcement, June 12, 1980

?

2.

D. G. Eisenhut (NRR), letter "To All Operating Boiling Water j

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 Testing 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, "Staf f Report and Evaluation of Supplement 4 to IE Bulletin 80-17" June 10, 1981 nklin Research Crater A Okemon of The fnsues m

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APPENDIX A NRC STAFF'S MODEL TECHNICAL SPECIFICATIONS

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Note: Applicable changes are marked by vertical lines in the margina U00 FrankHn Research C.e.n.ter 4 on

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RfACTivfTYCONTROLSYSTJM,,5 i

l LIMTTING CDNDITTOM 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.

3.

Otherwise, be in at least HOT SHUTDOW 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'OT SHUTCCWH within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

$URVEILLANCf RECUIPEPENT$

a 4.1.3.1.1 The scram discharge volume crain and vent valves sna11 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 RW and R$C3, all withdrawn control rods not recuired to have their directional control valves disarmed electrically or hydraulically shall be demonstrated OPERABLE by moving each control red at least one notch:

a.

At least once per 7 days, and b.

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

4.1.3.1.3 All control rods shall be canonstrated OPERABLE 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.

"Tnese valves may oe closed intermittently for testing under administrative controls.

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m TER-C5506-73 t!AciTVITr C xT20L sysTIMS C-iTR*L r.*O K1XIMtM SCXAM IN!!RTION T!MES t,IH!"iWO COWOIT10N FOR C7ErtAT70H 1

2.1. 2. 2 The maximum scru insertion time of each ::ntrol red from the fully witheravn position to notch position (6), based on de ener;ization of the scram pilot valve solencies as time :aro, shall not exceed (7.0) seconds.

AP*LICA3fLITr: CPERATICNAL CCHDITICH31 and 2.

ACTTON:

- Vith the maxistas scram insertion time of one or core control rods exceeding (7.0) seconds:

a.

Declare the c:ntrol red (s) with the slow insertion time incperable.

l and b.

Perfar: the Surveillance Requirements of Specification 4.1.3.2.c at least once per 50 cays wnen c eration is c6atinued with three or more control rods witn maximum scram insertion times in excess of l

(7.0) sqc=nds, or c.

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

SURVEILLANCE REcutREw!NTS 4.1.3.2 fhe maximum scram insertion time of the centrol rods shall be demon-strated througn measurement with reactor c:alant pressure greater than or

' equal to 950 psig and, during single consist red scram time tests, the emntrol rod drive pumps isolatad from the accumulat rs:

For all control rods prior to THERMAL POWER excaeding 4C% of RAf!D a.

THE.F.M. PC'a?.R following C3RE ALTERATIONS or aftar a reactor shutdown that is g-tatar than 120 days, b.

For spec.ifically affected individual cent n1 rods following r.aintenance on or = edification to the emntrol red =r control red drive system which could affect the scram insertion time of these specific corttrol ends, and' c.

For 1C% of the control rces, on a rotating basis, at least onca per 120 days of cperation.

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C TER-C5506-73 2/4.3 IN5"RWINTATION 2/4.2.1 Rr C~*2 PR*T CT7CN !Y.rTif' INSTRWENTATICH J

i LIFI ING C*CITICH FCR CFERATICN 1

3. 3.1 As a =fni v=, the reset:r protaction systas inst.:=in=atten cannels sa:en in Tule 3.3.1-1 shall be CPERA8LI with the RI;CTOR PR3TECTICH SY.T 7.33PCX3! TIME a.s shown in Ta.ble 3.3.1-2.

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&*L"' ABILITY: As shown in Tabit 3.3.1-1.

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Vith the nucte' of CPERABLE channels less than required by the Minist.n a.

r CPE?.A8LE Channels per Trip Systes requiracant for one trig systas, place at least one inopertble channel in the tripped c ndition within one hour.

With the nuncer of CPERABLE channels less than required by the Minimus C? ERA 3Li Channels.per Trip Systes requirement for both trip syst. ems, placs at least one inoperable enannel in at least ene tris systas". in the iM::ed ::ndition wiuin one hour and take : e ACT CN required by Ta:1e 2.3.1-1.

The :revisiens of 3 specification 3.0.3 are not a;plicule in CPE?.AT!CHAL C*H0!"!CH 5.

.CRVi!LLAN"! t!CUIRESENTS

4.3.1.1 fac reactar protection system instrumentation channel shall be ca=:r.s rt.44 C7EF.A8Li by ue per"srsance of 2e CMANNEL CHECX, CHANNE'.

FutCTICNAL 775T anc CHANNEL CAL 3 RAT %CH sperations for the Q7E?.ATICHAL j

ClCMDITICHS and at the frequencies, shown in Tatle 4.3.1.1-1.

4.3.1.2 LOGIC SYSTDf FUNCTIONAL TESTS and simulated automatie operation of ali cannels shall be performed at least onca per 18 sonths.

4.3.1.3 The P.!ACTCR PROTICT"CH SY3 TEM RESPCMSZ i!ME of each reactor trfp.

f::n:ti:n sh wn in Tacle 3.3.1-2 shall be daconstrated to be wfuin its limit at least :nca ;er 13 months. Ea:n test sna11 incluce at least one logic train entanel ;se functI:ic trains are tasted at least : :t ;ur 35 r.:ntas anc one s u:A tha. ::n lo-n suca taat all cannels art tas.ad at least once ever N ti.:ss l3 ::ntas where N is the tatal nu=ser of reduncant channels in a,.y spt:T fic ata: :t trip function.

.: :su :r.ac.aeis are incoeraba in one *ri; rystas, select at least one

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i T!!LE 3. 3.1-1 ' (Continued) 8.117 *tOTICTION SYSTEM THSTRUMENTATION ACTION (CI N 1 In 07ERAUCNR CONDmCH 2, be in at least MCT SMITOSW within 8 he:rs.

In C?ERATTCN4. CONom0N 5, suspend al,1 operations involving CCRE ALTE*ATION5* and fully insert all f asertable control rods within ore haar.

ACUN2 Lock the reactor mode switch in the Shutdown position within one tour.

A; TION 3 8e f a at leas-STARTUP within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

A*TI;N 4 In 0: ERA" ION 4 CONDITIDH 1 or 2, be in at least HDT SHUTDOW within 6 neurs.

In 0?E?A"IONA' CONDITICN S, suspend all operations involving CORE ALTIRATI:N5" and fully insert all insertable :entrol roas witnin ore nose.

1:T*:N 5 Se i: at least NOT EMUTOCW within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

A~~I*N $

Se f: STAAnlP vith the main staas line isciation valves closed within 2 houn or in at least HOT SHUTDOW within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

A:"!:N7 I..itiate a retuction in THERFAL PCWIR within 15 mTeutes and i

redu:s t:rt:ine first stage pressure to < (:50) psig, acufvelent to T3E?NL PC'TR 1ess than (30)% of RATI.D THERMAL PCWER, within 2 he:rs..

A:U:N$

In C?DAUCNR C2HOITION 1 or 2, be in at least. HDT SHUTCOW within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

In OFI?. CONE CDHDm0N 3 or 4, verify all insertable ::ntrol rods t.c le fully inserted wit \\in one hour.

In 0?PAUOME CON 0m0N $, suspend all eparations fr.volving CORE ALTIRATI N5* and fully insert all insertable' control reds within are her.

A: TION 9 fr> 0?PA"IOMR CON 3mCN 1 or 2 ha in at least N:7 SHUT:CW within 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months .

In OPE?AUORAL CCNDMCN 3 or 4, lock the ree= tor mode switch ir, tha 5.5.rtdo.it position within one hour.

In C?UA"!ONA*. CONDmCN 5, suspand all c;aratTons involving

. ::RI ALU?.ATION5* and fully insert, all insartable centrol rods within see a :r.

'ia:::. movement of I?.w., S?.v. se special =cvable detectors, or replacement of

_??.M strings provitec !?.w. i:str =entation is CPE?.AELE per Specificatica 3.9.2.

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tTR-C5 506-73 TU LE 3.3.1-1 (Continued) f REA* TOR MOTEC 10W SYSUM INSTRLHENTATION TA3Li NOTATIONS

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(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 f ar required survellian=e without placing the trip systes in the tripped concision provided at least one OPERAaLS channel in the same trip systas is sonitoring that pa'raseter.

b)

The ' shorting linW shall be removed from the RPS circuitry prior to and during the time any control rod is withdrawn" and shutdown mergin demonstrations performed per specification 3.10.1.

(c) An APRM channel is inoperable if there are less than 2 LPRM inputs per level or less tha.n (u) L7AM inputs to an APRM channel.

(d) These functions are not required to be OPERA 3Li when the reactor pressurs vessel nead is ur. bolted or removed per specification 3.10.1.

(e) This function small be aut==atically typassed when the reactor ::da switch is not in the hn position.

(f) This function is not required to be OPERABLE when PRIMARY CONTAINw.ENT INTIGRITY is not esquired.

(;) Also actuates the stanchy gas treatment system.

i (h) ' With any control rod withdrawn. Not applicable to centrol mes removed per Specification 3.9.10.1 or 3.9.10.2.

(f) These functiens are actsmatically bypassed vnen turbine first stage pressure is < (250) psig, equivalent to THERMAL POWR less than (30):l of RATED THEKMAL PC'ER.

4 (j) Also actuates the ECC-RPT system.

"Not recares 1=r control rods ramoved per Specification 3.9.10.1 or 3.9.10.2.

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3. 3. 5.

The contr:1 red withdrawal block instrumentation channels shown in it.ble 3.3.5-1 shall be CPERAELE win their trip set;oints set cansistant du

..: values sh:r.n in the Trip setpoint calur.n of itblg 3.3.5-2.

A25tICA!!LITY: As 'shown in' Table 3.3.6-1.

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A"T*:N:

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setpoint less conservative than us value shown in the A11ow6bie Values calumn of Taale 3.3.5-2, declare the channel inoperable until the channel is restored to CPEREL! status with its trip setpoint adjusted consistant with the Trip 5etpoint value.

b.

Vita the number of CPEMBLI channel.qJsss uan requirsd by the Minimu: CPEMELE Ohannels per Trip Functi:n t:;uirtment, taka the A0n:N. scuired by 74:le 3.3.5-b c.

The ;rsvisions of Specification 3.0.3 tre a:t a;;11czble in OPERA-TICNAL CONDITICH $.

CtVEILLANCE REOUIREw!NTS

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4. 3. 5 !ach of the amove required control red wiudrawal block trip systems a..2 instr =entation enannels shall be cascastrated OPEu2LI by us perf:r=ance

f tne CMANNEL CHECX, CMANNEL FUNCTIONAL TEIT anc CdANNE!. GLIERATION caera-ti:es fer the 07EMTICHAL CONDITIONS and at the frequencies shown in Tacle 4.3.5-1.

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TA!L! 3.3.5-1 (Continued)

C:: hit 0L R00 VTTHORAWat. *LOCX INSTE'?!NTATT04 ACTTCN Take the ACTICN required by Specification 3.1.4.3.

A:-*:N 60 Vith the nu=her of CPERABLE Channels:

A:T" N 81 a.

One less than required by the Midi,=um OPERABLE Channels per Trip function requirteent, restore the inoperable channel ta OPERABLE status wicin 7 days or place the ineperable channel in the tripped candition vicin the next hour.

b. ' Two or mort less than esquired by the Mini =um CPERA3LE Channels per Trip Function requirement, pt.aca at laast one inoperable channel in the tripped candition within one hour.

Vith us numeer of OPE?.A!Li :nannels less than required by the AZ:N f2 Mini =um CPE?.A3Li Channels ;~er Trip Fun: tion requirement, place the ineparable :nannel in the tri;;ed ::nditi:n vicin ene heur.

NC~is Vita TMi??AL PCVER 1 (20)lll cf RATID THE??AL PCkER.

With mort can ene control rod withertwn. Not ap:lica$le ta contral rods res:vec per !peciffcation 3.9.10.1 or 3.9.10.2 t.

no RIM shall b~s automatically bypassed when a ;sripheral czntrol rod is,

selec ed.

This function shall be automatically bypassed if detector count rate is

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

nis function shall be automatically bypused wnen the associated IRM

nar.nels ars on range 8 or higner, f.

nis function shall be aut:matica11y bypassed when tne IKM :hannels art en ange 3 or higner.

Bis function shall be automatically bypassed wnen the IRM :nanne'Is ire e.

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a.

b.

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Tha ;=;: sad sese:: mat.cul: 23 :: in::::::s:s tsviss: Tur.91:s1 50 sci'1:sti:ns in :ss ense t: you; C;1y 7, 1950, *.s:: :.

?s :::::ssc in.nical 5:s:ifi:sti:ns will s:isn;--in " :.-i o :visi: s '::

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s:sti n ty := vicing surveillance :scuiransets on te s::sm :is:ns;;s v:1::s.an: s..c

sin vslves.

In scdition, ths ::sssurs 311sf v31ves srs added to tPs existi.'g t.tries

f. = ntal. ment !seliti:n valvss and trelu sc in te

.- 11 saris t'.1r :

i

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c:: ::s3 it in:::sse tPa ;;; stility of :3vi:; sly sr.al :s:: ac:!:sa:s n cnifun :icns. f.targins of safety are increased by the ac-ition of 13:8.e.110 sureeilise.:s on tr.ase valvas *nich baceae contii. ant isolati n.sivas c.;;i.;

.tr.s tirs ;arica f:110.in; a s::am and t,afore tr.s ac:sm is :aset.

TM elant ?sview Scard 2nd the Safety.:sview Scar:. ave ravis=,1 t~e pt:: st: Onar;ts to the Tecnnical Scacificaticas and tne 03515, stated s?cve.

Por t.cs pr:::ssa c..angss, and have concluded that thsy do not involve an untsvisnac safsty cuestien.

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1 APPENDIX C l

s GEORGIA POWER CCMPANY LETTER OF OC10BER 1, 1981

-i f

WITH RESPONSE TO RFI FOR i

EDWIN I. HATCH NUCLEAR PLANT UNITS 1 AND 2 h

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3'"*8 Octocar 1, 1981 GeorgiaPower

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J. T. Besamen. Jr.

Vee beteent anc Generai uriger

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Mtc 00CXETS 50-321. 50-344 Nj OPUtATING LICENSES OPR-57, teF-5

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EDWIN I. HATCH NUC: JAR PLANT UNITS 1, 2 i

RE3PONSE TO FRANC.IN RE5lARCN CDm!R RECUEST CDPCHRNING SCRAM OISCH22 SYSTEM TECHNICAL :r.ur M.ATICNS QENTLDEN:

Your lettar dated Sectancer 1,1981, conveyed to Cecr;ia Power concany a recuest fer acciticnal information frem Franklin Researon Cantar (FRC) concerning cur Fectuary 26, 1981, suomittal of graccsed mccificaticns to the Tecnnical Scecificaticns regarcing the sc::am disonarge volune and asscciated inst:unents. The following infczmation is

~~1'=d in resconse to the FRC recpJest:

Itse 1 The model Tecnnical Specifications contained in your.kly 7,

1980, lettar placed the scram CDenerge volume vent and drain valves in section 3/4,1.3.1 of the model Technical Specifications; " Control Rod Operability".- Itse 1 of the FRC recuest asked for a reference to the section of the Tecnnical Specifications wners tne requestad enange is incorporated.

{

Our Fearuary 26,1981,1setar proposed that these valves be placect in the tablas of power operstad containment isolation valves instand of the

" Cartrol Rod Operability" section. These valves do not affect control rod coarseility at Plant Hatan. The plant unicus geometry of this system at Plant Hatch alloes free communication eetween the scram level switanes and the scram discharge volume (SDV).

Thus, tne level suitenas, not the vent and ctrain valves, protect the scram functicn, and in a sense control rod coerability, by providing assurance tnat the SDV is septy. The vent and crain valves are incertant, however, insofar as i

l they provide a contaitsuant pressure bouncary curing tne time that a scram is sealad-in. For this reason we have acoan to place the valves in the tables of containment isolation valves.

The surveillance roczai.e are therefore citTerent tnan those proposed by the econ 1 j

Tecmical Specification in order to be consistent with the retnairements for otner comparacia containment isolation valves.

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PDR C-1 ranklin Researc

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- Cecr3faPowerA Director of Nucisar Reacter Pr'1*m

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U. S. t4mlaar P r 'far m w==4 cn Octcoor 1, 1981 Page Two Itsee 2 and 3 As indicated in our Octoeer 10, 1980, lattar the scram leve[ switmes tre currently covered by Tecnnical Specifications c6 eacn unit. For Unit 1, please refer to Specifications 3.1 and 4.1, Tacles 3.1-1 and 4.14, item 7.

Fct Unit 2, the accrecriata reference is Specification 3/4.3.1, taolas 3.3.1-1 and A.3.1-1, itam S.

The instrument functional test frecuency for Unit.1 is cree ever/ t.w months as initially accroved by the Cennission en issuarce of the Unit 1 Operating t.icanse.

We nave not creocsed to.ccify tais specification.

1 Itsms a. 5 and 6 The SDV rod micek setecint and' surveillance recui.wts are scocified i

in Unit 2 Tecnnical Specification Section 3.3.5 and Tacias 3.3.5-1, J.3.5-2 and 4.3.5-1.

In reytowirq the Tecnnical Specifications for our l

Feoruary 26, 1981 semaittal, the ansance of a ccacarabis specification l

in the LJtit 1 Temnical" Specifications was not noted. We agree that it is appropriate to specify the limits and surveillance requirements for the EN rod block alarm switch and will propose an amenonent to the Unit -

1 license to incorporate requirements similar to these contained in our (21112 Specification referenced soove.

very truly yours, l

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J. T. Beckham, Jr.

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R. F. Rogers, III l

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DllVUU Frankan Research Center en -

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NUCLEAR REGULATORY COMMISSION

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 34 TO FACILITY OPERATING LICENSE NO. NPF-5 GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA CITY OF DALTON, GEORGIA EDWIN I. HATCH NUCLEAR PLANT, UNIT NO. 2 DOCKET NO. 50-366 Introduction As a result of events involving common cause failures of Scram Discharge Volume (SDV) limit switches and SDV drain valve operability, the NRC staff issued IE_ Bulletin 80-14 on June 12, 1980.

In addition, the NRC staff sent a letter dated July 7,1980, to all operating BWR licensees requesting that they propose Technical Specification (TS) changes to provide surveillance requirements for SDV vent and drain valves and Limiting Conditions for Oceration (LCO)/ surveillance requirements on SDV limit switches. tiodel TSs were enclosed with this letter to provide guidance to licensees for preparation of the reouested submittals.

Evaluation The enclosed report (TER-C5506-78) was prepared for us by Franklin Research Center (FRC) as part of a technical assistance contract program. Their report provides their Technical Evaluation of the compliance of Georgia Power Company's (the licensee) submittal with NRC provided criteria.

FRC has concluded that the licensee's response does not meet the explicit requirements of paragraph 3.3-6 and Table 3.3.6-1 of the NRC staff's Model TSs. However, the FRC report concludes that technical bases are defined on page 50 of the staff's " Generic Safety Evaluation Report BWR Scram Discharge 1

System," December 1, 1980, for this departure from the explicit requirements of the Model TSs. We conclude that these technical bases justify a deviation from the explicit requirements of the Model TSs.

The licensee proposed in its February 26, 1981, submittal to list the SDV vent and drain valves as containment isolation valves and to perform only the surveillance required for containment' isolation valves. The FRC report notes that it found this proposed surveillance to be unacceptable.

It also notes j

that in discussions on this subject with FRC, the licensee orally agreed to revise its proposed surveillance requirements to meet the NRC staff Model TS requirements for surveillance of these SDV vent and drain valves.

j Thus, FRC has concluded that the licensee's proposed TSs (as revised by f

subsequent discussion with the licensee) meet our criteria, j

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-4 Subsequently 24, 1983, the Commission required that the licensee 1)y, by Order dated June install the long term BWR scram discharge modifications in

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conformance with the staff's December 1,1980 Generic SER on Scram Discharge

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Systems before December 31, 1983 and 2) submit TS changes required for operation with the modified system at least 3 months prior to the required implementation date.

By letter dated September 19, 1983, the licensee submitted proposed TS revisions in accordance with its previous oral commitment, as discussed above. We have reviewed these proposed revisions and find that they are consistent with this previous commitment that provided the bases for J

acceptance in the FRC report and conclude that they are acceptable. The licensee also proposed to add an LC0 for the operability of SDV vent and i

drain valves. This LC0 would require the plant to be placed in Hot Shutdown in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months if any SDV vent or drain valve is inoperable. We find this proposed LCO is consistent with the NRC staff Model TSs and conclude that it is acceptable.

By letter dated Cetober 3,1983, the licensee has withdrawn its February 26, 1981, submittal that requested that the SDV vent and drain valves be listed in containment isolation valve tables and be required to meet only containment isolaticn valve surveillance requirements. These surveillance requirements are now provided by the licensee's proposed TSs submitted in its September 19, 1983, letter as discussed above.

In its September 19, 1983, submittal, the licensee has also proposed to add TS requirements, including trip setpoint, LCO, Action Statement and surveillance requirements, for the new diverse SDV high water level scram instrumentation (Thermal Level Sensors) to the Reactor Protection System instrumentation tables. The new instruments have been given the same requirements as the original level switches which were found acceptable in the FRC report. We conclude that this proposed addition is acceptable.

By letter dated December 14, 1983, the licensee proposed to change the 30 second closure time requirement for the SDV vent and drain valves as proposed in its September 19, 1983 submittal to 60 seconds for the inboard vent and drain valves and 120 seconds for the outboard vent and drain valves. This proposal deviates from the acceptability guidelines of 30 seconds closure time provided by the staff in its Generic Safety Evaluation Report on Scram Discharge Systems dated December 1, 1980. The staff is currently reviewing the licensee's justification for this latest proposed change.

In the interim, the staff has concluded that there is reasonable assurance of safe operation of the plants based on implementation of the short-term corrective measures noted in the June 24, 1983 Order; and the long-term corrective measures noted herein.

Based upon our review of the contractor's report and discussions with the reviewer and on our review, as discussed above, o' the ' licensee's subsequent submittals augmenting the information reviewed by the contractor, we conclude i

that except for the SDV vent and drain valve closure requirements as i

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discussed above, the licensee's proposed TSs satisfy our requirements for surveillance of 50V vent and drain valves and for LCOs and surveillance requirements for 50V limit switches. Consequently, we find the licensee's proposed TSs, except for the SDV vent and drain valve closure times, acceptable.

Environmental Consideration We have determined that the amendment does not authorize a change in effluent types or total amounts nor an increase in power level and will not result in any significant environmental impact. Having made this determination, we have further concluded that the amendment involves an action which is insignificant from the standpoint of environmental impact and, pursuant to 10 CFR 651.5(d)(4), that an environmental impact statement, or negative declaration and environmental impact appraisal need not be prepared in I

connection with the issuance of the amendment.

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

(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and.(2) such activities will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the ccmmon defense and security or to the health and safety of the public.

Dated: January 4, 1984 The following NRC personnel contributed to this Safety Evaluation:

Ken Eccleston and George Rivenbark i

3 4

____--__--______n

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]

ML20244B666

Text

SUMMARY

SUMMARY

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