ML20209J125

Conformance to Reg Guide 1.97,James a Fitzpatrick Nuclear Power Plant

ML20209J125
Person / Time
Site:	FitzPatrick
Issue date:	10/31/1985
From:	Udy A EG&G IDAHO, INC.
To:	NRC
Shared Package
ML20209J124	List: ML20209J125
References
CON-FIN-A-6483, RTR-NUREG-0737, RTR-NUREG-737, RTR-REGGD-01.097, RTR-REGGD-1.097 GL-82-33, TAC-51090, NUDOCS 8511110362
Download: ML20209J125 (17)

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i CONFORMANCE TO RrGULATORY GUIDE 1.97 JAMES A. FITZPATRICK NUCLEAR POWER PLANT

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i EG&G Idaho, Inc.

Idaho Falls, Idaho 83415 t i il 3

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Prepared for the l f i U.S. Nuclear Regulatory Commission j Washinoton, D.C. 20555 . ,

Under 00E Contract No. DE-AC07-761001570 l'

FIN No. A6483 .

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ABSTRACT This EG&G Idaho, Inc., report reviews the submittal for Reculatory i Guide 1.07, Revision 2, of the James A. FitzPatrick Nuclear Powar Plant and identi#ies areas of nonconformance to the renulatnry cuite. Exceotions to

. Reculatorv Guide 1.07 are evaluated and those areas where su'ficient basis foracceptabilityisnotprovidedareidentified.

FOREWORD This report is supplied as part of the "Prcoram for Evaluatino Licensee / Applicant Conformance to RG 1.97 " beino conducted for the ,

U.S. Nuclear Reculatory Commission, Office of Nuclear Reactor Reculation, Division of Systems Inteoration, by EG&G Idaho, Inc., NRC Licensina Supoort Section.

The U.S. Nuclear Reculatory Commission funded the work under authorization B&R 20-19-10-11-3.

Dock?t No, c0 333 TAC No. 51090 11 1

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$ CONTENTS i I

( ABSTP.ACT.............................................................. 11 11

', FOREWORD ..............................................................

i j 1. INTRODUCTION ..................................................... 1 2

2. RE V I E W REQU I RE ME NT S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

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3. EVALUATION ....................................................... 4 1

3.1 Adherence to Reoul atory Guide 1.97 . . . . . . . . . . . . . . . . . . . . . . . . . 4 3.2 Typ e A Y ar i ab l e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 i

3.3 Exceptions to Reculatory Guide 1.97 ........................ 4

4. CONCLUSIONS ...................................................... 12

] 5. REFERENCES ....................................................... 13 1,

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d CONFORMANCE TO REGULATORY GUIDE 1.97 JAMES A. FITZPATRICK NUC'. EAR POWER PLANT

1. INTRODUCTION On December 17, 1982, Generic Letter No. 82-33 (Reference 1) was issued by D. G. Eisenhut, Director of the Division of Licensina, Nuclear Reactor Regulation, to all licensees of operatino reactors, aoplicants for operating licenses, and holders of construction oermits. .This letter included additional clarification regarding Regulatory Guide 1.97, Revision 2 (Reference 2), relating to the requirements for emergency response capability. These reqJirements have been published as Supplement No. I to NUREG-0737, "TMI Action Plan Requirements" (Reference 3).

The New York Power Authority, the licensee for the James A.

FitzPatrick Nuclear Power Plant, provided a response to Section 6.2 of the generic letter on November 30, 1984 (Reference 4).

This report provides an evaluation of that material.

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2. REVIEW REQUIREMENTS Section 6.2 of NUREG-0737, Supplement No.1, sets forth the documentation to be submitted in a report to the NRC describing how the licensee complies with Reculatory Guide 1.97 as applied to emergency response facilities. The submittal should include documentation that provides the following information for each variable shown in the applicable table of Regulatory Guide 1.97:

1. Instrument rance

2. Environmental cualification

3. Seismic qualification 4 Quality assurance

5. Redundante and sensor location

6. Power supply

7. Location of display

8. Schedule of installation or upgrade Furthermore, the submittal should identify deviations froa the regulatory ggide and provide supporting justification or alternatives.

Subsequent to the issuance of the ceneric letter, the NRC held regional meetings in February and March 1983, to answer licensee and .

applicant questions and concerns reaardino the NRC policy on this subject.

At these meetings, it was noted that the NRC review would only address exceptions taken to Regulatory Guide 1.97, Furthermore, where licensees or .

applicants' explicitly state that instrument systems conform to the regulatory guide, it was noted that no further staff review would be

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e, necessary. Therefore, this report only addresses exe.eptions to Regulatory Guide 1.97. The followino evaluation is an audit of the licensee's submittal based on the review policy described in the NRC regional meetinos.

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3. EVALUATION The licensee provided a response to Section 6.2 of NRC Generic Letter 82-33 on November 30, 1984. This response describes the licensee's position on post-accident monitoring instrumentation. This evaluation is based on that material, i

3.1 Adherence to Reoulatory Guide 1.97 The licensee states that in most cases they meet the recommendations of Regulatory Guide 1.97. Technical justification was supplied for deviations that were identified and where instrumentation was not modified to provide compliance. The licensee states that modifications beino made to implement Regulatory Guide 1.97 will be complete within 30 days after the end of the 1987 (Reload 8/ Cycle 9) refuelino outaae or December 31, 1987, whichever is later. The conformatory order, which is in the process of being issued, chances December 31, 1987 to April 30, 1088.

Therefore, we conclude that the licensee has orovided an explicit commitment on conformance to Renulatory Guide 1.97. Exceotions tn and deviations from the reculatory guide are noted in Section 3.3.

3.2 Type A Variables Regulatory Guide 1.97 does not specifically identify Type A variables, i.e., those variables that provide in'ormation required to permit the control room operator to take specific manually controlled safety actions.

The licensee classifies the following instrumentation as Type A.

1. Reactor coolant system pressure

2. Coolant level in reactor vessel

3. Suppression pool water temperature ,

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4 Suppression pool water level

5. Drywell pressure

6. Residual heat removal system flow

7. Drywell temperature

8. Suppression chamber pressure ,.

9. Residual heat removal service water system flow

10. Containment hydrocen concentration

11. Containment oxygen concentration

12. Core spray system flow

13. Core spray system pressure The above variables either meet or will be modified to meet the Cateoory I requirements consistent with the requirements for Type A variables.

3.3 Exceptions to Reculatory Guide 1.07 The licensee identified deviations and exceptions from Reculatory Guide 1.97. These are discussed in the following paracraphs.

3.3.1 Neutron Flux .

The licensee has supplied Catecory 2 instrumentation for this variable, some of which is not environmentally qualified. Regulatory Guide 1.97 specifies Category 1 instrumentation.

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The licensee states that ricorous Cateaory 1 desion criteria is not justified. They also state that modifications are not needed since boron samplino and control rod drive (CRD) position indication provide adequate backup. These are not Catecory 1. The licensee also states that they consider that none of the neutron flux equipment requires qualification per the Environmental Qualification Rule,10 CFR 50.d9. 10 CFR 50.49(b) 3) requires environtrantal qualification of post-accident monitoring equipment. '

In the process of our review of the neutror, flux instrumentation for boiling water reactors, we note that the mechanical drives of the detectors have not sati J ied the environmental qualification requirement of Regulatory Guide 1.97. A Category I system that meets all the criteria of Regulatory Guide 1.97 is an industry development item. Based on our review, we conclude that the existina instrumentation is acceptable for interim operation. The licensee should follow industry development of this equipment, evaluate newly developed equipment, and install Cateoory 1 instrumentation when it becomes available.

3.3.2 Drywell Sump t.evel Drywell Drain Sumos Level Regulatory Guide 1.07 recomends Cateaory 1 instrumentation with indication from the bottom to the top of each sump. The licensee has Category 1 drywell level instrumentation with a range of 0 to 100 feet. A single narrow range instrument measuring from 19 to 45 inches is provided for each sump. The transmitter for one sump is environmentally and seismically qualified. Otherwise the narrow range channels are Category 3. The deviatinn for this variable is in the cateoory of the supplied instrumentation. This instrumentation does not cause any automatic or operator initiated safety-related functions. The sumo systems .

are automatically isolated by an accident signal as part of containment isolation. This prevents the pump-out of the sump contents.

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We conclude that the instrumentation provided by the licensee will provide appropriate monitoring of the parameters of concern. This is based on (a) for small leaks, the instrumentation is not expected to experience harsh environments during operation, (b) for laraer leaks, the sumps fill promptly and the sumo drain lines isolate due to the increase in drywell pressure, thus negating the drywell sump level and drywell drain sumps I

level instrumentation, and (c) this instrumentation neither automatically initiates nor alerts the operator to initiate operation of a safety-related system in a post-accident situation. Therefore, we find .the instrumentation provided acceptable.

3.3.3 Radiation Level in Circulatino Primary Coolant The licensee indicates that measurements to indicate fuel claddino 4 failure are provided by the following instruments:

1. High radiation sampling system

2. Condenser off-gas radiation monitors

3. Main steamline radiation monitors

4. Primary containment radiation monitors

5. Containrent hydrogen concentration monitors

6. Area radiation monitors Based on the alternate instrumentation provided by the licensee, we conclude that the instrumentation supplied for this variable is adequate .

. and, therefore, acceptable.

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3.3.4 Radiation Exposure Rate Regulatory Guide 1.47, Revision 2, specifies Cateoory 2 instrumentation for this variable with a rance 10'I to 10 R/hr. The licensee has provided instrumentation for this variable with a rance of 3

10-I to 10 mR/hr.

I The licensee states that there is no requirement to enter areas monitored by these instruments in a post-accident situation,*and that existing Category 3 radiation exposure rate monitors (rather than Category 2) that have ranges lower than recomended by Regulatory Guide 1.97 are acceptable. The licensee states that it is impractical to detect primary containment breach by use of these monitors. The licensee determines the habitability of the secondary containment by a combination of airborne activity samples and local ridiological survey instruments for beta and gamma dose rates.

Reculatory Guide 1.97, Revision 3 (Rtference 5), chances this variable to Category 3. Therefore, the only deviation at the FitzPatrick station for this variable is the rance supplied. The licensee has not shown any analysis of radiation levels expected for the monitor locations.

The licensee should show that the existing radiation exposure rate monitors have ranges that encompass the expected radiation levels in their locations.

3.3.5 Low Pressure Coolant Injection System Flow Residual Heat Removal System Flow Regulatory Guide 1.97 reconsnends instrumentation for these variables ,

with a range from 0 to 110 percent of design flow. The licensee's instrumentation has a range from 0 to 108 percent of design flow. The licensee did not provide justification for this deviation, but states that the existing range is close to 110 percent of design flow and is adequate to determine pumo runout flow rate during an accident. -

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The existing rance is adequate to provide the necessary accident and

! post-accident information. Therefore, these are acceptable deviations'from Regulatory Guide 1.97.

3.3.6 Standby Liquid Control System (SLCS) Flow

Regulatory Guide 1.97 recommends instrumentation with a range of 0 to 110 percent design flow for this variable. The licensee indicates that a flow measuring devices for this manually initiated system'are not i *provided. However, the flow could be verified by the followino

1. Fump discharge header pressure

2. Pump runnino licht

3. Level of the SLCS storage tank '

4. Reactivity chance in the reactor as measured by neutron flux

5. Squib valve continuity and ready indicatino lights l

We find the above instrumentation valid as an alternative indication of

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SLCS flow.

j 3.3.7 Standby Liould Control System Storaae Tank Level ,

l The licensee's instrumentation for this variable is not environmentally qualified. No justification for this deviation was aiven.  !

Environmental qualification has been clarified by the Environmental .

. Qualification. Rule, 10 CFR 50.49. We. conclude that Reaulatory Guide 1.97 has been superseded by a regulatory requirement. Any exception to this rule is beyond the scope of this review and should be addressed in j accordance with 10 CFR 50.49, t

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3.3.8 Residual Heat Removal (RHR) Service Water Flow Regulatory Guide 1.97 recomends instrumentation with a rance from 0 to 110 percent of desion flow for the variable coolina water flow to engineered safety feature (ESF) system components. The licensee's instrumentation measures only to desian flow. No justification for this I

deviation was given.

The licensee should either show that this rance is adequate for all accident and post-accident conditions, or provide the recommended rance.

3.3.9 Hiah Radioactivity Licuid Tank Level The licensee's instrumentation for this variable has it's display in the radwaste control room. The licensee states that any tank inputs are isolated with an accident sianal and that control room operators must verify the tank storace capacity with the radwaste operators. Based on this, we find this deviation in display location acceptable.

3.3.10 Reactor Buildino or Secondary Containment Area Radiation Reculatory Guide 1.97 recommends Catecory 2 instrumentation with a range of 10'I to 10# R/hr for the FitzPatrick Mark I containment. The licensee has supplied Cateaory 3 instrumentation with a rance of 10-I to 3

10 mR/hr.

The licensee states that there is no requirement to enter areas monitored by these instruments in a post-accident situation, and that the existina Category 3 instruments (rather than Cateoory 2), that have a range lower than recommended by Regulatory Guide 1.97, are acceptable. The ,

licensee states that it is impractical to detect primary containment breach by use of these monitors. The licensee determines the habitability of the l secondary containment by a combination of airborne activity samples and local radiological survey instruments for beta and gamma dose rates.

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i Based on the justification provided by the licensee, we find that i Category 3 instrumentation is acceptable. However, the licensee has not

shown any analysis of radiation levels expected for the monitor locations. ,

The licensee should show that the existing instruments have ranges that

! encompass the expected radiation levels in their locations.

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4. CONCLUSIONS Based on our review, we find that the licensee either conforms to or is justified in deviating from Reculatory Guide 1.97, with the followino exceptions:

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1. Neutron flux--the licensee's present instrumentation is acceptable on an interim basis until Cateoory 1 instrumentation is developed and installed (Section 3.3.1). -

) 2. Radiation exposure rate--the licensee should show that the rances supplied for this variable encompass the radiation level at the instrument location (Section 3.3.4).

3. Standby liquid control system st'orage tank level--environmental qualification should be addressed in accordance with 10 CFR 50.49 (Section 3.3.7). -

4 RHR service water flow--the licensee should either show that the range is adequate or provide the recommended rance (Section 3.3.8).

5 .- Reactor buildino or secondary containment area radiation--the licensee should show that the rances supplied for this variable encompass the radiation level at the instrument location (Section 3.3.10).

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5. REFERENCES

1. NRC letter, D. G. Eisenhut to All Licensees of Operatina Reactors, Applicants for Operatina Licenses, and Holders of Construction Permits, " Supplement No.1 to NUREG-0737--Requirements for Emergency ,

Response Capability (Generic Letter No. 82-33)," December 17, 1982.

2. Instrumentation for Licht-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions Durino and Followino an Accident, i 1 Regulatory Guide 1.97, Revision 2, NRC, Office of Standards Development, December 1980.

3. Clarification of TMI Action Plan Requirements, Requirements for Emercency Response Caoability, NUREG-0737, Supplement No.1, NRC, Office of Nuclear Reactor Regulation, January 1083.

4. New York Power Authority letter, C. A. McNeill, Jr. to Director of Nuclear Reactor Reculation, NRC, " Supplement No. I to NUREG-0737 (Generic Letter 82-33), Regulatory Guide 1.97, Revision 2, Implementation Report," November 30, 1984, IPN-84-77.

5. Instrumentation for Licht-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions Dtirina and Followina an Accident, seculatory Guide 1.97, Revision 3, NRC, Office of Nuclear Reoulatory Research, May 1983.

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%',"3$'- BIBLIOGRAPHIC DATA SHEET EGG-EA-7040 ut etsSM#uCTeo48 om T-4 tve.54 J LS&WISh'%E 2 vif6t s%o Swaist LE Conformance to Regulatory Guide 1.97, James A.

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This EG&G Idaho, Inc. report reviews the submittal for the James A. Fitzpatrick Nuclear Power Plant and identifies areas of nonconformance to Regulatory Guide 1.97.

Exceptions to these guidelines are evaluated and those areas where sufficient basis for acceptability is not provided are identified.

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