ML20148C044

Rev 1 to Conformance to Reg Guide 1.97 - Fitzpatrick

ML20148C044
Person / Time
Site:	FitzPatrick
Issue date:	11/30/1987
From:	Udy A EG&G IDAHO, INC., IDAHO NATIONAL ENGINEERING & ENVIRONMENTAL LABORATORY
To:	NRC
Shared Package
ML20148B037	List: ML20148B033 ML20148B100 ML20148C044
References
CON-FIN-A-6483, RTR-REGGD-01.097, RTR-REGGD-1.097 EGG-EA-7040, EGG-EA-7040-R01, EGG-EA-7040-R1, TAC-51090, NUDOCS 8803220250
Download: ML20148C044 (20)

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EGG-EA-7040 TECHNICAL EVALUATION REPORT CONFORMANCE TO REGULATORY GUIDE 1.97--

FITZPATRICK Cocket No. 50-333 Alan C. Udy Published November 1987 Idaho National Engineering Laboratory EG&G Icabo, Inc.

Idaho Falls, Icabo 83415 Prepared for the U.S. Nuclear Regulatory Commission Washington. 0.C. 20555 Under 00E Centract No. DE-AC07-76IC01570 FIN No. A6483

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' ABSTRACT This EG&G Idaho, Inc., report reviews the submittals for Regulatory Guide 1.97, Revision 2, of the James A. Fit: Patrick Nuclear Power Plant and identifies areas of nonconformance to the regulatory guide. Exceptions to Regulatory Guide 1.97 are evaluated and those areas where sufficient basis for acceptacility is not provided are icentifiec.

Docket No. 50-333 TAC No. 51090 11

FOREWORD This report is supplied as part of the "Program for Evaluating Licensee / Applicant Conformance to RG 1.97," being conducted for the U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulatien, Division of Engineering and System Technology, by EG&G Idaho, Inc.,

Electrical, Instrumentation and Control Systems Evaluation Unit.

The U.S. Nuclear Regulatory Commission funded the work under autneri:atien B&R 20-19-10-11-3, FIN No. A6453.

Docket No. 50-333 TAC No. 51090 iii

CONTENTS A B S T RA C T . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 FOREWORD .............................................................. iii

1. INTRODUCTICN ..................................................... 1

2. R EV I EW R EQU I R EM E N T S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2

3. EVALUATION ....................................................... 4 3.1 Adherence to Regulatory Guide 1.97 .................... .... 4 3.2 Type A Variables ........................................... 4 3.3 Exceptions to Regulatory Guide 1.97 ........................ 5

4. CONCLUSIONS ...................................................... 13

5. REFERENCES ....................................................... 14 iv

CONFORMANCE TO REGULATORY GUIDE 1.97--

FITZPATRICK

1. INTRODUCTION Cn Cecumber 17, 1982, Generic Letter No. 82-33 (Reference 1) was issued by D. G. Eisenhut Director of the Division of Licensing, Nuclear Reactor Regulation, to all licensees of operating rear. tors, applicants for operating licenses, and holders of construction permits. This letter included additional clarification regarding Regulatory Guide 1.97, Revision 2 (Reference 2), relating to the requirements for emergency responst capability. These requirements 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.

Fity, Patrick Nuclear Power Plant, provided a response to Section 6.2 of the gereric letter on November _ 30, 1984 (Reference 4). Additional information was provided on June 28, 1985 (Reference 5), on December 24, 1985 (Reference 6) and on Fecruary 25, 1986 (Reference 7).

Tnis report prevides an evaluation of these submittals.

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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 iteensee complies with Regulatory 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 range

2. Environmental qualification

3. Seismic qualification 4 Quality assurance

5. Redundance and sensor location

6. Power supoly

7. Location of display

8. Schedule of installation or upgrade The submittal should identify deviations from the regulatory guide and provide supporting justification or alternatives.

Subsequent to the issuance of the generic letter, the NRC held regional rasetings in February and March 1983, to answer licensee and applicant questions and concerns regarding the NRC policy on this subje.ct.

At these meetings, it was noted that the NRC review would only address exceptions taken to Regulatory Guide 1.97. Where licensees or applicants explicitly state that instrument systems conform to the regulatory guice, it was noted that no further staff review would be necessary. Therefore, 2

4 this report only addresses exceptions to Regulatory Guide 1.97. The following evaluation is an audit of the licensee's submittals based on the review policy described in the NRC regional meetings, 3

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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 instrumentatior. Additional information was provided on June 23, 1985, Decembar 24, 1995, February 25, 1986 and June 9, 1987. This evaluation is based on these submittals.

3.1 Adherence to Reculatory Guide 1.97 The licensee states that in most cases they meet the recommendations of Regulatory Guice 1.97. Technical justification was supplied for deviations that were identified and wnere instrumentation was not modified to provide compliance. The licensee states that modifications being made to implement Regulatory Guide 1.97 will be complete within 30 days after the end of the 1987 (Reload 3/ Cycle 9) refueling outage or April 1,1988, whichever is later. Any required mootfications to the instrumentation for the standby liquid control system will be scheduled as part of the anticipated transient witnout scram (ATWS) ruling. Therefore, we conclude that the licensee has provided an explicit commitment on conformance to Regulatory Guide 1.97. Exceptions to and deviations from the regulatory guice 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 the information reouired to permit the control room operator to take soecific manually controlled safety actions.

The licensee classifies the following instrumentation as Type A.

1. Reactor coolant system pressure i

l 2. Coolant level in reactor vessel

3. Suppression pool water temperature 4

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4. Suppression pool water level

5. Drywell pressure

6. Residual heat removal system flow

7. Drywell temperaturt

8. Suppression chamber pressure

9. Residual heat removal service water system flow

10. Containment hydrogen 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 Category I recuirements consistent with the requirements for Type A variables.

3.3 Exceptions to Regulatory Guide 1.97 The licensee identified deviations and exceptions from Regulatory Guide 1.97. These are discussed in the following paragraphs.

3.3.1 Neutron Flux The licensee has supplied Category 2 instrumentation for this variable, some of which is not environmentally qualified. 73;!= tory Guide 1.97 specifies Category 1 instrumentation.

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The licensee states that rigorous Category 1 design criteria is not justified. They also state that modifications are not needed since boron sanpling and control rod drive (CRD) position indication provice adequate backup. These are not Category 1. The licensee also states that tney consider that none of the neutron flux equipment requires qualification per the Environmental Cualification Rule, 10 CFR 50.49. 10 CFR 50.49(b)(3)

, requires environmental qualification of post-accident monitoring equipment.

In the process of our review of the neutron flux instrumentation for boiling water reactors, we note that the mechanical drives of the detectors have not satisfied the environmental qualification requirement of Regulatory Guide 1.97. A Category I system that meets all the criteria of Regulatory Guide 1.97 has been an industry development item. Based on our review, we conclude that the existing instrumentation is acceptable for interim operation. The licensee has committed to follow industry development of this equipment (Reference 6) and evaluate newly developed equipment. The licensee states that they will consider installing Category 1 instrumentation when it becomes available.

3.3.2 Coolant Level in Reactor Regulatory Guide 1.97 recommends Category 1 instrumentation for this variable witn a range from the bottom of the core support plate to the centerline of the main sttamline. In Reference 8, the licensee identifies that their instrumentation will deviate from the regulatory guide requirements in the following areas:

1. Category 1 instrumentation will monitor to 224.5 inches above the top of active fuel (TAF) which is 63.5 inches belcw the centerline of the main staamlines, and

2. a Category 3 shutdown range char.nel will moniter above 224.5 inches above the TAF.

The Category 1 cnannels provide all safety system initiation signals.

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With all safety trips occurring before the range of tne Category 1 instrumentation would be exceeded, there are no additional manual operations to be taken should the level range of the Category 1 channels be exceeded. Reference 9 looks at the potential for overfilling the reactor vessel anc, for its analyses, shcws that the range of the Category 1 instrumentation will not be exceeded.

The licensee provided justification for not upgrading the Category 3 shutdown channel to Category 1. The justification considers the requirements for adding a redundant second channel (i.e., new containment penetration, new reactor vessel taps (as no existing taps would provide the recommended redundancy), and provisions to prevent reference leg flashing) that make such an upgrade extremely costly and impractical.

Based on the licensee's justification ard Category 3 instrumentation that$onitorslevelsabovethefuelconeandwiderangeinstruments,we find the instrumentation provided for this variable acceptable.

3.3.3 Drywell Sumo Level Orpell r Drain Sumos !.evel Regulatory Guide 1.97 recontends Category 1 instrumentatien with indication fecm the bottom to the top of each sump. The licensee nas Category 1 drywell level instrumentation with a range of 0 tc 100 feet. A single narrow range instrument measuring frcm 19 to 45 inches is provided fer each of two sumps. These sump transmitters are not environmentally qualified. This instrumentation does not cause any automatic or operator initiated safety-related functions. The sump systems are automatically isolated by an accident signal as part of containment isoiation. This prevents the pump-out of the sump contents. The drywell level instrumentation is usable after these relatively small sumps are full.

We conclude that the instrumentation provided by the licensee will provice appropriate monitoring of the parameters of concern. This is based on (a) for small leaks, the instrumentation is not expected to experience 7

i harsh environments during operation, (b) for larger leaks, the sumps fill promptly and the sump drain lines isolate cue to the increase in drywell pressure, thus negating the drywell equipment drain sump level and drywell 4 floor drain sump level instrumentation, and (c) this instrumentation neither autcmatically 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.

3.3.4 Radiation level in Circulating primary Coolant 1

The licensee indicates that measurements to indicate fuel cladcing failure are provided by tne following instruments:

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1. High radiation sampling system J

2. Condenser off gas ractation monitors 4

3. Main steamline radiation monitors 1

4. Primary containment radiation monitors i

5. Cor.tainment hyaregen concentration monitors

6. Area radiation monitors Based on the alternate instrumentation provided by the licensee, ne j

conclude that the instrumentation supplied for this variable is adequate j and, therefore, acceptable.

3.3.5 Radiation Exposure Rate a

i Regulatory Guide 1.97, Revision 2, specifies Category 2 instrumentation for this variable with a range 10'1 to 104 R/hr. The

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licensee has provided instrumentation for this variable with a range of 3 ~4 10'1 to 10 mR/hr (10 to 1 R/hr), located in thirty areas in and around the plant.

! The licensee states tnat there is no requirement to enter areas monitered by thesa instruments in a post-accident situati:n, and that existing Category 3 radiation exposure rate eenitors (rather than l Category 2) that have ranges lower tnan 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 secondary containment by a combination of aireorne activity samples and local radiological survey instruments for beta and gamma dose rates, j

j Regulatory Guide 1.97, Revision 3 (Reference 10), changes this variable to Category 3. Therefore, the only deviation at tne FitzPatrick station for this variable is the range supplied. The licensee has reported an analysis of radiation levels expected for the monitor locations. They state that at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following an accident, more than half of the j detectors would be in radiation fields of lest than 1 R/hr and would be on i scale. After 14 days, the average dose rate would be less than 1.7 R/hr.

The licensee states that they have procedures which, from the control room, provide for appropriate controlled access to vital plant areas. They further state that entry into the reactor building is not necessary in the snort or medium term. Any personnel entry would be preceded by surveys j

with portable instrumentation.

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i Should the instrument range be exceeded, backup instrumentation, including portable survey instruments, airborne sampling, the high range

! effluent monitoring system. high range radiation monitors located in other i areas of the plant, and the plant noble gas effluent monitors, will be used l by the licensee for long term release surveillance. Based on this, we find i

the licensee's instrumentation for this variable acceptable.

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3.3.6 Low Pressure Coolant Injection System Flow Residual Heat Removal System Flow Regulatory Guide 1.97 recommends instrumentation for these variables with a range from 0 to 110 percent of design flow. The licensee's instrumentation has a range frcm 0 to 105 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 pump runout flow rate during an accident.

The existing range is adeouate to provide the necessary accident and post-accident information. Tnerefore, these are acceptable deviations from Regulatory Guide 1.97.

3.3.7 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 flow measuring devices for this manually initiated system are not provided. However, the flow could be verified by the following:

1. Pump disenarge heaoer pressure

2. Pumo running light

3. Level of the SLCS storage tank 4 Reactivity change in the reactor as measured by neutron flux

5. Squib valve continuity and ready indicating lignts We find the above instrumentation valid as an alternative indication of SLCS flow.

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3.3.8 Standby Liouid Control System Storace Tank Level Regulatory Guide 1.97 recommends Category 2 instruraentation for this variable. This includes environmental qualification. The licensee's instrumentation for this variable is not environnentally qualified.

The licensee states that this instrumentation will be operating in a mild environment and that the design basis for the standby liquid control system recognizes that the system is designed as an alternate method of reactivity control without a concurrent LOCA or higher energy line break.

The licensee conforms to all the criteria (power supply, range, etc.)

identified under Category 2 instrumentation except for environmental qualification. This instrumentation is located in a mild environment.

Therefore, we find this instrumentation acceptable.

3.3.9 Residual Heat Removal (RHR) Service Water Flow 9egulatory Guide 1.97 recommends instrumentation with a range from 0 to :'.0 percent of design flow fer the variaele cooling water flow to engineered safety feature (ESF) system components. Reference 6 corrects an error in Reference 4, and states the range as 0 to 150 percent of design flow.

Based on this additional information, we find that tnis range is adequate for all accident and post-accident conditions. Therefore, the instrumentation provided by the licensee for this variable is acceptable.

3,3.10 Hioh Radioactisity Liouid Tank Level The licensee's instrumentation fer this variable has it's 31 splay in the radwaste control room. The licensee states that any tank inputs are isolated with an accident signal and that control roem operators must verify the tank storage capacity with the radwaste operators. Based on this, we find this deviation in display location acceptable.

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3.3.11 Reactor Building or Secondary Containment Area Radiation Regulatory Guide 1.97 recommends Category 2 instrumentation with a range of 10 ~1 to 10 R/hr for the Fit: Patrick Mark I containment. The licensee has supplied Category 3 instrumentation with a range of 10"I to 103 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 existing Category 3 instruments (rather than Category 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 seconcary containment by a combination of airborne activity samples and local radiological survey instruments for beta and gamma dose rates.

Additional instrumentation includes the plant noble gas effluent monitors, high-range radiatien monitors and the high-range effluent monitoring system.

The licensee states that the instrumentation for this variable is not needec, as the plant noble gas effluent monitors (which are Category 2 instrumentation) are more useful and practical in detecting or assessing primary contain. rent leakage. The licensee reports that the use of local radiation exposure rate monitors to detect breach or leakage through primary containment penetrations results in ambiguous indications. This is due to the radioactivity in the primary t.catainment, the radioactivity in the fluids flowing in emergency Core coolant system piping and the amount and location of fluid and electrical penetrations. The licensee concluces that the use of the plant noole gas effluent monitors is the prcoer way to accomplish the purpose of this variable. Therefore, the licensee concludes that the existing Category 3 instrumentation for tnis variable is adequate.

We find that the existing Category 3 instrumentation and ranges in corcert with the noble gas effluent monitors is acceptable.

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

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

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5. REFERENCES

1. NRC letter, D. G. Eisenhut to All Licensees of Operating Reactors, Applicants for Operating Licenses, and Holders of Constructicn Permits, "Supplement No. 1 to NUREG-0737--Requirements for Emergency l Response Capaoility (Generic Letter No. 82-33)," Decemoer 17, 1982. l f

2. Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During anc Following an Accident, Regulatory Guide 1.97 Revision 2. NRC, Office of Standarcs  ;

Development, December 1980. t

3. Clarification of TMI Action Plan Recuirements. Reouirements for i Emergency Responso Cacability, NUREG-0737, Supplement No. 1, NRC, .

OTfice of Nuclear Reactor Regulation, January 1983.

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4. Letter, New York Power Authority (C, A. McNeill, Jr.) to NRC ,

(O. B. Vassallo), "Supplement No. I to NUREG-0737 (Generic Letter  !

S2-33), Regulatory Guide 1.97, Revision 2, Implementation Repor* "

November 30,1984, JPN-84-77.  ;

5. Letter, New York Power Authority (J. C. Brons) to NRC (O. B. Vasta11o), j "Regulatory Guide 1.97 Post-accident Instrumentation," June 28, 1985, i JPN-85-53.

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6. Letter, New York Power Autnority, (J. C. Brons) to NRC (D. R. Muller),  !

"Emergency Response Capability--Conformance to Regulatory Guide 1.97, Revision 2," December 24, 1985, JPN-SS-91. I

7. Letter, New York Power Authority (J. C. Brons) su NRC (0, R. Muller),

"Regulatory Guide 1.97 Revision 2 Implementation Report-Correction Regarding Orywell Sump Level Instrumentation," Feervary 25, 1986, r JPN-S6-06. (

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8. Letter, New York Power Authority (J. C. Brons) to NRC, "Reactor Vessel Water Level Instrumentation," June 9, 1987, JPN-87-033. -

9. Analysis to Extend Ocerator Action Time for Alternate Shutcown Panels L in Suecort of Fit:Patrica. Comoltance to Aeoenoix R: General Electric  !

Nuclear Enorgy Business Operations Report M0E-137-0585, Revision 2, r ORF C61-00045, November 1985. l l

10. Instrumentation for Light-Water-Cooled Nuclear Power olants to Assess l Plant and Environs Concitions During ano Following an Accicent, I Regulatory Guide 1.97, Revision 3, NRC, Office of Nuclear Regulatory Research, May 1983.  ;

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CONFORMANCE TO REGULATORY GUIDE 1.97 --

FITZPATRICK

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Electrical. Instrumentation and Control Systems Evaluation Unit ****a'*'a EG&G Idaho, Inc.

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Division of Engineering and System Technology Technical Evaluation Report Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission . n ,co co. ... ,.- .-

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,J .#.. . .G , ,Jdp eeres -seat This EG5G Idaho. Inc. report reviews the submittals 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 acceotability is not provided are identified.

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