ML20235M118
| ML20235M118 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 06/30/1987 |
| From: | Udy A EG&G IDAHO, INC., IDAHO NATIONAL ENGINEERING & ENVIRONMENTAL LABORATORY |
| To: | NRC |
| Shared Package | |
| ML20235M106 | List: |
| References | |
| CON-FIN-A-6483, RTR-REGGD-01.097, RTR-REGGD-1.097 EGG-EA-7040, TAC-51090, NUDOCS 8707170060 | |
| Download: ML20235M118 (20) | |
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EGG-EA-7040 June 1987 i.
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INFORMAL REPORT T
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CONFORMANCE TO REGULATORY GUIDE 1.97 --
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U.S. NUCLEAR REGULATORY COMMISSION
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DISCLAIMER This book was prepared as an account of work sponsored by an agency of the United States Govemment. Neither the United States Government nor any agency thereof, nor any o' ;aeir employees, makes any warranty, express or imphed, or assumes any legal habihty or responsibility for the accuracy, completeness, or usefu! ness of any informatt.*ta, apparatus, product or process disclosed, or represents that its use would not ininnge pnvately owned ngers. References herein to any specife commercial product, process, or service by trade name, trademark, manufacturer, or otherwise, does not necessanly constitute of imply its endorsement, recommendation, or favonng by the United States Government or any agency thereof. The views and opinions of authors expressed herein oo not necessanly state or reflect those of the United States Government or any agency thereof.
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1 EGG-EA-7040 l
l TECHNICAL EVALUATION REPORT 1
CONFORMANCE TO REGULATORY GUIDE 1.97--
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FITZPATRICK l
l Occket No. 50-333 i
A, C. Udy Published June 1987 I
l Idaho National Engineering Laboratory EG&G Idaho, Inc.
Prepared for the U.S. Nuclear Regulatory Commission Washington, D.C.
20555 Under DOE Contract No. DE-AC07-76ID01570 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. FitzPatrick 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 acceptability is not provided are identified.
Docket No. 50-333 TAC No. 51090 11
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FOREWORD l
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 Regulation, Division of Engineering and System Technology, by EG&G Idaho, Inc., NRR and I&E Support Branch.
The U.S. Nuclear Regulatory Commission funded the work under authorization B&R 20-19-10-11-3, FIN No. A6483.
1 Docket No. 50-333 TAC No. 51090 iii
CONTENTS l
ABSTRACT.............................................................
ii FOREWORD..........................................................
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i 1.
INTRODUCTION....................................................
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2.
REVIEW REQUIREMENTS.............................................
2 3.
EVALUATION.......................................................
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3.1 Adherence to Regulatory Guide ~1.97.........................
4 3.2 Type A Variables...........................................
4 3.3 Exceptions to Regulatory Guide 1.97........................
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4.
CONCLUSIONS.....................................................
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REFERENCES.....................................................
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p CONFORMANCE TO REGULATORY GUIDE 1.97--
FITZPATRICK I
1.
INTRODUCTION I
On December 17, 1982, Generic Letter No. 82-33 (Reference 1) was issued by D. G. Eisenhut, Director of the Division of Licensing, Nuclear o
Reactor Regulation, to all licensees of operating reactors, applicants for operating licenses, and holders of construction permits.
This letter included additional clarification regarding Regulatory Guide 1.97, i
Revision 2 (Reference 2), relating to the requirements for emergency response capability.
These requirements have been published as Supplement No. 1 to NUREG-0737, "TMI Action Plan Requirements" (Reference 3).
, The New York Power Authority, the licensec for the James A.
FitzP& trick Nuclear Power Plant, provided a response to Section 6.2 of the generic 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 February 25, 1986 (Reference 7).
This recort provides an evaluacion of these submittals.
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2.
REVIEW REQUIREMENTS 1
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 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 1.
Instrument range i
2.
Environmental qualification 3.
Seismic qualification 4.
Quality assurance 5.
Redundance and sensor location i
6.
Power supply 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.
l Subsequent to the issuance of the generic letter, the NRC held regional meetings in February and March 1983, to answer licensee and applicant questions and concerns regarding the NRC policy on thia subject.
At these meetings, it was noted that the NRC review would only afdress exceptions taken to Regulatory Guide 1.97.
Wnere licensees or applicants explicitly state that instrument systems conform to the regulatorf guide, it was noted that no further staff review would be necessary.
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this report oaly 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.
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EVALUATION The licensee provided a response to Section 6.2 of NRC Generic l
I Letter 82-33 on November 30, 1984 This response describes the licensee's position on post-accident monitoring instrumentation. Additional
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information was provided on June 28, 1985, December 24, 1985 and February 25, 1986.
This evaluation is based on these submittals.
3.1 Adherence to Regulatory Guide 1.97 l
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 being made to implement Regulatory Guide 1.97 will be complete within 30 days after the end of the 1987 (Reload 8/ Cycle 9) refueling outage or April 1, 1988, whichever is later.
Any required modifications to the instrumentation for the standby liquid control system will be scheduled as part of the I
anticipated transient without 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
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guide are noted in Section 3.3.
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3.2 Tyoe A Variables Regulatory Guide 1.97 does not specifically identify Type A variables, i.e., those variables that provide the information 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 4
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 hydrogen concentration 11.
Containment oxygen concentration
- 12. Core spray system flow 13.
Core spray system pressure l
l The above variables either meet or will be modified to meet the Category 1 requirements consistent with the requirements for Type A variables.
3.3 Exceptions to Reculatory 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 licens a has supplied Category 2 instrumentation for this variable, some of which is not environmentally qualified.
Regulatory Guide 1.97 specifies Category 1 instrumentation.
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I The licensee states that rigorous Category 1 design' criteria is not justified.
They also state that modifications are not needed since boron sampling and control rod drive (CRD) position indication provide adequate backup.
These are not Category 1.
The licensee also states that they consider that none of the neutron flux equipment requires qualification per the Environmental Qualification Rule, 10 CFR 50.49.
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 1 system that meets all the criteria of Regulatory Guide 1.97 is an industry development item.
Based on our review, we conclude that the existing instrumentation is acceptable for
'Tterim 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 j
Category 1 instrumentation when it becomes available.
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l 3.3.2 Drywell Sump Level Drywell Drain Sumps Level l
Regulatory Guide 1.97 recommenas Category 1 instrumentation with l
indication from the bottom to the top of each sump. The licensee has Category 1 drywell level instrumen'.ation with a range of 0 to 100 feet.
A single narrow range instrument measuring from 19 to 45 inches is provided for each of two sumps.
These sump transmitters are not environmentally l
qualified.
This instrumentation does not cause any automatic or operator initiated safety-related functions. The sump systems are automatically l
1 isolated by an accident signal as part of containment isolation. This
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prevents the pump-out of the sump contents. The drywell level instrumentation is usable after these relatively small sumps are full.
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Me 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 j
harsh environments during operation, (b) for larger leaks, the sumps fill promptly and the sump drain lines isolate due to the increase in drywell pressure, thus negating the drywell equipment drain sump level and drywell floor drain sump level instr'amentation, 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.
1 3.3.3 Radiation Level in Circulating Primary Coolant The licensee indicates that measurements to indicate fuel cladding failure are provided by the following instruments:
1.
High radiation sampling system 2.
Londenser off gas radiation monitors 1
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Main steamline radiation monitors i
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Primary containment radiation monitors 5.
Containment hydrogen concentration monitors 6.
Area radiation monitors l
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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l 3.3.4 Radiation Exposure Rate Regulatory Guide 1.97, Revision 2, specifies Category 2
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instrumentation for this variable.with a range 10 to 10 R/hr.
The licensee has provided instrumentation for this variable with a range of
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~4 10 to 10 mR/hr (10 to 1 R/hr), located in thirty areas in and around the plant.
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 recommended by Regulatory Guide 1.97 are acceptable.
The licensee states that it is impractical to f
detect prima 0 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 radiological survey instruments for beta and gamma dose rates.
Regulatory Guide 1.97, Revision 3 (Reference 8), changes this variable to Category 3.
Therefore, the only deviation at the 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 detectors would be in radiation fields of less than 1 R/hr and would be on 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 short or medium term.
Any personnel entry would be preceded by surveys with portable instrumentation.
Should the instrument range be exceeded, backup instrumentation, including portable survey instruments, airborne sampling, the high' range u
effluent monitoring system, high range radiation monitors located in other 8
areas of-the plant, and the plant noble gas effluent monitors, will be used
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by the licensee for long term release surveillance.
Based on this, we find the licensee's instrumentation for this variable acceptable.
3.3.5 Low Pressure Coolant Injection System Flow Residual Heat Removal System Flow l
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 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 pump runout ficw rate during an accident.
The existing range 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 Liouid 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:
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Pump discharge header pressure 2.
Pump 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 lights 9
He find the above instrumentation valid as an alternative indication of SLCS flow.
3.3.7 Standby Licuid Control System Storage Tank Level Regulatory Guide 1.97 recommends Category 2 instrumentation for this '
l variable.
This includes environmental qualification.
The licensee's instrumentation for this variable 1: not environmentally qualified.
4 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.
1 3.3.8 Residual Heat Removal (RHR) Service Water Flow Regulatory Guide 1.97 recommends instrumentation with a range from 0 to 110 percent of design flow for the variable 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 this range is adequate for all accident and post-accident conditions.
Therefore, the instrumentation provided by the licensee for this variable is acceptable.
3.3.9 High Radioactivity liquid 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 signal and that control room operators must 10 i
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verify. the tank storage capacity with the radwaste operators.
Based on this, we find this deviation in display location acceptable.
3.3.10 Reactor Buildina or Secondary Containment Area Radiation I
Regulatory Guide 1.97 recommends Category 2 instrumentation with a
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l range of 10 to 10 R/hr for the FitzPatrick Mark 1 containment.
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licensee has supplied Category 3 instrumentation with a range of 10 to-10 mR/hr.
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l 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 i
1 secondary containment by a combination of airborne activity samples and j
local radiological survey instruments for beta and gamma dose rates.
Additional instrumentation includes the plant noble gas effluent monitors, i
high-range radiation monitors and the high-range effluent monitoring system.
The licensee states that the instrumentation for this variable is not needed, as the plant noble gas effluent monitors (which are Category 2 instrumentation) are more useful and practical in detecting or assessing primary containment 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.
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due to the radioactivity in the primary containment, the radioactivity in the fluids flowing in emergency core coolant system piping and the amount i
and location of fluid and electrical penetrations.
The licensee concludes that the use of the plant noble gas effluent monitors is the proper way to accomplish the purpose of this variable.
Therefore, the licensee concludes that the existing Category 3 instrumentation for this variable is adequate.
We find that the existing Category 3 instrumentation and ranges in concert with the noble gas effluent monitors is acceptable.
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CONCLUSIONS l'ased on our review, we find that the licensee either conforms to or is jurtified in deviating from Regulatory Guide 1.97, with the following exceptions:
1.
Neutron flux--the 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 I
1.
NRC letter, D. G. Eisenhut to All Licensees of Operating Reactors, Applicants for Operating Licenses, and Holders of Construction 4
Permits, " Supplement No I to NUREG-0737--Requirements for Emergency I
Response Capability (Generic Letter No. 82-33)," December 17, 1982.
2.
Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident, Regulatory Guide 1.97, Revision 2, NRC, Office of Standards Development, December 1980.
3.
Clarification of TMI Action Plan Requirements, Requi'rements for i
Emergency Response Capability, NUREG-0737, Supplement No. 1, NRC, Office of Nuclear Reactor Regulation, January 1983.
4.
Letter, New York Power Authority (C. A. McNeill, Jr ) to NRC (D. B. Vassallo), " Supplement No. I to NUREG-0737 (Generic Letter 82-33), Regulatory Guide 1.97, Revision 2, Implementation Report,"
November 30, 1984, JPN-84-77.
5.
Letter, New York Power Authority (J. C. Brons) to NRC (D. B. Vassallo),
" Regulatory Guide 1.97 Post-accident Instrumentation," June 28, 1985, JPN-85-53.
6.
Letter, New York Power Authority, (J. C. Brons) to NRC (D. R. Muller),
" Emergency Response' Capability--Conformance to Regulatory Guide 1.97,.
Revision 2," December 24, 1985, JPN-85-91.
7.
Letter, New York Power Authority (J. C. Brons) to NRC (D R. Muller),
" Regulatory Guide 1.97 Revision 2 Implementation Report-Correction Regarding Drywell Sump Level Instrumentation," February 25, 1986, JPN-86-06.
8.
Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident, Regulatory Guide 1.97, Revision 3, NRC, Office of Nuclear Regulatory Research, May 1983.
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FITZPATRICK e oaf t REPORT COMPL8T(Q MONTH VEAR June 1987
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P. O. Box 1625 A6483 Idaho Fall.s, ID 83415 iO. SPON5ORING ORGANeJ ATION NAME ANO MAILING ACORE55 treusede Cases I14. TYPE OF REPORT Division of Engineering and System Technology Technical Office of Nuclear Reactor Regulation Evaluation Reoort U.S. Nuclear Regulatory Commission Washington, DC 20555 12 SUPPLE ME N f AR Y NOT E R 13 AuSTR ACT (100 we,es or <eans This EG&G 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 acceptability is not provided are identifed.
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