ML20132D187
| ML20132D187 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 06/30/1985 |
| From: | Udy A EG&G IDAHO, INC. |
| To: | NRC |
| Shared Package | |
| ML20132D191 | List: |
| References | |
| CON-FIN-A-6483, RTR-NUREG-0737, RTR-NUREG-737, RTR-REGGD-01.097, RTR-REGGD-1.097 GL-82-33, NUDOCS 8507100210 | |
| Download: ML20132D187 (19) | |
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1 CONFORMANCE TO REGULATORY GUIDE 1.97 EDWIN I. HATCN NUCLEAR PLANT, UNIT NOS. 1 AND 2 A. C. Udy Published June 1985 EG&G Idaho, Inc.
Prepared for the U.S. Nuclear Regulatory Commission Washington, D.C.
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Under DOE Contract No. DE-AC07-76ID01570 FIN No. A6483 k
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l ABSTRACT This EG&G Idaho, Inc., report reviews the submittals for Regulatory Guide 1.97 for Unit Nos. I and 2 of the Edwin I. Hatch Nuclear Plant and identifies areas of nonconformance to the regulatory guide.
Exceptions to Regulatory Guide 1.97 are evaluatec and those areas where sufficient basis
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for acceptability is not provided are identified.
i FOREWORD This report 1.s supplied as part of the " Program for Evaluating Licensee /Appiteant Conformance to AG 1.97," being conducted for the U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, Division of Systems Integration, by EG&G Idaho, Inc., NRC Licensing Support Section.
The U.S. Nuclear Regulatory Commission funded the work under authorization B&R 2C-19-10-11-3.
l Docket Nos. 50-321 and 50-366 TAC Nos. 51096 and 51097 i
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CONTENTS ABSTRACT..........'...................................................
11 FOREWORD..............................................................
11 1.
INTRODUCTION.....................................................
1 2.
REVIEW REQUIREMENTS..............................................
2 3.
EVALUATION.......................................................
4 3.1 Adherence to Regul atory Guide 1. 97.........................
4 3.2 Type A Variables...........................................
4 3.3 Exceptions to Regulatory Guide 1.97........................
5 4;
CONCLUSIONS.y...................................................
15 5.
REFERENCES.......................................................
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CONFORMANCE TO REGULATORY GUIDE 1.97 EDWIN I. HATCH NUCLEAR PLANT, UNIT NOS. 1 AND 2 1.
INTRODUCTION 09 December 17, 1982, Generic Letter No. 82-33 (Reference 1) was issuec by D. G. Eisenhut, Director of the Division of Licensing, Nuclear 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, Revision 2 (Reference 2), relating to the requirements for emergency response capability. These requirements have been published as Supplement No. I to NUREG-0737, "TMI Action Plan Requirements" (Reference 3).
Georgia power Company, the Mcenses for the Edwin I. Hatch Nuclear Plant, provided a response to Itam 6 of the generic letter on February 21, 1984 (Reference 4). Additional information was provided on April 29, 1985 (Reference 5).
i This report provides an evaluation of this material.
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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 Regulatory Guide 1.97 as applied to emergency f
response facilities.
The submittal should include documentation that provices the following information for each variable shown in the applicable table of Regulatory Cuide 1.97.
1.
Instrument range 2.
Environmental qualification t
3.
Seismic qualification 4.
Quality assurance i
5.
Redundance and sensor location 6.
Power supply 7.
Location of display 8.
Schedule of installation or upgrade t
l The submittal should identidy deviations from the guidance in the regulatory guide and provida supporting justification or alternatives.
l Subsequent to the issuance of the generic *1etter, the NRC held regional meetings in February and March 1983, to answer licensee and applicant questions and corcerns regarding the NRC policy on this subject.
I At these meetings, it was noted that the NRC review would only address
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exceptions taken to Regula:ory Guide 1.97.
Where Itcensees or applicants l
explicitly state that instrument systems conform to the regulatory guide,
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it was noted that no further staff review would be necessary.
Therefore, 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.
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3.
EVALUATION The licensee provided a response to Item 6 of NRC Generic Letter 82-33, on February 21, 1984, and additional information on April 29, 1935.
These submittals describe the licensee's position on post-accident monitoring instrumentation.
This evaluation is based on that material.
3.1 Adherence to Reculatory Guide 1.97 The licensee has provided a review of their post-accident monitoring instrumentation that compares the instrumentation characteristics against the recommendations ef Regulatory Guide 1.97, Revision 2.
The licensee's reports identify what presently installed instrumentation meets the recommendations, where updated instrumentation will be installed to meet or exceed the recommendations and the licensee's position, justification or planned enhancements for deviations from the regulatory guide.
Therefore, we conclude that the licensee has provided an explicit ec=mitment on conformance to Regulatory Guide 1.97.
Exceptions to and deviations from the regulatory guide are noted in Section 3.3.
3.2 Tyoe A Variables Regulatory Guide 1.97 does not specifically identify Type A variables, i.e., those variables that provide 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.
Residual heat removal service water flow 2.
Hydrogen content in the drywell 3.
Oxygen content in the drywell 4.
Reactor pressure vessel pressure 4
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Reactor pressure vessel level 6.
Drywell temperature in the vicinity of the reactor pressure vessel level instrumentation reference leg i
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. Suppression poel temperature 8.
Diesel generator output voltage i
9.
Diesel generator output current 10.
Diesel generator output power 11.
Diesel generator battery voltage i
The above instrumentation meets the Category 1 recommendations consistent with the requirements for Type A variables.
3.3 Exceptions to' Regulatory Guide 1.97 I
The licensee identified deviations and exceptions from Regulatory Guide 1.97.
These are discussed in the following paragraphs.
I 3.3.1 Neutron Flux i
l Regulatory Guide 1.97 recommends Category 1 instrumentation for this variable.
The Itcensee has provided instrumentation that is not Category 1.
The licensee has stated that it does not meet the recommendations for environmental and seismic qualificatien, that th four source range channels have a common recorder, and that the six average power range (APRM) channels have four recorders between them, 4
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These channels also have dedicated hardwired indicators.
This information is also displayed on the safety parameter display system (SPDS).
Therefore, we find that the shared recorders are acceptable.
Tne licensee indicates that plant emergency operating procecures require the control room operator to take action should a validated neutron flux signal not be availaole. This action may include the actuation of the standby liquid control system.
The procedures direct the operator to take all action available to him for reactivity control.
The licensee states that the importance to safety--in terms of prevention and mitigation of a reactivity associated accident--of the key variable of reactivity justifies Category 3 type instrumentation with appropriate emergency procedures.
In the process of our review of the neutron flux instrumentation for boiling water reactors, we note that most mechanical drives of the detectors have not satisfied the environmental qualification recuirement of Regulatory Guide 1.g7.
A Category 1 system that meets all ne criteria of Regulatory Guide 1.97 is 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 and evaluate the installation of improved instrumentation when it becomes available.
3.3.2 RCS Soluble Boron Concentration Regulatory Guide 1.97 recommends instrumentation for this variable with a range of 0 to 1000 parts per million.
The licensee has on-line instrumentation with a range of 100 to 6500 parts per million.
Offsite grab sample analysis is also available.
The licensee takes exception to the guidance of Regulatory Guide 1.97 with respect to post-accident sampling capability.
This exception goes beyond the scope of this review and is being addressed by the NRC as part of the review of NUREG-0737, Item II.B.3.
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3.3.3 Coolant Level in the Reactor
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Regulatory Guide 1.97 recommends redundant Category 1 in trumentation
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for this variable with a range extending from the bottom of the core support plate to the centerline of the main steamline or the top of the j
vessel (whichever,isless).
The licensee has supplied instrumentation that covers from the ccre support plate to 76 inches above the top of the
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vessel; however, the shutdown vessel flooding range is not redundant.
Thus, the level from 60 inches above instrument zero to the centerline of the main steamline is not covered by redundant instruments.
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4 The licensee states that there are redundant high water level trips as i
part of the Category I normal operating range level channels.
These are set at 58 inches above instrument zero and isolate the main steamlines should the reactor vessel water level exceed this level. Also, injection water into the reactor vessel is terminated by these same trips.
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manual and automat 1c safety functions occur in the range covered by the normal operating range level channels.
q In order to comply with the single failure requirement of Regulatory Guide 1.97, an additional vessel penetration would be needed.
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centerline of the main steam 11nes is used as the upper end of the Regulatory Guide 1.97 recommended range in order to provide the operator I
with an indication of whether the reactor coolant has reached, and spilled into, the main steamlines.
i As previously noted, all manual and automatic safety functions are j
initiated in the range covered by the safety-related normal operating range l
l 1evel instrumentation.
The licensee has concluded that the existing i
j reactor coolant level instrumentation meets the intent of the regulatory guide and that no improvement in plant safety would be achieved by l
installing a redundant flooding range channel.
1 a sig f at nereas ant s f c ne h
asn channel of instrumentation above the normal operating range is acceptable.
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3.3.4 Drywell Sume Level Drywell Drain Sumos Level Regulatory Guice 1.97 recommends Category 1 instrumentation for these variables.
The licensee has supplied Category 3 instrumentation consisting of continuous level indication, rate of rise indication and nigh and nign-high level alarms (each alarm starts one sump pump).
Timers indicate the duration of sump pump operation for estimating the amount of leakage.
No safety-related system is actuated either automatically or manually as a result of the sump level.
The crywell sump systems are automatically isolated at the primary containment penetration should an accicent signal I
occur.
We conclude that the alternate instrumentation supplied by the licensee will provide appropriate monitoring for the parameters of c:ncern.
This is based on (a) for small leaks, the alternate instrumentation is not expected to experience harsh environments during cperation, (b) for larger leaks, the sumps fill promptly and the sump drain lines isolate due to the increase in drywell pressure, thus negating the crywell sump level and drywell drain sumps level instrumentation, (c) this instrumention neither autematically initiates nor alerts the operator to initiate operation of a safety-related system in a post-accident situation.
Therefore, we find the alternate Category 3 instrumentation provided acceptacle.
3.3.5 Radiation Level in Circulatino primary Coolant The licensee indicates that radiation level measurements to indicate fuel cladding failure are provided by the following:
1.
Condenser off gas radiation monitors 2.
Main steamline radiation monitors 8
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Primary containment radiation monitors p
4 Post-accident sampling system.
P Based on the alternate instrumentation provided by the licensee, we conclude that the instrumentation supplied for this variable is adequate, and therefore, ac:eptable.
3.3.6 Radiation Exposure Rate Regulatory Guide 1.'97, Revision 2, specifies Category 2 t
instrumentation for this variable with a range of 10I 4
to 10 R/hr.
The licensee has provided Category 3 radiation exposure rate monitors that have ranges that are lower than recommended by Regulatory Guide 1.97.
These are stated as being influenced by piped radioactive fluids.
The licensee concludes that this makes it impractical to detect primary containment
. breach by use of these monitors, and Category 3 instrumen,ation is suitable for this application.
The licensee states that the plant noble gas effluent monitors are j
adequate to monitor the effluent from the secondary containment.
The licensee determines the habitability of secondary containment by a i
combination of atmosphere sampling and portable radiation survey instruments, not fixed location radiation exposure' rate meters.
Regulatory Guide 1.97, Revision 3 (Reference 6), changes this variable to Category 3.
Therefore, the only deviation of the Hatch station for this variable is the range supplied for a given location.
The licensee has not shown any analysis of radiation levels expected for the monitor location.
The licensee indicates that local radiation exposure rate level increases are alarmed, and that those areas are then surveyed with portable instrumentation.
The Itcensee also states that entry into secondary containment is not expected for at least 30 days after an accident.
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Should the itstrumentatien range be exceeded, backup instrumentation, including portable survey instruments, atmosphere sampling and radiation monitors at the e, tin stack and the reactor building vents will be used by the licensee for release detection and assessment and long term release surveillance.
Bated on this, we find the licensee's instrumentation for this variable acceptable.
3.3.7 Sueorea:iop Chameer Scray Flow Drywell So; ay Flow Regulatory Guide 1.97 specifies Category 2 instrumentation for these variables with a range from 0 to 110 percent of design flow.
These two sprays are not provided with dedicated flow measurement channels.
- Instead, the residual heat removal flow element common to these two sprays and the containment spray is used.
The flow is controlled by the positten of a tnrettling valve.
Valve lineup, observable in :ne control room for the suppression chamcer spray, drywell spray, and the containment spray, shows which sprays have the indicated flow.
Pressure and temperature change: in the drywell and suppression chamber determine the effectiveness of the spray.
The licensee concludes that this flow measurement, and the suppression chamber and drywell temperature and pressure, accurately and reliably measure the effectiveness of the drywell and suppression chamber spray, We find this alternate method of monitoring this variable acceptable.
3.3.8 Drywell Atmosehere Temocrature Regulatory Guide 1.97 recommends instrumeqtation for this variable with a range of 40 to 440*F.
Unit 1 meets this recommendation.
Unit 2 has instrumentation for this variable with a range of 0 to 400'F.
The licensee states that the maximum drywell average temperature during a design basis event would be less than 340*F.
The maximum temperature for continued 10
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operation is 135'F, which provides a significant margin between the t
temperatures expected in the drywell and the measurement capabilities of I
the monitoring instrument.
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We find that the temperature range of this instrumentation is adequate l
to monitor this variable during all accident and post-accident conditions.
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i 3.3.9 Standbr.tisit Control System (SLCS) Flow I
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The licensee has elected not to implement this variable as recommended i
The justification given by the Itcensee is (a) the SLCS pump-discharge header pressure indication provides indication that the SLCS pump is operating, (b) the level indication in the sodium j
pentaborate soluttop. storage tank gives indication that flow is occurrirg, j
(c) the reactivity change in the reactor as measured by neutron flux f t an indication of flow, (d) the motor indicating lights and pump discharge j
pressure show system operation, and (e) the squib valve continuity indicating lights are an indication of flow.
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1 3.3.10 Cooline Water Temperature to ESF System Comoonents i
j The licensee states that remote cooling water temperature indication i
is not provided.
Each area with essential coolers is provided with local i
area (air) temperature indication which is available in the main control f
These temperature indications in conjunction with plant service room.
j water flow, and cooler status indication provide the operator with adequate indication as to the status of the cooling capaht11 ties to the ESF System components.
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The Final Safety Analysis Report (FSAR, Reference 7) for Unit 1 l
(Unit 2 is similar) Section 10.7 (Section 9.2-Unit 2) describes the plant j
service water system as a once through system.
The cooling water source is i
the Altamaha River.
Thus, the temperature of the cooling water is f
essentially the river water temperature.
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i We find that the diverse indication provided to monitor the operation of the plant service water system is acceptable for this variable.
3.3.11 Coolinc Water Flew to ESF System Comconents Regulatory Guide 1.97 recommends instrumentation for this variable I
with a range of 0 to 110 percent of design flow.
The licensee does not provide instrumentation that is a direct indication for this variable,
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i relying instead on the plant service water pump output pressure and system pressure instrumentation, pump running indication, valve position j
indication and equipment room temperature to indicate system operation, i
System leaks are detected by the equipment and floor drainage system and room temperature instrumentation.
I The licensee's emergency operating procedures are written around this diverse instrumentation.
The instrumentation will show loss of flow due to j
blockage or piping failure.
The licensee also states that any need to i
upgrade this Category 3 instrumentation will be addressed during the assessment phase of the detailed control room design review, Section 5 of j
NUREG-0737, Supplement No. 1.
Based on the diversity of the j
instrumentation provided, we find the instrumentation provided for this variable acceptable for interim operation.
The detailed control room design review is beyond the scope of this review.
The adequacy of this i
instrumentation in the long term will be addressed by the NRC in that review.
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3.3.12 Reactor Buildino or Secondary Containment Area Radiation i
Regulatory Guide 1.97 recommends Category *2 instrumentation for this l
variable with a range of 10 to 10 R/hr for Hatch's Mark I
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-2 containment. The licensee has some instruments with a range of 10 g,
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~1 10+2 mR/hr(10 to 10 R/hr), and some instruments with a range of 4
~3 I to 10 mR/hr(10 to 10 R/hr). All these instruments are Category 3 rather than the recommended Category 2.
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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 containment, the radioactivity in the fluids flowing in emergency core coolant system piping and the amount and location of fluid and electrical penetrations.
The licensee concludes that the use I
of the plant ncble 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 the existing Category 3 instrumentation and ranges will monitor the ranges of concern and, therefore, are acceptable.
3'.3.13 Noble Gas and Vent Flow Rate--Common Plant Vent Regulatory Guide 1.97 recommends Category 2 instrumentation for this
-6 variable with a range of 10 to 10*4 vC1/cc.
Each Hatch unit has normal and wide range instrumentation with dedicated indicators.
The safety parameter display system integrates the two sets of instrumentation for a composite range of 10 to 10+5 uti/cc.
The wide range instruments are Category 2.
The normal range instruments, which provide information for levels of less than 5 x 10-3 uC1/ce, are not environmentally qualified.
I Our examination of this instrumentation shows the normal range detectors located on the off gas stack in a mild environment.
j The indicators are in the control room. As the instrumentation is located in a mild environment, the Environmental Qualification Rule,10 CFR 50.49, is not applicable.
Therefore, we find the provided instrumentation acceptable.
3.3.14 Estimation of Atmospherte Stability Regulatory Guide 1.97 recommends instrumentation for this variable with a range of -9 to +18'F or an analogous range for alternative stability analysis.
The licensee has supplied instrumentation with a range of -10 to I
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-10*F.
The licensee justifies this, indicating that the range is based on RG 1.23, Rev.1, Table 1, Classification of Atmospheric Stability by Temeerature Changa With Height.
Table 1 of Regulatory Guide 1.23 (Reference 7) provides seven atmospheric stability classifications based on the difference in temperature per 100 meters elevation change.
These classifications range from extremely unstable to extremely stable.
Any temperature difference greater than +4*C or less than -2*C does nothing to tne stability classification.
The licensee's instrumentation includes this range.
Therefore, we find that'the instrumentation provided is acceptable to determine the atmospheric stability.
I 3.3.15 Accident Samoling (primary Coolant. Containment Air and Sumo The licensee deviates from this recommendation in two areas.
- First, their analysis capability is offsite, backed up with online equipment for i
boron and chloride content, hydrogen concentration and pH.
Thus, gross activity, gamma spectrum and oxygen content do not have.any onsite analysis l
capability.
Second, the licensee does not sample the sump. A sample is taken from the residual heat removal system which takes suction from the suppression pool which accepts overflow from the containment sump. A sample from the reactor coolant system can also be taken and used as representative of the suppression pool, as the suppression pool is the source of makeup water.
Additionally, the licensee has not indicated that sampling capability for the auxiliary building sumps is part of the station design.
The licensee takes exception to Regulatory Guide 1.97 with respect to post-accident sampling capability. This exception goes beyond the scope of this review and is being be addressed by the NRC as part of their review of NUREG-0737, Item II.B.3.
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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 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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REFERENCES 1.
NRC letter, D. G. Eisenhut to All Licensees of Operating Reactors, Applicants for Operating Licenses, and Holders of Construction Permits, " Supplement No. I to NUREG-0737--Recuirements for Emergency Response Capability (Generic Letter No. 82-33)," December 17, 1982.
2.
Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant ano Environs Concitions During and Followine an Accicent, Regulatory Guice 1.97, Revision 2, U.S. Nuclear Regulatory Commission (NRC), Office of Standards Development, December 1950.
3.
Clarification of TMI Action Plan Recuirements. Recuirements for Emergency Resconse Caca0111ty, NUREG-0737, Supplement No. 1, NRC, Office of Nuclear Reactor Regulation, January 1983.
4.
Georgia Power Company letter, L. T. Gucwa to Director of Nuclear Reactor Regulation, NRC, " Submittal of Report on Regulatory Guide 1.97," February 21, 1984, NEE-84-071.
1 5.
Georgia Power Company letter, L. T. Gucwa to Director of Nuclear Reactor Eagulation, NRC, " Response to Regulatory Guide 1.97 SER Open Items," Acril 29, 1985, NED-85-306.
6.
Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant anc Environs Concitions During and Following an Accicent, Regulatory Guice 1.97, Revision 3, NRC, Office of Nuclear Regulatory Research, May 1983.
7.
Onsite Meteorolooical Programs, Regulatory Guide 1.23 (Safety Guide),
NRC, February 17, 1972 or Meteorological Programs in Suecort of Nuclear Power Plants, Proposec Revision 1 to Regulatory Guice 1.23, NRC,:0ffice of Standards Development, September 1980.
8.
Final Safety Analysis Reoort, Edwin I. Hatch Nuclear Plant. Unit 1, Georgia Power Company, Atlanta, GA.
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