ML20214E343

Rev 1 to Conformance to Reg Guide 1.97,Nine Mile Point Nuclear Station,Unit 1

ML20214E343
Person / Time
Site:	Nine Mile Point
Issue date:	01/31/1986
From:	Udy A EG&G IDAHO, INC.
To:	NRC
Shared Package
ML17055C650	List: ML20214E343
References
CON-FIN-A-6483, RTR-REGGD-01.097, RTR-REGGD-1.097 EGG-EA-6880, EGG-EA-6880-R01, EGG-EA-6880-R1, TAC-51109, NUDOCS 8603170518
Download: ML20214E343 (23)

Text

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Attachment to

. , Safety Evaluation w

CONFORMANCE TO REGULATORY GUIDE 1.97 '

NINE MILE FOINT NUCLEAR STATION, UNIT NO. 1 A. C. Udy Published January 1986 EG&G Idaho, Inc.

Idaho Falls, Idaho 83415 Prepared for the U.S. Nuclear Regulatory Commission Washington, D.C. 20555 dnder DOE Contract No. DE-AC07-76ID01570 FIN No. A6483 e

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ABSTRACT This EG&G Idaho, Inc., repcrt reviews the submittal for Regulatory Guide.l.97, Revision 2, for the Nine Mile Point Nuclear Station, Unit No. 1.

• Any exceptions to Regulatcry Guide 1.97 are evaluated.

FOREWORD This report is suppliec as part of the " Program for Evaluating Licensee /A:plicar.t Cor.formance to RG 1.97," being conducted for the U.S.

Nuclear Regulatory Com.-issien, Office of Nuclear Reactor Regulation, Division of PWR Licensing - A, by EG&G Idaho, Inc., NRR and I&E Support Branch.

The U.S. Nuclear Regulatory Ccmmission funded the work under authori:ation B&R 20-19-10-11-3.

Docket No. 50-223 '-

TAC No. 51109 e

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CONTENTS

, AE5 TRACT .............................................................. 11

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FOREWGRD .............................................................. 11 2

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JNTRODUCTION .................................................~.... I 2.

REVIEW REQUIREMENTS .............................................. 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 . . . . . . . . . . . . . . . . . . . . . . . . . 4 4.

CON C LU S I Oil s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18 5.

REFERENCES ........................................................ 19 p

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• CONFORMANCE TO REGULATORY GUIDE 1.97 NINE MILE POINT NUCLEAR STATION, UNIT NO. 1

1. INTRODUCTION On December 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 reactors, applicants for operating licenses and holders of construction permits. This letter included additional clarification regarding Regulatory Gul'de 1.97, 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).

Niagara Mohawk Power Corporation, the licensee for the Nine Mile Point Nuclear Station, provided a response to Section 6.2 of the generic letter on April 2, 1984 (Reference 4). Additional information was submitted on October 18, 1985 (Reference 5) and December 6, 1985 (Reference 6).

This report provides an evaluation of this material.

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2. REVIEW REQUIREMENTS Section 6.2 cf NUREG-0737, Supplement No. 1, sets forth the '

documentation to be submitted in a report to the NRC describing how the licensee comolies with Regalatory Guide 1.97 as applied to emergenci response facilities. The submittal shculd include documentation that provides the following information for each variable shown in the applicaole table of Regulatory Guide 1.97. ,

1. Instrument range

2. Environmental qualification

3. Seismic cualificction

4. Quality assurance

5. Redundance and sensor location

6. Power supply

7. Location of display S. Schedule of installation er upgrade The submittal should identify deviations from the regulatory guide and provide supporting justification or alternatives.

Subsecuent to the issuance of the generic letter, the NRC held regional meetings on February and March 1983, to answer licensee and applicant questions and ccccerns regarding 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. Where licensees or applicants explicitly state that instrument systems conform to the regulatory guide it was noted -

that no further staff review would be necessary. Therefore, this 2

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. 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 Generic Letter 82-33 on April 2, 1984. The response describes the licensee's position on post-accident nonitoring instrumentation. Additional infer =ation was provided on October 18, 1985 and December 6, 1935. Tnis evaluation is cased on this material.

3.1 AdherenceteReculatoryGuide1.9h The licensee has provided a review of their post-accident monitcring instruceitation that ccmpares the instrunentation characteristics against the recommencations of Regulatcry Guide 1.97, Revision 2. Tne licensee concludes that those Regulatory Guide 1.97 variables that are applicable to the Nine Mile Point Nuclear Station, Unit No.1, are monitored and displayed in the control room. Those modifications ider.tified by the licensee to bring the instrum.entation into compliance with Regulatory Guide 1.97 have been completed. 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 guide are noted in Section 3.3.

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

The licensee has cetermined that there are no variables that are required for manual safety res;onse; that all safety-related plant prctective actions are done automatically. Therefore, there are no Type A variables at Unit No. I of the Nine Mile Point station.

3.3 Exceptions to Reculatory Guide 1.97 The licensee identified deviations and exceptions te Regulatory Guide 1.97. These are discussed in the following paragracns.

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3.3.1 Neutron Flux

' Regulatory Guide 1.97 recommends environmentally qualified instrumentation. The licensee has instrumentation for this variable that has not been environmentally qualified. The licensee states that protective action is initiated prior to exposure to a harsh environient.

4 The licensee states that the source and intermediate range monitor (SRM and IRM respectively) detectors are driven into the core immediately as a result of the scram signal. As neutron flux decays, the recorders are re-ranged and the detectors are inserted. Exposure to the post-accident steam filled containment during this period of detector insertion would be brief. The licensee states that this equipment normally operates in the containment when elevated temperatures (not a harsh environment) are present. There are 4 SRMs and 8 IRMs in addition to the average / linear power range monitors that are said to be able to detect inadvertent reactivity insertion in the post-accident situation.

In the process of our review of cautron flux instrumentation for boiling water reactors (BWRs), we note that the mechanical drives of the detectors have not satisfied the environn: ental 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 existing instrumentation is acceptable for interim operation. The licensee has committed to follow industry development of this equipment, evaluate newly developed equipment, and install Category 1 instrumentation when it becomes available.

3.3.2 Reactor Coolant System Soluble Boron Concentration ,

Regulatory Guide 1.97 recommends in;trumentation for this variable with a range from 0 to 1000 parts per million. The licensee has supplied e

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instrumentaticn for this variable with a range of 50 to 2000 parts per . .

million. Thus, the licensee cannot resolve between 0 and 50 parts per million.

The licensee deviates from the guidance of Regulatory Guide 1.97 with resoect to the range of this post-accident sampling capability. This ceviatien goes beyond tne scope of this review and is being addressed by the NRC as part of their review of NUREG-0727, Item II.B.3.

3.3.3 Dryweil pressure Regulatory Guice 1.97 reccomenos menitering the pressure in the da drywell.

The range recommended is from 12 psia to design pressure (62 psig). Category 1 instrumentation is reccmmended. The instrumentation identified in Reference 5 for t51s variable is the Category I primary containment pressure that has a range of -5 to 250 psig. These pressure cetectors are located in the drywell. Thus the recommendation of Regulatory Gaide 1.97 for the variable drywell pressure are satisfied.

3.3.4 Drywell Sump Level Drywell Drain Sumps level The licensee has supplied Category 3 instrumentation; a single channel with a span of 600 gallons for the drywell equipment drain tank and a single channel with a span of 200 gallons for the drywell floor drain tank. The liter.see states that (a) the drywell pressure and temperature are more appropriate to detect a breach of the reactor coolant system and (b) the sump level equipment does not initiate any automatic protective action and (c) the sump level equipment would not be available if containment were isolated.

We cenclude that the instrumentation supplied by the licensee'will provide appropriate monitering for the parameters of concern. This 1s -

based on ,(a) for small leaks, the instrumentation is not expected to

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experience harsh environments during operation, (b) for larger leaks, the sumps fill promptly and the sump drain lines isolate due to the increase in dry'w ell pressure, thus negating the drywell sump level and drywell drain sumps level instrumentation, (c) the drywell pressure and temperature can

. be used to detect leakage in the drywell, and (d) this instrumentation neither automatically initiates nor alerts the operator to initiate operation of a safety-related system in post-accident situation.

Therefore, we find the Category 3 instrumentation provided acceptable.

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3.3.5 Primary Contain ent Pressure Regulato*y Guide 1.97 recommends instrumentation for this variable with a range of from 10 psia to four times the design pressure of 62 psig (248 psig). The licensee, in Reference 4, identified instrumentation located in the drywell with a range of 0 to 250 psig. Instrumentation in the torus has a range of 0 to 4 psig.

For the torus instrumentation, the licensee states that instrumentation with a higher range is not necessary, even though the torus design pressure is 35 psig. This is because of vacuum breakers between the torus and the drywell that keep the torus within 3 psi of the drywell. Thus, the drywell pressure instrumentation is applicable to the torus pressure. In Reference 5, the licensee states that the drywell pressure instrumentation has been re-ranged to -5 to 250 psig. This satisfies the recommendations of Regulatory Guioe 1.97, and is acceptable.

3.3.6 Primary Containment Isolation Valve Position Regulatory Guide 1.97 recommends Category 1 instrumentation for this variable. Thus, environmental qualification, seismic qualification and redundancy are recommended for this instrumentation. The licensee provides instrumentation for these variables, however, deviations are iden't'ified, in l Reference 4, in the above criteria.

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In Refe ence 5, the licensee states that environmental and seismic '

cualification of these valves has been addressed as part of their enVir:nmental qualification program in response to the Environmental Qualification Rule, 10 CFR 50.49. Seismic qualification for those position '

indication switches that were not upgraded as a result of this program do meet the criginal plant seismic design criteria. We find this acceptable for Regulatory Guide 1.97 instrumentation.

From the infcreation provided, we find the licensee deviates from a strict interpretation of the Categcry I redundancy recommendation. Only the active valves have position indication (i.e., check valves have no cosition indication). Since redundant isolaticn valves are provided, we find that recundant indication per valve is not intended by the regulatory guide. Pcsition indication cf check valves is specifically excluded by Tacle 1 of Regulato y Guide 1.97. Therefore, we find that the reduncarcy for this variable is acceptable.

There are three vacuum relief valves from the reactor builcing to the teres. These valves have position switches that are not stated to be seismically qualified. They function as check valves. As such, we find that position indicatien for these valves is net required by Regulatory Gaide 1.97.

3.3.7 Radiatien Level in Circulatino Primary Coolant The licensee indicates that raciation level measurements to indicate fuel cladding failure are provided by monitoring containment radiation and

by utilizing the post-accident sampling system, which is being reviewed by the NRC as part of their review of NUREG-0737, Item II.B.3.

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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3.3.8 Suporession Pool Water i.evel

' Regulatory Guide 1.97 recommends instrumentation for this variable with a range from the bottom of the ECCS suction line to five feet above the

. normal water level. The licensee's instrumentation has a range from

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3 ft. 3 in, below the ECCS suction to 3 ft 3.5 in above the normal water level. The licensee states that even in the most extreme degraded conditions, the level of the suppression pool is not expected to increase by more than 1-1/2 feet. All remedial operator actions tof ontrol a rise in torus level will have been taken before this level is reached.

Therefore, we find the range to 3 ft. 8.5 in, above the normal water level acceptable.

3.3.9 Containment and Drywell Hydrocen Concentration Regulatory Guide 1.97 recommends instrumentation for this variable with j a range of 0 to 30 percent. The licensee's instrumentation for this' variable has a range of 0 to 20 percent.

The licensee states that for an inerted containment, the primary concern is oxygen concentration. This is because combustion could not occur if sufficient oxygen is not present. Therefore, the licensee states that a maximum range of 20 percent is acceptable.

The licensee deviates from Regulatory Guide 1.97 with respect to l

hydrogen concentration instrumentation. This deviation goes beyond the l scope of this review and has been addressed by the NRC as part of the

! review of NUREG-0737, Item II.F.1.6, and found acceptable.

t 3.3.10 Radiation Exposure Rate Regulatory Guide 1.97, Revision 2, specifies Category 2 instrumentation i

for this variable with a range of 10 ~1 to 104 R/hr. The licensee has

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provided instrumentation for this variable with ranges that vary, dependent i '

on location, from the recommended range. The ranges are 0.01 to 100 mR/hr, 0.1 to 1000 mR/hr and a dual range of 0.1 to 1000 mR/hr and 0.01 to 9

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The licensee has stated that containment breach is detected oy the noble gas effluent monitors, and that release assessment is better per' formed with portable radiation instruments and secondary containment sample analysis. The licensee concludes that Category 3 instrumentation is ,

adequate for the radiation exposure rate instrumentation. .

Reguistcry Guide 1.97, Revision 3 (Reference 7), changes this variable i to Category 3. Therefore the only deviation of the Nine Mile Point statien for this variable is the range supplied for a given locatipn. The licensee states that twenty-one of the thirty-three instruments have ranges that encompass the expected radiation levels in their lccations. The remaining instrumentation woulc be used after an accident after they were shcwn to be cperable.

Should the instrumentation be effscale, entry ir.to the area is prohibited.

Portable survey instruments, atmosphere sampling and radiation monitors in tne olant stack will be usec by the licensee fer release detection and assessment and for long term release surveillance. Based on this, we find the licensee's instrumentation for this variable acceptable.

3.3.11 Sueoression Chamber Spray Flow Drywell Soray 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 ficw measurement channels. Instead, a ficw element common to these two sprays is used in their' common header.

This measures the total system flew that is going to both of these sprays.

The licensee indicates that a predetermined portion of the total flow is delivered to each spray via a throttling valve. The licensee concludes that indication of total flow and valve position is sufficient to monitor the operatien of these sprays. We find this deviation acceptable.

3.3.12 Drywell Atmosphere Temperature t

Regulatory Guide 1.97 recommends instrumentation for this variable with a range of 40 to 440*F. .

The licensee has instrumentation for this variable 10

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with a range of 50 to 300*F, and states that the range is sufficient t3

. prov.ide the operator with information relative to the potential for flashing in the level sensing instrument lines.

. We. agree that the given range is sufficient to monitor the potential for flashing in the instrumentation lines for the reactor vessel level instrumentation.

Our examination of the Final Safety Analysis Report (F55R, Reference 8) shows tnat the maximum internal drywell design temperature is 310*F. The actual peak temperature would be less than this and of short duration.

Based on this, the licensee's upper limit of 300*F for the post-accident perico is sufficient.

The licensee states that in the 16 years since reactor startup, the drywell atmosphere temperature has remained above 50*F. During post-accident conditions, the licensee states that the heat sources are sufficient to insure that this instrumentation will remain onscale.

Therefore, we find the range of 50 to 300'F acceptable for this variable.

3.3.13 Standby Licuid Control System Flow Exception has been taken by the licensee to Regulatory Guide 1.97 for the variable standby liquid control system flow. The licensee states that proper functioning of the system can be verified by monitoring the pump discharge header pressure, tank level, neutron flux change and system valve position.

We find the above instrumentation valid .ns an alternative indication of standby licuid centrol system flow.

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3.3.14 _S_tandby Lieuid Cor. trol System Sterage Tank Level * ~

-Regulatory Guide 1.97 recommends instrumentation for this variable with a range frem the bottom to the tcp. The instrumentation supplied by the

• licensee for this variabic has a range frcm 350 to 4150 galions.

Inc licenset states that.the range of the liquid poison level indicater covers the minimum (2000 gallens) to maximum (4050 gallor.s) volume of licuid poisch maintained in the tank, as requirec by Techni, cal Specifications. The licensee concludes that the range is' sufficient for the operator to cetermine that the liquid poison system is cperating or:perly.

Ine range surplied correspends to the height of the pump suction inlet anc the tank overflow. Based en the licensee's justificatice, the deviation from the recomnended range is acceptable.

3.3.15 Residual Heat Removal System Flow Residual Heat Removal Heat Exchanger Outlet Tem:erature Regulatory Guide 1.97 recommends monitoring the residual beat removai system for flow (0 to 110 percent of cesign flow) and heat exchanger outlet temperature (32 to 350"F) with environcentally qualified instrumer.tation.

Unit No. I at Nine Mile Point has no direct indication of ficw rate for tnis variable. The licensee states that the shutdown cooling system flow is manually adjusted to maintain the cooldown rate below 300'F/nr. Thus.

flow is controlled by the shutdown cooling system temperature, j The licensee monitors the operation of the shutdown cooling system (SCS) with the following instrumentation Reactor vessel water level -

i SCS pump running indication -

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SCS heat exchanger tube side outlet temperature "SCS heat exchanger shell side inlet and outlet temperature.

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.' The licensee states that there is no flow disturbance that cannot be observed with the existing instrumentation. Additionally, the SCS is not expected to be operated during accident or immediate post-accident conditions. It would only be operated in the long term after the unit is in t normal stable shutdown cooling condition. The licens,ee states, in Reference 5, that the instrumentation for the shutdown cooling system has been addressed in accordance with 10 CFR 50.49, and environmental qualification found not applicable.

Based en the alternate instrumentation and the design function of the shutdown cooling system, we find the instrumentation supplied for these variables acceptable.

3.3.16 Coolina Water Temperature to ESF System Components Coolina Water Flow to ESF System Cemconents Unit No. I at Nine Mile Point does not use a separate cooling water system to cool these components, which according to the licensee consist of the core spray and the containment spray pumps. These pumps are cooled by recirculation of the discharge flow. Pump suctiun is from the torus. ,

Thus, the cooling water temperature is essentially the torus water temperature, and flow is coincident with pump operation. We find that these deviations are acceptable.

3.3.17 Hich Radioactivity Liquid Tank Level Regulatory Guide 1.97 recommends instrumentation for this tank with a range from top to bottom. The indicated range for this variable is 0 to 100 percent (corresponding to O to 166 in. height, whereas the tank is 180 in. high). The existing range is adequate to indicate storage volume.

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Therefore, this is an acceptabla deviation from Regulatory Guide 1.97.

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3.3.18 Enercency Ventilation Damoer Position "

Regulatory Guide 1.97 recommends Category 2 instrumentation to monitor the operation of the emergency ventilation dampers. The licensee has

• provided two types of instrumentation for this variable. First, there are damper. position indicator lights which are not Category 2. Second, system flow is monitorec by Categcry 2 instrumentation.

. Based on this diversity, we find that the deviation froy. Category 2 to Category 3 instrumentation for damper position indication 'is acceptable.

System operation can be observed by the Category 2 flow instrumentation.

3.3.19 Reactor Building or Seconcary Certainment Radiatior.

Reculatcry Guide 1.97 recommends Category 2 instrumentation for this variable with a range cf 10"I to 104 R/hr for the Nine Mile Point Mark I containment. The licensee has 34 instruments for this variable, some with a range of 10 -5 to 10 ~I R/hr, some with a range of 10~# to 1 R/hr and one with a range of 10~2 to 103R/hr. These instruments are Catescry 3 rather than Category 2. The ranges were chosen on a plant analysis of expected radiation levels. The licensee's position is that secondary containment area radiation is not an apprcpriate parameter to use for assessing primary containment leakage or detecting significant releases. The licensee also states that the reactor building ventilation l system is automatically isolated and the emergency ventilation system

{ initiated at an exposure rate of 20 mR/hr (2 x 10~2 R/hr). Tin licensee concludes that the existing Category 3 instrumentation f'or this variable is adequate.

Based on the above, we find that the existing Category 3 instrumentation and ranges are acceptable.

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3.3.20 Noble Gas and Vent Flow Rate--Common Plant Vent

, . Regulatory Guide 1.97 recommends instrumentation for this variable with a range from 10 -6 to 10'4 uti/cc and 0 to 110 percent design flow.

The licensee's instrumentation for this variable was not ccmpletely identified by Reference 4. This was due to the installation of new equipment for this variable. Reference 5 identifies the range as 2 x 10 -0 tc 10 5uti/cc wnich meets the recommended range and 15 to 110 percent for the vent flow. Below the 15% limit of this ins'trumentation the flow is no longer isokinetic, but samples can still be analyzed.

Additionally, flow would be above 15 percent with any one fan operating.

We find this instrumentation acceptable for this variable.

3.3.21 Particulates and Halocens--All Identified Plant Release Points Regulatory Guide 1.97 recommends instrumentation for this variable with a range frcm 10 -3 to 10+2 uCi/cc.

The licensee's instrumentation for this variable was not completely identified by Reference 4. This was due to the installation of new equipment for this variable. Reference 5 identifies the range as

-13 10 to 10 uti/cc for undiluted samples. With dilution, the recommended

! range is satisfied. Additionally, lab samples can be analyzed in this range. We find this instrumentation acceptable for this variable.

3.3.22 Plant and Environs Radioactivity .

Revisic, 2 of Regulatory Guide 1.97 recommends a multichannel gamma-ray spectrometer for this variable for isotopic analysis in release assessment and analysis.

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The licensee, in Reference 5, identifies the instrumentation for this

• variable and shows that it satisfies the recommendations of Regulatory

. Guide 1.97.

3.3.23 Estimatic, of Atmoseheric Stability j Regulttcry Guide 1.97 re.cox.iends instrumentation for this variable with a range of -9 to +1ScF or an analogcus range for alternative stability analysis. The licensee has supplied instrumentation with a range of -3 to

+20*F. Tne licensee has not previded justification fer the deviatien from

-9 to -8'F.

Table 1 of Raguiatcry Guide 1.23 (Reference 9) provides seven atmoscheric stability classificatiens based on the difference in temperstare per 100 meters evaluation cnange. These classifications cover from extremely utstrble to extremely stable. Ar.y temperature difference greater than +4*C cr less than -2'C does nothing to the stability classification. The licensee's instrumentation includes this range.

Therefore, we find that this instrumentation is acceptable to determine atmospheric stability.

3.3.24 Accicer.t Samoling (Primary Coolant, Containment Air and Sumol 1

The licensee's post-accident sampling system provides sampling anc analysis as recommended by the regulatory guide except for. the followir,9 deviations.

1. Boren content--the minimum observable concentration is 50 ppm.

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2. Chloride content--the minimum observable concentration is 0.1 ppm.

l 3. Dissolved hydrogen--the minimum observable concentration 4s 25 cc/kg.

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Dissolved oxygen--the minimum observable concentration is 0.1 ppm.

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5". Air hydrogen content--the minimum observable concentration is 0.1 percent.

6. ~ Air oxygen content--the minimum observable concentration is -

0.5 percent.

7. Coolant pH--the range is 2 to 12 rather than the recommended 1 to 13.

S. A grab sample frem the primary containment sump cannot be obtained.

The licensee deviates from Regulatory Guide 1.97 with respect to pcst-accident sampling capability. This deviation goes beyond the scope of this review and is being addressed by the NRC as part of their review cf NUREG-0737, Item II.B.3. -

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4. CONCL*JSIONS Based on cur review, we find that the licensee either conforms to or is '

justified in deviating from Regulatory Guide 1.97, with the following excepticr.: -

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heutren flux--tne licensee's present instru.entation is accep ao;e cr ar. interim basis until Category 1 instrumentati,on is developed and irstallec (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 Construction Permits,

" Supplement No. I 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 During and Following an Accident, Regulatory Guide 1.97, Revision 2, NRC, Office of Stancards Development, December 1980.

3. Clarification of TMI Action Plan Requirements. Requirements for Emergency Response Capability, NUREG-0737, Supplement No. 1, NRC, Of fice of Nuclear Reactor Regulation, January 1983.

4. Niagara Mohawk Power Corporation letter, C. V. Mangan to Director of Nuclear Reactor Regulation, NRC, "Nine Mile Point Unit 1, Docket No. 50-220, OPR-63," April 2, 1924.

5. Niagara Monawk Power Corporation letter, C. V. Mangan to Director of Nuclear Reactor Regulation, NRC, " Request for Additional Information Concerning Niagara Mohawk Power Corporation's Submittal on Section 6 of Supplement I to NUREG-0737, Regulatory Guide 1.97-Application to Emergency Response Facilities," October 18, 1985.

6. Niagara Mohawk Power Corporation letter, C. V. Mangan to Director of Nuclear Reactor Regulation, NRC, "Nine Mile Point Unit 1, Docket No. 50-220, DPR-63," December 6, 1985.

7. Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant Environs Conoitions During and Following an Accident, Regulatory Guide 1.97, Revision 3, NRC, Office of Nuclear Research, May 1983.

8. Final Safety Analysis Report. Nine Mile Point Nuclear Station, Niagara MonawK Fower Corporation, Syracuse, NY 13202, June 1967.

9. Onsite Meteorological Programs, Regulatory Guide 1.23 (Safety Guide 23), NRC, February 17, 1972 or Meteorological Programs in Support of Nuclear Power Plants, Proposed Revision 1 to Regulatory Guide 1.23, NRC, Office of Standards Development, September 1980.

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