ML20085L541

Conformance to Reg Guide 1.97:Nine Mile Point-1

ML20085L541
Person / Time
Site:	Nine Mile Point
Issue date:	09/30/1991
From:	Udy A EG&G IDAHO, INC.
To:	NRC
Shared Package
ML18038A497	List: ML20085L541
References
CON-FIN-A-6483, RTR-REGGD-01.097, RTR-REGGD-1.097 EGG-NTA-9161, TAC-M69209, NUDOCS 9111040220
Download: ML20085L541 (37)

Text

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.N EGG NTA-9161 TECHhlCAL EVALVATION REPCRT CONFORMANCE TO REGULATORY GUIDE 1.97: NINE MILE POINT-1 .-

Docket No. 50-220 Alan C. Udy Published September 1991 EG&G Idaho, Inc.

Idaho National Engineering Laboratory Idaho Falls, Idaho 83415 Prepared for the U.S. Nuclear Regulatory Comission Washington, D.C. 20555 Under DOE Contract No. DE-AC07-761001570 FIN No. A6483 TAC No. 69209 A

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SUMMARY

This EG&G Idaho, Inc., report documents the review of the Regulatory Guide 1.97, Revision 2, submittals for Unit Number 1 of the Nine Mile Point ,

Nuclear Station and identifies areas of nonconformance to the regulatory 4'-

guide. Exceptions to Regulatory Guide 1.97 are evaluated and those areas

- where sufficient basis for acceptability is not provided are identified.

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FIN No. A6483 B&R 120 19-15-02-0 Docket No. 50-220 TAC No. 69209 11 i siiim ri e i . .

4 PREFACE 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 .

4 of Systems Technology, by EG&G Idaho, Inc., Regulatory and Technical Assistance Unit.

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SUMMARY

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- ; ACE ............................................................... iii

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, s4gt. . INTRO  ;-: .................................................... i -

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2. WEVidh Re. 'i.EMENTS ............................................. 2

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EVAltlATION ...................................................... 4 h 3.1 Adheror ce La Regul atory Guide 1. 97 . . . . . . . . . . . . . . . . . . . . . . . . 4 s

3.2 Type A Variables .......................................... 4 ,

3.3 Exceptions to Regul atory Guide 1.97 . . . . . . . . . . . . . . . . . . . . . . 6

4. CONCLUSIONS ..................................................... 30

5. REFERENCES ..... ......................................... ...... 31 T

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CONFORMANCE TO REGULATORY GUIDE 1.97: NINE MILE POINT-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. lhis letter included additional clarification regarding Regulatory Guide 1.97, Revision 2 (Reference 2), relating to the requirements for emergency response capability.

These requirements sve been published as Supplement No. I to NUREG-0737, "TMI -

Action Plan Requirements" (Reference 3).

• Niagara Mohawk Power Corporation, the licensee for the Nine Mile Point Nuclear Station, provided a Unit I specific response to Item 6.2 of the generic letter on April 2, 1984 (Reference 4). The licensee provided additional information in submittals dated October 18, 1985 (keference 5), and December 6, 1985 (Reference 6). These submittals were the basis for a previous Technical Evaluation Report, EGG-NTA-6880 (Reference 7).

The licensee provided schedules on May 19, 1989 (Reference 8). The licensee supplied updated information for their Unit 1 instruinentation on July 31, 1989 (Reference 9). Reference 9 superseded the earlier information.

The licensee provided additional information on May 25, 1990 (Reference 'i) ,

October 29, 1990 (Reference 11), and August 26, 1991 (Reference 12).

This report, ca.0d on the recommendations of Regulatory Guide 1.97, Revision 2, compares the instrumtntation described in the licensee's Unit I submi+tals with these recommendations.

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l satisfied the plant's original seismic design basis criteria, it was acceptable for meeting the seismic criteria of Regulatory Guide 1.97. )

Therefore, this report addresses only those exceptions to Regulato'y --

Guide 1.97 identified by the licensee. Ths following evaluation is an audii, l of the licensee's submittals based on the review policy described in the NRC regional meetings.

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1. are monitored and controlled post-accident to mitigate the consequences of the event and to assure the accomplishment of plant safety functions,

2. are used to assess a 'need for manual operator action, or

3. are maintained, by operator action, either above or below an E0P specified value or limit.

The licensee states that E0P key parameters are Category 1 variables. Thus, the definition of E0P key parameters is inclusive of the definition for Type A variables. The licensee defines the following variables as E0P key .

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parameters.

1. neutron flux -- average power range monitors

2. coolant level in reactor

3. reactor coolant system (RCS) pressure

4. suppression pool (torus) water temperature

5. suppressian pool (torus) water level

6. drywell temperature '

7. drywell pressure

8. containment hydrogen concentration

9. containment oxygen concentration

10. drywell water level We note that drywell water level is not a Regulatory Guide 1.97 variable.

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9 The licensee is revising the power sources for this instrumentation to achieve divisional independence.

The ACUREX fuel zone transmitters (36-24A and 36-248) share a common reactor vessel low-end tap and sensing line. The licensee committed to train control room operators for a postulated break in this sensing line. The licensee lists indications useful in diagnosing this postulated event. The licensee lists alternate means for determining the reactor vessel water level if this occurs. The licensee states these instruments do not start any automatic actions or confirm any automatic actions. The instrument line is approximately one inch in diameter. This break size is an analyzed postulated ,

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event in-the Final Safety Analysis Report. Emergency Operating Procedure

• E0P-2, "RPV Control," includes actions needed to restore and maintain the RPV water level. Based on the licensee analysis and available contingencies, we find the design with a single vessel tap and sensing line acceptable.

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3.3.5 BWR Core Thermocouples 1

i Regulatory Guide 1.97 recommends Category 1 instrumentation for this l variable. However, Section 6.1.b of Supplement No. I to NUREG-0737 (Reference 3) excludes this instrumentation. Therefore, this variable does not require instrumentation.

l l 3.3.6 Reactor Coolant System (RCS) Pressure Regulatory Guide 1.97 recommends Category 1 instrumentation for this variable. Category 1 criteria include qualified isolation devices for

! transmission of signals to other equipment. The licensee's instrumentation had two problems in this area. First, the signal feeds the feedwater control syster, without using a qualified isolation device. Second, a switch has inputs from both channels. The switch feeds the selected signal to a common

! reco-der display with a span of 950 psig to 1050 psig. The licensee has rewired this switching network to maintain separation (Reference 13).

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4 The licensee lists nonclass lE power for the following components.

Closed contact Open contact Yalve 44.2-15-lLSC 44.2-15-lLSO 44.2-15A .

44.2-16-lLSC 44.2-16-lLSO 44.2 16A .

44.2-17-lLSC 44.2-17-lLSO 44.2-17A 44.2-18-lLSC 44.2-18-lLSO 44.2-18A 40-02-lLSC 40-02-lLSO 40-028 40-12-ILSC 40 12-ILSO 40-12B Reference 12 clarifies this situation. Nonclass IE power powers the ,

indication lamps that are part of the control switch. The primary containment '

isolation valve mimic display uses Class lE power for the position indication for these valves. We find this acceptable.

The mimic display needs no individual annotation of the Regulatory Guide 1.97 function, as the mimic's function is to display the status of containment isolation.

3.3.10 Radiation Level in Circulatinc Primary Coolant l

l lhe licensee states that the following radiation level measurements j indicate fuel cladding failure:

I o containment radiation level i

! o main steamline radiation level o off-yas radiation level o post-accident sampling system The NRC reviewed and approved the post-accident sampling system as part of their review of NUREG-0737, Item II.B.3.

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1. are monitored and controlled post-accident to mitigate the consequence 3 of the event and to assure the accomplishment of plant safety functions,

2. are used to assess a need for manual operator action, or

3. are maintained, by operator action, either above or below an E0P specifiei value or limit.

The licensee states that E0P key parameters are Category 1 variables. Thus, the definition of E0P key parameters is inclusive of the definition for Type A variables. The licensee defines the following variables as E0P key ,

parameters. ' '

l. neutron flux -- average power range monitor s

2. coolant lesel in reactor

3. reactor coolant system (RCS) pressure

4. suppression pool (torus) water temperature

5. suppression pool (torus) water level

6. drywell temperature

7. drywell pressure

8. containment hydrogen concentration

9. containment oxygen concentration

10. drywell water level-We note that drywell water levcl is not a Regulatory Guide 1.97 variable.

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8 i to 12.5 percent power, with manual switching between linear ranges. The IRMs  !

share, via switching, recorders with the APRMs. l The licensee's SRMs consist of'four channels. Two channels receive .

1 power from RPS bus 11; the remaining two channels receive power from RPS '

bus 12. The detectors are seismically qualified. The licensee states that environmental qualification is not required. All four SRM channels have indicators. The licensee also records two of the four channels. The range is I count per second to 10' counts per second. The licensee states that this range covers up to 10 percent of full reactor power, The licensee is a sponsoring utility of the Boiling Wat!r Reactor Owners Group (BWROG) appeal of the NRC staff position that directed the installation of upgraded, qualified neutron monitoring instrumentation. The licensee deferred plant specific implementation until the BWROG appeal is resolved.

The NRC is currently reviewing the BWROG appeal. Upon resolution of the appeal, the licensee shovid install instrumentation that complies with the resolution of the BWR0G appeal. We conclude that the existing instrumentation is acceptable for interim operation.

3.3.2 Control Rod Position Regulatory Guide 1.97 recommends Category 3 instrumentation for this variable to indicate full in or not full in. The licensee's instrumentation indicates steps 00 through 40. We find this alternate range acceptable.

3.3.3 Reactor Control System Soluble Boron Concentration Regulatory Guide 1.97 recommends sampling and analysis for this variable. It recommends resolution between zero and 1000 parts per million.

The licensee's post-accident sampling system can resolve between T0 parts per 7

The instrumentation is usable after showing instrument operability following an offscale excursion. The licensee uses portable survey instruments, containment atmosphere sampling, and radiaticn monitors in the plant stack for release detection and assessment and 'for long term surveillance. Based on the alternative indications, instrument capabilities, and Revision 3 of the regulatory guide, we find this instrumentation acceptable.

3.3.18 Effluent Radioactivity - Noble Galgi Regulatory Guide 1.97 recommends Category 2 instrumentation for this ,

variable with a range of 10 4 Ci/cc to 10' Ci/cc. The licensee indicates that the three RAGEMS channels provide this data. The licensee states that the instrumentation is in a mild environment. The instrumentation display is in the chem lab. The licensee lists two channels with a range of 10* Ci/cc to 10' Ci/cc. The remaining channel has a range of 10 4 Ci/cc to 10 Ci/cc. Section 6.2 of Supplement No. I to NUREG-0737 (Reference 3) makes allowance for displays in places other than the control room. The instrumentation is in a mild environment. A channel of this instrumentation exceeds the recommended range. Therefore, we find the provided instrumentation acceptable.

3.3.19 Suporession Chamber Soray Flow Drvwell Soray Flow Regulatory Guide 1.97 recommends Category 2 instrumentation for these two variables. The licensee's containment spray system consists of two redundant pumping and distribution trains. The licensee's instrumentation measures the flow at the output of each pump. Thus, the instrumentation measures the total system flow to both of the sprays included in the regulatory guide criteria. The licensee indicates that each spray header receives a fixed portion of the total flow. We find the total flow indication 17

3.3.21 d31n Steamline isolation Valves' leakaae Control System Pres 1urg Regulatory Guide 1.97 recommends Category 2 instrumentation for this variable. ThelicenseestitesthatNineMilePoint-1hasnoleakagecontrol system on the main steamline isolation valves. Therefore, this variable does not require instrumentation.

3.3.22 Primary System Safety Relief Valve Positions Regulatory Guide 1.97 recommends Category 2 instrumentation for this ,

variable. The licensee's valve tailpipe thermocouple system is Category 3. '

The'licerisee's acoustic monitoring system is Category 2. Thus, the licensee's acoustic monitoring instrumentation satisfies the regulatory guide criteria for this variable.

3.3.23 Reactor Core Isolation Coolina System Flow Regulatory Guide 1.97 recommends Category 2 instrumentation for this variable. The licensee states that Nine Mile Point-1 has no reactor core isolation cooling system. Therefore, this instrumentation is not required.

3.3.24 Hiah Pressure Coolant In.iection (HPCI) System Flow Regulatory Guide 1.97 recommends Category 2 instrumentation for this variable. The licensee states that the main feedwater pumps perform the HPCI function. The main feedwater flow has instrumentation that does not satisfy the envic car antal qualification criteria for Categor; 2 instrumentation.

The licensee clarified the environmental qualification requirements for the main feedwater flow instrumentation. Because the transmitters are in a mild post-accident environment, they do not require environmental 19

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The licensee is revising the power sources for this instrumentation to achieve divisional independence.

The ACUREX fuel zone transmitters (36 24A and 36-248) share a common reactor vessel low-end tap and sensing line. The licensee committed to train '

control room operators for a postulated break in this sensing line. The licensee lists indications useful in diagnosing this postulated event. The licensee lists alternate means for determining the reactor vessel water level if this occurs. The licensee states these instruments do not start any automatic actions or confirm any automatic actions. The instrument line is approximately one inch in diameter. Inis break size is an analyzed postulated ,

event in the Final Safety Analysis Report. Emergency Operating Procedure '

E0P-2, "RPV Control," includes actions needed to restore and maintain the RPV water level. Based on the licensee analysis and available contingencies, we find the design with a single vessel tap and sensing line acceptable.

3.3.5 HWB Core Thermocouples Regulatory Guide 1.97 recommends Category 1 instrumentation for this variable. However, Section 6.1.b of Supplement No. I to NUREG-0737 (Reference 3) excludes this instrumentation. Therefore, this variable does not require instrumentation.

3.3.6 Reactor Coolant System (RCS) Pressure Regulatory Guide 1.97 recommends Category 1 instrumentation for this variable. Category 1 criteria include qualified isolation devices for transmission of signals to other equipment. The licensee's instrumentation had two problems in this area. First, the sianhl feeds the feedwater control system without using a qualified isalation device. Second, a switch has inputs from both channels. The switch feeds the selected signal to a common recorder display with a span of 950 psig to 1050 psig. The licensee has rewired this switching network to maintain separation (Reference 13).

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automatically isolate ,at the primary containment nenetration should an accident occur. Drywell temperature, drywell prt uure, and reactor pressure vessel water level-can also show leakage from the reactor coolant system.

We conclude that the alternate instrumentation provided by the licensee will. provide the appropriate monitoring of the sumps for the parameters of concern. We base.this conclusion on the following.

1. For sinall leaks, the instrumentation _will not experience a harsh environment during operation and14111 show response to the leak.

2. For larger leaks, the sumps fill prompt 1; and the sump drain lines '

isolate due to the increase in drywell pressure, thus negating the drywell sump level and drywell drains sump level instrumentation.

3. The drywell pressure and temperature (both Category 1), as well as the reactor pressure vessel. water level (Category 1) are alternative indications of leakage in the drywell.

4. 'This instrumentation neith~er automat'i: ally starts nor alerts the operator to' start _ operation of-a safety-related system in a

post-accident situation.

< JTherefore, we-find the provided-alternate instrumentation-acceptable.  ?

3.3.9 Primary Containment Isolation Valve Position Regulatory Guide 1.97 recommends Category 1 indication of-the open-closed-position of:the primary coatainment isolation valves. Category 1 criteria include redundancy,? environmental and seismic: qualification, Class lE.

4 power, and labeling in the control room.- The licensee provides a

- comprehensive-listing of their containment isolation valves with a oc scription of the position monitoring components.

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The licensee lists nonclass IE power for the following components.

Closed contact Open contact Valve 44.2-15-ILSC 44.2-15-lLSO 44.2-15A 44.2-16-lLSC 44.2 16-lLSO 44.2-16A ,

44.2-17 lLSC 44.2-17-lLSO 44.2-17A 44.2-18-lLSC 44.2-18-llSO 44.2 18A 40-02-ILSC 40-02-ILSO 40-028 40-12-1LSC 40-12-lLSO 40-12B Reference 12 clarifies this situation. Nonclass lE power powers the -

.ndication lamps that are part of the control switch.

The primary containment p isolation valve c.imic display uses Class lE power for the position indication P for these vaives. We find this acceptable.

The mimic display needs no individual annotation of the Regulatcry 1 Guide 1.97 function, as the mimic's function is to display the status of containment isolation.

3.3.10 Radiation level in Circulatina Primary Coolant The licensee states that the following radiation level measurements -

indicate fuel cladoing failure:

o containment radiation level o main steamline radiation level o off-gas radiation lerel o post-accident sampling system The NRC reviewed and approved the post-accident sampling system as part of their review of NUREG-0737, Item II.B.3.

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G feet above the normal operating level. Therefore, we find the provided 1.25 foot to 14.75 feet suppression pool watei level instrumentation acceptable.

3.3.14 Containment and Drywell Hydrocen Concentration Regulatory Guide 1.97 recommends instrumentation for this variable with a range of zero to 30 percent. The licensee's instrumentation has a range of zero to 100 perce t. This represents either zero to 5 percent or zero to 20 percent, depending on the position of a selector switch. Each channel of ,

instrumc:itation has its own range selector switch. -

The licensee states that the primary concern for an inerted containment is the concentration of oxygen. Combustion could not occur if sufficient oxygen is not present.

The NRC reviewed and approved this instrumentation as part of their review of NUREG-0737, item II.F.1.6. We find this a good faith attempt [as defined in huREG-0737, Supplement No. 1, Section 3.7 (Reference 3)] to meet NRC requirements. Therefore, this instrumentation is acceptable.

3.3.15 Containment and Drywell Oxvaen Concentration Regulatory Guide 1.97 recommends instrumentation for this variable with a range of zero to 10 percent. The licensee's instrumentation has a range of zero to 100 percent. This represents either zero to 5 percent or zero to 25 percent, depending on the same selector switch used for hydrogen concentration. Each channel of instrumentation has its own range' selector switch.

While the zero to 5 percent span dcas not comply with the regulatory guide, the zero to 25 percent span does. Therefore, this instrumentation is acceptable.

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e We find the use of the above instrumentation valid as an alternate indication of SLCS flow.

3.3.28 Elandby Liauid Control System (SLCS) Storace Tank Level ,

Regulatory Guide 1.97 recommends instrumentation for this variable with a range from the bottom to the top of the tank. The licensee's

.. strumentation has a span of 350 gallons to 4150 gallons. The span

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corresponds to the pump suction inlet and the tank overflow vent. Plant Technical Specifications require a maximum volume of 4080 gallons in this tank. At 350 gallons, the tank is essentially empty. Additional pumping will not occur from below the pump suction line. Therefore, t.he 350 gallon to 4150 gallon range is acceptable.

The licensee notes that they are processing a modification to change this range. The licensee identified the new range in Reference 12 as zero to 2000 gallons. The licensee statas that they will use enriched baron. With enriched boron, the technical specifications require a minimum of 1185 gallons. The maximum expected level is 1500 callons. Based on these limits, we find the zero to 2000 gallon range acceptable, l-3.3.29 Rgt qual Heat Removal (RHR) System Flow i

Regulatory Guide 1.97 recommends Category 2 instruuentation for this j variable. The regulatory guide recommends a range of zero to 110 percent of design flow. Nine Mile Point-1 has no direct indication of RHR system flow.

The licensee states that the RHR function is part of the shutdown cooling L system. The shutdown cooling system operates after establishing a normal,

! stable shutdown cooling condition in the long term recovery. The immediate post-accident recovery does not use the RHR function.

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3.3.31 Coolina Water Temperature to Enaineered Safety Features System Comoonents l

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Regulatory Guide 1.97 recommends Category 2 instrumentatien for thit variable. The licensee indicates that the subgroup of engineered safety .

features components fo. this variable are the core spray pumps and the centainment spray pumps. These pumps receive cooling water from the recirculation of a portion of the pump dischargo flow. Pump scetion is from the suppression pool. Category 1 instrumentation monitors the suppression pool temperature.

As the cooling water temperature is essentially the same as the

! suppression pool water temperature, we find the licensee's instrumentation and l design for this variable acceptable.

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3.3.32 Coolina Water Flow to Enaineered Safety Features System Components Regulatory Guide 1.97 recomt.. ends Category 2 instrumentation for this

! variable. Nine Mile Point-1 does not use a separate cooling system to cool these components. The core spray pumps and the containment spray pumps receive cooling water from the recirculation of a portion of the pump discharge flow. Pump suction is from the suppression pool. Thus, cooling water flow to these components is coincident witn pump operation. No other components are the subject of this variable.

! Based on the described features of these components, we find this j exception from the Regulatory Guide 1.97 recommendations acceptable.

3.3.33 Blah Radioactivity liauid Tank level Regulatory Guide 1.97 recommends instrumentation for this variable with a rance from the top to the bottom of the tank. The instrument span is zero to 166 inches, with an indicator marked zero to 100 percent. The tank height 23 w- - - - . - - - . - . ..

alternate instrumentation used by the licensee with this back-up instrumentation, we find the compliment of instruments proposed to monitor the secnndary containment area radiation acceptable.

3.3.35 Noble Gases and Vent Flow Rate - Common Plant Vent Regulatory Guide 1.97 recommends Category 2 instrumentation for this variable. Category 2 criteria include environmental qualification. The regulatory guide recommends ranges of 10 4 Ci/cc to 10' Ci/cc (10' pCi/cc if including certain purge flows) and zero to 110 percent of vent design flow. .

The licensee's radiological assessment and gaseous effluent monitoring system (RAGEMS) measures from 10* Ci/cc to 10' Ci/cc. The display is in the chem lab. The licensee infers that coverage beyond this is not necessary; that the provided range meets the recommendations of the regulatory guide. Reference 5 describes acceptable flow instrumentation. Reference 12 indicates that the range of zero to 250,000 cubic feet per minute satisfies the recommendations of the regulatory guide. Additionally, Section 6.2 of Supplement No. I to NUREG-0737 (Reference 3) makes allowance for displays in places other than the control room. Therefore, we find the instrumentation supplied for this variable acceptable.

3.3.36 Particulates and Halooens-- All Identified Plant Release Perm 111 Regulatory Guide 1.97 recommends sampling with onsite analysis capability for this variable and measurement of flow, it recommends an analysis capability of 10' Ci/cc to 10 Ci/cc. The licensee indicates an 2 analysis capability of 10 4 Ci/cc to 10 Ci/cc. The RAGEMS flow instrument monitors the flow. In Reference 5, the licensee discusses the analysis of undiluted and diluted samples. With a six decade span for undiluted samples, and the capability to dilute the samples, we find the provided analysis capability acceptable.

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12. The minimum range recommended for hydrogen content is zero. The provided minimum limit is 0.1 percent. The minimum range recommended fcr oxygen content is zero. The provided minimum limit is 0.5 percent.

The licensee deviates from the Regulatory Guide 1.97 post-accident sampling capability recommendations. The NRC reviewed and approved the licensee's post-accident sampling facility as part of their review of NUREG-0737, item II.B.?.

3.3.40 Redundancy and Separation -

s Regulatory Guide 1.97 recommends protecting Category 1 instrument channels against potential single failures by applying the redundancy and separation criteria of Regulatory Guide 1.75 up to and including any isolation .

devices. Nine Mile Doint-1 was designed and constructed before the guidance of Regulatory Guide 1.75 was available.

The licensee acknowledges that their separation of divisional cables is not consistent. - The licensee determined that no single hazard source would render both redundant instrument loops inoperable for any variable. The licensee's cable routing design guideline, EDG-1300, provides design guidance for redundancy and separation for system modifications. Reference 11 gives details on the licensee's analysis of cable routing. Verification and validation activities document, evaluate, report, and resolve any separation anomalies identified. The licensee has determined that they will not lose both redundant astrument channels for any v:riable simultaneously due to a single event. The licensee has committed to provide redundancy and separation for modifications. The licensee has committed to maintain the existing redundancy and separation. Therefore, we find the licensee's redundancy and separation Design Criteria Document acceptable for Regulatory Guide 1.97.

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3.3.42 Instrument Voorade.1 The licensee is developing Design Criteria Documents, and associated personnel training, to assure meeting the design basis requirements of Regulatory Guide 1.97 separation, environmental qualification, seismic qualification, quality assuranco, and power sources (including fuse sizing and coordination and wiring sizing) for future modifications and designs. Thus, the licensee has a design modification procedure for instrumentation to assure

_ the incorporation of the recommendations of Regulatory Guide 1.97 into future instrumentation modifications. We find this commitment commendable.

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5. REFERENCES

1. Letter, NRC (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. Initrumentation for Liaht-Water-Cooled Nuclear Power Plants to Assess '

Plant and Environs Conditions Durina and Followina an Accident, Regulatory Guide 1.97, Revision 2, NRC, Office of Standards Development, December 1980.

3. Clarification of TMI Action Plan Reauirements. Reauirements for Emeraency Resoonse Capability, NUREG-0737, Supplement No."1, NRC, Office of Nuclear Reactor Regulation, January 1983.

4. Letter, Niagara-Mohawk Power Corporation-(C. V. Mangan) to NRC, April 2, .

1984.

5. Letter, Niagara Mohawk Power Corporation (C. V. Mangan) to NRC, "Raquest 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. Letter, Niagara Mohawk Power Corporation (C. V. Mangan) to NRC, December 6, 1985. '

7. Technical Evaluation Report, "Conformance to Regulatory Guide 1.97, Nine ,

Mile Point Nuclear Station, Unit No. 1," EGG-EA-6880, Revision 1, A. C.

Udy, January 1986. EC&G Idaho, Inc., Idaho National Eagineering -

Laboratory, Idaho Falls, ID 83415.

8. Letter, Niagara Mohawk Power Corporation (C. D. Terry) to _NRC, May 19,

-1989, -NMPil 0401.

9. Letter, Niagara Mohawk Power iorporation-(C. D. Terry) to NRC, July 31,

-1989.

10. Letter, Niagara Mohawk Power Corporation (C. D. Terry) to NRC, May 25, 1990, NMPILO507.

11. Letter, Niagara Mohawk Power Corporation (C. D. Terry) to P"'

-October 29,-1990, NMPll 0534.

12. _ Letter, Niagara Mohawk Power Corporation (C. D. Terry) to NRC, t_ August- 26,-1991, NMPll-0601.

{ 13. Letter, NRC Region I (J. P. Durr) to Niagara Mohawk Power Corporation:

(L.'Burkhardt'III), " Inspection No. 50-220/89-35," February 1, 1990.

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! Stember 1991 4 OR GR ANT NvM8tp A6483 b AUTHORt52 6 T vPE OF REPORT l Alan C. Udy Technical Evaluation

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8 Pl a FOad aON'O OR MING, MGMnkl mOska Regulatory and Technical Assistance EG&G Idaho,'Inc. j P.O. Box 1625 1 Idaho Falls, ID 83415-2409 et rea carror pre,.oc 4 AC O.,ssoa O*' ace or 8evea v 5 4.csee, meenwoe, Comne'.s.**

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, Office of Nuclear Reactor Regulation L U.S. Nuclear Regulatory Commission l Washington, DC 20555 a 10 SUPPLEMENT ARY NOTis

11. A857 R AC T t/oc.orm er arms This EG&G Idaho, Inc. , report documents the review of the Regulatory Guide 1.97, Revision l 2, submittals for Unit No.1 of the Nine Mile Point Nuclear Station, and identifies areas of nonconformance to the regulatory guide. Exceptions to Regulatory Guide 1.97 cre evaluated and.those areas where sufficient basis for acceptability ic, not provided are identified.  ;

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