ML20235N525

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Rev 1 to Final Conformance to Reg Guide 1.97,TMI-1, Informal Rept
ML20235N525
Person / Time
Site: Three Mile Island Constellation icon.png
Issue date: 06/30/1987
From: Stoffel J
EG&G IDAHO, INC., IDAHO NATIONAL ENGINEERING & ENVIRONMENTAL LABORATORY
To:
NRC
Shared Package
ML20235N513 List:
References
CON-FIN-A-6483, RTR-REGGD-01.097, RTR-REGGD-1.097 EGG-NTA-7079, EGG-NTA-7079-R01, EGG-NTA-7079-R1, NUDOCS 8707200114
Download: ML20235N525 (26)
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  • U.S. NUCLEAR REGULATORY COMMISSION No. DE A%7-761D0tS70 '

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DISCLAIMER The book was prepared as an account of work sporisored by an agency of the Unked l States Government. Neither the, United States Government nor any agency thereot, I ner any of their employees, makes any warranty, express or emphed, of assumes any l legal liabskty or responsibtfity for the accuracy, completeness, or usefulness of any in9ormation, apparatus, product or process disclosed, or represents that its uw would j

not infringe pnvately owned rights. References herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise, does not necessanly constitute or imply its endorsement, recommendation, or favonng by the United States Government or any agency thereof. The views and opinions of I authors expressed herein do not necessanly state or reflect those of the United States l Government or any agency thereof.

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TECHNICAL EVALUATION REPORT CONFORMANCE TO REGULATORY GUIDE 1.97 THREE MILE ISLAND-1 Docket No. 50-289 J. W. Stof fel Published June 1987 Idaho National Engineering Laboratory EG&G Idaho, Inc.

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Prepared for the U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Unde

  • DOE Contract No. DE-AC07-76ID01570 FIN No. A6483

l ABSTRACT This EG&G Idaho, Inc., report. reviews the submittal for Regulatory Guide 1.97, Revision 3, for Unit No. 1 of the Three Mile Island Nuclear Station and identifies areas of nonconformance to the regulatory guide.

Exceptions to Regulatory Guide.1.97 are evaluated and those areas where sufficient basis for acceptability ~is not provided are identified.

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Decket No. 50-289 TAC No. 51361 ii

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

The U.S. Nuclear Regulatory Commission funded the work under authorization B&R 20-19-10-11-3.

Docket No. 50-289 TAC No. 51361 iii

CONTENTS ABSTRACT .......... ................................................... 11 FOREWORD .............................................................. iii

. 1. INTRODUCTION ..................................... ............... 1

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 ........ .............. 5

4. CONCLUSIONS .. ................................................... 18
5. REFERENCES .............. ... .................................... 19 l

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I

F CONFORMANCE TO REGULATORY GUIDE 1.97 THREE MILE ISLAND-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 'eactors, 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).

GPU Nuclear Corporation, the licensee for the Three Mile Island Nuclear Station, provided a response to Section 6.2 of the generic' letter on October 1, 1984 (Reference 4). Additional information was submitted on June 5,1986 (Reference 5) with additional updated information on May 7, 1987 (Reference 6). These responses provide a comparison of the licensee's instrumentation to the recommendations of Revision 3 of Regulatory Guide 1.97 (Reference 7).

This report provides an evaluation of that material.

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2. 'REV!EW' REQUIREMENTS-t: ,

LSection 6.2 of NUREG-0737, Supplement No. 1, sets forth the~

documentation to be ' submitted.in a report;to the NRC describing how the

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-licensee complies with Regulatory Guide 1.97' as applied to emergency- 'E response facilities. The submittal.should-include documentation that provides the following'information for each variable shown in the 'I applicable table of Regulatory Guide 1.97.

1. Instrument range
2. Environmental qualification'
3. Seismic-qualification

'4. Quality' assurance

5. Redundance and sensor'iocation
6. Power supply
7. Location of display

'8. Schedule of installation or upgrade The submittal should identify any deviations from the recommendations of-Regulatory Guide.l.97 and provide supporting justification or alternatives for the deviations identified.

Subsequent to the issuance of the generic letter, the NRC held regiona? meetings in February and March 1983, to answer licensee and ,

applicant questions and concerns regarding the NRC policy on this subject.

At-these meetings, it was noted that the NRC review would only adcress exceptions taken to Regulatory Guide 1.97. Where licensees or applicants explicitly state that instrument systems conform to the regulatory guide, 7

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it was noted that no further staff review would be necessary. Therefore,

j. 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 6.2 of NRC Generic Letter l 82-33 on 0ctober.1,.1984. ' Additional information was submitted on ,

June 5,i1986 with additional updated information.on May 7, 1987. The.

  • responses describe:the licensee's-position on post-accident monitoring ,

instrumentation. This evaluation is based on that material.,

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3.1' Adherence to Reoulatory Guide 1.97 1 The. licensee has provided a review of their. post-accident monitoring- J l

, instrumentation that compares the instrumentation characteristics against -

the recommendations'of Regulatory Guide 1.97,. Revision 3. The review. lists  ;

the regulatory guide variables, showing either full' compliance, noncompliance with justification, or noncompliance with'a commitment and schedule.to. upgrade. The licensee states that all upgrade modifications are scheduled for completion by the second refueling outage after restart, >

designated' refueling outage 7R. 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 Type A Variables J Regulatory Guide 1.97 does not specifically identify Type A variables,

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i.e., those variables that provide the information required to permit the control room operator to take specific manually controlled safety actions.

The licensee classifies the following instrumentation as Type A. )

1. Reactor coolant system (RCS) cold leg water temperature >
2. RCS pressure

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3. Core exit temperature 4

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4. Degrees of subcooling 1'
5. Containment hydrogen concentration
6. Low pressure injection / decay heat removal system flow
7. Flow in high pressure injection system
8. Refueling water storage tank level
9. Steam generator level
10. Steam generator pressure
11. Auxiliary or emergency feedwater flow
12. Condensate storage tank water level This instrumentation meets the Category I recommendations consistent with the requirements for Type A variabies, except as noted in Section 3.3.

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

3.3.1 Reactor Coolant Systen (RCS) Soluble Baron Concentration Regulatory Guide 1.97 recommends on-line instrumentation with a range

- of 0 to 6000 ppm. The licensee has not provided this on-line instrumentation, but can obtain the information by utilizing the

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post-accident sampling system and on-site laboratory analysis.

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1 The license'e deviates from Regulatory Guide 1.97 with respect to post-accident sampling capability. This' deviation goes beyond the scope of j

this review and is being addressed by the.NRC.as part of their review of NUREG-0737, Item II.B.3.

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  • 3.3.2 RCS Cold Leg Water Temperature

' Regulatory Guide. l.97 recommends instrumentation with a range of 50 to 700*F Torithis variable. The licensee has supplied instrumentation with a'-

range of=50 to 650 F. The licensee. considers the existing range adequate

-based on-the maximum steam generator pressure of 1200 psig and.

corresponding. saturation temperature of 600 F. Therefore, the cold leg temperature would always be at or below this value.

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Based on the-licensee's statement that the instrumentation will remain

.on. scale for any anticipated event,-we find the range of this

-instrumentation acceptable.

3.3.3 RCS' Hot Leg Water Temperature Regulatory Guide .1.97 recommends instrumentation with a range of 0 to 700'F for this variable. The ficensee has supplied instrumentation with a range of 120 to 920 F. The licensee. states that at temperatures less'than 300 F, the plant will be in the decay heat removal mode, in cold shutdown, and this temperature is not then required. The decay heat removal system has additional temperature instrumentation to monitor the RCS in this temperature range. Category 1 core exit thermocouple also provide information below 120 F.  ;

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Based on the alternate instrumentation and the justification provided by' the licensee, we conclude that the instrumentation supp11 ed for this variable is adequate and, therefore, acceptable. ',

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3.3.4 RCS Pressure j Regulatory Guide 1.97 recommends instrumentation for this variable with a range from 0 to 3000 psig. In Reference 4, the licensee. stated that instrumentation with a range from 0 to 2500 psig is provided for this i variable. The licensee stated that no additional. operator action would be taken or performed with an extended range from 2500 to 3000 psig, and that the code safety valves on the pressurizer are set to relieve pressure at 2500 psig.

In Reference 5, the licensee states that they will be in compliance with the regulatory guide requirement by refueling outage 7R. We find this commitment acceptable.

3.3.5 Radiation Level in Circulating Primary Coolant The licensee indicates that radiation level measurements to indicate fuel cladding failure are provided by the following instruments:

1. Letdown line radiation monitors (during normal operation)
2. Post-accident sampling system.

The post-accident sampling system is available with the reactor isolated, and is being reviewed by the,NRC as part of their review of NUREG-0737, Item II.B.3.

I Based on the alternate instrumentation provided by the licensee, we conclude that the instrumentation supplied for this variable is adequate

. and, therefore, acceptable.

. 3.3.6 RHR Heat Exchanger Outlet Temperature Regulatory Guide 1.97 recommends Category 2 instrumentation for this variable with a range of 40 F to 350 F. As such, environmentally qualified 7

'i instrumentation is' required .in accordance with 10 CFR 50.49, LIn Reference 4,. the.' licensee states that the existing range of .0 to 300 F is

-sufficient-to cover all post-acc.ident conditions since decay heat removal operation.is initiated when the'RCS temperat'ure is below 300 F.

Based on the licensee's justification, we find this range adequate to '

monitor this variable during all accident and post-accident conditions, ,

however, the licensee did not provide justification for the environmental qualification deviation in Reference 4.

In Reference 5, the licensee states that this instrumentation has been incorporated on the TMI-1 environmental qualification master list. We find

'this acceptable.

3.3.7. ' Accumulator Tank Level and Pressure Regulatory Guide 1.97 recommends Category 2 instrumentation for this variable. The licensee has provided Category 3 instrumentation that, except'for environmental qualification, is Category 2. The licensee justifies this deviation in Reference 4 by stating that these instruments provide the operator information pertaining to tank status during normal operation, and that since the core flooding system is totally passive, no monitoring'of these parameters is required for any manual actions to mitigate the consequences of an accident. Reference 5 restated the licensee's position that Category 3 instrumentation is adequate to monitor this variable.

The existing instrumentation is not acceptable. .An environmentally qualified instrument is necessary to monitor the status of these tanks.

The licensee should designate either level or pressure as the' key variable -

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to determine accumulator discharge and provide instrumentation for that variable that meets the requirements of Regulatory Guide 1.97 and -

10 CFR 50.49.

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Y 3.3.8 Accumulator Tank Isolation Valve Position Regulatory. Guide 1.97 recommends Category 2-instrumentation for this-Lvariable. The licensee statesLthat these are motor operated valves. They:

are open for' reactor operation. The circuit breakers for these valves are

? open'and de-energized when the reactor is critical. Therefore, the licensee recommends that.this variable be reclassified as Category 3.

Based on .the licensee's, justification and the fact that these valves.

are open'and do not change. position during or following an accident, we consider. Category 3 instrumentation adequate for this variable.

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3.3,9 Boric Acid Chargina Flow

.The. licensee'does not have' instrumentation for this variable. The licensee states that the charging system is not part of the emergency core cooling system (ECCS). High pressure injection and low pressure. injection are the flow' paths of the ECCS that are monitored. Therefore, we find that this variable is not applicable at the Three Mile Island Station.

3.3.10 pressurizer. Level Regulatory Guide 1.97 recommends Category 1 instrumentation for this variable. The licensee ' considers pressurizer level instrumentation to be Category 2.and has not met the environmental qualification requirement for the temperature compensation elements.

The; justification provided by the licensee for the Category 1 deviation-is that pressurizer level is only used as an indicator to the

. operator that throttling of the high pressure injection flow is allowed.

Therefore, the licensee's position is that the pressurizer level is an

-- indication.of system operating status but is not a key variable. In Reference 5, the licensee submitted a table showing the effects of the loss 9

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L 'of a' temperature' compensation element.on the indicated. pressurizer level.

lThis table indicates.that if an element.is shorted at a pressurizer

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. temperature ~ greater than 100 F the pressurizer could be water solid, and indicate on scale. If ar, element fails open, at low pressurizer levels, level will indicate off-scale low.' If an element fails open at'high

pressurizer . levels, indication will be off-scale high before the ,

l pressurizer is:actually water solid, *

-It appears that the' licensee is addressing their justification for this deviation on' the' need for pressurizer level instrumentation for ccre cooling only. However, the purpose of this instrumentation, as stated in Regulatory Guide 1.97, is to ensure proper operation of the pressurizer.

-Pressurizer. level is a key variable' used to ensure proper operation of the pressurizer. The licensee has not provided sufficient justification for -

deviating from the regulatory guide' requirements for this variable. The licensee should : commit. to installing Category:1 Temperature Compensation elements for this variable.

3.3.11. Pressurizer Heater Status Regulatory Guide 1. 97 recommends instrumentation to- monitor the current drawn by the pressurizer heaters. The licensee's instrumental',on consists of on/off indication of the pressurizer heaters. The licensee considers this to be sufficient indication when used in conjunction with

- RCS' pressure.

Section II.E.3.1 of NUREG-0737 requires a number of the pressurizer heaters to have the capability of being powered by the emergency power sources. Instrumentation is to be provided to prevent overloading a diesel generator.

In Reference 5, the licensee has maintained the position that an -

on-off mode of indication is adequate to monitor this variable. The licensee states that the most direct and effective measure of heater performance is the response of reactor coolant pressure. The 10

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l licensee further states that the diesel current can be monitored with the diesel ammeters which enables the operator to determine (based on the known power consumption of the heaters) whether he can load the heaters without overloading the diesels.

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- We find the justification provided by the licensee unacceptable. The light indicating the pressurizer heater circuit breaker is closed does not

indicate that the heaters are in fact energized or what amount of heaters are working. A means of monitoring pressurizer heater current in the control room should be provided.

3.1.12 Quench Tank Temperature Regulatory Guide 1.97 recommends instrumentation for this variable with a range from 50 to 750'F. The installed instrumentation has a range of 0 to 275'F. The licensee states that.the tank is isolated with a reactor trip and that the existing temperature range is adequate to detect leakage into the tank.

In Reference 5, the licensee states that a relief valve set at 40 psig (saturation temperature 287'F) and a rupture disc set at 55 psig (saturation temperature 308'F) are installed on the tank. This means that-for a short period of time the temperature of the tank could be above the existing range. The licensee has comitted to provide the capability to monitor the complete postulated temperature range by refueling outage 7R.

We find this commitment acceptable.

3.3.13 SteamGeneratorLevel Regulatory Guide 1.97 recommends instrumentation for this variable

[ with a range that monitors the steam generator level from the tube sheet to the separators. In Reference 5, the licensee indicated compliance te the level recommendation and added additional clarification in Reference 6.

The clarification states that the TMI-1 steam generator is a once-through design and as such the heat exchange area egld be described as tube sheet 11

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tbtubesheet. The'l'icenseehas.redundantCategory1l full,rangelevel

, . indicators on each steam generator with a range.of.'0-640 inches.. The .

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instrument zero is actually 6 inches, above. the lower tube sheet. Should the; water level reach this level, the steam generator is essentially dry.

1 Therefor 2, we find the, lower tap'locat',on; acceptable.

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The licensee furtherl states that during design basis accidents their analyses; shows that the range of interest'is up to the aspirating port .

(376 inches above the lower tube sheet). Reference.6 identifies the-

installation of recorder LR1046/LR1054 in the control room for the steam n generator level. This recorder records the " start-up" ranse level'of.each o, steam generator, 0-388 inches,-which covers the range of interest for-design basis
accidents. L As this range includes the ' aspirating port, we
find the range acceptable.

.Ba' sed.on theljustification provided by the licensee that the instrumentation will' remain on scale and functional for the analyzed

transients.snd accidents we find the range provided' acceptable.

3.3.14 Safety / Relief Valve Positions or Main Steam Flow

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. Regulatory Guide 1.97. recommends Category 2 instrumentation for this variable. ~The licensee has provided Category 3 instrumentation.- The licensee states that Category 3 instrumentation is acceptable for this.

variable because they consider the key variables to determine tte steam generator (S0) safety / relief valve position'or main steam flow to be SG 1evel and SG pressure. Valve position indication is provided as backup instrumentation.

The' licensee consioers the valve position indication to be a backup for the Category 1 steam generator level and pressure instrumentation. As .

the regulatory guide allows backup instrumentation to be ' Category 3, we ,

find this deviation acceptable.

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3.3.15 Containment Spray Flow 1

R(:gulatory Guide 1.97 recommends Category 2' instrumentation for this

variable.. The licensee has provided instrumentation that, except for environmental qualification, is Category 2. The licensee did not submit justification for the environmental qualification deviation in Reference 4.

-- In Reference 5, the licensee states that this instrumentation has been incorporated in the TMI-1 environmental qualification master list. We find l this acceptable.

l 3.3.16 ' Heat Removal by the Containment Fan Heat Removal System Regulatory Guide 1.97 recommends Category 2 instrumentation for this variable to provide indication that the reactor building cooling system is l performing its design objective. In Reference 5, the licensee committed to install Category 2 instrumentation for this variable. In Reference 6, the licensee re-evaluated the instrumentation provided for this variable and

' determined that containment pressure is the key variable for this function with other indications for this variable used for backup instrumentation.

The Category 3 instrumentation used for backup at TMI-1 includes indication of river water pump motor breaker status, reactor building fan motor.

contractor status, indication of each reactor building emergency cooling i-coi! outlet pressure and annunciation of leak detection for each reactor building emergency cooling coil.

As the containment pressure is affected by both the containment fan heat removal system and the containment spray system, and is a function of break size and location, we find that the containment pressure does not I show conclusively that the containment fan heat removal system is

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. operating. While the Category 3 backup instrumentation listed by the licensee can indicate a problem with the system, a direct Category 2 method of monitoring the operation of the heat removal system is required. -

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3. 3 J ' Containment Atmosphere Temperature Regulatory Guide l'.97 recommends Category 2 instrumentation for this variable with a range from 40-to 400 F. The licensee has supplied Category 3 instrumentation with a range of 0 to 300 F. Their justification ,

for this deviation is that the primary variable required to show accident mitigation and containment integrity is reactor building pressure, a ,.

Category I variable. The licensee considers the containment atmosphere temperature to be a Category 3 variable, A The key variable used by the licensee for reactor building monitoring is the reactor building pressure, which is monitored by Category 1 instrumentation; the reactor building atmosphere temperature is a backup variable for reactor building accicent monitoring, and as such, is measured by Category 3 instrumentation.

We find that the licensee's application of Category 3 backup instrumentation is in accordance with the regulatory guide.

The licensee states that the presently installed 0 to 300 F containment temperature indicators provide sufficient range to monitor the entire spectrum of containment temperature transients as analyzed in the p FSAR.

Based on this justification, we find that the existing range is adequate to monitor this variable during all accident and post-accident conditions.

3.3.18 Containment Sump Water Temperature Regulatory Guide 1.97 recommends Category 2 instrumentation for this I l

variable. The licensee has supplied instrum6ntation that, except for ~

j environmental qualification, is Category 2. In Reference 4 the licensee )

states that the n:inimum available net positive suction Laad for the decay I heat removal pump is independent of sump temperature and no automatic or j l

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manual actions are initiated based on this temperature. No additional justification was provided by Reference 5.

The temperature of the sump water is useful to the operator in determining the amount of containment heat removed during recirculation.

. Therefore, an environmentally qualified means of determining the containment sump water temperature should be provided by the licer.see.

3.3.19 Letdown Flow-Out Regulatory Guide 1.97 recommends Category 2 instrumentation for this variable. The licensee does not consider this variable to be a post-accident Category 2 instrument, and has supplied Category 3 instrumentation. The licensee states that this variable is not required in the mitigation of an accident and that the letdown system is isolated by any accident requiring containment isolation.

As this variable is not utilized in conjunction with a safety system, we find that the instrumentation provided for this variable is acceptable.

3.3.20 Component Coolina Water Temperature to Enaineered Safety Feature (ESF) System Regulatory Guide 1.97 recommends Category 2 instrumentation for this variable. The licensee is supplying instrumentation that, except for environmental qualification, is Category 2. The licensee states that the decay heat removal heat exchanger outlet temperature provides an adequate measure of the decay heat removal closed cooling water system heat removal capability.

In Reference 5, the licensee states that this instrumentation is not

- located in a harsh environment, therefore, qualification to the requirements of 10 CFR 50.49 is not required. Thus, we find this instrumentation acceptable.

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3.3.21 ' Component Cooling Water Flow to ESF System Regulatory Guide 1.97 recommends Category 2 flow instrumentation for this variable. The licensee does not have instrumentation for this variable. The~ licensee justified this exception in Reference 4 by stating that,'since all decay heat and nuclear services closed cycle cooling systems component cooling water valves are manually operated and are ,

normally open, pump status and system temperature is sufficient indication for system operation.

In Reference 5, the licensee gave information on the availability of pump discharge pressure indication and low flow alarmsd n the control room in addition to the pump status and temperature indication. We conclude that the intent of Regulatory Guide 1.97 is met with the instrument,2 ion prov1ded. Therefore, we find this deviation acceptable.

3.3.22 Radioactive Gas Holdup Tank pressure Regulatory Guide 1.97 recommends control . room instrumentation for this variable with a range of 0 to 150 percent of design pressure. The licensee has local indication only. The licensee states that the design pressure for these tanks is 150 psig. When the pressure reacnes 82 psig, it initiates a local high pressure alarm. Also, the pressure can be indicated on a local indicator on cemand. At 85 psig, the relief valve opens and discharges to the auxili'ry a building, where it will be detected and indicated by the auxiliary building radiation monitor. Also, when the relief valve opens, it will annunciate in the common problem panel in the control room.

In Reference 4, the licensee did not state what the range of the local -

indicator is nor state that the instrumentation is accessible post-accident. In Reference 5, the licensee states that the range of the I

local instrument is 0-100 psig, whicn is adequate to monitor the tank pressure. The licensee also stated that the local indication is available on tne radioactive waste control panel which is accessible after an accident. We find this acceptable.

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3.3.23 Status of Standby Power and Other Energy Sources Important to Safety: ,

Regulatory ~ Guide'1.97 recommends Category 2 instrumentation for this variable. The. licensee has provided instrument. air instrumentation that,

,. except for environmental' qualification, is Category 2.

. In Reference 5, the licensee states that this instrumentation is not located in a harsh environment, therefore,. qualification to the requirements of 10 CFR 50,49 is not required. We find this acceptable.

I 3.3.24 Vent from Steam Generator Safety Relief Valves or Atmospheric Dump Valves The. instrumentation provided fur this variable has a range of l- 3.96 x 10 -2 to 980 pCi/cc. Regulatory Guide 1.97 recommends 10 ~I to 3

10;pCi/cc. The existing range does not envelop the upper end of the recommended range. ..The existing range deviates from the recommended range by 20 pCi/cc, but is adequate to provide tne necessary accident and post-accident information. Therefore, this is an acceptable deviation.from Regulatory Guide 1.97.

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4. CONCLUSIONS 4

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. RCS soluble boren concentration--the NRC is addressing this

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deviation as part of their review of.NUREG-0737. Item II.B.3 (Section 3.3.1).

l 2. . Accumulator tank level and pressure--the licensee should provide a level or pressure instrument for this variable that is

. environmentally qualified in accordance with Regulatory Guide 1.97 and 10 CFR 50.49 (Section.3.3.7).

3. Pressurizer level--the licensee should commit to installing I redundant Category 1 instrumentation for this variable.

(Section3.3.10).

4. Pressurizer heater status--the licens2e should provide the recommended instrumentation (Section 3.3.11).
5. Heat removal by the containment fan heat removal system--the licensee should provide' Category 2 instrumentation that will monitor the operation of this system (Section 3.3.16).

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6. Containment sump water temperature--the licensee should identify an environmentally qualified means of monitoring this variable i (Section 3.3.18).

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o 5.. REFERENCES

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1 -. NRC letter, D. ~ G. Eisenhut to All Licensees fof Operating . Reactors, Applicants 1for-Operating Licenses. and Holders of Construction Permits, " Supplement No. l'to'NUREG-0737--Requirements.for Emergency Response Capability (Generic Letter No. 82-33)," December- 17,-1982.

2.. Instrumentation for' Light-Water-Cooled Nuclear' Poser Plants to Assess Plant-and Environs Conditions Durino and Following an Accident, O'D Regulatory Guide.l 97, Revision 2, NRC, Office of Stancards

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Development. December 1980, ii.

d4 -' Clarification of TMI' Action Plan Requirements, Requirements for

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Emergency Response Capability, NUREG-0737 Supplement No. 1, NRC, -!

Office of. Nuclear Reactor Regulation, January'1983.
4. GPU Nuclear, Corporation letter, H.E D. Hukill to Office of Nuclear' l Reactor Regulation, NRC,;0ctober 1,.1984, Serial No. 5211-84-2252 L 5. GPU Nuclear Corporation letter,.H. D. Nukill to Office of Nuclear l' Reactor Regulation, NRC, " Emergency Response. Capability-Conformance to' l Regulatory Guide 1.97," June.5, 1986, 5211-86-2097.
6. GPU Nuclear Corporation letter, H. D. Hukill to Office'of Nuclear Reactor Regulation, NRC, " Emergency Response Capability-Conformance to-Regulatory Guide,1.97," May 7, 1987, 5211-87-2087.
7. Instrumentation' for Light-Water-Cooled Nu'elear Power Plants to' Assess Plant and Environs Conditions During and following an Accident,.

Regulatory Guide 1.97, Revision 3, NRC, Of fice of Nuclear Regulatory Research, May 1983.

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.n 19

WAC POAM 335 u s. NUCLEAA A80ukatomy contwissions t ateoAY NuMeta #dmyese Ar ff0C,aar var so, seears (2 46 O"J$ 518UOGRAPHIC DATA SHEET EGG-NTA-7079 Rev. 1 84i th8?MUC? son 6 om ?>t 84vtRSE C0hFOR NCE TO REGULATOR ( GUIDE 1.97, THREE MILE ISLAND-1

(, 4 DATE Rtroni CoMPLETeo

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.our g n An l# June 1987 i > rantonvino oncA=izatsom maut Amo MAiLi=c. Acoatss um=m te Coms e emossetitAsamons unit muusta NRR and I&E Support Branch EG&G Idaho, Inc. * *iN oa caANT avMata P. O. Box 1625 Idaho Falls, 10 83415 A6483 30 kPONSoHsNG ORGAN #4 Af ton hsAMt ANo MA44tNG AooAt&s stative/sp Capet ilt TYPtofPePomY Division of Engineering and Systems Technology Office Of Nuclear . Reactor Regulation ""'"'*""'"~'""""'

U.S. Nuclear Regulatory Commission Washington, DC 20555 13 BuPPLEMENT AR v %Of t$

13 Ah4TR ACT t/00 cores et 'ess)

This EG&G Idaho, Inc. report reviews the submittals for Three Mile Island, Unit No. 1, and identifies areas of nonconformance to Regulatory Guide 1.97. Exceptions to these guidelines are evaluated and'those areas where sufficient basis for acceptability is not provided are identified.

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