ML20136G180

Conformance to Reg Guide 1.97,WPPSS Nuclear Project 2

ML20136G180
Person / Time
Site:	Columbia
Issue date:	03/31/1985
From:	Udy A EG&G IDAHO, INC.
To:	NRC
Shared Package
ML17278A349	List: ML20136G180
References
CON-FIN-A-6493, RTR-REGGD-01.097, RTR-REGGD-1.097 TAC-59516, NUDOCS 8507190539
Download: ML20136G180 (15)

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CONFORMANCE TO REGULATORY GUIDE 1.97 WASHINGTON PUBLIC POWER SUPPLY SYSTEM, NUCLEAR PROJECT NO. 2 A. C. Udy i

Published March 1985 EG&G Idaho. Inc.

Idaho Falls. Idaho 83415 Prepared for the U.S. Nuclear Regulatory Commission Washington. 0.C. 20555 under DOE Contract No. DE-AC07-761001570 FIN No. A6493

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ABSTRACT This EG&G Idaho, Inc., report provides a review of the submittals for Regulatory Guide 1.97 for the Washington Public Power Supply System Nuclear Project No. 2. Any exceptions to these guidelines are evaluated and those areas where sufficient basis for acceptability is not provided are identified.

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 Systems Integration, by EG&G Idaho, Inc., NRC Licensing Support Section.

The U S. Nuclear Regulatory Commission funded the work under authorization B&R 20-19-40-41-3.

I Docket No. 50-397

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. CONTENTS 1

i ABSTRACT .........................................................'...... 11 FOREWORD ............................................................... ii i 1. INTRODUCTION ...................................................... 1 1

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2. REVIEW REQUIREMENTS ............................................... 2 I

3. EVALUATION ........................................................ 4 3.1 Adherence to Regulatory Guide 1.97 .......................... 4 1

3.2 Type A Variables ............................................ 4 3.3 Exceptions to Regulatory Guide 1.97 ......................... 5 3

4. CONCLUSIONS ....................................................... 10

5. REFERENCES ........................................................ 12 i

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CONFORMANCE TO REGULATORY GUIDE 1.97 WASHINGTON PUBLIC POWER SUPPLY SYSTEM. NUCLEAR PROJECT NO. 2 i

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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 Guide 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 ActionPlanRequirements"(Reference 3).

The Washington Public Power Supply System, the licensee for Nuclear Project No. 2, provided a response to the generic letter on April 15, 1983

] (Reference 4). The letter referred to Section 7.5.2.3e of the Final Safety 1

Analysis Report (Reference 5) for a review of the instrumentation provided for  ;

l Regulatory Guide 1.97.

This report provides an evaluation of this material.

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j . 2. REVIEW REQUIREMENTS j i

Section 6.2 of NUREG-0737. Supplement No. 1 sets forth the documentation to be submitted in a report to the NRC describing how the licensee complies to Regulatory Guide 1.97 as applied to emergency response facilities. The submittal should include documentation that provides the following information i

for each variable shown in the applicable table of Regulatory Guide 1.97.

1. Instrument range

2. Environmental qualification

3. Seismic qualification

4. Quality assurance i-l 1 5. Redundance and sensor location 1

6. Power supply i l 7. Location of display l ~

i 8. , Schedule of installation or upgrade.

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a l I Furthermore, the submittal should identify deviations from the regulatory l l guide and provide supporting justification or alternatives.  !

Subsequent to the issuance of the generic letter, the NRC held regional

seetings in February and March 1983, to answer licensee and applicant t questions and concerns regarding the NRC policy on this subject. At these l l meetings, it was noted that the NRC review would only address exceptions taken i j to Regulatory Guide 1.97. Furthermore, where licensees or applicants l explicitly state that instrument systems conform to the provisions of the ,

I guide. It was noted that no further staff review would be necessary.  !

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Therefore, this report only addresses exceptions to Regulatory Guide 1.97.

The following evaluation is an audit of the licensee's submittals based on the ,

review policy described in the NRC regional meetings.

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. 3. EVALUATION ThelicenseeprovidedaresponsetotheNRCGenericLetter85-33on

! April 15, 1983. This response referred to the Final Safety Analysis Report (FSAR)whichdescribesthelicensee'spositiononpost-accidentmonitoring 1

instrumentation. This evaluation is based on these materials.

3.1 Adherence to Regulatory Guide 1.97 The licensee states, in Section 7.5.2.2.3e of the FSAR, that the FSAR provides an item by item discussion on the instrumentation used to conform to l the guidance of Regulatory Guide 1.97. Therefore, it is concluded that the

) licensee has provided an explicit comunitment on conformance to Regulatory i

Guide 1.97, except for those exceptions that are justified as noted in l Section 3.3.

3.2 Type A Variables 1

, Regulatory Guide 1.97 does not specifically identify Type A variables.

l 1.e., those variables that provide information required to permit the control room operator to take specific manually controlled safety actions. The licensee classifies the following instrumentation as Type A.

1. Neutron flux .

2. Coolant level in reactor

3. Reactor coolant system pressure 4

4. Primary containment pressure i

All of the above variables meet the Category 1 requirements consistent with the requirements for Type A variables.

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1 3.3 Exceptions to Regulatory Guide 1.97 i

! The licensee identified the following deviations and exceptions to i Regulatory Guide 1.97. These are discussed in the following paragraphs.

1.1.1 Nantenn Flux j The instrumentation supplied by the licensee for this variable complies j with the range and the Category 1 recommendations except that the four source l and the eight intermediate range detector drive units that are not qualified j to Category 1 requirements. These drive units remove the detector from the i core when operating at power. They are only required post-accident to drive

! the detectors into the core. The source range detectors cover a range of 10-3 l to 10 percent of full power in the fully withdrawn position, 10-7 to l 10-3 percent of full power when fully inserted. This, according to the licensee, is sufficient to insure that the reactor is subcritical. There are eight similar intermediate range drive units and detectors which cover higher

! .cora power levels. The licensee states that if all the drive units failed,

! and the source range monitors remained out of core, the indicated range

(minimumof10-3 percentoffullpower)issufficienttoinsurethe sub-criticality of the reactor.

I l In the process of our review of the neutron flux instrumentation for I boilingwaterreactors(BWRs),wenotethatthemechanicaldrivesofthe j detectors have not satisfied the environmental qualification requirements of i P.egulatcry Guide 1.97. A Category 1 system that meets all the criteria of l Regulatory Guide 1.97 is an industry development item. Based on our review, j we conclude that the existing instrumentation is acceptable for interim i operation. The licensee should follow industry development of this equipment.

l evaluate newly developed equipment, and install Category 1 instrumentation l when it becomes available.

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l l 3.3.2 Coolant Level in Reactor j Regulatory Guide 1.97 recommends instrumentation with a range from the j bottom of the core support plate to either the top of the vessel or the i centerline of the main steamline. The instrumentation supplied by the a

{ licensee covers a range from 150 in, below the top of active fuel to +60 in.

! above the dryer skirt. We have insufficient information to determine if the 4

bottom of the core support plate or if the centerline of the main steamline is included in the supplied range. The licensee has not justified this deviation from the range recommendations.

j The licensee should provide the correlation between the supplied and the recommended ranges, identify any deviation and justify any deviation.

l 3.3.3 Drywell Sump Level l Orywell Drain Sumps Level i

Regulatory Guide 1.97 recommends instrumentation for this variable. The i licensee has not supplied instrumentation for this variable. The licensee ,

l indicates that in a post-accident situation the sump drain lines are isolated l and the sump overflow goes to the suppression pool via downcomers.

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The licensee has not provided acceptable justification for not providing

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l instrumentation for these variables. The sump level instrumentation is the i

primary means to determine identified and nonidentified leakage rates.

Operator actions are based on the source and extent of the leakage. The

licensee should provide information describing how the level of the drywell j and the drywell drain sumps are ascertained during and following an accident.

3.3.4 Radioactivity Concentration The licensee indicates that radiation level measurements to indicate fuel cladding failure are provided in the pre-isolation condition by the condenser 6

o.ff-gas radiation monitors and by the main steamline radiation monitors; ir the post-accident condition by the post-accident sampling system.

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

3.3.5 Suppression Pool Water Level l

The instrumentation supplied by the licensee for this variable covers a range of i25 in. of normal water level. This does not conform to the recommended range from the bottom of ECCS suction line to five feet above normal water level (for a Type C variable) or from the top of vent to top of weir well (for a Type D variable). The licensee has not justified this deviation from the range reconnendations.

The licensee should provide the correlation between the supplied and the recommended range and satisfactorily justify the deviations identified or provide the reconnended range.

3.3.6 Suppression Chamber Spray Flow

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Theresidualheatremoval(RHR)systemflowisusedforthisvariable.

The suppression pool spray derives its flow from the RHR system, with a throttling valve proportioning the flow between the suppression pool spray and ,

the drywell spray. The position of the throttling valve is controlled from the control room. Pressure and temperature changes in the suppression pool determine the effectiveness of the spray.

The licensee concludes that RHR flow and suppression chamber pressure accurately and reliably measure the effectiveness of the suppression chamber

spray. Additionally, the RHR system values positions are known is the control

! room. We find that this instrumentation is adequate for this variable.

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3.3.7 Dr.ywell Atmosphere Temperature The instrumentation supplied by the licensee for this variable covers a l range of 50 to 400*F. Regulatory Guide 1.97 recommends a range of 40 to 440*F for this variable. This deviation in range has not been justified. The i licensee should justify the deviation from the range recommendations, or J

j re-span the instrumentation to provide the range recommended by Regulatory j Guide 1.97.

i 3.3.8 Drywell Spray Flow The residual heat removal (RHR) system flow is used for this variable.

l The drywell spray derives its flow from the RHR system, with a throttling valve proportioning the flow between the suppression pool spray and the

] drywell spray. The position of the throttling valve is controlled from the I control room. Pressure changes in the drywell determine the effectiveness of

! the spray.

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j The licensee concludes that RHR flow and drywell pressure accurately and reliably measure the effectiveness of the drywell spray. Additionally, the I

RHR system valves positions are known in the control room. We find that this instrumentation is adequate for this variable.

I i 3.3.9 Residual Heat Removal Heat Exchancer Outlet Temoerature Regulatory Guide 1.97 recomends Category 2 instrumentation for this variable. The licensee has provided Category 3 instrumentation for this s i variable.

l l The licensee states that the supplied instrumentation is adequate for f monitoring this variable, however, they have not included the basis for this '

l conclusion. The licensee should provide justification for this deviation from l l the Category 2 recomendations, or upgrade the instrumentation to j Category 2.

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3.3.10 Cooling Water Temperature to ESF System Components Regulatory Guide 1.97 recommends a range of up to 200*F for this variable. The instrumentation supplied by the licensee for this variable has an upper limit of 150*F. The licensee has not justified this deviation from the range recommendations. The licensee should supply adequate justification for this deviation.

3.3.11 Plant and Environs Radioactivity (Portable Instrumentation)

I Regulatory Guide 1.97 recommends a multichannel gamma-ray spectrometer for this variable. The licensee has not provided instrumentation for this variable, nor justification for not providing this instrumentation. The

licensee should provide this instrumentation.

! 3.3.12 Estimation of Atmospheric Stability The instrumentation supplied by the licensee for this variable covers a range of 115*F instead of the range recommended by the regulatory guide. -9 to 18'F. The licensee has not justified this deviation from range recommendations between +15 to 18'F. -

Table 1 of Regulatory Guide 1.23 provides seven atmospheric stability classifications based on the difference in temperature per 100 meters i

elevation change. These classifications range from extremely unstable to extremely stable. Any temperature difference greater than +4 or less than

-2*C does nothing to the stability classification. Therefore, we find that this instrumentation is acceptable to determine the atmospheric stability.

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. 4. CONCLUSIONS Besed on our review, we find that the licensee either conforms to or is justified in deviating from Regulatory Guide 1.97, with the following

exceptions

1. Neutron flux--the licensee's present instrumentation is acceptable ,

j on an interim basis until Category 1 instrumentation is d2veloped I

and installed (Section 3.3.1).

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2. Coolant level in reactor--the licensee should provide additional information for this variable and justify any deviation (Section3.3.2).

3. Drywell sump level--the licensee should provide additional justification for not supplying this instrumentation

(Section3.3.3).

i 4. Drywell drains sump level--the licensee should provide additional l justification for not supplying this instrumentation j (Section3.3.3).

5. Suppression pool water level--the licensee should justify the existing range or should provide the recommended range (Section3.3.5).

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6. Drywell atmosphere temperature--the Itcensee should justify a deviation from the reconnended range or supply the recommended range <

l (Section3.3.7).

) 7. Residual heat removal heat exchanger outlet temperature--the licensee should provide justification for deviating from Category 2 reconnendations for this variable, or supply instrumentation that is

Category 2 (Section 3.3.9).

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. 8. Cooling water temperature to ESF system ccmponents--the licensee sould justify the deviation from the recommended range i

(Section3.3.10).

9. Plant and environs radioactivity (portable instrumentation)--the Mcensee should provide instrumentation for ;this variable t

(Section3.3.11).

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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 i

Capability (Generic Letter No. 82-33) " December 17, 1982.

2. Instrumentation for Licht-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conc itions Durinc and Following an Accident, Regulatory Guide 1.97, Revision 2. L.S. Nuclear Regulatory Commission

(NRC),OfficeofStandardsDevelopment, December 1980.

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

4. Washington Public Power Supply System (WPPSS) lettar, G. D. Ouchey to

Director of Nuclear Regulatory Regulation, NRC, " Emergency Response l Capability " April 15, 1983, G02-83-346.

i 5. WPPSS Nuclear Project No. 2. Final Safety Analysis Report, Amendment j N . 23, February 1982.

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