ML20238D021

Conformance to Reg Guide 1.97:Oyster Creek, Interim Technical Evaluation Rept

ML20238D021
Person / Time
Site:	Oyster Creek
Issue date:	07/31/1987
From:	Udy A EG&G IDAHO, INC., IDAHO NATIONAL ENGINEERING & ENVIRONMENTAL LABORATORY
To:	NRC
Shared Package
ML20238C994	List: ML20238D021
References
CON-FIN-A-6483, RTR-REGGD-01.097, RTR-REGGD-1.097 EGG-NTA-7277, EGG-NTA-7277-01, EGG-NTA-7277-1, TAC-51115, NUDOCS 8709100515
Download: ML20238D021 (20)

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EGG-NTA-7277 TECHNICAL EVALUATION REPORT l

OYSTER CREEK CONFORMANCE TO REGULATORY GUIDE 1.97:

Docket No. 50-219 Alan C. Udy Published July 1987 Idaho National Engineering Laboratory '

EG&G Idaho, Inc.

Idaho Falls, Idaho 83415 i

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.i Prepared for the U.S. Nuclear Regulatory Commission Washicgton, D.C. 20555 Under DOE Contract No. DE-AC07-761001570 FIN No. A6483 B709100515 870904 PDR ADOCK 05000219 P pm

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k ABSTRACT l

This EG&G Idaho, Inc., report reviews the submittal's for Regulatory Guide 1.97, Revision ~2, for the Oyster Creek Nuclear. Generating Station.

Any exceptions to Regulatory Guide 1.97 are evaTuated and those areas where sufficient basis for acceptability is not provided are identified.

Docket No. 50-219 TAC No. 51115 11 l

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l FOREWORO 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, Divisior. of Engineering and System Technology, by EG&G' Idaho, Inc. ,

Electrical, Instrumentation and Control Systems Evaluation Group.

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

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Docket No. 50-219 TAC No. 5;1'15 iii j i

l CONTENTS l

ABSTRACT .. .. . .... ........ ................... . ., ..., .. 11 j FOREWORD . ... ... .,.. . ... ... . ..... .. ... . ............ iii-

1. INTRODUCTION ..... ..... .. ... .... ......... ................ .. I u

2. REVIEW REQUIREMENTS . .............. . . ....... ............. . 2  !

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3. EVALUATION ........... ...... ................... ................ 4  !

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3.1 Adherence to Regulatory Guide 1.97 . ....................... 4 j 3.2 Type A Variables . . . . . ................................... 4 i 3.3 Exceptions to Regulatory Guide 1.97 ........................ 5 CONCLUSIONS ... ............................................... ..

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5. REFERENCES ...................................... ................ 18 4

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- e CONFORMANCE TO REGULATORY GUIDE.1.97i OYSTER CREEK

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 i

No. I to NUREG-0737, "TMI Action Plan Requirements" (Reference 3).

GpU. Nuclear, the licensee for the Oyster Creek Nuclear Generating l- Station, provided a response to Item 6.2 of the generic letter on '

June 13,1984 (Reference 4). This submittal was superseded by a submittal dated May 9, 1986 (Reference 5).

This report provides an evaluation of the material from the May 9, 1986 submittal.

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2. REVIEW REQUIREMENTS Item 6.2 cf NUREG-0737, Supplement No. 1, sets forth the required documentation to be submitted in a report to the NRC describing how the licensee complies to Regulatory Guice 1.97 as applied to emergency response facilities. The submittal should include documentation that provides the following information 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

5. Redundance and sensor location

6. Power supply

7. Location of display i

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8. Schedule of installation or upgrade l

The submittal should identify any deviations taken from the regulatory l guide recommendations and provide supporting justification-or alternatives l for the deviations identified.

Subsequent to the issuance of the generic letter, the NRC held l regional meetings in February and March 1983, to answer licensee and I applicant questions and concerns regarding the NRC policy on this subject.

At these meetings, it was noted that the NRC review would only address j 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, !

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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 upcated tneir response to Item 6.2 of NRC Generic Letter 82-33 on May 9, 1986. The response describes the licensee's position on post-accident monitoring instrumentation. This evaluation is based on that j material.

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3.1 Adherence to Regulatory Guide 1.97 The licensee has provided a review of their post-accident monitoring instrumentation that compares the instrumentation characteristics against tne recommendations of Regulatory Guide 1.97, Revision 2. The review compares the provided instrumentation to the instrumentation recommended by the regulatory guida, identifies instrumentation that will be modified to-meet the regulatory guide, and gives justification for-instrumentation that the licensee has determir,ed appropriate for Oyster Creek. The licensee has scheduled those modifications to be made duking either refueling outage 11R l or in accordance with the living schedule where an extended shutdown is_ not required to make the modifications. Therefore, we corclude that the

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licensee has provided an explicit commitment on conformance to Regulatory l 1

Guide 1.97. Exceptions to and deviations from the ragulatory guide are noted in Section 3.3.

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

The lite- identifies the following'as Type A variables.

1. Reactor pressure vessel (RPV) pressure

2. RPV level

3. Torus water temperature

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x 4 Torus water level

5. Drywell pressure 6 Drywell hydrogen concentration

7. Drywell oxygen concentration-These variables meet the Category 1 recommendations consistent with the requirements for Type A variables, except as noted in Section 3.3.

3.3 Exceptions to Regulatory Guide 1.97 l

l The licensee identified deviations and exceptions to Regulatory -

Guide 1.97. These are discussed in the followina paragraphs.

3.3.1 Neutron Flux Regulatory Guide 1.97 recommends Category 1 instrumentation for this variable with a range from 10 to 100 percent of full powee. Category 1 recommendations include independent Class 1E power sources. The licensee is upgrading their instrumentation with environmentally and seismically qualified source and intermediate range monitors. The average power range monitors (APRMs) are not scheduled to be upgraded to Category 1. The licensee also notes that the power source for both redundant channels is derived from the same diesel generator backed bus, with provisions to ,

supply power from the other diesel generator.

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The licensee states that the intermediate range monitors will monitor l

power levels up to 20 percent of full power, and that monitoring beyond this power level is not needed because the emergency operating procedures, that deal with anticipated transients without scram (ATWS) events only require action if the power level is greater than 2 percent or cannot be determined. As the licensee has not classified this variable as a Type A variable, we cannot accept this argument as valid. If operator action is dependent on this variable it should be a Type A variable. If thi? is not-5 l l

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a Type ~A variable, then power levels above the span of the intermediate 3

, range monitors may be needed in a post-transient analysis. We conclude that the licensee should upgrade the APRM's to Category 1.

The licensee's power source for this instrumentation is not acceptable. A bus fault could result in the loss of all redundant instrumentation, irregardless of tne provision to transfer the bus to the alternate diesel generator. Thf s is not in conformance with the single failure criteria. The licensee should provide independent power sources for the redundant instrument channels.

i 3.3.2 Coolant Level in Reactor Regulatory Guide 1.97 recommends Category 1 instrumentation for.this variable with a range from the bottom of the core support plate to the ,

centerline of the main steamlines. Category 1- recommendations include )

independent Class 1E power supplies and redundant channels of instrumentation. The licensee has instrumentation that, with 4 overlapping spans, covers a range from -150 to +180 inches (referenced to the top of active fuel). One of these spans is covered by a single channel, rather than by redundant channels of instrumentation. No justification was given for this deviation. Both redundant sets of instrumentation are powered by the same diesel generator backed source, with provisions to transfer to a power source that is backea up by the other diesel generator.

The range limit of -150 inches'is equated to 5 inches below the bottom l

of the fuel. We find this acceptable. The range limit of'+180 inches was I l defined as 5 inches below the isolation condenser steam lines. It_was not related to the centerline of the main steamline or the top of the vessel. i The licensee should provide additional justification for this limit of the range, relating what portions of the range are covered by redundant ,

t instrumentation and justifying that portion of the range covered by a j single channel. ]

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The licensee's power source for'this instrumentation is not I i

acceptable. -A bus fault could result'in the_ loss of all redundant  !

instrumentation, irregardless of the provision to transfer the bus to the alternate diesel generator. This is not in conformance with the single failure criteria. The licensee should provide independent power sources for the redundant instrument channels.

3.3.3 Reactor Coolant System Pressure Regulatory Guide 1.97 recommends redundant- power supplies' for- this instrumentation. The licersee's instrumentation is supplied power by a diesel generator backed power source, with provisions to transfer to a power source that is backed up by the other diesel generator. 3 The licensee's power source for this instrumentation is not acceptable. A bus fault could result in the loss of all redundant ,

instrumentation, regardless of the provision to' transfer the bus to the I 1

alternate diesel generator. This is not in conformance with the single i failure criteria. The licensee should provide independent power sources for the redundant instrument channels.

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{ Drywell Drain Sumps Level Regulatory Guide 1.97 recommends Category 1 instrumentation for these variables. The sumps at Oyster Creek use level switches to alarm in the control room and to initiate sump pump out. Timers indicate the duration of sump pump operation for estimating the amounts of leakage. No s6fety-related system is actuated either automatically or manually as a result of the sump level. The drywell sump systems are automatically isolated at the primary containment penetration should an accident occur.

Drywell temperature and pressure, and primary containment area radiation instrumentation, also can be used to show leakage from the reactor coolant .

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system.

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The licensee has also identified a common recorder for all 4 channels. The licensee states that redundant and separate environmentally qualified channels will be provided. Details were not provided.

The licen%e's power source for this instrumentation is not acceptable. A bus fault could result in the loss of all redundant instrumentation, irregardless of the provision to transfer the bus to the alternate diesel generator. This is not in conformance with the single failure criteria. The licensee should provide independent power sources for the redundant instrument channels.

The licensee should describe the modified instrumentation in sufficient detail to show that the criteria for Category 1 instrumentation are satisfied.

3.3.10 Standby Liqutd Control System Flow The licensee has elected net to implement this variable as recommended by Regulatory Guide 1.97. The justification given by the licensee is that  ;

tne sttndby liquic control system (SLCS) pump discharge pressure provides

[ Mdicat1an that the SLCS pump is operating and that the level indication in the SLCS storage :.ank gives indication that flow is occurring. In J

addition, the fo',10 wing instrumentation supplements the above:

1. Nettran flux

2. Ou.ap e.otor indicating lights

3. Squib velve continui.f indir,ating lights.

We find the above ir.strumentation valid as an alternative lndicition of i

SLCS flew.

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3.3.11 Standby Liauid Control System Storage Tank Level Regulatory Guide 1.97 recommends Category 2 1 instrumentation for this variable. The . licensee has Category 3 instrumentation, and uses it as a key variable in determining that standby liquid control system (SLCS) flow is occurring.

The licensee states that this instrumentation will be operating in a mild environment and that the current design basis for the SLCS recognizes that the system has a classification less than the importance to safety of the reactor protection system and the engineered safeguards systems.

The instrumentation is stated to be in a mild environment, however, the licensee has not verified that the level instrumentation is Category 2. The licensee should verify that this instrumentation is Category 2. .

3.3.12 Cooiing Water Temperature te Engineered Safety Features System Components The licensee has not provided instrumentation for this variable which is defined by the licensee as emergency service water temperature to the containment spray heat exchanger. The licensee states.that proper #

operation of the heat exchanger for the containment spray system is observed by the containment spray tystem flow (which is Category 2 instrumentation) and heat exchanger inlet and outlet temperature (which the licensee has not identified as Category 2 instrumentation).

We find this instrumention acceptable for this varjable; however, the licensee should verify that the temperature instrumentation is Category 2.

Additionally, the licensee should verify that the containment spray system is the only engineered safety feature that utilizes cooling water.

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3.3.13 Hich'Radioac ivity Liquid Tank Level

. Regulatory Guide 1.97 recommends a readout for this instrumentation in the main control room. The licensee's instrumentation does-not have a

. readout there, but is indicated in the radwaste control room. The licensee states that tr.is information can be relayed to the main control room by the use of telephones or radio links. This infers that the radwaste control room is habitable and manned following an accident.

Based on the licensee's , justification, we find that monitoring this i variable in the. main control room of the 0yster Creek Station is not necessary.

3.3.14 Status of Standby Pewer '

1 Regulatory Guide 1.97 recommends Category 2 instrumentation for this vaf'isble. The licensee's instrumentation is Category 3. C. justification was given for this deviation.

We conclude that the lic,ensee should upgrade this instrumentation to Category 2.

3.3.15 Reactor Buildina Area Radiation Regulatory Guide 1.97 recommends Category 2 instrumentation for this variable with a range of 10 to 104 R/hr. The licensee states that )

- this variable need not be implemented. The licensee reports that the use of local radiation exposure rate monitors to detect breach or leakage through primary containment penetrations results in ambiguous indications.

This is due to the radioactivity in the primary containment, the radioactivity in the fluids flowing in emergency core coolant system piping and the amount and the location of fluid and electrical penetrations. The licensee concludes that the use of the vent stack noble gas effluent monitors is the proper way to accomplish the detection of releases, release 13

assessment and the long term surveillance recommended for this variable.

The license has not shown that the vent stack noble gas effluent monitors cover the range recommended for this variable.

We find the licensee's use of the alternative vent stack noble gas effluent monitors, in principle, to be acceptable. However, the licensee should show that the vent stack noble gas effluent monitors cover the equivalent to the recommended 10 3 to 104 R/hr range.

3.3.16 Stack Noble Gas and Vent Flow Rate Regulatory Guide 1.97 recommends Category 2 instrumentation for this -

variable with ranges of 10 -6 to 10#pC1/cc (because of the drywell or standby gas treatment system purge being included) and 0 to 110 percent of design flow.

The licensee has noted the following deviations:

1. The equipment is not environmentally qualified; however, it is i located in a mild environment. Therefore, environmental qualification is not required per 10 CFR 50.49,

2. The licensee has not verified the use of a high reliability power ,

source for this instrumentation. No justification was given.

The licensee should provide a highly reliable power source for this instrumentation.

3. The licensee notes that they do not meet the recommended range limit of 10 pCi/cc. No justification was provided. The range limit of the instrumentation provided was not identified.

Therefore, we conclude that the licensee should either provide l the recommended range or identify the provided range and provide acceptable justification for the deviation.

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3.3.17 Accident Sampling (Reactor Coolant, Containment Air and Sumos)

The licensee's post-accident sampling system can obtain samples and provide the analyses within the ranges recommended for this variable from the reactor coolant and the containment air. The licensee dces not sample the containment, auxiliary bvilding or emergency ccre coolant system (ECCS) sumps as recommended by Regulatory Guide 1.97. The torus and reactor coolant are sampled. The drywell sump systems overflow to the torus.

The licensee deviates from Regulatory Guide 1.97 with respect to post-accident sampling capability. This deviation goes beyond the scope of this review and has been reviewed by the NRC as part of the review of NUREG-0737, Item II.B.3.

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4 CONCLUSIONS Based on our review, we find that the licensee either conferms to or is justified in deviating from Regulatory Guide 1.97, with the following exceptions:

1. Neutron flux--the licensee should upgrade the average power range monitors to Category 1; the licensee should provide independent Class IE power supplies for the redundant channels of instrumentation (Section 3.3.1). i

2. Coolant level in reactor--the_ licensee should provide additional information on tne range and the span that is covered by a single channel; the licensee should provide independent Class IE power supplies for the redundant channel,s of instrumentation (Section 3.3.2),

3. Reactor coolant system pressure--the licensee should provide independent Class IE power supplies for the redundant channels of instrumentation (Section.3.3.3),

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4. Containment effluent radioactivity--the licensee should show that i the range of this instrumentation is adequate for its purpose as listed in the regulatory guide (Section 3.3.6).

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5. Effluent radioactivity--the licensee should verify that the power supply is suitable for Category 2 instrumentation; the licensee .

should show that the range of this instrumentation is adequate for its purpose as Tisted in the regulatory guide (Section 3.3.6).

6. Radiation exposure rate--the licensee should show, by analysis, that the instrument range for a given location encompasses the maximum expected radiation level (Section 3.3.7).

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7. Containment-spray' throttling valve position--the' licensee should

. verify that this instrumentation is Category 2 (Section 3.3.8).

8. Torus water temperature--the licensee should provide independent Class 1E power supplies and show that.the Category 1 criteria are satisfied for this instrumentation (Section 3.3.9).

9. Standby liquid control system storage tank level--the licensee should verify that this instrumentation is Category 2 (Section

'3.3.11).

10. Cooling water temperature to engineered safety features system components--the licensee should verify that the containment spray hea. exchanger inlet and outlet temperature. instrumentation is Category 2, and verify that the containment spray system is the only engineered safety feature that utilizes cooling water (Section 3.3.12). ,,

11. Status of standby power--the licensee should upgrade this instrumentation to Category 2-(Section 3.3.14).

12. Reactor building area radiation--the licensee should provide additional justification for the exception of instrumentation for this variable (Section 3.3.15).

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13. Stack noble gas and vent flow rate--the licensee should provide a highly reliable power source for this instrumentation.

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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. 1 to NUREG-0737--Requirements for Emergency Response Capability (Generic Letter No. 82-33)," December 17, 1982.

2. Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident, Regulatory Guice 1.97, Revision 2, NRC, Of fice of Standards Development, December 1980.

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

4. GPU Nuclear letter, P. B. Fiedler to D. G. Eisenhut, NRC,

" Supplement 1 to NUREG-0737 Regulatory Guide 1.97 Response," June 13, 1984.

5. GPU Nuclear letter, P. B. Fiedler to J. A. Zwolinski, NRC,

" Supplement I tc NUREG-0737 Regulatory Guide 1.97 Response,"

May 9, 1986.

6. Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident, Regulatory Guide 197, Revision 3, NRC, Of fice of Nuclear Regulatory .

Research, May 1983.

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EG&G Idaho, Inc.

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,. m, o...o ....,.1.o .I... ... .....,, ,.c , . .. , v o. .. i, Division of Engineering and System Technology Office of Nuclear Reactor Regulation **"*""*' ""

U. S. Nuclear Regulatory Comission Washington, DC 20555 i

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This EG&G Idaho, Inc. report reviews the submittals for the Oyster Creek Nuclear Gencrating Station 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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