ML20215M504

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Conformance to Reg Guide 1.97,Cooper Nuclear Station
ML20215M504
Person / Time
Site: Cooper Entergy icon.png
Issue date: 01/31/1986
From: Udy A
EG&G IDAHO, INC.
To:
NRC
Shared Package
ML20215M500 List:
References
CON-FIN-A-6483, RTR-REGGD-01.097, RTR-REGGD-1.097 EGG-EA-6789, TAC-51082, NUDOCS 8610300289
Download: ML20215M504 (19)
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CONFORMANCE TO REGULATORY GUIDE 1.97 COOPER NUCLEAR STATION A. C. Udy Published January 1986 EG&G Idaho, Inc.

Idaho Falls, Idaho 83415 l

Prepared for the U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Under DOE Contract No. DE-AC07-76ID01570 FIN No. A6483 9 ge e

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ABSTRACT This EG&G Idaho, Inc., report reviews the submittals for Reyision 2 of Regulatory Guide 1.97 for the Cooper Nuclear Station and identifies areas of nonconformance to the regulatory guide. Exceptions to Regulatory Guide i.97 are evaluated and those areas where sufficient basis for acceptabilit is not provided are ioentifieo.

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

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

Docket No. 50-298 TAC No. 51082 4

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e CONTENTS .

ABSTRACT ..............................................................

t it FOREWORD .............................................................. ii l.

INTRODUCTION ..................................................... 1 ,

2. REVIEW REQUIREMENTS .............................................. 2
3. EVALUATION ....................................................... 4 3.1 Adherence to Regulatory Guide 1.97 ......................... 4 3.2 Typ e A V a r i a b l e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 3.3 Exceptions to Regulatory Guide 1.97 ........................ 5 4

CONCLUSIONS ...................................................... 13

5. REFERENCES ....................................................... 14 l 9

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CONFORMANCE TO REGULATORY GUIDE 1.97 -

COOPER NUCLEAR STATION 1

1. ' INTRODUCTION -

On December 17, 1982, Generic Letter No. 82-33 (Reference 1) was issued by D. G. Eisenhut, Director of the Division of Licensing, Nuclear Reactor Regulation, to all licensees of operating reactors, applicants for operating licenses and holders of construction permits. This letter included additional clarification regarding Regulatory Guide 1.97, Revision 2 (Reference 2), relating to the requirements for emergency response capability. Tnese requirements have been published as Supplement No. I to NUREG-0737, "TMI Action Plan Requirements" (Reference 3).

Nebraska Public Power District, the licensee for the Cooper Nuclear Station, provided a response to the generic letter on April 15, 1983 (Reference 4). The response to Section 6.2 of the generic letter was submitted on March 1, 1984 (Reference 5), and revised on April 16, 1984 (Reference 6), on March 6, 1985 (Reference 7) and on May 29, 1985 (Reference 8). Additional information was provided on May 2'4, 1985 (Reference 9) and on December 4, 1985 (Reference 10).

This report provides an evaluation of these submittals.

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2. REVIEW REQUIREMENTS -

r 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 li:ensee complies with Regulatory Gulae 1.97 as applied to emergency response facilities. The submittal should in".lude documentation that provides the following information for each variable shown in the applicable table of Regulatory Guide 1.97.

1. Instrument range
2. Environmer. al qualification
3. Seismic qualification
4. Quality assurance
5. Redundance and sensor location
6. Power supply
7. Location of display
8. Schedule of installation or upgrade.

l l The submittal should identify deviations from the regulatory guide and provide supporting justification or alternatives.

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

At these meetings, it was noted that the NRC review would only address exceptions taken to Regulatory Guide 1.97. Where licensees or applicants explicitly state that instrument systems conform to the regulatory guide it was noted that no further staff review would be necessary. Therefore, this 2

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report only addresses exceptions to Regulatory Guide 1.97. The following "

evaluation is an audit of the licensee's submittals based on the review policy described in the NRC regional meetings.

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3. EVALUATION The licensee p-ovided a response to Item 6.2 of the generic letter on March 1, 1984. This was revised on April 16', 1984 and on May 29, 1985.

Additional information was provided on May 24, 1985 and on -

December 4, 1985. The responses describe the licensee's position on

- -accident monitoring instrumentation. This evaluation is based on that ial.

3.1 Adherence to Regulatory Guide 1.97 t

The licensee has provided a review of their post-accident monitoring instrumentation that coroares the instrumentation characteristics against the recommendations of Regulatory Guide 1.97, Revision 2. Reference 5 provides the licensee's status and schedule for implementation of j Regulatory Guide 1.97 requirements. Reference 6 provides justification for those deviations from the regulatory guide requirements that are identif.ied by the licensee. Reference 10 consolidates all previous responses. A l confirmatory order cated August 29, 1985 (Reference 11) requires the  !

licensee to make those modifications necessary to bring about compliance i with Regulatory Guide 1.97 prior to the startup from cycle 13 (approximately Spring 1989). 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 Regulatory Guide 1.97 does not specifically identify Type A variables, 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 pressure vessel (RPV) level .
2. RPV water pressure 4 k i

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3. Drywell pressure -
4. Suppression pool water level
5. Suppression pool water temperature. ~

This instrumentation meets the Category 1 recocaendations consistent with the requirements for Type A variables.

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

3.3.1 Neutron Flux Regulatory Guide 1.97 recommends Category 1 instrumentation for.this variable with a range of 10-6 to 100 percent of full power. The licensee has supplied instrumentation for this variable with a range of 10-8 to 125 percent of full power, saying it will be implemented as Category 3.

Thus, the category of the instrumentation supplied deviates from what is recommended.

In the process of our review of the neutron flux instrumentation for boiling water reactors, we note that the detectors and thei,r cables have not satisfied the environmental qualification requirement of Regulatory Guide 1.97. A Category 1 system that meets all the criteria of Regulatory Guide 1.97 is an industry development item. Based on our review, we conclude that the existing instrumentation is acceptable for interim operation. The licensee has committed to follow industry development of this equipment, to evaluate newly developed equipment, and to .1 stall Category 1 instrumentation when it becomes available.

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3.3.2 Ccolant 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 steamline. The licensee has instrumentation with -

overlapping ranges that measure the level from -264.to +400 inches; however, from +60 inches to the centerline of the main steamline (123.25 inches) is provided with only one channel of instrumentation.

Additional Category 1 instrumentation will be provided to monitor the coolant level to 6 inches below the bottom of the fuel.

Tne upper water level channel will be Category 1 except for rcdundancy. It uses the reactor vessel head vent as a penetration. In order to comply with the single failure requirement of Regulatory Guide 1.97, an additional vessel penetration would be needed for a redundant reference column for a second upper water level range channel.

The centerline of the main steamlines is used as the' upper end of the Regulatory Guide 1.97 recommended range in order to provide the operator with an indication of whether the reactor coolant has reached, and spilled into, the main steamlines.

The licensee states that all manual and automatic safety functions are initiated in the range covered by the safety-related wide range level instrumentation. The licensee has concluded that the existing reactor coolant level instrumentation meets the intent of the regulatory guide and that only a marginal improvement in plant safety would be achieved by installing a redundant upper water level range channel.

We find that a second upper water level range channel would not result in a significant increase in plant safety. We conclude that the instrumentation being provided by the licensee for this variable is .

acceptable.

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3.3.3 Drywell Sumo Level -

Drywell Drain Sumps Level Regulatory Guide 1.97 recommends Category 1 instrumentation for these variables. The licensee proposes to install Category 3 instrumentation for -

these variables during the 1986 refueling outage. This instrumentation is the same that is used to initiate sump pump operation and alarm an excessive sump fill rate. No safety-related system is actuated either I 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 signal occur.

We conclude that the instrumentation supplied by the licensee will provide appropriate monitoring for the parameters of concern. This is based on (a) for small leaks, the instrumentation is not expected to experience harsh environments during operation, (b) for larger leaks, the sumps fill promptly and the sump drain lines isolate due to the increase in drywell pressure, thus negating the drywell sump level and drywell drain sumps level instrumentation, and (c) this instrumentation neither automatically initiates nor alerts the operator to initiate operation of a safety-related system in a post-accident situation. Therefore, we find the proposed Category 3 instrumentation acceptable.

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

1. Post-accident sampling system
2. Condenser off-gas radiation monitors
3. Main steamline radiation monitors 7
4. Primary containment radiation monitors *
5. Containment hydrogen concentration monitors Based on the alternate instrumentation provided by the licensee, we conclude that the instrumentation supplied for this variable is adequate ,

ar.d, therefore, acceptable.

3.3.-5 Containment and Drywell Hydroaen Concentration Regulatory Guioe 1.97 recommends redundant instrumentation for this variable with a range of 0 to 30 percent. The licensee has two instr,pmentation channels. Ranges are are O to 5 percent for one instrument and 0 to 10 percent and 0 to 20 percent for the other. The power sources are the same.  ;

The licensee states that this variable is implemented as a Category 3 ,

variable. They refer to a response to generic letter 84-09, on hydrogen  !

I recombiner capability. '

The NRC has reviewed the acceptability of this variable as part of their review of NUREG-0737, Item II.F.1.6 and found it acceptable.

3.3.6 Containment and Drywell Oxygen Concentration Regulatory Guide 1.97 recommends Category 1 instrumentation for this variable with a range of 0 to 10 percent. Thus, redundancy is recommended. The licensee presently has one channel of this instrumentation for the inerted containment, with three separate ranges.

In Reference 9, the licensee commits to installing instrumentation that is in full compliance with the regulatory guide during the 1988

') refueling outage. ,

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3.3.7 Radiation Exposure Rate -

Regulatory Guide 1.97, Revision 2, recommends Category 2 instrumentation for this variable with a range from 10-I to 104 R/hr.

The licensee has provided Category 3 radiation exposure rate monitors

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(rather than Category 2) that have ranges that are lower than recommended by Regulatory Guide 1.97 (either 10 ~I to 103 R/hr or 10 -5 to 10-I ,

R/hr). These are stated to be for the detection of significant releases, release assessment and long-term surveillance. The licensee states that when accessibility is re-established to service safety-related equipment, it is done by post-accident sampling and portable instrumentation.

Regulatory Guide 1.97, Revision 3 (Reference 12), changes this variable to Category 3. Therefore, the only deviation of the Cooper Station for this variable is the range supplied. From a radiological standpoint, if the radiation levels reach or exceed the upper limit of the range, personnel would not be permitted into the areas without portab.le mcnitoring (except for life saving). Based on the alternate portable monitoring instrumentation used by the licensee with this variable, we find the proposed ranges for the radiation exposure rate monitors acceptable.

3.3.8 Suppression Chamber Spray Flow Drywell Spray Flow Regulatory Guide 1.97 recommends instrumentation for these variables with a range of 0 to 110 percent of design flow. Sections 4.3.5 and 8.5 of the FSAR show that the existence of spray flow to the suppression pool and drywell can be established. This is done by use of the residual heat removal system flow in conjunction with valve lineup. These parameters are indicated in the control room. The licensee indicates that the

., effectiveness of these flows is indicated by pressure and temperature changes in the drywell and the suppression chamber. We find that this

. alternate method of monitoring this variable is acceptable.

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3.3.9 Drywell Atmosobere Temperature -

Revision 3 of Regulatory Guide 1.97 (Reference 12) recommends instrumentation for this variable with a range from 40 to 440*F. The _

licensee has provideo instrumentation for this variable with a range of 50

  • to 350*F. The maximum analyzed drywell temperature resulting from an accident is 340'. Therefore, the upper limit of 350* is satisfactory. The lower limit deviates from the recommended end of span by 10*F or 3 percent of the instrument span. This is insignificant considering instrument

, accuracy and that post-accident temperatures are elevated above normal.

The licensee states that no accidents result in drywell atmosphere temperatures between 40 and 50*F. Therefore, the range provided for this variable is acceptable.

3.3.10 Standby Liquid Control System (SLCS) Flow The licensee has elected not to implement this variable as recommen'ded ,

in Regulatory Guide 1.97. The justification provided by the licensee is ,

that the SLCS storage tank level and the positive displacement pump discharge header pressure indicate that flow is occurring. Other indications of system operation include reactivity change (as monitored by the neutron flux instrumentation), pump motor indicating lights and squib valve continuity lights.

We find the above instrumentation valid as an alternative indication of SLCS flow.

3.3.11 Standby Liouid Control System Storage Tank Level Regulatory Guide 1.97 recommends Category 2 instrumentation for this variable. The licensee has instrumentation that, except for environmental

  • qualification, is Category 2.

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The licensee justifies using this instrumentation based on the mild

3.3.12 Reactor Building or Secondary Containment Area Radiation ' -

Regulatory Guide 1.97 recomends Category 2 instrumentation for this variable with a range of 10- to 10 #R/hr for the Mark I containment.

The licensee has some instruments with a range of 10 -5 to 10- R/hr,

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and a channel with a range of 10 to 10 R/hr. All these instruments are Category 3 rather than the recommended Category 2.

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 the emergency core coolant system piping and the amount and location of fluid and electrical penetrations. The licensee concludes that the use of the plant noble gas effluent monitors is the proper way to accomplish the purpose of this variable. Therefore, the licensee concludes that the existing Category 3 instrumentation for this variable is adequate.

We conclude that the existing Category 3 instrumentation and ranges will monitor the ranges of concern and are therefore acceptable.

3.3.13 Plant and Environs Radiation Regulatory Guide 1.97 recomends instrumentation for this variable

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with ranges of 10 to 10 R/hr-photons and 10 to 10 rads /hr-beta and low energy photons. The licensee has instrumentation for thi.s variable with ranges of 10 -3 to 103R/hr and 10 -3 to 200 rads /hr. The licensee states that these ran.ms are adequate, based on the personnel radiation exposure that would be required to operate these instruments in fields that are higher than the provided range.' Based on this rational, we find the provided instrumentation acceptable.

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3.3.14 Accident Sampling (Primary Coolant, Containment Air and Sumo) -

The licensee's post-accident sampling system provides sampling and analysis as recommended by the regulatory guide, except for the following _

deviations. .

1. Chloride content--the minimum observable concentration is 10 ppb; the maximum range is 10 ppm rather than 20 ppm, but the sample is dilutable.
2. Oxygen content--the minimum observable concentration is 10 ppb; the maximum range is 1 ppm rather than 20 ppm, but the sample is dilutable.
3. Dissolved hydrogen or total gas--the licensee calculates this, however, they do not state how it is done, nor do they equate it to the recommended analysis range of 0 to 2000 cc (STP)/kg.
4. The licensee does not sample the sumps recommended by the cegulatory guide.
5. The licensee has not provided information (required by Section 6.2 of Reference 3) to show compliance to the regulatory guide for hydrogen and oxygen content of the containment air.

The licensee deviates from Regulatory Guide 1.97 with respect to post-accident sampling capability. These deviations go beyond the scope of this review and are being addressed by the NRC as part of their review of NUREG-0737, Item II.B.3.

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4. CONCLUSIONS 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

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exception:

l. Neutron flux--the licensee's present instrumentation is acceptable on an interim basis until Category 1 instrumentation is developed and installed (Section 3.3.1).

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5. REFERENCES .
1. NRC letter, D. G. Eisenhut to All Licensees of Operating Reactors, Applicants for Operating Licenses, and Holders of Construction Permits, " Supplement No.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 Followinc an Accident, ~

Regulatory Guide 1.97, Revision 2, U.S. Nuclear Regulatory Commission (NRC), Office of Standards Development, December 1980.

3. Clarification of TMI Action Plan Reouirements, Reouirements for Emeroency Resoonse cacability, NUREG-0737, Supplement No. I, NRC, Office of Nuclear Reactor Regulation, January 1983.

4 Nebraska Public Power District letter, J. M. Pilant to D. G. Eisenhut, NRC, " Response to NUREG-0737, Supplement 1," April 15, 1983, LOA 8300129.

5. Nebraska Public Power District letter, J. M. Pilant to D. G. Eisenhut, NRC, "NUREG-0737, Supplement 1 - Regulatory Guide 1.97,"

Marc 6 1, 1984, NLS8400073.

6. debraska Public Power District letter, J. M. Pilant to D. G. Eisenhut, NRC, "NUREG-0737, Supplement 1 - Regulatory Guide 1.97,"

April 16, 1984.

7. Nebraska Public Power District letter, J. M. Pilant to H. L. Thompson, Jr., NRC, "NUREG-0737, Supplement 1 - Regulatory Guide 1.97 Response, Revision V," March 6,1985.
8. Nebraska Public Power District letter, J. M. Pilant to H. L. Thompson NRC, "NUREG-0737, Supplement 1, Regulatory Guide 1.97 Response, Revision VI," May 29, 1985.
9. Nebraska Public Power District letter, J. M. Pilant to Director of Nuclear Reactor Regulation, NRC, "NUREG-0737, Supplement 1 -

Regulatory Guide 1.97," May 24, 1985.

10. Nebraska Public Power District letter, J. M. Pilant to R. M. Bernero, NRC, "NUREG-0737, Supplement No. 1 - Regulatory Guide 1.97 Response, Revision VII," December 4, 1985.
11. NRC letter, D. Vassallo to J. M. Pilant, Nebraska Public Power District, August 29, 1985.
12. Instrumente?. ion for Licht-Water-Cooled Nuclear Power Plants to Assess Plant Env1-ons conditions During and Following an Accioent, Regulatory .

Guide 1.97, Revision i, NRC, Office of Nuclear Regulatory Research, May 1983.

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,, iv a.. r... ~o r n 13 AGST .cr J.10 e. ass .* 'esas This EG&G Idaho, Inc., repo-t provides a review of the submittals for Regulatory Guide 1.97, Revision 2, for the Cooper Nuclear Station.

Exceptions to the guidelines of Regulatory Guide 1.97 are evaluated.

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