ML20236T387

SER Re Conformance to Reg Guide 1.97 Concerning post-accident Monitoring Instrumentation.Design Acceptable

ML20236T387
Person / Time
Site:	Davis Besse
Issue date:	11/25/1987
From:	Office of Nuclear Reactor Regulation
To:
Shared Package
ML20236T382	List: ML20236T381 ML20236T387
References
RTR-REGGD-01.097, RTR-REGGD-1.097 GL-82-33, TAC-51084, NUDOCS 8712010200
Download: ML20236T387 (5)

Text

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j ENCLOSURE 1 i

SAFETY EVALUATION REPORT . .

DAVIS-BESSE NUCLEAR POWER STATION DOCKET NO. 50-346 1' CONFORMANCE TO REGULATORY GUIDE 1.97

1.0 INTRODUCTION

AND

SUMMARY

Toledo Edison Company,-the Davis-Besse Nuclear Power Station licensee,- a was requested by: Generic Letter 82-33 to provide a report to the NRC: I describing how the post-accident monitoring instrumentation meets the .

guidelines of Regulatory Guide (RG) 1.97 as applied lto emergency response.-

facilities. The licensee's responso to.RG 1.97 was provided by letter dated June 28,'1984.

Detailed reviews and technical evaluations of.the~11censee's submittals-were performed by EG&G Idaho. Inc., under contract _to the NRC, with- l general supervision by the NRC staff. The initial' evaluation of this submittal was~ transmitted to the licensee.by letter dated.0ctober 16, 1985._This evaluation identified seventeen exceptions to Y 1.97 recommendations which were not acceptable. The licensee's supplemental information submitted January 27, 1986, and September 26, 1986, was reviewed, evaluated, and reported by EG&G in their-Technical Evaluation Report (TER), "Conformance to Regulatory Guide:1.97, Davis-Besse Nuclear.

Generating Station," dated November 1986 (enclosed). The staff has reviewed this report and concurs with the conclusion-that the licensee  ;

either conforms to or has-justified deviations taken from the guidance i of Regulatory Guide 1.97 for each post-accident monitoring variable; however, four excepted variables were stated.to be beyond the scope-of the- ,

RG 1.97 review and were being reviewed as part of the resolution of the' NUREG-0737. Item II.B. 3 issue. .j 2.0 EVALUATION CRITERIA. j Subsequent to the issuance of the. generic letter 82-33, the NRC held j regional meetings in February and March 1983 to answer the licensee's questions and concerns regarding NRC policy on RG 1.97. At these  ;

meetings, it was established that NRC review would address only exceptions taken to the guidance of RG 1.97.. Further, where:the licensee 1 explicitly stated that instrument. systems confonn to the provisions' of . {

the regulatory guide, no staff review would be necessary. Therefore 'the review performed and reported by EG&G only. addresses exceptions to the  !

guidance of the regulatory guide.- This safety evaluation addresses the i licensee's submittals based on the review policy described in the NRC~.

regional meetings and the conclusions of the review as reported by EG8G.

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b 3.0 EVALUATION In the Safety Evaluation issued October 16, 1985, the staff identified l seventeen exceptions to the recommendations of RG 1.97 which were found to be unacceptable. The licensee responded with supplemental information which was reviewed and evaluated by the staff's consultant, EGAG. The results of this review were reported in their Technical Evaluation Report ,

(TER), EA-6941, which found that the licensee had either conformed to the 1 RG 1.97 recommendations or had provided acceptable justification for i deviating from those recommendations for each post-accident monitoring variable. The resolution of four of these excepted variables, the RCS soluble boron concentration, radiation level in circulating primary coolant, analysis of primary coolant, and accident sampling (primary coolant, containment air and sump) were stated to be beyond the scope of  !

the RG 1.97 review and were being reviewed as part of the resolution of  !

the NUREG-0737, Item II.B.3 issue. The status of the resolution of these  !

items is reported below:

For the RCS soluble boron concentration variable, the licensee stated that on-line monitoring equipment was not provided but monitoring of this variable was accomplished by sampling with the post-accident sampling system (PASS) and analyzing the samples using onsite laboratory analysis.

In Safety Evaluation issued February 10, 1987, the staff found this methodology to be acceptable for meeting the requirements of NUREG 0737,  !

Item II.B.3 and the recommendations of RG 1.97, Revision 2. Revision 3 did not change these recommendations, therefore, the recommendations of Revision 3 of RG 1.97 are also met and this response is acceptable.

For the measurement of radiation level in circulating primary coolant, one of the identified means of measurement was the PASS which was stated to be under review by the staff as part of the NUREG-0737, Item II.B.3 itwe. In the staff's Safety Evaluation February 10, 1987, the staff fod d that the PASS met both the Item II.B.3 requirements and the-RG 1.97, Revision 2 recommendations and was, therefore, acceptable. The licensee states that isotopic analysis is employed instead of gross gamma counting and that their equipment and procedures enable the individual radionuclides to be identified and quantified over the range recommended

, in Revision 3 of RG 1.97. The staff finds t' 'icceptable.

For analysis of primary coolant, the licen we states that this l

information can be obtained from the PASS isotopic analysis performed to monitor the radiation level in the circulating primary coolant and that another analysis of the primary coolant is not needed. In the Safety Evaluation issued February 10, 1987, the staff found the PASS and its isotopic analysis met the requirements of NUREG-0737 Item II.B.3 and the recommendations of RG 1.97, Revision 2 and was acceptable. The staff finds that Revision 3 did not change these recommendations, therefore, the recommendations of Revision 3 of RG 1.97 are also met for this variable and this response is acceptable.

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I 9 j For the accident sampling (primary coolant, containment' air and sump)'  !

variable, the licensee states that the PASS provides the sampling and analysis capability recommended by RG 1.97 with the exception that-the .,

'ECCS pump room sump is not sampled, containment air'is not analyzed for '

.i' oxygen content, and sampling of containment. air for hydrogen and contain- ,

ment air activity measurement is performed by.'on-line. systems.:

The licensee states that'the containment' sump liquid can be sampled on ai ')"

long term post-accident basis using the PASS and that sampling the.ECCS-pump room sump would yield no additional infomation from that obtained from sampling the containment sump;.therefore,Lthe:ECCS pump room sump is-not-sampled. The' staff finds this deviation acceptable on'the basis.that

no new additional infomation would be obtained.'

The licensee states that the oxygen concentration of.the~ containment. air-samples is not measured since the containment.is not purged and inerted,.

therefore the measurement of:its oxygen concentration serves noLuseful purpose. The staff finds this deviation acceptable on the basis that' there is no demonstrated need-for the infomation.

The staff found measurement of containment' air hydrogen concentration:

l by online equipment acceptable because'it was backed up by capability for taking and analyzing grab samples. .This met the requirements for NUREG-0737. Item II.B.3 and will.also meet the recommendations of RG 1.97..

The licensee stated that containment air activity is monitored by'an on-line system. This statement was clarified in a telecon with the licensee wherein the NRC was infomed that containment. air monitoring'is  ;

performed continuously by the Class-1E. containment high range. radiation; I

monitor which measures iodine and particulate activity. In addition, the.-

, PASS is relied on to take'and analyze grab samples of containment air.

These samples are analyzed on-site for isotopic activity using equipment which meets RG 1.97 recommendations. The staff finds this acceptable based on the clarifying information supplied by the licensee in the 3 telecon and the licensee's commitment to provide written confirmation of .]

this clarifying information.

4.0 CONCLUSION

Based on the staff's review of the enclosed TER 'the licensee's l submittals, a telecon with the licensee, and the comitment to provide -

written confirmation of clarifying information provided in the telecon,- 1 the staff finds that the Davis-Besse design is acceptable with respect to 1

l. confomance to RG 1.97, Revision 3.  :

An appropriate implementation schedule will be developed by the project -;

manager through discussion'with the licensee. Once the schedule is  !

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established, the licensee is required to inform the. Commission, in writing, of any significant changes to the implemented system from that approved in the staff's safety evaluation and when the implementation has-actually,been completed.

' Dated: November 25, 1987

November 25, 1987

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Items 1 and 2'are to be completed during the.fifth refueling outage and I

. items 3 and 4 are to be completed by the sixth-refueling outage. You are

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requested to confirm the outstanding modifications and' implementation.

schedule, in writing, no later than December 31, 1987..

U$8IO8l$lgned by

' Albert W. De Agazio, Project Manager-  :

. Project Directorate III-1 Division of Reactor Projects -.III, IV, V & Special. I'rojects

Enclosure:

Safety Evaluation cc w/ enclosure: d See next page l l

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