ML20136G269

Conformance to Reg Guide 1.97,Shoreham Nuclear Power Station,Unit 1

ML20136G269
Person / Time
Site:	Shoreham File:Long Island Lighting Company icon.png
Issue date:	03/31/1985
From:	Vanderbeek R EG&G IDAHO, INC.
To:	NRC
Shared Package
ML20136G273	List: ML20136G269 ML20137G564
References
CON-FIN-A-6493, RTR-REGGD-01.097, RTR-REGGD-1.097 NUDOCS 8508050237
Download: ML20136G269 (16)

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CONFORMANCE TO REGULATORY GUIDE 1.97 SHOREHAM NUCLEAR POWER STATION, UNIT NO. 1 l R. VanderBeek l

Published March 1985 EG&G Idaho, Inc.

Idaho Falls Idaho 83415 Prepared for the U.S. Nuclear Regulatory Comission Washington, D.C. 20555 Under DOE Contract No. DE-AC07-76ID01570 FIN No. A6493 h0M d'b ,

ABSTRACT This EG&G Idaho, Inc., report reviews the submittal for Regulatory Guide 1.97, Revision 2, for the Shoreham Nuclear Power Station, Unit No. 1. Any exception 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.

Docket No 50-322 11 6--.__

TABLE OF CONTENTS ABSTRACT ............................................................. 11 FOREWORD ............................................................. 11

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 ..................................................... 12

5. REFERENCES ...................................................... 13 S

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CONFORMANCE TO REGULATORY GUIDE 1.97 SHOREHAM NUCLEAR POWER STATION, UNIT NO. 1 -

1. INTRODUCTION On December 17, 1982, Generic Letter No. 82-33 (Referer.ce 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. I to NUREG-0737, "TMI Action Plan Requirements" (Reference 3).

The Long Island Lighting Company, the applicant for the Shoreham Nuclear Power Station, provided a response to the generic letter on April 14, 1983 (Reference 4).

This report provides an evaluation of this submittal.

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2. REVIEW REQUIREMENTS 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 applicant complies to Regulatory Guide 1.97 as applied to emergency response facilities. The submittal should include documentation that provides the following information l for each variable shown in the applicable table of Regulatory Guide 1.97.

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1. Instrument range

2. Environmental 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.

Furthermore, 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 questions and concerns regarding.the NRC policy on this subject. At these meetings, it was noted that the NRC review would only address exceptions taken to Regulatory Guide 1.97. Furthermore, where licensees or applicants explicitly state that instrument systems conform to the provisions of the guide it was noted that no further staff review would be necessary.

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

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The following evaluation is an audit of the applicant's submittal based on the review policy described in the NRC regional meetings. -

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3. EVALUATION l The applicant provided a response to NRC Generic Letter 82-33 on April 14, 1983. This evaluation is based on that submittal.

3.1 Adherence to Regulatory Guide 1.97 The applicant stated in Appendix C of his submittal that Shoreham is in conformance with Regulatory Guide 1.97 to the extent discussed in the:

submittal. Within Table I of Appendix C, the applicant has listed the ..

Regulatory Guide 1.97 variable and its status. Therefore, we concluje that the applicant has provided an explicit commitment to conform to Regulatory Guide 1.97, except for those exceptions that were justified as 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 information required to permit the control room operator to take specific manually controlled safety actions. The applicant classifies the following instrumentation as Type A.

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1. Reactor pressure vessel water level

2. Reactor pressure vessel pressure

3. Drywell pressure

4. Sump level

5. Suppression pool water level

6. Drywell and suppression chamber oxygen concentration

7. Suppression pool water temperature.

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

3.3 Exceptions to Regulatory Guide 1.97 m The applicant identified deviations and exeptions from Regulatory Guide 1.97. These are discussed in the following paragraphs. A 3.3.1 Neutron Flux Exception has been taken by the applicant to the recommendation of Regulatory Guide 1.97 for the neutron flux variable. The applicant has .

(a) specified a range of 10-1 to 106 Ci instead of the recomended 10-6 to 100 percent full power, (b) specified that the Source Range Monitor (SRM) instrumentation does not comply with any of the recommendations of Regulatory Guide 1.97, and (c) specified a non-IE power source instead of the IE power ( $

source recommended by Regulatory Guide 1.97. The applicant states that the (

installed equipment is satisfactory for the interim period and system ~

modifications will be completed after the second refueling outage assuming timely resolution of generic issues to Regulatory Guide 1.97. _

The applicant has proposed some upgrading of the neutron flux instrumentation, however, they have not identified the extent or end results of their changes.

In the process of our review of the neutron flux instrumentation for boiling water reactors (BWRs), we note that the mechanical drives of the detectors have not satisfied the environmental qualification requirement of Regulatory Guide 1.97. A Category 1 system that meets all the cr.iteria of F[

Regulatory Guide 1.97 is an industry development item. Based on our review, [

we conclude that the existing instrumentation is acceptable for interim M operation. The licensee should' follow industry development of this equipment. [

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

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3.3.2 Reactor Water Level Regulatory Guide 1.97 specifies that the level range should be from the bottom of the core support plate to the lesser of the top of the vessel or the centerline of the main steamline. The applicant provides a fuel zone range of

-308 to -108 inches and a wide range of -150 to +60 inches. This range is more than seven feet lower than the recommended upper level of indication and, therefore, does not meet the recommendations of Regulatory Guide 1.97. The applicant states that further action to this system is pending the resolution of the Inadequate Core Cooling (ICC) detection issue. ._

The applicant takes exception to the guida'ce of .iegulatory Guide 1.97 with respect to the range of this instrumentation. This exception goes beyond the scope of this review and is being addressed by the NRC as part of their review of NUREG-0737, Item II.F.2 (Inadequate Core Cooling). The acceptance criteria for Item II.F.2 is the same as Category 1 for Regulatory Guide 1.97.

3.3.3 RCS Soluble Boron Concentration Exception has been taken by the applicant to the recommendation of Regulatory Guide 1.97 for the RCS Soluble Boron Concentration variable. The applicant specifies a range of 100 to 1100 ppe instead of the recommended range of 0 to 1000 ppm per Regulatory Guide 1.97. The applicant has not provided any justification for deviation.

The applicant takes exception to the guidance of Regulatory Guide 1.97 with respect to post-accident sampling capability. This exception goes beyond the scope of this review and is being addressed by the NRC as part of their review of NUREG-0737. Item II.B.3.

3.3.4 Drywell Sump Level and Drywell Drain Sump Level The applicant states that compliance to the recommended Type B, Category 1, and Type C, Category 1, requirements is not applicable for these 6

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variables. Theapplic2ntstatesthattheinstalledCategory3equipmentis acceptable as is. The supporting justification is that (a) the drywell -

pressure and temperature along with the primary containment area radiation can be used to provide indication of leakage in the drywell, (b) these variables are qualified to Category 1 or 2, and (c) the drywell sump systems are isolated for accident conditions.

We conclude that the instrumentation supplied by the applicant will provide appropriate monitoring for the parameters of concern. Based on (a) for small leaks, the instrumentation is not expectad 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, (c) the drywell pressure and temperature as well as the primary containment area radiation instrumentation can be used to detect leakage in the drywell, and (d) this instrumentation neither automatically initiates nor alerts the operator to initiate operation of a safety-related system in a post-accident situation, we find the Category 3 instrumentation provided acceptable.

3.3.5 Primary Containment Isolation Valve Position The applicant states that the recommendation of Regulatory Guide 1.97 for the primary containment isolation valve position is not applicable. The recommendation of the regulatory guide is a closed-not closed range. The applicant does not provide justification for not providing this valve position indication.

The applicant should either provide specific information explaining why the range is not applicable or provide the specified range.

3.3.6 Radiation Level in Circulating Primary Coolant The applicant states that alternate instrumentation is available for this variable and that the critical actions to be taken to prevent and mitigate a 7

gross breach of fuel cladding are (a) shut down the reactor, and (b) maintain the water level. Neither of these actions are influenced by the this -

variable.

Theapplicantindicatesthatthepost-accidentsamplingsyster.(PASS) provides a means of obtaining samples of the reactor coolant and that the primary containment atmosphere pressure and the radiation monitors in the steam jet air ejector and the main steamlines provide information on the status of fuel cladding when the plant is not isolated.

Based on the alternate instrumentation provided by the applicant, we 4 conclude that the instrumentation supplied for this variable is adequate, and therefore, acceptable.

3.3.7 Radiation Exposure Rate The applicant has elected not to implement this type C variable as recomended by Revision 2 of Regulatory Guide 1.97, the justification being .

that other means, such as noble gas monitoring, are better suited for breach detection.

Revision 3 of Regulatory Guide 1.97 (Reference 5) deletes this Type C variable from the recomended instrumentation. Therefore, the lack of the Type C instrumentation is acceptable. This variable is still recommended as a Type E variable. Refer to Section 3.3.13.

3.3.8 Suppression Chamber Spray Flow Drywell Spray Flow

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The residual heat removal (RHR) system flow is used for these variables. Both sprays derive their flow from the RHR system, with a throttling valve proportioning the flow between the two sprays. The position of the throttling valve is controlled from the control room. Pressure and temperature changes in the d*ywell and suppression chamber determine the effectiveness of the spray.

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l The applicant concludes that the RHR flow, the suppression chamber and drywell temperature and pressure, accurately and re11abili+.y measure the -

effectiveness of the drywell and suppression chamber sprays. Additionally, the position of the throttling valve is known in the control room. We find that this instrumentation is adequate for this variable.

3.3.9 Core Spray System Flow The applicant has not submitted the range of the core spray system flow. ..

The applicant should provide this information as required by Section 6.2 of NUREG-0737, Supplement No. 1.

3.3.10 RHR Heat Exchanger Outlet Temperature The applicant has provided a range of 40 to 400*F for this variable.

Revision 2 of Regulatory Guide 1.97 recommends 32 to 350*F.

Revision 3 of Regulatory Guide 1.97 changes the recomended range to 40 to 350*F. Therefore, the range provided by the applicant is acceptable.

3.3.11 Cooling Water Temperature to ESF System Components The applicant has provided a range of 40 to 320*F for this variable.

Revision 2 of Regulatory Guide 1.97 recommends 32 of 200*F.

Revision 3 of Regulatory Guide 1.97 changes the recommended range to 40 to 200*F. Therefore, the range provided by the applicant is acceptable.

3.3.12 Secondary Containment Area Radiation Regulatory Guide 1.97 recommends instrumentation for this variable. The applicant's position is that the secondary containment area radiation is not

an appropriate parameter to use to detect or assess primary containment leakage. Therefore, the applicant states that the reactor enclosure area -

radiation monitors are not required.

The applicant has not shown that this variable will be adequately monitored by alternate instrumentation. The applicant should provide the recomended instrumentation.

3.3.13 Radiation Exposure Rate Regulatory Guide 1.97 recommends that this Type E variable be monitored, whereas the applicant's position is that (a) the Shoreham design does not require access to any harsh environment area to service safety-related equipment during an accident and (b) portable radiation monitors will be provided to establish accessibility.

Access to equipment areas could be required after an accident even if the areas are not designed for equipment service during an accident. The applicant has not provided enough justification for not having this instrumentation. The licensee should provide the recommended instrumentation for this variable.

3.3.14 Accident Sampling Capability (Primary Coolant, Containment Air and Sump The applicant takes exception to the recommendations of Regulatory Guide 1.97 for this variable. For the gross activity range, the applicant specifies 8 decades, 0.1 mR/hr to 1 x 107 mR/hr, instead of the 10 pC1/ml to 10 C1/ml recommended;.and for the baron content, the applicant specifies a range of 100 to 1100 ppm instead of the recomended range of 0 to 1000 ppm.

The applicant provided no justification for theie deviations.

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l The licensee takes exception to the guidance of Regulatory Guide 1.97 with respect to post-accident sau:pling capability. This exception goes beyond -

the scope of 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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4. CONCLUSIONS Based on our review, we find that the applicant conforms to or is j justified in deviating from Regulatory Guide 1.97, with the following l exceptions.

1. Neutron flux--the applicant's present instrumentation is acceptable on an interim basis until Category 1 instrumentation is developed l

and installed (Section 3.3.1).

2. Primary containment isolation valve position--the applicant should either provide justification for not providing this indication or provide the range specified (Section 3.3.5).

3. Core spray system flow--the applicant should specify the range as required by Section 6.2 of NUREG-0737, Supplement No. 1 (Section 3.3.9).

4. Secondary containment area radiation--the applicant should install a satisfactory system for this variable (Section 3.3.12).

5. Radiation exposure rate--the applicant should install the recommendedinstrumentation(Section3.3.13).

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