ML19338C689

Technical Evaluation of Electrical,Instrumentation & Control Design Aspects of Tech Spec Change for Main Steamline Break Protection Sys of Beaver Valley Nuclear Power Plant,Unit 1.

ML19338C689
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Site:	Beaver Valley
Issue date:	08/14/1980
From:	EG&G, INC.
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O TECllNICAL EVALUATION OF THE ELECTRICAL, INSTRUMENTATION, AND CONTROL DESIGN ASPECTS OF

'lllE TECILNICAL SPECIFICATION CILtNGE FOR THE MAIN STEAMLINE BREAK PROTECTION SYSTEM 0F Tile BEAVER VALLEY NUCLEAR POWER PLANT, UNIT 1 (Docket No. 50-334)

James H. Cooper EG6G, Inc.-Energy Measurements Group, San Ramon Operations

1.0 INTRODUCTION

In a letter to the U.S. Nuclear Regulatory Commission (NRC) da'.cd 27 October 1978, the Duquesne Light Company requestc4 an amendnent to its operating license No. 35 to incorporate a new steam 1'_nc break protection system design for the Beaver Valley Nuclear Power Plant, Unit 1. The protection system changes provide protection against main steamline breaks and a range of loss-of-coolant accidents (LOCAS).

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' A description and discussion of the proposed change was presented to the NRC by the nuclear steam supply system designer (Westinghouse) and by the Licensee in Washington D.C. on February 23, 197y 'Adgitiona1 #" '"' 3 written information forms part of the data cvaluated. The protection system design has been reviewed and recommended foy approval as reported in the technical evaluation report EG6G 1183-4142.

The purpose of this report is to evaluate the electrical, instrumentation, and control (EI 6 C) design aspects of the proposed technical specification changeusingtgesafetyanalysisofthelicenseamendmentrequest,'IEEE Std.-279yff((*1, criteria and the Code of Federal Regulations, Title 10

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Part 50.

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2.0 DESCRIPTION

OF THE NEW MAIN STEAMLINE BREAK PROTECTION SYSTEM l

2.1 INTRODUCTION

In order to review the instrument changes to which this technical speci-fication change applies, it is first necessary to describe the reactor protection functions that are involved.

2.2 Tile NEW PROTECTION SYSTEM

! The new system is designed to protect the reactor in case of a main steamline break which would result in a sudden and large energy removal i from the secondary loop of the reactor cooling system. The energy loss would, in turn cause a drop in primary coolant temperature and pressure, and, because of the negative coefficient moderator would result in a positive reactivity effect. The licensee states in the safety analysis l 8008180 M d

that for a worst cast stuck rod condition, safety injection is required to unconditionally terminate power operation by the neutron poisoning effect of the boron of the safety injection solution.

2.3 __THE LICENSEE'S SUBMITTAL The licensee's submittal for a license change to incorporate a new main steamline break protection system included a safety analysis by the nucicar

steam supply system designer that demonstrated that the new system meets the required criteria of 10CFR50 and 10CFR100. In the meeting in Nashington D.C.

(Ref.4) the sta*cment was made by the NSSS designer Jiat the new instrument system is as comprehensive for protection as the former system, and that it is expected to be more reliable.

l 2.4 INSTRUMENT SYSTEM

The instrument system for the main steamline break protection system l consists of; the reactor trip system whose initiating signals are unchanged, l

the safety injection system with two additional initiating signals and the deletion of three initiating signals, the steam generator feedwater line isolation system which is unchanged and the main steam isolation stop valve trip system with the two initiating signals replaced by three new initiating signals. A new permissive, P-11 is also added with the change.

3.0 THE TECHNICAL SPECIFICATION CIMNGE EVALUATION .

The initiating signals for the plant parameters that are unchanged are covered by the existing plant technical specification. The new initiating signals developed in the safety analysis must be added to the technical speci-fication by the amendment chaage. The initiating signals added for safety injection are low steamline p; essure in any loop set at 500psig, and high containment pressure at 1.5psig. The channel check, calibration, test and surveillance medes are unchanged from the original.systen requirements. The initiating signals added for steam line isolation are low steamline pressure at 500psig, high negative steam pressure rate at 100psig/sec, and high containment pressure at 5psig. There are three channels per loop with two channels required to trip, and applicable in all three operating modes for all the added steam line isolation signals. The added signal set points are listed in the revised technical specification and the values are the ones used in the safety analysis. The set points and allowable values are in a plausible range to meet the described conditions.

The response times of the added signals are noted in the safety analysis and are added to the revised technical specification under the appropriate reactor safety function. The response times are in the same range as the

• ones replaced.

j The limiting condition for operation in the revised technical specification l primarily involves shutdown margin and (N-1) cooling loop operation which are not in the domain of the report, or will be reviewed for a subsequent application.

4 The new permissive, P-11 is an interlock for the engineered safety fe.*ures system and is set at a pressure of 2010psig for the pressurizer, which cor-responds approximately to full power.

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! 3 4,0 CONCLUSIONS In reviewing the revised tecimical specification it was difficult to follow the requirements with respect to shutdown margin, boration levels required

, and the corresponding operating mode for these levels. Since the (N-1) cooling loop operating mode is not being reviewed for approvial at this time, i references to two loop conditions in the original submittal add to the

, confusion. It is recommended that this aspect of the technical specification l be reviewed by the appropriate branch for consistancy.

The revised technical specification covers the plant variables and initiating

signals required in the safety analysis presented by the licensee. They are

! found to be of approproiate magnitude and redundancy to mitigate the conse-quences of a main steamline break accident, and approval is recommended.

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REFERENCES

1. EGGG 1183-4142: Technical Evaluation of the Eles:trical Instrumentation and Control Design Aspects of the Proposed Abin Steamline Creak Protectica Systen of the Beaver Valley Nucicar Power Plant Unit 1, April, 1930.

2. Duquesne. Light Company (C.M. Dunn) Ic ter to NRC (A. Schwencer),

"Duquesne Docket No. 50-324 License Amend cat Request, Beaver Valley Power Station, Unit No. 1", dated 27 October 1973.

3. Duquesne Light Company License Amendment Ecquest, Technical Specificatica Revision Beaver Valley Power Station Unit No.1, Revised August 24, 1973/

October IS, 1979.

4. Westinghousc/Duquesne Logic Diagrams, presented in Washingten D.C.,

23 Pebruary 1979.

S. Westinghouse WCAP-7672: Solid-State Logic Protectica Systc= Description, June-1971

6. h'estinghouse WCAP-7706: An Evaluation of Solid-State Logic Reacter Protection in Anticipated Tr nsients, July 1971.

7. Nestinghouse NCAP-7319: Nuclear Instru=entation Systen Isolatien A plifier, April 1975.

S. Westinghcuse NCAP-S904: Westin;heuse Ener ency Core Ccoline Systen Evaluation >bdel for Analyzing (N-1) Lcep Operaticn of Plants with Locp Isolation Valves, December 1976.

9. IEEE Std-279-1971: Criteria for Protecticn Systens for Nuclesr Power Cencratint Staticas.

10. Code of Federal Regulations Title 10, Part 50.46: Accettance Criteria for Encreency Core Coolinn Systems for L:cht Nater Nuclear Power l

4 Reactors, January 1, 1975.

11. Code of Federal Regulations Title 10. Part 50. Apnendix A: General Design Criteria for Nucicar Power Plants, January 1, 1975.

12. Duquesne Light Cc pany, Beaver Valicy Final Safety Evaluatica Report (FSAR).

October 1976.

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