ML20212P392

Safety Evaluation Report Related to the Operation of Beaver Valley Power Station,Unit 2.Docket No. 50-412.(Duquesne Light Company,Et Al)

ML20212P392
Person / Time
Site:	Beaver Valley
Issue date:	08/31/1986
From:	Office of Nuclear Reactor Regulation
To:
References
NUREG-1057, NUREG-1057-S02, NUREG-1057-S2, TAC-62881, NUDOCS 8609030113
Download: ML20212P392 (43)

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NUREG-1057 Supplement No. 2 Safety Evaluation Report related to the operation of Beaver Valley Power Station, Unit 2 Docket No. 50-412 Duquesne Light Company, et al.

1 U.S. Nuclear Regulatory

' Commission Offics of Nuclear Reactor Regulation August 1986 v" " %,

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NOTICE Availability of Reference Materials Cited in NRC Publications 4

Most documents cited in NRC publications will be available from one of the following sources:

1. The NRC Public Document Room,1717 H Street, N.W.

Washington, DC 20555

2. The Superintendent of Documents, U.S. Government Printing Office, Post Office Box 37082, I"

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3. The National Technical Information Service, Springfield, VA 22161 Although the listing that follows represents the majority of documents cited in NRC publications, it is not intended to be exhaustive.

Referenced documents available for inspection and copying for a fee from the NRC Public Docu-ment Room include NRC correspondence and internal NRC memoranda; NRC Office of Inspection and Enforcement bulletins, circulars, information notices, inspection and investigation notices; Licensee Event Reports; vendor reports and correspondence; Commission papers; and applicant and h

licensee documents and correspondence.

f The following documents in the NUREG series are availab'e for purchase from the GPO Sales p i Program: formal NRC staff and contractor reports, NRC-sponsored conference proceedings, and

(

N RC booklets and brochures. Also available are Regulatory Guides, NRC regulations in the Code of Federal Regulations, and Nuclear Regulatory Commission Issuances.

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reports and technical reports prepared by other federal agencies and reports prepared by the Atomic l

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NUREG-1057 Supplement No. 2 Safety Evaluation Report related to the operation of Beaver Valley Power Station, Unit 2 Docket No. 50-412 Duquesne Light Company, et al.

U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation August 1986

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1

t ABSTRACT This report, Supplement No. 2 to the Safety Evaluation Report for the applica-tion filed b.v the Duquesne Light Company et al. (the applicant) for a license to operate the Beaver Valley Power Station, Unit 2 (Docket No. 50-412), has been prepared by the Office of Nuclear Reactor Regulation of the U.S. Nuclear Regulatory C:1mmission.

This supplement reports the status of certain items that had not been resolved at the time the Safety Evaluation Report was published.

Beaver Valley 2 SSER 2 iii

TABLE OF CONTENTS Pag _e ABSTRACT.............................................................

iii 1

INTRODUCTION AND GENERAL DISCUSSION.............................

1-1 1.1 Introduction...............................................

1-1 3

DESIGN OF STRUCTURES, COMPONENTS, EQUIPMENT, AND SYSTEMS........

3-1 3.9 Mechanical Systems and Components..........................

3-1 3.9.3 ASME Code Class 1, 2, and 3 Components, Component Supports, and Core Support Structures................

3-1 4

REACTOR.........................................................

4-1 4.4 The rmal liyd ra ul i c De s i g n...................................

4-1 4.4.7 Inadequate Core Cooling Instrumentation.............

4-1 6

ENGINEERED SAFETY FEATURES......................................

6-1 6.2 Containment Systems........................................

6-1 6.2.2 Containment Heat Removal Systems.....................

6-1 6.2.7 Fracture Prevention of Containment Pressure Boundary............................................

6-1 7

INSTRUMENTATION AND CONTROLS....................................

7-1 7.1 Introduction...............................................

7-1 7.1.4 Specific Findings...................................

7-1 7.3 Engineered Safety Feature Systems..........................

7-3 7.3.3 Specific Findings...................................

7-3 9

AUXILIARY SYSTEMS...............................................

9-1 9.5 Other Auxiliary Systems....................................

9-1 9.5.1 Fire Protection Program.............................

9-1 9.5.7 Emergency Diesel Engine Lubricating Oil System......

9-1 Beaver Valley 2 SSER 2 v

TABLE OF CONTENTS (Continued)

Page 15-1 l'5 ACCIDENT ANALYSIS...............................................

15.4 Reactivity and Power Distribution Anomalies...............

15-1 15.4.2 Uncontrolled Rod Cluster Control Assembly (Rod)

Bank Withdrawal at Power...........................

15-1 15-1 15.8 Anticipated Transients Without Scram......................

17-1 17 QUALITY ASSURANCE...............................................

17-1 17.1 General...................................................

17-1 17.2 Organization.........

APPENDICES A

CONTINUATION OF CHRON0 LOGY OF NRC STAFF RADIOLOGICAL REVIEW OF BEAVER VALLEY POWER STATION, UNIT 2 B

BIBLIOGRAPHY D

ACRONYMS AND INITIALISMS E

NRC STAFF CONTRIBUTORS AND CONSULTANTS L

STAFF SAFETY EVALUATION REPORT ON CONFORMANCE TO GENERIC LETTER 83-28, ITEM 2.1 (PART 1), EQUIPMENT CLASSIFICATION (REACTOR TRIP SYSTEM COMPONENTS) AND VENDOR INTERFACE M

STAFF SAFETY EVALUATION REPORT ON CONFORMANCE TO GENERIC LETTER 83-28, ITEMS 3.1.3 (REACTOR TRIP SYSTEM COMPONENTS) AND 3.2.3 (ALL OTHER SAFETY-RELATED COMPONENTS), POSTMAINTENANCE TESTING REQUIREMENTS IN TECHNICAL SPECIFICATION THAT COULD DEGRADE SAFETY LIST OF TABLES Page 1-2 1.2 Oper issues.....................................................

1-3 1.3 Backfit issucs..................................................

1-4 1.4 Confirmatory issues.............................................

1-7 j

1. 5 License condition issues........................................

Beaver Valley 2 SSER 2 vi

I I

1 INTRODUCTION AND GENERAL DISCUSSION i

1.1 Introduction i

The Nuclear Regulatory. Commission (NRC) Safety Evaluation Report (NUREG-1057)

(SER) on the application of the Duquesne Light Company (DLC or the applicant) for a license to operate the Beaver Valley Power Station, Unit 2, was issued in October 1985.

Supplement 1 was issued in May 1986.

This is the second supplement to the SER.

The purpose of this second Supplemental Safety Evaluation Report (SSER 2) is to l

revise the SER by providing the results of the staff's review of new information j

subsequently submitted by the applicant.

The information provided in letters-referenced in this SSER must be acceptably documented in amendments to the Beaver Valley Unit 2 Final Safety Analysis Report (FSAR) by the applicant before 4

the unit is licensed.

Each section or appendix of this SSER is designated and titled so that it cor-responds to the section or appendix of the SER that has been affected by the staff's additional evaluation.

Except where specifically noted, the SSER does not replace the corresponding SER section or appendix.

Appendix A is a contin-i uation of the chronology of events, including correspondence, leading to the l

publication of this SSER.

Appendix B is a list of references cited'in this supplement.* Appendix D is a list of abbreviations used in this supplerent.

Appendix E is a list of the principal contributors to this SSER.

Appencices L and M, being added to the SER by this supplement, are staff evaluations of the i

applicant's conformance to (1) Item 2.1 (Part 1) and (2) Items 3.1.3 and 3.2.3, l

respectively, of Generic Letter 83-28.

No changes were made to other appendices.

Tables j

1.2,1.3,1.4, and 1.5, all corresponding to tables of the same numbers in the SER, provide summaries of the status of open, backfit, confirmatory, and 1

j license condition issues, respectively.

If the status of an issue has changed j

since issuance of the SER, details of the change are documented in this supple-l 1

ment.

The next supplement (SSER 3) is expected to be issued in November 1986.

Copies of this SSER are available for public inspection in the NRC Public Docu-4' ment Room at 1717 H Street N.W., Washington, D.C., and at the B. F. Jones Memorial

{

Library, 663 Franklin Ave., Aliquippa, Pa.

Copies of this SSER are also available for purchase from the sources indicated on the inside front cover of this report.

i The NRC Project Manager is Peter S. Tam.

Mr. Tam may be contacted by calling (301) 492-9409 or by writing to the following address:

j Peter S. Tam j

Division of PWR Licensing-A

}

U.S. Nuclear Regulatory Commission j

Washington, D.C. 20555 I

i

• Availability of all material cited is described on the inside front cover of I

this supplement, t

J Beaver Valley 2 SSER 2 1-1 i

Table 1.P.

Open issues Issue Status SER section (1) Preservice/ inservice testing Under review 3.9.6 (2) Pump and valve leak testing Updated in SSER 1 3.9.6 but remains open (3) Inadequate core cooling instrumenta-Closed in SSER 2 4.4.7 tion (Item II.F.2 of NUREG-0737)

(4) Preservice/ inservice inspection program i

(a) PSI Under review 5.2.4.3, 5.4.2.2 (b) ISI Updated in SSER 1 6.6 but remains open (5) Safe and alternate shutdown Unchanged from SER 9.5.1 (6) Management and organization Unchanged from SER 13.1 (7) Cross-training program Closed in SSER 1 13.2.1.2 (8) Emergency preparedness plan Unchanged from SER 13.3.3 (9) Initial test program Under review 14 (10) Control room design review Updated in SSER 1 18.1 but remains open (11) Safety parameter display system Updated in SSER 1 18.2 but remains open t

l Beaver Valley 2 SSER 2 1-2

Table 1.3 Backfit issues Issue Status

• SER section (1) Snow and ice load C

2.3.1 (2) Underestimation of atmospheric dispersion C

2.3.4, 15.4.8 conditions (X/Q) at exclusion area boundary and consequences of radioactive release (3) Potential for flooding from probable maximum C

2.4.2, 2.4.10 precipitation and Peggs Run (4) Steam generator level control and protection C2 7.3.3.12 (5) Motor-operated accumulator isolation valve C

8.3.1.12 (6) Spent fuel pool maximum heat load C

9.1. 3 (7) Fire suppression in the cable spreading room A

9.5.1.6 (8) Class 1E pcwer for lighting and communication C

9.5.2.1 systems (9) Application of GDC 5 to communication systems C

9.5.2.1 (10) Application of GDC 2 and 4 to communication C

9.5.2 systems (11) Application of GDC 4 to lighting systems C

9.5.3 (12) Illumination levels in excess of SRP criteria C

9.5.3 (13) Application of RG 1.26 to areas excluded by C

9.5.4-9.5.8 RG 1.26 (14) Air dryers for emergency diesel generator C

9.5.6 (15) Alarm for rocker arm lube oil reserve C

9.5.7 (16) Diesel lube oil fill procedure C

9.5.7

• A - Issue was discussed in appeal meeting, and partial resolution was ad-dressed in the SER (October 1985).

C - Closed in SER (October 1985).

C2 - Closed in SSER 2, confirmatory issue 49 opened.

Beaver Valley 2 SSER 2 1-3

Table 1.4 Confirmatory issues Issue Status SER Section (1) Operating procedures for continuous Unchanged from SER 2.2.2 communication links (2) Differential settlements of buried pipes Unchanged from SER 2.5.4.5 (3) Internally generated missiles (outside Unchanged from SER 3.5.1.1 containment)

(4) Internally generated missiles (inside Unchanged from SER 3.5.1.2 containment)

(5) Turbine missiles Unchanged from SER 3.5.1.3 (6) Analysis of pipe-break protection Unchanged from SER 3.6.1 outside containment (7) FSAR drawings of break locations Unchanged from SER 3.6.2 (8) Results of jet impingement effects Unchanged from SER 3.6.2 (9) Soil-structure interaction analysis Closed in SSER 1 3.7.3 (10) Design documentation of ASME Code Closed in SSER 2 3.9.3.1 components (11) Item II.D.1 of NUREG-0737 Under review 3.9.3.2 (12) Seismic and dynamic qualification of Under review 3.10.1 mechanical and electrical equipment (13) Pump and valve operability assurance Under review 3.10.2 (14) Environmental qualification of Under review 3.11 mechanical and electrical equipment (15) Peak pellet design basis Closed in SSER 1 4.2.1 (16) Discrepancies in the FSAR Closed in SSER 1 4.2.2 (17) Rod bowing analysis Closed in SSER 1 4.2.3.1(6)

(18) Fuel rod internal pressure Closed in SSER 1 4.2.3.1(8)

(19) Predicted cladding collapse time Closed in SSER 1 4.2.3.2(2)

(20) Use of the square-root-of-the-sum-of-Closed in SSER 1 4.2.3.3(4) the-squares method for seismic and loss-of-coolant-accident load I

calculation Beaver Valley 2 SSER 2 1-4

.. ~ _

l Table 1.4 (Continued)

Issue Status SER section (21) Analysis of combined loss-of-coolant-Under review 4.2.3.3(4) accident and seismic loads (22) Natural circulation test Updated in SSER 1 5.4.7.5 but remains open (23) Reactor coolant system high point vents Under review 5.4.12 (24) Blowdown mass and energy release Under review 6.2.1.3 analysis methodology (25) Containment sump 50% blockage assumption Updated in SSER 2 6.2.2 but remains open (26) Design modification of automatic reactor Under review 7.2.2.3 trip using shunt coil trip attachment (27) Automatic opening of service water Closed in SSER 1 7.3.3.10 system valves MOV113C and 113D (28) IE Bulletin 80-06 concerns Unchanged from SER 7.3.3.13 (29) NUREG-0737, Item II.F.1, accident Closed in SSER 1 7.5.2.2 monitoring instrumentation positions (30) Bypass and inoperative status panel Under review 7.5.2.4.

(31) Revision of the FSAR--cold leg accumu-Unchanged from SER 7.6.2.4 lator motor-operated valve position indication (32) Control system failure caused by Under review 7.7.2.3 malfunctions of common power source or instrument line (33) Confirmatory site visit (a)

Independence of offsite power Closed in SSER 1 8.2.2.3 between the switchyard and Class 1E system (b) Confirmation of the protective Closed in SSER 1 8.3.1.2 bypass l

(c) Verification of DG start and load Closed in SSER 1 8.3.1.8 l

bypass (d) DG load capability qualification Closed in SSER 1 8.3.1.9 test l

Beaver Valley 2 SSER 2 1-5

i l

Table 1.4 (Continued)

Issue Status SER section (33) Confirmatory site visit (Continued)

(e) Margin qualification test Closed in SSER 1 8.3.1.10 (f) Electrical interconnection between Closed in SSER 1 8.3.1.13 redundant Class 1E buses (g) Verification of electrical Closed in SSER 1 8.3.3.5 independence between power supplies to controls in control room and remote locations (34) Voltage analysis--verification of test Unchanged from SER 8.3.1.1 results i

(35) Documentatioit of description and analysis Unchanged from SER 8.3.3.7.1 of compliance with GDC 50 (36) Completion of plant-specific core damage Unchanged from SER 9.3.2.2 estimate procedure before fuel load (37) Training program for the operation and Unchanged from SER 9.5.4.1 maintenance of the diesel generators (38) Vibration of instruments and controls on Unchanged from SER 9.5.4.1 l

diesel generator (39) Surveillance of lube oil level in the Closed in SSER 2 9.5.7 diesel generator rocker arm tube oil reservoir (40) Solid waste process control program Unchanged from SER 11.4.2 (41) TMI Action Plan items (a)

III.D.1.1 Unchanged from SER 13.5.2 (b)

II.K.1.5 and II.K.1.10 Under review 15.9.2 15.9.3 (c)

II.K.3.5 Urider review 15.9.9 (d)

II.K.3.17 under review 15.9.11 (e)

II.K.3.31 Under review 15.9.14 (42) Plant-specific dropped rod analysis Closed in SSER 2 15.4.2 (43) Steam generator tube rupture Under review 15.6.3 Beaver Valley 2 SSER 2 1-6

Table 1.4 (Continued)

Issue Status SER section (44) Quality assurance program Closed in SSER 1 17.4 (45) Cross-training of Unit 1 & 2 operators Unchanged from 13.2.1.1 SSER 1 (46) Control room isolation on high radiation Unchanged from 7.3.3.9 signal SSER 1 (47) Review of procedures generation package Unchanged from 13.5.2 SSER 1 (48) Fire protection:

Amendment 12 review and Opened in SSER 2 9.5.1 site visit (49) Steam generator high-level trip as non-Opened in SSER 2 7.3, protection system 15.1.2 (50) Implementation letter on ICCI system Opened in SSER 2 4.4.7 l

Table 1.5 License condition issues License conoition Status SER section

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(1) Emergency response capability, Specifics provided 7.5.2.1 RG 1.97, Rev. 2 in SSER 1; under additional generic review l

l Beaver Valley 2 SSER 2 1-7 j

j

l 3 DESIGN OF STRUCTURES, COMPONENTS, EQUIPMENT, AND SYSTEMS 3.9 Mechanical Systems and Components 3.9.3 ASME Code Class 1, 2, and 3 Components, Component Supports, and Core Support Structures 3.9.3.1 Loading Combinations, Design Transients, and Stress Limits j

In the SER, the staff stated that it had not completed its review of design documents for selected pumps, valves, and piping.

This was identified as con-firmatory issue 10 in Table 1.4 of the SER.

After the SER was published in October 1985, the staff requested additional information from the applicant to clarify information obtained during the audit, which was briefly discussed in l

the SER.

In a letter dated January 3,1986, the applicant responded to all of l

the staf f's requests.

The staff's evaluation of this information is discussed l

below.

l The staff's design documentation audit was initiated by a meeting at the S&W office in Boston, Massachusetts, on April 4 and 5, 1984.

Tais initial meeting was supplemented by further exchanges of information between the staff's con-sultant, Oak Ridge National Laboratory, and the applicant via correspondence and telephone conversations.

The staff's audit consisted of an evaluation of the following specific types of compolints:

(1) ASME Class 3 service water pumps (2) ASME Class 2 and 3 motor-operated butterfly valves used in service water systems.

(3) a portion of the ASME Class 2 steam generator blowdown piping system (4) supports for piping systems In the process of conducting this audit, the staff and its consultant reviewed the following documents whic.h are applicable to the four types of components listed above and are available at the Stone and Webster Engineering Corporation (S&W) office:

(1) Service Water Pump Data (a) Pump Dosign Specification 2BVS-224, Revision 3, Addenda 1-4 (b) Seismic Analysis, Revision 2 - S&W File No. 2602.540-224-001C (c) Seismic Analysis Addendum 1 - S&W File No. 2602.540-224-036A i

l (d) Outline Drawing - S&W File No. 2002.540-224-002H i

(e) Outline Drawing - S&W File No. 2002.540-224-003I Beaver Valley 2 SSER 2 3-1 l

(2) Motor-0perated Butterfly Valve Data (a) Valve Design Specification 2BVS-76A, Revision 2, Addendum 1 (b) 25WS*MOV107 Seismic Analysis (with Addendum B) - S&W File No. 2606.450-76A-1118 (c) Seismic Functional Test (with Addendum A) - S&W File No. 2606.450-76A-114F.

(d) Valve Manufacturer's Drawing, S&W File No. 2006.450-76A-072K (e) Reactor Containment Ventilation System S&W Drawings (12241)

RB-160-8R RB-16M-4E RB-15Q-4 RB-15F-9 (3) Steam Generator Blowdown Piping and Piping Support Data (a) Stress Analysis Data Package for Steam Generator Blowdown System (BDG) - RM-100A, Revision 1 (b) Flow Diagram RM-100A-13 (c) Piping Drawings RP-99A-60, RP-990-58, and RP-99F-3F (d) Pipe Stress Calculation X99K, Revision 2 (e) Pipe Supports Welding Design Guide 2BVM-102, Revision 2 (f) Piping Engineering and Design Specification 2BVS-939, Revision 3, and addenda up to and including Addendum No. 5 (g) Shop Fabricated Piping Specification 2BVS-58, Revision 4 (h) Design and Fabrication of Power Plant Piping Supports 2BVS-59, Revi-sion 2, and addenda up to and including Addendum No. 6 (1) S&W Pipe Classes Design Specification 2BVS-939A, Revision 4, and Addendum No. 1 (j) Field Fabrication and Erection of Piping Specification 2BVS-920, Revision 7, and addenda up to and including Addendum No. 4 (k) Procedure for Preparation of System Design Information Required for Pipe Stress Analysis - 2BVM-45 dated June 6, 1983 On the basis of a review of the above design documentation, the information.in the January 3, 1986, letter, and subsequent clarifying telephone conversations on this issue, the staff arrived at the following conclusions:

Beaver Valley 2 SSER 2 3-2

(1) Design specifications required by the ASME Code have been prepared and contain a complete basis for construction of the components.

(2) The equivalents of design reports for ASME Code Class 2 and 3 components which are required by the Code have been prepared.

The input data used are traceable to and agree with the design specification, and the analyses show compliance with Code requirements.

(3) The design specifications include appropriate provisions to ensure adequate performance of components during their anticipated service, and to demon-strate that appropriate documentation has been received which shows com-pliance with the specifications.

On the basis of the above conclusions, the staff considers confirmatory issue 10 resolved.

Beaver Valley 2 SSER 2 3-3

4 REACTOR 4.4 Thermal-Hydraulic Design 4.4.7 Inadequate Core Cooling Instrumentation The SER stated that this was an open issue because the applicant had not supplied sufficient information.

By letter dated March 3, 1986, the staff for-mally requested information needed to complete its review.

By letters dated April 11 and July 31, 1986, the applicant submitted the requested information.

The applicant stated that emergency operating procedures, when written, will conform to approved emergency response guidelines.

A separate review of emer-gency operating procedures is conducted only if the applicant deviates from the approved guidelines.

The Beaver Valley Unit 2 inadequate core cooling (ICC) monitoring system will be installed and tested before fuel load, and will be calibrated before the plant achieves 5% of rated power.

The ICC monitoring system installed at Beaver Valley Power Station includes the following:

(1) Core Exit Thermocouple System The incore thermocouple / core cooling monitor has been installed to provide the plant operators improved information presentation and to improve the display of the status of core heat removal capability.

The system moni-tors all core exit thermocouples and calculates core subcooling margin using redundant channels of instrumentation and control room displays.

The core exit temperature monitoring system is a part of the incore instru-mentation system which consists of 51 Type K chromel-alumel thermocouples at fixed core outlet positions.

The core exit thermocouple monitoring system employs two redundant independent trains that monitor all 51 of the chromel-alumel core exit thermocouples (26 thermocouples on protection set III and 25 thermocouples on protection set IV).

The thermocouple wires exit the upper head through four penetrations.

The qualified thermocouple wires then proceed through a Swagelok to a qualified connector to facili-tate disconnecting the wires when the upper head is removed.

The wiring exits the reactor vessel cavity and is routed to qualified reference junc-tion boxes located inside the containment.

The compensated thermocouple signal wiring then penetrates the containment wall and connects to signal i

processing units which provide readouts for control room displays.

The thermocouples have a range of 0 to 2300 F.

The thermocouple information is provided as an output on two redundant plasma displays which serve as both primary and backup displays.

(2) Core Subcooling Margin Monitor The subcooling margin monitor calculates the margin to saturation tempera-ture using wide-range reactor coolant system (RCS) pressure, up to 3000 psi, Beaver Valley 2 SSER 2 4-1 l

L

and a core exit temperature value based upon the auctioneered high thermo-couple trisector average temperatures.

This represents bulk loop tempera-ture and is consistent with the Westinghouse Owners Group Dnetgency Response Guidelines (ERGS).

The calculated values for the subcooling margin are routed to redundant plasma displays and analog indicators.

The applicant stated that the cable routing from sensor input to display meets the guide-lines of Regulatory Guide (RG) 1.75 and that the RG 1.97 range-guidelines of 200 F subcooling and 35*F superheat are satisfied.

The monitoring system displays several levels of information including:

(a) bulk average trisector core exit thermocouple temperature trending, (b) a spatial map exhibiting the thermocouple temperature at its respective location in the core, (c) a core map showing minimum, average, and maximum trisector temperatures, (d) subcooling margin, (e) a detailed data list exhibiting thermocouple location, tag designation, and temperature, and (f) hot channel core exit temperature.

(3) Reactor Vessel Level Instrumentation System The reactor vessel level instrumentation system (RVLIS) consists of two redundant independent trains that monitor the reactor vessel water level.

Each train provides three vessel level indications:

narrow range, full range, and dynamic head. The narrow range measures from the middle of the hot leg to the top of the reactor vessel head.

The full range RVLIS reading provides an indication of reactor vessel water level from the bottom of the vessel to the top of the vessel during natural circulation conditions.

The dynamic head reading provides an indication of reactor core, internals, and outlet nozzle pressure drop for any combination of operating reactor coolant pumps.

Comparison of the measured pressure drop with the normal pressure drop provides an approximate indication of the relative void content of the circulating fluid.

The three RVLIS readings are routed to plasma displays in the control room.

The applicant stated that the cable routing from sensor input to display meets the guidelines of RG 1.75.

The range of the full-range channels corresponds to O to 120% of the reactor vessel height.

The staff has reviewed the applicant's submittals dated April 11 and July 31, 1986. The applicant's July 31, 1986 letter noted that installation and calibrations are scheduled for completion by September 1986. The staff will require that the ICCI system be fully operational with appropriate emergency operating procedures in place prior to fuel load, and that the system shall be fully calibrated prior to exceeding 5% rated power. The acceptability of the final design will be demonstrated by the successful completion of preoperational testing of the ICCI subsystems. An implementation letter report containing the following information is required to complete the staff's review of implementation approval of the installed ICCI system, and must be provided prior to exceeding 5% rated power. Confirmatory issue 50 is opened to track this action.

Beaver Valley 2 SSER 2 4-2

a.

Notification that the system installation, functional testing, and calibration are complete and test results are available for inspection, b.

Summary of applicant's conclusions based on test results, e.g.,

the system performs in accordance with design expectations and within design error tolerances; or description of deviations from design perfonnance specifications and basis for concluding that the deviations are acceptable, c.

Description of any deviations of the as-built system from previous design descriptions with any appropriate explanation.

d.

Confirmation that the emergency operating procedures (E0Ps) used for operator training are complete and conform to the technical content of NRC-approved E0P guidelines.

(3) The core exit temperature monitors, subcooling margin monitors, and reactor vessel level monitors meet the single-failure criteria of RG 1.53.

No single failure within the ICC instrumentation or its auxiliary support equipment will render the system inoperable.

On the basis of its review of FSAR Sections 4.4.6.5 and 7.5 and the additional information provided by the applicant, the staff finds that the design of the Beaver Valley Unit 2 ICC instrumentation system is in conformance with the guidelines of NUREG-0737, and is, therefore, acceptable.

This closes open l

issue 3.

l Beaver Valley 2 SSER 2 4-3 I

6 ENGINEERED SAFETY FEATURES

=-I 6.2 Containment Systems n-6.2.2 Containment Heat Removal Systems By letter dated June 6, 1986, the applicant responded to a staff concern iden-tified as confirmatory issue 25 in the SER.

In the SER, the staff recommended g

that the applicant provide a debris generation and transport. analysis to justify g

the 50% sump blockage assumption used in assessing the design and performance a

of the emergency sump.

The applicant stated that the plant-specific debris i

evaluation is addressed only in Revision 1 of RG 1.82 which is not applicable to Beaver Valley Unit 2; a 50% sump blockage assumption is specified in Revi-sion 0 of RG 1.82.

The applicant claims that Beaver Valley Unit 2 meets the g_

intent of Revision 0 of RG 1.82, and, therefore, requested that this SER con-firmatory issue be closed.

_L h

The staff does not agree with the applicant because the need to justify the 50%

sump blockage assumption is not based on Revision 1 of RG 1.82.

Rather, the E

need for additional analysis was prompted by the fact that Beaver Valley Unit 2 does not meet the guidance in Revision 0 of RG 1.82.

More specifically, Revi-i sion 0 of RG 1.82 recommends that the design coolant velocity at the inner screen should be about 0.2 ft/sec, but the average flow velocity at the sump T

screen is 0.31 ft/sec.

The higher flow velocity means more debris could be transported to the sump screens.

Therefore, a 50% sump blockage assumption may not be appropriate, and the results of the net positive suction head (NPSH) analysis presented in the FSAR may not be valid.

As a result, there is no

=

assurance that sufficient suction head will be available to the pumps in the emergency core cooling system (ECCS) and the containment spray system (CSS) following a loss-of-coolant acciden't (LOCA).

On the basis of the above discussion, the staff finds the applicant's response

=

E unacceptable, and the staff's concern expressed in confirmatory issue 25 remains y

unchanged.

L 6.2.7 Fracture Prevention of Containment Pressure Boundary p

In the SER, the staff concluded that the ferritic materials of the components of the Beaver Valley Unit 2 containment pressure boundary complied with General Design Criterion (GDC) 51 because the materials met the fracture toughness re-quirements for Class 2 components according to the Summer 1977 Addenda of Sec-tion III of the ASME Code.

The ASME Code fracture toughness requirements for Class 2 components are defined in Article NC-2330 of Section III of the ASME

[

Code.

In order to meet these ASME Code fracture toughness requirements for components thicker than 2-1/2 inches, the material's permissible lowest service

=

temperature must be less than the component's lowest service temperature.

In a letter dated April 11, 1986, the applicant provided results of its reevalua-

{

tion of the lowest service temperature for the feedwater lines.

The applicant Beaver Valley 2 SSER 2 6-1

=

concluded that materials 'in the line met the Class 2 Section III fracture tough-ness requirements except for the 3-1/2-inch-thick feedwater check valve covers.

The applicant replaced the covers with components in which permissible lowest service temperature was less than the revised lowest service temperature.

The staff, therefore, concludes that the replacement covers meet the fracture toughness requirements for Class 2 components of Section III of the ASME Code and that the applicant has satisfied the requirements of GDC 51.

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l Beaver Valley 2 SSER 2 6-2

7 INSTRUMENTATION AND CONTROLS 7.1 Introduction 7.1. 4 Specific Findings 7.1.4.5 Site Visit In the SER, the staff indicated that a site review will be performed to confirm that both the physical arrangement and instal.lation of electrical equipment are in accordance with the design criteria and descriptive information approved by the staff.

On May 6 and 7, 1986, the staff visited the site, primarily to ver-ify that the installation of the instrumentation and control equipment conformed to the applicable design criteria regarding physical separation between redun-dant safety rglated cirguits and between safety-related and non-safety-related circuits.

In apdition, the staff reviewed the Beaver Valley Unit 2 design to verify that the 'ivr.t.rumentation and control systems had been installed consist-ent with the design basis as described in Chapter 7 of the FSAR and as depicted in electrical schematics and elementary diagrams.

Areas of review included the control room, instrument racks, containment building, turbine building, remote shutdown panels, and die.sel generator rooms.

The results of the site visit are provided in the following discussion, as well as in Section 7.3.3.2 of this SSER.

Each building / room is listed followed by a discussion of the observations and conclusions made by the staff.

In general, the physical arrangement and installation of the instrumentation and control systems a9peared to be in accordance with the applicable design criteria.

(1) Control Rocm-The staff reviewed the general layout of the control room including the main control boards, annunciation, and controls.

Cable separation was reviewed within main control board panels.

The internal cabinet wiring was found to conform to the separation criteria defined in the Beaver Valley Unit 2 FSAR.

The staff reviewed the bypass and inoperable status panel and verified that the panel is designed to satisfy the requirements of IEEE Std. 279(1971) and Regulatory Guide (RG) 1.47 as discussed in the SER.

(2)

Instrument Rooms The staff reviewed the internal cabinet wiring within several protection system racks (process, input, logic, output).

The Beaver Valley Unit 2 protection system racks are typical Westinghouse 7300 Series composed mainly of separate bays dedicated to individual channels or logic trains.

Where redundant circuits existed within a cabinet or bay (input), the staff verified that the internal wiring conformed to the separation criteria defined in the FSAR.

Beaver Valley 2 SSER 2 7-1

(3) Containment Building The staff reviewed several safety-related instrument installations includ-ing the steam generator level, hot-and cold-leg resistance temperature detectors, and reactor coolant flow instruments.

In addition, the staff reviewed the emergency core cooling system (ECCS) cccumulator valves, the pressurizer pilot-operated relief valve (PORV) installation, and the steam generator PORV.

No concerns were identified.

The staff reviewed the test features (test valves, connections, etc.) of a typical safety-related differential pressure transmitter.

A typical test sequence was discussed with the applicant.

No concern was identified.

The staff reviewed the auxiliary feedwater system.

The review included the auxiliary feedwater pumps, flow-control valves, and piping.

No concern was identified.

(4) Turbine Building The staff reviewed the turbine stop valves and the turbine electrohydraulic control system.

Special attention was given to the routing and separation of the safety related cables associated with the turbine stop valve limit switches and low auto-stop oil pressure switches which provide inputs into the reactor protection system (see SER Section 7.2.1).

Although the tur-bine building is a non-safety-related building, the applicant has treated these cables as safety related (color coding, seismic mounting, separate conduits, etc.) within the structure.

No concerns were identified.

The staff also reviewed the cable routing and separation of the redundant circuits which would trip the turbine following a reactor trip (see SER Section 7.2.2.2).

The staff's review indicates that the applicant pro-vided an individual conduit for each circuit within the turbine building up to the turbine pedestal.

No concern was identified.

(5) Remote Shutdown Panel / Appendix R Panel Two redundant trains of instruments and controls are provided at remote shutdown panels.

The staff reviewed the panel arrangement and cable separation within the remote shutdown panel.

Where redundant circuits existed within a panel, the staff verified that the internal wiring con-formed to the separation criteria defined in the FSAR.

One train of instrumentation and controls is provided at the safe-shutdown panel.

During the site visit, the staff reviewed the panel arrangement, and no new concern was identified.

(6) Outside Tanks and Instrument Sensing Lines The staff reviewed several outside storage tanks including the refueling water storage tank and primary plant demineralized water storage tank.

Special attention was given to the use of heat tracing applied to instrument-sensing lines such as those for sensing tank levels; at the time of the site visit, no heat tracilig had been installed.

Beaver Valley 2 SSER 2 7-2

(7) Diesel Generator Building The staff reviewed one diesel generator room during this site visit.

No concern was identified.

(8) Sequencer Panel The staff reviewed one diesel generator sequencer panel during this site visit.

No concern was identified.

(9) Static Inverter / Batter / Room The staff reviewed one static inverter panel and one battery room.

The operation of the inverter and the transfer of sources were discussed.

No concern was identified.

7.3 Engineered Safety Feature Systems 7.3.3 Specific Findings 7.3.3.2 Test of Engineered Safeguards P-4 Interlock In the SER, the staff stated that the applicant was not installing a permanent voltage indicator to measure the P-4 interlock during testing at the reactor trip breaker cabinets.

During the site visit discussed above, the staff observed the P-4 interlock terminal blocks from which the voltage measurements would be taken and concluded that accidental shorting or grounding was unlikely to occur.

No concern was identified.

7.3.3.12 Steam Generator Level Control and Protection The SER provided details of resolution of backfit issue 4 regarding steam gen-erator level control.

However, the SER also stated that additional review of the applicability of 10 CFR 50.55a(h), which requires compliance with the edi-tion of IEEE-279 in effect on the formal docket date, was necessary.

In a letter dated November 22, 1985, the staff requested that the applicant revise the FSAR so that the steam generator high-level trip system is not repre-sented as an engineered safety feature (ESF) actuation system.

The applicant responded by letter dated December 20, 1985, indicating that the FSAR will be revised by Amendment 11 to clarify : hat high steam generator level is not required for protection but is used only as a convenient ending point for the snalysis in Section 15.1.2 of the FSAR.

The staff has reviewed the response and Amendment 11 and has identified addi-tional references to the high-level trip system in FSAR Chapter 15 analyses.

A second letter, dated June 30, 1986, was sent requesting further revision of the FSAR.

The revised FSAR sections will be reviewed by the staff and reported on in a future supplement.

Such review is tracked by confirmatory issue 49.

Because the applicant has clearly stated that the high-level trip system is not a protection system, the requirement of 10 CFR 50.55a(h) does not apply.

The subject backfit issue 4, therefore, is considered closed; the revision of FSAR Beaver Valley 2 SSER 2 7-3

i i

Chapters 7 and 15 is a result of the resolution of the backfit issue, and will be tracked by confirmatory issue 49.

The staff's June 30, 1086, letter provides details of needed revisions to the FSAR.

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i Beaver Valley 2 SSER 2 7-4

9 AUXILIARY SYSTEMS 9.5 Other Auxiliary Systems 9.5.1 Fire Protection Program In the SER, the staf f indicated that the staff would visit the site to examine the relationship of safety-related components, systems, and structures in specific plant areas both to combustible materials and to associated fire de-tection and suppression systems.

In addition, the applicant has recently sub-mitted FSAR Amendment 12, which, among other things, documents changes to fire protection commitments made by letters dated March 27 and May 14, 1985.

There-fore, confirmatory issue 48 is opened to track the site visit and the resolution of Amendment 12 information regarding fire protection.

9.5.7 Emergency Diesel Engine Lubricating Oil System In the SER, the staff stated that the emergency diesel engine lubricating oil system design is acceptable with the proviso that surveillance of the lube oil level in the rocker arm lube oil reservoir be included in the plant operating procedures. This proviso is the subject of confirmatory issue 39.

By letter dated March 26, 1986, the applicant submitted a copy of the log sheet which is used to document required periodic check of the emergency diesel engine rocker arm lube oil reservoir level.

This log sheet is part of the Beaver Valley Unit 2 operating manual.

On the basis of this evidence, the staff considers confirmatory issue 39 resolved.

Beaver Valley 2 SSER 2 9-1

t 15 ACCIDENT ANALYSIS

~

15.4 Reactivity and Power Distribution Anomalies 15.4.2 Uncontrolled Rod Cluster Control Assembly (Rod) Bank. Withdrawal at:

Power i

In a letter dated' April 1, 1986, the applicant informed the staff that Westing-house has completed the dropped rod analysis for Cycle 1 of Beaver Valley Unit 2.

This analysis was performed using the methodology in Westinghouse report WCAP-10297-PA, " Dropped Rod Methodology for Negative Flux Rate Trip Plants."

On the basis of this analysis, the restriction of operations above 90% of rated power, described in Section 15.4.3 of the SER, may be removed.

This resolves confirmatory issue 42.

15.8 Anticipated Transients Without Scram Status of Salem ATWS Event Issues On July 8,1983, the NRC issued Generic Letter (GL) 83-28 as a result of the anticipated transient without scram (ATWS) events at Salem Nuclear Generating Station.

This letter addressed actions to be taken by licensees and applicants to ensure that a comprehensive program of preventive maintenance and surveillance testing is implemented for the reactor trip breakers in pressurized water reactors.

The staff has completed its review of parts of the applicant's response to GL 83-28 and will document its results in appendices to the SER.

The following listing serves as a record of completion of staff review, and shows where indi-vidual safety evaluations may be found:

Item 1.1, Post-Trip Review (Appendix K, SSER 1)

Item 2.1, Equipment Classification and Vendor Interface (Reactor Trip System Components) (Appendix L, SSER 2)

Items 3.1.3 and 3.2.3, Postmaintenance Testing in Technical Specification That Could Degrade Safety (Appendix M, SSER 2)

Item 4.1, Trip System Reliability (Appendix J, SSER 1)

(

Items 4.2.1 and 4.2.2, Preventive Maintenance Program for Reactor Trip l

Breakers - Maintenance and Trending (Appendix J, SSER 1)

Status of Implementation of 10 CFR 50.62, ATWS Mitigation System Actuation

[

Circuitry (AMSAC) l

(

The Code of Federal Regulations [10 CFR 50.62(d)] requires the submittal of a schedule for meeting the requirements of 10 CFR 50.62(c).

By letter dated

(

September 30, 1985, the applicant committed to install the AMSAC to meet the f

Beaver Valley 2 SSER 2 15-1

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requirement.

Specifically, the applicant has proposed to adopt the Westing-house Owners Group (WOG)-developed AMSAC design.

The WOG generic design pack-age was submitteu on July 25, 1985, and was approved by the staff on July 7, 1986 (letter, C. E. Rossi to L. Butterfield).

3 The applicant committed to provide a proposed schedule for meeting the require-ments of the subject regulations 180 days after the issuance of the safety i

evaluation report on the WOG generic design.

The staff will report on this in a future supplement.

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Beaver Valley 2 SSER 2 15-2

t i

i 17 QUALITY ASSURANCE 17.1 General i

In_the SER, the NRC staff found the quality assurance (QA) program description contained in Section 17.2 of the FSAR acceptable for the operations phase of I

i Beaver Valley Unit 2.

Since then, the applicant has amended the FSAR through Amendment 12 with several organizational changes and typographical corrections.

The staff finds the FSAR up to Amendment 12 acceptable and the following l

information supersedes that offered in the SER.

17.2 Organization The applicant's organization responsible for the operation of Beaver Valley i

Unit 2 is shown in FSAR Figure 17.2-1.

It consists of three groups; namely, the l

Nuclear Group, the Power Supply Group, and the Administrative Services Group.

Each of these groups is headea by a Vice President who reports directly to the Chairman of the Board and the President.

Reporting directly to the Nuclear Group Vice President are the Nuclear Engineer-

[

ing and Construction General Manager, Nuclear Operations General Manager, Nuclear Service General Manager, and Quality Assurance Manager.

i l

The Nuclear. Engineering and Construction General Manager is responsible for performing engineering and construction activities, including administrative control over design activities associated with nuclear power plants whether conducted by the applicant or contracted to others, t

The Nuclear Operations General Manager is responsible for the startup, testing, operation, maintenance, and refueling of Beaver Valley Unit 2 and the direction of offsite department nuclear staff activities required to provide technical i

support to Beaver Valley Unit 2.

Responsibility for the safe and efficient operation of the plant in accordance with the guidance and requirements of the operating license and station manuals rests with this organization.

l

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The Nuclear Services General Manager is responsible for directing nuclear training, administrative services (which includes records management), budget and fuel contracts, radiological control, and nuclear safety.

The Quality Assurance Manager is responsible for quality control, preservice and inservice inspection, QA engineering and procurement, and operational QA i

activities.

Responsibility for establishing, managing, and measuring the effec-1 tiveness of the Operations QA Program rests with this organization.

The Opera-tions QA Program is established and managed by the QA Manager in accordance with applicable regulatory requirements and the quality assurance policy which is promulgated by the Chairman of the Board and President of the Duquesne LigTn,

)

Company (DLC).

The QA Manager has the authority to report quality matters to 1

any level necessary within DLC in order to establish effective corrective action.

j The QA and QC (quality control) personnel have sufficient authority and organi-i zational freedom from the pressures of cost and schedules to identify quality 1

Beaver Valley 2 SSER 2 17-1

problems; initiate, recommend, or provide solutions to quality problems through designated channels; verify implementation of solutions to quality problems; and control further processing, delivery, or installation of nonconforming items until these items have been properly dispositioned.

1 I.

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I Beaver Valley 2 SSER 2 17 2 I

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APPENDIX A CONTINUATION OF CHRONOLOGY OF NRC STAFF RADIOLOGICAL REVIEW OF BEAVER VALLEY POWER STATION, UNIT 2 1

April 3, 1986 Letter to applicant requesting additional information on Items 3.1.3 and 3.2.3 of Generic Letter 83-28.

April 4, 1986 Letter from applicant transmitting results of dropped rod analysis for fuel cycle 1.

April 9, 1986 Letter to applicant requesting additional information on proposed Technical Specifications.

April 9, 1986 Letter from applicant transmitting Inservice Testing Pro-grams for pumps and valves.

April 9, 1986 Letter from applicant transmitting additional information on safety parameter display system.

' April 9, 1986 Letter to applicant transmitting Special Nuclear Materials l

License No. SNM-1954.

April 10, 1986 Third meeting with applicant-to discuss status of WHIPJET (summary dated April 16, 1986).

April 11, 1986 Letter from appilcant responding to staff questior s of March 3, 1986, on inadequate core cooling instrumentation.

April 11, 1986 Letter from applicant addressing fracture toughness prop-erties of containment pressure boundary components with respect to GDC 51.

l April 15, 1986 Letter.from applicant providing additional information on bypassed and inoperable systems indication logic.

April 22, 1986 Lettgr to applicant requesting cooperation in several staff site' visits.

April 22, 1986 Letter to applicant transmitting interim report on reviews of reactor coolant pump trip criteria.

April 22, 1986 Letter from applicant transmitting information on loading of diesel generators, clarifying- August 26, 1985, letter.

l l

April 29, 1986 Letter from applicant requesting approval to use ASME Code l

Case N-318-2.

j I

t Beaver Valley 2 SSER 2 1

Appendix A

April 30, 1986 Letter from applicant discussing the role of Westinghouse in 10 CFR 50.55(e) reporting.

May 1, 1986 Letter to applicant transmitting formal staff position and request for additional information on WHIPJET.

May 6, 1986 Letter from applicant responding to staff questions dated April 9, 1986, regarding proposed Technical Specifications.

May 6, 1986 Letter from applicant transmitting information on initial test program.

May 9, 1986 Letter from applicant providing response to Generic Let-ter 86-04 regarding engineering expertise on shift.

May 13, 1986 Letter from applicant providing additional information on preservice inspection.

May 20, 1986 Fourth meeting with applicant to discuss status of WHIPJET (summary dated May 29, 1986).

May 21, 1986 Letter from applicant transmitting reports of several audits of the Engineering Assurance Program.

May 23, 1986 Letter from applicant submitting Equipment Qual!fication Report on environmental qualification of Class 1E electrical equipment.

May 23, 1986 Letter to applicant requesting additional information on control room isolation on high radiation.

May 27, 1986 Letter to applicant approving use of ASME Code Case N-318-2.

May 28, 1986 Letter from applicant transmitting proprietary and non-proprietary reports on effects of multiple feedwater control valves failing open.

May 28, 1986 Letter to applicant formally issuing SER Supplement 1.

June 4, 1986 Letter from applicant transmitting Seismic and Dynamic Pro-gram for Safety-Related Equipment.

June 6, 1986 Letter from applicant regarding containment sump 50% block-age assumption.

June 10, 1986 Letter from applicant responding to staff questions on de-sign documentation review of ASME Code components.

June 16, 1986 Letter from applicant responding to staff questions of April 3,1986, on Generic Letter 83-28 Items 3.1.3 and 3.2.3.

June 16, 1986 Letter from applicant responding to staff questions of April 9, 1986, on safety parameter display system.

Beaver Valley 2 SSER 2 2

Appendix A

l June 17, 1986 Letter to applicant providing results and conclusions of the inspection of review plans for Engineering Assurance i

Program (inspection took place from April 28 to April 30, 1986, in Stone & Webster Engineering Corporation's office).

June 19, 1986 Letter from applicant transmitting status report of SER open and confirmatory issues.

June 20, 1986 Letter from applicant submitting FSAR Amendment 12.

June 23, 1986 Letter from applicant stating that shunt trip attachment has been installed on trip breakers.

June 23, 1986 NRC memorandum documenting several meetings held during 1984 and 1985.

June 27, 1986 Letter from applicant certifying that Amendment 12 of l

FSAR has.been distributed.

I l

June 27, 1986 Letter to applicant requesting additional information re-garding Generic Letter 83-28 Items 2.2, 4.3, and 4.5.

June 27, 1986 Letter to applicant transmitting partial safety evaluation on Generic Letter 83-28 Item 2.1.

June 27, 1986 Letter from applicant addressing TMI Action Plan Item II.K.3.31, small-break LOCA.

June 27, 1986 Letter from applicant addressing SER open issue 9, initial testing program.

June 27, 1986 Letter from appilcant requesting approval to remove primary coolant loop pipe restraints under the newly revised GDC 4.

June 30, 1986 Letter to applicant requesting FSAR be revised so it did a

not call steam generator high level an engineered safety feature actuation.

July 31,1986 Letter from applicant providing additional information on instrumentation to detect inadequate core cooling.

l i

i l

Beaver Valley 2 SSER 2 3

Appendix A

APPENDIX B BIBLIOGRAPHY U.S. Nuclear Regulatory Commission, NUREG-0737, " Clarification of TMI Action Plan Requirements," November 1980; Supplement 1, January 1983.

--, NUREG-1000, " Generic Implications of ATWS Events at Salem Nuclear Power Plant," July 1983 (attached to Generic Letter 83-28 dated Juiy 8,1983).

--, Federal Register, Vol. 51, page 12502, April 11, 1986.

Westinghouse Corp., WCAP-10297-PA, " Dropped Rod Methodology for Negative Flux Rate Trip Plants," June 1983.

Industry Codes and Standards American Society of Mechanical Engineers, Boiler and Pressure Vessel Code,Section III (Summer 1977 Addenda to the 1977 Edition), Article NC-2330, " Test Requirements and Acceptance Standards," Class 1, 2, and 3.

Institute of Electrical and Electronics Engineers, Standard 279, " Criteria for Protection Systems for Nuclear Power Generating Stations," 1971 edition.

Beaver Valley 2 SSER 2 1

Appendix 8

APPENDIX D ACRONYMS AND INITIALISMS AMSAC ATWS mitigation system actuation circuitry ASME American Society of Mechanical Engineers ATWS anticipated transient without scram i

CFR Code of Federal Regulations l

CSS containment spray system DG diesel generator DLC Duquesne Light Company ECCS emergen;y core cooling system EDO Executive Director for Operations, NRC ERG Emergency Response Guidelines ESF engineered safety feature FR Federal Register FSAR Final Safety Analysis Report GDC General Design Criteri(on)(a)

GL Generic Letter i

ICC inadequate core cooling IE Office of Inspection and Enforcement, NRC IEEE Institute of Electrical and Electronics Engineers LOCA loss-of-coolant accident i

NPSH net positive suction head NRC U.S. Nuclear Regulatory Commission P03V pilot-operated relief valve QA quality assurance QC quality control RCS reactor coolant system RG Regulatory Guide RTS reactor trip system RVLIS reactor vessel level instrumentation system SER Safety Evaluation Report SSER Supplemental Safety Evaluation Report WOG Westinghouse Owners Group i

i Beaver Valley 2 SSER 2 1

Appendix D I

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APPENDIX E NRC STAFF CONTRIBUTORS AND CONSULTANTS Principal staff reviewers and consultants who contributed to this supplement are:

Reviewer's Name Title Review Branch

• W. L. Belke QA Engineer, Nuclear Quality Assurance **

H. Brammer Senior Mechanical Engineering Engineer F. Burrows Electrical Engineer Electrical, Instrumentation &

Control Systems M. Dunenfeld Senior Nuclear Engineer Reactor Systems B. Elliott Materials Engineer Engineering R. Giardina Mechanical Engineer Plant Systems A. Gilbertt Nuclear Engineer Reactor Systems R. Karsch Nuclear Engineer Reactor Systems D. Lasher Electrical Enginger Electrical, Instrumentation &

Control Systems C. Li Mechanical Engineer Plant Systems J. Mauck Senior Electrical Electrical, Instrumentation &

Engineer Control Systems Consultant's Name Organization R. Haroldsen EG&G Idaho, Inc.

S. Moore Oak Ridge National Laboratory E. Rodabaugh Rodabaugh Associates

• Division of PWR Licensing A, Office of Nuclear Reactor Regulation, except where indicated otherwise.

• • 0ffice of Inspection and Enforcement.

tAlso contributed to SSER 1, but her name was inadvertently ommitted in that document.

Beaver Valley 2 SSER 2 1

Appendix E

)

t APPENDIX L STAFF SAFETY EVALUATION REPORT ON CONFORMANCE TO GENERIC LETTER 83-28, ITE!! 2.1 (PART 1), EQUIPMENT CLASSIFICATION (REACTOR TRIP SYSTEM COMPONENTS)

AND VENDOR INTERFACE INTRODUCTION AND

SUMMARY

On February 25, 1983, both scram circuit breakers at Unit 1 of the Salem Nuclear Generating Station (Salem) failed to open upon an automatic reactor trip signal from the reactor protection system. This incident was terminated manually by the operator about 30 seconds after the initiation of the automatic trip signal.

The failure of the circuit breakers was determined to be related to the sticking of the undervoltage trip attachment.

Before this incident, on February 22, 1983, at Salem Unit 1, an automatic trip signal was generated based on steam generator low-low level during plant startup.

In this case, the reactor was tripped manually by the operator almost coincidentally with the automatic trip.

Following these incidents, on February 28, 1983, the NRC Executive Director for l

Operations (EDO) directed the staff to investigate and report on the generic implications of these occurrences at Salem Unit 1.

The results of the staff's inquiry into the generic implications of the Salem Unit 1 incidents are reported in NUREG-1000. " Generic Implications of the ATWS Events at the Salem Nuclear Power Plant." As a result of this investigation, the Commission (NRC) requested (by Generic Letter 83-28 dated July 8, 19831) all licensees of operating reac-tors, applicants for an operating license, and holders of construction permits to respond to generic issues raised by the analyses of theses two anticipated j

transient without scram (ATWS) events.

This report evaluates the response submitted by Duquesne Light Company, the licensee for Beaver Valley Power Station, Unit 1, and applicant for Unit 2, for Item 2.1 (Part 1) of Generic Letter 83-28.

The actual documents reviewed as part of this evaluation are listed in the references.

Item 2.1 (Part 1) requires licensees and applicants to confirm that all reactor trip system (RTS) components are identified, classified, and treated as safety related as indicated in the following statement:

Licensees and applicants shall confirm that all components whose functioning is required to trip the reactor are identified as safety-related on documents, procedures, and information handling systems l

used in the plant to control safety-related activities, including j

maintenance, work orders, and parts replacement, i

EVALUATION 4

Ouquesne Light Company (DLC) responded to the requirements of Item 2.1 (Part 1) with submittals dated November 4, 19832 and March 30, 19848 DLC stated in t

Beaver Valley 2 SSER 2 1

Appendix L

these submittals that a classification list was developed to identify components necessary to trip the reactor and that all such components were reviewed to verify that these components are classified as safety-related equipment.

Some specific backup or anticipatory trip devices were excepted; however, the staff has found these exceptions to be acceptable.

DLC further stated that these systems and components are identified as safety related on relevant plant docu-ments such as operating manuals, mechanical and electrical drawings, equipment specifications, and spare parts reports.

CONCLUSION On the basis of its review of these responses, the staff finds DLC's statements confirm that a program exists for identifying, classifying, and treating compo-nents that are required for performance of the reactor trip function as safety related.

This program meets the requirements of Item 2.1 (Part 1) of Generic Letter 83-28, and is, therefore, acceptable.

REFERENCES tNRC letter, D. G. Eisenhut to all licensees of operating reactors, applicants for operating license, and holders of construction permits, " Required Actions Based on Generic Implications of Salem ATWS Events (Generic Letter 83-28),"

July 8, 1983.

2 Letter, J. J. Carey, DLC, to D. G. Eisenhut, NRC, November 4, 1983.

3 Letter, E. J. Woolever, DLC, to D. G. Eisenhut, NRC, March 30, 1984.

Deaver Valley 2 SSER 2 2

Appendix L

APPENDIX M STAFF SAFETY EVALUATION REPORT ON CONFORMANCE TO GENERIC LETTER 83-28, ITEMS 3.1.3 (REACTOR TRIP SYSTEM COMPONENTS)

AND 3.2.3 (ALL OTHER SAFETY-RELATED COMPONENTS), POSTMAINTENANCE TESTING REQUIREMENTS IN TECHNICAL SPECIFICATION THAT COULD DEGRADE SAFETY INTRODUCTION AND

SUMMARY

Generic Letter 83-281 describes intermediate-term actions to be taken by licens-ees and applicants to address the generic issues raised by the two anticipated transient without scram (ATWS) events that occurred at Unit 1 of the Salem Nuclear Generating Station.

This report is an evaluation of the responses submitted by Duquesne Light Company (DLC), the applicant for the Beaver Valley Power Station, Unit 2, for Items 3.1.3 and 3.2.3 of Generic Letter 83-28.

The actual documents reviewed as part of this evaluation are listed in the references at the end of this appendix.

The requirements for these two items are identical, with the exception that Item 3.1.3 applies these requirements to the reactor trip system components and i

Item 3.2.3 applies them to all other safety-related components.

Because of this I

similarity, the responses to both items were evaluated together.

l REQUIREMENT i

Licensees and applicants shall identify, if applicable, any postmaintenance test requirements in existing Technical Specifications which can be demonstrated to degrade rather than enhance safety.

Appropriate changes to these test require-ments, with supporting justification, shall be submitted for staff approval.

EVALUATION l

The applicant for Beaver Valley Unit 2 responded to these requirements with sub-mittals dated November 4, 19832 and June 16, 19863 The applicant stated in the first submittal that for Unit 1 there were no postmaintenance testing require-l ments in Technical Specifications for either reactor trip system or other safety-l related components which degraded safety, and confirmed in the latter submittal I

that the conclusions reached in the first submittal also applied to Unit 2.

CONCLUSION On the basis of the applicant's statement that no postmaintenance test require-l ments were found in Technical Specifications that degraded safety, the staff l

finds the applicant's responses acceptable for Items 3.1.3 and 3.2.3 of Generic Letter 83-20.

Deaver Valley 2 SSER 2 1

Appendix M L

REFERENCES 1NRC letter, D. G. Eisenhat to all licensees of operating reactors, applicants for operating license, and holders of construction permits, " Required Actions Based on Generic Implications of Salem ATWS Events (Generic Letter 83-28),"

July 8, 1983.

2 Letter, J. J. Carey, DLC, to D. G. Eisenhut, NRC, November 4, 1983.

I Sletter, J. J. Carey, DLC, to H. R. Denton, NRC, June 16, 1986.

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L Beaver Valley 2 SSER 2 2

Appendix M

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Safety Evaluation Report Related to the Operation of Beaver Valley Power Station, Unit 2

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Office of Nuclear Reactor Regulation

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Technical Same as 7. above

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l Docket No. 50-412

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Supplement No. 2 to the Safety luation /eport for the application filed by Duquesne Light Company, et al., for lice e to operate the Beaver Valley Power i

Station, Unit 2 (Docket No. 50-412), lo ted in Beaver County, Pennsylvania, has been prepared by the Office of Nu r Reactor Regulation of the Nuclear Regulatory Commission. The purpose of s supplement is to update the Safety Cvaluation of (1) additional informat n ubmitted by the applicants since Supplement No. I was issued, and (2) patte% that the staff had under review when Supplement No. I was issued.

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UNITED STATES

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NUCLEAR REGULATORY COMMISSION Postaose nsseuo WASHINGTON, D.C. 20666 ofl7,*e -

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