ML20135C055

Method of Analysis & Safe Shutdown Philosophy, 10CFR50, App R Compliance Review

ML20135C055
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Site:	Summer
Issue date:	09/04/1985
From:	SOUTH CAROLINA ELECTRIC & GAS CO.
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NUDOCS 8509110271
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l SOUTH CAROLINA ELECTRIC AND GAS CO.

V. C. SUMMER NUCLEAR STATION 10 CFR 50 APPENDIX R COMPLIANCE REVIEW METHOD OF ANALYSIS AND SAFE SHUTDOWN PHILOSOPHY Enclosure to letter dated September 4,1985, from Mr. O.W. Dixon, Vice President, Nuclear Operations, SCE&G, to Mr. Harold R. Denton, Director, Office of Nuclear Reactor Regulation 8509110271 850904 ,

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TABLE OF CONTENTS 1.0 Introduction 1.1 Basis For This Report 1.2 Section Cross-Reference 2.0 General Methods of Analysis 2.1 Background and Development

'2.2 Analysis Technique 2.3 Success Tree Logic 2.4 Fire Area / Zones 2.5 Detailed Analysis 2.6 Conclusion 3.0 Chanaes in Safe Shutdown Eauipment List 3.1 Basis for Changes 3.2 Equipment Deleted 3.3 Equipment Added 4.0 Method of Analysis for Sourious Components 5.0 Philosophy on the Use of Local Control Switches (" Fire Switches")

5.1 Intended Use 5.2 Criteria and Design 5.3 List of Equipment for Local Control Switches and jumper Procedures 5.4 Comparison to Previous Submittal 6.0 Response to Concerns on IE Bulletin 85-09 7.0 Extent of Local Control for Fires not Reauirina Evacuation of the Control Room 8.0 Philosophy on Controllino Sourious Operation of Valves 8.1 General 8.2 Controf of SpecificValves I 8.3 Conclusion 4

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TABLE OF CONTENTS (CONT'D) 9.0 Circuits Associated By Common Enclosure I 9.1 General  !

9.2 Technique 9.3 First Report 9.4 Second Report 9.5 Conclusion 10.0 Circuits Associated Py Common Power Supply i

10.1 General 10.2 Objective 10.3 Analysis 10.4 Conclusion 11.0 Response to the Concern of NRC Bulletin in 84-09 Requirina Direct Readina Leve! Instruments on Tanks 12.0 Time Critical Safe Shutdown Functions and Time To Cold Shutdown 12.1 Introduction 12.2 Primary System Concerns 12.3 Secondary System Concerns 12.4 Time Line 12.5 - Cold 5hutdown 13.0 Emeroency Lichtina 14.0 Schedule for Completion of Modifications and Compensatory Action ,

to be Used Until Installation .

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1.0 INTRODUCTION

1.1 Basis For This Report During the -onduct of the Appendix R audit of the V. C. Summer Nuclear Station

~ Oune 3-7,1985), a number of concerns were expressed by the NRC auditors, as stated in audit report number 50-395/85-26. Subsequently, some of the concerns were discussed in a meeting on July 24 and 25,1985, between NRR, Region ll, and SCE&G. SCE&G has prepared this report to respond to those concerns discussed in the meeting and to provide sufficient information for both NRR and Region il to more completely evaluate the SCE&G programs.

The concerns are in a variety of areas, including the method of analysis used to identify potential deviations, a comparison of the FPER list of equipment with the list of equipment developed during the current review, the method of analysis of the spurious components, philosophy on the use of local fire switches, philosophy on preventing spurious opening of valves, a review of " associated circuits," responses to NRC IE bulletin 85-09, and a time estimate to achieve cold shutdown. For each concern, a conclusion has been reached which represents current SCE&G response to the expressed concern.

This setof responsesis being provided in letter form to allow the NRC and,in particular NRR the opportunity for timely resolution. The same material content will appear in a proposed revision to the V. C. Summer Nuclear Station FPER, currently scheduled for the second quarter of1986.

1.2 Section Cross-Reference Because of the length and detail in the audit report,395/85-26, and the wide ranging discussions that occurred during the July 24-25, meeting between SCE&G, NRR, and Region ll Staffs, a cross-reference between the audit, the meeting minutes, and this report is being provided in Table 1.2. This cross-reference is intended to aid the staff in its review and to provide a common basis for structuring its responses.

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r O O TABLE 1.2 SECTION CROSS-REFERENCE APPENDIX R SECTION IN JULY 24-25,1985 NRR AUDIT SECTION THIS REPORT MEETING REQUEST 395/85-26 2,3,4,5,7,8,9,10 Associated circuit analysis, revised plant shutdown j scheme 395/85-26-02 5,6,8 Fuses,lif ted leads, disconnects 395/85-26-07 9 Sampling methodology for common enclosure 395/85-26-08 11 Revised plant shutdown scheme 395/85-26-09 12 Summary of times and steps to clarify the timeline 395/85-26-10 7 Identification of areas that are local control areas 395/85-26-12 13 Emergency 8-hour lighting units 395/85-26-16 Separateletter Protectionof thestructural steel of the "M" board 14 Non scheduled outage modification performance, non-outage work schedule Separateletter Steam PORV modifications Separate letter Th/Tc modification 1-2

2.0 GENERAL METHOD OF ANALYSl5 (Audit Report item 395/85-26-01, Ref. 5.a. (1))

2.1 Backaround and Development As a result of NRC supplemental guidance in regard to 10 CFR 50, Appendix R, and other interpretive documents relating to Appendix R, SCE&G conducted a re-evaluation of the plant in order to show compliance with the supplemental guidance, identify all potential deviations from the supplemental guidance, and take the appropriate corrective action. The following discussion describes the method used for this analysis.

Due to the complex nature of this re-evaluation effort, special procedures were developed for each major activity prior to performance of the actual work. This ensured control of the work and provided a common base of understanding for all parties. It also provided a solid historical basis for future design control.

The re-evaluation effort was carried out in a carefully documented program that ensured that appropriate quality assurance considerations were employed. All proposed plant modifications will be made af ter an independent 10 CFR 50.59 analysis has been performed.

2.2 Analysis Technique The re-evaluation effort began in the summer of 1984. The first effort was for SCE&G engineers to independently evaluate plant systems against the 10 CFR 50, Appendix R, requirements for shutdown and to determine the systems needed to facilitate shutdown.

Then each system was evaluated to determme what mechanical components (pumps, valves, etc.) and instrument transmitters were needed for shutdown. This work culminated in a Mechanical Equipment List. After extensive review by SCE&G engineers and G/C, Inc., a Master Mechanical Equipment List was developed which formed the basis for the start of the analysis. An Electrical Equipment List was then developed, based on the Master Mechanical Equipment List, itemizing the required electrical equipment (switchgear, MCC's, etc.) needed to operate the identified mechanical components.

The next step was to decide upon the analysis technique to be used. SCE&G determined that two separate analyses were required to adequately determine the ability of the plant to achieve safe shutdown. These analyses were defined as the Compliance Review (CR) and the Normal Control Review (NCR) scenarios. The CR scenario postulates a fire in each area of the plant including the relay room, cable spreading rooms, or control room. Shutdown would tnen be accomplished using only essential equipment with no restriction on the use of manual operation of components such as valves. Equipment included in this evaluation would be protected against the consequences of a fire if it were needed for safe shutdown for that particular fire location. The NCR scenario postulates a fire anywhere in the plant that does not require evacuation of the control room. Shutdown would be accomplished using the normal control room controls for remote actuation of components.

The intent of the analysis was to determine any particular control schemes which could be degraded due to a fire in a particular location. For these locations, the operating procedure could direct the operator to use local manual control of the equipment. It is SCE&G's position that the CR satisfies the requirements of Appendix R and that the use of local control, at the level of MCC's or the actual component in the case of valves,is not a divergence from the regulation. The NCR scenario is an attempt by SCE&G to enable a more orderly shutdown from the control room when possible. SCE&G believes it is desirable to maximize the control of equipment from the control room for non control room fires, but does not see a l

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l requirement for backfits that are intrusive to safety systems when well reasoned procedures utilizing limited amounts of local control will give assurance that shutdown can be accomplished. Specifically, local control is not considered a " repair" within the context of Appendix R. Aiso, " free from fire damage

• refers to the function of a component or train, not to the ability to have convenient control from the control room. This position on the use of local control is the same as was described and accepted in SSER #3, pp. 9-11, dated Jan.1982.

2.3 SuccessTree Loaic in parallel to the development of the Electrical Equipment List, the Master Mechanical SuccessTreeswere being developed at SCE&G. These success trees define all of the components necessary to achieve safe shutdown with each train of equipment, while showing the various alternative systems and cross-connect possibilities. Each component has its own logic block which details the cable and all additional " support" and " supplemental" equipment needed for operation. Support equipment provides functions such as cooling, air, power, cooling water, etc. Supplemental equipment defines main control board panels, local control panels, termination cabinets, relay panels, etc., required for the function of the required equipment in order to define the complete success tree.

It should be emphasized that these success trees are logic diagrams, as opposed to flow diagrams. Thus, two normally open valves in series, which were required to remain open, were shown as series blocks on the logic diagram. Conversely, two valves in series which were normally closed, atleast one of which needed to remain closed for safe shutdown, were shown as parallel items on the logic diagrams. The Master Mechanical Success Trees show only the logic for mechanical components and instrument transmitters.

2.4 Fire Area / Zones The next step was to update the Fire Protection Evaluation Report (FPER) drawings which had previously identified fire areas. The update consisted of giving the areas and zones unique identifiers and more clearly defining boundaries, particularly for newly defined zones.

2.5 Detailed Analysis Once the Mechanical Equipment List, Electrical Equipment List, Master Mechanical Success Trees,and Fire Area / Zones were defined, the identification of required cables and support equipment was begun. Each mechanical and electrical component identified on the mechanical and electrical equipment lists was analyzed to determine all required cabling and support equipment. Equiomentidentified as needed for operation of a safe shutdown component and which was not previously identified in the mechanical or electrical equipment lists were then categorized as supplemental equipment. Cables were traced and categorized. Spurious equipment and cables were included in the analysis and were treated the same as required equipment (see Section 4.0 for a detailed discussion). Aporopriate data were input to the computer to create the various computer sorts which were then used as input to the success tree analysis. These sorts identified all safe shutdown equipment and required cables per fire area (CR scenario) and fire zone (NCR scenario).

The identified support and supplemental equipment along with the previously identified electrical and mechanical equipment constituted the basis for the development of the composite equipment list. This list ioentifies all equipment necessary to bring the plant to hot and cold shutdown and is categorized by whether it is needed for the CR scenario and/or the NCR scenario. A line entry in the composite equipment list readily reveals to which 2-2

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scenario (CR and/or NCR), to which supplement equipment, and to which phase of shutdown (initiate hot shutdown, maintain hot shutdown, and achieve cold shutdown) each item of safe shutdown equipment applies. This composite equipment list also contains other usefut information, such as equipment description, operation status, control location, and type.

Simultaneously with the preparation of the Composite Equipment List, the Composite Success Trees were developed. All identified support and supplemental equipment were inserted into the Master Mechanical and Electrical 5uccess Trees to develop the Composite Success Trees for both the CR and NCR scenarios.

The next effort in the analysis was to analyze the success trees for each fire area and zone in

order to determine whether one train of equipment would remain free from fire damage for b fire in that particular area or zone. This was done by taking copies of the success tree diagram and marking one copy for each fire area. When either a component, or a cable supporting a component, or a function shown elsewhere on the success tree was identified as being affected by the fire in the area, that component was marked as unavailable.

After the success tree was marked in this manner for a given fire area, it was reviewed to determine whether an unaffected path existed which could be used for safe shutdown. If an unaffected path was found,it was concluded that one train of equipment was free from fire damage; if no path was found,it was necessary to analyze the diagram to determine the cause of the system failures. These failures were defined as " potential deviations." These

" potential deviations" were individually analyzed, given a unique identifier, and documented by fire area. The point in the success tree diagram was annotated by circling the diagram, and the apparent solution was recorded in notes developed during the identification of potential deviations task.

The resolution of these potential deviations involved a more complete evaluation. This evaluation included walkdowns of fire areas to establish the exact location of components and cables, the existence of intervening combustibles, and the adequacy of the existing detection and suppression systems. Each potential deviation was studied in detail and an appropriate resolution was proposed. In many cases, these resolutions consisted of taking credit for existing one-hour-rated conduit and cable tray wraps, the use of existing repair procedures for cold shutdown, or the recognition of existing spatial separation with limited intervening combustibles. Each proposed resolution to a potential deviation was reviewed in detail prior toits final signoff.

Equipment that needed modification to ensure or enhance compliance was identified during the analysis and documented. Requests for plant modifications were developed from these documents and described in the May 29,1985, letter from SCE&G to Mr. H. R. Denton, N RR.

The next step was the development of Plant Emergency Procedures to bring the plant to stable hot shutdown and cold shutdown conditions. FEP-1.0 and FEP-1.1 are the procedures to be used in the event of a control room evacuation and to bring the plant to cold shutdown within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Additional procedures are being developed to provide specific direction for shutdown from the contr,ol room in the event of a major fire in general plant areas. These procedures are being based on the results of the analysis including the identified resolutions of potential deviations. These procedures will be revised once all plant modifications have been installed.

The last step was to develop exemption requests based on the identified resolutions of potential " deviations" as described in the May 29,1985, letter to Mr. H. R. Denton, NRR.

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2.6 Conclusion SCE&G considers that this method of analysis and the resulting documentation provides an adequate demonstration of the compliance of V. C. Summer Nuclear Station to the criteria of 10 CFR 50, Appendix R.

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3.0 CHANGES IN SAFE SHUTDOWN EQUIPMENT LIST (Audit Report item 395/85-26-01, Ref. 5.a (1))

3.1 Basis for Chanaes As' described in Section 2.0, one of the initial tasks in the reanalysis effort was to completely review the systems and equipment that should be used at V..C. Summer Nuclear Station for safe shutdown in the event of a major fire. As a result of this effort, a number of changes have been made to the list of equipment necessary for safe shutdown as presently published in the Fire Protection Evaluation Report (FPER), pages Q1-15 through Q1-22. The significant changes are described in paragraphs 3-2 and 3-3 below. A complete revised equipment list which includes mechanical equipment used for safe shutdown along with instrumentation and supporting electrical equipment will be included in a revision to the FPER, currently scheduled for the second quarter of 1986.

3.2 Ecuipment Deleted A comparison of equipment in the new equipment list and equipment list presently in the FPER has shown that the major change to the FPER involves equipment in the chemical and volume control system. Boration was previously to have been accomplished using boric acid tanks, pumps, and associated valves. Equipment formerly listed for boration will either be deleted (boric acide tanks and pumps) or secured to prevent spurioius operation (valves at

~ the volume control tank). Equipment forletdown has also been deleted since this is no longer considered to be required. Boration required for safe shutdown will now be accomplished by taking makeup water from the Refueling Water Storage Tank (RWST),

which provides a source of makeup water with 2000 ppm concentration of boric acid to the suction side of the charging pumps.

The appropriate section of the FSAR is excerpted below to demonstrate that the charging pump suction can be aligned to the RWST to provide boration makeup for safe shutdown:

9.3.4.3.1 Reactivity Control "An adequate quantity of boric acid is also available in the refueling water storage tank to achieve cold shutdown." "As backup to the normal boric acid supply, the operator can align the refueling water storage tank outlet to the suction of the charging pumps."

. By calculation,5CE&G has shown that the RWST at 2000 ppm Boric Acid, when being used to make up for shrinkage and leakage, provides sufficient shutdown margin to go to cold shutdown.

Table 3.2 provides a list of equipment which has been deleted from the list of safe shutdown equipment as published in the FPER. Notes to the table explain the basis for the deletions.

3.3 - Fouipment Added New equipment was also added to the safe shutdown equipment list for a variety of reasons including the following:

1. Equipment for which spurious actuation could be detrimental to safe shutdown, but operation of which is not required for safe shutdown.

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2. Equipment to facilitate local control

3. Alternative process instrumentation

4. Alternative mechanical equipment (backup D.G. Fuel Oil Transfer Pumps).

5. Equipment (valving) for cross-connecting " swing" channel equipment.

6. Process valving which was implied by association to equipment and, therefore, not previously explicitly listed.

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s TABLE 3.2 EQUIPMENT DELETED FROM FPER LIST OF SAFE SHUTDOWN EQUIPMENT COMPONENT FU NCTION NOTES XVG8104-C5 Boric Acid Tank Line Valves to Chg. Pps. 1 XVD8331-C5 Boric Acid Tank Line Valves to Chg. Pps. 1 XVD8329-C5 Boric Acid Tank Line Valves to Chg. Pps. 1 XVD8323A-C5 Boric Acid Tank Line Valves to Chg. Pps. 1 XVG83238-C5 Boric Acid Tank Line Valves to Chg. Pps. 1 XPP-13A-C5 Boric Acid Transfer Pump . 1 XPP-138-C5 Boric Acid Transfer Pump 1 XVT8152-C5 Letdown Isolation Valve 2 XVB-3110A-5W Reactor Bldg. Cooling Unit to Ind, Cooler 3 Isolation Valve XVB-31108-5W - Reactor Bldg. Cooling Unit to Ind. Cooler 3 Isolation Valve XVB-31,11 A-5W Reactor Bldg. Cooling Unit to Ind. Cooler 3 Isolation Valve ,

XVB-3111B-5W Reactor B!dg. Cooling Unit to Ind. Cooler 3 Isolation Valve XVB-3112A-5W Reactor Bldg. Cooling Unit to Ind. Cooler 3 Isolation Valve

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XVB-31128-5W Reactor Bldg. Cooling Unit to Ind. Cooler 3 Isolation Valve PT-455 PZR Press. 4 PT-455A - PZR Press. 4 PT-456 PZR Press 4 LT-161 Boric Acid Tank Level 1 LT-161 A - Boric Acid Tank Level l' LT-163 Boric Acid Tank Level 1 LT-163A Boric Acid Tank Level 1

- LT-106 Boric Acid Tank Level 1 LT-168 Boric Acid Tank Level 1 3-3

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COMPONENT FUNCTION NOTES 4

NI-36A Intermediate Range Nuclear instrument 5 LT-112 - VCT Level 5 I

LT-115 VCT Level 5 l

LT-470 PZR Relief Tank Level 5 LT-470A PZR Relief Tank Level 5 i

FT-150 Low Press Letdown Flow 5 FT-150A Low Press Letdown' Flow 5 FT-122 Charge Flow 6 FT-122A Charge Flow 6 PT-121 Charging Pressure 6 PT-121 A Charging Pressure 6

, FT-110 Emergency Boration 1 FT-110A ' Emergency Boration 1 TE-9201 Reactor Building Temp. 6 TE-9203 Reactor Building Temp. 6 NOTES FOR TABLE 3 2

1. Boric acid system not required. RWST is used both as a water source for primary shrinkage and to borate to cold shutdown.

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2. Letdown not required for safe shutdown in the event of a fire.

j- 3. Isolation of Industrial Cooling not required when using service water for containment heat removal.

4. RCS wide range pressure transmitters (PT-402,402A,403 and 403A) provide equivalent indication.

5. Instruments not required because associated components are not used.

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6. Instruments not required to monitor system operation; manual operation surveillance will be used.

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O s 4.0 METHOD OF ANAL.YSIS FOR SPURIOUS COMPONENTS (Audit Report item 395/85-26-01, Ref. 5 a. (1))

In the early stages of the analysis SCE&G engineers determined which mechanical components were classified as " spurious" and so annotated the mechanical equipment list.

Spurious operation is defined as inadvertent equipment operation such that safe shutdown may be adversely affected. Valves were classified as spurious if the normal position and safe shutdown position were the same, such that a fire-induced " hot short" would drive the valve into the non-safe shutdown position. Valves were classified as " required" if the normal position and safe shutdown position were not the same, such that the valve must change position to facilitate shutdown. Electrical switchgear breakers were classified as spurious if inadvertent tripping or opening was detrimental to safe shutdown.

Cables associated to the identified spurious equipment were then analyzed to determine if a credible cable failure could cause inadvertent operation. An evaluation was made of the various possible credible cable failure modes, and this evaluation was used as a basis for evaluating the classification of individual cables.

Individual conductors on control elementary diagrams were analyzed with the emphasis on ,

' " hot shorts." A " hot short" is actually the contact of a control conductor (under consideration) of one cable with a hot conductor in a second control cable (without grounding) such that a current path is established which could energize electrical devices (relays, solenoids, etc.). Energizing of these devices could in turn cause the spurious operation of the safe shutdown equipment. It should be noted that these hot shorts are 4 actually not short circuits, since only normal current is postulated to flow in lieu of a large I fault current as is normally associated with short circuits.

These cables were then classified as spurious and documented in the computer sorts of cable / equipment by fire area (CR) and fire zone (NCR), computer generated cable worksheets, and cable identification packages. " Spurious" cables were tr eated the same as

" required" cables during success tree analysis. (See Section 2.5). Cables identified in the identification of potential deviations task were checked against the computer sort > to determine if they were classified as " spurious." An analysis was then made in the resolution of potential deviations task (See Section 2.5) to prevent tnese spurious cables from adversely affecting safe shutdown. Resolutions of potential deviations for spurious cables resulted primarily in requests for plant modificaitons such as installing local control " fire switches,"

wrapping cables with one-hour rated fire retardant material, installing second power disconnects, armoring cable, and upgrading existing local control switches.

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5.0 PHILOSOPHY ON THE USE OF LOCAL CONTROL SWITCHES (" FIRE SWITCHES")

(Audit Report item 395/85-26-01, Ref 5 a (1) and 395/85-26-02 Ref. 5.a. (1))

5.1 Intended Use For fires in the control room, relay room, or cable spreading rooms, the control room may need to be evacuated and the shutdown directed from the control room evacuation panel (CREP). In this evcnt, the shutdown will be accomplished using Train "B" equipment with control of individual equipment accomplished at the CREP, at related switchgear and motor control centers, at the diesel generators, and by local manual operation of valves. To facilitate this shutdown, control switches are being added to the switchgear and motor control centers. Also, some existing local control switches are being upgraded (to prevent spurious operation).

Fires in other plant areas do not require control room evacuation, but in some cases, individual control room controls will be disabled. in these cases, an operator will be dispatched from the control room to operate the equipment using the local control switches or to manually operate valves.

5.2 Criteria and Desian For Train *B", these local control switches (" fire switches") will be provided on equipment to be operated in order to initiate hot standby. Local control switches will also be provided for equipment needed to maintain hot standby, if needed to ensure the timeliness of operation, consistent with the SCE&G timeline, or to ensure operator safety. Permanent jumper procedureswill be provided for equipment needed only to achieve cold shutdown or not needed in the first 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of shutdown.

In each case, these local controls will consist of two switches. A local transfer switch will be installed to isolate cables identified in the analysis as spurious or capable of disabling control.

This transfer switch also connects a second control power supply into the control circuit. A second switch will be provided to start /stop the equipment after isolation takes place. The following regulations will be referenced and shall be considered in the design and testing of thelocal controls: 10 CFR 50, Appendix R, R.G.1.22, R.G.1.47, and NRC generic letter 81-12.

5.3 List of Equipment for local Control Switches and jumper Procedures Table 5.3 lists all equipment where local control and transfer switches (" fire switches") are to be installed, equipment for which jumper procedures are to be used on a permanent basis, and equipment that must only be de-energized via local control without the need of tools and/or materials.

Temporary (interim) jumper procedures will be provided in order to justify continued operation until the local control switches are installed. These procedures perform the same function as the local control switches and in most cases without the need of materials. (i.e.,

manually trip control power breakers to isolate spurious cables, charge the closing spring if neces';ary, and then manustly close the switchgear breaker via the local control button.)

Equipment that must only be de-energized to preclude spurious operation or inadvertent diesel loading falls under the " existing local control" co!umn. Opening the breaker involves no more than manually tripping the control power breakers and then manually pushing the

" trip" button on the switchgear cubicle. This action requires no tools or other materials but 5-1

does require an operational procedure which is incorporated into the Fire Emergency Shutdown Procedures.

5.4 Compar; son to Previous Submittal SCE&G'sletter to Mr. Harold R. Denton, NRR of May 29,1985, included a descnption of proposed modifications for adding local control switches or implementing jumper procedures for a number of components. This list was summarized in the NRC audit report 50-395/85-26, dated August 20,1985, paragraph 3.a., pg. 6. This original list inctuded those components for which local control was contemplated. With this discussion, SCE&G is confirming its intent to install local control switches as indicated in Table 5.3. This table also involves four additional local control switches. Two of these additional switches are covered by Note 7 to Table 5.3. The other two are covered by Note 5.

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i TABLE 5.3 LIST OF EQUIPMENT FOR LOCAL CONTROL SWITCHE5 AND JUMPER PROCEDURES

i. EXISTING PLAN N ED, LOCAL PERMANENT LOCAL CONTROL NOTES EQUIPMENT JUMPER CONTROL (FOR (SEE NEXT TAG NO. DESCRIPTION PROC. SWITCHES TRIPPING) PAGE)

, MFN-978-AH Reactor Bldg. "B" Train

. Cooling Fan X 2

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MFN-97D-AH Reactor Bldg. "B" Train Cooling Fan X 2 l

XEG-1B-DG "B" Diesel Generator X X ~ 1,4 XFN-388-AH Chgr. Room Supply i Fan "B" X 3 XFN-45A-AH DG "B" 50% Supply Fan X 2,7 XFN-458-AH DG "B" 50% supply Fan X 2,7 XFN-468-VL Charging /51 Pump "B" Room Fan X 2,7 XFN-498-VL RHR/ Spray Pump ~B" Fan X 6 XFN-76-VL ESF1DB Switchgear Room Fan X 3 XFN-808-AH SWPH 5upply Fan "B" X 7 XFN-818-VL SW Booster Pump "B" Fan X 3 XFN-1068-VL CREP and "B" Speed Switch Room Supply Fan X 3 2

XFN-133-VL Aux. Bldg. "B" Switchgear Room Fan X 3

.! XHX-1 B-VU "B" Chiller X X 1,4 1

i XPP-1 B-CC CC Pump "B" X 2-

< XPP-318-RH Residual Heat Removal Pump "B" X 6 XPP-398-5W SW Pump "B" X 1 XPP-438-C5 Charging Pump "B" X 2

- XPP-458-5W SW Booster Pump "B" X 2 XPP-488-VU Chilled Water Pump ~B" X 2 4

X5WIDB, U 14 Charging /SI Pump "C" Feeder "B" X 4 i

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EXISTING PLANN ED, LOCAL PERMANENT LOCAL CONTROL EQUIPMENT JUMPER CONTROL (FOR TAG NO. DESCRIPTION PROC. SWITCHES TRIPPING) NOTES X5W-I DA-ES, U 1 Normal Offsite Breaker X 4 i X5W-I DA-ES, U 15 Alt. Offsite Breaker X 4 X5W-IDA-ES, U3 "A" D.G. Breaker X 4 X5W-1 DB-ES, U 1 Alt. Offsite Breaker X 4

, X5W-1 DB-ES, U 16 Normal Offsite Breaker X 4 X5W-1 DB1, U4C Tie Breaker to X5W183 X 4 X5W-1DB-ES, U4 X5W-1 EB-E5 Feeder Breaker X 2 X5W-1DB-ES, U6 FDR Breaker R. B. Spray PP 388 X 4 X5W-1DB-ES, U7 Unit Sub Feeder Breaker X 2 l X5W-1EB-ES, U3 FDR Breaker for Unit 3

Sub IEB1 X 2 X5W-1 DB1-ES, U4B ESF 480 Unit Sub Main Breaker X 2 j X5W-1DB1, U78 FDR Breaker R.B. Fan 968 X 5 X5W-1 DB 1, U7C FDR Breaker R.B. Fan 96D X 5 X5W-I DB2-ES, U48 ESF 480 Unit Sub Main Breaker X 2 X5W-1 EB1-ES, U48 ESF 480 Unit Sub Main Breaker X 2 NOTES FOR TABLE 5.3

1. Existing switch being uograded; needed to maintain hot standby.

2. fJeeded to maintain hot standby. >

3. Not needed during the first 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

. 4. De-energize (open) breaker, no tools needed.

5. De-energize (open) breaker to preclude spurious operation, no tools needed.

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6. Needed for cold shutdown only.-

7. Designated as a future modification, XFN-468-VL and XFN-808-AH are not referenced in the May 29,1985, letter to Mr. H. R. Denton, N RR.

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6.0 RESPONSE TO THE CONCERN OF IE BULLETIN 85-09 (NRR Meeting item and Audit Report item 395/85-26-02 Ref. 5.a. (1))

1. E. Information Notice No. 85-09 addresses the potential deficiencies, specifically redundant fusing,in isolation switch design.

All existing isolation switches and planned modifications (where isolation switches are to be installed) will adequately isolate the circuit and will have (or presently do have) redundant fuses such that operability of equipment is maintained to achieve and maintain hot shutdown in accordance with 10 CFR 50, Section lit.G.1 and IEB 85-09. With the installation of the redundant fusing the equipment can be operated without fuse replacement or circuit breaker resettir.g.

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7.0 EXTENT OF LOCAL CONTROL FOR FIRES NOJ_REOUIRING EVACUATION OF THE CONTROL ROOM (Audit Report item 395/85-26-01, Ref. 5.a. (1), and item 395/85-26-10, Ref. 5.a. (3) (c))

The effect of fires in all fire areas of the plant, including the control room, relay room, and

. cable spreading rooms, has been extensively analyzed. The analysis has emphasized

- shutdown using only essential equipment with no restriction on the use of manual operation of components, such as valves, and has been termed the Compliance Review (CR) scenario.

This philosophy is based on the NRCinternal guidance given in SEC-1-83-269, Attachment C, paragraph b.

' As part of CR, control circuits for "B" Train shutdown equipment, such as fans and pumps, were traced to the control room evacuation panel (CREP), switchgear, and motor control centers. Control circuits for shutdown equipment, such as fans and pumps, powered by the "A" Train were traced to the control room. Control cables for valves susuptible to detrimental spurious operation were also traced. However, other valves, regardless of their

. power source, were not included in the control circuit analysis because they will be operated manually. Where necessary, modifications to the plant equipment and/or operating procedures are being implemented to ensure compliance with the latest interpretations of 10 CFR 50, Appendix R.

For fires in the control room, relay room, and cable spreading rooms which require control room evacuation, shutdown control will be directed from the CREP, with equipment operated from the CREP, from switchgears, or locally. Valves will be manually operated by hand wheels or by isolation and venting of control air.

For fires located in plant areas other than the control room, relay room, and cable spreading rooms, the control room will not be evacuated. Shutdown will be directed from the control room,with the majority of equipment being operated from the main control board. Fires in some areas may render some of the control room controls inoperable for the designated train of safe shutdown equipment, while disabling the alternate train of shutdown equipment. In these cases, CR has demonstrated that the designated train of equipment will remain operable and control can be achieved from the CREP, locally from either switchgear or motor control centers, or by manual operation for valves, as required.

As a part of the Normal Control Review (NCR) scenario analysis, all control circuits were traced to the control room for all shutdown equipment, including pumps, fans, and valves.

When completed, the NCR scenario will identify exactly which shutdown components will have to be locally controlled. Upon completion of this additional analysis, any local control actions identified will be incorporated in the operating procedures.

However, the CR 5cenario has indicated that only fires in a few fire areas would require some components to be manually operated, and this will be a very limited number of components for any one fire area. Table 7.0 is a list of fire areas versus the anticipated extent of local control (electrical and/or manual) based on the results of the CR analysis. An initial version of Table 7.0 was prepared by SCESG during the audit and is summarized on pages 5 and 6 of the audit report. After con pletion of the NCR analysis,a final version of the information presented in Table 7.0 will be incorporated in the FPER.

Based on the CR analysis, the following are specific examples of the' extent of local control where Table 7.0 indicates that " minor local control" will be required:

7-1

_ _ _ m . - . _ . - . _ _ _ _ .. _ _ _ . _. . _ .

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- a. For fire area AB 1, a fire at the 400' elevation would require local manual operation of the RWST suction valves before charging could be restored, since a fire in this area would locally damage valve cabling. All other equipment and valves would be controlled from the main control board.

I b. For fire area 18-25, a fire in the vicinity of the service water booster pump "B" requires .

- the use of service booster pump "A" The power cables for the "A" pump are protected [

I with Kaowool wraps; however, the control cables for a pressure interlock with the main j service water pump are not wrapped. The refore, local manual operation of the pump ,

i breaker, including the use of a jumper prc cedure, will be required for this pump. All '

i other equipment and valves would be co1 trolled from the main control board. ,

c. Similarly, for a fire in the vicinity of the feedwater regulating valves in fire area 18-25, [

valve control cables may be affected. Therefore,in this case,it would be necessary to pull ,

, fuses in the control board for these valves to ensure that they fail open. Feedwater

] would then be supplied from the turbine driven pump using the pump local control.~ All other equipment and valves would be controlled from the main control board.

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TABLE 7.0 LIST OF FIRE AREAS VS. ANTICIPATED EXTENT OF LOCAL CONTROL PREFERRED EQUIPMENT FIRE TRAIN TO ANTICIPATED EXTENT AREA BE USED OF LOCAL CONTROL AB-1 A or B* Control from Control Room, minor local control CB-1 A Control from Control Room, minor local control CB-2 A Control from Control Room, minor local control CB-3 Operator Choice Control from Control Room, no local control (no Appendix R Equipment or Cable)

CB-4 8 Control Room Evacuation, major local control (Cable Spreading Room)

CB-5 B Control from Control Room, minor local control CB-6 B Control Room Evacuation, major local control (Relay Room)

CB-7 Operator Choice Control from Control Room, no local control (no Appendix R Equipment or Cable)

CB-8 Operator Choice Control from Control Room, no local control (no Appendix R Equipment or Cable)

CB-9 Operator Choice Control from Control Room, no local control (no Appendix R Equipment or Cable)

CB-10 8 Control from Control Room, minor local control CB-11 Operator Choice Control from Control Room, no local control (no Appendix R Equipment or Cable)

CB-12 A Control from Control Room, minor local control CB-13 Operator Choice Control from Control Room, no local control (no Appendix R Equipment or Cable)

CB-14 Operator Choice Control from Control Room, no local control (no Appendix R Equipment or Cable)

CB-15 B Control Room Evacuation, major local control (Cable

. Spreading Room)

CB-16 Operator Choice Control from Control Room, no local control CB-17 8 Control Room Evacuation, major local control (Control Room)

• Choice of equipment train depends on the location of the fire within the fire area.

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TABLE 7 0 (CONT'D)

PREFERRED EQUIPMENT FIRE TRAIN TO . ANTICIPATED EXTENT AREA BE USED OF LOCAL CONTROL CB-18 B Control from Control Room, mittor local control CB-19 Operator Choice Control from Control Room, no local control (no Appendix R Equipment or Cable)

' CB-20 A Control from Control Room, minor local control CB-21 Operator Choice Control from Control Room, no local control (no Appendix R Equipment or Cable)

CB-22 B Control from Control Room, no local control CB A Control from Control Room, no local control CB-24 Operator Choice Control from Control Room, no local control (no Appendix R Equipment or Cable)

DG-1 B Control from Control Room, no local control DG-2 A Control from Control Room, no local control FH-1 A Control from Control Room, no local control 18-1 Operator Choice Control from Control Room, no local control (no Appendix R Equipment or Cable) 18-2 B Control from Control Room, no local control 18-3 8 Control from Control Room, no local control IB-4 A Control from Control Room, no local control 1B-5 B Control from Control Room, no local control IB-6 A Control from Control Room, no local control IB-7 A or B* Control from Control Room, no local control IB-8 Operator Choice Control from Control Room, no local controf IB-9 A Control from Control Room, no local control 1B-10 A Control from Control Room, no local control IB-11 A Control from Control Room, no local control IB-12 A Control from Control Room, rio local control 18-13 Operator Choice Control from Control Room, no local control 18-1 4 8 Control from Control Room, no local contros 18-1 5 A . Control from Control Room, no local control 18-1 6 8 Control from Control Room, no local control

-

• Choice of equipment train depends on the location of the fire within the fire area.

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TABLE 7.0 (CONT'D)

PREFERRED EQUIPMENT FIRE TRAIN TO ANTICIPATED EXTENT AREA BE USED OFLOCALCONTROL 18-17 A Control from Control Room, no local control 18-1 8 B Control from Control Room, no local control 18-19 A Centrol from Control Room, no local control 18-2 0 B Control from Control Room, no local controf 18-21 Operator Choice Control from Control Room, no local control 18-22 'A Control from Control Room, no local control IB-23 B Control from Control Room, no local control 18-2 4 - B Control from Control Room, no local control 18-25 A or B* Control from Control Room, moderate local control 18-2 6 B Control from Control Room, no local control M H A or B* Control from Control Room, no local control M H-8 Operator Choice Control from Control Room, no local control (no Appendix R Equipment or Cable)

MH-9 Operator Choice Control from Control Room, no local control (no Appendix R Equipment or Cable)

MH-11 Operator Choice Control from Control Room, no local control (no Appendix R Equipment or Cable)

RB-1 A or B* Control from Control Room, for hot standby, Reactor Building entry required for " Cold Shutdown" SWPH-1 B Control from Control Room, no local control SWPH 2 A Control from Control Room, no Iocal control SWPH-3 A Control from Control Room, no local control SWPH-4 A or B* Control from Control Room, no local control SWPH-5 A or B* Control from Control Room, no local control TB-1 Operator Choice Control from Control Room, very minor local control

' YD-1 Operator Choice Control from Control Room, no local control YD-A . Operator Choice Control from Control Room, no local control

• Choice of equipment train depends on the location of the fire within the fire area.

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8.0 PHILOSOPHY ON CONTROLLING SPURIOUS OPERATION OF VALVES (Audit Report item 395/85-26-01 and 395/85-26 02, Ref. 5.a. (1) and 5.a. (2)(b))

8.1 General The possible consequences of the spurious operation of certain valves require that the spurious operation be prevented or corrected on a priority basis. Of immediate concern are tha reactor coolant system Hi-Lo pressure boundary valves, other valves which can result in loss of reactor coolant inventory, and valves which can result in uncontrolled steam dumping.

For other valves, more time is available for correction of spurious operation.

The reactor coolant system Hi-Lo pressure boundary valves are all 480 volt ac motor operated and cannot spuriously open, since power to the motors has been disconnected during normal plant operation.

The remainder of the valves for which spurious operation must be corrected on a priority basis are air operated and controlled by one or more solenoid valves For valves controlled by a single solenoid, the valve power must be disconnected and the cabling to the solenoid must be protected with a grounded shield (with sufficient short circuit ampacity). This is necessary to prevent a " hot short" from spuriously operating these valves.

Valves controlled by two or more solenoids, where de-energizing any one solenoid puts the valve in the safe position, only require that the valve power be disconnected. The cabling to the solenoids does not require shielding, since two or more " hot shorts" simultaneously would be required to spuriously operate the valve; and for non-reactor coolant pressure boundary valves, multiple hot shorts are not considered credible.

Similarly, for situations involving two normally closed valves in series (with individual solcnoids) where at least one must be kept closed, disconnecting the power to the solenoids is sufficient. Two hot shorts, one to each solenoid, would be required to cause the flow path to be opened.

The required power disconnection will be accomplished in the main control board through a fuse removal procedure. This may appear to be in conflict with the NRC internal guidance provided in SEC-1-83-269, Attachment C, paragraph b. However, the fuses are easily accessible and safely removable; SCE&G considers this equivalent to the operation of switches. A human factors review will be conducted to ensure that the fuses can be easily identified and to ensure the fuse can be removed without tools, or ensure that tools are provided within the control room.

In addition, a secondary means of disconnecting the solenoid power will be provided ;n a separate fire area, for use in the unlike!y event a fire occurs in the control room and requires immediate evacuation. This secondary means of disconnecting power consists of switches which will be installed in the termination cabinets located in the cable spreading room, which is a separate fire area from the main control room. Human factors will also be considered in the design of these switches, to ensure that they can be readily located and opened. In the interim, a wire disconnection procedure equivalent to the switches has been developed (FEP-1.0, Attachment X).

For motor operated valves where suf ficient time is available, spurious operation will be controlled by opening the cubicle breaker in the MCC and then manually repositioning the valve locally by operating the hand wheel. The fire emergency safe shutdown procedure, 8-1

FEP-1.0, directs the tripping of the MCC cubicle breakers in a timely manner and provides separate instructions to manually reposition valves for which spurious repositioning could be detrimental to safe shutdown.

8.2 Control of Specific Valves The following is a detailed description of the method of controlling spurious action for each of the valveswhere time constraints prohibit the use of local manual operation. Table 8.2 also provides a listing of these valves, including notations indicating which are reactor coolant system Hi-Lo pressure boundary valves, and which valves require plant modification as described in t he May 29,1985, letter to Mr. Harold Denton, NRR:

1. The Reactor Vent isolation and the Residual Heat Removal inlet Isolation valves, that have been categorized as Hl-LOW reactor coolant pressure boundary valves are three phase 480 VAC motor operated. Three phase to three phase hot shorts of the proper sequence in the motor power supply circuit of these valves are not considered credible.

The power supply breakers of these valves are kept open and locked during normal plant operation. Thus spurious operation of these valves cannot occur.

7

2. The pressurizer PORV's are single solenoid / air operated valves Spurious activation of the solenoid could result in loss of reactor coolant inventory. These valves will be fully protected from spurious signals, by installing shielded control cables to their solenoids and by disconnecting the power to those solenoids as described in 81.

3. The pressurizer Spray Header Isolation valve is located inside the reactor building and is controlled by a single solenoid. Spurious operation of this valve after charging has been i restored could result in reactor coolant system depressurization. Thelocation of the i valve would require an excessive amount of time for manual correction of spurious actuation. Therefore, spurious actuation will be controlled prior to starting the charging pump, by shielding the solenoid cable and power disconnect as described in 8.1 above.

4. The Main Steam Isolation and the Main Steam isolation Bypass valves are controlled by

two solenoids and fait closed upon loss of power to either of the solenoids, even if the l operator controlling solenoid is spuriously energized. Therefore, the use of the power disconnects described in 8.1 is sufficient to prevent these valves from opening.

! 5. The steam generator PORV's provide the steam dump function for normal plant operation and are controlled by six solenoids each. Two or more solenoids must be spuriously activated to open the valve. Their spurious operation will be controlled in the same manner as described in Paragraph 4 above, upon completion of a control logic change which is being proposed to the NRC staff in a separate report. (Note: This is a change from the modifications described in the May 29,1985 letter to Mr. Harold R.

Denton.)

6. The pair of Excess Letdown Isolation valves and the pair of Letdown isolation valves are installed in series and closure of either one valve of the pair accomplishes the required isolation. Thus, although single solenoid, these valves require no shielding of cables and

, will be " failed" closed as described in 8.1.

7. The Normal and Alternate Charging Header isolation valves are each operated by a single l solenoid and normally fail open. Since these valves are in separate charging paths, spurious closure of one valve will not prevent charging. Thus, a single hot short cannot i

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FEP-1.0, directs the tripping of the MCC cubicle breakers in a timely manner and provides separate instructions to manually reposition valves for which spurious repositioning could be detrimental to safe shutdown.

8.2 Control of Specific Valves The following is a detailed description of the method of contro ling spurious action for each of the valves where time constraints prohibit the use of local manual operation Table 8.2 also provides a listing of these valves, including notations indicating which are reactor coolant system Hi-Lo pressure boundary valves, and which valves require plant modification as described in t he May 29,1985, letter to Mr. Harold Denton, NRR:

1. The Reactor Vent isolation and the Residual Heat Removal Inlet Isolation valves, that have been categorized as Hi-LOW reactor coolant pressure boundary valves are three phase 480 VAC motor operated. Three phase to three phase hot shorts of the proper sequence in the motor power supply circuit of these valves are not considered credible.

The powar supply breakers of these valves are kept open andlocked during normal plant operation. Thus spurious operation of these valves cannot occur.

2. The pressurizer PORV's are single solenoid / air operated valses. Spurious activation of the solenoid could result in loss of reactor coolant inventory. These valves will be fully protected from spurious signab, by installMg shielded control cables to their solenoids and by disconnecting the power to the .>lenoids as described in 81.

3. The pressurizer Spray Header Isolation valve is located inside the reactor building and is controlled by a single solenoid. Spurious operation of this valve after charging has been restored could result in reactor coolant system depressurization. The location of the valve would require an excessive amount of time for manual correction of spurious actuation. Therefore, spurious actuation will be controlled prior to starting the charging pump, by shielding the solenoid cable and power disconnect as descnbed in 8.1 above.

4. The Main Steam isolation and the Main Steam isolation Bypass valves are controlled by two solenoids and fait closed upon loss of power to either of the solenoids, even if the operator controlling solenoid is spuriously energized. Therefore, the use of the power disconnects described in 8.1 is sufficient to prevent these valves from opening.

5. The steam generator PORV's provide the steam dump function for normal plant operation and are controlled by six solenoids each. Two or more solenoids must be spuriously activated to open the valve. Their spurious operation will be controlled in the same manner as described in Paragraph 4 above, upon completion of a control logic change which is being proposed to the NRC staff in a separate report. (Note: Thisis a change ftom the modifications described in the May 29,1985 fetter to Mr. Harold R.

Denton.)

6. The pair of Excess Letdown isolation valves and the pair of Letdown isolation valves are irutalled in series and closure of either one valve of the pair accomplishes the required isolation. Thus, although single solenoid, these valves require no shielding of cables and will be "farled" c'osed as described in 8.1.

7. The Normal and Alternate Charging Header Isolation valves are each operated by a single solenoid and normally f ail open. Since these valves are in separate charging paths, spurious closure of one valve w;ll not prevent charging. Thus, a single hot short cannot 8-2 6

3

prevent charging. Therefore, power disconnects as described in 8.1 will be sufficient to ensure charging.

8.3 Conclusion SCE&G is confident that the methods described above will control the spurious operation of valves in a safe and timely manner such that the plant will not reach an unrecoverable condition in the event of a major fire.

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TABLE 8 2 LIST OF VALVES (FOR WHICH SPURIOUS OPERATION CANNOT BE CORRECTED BY LOCAL MANUAL ACTION)

Hi-LO INCLUDED VALVE PRESSURE METHOD TO CONTROL IN 5/29/85 EOUIPMENT NO. DESCRIPTION BOUNDARY SPU RIOUS OPERATION LETTER TO NRR XVT 8095A-RC Reactor Vent isolation Valve X MCC Breakerlocked open Note 1 XVT-80958-RC Reactor Vent isolation Valve X MCC Creaker locked open Note 1 XVT 8096A RC Reactor Vent isolation Valve X MCC Breakerlocked open Note 1 XVT-80968-RC Reactor Vent loslation Valve X MCC Breaker locked open Note 1 XVG-8701 A-RH RHR LP1 Inlet Isolation Valve X MCC Breaker locked open Note 1 XVG-87018-RH RHR LP2 Inlet isolation Valve X MCC Breaker locked open Note 1 XVG-8702A-RH RHR LPI Inlet isolation Valve X MCC Breaker locked open Note 1 XVG-87028-RH RHR LP2 Inlet isolation Valve X MCC Breaker locked open Note 1 IPV-444B-RC Pressurizer PORV Note 2 Pull Fuses or Open 2nd Power Disc. X and Shield Cables IPV-445A-RC Pressurizer PORV Note 2 Pull Fuses or Open 2nd Power Disc. X and Shield Cables IPV-4458-RC Pressurizer PORV Note 2 Pull Fuses or Open 2nd Power Disc. X and Shield Cables XCT-8145-C5 Aux. 5 pray Line Isolation Valve Pull Fuses of Open 2nd Power Disc. X j and Shield Cables IPV-2000-M5 Main Steam PORV - Note 3 Pull Fuses or Open 2nd Power Disc. X IPV-2010-M5 Main Steam PORV - Note 3 Pull Fuses or Open 2nd Power Disc. X 8-4

TABLE 8.2 (CONT'D)

Hi-LO INCLUDED VALVE .

PRESSURE METHOD TO CONTROL- IN S/29/85 EOUIPMENT NO. DESCRIPTION BOUNDARY SPURIOUS OPERATION LETTER TO NRR iPV-2020-MS . Main Steam PORV - Note 3 Pull Fuses or Open 2nd Power Disc. X XVM-2801 A-MS Main Steam LP 'A' isolation Valve Pull Fuses or Open 2nd Power Disc. .X XVM-2801B-MS Main Steam LP 'B' isolation Valve Pull Fuses or Open 2nd Power Disc. X XVM-2801C-MS Main Steam LP 'C' isolation Valve Pull Fuses or Open 2nd Power Disc. X XVT-2869A-MS Main Steam 'A' Bypass isolation Pull Fuses or Open 2nd Power Disc. X Valve -

XVT-28698-MS . . Main Steam 'B' Bypass isolation Pull Fuses or Open 2nd Power Disc. X Valve XVT-2869C-MS Main Steam 'C' Bypass isolation Pull Fuses or Open 2nd Power Disc. X

- Valve

. XVT-8146-CS - Normal Char. Header Isolation Pull Fuses or Open 2nd Power Disc. Note 4 -

Valve XVT-8147-CS Aux. Char. Header isolation Pull Fuses or Open 2nd Power Disc. Note 4 Valve XVT-8153-CS Excess Letdown isolation Valve Pull Fuses or Open 2nd Power Disc. X

XVT-8154-CS Excess Letdown isolation Valve Pull Fuses or Open 2nd Power Disc. X ILV-459-CS Letdown Line Isolation Valve - Pull Fusesin MCB orin CREP Note 1

~ Note S ILV-460-CS Letdown Line isolation Valve- Pull Fusesin MCB or in CREP Note 1 Note 5 8-5

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I NOTES FOR TABLE 8 2 4

1. These valves were not discussed in 5/29/85 f etter, since no plant modifications are required.

l 2. SCE&G is treating the pressurizer PORV's as reactor coolant system Hi-Lo pressure boundary valves, since spurious operation could result in loss of reactor coolant inventory.

3. SCE&G plans to revise the control logic for the main steam PORV's so that two independent hot

,' shorts would be required for spurious actuation.

4. These valves were added to the list of valves requiring power disconnects, based on a review of the analysis completed subsequent to the 5/29/85 letter to Mr. Harold R. Denton, NRR.

5. There is sufficient time before spurious actuation would lead to an unrecoverable condition to permit the control room evacuation panel (CREP) to be used as a secondary power disconnection

location.

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9.0 CIRCUITS ASSOCIATED BY COMMON ENCLOSURE (Audit Report item 395/85-26-07, Ref. 5.a. (2) (c) and 395/85-26-01, Ref. 5.a. (1))

9.1 General in NRC generic letter 81-12, one of the concerns raised was the possibility that " cables associated by common enclosure" could be subject to fire-induced faults which would be detrimental to safe shutdown. The concern was that a fire in a specific fire area would be propagated into another fire area by cables lacking adequate overcurrent protection.

Two reports were prepared to demonstrate the adequacy of overcurrent protection. The objective was to demonstrate that adequate overcurrent protection exists for all cables which could be associated by common enclosure.

9.2 Technique Each report contained a detailed discussion of the plant design criteria and practices which provided all plant cabling with adequate overcurrent protection. In order to demonstrate conclusively that these design criteria and practices were followed, each report also

contained an evaluation of the as-built conditions using a statistically valid sampling technique.

The statistical procedure used was suitable to demonstrate a 95% confidence level that 95%

of the power circuits have adequate overcurrent protection. The technique and acceptance criteria are similar to those used to demonstrate the adequacy of pipe supports. The procedure employs a hypergeometric distribution, and the sample size was checked to verify thatitwould be sufficiently large to provide a 95% confidence level. A published table of random numbers was used to select the sample of circuits from a population which was arranged by circuit number in an alphanumeric order by system and sequence (circuit numbers are by system and number designation, i.e., AHJ606X), and then sequentially numbered.

9.3 First Report For the first report, the statistical evaluation used a population of circuits that included all power cables in the plant, Class IE associated, and Non-Class 1E. As a practical matter, only power circuits larger than 10 AWG were considered because smaller circuits are not connected to significant sources of fault current. This population size consisted of 1834 circuits from wnich a random sample of 59 circuits were drawn. Each of the circuits in this

sample was then checked to confirm that overcurrent protection based on the project design parameters was provided for the circuit and that this protection was adequate for the cable size. This protection was documented on a worksheet for each circuit. Each circuit was analyzed to determine that the auto-ignition temperature of the cable could not be reached because the device operating time at the available fault current was shorter than the time to auto-ignition. All the circuits in the sample met the requirements for short circuit protection 4

and long term ampacity. This provides 95% confidence that at least 95% of the circuits in the population are provided with overcurrent protection by design. Values of the circuit breaker rating, trip setting, operating time, cable auto-ignition time, and the conclusion of adequacy are all documented in the report.

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9.4 Second Report For the second report, the statistical evaluation used a population limited to Class I E and 1 E-associated circuits (from the IEEE 384 point of view). Since the safe shutdown systems for the V. C. Summer nuclear system are all safety related, and since Non-Class IE circuits are run in raceways separated from Class IE and1E-associated circuits, the Non-Class IE circuits have no potential of becoming ' associated by common enclosure" for the purposes of Appendix R compliance. A large number of circuits such as turbine building welding circuits which have no relevance to safe shutdown were thus eliminated from consideration by this restriction to Class IE and 1E-associated circuits. The resulting population consisted of 321 power circuits from which a sample of 59 circuits were drawn. The sample circuits were evaluated as described in paragraph 9.3 for short circuit and overload protection. All the circuitsin the sample met the criteria for adequacy, leading to the conclusion that there was 95%

confidence that at least 95% of the circuits in the population are provided with overcurrent protection by design. Values of circuit breaker rating, trip setting, operating time, cable auto-ignition time, and the conclusion of adequacy are all documented in the report.

9.5 Conclusion Based on the results of both analyses, there is verification that the overcurrent design requirements were applied to the plant circuits and that there is a 95% confidence level that 95% of the circuits connected to significant sources of short circuit current are adequately protected against continuous overcurrent and short circuit currents. Consequently, circuits which are associated by common enclosure are adequately protected.

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10.0 CIRCUlTS ASSOCIATED BY COMMON POWER SUPPLY (Audit Report item 395/85-26-01, Ref. 5.a. (1))

.10.1 General Generic letter 81-12 describes a concern that power circuits required for safe shutdown be protected against fire-induced failures of circuits with which they are associated by virtue of the fact that they have a common power source.

T 10.2 Objective Circuits identified as AC and DC power supply circuits for the safe shutdown equipment for

. Appendix R associated by common power supply were analyzed. The equipment reviewed included that required for the " compliance review scenario." The objectof the analysis was to demonstrate that equipment served by a common power supply and thereby " associated" j was protected from loss of the power supply by coordination of the circuit treaker trip settings with the power supply breaker trip settings. This prevents a fire-related short circuit in an " associated circuit" feeder from tripping the main power supply breaker and, thereby, disconnecting power to the " required circuits."

10.3 Analysis The method of the analysis was to first identify the source power circuit breakers and their

" associated" breakers. After that determination, a simple one-line diagram was prepared identifying the frame size and trip setting of each breaker. From this data and the manufacturer circuit breaker trip characteristic curves, coordination curves were prepared to demonstrate visually the amount of coordination existing between the associated circuit breakers. A complete report of this coordination study was prepared and made part of the Appendix R review documentation.

10.4 Conclusion The results of the analysis indicated a high degree of coordination between the protective devices for the associated circuits of interest and the main protective devices for required power sources. Several cases for which the degree of coordination was insufficient were identified, and suitable new trio setting values for the circuit breakers were established.

Installation and maintenance work requests were initiated to correct these.

! This review demonstrated that the circuit breakers were coordinated in accordance with accepted design practices and that required power sources will be adequately protected from fire induced faults on circuits " associated by common power supply."

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a 11.0 RESPONSE TO THE CONCERN OF NRC BULLETIN lEB 84-09 REQUIRING DIRECTREADING LEVEL INSTRUMENTS ON ALL TANKS (Audit Report item 395/85-26-08, Ref. 5.a. (3) (a) 3) 1.E. Bulletin 84-09 provides guidance for power reactor facilities conducting analyses and/or making modifications to implement requirements of 10 CFR 50, Appendix R.

Attachment 1,l.E.B 84-09, page 6 of 9, dated 2/13/84 and 10CFR 50, Appendix R, Section lit L2.d. require direct reading of process variables necessary to perform and control the reactor shutdown function (including instrumentation needed for level indication on all tanks used).

Currently, a direct reading level instrument is not available for the Refueling Water Storage Tank. SCE&G proposes to implement a modification providing a direct reading pressure sensor type level gauge to the RWST.

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12.0 TIME CRITICAL SAFE SHUTDOWN FUNCTIONS AND TIME TO COLD SHUTDOWN (Audit Report item 395/85-26-01, Ref. 5.a. (1) and item 395/25-26-09.

Ref. 5.a. (3)(b))

12.1 Introduction One concern associated with Appendix R Compliance is the ability of the shutdown procedures to mitigate possible abnormal system transients on a time scale consistent with not reaching an unrecoverable condition; Also of concern is the ability to achieve cold shutdown within time criteria of Appendix R.

12.2 Primarv System Concerns For the primary coolant system the concern was the need to restore primary coolant makeup.

Several potential situations were considered which limit the time available to restore primary makeup. If one power operated relief valve (PORV) spuriously opens,it must be reclosed in time to ensure that sufficient inventory remains to allow for the maximum technical specification leakage until makeup can be restored. Similarly,if the reactor coolant pump seals heat up and warp, causing increased leakage to the levels postulated in WCAP 10541, along with leakage at the technical specification limits, mak eup must be restored before the pressurizer empties.

12.3 Secondarv system Concerns For the main steam system, two similar situations were considered. If one of the main steam PORV's spuriously opens, it must be reclosed in time to ensure that the primary shrinkage does not empty the pressurizer. Similarly, with steam relieI removing decay heat, an emergency feedwater pump must be started to provide emergency feedwater before the steam generators go dry.

12.4 Time Line i

Table 12.4 is a tabulation of these concerns, the maximum time for mitigation, and the time to complete mitigating action. The times to complete these mitigating actions are based on situations requiring control room evacuation and can be correlated with the timeline presented in Figure 12.4. For fires not requiring control room evacuation, the times for mitigation will be significantly less.

Other situations were also analyzed such as the time available to isolate letdown, and these were found to be considerably longer than those considered to be critical.

It should also be noted that the times listed in Table 12.4 represent the times required for mitigation using Procedure FEP-I.O. After completion of the planned modifications to the plant, which will provide local control switches and related facilities as described in the May 29,1985, submittal, the times for mitigation will be somewhat shorter since the interim (compensatory) jumpering procedures will no longer be required.

l l 12-1

12.5 . Cold Shutdown For conditions requiring control room evacuation, fire emergency safe shutdown procedur{:[,

FEP-1.1 is under development. It is presently estimated that this procedure can achieve shutdown within 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> from the start of the event.

For conditions not requiring control room evacuation, fire emergency safe shutdown procedures are to be developed. However, only a few areas of the plant will require repairs to achieve cold shutdown and these repairs are very limited. Therefore,it is anticipated that the plant can be ready to start cool down in significantly less than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> for a fire in any

., area of the plant.

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!- 12-2

TABLE 12.4 TIME FOR CRITICAL SAFE SHUTDOWN FUNCTIONS MITIGATING ACTION FUNCTIONAL CRITICAL TIME FOR (PROCEDURE - TIME TO COMPLETE CONCERN FU NCTIONAL CONCERN REFERENCE) MITIGATING ACTION

1. PORV spuriously 3.0 minutes (to level Pull fusesin MCB of immediate for fuse opens that provides for 2.2 open switchesin cable removal / switch hours of 3 gpm per spreading room operation and pumpleakage before (Att. lli, Step 1.0) and 90 minutes for pressurizer empties) restore charging charging (Step 10.0)

2. RCP sealleakage 90 minutes (with Restore charging Approximately above tech. specs. . leakage per WCAP through seals 90 minutes or less 10541 before (Step 10.0) pressurizer empties)

3. Primary coolant 2.7 minutes (before Pull fusesin MCB or immediate for fuse shrinkage due to pressurizer empties). open switchesin cable removal / switch spurious opening Based on conservative spreading room operation of main steam Chapter 15 analyses; (Att.111, Step 1.0)

PORV actual cooldown is expected to be much slower.

4. Steam generator Greater than 30 Provide feedwater Apprcximately dryout minutes (for three using turbine driven 25 milutes or less steam generators dry emergency feedwater with no feedwater pump (Att. IV, Step S.0) starting at low-low level)(NUREG-0611) e i

i h

l 12-3

,4 4

--y- - e,-,,- . . - .-

..y,- a- -- , , - - - _ . - - . , - - .. + - , ,- ,,, ,,-- , , . - - , , . , - y - - - , -, ,.-

O FIGURE 12 4 o OPERATOR TIME LINE ,

MAJOR EVENTS Presssurizer and steam PORV's isolated g Control at CRE P,7.2 k v buses de-energized ,6 Emergency feedwater water flow established ,, .,6 Diesel generator ready for loading a6 RCP seat injection established .

. a OPERATOR ACTIONS CONTROL ROOM SUPERVISOR implement Emergency Procedure FEP-t 0 Establish ccimmunication and control at CRE P, isolate letdown Establish steam generator level control Coordinate hot standby Establ6h seal *njection And prepare for cold shutdown NUCLEAR REACTOR OPERATOR AT CONTROLS Trip reactor -

De-energize 7 2 h v switchgear and strip loads Check d:esel cooling and start diesel "B" Supervise diesel generator "B" loading

" A* NUCLE AR RE ACTOR OPERATOR isolate oressurizer and steam PORV's isolate steam generator blowdown isolate seal water return Establish service water Establah RCP seal injection SHlFT TECHN: CAL ADVISOR Establish steam supply to emergency teedwater pump turbene Establish Emerg. Feed Pum p turbine centrol for 5G levei control Maintain steam generator levels MA!NTENANCE PERSON Proceed to CRE P Assist in stripping loads OPERATOR RELE ASED FROM BRIGADE Assist in establish ng RCP seal injection and task 5 as assigned 0 10 20 30 40 50 60 70 80 90 REFER TO PROCEDURE FEP-10 START MINUTES f OR DE TAILS OFEVENT 12-4

13.0 EMERGENCY LIGHTING (Audit Report item 395/85-26-12, Ref. 7)

During the NRC audit, questions were raised about the intention of SCE8G to utilize 1 hand-held lighting as a permanent solution to providing required light levels for a few iso'ated locations required by our safe shutdown scheme. As a result of discussions at that j

time and at a subsequent meeting between NRR, Region 11 and SCE&G on July 23 and 24,

~

1985, SCE&G now proposes to expand still further the amount of emergency lighting to be added to V. C. Summer Nuclear Station. Table 13.0 summarizes the lighting to be added. The entries marked with an asterisk (*) are additions to the list of additional lighting reviewed by the Region !! auditors. With their inclusion, SCE8G will be in full compliance with Section J of j Appendix R. Adequate lighting will be provided to perform all required functions including 1

ingress, egress, and operation of components without resort to hand-held lights.

Two comments should be made: One, the yard lighting will be provided to illuminate the i external entrances and exits of the affected structures, Turbine Building, Service Water 7

Pumphouse, and Circulating Water Pumphouse. No lighting is planned for general yard 4

areas, since the provision of entrance lights to illuminate doors and stairwells and to serve as -

guide beacons is suf ficient for safe travel between structures. Two, procedures will continue to call for hand-held lighting as a " belt and suspenders

• approach to safety.

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I TABLE 13 0 EMERGENCY LIGHTING ADDITIONS BUILDINGS /

ELEVATIONS )

AB 388,397 -

4 dual and 2 single lamp assemblies AB 400 -

1 duallamp assembly AB 412 -

2 duallamp assemblies West Pen 412 -

1 duallamp assembly East Pen 412 -

1 duallamp assembly IB 412 -

2 dual and 2 singlelamp assemblies TB 412 -

2 dual and I single lamp assembly

• CB 412 - 1 duallamp assembly AB 424 -

1 single lamp assembly IB424 - 2 duallamp assemblies

• DG 427 -

1 duallamp assembly AB 436 - 5 dual and I single lamp assembly West Pen 436 - 1 duallamp assembly East Pen 436 -

1 duallamp assembly IB 436 -

3 duallamp assemblies

• TB 436 -

2 dual and I singlelamp assembly IB 463 -

1 duallamp assembly

• TB 463 -

2 duallamp assemblies

• CB 463 - 1 duallamp assembly

• AB 485 - 1 dual lamp assembly SWPH 1 duallamp assembly CWPH -

1 single lamp assembly STAIRTOWER$

CB North Stairtower - 2 dual lamp assemblies AB North Stairtower -

3 duallamp assemblies (

AB West Stairtower - 2 dual lamp assemblies IB East Stairtower -

1 dual and I single lamp assembly IB West Stairtower -

2 dual lamp assemblies YAR D (436')

• Yard Lighting -

3 duallamp assemblies

• Post audit additions to preclude the need for deviation requests.

NOTE: All proposed units are 8-hour battery operated units 13 2

+=i.

14.0 SCHEDULE FOR COMPLETION OF MODIFICATIONS AND COMPENSATORY ACTION TO BE USED UNTilINSTALLATION (NRR Meeting item)

Table 14.0 outlines the schedule of implementation and summarizes compensatory action for planned modification associated with the reanalysis of V. C. Summer Nuclear Station to 10 CFR 50, Appendix R. The modifications have been proposed to resolve Associated Circuits of Concern issues ano enhance compliance with the recent Appendix R guidelines set forth in generic letters 81-12,83-33,85-01, and other interpretative documents. It should be noted that some modifications listed under the "MRF DESCRIPTION" column are actually combinations of audit open items, while some other modifications listed are more finely broken down. This is only a result of design and construction scheduling in order to efficiently process the modifications.

These plant modifications were identified and developed during the recent reanalysis effort, were described in the May 29,1985, letter (to Mr. H. R. Denton, NRR), and were reviewed during the NRCaudit at the plant site.

Several of the modifications require that the plant be off-line and, therefore, need to be scheduled during outage periods. The first such outage is the second refueling outage, scheduled for the fall of 1985. The short lead time to this outage, along with the need for a controlled approach to assure plant reliability and safe operation, have placed engineering and procurement constraints that preclude completion of most off-line related work during

. this refueling outage. SCE&G, therefore, will attempt to complete all work by the end of the third refueling outage (as referenced in the chart), scheduled for the spring of 1987. Non-outage work (which can be done during normal operation)is scheduled between now and the third refueling outage. Outage related work will progress as far as possible subject to constraints imposed by adherence to technical specifications and will be completed as quickly as possible taking advantage of any forced (unplanned) outages of sufficient duration.

The COMPENSATORY ACTION column summarizes the actions that have been taken to justify continued operation of the plant untilinstallation. The May 29,1985, letter to Mr. H. R. Denton of the NRR fully descr bes all plant modifications and interim compensatory action. Action of a procedural nature will be (or already has been) incorporated into the fire emergency safe shutdown procedures FEP-1.0 and FEP-1.1. A roving fire watch has been

[

initiated to provide assurance that the SCE&G interim procedures will not be subject to a challenge by a significant fire. This fire watch provides surveillance of the general floor areas of all fire areas and fire zones once every two hours except zones completely contained in high-rad, airborne, contaminated, or confined spaces. For a detailed explanation of the roving fire watch, see the June 21,1985, letter to Dr. J. Nelson Grace of the NRC.

14-1

TABLE 14 0 V C SUMMER NUCLFAR STATION ,5 SCHEDULE OF MODIFICATION IMPLEMENTATION ESTIMATES MRF SCE&G

SUMMARY

OF NRC AUDIT REPORT DESCRIPTION Ol AGE CONST. COMPENSATORY ENGR. MATERIAL ITEM #

COMPL. DELIVERY COMPL. DATE ACTION MRF #

DATE DATE 395/85-26-02 Add 2nd power disconnect for air Yes 4/86* 4/86* 3rd Refuel Interim Wire Cutting Procedure Para 5 a (1)(a) opesated valves (2nd Qtr. '87) thru 5 a (1)(e) 20784 627 395/85-26-02 Add Ihysite protectors to current No Design "9 #'

Para 5 a (1)(g) transloemers Issued 12/85* 4/86 20785-584 496/85-26 02 Add Fire 5 witches to 480V Yes 4/86* 4/86* 3rd Refuel Interim jumper Procedure Para 5 a (1) switchgear 20786 630 (2nd Qtr. '87) 395/85-26-02 Upgrade DG-8 Controls Yes 9/22/85* 9/22/85* 2nd Refuel Interim Jumper Procedure Para 5 a (1)(1) 20788-579 (4th Qtr. *BS) e Add Fie- Switches to 7.2kV Yes 3/86* 3rd Refuel Interim Jumper Procedure i 395/85-26-02 3/86*

N Para 5 a (1) switchgear 20789-631 (2nd Qtr. '87) 395/85-26-02 Upgrade Fire Switch for 5.W. Pp. - 7/86* 11/86* 3rd Refuel Roving Fire Watch Para 5 a (1) "B" and add Fire Switch for CC (2nd Qtr. '87)

Pp ~B* 20790 395/85-26-02 Upgrade Chiller 8 Contr. Transfer No 12/31/85* 12/31/85* 3/1/86 Interim Jumper Procedure Para 5 a (1) Swit(h 20791-614 Armor cables for solenoid valves Yes 3rd Refuel Interim Wire Cutting Procedure 395/85-26-02 10/19/86* 10/19/86*

Para 5 a (1)(a) (see also 20734) 20800-628 (2nd Qtr. '87) and 5 a (1)(e) 20784 395/85-26-02 Change RC Loop Tu/Te Yes To be 2/87* 3rd Refuel Operational Procedure Para 5 a (1)(h) instrumentation to one power completed as (2nd Qtr. '87) train per loop a resolution 20801 to R.G 1.97 issues (9/86*)

• These are tentative dain and subject to change as detailed planning proceeds.

14-2

.m

TABL E 14 0 (CONTINUED) .

o V. C SUMMER NUCLEAR STATION **

SCHEDULE OF MODiflCATION IMPLEMENTATION ESTIMATES

" ^ ' SCE&G $UMMARY OF DESCR PTION OUTAGE CONST. COMPENSATORY REPORT ENGR. MATERIAL REQ DELIVERY COMPL. DATE ACTION ITEM # COMPL.

MRF #

DATE DATE 395/85-26-02 Remov. ables DGM21B and No 9/22/85* N/A 2nd Refuel interim Lead-Lifting Procedure Para 5 a (1)(f) DGM11B from tower cable (4th Qtr. '85) spreadmq rcom 10788 579 install ..ddi tional sel f-contained No 1/1/86* 2nd Qtr. '86 Existing Emergency Lighting 395/85-26 12 1/1/86* Supplemented by Hand Held emergenty hghung units 20840 Lighting 395/85-26-04 Protect (unduits for N1-31/N!-32 N 2/86* 151 Qtr. '86 Roving Fire Watch 12/31/85*

or mstall transfer switch for N!-33 31968 w

Addebon of direct reading level No 9/86* 4th Qtr. *86 Roving Fire Watch I 9/86*

W gauge for RWST Revision of power circuit breaker Yes 9/27/85* N/A 2nd Refuel Roving Fire Watch 395/85-26-05 overcurr ent sethngs for (4th Qtr. '85) coordination 20846 Add f u e switches to 480V MCC's, Ye5 10/86* 3rd Refuel Interim Jumper Procedure 395/85-26-02 9/86*

XFN 45A-AH, XFN 458-AH, XFN- (2nd Qtr. '87) 468-VL. XFN 808-VL _20806 Wrap Tray 3088 31971 No 12/31/85* 2/86* 151 Qtr. '86 Roving Fire Watch 395/85-26-03 Upgrade 5 W 8 Pp 45A barrier No 2/86* 3/86* 2nd Qtr. '86 Roving Fire Watch 395/85-26-16 (supporig 20895 Change M 5 PORV controllogic No 9/86* N/A 4 th Qtr. '86 interim Valve Cutting Procedure 395/85 26-02 Para 5 a (1)(b) o pre- 1980 design (IPV-2000, 2010. 2010-MS)

- Upgr..de the FPER No 6/86* N/A 2nd Qtr. '86 -

• These are tentative datn and subject to change as detailed planning proceeds 14-3

. . . .