ML20249A122

Forwards Unsolicited Info from San Onofre Licensee Re Emergency Diesel Generator Delegation Allowed Outage Time Extension TS Request

ML20249A122
Person / Time
Site:	San Onofre
Issue date:	06/11/1998
From:	Ian Jung NRC (Affiliation Not Assigned)
To:	NRC (Affiliation Not Assigned)
References
NUDOCS 9806160081
Download: ML20249A122 (54)

Text

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NUCLEAR REGULATORY COMMISSION o WASHINGTON, D.C. 20555-0001

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June 11,1998 MEMORANDUM TO: Docket File FROM: lan C. Jung, Reliability and Risk Analyst b Y Probabilistic Safety Assessment Branch Division of Systems Safety and Analysis

SUBJECT:

UNSOLICITED INFORMATION FROM THE SAN ONOFRE LICENSEE REGARDING EMERGENCY DIESEL GENERATOR DELEGATION ALLOWED OUTAGE TIME EXTENSION TS REQUEST On June 4,1998, the San Onofre 2 and 3 licensee has provided the NRC with unsolicited information in the form of electronic mail and facsimile. It contains probabilistic risk analysis information for the staff review of the proposed emergency diesel generator allowed outage time extension (See the attached). This memorar:dum is intended to provide this information in the public domain. The licensee has also indicated that the same information would soon be f.-

I submitted to the NRC staff through normal channel. '

cc: Docket File (50-361/362)

PUBLIC CONTACT: 1. Jung, SPSB/DSSA 415-1837

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Additional Information Needed for Review of the San Onofre Emergency Diesel Generator Allowed Outage Time Extension Request QUESTION #1: Consistent with the Drafl Regulatory Guide 1.174,"An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the

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Current Licensing Basis," the staff expects that the scope and quality of the analysis conducted to justify the proposed changes be appropriate for the nature and scope of the change, and be based on the as-built and as-operated and maintained plant. Further, the current plant-specific PRA supporting the licensee proposal has been subjected to quality controls. The staff finds that your current PRA is different from the original Individual Plant Examination (IPE). Extensive review i of the Joint Application Report for Emergency Diesel Generators (EDGs), responses to the staff requests for additional information, your plant-specific submittal, and your IPE has been performed to examine your PRA. The review resulted in the need for the following additional information for the staff evaluation of your plant-specific submittal.

a. Describe any independent peer reviews performed on your most updated PRA. What quality control process do you have in order to maintain your current PRA representing the as-built and as-operated plant?

SCE Response: An extensivepeer review ofthe San Onofre Living PRA was conductedin i late 1996. Immediatelyfollowing thepeer review, San Onofre institutedaprocedurali:ed process to track, independently review, and document changes to the Living PRA model.

A briefsummary cfthatpeer review isprovided in Attachment #1. Since that time, the San Onofre Living PRA has been controlled via a change process delineated in procedure PRA-REV-001, "PRA Model Revisions", Revision 4, which isprovided as Attachment #2. This procedure establishes administrative controls andprovides l guidelinesforperforming updates to maintain the Living PRA models currentfor San l Onofre Units 2 and 3. Thisprocedure requires the review ofplant design and operation documents, such as-Abnormal Operating Instructions, Design Change Packages, Operating Procedures, etc., to determine ofthe PRA is impacted by the changes. The Living PRA modelis updated to reflect plant design and operating changes as near real time as practical. Initiating events and component reliability / availability data are updated at periods no greater than once every two refiteling cycles.

b. Provide the following for staff revie,v of your current PRA:

(a) a copy of the LOOP /SBO event tree, SCE Response: Enclosedin Attachment #3 are copies ofthe SONGS 2/3 Loss of Offsite Power andStation Blackout Event Trees.

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' (b)- a list of dominant SBO sequences, SCE Response: Enclosed in Attachment #4 are the top 50 cutsetsfor Loss of Ofsite Power / Station Blackout events.

- (c) major modeling assumptions for LOOP /SBO events, and i

SCE Respon.se: Major PRA modeling assumptions with respect to LOOP /SBC events include thefollowing:

a) . During an SBO event, the Class 1E 125VDC batteries on busses A and B can supplypower up to 90 minutes without loadshedding and 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> with loadshedding. The batteries on the other two busses, C & D, can ,

providepower up to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

b) Although the turbine-driven auxiliaryfeedwaterpump may initially be available during an SBO event, it is conservatively assumed that the loss ofthe vital batteries and resulting loss ofsteam generator level indication willlead to core damage.

I c) ' Credit is takenfor the engineered diesel generator cross'-tiefollowing an SBO.

(d) major differences from your original IPE.

SCE Response: Enclosed as Attachment #5, " Updates to the San Onofre IPE PRA l Model", is a list ofsigmficant changes to the PRA model that have occurredsince the originalIPE submittal. While the number ofchanges is substantial, there have been no structural changes to the electricalpower system model since the JPE. The only major changesfrom the IPE affecting SBO were: (1) the revision  ;

ofthe RCP seal LOCA model to reflect more conservative ABB/CE assumptions l

and (2) increasing creditfor the EDG cross-tie capability. The EDG cross-tie has '

always been credited in the San Onofre IPE, however, the likelihood ofits success l will be improved sigmficantly with the installation ofthe engineered EDG cross-tie modification.' The overall core damage impact ofthese model changes has  ;

been a reduction in the average CDFfrom 3.0E-5/yr reported in the 1993 SONGS 2/3 IPE to the current average CDF of2.5E-05/yr. The Level 2 impact ofthese changes has not been quantifiedsince LERF was not calculated at the time ofthe JPE submittal and CCFP is no longer calculatedfor the present model. l

- c. ' Both the Joint Application Report (CE NPSD-9%) and your plant-specific submittal ums Page 2 of 5 c:snetwp6sjungagr3.wp  !

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qualitatively addressed the Level 2 risk associated with the proposed extended EDG AOT. However, quantitative Level 2 risk information is also an important consideration when evaluating the risk associated with your proposed EDG AOT extension request.

The conditional containment failure probabilities (CCFPs) and large early release frequency (LERF) for both the current and proposed cases calculated from your updated PRA should be submitted for staff review. In addition, incremental conditional large early release probability (ICLERP) for a single 14-day AOT should also be submitted for staff evaluation of the risk associated with a single EDG outage.

SCE Response: Utili:ing the methodology described in CE NPSD-966, the current and proposed AOT extension internal events LERF values and single A OTICLERP were generated. See Attachment #6, Tables 1-1,1-2, and 1-3, for the results ofthese LERF andICLERP calculations. Per Draft Reg. Guide 1.174, the resultant increase in LERF due to this ACTextensionfalls within acceptance Region ill, and therefore, is considered acceptable. In addition, the single AOTICLERP value is much less than the 5.0E-08 ICLERP acceptance guideline published in NRC Draft Reg. Guide DG-1065. CCFP is no longer calculatedfor the San Onofre PRA models and is not a risk metric in draft risk-informed Reg. Guides DG-1061 or DG-1065.

d. No RCP seal loss of coolant accident (LOCA) was assumed during an SBO in your IPE.

Provide your engineering basis for making that assumption. Provide a discussion of how your PRA modeled the potential RCP LOCA during an SBO in your most updated PRA, if modeled differently from the IPE. If you have performed any sensitivity studies on potential seal failures during an SBO, provide the results. Explain why the potential RCP seal LOCA during an SBO is a negligible risk contributor for the proposed changes.

SCE Response: The San Onofre IPE model assumed that no RCP seal LOCA would occur during an SBO based on information provided in Combustion Engineering 's response to draft NUREG-1032, (CE NPSD-340, " Evaluation ofStation Blackout Accidents at Nuclear Power Plants ", March 1986). This doci na npresented the results ofSBO simulated tests as well as documentation ofRCP seal performance during actual loss ofCCWevents at CEplants. A simulatedstation blackout test at RCS temperature andpressure was performed with a prototype cartridge ofsimilar design as those used at San Onafre (4., cal stages, each ofwhich is designed to withstandfidl RCSpressure and temperature). The test was runfor 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br /> without loss ofsealfamction. A second test ran aprototype RCPfor 30 minutes without component cooling water cooling. During the 30 minute test, sealfimetion was maintained. The results ofthese tests are consistent with actual operating historical experience. As documented in the CE NPSD-340, these have been a number ofinstances in which these types ofpumps were runfor longperiods oftime without CCW cooling. These instances, none ofwhich resultedin loss ofseal function, were more severe than an SBO event since thepump would not be running following an SBO event and thus would not be generating additional seal heat. Based on this information, San Onofre did not consider RCP seal LOCA to be a credible event if 6rm s Page 3 of 5 cancewr6tjungdy3 wp

l l-thepumps are not running.

However, after the IPE was submitted, San Onofre revised the Living PRA model to include the potentialfor an RCP seal LOCA (1.5E-3) even ifthe RCPs are tripped within 30 minutes based on an assessment by ABB/CE ofall CEplant RCPs (CEOG Letter / Report CEOG-92-3203, " Additional Analysis ofRCP Seal Failure Given a Loss of

( Cooling" June 15,1993). Since no RCP sealfailure has occurred in a CE PWR, this l value was based on a conservative extrapolation ofsealleakage events (< 3 gpm) when l' one or more stages were observedto befailedor degraded. A more recent examination ofthis experience indicates that RCP sealfailureprobabilities are more than an order of magnitude lower than that used in the current SCE model.

e. For multiple unit sites, potential dual-unit initiators, single-unit initiators that propagate l to the other unit, and shared systems can be important in PRA. Explain how 3 sur current PRA addresses these issues and provide a quantitative or qualitative discussion that these issues either have been adequately taken into account in your analysis or have a n gligible risk impact due to the proposed changes.

SCE Response: Dual-unit initiators were evaluated aspart ofthe IPE effort and were determined to not contribute measurably to the core damage risk. Dual-unit initiators l

[ were considered in the EDG A OT extension analysis and were also not considered to }

. measurably contribute to core damage risk. The common systems in the Living PRA L model shared by Units 2 and 3 include instrument air, ' normal chilled water, emergency chilled water, turbine plant cooling water, and vital 4k Vpower (onlyfor secondary and 1 emergency AC sources). The potentialfor maintenance unavailability orfailures in these common systems' to impact both units is explicitly modeled in the San Onofre Safety

' Monitor. The only single unit initiatorpotentially impacted by the EDG AOT extension is aplant centered loss ofofsitepower at one unit. This event requires a number of l multiplefailures whose likelihood is negligible topropagate to the oth.:r unit and cause a core damage event. The breakerprotection scheme at each unit ensures that any electricalfaults on one unit will not propagate to the other unit. The San Onofre 2B l Configuration Risk Management Program will ensure during an extended outage ofa EDG that no other EDGs at either unit will voluntarily be removedfrom service and the EDG cross-tie capability will be available.

QUESTION #2: Your IPEEE analysis shows that the average CDF contributed by external events exceeds that contributed by internal events. Discuss the impact on external event risk due to the proposed changes either quantitatively or qualitatively, and justify that the impact on the

external event risk would not result in a different overall risk perspective.

SCE Response: The SONGS 2B IPEEE analysis determined the average CDF contributed by external events to be 3.3E-5 peryear. This compares to the SONGS 2B IPE analysis which calculatedan internal events average CDFof3.0E-5peryear. The contribution to core damage L

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riskfrom external events is dominated by seismic andfire initiating events. High winds, floods, and other ha:ards were notfound to contribute to the overall core damageprobability. The single 14 day ACT core damage riskfor seismic events was anal):ed andfound to be less than the NRC acceptance criteria ofSE-7. The impact ofthe EDG AOTextension onfire initiators was evaluated andfound to be much less likely than those causes anal > ed in the internal events PRA. Therefore, the inclusion ofexternal events in the analysis does not change the overall acceptability ofthe AOTextension.

QUESTION #3: According to your submittal, the cross-tie capability, recently installed at your site and modeled in your current PRA, significantly reduces the SBO risk contribution from 1.8E-6/yr to 5.91E-8/yr. The cross-tie was worth over 1.7E-6/yr for the SBO CDF. You have also indicated that the cross-tie reduced the total CDF by 3.4E-6/yr (11% reduction in total CDF). The staff expects that the cross-tie capability would dominantly contribute to the reduction in the SBO risk; however, your submittal indicates that the SBO CDF reduction accounts for only 50% of the total CDF reduction. Please explain how the cross-tie contributes to the remaining non-SBO risk.

SCE Response: The values referred to above were reported in response to two different NRC questions in a CEOG response datedMarch 26,1997 to a requestfor additional information (RAl) on the EDG ACTExtension. The responses to the two questions were not identical due to subtle diferences in the NRC questions. In question 6, the NRC requested the change in risk due to crediting any alternate AC or EDG cross-tie capability. In question 7, the NRC requested the change in riskfrom the IPE analysis due to crediting any alternate AC or EDG cross-tie capability. These changes in risk are not identical since limited cred.tfor the EDG cross-tie was assumed in the IPE. San Onofre has had the capability (includingprocedures) to cross-tie the EDGs since the late 1980s and as such it was credited in the JPE. However, duringpreparation ofthe IPEEE, San Onofre determined that upgrading the EDG cross-tie capability would sigmficantly reduce the contributionfrom core damage due to seismic events. Therefore, the difference between the two values above reflects the reduction in internal events core damage riskfrom implementing the limited EDG cross-tie capability assumed in the JPE. Therefore, the EDG cross-tie does not contribute to any other events other than SBO.

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SAN ONOFRE UNITS 2/3 LIVING PRA MODEL PEER REVIEW 1

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Attachment #1 San Onofre Units 2/3 Living PRA Model Peer Review A comprehensive independent peer review of the SONGS 2/3 Level 1 and Level 2 internal events living PRA for full power and shutdown operations was conducted between August 1996 and April 1997 by Dr. Parviz Moieni of SCIENTECH, Inc.. The review was mainly based on the guidance provided in the PRA procedure guides such as NUREG/CR-2300 and NUREG/CR-4550 as well as PRA applications documents such as EPRI TR-105396 and NUREG-1489. A database (called as the PRA Review Punch List) was developed by SCE as a mechanism to systematically track the review comments. The scope of the peer review is outlined in detail below.

I. System Fault Trees The following system fault trees with their assumptions and associated basic event (BE) calculation files were reviewed and comments were entered in the SONGS PRA Review Punch List:

Auxiliary Feedwater(AFW) a Low Pressure Safety Injection (LPSI)

Containment Spray / Containment Emergency Cooling (CS/ CEC)

High Pressure Safety Injection (HPSI)

Heating, Ventilation and Air Conditioning (HVAC)

Component Cooling Water (CCW)

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Chemical and Volume Control System (CVCS)

Main Feedwater(MFW) and Condensate Main Steam System (MSS)

Electric Power (EP)

Instrument Air (IA)

= Plant Protection System ( PPS)

Safety Injection Tank System (SIT)

Reactor Coolant System (RCS) Pressure Control Containment Isolation System (CIS)

Saltwater Cooling System (SWC)

The above list re; resents all the system fault trees in the SONGS 2/3 Living PRA.

II. Event Trees The following event trees with their assumptions and associated BE calculation files were reviewed and comments were entered in the SONGS PRA Review Punch List:

1 Attachment #1 San Onofre Units 2/3 Living PRA Model Peer Review l l

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. Loss of Power Conversion System (PCS)

. Transients with PCS Initially Available (TF) I Loss of Offsite Power (LOP)

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Station Blackout (SBO)

Main Steam Line Break (SLB)

. Large LOCA (LL)

Medium LOCA (ML)

Small LOCA (SL)

Small Small LOCA (SSL)

= Steam Generator Tube Rupture (SGR)

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Loss of125V DC Bus (LDC)

• Loss of Component Cooling Water (CCW)

. Interfacing System LOCA (VL) a Reactor Pressure Vessel Rupture (VR)

• Anticipated Transient Without Scram (TWS)

. Internal Flooding Analysis The above list represents all the event trees in the Level-1 SONGS 2/3 Living PRA. )

III. Basic Event (BE) Calculation Files As part of the SONGS 2/3 Individual Plant Examination (IPE) study, a large number of basic event (BE) calculation files had been developed to support a variety of tasks such as human reliability analysis (HRA) and common cause failure (CCF) analysis. All the BE calculation files related to the following topics were reviewed and comments were entered in the SONGS PRA Review Punch List.

• Fault tree analysis

• Event tree analysis

. CCF analysis

• Pre-initiating event operator actions (Type A) HRA

. Post-initiating event operator actions (Type C) HRA

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• Plant-specific equipment data analysis (i.e., Bayesian update of equipment failure rates)

• Plant-specific maintenance unavailability calculations i

Over 200 BE calculation files were reviewed.

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Attachment #1 San Onofre Units 2/3 Living PRA Model Peer Review IV. Level-2 Extended Event Trees (EETs)

Tl e following Level-2 extended event trees with their assumptions were reviewed and comments were entered in the SONGS PRA Review Punch List:

Loss of Power Conversion System (P1E and P2E)

Transients with PCS Initially Available (TlE and T2E)

Loss of Offsite Power (LPE)

Station Blackout (SBE)'

Main Steam Line Break (MSE)

. Large LOCA (LLE)

Medium LOCA (MLE)

Small LOCA (SLE)

Small Small LOCA (SSE)

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Steam Generator Tube Rupture (G1E, G2E, and G3E)

Loss of125V DC Bus (LDE)

. Loss of Component Cooling Water (CWE)

. Interfacing System LOCA (VLE)

Reactor Pressure Vessel Rupture (VRE)

. Anticipated Transient Without Scram (AWE)

The above list represents all the extended event trees in the Level-2 SONGS 2/3 Living PRA.

V. Shutdown PRA Fault Trees The following shutdown PRA fault trees with their assumptions and associated BE calculation files were reviewed and comments were entered in the SONGS PRA Review Punch List:

. Spent Fuel Pool Cooling System (SXSDR)

. CS/LPSI for Spent Fuel Pool Cooling (TXSDR)

. Spent Fuel Pool (SFP) Makeup via SPF Makeup Pump P011 (YXSDR)

. SFPC Pumps for SFP Inventory Makeup (SMSDR)

. Shutdown Cooling System (AXSDR)

. Charging System - for Injection (DXSDR)

• HPSI for PCS Inventory Makeup (HXSDR)

• LPSI for PCS Inventory Makeup (VXSDR)

.. . Containment Spray for Inventory Makeup (NXSDR) l l

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Attachment #1 San Onofre Units 2/3 Living PRA Model Peer Review l

Component Cooling Water (EXSDR)

Ileating, Ventilation and Air Conditioning (MXSDR) l .

Saltwater Cooling System (PXSDR)

Electric Power (UXSDR)

Instrument Air (BXSDR)

Plant Protection System (KXSDR)

The above list represents all the system fault trees for the SONGS 2/3 shutdown risk assessment.  !

VI. Shutdown PRA Event Trees All the shutdown PRA event trees with their assumptions and associated BE calculation files were reviewed and comments were entered in the SONGS PRA Review Punch List. A total of 32 event trees associated with 4 initiating events and 8 plant operating configurations had been  !

developed in the shutdown PRA. The 4 initiating events are:

. Loss of Operating Shutdown Cooling (SDC) Train Loss of RCS Inventory Plant-Centered Loss of Offsite Power Grid-Related Loss of Offsite Power The 8 plant operating configurations are:

Configuration 1 - RCS Water Level l' Below RV Flange, Fuel in the Core Configuration 2 - RCS Operation, Fuel Officading in Progress Configuration 3 - Spent Fuel Pool (SFP) Operation, Fuel Offloaded in the SFP

= Configuration 4 - SFP Operation with CS/SFP Cross-Tie Inservice, Fuel Offloaded in the SFP Configuration 5A - SFP Operation with CS/SFP Cross-Tie Inservice, Fuel Reloading in Progress Configuration 5B - RCS Operation, Fuel Reloading in Progress

. Configuration 6 - RCS Water Level l' Below RV Flange, Fuel Reloaded

. Configuration 7 - Midloop Operations, Fuel Reloaded The event trees for spent fuel pool were not available for review.

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Attachment #1 San Onofre Units 2/3 Living PRA Model Peer Review 1 VII. Shutdown HRA Worksheets A sample (three) of the human reliability analysis (HRA) BE calculation files related to the shutdown PRA were reviewed and comments were entered in the SONGS PRA Review Punch List.

VIII. Seismic-Related HRA Worksheets The seismic HRA methodology described in Section 3.6.4 of the IPE for External Events (IPEEE) study was reviewed. Also, the following seismic-related HRA BE calculation files were reviewed and comments were entered in the SONGS PRA Review Punch List:

• Operator failure to reset relays for battery chargers (OP64)

=- Operator failure to failure to open SWC pump discharge valve and start the redundant SWC pump (OP91)

Operator failure to reset and start the DG following relay or process switch chatter (OP13) )

Operator failure to align fire truck for CCW makeup given primary plant makeup l (PPMU) tank is unavailable (OPFIRETNKR)

• Operator failure to respond to high temperature alarm in the switchgear/ distribution room and align portable ventilation (OPALTVENT) l l

The above list represents all the seismic-related HRA BE calculation files for SONGS 2/3 IPEEE l study.

- IX. Fire-Related HRA Worksheets i The fire HRA methodology and the following fire-related HRA BE calculation files were reviewed and comments were entered in the SONGS PRA Review Punch List:

= Operator failure to activate fire procedure SO23-13-2 (Shutdown from Outside the Control Room)

• Operator failure to manually control Train A AFW/HPSI/CS from outside the control room following a control room fire i

Attachment #1 San Onofre Units 2/3 Living PRA Model Peer Review Operator failure to manually start the diesel generator (DG) from the DG room following a control room fire Operator failure to manually start the diesel generator (DG) from the control room following a relay room fire The above list represents all the fire-related lira BE calculation files for SONGS 2/3 IPEEE study.

ATTACHMENT #2 i

1 SAN ONOFRE UNITS 2/3 l

PRA MODEL REVISION PROCEDURE l

1 Attachment #2 PRA-REV-001 Revision 4,12/17/96 I

.PRA MODEL REVISION j

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Prepared By:

Author Date l

Reviewed By:

NSG Supervisor Date

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i i I TABLE OF CONTENTS SECTION PAGE h

1.0 PUMOSE .......................................................................................................9 1.

2.0 . APPLICABILITY ..........................................................................................2 i

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3.0 REFERENCES

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i 4.0 RESPONSIBILITIES ..............................................................................................4

5. ..................................................................................................

6. .......................................................................................................

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Attachment #2 PRA-REV-001 L

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1.0 PURPOSE 1.1 This procedure establishes the administrative controls and provides the guidelines l . for performing updates to maintain current the PRA models and their documentation f,r San Onofre Nuclear Generating Station Units 2 and 3.

1.2 This procedure applies to all personnel who perform or support the updates of the l documents and models mentioned in section 2 of this procedure.

1.3 This procedure implements QA Procedure N3.06 Revision 1; Plant PRA Model Revision.

1.4 The procedure is in accordance with the fbilowing sections of the Topical Quality  !

Assurance Manual:

1-C Quality Planning (Instructions and Procedures) 1-D Document Management L i-J Electronic Data Processing Controls 5-A Procedures and Instructions -

2.0 Ai) PLACABILITY

'2,1 This procedure is applicable to the update / revision of all the SONGS 2/3 Living PRA models. The SONGS 2/3 Living PRA models includes all documentation, models, and software pertaining to the internal and external initiating events  !

analysis for full power and other' operating modes.

3.0 REFERENCES

i 3.1 Individual Plant Examination Report for San Onofre Nuclear Generating Station i Units 2 and 3 in Response to Generic Letter 88-20; Submittal Document, April 1993, and the Computer Model.  !

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3.0 REFERENCES

(Continued) 3.2 Individual Plvit Examination of External Events for San Onofre Nuclear l Generating Station Units 2 and 3 in Response to Generic Letter 88-20, Supplement 4; Submittal Document, December 1995, and the Computer Model.

3.3 ' San Onofre Topical Quality Assurance Manual l, 3.4 PSA Applications Guide, EPRI TR-105396, August 1995.

3.5 PRA Procedure PRA-IPE-010, PRA Living Database Update, Maintenance and Application. l l

3.6 Individual Plant Examination Procedure PRA-IPE-003, System Fault Tree Development and Analysis.

3.7- I"di Adual Plant Examination Procedure PRA-IPE-004, Initiating Event Data Collection and Analysis.

1 3.8 Individual Plant Examination Procedure PRA-IPE-005, Accident Seg 4ence l Analysts. i i

3.9 Individual Plant Examination Procedure PRA-IPE-006, Data Collection and Analysis.

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3.10 Individual Plant Examination Procedure PRA-IPE-006A, Supplemental Guidance  !

on Dependent Failures. 1 3.11 Individual Plant Examination Procedure PRA-IPE-007, Human Reliability Analysis.

3.12 Individual Plant Examination Procedure PRA-IPE-008, Internal Flood Analysis.

3.13 Individual Plant Examination Procedure PRA-IPE-009, Plant Model  !

Ouantincation.

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3.0 REFERENCES

(Continued)

3.14 Individual Plant Examination Procedure PRA-IPE-009A, Sensitivity Analysis.

3.15 QA Procedure N5.06, General Document Review Procedure 3.16 S0123-XXIV-5.1, Engineering, Construction and Fuel Services Software Quality

j. Assurance 4.0 RESPONSIBILITIES 4.1 The Supervisor, Nuclear Safety Group (NSG) is responsible for implementing and updating as necessary, this procedure. The NSG Supervisor or his designee is responsible for assigning responsibility of reviews and appropriate actions in this procedure to PRA Engineers and must review and approve any changes to the PRA recommended by the PRA Engineers.

4.2 Nuclear Safety Group staff trained in PRA technology (hereafter referred to as l PRA Engineers) are responsible for gathering the information on plam changes l and document revisions, performing the screening and review of plant changes as l they relate to the Living PRA models, databases, and associated documents, and for providing the recommendations regarding the necessity to incorporating the changes.

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Attachment #2 PRA-REV-001 Revision 4,12/17/96 l 4.'0 ' RESPONSIBILITIES - (Continued) lt l Additionally, the PRA Engineers are also responsible for performing the l necessary analysis and calculations for the purpose ofincorporating the changes,

! in accordance with the procedures controlling the existing documents, software and computer models.

l- 5.0 ' PROCEDURE I

L The PRA documents may be updated for various reasons. These reasons include plant modifications, error corrections, procedural change s, plant data update, improved methods, etc. The update frequency for changes other than data will be as near real time as practical.- Initiating events and component reliability / availability data will be updated at periods no longer than once every two refueling cycles.

p The following provides guidelines for the steps taken to ensure that updates are l

performed per the quality assurance requirements:

5.1.a Monitoring and Collecting Plant Change Information. Table.1 provides a list of documents that shall be reviewed by PRA Engineers as they are issued by CDM, l- for their potential impact on the PRA. The Nuclear Safety Group should be on l the distribution list for all of these documents (either hardcopy or electronic). The l Supervisor, Nuclear Safety Group or his dee' nee is responsible for designating a PRA Engineer to review each document. b a documents issued by CDM should be reviewed on a near real time basis as they are obtained to determine potential L impact and priority for incorporation. An Information Evaluation Form (Attachment 1) shall be completed for each document reviewed.

The collection ofinformation to support the update ofinitiating event and

! component reliability / availability data will utilize References 3.7 and 3.9 and be

' documented in accordance with those procedures.

i-I ~ 5.1.b - Other Causes of PRA Model Revision. In addition to plant change information, L other sources could cause a PRA model revision including: PRA review comment, plant specific data update, change in NRC requirements, change in NSG procedures. In these cases, ar. Information Evaluation Form is unnecessary since it has already been determined that a revision to the PRA is required.

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Attachment #2 PRA-REV-001 -

Revision 4,12/17/96 5.0 PROCEDURE (Continued) 5.2 Screen New Information. Table 1 provides guidance on which key elements of the PRA may be impacted by the changes defined in the documents gathered in step 5.1. Table 2 defines the key elements of a PRA. The review in this step only determines whether the information has the potential to impact the PRA.

If the change is believed to have an impact on the ranking of the top 200 components in the latest risk increase or risk achievement worth list for plant l components, then the change shall be identified as high priority and implemented before other pending lower priority changes. If there is no potential impact on the PRA, the completed Information Evaluation Form (Attachment 1) shall be l  : reviewed by the Supervisor, Nuclear Safety Group or his designee and then the L form given to the Secretary, Nuclear Safety Group for filing.

The plant change information reviewed in this step can be discarded if the change -

does not affect the PRA.

l l

The review ofinformation to support th.e update ofinitiating event and component reliability / availability data shall utilize References 3.7 and 3.9.

5.3 Undate Required Evaluation. If the reviewed plant change documents have a potential to impact the PRA, then the information must be evaluated further. The

, evaluation and the proposed change is given an unique number by the Secretary, 1;  ; Nuclear Safety Group. The numbering scheme for the PRA Change Packages

'should be PRACP-YY-XXXX where XXXX is a consecutive number from 1 to o -

9999 in each year YY.gThis number is recorded on the Information Evaluation Form as well as in the PRA Change Package form (Attachment 2) by the I

. Secretary, Nucicar Safety Group.

t L: The information may have an impact on one key element of the PRA and the change in this element may result in cascading impacts on other key elements of the PRA. Table 2 provides a guide of how changes in each element may impact other elements of the PRA. Only the primary impact of each element on other elements has been indicated, not the cascading impacts.

9 of 19 a

L o- -_ J

l l Attachment #2 PRA-REV-001 Revision 4,12/17/96 5.0 PROCEDURE (Continued)

A PRA Engineer, designated by Supervisor, Nuclear Safety Group or his designee shall review the subject plant documentation and prepare a complete PRA Change Package that includes details of all the changes required to each key element of

, the PRA (computer model and documentation). The package must contain all the calculations and identify all the changes to be implemented. Table 3 provides a recommended format for the PRA Change Package. The calculations for each key PRA element will be based on their respective procedures (Ref. 35 through 3.14).

Attachment 3 provides the list of the key PRA elements and their governing procedures.

The proposed changes may be implemented only after the review of the entire PRA Change Package by the Supervisor, Nuclear Safety Group or his designee.

The PRA Change Package must include the PRA computer model and the j l appropriate marked up data in the Living PRA Database. The review of the L Change Package must be conducted, and comments and resolutions recorded in accordance with the QA requirements of Reference 3.15.

The Supervisor, Nuclear Safety Group shall determine if a PRA Peer Review j Panel and/or Expert Review Panel review of the subject PRA Change Package is required prior to approval for implementation.

- 5.4 ' Perform Update. Only after approval of the PRA Change Package by the Supervisor, Nuclear Safety Group may the proposed change be implemented. The Supervisor, Nuclear Safety Group or his designee shall designate a PRA Engineer to implement the change in the PRA models and documentation. The updating i

( must be performed on a model and documentation that is copied from the current controlled model, leaving the current controlled model and documentation intact.

The PRA software to be used for analysis of the updated model must be installed on a PC, verified, validated and the installation report completed in accordance l with the QA requirements of Reference 3.16. '

l L

10 of 19 l

l

. _ _ - = _ _ _ _ .__ ___ _ -___ ___ -__- - _ - _ -__-__ - _ _ _ - -_ _-__ - - __-

Attachment #2 PRA-REV-001 Revision 4,12/17/96 5.0 PROCEDURE (Continued)

I 5.5 Review Impact on PRA. After the PRA Engineer has implemented the changes to the relevant models and documents, the implementation of the changes and the results must be reviewed by the Supervisor, Nuclear Safety Group or his designee.

The review of the results will be based upon Appendix B of Reference 3.4 and any subsequent NRC guidance issued regarding PRA content and quality. The review comments shall be ' documented in accordance with the QA requirements of Reference 3.15.

5.6 Install Updated Model and File PRA Change Package. Only upon approval of the update by the Supervisor, Nuclear Safety Group, vri!! the existing controlled model be rep %ed by the new updated model. Once installed for use as the current centrolled model, the new updated model and documentation shall be backed up in accordance with Records requirements in Section 6.0. The completed PRA Change Package shall be given to the Secretary, Nuclear Safety  !

Group for filing.

6.0 RECORDS 6.1 A copy of the entire Living PRA model documentation shall be maintained in electronic format accessable to the Nuclear Safety Group. The Supervisor,:

Nuclear Safety Group shall designate a person to maintain the documentation.

I 6.2 All Information Evaluation forms and PRA Change Packages shall be maintained i in the Nuclear Safety Group work area and a copy provided to CDM. The information evaluated in the Information Evaluations (e.g., DCPs, EOls, etc.)

need not be retained since it is always available from CDM. The Secretary, Nuclear Safety Group shall be responsible for maintaining the documentation in

.this step.

I 11 of 19 i

Attachment #2 PRA-REV-001 Revision 4,12/17/96 6.0 RECORDS (Continued) 6.3 An electronic capy of the current Living PRA Models, databases, and PRA analysis software shall be maintained on a network drive or other suitable means to be available to Nuclear Safety Group PRA Engineers. A backup copy of the current Living PRA Models and databases shall be retained on suitable electronic storage media (e.g., tapes). The electronic storage media shall be clearly labeled with the revision numbers and dates. All prior versions of the Living PRA Models and databases shall be retained on suitable storage media in the Nuclear Safety Group work area for the remaining life of the plants. The Supervisor, Nuclear Safety Group or his designee shall be responsible for implementing this step.

12 of 19

Attachment #2 PRA-REV-001 Revision 4,12/17/96 Figure 1. MAINTENANCE OF Tile PRA MODELS Monitor and Cellect Plant Chai.ge Information _

l Screen New Information l

Update Required?

If no, Stop here.

I Perform Update l

Assess Impact on PRA l

Evaluate Results l

Install Updated Model l

END l

l 13 of 19 I 1

l t

L___ _ - -_ . _

Attachment #2 i 1

PRA-REV-001 Revision 4,12/17/96 TABLE 1. SOURCES AND DOCUMENTS TO BE REVIEWED j 1

l Key PRA l I'" *"

Type Description d

(See Table 2)

AOI Abnormal Operating Instruction 4,5,8 DCP Design Change Package 2,3,4,5,7,8,9, 10 DCN Design Change Notice 2,c,4,5,7,8,9, 10 EOI Emergency Operating Instruction 4,5,8 FCN Field Change Notice 2,3,4,5,7,8,9, 10 FIDCN Field Interim Design Change Notice 2,3,4,5,7,8,9, 10 I IDCN Interim Design Change Notice 2,3,4,5,7,8,9, 10 OP Operating Procedure 4,5,8 I PCN Procedure Change Notice 4,5,8 LER Licensee Event Report 2, 4 l l

l 1

l 1

14 of 19

i Attachment #2 l PRA-REV-001 Revision 4,12/17/96 l TABLE 2. KEY ELEMENTS OF A LIVING PRA Other Primary PRA I leinents Impacted K l Description of Key Element Element No. Key Element Modified 1 Basic Event Database (in software) --

4 2 Initiating Events 1 ,

3 Dependency Matrices 2,4,5 l 4 Event Tree and Plant Models 5 5 System Success Criteria / System Analysis 1, 2, 4 l

6 Component Failure and Unavailability Data Analysis 2, 5 7 Thermo-Hydraulic Analysis 4,5,8 8 Human Actions Analysis - 1, 2, 4, 5 9 Seismic Analysis 1 10 Fire Analysis 1 11 Other External Events 1 12 Containment Model 1 13 Release Model 1 l

15 of 19

Attachment #2 PRA-REV-001 Revision 4,12/17/96 TABLE 3. FORMAT FOR PRA CIIANGE PACKAGE TITLE SHEET TABLE OF CONTENTS

1.0 INTRODUCTION

2.0 MODEL CHANGES FOR REVISION 3.0 RESULTS 4.0 VERIFICATION VALIDATION Appendix A Databases for Model Appendix B Detailed calculations, models, from Section 2.0 1

1 i

)

16 of 19 l

l.

I.

L i -

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Attachment #2 PRA-REV-001 l Revision 4,12/17/96 TABLE 4. DOCUMENTS USED IN TIIE PRA MODELS FOR INITIATING EVENTS AND COMPONENT DATA l INITIATING EVENTS

1. Monthly Reports 1

4 l 2. LERs l

3. Systems Analyses Doeuments for Loss of Support Systems Initiated Events

4. Seismic Analysis Reports from Section 3.10 in Reference 3.2, i i

5. Fire Analysis Report from Section 4.9 in Reference 3.2.

I

6. Offsite Hazards Reports l

PLANT COMPONENT DATA ANALYSIS

1. Work Orders / Action Reports

2. Maintenance Requests / Maintenance Ordersfrag-out Orders

3. Trouble Reports / Discrepancy Reports

4. Component Operating Experience

5. Component Histories

6. Monthly Reports 17 of 19

i 1

Attachment #2 PRA-REV-001 ,

Revision 4,12/17/96 TABLE 4. DOCUMENTS USED IN TIIE PRA MODELS FOR INITIATING EVENTS l- AND COMPONENT DATA (continued)

7. LERs

8. Plant Power History

9. Control Room Logs

10. System Operating Procedures l

l l

l  !

j 1

i l

l l

l l

1 18 of 19 1

i

l I

l Attachment #2 PRA-REV-001 Revision 4,12/17/96 ATTACilMENT 1 INFORMATION EVALUATION FORM Information Identification No: l Impr.et on PRA:

1 l IfNo 0:

I i

Provide form to Secretary, Nuclear Safety Group for filing. l l

If Yes 0:

- Assigned PRA Change Package No.

(PRACP-YY-XXXX)

Potentially affects ranking of top 200 components on latest plant component risk increase listing?

l Yes O No O 1

l. If Yes, this is a high priority change and should be implemented I

prior to pending non-high priority changes.

l.  !

Reviewed b r: Date: j Approved by: Date: Supervisor, Nuclear (

1 l Safety Group 19 of 19 k'

l ui l N-  :

Attachment #2 PRA-REV-001 Revision 4,12/17/96 ATTACHMENT 2 PRA CHANGE PACKAGE l

Title:

Punch List item #:

(inapplicable)

Change No. PRACP-(PRACP-YY-XXXX) where XXXX is consecutive from 1 to 9999 Reason for change:

Proposed change:

t Prepared by: Date:

PRA Engineer NSG Supervisor Review:

PRA Review Panel Required: No O Yes O.

Expert Review Panel Required: No O Yes O Proposed Change reviewed by: Date:

Supervisor, Nuclear Safety Group or designee Change Implemented by: Date:

. PRA Engineer -

l PRA Living Database Updated: Date:

PRA Engineer l- Implementation Reviewed by: Date:

Supervisor, Nuclear Safety Group or designee (Ensure Punch List Database resolution date entered if applicable.)

, Approved by: Date:

Supervisor, Nuclear Safety Group 20 of 19 1

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Attachment #2 PRA-REV-001 Revision 4,12/17/96 i

1 l

l l

I l

t i

21 of 19

i l

Attachment #2 PRA-REV-001 Revision 4,12/17/96 ATTACHMENT 3 RELEVANT PROCEDURES FOR EVALUATION OF KEY PRA ELEMENTS pg Applicable PRA Description Procedures from Element No.

l Sectmn 2 1 Database for plant model References 3.5 through 3.14 2 Initiating Events Reference 3.7 3 Dependency Matrices References 3.5, 3.6, 3.8 4 Event Tree and Plant Models Reference 3.8 5 System Success Criteria / System Analysis References 3.5, 3.6 6 Basic Events Database and Data Analysis Reference 3.9, 310 7 Thermo-Hydraulic Analysis --

8 Human Actions Analysis Reference 3.11 9 Seismic Analysis --

10 Fire Analysis --

11 Other External Events --

12 Containment Model --

13 Release Model --

14 Plant Model Quantification Reference 3.13 15 Sensitivity Analysis Reference 3.14 1

l 22 of 19 I

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I ATTACHMENT #3 l SAN ONOFRE UNITS 2/3 LOSS OF OFFSITE POWER & STATION BLACKOUT EVENT TREES l

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ATTACHMENT #4 SAN ONOFRE UNITS 2/3 LOSS OF OFFSITE POWER and

/ STATION BLACKOUT EVENTS TOP 50 CUTSETS l

f l

l Attachment #4 Top 50 LOP /SBO Cutsets Pagei Safety Monitor 5.27a Configuration

Title:

LOP & SBO INITIATING EVENT CUTSETS (CROSS-TIE MOD CREDITED)

Daterfime of Calculation: 6/1/98 4:24:12 PM Unit 2 Cutsets from last calculation:

Cutsets sorted by core damage frequency CLISet Nc.ter: 1 CutSet Frequency = 6.10E-07 INIT-LOP 1.07E-01 INITIATING EVENT- LOSS OF O'TSITE POWER L-ilCNSBOMANU 6.50E-02 OP FT MAN CTL AFW FLOW VALVES W/l 2.5 TO 8.0 HOURS M-IICBCBFAILU 5.00E-01 OPER FT RESPOND TO BAT 1 CHRG B2 FAILURE GIVEN PREV RESP Fall M-HCillTEMP-U l.00E+00 NO OPER RESPONSE TO ECCS 111 TEM" ALARM GIVEN PREV RESP Fall M-IICNOSIAS-U l.50E-03 NO OPRT RESPONSE TO HIGH TEMPERATURE ALARM-ESF SWGR/DIST U-HCOSPWR60U l.17E-01 NONRECOVERY OF OFFSITE POWER IN 60 MINUTES CutSet Number: 2 CutSet Frequency = 4.04E-07 INIT-LOP 1.07E-01 INITIATING EVENT - LOSS OF OFFSITE POWER L-HCTP140-U 4.30E-02 OPER FT MANUALLY OPERATE TD AFW PUMP (NO DC PWR @ 8H)

M-HCBCBFAILU 5.00E-U 1 OPER FT RESPOND TO BATT CllRG B2 FAILURE GIVEN PREV RESP FAIL l M-IICHITEMP-U l.00E+00 NO OPER LESPONSE TO ECCS Hi TEMP ALARM GIVEN PREV RESP Fall 1 M-HCh0SIAS-U l.50E-03 NO OPRT RESPONSE TO HIGil TEMPERATURE ALARM-ESF SWGR/DIST U-IICOSPWR60U l.17E-01 NONRECOVERY OF OFFSITE POWER IN 60 MINUTES l

CutSet Number: 3 CutSet Frequency = 3.15E-07 INIT-LOP 1.07E-01 INITIATING EVENT - LOSS OF OFFSITE POWER L-IICNROSP811U l.00E+00 ADJ FACTOR FOR OSP RECOVERY WITil Till FAILURES DURING LOP ,

L-liCCSTMU--U 3.00E-05 OPERATOR FAILS TO PROVIDE T121 (CST) MAKEUP PER PROCEDURE U-HCOSPWRH1U 9.80E-02 NONRECOVERY OF OFFSITE POWER AT T=8 liOURS

)

CutSet Number: 4 CutSet Frequency = 1.07E-07 INIT-LOP 1.07E-01 INITIATING EVENT- LOSS OF OFFSITE POWER L-HCTP140-U 4.30E-02 OPER FT MANUALLY OPERATE TD AFW PUMP (NO DC PWR @ 811) l' L-MP504--IDR 9.57F-03 MTR-DRIVEN PP 504 FT RUN FOR 24 HR - TRAIN B U-C2A0418-P 3.00E-03 BKR (CNTL) 4160V A0418 FT OPEN ON DEMAND U-OOPCLOOP-Z 8.12E-01 LOOP EVENT PLANT. CENTERED LOOP- NO NON-lE PWR AVAILABLE Cutset Number: 5 Cutset Frequency = 9.67E-08 INIT LOP 1.07E-01 INITIAT!NG EVENT - LOSS OF OFFSITE POWER IAP140-lDR I.03E-02 TUr.BINE DRIVEN PP 140 FT RUN rOR 24 H L-TP140NRIDR 1.00E+00 TURBINE DRIVEN PP 140 FT TO RUN IS NOT RECOVERABLE M-HCBCBFAILU 5.00E-01 OPER FT RESPOND TO BATT CHRG B2 FAILURE GIVEN PREV RESP FAIL M-IICHITEMP-U l.00E+00 NO OPER RESPONSE TO ECCS Hi TEMP ALARM GIVEN PREV RESP FAIL M-IICNOSIAS-U l.50E-03 "O OPRT RESPONSE TO lilCH TEMPERNfURE ALARM-ESF SWGR/DIST ,

U-HCOSPWR60U l.17E-01 NONRECOVERY OF OFI SITE POWER IN 60 MINIJTES

Attachment #4 Top S0 LOP /SBO Cutsets Page 2 CutSet Number: 6 CutSet Frequency = 7.66E-08 INIT LOP 1.07E-01 INITIATING EVENT - LOSS OF OFFSITE POWER U-C2A0418--P 3 00E-03 BKR (CNTL) 4160V A0418 FT OPEN ON DEMAND U-C2A0618-P 3.00E-03 BKR (CNTL)4160V A0618 FT OPEN ON DEMAND U-HCOSPWRHlU 9.80E-02 NONRECOVERY OF OFFSITE POWER AT T=8 HOURS U-OOPCLOOP-Z 8.12E-01 LOOP EVENT PLANT. CENTERED LOOP - NO NON-1E PWR AVAILABLE CutSet Number: 7 CutSet Frequency = 6.54E-08 IN:T-LOP 1.07E-01 INITIATING EVENT - LOSS OF OFFSITE POWER l-TP140--M 6.97E-03 TURBINE-DRIVEN PP 140 OOS FOR MAINTENANCE M-HCBCBFAILU 5.00E-01 OPER FT RESPOND TO BATT CHRG B2 FAILURE GIVEN PREV RESP FAIL M-HCHITEMP-U l.00E+00 NO OPER RESPONSE TO ECCS HI TEMP ALARM GIVEN PREV RESP FAIL M-IICNOSIAS-U l.50E-03 NO OPRT RESPONSE TO HIGH TEMPERATURE ALARM-ESF SWGR/DIST U-HCOSPWR60U l.17E-01 NONRECOVERY OF OFFSITE POWER IN 60 MINUTES CutSet Number: 8 CutSet Frequency = 5.38E-08 INIT-LOP 1.07E-01 INITIA FING E'/ENT - LOSS OF OFFSITE POWER L-HCHV4716PU l.00E+00 OPERATOR FA LS TO MANUALLY OPEN HV4716 LOCALLY AFTER FAULT-L-MV4716--P 5.73E-03 MTR-OPERATED VLV 4716 I T OPEN ON DEMA'ND M-HCBCBFAILU 5.00E-01 OPER FT RESPOND TO BATT CHRG B2 FAILURE GIVEN PREV RESP Fall M-HCHITEMP-U l.00E+00 NO OPER RESPONSE TO ECCS HI TEMP ALARM GIVEN PREV RESP FAIL M-hCNOSIAS-U l.50E-03 NO OPRT RESPONSE TO HIGH TEMPERATURE ALARM-ESF SWGR/DIST U HCOSPWR60U l.17E-01 NONRECOVERY OF OFFSITE POWER IN 60 MINUTES Cutset Number: 9 CutSet Frequency = 4.00E-08 INIT-LOP 1.07E-0 l INITIATING EVENT - LOSS OF OFFSITE POWER L-TP140NR-S 4.26E-03 Pl40 FAILS TO START AND IS NOY RECOVERABLE M-HCBCBFAILU : 5.00E-01 OPER FT RESPOND TO BATT CHRG B2 FAILUkE GIVEN PREV RESP FAIL M-HCHITEMP-U l.00E+00 NO OPER RESPONSE TO ECCS HI TEMP ALARM GIVEN PREV RESP FAIL M-HCNOSIAS-U l.50E-03 NO OPRT RESPONSE TO HIGH TEMPERATURE ALARM-ESF SWGR/DIST U-hCOSPWR60U l.17E-01 NONRECOVERY OF OFFSITE POWER IN 60 MINUTES CutSet Number 10 Cutset Frequency = 3/e5E-08 INIT-LOP 1.07E-01 INITIATING EVENT- LOSS OF OFFSITE POWER

. L-HCTP140-U 4.30E-02 OPER FT MANUALLY OPERATE TD AFW PUMP (NO DC PWR @ 8H)

L-MP504-M 3.35E-03 MTR-DRIVEN PP 504 OOS FOR MAINTENANCE - TRAIN B U-C2A0418-P 3.00E-03 BKR (CNTL)4160V A0418 FT OPEN ON DEMAND U-OOPCLOOP-Z 8.12E-Oi LOOP EVENT PLANT. CENTERED LOOP - NO NON-lE PWR AVAILABLE

' CutSet Number: 11 CutSet Frequency = 3.36E-08

- INIT-LOP 1.07E-01 INITIATING EVENT- LOSS OF OFFSITE POWER L-C2A0603-N 3.00E-03 BKR (CNTL) ;160V 2A0603/3A0612 FT CLOSE ON DEMAND L-HCTP140-U 4.30E-02 OPER FT MANUALLY OPERATE TD AFW PUMP (NO DC PWR @ 8H)

U-C2A3418-P 3.00E-03 BKR (CNTL)4160V A0418 FT OPEN ON DEMAND U-OOPCLOOP-Z 8.12E-Ol' I.OOP EVENT PLANT. CENTERED LOOP NO NON 1E PWR AVAILABLE L _- _ - - _ - _ _ - - - _ _ - .-.

Attachment #4 Top 50 LOP /SBO Cutsets Page 3 l CutSet Number: 12 CutSet Frequency = 2.70E-08 INIT-LOP 1.07E-01 INITIATING EVENT - LOSS OF OFFSITE POWER L-HCliV4716PU l.00E+00 OPERATOR FAILS TO MANUALLY OPEN HV4716 LOCALLY AFTER FAULT L-MV4716---M 2.88E-03 MTR-OPTD VLV 4716 OOS FOR MAINTENANCE M-HCBCBFAILU 5.00E-01 OPER FT RESPOND TO BATF CHRG B2 FAILURE GIVEN PREV RESP FAIL M-HCillTEMP-U l.00E+00 NO OPER RESPOI4SE TO ECCS Hi TEMP ALARM GIVEN PREV P. ESP FAIL M-HCNOSIAS-U l.50E-03 NO OPRT RESPONSE TO HIGli TEMPERATURE ALARM-ESF SWGR/DIST U-HCOSPWR60U l.17E-01 NONRECOVERY OF OFFSITE POWER IN 60 MINUTES CutSet Number: 13 CutSet Frequency = 2.57E-08 INIT-LOP 1.07E-01 INITIATING EVENT. LOSS OF OFFSITE POWER L-MP504-IDR 9.57E-03 MTR-DRIVEN PP 504 FT RUN FOR 24 HR - TRAIN B L-TPl40-lDR l.03E-02 TURBINE DMVEN PP 140 FT RUN FOR 24 H L-TP140NRIDR 1.00E+00 TURBINE DRIVEN PP 140 FT TO RUN IS NOT RECOVERABLE U-C2A0418-P 3.00E-03 BKR (CNTL)4160V A0418 FT OPEN ON DEMAND U-OOPCLOOP-Z 8.12E-01 LOOP EVFNT PLANT- CENTERED LOOP - NO NON-1E PWR AVAILABLE CutSet Number: 14 CutSet Frequency = 2.54E-08 INIT-LOP 1.07E-01 INITIATING EVENT - LOSS OF OFFSITE POWER l 1

H-HASMPLT-X 7.90E-04 EMERGENCY SUMP LATCH LEFT OPEN I Y-OORV-DEM-Z l .00E-01 POST-TRIP PRESSURIZER SAFETY VALVES DEMANDED Y-RV0200---N 3.00E-03 RELIEF VLV SPRING LOAD 0200 FT CLOSE ON DEM CutSet Number: 15 CutSet Frequency-2.54E-08 INIT-LOP 1.07E-01 INITIATING EVENT- LOSS OF OFFSITE POWER H-HASMPLT--X 7.90E-04 EMERGENCY SUMP LATCH LEFT OPEN Y-OORV-DEM-Z l.00E-01 POST-TRIP PRESSURIZER SAFETY VALVES DEMANDED Y-RV0201--N 3.00E-03 RELIEF VLV SPRING LOAD 0201 FT CLOSE ON DEM l CutSet Number: 16 CutSet Frequency = 2.02E-08 INIT-LOP 1.07E-01 INITIATING EVENT LOSS OF OFFSITE POWER

( L-HCTP:40--U 4.30E-02 OPER FT MANUALLY OPERATE TD AFW PUMP (NO DC PWR @ 8H) )

l L-MP504--S 1.80E-03 MTR-DRIVEN PP 504 FT STAl?T ON F EMAND - TRAIN B i U-C2A0418-P 3.00E-03 BKR (CNTL) 4160V A0418 FT OPEN ON DEMAND l ]

U-OOPCLOOP-Z 8.12E-01 LOOP EVENT PLANT- CENTERED LOOP - NO NON-1E PWR AVAILABLE '

i CutSet Nuttber 17 CutSet frequency = 1.88E-08 INIT-LOP 1.07E-01 INITIATING EVENT - LOSS OF OFFSITE POWER L-SV4700--J 2.00E-03 SOLENOID VALVE 4700 FT ACT/DE-ACT ON DEMAND M-HCBCBFAILU 5.00E-01 OPER FT RESPOND TO BATT CilRG B2 FAILURE GIVEN PREV RESP Fall M-HCHITEMP-U l.00E+00 NO OPER RESPONSE TO ECCS Hi TEMP ALARM GIVEN PREV RESP FAIL M-IICNOSIAS-U l.50E-03 NO OPRT RESPONSE TO HIGli TEMPERATURE ALARM-ESF SWGR/DIST U-HCOSPWR60U l.17E-01 NONRECOVERY OF OFFSITE POWER IN 60 MINGTES CutSet Number: 18 CutSet Frequency = 1.88E-08 I

_ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ . . _ _ _ _ ..____..___________...__________J

Attachment #4 Top 50 LOP /SHO Cutsets INIT-LOP 1.07E-01 INITIATING EVENT - LOSS OF OFFSITE POWER ll-MP019---M 3 34E-02 PUMP 019 UNAVAILABLE DUE TO MAINTENANCE / TEST M-CllE-336--M l.75E-02 CHILLER E-336 OOS FOR MAINTENANCE - LOOP A ESF IIVAC Y-OORV-DEM-Z l.00E-01 POST-TRIP PRESSURIZER SAFETY VALVES DEMANDED Y-RV0201--N 3.00E-03 RELIEF VLV SPRING LOAD 0201 FT CLOSE ON DEM Page 4 CutSet Number: 19 CutSet Frequency = 1.88E-08 INIT-LOP 1.07E-01 INITIATING EVENT- LOSS OF OFFSITE POWER H-MP019--M 3.34E-02 PUMP 019 UNAVAILABLE DUE TO MAINTENANCE / TEST M-CHE-336--M 1.75E-02 CHILLER E-336 OOS FOR MAINTENANCE - LOOP A ESF IIVAC Y-OORV-DEM-Z l.00E-01 POST-TRIP PRESSURIZER SAFETY VALVES DEMANDED Y-RV0200--N 3.00E-03 RELIEF VLV SPRING LOAD 0200 ET CLOSE ON DEM CutSet Number: 20 CutSet Frequency = 1.81E-08 INIT-LOP LO7E-01 INITIATING EVENT - LOSS OF OFFSITE POWER L-HCNSBOMANU 6.50E-02 OP FT MAN CTL AFW FLOW VALVES W/I 2.5 TO 8.0 HOURS

. M-CllCC0001-S 4.44E-04 COMMON CAUSE FAILURE -CH-S-->E-335,E-336 M-HCGASFANSZ 5.00E-02 OPERATOR FAILS TO ESTABLISH ALTERNATE COOLING IN ESF ROOMS U HCOSPWR60U l.17E-01 NONRECOVERY OF OFFSITE POWER IN 60 MINUTES I

CutSet Number: 21 CutSet Frequency = 1.80E-08 INIT LOP 1.07E-01 ' INITIATING EVENT - LOSS OF OFFSITE POWER L-MPl41--IDR 6.72E-03 MTR-DRIVEN PP 141 FT RUN FOR 24 HR - TRAIN A L-TP140-1 DR 1.03E-02 TURBINE DRIVEN PP 140 FT RUN FOR 2411 L-TPl40NRIDR 1.00E+00 TURBINE DRIVEN PP 140 FT TO RUN IS NOT RECOVERABLE U-C2A06 8--P 3.00E-03 BKR(CNTL) 4160V A0618 FT OPEN ON DEMAND U-OOPCLOOP-Z 8.12E-01 LOOP EVENT PLANT- CENTERED LOOP - NO NON-lE PWR AVAILABLE l CutSet Number: 22 CutSe: Frequency = 1.74E-08 INIT-LOP 1.07E-01 INITIATING EVENT- LOSS OF OFFSITE POWEE

L-MP504-1DR 9.57E-03 MTR-DRIVEN PP 504 FT RUN FOR 24 HR - TRAIN B L-TP140-M 6.97E-03 TURBINL-DRIVEN PP 140 OOS FOR MAINTENANCE

! U-C2A0418--P 3.00E-03 BKR (CNTL) 4160V A04)8 FT OPEN ON DEMAND

! U-OOPCLOOP-Z 8.12E-01 LOOP EVENT PLANT CENTERED LOOP- NO NON-1E PWR AVAILABl.E

, CutSet Number: 23 CutSet Frequency = 1.63E-08 l INIT-LOP 1.07E-01 INITIATING EVEN F - LOSS OF OFFSITE POWER L-TK120--D 1.30E.06 CONDENSATE STORAGE TANK T121 UNAVAILABLE U-HCOSPWR60U l.17E-01 NONRECOVERY OF OFFSITE POWER IN 60 MINUTES CutSet Number: 24 Cutset Frequency = 1.63E-08 INIT-LOP 1.07E-01 INITIATING EVENT - LOSS OF OFFSITE POWER L-TK121-D - 1.30E-06 CONDENSATE STORAGE TANK UNAVAILABLE U-HCOSPWR60U l.17E-01 NONRECOVERY OF OFFSITE POWER IN 60 MINUTES CutSet Number: 25 CutSet Frequency = 1.57E-08 INIT-If.' l.07E-01 INITIATINC EVENT- LOSS OF OFFSITE POWER H-MP010--iDR 2.80E-02 MTR DRIVEN PP 019 FT RUN FOR 24 IIR M-CHE-336-M L75E-02 CHILLER E-336 OOS FOR MAINTENANCE - LOOP A ESF HVAC Y-OORV-DEM-Z l.00E-01 POST-TRIP PRESSURIZER SAFETY VALVES DEMANDED Y-RV0201-N 3.00E-03 RELIEF VLV SPRING LOAD 0201 FT CLOSE Oh DEM

Attachment #4 Top 50 LOP /SBO Cuisets {

CutSet Number. 26 CutSet Frequency = 1.57E-08 '

INIT-LOP 1.07E-01 INITIATING EVENT- LOSS OF OFFSITE PGWER H-MP019--I DR 2.80E-02 MTR-DRIVEN PP 019 FT RUN FOR 24 HR M-CliS-336--M 1.75E-02 CHILLER E-336 OOS FOR MAINTENANCE - LOOP A ESF HVAC Y-OORV-DEM-Z l .00E-01 POST-TRIP PRESSURIZER SAFETY VALVES DEMANDED Y RV0200-- N 3.00E-03 KELIEF VLV SPRING LOAD 0200 FT CLOSE ON DEM Page 5 CutSet Number: 27 CutSet Frequency = 1.48E-08 INIT-LOP 1.07E-01 INITIATING EVENT - LOSS OF OFFSITE POWER L-HCNSBOMANU 6.50E-02 OP FT MAN CTL AFW FLOW VALVES W/l 2.5 TO 8.0 HOURS M-CHE-336--M 1.75E-02 CHILLER E-336 OOS FOR MAINTENANCE. LOOP A ESF HVAC M-HCGASFANSZ 5.00E-02 OPERATOR FAILS TO ESTABLISH ALTERNATE COOLING IN ESF ROOMS U-C2A0618-P 3.00E-03 BKR (CNTL) 4160V A0618 FT OPEN ON DEMAND '

U OOPCLOOP-Z 8.12E-01 LOOP EVENT PLANT- CENTERED LOOP - NO NON-1E PWR AVAILABLE l

CutSet Number: 28 CutSet Frequency = 1.48E-08 INIT-LOP 1.07E-01 INITIATING EVENT- LOSS C'rOFFSITE POWER L-HCNSBOMANU 6.50E-02 OP FT MAN CTL AFW FLOW VALVES W/I 2.5 TO 8.0 HOURS M-CHE-335--M 1.75E-02 CillLLER E-335 OOS FOR MAINTENANCE - LOOP B ESF HVAC M-HCGASFANSZ 5.00E-02 OPERATOR FAILS TO ESTABLISH ALTERNATE COOLING IN ESF ROOMS I U-C2A0418--P 3.00E-03 BKR (CNTL) 4160V A0418 FT OPEN ON DEMAND U-OOPCLOOP-Z 8.12E-Oi &OOP EVENT PLANT- CENTERED LOOP- NO NON-1E PWR AVAILABLE CutSet Number: 29 CutSet Frequency = 1.43E-08 INIT-LOP 1.07E-01 INITIATING EVENT - LOSS OF OFFSITE POWER L-HCHV4716PU l.00E+00 OPERATOR FAILS TO MANUALLY OPEN HV4716 LOCALLY AFTER FAULT L-MP504-1DR 9.57E-03 MTR-DRIVEN PP 504 FT RUN FOR 24 HR - TRAIN B L-MV4716.--P 5.73E-03 MTR-OPERATED VLV 4716 FT OPEN ON DEMAND l U-C2A0418-P 3.00E-03 BKR(CNTL)4160V A0418 FT OPEN ON DEMAND i U-OOPC.OOP-Z S.12E-01 LOOP EVENT PLANT- CENTERED LOOP - NO NON-1E PWR AVAILABLE CutSet Number: 30 Cutset Frequency = 1.43E-08 INIT-LOP 1.07E-01 INITIATING EVENT- LOSS OF OFFSITE POWER M-CHCC0001-S 4.44E-04 COMMON CAUSE FAILURE -CH-S-->E-335,E-336 Y OORV-DEM-Z l.00E-01 POST-TRIP PRESSURIZER SAFETY VALVES DEMANDED Y-RV0201--N 3.00E-03 RELIEF VLV SPRING LOAD 0201 FT CLOSE ON DEM CutSet Number: 31 CutSet Frequency = 1.43E-08 INIT-LOP 1.07E-01 INITIATING EVENT - LOSS OF OFFSITE POWER M-CHCC0001-S 4.44E-04 COMMON CAUSE FAILURE -CH-S-->E-? '5,E-336 Y-OORV-DEM-Z l.00E-01 POST-TRIP PRESSURIZER SAFETY VALVES DEMANDED Y-RV0200-N 3.00E-03 RELIEF VLV SPRING LOAD 0200 FT CLOSE ON DEM CutSet Number: 32 CutSet Frequmcy = 1.39E-08 INIT-LOP 1.07E-01 INITIATING EVENT - LOSS OF OFFSITE POWER H-MPCC0001-5 4.33E-04 COMMON CAUSE FA! LURE -MP-S -> 017,018, AND 019 Y-OORV-DEM-Z l.00E ')1 POST-TRIP PRESSURIZER SAFETY VALVES DEMANDED Y-RV0200-N 3.00E-03 RELIEF VLV SPRING LOAD 0200 FT CLOSE ON DEM

Attachment #4 Top 50 LOP /SBO Cutsets I

CutSet Number: 33 Cutbet 1 requency = 1.39E-08 l INIT LOP 1.07E-01 INITIATING EVENT - LOSS OF OFFSITE POWER H-MPCC0001-S 4.33E-04 COMMON CAUSE FAILURE-MP-S > 017,018, AND 019 Y .OORV-DEM-Z l.00E-01 POST-TRIP PRESSURIZER SAFETY VALVES DEMANDED Y-RV0201--N 3.00E-03 RELIEF VLV SPRING LOAD 0201 FT CLOSE ON DEM Page 6 Cut:',et Number: 34 CutSet Frequency = 1.38E-08  :

INIT-LOP 1.07E-01 INITIATING EVENT- LOSS OF OFFSITE POWER L-MPCC0001-S 1.07E-04 COMMON CAUSE FAILURE -MP-S > 141 AND 504 (AFW MDPS)  ;

L-TPl40-IDR 1.03E-02 TURDINE DRIVEN PP 140 FT RUN FOR 24 H  !

L-TP140NR1DR 1.00E+00 TURBINE DRIVEN PP 140 FT TO RUN IS NOT RECOVERABLE U-HCOSPWR60U l.17E-01 NONRECOVERY OF OFFSITE POWER IN 60 MINUTES CutSet Number: 35 CutSet Frequency = 1.22E-08 l INIT-LOP 1.07E-01 INITIATING EVENT - LOSS OF OFFSI TE POWER L-MPl41--IDR 6.72E-03 MTR-DRIVEN PP 141 FT RUN FOR 24 HR - TRAIN A L-TP140--M 6.97E-03 TURBINE-DRIVEN PP 140 OOS FOR MAINTENANCE U-C2A0618--P 3.00E-03 BKR(CNTL)4160V A0618 FT OPEN ON DEMAND U-OOPCLOOP-Z 8.12E-01 LOOP EVENT PLANT- CENTERED LOOP - NO NON-1E PWR AVAILABLE CutSet Number: 36 CutSet Frequency = 1.20E-08 INIT-LOP 1.07 E-01 INITIATING EVENT - LOSS OF OFFSITE POWER L-HCTP140--U 4.30E-02 OPER FT MANUALLY OPEP ATE TD AFW PUMP (NO DC PWR @ 8H)

M-CHCC0001-S 4.44E-04 COMMON CAUSE FAILURE -CH-S- >E-335,E-336 M-HCGASFANS2 5.00E-02 OPERATOR FAILS TO ESTABLISH ALTERNATE COOL ING IN ESF ROOMS U-HCOSPWR60U l.17E-01 NONRECOVERY OF OFFSITE POWER IN 60 MINUTE ,

CutSet Number: 37 CutSet Frequency = 1.12E-08 INIT-LOP 1.07E-01 INITIATING EVENT - LOSS OF OFFSITE POWER L-CV547--N 1.00E-03 CHECK VLV 547 FT CLOSE ON DEMAND leHCTP140--U 4.30E-02 OPER FT MANUALLY OPERATE TD AFW PUMP (NO DC PWR @ 8H)

' ~ C2A0418-P.

3.00E-03 BKR (CNTL)4160V A0418 FT OPEN ON DEMAND

! u-OOPCLOOP-Z 8.12E-01 LOOP CVENT PLANT- CENTERED LOOP - NO NON-1E PWR AVAILABLE

! CutSet Number. 38 CutSet Frequency = 1.10E-08 INIT-LOP 1.07E-01 INITIATING E'/ENT - LOSS OF OFFSITE POWER K-HCEFAS--U l.00E+00 OP FT MANUALLY ACTUATE EFAS PER PROCDURE K-OOEFAS-2BZ 9.80E-04 EFAS 2B INDEPENDENT SIGNAL FAILURE L-HCTP140-U 4.30E-02 OPER FT MANUALLY OPERATE TD AFW PUMP (NO DC PWR @ 8H)

U-C2A0418-P 3.00E-03 BKR(CNTL)4160V A0418 FT OPEN ON DEMAND U-OOPCLOOP-Z 8.12E-01 LOOP EVENT PLANT- CENTERED t.OOP - NO NON-1E PWR AVAILABLF CutSet Number: 39 CutSet Frequency = 1.07E-08 INIT-LOP 1.i T-01 INITI ATING SVENT - LOSS OF OFFSITE POWER M-HCBCBFAILU 5. si-01 OPER FT RESPOND TO BATT CHRG B2 FAILURE GIVEN PREV RESP Fall M-HCHITEMP-U l.00E+00 NO OPER RESPONSE TO ECCS Hi TEMP ALARM GIVEN PREV RESP Fall

Attachment #4 Top 50 LOP /SBO Cutsets M-HCNOSIAS-U l.50E-03 NO OPRT RESPONSE TO lilGli TEMPERATURE ALARM-ESF SWGR/DIST T-AVCC0001-N 1.14E.03 COMMON CAUSE FAILURE-AV-N > 8419 AND 8421 T-HCSOADV--U l.00E+00 OPERATOR FAILS TO ISOLATE STUCK OPEN ADV U-ilCOSPWR60U l.17E-01 NONRECOVERY OF OFFSITE POWER IN 60 MINUTES CutSet Number: 40 CutSet Frequency = 1.06E-08 INIT-LOP 1.07E-01 INITIATING EVENT- LOSS OF OFFSITE POWER L-MP504--I DR 9.57E-03 MTR-DRIVEN PP 504 FT RUN FOR 24 HR - TRAIN B L-TPl40NR--S 4.26E-03 P140 FAILS TO START AND IS NOT RECOVERABLE U-C2A0418--P 3.00E-03 BKR(CNTL)4160V A0418 FT OPEN ON DEMAND U-OOPCLOOP-Z 8.12E-01 LOOP EVENT PLANT- CENTERED LOOP - NO NON-1E PWR AVAILABLE Page 7 CutSet Number: 41 CutSet Frequency = 1.00E-08 INIT-LOP 1.07E-01 INITI ATING EVENT - LOSS OF OFFSITE POWER L-HCHV4716PU l.00E+00 OPERATOR FAILS TO MANUALLY OPEN HV4716 LOCALLY AFTER FAULT L-MP141--I DR 6.72E-03 MTR-DRIVEN PP 141 FT RUN FOR 24 HR - TRAIN A L-MV4716--P 5.73E-03 MTR-OPERATED VLV 4716 FT OPEN ON DEMAND U-C2A0618-P 3.00E-03 BKR(CNTL)4160V A0618 FT OPEN ON DEMAND U-OOPCLOOP-Z 8.12E-01 LOOP EVENT PLANT- CENTERED LOOP - NO NON-lE PWR AVAILABLE CutSet Number: 42 CutSet Frequency = 9.81E-09 INIT-LOP 1.07E-01 INITIATING EVENT - LOSS OF OFFSITE POWER L-HCTP140-U 4.30E-02 OPER FT MANUALLY OPERATE Tn AFW PUMP (NO DC PWR @ 8H)

M-C11E-336--M 1.75E-02 CHILLER E-336 OOS FOR MAINTEF ANCE- LOOP A ESF HVAC M-HCGASFANSZ 5.00E-02 OPERATOR FAILS TO ESTABLISil ALTERNATE COOLING IN ESF ROOMS U-C2A0618-P 3.00E-03 BKR (CNTL)4160V A0618 FT OPEN ON DEMAND U-OOPCLOOP-Z 8.12E-01 LOOP EVENT PLANT- CENTERED LOOP - NO NON-1E PWR AVAILABLE CutSet Number: 43 CutSet Frequency = 9.81E-09 INIT-LOP 1.07E-01 INITIATING EVENT - LOSS OF OFFSITE POWER L-HCTP140-U 4.30E-02 OPER FT MANUALLY OPERATE TD AFW PUMP (NO DC PWR @ 8H)

M-CHF-335-M 1.75E-02 CHILLER E 335 OOS FOR MAINTENANCE - LOOP B ESF HVAC M-HCGASFANSZ 5.00E-02 OPERATOP FAILS TO ESTABLISH ALTERNATE COOLING IN ESF ROOMS U-C2A0418--P 3.00E-03 BKR (CNTL) 4160V A0418 FT OPEN ON DEMAND U-OOPCLOOP-Z 8.12E-01 LOOP EVENT PLANT- CENTERED LOOP - NO NON-1E PWR AVAILABLE i l

CutSet Number: 44 CutSet Frequency = 9.34E-09 l INIT-LOP 1.07E-01 INITIATING EVENT - LOSS OF OFFSITE POWER L MPCC0001-S 1.07E-04 COMMON CAUSE FAILURE -MP-S -> 141 AND 504 (AFW MDPS)

L-TP140---M 6.97E-03 TURBINE-DRIVEN PP 140 OOS FOR MAINTENANCE U-HCOSPWR60U 1.17E-01 NONRECOVERY OF OFFSITE POWER IN 60 MINUTES CutSet Number: 45 CutSet Frequency = 9.23F 09 INIT-LOP 1.07E-01 INITIATING EVENT- LOSS OF OFFSITE POWER l L.HV4762---M 4.87E-02 CONTROL VALVE 4762 OOS DUE TO MAINTENANCE l l

L-HV4763--M 4.87E-02 CONTROL VALVE 4 /63 CLOSED DUE TO MAINTENANCE i l L-MVCC0012-P 3.llE-04 COMMON CAUSE FAILURE -MV-P-> 4705,4706,4712 AND 4713 i U-HCOSPWR60U l .17E-01 NONRECOVERY OF OFFSITE POWER IN 60 MINUTES CutSet Number: 46 CutSet Frequency = 8.99E-09 l

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Attachment #4 Top 50 LOP /SBO Cutsets INIT-LOP 1.07E-01 INITIATING EVENT - LOSS OF OFFSITE POWER

. L-MP141---M 3.35E-03 MTR-DRIVEN PP 14100S FOR MAINTENANCE TRAIN A L-TPl40--LOR 1.02-02 TURBINE DRIVEN PP 140 FT RUN FOR 24 H L-TPl40NR1DR 1.00E+00 TURBINE DRIVEN PP 140 FT TO RUN IS NOT RECOVERABLE U-C2A0618--P 3.00E-03 BKR (CNTL) 4160V A0618 FT OPEN ON DEMAND U-OOPCL OOP-Z 8.12E-01 LOOP EVENT PLANT- CENTERED LOOP - NO NON-l E PWR AVAILABLE

. CutSet Number: 47 CutSet Frequency = 8.99E-09 INIT-LOP 1.07E-01' INITIATING EVENT- LOSS OF OFFSITE POWER L-MP504--M 3.35E-03 MTR-DRIVEN PP 504 OOS FOR MAINTENANCE - TRAIN B L-TPl40-IDR- 1.03E-02 TURBINE DRIVEN PP 140 FT RUN FOR 24 H L-TPl40NRIDR 1.00E+00 TURBINE DRIVEN PP 140 FT TO RUN IS NOT RECOVERABLE U-C2A0418--P. 3.00E-03 BKR(CNTL)4160V A0418 FT OPEN ON DEMAND U-OOPCLOOP-Z 8.12E-01 LOOP EVENT PLANT- CENTERED LOOP - NO NON-1E PWR AVAILABLE '

Page 8 CutSet Number: 48 Cutset Frequency = 8.29E-09 INIT LOP 1.07E-01 INITIATING EVENT - LOSS OF OFFSITE POWER L-MPl41-lDR 6.72E-03 MTR-DRIVEN PP 141 FT RUN FOR 24 HR-TRAIN A i L-MP504--!DR 9.57E-03 MTR DRIVEN PP 504 FT RUN FOR 24 HR TRAIN B l

. L-TPl40-1DR 1.03E-02 TURBINE DRIVEN PP 140 FT RUN FOR 24 H j L-TPl40NRIDR 1.00E+00 TURBINE DRIVEN PP 140 FT TO RUN IS NOT RECOVERABLE '

U-HCOSPWR60U l.17E-Ol' NONRECOVERY OF OFFSITE POWER IN 60 MINUTES CutSet14 umber: 49 CutSet Frequency = 7.68E-09 I INIT-LOP 1.07E-Ol' INITIATING EVENT - LOSS OF OFFSITE POWER L-HCHV4716PU l.00E+00 OPERATOR FAILS TO MANUALLY OPEN HV4716 LOCALLY AFTER FAULT L-MPCC0001-S 1.07E-04 COMMON CAUSE FAILURE -MP-S - > 14i AND 504 (AFW MDPS) I L-MV4716--P 5.73E-03 MTR-OPERATED VLV 4716 FT OPEN ON DEMAND l U-HCOSPWR60U l.17E-01 NONRECOVERY OF OFFSITE POWER IN 60 MINU11?S CutSet Number: 50 CutSet Frequency = 7.21E-09 '

INIT LOP 1.07E-01 INITIATING EVENT - LOSS OF OFFSITE POWER L-HCOVRSP-U 9.00E-02 OP FAILS TO RESET OVRSPD TRIP L-TP140RR-S 8.53E-03 TD AFW PUMP TRIPS ON OVRSPD AND IS RECOVERABLE i TEM U E+ N OP R RESPONSE O C S1 ITEMP A A GIVEN P V RESP FAIL M-HCNOSIAS-U l.50E-03 , NO OPRT RESPONSE TO HIGH TEMPERATURE ALARM-ESF SWGR/DIST U-HCOSPWR60U l.17E-01 NONRECOV'iRY OF OFFSITE POWER IN 60 MINUTES RESULTS: i L

Unit 2 core damage risk = 2.98E-06/yr Unit 2 otTsit e large, early release risk = 5.24E-08/yr b

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l ATTACHMENT #5 SAN ONOFRE UNITS 2/3 i SAN ONOFRE PRA MODEL UPDATES l

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Attachment #5 San Onofre Units 2/3 q i

UPDATES TO THE SAN ONOFRE IPE PRA MODEL The San Onofre Individual Plant Examination (IPE) was submitted to the NRC in May 1993. Since that time, San Onofre has maintained a living PRA model current v,ith plant design and operation.

Also, use of the living PRA model in the Safety Monitor has resulted in a hundreds of modification to improve the capability of the model to assess the risk of specific plant configurations. Most of the changes indicated below were included iri the model used for the San Onofre Individual Plant Examination for External Events submitted to the NRC in December 1995. Listed below are the major changes made to the living PRA mode since the IPE submittoi to the NRC.

1. . Updated plant specific failure data. .

2. Updated plant specific test and maintenance data.

]

3.

Revised the reactor coolant pump (RCP) seal LOCA model to incorporate more conservative RCP seal failure assumptions based on updated CE guidance on the potential for seal failure l given a loss of component cooling water. The revisions considered the potential for seal failure to be approximately 1.5E-3 in the event ofloss of component cooling water and 1.0 in the event of a loss of component cooling water coupled with a failure to trip the RCPs within 30 minutes.

4. Added more detail to loss ofnormal ESF switchgearand distribution room HVAC scenarios, l

5. Added a model for control room HVAC.  !

6. . Added a more detailed model of plant protection systems.

7. Improved modeling ofintake and heat exchanger heat treatment configurations.  !

8.' Revised LOCA initiating event frequencies to be consistent with CE Owners Group standards.

9. Incorporated common cause failure events directly into the fault trees to ensure truncation 'l ofindependent failure does not occur.  !

10. . Added detailed fault tree modeling for support system initiators.

I1. l Added dominant sequences from seismic, fire, and flooding PRAs. _  ;

12. . Revised interfacing system LOCA modeling to account for plant specific features (e.g.,  !

pressure indicators and alarm on piping between RCS boundary isolation valves).

13. Added Level 2 PRA sequences having contributions above 0.5% to the Safeiy Monitor.

14. Revised post-initiator operator recovery actions based on more detailed analyses.

15. 'Added all high safety significant SSCs to the Safety Monitor. j

16. ~ Incorporated switchyard maintenance impact in model.

' 17. Revised secondary heat removal model to credit feedwater makeup and steaming of a faulted and well as an unfaulted steam generatorconsistent with design features and the emergency ,

operator instructions.

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ATTACHMENT #6 SAN ONOFRE UNITS 2/3 INTERNAL EVENTS QUANTITATIVE LEVEL 2 RISK RESULTS l

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Attachment #6 Internal Events Level 2 Risk Results Table 1-1 CORRECTIVE MAINTENANCE DIESEL GENERATOR.AOT CONDITIONAL IN~ERNAL EVENTS RISK CONTRIBUTION PARAMETER Core Damage Large Early Frequency Release Frequency EDG Success Criteria 1 of 41 Present AOT, days 3 Proposed AOT,. days. 14 A1. Conditional Risk, per year 5.85E-05 7.13E-07 (1 EDG-unavailable).

A2. Conditional Risk, per year 2.67E-05 4.73E-07 (1 EDG never out for T/M)

A3. Increase in Risk, per year 3.18E-05 2.40E-07 A4. Single AOT Risk, Current 2.61E-07 1.97E-09 AS. Single AOT Risk, Propored 1.22E-06 f/ . 21E- 0 9 14 day A6. Downtime Frequency, per year 0.63 per diesel. j

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A7. Yearly.AOT Risk, Curront, 1.65E-07 1.24C-09 1

.per yr/ diesel I A8. Yearly AOT' Risk, Proposed, 7.68E-0~ 5.80E-09

. per yr/ diesel A9. Actual-Duration, hrs / event 23.8 A10. Single AOT Risk (based on 8.64E-08 6.52E-10 I actual data)

All. Yearly AOT Risk /yr/ diesel 5.44E-08 4.11E-10 (based on actual data)  ;

i 1-Utilizing EDG cross-tie under 10CFR50.54X for non-LOCA events.

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l Attachment #6 Internal Events Level 2 Risk Results Table 1-2 PREVENTIVE MAINTENANCE DIESEL GENERATOR AOT CONDITIONAL INTERNAL EVENTS RISK CONTRIBUTION PARAMETER Core Damage Large Early Frequency Release Frequency EDG Success Criteria 1 of 41 Present AOT, days 3 Proposed AOT, days 14 Bl. Conditional Risk, per year 2.83E-05 4.95E-07 (1 EDG unavailable)

B2. Conditional Risk, per year 2.67E-05 4.73E-07 (1 EDG Never out for T/M)

B3. Increase in Risk, per year 1.60E-06 2.20E-08 B4, Single AOT Risk, Current 1.32E-08 1.81E-1; B5. Single AOT Risk, Proposed 6.14E-08 8.443-10 14 day l l

B6. Downtime Frequency, per year 1.0  !

per diesel B7. Yearly AOT Risk, Current, per 1.32E-08 1.81E-10 yr/ diesel B8. Yearly AOT Risk, Proposed, 6.14E-08 8.44E-10 per yr/ diesel 1 B9. Proposed Downtime 224 1 (avg hrs / train /yr)

B10. Proposed Duration, hrs / event 224 B11. Single AOT Risk (based on 6.14E-08 8.44E-10 )

proposed 14 day AOT) i i

B12. Yearly AOT Risk /yr/ diesel l

(based on average annual 4.09E-08 5.63E-10 scoposed downtime) i 1 Utilizing EDG cross-tie under 10CFR50.54X for non-LOCA events.

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Attachment #6  !

Internal Events Level 2 Risk Results !

. Table 1-3 AVERAGE RISK RESULTS DIESEL GENERATOR AOT CONDITIONAL INTERNAL EVENTS RISK CONTRIBUTION PARAMETER Core Damage Large Early Frequency Release Frequency  ;

EDG Success Criteria' 1 of 41 j Present.AOT,. days. 3 Proposed AOT, days 14 I

C1. Proposed Downtime,~ hrs /yr/ diesel 239 l I

C2. Average Risk (current), 2.67E-05 4.73E-07 j per yr j l

C3. Average ' Risk (proposed) , 2.68E-05 4.73E-07 I per yr l

C4. Change factor from baseline < 1.004 < 1.001 Risk l

1 Utilizing EDG cross-tie under 10CFR50.54X for non-LOCA events.

2 CE NPSD-996 incorrectly indicated the proposed downtime as 220 hr/yr/ diesel. j I

l 1

i

. _.