October 9, 2014 A SCANA COMPANY RC-14-0158 · LERF 1.09E-07 3.19E-08 2.66E-07 PRA RAI 100 In letters dated October 10, 2012 (ADAMS Accession No. ML12297A218), April 1, 2013 (ADAMS

Thomas D. GatlinVice President, Nuclear Operations

803.345.4342

October 9, 2014A SCANA COMPANY RC-14-0158

U.S. Nuclear Regulatory CommissionDocument Control DeskWashington, D.C. 20555-0001Attn: S. A. Williams

Dear Sir or Madam:

Subject: VIRGIL C. SUMMER NUCLEAR STATION (VCSNS) UNIT 1.DOCKET NO. 50-395OPERATING LICENSE NO. NPF-12LICENSE AMENDMENT REQUEST-LAR-06-00055LICENSE AMENDMENT REQUEST TO ADOPT NFPA 805RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION

References:

1. Thomas D. Gatlin, SCE&G, Letter to NRC Document Control Desk, LicenseAmendment Request - LAR-06-00055, "License Amendment Request toAdopt NFPA 805 Response to Request for Additional Information". datedNovember 15, 2011 (RC-11-0149)

2. Shawn A. Williams, NRC, to Thomas D. Gatlin, SCE&G, "Virgil C. SummerNuclear Station Unit - 1 (VCSNS) - Request for Additional InformationRegarding License Amendment Request to Adopt National Fire ProtectionAssociation Standard 805 (TAC NO. ME7586)" dated July 11, 2014[ML14182A473]

South Carolina Electric & Gas Company (SCE&G), acting for itself and as agent for -South Carolina Public Service Authority pursuant to .10 CFR 50.90,. submitted LicenseAmendment Request (LAR) per Reference 1 and several responses to-Requests.for..Additional Information (RAI) to adopt NFPA 805. NRC review and audit of theserequests determined that additional information was required and a RAI was issuedper Reference 2. Attachment 1 contains SCE&G's response to these RAIs.

In addition to the response to the RAI, SCE&G was requested to revise Attachment Sand the license condition with the current implementation status and completion- time.Attachment 2 contains a revised Attachment S, Attachment 3 provides the revised list ofcommitments, Attachment 4 provides a marked up license condition and Attachment 5provides the typed revised License Condition. SCE&G is also providing a proposeddate change for completing the NFPA 805 Penetration Seal Documentation to the endof 2015.

Virgil C. Summer Station . Post Office Box 88. Jenkinsville, SC .29065. F (803) 941-9776 O e

Document Control DeskLAR-06-00055RC-14-0158Page 2 of 2

If you have any questions about this submittal, please contact Bruce L. Thompson at(803) 931-5042.

I certify under penalty of perjury that the foregoing is correct and true.

Executdd on ThorO D. Gatlin

RLP/TDG/wm

Attachment 1 - Probabilistic Risk Assessment (PRA) Request for AdditionalInformation (RAI) Response

Attachment 2 - Attachment SAttachment 3 - List of Regulatory CommitmentsAttachment 4 - Marked up License ConditionAttachment 5 - Retyped License ConditionAttachment 6 - Operating License & Technical Specification Changes

c: K. B. MarshS. A. ByrneJ. B. ArchieN. S. CamsJ. H. HamiltonJ. W. WilliamsW. M. CherryV. M. McCreeS. A. WilliamsNRC Resident InspectorS. E. JenkinsPaulette LedbetterK. M. SuttonNSRCRTS (CR-06-00055)File (813.20)PRSF (RC-14-0158)

Document Control DeskAttachment 1LAR-06-00055RC-14-0158Page 1 of 15

VIRGIL C. SUMMER NUCLEAR STATION UNIT IDOCKET NO. 50-395

OPERATING LICENSE NO. NPF-12ATTACHMENT I

Probabilistic Risk Assessment (PRA) Request for Additional Information (RAI)Response


Probabilistic Risk Assessment RAI 64.01

In a letter dated October 10, 2012 (Agency wide Documents Access and ManagementSystem (ADAMS) Accession Number ML12297A218), the licensee responded to PRARAI 64 and provided the results of a sensitivity analysis from applying a variancemethod that incorporates component failure rate error factors in developing the meaninterfacing system loss of coolant accident (ISLOCA) frequency. The results showed anincrease by a factor of 3.7 in internal events core damage frequency (CDF) to 5.OE-05per year and an increase by a factor of 145 in internal events large early releasefrequency (LERF) to 3.7E-05 per year. The licensee stated that these increases aredominated by the variance. in the rupture failure rate for motor-operated valves (MOVs).The NRC staff notes that while the ISLOCA frequency calculation method utilized in theinternal events probabilistic risk assessment (IEPRA) has no impact on the Fire PRA(FPRA), the large increase in CDF and LERF shown in this sensitivity analysis results inthe total CDF (including internal events, fire events, and seismic events) and LERFexceeding the Regulatory Guide (RG) 1.174, "An Approach for Using Probabilistic RiskAssessment-in Risk-Informed Decisions on Plant Specific Changes to the LicensingBasis," Revision 2, May 2011 (ADAMS Accession No. ML1 00910006) risk guidelines forRegion II so that only very small risk increases are allowed. Assess the relevance ofmore recent data on rupture failure rate of MOVs to the analysis, including updating thecontribution to total CDF and LERF from this issue.

Alternately, additional analysis or modifications may be necessary to meet RG 1.174guidelines. Provide a discussion of the potential changes to the overall analysis,including any plant changes needed to meet RG 1.174 guidelines as developed inresponse to PRA RAI 98, as well as the updated risk measures, CDF and LERF

Response

The question about variance in the ISLOCA frequency was originally raised in the peerreview of the SCE&G PRA in 2002 (F&O IE-06). The answer to this issue involved aspread sheet using a factored approach that gave a mean value for ISLOCA frequencywith a large variance as described in the question above.

The ISLOCA model for SCE&G underwent an update after F&O IE-06 was addressed.This update changed the ISLOCA model from a single number calculated outside of thefault tree (effectively a module) to explicit modeling of ISLOCA failure modes within thefault tree. These failure modes are represented by basic events with assigned errorfactors. The basic events are correlated by "type codes" for uncertainty studies.


The new ISLOCA model is based on WCAP-17154-P. Recovery for ISLOCA flow pathswas moved from the initiating event frequency calculation into the ISLOCA event treewhich resulted in higher initiating event frequencies. Also, the original ISLOCAmodeling requirement for Reactor Coolant System (RCS) depressurization and ReactorBuilding (RB) sump recirculation to reach a successful end state was revised to requirerefilling the Refueling Water Storage Tank (RWST). Moving the recoveries to the eventtree offset the increase in initiating event frequency. Refilling the RWST is more reliablethan RB sump recirculation. Therefore the net result was a reduction in ISLOCACDF/LERF contribution.

With the new model, it is possible run a monte-carlo simulation on the ISLOCA cutsetfile that produces uncertainty bounds and gives a new mean value for the ISLOCAinitiating event frequency based on the uncertainty bounds. The result of this calculationis that the monte-carlo ISLOCA initiating event mean (5.59E-05/yr) is very close to thevalue calculated by the fault tree (5.56E-05/yr).

50%-.O 5.3%-M

%%-I 72~E.M0 '1000

J

4.5E-5 5.E5 5.45 61-5 6.5E- 715 7.E~ eE-5 aS- 91-5 ai5E5 1.E-4 .-

fie,Aency / Fiobabfiy

I G,~.. I I ~ I II43wa


The individual ISLOCA component failure uncertainties are correlated in the overalluncertainty analysis for CDF and LERF for the new model. The CDF and LERF resultsfor a monte-carlo analysis using 10,000 simulations are shown in the table below anddemonstrate compliance with RG 1.174 risk guidelines.

Mean 5% 95%CDF 4.15E-06 1.73E-06 8.66E-06LERF 1.09E-07 3.19E-08 2.66E-07

PRA RAI 100

In letters dated October 10, 2012 (ADAMS Accession No. ML12297A218), April 1, 2013(ADAMS Accession No. ML13092A333), November 26, 2013 (ADAMS Accession No.ML13333A282), January 9, 2014 (ADAMS Accession No. ML14013A074), and May 2,2014 (ADAMS Accession No. ML14125A274), the licensee responded to PRA RAIs 08,66, 95, 10.02, and 10.03 and described the methodology to evaluate the risk of maincontrol room (MCR),abandonment scenarios, including the fault tree that is used tomodel accident scenarios with MCR evacuation followed by a failure to control thereactor remotely from the Control Room Evacuation Panel (CREP) and associatedrecovery actions. The NRC staff noted that the MCR evacuation logic includes basicevents for 1) evacuation due to the loss of habitability in the MCR, 2) evacuation due toloss of control (inability to successfully shutdown from the MCR), 3) pre-matureevacuation when control from the MCR is still available, 4) failure to evacuate whencontrol from the MCR is lost, and 5) failure to successfully shutdown followingabandonment (with and without station blackout [SBO]). In addition to abandonment onloss of habitability in the MCR, abandonment on loss of control scenarios are evaluatedin Fire Areas CB04, CB06, CB15, and CB17 (MCR).

Furthermore, regarding loss of MCR habitability scenarios, the licensee stated thatbecause the MCR abandonment logic acts by Boolean reduction in combination with therest of the model, it cannot be used to develop a CCDP or CLERP for a given scenario.The licensee further stated that it is not independent of the rest of the model, andcannot be quantified independently for the scenarios to yield an abandonment-onlyCCDP or CLERP at the scenario level. As a result, the licensee provided the entirerange of CCDPs for each of the four abandonment areas, which included the CCDPs forboth abandonment and non-abandonment scenarios. However, the updated LARAttachment G (submitted in a letter dated February 25, 2014, ADAMS Accession No.ML14063A455) describes two human error probabilities (HEPs) representing failure tosuccessfully implement the new MCR abandonment procedure: 1) 6.2E-02 for non-SBO scenarios and 2) 9.97E-02 for SBO scenarios. In addition, in a letter datedOctober 10, 2012 (ADAMS Accession No. ML12297A218), the licensee responded to


Fire Modeling RAI 01 .c and indicated that the probability of having to abandon the MCRdue to habitability is about 9E-02 which suggests that the CCDP for MCR abandonmentscenarios due to loss of habitability is less than 1 E-02. Provide the following:

a) The results (i.e., CDF, LERF, ACDF and ALERF) of a sensitivity studythat removes credit for Main Control Room (MCR) abandonment onloss of control (i.e., only credits remote shutdown due to loss ofhabitability in the MCR due to a fire). As a part of the response, youmay also present additional information that supports the analysis forloss of control abandonment. This additional information shouldinclude detailed discussion of the MCR evacuation fault tree logic,justification for each basic event probability, description andjustification for the criteria used to make the abandonment on.loss ofcontrol decision, and a characterization of the types of scenarios andrange of GCDPs for MCR abandonment scenarios due to loss ofcontrol only.

RESPONSE

SCE&G has previously provided detailed information on the completeness of theabandonment model and what it covers. We will expand that response by providing anactual image of the fault tree structure used and an explanation of the function of eachgate and how it operates. Some of the gates have extensive systems modeling underthe gate, so-these will not be provided in detail. Rather, a narrative discussion of whatis modeled under these gates will be provided.

Regarding justification for each basic event probability, we understand this to mean onlythose items that are specific to the abandonment model, which are the human errorprobabilities. The equipment random failure probabilities, the fire-induced directequipment failures, and the hot short failure probabilities that are used in theabandonment model are handled identically to all the other such failures in the overallplant model, and will not be discussed.

Regarding the justification of the criteria for making the abandonment on loss of controldecision, a detailed human reliability analysis (HRA) was conducted using NUREG-1921 accepted methods to evaluate MCR abandonment, including the decision toabandon, modeled by the human failure event (HFE) Main Control Room Evacuation(MCREVAC).

Since the issue is not whether the MCR is ever abandoned, but whether it is abandonedin time, a timeline that has been established (using the intended new procedures) toassess the following key timing elements that provide input to the calculation of thehuman error probability (HEP) for the cognitive HFE:


o Procedure validation and PRA evaluation were used to determine how long theoperators have to evacuate the control room.* The PRA has evaluated that since LOCA events are assumed not to be

occurring for sequences where control room evacuation is successful, thecrew has much longer to evacuate based on the PRA scenario as long asit is physically safe to do so. The available time for abandoning the MCRis based on the following:

• It is assumed that all actions required to evacuate the control roomand establish control of Emergency Feedwater at the Control RoomEvacuation Panel (CREP) must be completed within a system timewindow (T-sw) of 45 minutes, whether due to loss of control or loss ofhabitability. The basis for this timeframe is as follows: There is nocurrently available plant specific T/H calculation specifically fordetermining feedwater recovery time. However, there is a calculationfor establishing the T-sw for bleed-and-feed. This calculation usedthe time to steam generator dryout as the T-sw, which is 45 minutes.This is a conservative timeframe for recovering feedwater, since therewould still be quite a large amount of water above the fuel. Althoughthere is no plant-specific calculation for-SCE&G, WCAP-16141 hasgeneric calculations for Westinghouse plants. It shows that the timeto core uncovery is 2.3 hours for a generic 3-loop plant with nofeedwater at T=O and 21gpm/pump leakage (i.e., non-LOCAconditions, which apply in this case because Alternate Shutdown(ASD) credit is only applied for non-LOCA sequence). This supports aconclusion that the T-sw for SCE&G for recovering feedwater must besubstantially greater than 45 minutes. In this analysis, we used the 45minute steam generator dryout time from the plant specific calculationfor the T-sw for completing all actions required to achieve a safe-and-stable condition following abandonment. While this is clearlyconservative, the HRA for the cognitive decision to abandon the MCRon loss of control showed that using a longer T-sw would not affectthe HEP because the Cause Based Decision Tree Method (CBDTM)controls the HEP anyway. The analysis of the execution time forcompleting all necessary actions from the point at which the AbnormalOperating Procedure (AOP-900.2) attachments have been handed tothe operators, which was determined by detailed walkthrough of theprocedures, is on the order of (but less than) 30 minutes, 30 minuteswill be used to bound this action time.

Since 30 minutes is required for execution, and 45 minutes isassumed to be available, 15 minutes is judged to be a conservativetime window for exiting the control room and initiating theabandonment attachments.


o Operator interviews and walkthroughs were used to determine how long it takesfrom receipt of a fire alarm to perform the visual confirmation; this is taken as thedelay time.* The sequence starts at the point where fire damage causes equipment

failures and plant trip. It is assumed plant trip and loss of systems doesnot occur until the onset of fire damage, since the only scenarios that canlead to loss of control are fires that cause significant damage.

The delay time was estimated as the time required for the ReactorOperator (RO) to verify the fire location. Based on three separate timedwalkdowns performed by SCE&G Operations, the time assessed was 1.5minutes. By this time the Control Room Supervisor (CRS) will have all theinformation needed per the AOP to make the decision to abandon (that is,the operator will have a report on the fire location, its severity, and the fireinduced failures (e.g., spurious actuations) will have manifestedthemselves in the MCR.

o The AOP specifies the loss of control abandonment criteria as follows: "Inabilityto operate equipment from the MCB, visible MCB panel damage, or spuriousequipment operation is observed AND the fire is of such a nature that there isconcern about maintaining the ability to safely control the plant". Operatorinterviews to determine the time taken for the diagnosis once the delay time hasended and given these conditions resulted an estimate of 1.5 minutes was used.

o Additional time was included for the issuance by the CRS of procedureattachments and direction to obtain keys to the operators who would beimplementing actions at the Control Room Evacuation Panel (CREP) or locally inthe plant. This time data was obtained through a separate timed walkthrough ofCRS actions by SCE&G operations and included time to don Self-ContainedBreathing Apparatus (SCBA). The time data was included in the calculation ofthe cognitive HEP as what is usually termed the "manipulation time" and wasevaluated as requiring 8.5 minutes. This is the time between the decision toabandon and the actual start of the execution actions.

The timing will be verified during the implementation phase, prior to self-approval.Simulator exercises to observe the diagnosis process will be conducted after theoperators are trained on the new procedures.

The HRA Calculator uses this timing along with analyst evaluations based uponoperator interviews and observations during walkthroughs as input information tocalculate an HEP using several methods. The final HEP is generally based upon thehighest value (unless the analyst had a compelling reason to specify a particular


method). For the cognitive HFE MCREVAC, the CBDTM yielded a slightly higher valuethan the Human Cognitive Reliability (HCR) method based on the assessment that theAOP-900.2 does not "present all information required to identify the actions directed,"meaning that the procedure is not completely prescriptive and the CRS must still makea decision based on conditions and experience that abandonment is required.

Based on the decision to abandon modeled by the cognitive HFE, two cases forprocedure-directed execution actions are then implemented in the VCSNS MCRabandonment model depending upon whether AC power recovery is required or not.Further binning of execution actions based on fire location and equipment impacts wasnot considered necessary.

Finally, it is noted that there is no specific guidance on performing the HRA andmodeling for abandonment on loss of control. However, SCE&G is confident thatimplementing the approach discussed above is within the current capabilities of bothPRA modeling and HRA quantification techniques, and-that this will constitute anacceptable analysis for these scenarios. Further, it should be noted that the strategyimplemented by SCE&G in the new procedures is designed to encourage abandonmentin those cases where it affords the best opportunity to reach a safe, stable state.SCE&G has committed to a number of physical plant modifications and has putsubstantial effort into improving and validating its main control room abandonmentprocedures and training with the specific intent of engendering confidence in theabandonment process. For these reasons, SCE&G believes that performing asensitivity study removing all credit for abandonment on loss of control is not productive.

With regard to providing ranges of Conditional Core Damage Probability (CCDPs) forloss of control abandonment scenarios, SCE&G has explained in previouscorrespondence that the way the model is built, with the abandonment fully integratedinto the fault trees, there are no designated abandonment scenarios. Rather, there areabandonment cutsets within scenarios, and for the control room these could includeboth loss of control and loss of habitability, along with non-abandonment cutsets. Thus,CCDPs that only include the abandonment part of the CCDP do not exist. SCE&Gproposes the following work-around using cutset manipulation.

In order to promote a complete understanding of the frequencies of core damageresulting from abandonment cases, there will be three frequencies and probabilitiesreported, based on the following construct.

CDF = Fs * CCDP

CCDP = CLOCP * CAFP

Where:

CDF = Core Damage Frequency


Fs = Frequency of the Fire Scenario (the frequency that the fire starts in a particularignition source and effects a particular set of targets)

CCDP = Conditional Core Damage Probability (probability of core damage given theoccurrence of the fire scenario)

CLOCP = Conditional Loss of Control Probability (probability that the scenario leads to aloss of control from the control room, requiring abandonment - applies to fires in thefour areas (the control complex). Note that some licensees have been combining thisvalue with the fire scenario frequency (Fs) and calling the resultant product the"abandonment scenario frequency."

CAFP = Conditional Abandonment Failure Probability (probability that achieving safeshutdown fails given that abandonment is requirpd). Note that some licensees (thosewho calculate the "abandonment scenario frequency" mentioned above) have beenreporting this value as the CCDP for abandonment scenarios.

SCE&G prepared a control room abandonment results table providing CCDP, CLOCP,and CAFP ranges so that the scenarios where loss of control abandonment could becalled for, using the following approach:

o For the control room scenarios, the probability of a loss of habitability was set tozero in the cutset editor. This eliminates the loss of habitability-contribution.

0 For all the scenarios in the control complex, the scenario frequency was set to1.0. This gives the scenario CCDP. Note that this can include both loss ofcontrol and non-loss of control cases.

Some additional manipulations of the cutsets were necessary because the cutsetsassociated with these CCDPs are a mix of non-abandonment, failure to abandon, failureto successfully execute after abandonment, and equipment impacts precluding successof abandonment. The manipulations are discussed below.

The cutsets for the scenarios where abandonment can be credited were compiled andevaluated. There are several unique identifiers (flags) used in the cutsets that allow theanalyst to have a better understanding of the possible outcomes when fires occur inareas where remote shutdown can be credited.

o FLAG-EQUIP-REMOTE-FAIL - This flag is used to identify scenarios whereequipment that is required to perform remote shutdown is damaged in the fire.

o REMOTE-FAIL-EVENT - This identifier is used for accident sequences whereremote shutdown cannot be credited (i.e., LOCA, secondary side break, etc.)

o FORCED-EVAC-CB1701 - This identifier is for habitability scenarios (CLOHP)and is discussed in part b) of this RAI.


These identifiers were used to determine the portion of the risk associated with thedifferent types of abandonment scenarios.

To determine the ranges for CLOCPs and CAFPs, the cutsets that do NOT include aCLOHP flag for habitability were analyzed for the following categories:

o Scenarios where remote shutdown is not possible due to the accident type(cutsets including the flag REMOTE-FAIL-EVENT). In this case, CCDP=CLOCPbecause CAFP=I. See Tables 1 and 2 for results.

o Scenarios where the MCR is abandoned due to a loss of control and the remoteshutdown still fails due to a loss of equipment that is required to be available forremote shutdown (cutsets including FLAG-EQUIP-REMOTE-FAIL). See Tables3 and 4 for results.

o Scenarios where the MCR should be abandoned on loss of control, but either theoperators fail to abandon or operators fail to shut down the plant successfully,despite the availability of required equipment at the remote shutdown panel. Inthis case CAFP=HEP and CLOCP=CCDP/CAFP. See Tables 5 and 6 forresults.

The three conditional probability value ranges are provided for each of the above cases.In addition, the total frequency of the scenarios in each case (that is, the sum of thescenario frequencies that can lead to each case) is reported. This value is not equal tothe sum of the values from the three loss of control abandonment cases mentionedabove because there is double counting involved. The reason for the double counting isthat there is not necessarily a one-to-one correspondence between each scenario andthose cases. That is, any given scenario could possibly have two or three outcomes ifthere are failure probabilities involved other than just direct fire failures (e.g., somecutsets could lead to failure due to human actions, some cutsets could lead to failuredue to equipment failures, and some cutsets could lead to failure due to conditionsbeyond the capability of alternate shutdown) and therefore that scenario frequencywould appear in all three lists.

As can be seen from the results, the maximum CAFP value in Tables 1 through 4 is 1.0,indicating that there are some scenarios in each area where abandonment is aguaranteed failure either due to fire-induced conditions that preclude successfulalternate shutdown or fire-induced equipment failures that preclude successful alternateshutdown. Tables 5 and 6 do not have CAFP values of 1.0, but that makes sensebecause these cases are, by definition, limited to those conditions where remoteshutdown is possible and feasible. The range of CAFPs results from variations in thecomplexity of the response for the scenario (i.e., the extent of the functional failures).


Table 1. CDF Results for MCR Abandonment Scenarios Where Remote Shutdownis not Possible Due to Accident Type

Max Min Max MinFire Area Frequency CCDP CCDP CLOCP CLOCP CAFP

CB04 1.1E-06 9.6E-02 3.5E-05 9.6E-02 3.5E-05 1.OE+00CB06 4.8E-06 1.2E-01 1.2E-06 1.2E-01 1.2E-06 1.OE+00CB15 1.2E-05 2.6E-01 1.4E-07 2.6E-01 1.4E-07 1.OE+00CB17.01 1.4E-05 1.7E-01 2.2E-06 1.7E-01 2.2E-06 1.OE+00CB17.02 1.6E-07 5.7E-03 1.2E-03 5.7E-03 1.2E-03 1.OE+00CB17.03 1.2E-07 5.7E-03 1.2E-03 5.7E-03 1.2E-03 1.OE+00

Table 2. LERF Results for MCR Abandonment Scenarios Where RemoteShutdown is not Possible Due"to Accident Type

Max Min Max MinFire Area Frequency CLERP CLERP CLOCP CLOCP CAFPCB04 2.3E-08 1.5E-03 8.8E-09 1.5E-03 8.8E-09 1.OE+00CB06 1.OE-08 1.5E-03 2.9E-10 1.5E-03 2.9E-10 1.OE+00CB15 4.OE-08 2.9E-03 2.8E-10 2.9E-03 2.8E-10 1.OE+00CB17.01 6.7E-08 2.9E-03 4.9E-07 2.9E-03 4.9E-07 1.OE+00C1317.02 1.5E-10 5.8E-06 1.1E-06 5.8E-06 1.1E-06 1.OE+00CB17.03 1.OE-10 5.8E-06 1.1E-06 5.8E-06 1.1E-06 1.OE+00

Table 3. CDF Results for MCR Abandonment Scenarios With a.Loss of ControlFire Max Min Max Min Max MinArea Frequency CCDP CCDP CLOCP CLOCP CAFP CAFP

CB04 9.3E-08 5.7E-03 5.7E-03 6.OE-02 6.OE-02 1.OE+00 2.OE-02CR06 1.6E-06 9.6E-02 2.7E-06 1.OE+00 2.7E-06 1.OE+00 2.0E-02CB15 1.7E-06 9.6E-02 7.7E-06 1.OE+00 7.7E-06 1.OE+00 2.OE-02CB17.01 4.6E-06 9.6E-02 7.5E-06 1.OE+00 7.5E-06 1.OE+00 2.OE-02CB17.02 0.OE+00 0.0E+00 0.OE+00 O.OE+00 0.OE+00 0.OE+00 0.OE+00CB17.03 0.OE+00 0.OE+00 0.OE+00 0.OE+00 0.OE+00 0.OE+00 0.OE+00

Table 4. LERF Results for MCR Abandonment Scenarios With a Loss of ControlFire Max Min Max Min Max MinArea Frequency CLERP CLERP, CLOCP CLOCP CAFP CAFP

CB04 2.1E-08 1.3E-03 1.3E-03 1.3E-02 1.3E-02 1.OE+00 2.OE-02CR06 3.4E-09 5.5E-05 3.8E-08 5.8E-04 1.9E-09 1.OE+00 2.OE-02CB15 3.4E-08 1.6E-03 3.7E-08 1.7E-02 1.9E-09 1.OE+00 2.OE-02C117.01 6.2E-08 1.6E-03 3.9E-08 1.7E-02 2.8E-09 1.OE+00 2.OE-02CB17.02 0.OE+00 0.OE+00 0.OE+00 0.0E+00 0.OE+00 0.OE+00 0.OE+00CR17.03 0.OE+00 0.OE+00 0.OE+00 0.OE+00 0.OE+00 0.OE+00 0.OE+00


Table 5. CDF Results for MCR Abandonment Scenarios With a Failure to ShutdownSuccessfully

Fire Max Min Max Min MaxArea Frequency CCDP CCDP CLOCP CLOCP CAFP Mi CAFP

CB04 9.3E-08 6.OE-03 6.OE-03 6.OE-02 6.OE-02 1.1E-01 1.1E-01CB06 1.6E-06 6.3E-02 2.9E-06 1.OE+00 4.7E-05 1.1E-01 6.3E-02CB15 1.7E-06 1.OE-01 8.3E-06 1.OE+00 1.3E-04 1.1E-01 6.3E-02CB17.01 4.6E-06 6.3E-02 8.1E-06 1.OE+00 1.3E-04 1.1E-01 6.3E-02CB17.02 O.OE+00 O.QE+00 O.OE+00 O.OE+00 O.OE+00 O.OE+00 O.OE+00CB17.03 O.OE+00 O.OE+00 O.OE+00 O.OE+00 0.0E+00 O.OE+00 O.OE+00

Table 6. LERF Results for MCR Abandonment Scenarios With a Failure to ShutdownSuccessfully

Max Min Max Min Max MinFire Area Frequency CLERP CLERP CLOCP CLOCP CAFP CAFPCB04 2.1E-08 1.3E-03 1.3E-03 1.3E-02 1.5E-09 1.1E-01 1.IE-01CB06 3.4E-09 3.6E-05 2.OE-09 1.2E-04 3.2E-10 1.1E-01 6.3E-02CB15 3.4E-08 1.7E-03 2.OE-09 1.3E-02 9.1E-10 1.1E-01 6.3E-02CB17.01 6.2E-08 1t0E-03 3.3E-09 1.3Eý-02 3.OE-10 1.1E&01 6.3E-02CB17.02 O.OE+00 O.OE+00 O.OE+00 O.OE+00 O.OE+00 O.OE+00 O.OE+00CB17.03 O.OE+00 O.OE+00 O.OE+00 O.OE+00 O.OE+00 O.OE+00 O.OE+00


b) Characterize the types of scenarios and range of CCDPs for MCRabandonment scenarios due to loss of habitability only. If fire-inducedfailures of MCR functions are not considered in abandonmentscenarios, provide justification for their exclusion. Describe howcredited abandonment actions from the abandonment procedureaddress loss of function or spurious actions that may occur as a resultof a fire leading to abandonment. If abandonment actions do notaccount for these effects then provide an updated response to PRARAI 98 that incorporates fire-induced failures in the modeling ofabandonment scenarios. In the response, address MCR abandonmentscenarios due to loss of habitability and loss of control separately.

RESPONSE

Similar to the discussion above with regard to CCDPs for loss'of habitabilityscenarios, SCE&G has explained in previous correspondence that the way themodel is built, with the abandonment fully integrated into the fault trees, there areno designated abandonment scenarios. Rather, there are abandonment cutsetswithin scenarios, and for the control room these could include both loss of controland loss of habitability, along with non-abandonment cutsets. Thus, CCDPs thatonly include the abandonment part of the CCDP do not exist: SCE&G proposesthe following work-around using cutset manipulation.

The equations shown are essentially the same as in part (a):

CDF = Fs * CCDP

CCDP = CLOHP * CAFP

Where:

CLOCP is replaced with CLOHP

CLOHP = Conditional Loss of Habitability Probability (probability that thescenario leads to a loss of habitability in the control room, requiringabandonment, applies to fires in the main control room only). Note that somelicensees have been combining this value with the fire scenario frequency (Fs)and calling the resultant product the "abandonment scenario frequency."

SCE&G prepared a control room abandonment results table providing CCDP,CLOHP, and CAFP ranges for the scenarios where loss of habitabilityabandonment could be called for, using the following approach:


o CLOHP is a basic event in the model. Its value can be extracted directlyfrom the CAFTA database.

o The cutsets were trimmed to include only the loss of habitability (LOH)cutsets by identifying cutsets with the LOH flag and deleting all of theothers.

o The CLOHP basic event was returned to its original value in the model.This generated CCDPs for each scenario for the loss of habitabilitycutsets.

Regarding the consideration of fire-induced failure, spurious actuation, andrandom failure and how their impacts are addressed in the model, all of these areincluded and it has been explained in prior submittals. The additionalexplanation in the response to item (a) above will provide any additionalclarification needed, since these are treated the same way for both loss of controland loss of habitability abandonment.

As noted in the discussion in (a), there are several unique identifiers (flags) usedin the cutsets that allow the analyst to have a better understanding of thepossible outcomes when fires occur in areas where remote shutdown can becredited. In this case, the only flag of interest is:

o FORCED-EVAC-CB1701 - This identifier (and similar identifiers forCB1702 and CB1703) designates a loss of habitability in the MCR. Theseevents are also defined as CLOHP. See Tables 7 and 8 for results.

To determine the portion of the risk associated with habitability (CLOHP), thecutsets that include the identifiers used to flag forced evacuation (like FORCED-EVAC-CB1701) were filtered and the CCDPs for those cutsets summed for eachscenario. This can then be compared to the overall CCDP of each scenario todetermine the portion of the risk associated with forced evacuation due tohabitability issues. SCE&G has provided a range of CLOHP, CAFP, and CCDPvalues for these scenarios, as well as a total frequency of LOH. In this case,SCE&G did not break down the values by sub-type as for the Loss of Control(LOC) case discussed in (a), but is reported as single ranges covering all LOHscenarios.


Table 7. CDF Results for MCR Abandonment Scenarios With a Loss ofHabitabilit

Max Min Max MinFire Area Frequency CCDP CCDP CAFP CAFP CLOHPCB17.01 1.8E-05 1.8E-02 1.4E-02 2.OE-01 1.6E-01 9.1E-02CB17.02 1.6E-07 1.6E-02 8.5E-04 2.OE-01 1.OE-02 8.2E-02CB17.03 1.2E-07 1.6E-02 8.5E-04 2.0E-01 1.OE-02 8.2E-02

Table 8. LERF Results for MCR Abandonment Scenarios With a Loss ofHabitability

Max Min Max MinFire Area Frequency CLERP CLERP CAFP CAFP CLOHP

CB17.01 7.OE-08 2.4E-04 6.OE-06 2.6E-03 6.6E-05 9.1E-02CB17.02 1.5E-10 1.4E-05 I 8.4E-07 1.7E-04 1.0E-05 8.2E-02CB17.03 1.OE-10 1.4E-05 8.4E-07 1.7E-04 1.OE-05 8.2E-02


VIRGIL C. SUMMER NUCLEAR STATION UNIT 1DOCKET NO. 50-395

OPERATING LICENSE NO. NPF-12ATTACHMENT 2

Attachment S(PRA RAI 101)

PO-O" =-- -St:6-Ma" RC-14-0158 Attachment S

RC-14-0158 Attachment S

S. Plant Modifications and Items to be Completed During Implementation

Plant Modifications, Items to be Completed During Implementation and Items Completed Page 1Plant Modifications, Items to be Completed During Implementation and Items Completed Page 1

-- a RC-14-0158 Attachment S

Table S-1, Plant Modifications Committed, provided below includes a description of the modifications, along with the followinginformation:

* Item ECR number,

* Risk ranking of the modification,

* Location of the modification,* Problem statement,

* Proposed change,

* An indication if the modification is currently included in the Fire PRA,

* Compensatory Measure in place,* A risk-informed characterization of the modification and compensatory measure, and

* The modification completion date.

Table S-1 Plant Modifications Committed

InItem Rank Location Problem Statement Proposed Change Fire omp Risk InformedPA Measure Characterization Completion

PRA

ECR50784: Low As Defined Additional insights gained Provide protection of Yes Yes Protection in the form of 3 2015NFPA 805 during performance of tubing/ circuits from the hour to 1 hour fire rating is

Circuit/ Tubing NFPA 805 analysis effects of fire. being provided for selectProtection defining circuit and circuits identified through

equipment interactions, the NSCA.

ECR50799: Medium RB412 Improvement in station Provide lower leakage Yes None Alternate Seal Injection 2015NFPA 805 RCP equipment to address Loss RCP Seals [Outage]. obviates much of the

Seal of Seal Cooling/ LOCA benefit of this modification.Replacement scenarios for RCP Seals. This would be ranked

"High" if not for AlternateSeal Injection.

ECR50810: High As Defined Fire protection feature Provide mitigation Yes Yes A sensitivity study for the 2015NFPA 805 enhancements. strategies to address fire PRA showed that this

Hazard fire initiators or limit fire modification was highlyProtection propagation. important.

Plant Modifications, Items to be Completed During Implementation and Items Completed Page 2Plant Modifications, Items to be Completed During implementation and Items Completed Page 2


Table S-1 Plant Modifications CommittedInItem Rank Location Problem Statement Proposed Change Fire Comp Risk Informed

Measure Characterization CompletionPRA

ECR71588: Low Various Improve documentation of Document updates to Yes None Integrity of fire barriers is 2015NFPA 805 penetration seal designs to include improved maintained by the quality of

Penetration Seal penetration tests. penetration details and penetration sealDocumentation alignment with vendor installations vs. fire test

tests. configurations (important tofire scenario development).

ECR50856: Medium As Defined Improve availability and Provide alternate No- None Communication is implicitly 2015NFPA 805 reliability of station backup, protected considered in credit for Fire

Communication communication system(s) communication system PRA operator actions.during fire scenarios, to support fire event. However, many are

performed in the controlroom wherecommunication is notthreatened by fire.

Note: ECR71588 is not a plant modification. This ECR was added to Table S-1 to emphasize the importance and size of the scope.

Plant Modifications, Items to be Completed During Implementation and Items Completed Page 3

4rýscz&e. RC-14-0158 Attachment S

Table S-2, Implementation Items, provided below includes:those items (procedure changes, process updates, and training toaffected plant personnel) that will be completed prior to the full implementation of new NFPA 805 fire protection program.Completion dates are as follows: items 3, 6, 7, 8, 10, 12, 13, 14, 17, 19 and 21 will be completed within 180 days of NRCapproval; items 1, 2, 4 and 11 will be completed by 12/31/2015; and items 5, 15, 16, 18, 20, 22 and 23 will be completed by3/31/2016.

Table S-2 Implementation Items

Item Primary NFPA 805 Description LAR Section I Source Corrective Action

No. Code Section1 3.2 FP Plan Table B-1 Open Items - Revise Fire Protection Program 4.1.2 and Attachment A CR1 1-03925/01

Administrative procedures (e.g. FP Program Plan, Transient MaterialControl, Compensatory Measures) as needed for implementation ofNFPA 805 Program as defined in Attachment A.

2 3.2.3 Procedures Table B-1 Open Items - Revise Fire Protection Preventive 4.1.2 and Attachment A CR1 1-03925/02Maintenance and Surveillance procedures to improve alignment toscope and frequencies associated with NFPA Code requirements asdefined in Attachment A and NFPA Code of Record Document.

3 3.3 Prevention Table B-1 Open Items - Revise Fire Protection Program Technical 4.1.2 and Attachment A CR1 1-03925/03procedures (e.g. Electrical Cable, Insulation Materials, InteriorFinishes) as needed for implementation of NFPA 805 Program asdefined in Attachment A.

4 2.6 Monitoring Table B-1 - Enhance VCSNS Condition Monitoring Program to 4.1.2 and Attachment A CR1 1-03925/04include NFPA 805 elements. (NFPA 805 Sections 3.2.3(3), 2.6)

5 3.4.2 Pre-Fire Plans & 4.3 Table B-1 - Update Fire Pre Plans to include NFPA 805 elements, 4.1.2 / 4.4.2 and Attachment A/E CR1 1-03925/05Radiation Release Fire PRA and Radiological Release elements. (NFPA 805 Section

3.4.2)6 3.4 Industrial Fire Brigade Table B-1 - Enhance VCSNS Fire Brigade Member qualification to 4.1.2 and Attachment A CR1 1-03925/06

include NFPA 805 elements. (NFPA 805 Section 3.4.1)

7 3.4.3 Training and Drills Table B-1 - Enhance VCSNS Emergency Response training 4.1.2 / 4.4.2 and Attachment A/E CR1 1-03925/07program to include NFPA 805 elements. (NFPA 805 Sections 3.4.3,3.4.4 and 3.4.5)

8 3.4.6 Communications Table B-1 - Complete communications study and define strategies to 4.1.2 and Attachment NG CR1 1-03925/08ensure viable communications exists to support the fire brigade andother plant personnel during the course of a fire emergency. (NFPA805 Section 3.4.6)




Item Primary NFPA 805 Description LAR Section / Source Corrective Action

No. Code Section9 N/A - removed prior to LAR submittal.

10 3.8.2 Detection Table B-1 - Rework any smoke detectors found to not in compliance 4.1.2 and Attachment A CR11-03925/10with NFPA 72E. (NFPA 805 Section 3.8.2)

11 3.8.2 Detection & 3.11 Table B-l/ B-3 - Update Surveillance procedures for "Required" Fire 4.1.2 and Attachment A/C CR1 1-03925/11Passive FP Features Barriers and ERFBS defined in the NSCA and Fire PRA. (NFPA 805

Sections 3.8.2 and 3.11)

12 3.11 Passive FP Features Table B-1 - Update Station Fire Barrier Penetration sealing details to 4.1.2 and Attachment A CR1 1-03925/12improve alignment with test protocols acceptable to the AuthorityHaving Jurisdiction. (NFPA 805 Sections 3. 1)

13 2.7.2 Configuration Control NFPA 805 - Complete update to Fire PRA procedures to manage 4.7 and Attachment B CRI1-03925/13& 3.2.3 Procedures configuration control of NFPA 805 Analysis documents. (NFPA 805

Section 2.7.2)14 2.7.2 Configuration Control NFPA 805 - Complete update to Engineering procedures to manage 4.7 and Attachment B CR1 1-03925/14

& 3.2.3 Procedures configuration control of NFPA 805 Analysis documents. (NFPA 805Section 2.7.2)

15 1.4 Performance Objectives TR08620-312 - Update of station operating procedures, including the 4.2.4 and Attachment C CR1 1-03925/15& 3.4.2 Pre-Fire Plans conducting associated training (which are not modification related) to

incorporate insights and the change in operational shutdown strategyin response to a fire at the station.

16 1.4 Performance Objectives TR07800-008 - Completion of Administrative procedures and 4.3.2 and Attachment D CR11-03925/16& 3.3.1 FP Operational documents to support the implementation of:the non-power modes ofActivities plant operating states for implementation of NFPA 805.

17 2.7.3.4 Qualification of NFPA 805 - Complete the identification of Training qualifications 2.7.3.4 and 4.7.2 CR1 1-03925/17Users & 3.2.1 Intent including the training of technical personnel responsible for update

and maintenance of the NFPA 805 Analysis. (NFPA 805 Section2.7.3.4)

18 2.7.1.2 FPP Design Basis NFPA 805 - Complete the development and issuance of the Fire 4.7.1 CR11-03925/18Documents & 3.2.3 Safety Analysis (FSA) to summarize area results and insights fromProcedures the NFPA 805 Analysis. (NFPA 805 Section 2.7.1.2)

19 3.4.4 Fire Brigade Table B-1 - Improve controls on procurement of FP Equipment to 4.1.2 and Attachment A CR1 1-03925/22Equipment ensure consistency with NFPA Standards




Item Primary NFPA 805 Description LAR Section / Source Corrective Action

No. Code Section20 2.7.2 Configuration Control Resolve (including timing) for 8 hour Emergency Lighting with the CR1 1-03925/25

& 3.2.3 Procedures elimination of Operator Manual Actions [except for Control RoomEvacuation]

21 Incorporate the recommendations of TR07800-01 1, "Seismic-Fire PRA RAI 35 CR 11-03925/26Interactions" into plant procedures and training.

22 Validate/update the Fire PRA model to reflect the as built PRA RAI 13 CR1 1-03925/27modifications and completed implementation items. This will alsoinclude a verification that the resulting change in risk is either lessthan that currently estimated in Attachment W or is within theguidance of RG 1.174. Should the validation/update of the modelreflecting as-built modifications and completed implementation itemsprovide results that do not meet RG 1.174 risk guidelines, actions willbe taken to restore compliance with the guidelines. These actionsmay include re-analysis, additional modeling, procedure changes oreven hardware changes to the plant. The course of action taken willbe specific to the issue that is causing RG 1.174 not to be met.

23 Upon completion of the project upgrading the PRA model, VCSNS PRA RAI 54 CR11-03925/31will work with the PWROG and schedule a full scope peer review.

Note: Changes to station procedures and Training associated with station hardware modifications (Table S-1 -and Table S-3) normally coincide with scheduled turnover ofthe equipment to the VCSNS Operation's organization, and are not included in the above table.


-- a RC-14-0158 Attachment SRC-14-0158 Attachment S

Table S-3, Plant Modifications Completed, provided below includes a description of the modifications, along with the followinginformation:

* Item ECR number,* Risk ranking of the modification,

* Location of the modification,* Problem statement,

* Plant change,* An indication if the modification is currently included in the Fire PRA,

* Compensatory Measure in place,* A risk-informed characterization of the modification and compensatory measure, and

* The modification completion year.

Table S-3 Plant Modifications Completed.

InItem Rank Location Problem Statement Plant Change Fire CoMrnp Risk InformedPA Measure Characterization Completion

PRA

ECR50577: Low Yard Operator manual action Provide auto start Yes CLB/ FEP Instrument air importance 2012NFPA 805 required to start Diesel capability for the Diesel in the internal events

Instrument Air Driven Air Compressor Driven Air Compressor model is associated withRecovery (Eliminate OMA). (XAC0014). Steam Generator Tube

Rupture. It is not asimportant for fire scenarios.

ECR50780: High AB Improvement in station Provide addition high Yes None A sensitivity study for the 2013Alternate Seal equipment to address Loss pressure pump/ Diesel fire PRA showed that this

Injection (MSPI) of Seal Cooling/ LOCA Generator to mitigate modification was highlyscenarios for RCP Seals. loss of RCP seal important.

cooling (NFPA 805Credit).

Plant Modifications, Items to be Completed During Implementation and Items Completed Page 7Plant Modifications, Items to be Completed During Implementation and items Completed Page 7


Table S-3 Plant Modifications CompletedInItem Rank Location Problem Statement Plant Change Fire Comp Risk Informed

Measure Characterization CompletionPRA

ECR50800: High TB436 Address vulnerability of Reroute 115kV Feed to Yes CLB/ FEP A sensitivity study for the 2014NFPA 805 IDA loss of the 230kV and ESF bus IDA (Risk) fire PRA showed that this115kV Supply 115kV feed from 1DX to [Outage]. modification was highly

Reroute 1DA and 1DB (ESF important.Busses) due to a single TBfire.

ECR50811: High CB Improve early indicationsof Provide Incipient Yes None A sensitivity study for the 2013NFPA 805 fire precursors in key risk Detection System at the fire PRA showed that thisIncipient significant areas of the top of selected modification was highlyDetection plant. electrical panels in the important.

Relay and Upper CableSpreading Rooms.

ECR50812: High CB Disconnect switches could Protect or reroute the Yes Yes The PRA showed that 2014

NFPA 805 not mitigate spurious disconnect switch spurious operation of theseDisconnect operation for all potential cables. components was a

Switch Rework circuit failure conditions. significant risk contributor.

Plant Modifications, Items to be Completed During Implementation and Items Completed Page 8Plant Modifications, Items to be Completed During Implementation and Items Completed Page 8




List of Regulatory Commitments

The following table identifies those actions committed to by SCE&G, Virgil C. SummerNuclear Station in References 1 and 3 as modified by Attachment S, Tables S-1 andS-3 of this document. Any other statements in this submittal are provided forinformation purposes and are not considered to be commitments. Please directquestions regarding these commitments to Bruce L. Thompson, Manager, NuclearLicensing, (803) 931-5042.

- Commitment Due Date/Event

ECR50577: NFPA 805 Instrument Air CompletedRecovery (2012)Provide auto start capability for the DieselDriven Air Compressor (XAC0014).

ECR50780: Alternate Seal Injection (MSPI) Completed

Provide addition high pressure pump/Diesel (2013)Generator to mitigate loss of RCP seal cooling(NFPA 805 Credit).

ECR50784: NFPA 805 Circuit/Tubing 2015Protection

Provide protection of tubing/circuits from theeffects of fire.

ECR50799: NFPA 805 RCP Seal 2015ReplacementProvide lower leakage RCP Seals [Outage].

ECR50800: NFPA 805 1DA 115kV Supply CompletedReroute (2014)Reroute 115kV Feed to ESF bus 1 DA (Risk)[Outage].


Commitment Due Date/EventECR50810: NFPA 805 Hazard Protection 2015Provide mitigation strategies to address fireinitiators or limit fire propagation.ECR5081 1: NFPA 805 Incipient Detection CompletedProvide Incipient Detection System at the top (2013)of selected electrical panels in the Relay andUpper Cable Spreading Rooms.

ECR50812: NFPA 805 Disconnect Switch CompletedRework (2014)Protect or reroute the disconnect switchcables.

ECR71588: NFPA 805 Penetration Seal 2015Documentation

Document updates to include improvedpenetration details and alignment with vendortests.ECR50856: NFPA 805 Communication 2015

Provide alternate backup, protectedcommunication system to support fire event.

Implementation Items listed in Enclosure 2, Table S-2, Implementation Items,Attachment S, Table S-2. provided below includes those items

(procedure changes, processupdates, and training to affectedplant personnel) that will becompleted prior to the fullimplementation of new NFPA 805fire protection program. Completiondates are as follows: items 3, 6, 7, 8,10, 12, 13, 14, 17, 19, and 21 willimplemented within 180 day of NRCapproval; items 1, 2, 4, and 11 willbe complete by 12/31/2015; anditems 5, 15, 16, 18, 20, 22, and 23will be complete by 3/31/2016.


VIRGIL C. SUMMER NUCLEAR STATION UNIT IDOCKET NO. 50-395


Marked Up License Condition

RG 110149 RC-14-0158 Attachment M~-44-OMg RC-14-0158 Attachment M

M. License Condition Changes4 Pages Attached

1 1Page

a RG-•0•a49 RC-14-0158 Attachment M

Replace the current VCSNS fire protection license condition 2.c (18) with the standardlicense condition in Regulatory Position 3.1 of RG 1.205, Revision 1, modified as shownbelow. In support of this change, VCSNS has developed a Fire Probabilistic RiskAssessment (Fire PRA) during the course of its observation of VCSNS's transition to NFPA805. Outstanding high level findings from the Fire PRA Peer review are included inAttachment V.

South Carolina Electric & Gas Company shall implement and maintain in effect allprovisions of the approved fire protection program that comply with 10 CFR 50.48(a) and 10CFR 50.48(c), as specified in the licensee amendment request dated November 15, 2011and as approved in the safety evaluation report dated . Except where NRCapproval for changes or deviations is required by 10 CFR 50.48(c), and provided no otherregulation, technical specification, license condition or requirement would require prior NRCapproval, the licensee may make changes to the fire protection program without priorapproval of the Commission if those changes satisfy the provisions set forth in 10 CFR50.48(a) and 10 CFR 50.48(c), the change does not require a change to a technicalspecification or a license condition, and the criteria listed below are satisfied.

Risk-Informed Changes that May Be Made Without Prior NRC Approval

A risk assessment of the change must demonstrate that the acceptance criteria below aremet. The risk assessment approach, methods, and data shall be acceptable to the NRCand shall be appropriate for the nature and scope of the change being evaluated; bebased on the as-built, as-operated, and maintained plant; and reflect the operatingexperience at the plant. Acceptable methods to assess the risk of the change may includemethods that have been used in the peer-reviewed fire PRA model, methods that havebeen approved by NRC through a plant-specific license amendment or NRC approval ofgeneric methods specifically for use in NFPA 805 risk assessments, or methods that havebeen demonstrated to bound the risk impact.

a. Prior NRC review and approval is not required for changes that clearly result in adecrease in risk. The proposed change must also be consistent with the defense-in-depthphilosophy and must maintain sufficient safety margins. The change may be implementedfollowing completion of the plant change evaluation.b. Prior NRC review and approval is not required for individual changes that result in a risk

-7 -8

increase less than 1 x 10 /year (yr) for CDF and less than 1 x 10 /yr for LERF. The proposedchange must also be consistent with the defense-in-depth philosophy and must maintainsufficient safety margins. The change may be implemented following completion of theplant change evaluation.

21 Page

S RG-.10-49 RC-14-0158 Attachment M

Other Changes that May Be Made Without Prior NRC Approval(1) Changes to NFPA 805, Chapter 3, Fundamental Fire Protection Program

Prior NRC review and approval are not required for changes to the NFPA 805, Chapter3, fundamental fire protection program elements and design requirements for which anengineering evaluation demonstrates that the alternative to the Chapter 3 element isfunctionally equivalent or adequate for the hazard. The licensee may use anengineering evaluation to demonstrate that a change to NFPA 805, Chapter 3, elementis functionally equivalent to the corresponding technical requirement. A qualified fireprotection engineer shall approve the engineering evaluation and conclude that thechange has not affected the functionality of the component, system, procedure, orphysical arrangement, using a relevant technical requirement or standard.

The licensee may use an engineering evaluation to demonstrate that changes to certainNFPA 805, Chapter 3, elements are acceptable because the alternative is "adequate forthe hazard." Prior NRC review and approval would not be required for alternatives tofour specific sections of NFPA 805, Chapter 3, for which an engineering evaluationdemonstrates that the alternative to the Chapter 3 element is adequate for the hazard. Aqualified fire protection engineer shall approve the engineering evaluation and concludethat the change has not affected the functionality of the component, system, procedure,or physical arrangement, using a relevant technical requirement or standard. The fourspecific sections of NFPA 805, Chapter 3, are as follows:

* Fire Alarm and Detection Systeriis (Section 3.8);* Automatic and Manual Water-Based Fire Suppression Systems (Sect!on 3.9);* Gaseous Fire Suppression Systems (Section 3.10); and,* Passive Fire Protection Features (Section 3.11).

(2) Fire Protection Program Changes that Have No More than Minimal Risk Impact

Prior NRC review and approval are not required for changes to the licensee's fireprotection program that have been demonstrated to have no more than a minimal riskimpact. The licensee may use its screening process as approved in the NRC safetyevaluation dated . The licensee shall ensure that fire protectiondefense-in-depth and safety margins are maintained when changes are made to thefire protection program.

Transition License Conditions

(1) Before achieving full compliance with 10 CFR 50.48(c), as specified by (2) below, risk-informed changes to the licensee's fire protection program may not be made without priorNRC review and approval unless the change has been demonstrated to have no more thana minimal risk impact, as described in (2) above.(2) The licensee shall implement the following modifications to its facility to comnplete thetransition to full compl.Iiae with 10 CFR 50.48(-G) by December 31, 2015:

31 Page

R-G--1-O49 RC-14-0158 Attachment M

& ECR50577: NFPA 805 Instrument Air Recovery* ECR50780: Alternate Seal Injection (MSPI)0 ECR50784: NFPA 805 Circuit/Tubing Protection0 ECR50799: NFPA 805 RCP Seal Replacement* ECR50800: NFPA 805 1DA 115kV Supply Reroute* ECR50810: NFPA 805 Hazard Protection0 ECR50811: NFPA 805 Incipient Detection0 ECR50812: NFPA 805 Disconnect Switch Reworka ECR70588 ECR71588: NFPA 805 Penetration Seal Documentation0 ECR71553: NFPA 805 Communication

(3) The licensee shall complete implementation items in Table S-2 submitted in letterRC-14-0158 dated October 9, 2014, by the completion dates specified in Table S-2 to complete transition to full compliance with 10 CFR 50.48(c) by March 31,2016.

(4) The licensee shall maintain appropriate compensatory measures in place untilcompletion of the modifications delineated Transition License Conditions items 2 and 3above.

License condition 2.c (18) shall be superseded upon full implementation of the NFPA 805license condition:

Fire Protection System (Section 9.5.1, SSER 4)

Virgil C. Summer Nuclear Station shall implement and maintain in effect allprovisions of the approved fire protection program as described in the Final SafetyAnalysis Report for the facility, and as approved in the Safety Evaluation Report(SER) dated February 1981 (and Supplements dated January 1982 and August1982) and Safety Evaluations dated May 22, 1986, November 26, 1986, and July27, 1987 subject to the following provisions:

The licensee may make changes to the approved fire protection programwithout prior approval of the Commission only if those changes would notadversely affect the ability to achieve and maintain safe shutdown in theevent of fire.

No other license conditions need to be revised or superseded.

VCSNS implemented the following process for determining that these are the onlylicense conditions required to be either revised or superseded to implement the newFPP which meets the requirements in 10 CFR 50.48(a) and 50.48(c):

A review was conducted of the VCSNS Facility Operating License NPF-12, byVCSNS licensing staff and NFPA 805 Transition Team. The review was performedby reading the Operating License and performing electronic searches. In addition,outstanding LARs that have been submitted to the NRC were also reviewed forpotential impact on the license conditions.

Refer to Enclosure 3 for the proposed VCSNS Facility Operating License NPF-12markups and retyped pages.

41 Page




Operating License & Technical Specification Changes

Attachment to License Amendment No. LAR-06-00055To Facility Operating License No. NPF-12

Docket No. 50-395

Replace the following pages of the Operating License and Appendix A TechnicalSpecifications with the attached revised pages. The revised pages are identified byamendment number and contain marginal lines indicating the areas of change.

Remove Pages Insert PagesOL Page 7 OL Page 7._.

OL Page 7aOL Page 7b

OL Page 8 OL Page 8TS Page 6-11 TS Page 6-11

(Submitted with RC-1 1-0149)

-7-

b. In the event that one-third thickness semi-circular reference flawscannot be detected and discriminated from inherent anomalies,the entire volume of the weld shall be examined during theinservice inspection.

c. The reporting of the inservice inspection examination results shallbe documented in a manner to define qualitatively whether, theweldment and the heat affected zone and adjacent base metal onboth sides of the weld were examined by ultrasonic angle beamtechniques.

(9) Design Description - Control (Section 4.3.2. SER)

SCE&G is prohibited from using part-length rods during power operation.

(13) Deleted

(14) Deleted

(15) Deleted

(16) Cable Tray Separation (Section 8.3.3, SSER 4)

Prior to startup after the first refueling outage, SCE&G shall implementthe modifications to the cable trays discussed in Section 8.3.3 ofSupplement No. 4 to the Safety Evaluation Report or demonstrate to theNRC staff that faults induced in non-class 1 E cable trays will not result infailure of cable in the adjacent Class 1 E cable trays.

(17) Alternate Shutdown System Section 9.5.1, SSER 4)

Prior to startup after the first refueling outage, SCE&G shall install asource range neutron flux monitor independent of the control complex aspart of the alternate shutdown system.

(18) Fire Protection System

South Carolina Electric & Gas Company shall implement and maintain ineffect all provisions of the approved fire protection program that complywith 10 CFR 50.48(a) and 10 CFR 50.48(c), as specified in the licenseeamendment request dated November 15, 2011 and as approved in thesafety evaluation report dated . Except where NRCapproval for changes or deviations is required by 10 CFR 50.48(c), andprovided no other regulation, technical specification, license condition orrequirement would require prior NRC approval, the licensee may makechanges to the fire protection program without prior approval of theCommission if those changes satisfy the provisions set forth in 10 CFR50.48(a) and 10 CFR 50.48(c), the change does not require a change to atechnical specification or a license condition, and the criteria listed beloware satisfied.

Renewed Facility Operating License No. NPF-12

-7a-

Risk-Informed Changes that May Be Made Without Prior NRCApproval

A risk assessment of the change must demonstrate that the acceptancecriteria below are met. The risk assessment approach, methods, anddata shall be acceptable to the NRC and shall be appropriate for thenature and scope of the change being evaluated; be based on the as-built, as-operated, and maintained plant; and reflect the operatingexperience at the plant. Acceptable methods to assess the risk of thechange may include methods that have been used in the peer-reviewedfire PRA model, methods that have been approved by NRC through aplant-specific license amendment or NRC approval of generic methodsspecifically for use in NFPA 805 risk assessments, or methods that havebeen demonstrated to bound the risk impact.

a. Prior NRC review and approval is not required for changes thatclearly result in a decrease in risk. The proposed change mustalso be consistent with the defense-in-depth philosophy and mustmaintain sufficient safety margins. The change may beimplemented following completion of the plant change evaluation.

b. Prior NRC review and approval is not required for individualchanges that result in a risk increase less than 1 xl0-7/year (yr) forCDF and less than 1 x10-8/yr for LERF. The proposed changemust also be consistent with the defense-in-depth philosophy andmust maintain sufficient safety margins. The change may beimplemented following completion of the plant change evaluation.

Other Changes that May Be Made Without Prior NRC Approval

a. Changes to NFPA 805, Chapter 3, Fundamental Fire ProtectionProgram

Prior NRC review and approval are not required for changes to theNFPA 805, Chapter 3, fundamental fire protection programelements and design requirements for which an engineeringevaluation demonstrates that the alternative to the Chapter 3element is functionally equivalent or adequate for the hazard. Thelicensee may use an engineering evaluation to demonstrate that achange to NFPA 805, Chapter 3, element is functionallyequivalent to the corresponding technical requirement. A qualifiedfire protection engineer shall approve the engineering evaluationand conclude that the change has not affected the functionality ofthe component, system, procedure, or physical arrangement,using a relevant technical requirement or standard.

The licensee may use an engineering evaluation to demonstratethat changes to certain NFPA 805, Chapter 3, elements areacceptable because the alternative is "adequate for the hazard."Prior NRC review and approval would not be required foralternatives to four specific sections of NFPA 805, Chapter 3, forwhich an engineering evaluation demonstrates that the alternative


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to the Chapter 3 element is adequate for the hazard. A qualified fireprotection engineer shall approve the engineering evaluation andconclude that the change has not affected the functionality of thecomponent, system, procedure, or physical arrangement, using arelevant technical requirement or standard. The four specificsections of NFPA 805, Chapter 3, are as follows:

" Fire Alarm and Detection Systems (Section 3.8);" Automatic and Manual Water-Based Fire Suppression

Systems (Section 3.9);" Gaseous Fire Suppression Systems (Section 3.10); and," Passive Fire Protection Features (Section 3.11).

b. Fire Protection Program Changes that Have No More than MinimalRisk Impact

Prior NRC review and approval are not required for changes to thelicensee's fire protection program that have been demonstrated tohave no more than a minimal risk impact. The licensee may use itsscreening process as approved in the NRC safety evaluation dated

_ The licensee shall ensure that fireprotection defense-in-depth and safety margins are maintainedwhen changes are made to the fire protection program.

Transition License Conditions

a. Before achieving full compliance with 10 CFR 50.48(c), as specifiedby (2) below, risk-informed changes to the licensee's fire protectionprogram may not be made without prior NRC review and approvalunless the change has been demonstrated to have no more than aminimal risk impact, as described in (2) above.

b. The licensee shall implement the following modifications by

December 31, 2015:

" ECR50577: NFPA 805 Instrument Air Recovery" ECR50780: Alternate Seal Injection (MSPI)" ECR50784: NFPA 805 Circuit/Tubing Protection" ECR50799: NFPA 805 RCP Seal Replacement" ECR50800: NFPA 805 1DA 115kV Supply Reroute" ECR50810: NFPA 805 Hazard Protection" ECR5081 1: NFPA 805 Incipient Detection" ECR50812: NFPA 805 Disconnect Switch Rework" ECR71588: NFPA 805 Penetration Seal Documentation" ECR71553: NFPA 805 Communication

c. The licensee shall complete implementation items in Table S-2submitted in letter RC-14-0158 dated October 9, 2014, by the


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completion dates specified in Table S-2 to complete transition tofull compliance with 10 CFR 50.48(c) by March 31, 2016.

d. The licensee shall maintain appropriate compensatory measuresin place until completion of Transition License Condition items 2and 3 above.

(19) Instrument and Control Vibration Tests for Emergency Diesel EngineAuxiliary Support Systems (Section 9.5.4, SER)

Prior to startup after the first refueling outage, SCE&G shall either providetest results and results of analyses to the NRC staff for review andapproval which validate that the skid-mounted control panels andmounted equipment have been developed, tested, and qualified foroperation under severe vibrational stresses encountered during dieselengine operation, or SCE&G shall floor mount the control panelspresently furnished with the diesel generators separate from the skid on avibration-free floor area.

(20) Solid Radioactive Waste Treatment System (Section 11.2.3, SSER 4)

SCE&G shall not ship "wet" solid wastes from the facility until the NRCstaff has reviewed and approved the process control program for thecement solidification system.

(21) Process and Effluent Radiological Monitoring and Sampling Systems(Section 11.3, SSER 4)

Prior to startup after the first refueling outage, SCE&G shall install andcalibrate the condensate demineralizer backwash effluent monitorRM-L11.

(22) Core Reactivity Insertion Events (Section 15.2.4, SSER 4)

For operations above 90% of full power, SCE&G shall control the reactormanually or the rods shall be out greater than 215 steps until writtenapproval is received from the NRC staff authorizing removal of thisrestriction.

(23) NUREG-0737 Conditions (Section 22)

SCE&G shall complete the following conditions to the satisfaction of theNRC staff. Each item references the related subpart of Section 22 of theSER and/or its supplements.

a. Procedures for Transients and Accidents (1.C.1, SSER 4)

Prior to startup after the first refueling outage, SCE&G shallimplement emergency operating procedures based on guidelinesapproved by the NRC staff.
