Text

Response to Request for Additional Information Regarding Severe Accident Mitigation Alternatives
	ML092860615
Person / Time
Site:	Crystal River
Issue date:	10/09/2009
From:	Franke J; Progress Energy Florida
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	3F1009-03, TAC ME0278
	Download: ML092860615 (77)
	v • d • e

1 Progress Energy Crystal River Nuclear Plant Docket No. 50-302 Operating License No. DPR-72 Ref: 10 CFR 54 October 9, 2009 3F1009-03 U'-'.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001

Subject:

Crystal River Unit 3 - Response to Request for Additional Information Regarding Severe Accident Mitigation Alternatives for Crystal River Unit 3 Nuclear Generating Plant License Renewal Application (TAC NO. ME0278)

References:

(1) CR-3 to NRC letter, 3F1208-01, dated December 16, 2008, "Crystal River Unit 3 - Application for Renewal of Operating License" (2) NRC to CR-3 letter, dated August 10, 2009, "Request for Additional Information Regarding Severe Accident Mitigation Alternatives for Crystal River Unit 3 Nuclear Generating Plant License Renewal Application (TAC NO. ME0278)"

Dear Sir:

On December 16, 2008, Florida Power Corporation (FPC), doing business as Progress Energy Florida, Inc. (PEF), requested renewal of the operating license for Crystal River Unit 3 (CR-3) to extend the term of its operating license an additional 20 years beyond the current expiration date (Reference 1). Subsequently, the Nuclear Regulatory Commission (NRC), by letter dated August 10, 2009, provided a request for additional information (RAI) concerning the CR-3 License Renewal Application (Reference 2). The Enclosure to this letter provides the response to Reference 2.

No new regulatory commitments are contained in this submittal.

If you have any questions regarding this submittal, please contact Mr. Mike Heath, Supervisor, License Renewal, at (910) 457-3487, e-mail at mike.heath@pgnmail.com.

onA. Franke Vice President Crystal River Unit 3 JAF/dwh

Enclosure:

Response to Request for Additional Information xc: NRC CR-3 Project Manager NRC License Renewal Project Manager NRC Regional Administrator, Region II Senior Resident Inspector Progress Energy Florida, Inc.

Crystal River Nuclear Plant 15760 W. Power Line Street Crystal River, FL 34428

U. S. Nuclear Regulatory Commission Page 2 of 2 3F1009-03 STATE OF FLORIDA COUNTY OF CITRUS Jon A. Franke states that he is the Vice President, Crystal River Nuclear Plant for Florida Power Corporation, doing business as Progress Energy Florida, Inc.; that he is authorized on the part of said company to sign and file with the Nuclear Regulatory Commission the information attached hereto; and that all such statements made and matters set forth therein are true and correct to the best of his knowledge, information, ief.

JA. Franket kie President Crystal River Nuclear Plant iT*he foregoing document was acknowledged before me this day of 2009, by Jon A. Franke.

662Yý42L, 2 &A)AAkd/YL Signature of Notary Public State of Florida

y ..... CAROLYN I. OD $MA (Print PU -*m0_ 1Sbatift (Print, type, or sTamprC5m-nrss-r ,d Name of Notary Public)

Personally / Produced Known -OR- Identification

PROGRESS ENERGY FLORIDA, INC.

CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50 - 302 / LICENSE NUMBER DPR - 72 ENCLOSURE RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION

U. S. Nuclear Regulatory Commission Enclosure 3F1`009-03 Page 1 of 74 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION

1. Provide the following information regarding the Level 1 Probabilistic Safety Assessment (PSA) used for the Severe Accident Mitigation Alternatives (SAMA) analysis:

a. Section E.2.1 provides a detailed description of the PSA model changes made since the IPE Level 1 model.

i. Provide the core damage frequency (CDF) and large early release frequency (LERF) for each version of the PSA Model of Record (MOR) to demonstrate how changes in the PSA model impacted the calculated CDF and LERF.

ii. For each version of the PSA, identify the model changes listed in Section E.2.1 that most impacted the change in CDF and LERF.

b. Section E.2.2.1 states that the MOR 2006 PSA model used for the SAMA analysis reflects Crystal River Nuclear Power Plant (CR-3) as designed and operated up to April 2006. Identify any changes to the plant (physical and procedural modifications) since April 2006, that could have a significant impact on the results of the PSA and/or the SAMA analyses. Provide a qualitative assessment of their impact on the PSA and on the results of the SAMA evaluation.

c. Section E.2 states that an industry peer review was performed on the MOR 2000 PSA model and that all Level A, B, C, and D Facts and Observations (F&Os) have been addressed and closed. Section E.2.1.1.10 further states that the Level 2 PSA was not completed in time to support the industry Peer Review. In light of this, and the fact that the peer review of the PSA was performed several years prior to the development of the MOR 2006 PSA model used for the SAMA analysis, provide a description of the quality controls applied to the development of the MOR 2006, Level 1 and 2 PSA model.

Identify and discuss any additional internal and external reviews. Describe any significant review comments, their resolution, and the potential impact of any unresolved comments on the results of the SAMA analysis.

d. Figure E.2-1 provides the contribution to CDF by initiator as a percentage of the internal events CDF (4.99E-06/yr). Provide the actual numerical value for the CDF contribution for each initiator that sums to the total internal events CDF.

Response

1.a.i The changes to the CDF and LERF for the CR-3 PRA history are summarized in the table below:

U. S. Nuclear Regulatory Commission Enclosure 3F1009-03 Page 2 of 74 CDF LERF Ref IPE 1.4E-5 1.2E-6' I MOROO 3.4E-6 NC 2 MOROI 5.1E-6 NC 3 MOR02 6.83E-6 3.59E-7 4 MOR03 7.49E-6 3.42E-7 5 MORO3a 7.49E-6 3. 72E-7 6 MORO3b 5.40E-6 3.98E-7 7 MOR06 4.99E-6 3.69E-7 8 NC - Not Calculated IIPE LERF is a combination of Containment Fails early and Containment bypassed.

References:

1. Crystal River 3 Individual Plant Examination, March 1993

2. CR-3 PSA Model of Record, N01-0002, Rev. 0

3. CR-3 PSA Model of Record, N01-0002, Rev. 1

4. CR-3 PSA Model of Record, N01-0002, Rev. 2

5. PSA Model of Record, P02-0001, Rev. 0

6. PSA Model of Record, P02-0001, Rev. 1

7. PSA Model of Record, P02-0001, Rev. 2

8. PSA Model of Record, P02-0001, Rev. 3 1.a.ii The major items that impacted the PRA model for each model revision are listed below:

MORO0

" Added Backup ES Transformer(BEST) - ("A" and "B" safeguards trainspowered from separatetransformers)

FeedwaterPump 7 (FWP-7)powered with diesel generatorMTDG-1

Appendix R chiller installed

InstalledAlternate A C diesel, which could power an Essential Bus MOR01

Updates to the timing for post initiatorevents and dependency.

Revised the Steam Generator Tube Rupture (SGTR) binning per revised Event Tree.

MOR02

Added IE_F6A based upon internalflooding analysis revision.'

Updated the Post HRA values and the dependency analysis.

Revised the binning and updated the sequences for the level 2 analysis MOR03

New initiating event fault trees were added for Loss of Service Water (TIO) and Loss of Make-up (T16).

" Updated mutually exclusive events

U. S. Nuclear Regulatory Commission Enclosure 3F1009-03 Page 3 of 74 MORO3a

Updated table 8 bin definitions. Bin 12 is now late, medium (from early, low)

MOR03b

PORV block valve alignment

" Added HHUTHR1 Y, (Operatorfails to control HPI following spurious actuation)

Added HHUTHR2Y, (Operatorfails to control HPI before liquid relief)

MOR06

The major model changes were the installation of an alternate diesel generator EGDG-1C, the removal of MTDG-1, and the ability to align unit buses from the auxiliarytransformer.

1.b A working model was quantified July 2009 and has a CDF of 3. 63E-06 and a LERF of 1.82E-07. There have been no plant changes that have a significant impact to the PRA model since MOR06. The major changes to the model from the issue of MOR06 to July 2009 have been to comply with Regulatory Guide (RG) 1.200 and to include the addition of potential multiple spurious operation (MSO) events. None of these changes are expected to have a significant impact on the results of the SAMA evaluation.

1.c Quality control of the PRA model is prescribed in an administrative procedure for updates to the PSA Model. The administrative procedure outlines the methodology to ensure the PRA model is maintained current with the changes to the plant. The PSA model update procedure covers model update administration, implementation, and tracking of the errorand improvement opportunities.

A full scope PRA self assessment (except flooding) of MOR06 was performed in 2007 to the requirements of the RG 1.200 standard. The requiredmodel changes from the facts and observations were incorporatedin the working model in 2008. Also, a limited scope peer review was performed in 2009 covering a portion of the Technical Elements of the internal events, at-power PRA. The Technical Elements included in the focused review were Initiating Events Analysis (IE), Quantification (QU) (partial), and LERF Analysis (LE). The findings from these reviews should not have a significant impact on the results of the SAMA analysis.

1.d The following table gives the contribution of the IE to the CDF.

Initiating Event IE-12 %E-12 IE A LARGE BREAK LOCA 1.74E-07 3.5%

IEFIG RUPTURE OF CIRC WATER EXP JOINTS ON TURB BLD EL 95 (FIRE ZONE TB- 7.87E-09 0.2%

95-400A) 7.87-09 0.2 IEFIB FIRE SERVICE PIPING RUPTURE INFIRE ZONE AB-1 19-6E (SPRAY) 7.03E-10 0.0%

IE F3 PIPE RUPTURE INFIRE ZONE TB-1 19-400E 2.11E-10 0.0%

IE F4 RUPTURE OF BWSTPIPING INDECAYHEATPITB (FIRE ZONE AB-75-4) 1.60E-10 0.0%

IEF5 RUPTURE OF BWSTPIPING INDECAY HEATPITA (FIRE ZONE AB-75-5) 2.68E-11 0.0%

IEF6A PIPE RUPTURE ON ELEVATION 95 OF THE AUX BLDG (FIRE ZONE AB-95-X) 3.83E-07 7.7%

IEF9A PIPE RUPTURE ON ELEVATION 95 OF TURB BLDG (FIRE ZONE TB-95-400A) 1.34E-09 0.0%

IE_F9B PIPE RUPTURE ON ELEVATION 95 OF TURB BLDG (FIRE ZONE TB-95-400A) - 4.61E-09 0.1%

U. S. Nuclear Regulatory Commission Enclosure 3F1009-03 Page 4 of 74 Initiating Event 1E-12 %E-12 (SPRAY)

IEM MEDIUM BREAK LOCA 1.09E-07 2.2%

IER STEAM GENERATOR TUBE RUPTURE 3.53E-07 7.1%

IES SMALL BREAK LOCA 1.52E-06 30.5%

IET1 REACTOR TRIP 2.73E-07 5.5%

IETIO LOSS OF NSCCC 1.07E-07 2.1%

IET11 LOSS OF INTAKE 3.14E-07 6.3%

IET12 LOSS OF'A'DCPOWER BUS 1.62E-08 0.3%

IET13 LOSS OF'B'DCPOWER BUS 5.17E-08 1.0%

IET14 LOSS OF "C"BATTERY BACKED BUS 7.74E-08 1.6%

IET15 LOSS OF STARTUP/BACKUP ES TRANSFORMER 1.29E-09 0.0%

IET16 LOSS OF MAKEUP 1.53E-07 3.1%

IET2 LOSS OF MAIN FEEDWATER 1.22E-07 2.4%

IET3 LOSS OF OFFSITEPOWER 3.03E-07 6.1%

IET4 EXCESSIVE FEEDWATER 4.75E-08 1.0%

IET5 STEAM/FEEDLINE BREAK 2. 72E-08 0.5%

IET7 SPURIOUS ES ACTUATION 6.34E-08 1.3%

IET8 LOSS OF 4160V ES BUS 3A 2.63E-07 5.3%

IET9 LOSS OF 4160V ES BUS 3B 6.42E-08 1.3%

IEV ISLOCA - DHR DROP LINE AND INJECTION LINES 5.14E-08 1.0%

IEZ REACTOR VESSEL RUPTURE 5.OOE-07 10.0%

Total 4.99E-06

2. Provide the following information relative to the Level 2 analysis:

a. Describe how the Level 2 model used for the SAMA analysis differs from the Individual Plant Examination (IPE) backend analysis.

b. Table E.5-2, Level 2 Importance List Review for Risk Reduction Worth (RRW) Greater than 1.02, presents the basic events with an RRW greater than 1.02 for LERF sequences. Not counting flags, split fractions and initiating events only five basic events are presented. Explain why there are so few basic events with an RRW greater than 1.02 for LERF sequences. Specifically address why there are no loss of offsite power related events. Also, provide the Risk Reduction Worth values for each entry in Table E.5-2.

c. Section E.2.1.3.10, states that certain sequences (i.e. MK, SK, RK ATWS) were deleted from the Level II PSA model due to low frequency. Provide the cutoff frequency used to delete these sequences. Also, define MK, SK, and RK ATWS.

d. Provide a breakdown of the CDF and population dose by containment release mode (e.g., intact containment, containment isolation failure, containment bypass - Steam Generator Tube Rupture [SGTR], containment bypass - ISLOCA, IVR).

U. S. Nuclear Regulatory Commission Enclosure 3F1009-03 Page 5 of 74

e. In the discussion of the Level 2 analysis (Sections E.2.2.2, E.2.2.3 and E.3.4), the process used to map Level 1 results into the Level 2 model and group the containment event tree (CET) end states into release categories is not clear.

i. Provide a description of the process used to map the Level 1 results into the Level 2 analysis. Describe the plant damage states and how they were applied.

ii. Provide a description of the process used to group the containment event tree (CET) end states into release categories. Provide a typical CET showing release categories assigned to each end state.

iii. Identify and describe the number of Modular Accident Analysis Program (MAAP) calculations made to obtain the fission product release fractions for each release category. Provide an example of the weighting calculation for a representative CET sequence. Also, identify the version of MAAP used in the SAMA evaluation.

iv. Section E.5.1.2 explains that "...even though Release Categories 3B and 4C were not contributors to LERF, they were large contributors to Level 3 offsite consequence" and that a review was performed to determine if further risk dominant basic events could be identified. Section E.5.1.2 goes on to explain that no new dominant risk contributors not already identified from Level 1 results were found.

Whereas Table E.5-2 identifies the LERF sequences having RRW values greater than 1.02, no corresponding table or information is presented that shows how risk importance events contributing to Release Categories 3B or 4C were reviewed.

Identify the risk contributors from the Level 1 review that are also dominant contributors to Release Categories 3B and 4C and clarify why these are the dominant risk contributors.

Response

2.a The general concepts used to develop the Level 2 in MOR06 is similar to that used in the IP; however, the analysis has been completely revised using a different set of analysis tools and methods consistent with other Progress Energy plant PRAs. The process of binning and assigning plant damage states is similar although the current method uses a single top fault tree solution whereas the IPE solved the Level I and Level 2 sequences separately and combined them with post processing tools. The IPE also used STCP and CONTAIN for T-H and containment analysis while MOR06 is based upon MAAP.

2.b Table E.5-2 was intended to identify only those events unique to both LERF and those Release Categories (RCs) responsible for a large contribution to the total offsite dose (i.e., RCs 3B and 4C). However, since it appears that there were only a few basic events identified that were unique to both LERF and these RCs, the following table was generated to identify all events in any cutset that was associated with a Plant Damage State (PDS) that has a possible contribution to these release categories. For example, even though an event may not be a specific contributor to LERF, but it belonged to a cutset that was identified with a PDS that had some non-zero split fraction assigned to LERF, or RCs 3B or 4C, it was listed in the below table to identify whether it was previously assessed as part of the Level 1 importance list review. In all cases, the

U. S. Nuclear Regulatory Commission Enclosure 3F1009-03 Page 6 of 74 events listed below that could have a possible contribution to LERF or RCs 3B or 4C were all assessedas part of the Level 1 review in Table E.5-1.

Loss of offsite power-related events were addressed as part of the Level 1 importance list review, e.g., SAMA 18 dealt with the provision of an additional EDG to mitigate common cause failure events. For the Crystal River Nuclear Power Plant (CR-3), the loss of offsite power events were not significant contributors to LERF and the other release categories responsible for relatively high contributions to offsite dose.

For CR-3, LERF is mainly dominated by SGTR and ISLOCA events. The number of events unique to LERF with RRW values greater than 1.02 is small due to the fact that ISLOCA is modeled as a point estimate, while SGTR sequences are dominated by operatoraction failures to cool down and depressurize the reactor coolant system, e.g.,

RHUPORVY - OPERATORS FAIL TO OPEN PORV FOR PRESSURE RELIEF. Also, support systems, such as HPI and AFW, have sufficient redundancy and a good "defense-in-depth" design. such that these support systems are not significant contributorsto LERF.

U. S. Nuclear Regulatory Commission Enclosure 3F1 009-03 Page 7 of 74 SAMA RAI 2.b Response Event Name Probability 1 Red W Description Potential SAMAs FLGX 1.00E+00 1.75 TAG EVENT - LONG TERM COOLING (HPR/LPR/REFILL) Already accounted for in the Level 1 importance list IES 5.OOE-04 1.45 SMALL BREAK LOCA Already accounted for in the Level 1 importance list FLG HVAC 1.OOE+00 1.37 HVAC REQUIRED DUE TO AVAILABILITY OF AC POWER Already accounted for in the Level 1 importance list FLGQHUEFWMR 1.OOE+00 1.33 OPERATORS FAIL TO MANUALLY OPEN CONTROL VALVE Already accounted for in the Level 1 importance list QSPLHVAC 5.00E-01 1.33 SPLIT FRACTION - VALVES FAIL CLOSED ON LOSS OF HVAC Already accounted for in the Level 1 importance list FLG SW 1.OOE+00 1.24 TAG EVENT - LOSS OF NORMAL SW Already accounted for in the Level 1 importance list JHUCHP2R 1.00E+00 1.14 OPERATORS FAIL TO USE DEDICATED CHILLED WATER SYSTEM Already accounted for in the Level 1 importance list IEZ 5.00E-07 1.12 REACTOR VESSEL RUPTURE Already accounted for in the Level 1 importance list QPMFWP7M 2.03E-02 1.12 FWP-7 IN MAINTENANCE Already accounted for in the Level 1 importance list JHUCHP2Z 5.OOE-02 1.11 JHUCHP2R Already accounted for in the Level 1 importance list QHUFW7EY 1.OOE+00 1.10 OPERATORS FAIL TO START FWP-7 BEFORE PORV LIFTS Already accounted for in the Level 1 importance list RHUPORVY 5.O0E-01 1.09 OPERATORS FAILS TO OPEN PORV FOR PRESSURE RELIEF Already accounted for in the Level 1 importance list

F6A IE 2.63E-03 1.08 PIPE RUPTURE ON ELEVATION 95 OF THE AUX BLDG (FIRE ZONE AB- Already accounted for in the Level 1 importance list

_EF6A_2.63E-03_1.08 95-X) Alreayaccuntedfor___theLevel1_imprtanclis IER 3.OOE-03 1.07 STEAM GENERATOR TUBE RUPTURE Already accounted for in the Level 1 importance list QMMEFP3F 3.29E-02 1.07 EFP-3 PUMP TRAIN FAILS TO RUN Already accounted for in the Level 1 importance list lE T3 7.27E-03 1.07 LOSS OF OFFSITE POWER Already accounted for in the Level 1 importance list HHUHPRCY 4.40E-04 1.07 OPERATORS FAIL TO SWITCH FROM HIGH PRESSURE INJECTION TO Already accounted for in the Level 1 importance list RECIRCULATION Alradyaccuntdfr___thLeel__imortnce __s IET11 1.16E-04 1.07 LOSS OF INTAKE Already accounted for in the Level 1 importance list lE T8 3.21E-03 1.06 LOSS OF 4160V ES BUS 3A Already accounted for in the Level 1 importance list HHUMPSBY 1.OOE+00 1.06 OPERATOR FAILS TO START STANDBY MAKEUP PUMP Already accounted for in the Level 1 importance list RHUCOOLY 5.80E-04 1.06 OPERATORS FAIL COOLDOWN VIA OTSG Already accounted for in the Level 1 importance list QHUFWP7Y 5.60E-03 1.06 OPERATORS FAIL TO START FWP-7 Already accounted for in the Level 1 importance list RMMRCVSC 2.50E-02 1.06 SAFETY RELIEF VALVE FAILS TO CLOSE (STEAM RELIEF) Already accounted for in the Level 1 importance list SPMRW3BM 8.60E-03 1.06 RWP-3B IN MAINTENANCE Already accounted for in the Level 1 importance list ZHUCOM2Z 2.80E-01 1.05 COND PROB OF RHUPORVY GIVEN RHUCOOLY Already accounted for in the Level 1 importance list IE T1 1.1OE+00 1.05 REACTOR TRIP Already accounted for in the Level 1 importance list APWNR01R 6.40E-01 1.04 BOTH EDGS FTS, BOTH EFPS FTS Already accounted for in the Level 1 importance list FLGTBQR 1.OOE+00 1.04 TAG EVENT - STUCK OPEN RELIEF AFTER B Already accounted for in the Level 1 importance list LPM001BM 1.03E-02 1.04 DHP-1B TRAIN IN MAINTENANCE Already accounted for in the Level 1 importance list HHUINJAY 5.OOE-01 1.04 OPERATORS FAIL TO SWITCH MUV-23/24 TO BACKUP POWER Already accounted for in the Level 1 importance list

U. S. Nuclear Regulatory Commission Enclosure 3F1009-03 Page 8 of 74 SAMA RAI 2.b Response Event Name Probability [Red W Description Potential SAMAs QMMEFP2F 3.37E-02 1.04 EFP-2 FAILS TO CONTINUE TO RUN Already accounted for in the Level 1 importance list IEA 5.OOE-06 1.04 LARGE BREAK LOCA Already accounted for in the Level 1 importance list LPM001AM 1.03E-02 1.04 DHP-1A TRAIN IN MAINTENANCE Already accounted for in the Level 1 importance list SCCHDABF 2.39E-04 1.04 COMMON CAUSE FAILURE OF HXs DCHE-1A AND DCHE-1 B PLUGGED Already accounted for in the Level 1 importance list LMMDHPBF 4.90E-03 1.04 FAILURE OF DHP-1B AND ITS VALVES Already accounted for in the Level 1 importance list SPLT RA 5.OOE-01 1.03 SGTR OCCURS ON OTSG-A <SPLIT FRACTION> Already accounted for in the Level 1 importance list SPLT RB 5.OOE-01 1.03 SGTR OCCURS ON OTGS-B <SPLIT FRACTION> Already accounted for in the Level 1 importance list LMMDV43F 4.65E-03 1.03 FAILURE OF TRAIN B RECIRC VALVE DHV-43 Already accounted for in the Level 1 importance list LMMDV12F 4.65E-03 1.03 TRAIN B RECIRC VALVE DHV-12 FAILS Already accounted for in the Level 1 importance list QMMCST 6.46E-03 1.03 FAILURE OF CST WATER SUPPLY Already accounted for in the Level 1 importance list LMMDHPAF 4.90E-03 1.03 FAILURE OF DHP-1A AND ITS VALVES Already accounted for in the Level 1 importance list QHUEFW9Y 2.70E-03 1.03 OPERATORS FAIL TO RAISE OTSGs LEVEL Already accounted for in the Level 1 importance list H SPLT B 7.OOE-01 1.03 FRACTION OF SLOCAS IN COLD LEG LOCATIONS REQUIRING Already accounted for in the Level 1 importance list HSPLTB 7.00E-01 1.03 SECONDARY COOLING IE T16 1.OOE+00 1.03 LOSS OF MAKEUP Already accounted for in the Level 1 importance list FLG TQR 1.OOE+00 1.03 TAG EVENT - STUCK OPEN RELIEF Already accounted for in the Level 1 importance list FHUF6A1Y 1.90E-03 1.03 OPERATOR FAILS TO ISOLATE FLOOD F6A (CASE 1) Already accounted for in the Level 1 importance list LMMDV42F 4.65E-03 1.03 FAILURE OF TRAIN A RECIRC VALVE DHV-42 Already accounted for in the Level 1 importance list SPMRW3AM 8.60E-03 1.03 RWP-3A IN MAINTENANCE' Already accounted for in the Level 1 importance list LMMDV1 1F 4.65E-03 1.03 TRAIN A RECIRC VALVE DHV-1 1 FAILS Already accounted for in the Level 1 importance list LHULPRCY 2.50E-02 1.03 OPERATORS FAIL TO GO TO LOW PRESSURE RECIRCULATION Already accounted for in the Level 1 importance list RCCDRODA 1.00E-06 1.03 MECH FAILURE OF ENOUGH CONTROL RODS TO DROP Already accounted for in the Level 1 importance list HHUMBACY 1.00E+00 1.03 OPERATORS FAIL TO SWITCH MUP-1 B POWER SOURCE IN Already accounted for in the Level 1 importance list FLG PHURMFWR 1.OOE+00 1.03 OPERATORS FAIL TO RECOVER MFW Already accounted for in the Level 1 importance list ZHUCOM1Z 2.80E-01 1.03 COND PROB OF RHUPORVY GIVEN QHUEFW9Y Already accounted for in the Level 1 importance list This is a flag identifying battery depletion sequences, not FLGBATFA 1.OOE+00 1.03 FLAG EVENT - AC NOT RESTORED BEFORE DPBA-1A DEPLETES a basic event. However, a SAMA has been proposed to (APPROX. 4 HRS) provide a portable battery charger to extend battery life (SAMA 2).

SMMDHCCB 2.73E-03 1.03 DHCCC TRAIN B FAULTS Already accounted for in the Level 1 importance list SMMRW3BF 2.69E-03 1.03 RWP-3B PUMP TRAIN FAILS TO OPERATE Already accounted for in the Level 1 importance list SPMDHCBM 4.OOE-03 1.03 DHCCC TRAIN B IN MAINTENANCE Already accounted for in the Level 1 importance list ADGES3BM 3.37E-02 1.03 EGDG-1B IN MAINTENANCE Already accounted for in the Level 1 importance list

U. S. Nuclear Regulatory Commission Enclosure 3F1009-03 Page 9 of 74 SAMA RAI 2.b Response Event Name Probability Red W Description Potential SAMAs HCCMV44N 1.43E-04 1.02 COMMON CAUSE FAILURE OF MUV-23, MUV-24, MUV-25, AND MUV-26 Already accounted for in the Level 1 importance list HCCMV44N_______ 1.3E0 10 TO OPEN QHUEFT2Y 7.70E-04 1.02 OPERATORS FAIL TO CROSSTIE EFW SOURCES Already accounted for in the Level 1 importance list PMMICSBH 7.45E-02 1.02 OTSG B LEVEL CONTROL FAULTS Already accounted for in the Level 1 importance list IE T10 1.OOE+00 1.02 LOSS OF NSCCC Already accounted for in the Level 1 importance list PMMICSAH 7.45E-02 1.02 OTSG A LEVEL CONTROL FAULTS Already accounted for in the Level 1 importance list IEM 4.00E-05 1.02 MEDIUM BREAK LOCA Already accounted for in the Level 1 importance list ADGEG1CF 7.68E-02 1.02 EGDG-1C FAILS TO RUN Already accounted for in the Level 1 importance list HHUINJBY 5.OOE-01 1.02 OPERATORS FAIL TO SWITCH MUV-25/26 TO BACKUP POWER Already accounted for in the Level 1 importance list IET2 2.40E-01 1.02 LOSS OF MAIN FEEDWATER Already accounted for in the Level 1 importance list AHUEG1CY 5.OOE-01 1.02 OPERATORS FAIL TO START AND ALIGN EGDG-1 C Already accounted for in the Level 1 importance list MTC 2.50E-01 1.02 MTC GREATER THAN 95% Already accounted for in the Level 1 importance list

U. S. Nuclear Regulatory Commission Enclosure 3F1009-03 Page 10 of 74 2.c The contribution from the MK, SK, and RK sequences in the MOROI model was 7E-9 or less and was considered small enough to remove from the event tree. No specific cutoff frequency was listed in the calculation. Below is the description of the sequences.

MK Medium LOCA followed by a failure to trip (event K).

SK Small LOCA with failure of the reactorto trip.

RK SGTR with failure of the reactorto trip.

2. d A breakdown of CDF and the population dose by release category was given for each SAMA in Section E.6, and each release category is described in detail in Table E.2-3.

The CDF and population dose by containment release mode is given in Table E.3-7.

2.e.i CalculationP02-0011, Revision 2, provides the description of how the Level I results are mapped into the Level 2 analysis. The following paragraphs are taken from the calculation.

The CSET and the assignment of core damage and plant damage state bins are accomplished using a CAFTA model and a rule-basedrecovery file. The CAFTA model is joined with the core damage model and containment system models to create a plant damage state model. The core damage model is described in anotherdocument. The model used to generate the PDSs is described herein.

The model is comprised of two parts. The first assigns the core damage event tree endstates to the appropriatecore damage bin. The model solution combines the accident sequences developed in the core damage model with an appropriateflag event representingthe core damage bin (e.g., FLCDB1). This tags each sequence with its associatedcore damage bin.

The second part of the model develops the possible outcomes of the CSET (states A through S). The model is developed with the associated success logic inserted into the different states. This precludes core damage cut set replication in more than one plant damage state. The model includes state A (all working) to define a complete solution. Each CSET is assigned a flag event to provide a tag (FLPDSA).

Combining the CSET solutions with the core damage bin solutions provides the plant damage state solution with two tags (CDB and CSET state). A mutually exclusive gate deletes the PDS states for the interfacing system LOCA case since it is only assigned to PDS S.

The rule-based recovery file is then applied which adds the associated plant damage state (FLPDSIA) and deletes the individual events (FL_CDB1 and FLPDSA). This defines the complete PDS definition. The rule-based recovery file then uses the "class"statement to assign the results to individual classes.

This allows the results to be reported in terms of the plant damage state to support the level 2 analysis. The results are then available for review and use in support of the level 2 analysis. The fault trees, mutually exclusive events, and recovery rules are maintainedas controlledfiles as part of the Model of record.

U. S. Nuclear Regulatory Commission Enclosure 3F1009-03 Page 11 of 74 2.e.ii CalculationP02-0012, Revision 2, provides the process used to group the containment event tree end states into release categories. The following paragraphsare taken from the calculation.

The end states of the CET correspond to the outcome of possible severe accident sequences. Each end point defines a different containment state with an associated radionuclide release. Simplifications can be attained by grouping sequences with similarrelease characteristicsinto a release category that can be applied to the containment end states. A set of release categoriesis defined such that all accidents assigned to the same category are assumed to have the same set of release fractions.

The main characteristics of the containment end states considered when developing these release categories were.

Release Energy

Containment Isolation Failure Size

Timing of The Release

Isotopic Composition POs IVR H"R Ce I CI IFL ECF DEC I L LOP RVP SR I Nae RLEBS*

PLANTDAMAGE IN-VESSEL SO'SURN CONTAINMENT C-TT CONTBAIMAENIEARLY EX-VCESSEL LATE BATE LATE CRDIBOND5 CATEG(

STATEENTRY RECOVERY AFTERIVR BYPASSED ISOLATION EAI(AGE RATI COhTAINLENT COIONG RSSAPORZAT S HYDROGENCONTABBSENT 8BITG POINT FAILURE BURN FAiLUR. ON

¢BISOt.ATED SCRUBBED NLARGES 9 ILE AE E (.)75 RC-I M A R570 RC.7I NO N2*JRN[NOT SCRUBBE 77 RC-2 KOHBUN 2T- 76II RU-R.

LEW E SMALL r -

SCRUBED ; Rc-4 WROCUR$

NO 55 RU-lU]

5YA'LARGE LEAKAGE - l R-62 RC-5c SCRUBBE BRNFSSC 83 RC-2 RANSF TE Figure 8. CR3 Containment Event Tree (page 1 of 4) 2.e.iii The following table summarizes the mapping of the Level I sequences to the Level 2 release categories. Note that a single MAAP run was designated for each release category based on the representative case defined for the release category; no weighting of multiple MAAP cases was required.

The version of MAAP used in the SAMA evaluation is MAAP4.0.6, with executable dated December 13, 2005, 2:53:06 PM.

U. S. Nuclear Regulatory Commission Enclosure 3F1009-03 Page 12 of 74 SAMA RAI 2.e.iii Response Representative MAAP Level 2 Case Descriptions Case Release C Cat.

I Sequ. Representative Case Description MAAP Case Description MAAP ID 1 IC-1 SXP Small LOCA + SSHR success + injection success + no SMALL LOCA, W/ HPI INJECTION W/O RECIRC, IC1_CR3_406 recirculation - oper fails to do recirc, cc and random failures of SSHR AVAILIBLE, NO CONT SPRAYS AND RBCU'S both DH trains (Successful isolation, sprays fail in injection (FANS) AVAILABLE and fans succeed) 2 RC-1 TQX_P Transient + SSHR success + safety valve lifts and sticks open TRANSIENT AND SORV, W/ HPI INJECTION W/O RECIRC, RClCR3_406

+ high head injection success + failure to switch to recirc - SSHR AVAILABLE, NO CONT SPRAYS AND opers fail to prevent PZ overfill, fail to recirc (Successful RBCU'S(FANS) OFF @ VESSEL FAILURE isolation, sprays fail in injection and fans succeed) 3 RC-1A SBPP Small LOCA + operators fail to raise SG level (=inadequate SMALL LOCA AND PORV FAILS CLOSED, W/ HPI RC1a CR3_406 SSHR) + PORV fails to open - oper fails to control level, INJECTION, PORV (Successful isolation, sprays fail in injection and fans NO SSHR, NO CONT SPRAYS AND RBCU'S (FANS) succeed) OFF @ VESSEL FAILURE 4 RC-1B TBL1 UP Transient + SSHR failure + injection failure - HVAC and AFW TRANSIENT, W1O INJECTION, W/O SSHR, RC1 b_CR3_406 control probs (Successful isolation, sprays fail in injection and NO CONT SPRAYS AND RBCU'S(FANS) OFF @

fans succeed) VESSEL FAILURE, 4" DIAMETER CONTAINMENT FAILURE 5 RC-1BA TBL1 U_P Transient + SSHR failure + injection failure - HVAC and AFW TRANSIENT, W/O INJECTION, W/O SSHR, RC1b_CR3_406 control probs (Successful isolation, sprays fail in injection and NO CONT SPRAYS AND RBCU'S(FANS) OFF @

fans succeed) VESSEL FAILURE, 4" DIAMETER CONTAINMENT FAILURE 6 RC-2 TBL1 U_P Transient + SSHR failure + injection failure - HVAC and AFW TRANSIENT, W/O INJECTION, W/O SSHR, RC2_CR3_406 control probs (Successful isolation, sprays fail in injection and NO CONT SPRAYS AND RBCU'S(FANS) OFF @

fans succeed) VESSEL FAILURE, 1' DIAMETER CONTAINMENT FAILURE 7 RC-2B TQX_P Transient + SSHR success + safety valve lifts and sticks open TRANSIENT AND SORV, W/ HPI INJECTION W/O RECIRC, RC2bCR3_406

+ high head injection success + failure to switch to recirc - SSHR AVAILABLE, NO CONT SPRAYS AND opers fail to prevent PZ overfill, fail to recirc (Successful RBCU'S(FANS) OFF @ VESSEL FAILUR, 4" DIAMETER isolation, sprays fail in injection and fans succeed) CONTAINMENT FAILURE 8 RC-3 TBL1U P Transient + SSHR failure + injection failure - HVAC and AFW TRANSIENT, W/O INJECTION, W/O SSHR, RC3_CR3_406 (Note 1) control probs (Small isolation failure, sprays fail in injection W/ CONT SPRAYS AND RBCU'S(FANS) OFF @

and fans succeed) VESSEL FAILURE, 4" DIAMETER CONTAINMENT FAILURE 9 RC-3B TBL1U_P Transient + SSHR failure + injection failure - HVAC and AFW TRANSIENT, W/O INJECTION W/O SSHR, RC3b_CR3_406 control probs (Small isolation failure, sprays fail in injection W/O CONT SPRAYS AND RBCU'S(FANS) OFF @

and fans succeed) ,VESSEL FAILURE, 4" DIAMETER CONTAINMENT FAILURE 10 RC-4C RCP_P SGTR + failure to cooldown/depressurize using secondary SGTR WITH STUCK-OPEN S/G PORV, NO SSHR, RC4cCR3_406 side + failure to depressurize using PORV - opers fail to NO CONT SPRAYS AND RBCU'S(FANS) ARE I cooldown, open PORV (Containment bypass) AVAILABLE 11 RC-5C ISLOCP LLOCA outside containment + injection failure - DHR drop line ISLOCA, W/ FAILURE OF ALL INJECTION, RC5cCR3_406 (Containment bypass) NO CONT SPRAYS AND RBCU'S(FANS) ARE AVAILABLE Note 1: TBLX is the correct sequence that was meant to be referenced. Although this was a typographical error in the original SAMA report, the MAAP run for this Release Category was performed correctly. The TBLX sequence represents a transient with an extended loss of heat removal via the SGs. Feed-and-bleed cooling succeeds, but feedwater cannot be restored in the long term, and HPR fails. This scenario eventually leads to core damage.

U. S. Nuclear Regulatory Commission Enclosure 3F1009-03 Page 13 of 74 2.e.iv See response to RAI 2.b.

Table E.3-7 lists the dose-risk (p-rem/yr) for each release category, which reveals that categories3B and 4C accounted for 75% of the total dose-risk.

3. Provide the following information with regard to the treatment and inclusion of external events in the SAMA analysis:

a. Provide fire CDF by fire zone/area and the total fire CDF for CR-3. If the fire CDF is different than that reported in the Individual Plant Examination for External Events (IPEEE), provide an explanation for the differences.

b. For each of the dominant fire areas, explain what measures, if any, have already been taken (since the IPEEE) to reduce fire risk. Include in the response specific improvements to fire detection systems, enhancements to fire suppression capabilities, changes that would improve cable separation, and improvements to processes/procedures for monitoring and controlling the quantity of combustible materials in critical areas.

c. The SAMA analysis assumes that risks posed by external and internal events is approximately equal (page E.5-7). Based on this assumption, the estimated benefit from reduction of internal event risk was doubled to account for a corresponding reduction in external event risk. However, Section 1.4 of the IPEEE identifies the calculated CDF from fires to be 4.2E-05 per year, a factor of 8.5 greater than the internal events CDF (4.95E-06 per year) used in the SAMA analysis. Furthermore, while a seismic CDF was not developed for the IPEEE, the U.S. Nuclear Regulatory Commission (NRC) staff estimates the seismic CDF for CR-3 to be about 1.2E-05 per year using the approximation method described in a paper by Robert P. Kennedy, "Overview of Methods for Seismic PRA and Margin Analysis Including Recent Innovations" and using updated 2008 seismic hazard curve data from the U.S. Geologic Survey (USGS). Based on this, provide justification for why a multiplier of 12 [(4.2E-05 + 1.2E-05) / 4.95E-06 + 1]

shouldn't be used to account for the additional risk of all external events (seismic, fire, high winds, etc.) rather than the multiplier of two used in the SAMA analysis.

d. Provide an assessment of the impact on the initial and final SAMA screenings if the internal events risk reduction estimate is increased by a factor of 12 (or a smaller factor for which sufficient basis can be provided). Provide a Phase II analysis for any Phase I SAMAs that were screened out in the Environmental Report (ER) but would not have been with the higher factor.

e. Section E.5.1.6 of the ER notes that at the time of the 1997 IPEEE submittal, the plan was for CR-3's plant-specific response to unresolved safety issue (USI) A-46, "Seismic Qualification of Equipment in Operating Plants," to sufficiently address seismic risk. The USI A-46 safety evaluation report (SER) in 2000 identified three topics that required additional work to resolve: 1) one equipment seismic capacity outlier, 2) five outliers associated with differences between the caveats in generic implementation procedure (GIP)-2 and those in the plant-specific procedure (PSP), and 3) revision of abnormal procedure (AP)-961. The USI A-46 SER further states that each of these issues is being tracked in the CR-3 Corrective Action Program. ER Section E.5.1.6 is silent as to the

U. S. Nuclear Regulatory Commission Enclosure 3F1009-03 Page 14 of 74 status of these issues. Clarify the resolution status of each of these unresolved issues. If still unresolved, identify and evaluate seismic SAMAs to address each unresolved issue.

Response

3.a Table 1.4-1 was extracted from the CR-3 IPEEE report. This table represents the majority of the core damage risk due to fire for CR-3 and is associated with fires occurring in fire zones within the control complex. Table 1.4-1 lists the fire zones which have a core damage frequency due to fire of greaterthan lx10-6 per year, along with the control room and cable spreading room.

Table 1.4-1 Fire Zone Core Damage Frequencies ZAon. ., DO$CrptOn : .. COF CC-108-106 BATTERY CHARGER ROOM 3A 1.49E-O5 CC-108-108 4160V ES SWITCHGEAR BUS ROOM 3A 7.31E-06 CC-108-107 4160V ES SWITCHGEAR BUS ROOM 3B 6,79E-06 CC-124-117 480V ES SWITCHGEAR BUS ROOM 3A 3,79E-06 CC-108-105 BATTERY CHARGER ROOM 35 2.72E-06 CC-108-102 HALLWAY AND REMOTE SHUTDOWN ROOM 2.66E-06 CC-124-111 CRD &COMMUNICATION EQUIP ROOM 1.58E-06 CC-108-109 INVERTER ROOM 38 1.45E-0 CC-145-118B CONTROL ROOM 5.7E-07 CC-134-118A CABLE SPREADING ROOM 9,9E-08 The fire CDF beyond the IPEEE has not been performed.

3.b Since completion of the IPEEE review by the NRC staff, various plant modifications that may impact fire risk have been undertaken; however, no quantification of the fire risk impact has been performed. A specific fire PRA will be completed as part of the ongoing CR-3 transitionto NFPA 805, "PerformanceBased Standardfor Fire Protection for Light Water Reactor Electric Generating Plants", 2001 edition. Specific fire protection related modifications that have been performed include:

0 Installation of Emergency Lighting 0 Improve separationof electricalcables,

Improved administrativecontrols of transientcombustibles,

Modifications resulting from the ongoing NFPA Code review:

o Fire Detector Upgrade (in progress),

o Suppression system upgrades, e.g., improvement of sprinkler systems in the Fire Pump House and Control Complex 95 ft. elevation, o Upgradedprogrammaticcontrols for penetration seals.

3. c CR-3 has not performed a seismic analysis and while the approximation method described in "Overview of Methods for Seismic PRA and Margin Analysis Including Recent Innovations" is a conservative method, it is beyond the scope of the SAMA analysis to develop a seismic PRA.

U. S. Nuclear Regulatory Commission Enclosure 3F1009-03 Page 15 of 74 CR-3 is in the process of developing an updated fire PRA, but the model is incomplete and updated CDF information is not available. The level of effort required to provide quantitativejustifications for reducing the fire CDFs documented in the IPEEE is also considered to be beyond the scope of the SAMA analysis.

As a result, no exceptions are taken to the proposed use of a multiplier of 12 to account for external events contributorsat CR-3.

3.d As documented in the response to RAI 3.c, the resourcesrequired to provide a detailed, quantitativejustification for an external events multiplierof less than 12 are not available.

This response provides an assessment of the impact of using a multiplier of 12 to account for the external events contributorsin the SAMA analysis.

With regard to the impact of the largerexternal events multiplier on the Phase I and II analyses, both phases would be affected. In the Phase I analysis, the larger external events multiplier would increase the value of the MMACR, which could result in the retention of SAMAs that were previously screened from further analysis. In the Phase II analysis, the larger external events multiplier would increase the SAMA specific averted cost-risk values, which may result in the identification of additional cost beneficial SAMAs. In order to address the impact on the "final"SAMA screenings, the impact of the use of the 9 5th percentile PRA results is also examined in conjunction with the increasedexternal events multiplier.

Phase I Impact While the use of the external events multiplier of 12 will increase the MMACR and may prevent the screening of some of the higher cost modifications, the impact on the overall SAMA results due to the retention of the higher cost SAMAs for Phase II analysis is typically small. This is due to the fact that the benefit obtained from the implementation of those SAMAs must be extremely large in order to be cost beneficial. Because the MMACR calculations scale linearly with external events multiplier, the revised MMACR can be determined by multiplying the original MMACR by the ratio of the new external events multiplier to the originalexternal events multiplier:

MMACR12EEx = MMACRbase x 12/2 = $682, 000 x 6 = $4,092,000 As documented in Section E.7.2 of the ER, the use of the 9 5 th percentile PRA results will increase the MMACR by a factor of 2.18, which results in a value of $8,920,560

($4,092,000 x 2.18 = $8,920,560). The SAMAs that were initially screened in the Phase I analysis have been re-examined using the revised 9 5 th percentile MMACR of

$8,920,560 to identify SAMAs that would be retained for the Phase II analysis. Those SAMAs that were previously screened because their Costs of implementation exceeded

$500,000 are now retained if their costs of implementation are less than $8,920,560 million (Section E.5.2 of the ER identifies that $500,000 was used as the Phase I screening value for CR-3). Because some of the SAMAs that were originally screened in the Phase I analysis used rough cost estimates that only indicated that they would exceed certain values, revised cost estimates have been assigned to these SAMAs to aid in this evaluation. The following table provides the updated Phase I screening results for those SAMAs that were initially screened in the Phase I analysis:

U. S. Nuclear Regulatory Commission Enclosure 3F1009-03 Page 16 of 74 SAMA Number SAMA Title SAMA Description Cost Estimate Retaine Retained Numbe for Phase H 8 Temp pump to Provide a temporarypump or $500,000 Yes replace RWP pump of alternatedesign and suction supply which can be aligned to supply cooling water in lieu of RWP 26 Install separate Install separateindependent Browns Ferry SAMA Yes and independent cooling for EFIC (considerDC analysis estimated the EFIC cooling power, self-powered fans) cost of installing a system redundanttrain of room cooling for RHR, CS, and the EDGs to be $6 million per unit. For CR-3, the cooling train would have to be independent and battery powered.

Because the TVA estimate addresses 3 areascompared with the single area for CR-3, the estimate has been divided by 3. A cost estimate of $2 million is used for this application.

14 Automate SG This is in respect to HRA event The Farley SAMA Yes level control QHUEFW9Y, "Operatorsfail to analysis estimated that a requirementsfor raise O TSGs level." This is digital Feedwater SBLOCA needed for small LOCA Upgrade would cost response. Proposed SAMA, about $900,000. This is automate level control. considered to be similar in scope to automating SG level control for small LOCAs.

37 DH HX strainers Add removable strainers $600,000 Yes aheadof heat exchangers 1 Automate Automate alignment of $1,000,000 Yes EFIC/inverter dedicated chilled water system backup cooling to cool inverters and EFIC when required 7 New AFW Add a new independentAFW $5,000,000 Yes suction source source and pump and pump 16 Enhance With respect to SMMDHCCB, $5,000,000 Yes procedures and "DHCCCtrain B faults." (same make design for A, same for similarfailures changes as in other systems). Proposed required to SAMA: Proceduralize facilitate crosstying between train A and crosstying trains train B of MU, DH, DHCC of DH, DHCC, trains at appropriatesuction /

I etc. dischargepoints.

U. S. Nuclear Regulatory Commission Enclosure 3F1009-03 Page 17 of 74 SAMA Number SAMA Title SAMA Description Cost Estimate RetaineT Retained Nube for Phase HI 18 Add another With respect to BE The cost of installingan Yes EDG ADGES3BM, "EGDG-IB in additionalEDG ranges maintenance." (same for A) greatly depending on the Proposed SAMA: add another application,size, and the EDG. organization implementing the change. Forexample, the Farley SAMA analysis used an estimate of $74,500,000 while the PrairieIsland SAMA analysis estimated a cost of

$8,000,000 for a single SBO EDG. While the PrairieIsland application is for a swing EDG that may not include the auto start capability that may be desirable for preventing RCP seal LOCAs, it is used as a lower bound estimate.

13 Add additional This is in respect to BE No industry estimates Yes, for train of DH, of LPMOOIA, DHP-1A train in have been identified for demonstration diverse design maintenance. (similarfor B installing a complete, purposes train) Add an additionaltrain diverse train of DH, but or "maintenance"train of the Calvert Cliffs SAMA diverse design. analysis includes an estimate of $5,000,000-

$10, 000, 000 for the installationof an independent HPSI pump. An independent HPSI pump does not address the heat removal function of a DH train or the diverse piping requirements, but

$10,000,000 is used as a lower bound estimate for this analysis.Even though it is expected that the actual implementation cost may be closer to $50 million, this SAMA has been retained for Phase II analysis to clearly demonstrate its low averted cost-risk.

U. S. Nuclear Regulatory Commission Enclosure 3F1009-03 Page 18 of 74 SAMAF AMA SAMA Title ]IFRetained SAMA Description Cost Estimate fReased Number I for Phasell 52 Install parallel This is in respect to BE This requires an Yes flowpath for DHR LMMDHRSF, which was an additionalinterface in the drop line important contributor(RRW > RCS with MOVs and 1.02) for sequences leading to controls in the MCR.

Release Category 4C This is similarin nature to the cross-ties proposed in SAMA 16 and the $5,000,000 estimate is also applied to this SAMA.

All of the SAMAs that were previously screened in the Phase I analysis would be retained for Phase II evaluation when the external events multiplier of 12 is used in conjunction with the 9 5 th percentile PSA results.

PhaseII Impact The same process used to scale up the MMACR to account for the external events multiplier of 12 can be used for the averted cost-risk estimates for each of the Phase II SAMAs. This includes those SAMAs that were originally evaluated in the ER's Phase II analysis as well as the SAMAs that were originallyscreened in the Phase I analysis. For the SAMAs that were originally screened in the Phase I analysis, however, Phase II quantificationsare required.

The exception to this is that SAMA 49 must be re-quantified using the assumptions about the external events contributionsidentified in RAI question 3c.

These quantificationsand their results are provided below. It should be noted that the intermediate and final results documented in this RAI response have been rounded for presentation purposes. The actual calculations that were performed for the Phase II quantifications may include additional significant digits that are not apparent in the documentation. Consequently, it may not be possible to exactly reproduce the Phase II evaluations using the intermediate results that are presented in this response. For example, the sum of the baseline dose-risk across all release categories is 3.77 person-rem/yr when summing the values presented in the results table; however, because extra digits not shown in the table were used in the actual summation, a slightly higher result of 3.79 person-rem/yr was obtained. These differences are negligible.

SAMA 49: Upgrade Fire Compartment Barriers The original quantification of SAMA 49 was based on the assumption that the internal and external events risks were approximately equal and that the external events contributions were distributed among the different external events contributors in proportion to their CDFs. The basis of the external events multiplier of 12 in RAI question 3c is that the external events CDF is directly proportionalto the internal events component of the MACR.

U. S. Nuclear Regulatory Commission Enclosure 3F1009-03 Page 19 of 74 It follows, therefore, that the averted cost-risk for SAMA 49 can be estimated by multiplying the internal events MACR by the ratio of the Battery ChargerRoom 3A fire CDF to the internalevents CDF:

$341,000 x 1.49E-05 / 4.95E-06 = $1,026,444 This result, in conjunction with the $150,000 Cost of Implementation (COI) that was estimated for this SAMA in the ER, can be used to recalculate the net value for SAMA 49:

SAMA 49 Net Value Averted CO1 Net Value Cost-Risk

$1,026,444 $150,000 $876,444 The net value for this SAMA is the averted cost-risk minus the COl, or $876,444

($1,026,444 - $150,000 = $876,444).

SAMA 8: Provide a Temporary Pump to Provide a Backup Supply of Cooling Water in Lieu of Raw Water Pump The Nuclear Services and Decay Heat Seawater (RW) system is comprised of two sub-systems, the Nuclear Services Sea Water (SW) portion (cooling the Nuclear Service Closed Cycle Cooling system) and the Decay Heat portion (cooling the decay heat closed cycle cooling system - DC and eventually the decay heat system - DH). The function for RW-DC is what is addressedby this particularSAMA.

The RW-DC system is safety-related and serves as the primary means of transferring heat from the DH system to the ultimate heat sink (Gulf of Mexico). The RW-SW is an open system comprised of two trains each providing cooling to the associatedDC train.

Although the RW-DC and RW-SW systems are two different systems, the pumps share a common suction pit (one per train) and inlet piping along with common piping returning the water to the ultimate heat sink (see Figure 1).

The RW-DC pumps take suction from the pit and discharge through the tube side of a single pass tube and shell heat exchangers (DCHE-1A and DCHE-1B). This provides cooling for the Decay Heat Closed Cycle Cooling system. After the water leaves the heat exchangers, it is returned to the ultimate heat sink through an undergroundflume to the discharge canal. Each train is designed to supply 100% of the cooling water needed for removal of the heat load from essential equipment during emergency conditions.

The primary function of the RW-DC system is to remove reactor decay heat following normal plant shutdown. 'When the DH system is placed in service during a normal plant cooldown, the correspondingRW-DC train is placed in service. The system is aligned for operation while in standby; therefore, once the RW-DC pump is started and the pump discharge check valve opens, the system provides the needed cooling. For emergency operation, the RW-DC pumps receive an ES start signal on either low RCS pressure or high reactor building pressure. The RW-DC system is in an ES alignment while in

U. S. Nuclear Regulatory Commission Enclosure 3F1009-03 Page 20 of 74 standby and upon receiving the actuationsignal, the RW-DC pumps begin supplying the requiredflow.

The purpose of this particularSAMA is to provide a backup raw water pump capable of supplying sea water to either of the DH system coolers (DCHE-IA(B)). Since the pump is assumed to be of a different design, it was assumed that the motive power for this alternate pump would be supplied by an air-cooled diesel engine, so as not to rely on any of the support dependencies normally used by the RW-DC pumps RWP-3A(B). The suction source for this pump would come from the same common suction pit as is used by the normal standby pumps. A crosstie just upstream of the RWP-3A(B) pump discharge valves would need to be installed with normally closed manual valves between the two trains, such that this alternate diesel-driven pump could supply either, but not both, of the two DHCCC trains (see Figure 1). A connection to this crosstie from the discharge of the proposedbackup RW-DC pump would need to also be installed, the discharge of which would also consist of a manual discharge valve (normally open) and a check valve to prevent backflow from any of the already installed RW-DC pumps should they be in a running condition.

Assumptions

1. The HEP event that was used for this SAMA analysis was assumed to be similar in nature to the event used for SAMA 10 in Section E.6.4 of the Environmental Report, which was assigned a failure probability on the order of IE-2.

2. Any dependencies between the new HEP event postulated for this SAMA and other existing HEPs in the PRA model were assumed to be negligible. This is a conservative assumption that will tend to bound the risk benefits.

3. In modeling this SAMA, it was assumed that the component failure probabilitiesobtained from NUREG/CR-6928 would be applicable.

PRA Model Changes to Model SAMA To calculate the consequences of implementation of this SAMA, the logic beneath gates S800 and S800AH was modified to account for this proposed backup RW-DC system.

Also, to separate the decay heat cooler faults from the RWP failures that were under OR gate S920, two new gates were created, i.e., S920_1 for non-DCH faults and S920_2 for DCH faults. That is, a backup RW pump would not be able to mitigate any failures associated with heat exchanger faults associated with components DCHE-1A and DCHE-1B. Similarly, the logic structure under OR gate S800_X for recirculation mode was also modified to allow recovery of RW-DC using this backup RWpump.

To better illustrate the logic changes made, Figures 2a through 2c represent that portion of the original logic of the CR-3 PRA model that was considered applicable to this SAMA, and Figures 3a through 3c represent the additional and modified logic that was used to simulate implementation of this SAMA.

The table below shows the new basic events and theirprobabilitiesthat were included in the PRA model to represent this SAMA implementation:

U. S. Nuclear Regulatory Commission Enclosure 3F1009-03 Page 21 of 74 SAMA 8 New Basic Events Basic Event Description Probability Comments SHUSAMA8X FAILURE OF OPERATOR TO ALIGN FOR SAMA 8New Basic Events IE-02 HEP assigned a failure probabilitybased on A L TERNA TE RWP OPERATION what was used for SAMA 10.

SPMSAMA8F FAILURE OFALTERNATE RWP TO RUN 2.28E-03 Failure to run based on (NRC 2007) and 24 (DIESEL-DRIVEN) hour mission time.

SPMSAMA8A FAILURE OFALTERNATE RWP TO START 3.88E-03 Failureprobabilitybased on (NRC 2007).

(DIESEL-DRIVEN)

SXVSAMA8N FAILURE OF MANUAL VALVE TO OPEN 7.43E-04 Failureprobability based on (NRC 2007).

U. S. Nuclear Regulatory Commission Enclosure 3F1009-03 Page 22 of 74

'SAMA B co(

0 e- 1ja L) i,,

d FIGURE 1 SAMA 8 CONCEPTUAL DESIGN

U. S. Nuclear Regulatory Commission Enclosure 3F1009-03 Page 23 of 74 K

FIGURE 2a DEPICTION OF ORIGINAL FAULT TREE LOGIC FOR GATE S800

U. S. Nuclear Regulatory Commission Enclosure 3F1009-03 Page 24 of 74 FIGURE 2b DEPICTION OF ORIGINAL FAULT TREE LOGIC FOR GATE S800AH

U. S. Nuclear Regulatory Commission Enclosure 3F1009-03 Page 25 of 74 FIGURE 2c DEPICTION OF ORIGINAL FAULT TREE LOGIC FOR GATE S800_X

U. S. Nuclear Regulatory Commission Enclosure 3F1009-03 Page 26 of 74 F

FIGURE 3a DEPICTION OF FAULT TREE LOGIC USED TO SIMULATE IMPLEMENTATION OF SAMA 8 FOR GATE S800

U. S. Nuclear Regulatory Commission Enclosure 3F1009-03 Page 27 of 74 FIGURE 3b DEPICTION OF FAULT TREE LOGIC USED TO SIMULATE IMPLEMENTATION OF SAMA 8 FOR GATE S800AH

U. S. Nuclear Regulatory Commission Enclosure 3F1009-03 Page 28 of 74 I---I FIGURE 3c DEPICTION OF FAULT TREE LOGIC USED TO SIMULATE IMPLEMENTATION OF SAMA 8 FOR GATE S800_X

U. S. Nuclear Regulatory Commission Enclosure 3F1009-03 Page 29 of 74 Results of SAMA Quantification Implementation of this SAMA yielded a moderate reduction in CDF, but only a relatively small change in the Dose-Risk, and Offsite Economic Cost-Risk (OECR). The results are summarized in the following table for CR-3:

CDF Dose-Risk OECR Base Value 4.95E-06 3.79 $6,624 SAMA Value 3.86E-06 3.75 $6,582 PercentChange 22.1% 1.1% 0.6%

A further breakdown of the Dose-Risk and OECR information is provided in the table below accordingto release category:

Release ICI RC1 RCIA RCIB RCIAB RC2 RC2B RC3 RC3B RC4C RC5C Total Category FrequencYBASE 4. 10E-06 2.44E-08 4.71E-10 1.59E-08 1.25E-09 8.43E-10 3.46E-09 2.46E-07 1.57E-07 3.44E-07 5.15E-08 4.95E-06 FrequencYsAMA 3.04E-06 1.89E-08 4.02E-10 1.30E-08 9.88E-10 6.82E-10 2.57E-09 2.29E-07 1.56E-07 3.43E-07 5.13E-08 3.86E-06 Dose-RiskBAsE 0.04 0.01 0.00 0.03 0.00 0.00 0.01 0.07 0.30 2.57 0.74 3.79 Dose-RisksAMA 0.03 0.00 0.00 0.03 0.00 0.00 0.01 0.07 0.30 2.56 0.74 3.75 OECRBASE $0 $1 $0 $29 $2 $7 $27 $15 $679 $4,855 $1,009 $6,624 OECRsAMA $0 $0 $0 $24 $2 $6 $20 $14 $673 $4,836 $1,006 $6,582 This information was used as input to the cost-benefit calculation. The results of this calculation are provided in the following table:

SAMA 8 Net Value Base Case Revised Averted COl Net Value Cost-Risk Cost-Risk Cost-Risk

$4,092,000 $3,734,028 $357,972 $500,000 -$142,028 The SAMA 8 results indicate a moderate reduction in CDF with relatively small changes in dose-risk and offsite economic consequences. Even though the cost of implementation is $500,000, the net value for this SAMA is -$142,028 ($357,972 -

$500,000 = -$142,028).

SAMA 26: Install a Separate and Independent EFIC Room Coolinq System The EFW Initiation and Control (EFIC) system provides for automatic initiation of the Emergency Feedwater System when needed and contains the logic circuitry responsible for appropriate feedwater valve isolation and alignment. Once EFW is initiated, its controls are governed by the EFIC system.

If the HVAC system responsible for cooling the rooms containing the EFIC cabinets and electronic components is unavailable, the system is assumed to fail due to overheating.

This SAMA provides for an alternate and independent backup room cooling system for both trains of the EFIC system.

U. S. Nuclear Regulatory Commission Enclosure 3F1009-03 Page 30 of 74 Assumptions For both EFIC rooms A and B, a single point estimate of the unavailability for each postulated backup system was assumed that would take into account all possible failure modes, which would include mechanical,electrical,and operatoractions.

PRA Model Changes to Model SAMA To calculate the consequences of implementation of this SAMA, a new AND gate (Q141_SAMA26) was inserted one level above OR gate Q141. Gate Q141_SAMA26 is modeled as an input to OR gate Q032 and OR gate Q082. The inputs to AND gate Q141_SAMA26 are the original Q141 gate logic and the new postulated SAMA event SAMA26-EFIC-B, which represents the unavailabilityfor the EFIC room B HVAC backup system.

Likewise, for train A of EFIC, a new AND gate was created (Q140_SAMA26), which contained the original Q140 gate logic and the new SAMA event SAMA26-EFIC-A as inputs. Gate Q140_SAMA26 is modeled as an input to AND gate Q149.

SAMA 26 New Basic Events Basic Event Description Probability Comments SAMA26-EFIC-A HVAC BACKUP FOR EFIC IE-02 Assumed unavailability to account ROOM A FAILS for all possible failure mechanisms, including operatoractions, for this HVAC backup system.

SAMA26-EFIC-B HVAC BACKUP FOR EFIC IE-02 Assumed unavailability to account ROOM B FAILS for all possible failure mechanisms, including operatoractions, for this HVAC backup system.

Results of SAMA Quantification Implementation of this SAMA yielded a moderate reduction in CDF, but only a relatively small change in the Dose-Risk, and Offsite Economic Cost-Risk (OECR). The results are summarized in the following table for CR-3:

CDF Dose-Risk OECR Base Value 4.95E-06 3.79 $6,624 SAMA Value 3. 73E-06 3.52 $6,151 Percent Change 24.5% 7.1% 7.1%

A further breakdown of the Dose-Risk and OECR information is provided in the table below accordingto release category:

U. S. Nuclear Regulatory Commission Enclosure 3F1009-03 Page 31 of 74 Release IC1 RC1 RCIA RCIB RCIAB RC2 RC2B RC3 RC3B RC4C RC5C Total Category FrequencyBASE 4. 10E-06 2.44E-08 4.71E-10 1.59E-08 1.25E-09 8.43E-10 3.46E-09 2.46E-07 1.57E-07 3.44E-07 5.15E-08 4.95E-06 FrequencysAmA 3.15E-06 2.37E-08 3.90E-10 1.06E-08 6.57E-10 4.64E-10 3.OOE-09 9.61E-08 5.26E-08 3.45E-07 5.15E-08 3.73E-06 Dose-RiskBASE 0.04 0.01 0.00 0.03 0.00 0.00 0.01 0.07 0.30 2.57 0.74 3.79 Dose-RisksAMA 0.03 0.00 0.00 0.02 0.00 0.00 0.01 0.03 0.10 2.57 0.74 3.52 OECRBASE $0 $1 $0 $29 $2 $7 $27 $15 $679 $4,855 $1,009 $6,624 OECRsAMA $0 $1 $0 $20 $1 $4 $23 $6 $227 $4,860 $1,009 $6,151 This information was used as input to the cost-benefit calculation. The -results of this calculation are provided in the following table:

SAMA 26 Net Value Base Case Revised Averted CO/ Net Value Cost-Risk Cost-Risk Cost-Risk

$4,092,000 $3,536,400 $555,600 $2,000,000 -$1,444,400 The SAMA 26 results indicate a moderate reduction in CDF with relatively small changes in dose-risk and offsite economic consequences. Given the cost of implementation of

$2,000,000, the net value for this SAMA is -$1,444,400 ($555,600 - $2,000,000 =

-$1,444,400).

SAMA 14: Automate Steam GeneratorLevel Controls for Small LOCA Operators are directed by procedure to establish OTSG levels at 80-90% (or achieve adequate flow) given a loss of adequate subcooling margin. The relevant sequences for this operatoraction involve a small-break LOCA where not enough water is lost out of the break to remove decay heat. Therefore, enhanced primary-to-secondary heat transfer is needed to remove the balance of the decay heat. For other scenarios, such as a loss of all RCPs, EFIC automatically controls to the natural circulation level. The operators are directed to raise secondary water level in the steam generatorsto 80-90%

by selecting the ISCM setpoint. If the level is not achieved, the minimal criterion is to achieve the flow rates dictated by the EOP's. To improve the reliabilityof this action, this SAMA simulates the use of automatic controls to accomplish this action of raising OTSG water level.

Assumptions To simulate the use of automatic controls, it was assumed that the operator failure probabilitycould be reduced to simulate the use of more reliable automaticcontrols.

PRA Model Chanties to Model SAMA The only change made to the PRA model to simulate implementation of this SAMA was the reduction of the failure probabilityof QHUEFW9Y from 2. 7E-3 to 2. 7E-6. This failure reduction conservatively bounds the risk benefit by implying that automatic controls are three orders of magnitude more reliable than an operatorusing manual control.

U. S. Nuclear Regulatory Commission Enclosure 3F1009-03 Page 32 of 74 Results of SAMA Quantification Implementation of this SAMA yielded a small reduction in CDF, Dose-Risk, and Offsite Economic Cost-Risk (OECR). The results are summarized in the following table for CR-3:

CDF Dose-Risk OECR Base Value 4.95E-06 3.79 $6,624 SAMA Value 4.79E-06 3.78 $6,610 Percent Change 3.2% 0.3% 0.2%

A further breakdown of the Dose-Risk and OECR information is provided in the table below according to release category:

Release ICI RCI RCIA RCIB RCIAB RC2 RC2B RC3 RC3B RC4C RC5C Total Category FrequencyBASE 4. 1OE-06 2.44E-08 4.71E-10 1.59E-08 1.25E-09 8.43E-10 3.46E-09 2.46E-07 1.57E-07 3.44E-07 5.15E-08 4.95E-06 FrequencysAMA 3.95E-06 2.31E-08 3.30E-10 1.57E-08 1.23E-09 8.42E-10 3.27E-09 2.45E-07 1.57E-07 3.44E-07 5.13E-08 4.79E-06 Dose-RiskaASE 0.04 0.01 0.00 0.03 0.00 0.00 0.01 0.07 0.30 2.57 0.74 3.79 Dose-RisksAMA 0.04 0.00 0.00 0.03 0.00 0.00 0.01 0.07 0.30 2.57 0.74 3.78 OECRBASE $0 $1 $0 $29 $2 $7 $27 $15 $679 $4,855 $1,009 $6,624 OECRSAMA $0 $1 $0 $29 $2 $7 $25 $15 $677 $4,849 $1,005 $6,610 This information was used as input to the cost-benefit calculation. The results of this calculation are provided in the following table:

SAMA 14 Net Value Base Case Revised Averted COl Net Value Cost-Risk Cost-Risk Cost-Risk

$4,092,000 $4,037,472 $54,528 $900,000 -$845,472 The SAMA 14 results indicate a small reduction in CDF, dose-risk and offsite economic consequences. Given a cost of implementation of $900,000, the net value for this SAMA is -$845,472 ($54,528 - $900,000 = -$845,472).

SAMA 37: Install Removable Strainers Upstream of Decay Heat Coolers to Mitigate Common Cause Pluagging Concerns The Decay Heat Seawater (RW) system is responsible for providing cooling water to the decay heat closed cycle heat exchangers DCHE-1A and DCHE-1B, which serve as the ultimate heat sink for removal of decay heat during shutdown cooling or emergency conditions. Since DCHE-1A and DCHE-1B are susceptible to common cause failure due to plugging from debris, this SAMA provides an option to remediate this possible failure via use of removable strainersupstream of the heat exchangers to capture any possible debris and prevent the heat exchangers from being plugged.

U. S. Nuclear Regulatory Commission Enclosure 3F1009-03 Page 33 of 74 One possible design for these upstream strainers would be that of a duplex arrangement, in which it is possible for the operatorto swap from one in-line strainerto another without having to breach the decay heat seawatersystem.

Assumptions To reduce the susceptibility of both DCHE-1A and DCHE-1B to plugging failure, it was assumed that an equipment operator could easily swap to a clean strainer, e.g., via duplex strainerarrangement,for one or both of the affected decay heat seawater trains with enough time available to avert any possible core damage.

PRA Model Changes to Model SAMA The only change made to the PRA model to simulate implementation of this SAMA was the reduction of the failure probability of SCCHDABF from 2.39E-4 to 2.39E-6. This failure reduction conservatively bounds the risk benefit by implying that common cause plugging of both DCHE-1A and DCHE-IB can be reduced by two orders of magnitude by means of an equipment operator manually controlling strainer configuration to avoid debris from stopping the flow of cooling water to these heat exchangers.

Results of SAMA Quantification Implementation of this SAMA yielded a small reduction in CDF, Dose-Risk, and Offsite Economic Cost-Risk (OECR). The results are summarized in the following table for CR-3:

CDF Dose-Risk OECR Base Value 4.95E-06 3.79 $6,624 SAMA Value 4.78E-06 3.79 $6,619 PercentChange 3.4% 0.1% 0.1%

A further breakdown of the Dose-Risk and OECR information is provided in the table below according to release category:

Release ICI RC1 RCIA RCIB RCIAB RC2 RC2B RC3 RC3B RC4C RC5C Total Category FrequencyBAsE 4. 10E-06 2.44E-08 4.71E-10 1.59E-08 1.25E-09 8.43E-10 3.46E-09 2.46E-07 1.57E-07 3.44E-07 5.15E-08 4.95E-06 FrequencysAMA 3.94E-06 2.29E-08 4.55E-10 1.57E-08 1.24E-09 8.42E-10 3.26E-09 2.45E-07 1.57E-07 3.44E-07 5.16E-08 4.78E-06 Dose-RiskBASE 0.04 0.01 0.00 0.03 0.00 0.00 0.01 0.07 0.30 2.57 0.74 3.79 Dose-RisksAMA 0.04 0.00 0.00 0.03 0.00 0.00 0.01 0.07 0.30 2.57 0.74 3.79 OECRsASE $0 $1 $0 $29 $2 $7 $27 $15 $679 $4,855 $1,009 $6,624 OECRsAMA $0 $0 $0 $29 $2 $7 $25 $15 $678 $4,850 $1,012 $6,619 This information was used as input to the cost-benefit calculation. The results of this calculation are provided in the following .table:

U. S. Nuclear Regulatory Commission Enclosure 3F1009-03 Page 34 of 74 SAMA 37 Net Value Base Case Revised Averted COl Net Value Cost-Risk Cost-Risk Cost-Risk

$4,092,000 $4,037,916 $54,084 $600,000 -$545,916 The SAMA 37 results indicate a small reduction in CDF, dose-risk, and offsite economic consequences. Based on the $600,000 cost of implementation, the net value for this SAMA is -$545,916 ($54,084 - $600,000 = -$545,916).

SAMA 1: Automate EFIC/InverterBackup Coolinq AI~gnment During normal operation, the control complex Chilled Water (CH) system is in operation with one pump and one chiller in operation. The control complex CH is normally aligned to provide cooling water to such heat loads as the Control Complex ventilation system cooling coils, Reactor Building penetration HVAC cooling coils, and the, EFIC room HVAC cooling coils (AHHE-43, 44).

Another chilled water system, the "Appendix R" Chilled Water subsystem, consists of a chiller, a chilled water pump, a chilled water surge tank, and the isolation and control valves required for system operation. Chilled water is pumped from the aligned heat loads through the chiller, and back to the aligned heat loads. During normal operation, the Appendix R CH subsystem supplies Turbine Building switchgear room air handling unit cooling coils (AHHE-IOA, lOB), but can also be manually aligned to cool the EFIC room HVAC cooling coil (AHHE-44). Certain multiple failures of the control complex CH system are assumed to lead to failure of the EFIC system, which eventually results in loss of all EFW An available option to cool critical equipment important to the operation of EFW is the use of this Appendix R CH system.

The purpose of this SAMA is to investigate whether improving the capability of the Appendix R Chilled Water subsystem by assuming automatic realignment will provide a cost beneficial option.

Assumptions It was assumed that a reduction in the failure probability of a single human error probability (HEP) event would be a satisfactory simulation of the addition of automatic controls and realignment capability for the Appendix R Chilled Water subsystem. This implies that any unavailability representing mechanical, electrical, and other support systems necessary for this SAMA could be adequately estimated via this "lumped parameter"approach in determining the possible risk benefits.

PRA Model Changes to Model SAMA To calculate the consequences of implementation of this SAMA, the failure probabilityof HEP event JHUCHP2R in the PRA model was reduced from 1.0 to IE-4. Also, to prevent the use of JHUCHP2Z, which is a recovery event for this human action in the quantification recovery rules file, the appropriatecommand lines in the recovery rules

U. S. Nuclear Regulatory Commission Enclosure 3F1009-03 Page 35 of 74 were delineated as comments so as to prevent appending this non-recovery event with a probability0. 05 to any newly generated cutsets for this SAMA quantification.

Results of SAMA Quantification Implementation of this SAMA yielded a small reduction in CDF, Dose-Risk, and Offsite Economic Cost-Risk (OECR). The results are summarized in the following table for CR-3:

CDF Dose-Risk OECR Base Value 4.95E-06 3.79 $6,624 SAMA Value 4.33E-06 3.57 $6,218 PercentChange 12.4% 5.8% 6.1%

A further breakdown of the Dose-Risk and OECR information is provided in the table below according to release category:

Release ICI RC1 RCIA RCIB RCIAB RC2 RC2B RC3 RC3B RC4C RC5C Total Category FrequencyBASE 4. 10E-06 2.44E-08 4.71E-10 1.59E-08 1.25E-09 8.43E-10 3.46E-09 2.46E-07 1.57E-07 3.44E-07 5.15E-08 4.95E-06 FrequencysAMA 3.70E-06 2.43E-08 4.50E-10, 1.41E-08 1.04E-09 6.48E-10 3.25E-09 1.30E-07 6.56E-08 3.45E-07 5.15E-08 4.33E-06 Dose-RiSkBASE 0.04 0.01 0.00 0.03 0.00 0.00 0.01 0.07 0.30 2.57 0.74 3.79 Dose-RisksAMA 0.04 0.00 0.00 0.03 0.00 0.00 0.01 0.04 0.13 2.57 0.74 3.57 OECRSASE $0 $1 $0 $29 $2 $7 $27 $15 $679 $4,855 $1,009 $6,624 OECRsAMA $0 $1 $0 $26 $2 $6 $25 $8 $283 $4,858 $1,010 $6,218 This information was used as input to the cost-benefit calculation. The results of this calculation are provided in the following table:

SAMA 1 Net Value Base Case Revised Averted CO1 Net Value Cost-Risk Cost-Risk Cost-Risk

$4,092,000 $3,750,708 $341,292 $1,000,000 -$658,708 The SAMA 1 results indicate a small reduction in CDF, dose-risk, and offsite economic consequences. Based on a cost of implementation of $1,000,000, the net value for this SAMA is -$658,708 ($341,292 - $1,000, 000 = -$658,708).

SAMA 7: Install a Separate and Independent AFW Source and Pump as a Backup to the EFW System This SAMA investigates the implementation of a separate and independent train of auxiliary feedwater (AFW). Given that EFW pumps EFP-1, -2, and -3 of the EFW system and FWP-7 of the existing AFW system are all unavailable, this SAMA would provide another means of providing feedwater to both steam generators during

U. S. Nuclear Regulatory Commission Enclosure 3F1009-03 Page 36 of 74 emergency conditions. This backup AFW system is postulated to have an independent suction source capable of satisfying long-term cooling requirements and rely on support systems that are totally independent of existing plant systems, such that common cause failure mechanisms with other plant equipment would be practicallynon-existent.

Assumptions It was assumed that the unavailability of this alternate AFW backup system could be representedby a single basic event. That is, any unavailabilityrepresentingmechanical, electrical, and other support systems necessary for this SAMA could be adequately estimated via this "lumped parameter" approach in determining the possible risk benefits.

PRA Model Chances to Model SAMA To calculate the consequences of implementation of this SAMA, a new event (SAMA7-AFW-BU) was added to AND gates @BS01 (LOSS OF ALL PRIMARY-SECONDARY COOLING) and @B03 (FAILURE OF EFW / AFW) that represents the overall unavailability of this postulated independent AFW backup system. An unavailabilityof 1E-03 was chosen to conservatively bound the risk benefit that might be achieved given the implementation of such a system.

SAMA 7 New Basic Event Basic Event Description Probability Comments SAMA7-AFW-BU FAILURE OF ALTERNATE AFW IE-03 Assumed unavailabilityto account for all BACKUP SYSTEM (SAMA 7) possible failure mechanisms, including operatoractions, for this backup AFW system.

Results of SAMA Quantification Implementation of this SAMA yielded a moderate reduction in CDF, but only a relatively small change in the Dose-Risk, and Offsite Economic Cost-Risk (OECR). The results are summarized in the following table for CR-3:

CDF Dose-Risk OECR Base Value 4.95E-06 3.79 $6,624 SAMA Value 3.05E-06 3.36 $5,885 PercentChange 38.4% 11.3% 11.2%

A further breakdown of the Dose-Risk and OECR information is provided in the table below according to release category:

U. S. Nuclear Regulatory Commission Enclosure 3F1009-03 Page 37 of 74 Release IC1 RC1 RCIA RCIB RCIAB RC2 RC2B RC3 RC3B RC4C RC5C Total Category FrequencyBAsE 4. 1OE-06 2.44E-08 4.71E-10 1.59E-08 1.25E-09 8.43E-10 3.46E-09 2.46E-07 1.57E-07 3.44E-07 5.15E-08 4.95E-06 FrequencysAMA 2.61E-06 2.15E-08 2.73E-10 5.33E-09 1.04E-10 2.42E-10 2.56E-09 6.88E-09 7.26E- I1 3.43E-07 5.15E-08 3.05E-06 Dose-RiSkBASE 0.04 0.01 0.00 0.03 0.00 0.00 0.01 0.07 0.30 2.57 0.74 3.79 Dose-RisksAMA 0.03 0.00 0.00 0.01 0.00 0.00 0.01 0.00 0.00 2.57 0.74 3.36 OECRBAse $0 $1 $0 $29 $2 $7 $27 $15 $679 $4,855 $1,009 $6,624 OECRsAMA $0 $0 $0 $10 $0 $2 $20 $0 $0 $4,842 $1,010 $5,885 This information was used as input to the cost-benefit calculation. The results of this calculation are provided in the following table:

SAMA 7 Net Value Base Case Revised Averted COI Net Value Cost-Risk Cost-Risk Cost-Risk

$4,092,000 $3,220,956 $871,044 $5,000,000 -$4,128,956 The SAMA 7 results indicate a moderate reduction in CDF with relatively small changes in dose-risk and offsite economic consequences. Based on a cost of implementation of

$5,000,000, the net value for this SAMA is -$4,128,956 ($871,044 - $5,000,000

-$4,128,956).

SAMA 16: Implement Desiqn and Procedural Changes for Crosstying Trains of the Decay Heat and Decay Heat Closed Cooling Systems The function of the Decay Heat (DH) system is to provide the ability to inject borated water into the reactorcoolant system given a LOCA condition. It also provides the ability to recirculatecoolant that accumulates in the reactorbuilding sump following depletion of the Borated Water Storage Tank (BWST). In sump recirculationmode, at depressurized conditions, the DH pumps can inject water directly back into the reactorcoolant system, or at higher pressures, to the suction of the Makeup (MU) pumps for high pressure recirculation. Reactor decay heat is removed via the DH coolers DHHE-1A and DHHE-lB. These heat exchangers are in turn cooled by the Decay Heat Closed Cycle Cooling (DHCCC) system via heat exchangers DCHE-1A and DCHE-1B.

Based on the following basic events obtained from a review of the Level 1 CDF importance list, this SAMA attempts to simulate the ability to cross-connect the DH and DHCCC systems to improve the continued availabilityof these systems. It is important to note that even though some of the listed events exhibited a RRW of < 1.02, they were nonetheless included for the purpose of showing train symmetry.

U. S. Nuclear Regulatory Commission Enclosure 3F1009-03 Page 38 of 74 Basis Event RRW Description LMMDV42F 1.031 FAILURE OF TRAIN A RECIRC VALVE DHV-42 LMMDV43F 1.033 FAILURE OF TRAIN B RECIRC VALVE DHV-43 LMMDV11F 1.031 TRAIN A RECIRC VALVE DHV-1 1 FAILS LMMDV12F 1.033 TRAIN B RECIRC VALVE DHV-12 FAILS SMMDHCCA 1.018 DHCCC TRAIN A FA UL TS SMMDHCCB 1.025 DHCCC TRAIN B FAULTS SPMDHCAM 1.013 DHCCC TRAIN A IN MAINTENANCE SPMDHCBM 1.025 DHCCC TRAIN B IN MAINTENANCE Assumptions In order to attempt to maximize the perceived benefit of being able to crosstie the DH and DHCCC systems, it was assumed that the risk benefit may be bounded by logically combining existing events under new logical AND gates. That is, no considerationwas given to any new operatoror manual actions that might tend to deflate the risk benefit.

PRA Model Changes to Model SAMA To calculate the consequences of implementation of this SAMA, the following logic changes were made to the PRA model:

" A new AND gate labeled SAMA16-LO01 using existing modular events LMMDV42F and LMMDV43F as inputs was created. This AND gate was then used to replace these events where they originallyexisted in the PRA model.

" A new AND gate labeled SAMA16-L002 using existing modular events LMMDV11F and LMMDV12F as inputs was created. This AND gate was then used to replace these events where they originallyexisted in the PRA model.

A new AND gate labeled SAMA16-SOO1 using existing modular events SMMDHCCA and SMMDHCCB as inputs was created. This AND gate was then used to replace these events where they originallyexisted in the PRA model.

To simulate the ability of being able to cross-connect DHCCC trains when one is in maintenance, each maintenance event (SPMDHCAM for train A and SPMDHCBM for train B) was reduced by a factor of 10 since it would be expected that some, but not all, maintenance activities on the DHCCC system would afford the opportunity of using the opposite train's equipment.

Results of SAMA Quantification Implementation of this SAMA yielded a small reduction in CDF, Dose-Risk, and Offsite Economic Cost-Risk (OECR). The results are summarized in the following table for CR-3:

CDF Dose-Risk OECR Base Value 4.95E-06 3.79 $6,624 SAMA Value 4.14E-06 3.76 $6,603 Percent Change 16.4% 0.7% 0.3%

U. S. Nuclear Regulatory Commission Enclosure 3F1009-03 Page 39 of 74 A further breakdown of the Dose-Risk and OECR information is provided in the table below according to release category:

Release ICi RC1 RCIA RCIB RCIAB RC2 RC2B RC3 RC3B RC4C RC5C Total Category FrequencyBAsE 4. 10E-06 2.44E-08 4.71E-10 1.59E-08 1.25E-09 8.43E-10 3.46E-09 2.46E-07 1.57E-07 3.44E-07 5.15E-08 4.95E-06 FrequencysAMA 3.32E-06 1.81E-08 3.79E-10 1.47E-08 1.18E-09 8.16E-10 2.57E-09 2.25E-07 1.57E-07 3.44E-07 5.13E-08 4.14E-06 Dose-RiskBASE 0.04 0.01 0.00 0.03 0.00 0.00 0.01 0.07 0.30 2.57 0.74 3.79 Dose-RiSksAMA 0.03 0.00 0.00 0.03 0.00 0.00 0.01 0.07 0.30 2.57 0.74 3.76 OECRBASE $0 $1 $0 $29 $2 $7 $27 $15 $679 $4,855 $1,009 $6,624 OECRsAMA $0 $0 $0 $27 $2 $7 $20 $14 $677 $4,850 $1,006 $6,603 This information was used as input to the cost-benefit calculation. The results of this calculation are provided in the following table:

SAMA 16 Net Value Base Case Revised Averted COl Net Value Cost-Risk Cost-Risk Cost-Risk

$4,092,000 $3,829,788 $262,212 $5,000,000 -$4,737,788 The SAMA 16 results indicate a small reduction in CDF, dose-risk and offsite economic consequences. Based on a cost of implementation is $5, 000,000, the net value for this SAMA is -$4, 737,788 ($262,212 - $5,000,000 = -$4, 737,788).

SAMA 18: Install an Additional Emerqency Diesel Generator to Provide an Additional A C Power Source This SAMA investigates the proposed installation of an additional emergency diesel generator (EDG) to help reduce the contribution to core damage due to scenarios involving the loss of backup AC power. Implementation of this SAMA will be modeled in the PRA as a new event with an overall unavailability that is meant to capture all mechanical, electrical, and operatorfailures, while also creating a bounding estimate of the perceived risk benefit. This additional EDG is assumed to be independent of other plant support systems, so as to minimize any common cause failure mechanisms, e.g.,

use of an air-cooled diesel engine instead of one requiringa heat exchanger cooled by the plant's component cooling water system.

Assumptions

1. It was assumed that the unavailabilityof this backup EDG could be represented by a single basic event. That is, any unavailability representing mechanical, electrical, and other support systems necessary for this SAMA could be adequately estimated via this "lumped parameter"approach in determining the possible risk benefits.

2. As a means of simplifying the revised model logic, while at the same time providing a bounding estimate of the risk benefit for this SAMA analysis, it was

U. S. Nuclear Regulatory Commission Enclosure 3F1009-03 Page 40 of 74 assumed that this backup EDG could hypothetically be used to supply both 4,160 VAC Engineered Safeguards buses simultaneously.

PRA Model Changes to Model SAMA To calculate the consequences of implementation of this SAMA, the following PRA model logic was utilized.;

A new AND gate for train A was created (SAMA18 001) with inputs that include existing OR gate A401 and the SAMA event SAMA18-EDG-BU, which representsthe unavailabilityfor this backup EDG.

A new AND gate for train B was created (SAMA18 002) with inputs that include existing OR gate A451 and event SAMA18-EDG-BU.

Gate SAMA18_001 is used as an input to existing gates A002 (4160V ES BUS 3A SUPPLY FAULTS) and SDR072 (LOSP occurs with no power from DG 3A).

Gate SAMA18_002 is used as an input to existing gates A052 (4160V ES BUS 3B SUPPLY FAULTS) and SDR074 (LOSP occurs with no power from DG 3B).

An unavailability of 1E-03 was chosen for event SAMA18-EDG-BU to conservatively bound the risk benefit that might be achieved given the installation of a new backup EDG.

SAMA 18 New Basic Event Basic Event Description Probability Comments SAMA18-EDG-BU FAILURE OF BACKUP EDG (SAMA 18) 1E-03 Assumed unavailabilityto account for all possible failure mechanisms, including operatoractions, for this backup AC power source.

Results of SAMA Quantification Implementation of this SAMA yielded a small reduction in CDF, Dose-Risk, and Offsite Economic Cost-Risk (OECR). The results are summarized in the following table for CR-3:

CDF Dose-Risk OECR Base Value 4.95E-06 3.79 $6,624 SAMA Value 4.59E-06 3.62 $6,313 PercentChange 7.2% 4.5% 4.7%

A further breakdown of the Dose-Risk and OECR information is provided in the table below accordingto release category:

U. S. Nuclear Regulatory Commission Enclosure 3F1009-03 Page 41 of 74 Release IC1 RCI RCIA RCIB RCIAB RC2 RC2B RC3 RC3B RC4C RC5C Total Category FrequencyBAsE 4. 1OE-06 2.44E-08 4.71E-10 1.59E-08 1.25E-09 8.43E-10 3.46E-09 2.46E-07 1.57E-07 3.44E-07 5.15E-08 4.95E-06 FrequencysAMA 3.89E-06 2.43E-08 4.54E-10 1.05E-08 6.50E-10 6.92E-10 3.35E-09 1.73E-07 8.95E-08 3.44E-07 5.14E-08 4.59E-06 Dose-RiskBASE 0.04 0.01 0.00 0.03 0.00 0.00 0.01 0.07 0.30 2.57 0.74 3.79 Dose-RisksAMA 0.04 0.00 0.00 0.02 0.00 0.00 0.01 0.05 0.17 2.57 0.74 3.62 OECRBASE $0 $1 $0 $29 $2 $7 $27 $15 $679 $4,855 $1,009 $6,624 OECRsAMA $0 $1 $0 $19 $1 $6 $26 $11 $387 $4,855 $1,008 $6,313 This information was used as input to the cost-benefit calculation. The results of this calculation are provided in the following table:

SAMA 18 Net Value Base Case Revised Averted COl Net Value Cost-Risk Cost-Risk Cost-Risk

$4,092,000 $3,864,948 $227,052 $8,000,000 -$7,772,948 The SAMA 18 results indicate a small reduction in CDF, dose-risk and offsite economic consequences. Based on a cost of implementation is $8,000,000, the net value for this SAMA is -$7, 772,948 ($227,052 - $8,000,000 = -$7, 772,948).

SAMA 13: Add an Additional Train of Decay Heat Removal Coolinq of Diverse Desiqn The decay heat removal (DHR or DH) system provides normal heat removal operation following plant cool down and provides low pressure makeup following a loss of coolant accident (LOCA) in both the low pressure injection (LPI) and both low and high pressure recirculation(LPR, HPR) modes -ofoperation.

The DH system consists of two pumps, two heat exchangers, and a borated water storage tank (BWST). Interconnecting piping, motor-operated control and isolation valves are required for normal and emergency system operation.

The DH system is comprised of two independent and redundant cooling trains. Each train is capable of providing 100% of the heat removal requirementsfor a normal reactor shutdown, as well as for LOCA emergency cooling. Each train may take suction from the BWST, the reactorbuilding sump, or the Reactor Coolant System (RCS) B hot leg via the decay heat drop line. Each DH heat exchanger is cooled by its own decay heat closed cycle cooling (DHCCC)system train.

This SAMA investigates the proposed installationof an additionaltrain of DHR of diverse design to increase reliability and availability of the DHR function, and also to minimize any common cause related failure mechanisms. Implementation of this SAMA will be modeled in the PRA as a new event with an overall unavailability that is meant to capture all mechanical, electrical, and operator failures, while also creating a bounding estimate of the perceived risk benefit. This additional train of DHR is assumed to be independent of other plant support systems, e.g., use of a diesel-driven air-cooledpump and a DHR heat exchanger with a different means of heat removal, such as via radiative

U. S. Nuclear Regulatory Commission Enclosure 3F1009-03 Page 42 of 74 cooling instead of the conventional convection method. Multiple available suction sources for primary inventory control would also be part of this SAMA design so as to increase the success of a borated water source, ratherthan solely relying on the single BWST that currently exists.

Assumptions It was assumed that the unavailabilityof this backup DHR train could be represented by a single basic event. That is, any unavailability representing mechanical, electrical, and other support systems necessary for this SAMA could be adequately estimated via this "lumpedparameter"approach in determining the possible risk benefits.

PRA Model Changes to Model SAMA To calculate the consequences of implementation of this SAMA; the following PRA model logic was utilized:

The new backup DHR SAMA event (SAMA13-DHR-BU) was used as a new input to the following AND gates:

o H3630 - LOSS OF SUCTION FLOW TO MUP-1A o H3730 - LOSS OF SUCTION FLOW TO MUP-1B o L102 - LOSS OF FLOWFROM BOTH LPI TRAINS

The following new gate logic was added to the PRA model:

o A new AND gate was created (SAMA 13001) with existing OR gate LO01 and new SAMA event SAMA13-DHR-BU as inputs. Gate SAMA13_001 is used as an input to OR gate @AU1.

o A new AND gate was created (SAMA 13002) with existing OR gate L201 and new SAMA event SAMA13-DHR-BU as inputs. Gate SAMA13_002 is used as an input to OR gate @GR02.

An unavailability of 1E-03 was chosen for event SAMA13-DHR-BU to conservatively bound the risk benefit that might be achieved given the installation of a new diverse train of DHR.

SAMA 13 New Basic Event Basic Event Description Probability Comments SAMA13-EDG-BU FAILURE OF BACKUP DHR (SAMA 13) IE-03 Assumed unavailabilityto account for all possible failure mechanisms, including operatoractions, for this backup AC power source.

Results of SAMA Quantification Implementation of this SAMA yielded a moderate reduction in CDF, but only a relatively small change in the Dose-Risk, and Offsite Economic Cost-Risk (OECR). The results are summarized in the following table for CR-3:

U. S. Nuclear Regulatory Commission Enclosure 3FY1009-03 Page 43 of 74 CDF Dose-Risk OECR Base Value 4.95E-06 3.79 $6,624 SAMA Value 3.36E-06 3.61 $6,342 Percent Change 32.1% 4.8% 4.3%

A further breakdown of the Dose-Risk and OECR information is provided in the table below according to release category:

Release ICI RCI RCIA RCIB RCIAB RC2 RC2B RC3 RC3B RC4C RC5C Total Category FrequencyBASE 4. IOE-06 2.44E-08 4.71E-10 1.59E-08 1.25E-09 8.43E-10 3.46E-09 2.46E-07 1.57E-07 3.44E-07 5.15E-08 4.95E-06 FrequencysAMA 2.60E-06 1.09E-08 2.88E-10 1.42E-08 1.19E-09 8.43E-10 1.59E-09 1.99E-07 1.57E-07 3.26E-07 5.14E-08 3.36E-06 Dose-RisksAsE 0.04 0.01 0.00 0.03 0.00 0.00 0.01 0.07 0.30 2.57 0.74 3.79 Dose-RisksAMA 0.03 0.00 0.00 0.03 0.00 0.00 0.01 0.06 0.30 2.43 0.74 3.61 OECRBASE $0 $1 $0 $29 $2 $7 $27 $15 $679 $4,855 $1,009 $6,624 OECRsAMA $0 $0 $0 $26 $2 $7 $12 $12 $678 $4,595 $1,008 $6,342 This information was used as input to the cost-benefit calculation. The results of this calculation are provided in the following table:

SAMA 13 Net Value Base Case Revised Averted COl Net Value Cost-Risk Cost-Risk Cost-Risk

$4,092,000 $3,487,992 $604,008 $10,000,000 -$9,395,992 The SAMA 13 results indicate a moderate reduction in CDF with relatively small dose-risk and offsite economic consequences. Based on a cost of implementation of

$10,000,000, the net value for this SAMA is -$9,395,992 ($604,008 - $10,000,000 =

-$9,395,992).

SAMA 52: Install a ParallelFlow Path for the Decay Heat Removal Drop Line During power operations, the Decay Heat (DH) system is normally in standby and aligned for automatic actuation in the Low Pressure Injection (LPI) mode of operation.

This configuration allows the system to automatically align to deliver water from the BWST to the reactorvessel through the core flood nozzle penetrations. During a normal plant cooldown, the DH system is placed in operation after reactorcoolant temperature has been reduced to 280°F. The system takes suction from the decay heat drop line and delivers it through the DH heat exchangers and back through the Core Flood (CF) nozzle penetrations.

This SAMA investigates the proposed installation of an additional parallel flow path similar to the existing DH drop line. Implementation of this SAMA will be modeled in the PRA as a new event with an overall unavailability that is meant to capture all

U. S. Nuclear Regulatory Commission Enclosure 3F1009-03 Page 44 of 74 mechanical, electrical, and operatorfailures, while also creating a bounding estimate of the perceived risk benefit.

Assumptions

1. It was assumed that the unavailability of this parallel flow path could be represented by a single basic event. That is, any unavailability representing mechanical, electrical, and other support systems necessary for this SAMA could be adequately estimated via this "lumped parameter"approach in determining the possible risk benefits.

2. Since the additional flow path can also act as an additional ISLOCA contributor, to account for a possible increase in offsite risk, the ISLOCA initiating frequency was increasedby a factor of 1.2 (20% increase). Therefore, the initiating event frequency for IE_V (ISLOCA - DHR DROP LINE AND INJECTION LINES) was redefined as (1.2

5.14E-08) = 6. 17E-08.

PRA Model Changes to Model SAMA To calculate the consequences of implementation of this SAMA, the following PRA model logic was utilized:

A new AND gate was created (SAMA52_001) with inputs that include existing OR gate L207 and the SAMA event SAMA52-DROP-BU, which represents the unavailabilityfor this backup drop line.

" OR gate L207 was originally used as an input to gates A1604, A1607, and L201, but AND gate SAMA52_001 is now used in place of OR gate L207 as an input to these gates.

An unavailability of IE-03 was chosen for event SAMA52-DROP-BU to conservatively bound the risk benefit that might be achieved given the installation of a new parallel drop line.

SAMA 52 New Basic Event Basic Event Description Probability Comments SAMA52-DROP-BU FAILURE OF BACKUP IE-03 Assumed unavailabilityto account for all DROPLINE (SAMA 52) possible failure mechanisms, including operatoractions, for this backup DH drop line.

Results of SAMA Quantification Implementation of this SAMA yielded a very small reduction in CDF, a slight increase in dose risk, and a very small decrease in Offsite Economic Cost-Risk (OECR). The results are summarized in the following table for CR-3:

U. S. Nuclear Regulatory Commission Enclosure 3F1009-03 Page 45 of 74 CDF Dose-Risk OECR Base Value 4.95E-06 3.79 $6,624 SAMA Value 4.94E-06 3.82 $6,606 Percent Change 0.2% -0. 8% (increase) 0.6%

A further breakdown of the Dose-Risk and OECR information is provided in the table below according to release category:

Release IC1 RC1 RCIA RCIB RCIAB RC2 RC2B RC3 RC3B RC4C RC5C Total Category FrequencyBASE 4. 10E-06 2.44E-08 4.71E-10 1.59E-08 1.25E-09 8.43E-10 3.46E-09 2.46E-07 1.57E-07 3.44E-07 5.15E-08 4.95E-06 FrequencysAMA 4. IOE-06 2.44E-08 4.70E-10 1.59E-08 1.25E-09 8.43E-10 3.46E-09 2.45E-07 1.57E-07 3.29E-07 6.18E-08 4.94E-06 Dose-RiskBASE 0.04 0.01 0.00 0.03 0.00 0.00 0.01 0.07 0.30 2.57 0.74 3.79 Dose-RisksAMA 0.04 0.01 0.00 0.03 0.00 0.00 0.01 0.07 0.30 2.46 0.89 3.82 OECRBASE $0 $1 $0 $29 $2 $7 $27 $15 $679 $4,855 $1,009 $6,624 OECRsAMA $0 $1 $0 $29 $2 $7 $27 $15 $677 $4,637 $1,210 $6,606 This information was used as input to the cost-benefit calculation. The results of this calculationare provided in the following table:

SAMA 52 Net Value Base Case Revised Averted COl Net Value Cost-Risk Cost-Risk Cost-Risk

$4,092,000 $4,097,508 -$5,508 $5,000,000 -$5,005,508 The SAMA 52 results indicate a very small reduction in CDF and Offsite Economic consequences with a slight increase in dose risk. For this SAMA, the risk reductions based on the availability of the alternate DHR drop line were counterbalancedby the increase in risk related to the introduction of an additional high pressure/low pressure interface. The cost of implementation has been estimated to be $5,000,000, which results in a net value of -5,005,508 (-$5,508 - $5, 000,000 = -$5,005,508).

Results Summary The results of the Phase II quantificationsusing the external events multiplier of 12 and the point estimate PRA results are summarized in the table below.

U. S. Nuclear Regulatory Commission Enclosure 3F1009-03 Page 46 of 74 Phase IIResults Summary (EE Multiplier of 12, Point Estimate PRA results)

Averted Averted Net Value SAMA Cost of Cost Risk NEt V Cost Risk ID Implementation (EE x 2, (EE x 2, PE (EE x 12, PE (EE x 12, PE PEPRA) PRA) PRA) PRA) 34 $50,000 $94,706 $44,706 $568,236 $518,236 33 $50,000 $15,384 -$34,616 $92,304 $42,304 9 $50,000 $16,128 -$33,872 $96,768 $46,768 10 $50,000 $29,502 -$20,498 $177,012 $127,012 38 $50,000 $11,998 -$38,002 $71,988 $21,988

$350,000 $23,490 -$326,510 $140,940 -$209,060 6 $400,000 $19,782 -$380,218 $118,692 -$281,308 5 $500,000 $39,040 -$460,960 $234,240 -$265,760 17 $500,000 $27,332 -$472,668 $163,992 -$336,008 11 $250,000 $8,894 -$241,106 $53,364 -$196,636 15 $300,000 $24,684 -$275,316 $148,104 -$151,896 4 $250,000 $30,058 -$219,942 $180,348 -$69,652 35 $700,000 $259,090 -$440,910 $1,554,540 $854,540 51 $100,000 $76,776 -$23,224 $460,656 $360,656 49 $150,000 $89,132 -$60,868 $1,026,444 $876,444 8 $500,000 NA NA $357,972 -$142,028 26 $2,000,000 NA NA $555,600 -$1,444,400 14 $900,000 NA NA $54,528 -$845,472 37 $600,000 NA NA $54,084 -$545,916 1 $1,000,000 NA NA $341,292 -$658,708 7 $5,000,000 NA NA $871,044 -$4,128,956 16 $5,000,000 NA NA $262,212 -$4,737,788 18 $8,000,000 NA NA $227,052 -$7,772,948 13 $10,000,000 NA NA $604,008 -$9,395,992 52 $5,000,000 NA NA -$5,508 -$5,005,508 Use of the external events multiplier of 12 in place of the multiplier of 2 results in the reclassificationof 7 of the original Phase II SAMAs as cost beneficial (SAMAs 33, 9, 10, 38, 35, 51, and 49). Of the 10 SAMAs that were retainedfor Phase II analysis based on the transition to the external events multiplierof 12, none were cost beneficial.

In order to assess the impact on the "final" SAMA screenings, the use of the 9 5th percentile PRA results must be considered. As documented in Section E. 7.2 of the ER, the averted point estimate based cost-risk values can be multiplied by 2.18 to account for the impact of the use of the 9 5th percentile PRA results. The updated net values are summarized in the following table:

U. S. Nuclear Regulatory Commission Enclosure 3F11009-03 Page 47 of 74 Phase IIResults Summary (EE Multiplier of 12, 95th Percentile PRA results)

Averted Averted Net Value SAMA Cost of Cost Risk Net Value Cost Risk (EE x 12, ID Implementation (EE x 12, PE (EE x 12, 95th 95th PE PRA) Percentile) Percentile) 34 $50,000 $568,236 $518,236 $1,238,754 $1,188,754 33 $50,000 $92,304 $42,304 $201,223 $151,223 9 $50,000 $96,768 $46,768 $210,954 $160,954 10 $50,000 $177,012 $127,012 $385,886 $335,886 38 $50,000 $71,988 $21,988 $156,934 $106,934 3 $350,000 $140,940 -$209,060 $307,249 -$42,751 6 $400,000 $118,692 -$281,308 $258,749 -$141,251 5 $500,000 $234,240 -$265,760 $510,643 $10,643 17 $500,000 $163,992 -$336,008 $357,503 -$142,497 11 $250,000 $53,364 -$196,636 $116,334 -$133,666 15 $300,000 $148,104 -$151,896 $322,867 $22,867 4 $250,000 $180,348 -$69,652 $393,159 $143,159 35 $700,000 $1,554,540 $854,540 $3,388,897 $2,688,897 51 $100,000 $460,656 $360,656 $1,004,230 $904,230 49 $150,000 $1,026,444 $876,444 $2,237,648 $2,087,648 8 $500,000 $357,972 -$142,028 $780,379 $280,379 26 $2,000,000 $555,600 -$1,444,400 $1,211,208 -$788,792 14 $900,000 $54,528 -$845,472 $118,871 -$781,129 37 $600,000 $54,084 -$545,916 $117,903 -$482,097 1 $1,000,000 $341,292 -$658,708 $744,017 -$255,983 7 $5,000,000 $871,044 -$4,128,956 $1,898,876 -$3,101,124 16 $5,000,000 $262,212 -$4,737,788 $571,622 -$4,428,378 18 $8,000,000 $227,052 -$7,772,948 $494,973 -$7,505,027 13 $10,000,000 $604,008 -$9,395,992 $1,316,737 -$8,683,263 52 $5,000,000 .-$5,508 -$5,005,508 -$12,007 -$5,012,007 When the external events multiplier of 12 is used in conjunction with the 9 5 th percentile PRA results, 4 additional SAMAs have positive net values (5, 15, 4, and 8). The total number of SAMAs that could be classified as cost, beneficial when the external events multiplier of 12 is used in conjunction with the 9 5 th percentile PRA results is 12 (SAMAs 34, 33, 9, 10, 38, 5, 15, 4, 35, 51, 49, and 8). The use of the 9 5h percentile PRA results is not, however, considered to provide the most realistic assessment of the cost effectiveness of a SAMA.

U. S. Nuclear Regulatory Commission Enclosure 3F1009-03 Page 48 of 74 3.e Status of three USI A-46 topics:

1) Safety Evaluation Report for Unresolved Safety Issue (USI) A-46 Program Implementation at Crystal River Unit 3 (TAC No. M69440), identified two USI A-46 outliers, MTSW-3A, 480V Turbine Auxiliary Bus A, and MTSW-3C, 480V Reactor Auxiliary Bus A. Resolution of these two outliers had been deferred, and they were included in the CR-3 Corrective Action Program. All field work associated with this equipment has been completed, and this action was closed in October 2001.

2) The five outliers associated with differences between the caveats in generic implementation procedure (GIP)-2 and those in the plant-specific procedure (PSP) were resolved and the corrective action program action was closed in 2001.

3) Abnormal procedure (AP)-961 has been revised and this action was closed in October 2000.

4. Provide the following information concerning the MACCS2 analyses:

a. Section 2.12.1 of the Environmental Report states that "Progress Energy plans to increase CR-3's licensed power level and electrical output by approximately 20 percent in an Extended Power Uprate (EPU)scheduled to be carriedout during fall 2009 and fall 2011 refueling outages." Operation at this higher power level could impact the results of the SAMA evaluation due to the higher fission product inventory and replacement power costs associated with the EPU. Provide a revised SAMA analysis (baseline and uncertainty) assuming operation at the uprated power level.

b. Section E.3.2 states that county growth rates were applied to the year 2000 population to develop the SECPOP2000 population sector distribution.

i. Section E.3.2 does not discuss transient population. Clarify whether transient population was considered in the analysis.. If a transient population was not considered, either provide a justification/rationale for not including this or estimate the potential impact on the population dose risk and the SAMA evaluation.

ii. Provide the year 1990 Emergency Planning Zone (EPZ) population used for the evacuation study.

c. The MACCS2 analysis yielded a total population dose risk (PDR) and off-site economic cost risk (OECR) of 3.98 person-rem/year and 6,950 $/year, respectively, as reported in Table E.3-7. However, per Section E.4.6, the Phase I and II SAMA evaluations utilized a PDR and OECR of 3.79 person-rem/year and 6,624 $/year, respectively. Clarify the discrepancy and, if necessary, provide a revised SAMA evaluation.

Response

4. a The PRA model for CR-3 uprate is still in development and as a result information on the impact of the SAMA analysis due to the power uprate is not available.

U. S. Nuclear Regulatory Commission Enclosure 3F1009-03 Page 49 of 74 4.b.i The transientpopulation was not included in the analysis because its effects on the dose and cost risk are inconsequential. This was demonstrated by performing a conservative Level-3 calculation that included transients. Transients were identified at two State Parks within the EPZ, Crystal River Preserve State Park and Crystal River Archeological State Park. Crystal River Preserve State Park is immediately south of the CR-3 site/Crystal River Energy Complex, and lies almost entirely within the EPZ. Crystal River Archeological State Park is a small property completely surroundedby the Crystal River Preserve State Park and is accessed using the same roads. It is completely within the EPZ. The monthly numbers of visitors to the parks for July 2007 to June 2009 were obtained from the Florida DEP Recreation and Parks Management. There were 1708 visitors per day during March 2009, the maximum visitation month during those years.

Those 1708 visitors were adjusted to an equivalent year-2036 transient population of 3200 using the same growth rate as the residentialpopulation near the parks. It is noted that many of the park visitors would be EPZ residents and are thus double counted in this analysis. Although no other transient population of similar size or larger was identified within the EPZ, the residentialpopulation within this zone was increased by 10% to bound the presence of miscellaneous transients. The base case evacuation speed was reduced by 19.4% to reflect the like increase in EPZ population resulting from the assumed transients; this assumes that all evacuation routes would be saturated.

These conservative assumptions result in an increase in the base case dose-risk and cost-risk of 2.0% and O.7%. Considering only the documented maximum monthly park visitations, the dose- and cost-risks increase by 1. 1% and 0. 1% from the base case.

These small increasesare not of consequence in the SAMA evaluation.

4.b.li The 1990 Emergency PlanningZone (EPZ) population is 15,065.

4. c The apparent discrepancy may best be explained using the following information that was provided in Section E.2.2.3 of the SAMA report:

Although the same PRA model was used as the model-of-record (MOR2006) for quantification of the proposed Phase 2 SAMAs, the reportedbase value for CDF (4.95E-6) was slightly different due to the SAMA quantificationsbeing performed at a higher truncation limit of 1E-11 for a more efficient evaluation of multiple PRA model changes. The model-of-record result for CDF (4.99E-6) was performed at a truncation of 1E-12, which would tend to yield a slightly higher value for CDF. Additionally, two different yet valid methods of quantificationwere used. The model-of-record results were produced using EOOS software and the SAMA quantifications were performed using PRAQUANT software. In using PRAQUANT, each of the core damage accident sequences were individually quantified to retain plant damage states in order to account for all Level 2 release categories. At any event, the important aspect to note is that all SAMA calculations made use of the same method of quantification so that the relative cost difference between proposed SAMAs and the base MMACR value were kept consistent to give an appropriaterelative basis for comparison.

U. S. Nuclear Regulatory Commission Enclosure 3F1009-03 Page 50 of 74

5. Provide the following with regard to the SAMA identification and screening process:

a. Section E.5.1.7 identifies the 4.16 kV Switchgear Bus Room 3A, along with Battery Charger Room 3A, as being significant contributors to the fire CDF based on the IPEEE (i.e., 17% and 36%, respectively). IPEEE Section 1.4 also identifies 4.16 kV Switchgear Bus Room 3B as having a CDF similar to Room 3A. The uncertainty analysis for SAMA 49, Upgrade Fire Barriers in Battery Charger Room 3A, shows this SAMA to be cost beneficial. Provide justification for why a SAMA(s) for the 4.16 kV Switchgear Bus Rooms 3A and 3B should not be considered and evaluated.

b. Table E.5-1, Level 1 Importance List Review, identifies SAMA 5 as a mitigation strategy for event QHUFWP7Y, however Section E.6.8 does not identify that event as being mitigated by SAMA 5. Clarify this discrepancy.

c. Table E.5-1, Level 1 Importance List Review, identifies event HHUMPSBY, OPERATORS FAIL TO START STANDBY MAKEUP PUMP, as having an RRW value of 1.059 and a failure probability of 1.OE+00. It is further stated that a SAMA was not formulated because the current procedures and training are believed to be adequate to start and align the standby makeup pumps. Explain why the failure probability for this event is 1.OE+00 if the procedures and training related to this event are adequate.

Provide further justification for why a SAMA that improves procedures and training, or provides for a hardware modification, is not applicable.

d. Table E.5-1, Level 1 Importance List Review, identifies event APWNR01 R, BOTH EDGS FTS, BOTH EFPS FTS, as having an RRW value of 1.044 and a probability of 6.40E-01.

This failure denotes the likelihood that AC power will not be recovered in time for specified failures. It was further stated that "No specific SAMA was identified to change the AC power non-recovery value but a SAMA was identified to provide an additional EDG". The SAMA identified was SAMA 18 at an estimated cost of more than

$5,000,000.

i. Provide further justification for why a SAMA to enhance procedures and training is not considered.

ii. Provide an evaluation of the costs and benefits of providing AC power from one of the other Crystal River power plants (Crystal River 1 and 2).

e. Table E.5-1, Level 1 Importance List Review, identifies event HHUMBACY, OPERATOR FAILS TO SWITCH MUP-1B POWER SOURCE, as having an RRW value of 1.027 and a probability of 1.OE+00. This event is described as failure to locally swap power supply to the "swing" pump. Proposed SAMA 15 is described as providing remote control room capability to realign power to pump MUP-1 B. The cost of this SAMA was estimated to be $400,000 and was determined not to be cost beneficial. Provide an evaluation of the costs and benefits of developing local manual swap-over procedures and training (or enhancing procedures and training if they exist) in lieu of SAMA 15.

f. Table E.5-1 identifies several initiating events for which no SAMA was identified to reduce probability, but for which "basic events relating to mitigation are addressed separately." Identify the basic events and associated SAMAs that mitigate these initiating events, including: IE_S (small break LOCA), IE_R (steam generator tube

U. S. Nuclear Regulatory Commission Enclosure 3F1009-03 Page 51 of 74 rupture), IE_T1 I (loss of intake), IE.T3 (loss of offsite power), IE_T1 (reactor trip), IET8 (loss of 4160V ES Bus 3A), IEA (large break LOCA), IET16 (loss of makeup), IET2 (loss of main feedwater), and IE_T10 (loss of NSCCC).

g. Section E.5.1 states that industry Phase II SAMAs were reviewed for potential applicability to CR-3, but does not identify the specific nuclear plants reviewed. Section E.1 1 references the SAMA analyses for Calvert Cliffs, Robinson, and Brunswick nuclear power plants. Clarify if these were the plants for which the industry Phase II SAMA review was performed, and if there were other plants included in the review. Also clarify whether any of the Phase I SAMAs were identified as a result of the CR-3 review of these other SAMA analyses.

h. Section 4.20 states that approximately 25 Phase I SAMAs were identified for consideration in the SAMA analysis. Table E.5-3 lists and describes each of these Phase I SAMAs. The SAMA numbers for these SAMAs range from 1 to 52, are not consecutive, and do not correspond to the SAMA ID Numbers for the industry SAMAs identified in Addendum 1. This suggests that a pre-screening of the identified SAMAs occurred prior to the Phase I screening. Clarify the process used to develop the Phase I SAMAs. Furthermore, on p. E.7-3 it is stated that the Phase I screening process involved qualitative disposition of 9 SAMAs. Based on review of Table E.5-3 and the discussion in Section E.5.2, it appears that: (1) the Phase I screening was quantitative rather the qualitative, and (2) 10 SAMAs were screened out versus 9. Clarify the Phase I screening process.

Response

5.a As documented in E.7.2.3 of the ER, SAMA 49 was only cost beneficial when the 9 5 th percentile PRA results were considered. The corresponding averted cost-risk was determined to be $194,594, which was $44,594 greater than the estimated implementation cost of $150,000. SAMAs for 4160V ES Switchgear Bus Rooms 3A and 3B were not considered in the ER given that they would clearly not have been cost beneficial based on the external events contributions that were assumed in the ER. For example, the 4160V ES Switchgear Bus Room 3A CDF (7.31E-06/yr) is only 49 percent of the Battery ChargerRoom 3A CDF (1.49E-05/yr), which would correlate to an averted cost-risk of $95,469 given that the averted cost-risk scales linearly with CDF for the external events SAMAs.

If a larger external events contribution is assumed (as suggested in RAI question 3c),

then the fire scenarios could no longer be considered "low contributors"and it would not necessarily be possible to preclude the SAMAs for the 4160V ES Switchgear Bus Rooms from consideration. However, the results of the IPEEE are based on very limited credit for the existing Thermo-Lag fire barriers. The credit taken for Thermo-Lag is key to determining whether or not a SAMA to improve the fire barrierswould be effective. If credit is taken for the existing Thermo-Lag barriers,a SAMA to improve the fire barriers would have very little impact on the CDF.

The benefit of the fire barriers in the 4160V ES Switchgear Bus Rooms, whether they are Thermo-Lag or "improved"fire barriers, would be to prevent damage to the cables from the opposite division that are routed through the room. A fire in one of the switchgearrooms will result in the loss of one division of power, so improvements to the

U. S. Nuclear Regulatory Commission Enclosure 3F1009-03 Page 52 of 74 fire barriers for cables within the division of the initiating fire would have limited effectiveness, but protecting cables from the opposite division could have a large impact on risk. For cases in which fire suppression fails, the cross-division cables would be failed regardlessof the fire barriersused; however, when fire suppression is successful, the fire barriersare an effective means of preventing damage. For the CR-3 IPEEE, the dominant contributorsare cases in which fire suppression was successful because such limited credit was taken for Thermo-Lag (a high probabilityof cable damage is combined with the high probability of suppression system success). Current fire modeling techniques would credit Thermo-Lag for preventing cable damage in these scenarios and indicate that the risk from cross-divisionalcable damage is overstatedin the IPEEE.

Fire barrierimprovements in the 4160V ES Switchgear Bus Rooms could be suggested as SAMAs and they could be shown to be cost effective using the external events multiplier of 12 and the IPEEE Thermo-Lag assumptions; however, the IPEEE is not considered to be the most appropriate tool for such an evaluation. CR-3 is currently updating the plant's fire model and the insights from the model will be used to identify areas for potential improvements, which may include the fire barriers in the 4160V ES SwitchgearBus Rooms.

5.b See response to RAI 6.d.

5. c HHUMPSBY is assigned a value of 1. OE+00 in the cutset file after the HRA recovery rule file has evaluated the dependency with other operatoractions and assigned the recovery factor. The actual HRA values assigned to HHUMPSBY is given in the table below based upon various situations.

HHUMPSBY OPERATORS FAIL TO START NON-ES SELECTED 5e-1 MAKEUP PUMP HHUMPSBY (T) OPERATORS TO START FAIL PUMP MAKEUP NON-ES SELECTED (TRANSIENTS) 8.6e-3 In general, SAMAs to "improve procedures and training"are not useful unless a specific problem with the procedures/traininghas been identified.

The CR-3 operators are already trained on the use of the makeup (HPI) pumps and no procedure deficiencies have been identified. The relatively high failure probability for the non-transient version of the action is based on the limited time that is available to perform the pump start and is not a function of the quality of the plant procedures and trainingprograms.

For the transient version of the action, the HRA documentation identifies that explicit guidance is not provided to start the non-ES selected makeup pump, but that the EOPs direct the operators to ensure that at least one train of HPI is running. Starting the standby pump would satisfy the EOP requirement to ensure one train of HPI is running; therefore, the HEP meets the requirementsof the EOPs.

In conclusion, enhancing the plant procedures and/or training program would have a negligible impact on the HEPs for startingthe standby makeup pump.

U. S. Nuclear Regulatory Commission Enclosure 3F1009-03 Page 53 of 74 5.d.i Procedure changes are low cost SAMAs that can be an effective means of reducing risk; however, a procedure change (or improved training)must target a specific weakness in order for it to impact the plant. No such weaknesses have been identified in the,power restoration procedures at CR-3. In addition, even if changes were made to the training program or procedures to address a specific issue, the power recovery probability used in the PRA is data based and the improvements would not be reflected in the PRA model. Consequently, there would be no measurable benefit related to any potential improvements relatedto procedure/trainingenhancements.

5.d.ii The other Crystal River power plants, Units 1, 2, 4, and 5, already provide power to the CR-3 switchyard. In order to gain any additionalbenefit from these plants, a dedicated line would have to be laid in conjunction with the installationof a new transformer. New connections to the CR-3 emergency buses would also have to be added. Because the loss of offsite power events that would require the use of the dedicated line would likely be caused by a weather related event, the dedicated line would have to be buried to ensure power would be available in such an event.

The Calvert Cliffs SAMA analysis estimated the cost of installing a connection to an alternate offsite power source to be significantly greaterthan $25,000,000. An alternate SAMA in the Calvert Cliffs analysis estimated that the cost of burying the offsite power lines would also be significantly greater than $25,000,000. While the Calvert Cliffs SAMA analysis does not specify the length of the new transmission line that was the basis for the $25,000,000 implementation cost, the CR-3 configuration has been reviewed and the Calvert Cliffs implementation cost is not unreasonable for CR-3. The major factors considered in the CR-3 design include:

The addition of one or more transformers to reduce the voltage of the offsite source to 4kV AC,

There are plans to decommission Units I and 2, implying that the dedicated line would have to be run from either Unit 4 or 5, o Units 4 and 5 are fartherfrom Unit 3 than Units I and 2, o A significant amount of asphalt and concrete exist where the buried line would be installed, which complicates cable burial.

None of the other Crystal River units are black start units, which means that additional steps would have to be taken to ensure that the dedicated line would be available when it was required.

With regard to the potential benefit of implementing this SAMA, it would be essentially the same as installing an additional EDG (SAMA 18). The response to RAI 3.d documents that the averted cost-risk for SAMA 18 is only $494,973 even when an external events multiplier of 12 is used in conjunction with the 9 5 th percentile PRA results. Even if the cost of installing the buried, dedicated line from Crystal River Unit 4 (or 5) was assumed to be half of the Calvert Cliffs estimate, the net value would be highly negative:

$494,973 - $12,500, 000 = -$12,005,027 This'type of change would not be cost beneficial for CR-3.

U. S. Nuclear Regulatory Commission Enclosure 3F1009-03 Page 54 of 74

5. e During the quantification process, the non independent HRA events are assigned a value of 1.0. The recovery factor that is added to the cutset is based upon which HRA events appear and their dependency upon each other. The MOR06 value for HHUMBACY is 2.8E-1.

In general, SAMAs to "improve procedures and training"are not useful unless a specific problem with the procedures/traininghas been identified.

The CR-3 operators are already trained on the action to align the alternatepower source to MUP-1B and no procedure deficiencies have been identified. The high failure probability for the action is based on the long manipulation time relative to the time available rather than the quality of the plant procedures and training programs.

Enhancements to the procedures and training program would have a negligible impact on the HEP and as suggested in SAMA 15, the most effective means of reducing the HHUMBACY HEP would be to simplify the alignment process through physical changes.

5.f In order to document the correlationbetween the initiating events identified in this RAI to the basic events that were included in Table E.5-1 of the ER, a new table (Table RAI-5.f) has been created. This table maps each of these initiating events to the basic events includedin Table E.5-1.

The mapping process was performed by isolating the cutsets relevant to each of the identified initiating events and extracting all of the associated basic events with RRW values greaterthan 1.02. In all cases, the basic events that were identified had already been dispositionedin Table E.5-1 of the ER. Table RAI-5.f provides a list of these basic events, grouped by initiating event, and the original ER text related to the event disposition.

U. S. Nuclear Regulatory Commission Enclosure 3F1009-03 Page 55 of 74 Table RAI-5.f: Correlation of Basic Events to Initiating Events Associated Initiator Basic Event RRW Description SAMA Disposition IES RHUPORVY 1.09 OPERATORS FAILS TO OPEN Important action for LOOP, LOFW, etc. Related SAMA identified to auto PORV FOR PRESSURE RELIEF open PORV: SAMA 35 OPERATORS FAIL TO SWITCH Operator action to switch to recirculation. Related SAMA has been

__-__1 NEON TO RECIRC INJECTION TO RECIRCULATION identified to automate switchover: SAMA 3.

IES SPMRW3BM 1.06 RWP-3B IN MAINTENANCE Unavailability / failure of raw water pump. A SAMA has been identified to supply water to the system from an alternate source: SAMA 8.

This is a basic event representing a decay heat removal pump being out IES LPM001 BM 1.04 DHP-1 B TRAIN IN MAINTENANCE of service for testing and maintenance. A SAMA was proposed to provide a diverse or maintenance spare train: SAMA 13.

This is a basic event representing a decay heat removal pump train IES LPM001AM 1.04 DHP-1A TRAIN IN MAINTENANCE being failed or out of service. Proposed SAMA: provide a redundant/

diverse spare or a maintenance spare DH train: SAMA 13.

This is a basic event representing a decay heat removal pump train IES LMMDHPBF 1.04 FAILURE OF DHP-1 B AND ITS being failed or out of service. Proposed SAMA: provide a redundant/

VALVES diverse spare or maintenance spare DH train, which could also be substituted for a failed train: SAMA 13.

This is a basic event representing a decay heat removal pump train IE_S LMMDHPAF 1.04 VALVES being failed or out of service. Proposed SAMA: add an additional DH train: SAMA 13.

Basic event representing inability to open sump valve for recirculation.

IES LMMDV43F 1.04 VALVE DHV-43 Proposed SAMAs: proceduralize either manual operation of the valve or crosstying of LHI suction: SAMA 16, 33.

TRAIN B RECIRC VALVE DHV-12 Basic event represents inability to open valve to supply HHSI/ MUP from IES LMMDV12F 1.03 FAILS LHI. Proposed SAMA: proceduralize either valve or crosstying of MUP suction: SAMA 9,manual operation of the 16 Basic event representing inability to open sump valve for recirculation.

IE_S LMMDV42F 1.03 VALVE DHV-42 Proposed SAMA: proceduralize either manual operation of the valve or crosstying of LHI suction: SAMA 16, 33.

TRAIN A RECIRC VALVE DHV-1 1 Basic event represents inability to open valve to supply HHSI/ MUP from IES LMMDV11F 1.03 FAILS LHI. Proposed SAMA: proceduralize either manual operation of the valve or crosstying of MUP suction: SAMA 9, 16.

IES SPMRW3AM 1.03 RWP-3A IN MAINTENANCE Unavailability / failure of raw water pump. A SAMA has been identified to I I supply water to the system from an alternate source: SAMA 8.

U. S. Nuclear Regulatory Commission Enclosure 3F1009-03 Page 56 of 74 ThhIA RAI-Sf Corr~I~tinn of R~cia~ FvAnt~ to Initi~tinn Fvnnt~

Initiator Associated RRW Description SAMA Disposition

_______ Basic Event RWDecito 3OPERATORS FAIL TO RAISE This is an operator action required to ensure adequate "boiler-IE_S QHUEFW9Y 1.03 OTSGs LEVEL condenser" mode cooling during small LOCAs. Suggested SAMA:

automate level control setpoint change: SAMA 14 This is a conditional operator error probability, the likelihood that COND PROB OF RHUPORVY operators will fail to open a PORV given that they failed to raise OTSG IES ZHUCOM1Z 1.03 GIVEN QHUEFW9Y level. A SAMA has been identified to automate the change in OTSG level setpoint. Automating that action would allow greater focus on the second action. No further SAMA is suggested.

IES SMMDHCC13 1.03 DHCCC TRAIN B FAULTS Module containing various decay heat closed cooling system failures.

IS S C.Proposed SAMA: proceduralize crosstying of DHCC trains: SAMA 16.

IES SMMRW3BF 1.03 RWP-3B PUMP TRAIN FAILS TO Unavailability / failure of raw water pump. A SAMA has been identified to OPERATE supply water to the system from an alternate source: SAMA 8.

IER FLG X 1.79 TAG EVENT - LONG TERM This is not a basic event, but a tag identifying sequences involving recirc/

COOLING (HPR/LPR/REFILL) refill. Basic events for those sequences are addressed separately.

IER RHUPORVY 1.09 OPERATORS FAILS TO OPEN Important action for LOOP, LOFW, etc. Related SAMA identified to auto PORV FOR PRESSURE RELIEF open PORV: SAMA 35 Basic event for EFW pump FTR. SAMA related to EFW / AFW has been

- Midentified, to provide an independent train: SAMA 7.

OPERATORS FAIL TO START FWP- Oper. action to start AFW pump FWP-7. SAMAs related to EFW/AFW IER QHUFWP7Y 1.06 7have been identified, to provide autostart of FWP-7, and to provide an

- 7 additional train of AFW/ EFW: SAMAs 5, 7.

can be improved E6OPERATORS FAIL COOLDOWN VIA Operator action to cool down on SGTR. Someactions IE R RHUCOOLY 1.06 OTSG by improving procedures and training, however the CR-3 procedures and training are believed to be adequate.

This is the conditional probability of failure to open a PORV given failure IER ZHUCOM2Z 1.05 COND PROB OF RHUPORVY to initiate cooldown. No SAMA was identified to reduce the likelihood of GIVEN RHUCOOLY failure at one action presuming that operators failed to take another action.

IER QMMEFP2F 1.04 EFP-2 FAILS TO CONTINUE TO A SAMA has been idenftifed related to AFW / EFW, to provide an

- RUN additional train: SAMA 7.

IE_R SPLT_RA 1.04 SGTR OCCURS ON OTSG-A <SPLIT Split fraction, no SAMA required.

SGFRACTION>

IER SPLT_RB 1.04 SGTR OCCURS ON OTGS-B <SPLIT Split fraction, no SAMA required.

_ _ FRACTION>

U. S. Nuclear Regulatory Commission Enclosure 3F1009-03 Page 57 of 74 Table RAI-5.f: Correlation of Basic Events to Initiating Events Initiator Associated RRW Description SAMA Disposition Basic Event RWDecito IER SPLT RA 1.04 SGTR OCCURS ON OTSG-A <SPLIT Split fraction, no SAMA required.

FRACTION>____________________ ____

This module contains basic event failures which could lead to OTSG IE_R PMMICSAH 1.03 OTSG A LEVEL CONTROL FAULTS overfeed. Proposed SAMA: add redundant /diverse level controls:

SAMA 17.

This module contains basic event failures which could lead to OTSG IE_R PMMICSBH 1.03 OTSG B LEVEL CONTROL FAULTS overfeed. Proposed SAMA: add redundant /diverse level controls:

SAMA 17.

COMMON CAUSE FAILURE OF Common-cause failure of makeup valves. Proposed SAMA:

IE_R HCCMV44N 1.03 MUV-23, MUV-24, MUV-25, AND Proceduralize manual operation of these valves, which would address MUV-26 TO OPEN most modes of common-cause failure: SAMA 10.

This module contains basic event failures which could lead to OTSG IE_R PMMICSCC 1.02 ICS COMMON MODE FAULTS overfeed. Proposed SAMA: provide redundant /diverse level controls (SAMA 17).

This is not a basic event but a tag identifying sequences where HVAC is IE_T1 1 FLG_HVAC 1.37 VAILABILITY OF AC POWER required, primarily to provide cooling for EFW controls. Basic events for VAY those sequences are addressed separately.

This is not a basic event but a tag identifying sequences where HVAC is failed, control systems are potentially failed, and operator actions to IE Ti1 FLG QHUEFWMR 1.33 OPERATORS FAIL TO MANUALLY manually operate valves might be helpful. Events for those sequences OPEN CONTROL VALVE are addressed separately, however a SAMA is proposed to provide procedures and training for manual operation of the affected valves (EFV-55, -56, -57, -58): SAMA 34 SPLIT FRACTION - VALVES FAIL This is not a basic event but a split fraction. Related basic events are E CLOSED ON LOSS OF HVAC addressed separately.

This is not a basic event, but a tag identifying sequences involving loss IE_T11 FLG_SW 1.24 SW of service water. Basic events for those sequences are addressed separately.

Operator action related to importance of HVAC / cooling to EFW / EFIC.

OPERATORS FAIL TO USE Related SAMAs have been identified to provide automated replacement IE_T11 JHUCHP2R 1.14 DEDICATED CHILLED WATER of some of the functions, to manually perform some of the functions SYSTEM potentially lost (operate EFV-55, ...- 58), or to provide a substitute for the I_ __ potentially affected AFW/ EFW equipment: SAMAs 1, 26, 34.

U. S. Nuclear Regulatory Commission Enclosure 3F1009-03 Page 58 of 74 Table RAI-5.f: Correlation of Basic Events to Initiating Events Associated Initiator Basic Event RRW Description SAMA Disposition OPERATORS FAIL TO START FWP- Operator action to manually align FWP-7 AFW pump. Related SAMAs IE_T1 1 QHUFW7EY 1.12 7 BEFORE PORV LIFTS have been AFW/

identified to provide autostart of FWP-7 as well as to install an alternative EFW train with automatic start: SAMAs 4, 7.

AFW Pump FWP-7, related SAMA has been identified to provide an IETi 1 QPMFWP7M 1.11 FWP-7 IN MAINTENANCE alternate AFW/EVW train. Also a SAMA has been identified to reduce maintenance downtime for FWP-7: SAMAs 5, 7.

IETl1 JHUCHP2Z 1.11 JHUCHP2R Sequence-specific substitution for JHUCHP2R. See discussion related to that action.

Basic event for EFW pump FTR. SAMA related to EFW / AFW has been T 1identified, to provide an independent train: SAMA 7.

SAFETY RELIEF VALVE FAILS TO SRV FTC. No SAMA directly related to this event was identified however IE-Ti1 RMMRCVSC 1.07 CLOSE (STEAM RELIEF) SAMAs related to mitigating systems have been identified.

6OPERATORS FAIL TO START FWP- Oper action to start AFW pump FWP-7. SAMAs related to EFW/AFW I7ETll QHUFWP7Y 1.06 7 have been identified, to provide autostart of FWP-7 and to provide an additional train of AFW/ EFW: SAMAs 5, 7.

STAG EVENT - STUCK OPEN RELIEF This is a tag event intended to identify sequences involving a stuck-open IETA 1 FLGTBQR 1.05 BIrelief OPEN valve. The basic events related to those sequences are evaluated FTER B I0 separately.

IETi 1 QMMEFP2F 1.04 EFP-2 FAILS TO CONTINUE TO A SAMA has been identified related to AFW / EFW, to provide an RUN additional train: SAMA 7.

Module representing various CST failure modes. Proposed SAMA:

IETi 1 QMMCST 1.03 FAILURE OF CST WATER SUPPLY proceduralize use of alternate water sources in event of CST failure:

SAMA 38.

HVAC REQUIRED DUE TO This is not a basic event but a tag identifying sequences where HVAC is IE_T3 FLG_HVAC 1.37 AVAILABILITY OF AC POWER required, primarily to provide cooling for EFW controls. Basic events for those sequences are addressed separately.

This is not a basic event but a tag identifying sequences where HVAC is failed, control systems are potentially failed, and operator actions to IET3 FLGQHUEFWMR 1.33 OPERATORS FAIL TO MANUALLY manually operate valves might be helpful. Events for those sequences OPEN CONTROL VALVE are addressed separately, however a SAMA is proposed to provide procedures and training for manual operation of the affected valves (EFV-55, -56, -57, -58): SAMA 34 IET3 QSPLHVAC 1.33 SPLIT FRACTION - VALVES FAIL This is not a basic event but a split fraction. Related basic events are

. CLOSED ON LOSS OF HVAC addressed separately.

U. S. Nuclear Regulatory Commission Enclosure 3F1009-03 Page 59 of 74 Table RAI-5.f: Correlation of Basic Events to Initiating Events Associated Initiator Basic Event RRW Description SAMA Disposition This is not a basic event, but a tag identifying sequences involving loss ITAG EVENT - LOSS OF NORMAL of service water. Basic events for those sequences are addressed SSW separately.

Operator action related to importance of HVAC / cooling to EFW / EFIC.

OPERATORS FAIL TO USE Related SAMAs have been identified to provide automated replacement IET3 JHUCHP2R 1.14 DEDICATED CHILLED WATER of some of the functions, to manually perform some of the functions SYSTEM potentially lost (operate EFV-55, ....-58), or to provide a substitute for the potentially affected AFW/ EFW equipment: SAMAs 1, 26, 34.

Basic event for EFW pump FTR. SAMA related to EFW / AFW has been T Qidentified, to provide an independent train: SAMA 7.

OPERATOR FAILS TO START Operator action to start / align standby makeup pump. Some actions IET3 HHUMPSBY 1.06 STANDBY MAKEUP PUMP can be improved by improving procedures and training, however the CR-3 procedures and training are believed to be adequate.

This is a calculated value denoting the likelihood that AC power will not IE T3 APWNR01R 1.04 BOTH EDGS FTS, BOTH EFPS FTS be recovered in time for the specified failures. No specific SAMA was T Aidentified to change the AC power nonrecovery value but a SAMA was identified to provide an additional EDG: SAMA 18.

IET3 HHUINJAY 1.04 OPERATORS FAIL TO SWITCH Operator action to supply backup power to high head injection valves. A MUV-23/24 TO BACKUP POWER SAMA was identified to proceduralize manual alignment: SAMA 10.

IE T3 ADGES3BM 1.02 EGDG-1B IN MAINTENANCE Proposed SAMA: add another EDG (SAMA 18).

IE T3 ADGEG1CF 1.02 EGDG-1C FAILS TO RUN Proposed SAMA: add another EDG (SAMA 18).

IET3 HHUINJBY 1.02 OPERATORS FAIL TO SWITCH Proposed SAMA, proceduralize manual operation of these valves (SAMA MUV-25/26 TO BACKUP POWER 10).

IET3 AHUEGICY 1.02 OPERATORS FAIL TO START AND Proposed SAMA: add another EDG (SAMA 18).

ALIGN EGDG-1C__________________ _____

IETi FLGX 1.79 TAG EVENT - LONG TERM This is not a basic event, but a tag identifying sequences involving recirc/

-_-_COOLING (HPR/LPR/REFILL) refill. Basic events for those sequences are addressed separately.

This is not a basic event but a tag identifying sequences where HVAC is IE_T1 FLGHVAC 1.37 AVAILABILITY OF AC POWER required, primiarily to provide cooling for EFW controls. Basic events for those sequences are addressed separately.

U. S. Nuclear Regulatory Commission Enclosure 3F1009-03 Page 60 of 74 Table RAI-5.f: Correlation of Basic Events to Initiating Events Associated Initiator Basic Event RRW Description SAMA Disposition This is not a basic event but a tag identifying sequences where HVAC is failed, control systems are potentially failed, and operator actions to IETi FLGQHUEFWMR 1.33 OPERATORS FAIL TO MANUALLY manually operate valves might be helpful. Events for those sequences OPEN CONTROL VALVE are addressed separately, however a SAMA is proposed to provide procedures and training for manual operation of the affected valves (EFV-55, -56, -57, -58): SAMA 34 IETi QSPLHVAC 1.33 SPLIT FRACTION - VALVES FAIL This is not a basic event but a split fraction. Related basic events are CLOSED ON LOSS OF HVAC addressed separately.

This is not a basic event, but a tag identifying sequences involving loss IETTAG EVENT - LOSS OF NORMAL of service water. Basic events for those sequences are addressed SW E1 separately.

Operator action related to importance of HVAC / cooling to EFW / EFIC.

OPERATORS FAIL TO USE Related SAMAs have been identified to provide automated replacement IE_T1 JHUCHP2R 1.14 DEDICATED CHILLED WATER of some of the functions, to manually perform some of the functions SYSTEM potentially lost (operate EFV-55, ...- 58), or to provide a substitute for the potentially affected AFW/ EFW equipment: SAMAs 1, 26, 34.

IETi QHUFW7EY 1.12 OPERATORS FAIL TO START FWP- Operator action to manually align FWP-7 AFW pump. Related SAMAs

-2 7 BEFORE PORV LIFTS have been identified to provide autostart of FWP-7 as well as to install an alternative AFW/ EFW train with automatic start: SAMAs 4, 7.

AFW Pump FWP-7, related SAMA has been identified to provide an IET1 QPMFWP7M 1.11 FWP-7 IN MAINTENANCE alternate AFW/EVW train. Also a SAMA has been identified to reduce maintenance downtime for FWP-7: SAMAs 5, 7.

IET1 JHUCHP2Z 1.11 JHUCHP2R Sequence-specific substitution for JHUCHP2R. See discussion related

-__to that action.

Basic event for EFW pump FTR. SAMA related to EFW / AFW has been T Qidentified, to provide an independent train: SAMA 7.

OPERATORS FAIL TO SWITCH Operator action to switch to recirculation. Related SAMA has been INETO TO RECSRC identified to automate switchover: SAMA 3.

INJECTION TO RECIRCULATION IETi RMMRCVSC 1.07 SAFETY RELIEF VALVE FAILS TO SRV FTC. No SAMA directly related to this event was identified however

- CLOSE (STEAM RELIEF) SAMAs related to mitigating systems have been identified.

IETi QHUFWP7Y 1.06 OPERATORS FAIL TO START FWP- Oper action to start AFW pump FWP-7. SAMAs related to EFW/AFW 17 have been identified, to provide autostart of FWP-7 and to provide an

- 7 additional train of AFW/ EFW: SAMAs 5, 7.

U. S. Nuclear Regulatory Commission Enclosure 3F1009-03 Page 61 of 74 Table RAI-5.f: Correlation of Basic Events to Initiatina Events Associated Initiator Basic Event RRW Description SAMA Disposition OPERATOR FAILS TO START Operator action to start / align standby makeup pump. Some actions IE_T1 HHUMPSBY 1.06 STANDBY MAKEUP PUMP can be improved by improving procedures and training, however the CR-3 procedures and training are believed to be adequate.

IE T1 SPMRW3B3M 1.06 RWP-3B IN MAINTENANCE Unavailability / failure of raw water pump. A SAMA has been identiifed to T Psupply water to the system from an alternate source: SAMA 8.

5TAG EVENT - STUCK OPEN RELIEF This is a tag event intended to idenfiy sequences involving a stuck-open TAG EVENT FTER 5 B 1.05STUCK OPEN REL relief valve.

separately. The basic events related to those sequences are evaluated IETi QMMEFP2F 1.04 EFP-2 FAILS TO CONTINUE TO A SAMA has been idenftifed related to AFW / EFW, to provide an

-_RUN additional train: SAMA 7.

IETi SCCHDABF 1.04 COMMON CAUSE FAILURE OF HXs Proposed SAMA, add removable strainers ahead of heat exchangers:

-_DCHE-1A AND DCHE-1B PLUGGED SAMA 37.

IE_T1 FLGPHURMFWR 1.03 OPERATORS FAIL TO RECOVER This is a flag event. Related basic events are considered separately.

__________ MFW IE_T1 RCCDRODA 1.03 MECH FAILURE OF ENOUGH Part of ATWS initiating event logic. No relevant SAMA identified.

_____CONTROL RODS TO DROP OPERATORS FAIL TO CROSSTIE SAMAs have been identified to provide additional makeup / suction E EFW SOURCES supplies to AFW and EFW: SAMAS 7, 38.

This module contains basic event failures which could lead to OTSG IE_T1 PMMICSAH 1.03 OTSG A LEVEL CONTROL FAULTS overfeed. Proposed SAMA: add redundant/diverse level controls:

SAMA 17.

This module contains basic event failures which could lead to OTSG IE_T1 PMMICSBH 1.03 OTSG B LEVEL CONTROL FAULTS overfeed. Proposed SAMA: add redundant /diverse level controls:

SAMA 17.

IE T1 ADGES3BM 1.02 EGDG-1B IN MAINTENANCE Proposed SAMA: add another EDG (SAMA 18).

IETi HHUINJBY 1.02 OPERATORS FAIL TO SWITCH Proposed SAMA, proceduralize manual operation of these valves (SAMA MUV-25/26 TO BACKUP POWER 10).

Essentially a split fraction identifying the fraction of the time the IE_T1 MTC 1.02 MTC GREATER THAN 95% moderator temperature coefficient is too high to sufficiently limit an ATWS event. No SAMA identified.

This module contains basic event failures which could lead to OTSG IE_T1 PMMICSCC 1.02 ICS COMMON MODE FAULTS overfeed. Proposed SAMA: provide redundant /diverse level controls (SAMA 17).

IE_T1 AHUEG1CY 1.02 OPERATORS FAIL TO START AND Proposed SAMA: add another EDG (SAMA 18).

________ [ALIGN EGDG-iC

U. S. Nuclear Regulatory Commission Enclosure 3F1009-03 Page 62 of 74 Table RAI-5.f: Correlation of Basic Events to Initiating Events Associated Initiator Basic Event RRW Description SAMA Disposition IET8 FLGX 1.79 TAG EVENT - LONG TERM This is not a basic event, but a tag identifying sequences involving recirc/

-_-_COOLING (HPR/LPR/REFILL) refill. Basic events for those sequences are addressed separately.

HVAC REQUIRED DUE TO This is not a basic event but a tag identifying sequences where HVAC is IE_T8 FLG_HVAC 1.37 AVAILABILITY OF AC POWER required, primiarily to provide cooling for EFW controls. Basic events for those sequences are addressed separately.

HVAC REQUIRED DUE TO This is not a basic event but a tag identifying sequences where HVAC is IE_T8 FLG_HVAC 1.37 VAILABILITY OF AC POWER required, primiarily to provide cooling for EFW controls. Basic events for LA L those sequences are addressed separately.

This is not a basic event but a tag identifying sequences where HVAC is failed, control systems are potentially failed, and operator actions to IE T8 FLG QHUEFWMR 1.33 OPERATORS FAIL TO MANUALLY manually operate valves might be helpful. Events for those sequences OPEN CONTROL VALVE are addressed separately, however a SAMA is proposed to provide procedures and training for manual operation of the affected valves (EFV-55, -56, -57, -58): SAMA 34 IET8 QSPLHVAC 1.33 SPLIT FRACTION - VALVES FAIL This is not a basic event but a split fraction. Related basic events are CLOSED ON LOSS OF HVAC addressed separately.

Operator action related to importance of HVAC / cooling to EFW / EFIC.

OPERATORS FAIL TO USE Related SAMAs have been identified to provide automated replacement IE_T8 JHUCHP2R 1.14 DEDICATED CHILLED WATER of some of the functions, to manually perform some of the functions SYSTEM potentially lost (operate EFV-55, ....-58), or to provide a substitute for the potentially affected AFW/ EFW equipment: SAMAs 1, 26, 34.

IET8 QHUFW7EY 1.12 OPERATORS FAIL TO START FWP- Operator action to manually align FWP-7 AFW pump. Related SAMAs

- 7 BEFORE PORV LIFTS have been identified to provide autostart of FWP-7 as well as to install an alternative AFW/ EFW train with automatic start: SAMAs 4, 7.

-AFW Pump FWP-7, related SAMA has been identified to provide an IET8 QPMFWP7M 1.11 FWP-7 IN MAINTENANCE alternate AFW/EVW train. Also a SAMA has been identified to reduce maintenance downtime for FWP-7: SAMAs 5, 7.

IET8 JHUCHP2Z 1.11 JHUCHP2R Sequence-specific substitution for JHUCHP2R. See discussion related to that action.

IET8 QMMEFP3F 1.07 EFP-3 PUMP TRAIN FAILS TO RUN Basic event for EFW pump FTR. SAMA related to EFW / AFW has been identified, to provide an independent train: SAMA 7.

IET8 RMMRCVSC 1.07 SAFETY RELIEF VALVE FAILS TO SRV FTC. No SAMA directly related to this event was identified however

- CLOSE (STEAM RELIEF) SAMAs related to mitigating systems have been identified.

U. S. Nuclear Regulatory Commission Enclosure 3F1009-03 Page 63 of 74 Table RAI-5.f: Correlation of Basic Events to Initiatina Events Associated Initiator Basic Event RRW Description SAMA Disposition IET8 QHUFWP7Y 1 06 OPERATORS FAIL TO START FWP- Oper action to start AFW pump FWP-7. SAMAs related to EFW/AFW 7

1 have been identified, to provide autostart of FWP-7 and to provide an additional train of AFW/ EFW: SAMAs 5, 7.

IE T8 SPMRW3BM 1.06 RWP-3B IN MAINTENANCE Unavailability / failure of raw water pump. A SAMA has been identiifed to T Psupply water to the system from an alternate source: SAMA 8.

This is a tag event intended to idenfiy sequences involving a stuck-open IET8 FLG_TBQR 1.05 FATER B relief valve. The basic events related to those sequences are evaluated

-___,FTERBseparately.

IET8 QMMEFP2F 1.04 EFP-2 FAILS TO CONTINUE TO A SAMA has been idenftifed related to AFW / EFW, to provide an RUN additional train: SAMA 7.

This is a basic event representing a decay heat removal pump being out IE_T8 LPM001 BM 1.04 DHP-1 B TRAIN IN MAINTENANCE of service for testing and maintenance. A SAMA was proposed to provide a diverse or maintenance spare train: SAMA 13.

IET8 HHUINJAY 1.04 OPERATORS FAIL TO SWITCH Operator action to supply backup power to high head injection valves. A IMUV-23/24 TO BACKUP POWER SAMA was identified to proceduralize manual alignment: SAMA 10.

Module representing various CST failure modes. Proposed SAMA:

IET8 QMMCST 1.03 FAILURE OF CST WATER SUPPLY proceduralize use of alternate water sources in event of CST failure:

SAMA 38.

IE_T8 FLGPHURMFWR 1.03 OPERATORS FAIL TO RECOVER This is a flag event. Related basic events are considered separately.

______________ MFW MECH FAILURE OF ENOUGH IE_T8 RCCDRODA 1.03 CONTROL RODS TO DROP Part of ATWS initiating event logic. No relevant SAMA identified.

IET8 HHUMBACY 1.03 OPERATORS FAIL TO SWITCH Failure to locally swap power supply to "swing" pump. Proposed SAMA:

MUP-1B POWER SOURCE IN provide remote switching capability: SAMA 15.

Essentially a split fraction identifying the fraction of the time the IE_T8 MTC 1.02 MTC GREATER THAN 95% moderator temperature coefficient is too high to sufficiently limit an ATWS event. No SAMA identified.

IE A FLG X 1.79 TAG EVENT - LONG TERM This is not a basic event, but a tag identifying sequences involving recirc/

- COOLING (HPRPLPR/REFILL) refill. Basic events for those sequences are addressed separately.

IEA SPMRW3BM 1.06 RWP-3B IN MAINTENANCE Unavailability / failure of raw water pump. A SAMA has been identified to supply water to the system from an alternate source: SAMA 8.

This is a basic event representing a decay heat removal pump being out IEA LPM001BM 1.04 DHP-1 B TRAIN IN MAINTENANCE of service for testing and maintenance. A SAMA was proposed to provide a diverse or maintenance spare train: SAMA 13.

U. S. Nuclear Regulatory Commission Enclosure 3F1009-03 Page 64 of 74 Table RAI-5.f: Correlation of Basic Events to Initiating Events Associated Initiator Basic Event RRW Description SAMA Disposition This is a basic event representing a decay heat removal pump train IE_A LPM001AM 1.04 DHP-1A TRAIN IN MAINTENANCE being failed or out of service. Proposed SAMA: provide a redundant!

diverse spare or a maintenance spare DH train: SAMA 13.

This is a basic event representing a decay heat removal pump train IEA LMMDHPBF 1.04 FAILURE OF DHP-1 B AND ITS being failed or out of service. Proposed SAMA: provide a redundant/

VALVES diverse spare or maintenance spare DH train, which could also be substituted for a failed train: SAMA 13.

IEA SCCHDABF 1.04 COMMON CAUSE FAILURE OF HXs Proposed SAMA, add removable strainers ahead of heat exchangers:

-DCHE-1 A AND DCHE-1 B PLUGGED SAMA 37.

FAILURE OF DHP.1A AND ITS This is a basic event representing a decay heat removal pump train IEA LMMDHPAF 1.04 being failed or out of service. Proposed SAMA: add an additional DH VALVES train: SAMA 13.

Basic event representing inability to open sump valve for recirculation.

IEA LMMDV43F 1.04 VALVE DHV-43 Proposed SAMAs: proceduralize either manual operation of the valve or crosstying of LHI suction: SAMA 16, 33.

IE A SPMRW3AM 1.03 RWP-3A IN MAINTENANCE Unavailability / failure of raw water pump. A SAMA has been identified to

- Psupply water to the system from an alternate source: SAMA 8.

Basic event representing inability to open sump valve for recirculation.

IEA LMMDV42F 1.03 VALVE DHV-42 Proposed SAMA: proceduralize either manual operation of the valve or V Dcrosstying of LHI suction: SAMA 16, 33.

1IEA LHULPRCY 1.03 OPERATORS FAIL TO GO TO LOW Failure of operators to align plant for recirculation. Proposed SAMA:

PRESSURE RECIRCULATION automate switchover to recirculation: SAMA 3.

IEA SPMDHCBM 1.03 DHCCC TRAIN B IN MAINTENANCE Module containing various decay heat closed cooling system failures.

-_ Proposed SAMA: proceduralize crosstying of DHCC trains: SAMA 16.

HVAC REQUIRED DUE To This is not a basic event but a tag identifying sequences where HVAC is IE_T16 FLGHVAC 1.37 HVAILABILITY OF AC POWER required, primarily to provide cooling for EFW controls. Basic events for those sequences are addressed separately.

This is not a basic event but a tag identifying sequences where HVAC is failed, control systems are potentially failed, and operator actions to IET16 FLGQHUEFWMR 1.33 OPERATORS FAIL TO MANUALLY manually operate valves might be helpful. Events for those sequences OPEN CONTROL VALVE are addressed separately, however a SAMA is proposed to provide procedures and training for manual operation of the affected valves (EFV-55, -56, -57, -58): SAMA 34 IET16 QSPLHVAC 1.33 SPLIT FRACTION - VALVES FAIL This is not a basic event but a split fraction. Related basic events are

- CLOSED ON LOSS OF HVAC addressed separately.

U. S. Nuclear Regulatory Commission Enclosure 3F1009-03 Page 65 of 74 Table RAI-5.f: Correlation of Basic Events to Initiating Events Associated Initiator Basic Event RRW Description SAMA Disposition This is not a basic event, but a tag identifying sequences involving loss ET16TAG EVENT - LOSS OF NORMAL of service water. Basic events for those sequences are addressed SW 1 separately.

Operator action related to importance of HVAC / cooling to EFW / EFIC.

OPERATORS FAIL TO USE Related SAMAs have been identified to provide automated replacement IE_T16 JHUCHP2R 1.14 DEDICATED CHILLED WATER of some of the functions, to manually perform some of the functions SYSTEM potentially lost (operate EFV-55, ....-58), or to provide a substitute for the potentially affected AFW/ EFW equipment: SAMAs 1, 26, 34.

IET16 QHUFW7EY 1.12 OPERATORS FAIL TO START FWP- Operator action to manually align FWP-7 AFW pump. Related SAMAs

. 7 BEFORE PORV LIFTS have been identified to provide autostart of FWP-7 as well as to install an alternative AFW/ EFW train with automatic start: SAMAs 4, 7.

AFW Pump FWP-7, related SAMA has been identified to provide an IET16 QPMFWP7M 1.11 FWP-7 IN MAINTENANCE alternate AFW/EVW train. Also a SAMA has been identified to reduce maintenance downtime for FWP-7: SAMAs 5, 7.

IE T16 JHUCHP2Z 1.11 JHUCHP2R Sequence-specific substitution for JHUCHP2R. See discussion related

- Cto that action.

IET16 QMMEFP3F 1.07 EFP-3 PUMP TRAIN FAILS TO RUN Basic event for EFW pump FTR. SAMA related to EFW / AFW has been identified, to provide an independent train: SAMA 7.

IET16 RMMRCVSC 1.07 SAFETY RELIEF VALVE FAILS TO SRV FTC. No SAMA directly related to this event was identified however CLOSE (STEAM RELIEF) SAMAs related to mitigating systems have been identified.

6OPERATORS FAIL TO START FWP- Oper action to start AFW pump FWP-7. SAMAs related to EFW/AFW T16 QHUFWP7Y 1.06 7E hhave been identified, to provide autostart of FWP-7 and to provide an additional train of AFW/ EFW: SAMAs 5, 7.

OPERATOR FAILS TO START Operator action to start / align standby makeup pump. Some actions IE_T16 HHUMPSBY 1.06 STANDBY MAKEUP PUMP can be improved by improving procedures and training, however the CR-3 procedures and training are believed to be adequate.

This is a tag event intended to identify sequences involving a stuck-open IET16TAG EVENT - STUCK OPEN RELIEF relief valve. The basic events related to those sequences are evaluated 0FTER B separately.

IET16 QMMEFP2F 1.04 EFP-2 FAILS TO CONTINUE TO A SAMA has been identified related to AFW / EFW, to provide an

- RUN additional train: SAMA 7.

Module representing various CST failure modes. Proposed SAMA:

IET16 QMMCST 1.03 FAILURE OF CST WATER SUPPLY proceduralize use of alternate water sources in event of CST failure:

I I_ I_ SAMA 38.

U. S. Nuclear Regulatory Commission Enclosure 3F1009-03 Page 66 of 74 Table RAI-5.f: Correlation of Basic Events to Initiating Events Associated Initiator Basic Event RRW Description SAMA Disposition IE_T16 RCCDRODA 1.03 MECH FAILURE OF ENOUGH Part of ATWS initiating event logic. No relevant SAMA identified.

CONTROL RODS TO DROP______________ _________

IET16 QHUEFT2Y 1.03 OPERATORS FAIL TO CROSSTIE SAMAs have been identified to provide additional makeup / suction EFW SOURCES supplies to AFW and EFW: SAMAS 7, 38.

This module contains basic event failures which could lead to OTSG IET16 PMMICSAH 1.03 OTSG A LEVEL CONTROL FAULTS overfeed. Proposed SAMA: add redundant /diverse level controls:

SAMA 17.

This module contains basic event failures which could lead to OTSG IE_T16 PMMICSBH 1.03 OTSG B LEVEL CONTROL FAULTS overfeed. Proposed SAMA: add redundant /diverse level controls:

SAMA 17.

This module contains basic event failures which could lead to OTSG IE_T16 PMMICSCC 1.02 ICS COMMON MODE FAULTS overfeed. Proposed SAMA: provide redundant/diverse level controls (SAMA 17).

IET2 FLGX 1.79 TAG EVENT - LONG TERM This is not a basic event, but a tag identifying sequences involving recirc/

COOLING (HPR/LPR/REFILL) refill. Basic events for those sequences are addressed separately.

2OPERATORS FAIL TO START FWP- Operator action to manually align FWP-7 AFW pump. Related SAMAs IE7T2 QHUFW7EY 1.12 7BEORER ORSFIT LIFTS have been identified to provide autostart of FWP-7 as well as to install an alternative AFW/ EFW train with automatic start: SAMAs 4, 7.

AFW Pump FWP-7, related SAMA has been identified to provide an IET2 QPMFWP7M 1.11 FWP-7 IN MAINTENANCE alternate AFW/EVW train. Also a SAMA has been identified to reduce maintenance downtime for FWP-7: SAMAs 5, 7.

Basic event for EFW pump FTR. SAMA related to EFW / AFW has been E identified, to provide an independent train: SAMA 7.

IEOT2 HHUHPRCY OPERATORS FAIL TO SWITCH .

1.07 FROM HIGH PRESSURE Operator action to switch to recirculation. Related SAMA has been I-____ INJECTION JINJECTION TO TO RECSRC identified to automate switchover: SAMA 3.

RECIRCULATION IET2 RMMRCVSC 1.07 SAFETY RELIEF VALVE FAILS TO SRV FTC. No SAMA directly related to this event was identified however

- CLOSE (STEAM RELIEF) SAMAs related to mitigating systems have been identified.

This is a tag event intended to identify sequences involving a stuck-open IET2TAG EVENT - STUCK OPEN RELIEFrelief valve. The basic events related to those sequences are evaluated FTER B 5 separately.

IET2 QMMEFP2F 1.04 EFP-2 FAILS TO CONTINUE TO A SAMA has been identified related to AFW / EFW, to provide an RUN additional train: SAMA 7.

IET2 SCCHDABF 1.04 COMMON CAUSE FAILURE OF HXs Proposed SAMA, add removable strainers ahead of heat exchangers:

-_DCHE-1A AND DCHE-1B PLUGGED SAMA 37.

U. S. Nuclear Regulatory Commission Enclosure 3F1009-03 Page 67 of 74 Table RAI-5.f: Correlation of Basic Events to Initiating Events Associated Initiator Basic Event RRW Description SAMA Disposition Module representing various CST-failure modes. Proposed SAMA:

IET2 QMMCST 1.03 FAILURE OF CST WATER SUPPLY proceduralize use of alternate water sources in event of CST failure:

SAMA 38.

IET2 RCCDRODA 1.03 MECH FAILURE OF ENOUGH Part of ATWS initiating event logic. No relevant SAMA identified.

_____CONTROL RODS TO DROP IET2 QHUEFT2Y 1.03 OPERATORS FAIL TO CROSSTIE SAMAs have been identified to provide additional makeup / suction EFW SOURCES supplies to AFW and EFW: SAMAS 7, 38.

This module contains basic event failures which could lead to OTSG IE_T2 PMMICSAH 1.03 OTSG A LEVEL CONTROL FAULTS overfeed. Proposed SAMA: add redundant/diverse level controls:

SAMA 17.

This module contains basic event failures which could lead to OTSG IET2 PMMICSBH 1.03 OTSG B LEVEL CONTROL FAULTS overfeed. Proposed SAMA: add redundant /diverse level controls:

SAMA 17.

COMMON CAUSE FAILURE OF Common-cause failure of makeup valves. Proposed SAMA:

IE_T2 HCCMV44N 1.03 MUV-23, MUV-24, MUV-25, AND Proceduralize manual operation of these valves, which would address MUV-26 TO OPEN most modes of common-cause failure: SAMA 10.

Essentially a split fraction identifying the fraction of the time the IE_T2 MTC 1.02 MTC GREATER THAN 95% moderator temperature coefficient is too high to sufficiently limit an ATWS event. No SAMA identified.

This module contains basic event failures which could lead to OTSG IE_T2 PMMICSCC 1.02 ICS COMMON MODE FAULTS overfeed. Proposed SAMA: provide redundant /diverse level controls (SAMA 17).

7HVAC REQUIRED DUE TO This is not a basic event but a tag identifying sequences where HVAC is IET10 FLG_HVAC 1.37 AVAILABILITY OF AC POWER required, primarily to provide cooling for EFW controls. Basic events for those sequences are addressed separately.

This is not a basic event but a tag identifying sequences where HVAC is failed, control systems are potentially failed, and operator actions to IET10 FLGQHUEFWMR 1.33 OPERATORS FAIL TO MANUALLY manually operate valves might be helpful. Events for those sequences OPEN CONTROL VALVE are addressed separately, however a SAMA is proposed to provide procedures and training for manual operation of the affected valves (EFV-55, -56, -57, -58): SAMA 34 IET1O QSPLHVAC 1.33 SPLIT FRACTION - VALVES FAIL This is not a basic event but a split fraction. Related basic events are

- CLOSED ON LOSS OF HVAC addressed separately.

U. S. Nuclear Regulatory Commission Enclosure 3F1009-03 Page 68 of 74 Table RAI-5.f: Correlation of Basic Events to Initiating Events Initiator Associated RRW Description SAMA Disposition Basic Event RWDecito TAG EVENT - LOSS OF NORMAL This is not a basic event, but a tag identifying sequences involving loss IE_-T10 FLGSW 1.24 SW of service water. Basic events for those sequences are addressed separately.

Operator action related to importance of HVAC / cooling to EFW / EFIC.

OPERATORS FAIL TO USE Related SAMAs have been identified to provide automated replacement IET10 JHUCHP2R 1.14 DEDICATED CHILLED WATER of some of the functions, to manually perform some of the functions SYSTEM potentially lost (operate EFV-55, ...- 58), or to provide a substitute for the Potentially affected AFW/ EFW equipment: SAMAs 1, 26, 34.

OPERATORS FAIL TO START FWP- Operator action to manually align FWP-7 AFW pump. Related SAMAs IET1- QHUFW7EY 1.12 7 BEFORE PORV LIFTS have been identified to provide autostart of FWP-7 as well as to install an alternative AFW/ EFW train with automatic start: SAMAs 4, 7.

,FW Pump FWP-7, related SAMA has been identified to provide an IET10 QPMFWP7M 1.11 FWP-7 IN MAINTENANCE alternate AFW/EVW train. Also a SAMA has been identified to reduce

_maintenance downtime for FWP-7: SAMAs 5, 7.

IET10 JHUCHP2Z 1.11 JHUCHP2R Sequence-specific substitution for JHUCHP2R. See discussion related

-_ to that action.

OPERATORS FAIL TO START FWP- Oper action to start AFW pump FWP-7. SAMAs related to EFW/AFW IET10 QHUFWP7Y 1.06 have been identified, to provide autostart of FWP-7 and to provide an additional train of AFW/ EFW: SAMAs 5, 7.

OPERATOR FAILS TO START Operator action to start / align standby makeup pump. Some actions IE_T10 HHUMPSBY 1.06 STANDBY MAKEUP PUMP can be improved by improving procedures and training, however the CR-3 procedures and training are believed to be adequate.

IE T10 SPMRW3BM 1.06 RWP-3B IN MAINTENANCE Unavailability / failure of raw water pump. A SAMA has been identified to T 0supply water to the system from an alternate source: SAMA 8.

Module representing various CST failure modes. Proposed SAMA:

IET10 QMMCST 1.03 FAILURE OF CST WATER SUPPLY proceduralize use of alternate water sources in event of CST failure:

SAMA 38.

Module containing various decay heat closed cooling system failures.

T MProposed SAMA: proceduralize crosstying of DHCC trains: SAMA 16.

IET10 SMMRW3BF 1.03 RWP-3B PUMP TRAIN FAILS TO Unavailability / failure of raw water pump. A SAMA has been identified to OPERATE supply water to the system from an alternate source: SAMA 8.

IET1O SPMDHCBM 1.03 DHCCC TRAIN B IN MAINTENANCE Module containing various decay heat closed cooling system failures.

I I_ Proposed SAMA: proceduralize crosstying of DHCC trains: SAMA 16.

U. S. Nuclear Regulatory Commission Enclosure 3F1009-03 Page 69 of 74 5.g The SAMA analyses for the referenced plants were reviewed to determine any insights that might be gained from what other utilities had proposed as potential cost-beneficial SAMAs. This review that was subjectively performed did not reveal any new insights than what had already been identified from the importance list review regardingpotential SAMAs that would be of benefit to Crystal River Unit 3. As such, no detailed explanation of this process was necessary since it was not considered to provide any additional value to this SAMA analysis. The most important insight in identification of SAMAs for Crystal River Unit 3 was from the review of the plant-specific PRA importance lists that provided more valuable information with regardto assessingpotential SAMAs that would be cost beneficial.

The one industry SAMA that was chosen to apply to Crystal River Unit 3 was SAMA 49, which involved the improvement of fire barriersto reduce the risk due to fire hazards in the plant. This was identified from SAMA ID #248 in Addendum 1.

5.h None of the SAMA ID numbers for the industry SAMAs identified in Addendum I were used. The listing of SAMA IDs was initially meant to consist of a consecutive numbering scheme, but during the review process of the basic event importance lists during the Phase I process, there were some SAMA numbers that were either subsumed into other identified SAMAs or determined that no SAMA was necessary, such as due to the event being a logical flag. However, to prevent having to renumber the SAMAs already identified, it was determined to not renumber the entire list so as to prevent possible configuration management errors when working with other personnel across different organizations.

According to Table E.5-3, the actual total number of Phase I SAMAs that were finally identified as being qualitatively screened was in fact ten. It was quite possible that earlier in the process that perhaps only nine were identified as candidates for being screened,but that an additionalSAMA was later added to the list and the description on page E. 7-3 not updated.

The Phase I screening process involved a comparison of the implementation costs with the perceived risk benefit and was explained in Section E.5.2. For convenience, the main points of this review process are reiteratedbelow:

a) Applicability to the Plant: If a proposed SAMA does not apply to the CR-3 design, it is not retained.

b) Engineering Judgment: Using extensive plant knowledge and sound engineering judgment, potential SAMAs are evaluated based on their expected maximum cost and dose benefits; those that are deemed not beneficial are screened from further analysis.

The SAMA identification process is focused on the identification of plant enhancements to address plant specific risk. Accordingly, the CR-3 PRA was the primary source of information used to develop the SAMA list. Specifically, the importance list is analyzed item by item using cutset analysis to determine the main risk contributorsfor each basic event and methods to mitigate the main risk contributors for each basic event are devised. An industry SAMA list, such as the one provided in NEI 05-01, is typically consulted to aid in the development of SAMAs. In many cases, other plants have

U. S. Nuclear Regulatory Commission Enclosure 3F1009-03 Page 70 of 74 identified general types of plant enhancements that could be used to address the issues raised in the importance list review. These general enhancements are then tailored to address plant specific issues, as required, and included on the SAMA list. This practice reduces the effort required for SAMA development and helps to ensure potentially reasonable changes are not overlooked in the SAMA development process. In many cases, however, the industry SAMA list does not include a plant enhancement that is applicable to the risk contributorin question and an entirely new SAMA is developed based on plant-specific needs. The result of the process is a list of SAMAs that can impact the important risk contributorsfor the plant.

In addition to the importance list review, which is performed for both the Level 1 and Level 2 contributors,previous plant specific risk analyses (including the IPE and IPEEE) are reviewed to determine if any previously identified plant improvements remain unimplemented. Any unimplemented plant enhancement would be a candidate for considerationas a potential SAMA.

Beyond the IPEEE plant enhancement review noted above, the results of the external events analyses were reviewed to determine if any other potentially cost effective plant enhancements may exist that were not identified during the IPEEE process. This review is generally difficult given that IPEEEs typically lack detailed quantitative information for many of the external events initiators, but the major risk contributors are examined to identify the types of changes that could be used to mitigate risk.

The potentially cost effective SAMAs from a set of selected submittals are also reviewed to identify potentially cost beneficial changes that may have been overlooked in the plant specific SAMA identification process. The majority of the sites chosen for this type review are usually of a similardesign to the plant being analyzed as the SAMAs have a better chance of being relevant; however, at least one dissimilar plant is included to introduce an alternate set of potential changes. There is no formal review process used to evaluate the cost effective SAMAs from these plants; the analyst qualitatively assesses the SAMAs to identify changes that impact risk areas that were not the focus of the plant specific importance list review. The objective is to identify reasons why those types of changes are not relevant to the plant being analyzed. In addition, SAMAs that address common risk areas using different methods are also considered to determine if they could be used in place of an existing SAMA. The use of these industry SAMA analyses is similar to the use of the generic SAMA list, but it provides a means of maintaining a link to the latest industry thinking without forcing a formal analysis of an ever growing SAMA list.

In summary, the CR-3 SAMA identification process primarily used the PRA to focus resources on developing plant changes that would most effectively reduce plant risk.

The process also relies on previous industry analyses to gain further insight and help ensure other important applicable SAMA designs are not overlooked. This is considered to be the most effective and prudent method of generatinga plant's SAMA list.

U. S. Nuclear Regulatory Commission Enclosure 3F1009-03 Page 71 of 74

6. Provide the following with regard to the Phase II cost-benefit evaluations:

a. Section E.6 introduction states that CR-3 specific implementation cost estimates were developed by plant personnel, and footnote 1 to Table E.5-3 states that "Cost estimates provided/validated by CR-3." Section E.5.1.1 further states that procedural changes have previously been estimated to cost about $50,000. Beyond this, no further basis is provided for implementation cost estimates. Provide a general explanation of the basis for CR-3-specific SAMA implementation cost estimates developed by plant personnel.

b. For a number of the Phase II SAMAs listed in Table E.6-1, the information provided does not sufficiently describe the associated modifications and what is included in the cost estimate. Provide a more detailed description of both the modifications and the cost estimates for Phase II SAMAs 4, 5, 15, 35, and 49.

c. Analysis of SAMA 49, Upgrade Fire Barriers in Battery Charger Room 3A (Section E.6.15), assumes a 13.1 percent risk reduction. This is based on two assumptions: 1) the contribution of Battery Charger Room 3A to the total (external + internal) CDF is 26.3 percent and 2) the external event CDF approximately equals the internal events CDF.

However, IPEEE Section 1.4 indicates that the CDF for this room (1.49E-05 per year) is 149 percent of the internal event (including internal flooding) risk (about 1.OE-05 per year). Justify the benefit estimate for this SAMA (see related RAIs 3.a - 3.d).

d. Table E.5-1, Level 1 Importance List Review, identifies potential SAMAs 5 and 7 to address event QHUFWP7Y, OPERATORS FAIL TO START FWP-7. But Table E.5-2, Level 2 Importance List Review for RRW Greater than 1.02, identifies potential SAMA 4 to address this same event, QHUFWP7Y. Neither SAMA 4 nor SAMA 5's benefit evaluation considers the risk reduction related to event QHUFWP7Y. Since SAMA 5 improves maintenance unavailability, it appears that SAMA 4, Automatic Start of Auxiliary Feedwater Pump (FWP-7) When Required, is the appropriate SAMA to address QHUFWP7Y. Clarify which SAMA(s) were considered to address event QHUFWP7Y, and explain why the risk reduction associated with mitigation of event QHUFWP7Y is not credited in the analysis of SAMA 4.

e. SAMA 16 (enhance procedures and make design changes as required to facilitate crosstying trains of DH, DHCC, etc.) has an estimated cost of $5M. This cost appears high for what appears to be mostly a procedure issue. Justify the cost estimate for this SAMA.

f. Section E.7.2.1 states that no additional Phase I SAMAs were retained for further analysis as a result of the uncertainty analysis using the 95th percentile CDF. Using the 95th percentile CDF results in a Modified Maximum Averted Cost-Risk (MMACR) of $1.4 million ($682,000 x 2.1). This is more than a factor of 2 greater than the cost estimates for Phase I SAMAs 8, 14, 26, 37, and 52. Provide a Phase II evaluation of these SAMAs.

Response

6. a Initial cost estimates were developed during the identification of Phase 1 and Phase 2 SAMAs to establish the list of potentially cost effective SAMAs. These were forwarded to CR-3 engineeringpersonnel for further review. This review consisted of an evaluation

U. S. Nuclear Regulatory Commission Enclosure 3F1009-03 Page 72 of 74 of each SAMA from the perspective of what actions would be required in each case to actually implement it. This involved determining what new equipment would have to be installed, how it would be housed, how it would be interconnected with plant systems, and generally how large or small the SAMA implementation would be. Based on this scoping effort, an estimate was provided based on benchmarking to other projects of similar size. This effort was not expected to be precise, but was expected to provide a ballparkevaluation of the size and scope of the proposed SAMA.

6.b SAMA 4 evaluates the risk benefit of changing Auxiliary Feedwater Pump FWP-7 from manual to automatic operation for providing a back-up means of supplying feedwater to the OTSGs in the event the automated EFW system in unavailable. Providing automatic operation of this pump would require detection of loss of main feedwater and low flow from the EFW flow instrumentation. It further assumes that normal power is available and does not include auto-startingthe backup power source.

SAMA 5 addresses possible improvement in Auxiliary Feedwater Pump FWP-7 unavailability. To better evaluate this SAMA, a 7-year unavailability review was performed to determine the significant contributors to pump unavailability. Excluding a modification performed to install an alternate power feed to the pump, unavailability was 0.9 during the period and was mostly associated with surveillances. One significant failure was identified with a bearing failure in 2000 and the cause for that failure was addressed. With such a low unavailability, there were no readily identifiable actions short of major equipment replacements that would materiallyimpact pump operation and it would not be certain that major equipment replacements would have a positive impact.

The estimate for this SAMA assumed that major hardware modifications would be required.

SAMA 15 simulates the ability to remotely realign the power supply for make-up pump MUP-1B in lieu of local manipulations from outside the control room. The make-up and purification (MUP) system provides for inventory and water chemistry control of the reactor coolant, and for emergency makeup (high pressure injection or HPI). The system consists of three makeup pumps that are powered from two trains of engineered safeguard (ES) 4160 VAC electrical buses. MUP-1A is powered from trainA of electrical power and MUP-1C is powered from train B. MUP-1B acts as a swing pump that can be powered from either 4160 VAC bus, but must be manually realigned. The $300,000 estimated cost for this SAMA is based on the need to install equipment that would allow remote alignment of eithertrain to MUP-1C including controls and breakersnecessary to operate the equipment in a mannerthat protects both trains.

SAMA 35 attempts to automate the process of cooling down the plant and performing what the operators would normally do when opening the PORV for manual pressure control. The RCS pilot-operatedelectromatic relief valve (PORV) is normally designed to open to relieve RCS pressure during overpressure conditions, due to exceeding a pressure setpoint or by remote operation. A solenoid energizes to open the PORV, and deenergizes to allow the PORV to close. In certain plant scenarios,such as during plant transients that cause an excessive increase in RCS pressure, e.g., loss of main feedwater, the operator may be required to manually open the PORV from the control room to prevent challenging the safety relief valves. The cost for this SAMA was based on the need to provide safety related power and the complexity of the automated controls required to provide adjustable band control of the PORVs.

U. S. Nuclear Regulatory Commission Enclosure 3F1009-03 Page 73 of 74 SAMA 49 addresses upgrading the fire barrierprotection in Battery ChargerRoom 3A in the same manner that Battery Charger Room 3B had been upgraded to meet the requirements of 10 CFR 50 Appendix R. As stated in E.6.15, the $150,000 cost for this upgrade was a benchmark estimate based on previous work to upgrade Battery Charger Room 3B.

6.c As stated in RAI question 6.b, the benefit for SAMA 49 was based, in part, on the assumption that the internal and external events risks are approximately equal. If it is assumed, as suggested in RAI question 3.c, that the external events multiplier should be based directly on the external events CDF, then the averted cost-risk would be larger than what was estimated in the ER.

As documented in the response to RAI 3.d, the averted cost-risk for SAMA 49 could be estimated by multiplying the internal events MACR by the ratio of the Battery Charger Room 3A fire CDF to the internalevents CDF:

$341,000 x 1.49E-05 / 4.95E-06 = $1,026,444 Note that the internal events CDF of 4.95E-06/yr includes internal flooding contributions.

Given that this SAMA was already determined to be cost beneficial based on an implementation cost of $150,000, re-quantifying the averted cost-risk of this SAMA in this way would not alter the conclusions about SAMA 49; it still has a positive net value

($1,026,444 - $150,000 = $876,444).

6.d Although SAMAs 4, 5, and 7 are all related to improving the reliabilityand redundancy of the EFW system, it would be more appropriateto identify event QHUFWP7Y with SAMA 4, which was created to postulate auto-startof FWP-7. To model SAMA 4, the failure probability for HEP event QHUFW7EY (OPERATORS FAIL TO START FWP-7 BEFORE PORV LIFTS) was reduced from a value of 1.0 to 1E-5 in the PRA basic event database. Also, in the recovery rules file, the recovery event QHUFW7EZ was commented out so as not to append this recovery event with a failure probabilityof 2.6E-02 to cutsets containing QHUFW7EY. Since event QHUFW7EY had a higher RRW value (RRW=I. 115) than QHUFWP7Y (RRW=1.063), it was implied that the latter event would be bounded by evaluation of the former event.

6.e Implementation of SAMA 16 would involve both changes to operating procedures and modifications to plant safety systems as noted in the discussion of SAMAs 16 and 52 under SAMA RAI 3.d above. Therefore, the cost is increased above that required for a procedure issue alone.

6.f The Phase II evaluations of SAMAs 8, 14, 26, 37, and 52 are documented in the response to RAI question 3.d using the external events multiplier of 12. The results are summarized in the table below for both the point estimate PIRA results and the 95 th percentile PRA results.

U. S. Nuclear Regulatory Commission Enclosure 3F1009-03 Page 74 of 74 PhaseII Results Summary for RAI 6.f (EE Multiplier of 12)

SAMA Cost of Averted Net Value Averted Net Value ID Implementation Cost (EE x 12 PE Cost Risk (EE x 12, 95th Risk PRA) (EE x 12, 95th Percentile)

(EE x 12 Percentile)

PE PRA) 8 $500,000 $357,972 -$142,028 $780,379 $280,379 26 $2,000,000 $555,600 -$1,444,400 $1,211,208 -$788,792 14 $900,000 $54,528 -$845,472 $118,871 -$781,129 37 $600,000 $54,084 -$545,916 $117,903 -$482,097 52 $5,000,000 -$5,508 -$5,005,508 -$12,007 -$5,012,007 Of the SAMAs identified for evaluation in this RAI question, none are potentially cost beneficial when the point estimate PRA results are used in conjunction with the external events multiplier of 12. When the 9 5 th percentile PRA results are applied in conjunction with the external events multiplier of 12, only SAMA 8 was determined to be potentially cost beneficial.

7. Section 4.20 states that "Progress Energy will consider the four SAMAs using the appropriate CR-3 design process." Describe the "CR-3 design process" and clarify how the four SAMAs, and any other SAMAs determined to be potentially cost-beneficial in response to these RAIs, are evaluated using this process.

Response

An action within the CR-3 Corrective Action Program will be used to track the evaluation of the potentially cost-beneficial SAMAs identified by the CR-3 License Renewal Environmental Report and subsequent RAl responses. This evaluation will provide a more detailed analysis of the actions required to implement the SAMAs and the costs for implementation. Those SAMAs that remain as potentially cost-beneficial SAMAs will be forwarded to the Project Review Group (PRG). The PRG will determine which of the potentially cost-beneficial SAMAs merit further study or implementation. If all or some of the potentially cost-beneficial SAMAs are accepted by the PRG they will be entered into the Long Range Plan and tracked from that point as a project.

v t e Letter Sequence Response to RAI
TAC:ME0278, Elimination of Hydrogen Recombiners and Change to Hydrogen and Oxygen Monitors (Approved, Closed)
Initiation Request, Request Acceptance Supplement Administration Withholding Request Acceptance, Withholding Request Acceptance Meeting, Meeting Questions 20 April 2009 3 August 2009 10 August 2009 20 August 2009 14 August 2009 14 August 2009 11 September 2009 20 August 2009 31 August 2009 31 August 2009 31 August 2009 2 September 2009 22 September 2009 15 October 2009 3 November 2009 27 October 2009 1 December 2009 30 November 2009 2 February 2010 21 May 2010 2 June 2010 8 July 2010 8 July 2010 14 October 2010 8 November 2010 16 November 2010 31 March 2011 10 June 2011 30 June 2011 8 November 2011 Answers 9 October 2009 11 May 2009 11 September 2009 11 September 2009 18 September 2009 30 September 2009 2 October 2009 13 October 2009 22 October 2009 5 November 2009 12 November 2009 3 December 2009 30 December 2009 27 January 2010 3 March 2010 1 April 2010 5 November 2009 1 April 2010 21 June 2010 28 June 2010 28 June 2010 9 August 2010 12 November 2010 23 November 2010 16 December 2010 29 December 2010 25 February 2011 15 March 2011 26 April 2011 15 June 2011 20 July 2011 29 November 2011 Results Approval, Approval, Approval Other: 3F1109-06, State of Florida Department of Environmental Protection, Conditions of Certification, 3F1209-04, Response to Follow-up to Progress Energy RAI Responses on CR-3 SAMA Evaluation, ML090090233, ML090210171, ML090350657, ML090420362, ML090780840, ML091200415, ML091660559, ML091820465, ML092600889, ML092610309, ML093200023, ML093520079, ML100050166, ML100980517, ML100980519, ML100980520, ML100980523, ML100980524, ML100980525, ML100980526, ML100980528, ML101110235, ML101320401, ML101460577, ML110250183, ML110410597, ML110460629, ML110820185, ML110880294, ML11137A009, ML11140A100, ML11180A275, ML11343A639, ML12117A260, ML12213A058, ML12346A011
MONTHYEAR2009-06-10 [Table View]

3F1009-03, Response to Request for Additional Information Regarding Severe Accident Mitigation Alternatives

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