ML022180805

Results of the Palisades Nuclear Plant SDP Phase 2 Notebook Benchmarking Visits

ML022180805
Person / Time
Site:	Palisades
Issue date:	08/05/2002
From:	Reinhart F Division of Systems Safety and Analysis
To:	Carpenter C NRC/NRR/DIPM
Franovich R, NRR/DSSA/SPSB, 415-3361
References
Download: ML022180805 (17)

Text

August 5,2002 NOTE TO: Cynthia Carpenter, Chief Inspection Program Branch Division of Inspection Program Management Office of Nuclear Reactor Regulation Patrick D. OReilly Operating Experience Risk Applications Branch Division of Risk Analysis and Applications Office of Nuclear Regulatory Research FROM: Mark F. Reinhart, Chief /RA/ Mark Caruso for Licensing Section Probabilistic Safety Assessment Branch Division of Systems Safety and Analysis Office of Nuclear Reactor Regulation

SUBJECT:

RESULTS OF THE PALISADES NUCLEAR PLANT SDP PHASE 2 NOTEBOOK BENCHMARKING VISITS During December, 2001, NRC staff and a contractor visited the Palisades site to compare the Palisades Significance Determination Process (SDP) Phase 2 notebook and licensees risk model results to ensure that the SDP notebook was generally conservative. The Palisades PSA did not include external initiating events so no sensitivity studies were performed to determine any impact of these initiators on SDP color determinations. In addition, the results from analyses using the NRCs draft Revision 3i Standard Plant Analysis Risk (SPAR) model for Palisades were also compared with the licensees risk model. The results of the SPAR model benchmarking effort will be documented in a separate trip report to be prepared by the Office of Research.

In the review of the Palisades SDP notebook for the benchmark efforts, it was determined that some changes to the SDP worksheets were needed to reflect how the Palisades plant is currently designed and operated. Thirty one hypothetical inspection findings were processed through the SDP notebook. Results from this effort indicated that the total risk impacts modeled in the SDP notebook were underestimated by 10 percent, overestimated by 39 percent, and adequately estimated by 52 percent. The reviewers found that if 24 fixes were made to the SDP notebook, the results would be 3 percent underestimation and 32 percent overestimation of risk impacts.

Attachment A describes the process and results of the comparison of the Palisades SDP Phase 2 Notebook and the licensees PSA.

CONTACT: M. Franovich, SPSB/DSSA/NRR 301-415-3361 Attachments: As stated

August 5, 2002 NOTE TO: Cynthia Carpenter, Chief Inspection Program Branch Division of Inspection Program Management Office of Nuclear Reactor Regulation Patrick D. OReilly Operating Experience Risk Applications Branch Division of Risk Analysis and Applications Office of Nuclear Regulatory Research FROM: Mark F. Reinhart, Chief /RA/ Mark Caruso for Licensing Section Probabilistic Safety Assessment Branch Division of Systems Safety and Analysis Office of Nuclear Reactor Regulation

SUBJECT:

RESULTS OF THE PALISADES NUCLEAR PLANT SDP PHASE 2 NOTEBOOK BENCHMARKING VISITS During December, 2001, NRC staff and a contractor visited the Palisades site to compare the Palisades Significance Determination Process (SDP) Phase 2 notebook and licensees risk model results to ensure that the SDP notebook was generally conservative. The Palisades PSA did not include external initiating events so no sensitivity studies were performed to determine any impact of these initiators on SDP color determinations. In addition, the results from analyses using the NRCs draft Revision 3i Standard Plant Analysis Risk (SPAR) model for Palisades were also compared with the licensees risk model. The results of the SPAR model benchmarking effort will be documented in a separate trip report to be prepared by the Office of Research.

In the review of the Palisades SDP notebook for the benchmark efforts, it was determined that some changes to the SDP worksheets were needed to reflect how the Palisades plant is currently designed and operated. Thirty one hypothetical inspection findings were processed through the SDP notebook. Results from this effort indicated that the total risk impacts modeled in the SDP notebook were underestimated by 10 percent, overestimated by 39 percent, and adequately estimated by 52 percent. The reviewers found that if 24 fixes were made to the SDP notebook, the results would be 3 percent underestimation and 32 percent overestimation of risk impacts.

Attachment A describes the process and results of the comparison of the Palisades SDP Phase 2 Notebook and the licensees PSA.

CONTACT: M. Franovich, SPSB/DSSA/NRR 301-415-3361 Attachments: As stated G://SPSB/Franovich/PAlisadesbench.wpd OFFICE *SPSB SPSB:SC *RIII NAME MFranovich:nyc MReinhart Mark Caruso for SBurgess DATE 06/20/02 06/05/02 07/03/02 OFFICIAL RECORD COPY

SUMMARY

REPORT ON BENCHMARKING TRIP TO PALISADES NUCLEAR PLANT (Dec. 11-12, 2001)

G. Martinez-Guridi Brookhaven National Laboratory (BNL)

Energy Sciences and Technology Department Upton, NY 11973 June 17, 2002 Attachment A

Table of Contents Page

1. Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1

2. Summary Results from Benchmarking . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2

3. Proposed Revisions to Rev-0 SDP Notebook . . . . . . . . . . . . . . . . . . . . . . . . . . 8 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 List of Tables Page Table 1. Comparison of Sensitivity Calculations between SDP Phase 2 Worksheets and Palisades RAWs . . . . . . . . . . . . . . . . . . . . . . 4 Table 2. Comparative Summary of the Benchmarking Results . . . . . . . . . . . . . . . 7

-iv-

1. Introduction A benchmarking of the Palisades Nuclear Plant (PNP) Significance Determination Process (SDP)

Risk-Informed Inspection Notebook was conducted during a plant site visit on December 11-12, 2001. NRC staff (Sonia Burgess, Mike Franovich and See Meng Wong) supported by BNL staff (G. Martinez-Guridi) participated in this benchmarking exercise.

In preparation of the plant site visit, BNL staff reviewed the Rev-0 PNP SDP notebook and evaluated a set of hypothetical inspection findings using the Rev-0 SDP worksheets, plant system diagrams and information in the licensees updated PSA.

The major activities performed during this plant site visit were:

1. Discussed licensees comments on the Rev-0 SDP notebook.

2. Obtained listings of the Risk Achievement Worth (RAW) values for basic events of the internal events PRA model.

3. Identified a target set of basic events for the benchmarking exercise.

4. Performed benchmarking of the Rev-0 SDP worksheets with considerations of the licensees proposed modifications to the SDP notebook.

5. Identified areas of discrepancies and reviewed the licensees PSA model to determine the underlying reasons. Proposed additional changes to the SDP notebook if appropriate.

The benchmarking exercise provided insights for significant improvement to the SDP notebook.

In 29 of the 31 cases evaluated, the revised SDP notebook should obtain either a match or one order of magnitude (color) higher than the licensees PSA. As described below, in one of the two remaining cases (failure of one MSIV), the revised SDP notebook correctly obtains more than one order of magnitude (color) difference than the licensees PSA. In the other remaining case (Diesel generator 1-1 fails to start), the revised SDP notebook underestimates by one color. A brief description of these two cases follows:

1. Failure of one MSIV. We initially obtained a red color because the SDPs worksheet for Main Steam Line Break (MSLB) Outside Containment considers that 2/2 MSIVs are required to close to prevent pressurized thermal shock (PTS); the SDP worksheet assumes that PTS leads to core damage. The licensee is currently analyzing MSIV failures and their potential to cause PTS. Currently, the licensees PRA model does not include PTS due to MSIV failures. Therefore, the failure of an MSIV is currently evaluated as a green by the licensee. Subsequent to the benchmark trip, the SDP MSLB worksheet and event tree were revised to make the sequence that could lead to PTS similar and consistent with the SDP notebook treatment of PTS for other affected Combustion Engineering plants. This change resulted in a yellow color.

2. Diesel generator 1-1 fails to start. The licensees PRA model obtained a red color, while the SDP notebook got a yellow. The cause of the difference is the rounding error that is inherent in SDPs evaluations. This error is discussed in detail in the next chapter. Similar underestimations when evaluating a diesel generator have been observed in other benchmarking exercises, and a solution will be considered.

2. Summary Results From Benchmarking This Section provides the results of the benchmarking exercise. The results of benchmarking analyses are summarized in Table 1. Table 1 consists of six column headings. In the first column, the out-of-service components, including human errors are identified for the case analyses. The second column contains the basic event(s) selected to model the failure in the first column. The third and fourth columns show the RAW values and the associated colors based on the licensees latest PSA model. The colors assigned for significance characterization from using the Rev-0 SDP worksheets before incorporation of the licensees comments are shown in the fifth column. Finally, the colors assigned for significance characterization from using the SDP worksheets after incorporation of the licensees comments are shown in the sixth column.

Table 1 is divided into three sections. The first section, titled No match, contains the two events for which a match could not be obtained; they are discussed next.

1. Failure of one MSIV. As mentioned in the previous chapter, we obtained a Yellow color for this failure because the SDPs event tree logic and worksheet for Main Steam Line Break (MSLB) Outside Containment considers that 2/2 MSIVs are required to close or isolation of feedwater to the affected steam generator to prevent pressurized thermal shock (PTS);

the SDP worksheet assumes that PTS leads to core damage. The licensee is currently analyzing MSIV failures and their potential to cause PTS. Currently, the licensees PRA model does not include PTS due to MSIV failures. Therefore, the failure of an MSIV is currently evaluated as a green by the licensee.

2. Diesel generator 1-1 fails to start. The licensees PRA model obtained a red color, while the SDP notebook got a yellow. The cause of the difference is the rounding error that is inherent in SDPs evaluations. For example, a station blackout sequence consists of LOOP

• EAC

• REC4 where LOOP is the initiating event, EAC is Emergency AC Power (diesel generators), and REC4 is Recovery of AC Power in < 4 hrs. Assuming the probabilistic values for these elements of the sequence are as follows:

Element Licensee SDP LOOP 3 E-2 2 EAC (given one diesel generator is unavailable) 2.5 E-2 2 REC4 2.0 E-1 1 Total 1.5 E-4 5 From this table, the licensees estimate is about one order of magnitude larger than the SDPs.

Similar underestimations when evaluating a diesel generator have been observed in other benchmarking exercises, and a solution will be considered.

Table 1's second section is entitled Match or SDP one magnitude higher than licensees PSA:

Hardware failures, contains 24 hardware failures for which the revised SDP notebook obtained either a match or one order of magnitude (color) higher than the licensees PSA. Table 1's third section is entitled Match or SDP one magnitude higher than licensees PSA: Human Errors, and contains five Human Errors for which the revised SDP notebook obtained either a match or one order of magnitude (color) higher than the licensees PSA.

To evaluate a failure in the CCW system, a new worksheet was developed before the visit, and it was revised during the visit.

At the beginning of the visit, we also intended to evaluate the unavailability of the turbine bypass valve (TBV) and one Steam Generator safety valve. However, due to the redundancy available for steam relief from the steam generators, we decided to remove these components from the worksheets.

A comparative summary of the benchmarking results is provided on Table 2. Table 2 shows the number of cases where the SDP was more or less conservative, or the SDP matched the outcome from the licensees PRA model. The associated percentage of differences found for the 31 analyzed cases are also shown on Table 2. It is concluded that the pre-visit SDP Notebook could capture at least 90% of the significant inspection findings (see Table 2 summation of the cases matched and overestimated). The revised SDP notebook captured 97% of the true significance of inspection findings (either true color or more conservative).

Table 1 Comparison of Sensitivity Calculations Between SDP Phase 2 Worksheets and Palisades RAWs CDF = 6.16E-5 / year, White = 1.02, Yellow = 1.16, Red = 2.61 TRUNCATION=1E-9 / year Description Basic Event Name RAW Plant SDP SDP CDF Before After Color No match 1 MSIV M-AVMB-CV-0501 1.01 Green Red Yellow (SGA)

Diesel generator 1-1 fails E-DGME-K-6A 3.11 Red Yellow Yellow to start Match or SDP one magnitude higher than licensees PSA: Hardware failures Diesel generator 1-2 fails E-DGME-K-6B 3.56 Red Yellow Red to start Battery charger D-BCMT-ED-15 1.88 Yellow Yellow Yellow D-BCMT-ED-16 D-BCMT-ED-17 D-BCMT-ED-18 EDG1-1 and EDG1-2 E-DGCC-K-6AB-ME 21.3 Red Red Red common cause fail to start Compressor C-2B fails to I-CMME-C-2B 1.02 White Red Yellow start Charging pump P-55A G-PMME-P-55A 1.07 White Green White fails to start AFW pump P-8C fails to A-PMME-P-8C 1.21 Yellow Red Red start AFW turbine pump P-8B A-PMME-P-8B 1.19 Yellow Red Red fails to start Loss of Condensate pump IE-LOCPA 1.0 Green Green Green P-2A P-54A Containment Spray S-PMME-P-54A 2.17 Yellow Yellow Yellow pump fails to start Description Basic Event Name RAW Plant SDP SDP CDF Before After Color Common cause 3 S-PMCC-P54ABC-ME 42.3 Red Red Red Containment Spray pumps fail to start Battery D02 fails to D-BYMB-ED-02 3.73 Red Red Red discharge on demand HPSI pump P-66A fails to H-PMME-P-66A 3.44 Red Red Red start 1 pump of LPSI L-PMME-P67A 1.01 Green Yellow Green Common cause failure of O-RVCC-PORVS-MA 2.4 Yellow Red Red both PORVs to not open Turbine fails to trip with TTF 1.0 Green White White ATWS Primary Safety Relief RVO 1.0 Green White White Valves FTO Loss of Service Water IE-LOSWS 5.12 Red Red Red System Loss of 2400V bus 1D IE-LOBUS1D 1.14 White White White ARS: Failure of SIS(X) R-REMB-SIS-X2 1.29 Yellow Red Red signal to even-numbered R-REMB-SIS-X4 relays R-REMB-SIS-X6 R-REMB-SIS-X8 1 ADV (FTO) B-AVMA-CV-0780 1.09 White Yellow Yellow 1 Fan cooler V-FNME-V-1A 1.0 Green Green Green 1 SIT T-CVMA-CK-ES3117 1.0 Green Green Green Pump 7B of SW U-PMME-P-7B 2.02 Yellow Yellow Yellow 1 pump of CCW C-PMME-P-52B 1.01 Green White Green Match or SDP one magnitude higher than licensees PSA: Human Errors Fails to makeup to CST A-OOOT-CSTMKUP 5.56 Red Red Red Operator fails to W-AVOA-PZR-SPRAY 22.7 Red Red Red depressurize PCS with Pzr Spray / Aux Spray Description Basic Event Name RAW Plant SDP SDP CDF Before After Color Fails to initiate once H-ZZOA-OTC-INIT 2.67 Red Red Red through cooling Recovery of AC Power in R2H 1.0 Green White White

< 1 hr (REC1)

Recovery of AC Power in R4H 1.23 Yellow Yellow Yellow

< 4 hrs (REC4)

Table 2 : Comparative Summary of the Benchmarking Results Total Number of SDP Notebook SDP Notebook Cases Compared Before (Rev 0) After (Rev 1)

Number of Cases Percentage Number of Cases Percentage SDP: Less 3 10% 1 3%

Conservative SDP: More 12 39% 10 32%

Conservative SDP: Matched 16 52% 20 65%

Total 31 100% 31 100%

3. Proposed Revisions to Rev-0 SDP Notebook Based on insights gained from the plant site visit, a set of revisions are proposed for the Rev-0 SDP notebook. The proposed revisions are based on licensee comments on the Rev-0 SDP notebook, better understanding of the current plant design features, consideration of additional recovery actions, use of revised Human Error Probabilities (HEPs) and initiator frequencies, and the results of benchmarking.

3.1 Specific Changes to the Rev-0 SDP Notebook for the Palisades Nuclear Plant The licensee provided several comments on the Rev-0 SDP Notebook. In addition, several major revisions that directly impacted the color assignments by the SDP evaluation were discussed with the licensee and their resolutions were identified in the meeting. The suggested changes that had an impact on the evaluation of the worksheets were incorporated during the visit, including revised HEPs and initiator frequencies. The remaining changes dealt mainly with updated footnotes to the dependency matrix and the worksheets. The proposed revisions are discussed below:

1. Loss of CCW (LCCW) was added to row II of Table 1, Categories of Initiating Events for Palisades Power Plant. A worksheet for Loss of CCW also was developed, and Table 2, Initiators and System Dependency for Palisades Nuclear Power Plant was updated to indicate the new worksheet to be used for CCW failures.

On loss of CCW, the RCPs must be tripped by the operators within 10 minutes after the onset of the loss of cooling to the RCPs, but this event is not currently included in the licensees PRA model. However, after the benchmarking visit, and per SDP assumptions, we included this event in the SDPs worksheet and event tree, with the generic credit = 3, and leading to a small LOCA.

2. A generic change for all worksheets and associated event trees, except LOOP and LIA, was to remove the safety function Make Up to CST (MUCST) because this make up is provided automatically, and no operator action is required. In the LOOP and LIA scenarios the make up requires operator actions.

3. A generic change for many worksheets and associated event trees is that, since the plant can achieve decay heat removal using AFW during the mission time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, it is not necessary to ask for shutdown cooling (SDC) once AFW was successful. The worksheets and associated event trees were revised, and the dependencies for LPSI/SDC in Table 2, Initiators and System Dependency for Palisades Nuclear Power Plant, were updated.

4. The licensee clarified that to implement once-through-cooling (feed and bleed), the operators open the block valves by turning a control switch in the control room. The PORVs are opened by putting them in the OPEN position from the control room, causing a DC relay to energize. The licensees HEP for this action is 2.9E-3, and we assigned a credit = 3. The worksheets containing feed and bleed were updated, as well as their footnotes.

5. The licensee models high-pressure recirculation using two human actions: failure to align subcooling valves (HEP = 4.8E-3, event H-AVOA-HISUBCLG), and failure to enable ESS recirc valves (HEP = 2.6E-4, event Y-AVOB-RAS-VLVS). We assigned a credit of an operator action = 2.

6. Power Conversion System (PCS). One change implemented for the worksheets using the PCS was a success criteria of (1/2 MFW trains with 1/2 condensate trains) or (1/2 condensate trains with 1/4 ADVs). In Table 2, Initiators and System Dependency for Palisades Nuclear Power Plant, the Initiating Event column for the ADVs was updated according to the modifications to several worksheets. The footnote for the ADVs was modified to indicate that Nitrogen provides a supply to operate the ADVs.

7. Table 2, Initiators and System Dependency for Palisades Nuclear Power Plant. Footnote 1 was updated with the plant internal event CDF = 6.16E-5 / year.

8. Table 2, Initiators and System Dependency for Palisades Nuclear Power Plant. A new footnote was added to indicate the subsystems comprising the Actuating Relays System (ARS).

9. Table 2, Initiators and System Dependency for Palisades Nuclear Power Plant. The turbine bypass valve (TBV) and Steam Generator safety valves were removed because the safety functions containing these components were deleted.

10. Table 2, Initiators and System Dependency for Palisades Nuclear Power Plant. The Support Systems column for AFW was updated to include the Condensate Storage Tank (CST) instead of the Demineralized water storage tank. The CST is the primary suction source for the auxiliary feedwater (AFW) system. The demineralized water storage tank is the primary makeup source for the CST.

11. Table 2, Initiators and System Dependency for Palisades Nuclear Power Plant. A new row for Condensate Storage Tank (CST) was created to show its dependency on the Demineralized water storage tank and backup sources of water: the primary system makeup tank, fire protection system, and the service water system. The SW can provide suction water for only one (motor-driven) AFW pump (P-8C).

12. Table 2, Initiators and System Dependency for Palisades Nuclear Power Plant. The Initiating Event column for AFW was updated to All except MLOCA, LLOCA.

13. Table 2, Initiators and System Dependency for Palisades Nuclear Power Plant. Added a footnote for HPSI indicating that, according to the licensee, HPSI pumps only need CCW cooling when the temperature of the pumped water is higher than 325 degrees F.

14. Medium LOCA (MLOCA). The licensees PRA model credits Low Pressure Injection (LPI) if Early Inventory, HP Injection (EIHP) failed and PCS/AFW is successful to depressurize to the LPIs injection pressure. High Pressure Recirculation (HPR) is the preferred path for recirculation. Therefore, we conservatively did not give credit to the mitigation strategy of depressurizing to the LPIs pressure and using LPI. Hence, we removed the safety functions RCS Depressurization (RCSDEP), Low Pressure Injection (LPI), and Low Pressure Recirculation (LPR).

15. Large LOCA (LLOCA). The licensee indicated that it will use High Pressure Recirculation (HPR), as opposed to Low Pressure Recirculation (LPR). The safety function Low Pressure Recirculation (LPR) was removed.

16. Loss of Offsite Power (LOOP). The licensee indicated that the HEP for Recovery of AC Power in < 4 hrs (REC4) is larger than 0.05, so we changed the credit to operator action

= 1, instead of 2.

The licensees PRA model credits the MDAFW pumps if AC power is recovered within 30 minutes.

17. Steam Generator Tube Rupture (SGTR). The licensee indicated that when Secondary Heat Removal (AFW) and Pressure Equalization (EQ) are successful, water makeup still has to be provided to the vessel. For this makeup, the plant can use either the HPSI pumps or 3/3 charging pumps. A new safety function, Charging Pumps (CHAPUM), was added to give credit to the charging pumps for this purpose. The licensees HEP for aligning charging is 2.5E-4. Therefore, it is an operator action limited by hardware failure.

For Pressure Equalization (EQ), the licensee indicated that auxiliary spray can also be used.

The licensee indicated that the HEP for RCS Depressurization (RCSDEP) corresponds to a credit = 3, instead of 1.

18. Anticipated Transients without Scram (ATWS). The licensee indicated that the HEP for Emergency Boration (CVCS) corresponds to a credit = 1, instead of 2.

The licensees current analysis indicates that after the charging system injects sufficient boron to shutdown the reactor, one AFW pump is sufficient to remove decay heat. When new information on secondary heat removal requirements is available, the worksheet will be revised per communication to Regional SRA.

19. Main Steam Line Break (MSLB) Outside Containment. The licensee is currently analyzing MSIV failures and their potential to cause pressurized thermal shock (PTS). Currently, the licensees PRA model does not include PTS due to MSIV failures. During the visit, we considered that both MSIVs have to close to prevent PTS, and this success criteria was consolidated in the safety function Closure of Both MSIVs (MSIV); PTS could also be prevented by isolation of feedwater to the affected SG and hence the event tree logic and worksheet were revised.

20. Loss of SW (LSW). The RCPs must be tripped by the operators within 10 minutes after the onset of the loss of cooling to the RCPs, but this event is not currently included in the licensees PRA model. However, after the benchmarking visit, and per SDP assumptions, we included this event in the SDPs worksheet and event tree, with the generic credit = 3, and leading to a small LOCA.

21. Loss of IA (LIA). Loss of motive air fails the AFW injection valves open. The valves of pumps P-8A (motor-driven pump) & B (steam-driven pump) have nitrogen backup, which lasts for at least 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. During these 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> the plant has three alternatives: 1) recover Instrument Air, 2) the valves can be operated manually, or, 3) the licensee considers that the plant can reach entry conditions to SDC; operator actions are then needed to align SDC. Accordingly, the TDP of AFW was credited; it requires operator action (credit = 1),

which is limited by hardware failure, 1 ASD train.

Loss of IA also fails CV-3025 closed, and the SDC mode of heat removal requires operator actions. Loss of IA also fails the turbine bypass valve closed. Nitrogen provides a supply to operate the ADVs.

Make up water to CST can be provided from (1) demineralized water storage tank, (2) primary system makeup tank, (3) Service water, or (4) fire protection water. However, the valves of these sources of water to the CST are supported by instrument air, or can be locally manually operated, or are manual valves. On loss of instrument air, all valves can be locally manually operated, but the human actions required to provide make up to the CST are expected to be more involved since there is no automatic capability to open the valves. Hence, the safety function Make Up to CST (MUCST) was kept in this worksheet.

The licensees HEP is 2.6E-3.

22. Loss of One DC Bus (LDCB). CV-3025 fails closed and the shutdown heat exchangers cannot be used for heat removal. However, manual operation of this valve is possible, enabling the SDC mode of heat removal.

There are two battery chargers in each DC train. One charger is active, and the other is in standby. The licensee is currently analyzing whether a battery charger is capable of handling the starting of all safety loads. A footnote was added in Table 2, Initiators and System Dependency for Palisades Nuclear Power Plant.

23. Loss of One Engineered Safeguards AC Bus (LACB). In addition to the equipment lost that was identified in the Rev-0 SDP Notebook, two SW pumps are lost. Operator action is needed to reduce SW loads, so there is enough flow of SW to support SDC. SDC is not modeled because AFW can provide decay heat removal during the mission time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

24. Other footnotes containing minor clarifications were added or updated in Tables 1 and 2 and in the worksheets.

3.2 Generic Change in IMC 0609 for Guidance to NRC Inspectors No specific recommendation for changes to IMC 0609 was identified as a result of this benchmarking exercise. However, the following item was identified that can further streamline the process:

1. Additional training may be required for NRC inspectors to clarify the SDP evaluation rules and present to them a more complete set of examples.

3.3 Generic Change to the SDP Notebook No generic change was identified.

Attachment 1 List of Participants Sonia Burgess (NRC/Region III)

Mike Franovich (NRC/NRR)

See Meng Wong (NRC/NRR)

Bob White (NMC/PSA)

Mark Cimock (NMC/PSA)

Brian Brogan (NMC/PSA)

Kevin Shields (NMC/PSA)

Andy Spotts (NMC/PSA)

Scott Beck (INEEL)

G. Martinez-Guridi (BNL)