ML13227A003

Response to NRC Request for Additional Information Regarding the Environmental Review of the Sequoyah Nuclear Plant, Units 1 and 2, License Renewal Application

ML13227A003
Person / Time
Site:	Sequoyah
Issue date:	07/17/2013
From:	James Shea Tennessee Valley Authority
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
TAC MF0057, TAC MF0058
Download: ML13227A003 (107)

Text

L44 130717 001 Tennessee Valley Authority, 1101 Market Street, Chattanooga, Tennessee 37402 July 17, 2013 10 CFR Part 51 ATTN: Document Control Desk U,S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Sequoyah Nuclear Plant, Units 1 and 2 Facility Operating License Nos. DPR-77 and DPR-79 NRC Docket Nos. 50-327 and 50-328

Subject:

Response to NRC Request for Additional Information Regarding the Environmental Review of the Sequoyah Nuclear Plant, Units I and 2, License Renewal Application (TAC Nos. MF0057 and MF0058)

References:

1. TVA Letter to NRC, "Sequoyah Nuclear Plant, Units 1 and 2 License Renewal," dated January 7, 2013 (ADAMS Accession No. ML13024A004)

2. NRC Letter to IVA, "Requests for Additional Information for the Environmental Review of the Sequoyah Nuclear Plant, Units 1 and 2, License Renewal Application," dated May 10, 2013 (ADAMS Accession No.

MLI13119A083)

3. NRC Letter to TVA, "Revised Requests for Additional Information for the Environmental Review of the Sequoyah Nuclear Plant, Units 1 and 2, License Renewal Application," dated June 7, 2013 (ADAMS Accession No.

ML13136A358)

By letter dated January 7, 2013 (Reference 1), Tennessee Valley Authority (TVA) submitted an application to the Nuclear Regulatory Commission (NRC) to renew the operating license for the Sequoyah Nuclear Plant, Units 1 and 2. The request would extend the license for an additional 20 years beyond the current expiration date. By letter dated May 10, 2013 (Reference 2), the NRC forwarded a request for additional information (RAI).

Subsequently, the NRC revised the RAI by letter dated June 7, 2013 (Reference 3). The required response date for RAI question number 7 was June 24, 2013. Mr. David Drucker, the NRC Environmental License Renewal Project Manager has given a verbal extension to July 17, 2013 for RAI question number 7.

P&tned on~rftycWe pa~W"P9

U.S. Nuclear Regulatory Commission Page 2 July 17. 2013 to this letter provides TVA's response to the Reference 3 RAI question number 7.

There are no new regulatory commitments contained in this submittal.

Consistent with the standards set forth in 10 CFR 50.92(c), TVA has determined that the additional information, as provided in this letter, does not affect the no significant hazards considerations associated with the proposed application previously provided in Reference 1.

Please address any questions regarding this submittal to Henry Lee at (423) 843-4104.

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

Executed on this 17t day of July, 2013.

7 epully,

/. W. Shea (ice President, Nuclear Licensing

Enclosures:

1. TVA Responses to NRC Request for Additional Information cc (Enclosures):

NRC Regional Administrator - Region II NRC Senior Resident Inspector - Sequoyah Nuclear Plant

NRC RAI 7.a.i

7. Severe Accidents MitigationAlternatives (SAMA)

a. Provide the following information regardingthe ProbabilisticRisk Assessment (PRA) used for the SAMA analysis. Basis: Applicants for license renewal are requiredby 10 CFR 51.53(c) (3) (ii)(L) to consider SAMAs if not previously considered in an environmentalimpact assessment, relatedsupplement, or environmental assessment for the plant. As part of its review of the SQN SAMA analysis, NRC staff evaluates the applicant'streatment of internal events and calculation of core damage frequency in the Level 1 PRA model. The requestedinformation is needed in order for the NRC staff to reach a conclusion on the sufficiency of the applicant'sLevel 1 PRA model for supporting the SAMA evaluation.

i. ER Section E.1.1 states that the SQN PRA models reflect the SQN configuration and design as of November 30, 2009.

Identify and discuss any completed or planned changes to the plant design or operationsince these dates that might impact the SAMA analysis.

TVA Response As part of the design change process, plant design changes (i.e., Design Change Notices (DCNs)) are evaluated for their potential effect on the PRA model. If a DCN is evaluated as potentially affecting the PRA model, a PRA evaluation is performed to determine the potential change in risk, and the DCN is entered into the PRA modeling change database for inclusion in the next modeling update (provided the DCN has been installed in the plant). Any changes that would incur a core damage frequency (CDF) or large early release fraction (LERF) change of

+/-10% would result in a PRA model change after the DCN was installed.

The following DCNs from the PRA model change database qualified for their consideration for their effect on the SAMA analysis. As a result of that review, the following is a summary of those completed DCNs that had the possibility of affecting the SAMA analysis and their effect on the PRA model results.

Multiple Circuit Failure (Multiple Spurious Operation) Desigqn Changes Regulatory Guide (RG) 1.189 (Revision 2) states that post fire safe-shutdown circuit analysis should address all possible fire-induced failures that could affect the safe-shutdown success path, including multiple spurious operations (MSOs). A review of the SQN Circuit analysis considering RG 1.189 (Revision 2) guidance identified necessary design changes.

There were a total of 17 DCNs performed to resolve multiple circuit fault concerns at Sequoyah for Units 1 and 2. These modifications were discussed with the NRC in a meeting on September 5, 2012, to advise the NRC of the status of TVA plans and actions to date to resolve multiple circuit fault concerns involving undesired operation of systems and components at Sequoyah Units 1 and 2. These DCNs were reviewed for their impact on the PRA model and the SAMA analysis. DCNs 22616 and 22617 to replace the SQN USSTs are discussed below.

Of the remaining DCNs only 3 affected the internal events PRA model. These modifications to the plant included changes to the electrical boards supplying power to components, and the installation of an air to close valve to prevent steam generator overfill in an Appendix R fire. The El- 1 of 87

CAFTA revision 0 PRA model was revised to reflect these changes and was requantified. The results are listed below:

PRA Model Results for MSO Modifications Risk MetricI Baseline CDF/LERF New CDFILERF [Change in CDFILERF Ul CDF 3.0253E-05 3.0256E-05 3.000E-09 Ul LERF 4.3899E-06 4.3900E-06 1.000E-10 U2 CDF 3.5901E-05 3.5903E-05 2.000E-09 U2 LERF 4.6332E-06 4.6333E-06 1.000E-10 It can be concluded from the very small changes in CDF and LERF that the SAMA would not be impacted by the MSO DCNs.

Unit Station Service Transformer Modifications (DCN 22616 and 22617)

These DCNs installed new Unit Service Station Transformers (USSTs) on Units 1 and 2 to serve as the supply of offsite power to Units 1 and 2. The DCN for Unit 2 is complete and the DCN for Unit 1 is scheduled for the fall 2013 outage (U1R19). Currently each unit is normally supplied by the Common Station Service Transformers (CSSTs) A and C, with B CSST serving as a spare. After the completion of the Unit 1 DCN, the units will be supplied by USSTs during normal operation and the CSSTs will serve as the safety related offsite power source. As noted in the evaluation SQN-0-12-078, the effect on risk after the entire change is completed for both units is minimal, i.e., the delta CDF value is 1E-9 and the delta LERF value is 1E-10. Therefore these DCNs have no effect on the SAMA analysis.

Replacement Steam Generators for Unit 2 The current PRA model success criteria were developed based on the original Unit 2 steam generators (SGs), which is conservative. The Unit 2 SGs were replaced during the fall 2012 refueling outage. The replacement SGs (RSGs) have better heat transfer characteristics but not to the extent that the safety analysis would need to be altered; that is, the current UFSAR Chapter 15 analysis remains valid. Therefore the RSGs have no adverse effect on the SAMA analysis.

Planned Changes With Regard to Flood Mitigation By letter from TVA to NRC, dated April 16, 2013 (Reference 1), as updated by letter from TVA to NRC, dated July 1, 2013 (Reference 2), TVA committed to implementing design changes at SQN to improve external flood mitigation. As noted in Reference 2, TVA has committed to installing "newly designed means for removing reactor decay heat [during flooding scenarios]

and providing reactor coolant system makeup for both SQN Units." Although these design changes will not be evaluated for PRA impact until the design process has progressed further, TVA anticipates that the installation of these components would decrease CDF and LERF for events related to external flooding.

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References:

1. Letter, TVA to NRC, "Commitment to Install Improved Food mitigation Systems," dated April 16, 2013 (ADAMS Accession No. ML13108A107)

2. Letter, TVA to NRC, "Progress Update on Improved Flood Mitigation System Project,"

dated July 1, 2013 NRC RAI from a Telecom with TVA On June 18, 2013, in a TVA telecom with the NRC SAMA staff, the NRC wanted TVA to "Confirm that the SQN fuel inventory is applicable to the fuel cycle expected for SQN during the period of extend operation (PEO)for the license renewal."

TVA Response TVA confirms that the SQN fuel inventory is applicable to the fuel cycle expected for SQN during the PEO for the license renewal.

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NRC RAI 7.a.ii ii. Describe the primary reasonsfor the differences in major core damage frequency (CDF)contributorsbetween the two units as shown in Tables E. 1-1 and E.1-2 as well as the significant differences in SAMA CDF reductions between the two units as shown in Tables E£2-1 and E.2-2.

TVA Response The following table presents the different initiating event groups in the PRA model. To compare the two contributions, the availability factor for both models was set to "true," which removes any difference in values based on the different availability factors for each unit.

Unit I Unit 2 Difference Group Frequency Group Frequency (Unit 1- Unit 2)

Core Power Excursion 1.48E-08 Core Power Excursion 1.33E-08 1.51 E-09 Electric Board Room Electric Board Room BoardERoom3.57E-06 none Flooding Flooding Excessive Main Feedwater 9.86E-08 Excessive MFW 8.94E-08 9.19E-09 (MFW)

Inadvertent SI 6.27E-09 Inadvertent SI 5.83E-09 4.43E-10 Internal Flooding 1.52E-05 Internal Flooding 2.23E-05 -7.1OE-06 ISLOCAs 3.09E-08 ISLOCAs 3.09E-08 none Large LOCA 6.69E-09 Large LOCA 6.60E-09 8.84E-1 1 Loss of 480V Power 8.70E-09 Loss of 480V Power 5.74E-09 2.97E-09 Loss of all Component 3.94E-06 Loss of all CCS 3.43E-06 5.05E-07 Cooling System (CCS)

Loss of Component Cooling System (CCS) 1.02E-06 Loss of CCS Train A 8.45E-07 1.75E-07 Train A Loss of Vaum1.80E-07 Condenser Loss of Condenser Vaum9.87E-08 8.18E-08 Vacuum Vacuum Loss of Instrument Board 8.17E-07 Loss of Instrument Board 6.24E-07 1.92E-07 Loss of Offsite Power 7.68E-07 LOOP 4.49E-07 3.19E-07 (LOOP)

Loss of Plant Air 7.06E-08 Loss of Plant Air 4.47E-08 2.59E-08 Loss of RCP 1.07E-07 Loss of RCP 1.04E-07 2.61 E-09 Loss of Train A ERCW O.OOE+00 Loss of Train A ERCW O.OOE+00 none Loss of Unit Board 1.24E-08 Loss of Unit Board 9.42E-09 2.94E-09 Loss of Vital Battery Board 6.18E-08 Loss of Vital Battery 5.48E-08 7.03E-09 Board Medium LOCA 2.85E-08 Medium LOCA 2.85E-08 -8.20E- 11 MSIV Closures 7.47E-08 MSIV Closures 5.41 E-08 2.05E-08 El- 4 of 87

Unit I Unit 2 Difference Group Frequency Group Frequency (Unit 1- Unit 2 Non-Isolable LOCA 5.25E-09 Non-Isolable LOCA 6.13E-09 -8.74E-10 Partial Loss of MFW 4.57E-07 Partial Loss of MFW 4.47E-07 1.03E-08 Reactor Trip 1.03E-06 Reactor Trip 1.01E-06 2.18E-08 Small LOCA 4.34E-07 Small LOCA 4.90E-07 -5.63E-08 SG Pressure Over Relief 4.31E-09 SG PORV Fails Open 4.21E-09 1.06E-10 Valve (PORV) Fails Open SG Tube Rupture 3.28E-08 SG Tube Rupture 4.41 E-08 -1.13E-08 Steam Line Break Inside 1.34E08 Steam Line Break Inside 1.40E-08 -5.81E-10 Containment Containment Steam Line Break Outside 1.48E-06 Steam Line Break Containment Outside Containment Stuck Open Safety Relief 2.56E-06 Stuck Open Safety Relief 2.72E-06 -1.61 E-07 Valve Valve Total Loss of ERCW 0.OOE+00 Total Loss of ERCW 0.OOE+00 none Total Loss of MFW 5.85E-07 Total Loss of MFW 3.22E-07 2.64E-07 Turbine Trip 5.80E-07 Turbine Trip 5.67E-07 1.29E-08 Vessel Rupture 3.22E-08 Vessel Rupture 3.22E-08 none The two events in the above table with the largest CDF contribution difference between Unit 1 and Unit 2 CDFs, Internal Flooding and Loss of All CCS, are discussed below:

Internal Flooding The largest contributor to the risk difference between the two units is Internal Flooding. For the purposes of this analysis the Electric Board Room Floods were separated out. The Unit 2 internal flooding results are higher than the Unit 1 internal flooding results due to asymmetries in the plant routing of piping.

Specifically, on elevation 690 of the Auxiliary Building (AB) there is piping that is routed above the two "Boric Acid and Auxiliary Feedwater (AFW) Space Coolers" that could affect both of the motors on these coolers simultaneously. This flood would cause, in the PRA model, a failure of the AFW area cooling, which would cause a total loss of motor driven AFW for Unit 2. These flooding events contribute to 21% of the CDF on Unit 2 and only 0.2% of the CDF on Unit 1.

Loss of All CCS The total loss of Component Cooling System (CCS) initiating event contribution is higher for Unit 1 than for Unit 2. The difference between the two units is based on the number of valves that are in the flow path from the 2A Essential Raw Cooling Water (ERCW) header to the CCS heat exchangers. As can be seen in the following figure, there are three additional valves that could plug/transfer closed that do not exist in the flow path for the 2A CCS heat exchangers (1 -FCV-067-0478, 1-FCV-067-0223, and 2-FCV-067-0223).

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-ly ER-WSUPPLY IIEAER IS6 A POWER REMOVED HX 2A2 Comparison of SAMA Candidates The difference in risk reduction for the following SAMAs is based on the internal flooding affects discussed earlier. For Unit 2 there is a higher chance of having an internal flooding event that disables the area cooling for both motor driven AFW pumps (MDAFWP). Because the fault tree logic only requires an additional failure of the turbine driven AFW pump to enter into a feed and bleed sequence, the majority of accident sequences on Unit 2 involve failures of the human actions and associated hardware for feed and bleed going to recirculation.

0 SAMA 032 - This SAMA would remove the failure to align high pressure recirculation. In Unit 2, there are more accident sequences that involve this human action than on Unit 1, therefore Unit 2 has a greater reduction in risk for this SAMA.

• SAMA 068 - This SAMA would install an additional means of providing flow to the SGs.

As discussed, because there are more accident sequences on Unit 2 that involve feed and bleed, the new pump would inhibit some of these accident sequences and would therefore affect Unit 2 more than Unit 1.

• SAMAs 105 and 106 - These SAMAs would provide for additional Refueling Water Storage Tank (RWST) inventory due to the Containment Spray (CS) pumps not draining the RWST during the injection phase of a safety injection (SI) signal. The effect of these SAMAs would result in a larger risk reduction on Unit 2 as more accident sequences on Unit 2 are related to feed and bleed.

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SAMA 160 - This SAMA would allow for ventilation for specific areas to be enhanced using temporary Heating Ventilation Air Condition (HVAC). The Unit 2 model has a greater reduction in risk due to the flooding events not necessarily affecting all of the MDAFWPs. The temporary HVAC would preclude the use of the modeled area coolers and allow for fewer feed and bleed accident sequences.

SAMA 249 - This SAMA would allow for the RWST to be refilled fast enough to preclude the need to transfer to recirculation. All feed and bleed sequences in the SQN PRA model require operator actions to establish piggy-back flow from the RHR pumps to the SI or Centrifugal Charging Pumps (CCP). Because Unit 2 has more accident sequences with failure to perform recirculation (specifically from feed and bleed events), the effect of this SAMA is greater for Unit 2 than for Unit 1.

SAMA 275 - This SAMA would remove the influence of the internal flooding initiating events on elevation 690 of SQN. As discussed for SAMAs 068 and 160, keeping the AFW space coolers available during such events would reduce the number of feed and bleed sequences, which have a greater affect on Unit 2 than on Unit 1.

SAMA 289 - This SAMA would increase the redundancy in the space/area cooling on elevation 690 of the AB. As discussed for SAMAs 68 and 160, the increased availability of the AFW pumps decreases the number of accident sequences related to feed and bleed, which are a more important contributor to Unit 2 risk than Unit 1.

NRC RAI 7.a.iii iii. It is noted that the loss of offsite power (LOOP) initiatorcontributes only 1 to 2% to the CDF while station blackout (SBO) contributes 10 to 13%. Also the Level I importance analysis does not include any events for failure of the emergency diesel generators.

Explain the reasons why the LOOP contributionto CDFis so small compared to the SBO contributionto CDFand why the failure of the emergency diesel generators is so small a contributorto the Level 1 results that it does not appearin the Level 1 importanceanalysis results.

TVA Response The station blackout (SBO) contribution was calculated by evaluating gate SBOSEQS_U1/2_L2.

This gate calculates the total contribution to station blackout from all sources, not just the loss of offsite power (LOOP) initiating events. Station blackout, as noted, is a gate that accounts for all failures of both 6.9kV shutdown boards. There are other additional cutsets where there were failures of breakers and boards causing both shutdown boards to be de-energized, which is also considered an SBO. The total contribution for LOOP events to this gate is fifteen percent, with the remainder of the contribution coming from internal flooding events.

Diesel generator (DG) importance to the Level 1 PRA model is low for two primary reasons.

The first is that the success criteria for the ERCW system (which is closely linked to the DGs due to cooling for the DGs being performed by the ERCW pumps) varies based on the type of accident. For all events that do not require containment spray heat exchangers, i.e. non LOCA or steam line breaks inside containment, only one ERCW pump per train is necessary. So flow requirements are low enough that only one A train pump and one B train pump can provide enough cooling to all loads. Therefore, to get a core damage sequence from a LOOP requires El- 7 of 87

failure of all ERCW pumps. If there is a LOCA event then the success criteria changes to the design bases of one pump per header (1A, 2A, 1 B, and 2B), or four pumps.

In addition to the variable ERCW success criteria, SQN has an installed utility bus that allows the plant to power shutdown boards from other shutdown boards. This process is only used at SQN when one unit has a total loss of onsite power (failure of both DGs) and the other unit has both DGs available. Operators can de-energize one of the boards on the unit with both boards available and energize a board on the unit without any power.

NRC RAI 7.a.iv iv. Discuss the modeling of consequentialLOOP events in the SQN PRA and, if not considered, estimate the impact of this omission on the selection of cost beneficial SAMAs.

TVA Response Consequential LOOP was not modeled in the SQN PRA model. To assess the effect of consequential LOOP the fault tree model was updated by adding two new basic events to represent consequential LOOP from a LOCA and consequential LOOP for all other reactor trips.

The conditional loop probabilities were developed using References 1 and 2.

For all LOCA events, the consequential LOOP probability was set to 0.02 and for all other trips the LOOP probability was set to 0.002.

The increase in risk for consequential LOOP is 0.28%.

As stated at the end of TVA response to RAI 7.a.i, future design changes will be made to improve external flooding mitigation. These additional onsite resources will be designed to supply water to both units; thus, help mitigate LOOP events. Therefore, the contribution of consequential LOOP will be addressed in the response to RAI 7.a.i.

References

1. WCAP-1 6304-P, "Strategy for Identification and Treatment of Modeling Uncertainties in PSAs: Applications to LOCA and LOOP," Westinghouse Electric Corporation,Revision 0.

2. G. Martinez-Guridi and J. Lehner, "Generic Probability of a LOOP after a Large LOCA,"

Brookhaven National Laboratory, November 2, 2006 [ADAMS Number ML071430462].

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NRC RAI 7.a.v

v. ER Section E. 1.4.5 describes the CAFTA Revision 0 as a complete revision of the model that involved converting the model from the RISKMAN software platform into CAFTA format.

1. Clarify the extent to which models, data, success criteria, etc. from the prior version were utilized in the CAFTA version.

2. Was there a peer review of any RISKMAN PRA versions?

TVA Response 7.a.v.l: The CAFTA Revision 0 model was a complete reconstitution of the PRA model for SQN. There was no real conversion from the previous RISKMAN models, and information from previous models was not utilized. The SQN model started with the basic outline from the WBN Unit 1 CAFTA model, and recreated the success criteria, accident sequences, systems analysis, human reliability analysis, level 2 analysis, internal flooding, and data to match the configuration and operation at SQN. The PRA model is a dual unit model for both Units 1 and 2. The guidance of RG 1.200 Revision 2 (Reference 1) and ASME/ANS RA-Sa -2009 (Reference 2) were followed and the model was peer reviewed by a Pressurized Water Reactor Owner Group (PWROG) review team to Capability Category II of the RG and the ASME Standard. The major technical elements were reassessed in the creation of this model.

7.a.v.2: Previously there were four versions of the SQN RISKMAN model. There was a peer review performed by the PWROG of the 3rd version of the RISKMAN model (Revision 2) in November 2000; however, the ASME PRA standard had not been developed at that time, and peer reviews were based on engineering judgment. As stated above, the CAFTA Revision 0 model was a complete reconstitution of the PRA for SQN and previous versions of the RISKMAN models were not utilized as input.

References:

1. USNRC RG 1.200, Revision 2, "An Approach For Determining The Technical Adequacy Of Probabilistic Risk Assessment Results For Risk-Informed Activities," March 2009

2. ASME/ANS RA-Sa-2009, "Standard for Level 1/Large Early Release Frequency Probabilistic Rick Assessment for Nuclear Power Plant Applications" El- 9 of 87

NRC RAI 7.a.vi vi. Identify any systems shared between the two SQN units and how the availability/unavailabilityof these systems is incorporatedin the SQN PRAs during normal operation, shutdown, and dual unit trips.

TVA Response Shared Systems List The following systems are shared between the two units in the PRA model.

• ERCW System

" CCS System (only the B header)

• Plant Control Air

• Aux-Control Air

" Electric Power (off site power supplies)

• Raw Cooling Water (RCW)

Normal Operation During normal operation the above systems are available and providing their required function.

" ERCW system - provides heat removal from the CCS heat exchangers, room/area cooling to those components requiring heat removal, and CCP pump gear oil cooling.

Normal testing and maintenance of the ERCW pumps is included within the model.

" CCS system - provides heat removal for letdown, cools the RCP motors, and provides mechanical seal cooling for the operating CCP. Normal testing and maintenance of the CCS pumps and heat exchangers are modeled within the PRA. The plant can align the B CCS pumps to either the A train for the respective unit, or to the B train of CCS. The PRA model includes basic events for the alignment configuration for these pumps as well as test and maintenance events for these pumps.

There are two different success criteria built into the PRA model for the CCS model based on the ERCW inlet temperature. For conditions where the inlet temperature, is greater than 70 0 F, the CCS heat exchanger success criteria changes to require both of the CCS heat exchangers.

• Plant Control Air - provides air to those air-operated valves (AOVs) that are required for normal operation such as main feedwater (MFW) level control valves (LCV).

• Aux-Control Air - provides no function for normal plant operation. There are test and maintenance events built into the PRA model for these compressors.

" Electric Power - provides electric power to essential and non-essential loads. Some of the redundant power supplies, specifically the CSSTs, have test and maintenance events built into the PRA model.

• RCW - provides heat removal for secondary side components, provides area cooling for the reactor protection MG sets, and provides cooling to the ice condenser chillers. The RCW system also provides cooling to station air compressors.

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Normal Shutdown

• ERCW system - Performs the same functions as during normal operation

" CCS system - Performs the same functions as during normal operation except that when the plant is on shutdown cooling it provides heat removal from the RHR heat exchangers. If the operating unit were to have an accident while the B train of CCS was performing the RHR heat removal function for the outage unit, that header of the CCS would not be available to the accident unit for use in mitigating the accident. This is addressed in the PRA model by basic event FLGO070_UNITMODE56_RHRTRN_B

• Plant Control Air - Performs the same functions as during normal operation

• Aux-Control Air - Performs no function for normal shutdown

• Electric Power - Performs the same functions as during normal operation

" RCW - Performs the same functions as during normal operation Dual Unit Trips

" ERCW system - Performs the same functions as during normal operation. A dual unit trip, such as a LOOP, does not change the success criteria for the ERCW system.

" CCS System - Performs the same functions as during normal operation, in the event of a dual unit trip, the CCS system has the capability of providing heat removal for all loads on the A train for each unit. The B train can be used for both units, provided that recirculation or shutdown cooling is not required for both units.

• Plant Control Air - In the event of a dual unit trip, the functions would be the same as normal operation. If the event were a LOOP, then station air compressors A and B could be started manually.

• Aux-Control Air - Performs no function in the event of a dual unit trip not related to losing all offsite power. In the event of a LOOP, the aux air compressors provide air to the essential loads within the auxiliary/reactor building. Normal test and maintenance terms are included within the model to address availability issues associated with these components.

• RCW - Provides cooling to the station air compressors and the secondary side. In the event of a dual unit trip due to a LOOP, the RCW system does not provide cooling as it cannot be energized from the shutdown boards.

Dual Unit Initiating Event Models Dual unit initiating events in the SQN PRA model are modeled as affecting both units. Thus, for a LOOP event, both units lose power. Therefore, the success criteria for the shared systems are always the most restrictive. The response to RAI 7.a.x discusses the modeling of dual unit initiators.

PRA System Modeling Each system within the PRA model, including the shared systems, was modeled individually to represent the system's success criteria. For the shared systems, the individual fault tree includes both Unit 1 and Unit 2 components with the appropriate normal operating alignments.

The individual fault trees were then combined and are incorporated in the PRA model for each unit. This allows components that are in a shared system to propagate correctly to either or El-11 of 87

both units. The only system that is shared between the two units that can be impacted by an outage is the B train of CCS. The unavailability of the system during an outage is accounted for within the fault tree by the use of a flag event. The CCS flag event fails the B train for the non-outage unit, if the outage unit is relying on the B train for cooldown, as stated above in the Normal Shutdown section. There is no other impact to unavailability due to one unit being in an outage.

For example, the CCS system has five pumps between both units. The 1A and 2A CCS pumps are normally aligned to the 1A and 2A trains respectively. CCS pump C-S is a common pump between both units and is normally aligned to the common B train of CCS. The 1 B CCS pump can be aligned to either the 1A train or the B train, depending on where it is needed. Similarly, the 2B CCS pump can be aligned to either the 2A train or the B train as needed. The CCS system model includes all three trains (1A, 2A, and B) and the shared dependencies between units. The system model was then integrated in both the Unit 1 and Unit 2 models, so the appropriate dependencies are accounted for on both units. Therefore, if the 1B pump is removed due to test or maintenance, it will affect both units due to the common B train dependency.

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NRC RAI 7.a.vii vii. It is stated that no changes were made to the Level 1 CAFTA Rev. 0 PRA to produce the SQN SAMA Level I model. The peer review is stated to have been performed on the January 14, 2011 version of the SQN CAFTA PRA, while the Rev. 0 model is given in Table E. 1-17 as having an issue date of June 3, 2011. Apparently, the peer review was performed on an earlier version of the CAFTA PRA.

1. Provide the CDFand large early release frequency (LERF) for the model that was peer reviewed.

2. Confirm that all of the changes described in the resolution of the peer review findings in Section E. 1.4.7 have been implemented in the Revision 0 model and a summary of the more significant changes to the peer reviewed model to make the Revision 0 model.

TVA Response 7.a.vii.l: The pre-peer review model had the following CDF and LERF values:

Unit CDF LERF Unit 1 6.50E-05 2.54E-06 Unit 2 6.25E-05 3.1OE-06 7.a.vii.2: The changes described in the resolution of the peer review findings in Table E.1.4.7 have been implemented in the Revision 0 model. After the peer review, the following changes were made to resolve specific Findings and Observations (F&Os).

• F&O 1-15 (667) was resolved by creating new core damage sequences within the PRA model. These sequences specifically account for failure of pressure relief in a transient.

This was conservatively modeled as leading to core damage, because adequate secondary side relief was not available to reduce the pressure transient. In addition, all LOOP recovery basic events were recalculated using an additional 30 minutes to perform the actions.

• F&O 2-3 (670) was resolved by performing Bayesian updating of the significant contributors to risk. Specifically, 11 type codes were updated with success data only (no failures) in the Revision 0 of the CAFTA PRA model.

• F&O 3-7 (676) was partially resolved by changes to the PRA model; specifically, item two required that the 480 gpm seal LOCA events were classified as small break LOCAs and item four required the addition of the cold leg accumulators to the success pathways for large LOCAs. The other areas of the F&O did not affect the model significantly.

The following changes were made during a structured review of the PRA model to assure technical adequacy of the PRA model:

• Recalculated the human error probabilities to account for dependency factors on recovery actions for both cognitive and execution steps.

• Incorporated crosstie logic to model the ability for one shutdown board to power another shutdown board. The logic was developed so that if one unit has lost all shutdown boards and the other unit has both shutdown boards available, power can be transferred over to the unpowered unit's shutdown board.

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• Revised the PRA model to account for ERCW success criteria refinement. Based on the flow balancing calculations, the ERCW success criterion is one pump per train (A and B) for both units (i.e., two out of eight pumps) as long as CS is not required.

• Added mutually exclusive logic to remove operator actions to trip the RCPs showing up in the same cutsets as LOOP.

Added mutually exclusive logic to prevent the block valve isolation events from showing up in the same cutsets, i.e., A and B block valve basic events. The model already has a basic event that simulates both block valves being closed.

Credited manual operation of the atmospheric relief valves (ARV) as a backup to the control room action. This manual action HAMARV was added to the long term heat removal gates (UI/2_LTHR).

Added mutually exclusive logic for test and maintenance of the CCS heat exchangers and the summer flag. Maintenance is prohibited during the "summer" season defined in the PRA model.

Refined internal flooding sequences, to limit the overly conservative effects of flooding events. The new EPRI standard for pipe rupture frequencies was issued and incorporated into the model. This new standard contains leak before break factors which are credited in the highest value flooding sequences.

Following incorporation of the changes, the Model of Record was issued with the following CDF and LERF values.

Unit CDF LERF Unit 1 3.02E-05 4.39E-06 Unit 2 3.59E-05 4.63E-06 NRC RAI 7.a.viii.1 viii. For the peer review results discussed in ER Section E. 1.4.7:

1. For Finding 1-10; what is the significance on the results of the SAMA analysis of the assumption that the operatoris successful in providing feedwater to the ruptured steam generator?

TVA Response Because the SAMA analysis does not credit radionuclide scrubbing for a ruptured SG, the operator action to provide water to the ruptured SG does not have an impact on the SAMA analysis.

NRC RAI 7.a.viii.2

2. For Finding 1-14, clarify if the data used for the Rev. 0 PRA has been updated to exclude the post maintenance test data. If not, what is the impact on the results?

TVA Response The post maintenance testing information was excluded from the success data credited as part of the data analysis calculation.

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NRC RAI 7.a.viii.3

3. For Finding 4-3, was the impact of spraying from glycol system failures included in the analysis?

TVA Response The glycol system was included within the PRA model as a potential flooding initiating event.

Glycol only exists as an initiating event in 759.0-AO1 and A03. As discussed in the peer review resolution, the glycol system is a limited volume system, i.e., it would not cause failures by submergence. The potential exists for the glycol system to spray down the Solid State Protection System (SSPS) cabinets in the control rod drive Motor Generator-set rooms.

However, these flood areas were both screened from inclusion within the PRA model because the entire contribution of the flooding area has a CDF less than 10-9 (IFQU-A3).

NRC RAI 7.a.viii.4

4. Finding 1-15 identifies certain deficiencies in the generaltransient event tree including lack of a separatetree for station blackout (SBO) events and not addressingthe operation of systems such as charging and auxiliary feedwater (AFW) following power recovery.

Provide more information on this finding and its resolution to address these two issues.

TVA Response The general transient event tree was modified to address the comments made in Finding 1-15 concerning the availability of pressure relief. All transients now require pressure relief. The failures of the pressure relief would lead to core damage if there is not enough AFW flow available.

The SBO event tree was not added to the model. As discussed in the F&O 1-15 resolution:

Evaluation of systems failures post power recovery was not included given that this would take significant modification of the model structure with no appreciable enhancement to the PRA insights. Adding this type of logic would produce few cutsets (if any) or change the importance of key modeled systems given that the LOOP cutsets that would be generated would be identical to more benign transients (such as manual scram) because the same systems would have to fail to arrive at core damage in the transient as in the LOOP event.

To estimate the effect of not including those additional failures, the SBO cutsets were modified to determine the frequency of successful offsite power recovery. The SBO cutsets were analyzed for their non-recovery probabilities. These values were then adjusted to one minus the non-recovery of offsite power probability to calculate the recovery probability. The cutsets were updated with these new recovery probabilities and the frequency of successful offsite power recovery was calculated. To estimate the effect of additional hardware failures, the conditional probability of a seal LOCA, which is the result of all SBO events, was then multiplied by the frequency of successful offsite power recovery following an SBO. The contribution for the additional hardware failures was estimated to be 2.53E-08, i.e., an increase in CDF risk of 0.084%. Therefore, the inclusion of the SBO system failures post power recovery has a negligible effect on the CDF or the SAMA analysis.

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NRC RAI 7.a.ix ix. Anticipated transient without scram (A TWS) sequences make up a relatively high and significant contribution to CDF and are dominant contributorsto offsite risk.

Discuss the reasons for this high frequency and potential additionalSAMAs to mitigate this risk.

TVA Response TVA PRA model was developed using the methodology outlined in WCAP-1 5831-P-A.

Consistent with that methodology, the dominant ATWS sequences have two particular types of sequences. The most common are ATWS events that are dominated by unfavorable exposure times (UET); the second are events with power dependencies.

The UET is a point in the operating cycle where the reactivity is great enough such that, given an ATWS situation, the amount of pressure relief available is not sufficient to prevent exceeding the design pressure for the primary system. The SAMA analysis assumes that there is a 0.1 probability of having the UET.

The second type of ATWS events are those where power dependencies are failed and those failures cause failures of the automatic reactor trip protection logic or failure of the power supplies leading to opening of the PORVs.

The UET, identified as FLG_UI/U2_UET, was already considered in the PRA model. The WCAP defines four actual intervals where the UET exists. The UET ATWS event sequences were modeled correctly.

Conversely, the power dependent ATWS events are modeled incorrectly. The majority of the cutsets involve a battery depletion flag, due to failure of the power to the battery chargers.

Because an ATWS event occurs early in the accident, the power supplies to the PORVs would be available as long as there is no fault with the batteries or the battery board.

To resolve this error, the following changes were made to the ATWS modeling in Revision 001 of the CAFTA PRA. In ATWS event tree, the FLG Ul UET was set to false because the WCAP analysis and the Revision 0 model already had pressure relief intervals for situations where there is always inadequate pressure relief. The basic event U1/2_PRCI1 (with a probability of 0.11) already accounts for the time frame during which pressure relief is inadequate.

The second change made to address ATWS situations was to correctly model the power dependencies for the reactor protection system. The loss of AC power to the shutdown boards would cause an automatic reactor trip; additionally, the DC solenoid on the reactor trip breakers would force open the breakers on a loss of DC power. The reactor trip breakers could still mechanically fail to open. However, the requirement for breaker control power to be available to open the reactor trip breakers is no longer considered. Therefore a majority of the previous ATWS initiating events, such as %0_ELBFLOOD, are no longer ATWS events due to the breakers opening on the loss of DC power.

Incorporating these changes into the new PRA model results in an ATWS contribution of 2% for Unit 1 and 2.3% for Unit 2. The dominant contributor to ATWS is now failure of the control rods to insert due to mechanical reasons (UlCRI and U2_CRI). This event was addressed by SAMA 277 and determined to not be cost beneficial. Two other SAMAs (136 and 137) which addressed the ability to trip the motor generator breakers and the ability to remove power from El- 16 of 87

the control rods were also evaluated and determined to be not cost beneficial. No additional SAMAs were identified.

NRC RAI 7.a.x

x. Discuss the modeling of dual unit initiatorssuch as loss of offsite power or failure of shared support systems.

TVA Response The SQN PRA model was developed as a dual unit model. All shared systems are modeled with the most restrictive success criteria based on a dual unit initiating event. The following discusses the modeling of the various shared systems:

ERCW The ERCW system provides the ultimate heat sink. LOOP initiating events affect the success criteria for this system. The ERCW pumps that can be supplied from one shutdown board are restricted to one pump when no offsite power is available (to prevent DG overloading). In the ERCW fault tree, only one pump can be selected to "restart" following a LOOP.

The ERCW system has three dual unit initiating events, a total loss of the ERCW system, loss of the A train of ERCW, and loss of the ERCW 2A-A header. Each of these events affects portions of the ERCW system and is addressed via the support system dependencies built into the fault tree models. For example, the loss of the 2A-A header initiating event only affects the availability of that header; the Unit 1 system would still have both the ERCW headers (1A-A and 1 B-B), and Unit 2 would still have the availability of the 2B-B ERCW header.

Station and Control Air The station and control air compressors provide air to non-essential loads. The loss of this system would make all non-essential components fail to their loss of air position. This effect requires the auxiliary air compressors to activate and accept the essential air loads, specifically to the AFW LCVs. If a loss of station air event were to occur, the station and control air compressor headers would isolate from the auxiliary air compressors. This effect is captured by the fault tree logic and dependencies.

CCS The only shared system that is uniquely modeled is the CCS system. The CCS system is a "pseudo" shared system where the A train of CCS is unit specific, but the B train is shared.

Because only the B train is shared, the failure of this system would be the only event that would be "shared." However, the loss of the B train of CCS does not require a shutdown, so while one unit could have a total loss of CCS, the other unit would still have the A train available and not require a shutdown.

LOOP The PRA model treats all LOOP initiating events as being dual unit events. The current power lineup requires the 161kV switchyard to power the plant loads. After the LOOP initiating event occurs, the support systems' success criteria are based on the design basis for each system.

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NRC Question 7.a.xi xA Forboth units, the internalflooding initiatorcontributedmore than half of the CDF. Discuss any influence of the occurrence of internalflooding at one unit with the PRA modeling at the other unit.

TVA Response The internal flooding PRA model was developed by using the Internal Flooding database which was then fed into the EPRI FRANX code. This database allowed for sequences that would be dual unit initiating events so that each unit's FRANX file would be identical.

For example, a flooding event from the ERCW supply header 2A-A in the Unit 2 Turbine Driven AFW pump (TDAFWP) room would be a dual unit initiating event. The flood would also affect the TDAFWP for Unit 2 at the start of the flooding event. In this example the Unit 1 and Unit 2 FRANX files would both have a sequence created with identical effects. This assures that the effects on the units are evaluated for each flooding sequence.

In addition, there are situations where a reactor trip on one unit is not necessary unless there is a subsequent failure to isolate. Ifthe B ERCW discharge header were to rupture Unit 1 would be required to trip the reactor, while Unit 2 would not be required to trip the reactor. If the flood was not isolated in time, the RHR and CS pumps would become submerged requiring a Tech Spec 3.0.3 shutdown. These sequences were conservatively assumed to be a reactor trip on both Units at the start of the event.

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NRC RAI 7.b.i

b. Provide the following information relative to the Level 2 analysis. Basis: Applicants for license renewal are required by 10 CFR 51.53(c)(3)(ii)(L) to consider SAMAs if not previously considered in an environmentalimpact assessment, related supplement, or environmental assessment for the plant. As part of its review of the SQN SAMA analysis, NRC staff evaluates the applicant'streatment of accident propagationand radionucliderelease in the Level 2 PRA model. The requested information is needed in order for the NRC staff to reach a conclusion on the adequacy of the applicant'sLevel 2 PRA model for supporting the SAMA evaluation.

i. Provide a brief history of the current SQN Level 2 PRA model development.

TVA Response The SQN Level 2 model used the Watts Bar Nuclear Plant (WBN) Level 2 model developed by Westinghouse as its starting point. This model utilized guidance contained in NUREG/CR-6595

("An Approach for Estimating the Frequencies of Various Containment Failure Modes and Bypass Events") and WCAP-16657-P ("SAMG Template for Level 2 PRA") in its development.

This model was developed from the ground up when the PRA model was converted from RISKMAN into CAFTA format.

The containment event trees (CET) developed for the WBN model consisted of one CET for SBO sequences and one CET for non-SBO sequences. The primary difference between the SBO and non-SBO CETs was that in the SBO CET, the containment air return fans (ARF) and hydrogen igniters are treated as having failed. The SQN CETs retain the logical branch structure of the original. WBN trees with the following exceptions:

1. The logic branches regarding ARF failure and hydrogen igniter failure were removed from the SBO CETs because this equipment was always treated as having failed for SBO sequences, and retaining them would result in several branches that produced no results for SBO.

2. The SQN fault tree logic explicitly includes the negation of the success path branches of the CETs when solving the fault tree logic for each sequence, which was not the case in the WBN Level 2 model.

3. The Hydrogen Mitigation System consists of two groups of 34 igniters distributed throughout the containment. Each group of igniters receives its power from the Class 1 E (essential) AC auxiliary power system. One group of igniters receives its power from the train A diesel and the other group from the train B diesel.

4. SQN has two black-out DGs (2 MWe each) which can be used to power one group of hydrogen igniters on each unit in the event of an SBO.

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NRC RAI 7.b.ii ii. In several places in the ER, the accuracy of the Level 2 model is discussed.

For example, in Section E. 1.2.1 it is stated:

"However,the quantificationof the non-LERF end states is not as accurate as would be obtained from a rigorousLevel 2 model."

and "The event tree nodes and split fractions were reviewed to ensure that the consequences, in terms of release frequencies, would be largerthan would be expected with a fully developed Level 2 model."

and in Section E.1.2.3.2 "Quantificationof the SQN SAMA Model results in release frequencies that are over predicted..."

Provide a discussion of the bases for these statements and the reasons for the inaccuraciesas well as steps taken to insure that any inaccuraciesdo not adversely impact the SAMA identificationor cost-benefit analysis process.

TVA Response As discussed in Sections E.1.2.1 and E.1.2.3.2, it was determined that the SQN Level 2 Model of Record overestimated the release category frequency results. Because the original focus for the development of the Level 2 model was LERF, the remaining release end states had not been quantified and reviewed in detail. Therefore, modifications were made to the existing Level 2 model in order to improve the quantification of the other end-states. This included incorporation of sequence success logic for each Level 2 sequence directly into the Level 2 fault tree by adding the negation of the success branches of each CET sequence. In addition, it was determined that the original logic did not correctly model the failure of containment isolation for SBO sequences because SBO conditions did not result in large isolation releases as intended.

A minor fault tree revision corrected the logic. The models used for the SAMA analysis includes these modifications and corrections.

The inclusion of these modifications resulted in a significant reduction in the overestimation of the frequencies, and the sum of all release category frequencies, excluding the intact category, is slightly less than the total CDF. However, the response to RAI 7.b.iii indicates that when the intact frequency is included, the total release frequency is between 43% and 46% higher than the CDF.

The total release frequency estimate is greater than the total CDF due to the treatment of the success branches within the event trees. When there are cutsets with high-probability events (i.e., 0.9 or above), as there are in many of the success branches of the CETs, the results are overestimated due to the fact that the FTREX/CAFTA cutset quantification software uses the Tmin cut upper bound" approximation when solving the cutsets. This provides a close approximation to the top event probability when the probabilities of basic events are small, but provides a conservative (higher) result when the basic event probabilities are large. Such is the case with several of the probabilities in the CET branches, so the overall release frequencies are overestimated. A sensitivity was run in which the CET branches with probabilities greater than 0.9 were set to "true" rather than their original values during quantification. The sensitivity study resulted in an overall release frequency (which includes the Intact, SERF, LERF and the late release end states) that reduced the over-estimation of the release states by 7%. The results of the sensitivity study showed Unit 1 total release frequency calculation was reduced El- 20 of 87

7% from the original value; the Unit 2 total release frequency calculation was reduced 5% from the original value.

NRC RAI 7.b.iii iii. What is the frequency for the intact containment event tree (CET) end state?

TVA Response The intact Level 2 end states were calculated using a truncation of 1 E-1 2 for Unit 1 and Unit 2.

CDF Intact  % of CDF Frequency Unit 1 3.03E-05 1.45E-05 48%

Unit 2 3.59E-05 2.38E-05 66%

Note that the intact containment end state and the other end states are overestimated. The frequencies for the intact CET estimate, the SERF, LERF and the late end states, when added together, are greater than the total CDF due to the treatment of the success branches within the event trees. When there are cutsets with high-probability events (i.e., 0.9 or above), as in many of the success branches of the CETs, the results are overestimated due to the fact that the FTREX/CAFTA cutset quantification software uses the "min cut upper bound" approximation when solving the cutsets. This provides a close approximation to the top event probability when the probabilities of basic events are small, but provides a conservative (higher) result when the basic event probabilities are large. Such is the case with several of the probabilities in the CET branches, so the overall release frequencies are overestimated. A sensitivity was run in which the CET branches with probabilities greater than 0.9 were set to "true" rather than their original values during quantification. The sensitivity study resulted in an overall release frequency (which includes the Intact, SERF, LERF and the late release end states) that reduced the over-estimation of the release states by 7%.

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NRC RAI 7.b.iv iv. While SEQSOR was used to determine the fission product release fractions, a phenomenologicalbased code such as MAAP is necessary for other inputs, such as available operatoraction time windows and the evaluation of accident progressionto determine containment failure probabilitiesand release timing.

Briefly describe the scope of these additionalanalysis performed for the SQN PRA and identify the code or codes used including the version.

TVA Response To support the Level 2 model development, the SQN PRA model uses MAAP 4.0.7. MAAP was used in the Level 2 calculation to provide the following:

" Calculated time to Vessel failure

• Ex-vessel cooling success

• Seal table - molten core interaction

• Uncertainty associated with the availability of the ice condenser

• Modeling of the availability of the containment ARFs

• Core damage stopping prior to vessel failure

• Time to hydrogen detonation

• Hydrogen concentrations

" Direct containment heating

• Timing of early containment vessel failure

• Effectiveness of containment heat removal

• Base mat melt through timing

• Timing of operator action timing El- 22 of 87

NRC RAI 7.b.v

v. The discussion of the BLERF end state in ER Section E.1.2.1 states "A bypass release does not have an opportunity to undergo scrubbing within the containment. However, the SGTR tube rupture cases may have an opportunity for scrubbing."

The steam generator tube rupture (SGTR) initiating event does not appearin the Level II importance analysis results.

Discuss further the reasonsfor the apparentlack of importance of SG TR initiatorsand treatment of SGTRs (both those as an initiatingevent and those induced) including the release categorieswhich incorporateboth scrubbed and unscrubbed SGTRs.

TVA Response The level 1 event tree for SGTRs was developed using a consensus model for such events developed by Westinghouse (i.e., WCAP 15955). This accident sequence was then modified by crediting additional plant specific considerations that applied to SQN. The change that resulted in the biggest difference was the crediting of manual actions to depressurize/cooldown the primary side. The generic analysis does not credit the use of manual hand wheels to open the SG atmospheric relief valves. However, the SQN model does credit these valves due to the manual actions, which allows for greater redundancy in the de-pressurize/cooldown node of the event tree. Updated data from NUREG/CR-6928 was used for the SGTR initiating event as well, which reduces the frequency.

The SGTR initiating event is modeled as four individual initiators, i.e., one for each SG. These events do not appear in the Level I or Level 2 importance measures included in the SAMA risk significant terms tables because the initiators did not meet the review criteria of 1.005 for significance nor did they exceed the equivalent Risk Reduction Worth (RRW) of a $50,000 simple procedure change when uncertainty is considered (-1.003).

For Unit 1, each of these initiators has an RRW of 1.0002 for the Level 1 model. The corresponding Level 2 (LERF) RRW is 1.001.

The pressure induced SGTR event RRW is 1.0019 which does not meet the review criteria of significance. The thermal induced rupture event (TISGTRNOSBO U1_L2) with an RRW of 1.032 meets the review criteria for significance and was included in the Level 2 list of important events.

These events are addressed as part of Release Category Ill. The SEQSOR evaluation of Release Category III assumed no scrubbing of releases in the SG.

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NRC RAI 7.b.vi vi. The subcategory definitions given in ER Tables E. 1-13 and E. 1-14 appearto exclude some combinations of the listed characteristicssuch as SBO versions of Release Categories(RCs) I and Ill and transientcontributors similar to RC Va.

Provide a discussion of the development of the release categoriesand of the logic used to insure that all containment failure CET endpoints are incorporated.

TVA Response The subcategory definitions are provided as a description of the dominant characteristics of the binned sequences. The subcategories were created as a means of grouping releases which have similar characteristics that are important for characterizing the releases (e.g., timing of release, status of containment, RPV pressure). The results of the release category quantification helped define the categorization.

For example, one could reasonably expect that there could be an SBO version of RC I_b, but no SBO releases similar to RC Ib resulted from the quantification. As discussed in the response to RAI 7.b.ii, SBO sequences result in a large containment isolation failure. Therefore, all SBO sequences are evaluated as isolation failure end states (RC IIa and II_b) and there are no SBO sequences in release categories RC I_a, I_b, and I_c. While this is the case for SBO sequences, there may be some minor contribution of transient events in release categories described as LOCA categories and minor LOCA contribution in transient categories.

Each CET failure endpoint is assigned to one of the major release categories (i.e., LERF, ILERF) and all are incorporated, except for the intact category (see response to RAI 7.b.ii).

Overall, the categorization utilized is reasonable and provides results that would not be significantly different or less conservative than one with more categories which specifically addressed the minor contributing sequences.

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NRC RAI 7.b.vii vii. Provide a discussion of the representativeaccident scenarios used for the determinationof the release characteristicsfor each of the release categories including:

1. A description of each scenario

2. The bases for the selection of the representativescenarios

3. Steps taken to insure that the benefit of a SAMA is not underestimated for situations in which a SAMA impacts scenariosthat could have a lower (non-dominant)frequency but significantly larger consequence than that for the representativescenario TVA Response Table 7.b.vii below provides a listing of the release categories and their associated Level 2 sequences.

The representative accident scenarios were selected based on the definitions of the release categories and on their frequency contribution to those release categories. In order to ensure that the effects of SAMAs were not underestimated, the input parameters for SEQSOR were conservatively selected for each release category. The other release characteristics (e.g., time of release, warning time, release energy) that were input to MACCS were also conservatively selected. While this resulted in conservative consequences, it reduced the possibility that a low frequency scenario with significantly larger consequences would be under represented in a SAMA evaluation. Each of the release categories are individually addressed below.

RCIa Sequence HLERF-001 is the Level 2 sequence for this category. The Level 1 sequence is a small LOCA with successful cold leg injection via the CVCS pumps, AFW initially available, and failure to transfer to high pressure recirculation followed by the failure of operators to depressurize for low pressure injection. The containment accident progression includes AC power available (no SBO), failure to depressurize vessel, successful air recirculation fans (ARF) and successful igniter operation. The containment fails early following vessel breach which occurs at high pressure.

RCIb Sequence LLERF-020 is the dominant Level 2 sequence for this category. The dominant Level 1 sequence is a small LOCA with successful cold leg injection via the CVCS pumps, AFW initially available, failure to transfer to high pressure recirculation, and successful RCS depressurization followed by failure of low pressure recirculation via RHR. The containment accident progression includes AC power available, RCS at low pressure, successful ARF and failed igniters. The containment fails early following vessel breach which occurs at low pressure. The small LOCA sequence represents 93% of the total release frequency for this release category. Level 2 sequence LLERF-020 is -80% of the total frequency while LLERF-019 is -16%. The only difference in the accident progression of the two is that the igniters are successful in LLERF-019 and containment fails due to an ex-vessel steam explosion.

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RCIc Sequence LLERF-014 is the dominant Level 2 sequence for this category. The dominant Level 1 sequence is a general transient scenario with successful reactor trip, successful operation of PORVs, no loss of RCP seal cooling, failure of decay heat removal via SGs and successful implementation of bleed and feed cooling with charging pumps followed by failure of high pressure recirculation which leads to core damage. The containment accident progression includes AC power available, AFW is not available, successful depression of RCS, successful ARF and failure of igniters. The containment fails early following vessel breach which occurs at low pressure. The dominant Level 1 sequence represents -80% of the release frequency. A similar transient Level 1 sequence accounts for -18% of the release frequency. The only difference between the two is that implementation of bleed and feed fails in the second (thereby making high pressure recirculation irrelevant), so these sequences are very similar. Level 2 sequence LLERF-014 accounts for -83% of the release frequency and LLERF-013 is -17%.

The only difference between these two sequences is that the igniters are successful in LLERF-013 and containment fails due to ex-vessel steam explosion.

RC II a and RC II b These two release categories are developed from Level 2 sequence ILERF-002. The basic characteristics of this Level 2 sequence are an SBO and a large isolation failure. The failure to isolate occurs because the SBO results in loss of power to the reactor coolant pump seal return line containment isolation valves. Because the valves are motor operated, they remain open.

Approximately 85% of the SBO contribution is from internal flooding initiators and remaining is primarily from LOSP initiators with failure of the on-site power systems. It should be noted that SQN has a procedure to manually isolate the reactor coolant pump seal return line penetration.

However, this procedure was not credited in the PRA. If this procedure were credited, the Maximum Averted Cost Risk would be reduced and some SAMAs (e.g., those associated with internal flooding events which cause an SBO) that are currently potentially cost beneficial may no longer be cost beneficial. Crediting the new RCP seals in the PRA model would have a similar effect and would likely reduce the overall importance of the procedure from a risk perspective but they would not eliminate the need for manual isolation of the seal return line.

The results of the sequence quantification are divided into two subcategories based on the two types of Level 1 scenarios in the results, ATWS and LOCA. The RC II a category is characterized as an ATWS event followed by successful automatic rod insertion, turbine trip and failure of AFW. The reactor is at high pressure at the time of core damage. This represents

-84% of the ILERF-002 release frequency. The RC II b category is characterized as a very small LOCA followed by inability to bleed and feed (SBO) and failure of AFW. The RPV is at high pressure at the time of core damage. This group represents -16% of the release frequency.

RC II c and RC II d These two release categories are developed from Level 2 sequence ILERF-001. The basic characteristics of the Level 2 sequence is AC power available (i.e., no SBO) and a large isolation failure occurs. The failure to isolate occurs because of loss of power or control circuits to the reactor coolant pump seal return line containment isolation valves. The loss of power to El- 26 of 87

the valves typically occurs due to a transient initiated scenario which includes electrical component hardware failures. The results of the sequence quantification are divided into two subcategories based on the two dominant types of Level 1 scenarios in the results, LOCA and transients. The RC II c category is characterized as a small LOCA with successful cold leg injection via the CVCS pumps, the AFW is initially available, failure to transfer to high pressure recirculation, and successful RCS depressurization followed by failure of low pressure recirculation via RHR. This represents - 58% of the ILERF-001 release frequency. The RC II d category is characterized as a general transient scenario with successful reactor trip, successful operation of PORVs, no loss of RCP seal cooling, stuck open PORV and failure of the AFW.

This represents - 42% of the release frequency.

RC III Sequence BLERF-003 is the dominant Level 2 sequence for this category. The basic characteristics of this Level 2 sequence are AC power is available and there is a large bypass of the containment. The dominant Level 1 sequence involves the occurrence of an ATWS during UET core conditions. Core damage is assumed under these conditions as is containment failure. Level 2 sequence BLERF-003 represents -70% of the release frequency and the most challenging scenario for containment performance. Overall, ATWS scenarios are -62% of RC III and SGTRs, both initiators and induced tube ruptures, are 34%. The remaining frequency is due to interfacing system LOCA initiators.

RC IV a Sequence LATE-042 is the dominant Level 2 sequence for this category. The dominant Level 1 scenario is a small LOCA with successful cold leg injection via the CVCS pumps, AFW initially available, failure to transfer to high pressure recirculation, and successful RCS depressurization followed by failure of low pressure recirculation via RHR. The containment accident progression includes AC power available, RCS at low pressure, successful ARF and successful igniters.

The containment fails late because of the failure of containment heat removal. The small LOCA sequence accounts for over 95% of the category's total release frequency. Level 2 sequence LATE-042 is 80% of the total release frequency while LATE-044 is -20%. The only difference in the accident progression of the two is that the igniters fail in LATE-044 but the resulting detonation is not large enough to fail containment. As with LATE-042, containment failure in LATE-044 is due to containment heat removal failure.

RC IV b Sequence LATE-034 is the Level 2 sequence for this category. The dominant Level 1 sequence is a general transient scenario with a successful reactor trip, successful operation of PORVs, no loss of RCP seal cooling, failure of decay heat removal via SGs and successful implementation of bleed and feed cooling with charging pumps followed by failure of high pressure recirculation which leads to core damage. The containment accident progression includes AC power available, AFW is not available, successful depression of RCS, successful ARF and success of igniters. The containment fails due to failure of containment heat removal.

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RC Va and RC V b These two release categories are developed from Level 2 sequence SERF-001. The basic characteristics of this Level 2 sequence are AC power is available and a small isolation failure occurs. The results of the sequence quantification are divided into two subcategories based on the two dominant types of Level 1 scenarios in the results, LOCA and transients. The RC V a category is characterized by a small LOCA with successful cold leg injection via the CVCS pumps, AFW initially available, failure to transfer to high pressure recirculation, and successful RCS depressurization followed by failure of low pressure recirculation via RHR. The containment failure mode is by small containment bypass. This represents -65% of the SERF-001 release frequency. The remainder of the release frequency is assigned to RC V b and characterized as a transient scenario with successful reactor trip, successful operation of PORVs, no loss of RCP seal cooling, failure of decay heat removal via SGs and successful implementation of bleed and feed cooling with charging pumps followed by failure of high pressure recirculation which leads to core damage. Core damage and vessel failure occur at high pressure. The containment failure mode is by small containment bypass.

Table 7.b.vii, U1 Final Release Categories and Frequencies Release Sequence Category Dominant Level 2 Release Category Sequences Frequency Frequency I a - LER; RPV Hi Pressure, LOCA, non-SBO HLERF-001 4.06E-08 4.06E-08 LLERF-001 2.04E-08 LLERF-002 1.84E-08 I b - LER; RPV Lo Pressure, LOCA, non-SBO LLERF-01 9 1 .59E-07 9.70E-07 LLERF-020 7.72E-07 LLERF-013 4.59E-08 I c - LER; RPV Lo Pressure, non-SBO Transient 2.65E-07 LLERF-014 2.19E-07 II a - Isolation LER; SBO, ATWS ILERF-002 (83.83%) 3.26E-06 3.26E-06 l1b- Isolation LER; SBO, LOCA ILERF-002 (16.17%) 6.29E-07 6.29E-07 IIc- Isolation LER; non-SBO, LOCA ILERF-001 (57.6%) 6.47E-08 6.47E-08 II d - Isolation LER; non-SBO, Transient ILERF-001 (42.4%) 4.77E-08 4.77E-08 BLERF-001 1.83E-07 III - Bypass LER; non-SBO BLERF-002 1.1OE-08 6.43E-07 BLERF-003 4.48E-07 LATE-042 1.43E-05 IV a - Late Release; non-SBO, LOCA LATE-044 3.65E-06 1.79E-05 IV b - Late Release; non-SBO Transient LATE-034 2.23E-06 2.23E-06 V a - SER; Lo Pressure, LOCA SERF-001 (64.7%) 2.08E-06 2.08E-06 V b-SER; High Pressure, LOCA/Transient SERF-001 (35.3%) 1.13E-06 1.13E-06 El- 28 of 87

NRC RAI 7.b.v.i, viii. Provide a listing of the characteristicsused to describe each representative scenario for each RC as input to the SEQSOR emulator.

TVA Response SQN used the same SEQSOR emulator that WBN-2 used for its SAMA analysis. The inputs used for WBN-2 were also used as an example for identifying the appropriate scenario characteristics and inputs for the SQN analysis.

The accident characteristics and number of options available for input to SEQSOR are listed below.

• Early releases from fuel to RCS (low or high zirconium oxidation);

• Fraction of core released from RCS before vessel breach (six conditions-primarily related to reactor pressure and type of bypass);

" Fraction of containment contents released in the early and late phase (nine containment failure modes);

• Containment state (wet or dry);

" Sprays during early phase (three conditions-related to timing of containment failure in relation to vessel breach and RCS pressure at time of vessel breach);

" Number of holes in the RCS if vessel failure (one or two);

" DCH release (two conditions-related to vessel pressure);

" Core-concrete interaction pool scrubbing (two conditions-related to volume of water in containment cavity);

• SGTR scrubbing (no credit was taken for scrubbing in the SG);

• Ice condenser decontamination factor for early phase (three conditions related to ARF status);

• Ice condenser decontamination factor for late phase (four conditions related to ARF status);

" Fraction of core mass ejected from high pressure mass ejections (four conditions-related to magnitude of mass ejected) ;

• Early phase ice condenser bypass (three conditions-related to magnitude of bypass);

and

• Late phase ice condenser bypass (three conditions-related to magnitude of bypass)

Table 7.b.v.iii provides a summary tabulation of the SEQSOR inputs.

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Table 7.b.v.iii, SEQSOR Emulator Inputs SEQSOR Parameter I. Large Early II. Isolation Large Early III. Bypass IV. Late V. Small Early la lb Ic Ila lb lIc lid Ill IVa IVb Va Vb Accident non-SBO non-SBO non-SBO SBO SBO non-SBO non-SBO non-SBO non-SBO non-SBO non-SBO Characteristics' LOCA LOCA TRANS ATWS LOCA LOCA TRANS TRANS LOCA TRANS LOCA non-SBO Percentage Contributiona 3% 76% 21% 81% 16% 2% 1% 100% 89% 11% 65.00% 35%

RCS Pressure (Hi or Hi Lo Lo Hi Hi Lo Hi Event V Lo Lo Lo Hi Lo) I I I ARFS Success or S S F F Avail Avail N/A S S S S Failed (S or F) _ I I Zirconium Oxidation Low RCS Preeassue at________ ____

RCS Pressure at 100% Hi 100% Lo 100% Lo 100% Hi 100% Hi 100% Lo 100% Hi 100% Hi RCS ReleaseIIIII 100% Lo 100% Lo 100% Lo 100% Hi SGTR Secondary Reclosed SRV's Reclosing Containment Failure 100% 100% 100% 100% 100% 100% 100% 100% Event 100% Late 100% Late 25x 25x Design Mode Rupture Rupture Rupture Event V Event V Event V Event V V Rupture Rupture Design Leak Rate Lower Lower Lower Leak Rate 2 2 Conaimet Sat1Wt 100% 100% 100% Wet 100% ry 3 100% 100% Wet 2 100% Dry2 100% Dry3 Wet 2

Wet 2

. ru 100% Dry2 100% Wet2 100% Dry3 100% D100 Sprays for RCS CF at VB Other Other 100% 100% Other CF at VB Other Other Other Other Release 4 HP 100% w/Sprays w/Sprays None None with HP 100% 100% Nne w/Sprays w/Sprays w/Sprays w/Sprays Sprays Sprays Sprays Holes in RCS 2 Ice-Condenser 100% 100% 100% 100% 100%

Decontamination Fans priar F 100% 100% No 100% No Fans Fans 100% No 100% Fans 100% Fans Fans i 100% Fans Iacondeminseras 100% 10F prior to CF a Cr ao Fans Fans prior to prior to Fans prior to CF prior to to Fpri prior to CF Factor, RCS Release taoCF CFp a ta CF CF . CF ta CF Ice Condenser 100% Nne Bypass (RCS)

Ice-Condenser DeConden 100%

100% 100% Fans 100%

100% 100% No 100%

10% 100%

10 1one 100%

100% No Fans Fans 100% No 100% Fans 100% Fans 10%p. 100% Fans Factor, CCi Release ta CF prior to CF ta CF Fans Fans prior to prior to Fans prior to CF prior to CF ta CF prior to CF Factor, ____ReleasetoCFto CF CF CF I _to CF Fraction of Mass Law, Medium, High (.33,.33,.33)

Ejected from Core Low,_Medium,_High (.33,_.33,_.33)

Core-Concrete 100% 100% 100% 100% 100% 100% 100%

Interaction 100% Full 100% Full Empty Empty Full Full Empty 100% Empty 100% Full Empty 100% Full Empty Decontamination Cavity Cavity Cavity Cavity Cavity Cavity Cavity Cavity Cavity Cavity Cavity Cavity Ice Condenser Nane Bypass (CCI)

Notes 1 - Items listed in italics are characteristics of the release categories, but are not actual inputs to SEQSOR.

2 - LOCA sequences assumed to be wet.

3 -Transient sequences assumed to be dry.

4 - "Other" refers to conditions other than containment failure with vessel at high pressure.

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NRC RAI 7.b.ix ix. Provide a discussion of the steps taken to ensure the technical adequacy of the SQN Level 2 model and analysis.

TVA Response The changes made to the Level 2 model for the SAMA analysis were documented in an Enercon calculation performed in accordance with Enercon procedures. This calculation was subjected to an internal review and a separate peer review by an individual with extensive SAMA and SAMA Level 2 experience prior to its submittal to TVA for review. All comments were incorporated prior to the final approval of the calculation. The changes were added to the TVA model change tracking program and subsequently were incorporated into the SQN Level 2 Model of Record in accordance with TVA procedures.

NRC RAI 7.b.x

x. Provide an expanded discussion of the mapping of Level 1 sequences to the Level 2 analysis.

Describe how the plant damage state (PDS) bins are utilized in the Level 2 analysis.

TVA Response The Level 1 to Level 2 model logic was performed using the existing WBN PRA model as a starting point.

The Level 2 binning criteria used was the same as that used for the WBN model, although the individual accident sequences included in each bin may differ. The steps used to develop the Level 2 model were as follows:

• Develop Level 1 Event Trees with plant damage states (PDS) assigned to the core damage (CD) sequences. These PDS reflect the containment integrity (bypassed or not bypassed), pressure of the reactor vessel (high or low), and the status of the SGs (wet or dry).

" Develop Level 2 Event Trees. Two unique CETs were developed; SBO and non-SBO.

Each end state of the CETs are classified based on the release category (i.e., LERF, SERF, Late, Intact)

" Each Level 1 sequence that resulted in core damage was assigned to one of five Level 2 bins based on the PDS. All level 1 CD sequences are directly linked into the level 2 CETs and the PDS bins were used at the various branch points of the CETs to screen out sequences not applicable to that particular branch. The level 1 CD sequences are exactly the same as in the Level 1 PRA model. Appendix E, Section E. 1.2.2 of the Environmental Report, provides tables of the relationship between the Level 2 bins and the PDS bins and the Level 1 CD sequences associated with each PDS.

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NRC RAI 7.b.xi xi It is noted that ER Section E. 1.2.3.2 indicates that RC V includes small containment isolation failures and that Tables E.1-13 and E.1-14 indicate the frequency of this RC makes up approximately 10% of the total frequency of all RCs and is largerthan the frequency of RC I. Because it is expected that a small isolation failure would not prevent large early containment failure due to early failure causes such as hydrogen detonation or direct containment heating, describe the Level 2 modeling of small isolation failures to show that the potential for large early releases is properly considered for small isolation failure sequences.

TVA Response Release Category V in the SQN Level 2 PRA represents small early releases (SERFs). It was noted that there is a potential for small containment isolation failures to result in large early containment failures due to hydrogen detonation or direct containment heating. A sensitivity analysis was performed to determine the effect of this potential on the SAMA results. The analysis redistributed the small early release accident sequences (RC V) into the remaining release categories with the exception of large early releases due to containment failure (RC II).

The small early release frequency was proportionally redistributed to release categories I (LERF), III (Bypass) and IV (Late) based on their relative magnitudes for each SAMA. This resulted in a SERF of zero, no change to release category II, and an increase in the frequency of release categories 1,111 and IV.

Unit I Results A number of Phase II SAMA candidates were found to be potentially cost-beneficial for mitigating the consequences of a severe accident at SQN Unit 1 during the original SAMA analysis. Because an increase in cost benefit will not change the original conclusions for these SAMA candidates, the effects of the SERF redistribution analysis are not presented here. It should be noted that SAMA 087, which was previously identified as Not Cost-Beneficial in the original submittal, has since been identified as Potentially Cost-Beneficial. Therefore, SAMA 087 is not included in this sensitivity.

The SERF redistribution analysis resulted in an increase to the assessed benefit of less than 15% for seventeen SAMA candidates. The revised benefit of these seventeen candidates remained below the estimated cost of implementation. For example, the assessed benefit of SAMA 277 increased from $156,357 to $177,595 (14%).

Of the twelve SAMA candidates with an increased benefit of greater than 15%, six SAMAs have an assessed benefit of less than $35,000 (some significantly less). Therefore, a small increase in benefit resulted in a large percentage increase due to the relatively low initial benefit. The "not cost-beneficial" status of these SAMAs did not change due to this analysis.

The remaining six SAMA candidates with an increased benefit of greater than 15% are summarized in the table below. After the redistribution of the SERF frequency, these six candidates remain not cost-beneficial because the averted cost risk when considering the 9 5 th percentile uncertainty is below the cost of implementation.

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Unit I SAMA Original Revised Benefit 95t" Percentile Cost of Benefit due to SERF Uncertainty Implementation Reallocation Sensitivity SAMA 032 - Add the ability to automatically align emergency core cooling $457,637 $563,148 $1,407,870 $2,100,000 system to recirculation mode upon RWST depletion.

SAMA 188 - Implement modifications to the compressed air system to $466,910 $559,936 $1,399,840 $2,782,200 increase the capacity of the system.

SAMA 276 - Replace one or more existing SG ARVs $49,275 $61,149 $152,873 $1,223,000 with a Valve of different design or manufacturer SAMA 278 - Improve the reliability of the RHR pumps and improve maintenance $105,811 $130,169 $325,423 $345,095 procedures to reduce potential for common cause failure.

SAMA 284 - Reduce the probability that the pressurizer safety relief $88,946 $110,183 $275,458 $1,566,800 valves fail to reclose following a water pressure relief event.

SAMA 287 - Protect, re-route, or modify circuits to upgrade core damage $397,595 $475,641 $1,189,103 $2,000,000 mitigation capability for fires that result in MCR evacuation.

Unit 2 Results A number of Phase II SAMA candidates were found to be potentially cost-beneficial for mitigating the consequences of a severe accident at SQN Unit 2 during the original SAMA analysis. Because an increase in cost benefit will not change the original conclusions for these SAMA candidates, the effects of the SERF redistribution analysis are not presented here. It should be noted that SAMA 087, which was previously identified as Not Cost-Beneficial in the original submittal, has since been identified as Potentially Cost-Beneficial. Therefore, SAMA 087 is not included in this sensitivity.

The SERF redistribution analysis resulted in an increase to the assessed benefit for seventeen SAMA candidates that is less than or equal to 15% of the original benefit. The revised benefit of these seventeen candidates remained below the estimated cost of implementation. For example, the assessed benefit of SAMA 103 increased from $397,065 to $445,756 (12%).

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Of the eleven SAMA candidates with an increased benefit of greater than 15%, six SAMAs have an assessed benefit of less than $10,000. Therefore, a small increase in benefit resulted in a large percentage increase due to the relatively low initial benefit. The "not cost-beneficial" status of these SAMAs did not change due to this analysis.

The remaining five SAMA candidates with an increased benefit of greater than 15% are summarized in the table below. The cost of implementation for these candidates is higher than the benefits when considering the 9 5 th percentile uncertainty, so these five remain not cost-beneficial.

Unit 2 SAMA Original Revised Benefit 95th Percentile Cost of Benefit due to SERF Uncertainty Implementation Reallocation Sensitivity SAMA 068 - Add a motor- $1,303,463 $1,564,829 $3,912,073 $10,000,000 driven feedwater pump.

SAMA 188 - Implement modifications to the compressed air system to $424,175 $514,841 $1,287,103 $2,782,200 increase the capacity of the system.

SAMA 276 - Replace one or more existing SG ARVs $56,157 $71,267 $178,168 $1,223,000 with a valve of different design or manufacturer SAMA 284 - Reduce the probability that the pressurizer safety relief $60,875 $75,986 $189,965 $1,566,800 valves fail to reclose following a water pressure relief event.

SAMA 287 - Protect, re-route, or modify circuits to upgrade core damage $308,398 $600,351 $1,500,878 $2,000,000 mitigation capability for fires that result in MCR evacuation.

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NRC RAI 7.c.i

c. Provide the following information with regard to the treatment and inclusion of external events in the SAMA analysis. Basis: Applicants for license renewal are requiredby 10 CFR 51.53(c) (3)(ii)(L) to consider SAMAs if not previously considered in an environmental impact assessment, related supplement, or environmental assessment for the plant. As part of its review of the SQN SAMA analysis, NRC staff evaluates the applicant's treatment of external events in the Level 1 PRA model.

The requested information is needed in order for the NRC staff to reach a conclusion on the sufficiency of the applicant'sLevel I PRA model for supporting the SAMA evaluation.

i. The Technical Evaluation Report (TER) of the SQN individual plant examination of external events (IPEEE)concludes that there are several weaknesses in the fire analysis that could lead to optimistic results. These are the inappropriatecombining of severity factors and non-suppression probabilitiesand the assumption of independence of several human actions in the main control room fire analysis. Also, it was observed that the cable spreading room was screened out due to lack of fire sources. While the latter appears to not be strictly true, the cable spreading room fire analyzed in the SQN IPEEE assumes no safe shutdown equipment failure.

Discuss these observations and the impact of them on the SQN fire CDF.

TVA Response The first weakness identified in the Technical Evaluation Report (TER, February 2001) was a concern over the potential inappropriate combination of severity factors and non-suppression probabilities. The TER states:

A severity factor is used in conjunction with failure of automatic and manual suppression possibilities. This approach has led to optimistic results for those compartments that could not be screened out in initialstages of the analysis.

This concern was first introduced in the IPEEE RAI (August 2000), which states:

In Step 3 of Phase II (please refer to Section 5 of the SQN fire IPEEEsummary report),

the submittal has introduced a "severity factor"to compute the conditional probabilityof damage to safe shutdown cables and equipment in a compartment. This "severityfactor" is then multiplied by the failure probabilitiesof automaticand manual suppression.As discussed in Reference 1, severity factors should not be multiplied by non-suppression probabilities,since the two probabilitiesare based on a common pool of event data. That is, the potential for a large fire is dependent upon failure of fire suppression;therefore, the methodology employed in IPEEEfire analysis effectively results in double counting suppression efforts.

For each case that the automatic fire suppression was credited in conjunction with the fire severity factors, please explain why such credit does not constitute double counting for suppression.Please note that a re-analysis should be provided if the multiplicationof the severity factors and the non-suppressionprobabilitiescannot be adequatelyjustified.

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The response from TVA to the (October 2000) RAI goes through a detailed explanation as to how the IPEEE fire analysis does not double count suppression efforts.

The WBN Unit 1 IPEEE RAI 6 included the same concern regarding the potential for double counting suppression efforts. The WBN Unit 1 IPEEE fire methodology is the same as the methodology used for SQN. The RAI responses for WBN and SQN explained how the methodology does not double count suppression efforts. The NRC did not identify this issue as a weakness nor did the NRC state that it led to optimistic results in the WBN Unit 1 TER.

The inclusion of severity factors in the SQN IPEEE TER introduces more realism to the fire risk assessment, and does not lead to optimistic results.

The second weakness identified in the TER was a concern over the potential for incorrectly assuming independence between human actions in the main control room fire:

In the main control room fire analysis, several events that include human actions as an importantpart of the events were assumed to be independent from one another.This assumption was not properly supported in the submittal or the response to the RAIs and may have led to optimistic evaluation of the CDF for the main control room.

During the preparation of the IPEEE RAI response mentioned above, TVA discovered that the control room fire was a special case in which the severity factor and non-suppression probability should not be combined. In the RAI response, the severity factor was removed from the main control room fire which resulted in an unacceptable CDF. Subsequently, control room recovery was credited in accordance with the EPRI Fire Probabilistic Risk Assessment Implementation Guide, resulting in a CDF below the 1 E-06 threshold.

It is assumed that the identified weakness refers to the control room assessment performed in the RAI response. Although there is little explanation in the IPEEE and the IPEEE RAIs that justifies the independence of the control room events, the two events of concern (control room evacuation and control room recovery) are not strictly human actions. The values used in the assessment for these two events are actually Conditional Core Damage Probability (CCDP) that incorporate hardware failures as well. The human actions included in the CCDPs are believed to be entirely independent because evacuation takes place within 15 minutes and recovery takes place after 60 minutes. Based on the Shutdown from Auxiliary Control Room procedure (AOP-C.04) the operators move from the main control room to the auxiliary control room and then potentially recover the control room, re-entering it after clearances have been verified.

Based on the information provided in this response, the weakness identified by the NRC in the IPEEE TER is actually a lack of adequate documentation of the assumption and did not lead to optimistic results.

The TER of the SQN IPEEE made an observation that the screening of the cable spreading room may have resulted in missing important lessons about the effects of a cable spreading room fire:

The cable spreading room was screened out citing lack of significant fire ignition sources and the presence of automatic detection and suppression capability. This rationale for screening an important compartment is optimistic since it discounts the potential of small fire scenariosthat may damage an importantset of cables or impact operators' effectiveness in the main control room. Importantlessons about the effects of cable damage in the cable spreadingroom on plant risk may have been missed.

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Based on SQN-IPEEE-005, Section 6.2, an extensive fire in the cable spreading room is assumed to not be credible due to the lack of fire ignition sources and the presence of fire detection and suppression capability. However, the conditional CDF associated with the failure of all non-credited plant equipment (including failure of bleed and feed cooling) is applied in order to maintain conservatism in the assessment. A screening assessment, such as that performed for the IPEEE, does not provide detailed insights as to the effects of particular scenarios. A more detailed assessment yields lower CDF results as more conservatisms are removed from the assessment.

For example, utilizing the process employed for other control building areas (Auxiliary Instrument Rooms) in SQN-IPEEE-005 yields the following results:

Event Tree:

Fire Automatic Fire Severity Suppression Brigade 7.01 E-03 0.85 5.96E-03 minor damage minor 0.15 0.96 1.01 E-03 minor damage severe yes 0.04 0.9 3.79E-05 multiple damage no yes 0.1 4.21 E-06 extensive damage no El- 37 of 87

Evaluation of Cable Spreading Room:

Case Prob of Core (F1CDFxP2)

Description Frequency Damage Case F1 P2 Case 1 Minor Fire, suppressed 6.97E-03 3.29E-06 2.29E-08 Case 2 Significant fire with manual suppression 3.79E-05 3.71E-04 1.40E-08 Case 3 Severe fire, assumed to require control room evacuation 4.21 E-06 0.074 3.11E-07 Total: 7.01 E-03 3.48E-07 3.48E-07 Total: 7.01 E-03 Therefore, using a slightly more detailed approach than what was originally done for the cable spreading room yields a slightly lower CDF. It is understood that the FIVE methodology utilized in the IPEEE is essentially a screening process that does not give detailed insights regarding specific fire scenarios. However, TVA does not agree that the results are optimistic as stated in the TER. Other known plants that have completed a NUREG/CR-6850 fire PRA have comparable CDF results for their cable spreading rooms.

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NRC RAI 7.c.ii ii. Provide a summary of the conservatisms and non-conservatismsin the SQN IPEEE fire analysis in light of the above TER review and the recent research and guidance reported in NUREG/CR-6850, specifically in the areas of hot short probabilities,fire ignition frequencies, and non-suppression probabilities,that indicates the fire analysis methodologies utilized for the IPEEEs may underestimate fire risk.

Discuss the impact on the evaluation of potential SAMAs for fire risk contributorsin addition to the use of the external events multiplier.

TVA Response The RAI states that NUREG/CR-6850 was the fire analysis methodologies utilized for the Individual Plant Examinations for External Events (IPEEE) may underestimate fire risk.

However, the use of NUREG/CR-6850 methods may result in an overestimate of the fire risk.

See Roadmap for Attaining Realism in Fire PRAs, NEI, December 2010 at ADAMS Accession No. ML110210990, which concludes, "Based on the results and insights from industry fire PRAs, it has been identified that the methods described in NUREG/CR-6850/EPRI TR-1011989 contain excess conservatisms that bias the results and skew insights. While the prior FAQ process made some incremental progress in addressing areas of excessive conservatisms, many more remain in need of enhancement." Thus, the results of the initial NUREG/CR-6850 analyses should not be used to draw conclusions about the IPEEE fire risk estimates.

However, in order to adequately address the concern related to fire analysis conservatisms and non-conservatisms specifically associated with the evaluation of potential SAMAs and the use of the external events multiplier, which is described below. It is important to note the bases of the external events multiplier.

The external events multiplier for SQN Unit I is given by:

EE Multiplier = (Internal Event CDF + Fire CDF + Seismic CDF) / Internal Event CDF

= (2.96E-05 + 5.83E-06 + 5.1E-05) / 2.96E-05 where:

Internal Event CDF = 2.96E-05 Fire CDF = 5.83E-06 Seismic CDF = 5.1E-05 The external events multiplier for SQN Unit 2 is given by:

EE Multiplier = (Internal Event CDF + Fire CDF + Seismic CDF) / Internal Event CDF

= (3.51 E-05 + 5.83E-06 + 5.1E-05) /3.51E-05 where:

Internal Event CDF = 3.51E-05 Fire CDF = 5.83E-06 Seismic CDF = 5.1E-05 El- 39 of 87

As can be seen from the bases of the external events multipliers for SQN Units 1 and 2, the seismic portion of the external events dominates the multiplier. Approximately 90% of the multiplier is based on seismic events and approximately 10% is due to the fire events.

Therefore, the overall effect of fire analysis conservatisms and non-conservatisms has little effect on the evaluation of potential SAMAs.

Furthermore, there are conservatisms built into the SAMA evaluations. For example, SAMA 287 directly applies to fire CDF and conservatively applies the external event multiplier (even though it is more heavily weighted by seismic risk) to ensure a bounding determination of benefit. Also, a sensitivity analysis was performed using the 95th percentile uncertainty associated with the model. As indicated in the results, this sensitivity case yielded no change to the outcome of the SAMA evaluation. Therefore, potential non-conservatisms in the fire model are offset by the conservatisms in the SAMA methodology.

Hot shorts at SQN have been addressed by the implementation of plant modifications to resolve MSO issues. These modifications were initiated based on the guidance provided in RG 1.189, Rev. 2 for fire-induced circuit failures that could cause MSOs.

NRC RAI 7.c.iii iii. Provide a discussion of the impact of the recent external flooding developments on the conclusions that external flooding is not a significant contributorto external risk at SQN and need not be considered further in the SAMA assessment.

TVA Response As discussed in the response to RAI 7.a.i, design changes are planned to improve external flooding mitigation. TVA anticipates that the completion of these changes would decrease CDF for events related to external flooding; this action will supersede any cost beneficial SAMAs that could be proposed.

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NRC RAI 7.d.i

d. Provide the following information relative to the Level 3 PRA analysis. Basis:

Applicants for license renewal are requiredby 10 CFR 51.53(c) (3) (ii)(L) to consider SAMAs if not previously consideredin an environmental impact assessment, related supplement, or environmental assessment for the plant. As part of its review of the SQN SAMA analysis, NRC staff evaluates the applicant'sanalysis of consequences in the Level 3 PRA model. The requestedinformation is needed in order for the NRC staff to reach a conclusion on the acceptabilityof the applicant's Level 3 PRA model for supporting the SAMA evaluation.

i. ER Table E. 1-18 provides the estimated population distribution within a 50-mile radius for 2041.

Describe the currentpopulation distribution surroundingSQN.

TVA Response Table E.1-18 from the Environmental Report is provided below for reference.

Table E.1 Estimated Population Distribution within a 50-Mile Radius 0 to 11 to 21 to 31 to 41 to Direction 10 Miles 20 Miles 30 Miles 40 Miles 50 Miles Total N 3,674 10,760 5,305 6,188 28,425 54,352 NNE 1,156 12,200 10,908 12,871 15,776 52,911 NE 2,975 4,672 9,844 11,761 25,876 55,128 ENE 2,290 7,756 14,687 44,628 29,841 99,202 E 4,117 39,238 10,640 5,293 6,544 65,832 ESE 4,386 66,734 7,740 1,886 24,228 104,974 SE 3,871 9,759 11,793 5,974 15,482 46,879 SSE 6,634 13,341 31,004 83,296 18,154 152,429 S 13,034 52,280 43,407 35,950 22,566 167,237 SSW 15,433 128,791 59,809 27,398 17,774 249,205 SW 13,629 129,629 42,543 19,124 13,594 218,519 WSW 29,000 53,021 7,940 22,131 13,628 125,720 W 14,406 5,338 11,446 8,050 12,882 52,122 WNW 6,887 4,919 7,022 8,123 5,447 32,398 NW 6,253 2,513 3,866 2,012 20,243 34,887 NNW 4,110 1,711 5,950 5,601 8,241 25,613 Totals 131,855 542,662 283,904 300,286 278,701 1,537,408 These values were obtained from SECPOP2000 v3.13.1 by using a population multiplier.

The population multiplier was calculated using the results of the population data calculation El- 41 of 87

based on the 2000 census data and transient population data. The 2011 data for permanent population, transient population, and combined populations from the above mentioned calculation are shown below:

Permanent 2011 Population 0 to 11 to 21 to 31 to 41 to Direction 10 Miles 20 Miles 30 Miles 40 Miles 50 Miles Total N 2,607 7,800 3,432 4,335 22,281 40,455 NNE 817 8,158 7,585 9,395 11,817 37,772 NE 1,574 3,379 6,938 9,056 18,556 39,503 ENE 1,952 5,385 9,825 30,363 22,951 70,476 E 2,111 28,271 6,924 3,169 4,958 45,433 ESE 2,901 45,852 5,944 820 17,722 73,239 SE 2,945 6,474 8,880 6,970 12,184 37,453 SSE 4,972 11,146 23,978 74,358 19,897 134,351 S 9,021 42,720 35,751 19,608 11,252 118,352 SSW 8,923 95,590 50,028 18,645 11,756 184,942 SW 11,420 82,065 24,702 14,703 9,765 142,655 WSW 18,950 28,360 5,436 13,744 8,888 75,378 W 8,414 3,097 6,953 5,144 8,549 32,157 WNW 4,117 5,013 4,544 4,945 3,813 22,432 NW 3,812 1,920 2,793 1,514 14,401 24,440 NNW 2,640 1,017 4,275 3,826 5,711 17,469 Totals 87,176 376,247 207,988 220,595 204,501 1,096,507 Based on 2000 Census Data E1- 42 of 87

Transient 2011 Population 0 to 11 to 21 to 31 to 41 to Direction 10 Miles 20 Miles 30 Miles 40 Miles 50 Miles Total N 232 685 299 376 1,911 3,503 NNE 73 713 660 818 1,022 3,286 NE 140 295 605 789 1,601 3,430 ENE 173 469 855 2,642 1,967 6,106 E 185 2,465 613 280 450 3,993 ESE 255 3,998 528 72 1,538 6,391 SE 260 564 751 580 1,009 3,164 SSE / 442 969 2,004 6,199 1,654 11,268 S 803 3,697 2,972 1,641 940 10,053 SSW 794 8,384 4,180 1,564 951 15,873 SW 1,017 7,304 2,119 1,068 364 11,872 WSW 1,687 2,513 478 1,075 532 6,285 W 749 269 610 453 747 2,828 WNW 367 429 392 435 334 1,957 NW 339 166 241 134 1,266 2,146 NNW 235 89 373 339 499 1,535 Totals 7,751 33,009 17,680 18,465 16,785 93,690 Based on State Tourism Data El- 43 of 87

Total 2011 Population 0 to 11 to 21 to 31 to 41 to Direction 10 Miles 20 Miles 30 Miles 40 Miles 50 Miles Total N 2,839 8,485 3,731 4,711 24,192 43,958 NNE 890 8,871 8,245 10,213 12,839 41,058 NE 1,714 3,674 7,543 9,845 20,157 42,933 ENE 2,125 5,854 10,680 33,005 24,918 76,582 E 2,296 30,736 7,537 3,449 5,408 49,426 ESE 3,156 49,850 6,472 892 19,260 79,630 SE 3,205 7,038 9,631 7,550 13,193 40,617 SSE 5,414 12,115 25,982 80,557 21,551 145,619 S 9,824 46,417 38,723 21,249 12,192 128,405 SSW 9,717 103,974 54,208 20,209 12,707 200,815 SW 12,437 89,369 26,821 15,771 10,129 154,527 WSW 20,637 30,873 5,914 14,819 9,420 81,663 W 9,163 3,366 7,563 5,597 9,296 34,985 WNW 4,484 5,442 4,936 5,380 4,147 24,389 NW 4,151 2,086 3,034 1,648 15,667 26,586 NNW 2,875 1,106 4,648 4,165 6,210 19,004 Totals 94,927 409,256 225,668 239,060 221,286 1,190,197 El- 44 of 87

NRC RAI 7A.11 ii. ER Section E. 1.5.2.6 states meteorologicaldata from 2005 resulted in the highest release quantities.

1. Confirm that the highest release quantities correspondto the highest public dose risk and highest averted cost risk.

2. Describe the modeling of precipitationevents, including boundary precipitationmodeling, and precipitationinfluence on calculated doses.

3. Quantify the amount of missing meteorologicaldata in 2003, 2004, and 2005, which were estimated using data interpolationor replacement.

TVA Response

7. d.ii.1: The highest release quantities correspond to the highest public dose risk and highest modified maximum averted cost risk (MMACR) as shown in the tables below:

Unit 1 Met Data Sensitivity Economic Affected Met Dose Risk Risk Land MMACR Data Year (person-rem/yr) ($/yr) (hectares) ($/yr) 2003 3.97E+01 8.94E+04 5.67E+04 7,152,416 2004 4.16E+01 9.40E+04 5.35E+04 7,414,600 2005 4.50E+01 9.70E+04 5.76E+04 7,720,482 Unit 2 Met Data Sensitivity Economic Affected Met Dose Risk Risk Land MMACR Data Year (person-rem/yr) ($/yr) (hectares) ($/yr) 2003 3.87E+01 8.59E+04 5.67E+04 6,554,441 2004 4.04E+01 9.03E+04 5.35E+04 6,772,713 2005 4.39E+01 9.31 E+04 5.76E+04 7,046,951 7.d.ii.2: The boundary weather conditions included a mixing height of 1000 m, stability class D, a rain rate of 5 mm/hr, and wind speed of 5 m/s. The SQN WinMACCS evaluation produced a conservative result by forcing the plume to wash out in the last grid interval in accordance with NUREG-1150. Further, the rain densities and rain intensities utilized in the SQN WinMACCS evaluation were in accordance with NUREG-1 150.

Modeling of the rain rate that was used for the boundary weather conditions was performed for selected Release Categories using rain rates of 5 mm/hr (base case), and 0 mm/hr and 10 mm/hr sensitivities. The sensitivity results were within +/-1% of population dose of the base case.

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Economic costs were within approximately +/-10% of the base case. As rain increased, population dose and economic cost increased. As rain decreased, population dose and economic cost decreased. Further, based on the meteorological data for the three years evaluated, the average rainfall per hour (during hours of rain) was 2.77 mm/hr for 2003, 3.04 mm/hr for 2004, and 1.75 mm/hr for 2005. Therefore, the 5 mm/hr rain rate was determined to be a conservative value for this assessment.

7.d.ii.3: For the year 2003, 8,487 data points out of a total of 8,760 data points (96.9%) were available. For the year 2004, 8,558 data points out of a total of 8,784 data points (97.4%) were available, although, the 366th day of 2004 was not used in the Met File for WinMACCS. For the year 2005, 8,692 data points out of a total of 8,760 data points (99.2%) were available.

When only one hour of data was missing, values were interpolated based on the values immediately before and after the data gap. When more than one hour of data was missing in series, the data was replaced with data from days with similar meteorological conditions immediately before and after the missing data, in accordance with Atkinson & Lee, "Procedures for Substituting Values for Missing NWS Meteorological Data for use in Regulatory Air Quality Models." (See http://www.epa.gov/scram001/surface/missdata.txt)

NRC RAI 7.d.iii iii. As described in ER Section E. 1.5.2.7, the maximum preparationtime of 105 minutes was applied for the 10-mile emergency planning zone. This value was described to include 75 minutes for notification and 30 minutes for preparation.

1. Provide information to support the total delay time to leave of 105 minutes.

2. Clarify if the evacuation analysis considered generic information for the average evacuation speed or site specific information based on the current (orprojected) number of people requiredto evacuate via existing roads.

TVA Response 7.d.iii.l: Annex H of the Tennessee Multi-jurisdictional Radiological Emergency Response Plan (MJRERP), entitled "Evacuation" contains site specific information relative to exposure control and evacuation. Annex H of the emergency response plan is specific to SQN. This Annex provides evacuation route estimates and the bases for evacuation times.

Assumption E of Appendix 1, Evacuation Time Estimate - 10 Mile EPZ (Page H-6) of Annex H, Evacuation, of the Tennessee MJRERP, states the following:

A maximum preparationtime of 105 minutes was based on times for people in Southeast Hamilton County to get home and have approximately 30 minutes to prepare to leave.

Because this value was assumed in the Tennessee MJRERP, it was reasonable to use this value in this evaluation.

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7.d.iii.2: Annex H of the Tennessee MJRERP contains site specific information relative to exposure control and evacuation. Annex H of the emergency response plan is specific to SQN.

This Annex provides evacuation route estimates and the bases for evacuation times.

Assumption D of Appendix 1, Evacuation Time Estimate - 10 Mile EPZ (Page H-6) of Annex H, assumes that the travel speed on major evacuation routes is 20 mph (8.94 m/s). Further, it states that most of the evacuation routes are local roads with low capacities. With the expected high evacuation volumes, the road-way would be under forced flow conditions; therefore the travel speeds would be lower.

The site specific evacuation analysis performed assumed an evacuation speed of 2.2 m/s (4.92 mph) to account for forced flow conditions and increased projected population for the year 2041 within the 10-mile emergency planning zone. The travel speed estimated in the Annex H of the Tennessee MJRERP (8.94 m/s) was reduced by approximately 50% to account for forced flow conditions, where travel speed would be lower (-4.4 m/s). This evacuation speed was then reduced by 50% to 2.2 m/s to account for the projected population growth within the 10-mile emergency planning zone. Reducing the evacuation speed by 50% to account for the projected population growth was determined to be conservative as the population within the 10-mile emergency planning zone was only projected to increase by 39% between 2011 and 2041.

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NRC RAI 7.d.iv iv. Evacuationsensitivity is presentedin ER Table E. 1-23. Small effects on dose were shown for evacuation fractions of 90, 95 and 99.5 percent (or 10, 5, and 0.5 percent of individuals who do not evacuate). Consideringroughly half of the calculatedpopulation dose risk is attributed to late releases,

1. explain the causes of the reportedsmall effects on dose from evacuation.

2. Briefly describe where members of the public would be evacuatedin terms of the distance from the plant.

3. Clarify if radiologicalexposure to evacuees at those destinations was accounted for in the analysis TVA Response 7.d.iv.l: The dose to the population for this accident sequence is primarily to those outside of the EPZ as shown in the table below:

SQN IVa - COHORT 1-3 Population Likelihood Population Population Evacuatin Dose Dose Risk Dose Risk Fraction (person-rem) (events/yr) (person-rem/yr)  %

0 - 10 mi 1.01E+05 1.81E+00 9.33%

90% 1.792E-05 10 - 50 mi 9.82E+05 1.76E+01 90.67%

0 - 10 mi 7.76E+04 1.39E+00 7.32%

95% 1.792E-05 10 - 50 mi 9.82E+05 1.76E+01 92.68%

0 - 10 mi 5.62E+04 1.01E+00 5.41%

99,5% 1.792E-05 10 - 50 mi 9.82E+05 1.76E+01 94.59%

These results were obtained by summing COHORT 1 (evacuating fraction of the public),

COHORT 2 (non-evacuating fraction of the public), and COHORT 3 (NRC best practices CHRONC input).

The CHRONC file includes doses to the population based on food ingestion, water ingestion, water wash off, etc.

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The population dose results due to the immediate release (i.e., COHORT 1 and COHORT 2) are shown below:

SQN IVa - COHORT I & 2 Population Likelihood Population Population Evacuating Distances Dose Dose Risk Dose Risk Fraction (person-rem) (events/yr) (person-rem/yr)  %

0 - 10 mi 4.86E+04 8.71 E-01 6.85%

90% 1.792E-05 10 - 50 mi 6.61E+05 1.18E+01 93.15%

0 - 10 mi 2.43E+04 4.36E-01 3.55%

95% 1.792E-05 10 - 50 mi 6.61 E+05 1.18E+01 96.45%

0 - 10 mi 2.43E+03 4.35E-02 0.23%

99.5% 1.792E-05 10 - 50 mi 6.61 E+05 1.18E+01 99.63%

Note: The results above are for release category IVa. Therefore, the summation will not be equivalent to the results in ER Table E.1-23 which sums all release categories.

The results from using only COHORT 1 and 2 show that the dose to the public in the EPZ based on the non-evacuating fraction is proportional to the fraction of the population that does not evacuate.

The evacuating fraction makes up only a small portion of the total dose to the entire population within 50 miles of the plant. This is in part due to the small resident and transient population within the EPZ and the relatively large population (i.e., Chattanooga) within 50 miles of SQN.

7.d.iv.2: The members of the public are assumed to be evacuated to a number of locations as shown in the drawing below from the Tennessee MJRERP, SQN Addendum:

HAMILTON COUNTY SHELTERS:

1. Brainerd High School

2. East Ridge High School

3. East Ridge Middle School

4. Dalewood Middle School

5. Chattanooga High School

6. Orchard Knob Middle School

7. Howard School of Academics and Technology BRADLEY COUNTY SHELTERS:

9. Ocoee Middle School El- 49 of 87

MEIGS COUNTY SHELTERS:

10. Meigs County High School RHEA COUNTY SHELTERS:

11. Rhea Central Elementary School

12. Rhea County High School SEQUATCHIE COUNTY SHELTER:

13. Sequatchie County High School SHELTER LOCATIONS I0 Ia 20 A SHELT'ER SCALEOF MILES El- 50 of 87

7.d.iv.3: The radiological exposure to the population within the EPZ is controlled in accordance with the emergency plan (Annex G). The population within the EPZ (10 miles) would be evacuated to the above locations if possible. Once there, the evacuees would remain in place and would continue to receive dose.

NRC RAI 7.e.i

e. Provide the following information with regardto the selection and screening of Phase I SAMA candidates. Basis: Applicants for license renewal are requiredby 10 CFR 51.53(c)(3)(ii)(L) to consider SAMAs if not previously considered in an environmental impact assessment, related supplement, or environmental assessment (EA) for the plant. As part of its review of the SQN SAMA analysis, NRC staff evaluates the applicant's basis for the selection and screening Phase I SAMA candidates. The requested information is needed in order for the NRC staff to reach a conclusion on the adequacy of the applicant's Phase I SAMA selection and screening process for the SAMA evaluation.

i ER Section E.1.1 indicates that the results of the importance analysis were reviewed down to a risk reduction worth (RRW) of 1.005 for the identification of potentialcost beneficial SAMAs. This correspondsto a potentialmaximum benefit including uncertainty of $97K for SQN Unit I and $88K for Unit 2.

This precludes potentialsimple procedure changes that accordingto Section E.2.3 might cost $50K.

Discuss the potential for added candidate SAMAs down to a RRW correspondingto a simple procedure change.

TVA Response For Unit 1, the $50K value for a simple procedure change would correspond to an RRW of slightly greater than 1.0026. The corresponding RRW value for Unit 2 is 1.003. The additional basic events with an RRW between 1.005 and the values above are discussed in the tables below for each unit. No new low cost (i.e., simple procedure change) SAMA candidates were identified as a result of this review.

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Unit I Event (Unit 1, Level 1) Red W Description Disposition UI_2BLOCK 1.0049 PROBABILITY THAT 2 2 PORVs being blocked at the same time would be a result of coincident PORVS ARE BLOCKED maintenance. The SQN Outage and Site Scheduling Directive Manual is intended to minimize these type of events for systems important to PRA.

Given the existing low probability of this event, significant improvements to the Outage and Site Scheduling Manual would likely cost in excess of

$100,000 which is greater than the RRW for this event. One or more block valves inoperable has required actions and completion times per LCO 3.4.3.2.

HACD2 1.0047 Perform cooldown with AFW This term represents the failure of the human action to cooldown with AFW and steam dumps and steam dumps. Phase II SAMAs 103 and 283 to increase training and improve awareness of important operator actions have been evaluated.

While this event was not included in these evaluations, SAMA 283 was determined to be potentially cost beneficial and the event could be added to those addressed in that SAMA.

MOCXC1FCV_07000041E 1.0047 FCV-70-4 TRANSFERS This term represents CCS flow control valve transferring close resulting in the CLOSED loss of CCS. Phase II SAMA 45 to enhance procedural guidance for use of cross-tied component cooling pumps has been evaluated.

MOCXC1FCV_07000081E 1.0047 FCV 70-8 TRANSFERS This term represents CCS flow control valve transferring close resulting in the CLOSED loss of CCS. Phase II SAMA 45 to enhance procedural guidance for use of cross-tied component cooling pumps has been evaluated.

MOCXC1 FCV_07000251E 1.0047 FCV 70-25 TRANSFERS This term represents CCS flow control valve transferring close resulting in the CLOSED loss of CCS. Phase II SAMA 45 to enhance procedural guidance for use of cross-tied component cooling pumps has been evaluated.

AHUFRl CLR_0300191 1.0046 CCS PUMPS AND AFW This term represents the failure of CCS pumps and AFW pumps space PUMPS SPACE COOLER B coolers. Phase II SAMAs 268 and 289 evaluate the cooling requirements for FAILS TO RUN the CCS/AFW pump space coolers and the potential for installing backup cooling to these coolers. As this is a hardware failure, no additional low cost procedural SAMA was identified.

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Event (Unit 1, Level 1) Red W Description Disposition SHESUM 1.0046 Remove Drain Plugs from This term represents the human action to remove drain plugs from the Refueling Canal After refueling canal after refueling. Phase II SAMAs 103 and 283 to increase Refueling training and improve awareness of important operator actions have been evaluated. While this event was not included in these evaluations, SAMA 283 was determined to be potentially cost beneficial and the event could be added to those addressed in that SAMA.

TM_U1PMP0740010 1.0046 RHR PUMP 1B-B IS This term represents the maintenance unavailability of RHR Pump lB-B.

UNAVAILABLE DUE TO Phase II SAMA 278 improving the reliability of the RHR pumps has been TEST OR MAINTENANCE evaluated. No additional low cost procedural SAMA was identified to address reliability of this component.

EDGFR1GEN10821AA 1.0044 DG 1A-A Fails to Run This event represents the failure of the Unit 1 A-A DG to run for the remaining mission time. Phase II SAMAs 14, 161 and 254 to improve the reliability and availability of EDGs have also been evaluated. As this is a hardware failure no additional low cost procedural SAMA was identified.

UlPR-INS 1.0044 Insufficient pressure relief This term represents insufficient pressure relief through PORVs. Phase II SAMAs 103 and 283 to increase training and improve awareness of important operator actions, including actions to depressurize, have been evaluated. No additional low cost procedural SAMA was identified to address reliability and capacity of these components.

AHUFR1CLR_0300190 1.0043 CCS PUMPS AND AFW This term represents the failure of CCS pumps and AFW pumps space PUMP SPACE COOLER A coolers. Phase II SAMAs 268 and 289 evaluate the cooling requirements for FAILS TO RUN the CCS/AFW pump space coolers and the potential for installing backup cooling to these coolers. As this is a hardware failure no additional low cost procedural SAMA was identified.

PMAFD1 PMP_00300128 1.0043 MDAFWP FAILS TO START This term represents MDAFWP 1B-B failing to run. Phase I SAMA 223 to SQN-1-3-128-B improve the reliability of the AFW pumps and valves has been implemented.

Phase II SAMA 068 to add a motor driven feedwater pump has been evaluated. As this is a hardware failure no additional low cost procedural SAMA was identified.

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Event (Unit 1, Level 1) Red W Description Disposition EDGFR1GEN10821B_B 1.0042 DG 1B-B Fails to Run This event represents the failure of the Unit 1 B-B DG to run for the remaining mission time. Phase II SAMAs 14, 161 and 254 to improve the reliability and availability of EDGs have also been evaluated. As this is a hardware failure no additional low cost procedural SAMA was identified.

%714.0-AO1-1_067_MOB 1.0041 ERCW MAJOR FLOOD This event represents the initiator for a major ERCW flood event in partition 1 EVENT IN 714.0-Al-1 FROM of room 714.0-Al. This initiator potentially results in loss of component DISCHARGE B - Initiator cooling. Phase II SAMA 279 to improve internal flooding response procedures and training to improve the response to internal flooding events has been evaluated.

%714.0-AO1-2_067_M01B 1.0041 ERCW MAJOR FLOOD This event represents the initiator for a major ERCW flood event in partition 2 EVENT IN 714.0-Al-2 FROM of room 714.0-Al. This initiator potentially results in loss of component DISCHARGE B - Initiator cooling. Phase II SAMA 279 to improve internal flooding response procedures and training to improve the response to internal flooding events has been evaluated.

%690.0-AO1-1_067_F_0A 1.004 ERCW FLOOD EVENT IN This event represents the initiator for a ERCW flood event in partition 1 of 690.0-Al-1 FROM A room 690.0-Al. This initiator results in loss of component cooling. Phase II DISCHARGE - Initiator SAMA 279 to improve internal flooding response procedures and training to improve the response to internal flooding events has been evaluated.

%690.0-A01-1_067_F_1A 1.004 ERCW FLOOD EVENT IN This event represents the initiator for a ERCW flood event in partition 1 of 690.0-Al-1 FROM 1A-A room 690.0-Al. This initiator results in loss of component cooling. Phase II HEADER - Initiator SAMA 279 to improve internal flooding response procedures and training to improve the response to internal flooding events has been evaluated.

UO06_250_BCHFR_ 1.004 CCF of all components in This event represents the common cause failure of 250V Vital Battery VBB ALL group Chargers. Phase II SAMA 218 to improve the reliability of power supplies has

'UO 06 250_BCHFRVBB' been evaluated. As this is a hardware failure no additional low cost

_ procedural SAMA was identified.

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Event (Unit 1, Level 1) Red W Description Disposition

%734.0-A1 3-1_067_FOB 1.0039 ERCW FLOOD EVENT IN This event represents the initiator for a ERCW flood event in partition 1 of 734.0-A13-1 FROM room 734.0-Al 3. This initiator potentially results in loss of component DISCHARGE HEADER B - cooling. Phase II SAMA 279 to improve internal flooding response Initiator procedures and training to improve the response to internal flooding events has been evaluated.

%734.0-A02_026_S 1.0036 HPFP SPRAY EVENT IN This event represents the initiator for a Hi Pressure Fire Pump (HPFP) flood 734.0-A2 - Initiator event in room 734.0-A2. This initiator results in loss the 6.9kV and the 480V Shutdown Boards. Phase II SAMA 014 evaluates installing a gas turbine generator to provide additional independent power to the shutdown boards.

Phase II SAMA 279 to improve internal flooding response procedures and training to improve the response to internal flooding events has been evaluated.

INVFRINVB250QLD 1.0036 Failure of the Normal Inverter This event represents the failure of a normal 250V Inverter. Phase II SAMA 1-1 1-INVB-250-QL-D 218 to improve the reliability of power supplies has been evaluated. As this is a hardware failure no additional low cost procedural SAMA was identified.

%669.0-AO1067_FOB 1.0035 ERCW FLOOD EVENT IN This event represents the initiator for a ERCW flood event in room 669.0-Al.

669.0-Al FROM This initiator potentially results in loss of component cooling train 1A. Phase DISCHARGE HEADER B - II SAMA 279 to improve internal flooding response procedures and training to Initiator improve the response to internal flooding events has been evaluated.

AHUFD1CLR_0300175 1.0035 Air Handling Unit (Standby) This event represents the failure of the standby RHR Pump Room A AHU to Fails to Start (1-CLR-030- run. Phase II SAMA 160 to implement procedures for temporary HVAC, for 0175) rooms including RHR Pump Room A, has been evaluated. As this is a hardware failure no additional low cost procedural SAMA was identified.

%1SLOCA-CL1 1.0034 SMALL LOCA ON COLD This event represents a small break LOCA on cold leg 1. Plant response to a LEG 1 small break LOCA would be to align high pressure recirculation and depressurize. Phase II SAMA 032 to automatically align high pressure recirculation has been evaluated. Phase II SAMAs 103 and 283 to increase training and improve awareness of important operator actions have been evaluated.

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Event (Unit 1, Level 1) Red W Description Disposition

%1SLOCA-CL4 1.0034 SMALL LOCA ON COLD This event represents a small break LOCA on cold leg 4. Plant response to a LEG 4 small break LOCA would be to align high pressure recirculation and depressurize. Phase II SAMA 032 to automatically align high pressure recirculation has been evaluated. Phase II SAMAs 103 and 283 to increase training and improve awareness of important operator actions have been evaluated.

%690.0-AO1-3_067_S 1.0034 ERCW SPRAY EVENT IN This event represents the initiator for an ERCW spray event in partition 3 of 690.0-Al - Initiator room 690.0-Al. This initiator results in loss of component cooling pump B as well as the CCS/AFW area Space Coolers. Phase II SAMA 288 to install spray protection on CCS Pumps and CCS/AFW Space Coolers has been evaluated.

%1SLOCA-CL2 1.0033 SMALL LOCA ON COLD This event represents a small break LOCA on cold leg 2. Plant response to a LEG 2 small break LOCA would be to align high pressure recirculation and depressurize. Phase II SAMA 032 to automatically align high pressure recirculation has been evaluated. Phase II SAMAs 103 and 283 to increase training and improve awareness of important operator actions have been evaluated.

%1SLOCA-CL3 1.0033 SMALL LOCA ON COLD This event represents a small break LOCA on cold leg 3. Plant response to a LEG 3 small break LOCA would be to align high pressure recirculation and depressurize. Phase II SAMA 032 to automatically align high pressure recirculation has been evaluated. Phase II SAMAs 103 and 283 to increase training and improve awareness of important operator actions have been evaluated.

%690.0-AO1-3_024_S 1.0033 RCW SPRAY EVENT IN This event represents the initiator for a RCW spray event in partition 3 of 690.0-Al - Initiator room 690.0-Al. This initiator results in loss of component cooling pump B as well as the CCS/AFW area Space Coolers. Phase II SAMA 288 to install spray protection on CCS Pumps and CCS/AFW Space Coolers has been evaluated.

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Event (Unit 1, Level 1) Red W Description Disposition

%1LRCP 1.0032 Loss of 1 or More This event represents the loss of 1 or more RCPs/Primary Flow. Given the RCPs/Primary Flow complexity of the RCPs and primary system, the RRW (1.0032) of this event, and the corresponding benefit of eliminating this initiator ($63,000), there are no simple plant procedure changes or minor hardware changes that would result in a potentially cost beneficial SAMA for this initiator.

%662.0-1_024_S 1.0031 RCW SPRAY EVENT IN This event represents the initiator for a RCW spray event in room 662.0-1.

662.0 Initiator Phase II SAMA 279 to improve internal flooding response procedures and training to improve the response to internal flooding events has been evaluated.

EDGFR_SOK_2 1.0031 EDGFR State of Knowledge This term is a factor used to address the state of knowledge correlation for Factor for a group of 2 two DGs failing to run. It is used to adjust the frequency of cutsets which contain two EDG fail to run events. Phase II SAMAs 14, 161 and 254 to improve the reliability and availability of EDGs have also been evaluated.

HART1 1.003 Manually trip reactor, given This event represents the failure of operators to manually trip the reactor if SSPS fails automatic trips fail. Phase II SAMAs 136, 137, 103, 277 and 283 to add additional methods of tripping reactor and to increase training and improve awareness of important operator actions have been evaluated. While this event was not included in these evaluations, SAMA 283 was determined to be potentially cost beneficial and the event could be added to those addressed in that SAMA.

TMODRYA0320001 1.003 ACAS TRAIN A DRYER This term represents the probability of ACAS Train A Dryer being in MAINTENANCE maintenance. Phase II SAMA 188 to implement modifications to the compressed air system to increase their reliability, increase capacity, and decrease their time in maintenance has been evaluated. No additional low cost procedural SAMA was identified to address reliability of this component.

UOERW08POEFRALL 1.003 CCF of all components in This term represents the common cause failure of all components in the group 'UO_ERW08POEFR' ERCW Pumps group. Phase II SAMA 046 for adding a service water pump has been evaluated. As this is a hardware failure no additional low cost procedural SAMA was identified.

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Event (Unit 1, Level 1) Red W Description Disposition X-WI-SBO7A 1.003 0 This term represents the failure to recover offsite power to the unit following a weather related LOOP. This term is applied for SBO sequences where both EDGs failed to start, battery failure occurs at 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, RCP seal leak is 21 gpm, AFW operating, and depressurization has not occurred. Phase II SAMAs 70, 167, 215, 226 and 240 to improve coping capability during an SBO have been evaluated. However, the recovery event is dependent on conditions and actions outside the control of the plant and therefore no additional low cost procedure SAMAs were identified.

%1EXMFW 1.0029 Excessive MFW This term represents an excessive MFW initiating event. Phase I SAMA 065 to install a digital feedwater upgrade has already been implemented at SQN.

There have been no excessive MFW events at SQN. The initiating event probability is derived from NUREG/CR-6928. The cost of modifications to the digital feedwater system would be greater than the potential benefit in reducing the frequency of an excessive feedwater initiating event.

EDGFR2GEN20822AA 1.0029 DG 2A-A Fails to Run This event represents the failure of the Unit 2 A-A DG to run for the remaining mission time. Phase II SAMAs 14, 161 and 254 to improve the reliability and availability of EDGs have also been evaluated. As this is a hardware failure no additional low cost procedural SAMA was identified.

HASE1 1.0029 Stop RCPs on Phase B This term represents the failure of the human action stop RCPs on Phase B isolation (Non-LOCA Initiator) isolation for non-LOCA initiators. Phase II SAMAs 103 and 283 to increase training and improve awareness of important operator actions have been evaluated. While this event was not included in these evaluations, SAMA 283 was determined to be potentially cost beneficial and the event could be added to those addressed in that SAMA.

INVFR1INVB250QNE 1.0029 Failure of the Inverter 1-11 1- This event represents the failure of a 250V Inverter. Phase II SAMA 218 to INVB-250-QN-E improve the reliability of power supplies has been evaluated. As this is a hardware failure no additional low cost procedural SAMA was identified.

TMODRYA0320002 1.0029 ACAS TRAIN B DRYER This term represents the probability of ACAS Train B Dryer being in MAINTENANCE maintenance. Phase II SAMA 188 to implement modifications to the compressed air system to increase their reliability, increase capacity, and decrease their time in maintenance has been evaluated. No additional low cost procedural SAMA was identified to address reliability of this component.

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Event (Unit 1, Level 1) Red W Description Disposition UOCAS02CMPSD_1_2 1.0029 CCF of two components: This term represents the common cause failure of ACAS compressor B failing CMPSD0CMP_0320060 & to start. Phase II SAMAs 70, 87, 188 to increase reliability of air systems and CMPSDOCMP_0320086 coping capability on their loss have been evaluated. As this is a hardware failure no additional low cost procedural SAMA was identified.

X-WI-1 LOOPDGALL 1.0029 0 This term represents the failure to recover offsite power to the unit following a weather related LOOP and the common cause failure of all DGs. Phase II SAMAs 14, 70, 167, 215, 226 and 240 to improve coping capability during an SBO have been evaluated. However, the recovery event is dependent on conditions and actions outside the control of the plant and therefore no additional low cost procedure SAMAs were identified.

X-WI-SBO7E 1.0029 0 This term represents the failure to recover offsite power to the unit following a weather related LOOP. This term is applied for SBO sequences where both EDGs failed to run due to common cause, battery failure occurs at 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, RCP seal leak is 182 gpm, AFW operating and depressurization has not occurred.

Phase II SAMAs 14, 70, 167, 215, 226 and 240 to improve coping capability during an SBO have been evaluated. However, the recovery event is dependent on conditions and actions outside the control of the plant and therefore no additional low cost procedure SAMAs were identified.

EDGFD1GEN0821AA 1.0028 DG Fails to Start and Run This event represents the failure of the Unit 1 A-A DG to start and run for one First Hour hour. Phase II SAMAs 14 and 254 to improve the reliability and availability of EDGs have also been evaluated. As this is a hardware failure no additional low cost procedural SAMA was identified.

EDGFD1GEN0821B_B 1.0028 DG fails to start and run first This event represents the failure of the Unit 1 B-B DG to start and run for one hour hour. Phase II SAMAs 14 and 254 to improve the reliability and availability of EDGs have also been evaluated. As this is a hardware failure no additional low cost procedural SAMA was identified.

TMOCMP_0320060 1.0027 COMPRESSOR TRAIN A IN This term represents the probability of compressor train A being in MAINTENANCE maintenance. Phase II SAMA 188 to implement modifications to the compressed air system to increase their reliability, increase capacity, and decrease their time in maintenance has been evaluated. No additional low cost procedural SAMA was identified to address reliability of this component.

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Event (Unit 1, Level 1) Red W Description Disposition TMOCMP_0320086 1.0027 COMPRESSOR TRAIN B IN This term represents the probability of compressor train B being in MAINTENANCE maintenance. Phase II SAMA 188 to implement modifications to the compressed air system to increase their reliability, increase capacity, and decrease their time in maintenance has been evaluated. No additional low cost procedural SAMA was identified to address reliability of this component.

UO06_250_BCHFR_ 1.0027 CCF of three components: This event represents the common cause failure of 250V Vital Battery VBB 1 3 4 BCHFROCHGB250QED & Chargers. Phase II SAMA 218 to improve the reliability of power supplies has VBB1_ BCHFROCHGB250QGE & been evaluated. As this is a hardware failure no additional low cost BCHFROCHGB250QH procedural SAMA was identified.

UlPRCI1 1.0027 Interval 1 for PRC (pres relief This term represents insufficient pressure relief through PORVs. Phase II always insufficient) SAMAs 103 and 283 to increase training and improve awareness of important operator actions, including actions to depressurize, have been evaluated. No additional low cost procedural SAMA was identified to address the reliability and capacity of these components.

%OLOSP-GR 1.0026 LOOP (Grid Related) This term represents the LOOP to the unit due to transmission grid related issues or failures. Phase II SAMAs 14, 70, 167, 215, 226 and 240 to improve coping capability during an SBO have been evaluated. However, the recovery event is dependent on conditions and actions outside the control of the plant and therefore no additional low cost procedure SAMAs were identified.

UIXPORV2 1.0026 BLOCK VALVE CLOSED TO This term represents the probability of a block valve being closed to isolate a ISOLATE LEAKING PORV leaking PORV. Phase II SAMA 284 to reduce probability that the pressurizer safety relief valves fail to close was evaluated.

X-WI-SBO7B 1.0026 0 This term represents the failure to recover offsite power to the unit following a weather related LOOP. This term is applied for SBO sequences where one EDG failed to start and the other failed to run, battery failure occurs at 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, RCP seal leak is 182 gpm, AFW operating and depressurization has not occurred.

Phase II SAMAs 14, 70, 167, 215, 226, 229, and 240 to improve coping capability during an SBO have been evaluated. However, the recovery event is dependent on conditions and actions outside the control of the plant and therefore no additional low cost procedure SAMAs were identified.

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Unit 2 Event (Unit 2, Level 1) Red W Description Disposition AHUFD_SOK_2 1.0049 AHUFD State of This term is a factor used to address the state of knowledge correlation for Knowledge Factor for a two air handling units. It is used to adjust the frequency of cutsets which group of 2 contain two air handling unit events. Phase II SAMA 160 to improve coping capability for the loss of room cooling has been evaluated. SAMA 268 to assess the need for the CCS/AFW Space Coolers (the most important AHUs) has been evaluated. SAMAs 275, 288, 289 to increase availability of space coolers have been evaluated.

%690.0-A01-2_024_S 1.0048 RCW SPRAY EVENT IN This event represents the initiator for a RCW spray event in partition 2 of 690.0-Al Initiator room 690.0-Al. This initiator results in loss of component cooling pump A as well as the CCS/AFW area Space Coolers. Phase II SAMA 288 to install spray protection on CCS Pumps and CCS/AFW Space Coolers has been evaluated.

%690.0-AO1-4_026_M 1.0048 HPFP MAJOR FLOOD IN This event represents the initiator for a HPFP flood event in partition 4 of 690.0-AO1 Initiator room 690.0-Al. This initiator results in loss of MDAFWP 2B-B. Phase II SAMA 279 to improve internal flooding response procedures and training to improve the response to internal flooding events has been evaluated.

%690.0-AO1-1_026_S 1.0046 HPFP SPRAY EVENT IN This term represents the initiator for a HPFP spray event in partition 1 of 690.0-Al - Initiator room 690.0-Al in the AB. This initiator results in loss of MDAFWP 2A-A.

Phase II SAMA 275 to install spray shields on the AFW pumps and AFW/Boric Acid Tank (BAT) Space Cooler has been evaluated.

%690.0-AO1-1_067_S 1.0046 ERCW SPRAY EVENT IN This event represents the initiator for an ERCW spray event in partition 1 of 690.0-Al - Initiator room 690.0-Al. This initiator results in loss of MDAFWP 2A-A. Phase II SAMA 275 to install spray shields on the AFW pumps and AFW/BAT Space Cooler has been evaluated.

SSIOP 1.0046 Terminate SI to prevent This term represents the failure of operators to terminate SI. Phase II PORV water challenge SAMA 283 to increase the awareness of important human actions was evaluated. While this event was not included in these evaluations, SAMA 283 was determined to be potentially cost beneficial and the event could be added to those addressed in that SAMA.

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COMBINATION_2674 1.0045 HEP dependency factor for This term is a factor that accounts for the dependency between multiple HARR1 ,SSIOP human failure events (HARR1, SSIOP) that occur in the same cutset.

Phase II SAMA 283 to improve awareness for important human actions including combinations has been evaluated. This combination was analyzed in SAMA 283.

SHETDA 1.0045 Turbine Driven AFW This term represents the TDAFW Isolation test error in which an operator Isolation Test Error isolates the flowpath to the SGs from TDAFW. There is no indication in the MCR of isolation valve positioning and it could be left closed post check valve testing, preventing flow to a SG. Procedures and checklists are in place to ensure the isolation valves are in the required position post testing.

Phase II SAMA 283 to improve awareness for important human actions has been evaluated. While this event was not included in these evaluations, SAMA 283 was determined to be potentially cost beneficial and the event could be added to those addressed in that SAMA.

%690.0-AO1-2_067_FOB 1.0044 ERCW flood event in This event represents the initiator for an ERCW flood event in partition 2 of 690.0-AO1-2 from the B room 690.0-Al. This initiator results in loss of component cooling. Phase II discharge header with no SAMA 279 to improve internal flooding response procedures and training to isolation. - Initiator improve the response to internal flooding events has been evaluated.

%714.0-AO1-2_067_MOA 1.0042 ERCW MAJOR FLOOD This event represents the initiator for a major ERCW flood event in room EVENT IN 714.0-A1-2 714.0-Al. This initiator results in loss of CCS Train 2A. Phase II SAMA 279 FROM DISCHARGE A - to improve internal flooding response procedures and training to improve the Initiator response to internal flooding events has been evaluated.

BCHFROCHGB250QJG 1.0041 Charger IV 0-chgb-250-qj-g This term represents the failure of 250V battery charger. Phase II SAMA 218, to increase the reliability of power supplies has been evaluated. As this is a hardware failure no additional low cost procedural SAMA was identified.

PMAFD2PMP_0030128 1.0041 MDAFWP B FAILS TO This term represents MDAFWP 1B-B failing to run. Phase I SAMA 223 to START ON DEMAND improve the reliability of the AFW pumps and valves has been implemented.

Phase II SAMA 068 to add a motor driven feedwater pump has been evaluated. As this is a hardware failure no additional low cost procedural SAMA was identified.

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%690.0-A01-4_059_S 1.004 DEMIN WATER SPRAY This event represents the initiator for a demin water spray event in partition EVENT IN 690.0-Al - 4 of room 690.0-Al. This initiator results in loss of MDAFW B as well as the Initiator BAT/AFW area Space Coolers. Phase II SAMA 275 to install spray shields on the AFW pumps and AFW/BAT Space Cooler has been evaluated.

MOCXC2FCV_07000041E 1.004 FCV-70-4 TRANSFERS This term represents CCS flow control valve transferring close resulting in CLOSED the loss of CCS. Phase II SAMA 45 to enhance procedural guidance for use of cross-tied component cooling pumps has been evaluated. As this is a hardware failure no additional low cost procedural SAMA was identified.

MOCXC2FCV_07000151E 1.004 FCV 70-15 TRANSFERS This term represents CCS flow control valve transferring close resulting in CLOSED the loss of CCS. Phase II SAMA 45 to enhance procedural guidance for use of cross-tied component cooling pumps has been evaluated. As this is a hardware failure no additional low cost procedural SAMA was identified.

MOCXC2FCV_07000161E 1.004 FCV 70-16 TRANSFERS This term represents CCS flow control valve transferring close resulting in CLOSED the loss of CCS. Phase II SAMA 45 to enhance procedural guidance for use of cross-tied component cooling pumps has been evaluated. As this is a hardware failure no additional low cost procedural SAMA was identified.

SHESUM 1.0039 Remove Drain Plugs from This term represents the human action to remove drain plugs from the Refueling Canal After refueling canal after refueling. Phase II SAMAs 103 and 283 to increase Refueling training and improve awareness of important operator actions have been evaluated. While this event was not included in these evaluations, SAMA 283 was determined to be potentially cost beneficial and the event could be added to those addressed in that SAMA.

TMU2PMP0740010 1.0039 RHR Pump 2B-B in Test This term represents the maintenance unavailability of RHR Pump 2B-B.

and Maintenance Phase II SAMA 278 improving the reliability of the RHR pumps has been evaluated. As this is a hardware failure no additional low cost procedural SAMA was identified.

U2_PR-INS 1.0038 Insufficient pressure relief This term represents insufficient pressure relief through PORVs. Phase II SAMAs 103 and 283 to increase training and improve awareness of important operator actions, including actions to depressurize, have been evaluated. No additional low cost procedural SAMA was identified to address the reliability and capacity of these components.

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%690.0-AO1-2_067_F_2A 1.0036 ERCW flood event in This event represents the initiator for an ERCW flood event in partition 2 of 690.0-AO1-2 from the 2A-A room 690.0-Al. Phase II SAMA 279 to improve internal flooding response header with no isolation. - procedures and training to improve the response to internal flooding events Initiator has been evaluated.

%2SLOCA-CL2 1.0035 SLOCA ON COLD LEG 2 This event represents a small break LOCA on cold leg 2. Plant response to a small break LOCA would be to align high pressure recirculation and depressurize. Phase II SAMA 032 to automatically align high pressure recirculation has been evaluated. Phase II SAMAs 103 and 283 to increase training and improve awareness of important operator actions have been evaluated.

%2SLOCA-CL3 1.0035 SLOCA ON COLD LEG 3 This event represents a small break LOCA on cold leg 3. Plant response to a small break LOCA would be to align high pressure recirculation and depressurize. Phase II SAMA 032 to automatically align high pressure recirculation has been evaluated. Phase II SAMAs 103 and 283 to increase training and improve awareness of important operator actions have been evaluated.

%662.0-1_026_M-1 1.0035 HPFP MAJOR FLOOD This event represents the initiator for a flood in the Turbine Building which EVENT IN 662.0-1 DOES originates from High Pressure Fire Protection (HPFP) system piping. This NOT AFFECT THE TB flood does not affect the Turbine Building electrical distribution boards.

DISTRIBUTION BOARDS - Phase II SAMA 285, to protect equipment in the Turbine Building from I internal flooding effects has been evaluated. Phase II SAMA 279 to improve internal flooding response procedures and training to improve the response to internal flooding events has been evaluated.

%690.0-AO1-4_070_S 1.0034 CCS SPRAY EVENT IN This event represents the initiator for a CCS spray event in partition 4 of 690.0-Al - Initiator room 690.0-Al. This initiator results in loss of MDAFWP 2B-B as well as the BAT/AFW area Space Coolers. Phase II SAMA 275 to install spray shields on the AFW pumps and AFW/BAT Space Cooler has been evaluated.

UO06_250_BCHFRVBBALL 1.0034 CCF of all components in This event represents the common cause failure of 250V Vital Battery group Chargers. Phase II SAMA 218 to improve the reliability of power supplies

'UO 06_250_BCHFRVBB' has been evaluated. As this is a hardware failure no additional low cost procedural SAMA was identified.

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%714.0-AO1-2_067_F lA 1.0031 ERCW FLOOD EVENT IN This event represents the initiator for an ERCW flood event in room 714.0-714.0-Al-2 FROM 1A-A Al. Phase II SAMA 279 to improve internal flooding response procedures HEADER - Initiator and training to improve the response to internal flooding events has been evaluated..

TM_2PMP_062104 1.0031 CCP B in Maintenance This term represents the time that CCP B is in maintenance. Given the low RRW of this basic event, corresponding to approximately $50,000 for eliminating all maintenance, there would be no minor procedural changes that would result in a cost beneficial SAMA for improving the amount of time this pump is in maintenance.

AHUFD2CLR_0300175 1.003 Air Handling Unit (Standby) This event represents the failure of the standby RHR Pump Room A AHU to Fails to Start (2-CLR-030- run. Phase II SAMA 160 to implement procedures for temporary HVAC, for 0175) rooms including RHR Pump Room A, has been evaluated. As this is a hardware failure no additional low cost procedural SAMA was identified.

AHUFR1CLR_0300191 1.003 CCS PUMPS AND AFW This term represents the failure of CCS pumps and AFW pumps space PUMPS SPACE COOLER coolers. SAMA 268 and SAMA 289 evaluate the cooling requirements for B FAILS TO RUN the CCS/AFW pump space coolers and the potential for installing backup cooling to these coolers. As this is a hardware failure no additional low cost procedural SAMA was identified.

U2_XPORV1 1.003 BLOCK VALVE CLOSED This term represents the probability of a block valve being closed to isolate TO ISOLATE LEAKING a leaking PORV. Phase II SAMA 284 to reduce probability that the PORV pressurizer safety relief valves fail to close was evaluated.

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NRC RAI 7.e.ii ii. As stated in Section E. 1.1.1 and the SQN IPEEE Safety Evaluation Report (SER), the limiting plant component (HCLPF)is 0. 23g which is less than the review level earthquake (RLE) of 0.3g. The TER (Table 4.1) supporting the SER indicates there are 12 components with HCLPFs below the RLE. While the NRC concluded that the SQN IPEEE meets the intent of to GL 88-20, Supplement 4, the result above indicates that there are some components which should be examined for the identificationof potential cost-beneficial SAMAs.

The TVA responses to the Fukushima Near-Term Task Force Report Recommendation 2.3: Seismic Response Report states that "The statuses of all IPEEE outliers which were not corrected through physical modification were resolved through re-calculationof the appropriateHCLPFcapacities. All IPEEE outliers are now resolved and have minimum HCLPF Capacitiesabove 0.3g."

1. Discuss the actions taken on these 12 items, a) the final HCLPF values, if available and b) the potential for cost-beneficialSAMAs for these SQN components.

2. Also, provide a discussion of the current status of the seismic reevaluationand walkdown activities being undertaken in response to the Fukushima Dai-ichi event.

TVA Response A technical review was conducted of the HCLPF capacities for those components that, during the IPEEE, were less than 0.3 g.

Each of the 12 noted components were re-analyzed. Eleven of the components met the 0.3 g requirement. One component, the 480V shutdown transformer, required a minor modification to the anchorage, which is complete.

TVA has reviewed the other low margin outliers in the calculation. No other minor modifications that would be cost beneficial for the reduction of seismic risk were identified.

The seismic design of SQN will be further evaluated by the ongoing Fukushima project requirements. Also, the improved external flooding mitigation provided by installing additional equipment to provide secondary feedwater and RCS makeup to both Units, all housed in a hardened bunker building (refer to the end of TVA response to RAI 7.a.i) will provide mitigation capability for a seismic event risk reduction.

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NRC RAI 7.e.iii iii. Review of the fire CDF values in ER Table E. 1-16 indicates that the CDFfor the top 14 fire areas is greaterthan that which if mitigated would have a benefit of the minimum hardware cost of $100,000. While four of these fire areas are addressedby SAMA 287, provide a discussion of the potential for cost beneficial SAMAs for the other fire areas.

TVA Response As indicated in the RAI, four of the 14 top fire areas from the SQN fire IPEEE were addressed in SAMA 287. The remaining ten areas have fire CDF ranging from approximately 1 E-07 to 9.8E-07/yr. It is not possible to eliminate all of the fire risk from each of these areas, but it may be possible to significantly reduce the contribution from severe fires. Therefore, the benefit of potential SAMAs to remove or significantly reduce the risk of severe fires was estimated by assuming severe fires account for 30% of the fire risk and eliminating it from the fire CDF for each area. Assuming that the benefit is proportional to the change in CDF, benefits for these ten fire areas range from approximately $4,200 to $26,000. The minimum hardware modification of $100,000 is well above the benefit of each individual fire area and therefore, SAMAs on an individual fire area bases are not cost beneficial. The modification for a SAMA to reduce the risk of fires is expected to involve fire wrap of cable, re-routing of cable, addition of fire suppression or some combination of the three for each of the fire areas. Therefore, a SAMA modification to address all of the fire areas is expected to be much greater than the minimum hardware cost based on estimates for SAMA 287, and is also considered not cost beneficial. A similar evaluation for Unit 2 results in similar results and the same conclusions. No additional low cost SAMAs were identified.

NRC RAI 7.e.iv iv. One of the screening criteriagiven in ER Section E.2.2 is:

Excessive Implementation Cost: If the estimated cost of implementation is greaterthan the modified Maximum Averted Cost-Risk, the SAMA cannot be cost beneficial and is screened from further analysis.

If the uncertaintymultiplier of 2.5 was not consideredin performing this screening discuss the impact on the screeningresults.

TVA Response The uncertainty multiplier of 2.5 was not considered during the Phase I SAMA candidate screening. However, the selection of the screening criteria "Excessive Implementation Cost" was limited to cases where the estimated cost of implementation approached or exceeded the modified Maximum Averted Cost Risk. An additional review of the Phase I SAMA candidates screened as "Excessive Implementation Cost" was performed to identify any candidates that should be reconsidered with the uncertainty multiplier of 2.5 applied.

The Phase I candidates were grouped into categories based on how the potential SAMA would affect the plant. The categories were SGTR, Injection Capabilities, Containment Response/Venting, Reactor Vessel, and AC/SBO. A bounding maximum potential benefit was developed for each of the categories. These potential benefits, including the uncertainty multiplier, were used to perform a review of the SAMA candidates that had previously been screened based on high implementation costs. The results of the review are below.

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SGTR The Phase I SAMA candidates screened as "Excessive Implementation Costs" that prevent or mitigate the consequences of SGTR remain not cost beneficial even when applying the 95th percentile uncertainty to the bounding benefit.

The maximum benefit for the SGTR SAMA candidates was determined to be

$637,654 ($1,594,135 with 9 5th percentile uncertainty). This benefit was conservatively estimated by eliminating all releases from Release Categories III and V.

Injection The Phase I SAMA candidates screened as "Excessive Implementation Costs" that improve injection capabilities remain not cost beneficial even when applying the 95th percentile uncertainty to the bounding benefit. The maximum benefit for the injection SAMA candidates was determined to be $27,323 ($68,307 with 9 5 th percentile uncertainty). This benefit was conservatively estimated by eliminating the top events for failure of SI.

Containment The Phase I SAMA candidates screened as "Excessive Implementation Costs" that reduce the probability of containment failure remain not cost beneficial even when applying the 95th percentile uncertainty to the bounding benefit. The maximum benefit for the containment SAMA candidates was determined to be

$1,070,829 ($2,677,072 with 9 5 th percentile uncertainty). This benefit was conservatively estimated by reducing the frequency of all large early releases (Release Categories I and II) by one-half.

Reactor The Phase I SAMA candidates screened as "Excessive Implementation Costs" that reduce the probability of reactor vessel failure remain not cost beneficial even when applying the 95th percentile uncertainty to the bounding benefit.

While, the maximum benefit for the reactor SAMA candidates was not quantified, the cost of implementing these SAMA candidates are very high, especially for an existing plant, and would remain not cost-beneficial.

AC/SBO The Phase I SAMA candidates screened as "Excessive Implementation Costs" that reduce the probability of SBO and LOOP events remain not cost beneficial even when applying the 95th percentile uncertainty to the bounding benefit.

The maximum benefit for the SBO/LOOP SAMA candidates was determined to be $588,067 ($1,470,168 with 9 5th percentile uncertainty). This benefit was conservatively estimated by eliminating the failure of the Common Service Station Transformers, the 6.9 kV Shutdown Boards, and the Safety Related 125V DC Vital Boards.

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NRC RAI 7.e.v

v. The dispositionsgiven in the correlationof risk significant terms to SAMAs in ER Tables E.1-3 and E. 1-4 do not include Phase II SAMAs for some of the events. While it is stated that Phase I SAMAs have been implemented (to the extent that the implementations are reflected in the current PRA), the following events appearsignificant enough to warrant further review for additionalSAMAs that mitigate the specific failures representedby these events.

1. For events PTSFD1PMP_0030142and PTSFR1PMP_0030142, representingthe random failure of the turbine driven auxiliary feed water pump to start or run, 4 Phase I SAMAs to improve pump reliabilityare stated to have been implemented. These failures contribute about 8% and 1% of the CDF, respectively.

Discuss the implemented SAMAs and the potential for other candidate SAMAs to mitigate these failures.

2. For event AFWOP3, representingthe failure of operatorsto depressurizeand cool down the vessel in orderfor low pressure injection to be used following a small or medium loss of coolant accident with failure of high pressure recirculation,three Phase I SAMAs to improve the capacity to cool down and depressurizationare stated to have been implemented.

Discuss the implemented SAMAs and the potential for other candidate SAMAs to mitigate this operatorerror.

3. For event TM_ 1PMP_003001AS, representingthe maintenance unavailabilityof the turbine driven AFW pump, two Phase I SAMAs to improve the reliability of the AFW turbine driven pump have been implemented.

Discuss the implemented SAMAs and the potential for other candidate SAMAs to mitigate this unavailability.

.4. For event TM_ 1PMPO030118A, representingthe maintenance unavailabilityof motor driven AFW Pump 1A-A, one Phase I SAMA to improve the reliability of the AFW pumps and valves has been implemented.

Discuss the implemented SAMA and the potential for other candidate SAMAs to mitigate this unavailability.

5. For event PMAFDIPMP_00300118,representingthe random failure of motor driven AFW Pump lA-A failing to start, two Phase I SAMAs are stated to have been implemented.

Discuss the implemented SAMAs and the potential for other candidate SAMAs to mitigate this failure.

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TVA Response The events included in Items 7.e.v.1 and 7.e.v.5 represent hardware failures of the MDAFWPs while those in Items 7.e.v.3 and 7.e.v.4 represent testing and maintenance unavailability of AFW pumps The events all identify SAMA 223 "Improve reliability of AFW pumps and valves" as an applicable SAMA. This Phase I SAMA was dispositioned as follows:

"The SQN AFW systems meet reliability and unavailability goals established in the maintenance rule program. To improve reliability there are initiatives to upgrade the Terry Turbine Governor Controls and Governor Valve stem material; obtain spares for MDAFWP, TDAFWP, MDAFWP motor; replace Bailey 550 transmitters to increase the reliability of holding as found and as left tolerances. Therefore, implementation of this SAMA is an ongoing process at SQN."

A new reactor decay heat removal system will be installed to address external flooding issues (See the response to RAI 7.a.i). This system includes newly designed means for removing reactor decay heat [during flooding scenarios] and providing reactor coolant system makeup for both SQN Units. The installation of this system will significantly reduce the importance of the listed AFW events.

For RAI 7.e.v.2, event, AFWOP3 references SAMA 42 "Make procedure changes for RCS depressurization." The disposition of SAMA 42 was as follows:

"TVA procedures, FR-C1, Inadequate Core Cooling, FR-C2, Degraded Core Cooling, and SAG-2, Depressurize the RCS, contain guidance to depressurize the RCS and allow RCS makeup from low pressure injection sources. Therefore, SAMA 42 is determined to be already implemented."

The above procedures are consistent with the latest guidance from the Westinghouse Owners Group. In addition, human failure event AFWOP3 was addressed in Phase II SAMAs 103 "Institute simulator training for severe accident sequences" and 283 "Initiate frequent awareness training for plant operators/ maintenance/ testing staff on important human actions, including dependent (combination) events, for plant risk."

Therefore, the above basic events have been addressed by Phase II SAMAs or will be addressed by ongoing plant improvement initiatives.

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NRC RAI 7.e.vi vi. For basic events %1RTIE and %1TTIE representinga general reactortrip and a turbine trip, respectively, it is stated in ER Tables E.1-3 and E.1-4 that Phase II SAMA 218, to increase the reliabilityof power supplies, has been evaluated.

Discuss the potential for other SAMAs to reduce the general reactortrip and turbine trip frequency.

TVA Response One potential SAMA for reducing the general reactor trip and turbine trip frequencies is the implementation of a trip reduction program focused on these two initiators. Given current nuclear plant trends, it is estimated that such a program could only result in a decrease in frequency of at most 20%. When a 20% reduction in frequency is applied to both initiators, the Unit 1 internal events CDF is reduced by approximately 2.9E-07/yr. Assuming the maximum averted risk cost is proportional to the CDF, the reduction in CDF translates into a benefit of approximately $26,000 ($65,000 when 9 5th percentile uncertainty is considered). Similar results are obtained for Unit 2. Based on previous TVA experience with the development and implementation of reliability studies, a trip reduction program is estimated to cost between

$550,000 and $1,250,000. The higher range of the cost estimate includes the development of detailed modeling of reactor and turbine trip failures. Based on the estimated cost to implement a trip reduction program and the minimal benefit gained, this potential SAMA would not be cost beneficial.

NRC RAI 7.e.vii vii. One source of Phase I SAMAs is indicated to be the October 2010 Watts Bar Unit 2 severe accident management design alternativesSAMDA submittal.

Considerany additionalcost-beneficial SAMAs (SAMDAs) identified during the review of this submittal as documented in TVA responses to RAIs and the draft NRC staff review of the SAMDA analysis (if available).

TVA Response The initial list of potential Phase I SAMAs for SQN was based on the WBN Unit 2 SAMDA Phase I list. The only difference was that items on the WBN Unit 2 SAMDA Phase I list because of risk reduction worth from the WBN Unit 2 PRA, were not included in the initial list of potential Phase I SAMAs for SQN. The initial list of potential Phase I SAMAs for SQN was expanded by including SAMAs for important events from the SQN PRAs and potentially cost beneficial SAMAs identified from reviews of other plant's SAMA evaluations.

The WBN Unit 2 responses to RAIs were also reviewed to identify any applicable potentially cost beneficial SAMAs. The RAI response dated September 16, 2011, (Adams No. ML11264A052) contains a consolidated list of SAMAs for the WBN Unit 2 analysis. Final approval of the WBN Unit 2 SAMDA has not been given at this time, but the latest draft of NUREG-0498, Supplement 2 (ADAMS Accession No. ML11298A095) dated September 2011 was also reviewed. No new additional SAMA candidates for SQN were identified as a result of these reviews.

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NRC RAI 7.f.i

f. Provide the following information with regardto the Phase II cost-benefit evaluations.

Basis: Applicants for license renewal are required by 10 CFR 51.53(c)(3)(ii)(L) to consider SAMAs if not previously considered in an environmental impact assessment, related supplement, or environmental assessment for the plant. As part of its review of the SQN SAMA analysis, NRC staff evaluates the applicant'scost benefit analysis of Phase II SAMAs. The requested information is needed in order for the NRC staff to reach a conclusion on the acceptabilityof the applicant'scost estimations for individualSAMAs and cost benefit evaluation.

i. Identify what is included and what is not included in the SQN specific cost estimates including such things as contingency, replacement power, lifetime maintenance, etc.

TVA Response The cost estimates for SQN SAMA candidates were developed by individuals who previously worked in TVA engineering, operations, and maintenance. The cost estimates assumed 2012 dollars and included contingency costs and capital overhead. Escalation, replacement power costs, and inflation were not considered in the estimate. Cost estimates from past projects were used when applicable. For cost estimates that were not based directly from past projects, itemized cost estimates were developed, where applicable and appropriate. Specific hardware costs from recent projects such as piping, valves, electrical cable, and switchgear were used when applicable. Additional costs for the lifetime maintenance costs of SAMA related hardware was not considered when developing implementation costs. Procedure change costs were developed based on man-hours for all the processing needed to update the procedures. The implementation cost for SAMAs requiring recurring procedure maintenance and training also included the lifetime cost of annual procedure reviews and increased training in the estimates. Engineering estimates were based on typical man-hour costs for design changes. Training costs were developed based on the man-hours needed to prepare operator training materials. Simulator modification costs were based on estimated cost to add event scenarios or to modify the simulator. Cost input was received from the electrical, mechanical, and civil disciplines as required. The cost estimates were reviewed by the project manager and/or the discipline engineering managers when warranted. Unless otherwise stated, all costs are those for a single unit.

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NRC RAI 7.f.ii ii. Discuss the impact of sharing engineering and design cost between units for the SQN specific costs as well as the costs based on otherplant's SAMA analyses.

TVA Response For plant modifications that would provide benefit to both units (e.g., SAMA 286: Install Flood Doors to Prevent Water Propagation in the Electric Board Room), the averted cost risk from Unit 1 and Unit 2 were combined to provide a total averted cost risk for SQN. The implementation cost for these SAMAs could be shared between the two units.

For the remaining Phase II SAMA candidates, a review was performed of the Non-Cost Beneficial SAMAs. These SAMA candidates fell into one of three categories.

1. The SAMA was determined to be potentially cost beneficial as a result of the sensitivity analyses. The sensitivity analyses identified nine additional SAMA candidates as Potentially Cost Beneficial for one of the two units (SAMAs 032, 087, 088, 160, 249, 275, 285, 286, 288, and 289). With the exception of SAMA 160 (Temporary Ventilation), the engineering and design costs of these candidates could likely be shared between the two units.

2. The Unit 1 and Unit 2 combined internal and external benefit, or averted cost, of the SAMA implementation is less than the estimated cost of implementation for a single unit.

Therefore, even when conservatively combining the 9 5 th percentile sensitivity benefits for these SAMAs, implementation remains non cost-beneficial.

3. Five Phase II SAMA candidates do not fall into either of the categories above. They are SAMA 109 (Install a passive hydrogen control system), SAMA 136 (Install motor generator set trip breakers in control room), SAMA 137 (Provide capability to remove power from the bus powering the control rods), SAMA 218 (Improve reliability of power supplies to reduce reactor trip), and SAMA 278 (Improve reliability of the RHR pumps and improve maintenance procedures to reduce potential for common cause failure).

The 95th percentile sensitivity benefits for SAMA 109 are $2,232,324 for Unit 1 and

$2,028,665 for Unit 2 for a combined averted cost risk of $4,260,989. The estimated cost of implementation for SAMA 109 for a single unit is $3,736,000. There may be some design and procedure development costs that could be shared between the two units. However, these costs would not be significant enough to make implementation of this SAMA cost-beneficial at either unit. The hardware, installation and lifetime maintenance cost associated with a passive hydrogen control system would exceed the 95th percentile benefit of each unit such that implementation of this SAMA, even when considering shared costs between the units, would not be cost-beneficial.

The 9 5 th percentile sensitivity benefits for SAMA 136 are $64,009 for Unit 1 and $56,183 for Unit 2 for a combined averted cost $120,192. Due to the low benefit for SAMA 136, the estimated cost was not an actual cost estimate, but was the minimum assumed hardware cost for any plant change (i.e., $100,000). The actual estimated cost for implementation of this SAMA (from a WBN 2 estimate) is $241,795, which is much greater than the Unit 1 and Unit 2 combined averted cost risk. Sharing engineering and design costs for SAMA 136 would not change the cost-beneficial status of this SAMA.

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Because the same analysis was used for SAMAs 136 and 137, the 9 5 th percentile sensitivity benefits for SAMA 137 are $64,009 for Unit 1 and $56,183 for Unit 2 for a combined averted cost $120,192. Similar to SAMA 136, because the benefit for SAMA 137 was so low, the estimated cost was not an actual cost estimate, but was the minimum assumed hardware cost for any plant change (i.e., $100,000). In addition to training (a cost which would be recurring for the life of the plant) and procedure changes, implementation of this SAMA would require reevaluation of the loss of power to other loads on the unit boards. This reevaluation, increased training and procedure change costs would easily drive the true implementation cost for this SAMA well above the combined averted cost risk of $120,192. The original cost estimate of $100,000 was chosen as a minimum cost because of the low benefit for implementation of this SAMA.

Sharing engineering and design costs for SAMA 137 would not change the cost-beneficial status of this SAMA.

The 9 5 th percentile sensitivity benefits for SAMA 218 are $419,814 for Unit 1 and

$351,756 for Unit 2 for a combined averted cost $771,570. The implementation cost of SAMA 218 is comprised of the analysis, hardware costs, engineering, procedures and implementation. While there may be some shared costs between the units for engineering, analysis and procedures, each unit would still require unit specific analysis and procedure changes. Additionally, the costs associated with hardware and implementation, are estimated to comprise 75% of the total costs. Conservatively assuming that all of the remaining 25% of the implementation costs could be shared between the two units (a 12.5% decrease in the total cost for each unit), implementation of this SAMA would still remain not cost-beneficial.

There may be some aspects of implementing SAMA 278 (i.e., costs associated with revising and enhancing maintenance procedures) that could be shared between the units. However, given the conservative nature of this SAMA analysis, the benefit compared to the cost, and the likelihood that some hardware would still be required to improve the reliability of the RHR pumps, this SAMA would also remain not cost-beneficial. The analysis for SAMA 278 was performed by reducing by 50 percent the fail to run, fail to start, common cause, and unavailability due to maintenance events for all of the RHR pumps. This conservative analysis resulted in 95th percentile sensitivity benefits of $264,526 and $290,579 for Units 1 and 2, respectively, for a total combined benefit of $555,105. If a significant portion of the annual costs for implementation of this SAMA could be shared, then the SAMA may potentially be cost-beneficial. However, due to planned modifications to improve external flooding mitigation (refer to RAI 7.a.i),

the benefit of this SAMA will be significantly reduced because of the additional train of decay heat removal being installed. Therefore, the cost beneficial status of this SAMA is not changed due to shared engineering and design costs.

NRC RAI 7.f.iii iii. Provide the release category frequencies for each Phase II SAMA.

TVA Response Release Category Frequencies for each unit are presented below.

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Unit 1 Release Category Frequencies Total SAMA No RC-la RC-Ib RC-lc RC-Ila RC-IIb RC-lIc RC-IId RC-III RC-IVa RC-IVb RC-Va RC-Vb CDF Total Release Frequency BaseCase 4.06E-08 9.70E-07 2.65E-07 3.26E-06 6.29E-07 6.47E-08 4.77E-08 6.43E-07 1.79E-05 2.23E-06 2.08E-06 1.13E-06 2.96E-05 2.93E-05 8 N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A 14 4.06E-08 9.70E-07 2.65E-07 3.26E-06 2.58E-08 5.75E-07 4.69E-08 6.43E-07 1.79E-05 2.24E-06 2.08E-06 1.13E-06 2.95E-05 2.92E-05 32 2.21E-08 9.52E-07 8.11E-08 3.26E-06 6.29E-07 6.19E-08 4.28E-08 5.05E-07 1.79E-05 2.21E-06 2.08E-06 6.33E-07 2.56E-05 2.84E-05 45 4.06E-08 9.49E-07 2.65E-07 3.26E-06 6.29E-07 5.74E-08 4.82E-08 6.43E-07 1.76E-05 2.20E-06 2.05E-06 1.13E-06 2.94E-05 2.89E-05 46 4.06E-08 9.69E-07 2.65E-07 3.26E-06 5.04E-07 6.46E-08 4.77E-08 6.43E-07 1.78E-05 2.22E-06 2.07E-06 1.13E-06 2.94E-05 2.91E-05 55 4.06E-08 9.65E-07 2.65E-07 3.26E-06 2.84E-07 4.25E-08 4.98E-08 6.43E-07 1.74E-05 2.22E-06 1.98E-06 1.16E-06 2.87E-05 2.83E-05 56 4.06E-08 9.65E-07 2.65E-07 3.26E-06 2.84E-07 4.25E-08 4.98E-08 6.43E-07 1.74E-05 2.22E-06 1.98E-06 1.16E-06 2.87E-05 2.83E-05 68 3.99E-08 9.49E-07 1.24E-08 3.26E-06 6.29E-07 5.92E-08 4.17E-09 3.90E-07 1.74E-05 1.20E-07 2.02E-06 4.32E-07 2.36E-05 2.53E-05 70 4.05E-08 9.70E-07 4.03E-08 3.26E-06 6.29E-07 6.47E-08 4.51E-08 4.77E-07 1.79E-05 2.21E-06 2.08E-06 9.07E-07 2.78E-05 2.86E-05 71 4.06E-08 9.70E-07 2.65E-07 3.26E-06 2.63E-07 6.47E-08 4.77E-08 6.43E-07 1.79E-05 2.23E-06 2.08E-06 1.13E-06 2.92E-05 2.89E-05 79 4.06E-08 9.70E-07 2.65E-07 3.26E-06 6.29E-07 6.47E-08 4.77E-08 6.43E-07 1.79E-05 2.23E-06 2.08E-06 1.13E-06 2.96E-05 2.93E-05 83 4.06E-08 9.70E-07 2.65E-07 3.26E-06 6.29E-07 6.38E-08 4.77E-08 6.43E-07 1.79E-05 2.23E-06 2.08E-06 1.13E-06 2.96E-05 2.93E-05 87 4.05E-08 9.70E-07 7.33E-08 3.26E-06 6.29E-07 6.47E-08 4.51E-08 4.99E-07 1.79E-05 2.21E-06 2.08E-06 8.87E-07 2.77E-05 2.87E-05 88 2.80E-08 9.60E-07 2.65E-07 3.26E-06 6.29E-07 6.45E-08 4.77E-08 6.42E-07 1.79E-05 2.23E-06 2.08E-06 1.87E-06 2.86E-05 2.91E-05 103 3.73E-08 8.64E-07 2.76E-07 3.26E-06 6.29E-07 6.27E-08 4.57E-08 6.16E-07 1.67E-05 2.12E-06 1.97E-06 1.05E-06 2.80E-05 2.76E-05 105 3.13E-08 9.61E-07 1.73E-07 3.26E-06 6.29E-07 6.33E-08 4.50E-08 5.45E-07 1.79E-05 2.22E-06 2.08E-06 8.81E-07 2.76E-05 2.88E-05 106 3.13E-08 9.61E-07 1.73E-07 3.26E-06 6.29E-07 6.33E-08 4.50E-08 5.45E-07 1.79E-05 2.22E-06 2.08E-06 8.81E-07 2.76E-05 2.88E-05 109 4.06E-08 1.79E-07 4.59E-08 3.26E-06 6.29E-07 6.47E-08 4.77E-08 6.43E-07 1.43E-05 2.23E-06 2.08E-06 1.13E-06 2.96E-05 2.46E-05 111 4.06E-08 9.70E-07 2.65E-07 3.26E-06 6.29E-07 6.47E-08 4.77E-08 6.15E-07 1.79E-05 2.23E-06 2.08E-06 1.13E-06 2.96E-05 2.93E-05 112 4.06E-08 9.70E-07 2.65E-07 3.26E-06 6.29E-07 6.24E-08 4.69E-08 6.43E-07 1.79E-05 2.23E-06 2.08E-06 1.13E-06 2.96E-05 2.93E-05 136 4.03E-08 9.68E-07 2.65E-07 3.26E-06 6.29E-07 6.47E-08 4.64E-08 6.19E-07 1.79E-05 2.23E-06 2.08E-06 1.13E-06 2.95E-05 2.93E-05 137 4.03E-08 9.68E-07 2.65E-07 3.26E-06 6.29E-07 6.47E-08 4.64E-08 6.19E-07 1.79E-05 2.23E-06 2.08E-06 1.13E-06 2.95E-05 2.93E-05 147 4.06E-08 9.69E-07 2.65E-07 3.26E-06 6.29E-07 6.47E-08 4.77E-08 6.43E-07 1.79E-05 2.23E-06 2.07E-06 1.13E-06 2.96E-05 2.93E-05 160 4.01E-08 7.95E-07 2.65E-07 3.26E-06 6.29E-07 3.90E-08 4.94E-08 6.38E-07 1.52E-05 2.22E-06 1.74E-06 1.13E-06 2.69E-05 2.60E-05 161 4.06E-08 9.70E-07 2.65E-07 3.26E-06 4.72E-07 6.37E-08 4.32E-08 6.43E-07 1.79E-05 2.22E-06 2.08E-06 1.13E-06 2.95E-05 2.91E-05 El- 75 of 87

167 4.06E-08 9.70E-07 2.65E-07 3.26E-06 6.29E-07 6.47E-08 4.77E-08 6.43E-07 1.79E-05 2.23E-06 2.08E-06 1.13E-06 2.96E-05 2.93E-05 188 2.80E-08 9.60E-07 4.03E-08 3.26E-06 6.17E-07 6.45E-08 4.38E-08 4.69E-07 1.79E-05 2.20E-06 2.08E-06 7.14E-07 2.63E-05 2.84E-05 215 3.59E-08 9.99E-08 2.63E-07 3.26E-06 2.99E-07 2.02E-08 4.98E-08 6.43E-07 1.82E-06 2.22E-06 3.85E-07 1.09E-06 1.55E-05 1.02E-05 218 4.03E-08 9.62E-07 2.65E-07 3.26E-06 6.23E-07 6.17E-08 4.60E-08 6.42E-07 1.72E-05 2.21E-06 1.98E-06 1.12E-06 2.88E-05 2.85E-05 226 4.06E-08 9.65E-07 2.65E-07 3.26E-06 2.79E-07 4.22E-08 4.98E-08 6.36E-07 1.74E-05 2.22E-06 2.01 E-06 1.13E-06 2.87E-05 2.83E-05 239 4.06E-08 9.70E-07 2.65E-07 3.26E-06 6.29E-07 6.47E-08 4.77E-08 6.15E-07 1.79E-05 2.23E-06 2.08E-06 1.13E-06 2.96E-05 2.93E-05 240 4.06E-08 9.65E-07 2.65E-07 3.26E-06 2.79E-07 4.22E-08 4.98E-08 6.36E-07 1.74E-05 2.22E-06 2.01E-06 1.13E-06 2.87E-05 2.83E-05 249 3.13E-08 9.61E-07 1.73E-07 3.26E-06 6.29E-07 6.33E-08 4.50E-08 5.45E-07 1.79E-05 2.22E-06 2.08E-06 8.81E-07 2.76E-05 2.88E-05 254 4.06E-08 9.70E-07 2.65E-07 3.26E-06 5.72E-07 6.45E-08 4.61E-08 6.43E-07 1.79E-05 2.22E-06 2.08E-06 1.13E-06 2.95E-05 2.92E-05 268 4.01E-08 5.83E-07 2.40E-07 3.26E-06 6.29E-07 5.OOE-08 3.54E-08 6.29E-07 9.70E-06 2.60E-07 1.24E-06 8.65E-07 2.09E-05 1.75E-05 275 4.02E-08 8.27E-07 2.56E-07 3.26E-06 6.29E-07 6.20E-08 4.59E-08 6.38E-07 1.57E-05 1.99E-06 1.87E-06 1.04E-06 2.72E-05 2.64E-05 276 3.43E-08 9.66E-07 2.65E-07 3.26E-06 6.29E-07 6.27E-08 4.77E-08 6.28E-07 1.79E-05 2.23E-06 2.08E-06 1.07E-06 2.91 E-05 2.92E-05 277 3.72E-08 9.65E-07 2.65E-07 3.26E-06 6.29E-07 6.47E-08 4.75E-08 5.OOE-07 1.79E-05 2.23E-06 2.08E-06 1.10E-06 2.92E-05 2.91E-05 278 4.06E-08 9.38E-07 2.65E-07 3.26E-06 6.29E-07 6.13E-08 4.77E-08 6.43E-07 1.77E-05 2.22E-06 1.95E-06 1.13E-06 2.86E-05 2.89E-05 279 3.90E-08 6.98E-07 1.46E-07 3.26E-06 6.29E-07 6.46E-08 4.57E-08 5.56E-07 1.66E-05 2.21 E-06 2.08E-06 9.36E-07 2.80E-05 2.73E-05 283 3.73E-08 8.64E-07 2.76E-07 3.26E-06 6.29E-07 6.27E-08 4.57E-08 6.16E-07 1.67E-05 2.12E-06 1.97E-06 1.05E-06 2.80E-05 2.76E-05 284 3.78E-08 9.64E-07 2.65E-07 3.26E-06 6.29E-07 5.46E-08 4.77E-08 6.43E-07 1.77E-05 2.23E-06 2.01E-06 1.10E-06 2.89E-05 2.89E-05 285 4.06E-08 9.62E-07 8.24E-08 3.26E-06 6.29E-07 6.31E-08 4.51E-08 5.07E-07 1.71E-05 2.21E-06 1.98E-06 9.04E-07 2.70E-05 2.78E-05 286 4.06E-08 9.70E-07 2.65E-07 1.05E-08 6.30E-07 6.48E-08 8.62E-08 6.44E-07 1.79E-05 2.24E-06 2.08E-06 1.13E-06 2.64E-05 2.61E-05 287 3.85E-08 9.20E-07 2.52E-07 3.09E-06 5.97E-07 6.14E-08 4.53E-08 6.1OE-07 1.70E-05 2.11E-06 1.97E-06 1.07E-06 2.81E-05 2.78E-05 288 4.01E-08 7.56E-07 2.60E-07 3.26E-06 6.29E-07 6.05E-08 4.75E-08 6.41E-07 1.46E-05 2.OOE-06 1.77E-06 1.10E-06 2.70E-05 2.52E-05 289 4.03E-08 7.18E-07 2.47E-07 3.26E-06 6.29E-07 5.64E-08 4.08E-08 6.32E-07 1.22E-05 6.76E-07 1.48E-06 9.20E-07 2.32E-05 2.09E-05 El- 76 of 87

Unit 2 Release Category Frequencies Total SAMA No RC-la RC-Ib RC-Ic RC-lia RC-IIb RC-IIc RC-IId RC-III RC-IVa RC-Ivb RC-Va RC-Vb CDF Total Release Frequency BaseCase 4.55E-08 9.53E-07 3.86E-07 3.30E-06 3.26E-07 6.26E-08 6.76E-08 7.41E-07 1.74E-05 1.24E-06 2.01E-06 1.92E-06 3.51E-05 2.85E-05 8 N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A 14 4.55E-08 9.53E-07 3.86E-07 3.30E-06 1.36E-08 3.46E-07 6.84E-08 7.41E-07 1.74E-05 1.26E-06 2.01E-06 1.93E-06 3.51E-05 2.85E-05 32 2.55E-08 9.34E-07 6.72E-08 3.30E-06 3.26E-07 5.98E-08 5.35E-08 5.14E-07 1.74E-05 5.72E-07 2.01 E-06 5.12E-07 2.39E-05 2.58E-05 45 4.55E-08 9.29E-07 3.86E-07 3.30E-06 3.26E-07 6.21 E-08 6.76E-08 7.41 E-07 1.70E-05 1.24E-06 1.98E-06 1.92E-06 3.49E-05 2.80E-05 46 4.55E-08 9.52E-07 3.86E-07 3.30E-06 2.99E-07 6.25E-08 6.76E-08 7.41E-07 1.73E-05 1.23E-06 2.OOE-06 1.92E-06 3.50E-05 2.83E-05 55 4.55E-08 9.46E-07 3.86E-07 3.30E-06 2.61E-07 3.99E-08 7.02E-08 7.41E-07 1.66E-05 1.23E-06 1.91E-06 1.92E-06 3.43E-05 2.75E-05 56 4.55E-08 9.46E-07 3.86E-07 3.30E-06 2.61E-07 3.99E-08 7.02E-08 7.41E-07 1.66E-05 1.23E-06 1.91E-06 1.92E-06 3.43E-05 2.75E-05 68 4.48E-08 9.31E-07 1.50E-08 3.30E-06 3.26E-07 5.67E-08 6.31E-09 4.07E-07 1.69E-05 5.78E-08 1.95E-06 5.15E-07 2.34E-05 2.45E-05 70 4.55E-08 9.52E-07 1.59E-07 3.30E-06 3.26E-07 6.25E-08 6.12E-08 5.73E-07 1.74E-05 1.23E-06 2.01E-06 1.70E-06 3.33E-05 2.78E-05 71 4.55E-08 9.53E-07 3.86E-07 3.30E-06 3.18E-07 6.26E-08 6.58E-08 7.41E-07 1.74E-05 1.24E-06 2.01E-06 1.92E-06 3.51E-05 2.85E-05 79 4.55E-08 9.53E-07 3.86E-07 3.30E-06 3.26E-07 6.26E-08 6.76E-08 7.41E-07 1.74E-05 1.24E-06 2.01E-06 1.92E-06 3.51E-05 2.85E-05 83 4.55E-08 9.52E-07 3.86E-07 3.30E-06 3.26E-07 6.18E-08 6.76E-08 7.41E-07 1.74E-05 1.24E-06 2.00E-06 1.92E-06 3.51E-05 2.84E-05 87 4.54E-08 9.53E-07 1.92E-07 3.30E-06 3.23E-07 6.24E-08 6.17E-08 5.93E-07 1.74E-05 1.23E-06 2.01E-06 1.68E-06 3.31E-05 2.79E-05 88 3.21E-08 9.42E-07 3.86E-07 3.30E-06 3.26E-07 6.23E-08 6.76E-08 7.31E-07 1.74E-05 1.24E-06 2.01E-06 1.79E-06 3.40E-05 2.83E-05 103 4.20E-08 8.45E-07 3.48E-07 3.30E-06 3.26E-07 6.07E-08 6.43E-08 7.03E-07 1.61E-05 1.12E-06 1.90E-06 1.74E-06 3.28E-05 2.66E-05 105 3.54E-08 9.43E-07 2.26E-07 3.30E-06 3.26E-07 6.12E-08 6.02E-08 5.98E-07 1.74E-05 9.05E-07 2.01E-06 1.22E-06 2.95E-05 2.71E-05 106 3.54E-08 9.43E-07 2.26E-07 3.30E-06 3.26E-07 6.12E-08 6.02E-08 5.98E-07 1.74E-05 9.05E-07 2.01E-06 1.22E-06 2.95E-05 2.71E-05 109 4.55E-08 1.77E-07 9.45E-08 3.30E-06 3.26E-07 6.26E-08 6.76E-08 7.41E-07 1.38E-05 1.24E-06 2.01E-06 1.92E-06 3.51E-05 2.38E-05 111 4.55E-08 9.53E-07 3.86E-07 3.30E-06 3.26E-07 6.26E-08 6.76E-08 7.12E-07 1.74E-05 1.24E-06 2.01E-06 1.92E-06 3.51E-05 2.84E-05 112 4.55E-08 9.53E-07 3.86E-07 3.30E-06 3.26E-07 5.96E-08 6.64E-08 7.41E-07 1.74E-05 1.24E-06 2.01E-06 1.92E-06 3.51E-05 2.85E-05 136 4.53E-08 9.51E-07 3.86E-07 3.30E-06 3.26E-07 6.26E-08 6.67E-08 7.16E-07 1.74E-05 1.24E-06 2.01E-06 1.92E-06 3.50E-05 2.84E-05 137 4.53E-08 9.51E-07 3.86E-07 3.30E-06 3.26E-07 6.26E-08 6.67E-08 7.16E-07 1.74E-05 1.24E-06 2.01E-06 1.92E-06 3.50E-05 2.84E-05 147 4.55E-08 9.52E-07 3.86E-07 3.30E-06 3.26E-07 6.26E-08 6.76E-08 7.41E-07 1.74E-05 1.24E-06 2.00E-06 1.92E-06 3.51E-05 2.84E-05 160 4.55E-08 9.32E-07 3.79E-07 3.30E-06 3.26E-07 4.06E-08 5.07E-08 7.32E-07 1.71E-05 8.42E-07 1.89E-06 1.81E-06 3.33E-05 2.74E-05 161 4.55E-08 9.53E-07 3.86E-07 3.30E-06 2.42E-07 6.26E-08 6.11E-08 7.41E-07 1.74E-05 1.24E-06 2.01E-06 1.92E-06 3.50E-05 2.84E-05 El- 77 of 87

167 4.55E-08 9.53E-07 3.86E-07 3.30E-06 3.26E-07 6.26E-08 6.76E-08 7.41E-07 1.74E-05 1.24E-06 2.01E-06 1.92E-06 3.51E-05 2.85E-05 188 3.21E-08 9.42E-07 1.59E-07 3.30E-06 3.25E-07 6.24E-08 6.05E-08 5.56E-07 1.74E-05 1.23E-06 2.01E-06 1.50E-06 3.17E-05 2.76E-05 215 4.OOE-08 1.01E-07 3.85E-07 3.30E-06 2.64E-07 1.83E-08 7.02E-08 7.41E-07 1.70E-06 1.23E-06 3.72E-07 1.86E-06 2.16E-05 1.01E-05 218 4.52E-08 9.45E-07 3.86E-07 3.30E-06 3.23E-07 5.98E-08 6.60E-08 7.39E-07 1.68E-05 1.23E-06 1.92E-06 1.92E-06 3.43E-05 2.77E-05 226 4.55E-08 9.46E-07 3.86E-07 3.30E-06 2.60E-07 3.96E-08 7.02E-08 7.34E-07 1.66E-05 1.23E-06 1.91 E-06 1.92E-06 3.42E-05 2.75E-05 239 4.55E-08 9.53E-07 3.86E-07 3.30E-06 3.26E-07 6.26E-08 6.76E-08 7.12E-07 1.74E-05 1.24E-06 2.01E-06 1.92E-06 3.51E-05 2.84E-05 240 4.55E-08 9.46E-07 3.86E-07 3.30E-06 2.60E-07 3.96E-08 7.02E-08 7.34E-07 1.66E-05 1.23E-06 1.91 E-06 1.92E-06 3.42E-05 2.75E-05 249 3.54E-08 9.43E-07 2.26E-07 3.30E-06 3.26E-07 6.12E-08 6.02E-08 5.98E-07 1.74E-05 9.05E-07 2.01E-06 1.22E-06 2.95E-05 2.71E-05 254 4.55E-08 9.53E-07 3.86E-07 3.30E-06 3.OOE-07 6.25E-08 6.53E-08 7.41E-07 1.74E-05 1.24E-06 2.01E-06 1.92E-06 3.51E-05 2.84E-05 268 4.49E-08 4.89E-07 3.84E-07 3.30E-06 3.26E-07 4.88E-08 6.73E-08 7.40E-07 8.14E-06 1.01E-06 1.08E-06 1.91E-06 2.76E-05 1.75E-05 275 4.51E-08 8.07E-07 3.OOE-07 3.30E-06 3.26E-07 5.98E-08 5.88E-08 6.83E-07 1.52E-05 1.24E-06 1.80E-06 1.34E-06 2.89E-05 2.51E-05 276 3.62E-08 9.46E-07 3.86E-07 3.30E-06 3.26E-07 5.97E-08 6.76E-08 7.26E-07 1.74E-05 1.24E-06 2.01 E-06 1.84E-06 3.44E-05 2.83E-05 277 4.24E-08 9.48E-07 3.86E-07 3.30E-06 3.26E-07 6.26E-08 6.74E-08 5.96E-07 1.74E-05 1.24E-06 2.01 E-06 1.89E-06 3.47E-05 2.83E-05 278 4.55E-08 9.18E-07 3.86E-07 3.30E-06 3.26E-07 5.91E-08 6,76E-08 7.41E-07 1.71E-05 1.23E-06 1.88E-06 1.92E-06 3.41E-05 2.80E-05 279 4.34E-08 6.68E-07 2.00E-07 3.30E-06 3.26E-07 6.22E-08 5.99E-08 6.10E-07 1.59E-05 9.OOE-07 1.99E-06 1.28E-06 2.99E-05 2.54E-05 283 4.20E-08 8.45E-07 3.48E-07 3.30E-06 3.26E-07 6.07E-08 6.43E-08 7.03E-07 1.61E-05 1.12E-06 1.90E-06 1.74E-06 3.28E-05 2.66E-05 284 4.23E-08 9.47E-07 3.86E-07 3.30E-06 3.26E-07 5.20E-08 6.76E-08 7.41E-07 1.73E-05 1.24E-06 1.96E-06 1.89E-06 3.45E-05 2.82E-05 285 4.55E-08 9.44E-07 2.02E-07 3.30E-06 3.26E-07 6.06E-08 6.18E-08 6.03E-07 1.66E-05 1.23E-06 1.90E-06 1.69E-06 3.24E-05 2.70E-05 286 4.55E-08 9.53E-07 3.86E-07 1.06E-08 3.25E-07 8.33E-08 9.57E-08 7.42E-07 1.74E-05 1.25E-06 2.01E-06 1.93E-06 3.19E-05 2.52E-05 287 4.35E-08 9.11E-07 3.69E-07 3.16E-06 3.11E-07 5.99E-08 6.46E-08 7.08E-07 1.66E-05 1.18E-06 1.92E-06 1.84E-06 3.35E-05 2.72E-05 288 4.50E-08 7.36E-07 3.86E-07 3.30E-06 3.26E-07 5.83E-08 6.76E-08 7.41 E-07 1.40E-05 1.24E-06 1.70E-06 1.92E-06 3.27E-05 2.46E-05 289 4.52E-08 6.90E-07 3.86E-07 3.30E-06 3.26E-07 5.47E-08 6.76E-08 7.41E-07 1.16E-05 1.24E-06 1.40E-06 1.92E-06 3.03E-05 2.17E-05 El- 78 of 87

NRC RAI 7.f.iv iv. The benefit for SAMA 8 (Increase trainingon response to loss of two 120V AC buses) was determined by eliminating the inadvertent actuation of safety injection.

1. Identify any other impacts of the loss of the two buses that would benefit from the training.

2. Also, the RRW for loss of a single bus is given, but there is no value for the common cause failure of both buses. Discuss this omission.

TVA Response A failure of an electrical bus is considered to be a passive event, i.e., not related to mechanical failures that are normally associated with common cause. TVA modeling practice is not to model passive common cause failures. Therefore, in the transient response model, common cause failures of the 120VAC busses were not modeled. With respect to the initiating events, failure of one 120VAC board will cause a reactor trip. The TVA PRA does not model multiple initiating events within a cutset. Therefore, no RRW was calculated associated with the failure of two busses.

Additional analyses, with various assumptions regarding 120V AC busses, were performed to assess the benefit of increased training upon loss of two busses. In most analyses, the averted cost risk exceeded $50,000. Therefore, this SAMA candidate will be retained for consideration as a potentially cost-beneficial SAMA.

NRC RAI 7.f.v

v. The impact of adding the gas-turbinein SAMA 14 is only a 0.35% and 0.1%

reduction in CDF. Explain why this so small consideringthat SBO is about 10% of the CDF.

TVA Response Installing a gas turbine generator would not eliminate all SBO sequences, but would increase the availability of the DGs, and therefore increase the availability of on-site AC power (namely to the shutdown boards). To estimate the benefit of installing a gas turbine generator, a new event (failure of the gas turbine generator), was added to the DG Supply logic so that failure of DG and the gas turbine generator are required to cause loss of power to the shutdown boards.

There is a high level of redundancy at SQN between each units' shutdown boards, with cross-ties between the two unit's shutdown boards. Therefore, the benefit gained from installing a gas turbine generator is minimal. The contribution to CDF of an SBO (approximately 10% in the SAMA model) is greatly overstated due to the inclusion of internal flooding events in the SBO contribution. Adding a gas turbine generator would not eliminate internal flooding SBO accident sequences. The PRA model is currently being reevaluated to determine the SBO contribution to CDF if internal flooding is not included. This revised value is expected to be approximately 2%. Due to the existing redundancy between each unit's shutdown boards and the internal flooding contribution to SBO, adding a gas turbine generator would not be cost beneficial.

El- 79 of 87

NRC RAI 7.f.vi vi. The title for SAMA 68 on ER page E-122 indicates that an auxiliary feed pump is being added while the text and the title of the SAMA in Table E.2-1 indicates that it is a normal motor driven feed pump. Clarify the intent of this SAMA and revise the discussion of this SAMA to distinguish between main feedwater pumps and AFW pumps.

TVA Response SAMA 68 was conservatively evaluated by eliminating the loss of Main Feedwater (MFW) initiating events and reducing the likelihood of loss of AFW. For MFW, the total and partial loss of MFW initiating events was eliminated. For AFW, a new independent AFW pump with no power dependencies was added to the model to simulate another source of feedwater. While the intent of this SAMA was to increase the availability of MFW, the analysis was conservatively performed to increase the availability of both MFW and AFW. Update of the SQN ER before NRC issues the Final SEIS would include clarification of the SAMA 68 analysis performed.

NRC RAI 7.f.vii vii. For SAMA 70 (Install accumulatorsfor turbine-drivenAFW pump flow control valves), it is indicatedthat a bounding analysis was performed by eliminating the failure of the existing flow control valves. Confirm that this analysis included the failure due to lack of air.

TVA Response The analysis assumed no failure of the human action to restore TDAFWP speed control following initiator and loss of air (HAFR2). The analysis was performed in this manner in order to not introduce excessive benefit from increased reliability of air outside of the TDAFWP control valves. An additional analysis has been performed to completely eliminate all failures of the AFW LCVs (including air and human actions). The results of the sensitivity analyses showed that the Unit 1 benefit increased from $348,010 to $397,595. The Unit 2 benefit increased from $311,460 to $325,625. Note that the original analysis determined that this SAMA is potentially cost beneficial for both units and no revision to the previous analysis is currently required.

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NRC RAI 7.f.viii viii. For SAMA 83 (Add a switchgearroom high temperature alarm), it is stated that a bounding analysis was performed by eliminating the failure of the ventilation fans in the 480V TransformerRoom, thereby maintaininga proper temperature in the room. Confirm that this room is the only one impacted by loss of switchgearheating, ventilation, and air conditioning (HVAC).

TVA Response The 480V Transformer Rooms are the only rooms affected by loss of switchgear HVAC.

Other rooms containing switchgear were also evaluated and it was determined that HVAC was not required for successful switchgear operation during the PRA mission time. There are additional rooms with HVAC requirements, however, these are mechanical equipment rooms and were evaluated in SAMA 160. The rooms evaluated in SAMA 160 include: TDAFWP Rooms; RHR Pump Rooms A and B; SI Pump Rooms A and B; CS Room; CCP Cooler Rooms A and B; and Space Coolers A and B for Boric Acid Transfer Pump and AFW Pumps.

SAMA 160 was determined to be potentially cost-beneficial for Unit 1 and potentially cost-beneficial, when including sensitivities, for Unit 2. Ventilation for the Emergency DGs is considered in SAMA 161.

NRC RAI 7.f.ix ix. For SAMA 103 (Institute simulatortraining for severe accident scenarios), it is stated that a bounding analysis was performed by reducing the failure probabilityof important human actions and that the human errorprobability (HEP)dependency factors for importanthuman actions were also improved.

Identify the HEPs reduced and the amount of the reduction.

TVA Response The following human actions were reduced by 10% to analyze increased operator training.

AFWOP3 Depressurize/cooldown to LP injection after a small or medium LOCA with failure HACD1 Perform cooldown with MFW, following AFW failure HAFRI Restore MDAFW LCV control following initiator and loss of air HAFR2 Restore TDAFWP speed control following initiator and loss of air HAHH1 Place Hydrogen igniters in service HAMARV Hand wheel Operation of the SG ARVs S/G 1&4 HAPRZ Depressurization of the RCS using pressurizer PORVs (Level 2 ONLY)

HAPRZ-SUC SUCCESS-INTENTIONAL OR UNINTENTIONAL RCS DEPRESS PRE I-SGTR (NON-SBO SEQUENCE)

HAOB2 Establish RCS Bleed and Feed cooling given no CCPS running HARRI Align high pressure recirculation, given auto swap over works HART1 Manually trip reactor, given SSPS fails El- 81 of 87

HASE2 Trip RCPs on loss of Component Cooling Water SHECLR-1 Mis-positioning ERCW valves, blocking flow to the CCS/AFW and the BAT/AFW space SHECLR-2 Mis-positioning ERCW valves, blocking flow to the CCS/AFW and the BAT/AFW space SSIOP Terminate SI to prevent PORV water challenge The following HEP dependency factors were revised to reflect the 10% improved human action(s) above.

COMBINATION_1310 HEP dependency factor for HAHH1,HAFR2,HAOB2,HAFR1 COMBINATION_1370 HEP dependency factor for HAHH1,HAFR2,HARR1,HAFR1 COMBINATION_1458 HEP dependency factor for HAPRZ,HAFR2,HAOB2,HAFR1 HEP dependency factor for COMBINATION_1573 HAPRZ,HAFR2,HARR1 ,HAFR1 COMBINATION_2309 HEP dependency factor for HASE2,HAHH1 COMBINATION_1408 HEP dependency factor for HAFR2,HAOB2,HAFR1 COMBINATION_1593 HEP dependency factor for HAFR2,HARR1 ,HAFR1 COMBINATION_1942 HEP dependency factor for HARR1 ,HAFR1 COMBINATION_210 HEP dependency factor for HARR1 ,AFWOP3,HAMARV COMBINATION_213 HEP dependency factor for HASE2, HARR1, AFWOP3, HAMARV COMBINATION_2322 HEP dependency factor for HAOB2,HAMARV COMBINATION_2365 HEP dependency factor for HARR1 ,HAMARV COMBINATION_2674 HEP dependency factor for HARR1 ,SSIOP COMBINATION_919 HEP dependency factor for HARR1 ,HACD1 El- 82 of 87

NRC RAI 7.f.x

x. For SAMA 161 (Provide backup ventilation for the emergency diesel generatorrooms, should their normal HVAC supply fail), the cost is given as

$1M with the source being the minimum hardwarecost. The minimum hardwarecost is $I OOK. Explain this discrepancy.

TVA Response An SQN plant specific estimate was developed for SAMA 161. The source column should be listed as "SQN Estimate." Update of the SQN ER before NRC issues the Final SEIS will include clarification to state the appropriate plant specific estimate.

NRC RAI 7.f.xi xi. For SAMA 188 (Implement modifications to the compressed air system to increase the capacity of the system), describe the modification in more detail to support the cost estimate of $2.8M. Note that the cost for SAMA 87 involving replacingthe service and instrument aircompressors is $900K.

Explain this difference.

TVA Response The difference in the cost estimate comes from SAMA 188 increasing the capacity of the compressed air system. SAMA 87 does not increase the capacity of the compressors, but replaces them with more reliable compressors. The SQN cost estimate for SAMA 188 involved replacing the existing compressor with a larger compressor. The estimate also included installing more reliable compressors which have self-contained air cooling by shaft driven fans in order to eliminate the system dependency on ERCW for cooling water and to avoid potential loss of the compressed air system due to a loss of ERCW.

NRC RAI 7.f.xii xii. If the results of SAMA 268 (Perform an evaluation of the CCS/AFW area cooling requirements)indicate that area cooling is required,will alternative SAMAs, other than SAMA 289 (Install backup cooling system for CCS/AFW space coolers), be added to mitigate these area cooling failures?

TVA Response SAMA 268 originated from the SQN IPE and is used only to perform an evaluation of the CCS/AFW Area Cooling Requirements. This evaluation will establish whether or not space coolers are required for the PRA mission time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Eliminating the dependency of the CCS and AFW pumps on these space coolers would increase the availability of the pumps in the PRA model. Because SAMA 268 is strictly an evaluation, SAMA 289 was analyzed to determine the benefit of installing a backup cooling system for the CCS/AFW Space Coolers.

SAMA 289 was determined to be potentially cost-beneficial in the sensitivity analyses.

SAMA 289 evaluated installing a backup cooling system for the CCS/AFW Space Coolers.

Therefore, SAMA 289 was determined to be potentially cost-beneficial regardless of the results of the CCS/AFW Area Cooling Requirements evaluation (SAMA 268).

El- 83 of 87

If the analysis performed to support SAMA 268 determines that the CCS/AFW Space Coolers are not required, then SAMA 289 will no longer need to be considered nor will it be cost-beneficial.

NRC RAI 7.f.xiii xiii. In ER Tables E.2-1 and E.2-2, the source for the cost of SAMA 284 (Improve reliabilityof pressurizersafety relief valves) is given as "Minimum Hardware Cost" while the value is given as $1.6M versus the stated minimum hardware cost of $1OOK. Correct the source for this cost and describe what makes up this cost.

TVA Response An SQN site-specific estimate was developed for the cost estimate of SAMA 284. The ER reference for the SAMA 284 Minimum Hardware Cost is incorrect. Update of the SQN ER before NRC issues the Final SEIS will include the revised SAMA 284 cost estimate.

The cost estimate for implementation of this SAMA is comprised of Engineering (calculation, DCN and Procurement Specification), Hardware (three valve components), and Installation and Modification Labor (Mechanical, Support Craft, Post Mod Testing, Procedures, Training, and Simulator Modifications).

NRC RAI 7.f.xiv xiv. Explain how the reduction in populationdose and off-site economic cost was determined for SAMA 287 involving mitigating internal fire events.

TVA Response SAMA 287 assumed the conditional core damage probability for four fire compartments was reduced by a factor of 10. The reduction in CDF was applied to the base total CDF (internal and external events) which was then used to determine the percentage reduction in CDF for implementation of SAMA 287. The evaluation assumed that all release category frequencies were reduced by the same percentage as CDF. The reduced CDF and release category frequencies were then used to determine the reduction in population dose and off-site economic cost in a manner similar to all other SAMAs.

El- 84 of 87

NRC RAI 7.a.i

g. Forcertain SAMAs consideredin the SQN EnvironmentalReport, there may be lower-cost or more effective alternatives that could achieve much of the risk reduction. In this regard,provide an evaluation of the following SAMAs. Basis:

Applicants for license renewal are requiredby 10 CFR 51.53(c)(3) (ii)(L) to consider SAMAs if not previously considered in an environmental impact assessment, related supplement, or environmental assessment for the plant. As part of its review of the SQN SAMA analysis, NRC staff considers additionalSAMAs that may be more effective or have lower implementation costs than the other SAMAs evaluated by the applicant. The requestedinformation is needed in orderfor the NRC staff to reach a conclusion on the adequacy of the applicant's determination of cost beneficial SAMAs.

i. Forbasic event HASE2 (and others) involving reactorcoolant pump (RCP) seal cooling failures, consider automaticallytripping of RCPs on loss of CCW TVA Response The RRW of HASE2 for Unit 1 and Unit 2 are 1.264 and 1.21, respectively. Given the high importance of RCP seal cooling in the SQN PRA model, tripping the RCPs upon loss of CCS is one of the most important actions. An RRW of 1.264 for Unit 1 corresponds to an averted cost of approximately $1,600,000 before the uncertainty multiplier is applied. An RRW of 1.21 for Unit 2 corresponds to an averted cost of approximately $1,220,000 before the uncertainty multiplier is applied. The cost of installing a system to automatically trip the RCPs upon loss of seal cooling is estimated to be $1,500,000, not including the lifetime costs of the system.

Therefore, a new SAMA candidate to address important human action HASE2, to automatically trip the RCPs upon loss of seal cooling, will be evaluated for consideration as a potentially cost-beneficial SAMA.

NRC RAI 7.q.ii ii. For SAMA 289 involving installing backup cooling for the component cooling water system (CCS)/AFW space coolers, consider opening doors and/or stage portable fans, etc. unless this is addressedby SAMA 160.

TVA Response SAMA 268 (Perform an Evaluation of the CCS/AFW Area Cooling Requirements) would determine whether the CCS/AFW space coolers are required for the PRA mission time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. If required for the PRA mission time, the analysis would provide an assessment of the time available to perform mitigating actions and the viability of potential mitigating actions.

Thus, the results of this analysis are necessary to determine if lower cost compensatory actions such as opening doors and/or staging portable fans to mitigate the loss of the CCS/AFW space coolers are viable and should be considered for implementation. As stated in response to RAI 7.f.xii., the results of the evaluation performed in support of SAMA 268 will determine future actions. If the analysis determines that the CCS/AFW Space Coolers are not required, then SAMA 289 will no longer be cost-beneficial. If the CCS/AFW Space Coolers are required, then the SAMA 268 analysis results would provide the information needed to determine if opening doors and/or staging portable fans is an acceptable means to provide the required ventilation/cooling. SAMA 289 (Install Backup Cooling System for CCS/AFW Space El- 85 of 87

Coolers) was determined to be potentially cost-beneficial in the sensitivity analysis for both units.

NRC RAI 7.q.ii iii. ForEvents %690.O-A01-1_067_S and %669.0-AO0_067_S representingthe initiatorfor emergency raw cooling water (ERCW) spray events in room 690.0-A l and room 669.0-AO1 in the auxiliary building which results in the loss of AFW pumps, consider installing spray shields for these events.

TVA Response Event %690.0-AO1-1_067_S is analyzed with a spray shield in the evaluation of SAMA 275.

SAMA 275 (Install Spray Protection on Motor-Driven AFW Pumps and Space Coolers) resulted in potentially cost beneficial implementation of this SAMA. The RRW of event

%669.0-A01_067_S is relatively low (1.0054 for Unit 1 and 1.0061 for Unit 2). Eliminating these spray initiators which spray on the turbine driven AFW pump would result in an approximately $100,000 benefit for each unit. The costs associated with engineering, design, hardware, and installation of spray shields to protect the TDAFWP is estimated to be approximately $400,000. While there may be some shared costs between the two units for design and engineering of the shields, the turbine driven pumps are located on opposite sides of the AB on elevation 669.0. Because each pump would require a spray shield, it would not be cost-beneficial to protect the turbine driven pumps from sprays.

NRC RAI 7._.iv iv. For SAMA 71 (Install a new condensate storage tank), discuss the alternative of using a portable pump to provide water for the AFW system.

TVA Response The analysis for SAMA 071 was conservative by assuming that long term makeup to the CST was always available.

If a portable AFW pump is available, the possible failure mechanisms of this pump to provide water to the AFW system (power, discharge lines/valves, and makeup source) were not introduced to the model.

With this conservative assumption, the benefit for Unit 1 was approximately $179,000 and was negligible for Unit 2. With no benefit to Unit 2 from a portable pump, the full cost would be allocated to Unit 1 for hardware, procedures and training, and would likely result in the portable pump not being cost beneficial.

In accordance with Interim Staff Guidance JLD-ISG-2012-01 (FLEX), portable pumps, hoses and generators would likely be available at the plant for use in makeup to the AFW system.

As stated at the end of TVA response to RAI 7.a.i, future SQN design changes will be made to improve external flooding mitigation. TVA has committed to installing "newly designed means for removing reactor decay heat [during flooding scenarios] and providing reactor coolant system makeup for both SQN Units.

El- 86 of 87

NRC RAI 7.q.v

v. For SAMA 161 (Providebackup ventilation for the emergency diesel generatorrooms, should their normal HVAC supply fail), consider using temporary ventilation, opening doors, etc.

TVA Response Eliminating the failure of the dampers and exhaust fans which provide ventilation to the EDGs and Electric Board Room resulted in a Unit 1 and Unit 2 combined averted cost risk of

$119,355. While the costs of opening doors would be lower than temporary ventilation, analyses would be required in both scenarios to ensure that sufficient cooling could be achieved from opening doors and/or providing temporary ventilation. Analyses are also necessary to determine whether there is sufficient time available to perform these actions because the EDG rooms would likely heat up rapidly upon loss of ventilation. A similar area heat-up analysis (e.g., SAMA 268) has been estimated to cost about $300,000. In addition to hardware costs, procedure changes and training that would be required, the additional analysis to determine the viability of temporary ventilation would increase the costs such that implementation of this SAMA would remain not cost beneficial.

NRC RAI 7.q.v, vi. Provide an evaluation of a SAMA to purchase or manufacture a "gagging device" that could be used to close a stuck-open steam generatorsafety valve for a SGTR event priorto core damage.

TVA Response A potential cost beneficial SAMA evaluation will be performed to purchase or manufacture a "gagging device" and develop a procedure or work order that could be used to close a stuck-open SG safety valve for a SGTR event prior to core damage.

El- 87 of 87

NPG-SPP-03.10 Licensing Transmittal to NRC Summary and Concurrence Sheet The purpose of this concurrence sheet is to assure the accuracy & completeness of TVA submittals to the NRC. All blocks on this form must be comoleted.

DATE DUE: 06/10/13 (C) I(A) PREPARER: Henry Lee, Mike Walker, Gary Smith

SUBJECT:

SQN LR SAMA RAI 0002 RESPONSE TO MLI3119A083.SAMA. RECEIVED ON 5110.30DAY REGULATORY CONSIDERATIONS (B) Oath or Affirmation 0 YES (I NO (E) Posting Requirements []YES 0 NO (C) Licensing Verification 0 YES El NO (F) FSAR IMPACT El YES 0 NO (D) New Commitments E] YES Z NO (G) OGC Consultation 0 YES El NO (H) LIC-109 Review El YES 0 NA (I) Engineering Independent 0 YES

• E] NO Technical Review (ITR)

• Engineering (only): BNorman, CGuey, KVincent, or Design Change N/A LChristianson, JMcdonald, DLundy INDEPENDENT TECHNICAL REVIEWER (J) Licensing (only): Jonathan Johnson / ,-

PEER CHECKER ,

CONCURRENCE (K) K 7//1 A concurrence signature reflects that the signatory has assured that the submittal is appropriate and consistent with TVA Policy, applicable commitments are approved for implementation and supporting documentation for submittal completeness and accuracy has been prepared..

NAME ORGANIZATION Schrull, Edward D Corp Licensing Vance, Scott A Tva OGC Walker, Michael J SQN PRA Mgr Guey, Ching Corp PRA Sr Mgr Noe, P Todd SQN Eng Dir McBrearty, Michael SQN Licensing Mgr __________________

Marshall, Thomas B SON Dir, Safety/Licens ___________________

Carlin, John T SQN Vice President (L) Licensing (only)

NPG-SPP-03.10 Licensing Transmittal to NRC Summary and Concurrence Sheet The purpose of this concurrence sheet is to assure the accuracy & completeness of TVA submittals to the NRC. All blocks on this form must be completed.

DATE DUE: 06/10/13 (C) I(A) PREPARER: Henry Lee, Mike Walker, Gary Smith

SUBJECT:

SON LR SAMA RAI 0002 RESPONSE TO MLI3119A083.SAMA. RECEIVED ON 5/10, 30DAY REGULATORY CONSIDERATIONS (B) Oath or Affirmation 0 YES [] NO (E) Posting Requirements F] YES 0 NO (C) Licensing Verification 0 YES 0] NO (F) FSAR IMPACT El YES 0 NO (D) New Commitments FJ YES 0 NO (G) OGC Consultation 0 YES El NO (H) LIC-109 Review El YES 0 NA (I) Engineering Independent 0 YES* O NO Technical Review (ITR)

• Engineering (only): BNorman, CGuey, KVincent, or Design Change N/A LChristianson, JMcdonald, DLundy INDEPENDENT TECHNICAL REVIEWER (J) Licensing (only): Jonathan Johnson /

PEER CHECKER CONCURRENCE (K)

A concurrence signature reflects that the signatory has assured that the submittal is appropriate and consistent with TVA Policy, applicable commitments are approved for implementation and supporting documentation for submittal completeness and accuracy has been prepared.

NAME ORGANIZATION SIGNATURE DATE Schrull, Edward D Corp Licensing Reviewed Vance, Scott A Tva OGC Reviewed Walker, Michael J SON PRA Mgr Guey, Ching Corp PRA Sr Mgr Signed Noe, P Todd SON Eng Dir Signed McBrearty, Michael SON Licensing Mgr Signed Marshall, Thomas B SON Dir, Safety/Licens Signed Carlin, John T SON Vice President (L) Licensing (only) Henry Lee / (Quality Reviewer)

NPG-SPP-03.10 Licensing Transmittal to NRC Summary and Concurrence Sheet The purpose of this concurrence sheet is to assure the accuracy & completeness of TVA submittals to the NRC. All blocks on this form must be comoleted.

DATE DUE: 06/10/13 (C) I(A) PREPARER: Henry Lee, Mike Walker, Gary Smith

SUBJECT:

SQN LR SAMA RAI 0002 RESPONSE TO MLI3119A083.SMA, RECEIVED ON 5/10, 30DAY REGULATORY CONSIDERATIONS (B) Oath or Affirmation 0 YES [: NO (E) Posting Requirements El YES 0 NO (C) Licensing Verification 0 YES 0l NO (F) FSAR IMPACT El YES 0 NO (D) New Commitments El YES 0 NO (G) OGC Consultation 0 YES E] NO (H) LIC-109 Review El YES 0 NA (I) Engineering Independent 0 YES* [] NO Technical Review (ITR)

Engineering (only): BNorman, CGuey, KVincent, or Design Change N/A LChristianson, JMcdonald, DLundy INDEPENDENT TECHNICAL REVIEWER (J) Licensing (only): Jonathan Johnson /

PEER CHECKER CONCURRENCE (K)

A concurrence signature reflects that the signatory has assured that the submittal is appropriate and consistent with TVA Policy, applicable commitments are approved for implementation and supporting documentation for submittal comoleteness and accuracy has been oreoared.

NAME ORGANIZATION SIGNATURE DATE Schrull, Edward D Corp Licensing Vance, Scott A Tva OGC Walker, Michael J SQN PRA Mgr Guey, Ching Corp PRA Sr Mgr ,4* -'-" "*ff*-Lo.

Noe, P Todd SQN Eng Dir McBrearty, Michael SQN Licensing Mgr Marshall, Thomas B SQN Dir, Safety/Licens Carlin, John T SQN Vice President (L) Licensing (only) Henry Lee (Quality Reviewer)

NPG-SPP-03.10 Licensing Transmittal to NRC Summary and Concurrence Sheet The purpose of this concurrence sheet Is to assure the accuracy &completeness of TVA submittals to the NRC. All blocks on this form must be completed.

DATE DUE: 06/10/13 (C) (A) PREPARER: Henry Lee, Mike Walker, Gary Smith

SUBJECT:

SQN LR SAMA RAI 0002 RESPONSE TO MLI3119A083.SAMA, RECEIVED ON 5/10, 30DAY REGULATORY CONSIDERATIONS (B) Oath or Affirmation 0 YES [] NO (E) Posting Requirements El YES 0 NO (C) Licensing Verification 0 YES El NO (F) FSAR IMPACT El YES 0-NO (D) New Commitments 0l YES 0 NO (G) OGC Consultation 0 YES El NO (H) LIC-109 Review El YES 0 NA (I) Engineering Independent E YES

• E] NO Technical Review (ITR)

Engineering (only): BNorman, CGuey, KVincent, or Design Change N/A LChristianson, JMcdonald, DLundy_

INDEPENDENT TECHNICAL REVIEWER (J) Licensing (only): Jonathan Johnson I PEER CHECKER CONCURRENCE (K)

A concurrence signature reflects that the signatory has assured that the submittal is appropriate and consistent with TVA Policy, applicable commitments are approved for implementation and supporting documentation for muihmittal comoleteness and accuracy has been Drenared.

NAME ORGANIZATION SIGNATURE DATE Schrull, Edward D Corp Licensing Vance, Scott A Tva OGC Walker, Michael J SQN PRA Mgr Guey, Ching Corp PRA Sr Mgr Signed Noe, P Todd SQN Eng Dir McBrearty, Michael SQN Licensing Mgr Marshall, Thomas B SQN Dir, Safety/Licens Carlin, John T SQN Vice President (L) Licensing (only) / (Quality Revie4ko" '

NPG-SPP-03.10 Licensing Transmittal to NRC Summary and Concurrence Sheet The purpose of this concurrence sheet is to assure the accuracy & completeness of TVA submittals to the NRC. All blocks on this form must be completed.

DATE DUE: 06/10/13 (C) I(A) PREPARER: Henry Lee, Mike Walker, Gary Smith

SUBJECT:

SQN LR SAMA RAI 0002 RESPONSE TO ML13119A083.SAMA, RECEIVED ON 5/10, 30DAY REGULATORY CONSIDERATIONS (B) Oath or Affirmation 0 YES El NO (E) Posting Requirements El YES 0 NO (C) Licensing Verification 0 YES El NO (F) FSAR IMPACT El YES 0 NO (D) New Commitments El YES 0D NO (G) OGC Consultation 0 YES El NO (H) LIC-109 Review El YES 0 NA (I) Engineering Independent 0 YES

• El NO Technical Review (ITR)

Engineering (only): BNorman, CGuey, KVincent, or Design Change N/A LChristlanson, JMcdonald, DLundy INDEPENDENT TECHNICAL REVIEWER (J) Licensing (only): Jonathan Johnson S PEER CHECKER 0 CONCURRENCE (K)

A concurrence signature reflects that the signatory has assured that the submittal is appropriate and consistent with TVA Policy, applicable commitments are approved for implementation and supporting documentation for submittal completeness and accuracy has been prepared.

NAME ORGANIZATION SIGNATURE DATE Schrull, Edward D Corp Licensing Vance, Scott A Tva OGC Walker, Michael J SQN PRA Mgr Guey, Ching Corp PRA Sr Mgr Noe, P Todd SQN Eng Dir 113//.

McBrearty, Michael SQN Licensing Mgr 'kJk',____,____._;>

S Q N D ir , S a fet y/ L ic e n s __

_ _ _ _-__ _ _ _ _ _ _ _ ,, - _

Ma rs h a ll, Th o m a s B _ _ _ _ _ _ _ _ _ _ _

Carlin, John T SQN Vice President (L) ,Liesn (OY Heny Le..(Qaliy. Rvieer (L) Licensing (only) Henry Lee / (Quality Reviewer)

SCF 2 WA~p Major changes/comments in Q #

Minor editorial edits in Q#

No comment in Q#

From: Wadkins, George Sent: Wednesday, July 10, 2013 4:47 PM To: Lee, Henry

Subject:

RE: Set 2.Sama Peer Check status?

Based on what I have been given, ye*s.

From: Lee, Henry Sent: Wednesday, July 10, 2013 4:17 PM To: Wadkins, George

Subject:

RE: Set 2.Sama Peer Check status?

Does this mean you are done with the Peer Ck for SAMA?

From: Carey, Christopher Sent: Friday, July 12, 2013 2:35 PM To: Lee, Henry

Subject:

RE: NRC comments for the 7/1/13 SQN SAMA RAI response revision

Henry, Yes, the statement in red below is true and accurate.

For the Verification package:

The fuel fission product inventory is primarily determined by the reactor power, operating cycle length and initial fuel enrichment (see attached calculation).

TVA does not have plans to change the operating cycle length or power level and consequently initial fuel enrichment. -

I can get Jim Lemons (General Manager, Reactor Engineering and Fuels) to sign this statement when the final verification package is ready.

Chris Carey Manager, Nuclear Safety Analysis Tennessee Valley Authority Phone: 423-751-2535 Cell: 423-290-4715 email: ccarey@tva.gov Henry From: Lee, Henry Sent: Wednesday, July 03, 2013 12:46 PM To: Carlin, John Thomas Cc: Pollock, Douglas P; Lundy, Dennis L

Subject:

FW: NRC comments for the 7/1/13 SQN SAMA RAI response revision If we state (in red) then NRC would be satisfied.

Can we truthfully & accurately say the statement below under Oath & Affirmation?

NRC RAI from a Telecom with TVA On June 18, 2013, in a SAMA TVA telecom with the NRC staff, the NRC made the statement "Confirm that the SQN fuel inventory is applicable to the fuel cycle expected for SQN during the period of extend operation for the license renewal."

TVA Response TVA confirms that the SQN fuel inventory is applicable to the fuel cycle expected for SQN during the period of extend operation for the license renewal From: Carey, Christopher Sent: Tuesday, July 02, 2013 9:20 AM To: Lee, Henry

Subject:

RE: Set 2. SAMA RAI response

Enclosure:

Due 7/9 Looks good.

Chris Carey Manager, Nuclear Safety Analysis Phone: 423-751-2535 Cell: 423-290-4715 email: ccarey@tva.gov From: Pollock, Douglas P

Sent: Tuesday, July 02, 2013 7:24 AM To: Lee, Henry Cc: Carey, Christopher

Subject:

RE: Set 2. SAMA RAI response

Enclosure:

Due 7/9 I'm ok with the response on the last page of the enclosure

Thanks, Doug Pollock TVA Nuclear Safety Analysis Program Manager LP4G-C, 423-751-4312 Pager - 14086 Cell - 423-505-9561 From: Lee, Henry Sent: Monday, July 01, 2013 10:07 AM To: Lundy, Dennis L; Williams, Evelyn Marie; Wilson, Charles L; Adkins, Gary M; Guey, Ching; Walker, Michael J; McDonald, Jeffrey A; Vincent, Keith Aaron; Christiansen, Lance; Jiang, Hongbing; 'gsmith@enercon.com';

'mamorris@enercon.com'; 'dluchsinger@enercon.com'; 'bnorman@enercon.com'; 'dmosby@enercon.com';

'ccramer@enercon.com'; Pollock, Douglas P; Carey, Christopher; Brown, David Michael Cc: Nelson, Blake Jon; Vance, Scott A; Vigluicci, Edward J; Shea, Joseph W; Schrull, Edward Dustin; Wadkins, George; Goodin, Donald V II; Lundy, Dennis L; Adkins, Gary M; Marshall, Thomas B; McBrearty, Michael; Noe, P Todd

Subject:

Set 2. SAMA RAI response

Enclosure:

Due 7/9

All, Here is the latest SAMA RAI Response ("Enclosure") update.

Comments from SQN Staff, Ed S., Blake/OGC, NRC SAMA staff (from 3 NRC telecoms) are incorporated in the attached version.

The entire response has been revised & tweaked to incorporate feedback from the NRC SAMA staff. Please review your response closely to ensure that the intent/meaning of your original response has NOT been altered due to my mistake.

Mike W. & Gary S: Please use this version to build your SAMA validation package.

Doug & Chris C: Please the review the final response, on the last page of the "Enclosure" OGC/Gary S/Mike W./Ching: I need all three of you to review the entire SAMA response again.

7.a.i: I add this note. Is this ok?

Note: "Minimal effect" means that it is not worth the time and effort to refine the PRA model and calculate the exact difference for the internal events model 7.b.i Review the redraft 1V parag. Of the response 7.b.ii See OGC comments below TVA Response As discussed in Sections E. 1.2.1 and E. 1.2.3.2, it was determined that the SQN Level 2 Model of Record overestimated the release category frequency results. Because the original focus for the development of the Level 2 model was LERF, the remaining release end states had not been quantified and reviewed in detail.

Therefore, modifications were made to the existing Level 2 model in order to improve the quantification of the other end states. This included incorporation of sequence success logic for each Level 2 sequence directly into the Level 2 fault tree by adding the negation of the success branches of each CET sequence. In addition, it was determined that the original logic did not correctly model the failure of containment isolation for SBO scenarios because SBO conditions did not result in large isolation releases as intended. A minor fault tree revision corrected the logic. The models used for the SAMA analysis includes these modifications and corrections.

Excluding the intact category, these modifications resulted in a significant reduction in the overestimation of the frequencies and the sum of all of the release category frequencies, which was slightly less than the total CDF.

However, the response to RAI 7.b.iii indicates that when the intact frequency is included, the total release frequency is approximately 50% higher than the CDF. The total release frequency estimate is greater than the total CDF due to the treatment of the success branches within the event trees. When there are cutsets with high-probability events (e.g. 0.9 or above), as there are in many of the success branches of the CETs, the results are overestimated due to the breakdown of the rare event approximation. Essentially, the cutsets resulting from CETs branches with branch probabilities close to one produce cutsets that are non-minimal, resulting in some double-counting.

NRC RAI 7.1x J. For SAMA 161 (Providebackup ventilation for the emergency diesel generatorrooms, should their normal HVAC supply fail), the cost is given as $1M with the source being the minimum hardwarecost. The minimum hardware cost is $100K. Explain this discrepancy.

TVA Response

A SQN plant specific estimate was developed for this SAMA. The source column should be listed as "SQN Estimate". The SQN ER annual update would include clarification to state the appropriate plant specific estimate.

RAI 7.e.vii Review the response Ms. Joe W.Shea VIV PtaMefi. Nuclear Licensing Tertrnaee Valley Aulhority P.O. B0= 2000 6oSdy-Vi*y, TNl37384

SUBJECT:

PROJECT MANAGER CKANGE FOR THE LICENSE RENEWAL OF SEQUOYAk NUCLEAR PLANT. UNITS I AND 2 (TAC NOS. MFO057 AND MFOO" Oew M. Shew&

This lehter provides notlfbcation df a change to the oftely end envronmental project fmag

• Waigped to the Sequoyh Nucle Pla (SQN), UVIhiI and 2, IlceMe wtg*"I FloW*,

Elfective May 5, 2013. Mr. Richard Ptasse wi be "hlesly project mnWnage and Mr CavtM Drdwvrwll be the environmeral projec managet for SON.

All coemspand*n*o for the Pirvid Ulnagar aodated wYth the sofely and emlrenrnental ruww for SON Deem renewal should be wi to Mr. Richad Please, safety ptojea manager,

=01 Mr. David Druckw, ewrmmlbntai project managor, respbcively, at Mte tflwing aeddfes.

NRC RAI 7.a.i at the end of the TVA response On June 18, 2013, in a telecom with the SAMA staff, NRC made the statement "Confirm that the SQN fuel inventory is applicable to the fuel cycle expected for SQN during the period of extend operation for the license renewal."

TVA Response TVA reviewed NEI 05-01 pg. 14 and it discussed methods to determine the source term for the SAMA analysis.

The SQN source term is based on current licensed power and an 18 month operating cycle.

During the period of license renewal the current power level and the length of the operating cycle are expected to remain the same. Therefore, TVA confirms that this inventory is applicable to the fuel cycle expected for SQN during the period of extend operation for the license renewal.

From: Brown, David Michael Sent: Wednesday, June 19, 2013 10:45 AM To: Carey, Christopher; Pollock, Douglas P; Lee, Henry

Subject:

RE: SQN SAMA Fuel inventory

One addition below in red.

Thanks Doug and Chris From: Carey, Christopher Sent: Wednesday, June 19, 2013 10:42 AM To: Pollock, Douglas P; Lee, Henry Cc: Brown, David Michael

Subject:

RE: SQN SAMA Fuel inventory I looked at NEI 05-01 pg. 14 and it discussed methods to determine the source term for the SAMA analysis. The NRC question comes straight from the last sentence in the NEI discussion so I suggest the following answer:

The SQN source term is based on current licensed power and an 18 month operating cycle. During the period of license renewal the current power level and the length of the operating cycle are expected to remain the same. Therefore, TVA confirms that this inventory is applicable to the fuel cycle expected for SQN during the period of extend operation of the license renewal.

Chris Carey Manager, Nuclear Safety Analysis Tennessee Valley Authority Phone: 423-751-2535 Cell: 423-290-4715 email: ccarey@tva.gov This electronic message transmission contains information which may be TVA SENSITIVE, TVA RESTRICTED orTVA CONFIDENTIAL. Any misuse or unauthorized disclosure can result in both civil and criminal penalties. If you are not the intended recipient, be aware that any disclosure, copying, distribution or use of the content of this information is prohibited. If you have received this communication in error, please notify me immediately by email and delete the original message.

From: Pollock, Douglas P Sent: Wednesday, June 19, 2013 10:38 AM To: Lee, Henry Cc: Carey, Christopher; Brown, David Michael

Subject:

RE: SQN SAMA Fuel inventory Henry, David asked me to look into this question The writeup in the Environmental report isn't quite correct. The core inventory isn't a function of MWe, it's a function of MWt and that is probably why we are getting this question.

WBN is licensed for 3459 MWt and SQN is licensed for 3455 MWt. So in reality, WBN would have a slightly high core inventory. In Design basis space for radiological consequences, we have assumed the same parameters and use the same core inventory between WBN and SQN.

So to answer the question, I would say they are higher than what we have analyzed and so they would be conservative to use. I also noted that they listed a number of isotopes that we don't analyze, so if we've represented that we have determined the impact from those isotopes, then we may have a problem and would want to revise that table.

Let me know if you have any questions or need any additional info.

Thanks, Doug Pollock TVA Nuclear Safety Analysis Program Manager LP4G-C, 423-751-4312 Pager - 14086 Cell - 423-505-9561 This electronic message transmission contains information which may be TVA SENSITIVE, TVA RESTRICTED or TVA CONFIDENTIAL. Any misuse or unauthorized disclosure can result in both civil and criminal penalties. If you are not the intended recipient, be aware that any disclosure, copying, distribution or use of the content of this information is prohibited. If you have received this communication in error, please notify me immediately by email and delete the original message.

From: Brown, David Michael Sent: Wednesday, June 19, 2013 9:27 AM To: Brown, David Michael; Pollock, Douglas P Cc: Carey, Christopher

Subject:

RE: SQN SAMA Fuel inventory

Doug, An additional piece of info is that the NRC mentioned NEI 0501 page 14 and the MAXFUEL code while discussing this question.

David From: Brown, David Michael Sent: Wednesday, June 19, 2013 8:08 AM To: Pollock, Douglas P Cc: Carey, Christopher

Subject:

FW: SQN SAMA Fuel inventory Importance: High

Doug, If it is ok with Chris, can you help me with this today? Apparently the NRC asked this question during the initial review of the SQN License Renewal and Corporate Licensing referred the question to me.

David From: Lee, Henry Sent: Tuesday, June 18, 2013 4:00 PM To: Brown, David Michael

Subject:

SQN SAMA Fuel inventory Importance: High See the attached document for the quote that came from Bob Schmit below.

Can you provide an answer to the NRC question?

Henry From: Dozier, Jerry [mailtoiJerry.Dozier@nrc.gov]

Sent: Tuesday, June 18, 2013 3:52 PM To: Drucker, David

Subject:

FW: SQN SAMA Fuel inventory Importance: High From: erschmidt36@comcast.net [1]

Sent: Tuesday, June 18, 2013 3:50 PM To: Dozier, Jerry; Roland Benke

Subject:

SQN SAMA Fuel inventory Importance: High Section E.1.5.2.8 of the License Renewal SQN ER states: "The core inventory was taken from Watts Bar Unit 1, and multiplied by the ratio of power output from SQN Unit 1 to the power output of Watts Bar Unit 1 (1123 MWe)."

NRC Question: Confirm that this inventory is applicable to the fuel cycle expected for SQN during the period of extend operation of the license renewal. If not provide justification for the use of this inventory for the SAMA analysis.

E. R. (Bob) Schmidt 301-473-7010

A (&zThVVF)

NPG Standard Managing TVA's Interface With NRC NPG-SPP-03.10 Programs and Rev. 0000 Processes Page 40 of 57 Appendix F (Page I of 5)

Licensing Submittal Format and Content 1.0 ATTRIBUTES These attributes should be considered during development of a submittal to ensure that the submittal is appropriate for the issue.

A. CORPORATE/SITE LICENSING MANAGER 03 1. Utilize NPG-SPP-18.2.2, Appendix C, "Human Performance Tools for Managers and Supervisors," and NLDP-7, "Licensing Personnel Training and Qualification," when assigning critical licensing functions to Preparer(s). [PER 490611-042 and -016]

o3 2. Complete NPG-SPP-03.10-2 correspondence review form and return to appropriate Management Assistant for filing in EDMS. Actions noted on NPG-SPP-03.10-2 form should be consistent with intended action. [PER 490611-042]

B. BASIS FOR SUBMITTAL 0 1. NPG-SPP-03.10-2 correspondence review completed, filed In EDMS, and consistent with intended action.

E3 2. Detailed review of NRC correspondence complete and all requirements for response identified.

FORMAT SThe standard font for licensing submittals is uArialn" The standard font size is 11.

./ Different fonts or sizes may be used when appropriate, but fonts and sizes used shall ensure legibility of the document.

The standard format for licensing submittal letters should follow the Correspondence Guidelines located on the Licensing homepage under "Correspondence Guidelines."

D. COVER LETTER should include the following attributes and features (as applicable):

.$ Free of grammatical, spelling, and typographical errors.

• 2. Regulations providing the basis for the submittal annotated in the upper right corner of the submittal cover letter.

NPG Standard Managing TVA's Interface With NRC NPG-SPP-03.10 Programs and Rev. 0000 Processes Page 41 of 57 Appendix F (Page 2 of 5)

Licensing Submittal Format and Content 1.0 ATTRIBUTES (continued)

(* Descriptive title.

Include any applicable NRC TAC numbers in the descriptive title.

SClear description of the regulatory processes, including reference to applicable regulations (examples: 10 CFR 50.59, §50.55a, §50.51,

§50.30, §50.90, §50.71, §50.73, §50.54(f), etc.).

~ Docket numbers for all units affected by the submittal. [PER 622194-001]

Clear summation including:

O."Purpose of the submittal.

01 Brief explanation of the content

c. When completion of NRC review action is needed.

0 8. Include a clear description of any deviations from industry precedence-4.- State what is different.

References to prior correspondence, meetings, telephone calls, etc.

o0q,ff} [0 10. Clear statement of proprietary information and rationale for withholding from the public. Specific guidance for required marking of the cover letter and other portions of the document to be withheld from public disclosure is contained in 10 CFR 2.390 and NRC RIS 2004-11.

Note any commitments made and when actions will be completed or add a "no commitment" statement to cover letter if none. (Not required for LERs or routine reports.)

0 12. Discussion of the risk-informed nature of the submittal, if applicable.

Af4*¢ 0 13. Citation of applicable regulations ifthe amendment is being filed as exigent or emergency.

NPG Standard Managing TVA's Interface With NRC NPG-SPP-03.10 Programs and Rev. 0000 Processes Page 42 of 57 Appendix F (Page 3 of 5)

Licensing Submittal Format and Content 1.0 ATTRIBUTES (continued)

4. Oath or affirmation are required for the signature on the cover letter if the

. submittal involves license applications or amendments per 10 CFR 50.30(b), FSAR amendments per 10 CFR 50.71 (e)(2), any response that NRC requires a response under 10 CFR 50.54(f) or other response to NRC requests for which the regulation specifies signature under oath or affirmation.

5.Request for Additional Information (RAI) responses must be submitted under oath or affirmation if the original submittal to NRC required oath or affirmation.

O ENCLOSURES should include the following attributes and features:

0 . Free of grammatical, spelling, and typographical errors.

W 2. Overall level of detail is consistent with degree of complexity of the issue.

. Technical analysis is thorough and includes a clearly written justification.

(* . Regulatory analysis is technically accurate and thorough.

, Order headings and subheadings logically and state the content.

Organize thought process to tell "the whole story." (stand-alone document)

S Section into distinct pieces, such as historical, technical, etc.

• . Define and explain technical terms.

O. 9. Anticipate NRC questions and address them in the submittal. (Based on a review of RAIs for precedent examples).

O3 10. Write PROPOSED CHANGE AND REGULATORY SAFETY ANALYSIS sections which are suitable for use in the NRC staff's safety evaluation report.

NPG Standard Managing TVA's Interface With NRC NPG-SPP-03.10 Programs and Rev. 0000 Processes Page 43 of 57 Appendix F (Page 4 of 5)

Licensing Submittal Format and Content 1.0 ATTRIBUTES (continued)

Consider submitting drawings or figures for clarification (or reference 4/4 0[] 11. FSAR figures).

413-- 12. Write supplements to original submittals so they stand alone.

Marking of pages (at the top of the page) or paragraphs (adjacent with the 013. paragraph) regarding the need to withhold from public disclosure and basis for withholding. Detailed instructions on the requirements for marking for this purpose are contained in 10 CFR 2.390(b)(1)(i).

/J* I'-I14.

Ensure there is a separate enclosure with a List of Commitments, if the submittal contains commitments. [PER 594050-001]

O 15. Records have been properly designated and submitted to EDMS.

SPECIFIC ENCLOSURE CONSIDERATIONS - In addition to the above, there are specific submittal processes prescribed in procedures that shall be applied as follows:

O 1. If outgoing correspondence is a LICENSE AMENDMENT REQUEST (LAR), it meets the requirements established in NLDP-1, "Technical Specifications/Licenses and Amendments." The industry template for LARs is provided in NLDP-1, Appendix A. Other formats may be used to facilitate TS amendment requests for specific applications as developed by industry groups or vendors, or when determined by Licensing management. These include the Consolidated Line Items Improvement Process (CLIIP). A CLIIP should not deviate from the approved CLIIP unless the deviation is clearly identified as a deviation and the basis for the deviation is provided. Deviations should be coordinated with the appropriate NRC Project Manager.

O3 2. If outgoing correspondence is a RELIEF REQUEST or ASME CODE REQUEST FOR RELIEF, it meets the requirements of NPG-SPP-09.1, "ASME Code and Augmented Programs," Appendix A, which provides the preferred industry templates.

O3 3. If outgoing correspondence is a FSAR AMENDMENT, it meets the requirements of NLDP-5, "FSAR Management" and NLDP-9, u"ndependent Spent Fuel Storage Installation (ISFSI) FSAR Management Process."

NPG Standard Managing TVA's Interface With NRC NPG-SPP-03.10 Programs and Rev. 0000 Processes Page 44 of 57 Appendix F (Page 5 of 5)

Licensing Submittal Format and Content 1.0 ATTRIBUTES (continued) 0 4. If outgoing correspondence is a LICENSE EVENT REPORT (LER), it meets the requirements of NPG-SPP-03.5, "Regulatory Reporting Requirements," and NUREG-1022, "Event Reporting Guidelines."

G. LICENSING TRANSMITTAL TO NRC

SUMMARY

AND CONCURRENCE SHEET (NPG-SPP-03.10-1)

• . Items A through I of Form NPG-SPP-03.10-1 have been appropriately dispositioned per the guidance and references on Attachment 1, pages 2 and 3. These responsibilities are delegated to the LE/LPM and reviewed by the Peer Checker.

e 2. Form NPG-SPP-03.10-1, "Licensing Transmittal to NRC Summary and Concurrence Sheet," has been included as part of the regulatory response package described in Section 3.2.3C.13.

( . Form NPG-SPP-03.10-1, "Licensing Transmittal to NRC Summary and Concurrence Sheet," demonstrates communication of submittal to affected parties.

. i Letter/Submittal Package includes sufficient verification developed in accordance with NPG-SPP-03.10, Appendix F, when e

verification is specified in Form NPG-SPP-03.10-1, "Licensing Transmittal to NRC Summary and Concurrence Sheet," Item C.

ol 5. Form NPG-SPP-03.10-1, uLicensing Transmittal to NRC Summary and Concurrence Sheet," Item L, Final Quality Review is complete, indicating submittal is ready for signature and release to NRC NOTE Checklist items abov should be completed before the submittal is released to NRC. Item six below is a required action th t can only be completed after the submittal is signed.

6. Form NPG-SPP-03.10-1, "Licensing Transmittal to NRC Summary and Concurrence Sheet," has been submitted to EDMS as part of the c respondence record.