ML070470644
| ML070470644 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 02/12/2007 |
| From: | Korsnick M Constellation Energy Group |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| Download: ML070470644 (22) | |
Text
11 Maria Korsnlck R.E. Ginna Nuclear Power Plant, LLC Site Vice President 1503 Lake Road Ontario, New York 14519-9364 585.771.5200 585.771.3943 Fax maria.korsnick@ costellation.com Generation Group Energyf Constellation February 12, 2007 U. S. Nuclear Regulatory Commission Washington, DC 20555 ATTENTION: Document Control Desk
SUBJECT:
R.E. Ginna Nuclear Power Plant Docket No. 50-244 Response to Request for Additional Information and Revised Mark-up Regarding Service Water Pump Operability Requirements for the R.E. Ginna Nuclear Power Plant Reference 1: Letter to USNRC Document Control Desk from Mary Korsnick (Ginna), License Amendment Request: Change to Technical Specification 3.7.8, Service Water (SW) from an Electrical Train Based to a Pump Based Specification, dated September 29, 2006 Reference 2: Letter from Patrick Milano (NRC) to Mary Korsnick (Ginna), Request for Additional Information Regarding Service Water Pump Operability Requirements, R.E. Ginna Nuclear Power Plant, dated December 13, 2006 (TAC No. MD3118)
Reference 3: Letter to USNRC Document Control Desk from Mary Korsnick (Ginna),
Addendum to September 29, 2006 Submittal RE: Change to Technical Specification 3.7.8, Service Water (SW) from an Electrical Train Based to a Pump Based Specification, dated December 7, 2006.
On September 29, 2006 R.E. Ginna Nuclear Power Plant, LLC (Ginna LLC) submitted a request to change Technical Specification (TS) 3.7.8, Service Water (SW) from an Electrical Train Based to a Pump Based Specification (Reference 1). On December 13, 2006 NRC issued a Request for Additional Information (RAI) regarding that submittal (Reference 2). Reference 3 contains information relevant to the RAI. Attachment 1 to this letter contains the Ginna LLC response to the RAI, Attachment 2 contains a new commitment and Attachment 3 contains a revised mark-up of the proposed change to TS 3.7.8. The revised mark-up provides consistency with other T
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Document Control Desk February 12, 2007 Page 2 sections by clarifying the wording of Condition D to address more than three (3) SW pumps inoperable.
Should you have questions regarding the information in this submittal, please contact Mr. Robert Randall at (585) 771-5219 or Robert.Randall@constellation.com.
Very truly yours Mary G. Korsnick STATE OF NEW YORK
- TO WIT:
COUNTY OF WAYNE I, Mary G. Korsnick, begin duly sworn, state that I am Vice President, R.E. Ginna Nuclear Power Plant, LLC (Ginna LLC), and that I am duly authorized to execute and file this request on behalf of Ginna LLC. To the best of my knowledge and belief, the statements contained in this document are true and correct. To the extent that these statements are not based on my personal knowledge, they are based upon information provided by other Ginna LLC employees and/or consultants. Such information has been reviewed in accordance with company practice and I believe it to be reliable.
11/At k"6 9.
Subsc*ibed and sworn before me a Notary Public in and for the State of New York and County of//I L/., this /A. day of Fed.2rZhL, 2007.
WITNESS my Hand and Notarial Seal:
Notary Public U1ER M*A/ONl PStadtecNewYork My Commission Expires: No.01M!6017Th5 1A -.D2 -/
Date
Document Control Desk February 12, 2007 Page 3 MK/MR Attachments: (1) Response to Request for Additional Information (2) List of Regulatory Commitments (3) Proposed Technical Specification Changes (Revised Mark-up) cc: S. J. Collins, NRC D.V. Pickett, NRC Resident Inspector, NRC (Ginna)
P.D. Eddy, NYSDPS J. P. Spath, NYSERDA
Attachment (1)
Lý - - -
Response to Request for Additional Information I - I Attachment (1)
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Attachment (1)
Response to Request for Additional Information To complete its review, the Nuclear Regulatory Commission (NRC) staff requested the following additional information:
Question 1 Regulatory Guide (RG) 1.177, "An Approach for Plant-Specific, Risk-Informed Decision making: Technical Specifications," indicates in Sections 2.1 and 2.2 that licensees should address compliance with current regulations and traditional engineering considerations, respectively, when requesting risk-informed changes to TS requirements. Provide this information consistent with the guidance provided in RG 1.177.
Note: In a conference call on December 13, 2006 the NRC Staff requested that the response to this question be formatted to match the RG format. To aid the reader the text below in italics is taken directly from RG 1.177.
2.1 Compliance with CurrentRegulations:
In evaluatingproposedchanges to TS, the licensee must ensure that the currentregulations, orders, and license conditions are met, consistent with Principle1 of risk-informedregulation.
The NRC regulationsspecific to TS are stated in 10 CFR 50.36, "TechnicalSpecifications."
Additional information with regardto the NRC's policies on TS is containedin the "FinalPolicy Statement on TechnicalSpecificationImprovements for Nuclear Power Reactors" (58 FR 39132) ofJuly 22, 1993 (Ref 9). These documents define the main elements of TS andprovide criteria for items to be included in the TS. The finalpolicy statement and the statement of considerations for 10 CFR 50.36 of July 19, 1995 (Ref.10), also discuss the use ofprobabilisticapproachesto improve TS. Regulationsregardingapplicationfor and issuanceof license amendments are found in 10 CFR 50.90, 50.91, and 50.92. In addition,the licensee should ensure that any discrepanciesbetween the proposedTS change and licensee commitments are identifiedand consideredin the evaluation.
Response
On page 10 of Attachment (1) to Reference 1, Ginna stated that "...such activities will be conducted in compliance with the Commission's regulations..." To validate that statement the relevant regulation, 10CFR50.36, was reviewed and it was determined that the requested amendment will not violate the provisions of that regulation. Specifically, the change was focused on the Limiting Condition for Operation (LCO), with a minor wording change to a Surveillance Requirement (SR) for clarification. Neither of these changes will result in a non compliance with the associated sections of 10CFR50.36. In addition, a search of Ginna Attachment (1)
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commitments indicated no regulatory commitments which conflict with the proposed amendment.
2.2 TraditionalEngineeringConsiderations
2.2.1 Defense in depth The engineeringevaluation conductedshould determine whether the impact of the proposedTS change is consistent with the defense-in-depthphilosophy. In this regard,the intent of the principleis to ensure that the philosophy of defense in depth is maintained,not to prevent changes in the way defense in depth is achieved The defense-in-depthphilosophy has traditionallybeen applied in reactordesign and operationto provide multiple means to accomplish safetyfunctions andprevent the release of radioactivematerial.It has been and continues to be an effective way to accountfor uncertaintiesin equipment andhuman performance. When a comprehensive risk analysiscan be performed, it can be used to help determine the appropriateextent ofdefense in depth (e.g., balance among core damage prevention, containmentfailures,and consequence mitigation) to ensure protection ofpublic health and safety. When a comprehensive risk analysis is not or cannot be performed, traditional defense-in-depth considerationsshould be used or maintainedto accountfor uncertainties.The evaluationshould consider the intent of the generaldesign criteria,nationalstandards,and engineeringprinciplessuch as the singlefailure criterion.Further,the evaluationshould consider the impact of the proposedTS change on barriers(both preventive andmitigative) to core damage, containmentfailure or bypass, and the balance among defense-in-depth attributes.
As stated earlier,the licensee should select the engineeringanalysis techniques, whether quantitativeor qualitative,traditionalorprobabilistic,appropriateto the proposedTS change.
The licensee should assess whether the proposedTS change meets the defense-in-depth principle.Defense in depth consists of a number of elements as summarized below. These elements can be used as guidelinesfor assessing defense in depth. Other equivalent acceptance guidelines may also be used.
Consistency with the defense-in-depthphilosophy is maintainedif:
A reasonablebalance among prevention of core damage,prevention of containment failure,and consequence mitigationis preserved,i.e., the proposedchange in a TS has not significantly changedthe balance among theseprinciples ofprevention and mitigation, to the extent that such balance is needed to meet the acceptance criteriaof the specific design basis accidentsand transients,consistent with 10 CFR 50.36. TS change requests should considerwhether the anticipatedoperationalchanges associated with a TS change could introduce new accidents or transientsor could increase the likelihood ofan accident or transient(as is requiredby 10 CFR 50.92).
Response
10CFR50.36 states in part, "(2) Limiting conditionsfor operation.(i) Limiting conditions for operation are the lowest functional capability or performance levels of equipment required for safe operation of the facility. When a limiting condition for operation of a Attachment (1)
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nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the technical specifications until the condition can be met." The safety related function of the SW System is to provide cooling for safety related equipment, mitigate the containment response effects of a Main Steam Line Break (MSLB) and design basis Loss of Coolant Accident (LOCA), and provide long term containment and core cooling in the event of a LOCA. The most limiting phase of either accident for SW availability is the recirculation phase of the design basis Loss of Coolant Accident (LOCA). Two SW pumps are required for this phase to ensure appropriate SW flow to all components. Reference 1, Attachment 1, Table 3 demonstrates that the proposed change ensures that two SW pumps will be available for accident mitigation, using industry accepted single failure assumptions. Therefore the proposed change maintains defense in depth consistent with 10CFR50.36 requirements, without a shift in the balance between prevention and mitigation. 10CFR50.92 requirements are addressed in Reference 1, Attachment 1, Section 5, No Significant Hazard Determination.
Over-relianceon programmaticactivities to compensatefor weaknesses in plant design is avoided, e.g., use of high reliabilityestimates that areprimarilybased on optimistic program assumptions.
Response
Reference 1, Attachment 2, Section 1.1.1.1 details the availability and reliability data assumed in the risk analysis. The information indicates that the proposed change in allowed outage time will not cause future unavailability to increase, largely because all planned maintenance is currently performed on line. SW pump reliability estimates are realistic, as they are based on generic industry data combined with actual plant data.
System redundancy,independence, and diversity are maintainedcommensurate with the expectedfrequency and consequences of challenges to the system, e.g., there are no risk outliers. The following items should be considered.
Whether there are appropriaterestrictionsin place to precludesimultaneous equipment outages that would erode the principles of redundancyand diversity,
Response
Reference 1, Attachment 1, Table 3 indicates the built in protection of redundancy due to restrictions relating to Diesel Generator availability, thereby protecting the redundancy of the system from a loss of power scenario. Ginna's Safety Function Determination Program required by TS 5.5.14 also ensures that there is no loss of safety function.
Additionally, Ginna's work control process maintains redundancy through extensive work planning, scheduling and review. On line risk monitors are used in the planning and scheduling process to ensure that the overall plant risk is maintained within acceptable levels in accordance with 10CFR50.65. On-line work is scheduled, monitored and controlled such that key system functions are Attachment (1)
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protected by a Defense-in-Depth strategy and the plant is maintained within an acceptable risk envelope by incorporating results of risk management tools. On-line maintenance is scheduled and performed such that unavailability time is minimized.
Whether compensatory actions to be taken when entering the modifiedAOTfor pre-plannedmaintenanceare identified,
Response
Compensatory measures are not required by the TS itself because the requirements are structured to maintain the necessary redundancy and diversity.
However, as an added layer of conservatism, Ginna's scheduling and risk/work management procedures identify and implement compensatory measures where appropriate.
Whether voluntary removal of equipmentfrom service duringplant operation should not be scheduledwhen adverse weatherconditions are predictedor at times when the plant may be subjected to other abnormalconditions
Response
Ginna's scheduling and risk/work management procedures evaluate the potential impact of severe weather or other external conditions relative to the proposed scheduled maintenance. The impact is evaluated if such weather conditions are imminent or have a high probability of occurring during the planned out-of-service duration. Consideration is given to rescheduling or expediting the return to service of equipment important to safety.
Whether the impact of the TS change on the safetyfunction should be taken into consideration.Forexample, what is the impact of a change in the AOTfor the low-pressure safety injection system on the overall availabilityand reliabilityof the low-pressure injectionfunction?
Response
Per Reference 1, Attachment 2, Section 1.1.1.1, a review of the historical unavailability of the SW pumps for the four year data window used to develop the above data indicates that all but 78.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> (or 2.1%) of SW pump unavailability occurred while the plant was on line. Further, the entire 78.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> was unplanned maintenance, such that all planned maintenance of the SW pumps occurred on-line. This is due to the fact that the current Technical Specification (using the pre-uprate definition of train operability) allowed a single SW pump to be removed from service indefinitely, provided the other pump powered by the same electrical train was operable. Consequently, scheduled maintenance (as well as most corrective maintenance) was performed on-line. Since all planned maintenance was previously performed on-line, no increase in on-line SW pump unavailability is planned or anticipated.
Attachment (1)
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Defenses againstpotential common causefailures are maintainedand the potentialfor introductionof new common causefailure mechanisms is assessed,e.g., TS change requests shouldconsider whether the anticipatedoperationalchanges associatedwith a change in an AOT or STI could introduce any new common causefailure modes not previously considered.
Response
The method or mode of operation or the configuration of the SW system is not changed by this amendment. The system will continue to be operated, maintained and tested in the same manner as before. Therefore, no new common cause mechanisms are introduced. Additionally, Ginna is committing to an administrative requirement to evaluate for common cause in the event of a pump failure (Attachment 3).
Independence ofphysical barriersis not degraded,e.g., TS change requests should address a means of ensuring that the independence of barriershas not been degradedby the TS change (e.g., when changing TSfor containment systems).
Response
The relationship of the SW system to the individual barriers has will not change as a result of this amendment. The SW system will continue to support core cooling and containment cooling as before. Reference 1, Attachment 1 (specifically Table 3),
demonstrate that adequate SW will be available to support the system design function.
Defenses againsthuman errors are maintained,e.g., TS change requests should consider whether the anticipatedoperation changes associatedwith a change in an AOT or STI could change the expected operatorresponse or introduceany new human errorsnot previously considered,such as the changefrom performingmaintenanceduring shutdown to performing maintenance at power when different personnel anddifferent activitiesmay be involved.
Response
Operator response is not expected to change during normal, abnormal or emergency operating conditions. Any minor procedure changes will be addressed, and appropriate training conducted, as part of the change implementation process. Also, and as stated above, normal pump maintenance is currently performed on-line, which is not expected to change as a result of the amendment. Therefore, defense against human errors is maintained.
The intent of the GeneralDesign Criteriain Appendix A to 10 CFR Part50 is maintained.
Response
The following General Design Criteria were reviewed to ensure the intent of each criteria is maintained, considering the proposed amendment:
Attachment (1)
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IV. Fluid Systems Criterion 34--Residual Heat Removal (RHR)
This system is indirectly supported by the SW system via the Component Cooling Water (CCW) System. The change will not impact the ability of the SW system to support the RHR system. The RHR system will continue to receive sufficient cooling to perform its function in both the accident mode or during normal operation, given industry accepted power and single failure assumptions.
Criterion 35--Emergency Core Cooling System (ECCS)
This system is supported by the SW system. The change will not impact the ability of the SW system to support the ECCS system. The ECCS system will continue to receive sufficient SW flow to perform its function in both the accident mode or during normal operation, assuming industry accepted power and single failure assumptions.
Criterion 38--Containment heat removal This system, specifically the Containment Recire Fan Coolers (CRFCs), is supported by the SW system. The change will not impact the ability of the SW system to support the CRFCs. Analysis shows that two SW pumps are required to support the all cooling loads during Loss of Coolant Accidents. The SW system will continue meet this requirement assuming industry accepted power and single failure assumptions.
Criterion 44--Cooling water The ability of the SW system to transfer heat from structures, systems and components important to safety to the ultimate heat sink is not diminished by this change. Reference 1, Attachment 1, Table 3 demonstrates that the system maintains suitable redundancy (i.e., two pumps available) assuming off-site power is unavailable, and a single failure. System interconnections, leak detection and isolation capabilities are not affected by this change.
Criterion 45--Inspection of cooling water system The ability to inspect, or frequency or method of inspection, of the SW system is not affected by this change.
Criterion 46--Testing of cooling water system The ability to test, or frequency or method of testing, is not affected by this change.
2.2.2 Safety Margins Attachment (1)
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The engineeringevaluation conductedshould assess whether the impact of the proposed TS change is consistent with the principle that sufficient safety margins are maintained(Principle 3). An acceptable set ofguidelinesfor making that assessment are summarized below. Other equivalent decision guidelines are acceptable.
Sufficient safety margins are maintainedwhen:
Codes and standards(e.g., American Society of MechanicalEngineers (ASME), Institute ofElectricalandElectronicEngineers (IEEE)or alternativesapprovedfor use by the NRC are met, e.g., the proposed TS AOT or STI change is not in conflict with approved Codes and standardsrelevant to the subject system.
Response
A search was conducted of the Ginna UFSAR, the Ginna Service Water System Reliability Optimization Program (SWSROP), and NRC Generic Letter 89-13. The applicable code or standard potentially affected by this change is ANSIIANS-58.9, Single Failure Criteria for Light Water Reactor Safety-Related Fluid Systems. Reference 1, Attachment 1, Table 3 demonstrates the different failure combinations and associated action statements. The summary for that section states, "The technical evaluation demonstrates that the required number of SW pumps will be available at all times given single failure criteria, or immediate action will be taken to place the plant in a safe mode." Therefore the safety margin relative to codes and standards is maintained.
Safety analysisacceptance criteriain the FinalSafety Analysis Report (FSAR) are met, orproposedrevisionsprovide sufficient margin to accountfor analysis anddata uncertainties,e.g., the proposedTS AOT or STI change does not adversely affect any assumptions or inputs to the safety analysis, or, if such inputs are affected,justificationis providedto ensure sufficient safety margin will continue to exist. For TS AOT changes, an assessmentshould be made of the effect on the FSAR acceptancecriteriaassuming the plant is in the AOT (i.e., the subject equipment is inoperable)andthere are no additional failures.Such an assessment should result in the identificationof all situations in which entry into the proposedAOT could result in failure to meet an intendedsafetyfunction.
Response
This assessment was performed in Reference 1, Attachment 1. Table 3 demonstrates that the minimum number of SW pumps assumed to be operating in the accident analysis (2) is maintained for any combinations resulting in an AOT, assuming no additional failures.
If not, immediate action is taken to place the plant in a safe mode. Therefore, the safety analysis criteria is maintained following this change.
Question 2 In Section 1.1.7.4 of Attachment 2 of the application, credit is taken in the risk analysis for repair and recovery of an Inoperable SW pump. The NRC staff does not consider it appropriate to assume recovery of inoperable equipment as part of the justification that the risk associated with that equipment being inoperable is small. Therefore, provide revised risk Attachment (1)
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calculations that do not assume any credit for repair of inoperable SW pumps.
Response
SW pump repair is not credited post-trip. The recoveries are only applied to the 'Loss of Service Water' initiating event fault tree model, not the post-initiator failure of SW pumps. It was conservatively considered that any shutdowns caused by exceeding the allowed out-of-service times would result in a loss of service water initiating event. To develop a realistic initiating event frequency, it was considered that a failed SW pump could potentially be repaired prior to the failure of a second SW pump. As these recoveries are considered pre-trip with no other initiating event in progress, all available plant resources could, and likely would, be focused on restoring one of the SW pumps to service, as opposed to a post-transient situation where resources would be focused on safely shutting down the plant and maintaining it in a safe condition. However, as noted in section 1.1.9.3, a sensitivity analysis of these non-recovery factors indicates that the overall results of the analysis have very little sensitivity to these events. Therefore, to eliminate any concerns regarding the appropriateness of these recovery events, they have been eliminated from the analysis (i.e., their failure probability has been set to 1.0), and the Incremental Conditional Core Damage Probability (ICCDP) values have been re-calculated.
The results of the updated ICCDP analysis, which also incorporates increased common cause failure probabilities (see RAI 5), an explicit seismic risk evaluation (see RAI 6),
and the Incremental Conditional Large Early Release Probability (ICLERP) analysis requested in RAI 3 (which also accounts for the responses to RAI's 2, 5, and 6), are presented below.
Attachment (1)
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ICCDP Single Pump Impacts SW Pump OOS 14 Day AOT A 1.006E-08 B 9.977E-09 C 1.006E-08 D 9.977E-09 Two Pump Impacts SW Pumps OOS 72 Hour AOT A&B 5.416E-08 A &C 5.088E-08 A&D 5.416E-08 B&C 5.416E-08 B&D 5.088E-08 C&D 5.416E-08 ICLERP Single Pump Impacts SW Pump OOS 14 Day AOT A 8.887E-11 B 8.882E-11 C 8.887E-11 D 8.822E-11 Two Pump Impacts SW Pumps OOS 72 Hour AOT A&B 6.740E-10 A&C 6.329E-10 A&D 6.740E-10 B&C 6.740E-10 B&D 6.321E-10 C&D 6.740E-10 As can be seen by comparing these results to the original submittal results in Reference 1, all ICCDP values have increased. However, all are still well below the 5E-07 criteria.
Similarly, all ICLERP values are well below the 5E-08 criteria.
Question 3 In Section 1.1.3 of Attachment 2 no specific incremental conditional large Attachment (1)
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early release probability (ICLERP) calculations were performed. The stated basis was that even if 100% of the sequences contributing to the incremental conditional core damage probability (ICCDP) became large early releases, all cases calculated would be at least a factor of 10 below the RG 1.177 criteria for ICLERP. This implicitly assumes that any increase in the large early release frequency (LERF) would only arise from sequences contributing to increases in core damage frequency (CDF), and not from existing core damage sequences, which would now become large early releases. Justify this assumption, or provide ICLERP analysis for the proposed changes.
Response
An ICLERP analysis for the proposed change has been performed. The results of this ICLERP analysis, taking into account the removal of pre-trip SW pump recoveries (RAI 2), increased common cause failure probabilities (RAI 5), and an explicit seismic risk evaluation (RAI 6), are presented in the above response to RAI 2.
Question 4 In Section 1.1.3 of Attachment 2, the licensee does not identify whether the evaluation of ICCDP was based on a zero maintenance model or on a average maintenance model. Identify the basis of the ICCDP and ICLERP calculations.
Response
The evaluation of ICCDP and ICLERP evaluation are based on an average maintenance model. That is, the unavailability of the SW pumps is explicitly addressed, while all other maintenance unavailability events have been set to their average values.
Question 5 Section 1.1.3 of Attachment 2, the licensee has not provided separate risk calculations for planned vs. unplanned maintenance assuming a higher common-cause failure rate, consistent with RG 1.177, Appendix A, Section A.1.3.2. Provide these calculations of ICCDP and, if necessary (based on the response for RAI 2), of ICLERP.
Response
A revised analysis of ICCDP and an analysis of ICLERP have been performed to address this issue. For the updated analysis, increased common cause failure rates for the SW pumps failing to start were used, assuming that the unplanned maintenance, for either one or two pumps out of service, was caused by a failure or failures of a stand-by SW pump to start. The increased common cause failure rates used are the conditional common cause failure rates, given that one or two pumps have already failed. To address the Attachment (1)
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potential for common cause failure of operating pumps, administrative controls will be put in place to require that if one SW pump is inoperable due to equipment failure, and a second SW pump fails before the first pump is returned to service, then an evaluation of possible common cause and determination of the operability of the remaining pumps will be performed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the second failure.
A sensitivity analysis was performed to demonstrate that the risk associated with two pumps failed with the remaining pumps potentially subject to the same common cause failure is acceptable for a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> time period. For all combinations of two pumps out of service, the higher common-cause failure rates provide at least 33 hours3.819444e-4 days <br />0.00917 hours <br />5.456349e-5 weeks <br />1.25565e-5 months <br /> of exposure before the ICCDP exceeds 5.0E-07 or the ICLERP exceeds 5E-08. Thus, even if the full 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> were used, and then the plant entered a TS required shutdown, the plant would be at hot shutdown within 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />, prior to the accumulated risk exceeding the allowed thresholds.
Based on the above requirement, the revised ICCDP analysis and the ICLERP analyses include increased common cause failure to run rates for a single SW pump out of service, or two SW pumps out of service where one is for planned maintenance. This is necessary-since an explicit determination that no common cause failure mode exists is not required for these cases. The results of this revised analysis, incorporating the increased fail to start and fail to run common cause failure rates are included in the updated ICCDP and ICLERP values presented in the response to RAI 2.
Question 6 Section 1.1.5 of Attachment 2, the Ginna probabilistic safety assessment (PSA) model covers both internal and external events, including fire and external flooding. However, the licensee has not provided any quality information for the fire and external flooding models, nor provided any disposition of other external events which are not quantitatively addressed by the Ginna PSA model. Since the potential risk increase due to the proposed changes is due to loss-of-coolant accidents (LOCAs) of size equivalent to 2-inches diameter or greater (Section 1.1.7.2), confirm that the external events included in the quantitative risk assessment model (i.e., fire and external floods) do not potentially induce LOCAs greater than 2 inches in diameter requiring two SW pumps for mitigation. Otherwise, provide quality information for the external events probabilistic risk assessment model used to support the risk evaluation. In either case, the licensee should disposition seismic risk and other external events risk for this application.
Response
Although external floods are addressed in the Ginna PSA model, they do not have the capability to cause a LOCA event of 2" or greater. In order for a LOCA of 2" or greater to be caused by a transient initiator, both power operated relief valves (PORVs) must be challenged due to the RCS pressure excursion caused by the plant trip, and then both PORVs must fail to reseat. Only certain plant trip events create a pressure excursion in Attachment (1)
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the reactor coolant system (RCS) which challenges the PORVs (e.g., large loss of electrical load, complete loss of feedwater, locked RCP rotor, etc.). An external flooding event at Ginna is a slowly developing event (i.e., over several hours) resulting from rising levels in Lake Ontario, or in Deer Creek which runs by the plant. By procedure, prior to water levels rising to the point that plant equipment is impacted, a controlled load reduction is required. The plant should then be at or near hot shutdown conditions prior to any potential equipment impact from the flood. Thus, a plant trip caused by an external flooding event would not challenge the PORVs. Even if the plant was at or near 100%/o power and operators manually tripped the plant due to imminent impact of rising water on plant equipment, the PORVs would not be challenged since secondary side heat removal capability is more than adequate to remove decay heat levels, preventing primary side heat-up. Thus, for external flooding events a PORV challenge is precluded, and a LOCA of 2" or greater will not result.
Ginna fire events can potentially induce a LOCA greater than 2 inches in diameter. The fire events can result in a reactor trip which challenges two PORVs. If both PORVs fail to reseat, a LOCA of equivalent size of> 2" will result. No peer review has been performed on the fire portion of the Ginna PSA, in part due to the fact that there is no approved standard for fire PRA quality. However, a significant internal review of the fire modeling has taken place. Any time that the Ginna PSA is updated, the model is quantified, and the resulting cutsets are examined for reasonableness. Further, since the fire PSA uses the existing top logic from the internal events PRA, which has been subjected to a Peer Review, once the fire impacts have been incorporated into the model correctly, the validity of the top logic is ensured. Specifically for this submittal, the cutset files were reviewed to ensure that the expected scenario (the transient induced PORV LOCAs described above) appeared. Additionally, fires have a very minor impact on this submittal. For both the one pump and two pumps out of service cases, fire events contribute less than 1% of the overall ICCDP and ICLERP. Thus, the quality of the fire portion of the PSA is considered adequate for the purposes of this submittal.
Ginna Station does not have a seismic PSA. However, to address this issue, a seismic analysis has been performed, similar to the one provided in Ginna's Extended Power Uprate (EPU) submittal, to assess the risk from a seismic event as a result of the AOT extension. As the delta risk calculations are only applicable to LOCAs greater than 2",
the additional fragility of the RCS is considered. As with the EPU evaluation, a seismically induced loss of off-site power (LOOP) (with no recovery [i.e. 24 hr LOOP duration]) is calculated to occur at a frequency of 1E-4 per yr. Further, the RCS is considered a seismically rugged structure with a High-Confidence-Low-Probability-of-Failure (HCLPF) of 0.3 g. This equates to a conditional probability of 2.8% that a LOCA occurs given a seismically induced LOOP. This LOCA frequency is parsed over the spectrum of LOCA break sizes. The contribution of the seismic to the SW license amendment request is approximately 9% for the single pump out of service cases, and approximately 17% for the two pumps out of service cases. Given this relatively small contribution, the very conservative modeling used, and the significant margin available Attachment (1)
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below the ICCDP and ICLERP thresholds, this level of detail is considered adequate.
The results of this seismic analysis are included in the updated ICCDP and ICLERP values presented in the response to RAI 2.
Question 7 In Section 1.1.2 of Attachment 2, it states that the Ginna PSA model, revision 6.2, has a calculated internal and external event CDF of less than 1E-4 per year, and therefore justifies the application of RG 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," criteria for risk metrics. Provide the specific value of the CDF and LERF of the baseline model.
Response
The baseline values for CDF and LERF in revision 6.2 of the Ginna PSA model are 8.82E-05/yr and 4.13E-06/yr, respectively. This includes internal events (including floods), fires, shutdown risk, external floods, and the seismic evaluation discussed in the response to question 6, above. Of these values, 18% of CDF and 14% of LERF is shutdown risk.
Question 8 In Section 1.1.10.1 of Attachment 2 it states that 33 of 35 B-level facts and observations (F&Os) from the licensee's 2002 peer review have been addressed. However, the NRC staff review of the detailed information on F&Os identified 36 B-level F&Os. It is not apparent which of the B-level F&Os have not yet been completed. Therefore, specifically identify the F&Os not yet resolved and provide a disposition for each of these F&Os relevant to this application.
Response
This information was provided in Reference 3.
Question 9 In Section 1.3 of Attachment 2, the licensee identifies Sections 1.3.4 and 1.3.5 to address the risk of level 2 and external events in configuration risk management program. These sections are missing from the submittal.
Provide this missing information.
Response
The missing sections were provided in Reference 3.
Attachment (1)
Page 14 of 14
Attachment (2)
List of Regulatory Commitments
Attachment (2)
List of Regulatory Commitments The following table identifies those actions committed to by R.E. Ginna Nuclear Power Plant, LLC in this document. Any other statements in this submittal are provided for information purposes and are not considered to be regulatory commitments. Please direct questions regarding these commitments to Robert Randall at (585) 771-5219, or Robert. Randallconstellation.com.
REGULATORY COMMITMENT DUE DATE Administrative controls will be put in place to require that if one SW pump is inoperable due Upon implementation of the TS amendment to equipment failure, and a second SW pump fails before the first pump is returned to service, an evaluation of possible common cause and determination of the operability of the remaining pumps will be performed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the second failure.
Attachment (3)
Proposed Technical Specification Changes (Revised Mark-up)
-'a
SW System 3.7.8 3.7 PLANT SYSTEMS 3.7.8 Service Water (SW) System LCO 3.7.8 T-we SW #ain and the SW loop header shall be OPERABLE.
APPLICABILITY: MODES 1, 2, 3, and 4.
ACTIONS (Eýý)
CONDITION REQUIRED ACTION A. One SWtý!n Inoperable. A.1 Restore SW 4-ia~to OPERABLE status.
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,*. Required Action and Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> C associated Completion Time of Condition Anot met.
or- Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> C-
.4 S F-Two SW h or loop Z.l header inoperable.
D D
-NOTE-T-Ar-ec of- rviorlfe 7M ) Enter applicable conditions and Required Actions of LCO 3.7.7, "CCW System,"
for the component cooling water heat exchanger(s) made Inoperable by SW.
Enter LCO 3.0.3. Immediately
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- 13. Two -5W P ot ny & ) e e) 0 P 6 kA U L E esto lu S R.E. Ginna Nuclear Power Plant 3.7.8-1 Amendment SW System 3.7.8 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.8.1 Verify screenhouse bay water level and temperature 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> are within limits.
- NOTE -
SR 3.7.8.2 Isolation of SW flow to individual com onents does not render the SW loop header rable.
.op Verify each SW manual, power perated, and 31 days automatic valve in the SW flow path and loop header that is not locked, sealed, or otherWise secured in position, is in the correct position.
SR 3.7.8.3 Verify all SW loop header cross-tie valves are locked 31 days in the correct position.
SR 3.7.8.4 Verify each SW automatic valve in the flow path that is 24 months not locked, sealed, or otherwise secured In position, actuates to the correct position on an actual or simulated actuation signal.
SR 3.7.8.5 Verify each SW pump starts automatically on an 24 months actual or simulated actuation signal.
R.E. Ginna Nuclear Power Plant 3.7.8-2 Amendment 4-
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