ML072880107: Difference between revisions

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#REDIRECT [[RS-07-136, Braidwood, Units 1 and 2, Byron, Units 1 and 2, Response to NRC Request for Additional Information Concerning Application for License Amendment to Implement WCAP-14333 and WCAP-15376]]
| number = ML072880107
| issue date = 10/12/2007
| title = Braidwood, Units 1 and 2, Byron, Units 1 and 2, Response to NRC Request for Additional Information Concerning Application for License Amendment to Implement WCAP-14333 and WCAP-15376
| author name = Simpson P R
| author affiliation = Exelon Generation Co, LLC, Exelon Nuclear
| addressee name =
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000454, 05000455, 05000456, 05000457
| license number = NPF-037, NPF-066, NPF-072, NPF-077
| contact person =
| case reference number = RS-07-136, TAC MD4009, TAC MD4010, TAC MD4011, TAC MD4012, WCAP-14333, WCAP-15376
| document type = Letter
| page count = 17
| project = TAC:MD4009, TAC:MD4010, TAC:MD4011, TAC:MD4012
| stage = Response to RAI
}}
 
=Text=
{{#Wiki_filter:RS-07-136 October 12, 2007 wvdw txel oncerp.co m U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Braidwood Station, Units 1 and 2 Facility Operating License Nos. NPF-72 and NPF-77 NRC Docket Nos. 50-456 and 50-457 Byron Station, Units 1 and 2 Facility Operating License Nos. NPF-37 and NPF-66 NRC Docket Nos. 50-454 and 50-455 Exel6n Nuclear
 
==Subject:==
Response to NRC Request for Additional Information Concerning Application for License Amendment to Implement WCAP-14333 and WCAP-15376 References
: 1. Letter from D. M. Benyak (Exelon Generation Company, LLC) to U.S. NRC, "Application for License Amendment: Implementation of WCAP-14333 and WCAP-15376, Reactor Trip System Instrumentation and Engineered Safety Feature Actuation System Instrumentation Test Times, Completion Times and Surveillance Test Intervals," dated January 7, 2007 2. Letter from R. F. Kuntz (U. S. NRC) to C. M. Crane (Exelon Generation Company, LLC), "Byron Station, Unit Nos. 1 and 2, and Braidwood Station, Units 1 and 2 - Request for Additional Information Related to License Amendment Request to Revise Technical Specification Requirements for Select Reactor Trip System, Engineered Safety Feature Actuation System, and Containment Ventilation Isolation, Instrumentation (TAC Nos. MD4009, MD4010, MD4011, and MD4012)," dated August 17, 2007 In Reference 1, Exelon Generation Company, LLC (EGC) submitted a request to amend Appendix A, Technical Specifications (TS) for Facility Operating License Nos. NPF-72, NPF-77, NPF-37, and NPF-66 for Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2, respectively. The proposed amendment would revise the TS requirements for selected Reactor Trip System (RTS) instrumentation, Engineered Safety Feature Actuation System (ESFAS) instrumentation, and Containment Ventilation Isolation Instrumentation to adopt Completion October 12, 2007 U. S. Nuclear Regulatory Commission Page 2 Time, test bypass time, and Surveillance Test Interval changes. The proposed changes are based on Westinghouse Electric Company LLC topical reports WCAP-14333-P-A, Revision 1, "Probabilistic Risk Analysis of the RPS and ESFAS Test Times and Completion Times," and WCAP-15376-P-A, Revision 1, "Risk-Informed Assessment of the RTS and ESFAS Surveillance Test Intervals and Reactor Trip Breaker Test and Completion Times." In Reference 2, the NRC provided a request for additional information (RAI) related to the proposed license amendment. In response to this request, EGC is providing the information in Attachment 1 to this letter. There is one new regulatory commitment and two revised regulatory commitments described in this response. The new and the revised regulatory commitments are described in Attachment
: 2. Any other statements in this submittal are provided for information purposes and are not regulatory commitments. Should you have any questions related to this letter, please contact Mr. John L. Schrage at (630) 657-2821. I declare under penalty of perjury that the foregoing is true and correct. Executed on the 12`h day of October 2007. Patrick R. Simpson Manager - Licensing Attachments
: R 1. Response to Request for Additional Information, License Amendment Request to Implement WCAP-14333 and WCAP-15376
: 2. Summary of Regulatory Commitments cc: NRC Regional Administrator, Region III NRC Senior Resident Inspector, Braidwood Station NRC Senior Resident Inspector, Byron Station ATTACHMENT 1 Response to Request for Additional Information License Amendment Request to Implement WCAP-14333 and WCAP-15376 NRC Request 1 The analysis for WCAP-14333/WCAP-15376 assumed that maintenance on master and slave relays, logic cabinets, and analog channels while at power occurs only after a component failure, and that preventive maintenance does not occur. The topical report does not preclude the practice of at-power preventive maintenance but limits the total time a component is unavailable due to corrective or preventive maintenance to the values used in the analysis. If preventive maintenance is to be performed at Byron/Braidwood Stations Units 1 and 2, confirm that the unavailability for components evaluated in WCAP-14333 are consistent with the plant specific estimates at Byron/Braidwood and do not exceed those assumed in the analysis. See the submittal, Attachment 6, Page 1, "Implementation Guidelines," Table 1, "Analog Channel Calibration" as an example. Response Exelon Generation Company, LLC (EGC) does not perform at-power preventative maintenance on master and slave relays, logic cabinets, and analog channels. NRC Request 2 Provide an assessment of external events risk impact including, seismic, fire, and external floods/high wind with respect to the proposed completion times (CT) and surveillance test interval (STI) extensions, per Regulatory Guide (RG) 1.174 Section 2.2.4, "Acceptance Guidelines" and RG 1.177 Section 2.3.2, "Scope of the Probabilistic Risk Assessment (PRA) for technical specification (TS) Applications
." Include any seismic vulnerabilities associated with instrumentation/logic systems or components and risk impacts on fire screening criteria and quantification of fire sequences in unscreened areas. In addition, confirm that the combined total CDF from external and internal events remains less than the RG 1.174 base CDF of 1 E-4 per year. Response In the Braidwood Station and Byron Station Individual Plant Examination of External Events (IPEEE), the dominant risk contribution was from internal fire initiating events. The fire risk quantification approach that was used involved the application of fire-initiating events and fire scenario-related effects on mitigating systems to pertinent internal events in at-power PRA accident sequences. The fire damage impact is typically assigned a probability of 1.0. Thus, any mitigating equipment that might be affected by the postulated fire is assumed to fail in the applicable internal events model scenario (i.e., those scenarios used to represent the plant response to the fire-initiating event). The fire risk is generally driven by this conservative fire damage assumption. The effect of the proposed change will be associated with the test and maintenance unavailabilities for the Reactor Trip System (RTS) and Engineered Safety Feature Actuation System (ESFAS) that are used in the internal events model. Thus, the impact of the proposed changes on the test and maintenance unavailabilities are expected to be a second order effect (i.e., smaller) for fire core damage sequences, relative to the effect in the internal events model. Other potential contributors, such as seismic effects and high winds, were within acceptable limits, as discussed below. The seismic analyses in the Braidwood Station and Byron Station IPEEEs were based on the seismic margin assessment. No significant seismic concerns were identified, and as such, both Page 1 ATTACHMENT 1 Response to Request for Additional Information License Amendment Request to Implement WCAP-14333 and WCAP-15376 Stations possess significant seismic margin. The IPEEEs did identify some seismic outliers (i.e., control room lighting diffuser panels and electrical cabinet interactions). These seismic outliers were resolved with the implementation of design modifications in 2003. The proposed change to the RTS, ESFAS, and Containment Ventilation Isolation Instrumentation Completion Times (CTs) and Surveillance Test Intervals (STIs) has a negligible effect on the seismic risk profile at both stations. Evaluation of high winds, external floods, and other external events in the Braidwood Station and Byron Station IPEEEs have been completed in accordance with the requirements of NUREG-1407, "Procedural and Submittal Guidance for the Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities
." These evaluations indicated that both Stations conform to the acceptance criteria defined in NUREG-0800, Revision 1, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants," with no potential vulnerabilities. The proposed change to the RTS, ESFAS, and Containment Ventilation Isolation Instrumentation CTs and STIs do not impact these conclusions. Since EGC does not currently have quantitative external event models or internal fire risk models for either Braidwood Station or Byron Station that are sufficiently developed and peer reviewed to enable EGC to combine the risk contribution from external events with that of internal events, EGC has utilized a common approach to estimate the risk from external events and combine this estimate with the risk from internal events. This approach assumes that the risk from external events is equal to the risk from internal events. Additionally, EGC has assumed that the fractional LERF contribution from the internal events model also provides a reasonable estimate of the LERF impact from external events. EGC has recently approved a revised Braidwood Station and Byron Station PRA. In part, this new model incorporates a revision to the internal flooding analysis, specifically a revision to the internal flooding Human Reliability Analysis (HRA) to meet the requirements of Regulatory Guide 1.200, Revision 0, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities
." Utilizing this revised model, in concert with the approach described above for estimating the risk contribution from external events, the combined total CDF from internal and external events is below the 1 E-04 threshold in RG 1.174, Revision 1 for both Braidwood Station and Byron Station. These results assume that, in part, each Station has implemented, or will implement enhanced internal flood response procedures prior to implementation of the proposed changes. At the current time, Braidwood Station has implemented these procedures. EGC will implement the enhanced internal flood response procedures at Byron Station prior to implementation of the proposed changes. [Regulatory Commitment]
ATTACHMENT 1 Response to Request for Additional Information License Amendment Request to Implement WCAP-14333 and WCAP-15376 NRC Request 3 Provide a discussion on the following aspects of probability risk assessment (PRA) quality for Byron/Braidwood. NRC Request 3.a Provide the date of Byron and Braidwood Stations PRA industry peer review and date of certification. Provide the results of the Peer review including disposition of "A" and "B" facts and observations (F&Os). Response The Braidwood Station PRA was subjected to a Westinghouse Owners' Group (WOG) peer review in September 1999. The Byron Station PRA was subjected to a separate (WOG) peer review in July 2000. No peer reviews have taken place since that date for either site. Since those peer reviews, the PRA model (the Braidwood Station and Byron Station PRA models are very similar and exist in an integrated model) has undergone both scheduled and interim model updates. These model updates were implemented in June 2001, February 2002, December 2002, May 2003, June 2003, and July 2005. In addition, EGC recently completed a model update during preparation of this response (i.e., October 2007). EGC transmitted the results of the Braidwood Station and Byron Station PRA peer reviews, including disposition of "A" and "B" facts and observations (F&Os), as part of various risk-informed submittals. The transmittal letters for previous submittals of peer review findings to the NRC are listed below. Letter from R. M. Krich (Commonwealth Edison Company) to USNRC, "Response to a Request for Additional Information related to 'Request for Amendment to Technical Specifications - Extension of Allowable Completion Times and Surveillance Requirement Change for Emergency Diesel Generators'," dated July 7, 2000 Letter from K. A. Ainger (EGC) to USNRC, "Response to a Request for Additional Information Regarding a Technical Specification Change Request - Extension of Completion Time for Instrument Bus Inverters," dated June 20, 2003 Letter from Letter from K. A. Ainger to USNRC, "Request for Additional Information Regarding a License Amendment for a One-Time Extension of the Essential Service Water Train Completion Time," dated December 5, 2003 Letter from D. M Benyak (EGC) to USNRC, "Additional Information Supporting Risk-Informed Inspection Relief Request," dated April 3, 2007 The few remaining open A and B F&Os from the Braidwood Station and Byron Station peer reviews are listed in Table 1. The significance of the open F&Os with respect to the proposed changes is provided in the far-right column. F&Os that have been fully addressed from a technical perspective in the recently completed update no longer have any potential impact on the proposed changes, and are noted as such. These F&Os are still listed in the table since the associated PRA updating requirements evaluation (URE) tracking activities have not yet been closed, as of the date of this response.
ATTACHMENT 1 Response to Request for Additional Information License Amendment Request to Implement WCAP-14333 and WCAP-15376 NRC Request 3.b The plant-specific PRA reflects the as-built, as-operated plant. Response The Braidwood Station and Byron Station PRA model that was used to validate the applicability of the WCAP-14333 and WCAP-15376 analyses to the RTS and ESFAS instrumentation for each respective station was approved in September 2005, and reflected the as-built and as-operated plants, as of that time. Subsequent to submittal of this license amendment request, EGC approved a revised Braidwood Station and Byron Station PRA model. This new PRA model incorporates recent design and operational changes. EGC has reconfirmed the applicability of WCAP-14333 and WCAP-15376 to Braidwood Station and Byron Station, using data from the recently completed PRA model revision. The impact of plant design or operational modifications that are not reflected in the PRA model that was used to validate the applicability of the WCAP-14333 and WCAP-15376 analyses are discussed in the response NRC Request 3.f below. From a programmatic perspective, the EGC risk management process ensures that the applicable PRA model remains an accurate reflection of the as-built and as-operated plants. This process is defined in the EGC Risk Management program, which consists of a governing procedure (ER-AA-600, "Risk Management")
and subordinate implementation procedures. EGC procedure ER-AA-600-1015, "FPIE PRA Model Update" delineates the responsibilities and guidelines for updating the full power internal events PRA models at all active EGC nuclear generation sites. EGC transmitted Revision 7 of this procedure to the NRC in a letter dated April 3, 2007, as part of a risk-informed Inservice Inspection Relief Request for Byron Station. The overall EGC Risk Management program, including ER-AA-600-1015, defines the process for implementing regularly scheduled and interim PRA model updates, for tracking issues identified as potentially affecting the PRA models (e.g., due to changes in the plant, errors or limitations identified in the model, industry operational experience), and for controlling the model and associated computer files. To ensure that the current PRA model remains an accurate reflection of the as-built, as-operated plants, the following activities are routinely performed: Design changes and procedure changes are reviewed for their impact on the PRA model. New engineering calculations and revisions to existing calculations are reviewed for their impact on the PRA model. Maintenance unavailabilities are captured, and their impact on CDF is trended. Plant specific initiating event frequencies, failure rates, and maintenance unavailabilities are updated approximately every four years. In addition to these activities, prior to approval of a PRA model, the requirements of the EGC PRA quality assurance programs and procedures that are described in the response to NRC Request 3.d below are followed for model review.
ATTACHMENT 1 Response to Request for Additional Information License Amendment Request to Implement WCAP-14333 and WCAP-15376 NRC Request 3.c Applicable PRA updates conducted since completion of individual plant examination (IPE) and IPEEE, and the status of any improvements identified by the IPE and IPEEE. Response The response to NRC Request 3.a above delineates the PRA updates that EGC has completed since completion of the initial IPE and IPEEE. The response to NRC Request 2 above describes the results of the Braidwood Station and Byron Station IPEEE, including the status of any improvements identified by the IPEEE. The Braidwood Station and Byron Station initial IPE identified a significant risk contribution from internal flooding scenarios at each Station involving a dual unit loss of essential service water from flooding. The risk from internal flooding was substantially addressed and mitigated through several design changes and implementation of internal flooding response procedures, as discussed in the response to NRC Request 2 above. The design changes included the installation of equipment that enables both Stations to use the fire protection system as an alternate means of providing cooling to the centrifugal charging pumps, which, in turn, would prevent loss of all reactor coolant pump (RCP) seal cooling (and the potential for an induced RCP seal loss of coolant accident) following a loss of essential service water. In addition, the discharge of the essential service water system at Braidwood was extended above water level to prevent the potential for draining the cooling lake into the Auxiliary Building, following a rupture of the essential service water system piping. This design change was not needed at Byron as the essential service water system was originally designed with a discharge above water level. NRC Request 3.d Reference PRA quality assurance programs/procedures, including expected PRA revision schedules. Response The Braidwood Station and Byron Station PRA model is developed, controlled, and maintained in accordance with EGC risk management procedure ER-AA-600. Subordinate EGC procedures provide the guidance for particular risk management and PRA quality and maintenance activities. This guidance includes: Documentation of the PRA model, PRA products, and bases documents. The approach for controlling electronic storage of Risk Management (RM) products including PRA update information, PRA models, and PRA applications. Guidelines for updating the full power, internal events (FPIE) PRA models for all active EGC nuclear generation sites.
ATTACHMENT 1 Response to Request for Additional Information License Amendment Request to Implement WCAP-14333 and WCAP-15376 Guidance for use of quantitative and qualitative risk models in support of the On-Line Work Control Process Program for risk evaluations for maintenance tasks (corrective maintenance, preventive maintenance, minor maintenance, surveillance tests and modifications) on SSCs within the scope of the Maintenance Rule (10CFR50.65 (a)(4)). In accordance with this guidance, regularly scheduled PRA model updates nominally occur on an approximately 4-year cycle; longer intervals may be justified if it can be shown that the PRA continues to adequately represent the as-built, as-operated plant. EGC is currently implementing a new revision to the Braidwood Station and Byron Station PRA model. Therefore, the next planned model update is 2011. The criteria for initiating interim model updates (i.e., updates between the regularly-scheduled updates) are provided in the EGC Risk Management procedures. NRC Re quest 3.e PRA adequacy and completeness with respect to evaluating the proposed CT, bypass test time under Tier 3 configuration risk management. Response EGC has reviewed the Braidwood Station and Byron Station PRA model to confirm its adequacy and completeness with respect to Tier 3 evaluations. Significant portions of the reactor trip and engineered safety features actuation signals are modeled in the PRA model. The PRA model is used directly as one input to the Braidwood Station and Byron Station Configuration Risk Management Programs (CRMPs). The Braidwood Station and Byron Station CRMP is a subset of the work management process, and ensures that configuration risk (i.e., probabilistic and/or deterministic) is assessed and managed, prior to initiating any maintenance activity, consistent with the requirements of 10 CFR 50.65. The CRMP also ensures that risk is reassessed if an emergent condition results in a plant configuration that has not been previously assessed. The CRMP is described in Appendix T of the Braidwood Station and Byron Station Technical Requirements Manual (TRM). Probabilistic risk assessments of online configurations are performed using the latest approved revision of the level 1 PRA model. Deterministic defense-in-depth evaluations of key safety functions are performed for online and shutdown configurations using a safety function assessment module. Deterministic evaluations of plant configurations that result in a change to initiating event frequency and/or a decrease in mitigation capability are performed using a plant transient assessment module. The procedures that implement the CRMP at Braidwood Station and Byron Station ensure that representative signals are used, to the appropriate depth, to perform Tier 3 evaluations of the RTS and ESFAS configuration impact. These procedures require that signals that are not explicitly modeled will either be added to the model, or addressed by surrogates that are modeled. This action will result in appropriate modeling in the PRA to complete Tier 3 evaluations. It is expected that in most cases, with one RTS or ESFAS channel unavailable, ATTACHMENT 1 Response to Request for Additional Information License Amendment Request to Implement WCAP-14333 and WCAP-15376 risk will be Green, with the exception of those variables modeled in the Reactivity Control safety function deterministic assessment. NRC R equest 3.f Plant design or operational modifications not reflected in the WCAP-14333/WCAP15376 PRA used in this application that are related to or could impact this application. Justify the acceptability of not including these modifications in the PRA as part of this application. Response The Braidwood Station and Byron Station PRA model that was used to validate the applicability of the WCAP-14333 and WCAP-15376 analyses to the RTS and ESFAS instrumentation for each respective station was approved in September 2005, and reflected the as-built and as-operated plants, as of that time. The EGC risk management process ensures that the applicable PRA model remains an accurate reflection of the as-built and as-operated plants. The PRA quality assurance program that is described in the response to NRC Request 3.d above, is utilized for verification and validation of the PRA model. Subsequent to submittal of this license amendment request, EGC approved a revised Braidwood Station and Byron Station PRA model. This new PRA model incorporates recent design and operational changes. The most significant of these changes was the partial completion of a modification at Braidwood Station that replaced several 120V AC manual transfer switches with automatic transfer switches. This partial implementation of a modification at Braidwood Station improved the reliability of the RTS and ESFAS instrumentation systems. The completion of the modification at Braidwood Station, and the implementation of the modification at Byron Station will also enhance and improve this reliability. In addition, for Byron Station only, the service air compressors were upgraded with air-cooled compressors. This eliminated a support system dependency. The incorporation of this design change had no impact on the reliability of the RTS and ESFAS instrumentation systems. EGC has reconfirmed the applicability of WCAP-14333 and WCAP-15376 to Braidwood Station and Byron Station, using data from the recently completed PRA model revision. This reconfirmation indicated that the conformance of Braidwood Station and Byron Station with the WCAP acceptance criteria, using the PRA model at that time, bounds the results using the most recent model. The most recent model reflects the current (i.e., October 2007) design and operational configuration for each Station. NRC Request 4 Consistent with RG 1.174, the cumulative risk of the present TS change in light of past applications (or additional applications planned or under review) should be understood. Cumulative risks were addressed for Byron/Braidwood Stations Units 1 and 2 with respect to the implementation of WCAP-10271, WCAP-14333, and WCAP-15376 only. Provide an evaluation Page 7 ATTACHMENT 1 Response to Request for Additional Information License Amendment Request to Implement WCAP-14333 and WCAP-15376 of cumulative risk impact of previous and additional TS applications under development or review per RG 1.174, Section 3.3.2, as applicable to the implementation of WCAP-14333 and WCAP-15376. Response The Braidwood Station and Byron Station risk management processes require that the risk impact of each permanent TS application be evaluated after each periodic PRA update. However, the cumulative risk impact from previous TS applications is not a straightforward calculation, since the risk impacts of the various applications do not necessarily occur simultaneously. In addition, the analyses supporting the various applications were performed with PRA updates reflecting different sets of plant modifications and operating experiences. Table 2 provides the calculated delta CDF and delta LERF values for each risk-informed TS application from the last complete periodic update. For the reasons noted above, summation of these values is not representative of cumulative risk. However, Table 2 provides the sum of the impacts over all the changes, for comparative purposes. NRC Reque st 5 Does the configuration risk management program at Byron/Braidwood provide modeling of the reactor trip and ESFAS systems and components addressed by WCAP-15376 and WCAP-14333 when performing Tier 3 evaluations?
Discuss how signals or components not specifically modeled will be addressed. Response The response to NRC Request 3.e above describes the Braidwood Station and Byron Station CRMP, with respect to modeling the appropriate reactor trip and ESFAS systems and components for Tier 3 evaluations. NRC R ectuest 6 Discuss the omission of RCS pressure relief systems from the regulatory commitments listed in Attachment 5, Page 1 of the submittal, see WCAP-15376 (Tier 2). Response Upon further review, EGC will revise the two applicable commitments from the original license amendment request (i.e., in a letter from D. M. Benyak (Exelon Generation Company, LLC) to U. S. NRC, dated January 7, 2007), to include the RCS pressure relief system. The revision of these two commitments is described in Attachment
: 2. [Regulatory Commitments]
ATTACHMENT 1 Response to Request for Additional Information License Amendment Request to Implement WCAP-14333 and WCAP-15376 Table 1 "A" and "B" F&Os Cert. Element F&O Element Description Level F&O Description Status Resolution Impact DA-10 Common cause B [Braidwood F&O DA-5] Resolved The asymmetric modeling None groups to which Reviewers do not agree in Rev. 6 assumption was removed and the the common with the justification for of the model logic was changed to consider cause failure asymmetric modeling of Model symmetric common cause among all probability applies the emergency diesel four EDGs (two for each plant). have been derived generators (EDGs). based on sound judgment and are [Byron F&O DA-05B] documented. Reviewers do not agree with the rationale provided as a resolution to the Braidwood finding. HR-2 Human reliability B [Braidwood F&O HR-2] Resolved Subsequent updates to the HRA None analysis (HRA) is The modified cause in Rev. 6 methodology and documenation consistent with based decision tree of the eliminated the conflicts between industry practice. (CBDT) method used Model CBDT and human cognitive reliability for the HRA apparently modeling. The CBDT methods from deviates from standard Quad Cities were used, which industry practice. subsequently received favorable peer review certification.
ATTACHMENT 1 Response to Request for Additional Information License Amendment Request to Implement WCAP-14333 and WCAP-15376 Table 1 "All and "B" F&Os Cert. Element F&O F&O Description Status Resolution Impact Element Description Level HR-10, Assessment of B [Braidwood F&O HR-4] Resolved Operator interviews were conducted None 22 plant procedures The steam generator in Rev. 6 as part of the Rev. 6 changes to the and plant specific tube rupture (SGTR) of the HEPs. Entry inot FR-H.1 was operating event tree structure and Model confirmed and timing of actions was experience are HRA do not reflect the verified. explicitly included circumstances around in the identification entering FR-H.1. The and quantification success criteria, event process for the tree structure, and HRA human should be modified to interactions (His). reflect accident The models and sequence mitigation analysis are dictated by the consistent with the emergency operating operating procedures (EOPs). procedures and training.
ATTACHMENT 1 Response to Request for Additional Information License Amendment Request to Implement WCAP-14333 and WCAP-15376 Table 1 "A" and "B" F&Os Cert. Element Element Description F&O Level F&O Description Status Resolution Impact HR-11 The symptoms B [Byron F&O HR-07B] Resolved The manual reactor trip operator None available during The HRA for manual in Rev. 6 action is conditioned on the basis of the postulated Reactor Protection of the whether the failure was due to accident sequence System (RPS) and Model actuation logic failure versus signal are evaluated and Engineered Safety failure. However, the operator action input into the HRA Feature Actuation to initiate ESFAS manually was not process. System (ESFAS) modified. actions do not explicitly The Braidwood and Byron Station account for degradation procedures require manual actuation of plant monitoring (i.e., of Safety Injection (SI) regardless of different HEP whether or not signals are present. depending on whether In addition, the procedures require or not failure of auto- manual actuation of individual actuation of equipment components if auto actuation fails, was due to equipment regardless of the status of signal failure or signal failure). inputs. The HEPs in the model reflect both the Sl actuation and the individual component actuation as well as the conditional probability between the two.
ATTACHMENT 1 Response to Request for Additional Information License Amendment Request to Implement WCAP-14333 and WCAP-15376 Table 1 "All and "B" F&Os Cert. Element F&O F&O Description Status Resolution Impact Element Description Level HR-11, The symptoms B [Byron F&O HR-02B] Resolved As part of the Rev. 6 model None 14, 18, available during Although a significant in Rev. 6 revisions, all significant operator 20 the postulated effort has been of the actions were reviewed with accident sequence undertaken to gain Model Operations and Training personnel are evaluated and operator input into to assess the timong for execution input into the HRA evaluating HEPs, the and cognition, as well as the process. reviewers found the procedural guidance for the actions. Operator actions effort to date had not have been fully addressed reviewed by the observations from the operating staff and Braidwood peer their impact is certification (F&O included in the HR-5). HRA evaluation Specifically cited is the The performance lack of operator shaping factor for interviews to validate time available for input assumptions, an action and the timing and logic of time required to certain key operator take an action are actions. developed on a The reviewers also plant specific acknowledged that the basis. sensitivity analyses The time required performed showed that to complete the the overall results are actions is based not overly sensitive to on observation or the operator action operations staff modeling. input.
ATTACHMENT 1 Response to Request for Additional Information License Amendment Request to Implement WCAP-14333 and WCAP-15376 Table 1 "All and "B" F&Os Cert. Element Element Description F&O Level F&O Description Status Resolution Impact QU-27, A search is A [Braidwood F&O QU-7] Open Formal guidance for addressing this The circumstance described by 28 performed for Only parametric issue is still being developed by NRC this WOG PEER review finding unique or unusual uncertainty analyses and EPRI. is not expected to adversely sources of have been performed. affect the conclusions made for uncertainty not the proposed RPS and ESFAS present in the extended surveillance test typical or generic interval (STI) and completion plant analysis. times (CTs) because of the If there are conservatism used in the unusual sources Braidwood Station and Byron of uncertainty, Station PRA model. In addition, special sensitivity this PRA model is not used to evaluations or calculate changes in CDF, quantitative LERF, or ICCDP to justify the uncertainty proposed changes. assessments are performed to support the base conclusion and future applications.
ATTACHMENT 1 Response to Request for Additional Information License Amendment Request to Implement WCAP-14333 and WCAP-15376 Table 2 Estimated Cumulative Risk Item no. LAR Calculated ACDF Calculated A LAR Application (internal events) LERF Date internal events 1 9/1/2000 Small decreases in risk for both Bryon Emergency Diesel Generator and Braidwood with the compensatory CT Extension measure credited. 2 11/19/2003 7E-10 Byron 4E-09 Byron Inverter CT Extension 9E-10 Braidwood 7E-09 Braidwood 3 2/14/2006 9E-08 Byron 1 E-09 Byron Risk Informed ISI Relief 1 E-07 Braidwood 1 E-09 Braidwood Request 4 4/7/2007 N/A 2E-07 Byron ILRT STI extension (i.e., from 2E-07 Braidwood once in 10 years to once in 15 ears 5 1/7/2007 < 1 E-06 Byron < 1 E-07 Byron RTS/ESFAS Instrumentation CT < 1 E-06 Braidwood
< 1 E-07 Braidwood and STI Extension 6 Planned < 1 E-08 Byron < 1 E-08 Byron RCP Flywheel Submittal: < 1 E-08 Braidwood
< 1 E-08 Braidwood Earl 2008 Approximate Sum 1 E-06 Byron 3E-07 Byron Cumulative Risk 1 E-06 Braidwood 3E-07 Braidwood ATTACHMENT 2 Summary of Regulatory Commitments License Amendment Request to Implement WCAP-14333 and WCAP-15376 The following table identifies those actions committed to as part of the EGC response an NRC Request for Additional Information that was provided in a letter from R. F. Kuntz (U. S. NRC) to C. M. Crane (Exelon Generation Company, LLC), dated August 17, 2007. The first commitment applies to Byron Station and is a new commitment. The two remaining commitments apply to both Braidwood Station and Byron Station and are revisions of the same commitments that were made in the original license amendment request. Any other statements in this submittal are provided for information purposes and are not regulatory commitments. Commitment Committed Due Date/Event Date One-Time Programmatic Action Yes/No Yes/No EGC will implement enhanced internal flood Prior to No Yes response procedures at Byron Station. implementation of the revised TS. Revision of Or*wjW.: Agtulatory Prior to No Yes Commitment implementation of the revised EGC will implement administrative controls TS. to ensure that activities that degrade the availability of the RCS pressure relief system, auxiliary feedwater system (AFW), AMSAC, or turbine trip should not be scheduled when a logic train or an RTB train is inoperable for maintenance. Revision of Orilgow ulatorv Prior to No Yes Commitment implementation of the revised EGC will implement administrative controls TS. to ensure that activities that result in the inoperability of electrical systems (e.g., AC and DC power) and cooling systems (e.g., essential service water and component cooling water) that support the RCS pressure relief system, AFW system, AMSAC, turbine trip, one complete train of ECCS, and the available reactor trip and ESFAS actuation functions should not be scheduled when a logic train or an RTB train is inoperable for maintenance. That is, one complete train of a function that supports a complete train of a function noted above must be available}}

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