RS-16-197, License Amendment Request for a One-Time Extension of the Essential Service Water (SX) Train Completion Time to Support 2A SX Pump Repair

License Amendment Request for a One-Time Extension of the Essential Service Water (SX) Train Completion Time to Support 2A SX Pump Repair

ML16274A474
Person / Time
Site:	Braidwood
Issue date:	09/30/2016
From:	Gullott D Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RS-16-197
Download: ML16274A474 (129)

Text

Exelon G Warr Warr Winfield Winfield Road IL 60555 www.exeloncorp.com 10 CFR 50.90 RS-16-197 September 30, 2016 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Braidwood Station, Unit 2 Renewed Facility Operating License No. NPF-77 NRC Docket No. STN 50-457

Subject:

License Amendment Request for a One-Time Extension of the Essential Service Water (SX) Train Completion Time to Support 2A SX Pump Repair

References:

1. NRC Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," Revision 2, dated May 2011

2. NRC Regulatory Guide 1.177, "An Approach for Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications," Revision 1, dated May 2011

3. NRC Regulatory Guide 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision 2, dated March 2009 In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Exelon Generation Company, LLC (EGC) is requesting a license amendment to Renewed Facility Operating License No. NPF-77 for Braidwood Station Unit 2.

This amendment request proposes a new Required Action A.2 that increases the Completion Time (CT) currently specified in Required Action A.1, "Restore unit-specific SX train to OPERABLE status," associated with Technical Specifications (TS) Section 3.7.8, "Essential Service Water (SX) System," from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br />. This proposed change will only be used one time during a planned 2A SX pump repair scheduled to start the week of November 27, 2016.

The current TS Limiting Condition for Operation (LCO) 3.7.8 requires that two unit-specific SX trains (i.e., the "A" train and "B" train) must be operable in Modes 1, 2, 3 and 4 and that one opposite-unit SX train must be operable for unit-specific support. Condition A allows one unit-specific SX train to be inoperable with a CT of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. An extension of the CT to 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br /> (i.e., an increase of 128 hours0.00148 days <br />0.0356 hours <br />2.116402e-4 weeks <br />4.8704e-5 months <br />) is needed to preemptively repair and replace components of the

September 30, 2016 Nuclear Regulatory Commission Page 2 2A SX pump due to observed performance degradation. This evolution is not a typical maintenance activity that can be performed within the existing 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> CT window and current planning estimates and past maintenance history have shown that the planned pump repairs cannot be performed within the current CT.

Repair of the 2A SX pump will be conducted during a planned 2A SX system outage; however, due to the system configuration necessary to perform the repair of the degraded SX pump, the 2A SX train will be declared inoperable resulting in Braidwood Station, Unit 2 entering Condition A with a 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> CT. Consequently, not being able to complete the 2A SX pump repair in the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> CT would require a Unit 2 shutdown in accordance with TS 3.7.8, Required Action C.1 and C.2.

The proposed change to the CT has been evaluated using the risk-informed processes described in Regulatory Guide (RG) 1.174 (Reference 1) and RG 1.177 (Reference 2). The risk associated with the proposed changes was determined to be acceptable.

The attached request is subdivided as shown below.

Attachment 1 provides an evaluation of the proposed changes.

Attachment 2 includes the marked-up TS pages with the proposed changes indicated.

Attachment 3 includes the marked-up TS Bases pages with the proposed changes indicated.

The TS Bases pages are provided for information only and do not require NRC approval.

Attachment 4 provides a summary of the regulatory commitments contained in this letter.

Attachment 5 provides the supporting risk-informed evaluation of the requested change including an evaluation of the technical adequacy of the PRA in accordance with RG 1.200 (Reference 3).

EGC requests approval of the proposed license amendment by November 17, 2016, to support the planned repair of the 2A SX pump during the scheduled work window. Attempting to replace the 2A SX pump without prior NRC approval to extend TS 3.7.8, Condition A CT to 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br /> would be inappropriate since, as noted above, the pump repair cannot be completed in the current 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> CT window. The shutdown of Unit 2 would result in an unnecessary operational transient since the 2B SX train and an opposite unit SX train (i.e., 1A or 1 B train required by TS LCO 3.7.8.b) would be operable.

Considering recent changes in pump performance and the need to address the degrading performance in a timely manner, EGC could not have reasonably avoided this request for an expedited NRC review and approval. The proposed change results in an overall integrated safety improvement by eliminating a degraded material condition and avoiding an unnecessary plant shutdown and startup. Once approved, the amendment will be implemented prior to the November 27, 2016 SX system work window.

The proposed change has been reviewed by the Braidwood Station Plant Operations Review Committee in accordance with the requirements of the EGC Quality Assurance Program.

September 30, 2016 Nuclear Regulatory Commission Page 3 In accordance with 10 CFR 50.91, paragraph (b), EGC is notifying the State of Illinois of this application for license amendment by transmitting a copy of this letter and its attachments to the designated State Official.

Regulatory Commitments are contained in Attachment 4 to this letter. Should you have any questions concerning this letter, please contact Joseph A. Bauer at (630) 657-2804.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 30th day of September 2016.

Respectfully, David M. Gullott Manager Licensing Exelon Generation Company, LLC Attachments: 1. Evaluation of Proposed Changes

2. Proposed Technical Specification Pages for Braidwood Station

3. Proposed Technical Specification Bases Pages for Braidwood Station (For Information Only)

4. Summary of Regulatory Commitments

5. BW-LAR-008, "Risk Assessment Input for the Braidwood One-Time Technical Specification Change for the Essential Service Water Pump 2A Completion Time from 72 to 200 Hours," Revision 0 cc: Regional Administrator NRC Region III NRC Senior Resident Inspector Braidwood Station Illinois Emergency Management Agency Division of Nuclear Safety

ATTACHMENT 1 Evaluation of Proposed Changes 1.0

SUMMARY

DESCRIPTION

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria 4.2 Precedence 4.3 No Significant Hazards Consideration 4.4 Conclusions

5.0 ENVIRONMENTAL CONSIDERATION

Page 1 of 24

ATTACHMENT 1 Evaluation of Proposed Changes 1.0

SUMMARY

DESCRIPTION In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Exelon Generation Company, LLC (EGC) is requesting an amendment to Facility Operating License No. NPF-77 for Braidwood Station, Unit 2. The amendment request proposes a new Required Action A.2 that increases the Completion Time (CT) currently specified in Required Action A.1, "Restore unit-specific SX train to OPERABE status,"

associated with TS Section 3.7.8, "Essential Service Water (SX) System," from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br />. This proposed change will only be used one time during a Braidwood Station planned 2A SX pump repair window scheduled to start the week of November 27, 2016.

2.0 DETAILED DESCRIPTION The current TS Limiting Condition for Operation (LCO) 3.7.8 requires that two unit-specific SX trains (i.e., the "A" train and "B" train) must be operable in Modes 1, 2, 3 and 4 and that one opposite-unit SX train must be operable for unit-specific support. Condition A allows one unit-specific SX train to be inoperable with a CT of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. An extension of the CT to 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br /> (i.e., an increase of 128 hours0.00148 days <br />0.0356 hours <br />2.116402e-4 weeks <br />4.8704e-5 months <br />) is needed to repair and replace components of the 2A SX pump due to observed performance degradation. This CT extension for one train of SX to be inoperable from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br /> is needed to preemptively replace the 2A SX pump rotating elements, including the pump shaft and impeller (hereafter, "rotating element") and perform potential weld repairs to the cutwater region of the pump casing. The extension of the CT to 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br /> is supported by the risk assessment summarized below in Section 3.0 and detailed in Attachment 5. This evolution is not a typical maintenance activity that can be performed within the existing 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> CT and current planning estimates and maintenance history have shown that these pump components cannot be replaced within the current CT.

Repair of the 2A SX pump will be conducted during a planned SX System outage; however, due to the system configuration necessary to perform the repair of the degraded 2A SX pump, the 2A SX train will be declared inoperable resulting in Braidwood Station, Unit 2 entering TS 3.7.8 Condition A with a 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> CT. Consequently, not being able to complete the 2A SX pump repair in the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> CT would require a Unit 2 shutdown, in accordance with TS 3.7.81 Required Action C.1 and C.2. Therefore, the proposed change increases, on a one-time basis, the CT to restore an inoperable unit-specific SX train from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br />.

In order to repair the 2A SX pump, the rotating element will be replaced along with potential weld repairs to the cutwater region of the pump casing. As such, a note is proposed to be added to the existing Required Action A.1 of TS 3.7.8 stating that Required Action A.1 does not apply to Unit 2 while the 2A SX pump repair outage is ongoing. A new Required Action A.2 with associated note is proposed that specifies use of the 200 hour0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br /> CT is only applicable to Unit 2 during the 2A SX pump repair during the November 27, 2016 work window. The proposed changes are shown on the marked-up Braidwood Station TS pages included in Attachment 2.

In addition, an informational copy of the associated TS Bases pages with marked-up changes is provided in Attachment 3.

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ATTACHMENT 1 Evaluation of Proposed Changes The scheduled pump repair is anticipated to take 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br /> which includes contingency time to account for potential unplanned delays, such as difficulties with system isolation (based on previous experience); and other emergent issues. If the work exceeds the proposed 200 hour0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br /> CT, the unit will be shutdown in accordance with TS 3.7.8 Condition C. A unit shutdown in accordance with TS 3.7.8 Condition C, due to the 2A SX train being inoperable for 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br />, would be no different in terms of plant response and operator actions associated with a unit shutdown in accordance with the current TS CT of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, due to one unit-specific SX train being out of service. Additionally, in the event that the 213 SX pump becomes inoperable during repair of the 2A SX pump, Unit 2 will be shutdown in accordance with the TS.

3.0 TECHNICAL EVALUATION

System Description

The SX System provides a heat sink for the removal of process and operating heat from safety related components during a Design Basis Accident (DBA) or transient. During normal operation, and a normal shutdown, the SX System also provides this function for various safety related and non-safety related components.

The unit-specific SX System consists of two separate, electrically independent, 100% capacity, safety related, cooling water trains. Each train consists of a 100% capacity pump, piping, valving, and instrumentation. Reference Figure 1 for a simplified drawing of the SX system.

Note that this figure is provided for informational purposes and is not controlled as a design document.

The pumps and valves are remote and manually aligned, except in the unlikely event of a Loss of Coolant Accident (LOCA). The pumps are automatically started upon receipt of a safety injection signal or an undervoltage on the Engineered Safety Features bus, and all essential valves are aligned (automatically or manually) to their post-accident positions (Diesel Generator (DG) SX cooling water supply valves are opened once the DG has reached sufficient speed).

The SX System is the safety-related backup water supply to the Auxiliary Feedwater System and backup supply to the Fire Protection System.

The SX System includes provisions to crosstie the trains (unit-specific crosstie refer to Figure 1), as well as provisions to crosstie the units (opposite-unit crosstie refer to Figure 1).

The opposite-unit crosstie valves (1 SX005 and 2SX005) must both be open to accomplish the opposite-unit crosstie. The opposite-unit crosstie flowpath along with an opposite-unit SX pump are capable of providing backup cooling in the event of a loss of all SX on one unit. The system is normally aligned with the unit-specific crosstie valves open and the opposite-unit crosstie valves closed. Additional information about the design and operation of the SX System, along with a list of the components served, is presented in the UFSAR, Section 9.2.1. Some of the functions served by the SX System are the removal of decay heat from the reactor via the Component Cooling Water (CC) System, the removal of heat from containment via the Reactor Containment Fan Coolers (RCFC), and cooling of the DGs.

The entire SX System is designated Safety Category 1, Quality Group C, including the supply lines, pumps, and return lines with the exception of the discharge extension lines that extend the SX return above the Braidwood Ultimate Heat Sink surface to prevent flooding. The SX supply Page 3 of 24

ATTACHMENT 1 Evaluation of Proposed Changes and discharge lines are either below or incorporated in the turbine building base mat. They are not inside the turbine building and, the lines are adequately protected from any occurrence within the turbine building. The turbine building base mat is designed and constructed such that it will not suffer gross failure or collapse during either a safe shutdown earthquake or a design basis tornado.

The SX pumps are located in the auxiliary building basement with the 1A and 2A pumps in one compartment and the 1 B and 2B pumps located in a separate adjacent compartment. Entrance to each compartment is via a watertight door and each compartment is constructed to prevent flooding due to a pipe failure from affecting the equipment in the opposite train SX compartment.

Safety Analysis The design basis of the SX System is for one train, in conjunction with the CC System and a 100% capacity containment cooling system, to remove core decay heat following a design basis LOCA as discussed in the UFSAR, Section 6.2. This prevents the containment sump fluid from increasing in temperature during the recirculation phase following a LOCA and provides for a gradual reduction in the temperature of this fluid as it is supplied to the Reactor Coolant System by the Emergency Core Cooling System pumps. The SX System is designed to perform its function with a single failure of any active component, assuming the loss of offsite power.

The SX System, in conjunction with the CC System, also cools the unit from Residual Heat Removal (RHR) entry conditions, as discussed in the UFSAR, Section 5.4.7, to Mode 5 during normal and post-accident operations. The time required for this evolution is a function of the number of CC and RHR System trains that are operating. One SX train is sufficient to remove decay heat during subsequent operations in Modes 5 and 6.

Generic Letter 91-13 included risk-based recommendations for enhancing the availability of SX Systems, in the case of a loss of all SX to a particular unit. Crediting the opposite-unit SX System with an opposite-unit pump and the opposite-unit crosstie valves, was a part of the response to this Generic Letter.

The unit-specific SX System satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii). The opposite-unit SX System satisfies Criterion 4 of 10 CFR 50.36(c)(2)(ii).

Need for Amendment As noted above, the current TS LCO 3.7.8 requires that two unit-specific SX trains (i.e., the "K and "B" trains) be operable in Modes 1, 2, 3, and 4. Condition A allows one unit-specific SX train to be inoperable with a CT of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. An extension of the CT to 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br /> is needed to repair the 2A SX pump.

An adverse trend in pump differential pressure (DP) has been observed during Inservice Testing (IST) Program testing for the 2A SX pump. The declining trend in DP performance is most likely attributable to wear and erosion of the pump rotating element and casing based on prior repairs/replacements of the SX pumps. A review of the equipment maintenance history shows that the SX pumps have been repaired/replaced in a 15-20 year timeframe. The current 2A SX pump has been in service for approximately 14.5 years. The rotating element replacement and Page 4 of 24

ATTACHMENT 1 Evaluation of Proposed Changes casing repair were originally scheduled to be performed during the upcoming Unit 2 refueling outage in spring 2017, at a pump age of 15 years.

During a recent IST comprehensive pump test performed on September 7, 2016, there was evidence of further degradation in the pump DP performance. Due to questions regarding the validity of the test methodology, the test was re-performed on September 22, 2016. This test confirmed the pump performance was in the DP Alert Range requiring an increase in the comprehensive test frequency in accordance with the IST Program. A review of other pump performance parameters (i.e., pump and motor vibrations, motor amperage) have not identified any other degraded trends.

Braidwood Station has subsequently determined that a preemptive repair of the 2A SX pump is warranted to address the declining performance. Repair of an SX pump is not a typical maintenance activity and planning estimates and maintenance history have shown that completion of this evolution cannot be assured within the existing 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> CT.

SX Pump Repair/Replacement History All Braidwood Station SX pumps have been replaced once thus far over the operating life of the plant. The SX pumps have exhibited an operating life of approximately 15-20 years. All pumps are tested quarterly, but have had no major maintenance since each pump's replacement. See below table for a summary of SX pump status.

The 1A SX pump was replaced in spring 2006 (currently approximately 10.5 years old), well within the historical operating life of an SX pump. The 1A SX pump is currently operating with sufficient margin; an average margin of 3.4 psid within the last year.

The 1 B SX pump was replaced in spring 2003 (currently approximately 13.5 years old). This pump has been operating with minimal margin recently. Consequently, the rotating element will be replaced during Refueling Outage Al R19 (in September/October 2016) to increase the margin. Note that the previous 1 B SX pump was also repaired (i.e., replaced the rotating element and performed a weld repair of the cutwater region of the casing) in spring 2000.

The 2A SX pump was replaced in spring 2002 (currently approximately 14.5 years old). The pump observed DP performance has shown a declining trend and the pump entered the DP Alert Range on September 22, 2016, requiring an increase in the comprehensive test frequency in accordance with the IST Program.

The 2B SX pump was replaced in spring 2008 (currently approximately 8.5 years old), well within the historical operating life of an SX pump. The 2B SX pump is currently operating with sufficient margin; an average margin of 3.5 psid within the last year.

The 1A and 2B SX pumps show stable trends during the quarterly tests with sufficient margin.

Subsequently, it is concluded that the 1A and 2B SX pumps are in acceptable operating condition with sufficient margin on each pump; and as noted, the 1 B SX pump is being repaired in the Fall 2016 Refueling Outage. Repair of the 2A SX pump is the subject of this request and is further discussed below.

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ATTACHMENT 1 Evaluation of Proposed Changes SX Pump Status Summary Pump Replacement Average DP Margin Approximate Pump History Over Previous Year Current Pump Age (years) 1A SX Pump Spring 2006 3.4 psid 10.5

• 1 B SX Pump Spring 2003 1.3 psid 13.5 2A SX Pump Spring 2002 1.3 psid 14.5 213 SX Pump Spring 2008 j 3.5 psid 8.5 "Note: As mentioned above, there is a planned rotating element replacement for the 1 B SX Pump in Refueling Outage Al R19 (October 2016) 2A SX Pump Repair Overview The 2A SX pump was replaced in spring 2002. The removed pump's rotating element showed signs of general corrosion, including heavy damage to the suction vanes. The pump casing, including the cutwater region, also required weld repairs. The most likely cause of the wear to the removed 2A SX pump was due to erosion/corrosion intensified by the material of the pump internals (i.e., carbon steel) and the medium being pumped (i.e., raw water). Therefore, given the age of the current in-service 2A SX pump, it is reasonable to expect that similar wear has occurred. Repair of the 2A SX pump by replacing the pump rotating element and performing potential weld repairs on the pump casing (i.e., cutwater region) will improve design margin. Based on a similar repair made to the 1 B SX pump (completed in spring 2000),

Braidwood Station expects to regain approximately 2 psid of margin. Given this increase in margin, the 2A SX pump will continue to meet its design requirements. The 2A SX pump replacement will be scheduled for a future refueling outage based on actual pump performance.

Disassembly and repair of the pump by replacing the rotating element and potential weld repairs on the cutwater region of the pump casing involves a number of major steps requiring an estimated duration of 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br /> to restore the 2A SX pump to service. To support repair of the 2A SX pump, the flood seal for the "A" SX pump room (containing the 1A and 2A SX pumps),

will be temporarily removed. Station barrier impairment procedures will be implemented to address this temporary configuration.

The required major steps to repair the 2A SX pump are listed below. The detailed schedule identifies activities that will be performed in parallel.

Scheduled maintenance activities are:

• Place out of service clearance orders, isolate and drain system

• Implement station procedures to address temporary removal of the "A" SX pump room flood seal

• Disassemble pump, lube oil piping, coupling, lube oil pump and bearings

• Unbolt and lift pump upper case and remove rotating element

• Potential weld repair of casing

• Set new rotating element in place, reassemble pump casing

• Align pump to motor Page 6 of 24

ATTACHMENT 1 Evaluation of Proposed Changes

• Couple pump

• Complete alignment

• Clear out of service and perform post maintenance tests The work will be performed on a 24-hour/day schedule and centrally coordinated by the staffed Outage Control Center (OCC). All parts will be on site and available prior to the start of the 2A SX pump work window.

Risk Evaluation The proposed TS changes have been evaluated using the risk-informed processes described in Regulatory Guide (RG) 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," dated May 2011 (Reference 5) and RG 1.177, "An Approach for Plant-Specific, Risk-informed Decisionmaking:

Technical Specifications," dated May 2011 (Reference 6). The risk assessment, detailed in , supports a one-time CT extension for the 2A SX pump from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br />.

Reference 5 describes a risk-informed approach for assessing the nature and impact of proposed permanent licensing-basis changes by considering engineering issues and applying risk insights. This RG also provides risk acceptance guidelines for evaluating the results of such evaluations.

Reference 6 describes a risk-informed approach specifically for assessing proposed permanent TS changes in CTs. This RG also provides risk acceptance guidelines for evaluating the results of such evaluations.

In implementing risk-informed decisionmaking under References 5 and 6, TS changes are expected to meet a set of five key principles. These principles include consideration of both traditional engineering factors (e.g., defense in depth and safety margins) and risk information. provides the risk-informed evaluation of the proposed change in the SX CT that considers each one of these key principles.

• The proposed change meets the current regulations unless it is explicitly related to a requested exemption.

• The proposed change is consistent with the defense-in-depth philosophy.

• The proposed change maintains sufficient safety margins.

• When proposed changes result in an increase in core damage frequency (CDF) or risk, the increases should be small and consistent with the intent of the Commission's Safety Goal Policy Statement.

• The impact of the proposed change should be monitored using performance measurement strategies.

The numeric results are summarized in the following table. Since this is a one-time change EGC did not report Delta CDF and Delta Large Early Release Frequency (LERF) values; that would normally be part of a permanent change. The results of the risk evaluation are compared Page 7 of 24

ATTACHMENT I TACHMENT 1 Evaluation of Proposed Changes in Table I with the risk acceptance guidelines described in Attachment 5. The values for the Incremental Conditional Core Damage Probability (ICCDP) and the Incremental Conditional Large Early Release Probability (ICLERP) demonstrate that the proposed one-time unit-specific SX train CT extension has a small quantitative impact on plant risk.

Table I COMPARISON OF INDIVIDUAL HAZARD GROUP RESULTS TO ACCEPTANCE GUIDELINES Figure of Merit Value Acceptance Guideline Below Acceptance Guideline Internal Events and Internal Floods ICCDP 7.2E-08 <1.0E-06, or <1.0E-5(1) Yes ICLERP 1.1 E-09 <1.0E-07, or <1.0E-6(2) Yes Internal Fires ICCDP 2.6E-06 <1.0E-06, or <1.0E-5(1) Yes(i)

ICLERP 3.9E-08 <1.0E-07, or <1.0E-6(2) Yes Other Hazard Groups ICCDP Negligible <1.0E-06, or <1.0E-5(1) Yes ICLERP Negligible <1.0E-07, or <1.0E-6(2) Yes Total Values ICCDP 2.7E-06 <1.0E-06, or <1.0E-5(1) Yes(i)

ICLERP 4.0E-08 <1.0E-07, or <1.0E-6(2) Yes Per RG 1.177 a value between 1 E-06 and 1 E-05 may be deemed acceptable with effective compensatory measures implemented to reduce the sources of increased risk for a one-time change.

(2)

Per RG 1.177 a value between 1 E-07 and 1 E-06 may be deemed acceptable with effective compensatory measures implemented to reduce the sources of increased risk for a one-time change.

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ATTACHMENT 1 Evaluation of Proposed Changes Table 2 provides a summary of the approach and results of the evaluation of each of the potential risk contributors. These analyses demonstrate that the risk impact of the proposed one-time CT extension for the 2A SX pump repair is small and below the acceptance guidelines.

Table 2

SUMMARY

OF RISK INSIGHTS FOR SX 2A CT EXTENSION RISK CONTRIBUTOR APPROACH INSIGHTS Internal Events Quantify ICCDP & ICLERP for 0 Base risk well within planned configuration acceptance guidelines

• ICCDP < 1E-6

• Compensatory measures

• ICLERP < 1 E-7 further reduce risk If exceeded compare to acceptance guidelines with risk management actions implemented to reduce sources of risk

• ICCDP < 1 E-5

• ICLERP < 1 E-6 Internal Fire Qualitatively and quantitatively ICCDP and ICLERP within evaluated: acceptance guidelines with

• Identify fire scenarios risk management actions to impacted by reduce risk sources.

configuration

• Internal events

• Estimate fire risk impacts compensatory measures due to configuration and apply to fire scenarios quantify ICCDP and 0 Additional Fire-related ICLERP compensatory measures

• Identify compensatory identified measures Seismic Qualitatively evaluated.

• Seismic risk impacts negligible High Winds Qualitatively evaluated.

• High winds risk impacts negligible

• High winds risk reduced with compensatory measures for internal events and fire Page 9 of 24

ATTACHMENT I Table 2

SUMMARY

OF RISK INSIGHTS FOR SX 2A CT EXTENSIOK RISK CONTRIBUTOR APPROACH INSIGHTS Other External Hazards Qualitatively evaluate each

• Other External Event risks hazard based on the BW IPEEE were found to be negligible and a re-examination for the contributors specific configuration with SX 2A inoperable.

Overall At-Power Risks Quantify ICCDP & ICLERP for Quantitative guidelines for planned configuration with normal work controls normal work controls challenged, but acceptable

• ICCDP < 1E-6 with risk management actions implemented.

ICLERP < 1 E-7 If exceeded compare to acceptance guidelines with risk management actions implemented to reduce sources of risk

• ICCDP < 1E-5

• ICLERP < 1E-6 Regulatory Guide 1.177 specifies an approach and acceptance guidelines for the evaluation of plant licensing basis changes. RG 1.177 identifies a three-tiered approach for the evaluation of the risk associated with a proposed TS change as identified below:

Tier 1 is an evaluation of the plant-specific risk associated with the proposed TS change, as shown by the change in core damage frequency (CDF) and incremental conditional core damage probability (ICCDP). Where applicable, containment performance should be evaluated on the basis of an analysis of large early release frequency (LERF) and incremental conditional large early release probability (ICLERP). The acceptance guidelines given in RG 1.177 for determining an acceptable permanent TS change is that the ICCDP and the ICLERP associated with the change should be less than 1 E-06 and 1 E-07, respectively. RG 1.177 also addresses risk metric requirements for one-time TS changes.

• Tier 2 identifies and evaluates, with respect to defense-in-depth, any potential risk-significant plant equipment outage configurations associated with the proposed change. Reasonable assurance should be provided that risk-significant plant equipment outage configurations will not occur when equipment associated with the proposed TS change is out-of-service.

• Tier 3 provides for the establishment of an overall configuration risk management program (CRMP) and confirmation that its insights are incorporated into the decision-making process before taking equipment out-of-service prior to or during the CT. Compared with Tier 2, Tier 3 provides additional coverage based on any additional risk significant configurations that may be encountered during Page 10 of 24

ATTACHMENT 1 Evaluation of Proposed Changes maintenance scheduling over extended periods of plant operation. Tier 3 guidance can be satisfied by the Maintenance Rule (10 CFR 50.65(a)(4)), which requires an assessment and management of the increase in risk that may result from activities such as surveillance, testing, and corrective and preventive maintenance.

The risk analysis provided in Attachment 5 supports the Tier 1 element of RG 1.177, specifically the comparison of the results with the acceptance guidelines for ICCDP and ICLERP associated with changing a TS CT. The Tier 2 and Tier 3 elements are addressed below.

Tier 2: Avoidance of Risk-Significant Plant Configurations The following compensatory measures all serve to lessen the calculated increase in the core damage and large early release risk when the 2A SX pump is out-of-service.

The risk-informed evaluation identified a number of compensatory measures that will be implemented during the planned 2A SX pump maintenance configuration to assure the risk impacts are acceptably low. These are discussed in detail in Attachment 5 and summarized below. The compensatory measures below are considered to be regulatory commitments and are included in Attachment 4.

The assessment of risk from internal events and internal fires identified the following actions as important compensatory measures that will help to reduce the overall risk during the performance of the extended CT:

1. There will be no elective maintenance work on the remaining SX pumps, (1A, 1 B, 2B) during the 2A SX extended CT. Additionally, this equipment will be protected for this one-time outage. This supports the maintenance assumptions in the risk analysis.

2. There will be no elective maintenance work on the emergency diesel generators (1A, 1 B, 2A, 2B) during the 2A SX extended CT. Additionally, this equipment will be protected for this one-time outage. This supports the maintenance assumptions in the risk analysis and also supports mitigation of a loss of offsite power during the maintenance window.

3. There will be no elective maintenance work on the Unit 2 auxiliary feed pumps (2A, 2B). This equipment will be protected for the one-time outage. This supports the maintenance assumptions in the risk analysis.

4. There will be no elective maintenance on the 1/2SX16A/B (i.e., RCFC SX inlet valves) and 1/2SX27A/B (RCFC SX outlet valves) due to interlocks that could prevent use of the remaining SX pumps. This supports the maintenance assumptions in the risk analysis.

5. There will be no elective maintenance on the 211, 2121 213, or 214 instrument busses or their associated inverters and transformers. Additionally, this equipment will be protected for the one-time outage. This supports the maintenance assumptions in the risk analysis.

6. There will be no elective maintenance on the startup feedwater pump, 2FW02P.

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ATTACHMENT I

7. There will be no elective maintenance activities on the Unit 2 Station Auxiliary Transformers.

8. The extended weather forecast' will be examined to ensure severe weather conditions are not predicted prior to entry into the CT. In the event of an unforeseen severe weather condition due to rapidly changing conditions, such as severe high winds, a briefing with crew operators will be performed to reinforce operator actions and responses in the event of a loss of offsite power.

9. Fire Risk Management Actions (RMAs) applicable for the SX 2A pump will be completed per OP-AA-201-012-1001 "OPERATIONS ON-LINE FIRE RISK MANAGEMENT" (these actions protect against fire impacting key redundant equipment).

10.Operations will hold briefings on the following actions:

o On a loss of all Reactor Coolant Pump (RCP) seal cooling, Operations trips RCPs in time to prevent damage to the Shutdown Seals relied on for extended loss of seal cooling events.

o On a post-trip loss of AF, Operations initiates flow from either the motor-driven feedwater pump (2FW01 PA) or the startup feedwater pump (2FW02P) to at least one SG prior to reaching dry SG conditions.

o Operators manually throttle 0/2SX007 valves when the Residual Heat Removal (RHR) heat exchangers are used for ECCS recirculation.

o On a loss of Unit 2 SX, Operations opens the 1/2SX005 valve(s) to crosstie SX between the units.

o Operations refills the diesel-driven AF day tank from the 125K Fuel Oil Storage Tank in order to maintain operation of the diesel-driven AF pump.

o On loss of Vital Instrument Bus (120 VAC) 111 or 114, Operations opens the AF flow control valves 2AF005A-D ("A" train) or 2AF005E-H ("B" train) by locally failing air to the valve operators, then Operations throttles 2AF01 3A-D ("A" train) or 2AF01 3E-H ("B" train) from the Main Control Room (MCR) to control SG levels.

11. Prior to entering the TS 3.7.8 Action Statement for repair of the 2A SX pump, an operating crew shift briefing and pre job walkdowns will be conducted to reduce and manage transient combustibles and to alert the staff about the increased sensitivity to fires in the following fire zones during the extended SX 2A outage window. Operating crew shift briefings will continue to be conducted every shift throughout the duration of the CT period. Additionally, planned hot work activities in the following fire zones will be prohibited during the time within the extended SX 2A CT. In the event of an emergent issue requiring hot work in one of the listed zones, additional compensatory actions will be developed to minimize the risk of fire. The listed fire zones were identified based on risk significance in the FPRA results (generally zones with Division 2 equipment that impact SX). (The purpose of these walkdowns is to reduce the likelihood of fires in these zones by limiting transient combustibles, ensuring transients, if required to be present, are located away from fixed ignition sources and eliminating or Page 12 of 24

ATTACHMENT 1 Evaluation of Proposed Changes isolating potential transient ignition sources, e.g., energized temporary equipment and associated cables).

Fire Zone(l) Fire Zone Description 5.1-2 Division 22 ESF Switchgear Room 5.1-1 Division 12 ESF Switchgear Room 1

3.2-0 Auxiliary Building El. 439 -011 11.4-0 Auxiliary Building General Area, El. 383' Division 22 Containment Electrical Penetration 11.6-2 Area, El. 426' 11.2C-2 Containment Spray Pump 2B Room 11.1B-0 Unit 2 Auxiliary Building Basement El. 330' 18.10D-2 Unit Auxiliary Transformer 241-2 18.10E-2 System Auxiliary Transformers 242-1/242-2 (1) For larger fire zones walkdowns may be focused on specific fire sensitive areas within the larger firezones. Walkdowns are judged as not being required for areas with continuous operator occupation (e.g. MCR). Fire Risk Management Actions (RMAs) where they occur may address the need for walkdowns in some of these areas. ALARA principles apply when reviewing radiological areas such as RHR.

Since the first compensatory measure that will be taken while in the extended CT is that certain other PRA-modeled equipment will not be voluntarily taken out-of-service, risk-significant plant configurations are inherently avoided. Additionally, should an emergent condition arise such that plant equipment becomes unavailable, in addition to the planned out-of service equipment, the associated risk will be assessed and managed in accordance with the Tier 3 program discussed below.

Tier 3: Risk-Informed Configuration Risk Management Consistent with 10 CFR 50.65(a)(4), Braidwood has developed a CRMP governed by station procedures that ensures the risk impact of equipment out of service is appropriately evaluated prior to performing any maintenance activity. This program requires an integrated review to uncover risk-significant plant equipment outage configurations in a timely manner both during the work management process and for emergent conditions during normal plant operation.

Appropriate consideration is given to equipment unavailability, operational activities like testing or load dispatching, and weather conditions. Braidwood currently has the capability to perform a configuration dependent assessment of the overall impact on risk of proposed plant configurations prior to, and during, the performance of maintenance activities that remove equipment from service. Risk is re-assessed if an equipment failure/malfunction or emergent condition produces a plant configuration that has not been previously assessed.

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ATTACHMENT 1 Evaluation of Proposed Changes For planned maintenance activities, an assessment of the overall risk of the activity on plant safety, including benefits to system reliability and performance, is currently performed prior to scheduled work. The on-line assessment is controlled by EGC procedure WC-AA-101, "On-Line Work Control Process," and includes the following considerations:

• Maintenance activities that affect redundant and diverse structures, systems, and components (SSCs) that provide backup for the same function are minimized.

• The potential for planned activities to cause a plant transient are reviewed and work on SSCs that would be required to mitigate the transient are avoided.

• Work is not scheduled that is highly likely to exceed a TS or Technical Requirements Manual completion time requiring a plant shutdown. For activities that are expected to exceed 50% of a TS Completion Time, compensatory measures and contingency plans are considered to minimize SSC unavailability and maximize SSC reliability.

• For Maintenance Rule Program High Safety Significant SSCs, the impact of the planned activity on the unavailability performance criteria is evaluated.

• As a final check, a quantitative risk assessment is performed to ensure that the activity does not pose any unacceptable risk. This evaluation is performed using the impact on both CIDF and LERF. The results of the risk assessment are classified by a color code based on the increased risk of the activity as shown below.

Color Meaning Plant Impact and Required Action Green Non-risk significant Small impact on plant risk Requires nos ecific actions Yellow Non-risk significant with non- Impact on plant risk quantitative factors applied Limit unavailability time or take compensatory actions to reduce plant risk Orange Potentially risk-significant Significant impact on plant risk Requires senior management review and approval prior to entering this condition Requires compensatory measures to reduce risk including contingency plans All entries will be of short duration Red Risk-significant Not entered voluntarily If this condition occurs, immediate and significant actions taken to alleviate the problem Emergent work is reviewed by shift Operations to ensure that the work does not invalidate the assumptions made during the work management process. EGC's risk management procedure has been implemented at Braidwood. This procedure defines the requirements for ensuring that the PRA model used to evaluate on-line maintenance activities is an accurate model of the current plant design and operational characteristics. Plant modifications and procedure changes are monitored, assessed, and dispositioned. Evaluation of changes in plant configuration or PRA model features are dispositioned by implementing PRA model changes or by the qualitative assessment of the impact of the change on the configuration risk management Page 14 of 24

ATTACHMENT I Evaluation of Proposed Changes tool. Changes that have potential risk impact are recorded in an update requirements evaluations (URE) log for consideration in the next periodic PRA model update.

The reliability and availability of the SX pumps are monitored under the Maintenance Rule (MR)

Program. If the pre-established reliability or availability performance criteria are exceeded for the SX pumps, they are considered for 10 CFR 50.65, "Requirements for monitoring the effectiveness of maintenance at nuclear power plants," paragraph (a)(1) actions, requiring increased management attention and goal setting in order to restore their performance (i.e.,

reliability and availability) to an acceptable level. The performance criteria are risk-informed and, therefore, are a means to manage the overall risk profile of the plant. An accumulation of large core damage probabilities over time is precluded by the performance criteria.

The SX pumps are all currently in the 10 CFR 50.65(a)(2) MR category (i.e., the SX pumps are meeting established performance goals). Repair of the 2A SX pump is not anticipated to result in exceeding the current established MR criteria for the SX pumps.

Plant modifications and procedure changes are monitored, assessed and dispositioned.

Evaluation of changes in plant configuration or PRA model features are dispositioned by implementing PRA model changes or by qualitatively assessing the impact of the changes on the CRMP assessment tool. Procedures exist for the control and application of CRMP assessment tools, and include a description of the process when the plant configuration of concern is outside the scope of the CRMP assessment tool.

The goals of a CRMP are to ensure that risk significant plant configurations will not be inadvertently entered for planned maintenance activities, and appropriate actions will be taken should unforeseen events place the plant in a risk-significant configuration during the proposed 2A SX pump extended CT.

This request has been evaluated consistent with the key principles identified in RG 1.174 for risk informed changes to the licensing basis and demonstrates that the risk from the proposed change is acceptably small. The evaluation with respect to these principles is summarized below.

The proposed change meets the current regulations unless it is explicitly related to a requested exemption or rule change.

This LAR itself does not propose to deviate from existing regulatory requirements, and compliance with existing regulations is maintained by the proposed one time change to the plant's TS requirements.

The proposed change is consistent with the defense-in-depth philosophy.

The configuration to be entered decreases the redundancy of the SX system due to the removal of one of the SX pumps from service. The reduced redundancy increases the potential for the plant to lose SX cooling to plant equipment; however, the current plant design and supporting analyses demonstrate that the plant has sufficient capability to prevent and mitigate a loss of SX.

Page 15 of 24

ATTACHMENT 1 Evaluation of Proposed Changes Defense-in-depth is maintained during the configuration. Compensatory measures are identified to strengthen the level of defense-in-depth and reduce overall risk.

The proposed change maintains sufficient safety margins.

The proposed TS change is consistent with the principle that sufficient safety margins are maintained based on the following:

• Codes and standards (e.g., American Society of Mechanical Engineers (ASME), Institute of Electrical and Electronic Engineers (IEEE) or alternatives approved for use by the NRC). The proposed change is not in conflict with approved codes and standards relevant to the SX system.

• While in the proposed configuration, safety analysis acceptance criteria in the UFSAR are met, assuming there are no additional failures.

When proposed changes result in an increase in core damage frequency or risk, the increases should be small and consistent with the intent of the Commission's Safety Goal Policy Statement A risk evaluation was performed that considers the impact of the proposed change with respect to the risks due to internal events, internal fires, seismic events and other external hazards. The evaluation of the quantitative impacts of internal event risks due to the planned configuration demonstrate that the impact on the likelihood of core damage and large early release is well below the risk acceptance guideline. The fire evaluation determined that the impact on the likelihood of fire-related core damage is also below the risk acceptance guideline with sufficient compensatory measures. The risk associated with seismic events and other external hazards are either not impacted by the change or are bounded by the risk from internal events.

The impact of the proposed change should be monitored using performance measurement strategies.

EGC's Configuration Risk Management Program will effectively monitor the risk of emergent conditions during the period of time that the proposed change is in effect. This will ensure that any additional risk increase due to emergent conditions is appropriately managed.

Page 16 of 24

Evaluation of Proposed Changes 4.1 Applicable Regulatory Requirements/Criteria 10 CFR 50.36(c) provides that TS will include Limiting Conditions for Operation (LCOs) which 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 nuclear reactor is not met, the licensee will shut down the reactor or follow any remedial action permitted by the technical specifications until the condition can be met." The design of the unit-specific SX System must satisfy the requirements of 10 CFR 50.36, "Technical Specifications," paragraph (c)(2)(ii),

Criterion 3; and the design on the opposite-unit SX Systems must satisfy the requirements of 10 CFR 50.36, paragraph (c)(2)(ii), Criterion 4. These requirements state the following:

(ii) A technical specification limiting condition for operation of a nuclear reactor must be established for each item meeting one or more of the following criteria:

Criterion 3. A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

Criterion 4. A structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.

The proposed changes involve extensions of the affected CT from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br />. The LCOs themselves remain unchanged, as do the required remedial actions or shutdown requirements in accordance with 10 CFR 50.36(c). Therefore, the proposed changes are consistent with the current regulations.

Although not the direct subject matter of this requested amendment, the following 10 CFR 50, Appendix A, General Design Criteria apply to the SX System covered by the proposed changes in this amendment application.

CRITERION 2 DESIGN BASES FOR PROTECTION AGAINST NATURAL PHENOMENA "Structures, systems, and components important to safety shall be designed to withstand the effects of natural phenomena such as earthquakes, tornadoes, hurricanes, floods, tsunami, and seiches without loss of capability to perform their safety functions. The design bases for these structures, systems, and components shall reflect: (1)

Appropriate consideration of the most severe of the natural phenomena that have been historically reported for the site and surrounding area, with sufficient margin for the limited accuracy, quantity, and period of time in which the historical data have been accumulated, (2) appropriate combinations of the effects of normal and accident conditions with the effects of the natural phenomena and (3) the importance of the safety functions to be performed."

Page 17 of 24

ATTACHMENT 1 Evaluation of Proposed Changes Structures, systems, and components important to safety shall be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, including loss-of-coolant accidents. These structures, systems, and components shall be appropriately protected against dynamic effects, including the effects of missiles, pipe whipping, and discharging fluids, that may result from equipment failures and from events and conditions outside the nuclear power unit. However, dynamic effects associated with postulated pipe ruptures in nuclear power units may be excluded from the design basis when analyses reviewed and approved by the Commission demonstrate that the probability of fluid system piping rupture is extremely low under conditions consistent with the design basis for the piping."

CRITERION 5 SHARING OF STRUCTURES, SYSTEMS, AND COMPONENTS "Structures, systems, and components important to safety shall not be shared among nuclear power units unless it can be shown that such sharing will not significantly impair their ability to perform their safety functions, including, in the event of an accident in one unit, an orderly shutdown and cooldown of the remaining units."

CRITERION 34 - RESIDUAL HEAT REMOVAL "A system to remove residual heat shall be provided. The system safety function shall be to transfer fission product decay heat and other residual heat from the reactor core at a rate such that specified acceptable fuel design limits and the design conditions of the reactor coolant pressure boundary are not exceeded. Suitable redundancy in components and features, and suitable interconnections, leak detection, and isolation capabilities shall be provided to assure that for onsite electric power system operation (assuming offsite power is not available) and for offsite electric power system operation (assuming onsite power is not available) the system safety function can be accomplished, assuming a single failure."

CRITERION 35 EMERGENCY CORE COOLING "A system to provide abundant emergency core cooling shall be provided. The system safety function shall be to transfer heat from the reactor core following any loss of reactor coolant at a rate such that (1) fuel and clad damage that could interfere with continued effective core cooling is prevented and (2) clad metal-water reaction is limited to negligible amounts.

Suitable redundancy in components and features, and suitable interconnections, leak detection, isolation, and containment capabilities shall be provided to assure that for onsite electric power system operation (assuming offsite power is not available) and for offsite electric power system operation (assuming onsite power is not available) the system safety function can be accomplished, assuming a single failure."

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ATTACHMENT I "A system to remove heat from the reactor containment shall be provided. The system safety function shall be to reduce rapidly, consistent with the functioning of other associated systems, the containment pressure and temperature following any loss-of coolant accident and maintain them at acceptably low levels.

Suitable redundancy in components and features, and suitable interconnections, leak detection, isolation, and containment capabilities shall be provided to assure that for onsite electric power system operation (assuming offsite power is not available) and for offsite electric power system operation (assuming onsite power is not available) the system safety function can be accomplished, assuming a single failure."

CRITERION 44 - COOLING WATER "A system to transfer heat from structures, systems, and components important to safety, to an ultimate heat sink shall be provided. The system safety function shall be to transfer the combined heat load of these structures, systems, and components under normal operating and accident conditions.

Suitable redundancy in components and features, and suitable interconnections, leak detection, and isolation capabilities shall be provided to assure that for onsite electric power system operation (assuming offsite power is not available) and for offsite electric power system operation (assuming onsite power is not available) the system safety function can be accomplished, assuming a single failure."

CRITERION 45 INSPECTION OF COOLING WATER SYSTEM "The cooling water system shall be designed to permit appropriate periodic inspection of important components, such as heat exchangers and piping, to assure the integrity and capability of the system."

CRITERION 46 TESTING OF COOLING WATER SYSTEM "The cooling water system shall be designed to permit appropriate periodic pressure and functional testing to assure (1) the structural and leaktight integrity of its components, (2) the operability and the performance of the active components of the system, and (3) the operability of the system as a whole and, under conditions as close to design as practical, the performance of the full operational sequence that brings the system into operation for reactor shutdown and for loss-of-coolant accidents, including operation of applicable portions of the protection system and the transfer between normal and emergency power sources."

Braidwood Station UFSAR Section 3.1 documents Braidwood Station's compliance with the NRC GDC. The changes in CT associated with this amendment request do not impact the compliance with these GIDCs as documented in the UFSAR.

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ATTACHMENT 1 Evaluation of Proposed Changes Generic Letter 91-13 "Request for Information Related to the Resolution of Generic Issue 130,

'Essential Service Water System Failures at Multi-Unit Sites,' Pursuant to 10 CFR 50.54(f)"

This Generic Letter (GL) provided the industry with an evaluation of a generic safety issue related to the loss or failure of essential service water systems at multi-unit sites. The GL concluded that licensees should evaluate administrative type improvements that would significantly enhance the availability of the SX System at plants like Braidwood Station. In response to this GL, the Braidwood Station SX TS 3.7.8 was amended consistent with the GL recommendations; specifically adding the opposite unit's SX pump to the TS LCO and added associated Conditions, Required Actions, and Completion Times. This one-time change to the 2A SX train TS CT does not impact the response and evaluation associated with GL 91-13.

There are no changes being proposed in this amendment application such that commitments to the regulatory requirements and guidance documents above would come into question. The description of the SX System in the UFSAR remains valid during implementation of the proposed extension of the TS CT for the 2A SX pump repair.

4.2 Precedence The NRC has previously reviewed requests for TS changes in support of one-time extensions of LCO completion times as documented in the following approved amendments.

EGC submitted a one-time license amendment request (LAR) on September 10, 2015 to extend the Shutdown Service Water 2A subsystem CT for Clinton Power Station. The request proposed to extend the CT to restore the Division 2 Shutdown Service Water subsystem to operable status from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 7 days to allow replacement of the Division 2 SX pump. The NRC approved the LAR for Clinton Power Station in Amendment No. 207 on October 22, 2015 (Reference 8).

Additionally, EGC submitted a one-time LAR on June 11, 2003 to extend the SX train CT for Braidwood Station and Byron Station. The request proposed to extend the CT for Condition A from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 144 hours0.00167 days <br />0.04 hours <br />2.380952e-4 weeks <br />5.4792e-5 months <br /> to support SX pump suction valve replacement. The NRC approved the LAR in Amendment No. 130 for Braidwood Station and Amendment No. 136 for Byron Station on March 18, 2004 (Reference 9).

4.3 No Significant Hazards Consideration In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Exelon Generation Company, LLC (EGC) is requesting a license amendment to Facility Operating License No. NPF-77 for Braidwood Station, Unit 2. This amendment request proposes a one-time extension of the Completion Time (CT) to restore the Unit 2 SX "A" train to Operable status associated with Technical Specifications (TS) Limiting Condition for Operation (LCO) 3.7.8, "Essential Service Water (SX) System," from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br />. This proposed change will only be used one time during a Braidwood Station planned 2A SX pump repair scheduled to start the week of November 27, 2016.

The current TS Limiting Condition for Operation (LCO) 3.7.8 requires that two unit-specific SX trains (i.e., the "A" train and "B" train) must be operable in Modes 1, 2, 3 and 4 and that one Page 20 of 24

ATTACHMENT I opposite-unit SX train must be operable for unit-specific support. Condition A allows one unit-specific SX train to be inoperable with a CT of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. An extension of the CT to 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br /> (i.e., an increase of 128 hours0.00148 days <br />0.0356 hours <br />2.116402e-4 weeks <br />4.8704e-5 months <br />) is needed to repair and replace components of the 2A SX pump due to observed performance degradation. This CT extension for one train of SX to be inoperable from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br /> is needed to preemptively replace the 2A SX pump rotating elements, including the pump shaft and impeller and perform potential weld repairs to the cutwater region of the pump casing due to observed performance degradation. This evolution is not a typical maintenance activity that can be performed within the existing 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> CT and current planning estimates and maintenance history have shown that these pump components cannot be replaced within the current CT. Consequently, not being able to complete the 2A SX pump repair in the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> CT would require Braidwood Station to be shutdown in accordance with TS 3.7.8, Required Action C.1 and C.2.

The unit-specific SX System consists of two separate, electrically independent, 100% capacity, safety related, cooling water trains. Each train consists of a 100% capacity pump, piping, valving, and instrumentation. The pumps and valves are remote and manually aligned, except in the unlikely event of a loss of coolant accident (LOCA). The pumps are automatically started upon receipt of a safety injection signal or an undervoltage on the Engineered Safety Features bus, and all essential valves are aligned (automatically or manually) to their post accident positions. The SX System is also the backup safety-related water supply to the auxiliary feedwater system and the backup water supply to the fire protection system.

The proposed changes have been evaluated using the risk-informed processes described in Regulatory Guide (RG) 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," dated May 2011 and RG 1.177, "An Approach for Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications," dated May 2011. The risk associated with the proposed changes was shown to be acceptable.

According to 10 CFR 50.92, "Issuance of amendment," paragraph (c), a proposed amendment to an operating license involves no significant hazards consideration if operation of the facility in accordance with the proposed amendment would not:

(1) Involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) Involve a significant reduction in a margin of safety.

EGC has evaluated the proposed change to the TS for Braidwood Station, using the criteria in 10 CFR 50.92, and has determined that the proposed change does not involve a significant hazards consideration. The following information is provided to support a finding of no significant hazards consideration.

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ATTACHMENT 1 Evaluation of Proposed Changes

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

03= 0~E The proposed changes have been evaluated using the risk informed processes described in RG 1.1 74, "An Approach for Using Probabilistic Risk Assessment in Risk-informed Decisions on Plant-Specific Changes to the Licensing Basis," Revision 2 dated May 201 1, RG 1. 177, "An Approach for Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications,"

Revision 1 dated May 2011 and NRC Regulatory Guide 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-informed Activities,"

Revision 2 dated March 2009. The risk associated with the proposed change was found to be acceptable.

The previously analyzed accidents are initiated by the failure of plant structures, systems, or components. The SX System is not considered an initiator for any of these previously analyzed events. The proposed change does not have a detrimental impact on the integrity of any plant structure, system, or component that initiates an analyzed event. No active or passive failure mechanisms that could lead to an accident are affected. The proposed change will not alter the operation of, or otherwise increase the failure probability of any plant equipment that initiates an analyzed accident. Therefore, the proposed change does not involve a significant increase in the probability of an accident previously evaluated.

The proposed change does not alter or prevent the ability of structures, systems, and components (SSCs) from performing their intended function to mitigate the consequences of an initiating event within the assumed acceptance limits. The proposed change does not require any physical change to any plant SSCs nor does it require any change in systems or plant operations. The proposed one-time increase in the CT is consistent with the philosophy of the current TS LCO which allows one SX train to be inoperable for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. This change only extends the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> CT to 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br /> which has been shown to be acceptable from a risk perspective. The minimum equipment required to mitigate the consequences of an accident and/or safely shut down the plant will be Operable or available during the extended CT. The proposed change is consistent with the safety analysis assumptions and resultant consequences. Based on the above, the proposed change does not involve a significant increase in the consequences of an accident previously evaluated.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No The proposed changes do not involve the use or installation of new equipment and the currently installed equipment will not be operated in a new or different manner. No new or different system interactions are created and no new processes are introduced. The proposed changes will not introduce any new failure mechanisms, malfunctions, or accident initiators not already considered in the design and licensing bases. Based on this evaluation, the proposed change Page 22 of 24

ATTACHMENT 1 does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No The proposed change does not alter any existing setpoints at which protective actions are initiated and no new setpoints or protective actions are introduced. The design and operation of the SX System remains unchanged. The risk associated with the proposed increase in the time the 2A SX pump is allowed to be inoperable was evaluated using the risk informed processes described in RG 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," Revision 2 dated May 2011, RG 1.177, "An Approach for Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications," Revision 1 dated May 2011 and NRC Regulatory Guide 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision 2 dated March 2009. The risk was shown to be acceptable.

Based on this evaluation, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, EGC concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of no significant hazards consideration is justified.

4.4 Conclusions In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

5.0 ENVIRONMENTAL CONSIDERATION

EGC has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, "Standards for Protection Against Radiation." However, the proposed amendment does not involve: (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22, "Criterion for categorical exclusion; identification of licensing and regulatory actions eligible for categorical exclusion or otherwise not requiring environmental review,"

Paragraph (c)(9). Therefore, pursuant to 10 CFR 51.22, Paragraph (b), no environmental impact statement or environmental assessment needs be prepared in connection with the proposed amendment.

Page 23 of 24

ATTACHMENT 1 Evaluation of Proposed Changes

6.0 REFERENCES

1. Braidwood UFSAR, Section 9.2.1, "Station Service Water System"

2. Braidwood UFSAR, Section 6.2, "Containment Systems"

3. Braidwood UFSAR, Section 5.4.7, "Residual Heat Removal System"

4. NRC Generic Letter 91-13, Request for Information Related to the Resolution of Generic Issue 130, "Essential Service Water System Failures at Multi-Unit Sites," Pursuant to 10 CFR 50.54(f)

5. Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," Revision 2, dated May 2011

6. NRC Regulatory Guide 1.177, "An Approach for Plant-Specific, Risk-informed Decisionmaking: Technical Specifications," Revision 1, dated May 2011

7. NRC Regulatory Guide 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-informed Activities," Revision 2, dated March 2009

8. Letter from U. S. NRC to B. Hanson (Exelon Generation Company, LLC), "Clinton Power Station, Unit 1 Issuance of Amendment Related to One-Time Extension of Completion Time for Shutdown Service Water (CAC No. MF6705)(RS-1 5-264)," dated October 22, 2015

9. Letter from U. S. NRC to C.M. Crane (Exelon Generation Company, LLC), "Issuance of Amendments Re: One-Time Change to the Completion Time for Restoration of a Unit Specific Essential Service Water Train (TAC NOS. M139547, M139548, M139545, and MB9546)," dated March 18, 2004 Page 24 of 24

. ~ SX CONTROLS AND INDICATIONS 1 A>V 1PM061 Sta vs" swiarh Pomp Amw I S)MIAB Amt LO.too lisle ISX004 PV disdrtuI Inge 1 SX005 Pampa disahor Hdr scrap I SX146 Rehm Hdr how I SX10/11/136 (Robin Htir X-ie)

ISX0331034 (P-W H& X-tie) 1SX112/114AB & 147A/B mftioZn OSX165A/B (OPMOIJ) p 0SXI15A 1

I B=

SX Pmwfi 1 Resote%ocal awito}nes Puatp 8 l I

OSX165A Sbwttswp awitdtes Rattmu Hd r amp 1 0 1150 ' Power I

]teotovad ALARMS j1B 1

1 I

10 1AF024 IA AF ppReeic 1B AF pp Raeie Pwsp Dsach Tanp Puetp Discit Pmts Wasp Suction Press Law Himh 90 paig 10 prig 9eF Low 35-F p Il 1 1 Strainer DP 9 paid or > 4 ashossw of coataraas I I~ bKkwath I Labe Oil Press LOW 6 paig I IC I 49-r r _ -* Injection Chtam p Suchien vtv Pit Lv1 Hi p CHEMICALS I) Sodium HypocLiarile lo oonid fangs and alpe plix is added at LSH prior 10 pomp and ,taper

2) Baoside to ia, , effidisicy of Sadism Hypoddoriiw 12C 3) Sift Disp:ast / Scsic Ltibilor / Convoim ieLbilor SX LUBE OIL PUMPS

1) Supply oil pretaane as start

2) Atoo starts an low LO. press said asst' of the 1bYow:g:

a) SX pump breaks cbaed

- IB S?i Cub Co b) WW Bs UV Sequmoer SSPS 11617 aaanhsa~

1B SX pp Oil c) _PM06J CS in CLOSE or if local seiec/ed, RSP CS in CLOSE I 411 ,I 1I2A _ _ 11=658 STRAINERS (I per Pump)

Ldw Screen Pbwar t) 24,000 Was copecR7 e 1-5 paid Howe P-OmJa lieaaoted 2) Nasal DP is 4 paid

3) Beckwaahes at SOB Spot for 3 this.

4) Aab Badtwaab err:

a) Hi DP o6.5 psid b) Tmar (0 - 30 )tows), Narsdly sat at 8 boss 2A SX Cub Co 1. 22 N 2A SX pp Oi'. Cncder A SX OPERATION ON UNIT I SI 22 1) 2nd pomp had both lobe al pps start.

PURPOSE OSXIn 1 21 rr

2) ISX016A/B ops oo SI aiPn ISX027A/B open wbea law speedbk3r donor

3) Cesdaiases WO isdatea 2A SX pp 1)Provide coding weir to safely rdased o*Wment

4) SX to VP CJr7ar bypwed and isaialed:

and equipment cum" b tiro safe ahtkbwn of the PUMPS CUBICLE COOLERS 1SX112AB & ISXII4A/B chase Reactor during normal or accident conditfcas. Ddhw 24,000 Spot oasts and wo ean4oilad fins the / ) Used so cod safety related pump foams I SX 147AM opal DESIGN BASIS 2) Provide badmp source of water to the Auxiliary MCR or 66RSDP.They wopowatad Sam* 2) Fans are controlled by a ttmnostat 5) RCPCs swap or shat (if in slandry)in law speed

- SX Pond is uhimab Heat Sink whkh is aimed b pmW Feed Pumps in the avert the Condensab StOsago IA But 141 3) SX always flow dwough Hx and fir cycks 6) ISX169AB opens at 280 spin to happily cooling for 30 days with postulated LOCA in one Unit and Tank (CST) is not available. 4) Designed b maintain room temp < 12n

3) Provide otfaly-edited bwk-W satrap of waEa to 1B Bs 142 D/G an not sign Wm unit being alastdawn. This pieveam adverse environmental

- SX Had Sink doa*nW to ddtver water at maximum of 100-F the Fie Prutectior System.

7) ISX178 opens at 350 rps b sappiy eonditiomp and allows proper opersh rt-mtdw wad expected meteorological oordhiau.

PUMP TRIPS Dial Amt food pans

- During the recirculation phase of a Imp break LOCH, the system PUMP MANUAL START I} LowStr6onPtaalotuaf<ipaig TS can be split into two separate pains wing 60 common CC had 1) Suction valve open (SX001") ltar> 1Sssaxsstds COOLER FAN I) 3.7.8 SX: Two Unit SX spa'fi iuia, and ease Opposite exchaup w to moot single 6ailum alkir (passive component fiikue). 2) RCFC inlet open (SX016") 1) Auto starts and stops on tomperatwc.

Unit fait suet be err , W la MODES 14

3) RCFC outlet gt11t (SlDDx7A/I) 3) blond 2) Auto starts an a start of its misted pump

2) 3.7.9 UHS: Lvvd ?~90' (SS9'-0' ea ~lJ ssscw r), <lOrlc

4) Lube dl > 50 4) 8W Bus Undm-oltalec and aLopa after pwupp is shutdown and DESIGN CHARACTERISTICS tows tempaauac drops is less than svwaV UHS wnp, Battoot < 584' in MODES 1.4.

- SX System designed w provide two redundant fltll RCFC'S Refs e :

cepwity bops in each trait for cooling of essential PUMPAUTU START 1

1) 04 Star UV with suction tatty" (SMIAM) 4 RCFC s in two thins AX, B/D, M-42, I26 hoot lads. normally all aligned. Each fan has rwo RCFC VALVE INTERLOCKS 20E-1-4*MSX seeies.

- SX ptmups on lowed level ofAter. Building (330' level) speedi normally run m high Low 1) _SX016AB Auto Open on SI for sufficient NPSH.

Pimps are ESP powered. A Bus _41 _B Bw 42 STARTING DUTIES

1) A motor is atbwtd 2 starts without any conditions if aft blot speed is for FSF actuation. 2) _SX027AlB Auto Open if eidtcr tram's related RCFC High or Low Speed Figure 1 Essential Ser vice Hem Sink is safety rolatod, detigpcd to withstand flooding or Safe Shutdown Bartigwlm.

start is at a Cold Start ccmdition. A Cold Start is defimd as follows: The motor has SAT IDLE' for GRUATER THAN 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />.

On SI, any RCFC running in high speed breaker is closed.

3) Naidmr valve cm be closed if associated SX-1, ESSENTIAL trips and rcrtatia 20 seconds later in low SX Pad will be intact if Main Coding Pond is lost to 2) AN other ,urt9 must mw riot Start conditions prior to attempting speed. This prevents motor overloading SX pump is running. SERVICE WATER dike fYikue. a start. A I lot Start is defirwd as F.I71(ER Ste motor low RUN for due to the advrm conditions expected 4) Both valves will stroke closed if associated Two auction lima fiom Lake Saloon House provide flaw GRFAT1:R THAN 20 minutes OR, the motor has SAT MIX for in omutinment when St is required. Any SX pump is shutdown and the associated January 6, 2015, REV. 18 10 SX Pumps and two return littcs direct flow back to SX Pond GREATER THAN 45 minutes. If the NOW has nun )'ir less than Sl or Fm Running Sigma is present but SX Pad is excavated six feat lower than rat of Malt Coding Pond. 20 toloo s,It most SIT IDLE' far GRB/1TTliL TtMN 45 tolooks. fan mrmin8 is IoW spoai stays in low.

wifl immediately re-opon.

FOR TRAINING USE ONLY Md

ATTACHMENT Proposed Technical Specification Pages for Braidwood Station RESIVED TS, PAGES 3.7.8-1 3.7.8-2

SX System 3.7 PLANT SYSTEMS 3.7.8 Essential Service Water (SX) System LCO 3.7.8 The following SX trains shall be OPERABLE:

a. Two unit-specific 3X trains; and

b. One opposite-unit SX train for unit-specific support.

APPLICABILITY: MODES 1, 2, 3, and 4.

CONDITION l REQUIRED ACTION COMPLETION TIME I

A. One unit-specific 5X train inoperable. 1. Enter applicable Condition5 and Required Actions of LCO 3.8.1. "AC Sources-Operating," for Emergency Diesel Generatormak inoperable by SX.

Condition~ and Required Actions of LCO 3.4.6.

"RC5 Loops-MODE 4," for Residual Heat Removal loops made inoperable by SX.

BRAIDWOOD UNITS 1 & 2 3'7.8 1 Amendment XXX

SX System 3.7.8 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.1 -------- NOTE --------

Not applicable to

)ni t 2 duri nc repaj r of the 2A SX Dt,mi1 utli ng the Unit 2 planned SX System oLrt;aae starti nq the wQek of November 27, M IL Restore unit-specific 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> SX train to OPERABLE status, Restore unit-specific 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br /> SX train to OPERABLE status. 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />

- Dei~actA.l --------- NOTES --------

1Enter applicable Conditions and B. Opposite-unit SX train B.1 Restore opposite-unit 7 days Required Actions of LCO 3.8.1, "AC inoperable. I SX train to OPERABLE Sources-status. Operating," for Emergency Diesel Generator made inoperable by SX.¶ C. Required Action and 7 ¶ C.1 Be in MODE 3. 1 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 2Enter applicable Conditions and associated Completion Required Actions of LCO 3.4.6, Time of Condition A or AND i "RCS Loops-MODE 4," for Residual B not met. Heat Removal loops made inoperable by SX.1 C.2 Be in MODE 5. 136 hours0.00157 days <br />0.0378 hours <br />2.248677e-4 weeks <br />5.1748e-5 months <br /> unit-specific SX train to F~o status.

BRAIDWOOD UNITS 1 & 2 3.7.8 2 Amendment XXX

ATTACHMENT 3 Proposed Technical Specification Bases Pages for Braidwood Station (For Information Only)

RESIVED BASES PAGES B3.7.8-4 B3.7.8-5

SX System B 3.7.8 BASES APPLICABILITY In MODES 1, 2 1 3, and 4, the unit-specific SX System is a normally operating system that is required to support the OPERABILITY of the equipment serviced by the SX System and required to be OPERABLE in these MODES.

While a specific unit is in MODES 1, 2 9 3, or 4, the opposite-unit SX System must be available (independent of the opposite unit's MODE or condition) for unit-specific support. This minimizes the risk associated with loss of al unit-specific SX.

In MODES 5 and 6 the OPERABILITY requirements of the unit-specific SX System are determined by the systems it supports and there are no opposite-unit SX System requirements.

ACTIONS A.1 and A.2 If one unit-specific SX train is inoperable, action must be taken to restore OPERABLE status within ~2 h9HP.s. In this Condition, the remaining OPERABLE SX train is adequate to perform the heat removal function. However, the overall reliability is reduced because a single failure in the OPERABLE SX train could result in loss of the SX System function in the short term.

The 72 h9HP GGmPlletieH Time is based 9H the pedHRdaHt Gapabi!#4es affepded by the OPERABLE tr-aiR, and the lew ppebability Gf a PBA 9G r1l 1 y%y%_1 I Vill dur-iHq this time per

,ied..

Required Action A.2 requires restoring the unit-specific SX BRAIDWOOD UNITS 1 & 2 B 3.7.8 4 Revision XX

SX System B 3.7.8 ACTIONS (continued)

Required Actions A.1 and A.2 4have been modified by two Notes. The first Note indicates that the applicable Conditions and Required Actions of LCO 3.8.1, "AC Sources-Operating," should be entered if an inoperable SX train results in an inoperable emergency diesel generator. The second Note indicates that the applicable licable Conditions and Required Actions of LCO 3.4.6, "E Loops-MODE 4," should be entered if an inoperable SX train results in an inoperable decay heat removal train. These are exceptions to LCO 3.0.6 and ensure the proper actions are taken for these components.

B.1 If the opposite-unit SX train is not OPERABLE for unit-specific support, action must be taken to restore OPERABLE status within 7 days. In this Condition, if a complete loss of unit-specific SX were to occur, the SX System function would be lost. The 7 day Completion Time is based on the capabilities of the unit-specific SX System and the low probability of a DBA with a loss of all unit-specific SX occurring during this time period.

C.1 and C.2 If the unit-specific SX train or the opposite-unit SX train cannot be restored to OPERABLE status within the associated Completion Time, the unit must be placed in a MODE in which the LCO does not apply. To achieve this status, the unit must be placed in at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging plant systems.

BRAIDWOOD UNITS 1 & 2 B 3.7.8 5 Revision XX

ATTACHMENT 4 Summary of Regulatory Commitments

The following table identifies commitments made in this document. (Any other actions discussed in the submittal represent intended or planned actions. They are described to the NRC for the NRC's information and are not regulatory commitments.)

COMMITMENT TYPE COMMITMENT COMMITTED DATE ONE-TIME PROGRAM-OR "OUTAGE" ACTION MATIC (YES/NO) (YES/NO)

Implement the compensatory Upon implementation Yes No measures (items 1 through 11) of the one-time listed in Attachment 1 of RS extension of the SX 197, Section 3.0, "Technical train Completion Time.

Evaluation," Tier 2 Discussion

ATTACHMENT 5 BW-LAR-008, Revision 0 "Risk Assessment Input for the Braidwood One-Time Technical Specification Change for the Essential Service Water Pump 2A Completion Time from 72 to 200 Hours"

..ter r ExeLon Generation.

Braidwood PRA APPLICATION NOTEBOOK Risk Assessment Input for the Braidwood One-Time Technical Specification Change for the Essential Service Water Pump 2A Completion Time from 72 Hours to 200 Hours REVISION 0

11s/ DOCUMENTATION NO: IIW-LAIt-M Ii PAGE 90. 2 ~

STATION: Srafdwaod UNff(S) AFFECTID: Anil 2 TMJ!: Rlsk Asseement Input fbr the Dratdwood One-Tarn T*ohnkA 8peiflradon Cheap tsar the Emmoal Swvks Winter ftmp 2A Compledon Ting bans 72 Noum to 200 "Ours

SUMMARY

TN% asaeserrr d it p&tmlad In support of " Lio~ A wndmernt Pwgwst

{L submMad for a one-lir" change fa extend the CompleOw 71me (CT) for the tank .2 Eaw0al Servhce Walsr A pump from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br /> In order to allow for mpalm of the pump tram. Saw the oornpene tiory measure:In Salton 5.4.1 OW support this ftk sna"lo and

-cortstjhA QoW risk manapmenl salkw o for this 9X 2A ouWpe.

The risk ssee+mment le perforn ad in aomdenw with ER-hMM100, Rev. a, Risk Mehim NOEO and LAR jRef. I Rjj Review regLgred sfler I X I lnbmsl RM tmumMnftforr I I Erna!RM Docurantadoa E16cttonk Calw tfon Dais Ffts; '8' 4.AR-t7f &Adachrmnts.*8,149M&V-261&11:02AM lh th~t__of,R,Itrwr: f X J Dal,lled ] Alter ywk[ I RaM+lr+w o(Emmal Docwmnt This RM dcocurnentrtion supersedes: WA in Its errif".

Papamd by: Jeff SchappauM I _ .**r 1 Marne ~ sionat ew,. DIM rod by; Joe Edwn 91.261do memo Pnparod by: Usarns Farradj i If--2C , &

(Fire Ex wnaf Events) inn >g r ?ate 1.1me Prsp&nW by: _ Eric Tharrvs ur te_

r 1 jechnical Adequacy) 8~B re Marne RwNOFW by: .~ tow* 131 ~

Name &gnalure

Reviewed by: Heather Fax Name Signature Date Reviewed by: Clint Pierce I = t--'° 1

s. ` F 2d f (Fire External Eventsj Signature Date Name Reviewed by: Anthony Hable Name

~, 9126/2016 Approved by: Brandon Irvin ~'"

Name .,...,...,... Date

Section Page

1.0 INTRODUCTION

............................................................................................... 1-1 1.1 PURPOSE .............................................................................................. 1-1

1.2 BACKGROUND

...................................................................................... 1-1 1.2.1 Technical Specification Changes .................................................1-1 1.3 REGULATORY GUIDES ........................................................................ 1-2 1.3.1 Regulatory Guide 1.200, Revision 2 .............................................1-2 1.3.2 Regulatory Guide 1.174, Revision 2..... ..... -3 1.3.3 Regulatory Guide 1.177 Revision 1 ..............................................1-3 1.3.4 Acceptance Guidelines ................................................................ 1-5 1.4 SCOPE ................................................................................................... 1-7 1.5 BRAIDWOOD PRA MODELS ................................................................. 1-8 3.0 RISK ANALYSIS ................................................................................................ 3-1 3.1 ASSESSMENT OVERVIEW AND ASSUMPTIONS ............................... 3-1

3. 1.1 Overview ...................................................................................... 3-1 3.1.2 Assumptions ................................................................................ 3-3 3.2 INTERNAL EVENTS ............................................................................... 3-5 3.2.1 FPIE PRA Evaluation and Results ............................................... 3-5 3.2.2 Opposite Unit Impact .................................................................. 3-16 3.2.3 Peer Review Finding IFSO-A4-01 Sensitivity Analysis ............... 3-16 3.3 EXTERNAL EVENTS ............................................................................ 3-19 3.3.1 Assessment of Relevant Hazard Groups ................................... 3-19 3.3.2 Internal Fires .............................................................................. 3-19 3.3.3 Seismic ...................................................................................... 3-25 3.3.4 High Winds ................................................................................. 3-26 3.3.5 External Flood ............................................................................ 3-27 3.3.6 Other External Hazards and Conclusions .................................. 3-27 3.4 RESULTS COMPARISON TO ACCEPTANCE GUIDELINES .............. 3-28 3.5 UNCERTAINTY ASSESSMENT ........................................................... 3-30 3.5.1 Parametric Uncertainty ............................................................... 3-30 3.5.2 Model Uncertainty ...................................................................... 3-31 3.5.3 Completeness Uncertainty ......................................................... 3-32 3.5.4 Uncertainty Analysis Conclusions .............................................. 3-33 3 .6 RISK

SUMMARY

.................................................................................1. 3-34 4.0 TECHNICAL ADEQUACY OF PRA MODEL ................................................:....4-1 4.1 PRA QUALITY OVERVIEW .................................................................... 4-1 4.2 SCOPE ................................................................................................... 4-3

4.3 FIDELITY

PRA MAINTENANCE AND UPDATE ....................................4-4 4.4 STANDARDS .......................................................................................... 4-6 4.5 PEER REVIEW AND PRA SELF-ASSESSMENT .................................. 4-6 9/26/16

Braidwood SX 2A CT Extension 4.6 APPROPRIATE PRA QUALITY .............................................................. 4-7 4.6.1 Plant Changes Not Yet Incorporated into the PRA Model ............ 4-7 4.6.2 Consistency with Applicable PRA Standards .............................4-11 4.6.3 Relevant Peer Review Findings ................................................. 4-13 4.6.4 Identification of Key Assumptions ..............................................4-16 4.6.5 Fire PRA Peer Review Results and F&Os ................................. 4-17 4.7 GENERAL CONCLUSION REGARDING PRA CAPABILITY ...............4-34 5.0

SUMMARY

AND CONCLUSIONS .................................................................... 5-1 5.1 SCOPE INVESTIGATED ........................................................................ 5-1 5.2 PRA QUALITY ........................................................................................ 5-2 5.3 QUANTITATIVE RESULTS VS. ACCEPTANCE GUIDELINES .............5-3

5.4 CONCLUSION

S ..................................................................................... 5-3 5.4.1 Compensatory Measures ............................................................. 5-4

6.0 REFERENCES

.................................................................................................. 6-1 9/26/16

1.0 INTRODUCTION

1.1 PURPOSE The purpose of this analysis is to assess the acceptability, from a risk perspective, of a change to extend the Braidwood Station completion time (CT) for the 2A Essential Service Water (SX) Pump Train in Tech Spec (TS) 3.7.8 from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br /> in order to allow for replacement of the pump. These proposed changes are requested to be effective only during a one-time outage in the Fall of 2016.

The analysis follows the guidance provided in Regulatory Guide 1.200 Revision 2 [Ref.

1], "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-informed Activities."

1.2 BACKGROUND

1.2.1 Technical Specification Changes Since the mid-1980s, the NRC has been reviewing and granting improvements to TS that are based, at least in part, on probabilistic risk assessment (PRA) insights. In its final policy statement on TS improvements of July 22, 1993, the NRC stated that it ...

. . . expects that licensees, in preparing their Technical Specification related submittals, will utilize any plant-specific PSA or risk survey and any available literature on risk insights and PSAs... Similarly, the NRC staff will also employ risk insights and PSAs in evaluating Technical Specifications related submittals. Further, as a part of the Commission's ongoing program of improving Technical Specifications, it will continue to consider methods to make better use of risk and reliability information for defining future generic Technical Specification requirements.

The NRC reiterated this point when it issued the revision to 10 CFR 50.36, "Technical Specifications," in July 1995. In August 1995, the NRC adopted a final policy statement on the use of PRA methods in nuclear regulatory activities that encouraged greater use of PRA to improve safety decision-making and regulatory efficiency. The PRA policy statement included the following points:

1-1 9126/16

Braidwood SX 2A CT Extension

1. The use of PRA technology should be increased in all regulatory matters to the extent supported by the state of the art in PRA methods and data and in a manner that complements the NRCs deterministic approach and supports the NRCs traditional defense-in-depth philosophy.

2. PRA and associated analyses (e.g., sensitivity studies, uncertainty analyses, and importance measures) should be used in regulatory matters, where practical within the bounds of the state of the art, to reduce unnecessary conservatism associated with current regulatory requirements.

3. PRA evaluations in support of regulatory decisions should be as realistic as practicable and appropriate supporting data should be publicly available for review.

4. The Commission's safety goals and subsidiary numerical objectives are to be used with consideration of uncertainties in making regulatory judgments...

The movement of the NRC to more risk-informed regulation has led to the NRC identifying Regulatory Guides and associated processes by which licensees can submit changes to the plant design basis including Technical Specifications. Regulatory Guides 1.174 [Ref. 2] and 1.177 [Ref. 3] both provide processes to incorporate PRA input for decision makers regarding a Technical Specification modification.

1.3 REGULATORY GUIDES Three Regulatory Guides provide primary inputs to the evaluation of a Technical Specification change. Their relevance is discussed in this section.

1.3.1 Regulatory Guide 1.200, Revision 2 Regulatory Guide 1.200, Revision 2 [Ref. 1] describes an acceptable approach for determining whether the quality of the PRA, in total or the parts that are used to support an application, is sufficient to provide confidence in the results, such that the PRA can be used in regulatory decision-making for light-water reactors. This guidance is 1-2 9/26/16

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intended to be consistent with the NRCs PRA Policy Statement and more detailed guidance in Regulatory Guide 1.174.

It is noted that RG 1.200, Revision 2 endorses Addendum A of the ASIVIE/ANS PRA Standard [Ref. 5] as clarified in Appendix A of RG 1.200, Revision 2.

1.3.2 Regulatory Guide 1.174, Revision 2 Regulatory Guide 1.174 [Ref. 2] specifies an approach and acceptance guidelines for use of PRA in risk informed activities. RG 1.174 outlines PRA related acceptance guidelines for use of PRA metrics of Core Damage Frequency (CDF) and Large Early Release Frequency (LERF) for the evaluation of permanent TS changes. The guidelines given in RG 1.174 for determining what constitutes an acceptable permanent change specify that the ACDF and the ALERF associated with the change should be less than specified values, which are dependent on the baseline CDF and LERF, respectively.

RG 1.174 also specifies guidelines for consideration of external events. External events can be evaluated in either a qualitative or quantitative manner.

Since this LAR is for a one-time TS change, the ACDF and the ALERF of RG 1.1.74 do not specifically apply.

1.3.3 Regulatory Guide 1.177 Revision 1 Regulatory Guide 1.177 [Ref. 3] specifies an approach and acceptance guidelines for the evaluation of plant licensing basis changes. RG 1.177 identifies a three-tiered approach for the evaluation of the risk associated with a proposed TS change as identified below:

• Tier 1 is an evaluation of the plant-specific risk associated with the proposed TS change, as shown by the change in core damage frequency (CDF) and incremental conditional core damage probability 1-3 9/26/16

Braidwood SX 2A CT Extension (ICCDP). Where applicable, containment performance should be evaluated on the basis of an analysis of large early release frequency (LERF) and incremental conditional large early release probability (ICLERP). The acceptance guidelines given in RG 1.177 for determining an acceptable permanent TS change is that the ICCDP and the ICLERP associated with the change should be less than 1 E-06 and I E-07, respectively. RG 1.177 also addresses risk metric requirements for one-time TS changes, as outlined in Section 1.3.4 of this risk assessment.

• Tier 2 identifies and evaluates, with respect to defense-in-depth, any potential risk-significant plant equipment outage configurations associated with the proposed change. The licensee should provide reasonable assurance that risk-significant plant equipment outage configurations will not occur when equipment associated with the proposed TS change is out-of-service.

• Tier 3 provides for the establishment of an overall configuration risk management program (CRMP) and confirmation that its insights are incorporated into the decision-making process before taking equipment out-of-service prior to or during the CT. Compared with Tier 2, Tier 3 provides additional coverage based on any additional risk significant configurations that may be encountered during maintenance scheduling over extended periods of plant operation. Tier 3 guidance can be satisfied by the Maintenance Rule (10 CFR 50.65(a)(4)), which requires a licensee to assess and manage the increase in risk that may result from activities such as surveillance, testing, and corrective and preventive maintenance.

This risk analysis supports the Tier 1 element of RG 1.177, specifically the comparison of the results with the acceptance guidelines for ICCDP and ICLERP associated with 1-4 9/26/16

Braidwood SX 2A CT Extension changing a Technical Specification Completion Time. Other portions of the LAR submittal will address Tier 2 and Tier 3 elements.

1.3.4 Acceptance Guidelines Risk significance in an LAR is determined by comparison of changes in Core Damage Frequency (CDF) and Large Early Release Frequency (LERF) and values of Incremental Conditional Core Damage Probability (ICCDP) and Incremental Conditional Large Early Release Probability (ICLERP) produced by a permanent change to either the plant design basis or Technical Specifications to the guidelines given in Regulatory Guide 1.174 and Regulatory Guide 1.177. Reg. Guide 1.174 specifies the acceptable changes in CDF and LERF for permanent, changes. Reg. Guide 1.177 specifies the acceptable ICCDP and ICLERP for permanent changes, usually associated with changing CT.

Also Reg. Guide 1.177 directly addresses the risk metric requirements for one-time TS changes, as reproduced below:

"For one-time only changes to TS CTs, the frequency of entry into the CT may be known, and the configuration of the plant SSCs may be established. Further, there is no permanent change to the plant CDF or LERF, and hence the risk guidelines of Regulatory Guide 1.174 cannot be applied directly. The following TS acceptance guidelines specific to one-time only CT changes are provided for evaluating the risk associated with the revised CT.*

1. The licensee has demonstrated that implementation of the one-time only TS CT change impact on plant risk is acceptable (Tier 1):

• ICCDP of less than 1. OX 10-6 and an ICLERP of less than 1. OX 10"',or

• ICCDP of less than 1. OX 10-5 and an ICLERP of less than 1. OX 10-6 with effective compensatory measures implemented to reduce the sources of increased risk.

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2. The licensee has demonstrated that there are appropriate restrictions on dominant risk-significant configurations associated with the change (Tier 2).

3. The licensee has implemented a risk-informed plant configuration control program. The licensee has implemented procedures to utilize, maintain, and control such a program (Tier 3). "

Based on the available quantitative guidelines for other risk-informed applications, it is judged that the quantitative criteria shown in Table 1-1 represent a reasonable set of acceptance guidelines. For the purposes of this evaluation, these guidelines demonstrate that the risk impacts are acceptably low. This combined with effective compensatory measures to maintain lower risk will ensure that the TS change meets the intent of small risk increases consistent with the Commission's Safety Goal Policy Statement.

Table 1 -1 PROPOSED RISK ACCEPTANCE GUIDELINES RISK ACCEPTANCE BASIS GUIDELINE ICCDP < 1 E-6, or ICCDP is an appropriate metric for assessing risk impacts of out of service equipment per RG 1.177. This guideline is specified in Section 2.4 of RG 1.177.

ICCDP < 1 E-5 with effective compensatory measures implemented to reduce the sources of increased risk ICLERP < 1E-7, or ICLERP is an appropriate metric for assessing risk impacts of out of service equipment per RG 1.177. This guideline is specified in Section 2.4 of RG 1.177.

ICLERP < 1 E-6 with effective compensatory measures implemented to reduce the sourcesof increased risk 1-6 9/26/16

Braidwood SX 2A CT Extension This section addresses the requirements of RG 1.200, Revision 2 Section 3.1 which directs the licensee to define the treatment of the scope of risk contributors (i.e., internal initiating events, external initiating events, and modes of power operation at the time of the initiator). Discussion of these risk contributors are as follows:

• Full Power Internal Events (FPIE) The Braidwood PRA model used for this analysis includes a full range of internal initiating events (including internal flooding) for at-power configurations. The Essential Service Water system (SX) is credited in the PRA for support of cooling for Safety-Related cooling loads including the emergency diesel generators (DG), component cooling water (CC), auxiliary feedwater (AF), and various room coolers. The FPIE model is further discussed in Section 1.5.

• Low Power Operation - The FPIE assessment is judged to adequately capture risk contributors associated with low power plant operations.

The FPIE analysis assumes that the plant is at full power at the time of any internal events transient, manual shutdown, or accident initiating event. This analytic approach results in conservative accident progression timings and systemic success criteria compared to what may otherwise be applicable to an initiator occurring at low power. As such, low power risk impacts are not discussed further in this risk assessment.

• Shutdown / Refueling Braidwood does not have a shutdown PRA model, but instead relies upon deterministic methodology to assess defense-in-depth of key safety functions. The intent is for the unit to remain at-power for the duration of the extended CT.

• Internal Fires Braidwood currently has an interim, peer-reviewed fire PRA model. The Braidwood working Fire PRA [Ref. 10] is used to provide both quantitative and qualitative insights to the analysis of the SX 2A CT extension (refer to Section 3.3.2).

• Seismic - Braidwood does not currently maintain a Seismic PRA. A qualitative assessment is performed in this analysis (refer to Section 3.3.3).

• High Winds Braidwood does not have a high winds or tornado PRA.

A qualitative assessment is performed in this analysis (refer to Section 3.3.4).

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• External Flood Braidwood does not have an external flood PRA. A qualitative assessment is performed in this analysis (refer to Section 3.3.5).

• Other External Events - Other external event risks were assessed in the Braidwood IPEEE study [Ref. 13] and found to be insignificant risk contributors. These conclusions are revisited for this SX 2A CT extension assessment (refer to Section 3.3.6).

This section addresses the requirements of Section 3.1 of RG 1.200, Revision 2 [Ref. 1]

which directs the licensee to identify the portions of the PRA used in the analysis.

The PRA analysis uses the BB01 1 M full power internal events (FPIE) Level 1 Core Damage Frequency (CDF) model and the associated Level 2 Large Early Release Frequency (LERF) model to calculate the risk metrics [Ref. 7]. The PRA analysis also uses the fire model BB01 1 b-FL-B [Ref. 10] to calculate the risk metrics for full power internal fires to develop quantitative and qualitative risk insights. Section 3.2 details the internal events analysis using the FPIE PRA, and Section 3.3 details the fire risk assessment.

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Braidwood SX 2A CT Extension T

The analysis and documentation utilizes the guidance provided in RG 1.200, Revision 2.

The guidance in RG 1.200, Revision 2 indicates that the following steps should be followed to perform this study:

1. Per Section 3. of RG 1.200, include the following information regarding the PRA

a. Describe the SSCs, operator actions, and operational characteristics affected by the application and how these are implemented in the PRA model.

b. Provide a definition of the acceptance guidelines used for the application.

2. Per Section 3.1 of RG 1.200, identify the scope of risk contributors addressed by

a. If not full scope (i.e. internal and external), identify appropriate compensatory measures or provide bounding arguments to address the risk contributors not addressed by the model.

3. Per Section 3.2 of RG 1.200, identify the parts of the PRA used to support the application

a. Identify the logic model elements onto which the relevant SSCs, operator actions, and operational characteristics are mapped to the PRA model.

b. Identify the relevant accident sequences that are impacted by the changes identified in the first group.

4. Per Section 3.3 and 4.2 of RG 1.200, demonstrate the Technical Adequacy of the PRA 2-1 9/26/16

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a. Identify plant changes (design or operational practices) that have been incorporated at the site, but are not yet in the PRA model and justify why the change does not impact the PRA results used to support the b.Document that the parts of the PRA used in the decision are consistent with applicable standards endorsed by the Regulatory Guide. Provide justification to show that where specific requirements in the standard are not met, it will not unduly impact the results.

c.Document peer review findings and observations that are applicable to the parts of the PRA required for the application, and for those that have not yet been addressed justify why the significant contributors would not be impacted.

d.Identify key assumptions and approximations relevant to the results used in the decision-making process.

5. Per Section 4.2 of RG 1.200, summarize the risk assessment methodology used to assess the risk of the application

a. Include how the PRA model was modified to appropriately model the risk impact of the change request.

Table 2-1 summarizes the RG 1.200 identified actions and the corresponding location of that analysis or information in this report.

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RG 1.200 Actions Report Section 1 a. Describe the SSCs, operator actions, and operational characteristics Section 1.5 and affected by the application and how these are implemented in the PRA model.

Section 3.1.1 1 b. Provide a definition of the acceptance guidelines used for the application. Section 1.3.4

2. Identify the scope of risk contributors addressed by the PRA model. Section 1.4 2a. If not full scope (i.e., internal and external events), identify appropriate Section 3.3 compensatory measures or provide bounding arguments to address the risk contributors not addressed by the model.

3. Identify the parts of the PRA used to support the application Section 1.5 and Section 3 3a. Identify logic model elements that are mapped to the PRA model Section 3.1 and Section 3.2 3b. Identify the accident sequences impacted by those changes. Section 3

4. Demonstrate the Technical Adequacy of the PRA. Section 4 4a. Identify plant changes (design or operational practices) that have been Section 4.6.1, incorporated at the site, but are not yet in the PRA model and justify why the Table 4-1 change does not impact the PRA results used to support the application.

4b. Document that the parts of the PRA used in the decision are consistent with Section 4.6.2, applicable standards endorsed by the RG. Provide justification to show that Table 4-2 where specific requirements in the standard are not met, it will not unduly impact the results.

4c. Document PRA peer review findings and observations that are applicable to Section 4.6.3, the parts of the PRA required for the application, and for those that have not yet Table 4-3 been addressed justify why the significant contributors would not be impacted.

4d. Identify key assumptions and approximations relevant to the results used in Section 3.1 and the decision-making process. Section 3.5

5. Summarize the risk assessment methodology used to assess the risk of the Section 1.5 and application. Include how the PRA model was modified to appropriately model Section 3 the risk impact of the change request. I I 2-3 9/26/16

Braidwood SX 2A CT Extension This section evaluates the plant-specific risk associated with the proposed TS change, based on the risk metrics of CDF, ICCDP, LERF, and ICLERP.

3.1 ASSESSMENT OVERVIEW AND ASSUMPTIONS

3. 1.1 Overview This analysis is performed for unavailability of the SX 2A pump. The PRA analysis involves identifying the system and components or maintenance activities modeled in the PRA which are most appropriate for use in representing the extended CT configurations and comparing the results to the baseline. The base risk metrics for the FPIE PRA and the FPRA are established in Table 3.1-1.

Table 3.1 -1 BRAIDWOOD FPIE PRA CDF AND LERF BASE RISK METRICS Risk T BB011 b4- Unit 1 BB011b4- Unit 2(/y FPIE CDF 1.96E-5 1.94E-5 FPIE LERF 9.60E-7 9.52E-7 Risk Metric BB011b4-Unit BB011b4- Unit 2 (/irj~

Fire CDF 4.25E-5 5.32E-5 Fire LERF 7.73E-6 5.96E-6 The general configuration for the extended CT is Braidwood at-power on both units with the SX 2A pump train out of service. The planned maintenance is expected to focus on repair of the rotating element/impeller of the pump with a contingency to replace the entire pump within the requested extended CT. The pump maintenance will be done in a workweek where the pump maintenance will be the focus of the week and there will not be significant concurrent maintenance work. The opposite division train (213) and the Unit 1 SX trains (1A and 1 B) will be protected. Additionally, all station emergency 3-1 9/26/16

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• diesel generators will also be protected. A complete list of protected equipment and other compensatory measures is discussed in Section 5.4.1.

Initially, the PRA model was quantified using the base "average test and maintenance" PRA model with the SX 2A out for maintenance. The average test and maintenance model represents baseline assumed maintenance frequencies for all components with the exception of Technical Specification violations that are normally excluded in the disallowed maintenance logic in the base PRA model. After analyzing the risk results, other maintenance terms became candidates for restricting elective maintenance to help reduce the overall risk associated with the extended CT. Restricted maintenance is discussed further in Section 3.1.2. This configuration is represented in the PRA as shown below in Table 3.1-2.

Table 3.1-2 SX 2A PUMP EXTENDED CT CONFIGURATION REPRESENTATION BASIC EVENT DESCRIPTION VALUE 2SX01 PA ----- PMMM SX PUMP 2A UNAVAILABLE DUE TO MAINTENANCE TRUE FLAG-SX-PUMP-26 SX PUMP 2B IS IN STANDBY (FLAG) FALSE(')

FLAG-SX-PUMP-2A SX PUMP 2A IS IN STANDBY (FLAG) TRUE(')

2SX01 PB-----PMMM (2)

SX PUMP 2B UNAVAILABLE DUE TO MAINTENANCE FALSE 1SX01PA ----- PMMM SX PUMP 1A UNAVAILABLE DUE TO MAINTENANCE (2)

FALSE 1SX01PB ----- PMMM SX PUMP 1B UNAVAILABLE DUE TO MAINTENANCE (2)

FALSE DIESEL GENERATOR 2A UNAVAILABLE DUE TO 2DG2A ------- DGMM FALSE(')

MAINTENANCE AT POWER DIESEL GENERATOR 2B UNAVAILABLE DUE TO (3) 2DG2B ------- DGMM FALSE MAINTENANCE AT POWER DIESEL GENERATOR 1A UNAVAILABLE DUE TO (3) 1 DG1A-------DGMM FALSE MAINTENANCE AT POWER DIESEL GENERATOR 1 B UNAVAILABLE DUE TO (31 1 DG1 B-------DGMM FALSE MAINTENANCE AT POWER AF MOTOR-DRIVEN PUMP 2AF01 PA UNAVAILABLE DUE (3) 2AF01 PA ----- PMMM FALSE TO MAINTENANCE AF DIESEL-DRIVEN PUMP 2AF01PB UNAVAILABLE DUE (3) 2AF01PB ----- PDMM FALSE TO MAINTENANCE SX PUMP A MIN FLOW PATH VIA SX 16A\ 27A (RCFC (3) 1SX016A027A-MVMM FALSE 1A\1 C) IS ISOLATED SX PUMP B MIN FLOW PATH VIA SX 16B\ 27B (RCFC 1SX016B027B-MVMM FALSE(3) 1 B\1 D) IS ISOLATED 3-2 9/26/16

Braidwood SX 2A CT Extension BASIC EVENT DESCRIPTION VALUE SX PUMP A MIN FLOW PATH VIA SX 16A\ 27A (RCFC 2SX016A027A-MVMM FALSE"'

2A\2C IS ISOLATED SX PUMP 2B MIN FLOW PATH VIA SX 16B\ 27B (RCFC 2SX016B027B-MVMM FALSE("

2B\2D) IS ISOLATED 21P211 ------ IXMM INVERTER 211 UNAVAILABLE DUE TO MAINTENANCE FALSE 21P212 ------ IXMM INVERTER 212 UNAVAILABLE DUE TO MAINTENANCE FALSE 21P213 ------ IXMM INVERTER 213 UNAVAILABLE DUE TO MAINTENANCE FALSE 21P214 ------ IXMM INVERTER 214 UNAVAILABLE DUE TO MAINTENANCE FALSE 480-120V TRANSFORMER 211 UNAVAILABLE DUE TO 21P211 ------ TRMM FALSE(')

MAINTENANCE 480-120V TRANSFORMER 212 UNAVAILABLE DUE TO 21P212 ------ TRMM FALSE(')

MAINTENANCE 480-120V TRANSFORMER 213 UNAVAILABLE DUE TO 21P213 ------ TRMM FALSE(')

MAINTENANCE 480-120V TRANSFORMER 214 UNAVAILABLE DUE TO 21P214 ------ TRMM FALSE(')

MAINTENANCE Notes to Table 3.1-2:

(1)The Braidwood PRA model BB011 b4 uses model flags to set the default alignment of the SX pumps.

Setting the flag to TRUE places the pump in standby, while setting the flag to FALSE indicates to the model that the pump should be running.

(2)During the proposed maintenance of the SX 2A, there is to be no elective maintenance of the 1A, 1 B, or 26 SX pumps.

(3)These terms represent additional elective maintenance that will be restricted during the execution of the SX 2A extended CT.

3.1.2 Assumptions The following assumptions are used in quantifying the plant risk due to the one-time SX KMI~oMill

• The SX 2A Pump CT is assumed to increase from its current duration of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to a proposed duration of 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br />.

• The base analysis in this risk assessment assumes one entry per year into the proposed CT for purposes of calculating changes in annual CDF. This is consistent with the current plans to enter the extended CT only once for a pump replacement/repair.

• This risk assessment does not credit the averted online risk due to a forced shutdown that would be required due to exceeding the existing CT.

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• The FRE PRA analysis assumes unavailability of the SX 2A pump via its corresponding maintenance basic event.

• No elective maintenance will be performed on the SX 1A, 1 B, or 2B pumps. These maintenance terms are set to FALSE for the quantification.

• There will be no elective maintenance work on the 1A, 113 7 2A, or 2B emergency diesel generators during the SX 2A extended CT. These maintenance terms are set to FALSE for the quantification.

• There will be no elective maintenance work on the Unit 2 auxiliary feed (AF) pumps. These maintenance terms are set to FALSE for the quantification.

• There will be no elective maintenance on the SX 16A/B or SX 27A/B on either unit due to interlocks that could prevent use of the remaining SX pumps. These maintenance terms are set to FALSE for the quantification.

• There will be no elective maintenance on the 211, 212, 213, or 214 instrument busses or their associated inverters and transformers. The inverter and transformer maintenance terms are set to FALSE for the quantification.

• Additional elective maintenance activities will be prohibited during the repair as compensatory measures to reduce plant risk that are not included in the quantification results. The complete list of restricted maintenance, protected equipment, and additional compensatory measures is summarized in Section 5.4.1.

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Braidwood SX 2A CT Extension 3.2 INTERNAL EVENTS The proposed technical specification change involves unavailability of the SX 2A pump.

The revised CIDF and LERF values for the CT configurations are obtained by re-quantifying the base PRA model with all of the identified events set as shown in Table 3.1-2. The SX 2A maintenance term was set to TRUE using a flag file, while disallowed maintenance terms were set to FALSE.

The evaluation of ICCIDP and ICLERP for the SX 2A CT change is determined as shown below:

The ICCIDP associated with SX 2A pump being OOS using the new CT is given by ICCDPSX 2A = (CDFSX 2A - CDFBASE) X CTNEW [Eq. 3-1]

where CDFsx 2A= the annual average CIDF calculated with the SX 2A equipment 00S CIDFBASE = baseline annual average CIDF with average unavailability for all equipment. This is the CIDF result of the baseline PRA.

CTNEW= the new extended CT (in units of hours, e.g. 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br />

• 1 year / 365 days / 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> = 2.28E-2 years)

Note: ICCIDP is a dimensionless probability.

Risk significance relative ICLERP is determined using equations of the same form as noted above for ICCIDP.

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Braidwood SX 2A CT Extension Since this evaluation is for a one-time Tech Spec CT allowance, the lCCDP and ICLERP are the only meaningful metrics as there is no permanent change in plant risk after this one-time CT extension.

The relevant inputs to Equation 3-1 (and the equivalent for LERF) are shown in Table 3.2-1 below. The corresponding output parameters from the equations above are then provided in Table 3.2-2. The analysis is performed for CDF and LERF from the internal events and internal floods Unit 2 PRA model.

Table 3.2-1 FPIE Risk Assessment Input Parameters and Results for Unit 2 Input Parameter Value CDFBASE 1.94E-05/yr(l)

CDFSX2A 2.25E-05/yr(l)

LER BASE 9.52E-07/yr (2)

LERFSX2A 1.00E-06/yr (2)

C NEW 2.28E-02/yr (i) basea on a truncation of it-iu (2) Based on a truncation of 1 E-11 Table 3.2-2 FPIE PRA Risk Assessment Base Outout Results for Unit 2 Risk Metric Value ICCDPSX2A 7.2E-08 ICLERPSX2A 1.1 E-09 In addition to the CDF/LERF calculations, a sequence review is performed as directed by ER-AA-600-1046 [Ref. 19]. This analysis consists of determining if significant changes to accident sequences exist due to the extended CT configuration.

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As shown in Table 3.2-3, for SX 2A OOS, general transient sequences contribute 41.5%

of the risk associated with the SX 2A Pump OOS sequence quantification. Sequences involving small loss of coolant accidents (SLOCA) contribute 30.1 % of the risk associated with the SX 2A Pump OOS sequence quantification, followed by loss of offsite power (LOOP) sequences at 11.1 %. These results indicate that the Transient (reactor trip), SLOCA and LOOP sequences dominate the source of station risk for the configuration with the SX 2A pump being out-of-service. In comparison to the base case results, the majority of the sequence results (with the exception of LOOP sequences) were similar to the base case results.

Table 3.2-3: Comparison of Sequence Contributions for the 2A SX Pumn OOS Case Sequence 2A SX OOS Base Case

% Contribution Group CDF Contribution MAN 1.02E-05 41.5% 42.5%

2SLOC 7.41E-06 30.1% 30.5%

2LOOP 2.74E-06 11.1% 6.1%

2SGTR 1.68E-06 6.8% 7.9%

2MLOC 1.48E-06 6.0% 7.0%

21 LOC 3.40E-07 1.4% 1.6%

2SLBI 2.21E-07 0.9% 1.0%

2LODC 2.06E-07 0.8% 1.8%

2ATWS 1.60E-07 0.6% 0.8%

2SLBO 1.53E-07 0.6% 0.7%

2LLOC 1.20E-08 0.0% 0.1%

= 2XLOC 2.32E-09 0.0% 0.0%

The second aspect of the SX 2A risk characterization that can be taken from the internal events model is the type of initiating event(s) that contribute to the CIDF associated with SX 2A being out of service. As shown in Table 3.2-4, the largest contribution comes from a loss of SX, followed by SLOCA and loss of offsite power events. Note that the Loss of SX, Loss, of Component Cooling Water, and Loss of AC Power (AP) (non-LOOP) events are evaluated through the Transient event trees for sequence quantification. These results are consistent with the results of the sequence analysis.

These insights indicate that transient/reactor trip, SLOCA, LOOP, and other initiators all have the potential to create a demand for SX.

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Braidwood SX 2A CT Extension Table 3.2-4: CIDIF Contribution by Initiating vent Group Initiatins Event Groue  % Contribution Base Case Contribution Loss of SX 34.3% 37.5%

Small LOCA 13.4% 16.1%

LOOP & DLOOP 11.4% 2.0%

SGTR 7.1% 8.6%

Internal Flooding 6.8% 7.7%

Loss of AP 6.7% 1.4%

Loss of CCW 6.6% 8.0%

Medium LOCA 6.4% 7.6%

Other 5.0% 7.0%

General Transient & LMFW 2.3% 3.9%

In addition, the dominant cutsets were reviewed. The top 20 cutsets (representing 50% of CIDF contribution) for the SX 2A unavailable configuration are shown in Table 3.2-5 3-8 9/26/16

Braidwood SX 2A CT Extension Table 3.2-5: Too 20 CDF Cutsets for the SX 2A Pijmn 00S ['_nnfimirntinn Cutset # Cutset Prob. Event Prob Event Event Description 1 2.90E-06 9.60E-01 %SXIE INDICATOR FOR SX INITIATING EVENT 2.16E-04 OSX01AB2AB-CPMFRIE FAILURE OF ALL SX PUMPS (1A/1B/2A/2B) TO RUN DUE TO CCF 4/4 1.40E-02 2FW-FWR --- EHSYOA OPERATORS FAIL TO EXECUTE FW RESTORATION 2 1.29E-06 9.60E-01 %SXIE INDICATOR FOR SX INITIATING EVENT 2.16E-04 OSX01AB2AB-CPMFRIE FAILURE OF ALL SX PUMPS (1A/1 B/2A/2B) TO RUN DUE TO CCF 4/4 6.25E-03 2AP-BOTHSAT-TRMM BOTH U2 SAT OOS FOR TM - 241 PWR VIA 141; 242 PWR VIA 142; 256 - 259 ON UAT 3 1.19E-06 6.98E-04 %RC-SLOC2-N-PSIE SMALL LOCA INITIATING EVENT (NOWISOLABLE) 3.40E-03 OPERATORS FAIL TO LOCALLY THROTTLE SX007 OSX007-ES13HMVOA TO CC HXS 5.00E-01 FLAG-CCHTXO-U1 CCW HTX 0 ALIGNED TO UNIT 1 4 1.19E-06 6.98E-04 %RC-SLOC2-N-PSIE SMALL LOCA INITIATING EVENT (NOWISOLABLE) 3.40E-03 OPERATORS FAIL TO LOCALLY THROTTLE SX007 OSX007-ES13HMVOA TO CC HXS 5.00E-01 FLAG-CCHTXO-U2 CCW HTX 0 ALIGNED TO UNIT 2 5 6.78E-07 3.99E-04 %RC-MLOC2 --- PMIE MEDIUM LOCA INITIATING EVENT 3.40E-03 OPERATORS FAIL TO LOCALLY THROTTLE SX007 OSX007-ES13HMVOA TO CC HXS 5.00E-01 FLAG-CCHTXO-U1 CCW HTX 0 ALIGNED TO UNIT 1 6 6.78E-07 3.99E-04 %RC-MLOC2--PMIE MEDIUM LOCA INITIATING EVENT 3-9 9/26/16

Braidwood SX 2A CT Extension Table 3.2-5: Top 20 CDF Cutsets for the SX 2A Pump OOS Confiquration Cutset # Cutset Prob. Event Prob Event Event Description OPERATORS FAIL TO LOCALLY THROTTLE SX007 3.40E-03 OSX007-ES13HMVOA TO CC HXS 5.00E-01 FLAG-CCHTXO-U2 CCW HTX 0 ALIGNED TO UNIT 2 UNIT 2 MAJOR FLOOD (>3,700GPM) FROM NORMAL 7 5.95E-07 4.23E-04 %FL2WSM3AO ---- T1 SERVICE WATER INTO AUX BLDG - COMMON 3.90E-03 FLMITIG-M3-T1 WS RECOV OF LOSS OF SX SEAL LOCA (COND PROS 3.60E-01 2FP-PRI-7D-HMVRA OF 2FP-PRI-7D-HMVRA + 0.21 SEAL FAIL 8 5.10E-07 6.98E-04 %RC-SLOC2-N-PSIE SMALL LOCA INITIATING EVENT (NOWISOLABLE) 7.30E-04 2RH-SP-X ---HPMOA OPERATORS FAIL TO STOP RH PUMPS 9 4.24E-07 9.60E-01 %APIE INDICATOR FOR AP INITIATING EVENT 2.21 E-03 2AP242 ------ BSLPIE BUS 242 FAILS JOINT HEP FOR OSX-XTIE--HMVOA AND 2AF-2.00E-04 2RX-JHEP26-HOADA AF005--HAVOA "10 3.33E-07 9.60E-01 %SXIE INDICATOR FOR SX INITIATING EVENT 2.16E-04 FAILURE OF ALL SX PUMPS (1A/1 B/2A/2B) TO RUN OSX01AB2AB-CPMFRIE DUE TO CCF 4/4 2.40E-03 CONDITIONAL PROBABILITY OF DLOOP GIVEN OAP-DLOOP-GT GENERAL TRANSIENT 6.70E-01 FRACTION OF CONDITIONAL LOOPS THAT ARE OAP-DLOOP-SC SWITCHYARD-CENTERED 11 2.78E-07 8.41 E-04 3.30E-04

%RC-SGTR2-A-HXIE 2RX-JHEP28-HOADA STEAM GENERATOR TUBE RUPTURE IN S/G 2A JOINT HEP FOR 2RC-DS-SGTRHDVOA AND 2RC-LCD ---- HSYOA 7

3-10 9/26/16

Braidwood SX 2A CT Extension Table 3.2-5: Top 20 CDF Cutsets for the SX 2A Puma OOS Confiauration Cutset # Cutset Prob. Event Prob Event Event Description 12 2.78E-07 8.41E-04 %RC-SGTR2-B-HXIE STEAM GENERATOR TUBE RUPTURE IN SIG 26 JOINT HEP FOR 2RC-DS-SGTRHDVOA AND 2RC-3.30E-04 2RX-JHEP28-HOADA LCD ---- HSYOA 13 2.78E-07 8.41E-04 %RC-SGTR2-C-HXIE STEAM GENERATOR TUBE RUPTURE IN S/G 2C JOINT HEP FOR 2RC-DS-SGTRHDVOA AND 2RC-3.30E-04 2RX-JHEP28-HOADA LCD ---- HSYOA 14 2.78E-07 8.41 E-04 %RC-SGTR2-D-HXIE STEAM GENERATOR TUBE RUPTURE IN S/G 2D JOINT HEP FOR 2RC-DS-SGTRHDVOA AND 2RC-3.30E-04 2RX-JHEP28-HOADA LCD ---- HSYOA 15 2.57E-07 FREQ OF EXPOSING RHR PUMP DISCHARGE 9.16E-07 %RCS-RHR-DISCHIE HEADERS TO RCS PRESSURE 2.80E-01 CONDITIONAL PROB OF LEAK 800 GPM GIVEN LEAK-800-150 I I I LEAK IS AT LEAST 150 GPM W =

16 2.55E-07 9.60E-01 %CCIE INDICATOR FOR CC INITIATING EVENT 4.96E-04 CCW PUMPS 2CC01 PA & 2CC01 PB FAIL TO RUN 2CC01 PA-B--CPMFRIE DUE TO CCF 2/4 JOINT HEP FOR 2CV-ALL ---- HPMOA AND 2RC-5.10E-03 2RX-JHEP87-HOADA PMTRI PAHSYOA 5.00E-01 FLAG-CCHTX0-U1 CCW HTX 0 ALIGNED TO UNIT 1 UNIT 2 SEAL LOCA >21GPM RANDOMLY OCCURS -

2.10E-01 SEAL-U2-TRANS NON-LOOP SEQUENCES 17 2.55E-07 9.60E-01 %CCIE INDICATOR FOR CC INITIATING EVENT 4.96E-04 CCW PUMPS 2CC01 PA & 2CC01 PB FAIL TO RUN 2CC01 PA-B--CPMFRIE DUE TO CCF 2/4 JOINT HEP FOR 2CV-ALL----HPMOA AND 2RC-5.10E-03 2RX-JHEP87-HOADA PMTRIPAHSYOA 3-11 9/26/16

Braidwood SX 2A CT Extension Table 3.2-5: Top 20 CDF Cutsets for the SX 2A Puma OOS Confiauration Cutset # Cutset Prob. Event Prob Event Event Description 5.00E-01 FLAG-CCHTXO-U2 CCW HTX 0 ALIGNED TO UNIT 2 UNIT 2 SEAL LOCA >21 GPM RANDOMLY OCCURS -

2.10E-01 SEAL-U2-TRANS 6 m 18 1

2.28E-07 1 __ __

9.60E-01 I

%SXIE I NON-LOOP SEQUENCES INDICATOR FOR SX INITIATING EVENT FAILURE OF ALL SX PUMPS (1A/1 B/2A/2B) TO RUN 2.16E-04 OSX01AB2AB-CPMFRIE DUE TO CCF 4/4 OPERATORS FAIL RECOGNIZE THE CUE TO 1.10E-03 2FW-FRH1 --- HSGOA SECONDARY COOLING 19 2.12E-07 9.60E-01 %APIE INDICATOR FOR AP INITIATING EVENT mi 2.21 E-03 2AP242 ------ BSLPIE BUS 242 FAILS JOINT HEP FOR OSX005-----HMVOA AND 2AF-AF005-2.00E-04 2RX-JHEP76-HOADA

-HAVOA 5.00E-01 FLAG-CCHTXO-U1 CCW HTX 0 ALIGNED TO UNIT 1 20 2.11 E-07 9.60E-01 %SXIE INDICATOR FOR SX INITIATING EVENT 1.57E-05 OSX-ALL ---- CSRPGIE SX STRAINERS - PLUGGED DUE TO CCF (4/4) 1.40E-02 2FW-FWR --- EHSYOA OPERATORS FAIL TO EXECUTE FW RESTORATION 3-12 9/26/16

Braidwood SX 2A CT Extension Consistent with the contribution identified in Table 3.2-1 by Sequence and Table 3.2-2 by Initiating Event, the top cutsets involve loss of SX, Loss of AP and Small LOCA events. A further review of the cutsets from Table 3.2-5 identifies the following operator actions as being important to the assessment, shown in Table 3.2-6.

Table 3.2-6: Significant Operator Actions From Cutset Reviews Basic Event Description 2FW-FWR --- EHSYOA OPERATORS FAIL TO EXECUTE FW RESTORATION OPERATORS FAIL TO LOCALLY THROTTLE SX007 OSX007-ES13HMVOA TO CC HXS Operating Crew briefings to identify and review these actions for the duration of the extended CT would be prudent. In addition to the operator actions, one maintenance unavailability event associated with the U2 SATs was also identified. A recommendation to the station to restrict any elective maintenance on the Unit 2 SATs as a compensatory action has been made though this action has not been incorporated into the PRA results.

Table 3.2-7 provides a review of basic event importance for the SX 2A pump unavailability case. This table shows basic events with more than 1 % contribution to CDF. For this table, FLAG events, alignment events, initiating events and so forth were excluded from this list.

Table 3.2-7: Basic Events with Greater than 1% CDF Contribution Event Description FV - CDF OPERATORS FAIL TO LOCALLY THROTTLE SX007 OSX007-ES13HMVOA 1.83E-01 TO CC HXS 2FW-FWR --- EHSYOA OPERATORS FAIL TO EXECUTE FW RESTORATION 1.45E-01 JOINT HEP FOR 2CV-ALL ---- HPMOA AND 2RC-2RX-JHEP87-HOADA 1.01 E-01 PMTRIPAHSYOA BOTH U2 SAT OOS FOR TM - 241 PWR VIA 141; 242 2AP-BOTHSAT-TRMM 6.43E-02 PWR VIA 142; 256 - 259 ON UAT JOINT HEP FOR 2RC-DS-SGTRHDVOA AND 2RC-2RX-JHEP28-HOADA 4.72E-02 LCD ---- HSYOA 3-13 9/26/16

Braidwood SX 2A CT Extension Table 3.2-7: Basic Events with Greater than 1% CDF Contribution Event Description FV - CDF RECOV OF LOSS OF SX SEAL LOCA (COND PROB OSX-XTIE-D-HMVRA 3.36E-02 OF OSX-XTIE-D-HMVRA + 0.21 SEAL FAIL 2RC-PMTRIPAHSYOA OPS FAIL TO TRIP RCP TO PROTECT SDS 3.29E-02 JOINT HEP FOR OSX-XTIE --- HMVOA AND (2FP-PRI-2RX-JHEP22-HOADA 3.12E-02 7X-HMVOA OR 2CV-ALL ---- HPMOA DIESEL GENERATOR 2B UNAVAILABLE DUE TO 2DG2B ------- DGMM 2.93E-02 MAINTENANCE AT POWER RECOV OF LOSS OF SX SEAL LOCA (COND PROS 2FP-PRI-7D-HMVRA 2,90E-02 OF 2FP-PRI-7D-HMVRA + 0.21 SEAL FAIL FAILURE TO MITIGATE >3700 WS FLOOD FOR T1 FLMITIG-M3-T1-WS 2,79E-02 SCENARIO RECOV OF LOSS OF SX SEAL LOCA (COND PROB 2CV-ALL-D--HPMRA 2,55E-02 OF 2CV-ALL-D-HPMRA + 0.21 SEAL FAIL 2RH-SP-X---HPMOA OPERATORS FAIL TO STOP RH PUMPS 2.54E-02 2DG2B ------- DGFS DG 2B FAILS TO START RANDOMLY 2.41 E-02 JOINT HEP FOR OSX-XTIE --- HMVOA AND 2AF-AF005-2RX-JHEP26-HOADA 1.95E-02

-HAVOA OPERATORS FAIL RECOGNIZE THE CUE TO 2FW-FRH1 --- HSGOA 1.74E-02 SECONDARY COOLING POWER XTIE FAILS EARLY (AF FAILED TIME TO 2AP-XTE-SBOHHBOA 1.58E-02 XTIE X30 MIN) SBO XTIE OSX-XTIE --- HMVOA OPERATORS FAIL TO OPEN SX UNIT XTIE VALVES 1.46E-02 FAILURE OF THE SDS TO ACTUATE AND INITIALLY ORC-SDSFAIL-SLOO 1.44E-02 SEAL 2AF01 PA-B--CPMFR AF PUMPS FAIL TO RUN DUE TO CCF (2/2) 1.40E-02 MFW MD START UP PUMP FW02P UNAVAILABLE 2FW02P ------ PMMM 1.38E-02 DUE TO MAINTENANCE XTIE FAILS FOR SEQUENCES WITH AF AVAIL FOR 2AP-XTL-SBOHHBOA 1.37E-02 FIRST FOUR HOURS (SBO XTIE 2DG2B ------- DGFR DG 2B FAILS TO RUN 1.21 E-02 SEQUENCER FAILS IN A MANNER THAT FAILS THE 2AP242SQCMB-SQMF 1.04E-02 DIESEL The basic events with the largest contribution (>10% each) are associated with operator actions. The top two basic events have been identified through the cutset review performed above. The third action is a combination HEP event that consists primarily of the action to trip the RCPs to prevent damage to the shutdown seals. The basic events associated with the capability of the 213 DG to start and run comprise 5% contribution to CDF. From this review, the following items are identified:

3-14 9/26/16

-Braidwood SX 2A CT Extension

• Operating Crew briefings to review the action to trip the RCPs to preclude damage to the Shutdown Seals for the duration of the extended CT would be prudent.

• Prohibit planned maintenance activities on the 2B DG for the duration of the extended CT condition. Note that this action is credited in the SX 2A pump unavailable quantification.

• Prohibit planned maintenance activities on the motor-driven Feedwater pump, 2FW02P, for the duration of the extended CT. Note that this action is not credited in the SX 2A pump unavailable quantification.

Compensatory Action Summary from the FPIE PRA Evaluation The following compensatory actions have been identified through review of the FPIE PRA results and are summarized below:

• Perform Operating Crew briefings on the actions to restore main feedwater and throttle the SX007 valves as needed.

• Perform Operating Crew briefings on the action to trip the RCPs to preclude damage to the Shutdown Seals for the duration of the extended CT.

• Prohibit planned maintenance activities on the 1A, 1 By 2A, and 2B DGs for the duration of the extended CT.

• Prohibit planned maintenance activities on the SX 1A, 1 B, and 2B pumps.

• Prohibit planned maintenance activities on the motor-driven Feedwater pump, 2FW02P, for the duration of the extended CT. Note that this action is not credited in the SX 2A pump unavailable quantification.

• Prohibit planned maintenance activities on the Unit 2 SATs for the duration of the extended CT. Note that this action is not credited in the SX 2A pump unavailable quantification.

3-15 9/26/16

Braidwood SX 2A CT Extension 3.2.2 Opposite Unit Impact Due to the crosstie capability of SX, the proposed extended CT was also analyzed for impact on Unit 1. As expected, the Unit 2 results for ICCDP and ICLERP were limiting.

The results for Unit 1 are shown below in Table 3.2-8. The results reported include all restricted maintenance terms as listed in Table 3.1-2.

Table 3.2-8 FPIE PRA Risk Assessment Output Results For Unit I Risk Metric Value ICCDPSX 2A U1 1.2E-08 F7IC ERPSX 2A UI 4.0E-10 3.2.3 Peer Review Finding IFSO-A4-01 Sensitivity Analysis A disposition of peer review F&Os was performed as part of the PRA Model Adequacy study in Section 4. This review determined that one Finding, IFSO-A4-01, had the potential to impact the extended CT results. IFSO-A4-01 reads as follows:

Effect of plant-specific maintenance practices on internal flooding: Though this may only be a documentation issue, in absence of supporting information for not considering maintenance-induced flooding, an increase in internal flood frequencies of approximately 1. 45 could apply.

To ensure this Finding does not have a significant impact on the results, a sensitivity analysis was performed by increasing all internal flooding frequencies by a factor of 1.45 (with an adjustment of 0.95 to account for plant capacity factor). This adjustment was made in the PRA model database and renamed as BW01 1 b4-FloodSens.rr. Using this adjusted database, the ICCDP and ICLERP calculations were re-performed for both the base case and the SX 2A extended CT configuration shown in Table 3.1-2. These results are shown in Tables 3.2-9 and 3.2-10.

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Table 3.2-9 FPIE Risk Assessment Input Parameters and Results for IFSO.A4.01 Sensitivity Case I nput

-- Parameter Value CDFBASE, FL SENS 1.99E-05/yr(l)

CDFSX 2A, FL SENS 2.31 E-05/yr(l)

LERFBASE, FL SENS 9.71 E-07/yr (2)

LERFSX 2A, FL SENS 1.02E-06/yr (2)

CTNEW,FLSENS 2.28E-02/yr (1) Based on a truncation of 1F-lo (2) Based on a truncation of 1 E-11 Table 3.2-10 FPIE PRA Risk Assessment Output Results for IFSO-A4-01 Sensitivity Case Risk Metric Value ICCDPFL-SENS 7.3E-08 ICLERPFL-SENS 1.1 E-09 The sensitivity results were then compared to the original results to determine significance of the impact. This is shown in Table 3.2-11 Table 3.2-11 Percent Change for IFSO-A4-01 Sensitivity Case SX 2A Flooding Percent Extended CT Sensitivity Change lCCDP 7.2E-08 7.3E-08 1.4%

ICLERP 1.1 E-09 1.1 E-09 1.5%

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Braidwood SX 2A CT Extension The sensitivity analysis results show a minimal change in lCCDP and ICLERP.

Therefore, it is determined that this open F&O does not have a significant impact on the evaluation for extending the SX 2A CT.

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3.3 EXTERNAL EVENTS 3.3.1 Assessment of Relevant Hazard Groups The purpose of this portion of the assessment is to evaluate the spectrum of external event challenges to determine which external event hazards should be explicitly addressed as part of the Braidwood SX 2A CT extension risk assessment.

The at-power PRA models used for this analysis include:

• internal events and internal floods, and

• internal fires.

In addition, the seismic, high winds, external floods, and other hazard groups are addressed qualitatively. It is noted that it is unnecessary to evaluate the low-power and shutdown contribution to the base CIDF and LERF since the change being proposed involves performance of the repair while at-power. Because a detailed low power and shutdown PRA model has not been developed for this plant, the analysis conservatively omits the risk reduction that would typically occur with the unit shutdown.

3.3.2 Internal Fires The impact on the internal fire risk profile due to the proposed CT extension is evaluated using the Braidwood Interim FPRA [Ref. 10], BB01 1 b-FL-B. The Braidwood FPRA is an interim implementation of NUREG/CR-6850 and other approved methodologies in that the model documentation has not been finalized. However, the model has undergone peer review by a review team assembled by the PWR Owners Group and changes as a result of Findings and Observations from the peer review impacting model quantification have been incorporated in the model. The final documentation for this model is expected to be issued by the end of 2016. Therefore, the FPRA is judged useful to develop both quantitative and qualitative insights for this risk assessment.

3-19 9/26/16

Braidwood SX 2A CT Extension The same process in Section 3.2 that was used for the ME model has also been used with the FPRA model results. The basic event changes for the equipment configuration during the extended CT are as shown in Table 3.1-2 for the SX A outage. The relevant inputs to Equation 3-1 are shown in Table 3.3-1 below. The corresponding output parameters from the equation above are then provided in Table 3.3-2. Note that equations apply to fire LERF as well and the relevant inputs are also shown in Table 3.3-1 with the output parameters provided in Table 3.3-2.

The fire risk insights and compensatory measures are focused on CDF since the results indicate that the impact on CDF risk measures is more significant than that associated with the fire impact on LERF risk. The ICFCDP due to fire is larger than for internal events; however, there is still considerable margin (i.e., more than a factor of 3) to the acceptance guidelines of 1.0E-5 (with implementation of effective compensatory measures).

Table 3.3-1 FIRE Risk Assessment Input Parameters Input Parameter Value FCDFBASE 5.32E-5/yr(l)

FCDFSX A 1.66E-4/yr(l)

FLERFBASE 5.96E-6/yr(l)

FLERFSX A 7.65E-6/yr(l)

CTNEW 2.28E-02/yr (1) Based on a truncation of 1 E-11 for CDF and 1 E-12 for LERF 3-20 9/26/16

Table 3.3-2 FIRE PRA Risk Assessment Base Outnut Results Risk Metric Value ICFCDPSX 2A 2.6E-06 ICFLERPSX2A 3.9E-08 A review of cutsets and importance measures is also performed to help understand the FPRA results. The FPRA results for CDF with SX 2A OOS indicate that the top cutsets are associated with failure to trip the RCPs or the loss of RCP seal cooling.

Significant Fire Zones and Compensatory Measures The fire CDF results from the SX 2A OOS case identified the fire zones that could result in an increased likelihood of core damage during the extended SX 2A outage window.

The fire zones with a contribution of greater than or equal to 1 % of the SX 2A OOS risk are listed in Table 3.3-3. These fire zones would potentially benefit from additional compensatory measures that could further reduce the risk of fires in these zones.

From a qualitative perspective, the high fire risk contributors for when SX 2A is unavailable are expected from fire scenarios that impact the opposite division of SX (SX 213). This is in line with the top fire zones listed in Table 3.3-3 where fire zones from Division 2 are shown to be most important during the SX 2A outage window.

Table 3.3-3 Fire CDF SX 2A OOS Significant Fire Zones Fire Fire Zone Description Importance Zone Contribution 5.1-2 Division 22 ESF Switchgear Room 26%

5.1-1 Division 12 ESF Switchgear Room 10%

3.2-0 L Auxiliary Building El. 439'-0" 9%

3-21 9/26/16

cli~ l il il ~ l il i~il il il il il l l l l I I I III III ~ I' L III 111 0 A A i

• Table 3.3-3 Fire COF SX 2A OOS Sianificant Fire Zones Fire Fire Zone Description Importance Zone Contribution 11.4-0 Auxiliary Building General Area, El. 383' 6%

11.6-2 Division 22 Containment Electrical Penetration Area, El. 426' 5%

11.2C-2 Containment Spray Pump 2B Room 4%

11.16-0 Unit 2 Auxiliary Building Basement El. 330' 4%

18.10D-2 Unit Auxiliary Transformer 241-2 1%

18.10E-2 System Auxiliary Transformers 242-1/242-2 1%

Heightened awareness in the form of shift briefs or pre-job walkdowns will be implemented to reduce and manage transient combustibles prior to entrance into the extended CT. Additionally, hot work will be limited in these areas during the extended SX 2A outage window. This heightened awareness when combined with the other compensatory actions will reduce the potential for core damage from postulated fire scenarios.

As part of the Braidwood Configuration Risk Management Program (CRMP), Risk Management Actions (RMAs) were identified to reduce the fire risk when equipment with an appreciable impact on core damage mitigation is taken out-of-service. The BW CRM Program includes RMAs for when SX 2A is taken OOS for longer than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> and are documented in BW-CRM-1 15, Revision 1 [Ref. 11 ]. For fire zones with high contribution, as specified in Table 3.3-3 above, RMAs that will be performed include maintaining detection and suppression systems, minimizing transient combustibles, maintaining fire zone barriers, and prohibiting hot work and temporary heat sources, prohibiting maintenance activities on certain panels, and avoiding switching at certain panels (as applicable, to the maximum extent possible without jeopardizing plant safety).

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Significant Operator Actions and Compensatory Measures The fire CDF results from the SX 2A OOS case identified the operator actions, if failed, that could result in an increased likelihood of core damage during the extended SX 2A outage window. The top five (5) operator actions with the greatest contribution are listed in Table 3.3-4.

Table 3.3-4 Fire CDF SX 2A OOS Sianificant Operator Actions Operator Action Description ntribution Contribution 2RC-PMTRIPAHSYOA-F (see independent action below for a more detailed description);

COMB0236-2 24%

OSX-XTIE --- HIVIVOA-F (see independent action below for more detailed description) 2RC-PMTRIPAHSYOA-F OPS FAIL TO TRIP RCP TO PROTECT SDS 18%

2AF-AF005--HAVOA-F (see independent action below for more detailed description);

COMB0210-2 15%

OSX-XTI E --- HIVIVOA-F (see independent action below for more detailed description)

OPERATORS FAIL TO OPEN SX UNIT XTIE OSX-XTI E --- HIVIVOA-F 3%

VALVES OPERATORS FAIL TO REFILL DDAFP FUEL OIL 2AF01PB-FO-HXVOA-F DAY TANK FROM STORAGE TANK - FIRE 3%

2AF-AF005--HAVOA-F (see independent action above for a more detailed description);

COMB0127-2 3%

OSX005 ----- HIVIVOA-F (OPERATOR ACTION TO OPERATORS FAIL TO RECOVER SX005 X-TIE MOVs UPON LOSS OF POWER)

OPERATORS FAIL TO OPEN AF005 VALVES 2AF-AF005--HAVOA-F (LOCALLY FAIL AIR) - FIRE 2%

2RC-PMTRIPAHSYOA-F (see independent action above for a more detailed description);

COMB0602-2 1%

2FP-PRI-7X-HMVOA-F (OPERATORS FAIL TO ALIGN FP SEAL COOLING - SX NON-PIPE FAILURE INITIATOR) I 3-23 9/26/16

Braidwood SX 2A CT Extension The significant operator actions are related to RCP trip/seal cooling, SX unit cross-tie, refueling the diesel-driven AFW day tank, and manual operation of AFW control valves.

These operator actions confirm the impact of fire on SX potentially resulting in failure of RCP thermal barrier cooling and cooling for seal injection pumps. Operator briefings on the importance of these actions will be performed prior to entering the SX 2A OOS configuration.

Summary of Compensatory Measure Impacts on Important Fire Zones Based on a review of results from the fire PRA contributors, the following compensatory actions are highlighted as important to reduce the risk from fire events during the performance of the extended CT:

• Heightened awareness in the form of shift briefs or pre job walkdowns to reduce and manage transient combustibles prior to entrance into the extended CT completion time will be used to alert the staff about the increased sensitivity to fires in the important zones shown in Table 3.3-3 during the extended SX 2A outage window. Additionally, hot work should be limited or prohibited in these areas during the extended SX 2A outage window.

• As part of the Braidwood Configuration Risk Management (CRM) Program, Risk Management Actions (RMAs) were identified to reduce the fire risk when equipment with an appreciable impact on core damage mitigation is taken out-of-service. The Braidwood CRM Program includes RMAs for when the SX 2A pump is taken OOS and these are documented in BW-CRM-1 15, Revision 1

[Ref. 11]. For fire zones with high contribution, as specified in Table 3.3-3 above, RMAs that will be performed include maintaining detection and suppression systems, minimizing transient combustibles, maintaining fire zone barriers, and prohibiting hot work and temporary heat sources, prohibiting maintenance activities on certain panels, and avoiding switching at certain panels (as applicable, to the maximum extent possible without jeopardizing plant safety).

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• Heightened awareness in the form of shift briefs or procedure reviews to better understand the risk significant operator actions will be provided during the SX 2A OOS timeframe.

The Fire PRA risk for the SX 2A OOS condition discussed in this section will be reduced below reported values through implementation of these additional controls.

Opposite Unit Impact Due to the crosstie capability of SX, the proposed extended CT was also analyzed for impact on Unit 1. As expected, the Unit 2 results for ICCDP and ICLERP were limiting.

The results for Unit 1 are shown below in Table 3.3-5. The results reported include all restricted maintenance terms as listed in Table 3.1-2.

Table 3.3-5 Fire PRA Risk Assessment Output Results for Unit I Risk Metric Value ICFCDPSX2AUl 2.8E-08 ICFLERPSX2AU1 1.5E-09 3.3.3 Seismic There is no quantitative Seismic PRA Model of Record for Braidwood. A Phase 1 seismic PRA (SPRA) model using non-plant-specific SSC fragility information was completed in 2013. However, this model, using the representative SSC fragility information, is not intended for quantitative evaluations. Thus a qualitative assessment must be performed. The Phase 1 seismic PRA model risk insights documented in BB-PRA-021.021.01, Revision 0, "Level 1 Seismic Quantification Notebook" [Ref. 12] was reviewed and considered in this qualitative assessment.

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Braidwood SX 2A CT Extension A fundamental concept of the seismic modeling is that similar components at the same location of a building will experience similar seismic forces. For Braidwood, station, all four SX pumps are located on the basement elevation of the Auxiliary Building and would be expected to experience the same seismic acceleration from a seismic event.

In addition, with each pump being the same style, make, and model, the pumps would be expected to have the same fragility. Therefore, any seismic event that would result in failure of one SX pump would be expected to result in the failure of all SX pumps.

Based on this information, assuming that the SX pumps are seismically correlated is reasonable. For the purposes of a seismic risk evaluation, the SX pumps can be treated as completely seismically correlated (i.e., a seismic event that can cause failure of one SX pump is likely to cause failure of all SX pumps).

Seismic hazards are estimated to be negligible contributors in this risk evaluation for the SX 2A pump being unavailable and are not included in the quantified risk results. The additional risk due to a seismic event is qualitatively evaluated as low and would not have a significant impact on the overall results or conclusion for this risk evaluation.

Motor driven pumps generally have high seismic capacity and a relatively low probability of failure compared to other components during a seismic event.

Therefore, unavailability of the SX 2A pump would not have a significant impact on overall seismic risk. This assessment concludes that seismic risk can be appropriately screened as a non-significant contributor to the risk of the proposed CT extension.

3.3.4 High Winds Similarly to the Seismic discussion above, Braidwood station does not have a peer reviewed high winds hazard PRA model. The impact of the proposed completion time extension will be addressed qualitatively for high winds hazards.

As noted in the Braidwood IPEEE report [Ref. 13] the Braidwood station design is such that risk from high winds and tornadoes is not significant based on the screening 3-26 9/26/16

Braidwood SX 2A CT Extension assessments performed in [Ref. 13] which illustrated compliance with NRC SRP requirements. This assessment concludes that high winds risk can be appropriately screened as a non-significant contributor to the risk of the proposed CT extension.

3.3.5 External Flood From the Braidwood IPEEE [Ref. 13], external floods were screened out as an acceptably small risk. Additionally, significant external floods would be a slow developing event which would be expected to allow time for restoration of the out of service SX 2A pump prior to presenting a significant challenge. The redundant SX pump room (i.e.1B and 2B SX pumps) is equipped with dual watertight doors which provides further mitigation of internal and external flooding.

3.3.6 Other External Hazards and Conclusions As noted in the IPEEE [Ref. 13], the risk impact from Transportation and Nearby Facilities is negligible and has been screened from further consideration as documented in the SER and UFSAR for Braidwood station.

In evaluating the risk impact associated with nearby facilities, [Ref. 13] identified that the NRC SER concluded that the facilities do not represent a significant risk to the plant.

Based on the conclusions documented in [Ref. 13], the risk impact from Transportation and Nearby Facilities is insignificant and can be screened from further consideration.

As discussed in [Ref. 13], external events characterized as "high winds, floods and other external events" in NUREG 1407 [Ref. 14] have been evaluated using the guidance provided in that document. Those evaluations have been madeon the basis of the UFSAR and NRC's SER for Braidwood station. The external events discussed above have uniformly been found to be of negligible risk significance for Braidwood station. This assessment concludes that the risk from the external events assessed 3-27 9/26/16

III! 1 11111 pillpiIIIIIIII I 1111 Ni l L*I I 111 0 A

& i I I i i' .I #I above can be appropriately screened as a non-significant contributor to the risk of the I Go 3.4 RESULTS COMPARISON TO ACCEPTANCE GUIDELINES Table 3.4-1 shows a comparison of the individual hazard group core damage risk metrics to the acceptance guidelines defined in Section 1.3.4.

Table 3.4-1 COMPARISON OF INDIVIDUAL HAZARD GROUP RESULTS TO ACCEPTANCE GUIDELINES Figure of Merit Value Acceptance Guideline Below Acceptance Guideline Internal Events and Internal Floods ICCDP 7.2E-08 <1.0E-06, or <1.0E-5(1) Yes ICLERP 1.1 E-09 <I.0E-07, or <1.0E-6(2) Yes Internal Fires ICCDP 2.6E-06 <1.0E-06, or <1.0E-5(1) Yes(i)

ICLERP 3.9E-08 <1.0E-07, or <1.0E-6(2) Yes Other Hazard Groups ICCDP Negligible <1.0E-06, or <1.0E-5(1) Yes ICLERP Negligible <1.0E-07, or <1.OE-6 (2) Yes Total Values ICCDP 2.7E-06 <I.0E-06, or <1.0E-5(1) Yes(i)

ICLERP 4.0E-08 <1.0E-07, or <1.0E-6(2) Yes Per RG 1.177 a value between 1E-06 and 1E-05 may be deemed acceptable with effective compensatory measures implemented to reduce the sources of increased risk.

(2)

Per RG 1.177 a value between 1E-07 and 1E-06 may be deemed acceptable with effective compensatory measures implemented to reduce the sources of increased risk.

The results indicate that the acceptance guideline values for a one-time extension are not exceeded for the ICCDP and ICLERP risk metrics. The internal events ICCDP and ICLERP results are far below the threshold for the acceptance guidelines, while the total 3-28 9/26/16

Braidwood SX 2A CT Extension values for lCCDP and ICLERP fall within the acceptable range with effective compensatory measures. Additional compensatory measures would potentially reduce risk further, such as protected equipment and heightened awareness of important operator actions and high risk fire zones. These additional measures are not accounted for in the quantification.

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'L30raidwood SX 2A CT Extension 3.5 UNCERTAINTY ASSESSMENT This section evaluates epistemic uncertainties that could impact the SX 2A CT extension assessment. Epistemic uncertainty is generally categorized into three types parameter, model, and completeness uncertainty. These are each discussed in the sections which follow.

3.5.1 Parametric Uncertainty Consistent with the ASME/ANS PRA Standard, quantitative parametric uncertainty analyses for both CDF and LERF were evaluated to determine if the point estimates calculated by the BB011b PRA model appropriately represent the mean. The uncertainty analysis did not have any F&Os from the most recent peer review related to parametric uncertainty. The results of the parametric uncertainty analyses confirm that the point estimate is a sufficient representation of the mean to represent the mean for the calculation of the changes in the risk metrics for this application. Please note that the Braidwood models BB011a (periodic update), BB011b (interim update) and BB01 1 b4 (ASM for RCP Shutdown Seals) all use the same plant-specific and generic data for the development of random failure probabilities and maintenance unavailabilities with the exception of the additions made to address the revised RCP Seal LOCA modeling. The parametric uncertainty conclusions are considered applicable to all three models.

The same conclusion is applicable to the Fire PRA model uncertainty which was evaluated in the peer review and demonstrated a small variation in uncertainty relative to the point estimate. This same conclusion is considered applicable to the SX pump out of service condition evaluated in this evaluation.

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3.5.2 Model Uncertainty An evaluation of model uncertainty also exists for the BB011b PRA model. The documentation of the BB01 1 b4 application specific model [Ref. 7] adds to this evaluation of model uncertainty via identification of key modeling assumptions related to the Shutdown Seals. These two sources of model uncertainty were reviewed for impacts on this specific application.

From the BB01 1 b model uncertainty evaluation, the assumed alignment of SX pumps was identified as a potential model uncertainty. Since this evaluation applies the actual SX pump alignments that will be in place for Unit 2, this model uncertainty is not expected to impact the evaluation.

From the BB01 1 b4 application-specific model documentation, the key model uncertainty is the modeling of the Shutdown Seals in the PRA model. Because a loss of SX can impact RCP seal cooling and lead to a challenge to the RCP seals, the modeling of the Shutdown Seals and associated human actions is identified as a potentially key uncertainty for this application. The modeling follows the guidance in PWROG-14001-P, Revision 1 [Ref. 20], and PWROG-14006-P, Revision O-B [Ref. 21]. In addition, the analysis incorporates the technical issues from Westinghouse Technical Bulletin TB 1 [Ref. 22], which addresses required actions to maintain the No. 2 RCP seal integrity following loss of all seal cooling scenarios. The logic model for the existing non-SIDS seals is based on WCAP-16141 [Ref. 23], also known as the WOG-2000 model.

Therefore, the seal modeling follows the latest industry guidance, so the results are evaluated to realistically represent the as-built, as-operated plant.

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Braidwood SX 2A CT Extension 3.5.3 Completeness Uncertainty Section 3.3 addresses those hazard groups not included in the FPIE PRA. With the exception of internal fire, the majority of those hazard groups were qualitatively determined to have negligible impact on plant risk for the SX 2A CT extension.

As discussed in Section 3.3, the Braidwood Fire PRA used in this SX 2A CT assessment is an interim implementation of NUREG/CR-6850 and other approved methodologies. The Braidwood FPRA is judged sufficiently complete to provide useful risk insights for applications, including this SX CT extension. The interim FPRA model was utilized to obtain quantitative risk metric results, but more importantly it helped to identify those fire areas that were subject to increased risk from fire during the extended CT for consideration of potential compensatory measures.

The FPRA model utilized for the assessment includes a full scope representation from the risk of fire for all Braidwood site fire areas. The selection of the global plant analysis boundary and the criteria for including/excluding plant areas are consistent with the current NUREG/CR-6850 guidance and methods. Therefore, the scope of areas included is sufficient for this application.

Fire scenario development in the FPRA model includes fixed ignition sources and transient sources, consistent with the guidance in NUREG/CR-6850. The potential for Main Control Room abandonment due to environmental conditions is also included in the model based on a CFAST model of the Main Control Room Complex spaces.

Specific consideration of hot gas layer (HGQ and Multi-Compartment Analysis (MCA) is also included in the FPRA.

The potential for multiple spurious operations is included in the fire model, based on an expert panel evaluation of the generic PWR MSO scenarios and consideration of plant specific scenarios.

Instrumentation has been explicitly included in the fire PRA.

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0 A A

• Based on the above information, there is no major form of completeness uncertainty that is judged to change the results of this assessment (i.e., ICCDP and ICLERP).

3.5.4 Uncertainty Analysis Conclusions As previously indicated, the uncertainty analysis addresses the three generally accepted forms of uncertainty - parameter, model, and completeness uncertainty. The conclusions from these assessments are as follows.

Parameter Uncertainty The parameter uncertainty assessment indicated that the use of the point estimate results directly for this assessment is acceptable.

Model Uncertainty The model uncertainty assessment highlighted the following attributes as related to uncertainty being important to address with potential compensatory measures:

• The alignment of SX is fixed for this evolution, eliminating uncertainty for alignment of the SX system.

• The model utilizes the latest industry guidance for modeling of the RCP safe shutdown seals, so results are expected to adequately represent the plant as-built, as-operated.

Completeness Uncertainty There is no major form of completeness uncertainty that would impact the results of this assessment. Although the fire model is an interim implementation of NUREG/CR-6850, it is essentially complete and appropriate for developing useful insights for the development of compensatory actions.

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Braidwood SX 2A CT Extension This analysis demonstrates with reasonable assurance that the proposed TS change is within the current risk acceptance in RG 1.177 for one-time changes. As shown in Table 3.4-1, there is significant margin between the calculated FPIE and FPRA risk metrics and the acceptance criteria considering the implementation of effective compensatory measures. The quantitative results combined with effective compensatory measures to maintain lower risk ensure the proposed TS change meets the intent of the ICCDP and ICLERP acceptance guidelines.

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The 2015 application specific PRA model (BB01 1 b4) is the most recent evaluation of the risk profile at BW for FPIE challenges. This model was developed to incorporate the installation of RCP safe shutdown seals at Braidwood. The BW PRA modeling is highly detailed, including a wide variety of initiating events, modeled systems, operator actions, and common cause events. The PRA model quantification process used for the BW PRA is based on the event tree / fault tree methodology, which is a well-known methodology in the industry.

Exelon employs a multi-faceted approach to establishing and maintaining the technical adequacy and plant fidelity of the PRA models for all operating Exelon nuclear generation sites. This approach includes both a proceduralized PRA maintenance and update process, and the use of self-assessments and independent peer reviews. The following information describes this approach as it applies to the BW PRA.

4.1 PRA QUALITY OVERVIEW The quality of the BW FPIE PRA is important in making risk-informed decisions. The importance of the PRA quality derives from NRC Policy Statements as implemented by RGs 1.174 and 1.177, rule-making and oversight processes. These can be briefly summarized as follows using the words of the NRC Policy Statement (1995):

1. "The use of PRA technology should be increased in all regulatory matters to the extent supported by the state-of-the-art ... and supports the NRCs traditional defense-in-depth philosophy."

2. "PRA ... should be used in regulatory matters ... to reduce unnecessary conservatism..."

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Braidwood SX 2A CT Extension

3. "PRA evaluations in support of regulatory decisions should be ... realistic ... and appropriate supporting data should be publicly available for reviews."

4. "The Commission's safety goals ... and subsidiary numerical objectives are to be used with appropriate consideration of uncertainties in making regulatory judgments..."

5. "Implementation of the [PRA] policy statement will improve the regulatory process in three ways:

- Foremost, through safety decision making enhanced by the use of PRA insights,

- Through more efficient use of agency resources, and

- Through a reduction in unnecessary burdens on licensees."

PRA quality is an essential aspect of risk-informed regulatory decision making. In this context, PRA quality can be interpreted to have five essential elements:

• Scope (Section 4.2): The scope (i.e., completeness) of the FPIE PRA.

The scope is interpreted to address the following aspects:

- Challenges to plant operation (Initiating Events):

Internal Events (including Internal Floods)

External Hazards Fires

- Plant Operational states:

Full Power M Wo" 6=M T 0 0 Shutdown 4-2 9/26/16

- The metrics used in the quantification:

Level 1 PRA - CIDF Level 2 PRA - LERF Level 3 PRA - Health Effects

• Fidelity (Section 4.3): The fidelity of the PRA to the as-built, as-operated plant.

• Standards (Section 4.4): ASME/ANS PRA Standard [Ref. 4 and Ref 5] as endorsed by the NRC in Regulatory Guide 1.200 [Ref. 1].

• Peer Review (Section 4.5):1 An independent PRA peer review provides a method to examine the PRA process by a group of experts. In some cases, a PRA self-assessment using the available PRA Standards endorsed by the NRC can be used to replace or supplement this peer review.

• Appropriate Quality (Section 4.6): The quality of the PRA needs to be commensurate with its application. In other words, the needed quality is defined by the application requirements.

4.2 SCOPE The BW PRA is a full power, internal events (FPIE) PRA that addresses both CIDF and LERF. The quantitative insights from the FPIE PRA are directly applicable to the SX 2A CT Extension PRA application. This scope is judged to be adequate to support the SX 2A CT PRA application. Consideration of other modes of operation is addressed in Section 1.4 and an evaluation of other potential hazard groups is included in Section 3.3 Because not all PRA standards are available to define the appropriate elements of PRA quality for all applications, the NRC has adopted a phased implementation approach.

This phased approach uses available PRA tools and their quantitative results where standards are available and endorsed by the NRC. Where standards are not yet 4-3 9/26/16

Braidwood SX 2A CT Extension available or endorsed, this approach uses qualitative insights or bounding approaches as needed.

The quality assessment performed in this section confirms the adequacy of the FPIE PRA. This quality assessment does not address the risk implications associated with low power or shutdown operation, nor does it address the quality assessment of external events. However, the results of the analysis for these other contributors have been used to obtain additional insights for potential compensatory measures and otherwise do not change the conclusions of the assessment.

Completion of a Fire PRA peer review within the past year and resolution of F&Os impacting risk quantification ensure the quality of the Fire PRA model used in this assessment. Completion of model documentation is expected by the end of this year.

The Exelon risk management process for maintaining and updating the PRA ensures that the PRA model remains an accurate reflection of the as-built and as-operated plants. This process is defined in the Exelon Risk Management program, which consists of a governing procedure (ER-AA-600, "Risk Management" [Ref. 15]) and subordinate implementation procedures. Exelon procedure ER-AA-600-1015, "FPIE PRA Model Update" [Ref. 16] delineates the responsibilities and guidelines for updating the full power internal events PRA models at all operating Exelon nuclear generation sites. The overall Exelon 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 operating 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:

4-4 9/26/16

Braidwood SX 2A CT Extension

• Design changes and procedure changes 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, Exelon risk management 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 PRA models for Exelon nuclear generation sites.

• 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 systems, structures, and components (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 a four year cycle; shorter intervals may be required if plant changes, procedure enhancements, or model changes result in significant risk metric changes.

4-5 9/26/16

.I EA ~~ IWA, "k The ASME/ANS PRA Standard provides the basis for assessing the adequacy of the Braidwood PRA as endorsed by the NRC in RG 1.200, Revision 1. The predecessor ts:

the ASME/ANS PRA Standard was NEI 00-02 which identified the critical internal events PRA elements and their attributes necessary for a quality PRA.

There are three principal ways of incorporating the necessary quality into the PRA in addition to the maintenance and update process. These are the following:

• A thorough and detailed investigation of open issues and the implementation of their resolution in the PRA.

• A PRA Peer Review to allow independent reviewers from outside to examine the model and documentation. The ASME/ANS PRA Standard specifies that a PRA Peer Review be performed on the PRA.

0 The use of the ASME/ANS PRA Standard to define the criteria to be used in establishing the quality of individual PRA elements.

There have been several assessments to support a conclusion that the Braidwood PRA model adequately meets the PRA standard such that it can be used to support risk applications in accordance with Regulatory Guide (RG) 1.200 Revision 2.

The Braidwood PRA model for internal events received a formal industry peer review in July 2013 [Ref. 9] against Addendum B of the ASME/ANS PRA Standard.

In the SE that was issued on February 24, 2011 from the NRC [Ref. 24] for implementation of the surveillance frequency control program (SFCP), which allows for relocation of surveillance test intervals to a licensee-controlled program, the following concluding statement was included regarding the quality of the Braidwood PRA model:

4-6 9/26/16

Braidwood SX 2A CT Extension Based on the licensee's assessment using the applicable PRA standard and RG 1.200, the level of PRA quality, combined with the proposed evaluation and disposition of gaps, is sufficient to support the evaluation of changes proposed to surveillance frequencies within the SFCP, and is consistent with Regulatory Position 2.3.1 of RG 1.177.

It should be noted that PRAs can be used in applications despite not meeting all of the Supporting Requirements of the ASME/ANS PRA Standard. This is well recognized by the NRC and is explicitly stated in the ASME/ANS PRA Standard.

4.6 APPROPRIATE PRA QUALITY The PRA is used within its limitations to augment the deterministic criteria for plant operation. This is confirmed by the PRA Peer Review and the PRA Self-Assessment.

As indicated previously, RG 1.200 also requires that additional information be provided as part of the LAR submittal to demonstrate the technical adequacy of the PRA model used for the risk assessment. Each of these items (plant changes not yet incorporated in to the PRA model, consistency with applicable PRA Standards, relevant peer review findings, and the identification of key assumptions) is discussed below.

4.6.1 Plant Changes Not Yet Incorporated into the PRA Model A PRA updating requirements evaluation (URE) is Exelon's PRA model update tracking database. These UREs are created for all issues that are identified with a potential to impact the PRA model. The URE database includes the identification of those plant changes that could impact the PRA model. A review of the current open items in the URE database associated with plant changes for Braidwood is summarized in Table 4-1 along with an assessment of the impact for this application.

The results of the assessment documented in Table 4-1 are that none of the plant changes have any measurable impact on the SX CT extension request.

4-7 9/26/16

Table 4-1 ImDact on the Braidwood PRA Model of Plant Chanaes Since the Last Model Update Impact on the URE Number Description Disposition Application?

ATWS Model revision in support of Power Uprate BB-0230 No ATWS is insignificant contributor to this application Project Measurement Uncertainty Recapture (MUR) Power BB-0815 Uprate - ECs 378382 & 378383 (BYR), 378380 & No MUR represents an insignificant impact to results 378381 (Bwd)

Change in Steps in 1(2)BwEP-3, Steam Generator BB-0903 No No impact on HRA for this application Tube Rupture HELB Removal of VX dampers and implement HELB BB-0963 No Current model is conservative dampers EC 388397 & EC 388442 Byron/Braidwood VA Recirculation Duct added to BB-0986 No No impact on this application AUX BLD BY ECs 391071(2), BW ECs 388703,389632 New Revision of 1(2)BWOA ELEC-3 Rev 102, analyze BB-1002 No No impact on HRA for this application.

for HRA impact CC pump discharge check valves replaced. The current valve is a Velon swing check with welded ends, while the new valve is a Crane nozzle check BB-1006 No Negligible impact to this application with flanged ends. The PRA uses plant-specific data for these check valves, which will no longer apply once the valves are replaced.

Annunciator window added in the control room BB-1012 No No impact to this application, model is conservative which provides alarm for the VA monitoring system.

BB-1014 ALTERNATE SX SUPPLY TO 1/2SX04P PUMP SUCTION No Model is conservative 4-8 9/26/16

Braidwood SX 2A CT Extension Table 4-1 Impact on the Braidwood PRA Model of Plant Chanaes Since the Last Model Update Impact on the URE Number Description Disposition Application?

Review new subtask in SWOP RH-6 for HRA effect, BWOP RH-6 has a new subtask added that alters the BB-1083 No Current HEP value is conservative HEP for 1RH-NR-SGTRHSYOA, OPERATORS FAIL TO ESTABLISH NORMAL RH SHUTDOWN COOLING Review new step in BWEP ES-1.3 for HRA effect. Step 12 was added to Rev 202 of ES-1.3 to specifically direct the operators to verify that CLR has been established. The HEP calculation associated with CLR alignment (1S1-HPR ---- HSYOA) currently uses a BB-1084 No Current HEP value is conservative generic execution recovery as the verification procedure step did not exist at the time the calculation was developed. Applying step 12 as an execution recovery would lower the HEP to the mid E-3 range BB-1087 Modifications for FLEX No No impact to FPIE or FPRA, results are conservative Multiple EC - MCR Fire Modifications, modifications are being made to respond to the NRC discovery of a BB-1094 No No impact to FPIE, FPRA is conservative circuit design deficiency for PORV response to a design basis MCR fire.

Relief valve along CV charging line introduces Maximum flow diversion is 20 gpm, sensitivity run in BB-1097 potential flow diversion during injection or as a LOCA No MAAPS showing negligible impact to success criteria for pathway after downstream check valve failure. ECCS injection 4-9 9/26/16

Braidwood SX 2A CT Extension Table 4-1 Impact on the Braidwood PRA Model of Plant Changes Since the Last Model Update URE Number Description Impact on the Disposition Application?

EH reservoir control block assembly remove and No impact to FPIE, modification has not yet been BB-1099 replace NO installed (A2R19)

Fire mitigating actions to remove PORV control BB-1100 No No impact to FPIE, FPRA is conservative power fuses 4-10 9/26/16

As indicated above, a formal peer review against Addendum A of the ASME/ANS PRA Standard was performed in July 2013. The results of that review lead to the identification of the Braidwood PRA as not meeting Capability Category 11 for a small number of Supporting Requirements (SRs) listed below. These SRs are summarized in Table 4-2 along with an evaluation of their impact on the base model and this application.

The FPIE PRA model of record for this evaluation is Revision BB01 1 b4 as documented in BB-ASM-002, Application Specific Model Notebook RCP Shutdown Seals [Ref. 7].

This application-specific model to incorporate the RCP Shutdown Seals is based on BB-PRA-014, Quantification Notebook, Revision BB011b [Ref. 8]. A peer review of the BB011b model was performed in July 2013 to assess the technical adequacy of the internal events and internal flooding models. The Peer Review report is documented in LTR-RAM-11-13-067-NP [Ref. 9].

LTR-RAM-11-13-067-NP identified six supporting requirements that were evaluated as not being met. In addition there were 10 supporting requirements that were assessed as being at Capability Category I. Table 4-2 provides a listing of the Not Met and Category I supporting requirements (SR) and an assessment of the impact on the evaluation presented here.

Table 4-2:

Bvron / Braidwood Not Met and Canabilitv Cateaory I Sunnortina Requirements Supporting Capa bility Evaluation Impact Requirement Category These SRs are associated with the counting of failures and demands in the development of failure probabilities. Changes in r%c DA-05 Not Met random failure probabilities are not expected to impact the calculation of the change in risk due to the component unavailability in this application.

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Braidwood SX 2A CT Extension Table 4-2:

Bvron I Braidwood Not Met and Capability Cateaory I Supporting Requirements Supporting Capability Evaluation Impact Requirement Catego!j These SRs are associated with the counting of failures and demands in the development of failure probabilities. Changes in DA-C6 Not Met random failure probabilities are not expected to impact the calculation of the change in risk due to the component unavailability in this application.

This SIR deals with not including other than at-power events in the development of Initiating Events. A subsequent review has IE-A7 Not Met confirmed that no initiating events are missing. This SIR does not impact the results of the evaluation.

The assessment of this SIR identifies that certain operator action timing information and the potential impact on cues was not IFQU-A6 Not Met addressed in the flooding assessment. The operator actions associated with the components addressed by this assessment are not impacted by the internal flooding actions.

The assessment of this SIR is related to the need for additional discussion of the inputs into the flood scenario development. Lack IFSN-133 Not Met of this additional discussion does not impact the results provided in this evaluation.

This assessment for this SIR relates to a need for increased discussion of limitations in the LERF analysis that could impact LE-G5 Not Met applications. There are no specific limitations in the LERF analysis that impact this application.

This SIR relates to having a structured review with Operations and training personnel to confirm the interpretation of the procedure is HR-E3 cc I consistent with training expectations and plant use. This is not expected to have an impact on the results of this evaluation.

This SIR relates to having a structured review with Operations and training personnel to confirm the interpretation of the procedure is HR-E4 cc I consistent with training expectations and plant use. This is not expected to have an impact on the results of this evaluation.

This SIR addresses interviews of plant personnel to determine if potential initiating events are missing from the PRA. The components being evaluated here do not have an impact on the IE-A8 cc I modeled Initiating Events. A subsequent review has confirmed that no initiating events are missing. This SIR does not impact the results of the evaluation.

This SIR addresses review of plant-specific precursors to determine if potential initiating events are missing from the PRA. The components being evaluated here do not have an impact on the IE-A9 cc I modeled Initiating Events. A subsequent review has confirmed that no initiating events are missing. This SIR does not impact the results of the evaluation.

The assessment of this SIR relates to a lack of inclusion of Braidwood specific OE in the development of internal flooding IFEV-A6 cc I frequencies. The inclusion of this OE into the internal flooding results is not expected to have a significant impact on the model results or the results for this evaluation.

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ipil 111pliplill 117W l

Table 4-2:

Bvron I Braidwood Not Met and Caoabilitv Category I Su000rtina Reauirements Supporting Capa bility Evaluation Impact Requirement Category The assessment of this SR relates to a lack of inclusion of HELB contributions to internal flooding. The inclusion of qualitative HELB lFSN-A6 cc I considerations to the internal flooding results is not expected to have a significant impact on the model results or the results for this evaluation.

These SRs relate to the identification and documentation of significant contributors to LERF. The inclusion of this LE-F1 cc I documentation does not impact the ability of the LERF model to provide accurate results. This SR does not impact the evaluation results.

These SRs relate to the identification and documentation of significant contributors to LERF The inclusion of this LE-G3 cc I documentation does not impact the ability of the LERF model to provide accurate results. This SR does not impact the evaluation results.

This SR relates to providing a comparison of quantification results to those from other plants. Since this assessment provides an QU-D4 cc I evaluation on the delta risk impact, this SIR will not have an impact on the evaluation results.

This SR relates to the assessment and development of beyond 24-hour mission times. The inclusion of these beyond 24-hour mission SC-A5 cc I times is not expected to have a significant impact on the PRA model or on the results from this evaluation.

In summary, there were 16 Supporting Requirements that were judged to be "Not Met" or only meeting Capability Category I. As indicated in Table 4-2, these gaps have very limited or no impact on the model results and also have very limited or no impact on the SX CT extension request.

4.6.3 Relevant Peer Review Findings RG 1.200, Revision 2 provides the following guidance with respect to meeting the ASME/ANS PRA Standard requirements and hence to the quality of a PRA model:

If the requirement has been met for the majority of the systems or parameter estimates, and the few examples can be put down to mistakes or oversights, the requirement would be considered to be met. If, however, there is a systematic failure to address the requirement (e.g.

component boundaries have not been defined anywhere), then the requirement has not been complied with. In either case, the examples of 4-13 9/26/16

Braidwood SX 2A CT Extension noncompliance are to be (1) rectified or demonstrated not to be relevant to the application, and (2) documented.

The results of the July 2013 peer review are also used to identify the relevant peer review findings for the PRA model used for this assessment. These findings are summarized in Table 4-3 along with an assessment of the impact for the base model development. The associated F&Os with the "Not Met" or "Capability Category I" issues identified above are not repeated in this assessment. Table 4-3, therefore, only includes those "Findings" that are associated with SRs that were otherwise assigned to be at least Capability Category 11 from the peer review consistent with the RG-1.200 Table 4-3:

Byron / Braidwood Finding F&Os from CC-11 or Better Supporting Reauirements Finding Evaluation Impact F&O The SIR requires that each system be analyzed systematically. There are systems or trains noted as screened and not causing a Plant Trip where there is not a IE-A5-01 detailed explanation or justification for no plant trip occurring. A subsequent review has confirmed that no initiating events are missing. This Finding does not impact the results of the evaluation.

Initiating Event SLOCA frequency does not include RCP Seal LOCA Frequency IE-C1-01 input. SLOCA frequency was re-examined to include random RCP seal failures in the current model used for this application.

Potential failure of containment sump suction screens due to debris clogging is not AS-133-01 represented in the fault tree. This is not expected to be a significant effect since dominant sequences often already have failure of recirculation for other reasons.

Additional justification based on analysis other than MAAP is needed to justify large LOCA success criterion for RCS inventory control early. The Large LOCA success I SC-B1-01 criteria is acceptable, but needs additional justification. Therefore, this Finding does not impact the results of the evaluation.

For loss of CC or SX, assumption of 30 minutes to trip RCP is not justifiable. Along SC-132-01 with the new RCP Shutdown Seal modeling, the timing for operator actions has been updated. within the current model used for this application.

The failure probability of S18806 due to failure to remain open may rise above the 1% threshold per SY-A15 or even dominate other failure modes in the system.

Quantification of the system gate 2S1-PUMPS-HPI has a value of 9.6E-4, while the SY-Al 5-01 probability of 518806 spuriously closing would be 4.45E-08

• 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> = 1.07E-6, which is much less than the 1% threshold, so the impact of this Finding would be minimal.

HVAC dependency is not included for the switchgear and battery rooms. Modeling SY-1312-01 of the necessary dependencies is included in the current model.

SY-C2-01 Inconsistent references to battery life. This is a documentation issue only.

QU-D1-01 Documentation of cutset reviews. This is a documentation issue only.

4-14 9/26/16

Table 4-3:

Byron I Braidwood Finding F&Os from CC-11 or Better Supporting Requirements Finding Evaluation Impact F&O Review of Braidwood procedures for pre-initiator HRA in addition to Byron. Review HR-A1-01 of Braidwood-specific procedures showed no differences that would affect the PRA model, so this Finding has no impact on this evaluation.

It was found that there is only one bleed and feed cooling execution HEP developed, which uses thermal/hydraulic analysis timing based on non-loss of main feed water initiators. During an evaluation of the impact of this Finding, the highest HR-G4-01 cutset containing Loss of MFW and the HEP is in the form of one of its JHEPs (JHEP64) and occurs in the 1E-8 range at BW. Therefore, slight increases in the HEP or related JHEPs are not expected to have a significant effect on the model or this application.

Testing events for several systems are stated to satisfy the requirements for inclusion of testing in unavailability of the systems, yet the data are then stated to rely on values from the IPE study and not updated. Since the peer review, data DA-C3-01 was reviewed and sources were identified for most test & maintenance basic events in the current model. Inaccurate information was removed from Data Notebook regarding IPE as a source of current data.

Additional justification for and documentation of the final data used for the failure or DA-C3-02 unavailability data calculations is needed. This is a documentation issue only.

The use of plant-specific data from Byron for Braidwood or from Braidwood for Byron does not meet the requirement for the use of plant-specific data and should DA-C3-04 be avoided. Since any differences among these plants is minimal among those events using opposite-site data, the impact of this Finding is minimal.

SR-DA-D7 requires that if common cause events deleted from common cause population of estimate formula due to non-applicability events in the total population also have to be screened and deleted if non-applicable. It was noted that one DA-D7-01 common cause event excluded from the common cause group was not excluded from the total population. Changes in the one CCF term identified by the review team do not have a large overall effect on model results.

Document the processes used for data parameter definition, grouping, and DA-E2-01 collection. This is a documentation issue only.

IFSO-A1-01 Improve identification of flood sources. This is a documentation issue only.

Flood source from within SX valve pit. An analysis was performed to show that the bolts securing the deck plates for the SX valve pits can indeed withstand the IFSO-A1-02 hydrostatic pressure of water from the lake. In addition, a quantitative what-if analysis shows that the impact is expected to be in the E-8 range since it is only a few feet of piping. The impact of this Finding is minimal.

Review screening analysis for internal flooding. The screening criterion for miscellaneous buildings outside the power block was corrected to be associated IFSO-A3-01 with Supporting Requirement IFSN-Al2 in Table B.2-1. This is a documentation issue only.

Effect of plant-specific maintenance practices on internal flooding. Though this may only be a documentation issue, in absence of supporting information for not considering maintenance-induced flooding, an increase in internal flood frequencies IFSO-A4-01 of approximately 1.45 could apply. This Finding was determined to have a potential I impact on both the base model and the SX CT model. A sensitivity analysis was ertormed in Section 3.2.3 4-15 9/26/16

Braidwood SX 2A CT Extension

• 4-3:

Byron / Braidwood Finding F&Os from CC-11 or Better Supporting Finding Evaluation Impact F&O Correct flooding initiating events for critical factor. Without the plant criticality factor, the results for each calculation are slightly conservative, and therefore have IFEV-A5-01 a minor impact on the model. This may work to offset some of the impact from IFSO-A4-01.

Correct inconsistent flood I E frequencies to consistently account for piping at a data IFEV-A5-02 threshold such as 6". The impact of these corrections is minor, and shifts frequen ies to less severe scenarios, so the impact of this Finding is minor.

Address indirect effects of flooding, such as jet impingement. A subsequent analysis provides a discussion of the jet impingement forces imposed on insulated pipes surrounded by lagging due to a pipe rupture. The analysis concluded that the IFQU-A9-01 impingement forces for moderate energy water sources are insufficient to perforate the aluminum lagging and threaten damage to equipment located at a distance from the water source. Therefore, this Finding has no impact on the application.

Documentation for the Braidwood Station is not provided in the Internal Flooding IFQU-B1-01 Notebook. This is a documentation issue only. -1 In summary, there were 24 additional peer review findings not already encompassed within the entries in Table 4-2. As indicated in Table 4-3, however, most of these remaining open items have no or very limited impact on the model results and also should have no or very limited impact on most applications of the model. The one issue related to Internal Flooding frequencies (IFS(-A4-01) is addressed with a simple sensitivity assessment performed in Section 3.2.3.

4.6.4 Identification of Key Assumptions Key assumption identification is addressed by the Uncertainty Assessment in Section 3.5, specifically the Model Uncertainty in Section 3.5.2 and Completeness Uncertainty in Section 3.5.3. The key assumptions that introduce uncertainties for this application are summarized in Section 3.5.4.

4-16 9/26/16

4.6.5 Fire PRA Peer Review Results and F&Os The FPRA peer review for Braidwood was performed by the PWR Owners Group in May 2015. The results of this peer review are summarized in Tables 4-4 (Not Met and Capability Category I Supporting Requirements) and 4-5 (Finding F&Os from CC-11 or Better Supporting Requirements) below. As noted in the evaluation of the supporting requirements and F&Os below, the issues identified have been addressed in the current quantification to the extent that they impact the quantification results.

Table 4-4:

Braidwood Not Met and Capability Category I Supporting Requirements F&O QUANT SIR DESCRIPTION ACTION TAKEN NO. IMPACT APPENDIX C OF THE BW CABLE SELECTION Based on information provided there is lack of NOTEBOOK (BW-PRA-details to meet capability category 11/111 as the 021.03) WAS UPDATED TO only statement is in Section 3.8 of the PROVIDE SPECIFIC

'Braidwood Fire PRA Cable Selection Notebook REFERENCES AND (BW-PRA-021.03), Rev 0': "The BW Fire PRA ADDITIONAL DETAILS reviewed the electrical coordination PERTAINING TO THE CS-131 16-4 calculations for applicability to the Fire PRA. FALSE BREAKER COORDINATION These were reviewed for each of the credited OF THE CREDITED POWER power supplies in the model." This did not SUPPLIES. THE INCLUSION provide Analysis or Identified any additional OF THIS ADDITIONAL requirements only stated that it was reviewed.

INFORMATION DID NOT CHANGE THE RESULTS OF (This F&O originated from SR CS-131)

THE BREAKER COORDINATION.

HLR-QU-D7 requires review of importance of components and basic events to determine Documentation upgrade FQ-E1 19-9 that they make logical sense. FALSE only. No impact on quantification.

(This F&O originated from SR FQ-El) 4-17 9/26/16

Table 4-4:

Braidwood Not Met and Capability Category I Supporting Requirements F&O QUANT SIR DESCRIPTION ACTION TAKEN NO. IMPACT Document the relative contribution of contributors to LERF.

Documentation upgrade (This F&O originated from SR FQ-Fl)

FQ-E1 19-8 FALSE only. No impact on quantification.

• • • REMOVE THIS LINK TO SR FQ-El***

Nonsignificant accident cutsets Documentation upgrade FQ-E1 19-17 FALSE only. No impact on (This F&O originated from SR FQ-El) quantification.

Perform a quantification evaluation for the contribution of contributors to LERF for the Documentation upgrade FQ-E1 19-16 FPRA. FALSE only. No impact on quantification.

(This F&O originated from SR FQ-El)

An internal event HFE is being used in the fire PRA in addition with the fire event HFE See HRA F&Os. Updated FQ-E1 19-10 version. TRUE HFEs as required.

(This F&O originated from SR FQ-El)

Limitations on knowledge of severe accident phenomenology as well as level-2 PRA Documentation upgrade modeling to capture severe accident FQ-F1 19-14 FALSE only. No impact on progression has not been provided.

quantification.

(This F&O originated from SR FQ-Fl)

There is no discussion of asymmetries from Documentation upgrade QU-F2(l).

FQ-F1 19-1 FALSE only. No impact on quantification.

(This F&O originated from SR FQ-F1) 4-18 9/26/16

Braidwood SX 2A CT Extension Table 4-4:

Braidwood Not Met and Capability Category I Supporting Requirements F&O QUANT SR DESCRIPTION ACTION TAKEN NO. IMPACT There is no document of the importance measures for Braidwood Unit 2 CDF/LERF from Documentation upgrade FQ-F1 19-11 QU-F2(j). FALSE only. No impact on quantification.

(This F&O originated from SR FQ-Fl)

The definition of significance is not being used.

There is no discussion of documentation of significance definitions (basic event, cutsets, Documentation upgrade FQ-F1 19-13 and accident sequences) as required by QU- FALSE only. No impact on F6. quantification.

(This F&O originated from SR FQ-Fl)

Document the process used to identify plant damage states and accident progression Documentation upgrade FQ-F1 19-15 contributors. FALSE only. No impact on quantification.

(This F&O originated from SR FQ-Fl)

There is no discussion of the quantification Documentation upgrade process limitations as required in QU-F5.

FQ-F1 19-12 FALSE only. No impact on quantification.

(This F&O originated from SR FQ-Fl)

Probabilities of suppression unavailability are Data used is based on credited, but not plant-specific information. NUREG/CR-6850. A review FSS-D7 14-6 FALSE of plant data confirmed no (This F&O originated from SR FSS-D7) outlier behavior.

The Human Reliability Analysis notebook Instrumentation are Appendix A identifies the cues explicitly modeled in the (instrumentation and other indications) Byron and Braidwood fire required for significant operator actions used model. Table 3-4 in the HRA HRA 15-6 FALSE in the Fire PRA. The instrumentation required notebooks summarizes the is not consistent with the indicators in instrumentation for Appendix A Table A-1 of the Equipment operator actions in the Fire Selection notebook. PRA model.

4-19 9/26/16

Table 4-4:

Braidwood Not Met and Capability Category I Supporting Requirements F&O QUANT SR DESCRIPTION ACTION TAKEN NO. IMPACT Plant fire response procedures were not reviewed to identify possible new fire-specific Braidwood fire procedures safe shutdown actions for the FPRA, including have been reviewed and HRA-A2 18-2 FALSE MCR abandonment. documented in Section 2.2.3 of the HRA Notebooks (This F&O originated from SR HRA-A2)

Errors of commission, due to a single spurious fire-induced failure, were not reviewed with Operator interviews have Braidwood operators.

been performed at Braidwood and scenarios The procedures and sequence of events were have been discussed with not reviewed in detail with Braidwood HRA-A4 18-4 FALSE operators in detail. The operations and training personel to confirm outcome of the operator that the interpretation of the procedures was interviews is presented in consistent with plant observations and Appendix D of the HRA training procedures.

Notebook.

(This F&O originated from SR HRA-A4)

With respect to documenting the FPRA, the Fire PRA HRA Notebook (BW-PRA-021.09) contains inconsistencies and discrepancies, including:

Documentation, HRAC HEP Only Unit 1 is documented, although some values, and CAFTA model differences between the two units exist in the HEP values have been FPRA database.

HRA-El 18-7 FALSE reviewed and inconsistencies have been Some combination HEPs in Table 4.1 are resolved. Correct values are different from those in the FPRA.

used in the CAFTA model.

Some documented screening HEPs are not in the FPRA.

(This F&O originated from SR HRA-El) 4-20 9/26/16

Braidwood SX 2A CT Extension Table 4-4:

Braidwood Not Met and Capability Category I Supporting Requirements

&OO QUANT SR DESCRIPTION ACTION TAKEN R . IMPACT Plant-specific fire events were not compared No outlier events were against the fire events used to develop the identified, therefore, the IGN-A4 20-4 generic ignition frequencies in NUREG-2169. TRUE use of NUREG-2169 is considered appropriate.

(This F&O originated from SR IGN-A4)

The instrumentation and indications required for cues identified in the HRA notebook are See HRA F&Os. All required PRM-1311 15-14 not included in the Fire PRA model. TRUE cues are addressed in the FPRA model.

(This F&O originated from SR PRM-1311)

BW-PRA-021.05, Fire PRA Plant Response Model notebook Rev 0, Section 3.1.8 and Appendix B. Appendix B documents the review of the containment paths and the basis for screening or including the individual Documentation upgrade pathways. Additional pathways were PRM-1315 15-15 FALSE only. No impact on identified that should be included in the Fire quantification.

PRA model. However, these pathways appear to not be included under gate 1(2)-CONT-ISOLATION.

(This F&O originated from SR PRM-1315)

Internal events F&O IE-A5-01 does not appear to have been addressed for the Fire PRA Documentation upgrade PRM-132 15-9 analysis. FALSE only. No impact on quantification.

(This F&O originated from SR PRM-132)

Internal events F&O SC-A5-01 does not appear to have been addressed for the Fire PRA Documentation upgrade PRM-132 15-12 analysis. FALSE only. No impact on quantification.

(This F&O originated from SR PRM-B2) 4-21 9/26/16

Table 4-4:

Braidwood Not Met and Capability Category I Supporting Requirements F&O QUAN T SR DESCRIPTION ACTION TAKEN NO. IM PACT Internal events F&O HR-G4-01 does not appear to have been addressed for the Fire Documentation upgrade PRIVI-132 15-11 PRA analysis. FALSE only. No impact on quantification.

(This F&O originated from SR PRM-B2)

Internal events F&O SC-132-01 does not appear to have been addressed for the Fire PRA Documentation upgrade PRIVI-B2 15-10 analysis. FALSE only. No impact on quantification.

(This F&O originated from SR PRIVI-132)

Some anomalies were observed in the database that was used for the uncertainty Anomalies were addressed analysis which was different from that used in and do not impact UNC-Al 18-12 FALSE the FPRA. conclusions of uncertainty analysis.

(This F&O originated from SR UNC-Al)

Some documented assumptions were not identified as potential sources of uncertainty and characterized to identify the effect on the Documentation upgrade UNC-A2 18-13 PRA model and to permit the potential impact FALSE only. No impact on on the results to be understood. quantification.

(This F&O originated from SR UNC-A2)

Based on similarities between the two plants, some elements of the Braidwood FPRA reflect the application of work performed for Byron, Documentation upgrade UNC-A2 18-14 without identifying this as a potential source FALSE only. No impact on of uncertainty. quantification.

(This F&O originated from SR UNC-A2) 4-22 9/26/16

Braidwood SX 2A CT Extension Table Braidwood Finding F&Os from CC-11 or Better Supporting Requirements SR F&O DESCRIPTION ACTION TAKEN IMPACT A REVIEW AGAINST THE Per review of Braidwood Fire PRA Detailed REQUIRED CABLES IDENTIFIED IN Circuit Analysis Notebook (BWPRA-021.03.01), BW(BY)-PRA-021.03.01 WAS Rev 0" and "Braidwood Fire PRA Cable PERFORMED. ANY CABLES THAT Selection Notebook (BW-PRA-021.03), Rev 0" WERE MISSING FROM THE CABLE CS-A1 16-1 TRUE the captured cables did not include all TO COMPONENT DATA WERE required cables. ADDED INTO THE FIRE PRA MODEL. THE CABLES THAT WERE (This F&O originated from SR C IDENTIFIED TO BE MISSING SHOULD HAVE BEEN INCLUDE THE BE MAPPING FOR BOTH BY AND BW HAS BEEN UPDATED TO Appendix A Table A-1 contains some REFLECT THE NECESSARY CODING dispositions, bases, and notes that are AND DISPOSITIONS CHANGES.

ES-A4 15-2 inconsistent, vague, or inaccurate. TRUE THE NEW BE MAPPING CAN BE FOUND IN APP A OF THE This F&O originated from SR ES-A4

()

EQUIPMENT SELECTION NOTEBOOK (BW(BY)-PRA-021.02.

Appendix C of the BW Equipment In Appendix C Table C-1, component OSX007, Selection notebook (BW-PRA-the disposition does not appear correct. 021.02) has been updated to ES-131 15-4 FALSE reflect the modeling of this (This F&O originated from SR ES-131) component. The disposition was changed to "modeled in PRA".

4-23 9/26/16

Table Braidwood * *F&Os fromor Better Requirements F&OO QUANT SR DESCRIPTION TAKEN IMPACT THE DAMPER MAPPING HAS Appendix C Table C-1 contains several fire BEEN UPDTED TO REFLECT THE damper components contain notes that PRA MODELING. APP C OF THE indicate that there may be open items for ES-61 15-5 FALSE EQUIPMENT SELECTION these components.

NOTEBOOK HAS BEEN UPDATED TO REFLECT THESE CHANGES (This F&O originated from SR ES-131)

(BW-PRA-021.02).

THE SECTION REFERENCES HAVE There are several incorrect references within BEEN UPDATED TO REFLECT THE the report that should be corrected. APPROPRIATE SECTION OR ES-D1 15-8 FALSE APPENDIX NUMBERS. THESE (This F&O originated from SR ES-D1) UPDATES WERE MADE IN BW-PRA-021.02.

The use of modules in the fire risk INCLUDED STATEMENT IN quantification was not documented for SECTION 3.2 OF BW(BY)-PRA-FQ-131 19-2 interpretation. FALSE 021.11 CLARIFYING THAT MODULES ARE NOT USED IN THE (This F&O originated from SR FQ-131) FIRE PRA QUANTIFICATION.

4-24 9/26/16

Braidwood SX 2A CT Extension Table 4-5:

Braidwood Finding F&Os from CC-11 or Better Supporting Requirements F&O QUAN T SR DESCRIPTION ACTION TAKEN NO. IM PACT The criteria of the truncation limit for CDF/LERF of being less than 5% for the final change is not met and there is a potential for Updated quantification has FQ-131 19-3 accident sequences being inadvertently TRUE confirmed convergence.

eliminated.

(This F&O originated from SR FQ-131)

Identfication and correction of mutually exclusive events in the results has not been Documentation upgrade only.

FQ-131 19-4 defined. FALSE No impact on quantification.

(This F&O originated from SR FQ-Fl)

When locating transient fires in PAUs, plant personnel indicated that transient scenarios A sensitivity analysis was were only placed in areas with targets such as performed placing transient conduits or cable trays. Transient fires were ignition sources adjacent to fixed FSS- not placed in areas where the only targets 14-3 FALSE ignition sources (panels). The Al were PRA equipment. The risk contribution impact on risk was small using a from the areas excluded from transient conservative estimate of the ignition sources could possibly be significant.

impact.

(This F&O originated from SR FSS-Al) 4-25 9/26/16

Braidwood SX 2A CT Extension Table 4-5:

Braidwood Finding F&Os from CC-11 or Better Supporting Requirements F&O QUAN T SIR DESCRIPTION ACTION TAKEN NO. IM PACT THE MCA INTERACTIONS HAVE BEEN REVIEWED TO IDENTIFY WHERE THE SCREENING CRITERIA WAS APPLIED INCORRECTLY. IN The screening criteria was not applied FSS- THOSE CASES THE INTERACTIONS 14-4 correctly in one multi compartment scenario. TRUE G3 WERE ADDED BACK TO THE LIST (This F&O originated from SR FSS-G3)

OF INTERACTIONS TO BE ANALYZED. THE UPDATED INTERACTION LIST IS PROVIDED IN APP A OF THE The methodology for assigning ignition frequency and targets for cable fires due to FSS- Documentation upgrade only.

14-5 welding is not described in the scenario FALSE H1 No impact on quantification.

development notebook.

(This F&O originated from SR FSS-H1)

This is a generic F&O related to the overall FSS- documentation. Documentation upgrade only.

14-7 FALSE H1 No impact on quantification.

(This F&O originated from SR FSS-H1)

The Human Reliability Analysis notebook Instrumentation are explicitly Appendix A identifies the cues - modeled in the Byron and (instrumentation and other indications) Braidwood fire model. Table 3-4 required for significant operator actions used in the HRA notebooks HRA 15-6 FALSE in the Fire PRA. The instrumentation required summarizes the instrumentation is not consistent with the indicators in for operator actions in the Fire Appendix A Table A-1 of the Equipment PRA model. Document Selection notebook. clarification only.

4-26 9/26/16

Table 4-6:

Braidwood Finding F&Os from CC-11 or Better Supporting Requirements F&O QUANT SR DESCRIPTION ACTION TAKEN NO. IMPACT 1RC-EB-ATWSHSYOA-F and 1RT-RX-ATWSHRBOA-F are set to 1.0 Some FPIE actions were carried forward into in the fire model.

the FPRA with no clear determination of HRA- relevance and validity within the context of 18-1 TRUE Al the fire scenarios.

Internal events single HFEs and joint HFEs have been removed (This F&O originated from SR HRA-Al) from the fire model and only fire versions are credited.

The identification of the key human response actions was not always based on Braidwood-HRA- Documentation upgrade only.

18-3 specific procedures. FALSE Al No impact on quantification.

(This F&O originated from SR HRA-Al)

Risk significant HFEs set to screening values in the internal events model have been The FPRA credits some operator actions which reviewed and specific fire HFEs were not modified through screening or HRA- have been developed.

18-9 detailed analysis for applicable fire effects. TRUE 131 The assessment column in Table (This F&O originated from SR HRA-131) 5-1 of the Fire HRA Notebook identifies the new HFEs developed for the fire model.

4-27 9/26/16

Braidwood SX 2A CT Extension Table 4-5:

Braidwood Finding F&Os from CC-11 or Better Supporting Requirements F&O QUAN T SIR DESCRIPTION ACTION TAKEN NO. IM PACT There is no explicit listing of the particular instruments and indications that are credited to provide the required cues and no Table 3-4 of the HRA Notebook HRA- correlation of those cues to the instruments provides a list of instruments 18-11 FALSE B3 listed on the T-sheets as reliable for particular credited for each operator fires. action.

(This F&O originated from SR HRA-133) 1AF-START--HPMOA-F is based on the 660003 run that has feed water unavailable.

The development of some HFEs credited support systems that are inconsistent with HRA-18-5 assumptions for the FPRA. TRUE B3 1FW-FRH1---HSGOA-F assumed that IVIFW is not available. It (This F&O originated from SR HRA-133) discusses the possibility to start AFW to avoid CD after 55 min but this is not credited.

The fire HRA approach is The definitions of some specific high-level consistent with the internal tasks do not describe cross-unit performance events HRA approach. Cross unit factors (e.g., availability of cues, HFE applicability is described in HRA-18-6 communication, manpower) for credited train- FALSE FPIE HRA Notebook 5.2.1. Text at B3 level recoveries. the end of paragraph 2.1.3 of the fire HRA notebook has been (This F&O originated from SR HRA-133) added to explain cross-unit impact.

4-28 9/26/16

Table 4-6:

Braidwood Finding F&Os from CC-11 or Better Supporting Requirements F&O ACTION TAKEN SIR NO.

DESCRIPTION IMPACT IMPA Detailed analysis was not performed for some HFEs that were dispositioned in Table 5-1 as All risk significant HFEs have been "Not risk significant in FPRA model" but for revised and detailed analysis HRA-18-10 which the screening HEP met one or more TRUE performed for HFE that had C1 criteria for being considered risk significant. screening values in the internal events model.

(This F&O originated from SR HRA-C1)

Some HFEs are included in the FPRA with no All HFEs have been revised and basis to establish the relevant fire-related HRA- HFEs that are credited in the 18-8 effects for the associated HEPs. TRUE C1 model consider fire-related effects.

(This F&O originated from SR HRA-C1)

THE COMPONENT COUNT IN 11.4A-1 AND 11.4A-2 HAVE BEEN UPDATED TO THE CORRECT COUNTS BASED ON THE EQUIPMENT LOCATED IN THE During the Peer Review walkdown, three PAU. THIS IS DOCUMENTED IN rooms were checked and found to have errors APPENDIX A OF BW-PRA-021.06 IGN- in ignition source counting: 11.4A-1, 11.4A-2, AND APPENDIX A OF BW-PRA-20-1 TRUE A7 11.6-0. 021.10. A REVIEW WAS ALSO COMPLETED ON OTHER IGNITION (This F&O originated from SR IGN-A7) SOURCES TO IDENTIFY IF THIS ISSUE WAS FOUND ELSEWHERE.

DISCREPANCEIS IDENTIFIED WERE CORRECTED AS NECESSARY. THIS REVIEW WAS DONE FOR BOTH BY AND BW.

4-29 9/26/16

Braidwood SX 2A CT Extension Table 4-6:

Braidwood Finding F&Os from CC-11 or Better Supporting Requirements F&O QUAN T SIR DESCRIPTION ACTION TAKEN NO. IM PACT DISCUSSION ADDED IN SECTION Figure 4.1 does not indicate any engineering 3.1 FOR MANHOLES, DRAWINGS manholes as PAUs.

PP-B6 20-5 TRUE ADDED AS APPENDIX C, OF NOTEBOOK PP (BY/BW)-PRA-(This F&O originated from SR PP-136) 021.01).

ALL BUILDINGS HAVE BEEN IDENTIFIED WITHIN THE PLANT A review of Figure 4-1 of PP shows some LAYOUT DRAWINGS. THIS possible buildings or structures that are not INCLUDES ALL STRUCTURES THAT PP-Cl 20-2 identified. TRUE MAY BE RELATED TO TANKS OR ELECTRICAL TOWERS. SEE FIGURE (This F&O originated from SR PP-Cl) 4-1, FIGURE 4-2, AND TABLE 4-1 OF THE BW PLANT PARTITIONING NOTEBOOK (BW-PRA-021.01).

Appendix B of the PP notebook gives the features of the partitioning elements, but it does not provide details of walkdowns. Plant Documentation upgrade only.

PP-C3 20-3 personnel say they have the walkdowns FALSE No impact on quantification.

documented on an IPAD.

(This F&O originated from SR PP-C3)

THE LOGIC FOR MS017 HAS BEEN UPDATED TO REFLECT THAT A Logic for spurious opening of pressurizer SINGLE PORV SPURIOUSLY PORVs (MSO-17) should be modified.

OPENING CAN LEAD TO A SMALL PRM- LOCA. THIS CHANGE HAS BEEN 15-17 Logic for spurious opening of steam generator TRUE B4 DOCUMENTED IN APPENDIX A OF relief valves (MSO-22) should be modified.

BW-PRA-021.05.

(This F&O originated from SR PRIVI-134)

SEE RESPONSE TO F&O PRM-134 (21-7) FOR MS022 CHANGES.

4-30 9/26/16

Table Braidwood

• i F&Os from or Better Requirements SR DESCRIPTION ACTION TAKEN NO IMUAN T M

THE LOGIC MODEL HAS BEEN UPDATED TO ENSURE THAT ALL SLBO LOCATIONS INLCUDE THE Logic for spurious opening of steam generator MS022 LOGIC AS AN INPUT.

relief valves (MSO-22) should be modified.

PRM- MS022 WAS REMOVED FROM 21-7 (Appendix A change #23G) TRUE B4 ALL SBLI LOCATIONS AS IT IS MORE APPROPRIATE TO MODEL (This F&O originated from SR PRM-134)

THIS UNDER SBLO. THIS CHANGE IS DOCUMENTED IN APPENDIX A OF BY(BW)-PRA-021 PRM notebook Section 3.1.5 states that no changes are made to the logic, but changes PRM- Documentation upgrade only.

15-13 are made to the logic for fire-specific failures. FALSE C1 No impact on quantification.

(This F&O originated from SR PRM-C1)

THE QUALITATIVE SCREENING CRITERIA HAS BEEN REVISED TO Section 3.3 of the Qualitative Screening ADDRESS THE CONCERN OF Notebook (BW-PRA-021.04) describes the SCREENING PADS WITHOUT QLS- application of all the screening criteria to all ADEQUATE CRITERIA MET.

17-1 FALSE A3 PAUs. Appendix A lists the screening bases. APPENDIX A AND APPENDIX B OF Appendix B lists all PAUs from the Plant THE QUALITATIVE SCREENING Partitioning Notebook (BW-PRA-021.01) and NOTEBOOK (BW-PRA-021.04)

HAVE BEEN UPDATED APPROPRIATELY.

4-31 9/26/16

Braidwood SX 2A CT Extension Table 4-6:

Braidwood Finding F&Os from CC-11 or Better Supporting Requirements F&O QUAN T SR DESCRIPTION ACTION TAKEN NO. IM PACT In Section 4.1.1, various sections of the IPEEE are quoted skipping potential issues that were Section 4.1.1 of BW(BY)-PRA-identified. A conclusion is made based entirely 021.13 has been updated to on the IPEEE work with no evaluations or SF-Al 17-2 FALSE include additional references that discussions about the past potential issues indicate the results and findings with respect to the current FPRA work.

from seismic walkdowns.

(This F&O originated from SR SF-Al)

Section 4.1.2 addressed fire water and carbon dioxide suppression systems but not Halon.

IPEEE Section 4.9.3.1.2 was quoted as support Documentation upgrade only.

SF-A2 17-3 FALSE for Spurious Actuation of a Fire Water System. No impact on quantification.

(This F&O originated from SR SF-A2)

Section 4.1.3 describes the three suppression systems: water, carbon dioxide, and Halon.

However, the potential for common cause failure is only described for water. In addition, Section 4.1.3 of BW(BY)-PRA-a "not insignificant potential exists" for a 021.13 has been updated to SF-A3 17-4 common cause failure of the fire protection FALSE include a complete discussion of pumps due to loss of the Lake Screen House is the potential for common cause mentioned in the FPRA. But there is no failure of all suppression systems.

assessment of the risk to the plant.

(This F&O originated from SR SF-A3)

In Section 4.1.4, the concluding paragraphs reference the conclusions of Sections 4.1.1 Section 4.1.4 of BW-PRA-012.13 and 4.1.2 but not 4.1.3 to justify the has been updated to reference SF-A4 17-5 effectiveness of the plant seismic response FALSE section 4.1.3. This provides the procedures. discussion of common cause between the different systems.

(This F&O originated from SR SF-A4) 4-32 9/26/16

Table 4-6:

Braidwood Finding F&Os from CC-11 or Better Supporting Requirements F&O QUAN T SIR DESCRIPTION ACTION TAKEN NO. IM PACT The BW Seismic-Fire evaluation is documented in BW-PRA-021.13. Some documentation deficiencies have been identified in Sections 3.1 and 4.1. Section 3.1.1 doesn't discuss how seismic fire ignition sources were identified; all that is described is the Seismic PRA generally.

Section 3.1.2 discusses how fire suppression was assessed but not fire detection. Section 3.1.3 incorrectly states that the fire suppression system is made up of a single fire water system. There are carbon dioxide and automatic Halon systems. Also, there is no discussion about how the evaluation was performed.

Section 4.1.1 quotes various sections from the IPEEE with no explanation about why these specific sections are pertinent to seismic fire ignition source assessments. IPEEE Section 3.4.7.2 is the most applicable section from IPEEE. But the excerpt missed the discussion on seismically induced fires. Moreover, this Revised Section 4 to add in excerpt is repeated in Sec. 4.1.1 with no discussions related to findings discussion about the potential issues that from the IPEEE. In most cases the SF-131 17-6 were described in the IPEEE. Section 4.9.3.1 FALSE direct quotes have been from the IPEEE is quoted with "Response" removed, and specific excerpts from sections 4.9.3.1.1, 4.9.3.1.2, and conclusions have been made 4.9.3.1.3 (Seismically-Induced Fires, Seismic based on the IPEEE.

Actuation of Fire Suppression Systems, and Seismic Degradation of Fire Suppression Systems, respectively). It is unclear why fire suppression is discussed in this section. Also, the excerpts skipped the discussion of issues that were described in the IPEEE. It's unknown whether those issues are significant for this current Fire PRA assessment. In Section 4.1.2, the paragraph after the reference to IPEEE 4.9.3.1.2 in "Smoke Detector Actuation during a Seismic Event" is out of place; this paragraph belongs in the discussion on Spurious Actuation. In Section 4.1.4 there is the typo "liaison" that should be "Haise." Section could be improved upon by ad*Mmore detail 9/26116 on the firefighting equipment, estimated response times, and any outside agency r#tr%r%r%r+ ;; lv% n.4*4;+;^r% +kr% f%r%^v%;r%rT

Braidwood SX 2A CT Extension Table 4-6:

Braidwood Finding F&Os from CC-11 or Better Supporting Requirements F&O QUAN T SIR DESCRIPTION ACTION TAKEN NO. IM PACT Based on information provided there is lack of APPENDIX C OF THE BW CABLE details to meet capability category 11/111 as the SELECTION NOTEBOOK (BW-PRA-only statement is in Section 3.8 of the 021.03) WAS UPDATED TO

'Braidwood Fire PRA Cable Selection Notebook PROVIDE SPECIFIC REFERENCES (BW-PRA-021.03), Rev 0': "The BW Fire PRA AND ADDITIONAL DETAILS reviewed the electrical coordination PERTAINING TO THE BREAKER CS-131 164 calculations for applicability to the Fire PRA. FALSE COORDINATION OF THE These were reviewed for each of the credited CREDITED POWER SUPPLIES. THE power supplies in the model." This did not INCLUSION OF THIS ADDITIONAL provide Analysis or Identified any additional INFORMATION DID NOT CHANGE requirements only stated that it was reviewed.

THE RESULTS OF THE BREAKER COORDINATION.

(This F&O originated from SR CS-131) J The Braidwood PRA maintenance and update processes and technical capability evaluations provide a robust basis for concluding that the PRA is suitable for use in risk-informed licensing actions, specifically in support of the requested extended CT for the SX system.

Previously identified gaps to specific requirements in the ASMEANS PRA Standard have been reviewed to determine which gaps might merit application-specific sensitivity studies in the presentation of the application results. One sensitivity study was performed for internal flooding initiators per Finding F&O IFSO-A4-01 (documented in Section 3.2.3) and was found to not have significant impact on this SX CT extension request.

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Braidwood SX 2A CT Extension 5.0

SUMMARY

AND CONCLUSIONS 5.1 SCOPE INVESTIGATED This analysis evaluates the acceptability, from a risk perspective, of a change to the Braidwood TS for the 2A Essential Service Water Train (SX 2A) for a one-time increase of the CT from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br /> when SX 2A is inoperable.

The analysis examines a range of risk contributors as shown in Table 5-1.

Table 6-1

SUMMARY

OF RISK INSIGHTS FOR SX 2A CT EXTENSION RISK CONTRIBUTOR APPROACH INSIGHTS Internal Events Quantify ICCDP & ICLERP for

• Base risk well within planned configuration acceptance guidelines

• ICCDP < 1 E-6

• Compensatory measures

• ICLERP < 1 E-7 further reduce risk If exceeded compare to acceptance guidelines with risk management actions implemented to reduce sources of risk

• ICCDP < 1 E-5

• ICLERP < 1 E-6 Internal Fire Qualitatively and quantitatively 0 ICCDP and ICLERP within evaluated: acceptance guidelines with

• Identify fire scenarios risk management actions to impacted by reduce risk sources.

configuration

• Internal events

• Estimate fire risk impacts compensatory measures due to configuration and apply to fire scenarios quantify ICCDP and 0 Additional Fire-related ICLERP compensatory measures

• Identify compensatory identified measures Seismic Qualitatively evaluated.

• Seismic risk impacts negligible 5-1 9/26/16

Table 6-1

SUMMARY

OF RISK INSIGHTS FOR SX 2A CT EXTENSION RISK CONTRIBUTOR APPROACH INSIGHTS High Winds Qualitatively evaluated. 0 High winds risk impacts negligible 0 High winds risk reduced with compensatory measures for internal events and fire Other External Hazards Qualitatively evaluate each 0 Other External Event risks hazard based on the BW IPEEE were found to be negligible and a re-examination for the contributors specific configuration with SX 2A inoperable.

Overall At-Power Risks Quantify ICCDP & ICLERP for 0 Quantitative guidelines for planned configuration with normal work controls normal work controls challenged, but acceptable

• ICCDP < 1 E-6 with risk management actions implemented.

ICLERP < 1 E-7 If exceeded compare to acceptance guidelines with risk management actions implemented to reduce sources of risk

• ICCDP < 1E-5

• ICLERP < 1 E-6 The PRA quality has been assessed and determined to be adequate for this risk application, as follows:

• Scope - The BW PRA modeling is highly detailed, including a wide variety of initiating events, modeled systems, operator actions, and common cause events. The PRA has the necessary scope to appropriately assess the pertinent risk contributors.

• Fidelitv The BW PRA model (BB01 1 b4) is the most recent evaluation of the risk profile at BW for FPIE challenges. The PRA reflects the as-built, as-operated plant.

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Braidwood SX 2A CT Extension

• Standards The PRA has been reviewed against the ASME/ANS PRA Standard and the PRA elements are shown to have the necessary attributes to assess risk for this application.

• Peer Review - The PRA has received a peer review. Based on addressing the peer review results and subsequent gap analyses to the current standards, the PRA is found to have the necessary attributes to assess risk for this application.

• Appropriate Quality The PRA quality is found to be appropriate to assess risk for this application.

5.3 QUANTITATIVE RESULTS VS. ACCEPTANCE GUIDELINES As shown in Table 3.4-1 this analysis demonstrates with reasonable assurance that the proposed TS change is within the current risk acceptance guidelines in RG 1.177 for one-time changes. This combined with effective compensatory measures to maintain lower risk ensures that the TS change meets the intent of the ICCDP and ICLERP acceptance guidelines.

5.4 CONCLUSION

S This analysis demonstrates the acceptability, from a risk perspective, of a change to the BW TS for the 2A Essential Service Water Pump to increase the CT from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br /> when SX 2A is inoperable.

A PRA technical adequacy evaluation was also performed consistent with the requirements of ASME/ANS PRA Standard and RG 1.200, Revision 2. Additionally, a review of model uncertainty was performed with this application. None of these identified sources of uncertainty were significant enough to change the conclusions from the risk assessment results presented here.

5-3 9/26/16

However, the assessment of risk from internal events and internal fires did help to identify the following actions as important compensatory measures that will help to reduce the overall risk during the performance of the extended CT:

5.4.1 Compensatory Measures

• There will be no elective maintenance work on the remaining SX pumps (1A, 1 B, 2B) during the SX 2A extended CT. Additionally, this equipment should be protected for this one-time outage. This supports the maintenance assumptions in the analysis.

• There will be no elective maintenance work on the emergency diesel generators (1A, 1 B, 2A, 213) during the SX 2A extended CT.

Additionally, this equipment should be protected for this one-time outage. This supports the maintenance assumptions in the analysis and also supports mitigation of a loss of offsite power during the maintenance window.

• There will be no elective maintenance work on the Unit 2 auxiliary feed pumps (2A, 2B). This equipment should be protected for the one-time outage. This supports the maintenance assumptions in the analysis.

• There will be no elective maintenance on the SX 16A/B or SX 27A/B on either unit due to interlocks that could prevent use of the remaining SX pumps. This supports the maintenance assumptions in this analysis.

• There will be no elective maintenance on the 211, 212, 213, or 214 instrument busses or their associated inverters and transformers.

Additionally, this equipment should be protected for the one-time outage. This supports the maintenance assumptions in the analysis.

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Braidwood SX 2A CT Extension

• There will be no elective maintenance on the motor-driven feedwater pump, 2FW02P.

• There will be no elective maintenance activities on the Unit 2 SATs.

• The extended weather forecast will be examined to ensure severe weather conditions are not predicted prior to entry into the CT. In the event of an unforeseen severe weather condition due to rapidly changing conditions, such as severe high winds, a briefing with crew operators will be performed to reinforce operator actions and responses in the event of a loss of offsite power.

• Fire Risk Management Actions (RMAs) applicable for SX 2A from BW-CRM-1 15 Rev 1 [Ref. 11 ] will be completed per OP-AA-201-012-1001 "OPERATIONS ON-LINE FIRE RISK MANAGEMENT" [Ref. 18].

(These actions protect against fire impacting key redundant equipment)

• Operations will hold briefings on the following actions prior to entry into the extended CT.

o Tripping the RCPs to preclude damage to the Shutdown Seals.

o Restoring main feedwater and throttling the SX007 valves as needed.

o Opening the SX crosstie valves, tripping the RCPs to preclude damage to the Shutdown Seals, refilling the diesel-driven auxiliary feedwater pump fuel oil day tank, and operation of the AF005 valves in event of fire as shown in Table 3.3-4.

• Shift briefs and pre job walkdowns to reduce and manage transient combustibles prior to entrance into the extended CT will be used to alert the staff about the increased sensitivity to fires in the following fire zones during the extended SX 2A outage window. Additionally, any hot work activities in the following fire zones will be prohibited during 5-5 9/26/16

Braidwood SX 2A CT Extension the time within the extended SX 2A CT. The listed fire zones were identified based on risk significance in the FPRA results (generally zones with Div 2 equipment that impact SX). (The purpose of these walkdowns is to reduce the likelihood of fires in these zones by limiting transient combustibles, ensuring transients, if required to be present, are located away from fixed ignition sources and eliminating or isolating potential transient ignition sources, e.g., energized temporary equipment and associated cables)

Fire Zone(l) Fire Zone Description 5.1-2 Division 22 ESF Switchgear Room 5.1-1 Division 12 ESF Switchgear Room 3.2-0 Auxiliary Building El. 439'-0" 11.4-0 Auxiliary Building General Area, El. 383' Division 22 Containment Electrical Penetration Area, 11.6-2 El. 426' 11.2C-2 Containment Spray Pump 2B Room 11.IB-0 Unit 2 Auxiliary Building Basement El. 330' 18.10D-2 Unit Auxiliary Transformer 241-2 18.10E-2 System Auxiliary Transformers 242-1/242-2 (1) For larger fire zones walkdowns may be focused on specific fire sensitive areas within the larger firezones. Walkdowns are judged as not being required for areas with continuous operator occupation (e.g. MCR). Fire Risk Management Actions (RMAs) where they occur may address the need for walkdowns in some of these areas. ALARA principles apply when reviewing radiological areas such as RHR.

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Braidwood SX 2A CT Extension

6.0 REFERENCES

[1] Regulatory Guide 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk Informed Activities," Revision 2, March 2009.

[2] Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-informed Decisions on Plant-Specific Changes to the Licensing Basis,"

Revision 2, May 2011.

[3] Regulatory Guide 1.177, "An Approach for Plant-Specific, Risk-informed Decisionmaking: Technical Specifications," Revision 1, May 2011.

[4] "Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications," ASME RA-S-2002, Addenda RA-Sa-2003, and Addenda RA-Sb-2005, December 2005.

[5] "Addenda to RA-S-2008, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications," ASME/ANS RA-Sa-2009, February 2009.

[6] Regulatory Guide 1.182, "Assessing and Managing Risk Before Maintenance Activities at Nuclear Power Plants," May 2000.

[7] BB-ASM-002, Application Specific Model Notebook RCP Shutdown Seals, Rev.

0, September 2015.

[8] BB PRA-014, "Byron/Braidwood PRA Quantification Notebook," Rev. BB011b, September 2012.

[9] LTR-RAM-11-13-067-NP, Rev. 0, "Byron / Braidwood Nuclear Plants RG 1.200 Internal Events and Internal Flooding PRA Peer Review Report",November 12, 2013.

[10] Exelon Risk Management Team, BW Fire PRA Fire Risk Quantification Notebook, BW-PRA-021.1 1, Rev. 0, May 2015.

[11] Exelon Risk Management Team, Development of Risk Management Actions for the Inclusion of Fire Insights into Braidwood Configuration Risk Management Program, BW-CRM-1 15, Rev. 1, February, 2014.

[12] BB-PRA-021.021.01, Revision 0, "Level 1 Seismic Quantification Notebook",

December 2013 6-1 9/26/16

I -.Tn,

• TM_I*

[13] "Individual Plant Examination of External Events For Severe Accident Vulnerabilities Submittal Report", June 1997, Braidwood Station

[14] NUREG-1407, "Procedural and Submittal Guidance for the Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities June 1991

[15] ER-AA-600, Risk Management, Revision 7.

[17] USNRC, "Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-informed Decisionmaking," NUREG-1855, March 2009.

[18] Exelon, OP-AA-201-012-1001, "OPERATIONS ON-LINE FIRE RISK MANAGEMENT", Revision 1.

[19] Exelon, ER-AA-600-1046, "Risk Metrics NOED and LAR", Revision 6.

[20] PWROG-14001-P, Revision 1, PRA Model for the Generation III Westinghouse Shutdown Seal, Risk Management Committee, PA-RMSC-0499R2, July 2014.

[21] PWROG-14006-P, Revision O-B, Implementation Guide for the Generation III Westinghouse Shutdown Seal, Risk Management Committee, PA-RMSC-0499R1 , December 2014.

[22] Westinghouse Technical Bulletin TB-15-1, Reactor Coolant System Temperature and Pressure Limits for the No. 2 Reactor Coolant Pump Seal, March 3, 2015.

[23] WCAP-16141, "RCP Seal Leakage PRA Model Implementation Guidelines for Westinghouse PWRs," August 2003.

[24] USNRC, "Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Amendment No 165 to Facility Operating License No. NPF-72 and Amendment No. 165 to Facility Operating License No. NPF-77," February 24, 2011 6-2 9/26/16