ML100840752

Request for Amendment to Technical Specification 3.1.7, Standby Liquid Control (SLC) System.

ML100840752
Person / Time
Site:	Peach Bottom
Issue date:	03/24/2010
From:	Cowan P Exelon Generation Co, Exelon Nuclear
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML100840752 (107)

Text

Exelon Nuclear www.exeloncorp.com 200 Exelon Way Nuclear Kennett Square, PA 19348 10 CFR 50.90 March 24, 2010 u.s. Nuclear Regulatory Commission ATIN: Document Control Desk Washington, D.C. 20555-0001 Peach Bottom Atomic Power Station, Units 2 and 3 Renewed Facility Operating License Nos. DPR-44 and DPR-56 NRC Docket Nos. 50-277 and 50-278

Subject:

Request for Amendment to Technical Specification 3.1.7, IIStandby Liquid Control (SLC) System ll

References:

1) Letter from M. A. Satorius (U. S. NRC) to C. M. Crane (Exelon Generation Company, LLC), IINotice of Enforcement Discretion for Exelon Generation Company LLC Regarding Quad Cities Nuclear Power Station, Unit 1 (NOED 06-3-01 ), II dated October 18, 2006

2) Letter from M. A. Satorius (U. S. NRC) to C. M. Crane (Exelon Generation Company, LLC), IINotice of Enforcement Discretion for Exelon Generation Company LLC Regarding Dresden Nuclear Power Station, Unit 2 (NOED 07-3-01; TAC MD4044),1I dated January 24,2007 In accordance with 10 CFR 50.90, IIApplication for amendment of license or construction permit, II Exelon Generation Company, LLC (EGC) requests an amendment to Appendix A, Technical Specifications (TS) of Renewed Facility Operating License Nos. DPR-44 and DPR-56 for Peach Bottom Atomic Power Station (PBAPS), Units 2 and 3.

The proposed amendment revises Technical Specification (TS) 3.1.7, IIStandby Liquid Control (SLC) System,1I to extend the Completion Time (CT) for Condition C (i.e., IITwo SLC subsystems inoperable for reasons other than Condition A. II ) from eight hours to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

In References 1 and 2, the NRC exercised discretion to not enforce compliance with the actions required in TS 3.1.7, Condition B for Quad Cities Nuclear Power Station and Dresden Nuclear Power Station, respectively. These notices of enforcement discretion (NOEDs) provided a 72-hour extension to the eight-hour CT specified in Required Action B.1. This extension enabled each site to avoid a TS-required shutdown while implementing repair and restoration activities for the SLC System.

EGC has utilized the guidance in Regulatory Guide 1.174, IIAn Approach for Using Probabilistic Risk Assessment In Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, II to develop the technical basis of this license amendment request. The EGC analysis demonstrates, with reasonable assurance, that the proposed license amendment satisfies the risk acceptance guidelines in Regulatory Guide 1.174 and Regulatory Guide 1.177, "An Approach for Plant-Specific, Risk-Informed Decision-making: Technical Specifications." The proposed license amendment meets the intent of "very small" risk increases consistent with the NRC's Safety Goal Policy Statement.

u.s. Nuclear Regulatory Commission March 24, 2010 Page 2 EGC Probabilistic Risk Assessment (PRA) maintenance, update processes, and technical capability evaluations provide a robust basis for concluding that the PRA is suitable for use in risk-informed licensing actions. Additionally, a PRA technical adequacy evaluation was performed consistent with the requirements of Regulatory Guide 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision 1.

This request is subdivided as follows:

• Attachment 1 provides a description and evaluation of the proposed changes.

• Attachment 2 provides the marked-up PBAPS TS page with the proposed change indicated.

• Attachment 3 provides the marked-up PBAPS TS Bases pages with the proposed changes indicated. This attachment is provided for information only.

• Attachment 4 provides the risk assessment that supports the proposed TS change for PBAPS (Le., RM Documentation PB-LAR-05, Revision 0).

The proposed amendment has been reviewed and approved by the PBAPS Plant Operations Review Committee and the Nuclear Safety Review Board in accordance with the requirements of the Quality Assurance Program and ECG procedures. EGC requests approval of the proposed amendment by March 24, 2011, with implementation within 60 days of issuance.

In accordance with 10 CFR 50.91, "Notice for public comment, EGC is notifying the State of II Pennsylvania of this application for amendment by transmitting a copy of this letter and its attachments to the designated State Official.

There are no regulatory commitments contained within this letter. If you have any questions or require additional information, please contact Mr. Doug Walker at (610) 765-5952.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 24th day of March 2010.

Respectfully,

~~i ~1dJw~

Pamela B. Cowan Director - Licensing and Regulatory Affairs Attachment 1: Evaluation of Proposed Amendment Attachment 2: Proposed Markup of PBAPS Technical Specification 3.1.7 Attachment 3: Proposed Markup of PBAPS Technical Specification Bases B3.1 .7 Attachment 4: RM Documentation No. PB-LAR-05, Revision 0 cc: USNRC Region I, Regional Administrator USNRC Project Manager, NRR - Peach Bottom Atomic Power Station USNRC Senior Resident Inspector - Peach Bottom Atomic Power Station S. T. Gray, State of Maryland R. R. Janati, Bureau of Radiation Protection, Commonwealth of Pennsylvania

ATTACHMENT 1 Evaluation of Proposed Amendment 1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION

4.0 REGULATORY EVALUATION

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

6.0 REFERENCES

Page 1 of 19

ATTACHMENT 1 Evaluation of Proposed Amendment 1.0

SUMMARY

DESCRIPTION In accordance with 10 CFR 50.90, "Application for amendment of license or construction permit," Exelon Generation Company, LLC (EGC) requests an amendment to Facility Operating License Nos. DPR-44 and DPR-56, for Peach Bottom Atomic Power Station (PBAPS) Units 2 and 3. The proposed amendment changes Technical Specification (TS) 3.1.7, "Standby Liquid Control (SLC) System," by extending the Completion Time (CT) for two inoperable SLC subsystems from eight hours to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

PBAPS TS LCO 3.1.7 requires the operability of two SLC subsystems when the reactor is in Modes 1, 2, and 3. In Modes 1 and 2, the SLC System satisfies the requirements of 10 CFR 50.62, "Requirements for reduction of risk from anticipated transients without scram (ATWS) events for light-water-cooled nuclear power plants," and Atomic Energy Commission (AEC)

Criterion 27, 28, 29 and 30. Also in Modes 1 and 2 and additionally in Mode 3, the SLC System helps ensure that offsite doses remain within the limits of 10 CFR 50.67, "Accident source term" following a loss-of-coolant accident (LOCA) involving significant fission product releases.

TS 3.1.7, Condition C, and the associated Required Action C.1 address the inoperability of both SLC subsystems. Specifically, Required Action C.1 requires restoration of one SLC subsystem to operable status, with a CT of eight hours. If Required Action C.1 cannot be satisfied within the CT, Condition 0 and associated Required Actions 0.1 and 0.2 require the reactor to be in Mode 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and Mode 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

The current CT for Required Action C.1 is based on the low probability of a design basis accident or transient occurring concurrent with the failure of the control rods to shut down the reactor. Consistent with this current basis, the proposed TS CT change is based upon a risk-informed assessment that evaluates the probability and consequences of transients, accidents, and severe accidents, including the design basis accident and transients occurring concurrent with control rod insertion failure.

EGC has utilized the guidance 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," Revision 1, to develop the risk assessment for this proposed change. The EGC assessment demonstrates, with reasonable assurance, that the proposed license amendment satisfies the risk acceptance guidelines in RG 1.174 and RG 1.177, "An Approach for Plant-Specific, Risk-Informed Decision-making: Technical Specifications" Revision O. The proposed license amendment meets the intent of very small" risk increases consistent with the NRC's Safety Goal Policy Statement.

In addition to evaluating the risk impact, EGC has evaluated the proposed change to determine whether the impact of the change is consistent with the intent of the defense-in-depth philosophy and the principle that sufficient safety margins are maintained (i.e., consistent with the requirements of RG 1.177, Section C, "Regulatory Position," paragraph 2.2, "Traditional Engineering Considerations").

EGC has also determined that the EGC Probabilistic Risk Assessment (PRA) maintenance, update processes, and technical capability evaluations provide a robust basis for concluding that the EGC PRA is suitable for use in risk-informed licensing actions. EGC conducted a PRA technical adequacy evaluation consistent with the requirements of Regulatory Guide 1.200, "An Page 1 of 19

ATTACHMENT 1 Evaluation of Proposed Amendment Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities,lI Revision 1.

2.0 DETAILED DESCRIPTION The proposed amendment revises the CT for PBAPS TS 3.1.7, Required Action C.1 from eight hours to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

PBAPS TS 3.1.7 Action (Units 2 & 3)

ACTIONS (continued)

CONDITION REQU I RED ACTI ON COM PLETI ON TI ME M~

C. Two SLC subsystems C.1 Restore one SLC  !~ hours inoperable for reasons subsystem to OPERABLE ~-~

other than status. J2J Condition A.

The SLC System is designed to provide the capability of bringing the reactor, at any time in a fuel cycle, from full power and minimum control rod inventory to a subcritical condition with the reactor in the most reactive, xenon free state without taking credit for control rod movement.

The SLC System satisfies the requirements of 10 CFR 50.62, IIRequirements for reduction of risk from anticipated transients without scram (ATWS) events for light-water-cooled nuclear power plants. II PBAPS TS LCO 3.1.7 requires the operability of two SLC subsystems when the reactor is in Modes 1,2, and 3. TS 3.1.7, Condition C, and the associated Required Action C.1 address the inoperability of both SLC subsystems. Specifically, Required Action C.1 requires restoration of one SLC subsystem to operable status, with a CT of eight hours. If Required Action C.1 cannot be satisfied within the CT, Condition D and associated Required Actions D.1 and D.2 require the reactor to be in Mode 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and Mode 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

In October 2006 and January 2007, EGC requested Notices of Enforcement Discretion (NOEDs) for Quad Cities Nuclear Power Station (QCNPS), Unit 1, and Dresden Nuclear Power Station (DNPS), Unit 2, respectively, to allow sufficient time for the repair of minor SLC System tank leaks. The NRC granted these NOEDs, allowing an additional 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to the original 8-hour CT required for the dual-train inoperability of the SLC Systems (References 1 and 2).

The purpose of this proposed change is to adopt a permanent, risk-informed CT extension, thus minimizing the potential for thermal transients associated with placing PBAPS Units 2 and 3, in Cold Shutdown (Le., Mode 4). The integrity of the reactor vessel and other components of the primary system of a nuclear plant can be adversely affected by the number of thermal transients that they are subjected to during their lifetime. As each additional thermal transient can affect this integrity, it is prudent to avoid such transients.

Page 2 of 19

ATTACHMENT 1 Evaluation of Proposed Amendment

3.0 TECHNICAL EVALUATION

The proposed change is consistent with the principle that adequate defense-in-depth is maintained, that sufficient safety margins are maintained, and that increases in risk are 'very small" and meet the acceptance guidelines in RG 1.74, RG 1.77, and the NRC's Safety Goal Policy Statement. This consistency is described below, as well as in Attachment 4.

3.1 System Description The SLC System consists of a boron solution storage tank, two positive displacement pumps, two explosive valves that are provided in parallel for redundancy, and associated piping and valves used to transfer borated water from the storage tank to the reactor pressure vessel (RPV). The borated solution is discharged near the bottom of the core shroud, where it then mixes with the cooling water rising through the core. A smaller tank containing demineralized water is provided for testing purposes.

The performance objective of the SLC System is to provide an alternative to the highly reliable control rod drive (CRD) scram system for reactivity control. The SLC System provides the capability of bringing the reactor from full power to a cold, xenon free shutdown condition assuming that none of the withdrawn control rods can be inserted.

This is accomplished by injecting a sufficient quantity of sodium pentaborate solution into the reactor core.

To meet this objective, it is necessary to inject a quantity of boron that produces a reactivity change equivalent to a concentration of 660 ppm of natural boron for PBAPS.

The shutdown analysis assumes a sodium pentaborate solution with enriched boron is used. A 9.82% enriched sodium pentaborate solution is also used to satisfy the requirements of 10 CFR 50.62.

The SLC System is manually initiated from the main control room, as directed by the emergency operating procedures, if and when the operator determines the reactor cannot be shut down, or kept shut down, with the control rods. The SLC System is used in the event that not enough control rods can be inserted to accomplish shutdown and cooldown in the normal manner.

The SLC System is also required to be operable in Modes 1, 2, and 3 to ensure that offsite doses remain within 10 CFR 50.67, "Accident source term, limits following a II LOCA involving significant fission product releases. The SLC System is credited for maintaining pH balance in the suppression pool at or above 7 following a LOCA to ensure that iodine will be retained in the suppression pool water. Additional redundancy for the control of suppression pool pH control following a LOCA is established by the PBAPS Emergency Operating Procedures (EOPs). The EOPs describe the actions and criteria for alternate means of manual addition of boron should the SLC System fail to operate as designed. Therefore, the proposed SLC CT extension will not impact the ability of PBAPS to comply with the requirements of 10 CFR 50.67.

Page 3 of 19

ATTACHMENT 1 Evaluation of Proposed Amendment 3.2 Defense-in-Depth The control rods are the primary reactivity control system for the reactors at PBAPS. In conjunction with the Reactor Protection System (RPS), the control rods provide the means for reliable control of reactivity changes to ensure that fuel design limits are not exceeded. Operability of the control rods is governed by TS 3.1.3, "Control Rod OPERABILITY," and the control rods are demonstrated operable by the performance of TS Surveillance Requirements 3.1.3.1 through 3.1.3.5. These specifications assure that the insertion capability of the control rods is maintained in the event of an accident or transient, thus meeting the assumptions used in the safety analysis.

Scram reliability is the object of a number of features in the system, including:

Two sources of scram energy (accumulator and reactor pressure) that complement each other for each control rod drive whenever the reactor is operating.

Each control rod drive mechanism is equipped with specific scram valves and pilot valves so that only one control rod drive can be affected by a scram valve failure.

Under scram conditions the control rod drive mechanism develops adequate force, providing a large margin to overcome possible friction.

The scram system is designed so that the scram signal overrides all other operating signals.

The scram valves fail open on loss of either air or electrical power. Hence, failure of the valves' air system or electric system will produce, rather than prevent, a scram.

The Alternate Rod Insertion (ARI) system provides a separate set of backup scram valves in the event that the normal scram path cannot be initiated by RPS. The ARI system provides a means of control rod insertion which is motivated mechanically by the normal hydraulic control units and control rod drives but which utilizes totally independent and diverse logic from RPS. The ARI system energizes backup valves which vent the scram air header. This feature minimizes the impact of individual scram valve and/or pilot valve failures.

The scram discharge volume high level instrument provides a signal to the RPS logic to initiate an automatic scram prior to scram discharge volume inventory reaching a level which would inhibit a reactor scram.

As noted above, operability of the trip function of the control rods is demonstrated by specific Surveillance Requirements. For the control rod scram function to fail when a valid signal is sent, a diverse number of failures would have to occur in order in prevent the scram valves from opening. Also as noted above, the ARI system would be available as a separate means for reactor shutdown in the event that the normal scram path cannot be initiated by the RPS.

In addition to the ARI system, the ATWS Recirculating Pump Trip (RPT) provides an additional means for rapid power reduction in the event that the normal scram path cannot be initiated by RPS. In this case, the automatic trip of the reactor recirculation pumps causes a quick reduction in core flow which increases core void generation.

These increased voids introduce negative reactivity thus decreasing the reactor power.

The quick power reduction helps bring reactor pressure, neutron flux, and fuel surface Page 4 of 19

ATTACHMENT 1 Evaluation of Proposed Amendment heat flux down rapidly enough to limit the peak pressure, clad oxidation, and peak fuel enthalpy.

The proposed change to the SLC System CT does not affect the redundancy, independence, or diversity of the RPS and ARI systems, or the RPT. These systems and instrumentation remain operable to mitigate the consequences of any previously analyzed accident. In addition to the TS 3.1.3 requirements for control rod operability, the EGC Work Management and Maintenance Rule (Le., 10 CFR 50.65(a)(4)) programs provide controls and assessments to minimize the probability of simultaneous outages of redundant trains and ensure system reliability. The proposed SLC System CT extension does not involve any change to plant equipment or system design functions.

This proposed TS CT extension does not change the design function of the SLC System and does not affect the system's ability to perform its design function. As such, the proposed change complies with the defense-in-depth principles described in RG 1.174, paragraph 2.2.1.1 and RG 1.177, paragraph 2.2.1. These principles, and the impact of the proposed change on each, are described below.

A reasonable balance is preserved between prevention of core damage, prevention of containment failure, and consequence mitigation.

The proposed SLC System CT extension does not affect the ability of SLC, or any system, to prevent core damage, prevent containment failure, or mitigate the consequences of an accident. The proposed change has only a very small impact on risk. The proposed change does not compensate for this risk impact with an assumption of improved containment integrity, nor does this proposed change degrade containment integrity and compensate with an assumption of improved core damage prevention.

Over-reliance on programmatic activities to compensate for weaknesses in plant design is avoided.

Plant design for both the primary (Le., RPS and ARI/RPT) and alternate (Le., SLC System) reactivity control systems at PBAPS is robust. The proposed SLC System CT extension does not require nor rely upon programmatic activities. During the extended CT, the dual-channel RPS, in concert with the control rods, ensures reliable and automatic control of reactivity changes to ensure that fuel design limits are not exceeded. The scram system is designed so that the scram signal overrides all other operating signals. Upon loss of either instrument air or electrical power, the scram valves will fail open. Hence, failure of the valves' air system or electric system will produce, rather than prevent, a scram.

System redundancy, independence, and diversity are maintained commensurate with the expected frequency and consequences of challenges to the system.

The redundancy, independence, and diversity of the RPS, the control rods, and the control rod drive system are not affected during the extended 72-hour SLC System CT. Since entry into the dual-train SLC System CT will typically be due to equipment failure, the EGC Configuration Risk Management Program (CRMP) will assess the emergent condition and direct activities as appropriate from a risk management perspective.

Page 5 of 19

ATTACHMENT 1 Evaluation of Proposed Amendment Additional redundancy for both reactivity control and suppression pool pH control is established by the PBAPS Emergency Operating Procedures (EOPs). The EOPs describe the actions and criteria for alternate means of manual addition of boron, should the SLC System fail to operate as designed.

Defenses against potential common cause failures are maintained and the potential for introduction of new common cause failure mechanisms is assessed.

The extended SLC System CT does not change the design function of the SLC System. Therefore, the proposed change does not affect existing common cause failure mechanisms. In addition, the operating environment and operating parameters for the SLC System, the RPS system, the control rods, and the control rod drive system remain constant; therefore, new common cause failure modes are not expected. Consequently, no new potential common cause failure mechanisms have been introduced by the proposed change.

Independence of barriers is not degraded.

The extended CT does not provide a mechanism that degrades the independence of fission product barriers (Le., fuel cladding, the reactor coolant system, or containment).

Defenses against human errors are maintained.

The risk assessment for the extended SLC System CT does not credit nor require new operator actions. Therefore, the proposed change does not impact defense-in-depth against human error.

3.3 Safety Margin Assessment The proposed SLC System CT extension does not involve a reduction in the margin of safety. The margin of safety is established through the design of the plant structures, systems, and components, the parameters within which the plant is operated, and the setpoints for the actuation of equipment relied upon to respond to an event. The proposed amendment does not modify the safety limits or setpoints at which protective actions are initiated. Safety margins applicable to the SLC System include pump capacity, boron concentration, boron enrichment, and system response timing. Since this proposed TS amendment does not change the SLC System design, but only extends a CT, safety margins are not challenged.

3.4 Risk Assessment The CT is defined as part of the limiting condition for operation (LCO), and is intended to allow sufficient time to repair failed equipment while minimizing the risk associated with the loss of the component function. An extension of the CT increases the unavailability of a component due to the increased time the component is out-of-service for maintenance. The CT risk is reflected in the core damage frequency (CDF) and the large early release frequency (LERF) by adjusting the component unavailability due to maintenance.

Page 6 of 19

ATTACHMENT 1 Evaluation of Proposed Amendment The proposed CT extension for the dual-train inoperability of the PBAPS SLC System provides additional time to complete test and maintenance activities while at power, potentially reducing the number of forced outages related to compliance with the existing CT.

EGC completed risk assessments for PBAPS using the respective full power internal events, Level 1 CDF models and the associated Level 2 LERF models. These risk assessments are provided in Attachment 4. The risk assessments were performed in accordance with the requirements in RG 1.174, RG 1.177, and RG 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk Informed Activities," Revision 1. The results of these risk assessments are discussed below.

3.4.1 Regulatory Standards The RG 1.174 acceptance guidelines for a permanent TS change specify that the delta (a)CDF and the aLERF associated with the change should be less than specified acceptable values, which are dependent on the baseline CDF and LERF. These specified acceptable values are presented for two ranges of risk impacts: those described as "small changes," and those described as livery small changes. EGC utilized the acceptance guidelines for livery small changes" II in the risk assessments for the proposed PBAPS changes.

The RG 1.174 acceptance guidelines prescribe that the risk metrics of aCDF and aLERF be less than 1.0E-06/yr and 1.0E-07/yr, respectively, to establish a "very small" risk increase with no additional compensatory measures required.

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

RG 1.177 identifies a three-tiered approach for the evaluation of the risk associated with a proposed TS change.

Tier 1, PRA Capability and Insights Tier 1 is an evaluation of the plant-specific risk associated with the proposed TS change as shown by the change in CDF and incremental conditional core damage probability (ICCDP). Where applicable, containment performance should be evaluated on the basis of an analysis of LERF and incremental conditional large early release probability (ICLERP). The acceptance guidelines given in RG 1.177 for determining an acceptable TS change are that the ICCDP and the ICLERP associated with the change should be less than 5E-07 and 5E-08, respectively.

Tier 2, Avoidance of Risk Significant Plant Configuration Tier 2 identifies and evaluates, with respect to defense-in-depth, any potential risk-significant plant equipment outage configurations associated with the proposed change. As such, procedures 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.

Page 7 of 19

ATTACHMENT 1 Evaluation of Proposed Amendment Tier 3, Risk-Informed Configuration Risk Management Tier 3 provides for the establishment of an overall 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.

RG 1.200, Revision 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 intended to be consistent with the NRC's PRA Policy Statement and more detailed guidance in RG 1.174.

RG 1.200, Revision 1, endorses Addendum B of the American Society of Mechanical Engineers (ASME) Standard RA-S-2002, "Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications," Addenda RA-Sa-2003, and Addenda RA-Sb-2005, as applicable to full power internal event (FPIE) PRA models.

Since that time, the new ASME/American Nuclear Society (ANS) Standard RA-Sa-2009, "Addenda to RA-S-2008, Standard for Level1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications,"

has been released. Although this standard is presently issued and endorsed by RG 1.200, Revision 2, neither of these documents adds further requirements that impact the results of the SLC System CT risk assessment.

3.4.2 Tier 1, PRA Capability and Insights As stated in RG 1.177, Tier 1 is an evaluation of the impact of the proposed TS change on CDF, ICCDP, and, when appropriate, LERF and ICLERP considering PRA validity, and PRA insights and findings. Table 4.4.2-1 below provides the plant-specific risk associated with the proposed PBAPS TS change using the FPIE PRA models and based on the risk metrics of ~CDF, ICCDP, ~LERF, and ICLERP.

Page 8 of 19

ATTACHMENT 1 Evaluation of Proposed Amendment Table 4.4.2-1 PBAPS Risk Assessment Summary Results Hazard ACDF ICCDP ALERF ICLERP FPIE 2.3E-08/yr 2.3E-08 4.1 E-10/yr 4.1 E-10 Acceptance Criteria <1.0E-06/yr <5.0E-07 <1.0E-07/yr <5.0E-08 External Events (1 ) (1 ) (1 ) (1 )

(1) In accordance with RG 1.174, paragraph 2.2.5.5, "Comparisons with Acceptance Guidelines, II EGC performed a qualitative assessment of external event risk associated with the proposed PBAPS SLC System CT extension (Le., as described below and in Appendix A of Attachment 4) to demonstrate that the changes in risk remain within the acceptance guidelines.

The base results of the risk assessment, as summarized in Table 4.4.2-1 above, indicate that the ~CDF, ICCDP, ~LERF, and ICLERP risk metric values for the proposed change are below the acceptance guidelines as defined in RG 1.174 and RG 1.177. This analysis demonstrates that the proposed TS change.

satisfies the risk acceptance guidelines in RG 1.174 and RG 1.177, and therefore meets the intent of "very small" risk increases consistent with the NRC's Safety Goal Policy Statement.

As part of the risk assessments, EGC performed a sensitivity analysis to determine the maximum allowable CT prior to exceeding the livery smallII acceptance criteria. For this sensitivity, ICCDP and ICLERP were set to their maximum allowable values in RG 1.177, and the CTNEW allowable was calculated. ICLERP was determined to be the bounding parameter, and a CTNEW value of 1596 hours0.0185 days <br />0.443 hours <br />0.00264 weeks <br />6.07278e-4 months <br /> for PBAPS was calculated. This value represents significant margin relative to the proposed CT extension.

The PBAPS risk assessment also included a qualitative assessment of external event risks in accordance with RG 1.174, paragraph 2.2.5.5, IIComparisons with Acceptance Guidelines. 1I This qualitative assessment is summarized below and described in Appendix A of Attachment 4.

Internal Fires The impact on the internal fires risk profile due to the proposed change was evaluated using the following information sources:

• NUREG/CR-6850, IIEPRI Report 1011989, 'Fire PRA Methodology for Nuclear Power Facilities', II September 2005

• PECO Energy, "Peach Bottom Units 2 and 3 Individual Plant Examination for External Events," May 1996

• Boiling Water Reactor Owners' Group (BWROG), IIAssessment of NRC Information Notice 2007-07,11 October 16, 2007 (Le., Appendix C of Attachment 4)

Page 9 of 19

ATTACHMENT 1 Evaluation of Proposed Amendment The assessment concluded that fire hazards can be appropriately screened as non-significant contributors to the risk assessment of the proposed SLC System CT because of the low frequency of a fire coupled with a failure to scram.

Seismic The impact on the seismic risk profile for PBAPS due to the proposed change was evaluated using the following information sources:

• PECO Energy, "Peach Bottom Units 2 and 3 Individual Plant Examination for External Events," May 1996

• NUREG-1150, "Severe Accident Risks: An Assessment for Five U.S.

Nuclear Power Plants, December 1990 II The assessment concluded that the seismic hazard can be appropriately screened as a non-significant contributor to the risk assessment of the proposed change.

Other External Hazards Other external event risks such as external floods, severe weather, high winds or tornados, transportation accidents, nearby facility accidents, turbine missiles, and other miscellaneous external hazards were also considered in the PBAPS Individual Plant Examination for External Events (I PEEE) listed above. No significant quantitative contribution from these external events was identified by the IPEEE evaluations. As such, other external hazards are appropriately screened as non-significant contributors to the risk assessment of the proposed CT.

Consistent with the ASME PRA Standard, quantitative parametric uncertainty analyses for both CDF and LERF were evaluated to determine if the point estimates calculated by the PRA model appropriately represent the means for the risk metrics that were evaluated. The results of these analyses are summarized in Appendix B of Attachment 4.

The parametric uncertainty analysis supports the use of the point estimate to represent the mean for the calculation of the changes in the risk metrics for the extended CT.

An assessment of modeling uncertainties is also documented in Appendix B of Attachment 4. This assessment includes site-specific modeling uncertainty evaluations for the PRA Base Case and an examination of the specific cutsets that affect the change in the CDF risk metric associated with the change in the SLC System CT extension. The results of the modeling uncertainty assessments do not change the conclusions of this risk assessment for the proposed SLC System CT changes.

3.4.3 Tier 2, Avoidance of Risk Significant Plant Configurations Tier 2 requires an examination of the need to impose additional restrictions when operating under the proposed CT in order to avoid risk-significant equipment outage configurations. Consistent with the guidance in Regulatory Position C.2.3 Page 10 of 19

ATTACHMENT 1 Evaluation of Proposed Amendment of RG 1.177, and as part of the PBAPS risk assessment (Le., Attachment 4),

EGC performed an evaluation of equipment according to its contribution to plant risk while the equipment covered by the proposed CT change is out-of-service for test or maintenance (Le., site-specific modeling uncertainty evaluations for the PRA base case and an examination of the specific cutsets that affect the change in the CDF risk metric associated with the change in the SLC System CT extension).

This evaluation is provided in Attachment 4, Appendix B, IIUncertainty Analysis, II section B.2, IIModel Uncertainties Associated with SLC System Out of Service. 1I This evaluation indicates that the scram system hardware failure is the most important contributor to the LlCDF assessment for the SLC System out-of-service case.

Since TS 3.1.7, Condition C, is typically entered due to SLC System equipment failure, the Tier 3 CRMP discussed below will assess the emergent condition, including the impact of any additional out-of-service equipment. With both SLC subsystems inoperable, the PBAPS on-line risk would be depicted as 1I0range,II based on the deterministic assessment portion of the CRMP. In this condition, station procedures require senior management review and approval to remove equipment from service, as well as implementation of compensatory measures to reduce risk, including contingency plans.

3.4.4 Tier 3, Risk-Informed Configuration Risk Management Tier 3 requires a proceduralized process to assess the risk associated with both planned and unplanned work activities. The objective of the third tier is to ensure that the risk impact of out-of-service equipment is evaluated prior to performing any maintenance activity. As stated in Section 2.3 of RG 1.177, a viable lI program would be one that is able to uncover risk-significant plant equipment outage configurations in a timely manner during normal plant operation. II The third-tier requirement is an extension of the second-tier requirement, but addresses the limitation of not being able to identify all possible risk-significant plant configurations in the Tier 2 evaluation.

EGC has developed and implemented a CRMP at PBAPS. The CRMP is governed by station procedures that ensure the risk impact of out-of-service equipment is appropriately evaluated prior to performing any maintenance activity. These procedures require 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 such as testing or load dispatching, and weather conditions.

PBAPS 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.

Page 11 of 19

ATTACHMENT 1 Evaluation of Proposed Amendment For planned maintenance activities, an assessment of the overall risk of the activity on plant safety is currently performed prior to scheduled work. The assessment 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 is reviewed, and work on SSCs that is important in mitigating the transient is avoided.

• Work is not scheduled that is highly likely to exceed a TS or Technical Requirements Manual (TRM) CT requiring a plant shutdown.

• For Maintenance Rule high risk 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 CDF and LERF. The results of the risk assessment are classified by a color code based on the increased risk of the activity. As postulated risk for the activity increases, appropriate actions are required and implemented. 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 PRA risk management procedure ER-AA-600-1042, "On-Line Risk Management," provides guidance for the Risk Management Support of the On-Line Work Control Process Program. EGC's PRA risk management procedure ER-AA-600 defines the requirements for maintaining and updating the PRA model used to evaluate on-line maintenance activities.

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 PRA assessment 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 SLC System, RPS, control rods, control rod drives and the ARI system are monitored under the Maintenance Rule Program.

If the pre-established reliability or availability performance criteria is exceeded for an instrumentation component, that component is considered for 10 CFR 50.65, IIRequirements for monitoring the effectiveness of maintenance at nuclear power plants, II paragraph (a)(1) actions, requiring increased management attention and goal setting in order to restore performance (Le., 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.

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.

Page 12 of 19

ATTACHMENT 1 Evaluation of Proposed Amendment 3.4.5 Technical Adequacy of PRA Model As stated in Section 1.0 above, RG 1.200, Revision 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 as an input in regulatory decision-making.

With respect to the risk assessment for the proposed SLG System GT extension, EGG has documented this determination of PRA quality in Attachment 4. Table 2-1 of the attachment provides a "RG 1.200 Analysis Actions Roadmap." This roadmap cross references the required RG 1.200 actions to sections in the site-specific attachments that address the actions which are summarized below.

EGG employs a multi-faceted approach to establishing and maintaining the technical adequacy and plant fidelity of the PRA models for all operating EGG 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 EGG 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 EGG Risk Management program which consists of a governing procedure (Le., ER-AA-600, "Risk Management") and subordinate Technical & Reference Material (T&RM) documents. EGG T&RM ER-AA-600-1015, "FPIE PRA Model Update,"

delineates the responsibilities and guidelines for updating the full power internal events PRA models at all operating EGG nuclear generation sites.

The overall EGG Risk Management program, including ER-AA-600-1 015, defines the process for implementing regularly scheduled and interim PRA model updates for tracking issues identified as potentially affecting the PRA models (e.g., changes in the plant, errors or limitations identified in the model, industry operating experience), and for controlling the model and associated computer files.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria 10 CFR 50.62, II Requirements for reduction of risk from anticipated transients without scram (ATWS) events for light-water-cooled nuclear power plants" 10 GFR 50.62 (c)(4) states that boiling water reactors are required to have a standby liquid control (SLG) system with the capability of injecting, into the reactor pressure vessel (RPV), a borated water solution with a flow rate, boron concentration, and boron-10 enrichment that would be necessary to ensure that the resulting reactivity control is at least equivalent to that resulting from injection of 86 gallons per minute of 13 weight percent sodium pentaborate decahydrate solution at the natural boron-l0 isotope abundance into a 251-inch inside diameter reactor pressure vessel for a given core design. Furthermore, the SLG System and its injection location must be designed to perform its function in a reliable manner. The proposed change will not impact the ability of the PBAPS SLG System to ensure compliance with these requirements.

Page 13 of 19

ATTACHMENT 1 Evaluation of Proposed Amendment 10 CFR 50.67, IIAccident source term ll 10 CFR 50.67.b(1) provides guidance to licensees with respect to revision of the licensee's current accident source term in design basis radiological consequence analyses. Specifically, the regulation states that in order to revise the accident source term, a licensee shall apply for a license amendment under 10 CFR 50.90 and that the application shall contain an evaluation of the consequences of applicable design basis accidents previously analyzed in the safety analysis report.

By letter dated July 13, 2007, EGC requested an amendment to the PBAPS TSs regarding the adoption of an alternate source term (AST) methodology. The NRC approved the requested license amendment by letter and safety evaluation (SE) dated September 05,2008. As part of the proposed AST methodology, EGC proposed the use of the SLC System to inject sodium pentaborate into the RPV following a LOCA in order to maintain suppression pool pH above 7 (Le., in order to ensure against re-evolution of elemental iodine). As such, the SLC System is required to be operable in Modes 1,2 and 3 to ensure that offsite doses remain within the limits of 10 CFR 50.67, IIAccident ll source term following LOCA involving significant fission product releases. However, additional redundancy for the control of suppression pool pH control following a LOCA is established by the PBAPS Emergency Operating Procedures (EOPs). The EOPs describe the actions and criteria for alternate means of manual addition of boron should the SLC System fail to operate as designed. Therefore, the proposed SLC System CT extension will not impact the ability of PBAPS to comply with the requirements of 10 CFR 50.67.

AEC Criterion 27 - Redundancy of Reactivity Control AEC Criterion 27 specifies IIAt least two independent reactivity control systems, preferably of different principles, shall be provided. II The SLC System shall be designed to provide backup reactivity control that is redundant with, but independent of, the control rods to conform with AEC Criterion 27. Therefore, the proposed SLC System CT extension will not impact the ability of PBAPS to comply with the requirements of AEC Criterion 27.

AEC Criterion 28 - Reactivity Hot Shutdown Capability AEC Criterion 28 specifies IIAt least two of the reactivity control systems provided shall independently be capable of making and holding the core subcritical from any hot standby or hot operating condition, including those resulting from power changes, sufficiently fast to prevent exceeding acceptable fuel damage limits. 1I The SLC System shall be designed with the capability to establish and maintain the reactor subcritical from the steady-state rated operating condition at any time in the core life, sufficiently fast to prevent exceeding acceptable fuel damage limits and independently of the control rods, to conform with AEC Criterion 28.

Therefore, the proposed SLC System CT extension will not impact the ability of PBAPS to comply with the requirements of AEC Criterion 28.

Page 14 of 19

ATTACHMENT 1 Evaluation of Proposed Amendment AEC Criterion 29 - Reactivity Shutdown Capability AEC Criterion 29 specifies "At least one of the reactivity control systems provided shall be capable of making the core subcritical under any condition (including anticipated operational transients) sufficiently fast to prevent exceeding acceptable fuel damage limits. Shutdown margins greater than the minimum worth of the most effective control rod when fully withdrawn shall be provided. II The SLC System shall be designed with the capability to establish and maintain the reactor subcritical with a minimum concentration of boron-10 in the RPV of 121 PPM in the event of a loss of reactivity control systems and ATWS events sufficiently fast at a flow rate of about 50 GPM (non-ATWS), a boron equivalent in control capacity to 86 GPM of 13 weight percent natural boron (ATWS) and a maximum boron-10 solution concentration of 9.82 percent, a pump discharge pressure of 1255 PSIG to prevent exceeding acceptable fuel damage limits, and a boron-1 0 solution saturation temperature not exceeding 43°F to conform with AEC Criterion 29.

Therefore, the proposed SLC System CT extension will not impact the ability of PBAPS to comply with the requirements of AEC Criterion 29.

AEC Criterion 30 - Reactivity Holddown Capability AEC Criterion 30 specifies "At least one of the reactivity control systems provided shall be capable of making and holding the core subcritical under any conditions with appropriate margins for contingencies. II The SLC System shall be designed with the capability for controlling the reactivity difference between the steady-state, rated operating condition, and the cold shutdown condition, including shutdown margin, at any time in the core life to conform with AEC Criterion 30.

Therefore, the proposed SLC System CT extension will not impact the ability of PBAPS to comply with the requirements of AEC Criterion 30.

Regulatory Guide 1.174, IIAn Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis" Regulatory Guide 1.177, IIAn Approach for Plant-Specific, Risk-Informed Decision-ll making: Technical Specifications Regulatory Guide1.200, IIAn Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, II Revision 1 Regulatory Guide (RG) 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, specifies II risk-informed acceptance guidelines for a permanent TS change. These acceptance guidelines are presented for two ranges of risk impacts, those described as "small changes" and those described as livery small changes. II The RG 1.174 acceptance guidelines prescribe that the risk metrics of delta (8) CDF and 8LERF be less than 1.0E-06/yr and 1.0E-07/yr, respectively, to establish a "very small" risk increase with no additional compensatory measures required. RG 1.174, paragraph Page 15 of 19

ATTACHMENT 1 Evaluation of Proposed Amendment 2.2.5.5, "Comparisons with Acceptance Guidelines," also specifies guidelines for consideration of external events, and stipulates that external events can be evaluated in either a qualitative or quantitative manner.

RG 1.177, "An Approach for Plant-Specific, Risk-Informed Decision-making: Technical Specifications," identifies a three-tiered approach for the evaluation of the risk associated with a proposed TS change.

RG 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision 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.

The proposed change complies with the acceptance guidelines and requirements of RG 1.174, RG 1.177, and RG 1.200 to demonstrate a very small change in risk.

Regulatory Summary 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 NRC'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.

4.2 Precedent

• Dresden Nuclear Power Station I Quad Cities Nuclear Power Station, License Amendment Request (LAR), November 10,2009 (ML093140516)

Description:

The amendment revises the Technical Specifications (TS) Section 3.1.7, Standby Liquid Control (SLC) System, to extend the completion time (CT) for Condition B (Le., "Two SLC subsystems inoperable.") from eight hours to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

This LAR is currently under review by the NRC.

4.3 No Significant Hazards Consideration 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; (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.

Exelon Generation Company, LLC (EGC) has evaluated the proposed changes to the Technical Specifications (TS) for Peach Bottom Atomic Power Station (PBAPS), Units 2 and 3 using the criteria in 10 CFR 50.92 and has determined that the proposed changes Page 16 of 19

ATTACHMENT 1 Evaluation of Proposed Amendment do not involve a significant hazards consideration. EGC is providing the following information to support a finding of no significant hazards consideration.

(1) Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed amendment revises Technical Specification (TS) 3.1.7, IIStandby Liquid Control (SLC) System,1I to extend the completion time (CT) for Condition C (Le., IITwo SLC subsystems inoperable for reasons other than Condition A:') from eight hours to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

The proposed change is based on a risk-informed evaluation performed in accordance with Regulatory Guides (RG) 1.174, IIAn Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions On Plant-Specific Changes to the Licensing Basis, II and RG 1.177, IIAn Approach for Plant-Specific, Risk-Informed Decision-making: Technical Specifications. II The proposed amendment modifies an existing CT for a dual-train SLC System inoperability. The condition evaluated, the action requirements, and the associated CT do not impact any initiating conditions for any accident previously evaluated. .

The proposed amendment does not increase postulated frequencies or the analyzed consequences of an Anticipated Transient Without Scram (ATWS).

Requirements associated with 10 CFR 50.62 will continue to be met. In addition, the proposed amendment does not increase postulated frequencies or the analyzed consequences of a large-break loss-of-coolant accident for which the SLC System is used for pH control. The new action requirement provides appropriate remedial actions to be taken in response to a dual-train SLC System inoperability while minimizing the risk associated with continued operation.

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

(2) Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed amendment revises TS 3.1.7 to extend the CT for Condition C from eight hours to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The proposed amendment does not involve any change to plant equipment or system design functions. This proposed TS amendment does not change the design function of the SLC System and does not affect the system's ability to perform its design function. The SLC System provides a method to bring the reactor, at any time in a fuel cycle, from full power and minimum control rod inventory to a subcritical condition with the reactor in the most reactive xenon free state without taking credit for control rod movement.

Required actions and surveillance requirements are sufficient to ensure that the SLC System functions are maintained. No new accident initiators are introduced by this amendment. Therefore, the proposed amendment does not create the possibility of a new or different kind of accident from any previously evaluated.

Page 17 of 19

ATTACHMENT 1 Evaluation of Proposed Amendment (3) Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No.

The proposed amendment revises TS 3.1.7 to extend the CT for Condition C from eight hours to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The proposed amendment does not involve any change to plant equipment or system design functions. The margin of safety is established through the design of the plant structures, systems, and components, the parameters within which the plant is operated and the setpoints for the actuation of equipment relied upon to respond to an event.

Safety margins applicable to the SLC System include pump capacity, boron concentration, boron enrichment, and system response timing. The proposed amendment does not modify these safety margins or the setpoints at which SLC is initiated, nor does it affect the system's ability to perform its design function. In addition, the proposed change complies with the intent of the defense-in-depth philosophy and the principle that sufficient safety margins are maintained consistent with RG 1.177 requirements (Le., Section C, "Regulatory Position,"

paragraph 2.2, Traditional Engineering ConsiderationsII). Therefore, the proposed amendment does not involve a significant reduction in a margin of safety.

4.4 Conclusions Based on the above analysis, EGC concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c),

and, accordingly, a finding of IIno significant hazards consideration II is justified.

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, IIStandards for Protection Against Radiation. 1I 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, IICriterion for categorical exclusion; identification of licensing and regulatory actions eligible for categorical exclusion or otherwise not requiring environmental review, II paragraph (c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

6.0 REFERENCES

1. Letter from M. A. Satorius (U. S. NRC) to C. M. Crane (Exelon Generation Company, LLC), IINotice of Enforcement Discretion for Exelon Generation Company LLC Regarding Quad Cities Nuclear Power Station, Unit 1 (NOED 06-3-01), II dated October 18, 2006 Page 18 of 19

ATTACHMENT 1 Evaluation of Proposed Amendment

2. Letter from M. A. Satorius (U. S. NRC) to C. M. Crane (Exelon Generation Company, LLC), "Notice of Enforcement Discretion for Exelon Generation Company LLC Regarding Dresden Nuclear Power Station, Unit 2 (NOED 07-3-01; TAC MD4044)," dated January 24,2007

3. Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis", Revision 1, November 2002

4. Regulatory Guide 1.177, "An Approach for Plant-Specific, Risk-Informed Decision-making: Technical Specifications Revision 0, August 1998 II

5. Regulatory Guide 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, Revision 1, August II 2006

6. PECO Energy, "Peach Bottom Units 2 and 3 Individual Plant Examination for External Events," May 1996 Page 19 of 19

ATTACHMENT 2 License Amendment Request Peach Bottom Atomic Power Station Units 2 and 3 Docket Nos. 50-277 and 50-278 Proposed Markup of PBAPS Technical Specification 3.1.7 Unit 2 Unit 3 3.1-21 3.1-21

SLC System 3.1.7 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME

~ur~

C. Two SLC subsystems C.1 Restore one SLC inoperable for reasons subsystem to OPERABLE other than status.

Condition A. 72 D. Required Action and D.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time not met. AND D.2 Be in MODE 4. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.7.1 Verify level of sodium pentaborate solution 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in the SLC tank is ~ 46%.

SR 3.1.7.2 Verify temperature of sodium pentaborate 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> solution is ~ 53°F.

SR 3.1.7.3 Verify temperature of pump suction piping 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is ~ 53°F.

SR 3.1.7.4 Verify continuity of explosive charge. 31 days (continued)

PBAPS UNIT 2 3.1-21 Amendment No. 269

SLC System 3.1. 7 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME

~our~

C. Two SLC subsystems C.1 Restore one SLC inoperable for reasons subsystem to OPERABLE other than status. 72 Condition A.

D. Required Action and 0.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time not met. AND 0.2 Be in MODE 4. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.7.1 Verify level of sodium pentaborate solution 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in the SLC tank is ~ 46%.

SR 3.1.7.2 Verify temperature of sodium pentaborate 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> solution is ~ 53°F.

SR 3.1.7.3 Verify temperature of pump suction piping 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is ~ 53°F.

SR 3.1.7.4 Verify continuity of explosive charge. 31 days (continued)

PBAPS UNIT 3 3.1-21 Amendment No. 273

ATTACHMENT 3 License Amendment Request Peach Bottom Atomic Power Station Units 2 and 3 Docket Nos. 50-277 and 50-278 Proposed Markup of PBAPS Technical Specification Bases B3.1.7 Unit 2 Unit 3 B 3.1-43 B 3.1-43 B 3.1-47 B 3.1-47

SLC System B 3.1.7 BASES ACTIONS B.1 (continued) availability of an OPERABLE subsystem capable of performing the intended SLC System function and the low probability of a DBA or severe transient occurring concurrent with the failure of the Control Rod Drive (CRD) System to shut down the plant.

The second Completion Time for Required Action B.1 establishes a limit on the maximum time allowed for any combination of concentration out of limits or inoperable SLC subsystem during any single contiguous occurrence of failing to meet the LCO. If Condition B is entered while, for instance, concentration is out of limits, and is subsequently returned to within limits, the LCO may already have been not met for up to 3 days. This situation could lead to a total duration of 10 days (3 days in Condition A, followed by 7 days in Condition B), since initial failure of the LCO, to restore the SLC System. Then concentration could be found out of limits again, and the SLC subsystem could be restored to OPERABLE. This could continue indefinitely.

This Completion Time allows for an exception to the normal "time zero" for beginning the allowed outage time "clock,"

resulting in establishing the "time zero" at the time the LCO was initially not met instead of at the time Condition B was entered. The 10 day Completion Time is an acceptable limitation on this potential to fail to meet the LCO indefinitely .

. ~~lboth SLC subsystems Condition A, at least 0

~ inoperable for reasons other than subsystem must be restored to ERABLE status within[ft]hours. The allowed Completion Time o ~ hours is considered acceptable given the low probability of a DBA or transient occurring concurrent with the failure of the control rods to shut down the ;eact~.

(Ref. 4) 0.1 and 0.2 If any Required Action and associated Completion Time is not met, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be (continued)

PBAPS UNIT 2 B 3.1-43 Revision No. 75

SLC System B 3.1.7 BASES SURVEILLANCE SR 3.1.7.9 (continued)

REQUIREMENTS Surveillance when performed at the 24 month Frequency; therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

SR 3.1.7.10 Enriched sodium pentaborate solution is made by mlxlng granular, enriched sodium pentaborate with water. In order to ensure the proper B-10 atom percentage (in accordance with Table 3.1.7-1) is being used, calculations must be performed to verify the actual B-10 enrichment within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after addition of the solution to the SLC tank. The calculations may be performed using the results of isotopic tests on the granular sodium penta borate or vendor certification documents. The Frequency is acceptable considering that boron enrichment is verified during the procurement process and any time boron is added to the SLC tank.

REFERENCES 1. 10 CFR 50.62.

2. UFSAR, Section 3.8.4.

3. 10 CFR 50.67.

/

4. RM Documentation No. PB-LAR-05, Revision 0, "Risk Assessment Input for Peach Bottom Technical Specification Change for Standby Liquid Control System Completion Time from 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />" November 16, 2009 PBAPS UNIT 2 B 3.1-47 Revision No. 75

SLC System B 3.1.7 BASES ACTIONS B.1 (continued) availability of an OPERABLE subsystem capable of performing the intended SLC System function and the low probability of a DBA or severe transient occurring concurrent with the failure of the Control Rod Drive (CRD) System to shut down the plant.

The second Completion Time for Required Action B.1 establishes a limit on the maximum time allowed for any combination of concentration out of limits or inoperable SLC subsystem during any single contiguous occurrence of failing to meet the LCD. If Condition B is entered while, for instance, concentration is out of limits, and is subsequently returned to within limits, the LCD may already have been not met for up to 3 days. This situation could lead to a total duration of 10 days (3 days in Condition A, followed by 7 days in Condition B), since initial failure of the LCD, to restore the SLC System. Then concentration could be found out of limits again, and the SLC subsystem could be restored to OPERABLE. This could continue indefinitely.

This Completion Time allows for an exception to the normal "time zero" for beginning the allowed outage time "clock,"

resulting in establishing the "time zero" at the time the LCD was initially not met instead of at the time Condition B was entered. The 10 day Completion Time is an acceptable limitation on this potential to fail to meet the LCD indefinitely.

subsys ms are inoperable for reasons other than Conditi A, at leas ne subsystem must be restored to OPER E status within hours. The allowed Completion Time of[ft]hours is considered acceptable given the low probability of a DBA or transient occurring concurrent with the failure of the control rods to shut down the reactor.

0.1 and 0.2 ~

If any Required Action and associated Completion Time is not met, the plant must be brought to a MODE in which the LCD does not apply. To achieve this status, the plant must be (continued)

PBAPS UNIT 3 B 3.1-43 Revision No. 76

SLC System B 3.1.7 BASES SURVEILLANCE SR 3.1.7.9 (continued)

REQUIREMENTS Surveillance when performed at the 24 month Frequency; therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

SR 3.1.7.10 Enriched sodium penta borate solution is made by mlxlng granular, enriched sodium penta borate with water. In order to ensure the proper B-I0 atom percentage (in accordance with Table 3.1.7-1) is being used, calculations must be performed to verify the actual B-I0 enrichment within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after addition of the solution to the SLC tank. The calculations may be performed using the results of isotopic tests on the granular sodium pentaborate or vendor certification documents. The Frequency is acceptable considering that boron enrichment is verified during the procurement process and any time boron is added to the SLC tank.

REFERENCES 1. 10 CFR 50.62.

2. UFSAR, Section 3.8.4.

3. 10 CFR 50.67.

4. RM Documentation No. PB-LAR-05, Revision 0, "Risk Assessment Input for Peach Bottom Technical Specification Change for Standby Liquid Control System Completion Time from 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />" November 16, 2009 PBAPS UNIT 3 B 3.1-47 Revision No. 76

ATTACHMENT 4 License Amendment Request Peach Bottom Atomic Power Station Units 2 and 3 Docket Nos. 50-277 and 50-278 RM Documentation No. PB-LAR-05, Revision 0

ATTACHMENT 4 License Amendment Request Peach Bottom Atomic Power Station Units 2 and 3 Docket Nos. 50-277 and 50-278 RM Documentation No. PB-LAR-05, Revision 0

RM DOCUMENTATION NO. PB-lAR..05 REV: 0 PAGE NO.1 STATION: Peach Bottom Atomic Power Station UN1T~)AFFECTED: 2and3 TITLE: Risk Assessment Input fOl" Peach Bottom Technical Specification Change for Standby Liquid Control System Completion Time from 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />

SUMMARY

This assessment is petformed in support of the license amendment request (tAR) submittal to extend the Technical Specification 3.1.7, Condition C Completion Time (CT) for the Standby Liquid Control (8 LC) System from 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

The risk assessment is perfonned in accordance with ER-AA-600.. 1046, Rev. 4, Risk fvIetrics - NOED andLAR No UREs have been created as a result of this application.

[ I Review required after periodic Update

[ X] Internal RM Documentation [ ] External RM Documentation Electronic Calculation Data Files:

Method of Review: [ X] Detailed [ ] Alternate [ ] Review of External Document This R1VI documentation supersedes: _ _..; .,;N;,.;A..;....

/ in its entirety.

Prepared by: ___R_._A_._N_ara~in / /td~ I { Q!Jt/l(J9e(

Sign Date Prepared by: ~_~D_._E_._V~an_o~v_e_r_ _ / ----::O::::..--=E:.---'V:--.,;;..a:.--_-J!'_'"'_-'-_v-J I /0/ i b /2. 00 '7 Sign D~~

Reviewed by: R. A. Hill £c:rZ4 Stgn J 1f!?lf'/~r Date Reviewed by: _ _G_._A_,T_e_a~ga_rd_e_n _ _ I ~ 2~~ I Sign ,

/c,/t S;(z.ou '1 Date Approved by: E. T. Burns

--~----_.-

JCT~ Sign.

I /0-/6-0'(

Date C467090020-8958-1 /8/201 0

RM DOCUMENTATION NO. PB-LAR-05 REV: 0 PAGE NO.1 STATION: Peach Bottom Atomic Power Station UNIT(S) AFFECTED: 2 and 3 TITLE: Risk Assessment Input for Peach Bottom Technical Specification Change for Standby Liquid Control System Completion Time from 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />

SUMMARY

This assessment is performed in support of the license amendment request (LAR) submittal to extend the Technical Specification 3.1.7, Condition C Completion Time (CT) for the Standby Liquid Control (SLC) System from 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

The risk assessment is performed in accordance with ER-AA-600-1046, Rev. 4, Risk Metrics - NOED andLAR No UREs have been created as a result of this application.

[ ] Review required after periodic Update

[ X] Internal RM Documentation [ ] External RM Documentation Electronic Calculation Data Files:

Method of Review: [ X] Detailed [ ] Alternate [ ] Review of External Document This RM documentation supersedes: _ _N/A in its entirety.

Prepared by: D.E.VanoverlR A Narain / /

Sign Date Reviewed by: R A Hill Sign Date Reviewed by: G A Teagarden Sign Date Reviewed by: G T Zucal (External Events Impact)

Sign Date Approved by: E T Bums Sign Date C467090020-8958-1 /8/2010

Peach Bottom SLC CT Extension TABLE OF CONTENTS Section

1.0 INTRODUCTION

2 1.1 Purpose 2 1.2 Background 2 1.3 SLC Technical Specifications 3 1.4 Regulatory Guides 3 1.5 Scope 6 1.6 Peach Bottom PRA Model 6 2.0 ANALYSIS ROADMAP AND REPORT ORGANIZATION 8 3.0 TIER 1 RISK ANALYSIS 9 3.1 Key Assumptions 9 3.2 Internal Events 10 3.3 Results Comparison to Acceptance Guidelines 12 3.4 External Events 12 3.5 Uncertainty Assessment. 14 3.6 Risk Summary 14 4.0 TECHNICAL ADEQUACY OF THE PRA MODEL 16 4.1 PRA Quality Overview 16 4.2 Scope 18

4.3 Fidelity

PRA Maintenance and Update 18 4.4 Standards 19 4.5 Peer Review and PRA Self-Assessment 19 4.6 Appropriate PRA Quality 21 4.7 General Conclusion Regarding PRA Capability 34 5.0

SUMMARY

AND CONCLUSIONS 35 5.1 Scope Investigated 35 5.2 PRA Quality 35 5.3 Quantitative Results vs. Acceptance Guidelines 36 5.4 Conclusions 36

6.0 REFERENCES

38 APPENDICES A EXTERNAL EVENT ASSESSMENT B UNCERTAINTY ANALYSIS C BWROG ASSESSMENT OF NRC INFORMATION NOTICE 2007-07 1 C467090020-8958-1 /8/2010

Peach Bottom SLC CT Extension

1.0 INTRODUCTION

1.1 PURPOSE The purpose of this analysis is to assess the acceptability, from a risk perspective, of a change to the Peach Bottom Atomic Power Station (PBAPS) Technical Specification (TS) for the Standby Liquid Control (SLC) system to increase the Completion Time (CT), sometimes called the allowed outage time (AOT), from 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> when both SLC subsystems (Le., both trains) are inoperable. An extension will provide flexibility during power operation in the performance of corrective maintenance, preventive maintenance, and surveillance testing of SLC system components that would cause the system to be inoperable. Consistent with the NRC's approach to risk-informed regulation, Exelon Generation Company, LLC (Exelon) has identified a particular TS requirement that is very restrictive in its nature and, if relaxed, has a minimal impact on the safety of the plant. The PBAPS analysis is consistent with similar analyses being conducted for all Exelon Boiling Water Reactor (BWR) plants that currently have an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> CT for the SLC system.

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. 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 2 C467090020-8958-1/8/2010

Peach Bottom SLC CT Extension and data and in a manner that complements the NRC's deterministic approach and supports the NRC's 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.

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.

PBAPS, other Exelon plants, and numerous other commercial nuclear plants in the industry have used these risk-informed guidelines to support both permanent and one-time CT extensions for EDGs, and other systems.

1.2.2 Exelon SLC Experiences In October 2006 (Quad Cities) and January 2007 (Dresden) Exelon requested Notices of Enforcement Discretion (NOEDs) for SLC System Tank leaks allowing an additional 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to the original a-hour completion time required for a dual-train inoperability.

These NOEDs were approved by the NRC. An extended CT would preempt the need for such NOEDs.

1.3 SLC TECHNICAL SPECIFICATIONS The proposed TS change involves extending the CT for TS 3.1.7 Condition C from a hours (current TS) to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> (proposed TS). Condition C is the situation where both SLC subsystems are inoperable. Technical Specification requirements for other SLC conditions will remain unchanged. For PBAPS, the TS Condition C applies to Modes 1 and 2.

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

3 C467090020-8958-1/8/2010

Peach Bottom SLC CT Extension 1.4.1 Regulatory Guide 1.174 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 ilCDF and the ilLERF associated with the change should be less than specified values, which are dependent on the baseline CDF and LERF, respectively. These specified values of ilCDF and ilLERF are given in RG 1.174 Figures 3 and 4, respectively. These values are presented for two ranges of risk impacts, those described as "small changes" and those described as "very small changes". The acceptance guidelines for "very small changes" are utilized in this risk assessment.

Based on the PB205C (Le., Peach Bottom Unit 2 PRA model from the 2005 PRA update, Revision C) baseline internal events CDF of 3.9E-6/yr and LERF of 1.8E-7/yr for PBAPS, the RG 1.174 minimum acceptance guidelines prescribe that the risk metrics of ilCDF and ilLERF be less than 1.0E-06/yr and 1.0E-07/yr, respectively, to establish a very small risk increase with no additional compensatory measures required.

Note that the Unit 2 model is used for this risk assessment. The results for Unit 3 would be similar (the baseline CDF for Unit 3 is slightly lower at 3.6E-6/yr).

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

1.4.2 Regulatory Guide 1.177 Regulatory Guide 1.174 [Ref. 2] 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 frequency (ICLERP). The acceptance guidelines given in RG 1.177 for determining an acceptable TS change is that the ICCDP and the ICLERP associated with the change should be less than 5E-07 and 5E-08, respectively.

• 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 4 C467090020-8958-1 /8/20 10

Peach Bottom SLC CT Extension 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 acceptance guidelines for ICCDP and ICLERP for permanent changes associated with changing a Technical Specification Completion Time. Other portions of the LAR submittal will address Tier 2 and Tier 3 elements.

1.4.3 Regulatory Guide 1.200. Revision 1 Regulatory Guide 1.200, Rev. 1 [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 intended to be consistent with the NRC's PRA Policy Statement and more detailed guidance in Regulatory Guide 1.174.

It is noted that RG 1.200 Rev. 1 endorses Addendum B of the ASME PRA Standard

[Ref. 5] applicable to full power internal event (FPIE) PRA models. Since that time, the new ASMEIANS Combined PRA Standard [Ref. 26] has been released. Although the Combined Standard is presently issued and endorsed by RG 1.200 Revision 2 [Ref. 27],

neither of these document revisions impact this analysis.

1.4.4 Acceptance Criteria Based on the guidance provided in Regulatory Guides 1.174 and 1.177, the following quantitative PRA related acceptance criteria are utilized in this risk analysis:

• L\CDF < 1.0E-06/yr

• L\LERF < 1.0E-07/yr

• ICCDP < 5.0E-07

• ICLERP < 5.0E-08 5 C467090020-8958-1/8/2010

Peach Bottom SLC CT Extension 1.5 SCOPE This section addresses the requirements of RG 1.200, Rev. 1 Section 3.2 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 PBAPS PB205C PRA model used for this analysis includes a full range of internal initiating events (including internal flooding) for at-power configurations. The SLC system is credited in the PRA for criticality control. The FPI E model is further discussed in Section 1.6.

• Low Power Operation - The FPIE assessment is judged to adequately capture risk contributors associated with low power plant operations. The FPI E 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 - In consideration of shutdown and refueling modes (Le., Modes 3,4, and 5), the SLC TS does not apply. As such, shutdown risk impacts are not discussed further in this risk assessment.

• Internal Fires - An interim fire PRA is available for PBAPS. The PBAPS interim Fire PRA [Ref. 10], and a BWROG assessment [Ref. 19] are used to provide qualitative and semi-quantitative insights to the analysis (refer to Section 3.4.1).

• Seismic - Consistent with most sites, PBAPS does not currently maintain a Seismic PRA. A qualitative assessment is performed in this analysis (refer to Section 3.2) based on insights from the PBAPS IPEEE study

[Ref. 11] and other industry studies.

• Other External Events - Other external event risks were assessed in the PBAPS IPEEE study [Ref. 11] and found to be insignificant risk contributors (refer to Section 3.4.3).

1.6 PBAPS PRA MODEL This section addresses the requirements of Section 3.1 of RG 1.200, Rev. 1 which directs the licensee to identify the portions of the PRA used in the analysis.

The PRA analysis for the TS change uses the PBAPS PB205C full power internal events Level 1 Core Damage Frequency (CDF) model and the associated Level 2 Large Early Release Frequency (LERF) model to calculate the risk metrics.

6 C467090020-8958-1 /8/2010

Peach Bottom SLC CT Extension This risk assessment applies to both PBAPS Unit 2 and Unit 3. The models for both units are maintained individually but are very similar and the risk impact of this TS change is minor. Given the higher risk metrics for Unit 2, this analysis can be considered bounding for Unit 3. Table 1-1 shows the CDF and lERF risk metrics for both units.

Table 1-1 COMPARISON OF UNIT 2 AND UNIT 3 RISK METRICS (FULL POWER INTERNAL EVENTS MODEL)

Risk Metric Unit 2 Unit 3 Percent Difference CDF 3.9068E-06 3.6352E-06 6.95%

LERF 1.7761 E-07 1.6453E-07 7.36%

The CDF and lERF for both units are very similar. Given that the risk metrics of Unit 2 are slightly higher, the analysis can be considered bounding for Unit 3. Therefore, the use of the PB205C Unit 2 PRA model to reflect the risk impact of this TS change on either unit is reasonable and acceptable.

This analysis is specific to the SlC System and therefore the SlC system fault tree model is the only portion of the PB205C PRA model modified for this risk application.

The PRA analysis involved identifying the system and components or maintenance activities modeled in the PRA which are most appropriate for use in setting both subsystems of SlC to be inoperable. As discussed later in Section 3.1, the model parameter SSYS-SlCTM2 "SlC SYSTEM UNAVAILABLE DUE TO TESTING," was selected as an appropriate parameter to adjust to make the entire SlC system unavailable in the PRA (to reflect SlC inoperable and entry into TS 3.1.7, Condition C).

No other aspect of the PB205C PRA model required adjustment for this risk application.

The entire PB205C PRA model is quantified for this assessment using the "average maintenance" PRA model (Le., no portions of the at-power internal events PB205C model were excluded or zeroed out of the quantification).

7 C467090020-8958-1 /8/20 10

Peach Bottom SLC CT Extension 2.0 ANALYSIS ROADMAP AND REPORT ORGANIZATION The analysis and documentation utilizes the guidance provided in RG 1.200, Revision 1

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

Table 2-1 RG 1.200 ANALYSIS ACTIONS ROADMAP RG 1.200 Actions Report Section

1. Identify the parts of the PRA used to support the application Section 3 1.a Systems, structures and components (SSCs), operational Section 3.2 characteristics affected by the application, and how these are implemented in the PRA model 1.b Acceptance criteria used for the application Section 1.4.4

2. Identify the scope of risk contributors addressed by the PRA model. If Section 1.5 not full scope (Le., internal and external events), identify appropriate compensatory measures or provide bounding arguments to address the risk contributors not addressed by the model.

3. Summarize the risk assessment methodology used to assess the risk Section 3 of the application. Include how the PRA model was modified to appropriately model the risk impact of the change request.

4. Demonstrate the Technical Adequacy of the PRA. Section 4 4.a Identify plant changes (design or operational practices) that have Section 4.6.1 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 application.

4.b Document that the parts of the PRA used in the decision are Section 4.6 consistent with applicable standards endorsed by the RG (currently, in RG 1.200 Rev. 1. RG 1.200 Rev. 1 addresses the internal events ASME PRA standard). Provide justification to show that where specific requirements in the standard are not met, it will not unduly impact the results.

4.c Document PRA peer review findings and observations that are Section 4.5 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.

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

8 C467090020-8958-1/8/20 10

Peach Bottom SLC CT Extension 3.0 TIER 1 RISK ANALYSIS 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 KEY ASSUMPTIONS The following inputs and general assumptions are used in estimating the plant risk due to the proposed SLC System CT extension.

a. The SLC System CT is assumed to increase from its current duration of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to a proposed duration of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

b. The base analysis in this risk assessment assumes one entry per year into the proposed CT. The duration of the proposed CT is assumed to be adequate for performing the majority of corrective maintenance, preventive maintenance, and surveillance testing on-line. An examination of SLC rolling unavailability for the period from 1/1/05 to 12/31/08 showed that for Unit 2, Train A was unavailable for 43.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />, and Train B was unavailable for 14.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />. Unit 3 unavailabilities were 30.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> for Train A, and 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> for Train B. These unavailabilities are all below the proposed 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> window. Thus, any impact from extending the CT is assumed to be negligible and it is conservatively assumed that the outage will not be entered more than once a year.

Additionally, Configuration Risk Management at PBAPS is governed by the Maintenance Rule (10 CFR 50.65(a)(4)). A sensitivity analysis of the risk associated with entering the CT was performed and indicated that the SLC system could be taken out of service for up to 1596 hours0.0185 days <br />0.443 hours <br />0.00264 weeks <br />6.07278e-4 months <br /> before the very small risk increase metrics of RG 1.174 and RG 1.177 are exceeded.

This represents a significant margin compared to the proposed 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> CT. As stated above, the historical analysis of unavailability data shows that the SLC system does not exceed this ceiling value.

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

d. The model manipulations were performed on the Unit 2 model. The results of the Unit 2 analysis can be considered bounding for Unit 3.

9 C467090020-8958-1 /8/201 0

Peach Bottom SLC CT Extension 3.2 INTERNAL EVENTS The PBAPS PB205C PRA model(1) [Ref. 4] was examined to determine which PRA basic event to modify to reflect the coincident unavailability of both SLC subsystems.

The applicable basic event for the 2005C PRA model was identified as SSYS-SLCTM2 "SLC SYSTEM UNAVAILABLE DUE TO TESTING," This event is appropriate because it fails both SLC subsystems and no other equipment in the model.

Event SSYS-SLCTM2 was set to a binary logic value of "TRUE" (using a quantification flag file) and the entire PB205C model was requantified using the same PRA software codes and revisions as used for the base PB205C model [Ref. 4]. These configuration specific CDF and LERF values are used in conjunction with the base PB205C values to calculate the risk impacts of the proposed TS change.

The calculations of ~CDF, ICCDP, ~LERF and ICLERP for the CT change are determined as shown below.

The ~CDF to be compared to the RG 1.174 acceptance guidelines is given by (as defined by [Ref. 21]):

~CDF = CDFNEW - CDFsASE [Equation 3-1]

~CDF is the difference between the annual average CDF with the CT extended and the CDF with the current CT. The ~CDF has units of "per reactor year."

In the above equation, CDFNEW is equal to:

CDFNEW =CTSLC-OOS

• CDFsLC-OOS + [(1-CTSLC-OOS)

• CDFsASE] [Equation 3-2]

(1) The PB205C baseline model used in the calculations contains the average maintenance associated with system trains.

10 C467090020-8958-1 /8/2010

Peach Bottom SLC CT Extension Where:

CDFsLC-OOS = the annual average CDF calculated with both SLC subsystems out of service (SSYS-SLCTM2 set to True)

CDFsASE = baseline annual average CDF with average unavailability for all equipment. This is the CDF result of the PB205C baseline PRA.

CT~Lc-ooS = the new extended CT as an annual unavailability (Le., 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> / 8760 hour0.101 days <br />2.433 hours <br />0.0145 weeks <br />0.00333 months <br />s/year = 8.2E-03 yr)

CTSLC-OOS = the new extended CT as a probability (Le., 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> / 8760 hours0.101 days <br />2.433 hours <br />0.0145 weeks <br />0.00333 months <br /> = 8.2E-03)

The ICCDP associated with the SLC System being out of service using the new CT is given by:

ICCDP(l) = (CDFsLc-oos - CDFsASE) X CT~Lc-oos [Equation 3-3]

Risk significance relative to ~LERF and ICLERP(1) is determined using equations of the same form as noted above for ~CDF and ICCDP.

The relevant input parameters for the base quantification of this risk analysis are summarized in Table 3.2-1. The corresponding base risk metric results for this risk analysis (based on quantification of the PB205C model and use of the above equations) are provided in Table 3.2-2.

Table 3.2-1 RISK ASSESSMENT INPUT PARAMETERS Input Parameter Value Reference CDF sAsE 3.9E-06/yr PB205C PRA [Ref. 4]

LERF sAsE 1.8E-07/yr PB205C PRA [Ref. 4]

CTSLC-OOS 8.2E-03 One 72-hr TS 3.1.7 Condition C entry assumed per year (Le., 72 hr/8760 hrs).

(1) ICCDP and ICLERP are probabilities, i.e., no units.

11 C467090020-8958-1 /8/2010

Peach Bottom SLC CT Extension Table 3.2-2 RISK ASSESSMENT BASE RESULTS Risk Metric Value Acceptance GUideline~

CDFsLC-OOS 6.6E-06/yr N/A CDFNEW 3.9E-06/yr N/A 8CDF 2.3E-08/yr <1.0E-06/yr ICCDP 2.3E-08 <5.0E-07 LERFsLc-oos 2.3E-07/yr N/A LERFNEW 1.8E-07/yr N/A 8LERF 4.1 E-1 O/yr <1.0E-07/yr ICLERP 4.1E-10 <5.0E-08 3.3 RESULTS COMPARISON TO ACCEPTANCE GUIDELINES As can be seen from Table 3.2-2, the base results of the risk assessment indicate that the ilCDF, ICCDP, ilLERF, and ICLERP risk metric values are below the acceptance guidelines as defined in RG 1.174 and RG 1.177. In addition, quantitative sensitivity cases for model uncertainties are provided in Appendix B.

This analysis demonstrates that the proposed TS change satisfies the risk acceptance guidelines in RG 1.174 and RG 1.177, and therefore meets the intent of very small risk increases consistent with the Commission's Safety Goal Policy Statement.

A sensitivity analysis was performed to determine the maximum allowable CT before exceeding the acceptance criteria for very small risk increases. For this sensitivity, ICCDP and ICLERP were set to their maximum allowable values in RG 1.177, and the CTNEW allowable was calculated. ICLERP was determined to be the bounding parameter, and a CTNEW of 1596 hours0.0185 days <br />0.443 hours <br />0.00264 weeks <br />6.07278e-4 months <br /> was calculated. This represents a significant margin compared to the proposed 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> CT.

3.4 EXTERNAL EVENTS A qualitative assessment of external event risks is provided. Further details are found in Appendix A.

12 C467090020-8958-1 /8/2010

Peach Bottom SLC CT Extension 3.4.1 Internal Fires The impact on the internal fires risk profile due to the proposed CT is evaluated using the following information sources:

• NUREG/CR-6850 [Ref. 18]

• PBAPS interim FPRA [Ref. 10]

• BWROG Assessment of Fire-Induced Failure to Scram [Ref. 19]

The internal fires risk impact assessment is discussed in Appendix A.4. The assessment concluded that fire hazards can be appropriately screened as non-significant contributors to the risk assessment of the proposed SLC CT because of the low frequency of a fire coupled with a failure to scram.

3.4.2 Seismic Exelon does not currently maintain a seismic PRA for PBAPS. The impact on the seismic risk profile due to the proposed CT is evaluated using the following information source:

• PBAPS IPEEE [Ref. 11]

• NUREG-1150 [Ref. 23] for PBAPS The seismic risk impact assessment is discussed in Appendix A.3. The assessment concluded that seismic can be appropriately screened as a non-significant contributor to the risk assessment of the proposed CT.

3.4.3 External Floods and Other External Hazards In addition to internal fires and seismic events, the PBAPS IPEEE analysis of high winds, external floods, and other external hazards (HFO) was accomplished by reviewing the plant environs against regulatory requirements regarding these hazards.

Since both PBAPS units were designed (with construction started) prior to the issuance of the 1975 Standard Review Plan (SRP) criteria, PECO (now Exelon) performed a plant hazard and design information review for conformance with the SRP criteria. For seismic and fire events that were not screened out, additional analyses were performed to determine whether or not the hazard frequency was acceptably low. HFO events were screened out by compliance with the 1975 SRP criteria. As such, these hazards were determined in the PBAPS IPEEE to be negligible contributors to overall plant risk.

As such, external flooding and other external hazards are appropriately screened as a non-significant contributor to the risk assessment of the proposed CT (refer to Appendix A.2).

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Peach Bottom SLC CT Extension 3.5 UNCERTAINTY ASSESSMENT 3.5.1 Parametric Uncertainty Consistent with the ASME PRA Standard, quantitative parametric uncertainty analyses for both CDF and LERF are evaluated to determine if the point estimates calculated by the PRA model appropriately represent the mean. The results of these analyses are summarized in Appendix B.3.

The parametric uncertainty analysis shown in Appendix B.3 supports the use of the point estimate to represent the mean for the calculation of the changes in the risk metrics for the extended CT.

3.5.2 Modeling Uncertainty An assessment of modeling uncertainties is documented in Sections B.1 and B.2.

• Section B.1 provides PBAPS specific modeling uncertainty evaluations for the Base Case.

• Section B.2 provides an examination of the specific cutsets that affect the change in the CDF risk metric associated with the change in the SLC CT.

The results of the modeling uncertainty assessments do not change the conclusions of this risk assessment for the proposed SLC CT changes.

3.6 RISK

SUMMARY

As discussed above and as summarized in Table 3.6-1, the FPIE quantitative evaluation results are well below the risk acceptance guidelines of RG 1.174 and RG 1.177.

External events evaluations are discussed in Appendix A and do not change the results or conclusions of this risk assessment. As such, this risk evaluation demonstrates that the proposed TS change can be made with a very small risk increase.

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Peach Bottom SLC CT Extension Table 3.6-1 RISK ASSESSMENT

SUMMARY

RESULTS Hazard ACDF ICCDP ALERF ICLERP FPIE 2.3E-08/yr 2.3E-08 4.1E-10/yr 4.1 E-10 Acceptance Criteria

<1.0E-06/yr <5.0E-07 <1.0E-07/yr <5.0E-08 Fire (1 ) (1 ) (1 ) (1 )

Seismic (1 ) (1 ) (1 ) (1 )

(1) Evaluated and determined not to change the conclusions of the FPIE risk analysis.

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Peach Bottom SLC CT Extension 4.0 TECHNICAL ADEQUACY OF PRA MODEL The 2005C update to the PBAPS PRA model (PB205C) is the most recent evaluation of the risk profile at PBAPS for FPI E challenges. The PBAPS 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 PBAPS 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 PBAPS PRA.

4.1 PRA Quality Overview The quality of the PBAPS 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 NRC's traditional defense-in-depth philosophy. "

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

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 [PRAJ policy statement will improve the regulatory process in three ways:

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Peach Bottom SLC CT Extension

- 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 (Le., 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

>> Low Power

>> Shutdown

- The metrics used in the quantification:

>> Level 1 PRA - CDF

>> 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): ASMEIANS PRA Standard [Ref. 5] as endorsed by the NRC in Regulatory Guide 1.200 [Ref. 1].

• Peer Review (Section 4.5): 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.

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Peach Bottom SLC CT Extension 4.2 SCOPE The PBAPS PRA is a full power, internal events (FPIE) PRA that addresses both CDF and LERF. The quantitative insights from the FPIE PRA are directly applicable to the SLC CT Extension PRA application. This scope is judged to be adequate to support the SLC CT PRA application.

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 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 FPI E PRA. This assessment does not address the risk implications associated with low power or shutdown operation or with external events (including fire).

4.3 FIDELITY

PRA MAINTENANCE AND UPDATE 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 ll consists of a governing procedure (ER-AA-600, IIRisk Management and subordinate

)

implementation procedures. Exelon procedure ER-AA-600-1015, IIFPIE PRA Model ll Update 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:

• Design changes and procedure changes are reviewed for their impact on the PRA model.

• New engineering calculations and revisions to existing calculations are reviewed for their impact on the PRA model.

• Maintenance unavailabilities are captured, and their impact on CDF is trended.

• Plant specific initiating event frequencies, failure rates, and maintenance unavailabilities are updated approximately every four years.

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Peach Bottom SLC CT Extension 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.

4.4 STANDARDS The ASME PRA Standard [Ref. 5] provides the basis for assessing the adequacy of the PBAPS PRA as endorsed by the NRC in RG 1.200, Rev. 1 [Ref. 1]. The predecessor to the ASME PRA Standard was NEI 00-02 which identified the critical internal events PRA elements and their attributes necessary for a quality PRA.

4.5 PEER REVIEW AND PRA SELF-ASSESSMENT 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 PRA Standard [Ref.

5] specifies that a PRA Peer Review be performed on the PRA.

• The use of the ASME PRA Standard to define the criteria to be used in establishing the quality of individual PRA elements 19 C467090020-8958-1 /8/2010

Peach Bottom SLC CT Extension A summary of the results from the independent PRA Peer Review and the ASME PRA Standard self-assessment review are included along with the resolution of the review comments.

Several assessments of technical capability have been made and continue to be planned for the PBAPS PRA model. A chronological list of the assessments performed includes the following:

• An independent PRA peer review was conducted under the auspices of the BWR Owners' Group in 1998, following the Industry PRA Peer Review process [Ref. 6]. This peer review included an assessment of the PRA model maintenance and update process.

• In 2004, prior to the 2005 PRA update, a Self Assessment ("Gap") analysis was performed against the available version of the ASME PRA Standard, Addendum A [Ref. 5] and the draft version of Regulatory Guide 1.200, DG-1122 [Ref 22]. In 2006, an assessment of the extent to which the previously defined gaps had been addressed was performed in conjunction with a PRA model update.

• During 2005 and 2006 the PBAPS, Units 2 and 3 PRA model results were evaluated in the BWR Owners' Group PRA cross-comparisons study performed in support of implementation of the mitigating systems performance indicator (MSPI) process.

• As part of the next PRA model update in 2010, the gap analysis is expected to be updated to reflect pertinent changes to both the PRA Standard and Regulatory Guide 1.200.

A summary of the disposition of 1998 Industry PRA Peer Review facts and observations (F&Os) for the PBAPS, Units 2 and 3 PRA models was documented as part of the statement of PRA capability for MSPI in the PBAPS MSPI Basis Document [Ref 7]. As noted in that document, there were no significance level A F&Os from the peer review, and all significance level B F&Os were addressed and closed out with the completion of the current PB205 and PB305 models of record. Also noted in that submittal was the fact that, after allowing for plant-specific features, there are no MSPI cross-comparison outliers for PBAPS (refer to the third bulleted item above).

4.5.1 PRA Peer Review Overview An independent peer review of the Peach Bottom PRA was performed in 1998 following the review guidelines of the BWR Owner's Group (a predecessor to the ASME PRA Standard). All of the significance level "A" and "B" F&Os have been resolved.

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Peach Bottom SLC CT Extension PRAs can be used in applications despite not meeting all of the Supporting Requirements of the Combined ASMEIANS PRA Standard. This is well recognized by the NRC and is explicitly stated in the Combined ASMEIANS PRA Standard and RG 1.174. RG 1.174 states the following in Section 2.2.6:

There are, however, some applications that, because of the nature of the proposed change, have a limited impact on risk, and this is reflected in the impact on the elements of the risk model.

The proposed SLC CT Extension PRA application may not require more than Capability Category I for some SRs. It is also acknowledged that for PRAs with SRs ranked as "Not Met," the PRA may be used for PRA applications but may require additional justification and support to allow their use. Finally, it is judged that no PRA has Capability Category III for all of its SRs, nor is this currently expected as part of the NRC PRA Quality Program.

4.5.2 Self-Assessment Overview A Self-Assessment ("Gap" Analysis) for the 2002 PBAPS, Units 2 and 3 PRA models (PB202 and PB302, respectively) was completed in January 2004 in preparation for the 2005 PRA update. This Gap Analysis was performed against PRA Standard RA-S-2002 [Ref. 5] and associated NRC comments in draft regulatory guide DG-1122, the draft version of Regulatory Guide 1.200 Revision O. This gap analysis defined a list of 83 supporting requirements from the Standard for which potential gaps to Capability Category II of the Standard were identified. For each such potential gap, a PRA updating requirements evaluation (URE) (Exelon model update tracking database) was documented for resolution.

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 21 C467090020-8958-1 /8/2010

Peach Bottom SLC CT Extension URE database associated with plant changes for PBAPS 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 is that none of the plant changes have any measurable impact on the SLC CT extension request.

4.6.2 Consistency with Applicable PRA Standards As indicated above, an independent peer review of the PBAPS PRA was performed in 1998 following the review guidelines of NEI 00-02 (the predecessor to the ASME PRE Standard). All of the significance level "A" and "B" F&Os have been resolved.

The results of the 1998 peer review are also used to identify the relevant peer review findings for the PRA model used for this assessment.

The self-assessment provides the connection between the PRA and the ASME PRA Standard by also considering the PRA Peer Review comments.

Table 4-2 summarizes the evaluation of the identified "gaps" from the self-assessment and their impact on the SLC CT extension request.

In summary, of the 21 gaps identified and evaluated in Table 4-2, none have a measurable impact on the SLC CT extension request.

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Peach Bottom SLC CT Extension Table 4-1 IMPACT OF PLANT CHANGES SINCE THE LAST UPDATE ON THE PBAPS PRA MODEL URE Number Plant change Impact on the PBAPS PRA I~pact on the Application PB2008-002 A change was made in SE-11 (loss of No impact. A detailed review of the proposed changes to SE-11.1 as No impact.

offsite power procedure) for documented in the ECR was performed and compared to the human placement of the Load Tap Changer reliability assessment. The changes and movement of the steps has no to be maintained in manual at an numerical impact on the Human Error Probability evaluation. However, optim um tap position of 25 for the the next update of the HRA Notebook should include reference to the entire SBO load sequence. revised steps that will be included in revision to SE-11.1.

PB2007-024 An Engineering Change Request No impact. Conversion of the cable spreading room and computer room No impact.

(ECR) was created for converting the Cardox fire suppression control systems from automatic to manual cable spreading room and computer does not impact the PBAPS Fire PRA. Automatic suppression is not room Cardox fire suppression control credited for fires in the cable spreading room or computer room, which systems from automatic systems to are evaluated in the analysis of Fire Compartment 25.

manual systems.

PB2006-040 An ECR was developed to disable the Minimal impact. The disabling of the alarm and system trouble alarm No impact.

alarm and system trouble alarm functions should not impact the turbine trip frequency since the trip functions associated with the unit 2 setpoints are not being modified by this change.

main turbine differential expansion In any event, the Turbine Trip Initiating Event Frequency is based on detector. No turbine trips will be plant-specific data collection efforts, and will be updated as part of the disabled. normal PRA model update process.

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Peach Bottom SLC CT Extension Table 4-2 Status of Identified Gaps to Capability Category II of the ASME PRA Standard Applicable Importance to Item Description of Gap SRs Current Status I Comment Application Gap #1 Update the PBAPS ISLOCA evaluation to be consistent IE-C12 Open - ISLOCA update has not yet Not significant with more recent Exelon evaluations. been performed. However, the given that the current ISLOCA values are current reasonably conservative compared to approach is other sites that have utilized the more reasonably detailed methodology. conservative.

Gap #2 Interview plant operations, maintenance, engineering, and IE-A6 Open - Although this would be an None. Category safety analysis personnel for the purpose of identifying enhancement to the IE Notebook, it is I is met and potentiallEs that may have been overlooked. Alternatively, not judged as a high priority. The appropriate for have such personnel review Section 2 of the PBAPS PRA current IE evaluation provides this application.

IE Notebook and provide comments to this effect. thorough documentation of the Incorporate results of these interviews/reviews as an Initiating Events considered in the appendix to the IE Notebook or as a set of appropriate PBAPS model that is consistent with sentences (with references) to Section 2 of the IE other BWRs.

Notebook. Note that Cat I for this SR does not require the performance of interviews for this purpose.

Gap #6 Development of a PBAPS PRA Dependency Matrix SC-A4 Partially resolved - Although a specific None.

Notebook such that it becomes the notebook describing the dependency matrix notebook has not Dependencies approach to treatment of all the various type of been prepared for PBAPS, each of the are modeled.

dependencies throughout the PRA should be considered. system notebooks includes a This is simply a This can be accomplished by summarizing how all the description of all dependencies and baseline PRA various aspects of dependencies are treated and where the includes a detailed dependency model associated analyses for the dependencies (e.g., supporting matrix. Additionally, accident documentation walkdown information, room cooling assumptions, water sequence dependencies as a function consideration.

supply duration, HRA, CCF) are documented. of initiating event category are discussed in the event tree notebook.

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Peach Bottom SLC CT Extension Table 4-2 Status of Identified Gaps to Capability Category II of the ASME PRA Standard Applicable Importance to Item Description of Gap SRs Current Status I Comment Application Gap #11 Provide descriptions of the limitations of thermal hydraulic SC-C2 Open - Included in practice. MAAP None. The analyses with respect to their use in the PRA (bases for was not utilized outside the bounds of model is not success criteria, HRA timing, etc.) and ensure the known acceptability. Otherwise, used beyond its application is within the limits of the code. Assessments of awaiting guidance from EPRI and known the capability limitations may be limited to the specific endorsement from NRC. limitations. This application of the calculation. is a docum entation consideration only.

Gap #25 To meet the requirements of SR HR-A1 and HR-B1, the HR-A1 Open - Pre-initiator errors are Not significant.

following would be developed as supporting documentation HR-B1 included for some risk significant Capability for PBAPS: systems (Le. HPCI, RCIC, LPCS, and Category I is

- A list of the PRA systems to consider for test and SLCS) on a generic basis. believed to be maintenance actions The performance of a detailed met for HR-B1.

process for identifying and screening The pre-initiator

- Rules for identifying and screening test and maintenance assessm ent that actions from the PRA test and maintenance pre-initiators is judged to have a minimal impact on exists is

- A list of procedures reviewed, the potential test and the results of the model. adequate for this maintenance actions associated with the procedures, and application. Pre-the disposition of the action (screened or evaluated). initiator human actions do not contribute significantly to the risk significance results for this application.

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Peach Bottom SLC CT Extension Table 4-2 Status of Identified Gaps to Capability Category II of the ASME PRA Standard Applicable Importance to Item Description of Gap SRs Current Status I Comment Application Gap #26 To meet the requirements of SR HR-A2, the following HR-A2 Partially Resolved - The process did Not significant.

would be developed as supporting documentation for not include a procedure review but did The pre-initiator PBAPS: include a review of the need for assessment that transmitter/trip unit components to exists is

- A list of the PRA systems to consider for mis-calibration function properly, or where false adequate for this actions signals could prematurely terminate application,

- Rules for identifying and screening mis-calibration actions the system function. given the from the PRA evaluation that

- A list of procedures reviewed, the potential mis-calibration has been actions associated with the procedures, and the disposition performed and of the action (screened or evaluated). reflected in the model.

Gap #27 To meet the requirements of SR HR-A3, the following HR-A3 Partially Resolved - The process did Not significant.

would be developed as supporting documentation for not include a procedure review but did The pre-initiator PBAPS: include a review of the need for assessment that

- A list of the PRA systems to consider for common cause transmitter/trip unit components to exists is mis-calibration actions function properly, and common cause adequate for this mis-calibrations were also included. application,

- Rules for identifying and screening common cause mis- given the calibration actions from the PRA evaluation that

- A list of procedures reviewed, the potential common has been cause mis-calibration actions associated with the performed and procedures, and the disposition of the action (screened or reflected in the evaluated). model.

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Peach Bottom SLC CT Extension Table 4-2 Status of Identified Gaps to Capability Category II of the ASME PRA Standard Applicable Importance to Item Description of Gap SRs Current Status I Comment Application Gap #31 Establish the 'significant' pre-initiator HFEs based on the HR-D2 Partially Resolved - Pre-initiators None. The DG-1122 definition, and re-quantify the balance of the were included in the system models updated PRA significant HFEs using the methodologies outlined in as described in the system notebooks model meets PB02AF-003. that were created as part of the 2005 HR-D2 at update. The process included a Capability review of the need for transmitter/trip Category I, unit components to function properly, which is or where false signals could sufficient for this prematurely terminate the system application.

function.

However, not all significant pre-initiators were evaluated with a detailed HEP analysis. Rather, they l

were assigned a Itype based on the transmitter it is associated with, and the types were assigned an HEP value based on the limited set of detailed pre-initiator evaluations that were performed as described in the HRA notebook.

Gap #54 Document and employ the methodology used for DA-C6 Open - For the most part, the Not significant.

determining the standby component number of demands to estimated demands were determined The model is include plant specific: a) surveillance tests, b) maintenance from the Maintenance Rule database, reasonably acts, c) surveillance tests or maintenance on other but a confirmation that it is collected consistent with components, d) operational demands. Additional demands exactly consistent with the DA-C6 data from the from post-maintenance testing should not be included. requirements has not been performed. plant MR This is judged to have a minimal database, which impact on the Bayesian updated is adequate for reliability values utilized in the model. this application.

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Peach Bottom SLC CT Extension Table 4-2 Status of Identified Gaps to Capability Category II of the ASME PRA Standard Applicable Importance to Item Description of Gap SRs Current Status I Comment Application Gap #31 Establish the 'significant' pre-initiator HFEs based on the HR-D2 Partially Resolved - Pre-initiators None. The DG-1122 definition, and re-quantify the balance of the were included in the system models updated PRA significant HFEs using the methodologies outlined in as described in the system notebooks model meets PB02AF-003. that were created as part of the 2005 HR-D2 at update. The process included a Capability review of the need for transmitter/trip Category I, unit components to function properly, which is or where false signals could sufficient for this prematurely terminate the system application.

function.

However, not all significant pre-initiators were evaluated with a detailed HEP analysis. Rather, they were assigned a 'type' based on the transmitter it is associated with, and the types were assigned an HEP value based on the lim ited set of detailed pre-initiator evaluations that were performed as described in the HRA notebook.

Gap #55 To be consistent with SR DA-C8, the PBAPS PRA would DA-C8 Open - Note that Category I allows for None. Capability need to be enhanced to include reviews of operating estimates of standby status estimates Category I is experience to determine the times that components were in as an acceptable approach. met, which is standby. adequate for this application.

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Peach Bottom SLC CT Extension Table 4-2 Status of Identified Gaps to Capability Category II of the ASME PRA Standard Applicable Importance to Item Description of Gap SRs Current Status I Comment Application Gap #58 Ensure that the enhancements associated with DA-C4 DA-C11 Open - The maintenance rule data is Not significant.

include the guidance regarding the definition of used directly, but a confirmation that it The model is maintenance hours that is provided in SR DA-C11 and that collected exactly consistent with the reasonably the counting of unavailability hours follows that definition. DA-C11 requirements has not been consistent with performed. This is judged to have a data from the minimal impact on the unavailability plant MR hours used in the model. database, which is adequate for this application.

Gap #59 Ensure that the enhancements associated with DA-C4 DA-C12 Open - The maintenance rule data is Not significant.

include 1) the guidance regarding the treatment of used directly, but a confirmation that it The model is maintenance hours vs. plant operational status that is collected exactly consistent with the reasonably provided in SR DA-C12 (and ensure that the counting of DA-C11 requirements has not been consistent with unavailability hours follows that definition); and 2) perform performed. This is judged to have a data from the interviews of maintenance staff for equipment with minimal impact on the unavailability plant MR incomplete or limited maintenance information. hours used in the model. database, which is adequate for this application.

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Peach Bottom SLC CT Extension Table 4-2 Status of Identified Gaps to Capability Category II of the ASME PRA Standard Applicable Importance to Item Description of Gap SRs Current Status I Comment Application Gap #60 To be consistent with SR DA-C13, the PRA should include DA-C13 Partially resolved - Model includes Not significant.

an examination of coincident outage times for redundant coincident outage times for a few The model is equipment (both intra- and inter-system) and incorporate pertinent combinations (e.g. reasonably the results into the modeling and documentation. However, HPCI/RCIC, RHR Loops), but since consistent with it is judged that it is not practical to model all potential no known overlap existed for these known plant combinations of coincident maintenance unavailability combinations, an arbitrarily small operating values, and that a review of maintenance experience would value (1.0E-5) was assigned. practice and not be sufficient to allow the prediction of the dominant risk experience. An contributor combinations. As such, the approach suggested exhaustive is to identify dom inant risk contributor com binations based It is judged that the incorporation of assessm ent is on knowledge of the accident sequences modeling, and coincident maintenance terms will not needed to model such combinations of coincident maintenance have a minimal impact on the results support use of outages in the fault tree logic. A review of recent of the model. the PRA for this maintenance experience would then be performed to application.

identify events of coincident maintenance outages for these combinations to support probability estimation for the events.

Gap #65 During the plant specific data update, ensure the data used DA-D7 Partially resolved - The Com ponent None. The reflects the current design and operating conditions. Data Notebook includes development process used is Include guidance in the documentation related to updating of the updated plant-specific data appropriate.

data when changes are made to equipment or operating evaluation. Specific guidance on This is a conditions. updating data when changes are documentation made is not provided, but providing issue only.

these definitions should not have an impact on the quantitative results from the PRA model.

30 C467090020-8958-1/8/2010

Peach Bottom SLC CT Extension Table 4-2 Status of Identified Gaps to Capability Category II of the ASME PRA Standard Applicable Importance to Item Description of Gap SRs Current Status I Comment Application Gap #67 The PBAPS PRA appropriately includes a number of IF-F* Open - Internal Flood analysis being Not significant.

internal flood initiators and associated event trees (refer to updated in 2008. The updated Section 9 of the main documentation). The internal flooding internal flooding analysis needs to be expanded into a single analysis that has comprehensive analysis, and updated where appropriate. not yet been Flooding documentation needs to be upgraded especially integrated into for walkdowns and descriptions of calculations supporting the updated the quantitative analysis. model indicates that the contribution from internal flooding initiators to the internal events CDF and LERF risk metrics are still relatively small (Le., <10%

total contribution to CDF and LERF).

Gap #68 Identify the PRA modeled SSCs in flood areas per IF-A* Open -Internal Flood analysis being Not significant.

requirements of IF-A2 and IF-A3. updated in 2008. See Gap #67.

Gap #69 Identify and document potential flood sources for areas that IF-B* Open - Internal Flood analysis being Not significant.

do not screen out per the requirements in IF-B1, B2, B3, updated in 2008. See Gap #67.

and B4.

Gap #70 Identify and document scenarios, propagation paths, and IF-C* Open - Internal Flood analysis being Not significant.

affected SSCs per the requirements in IF-C1, C2, C3, and updated in 2008. See Gap #67.

C4.

31 C467090020-8958-1 /8/2010

Peach Bottom SLC CT Extension Table 4-2 Status of Identified Gaps to Capability Category II of the ASME PRA Standard Applicable Importance to Item Description of Gap SRs Current Status I Comment Application Gap #71 Identify human actions for flood mitigation and incorporate IF-C6 Open - Internal Flood analysis being Not significant.

into model per IF-C6 standards updated in 2008. See Gap #67.

Gap #72 Review and update flood frequencies per IF-02, 03, 04, IF-O* Open - Internal Flood analysis being Not significant.

and 05. updated in 2008. See Gap #67.

Gap #77 The uncertainty analysis could be further enhanced by QU-E2 Partially resolved - Sensitivity studies See Add #2 providing a discussion of the guidelines used to review included as part of the evaluation in below.

results and identify important contributors to uncertainty. Section 4.5 of the PB PRA Summary Use of a systematic process of identifying these areas and Notebook, but the choice of evaluating them may improve the overall quality of the sensitivities could be judged as not a analysis. systematic process. However, the QU-E2 SR definition has since changed - refer to Add #2 below.

Gap #80 Include an assessment of the significance of assumptions QU-F4 Open - Identification of key See Add #2 on the quantitative results. assumptions will be application below.

specific. Also, the QU-F4 SR has been redefined.

Gap #83 Strict reading of LE-E4 would indicate that the following LE-E4 Open - Level 2 Analysis is being Not significant.

enhancements to the LERF analysis and associated updated as part of the ongoing 2009 LERF is not an documentation would need to be made to comply with the PRA update process. important Standard: contributor in

- Explicitly assess dependencies among Level 2 HEPs (and this assessment.

combinations of Level 2 HEPs with Level 1 HEPs) The change in COF alone is

- Perform quantitative sensitivity studies of the LERF below the LERF analysis- Perform quantitative uncertainty assessment of acceptance the LERF analysis. guidelines.

32 C467090020-8958-1/8/2010

Peach Bottom SLC CT Extension Table 4-2 Status of Identified Gaps to Capability Category II of the ASME PRA Standard Applicable Importance to Item Description of Gap SRs Current Status I Comment Application Add #1 Addendum B of the ASME PRA Standard added SRs to QU-F6 Open - These new SRs will be None. This is a document the quantitative definition used for significant LE-G6 addressed during the next full PRA documentation basic event, significant cutset, significant accident model update, but providing these issue. The sequence, and significant accident progression sequence definitions should not have an impact model is not in the COF and LERF analysis. on the quantitative results from the being changed PRA model. to address this item.

Add #2 Several SRs associated with treatment of model QU-E1 Open - These recently redefined SRs An initial uncertainty and related model assumptions have been QU-E2 will be addressed during the next full assessment recently redefined. NRC has issued a clarification to its PRA model update after the NRC and based on the endorsement of the PRA Standard. NRC and EPRI are QU-E4 EPRI guidance becomes available. final EPRI currently preparing guidance on an acceptable process for QU-F4 guidance for the meeting these requirements. base PRA IE-03 model has been AS-C3 performed. The SC-C3 results of that assessment are SY-C3 factored into the HR-13 identification of potentially key OA-E3 assumptions for IF-F3 this application LE-G4 as described in Appendix B of this report.

33 C467090020-8958-1 /8/2010

Peach Bottom SLC CT Extension 4.7 GENERAL CONCLUSION REGARDING PRA CAPABILITY The PBAPS 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 SLC system.

Previously identified gaps to specific requirements in the ASME PRA Standard have been reviewed to determine which gaps might merit application-specific sensitivity studies in the presentation of the application results. No gaps were identified as needing specific sensitivity studies for this SLC CT extension request.

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Peach Bottom SLC 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 PBAPS TS for the SLC system to increase the CT from 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> when both SLC subsystems (Le., both trains) are inoperable.

The analysis examines a range of risk contributors as follows:

• The PBAPS FPIE PRA model is used to quantitatively address risk impacts.

• The FPI E assessment is judged to adequately capture risk contributors associated with low power plant operation.

• The SLC TS only applies to Modes 1 and 2. Shutdown and refueling modes (Modes 3, 4 and 5) are not applicable to the SLC TS.

• The interim Fire PRA model and other fire studies (e.g., NUREG/CR-6850) are used to provide qualitative and semi-quantitative insights, determining that fire hazards are negligible contributors.

• Seismic risk contributors are determined to be negligible based on qualitative insights from the NUREG-1150 study.

• Other External Event risks were found to be negligible contributors based on the PBAPS IPEEE.

5.2 PRA QUALITY The PRA quality has been assessed and determined to be adequate for this risk application, as follows:

• Scope - The PBAPS 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.

• Fidelity - The PBAPS PRA model (PB205C) is the most recent evaluation of the risk profile at PBAPS for FPIE challenges. The PRA reflects the as-built, as-operated plant.

• Standards - The PRA has been reviewed against the ASME PRA Standard

[Ref 5] 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 35 C467090020-8958-1 /8/20 10

Peach Bottom SLC CT Extension 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 commensurate with that needed to assess risk for this application.

5.3 QUANTITATIVE RESULTS VS. ACCEPTANCE GUIDELINES As shown in Table 5.3-1 below, the base results of the risk assessment indicate that the

~CDF, ICCDP, ~LERF, and ICLERP risk metric values are below the acceptance guidelines as defined in the corresponding risk significance guidelines from RG 1.174 and RG 1.177.

This analysis demonstrates that the proposed TS change satisfies the risk acceptance guidelines in RG 1.174 and RG 1.177, and therefore meets the intent of very small risk increases consistent with the Commission's Safety Goal Policy Statement.

Table 5.3-1 RISK ASSESSMENT BASE RESULTS Risk Metric Value(1) Acceptance Guidelines Reference

~CDF 2.3E-08/yr <1.0E-06/yr RG 1.174 ICCDP 2.3E-08 <5.0E-07 RG 1.177

~LERF 4.1E-10/yr <1.0E-07/yr RG 1.174 ICLERP 4.1 E-10 <5.0E-08 RG 1.177

5.4 CONCLUSION

S This analysis demonstrates the acceptability, from a risk perspective, of a change to the PBAPS TS for the SLC system to increase the CT from 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> when both SLC subsystems (Le., both trains) are inoperable.

This analysis demonstrates that the proposed TS change satisfies the risk acceptance guidelines in RG 1. 174 and RG 1. 177. This meets the intent of very small risk increases consistent with the Commission's Safety Goal Policy Statement.

Additionally, a PRA technical adequacy evaluation was performed consistent with the requirements of ASME PRA Standard, Addendum Band RG 1.200, Revision 1. This 36 C467090020-8958-1 /8/2010

Peach Bottom SLC CT Extension included a process to identify potential key sources of model uncertainty and related assumptions associated with this application. This resulted in the identification of issues that could both decrease and increase the calculated risk metrics. None of these identified sources of uncertainty were significant enough to change the conclusions from the risk assessment results presented here.

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Peach Bottom SLC CT Extension

6.0 REFERENCES

[1] RG 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk Informed Activities," Revision 1, January 2007.

[2] RG 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," Revision 1, November 2002.

[3] RG 1.177, "An Approach for Plant-Specific, Risk-Informed Decisionmaking:

Technical Specifications," August 1998.

[4] Exelon Risk Management Team, PB-PRA-013, Peach Bottom Atomic Power Station Probabilistic Risk Assessment Summary Notebook, PB205 and PB305 Models, Revision 1, July 2006.

[5] "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.

[6] Boiling Water Reactors Owners' Group, "BWROG PSA Peer Review Certification Implementation Guidelines," Revision 3, January 1997.

[7] Peach Bottom MSPI Basis Document, Rev. 2, March 27,2007.

[8] "Peach Bottom PRA Peer Review Using ASME PRA Standard Requirements,"

January 1998.

[9] Treatment of Parameter and Model Uncertainty for Probabilistic Risk Assessments, EPRI Report 1016737, Palo Alto, CA.

[10] Exelon Risk Management Team, Peach Bottom Atomic Power Station Fire Risk Analysis Summary Report, Revision 4, P0467060015-2778, April 2007.

[11] PECO Energy, "Peach Bottom Units 2 and 3 Individual Plant Examination for External Events," May 1996.

[12] "PRA Procedures Guide", NUREG/CR-2300, September 1981.

[13] "Analysis of Core Damage Frequency: Peach Bottom, Unit 2, External Events,"

NUREG/CR-4550, Volume 4, Revision 1, Part 3, Table 4.14, page 4-83.

[14] NUREG/CR-5042, "Evaluation of External Hazards to Nuclear Power Plants in the United States," December 1987.

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Peach Bottom SLC CT Extension

[15] Kennedy, R.P., et aI., "Capacity of Nuclear Power Plant Structures to Resist Blast Loading, II Sandia National Laboratories, NUREG/CR-2462, September 1983.

[16] NUREG/CR-5500, "Reliability Study: General Electric Reactor Protection System, 1984-1995, Volume 3" May 1999.

[17] Gorham, E.D., et aI., "Evaluation of Severe Accident Risks: Methodology for the Containment, Source Term, Consequence, and Risk Integration Analyses",

NUREG/CR-4551, December 1993.

[18] NUREG/CR-6850, EPRI Report 1011989, "Fire PRA Methodology for Nuclear Power Facilities", September 2005.

[19] Gorman, Thomas, BWR Owners; Group Assessment of IN 2007-07,10/16/2007

[20] "Guidance for Post-Fire Safe Shutdown Analysis", NEI 00-01, Rev. 2.

[21] Exelon, ER-AA-600-1046, "Risk Metrics - NOED and LAR", Revision 4.

[22] DG-1122, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk Informed Activities, II Draft Reg Guide, 2002.

[23] "Severe Accident Risks: An Assessment for Five U.S. Nuclear Power Plants",

NUREG-1150, December 1990.

[24] NUREG/CR-5088, IIFire Risk Scoping Study: Investigation of Nuclear Power Plant Fire Risk, Including Previously Unaddressed Issues,1I U.S. Nuclear Regulatory Commission, January 1989.

[25] FAQ 08-0051, "Hot Short Duration," June 2008, Draft, ADAMS Doc. #

ML083400188.

[26] ASMEIANS RA-Sa-2009, "Addenda to RA-S-2008, Standard for Level1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications," February 2009.

[27] RG 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk Informed Activities," Revision 2, March 2009.

39 C467090020-8958-1 /8/2010

Peach Bottom SLC CT Extension Appendix A External Event Assessment A.1 INTRODUCTION This appendix discusses the external events assessment in support of the PBAPS SLC System CT extension risk assessment. This appendix uses as the starting point of this assessment the external event work documented in the PBAPS IPEEE [Ref. A-1].

A.2 EXTERNAL EVENT ASSESSMENT The purpose of this portion of the assessment is to screen the spectrum of external event challenges to determine which external event hazards should be explicitly addressed as part of the PBAPS SLC System CT extension risk assessment.

Seismic There is no currently maintained quantitative Seismic PRA for Peach Bottom. Section A.3 discusses seismic ATWS insights from the Peach Bottom IPEEE [Ref. A-1] and NUREG-1150.

Internal Fires This internal fire assessment is based on the Peach Bottom Interim Fire PRA Model

[Ref. A-3] and generic assessments in NUREG/CR-6850 and the BWROG assessment of IN 2007-07. This assessment is discussed in Section A.4.

Other External Hazards In addition to internal fires and seismic events, the Peach Bottom IPEEE analysis of high winds, external floods, and other external hazards (HFO) was accomplished by reviewing the plant environs against regulatory requirements regarding these hazards.

Since both PBAPS units were designed (with construction started) prior to the issuance of the 1975 Standard Review Plan (SRP) criteria, PECO (now Exelon) performed a plant hazard and design information review for conformance with the SRP criteria. For seismic and fire events that were not screened out, additional analyses were performed to determine whether or not the hazard frequency was acceptably low. HFO events were screened out by compliance with the 1975 SRP criteria. As such, these hazards A-1 C467090020-8958-1 /8/2010

Peach Bottom SLC CT Extension were determined in the Peach Bottom IPEEE to be negligible contributors to overall plant risk.

No significant quantitative contribution from these external events was identified by the IPEEE evaluations. The compensatory actions and risk insights in this LAR are also judged applicable to qualitatively reduce the risk associated with these events.

Conclusions of Screening Assessment Given the foregoing discussions, external hazards are assessed to be negligible contributors to plant risk. Explicit treatment of these other external hazards is not necessary for most PRA applications (including the SLC System CT extension risk assessment) and would not provide additional risk-informed insights for decision making.

Further information is presented in this appendix to further justify the screening of Fire and Seismic hazards.

A.3 SEISMIC ASSESSMENT There is no currently maintained quantitative Seismic PRA for PBAPS. The following sections discuss seismic ATWS insights from the Peach Bottom IPEEE and NUREG-1150.

A.3.1 Peach Bottom Seismic IPEEE Overview The objective of the IPEEE [Ref A-1] Seismic Margin Analysis (SMA) was to rank each plant component in terms of its seismic capacity. In general, PBAPS equipment was found to be seismically rugged. The IPEEE did not evaluate specific seismic impacts associated with the SLC system, but seismic impacts on SLC system components would be similar to the seismic impacts on CRD. Per other studies (see below), seismic induced ATWS sequences are generally found to be negligible contributors.

A.3.2 PBAPS NUREG-1150 Seismic Overview The NUREG/CR-4551 study completed an update of the NUREG-1150 severe accident analysis for five nuclear power plants, including the Peach Bottom Atomic Power Station. This analysis addressed both internal and external events, including seismic initiators.

The NUREG/CR-4551 PBAPS seismic analysis screened seismic-induced ATWS accident sequences as non-significant contributors <<1%) to the plant seismic CDF.

A-2 C467090020-8958-1/8/2010

Peach Bottom SLC CT Extension A.3.3 Seismic Risk Impact Conclusion Based on the preceding discussions, it is concluded that the risk of a seismically induced ATWS is negligible; therefore the risk a coincident inoperability of both SLC subsystems along with the seismically induced ATWS is further negligible. Based on these results, it is judged that seismic-induced ATWS accident sequences are non-significant contributors to the Peach Bottom seismic CDF.

A.4 INTERNAL FIRES ASSESSMENT This internal fire assessment is based on the interim PBAPS Fire PRA (FPRA) model developed in 2007 and other industry efforts.

A.4.1 PBAPS Interim Fire PRA The current PBAPS FPRA [Ref. A-3] is an interim implementation of NUREG/CR-6850; that is, not all tasks identified in NUREG/CR-6850 are yet completely addressed or implemented due to the changing state-of-the-art of industry at the time of the 2007 PBAPS FPRA development.

NUREG/CR-6850 task limitations and other precautions regarding the FPRA upgrade for PBAPS are as follows:

• Multiple Spurious Operation (MSO) Review (NUREG/CR-6850 Task 2)

- At the time of the 2007 PBAPS FPRA the BWR Owners' Group was developing a generic list of MSOs to be considered. No expert panel was used to identify specific MSO scenarios not already inherently addressed in the PRA. At future updates the BRWOG list should be reviewed, an expert panel should convene, and the results of each incorporated as necessary.

• Instrumentation Review (NUREG/CR-6850 Task 2) - The new requirements of NUREG/CR-6850 regarding the explicit identification and modeling of instrumentation required to support PRA credited operator actions is not addressed. The industry treatment for this task is still being developed.

• The Balance of Plant (NUREG/CR-6850 Task 2) - The BOP is not fully treated. BOP support system failure is conservatively assumed in most areas. Additional modeling could be conducted to reduce the fire CDF due to this assumption if time and funding is available in future updates.

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Peach Bottom SLC CT Extension

• Large Early Release Frequency (LERF) (NUREG/CR-6850 Task 2) -

LERF is not considered. LERF is expected to be addressed in future updates.

• Limited Analysis Iterations (NUREG/CR-6850 Task 9-12) - The process of conducting a FPRA is iterative, identifying conservative assumptions and high risk compartments and performing analyses to refine the assumptions and reduce those compartment risks. The ability to conduct iterations is limited based on resources. The scenarios developed for the 2007 PBAPS FPRA may benefit from further refinement as necessary for application or for future updates.

• Multi-Compartment Review (NUREG/CR-6850 Task 11) - This subtask reviews the fire analysis compartment boundaries to ensure they are sufficiently robust to prevent the spread of fire between FPRA analysis compartments or that such propagations are adequately addressed by the developed scenarios. The design and plant layout of PBAPS make fire propagation to multiple compartments unlikely compared to the fire risk in individual compartments. Therefore, an explicit multi-compartment review was not performed.

• Seismic Fire Interactions (NUREG/CR-6850 Task 13) - This task reviews previous assessments to identify any specific interaction between suppression system and credited components or adverse impact of fire protection system interactions that should be accounted for in the FPRA. This has not been performed to support this FPRA.

• Uncertainty and Sensitivity Analysis (NUREG/CR-6850 Task 15) - This task explores the impacts of possible variation of input parameters used in the development of the model and the inputs to the analysis on the FPRA results. This task is not currently addressed because the industry is still developing an appropriate methodology.

Some limitations of these items are:

• Item 1(MSO), represents a source of additional fire CDF contribution (Le., if the BWROG MSO list includes MSOs not addressed in this update).

• Item 2 (Instrumentation Review) represents a potential additional fire CDF contribution that cannot be estimated at this time since the methodology is not established.

• Items 3 (BOP) and 8 (Uncertainty) are potential sources of conservatism in the results.

• Item 4 (LERF) is a future scope issue not affecting the fire CDF model.

• Items 5 (Iterations) and 6 (Multi-compartment) represent modeling A-4 C467090020-8958-1 /8/2010

Peach Bottom SLC CT Extension assumptions that should be reviewed with each FPRA application to determine their applicability and/or potential impact on the decision.

• Item 7 (Seismic) is a FPRA application completeness issue for which the methodology is not yet established.

It is noted that the PBAPS FPRA model is a conservative model, resulting in a skewing of the total reported plant CDF towards the upper bound. This is contrasted with the internal events plants PRA which provides a best estimate (mean) CDF.

Given the above, the PBAPS FPRA model is judged to provide a meaningful representation of fire CDF contributors, and is appropriate for use in risk-informed decision-making, to the extent that these limitations are recognized and addressed in each application, as appropriate. The model is, however, "interim" due to the stated limitations.

The PBAPS FPRA model did not credit SLC as a safe shutdown system. ATWS sequences were also not developed in creation of the event tree model:

'~ TWS sequences are not postulated for fire events. The node itself is not modified, but the transfer is not developed for the failure path of the node.

Failures of the node do not contribute to CCDP." [Ref. A-3].

Based on the preceding discussion, the SLC system does not contribute to the overall fire PRA model results as fire challenges that also result in ATWS were screened from incorporation into the fire PRA model consistent with the guidance provide in NUREG/CR-6850 as discussed below in Section A.4.2.

A.4.2 NUREG/CR-6850 Screening NUREG/CR-6850, Volume 2, Section 2.5.1 (page 2-7) [Ref. A-4] provides the following directions for selecting components and accident scenarios to be examined in an internal fire PRA:

"The types of sequences that could generally be eliminated from the PRA include the following... Sequences associated with events that, while it is possible that the fire could cause the event, a low-frequency argument can be justified. For example, it can often be easily demonstrated that anticipated transient without scram (A TWS) sequences do not need to be treated in the Fire PRA because fire-induced failures will almost certainly remove power from the control rods (resulting in a trip), rather than cause a "failure-to-scram" condition. Additionally, fire frequencies multiplied by the independent failure-to-scram probability can usually be argued to be small contributors to fire risk."

A-5 C467090020-8958-1 /8/2010

Peach Bottom SLC CT Extension As can be seen from the NUREG/CR-6850 excerpt above, fire-induced ATWS contributors are generally acknowledged as non-significant contributors to the fire risk profile.

A.4.3 BWROG Position on Fire-Induced Failure to Scram Fire scenarios that could threaten the function of the reactor protection system have been addressed in a BWROG assessment (refer to Appendix C) of NRC Information Notice 2007-07. [Ref. A-2] The assessment outlines the types of scenarios in which a fire could energize a circuit through a "hot short" that would compromise scram capabilities. The assessment also indicates that there are multiple actions that would have to occur in conjunction to the very specific fire scenarios for function to be lost.

The assessment concluded that these scenarios are of low-likelihood, low safety-significance, and have multiple layers of defense-in-depth which would either prevent the condition, or adequately mitigate it.

A.4.4 Fire Risk Impact Conclusion Based on the preceding discussions, it is concluded that fire-induced ATWS is a non-significant contributor to the plant risk profile and thus does not impact the decision-making of the proposed PBAPS SLC CT extension.

A-6 C467090020-8958-1 /8/2010

Peach Bottom SLC CT Extension REFERENCES

[A-1] PECO Energy, "Peach Bottom Units 2 and 3 Individual Plant Examination for External Events," May 1996.

[A-2] Gorman, Thomas, BWROG Assessment of IN 2007-07,10/16/2007.

[A-3] Exelon Risk Management Team, Peach Bottom Atomic Power Station Fire Risk Analysis Summary Report, Revision 4, P0467060015-2778, April 2007.

[A-4] NUREG/CR-6850, EPRI Report 1011989, "Fire PRA Methodology for Nuclear Power Facilities", September 2005.

A-7 C467090020-8958-1 /8/2010

Peach Bottom SLC CT Extension AppendixB Uncertainty Analysis This appendix evaluates uncertainties that could impact the SLC CT extension assessment. Section 8.1 and 8.2 evaluate model uncertainties. Section 8.3 evaluates parametric uncertainty.

• Section 8.1 provides P8APS specific modeling uncertainty evaluations for the 8ase Case.

• Section 8.2 provides an examination of the specific cutsets that affect the change in the CDF risk metric associated with the change in the SLC CT.

8.1 MODEL UNCERTAINTIES

SUMMARY

Postulated key modeling uncertainties are identified through a systematic structured process. Table 8-1 presents the candidate key modeling uncertainties for the P8205C model. The five modeling uncertainties that rise to the definition of a key model uncertainty are summarized in Table 8-2 along with the impacts on the CDF risk metric.

It is noted that none of these cases evaluates modeling issues associated with the SLC system or ATWS sequences.

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Peach Bottom SLC CT Extension TABLE B-1

SUMMARY

OF SENSITIVITY CASES TO IDENTIFY RISK METRIC CHANGES ASSOCIATED WITH CANDIDATE MODELING UNCERTAINTIES PERCENT CHANGE FROM CASE CDF BASE DESCRIPTION Base 4.06E-06 0% Base Case 1-1 3.31E-06 -18.5% The im pact of reducing the loss of offsite power initiating event frequency to 0.030/yr 1-2 4.21E-06 3.70/0 The impact of increasing the loss of offsite power initiating event frequency to 0.056/yr 2-1 3.49E-06 -14.1% The impact of reducing the 'mission time' for the Emergency Diesel Generators to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 2-2 7.92E-06 95.1% The impact of increasing the 'mission time' for the Emergency Diesel Generators to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 3-1 3.25E-06 -20.0% The impact setting all common cause failures to tti their 5 percentile value 3-2 6.67E-06 64.30/0 The impact setting all common cause failures to th their 95 percentile value 4-1 2.13E-06 -47.5% The impact of removing all dependent HEPs from the model 4-2 1.02E-06 -74.9% The impact of removing all dependent and independent HEPs from the model 5 3.95E-06 -2.7% The impact of allowing for the potential of ECCS success following containment venting 6 3.57E-06 -12.1% The impact of allowing for the potential of fire system injection to the RPV 7 6.19E-06 +52.5% The impact of not crediting use of the Conowingo SBO Line in non-SBO scenarios 8-1 3.15E-06 -22.4% The im pact of setting the dependent operator action between aligning the SBO Line and cross-tieing the 4kV buses to its lower bound value of 5.5E-4 B-2 C467080016-841 0-1/8/2010

Peach Bottom SLC CT Extension TABLE B-1

SUMMARY

OF SENSITIVITY CASES TO IDENTIFY RISK METRIC CHANGES ASSOCIATED WITH CANDIDATE MODELING UNCERTAINTIES PERCENT CHANGE FROM CASE CDF BASE DESCRIPTION 8-2 4.71E-06 +16.0% The impact of setting the dependent operator action between aligning the SBO Line and cross-tieing the 4kV buses to its upper bound value of 7.5E-3 9 3.76E-06 -7.4% The impact of allowing credit for ECW open loop mode cooling 10 3.73E-06 -8.1% The impact of allowing credit for use of SE-11, Attachment W in non-Loop scenarios Table B-2 FIVE KEY MODELING UNCERTAINTY CASES CASE CDF PERCENT CHANGE DESCRIPTION FROM BASE 2-2 7.92E-06 95.1% The impact of increasing the 'mission time' for the Emergency Diesel Generators to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 3-2 6.67E-06 64.3% The impact setting all common cause failures to their th 95 percentile value 4-1 2.13E-06 -47.5% The impact of removing all dependent HEPs from the model 4-2 1.02E-06 -74.9% The impact of removing all dependent and independent HEPs from the model 7 6.19E-06 +52.5% The impact of not crediting use of the Conowingo SBO Line in non-SBO scenarios B-3 C467080016-841 0*1/8/2010

Peach Bottom SLC CT Extension B.2 MODEL UNCERTAINTIES ASSOCIATED WITH SLC SYSTEM OUT OF SERVICE To determine the relative importance of individual contributors for this SLC CT extension, the focus needs to be on the results of the CDF assessment for the SLC system out-of-service. To obtain insights regarding this change to the base case results, the first step is to take the SLC out-of-service case cutsets and remove the base case cutsets. This is done in CAFTA through the delete term function of the cutset editor. The result of this process are cutsets that are unique to the SLC out-of-service case and do not appear in the base case. These cutsets can be used to determine information regarding significant accident sequences or cutsets that drive the delta-CDF assessment.

Table B-3 presents the top ten cutsets for the delta-CDF assessment. Table B-4 presents the most important contributors to the delta-CDF assessment sorted by the Fussell-Vesely importance measure.

Tables B-3 and B-4 show that the Scram system hardware failure is the most important contributor for the SLC system out of service case. The top ten cutsets are exclusively failures of the Scram system associated with various initiating events. Of the events with a Fussell-Vesely greater than 2E-2 (>2% contribution to CDF), other than the Scram system failure and initiating events, these dominant contributors are primarily HEPs.

It can be concluded that the CDF is dominated by failures of the Scram system. The basic events used to model the Scram system failures are already considered in the base uncertainty assessment. The contributions of HEPs are also considered in the base uncertainty assessment.

Because of the large potential impact of the mechanical failure to scram probability on the assessment of the risk metrics for this application, it is prudent to perform a sensitivity recognizing the uncertainty in the mechanical common cause failure to scram probability.

This sensitivity is performed by including the 95% upper bound on the common cause mechanical scram failure probability in both the base case and the case with the SLC system set to TRUE.

The results of the sensitivity case are shown in Table B-5.

Based on the results of the sensitivity analysis, it is found that the acceptance criteria are all met even for this extreme assumption regarding the common cause mechanical scram failure probability.

8-4 C467080016-841 0-1/8/2010

Peach Bottom SLC CT Extension Table B-3 TOP TEN CUTSETS FOR CDF FOR THE SLC SYSTEM OUT OF SERVICE Cutset Event Event

1. Prob Prob Description 1 1.87E-06 8.89E-01 IEnR TURBINE TRIP 2.10E-06 CM RPS MECHANICAL FAIL TO BYPASS THE MSIV RPV LOW LEVEL 1.00E+00 FHUBLMSVDXI2 INTERLOCK (LEVEL 1) 2 2.23E-07 1.06E-01 IETCV LOSS OF CONDENSER VACUUM 2.10E-06 CM RPS MECHANICAL 3 1.38E-07 6.58E-02 IETF LOSS OF FEEDWATER 2.10E-06 CM RPS MECHANICAL 4 1.08E-07 5.15E-02 IETE LOSS OF OFFSITE POWER 2.10E-06 CM RPS MECHANICAL 5 8.74E-08 4.16E-02 IETI FREQUENCY OF IORV TRANSIENTS 2.10E-06 CM RPS MECHANICAL 6 7.65E-08 8.89E-01 IEnR TURBINE TRIP 2.10E-06 CM RPS MECHANICAL COGNITIVE ERROR FORLEVEL / POWER EARLY 4.10E-02 ZHUPWLVLDXI2 IN ANATWS 7 7.20E-08 3.43E-02 IETM MSIV CLOSURE 2.10E-06 CM RPS MECHANICAL 8 6.68E-08 8.89E-01 IEnR TURBINE TRIP OPEN SRV DURING NON-ISOLATION ATWS FAILS 3.58E-02 APHATWNIDKI2 TO RECLOSE 2.10E-06 CM RPS MECHANICAL 9 2.61 E-08 8.89E-01 IEnR TURBINE TRIP 2.10E-06 CM RPS MECHANICAL EXECUTION ERROR FORLEVEL / POWER EARLY 1.40E-02 ZHUFWHPLDXI2 IN AN ATWS 10 1.87E-08 8.89E-01 IEnR TURBINE TRIP 2.10E-06 CM RPS MECHANICAL MSIVS FAIL TO REMAIN OPEN DURING 1.00E-02 FPH--MSTDXI2 TRANSIENT 8-5 C467080016-841 0-1/8/2010

Peach Bottom SLC CT Extension Table B-4 BASIC EVENT IMPORTANCE MEASURES FOR CDF ASSESSMENT FOR SLC OUT OF SERVICE Event Name Probability Fus Ves Description CM 2.10E-06 1.00E+OO RPS MECHANICAL IETTR 8.89E-01 7.28E-01 TURBINE TRIP FAIL TO BYPASS THE MSIV RPV LOW LEVEL INTERLOCK FHUBLMSVDXI2 1.00E+OO 6.71 E-01 (LEVEL 1)

IETCV 1.06E-01 7.77E-02 LOSS OF CONDENSER VACUUM IETF 6.58E-02 4.83E-02 LOSS OF FEEDWATER IETE 5.15E-02 3.79E-02 LOSS OF OFFSITE POWER IETI 4.16E-02 3.05E-02 FREQUENCY OF IORV TRANSIENTS COGNITIVE ERROR FORLEVEL / POWER EARLY IN AN ZHUPWLVLDXI2 4.10E-02 2.75E-02 ATWS IETM 3.43E-02 2.52E-02 MSIV CLOSURE OPEN SRV DURING NON-ISOLATION ATWS FAILS TO APHATWNIDKI2 3.58E-02 2.40E-02 RECLOSE EXECUTION ERROR FORLEVEL / POWER EARLY IN AN ZHUFWHPLDXI2 1.40E-02 9.39E-03 ATWS IETBCCW 1.00E+OO 8.82E-03 LOSS OF TBCCW INITIATING EVENT FLAG IENSW 1.00E+OO 8.27E-03 LOSS OF SW INITIATING EVENT FLAG IERBCCW 1.00E+OO 7.73E-03 LOSS OF RBCCW INITIATING EVENT FLAG IEIAS 1.00E+OO 7.33E-03 LOSS OF IA INITIATING EVENT FLAG IE COMMON CAUSE FAILURE OF RBCCW PUMPS FAIL MPMXP010CRIE2 8.21 E-03 6.71 E-03 TO RUN FPH--MSTDXI2 1.00E-02 6.71 E-03 MSIVS FAIL TO REMAIN OPEN DURING TRANSIENT FPH-RBVSDXI2 1.00E-02 6.71 E-03 FAILURE OF REACTOR BLDG VENTILATION SYSTEM IES2 8.49E-03 6.23E-03 FREQUENCY OF SMALL LOCA IE COMMON CAUSE FAILURE OF TBCCW PUMPS FAIL TPMXP144CRIE2 8.21 E-03 6.02E-03 TO RUN OPERATORS FAIL TO START STANDBY NSW PUMP WHU--NSWDXI2 1.00E+OO 4.53E-03 (EARLY)

WPN--ABCCRI E2 5.89E-03 4.32E-03 CCF OF ALL 3 NSW PUMPS TO RUN CCF OF RBCCW AND TBCCW PUMPS TO RUN (INIT ZPM-TBRBCRIE2 3.33E-03 2.44E-03 EVENT)

ICMDK001 HRIE2 5.84E-01 2.44E-03 COMPRESSOR 2DK001 FAILS TO RUN WALAPOO4XY2 3.33E-01 2.21 E-03 FRACTION OF STANDBYTIME FOR NSW PUMP 2APOO4 IFL14ALLCEIE2 4.08E-03 2.15E-03 IE CCF PLUGGING OF 2A/B/CF014 FILTER/SCREENS IEFL42A2 2.41 E-03 1.97E-03 TURB BLDG FLOOD CAUSES PLANT TRIP CONSEQUENTIAL LOSS OF OFFSITE POWER GIVEN EPHOSPTRIPO 2.40E-03 1.62E-03 PLANT TRIP IETDC2B 1.50E-03 1.62E-03 LOSS OF 125V DC CHANNEL 2B 8-6 C467080016-841 0-1/8/2010

Peach Bottom SLC CT Extension Table B-4 BASIC EVENT IMPORTANCE MEASURES FOR CDF ASSESSMENT FOR SLC OUT OF SERVICE Event Name Probability Fus Ves Description ICM-ABCDCRIE2 2.03E-03 1.49E-03 IA COMPRESSORS A, B, C, AND D CCF [4 OF 4]

WALBP004XY2 3.33E-01 1.46E-03 FRACTION OF STANDBYTIME FOR NSW PUMP 2BP004 WALCP004XY2 3.33E-01 1.46E-03 FRACTION OF STANDBYTIME FOR NSW PUMP 2CP004 IETACBUSE12 1.68E-03 1.42E-03 LOSS OF 4KV AC BUS E12 (20A15)

IETACBUSE32 1.68E-03 1.38E-03 LOSS OF 4KV AC BUS E32 (20A17)

IETACBUSE42 1.68E-03 1.38E-03 LOSS OF 4KV AC BUS E42 (20A18)

IETACBUSE22 1.68E-03 1.37E-03 LOSS OF 4KV AC BUS E22 (20A16)

IETDC2A 1.50E-03 1.22E-03 LOSS OF 125V DC CHANNEL 2A IETDC2C 1.50E-03 1.22E-03 LOSS OF 125V DC CHANNEL 2C IETDC2D 1.50E-03 1.22E-03 LOSS OF 125V DC CHANNEL 2D ICMAK001 HRIE2 5.84E-01 9.90E-04 COMPRESSOR 2AK001 FAILS TO RUN IEFL62 1.20E-03 9.80E-04 RBCCW RM-HPCI,RCIC,B CS,B RHR LOOPS FAIL ICMBK001 HRIE2 5.84E-01 8.88E-04 COMPRESSOR 2BK001 FAILS TO RUN IFLAF014HEIEO 8.39E-02 8.63E-04 PLUGGING OF OAF014 FILTER/SCREEN ICMCK001 HRIE2 5.84E-01 8.35E-04 COMPRESSOR 2CK001 FAILS TO RUN WPN--ABXCRIE2 1.84E-03 6.89E-04 CCF OF A AND B NSW PUMPS TO RUN (INIT EVENT)

WPN--AXCCRIE2 1.84E-03 6.89E-04 CCF OF A AND C NSW PUMPS TO RUN (INIT EVENT)

WPN--XBCCRIE2 1.84E-03 6.89E-04 CCF OF BAND C NSW PUMPS TO RUN (INIT EVENT)

TBP--VLVCWI2 1.00E-03 6.60E-04 TURBINE BYPASS VALVES COMMON CAUSEFAILURE ICM-ABCXCRIE2 1.23E-03 6.43E-04 IA COMPRESSORS A, B, AND C CCF [3 OF 4]

WCV1762ADNI2 1.00E-03 5.96E-04 CHECK VALVE 2-30-1762A FAILS TO CLOSE WCV1762BDNI2 1.00E-03 5.96E-04 CHECK VALVE 2-30-1762B FAILS TO CLOSE WCV1762CDNI2 1.00E-03 5.96E-04 CHECK VALVE 2-30-1762C FAILS TO CLOSE ICM-ABXDCRIE2 1.23E-03 5.69E-04 IA COMPRESSORS A, B, AND D CCF [3 OF 4]

ICM-AXCDCRIE2 1.23E-03 5.67E-04 IA COMPRESSORS A, C, AND D CCF [3 OF 4]

ICM-XBCDCRIE2 1.23E-03 5.67E-04 IA COMPRESSORS B, C, AND D CCF [3 OF 4]

WPNAP004HRIE2 2.44E-01 5.19E-04 MOTOR DRIVEN PUMP NSW 2AP004 FAILS TO RUN WPNBP004HRIE2 2.44E-01 4.60E-04 MOTOR DRIVEN PUMP NSW 2BP004 FAILS TO RUN WPNCP004HRIE2 2.44E-01 4.60E-04 MOTOR DRIVEN PUMP NSW 2CP004 FAILS TO RUN MALAP010XX2 5.00E-01 3.88E-04 FRACTION OF RUNNINGTIME FOR RBCCW PUMP2AP010 MALBP010XX2 5.00E-01 3.88E-04 FRACTION OF RUNNINGTIME FOR RBCCW PUMP2BP010 MPMAP010HRIE2 4.29E-02 3.84E-04 RECW PUMP 2AP10 FAILS TO CONTINUE TO RUN MPMBP010HRIE2 4.29E-02 3.84E-04 RBCCW PUMP 2BP10 FAILS TO CONTINUE TO RUN MPMBP010TM2 1.00E-02 3.57E-04 RBCCW PUMP B UNAVAILABLE DUE TO TEST/MAINT B- TEST/MAINT. FACTOR TO CORRECT FOR A- RUN MALBP010TM2 2.00E+00 3.53E-04 FACTOR MALAP010TM2 2.00E+00 3.53E-04 A- TEST/MAINT. FACTOR TO CORRECT FOR B- RUN 8-7 C467080016-841 0-1/8/2010

Peach Bottom SLC CT Extension Table B-4 BASIC EVENT IMPORTANCE MEASURES FOR CDF ASSESSMENT FOR SLC OUT OF SERVICE Event Name Probability Fus Ves Description FACTOR MPMAP010TM2 1.00E-02 3.53E-04 RBCCW PUMP A UNAVAILABLE DUE TO TEST/MAINT EHB-0202DN12 5.00E-04 3.32E-04 CIRCUIT BREAKER 252-0202 (22) NO-FO ETR--2SUTMO 4.88E-04 3.24E-04 2SU FEED OUT FOR MAINTENANCE ICM-AXCXCRIE2 3.72E-03 2.82E-04 IA COMPRESSORS A AND C CCF [2 OF 4]

ICM-XBCXCRIE2 3.72E-03 2.82E-04 IA COMPRESSORS BAND C CCF [2 OF 4]

ICM-ABXXCRIE2 3.27E-03 2.56E-04 IA COMPRESSORS A AND B CCF [2 OF 4]

INSTRUMENT AIR COMPRESSOR 2DK001 FAILS TO ICMDK001 DSI2 2.00E-02 2.31E-04 START FLOOD AREA 42 CASE B - ENOUGH TO FAIL COND AND IEFL42B2 3.01 E-04 2.21 E-04 13KV FLOOD AREA 42 CASE C - ENOUGH TO FAIL IEFL42C2 3.01 E-04 2.21 E-04 CONDENSATE INSTRUMENT AIR COMPRESSOR 2AK001 FAILS TO ICMAK001 DSI2 2.00E-02 1.87E-04 START HEAT EXCHANGER 2AE018 AND 2BE018 CC MHX-E018CWIE2 2.23E-04 1.78E-04 RUPTURE/FAILURE INSTRUMENT AIR COMPRESSOR 2BK001 FAILS TO ICMBK001 DSI2 2.00E-02 1.78E-04 START HEAT EXCHANGER 2AE038 AND 2BE038 CC THX-E038CWIE2 2.23E-04 1.64E-04 RUPTURE/FAILURE INSTRUMENT AIR COMPRESSOR 2CK001 FAILS TO ICMCK001 DSI2 2.00E-02 1.57E-04 START 120V AC UPS 20D37 INVERTER / STATIC SWITCH EUP20D37HWI2 2.40E-04 1.56E-04 FAILURE ICM-AXXDCRIE2 3.72E-03 1.50E-04 IA COMPRESSORS A AND D CCF [2 OF 4]

ICM-XBXDCRIE2 3.72E-03 1.50E-04 IA COMPRESSORS BAND D CCF [2 OF 4]

ICM-XXCDCRIE2 3.72E-03 1.46E-04 IA COMPRESSORS C AND D CCF [2 OF 4]

FTU--ABCCRI2 2.02E-04 1.32E-04 PUMPS 2AP001 2BP001 AND 2CP001 [3 OF 31 ICMDK001TM2 1.00E-02 1.16E-04 COMPRESSOR 2DK001 UNAVAIL TESTING/MAINT.

APH--ARIDXI2 6.90E-05 1.12E-04 INDEPENDENT ARI HARDWARE FAILURES CE 3.70E-06 1.12E-04 RPS ELECTRICAL STUCK OPEN SRV FAILS TO RECLOSE ATLOW RPV APH--NRCDXI2 1.50E-01 1.11 E-04 PRESSURE APHNUMTMDXI2 1.60E+01 1.11 E-04 NUMBER OF SRV OPENINGS (MSIV CLOSURE EVENTS)

ARV--SRVDKI2 6.50E-03 1.11 E-04 SRV FAILS TO RECLOSE INITIALLY FAILURE TO RECOVER OSP IN 2.5 HRS / NORECOVERY NOOSP2E 5.50E-01 1.11 E-04 IN 0.5 HRS FAILURE TO RECOVER OFF-SITE POWER EARLY (30 NOOSPE 6.07E-01 1.11 E-04 MINUTES) 8-8 C467080016-841 0-1/8/2010

Peach Bottom SLC CT Extension Table B-4 BASIC EVENT IMPORTANCE MEASURES FOR CDF ASSESSMENT FOR SLC OUT OF SERVICE Event Name Probability Fus Ves Description RCVCL-1B 1.00E+00 1.11 E-04 RCV FOR DAMAGE CLASS 1B RCVSEQ-LP4-32 1.00E+00 1.11 E-04 RCV FOR SEQ LP4-32 IFL14ALLCEI2 4.08E-03 1.10E-04 CCF PLUGGING OF 2A/B/CF014 FILTER/SCREENS OPERATORS FAIL TO IMPLEMENT SE-11 ATTACHMENT MHUSE11 WDXI2 4.30E-02 1.03E-04 W ICMAK001TM2 1.00E-02 9.34E-05 COMPRESSOR 2AK001 UNAVAIL TESTING/MAINT.

ICMBK001TM2 1.00E-02 8.92E-05 COMPRESSOR 2BK001 UNAVAIL TESTING/MAINT.

ICMCK001TM2 1.00E-02 7.84E-05 COMPRESSOR 2CK001 UNAVAIL TESTING/MAINT.

JPHSPLEVHWIEO 9.99E-05 7.32E-05 LOW INTAKE POND LEVEL TALAP144XX2 5.00E-01 6.50E-05 FRACTION OF RUNNINGTIME FOR TBCCW PUMP2AP144 TALBP144XX2 5.00E-01 6.50E-05 FRACTION OF RUNNINGTIME FOR TBCCW PUMP2BP144 TPMAP144HRIE2 4.29E-02 6.50E-05 TBCCW PUMP 2AP144 FAILS TO CONTINUE TO RUN TPMBP144HREI2 4.29E-02 6.50E-05 TBCCW PUMP 2BP144 FAILS TO CONTINUE TO RUN MHU--RHXDXI2 1.00E-02 6.40E-05 OPERATOR FAILS TO ALIGN STANDBY RBCCW HX THU--THXDXI2 1.00E-02 6.40E-05 OPERATOR FAILS TO ALIGN STANDBY TBCCW HX EHB-0105DN12 5.00E-04 5.97E-05 CIRCUIT BREAKER 252-0105 (11) NO-FO WCV1762XCNI2 4.27E-05 5.91 E-05 CCF OF PUMP DISCHARGE CHECK VALVES TO CLOSE TEST/MAINTENANCE FACTOR TO CORRECT FOR WALAP004TM2 3.00E+00 5.86E-05 STANDBY FACTOR WPNAP004TM2 1.00E-02 5.86E-05 MOTOR DRIVEN PUMP NSW 2AP004 IN MAINTENANCE ETR--3SUTMO 4.88E-04 5.83E-05 3SU FEED OUT FOR TESTING MAINT.

BALAP039XXI2 5.00E-01 5.56E-05 FRACTION OF RUNNINGTIME FOR CRD PUMP 2AP039 BALBP039XXI2 5.00E-01 5.56E-05 FRACTION OF RUNNINGTIME FOR CRD PUMP 2BP039 OPERATOR FAILS TO CROSSTIE U21NSTRUMAIR IHUTRAINDXI2 1.00E-01 5.31 E-05 TRAINS FAILURE TO RECOVER EDG IN 2.5 HRS / NORECOVERY NODG2E 8.20E-01 5.03E-05 IN 0.5 HRS NODGE 1.00E+00 5.03E-05 FAILURE TO RECOVER DIESEL EARLY (30 MINUTES)

FLC-MSTRHWI2 7.20E-05 4.69E-05 FEEDWATER MASTER CONTROLLER FAILURE OPERATOR FAILS TO ALIGN CAD TANK TO UNIT 2 INS AHU--CADDXI2 3.60E-02 4.43E-05 IB I IFLBS013TM2 1.00E-03 4.28E-05 FILTER/DRYER PACKAGE 2BS013 TEST/MAl NT E324 TRANSFORMER BREAKER 152-1705 FAILS TO ECB-1705DN12 3.00E-04 4.10E-05 CLOSE E424 TRANSFORMER BREAKER 152-1806 FAILS TO ECB-1806DN12 3.00E-04 3.66E-05 CLOSE EDGOCG12TMO 1.58E-02 3.63E-05 DIESEL GENERATOR E3 - OCG12 TESTING OR MAINT.

TEST/MAINTENANCE FACTOR TO CORRECT FOR WALBP004TM2 3.00E+00 3.39E-05 STANDBY FACTOR 8-9 C467080016-8410-1/8/2010

Peach Bottom SLC CT Extension Table B-4 BASIC EVENT IMPORTANCE MEASURES FOR CDF ASSESSMENT FOR SLC OUT OF SERVICE Event Name Probability Fu~ VA~ . tion TEST/MAINTENANCE FACTOR TO CORRECT FOR WALCP004TM2 3.00E+00 3.39E-05 STANDBY FACTOR WPNBP004TM2 1.00E-02 3.39E-05 MOTOR DRIVEN PUMP NSW 2BP004 MAINTENANCE WPNCP004TM2 1.00E-02 3.39E-05 MOTOR DRIVEN PUMP NSW 2C004 IN MAINTENANCE ICV5986ADNI2 1.00E-03 3.37E-05 CHK VLV 36-45986A (WHEN OPEN) FAILS TO CLOSE ICV5986BDNI2 1.00E-03 3.37E-05 CHK VLV 36-45986B (WHEN OPEN) FAILS TO CLOSE EDGODG12TMO 1.58E-02 3.23E-05 DIESEL GENERATOR E4 - ODG12 TESTING OR MAINT.

MALAE018XX2 5.00E-01 3.20E-05 FRACTION OF TIME IN SERVICE FOR HX 2AE018 MALBE018XX2 5.00E-01 3.20E-05 FRACTION OF TIME IN SERVICE FOR HX 2BE018 MHXAE018HWIE2 8.72E-03 3.20E-05 HEAT EXCHANGER A 2AE018 RUPTURES/FAILS MHXBE018HWEI2 8.72E-03 3.20E-05 HEAT EXCHANGER B 2BE018 RUPTURES/FAILS TALAE038XX2 5.00E-01 3.20E-05 FRACTION OF TIME IN SERVICE FOR HX 2AE038 TALBE038XX2 5.00E-01 3.20E-05 FRACTION OF TIME IN SERVICE FOR HX 2BE038 THXAE038HWIE2 8.72E-03 3.20E-05 HEAT EXCHANGER A 2AE038 RUPTURES/FAILS THXBE038HWIE2 8.72E-03 3.20E-05 HEAT EXCHANGER B 2BE038 RUPTURES/FAILS A- TEST/MAl NT. FACTOR TO CORRECT FOR B- RUN TALAP144TM2 2.00E+00 3.17E-05 FACTOR B- TEST/MAl NT. FACTOR TO CORRECT FOR A- RUN TALBP144TM2 2.00E+00 3.17E-05 FACTOR TPMAP144TM2 1.01 E-03 3.17E-05 TBCCW PUMP A UNAVAILABLE DUE TO TEST/MAINT TPMBP144TM2 1.01 E-03 3.17E-05 TBCCW PUMP B UNAVAILABLE DUE TO TEST/MAINT MPMAP010DSI2 2.00E-03 3.14E-05 RBCCW PUMP 2AP010 FAILS TO START MPMBP010DSI2 2.00E-03 3.14E-05 RBCCW PUMP 2BP010 FAILS TO START TPMAP144DSI2 2.00E-03 3.14E-05 TBCCW PUMP 2AP144 FAILS TO START TPMBP144DSI2 2.00E-03 3.14E-05 TBCCW PUMP 2BP144 FAILS TO START WXV1700AHQIE2 2.63E-04 3.14E-05 MANUAL VALVE 30-21700A (N.O.) FAILS CLOSED WXV1700BHQI E2 2.63E-04 3.14E-05 MANUAL VALVE 30-21700B (N.O.) FAILS CLOSED WXV1700CHQIE2 2.63E-04 3.14E-05 MANUAL VALVE 30-21700C (N.O.) FAILS CLOSED WCV1762ADPI2 2.00E-04 2.39E-05 CHECK VALVE 2-30-21762A FAILS TO OPEN WCV1762BDPI2 2.00E-04 2.39E-05 CHECK VALVE 2-30-21762B FAILS TO OPEN WCV1762CDPI2 2.00E-04 2.39E-05 CHECK VALVE 2-30-21762C FAILS TO OPEN IAV0250DDPI2 2.00E-03 2.23E-05 AOV 80250D FAILS NC-FC FPNAP003TM2 1.24E-02 2.16E-05 CONDENSATE PUMP A OUT FOR MAINTENANCE EDGOCG12HRI0 1.32E-02 2.07E-05 DIESEL GENERATOR E3 - OCG12 FAIL TO RUN EARLY TPV80029DWIE2 2.00E-03 1.70E-05 PRESSURE CONTROL VALVE 34-80029 EDGODG12HRI0 1.32E-02 1.69E-05 DIESEL GENERATOR E4 - ODG12 FAIL TO RUN EARLY DTRLPCIBTM2 5.52E-03 1.52E-05 LPCI LOOP B IN MAINTENANCE 8-10 C467080016*841 0*1/8/2010

Peach Bottom SLC CT Extension Table B-4 BASIC EVENT IMPORTANCE MEASURES FOR CDF ASSESSMENT FOR SLC OUT OF SERVICE I Event Name I Probability I Fus Ves I Description I COMMON CAUSE FAILURE OF TBCCW PUMPS FAIL TO TPMXP144CRI2 2.25E-05 1.47E-05 RUN EXHOOX03HWI0 2.88E-05 1.33E-05 EMER AUX TRANSFORMER {00X03)LOSS OF FUNCTION WCV1762XCPI2 8.55E-06 1.18E-05 CCF OF PUMP DISCHARGE CHECK VALVES TO OPEN PCV-2428 FAILS TO ISOLATE 2CK001 FROMSERVICE AIR IPV02428DNI2 2.00E-03 1.10E-05 LOADS FPN--ABCCRI2 1.61 E-05 1.05E-05 PUMPS 2AP003 2BP003 AND 2CP003 [3 OF 3]

WPN--ABCCRI2 1.61 E-05 1.05E-05 CCF OF ALL 3 NSW PUMPS TO RUN IFLBS013HEI2 2.40E-04 1.03E-05 FILTER/DRYER PACKAGE 2BS013 FAILS ICMDK001 HRI2 2.40E-03 6.38E-06 COMPRESSOR 2DK001 FAILS TO RUN ZPM-TBRBCRI2 9.12E-06 5.95E-06 CCF OF RBCCW AND TBCCW PUMPS TO RUN ICV45993DNI2 1.00E-03 5.65E-06 CHK VLV 36-45993 (WHEN OPEN) FAILS TO CLOSE TPV80029DWI2 2.00E-03 5.32E-06 PRESSURE CONTROL VALVE 34-80029 EBSCHN2BTM2 1.00E-05 4.61E-06 125VDC CHANNEL 2B IN TESTING OR MAINTENANCE NAM-819BHWI2 7.20E-05 4.58E-06 LOGIC MATRIX C819 CHANNEL B FAILS EDGOCG12DSI0 4.44E-03 4.40E-06 DIESEL GENERATOR E3 - OCG12 FAIL TO START EDGODG12DSI0 4.44E-03 4.40E-06 DIESEL GENERATOR E4 -*,ODG12 FAIL TO START COMMON CAUSE FAILURE OF RBCCW PUMPS FAIL TO MPMXP010CSI2 1.22E-04 3.83E-06 START EBS4K-23CWI2 4.08E-08 3.82E-06 CCF OF 4 KV AC BUSES E22 AND E32 EDGOBG12HRI0 1.32E-02 3.81 E-06 DIESEL GENERATOR E2 - OBG12 FAIL TO RUN EARLY EDGOBG12TMO 1.58E-02 3.81 E-06 DIESEL GENERATOR E2 - OBG12 TESTING OR MAINT.

ICM-ABCDCRI2 5.57E-06 3.63E-06 IA COMPRESSORS A, B, C, AND 0 CCF [4 OF 4]

ITKDT006HFIE2 8.76E-04 2.62E-06 BACKUP INST. AIR RECEIVER 2DT006 FAILS APV-6529DW10 2.00E-03 2.46E-06 PRESS. CONT. VALVE PCV-6529 FAILS APV-7700DW10 2.00E-03 2.46E-06 PRESS. CONT. VALVE PCV-7700 FAILS MXV4200AHQIE2 2.63E-04 1.93E-06 MANUAL VALVE (NO) 35-24200A FAILS CLOSED MXV4200BHQIE2 2.63E-04 1.93E-06 MANUAL VALVE (NO) 35-24200B FAILS CLOSED MXV4203AHQIE2 2.63E-04 1.93E-06 MANUAL VALVE (NO) 35-24203A FAILS CLOSED MXV4203BHQIE2 2.63E-04 1.93E-06 MANUAL VALVE (NO) 35-24203B FAILS CLOSED MPMAP010HRI2 1.20E-04 1.89E-06 RECW PUMP 2AP10 FAILS TO CONTINUE TO RUN MPMBP010HRI2 1.20E-04 1.89E-06 RBCCW PUMP 2BP10 FAILS TO CONTINUE TO RUN TPMAP144HRI2 1.20E-04 1.89E-06 TBCCW PUMP 2AP144 FAILS TO CONTINUE TO RUN TPMBP144HRI2 1.20E-04 1.89E-06 TBCCW PUMP 2BP144 FAILS TO CONTINUE TO RUN EHB-0311 HOIO 7.20E-06 1.76E-06 CIRCUIT BREAKER 2SU-B (252-0311) NC-FO EHB-0313H010 7.20E-06 1.76E-06 CIRCUIT BREAKER 2SUB (252-0313) NC- FO EHB-SU25HOI0 7.20E-06 1.76E-06 CIRCUIT BREAKER SU25 (452-02) NC-FO 8-11 C467080016-841 0-1/8/2010

Peach Bottom SLC CT Extension Table 8-5 RISK ASSESSMENT SENSITIVITY RESULTS Acceptance Risk Metric Value Guidelines Reference

~CDF B.BE-OB/yr <1.0E-06/yr RG 1.174 ICCDP B.BE-OB <5.0E-07 RG 1.177

~LERF 1.7E-09/yr <1.0E-07/yr RG 1.174 ICLERP 1.7E-09 <5.0E-OB RG 1.177 8-12 C467080016-841 0-1/8/201 0

Peach Bottom SLC CT Extension 8.3 Parametric Uncertainty Consistent with the ASME PAA Standard, quantitative parametric uncertainty analyses for both CDF and LEAF have been performed and are summarized in this section. The results of the uncertainty analysis for the proposed CT are compared with the results of the uncertainty analysis performed for the 2005C PAA Update.

The parametric uncertainty analyses are performed using Monte Carlo simulation. The analysis is performed using the EPAI A&A workstation UNCEAT software.

8.3.1 Core Damage Frequency Parametric Uncertainty Distribution The resulting uncertainty distribution for the proposed CT configuration (Le., CDFsLc-oos) calculated by UNCEAT Version 2.3a for CDF is shown in Figure 8-1. The figure summarizes the following:

• Distribution statistics (e.g., mean, error factor, etc.)

• Probability density chart of the CDF The approximate error factor (or range factor) for the proposed CT is 2.5. This is equal to the error factor of the P8205C Model.

One of the critical aspects of the parametric uncertainty assessments is the desire to ensure that the point estimate calculation performed with the base PAA model (Le.,

using CAFTA) produces a point estimate result that is not too dissimilar from the true mean calculation when the correlation effect is accounted for.

Table 8-6 provides this comparison for the proposed CT model (CDFsLc-oos):

Table 8-6 PARAMETER UNCERTAINTY COMPARISON FOR CDF CDF Parameter CDF Result Code Point Estimate 6.6E-6/yr CAFTA Uncertainty Mean 6.7E-6/yr UNCEAT

~ 5.0E-8 -

This difference represents a very small perturbation on the point estimate CDF.

Therefore, it is concluded that the point estimate CDF calculated by CAFTA can be used to represent the mean CT CDF.

8-13 C467060024-7468-1/8/2010

Peach Bottom SLC CT Extension B.3.2 Large Early Release Frequency Parametric Uncertainty Distribution The same process as used for CDF is also used for LERF. The resulting uncertainty distribution calculated by UNCERT Version 2.3a for LERF is shown in Figure B-2. The figure summarizes the following:

• Distribution statistics (e.g., mean, error factor, etc.)

• Probability density chart of the LERF The approximate error factor (or range factor) for the LERF uncertainty distribution is 3.8 (calculated using SQR(95 % /5 % )) , as compared to the error factor of 4.9 for the PB205C model.

Table B-7 provides this comparison for the proposed CT model (Le., LERFsLc-oos):

Table B-7 PARAMETER UNCERTAINTY COMPARISON FOR LERF LERF Parameter CDF Result Code Point Estimate 2.3E-7/yr CAFTA Uncertainty Mean 2.3E-7/yr UNCERT Ll E -

8-14 C467060024-7468-1 /8/2010

Peach Bottom SLC CT Extension Figure B-1 CDF PARAMETRIC UNCERTAINTY DISTRIBUTION FOR THE PROPOSED COMPLETION TIME UNCERT 2.3a COREDAMAGE.CUT PB205C.BE Samples 50,000 Random Seed Auto 8-15 C467060024*7468-1/8/2010

Peach Bottom SLC CT Extension Figure B-2 LERF PARAMETRIC UNCERTAINTY DISTRIBUTION FOR THE PROPOSED COMPLETION TIME UNCERT 2.3a LERF-TOT.CUT PB205C.BE Samples 50,000 Random Seed Auto 8-16 C467060024-7468-1/8/2010

Peach Bottom SLC CT Extension AppendixC BWROG Assessment of NRC Information Notice 2007-07 The BWROG assessment of NRC Information Notice 2007-07 is provided in this appendix. This assessment discusses the low-likelihood scenario of fire-induced failure to scram. Refer to Section A.4.3 of this risk assessment.

C-1 C467090020-8958-1 /8/2010

Peach Bottom SLC CT Extension BWROG Assessment of NRC Information Notice 2007-07 1.0) Summary:

This assessment addresses the condition described by the NRC in NRC InfOlmation Notice 2007-07 and in the inspection report referenced therein.

The overall assessment of the condition described in NRC Information Notice 2007-07 by the BWROG is that it represents a condition with a low likelihood of occurrence, with low safety significance and with multiple layers of defense-in-depth currently in place each with the capability to either prevent the condition from occurring or to effectively mitigate the effects ofthe occurrence without consequence.

It is the position ofthe BWROG that all BWRs should have a manual operator action tied to their post-fire safe shutdown procedures instructing the operator to implement the requirements of EO-I 13 should the fire impact the ability to scram. This manual operator action should be endorsed by the NRC for use in both III.G.I and 2 areas, as well as, III.G.3 and IILL areas. The evaluation provided in this paper and the limited likelihood of occurrence ofthe condition are considered to be sufficient justification for concluding that this manual operator action is both feasible and reliable.

It is recommended that each BWR review this assessment and assure that their plant specific conditions are consistent with the measures described herein. As a minimum, each licensee should assure that the EOP action to implement the requirements of EO-113 is linked to their post-fire safe shutdown procedures.

2.0) Description of Issue:

NRC Information Notice 2007-07 postulates a condition where two (2) hot shorts could result in the failure of one of four control rods groups to insert during a manual scram from the Control Room. The IN further postulates that with the reactor in this condition the operator rapidly depressurizes the reactor and re-floods the reactor with cold water using a low pressure system. The IN further states:

'"By design, the negative reactivity, added by all four rod groups during a scram, provides adequate shutdown margin to offset the positive void and temperature reactivity [that] would have been added to the vessel [during such a shutdown sequence]".

3.0) Scram System Design

Description:

Typically, the Reactor Protection System (RPS) for a BWR consists of two (2) Trip Systems (A and B), each containing two Trip Channels (AI, A2, BI, B2) of sensors and logic. The four channels contain automatic scram logic for the monitored parameters listed below, each of which has at least one input to each ofthe logic channels:

• Scram Discharge Volume Water Level C-2 C467090020-8958-1/8/2010

Peach Bottom SLC CT Extension BWROG Assessment of NRC Information Notice 2007-07

• Main Steam Line Isolation Valve Position

• Turbine Stop Valve Position

• Turbine Control Valve Fast Closure

• Reactor Vessel Water Level

• Main Steam line Radiation

• Neutron Monitoring System

• Primary Containment Pressure

• Reactor Vessel Pressure The RPS automatic trip logic requires at least one channel in each trip system to be tripped in order to cause a scram. This is referred to as one-out-of-two-taken-twice trip logic.

The two RPS Trip Systems are independently powered from their respective RPS Buses.

The trip channels (AI, A2, Bl, B2) associated with each Trip System (A, B) operate the automatic scram Trip Logic Relays (K14 A-H). The RPS auto scram logic string is sometimes referred to as "trip actuator" or "actuation" logic because the output ofthe logic is what actually causes the control rods to scram by de-energizing the pilot scram solenoid valves.

The RPS circuits are a fail-safe design in that the circuits are normally energized, and the loss of power, including the loss of offsite power, will initiate the scram.

Once the scram has occurred, re-energization ofthe RPS logic will not, in and of itself, cause the control rod movement necessary to re-establish reactor criticality.

4.0) Evaluation:

The evaluation performed is divided into two sections. The first section performs a circuit analysis ofthe scram circuitry. This portion ofthe evaluation examines the scram circuitry in an effort to determine the set of hot shorts that, should they occur, have the potential to prevent one or more rod groups from inserting. The first section also addresses the significance ofthe postulated condition and the features currently in place with the capability to prevent or mitigate the effects of the condition. The second section addresses the implications for Appendix R Compliance given the required circuit design for this important safety system and given the potential ramifications ofthe hot shorts postulated in the first section.

4.1) Circuit Analysis:

Figures 1through 4 attached to this paper shows portions ofthe scram circuitry for a typical BWR. Three (3) separate cases involving up to two hot shorts are discussed in this paper.

C-3 C467090020-8958-1/8/2010

Peach Bottom SLC CT Extension BWROG Assessment of NRC Information Notice 2007-07 Case I: (Refer to Figure 1)

Case I attempted to identify the condition described in IN 2007-07. IN 2007-07 concluded that two (2) hot shorts were required to prevent a single rod group from scramming.

The BWROG, however, was unable to identify any circuitry where two (2) flre-induced hot shorts would prevent one of four scram rod groups from inserting.

The BWROG identifled that a single hot short in either ofthe divisionalized trip logics can prevent the scram of a single rod group. This fmding is different than the conclusion in IN 2007-07. The flnding ofthe BWROG assessment is a direct consequence ofthe lout of2 taken twice logic used in the design for the scram function.

The single hot short with the potential for preventing the scramming of a single rod group could occur in either the Trip System A or B Relay Panel. [Refer to Figure 1attached for a description ofthe location ofthe subject hot short, labeled as "Hot Short 1".] The hot short must occur prior to the operator scramming the reactor. The location ofthe hot short shown in Figure 1would be either in one of the Trip System Relay Panels or in a raceway carrying the circuit from the Trip System Relay Panel to the Scram Pilot Solenoid Valves. (Note: For some licensees, the relay panels are located in separate relay rooms outside ofthe main control room.)

For the hot short in this case to affect the reactivity function, it must remain in effect until such time when the operator depressurizes the reactor and begins re-flooding with a low pressure system. The Emergency Operating Procedures for a BWR instruct the operator not to depressurize the reactor until reactor level reaches the top of active fuel. In a typical BWR, it will take approximately 20 to 25 minutes of boil-off for reactor level to decrease to the top of active fuel.

Industry and NRC cable flre testing have shown that hot shorts last for only a few minutes prior to shorting to ground. [EPRI Testing determined the maximum duration of a hot short was 11.3 minutes. CAROLFIRE Testing determined that the maximum duration of a hot short was 7.6 minutes.]

Therefore, it appears unlikely that the required hot short could last for a sufficient amount of time that the impacted control rod group would fail to insert prior to the time when the EOPs directed the operator to depressurize the reactor.

Case II: (Refer to Figure 2)

Case II is one oftwo cases identified where two (2) flre-induced hot shorts could prevent a full scram. (Note: No conditions were identified where two (2) flre-induced hot shorts were required to prevent a single rod group from scramming.)

C-4 C467090020*8958*1 /8/2010

Peach Bottom SLC CT Extension BWROG Assessment of NRC Information Notice 2007-07 Refer to Figure 2 attached for the case where two (2) fIre-induced hot shorts could prevent a full scram.

This case postulates a condition where two hot shorts just below the manual scam switches for two trip channels can prevent a full scram. The postulated hot shorts could occur in either the main control room operating bench board or in a raceway carrying the trip circuit to one of the Trip System Relay Panels. The hot short will keep the K15 relays from de-energizing and this will subsequently keep the K14 relays energized. By keeping the K14 relays energized, as shown in Figure 1, none of the rod groups will de-energize and none will insert. Figure 2 shows the location ofthe two individual hot shorts. One affects the K15B relay and one affects the K15D relay. The K15 relays are de-energized by actuating the manual scram switches in the Control Room on the main control board. Keeping the K15 relays energized by the hot shots shown in Figure 2, will keep the K14 relays energized, as shown in Figures 3. Keeping the K14 relays energized, as shown in Figure 3, will prevent rod group insertion, as shown in Figure 1.

For this case, however, there are numerous other inputs into the scram logic that can override the effects of the hot short affecting the K15 relays. Refer to Figures 3 and 4 for the additional input signals to the scram function. For example, as shown on Figure 4, closure ofthe MSIVs or reactor level reaching the +13" level will override the effects ofthe hot shorts affecting the K15 relays and result in a de-energization ofthe K14 relays and full rod insertion.

Therefore, it appears unlikely that the required hot shorts, even ifthey were to co-exist, could prevent the scram and cause the reactivity transient described in the IN. This is true because the effect ofthe hot short would be overriddened by the reduction in reactor level that would be necessary before the operator would take the action to depressurize the reactor prior to making up with a low pressure system.

Case III: (Refer to Figure 3) (Limited to the Trip System Relay Panels)

Case III is similar to Case II. Hot shorts are postulated in the locations shown in Figure 3, the K14 relays will again remain energized. The energization ofthe K14 relays will prevent the scram for all rod groups.

For this case to occur, the fIre must sufficiently damage two separate circuits and the fIre induced damage must occur on each circuit simultaneously. Industry and NRC cable fIre testing have shown that hot shorts last for only a few minutes prior to shorting to ground [EPRI Testing determined the maximum duration of a hot short was 11.3 minutes. CAROLFIRE Testing determined that the maximum duration of a hot short was 7.6 minutes.]

C-5 C467090020-8958-1 /8/2010

Peach Bottom SLC CT Extension BWROG Assessment of NRC Information Notice 2007-07 for each ofthese postulated fire areas would be ineffective in preventing the occurrence ofthe condition. The condition postulated in Case I can only be mitigated by the use of a manual operator action consistent with the manual operator actions currently invoked under Emergency Operating Procedure, EO-113.

The conditions described for Cases II and III are similar. Neither ofthese cases represents a condition that is prevented by the type of redundant train separation invoked under Appendix R, since the postulated hot shorts occur within a single division.

Therefore, the provision of Appendix R cannot be used to address the conditions described in this paper. Re-design of the scram circuitry is not a viable option without compromising the design function ofthis important safety function. In addition to the features ofthe RPS system described above, the Alternate Rod Insertion (ARI) system (vents SCRAM air header), Backup Scram Solenoids (vents SCRAM air header), and Standby Liquid Control (SLC) system (inserts sodium pentaborate) provide additional redundant means to achieve reactor shutdown. For areas such as the main Control Room and the Relay Rooms, however, similar fire-induced impacts could be postulated.

This paper has highlighted one example of an area where verbatim compliance with the requirements of Appendix R is insufficient in preventing fire induced damage from potentially impacting safe shutdown. The BWROG believes that this case and, potentially, other like it are the reason why from the initial issuance of Appendix R that certain conditions were considered to be initial boundary conditions for the Appendix R Post-Fire Safe Shutdown Analysis. Assuming that the reactor is scrammed was one of those initial boundary conditions given for the Post-Fire Safe Shutdown Analysis. NRC Generic letter 86-10 in the Response to Question 3.8.4, Control Room Fire Considerations, endorsed the assumption of a reactor trip prior to evacuating the Control Room. Based on this and on the fail-safe nature ofthe reactor protection system, many licensees assumed and the NRC accepted that a reactor trip was an initial boundary condition for the start of the post-fire safe shutdown analysis, i.e. the plant is scrammed prior to the scram circuitry being damaged by the fire.

Although the BWROG believes that the prior industry position related to the scram is correct and its use provides for a safe plant design, the BWROG also recognizes that fires have some limited potential to impact the scram capability. As a precaution, it is the position ofthe BWROG that all BWRs should have a manual operator action tied to their post-fire safe shutdown procedures instructing the operator to implement the requirements of EO-l13 should the fire impact the ability to scram. This manual operator action should be endorsed by the NRC for use in both III.G.l and III.G.2 areas, as well as, III.GJ and III.L areas. The evaluation provided in this paper and the limited likelihood of occurrence of the condition are considered to be sufficient justification for the feasibility and reliability ofthis manual operator action.

e-G C467090020-8958-1 /8/2010

Peach Bottom SLC CT Extension BWROG Assessment of NRC Information Notice 2007-07 5.0) Risk Assessment:

Given the unlikely set of circumstances required for this condition to occur and to remain in effect until such time that it could pose a beyond design basis concern to the reactor, the risk associated with this issue is judged to be low.

6.0) Safety Assessment:

Given the fact that there are multiple barriers (circuit failure characteristics, design features, procedural guidance and rigorous operator training) in place to prevent the occurrence ofthis condition, the safety significance ofthis issue is also judged to be very low.

7.0) Conclusions and Recommendations:

This assessment addresses the condition described by the NRC in NRC Information Notice 2007-07 and in the inspection report referenced therein.

The overall assessment ofthe condition described in NRC Information Notice 2007-07 by the BWROG is that it represents a condition with a low likelihood of occurrence, with low safety significance and with multiple layers of defense-in-depth currently in place each with the capability to either prevent the condition from occurring or to effectively mitigate the effects ofthe occurrence without consequence.

It is the position ofthe BWROG that all BWRs should have a manual operator action tied to their post-fire safe shutdown procedures instructing the operator to implement the requirements of EO-I13 should the fire impact the ability to scram. This manual operator action should be endorsed by the NRC for use in both III.G.l and 2 areas, as well as, III.G.3 and IILL areas. The evaluation provided in this paper and the limited likelihood of occurrence ofthe condition are considered to be sufficient justification for concluding that this manual operator action is both feasible and reliable.

It is recommended that each BWR review this assessment and assure that their plant specific conditions are consistent with the measures described herein. As a minimum, each licensee should assure that the EOP action to implement the requirements of EO-113 is linked to their post-fire safe shutdown procedures.

Prepared by: Thomas A. Gorman Date: 10/16/2007 Thomas A. Gonnan, PE, SFPE Reviewed by: Gary Birmingham Date: 11/13/2007 Gary S. Binningham C-7 C467090020-8958-1 /8/2010

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C-8 C467090020-8958-1 /8/2010

Peach Bottom SLC CT Extension BWROG Assessment of NRC Information Notice 2007-07

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• Manual Scram Circuitry* Typical of two Trip Systems C-9 C467090020-8958-1 /8/2010

Peach Bottom SLC CT Extension BWROG Assessment of NRC Information Notice 2007-07 Refer to Figure 4 for the remaining set of contacts that affect the automatic scram function Hot Short #3 location ttvpical 2 per Trip Systems)

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Figure 3 - Reactor Auto-Scram Circuitry - Typical of four Trip Channels in two Trip Systems C-10 C467090020-8958-1/8/2010

Peach Bottom SLC CT Extension BWROG Assessment of NRC Information Notice 2007~07 F4lG CIIlEFJ 1

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C-11 C467090020-8958-1 /8/2010