Text

Request for Amendment to Technical Specification 3.1.7, Standby Liquid Control (SLC) System.
	ML100630136
Person / Time
Site:	Dresden, Clinton, Quad Cities
Issue date:	03/03/2010
From:	Hansen J; Exelon Generation Co, Exelon Nuclear
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	RS-10-028, TAC MD4044
	Download: ML100630136 (102)
	v • d • e

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!~d;i.r!we,elr,rco!pi ~ i 7 1 Nuclear 10 CFR 50.90 RS-10-028 March 3,2010 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001 Clinton Power Station, Unit 1 Facility Operating License Nos. NPF-62 NRC Docket No. 50-461

Subject:

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

References:

1) 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 Quad Cities Nuclear Power Station, Unit 1 (NOED 06-3-01)," dated October 18,2006

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 In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Exelon Generation Company, LLC (EGC) requests an amendment to Appendix A, Technical Specifications (TS) of Facility Operating License No. NPF-62 for Clinton Power Station (CPS), Unit 1.

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

March 3,2010 U. S. Nuclear Regulatory Commission Page 2 In References 1 and 2, the NRC exercised discretion to not enforce compliance with the actions required in TS 3.1.7, Condition C for Quad Cities Nuclear Power Station, Unit 1 and Dresden Nuclear Power Station, Unit 2, respectively. These notices of enforcement discretion (NOEDs) provided a 72-hour extension to the 12-hour CT specified in Required Action C.l (i.e., "Be in MODE 3"). This extension enabled each site to avoid a TS-required shutdown while implementing short-term repair and restoration activities for an emergent issue impacting SLC system operability. The purpose of this proposed license amendment request (LAR) is to adopt a permanent, risk-informed CT extension for CPS TS 3.1.7, Required Action B.l, thus minimizing the potential for thermal transients associated with placing CPS Unit 1 in Mode 3.

EGC has utilized the guidance in Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment In Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," to develop the technical basis for this M R . The EGC analysis demonstrates, with reasonable assurance, that the proposed LAR satisfies the risk acceptance guidelines in Regulatory Guide 1.I74 and Regulatory Guide 1.177, "An Approach for Plant-Specific, Risk-lnformed Decision-making: Technical Specifications." The proposed LAR meets the intent of very small risk increases consistent with the NRC's Safety Goal Policy Statement.

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-lnformed Activities," Revision 1.

This request is subdivided as follows:

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

o Attachment 2 provides a mark-up of the CPS TS page with the proposed change indicated.

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

o Attachment 4 provides the risk assessment that supports the proposed TS change for CPS (i.e., RM Documentation CL-LAR-01, Revision 1).

March 3,2010 U. S. Nuclear Regulatory Commission Page 3 The proposed amendment has been reviewed and approved by the CPS Plant Operations Review Committee and the Nuclear Safety Review Board in accordance with the requirements of the EGC Quality Assurance Program and procedures. EGC requests approval of the proposed amendment by March 3, 201 1, with implementation within 60 days of issuance.

In accordance with 10 CFR 50.91, "Notice for public comment," EGC is notifying the State of Illinois 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. John L. Schrage at (630) 657-2821.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 3'* day of March 2010.

Manager - Licensing : Evaluation of Proposed Amendment : Proposed Markup of CPS Technical Specification 3.1.7 : Proposed Markup of CPS Technical Specification Bases B 3.1.7 : RM Documentation No. CL-LAR-01, Revision 1

ATTACHMENT 1 Evaluation of Proposed Amendment

1.0 DESCRIPTION

2.0 PROPOSED CHANGE

3.0 BACKGROUND

4.0 TECHNICAL ANALYSIS

5.0 REGULATORY ANALYSIS

5.1 No Significant Hazards Consideration 5.2 Applicable Regulatory Requirementslcriteria

6.0 ENVIRONMENTAL CONSIDERATION

7.0 REFERENCES

Page 1 of 19

ATTACHMENT 1 Evaluation of Proposed Amendment

1.0 DESCRIPTION

In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Exelon Generation Company, LLC (EGC) requests an amendment to Facility Operating License Nos. NPF-62 for Clinton Power Station (CPS) Unit 1. 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 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

CPS TS LC0 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 "10 CFR 50, Appendix A, "General Design Criteria for Nuclear Power Plants," Criterion (GDC) 26, "Reactivity control system redundancy and capability." 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 B and the associated Required Action B.l address the inoperability of both SLC subsystems. Specifically, Required Action B.1 requires restoration of one SLC subsystem to operable status, with a CT of eight hours. If Required Action B.l cannot be satisfied within the CT, Condition C and associated Required Actions C.l and C.2 require the reactor to be in Mode 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and Mode 4 in 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

The current CT for Required Action B.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 1.174, "An Approach for Using Probabilistic Risk Assessment In Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," 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 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 NRC1sSafety 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 I.I 77, 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.ZOO, "An Page 2 of 19

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

2.0 PROPOSED CHANGE

The proposed amendment revises the CT for CPS TS 3.1.7, Required Action B.l from eight hours to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

3.0 BACKGROUND

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, "Requirements for reduction of risk from anticipated transients without scram (ATWS) events for light-water-cooled nuclear power plants."

CPS TS LC0 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 B and the associated Required Action B.l address the inoperability of both SLC subsystems. Specifically, Required Action B.l requires restoration of one SLC subsystem to operable status, with a CT of eight hours. If Required Action B.l cannot be satisfied within the CT, Condition C and associated Required Actions C.l and C.2 require the reactor to be in Mode 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and Mode 4 in 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

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 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to the original 12-hour CT for TS 3.1.7, Required Action C.l (i.e., "Be in MODE 3") for the emergent dual-train inoperability of the SLC systems (References 1 and 2).

The purpose of this proposed LAR is to adopt a permanent, risk-informed CT extension for CPS TS 3.1.7, Required Action B.l, thus minimizing the potential for thermal transients associated with placing CPS, Unit 1 in Mode 3. 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.

4.0 TECHNICAL ANALYSIS

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.I 74, RG 1.177, and the NRC's Safety Goal Policy Statement. This consistency is described below, as well as in Attachment 4.

Page 3 of 19

ATTACHMENT 1 Evaluation of Proposed Amendment 4.1 System Description 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 (which is at the peak of the xenon transient) 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 on anticipated transient without scram (ATWS).

The SLC system is also used to maintain suppression pool pH at or above 7 following a loss of coolant accident (LOCA) involving significant fission product releases.

Maintaining suppression pool pH levels at or above 7 following an accident ensures that iodine will be retained in the suppression pool water.

The SLC system consists of a boron solution storage tank, two positive displacement pumps, two explosive valves, which 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 preferred flow path of the boron neutron absorber solution to the reactor vessel is by the High Pressure Core Spray (HPCS) System sparger. The SLC piping is connected to the HPCS System just downstream of the HPCS manual injection isolation valve. An alternate flow path to the reactor vessel is provided by the SLC sparger near the bottom of the core shroud. This flow path is normally locked out of service by the SLC manual injection valve.

The SLC system is manually initiated from the main control room, as directed by the emergency operating procedures, if the operator believes 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 injects borated water into the reactor core to compensate for all of the various reactivity effects that could occur during plant operation. To meet this objective, it is necessary to inject a quantity of boron that produces a concentration equivalent to at least 1000 ppm of natural boron in the reactor core at 68°F. This is accomplished by the use of enriched boron (i.e., greater than or equal to 30 atom% boron 10). To allow for potential leakage and imperfect mixing in the reactor system, an additional amount of boron equal to 25% of the amount cited above is added. The concentration versus volume limits are calculated such that the required concentration is achieved accounting for dilution in the RPV with normal water level and including the water volume in the residual heat removal shutdown cooling piping and in the recirculation loop piping.

The control rods are the primary reactivity control system for the reactor at CPS. In conjunction with the Reactor Protection System (RPS), the control rods provide the means for reliable control of reactivity changes to ensure that, under conditions of normal operation, including anticipated operational occurrences, specified acceptable 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 (SRs) 3.1.3.1 and 3.1.3.3 through Page 4 of 19

ATTACHMENT I Evaluation of Proposed Amendment 3.1.3.5. This Specification, along with TS 3.1.4, "Control Rod Scram Times," and TS 3.1.5, "Control Rod Scram Accumulators," ensure that the performance of the control rods in the event of a Design Basis Accident (DBA) or transient meets the assumptions used in the safety analyses.

Scram reliability is ensured by a number of design and operational features:

An individual accumulator is provided for each control rod drive with sufficient stored energy to scram at any reactor pressure. The reactor vessel itself, at pressures above 600 psi, will supply the necessary force to insert a drive if its accumulator is unavailable.

Each drive mechanism has its own scram valves and a dual solenoid scram pilot valve therefore only one drive can be affected if a scram valve fails to open. Both pilot valve solenoids must be deenergized to initiate a scram.

The reactor protection system and the HCUs are designed so that the scram signal and mode of operation override all others.

The collet assembly and index tube are designed so they will not restrain or prevent control rod insertion during scram The scram discharge volume is monitored for accumulated water and the reactor will scram before the volume is reduced to a point that could interfere with a scram.

The alternate rod insertion (ARI) system provides an alternate means of exhausting the scram air header and closing the vent and drain valves of the scram discharge volume, thereby providing an additional reactor scram mechanism which is diverse, redundant and independent of the reactor protection system.

In addition to the ARI system, the ATWS Recirculating Pump Trip (RPT) system provides an additional means for rapid power reduction. The ATWS-RPT system initiates a recirculation pump trip, adding negative reactivity, following events in which a scram does not, but should occur, to lessen the effects of an ATWS event.

As noted above, operability of the trip function of the control rods is demonstrated by specific SRs. 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.

Operability of the ATWS system (i.e., the ARI system and the ATWS RPT system) is governed by TS 3.3.4.2, "Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RPT) Instrumentation," and is demonstrated operable by the performance of TS SRs 3.3.4.2.1 through 3.3.4.2.5.

The proposed change to the SLC CT does not affect the redundancy, independence, and diversity of the RPS, ARI, and the ATWS-RPT systems. These systems and instrumentation remain operable to mitigate the consequences of any previously analyzed accident. In addition to the TS requirements for control rod and ATWS system operability, the EGC Work Management and Maintenance Rule (i.e., 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 CT extension does not involve any change to plant equipment or system design functions.

Page 5 of 19

ATTACHMENT I Evaluation of Proposed Amendment 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.I 74, paragraph 2.2.1 .Iand RG 1.I 77, 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 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 (i.e., RPS and ARIIRPT) and alternate (i.e., SLC) reactivity control systems at CPS is robust. The proposed SLC CT extension does not require, nor rely upon programmatic activities to compensate for weaknesses in plant design. The four-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 CT.

Entry into the dual-train SLC CT will be assessed and managed in accordance with the EGC Configuration Risk Management Program (CRMP).

Additional redundancy for reactivity control is established by CPS procedures.

These procedures describe the actions and criteria for manual addition of boron into the reactor coolant system (i.e., via the reactor water cleanup system), should RPS, the control rods, the control rod drive system, and the SLC be unable to perform the specifed design functions.

ATTACHMENT 1 Evaluation of Proposed Amendment Defenses against potential common cause failures are maintained and the potential for introduction of new common cause failure mechanisms is assessed.

The extended SLC 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 failures modes are not expected.

Therefore, 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, (i.e., fuel cladding, the reactor coolant system, or containment).

Defenses against human errors are maintained.

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

4.3 Safety Margin Assessment The proposed SLC 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. Since this proposed TS amendment does not change the SLC system design, but only extends a CT, safety margins are not challenged.

4.4 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.

The proposed CT extension for the dual-train inoperability of the CPS 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.

Page 7 of 19

ATTACHMENT 1 Evaluation of Proposed Amendment EGC completed a risk assessment for CPS using the full power internal events, Level 1 CDF model and the associated Level 2 LERF model. This risk assessment is provided in Attachment 4. The risk assessment was performed in accordance with the requirements in RG 1. I74, RG 1.177, and RG 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk Informed Activities," Revision I.The results of these risk assessments are discussed below.

4.4.1 Regulatory Standards The RG 1.I 74 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 "very small changes". EGC utilized the acceptance guidelines for "very small changes" in the risk assessment for the proposed CPS TS change.

The RG 1.I 74 acceptance guidelines prescribe that the risk metrics of ACDF and ALERF be less than I.OE-O6lyr and I.OE-07/yr, respectively, to establish a very small risk increase with no additional compensatory measures required. RG 1.I74 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.I 77 identifies a three-tiered approach for the evaluation of the risk associated with a proposed TS change.

Tier I,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.I77 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, 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.

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 ASMEIAmerican Nuclear Society (ANS) Standard RA-Sa-2009, "Addenda to RA-S-2008, Standard for Level IILarge 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 CT risk assessment.

4.4.2 Tier 1: PRA Capability and Insights As stated in RG I.I 77, 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 CPS TS change using the FPIE PRA models and based on the risk metrics of ACDF, ICCDP, ALERF, and ICLERP.

Page 9 of 19

ATTACHMENT 1 Evaluation of Proposed Amendment Table 4.4.2-1 CPS Risk Assessment Summary Results Hazard ACDF ICCDP ALERF ICLERP FPlE 2.9E-081yr 2.9E-08 6.2E-091yr 6.2E-09 Acceptance < I .OE-OGIyr c5.OE-07 < I .OE-071yr e5.OE-08 Guideline External Events (1) (1) (1) (1)

(1) In accordance with RG I.I 74, paragraph 2.2.5.5, "Comparisons with Acceptance Guidelines,"

EGC performed a qualitative assessment of external event risk associated with the proposed CPS SLC CT extension (i.e., 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 ACDF, ICCDP, ALERF, and ICLERP risk metric values for the proposed change are below the acceptance guidelines as defined in RG 1.I 74 and RG 1.I 77. This analysis demonstrates that the proposed TS change satisfies the risk acceptance guidelines in RG 1.I74 and RG 1.I 77, 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 "very small" acceptance criteria. For this sensitivity, ICCDP and ICLERP were set to their maximum allowable values in RG I.I 77, and the CTNEW allowable was calculated. ICLERP was determined to be the bounding parameter, and a CTNEW value of 582 hours0.00674 days 0.162 hours 9.623016e-4 weeks 2.21451e-4 months was calculated. This value represents significant margin, relative to the proposed CT extension.

The CPS risk assessment also includes a qualitative assessment of external event risks in accordance with RG 1.174, paragraph 2.2.5.5, "Comparisons with Acceptance Guidelines."

This qualitative external events assessment used the external event analyses in the 1995 CPS Individual Plant Examination of External Events (IPEEE).

The qualitative external events assessment is described in Appendix A of Attachment 4, and summarized below.

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

NUREGICR-6850 , "EPRI Report 1011989, 'Fire PRA Methodology for Nuclear Power Facilities'," September 2005 Page 10 of 19

ATTACHMENT 1 Evaluation of Proposed Amendment CPS-PSA-021.06, "Clinton FPRA Summary and Quantification Report," Rev. 0, September 2008 Boiling Water Reactor Owners' Group (BWROG), "Assessment of NRC Information Notice 2007-07," October 16, 2007 (i.e., Appendix C of Attachments 4 and 5)

The assessment concluded that a fire-induced ATWS is a non-significant contributor to the plant risk profile and thus does not impact the proposed SLC system CT.

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

CPS Seismic Margins Assessment that was performed as part of the CPS IPEEE, and was consistent with the guidance in EPRl NP-6041, "A methodology for assessment of nuclear power plant seismic margin" NUREG-1150, "Severe Accident Risks: An Assessment for Five U.S.

Nuclear Power Plants," December 1990 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 CPS IPEEE analysis. No significant quantitative contribution from these external events was identified by the CPS 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 performed. The results of these analyses are summarized in Appendix B of Attachment 4.

An assessment of modeling uncertainties is also documented in Appendix B of Attachment 4. This assessment includes CPS-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 CT extension. The results of the modeling uncertainty assessments do not change the conclusions of this risk assessment for the proposed SLC CT changes.

4.4.3 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 Page 11 of 19

ATTACHMENT 1 Evaluation of Proposed Amendment outage configurations. Consistent with the guidance in Regulatory Position C.2.3 of RG 1.177, and as part of the CPS risk assessment (i.e., 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 (i.e., 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 CT extension).

This evaluation is provided in Attachment 4, Appendix B, "Uncertainty Analysis,"

section B.2, "Model Uncertainties Associated with SLC System Out of Service."

This evaluation indicates that the scram system hardware failure is the most important contributor to the ACDF assessment for the SLC system out-of-service case.

Entry into the dual-train SLC CT will be assessed and managed in accordance with the EGC CRMP. The CRMP will assess the emergent condition, including the impact of any additional out-of-service equipment. With both SLC subsystems unavailable, the CPS on-line risk would be depicted as "Orange,"

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.

4.4.4 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.I 77, "a viable program would be one that is able to uncover risk-significant plant equipment outage configurations in a timely manner during normal plant operation." 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 CPS. 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 like testing or load dispatching, and weather conditions. CPS 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 failurelmalfunction or emergent condition produces a plant configuration that has not been previously assessed.

Page 12 of 19

ATTACHMENT I 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 are reviewed, and work on SSCs that are important in mitigating the transient are avoided.

Work is not scheduled that is highly likely to exceed a TS or Operational Requirements Manual (ORM) Completion Time requiring a plant shutdown.

For Maintenance Rule high risk significant SSCs, the impact of the planned activity on the unavailability performance criteria is evaluated.

A quantitative risk assessment is performed for those SSCs modeled in the CPS PRA model 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 defines the requirements for ensuring that the PRA model used to evaluate on-line maintenance activities is an accurate model of the current plant design and operational characteristics.

Plant modifications and procedure changes are monitored, assessed, and dispositioned. Evaluation of changes in plant configuration or PRA model features are dispositioned by implementing PRA model changes or by the qualitative assessment of the impact of the change on the 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, "Requirements for monitoring the effectiveness of maintenance at nuclear power plants," paragraph (a)(l) actions, requiring increased management attention and goal setting in order to restore performance (i.e., reliability and availability) to an acceptable level. The performance criteria are risk-informed, and therefore are a means to manage the overall risk profile of the plant. An accumulation of large core damage probabilities over time is precluded by the performance criteria.

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 13 of 19

ATTACHMENT 1 Evaluation of Proposed Amendment 4.4.5 Technical Adequacy and Quality 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 SLC CT extension, EGC has documented this determination of PRA quality in Attachment 4. Table 2-1 of Attachment 4 provides a "RG 1.ZOO Analysis Actions Roadmap." This roadmap cross references the required RG 1.200 actions to the applicable sections in the attachment that address the actions, which are summarized below.

EGC employs a multi-faceted approach to establishing and maintaining the technical adequacy and plant fidelity of the PRA models for all operating EGC 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 EGC 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 EGC Risk Management program, which consists of a governing procedure (i.e., ER-AA-600, "Risk Management") and subordinate Technical & Reference Material (T&RM) documents. EGC 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 EGC nuclear generation sites.

The overall EGC Risk Management program, including ER-AA-600-1015, defines the process for implementing regularly scheduled and interim PRA model updates, for tracking issues identified as potentially affecting the PRA models (e.g., changes in the plant, errors or limitations identified in the model, industry operating experience), and for controlling the model and associated computer files.

Page 14 of 19

ATTACHMENT 1 Evaluation of Proposed Amendment

5.0 REGULATORY ANALYSIS

5.1 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 consequence of an accident previously evaluated; (2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) lnvolve a significant reduction in a margin of safety.

Exelon Generation Company, LLC (EGC) has evaluated the proposed changes to the Technical Specifications (TS) for Clinton Power Station (CPS), Unit 1 using the criteria in 10 CFR 50.92 and has determined that the proposed changes 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, "Standby Liquid Control (SLC) System," to extend the completion time (CT) for Condition B (i.e., "Two SLC subsystems inoperable.") from eight hours to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

The proposed change is based on a risk-informed evaluation performed in accordance with Regulatory Guides (RG) 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions On Plant-Specific Changes to the Licensing Basis," and RG 1.I 77, "An Approach for Plant-Specific, Risk-Informed Decision-making: Technical Specifications."

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 or a large-break loss-of-coolant accident for which the SLC system will be used for pH control. The extended CT provides additional time to implement actions in response to a dual-train SLC system inoperability, while also minimizing the risk associated with continued operation. Therefore, Page 15 of 19

ATTACHMENT 1 Evaluation of Proposed Amendment 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 B from eight hours to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . 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.

(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 B from eight hours to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . 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.

The proposed amendment does not modify the condition or point 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.I77 requirements (i.e., Section C, "Regulatory Position,"

paragraph 2.2, "Traditional Engineering Considerations").

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 "no significant hazards consideration" is justified.

Page 16 of 19

ATTACHMENT 1 Evaluation of Proposed Amendment Applicable Regulatory RequirementslCriteria 10 CFR 50.62, "Requirements for reduction of risk from anticipated transients without scram (ATWS) events for light-water-cooled nuclear power plants" 10 CFR 50.62 (c)(4) states that boiling water reactors are required to have a standby liquid control (SLC) 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-10 isotope abundance into a 251-inch inside diameter reactor pressure vessel for a given core design. Furthermore, the SLC 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 CPS SLC system to ensure compliance with these requirements.

10 CFR 50.67, "Accident source term" 10 CFR 50.67.b(1) provided 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 April 3, 2003, AmerGen Energy Company, LLC (i.e., the CPS licensee at that time) requested an amendment to the CPS TS 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 11, 2006. As part of the proposed AST methodology, EGC will use the SLC system to inject sodium pentaborate into the RPV following a LOCA in order to maintain suppression pool pH above 7 (i.e., in order to ensure against re-evolution of elemental iodine).

As such, the SLC will be required to be operable in Mode 3 to 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. However, additional redundancy for the addition of boron into the reactor coolant system is established by CPS procedures. The procedures describe the actions and criteria for manual addition of boron into the Reactor Core Isolation Cooling (RCIC) system tank, and the use of the RCIC system to inject the boron into the RPV, should RPS, the control rods, the control rod drive system, and the SLC be unable to perform the specifed design functions. Therefore, the proposed SLC CT extension will not impact the ability of CPS to comply with the requirements of 10 CFR 50.67.

Page 17 of 19

ATTACHMENT 1 Evaluation of Proposed Amendment 10 CFR 50, Appendix A, "General Design Criteria for Nuclear Power Plants,"

Criterion (GDC) 26, "Reactivity control system redundancy and capability" GDC 26 requires the provision of two independent reactivity control systems of different design principles. While one of the systems shall use control rods, the second reactivity control system shall be capable of reliably controlling the rate of reactivity changes resulting from planned, normal power changes (including xenon burnout) to assure acceptable fuel design limits are not exceeded. The proposed change will not impact the ability of the CPS SLC system to ensure compliance with this requirement.

Regulatory Guide (RG) 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis" RG I.177, "An Approach for Plant-Specific, Risk-Informed Decision-making:

Technical Specifications" RG I.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision I 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 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 "very small changes."

The RG 1.I 74 acceptance guidelines prescribe that the risk metrics of delta (A) CDF and ALERF be less than 1.OE-O6lyr and 1.OE-07/yr, respectively, to establish a very small risk increase with no additional compensatory measures required. RG 1.174, paragraph 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.I 77, "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 I. 174, RG 1.I 77, and RG 1.200 to demonstrate a very small change in risk.

Page 18 of 19

ATTACHMENT 1 Evaluation of Proposed Amendment Regulatory Summary Based on the considerations discussed above, ( I ) 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.

6.0 ENVIRONMENTAL CONSIDERATION

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

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

7.0 REFERENCES

1. 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 Quad Cities Nuclear Power Station, Unit 1 (NOED 06-3-01)," dated October 18, 2006

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 Page 19 of 19

ATTACHMENT 2 Proposed Markup of CPS Technical Specification 3.1.7 TS Page 3.1-20

SLC System 3.1.7 3.1 REACTIVITY CONTROL SYSTEMS 3.1.7 Standby Liquid Control (SLC) System LC0 3.1.7 Two SLC subsystems shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION A. One SLC subsystem A.l Restore SLC subsystem inoperable. to OPERABLE status.

B. Two SLC subsystems B.1 Restore one SLC inoperable. I subsystem to OPERABLE I status.

C. Required Action and C.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time not met.

I C.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months SURVEILLANCE REQUIREMENTS I

SURVEILLANCE FREQUENCY 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Verify available volume of sodium pentaborate solution is within the limits of Figure 3.1.7-1.

(continued)

CLINTON Amendment No. 167

Attachment 3 Proposed Markup of CPS Technical Specification Bases B 3.1.7 TS Bases Page

SLC System B 3.1.7 BASES ACTIONS A 2 (continued) remaining OPERABLE subsystem could result in reduced SLC System shutdown capability. The 7 day Completion Time is based on the availability of an OPERABLE subsystem capable of performing the intended SLC System function and the low probability of a Design Basis Accident (DBA) or severe transient occurring concurrent with the failure of the Control Rod Drive System to shut down the plant.

B.l 44 hours5.092593e-4 days 0.0122 hours 7.275132e-5 weeks 1.6742e-5 months . The allowed Completion Time of P If both SLC subsystems are inoperable, at east one subsystem must be restored to OPERABLE sta us within 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 v

C.l and C.2 If any Required Action and associated Completion Time is not met, the plant must be brought to a MODE in which the LC0 does not apply. To achieve this status, the plant must be brought to MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and MODE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.1.7.1, SR 3.1.7.2, and SR 3.1.7.3 REQUIREMENTS SR 3.1.7.1 through SR 3.1.7.3 are 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Surveillances, verifying certain characteristics of the SLC System (i.e.,

the volume and temperature of the borated solution in the storage tank, and temperature of the pump suction piping),

thereby ensuring the SLC System OPERABILITY without disturbing normal plant operation. These Surveillances ensure the proper borated solution and temperature, including the temperature of the pump suction piping, are maintained. Maintaining a minimum specified borated solution temperature is important in ensuring that the boron remains in solution and does not precipitate out in the (continued)

CLINTON Revision No. 10-5

SLC System B 3.1.7 BASES REFERENCES 1. 10 CFR 50.62,

2. USAR, Section 9.3.5.3.

3. Calculation IP-0-0012.

4. Calculation IP-0-0013.

5. Calculation IP-0-0014.

6. Calculation IP-0-0015.

7. Calculation IP-0-0016.

8. NUREG-1465, "Accident Source Terms for Light-Water Nuclear Power Plants, Final Report," February 1, 1995.

9. 10 CFR 50.67, "Accident Source Terms."

L

10. RM Documentation No. CL-LAR-01, Revision 1, "Risk Assessment Input for Clinton Technical Specification Change for Standby Liquid Control System Completion Time from 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ," December 28, 2009 CLINTON Revision No. 10-5

Attachment 4 RM Documentation No. CL-LAR-01, Revision 1

7 RM DOCUMENTATION NO. CL-LAR-01 REV: 1 PAGE NO. 1 STATION: Clinton UNIT(S) AFFECTED: N/A r

TITLE: Risk Assessment Lnput for Clinton Technical Specification Change for Standby Liquid Control System Completion Time from 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months I

SUMMARY

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

The risk assessment is performed in accordance with ER-AA-600-1 046, Rev. 4, Risk Metrics - NOED and LAR No UREs have been created as a result of this application.

I

[ I Review required after periodic Update

[X 1 In ternirl RM Documentation [ 1 External WE Documentation Electronic CalcuIation Data Files:

Method of Review: [ X ] Detailed [ ] Alternate [ ] Review of External Document This RM documentation supersedes: CL-LAR-01 Rev 0 in its entirety.

I Prepared by: L. K. LeeR A Narain / &A $!%!z-Z/ '~-/2+4~(Z,l1vhq Reviewed by: R A Hill 2 Sign q://Ohr(

Date Reviewed by: G A Teagarden 7a I z / / o ! ~ ?

Sign Date Reviewed by: V M Andersen (Model Owner) / Z /z3/i17 Date Approved by: E T Burns l-L-?$-O?

Sign Date

Clinton SLC CT Extension TABLE OF CONTENTS

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 I.6 Clinton PRA Model ....................................................................................7 2.0 ANALYSIS ROADMAP AND REPORT ORGANIZATION .................................... 8 3.0 TIER 1 RISK ASSESSMENT ................................................................................ 9 3.1 Key Assumptions .......................................................................................9 3.2 InternalEvents ........................................................................................ 10 3.3 Results Comparison to Acceptance Guidelines ....................................... 12 3.4 External Events........................................................................................ 13 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 .......................................................................................................17

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

...................................................................................................37 APPENDICES A EXTERNAL EVENT ASSESSMENT B UNCERTAINW ANALYSIS C BWROG ASSESSMENT OF NRC INFORMATION NOTICE 2007-07

Clinton SLC CT Extension 1.O INTRODUCTION

1. I PURPOSE The purpose of this analysis is to assess the acceptability, from a risk perspective, of a change to the Clinton 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 0.00222 hours 1.322751e-5 weeks 3.044e-6 months to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months when both SLC subsystems (i.e., 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 Generating Company (EGC) 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 Clinton analysis is consistent with similar analyses being conducted for all EGC Boiling Water Reactor (BWR) plants that currently have an 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 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 sunley 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 and data and in a manner that complements the NRC's deterministic approach and supports the NRC's traditional defense-in-depth philosophy.

Clinton SLC CT Extension

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.I 74 [Ref. 21 and 1. I 77 [Ref. 31 both provide processes to incorporate PRA input for decision makers regarding a Technical Specification modification.

Clinton, other EGC 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), EGC requested Notices of Enforcement Discretion (NOEDs) for SLC System Tank leaks allowing an additional 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to the original &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 completion time for TS 3.1.7 Condition B from 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months (current TS) to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months (proposed TS). Condition B is the situation where both SLC subsystems are inoperable. Technical Specification requirements for other SLC conditions will remain unchanged. For Clinton the TS Condition B applies to Modes 1 and 2 for reactivity control. Consideration of TS applicability for Modes 1, 2, and 3 for pH control is not addressed in this report.

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.

Clinton SLC CT Extension 1.4.1 Re~ulatowGuide 1.174 Regulatory Guide 1.I 74 [Ref. 21 specifies an approach and acceptance guidelines for use of PRA in risk informed activities. RG 1.I 74 outlines PRA related acceptance guidelines for use of PRA metrics of Core Damage Frequency (CDF) and Large Early Release Frequency (LERF) for the evaluation of permanent TS changes. The guidelines given in RG 1.174 for determining what constitutes an acceptable permanent change specify that the ACDF and the ALERF associated with the change should be less than specified values, which are dependent on the baseline CDF and LERF, respectively. These specified values of ACDF and ALERF are given in RG 1.I 74 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 CLO6C baseline internal events CDF of 5.6E-6lyr and LERF of 1.2E-7lyr for Clinton, the RG 1.174 acceptance guidelines prescribe that the risk metrics of ACDF and ALERF be less than 1.OE-OGlyr and 1.OE-07/yr, respectively, to establish a very small risk increase with no additional compensatory measures required.

RG 1.I 74 also specifies guidelines for consideration of external events. External events can be evaluated in either a qualitative or quantitative manner.

1.4.2 Requlatow Guide 1.177 Regulatory Guide 1.174 [Ref. 21 specifies an approach and acceptance guidelines for the evaluation of plant licensing basis changes. RG 1.I 77 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.I 77 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 assurance that risk-significant plant equipment outage configurations will not occur when equipment associated with the proposed TS change is out-of-service.

Clinton SLC CT Extension 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 Requlatow Guide 1.200, Revision 1 Regulatory Guide 1.ZOO, Rev. 1 [Ref. I ] , 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.I 74.

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

[Ref. 51 applicable to full power internal event (FPIE) PRA models. Since that time, the new ASMEIANS Combined PRA Standard [Ref. 261 has been released. Although the Combined Standard is presently issued and endorsed by RG 1.ZOO Revision 2 [Ref. 271, neither of these document revisions impact this analysis.

Acce~tanceCriteria 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:

ACDF < 1.OE-O61yr ALERF < 1.OE-071yr ICCDP < 5.OE-07 ICLERP < 5.OE-08

Clinton SLC CT Extension I.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 Clinton CLO6C 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 FPlE model is further discussed in Section 1.6.

Low Power Operation - The FPlE assessment is judged to adequately capture risk contributors associated with low power plant operations. The FPlE 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 / Refuelinq - In consideration of shutdown and refueling modes (i.e., 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 Clinton. The Clinton Interim Fire PRA [Ref. 101, and a BWROG assessment [Ref. 191 are used to provide qualitative and semi-quantitative insights to the analysis (refer to Section 3.4.1).

Seismic - Consistent with most sites, Clinton does not currently maintain a Seismic PRA. A qualitative assessment is performed in this analysis (refer to Section 3.4.2) based on insights from the Clinton IPEEE study

[Ref. 1I]and other industry studies.

Other External Events - Other external event risks were assessed in the Clinton IPEEE study [Ref. 1I ] and found to be insignificant risk contributors (refer to Sections 3.4.3 and 3.4.4).

Clinton SLC CT Extension 1.6 CLINTON 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 Clinton CLO6C 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. This analysis is specific to the SLC system and therefore the SLC system fault tree model is the only portion of the CLO6C PRA model modified for this risk application. The Clinton SLC system is a manually initiated system with two SLC pump required to meet the10 CFR 50.62 requirements for ATWS response. 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 ISC-1A-1B----M-- "SBLC A AND SBLC B IN COINCIDENT MAINTENANCE," 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 B).

No other aspect of the CLO6C PRA model required adjustment for this risk application.

The entire CLO6C PRA model is quantified for this assessment using the "average maintenance" PRA model (i.e., no portions of the at-power internal events CLO6C model were excluded or zeroed out of the quantification).

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

[Ref. I]. 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

.a Identify plant changes (design or operational practices) that have n incorporated at the site, but are not yet in the PRA model and fy why the change does not impact the PRA results used to support consistent with applicable standards endorsed by the RG (currently, in RG 1.ZOO Rev. I.RG 1.ZOO Rev. 1 addresses the internal events ASME PRA standard). Provide justification to show that where specific equirernents in the standard are not met, it will not unduly impact the ed for the application, and for justify why the significant 4.d Identify key assumptions and approximations relevant to the results Section 3.1 used in the decision-making process.

Clinton SLC CT Extension 3.0 TIER IRISK ASSESSMENT 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 0.00222 hours 1.322751e-5 weeks 3.044e-6 months to a proposed duration of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

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 past 24 months as of June 22, 2009 revealed that SLC Trains A and B were not both unavailable. Train A had unavailable 2.7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months , and Train B had not been unavailable. 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 Clinton 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 582 hours0.00674 days 0.162 hours 9.623016e-4 weeks 2.21451e-4 months 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 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 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.

Clinton SLC CT Extension 3.2 INTERNAL EVENTS The Clinton 2006C PRA model(') [Ref. 41 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 2006C PRA model was identified as 1SC-1A-I B----M--

"SBLC A AND SBLC B IN COINCIDENT MAINTENANCE." This event is appropriate because it fails both SLC subsystems and no other equipment in the model.

Event ISC-1A-I B----M-- was set to a binary logic value of "TRUE" (using a quantification flag file) and the entire CLO6C model was requantified using the same PRA software codes and revisions as used for the base CLO6C model [Ref. 41. These configuration specific CDF and LERF values are used in conjunction with the base CLO6C values to calculate the risk impacts of the proposed TS change.

The calculations of ACDF, ICCDP, ALERF and ICLERP for the CT change are determined as shown below.

The ACDF to be compared to the RG 1.I 74 acceptance guidelines is given by (as defined by [Ref. 211):

ACDF = CDFNEW - CDFBASE [Equation 3-11 ACDF is the difference between the annual average CDF with the CT extended and the CDF with the current CT. The ACDF has units of "per reactor year."

In the above equation, CDFNEW is equal to:

CDFNEW= CTSLC-00s

CDFSLC-00s + [(l-CTS~C-OOS )

CDFBASE] [Equation 3-21 Where:

CDFsLc-oos= the annual average CDF calculated with both SLC subsystems out of service (1SC-1A-1B----M-- set to TRUE)

CDFBAsE= baseline annual average CDF with average unavailability for all equipment. This is the CDF result of the CLO6C baseline PRA.

C T ~ = the ~ new

~ extended

- ~ CT ~ as ~ an annual unavailability (i.e., 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 1 8760 hour0.101 days 2.433 hours 0.0145 weeks 0.00333 months slyr = 8.2E-03 yr)

I The CLOGC baseline model used in the calculations contains the average maintenance associated with system trains.

Clinton SLC CT Extension CTsLc-oos= the new extended CT as a probability (i.e., 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months / 8760 hours0.101 days 2.433 hours 0.0145 weeks 0.00333 months = 8.2E-03)

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

[Equation 3-31 Risk significance relative to ALERF and ICLERP(') is determined using equations of the same form as noted above for ACDF 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 CLOGC model and use of the above equations) are provided in Table 3.2-2.

Table 3.2-1 RISK ASSESSMENT INPUT PARAMETERS

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

Clinton SLC CT Extension Table 3.2-2 RISK ASSESSMENT BASE RESULTS 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 ACDF, ICCDP, ALERF, and ICLERP risk metric values are below the acceptance guidelines as defined in RG 1.I 74 and RG 1.I 77. 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.I74 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.I 77, and the CTNEWallowable was calculated. ICLERP was determined to be the bounding parameter, and a CTNEWof 582 hours0.00674 days 0.162 hours 9.623016e-4 weeks 2.21451e-4 months was calculated. This represents a significant margin compared to the proposed 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months CT.

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

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

NUREGICR-6850 [Ref. 181 Clinton Interim FPRA [Ref. 101 BWROG Assessment of Fire-Induced Failure to Scram [Ref. 191 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 EGC does not currently maintain a seismic PRA for Clinton. The impact on the seismic risk profile due to the proposed CT is evaluated using the following information sources:

CPS IPEEE [Ref. 1I ]

NUREG-1150 [Ref. 231 The seismic risk impact assessment is discussed in Appendix A.3. The assessment concluded that seismic risk can be appropriately screened as a non-significant contributor to the risk assessment of the proposed CT.

3.4.3 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 IPEEE analysis. The Clinton site characteristics and design meet all the applicable criteria of the NRC Standard Review Plan (SRP). No significant quantitative contribution from these external events was identified by IPEEE evaluations (refer to Appendix A.2).

As such, other external hazards are appropriately screened as non-significant contributors to the risk assessment of the proposed CT.

Clinton SLC CT Extension 3.5 UNCERTAINTY ASSESSMENT 3.5.1 Parametric Uncertaint~

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.

Modeling Uncertaintv An assessment of modeling uncertainty is documented in Sections B.l and B.2. The results of these modeling uncertainty assessments are judged not to change the conclusions of this risk assessment for the proposed SLC CT change as they do not directly impact the SLC system or ATWS scenarios.

Section B.1 provides the Clinton specific modeling uncertainty evaluations for the Base 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.

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

3.6 RISK

SUMMARY

As discussed above and as summarized in Table 3.6-1, the FPlE 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.

Clinton SLC CT Extension Table 3.6-1 RISK ASSESSMENT

SUMMARY

RESULTS

('I Evaluated and determined not to change the conclusions of the FPlE risk analysis.

Clinton SLC CT Extension 4.0 TECHNICAL ADEQUACY OF PRA MODEL The 2006C update to the Clinton PRA model (CLOGC) is the most recent evaluation of the risk profile at Clinton for FPlE challenges. The Clinton 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 Clinton PRA is based on the event tree I fault tree methodology, which is a well-known methodology in the industry.

EGC employs a multi-faceted approach to establishing and maintaining the technical adequacy and plant fidelity of the PRA models for all operating EGC 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 Clinton PRA.

4. I PRA QUALITY OVERVIEW The quality of the Clinton FPlE 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 consen/atism..."

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

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

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

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

- Through more efficient use of agency resources; and

- Through a reduction in unnecessary burdens on licensees."

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

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

The scope is interpreted to address the following aspects:

- Challenges to plant operation (Initiating Events):

9 Internal Events (including Internal Floods) 9 External Hazards 9 Fires

- Plant Operational states:

9 Full Power 9 Low Power 9 Shutdown

- The metrics used in the quantification:

9 Level IPRA - CDF 9 Level 2 PRA - LERF 9 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. 51 as endorsed by the NRC in Regulatory Guide 1.200 [Ref. I ] .

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.

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

4.2 SCOPE The Clinton PRA is a full power, internal events (FPIE) PRA that addresses both CDF and LERF. The quantitative insights from the FPlE 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

Clinton SLC CT Extension available or endorsed, this approach uses qualitative insights or bounding approaches as needed.

The quality assessment performed in this section confirms the adequacy of the FPIE PRA. This 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 EGC 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 EGC Risk Management program, which consists of a governing procedure (ER-AA-600, "Risk Management") and subordinate implementation procedures. EGC procedure ER-AA-600-1015, "FPIE PRA Model Update" delineates the responsibilities and guidelines for updating the full power internal events PRA models at all operating EGG nuclear generation sites. The overall EGC Risk Management program, including ER-AA-600-1015, defines the process for implementing regularly scheduled and interim PRA model updates, for tracking issues identified as potentially affecting the PRA models (e.g., due to changes in the plant, errors or limitations identified in the model, industry 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.

In addition to these activities, EGC 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 EGC nuclear generation sites.

Clinton SLC CT Extension 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. 51 provides the basis for assessing the adequacy of the Clinton PRA as endorsed by the NRC in RG 1.200, Rev. 1 [Ref. I]. 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. Table 4-1 includes the continuing investigations by EGC of plant modifications and changes that could influence the risk spectrum.

A PRA Peer Review to allow independent reviewers from outside to examine the model and documentation. The ASME PRA Standard [Ref.

51 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.

Several assessments of technical capability have been made and continue to be planned for the Clinton 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 (BWROG) in 2000, following the Industry PRA Peer Review process [Ref. 61. This peer review included an assessment of the PRA model maintenance and update process.

Clinton SLC CT Extension A self-assessment analysis was performed against Addenda B of the ASME PRA Standard and the draft of Revision. Iof Regulatory Guide 1.200 (DG-1161).

During 2005 and 2006 the CPS PRA model results were evaluated in the BWROG PRA cross-comparisons study performed in support of implementation of the mitigating systems performance indicator (MSPI) process.

A current industry peer review of the Clinton PRA was conducted in the fourth quarter of 2009. Results of this review are still being processed.

A summary of the disposition of the BWROG PRA Peer Review facts and observations (F&Os) for the Clinton PRA models was documented as part of the statement of PRA capability for MSPl in the Clinton MSPl Basis Document [Ref. 71. As noted in that document, all five (5) of the significance level " A F&Os have been resolved and eighty-nine (89) of the ninety-two (92) significance level "B" F&Os have been resolved. The remaining three (3) open significance level "B" F&Os are insignificant.

4.5.1 Self-Assessment Overview A Self-Assessment of the 2003 CPS PRA was performed in support of the CPS 2006 PRA Update. This Gap Analysis was performed using Addenda B of the ASME PRA Standard (ASME RA-Sb-2005) and the draft of Revision 1 Regulatory Guide 1.200 (DG-1161). Potential gaps to Capability Category II of the Standard were identified and used to plan the Clinton 2006 PRA Update. Table 4-3 presents a discussion of the identified gaps and concludes that none impact this application.

4.5.2 PRA Peer Review Overview Table 4-2 presents the open significant PRA Peer Review findings. PRAs can be used in applications despite not meeting all of the Supporting Requirements of the ASMEIANS PRA Standard. This is well recognized by the NRC and is explicitly stated in the 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

Clinton SLC CT Extension Capability Category Ill for all of its SRs, nor is this currently expected as part of the NRC PRA Quality Program.

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 License Amendment Request (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 EGC's PRA model update tracking database. These UREs are created for all issues that are identified with a potential to impact the PRA model. The URE database includes the identification of those plant changes that could impact the PRA model. A review of the current open items in the URE database associated with plant changes for Clinton as well as items related to SLC or ATWS modeling 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.

Consistency with Applicable PRA Standards As indicated above, an independent peer review of the Clinton PRA was performed in 2000 following the review guidelines of NEI 00-02 (the predecessor to the ASME PRE Standard). All of the significance level " A F&Os have been resolved and all but three (3) of the significance level "B" have been resolved. The three open significance level "B" F&Os from the peer review are summarized in Table 4-2 along with an assessment of the impact for this application. None of the three are found to impact this application.

4.6.3 Relevant PRA Peer Review Findings As indicated above, a current industry peer review of the Clinton PRA is scheduled for the fourth quarter of 2009. However, a self-assessment against the PRA Standard and draft RG 1.200 was performed in support of the CPS 2006 PRA Update. Potential gaps to Capability Category II of the Standard were identified for treatment in the 2006 PRA Update. The identified gaps that were not closed by the 2006 PRA Update are summarized in Table 4-3 along with an assessment of the impact for this application.

Clinton SLC CT Extension Of the gaps identified and evaluated in Table 4-3, none have a measurable impact on the SLC CT extension request.

Clinton SLC CT Extension Table 4-1 IMPACT ON THE CLINTON PRA MODEL OF PLANT CHANGES SINCE THE LAST PRA UPDATE Non-significant impact.

the PRA, basic event DDCI EIACBD (prob 2.4E-5), although the breaker has no modeled function other than to remain closed.

This BE represents an inadvertent change of state for the breaker. A combination of switches and fuses could have a somewhat different failure rate, than a DC breaker. However Transformer is being replaced with 3 RATs will also result in changes to the low side buswork. This is generally covered nder EC 339047 "RAT Replacement Project-Transformers,

-Seg Bus, Relay Panels, Power and Control Cabling for RAT Voltage Side". The PRA model needs to have the offsite er logic split to correspond to its RAT supply. This is overed by URE CL2005-001. Another feature of this change is hat the fault protection scheme for the bus ductwork and for the ew RATs is being modified. This is generally below the level of etail considered in PRA models and can be considered in the ategory of events that would cause inadvertent opening (or

Clinton SL C CT Extension Table 4-1 IMPACT ON THE CLINTON PRA MODEL OF PLANT CHANGES SINCE THE LAST PRA UPDATE nd is enhanced to cover several situations that may be useful r addressing PRA accident sequences. Noteworthy topics clude: Running RClC without DC power, Starting Diesels ithout DC power for field flashing or air start solenoids, ntervals. For example CPS 9054.02 RClC Valve Operability as been modified to reduce the testing interval for 1E51F063 nd 1E51F064 from quarterly to cold shutdown test intervals.

his could have a minor impact on the data analysis for omponents. Note these two RClC valves do not have fail to on should be revised as OG-F-- FALSE INDICATION OF COND.

Clinton SLC CT Extension Table 4-1 IMPACT ON THE CLINTON PRA MODEL OF PLANT CHANGES SINCE THE LAST PRA UPDATE ource (from the same DC bus) and the same AC transformer ce and a static transfer switch that can swap between them.

result the basic PRA logic and power dependencies remain e same. The internal design of the inverters is different so may not have all the same the original inverters.

t be replaced in the short omewhat different from those modeled in the PRA: impact would be reflected in plant-specific transient initiator

. This procedure allows 4 condensate pump operation.

ecause this provides additional NPSH margin to feedpump ips, the plant operators have pretty much elected to run 4 umps all the time. Note one of the pumps (A or D depending on which one is selected) will automatically trip upon a bus sfer from UAT to RAT (such as occurs on a normal plant This prevents overloading the RAT. So reality is we run 4, ave 3 left immediately after a scram and we really only need to support post scram operation of the MDRFP. I think the es that there are only two CD and two CB pumps at others can be started; this modeling is

Clinton SLC CT Extension Table 4-1 IMPACT ON THE CLINTON PRA MODEL OF PLANT CHANGES SINCE THE LAST PRA UPDATE Rev. 030a. This action had previously been deleted

1. Procedure 3214.01 now provides steps for bypassing SA

2. Procedure 3214.01 now provides steps for bypassing IA system isolations for particular ring header isolation valves.

Clinton SLC CT Extension Table 4-1 IMPACT ON THE CLINTON PRA MODEL OF PLANT CHANGES SINCE THE LAST PRA UPDATE ome set of transient conditions that they previously would not.

world improvement, it is below the level of nt model (accounted for in the loss of IA IE some types of events, such as a turbine trip (and EC369429 evaluates the impact of this change on transient analysis). Since most of the PRA IEs involve a turbine trip, the ATWS RPT logic now may be backed up by the RR pump trip occurring with a

Clinton SL C CT Extension Table 4-1 IMPACT ON THE CLINTON PRA MODEL OF PLANT CHANGES SINCE THE LAST PRA UPDATE main feed breaker is being replaced, with a switch fuse combination. This may result in a slightly different failure rate than the original breaker. This change is expected to have a very minor change on calculated PRA results because inadvertent opening of passive electrical connections tend to have relatively low failure rates regardless of the device.

This EC replaces the main feed breaker for DC MCC 1DC17E with a switch and fuses combination that will fit into the existing DC breaker cubicle. The normally closed breaker is modeled in the PRA, basic event 1DCCB-DC1FIA-U-- (prob 1.2E-5),

although the breaker has no modeled function other than to remain closed.

This BE represents an inadvertent change of state for the breaker. A combination of switches and fuses could have a somewhat different failure rate, than a DC breaker. However because in both cases a passive function is modeled the failure 4, the switchyard breakers are included in the CPS offsite is unlikely to contribute existing cabling to

Clinton SL C CT Extension Table 4-2 IMPACT OF OPEN SIGNIFICANT PRA PEER REVIEW FINDINGS FOR THE CLINTON PRA MODEL Additional plant specific room heat-up calculations (or enhancements to existing slowly evolving nature of the phenomena and the fact that the control room is continuously manned and doors could be opened. The AC switchgear room modeling does not require room cooling, based on the large size of these Room cooling modeling assumptions are assessed as a sensitivity study and do not change the conclusions of this risk application (see Appendix B).

calculations for each pre-initiator. The impact on the model is non-significant, pre-initiator HEPs contribute The documentation of the internal flooding Documentation issue. No impact.

Clinton SL C CT Extension Table 4-3 IMPACT OF OPEN SELF-ASSESSMENT ITEMS FOR THE CLINTON PRA MODEL No Impact - documentation item.

Deferred: This documentation aspect has not been A rigorous explicit assessment of all the events in NUREG- incorporated into the CPS PRA notebooks. This work 1275 could be pursued (if determined that this is the true was performed for another BWR plant (review of intent of SR IE-A7); however, such an effort is judged not hundreds of events INPO SENs, SOERs, SERs, and to provide much benefit to the CPS IE analysis. NRC SECY letters on precursors) and no new initiating events were identified. It is expected that future industry studies will make provide this generic supported by room cooling calculation.

this risk application (see Appendix B).

Deferred: Room cooling calculations have not been performed at this time and are being considered in the

Clinton SLC CT Extension Table 4-3 IMPACT OF OPEN SELF-ASSESSMENT ITEMS FOR THE CLINTON PRA MODEL

1) A list of the PF% systems to consider for test and Deferred: The current methodology and documentation maintenance actions for identifying pre-initiator HEPs is judged adequate.

Any additional documentation to conform to inferred

2) Rules for identifying and screening test and maintenance actions from the PRA requirements of pre-initiator identification SRs would be documentation enhancement and not result in

3) A list of procedures reviewed, the potential test and increasing the number of pre-initiator HEPs included in maintenance actions associated with the procedures, the model. The CPS PRA includes over 100 pre-and the disposition of the action (screened or initiator HEPs in the model. Other BWRs that have attempted to rigorously follow these pre-initiator HEP

4) Identify T&M activities that require realignment of the SRs have resulted in explicitly modeling significantly system outside its normal operational or stand by less pre-initiator HEPs in the model.

Deferred: Future updates of the CPS PRA will consider explicit/specific pre-initiator HEP calculations. The as human-machine interface for both pre-initiator and post- current calculations are based on representative initiator human actions. procedures/practicesfor similar pre-initiator HEPs. The current estimates are generally higher in error rates Possible upgrade to the pre-initiator HRA to include than would be obtained if various explicit recovery specific quantifications for each pre-initiator HEP would be factors and testing frequencies were applied in specific strict compliance with the standard. This is not considered HEP calculations for each pre-initiator. The impact on necessary for most applications. It is recommended that the model is non-significant, pre-initiator HEPs n this item before contribute approximately 2% to the CLOGC CDF.

each application in

Clinton SL C CT Extension Table 4-3 IMPACT OF OPEN SELF-ASSESSMENT ITEMS FOR THE CLINTON PRA MODEL Non-significant impact.

documented appropriately. Review surveillance test Deferred: Current industry PRA efforts and PRA peer procedures and identify all failure modes that are fully reviews are having difficulty understanding the full intent tested by the procedures. Include data for the failure of this SR. Future updates of the CPS PRA will modes that are fully tested. The results of unplanned documentation and demands on equipment should also be accounted for.

incomplete or limited maintenance information and Deferred: Future updates of the CPS PRA will consider document appropriately. performance of interviews of plant personnel to supplement maintenance unavailability estimates for equipment with limited maintenance information. Any refinements to maintenance unavailabilities are judged ce information -

at the time of the

Clinton SLC CT Extension Table 4-3 IMPACT OF OPEN SELF-ASSESSMENT ITEMS FOR THE CLINTON PRA MODEL No Impact - documentation item.

Significant events and cutsets to this risk application are quantified using fault tree methods. identified in Appendix B of this report.

Otherwise, importance measures calculated and assessed to ensure results make logical sense. Deferred: This documentation aspect has not been incorporated into the CPS PRA notebooks. Initiating event fault trees are not linked into the accident sequence models. Documentation of the importance of a) Incorporate an overview of the quantification process. Deferred: Not all items incorporated into the CPS 2006 b) Provide a list of human actions and equipment failures PRA Update. Items (a) and (c) are incorporated into the (significant basic events) that cause accidents to be quantification documentation of the CPS 2006 update.

non-dominant. Items (b), (d) and (e) are documentation enhancements c) Refer to the Disposition for SR QU-E4 regarding for the base PRA and are maintained for consideration assumptions, sources of uncertainty and related for future updates.

sensitivity assessments.

d) Bases for the elimination of mutually exclusive events from the model need to be added.

e) Include cutsets segregated by accident sequence in the documentation. This is available but may not be needed in the formal documentation. This should await

Clinton SLC CT Extension 4.7 GENERAL CONCLUSION REGARDING PRA CAPABILITY The Clinton 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.

Clinton SLC CT Extension 5.0

SUMMARY

AND CONCLUSIONS

5. I SCOPE INVESTIGATED This analysis evaluates the acceptability, from a risk perspective, of a change to the Clinton TS for the SLC system to increase the CT from 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months when both SLC subsystems (i.e., both trains) are inoperable.

The analysis examines a range of risk contributors as follows:

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

The FPlE 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., NUREGICR-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 Clinton IPEEE and the NUREG-1150 study.

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

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

Scope - The Clinton PRA modeling is highly detailed, including a wide variety of initiating events, modeled systems, operator actions, and common cause events. The PRA has the necessary scope to appropriately assess the pertinent risk contributors.

Fidelitv - The Clinton PRA model (CLO6C) is the most recent evaluation of the risk profile at Clinton for FPlE challenges. The PRA reflects the as-built, as-operated plant.

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

[Ref. 51 and the PRA elements are shown to have the necessary attributes to assess risk for this application.

Peer Review - The PRA received a Peer Review in 2000. Based on the Peer Review results and the incorporation of Peer Review comments, the PRA is found to have the necessary attributes to assess risk for this application.

Clinton SLC CT Extension Appropriate Qualitv - 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 ACDF, ICCDP, ALERF, and ICLERP risk metric values are below the acceptance guidelines as defined in the corresponding risk significance guidelines from RG 1.I 74 and RG 1.I 77.

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(') Acceptance Guidelines ACDF 2.9E-081yr < 1.OE-O6Iyr ICCDP 2.9E-08 <5.OE-07 ALERF 6.2E-091yr < I .OE-07Iyr L

ICLERP 6.2E-09 <5.OE-08

5.4 CONCLUSION

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

This analysis demonstrates that the proposed TS change satisfies the risk acceptance guidelines in RG 1.I 74 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 RG 1.200, Revision 1. This 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.

Clinton SLC CT Extension REFERENCES RG 1.ZOO, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk lnformed Activities," Revision 1, January 2007.

RG 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-lnformed Decisions on Plant-Specific Changes to the Licensing Basis," Revision I , November 2002.

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

Technical Specifications," August 1998.

Exelon Risk Management Team, CPS-PRA-014, Clinton Power Station Probabilistic Risk Assessment Quantification Notebook, CLO6C, Revision 4, March 2007.

"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.

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

Clinton MSPl Basis Document, Rev. 3, December 12, 2008.

Drouin, M. et al., "Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decision Making", NUREG-1855, March 2009.

Treatment of Parameter and Model Uncertainty for Probabilistic Risk Assessments, EPRl Report 1016737, Palo Alto, CA: 2008.

CPS-PSA-021.06, "Clinton FPRA Summary and Quantification Report, Rev. 0, September 2008.

AmerGen, "Clinton Power Station Individual Plant Examination for External Events," September 1995 "PRA Procedures Guide", NUREGICR-2300, September 1981.

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

NUREGICR-4550, Volume 4, Revision 1, Part 3, Table 4.14, page 4-83.

NUREGICR-5042, "Evaluation of External Hazards to Nuclear Power Plants in the United States," December 1987.

Clinton SL C CT Extension Not Used.

NUREGICR-5500, "Reliability Study: General Electric Reactor Protection System, 1984-1995, Volume 3" May 1999.

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

NUREGICR-4551, December 1993.

NUREGICR-6850, EPRI Report 1011989, "Fire PRA Methodology for Nuclear Power Facilities", September 2005.

Gorman, Thomas, BWR Owners' Group (BWROG), "BWROG Assessment of IN 2007-07", 1011612007.

Not Used.

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

Not Used.

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

NUREG-1150, December 1990.

Not Used.

Electric Power Research Institute, "A Methodology for Assessment of Nuclear Power Plant Seismic Margin", EPRl NP-6041, October 1988.

ASMEIANS RA-Sa-2009, "Addenda to RA-S-2008, Standard for Level 1lLarge Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications," February 2009.

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

Clinton SLC CT Extension Appendix A External Event Assessment A. 1 INTRODUCTION This appendix discusses the external events assessment in support of the Clinton SLC system CT extension risk assessment. This appendix uses as the starting point of this assessment the external event work documented in the Clinton Individual Plant Examination of External Events (IPEEE) [Ref. A-I].

Because the effects of the SLC CT extension are evident only in the failure to scram (Anticipated Transients Without Scram (ATWS)) related sequences, the following examination of external events focuses on the ATWS accident sequence insights.

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

There is no currently maintained quantitative Seismic PRA for Clinton. Section A.3 discusses seismic ATWS insights from the CPS IPEEE and NUREG-1150.

Internal Fires This internal fire assessment is based on the Interim Clinton Fire PRA (FPRA) model developed in 2008 and generic assessments in NUREGICR-6850 and the BWROG assessment of IN 2007-07. This assessment is discussed in Section A.4.

Other External Hazards The Clinton plant design with respect to external flooding meets all the applicable criteria of the NRC Standard Review Plan (SRP). Core damage accidents induced by external flooding are negligible contributors to plant risk.

Other external event risks such as severe weather, high winds or tornados, transportation accidents, nearby facility accidents, turbine missiles, and other

Clinton SL C CT Extension miscellaneous external hazards were also considered in the IPEEE analysis. The Clinton site characteristics and design meet all the applicable criteria of the NRC SRP.

No significant quantitative contribution from these external events was identified by 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, other external hazards are assessed to be insignificant contributors to plant risk. Explicit treatment of the "other" external hazards is not necessary for most PSA 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 justify the screening of Fire and Seismic hazards for the SLC CT extension application.

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

A.3.1 Clinton Seismic IPEEE Overview Clinton performed a seismic margins assessment (SMA) as part of the IPEEE, following the guidance of EPRl NP-6041. [Ref. A-21 The SMA is a deterministic evaluation process that does not calculate risk on a probabilistic basis. No core damage frequency sequences were quantified as part of the seismic risk evaluation.

The conclusions of the Clinton seismic risk analysis are as follows: [Ref. A-I]

"No improvements to the plant were identified as a result of the Seismic Margins Assessment ... the plant was determined to be fully capable of attaining safe shutdown conditions afler the Review Level Earthquake (RLE). "

Based on a review of the Clinton IPEEE and the conclusions identified earlier in this assessment, the conclusions of the SMA are unaffected by the SLC CT extension. The SLC CT extension has no impact on the seismic qualifications of the SSCs.

A.3.2 Peach Bottom NUREG-1150 Seismic Overview The NUREGICR-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. Peach Bottom utilized the Seismic Margins Analysis as part of the IPEEE. It

Clinton SLC CT Extension is reasonably assumed that the seismic ATWS risk portion of the analysis is generically appropriate for all BWRs due to the similarity of CRD and SLC systems.

The NUREGICR-4551 Peach Bottom seismic analysis screened seismic-induced ATWS accident sequences as non-significant contributors ( 4 %) to the plant seismic CDF.

Based on the Peach Bottom results, it is judged that seismic-induced ATWS accident sequences are similarly non-significant contributors to Clinton plant seismic CDF.

A.3.3 Seismic Risk Impact Conclusion Based on the preceding discussions, it is concluded that the risk of a seismically induced ATWS is non-significant and does not impact the decision-making for the proposed CPS SLC CT extension.

A.4 INTERNAL FIRES ASSESSMENT This internal fire assessment is based on the Clinton Interim Fire PRA (FPRA) model developed in 2008 and generic assessments in NUREGICR-6850 [Ref. A-31 and the BWROG assessment of IN 2007-07 [Ref. A-51.

NUREGICR-6850 Screening NUREGICR-6850, Volume 2, Section 2.5.1 (page 2-7) [Ref. A-31 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 on'en be easily demonstrated that anticipated transient without scram (ATWS) 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. "

As can be seen from the NUREGICR-6850 excerpt above, fire-induced ATWS contributors are generally acknowledged as non-significant contributors to the fire risk profile.

Clinton SLC CT Extension Clinton Interim Fire PRA The current Clinton FPRA [Ref. A-41 is an interim implementation of NUREGICR-6850; that is, not all tasks identified in NUREGICR-6850 are yet completely addressed or implemented due to the changing state-of-the-art of industry at the time of the 2007-2008 Clinton FPRA development.

NUREGICR-6850 task limitations and other precautions regarding the 2007-2008 FPRA upgrade for Clinton are as follows:

Multiple Spurious Operation (MSO) Review (NUREGICR-6850 Task 2)

- MSOs are reviewed and considered; however, an expert panel is not used. At the time of the Clinton FPRA development the BWR Owners' Group was developing a generic list of MSOs to be considered. In future CPS FPRA updates this list will be reviewed and incorporated as necessary.

Instrumentation Review (NUREGICR-6850 Task 2) - The new requirements of NUREGICR-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 (NUREGICR-6850 Task 2) - The BOP is not fully treated. BOP support system failure is conservatively assumed.

Additional modeling could be conducted to reduce the fire CDF due to this assumption if time and funding is available in future updates.

Large Early Release Frequency (LERF) (NUREGICR-6850 Task 2) -

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

Limited Analysis Iterations (NUREGICR-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 2008 Clinton FPRA may benefit from further refinement as necessary for application or for future updates.

Multi-Compartment Review (NUREGICR-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 Clinton make fire propagation to multiple compartments unlikely compared to the fire risk in individual compartments.

Clinton SLC CT Extension Seismic Fire Interactions (NUREGICR-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.

Uncertainty and Sensitivity Analysis (NUREGICR-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:

ltem I(MSO), represents a source of additional fire CDF contribution (i.e., if the BWROG MSO list includes MSOs not addressed in this update).

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

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

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

ltems 5 (Iterations) and 6 (Multi-compartment) represent modeling assumptions that should be reviewed with each FPRA application to determine their applicability andlor potential impact on the decision.

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

Given the above, the 2008 Clinton 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.

Based on the interim CPS Fire PRA, fire-induced ATWS CDF is approximately 4E-8/yr

( 4 % of CPS FPRA CDF). This is approximately an order of magnitude lower than the CPS internal events ATWS CDF. As such, like NUREGICR-6850, the CPS interim FPRA shows that fire-induced ATWS is a non-significant contributor to the plant risk profile and does not impact the decision-making of the proposed CPS SLC CT extension.

Clinton SLC CT Extension 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-51 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 CPS SLC CT extension.

Clinton SLC CT Extension A.5 REFERENCES

[A-I] AmerGen, "Clinton Power Station Individual Plant Examination for External Events," September 1995

[A-21 A methodology for assessment of nuclear power plant seismic margin, EPRl NP-6041, Palo Alto, CA: 2001.

[A-31 NUREGICR-6850, EPRl Report 1011989, "Fire PRA Methodology for Nuclear Power Facilities", September 2005.

[A-41 CPS-PSA-021.06, "Clinton FPRA Summary and Quantification Report", Rev. 0, September 2008.

[A-51 Gorman, Thomas, BWR Owners Group (BWROG), "BWROG Assessment of IN 2007-07", 1011612007.

Clinton SLC CT Extension Appendix B Uncertainty Analysis This appendix evaluates uncertainties that could impact the SLC CT extension assessment. Section B.1 and B.2 evaluate model uncertainties. Section B.3 evaluates parametric uncertainty.

Section B.1 provides Clinton specific modeling uncertainty evaluations for the base 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.

Section 8.3 documents the parametric uncertainty analysis of the model used in this application.

B. 1 MODEL UNCERTAINTIES

SUMMARY

Postulated key modeling uncertainties are identified through a systematic structured process [Ref. 9-11, Table B-1 presents the candidate key modeling uncertainties for the CLO6C model. The five modeling uncertainties that can be considered important model uncertainty are summarized in Table 8-2 along with associated impacts on the CDF and LERF risk metrics.

It is noted that none of these five cases presented in Table 8-2 evaluates modeling issues associated with the SLC system or ATWS sequences.

Clinton SLC CT Extension Table B=1

SUMMARY

OF SENSITIVITY CASES TO IDENTIFY RISK METRIC CHANGES ASSOCIATED WITH CANDIDATE MODELING UNCERTAINTIES(~)

1A) Applicability of industry experience 5.66E-06 5.54E-06 1.21E-07 1.20E-07 to environmentally influenced events (i.e., loss of service water, LOOP, etc.) - Loss of Service Water 1B) Applicability of industry experience -6.2~-06(~) (6) -1.21E-07'~) (6) to environmentally influenced events (i.e., loss of service water, LOOP, etc.) - Loss of Intake Structure 1C) Applicability of industry experience 7.78E-06 4.79E-06 1.52E-07 1.09E-07 to environmentally influenced events (i.e., loss of service water, LOOP, etc.) - Severe and Extreme Weather Induced LOOP 2A) Treatment of Rare and Extremely 5.58E-06 5.57E-06 1.23E-07 I.19E-07 Rare Events - Excessive LOCA 2B) Treatment of Rare and Extremely 5.90E-06 5.46E-06 1.22E-07 1.20E-07 Rare Events - SW Flood in RB 3), 4), 6), 1I ) , 17), 24) Beyond Design 8.57E-06 4.51E-06 1.46E-07 1.11E-07 Basis Environment

5) and 8) Case A) Impact of LOOPISBO 6.59E-06 5.16E-06 1.20E-07 1.20E-07 conditions on allowable AC Recovery

5) and 8) Case B) Impact of LOOPISBO 5.57E-06 4.07E-06 1.20E-07 1.05E-07 conditions - DFP injection 7), 12), 18) Room Cooling Assumptions 1.77E-05 4.22E-06 1.31E-07 1.19E-07

9) & 15) Impact of venting on systems 5.92E-06 5.52E-06 1.20E-07 1.20E-07 (3) (3) (3) (3)

10) Time Dependency failures due to environmental conditions (3) (3) (3) (3)

13) Recirc Pump Seal Leakage

14) Suppression Pool Strainer 5.84E-06 5.48E-06 1.21E-07 1.20E-07 Performance

16) Treatment of Instrumentation 6.77~-06(~' -5.2~-06(~' -

3.13 ~ - 0 7 ( ~ ) 1 . 0 ~ - 0 7 ( ~ )

required for operator action

Clinton SLC CT Extension Table B-1

SUMMARY

OF SENSITIVITY CASES TO IDENTIFY RISK METRIC CHANGES ASSOCIATED WITH CANDIDATE MODELING UNCERTAINTIES(~)

Notes to Table B-I;

('I Compared with a CLO6C base CDF of 5.57E-61yr quantified with a 1E-I llyr truncation limit.

'2' Compared with a CLOGC base LERF of 1.20E-07lyr quantified with a 1E-12lyr truncation limit.

(3) Subsumed by Case 518.

'4) Based on installed CPS system for suppression pool cooling, this candidate modeling uncertainty identified for other BWRs is considered not to be quantitatively significant and does not lead to a key modeling uncertainty.

(5) Most of the sensitivity results were produced by manipulating the cutset results file. These results were produced by re-quantifying the entire model.

These lower bound cases not performed; interest is in the increase in CDF and LERF.

('I Estimate for 2006C is based on sensitivity case results using the 20068 model.

Clinton SLC CT Extension Table B-2 IMPORTANT MODELING UNCERTAINTY CASES Notes to Table 8-2;

'2' Compared with a CLO6C base CDF of 5.57E-6lyr quantified with a 1E-I l l y r truncation limit.

Compared with a CLO6C base LERF of 1.20E-071yr quantified with a 1E-121yr truncation limit.

(3) These changes in the risk metric are below 2.0, but they are retained for identification to the decision-makers.

Clinton SL C 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 out-of-service case cutsets and remove the base case cutsets. This is done in CAFTA through the delete term function of the cutest editor.

The result of this process is 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 determine the change in risk metrics, i.e., drive the delta-CDF assessment.

Table 9-3presents the top ten cutsets for the delta-CDF assessment. Table 8-4 presents the importance measures associated with the delta-CDF assessment.

Tables B-3 and 6-4show 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. The first nine represent single failures that lead to core damage. The tenth cutset is an electrical failure of the scram system coupled with operator failures to manually scram. Of the 76 events appearing in Table B-4, only 8 are basic events, with the rest being initiators.

This is due to the fact that the cutsets associated with the SLC system out-of-service are again predominantly single failures of the Scram system leading to core damage.

It can be concluded that the SLC out-of-service case 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.

Similarly, the LERF results are dominated by failures of the Scram system for the SLC out-of-service case. The LERF results provide similar insights to the CDF results insights.

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.

Clinton SLC CT Extension 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.

Clinton SLC CT Extension Table 613 TOP TEN CUTSETS FOR CDF FOR THE SLC SYSTEM OUT OF SERVICE

Clinton SLC CT Extension Table B-4 BASIC EVENT IMPORTANCE MEASURES FOR CDF ASSESSMENT FOR SLC OUT OF SERVICE

Clinton SL C CT Extension Table 8-4 BASIC EVENT IMPORTANCE MEASURES FOR CDF ASSESSMENT FOR SLC OUT OF SERVICE

%FLOOD12C

%FLOOD09A 2.20E-05 1.23E-05 Flood in Area A.2.2 - WO Line Break

%FLOOD16 2.20E-05 1.23E-05 Flood in Area A.4.6 - Any Break

%FLOOD08C 1.86E-05 1.04E-05 Flood in Area A.3.3 - RClC Line Break

%FLOOD15A 1.50E-05 8.40E-06 Flood in Area A.4.1 - SX-B Line Break

%FLOOD07A 7.50E-06 4.20E-06 Flood in Area A.2.3 - WO Line Break

%FLOOD07B 7.50E-06 4.20E-06 Flood in Area A.2.3 - Other Line Break

%FLOOD03A 7.45E-06 4.17E-06 Flood in Area A.1.9 - RClC Line Break

%FLOOD04A 7.45E-06 4.17E-06 Flood in Area A. I .I 0 - RClC Line Break

%FLOOD05A 7.45E-06 4.17E-06 Flood in Area A.2.5 - WO Line Break

%FLOOD13A 7.45E-06 4.1 7E-06 Flood in Area A.l .I- WS Line Break

%FLOOD27A 7.45E-06 4.1 7E-06 Flood in Area H.l . I - WS Line Break

%FLOOD28A 7.45E-06 4.1 7E-06 Flood in Area H.1.2 - WS Line Break

%FLOOD29A 7.45E-06 4.17E-06 Flood in Area H.1.3 - WS Line Break 1CTSYLRGPCFLLR-- 2.00E-01 2.67E-06 CONT. CATASTROPHIC FAILURE MODE 1CVPH-SMALLD-F-- 1.00E+00 2.67E-06 SMALL DIA VENTS ASSESSED AS UNSUCCESSFUL (4411.06 PROC SECT 2.3 & 2.4)

IN CONTAINMENT MOVIAOV FAILS CLOSED DUE TO ENVIRONM. STRESS (LEVEL 1CVPH-TEMPF--F-- 1.00E-02 2.67E-06 1) 1--RX-SPC-VACH-- 5.00E-07 2.67E-06 DEP HEP: OP FAILURE OF SPC (LATE) AND VACUUM PUMPS

Clinton SLC CT Extension Table B-5 RISK ASSESSMENT SENSITIVITY RESULTS

Clinton SLC CT Extension PARAMETRIC UNCERTAINTY Consistent with the ASME PRA Standard [Ref. 8-21, quantitative parametric uncertainty analyses for both CDF and LERF have been performed using an EPRl method

[Ref. B-I] 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 2006C PRA Update.

The parametric uncertainty analyses are performed using Monte Carlo simulation. The analysis is performed using the EPRl R&R workstation UNCERT software.

B.3.1 Core Damage Frequencv Parametric Uncertaintv Distribution The resulting uncertainty distribution for the proposed CT configuration (i.e., CDFsLc-oos) calculated by UNCERT Version 2.3a for CDF is shown in Figure B-I. It summarizes:

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, as compared to the error factor of 2.0 for the CLO6C base 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 PRA model (i.e., 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 B-6 provides this comparison for the proposed CT model case (i.e., CDFsLc-oos):

Table B-6 PARAMETER UNCERTAINTY COMPARISON FOR CDF k

CDF CDFSLC-oos Parameter Result Code Point Estimate 9.1 E-61yr CAFTA Uncertainty Mean 9.1E-6Iyr UNCERT A E -

The propagated uncertainty mean for CDFsLc-oosis the same as the CDFsLc-oospoint estimate calculation. If the CDFsLc-oospropagated uncertainty mean instead of the CLO6C CDFBAsEpropagated uncertainty mean were used to calculate the risk metrics, the results would not differ from those presented in Table 5.3-1.

Clinton SLC CT Extension B.3.2 Lar~e Earlv Release Frequencv (LERF) 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 8-2. 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 proposed CT for the LERF uncertainty distribution is 2.0 (calculated using SQR(95%/5%)), as compared to the error factor of 3.1 for the CLO6C base model.

Table B-7 provides a comparison of the PRA LERF point estimate and the propagated uncertainty mean for the proposed CT case (i.e., LERFsLc-oos):

Table B-7 PARAMETER UNCERTAINTY COMPARISON FOR LERF Point Estimate If the LERFsLc-oospropagated uncertainty mean (8.8E-7Iyr) and the CLOGC LERFBnsE propagated uncertainty mean (1.2E-7lyr) are used to calculate the risk metrics, the results would change in the second decimal place compared to the results shown in Table 5.3-1 (i.e., non-significant change).

Clinton SL C CT Extension Figure B-1 CDF PARAMETRIC UNCERTAINTY DISTRIBUTION FOR THE PROPOSED COMPLETION TIME UNCERT 2.3a COREDAMAGE.CUT CL206C-UNCERT.BE Samples 50,000 Random Seed Auto Relative Frequency Mean - M : 9.09E-06 5%-[ :3.44E-06 50% - x : 6.99E-06 95% - ] : 2.06E-05 Std Dev : 9.43E-06 1.E-4 Frequency !Probability

Clinton SL C CT Extension Figure 8-2 LERF PARAMETRIC UNCERTAINTY DlSTRlBUTlON FOR THE PROPOSED COMPLETION TIME UNCERT 2.3a LERF-TOT.CUT CLOGC-UNCERT.BE Samples 50,000 Random Seed Auto Relative Frequency Mean - M : 8.76E -07 5% - [ : 7.78E-08 50%- x : 3.54E -07 95%-] :3.1OE-06 StdDev :2.61E-06 F-+-+-Frequency

+ i Probability

Clinton SLC CT Extension

6.5 REFERENCES

[B-I] Treatment of Parameter and Model Uncertainty for Probabilistic Risk Assessments, EPRl Report 1016737, Palo Alto, CA: 2008.

[B-21 ASMEIANS RA-Sa-2009, "Addenda to RA-S-2008, Standard for Level 1lLarge Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications," February 2009.

B-I 6 C46709002O-8935-12/28/2009

Clinton SLC CT Extension Appendix C B WROG Assessment of NRC lnformation Notice 2007-07 The BWROG assessment of NRC lnformation 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.

Clinton 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 Information 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 of the occurrence without consequence.

It is the position of the 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-113 should the fire impact the ability to scram. This manual operator action should be endorsed by the NRC for use in both 1II.G.1 and 2 areas, as well as, III.G.3 and 1II.L areas. The evaluation provided in this paper and the limited likelihood of occurrence of the 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 fbrther 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 fbrther 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 (Al, A2, B1, 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 of the logic channels:

Scram Discharge Volume Water Level

Clinton 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 (Al, A2, B1, 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 of the logic is whd 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 of the 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 of the scram circuitry. This portion of the 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 ~ i ~ c a n ofc the e 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 of the hot shorts postulated in the first section.

4.1) Circuit Analysis:

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

Clinton 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) fire-induced hot shorts would prevent one of four scram rod groups from inserting.

The BWROG identified that a single hot short in either of the divisionalized trip logics can prevent the scram of a single rod group. This finding is different than the conclusion in IN 2007-07. The finding of the BWROG assessment is a direct consequence of the 1 out of 2 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 1 attached for a description of the location of the subject hot short, labeled as "Hot Short I".] The hot short must occur prior to the operator scramming the reactor. The location of the hot short shown in Figure 1 would 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 of the 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 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 11.3minutes. 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 11: (Refer to Figure 2)

Case I1 is one of two cases identified where two (2) fire-induced hot shorts could prevent a full scram. (Note: No conditions were identified where two (2) frre-induced hot shorts were required to prevent a single rod group from scramming.)

Clinton 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 shortsjust 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 subsequentlykeep 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 of the 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 K 15 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 of the MSIVs or reactor level reaching the t13" level will override the effects of the hot shorts affectingthe K15 relays and result in a de-energization of the K14 relays and full rod insertion.

Therefore, it appears unlikely that the required hot shorts, even if they were to co-exist, could prevent the scram and cause the reactivity transient described in the IN. This is true because the effect of the 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 111: (Refer to Figure 3) (Limited to the Trip System Relay Panels)

Case I11 is similar to Case 11. Hot shorts are postulated in the locations shown in Figure 3, the K14 relays will again remain energized. The energization of the 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.]

Clinton SLC CT Extension BWROG Assessment of NRC Information Notice 2007-07 Therefore, it appears unlikely that the required hot shorts would co-exist given that the time required for fire damage to the individual cables and fire propagation between relay compartments to occur.

For all of the cases discussed above, regardless of the number of fire-induced hot shorts postulated, the required hot short configuration must occur prior to the operator scramming the unit. For those configurations requiring more than a single hot short, the two hot shorts must exist coincidentally.

The hot short configurations 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 hel.

Additionally, the scenario described in the IN represents a condition more severe than many BWRs would experience due to the availability of additional safe shutdown system capability. Many BWRs also have high pressure systems available for alternative shutdown at their remote shutdown panel. For a BWR with a high pressure system safe shutdown capability, the time available prior to the need to reduce pressure reactor pressure for injection with either a low pressure system or for shutdown cooling would be extended by a number of hours.

Finally, operators for all BWRs are trained on the use of the Emergency Operating Procedures. EO- 113 for each BWR provides clear direction to the to either remove RPS power or the vent the SCRAM air header to achieve a full scram.

3.2) Implications for Appendix R Compliance:

For all plants the main operating bench board is in the main control room. At some plants, the relay panels are located in the main control room. In other plants the relay panels are located in a relay room separate from the main control room. For these latter set of plants, some classlfy the relay room as III.G.3 areas and some classify the relay room as 1II.G.1 and 2 areas.

This issue, therefore, has implications for redundant safe shutdown under Appendix R Section 1II.G.1 and 2 and for alternative and dedicated safe shutdown under the requirements of Appendix R Section III.G.3 and 1II.L.

With respect to Case I, it is clear that none of the methods available under III.G.2 would be effective in preventing the condition. Protection of the subject circuits with a 3 hour3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months fire rated barrier, with a one hour fire rated barrier with automatic suppression and detection or by separation of 20 feet with automatic suppression and detection and no intervening combustibles, would not prevent the occurrence of this event. Additionally, even if the relay panels for each of the four channels are located in separate controVrelay room in separate fire areas, the condition could still occur and 3-hour f i e rated barriers

Clinton SL C CT Extension BWROG Assessment of NRC Information Notice 2007-07 for each of these postulated fire areas would be ineffective in preventing the occurrence of the 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 I1 and I11 are similar. Neither of these 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 of this important safety function. In addition to the features of the 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 of the 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 of the 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- 113 should the fire impact the ability to scram. This manual operator action should be endorsed by the NRC for use in both II1.G. 1 and III.G.2 areas, as well as, III.G.3 and II1.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 of this manual operator action.

Clinton 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 of this condition, the safety significance of this 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 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 of the occurrence without consequence.

It is the position of the 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-113 should the fire impact the ability to scram. This manual operator action should be endorsed by the NRC for use in both II1.G. 1 and 2 areas, as well as, III.G.3 and 1II.L areas. The evaluation provided in t h s paper and the limited likelihood of occurrence of the 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-fke safe shutdown procedures.

Prepared by: Thomas A. Gorman Date: 10/16/2007 Thomas A. Gorman, PE, SFPE Reviewed by: Gary Birmingham Date: 11/13/2007 Gary S. Birmingham

Clinton SL C CT Extension El\VHQ6; &sasmctnt of NRC bfsrmtltioa ?dotire 2007-07

Clinton SLC CT Extension BWROG Assessment of NRC Information Notice 2007-07 No%: Thb canbination assumear RP6 dws not pick up mtht want signal, e.g. MStV Closure; f;$pr* Turbine Stop Valve Closure.

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Cnmaat ," 4mb f3 HARbAL $CRAW

& W P L Yt" f-yF-m*---~- -

"$"nipo"eI"$c"tL ,'d&?M -

Fioure 2 - Manual Scram Circuitry Tvoical of two Top Svstern~

Clinton SL C CT Exdension B'BYROG Assessment of NRC Informi&ionNotice 2007-07 Refer to Figure 4 for the remaining set of contacts that affect the automatic scram function Hot Shod #3 %tireation Ib~icaf2 $OWTrio x?k%a&

Firrure 3 Reactor Auto-Scram Circuitrv Tvpicd of four Trip Is in two Trio Svstem

Clinton SLC CT Extension BWROG Assessment of NRC Information Notice 2007-07 ce of Auto-Scram Circultw (tvglcal of 4 T r l Channels)

~

v t e Letter Sequence Request
TAC:MD4044, (Open)
Initiation Request, Request Acceptance... Administration Questions Answers Results Approval... Other: ML070240595
MONTHYEAR2010-02-22 [Table View]

RS-10-028, Request for Amendment to Technical Specification 3.1.7, Standby Liquid Control (SLC) System.

Text

Subject:

References:

1.0 DESCRIPTION

2.0 PROPOSED CHANGE

3.0 BACKGROUND

4.0 TECHNICAL ANALYSIS

5.0 REGULATORY ANALYSIS

6.0 ENVIRONMENTAL CONSIDERATION

7.0 REFERENCES

1.0 DESCRIPTION

2.0 PROPOSED CHANGE

3.0 BACKGROUND

4.0 TECHNICAL ANALYSIS

5.0 REGULATORY ANALYSIS

6.0 ENVIRONMENTAL CONSIDERATION

7.0 REFERENCES

SUMMARY

1.0 INTRODUCTION

4.3 Fidelity

SUMMARY

6.0 REFERENCES

1.2 BACKGROUND

SUMMARY

SUMMARY

4.3 FIDELITY

SUMMARY

5.4 CONCLUSION

SUMMARY

SUMMARY

SUMMARY

6.5 REFERENCES

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RS-10-028, Request for Amendment to Technical Specification 3.1.7, Standby Liquid Control (SLC) System.

Text

Subject:

References:

1.0 DESCRIPTION

2.0 PROPOSED CHANGE

3.0 BACKGROUND

4.0 TECHNICAL ANALYSIS

5.0 REGULATORY ANALYSIS

6.0 ENVIRONMENTAL CONSIDERATION

7.0 REFERENCES

1.0 DESCRIPTION

2.0 PROPOSED CHANGE

3.0 BACKGROUND

4.0 TECHNICAL ANALYSIS

5.0 REGULATORY ANALYSIS

6.0 ENVIRONMENTAL CONSIDERATION

7.0 REFERENCES

SUMMARY

1.0 INTRODUCTION

4.3 Fidelity

SUMMARY

6.0 REFERENCES

1.2 BACKGROUND

SUMMARY

SUMMARY

4.3 FIDELITY

SUMMARY

5.4 CONCLUSION

SUMMARY

SUMMARY

SUMMARY

6.5 REFERENCES

Description:

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