Text

Issuance of Amendments Technical Specification Change to Add Limiting Condition of Operation 3.0.8 on the Inoperability of Snubbers Using the Consolidated Line Item Improvement Process
	ML13017A381
Person / Time
Site:	Robinson
Issue date:	02/26/2013
From:	Billoch-Colon A; Plant Licensing Branch II
To:	William Gideon; Carolina Power & Light Co
	Billoch-Colon A
References
	TAC ME9236
	Download: ML13017A381 (27)
	v • d • e

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 February 26, 2013 Mr. William R. Gideon, Vice President Carolina Power & Light Company H.B. Robinson Steam Electric Plant, Unit 2 3581 West Entrance Road Hartsville, SC 29550

SUBJECT:

H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO.2 - ISSUANCE OF AMENDMENT REGARDING THE TECHNICAL SPECIFICATION CHANGE TO ADD LIMITING CONDITION OF OPERATION 3.0.8 ON THE INOPERABILITY OF SNUBBERS USING THE CONSOLIDATED LINE ITEM IMPROVEMENT PROCESS (TAC NO. ME9236)

Dear Mr. Gideon:

The Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment No. 232 to Renewed Facility Operating License No. DPR-23 for the H. B. Robinson Steam Electric Plant, Unit No.2 (HBRSEP). This amendment changes the HBRSEP Technical Specifications (TSs) in response to your application dated August 6, 2012.

The amendment modifies the TSs requirements for inoperable snubbers by adding Limiting Condition for Operation (LCO) 3.0.8. The amendment also makes conforming changes to TS LCO 3.0.1 to reference TS LCO 3.0.8. The proposed change is based on the NRC approved Technical Specification Task Force (TSTF) standard TS change TSTF-372, Revision 4. A notice of availability for this TS improvement using the consolidated line item improvement process was published by the NRC staff in the Federal Register on October 16, 2012 (77 FR 63347).

A copy of the safety evaluation is also enclosed. Notice of Issuance will be included in the Commission's Biweekly Federal Register Notice.

Sincerely,

~r,~(A Araceli T. Billoch Col6n, Project Manager Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-261

Enclosures:

1. Amendment No. 232 to DPR-23

2. Safety Evaluation cc w/enclosures: Distribution via ListServ

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 CAROLINA POWER & LIGHT COMPANY DOCKET NO. 50-261 H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO.2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 232 Renewed License No. DPR-23

1. The Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by Carolina Power & Light Company (the licensee), dated August 6,2012, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

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2. Accordingly, the license is amended by changes to the Technical Specifications, as indicated in the attachment to this license amendment; and paragraph 3.B. of Renewed Facility Operating License No. DPR-23 is hereby amended to read as follows:

B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 232 are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of the date of its issuance and shall be implemented within 60 days.

FOR THE NUCLEAR REGULATORY COMMISSION sie F. Quichocho, Chief lant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to Operating License No. DPR-23 and the Technical Specifications Date of Issuance: February 26, 2013

ATTACHMENT TO LICENSE AMENDMENT NO. 232 RENEWED FACILITY OPERATING LICENSE NO. DPR-23 DOCKET NO. 50-261 Replace the following pages of the Renewed Facility Operating License and Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove License DPR-23 License DPR-23 Page 3 Page 3 TSs TSs 3.0-1 3.0-1 3.0-3 3.0-3

- 3 neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; D. Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required any byproduct, source, or special nuclear material without restriction to chemical or physical form for sample analysis or instrument and equipment calibration or associated with radioactive apparatus or components; E. Pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by operation of the facility.

3. This renewed license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations: 10 CFR Part 20, Section 30.34 of 10 CFR Part 30, Section 40.41 of 10 CFR Part 40, Section 50.54 and 50.59 of 10 CFR Part 50. and Section 70.32 of 10 CFR Part 70; and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

A. Maximum Power Level The licensee is authorized to operate the facility at a steady state reactor core power level not in excess of 2339 megawatts thermal.

B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 232 are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications.

(1) For Surveillance Requirements (SRs) that are new in Amendment 176 to Final Operating License DPR-23, the first performance is due at the end of the first surveillance interval that begins at implementation of Amendment 176. For SRs that existed prior to Amendment 176, including SRs with modified acceptance criteria and SRs whose frequency of performance is being extended, the first performance is due at the end of the first surveillance interval that begins on the date the Surveillance was last performed prior to implementation of Amendment 176.

Renewed Facility Operating License No. DPR-23 Amendment No. 232

LCD Applicability 3.0 3.0 LIMITING CONDITION FOR OPERATION (LCD) APPLICABILITY LCD 3.0.1 LCOs shall be met during the MODES or other specified conditions in the Applicability, except as provided in LCD 3.0.2,3.0.7, and 3.0.8.

LCD 3.0.2 Upon discovery of a failure to meet an LCD, the Required Actions of the associated Conditions shall be met, except as provided in LCD 3.0.5 and LCD 3.0.6.

If the LCD is met or is no longer applicable prior to expiration of the specified Completion Time(s), completion of the Required Action(s) is not required unless otherwise stated.

LCO 3.0.3 When an LCD is not met and the associated ACTIONS are not met, an associated ACTION is not provided, or if directed by the associated ACTIONS, the unit shall be placed in a MODE or other specified condition in which the LCD is not applicable. Action shall be initiated within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to place the unit, as applicable, in:

a. MODE 3 within 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months ;

b. MODE 4 within 13 hours1.50463e-4 days 0.00361 hours 2.149471e-5 weeks 4.9465e-6 months ; and

c. MODE 5 within 37 hours4.282407e-4 days 0.0103 hours 6.117725e-5 weeks 1.40785e-5 months .

Exceptions to this Specification are stated in the individual Specifications.

Where corrective measures are completed that permit operation in accordance with the LCO or ACTIONS. completion of the actions required by LCD 3.0.3 is not required.

LCO 3.0.3 is only applicable in MODES 1.2.3. and 4.

LCO 3.0.4 When an LCO is not met, entry into a MODE or other specified condition in the Applicability shall only be made:

(continued)

HBRSEP Unit No.2 3.0-1 Amendment No. 232

LCO Applicability 3.0 3.0 LCO APPLICABILITY LCO 3.0.6 the loss of safety function exists are required to be entered.

(continued)

When a support system's Required Action directs a supported system to be declared inoperable or directs entry into Conditions and Required Actions for a supported system, the applicable Conditions and Required Actions shall be entered in accordance with LCO 3.0.2.

LC03.0.7 Test Exception LCO 3.1.8 allows specified Technical Specification (TS) requirements to be changed to permit performance of special tests and operations. Unless otherwise specified, all other TS requirements remain unchanged. Compliance with Test Exception LCOs is optional. When a Test Exception LCO is desired to be met but is not met, the ACTIONS of the Test Exception LCO shall be met. When a Test Exception LCO is not desired to be met, entry into a MODE or other specified condition in the Applicability shall be made in accordance with the other applicable Specifications.

LCO 3.0.8 When one or more required snubbers are unable to perform their associated support function(s), any affected supported LCO(s) are not required to be declared not met solely for this reason if risk is assessed and managed, and:

a. the snubbers not able to perform their associated support function(s) are associated with only one train or subsystem of a multiple train or subsystem supported system or are associated with a single train or subsystem supported system and are able to perform their associated support function within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; or

b. the snubbers not able to perform their associated support function(s) are associated with more than one train or subsystem of a multiple train or subsystem supported system and are able to perform their associated support function within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

At the end of the specified period the required snubbers must be able to perform their associated support function(s). or the affected supported system LCO(s) shall be declared not met.

HBRSEP Unit No.2 3.0-3 Amendment No. 232

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFfiCE Of NUQ.bEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 232 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-23 CAROLINA POWER & LIGHT COIVIPANY H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO.2 DOCKET NO. 50-261

1.0 INTRODUCTION

By letter to the U.S. Nuclear Regulatory Commission (NRC, the Commission) dated August 6, 2012 (Agencywide Documents Access and Management System Accession Number ML12229A047), Carolina Power & Light Company (the licensee), dOing business as Progress Energy Carolinas, Inc., submitted a license amendment request for changes to the H. B. Robinson Steam Electric Plant, Unit No.2 (HBRSEP), Technical Specifications (TSs).

The proposed change would revise the TSs to add Limiting Condition for Operation (LCO) 3.0.8 to address conditions where one or more snubbers are unable to perform their associated support function. A conforming change would also be made to TS LCO 3.0.1 to reference TS LCO 3.0.8. The proposed changes are based on Technical Specification Task Force (TSTF) change TSTF-372, Revision 4, which has been approved generically for the Standard Technical Specifications (STSs; NUREGs-1430 - 1434) by the NRC. The NRC staff published a notice of availability of this TS change in the Federal Register on May 4,2005 (70 FR 233252) as part of the Consolidated Line Item Improvement Process (CLlIP). The notice included a model safety evaluation (SE) that may be referenced by licensees in plant-specific applications to adopt the TSTF-372 changes. In its application, the licensee stated that the justifications presented in the model SE for TSTF-372 are applicable to HBRSEP and justify the proposed TS changes. The SE that follows is based on the model SE.

TSTF-372, Revision 4, is an improvement to the STSs that allows licensee's, through a license amendment, to add an LCO allowing a delay time for entering a supported system TS, when the inoperability is due solely to an inoperable snubber, if risk is assessed and managed. The postulated seismic event requiring snubbers is a low-probability occurrence, and the overall TS system safety function would still be available for the vast majority of anticipated challenges.

TSTF-372 was approved under the risk-informed TSs program. These initiatives are intended to maintain or improve safety through the incorporation of risk assessment and management techniques in the TSs, while reducing unnecessary burden and making TS requirements Enclosure 2

- 2 consistent with the Commission's other risk-informed regulatory requirements, in particular the Maintenance Rule.

In accordance with the approved TSTF-372, the proposed change adds LCO 3.0.8 to the licensee's TSs. LCO 3.0.8 allows licensees to delay declaring an LCO not met for equipment that is supported by snubbers unable to perform their associated support functions when the risk associated with the delay is assessed and managed. This new LCO 3.0.8 states:

When one or more required snubbers are unable to perform their associated support function(s), any affected supported LCO(s) are not required to be declared not met solely for this reason if risk is assessed and managed, and:

a. the snubbers not able to perform their associated support function(s) are associated with only one train or subsystem of a multiple train or subsystem supported system or are associated with a single train or subsystem supported system and are able to perform their associated support function within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; or

b. the snubbers not able to perform their associated support function(s) are associated with more than one train or subsystem of a multiple train or subsystem supported system and are able to perform their associated support function within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

At the end of the specified period the required snubbers must be able to perform their associated support function(s), or the affected supported system LCO(s) shall be declared not met.

Consistent with TSTF-372, a conforming change would also be made to LCO 3.0.1 to reference the new LCO 3.0.8. HBRSEP TS LCO 3.01 currently reads as follows:

LCOs shall be met during the MODES or other specified conditions in the Applicability, except as provided in LCO 3.0.2, and LCO 3.0.7.

LCO 3.0.1 would be revised to read as follows:

LCOs shall be met during the MODES or other specified conditions in the Applicability, except as provided in LCO 3.0.2,3.0.7, and 3.0.8.

2.0 REGULATORY EVALUATION

In Section 50.36 of Title 10 of the Code of Federal Regulations (10 CFR), the Commission established its regulatory requirements related to the content of the TSs. Pursuant to 10 CFR 50.36, TSs are required to include items in the following five specific categories:

(1) safety limits, limiting safety system settings, and limiting control settings; (2) LCOs; (3) surveillance requirements (SRs); (4) design features; and (5) administrative controls. The rule does not specify the particular requirements to be included in a plant's TSs. As stated in 10 CFR 50.36(c)(2)(i), "[I]imiting conditions for operation are the lowest functional capability or performance levels of equipment required for safe operation of the facility. When a limiting

- 3 condition for operation of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the technical specifications ... " HBRSEP TS Section 3.0, on "LCO and SR Applicability," provides details or general application rules for complying with the LCOs.

Snubbers are chosen in lieu of rigid supports in areas where restricting thermal growth during normal operation would induce excessive stresses in the piping nozzles or other equipment.

Although they are classified as component standard supports, they are not designed to provide any transmission of force during normal plant operations. However, in the presence of dynamic transient loadings, which are induced by seismic events as well as by plant accidents and transients, a snubber functions as a rigid support. The location and size of the snubbers are determined by stress analyses based on different combinations of load conditions, depending on the design classification of the particular piping.

Prior to the conversion to the improved STSs, TS requirements applied directly to snubbers.

These requirements included:

A requirement that snubbers be operable and in service when the supported equipment is required to be operable;

A requirement that snubber removal for testing be done only during plant shutdown;

A requirement that snubber removal for testing be done on a one-at-a-time basis when supported equipment is required to be operable during shutdown;

A requirement to repair or replace within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months any snubbers, found to be inoperable during operation in Modes 1 through 4, to avoid declaring any supported equipment inoperable;

A requirement that each snubber be demonstrated operable by periodic visual inspections; and

A requirement to perform operability tests on a representative sample of at least 10 percent of plant snubbers, at least once every 18 months during shutdown.

In the late 1980s, a joint initiative of the NRC and industry was undertaken to improve the STSs.

This effort identified the snubbers as candidates for relocation to a licensee-controlled document, based on the fact that the TS requirements for snubbers did not meet any of the four criteria in 10 CFR 50.36(c)(2)(ii) for inclusion in the improved STSs. The NRC approved the relocation without placing any restriction on the use of the relocated requirements. However, this relocation resulted in different interpretations between the NRC and the industry regarding its implementation.

The NRC has stated that since snubbers are supporting safety equipment that is in the TSs, the definition of OPERABILITY must be used to immediately evaluate equipment supported by a removed snubber and, if found inoperable, the appropriate TS required actions must be entered.

This interpretation has, in practice, eliminated the 72-hour delay to enter the actions for the

- 4 supported equipment that existed prior to the conversion to the improved STSs (the only exception is if the supported system has been analyzed and determined to be OPERABLE without the snubber). The industry has argued that since the NRC approved the relocation without placing any restriction on the use of the relocated requirements, the licensee-controlled document requirements for snubbers should be invoked before the supported system's TS requirements become applicable. The industry's interpretation WOUld, in effect, restore the 72-hour delay to enter the actions for the supported equipment that existed prior to the conversion to the improved STS. The industry's proposal would allow a time delay for all conditions, including snubber removal for testing at power.

The option to relocate the snubbers to a licensee-controlled document, as part of the conversion to improved STSs, has resulted in non-uniform and inconsistent treatment of snubbers. On the one hand, plants that have relocated snubbers from their TSs to licensee-controlled documents such as Technical Requirements Manual, etc., are allowed to change the requirements for snubbers under the auspices of 10 CFR 50.59, provided the requirements of 10 CFR 50.55a continue to be met, but they are not allowed a 72-hour delay before they enter the actions for the supported equipment. On the other hand, plants that have not converted to improved STSs have retained the 72-hour delay if snubbers are found to be inoperable, but they are only allowed to change TS requirements for snubbers by a license amendment request within the 10 CFR 50.55a requirements. It should also be noted that a few plants that converted to the improved STSs chose not to relocate the snubbers to a licensee-controlled document and, thus, retained the 72-hour delay. In addition, it is important to note that, unlike plants that have not relocated snubbers, plants that have relocated snubbers can perform functional tests on the snubbers at power (as long as they enter the actions for the supported equipment). Some potential undesirable consequences of this inconsistent treatment of snubbers are:

Performance of testing during crowded time period windows when the supported system is inoperable with the potential to reduce the snubber testing to a minimum since the snubber requirements relocated from TSs are controlled by the licensee;

Performance of testing during crowded windows when the supported system is inoperable with the potential to increase the unavailability of safety systems; and

Performance of testing and maintenance on snubbers affecting multiple trains of the same supported system during the 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months allotted before entering MODE 3 under LCO 3.0.3.

To remove the inconsistency in the treatment of snubbers among plants, TSTF-372 was approved by the NRC to allow a risk-informed TS change that introduces a delay time before entering the actions for the supported equipment, when one or more snubbers are found inoperable or removed for testing, if risk is assessed and managed. Such a delay time will provide needed flexibility in the performance of maintenance and testing during power operation and at the same time will enhance overall plant safety by:

Avoiding unnecessary unscheduled plant shutdowns and, thus, minimizing plant transition and realignment risks;

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Avoiding reduced snubber testing and, thus, increasing the availability of snubbers to perform their supporting function;

Performing most of the required testing and maintenance during the delay time when the supported system is available to mitigate most challenges and, thus, avoiding increases in safety system unavailability; and

Providing explicit risk-informed guidance in areas in which that guidance currently does not exist, such as the treatment of snubbers impacting more than one redundant train of a supported system.

3.0 TECHNICAL EVALUATION

1 The industry submitted TSTF-372, Revision 4, "Addition of LCO 3.0.B, Inoperabilityof Snubbers," in support of the proposed TS change. This submittal (Reference 1) documents a risk-informed analysis of the proposed TS change. Probabilistic risk assessment (PRA) results and inSights are used, in combination with deterministic and defense-in-depth arguments, to identify and justify delay times for entering the actions for the supported equipment associated with inoperable snubbers at nuclear power plants. This is in accordance with guidance provided in Regulatory Guides (RGs) 1.174 and 1.177 (References 2 and 3, respectively).

The risk impact associated with the proposed delay times for entering the TS actions for the supported equipment can be assessed using the same approach as for allowed completion time (CT) extensions. Therefore, the risk assessment was performed following the three-tiered approach recommended in RG 1.177 for evaluating proposed extensions in currently allowed CTs:

The first tier involves the assessment of the change in plant risk due to the proposed TS change. Such risk change is expressed (1) by the change in the average yearly core damage frequency (ACDF) and the average yearly large early release frequency (ALERF) and (2) by the incremental conditional core damage probability (ICCDP) and the incremental conditional large early release probability (ICLERP). The assessed ACDF and ALERF values are compared to acceptance guidelines, consistent with the Commission's Safety Goal Policy Statement as documented in RG 1.174, so that the plant's average baseline risk is maintained within a minimal range. The assessed ICCDP and ICLERP values are compared to acceptance guidelines provided in RG 1.177, which aim at ensuring that the plant risk does not increase unacceptably during the period the equipment is taken out of service.

The second tier involves the identification of potentially high-risk configurations that could exist if equipment in addition to that associated with the change were to be taken out of The following technical evaluation is presented in terms of the bounding assessment of this change for all commercial pressurized-water reactors (PWRs) nuclear power plants, including HBRSEP, performed as part of the approval of TSTF-372, Revision 4, and publication of the CLlIP notices.

- 6 service simultaneously, or other risk-significant operational factors such as concurrent equipment testing were also involved. The objective is to ensure that appropriate restrictions are in place to avoid any potential high-risk configurations.

The third tier involves the establishment of an overall configuration risk management program (CRMP) to ensure that potentially risk-significant configurations resulting from maintenance and other operational activities are identified. The objective of the CRMP is to manage configuration-specific risk by appropriate scheduling of plant activities and/or appropriate compensatory measures.

A simplified bounding risk assessment, which is also applicable to the license amendment request for HBRSEP, was performed to justify the proposed addition of LCO 3.0.8 to the TSs.

This approach was necessitated by (1) the general nature of the proposed TS changes (i.e.,

they apply to all plants and are associated with an undetermined number of snubbers that are not able to perform their function), (2) the lack of detailed engineering analyses that establish the relationship between earthquake level and supported system pipe failure probability when one or more snubbers are inoperable, and (3) the lack of seismic risk assessment models for most plants. The simplified risk assessment is based on the following major assumptions, which the NRC staff finds acceptable, as discussed below:

The accident sequences contributing to the risk increase associated with the proposed TS changes are assumed to be initiated by a seismically-induced loss-of-offsite power (LOOP) event with concurrent loss of all safety system trains supported by the out-of-service snubbers. In the case of snubbers associated with more than one train (or subsystem) of the same system, it is assumed that all affected trains (or subsystems) of the supported system are failed. This assumption was introduced to allow the performance of a simple bounding risk assessment approach with application to all plants. This approach was selected due to the lack of detailed plant-specific seismic risk assessments for most plants and the lack of fragility data for piping when one or more supporting snubbers are inoperable.

The LOOP event is assumed to occur due to the seismically-induced failure of the ceramic insulators used in the power distribution systems. These ceramic insulators have a high confidence (95 percent) of low probability (5 percent) of failure (HCLPF) of about 0.1 g, expressed in terms of peak ground acceleration. Thus, a magnitude 0.1 g earthquake is conservatively assumed to have 5 percent probability of causing a LOOP initiating event.

The fact that no LOOP events caused by higher magnitude earthquakes were considered is justified because (1) the frequency of earthquakes decreases with increasing magnitude and (2) historical data (References 4 and 5) indicate that the mean seismic capacity of ceramic insulators (used in seismic PRAs) , in terms of peak ground acceleration, is about 0.3g, which is significantly higher than the 0.1g HCLPF value. Therefore, the simplified analysis, even though it does not consider LOOP events caused by earthquakes of a magnitude higher than 0.1 g, bounds a detailed analysis that would use mean seismic failure probabilities (fragilities) for the ceramic insulators.

Analytical and experimental results obtained in the mid-1980s as part of the industry's "Snubber Reduction Program" (References 4 and 6) indicated that piping systems have

- 7 large margins against seismic stress. The assumption that a magnitude 0.1 g earthquake would cause the failure of all safety system trains supported by the out-of-service snubbers is very conservative, because safety piping systems could withstand much higher seismic stresses even when one or more supporting snubbers are out of service. The actual piping failure probability is a function of the stress allowable and the number of snubbers removed for maintenance or testing. Since the licensee-controlled testing is done on only a small (about 10 percent) representative sample of the total snubber population, typically only a few snubbers supporting a given safety system are out for testing at a time. Furthermore, since the testing of snubbers is a planned activity, licensees have flexibility in selecting a sample set of snubbers for testing from a much larger population by conducting configuration-specific engineering and/or risk assessments. Such a selection of snubbers for testing provides confidence that the supported systems would perform their functions in the presence of a design-basis earthquake and other dynamic loads and, in any case, the risk impact of the activity will remain within the limits of acceptability defined in risk informed RGs 1.174 and 1.177.

The analysis assumes that one train (or subsystem) of all safety systems is unavailable during snubber testing or maintenance (an entire system is assumed unavailable if a removed snubber is associated with both trains of a two-train system). This is a very conservative assumption for the case of corrective maintenance, since it is unlikely that a visual inspection will reveal that one or more snubbers across all supported systems are inoperable. This assumption is also conservative for the case of the licensee-controlled testing of snubbers, since such testing is performed only on a small representative sample.

In general, no credit is taken for recovery actions and alternative means of performing a function, such as the function performed by a system assumed failed (e.g., when LCO 3.0.B.b applies). However, most plants have reliable alternative means of performing certain critical functions. For example, feed and bleed (F&B) can be used to remove heat in most PWRs such as at the HBRSEP's facility, when auxiliary feedwater (AFW), the most important system in mitigating LOOP accidents, is unavailable. A 10 percent failure probability for recovery actions to provide core cooling using alternative means is assumed for Diablo Canyon, the only West Coast PWR plant with F&B capability, when a snubber impacting more than one train of the AFW system (I.e., when LCO 3.0.B.b is applicable) is out of service. This failure probability value is significantly higher than the value of 2.2E-2 used in Diablo Canyon's PRA. Furthermore, Diablo Canyon has analyzed the impact of a single limiting snubber failure, and concluded that no single snubber failure would impact two trains of AFW. No credit for recovery actions to provide core cooling using alternative means is necessary for West Coast PWR plants with no F&B capability, because it has been determined that there is no single snubber whose non-functionality would disable two trains of AFW in a seismic event of magnitude up to the plant's safe shutdown earthquake (SSE). It should be noted that a similar credit could have been applied to most Central and Eastern U.S. plants, such as HBRSEP's facility, but this was not necessary to demonstrate the low-risk impact of the proposed TS change due to the lower earthquake frequencies at Central and Eastern U.S. plants as compared to West Coast plants.

The earthquake frequency at the 0.1g level was assumed to be 1E-3/year for Central and Eastern U.S. plants and 1E-1/year for West Coast plants. Each of these two values

- 8 envelop the range of earthquake frequency values at the 0.1g level, for Eastern U.S. and West Coast sites, respectively (References 5 and 7).

The risk impact associated with non-LOOP accident sequences (e.g., seismically initiated loss-of-coolant accident (LOCA) or anticipated transient without scram sequences) was not assessed. However, this risk impact is small compared to the risk impact associated with the LOOP accident sequences modeled in the simplified bounding risk assessment. Non LOOP accident sequences, due to the ruggedness of nuclear power plant designs, require seismically-induced failures that occur at earthquake levels above 0.3g. Thus, the frequency of earthquakes initiating non-LOOP accident sequences is much smaller than the frequency of seismically-initiated LOOP events. Furthermore, because of the conservative assumption made for LOOP sequences that a 0.1g level earthquake would fail all piping associated with inoperable snubbers, non-LOOP sequences would not include any more failures associated with inoperable snubbers than would LOOP sequences.

Therefore, the risk impact of inoperable snubbers associated with non-LOOP accident sequences is small compared to the risk impact associated with the LOOP accident sequences modeled in the simplified bounding risk assessment.

The risk impact of dynamic loadings other than seismic loads is not assessed. These shock-type loads include thrust loads, blowdown loads, waterhammer loads, steamhammer loads, LOCA loads, and pipe rupture loads. However, there are some important distinctions between nonseismic (shock-type) loads and seismic loads which indicate that, in general, the risk impact of the out-of-service snubbers is smaller for nonseismic loads than for seismic loads. First, while a seismic load affects the entire plant, the impact of a nonseismic load is localized to a certain system or area of the plant. Second, although nonseismic shock loads may be higher in total force and the impact could be as much or more than seismic loads, generally they are of much shorter duration than seismic loads.

Third, the impact of nonseismic loads is more plant specific, and, thus, is harder to analyze generically than is the impact of seismic loads. For these reasons, licensees will be required to confirm, every time LCO 3.0.8.a is used, that at least one train of each system that is supported by the inoperable snubber(s) would remain capable of performing the system's required safety or support functions for postulated design loads other than seismic loads.

3.1 Risk Assessment Results and Insights The results and insights from the implementation of the three-tiered approach of RG 1.177 to support the proposed addition of LCO 3.0.8 to the TSs are summarized and evaluated in Sections 3.1.1 to 3.1.3 below.

3.1.1 Risk Impact The bounding risk assessment approach, discussed in Section 3.0, was implemented generically for all U.S. operating nuclear power plants. Risk assessments were performed for two categories of plants, Central and East Coast plants and West Coast plants, based on historical seismic hazard curves (earthquake frequencies and associated magnitudes). The first

- 9 category, Central and East Coast plants, includes the vast majority of the U.S. nuclear power plant population (Reference 7). For each category of plants, two risk assessments were performed:

The first risk assessment applies to cases where all inoperable snubbers are associated with only one train (or subsystem) of the impacted safety systems. It was conservatively assumed that a single train (or subsystem) of each safety system is unavailable. It was also assumed that the probability of non-mitigation using the unaffected redundant trains (or subsystems) is 2 percent. This is a conservative value, given that for core damage to occur under those conditions, two or more failures are required.

The second risk assessment applies to the case where one or more of the inoperable snubbers are associated with multiple trains (or subsystems) of the same safety systems. It was assumed in this bounding analysis, except for West Coast PWR plants, that all safety systems are unavailable to mitigate the accident. Credit for using F&B to provide core cooling is taken for plants having F&B capability (e.g., Diablo Canyon) when a snubber impacting more than one train of the AFW system is inoperable. Credit for one AFW train to provide core cooling is taken for West Coast PWR plants with no F&B capability (e.g., San Onofre), because it has been determined that there is no single snubber whose non-functionality would disable two trains of AFW in a seismic event of a magnitude up to the plant's SSE.

The results of the performed risk assessments, in terms of core damage and large early release risk impacts, are summarized in Table 1. The first row lists the conditional risk increase, in terms of CDF (core damage frequency), ~RCDF, caused by the out-of-service snubbers (as assumed in the bounding analysis). The second and third rows list the ICCDP and the ICLERP (incremental conditional large early release probability) values, respectively. For the case where all inoperable snubbers are associated with only one train (or subsystem) of the supported safety systems, the ICCDP was obtained by multiplying the corresponding ~RCDF value by the time fraction of the proposed 12-hour delay to enter the actions for the supported equipment. For the case where one or more of the inoperable snubbers are associated with multiple trains (or subsystems) of the same safety system, the ICCDP was obtained by multiplying the corresponding ~RCDF value by the time fraction of the proposed 12-hour delay to enter the actions for the supported equipment. The ICLERP values were obtained by multiplying the corresponding ICCDP values by 0.1 (Le., by assuming that the ICLERP value is an order of magnitude less than the ICCDP). This assumption is conservative, because containment bypass scenarios, such as steam generator tube rupture accidents and interfacing system LOCAs, would not be uniquely affected by the out-of-service snubbers. Finally, the fourth and fifth rows list the assessed ~CDF and ~LERF values, respectively. These values were obtained by dividing the corresponding ICCDP and ICLERP values by 1.5 (Le., by assuming that the snubbers are tested every 18 months, as was the case before the snubbers were relocated to a licensee-controlled document). This assumption is reasonable because (1) it is not expected that licensees would test the snubbers more often than what used to be required by the TSs, and (2) testing of snubbers is associated with higher risk impact than is the average corrective maintenance of snubbers found inoperable by visual inspection (testing is expected to involve significantly more snubbers out of service than corrective maintenance).

The assessed ~CDF and ~LERF values are compared to acceptance guidelines, consistent with the Commission's Safety Goal Policy Statement as documented in RG 1.174, so that the

- 10 plant's average baseline risk is maintained within a minimal range. This comparison indicates that the addition of LCO 3.0.8 to the existing TSs would have an insignificant risk impact.

Table 1:Bounding Risk Assessment Results for Snubbers Impacting a Single Train and Multiple Trains of a Supported System Central and East Coast Plants West Coast Plants Single Train Multiple Train Single Train Multiple Train LlRcDF/yr 1E-6 SE-6 1E-4 SE-4 ICCDP 8E-9 7E-9 8E-7 7E-7 ICLERP 8E-10 7E-10 8E-8 7E-8 LlCDF/yr SE-9 SE-9 SE-7 SE-7 LlLERF/yr SE-10 SE-1O SE-8 SE-8 The assessed LlCDF and LlLERF values meet the acceptance criteria of 1E-6/year and 1E-7/year, respectively, based on guidance provided in RG 1.174. This conclusion is true without taking any credit for the removal of potential undesirable consequences associated with the current inconsistent treatment of snubbers (e.g., reduced snubber testing frequency, increased safety system unavailability, and treatment of snubbers impacting multiple trains) discussed in Section 2.0 above, and given the bounding nature of the risk assessment.

The assessed ICCDP and ICLERP values are compared to acceptance guidelines provided in RG 1.177, which aim at ensuring that the plant risk does not increase unacceptably during the period the equipment is taken out of service. This comparison indicates that the addition of LCO 3.0.8 to the existing TSs meets the RG 1.177 numerical guidelines of SE-7 for ICCDP and SE-8 for ICLERP. The small deviations shown for West Coast plants are acceptable because of the bounding nature of the risk assessments, as discussed in Section 3.0.

The risk assessment results of Table 1 are also compared to guidance provided in the revised Section 11 of NUMARC 93-01, Revision 2 (Reference 8), endorsed by RG 1.182 (Reference 9),

for implementing the requirements of paragraph (a)(4) ofthe Maintenance Rule, 10 CFR SO.6S.

Such guidance is summarized in Table 2. Guidance regarding the acceptability of conditional risk increase in terms of CDF (i.e., LlRcDF) for a planned configuration is provided. This guidance states that a specific configuration that is associated with a CDF higher than 1E-3/year should not be entered voluntarily. Since the assessed conditional risk increase, LlRcDF , is significantly less than 1E-3/year, plant configurations including out-of-service snubbers and

- 11 other equipment may be entered voluntarily if supported by the results of the risk assessment required by 10 CFR 50.65(a)(4), by LCO 3.0.8, or by other TSs.

Table 2: Guidance for Implementing 10 CFR 50.65(a)(4)

LlRcDF Guidance Greater than 1E-3/year Configuration should not normally be entered voluntarily.

ICCDP Guidance ICLERP Greater than 1E-5 Configuration should not normally be entered Greater than voluntarily 1E-6 Assess non-quantifiable factors; Establish risk management actions 1E-7 to 1E-6 1E-6 to 1E-5 Less than 1E-6 Normal work controls Less than 1E-7 Guidance regarding the acceptability of ICCDP and ICLERP values for a specific planned configuration and the establishment of risk management actions is also provided in NUMARC 93-01. This guidance, as shown in Table 2, states that a specific-plant configuration that is associated with ICCDP and ICLERP values below 1E-6 and 1E-7, respectively, is considered to require "normal work controls." Table 1 shows that for the majority of plants (Le.,

for all plants in the Central and East Coast category) the conservatively assessed ICCDP and ICLERP values are over an order of magnitude less than what is recommended as the threshold for the "normal work controls" region. For West Coast plants, the conservatively assessed ICCDP and ICLERP values are still within the "normal work controls" region. Thus, the risk contribution from out-of-service snubbers is within the normal range of maintenance activities carried out at a plant. Therefore, plant configurations involving out-of-service snubbers and other equipment may be entered voluntarily if supported by the results of the risk assessment required by 10 CFR 50.65(a)(4), by LCO 3.0.8, or by other TSs. However, this simplified bounding analysis indicates that, for West Coast plants, the provisions of LCO 3.0.8 must be used cautiously and in conjunction with appropriate management actions, especially when equipment other than snubbers is also inoperable, based on the results of configuration-specific risk assessments required by 10 CFR 50.65(a)(4), by LCO 3.0.8, or by other TSs.

In its approval of TSTF-372, the NRC staff found that the risk assessment results supported the proposed addition of LCO 3.0.8 to the TSs. The risk increases associated with this TS change when adopted by licensees will be insignificant (based on guidance provided in RGs 1.174 and 1.177) and within the range of risks associated with normal maintenance activities. In addition, LCO 3.0.8 will remove potential undesirable consequences stemming from the current

- 12 inconsistent treatment of snubbers in the TSs, such as reduced frequency of snubber testing, increased safety system unavailability, and the treatment of snubbers impacting multiple trains.

3.1.2 Identification of High-Risk Configurations The second tier of the three-tiered approach recommended in RG 1.177 involves the identification of potentially high-risk configurations that could exist if equipment, in addition to that associated with the TS change, were to be taken out of service simultaneously. Insights from the risk assessments, in conjunction with important assumptions made in the analysis and defense-in-depth considerations, were used to identify such configurations. To avoid these potentially high-risk configurations, specific restrictions to the implementation of the proposed TS changes were identified.

For cases where all inoperable snubbers are associated with only one train (or subsystem) of the impacted systems (i.e., when LCO 3.0.B.a applies), it was assumed in the analysiS that there will be unaffected redundant trains (or subsystems) available to mitigate the seismically initiated LOOP accident sequences. This assumption implies that there will be at least one success path available when LCO 3.0.B.a applies. Therefore, potentially high-risk configurations can be avoided by ensuring that such a success path exists when LCO 3.0.B.a applies. Based on a review of the accident sequences that contribute to the risk increase associated with LCO 3.0.B.a, as modeled by the simplified bounding analysis (i.e., accident sequences initiated by a seismically-induced LOOP event with concurrent loss of all safety system trains supported by the out-of-service snubbers), the following restrictions was identified to prevent potentially high-risk configurations:

For PWR plants, at least one AFW train (including a minimum set of supporting equipment required for its successful operation) not associated with the inoperable snubber(s), must be available when LCO 3.0.B.a is used.

For cases where one or more of the inoperable snubbers are associated with multiple trains (or subsystems) of the same safety system (i.e., when LCO 3.0.B.b applies), it was assumed in the bounding analysis (except for West Coast plants) that all safety systems are unavailable to mitigate the accident. Credit for using F&B to provide core cooling is taken for plants having F&B capability (e.g., Diablo Canyon) when a snubber impacting more than one train of the AFW system is inoperable. Credit for one AFW train to provide core cooling is taken for West Coast PWR plants with no F&B capability (e.g., San Onofre) because it has been determined that there is no Single snubber whose nonfunctionality would disable more than one train of AFW in a seismic event of magnitude up to the plant's SSE. Based on a review of the accident sequences that contribute to the risk increase associated with LCO 3.0.B.b (as modeled by the simplified bounding analysis) and on defense-in-depth considerations, the following restrictions were identified to prevent potentially high-risk configurations:

LCO 3.0.B.b cannot be used at West Coast PWR plants with no F&B capability when a snubber whose nonfunctionality would disable more than one train of AFW in a seismic event of magnitude up to the plant's SSE is inoperable (it should be noted, however, that based on information provided by the industry, there is no plant that falls in this category);

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When LCO 3.0.8.b is used at PWR plants, at least one AFW train (including a minimum set of supporting equipment required for its successful operation) not associated with the inoperable snubber(s), or some alternative means of core cooling (e.g., F&B, firewater system or "aggressive secondary cooldown" using the steam generators) must be available.

3.1.3 Configuration Risk Management The third tier of the three-tiered approach recommended in RG 1.177 involves the establishment of an overall CRMP to ensure that potentially risk-significant configurations resulting from maintenance and other operational activities are identified. The objective of the CRMP is to manage configuration-specific risk by appropriate scheduling of plant activities and/or appropriate compensatory measures. This objective is met by licensee programs to comply with the requirements of paragraph (a)(4) of the Maintenance Rule (10 CFR SO.6S) to assess and manage risk resulting from maintenance activities, and by the TSs requiring risk assessments and management using (a)(4) processes if no maintenance is in progress. These programs can support licensee decision-making regarding the appropriate actions to manage risk whenever a risk-informed TS is entered. Since the 10 CFR SO.6S(a)(4) guidance (i.e., Reference 8), does not currently address seismic risk, licensees adopting this change must ensure that the proposed LCO 3.0.8 is considered with respect to other plant maintenance activities and integrated into the existing 10 CFR SO.6S(a)(4) process, whether the process is invoked by a TS or by (a)(4) itself.

3.1.4 Optional Changes and Variations In their application, the licensee states that, "Progress Energy is not proposing any variations or deviations from the TS changes described in the TSTF-372 Revision 4 or the NRC staff's model safety evaluation dated May 4, 200S."

3.2 Summary and Conclusions The option to relocate the snubbers to a licensee-controlled document, as part of the conversion to Improved STSs, has resulted in non-uniform and inconsistent treatment of snubbers. Some potential undesirable consequences of this inconsistent treatment of snubbers are:

Performance of testing during crowded windows when the supported system is inoperable, with the potential to reduce the snubber testing to a minimum (within the requirements of 10 CFR SO.SSa) since the relocated snubber requirements are controlled by the licensee;

Performance of testing during crowded windows when the supported system is inoperable, with the potential to increase the unavailability of safety systems; or

Performance of testing and maintenance on snubbers affecting multiple trains of the same supported system during the 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months allotted before entering MODE 3 under LCO 3.0.3.

To remove the inconsistency from the Robinson TSs, Carolina Power & Light Company, proposed to adopt TSTS-372 Revision 4, which is a risk-informed TS change that introduces a

- 14 delay time before entering the actions for the supported equipment when one or more snubbers are found inoperable or removed for testing. The delay time will provide needed flexibility in the performance of maintenance and testing during power operation and, at the same time, will enhance overall plant safety by (1) avoiding unnecessary unscheduled plant shutdowns, thus, minimizing plant transition and realignment risks; (2) avoiding reduced snubber testing, thus, increasing the availability of snubbers to perform their supporting function; (3) performing most of the required testing and maintenance during the delay time when the supported system is available to mitigate most challenges, thus avoiding increases in safety system unavailability; and (4) providing explicit risk-informed guidance in areas in which that guidance currently does not exist, such as the treatment of snubbers impacting more than one redundant train of a supported system.

The risk impact of the proposed TS changes for TST-372 was assessed generically following the three-tiered approach recommended in RG 1.177. A simplified bounding risk assessment was performed to justify the proposed TS changes. This bounding assessment assumes that the risk increase associated with the proposed addition of LCO 3.0.8 to the TSs is associated with accident sequences initiated by a seismically-induced LOOP event with concurrent loss of all safety system trains supported by the out-of-service snubbers. In the case of snubbers associated with more than one train, it is assumed that all affected trains of the supported system are failed. This assumption was introduced to allow the performance of a simple bounding risk assessment approach with application to all plants and was selected due to the lack of detailed plant-specific seismic risk assessments for most plants and the lack of fragility data for piping when one or more supporting snubbers are inoperable. The impact from the addition of the proposed LCO 3.0.8 to the TSs on defense-in-depth was also evaluated in conjunction with the risk assessment results.

Based on this integrated evaluation, the NRC staff concludes that the proposed addition of LCO 3.0.8 to the HBRSEP TSs would lead to insignificant risk increases, if any. Indeed, this conclusion is true without taking any credit for the removal of potential undesirable consequences associated with the current inconsistent treatment of snubbers testing, such as the effects of avoiding a potential reduction in the snubber testing frequency during the refueling outage and increased safety system unavailability.

Consistent with the NRC staff's approval and inherent in the implementation of TSTF-372, licensees must, as applicable, operate in accordance with the following stipulations:

1. Appropriate plant procedures and administrative controls will be used to implement the following Tier 2 Restrictions.

(a) At least one AFW train (including a minimum set of supporting equipment required for its successful operation) not associated with the inoperable snubber(s) must be available when LCO 3.0.8.a is used at PWR plants.

The licensee's assessment of Condition 1(a):

"Condition 1 (a) of the model safety evaluation applies to PWR plants and requires the availability of at least one train of AFW when LCO 3.0.8.a is used. The TS Bases for LCO 3.0.8 as

- 15 presented in TSTF-372 do not describe this requirement. The licensee's application includes this restriction. AFW is only required to be Operable in MODES 1, 2, and 3, and in MODE 4 when steam generator is being used for heat removal. During shutdown modes, an AFW train cannot be used for core cooling.

Most snubbers are tested during shutdown. The random testing of snubbers requires expansion of the test sample when failures are discovered. This may occur at a point in the outage in which AFW is not available, effectively prohibiting the use of LCO 3.0.B.a."

(b) At least one AFW train (including a minimum set of supporting equipment required for its successful operation) not associated with the inoperable snubber(s), or some alternative means of core cooling (e.g., F&B, fire water system or "aggressive secondary cooldown" using the steam generators), must be available when LCO 3.0.B.b is used at PWR plants.

The licensee's assessment of Condition 1(b):

"Condition 1 (b) of the model safety evaluation applies to PWR plants and placed conditions on the use of LCO 3.0.B.b that require the availability of at least one AFW train not associated with inoperable snubbers or some alternative means of core cooling such as feed and bleed, fire water system or "aggressive secondary cooldown". The TS Bases for LCO 3.0.B as presented in TSTF-372 do not describe the AFW or alternate core cooling source availability requirements of Item 1(b).

Progress Energy believes it is appropriate to modify the restriction of Condition 1 (a) pertaining to the use of LCO 3.0.B.a such that the use of LCO 3.0.B.a when the AFW is not required to be operable has the same flexibility with regards to reliance on the availability of alternate core cooling sources as LCO 3.0.B.b and to include the modified restriction on the use of LCO 3.0.B.a and the restriction on the use of LCO 3.0.B.b in the proposed TS Bases for LCO 3.0.B. The proposed TS Bases for LCO 3.0.B provides the following supplemental discussion and guidance for various modes of operation.

When applying LCO 3.0.B.a, at least one train of Auxiliary Feedwater (AFW) system must be OPERABLE during MODES when AFW is required to be OPERABLE. When applying LCO 3.0.B.a during MODES when AFW is not required to be OPERABLE, a core cooling method (such as Decay Heat Removal (DHR) system) must be available. When applying LCO 3.0.B.b, a means of core cooling must remain available (AFW, DHR, equipment necessary for feed and bleed operations, etc.).

Reliance on availability of a core cooling source during modes

- 16 where AFW is not required by TSs provides an equivalent safety margin for plant operations were LCO 3.0.8 not applied and meets the intent of Technical Specification Task Force Change Traveler TSTF-372, Revision 4, "Addition of LCO 3.0.8, Inoperability of Snubbers."

The licensee's above statement pertain to HBRSEP TS Bases. The staff's model safety evaluation requires that the licensee will establish TS Bases for LCO 3.0.8 as adopted with the applicable license amendment. The staff neither endorses nor approves changes to a licensee's TS Bases.

The licensee's adoption of specific TS changes are not affected.

(c) LCO 3.0.8.b cannot be used by West Coast PWR plants with no F&B capability when a snubber, whose non-functionality would disable more than one train of AFW in a seismic event of magnitude up to the plant's SSE, is inoperable.

The licensee's assessment of Condition 1(c):

"Condition 1(c) is applicable only to West Coast PWR plants and is therefore not applicable to HBRSEP Unit No.2."

(d) BWR plants must verify, every time the provisions of LCO 3.0.8 are used, that at least one success path, involving equipment not associated with the inoperable snubber(s), exists to provide makeup and core cooling needed to mitigate LOOP accident sequences.

The licensee's assessment of Condition 1(d):

"Condition 1(d) is applicable only to BWR plants and is therefore not applicable to HBRSEP Unit No.2."

(e) Every time the provisions of LCO 3.0.8 are used, licensees will be required to confirm that at least one train (or subsystem) of systems supported by the inoperable snubbers would remain capable of performing the system's required safety or support functions for postulated design loads other than seismic loads.

LCO 3.0.8 does not apply to nonseismic snubbers. In addition, a record of the design function of the inoperable snubber (Le., seismic vs. non-seismic), the implementation of any applicable Tier 2 restrictions, and the associated plant configuration shall all be available on a recoverable basis for staff inspection.

The licensee's assessment of Condition 1(e):

"Condition 1 (e) requires confirmation that at least one train (or subsystem) of systems supported by the inoperable snubbers would remain capable of performing their required safety or support functions for postulated design loads other than seismic loads. It also states that LCO 3.0.8 does not apply to non-seismic snubbers and identifies specific information that shall be available on a recoverable basis for staff inspection."

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2. When the licensee implements the provisions of LCO 3.0.8 for snubbers, which include delay times to enter the actions for the supported equipment when one or more snubbers are out of service for maintenance or testing, it must be done in accordance with an overall CRMP to ensure that potentially risk-significant configurations resulting from maintenance and other operational activities are identified and avoided, as discussed in the proposed TS Bases.

This objective is met by licensee programs to comply with the requirements of paragraph (a)(4) of the Maintenance Rule, 10 CFR 50.65, to assess and manage risk resulting from maintenance activities or when this process is invoked by LCO 3.0.8 or other TSs. These programs can support licensee decisionmaking regarding the appropriate actions to manage risk whenever a risk-informed TS is entered. Since the 10 CFR 50.65(a)(4) guidance (i.e.,

Reference 8), does not currently address seismic risk, licensees adopting this change must ensure that the proposed LCO 3.0.8 is considered in conjunction with other plant maintenance activities and integrated into the existing 10 CFR 50.65(a)(4) process. In the absence of a detailed seismic PRA, a bounding risk assessment, such as that utilized in this Safety Evaluation, shall be followed.

The licensee's assessment of Condition 2:

"Condition 2 directs that decision making must ensure that the proposed LCO 3.0.8 and seismic risk is considered in conjunction with maintenance activities. "

In its submittal, the licensee stated that it reviewed the NRC staff's evaluation, as well as the information provided to support TSTF-372, and has concluded that the justifications presented in the TSTF proposal and NRC staff model SE are applicable to HBRSEP and justify this amendment. Based on its own review, the NRC staff agrees. Therefore, the NRC staff concludes that the proposed TS changes for HBRSEP are acceptable.

Based on the above, the NRC staff concludes that the proposed LCO 3.0.8, which will be in Section 3.0 of the TS on LCO applicability, properly defines the rules and practices for the affected support LCOs for when one or more snubbers are unable to perform their associated support function(s). Therefore, the NRC further concludes that the proposed LCO meets the requirements of 10 CFR 50.36.

With the addition of LCO 3.0.8 to Section 3.0 of the TSs, there will be another LCO in that section, besides LCO 3.0.2 and LCO 3.0.7, that explains, in this case for snubbers, when LCOs do not have to be declared not met. Because of this, LCO 3.0.8 has to be listed in LCO 3.0.1 of Section 3.0 of TS. This is an administrative change that does not change any requirements in the TSs and is needed to identify the exceptions to TS 3.0.1. Based on these considerations, the NRC staff concludes that the addition of LCO 3.0.8 to LCO 3.0.1 meets 10 CFR 50.36, and is, therefore, acceptable.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the South Carolina State official was notified of the proposed issuance of the amendments. The State official had no comments.

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5.0 ENVIRONMENTAL CONSIDERATION

The amendments change a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no public comment on such finding on October 16, 2012 (77 FR 63347). Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.

6.0 CONCLUSION

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

7.0 REFERENCES

1. TSTF-372, Revision 4, "Addition of LCO 3.0.8, Inoperability of Snubbers," April 23, 2004.

2. Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," NRC, August 1998.

3. Regulatory Guide 1.177, "An Approach for Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications," NRC, August 1998.

4. Budnitz, R. J., et aI., "An Approach to the Quantification of Seismic Margins in Nuclear Power Plants," NUREG/CR-4334, Lawrence Livermore National Laboratory, July 1985.

5. Advanced Light Water Reactor Utility Requirements Document, Volume 2, ALWR Evolutionary Plant, PRA Key Assumptions and Groundrules, Electric Power Research Institute, August 1990.

6. Bier V. M., et aI., "Development and Application of a Comprehensive Framework for Assessing Alternative Approaches to Snubber Reduction," International Topical Conference on Probabilistic Safety Assessment and Risk Management PSA '87, Swiss Federal Institute of Technology, Zurich, August 30-September 4, 1987.

7. NUREG-1488, "Revised Livermore Seismic Hazard Estimates for Sixty-Nine Nuclear Power Plant Sites East ofthe Rocky Mountains," April 1994.

8. Nuclear Energy Institute, Revised SeCtion 11 of Revision 2 of NUMARC 93-01, May 2000.

9. Regulatory Guide 1.182, "Assessing and Managing Risk Before Maintenance Activities at Nuclear Power Plants," May 2000.

Principal Contributors: A. Billoch Colon G. Bedi R. Grover C. Schulten Date: February 26,2013

ML13017A381 "OGC concurrence not required per memo dated 10/29/07 (ML072980209)

OFFICE LPL2-2/PM LPL2-21LA(lT) LPL2-2/LA DSS/STSB DE/EPTB NAME ABiliochColon FKeith BClayton RElliott AMcMurtrav DATE 1115/13 1/25/13 1125113 2113/13 2121/13 i

OFFICE OGC" LPL2-2/BC LPL2-2/PM i NAME DATE N/A N/A JQuichocho 2/26/13

~chcolon 13

v t e Letter Sequence Approval
TAC:ME9236, Addition of LCO 3.0.8, Inoperability of Snubbers (Approved, Closed)
Initiation Request, Request Acceptance Administration Questions Answers Results Approval Other: ML12248A106
MONTHYEAR2012-09-11 [Table View]

ML13017A381