ML110490585

License Amendment, Issuance of Amendment Regarding Changes to Remove LCO 3/4.6.4, Snubbers, and Add LCO 3.0.8 on the Inoberability of Snubbers Using Consolidated Line Item Improvement Process

ML110490585
Person / Time
Site:	Nine Mile Point
Issue date:	03/10/2011
From:	Richard Guzman Division of Operating Reactor Licensing
To:	Belcher S Nine Mile Point
Guzman R, NRR/DORL, 415-1030
References
TAC ME3584
Download: ML110490585 (27)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 March 10, 2011 Mr. Samuel L. Belcher Vice President Nine Mile Point Nine Mile Point Nuclear Station, LLC P.O. Box 63 Lycoming, NY 13093

SUBJECT:

NINE MILE POINT NUCLEAR STATION, UNIT NO.1 - ISSUANCE OF AMENDMENT REGARDING CHANGES TO REMOVE LIMITING CONDITION FOR OPERATION (LCO) 3/4.6.4, "SNUBBERS," AND ADD LCO 3.0.8 ON THE INOPERABILITY OF SNUBBERS USING THE CONSOLIDATED LINE ITEM IMPROVEMENT PROCESS (TAC NO. ME3584)

Dear Mr. Belcher:

The Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment No. 207 to Renewed Facility Operating License No. DPR-63 for the Nine Mile Point Nuclear Station, Unit No.1 (NMP1), in response to your application dated March 18, 2010, (Agencywide Documents and Management System (ADAMS) Accession No. ML100830119).

This amendment revises the NIVlP1 Technical Specifications (TSs) for snubbers by removing TS 3/4.6.4, "Shock Suppressors (Snubbers)," relocating these requirements to a licensee-controlled document, and adding a new LCO 3.0.8, related to snubbers. In addition, the TS Table of Contents is revised to reflect these changes. The addition of LCO 3.0.8 is consistent with the industry Technical Specification Task Force (TSTF) Traveler TSTF 372-A, Revision 4, "Addition of LCO 3.0.8, Inoperability of Snubbers." A notice of the TSTF-372-A, Revision 4 TS improvement was published in the Federal Register on May 4, 2005 (70 FR 23252) as part of the Consolidated Line Item Improvement Process.

A copy of the related Safety Evaluation is enclosed. A Notice of Issuance will be included in the Commission's next regular biweekly Federal Register notice.

Sincerely, Richard V. Guzman, Senior Project Manager Plant Licensing Branch 1-1 Division of Operator Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-220

Enclosures:

1. Amendment No. 207 to DPR-63

2. Safety Evaluation cc w/encls: Distribution via Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 NINE MILE POINT NUCLEAR STATION, LLC (NMPNS)

DOCKET NO. 50-220 NINE MILE POINT NUCLEAR STATION, UNIT NO.1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 207 Renewed License No. DPR-63

1.

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

A.

The application for amendment by Nine Mile Point Nuclear Station, LLC (the licensee) dated March 18, 2010, 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.

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-63 is hereby amended to read as follows:

(2)

Technical Specifications The Technical Specifications contained in Appendix A, which is attached hereto, as revised through Amendment No. 207, is hereby incorporated into this license.

Nine Mile Point Nuclear Station, LLC shall operate the facility in accordance with the Technical Specifications.

- 2

3.

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

FOR THE NUCLEAR REGULATORY COMMISSION I!~~"'r. ;;;(.~

Nancy L. Salgado, Chief Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the License and Technical Specifications Date of Issuance: March 10, 2011

ATTACHMENT TO LICENSE AMENDMENT NO. 207 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-63 DOCKET NO. 50-220 Replace the following page of the Renewed Facility Operating License with the attached revised page. The revised page is identified by amendment number and contains marginal lines indicating the areas of change.

Remove Page Insert Page 3

3 Replace the following pages of Appendix A, Technical Specifications, with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Pages Insert Pages iii iii 27 27 259 259 260 260 261 262 263 263a

-3 (3)

Pursuant to the Act and 10 CFR Parts 30, 40, and 70 to receive, possess and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4)

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.

(5)

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 the operation of the facility.

C.

This renewed operating license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I:

Part 20, Section 30.34 of Part 30; Section 40.41 of Part 40; Section 50.54 and 50.59 of Part 50; and Section 70.32 of Part 70. This renewed license 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 also subject to the additional conditions specified or incorporated below:

(1)

Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 1850 megawatts (thermal).

(2)

Technical Specifications The Technical Specifications contained in Appendix A, which is attached hereto, as revised through Amendment No. 207, is hereby incorporated into this license. Nine Mile Point Nuclear Station, LLC shall operate the facility in accordance with the Technical Specifications.

(3)

Deleted Renewed License No. DPR 63 Amendment No. 191 to 206, 207

SECTION DESCRIPTION PAGE 3.3.4 Isolation Valves 4.3.4 Isolation Valves 143 3.3.5 Access Control 4.3.5 Access Control 151 3.3.6 Vacuum Relief 4.3.6 Vacuum Relief 153 3.3.7 Containment Spray 4.3.7 Containment Spray 159 3.4.0 Secondary Containment 164 Limiting Condition for Oeeration Surv!illance Regylrew!nts 3.4.1 Leakage Rate 4.4.1 Leakage Rate 165 3.4.2 Isolation Valves 4.4.2 Isolation Valves 168 3.4.3 Access Control 4.4.3 Access Control 170 3.4.4 Emergency Ventilation 4.4.4 Emergency Ventilation 173 3.4.5 Control Room Ventilation 4.4.5 Control Room Ventilation 178 3.5.0 Shutdown and Refueling 182 Limiting C2nditi2n for O~eration Surveillance Reguirements 3.5.1 Source Range Monitoring 4.5.1 Source Range Monitoring 183 3.5.2 Refueling Platform Interlock 4.5.2 Refueling Platform Interlock 186 3.6.0 General Reactor Plant 191 Limiting Condition for 0Reration Surveillance Regulrement!

3.6.1 Mechanical Vacuum Pump Isolation 4.6.1 Mechanical Vacuum Pump Isolation 192 3.6.2 Protective Instrumentation 4.6.2 Protective Instrumentation 194 3.6.3 Emergency Power Sources 4.6.3 Emergency Power Sources 255 3.6.4 (Deleted) 4.6.4 (Deleted) 259 AMENDMENT NO. 142, 176, 207 iii

3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY 3.0.1 When a system. subsystem, train, component or device is determined to be inoperable solely because its emergency power source is inoperable, or solely because its normal power source is inoperable. it may be considered operable for the purpose of satisfying the requirements of its applicable LCO, provided: (1) its corresponding normal or emergency power source is operable; and (2) all of its redundant system(s), subsystem(s), train(s), component(s) and device(s) are operable, or likewise satisfy the requirements of this specification. Unless both conditions (1) and (2) are satisfied, the unit shall be placed in a condition stated in the individual specification.

In the event LCO requirements cannot be satisfied because of circumstances in excess of those addressed in the specification, the unit shall be placed in a condition consistent with the individual specification unless corrective measures are completed that permit operation for the specified time interval as measured from initial discovery or until the reactor is placed in an operational condition in which the specification is not applicable.

3.0.2 through 3.0.7 - Reserved for Future Use 3.0.8 When one or more required snubbers are unable to perform their associated support function(s), any affected supported LCOes) 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 <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; 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 <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

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.

4.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY 4.0.1 SRs shall be met during the applicable reactor operating or other specified conditions for individual LCOs, unless otherwise stated in the SR. Failure to meet a surveillance, whether such failure is experienced during the performance of the surveillance or between performances of the surveillance. shall be failure to meet the LCO. Failure to perform a surveillance within the specified frequency shall be failure to meet the LCO except as provided in Specification 4.0.3. Surveillances do not have to be performed on inoperable equipment or variables outside specified limits.

4.0.2 Each SR shall be performed within the specified surveillance interval with a maximum allowable extension not to exceed 25 percent of the specified surveillance interval.

AMENDMENT NO. 142, 182, 207 27

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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 207 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-63 NINE MILE POINT NUCLEAR STATION. LLC NINE MILE POINT NUCLEAR STATION. UNIT NO.1 DOCKET NO. 50-220

1.0 INTRODUCTION

By letter dated March 18, 2010 (Agencywide Documents Access Management System (ADAMS) Accession No. ML100830119), Nine Mile Point Nuclear Station, LLC (NMPNS or the licensee) submitted a request for changes to the Nine Mile Point, Unit No.1 (NMP1) Technical Specifications (TSs). The proposed amendment would revise the NMP1 TSs for snubbers by removing TS 3/4.6.4, "Shock Suppressors (Snubbers);' relocating these requirements to a licensee-controlled document, and adding a new limiting condition for operation (LCO) 3.0.8 related to snubbers. In addition, the TS Table of Contents would be revised to reflect these changes This LCO establishes conditions under which TS systems would remain operable when one or more required snubbers are not capable of providing their related support function. The licensee stated that the proposed amendment is consistent with Nuclear Regulatory Commission (NRC) approved Technical Specification Task Force (TSTF) Improved Standard Technical Specifications (STS) Change Traveler, TSTF-372, Revision 4. The NRC staff published a notice of this TS improvement in the Federal Register on May 4,2005 (70 FR 23252) as part of the Consolidated Line Item Improvement Process (CUIP). The notice contains a model safety evaluation (SE). Since the licensee adopted TSTF-372, Revision 4, with minor plant-specific variations, the NRC staff has reproduced the model SE with plant-specific changes to reflect the licensee's application.

2.0 REGULATORY EVALUATION

In Title 10 of the Code of Federal Regulations (10 CFR) Section 50.36, the Commission established its regulatory requirements related to the content of TS. Pursuant to 10 CFR 50.36, TS are required to include items in the following five specific categories related to station operation: (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 planfs TS. As stated in 10 CFR 50. 36( c)(2)(i), the "Limiting conditions for operation are the lowest functional capability or performance levels of equipment required for safe operation of the facility. When an LCO of a

nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the technical specification..." NMP1 TS Sections 3.0 and 4.0, on "LCO and SR Applicability," provide details or ground 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 analysis based on different combinations of load conditions, depending on the design classification of the particular piping.

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

These requirements included:

• A requirement that snubbers be functional 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 <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 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 functional tests on a representative sample of at least 10% 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 STS.

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 4 criteria in 10 CFR 50.36(c)(2)(ii) for inclusion in the improved STS. 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 TS, 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 supported equipment that existed prior to the conversion to the improved STS (the only

- 3 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 STS, has resulted in non-uniform and inconsistent treatment of snubbers. On the one hand, plants that have relocated snubbers from their TS are allowed to change the TS requirements for snubbers under the auspices of 10 CFR 50.59, 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 STS have retained the 72-hour delay if snubbers are found to be inoperable, but they are not allowed to use 10 CFR 50.59 to change TS requirements for snubbers. It should also be noted that a few plants that converted to the improved STS 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, plants that have relocated can perform functional tests on the snubbers at power (as long as they enter the actions for the supported equipment) and at the same time can reduce the testing frequency (as compared to plants that have not relocated) if it is justified by 10 CFR 50.59 assessments. 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 that have been relocated from TS 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 <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> allotted before entering MODE 2 under LCO 3.0.3 (applicable to NUREG-1433, Standard TSs, General Electric Plants, BWRl4).

To remove the inconsistency in the treatment of snubbers among plants, the TSTF proposed 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

On April 23, 2004, the Nuclear Energy Institute (NEI) Risk Informed Technical Specifications Task Force (RITSTF) submitted a proposed change, TSTF-372, Revision 4, to the STS (NUREG 1430-1434) on behalf of the industry (TSTF-372, Revisions 1 through 3 were prior draft iterations). TSTF-372, Revision 4, is a proposal to add an STS LCO 3.0.8, 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.

This proposal is one of the industry's initiatives being developed under the risk-informed TS program. These initiatives are intended to maintain or improve safety through the incorporation of risk assessment and management techniques in TS, while reducing unnecessary burden and making TS requirements consistent with the Commission's other risk-informed regulatory requirements, in particular the Maintenance Rule in 10 CFR 50.65.

The proposed change adds a new LCO 3.0.8, to the TS. LCO 3.0.8 allows licensees to delay declaring an LCO not met for equipment, supported by snubbers unable to perform their associated support functions, when risk 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 <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; 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 <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

- 5 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.

The proposed TS change is further described in Section 2.0. The technical evaluation and approach used to assess its risk impact is discussed in Section 3.0 with the specific results and insights of the risk assessment discussed in Section 3.1. Section 3.2 summarizes the NRC staff's conclusions from the review of the proposed TS change.

The industry submitted TSTF-372, Revision 4, "Addition of LCO 3.0.8, Inoperability of 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 (f1CDF) and the average yearly large early release frequency (f1LERF) and (2) by the incremental conditional core damage probability (ICCDP) and the incremental conditional large early release probability (ICLERP). The assessed f1CDF and f1LERF 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 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 config uration 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 was performed to justify the proposed addition of LCO 3.0.8 to the TS. This approach was necessitated by (1) the general nature of the proposed TS changes (Le., 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%) of low probability (5%) of failure (HCLPF) of about 0.1g, expressed in terms of peak ground acceleration. Thus, a magnitude 0.1g earthquake is conservatively assumed to have 5% 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 magnitude higher than 0.1g, bounds a detailed analysis which would use mean seismic failure probabilities (fragilities) for the ceramic insulators.

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

- 7 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%) 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 gU 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 3.0.8b applies). However, boiling-water reactors (BWRs) typically have reliable alternative means of performing certain critical functions.

For example, if high pressure makeup (e.g., reactor core isolation cooling) and heat removal capability (e.g., suppression pool cooling) are unavailable in NMP1, reactor depressurization in conjunction with low pressure makeup (e.g., low pressure core spray) and heat removal capability (e.g., shutdown cooling) can be used to cool the core.

• The earthquake frequency at the 0.1g level was assumed to be 1 E-3/year for Central and Eastern US plants and 1 E-1/year for West Coast plants. Each of these two values envelop the range of earthquake frequency values at the 0.1g level, for Eastern US and West Cost 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 (ATWS) 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

- 8 level earthquake would fail all piping associated with inoperable snubbers, non LOOP sequences would not include any more failures associated with inoperable snubbers than 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 non-seismic (shock-type) loads and seismic loads which indicate that, in general, the risk impact of the out-of-service snubbers is smaller for non-seismic loads than for seismic loads. First, while a seismic load affects the entire plant, the impact of a non-seismic load is localized to a certain system or area of the plant. Second, although non-seismic 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 non-seismic loads is more plant specific, and thus harder to analyze generically, than for seismic loads. For these reasons, licensees will be required to confirm every time LCO 3.0.8 is used, that at least one train of each system that is supported by the inoperable snubber(s) would remain capable of performing their 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 TS are summarized and evaluated in the following Sections 3.1.1 to 3.1.3.

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 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%. This is a conservative value given that for core damage to occur under those conditions, two or more failures are required.

- 9

• 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 that all safety systems are unavailable to mitigate the accident, except for West Coast Pressurized-Water Reactor (PWR) plants. Credit for using feed and bleed (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 magnitude up to the plant's safe shutdown earthquake (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), l!.RCDF' caused by the out-of-service snubbers (as assumed in the bounding analysis). The second and third rows list the ICCDP (incremental conditional core damage probability) and the ICLERP (incremental conditional large early release probability) values, respectively. The ICCDP for the case where all inoperable snubbers are associated with only one train (or subsystem) of the supported safety systems, was obtained by multiplying the corresponding l!.RCDF value by the time fraction of the proposed 72-hour delay to enter the actions for the supported equipment. The ICCDP for the case where one or more of the inoperable snubbers are associated with multiple trains (or subsystems) of the same safety system, was obtained by multiplying the corresponding l!.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 since containment bypass scenarios, such as steam generator tube rupture accidents (only applicable to PWRs) and interfacing system LOCAs, would not be uniquely affected by the out of-service snubbers. Finally, the fourth and fifth rows list the assessed l!.CDF and l!.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 TS, and (2) testing of snubbers is associated with higher risk impact than 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 l!.CDF and l!.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 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 TS would have an insignificant risk impact.

-10 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 ISingle Train Multiple Train Single Train Multiple Train fdRCDF/yr 1E-6 SE-6 1E-4 SE-4 ICCDP BE-9 7E-9 BE-7 7E-7 ICLERP BE-1O 7E-10 BE-B 7E-B

~CDF/yr SE-9 SE-9 SE-7 SE-7 ftlLERF/yr SE-1O SE-1O ISE-B SE-B The assessed nCDF and nLERF values meet the acceptance criteria of 1 E-6/year and 1 E-7/year, respectively, based on guidance provided in RG 1.174. This conclusion is valid 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.B to the existing TS meets the RG 1.177 numerical guidelines of SE-7 for ICCDP and SE-B 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 2.0.

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 B), endorsed by RG 1.1B2 (Reference 9),

for implementing the requirements of paragraph (a)(4) of the 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., nRCDF) for a planned configuration is provided. This guidance states that a specific configuration that is associated with a CDF higher than 1 E-3/year should not be entered voluntarily. Since the assessed conditional risk increase, nRCDF' is significantly less than 1 E-3/year, plant configurations including out-of-service snubbers and other equipment may be entered voluntarily if supported by the results of the risk assessment required by 10 CFR SO.6S(a)(4), by LCO 3.0.B, or by other TS.

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

ARCDF Guidance Greater than 1 E-3/year Configuration should not normally be entered voluntarily ICCDP Guidance ICLERP Greater than 1 E-5 Configuration should not normally be entered voluntarily Greater than 1 E-6 1E-6to 1E-5 Assess non-quantifiable factors Establish risk management actions 1 E-7 to 1 E-6 Less than 1 E-6 Normal work controls Less than1E-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 (Reference 8). 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 (I.e., 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 TS. 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 TS.

The NRC staff finds that the risk assessment results support the proposed addition of LCO 3.0.8 to the TS. The risk increases associated with this TS change 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 inconsistent treatment of snubbers in the TS, 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

- 12 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 3.0.8a 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 3.0.8a applies. Therefore, potentially high-risk configurations can be avoided by ensuring that such a success path exists when LCO 3.0.8a applies. Based on a review of the accident sequences that contribute to the risk increase associated with LCO 3.0.8a, 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 were identified to prevent potentially high-risk configurations:

For BWR plants, one of the following two means of heat removal must be available when LCO 3.0.8a is used:

At least one high pressure makeup path (e.g., using high-pressure coolant injection (HPC1) or reactor core isolation cooling (RCIC) or equivalent) and heat removal capability (e.g., suppression pool cooling), including a minimum set of supporting equipment required for success, not associated with the inoperable snubber(s), or At least one low pressure makeup path (e.g., low-pressure coolant injection (LPCI) or containment spray (CS)) and heat removal capability(e.g., suppression pool cooling or shutdown cooling), including a minimum set of supporting equipment required for success, not associated with the inoperable snubber(s).

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 3.0.8b applies), it was assumed in the bounding analysis that all safety systems are unavailable to mitigate the accident. Based on a review of the accident sequences that contribute to the risk increase associated with LCO 3.0.8b (as modeled by the simplified bounding analysis) and defense-in-depth considerations, the following restriction was identified to prevent potentially high-risk configurations:

• When LCO 3.0.8b is used at BWR plants, it must be verified that at least one success path exists, using equipment not associated with the inoperable snubber(s), to provide makeup and core cooling needed to mitigate LOOP accident sequences.

- 13 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 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. 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, and by the TS 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 50.65(a)(4) guidance, the revised (May 2000) Section 11 of NUMARC 93-01, 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 50.65(a)(4) process whether the process is invoked by a TS or (a)(4) itself.

3.2 Relocation of TS 3/4.6A "Shock Suppressors (Snubbers)," to the NMP1 Station Procedures NMP1 's TS 3/4.6.4 currently contains requirements for snubber operability and surveillance testing. With one or more snubbers inoperable, the required TS Action is to replace or restore the inoperable snubber(s) to operable status or perform an engineering evaluation of the supported component within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Otherwise, the supported system is required to be declared inoperable. The licensee's license amendment request proposes to relocate TS 3/4.6.4, "Shock Suppressors (Snubbers)," to the NMP1 station procedures. The licensee states, "requirements for snubber operability and surveillance testing are not required by 10 CFR 50.36(c)(2)(ii) to be included in the TS. Relocating TS 3/4.6.4 to station procedures would permit snubber requirements to be revised in accordance with 10 CFR 50.59 and 10 CFR 50.55a without requiring a license amendment. Station procedures are controlled as described in the Updated Final Safety Analysis Report (UFSAR). Changes to station procedures are subject to review in accordance with 10 CFR 50.59."

As part of the development of the STS, the NRC staff determined that snubbers did not meet any of the four criteria in 10 CFR 50.36(c)2)(ii) for inclusion in the TSs (References 10, 11, and 12). Accordingly, the NRC staff has previously approved the relocation of requirements for snubbers to a licensee-controlled document such as the Technical Requirement Manual (TRM),

or a program document, and considers the licensee's request to relocate TS 3/4.6.4 to station procedures to be acceptable.

In addition, as stated in the licensee's license amendment request, changes to the station procedures are subject to review in accordance with 10 CFR 50.59, and snubber inservice testing and examination will be performed in accordance with Subsection ISTO of the ASME OM Code except where reliefs/alternatives have been approved in accordance with 10 CFR 50.55a.

The NRC staff notes that the licensee's stated actions will ensure that functionality and testing of snubbers will continue to be adequately assured after the relocation of the subject TS requirements, and is, therefore acceptable.

- 14 The licensee is still required to comply with the requirements of 10 CFR SO.SSa, including updating the snubber program at 120-month intervals and ensuring compliance with the requirements of the applicable ASME Code edition and addenda adopted by the licensee for the 120-month inspection interval, in accordance with 10 CFR SO.SSa(g)(4).

3.3 Summary and Conclusions In TSTF-372, the risk impact of the proposed TS changes was assessed 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 TS 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 TS on defense-in-depth was also evaluated in conjunction with the risk assessment results and was found to be acceptable.

Based on the integrated evaluation described above which bounds NMP1, the NRC staff concludes that the proposed addition of LCO 3.0.8 to the TS would lead to insignificant risk increases, if any. This conclusion is valid without taking any credit for the removal of potential undesirable consequences associated with the current inconsistent treatment of snubbers, such as the effects of avoiding a potential reduction in the snubber testing frequency and increased safety system unavailability.

Consistent with the NRC staff's approval and inherent in the implementation of TSTF-372, licensees interested in implementing LCO 3.0.8 must, as applicable, operate in accordance with the following stipulations:

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

a. 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. At least one high pressure makeup path (e.g., using high pressure coolant injection (HPCI) or reactor core isolation cooling (RCIC) or equivalent) and heat removal capability (e.g., suppression pool cooling), including a minimum set of supporting equipment required for success, not associated with the inoperable snubber(s), or At least one low pressure makeup path (e.g., low pressure coolant injection (LPCI) or core spray (CS>> and heat removal capability (e.g., suppression pool cooling or shutdown cooling), including a minimum set of supporting equipment required for success, not associated with the inoperable snubber(s).

- 15 In the licensee's license amendment request, NMP1 (a BWR-2 type reactor) stated that they will ensure appropriate plant procedures and administrative controls are revised to implement the above Tier 2 restrictions using the equivalent NMP1 plant specific systems which are:

High pressure:

Makeup - High Pressure Coolant Injection Heat Removal-Electromagnetic Relief Valves with Containment Spray in Torus Cooling Mode, or Emergency Condensers Low Pressure:

Makeup - Core Spray

b. Every time the provisions of LCO 3.0.8 are used licensees will be required to con'firm 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. LCO 3.0.8 does not apply to non-seismic snubbers. In addition, a record of the design function of the inoperable snubber (i.e., seismic vs. non-seismic), implementation of any applicable Tier 2 restrictions, and the associated plant configuration shall be available on a recoverable basis for NRC staff inspection.

Per the licensee's license amendment request, NMP1 will ensure appropriate plant procedures and administrative controls are revised to implement the above Tier 2 restrictions.

Should licensees implement 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 TS, 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 50.65(a)(4) guidance, the revised (May 2000)

Section 11 of NUMARC 93-01, 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 utilized in this SE, shall be followed.

The licensee's license amendment request states, "NMPNS will establish TS Bases for LCO 3.0.8 which provide guidance and details on how to implement the new requirements, LCO 3.0.8 requires that risk be managed and assessed. The Bases will also state that while the industry and NRC guidance on implementation of 10 CFR 50.65{a)(4), the Maintenance Rule,

- 16 does not address seismic risk, LCO 3.0.8 should be considered with respect to other plant maintenance activities, and integrated into the existing Maintenance Rule process to the extent possible so that maintenance on any unaffected train or subsystem is properly controlled, and emergent issues are properly addressed.

The risk assessment need not be quantified, but may be a qualitative assessment of the vulnerability of systems and components when one or more snubbers are not able to perform their associated support function.

The Bases for TS 3.0.8 will be established and maintained in accordance with TS 6.5.6, "Technical Specification (TS) Bases Control Program."

The NRC concludes that the licensee's approach discussed above is consistent with the requirements defined for BWRs in NRC-approved TSTF-372. In addition, based on the technical evaluation provided in Section 3.2, the NRC further concludes that the licensee's proposed change regarding relocation of TS 3/4.6.4 requirements to the NMP1 station procedures is acceptable. However, the licensee is still required to ensure that the snubber program remains in compliance with the applicable ASME Code edition and addenda and the requirements of 10 CFR 50.55a.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the New York State official was notified of the proposed issuance of the amendment. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes SRs.

The NRC staff has determined that the amendment involves 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 amendment involves no significant hazards consideration, and there has been no public comment on such finding (75 FR 39979, July 13, 2010). In addition, the NRC staff has determined that some of the changes are administrative in nature. Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9) and (c)(10). Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

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) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

- 17

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,"

USNRC, August 1998. (ML003740133).

3.

Regulatory Guide 1.177, "An Approach for Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications," USNRC, August 1998 (ML003740176).

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 of the Rocky Mountains," April 1994.

8.

NEI, Revised Section 11 of Revision 2 of NUMARC 93-01, "Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants," May 2000.

9.

Regulatory Guide 1.182, "Assessing and Managing Risk Before Maintenance Activities at Nuclear Power Plants," May 2000 (ML003699426).

10.

"Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors", 58 FR 39132, July 22,1993.

11.

"Technical Specifications," 60 FR 36953, July 19, 1995.

12.

"Waterford Steam Electric Station, Unit 3 - Issuance of Amendment Re: Relocation of Technical Specification 3.7.8 and addition of Limiting Condition for Operation 3.0.8 Regarding the Inoperability of Snubbers (TAC No. MD9672)," May 1, 2009 (ML090830472).

Principal Contributors: R. Grover G. Bedi Date: March 10, 2011

ML110490585

• SE rovided b memo. No sUbstantial chan es made. ** Concurrence via e-mail OFFICE LPL1-1/PM LPL1-1/LA DCIICPTB/BC NAME RGuzman SUttle DATE 3/9/11 3/9/11 March 10, 2011 Mr. Samuel L. Belcher Vice President Nine Mile Point Nine Mile Point Nuclear Station, LLC P.O. Box 63 Lycoming, NY 13093

SUBJECT:

NINE MILE POINT NUCLEAR STATION, UNIT NO.1 - ISSUANCE OF AMENDMENT REGARDING CHANGES TO REMOVE LIMITING CONDITION FOR OPERATION (LCO) 3/4.6.4, "SNUBBERS," AND ADD LCO 3.0.8 ON THE INOPERABILITY OF SNUBBERS USING THE CONSOLIDATED LINE ITEM IMPROVEMENT PROCESS (TAC NO. ME3584)

Dear Mr. Belcher:

The Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment No. 207 to Renewed Facility Operating License No. DPR-63 for the Nine Mile Point Nuclear Station, Unit No. 1 (NMP1), in response to your application dated March 18, 2010, (Agencywide Documents and Management System (ADAMS) Accession No. ML100830119).

This amendment revises the NMP1 Technical Specifications (TSs) for snubbers by removing TS 3/4.6.4, "Shock Suppressors (Snubbers)," relocating these requirements to a licensee-controlled document, and adding a new LCO 3.0.8, related to snubbers. In addition, the TS Table of Contents is revised to reflect these changes. The addition of LCO 3.0.8 is consistent with the industry Technical Specification Task Force (TSTF) Traveler TSTF 372-A, Revision 4, "Addition of LCO 3.0.8, Inoperability of Snubbers." A notice of the TSTF-372-A, Revision 4 TS improvement was published in the Federal Register on May 4, 2005 (70 FR 23252) as part of the Consolidated Line Item Improvement Process.

A copy of the related Safety Evaluation is enclosed. A Notice of Issuance will be included in the Commission's next regular biweekly Federal Register notice.

Sincerely, IRAJ Richard V. Guzman, Senior Project Manager Plant Licensing Branch 1-1 Division of Operator Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-220

Enclosures:

1. Amendment No. 207 to DPR-63

2. Safety Evaluation cc w/encls: Distribution via Listserv Distribution:

PUBLIC RidsOgcRp RidsOGCMailCenter RidsRgn 1 MailCenter LPL1-1 RtF RidsNrrDorlLPL1-1 RidsNrrLASLittle RidsNrrPMNineMilePoint R. Grover, NRR RidsNrrDirsltsb RidsNrrDorlDpr RidsAcrsAcnw_MailCenter G. Bedi, NRR RidsNrrDciCptb NRR-106 LPL1-1/BC DIRSIITSB/BC OGC Elliott*

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