ML101880333

Issuance of Amendment No. 206 Modification of TS Section 3.2.7.1 and 4.2.7.1

ML101880333
Person / Time
Site:	Nine Mile Point
Issue date:	07/26/2010
From:	Richard Guzman Plant Licensing Branch 1
To:	Belcher S Nine Mile Point
Guzman R, NRR/DORL, 415-1030
References
TAC ME2253
Download: ML101880333 (17)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 July 26, 2010 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 MODIFICATION OF TECHNICAL SPECIFICATION 3.2.7.1 AND 4.2.7.1 - PRIMARY COOLANT SYSTEM PRESSURE ISOLATION VALVES (TAC NO. ME2253)

Dear Mr. Belcher:

The Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment No. 206 to Renewed Facility Operating License No. DPR-63 for the Nine Mile Point Nuclear Station, Unit No.1, in response to your application dated September 18, 2009 (Agencywide Documents Access Management System (ADAMS) Accession No. ML092640681), as supplemented on October 15, 2009 (ADAMS Accession No. ML092940717) and April 14, 2010 (ADAMS Accession No. ML101110179).

This amendment revises the Technical Specifications (TSs) by modifying TS Section 3.2.7.1 and 4.2.7.1, "Primary Coolant System Pressure Isolation Valves," to incorporate requirements that are consistent with Section 3.4.5 of the Improved Standard Technical Specifications (STS),

NUREG-1433, Revision 3.0, "Standard Technical Specifications General Electric Plants, BWR/4."

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. 206 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. 206 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 September 18, 2009, as supplemented on October 15, 2009 and April 14, 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. 206, is hereby incorporated into this license.

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

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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.~

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: July 26, 2010

ATTACHMENT TO LICENSE AMENDMENT NO. 206 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 116 116 116a 117 117

-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. 206, 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, 192, 193, 194, 195, 196, 197, 198, 199,200,201,202,203,204,205206

LIMITING CONDITION FOR OPERATION REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVE (PIV) LEAKAGE Applicability:

Applies to the operating status of isolation valves for systems connected to the reactor coolant system.

Objective:

To increase the reliability of reactor coolant system pressure isolation valves thereby reducing the potential of an intersystem loss of coolant accident.

Specification:

---

NOTES-----------------------------

I 3.2.7.1

1. Separate specification entry is allowed for each flow path.

2.

Enter applicable specifications for systems made inoperable by PIVs.

a. The integrity of each pressure isolation valve shall be demonstrated. Valve leakage shall be within limit during the power operating and hot shutdown reactor operating conditions.

AMENDMENT NO.~, 206 SURVEILLANCE REQUIREMENT 4.2.7.1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVE (PIV) LEAKAGE Applicability:

Applies to the periodic testing of reactor coolant system pressure isolation valves.

Objective:

To increase the reliability of reactor coolant system pressure isolation valves thereby reducing the potential of an intersystem loss of coolant accident.

Specification:

---

NOTE------------------------------

Not required to be performed in the hot shutdown reactor operating condition.

a. The equivalent leakage of each reactor coolant system PIV shall be verified to be S 0.5 gpm per nominal inch of valve size up to a maximum of 5 gpm, at a reactor coolant system pressure

~ 1010 psig and S 1050 psig, at a frequency of 24 months.

116

LIMITING CONDITION FOR OPERATION

b.

If one or more flow paths with leakage from one or more PIVs is not within limit:

1. Isolate the high pressure portion of the affected system from the low pressure portion by use of one closed manual, deactivated automatic, or check valve within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, and

2. Isolate the high pressure portion of the affected system from the low pressure portion by use of a second closed manual, deactivated automatic, or check valve within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

Each valve used to satisfy Specifications b.1 and b.2 above must have been verified to meet Specification 4.2.7.1.a and be in the reactor coolant system pressure boundary or the high pressure portion of the system.

c.

If Specification 3.2.7.1.b cannot be met, an orderly shutdown shall be initiated within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and the reactor shall be in the cold shutdown condition within 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />.

SURVEILLANCE REQUIREMENT AMENDMENT NO. 206 116a

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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. 206 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 September 18, 2009 (Agencywide Documents Access Management System (ADAMS) Accession No. ML092640681), as supplemented on October 15, 2009 (ADAMS Accession No. ML092940717) and April 14, 2010 (ADAMS Accession No. ML101110179), 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 TSs by modifying TS Section 3.2.7.1 and 4.2.7.1, "Primary Coolant System Pressure Isolation Valves," to incorporate requirements that are consistent with Section 3.4.5 of the Improved Standard Technical Specifications (STS), NUREG-1433, Revision 3.0, "Standard Technical Specifications General Electric Plants, BWR/4."

The supplemental letters dated October 15,2009, and April 14, 2010, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the Nuclear Regulatory Commission (NRC) staff's original proposed no significant hazards consideration determination noticed in the Federal Register on October 14, 2009 (74 FR 52824).

2.0 REGULATORY EVALUATION

Section 182a of the Atomic Energy Act (the "Act") requires applicants for nuclear power plant operating licenses to include TSs as part of the license. The TSs ensure the operational capability of structures, systems, and components that are required to protect the health and safety of the public. The NRC's regulatory requirements related to the content of the TSs are contained in Title 10 of the Code of Federal Regulations (10 CFR) Section 50.36. That regulation requires that the TS include items in the following specific categories: (1) safety limits, limiting safety systems settings, and limiting control settings; (2) limiting conditions for operation (LCOs); (3) surveillance requirements (SRs); (4) design features; (5) administrative controls; (6) decommissioning; (7) initial notifications; and (8) written reports. However, the regulation does not specify the particular TS to be included in a plant's license.

- 2 In general, there are two classes of changes to TS: (1) changes needed to reflect modifications to the design basis (TS are derived from the design basis), and (2) voluntary changes to take advantage of the evolution in policy and guidance as to the required content and preferred format of TS over time. This amendment deals with the second class of changes.

In July 1993, the NRC issued the Final Policy Statement on TSs for Nuclear Power Reactors (58 FR 39132). The policy states, "LCOs which do not meet any of the criteria below may be proposed for removal from the Technical Specifications and relocation to licensee-controlled documents, such as the Final Safety Analysis Report (FSAR). The criteria may be applied to either standard or custom TSs... In accordance with this policy statement, improved STS have been developed and will be maintained for each nuclear steam supply system (NSSS) owners group. The commission encourages licensees to use the improved STS as the basis for plant specific technical specifications."

In accordance with 10 CFR 50.90, licensees can request amendments to their TS, either to remove items that do not meet the criteria of 10 CFR 50.36 for inclusion, or for other reasons.

These TS requirements may be relocated to other licensee-controlled documents.

10 CFR 50.36 specifies the criteria for including LCOs in the TS for commercial nuclear power reactors. According to 10 CFR 50.36(c)(2)(ii), an LCO must be established for items that meet one or more of the following criteria:

Criterion 1: Installed instrumentation that is used to detect, and indicate in the control room, a significant degradation of the reactor coolant pressure boundary.

Criterion 2: A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

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

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

SRs in 10 CFR 50.36 are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the LCOs will be met.

In June 2004, the NRC issued Revision 3 to NUREG-1433, "Standard Technical Specifications, General Electric Plants (BWR/4)." On May 6, 1991, the NRC issued Generic Letter (GL) 91-08, "Removal of Component Lists from Technical Specifications." This GL contains guidance on removing component lists from plant TS and relocating them to other licensee-controlled documents.

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3.0 TECHNICAL EVALUATION

3.1 Background

Pressure Isolation Valves (PIVs) isolate the high pressure reactor coolant system (RCS) from connected low pressure piping systems. PIVs may leak reactor coolant from the RCS to the lower pressure piping systems through normal wear or mechanical deterioration. The TS Sections 3.2.7.1 and 4.2.7.1 allow such leakage in quantities that do not compromise safety.

These limits ensure that leakage from the RCS to the low pressure systems, through the PIVs, does not overpressure the low pressure piping. If such a leak overpressured the low pressure piping, the result could be a loss-of-coolant accident outside of containment. Such an accident would degrade the ability for long-term core cooling.

3.2 Proposed Changes The proposed amendment request has three different types of changes to the NMP1 TS:

• Administrative Changes. Non-technical, administrative changes were intended to incorporate human-factors principles into the form and structure of the improved plant TS so that they would be easier to use for plant operations personnel. These changes are editorial in nature or involve the reorganization or reformatting of requirements without affecting technical content or operational requirements.

Relocated requirements. Existing TS requirements which do not satisfy or fall within any of the four specified criteria may be relocated to appropriate licensee-controlled documents.

• Less Restrictive Requirements. When requirements have been shown to provide little or no safety benefit, their removal from the TS may be appropriate. In most cases, relaxations in TS are the result of generic NRC actions, new NRC staff positions that have evolved from technological advancements and operating experience, or resolution of the Owners Groups' comments on the improved STS.

The proposed amendment revises TS Section 3.2.7.1, which is similar in concept to the LCO and Condition A for TS 3.4.5 of NUREG-1433, Revision 3.0, "Standard Technical Specifications General Electric Plants, BWR/4."

1. The title of the section is revised from "Primary Coolant System Pressure Isolation Valves" to "Reactor Coolant System Pressure Isolation Valve (PIV) Leakage," and the word "primary" is changed to "reactor" in both the Applicability and Objective paragraphs.

2. The following two new notes are added preceding Specification 3.2.7.1.a:

• Separate specification entry is allowed for each flow path.

Enter applicable specifications for systems made inoperable by PIVs.

- 4

3. Specification 3.2.7.1.a is revised by removing the reference to TS Table 3.2.7.1 (which lists the PIVs and associated leakage limits) and inserting "be within limit during the power operating and hot shutdown reactor operating conditions."

4. A new Specification 3.2.7.1.b is added to incorporate actions to be taken if one or more flow paths with leakage from one or more PIVs is not within limit. These actions and associated clarifying note are:

Isolate the high pressure portion of the affected system from the low pressure portion by use of one closed manual, deactivated automatic, or check valve within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, and Isolate the high pressure portion of the affected system from the low pressure portion by use of a second closed manual, deactivated automatic, or check valve within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

Each valve used to satisfy Specifications b.1 and b.2 above must have been verified to meet Specification 4.2.7.1.a and be in the reactor coolant system boundary or the high pressure portion of the system.

The proposed amendment includes the following revisions to TS Section 4.2.7.1, which are similar in concept to SR 3.4.5.1 in NUREG-1433:

1. The title of the section is revised from "Primary Coolant System Pressure Isolation Valves" to "Reactor Coolant System Pressure Isolation Valve (PIV) Leakage," and the word "primary" is changed to "reactor" in both the Applicability and Objective paragraphs.

2. The following new note is added preceding Specification 4.2.7.1.a:

Not required to be performed in the hot shutdown reactor operating condition.

3. Specification 4.2.7.1.a and the associated footnote (a) are deleted and replaced with the following:

• The equivalent leakage of each RCS PIV shall be verified to be S 0.5 gallons per minute (gpm) per nominal inch of valve size up to a maximum of 5 gpm, at an RCS pressure ~ 1010 pounds per square inch gauge (psig) and S 1050 psig, at a frequency of 24 months.

The proposed amendment removes TS Table 3.2.7.1 in its entirety. The list of PIVs (but not the maximum allowable leakage values and associated notes as reflected in the proposed Specification 4.2.7.1.a) will be relocated to the NIVI P1 Updated Final Safety Analysis Report (UFSAR). New proposed TS Bases that reflect the proposed TS changes will be processed in accordance with the NMP1 TS Bases Control Program (TS 6.5.6).

The changes are discussed in further detail below in Sections 3.3, 3.4, and 3.5.

- 5 3.3 Changes to TS Section 3.2.7.1 The title of the section is revised from "Primary Coolant System Pressure Isolation Valves" to "Reactor Coolant System Pressure Isolation Valve (PIV) Leakage," and the word "primary" is changed to "reactor" in both the Applicability and Objective paragraphs. This is an administrative change that more specifically reflects the purpose of the TS section and is consistent with NUREG-1433, Revision 3.0.

The following two new notes are added (1) separate specification entry is allowed for each flow path and (2) enter applicable specifications for systems made inoperable by PIVs. The notes provide clarification for the proper application of Specification 3.2.7.1. Note (1) clarifies that the specification can be entered separately for each flow path based on the independence of the flow paths. The NRC staff reviewed Specification 3.2.7.1 and concluded that the required actions in 3.2.7.1.b and 3.2.7.1.c provide appropriate compensatory measures for separate affected RCS PIV flow paths. Note (2) requires an evaluation of affected systems if a PIV is inoperable. The leakage may have affected system operability, or isolation of a leaking flow path with an alternate valve may have degraded the ability of the interconnected system to perform its safety function. As a result, the applicable Specifications for systems made inoperable by PIVs must be entered. This ensures appropriate remedial actions are taken, if necessary, for the affected systems. Therefore, the NRC staff finds this acceptable.

Reference to TS Table 3.2.7.1 (which lists the PIVs and associated leakage limits) is removed from Specification 3.2.7.1.a and is replaced with a requirement that valve leakage shall be within limit during the power operating and hot shutdown reactor operating conditions. The list of PIVs (but not the maximum allowable leakage values and associated notes) will be relocated to the NMP1 UFSAR, this change is discussed in Section 3.5 below.

A new Specification 3.2.7.1.b is added to incorporate actions to be taken if one or more flow paths with leakage from one or more PIVs is not within limit. NIVIP1 currently requires plant shutdown if any PIV leakage exceeds the specified maximum allowable leakage. The new actions will minimize the potential to initiate an unnecessary plant shutdown when sufficient flow path isolation can be provided. Isolating the high pressure portion of the affected system from the low pressure portion by use of one closed manual, deactivated automatic, or check valve within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> provides time to reduce leakage in excess of the allowable limit and to isolate the flow path if leakage cannot be reduced while corrective actions to reseat the leaking PIVs are taken. The 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> allows time for these actions and restricts the time of operation with leaking valves. Isolating the high pressure portion of the affected system from the low pressure portion by use of a second closed manual, deactivated automatic, or check valve within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> considers the time required to complete the action, the low probability of a second valve failing during this time period, and the low probability of a pressure boundary rupture of the low pressure piping when overpressurized to reactor pressure. These action completion times provide reasonable intervals for isolating the flow path and restoring a leaking PIV to operable status. This is a less restrictive change that is consistent with NUREG-1433, Revision 3.0 and will not adversely impact nuclear safety because the flow path will be sufficiently isolated, the period of time without redundant isolation capability will be limited, and the probability of a second valve failing during this time period is low.

- 6 3.4 Changes to TS Section 4.2.7.1 The title of the section is revised from "Primary Coolant System Pressure Isolation Valves" to "Reactor Coolant System Pressure Isolation Valve (PIV) Leakage," and the word "primary" is changed to "reactor" in both the Applicability and Objective paragraphs. This is an administrative change that more specifically reflects the purpose of the TS section and is consistent with NUREG-1433, Revision 3.0.

The following new note is added, "Not required to be performed in the hot shutdown reactor operating condition" to specification 4.2.7.1. This note permits entry into the hot shutdown condition for leakage testing at high differential pressures with stable conditions that are not possible in the cold shutdown and refueling conditions. This note is consistent with NUREG-1433, Revision 3.0.

Specification 4.2.7.1.a and the associated footnote are removed and replaced with, "The equivalent leakage of each reactor coolant system PIV shall be verified to be S 0.5 gpm per nominal inch of valve size up to a maximum of 5 gpm, at a reactor coolant system pressure

~ 1010 psig and S 1050 psiq, at a frequency of 24 months." Verification that reactor coolant system PIVs are S 0.5 gpm per nominal inch of valve size up to a maximum of 5 gpm is stated in TS Table 3.2.7.1 footnote (a)1. This change is relocating a footnote from Specification 3.2.7.1 Table 3.2.7.1 to Specification 4.2.7.1.a and is consistent with NUREG-1433, Revision 3.0. The added test pressure represents the RCS nominal operating pressure of 1030 psig with a tolerance of plus or minus 20 psig. The 24-month leakage test frequency is established in accordance with the American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code), 2004 Edition, which is the applicable code for the NMP1 Fourth 1O-year Inservice Test (1ST) interval. In accordance with the ASME OM Code, ISTC-1300, the PIVs are Category A valves since their seat leakage is limited to a specific maximum amount in the closed position for fulfillment of their required function. For Category A valves other than containment isolation valves, the ASME OM Code, ISTC-3630, specifies a test frequency of at least once every 2 years. Since the ASME OM Code is also the basis for preparation of the previously referenced 1STprogram, the proposed change to Specification 4.2.7.1.a is considered administrative in nature. Compliance with 10 CFR 50.55a, and the 1ST Program, is required by the NMP1 operating license. Programmatic controls for the 1ST Program are established in TS Section 6.5.4, "Inservice Testing Program." This change is consistent with NUREG-1433, Revision 3.0.

3.5 Removal of TS Table 3.2.7.1 The proposed amendment removes TS Table 3.2.7.1 in its entirety. The list of PIVs (but not the maximum allowable leakage values and associated notes as reflected in the proposed Specification 4.2.7.1.a) will be relocated to the NMP1 UFSAR. The NRC staff issued GL 91-08 to give guidance on relocating component lists from TS to licensee-controlled documents. A license amendment to relocate these component lists allows licensees to revise or update the lists in the future without having to apply for another license amendment. Changes to such lists, when relocated to licensee-controlled documents, are subject to review under the provisions of 10 CFR 50.59; therefore, a formal means of control over these lists still exists. to GL 91-08 contains specific issues that should be addressed by any request to remove component lists from plant TS. The issues that should be addressed include:

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1. an appropriate description of the scope of the components to which the TS requirements

apply,

2. an incorporation into the TS of any modifications or exceptions contained in notes to the table, and

3. a confirmation that the removed lists have been incorporated into appropriately controlled plant procedures.

With respect to issue 1 above, the scope of the PIVs to which the requirements of TS 3.2.7.1 and 4.2.7.1 apply was initially established in the NRC Order for Modification of License dated April 20, 1981 (ADAMS Accession No. ML010990058), and was subsequently modified by License Amendment No. 154, dated March 20, 1995 (ADAMS Accession No. ML011080725).

The list of PIVs in TS Table 3.2.7.1 will be relocated to the NMP1 UFSAR; however, the TS requirements relating to PIVs remain applicable. Therefore, removal of TS Table 3.2.7.1 and removal of references to the table from the text to TS 3.2.7.1 and 4.2.7.1 does not affect the scope of components to which the TS requirements apply.

With regard to Item 2 above, there are 3 footnotes in the TS Table 3.2.7.1 list of PIVs. The first footnote ("Leakage rates shall be limited to 0.5 gpm per nominal inch of valve diameter up to a maximum of 5 gpm") is incorporated into the proposed revision to Specification 4.2.7.1.a. The second footnote, "Test differential pressure shall not be less than 150 psid" is deleted. This test requirement was included in the NRC Order for Modification of License dated April 20, 1981, that initially established the PIV testing requirements. Inclusion of this detail in the TS is not required since the 1ST Program, required by 10 CFR 50.55a, provides the leak test requirements for these valves. The third footnote, "The observed leakage at the test differential pressure shall be adjusted to the functional maximum pressure differential" is also a detail that is not required to be included in the TS since the 1ST Program provides the leak test requirements for these valves. Consistent with NUREG-1433, Revision 3.0, this test provision is relocated to the TS Bases.

With regard to Item 3 above, NMPNS stated a regulatory commitment to ensure that the list of PIVs is relocated to the NMP1 UFSAR upon implementation of the license amendment.

Changes to the UFSAR are controlled in accordance with the provisions of 10 CFR 50.59.

Therefore, a formal means of control over these lists still exists.

3.6 Conclusion After reviewing the licensee's application, the NRC staff has determined that the proposed changes to the requirements specified in TS 3.2.7.1 and 4.2.7.1 for the RCS PIVs are consistent with NUREG-1433, Revision 3.0 and will continue to ensure that excessive leakage through these valves is properly identified and resolved. Testing in accordance with the 1ST Program will detect PIV leakage in excess of the established limits. When these limits are exceeded, required actions will initiate appropriate activities to minimize the impact of the leakage.

Inclusion of these actions will eliminate the potential to initiate an unnecessary plant shutdown when sufficient flow path isolation can be provided. Relocation of the list of PIVs from the TS to a licensee-controlled document (the UFSAR) in accordance with the guidance in GL 91-08 is an

- 8 administrative change that does not alter the TS requirements that are applicable to the PIVs.

Thus, the proposed changes will not have an adverse impact on nuclear safety.

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 (74 FR 52824, Oct. 14, 2009). 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.

Principal Contributors: K. Bucholtz R. Lake Date: July 26, 2010

ML092940717) and April 14, 2010 (ADAMS Accession No. ML101110179).

This amendment revises the Technical Specifications (TSs) by modifying TS Section 3.2.7.1 and 4.2.7.1, "Primary Coolant System Pressure Isolation Valves," to incorporate requirements that are consistent with Section 3.4.5 of the Improved Standard Technical Specifications (STS),

NUREG-1433, Revision 3.0, "Standard Technical Specifications General Electric Plants, BWR/4."

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, IRA!

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. 206 to DPR-63

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

PUBLIC RidsOgcRp RidsOGCMailCenter RidsRgn1 MailCenter LPL1-1 R/F RidsNrrDorlLPL1-1 RidsNrrLASLitt/e RidsNrrPMNineMilePoint KBucholtz, NRR RidsNrrDirsltsb RidsNrrDorlDpr RidsAcrsAcnw_MailCenter ADAMS Accession No'.. ML101880333

• SE provided by memo No substantial changes made NRR-106 OFFICE LPL1-1/PM LPL1-1/LA DCI/CPTB/BC DIRSIITSB/BC OGC LPL1-1/BC NAME RGuzman SLittie AMcMurtray RElliott*

MSmith NSalgado DATE 7/13/10 7/14/10 5/6/10 5/12/10 SE DTD 7/19/10 7/26/10