ML051920138

Issuance of License Amendment 226 Administrative Changes

ML051920138
Person / Time
Site:	Vermont Yankee
Issue date:	08/15/2005
From:	Richard Ennis NRC/NRR/DLPM/LPD1
To:	Kansler M Entergy Nuclear Operations
References
TAC MC5243
Download: ML051920138 (15)

Text

August 15, 2005 Mr. Michael Kansler President Entergy Nuclear Operations, Inc.

440 Hamilton Avenue White Plains, NY 10601

SUBJECT:

VERMONT YANKEE NUCLEAR POWER STATION - ISSUANCE OF AMENDMENT RE: ADMINISTRATIVE CHANGES (TAC NO. MC5243)

Dear Mr. Kansler:

The Commission has issued the enclosed Amendment No. 226 to Facility Operating License DPR-28 for the Vermont Yankee Nuclear Power Station, in response to your application dated December 6, 2004, as supplemented on June 14, 2005.

The amendment makes administrative and other miscellaneous changes to the license and Technical Specifications (TSs) including correction of references and deleting obsolete or redundant TS requirements and surveillances.

A copy of the related Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commissions biweekly Federal Register notice.

Sincerely,

/RA/

Richard B. Ennis, Senior Project Manager, Section 2 Project Directorate I Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-271

Enclosures:

1. Amendment No. 226 to License No. DPR-28

2. Safety Evaluation cc w/encls: See next page

Mr. Michael Kansler President Entergy Nuclear Operations, Inc.

440 Hamilton Avenue White Plains, NY 10601

SUBJECT:

VERMONT YANKEE NUCLEAR POWER STATION - ISSUANCE OF AMENDMENT RE: ADMINISTRATIVE CHANGES (TAC NO. MC5243)

Dear Mr. Kansler:

The Commission has issued the enclosed Amendment No. 226 to Facility Operating License DPR-28 for the Vermont Yankee Nuclear Power Station, in response to your application dated December 6, 2004, as supplemented on June 14, 2005.

The amendment makes administrative and other miscellaneous changes to the license and Technical Specifications (TSs) including correction of references and deleting obsolete or redundant TS requirements and surveillances.

A copy of the related Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commissions biweekly Federal Register notice.

Sincerely,

/RA/

Richard B. Ennis, Senior Project Manager, Section 2 Project Directorate I Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-271

Enclosures:

1. Amendment No. 226 to License No. DPR-28

2. Safety Evaluation cc w/encls: See next page DISTRIBUTION:

PUBLIC OGC PHearn PDI-2 Reading ACRS BPoole, OGC JUhle GHill (2) VBucci, OIG DRoberts TBoyce REnnis DLPM DPR CRaynor CAnderson, RGN-I Accession No.: ML051920138 Package: ML OFFICE PDI-2/PM PDI-2/LA IROB-A/SC OGC PDI-2/SC NAME REnnis CRaynor TBoyce JHull DRoberts DATE 7/18/05 7/18/05 7/22/05 8/3/05 8/11/05 OFFICIAL RECORD COPY

Vermont Yankee Nuclear Power Station cc:

Regional Administrator, Region I Ms. Carla A. White, RRPT, CHP U. S. Nuclear Regulatory Commission Radiological Health 475 Allendale Road Vermont Department of Health King of Prussia, PA 19406-1415 P.O. Box 70, Drawer #43 108 Cherry Street Mr. David R. Lewis Burlington, VT 05402-0070 Pillsbury, Winthrop, Shaw, Pittman, LLP 2300 N Street, N.W. Mr. James M. DeVincentis Washington, DC 20037-1128 Manager, Licensing Vermont Yankee Nuclear Power Station Ms. Christine S. Salembier, Commissioner P.O. Box 0500 Vermont Department of Public Service 185 Old Ferry Road 112 State Street Brattleboro, VT 05302-0500 Montpelier, VT 05620-2601 Resident Inspector Mr. Michael H. Dworkin, Chairman Vermont Yankee Nuclear Power Station Public Service Board U. S. Nuclear Regulatory Commission State of Vermont P.O. Box 176 112 State Street Vernon, VT 05354 Montpelier, VT 05620-2701 Director, Massachusetts Emergency Chairman, Board of Selectmen Management Agency Town of Vernon ATTN: James Muckerheide P.O. Box 116 400 Worcester Rd.

Vernon, VT 05354-0116 Framingham, MA 01702-5399 Operating Experience Coordinator Jonathan M. Block, Esq.

Vermont Yankee Nuclear Power Station Main Street 320 Governor Hunt Road P.O. Box 566 Vernon, VT 05354 Putney, VT 05346-0566 G. Dana Bisbee, Esq. Mr. John F. McCann Deputy Attorney General Director, Nuclear Safety Assurance 33 Capitol Street Entergy Nuclear Operations, Inc.

Concord, NH 03301-6937 440 Hamilton Avenue White Plains, NY 10601 Chief, Safety Unit Office of the Attorney General Mr. Gary J. Taylor One Ashburton Place, 19th Floor Chief Executive Officer Boston, MA 02108 Entergy Operations 1340 Echelon Parkway Ms. Deborah B. Katz Jackson, MS 39213 Box 83 Shelburne Falls, MA 01370

Vermont Yankee Nuclear Power Station cc:

Mr. John T. Herron Mr. Ronald Toole Sr. VP and Chief Operating Officer 1282 Valley of Lakes Entergy Nuclear Operations, Inc. Box R-10 440 Hamilton Avenue Hazelton, PA 18202 White Plains, NY 10601 Ms. Stacey M. Lousteau Mr. Danny L. Pace Treasury Department Vice President, Engineering Entergy Services, Inc.

Entergy Nuclear Operations, Inc. 639 Loyola Avenue 440 Hamilton Avenue New Orleans, LA 70113 White Plains, NY 10601 Mr. Raymond Shadis Mr. Brian OGrady New England Coalition Vice President, Operations Support Post Office Box 98 Entergy Nuclear Operations, Inc. Edgecomb, ME 04556 440 Hamilton Avenue White Plains, NY 10601 Mr. James P. Matteau Executive Director Mr. Michael J. Colomb Windham Regional Commission Director of Oversight 139 Main Street, Suite 505 Entergy Nuclear Operations, Inc. Brattleboro, VT 05301 440 Hamilton Avenue White Plains, NY 10601 Mr. William K. Sherman Vermont Department of Public Service Mr. John M. Fulton 112 State Street Assistant General Counsel Drawer 20 Entergy Nuclear Operations, Inc. Montpelier, VT 05620-2601 440 Hamilton Avenue White Plains, NY 10601 Mr. Jay K. Thayer Site Vice President Entergy Nuclear Operations, Inc.

Vermont Yankee Nuclear Power Station P.O. Box 0500 185 Old Ferry Road Brattleboro, VT 05302-0500 Mr. Kenneth L. Graesser 38832 N. Ashley Drive Lake Villa, IL 60046 Mr. James Sniezek 5486 Nithsdale Drive Salisbury, MD 21801

ENTERGY NUCLEAR VERMONT YANKEE, LLC AND ENTERGY NUCLEAR OPERATIONS, INC.

DOCKET NO. 50-271 VERMONT YANKEE NUCLEAR POWER STATION AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 226 License No. DPR-28

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

A. The application for amendment filed by Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, Inc. (the licensee) dated December 6, 2004, as supplemented on June 14, 2005, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions 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 Commissions 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 Commissions 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 3.B of Facility Operating License No. DPR-28 is hereby amended to read as follows:

(B) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 226, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of its date of issuance and shall be implemented within 30 days.

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

Darrell J. Roberts, Chief, Section 2 Project Directorate I Division of Licensing Project Management Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: August 15, 2005

ATTACHMENT TO LICENSE AMENDMENT NO. 226 FACILITY OPERATING LICENSE NO. DPR-28 DOCKET NO. 50-271 Replace the following pages of the Facility Operating License and 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.

Facility Operating License Remove Insert Page 3 Page 3 Technical Specifications Remove Insert Page 98 Page 98 Page 110 Page 110 Page 120 Page 120 Page 121 Page 121 Page 143 Page 143 Page 155a Page 155a Page 156 Page 156 Page 157 Page 157 Page 218 Page 218 Page 223 Page 223 Page 255 Page 255

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 226 TO FACILITY OPERATING LICENSE NO. DPR-28 ENTERGY NUCLEAR VERMONT YANKEE, LLC AND ENTERGY NUCLEAR OPERATIONS, INC.

VERMONT YANKEE NUCLEAR POWER STATION DOCKET NO. 50-271

1.0 INTRODUCTION

By letter dated December 6, 2004, as supplemented on June 14, 2005, Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, Inc. (Entergy or the licensee) submitted a request to amend the Vermont Yankee Nuclear Power Station (VYNPS) Facility Operating License (FOL) and Technical Specifications (TSs). The proposed amendment would make administrative and other miscellaneous changes to the FOL and the TSs including correction of references and deleting obsolete or redundant TS requirements and surveillances.

Specifically, the proposed amendment would make the following changes:

1) On page 3 of the FOL, License Condition 3.D, Records, would be deleted.

2) TS 4.7.C.1.a and TS 4.7.C.1.b would be deleted. Accordingly, TS 4.7.C.1.c would be renumbered to become TS 4.7.C.1.

3) TS 4.7.C.1.c (which would be renumbered as TS 4.7.C.1, as discussed above) would be revised to remove the current requirement that secondary containment be tested at each refueling outage prior to refueling.

4) TS 4.10.B.4 would be deleted.

5) TS 6.2.A.1 would be revised to change the title of the VYNPS quality assurance (QA) program manual.

6) TS 4.6.E.2 would be revised to correct a reference related to American Society of Mechanical Engineers (ASME) requirements for inservice testing (IST) of safety-related pumps and valves. Conforming changes would be made to the TS Bases for TS Sections 3.4/4.4, 3.5, 3.6/4.6, and 4.10.

7) TS 4.6.E.1 would not be changed; however, justification is provided to document a typographical error that was inadvertently corrected in a previous license amendment without markup or justification.

8) TS 3.7.C.5 and TS 4.7.C.5 would be deleted and the information would be relocated to the Updated Final Safety Analysis Report (UFSAR).

The supplement dated June 14, 2005, 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 or Commission) staff's original proposed no significant hazards consideration determination as published in the Federal Register on January 18, 2005 (70 FR 2888).

2.0 REGULATORY EVALUATION

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

(1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation (LCOs); (3) surveillance requirements (SRs); (4) design features; and (5) administrative controls. The regulation does not specify the particular requirements to be included in a plants TSs.

On July 22, 1993 (58 FR 39132), the Commission published a Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors (Final Policy Statement) which discussed the criteria to determine which items are required to be included in the TSs as LCOs. The criteria were subsequently incorporated into the regulations by an amendment to 10 CFR 50.36 (60 FR 36953). Specifically, 10 CFR 50.36(c)(2)(ii) requires that a TS LCO be established for each item meeting one or more of the following criteria:

Criterion 1: Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal 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.

In general, there are two classes of changes to TSs: (1) changes needed to reflect modifications to the design basis (TSs are derived from the design basis), and (2) changes to take advantage of the evolution in policy and guidance as to the required content and preferred format of TSs over time. In determining the acceptability of such changes, the NRC staff interprets the requirements of 10 CFR 50.36, using as a model the accumulation of

generically-approved guidance in the Improved Standard Technical Specifications (ISTS). For this review, the staff used NUREG-1433, Revision 3, Standard Technical Specifications, General Electric Plants BWR [boiling-water reactor]/4.

Within this general framework, licensees may remove material from their TSs if the material is not required to be in the TSs based on the NRC staffs interpretation of 10 CFR 50.36, including judgements about the level of detail required in the TSs. As discussed in the Final Policy Statement, the NRC staff reviews, on a case-by-case basis, whether enforceable regulatory controls are needed for the relocated material (e.g., 10 CFR 50.59). Licensees may revise the remaining TSs to adopt current ISTS format and content provided that plant-specific review supports a finding of continued adequate safety because: (1) the change is editorial, administrative, or provides clarification (i.e., no requirements are materially altered); (2) the change is more restrictive than the licensees current requirement; or (3) the change is less restrictive than the licensees current requirement, but nonetheless still affords adequate assurance of safety when judged against current regulatory standards.

3.0 TECHNICAL EVALUATION

3.1 License Condition 3.D Page 3 of the FOL, License Condition 3.D, Records, currently states Entergy Nuclear Operations, Inc. shall keep facility operating records in accordance with the requirements of the Technical Specifications. The licensee proposes to delete this license condition.

VYNPS Amendment No. 171, dated July 19, 1999, revised TS Section 6.0, Administrative Controls, to delete the sub-section pertaining to Plant Operating Records. The details contained in that TS section were relocated to the QA manual. The licensees request associated with Amendment No. 171 failed to identify that License Condition 3.D was linked to the TS requirements which were deleted (i.e., License Condition 3.D should have also been deleted at that time). The NRC staff concludes that since there are no longer any TS requirements associated with License Condition 3.D, the proposed change is administrative in nature and, therefore, the change is acceptable.

3.2 TS 4.7.C.1.a and TS 4.7.C.1.b TS 4.7.C.1.a and TS 4.7.C.1.b describe SRs for secondary containment testing that was required during the preoperational phase or during the first operating cycle for VYNPS. The licensee proposes to delete TS 4.7.C.1.a and TS 4.7.C.1.b since the testing has already been completed and the SRs no longer apply. The NRC staff concludes that the proposed change is acceptable since these TSs no longer provide any testing requirements applicable to current or future operation of VYNPS.

In addition, due to the deletion of TS 4.7.C.1.a and TS 4.7.C.1.b, the licensee proposed to renumber TS 4.7.C.1.c as TS 4.7.C.1. The NRC staff concludes that this change is administrative in nature and, therefore, is acceptable.

3.3 TS 4.7.C.1.c TS 4.7.C.1.c (which is to be renumbered as TS 4.7.C.1, as discussed above) currently requires that secondary containment be tested at least quarterly, and at each refueling outage prior to refueling. The licensee proposes to delete the requirement to test the secondary containment at each refueling outage prior to refueling.

The primary purpose of this SR is to demonstrate that secondary containment integrity is maintained and that the Standby Gas Treatment (SGT) System functions as designed.

NUREG-1433, SR 3.6.4.1.5, contains similar requirements as VYNPS TS 4.7.C.1. The frequency shown in NUREG-1433 for this SR is [18] months on a STAGGERED TEST BASIS for each SGT subsystem. The Bases for this SR state Operating experience has shown the

[secondary] containment boundary usually passes these Surveillance[s] when performed at the

[18] month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint. The NRC staff concludes that the existing quarterly test frequency in VYNPS TS 4.7.C.1 is sufficient to demonstrate secondary containment integrity. Therefore, the proposed change is acceptable.

3.4 TS 4.10.B.4 TS 4.10.B.4 currently reads: When it is determined that one Uninterruptible Power System or its associated Motor Control Center is inoperable, the requirements of Specification 4.5.A.4 shall be satisfied. The licensee proposes to delete TS 4.10.B.4 since TS 4.5.A.4 was deleted as part of VYNPS Amendment No. 209, issued August 14, 2002. The NRC staff concludes that, since TS 4.10.B.4 currently contains no requirements that can be implemented due to the previous deletion of TS 4.5.A.4, the proposed deletion of TS 4.10.B.4 is administrative in nature and, therefore, the change is acceptable.

3.5 TS 6.2.A.1 TS 6.2.A.1 provides requirements for onsite and offsite organizations associated with VYNPS operation and management. This TS currently reads, in part, that These requirements shall be documented in the Vermont Yankee Operational Quality Assurance Manual. The licensee proposes to change the title of the manual in TS 6.2.A.1 from Vermont Yankee Operational Quality Assurance Manual to Quality Assurance Program Manual. The change was necessitated by VYNPS transition from a site-specific QA program manual to an Entergy corporate QA program manual (reference Entergy letter BVY 05-014 dated March 2, 2005).

The NRC staff concludes that this change is administrative in nature and, therefore, is acceptable.

3.6 TS 4.6.E.2 and TS Bases for TS Sections 3.4/4.4, 3.5, 3.6/4.6, and 4.10.

TS 4.6.E.2 currently reads:

Operability testing of safety-related pumps and valves shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50, Section 50.55a(f), except where specific written relief has been granted by the NRC.

The licensee proposes to change TS 4.6.E.2 to read as follows:

Operability testing of safety-related pumps and valves shall be performed in accordance with the Code of Record as required by 10 CFR 50.55a, except where specific written relief has been granted by the NRC.

The licensees application provided the following basis for this change:

Technical Specification (TS) 4.6.E.2 currently requires Operability testing of safety related pumps and valves, also known as Inservice Testing (IST), to be performed in accordance with the requirements of Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50, Section 50.55a(f). The IST requirements were once part of ASME Section XI; however, this has since been relocated to a different ASME document. Specifically, the Code and Addenda currently required by 10 CFR 50.55a is actually the ASME Code for Operation and Maintenance of Nuclear Power Plants (OM Code). Accordingly, the proposed change would delete an incorrect reference and instead reference the Code of Record as required by 10 CFR 50, Section 50.55a. This change would not modify VYs IST program or the testing performed on safety related equipment, accordingly, the proposed change is considered administrative.

As discussed in an NRC Safety Evaluation (SE) dated December 22, 2003, VYNPS is currently in its fourth 10-year interval for the IST program. The interval began on September 1, 2003, and ends on August 31, 2013. The SE approved the use of the 1998 Edition through the 2000 Addenda of the ASME OM Code for VYNPSs fourth 10-year interval IST program. The NRC staff concludes that the proposed change to TS 4.6.E.2 is consistent with the VYNPS IST program and the requirements in 10 CFR 50.55a and, therefore, is acceptable.

The licensee has also proposed conforming changes to the TS Bases for TS Sections 3.4/4.4, 3.5, 3.6/4.6, and 4.10. The NRC staff has no objections to these changes.

3.7 TS 4.6.E.1 The second paragraph of TS 4.6.E.1 reads: Inservice inspection of piping, identified in NRC Generic Letter 88-01, shall be performed in accordance with the staff positions on schedule, methods, and personnel and sample expansion included in the Generic Letter or in accordance with alternate measures approved by NRC Staff. (emphasis added)

The typographical error in the word alternte was inadvertently corrected to alternate in VYNPS Amendment No. 219, dated April 14, 2004, without markup or justification. This license amendment will document the change. The staff concludes that this documentation of a typographical error is acceptable since this change is administrative and does affect the content of the TS. Note, the affected TS page (page 120) will be void of any revision bars since no actual change is being made to TS 4.6.E.1 at this time. The only change to the page will be a new amendment number.

3.8 TS 3.7.C.5 and TS 4.7.C.5 TS 3.7.C.5 currently requires that the core spray (CS) and low-pressure coolant injection (LPCI) lower compartment doors be closed at all times except during passage or when reactor coolant temperature is less than 212 <F. TS 4.7.C.5 requires the doors be checked closed daily. The licensee has proposed that these requirements be relocated to the VYNPS UFSAR.

The licensees application dated December 6, 2004, stated that the CS and LPCI lower compartment doors are related to flood protection and are not necessary for ensuring that secondary containment is operable (TS 3/4.7.C pertains to secondary containment integrity).

The licensees letter dated June 14, 2005, provided a response to an NRC request for additional information (RAI) to provide further justification for this proposed change. The licensees response to RAI question 1 (RAI 1) provided the following information regarding the design and licensing basis for the CS and LPCI lower compartment doors:

For VYNPS, the licensing basis requires the plant to achieve and maintain a hot safe shutdown condition even in the event of a rupture of a non-Class I system, component or pipe. Flooding which results from a rupture of any non-class I system drains to the sumps located in the lowest level (Elevation 213'-9") of the Reactor Building. If gross flooding were to occur, some of the water would collect in the torus area. From the torus area, the water would be pumped from the drain sumps to the floor drain collecting tank. Operating personnel would be alerted to the flooding of the torus area by the sounding of four drain sump high level alarms in the control room.

Each of the rooms at the northeast and southeast corners of the reactor building at the 213'-9" elevation contain a Core Spray and two Residual Heat Removal (RHR) pumps.

Each of these rooms has a watertight door installed at the entrance from the torus area.

The VYNPS Updated Final Safety Analysis Report, Section 4.8.5 states that theses (sic) rooms are isolated from the torus area by flood control doors so that a torus leak will not disable both RHR trains. Indication in the control room for internal flooding parameters includes alarms which monitor the normally closed watertight doors in the corner rooms.

The alarm system which monitors the position of the doors provides an audible alarm in the control room when either door is open. These alarms are installed to assure that equipment important to safety would not be damaged by flooding due to rupture of non-Class I components and/or pipes such that engineered safety features could not perform their design function; that is no single incident of a non-Class I system, component or pipe shall prevent safe shutdown of the facility.

Opening of these doors is controlled by VYNPS Administrative Procedure AP 0077, Barrier Control Process which provides administrative requirements for, but not limited to, breaching barriers and special features or blocking flooding paths at VYNPS.

The licensees response to RAI question 2 provided the following justification regarding the relocation of the TS 3.7.C.5 requirements based on the four criteria of 10 CFR 50.36(c)(2)(ii):

(1) The Core spray and LPCI pump lower compartment doors are related to flood protection and not the reactor coolant pressure boundary. The installed alarms discussed in response to RAI 1 are not used to detect, and indicate in the control

room, a significant abnormal degradation of the reactor coolant pressure boundary.

(2) The position of the Core Spray and LPCI pump lower compartment doors are not used as 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.

(3) The Core Spray and LPCI pump lower compartments are provided with watertight doors for the event of a torus rupture, not related to a design basis accident or transient. These doors are not used as part of the primary success path which functions or actuates to mitigate a design basis accident or transient.

(4) The watertight door TS LCO and Surveillance requirements, located in the Secondary Containment section of the TS, do not support operability of the secondary containment. They are a plant design feature that addresses concerns about a potential failure of the safety related torus impacting operation of redundant ECCS [emergency core cooling system] equipment. The failure of the torus and resulting flooding is not a design basis accident (DBA) or design basis transient (DBT) nor is the flooding event taken concurrent with a DBA or DBT. Requirements for protective design features, not related to design basis accidents or transients are typically described in the FSAR [final safety analysis report] and controlled under 10 CFR 50.59. Administrative controls will be maintained under 10 CFR 50.59 to ensure that the relocated requirements continue to be implemented at VYNPS. Entergy is not aware of any operating experience that would demonstrate that this design feature needs to be controlled in the TS nor is the equipment of the same safety significance as equipment that is included in the technical specifications that mitigates DBAs and DBTs.

The NRC staff finds that:

1) TS 3.7.C.5 does not meet the 10 CFR 50.36(c)(2)(ii) criteria for inclusion in the TSs based on the licensees justification presented above;

2) TSs 3.7.C.5 and 4.7.C.5, which are in the secondary containment section of the TSs, do not support operability of secondary containment based on the licensees description of the design and licensing basis for the lower compartment doors;

3) Similar flood control requirements do not exist in NUREG-1433.

4) The licensee has sufficient administrative controls in place to control the opening of the doors (i.e., Administrative Procedure AP 0077); and

5) The requirements of 10 CFR 50.59 provide the regulatory controls necessary for any future changes to the information being relocated to the UFSAR.

Based on the above five considerations, the NRC staff concludes that the proposed change is acceptable.

3.9 Technical Evaluation Conclusion Based on the considerations in SE Sections 3.1 through 3.8, the NRC staff concludes that the proposed amendment is acceptable.

4.0 STATE CONSULTATION

In accordance with the Commissions regulations, the Vermont 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 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 (70 FR 2888). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the 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 Commissions 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: P. Hearn R. Ennis Date: August 15, 2005