ML073340141

License Amendment, Issuance of Amendment Technical Specifications Regarding Control Room Envelope Habitability, Consistent with Technical Specification Task Force-448

ML073340141
Person / Time
Site:	FitzPatrick
Issue date:	01/03/2008
From:	Boska J NRC/NRR/ADRO/DORL/LPLI-1
To:	Balduzzi M Entergy Nuclear Operations
Boska J, NRR, 301-415-2901
References
TAC MD6136
Download: ML073340141 (20)

Text

January 3, 2008 Mr. Michael A. Balduzzi Sr. Vice President & COO Regional Operations, NE Entergy Nuclear Operations, Inc.

440 Hamilton Avenue White Plains, NY 10601

SUBJECT:

JAMES A. FITZPATRICK NUCLEAR POWER PLANT - ISSUANCE OF AMENDMENT RE: TECHNICAL SPECIFICATIONS REGARDING CONTROL ROOM ENVELOPE HABITABILITY, CONSISTENT WITH TECHNICAL SPECIFICATION TASK FORCE-448 (TAC NO. MD6136)

Dear Mr. Balduzzi:

The Commission has issued the enclosed Amendment No. 289 to Facility Operating License (FOL) No. DPR-59 for the James A. FitzPatrick Nuclear Power Plant. The amendment consists of changes to the FOL and the Technical Specifications (TS), in response to your application dated July 17, 2007, (Agencywide Document and Management System (ADAMS) accession no.

ML072050332) as supplemented on August 13, 2007 (ADAMS accession no. ML072400331) and supported by the FitzPatrick Generic Letter 2003-01 response dated September 27, 2004 (ADAMS accession no. ML042810225).

The amendment revises the FOL, the TS requirements related to control room envelope habitability in TS 3.7.3, Control Room Emergency Ventilation Air Supply System, TS Section 5.5, Programs and Manuals to adopt TSTF-448, Revision 3, Control Room Habitability, and the license conditions in Appendix C. This operating license improvement and TS revision was made available by the U.S. Nuclear Regulatory Commission on January 17, 2007, (72 FR 2022) 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,

/RA/

John P. Boska, Senior Project Manager Plant Licensing Branch I-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-333

Enclosures:

1. Amendment No. 289 to DPR-59

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

January 3, 2008 Mr. Michael A. Balduzzi Sr. Vice President & COO Regional Operations, NE Entergy Nuclear Operations, Inc.

440 Hamilton Avenue White Plains, NY 10601

SUBJECT:

JAMES A. FITZPATRICK NUCLEAR POWER PLANT - ISSUANCE OF AMENDMENT RE: TECHNICAL SPECIFICATIONS REGARDING CONTROL ROOM ENVELOPE HABITABILITY, CONSISTENT WITH TECHNICAL SPECIFICATION TASK FORCE-448 (TAC NO. MD6136)

Dear Mr. Balduzzi:

The Commission has issued the enclosed Amendment No. 289 to Facility Operating License (FOL) No. DPR-59 for the James A. FitzPatrick Nuclear Power Plant. The amendment consists of changes to the FOL and the Technical Specifications (TS), in response to your application dated July 17, 2007, (Agencywide Document and Management System (ADAMS) accession no.

ML072050332) as supplemented on August 13, 2007 (ADAMS accession no. ML072400331) and supported by the FitzPatrick Generic Letter 2003-01 response dated September 27, 2004 (ADAMS accession no. ML042810225).

The amendment revises the FOL, the TS requirements related to control room envelope habitability in TS 3.7.3, Control Room Emergency Ventilation Air Supply System, TS Section 5.5, Programs and Manuals to adopt TSTF-448, Revision 3, Control Room Habitability, and the license conditions in Appendix C. This operating license improvement and TS revision was made available by the U.S. Nuclear Regulatory Commission on January 17, 2007, (72 FR 2022) 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,

/RA/

John P. Boska, Senior Project Manager Plant Licensing Branch I-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-333

Enclosures:

1. Amendment No. 289 to DPR-59

2. Safety Evaluation cc w/enclosures: See next page Package No.: ML080080162 Amendment No.: ML073340141 Tech Spec No.: ML0

• SE input provided by memo OFFICE LPL1-1\\PM LPL3-1/LA LPL1-1\\LA ITSB\\BC LPL1-1\\BC NAME JPBoska BTully MOBrien for SLittle TKobetz*

DPickett for MKowal DATE 12/17/07 12/17/07 12/17/07 10/29/07 12/27/07 Official Record Copy

DATED: January 3, 2008 JAMES A. FITZPATRICK NUCLEAR POWER PLANT - ISSUANCE OF AMENDMENT RE:

TECHNICAL SPECIFICATIONS REGARDING CONTROL ROOM ENVELOPE HABITABILITY, CONSISTENT WITH TECHNICAL SPECIFICATION TASK FORCE-448 (TAC NO. MD6136)

PUBLIC LPL1-1 R/F RidsNrrDorl RidsNrrDorlLpl1-1 RidsOGCMailCenter GHill (2) (paper copies)

RidsNrrDirsItsb WCartwright, DIRS RidsAcrsAcnw&mMailCenter RidsNrrPMJBoska RidsNrrLASLittle (paper copy)

ECobey, RI RidsRgn1MailCenter RidsNrrDorlDpr TKobetz, DIRS cc: Plant Mailing list

FitzPatrick Nuclear Power Plant cc:

Mr. Michael R. Kansler President & CEO / CNO Entergy Nuclear Operations, Inc.

1340 Echelon Parkway Jackson, MS 39213 Mr. John T. Herron Sr. Vice President Entergy Nuclear Operations, Inc.

1340 Echelon Parkway Jackson, MS 39213 Sr. Vice President Engineering & Technical Services Entergy Nuclear Operations, Inc.

1340 Echelon Parkway Jackson, MS 39213 Mr. Peter T. Dietrich Site Vice President Entergy Nuclear Operations, Inc.

James A. FitzPatrick Nuclear Power Plant P.O. Box 110 Lycoming, NY 13093 Mr. Kevin J. Mulligan General Manager, Plant Operations Entergy Nuclear Operations, Inc.

James A. FitzPatrick Nuclear Power Plant P.O. Box 110 Lycoming, NY 13093 Mr. Oscar Limpias Vice President Engineering Entergy Nuclear Operations, Inc.

1340 Echelon Parkway Jackson, MS 39213 Mr. Joseph P. DeRoy Vice President, Operations Support Entergy Nuclear Operations, Inc.

1340 Echelon Parkway Jackson, MS 39213 Mr. John A. Ventosa GM, Engineering Entergy Nuclear Operations, Inc.

440 Hamilton Avenue White Plains, NY 10601 Mr. John F. McCann Director, Nuclear Safety & Licensing Entergy Nuclear Operations, Inc.

440 Hamilton Avenue White Plains, NY 10601 Ms. Charlene D. Faison Manager, Licensing Entergy Nuclear Operations, Inc.

440 Hamilton Avenue White Plains, NY 10601 Mr. Ernest J. Harkness Director, Oversight Entergy Nuclear Operations, Inc.

1340 Echelon Parkway Jackson, MS 39213 Mr. Michael J. Colomb Director of Oversight Entergy Nuclear Operations, Inc.

440 Hamilton Avenue White Plains, NY 10601 Director, Nuclear Safety Assurance Entergy Nuclear Operations, Inc.

James A. FitzPatrick Nuclear Power Plant P.O. Box 110 Lycoming, NY 13093 Mr. James Costedio Manager, Licensing Entergy Nuclear Operations, Inc.

James A. FitzPatrick Nuclear Power Plant P.O. Box 110 Lycoming, NY 13093

FitzPatrick Nuclear Power Plant cc:

Mr. William C. Dennis Assistant General Counsel Entergy Nuclear Operations, Inc.

440 Hamilton Avenue White Plains, NY 10601 Regional Administrator, Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Resident Inspector's Office James A. FitzPatrick Nuclear Power Plant U. S. Nuclear Regulatory Commission P.O. Box 136 Lycoming, NY 13093 Mr. Charles Donaldson, Esquire Assistant Attorney General New York Department of Law 120 Broadway New York, NY 10271 Mr. Paul Tonko President and CEO New York State Energy, Research, and Development Authority 17 Columbia Circle Albany, NY 12203-6399 Mr. John P. Spath New York State Energy, Research, and Development Authority 17 Columbia Circle Albany, NY 12203-6399 Mr. Paul Eddy New York State Dept. of Public Service 3 Empire State Plaza Albany, NY 12223-1350 Oswego County Administrator Mr. Steven Lyman 46 East Bridge Street Oswego, NY 13126 Supervisor Town of Scriba Route 8, Box 382 Oswego, NY 13126 Mr. James H. Sniezek BWR SRC Consultant 5486 Nithsdale Drive Salisbury, MD 21801-2490 Mr. Michael D. Lyster BWR SRC Consultant 5931 Barclay Lane Naples, FL 34110-7306 Mr. John Doering BWR SRC Consultant P.O. Box 189 Parker Ford, PA 19457

ENTERGY NUCLEAR FITZPATRICK, LLC AND ENTERGY NUCLEAR OPERATIONS, INC.

DOCKET NO. 50-333 JAMES A. FITZPATRICK NUCLEAR POWER PLANT AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 289 License No. DPR-59

1.

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

A.

The application for amendment by Entergy Nuclear Operations, Inc. (the licensee) dated July 13, 2007, as supplemented on August 13, 2007 and supported by the FitzPatrick Generic Letter 2003-01 response dated September 27, 2004, 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 and Additional Conditions in Appendix C as indicated in the attachments to this license amendment, and paragraphs 2.C.(2) and 2.F of Facility Operating License No. DPR-59 are hereby amended to read as follows:

(2)

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 289, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

F.

Additional Conditions The Additional Conditions contained in Appendix C, as revised through Amendment No. 289, are hereby incorporated into this license. ENO shall operate the facility in accordance with the Additional Conditions.

3.

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

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

Mark G. Kowal, Chief Plant Licensing Branch I-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the License and Technical Specifications Date of Issuance: January 3, 2008

ATTACHMENT TO LICENSE AMENDMENT NO. 289 FACILITY OPERATING LICENSE NO. DPR-59 DOCKET NO. 50-333 Replace the following pages of the License with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Page Insert Page 3

3 5

5 Replace the following pages of the 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 3.7.3-1 3.7.3-1 3.7.3-2 3.7.3-2 3.7.3-3 3.7.3-3 3.7.3-4 3.7.3-4 5.5-14 5.5-14 5.5-15 Replace the entire Appendix C Additional Conditions with the attached revised Appendix C. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Pages Insert Pages C-1 C-1 C-2

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 289 TO FACILITY OPERATING LICENSE NO. DPR-59 ENTERGY NUCLEAR OPERATIONS, INC.

JAMES A. FITZPATRICK NUCLEAR POWER PLANT DOCKET NO. 50-333

1.0 INTRODUCTION

By application dated July 17, 2007, (Agencywide Documents and Access Management System (ADAMS) Accession No. ML072050332) as supplemented by letter dated August 13, 2007, (ADAMS Accession No. ML072400331), and supported by the FitzPatrick Generic Letter 2003-01 response dated September 27, 2004, (reference 7), Entergy Nuclear Operations, Inc.

(the licensee) submitted a request for changes to the facility operating license (FOL) and the Technical Specifications (TS) for James A. FitzPatrick Nuclear Power Plant (JAF). The supplement, dated August 13, 2007, 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 as originally published in the Federal Register on September 11, 2007 (72 FR 51854).

The proposed amendment modifies the TS requirements related to control room envelope (CRE) habitability in TS 3.7.3, Control Room Emergency Ventilation Air Supply (CREVAS)

System, and adds a new TS 5.5.14, Control Room Envelope Habitability Program. In addition, the proposed amendment adds a license condition to Appendix C of the license. This license condition specifies the schedule for the initial performance of the new surveillance and assessment requirements for the Control Room Envelope Habitability Program. In addition, a typographical error in Appendix C of the license was corrected.

On August 8, 2006, the commercial nuclear electrical power generation industry owners group Technical Specifications Task Force (TSTF) submitted a proposed change, TSTF-448, Revision 3, to the improved standard technical specifications (STS) (NRC technical report designation (NUREGs) 1430-1434) on behalf of the industry (TSTF-448, Revisions 0, 1, and 2 were prior draft iterations). TSTF-448, Revision 3, is a proposal to establish more effective and appropriate action, surveillance, and administrative STS requirements related to ensuring the habitability of the control room envelope.

In NRC Generic Letter 2003-01 (Reference 1), licensees were alerted to findings at facilities that existing TS surveillance requirements for the Control Room Envelope Emergency Ventilation System (CREEVS) may not be adequate. Specifically, the results of American Society for Testing and Materials (ASTM) E741 (Reference 2) tracer gas tests to measure CRE unfiltered inleakage at facilities indicated that the differential pressure surveillance is not a reliable method for demonstrating CRE boundary operability. Licensees were requested to address existing TS as follows:

Provide confirmation that your technical specifications verify the integrity [i.e., operability]

of the CRE [boundary], and the assumed [unfiltered] inleakage rates of potentially contaminated air. If you currently have a differential pressure surveillance requirement to demonstrate CRE [boundary] integrity, provide the basis for your conclusion that it remains adequate to demonstrate CRE integrity in light of the ASTM E741 testing results.

If you conclude that your differential pressure surveillance requirement is no longer adequate, provide a schedule for: 1) revising the surveillance requirement in your technical specification to reference an acceptable surveillance methodology (e.g., ASTM E741), and 2) making any necessary modifications to your CRE [boundary] so that compliance with your new surveillance requirement can be demonstrated.

If your facility does not currently have a technical specification surveillance requirement for your CRE integrity, explain how and at what frequency you confirm your CRE integrity and why this is adequate to demonstrate CRE integrity.

To promote standardization and to minimize the resources that would be needed to create and process plant specific amendment applications in response to the concerns described in the generic letter, the industry and the NRC proposed revisions to CRE habitability system requirements contained in the STS, using the STS change traveler process. This effort culminated in Revision 3 to traveler TSTF-448, Control Room Habitability, which the NRC staff approved on January 17, 2007.

Consistent with the traveler as incorporated into NUREG-1434, the licensee proposed revising action and surveillance requirements in Specification 3.7.3, Control Room Emergency Ventilation Air Supply System, and adding a new administrative controls program, Specification 5.5.14, Control Room Envelope Habitability Program. The purpose of the changes is to ensure that CRE boundary operability is maintained and verified through effective surveillance and programmatic requirements, and that appropriate remedial actions are taken in the event of an inoperable CRE boundary.

Some editorial and plant specific changes were incorporated into this safety evaluation resulting in minor deviations from the model safety evaluation text in TSTF-448, Revision 3.

2.0 REGULATORY EVALUATION

2.1 Control Room and Control Room Envelope NRC Regulatory Guide 1.196, Control Room Habitability at Light-Water Nuclear Power Reactors, Revision 0, May 2003, (Reference 4) uses the term control room envelope in addition to the term control room and defines each term as follows:

Control Room: The plant area, defined in the facility licensing basis, in which actions can be taken to operate the plant safely under normal conditions and to maintain the reactor in a safe condition during accident situations. It encompasses the instrumentation and controls necessary for a safe shutdown of the plant and typically includes the critical document reference file, computer room (if used as an integral part of the emergency response plan), shift supervisor's office, operator wash room and kitchen, and other critical areas to which frequent personnel access or continuous occupancy may be necessary in the event of an accident.

Control Room Envelope (CRE): The plant area, defined in the facility licensing basis that in the event of an emergency, can be isolated from the plant areas and the environment external to the CRE. This area is served by an emergency ventilation system, with the intent of maintaining the habitability of the control room. This area encompasses the control room, and may encompass other non-critical areas to which frequent personnel access or continuous occupancy is not necessary in the event of an accident.

NRC Regulatory Guide 1.197, Demonstrating Control Room Envelope Integrity At Nuclear Power Reactors, Revision 0, May 2003, (Reference 5), also contains these definitions, but uses the term CRE to mean both. This is because the protected environment provided for operators varies with the nuclear power facility. At some facilities this environment is limited to the control room; at others, it is the CRE. In this safety evaluation, consistent with the proposed changes to the STS, the CRE will be used to designate both. For consistency, facilities should use the term CRE with an appropriate facility specific definition derived from the above CRE definition.

2.2 Control Room Emergency Ventilation Air Supply System The CREVAS system (the term used at JAF for the Control Room Envelope Emergency Ventilation System, CREEVS) provides a protected environment from which operators can control the unit, during airborne challenges from radioactivity, hazardous chemicals, and fire byproducts, such as fire suppression agents and smoke, during both normal and accident conditions.

The CREVAS system is designed to maintain a habitable environment in the control room envelope for 31 days of continuous occupancy after a Design Basis Accident (DBA) without exceeding a 5 rem whole body dose or its equivalent to any part of the body.

The safety-related function of the CREVAS system consists of two redundant subsystems, each capable of maintaining the habitability of the CRE. The CREVAS system is considered operable when the individual components necessary to limit operator exposure are operable in both subsystems. A CREVAS system subsystem is considered operable when the associated:

$ Fans are operable;

$ Pre-Filter, two high-efficiency particulate air filters and charcoal adsorbers are not excessively restricting flow, and are capable of performing their filtration functions;

$ Ductwork, valves, and dampers are operable, and air circulation can be maintained; and

$ CRE boundary is operable (the single boundary supports both subsystems).

The CRE boundary is considered operable when the measured unfiltered air inleakage is less than or equal to the inleakage value assumed by the licensing basis analyses of design basis accident consequences to CRE occupants.

2.3 Regulations Applicable to Control Room Habitability In Appendix A, General Design Criteria for Nuclear Power Plants, to 10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities, General Design Criteria (GDC) 1, 2, 3, 4, 5, and 19 apply to CRE habitability. Although the construction permit for JAF was issued prior to the publication of the Appendix A GDCs, the Updated Final Safety Analysis Report for JAF contains similar design criteria1. A summary of the applicable Appendix A GDCs follows.

GDC 1, Quality Standards and Records, requires that structures, systems, and components (SSCs) important to safety be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions performed.

GDC 2, Design Basis for Protection Against Natural Phenomena, requires that SSCs important to safety be designed to withstand the effects of earthquakes and other natural hazards.

GDC 3, Fire Protection, requires that SSCs important to safety be designed and located to minimize the effects of fires and explosions.

GDC 4, Environmental and Dynamic Effects Design Bases, requires SSCs important to safety to be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, including loss of coolant accidents (LOCAs).

GDC 5, Sharing of Structures, Systems, and Components, requires that SSCs important to safety not be shared among nuclear power units unless it can be shown that such sharing will not significantly impair their ability to perform their safety functions, including, in the event of an accident in one unit, the orderly shutdown and cooldown of the remaining units.

GDC 19, Control Room, requires that a control room be provided from which actions can be taken to operate the nuclear reactor safely under normal conditions and to maintain the reactor in a safe condition under accident conditions, including a LOCA. Adequate radiation protection 1 The following explains the use of GDC for JAF. The construction permit for JAF was issued by the Atomic Energy Commission (AEC) on May 20, 1970, and the operating license was issued on October 17, 1974. The plant design criteria for the construction phase are listed in the Updated Final Safety Analysis Report (UFSAR) Chapter 1.5, "Principal Design Criteria." The AEC published the final rule that added Appendix A to Title 10 of the Code of Federal Regulations (10 CFR) Part 50, "General Design Criteria for Nuclear Power Plants," in the Federal Register (36 FR 3255) on February 20, 1971, with the rule effective on May 21, 1971. In accordance with an NRC staff requirements memorandum from S. J. Chilk to J. M. Taylor, "SECY-92-223 - Resolution of Deviations Identified During the Systematic Evaluation Program," dated September 18, 1992 (ADAMS Accession No. ML003763736), the Commission decided not to apply the final GDC to plants with construction permits issued prior to May 21, 1971, which includes JAF. However, the JAF UFSAR, Chapter 16.6, "Conformance to AEC Design Criteria," evaluates JAF against the 10 CFR 50 Appendix A GDC. Also, the initial AEC safety evaluation of JAF, dated November 20, 1972, Chapter 14.0, stated "Based on our evaluation of the design and design criteria for the James A. FitzPatrick Nuclear Power Plant, we conclude that there is reasonable assurance that the intent of the General Design Criteria for Nuclear Power Plants, published in the Federal Register on May 21, 1971 as Appendix A to 10 CFR part 50, will be met." Therefore, the NRC staff reviews amendments to the JAF license using the 10 CFR 50 Appendix A GDC unless there are specific criteria identified in the UFSAR.

is to be provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of specified values.

Prior to incorporation of TSTF-448, Revision 3, the STS requirements addressing CRE boundary operability resided only in the following CRE ventilation system specifications:

$ NUREG-1430, TS 3.7.10, Control Room Emergency Ventilation System;

$ NUREG-1431, TS 3.7.10, Control Room Emergency Filtration System;

$ NUREG-1432, TS 3.7.11, Control Room Emergency Air Cleanup System;

$ NUREG-1433, TS 3.7.4, Main Control Room Environmental Control System; and

$ NUREG-1434, TS 3.7.3, Control Room Fresh Air (CRFA) System.

In these specifications, the surveillance requirement associated with demonstrating the operability of the CRE boundary requires verifying that one CRFA subsystem can maintain a positive pressure relative to the areas adjacent to the CRE during the pressurization mode of operation at a makeup flow rate. Facilities that pressurize the CRE during the emergency mode of operation of the CRFA system have similar surveillance requirements (SR). Other facilities that do not pressurize the CRE have only a system flow rate criterion for the emergency mode of operation. Regardless, the results of ASTM E741 (Reference 2) tracer gas tests to measure CRE unfiltered inleakage at facilities indicated that the differential pressure surveillance (or the alternative surveillance at non-pressurization facilities) is not a reliable method for demonstrating CRE boundary operability. That is, licensees were able to obtain differential pressure and flow measurements satisfying the SR limits even though unfiltered inleakage was determined to exceed the value assumed in the safety analyses.

In addition to an inadequate SR, the action requirements of these specifications were ambiguous regarding CRE boundary operability in the event CRE unfiltered inleakage is found to exceed the analysis assumption. The ambiguity stemmed from the view that the CRE boundary may be considered operable but degraded in this condition, and that it would be deemed inoperable only if calculated radiological exposure limits for CRE occupants exceeded a licensing basis limit; e.g., as stated in GDC 19, even while crediting compensatory measures.

NRC Administrative Letter (AL) 98-10, Dispositioning of Technical Specifications That Are Insufficient to Assure Plant Safety, (AL 98-10) states that the discovery of an improper or inadequate TS value or required action is considered a degraded or nonconforming condition, which is defined in NRC Inspection Manual Chapter 9900; see latest guidance in Regulatory Issue Summary (RIS) 2005-20 (Reference 3). Imposing administrative controls in response to improper or inadequate TS is considered an acceptable short term corrective action. The NRC staff expects that, following the imposition of administrative controls, an amendment to the inadequate TS, with appropriate justification and schedule, will be submitted in a timely fashion.

Licensees that have found unfiltered inleakage in excess of the limit assumed in the safety analyses and have yet to either reduce the inleakage below the limit or establish a higher bounding limit through reanalysis, have implemented compensatory actions to ensure the safety of CRE occupants, pending final resolution of the condition, consistent with RIS 2005-20.

However, based on Generic Letter (GL) 2003-01 and AL 98-10, the NRC staff expects each licensee to propose TS changes that include a surveillance to periodically measure CRE unfiltered inleakage in order to satisfy 10 CFR 50.36(d)(3), which requires a facility's TS to include surveillance requirements, which it defines as 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 limiting conditions for operation will be met. (Emphasis added.)

The NRC staff also expects facilities to propose unambiguous remedial actions, consistent with 10 CFR 50.36(d)(2), for the condition of not meeting the limiting condition for operation (LCO) due to an inoperable CRE boundary. The action requirements should specify a reasonable completion time to restore conformance to the LCO before requiring a facility to be shut down.

This completion time should be based on the benefits of implementing mitigating actions to ensure CRE occupant safety and sufficient time to resolve most problems anticipated with the CRE boundary, while minimizing the chance that operators in the CRE will need to use mitigating actions during accident conditions.

2.4 Adoption of TSTF-448, Revision 3, by James A. FitzPatrick Nuclear Power Plant Adoption of TSTF-448, Revision 3, will assure that the facility's TS LCO for the CREVAS system is met by demonstrating unfiltered leakage into the CRE is within limits; i.e., the operability of the CRE boundary. In support of this surveillance, which specifies a test interval (frequency) described in Regulatory Guide 1.197, TSTF-448 also adds TS administrative controls to assure the habitability of the CRE between performances of the ASTM E741 test. In addition, adoption of TSTF-448 will establish clearly stated and reasonable required actions in the event CRE unfiltered inleakage is found to exceed the analysis assumption.

The changes made by TSTF-448 to the STS requirements for the CRFA system and the CRE boundary conform to 10 CFR 50.36(d)(2) and 10 CFR 50.36(d)(3). Their adoption will better assure that Fitzpatrick's CRE will remain habitable during normal operation and design basis accident conditions. These changes are, therefore, acceptable from a regulatory standpoint.

3.0 TECHNICAL EVALUATION

The NRC staff reviewed the proposed changes against the corresponding changes made to the STS by TSTF-448, Revision 3, which the NRC staff has found to satisfy applicable regulatory requirements, as described above in Section 2.0. The emergency operational mode of the CREVAS system at the James A Fitzpatrick Nuclear Power Plant pressurizes the CRE to minimize unfiltered air inleakage. The proposed changes are consistent with this design.

3.1 Proposed Changes The proposed amendment would strengthen CRE habitability TS requirements by changing TS 3.7.3, Control Room Emergency Ventilation Air Supply (CREVAS) System, and adding a new TS administrative controls program on CRE habitability. Accompanying the proposed TS changes are appropriate conforming technical changes to the TS Bases. The proposed revision to the Bases also includes editorial and administrative changes to reflect applicable changes to the corresponding STS Bases, which were made to improve clarity, conform to the latest information and references, correct factual errors, and achieve more consistency among the STS NUREGs. Except for plant-specific differences, all of these changes are consistent with STS as revised by TSTF-448, Revision 3.

The NRC staff compared the proposed TS changes to the STS and the STS markups and evaluations in TSTF-448. The staff verified that differences from the STS were adequately justified on the basis of plant specific design or retention of current licensing basis. The NRC staff also reviewed the proposed changes to the TS Bases for consistency with the STS Bases and the plant specific design and licensing bases, although approval of the Bases is not a condition for accepting the proposed amendment. However, TS 5.5.11, TS Bases Control Program, provides assurance that the licensee has established and will maintain the adequacy of the Bases. The proposed Bases for TS 3.7.3 refer to specific guidance in Nuclear Energy Institute (NEI) 99-03, Control Room Habitability Assessment Guidance, Revision 0, dated June 2001 (Reference 6), which the NRC staff has formally endorsed, with exceptions, through Regulatory Guide 1.196, Control Room Habitability at Light-Water Nuclear Power Reactors, dated May 2003 (Reference 4).

3.2 Editorial Changes The licensee proposed editorial changes to TS 3.7.3, Control Room Emergency Ventilation Air Supply (CREVAS) System, to establish standard terminology, such as control room envelope (CRE) in place of control room, except for the plant-specific name for the CREEVS (Control Room Emergency Ventilation Air Supply (CREVAS) System), and radiological, chemical, and smoke hazards in place of various phrases to describe the hazards that CRE occupants are protected from by the CRFA System. These changes improve the usability and quality of the presentation of the TS, have no impact on safety, and therefore, are acceptable.

In addition to the above editorial changes, a typographical error in Appendix C of the license is being corrected. Under the additional conditions specified for Amendment No. 243, the reference should have stated the provisions of 10 CFR 50.59, rather than the provisions of 10 CFR 50.50. The NRC inadvertently introduced this typographical error when the NRC issued this page for Amendment 268. This change merely corrects a typographical error, and has no impact on safety, and therefore, is acceptable.

3.3 TS 3.7.3, Control Room Emergency Ventilation Air Supply (CREVAS) System Note: Only Evaluations 1, 5, and 6 of the Model Safety Evaluation published in the Federal Register on January 17, 2007 (72 FR 2022) are applicable to the James A. FitzPatrick Nuclear Power Plant.

Evaluation 1 The licensee proposed to revise the action requirements of TS 3.7.3, CREVAS, to acknowledge that an inoperable CRE boundary, depending upon the location of the associated degradation, could cause just one, instead of both CREVAS subsystems to be inoperable. This is accomplished by revising Condition A to exclude Condition B, and revising Condition B to address one or more CREVAS subsystems as follows:

$ Condition A One CREVAS subsystem inoperable for reasons other than Condition B.

$ Condition B One or more CREVAS subsystems inoperable due to inoperable CRE boundary in MODE 1, 2, or 3.

This change clarifies how to apply the action requirements in the event just one CREVAS subsystem is unable to ensure CRE occupant safety within licensing basis limits because of an inoperable CRE boundary. It enhances the usability of Conditions A and B with a presentation that is more consistent with the intent of the existing requirements. This change is an administrative change because it neither reduces nor increases the existing action requirements, and, therefore, is acceptable.

The licensee proposed to replace existing Required Action B.1, Restore control room boundary to OPERABLE status, which has a 24-hour Completion Time, with Required Action B.1, to immediately initiate action to implement mitigating actions: Required Action B.2, to verify, within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, that in the event of a DBA, CRE occupant exposures to radiological, chemical, and smoke hazards will not exceed limits; and Required Action B.3, to restore CRE boundary to operable status within 90 days.

The 24-hour Completion Time of new Required Action B.2 is reasonable based on the low probability of a DBA occurring during this time period, and the use of mitigating actions as directed by Required Action B.1. The 90-day Completion Time of new Required Action B.3 is reasonable based on the determination that the mitigating actions will ensure protection of CRE occupants within analyzed limits while limiting the probability that CRE occupants will have to implement protective measures that may adversely affect their ability to control the reactor and maintain it in a safe shutdown condition in the event of a DBA. The 90-day Completion Time is a reasonable time to diagnose, plan and possibly repair, and test most anticipated problems with the CRE boundary. Therefore, proposed Actions B.1, B.2, and B.3 are acceptable.

Evaluation 5 The licensee proposed to add a new condition to Action F of TS 3.7.3 that states, One or more CREVAS subsystems inoperable due to an inoperable CRE boundary during movement of recently irradiated fuel assemblies in the secondary containment or during operations with a potential for draining the reactor vessel (OPDRVs). The specified Required Actions proposed for this condition are the same as for the other existing condition for Action F, which states, Two CREVAS subsystems inoperable during movement of recently irradiated fuel assemblies in the secondary containment, or during OPDRVs. Accordingly, the new condition is stated with the other condition in Action F using the logical connector OR. The practical result of this presentation in format is the same as specifying two separately numbered Actions, one for each condition. Its advantage is to make the TS Actions table easier to use by avoiding having an additional numbered row in the Actions table. This new condition in Action F is needed because proposed Action B will only apply in Modes 1, 2, and 3. As such, this change will ensure that the Actions table continues to specify a condition for an inoperable CRE boundary during refueling and OPDRVs. Therefore, this change is administrative and acceptable.

Evaluation 6 In the emergency mode of operation, the CREVAS system isolates unfiltered ventilation air supply intakes, filters the emergency ventilation air supply to the CRE, and pressurizes the CRE to minimize unfiltered air inleakage past the CRE boundary. The licensee proposed to delete the CRE pressurization SR. This SR requires verifying that one CREVAS subsystem, operating in the emergency mode, can maintain a pressure of 0.125 inches water gauge, relative to the adjacent outside atmosphere during the pressurization mode of operation at a makeup flow rate between 900 standard cubic feet per minute (scfm) and 1100 scfm. The deletion of this SR is proposed because measurements of unfiltered air leakage into the CRE at numerous reactor facilities demonstrated that a basic assumption of this SR, an essentially leak tight CRE boundary, was incorrect for most facilities. Hence, meeting this SR by achieving the required CRE pressure is not necessarily a conclusive indication of CRE boundary leak tightness, i.e.,

CRE boundary operability. In its response to GL 2003-01, dated September 27, 2004, (ADAMS Accession No. ML042810225), the licensee reported that it had determined that the JAF CRE pressurization surveillance, SR 3.7.3.3, was an accurate predictor of unfiltered air inleakage to the CRE. The licensee did agree in their response to a request for additional information on their GL 2003-01 response (ADAMS Accession No. ML070390379) to replace their CRE surveillance tests with an inleakage measurement SR and a CRE Habitability Program in TS Section 5.5, in accordance with the approved version of TSTF-448. Based on the adoption of TSTF-448, Revision 3, the licensee's proposal to delete the current SR 3.7.3.3 is acceptable.

The proposed CRE inleakage measurement SR (proposed SR 3.7.3.3) states, Perform required CRE unfiltered air in leakage testing in accordance with the Control Room Envelope Habitability Program. The CRE Habitability Program (proposed TS 5.5.14) includes Requirements for determining the unfiltered air inleakage past the CRE boundary into the CRE in accordance with the testing methods and at the frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, Revision 0 (Reference 5). This guidance references ASTM E741 (Reference 2) as an acceptable method for ascertaining the unfiltered leakage into the CRE. The licensee has proposed to follow this method. Therefore, the proposed CRE inleakage measurement SR is acceptable.

3.4 TS 5.5.14, Control Room Envelope Habitability Program The proposed administrative controls program TS is consistent with the model program TS in TSTF-448, Revision 3. In combination with SR 3.7.3.3, this program is intended to ensure the operability of the CRE boundary, which as part of an operable CREVAS system will ensure that CRE habitability is maintained such that CRE occupants can control the reactor safely under normal conditions and maintain it in a safe condition following a radiological event, hazardous chemical release, or a smoke challenge. The program shall ensure that adequate radiation protection is provided to permit access and occupancy of the CRE under DBA conditions without personnel receiving radiation exposures in excess of 5 rem whole body dose or its equivalent to any part of the body for the duration of the accident.

A CRE Habitability Program TS acceptable to the NRC staff requires the program to contain the following elements:

Definitions of CRE and CRE boundary: This element is intended to ensure that these definitions accurately describe the plant areas that are within the CRE, and also the interfaces that form the CRE boundary, and are consistent with the general definitions discussed in Section 2.1 of this safety evaluation. Establishing what is meant by the CRE and the CRE boundary will preclude ambiguity in the implementation of the program.

Configuration control and preventive maintenance of the CRE boundary: This element is intended to ensure the CRE boundary is maintained in its design condition. Guidance for implementing this element is contained in Regulatory Guide 1.196 (Reference 4), which endorsed, with exceptions, NEI 99-03 (Reference 6). Maintaining the CRE boundary in its design condition provides assurance that its leak-tightness will not significantly degrade between CRE inleakage determinations.

Assessment of CRE habitability at the frequencies stated in Sections C.1 and C.2 of Regulatory Guide 1.197, Revision 0 (Reference 5), and measurement of unfiltered air leakage into the CRE in accordance with the testing methods and at the frequencies stated in Sections C.1 and C.2 of Regulatory Guide 1.197: This element is intended to ensure that the plant assesses CRE habitability consistent with Sections C.1 and C.2 of Regulatory Guide 1.197. Assessing CRE habitability at the NRC accepted frequencies provides assurance that significant degradation of the CRE boundary will not go undetected between CRE inleakage determinations.

Determination of CRE inleakage using test methods acceptable to the NRC staff assures that test results are reliable for ascertaining CRE boundary operability. Determination of CRE inleakage at the NRC accepted frequencies provides assurance that significant degradation of the CRE boundary will not occur between CRE inleakage determinations.

Measurement of CRE pressure with respect to all areas adjacent to the CRE boundary at designated locations for use in assessing the CRE boundary at a frequency of 18 months on a staggered test basis (with respect to the CREVAS subsystems): This element is intended to ensure that CRE differential pressure is regularly measured to identify changes in pressure warranting evaluation of the condition of the CRE boundary. Obtaining and trending pressure data provides additional assurance that significant degradation of the CRE boundary will not go undetected between CRE inleakage determinations.

Quantitative limits on unfiltered inleakage: This element is intended to establish the CRE inleakage limit as the CRE unfiltered infiltration rate assumed in the CRE occupant radiological consequence analyses of design basis accidents. Having an unambiguous criterion for the CRE boundary to be considered operable in order to meet LCO 3.7.3, will ensure that associated action requirements will be consistently applied in the event of CRE degradation resulting in inleakage exceeding the limit.

Consistent with TSTF-448, Revision 3, the program states that the provisions of SR 3.0.2 are applicable to the program frequencies for performing the activities required by program paragraph number c, parts (i) and (ii) (assessment of CRE habitability and measurement of CRE inleakage), and paragraph number d (measurement of CRE differential pressure). This statement is needed to avoid confusion. SR 3.0.2 is applicable to the surveillance that references the testing in the CRE Habitability Program. However, SR 3.0.2 is not applicable to Administrative Controls unless specifically invoked. Providing this statement in the program eliminates any confusion regarding whether SR 3.0.2 is applicable, and is acceptable.

Consistent with TSTF-448, Revision 3, proposed TS 5.5.14 states that (1) a CRE Habitability Program shall be established and implemented, (2) the program shall include all of the NRC staff required elements, as described above, and (3) the provisions of SR 3.0.2 shall apply to program frequencies. Therefore, TS 5.5.14, which is consistent with the model program TS approved by the NRC staff in TSTF-448, Revision 3, is acceptable.

3.5 Implementation of New Surveillance and Assessment Requirements by the Licensee The licensee has proposed license conditions regarding the initial performance of the new surveillance and assessment requirements. The new license conditions adopted the conditions in section 2.3 of the model application published in the Federal Register on January 17, 2007, (72 FR 2022). Plant-specific changes were made to the proposed license conditions. The proposed plant-specific license conditions are consistent with the model application, and are acceptable.

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 the 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 considerations, and there has been no public comment on the finding issued on September 11, 2007 (72 FR 51854). 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 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.

7.0 REFERENCES

1. NRC Generic Letter 2003-01, Control Room Habitability, dated June 12, 2003, (GL 2003-01).

2. ASTM E741 - 00, Standard Test Method for Determining Air Change in a Single Zone by Means of a Tracer Gas Dilution, 2000, (ASTM E741).

3. NRC Regulatory Issue Summary 2005-20: Revision to Guidance Formerly Contained in NRC Generic Letter 91-18, Information to Licensees Regarding Two NRC Inspection Manual Sections on Resolution of Degraded and Nonconforming Conditions and on Operability, dated September 26, 2005 (RIS 2005-20).

4. Regulatory Guide 1.196, Control Room Habitability at Light Water Nuclear Power Reactors, Revision 0, dated May 2003.

5. Regulatory Guide 1.197, Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors, Revision 0, May 2003.

6. NEI 99-03, Control Room Habitability Assessment Guidance, Revision 0, dated June 2001.

7. Entergy Letter NRC Generic Letter 2003-01 Control Room Habitability, Initial Summary Actions Report, dated September 27, 2004 (ADAMS Accession No. ML042810225)

Principal contributor: W. Cartwright, NRR/DIRS Date: January 3, 2008