ML14237A045

Issuance of Amendment No. 215, Revise Operating License and Technical Specifications to the Permanently Defueled Technical Specifications Consistent with Permanent Cessation of Reactor Operation

ML14237A045
Person / Time
Site:	Kewaunee
Issue date:	02/13/2015
From:	Wengert T J Plant Licensing Branch IV
To:	Heacock D A Dominion Energy Kewaunee
Huffman W
References
TAC MF1952
Download: ML14237A045 (113)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 Mr. David A. Heacock President and Chief Nuclear Officer Dominion Energy Kewaunee, Inc. Innsbrook Technical Center 5000 Dominion Boulevard Glen Allen, VA 23060-6711 February 13, 2015

SUBJECT:

KEWAUNEE POWER STATION -ISSUANCE OF AMENDMENT FOR PERMANENTLY SHUTDOWN AND DEFUELED TECHNICAL SPECIFICATIONS AND CERTAIN LICENSE CONDITIONS (TAC NO. MF1952)

Dear Mr. bleacock:

The U.S. Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment No. 215 to Renewed Facility Operating License No. DPR'"43 for the Kewaunee Power Station (KPS). The amendment revises the Renewed Facility Operating License and associated technical specifications (TSs) to conform to permanent shutdown defueled status of KPS. The amendment is in response to your application dated May 29, 2013, as supplemented by letters dated September 23, October 15, October 17, October 31, and November 7, 2013, and January 7, March 13, April 29, and October 6, 2014, and January 15, 2015. In the original May 29, 2013, amendment request, as supplemented, Dominion Energy Kewaunee, Inc. (DEK), proposed to modify the operating TSs at KPS to be consistent with the decommissioning (i.e:, permanently shutdown and defueled) status of the reactor. In general, the changes proposed by the original request eliminated those TSs applicable in operating modes or modes where fuel is em placed in the reactor vessel. Changes to other TS limiting conditions for operations, definitions, administrative controls, as well as several license conditions were also proposed.

In its March 13, 2014, supplemental letter, DEK stated that it had accelerated its schedule to transfer spent fuel from the KPS spent fuel pool to the KPS independent spent fuel storage installation.

Based on its new schedule, DEK requested expedited review and approval to delete certain TSs no longer applicable during the movement of irradiated fuel assemblies.

By letter dated June 9, 2014, the NRC issued Amendment No. 212 to the KPS Renewed Facility Operating License DPR-43. The amendment approved deletion of certain TSs no longer applicable to the movement of irradiated fuel assemblies, in response to DEK's March 13, 2014, supplemental request.

D. Heacock The NRC staff stated in its June 9, 2014, letter to DEK approving the license amendment that the remaining changes to the KPS TSs and license conditions are still under review. The staff has completed its evaluation of the proposed changes to the permanently defueled TSs as requested in DEK's letter dated May 29, 2013, as supplemented, and has included the staff's findings in the enclosed safety evaluation.

The May 29, 2013, amendment request, as supplemented, also proposed changes to the KPS

• license that DEK stated were either clarifications, redundant with other requirements, or no longer applicable based on the permanently shutdown and defueled status of KPS. The NRC staff has completed its review and assessment of the proposed license changes, as provided in the enclosed safety evaluation.

This amendment approves most of the proposed license changes requested by the licensee.

It also denies several of the proposed license changes. Specifically, the staff denies the proposed changes to Commission findings 1.B, 1.1, and 1.J of the license as discussed in Section 3.8.1 of the enclosed safety evaluation.

In addition, the NRC approves in part and denies in part the change proposed by DEK to license condition 2.B.(1) as discussed in Section 3.8.2 of the enclosed safety evaluation.

The Notice of Issuance will be included in the Commission's biweekly Federal Register notice. Sincerely, Docket No. 50-305

Enclosures:

1. Amendment No. 215 to Renewed Facility Operating License No. DPR-43 2. Safety Evaluation cc w/encls: Distribution via ListServ Wengert, Senior P

• ct Manager Plant Licensing IV-2 and Decommissioning Transition Branch Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

.) UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 DOMINION ENERGY KEWAUNEE.

INC. DOCKET NO. 50-305 KEWAUNEE POWER STATION AMENDMENT TO RENEWED FACILITY OPERATING LICENSE AND TECHNICAL SPECIFICATIONS Amendment No. 215 License No. DPR-43 1. The U.S. Nuclear Regulatory Commission (the Commission) has found that: A. The application for amendment by Dominion Energy Kewaunee, Inc., dated May 29, 2013, as supplemented by letters dated September 23, October 15, October 17, October 31, and November 7, 2013, and letters dated January 7, March 13, April 29, and October 6, 2014, and January 15, 2015, 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; 8. 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.

Enclosure 1 2. Accordingly, Renewed Facility Operating License No. DPR-43 is amended as indicated in the attachment to this license amendment.

Paragraph 2.8.(1) of the Renewed Facility Operating License No. DPR-43 is hereby amended to read as follows: (1) Pursuant to Section 104b of the Act and 10 CFR Part 50, "Licensing of Production and Utilization Facilities," to possess and use the facility at the designated location in Kewaunee County, Wisconsin in accordance with the procedures and limitations set forth in this renewed license; Paragraph 2.8.(2) of the Renewed Facility Operating License No. DPR-43 is hereby amended to read as follows: (2) Pursuant to the Act and 10 CFR Part 70, to possess at any time special nuclear material that was used as reactor fuel in accordance with the limitations for storage, as described in the Final Safety Analysis Report, as supplemented and amended; Paragraph 2.8.(3) of the Renewed Facility Operating License No. DPR-43 is hereby amended to read as follows: (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 sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; and possess any byproduct, source and special nuclear material as sealed neutron sources that was used for reactor startup; Paragraph 2.C.(1) of the Renewed Facility Operating License No. DPR-43 is hereby amended to read as follows: (1) Deleted; Paragraph 2.C.(2) of the Renewed Facility Operating License No. DPR-43 is hereby amended to read as follows: * (2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 215, are hereby incorporated in the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications. Paragraph 2.C.(3) of the Renewed Facility Operating License No. DPR-43 is hereby amended to read as follows: (3) Deleted Paragraph 2.C.(6) of the Renewed Facility Operating License No. DPR-43 is hereby amended to read as follows: (6) Deleted Paragraph 2.C.(14) of the Renewed Facility Operating License No. DPR-43 is hereby amended to read as follows: (14) Deleted Paragraph 2.E of the Renewed Facility Operating License No. DPR-43 is hereby amended to read as follows: E. Deleted; New license condition 3 of the Renewed Facility Operating License No. DPR-43 is hereby added to read as follows: 3. On February 25, 2013, Dominion Energy Kewaunee (DEK) certified that operations at Kewaunee Power Station would permanently cease in accordance with 10 CFR 50.82(a)(1

)(i). On May 14, 2013, DEK certified that the fuel had been permanently removed from the reactor vessel in accordance with 10 CFR 50.82(a)(1

)(ii). As a result, the 10 CFR 50 license no longer authorizes operation of the reactor, or the emplacement or retention of fuel in the reactor vessel. This license is effective as of the date of issuance and authorizes ownership and possession of Kewaunee Power Station until the Commission notifies the licensee in writing that the license is terminated.

The licensee shall: A. Take actions necessary to decommission the plant and continue to maintain the facility, including, where applicable, the storage, control and maintenance of the spent fuel, in a safe condition; and B. Conduct activities in accordance with all other restrictions applicable to the facility in accordance with the NRC regulations and the applicable provisions of the 10 CFR 50 facility license as defined in Section 2 of this license. . 3. This license amendment is effective as bf its date of issuance and shall be implemented within 90 days of the date of issuance.

Attachment:

Changes to the Renewed Facility Operating License and Technical Specifications FOR THE NUCLEAR REGULATORY COMMISSION Division of Operating Reactor Licensing, Office of Nuclear Reactor Regulation Date of Issuance: .february 13, 2015 ATTACHMENT TO LICENSE AMENDMENT NO. 215 RENEWED FACILITY OPERATING LICENSE NO. DPR-43 DOCKET NO. 50-305 Replace the following page of the Renewed Facility Operating License No. DPR-43 with the attached revised page. The revised page is identified by amendment number and contains marginal lines indicating the areas of change. Remove -3--4--5--6--7 -Renewed Facility Operating License Insert -3--4--5--6-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 ii iii 1.1-1 1.1-2 1.1-3 1.1-4 1.1-5 1.1-6 1.3-1 1.3-2 1.3-3 1.3-4 1.3-5 1.3-6 1.3-7 1.3-8 1.3-9 1.3-10 1.4-1 1.4-2 Technical Specifications 1.1-1 1.1-2 1.3-1 1.3-2 1.3-3 1.3-4 1.3-5 1.3-6 1.3-7 1.3-8 1.3-9 1.4-1 1.4-2 Remove 1.4-3 1.4-4 1.4-5 1.4-6 1.4-7 2.0-1 3.0-1 3.0-2 3.0-3 3.0-4 3.0-5 3.1.1-1 3.1.2-1 3.1.2-2 3.1.3-1 3.1.3-2 3.1.4-1 3.1.4-2 3.1.4-3 3.1.4-4 3.1.5-1 3.1.5-2 3.1.6-1 3.1.6-2 3.1.7-1 3.1.7-2 3.1.7-3 3.1.8-1 3.1.8-2 3.2.1-1 3.2.1-2 3.2.1-3 3.2.1-4 3.2.2-1 3.2.2-2 3.2.2-3 3.2.3-1 3.2.4-1 3.2.4-2 3.2.4-3 3.3.1-1 3.3.1-2 3.3.1-3 3.3.1-4 3.3.1-5 3.3.1-6 Insert 1.4-3 3.0-1 _ 3.0-2 Remove 3.3.1-7 3.3.1-8 3.3.1-9 3.3.1-10 3.3.1-11 3.3.1-12 3.3.1-13 3.3.1-14 3.3.1-15 3.3.2-1 3.3.2-2 3.3.2-3 3.3.2-4 3.3.2-5 3.3.2-6 3.3.2-7 3.3.2-8 3.3.2-9 3.3.3-1 3.3.3-2 3.3.3-3 3.3.3-4 3.3.4-1 3.3.4-2 3.3.6-1 3.3.6-2 3.3.6-3 3.3.6-4 3.4.1-1 3.4.1-2 3.4.2-1 3.4.3-1 3.4.3-2 3.4.3-3 3.4.3-4 3.4.4-1 3.4.5-1 3.4.5-2 3.4.6-1 3.4.6-2 3.4.7-1 3.4.7-2 3.4.7-3 3.4.8-1 3.4.8-2 3.4.9-1 Remove 3.4.9-2 3.4.10-1 3.4.10-2 3.4.11-1 3.4.11-2 3.4.11-3 3.4.12-1 3.4.12-2 3.4.12-3 3.4.12-4 3.4.13-1 3.4.13-2 3.4.14-1 3.4.14-2 3.4.14-3 3.4.15-1 3.4.15-2 3.4.15-3 3.4.16-1 3.4.16-2 3.4.17-1 3.4.17-2 3.5.1-1 3.5.1-2 3.5.2-1 3.5.2-2 3.5.2-3 3.5.3-1 3.5.3-2 3.5.4-1 3.6.1-1 3.6.2-1 3.6.2-2 3.6.2-3 3.6.2-4 3.6.3-1 3.6.3-2 3.6.3-3 3.6.3-4 3.6.3-5 3.6.3-6 3.6.4-1 3.6.5-1 3.6.6-1 3.6.6-2 3.6.6-3 Remove 3.6.7-1 3.6.7-2 3.6.8-1 . 3.6.8-2 3.6.9-1 3.6.10-1 3.7.1-1 3.7.1-2 3.7.1-3 .3.7.1-4 3.7.2-1 3.7.2-2 3.7.3-1 3.7.3-2 3.7.4-1 3.7.4-2 3. 7.5-1 3.7.5-2 3.7.5-3 3.7.5-4 3. 7.6-1 3.7.7-1 3.7.7-2 3.7.8-1 3.7.8-2 3.7.9-1 3.7.12-1 3.7.12-2 3.7.13-1 3.7.14-'1 3.7.15-1 3.7.15-2 3.7.15-3 3.7.16-1 3.8.1-1 3.8.1-2 3.8.1-3 3.8.1-4 3.8.1-5 3.8.1-6 3.8.1-7 3.8.1-8 3.8.1-9 3.8.1-10 3.8.4-1 3.8.4-2 3.7.13-1 3.7.14-1 3.7.15-1 3.7.15-2 Remove Insert 3.8.4-3 3.8.7-1 3.8.7-2 3.8.9-1 3.8.9-2 3.9.1-1 3.9.2-1 3.9.2-2 3.9.3-1 3.9.3-2 3.9.4-1 3.9.4-2 3.9.4-3 3.9.5-1 3.9.6-1 3.9.6-2 4.0-1 4.0-1 4.0-2 5.1-1 5.1-1 5.2-1 5.2-1 5.2-2 5.2-2 5.2-3 5.3-1 5.3-1 5.4-1 5.4-1 5.5-1 5.5-1 5.5-2 5.5-2 5.5-3 5.5-3 5.5-4 5.5-4 5.5-5 5.5-6 5.5-7 ---. 5.5-8 5.5-9 5.5-10 5.5-11 5.5-12 5.5-13 5.5-14 5.5-15 5.6-1 5.6-1 5.6-2 5.6-3 5.6-4 H. After weighing the environmental, economic, technical, and other benefits of the facility against environmental costs and considering available alternatives, the issuance of Renewed Facility Operating License No. DPR-43 is in accordance with 10 CFR Part 51 (formerly Appendix D of 1 O CFR Part 50) of the Commission's regulations and all applicable requirements have been satisfied; I. The receipt, possession, and use of byproduct, source, and special nuclear material as authorized by this renewed operating license will be in accordance with the Commission's regulations in 10 CFR Parts 30 and 70, including 10 CFR Sections 30.33, 70.23, and 70.31; and J. Actions have been identified and have been or will be taken with respect to (1) managing the effects of aging' during the period of extended operation on the functionality of structures and components that have been identified to require review under 10 CFR 54.21 (a)(1 ), and (2) time-limited aging analyses that have been identified to require review under 1 O CFR 54.21 (c), such that there is reasonable assurance that the activities authorized by this renewed operating license will continue to be conducted in accordance with the current licensing basis, as definec;f in 1 O CFR 54.3, for the facility, and that any changes made to the facility's current licensing basis in order to comply with 10 CFR.54.29(a) are in accordance with the Act and the Commission's regulations.

2. Renewed Facility Operating License No. DPR-43 is hereby issued to Dominion Energy Kewaunee, Inc., to read as follows: A. This license applies to the Kewaunee Power Station, a pressurized water nuclear reactor and associated equipment (the facility), owned by Dominion Energy Kewaunee, Inc. The facility is located in Kewaunee County, Wisconsin, and is described in the "Final Safety Analysis Report" as supplemented and amended, and in the Environmental Report as supplemented and amended. 8. Subject to the conditions and requirements incorporated herein, the.Commission hereby licenses Dominion Energy Kewaunee, Inc.: (1) Pursuant to Section 104b of the Act and 10 CFR Part 50, "Licensing of Production and Utilization Facilities," to possess and use the facility at the designated location in Kewaunee County, Wisconsin in accordance with the procedures and limitations set forth in this renewed license; (2) Pursuant to the Act and 1 O CFR Part 70, to possess at any time special nuclear material that was used as reactor fuel in accordance with the limitations for storage, as described in the Final Safety Analysis Report, as supplemented and amended; (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 sources for reactor instrumentation and radiation monitoring equipment calibration and as fission detectors in amounts as required; and possess any byproduct, source and special nuclear material as sealed neutron sources that was used for reactor startup; Renewed Operating License DPR-43 Amendment No. 215 (4) Pursuant to the Act and 1 O 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 calibration or associated with radioactive apparatus or components; (5) Pursuant to the Act and 1 O 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 co,nditions specified in the following Commission regulations in 1 O CFR Chapter 1: (1) Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; and (2) 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 (3) is subject to the additional conditions specified or incorporated below: (1) Deleted (2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 215, are hereby incorporated in the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications.

(3) Deleted (4) Physical Protection The licensee shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 1 O CFR 73.55 (51 FR 27817 and 27822) and to the authority of 1 O CFR 50.90 and 1 O CFR 50.54(p).

The combined set of plans, which contains Safeguards Information protected under 1 O CFR 73.21 is entitled: "Nuclear Management Coryipany Kewaunee Nuclear Power Plant Physical Security Plan (Revision O)" submitted by letter dated October 18, 2004, and supplemented by letter dated October 21, 2004, July 26, 2005, and May 15, 2006. The licensee shall fully implement and maintain in effect all provisions of the Commission-approved Kewaunee, Millstone, North Anna, and Surry Power Stations Cyber Security Plan (CSP), including changes made pursuant to the authority of 1 O CFR 50.90 and 1 O CFR 50.54(p).

The CSP was approved by License Amendment No. 210. (5) Deleted Renewed Operating License DPR-43 Amendment No. 215 (6) Deleted (7) Deleted (8) Deleted ' (9) Deleted (10) . Mitigation Strategy License Condition Develop and maintain strategies for addressing large fires and explosions and that include the following key areas: (a) Fire fighting response strategy with the following elements:

1. Pre-defined coordinated fire response strategy and guidance 2. Assessment of mutual aid fire fighting assets 3. Designated staging areas for equipment and materials

4. Command and control 5. Training of response personnel (b) Operations to mitigate fuel damage considering the following:

1 . Protection and use of personnel assets 2. Communications

3. Minimizing fire spread 4. Procedures for implementing integrated fire response strategy 5. Identification of readily-available pre-staged equipment

6. Training on integrated fire response strategy 7. Spent fuel pool mitigation measures (c) Actions to minimize release to include consideration of: 1. Water spray scrubbing

2. Dose to onsite responders (11) Seismic Analysis Methodology for Auxiliary Building Crane The licensee shall use the seismic analysis methodology submitted by letter dated July 7, 2008, supplemented on September 19, 2008, and March 17, 2009, and approved by the NRC staff in Amendment No. 205, for analysis of the Auxiliary Building crane. The licensee shall update the USAR to reflect this approval and in accordance with the schedule specified by 1 O CFR 50.71 (e). (12) Implementation of New and Revised Surveillance Requirements (a) For Surveillance Requirements (SRs) that are new in Amendment No. 207, the first performance is due at the end of the first surveillance interval, which begins on the date of implementation of that amendment.

Renewed Operating License DPR-43 Amendment No. 215 (b) (c) (d) For SRs that existed prior to Amendment No. 207 whose intervals of performance are being reduced, the first reduced surveillance interval begins upon completion of the first surveillance performed after implementation of that amendment.

For SRs that existed prior to Amendment No. 207 that have modified acceptance criteria, the first performance subject to the modified acceptance criteria is due at the end of the surveillance interval that began on the date the surveillance was last performed prior to the implementation of that amendment.

For SRs that existed prior to Amendment No. 207 whose intervals of performance are being extended, the first extended surveillance interval begins upon completion of the last surveillance performed prior to the implementation of that amendment.

(13) Removed Details and Requirements Relocated to Other Controlled Documents License Amendment No. 207 authorizes the relocation of certain technical specifications and operating license conditions, if applicable, to other licensee-controlled documents.

Implementation of that amendment shall include relocation of these requirements to the specified documents. ( 14) Deleted (15) Deleted (16) Spent Fuel Pool Neutron Absorber Material Surveillance Programs If all spent fuel assemblies have not been removed from the spent fuel pool by December 31, 2017, the licensee shall request, prior to that date, an amendment to the license, pursuant to 1 O CFR 50.90, to incorporate boron carbide and Baral surveillance programs (specified as Items 38 and 39 in Appendix A of NUREG-1958, "Safety Evaluation Report Related to the License Renewal of Kewaunee Power Station," dated January 2011) into the Technical Specifications.

D. The licensee shall comply with applicable effluent limitations and other limitations and monitoring requirements, if any, specified pursuant to Section 401 (d) of the Federal Water Pollution Control Act Amendments of 1972. E. Deleted Renewed Operating License DPR-43 Amendment No. 215 3. On February 25, 2013, Dominion Energy Kewaunee (DEK) certified that operations at Kewaunee Power Station would permanently cease in accordance with 10 CFR 50.82(a)(1

)(i). On May 14, 2013, DEK certified that the fuel had been permanently removed from the reactor vessel in accordance with 10 CFR 50.82(a)(1

)(ii). As a result, the 1 O CFR 50 license no longer authorizes operation of the reactor, or the emplacement or retention of fuel in the reactor vessel. This license is effective as of the date of issuance and authorizes ownership and possession of Power Station until the Commission notifies the licensee in writing *that the license is terminated.

The licensee shall: A. Take actions necessary to decommission the plant and continue to maintain the facility, including, where applicable, the storage; control and maintenance of the spent fuel, in a safe condition; and B. Conduct activities in accordance with all other restrictions applicable to the facility in accordance with the NRG regulations and the applicable provisions of the 1 O CFR 50 facility license as defined in Section 2 of this license. FOR THE NUCLEAR REGULATORY COMMISSION IRA! Eric J. Leeds, Director Office of Nuclear Reactor Regulation

Attachment:

Appendices A and B-Technical Specifications Date of Issuance:

February 24, 2011 Renewed Operating License DPR-43 Amendment No. 215 TABLE OF CONTENTS Page Number 1.0 USE ANDAPPLICATION 1.1 Definitions

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1.1-1 1 .2 Logical Connectors

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1 .2-1 1.3 Completion Times ......................................... , ......................... , ............................

1.3-1 1.4 Frequency

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1.4-1 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY

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3.0-1 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY

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..... 3.0-2 3. 7 PLANT SYSTEMS 3.7.13 Spent Fuel Pool Water Level ..........................................................................

3.7.13-1 3. 7.14

• Spent Fuel Pool Boron Concentration

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3. 7.14-1 3.7.15 Spent Fuel Pool.Storage

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3.7.15-1 4.0 DESIGN FEATURES 4.1 Site Location ..........

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4.0-1 4.3 Fuel Storage ...................

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4.0-1 5.0 ADMINISTRATIVE CONTROLS 5.1 Responsibility

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5.1-1. 5.2 Organization

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5.2-1 5.3 Facility Staff Qualifications

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5.3-1' 5.4 Procedures

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5.4-1 5.5 Programs and Manuals .................................

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5.5-1 5.6 Reporting Requirements

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5.6-1 5.7 High Radiation Area ..............................................................................................

5.7-1 Kewaunee Power Station Amendment No. 215 1.0 USE AND APPLICATION

1. 1 Definitions Definitions 1 .1 -----------------------------------------------------------N 0 TE----------------.--------------------------------

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The defined terms of this section appear in capitalized type and are applicable throughout these Technical Specifications and Bases. ACTIONS ACTUATION LOGIC TEST CERTIFIED FUEL HANDLER CHANNEL CALIBRATION CHANNEL CHECK Kewaunee Power Station Definition ACTIONS shall be that part of a Specification that prescribes Required Actions to be taken under designated Conditions within specified Completion Times. An ACTUATION LOGIC TEST shall be the application of various simulated or actual. input combinations in conjunction with each possible interlock logic state required for OPERABILITY of a logic circuit and the verification of the required logic output. The ACTUATION LOGIC TEST, as a minimum, shall include a continuity check of output devices. ACERTIFIED FUEL HANDLER is an individual who complies with provisions of the CERTIFIED FUEL HANDLER training program required by TS 5.3.2. A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds within the necessary range and accuracy to known values of the parameter that the channel monitors.

The CHANNEL CALIBRATION shall encompass all devices in the channel required for channel OPERABILITY.

Calibration of instrument channels with resistance temperature detector (RTD) or thermocouple sensors may consist of an inplace qualitative assessment of sensor behavior and normal calibration of the remaining adjustable devices in the channel. The CHANNEL CALIBRATION may be performed by means of any series of sequential, overlapping, or total channel steps. A CHANNEL CHECK shall be the qualitative assessment, by observation, of channel behavior during operation.

This determination shall include, where possible, comparison of the channel indication and status to other indications or status derived from independent instrument channels measuring the same parameter.

1.1-1 Amendment No. 215 CHANNEL OPERATIONAL TEST (COT) OPERABLE-OPERABILITY STAGGERED TEST BASIS TRIP ACTUATING DEVICE OPERATIONAL TEST (TADOT) Kewaunee Power Station Definitions 1; 1 A COT shall be the injection of a simulated or actual signal into the channel as close to the sensor as practicable to

• verify OPERABILITY of all devices in the channel required for channel OPERABILITY.

The COT shall include adjustments, as necessary, of the required alarm, interlock, and trip setpoints required for channel OPERABILITY such that the setpoints are within the necessary range and accuracy.

The COT may be performed by means of any series of sequential, overlapping, or total channel steps. A system, subsystem, train, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its specified safety function(s) are also capable of performing their related support function(s).

A STAGGERED TEST BASIS shall consist of the testing of one of the systems, subsystems, channels, or other designated components during the interval specified by the Surveillance Frequency, so that all systems, subsystems, channels, or other designated components are tested during n Surveillance Frequency intervals, where n is the total number of systems, subsystems, channels, or other designated components in the associated function. A T ADOT shall consist of operating the trip actuating device and verifying the OPERABILITY of all devices in the channel required for trip actuating device OPERABILITY.

The TADOT shall include adjustment, as necessary, of the trip actuating device so that it actuates at the required setpoint within the necessary accuracy.

The TADOT may be performed by means of any series of sequential, overlapping, or total channel steps. 1.1-2 Amendment No. 215.

Completion Times 1.3 1.0 USE AND APPLICATION 1 .3 Completion Times PURPOSE BACKGROUND DESCRIPTION The purpose of this section is to establish the Completion Time convention and to provide guidance for its use. Limiting Conditions for Operation (LCOs) specify minimum requirements for ensuring safe management of irradiated fuel. The ACTIONS associated with an LCO state Conditions that typically describe the ways in which the requirements of the LCO can fail to be met. Specified with each stated Condition are Required Action(s) and Completion Time(s). The Completion Time is the amount of time allowed for completing a Required Action. It is referenced to the time of dis,covery of a situation (e.g., inoperable equipment or variable not within limits) that requires entering an ACTIONS Condition unless otherwise specified, providing the facility is in a specified condition stated in the Applicability of the LCO. Required Actions must be completed prior to the expiration of the specified Completion Time. An ACTIONS Condition remains in effect and the Required Actions apply until the Condition no longer exists or the facility is not within the LCO Applicability.

If situations are discovered that require entry into more than one Condition at a time within a single LCO (multiple Conditions), the Required Actions for each Condition must be performed within the associated Completion Time. When in multiple Conditions, separate Completion Times are tracked for each Condition starting from the time of discovery of the situation that required entry into the Condition.

Once a Condition has been entered, subsequent trains, subsystems, components, or variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition, unless specifically,stated.

The Required Actions of the Condition continue to apply to each additional failure, with Completion Times based on initial entry into the Condition.

However, when a subsequent train, subsystem, component, or variable expressed in the Condition is discovered to be inoperable or not within limits, the Completion Time(s) may be extended.

To apply this Completion Time extension, two criteria must first be met. The subsequent inoperability:

a. Must exist concurrent with the first inoperability; and b. Must remain inoperable or not within limits after the first inoperability .is resolved.

Kewaunee Power Station 1.3-1 Amendment No. 215 Completion Times 1.3 1 .3 Completion Times DESCRIPTION (continued)

EXAMPLES The total Completion Time allowed for completing a Required Action to. address the subsequent inoperability shall be limited to the more restrictive of either: a. The stated Completion Time, as measured from the initial entry into the Condition, plus an additional 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />; or b. The stated Completion Time as measured from discovery of the subsequent inoperability.

The above Completion Time extensions do not apply to those Specifications that have exceptions that allow completely separate re-entry into the Condition (for each train, subsystem, component, or variable expressed in the Condition) and separate tracking of Completion Times based on this re-entry.

These exceptions are stated in individual Specifications.

The above Completion Time extension does not apply to a Completion Time with a modified "time zero." This modified "time zero" may be expressed as a repetitive time (i.e., "once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />," where the Completion Time is referenced from a previous completion of the Required Action versus the time of Condition entry) or as a time modified by the phrase "from discovery

... " The following examples illustrate the use of Completion Times with different types of Conditions and changing Conditions.

EXAMPLE 1 .3-1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. Required B.1 Verify .... 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Action and associated AND Completion Time not met. B.2 Restore .... 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> Kewaunee Power Station 1.3-2 Amendment No. 215 1.3 Completion Times EXAMPLES (continued)

Completion Times 1.3 Condition B has two Required Actions. Each Required Action has its own separate Completion Time. Each Completion Time is referenced to the time that Condition B is entered. The Required Actions of Condition B are to perform the verification required by ACTION B.1 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> AND perform the restoration required by ACTION B.2 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. A total of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is allowed for performing ACTION B.1 and a total of 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> (not 42 hours4.861111e-4 days <br />0.0117 hours <br />6.944444e-5 weeks <br />1.5981e-5 months <br />) is allowed for performing ACTION B.2 from the time that Condition B was entered. If ACTION B.1 is completed within 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />, the time allowed for completing ACTION B.2 is the next 33 hours3.819444e-4 days <br />0.00917 hours <br />5.456349e-5 weeks <br />1.25565e-5 months <br /> because the total time allowed for completing ACTION B.2 is 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. If Condition B is entered while performing the verification required by ACTION B.1, the time allowed for completing ACTION B.2 is the next 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. EXAMPLE 1.3-2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One pump A.1 Restore pump to 7 days inoperable.

OPERABLE status. B. Required B.1 Verify .... 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Action and associated AND Completion Time not met. B.2 Initiate .... 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> When a pump is declared inoperable, Condition A is entered. If the pump is not restored to OPERABLE status within 7 days, Condition B is also entered and the Completion Tf me clocks for Required Actions B.1 and B.2 start. If the inoperable pump is restored to OPERABLE status after Condition B is entered, Conditions A and B are exited, and therefore, the Required Actions of Condition B may be terminated.

Kewaunee Power Station 1.3-3 Amendment No. 215 1.3 Completion Times EXAMPLES (continued)

EXAMPLE 1.3-3 ACTIONS CONDITION A. One Function X train inoperable.

B. One Function Y train inoperable.

C. One Function X train inoperable.

AND* One Function Y train inoperable.

REQUIRED ACTION A.1 Restore Function X train to OPERABLE status. B.1 Restore Function Y train to OPERABLE status. C.1 Restore Function X train to OPERABLE status. OR C.2 Restore Function Y train to OPERABLE status. Completion Times 1.3 COMPLETION TIME 7 days 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 72 hours 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> When one Function X train and one Function Y train are inoperable, Condition A and Condition B are concurrently applicable.

The Completion Times for Condition A and Condition B are tracked separately for each train starting from the time each train was declared inoperable and the Condition was entered. A separate Completion Time is established for Condition C and tracked from the time the second train was declared inoperable (i.e., the time the situation described in Condition C was discovered).

Kewaunee Power Station 1.3-4 Amendment No. 215 1 .3 Completion Times EXAMPLES (continued)

Completion Times 1.3 If Required Action C.2 is completed within the specified Completion Time, Conditions B and C are exited. If the Completion Time for Required Action A.1 has not expired, operation may continue in accordance with Condition A.

• It is possible to alternate between Conditions A, B, and C in such a manner that operation could continue indefinitely without ever restoring systems to meet the LCO. However, doing so would be inconsistent with the basis of the Completion Times. Therefore, there shall be administrative controls to limit the maximum time allowed for any combination of Conditions that result in a single contiguous occurrence of failing to meet the LCO. These administrative controls shall ensure that the Completion Times for those Conditions are not inappropriately extended.

EXAMPLE 1 .3-4 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more A.1 Restore valve(s) to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> valves OPERABLE status. inoperable.

B. Required 8.1 Verify .... 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Action and associated AND Completion Time not met. 8.2 Initiate .... 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> A single Completion Time is used for any number of valves inoperable at the same time. The Completion Time associated with Condition A is based on the initial entry into Condition A and is not tracked on a per valve basis. Declaring subsequent valves inoperable, while Condition A is still in effect, does not trigger the tracking of separate Completion Times. Kewaunee Power Station 1.3-5 Amendment No. 215 1 .3 Completion Times EXAMPLES (continued)

Completion Times 1.3 Once one of the valves has been restored to OPERABLE status, the Condition A Completion Time is not reset, but continues from the time the first valve was declared inoperable.

The Completion Time may be extended if the valve restored to OPERABLE status was the first inoperable valve. The Condition A Completion Time may be extended for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> provided this does not result in any subsequent valve being inoperable for > 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. If the Completion Time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> (including the extension) expires while one or more valves are still inoperable, Condition B is entered. EXAMPLE 1.3-5 ACTIONS ---------------------------------------------N 0 TE -------------------------------------------Separate Condition entry is allowed for each inoperable valve. CONDITION REQUIRED ACTION COMPLETION TIME A. One or more A.1 Restore valve to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> valves OPERABLE status. inoperable.

B. Required B.1 Verify .... 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Action and associated AND Completion Time not met. B.2 Initiate .... 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> The Note above the ACTIONS Table is a method of modifying how the Completion Time is tracked. If this method of modifying how the Completion Time is tracked was applicable only to a specific Condition, the Note would appear in that Condition rather than at the top of the ACTIONS Table. Kewaunee Power Station 1.3-6 Amendment No. 215 1 .3 Completion Times EXAMPLES (continued)

Completion Times 1.3 The Note allows Condition A to be entered for each inoperable valve, and Completion Times tracked on a per valve basis. When a valve is declared inoperable, Condition A is entered and its Completion Time starts. If subsequent valves are declared inoperable, Condition A is entered for each valve and separate Completion Times start and are tracked for each valve. If the Completion Time associated with a valve in Condition A expires, Condition B is entered for that valve. If the Completion Times associated with subsequent valves in Condition A expire, Condition Bis entered separately for each valve and separate Completion Times start and are tracked for each valve. If a valve that caused entry into Condition B is restored to OPERABLE status, Condition Bis exited for that valve. Since the Note in this example allows multiple Condition entry and tracking of separate Completion Times, Completion Time extensions do not apply. EXAMPLE 1 .3-6 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One channel A.1 Perform SR 3.x.x.x. Once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> inoperable.

OR A.2 Verify .... 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> B. Required B.1 Initiate .... 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Action and associated Completion Time not met. Kewaunee Power Station 1.3-7 Amendment No. 215 1 .3 Completion Times EXAMPLES (continued)

Completion Times 1.3 Entry into Condition A offers a choice between Required Action A.1 or A.2. Required Action A.1 has a "once per" Completion Time, which qualifies for the 25% extension, per SR 3.0.2, to each performance after the initial performance.

The initial 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> interval of Required Action A.1 begins when Condition A is entered and the initial performance of Required Action A.1 must be complete within the first 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> interval.

If Required Action A.1 is followed, and the Required Action is not met within the Completion Time (plus the extension allowed by SR 3.0.2), Condition B is entered. If Required Action A.2 is followed and the Completion Time of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> is not met, Condition B is entered. If after entry into Condition B, Required Action A.1 or A.2 is met, Condition B is exited and operation may then continue in Condition A. EXAMPLE 1.3-7 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One A.1 Verify affected 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> subsystem subsystem isolated.

inoperable.

AND Once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter AND A.2 Restore subsystem 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to OPERABLE status. B. Required B.1 Verify .... 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Action and associated AND . Completion Time not met. B.2 Initiate .... 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> Kewaunee Power Station 1.3-8 Amendment No. 215 1.3 Completion Times EXAMPLES (continued)

Completion Times 1.3 Required Action A.1 has two Completion Times. The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Completion Time begins at the time the Condition is entered and each "Once per B hours thereafter" interval begins upon performance of Required Action A.1. If after Condition A is entered, Required Action A.1 is not met within either the initial 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or any subsequent 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> interval from the previous performance (plus the extension allowed by SR 3.0.2), Condition Bis entered. The Completion Time clock for Condition A does not stop after Condition B is entered, but continues from the time Condition A was initially entered. If Required Action A.1 is met after Condition B is entered, Condition B is exited and operation may continue in accordance with Condition A, provided the Completion Time for Required Action A.2 has not expired. IMMEDIATE When "Immediately" is used as a Completion Time, the Required Action COMPLETION TIME should be pursued without delay and in a controlled manner. Kewaunee Power Station 1.3-9 Amendment No. 2.15 Frequency 1.4 1 .0 USE AND APPLICATION 1.4 Frequency PURPOSE DESCRIPTION The purpose of this section is to define the proper use and application of Frequency requirements.

Each Surveillance Requirement (SR) has a specified Frequency in which the Surveillance must be met in order to meet the associated LCO. An understanding of the correct application of the specified Frequency is necessary for compliance with the SR. The "specified Frequency" is referred to throughout this section and each of the Specifications of Section 3.0, "Surveillance Requirement (SR) Applicability." The "specified Frequency" consists of the requirements of the Frequency column of each SR as well as certain Notes in the Surveillance column that modify performance requirements.

Sometimes special situations dictate when the requirements of a Surveillance are to be met. They are "otherwise stated" conditions allowed by SR 3.0.1. They may be stated as clarifying Notes in the Surveillance, as part of the Surveillance or both. Situations where a Surveillance could be required (i.e., its Frequency could expire), but where it is not possible or not desired that it be performed until sometime after the associated LCO is within its Applicability, represent potential SR 3.0.4 conflicts.

To avoid these conflicts, the SR (i.e., the Surveillance or the Frequency) is stated such that it is only "required.

when it can be and should be performed.

With an SR satisfied, SR 3.0.4 imposes no restriction.

The use of "met" or "performed" in these instances conveys specific meanings.

A Surveillance is "met" only when the acceptance criteria are satisfied.

Known failure of the requirements of a Surveillance, even without a Surveillance specifically being "performed," constitutes a Surveillance not "met." "Performance" refers only to the requiremenno specifically determine the ability to meet the acceptance criteria.

Kewaunee Power Station 1.4-1 Amendment No. 215 1 .4 Frequency EXAMPLES Frequency 1.4 The following examples illustrate the various ways that Frequencies are specified.

In these examples, the Applicability of the LCO (LCO not shown) is the applicable specified condition.

EXAMPLE 1 .4-1 / SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY Perform CHANNEL CHECK. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Example 1.4-1 contains the type of SR most often encountered in the Technical Specifications (TS). The Frequency specifies an interval (12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />) during which the associated Surveillance must be performed at least one time. Performance of the Surveillance initiates the subsequent interval.

Although the Frequency is stated as 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, an extension of the time interval to 1.25 times the stated Frequency is allowed by SR 3.0.2 for operational flexibility.

The measurement of this interval continues at all times, even when the SR is not required to be met per SR 3.0.1 (such as when the equipment is inoperable, a variable is outside specified limits, or the unit is outside the Applicability of the LCO). If the interval spedfied by SR 3.0.2 is exceeded while the unit is in a specified condition in the Applicability of the LCO, and the performance of the Surveillance is not otherwise modified (refer to Example 1.4-3), then SR 3.0.3 becomes applicable.

Kewaunee Power Station 1.4-2 Amendment No. 215 1.4 Frequency Frequency 1.4 EXAMPLES (continued)

If the interval as specified by SR 3.0.2 is exceeded while the unit is not in a specified condition in the Applicability of the LCO for which performance of the SR is required, then SR 3.0.4 becomes applicable.

The Surveillance must be performed within the Frequency requirements of SR 3.0.2, as modified by SR 3.0.3,.prior to entry into the specified condition or the LCO is considered not met (in accordance with SR 3.0.1) and LCO 3.0.4 becomes applicable.

EXAMPLE 1 .4-2 SURVEILLANCE REQUIREMENTS SURVEILLANCE Verify flow is within limits. FREQUENCY Once within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after entry into the applicable condition 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter Example 1 .4-2 has two Frequencies.

The first is a .one time performance Frequency, and the second is of the type shown in Example 1.4-1. The logical connector "AND" indicates that both Frequency requirements must be met Each time the applicable condition is entered, the Surveillance must be performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The use of "once" indicates a single performance will satisfy the specified Frequency (assuming no other Frequencies are connected by "AND"). This type cif Frequency does not qualify for the 25% extension allowed by SR 3.0.2. "Thereafter" indicates future performances must be established per SR 3.0.2, but only after a specified condition is first met (i.e., the "once" performance in this example).

If the applicable condition is exited, the measurement of both intervals stops. New intervals start upon entering the applicable condition.

Kewaunee Power Station 1.4-3 Amendment No. 215 LCO Applicability 3.0 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY LCO 3.0.1 LCO 3.0.2 LCO 3.0.3 LCO 3.0.4 LCO 3.0.5 LCOs shall be met during the specified conditions in the Applicability, except as provided in LCO 3.0.2. Upon discovery of a failure to meet an LCO, the Required Actions of the associated Conditions shall be met, except as provided in LCO 3.0.5. If the LCO is met or is no longer applicable prior to expiration of the specified Completion Time(s), completion of the Required Action(s) is not required unless otherwise stated. Deleted. When an LCO is not met, entry into a specified condition in the Applicability shall only be made: a. When the associated ACTIONS to be entered permit continued operation in the specified condition in the Applicability for an unlimited period of time; b. After performance of a risk assessment addressing inoperable systems and components, consideration of the results, determination of the acceptability of entering the specified condition in the Applicability, and establishment of risk management actions, if *appropriate; exceptions to this Specification are stated in the individual Specifications; or

• c. When an allowance is stated in the individual value, parameter, or other Specification.

• This Specification shall not prevent entry into specified conditions in the Applicability that are required to comply with ACTIONS. Equipment removed from service or declared inoperable to comply with ACTIONS may be returned to service under administrative control solely to perform testing required to demonstrate its OPERABILITY or the OPERABILITY of other equipment.

This is an exception to LCO 3.0.2 for the system returned to service under administrative control to perform the testing required to demonstrate.

OPERABILITY.

Kewaunee Power Station 3.0-1 Amendment No. 215 SR Applicability 3.0 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY SR 3.0.1 SR 3.0.2 SR 3.0.3 SR 3.0.4 SRs shall be met during the specified conditions in the Applicability for individual LCOs, unless otherwise stated in the SR. Failure to meet a Surveillance, whether such failure is experienced during the performance of the Surveillance or between performances of the Surveillance, shall be failure to meet the LCO. Failure to perform a Surveillance within the specified Frequency shall be failure to meet the LCO except as provided in SR 3.0.3. Surveillances do not have to be performed on inoperable equipment or variables outside specified limits. The specified Frequency for each SR is met if the Surveillance is performed within 1.25 times the interval specified in the Frequency, as measured from the previous performance or as measured from the time a specified condition of the Frequency is met. For Frequencies specified as "once," the above interval extension does not apply. If a Completion Time requires periodic performance on a "once per ... " basis, the above Frequency extension applies to each performance after the initial performance.

Exceptions to this Specification are stated in the individual Specifications.

If it is discovered that a Surveillance was not performed within its specified Frequency, then compliance with the requirement to declare the LCO not met may be delayed, from the time of discovery, up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or up to the limit of the specified Frequency, whichever is greater. This delay period is permitted to allow performance of the Surveillance.

A risk evaluation shall be performed for any Surveillance delayed greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and the risk impact shall be managed. If the Surveillance is not performed within the delay period, the LCO must immediately be declared not met, and the applicable Condition(s) must be entered. When the Surveillance is performed within the delay period and the Surveillance is not met, the LCO must immediately be declared not met, and the applicable Condition(s) must be entered. Entry into a specified condition in the Applicability of an LCO shall only be made when the LCO's Surveillances have been met within their specified

• Frequency, except as provided by SR 3.0.3. When an LCO is not met due to Surveillances not having met, entry into a specified condition in the Applicability shall only be made in accordance with LCO 3.0.4. This provision shall not prevent entry into specified conditions in the Applicability that are required to comply with ACTIONS. Kewaunee Power Station 3.0-2 Amendment No. 215

3. 7 PLANT SYSTEMS 3.7.13 Spent Fuel Pool Water Level Spent Fuel Pool Water Level 3.7.13 LCO 3.7.13 The spent fuel pool water level shall be 23 ft over the top of irradiated fuel assemblies seated in the storage racks. APPLICABILITY:

During movement of irradiated fuel assemblies in the spenJ fuel pool. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Spent fuel pool water level not within limit. A.1 Suspend movement of Immediately irradiated fuel assemblies in the spent fuel pool. SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.13.1 Verify the spent fuel pool water level is 23 ft above 7 days the top of the irradiated fuel assemblies seated in the storage racks. Kewaunee Power Station 3.7.13-1 Amendment No. 215 Spent Fuel Pool Boron Concentration 3.7.14 3. 7 PLANT SYSTEMS 3.7.14 Spent Fuel Pool Boron Concentration LCO 3.7.14 The spent fuel pool boron concentration shall be 240 ppm. APPLICABILITY:

When fuel assemblies are stored in the spent fuel pool and a spent fuel pool verification has not been performed since the last movement of fuel assemblies in the spent fuel pool. ACTIONS CONDITION REQUIRED ACTION A. Spent fuel pool boron A.1 Suspend movement of fuel concentration not within assemblies in the spent fuel limit. pool. AND A.2.1 Initiate action to restore spent fuel pool boron concentration to within limit. OR A.2.2 Initiate action to perform a spent fuel pool verification.

SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.7.14.1 Verify the spent fuel pool boron concentration is within limit. Kewaunee Power Station 3.7.14-1 COMPLETION TIME Immediately Immediately Immediately FREQUENCY 7 days Amendment No. 215 Spent Fuel Pool Storage 3.7.15 3. 7 PLANT SYSTEMS 3.7.15 Spent Fuel Pool Storage LCO 3.7.15 APPLICABILITY:

ACTIONS The combination of initial enrichment and burnup oLeach fuel assembly stored in the spent fuel pool shall be in accordance with the following:

• a. Irradiated fuel assemblies discharged prior to or during the 1984 refueling outage with a combination of burnup and initial nominal enrichment in the "Acceptable Domain" of Figure 3. 7.15-1 shall be stored in the transfer canal spent fuel pool or the north and south combined spent fuel pools; and b. Irradiated fuel assemblies discharged after the 1984 refueling outage, and irradiated fuel assemblies discharged prior to or during the 1984 refueling outage with a combination of burnup and initial nominal enrichment in the "Unacceptable Domain" of Figure 3. 7.15-1, shall be stored in the north and south combined spent fuel pools. Whenever any fuel assembly is stored in the spent fuel pool. CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of the LCO not met. A.1 Initiate action to move the . Immediately noncomplying fuel assembly to an acceptable location.

SURVEILLANCE REQUIREMENTS SR 3.7.15.1 SURVEILLANCE Verify by administrative means the initial enrichment and burnup of the fuel assembly is in accordance with Figure 3. 7.15-1. Kewaunee Power Station 3.7.15-1 FREQUENCY Prior to storing the fuel assembly in the spent fuel pool Amendment No. 215 Spent Fuel Pool Storage 3.7.15 35.0 Acceptable Domain , , I . I I I I (3.411, 30) I 30.0 *------1----1 1----i-------i----t-----

-I I I . I I 5 25.0 . B = 1.29731 x E 1 -.x E-

---!----!--I l I I (3 0, 24 t I 200 i 1-*--1 *------1---l-5 I I (2;,17.21)

I I I I 15*0 ---1--' (2.25, 13.37) . I i I l I I I I 10.0 ---,---[ ___ l ___ , _______ l _____ l __ _ (2.0, 9.24) I I I I I I I I ! I 5.0. ----------*-------------------L----*---1 I I I I I I I I I' I I I I I I 0.0 Bounding

2 2.2 2.4 2.6 2.8 3 3.2 3.4 Initial Enrichment (wt% U-235) Figure 3.7.15-1(page1of1)

Fuel Assembly Burnup Limits in the Spent Fuel Pools Kewaunee Power Station 3.7.15-2 Amendment No. 215 Design Features 4.0 4.0 DESIGN FEATURES 4.1 Site Location The Kewaunee Power Station is located on property owned by Dominion Energy Kewaunee Inc. at a site on the west shore of Lake Michigan, approximately 30 miles southeast of the city of Green Bay, Wisconsin.

The minimum distance from the center line of the reactor containment to the site exclusion radius as defined in 10 CFR 100.3 is 1200 meters. 4.2 Deleted 4.3 Fuel Storage 4.3.1 Criticality 4.3.1.1 The spent fuel storage racks are designed and shall be maintained with: a. Fuel assemblies having a maximum U-235 enrichment of 4.9776 weight percent; b. keff < 0.95 if fully flooded with unborated water, which includes an allowance for uncertainties as described in Section 9.5 of the USAR; c. A nominal 8.3 inch rack cell lattice spacing between fuel assemblies placed in the canal pool; d. A minimum 1 O inch center to center distance between fuel assemblies placed in the north and south pools; and e. New and spent fuel assemblies stored in the north and south pools and the canal pool in accordance with LCO 3.7.15, "Spent Fuel Pool Storage." 4.3.1.2 Deleted 4.3.2 Drainage The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 645 ft 2 inches (mean sea level). 4.3.3 Capacity The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 1205 fuel assemblies.

Kewaunee Power Station 4.0-1 Amendment No. 215 5.0 ADMINISTRATIVE CONTROLS 5.1 Responsibility Responsibility 5.1 5.1.1 The plant manager shall be responsible for overall plant operation and shall delegate in writing the succession to this responsibility during his absence. The plant manager or his designee shall approve, prior to implementation, each proposed test, experiment or modification to systems or equipment that affect nuclear safety. 5.1.2 The shift manager shall be responsible for the shift command function.

Kewaunee Power Station 5.1-1 Amendment No. 215 5.0 ADMINISTRATIVE CONTROLS . 5.2 Organization 5.2.1 Onsite and Offsite Organizations Organization 5.2 Onsite and offsite organizations shall be established for plant and corporate . management, respectively.

The onsite and offsite organizations shall include the positions for activities affecting safety of the nuclear fuel. a. Lines of authority, responsibility, and communication shall be defined and established throughout highest management levels, intermediate levels, and all operating organization positions.

These relationships shall be documented and updated, as appropriate, in organization charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent forms of documentation.

These requirements shall be documented in the quality . assurance program. The plant specific titles of those personnel fulfilling the responsibilities of the positions delineated in these Technical Specifications shall be maintained in appropriate plant documents.

b. The plant manager shall be responsible for overall safe operation of the plant and shall have control over those onsite activities necessary for safe storage and maintenance of the nuclear fuel. c. A specified corporate officer shall have corporate responsibility for overall plant nuclear safety and shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining, and providing technical support to the plant to ensure safe management of nuclear fuel. \, *, d. The individuals who train CERTIFIED FUEL HANDLERS, carry out health physics, or perform quality assurance functions may report to the appropriate onsite manager; however, these individuals shall have sufficient organizational freedom to ensure their ability to perform their assigned functions.

5.2.2 Facility Staff The facility staff organization shall include the following:

a. Each on duty shift shall be composed of at least the minimum shift crew composition shown in Table 5.2.2-1. b. Shift crew composition may be less than the minimum requirement of Table 5.2.2-1 for a period of not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, except in severe weather conditions, in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements.

• I Kewaunee Power Station 5.2-1 Amendment No. 215 5.2 Organization 5.2.2 Facility Staff (continued)

Organization 5.2 c. All fuel handling operations shall be directly supervised by a qualified individual.

d. The shift manager sha]I be a CERTIFIED FUEL HANDLER. e. An individual qualified in radiation protection procedures shall be onsite during fuel handling operations or movements of loads over storage racks containing fuel. Kewaunee Power Station 5.2-2 Amendment No. 215 Table 5.2.2-1 (page 1 of 1) Minimum Shift Crew Composition Organization 5.2 POSITION MINIMUM STAFFING CERTIFIED FUEL HANDLER Non-Certified Operator, Note: The Non-Certified Operator position may be filled by a CERTIFIED FUEL HANDLER. Kewaunee Power Station 5.2-3 Amendment No. 215 5.0 ADMINISTRATIVE CONTROLS 5.3 Facility Staff Qualifications Facility Staff Qualifications 5;3 5.3.1 Each member of the facility staff shall meet or exceed the minimum qualifications of ANSI N18.1-1971 for comparable positions, except for: a. The radiation protection manager who shall meet or exceed the recommendation of Regulatory Guide 1.8, Revision 1-R, September 1975, or their equivalent as further clarified in Attachment 1 to the NRG Safety Evaluation Report enclosed with Amendment No. 46, dated July 12, 1982. 5.3.2 An NRG approved training and retraining program for the CERTIFIED FUEL HANDLERS shall be maintained.

Kewaunee Power Station 5.3-1 Amendment No. 215 5.0 ADMINISTRATIVE CONTROLS 5.4 Procedures Procedures 5.4 5.4.1 Written procedures shall be established, implemented, and maintained covering the following activities:

a. The applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978; b. Deleted; c. Quality assurance for effluent and environmental monitoring;

d. Fire Protection Program implementation; and e. All programs specified in Specification 5.5. Kewaunee Power Station 5.4-1 . Amendment No. 215 Programs and Manuals 5.5 5.0 ADMINISTRATIVE CONTROLS 5.5 Programs and Manuals The following programs shall be established, implemented, and maintained.

5.5.1 Offsite Dose Calculation Manual (ODCM) a. The ODCM shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm and trip setpoints, and in the conduct of the radiological environmental monitoring program; and b. The ODCM shall also contain the radioactive effluent controls and radiological environmental monitoring activities, and descriptions of the information that should be included in the Annual Radiological Environmental Operating, and Radioactive Effluent Release Reports required by Specification 5.6.1 and Specification 5.6.2. c. Licensee initiated changes to the ODCM: 1 . Shall be documented and records of reviews performed shall be retained.

This documentation shall contain: a) Sufficient information to support the change(s) together with.the appropriate analyses or evaluations justifying the change(s);

and b) A determination that the change(s) maintain the levels of radioactive effluent control required by 1 O CFR 20.1302, 40 CFR 190, 10 CFR 50.36a, and 10 CFR 50, Appendix I, and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations;

2. Shall become effective after the approval of the plant manager; and 3. Shall be submitted to the NRG in the form of a complete, legible copy of the entire ODCM as a part of or concurrent with the Radioactive Effluent Release Report for the period of the report in which any change in the ODCM was made. Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (i.e., month and year) the change was implemented.

5.5.2 Deleted Kewaunee Power Station 5.5-1 Amendment No. 215 5.5 Programs and Manuals (continued) 5.5.3 Radioactive Effluent Controls Program Programs and Manuals 5.5 This program conforms to 1 O CFR 50.36a for the control of radioactive effluents and for maintaining the doses to members of the public from radioactive effluents . as low as reasonably achievable.

The program shall be contained in the ODCM, shall be implemented by procedures, and shall include remedial actions to be taken whenever the program limits are exceeded.

The program shall include the following elements:

a. limitations on the functional capability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology in the ODCM; b. Limitations on the concentrations of radioactive material released in liquid effluents to unrestricted areas, conforming to ten times the concentration*

values in Appendix B, Table 2, Column 2 to 1 O CFR 20.1001-20.2402;

c. Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20.1302 and with the methodology and parameters in the ODCM; d. Limitations on the annual and quarterly doses or dose commitment to a member of the public from radioactive materials in liquid effluents released from each unit to unrestricted areas, conforming to 10 CFR 50, Appendix I; e. Determination of cumulative dose contributions from radioactive effluents for the current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCM at least every 31 days. Determination of projected dose contribUtions from radioactive effluents in accordance with the methodology in the ODCM at least every 31 days; ' . f. Limitations on the functional capability and use of the liquid and gaseous effluent treatment systems to ensure that appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a period*of 31 days would exceed 2% of the guidelines for the annual dose or dose commitment, conforming to 10 CFR 50, Appendix I; g. Limitations on the dose rate resulting from radioactive material released in gaseous effluents from the site to areas at or beyond the site boundary shall be in accordance with the following: . 1. For noble gases: a dose rate::;; 500 mrem/yr to the whole body and a dose rate::;; 3000 mrem/yr to the skin; and 2. For iodine-131, iodine-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days: a dose rate::;; 1500 mrem/yr to any organ; Kewaunee Power Station 5.5-2 Amendment No. 215 Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.3 Radioactive Effluent Controls Program (continued) 5.5.4 5.5.5 5.5.6 5.5.7 5.5.8 5.5.9 5.5.10. h.* Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each unit fo areas beyond the site boundary, conforming to 10 CFR 50, Appendix I; i. Limitations on the annual and quarterly doses to a member of the public from iodine-131, iodine-133, tritium, and all radionuclides in particulate form with half lives > 8 days in gaseous effluents released from each unit to areas beyond the site boundary, conforming to 10 CFR 50, Appendix I; and j. Limitations on the annual dose or dose commitment to any member of the public, beyond the site boundary, due to releases of radioactivity and to radiation from uranium fuel cycle sources, conforming to 40 CFR 190. The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Radioactive Effluent Controls Program Surveillance Frequencies.

Deleted Deleted Deleted Deleted Deleted Deleted Storage Tank Radioactivity Monitoring Program This program provides controls for the quantity of radioactivity contained in unprotected outdoor liquid storage tanks. The liquid radwaste quantities shall be determined in accordance with Standard Review Plan, Section 15.7.3, "Postulated Radioactive Release due to Tank Failures." The program shall include: a. A surveillance program to ensure that the quantity of radioactivity contained in all outdoor liquid radwaste tanks that are not surrounded by liners, dikes, or walls, capable of holding the tanks' contents and that do not have tank overflows and surrounding area drains connected to the Waste Disposal *System is less than the amount that would result in concentrations less than the limits of 10 CFR 20, Appendix B, Table 2, Column 2, at the nearest potable water supply and the nearest surface water supply in an unrestricted area, in the event of an uncontrolled release of the tanks' contents.

Kewaunee Power Station 5.5-3 Amendment No. 215 Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.10 5.5.11 5.5.12 5.5.13 5.5.14 5.5.15 5.5.16 Storage Tank Radioactivity Monitoring Program (continued)

The provisions of SR 3.0.2 are applicable to the Storage Tank Radioactivity Monitoring Program Surveillance Frequencies.

Deleted Technical Specifications (TS) Bases Control Program This program provides a means for processing changes to the Bases of these Technical Specifications.

a. Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews. b. Licensees may make changes to Bases without prior NRC approval provided the changes do not require either of the following:

1. A change in the TS incorporated in the license; or 2. A change to the updated USAR or Bases that requires NRC approval pursuant to 1 O CFR 50.59. c. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the USAR. d. Proposed changes that meet the criteria of Specification 5.5.12.b above shall.be reviewed and approved by the NRC prior to implementation.

Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e).

Deleted Deleted Deleted Deleted Kewaunee Power Station 5.5-4 Amendment No. 215 5.0 ADMINISTRATIVE CONTROLS 5.6 Reporting Requirements Reporting Requirements 5.6 The following reports shall be submitted in accordance with 10 CFR 50.4. 5.6.1 Annual Radiological Environmental Operating Report The Annual Radiological Environmental Operating Report covering the operation of the facility during the previous calendar year shall be submitted by May 15 of each year. The report shall include summaries, interpretations, and analyses of trends of the results of the Radiological Environmental Monitoring Program for the reporting period. The material provided shall be consistent with the objectives outlined in the Offsite Dose Calculation Manual (ODCM), and in 10 CFR 50, Appendix I, Sections IV.B.2, IV.B.3, and IV.C. The Annual Radiological Environmental Operating Report shall include the results of analyses of all radiological environmental samples and of all environmental radiation measurements taken during the period pursuant to the locations specified in the table and figures in the ODCM, as well as summarized and tabulated results of these analyses and measurements in the format of the table in the Radiological Assessment Branch Technical Position, Revision 1, November In the event that some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for th_e missing results. The missing data shall be submitted in a supplementary report as soon as possible.

5.6.2 Radioactive Effluent Release Report The Radioactive Effluent Release Report covering the operation of the facility in the previous year shall be submitted by May 1 of each year in accordance with 1 O CFR 50.36a. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the facility.

The material provided shall be consistent with the objectives outlined in the ODCM and Process Control Program and in conformance with 10 CFR 50.36a and 10 CFR Part 50, Appendix I, Section IV.B.1. 5.6.3 Deleted 5.6.4 Deleted 5.6.5 Deleted Kewaunee Power Station 5.6-1 Amendment No. 215 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATING TO AMENDMENT NO. 215 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-43 DOMINION ENERGY KEWAUNEE.

INC. KEWAUNEE POWER STATION DOCKET NO. 50-305

1.0 INTRODUCTION

By letter dated February 25, 2013 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML 13058A065), Dominion Energy Kewaunee, Inc. (DEK, the licensee), submitted a certification to the U.S. Nuclear Regulatory Commission (NRC) indicating that it would permanently shut down Kewaunee Power Station (KPS) on May 7, 2013. On May 7, 2013, DEK permanently ceased power operations at KPS. On May 14, 2013, DEK certified that it had permanently defueled the KPS reactor vessel (ADAMS Accession No. ML 13135A209).

Consequently, pursuant to Part 50 of Title 10 of the Code of Federal Regulations (10 CFR), paragraph 50.82(a)(2), the KPS renewed facility operating license no longer authorizes operation of the reactor or emplacement or retention of fuel in the reactor vessel. By letter dated May 29, 2013, as supplemented by letters dated September 23, October 15, October 17, October 31, and November 7, 2013, and letters dated January 7, March 13, April 29, and October6, 2014, and a letter dated January 15, 2015 (ADAMS Accession Nos. ML 13156A037, ML13277A359, ML13294A091, ML13302B680, ML13308A326, ML13312A920, ML14009A393, ML 14084A384, ML 14126A005, ML 14281A757, and ML 15020A017 respectively), DEK requested an amendment tO the KPS Renewed Facility Operating License and Technical Specifications (TSs). Portions of the letter dated October 31, 2013, contain safeguards information and, accordingly, have been withheld from public disclosure.

The supplemental letters dated September 23, October 15, October 17, October 31, and November 7, 2013, and January 7, March 13, April 29, and October 6, 2014, and January 15, 2015, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination as published in the Federal Register on August 20, 2013 (78 FR 51224). In its original May 29, 2013, TS amendment request, DEK proposed numerous changes to conform the existing TSs to the permanently shutdown and defueled condition.

In general, the Enclosure 2 May 29, 2013, amendment request proposed to eliminate most of the previous operating TSs (that applied to KPS when it was authorized to operate) because these TSs are only applicable in operating modes or modes where fuel is em placed within the reactor vessel. Operation or emplacement of fuel into the reactor vessel in no longer authorized at KPS and, therefore, these TSs are no longer applicable or relevant based on the mode restrictions.

The proposed license amendment also requested changes to TS definitions and various organizational and program specifications.

In addition, the amendment requested certain license condition changes not directly related to the proposed TS changes. In its March 13, 2014, supplemental letter, DEK stated that it had accelerated its schedule to transfer spent fuel from the KPS spent fuel pool (SFP) to the KPS independent spent fuel storage installation (ISFSI). Based on its new schedule, DEK requested expedited review and approval to delete certain TSs no longer applicable during the movement of irradiated fuel assemblies.

By letter dated June 9, 2014 (ADAMS Accession No. ML 14111A234), the NRC issued Amendment No. 212 to the KPS Renewed Facility Operating License DPR-43. The amendment approved deletion of certain TSs no longer applicable to the movement of irradiated fuel assemblies, in response to DEK's March 13, 2014, supplemental request. The NRC staff stated in its June 9, 2014, letter to DEK approving amendment No. 212 that the remaining requested changes to the KPS TSs and license conditions remained under review by the NRC staff. The staff has completed its evaluation of the proposed changes to the permanently defueled TSs as requested in DEK's letter dated May 29, 2013, as supplemented, in this safety evaluation.

The May 29, 2013, amendment request, as supplemented, also proposed changes to the KPS license that DEK stated were either clarifications, redundant with other requirements, or no longer applicable based on the permanently shutdown and defueled status of KPS. The NRC staff has completed its review and assessment of the proposed license changes, as provided in the enclosed safety evaluation.

This amendment approves most of the proposed license changes requested by the licensee.

It also denies several of the proposed license changes. Specifically, the staff denies the proposed changes to Commission findings 1.B, 1.1, and 1.J of the license as discussed in Section 3.8.1 of the enclosed safety evaluation.

In addition, the NRC approves in part and denies in part the change proposed by DEK to license condition 2.8.(1) as discussed in Section 3.8.2 of the enclosed safety evaluation.

2.0 REGULATORY EVALUATION

2.1 Technical Specifications Section 182a of the Atomic Energy Act requires applicants for nuclear power plant operating licenses to include TSs as part of the application.

The NRC's regulatory requirements related to the content of the TSs are contained in 10 CFR 50.36, "Technical specifications." Pursuant to 10 CFR 50.36, each operating license issued by the Commission includes TSs and includes items in the following categories:

(1) safety limits, limiting safety systems settings and control settings, (2) limiting conditions for operation (LCOs), (3) surveillance requirements (SRs), (4) design features, (5) administrative controls, (6) decommissioning, (7) initial notification, and (8) written reports. Section 50.36 of 10 CFR provides four criteria to define the scope of equipment and parameters to be included in the TS LCOs. These criteria were developed for licenses authorizing operation (i.e., operating reactors) and focused on instrumentation to detect degradation of the reactor coolant system (RCS) pressure boundary, process variables and equipment, design features, or operating restrictions that affect the integrity of fission product barriers during design bases accidents or transients.

A fourth criterion refers to the use of operating experience and probabilistic risk assessment to identify and include in the TSs structures, systems, and components (SSCs) shown to be significant to public health and safety. A general discussion of the criteria that were used by the NRC staff in its evaluation to ensure that TS LCOs proposed for deletion are no longer required to be included in TSs is provided below. Criterion 1 of 10 CFR 50.36(c)(2)(ii)(A) states that TS LCOs must be established for "installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary." Since no fuel is present in the reactor or reactor coolant system at the KPS facility, this criterion is not applicable.

Criterion 2 of 10 CFR 50.36(c)(2)(ii)(B) states that TS LCOs must be established for 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." The purpose of this criterion is to capture those process variables that have initial values assumed in the design-basis accident (OBA) and transient analyses, and which are monitored and controlled during power operation.

The scope of OBAs applicable to a reactor permanently shut down and defueled is reduced from those postulated for an operating reactor, and most TSs satisfying Criterion 2 are no longer applicable.

There is one existing TS that defines the initial condition of the OBA associated with irradiated fuel movement that is discussed in Section 3.5 of this evaluation.

Criterion 3 of 10 CFR 50.36(c)(2)(ii)(C) states that TS LCOs for operation must be established for a structure, system, or component (SSC) "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." The intent of this criterion is to capture into TSs those SSCs that are part of the primary success path of a safety sequence analysis.

Also captured by this criterion are those support and actuation systems that are necessary for items in the primary success path to successfully function.

The primary success path of a safety sequence analysis consists of the combination and sequences of equipment needed to operate (including consideration of the single failure criterion), so that the plant response to OBAs and transients limits the consequences of these events to within the appropriate acceptance criteria.

There are no transients that continue to apply to permanently shutdown and defueled reactors.

The scope of applicable OBAs that apply to KPS are discussed in more detail in Section 3.0 of this safety evaluation.

Criterion 4 of 10 CFR 50.36(c)(2)(ii)(O) states that TS LCOs must be established for SSCs "which operating experience or probabilistic risk assessment has shown to be significant to public health and safety." The intent of this criterion is that risk insights and operating experience be factored into the establishment of TS LCOs. There are no longer any design-basis accidents at KPS that can result in a significant offsite radiological risk to public health and safety. The NRC staff notes that in the course of this evaluation, information contained in DRAFT NUREG-1625, "Proposed Standard Technical Specifications for Permanently Defueled Westinghouse Plants," March 1998 (ADAMS Accession No. ML082330233), was also

• considered.

This draft NUREG provides examples of decommissioning TSs for .Westinghouse pressurized water reactors (such as KPS) that the staff has previously found acceptable during TS reviews for permanently shutdown and defueled reactors.

2.2 Radiological Consequences Pursuant to the change process permitted by 10 CFR 50.59, "Changes, tests, and experiments," DEK has revised Chapter 14 of the KPS Updated Safety Analysis Report (USAR). Chapter 14 of the USAR describes the OBA and transient scenarios that could apply to KPS in its permanently sh,utdown and defueled status. Of these, DEK stated that there are no transients that continue to apply to KPS and that the only accident scenarios that could potentially apply to a permanently defueled facility would be a fuel handling accident (FHA), or an accident involving the rupture and release of reactor coolant system generated waste liquid or waste gas. The licensee stated that the reactor coolant system has been drained and vented. The liquid waste tanks storing the drained liquids from the reactor coolant system are located in the auxiliary building.

Any leakage from these tanks will be collected in the building sump and returned to the liquid waste system. Based on licensee analyses, there are no longer any radioactive liquid with a potential for being volatized.

Therefore, the storage of the remaining liquid waste from the reactor coolant system at KPS is no longer an applicable initiator or source of an accident.

The licensee stated that the only accident with potential offsite radiological consequences that remains applicable to KPS in the permanently shutdown and defueled conditio""'

is an FHA in the auxiliary building where the SFP is located. The FHA analysis for KPS shows that, following 90 days of decay time after reactor shutdown, the dose consequences from an FHA are acceptable without relying on SSCs to remain functional for accident mitigation during and following the event, provided that 23 feet of water is maintained above the irradiated fuel assemblies in the SFP. The NRC staff evaluated the radiological consequences of the postulated FHA OBA against the dose criteria specified in 10 CFR 50.67, "Accident source term," and using the guidance described in NRC Regulatory Guide (RG) 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," July 2000 (ADAMS Accession No. ML003716792).

RG 1.183 provides guidance to licensees on acceptable application of alternative source term (AST) submittals, including acceptable radiological analysis assumptions for use in conjunction with the accepted AST. By letter dated March 17, 2013, the NRC approved the implementation of the AST methodology for FHA dose consequence analysis at KPS by License Amendment No. 166 to Renewed Facility Operating License DPR-43 (ADAMS Accession No. ML030210062).

The submittal also included changes to the KPS TSs to reflect implementation of AST assumptions in accordance with 10 CFR 50.67. The FHA-specific dose acceptance criteria are specified in NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor] Edition," (SRP), Section 15.0.1, "Radiological Consequence Analyses Using Alternative Source Terms," July 2000 (ADAMS Accession No. ML003734190).

The dose acceptance criteria for the FHA are a Total Effective Dose Equivalent (TEDE) of 6.3 roentgen equivalent man (rem) at the exclusion area boundary (EAB) for the worst 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, 6.3 rem at the outer boundary of the low population zone (LPZ), and 5 rem in the control room (CR) for the duration of the accident.

3.0 TECHNICAL EVALUATION 3.1 Accident Analysis During normal power reactor operations, the forced flow of water through the RCS removes the heat generated by the reactor. The RCS, operating at high temperatures and pressures, transfers this heat through the steam generator tubes to the secondary system. The most severe postulated accidents for nuclear power plants involve damage to the nuclear reactor core and the release of large quantities of fission products to the RCS. Many of the accident scenarios postulated in the USAR involve failures or malfunctions of systems that could affect the reactor core. With the termination of reactor operations at KPS and the permanent removal of the fuel from the reactor core, such accidents are no longer possible.

The irradiated fuel is now stored in either the SFP or the KPS ISFSI. The reactor, RCS, and secondary system are no longer in operation and have no function related to the storage of the irradiated fuel. Therefore, the postulated accident$

involving failure or malfunction of the reactor, RCS, or secondary system are no longer applicable.

Chapter 14 of the KPS USAR describes transients and DBAs that are applicable to KPS during power operations.

Since KPS is permanently shut down and defueled, there are no longer any applicable transients.

The only DBAs that could potentially apply at this time to the permanently shutdown and defueled KPS reactor would be the.FHA, an accidental release of waste liquid, or

• an accidental release of waste gas. 3.2 Fuel Handling Accident In the AST evaluation supporting KPS power operation, the radiological consequence analysis evaluated the radiological consequences of a postulated FHA in the containment with no credit taken for containment isolation.

Since the assumptions and parameters used for an FHA inside containment are identical to those for an FHA in the auxiliary building, the resulting radiological consequences are the same regardless of the location of the accident.

Post-cessation of operations, an FHA onto the top of the core (or elsewhere within containment) is no longer possible and therefore no longer part of the licensing basis. However, an FHA in the SFP (which is located in the auxiliary building) is still possible at KPS, as long as spent fuel is stored in the SFP. The licensee defines the FHA in the SFP as the dropping of a spent fuel assembly onto the SFP floor or the racks that hold the spent fuel such that the cladding of all the fuel rods in one assembly ruptures.

The gap activity in the damaged rods is instantaneously released *into the SFP. The activity is assumed to pass through the 23 feet of required minimum water level over the top of the irradiated fuel assemblies in the SFP. It is postulated that the activity released from the SFP then mixes with the auxiliary building atmosphere before being released directly to the environment.

The FHA analysis postulates that the release from the SFP to the auxiliary building atmosphere is not mitigated en route to the environment.

This assumption is consistent with the current licensing basis FHA analysis, which does not credit the SFP ventilation system for accident mitigation.

The activity is assumed to be exhausted from the auxiliary building at a

• rate established to complete the release in 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> (consistent with RG 1.183). The NRC staff finds that the assumptions in the licensee's analysis are consistent with the current licensing basis FHA analysis, which does not credit the auxiliary building ventilation system for accident mitigation but assumes that it continues to operate to conservatively expel the activity to the environment or to the CR ventilation intake.

• The licensee assumed an overall decontamination factor of 200 for iodine in elemental and particulate forms in the SFP water with minimum water depth of 23 feet, consistent with the guidelines provided in RG 1.183. Also, a fission product decay period of 90 days cessation of operations) is assumed. The NRC staff finds the 90-day decay period assumption in the licensee's analysis to be conservative because KPS has been defueled since May 14, 2013, which is much greater than 90 days (nearly 21 months). Based on the actual time to date for spent fuel decay, the iodine source terms would be negligible.

No credit is taken for CR isolation or recirculation filtration in the FHA analysis.

Specifically, the Control Room Post Accident Recirculation system, support systems, or automatic actuation instrumentation (from the radiation monitor in the CR ventilation ductwork) are not credited in the DEK analysis.

No atmospheric dispersion is credited in the determination of CR dose. The NRC staff finds that the assumptions in the licensee's analysis are conservative.

The staff notes that the actual radiological plume will experience some dispersion in the environment in route to the CR intake. For calculation of doses at the EAB and LPZ, the licensee used current licensing basis atmospheric dispersion factors. The licensee's analysis of radiological consequences resulting from the postulated FHA for the permanently shutdown and defueled condition at KPS, concluded that the radiological consequences at the EAB, LPZ, and in the CR are within the dose criteria specified in 10 CFR 50.67 and accident-specific dose criteria described in SRP Section 15.0.1. The NRC staff has evaluated the licensee's analysis.

In performing this evaluation, the staff relied upon information provided by the licensee, as well as NRC staff experience in performing similar evaluations.

The staff reviewed the methods, parameters, and assumptions that the licensee used in its radiological dose consequence analyses and finds that they are consistent with the conservative guidance provided in RG 1.183. The FHA analysis assumption and parameters can be found in Table 1 of this safety evaluation.

The offsite atmospheric dispersion factors are in Table 2. The licensee's limiting calculated FHA dose results are given in Table 3. 3.3 AcGidental Release of Waste Liquid According the KPS USAR, postulated accidents that could result in the release of radioactive liquids are those that involve the rupture or leaking of system pipe lines or storage tanks containing waste liquids from the reactor coolant system. The reactor coolant system and associated holding tanks have been drained and vented. The remaining liquids are stored in liquid waste storage tanks in the auxiliary building.

All liquid waste and any leakage from the storage tanks or piping will be collected in the building sump to be pumped back into the liquid waste system. The building sumps and basement volume are sufficient to hold the full volume of a liquid wastes without overflowing to areas outside the building.

Should a complete failure of any waste tank occur, its contents will remain in the building.

Therefore, the potential hazard from waste liquid releases has been eliminated.

Since the KPS reactor is permanently shutdown, defueled, and placed in a long-term safe storage condition for decommissioning, liquid radioactive waste from operation of the reactor is no longer being generated.

The reactor coolant system and associated storage tanks and lines have been drained. The remaining liquids from reactor operation are stored in waste storage tanks in the auxiliary building that would contain any liquid leakage from the tanks. The NRC staff concludes that the storage of the remaining liquid waste from the reactor coolant system at KPS is no longe_r an applicable initiator or source of an accident and that an accident involving the liquid waste that could exceed the dose criteria specified in 10 CFR 50.67 is no longer possible.

3.4 Waste Gas Release Accidents The KPS licensing basis when it was operating included analyses of the radiological consequences of a rupture of a gas decay tank (GOT) and a rupture of the volume control tank (VCT). During reactor operation, the GDTs are used to store processed radioactive gases removed from the reactor coolant to allow for radioactive decay before the controlled release to the environment.

The VCT is a component in the plant's chemical and volume control systems that serves as a surge volume to balance differences in letdown and makeup flow rates while maintaining reactor coolant inventory.

Part of the reactor coolant (known as letdown) is removed from the RCS, cooled, filtered, demineralized, and degassed.

Since KPS is reactor is permanently shutdown and defueled, the GOT, VCT, and associated lines and tanks'are no longer in use and have been drained and purged of their contents, and ven,ted. The only remaining potential hazard from a radioactive waste release at KPS would be derived from any volatilized components in waste liquids. Licensee analyses of the remaining waste liquids indicate that the contents contain no radioactive gases with potential for being volatized while being stored, processed, or possibly leaked. Therefore, the potential hazard from volatized gases from processed or waste liquid releases has been eliminated.

The NRC staff concludes that a failure involving radioactive gas release from the waste liquid storage tanks that could exceed the dose criteria specified in 10 CFR 50.67 is no longer possible, and therefore, this hazard no longer exists. 3.5 Spent Fuel Pool Water Level Criterion 2 of 10 CFR 50.36(c)(2)(ii)(B) states that TS LCOs must be established for 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." The purpose of this criterion is to capture those process variables that have initial values assumed in the OBA. TS 3. 7 .13, "Spent Fuel Pool Water Level," specifies the TS required LCOs SRs that ensure the minimum water level in the SFP meets the assumptions of iodine decontamination factors following an FHA. OE K's analysis of the postulated FHA assumes that there is 23 feet of water between the top of the damaged fuel bundle and the fuel pool surface during the FHA. In the analysis, a FHA is defined as the dropping of a spent fuel assembly onto the SFP floor or racks such that the cladding of all the fuel rods in one assembly rupture. The gap activity in the damaged rods is instantaneously released into the SFP. The release occurs under 23 feet of water, which acts as a filter. The activity exhaust rate from the auxiliary building is established to complete the release in 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, as required by RG 1.183, but does not credit the auxiliary building ventilation for any mitigation of the release.

• Since the 23-foot water level of the SFP is an initial condition of the FHA OBA, it satisfies Criterion 2 for inclusion in TSs and is being retained as a TS for KPS in its permanently shutdown and defueled condition.

The amendment request by OEK does not involve any change to TS 3. 7 .13. The discussion in this evaluation of the SFP water level TS is provided only for completeness since the SFP water level is an important initial condition in the FHA analysis and should continue to be part of the KPS TSs. 3.6 Accident Analysis Summary As described above, the NRC staff reviewed the assumptions, inputs, and methods used by the licensee to assess the radiological consequences of OBAs based upon the permanently shutdown and defueled condition at KPS. The staff finds that the licensee used analysis methods and assumptions consistent with the conservative regulatory requirements and guidance identified in Section 2.2 of this safety evaluation.

The NRC staff compared the doses estimated by the licensee to the applicable criteria identified in Section 2.2. The licensee's limiting calculated OBA dose results are given in Table 3. The staff finds that the licensee has demonstrated with reasonable assurance that the estimates of the EAB, LPZ, and CR doses will comply with these criteria.

The NRC staff further finds that there is reasonable assurance that the KPS TSs, as modified by this license amendment, will continue to provide sufficient safety margins with adequate defense-in-depth to address postulated unanticipated events and to compensate for uncertainties in accident progression and analysis assumptions and parameters.

Therefore, the proposed license amendment is acceptable with respect to the radiological consequences of OBAs. Table 1 Parameters and Assumptions Used in Radiological Consequence Calculations Fuel Handling Accident*

Parameter Value Radial peaking factor 1.7 Source Term Decay 90 days Number of fuel assemblies damaged 1 Fuel pool water depth 23 feet Fuel gap fission product inventory:

50% of damaged fuel rods Noble gases excluding Kr-85 -------------------

5 percent Kr-85 -------------------------------------------

10 percent 1-131 -------------------------------------------

8 percent Other Halogens -----------------------------------

5 percent Remaining 50% of damaged fuel rods Noble gases excluding Kr-85 -------------------

10 percent Kr-85 -------------------------------------------

30 percent 1-131 -------------------------------------------

12 percent Other Halogens -----------------------------------

10 percent Fuel pool decontamination factors: Iodine 200 Noble gases 1 Duration of accident 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Control room: Isolation Not Credited Recirculation Not Credited Unfiltered lnflow 1 3000 cfm Atmospheric Dispersion Not Credited 1 Combined unfiltered normal intake and unfiltered inleakage for entire accident duration, which was varied between 400 -6,000 cubic feet per minute to maximize dose.

• The table above has been corrected from the table provided in the NRC staff's safety evaluation dated June 9, 2014 (ADAMS Accession No. ML 14111A234), due to an editorial error. The previous table listed alkali metal release fractions that were not used by the licensee.

In addition, fuel gap fission product inventory has been broken into two 50 percent fuel rod damage groups. The error in the information reported in previous evaluation did not affect the staff's conclusion. Table 2 Kewaunee Fuel Handling Accident Atmospheric Dispersion Factors (sec/m 3) Exclusion Area Low Population Control Room Time Period Boundary Zone Intake 0-2 hr. 2.232E-4 3.977E-5 Not credited Table 3 Calculated FHA Radiological Consequences EAB LPZ CR Calculated results, TEDE (90 day decay period) 0.001 rem 0.001 rem 1.9 rem Dose acceptance criteria, TEDE 6.3 rem 6.3 rem 5 rem 3. 7 Proposed TS Changes 3. 7 .1 Section 1.1, Definitions The licensee proposed deleting the following definitions because they pertain to an operating reactor. Since KPS is permanently shut down and defueled, the definitions have no relevance and no longer apply: AXIAL FLUX DIFFERENCE (AFD) -AFD shall be the difference in normalized flux signals between the top and bottom halves of a two section excore neutron detector.

CORE OPERATING LIMITS REPORT (COLR) -The COLR is the unit specific document that provides cycle specific parameter limits for the current reload cycle. These cycle specific parameter limits shall be determined for each reload cycle in accordance with Specification 5.6.3. Plant operation within these limits is addressed in individual Specifications.

DOSE EQUIVALENT 1-131 -DOSE EQUIVALENT 1-131 shall be that concentration of 1-131 that alone would produce the same dose when inhaled as the combined activities of specified iodine isotopes actually present. The determination of DOSE EQUIVALENT 1-131 shall be performed using ICRP-30, 1979, Supplement to Part 1, page 192-212, Table titled, "Committed Dose Equivalent in Target Organs or Tissues per Intake of Unit Activity." DOSE EQUIVALENT XE-133-DOSE EQUIVALENT XE-133 shall be that concentration of Xe-133 that alone would produce the same acute dose to the whole body as the combined activities of specified noble gas nuclides actually present. The determination of DOSE EQUIVALENT XE-133 shall be performed using effective dose conversion factors for air submersion listed in Table 111.1 of EPA Federal Guidance Report No. 12, 1993, "External Exposure to Radionuclides in Air, Water, and Soil." LEAKAGE -LEAKAGE [from the reactor coolant system] shall be [as summarized]:

a. Identified LEAKAGE [that is captured and conducted to collection systems or a sump or collecting tank, into the containment atmosphere, primary to secondary]

b. Unidentified LEAKAGE c. Pressure Boundary LEAKAGE MODE :.... A MODE shall correspond to any one inclusive combination of core reactivity condition, power level, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 [*]with fuel in the reactor vessel. *In conjunction with deletion of the term "MODE," TS Table 1.1-1, "MODES," is also being deleted. PHYSICS TESTS -PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation.

These tests are: a. Described in Chapter 13, "Initial Test and Operation," of the USAR; b. Authorized under the provisions of 10 CFR 50.59; or c. Otherwise approved by the Nuclear Regulatory Commission.

QUADRANT POWER TILT RA TIO (QPTR) -QPTR shall be the ratio of the maximum upper excore detector calibrated output to the average of the upper excore detector calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater. RATED THERMAL POWER (RTP) -RTP shall be a total reactor core heat transfer rate to the reactor coolant of 1772 MWt. SHUTDOWN MARGIN (SOM) -SOM shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming:

a. All rod cluster control assemblies (RCCAs) are fully inserted except for the single RCCA of highest reactivity worth, which is assumed to be fully withdrawn.

However, with all RCCAs verified fully inserted by two independent means, it is not necessary to account for a stuck RCCA in the SOM calculation.

With any RCCA not capable of being fully inserted, the reactivity worth of the RCCA must be accounted for in the determination of SOM; and b. In MODES 1 and 2, the fuel and moderator temperatures are changed to the nominal zero power level design temperature.

THERMAL POWER -THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant. The NRC staff examined the TS definitions proposed for deletion and concluded that all the terms listed above are only meaningful to a reactor authorized to operate. Since KPS is permanently shut down and defueled, the staff finds the deletion of these definitions from TSs acceptable.

In addition, the licensee proposed adding a definition for certified fuel handler. The licensee proposed to define a certified fuel handler as: CERTIFIED FUEL HANDLER -A CERTIFIED FUEL HANDLER is an individual who complies with provisions of the CERTIFIED FUEL HANDLER training program required by TS 5.3.2. Proposed TS 5.3.2 states an NRG-approved training and retraining program for the certified fuel handlers shall be maintained.

The NRC staff finds the definition of a certified fuel handler conforms to the usage contained in the Administrative Controls section of the KPS permanently defueled TSs and is consistent with the definition in 10 CFR Part 50 and is, therefore, acceptable.

3.7.2 Section 1.2, Logical Connectors The licensee is not proposing any changes to this section of the KPS defueled TS. 3.7.3 Section 1.3, Completion Times This section establishes the completion time convention throughout the TS and provides guidance for its use. The licensee has proposed to replace each reference to "operation of the unit" and "unit" with the new terminology, "management of irradiated fuel" and "facility," respectively, since operation of the unit is no longer permitted and safe management of irradiated fuel is the primary objective of the permanently defueled TS. In addition, the licensee proposed to delete references to "MODE," and "THERMAL POWER," to be consistent with the removal of these definitions from TSs and because these terms are no longer used in the required actions of the subsequent remaining LCO in the proposed KPS defueled TSs. The licensee also proposed to delete a portion of the explanation for Example 1.3-2 referring to LCO 3.0.3 since this LCO has been proposed for deletion and the information is no longer applicable to a permanently defueled reactor plant. Several examples of the proposed changes are shown with a strikethrough of the current wording and highlighting of the proposed changes: BACKGROUND Limiting Conditions for Operation (LCOs) specify minimum requirements for ensuring safe operation of the unit The ACTIONS associated with an LCO state Conditions that typically describe the ways in which the requirements of the LCO can fail to be met. Specified with each stated Condition are Required Action(s) and Completion Time(s). DESCRIPTION The Completion Time is the amount of time allowed for completing a Required Action. It is referenced to the time of discovery of a situation (e.g., inoperable equipment or variable not within limits) that requires entering an ACTIONS Condition unless otherwise specified, providing the umt §9Ult¥ is in a MODE or specified condition stated in the Applicability of the LCO. Required Actions must be completed prior to the expiration of the specified Completion Time. An ACTIONS Condition remains in effect and the Required Actions apply until the Condition no longer exists or the umt facility is not within the LCO Applicability

..... EXAMPLES The following examples illustrate the use of Completion Times with different types of Conditions and changing Conditions.

EXAMPLE 1.3-1 ACTIONS CONDITION B. Required Action and associated Completion Time not met.

ACTION 8.2 Be in MODE 5 COMPLETION TIME 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 36 hours Condition B has two Required Actions. Each Required Action has its own separate Completion Time. Each Completion Time is referenced to the time that Condition B is entered. 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. A total of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is allowed for 3 a total of 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> (not 42 hours4.861111e-4 days <br />0.0117 hours <br />6.944444e-5 weeks <br />1.5981e-5 months <br />) is allowed for reaching MODE 5 from the time that Condition B was entered. If 8:*1 MODE 3 is reached ted within 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />, the time allowed for ... ..

• ..

reaching 5

• .; ...*

.* next hours beca.use the total time allowed for reaching MODE 5 1s 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. EXAMPLE 1.3-2 When a pump is declared inoperable, Condition A is entered. If the pump is not restored to OPERABLE status within 7 days, Condition B is also entered and the Completion Time clocks for Required Actions B.1 and B.2 start. If the inoperable pump is restored to OPERABLE status after Condition B is entered, Conditions A and B are exited, and therefore, the Required Actions of Condition B may be terminated.

VVhen a second pump is declared inoperable while the first pump is still inoperable, Condition A is not re entered for the second pump. LCO 3.0.3 is entered, since the ACTIONS do not include a Condition for more than one inoperable pump. The Completion Time clock for Condition A does not stop after LCO 3.0.3 is entered, but continues to be tracked from the time Condition A was initially entered. VVhile in LCO 3.0.3, if one of the inoperable pumps is restored to OPERABLE status and the Completion Time for Condition A has not expired, LCO 3.0.3 may be exited and operation continued in accordance with Condition A. While in LCO 3.0.3, if one of the inoperable pumps is restored to OPERABLE status and the Completion Time for Condition A has expired, LCO 3.0.3 may be exited and operation continued in accordance with Condition B. The Completion Time for Condition B is tracked from the time the Condition A Completion Time expired. On restoring one of the pumps to OPERABLE status, the Condition A Completion Time is not reset, but continues from the time the first pump *.vas declared inoperable.

This Completion Time may be extended if the pump restored to OPERABLE status was the first inoperable pump. A 24 hour extension to the stated 7 days is allo*.ved, provided this does not result in the second pump being inoperable for > 7 days. The NRC staff has reviewed the proposed wording changes to the TS 1.3 Completion Times guidance and has determined that they are consistent with the transition from an operating reactor to a permanently shutdown and defueled facility with a primary safety focus of management of irradiated fuel. The proposed changes also remove references to operating modes or reactor thermal power that are no longer permitted following certification under the provisions of 10 CFR 50.82(a)(2).

In addition, the deletion of the example note referring to LCO 3.0.3 related actions is also appropriate since LCO 3.0.3 is being deleted as discussed in Section 3. 7.6 of this safety evaluation.

Therefore, the staff finds that the proposed changes to TS 1.3 are acceptable.

3.7.4 Section 1.4, Frequency This section defines the proper use and application of Frequency requirements throughout the TS. In this section, the licensee has proposed to delete the final paragraph in the description section, including the subparagraphs and associated examples.

The final paragraph of the TS 1.4 description section, regarding notes that modify the frequency of performance of some surveillances and the applicability of entry restrictions of SR 3.0.4, is being deleted in its entirety.

The subsequent subparagraphs regarding any of three conditions being satisfied, which are prefaced as a, b, and c, are also being deleted. None of the surveillances in the proposed TS contain notes that modify the frequency of performance or the conditions during which the acceptance criteria must be satisfied.

Therefore, this paragraph is not applicable to the proposed TS LCOs or SRs and may be deleted. Specifically, the following is being deleted: a. The Surveillance is not required to be met in the MODE or other specified condition to be entered; or b. The Surveillance is required to be met in the MODE or other specified condition to be entered, but has been performed within the specified Frequency (i.e., it is current) and is known not to be failed; or c. The Surveillance is required to be met, but not performed, in the MODE or other specified condition to be entered, and is knoi.vn not to be failed. Examples 1.4.3, 1.4 .4, 1.4.5, and 1.4.6 discuss these special situations.

In addition, the licensee has proposed to revise the terms "MODE," and "reactor power" and replace them with the terms referring to "entry into the applicable condition." The terms MODE and reactor power are no longer meaningful to the permanently shutdown and defueled condition at KPS. An example of the proposed changes are shown with a strikethrough of the current wording and highlighting of the proposed changes: EXAMPLE 1.4-2 SURVEILLANCE REQUIREMENTS SURVEILLANCE Verify flow is within limits. FREQUENCY Once within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after > 25% RTP

1"'V*cy<\.***

,. ,. *' * (_

.. 12P ,, ..* ,,,,,,, ***"'"***

.. ******' AND 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter Example 1.4-2 has two Frequencies.

The first is a one time performance Frequency, and the second is of the type shown in Example 1.4-1. The logical connector "AND" indicates that both Frequency requirements must be met. Each time reactor power is increased from a power level 2 25% RTP to RTP, the Surveillance must be performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The use of "once" indicates a single performance will satisfy the specified Frequency (assuming no other Frequencies are connected by "AND"). This type of Frequency does not qualify for the 25% extension allowed by SR 3.0.2. "Thereafter" indicates future performances must be established per SR 3.0.2, but only after a specified condition is first met (i.e., the "once" performance in this example).

If reactor po*.ver decreases to 2 RTP

• ' --.',V';:/>

the measurement of both intervals stops. New intervals start upon reactor power reaching 25%* RTP enferin *':tt1e

""'&... * . _g;._,,;, .... J21i:t .......

• ... The licensee also proposed to delete Examples 1.4-3, 1.4-4, 1.4-5, and 1.4-6 because these examples are not needed in a permanently defueled condition.

These examples used references to operating reactor modes that are no longer permitted at KPS. Specifically:

• Example 1.4-3 refers to surveillances to be performed after power is greater than or equal to 25 percent rated thermal power(% RTP).

• Example 1.4-4 and 1.4-5 refer to surveillances to be performed after reaching MODE 1.

• Example 1.4-6 refers to surveillances to be performed in MODES 1, 2, or 3. The licensee stated that the remaining Examples 1.4-1 and 1.4-2 (as revised) are sufficient to explain application of TS frequency requirements for the permanently defueled KPS TSs. The NRC staff has reviewed the proposed changes to TS Section 1.4 and has determined that they are ap'propriate for a permanently shutdown and defueled reactor. The proposed changes remove references to operating modes or reactor thermal powers that are no longer permitted following certification under the provisions of 10 CFR 50.82(a)(2).

The deletion of the surveillance note referring to MODE entry restrictions of SR 3.0.4 is also appropriate since none of the surveillances in the proposed remaining defueled TSs contain notes that modify the frequency of performance or the conditions during which the acceptance criteria must be satisfied.

Therefore, the proposed changes are acceptable.

3.7.5 Section 2.0, Safety Limits This section establishes safety limits, which preclude violation of the fuel design criteria and RCS design pressure.

TS 2.1, "Safety Limits" (SLs), contains two separate specifications:

• TS 2.1.1, Reactor Core SLs; and

• TS 2.1.2, Reactor Coolant System Pressure SL The TS SL section also contains TS 2.2, "Safety Limit Violations," which directs actions to be taken if an SL specified in TS 2.1 is violated.

The restrictions of the SLs promulgated in TS 2.1.1 prevent overheating of the fuel, cladding, and possible cladding perforation, which could result in the release of fission products to the reactor coolant. TS 2.1.1 is applicable in MODES 1 and 2. TS 2.1.2 promulgates requirements on parameters to protect the integrity of the RCS against overpressure.

TS 2.1.2 is applicable in MODES 1, 2, 3, 4, and 5. The licensee proposed to delete the SLs specified in Section 2.0 because they are not applicable to the permanently shutdown and defueled status of the plant. The licensee stated that the SL TSs limit important process variables that are necessary to reasonably protect the integrity of certain physical barriers required for safe operation of the reactor in MODES 1 through 5. However, 10 CFR 50.82(a)(2) prohibits operation of the reactor or placing fuel in the reactor vessel. Therefore, the SL TSs only address specific process variables that are no longer applicable to KPS. The NRC staff examined the Sls and their TS Bases. There are three Sls in Section 2.0: departure from nucleate boiling ratio (DNBR); peak fuel centerline temperature to ensure fuel and cladding integrity; and a maximum RCS pressure to ensure RCS integrity.

As stated in the Bases for DNBR, a limit is placed on the DNBR such that no fuel clad damage would occur as a result of normal operation and anticipated operational occurrences.

A limit is placed on peak fuel centerline temperature such that a hot fuel pellet in the core will not experience centerline fuel melting. The TS Bases for the maximum RCS pressure state that RCS integrity is an important barrier in the prevention of an uncontrolled release of fission products.

Because KPS has permariently shut down and defueled, and submitted certifications under the provisions of 10 CFR 50.82(a)(2), placing fuel in the reactor vessel and resuming*power operations are no longer authorized.

In this condition, the.re will be no DNBR or peak fuel centerline temperature to be monitored and there will be no challenge to RCS integrity.

Based on these findings, the staff concludes the Sls no longer apply. Therefore, the staff finds the deletion of the TS Sls acceptable.

3.7.6 Section 3.0, Limiting Conditions for Operation (LCO) and Surveillance Requirement (SR) Applicability This section contains the general requirements applicable to all LCOs and SRs and applies at all times unless otherwise stated in TSs. LCO 3.0.1 establishes the applicability statement within each individual TS as the requirement for when the LCO shall be met. The licensee proposed to delete the reference to "MODES" and references to LCO 3.0.7 and LCO 3.0.8. The licensee stated that reference to modes is no longer relevant since KPS is permanently shut down and defueled.

In addition, the deletion of references to LCOs 3.0.7 and 3.0.8 conforms to the request to delete theses LCOs from the KPS TSs as discussed below. LCO 3.0.2 establishes that upon discovery of a failure to meet an LCO, the associated actions shall be met. The licensee proposed to delete the reference to LCO 3.0.6 to conform to the request to delete TS LCO 3.0.6 from the KPS TSs as discussed below. LCO 3.0.3 establishes the actions thaf must be implemented when an LCO is not met and (1) an associated required action and completion time is not met and no other condition applies or (2) the condition of the unit is not specifically addressed by the associated actions. The licensee proposed to delete LCO 3.0.3 in its entirety since LCO 3.0.3 only applies in MODES 1, 2, 3, and 4. LCO 3.0.3 does not apply to KPS since the reactor is permanently shut down and defueled and TS modes are no longer relevant.

LCO 3.0.4 establishes limitations on changing modes or other specified conditions in the applicability when an LCO is not met. The licensee proposed to delete the reference to "MODES" since KPS is permanently shut down and defueled and TS modes are no longer relevant.

The licensee also proposed to delete the action that refers to shutting down of the unit since KPS is already permanently shut down. LCO 3.0.6 establishes an exception to LCO 3.0.2 for supported systems that have a support system LCO specified in the TS. In a DEK supplemental letter dated October 15, 2013 (ADAMS Accession No. ML 13294A091

), the licensee proposed to delete LCO 3.0.6 in its entirety.

The licensee states that there are no systems remaining in the proposed permanently defueled TSs that support or have commonality with any other TS systems (supported systems).

The licensee also stated that this conforms to its proposal to delete the safety function determination program, TS 5.5.13 as discussed in section 3.7.17.5 of this safety evaluation.

LCO 3.0.7 pertains to certain reactor physics special tests and operations required to be performed at various times over the life of the unit. LCO 3.0.7 is associated only with reactor physics test exceptions listed in LCO 3.1.8, which the licensee has proposed to delete. The licensee proposed to delete LCO 3.0.7 in its entirety since reactor physics testing is no longer relevant to and permanently shutdown and defueled reactor and to conform to its proposal to delete TS 3.1.8 as discussed in section 3.7.7 of this safety evaluation.

LCO 3.0.8 establishes conditions under which systems are considered to remain capable of performing their intended safety function when associated snubbers are not capable of providing their associated support function(s), The licensee proposed to delete LCO 3.0.8 in its entirety because all systems associated with snubbers are on systems associated with reactor operation and are no longer relevant at KPS given its permanently shutdown and defueled condition. , SR 3.0.1 establishes the requirement that SRs must be met during the modes or other specified conditions in the applicability for which the requirements of the LCO apply, unless otherwise specified in the individual SRs. This specification ensures that surveillances are performed to verify the operability of systems and components, and that variables are within specified limits. The licensee proposed to revise SR 3.0.1 to delete the reference to "MODES" since the reference to modes is no longer relevant for the permanently shutdown and defueled condition at KPS. SR 3.0.4 establishes the requirement that all applicable SRs must be met before entry into a MODE or other specified condition in the applicability.

The licensee proposed to delete the reference to "MODES" since KPS is permanently shut down and defueled and TS modes are no longer relevant.

The licensee also proposed to delete the action that refers to shutting down of the unit since KPS is already permanently shut down. The NRC staff has reviewed the proposed changes to the KPS TSs concerning LCO 3.0.1, LCO 3.0.3, LCO 3.0.4, SR 3.0.1. and SR 3.0.4 applicability and has determined that the changes . are consistent with the transition to a permanently shutdown and defueled facility.

Since 10 CFR 50.82(a)(2) prohibits the licensee from operating the plant or placing fuel in the reactor vessel, the references to "MODE," and the discussions about shutting down the unit, are no longer applicable, and the proposed changes to delete these references reflect the current KPS plant status and are appropriate. Concerning TS LCO 3.0.6, the NRC staff agrees that there will not be any systems in the permanently defueled TSs that are interrelated with other systems and that have TS LCOs (support and supported systems).

As a result, the conditions of LCO 3.0.6 no longer apply. Therefore, the staff finds that it is appropriate to delete LCO 3.0.6. Since LCO 3.0.6 is being deleted, the deletion of reference to LCO 3.0.6 in LCO 3.0.2 is also appropriate.

As discussed in Section 3.7.7 of this safety evaluation, since TS LCO 3.1.8 is being deleted, LCO 3.0.7 is no longer applicable.

Therefore, the proposed deletion of LCO 3.0.7 is acceptable.

Since LCO 3.0.7 is being deleted, the deletion of the reference to LCO 3.0.7 in LCO 3.0.1 is appropriate.

Snubbers are used on certain safety related systems covered in TS and the impact of inoperable snubbers on the TS operability of the associated system is addressed in TS 3.0.8. Because 10 CFR 50.82(a)(2) prohibits operation of the plant or placing fuel in the reactor vessel, all TS systems associated with snubbers are no longer required to be operable and are proposed for deletion from the permanently defueled TSs. As such, LCO 3.0.8 no longer applies to the proposed TSs. Therefore, the proposed deletion of LCO 3.0.8 is acceptable.

Since LCO 3.0.8 is being deleted, the deletion of reference to LCO 3.0.8 in LCO 3.0.1 is appropriate.

Based on the above, the proposed changes to LCO 3.0.1, LCO 3.0.2, LCO 3.0.3, LCO 3.0.4, LCO 3.0.6, LCO 3.0.7, LCO 3.0.8, SR 3.0.1, and SR 3.0.4 are acceptable.

3. 7. 7 Section 3.1, Reactivity Control Systems Section 3.1 of TSs, "Reactivity Control Systems," contains the LCOs, actions, and SRs that provide appropriate control of process variables, design features, or operating restrictions required to control reactivity in the reactor and, in turn, protect the integrity of a fission product barrier. This section contains the following LCOs:

• TS 3.1.1, "Shutdown Margin (SOM)"

• Applicability

-MODES 2, 3, 4, and 5 TS 3. 1.2, "Core Reactivity"

• Applicability

-MODES 1 and 2

• TS 3.1.3, "Moderator Temperature Coefficient (MTC)" *

• Applicability

-MODES 1, 2, and 3

• TS 3.1.4, "Rod Group Alignment Limits"

• Applicability

-MODES 1 and 2

• TS 3.1.5, "Shutdown Bank Insertion Limits"

• Applicability

-MODES 1 and 2

• TS 3.1.6, "Control Bank Insertion Limits"

• Applicability

-MODES 1 and 2

• TS 3.1.7, "Rod Position Indication"

• Applicability

-MODES 1 and 2 TS 3.1.8, "PHYSICS TESTS Exceptions

-MODE 2"

• Applicability

-MODE 2 The licensee proposed to delete all of the above TSs LCOs, since they are only applicable to an operating reactor and do not apply to the permanently shutdown and defueled condition of KPS. The NRC staff has reviewed these proposed changes and has determined that these TS are only needed to provide the LCOs and SRs necessary to maintain reactivity parameters of fuel loaded into a reactor vessel within the margins of conditions encountered during normal operation, anticipated occurrences, and for DBAs. The reactivity control systems TSs are only important for a reactor authorized to operate or retain irradiated fuel in the reactor vessel. However, because 10 CFR 50.82(a)(2) prohibits the licensee from operating the reactor or placing fuel in the reactor vessel, there is no reactor core and the reactivity control system are no longer relevant at KPS. The NRC staff has also reviewed the Section 3.1 reactivity control systems TSs proposed for deletion to ensure that these LCOs no longer satisfy the 10 CFR 50.36 criteria for inclusion in TSs, as described in Section 2.1 of this evaluation.

The staff notes that these TSs indicate modes for which the TS is applicable.

Modes, as defined in TSs, correspond to any one inclusive combination of core reactivity condition, power level, average reactor coolant temperature, and reactor vessel head closure bolt tensioning with fuel in the reactor vessel. The reference to modes for a permanently shutdown and defueled reactor, such as KPS, has no meaning and is not relevant.

Because DEK has submitted certifications pursuant to 10 CFR 50.82(a)(2), it is prohibited from operating the reactor or placing fuel in the reactor vessel and KPS is no longer in a configuration or a condition under which the TS modes apply. Therefore, the deletion of the Section 3.1 reactivity control TSs is acceptable.

3.7.8 Section 3.2, Power Distribution Limits Section 3.2 of TSs, "Power Distribution Limits, contains the LCOs, actions, and SRs that provide appropriate control of process variables, design features, or operating restrictions required to control power distribution in the reactor and in turn, protect the integrity of a fission product barrier. This section contains the following LCOs: TS 3.2.1, "Heat Flux Hot Channel Factor (F 0 (Z))"

• Applicability

-MODE 1 TS 3.2.2, "Nuclear Enthalpy Rise Hot Channel Factor (FrHr

• Applicability

-MODE 1 TS 3.2.3, "Axial Flux Difference (AFD)" *

• Applicability

-MODE 1

• TS 3.2.4, "Quadrant Power Tilt Ratio (QPTR)"

• Applicability

-MODE 1 The licensee proposed to delete all of the above TSs LCOs, since they are only applicable to an operating reactor and do not apply to the permanently shutdown and defueled condition of KPS. The NRC staff has reviewed these proposed changes and has determined that these TS are only needed to provide the LCOs and SRs necessary to maintain reactor power and heat generation within the margins of conditions encountered during normal operation, anticipated occurrences, and for DBAs. The power distribution limits TSs are only important for a reactor authorized to operate. However, because 10 CFR 50.82(a)(2) prohibits the licensee from operating the reactor or placing fuel in the reactor vessel, there is no reactor core generating power and the power distribution limits are no longer relevant at KPS. The NRC staff has also reviewed the Section 3.2 power distribution limits TSs proposed for deletion to ensure that these LCOs no longer satisfy the 10 CFR 50.36 criteria for inclusion in TSs, as described in Section 2.1 of this evaluation.

The staff notes that these TSs indicate modes for which the TS is applicable.

Modes, as defined in TSs, correspond to any one inclusive combination of core reactivity condition, power level, average reactor coolant temperature, and reactor vessel head closure bolt tensioning with fuel in the reactor vessel. The reference to modes for a permanently shutdown and defueled reactor, such as KPS, has no meaning and is not relevant.

Because DEK has submitted certifications pursuant to 10 CFR 50.82(a)(2), it is prohibited from operating the reactor or placing fuel in the reactor vessel and KPS is no longer in a configuration or a condition under which the TS modes apply. ' Therefore, the deletion of the Section 3.2 power distribution limit TSs is acceptable.

3.7.9 Section 3.3, Instrumentation Section 3.3 of TSs, "Instrumentation," contains the LCOs, actions, and SRs that provide for appropriate functional capability of sensing and control instrumentation required for safe operation of the facility.

This section contains the following LCOs:

• TS 3.3.1, "Reactor Protection System (RPS) Instrumentation"

• Applicability

-MODES 1, 2, 3, 4 and 5

• TS 3.3.2, "Engineered Safety Feature Actuation System (ESFAS) Instrumentation"

• Applicability

-MODES 1, 2, 3, and 4 TS 3.3.3, "Post Accident Monitoring (PAM) Instrumentation"

• Applicability

-MODES 1, 2, and 3 TS 3.3.4, "Dedicated Shutdown System"

• Applicability

-MODES 1, 2, and 3 TS 3.3.5, Previously deleted from the KPS TSs by NRC issued Amendment No. 212 dated June 9, 2014. TS 3.3.6, "Containment Purge and Vent Isolation Instrumentation"

• Applicability

-MODES 1, 2, 3, and 4 or during movement of irradiated fuel assemblies within containment

• TS 3.3.7, Previously deleted from the KPS TSs by NRC issued Amendment No. 212 dated June 9, 2014. The licensee proposed to delete all of the above TSs LCOs, since they are only applicable to an operating reactor and do not apply to the permanently shutdown and defueled condition of KPS. The NRC staff has reviewed the proposed change to TS 3.3, Instrumentation.

TS 3.3.1 is only necessary to maintain the ability of the reactor protection system to automatically initiate a reactor scram to preserve the integrity of the fuel cladding, preserve the integrity of the primary system barrier, and minimize the energy which must be absorbed, and prevent criticality following a loss-of-coolant accident (LOCA). TS 3.3.2 only concerns instrumentation designed to mitigate acCidents related to reactor operation.

TS 3.3.6 only concerns instrumentation designed to mitigate accidents related to reactor operation or during movement of irradiated fuel assemblies within containment.

TS 3.3.3 and TS 3.3.4 only concern instrumentation to ensure there is sufficient information available on selected unit parameters to monitor and assess unit status following an accident and allows operators to take manual actions specified in the emergency operating procedures.

None of the instrumentation addressed by these TSs is needed by a reactor that has permanently shutdown and defueled in accordance with 10 CFR 50.82(a)(2).

The NRC staff has also reviewed the instrumentation TSs proposed for deletion to ensure that these LCOs no longer satisfy the 10 CFR 50.36 criteria for inclusion in TSs, as described in Section 2.1 of this evaluation.

The staff notes that these TSs indicate modes for which the TS is applicable.

Modes, as defined in TSs, correspond to any one inclusive combinat.ion of core reactivity condition, power level, average reactor coolant temperature, and reactor vessel head closure bolt tensioning with fuel in the reactor vessel. The reference to modes for a permanently shutdown and defueled reactor, such as KPS, has no meaning and is not relevant.

Because DEK has submitted certifications pursuant to 10 CFR 50.82(a)(2}, it is prohibited from operating the reactor or placing fuel in the reactor vessel and KPS is no longer in a configuration or a condition under which the TS modes apply. The NRC staff has also considered the applicability of TS 3.3.6 during movement of irradiated fuel assemblies within containment.

KPS is permanently defueled with all fuel stored outside of containment.

In addition, and pursuant to 10 CFR 50.82(a)(2), KPS is prohibited from placing fuel in the reactor vessel. Therefore, there is no credible scenario that could result in movement of irradiated fuel assemblies within containment.

Based on the above, the deletion of the Section 3.3 instrumentation TSs is acceptable. 3.7.10 Section 3.4, Reactor Coolant System (RCS) The RCS TSs of Section 3.4, "Reactor Coolant System," contains the LCOs, Actions, and SRs that provide for appropriate control of process variables, design features, or operating restrictions needed for appropriate functional capability of RCS equipment required for safe operation of the facility.

This section contains the following LCOs: TS 3.4.1, "RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits"

• Applicability

-MODE 1

• TS 3.4.2, "RCS Minimum Temperature for Criticality"

• Applicability

-MODES 1 and 2

• TS 3.4.3, "RCS Pressure and Temperature (PIT) Limits"

• Applicability

-At all times

• TS 3.4.4, "RCS Loops -MODES 1 and 2"

• Applicability

-MODES 1 and 2

• TS 3.4.5, "RCS Loops -MODE 3"

• Applicability

-MODE 3

• TS 3.4.6, "RCS Loops -MODE 4"

• Applicability

-MODE 4

• TS 3.4.7, "RCS Loops -MODE 5, Loops Filled"

• Applicability

-MODE 5 with RCS Loops Filled TS 3.4.8, "RCS Loops -MODE 5, Loops Not Filled"

• Applicability

-MODE 5 RCS loops with not filled TS 3.4.9, "Pressurizer"

• Applicability

-: MODES 1, 2, and 3

• TS 3.4.10, "Pressurizer Safety Valves"

• Applicability

-MODES 1, 2, 3, and 4 TS 3.4.11, "Pressurizer Power Operated Relief Valves (PORVs)"

• Applicability

-MODES 1, 2, and 3 TS 3.4.12, "Low Temperature Overpressure Protection (L TOP) System"

• Applicability

-MODE 5 and MODE 6 when the reactor vessel head is on

• TS 3.4.13, "RCS Operational LEAKAGE"

• Applicability

-MODES 1, 2, 3, and 4

• TS 3.4.14, "RCS Pressure Isolation Valve (PIV) Leakage"

• Applicability

-MODES 1, 2, 3, and 4 (under certain conditions)

• TS 3.4.15, "RCS Leakage Detection Instrumentation"

• Applicability

-MODES 1, 2, 3, and 4 TS 3.4.16, "RCS Specific Activity"

• Applicability

-MODES 1, 2, 3, and 4 TS 3.4.17, "Steam Generator (SG) Tube Integrity"

• Applicability

-MODES 1, 2, 3, and 4 The licensee proposed to delete all of the Section 3.4, RCS TSs, since all except TS 3.4.3, are only applicable to an operating reactor and do not apply to a permanently shutdown and defueled reactor. Regarding TS 3.4.3, the licensee stated even though the applicability is "at all times," it is no longer needed at KPS and may be deleted because the reactor coolant pressure boundary will no longer be used as a fission product barrier, since the reactor vessel is permanently defueled and will no longer function as a barrier to fission product release. The NRC staff has reviewed these proposed changes and has determined that these TSs are only needed to provide the LCOs and SRs necessary to maintain functionality and integrity of the RCS pressure boundary.

These TSs contain requirements for various RCS parameters such as: thermal limitations for heatup and cooldown rates during plant operation in order to operate within the analyzed requirements for stress intensity and fatigue limits for the reactor vessel; pressurization, which established and maintained an equilibrium under saturated conditions for pressure control to prevent bulk boiling in the remainder of the RCS; coolant chemistry, which included limits on RCS activity to limit potential offsite doses due to postulated events and limits on RCS conductivity, chlorides, and pH to prevent stress-corrosion cracking; coolant leakage, which established primary system leakage limits to allow prompt identification and isolation of leaks before the integrity of the RCS pressure boundary was impaired; safety and relief valves, which specifies operability requirements for the safety and relief valves designed to prevent overpressurization of, and damage to, the primary system boundary; and structural integrity, which addresses the inservice inspection requirements of the primary system boundary components.

All of these TS are related to assuring the integrity of the RCS pressure boundary.

The RCS TSs are only important for a reactor authorized to operate or retain irradiated fuel in the reactor vessel. However, because 10 CFR 50.82(a)(2) prohibits the licensee from operating the reactor or placing fuel in the reactor vessel, the reactor coolant system is no longer functional or used in any capacity at KPS. The NRC staff has also reviewed the RCS TSs proposed for deletiq__n to ensure that these LCOs no longer satisfy the 10 CFR 50.36 criteria for inclusion in TSs, as described in Section 2.1 of this evaluation.

The staff notes that these TSs indicate modes for which the TS is applicable.

Modes, as defined in TSs, correspond to any one inclusive combination of core reactivity condition, power level, average reactor coolant temperature, and reactor vessel head closure bolt tensioning with fuel in the reactor vessel. The reference to modes for a permanently shutdown and defueled reactor, such as KPS, has no meaning and is not relevant.

Because DEK has submitted certifications pursuant to 10 CFR 50.82(a)(2), it is prohibited from operating the reactor or placing fuel in the reactor vessel and, therefore, KPS is no longer in a configuration or a condition under which the TS modes apply. Furthermore, because irradiated fuel has been permanently removed from the reactor pressure vessel, the RCS is no longer relevant as a fission product barrier. In addition, regarding the applicability of TS 3.4.3 at all times, the staff notes that the RCS is drained and vented, to the extent possible, and consequently there is no longer any concern about exceeding the RCS pressure and temperature limits specified in TS 3.4.3. Based on the above, the deletion of Section 3.4 RCS TSs is acceptable. 3.7.11 Section 3.5, Emergency Core Cooling Systems (ECCS) Section 3.5 of TSs, "Emergency Core Cooling Systems," contains the LCOs, Actions, and SRs that provide for appropriate functional capability of ECCS equipment required for mitigation of design basis accidents or transients so as to protect the integrity of fission product barriers.

This section contains the following LCOs: TS 3.5.1, "Accumulators"

• Applicability

-MODES 1, 2, and 3

• TS 3.5.2, "ECCS -Operating"

• Applicability

-MODES 1, 2, and 3 TS 3.5.3, "ECCS -Shutdown"

• Applicability

-MODE 4

• TS 3.5.4, "Refueling Water Storage Tank (RWST)"

• Applicability

-MODES 1, 2, 3, and 4 The licensee proposed to delete all of the above TSs LCOs, since they are only applicable to an operating reactor and do not apply to the permanently shutdown and defueled condition of KPS. The NRC staff has reviewed the proposed changes to the ECCS TSs and has determined that these TS are only needed to provide the LCOs and SRs necessary to maintain functionality of the systems that provide emergency cooling to the reactor core. These TSs includes multiple LCOs addressing the accumulators, safety injection

{high head), and residual heat removal (low head) subsystems, part of the ECCS, designed to provide adequate emergency cooling capability to the reactor in the event of a LOCA; and the refueling water storage tank designed to supply borated water to the ECCS during accident conditions.

All of these TS are related to providing cooling for a reactor vessel core. Since KPS is permanently shut down and defueled, there are no accidents of any kind that would require emergency core cooling and the accidents these systems and components were designed to mitigate are no longer possible.

The NRC staff reviewed the ECCS TSs proposed for deletion to ensure that these LCOs no longer satisfy the 1 O CFR 50.36 criteria for inclusion in TSs, as described in Section 2.1 of this evaluation.

The staff notes that these TSs indicate modes for which the TS is applicable.

Modes, as defined in TSs, correspond to any one inclusive combination of core reactivity condition, power level, average reactor coolant temperature, and reactor vessel head closure bolt tensioning with fuel in the reactor vessel. The reference to modes for a permanently shutdown and defueled reactor, such as KPS, has no meaning and is not relevant.

Because DEK has submitted certifications pursuant to 1 O CFR 50.82(a)(2), it is prohibited from operating the reactor or placing fuel in the reactor vessel and KPS is no longer in a configuration or a condition under which the TS modes apply. Based on the above, the deletion of the Section 3.5 ECCS TSs is acceptable. 3.7.12 Section 3.6, Containment Systems Section 3.6 of TSs, "Containment Systems," contains the LCOs, actions, and SRs that provide for appropriate control of process variables, design features, or operating restrictions required to protect the integrity of the containment as a fission product barrier; and appropriate functional capability of engineered safety features (ESF) equipment required for mitigation of DBAs or transients so as to protect the integrity of containment.

This section contains the following LCOs: ' *

• TS 3.6.1, "Containment"

• Applicability

-MODES 1, 2, 3, and 4

• TS 3.6.2, "Containment Air Locks"

• Applicability

-MODES 1, 2, 3, and 4

• TS 3.6.3, "Containment Isolation Valves"

• Applicability

-MODES 1, 2, 3, and 4

• TS 3.6.4, "Containment Pressure"

• Applicability

-MODES 1, 2, 3, and 4

• TS 3.6.5, "Containment Air Temperature"

• Applicability

-MODES 1, 2, 3, and 4

• TS 3.6.6, "Containment Spray and Cooling Systems"

• Applicability

-MODES 1, 2, 3, and 4

• TS 3.6.7, "Spray Additive System"

• Applicability

-MODES 1, 2, 3, and 4 TS 3.6.8, "Shield Building"

• Applicability

-MODES 1, 2, 3, and 4

• TS 3.6.9, "Vacuum Relief Valves"

• Applicability

-MODES 1, 2, 3, and 4

• TS 3.6.10, "Shield Building Ventilation System (SBVS)"

• Applicability

-MODES 1, 2, 3, and 4 The licensee proposed to delete all of the above TS LCOs, since they are only applicable to an operating reactor and do not apply to the permanently shutdown and defueled condition of KPS. The NRC staff has reviewed these proposed changes and has determined that these TS are only needed to provide the LCOs and SRs necessary to maintain functionality of the containment.

These TSs includes multiple LCOs addressing containment integrity including:

containment pressure, containment air temperature, and containment air locks, which forms part of the containment pressure boundary, and containment isolation valves, designed to isolate the containment in the event of a LOCA to prevent the release of fission products to the atmosphere; containment spray and cooling, which limit post-accident pressure and temperature in containment; spray additive system, which assists in reducing the iodine fission product inventory in the containment atmosphere; shield building, which surrounds the containment vessel and collects possible containment leakage following a LOCA; vacuum relief valves, which protect the containment vessel against negative pressure;-

and shield building ventilation, which reduces the radioactive content in the shield building following a design basis accident.

However, 10 CFR 50.82(a)(2) prohibits the licensee from operating the plant or placing fuel in the reactor vessel. Consequently, the operational conditions, transients, and accidents the containment SSCs were designed to contain are no longer possible.

The NRC staff has also reviewed the containment systems TSs proposed for deletion to ensure that these LCOs no longer satisfy the 10 CFR 50.36 criteria for inclusion in TSs, as described in Section 2.1 of this evaluation.

The staff notes that these TSs indicate modes for which the TS is applicable.

Modes, as defined in TSs, correspond to any one inclusive combination of core reactivity condition, power level, average reactor coolant temperature, and reactor vessel head closure bolt tensioning with fuel in the reactor vessel. The reference to modes for a permanently shutdown and defueled reactor, such as KPS, has no meaning and is not relevant.

Because DEK has submitted certifications pursuant to 10 CFR 50.82(a)(2), it is prohibited from operating the reactor or placing fuel in the reactor vessel and KPS is no longer in a configuration or a condition under which the TS modes apply. Based on the above, the deletion of Section 3.6 containment systems TSs is acceptable.

3.7.13 Section 3.7, Plant Systems Section 3.7 of TSs, "Plant Systems," contains the LCOs, actions, and SRs that provide for appropriate functional capability of plant equipment required for safe operation of the facility.

This section contains the following LCOs: TS 3.7.1, "Main Steam Safety Valves (MSIVs)"

• Applicability

-MODES 1, 2, and 3

• TS 3.7.2, "Main Steam Isolation Valves (MSIVs)"

• Applicability

-MODES 1, 2, and 3

• TS 3.7.3, "Main Feedwater Isolation Valves (MFIVs), Main Feedwater Regulation Valves (MFRVs), and MFRV Bypass Valves"

• Applicability

-MODES 1, 2, and 3 TS 3.7.4, "Steam Generator (SG) Power Operated Relief Valves (PORVs)"

• Applicability

-MODES 1, 2, 3, and 4 TS 3.7.5, "Auxiliary Feedwater (AFW) System"

• Applicability

-MODES 1, 2, 3, and 4 TS 3_7.6, "Condensate Storage Tanks (CSTs)"

• Applicability

-MODES 1, 2, 3, and 4 TS 3.7.7, "Component Cooling (CC) System"

• Applicability

-MODES 1, 2, 3, and 4 TS 3. 7 .8, "Service Water (SW) System"

• Applicability

-MODES 1, 2, 3, and 4

• TS 3.7.9, "Ultimate Heat Sink (UHS)"

• Applicability

-MODES 1, 2, 3, and 4 0 TS 3.7.10, Previously deleted from the KPS TSs by NRC issued Amendment No. 212 dated June 9, 2014.

• TS 3.7.11, Previously deleted from the KPS TSs by NRC issued Amendment No. 212 dated June 9, 2014.

• TS 3. 7 .12, "Auxiliary Building Special Ventilation (ASV) System"

• Applicability

-MODES 1, 2, 3, and 4

• TS 3.7.13, "Spent Fuel Pool Water Level"

• Applicability

-During movement of irradiated fuel assemblies in the spent fuel pool

• TS 3.7.14, "Spent Fuel Pool Boron Concentration"

• Applicability

-When fuel assemblies are stored in the spent fuel pool

• TS 3. 7 .15, "Spent Fuel Pool Storage

• Applicability

-When any fuel assembly is stored in the spent fuel pool

• TS 3. 7 .16, "Secondary Specific Activity"

• Applicability

-MODES 1, 2, 3, and 4 The licensee proposed to delete TS 3.7.1 though TS 3.7.9, TS 3.7.12, and TS 3.7.16, since these TSs are mode dependent and only applicable to an operating reactor. Therefore, these TSs do not apply to the* permanently shutdown and defueled condition of KPS. The licensee intends to retain TS 3. 7 .13, TS 3. 7 .14, and TS 3. 7 .15 but revise these TSs to delete the REQUIRED ACTIONS note that states that LCO 3.0.3 is not applicable.

LCO 3.0.3 is being deleted from the KPS TSs and removal of a reference to TS 3.0.3 is a conforming change. The licensee also proposed to delete reference to new fuel assemblies from LCO 3. 7 .15. b. New fuel is no longer stored onsite and License Condition 2.B.(2) is being revised to no longer allow receipt of new fuel. Therefore, the phrase "New fuel assemblies" is no longer applicable and may be deleted. The NRC staff has reviewed the proposed changes to TSs 3.7.1 though TS 3.7.9, TS 3.7.12, and TS 3.7.16 and has determined that these TS are only necessary to assure the operability of certain plant systems during reactor operation.

These TSs involve: main steam safety valves, which provide overpressure protection for the secondary system; main steam isolation valves, which isolate steam flow from the secondary side of the steam generator following a main steam line break; main feedwater isolation valves, MFRVs, and MFRV bypass valves, which isolate main feedwater flow to the secondary side of the steam generators following a high energy line break; steam generator power operated relief valves, which provide a method for cooling the unit should the condenser not be available; auxiliary feedwater system, which supplies feedwater to the steam generators upon the loss of the normal feedwater supply; condensate storage tanks, which provide the preferred source of water to the steam generators for removing decay and sensible heat from the RCS; component cooling system, service water system and ultimate heat sink, which provide a heat sink for the removal of process and operating heat from safety related components during a OBA or transient; auxiliary building special ventilation system, which filters potential radiological leakage that bypasses containment (into the auxiliary building) following a loss of coolant accident; and secondary specific activity, which specifies a limit on secondary coolant specific activity during power operation.

The above TSs are intended to protect the reactor and main turbine from potential operational transients.

However, 10 CFR 50.82(a)(2) prohibits the licensee from operating the reactor or placing fuel in the reactor vessel. Consequently, the operating, transient, and accident conditions these systems and components were designed to protect against are no longer possible.

The NRC staff has also reviewed the plant systems TS 3.7.1 though TS 3.7.9, TS 3.7.12, and TS 3.7.16, which have been proposed for deletion, to ensure that these LCOs no longer satisfy the 10 CFR 50.36 criteria for inclusion in TSs, as described in Section 2.1 of this evaluation.

The staff notes that these TSs indicate modes for which the TS is applicable.

Modes, as defined in TSs, correspond to any one inclusive combination of core reactivity condition, power level, average reactor coolant temperature, and reactor vessel head closure bolt tensioning with fuel in the reactor vessel. The reference to modes for a permanently shutdown and defueled reactor, such as KPS, has no meaning and is not relevant.

Because DEK has submitted certifications pursuant to 10 CFR 50.82(a)(2), it is prohibited from operating the reactor or placing fuel in the reactor vessel and KPS is no longer in a configuration or a condition under which the TS modes apply. Based on the above, the deletion of TSs 3.7.1 though TS 3.7.9, TS 3.7.12, and TS 3.7.16 is acceptable.

The NRC staff has reviewed the proposed deletion of the reference to LCO 3.0.3 in the required actions note in TS 3.7.13, TS 3.7.14, and TS 3.7.15 and has determined that it is consistent with the transition to a permanently defueled facility.

Since the licensee has permanently defueled the facility, 10 CFR 50.82(a)(2) prohibits the licensee from operating the plant or placing fuel in

• the reactor vessel. In addition, deletion of the required actions note that states, "LCO 3.0.3 is not applicable," is appropriate since LCO 3.0.3 is being deleted (see Section 3.7.6 of this safety evaluation).

Therefore, the staff finds the deletion of reference to LCO 3.0.3 in TS 3.7.13, TS 3.7.14, and TS 3.7.15 acceptable.

The NRC staff has reviewed the proposed change to TS 3.7.15 to delete reference to new fuel assemblies from LCO 3.7.15.b.

The staff notes that, License Condition 2.B.(2) is being revised to no longer allow receipt of new fuel (see Section 3.18 of this safety evaluation).

Since the facility license will no longer allow new fuel to be stored onsite, the phrase, "new fuel assemblies," is no longer applicable in TS 3.7.15. Based on the above, the staff finds the deletion of new fuel assemblies from LCO 3.7.15.b to be acceptable.

3.7.14 Section 3.8, Electrical Power Systems Section 3.8 of TSs, Electrical Power Systems," contains the LCOs, actions, and SRs related to the availability of alternating current (AC) and direct current (DC) electrical power supplies and distribution systems. This section contains the following LCOs: TS 3.8.1, "AC Sources -Operating"

• Applicability

-MODES 1, 2, 3 and 4 TS 3.8.2, Previously deleted from the KPS TSs by NRC issued Amendment No. 212 dated June 9, 2014

• TS 3.8.3, Previously deleted from the KPS TSs by NRC issued Amendment No. 212 dated June 9, 2014

• TS 3.8.4, "DC Sources -Operating"

• Applicability

-MODES 1, 2, 3 and 4 TS 3.8.5, Previously deleted from the KPS TSs by NRC issued Amendment No. 212 dated June 9, 2014

• TS 3.8.6, Previously deleted from the KPS TSs by NRC issued Amendment No. 212 dated June 9, 2014 TS 3.8.7, "Inverters

-Operating"

• Applicability

-MODES 1, 2, 3 and 4

• TS 3.8.8, Previously deleted from the KPS TSs by NRC issued Amendment No. 212 dated June 9, 2014 TS 3.8.9, "Distribution Systems -Operating"

• Applicability

-MODES 1, 2, 3 and 4

• TS 3.8.10, Previously deleted from the KPS TSs by NRC issued Amendment No. 212 dated June 9, 2014 The licensee proposed to delete electrical power systems TS 3.8.1, TS 3.8.4, TS 3.8. 7, and TS 3.8.9, since these TSs are mode dependent and only applicable to an operating reactor. Therefore, these TSs do not apply to the permanently shutdown and defueled condition of KPS. The NRC staff has reviewed the electrical power systems TS 3.8.1, TS 3.8.4, TS 3.8.7, and TS 3.8.9, which have been proposed for deletion, to ensure that these LCOs no longer satisfy the 10 CFR 50.36 criteria for inclusion in TSs, as described in Section 2.1 of this evaluation.

The staff notes that these TSs indicate modes for which the TS is applicable.

Modes, as defined in TSs, correspond to any one inclusive combination of core reactivity condition, power level, average reactor coolant temperature, and reactor vessel head closure bolt tensioning with fuel in the reactor vessel. The reference to modes for a permanently shutdown and defueled reactor, such as KPS, has no meaning and is not relevant.

Because DEK has submitted certifications pursuant to 1 O CFR 50.82(a)(2), it is prohibited from operating the reactor or placing fuel in the reactor vessel and KPS is no longer in a configuration or a condition under which the TS modes apply. The NRC staff notes that electrical power systems are not included in the proposed permanently defueled TSs since there are no TS requirements for components or systems that need electrical power: There are no active systems credited as part of the initial conditions of an analysis or as part of the primary success path for mitigation of the design bal5iS accidents that are credible with the KPS permanently shutdown and defueled.

The only electrically powered active system important for the storage of irradiated fuel is the SFP cooling system and support systems. The SFP cooling system did not meet the criteria in 10 CFR 50.36 for inclusion in the KPS TSs even when the reactor was authorized to operate. Based on the above, the deletion of TS 3.8.1, TS 3.8.4, TS 3.8.7, and TS 3.8.9 in Section 3.8, "Electrical Systems," is acceptable.

3.7.15 Section 3.9, Refueling Operations Section 3.9 of TSs, "Refueling Operations," contains the LCOs, actions, and SRs related to refueling operations.

This section contains the following LCOs:

• TS 3.9.1, "Boron Concentration"

• Applicability

-MODE 6

• TS 3.9:2, "Nuclear Instrumentation"

• Applicability

-MODE 6 TS 3.9.3, "Residual Heat Removal (RHR) and Coolant Circulation

-High Water Level"

• Applicability

-MODE 6

• TS 3.9.4, "Residual Heat Removal (RHR) and Coolant Circulation

-Low Water Level"

• Applicability

-MODE 6

• TS 3.9.5, "Refueling Cavity Water Level"

• Applicability

-During movement of irradiated fuel assemblies in containment

• TS 3.9.6, "Containment Penetrations"

• Applicability

-During movement of irradiated fuel assemblies in containment The licensee proposed to delete all of the above TSs LCOs, since these LCOs are only applicable for modes where the reactor is authorized to operate or store fuel in the reactor vessel. These modes do not apply to the permanently shutdown and defueled condition of KPS. The NRC staff has reviewed these proposed changes and has determined that these TS are only needed to assure the operability of plant systems required for refueling operations.

These TSs involve: boron concentration, which limits the boron concentration of the RCS, the fuel transfer canal, and the refueling cavity during refueling; nuclear instrumentation, which monitors the core reactivity condition during refueling operations; residual heat removal and coolant circulation

-high and low water level, which removes decay heat and sensible heat from the RCS, provides mixing of borated coolant, and prevents boron stratification; refueling cavity water level, which specifies a minimum water level of 23 feet above the top of the reactor vessel flange during movement of irradiated fuel assemblies within containment; and containment penetrations, which specifies requirements for containment closure during the conduct of refueling operations.

However, 1 O CFR 50.82(a)(2) prohibits the from operating the plant or placing fuel in the reactor vessel. Therefore, refueling operations are no longer permitted at KPS and the LCOs in TS Section 3.9 are no longer relevant.

The NRC staff has also reviewed the refueling operations TSs proposed for deletion to ensure that these LCOs no longer satisfy the 10 CFR 50.36 criteria for inclusion in TSs, as described in Section 2.1 of this evaluation.

The staff notes that these TSs indicate modes for which the TS is applicable.

Modes, as defined in TSs, correspond to any one inclusive combination of core reactivity condition, power level, average reactor coolant temperature, and reactor vessel head closure bolt tensioning with fuel in the reactor vessel. The reference to modes for a permanently shutdown and defueled reactor, such as KPS, has no meaning and is not relevant.

Because DEK has submitted certifications pursuant to 10 CFR 50.82(a)(2), it is prohibited from operating the reactor or placing fuel in the reactor vessel and KPS is no longer in a configuration or a condition under which the TS modes apply. Based on the above, the deletion of Section 3.9 refueling operations TSs is acceptable.

3.7.16 Section 4.0, Design Features This section contains a description of the design features of the facility.

Site Location The licensee proposed to retain this section in TSs without modification or revision.

Reactor Core This specification provides a general description of the number of and design mate.rial requirements for the fuel and control rod assemblies used in the reactor core. The licensee proposed to delete the design feature descriptions for fuel and control rod assemblies, since they are only applicable to an operating reactor and do not apply to the permanently shutdown and defueled condition of KPS. The NRC staff has reviewed the proposal to delete the reactor core fuel and control rod assemblies design features from TSs. Since 10 CFR 50.82(a)(2) prohibits the licensee from operating the reactor or placing fuel in the reactor vessel, the design features related to the reactor core fuel assemblies and control rods are no longer relevant at KPS. Therefore, the staff finds that the proposed deletion of the reactor core design feature acceptable.

Fuel Storage This specification provides a general description of the new and spent fuel storage racks and certain SFP design features.

The licensee proposed to delete the description of the new fuel storage rack design features in TS 4.3.1.2 since new fuel is no longer stored onsite and License Condition 2.B.(2) is being revised to no longer allow receipt of new fuel. The NRC staff has reviewed the proposed changes to remove the new fuel storage rack design features from TSs. The licensee currently has no new fuel stored onsite. Furthermore, since the licensee has permanently shut down and defueled the reactor pursuant to 10 CFR 50.82(a)(2), the licensee will no longer be receiving any new fuel onsite. In addition, License Condition 2.B.(2) is being revised to no longer allow receipt of new fuel. Since the facility license will no longer allow new fuel to be stored onsite, the requirements for new fuel storage racks are no longer applicable.

Based on the above, the staff finds the proposed changes to delete the new fuel storage rack design features from TS 4.3.1.2 to be acceptable.

3.7.17 Section 5.0, Administrative Controls The NRC staff reviewed the proposed changes to Section 5.0 of the KPS TSs as it relates to the guidance in the regulatory evaluation in Section 2.1 of this safety evaluation, as appropriate and

• applicable, in the following areas: Responsibility; Organization; Procedures, Programs and Manuals; and Reporting Requirements.

3. 7 .17 .1 TS 5.1: Responsibility TS 5.1.1 The current TS 5.1.1 states, in part, that: "The plant manager shall be responsible for overall unit operation and shall delegate in writing the succession to this responsibility during his absence." The licensee proposed the following change to TS 5.1.1 to address the permanently defueled and decommissioning status of the plant: The plant manager shall be responsible for overall plant operation and shall delegate in writing the succession to this responsibility during his absence. The NRC staff reviewed the proposed changes to TS 5.1.1 and found that the changes modified the scope of the position responsibility from "unit operation" to "plant operation." The staff finds that this change is acceptable because the overall management responsibilities are unchanged, and the description of the plant and the related responsibilities of the plant staff discussed in this technical specification have been updated to reflect that the reactor has been permanently shut down. TS 5.1.2 TS 5.1.2 identifies the responsibilities for the control room command function for the various operating modes, and describes delegation of authority requirements for the position.

The current TS states that: 'The shift manager shall be responsible for the control room command function.

During any absence of the shift manager from the control room while the unit is in MODE 1, 2, 3, or 4, an individual with an active Senior Operator license shall be designated to assume the control room command function.

During any absence of the shift manager from the CR while the unit is in MODE 5 or 6, an individual with an active Senior Operator license or Operator license shall be designated to assume the control room command function." The licensee proposed to change TS 5.1.2 to state: The shift manager shall be responsible for the shift command function.

As part of the amendment request, the licensee proposed to eliminate the mode dependency of the command function delegation requirements currently in the KPS TSs based on the permanently shutdown and defueled condition of the reactor. The NRC staff found that the KPS TSs define mode as corresponding to any one inclusive combination of core reactivity condition, power level, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1. Table 1.1-1 only applies with fuel in the reC!ctor vessel; therefore, due to the permanently defueled state of the reactor, operating modes are no longer applicable to KPS, and should not be used to determine delegation of authority decisions.

The licensee also proposed to change the command function for the shift manager from "control room command function"*to "shift command function." This change is editorial and is acceptable.

The proposed change also eliminates the requirements regarding delegation of authority of the shift command function when the shift manager leaves the control room (CR). As part of the shift command function, the shift manager is the facility emergency director and the CR is the location for emergency command and control (in accordance with the Kewaunee permanently defueled emergency plan). The staff notes that proposed TS 5.1.2 would permit the shift manager the flexibility to leave the control room while retaining his emergency director responsibility.

In addition, the proposed permanently defueled TS 5.2.2 does not require any other operations staff to be present in the CR. The NRC staff issued and RAI to understand how abnormal conditions (such as off-normal radiation level, fire, electrical anomalies, etc.), or emergencies, would be detected in the absence of any CR staffing.

The NRC staff requested DEK to provide information describing how operations staff would maintain awareness of plant conditions normally ascertained with control room instrumentation and alarms. The NRC staff also requested DEK to provide an assessment that demonstrates the permanently defueled emergency plan can be implemented, assuming both the shift manager and/or the non-certified operator are absent from the control room, without a reduction in effectiveness when compared to operations staff being present in the CR. DEK responded to the staff's request for additional information in a letter dated October 6, 2014, as supplemented by a letter dated January 15, 2015, indicating that DEK has implemented a *communications system modification, carried by both the shift manager and the non-certified operator, which provides mobile alarm notifications containing the same information as the control room annunciator system. All parameters that provide inputs for assessing the permanently defueled Emergency Action Level (EALs) are sent to these mobile electronic devices. In addition to potential EAL alarms, the device receives fire alarms, SFP trouble alarms, electrical alarms, and general alarms. The licensee states that besides this new mobile alarm notification system, the shift manager and the non-certified operator have alternate communications capability available with all other staff onsite using cellular communications devices and the Gai-Tronic public address system. When out of the control room, any calls to the operations staff are forwarded to their cellular phones. If the control room is unoccupied by operations staff when an ENS call is received from the NRC, the central alarm station operator will answer the call and immediately notify the shift manager via a continuous communications system capability DEK has also established between the security alarm stations and the shift manager regardless of the shift manager's location (see NRC exemption dated December 29, 2014; ADAMS Accession No. ML 14217A228)

In its October 6, 2014, submittal, DEK stated that shift manager and the non-certified operator are procedurally required to remain within defined areas while on shift. DEK has performed an assessment that determined the shift manager can return to the CR within 6 minutes from any authorized location onsite. The non-certified operator can return to the CR within 12 minutes from any authorized location onsite. The licensee states that there is ample time for the shift manager to return to the control room when an alarm is received on the mobile alarm notification system (or when contacted by alternate communications methods), assess the situation, and take appropriate actions including review and declaration of an emergency event within the allowed time of 30 minutes specified in the permanently defueled emergency plan. The NRC staff has concluded that because of the communications capability available to the shift manager, and the ability of the shift manager to quickly return to the CR, the shift manager can fulfill his shift command function and emergency response duties even when absent from the CR. Remaining in the CR is unnecessary since the shift manager constantly maintains awareness of plant conditions through a mobile alarm notification system and can make initial command decisions wherever he is located at the facility and still have ample time to return to the CR to complete his command functions.

The NRC staff also considered the licensee's assessment of the effectiveness of the operations staff to conduct emergency planning functions in accordance with the KPS permanently defueled emergency plan when operations staff are located outside of the control room. The NRC staff has concluded that if the shift manager was outside of the CR when an event occurred (even with the CR unoccupied and factoring in the maximum expected six-minute return time to the CR), there is sufficient margin to ensure review of and declaration of an emergency classification level would still be completed within the maximum allowed time of 30 minutes specified in the KPS permanently defueled emergency plan. Therefore, the staff finds that absence of the shift manager from the control room at KPS does not prevent the licensee from implementing its defueled emergency plan. Based on the reasons stated above, the NRC staff finds the changes to TS 5.1.2 acceptable.

3. 7 .17 .2 TS 5.2: Organization TS 5.2.1 -Onsite and Offsite Organizations The current TS states, in part, in the introduction to TS 5.2.1 that: "Onsite and offsite organizations sh.all be established for unit operation and corporate management, respectively.

The onsite and offsite organizations shall include the positions for activities affecting safety of the nuclear power plant." The licensee proposed the following changes to the TS 5.2.1 opening paragraph:

Onsite and offsite organizations shall be established for plant operation and corporate management, respectively.

The onsite and offsite organizations shall include the positions for activities affecting safety of the nuclear fuel. The NRC staff reviewed the introduction to TS 5.2.1 and found that it has been changed to reflect that the plant is permanently defueled.

As such, the introduction has been modified to be appropriate for activities associated with a decommissioning reactor. The NRC staff has reviewed the changes and finds them to be acceptable based on the plant's permanently shutdown and defueled status. TS 5.2.1.b The current TS states that: "The plant manager shall be responsible for overall safe operation of the plant and shall have control over those onsite activities necessary for safe operation and maintenance of the plant." The licensee proposed the following change to TS 5.2.1.b: The plant manager shall be responsible for overall safe operation of the plant and shall have control over those onsite activities necessary for safe storage and maintenance of the nuclear fuel. . The NRC staff reviewed the proposed revision to TS 5.2.1.b and finds that it reflects the permanently shutdown and defueled condition of the site with the safety focus on the .storage of spent fuel and the control of onsite activities.

As such, the proposed modification to this TS is appropriate for activities associated with a decommissioning reactor. Therefore, the proposed changes are acceptable.

TS 5.2.1.c The current TS states that: "A specified corporate officer shall have corporate responsibility for overall plant nuclear safety and shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining, and providing technical support to the plant to ensure nuclear safety." The licensee proposed the following changes to TS 5.2.1.c: A specified corporate officer shall have corporate responsibility for overall plant nuclear safety and shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining, and providing technical support to the plant to ensure safe management of nuclear fuel. The NRC staff reviewed the proposed revision to TS Section 5.2.1.c and finds that that it reflects the permanently shutdown and defueled condition of the site with the safety focus on the storage of spent fuel and the control of onsite activities.

As such, the proposed modification to this TS is appropriate for activities associated with a decommissioning reactor. Therefore, the proposed changes are acceptable.

TS 5.2.1.d The current TS states that: "The individuals who train the operating staff, carry out health physics, or perform quality assurance functions, may report to the appropriate onsite manager; however, they shall have sufficient organizational freedom to ensure their independence from operating pressures." The licensee proposed the following changes to TS 52.1.d: The individuals who train the CERTIFIED FUEL HANDLERS, carry out health physics or perform quality assurance functions may report to the appropriate onsite manager; however, they shall have sufficient organizational freedom to ensure their ability to perform their assigned functions.

The purpose of TS 5.2.1.d is to ensure that staff that train critical staff members, carry out health physics or perform quality assurance functions, can do so without being subject to pressure from schedule or budget. Prior to decommissioning, this included operations staff trainers.

The licensee proposed to change the TS from "operating staff' to "Certified Fuel Handlers," and from "independence from operating pressures" to "perform their assigned functions," to reflect the decommissioning status of the plant. The NRC staff finds that this is acceptable because the changes more accurately describe the requirements as a result of the plant's decommissioning status. 5.2.2 TS Unit Staff The licensee proposed to change the title of TS 5.2.2 from "Unit Staff' to "Facility Staff." In addition, the current introduction states: "The unit staff organization shall include the following:" The revised introduction would state: The facility staff organization shall include the following:

The NRC staff reviewed the proposed revision to TS 5.2.2 and concludes that changing the word "unit" to "facility" is editorial and reflects the permanent shutdown and defueled condition of the plant. Therefore, the staff finds the proposed change acceptable. TS 5.2.2.a. and b. -Unit Staff This TS 9iscusses the requirements for unit staff organization.

TS 5.2.2, paragraph a., currently states: "A non-licensed operator shall be assigned if the reactor contains fuel and an additional licensed operator shall be assigned if the reactor is operating in MODES 1, 2, 3, or 4." TS 5.2.2, paragraph b., currently states: "Shift crew composition may be less than the minimum requirement of 10 CFR 50.54(m)(2)(i) and 5.2.2.a and 5.2.2.e for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, except in severe weather conditions, in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements." Ttie licensee proposed to modify TS 5.2.2.a and TS 5.2.2.b as follows: a. Each on duty shift shall be composed of at least the minimum shift crew composition shown in Table 5.2.2.-1 b. Shift crew composition may be less than the minimum requirement of Table 5.2.2-1 for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements.

Table 5.2.2-1 Minimum Shift Crew Composition POSITION MINIMUM STAFFING CERTIFIED FUEL HANDLER 1 Non-Certified Operator 1 Note: Non-Certified Operator position may be filled by a CERTIFIED FUEL HANDLER. The NRC staff reviewed the proposed changes to TS 5.2.2.a. and b. to ensure that the changes reflect the scope of activities resulting from the permanent cessation of operations.

With the certifications submitted in accordance with 10 CFR 50.82(a)(2), the licensee is no longer authorized to operate the reactor or load fuel into the reactor vessel. As discussed in NRC letter dated May 21, 2014 (ADAMS Accession No. ML 14127A340), the requirements of 10 CFR 50.54(m) addressing licensed operator.staffing under various modes of operation no longer apply to facilities that have submitted certifications in accordance with 10 CFR 50.82(a)(2).

Therefore, proposed changes that remove requirements for licensed operators are acceptable.

In addition, the existing TSs indicate modes for which the TS is applicable.

Modes, as defined in TSs, correspond to any one inclusive combination of core reactivity condition, power level, average reactor coolant temperature, and reactor vessel head closure bolt tensioning with fuel .in the reactor vessel. The reference to modes for a'permanently shutdown and defueled reactor, such as KPS, has no meaning and is not relevant.

Because DEK has submitted certifications pursuant to 10 CFR 50.82(a)(2), it is prohibited from operating the reactor or placing fuel in the reactor vessel and KPS is no longer in a configuration or a condition under which the TS modes apply. The NRC staff, therefore, finds that proposed changes removing mode-dependent actions are also acceptable.

The NRC staff found that the proposed minimum crew complement is consistent with shift manning requirements for permanently shutdown sites with single SFPs. However, this TS differs from other permanently shutdown TSs in that there are no constraints that operations staff. must remain in the CR. The proposed TS would require a Certified Fuel Handler (CFH) [note that at KPS, the shift manager is a CFH] and a non-certified operator be on duty, but allow both to perform their duties outside the CR without a requirement that either remain in the control room. In its letter dated October 6, 2014, DEK discussed its basis for not specifying any TS constraints on staffing the CR at KPS. The licensee described the difference between the functions of the CR when operating and after the reactor was permanently shutdown and defueled.

Specifically, DEK stated that the CR at an operating reactor contains the controls and instrumentation necessary for complete supervision and response needed to ensure safe operation of the reactor and support systems during normal, off-normal, and accident conditions and, therefore, is the location of the shift command function.

Following permanent shutdown and removal of fuel from the reactor, operation of the reactor is no longer permitted and the CR no longer performs all the functions required while operating the reactor. There are no longer any activities at a permanently shutdown and defueled reactor that require a quick decision and response by operations staff in the CR. Regarding SFP safety, the licensee noted that control of SFP operations does not require a presence in the CR and resolution of most operational issues related to SFP function are more readily accomplished locally at the SFP. Because of the robust design of the SFP and the low decay heat load of the stored fuel, events involving the SFP are expected to evolve slowly and provide ample time for staff to return to the CR if needed. DEK states the primary functions of the CR at a permanently shutdown plant are monitoring, response, communications, and coordination.

The KPS CR provides a central location from where the shift command function can be conveniently performed because of the availability of existing communication systems and equipment, office computer equipment, and ready access to reference material.

The CR also provides a central location from which emergency response activities are coordinated.

Since many of the remaining system processes at a permanently shutdown and defueled reactor are controlled locally, DEK states that allowing the operating crew to leave the CR and to access other plant areas will optimize the efficiency of the shift crew in performing routine activities.

The licensee states that continuous presence of the shift crew in the CR is no longer necessary in order to effectively accomplish decommissioning activities, monitoring of spent fuel, and overall oversight of facility safety. As discussed in the NRC staff's evaluation of TS 5.1.2, the NRC staff requested additional information describing how operations staff would maintain awareness of plant conditions normally ascertained from the CR using control room instrumentation and alarms if no one was present *in the CR In DEK's October 6, 2014 RAI response, and as discussed in the NRC staff's evaluation of TS 5.1.2, DEK has a mobile alarm notification system that allows both the shift manager/CFH and the non-certified operator to retain situational awareness of the plant conditions at any authorized onsite location.

The licensee also described other diverse communications systems, such as cellular phones and the plant paging system, to ensure that the operations staff can communicate with both onsite personnel and offsite as effectively as in the CR 1 The NRC staff assessed the proposed TS 5.2.2 considering the absence of operations staff from the CR Based on the information provided in the licensee's supplemental letter dated October 6, 2014, the staff concludes that the mobile alarm notification system provides information equivalent to that provided by CR indications, and is sufficient to maintain a continuous awareness of important plant conditions, given the permanent shutdown and defueled condition of KPS, regardless of the onsite location of the operations staff. Furthermore, there are sufficient communications capabilities to ensure that both the shift supervisor and the non-certified operator can be contacted by multiple, diverse communications systems. The NRC staff finds that proposed revision to TS 5.2.2, in conjunction with the licensee's mobile communication systems, permit the operations staff procedural flexibility to leave the CR and directly observe and provide operational support, as needed throughout the plant, for those systems that continue to be used at a decommissioning reactor, including operations activities associated with the SFP or the handling of irradiated fuel. In addition, being free to leave the CR to conduct routine monitoring of plant conditions and equipment functions, as well as the ability to directly observe activities involving radiological decontamination is in the overall interest of safety of the facility . . For the reasons discussed above, the NRC staff finds that the proposed changes to TS 5.2.2.a and TS 5.2.2.b. are acceptable.

TS 5.2.2.c TS 5.2.2.c currently states that: "A radiation technologist shall be on site when fuel is in the reactor. The position may be vacant for not more than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, except in severe weather conditions, in order to provide for unexpected absence, provided immediate action is taken to fill the required position." This administrative TS is not applicable to a permanently shutdown and defueled reactor since fuel can never be place in the reactor. The NRC staff notes that a comparable TS 5.2.2.e was proposed by DEK in a supplemental letter dated March 13, 2014, to require an individual qualified in radiation protection procedures to be onsite during fuel handling operations or movements of loads over storage racks containing fuel. The NRC approved the proposed change to the KPS TS 5.2.2.e in Amendment No. 212 dated June 9, 2014. The approved change to TS 5.2.2.e is as follows: An individual qualified in radiation protection procedures shall be onsite during fuel handling operations or movements of loads over storage racks containing fuel. No additional changes are being proposed to TS 5.2.2.e since the issuance of Amendment No. 212. This discussion is provided for completeness only. DEK has proposed the following TS in place of the currently existing TS 5.2.2.c. All fuel handling operations.

shall be directly supervised by a qualified individual.

The licensee stated that this new requirement ensures that movement of irradiated fuel is only performed under the direct supervision of an individual who has been trained and qualified on the procedures, processes, requirements and standards for safe movement of irradiated fuel. In its response to a request for additional information in a submittal dated September 23, 2013, DEK stated that "the 'qualified individual' referred to in proposed TS 5.2.2.c is a Certified Fuel Handler. Certified Fuel Handlers are responsible for handling of nuclear fuel, as stated in the 'Shift Manager I Certified Fuel Handler Training Program Guide.' This document also explicitly requires that Certified Fuel Handlers be trained to supervise fuel movement as part of their qualification process. Additionally, proposed TS 5.1.2 places overall responsibility for facility shift operation with the shift manager, who is required to be a Certified Fuel Handler per proposed TS 5.2.2.d." The NRC staff notes that the shift manager, who is required to be a CFH by proposed TS 5.2.2.d is the staff position that is responsible for oversight, awareness, monitoring and authorization of fuel handling activities.

Unit staff qualified as CFHs have completed training in the safe conduct of decommissioning activities, safe handling and storage of spent fuel, and the appropriate response to plant emergencies.

The staff has found in its approval of the Certified Fuel Handler training program, as discussed in the evaluation of TS 5.2.2.d, provides the appropriate training for oversight of fuel handling operations.

In addition, the direct supervision of fuel handling operations by an individual who has been trained and qualified on the procedures, processes, requirements, and standards for safe movemenfof irradiated fuel will provide added assurance that this activity will be conducted in a safe manner. The new requirement establishes the level of qualification and supervision of fuel handling operations.

Therefore, the NRC staff concludes that proposed TS 5.2.2.c is acceptable. TS 5.2.2.d TS 5.2.2.d currently specifies the qualifications of plant operations management.

The current TS states: 'The operations manager or assistant operations manager shall hold a Senior Operator license." Revised TS 5.2.2.d, would state: d. The shift manager shall be a Certified Fuel Handler. DEK has proposed to revise TS 5.2.2.d to replace the requirement that the operations manager or assistant operations manager shall hold a Senior Operator license, with a requirement that the shift manager shall be a Certified Fuel Handler. DEK states that based on the permanently shutdown and defueled condition of KPS, licensed operators are no longer required.

Therefore, there is no longer a need for operations management staff to hold a Senior Operator license. DEK further states that replacing this with a requirement that the shift manager shall be a Certified Fuel Handler ensures that the senior individual on shift is appropriately trained and qualified, in accordance with the NRG-approved Certified Fuel Handler training program, to supervise shift activities.

The NRC staff reviewed the revised requirement and concluded that because the licensee has certified that it will no longer operate the unit, the requirements of 10 CFR 50.54(m) addressing licensed operator staffing no longer apply to facilities that have submitted certifications in accordance with 10 CFR 50.82(a)(2), as discussed in NRC letter dated May 21, 2014 (ADAMS Accession No. ML 14127 A340). The staff finds that the proposed TS establishes appropriate minimum qualification requirements for the shift manager position.

Shift managers qualified as CFHs have completed training in the safe conduct of decommissioning activities, safe handling and safe storage of spent fuel, and the appropriate response to plant emergencies.

In NRC letter dated May 12, 2014 (ADAMS Accession No. ML 14104A046), the staff approved DEK's shift manager/CFH training program recognizing that the licensee would use supervisors qualified as CFHs as the shift mangers at KPS. Therefore, the staff finds the proposed change to TS 5.2.2.d acceptable.

TS 5.2.2.e The NRC approved the proposed change to the KPS TS 5.2.2.e in Amendment No. 212 dated June 9, 2014. No additional changes are being proposed to TS 5.2.2.e since the issuance of Amendment No. 212. 3.7.17.3 TS 5.3: Unit Staff Qualifications The licensee proposed to change the title of TS 5.3 from "Unit Staff Qualifications" to "Facility Staff Qualifications" for editorial reasons consistent with usage throughout the TSs. The NRC staff reviewed the proposed revision to title TS 5.3 and concludes that changing the word "unit" to "facility" is an acceptable clarifying change that is editorial in nature such that the current intent of the requirement is unchanged.

Therefore, the staff finds the proposed change acceptable.

• TS 5.3.1 The current introduction in TS 5.3.1 states: "Each member of the unit staff shall meet ... " The revised introduction to TS 5.3.1 would state: Each member of the facility staff shall meet ... The licensee proposed to change unit staff to facility staff for editorial reasons consistent with usage throughout the TSs. The NRC staff reviewed the proposed revision to TS 5.3.1 and concludes that changing the word "unit" to facility" is an acceptable clarifying change that is editorial in nature such that the current intent of the requirement is unchanged.

Therefore, the staff finds the proposed change acceptable.

In addition, the licensee has also proposed to revise the requirements of TS 5.3.1.b. The TS currently states: "The education and experience eligibility requirements for operator license applicants, changes thereto, shall be those previously reviewed and approved by the NRC, specifically those referenced in NRC Safety Evaluation letter for Amendment 170, dated October 2, 2003." DEK has proposed to delete TS 5.3.1.b. DEK proposed to delete TS 5.3.1.b because the requirements for licensed operators no longer apply following submittal of the certifications required by 1 O CFR 50.82(a).

The NRC staff reviewed the proposed changes to TS 5.3.1.b. and determined that the education and experience eligibility requirements concerning licensed operators are no longer needed for KPS since it is not authorized to operate or place fuel in the reactor vessel. Because the licensee has certified that it will no longer operate the unit, the requirements of 10 CFR 50.54(m) addressing licensed operator staffing no longer apply to facilities that have submitted certifications in accordance with 10 CFR 50.82(a)(2), as discussed in NRC letter dated May 21, 2014 (ADAMS Accession No. ML 14127A340).

Therefore, the NRC staff finds that the licensee's proposed changes that remove requirements for licensed operators are acceptable. TS 5.3.2 TS 5.3.2 currently defines a licensed operator and licensed senior operator.

The TS states: "For the purpose of 10 CFR 55.4, a licensed Senior Operator and a licensed Operator are those individuals who, in addition to meeting the requirements of Specification 5.3.1, performed the functions described in 10 CFR 50.54(m)." DEK has proposed to delete TS 5.3.2. DEK proposed to delete current TS 5.3.2 because the requirements for licensed operators no longer apply following submittal of the certifications required by 10 CFR 50.82(a).

The NRC staff has determined that the requirements concerning licensed operators are no longer needed for KPS since it is not authorized to operate or place fuel in the reactor vessel. As discussed in NRC letter dated May 12, 2014, KPS has an NRC approved training program for a shift manager/CFH that qualifies the appropriate level of operations staff shift supervision at a decommissioning power reactor. In addition, the requirements of 10 CFR 50.54(m) addressing licensed operator staffing no longer apply to facilities that have submitted certifications in accordance with 10 CFR 50.82(a), as discussed in NRC letter dated May 21, 2014 (ADAMS Accession No. ML 14127A340).

Therefore, the NRC staff finds that the licensee's proposed changes that remove requirements for licensed operators are acceptable.

In place of the existing TS 5.3.2, the licensee proposed adding a new TS 5.3.2 requiring that a training and retraining program for the Certified Fuel Handler positions be maintained.

New TS 5.3.2 would state: An NRC approved training and retraining program for the CERTIFIED FUEL HANDLERS shall be maintained.

DEK states that the CFH Training Program ensures that the qualifications of fuel handlers are commensurate with the tasks to be performed and the conditions requiring response.

10 CFR 50.120, "Training and qualification of nuclear power plant personnel," requires training programs to be derived using a systems approach to training (SAT) as defined in 1 O CFR 55.4. The CFH Training Program provides adequate confidence that appropriate SAT based training of personnel who will perform CFH duties is conducted to ensure the facility is maintained in a safe and stable condition.

The NRC staff has determined that newly proposed TS 5.3.2 is consistent with the definition for a CFH in 1 O CFR 50.2, which states that a CFH is a non-licensed operator that has been qualified in accordance with an NRC-approved CFH training program. As noted previously, the CFH training and retraining program for KPS has been approved by the NRC staff. The NRC staff concludes that new TS 5.3.2 appropriately implements the 10 CFR Part 50 definition of a CFH. In accordance with 10 CFR 50.36(c)(5), "Administrative Controls" are provisions related to organization and management (among other items) that assure operation of the facility in a safe manner. The staff finds that the requirement for maintaining an approved training program for the CFH ensures that the CFH will be adequately qualified to supervise decommissioning activities safely. Therefore, the NRC staff concludes that the proposed new TS 5.3.2 is acceptable and appropriate to include in the TS administrative controls.

3. 7 .17.4 TS 5.4: Procedures TS 5.4.1.b TS 5.4.1 requires that written procedures be established, implemented, and maintained covering certain activities.

One of the activities requiring written procedures, TS 5.4.1, paragraph b., currently states: "The emergency operating procedures required to implement the requirements of NUREG-0737 .and NUREG-0737, Supplement 1, as stated in Generic Letter 82-33." DEK has proposed to delete TS 5.4.1.b. NUREG-0737, "Clarification of TMI Action Plan Requirements," November 1980 (ADAMS Accession No. ML051400209), and NUREG-0737, Supplement 1, "Clarification of TMI Action Plan Requirements:

Requirements for Emergency Response Capability," January 1983 (ADAMS Accession No. ML 102560009), as stated in Generic Letter 82-33, "Supplement 1 to NUREG-0737

-Emergency Response Capabilities," dated December 17, 1982 (ADAMS Accession No. ML031080548), incorporated into one document all Three Mile Island related items approved for implementation by the Commission at that time. This included the use of human factored, function oriented, emergency operating procedures to improve human reliability and the ability to mitigate the consequences of a broad range of initiating events for operating reactors, and subsequent multiple failures or operator errors, without the need to diagnose specific events. The licensee proposed to delete the requirement of TS 5.4.1.b. because the emergency operating procedures discussed therein only pertain to accidents and events resulting from reactor operation.

The licensee stated that the referenced procedures are no longer required for a permanently shutdown and defueled reactor. The NRC staff reviewed the proposed deletion of TS 5.4.1.b. and determined that NUREG-0737, as supplemented, implemented to the way reactor operators are trained, instrumentation information is presented, and procedures are structured, using human factors and function oriented approach to addressed operating events and accidents.

These accidents, and the associated emergency operating procedures to detect, respond to, and mitigate such accidents, concerned malfunctions of the reactor and its supporting systems and are not relevant to a permanently shutdown and defueled reactor, which is no longer authorized to operate or place fuel in the reactor vessel. Therefore, the NRC staff finds that deleting TS 5.4.1.b. is acceptable.

r 3.7.17.5 TS 5.5: Programs and Manuals This section defines the programs and manuals that are applicable to KPS. TS 5.5.2 Primary Coolant Sources Outside Containment ts 5.5.2., "Primary Coolant Sources Outside Containment," was established to minimize leakage from portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident.

The licensee proposed to delete this program since primary coolant systems have been drained at KPS and there are no longer any transient or accident conditions associated with primary coolant sources given the permanently shutdown and defueled condition of the plant. The NRC staff has determined that deletion of the Primary Coolant Sources Outside of Containment program from TSs is consistent with the transition to a permanently shutdown and defueled facility.

Since, in accordance with 10 CFR 50.82(a)(2), the licensee is prohibited from operating the plant or placing fuel in the reactor vessel, there are no longer any transients involving primary coolant outside of containment.

Therefore, the proposed deletion of this program appropriately reflects the change in plant status, and is acceptable.

TS 5.5.4 Component Cyclic or Transient Limit Program *TS 5.5.4, "Component Cyclic or Transient Limit Program," provides controls to track cyclic and transient occurrences to ensure that RCS components are monitored for fatigue evaluation based on a conservative estimate of the magnitude and frequency of the temperature and pressure transients resulting from normal operation, normal and abnormal load transients and accident conditions.

The licensee proposes to delete this program since the RCS components monitored by this program are no longer used at KPS in its permanently shutdown and defueled status. The NRC staff has determined that deletion of the Component Cyclic or Transient Limit Program from TSs is consistent with the transition to a permanently shutdown and defueled facility.

Since, in accordance with 10 CFR 50.82(a)(2), the licensee is prohibited from operating the plant or placing fuel in the reactor vessel, the RCS and reactor support systems are no longer in use. Consequently, the component cyclic or transient limit program is not relevant at KPS since the components monitored by the program are permanently out of service. The staff notes that structures, systems, and components related to the storage and cooling of spent fuel at KPS are not part of the program. Therefore, the proposed deletion of this program appropriately reflects the change in plant status. TS 5.5.5 Reactor Coolant Pump Flywheel Inspection Program TS 5.5.5, "Reactor Coolant Pump Flywheel Inspection Program," provides for the inspection of the reactor coolant pump flywheels.

The licensee proposed to delete this program since the reactor coolant pump flywheel is a component only used in support of reactor operation.

Inspection of the reactor coolant pump flywheel is not relevant to KPS since the licensee is no longer authorized to operate the reactor or emplace fuel in the reactor vessel. The NRC staff has determined that deletion of the Reactor Coolant Pump Flywheel Inspection Program from TSs is consistent with the transition to a permanently shutdown and defueled facility.

Since; in accordance with 10 CFR 50.82(a)(2), the licensee is prohibited from operating the plant or placing fuel in the reactor vessel, reactor coolant pumps are no longer used in support of any function at the facility.

Therefore, the proposed deletion of this program . appropriately reflects the change in plant status, and is acceptable.

TS 5.5.6 lnservice Testing Program TS 5.5.6, "lnservice Testing Program," establishes the controls for periodic testing of American Society of Mechanical Engineers (ASME) Code Class 1, 2, and 3 pumps and valves in accordance with the ASME Operation and Maintenance Code. The licensee proposed to delete this program since there are no longer any ASME Code Class 1, 2 or 3 pumps and valves in the KPS lnservice Testing Program that continue to operate an*d perform a specific function in mitigating the consequences of an accident due to the permanently shutdown and defueled status of the plant. Because the licensee is prohibited from operating the plant or placing fuel in the reactor vessel, in accordance with 10 CFR 50:82(a)(2), there are no longer any ASME Code class pumps and valves that remain in operation and are relied upon to mitigate a OBA As such, the lnservice Testing Program is no longer relevant to KPS given its permanently shutdown and defueled status. Therefore, the proposed deletion of this program appropriately reflects the change in plant status, and is acceptable.

TS 5.5.7 Steam Generator (SG) Program TS 5.5.7, "Steam Generator (SG) Program,

ensures that the steam generator tube integrity is maintained.

The licensee proposed to delete this program since KPS is permanently defueled and is not authorized to operate; therefore, the steam generator tubes will not be subjected to the temperature and pressure effects that the steam generator program was put in place to protect against. The NRC staff has determined that the Steam Generator Program is only relevant to an operating reactor for removing heat generated in the reactor core or stored in the reactor coolant system. Since, in accordance with 10 CFR 50.82(a)(2), the licensee is prohibited from operating the plant or placing fuel in the reactor vessel, the steam generators are no longer used in support of any function at the facility.

Therefore, the proposed deletion of this program appropriately reflects the change in plant status, and is acceptable.

TS 5.5.8 Secondary Water Chemistry Program TS 5.5.8, "Secondary Water Chemistry Program,

provides controls for monitoring secondary water chemistry to inhibit steam generator tube degradation and low pressure turbine disc stress corrosion cracking.

The licensee proposed to delete this program because the components that the program was established to protect using water chemistry control are associated with reactor operation.

With the licensee's decision to cease reactor operations, \ these components are no longer in operation and do not need protection from degradation or stress corrosion cracking.

The NRC staff has determined that the deletion of the Secondary Water Chemistry Program is consistent with the transition to a permanently shutdown and defueled facility.

Since, in accordance with 10 CFR 50.82(a)(2), the licensee is prohibited from operating the plant or placing fuel in the reactor vessel, the steam generators and turbine are no longer used in support of any function at the facility.

Therefore, the proposed deletion of this program appropriately reflects the change in plant status, and is acceptable.

TS 5.5.9 Ventilation Filter Testing Program (VFTP) TS 5.5.9, "Ventilation Filter Testing Program (VFTP)," provides confirmation that safety-related high-efficiency particulate air (HEPA) filters and safety-related charcoal adsorber filters are tested and perform within their system design parameters.

The licensee stated that the program only pertains to operating reactor support systems. The accident analysis applicable to the permanently shutdown and defueled condition (the FHA) does not rely on ventilation filters for accident mitigation.

' The NRC staff has determined that the Ventilation Filter Testing Program is only applicable to systems that have been evaluated by the staff and approved for deletion in preceding sections of this TS amendment request. The staff has also confirmed that the accident analysis applicable to the permanently shutdown and defueled condition (the FHA) does not rely on ventilation filters for accident mitigation.

Since, in accordance with 10 CFR 50.82(a)(2), the licensee is prohibited from operating the plant or placing fuel in the reactor vessel, there are no DBAs for KPS that require operable filters. Therefore, the proposed deletion of this program appropriately reflects the change in plant status, and is acceptable.

TS 5.5.10 Explosive Gas and Storage Tank Radioactivity Monitoring Program TS 5.5.10, "Explosive Gas and Storage Tank Radioactivity Monitoring Program," provides controls for potentially explosive gas mixtures contained in the Radioactive Waste Disposal System, and the quantity of radioactivity contained in gas storage tanks or fed into the offgas treatment system. The licensee stated that all KPS waste gas decay tanks have been vented and purged, and that no consequential residual radiation or radioactive material remains in these tanks. Therefore, the licensee proposed to delete this program as a result. The licensee responded to an RAI by the NRC staff concerning the control of tanks that may be used to store radioactive wastes during future active decommissioning and dismantlement activities at the site. In DEK supplemental letter dated October 15, 2013, the licensee stated that any tanks used to store radiative wastes generated in future active decommissioning would be outside of the scope of this program. The licensee subsequently proposed a modification to this program in a supplemental letter dated April 29, 2014. In response to the NRC staff's concerns, DEK proposed to retain a portion of the program to the extent required to support storage of liquid radioactive waste that might be generated during future decommissioning activities.

The proposed revised TS 5.5.10 would state: 5.5.10 \ Storage Tank Radioactivity Monitoring Program This program provides controls for the quantity of radioactivity contained in unprotected outdoor storage tanks. The liquid radwaste quantities shall be determined in accordance with the Standard Review Plan, Section 15.7.3, "Postulated Radioactive Release due to Tank Failures." The program shall include: a. A surveillance program to ensure that the quantity of radioactivity contained in all outdoor liquid radwaste tanks that are not surrounded by liners, dikes, or walls, capable of holding the tanks' contents and do not have tank overflows and surrounding area drains connected to the Waste Disposal System is less than the amount that would result in concentrations less than the limits of 10 CFR 20, Appendix B, Table 2, Column 2, at the nearest potable water supply and the nearest surface water supply in an unrestricted area, in the event of an uncontrolled release of the tanks' contents:

The provisions of SR 3.0.2 are applicable to the Storage Tank Radioactivity Monitoring Program Surveillance Frequencies.

The NRC staff has reviewed the proposed revision to the Storage Tank Radioactivity Monitoring Program. The staff finds the proposed changes prudent given the uncertainty in how future radwaste generated by flushing and cutting of radioactive systems will be stored and processed.

Therefore, the staff finds the revised TS 5.5.10 Storage Tank Radioactivity Monitoring Program acceptable.

TS 5.5.11 Diesel Fuel Oil Testing Program TS 5.5.11, "Diesel Fuel Oil Testing Program," pertains to the testing of both new and stored fuel oil used to supply and operate the EDGs. The accident analyses applicable to the permanently

• shutdown and defueled condition at KPS does not rely on EDGs for accident mitigation.

The requirement for EDGs, which are supported by the fuel oil being tested per this program, has already been deleted from TSs (Amendment No. 212 issued by NRC letter dated June 9, 2014; ADAMS Accession No. ML 14111A234).

The NRC staff determined that with KPS permanently shutdown and defueled and the irradiated fuel having decayed for a period greater than 90 days, the EDG fuel oil and lube oil system are not needed because there are no active systems or associated support systems credited as part of the initial conditions of an analysis or as part of the primary success path for mitigation of the FHA DBA. The staff confirmed that there are no other DBAs analyzed in Chapter 14 of the KPS USAR that rely on EDG as a support system. Therefore, the NRC staff finds the proposed deletion of the Diesel Fuel Oil Testing Program to be acceptable.

TS5.5.13 Safety Function Determination Program (SFDP) TS 5.5.13, "Safety Function Determination Program (SFDP)," ensures that a loss of safety function is detected and appropriate actions taken. Upon entry into TS LCO 3.0.6, an evaluation shall be made to determine if a loss of.safety function exists. The program implements the requirements of LCO 3.0.6. LCO 3.0.6 directs an evaluation in accordance with SFDP to determine if a loss of safety function exists based on the status of redundant TS safety systems and associated support systems (systems that support the functionality of the safety system) to ensure the appropriate required actions are taken to maintain overall reactor safety. With the termination of reactor operations at KPS and the permanent removal of the fuel from the reactor vessel, there are no active SSCs at KPS that are required for accident mitigation.

Therefore, none of the systems in the current SFDP at KPS meet the definition of a related SSC stated in 10 CFR 50.2 (with the exception of the passive SFP structure).

Therefore, the requirements of the safety function determination program, which directs train checks of multiple and redundant safety systems, no longer apply. The NRC staff has determined that there are no longer any active safety related SSCs that continue to function at KPS due to its permanently shutdown and defueled status. None of the systems that are currently in the SFDP remain in operation in a safety related capacity.

This is consistent with the removal of all active safety related systems from the KPS defueled TSs proposed by this licensing action. In addition, the safety function determination program is invoked by LCO 3.0.6, which is being deleted in its entirety, as discussed in Section 3.7.6 of this safety evaluation.

Therefore, the NRC staff finds the proposed deletion of the Safety Function Determination Program to be acceptable.

TS 5.5.14 Containment Leakage Rate Testing Program TS 5.5.14, "Containment Leakage Rate Testing Program,

was established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(0) and 10 CFR Part 50, Appendix J, Option B, as modified by approved exemptions.

The licensee proposed to delete this program since the containment is no longer relevant to a permanently shutdown and defueled reactor. The NRC staff has reviewed the proposed deletion of the Containment Leakage Rate Testing Program and has determined that this is acceptable because 10 CFR 50.54(0) excludes permanently defueled units from the requirements of 10 CFR Part 50, Appendix J. In addition, the staff notes that this action is consistent with the guidance in NRC Regulatory Guide 1.184, "Decommissioning of Nuclear Power Reactors," July 2000, Section 9.0, "Eliminated Regulatory Requirements," and the deletion appropriately reflects the condition of KPS as a permanently shutdown and defueled facility.

Therefore, the NRC staff finds the proposed deletion of the Containment Leakage Rate Testing Program to be acceptable.

TS5.5.15 Battery Monitoring and Maintenance Program TS 5.5.15, "Battery Monitoring and Maintenance Program," provides controls for safety-related battery restoration and maintenance.

The licensee proposed deletion of this program consistent with the deletion of the corresponding TS for direct current (DC) electrical systems and associated batteries.

The licensee stated that the KPS design basis accident analyses do not rely on batteries or any other safety related electrical systems for any accident mitigation.

The NRC staff determined that with KPS permanently shutdown and defueled and the, irradiated fuel having decayed for a period greater than 90 days, there are no longer any design basis accidents at KPS that can result in an offsite release exceeding regulatory limits. Consequently, there are no safety related electrical systems required at KPS and no need for safety-related batteries required to support the plant electrical systems. The staff confirmed that there are no other DBAs analyzed in Chapter 14 of the KPS USAR that rely on station batteriel?

as a support system. The evaluation of the TSs related to DC electrical power systems and battery operability is provided in Section 3.7.14 of this safety evaluation.

The staff has previously found that the applicable electrical power TSs that rely on battery support are no longer required at KPS and these TS were delete by NRC Amendment No. 212 ,issued by NRC letter dated June 9, 2014 (ADAMS Accession No. ML 14111A234).

Therefore, the staff finds the proposed deletion of the Battery Monitoring and Maintenance Program to be acceptable.

TS 5.5.16 Setpoint Control Program TS 5.5.16, "Setpoint Control Program," establishes the requirements for ensuring that setpoints for automatic protective devices are initially within and remain within the assumptions of the applicable safety analysis.

The program applies to TSs for the reactor protection system (LCO 3.3.1 ); engineered safety feature actuation system (LCO 3.3.2); loss of offsite power diesel generator start instrument (LCO 3.3.5); containment purge and vent isolation instrumentation (LCO 3.3.6); and CR post-accident recirculation actuation instrumentation (LCO 3.3.7). The licensee stated that these five TS LCOs to which this program applies are being deleted; therefore, the need for this program no longer exists. The instrumentation-related TSs that reference the setpoint control program are discussed in the Section 3.7.9 of this safety evaluation.

As discussed in Section 3.7.9, the NRC staff determined that these five TSs subject to the setpoint control program can be deleted from the KPS TSs. Therefore, the staff finds of the setpoint control program to be no longer necessary and the proposed deletion acceptable. 3. 7.17.6 TS 5.6 Reporting Requirements TS 5.6.1 Annual Radiological Environmental Operating Report The current TS states that the report covers" ... the operation of the unit during the previous calendar year ... " The licensee proposed to revise the TS description by replacing "unit" with "facility" such that the description will state" ... the operation of the facility during the previous calendar year ... " The NRC staff reviewed the proposed revision to TS 5.6.1 and concludes that changing the word "unit" to "facility" is an acceptable clarifying change that is editorial in nature such that the current intent of the requirement is unchanged.

Therefore, the staff finds the proposed change acceptable.

TS 5.6.2 Radioactive Effluent Release Report The current TS states that the report covers" ... the operation of the unit in the previous calendar year ... " In addition, the report shall summarize the effluent" ... waste released from the unit." The licensee proposed to revise the TS description by replacing "unit" with "facility" such that the description will state" ... the operation of the facility in the previous calendar year ... "and " ... waste released from the facility." The NRC staff reviewed the proposed revision to TS 5.6.2 and concludes that changing the word "unit" to facility" is an acceptable clarifying change that is editorial in nature such that the current intent of the requirement is unchanged.

Therefore, the staff finds the proposed change acceptable.

TS 5.6.3 Core Operating Limits Report (COLR) TS 5.6.3, "Core Operating Limits Report (COLR)," establishes the core operating limits prior to each reload cycle. The licensee proposed to delete this program since it is prohibited from reloading fuel into the KPS reactor core and the safety limits established by this report no longer apply. The NRC staff has determined that the proposed deletion of the COLR would appropriately reflect the permanently shutdown and defueled condition of the facility.

The COLR only applies to reactors authorized to operate. Since the licensee is prohibited from operating the reactor or placing fuel in the reactor vessel, the COLR is no longer necessary.

Therefore, the staff finds the proposed deletion of TS 5.6.3 acceptable.

TS 5.6.4 Post Accident Monitoring Report The Post Accident Monitoring Report is generated to document the inoperability and corrective actions of post accident monitoring instrumentation required by TS LCO 3.3.3. The licensee proposed to delete this program because there is no longer a need for the post accident monitoring instrumentation and the implementing TS 3.3.3 is being deleted. The staff has determined that the posfaccident monitoring instrumentation in TS 3.3.3, as discussed in Section 3.7.9 of this evaluation, is acceptable to be deleted in its entirety.

Therefore, the NRC staff finds the proposed deletion of the associated Post Accident Monitoring Report to be acceptable.

TS 5.6.5 Steam Generator Tube Inspection Report The Steam Generator Tube Inspection Report is associated with the TS 5.5.7 Steam Generator Program. KPS is permanently defueled and cannot operate; therefore, the steam generator*

tubes will not be subjected to the temperature and pressure effects that the steam generator program and associated inspection report was put in place to monitor and asses. The NRC staff has determined that the deletion of TS 5.5.7, steam generator program, as discussed in Section 3.7.17.5 of this safety evaluation, is acceptable.

Therefore, the NRC staff finds the proposed deletion of the associated Steam Generator Tube Inspection Report to be acceptable.

3.8 Changes to Renewed Facility Operating License 3.8.1 Changes to Commission Finding in Section 1 of the DEK license. The licensee proposes to delete Commission findings 1.8, 1.1, and 1.J of the license. 1. 8 currently states: 8. Construction of the Kewaunee Power Station (facility) has been substantially completed in conformity with Provisional Construction Permit No. CPPR-50, as amended, and the application, as amended, the provisions of the Act and the rules and regulations of the Commission.

DEK proposed to delete 1.8 because it has no bearing on the regulation of decommissioning activities at KPS. 1.1 currently states: I. The receipt, possession, and use of byproduct, source, and special nuclear material as authorized by this renewed operating license will be in accordance with the Commission's regulations in 10 CFR Parts 30 and 70, including 10 CFR Sections 30.33, 70.23, and 70.31; and DEK has proposed to delete 1.1 because it is redundant to license condition 2.8. The licensee states that possession and use of byproduct, source, and special nuclear material at KPS during decommissioning activities is adequately covered by license condition 2.8 and, therefore, 1.1 is not needed. 1.J currently states: J. Actions have been identified and have been or will be taken with respect to (1) managing . the effects of aging during the period of extended operation on the functionality of structures and components that have been identified to require review under 10 CFR 54.21 (a)(1 ), and (2) time-limited aging analyses that have been identified to require review under 1 O CFR 54.21 (c), such that there is reasonable assurance that the activities authorized by this renewed operating license will continue to be conducted in accordance with the current licensing basis, as defined in 10 CFR 54.3, for the facility, and that any changes made to the facility's current licensing basis in order to comply with 10 CFR 54.29(a) are in accordance with the Act and the Commission's regulations.

DEK has proposed to delete 1.J since 10 CFR 50.82(a)(2) prohibits operation of the KPS reactor once the certifications described therein are submitted, KPS will not operate during the period of extended operation.

Decommissioning of KPS is not dependent on the requirements of Part 54 for a renewed license. Therefore, requirements that are unique to a renewed license are not needed. DE.K refers to the items in Section 1 of the KPS license as license conditions.

The NRC staff notes that the items in Section 1 are not conditions on the license but, instead, constitute the Commission findings that provide the staff's licensing basis for issuing the KPS renewed facility operating license. While certain wording in Section 1 may be modified occasionally (e.g., when the name of the licensee changes), th_ese findings represent the historical documented basis for the staff's conclusions that the licensee meets other conditions referenced in the license and should remain unaltered in the license until license termination.

The staff notes that none of the Commission findings in Section 1 imposes any conditions or restrictions on the licensee.

Based on the above discussion, the NRC denies the requested changes to Commission findings 1. B, 1.1, and 1.J of the license. 3.8.2 Changes to License Condition 2.8.(1) Current license condition 2.8.(1) states: (1) Pursuant to Section 104b of the Act and 10 CFR Part 50, "Licensing of Production and Utilization Facilities," to possess, use, and operate the facility at the designated location in Kewaunee County, Wisconsin in accordance with the procedures and limitations set forth in this renewed license; Revised license condition 2.8.(1) would state: (1) Pursuant to Section 104b of the Act and 10 CFR Part 50, "Licensing of Production and Utilization Facilities," to possess the facility at the designated location in Kewaunee County, Wisconsin in accordance with the procedures and limitations set forth in this renewed license; DEK proposed that this license condition be revised to delete the words" ... use, and operate ... " because the licensee is no longer authorized to use or operate KPS. The NRC approves in part and denies in part the change proposed by DEK to license condition 2.8.(1). The current license condition allows DEK, pursuant to Section 104b of the Atomic Energy Act and 10 CFR Part 50, " ... to possess, use, and operate the facility ... set forth in this renewed license." For clarification, the licensee proposed to delete the words "use, and operate" from the license condition.

While the staff agrees that the licensee is no longer authorized to operate KPS as a nuclear reactor, the license still permits DEK to use (emphasis added) structures, systems, and components of the facility that may be required to ensure the safety and security of licensed material until license termination.

Therefore, the NRC denies removal of the word "use" from the proposed change, and instead, approves a revision to this conditioq_

that permits DEK to " ... possess and use the facility ... set forth in this renewed license." 3.8.3 Changes to License Condition 2.8.(2) Current license condition 2.8.(2) states: (2) Pursuant to the Act and 10 CFR Part 70, to receive, possess, and use at any time special nuclear material as reactor fuel in accordance with the limitations for storage and amounts required for reactor operation, as described in the Final Safety Analysis Report, as supplemented and* amended; Revised license condition 2.8.(2) would state: (2)

• Pursuant to the Act and 10 CFR Part 70, to possess at any time special nuclear material that was used as reactor fuel in accordance with the limitations for storage, as described in the Final Safety Analysis Report, as supplemented and amended; DEK proposed that this license condition be revised to delete the words "receive," "and use," "and the amounts required for reactor operation." DEK also proposed to add the words" ... that was used" immediately before the phrase" ... as reactor fuel. .. "within the license condition.

DEK stated that the receipt and use of special nuclear material as reactor fuel is no longer authorized for the permanently shutdown and defueled KPS. DEK stated that it is only authorized to possess existing reactor fuel at the site. The NRC staff has reviewed the proposed changes to this licensee condition and determined that, pursuant to 10 CFR 50.82(a)(2), the licensee is no longer authorized to receive or use reactor fuel at KPS. In addition, the licensee must still be allowed to possess the special nuclear material that is still present onsite as reactor fuel, in accordance with the specified limitations for storage as prescribed in the licensee condition.

Therefore, the proposed changes to license condition 2.B.(2) are acceptable.

3.8.4 Changes to License Condition 2.8.(3) Current license condition 2.B.(3) states: (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; Revised license condition 2.B.(3) would state: (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 sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; and possess any byproduct, source and special nuclear material as sealed neutron sources that were used for reactor startup; DEK proposed thatthis license condition be revised to delete the phrase " ... as sealed neutron sources for reactor startup, ... " DEK also proposed to add the phrase"; and possess any byproduct, source and special nuclear material as sealed neutron sources that was used for reader startup" to the end of the license condition.

DEK stated that the license condition has been revised to reflect authorization to only continue to possess those sealed neutron sources still present onsite that were used for reactor startup. Since KPS is permanently shut down and defueled, there is no longer any need to receive or use sealed neutron source for reactor startup. The NRC staff has reviewed the proposed changes to this licensee condition and determined that, consistent with 10 CFR 50.82(a)(2), the licensee has no need to be authorized to receive or use sealed neutron sources for reactor startup. However, since the licensee will continue to be allowed to possess the special nuclear material in the sealed neutron sources previously used for reactor startup, the proposed changes to license condition 2.8.(3) are acceptable.

3.8.5 Changes to License Condition 2.C.(1) Current license condition 2.C.(1) states: (1) Maximum Power Level The licensee is authorized to operate the facility at steady-state reactor core power levels not in excess of 1772 megawatts (thermal). Revised license condition 2.C.(1) would state: (1) Deleted. DEK stated that this license condition can be deleted because KPS is permanently shut down and defueled in accordance with 10 CFR 50.82(a)(2) and therefore power operation is no longer authorized.

The NRC staff has reviewed the proposed deletion of this licensee condition and determined that power operation is no authorized at KPS based on .its 10 CFR 50.82(a)(2) certifications of being permanently shutdown and defueled.

The licensee is not authorized to operate the facility at any power. Therefore, deletion of license condition 2.C.(1) is appropriate and acceptable.

3.8.6 Changes to License Condition 2.C.(3) Current license condition 2.C.(3) states: (3) Fire Protection The licensee shall implement and maintain in effect all provisions of the approved Fire Protection Program as described in the licensee's Fire Plan, and as referenced in the Updated Safety Analysis Report (USAR), and as approved in.the Safety Evaluation Reports, dated November 25, 1977, and December 12, 1978 (and supplement dated February 13, 1981 ), subject to the following provision:

The licensee may make changes to the approved Fire Protection Program without prior approval of the Commission, only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire. Revised license condition 2.C.(3) would state: (3) Deleted. DEK stated that this license condition is based on maintaining an operational fire protection program in accordance with 10 CFR 50.48, with the ability to achieve and maintain safe shutdown of the reactor in the event of a fire and is no longer applicable at KPS. However, many of the elements that are applicable for the operating plant fire protection program continue to be applicable during plant decommissioning.

During the decommissioning process, a fire protection program is required by 10 CFR 50.48(f) to address the potential for fires that could result in a radiological hazard. However, the regulation is applicable regardless of whether a requirement for a fire protection program is included in the facility license. Therefore, a license condition requiring such a program for a permanently shutdown and defueled plant is not needed. The NRC staff finds that license condition 2.C.(3), Fire Protection, is based on maintaining a fire protection program that provides reasonable assurance that the ability to achieve and maintain safe shutdown in the event of a fire in accordance with 10 CFR 50.48. Achieving and maintaining safe shutdown in the event of a fire is no longer applicable to the decommissioned fire protection program at KPS, since fuel has been removed from the reactor. However, elements of the fire protection program continue to be required during decommissioning to address fire events that could result in radiological hazards. 10 CFR 50.48(f) requires KPS to address potential for fires that could result in a radiological hazard. The licensee stated that the rule is sufficient to ensure that a program is maintained, and therefore; having a license condition that also requires a fire protection program for a permanently shutdown and defueled plant is not needed. On the basis of its evaluation, the NRC staff concludes that reliance on 10 CFR 50.48(f) is appropriate and the licensee's request to eliminate License Condition 2.C.(3) is. acceptable.

3.8.7 Changes to License Condition 2.C.(6) Current license condition 2.C.(6) states: (6) Steam Generator Upper Lateral Supports The design of the steam generator upper lateral supports may be modified by reducing the number of snubbers from four (4) to one (1) per steam generator.

Revised license condition 2.C.(6) would state: (6) Deleted. DEK stated that this license condition can be deleted because KPS is permanently shut down and defueled in accordance with 10 CFR 50.82(a)(2) and the steam generator lateral supports are no longer needed to support the decommissioning status of the plant. The NRC staff has reviewed the proposed deletion of this licensee condition and determined that steam generators lateral supports is a design feature that is not relevant to the decommissioning status of KPS. Therefore, deletion of license condition 2.C.(6) is acceptable.

3.8.8 Changes to License Condition 2.C.(14) Current license condition 2.C.(14) states: (14) Deferral of Certain Technical Specification Requirements Following implementation of License Amendment No. 207, the requirement for the reactor coolant system (RCS) Hot Leg A Temperature Indication to be OPERABLE as required by technical specification (TS) 3.3.3. and TS 3.3.4 may be deferred until startup after the first outage of sufficient duration to repair the RCS Hot Leg A Temperature Indication.

Specifically, TS Table 3.3.3-1, Function 3 will only require 1 channel to be OPERABLE, and TS Table B 3.3.4-1, Function 4.a will not be applicable.

Following the startup after the first outage of sufficient duration to repair the RCS Hot Leg A Temperature Indication, TS Table 3.3.3-1 Function 3 and TS Table B 3.3.4-1, Function 4.a requirements will be applicable.

Revised license condition 2.C.(14) would state: (14) Deleted. DEK stated that this license condition can be deleted because KPS is permanently shut down and defueled in accordance with 10 CFR 50.82(a)(2) and the RCS Hot Leg A Temperature Indication is no longer needed. The NRC staff has reviewed the proposed deletion of this licensee condition and determined that based on the permanently shutdown and defueled condition of KPS, the RCS is not relevant to the decommissioning status of KPS. As such, the RCS Hot Leg A Temperature Indication is no longer required and, therefore, deletion of license condition 2.C.(14) is acceptable.

3.8.9 Changes to License Condition 2.E and Proposed New License Condition 3 Current license condition 2.E states: E. This renewed operating license is effective as of the date of issuance and shall expire at midnight on December 21, 2033. Revised license condition

2. E would state: E. Deleted. DEK stated that this license condition can be deleted because KPS has permanently ceased operation.

10 CFR 50.82(a)(2) prohibits operation of the KPS reactor since the certifications described there.in have been docketed.

DEK has proposed that this license condition be replaced by new license condition 3, which comports to 10 CFR 50.51 in that the license authorizes ownership and possession of KPS until the Commission notifies the licensee in writing that the license is terminated.

The proposed new license condition to be used in place of license condition 2.E will be license condition

3. License condition 3 states: New License Condition 3 would state: 3. On February 25, 2013, Dominion Energy Kewaunee (DEK) certified that operations at Kewaunee Power Station would permanently cease in accordance with 10 CFR 50.82(a)(1

)(i). On May 14, 2013, DEK certified that the fuel had been permanently removed from the reactor* vessel in accordance with 10 CFR 50.82(a)(1)(ii).

As a result, the 10 CFR 50 license no longer authorizes operation of the reactor, or the emplacement or retention of fuel in the reactor vessel. This license is effective as of the date of issuance and authorizes ownership and possession of Kewaunee Power Station until the Commission notifies the licensee in writing that the license is terminated.

The licensee shall: A. Take actions necessary to decommission the plant and continue to maintain the facility, including, where applicable, the storage, control and maintenance of the spent fuel, in a safe condition; and B. Conduct activities in accordance with all other restrictions applicable to the facility in accordance with the NRC regulations and the applicable provisions of the 10 CFR 50 facility license as defined in Section 2 of this license. The NRC staff has reviewed the proposed deletion of licensee condition 2.E and the proposed new license condition 3 and determined that license condition 2.E, which documented the date of the expiration of the license, is no longer appropriate for the permanently shutdown condition of the plant. The proposed new license condition 3 documents the current condition of the plant and summarizes the actions and requirements applicable to the facility.

The proposed license condition 3 is consistent with the regulatory requirements applicable to the facility in the permanently shutdown and defueled condition, and consistent with a previously issued licensee condition for the permanently shutdown and defueled Millstone Unit 1. Based on the above, the proposed revision is acceptable.

3.8.10 Action on Other Proposed Changes to Renewed Facility Operating License In its original May 29, 2013, amendment request, DEK proposed changes to several other license conditions that were subsequently retracted by supplemental submittals or dispositioned in other licensing actions. For completeness, information on these license conditions are summarized below.

• In its May 29, 2013, submittal, DEK requested that current license condition 2.C.(4), Physical Protection, be revised to remove the requirement to implement a station Cyber Security Plan in accordance with 1 O CFR 73.54. Based on discussions between the NRC staff and the licensee, it was determined that the requirements of 10 CFR 73. 54 continue to apply to a permanently shutdown and defueled reactor that were licensed to operate on November 23, 2009, when the regulation became effective.

Consequently, in its submittal date January 7, 2014, DEK retracted its request to remove cyber security provisions from license condition 2.C.(4). In its May 29, 2013, submittal, DEK requested that current license condition 2.C.(10), Mitigating Strategy License Condition, be removed from its license. The licensee stated that regulations incorporating the Mitigating Strategy license condition specifically excluded reactors that has permanently shutdown and defueled under 10 CFR 50.54(hh)(3).

In consideration that KPS still has spent fuel stored in a SFP and in consideration of the continued need to provide adequate emergency response and protection against the design basis threat of radiological sabotage of the spent fuel stored in the SFP, the NRC staff requested additional information from the licensee to justify how the elimination of license condition 2.C.(10) would not result in operations that .would endanger the health and safety of the public or be inimical to the common defense and security.

Consequently, in its submittal dated April 29, 2014°, DEK retracted its request to remove license condition 2.C.(10).

In its May 29, 2013, submittal DEK requested that license condition 2.C.(15), License Renewal License Conditions, be removed from its license. DEK was concerned that certain license renewal activities required by license condition 2.C(15) could be required for KPS even though the reactor was permanently shutdown and not in a period of extend operation.

DEK requested a separate license amendment request to address the removal of this specific license condition in a letter to the NRC dated April 16, 2013. The NRC staff has processed that amendment request separately from the actions requested by DEK's May 29, 2013, submittal and approved the deletion of license condition

2. C. ( 15) in license amendment No. 213 issued to DEK on June 23, 2014.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Wisconsin 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. 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 which was published in the Federal Register on August 20, 2013 (78 FR 51224). 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 amendments.

6.0 CONCLUSION

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

K. Bucholtz D. Duvigneaud M. Keefe Date: February 13, 2015 D. Heacock The NRC staff stated in its June 9, 2014, letter to DEK approving the license amendment that the remaining changes to the KPS TSs and license conditions are still under review. The staff 1 has completed its evaluation of the proposed changes to the permanently defueled TSs as requested in DEK's letter dated May 29, 2013, as supplemented, and has included the staff's findings in the enclosed safety evaluation.

The May 29, 2013, amendment request, as supplemented, also proposed changes to the KPS license that DEK stated were either clarifications, redundant with other requirements, or no longer applicable based on the permanently shutdown and defueled status of KPS. The NRC staff has completed its review and assessment of the proposed license changes, as provided in the enclosed safety evaluation.

This amendment approves most of the proposed license changes requested by the licensee.

It also denies several of the proposed license changes. Specifically, the staff denies the changes proposed by DEK to Commission findings 1.B, 1.1, and 1.J of the license as discussed in Section 3.R 1 of the enclosed safety evaluation.

In addition, the NRC approves in part and der:iies in part the change proposed by DEK to license condition 2.B.(1) as discussed in Section 3:8.2 of the enclosed safety evaluation.

The Notice of Issuance will be included in the Commission's biweekly Federal Register notice. Docket No. 50-305

Enclosures:

Sincerely, IRA/* Thomas J. Wengert, Senior Project Manager Plant Licensing IV-2 and Decommissioning Transition Branch Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

1. Amendment No. 215 to Renewed

• Facility Operating License No. DPR-43 2. Safety Evaluation cc w/encls: Distribution via ListServ DISTRIBUTION:

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