View source for 3F0516-05, Final Safety Analysis Report Revision 38 & May 2016 10 CFR 50.59 and 10 CFR 72.48 Report

{{Adams
| number = ML16172A182
| issue date = 05/25/2016
| title = Final Safety Analysis Report Revision 38 & May 2016 10 CFR 50.59 and 10 CFR 72.48 Report
| author name = Reising R
| author affiliation = Duke Energy Florida, LLC
| addressee name = 
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000302
| license number = DPR-072
| contact person = 
| case reference number = 3F0516-05
| document type = Letter type:, Updated Final Safety Analysis Report (UFSAR)
| page count = 34
}}

=Text=
{{#Wiki_filter:Crystal River Nuclear Plant 15760 W. Power Line Street
                                                                                    . Crystal River, FL 34428 Docket 72-1035 Docket 50.302 Operating License No. DPA-72
                                                                             . 10 CFR 50.71(e) 10 CFR 50.59(d}(2) 10 CFR 72.48(d)(2)
May25, 2016 3F0516-05 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001

==Subject:==
Crystal River Unit 3 - Final Safety Analysis Report and 1o CFR 50.59 - 10 CFR 72.48 Report - May 2016

==Reference:==
: 1. CR-3 to NRC letter, "Crystal River Unit 3 - Certification of Permanent Cessation of Power Operations and that Fuel Has Been Permanently Removed from the Reactor," dated February 20, 2013. (ADAMS Accession No. ML13056A005)
: 2. CR-3 to NRC Letter, "Crystal River Unit 3 - Final Safety Analysis Report, Revision 34, and 10 CFR 50.59 Report, dated May 8, 2014 (ADAMS Accession No. ML14140A082)

==Dear Sir:==

In accordance with 10 CFR 50.71(e), Duke Energy Florida, LLC., previously known as Duke Energy Florida, Inc., (DEF), hereby submits Revision 38 to the Crystal River Unit 3 (CR-3) Final Safety Analysis Report. CD-ROMs are enclosed to the Document Control Desk the Regional Administrator (Region I) and the CR-3 NRC Project Manager. This revision replaces the previous revision of the FSAR in its entirety. FSAR text changes are indicated by revision bars on the outside right border of each page.
The FSAR revision includes material which describes the organization, modifications, system abandonments and other changes to CR-3 that have been implemented as of April 1, 2016. As required by 10 CFR 50.71 (e), a summary of changes made in FSAR, Revision 38 is provided in  (94 changes). Several of these changes were made after a 50.54(a) review was completed to assure no Quality Assurance commitments would be reduced.
Additionally, as required by 10 CFR 50.59(d)(2) and 10 CFR 72.48(d)(2), in Attachment 2, DEF is providing a summary of evaluations completed under 10 CFR 50.59 and 1o CFR 72.48 for changes made to the plant and Independent Spent Fuel Storage Installation (ISFSI) to the NRC. contains the 1o CFR 50.59 and 1o CFR 72.48 Report - May 2016, which includes a summary of all evaluations completed this reporting period under 10 CFR 50.59 with the exception of evaluations associated with changes, tests, or experiments that have not been fully implemented. Some of the completed 10 CFR 50.59 evaluations associated with plant changes (Modifications or system abandonments) required multiple revisions. The final 10 CFR 50.59 evaluation is being reported due to the cumulative nature of these changes. No 10 CFR 72.48 evaluations were completed for CR-3 ISFSI related changes during this reporting period.            j+ O        ~
No new regulatory commitments are made in this letter.                                                 Ne5p_

U.S. Nuclear Regulatory Commission                                                      Page 2 of 2 3F0516-05 If you have any questions regarding this submittal, please contact Mr. Phil Rose, Nuclear Regulatory Affairs, at (352) 563-4883.
I declare under penalty of perjury that the forgoing is true and correct. Executed on May 25, 2016 Sincerely,
     ;Wfl~
Ronald R. Reising, Senior Vice President Operations Support ARR/par Attachment 1 FSAR Revision 38 Change Summary Description Attachment 2 10 CFR 59.59 and 72.48 Report - May 2016 xc:     Regional Administrator, Region I NMSS Project Manager

                                             ---'!II DUKE ENERGY FLORIDA, INC.
CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302 I LICENSE NUMBER DPR-72 ATTACHMENT 1 FSAR REVISION 38 CHANGE 

==SUMMARY==
DESCRIPTION

U. S. Nuclear Regulatory Commission                                                  Attachment 1 3F0516-05                                                                            Page 1 of 17 FSAR REVISION 38 CHANGE 

==SUMMARY==
DESCRIPTION The Final Safety Analysis Report* (FSAR) revision reflects plant modifications, system abandonment activities, and information and analyses that constitute changes to the FSAR since the publication of FSAR Revision 34. Applicable figures and tables were included in these changes but were not individually identified. Several material changes to the FSAR have been made as the result of License Amendments. As part of implementation of License Amendments No. 246 and 247, FSAR changes were made that did not reflect any plant change but were approved by the amendment.
This FSAR revision includes changes made to incorporate the following:
A. Changes to plant engineering, programs, and revisions to analyses:
* FSAR Change Package 2013-05: This change revised FSAR Chapter 2, Site and Environment, Table 2-1, Chemical Storage Facilities at Crystal River Energy Complex, to add the new 1DOOL Argon Storage Tank for Crystal River Unit 3 (CR-3). This Table lists Hazardous chemicals and asphyxiants that had to be accounted for in the Cohtrol Room Habitability Analysis.
* FSAR Change Package 2013-21: This change was to remove the Auxiliary Steam interface with Units 1 and 2, which previously supplied Auxiliary Steam for Unit 3 startup activities. All discussion of Auxiliary Steam was removed from FSAR Chapter 1, Introduction and Summary, Section 1.9, Interactions Between Crystal River Unit 3 and the Four Fossil Fired Plants, and FSAR. Chapter 10, Steam and Power Conversion System, Section 10.2, System Design and Operation.
* FSAR Change Package 2014-16: This change revised the AC and DC Power descriptions in the FSAR to reflect the abandoned status of Diesel Generators 1A and 1C, as well as to align with the revised AC and DC configuration of the plant. The configuration of the plant was modified to reconfigure the electrical distribution system for permanent defueled/decommissioning status. This change affected FSAR Chapter 8, Electrical Systems, Sections 8.1, Design Bases, 8.2 Electrical System Design, 8.3, Tests and Inspections, and 8.4, Quality Control; and FSAR Chapter 9, Auxiliary and Emergency Systems, Sections 9.1, Makeup and Purification System, 9.3, Spent Fuel Cooling System, 9.5, Cooling Water Systems, 9.7; Plant Ventilation Systems, and 9.8, Plant Fire Protection Program.
* FSAR Change Package 2014-20: This change added a new seismic classification to the FSAR, Class 1*, and moved specific plant areas under this classification. This change affected FSAR Chapter 5, Containment System and Other Special Structures, Sections 5.1, Structural Design Classification and 5A, Other Class 1 Structures and Systems.
* FSAR Change Package 2014-21: This change added a discussion that would permit turning off both spent fuel cooling pumps in order to support maintenance activities, periods when forced cooling is not required, or to validate spent fuel pool heat-up rates.
This change affected FSAR Chapter 9, Auxiliary and Emergency Systems, Section 9.3, Spent Fuel Cooling.

U. S. Nuclear Regulatory Commission                                                Attachment 1 3F0516-05                                                                           Page 2of17
* FSAR Change Package 2014-22: This change revised the Fire Brigade make-up from five (5) members down to three (3) and increased reliance on offsite local fire departments for fire response, due to the decommissioning status of the Plant and the associated decreased risk. This change affected FSAR Chapter 7, Instrumentation and Controls, Section 7.4, Operating Control Stations and FSAR Chapter 9, Auxiliary and Emergency Systems, Section 9.8, Plant Fire Protection Program.
* FSAR Change Package 2014-34: This change revised the FSAR description of the Reactor Coolant System (RCS) to revise the system function and design information to only address the function of containing the activated core components/materials due to permanent removal of fuel from the reactor. This change affected FSAR Chapter 1, Introduction and Summary, Sections 1.3, Design Characteristics, and 1.4, Principal Architectural and Design Criteria; FSAR Chapter 3, Reactor, Section 3.2, Reactor Design; FSAR Chapter 4, Reactor Coolant System; and FSAR Chapter 11, Radioactive Waste and Radiation Protection, Section 11.3, Radiation Shielding.
* FSAR Change Package 2014-35: This change removed information related to Reactor Building external flood barriers and adds the mention of a concrete security barrier over the tendon access opening outside the Reactor Building to conform the FSAR to the plant configuration. This change affected FSAR Chapter 2, Site and Environment, Section 2.4, Hydrology.
* FSAR Change Package 2014-36: This change replaced information related to the Fire Protection Plan in FSAR Chapter 9, Auxiliary and Emergency Systems, Section 9.8, Plant Fire Protection Program, that was based on 10 CFR 50.48( c) with a fire protection plan based on 10 CFR 50.48(f). This change was made due to the decommissioning status of the plant using guidance from Regulatory Guide 1.191, Fire Protection for Nuclear Power Plants During Decommissioning and Permanent Shutdown.
* FSAR Change Package 2014-38: This change revised FSAR Chapter 1, Introduction and Summary, Section 1.1, Introduction, to document the NRC's acknowledgement of CR-3's certification of permanent cessation of operation and permanent removal of fuel from the reactor vessel. Additional discussion was included to clarify that system reclassification has occurred which takes precedence over safety classification identified in the FSAR, and that portions of this discussion would be removed after final system abandonment.

U.S. Nuclear Regulatory Commission                                                  Attachment 1 3F0516-05                                                                            Page 3 of 17
* FSAR Change Package 2014-40: This change updated the FSAR to incorporate changes approved by the NRC in the Certified Fuel Handler Training and Retraining Program and LAR #313, Revision 1, Revision to Improved Technical Specifications Administrative Controls for Permanently Defueled Conditions and Response for Request for Additional Information. This change affected FSAR Chapter 1, Introduction and Summary, Sections 1.3, Design Characteristics, 1.4, Principal Architectural and Design Criteria, and 1.7, Quality Program (Operational); FSAR Chapter 3, Reactor, Sections 3.2, Reactor Design, 3.4, Cycle 4 Conditions, 3.9, Cycle 8 Conditions, 3.10, Cycle 9 Conditions, 3.11, Cycle 1O Conditions, 3.12, Cycle 11 Conditions, 3.13, Cycle 12 Conditions, 3.14, Cycle 13 Conditions, 3.15, Cycle 14 Conditions, 3.16, Cycle 15 Conditions, 3.17, Cycle 16 Conditions, and 3.18, Cycle 17 Conditions; FSAR Chapter 4 Reactor Coolant System, Section 4.2, System Description and Operation, 4.3, System Design Evaluation, and 4.4, Tests and Inspections; FSAR Chapter 5, Containment System and Other Special Structures, Section 5.6, Testing; FSAR Chapter 7, Instrumentation and Controls, Section 7.1, Protection Systems; FSAR Chapter 9, Auxiliary and Emergency Systems, Section 9.6, Fuel Handling System; FSAR Chapter 10, Steam and Power Conversion Systems, Sections 10.6, Auxiliary Feedwater, and 10.7, Steam Generator Overfill; and FSAR Chapter 12, Conduct of Operations, Sections 12.2, Training, 12.3, Industrial Security, 12.4, Emergency Plan, 12.6, Plant Procedures, and 12. 7, Records.
* FSAR Change Package 2014-41: This change revised FSAR Chapter 9, Auxiliary and Emergency Systems, Section 9.3, Spent Fuel Cooling System, to incorporate the results of an analysis performed to assure structural integrity of the pool at temperatures up to 212 degrees Fahrenheit.
* FSAR Change Package 2014-58: This change revised the description of the plant UHF
* radio system power supplies from being backed up by the non-safety diesel generator to capable of being readily energized by the standby diesel generator and supported abandonment of the "C" Diesel Generator. This change affected FSAR Chapter 7, Instrumentation and Controls, Section 7.4, Operating Control Stations.
* FSAR Change Package 2014-59: This change revised the description details for the vital bus alternate source transformers. The plant replaced the regulating transformers with non-regulating transformers. The reconfigured plant electrical distribution system is justified due to the plant decommissioning and spent fuel storage status. This change affected FSAR Chapter 8, Electrical Systems, Section 8.2, Electrical System Design.
* FSAR Change Package 2014-60: This change revised a description of manual fire suppression equipment by removing the detail of how much hose is located at specific locations in the plant and surrounding buildings. This change affected FSAR Chapter 9, Auxiliary and Emergency Systems, Section 9.8, Plant Fire Protection Program.
* FSAR Change Package 2014-64: This change revised the description of the Spent Fuel Cooling System and the Control Complex Ventilation System to incorporate the new air cooled chillers. These chillers replace the plant cooling water sources from the plant intake used for spent fuel cooling with a pair of fifty percent capable chillers and the Control Complex Chillers were replaced with one air cooled chiller to allow for abandonment of the Raw Water Systems. This change affected FSAR Chapter 9, Auxiliary and Emergency Systems, Sections 9.3, Spent Fuel Cooling, 9. 7, Plant*
Ventilation Systems, and 9.12, References.

U. S. Nuclear Regulatory Commission                                                Attachment 1 3F0516-05                                                                           Page 4 of 17
* FSAR Change Package 2014-67:              This change revised the description of the Compressed Air System (IA) due to equipment replacement. The plant replaced the existing electric driven air compressors with two air cooled air compressors eliminating the need for cooling water, allowing for abandonment of the Raw Water Systems. This change. affected FSAR Chapter 9, Auxiliary and Emergency Systems, Section 9.10, Compressed Air.
* FSAR Change Package 2015-01: This change added the Alternate Dilution Flow Pump (RWP-6) discussion to the FSAR which permitted further abandonment of the Raw Water Systems. This is a lower flow pump that is used for providing dilution flow for permitted liquid releases. This change affected FSAR Chapter 11, Radioactive Waste and Radiation Protection, Section 11.2, Radioactive Waste Disposal Systems Summary.
* FSAR Change Package 2015-02: This change revised the section on site flooding to identify a lower flood level of 107 feet datum for non-safety related SSCs. A lower flood level permitted the plant to abandon structures and components used to provide flood protection. This change affected FSAR Chapter 2, Site and Environment, Sections, 2.4, Hydrology and 2.7, References.
* FSAR Change Package 2015-03: This change revised FSAR Chapter 12, Conduct of Operations, Section 12.4, Emergency Plan to reflect the implementation of the NRC approved Permanently Defueled Emergency Plan (PDEP). The Radiological Emergency Response Plan (RERP) was replaced with the PDEP and information that was no longer necessary was removed.
* FSAR Change Package 2015-11: This change added a commitment to FSAR Chapter 9, Auxiliary and Emergency Systems, Section 9.6, Fuel Handling System, that if all spent fuel assemblies have not been removed from the Spent Fuel Pool by December 31, 2019, CR-3 will, prior to that date, submit an amendment request pursuant to 10 CFR 50.90, to incorporate Baral and Carborundum Surveillance Programs into the CR-3 Permanently Defueled Technical Specifications. This was a CR-3 agreed upon action that supported NRC approval of Amendment #247.
* FSAR Change Package 2015-12: This change revised FSAR Chapter 9, Auxiliary and Emergency Systems, Table 9-18, Fire Protection Governing Documents, to remove the reference to Facility Operating License Condition 2.C(9). The removal of the Fire Protection License Condition was approved by the NRC in Amendment #247.
* FSAR Change Package 2016-05: This change revised FSAR Chapter 8, Electrical Systems, Figure 8-1, 500kV and 230kV Distribution, to reflect current plant configuration.
The changes included identifying that the disconnect switch between CR-3 and the 230kV switchyard is permanently open and identified the source of the 12kV alternate power for the plant.

U. S. Nuclear Regulatory Commission                                                 Attachment 1 3F0516-05                                                                            Page 5 of 17
: 8. Several editorial and clarification changes were made throughout the document. Each change was considered for 10 CFR 50.59 applicability using Crystal River Unit 3 and Duke Energy procedures:
* FSAR Change Package 2013-15: This change was an editorial change to clarify the name of several lines coming into the 230 kV switchyard in FSAR Chapter 8, Electrical Systems, Figure 8-1.
* FSAR Change Package 2015-04: This change revised the title of the individual responsible for the Special Nuclear Material (SNM) onsite from OTO Nuclear Engineer to Site SNM Custodian. This change was determined to be editorial as only the title of the position changed, not the duties. This change affected FSAR Chapter 12, Conduct of Operations, Section 12.7, Records.
* FSAR Change Package 2016-06: This change revised FSAR Chapter 1, Introduction and Summary, Table 1-3, Crystal River Unit 3 Quality Program Commitments, Regulatory Guide 1.123, "Quality Assurance Requirements for Control of Procurement of Items and Services for Nuclear Power Plants." This change corrected the numbering of the commitment clarifications and was determined to be an editorial change only.
C. Quality Assurance Program changes that did not reduce. commitments.
* FSAR Change Package 2013-18: This change is an editorial change to update departmental interfaces due to corporate organizational changes. This change affected FSAR Chapter 1, Introduction and Summary, Section 1.7, Quality Program (Operational). This change is an organizational change as provided for in 10 CFR 50.54(a)(3)(vi). This organizational revision does not impact individuals or organizations ability to perform QA functions. Those organizations continue to have the requisite authority and organizational freedom and sufficient independence to ensure that cost and schedule pressure does not compromise quality or nuclear safety considerations.
* FSAR Change Package 2014-10: This change deleted clarification 3 to Regulatory Guide 1.146, "Qualification of Quality Assurance Program Aud if Personnel for Nuclear Power Plants," and Paragraph 3.2 of ANSI N45.2.23. This clarification provided a grace period to the annual assessment for the maintenance of proficiency of lead auditors.
With the deletion of this clarification, the commitment reverts back to the words in the ANSI Standard. This assures that each lead auditor continues to be evaluated at least once every 12 months. This change is an administrative improvement that aligns with the ANSI Standard allowing the evaluation of proficiency to be completed for the work group.

U. S. Nuclear Regulatory Commission                                                  Attachment 1 3F0516-05                                                                             Page 6of17
* FSAR Change Package 2014-19: This change revised the organizational structure in the FSAR to reflect a revision to the Decommissioning Transition Organization. This change affected FSAR Chapter 1, Introduction and Summary, Section 1.7, Quality Program (Operational) and FSAR Chapter 12, Conduct of Operations, Sections 12.2, Training, 12.7, Records, and 12.8, Administrative Control.            This change is an organizational change as provided for in 10 CFR 50.54(a)(3)(vi). This organizational revision does not impact individuals or organizations ability to perform QA functions.
Those organizations continue to have the requisite authority and organizational freedom and sufficient independence to ensure that cost and schedule pressure does not compromise quality or nuclear safety considerations. The change did not reduce any
* quality commitments previously accepted by the NRC.
* FSAR Change Package 2014-57: This change revised the organizational structure of CR-3 to reflect the Duke Energy corporate and site structure for CR-3 while in decommissioning. This change affected FSAR Chapter 1, Introduction and Summary, Section 1.7, Quality Program (Operational), and FSAR Chapter 12, Conduct of Operations, Sections 12.1, Organization and Responsibility, and 12.2, Training. This change is an organizational change as provided for in 10 CFR 50.54(a)(3)(vi). This organizational revision does not impact individuals or organizations ability to perform QA functions. Those organizations continue to have the requisite authority and organizational freedom and sufficient independence to ensure that cost and schedule pressure does not compromise quality or nuclear safety considerations. The site organizational structure and responsibilities did not change. The proposed changes did not reduce any commitments previously accepted by the NRC.
* FSAR Change Package 2015-06: This change revised the CR-3 Quality Assurance Program Description (QAPD) in FSAR Chapter 1. Introduction and Summary, Section
: 1. 7 Quality Program (Operational), to address a concern identified by an industry organization (Nuclear Industry Evaluation Program - NIEP) and to include Augmented Quality in the Quality Assurance Plan Description (QAPD). The first change is to remove the statement, "Deviations from the QA Program shall be permitted only upon written authority from the Chief Nuclear Officer" from Section 1. 7.1 Introduction. This is considered a clarification to the QAPD that enhances the alignment with the 10 CFR 50.54(a) Regulations for the control of changes to the QAPD, as identified in FSAR Section 1. 7, and for intentional departures from License conditions in an emergency condition through the application of 10 CFR 50.54(x) and 10 CFR 50.54(y).
The second change added text for augmented quality to the QAPD This addition does not alter the application of 10 CFR 50 Appendix B to nuclear safety-related Structures, Systems, and Components. The QAPD still establishes clearly in writing, the authority and duties or persons or organizations performing activities affecting the safety-related functions of structures, systems, or components. The proposed change still meets the NRC requirements and does not reduce any quality commitments previously accepted by the NRC.

U.S. Nuclear Regulatory Commission                                                  Attachment 1 3F0516-05                                                                            Page 7 of 17
* FSAR Change Package 2015-08: This change revised the corporate and site organization to reflect the Duke Energy corporate structure. This change also reflects training program changes, and the reorganization of the Plant Nuclear Safety Committee (PNSC) due to decommissioning. This change affected FSAR Chapter 1, Introduction and Summary, Section 1.7, Quality Program (Operational); and FSAR Chapter 12, Conduct of Operations, Sections 12.2, Training, and 12.8, Administrative Control.
This change is an organizational change as provided for in 10 CFR 50.54(a)(3)(vi). This organizational revision does not impact individuals or organizations ability to perform QA functions. Those organizations continue to have the requisite authority and organizational freedom and sufficient independence to ensure that cost and schedule pressure does not compromise quality or nuclear safety considerations. This change does not reduce any quality commitments previously accepted by the NRC.
* FSAR Change Package 2015-1 O: This change revises FSAR Chapter 1, Introduction and Summary, Table 1-3, Crystal River Unit 3 Quality Program Commitments, to revise commitments to Regulatory Guide (RG) 1.88, "Collection, Storage, and Maintenance of Nuclear Power Plant Quality Assurance Records," and Regulatory Guide 1.123, "Quality Assurance Requirements for Control of Procurement of Items and Services for Nuclear Power Plants." The change to RG 1.88 is to add an exception that addresses management of electronic records and the change to RG 1.123 is to clarify existing exception 6.
The justification for the change to RG 1.88 made under 10 CFR 50.54(a) is that consistent with 50.54(a)(3(ii), the change to Table 1-3 uses a quality assurance alternative or exception approved by a NRC safety evaluation dated May 26, 2015 to Duke Energy Carolinas, ADAMS Accession No. ML15138A347. In the SE, the NRC staff evaluated the licensee's Quality Assurance Topical Report (QATR) submittal and concluded that the licensee's QA program continues to satisfy the requirements of Appendix B to 10 CFR 50. Specifically, the newer 2011 version of the Nuclear Information and Records Management Association (NIRMA) Guides provide additional implementing details that continue to meet the QA record requirements contained in Appendix B, Criterion XVI I. Since the CR-3 exception mirrors the exception added to the Duke Energy QATR, the proposed change is acceptable for CR-3.
The justification for the change to RG 1.123 made under 10 CFR 50.54(a), provides an acceptable method of compliance with the Appendix B requirements for control of purchased items and services. The predominant criteria of Appendix Bare Criteria I, IV, VII, XII, and XVII. As identified in NRC letter dated January 13, 2010 to Perry Johnson Laboratory Accreditation, Inc., Adams Accession No. ML100130016, implementing the clarification in the QA Program continues to satisfy the requirements of Appendix B to 10 CFR 50, "Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants." This change does not reduce any quality commitments previously accepted by the NRC.

U. S. Nuclear Regulatory Commission                                                  Attachment 1 3F0516-05                                                                             Page 8 of 17 D. Some changes are noteworthy because they involve removal of information from the FSAR.
These changes were evaluated per the guidance of Nuclear Energy Institute (NEI) 98-03, Revision 1, and determined to be appropriate:
* FSAR Change Package 2010-03: This change deleted a discussion of the Flammable Liquids Warehouse from the list of plant buildings that are protected by an automatic wet pipe fire suppression system, and to remove the standpipe from the list of fire hydrants in the warehouse area. This is to accommodate changes made in association with the Independent Spent Fuel Storage Installation, all modified portions are outside of the CR-3 Protected Area, and the fire service water for this building is supplied by Units 1 and 2.
This change affected FSAR Chapter 9, Auxiliary and Emergency Systems, Section 9.8 of the FSAR (Fire Protection).
* FSAR Change Package 2013-22: This change deleted a discussion of the Condenser Air Removal System (AR) from Chapter 11, Radioactive Waste and Radiation Protection, Section 11.2, Radioactive Waste Disposal Systems Summary. This system provided a means to monitor plant releases resulting from Primary to Secondary system leakage This change is to reflect abandonment of the AR system with the exception of valves used to isolate it from interfacing systems.
* FSAR Change Package 2013-24: This change deleted the containment isolation function from the FSAR. With the permanent removal of fuel from the reactor vessel and containment, the containment isolation valves no longer perform a fission product barrier function. This change affected FSAR Chapter 1, Introduction and Summary, Section 1.4, Principal Architectural and Design Criteria; and FSAR Chapter 5, Containment System and Other Special Structures, Section 5.3, Isolation System; and Section 5.6, Testing.
* FSAR Change Package 2013-25: This change deleted the discussion of the Circulating Water flood barriers and abandons the Circulating Water System (CW) and the Secondary Services Closed Cycle Cooling System (SC) due to the plant decommissioning status. This change affected FSAR Chapter 2, Site and Environment, Section 2.4, Hydrology; and FSAR Chapter 9, Auxiliary and Emergency Systems, Section 9.5, Cooling Water Systems.                                                     ,
* FSAR Change Package 2014-01: This change deleted the Condensate (CD) and Condensate Demineralizer (CX) systems due to abandonment activities. This change affected FSAR Chapter 1, Introduction and Summary, Section 1.9, Interactions between Crystal River Unit 3 and the Four Fossil Fired Plants; FSAR Chapter 10, Steam and Power Conversion System, Section 10.2, System Design and Operation; and FSAR Chapter 11, Radioactive Waste and Radiation Protection, Sections 11.2, Radioactive Waste Disposal Systems Summary; and 11.4, Radiation Monitoring System.
* FSAR Change Package 2014-02: This change removed Feedwater (FW), Main Steam (MS) and Re-heat Steam (RH) systems from the FSAR due to abandonment activities.
This change affected FSAR Chapter 4, Reactor Coolant System, Sections 4.2, System Description and Operation; FSAR Chapter 10, Steam and Power Conversion System, Sections 10.1, Design Bases, 10.2, System Design and Operation, 10.3, System Analysis, 10.4, Tests and Inspections, and 10.7, Steam Generator Overfill.

U. S. Nuclear Regulatory Commission                                                     Attachment 1 3F0516-05                                                                               Page 9of17
* FSAR Change package 2014-03: This change removed Figure 10-5, Auxiliary Steam as this system was completely abandoned. As part of complete abandonment, circuit breakers were left in the open position to completely de-energize this equipment. FSAR Chapter 10, Steam and Power Conversion System, Figure 10-5 is the only affected document.
* FSAR Change 2014-04: This change removed discussion associated with the Carbon Dioxide Gas (CO), Hydrogen Gas (HY) and Main Generator Gas (GG) systems due to abandonment activities. The Carbon Dioxide gas was used for fire suppression in specific areas of the plant - specifically oil lubricated bearings on turbines, the Hydrogen Gas was used for chemical addition to the Reactor Coolant System through the Make-up and Purification System, and the Generator Gas System used a different source of Hydrogen as coolant in the main generator. This change affected FSAR Table 2-1, Chemical Storage Facilities at Crystal River Energy Complex and FSAR Chapter 9, Auxiliary and Emergency Systems, Section 9.8, Plant Fire Protection Program.
* FSAR Change Package 2014-05: This change removed discussion of the Reactor Coolant Pump (RCP) Oil Collection System due to abandonment activities. The RCP Oil Collection System was designed to minimize oil leakage and corresponding fire risk for the RCPs. FSAR Section 9.8, Plant Fire Protection Program, was affected.
* FSAR Change Package 2014-07: This change removed all discussion of 10 CFR 50, Appendix R, Fire Protection Program for Nuclear Power Facilities Operating Prior to January 1, 1979, from the FSAR due to the CR-3 "Certification of Permanent Cessation of Operation and Permanent Removal of Fuel From the Reactor." This change affected FSAR Chapter 7, Instrumentation and Controls, Sections 7.4, Operating Control Stations; 7.5, Anticipated Transients without Scram; FSAR Chapter 8, Electrical Systems Section 8.2, Electrical System Design; FSAR Chapter 9, Auxiliary and Emergency Systems, Sections 9.5, Cooling Water Systems; 9.7, Plant Ventilation Systems; 9.8 Plant Fire Protection System; 9.9, Gas Tanks and Cylinders as Missiles; and Chapter 10, Steam and Power Conversion Systems, Section 10.6, Auxiliary Feedwater.
* FSAR Change Package 2014-08: This change removed references to the Electro-Hydraulic (EH) system due to this system being abandoned. The EH system was used to control the turbine generator system. This change affected FSAR Chapter 10, Steam and Power Conversion System, Sections 10.2, System Design and Operation; and 10.3, System Analysis.
* FSAR Change Package 2014-09: This change removed discussion and references to the Lubricating Oil (LO) system, the LO-Feedwater (LO-FW) system, LO-Turbine Building Seal Oil (LO-TB SO) and Turbine Generator System {TG) due to these systems being abandoned. This change affected FSAR Chapter 7, Instrumentation and Controls, Section 7.1, Protection Systems; FSAR Chapter 9, Auxiliary and Emergency Systems, Section 9.8, Plant Fire Protection Program; and FSAR Chapter 10, Steam and Power Conversion Systems, Sections 10.2, System Design and Operation; and 10.3, System Analysis.

U. S. Nuclear Regulatory Commission                                               Attachment 1 3F0516-05                                                                        Page 10of17
* FSAR Change Package 2014-11: This change removed discussion of the Decay Heat Closed Cycle Cooling System ventilation (AH-XF) due to system abandonment. This was a ventilation system used to provide cool air to the Decay Heat Closed Cycle System. This change affected FSAR Chapter 9, Auxiliary and Emergency Systems, Sections 9.5, Cooling Water Systems and 9.7, Plant Ventilation Systems.
* FSAR Change Package 2014-12: This change removed discussion of the Penetration Area Cooling System (AH-XP) due to abandonment of the system. This change affected FSAR Chapter 5, Containment System and Other Special Structures, Section 5.1, Structural Design Classification; and FSAR Chapter 9, Auxiliary and Emergency Systems, Section 9.7, Plant Ventilation Systems.
* FSAR Change Package 2014-13: This change removed discussion of the Reactor Coolant System Pressurizer Heaters to support system abandonment. This change affected FSAR Chapter 4, Reactor Coolant System, Sections 4.2, System Description and operation, 4.3, System Design Evaluation, 4.4, Tests and Inspections; FSAR Chapter 7, Instrumentation and Control, Section 7.3, Instrumentation, 7.4, Operating Control Stations; and FSAR Chapter 8, Electrical Systems, Section 8.2, Electrical System Design.
* FSAR Change Package 2014-14: This change removed all discussion of the Reactor Coolant Pump Power Monitors to support RCPPM system abandonment. This system would trip the reactor if an under-power or over-power condition existed at the pump to protect the core. This change affected FSAR Chapter 7, Instrumentation and Control, Sections 7.1, Protection System, 7.2, Control Systems, and 7.3, Instrumentation.
* FSAR Change Package 2014-15: This change removed discussion of (or makes historical) the Emergency Feedwater System (EF) and the Emergency Feedwater Initiation and Control System (EFIC) due to abandonment of the systems. This change affected FSAR Chapter 1, Introduction and Summary, Section 1.3, Design Characteristics; FSAR Chapter 2, Site and Environment, Section 2.5, Engineering Geology and Foundation Considerations; FSAR Chapter 5, Containment System and Other Special Structures, Sections 5.1, Structural Design Classification, 5.2, Reactor Building, 5.4, Other Class I Structures and Systems; FSAR Chapter 7, Instrumentation and Control, Sections 7.2, Control Systems, 7.4, Operating Control Stations; FSAR Chapter 8, Electrical Systems, Section 8.2, Electrical System Design; FSAR Chapter 9, Auxiliary and Emergency Systems, Section 9.7, Plant Ventilation Systems; and FSAR Chapter 10, Steam and Power Conversion Systems, Sections 10.5, Emergency Feedwater System, 10.6, Auxiliary Feedwater.
* FSAR Change Package 2014-62: This change removed discussion of the Non-Nuclear Instrumentation (NNI) Syste'm from FSAR Chapter 7, Instrumentation and Controls, Section 7.3.2, Process Instrumentation due to abandonment activities.              This instrumentation system was used to provide process controls for startup, operation and shutdown of the reactor system.

U.S. Nuclear Regulatory Commission                                                 Attachment 1 3F0516-05                                                                         Page 11of17
* FSAR Change Package 2014-18: This change deleted information pertaining to the Reactor Building equipment and personnel air locks due to abandonment activities. The Reactor Building integrity is no longer required for fission product containment. This change affected FSAR Chapter 5, Containment System and Other Special Structures, Section 5.2, Reactor Building, and Section 5.6, Testing.
* FSAR Change Package 2014-23:               This change removed the discussion of Uninterruptable Power Supply (UPS), Inverter, and Class 1E from the FSAR to support system abandonments. This change affected FSAR Chapter 7, Instrumentation and Controls, Sections 7.1, Protection Systems, 7.5, Anticipated Transients Without Scram; FSAR Chapter 8, Electrical Systems, Section 8.2, Electrical System Design; and FSAR Chapter 11, Radioactive Waste and Radiation Protection, Section 11.4, Radiation Monitoring System.
* FSAR Change Package 2014-24: This change removed discussion of the Loose Parts Monitoring System due to abandonment activities. This change affected' FSAR Chapter 7, Instrumentation and Controls, Sections 7.3, Instrumentation, and 7.4, Operating Control Stations.
* FSAR Change Package 2014-25: This change removed discussion related to the lncore Monitoring System due to system abandonment activities. This change affected FSAR Chapter 1, Introduction and Summary, Section 1.3, Design Characteristics, and FSAR Chapter 7, Instrumentation and Controls, Section 7.3, Instruments.
* FSAR Change Package 2014-26: This change removed discussion associated with the Chemical Addition System (CA) and Liquid Sampling System due to system abandonment. These systems were used to maintain Reactor Coolant System water quality as well as deliver the necessary chemicals to other systems. The Liquid Sampling System included the Post Accident Sampling System as well as other sampling points to determine water quality and in the event of an accident, the magnitude of core damage. This change affected FSAR Chapter 4, Reactor Coolant System, Sections 4.1, Design Basis, 4.2, System Description and Operation, and FSAR Chapter 9, Auxiliary and Emergency Systems, Sections 9.2, Chemical Addition and Liquid Sampling System, and 9.11, Post Accident Sampling System.
* FSAR Change Package 2014-27: This change removed discussion related to the Control Rod Drive System (DR) due to system abandonment activities. This change affected FSAR Chapter 1, Introduction and Summary, Sections 1.1, Introduction, 1.3, Design Characteristics, 1.4, Principle Architectural and Design Criteria; FSAR Chapter 3, Reactor, Sections 3.1, Design Bases, 3.2, Reactor Design, 3.3, Tests and Inspections; and FSAR Chapter 7, Instrumentation and Controls, Sections 7.1, Protection Systems, 7.2, Control Systems.
* FSAR Change Package 2014-28: This change revised references to remove the "A" Train of the Raw Water (RW) system due to system abandonment activities. The purpose of this revision is to conform the FSAR description of the equipment and systems to the current status of the plant. This change affected FSAR Chapter 4, Reactor Coolant System, Section 4.2, System Description and Operation; FSAR Chapter 5, Containment System and Other Special Structures, Section 5.1, Structural Design Classification; FSAR Chapter 9, Auxiliary and Emergency Systems, Sections 9.3, Spent Fuel Cooling System, 9.4, Decay Heat Removal System, and 9.5, Cooling Water Systems.

U. S. Nuclear Regulatory Commission                                              Attachment 1 3F0516-05                                                                       Page 12 of 17
* FSAR Change Package 2014-29: This change removed information and references related to the Building Spray (BS), Core Flood (CF), Decay Heat (DH), and Makeup and Purification (MU) Systems (including the discussions pertaining to the Emergency Core Cooling System) due to system abandonment activities. This change affected FSAR Chapter 2, Site and Environment, Section 2.4, Hydrology; FSAR Chapter 4, Reactor Coolant System, Sections 4.2, System Description and Operation, 4.3, System Design Evaluation; FSAR Chapter 6, Engineered. Safeguards - all Sections; FSAR Chapter 7, Instrumentation and Controls, Sections 7.1, Protection Systems, 7.3, Instrumentation; FSAR Chapter 8, Electrical Systems, Section 8.2, Electrical System Design; FSAR Chapter 9, Auxiliary and Emergency Systems, Sections 9.1, Makeup and Purification System, 9.3, Spent Fuel Cooling System, 9.4Decay Heat Removal, 9.5, Cooling Water Systems, 9.6, Fuel Handling System, and 9.11, Post Accident Sampling System; FSAR Chapter 10, Steam and Power Conversion System, Sections 10.2, System Design and Operation, 10.3, System Analysis, and 10.5, Emergency Feedwater System.
* FSAR Change Package 2014-30: This change removed all information related to the Anticipated Transients Without Scram System (ATWS) due to abandonment activities.
This change affected FSAR Chapter 7, Instrumentation and Controls, Section 7.5, Anticipated Transients Without Scram System.
* FSAR Change Package 2014-31: This change revised the FSAR to remove information related to the Reactor Protection System (RPS) due to abandonment activities. This change affected FSAR Chapter 1, Introduction and Summary, Section 1.4, Principle Architectural and Design Criteria; and FSAR Chapter 7, Instrumentation and Controls, Section 7 .1, Protection Systems, Section 7.2, Control Systems, and Section 7.4, Operating Control Stations.
* FSAR Change Package 2014-32: This change removed information related to the Nuclear Instrumentation System (NI) due to abandonment activities. This change affected FSAR Chapter 7,, Instrumentation and Controls, Sections 7.1, Protection Systems, and 7.3, Instrumentation.
* FSAR Change Package 2014-33: This change removed information pertaining to specific plant ventilation systems due to abandonment activities. These are the Intermediate Building Ventilation System (AH-XM), Turbine Building Ventilation System (AH-XN), and the Emergency Feedwater Pump Building Ventilation System (AH-XU).
This change affected FSAR Chapter 9, Auxiliary and Emergency Systems, Section 9.7, Plant Ventilation Systems.
* FSAR Change Package 2014-37: This change removed discussion of specific radiation monitors that were evaluated to be available for abandonment during the early phase of the Radiation Monitor System abandonment. This change affected FSAR Chapter 11, Radioactive Waste and Radiation Protection, Section 11.4, Radiation Monitoring System.
* FSAR Change Package 2014-39: This change removed all discussion of heat trace system references due to abandonment *activities. This change affected FSAR Chapter 4, Reactor Coolant System, Section 4.2, System Description and Operation; and FSAR Chapter 9, Auxiliary and Emergency Systems, Sections 9.2, Chemical Addition System and Liquid Sampling System, 9.4, Decay Heat Removal System, and 9.7, Plant Ventilation Systems.

U. S. Nuclear Regulatory Commission                                                   Attachment 1 3F0516-05                                                                            Page 13of17
* FSAR Change Package 2014-42: This change removed discussion related to the Containment Leak Rate Test System (LR) including discussion of the containment purge capability, due to abandonment activities. Local leak rate testing through valves and other penetrations was not included in this change package. This change affected FSAR Chapter 5, Containment System and Other Special Structures, Section 5.6, Testing; and FSAR Chapter 9, Auxiliary and Emergency Systems, Section 9.7, Plant Ventilation Systems.
* FSAR Change Package 2014-43: This change removed the discussion in FSAR Chapter 9, Auxiliary and Emergency Systems, Section 9.5, Cooling Water Systems, specific to the Traveling Screen for the Raw Water System due to abandonment activities.
* FSAR Change Package 2014-44: This change deleted reference to Regulatory Guide 1.97, "Criteria for Accident Monitoring Instrumentation for Nuclear Power Plants," and mention of the Post Accident Monitoring System due to the decommissioning status of CR-3. This change affected FSAR Chapter 6, Engineered Safeguards, Section 6.2, Reactor Building Spray System; FSAR Chapter 7, Instrumentation and Controls, Section 7.3, Instrumentation, 7.4, Operating Control Stations, and 7.5, Anticipated Transients without Scram Systems; and FSAR Chapter 11, Radioactive Waste and Radiation Protection, Section 11.4, Radiation Monitoring System.
* FSAR Change Package 2014-45: This change deleted portions of the FSAR Chapter 9, Auxiliary and Emergency Systems, Section 9.7, Plant Ventilation Systems, to allow for partial abandonment of the Emergency Diesel Generator Air Handling System (AH-XL) due to the abandonment of the A and C Diesel Generators. The B AH-XL system remains available to support the B Diesel Generator.
* FSAR Change Package 2014-46: This change removed information referring to the Reactor Building Ventilation System (AH-XA, and AH-XB) due to abandonment activities. This change affected FSAR Chapter 5, Containment System and Other Special Structures, Section 5.5, Ventilation and Purge Systems; and FSAR Chapter 9, Auxiliary and Emergency Systems, Section 9.7, Plant Ventilation Systems.
* FSAR Change Package 2014-47: This change removed discussion related to the charcoal filter portion of the plant ventilation system due to decay of the iodine isotopes that would be absorbed by these filters and abandonment activities. This change affected FSAR Chapter 1, Introduction and Summary, Section 1.4, Principal Architectural and Design Criteria; FSAR Chapter 5, Containment System and Other Special Structures, Section 5.5, Ventilation and Purge Systems; FSAR Chapter 9, Auxiliary and Emergency Systems, Sections 9.5, Cooling Water Systems and Section 9.7, Plant Ventilation Systems; FSAR Chapter 11, Radioactive Waste and Radiation Protection, Section 11.2, Radioactive Waste Disposal Systems Summary; and FSAR Chapter 14, Safety Analysis, Section 14.2, Standby Safeguards Analysis.
* FSAR Change Package 2014-48: This change removed all discussion of the Integrated Control System (ICS) from FSAR Chapter 7, Instrumentation and Controls, Section 7.2,
* Controls and Section 7.3, Instrumentation, due to abandonment activities. The ICS provided the proper coordination of the Reactor, Steam Generator Feedwater control, and Turbine control.

U. S. Nuclear Regulatory Commission                                                  Attachment 1 3F0516-05                                                                           Page 14 of 17
* FSAR Change Package 2014-50: This change removed discussion related to the Waste Gas System (>NG) and radioactive gases due to abandonment activities. The WG System has been vented, purged, and left open to the atmosphere to preclude accumulation of additional radioactive gasses. This change affected FSAR Chapter 4, Reactor Coolant System, Section 4.2, System Description and Operation; FSAR Chapter 6, Engineered Safeguards, Section 6.4, Engineered Safeguards Leakage and Radiation Considerations; FSAR Chapter 11, Radioactive Waste and Radiation Protection, Sections 11.1, Design Performance Objectives, 11.2, Radioactive Waste Disposal Systems Summary, and 11.3, Radiation Shielding.
* FSAR Change Package 2014-51: This change removed portions of FSAR Chapter 9, Auxiliary and Emergency Systems, Section 9.10, Compressed Air, to conform to the partial abandonment of the Instrument Air System (IA) and the Station Air System (SA).
* FSAR Change Package 2014-52: This change removed information related to the minimum tide hurricane and minimum depth of the intake canal. The reason for this change is that the volumetric water requirements for the plant are significantly less with CR-3 in permanent shutdown. This change affected FSAR Chapter 2, Site and Environment, Sections 2.4, Hydrology, and 2.7, References; and FSAR Chapter 9, Auxiliary and Emergency Systems, Sections 9.5, Cooling Water Systems, and 9.12, References.
* FSAR Change Package 2014-53: This change removed the discussion of the Engineered Safeguards System (ES) and the Engineered Safeguards Actuation System (ESAS) due to abandonment activities. This change affected FSAR Chapter 1, Introduction and Summary, Sections 1.2, Summary Plant Description, 1.3, Design Characteristics, and 1.4, Principal Architectural and Design Criteria; FSAR Chapter 4, Reactor Coolant System, Section 4.2, System Description and Operation, FSAR Chapter 5, Containment System and Other Special Structures, Sections 5.1, Structural Design Classification, 5.3, Isolation System, 5.4, Other Class I Structures and Systems, and 5.5, Ventilation and Purge Systems; FSAR Chapter 6, Engineered Safeguards, all Sections; FSAR Chapter 7, Instrumentation and Controls, Sections 7.1, Protection Systems, 7.2, Control Systems, 7.3, Instrumentation, and 7.4, Operating Control Stations; FSAR Chapter 8, Electrical Systems, Section 8.2, Electrical System Design; and FSAR Chapter 9, Auxiliary and Emergency Systems, Sections 9.1, Makeup and Purification System, 9.2, Chemical Addition System and Liquid Sampling System, 9.5, Cooling Water Systems, 9. 7, Plant Ventilation Systems, 9.10, Compressed Air, and 9.11, Post Accident Sampling System.
* FSAR Change Package 2014-54: This change removed the discussion of the Decay Heat Closed Cycle Cooling Water System (DC) due to abandonment activities. This change affected FSAR Chapter 4, Reactor Coolant System, Section 4.2, System Description and Operation; FSAR Chapter 5, Containment System and Other Special Structures, Section 5.1, Structural Design Classification; FSAR Chapter 9, Auxiliary and.
Emergency Systems, Section 9.5, Cooling Water Systems; and FSAR Chapter 11, Radioactive Waste and Radiation Protection, Section 11.4, Radiation Monitoring System.

U.S. Nuclear Regulatory Commission                                                    Attachment 1 3F0516-05                                                                            Page 15of17
* FSAR Change package 2014-55: This change removed discussion of the Industrial Cooling (Cl) System due to abandonment activities. This change affected FSAR Chapter 5, Containment System and Other Special Structures, Section 5.5, Ventilation and Purge Systems; FSAR Chapter 6, Engineered Safeguards, Section 6.3, Reactor Building Emergency Cooling System; and FSAR Chapter 9, Auxiliary and Emergency Systems,. Section 9.3, Spent Fuel Cooling System.
* FSAR Change Package 2014-56: This change removed the need to maintain self-contained breathing apparatus (SCBA) inside the main control room. The reason for this is that the potential of an event causing high radiation are sufficiently reduced to not require this additional form of protection for the control room staff due to the permanently defueled status of the plant. Additional changes included the reclassification of the Control complex as a Class Ill structure and to clarify that the safety function of shutting down the reactor no longer exists. This change affected FSAR Chapter 1, Introduction and Summary, Section 1.4, Principal Architectural and Design Criteria; and FSAR Chapter 7, Instrumentation and Controls, Section 7.4, Operating Control Stations.
* FSAR Change Package 2014-61: This change removed discussion about the primary processing chain of the Liquid Waste System (LW) due to abandonment activities. This change affected FSAR Chapter 11, Radioactive Waste and Radiation Protection, Section 11.2, Radioactive Waste Disposal Systems Summary.
* FSAR Change Package 2014-63: This change removed discussion related to the Control Complex Emergency Ventilation System (AH-XK) due to abandonment activities.
The normal duty Control Complex Ventilation is not affected by this change. This change affected FSAR Chapter 1, Introduction and Summary, Section 1.4, Principal Architectural and Design Criteria; FSAR Chapter 2, Site and Environment, Section 2.2, Site and Adjacent Areas; FSAR Chapter 7, Instrumentation and Controls, Sections 7.4, Operating Control Stations, 7.6, References; and FSAR Chapter 9, Auxiliary and Emergency Systems, Section 9.7, Plant Ventilation Systems.
* FSAR Change Package 2014-65: This change removed discussion related to the Remote Shutdown System (RSS) due to abandonment activities. This change affected FSAR Chapter 7, Instrumentation and Controls, Sections 7.3, Instrumentation, 7.4, Operating Control Stations, and 7.5, Anticipated Transients Without Scram System; FSAR Chapter 9, Auxiliary and Emergency Systems, Section 9.7, Plant Ventilation Systems; and FSAR Chapter 1O Steam and Power Conversion System, Sections 10.5, Emergency Feedwater System, and 10.6, Auxiliary Feedwater.
* FSAR Change Package 2014-66: This change removed information related to the Fire Service System (FS) due to partial system abandonment. Portions of the system abandoned include the water curtain for the Unit Auxiliary, Startup, Backup ES, and Main Step up Transformers; one Fire Service Tank, one of the two Diesel powered Fire Service Pumps, thermal fire detection equipment, and the fixed water spray fire suppression system. This change affected FSAR Chapter 8, Electrical Systems, Section 8.2, Electrical System Design; and FSAR Chapter 9, Auxiliary and Emergency Systems, Section 9.8, Plant Fire Protection Program.

U. S. Nuclear Regulatory Commission                                                 Attachment 1 3F0516-05                                                                          Page 16 of 17
* FSAR Change Package 2014-68: This change removed information related to the Nuclear Services Closed Cycle Cooling System (SW), Nuclear Service Seawater (RW-SW), and Decay Heat Seawater (RW-DC) Systems due to abandonment activities. This change affected FSAR Chapter 2, Site and Environment, Section 2.4, Hydrology; FSAR Chapter 5, Containment System and Other Special Structures, Section 5.1, Structural Design Classification; and FSAR Chapter 9, Auxiliary and Emergency Systems, Section 9.5, Cooling Water Systems.
* FSAR Change Package 2014-69: This change removed discussion related to the Chilled Water Cooling System for Control Complex and Penetration Cooling due to abandonment activities. This change affected FSAR Chapter 5, Containment System and Other Special Structures, Section 5.1, Structural Design Classification; FSAR Chapter 7, Instrumentation and Control, Section 7.4, Operating Control Stations; FSAR Chapter 9, Auxiliary and Emergency Systems, Sections 9.5, Cooling Water Systems, and 9.7, Plant Ventilation Systems.
* FSAR Change Package 2014-70: This change deleted information related to the Appendix R Chilled Water System (CH-AR), Appendix R Control Complex Dedicated Cooling System (AH-XT), and the "B" EFIC Room Cooling System (AH-XS) due to abandonment activities.      This change affected FSAR Chapter 9, Auxiliary and Emergency Systems, Section 9.7, Plant Ventilation Systems.
* FSAR Change Package 2015-05: This change Removed some information related to the Reactor Building Purge System (AH-XC) to reflect partial abandonment of the system. This change affected FSAR Chapter 5, Containment System and Other Special Structures, Sections 5.1, Structural Design Classification, 5.2, Reactor Building, 5.3, Isolation System, and 5.5, Ventilation and Purge Systems; FSAR Chapter 9, Auxiliary and Emergency Systems, Section 9.7, Plant Ventilation Systems; and FSAR Chapter 11, Radioactive Waste and Radiation Protection, Section 11.2, Radioactive Waste Disposal Systems Summary, and 11.4, Radiation Monitoring System.
* FSAR Change Package 2015-07: This change removed information in Chapter 2, Site and Environment, Section 2.3, Meteorology, related to the existing meteorological instrumentation and adds a discussion of alternative sources of meteorological information to be used if necessary. This change supported abandonment of the onsite meteorological instrumentation.
* FSAR Change Package 2015-09: This change removed information related to the Station Drain System (SD) to reflect the partial abandonment of the system. This change affected FSAR Chapter 2, Site and Environment, Section 2.4, Hydrology.
* FSAR Change Package 2015-13: This change removed discussion related to the Seismic Monitoring Instrumentation System (SI) due to abandonment activities. This change affected FSAR Chapter 2, Site and Environment, Section 2.5, Engineering Geology and Foundation Considerations; and FSAR Chapter 5, Containment System and Other Special Structures, Sections 5.1, Structural Design Classification.

U. S. Nuclear Regulatory Commission                                                Attachment 1 3F0516-05                                                                         Page 17of17
* FSAR Change Package 2016-01: This change removed portions of FSAR Chapter 1, Introduction and Summary, to correct portions that were missed in other abandonment packages and to reflect the SAFSTOR condition of the plant. This change affected Sections 1.1, Introduction, 1.2, Summary Plant Description, 1.4, Principal Architectural and Design Criteria, 1.8, Identification of Agents and Contractors, and 1.9, Interactions Between Crystal River Unit 3 and the Four Fossil Fired Plants.
* FSAR Change Package 2016-03: This change removed specific portions of FSAR Chapter 7, Instrumentation and Controls, and revised portions of Section 7.4, Operating Control Stations, to reflect current plant configuration and status. Specific discussions are no longer valid and can be eliminated since they were specific to operating requirements.

DUKE ENERGY FLORIDA, INC.
CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302 I LICENSE NUMBER DPR-72 ATTACHMENT 2 10 CFR 50.59 AND 10 CFR 72.48 REPORT- MAY 2016

U. S. Nuclear Regulatory Commission                                             Attachment 2 3F0516-05                                                                        Page 1 of 13 10 CFR 50.59 and 72.48 Evaluation Summaries Table of 10 CFR 50.59 Evaluations ID Number AR  676444           AC Vital Bus Partial Abandonment AR  687084           Shutdown of Spent Fuel Cooling AR  689505           Decay Heat System Abandonment AR  693221           Structural Evaluation of Spent Fuel Pool to 212 Degrees Fahrenheit AR  693870           Partial Abandonment of Instrument Air And Station Air Systems AR  699191           Retirement of PT-501, Intake Canal Survey AR  706204           FSAR Change to Eliminate SCBA Requirements AR  709569           Installation of Air Cooled Chillers AR  730617           Electrical System Changes for SAFSTOR 1 AR  746109           Flood Level Change AR  1984169          Abandonment of the Seismic Instrumentation System AR= Action Request No 10 CFR 72.48 evaluations were completed by CR-3 during this period.

U.S. Nuclear Regulatory Commission                                                  Attachment 2 3F0516-05                                                                           Page 2of13 ID Number:              AR 676444 Title                   AC Vital Bus Partial Abandonment Summary and Conclusions This evaluation is for the partial abandonment of the AC Vital Bus (VB) System for the decommissioning process.
Redundancy and reliability of power to the Vital Bus Distribution Panels and associated plant loads will be reduced with removal of battery backup and the inverters. On a loss of offsite power, the "A" ES Bus will not automatically be reenergized; it must be manually aligned to EGDG-1 B. Thus on a loss of offsite power, Vital Bus power is lost. "B" and "D" channels are restored automatically with EGDG-1 B, while "A" and "C" channels would be restored manually.
With the plant in the decommissioning status the design function to automatically repower both ES busses (and therefore, the associated Vital Busses) is no longer required. The credible accident and transient scenarios that existed for the plant in power operation during heat-up, reactor operation, and cool down are no longer possible. Therefore, the design functions required to respond to those events are no longer required per the limitations of 10 CFR 50.82(a)(2). Repowering the "B" ES bus will occur automatically, and repowering the "A" ES bus can be accomplished manually within minutes to hours, well within the time available to restore forced SFP cooling to protect the health and safety of the public.
The abandonment categorization envelope associated with OTO AR 676435 includes selected portions of the VB System, including all the inverters (VBIT-1A through -1 F), normal and battery inputs, as well as 120VAC Regulated Distribution Panels (VBDP-12 through -16), and associated transformers (VBTR-2A through -20 and VBTR-3A through VBTR-30). Required vital bus loads will be supplied using the alternate power transformers.
The Evaluation determined that no prior NRC approval is required to implement these activities.

U. S. Nuclear Regulatory Commission                                                  Attachment 2 3F0516-05                                                                            Page 3of13 ID Number:               AR 687084 Title                    Shutdown of Spent Fuel Cooling Summary and Conclusions The Spent Fuel (SF) Cooling system is designed to remove decay heat from the stored fuel assemblies and to maintain the water clarity in the SF pools. The SF system also limits radioactive fission product release to the outside environment following a fuel assembly rupture in the SF pools. The system assures that irradiated fuel assemblies in the SF Pools do not achieve a critical state. The main components in the SF system include two heat exchangers, two main coolant pumps, a demineralizer, two filters, a borated water recirculation pump, and instrumentation that monitors SF pool levels, temperatures, and pump flow rate._
This 10CFRS0.59 evaluates a change to FSAR section 9.3.1, Spent Fuel Cooling System Design Bases, where the following wording was added: "Due to the thermal storage capacity of the pool water, the spent fuel cooling system may be shut down for certain time periods to support maintenance activities; periods when forced cooling is not required; or to validate spent fuel pool heat-up rates." Operating procedure. instructions (OP-406) will be developed to implement this change.
FSAR section 9.3 was reviewed and it was determined that shutting down all the spent fuel cooling pumps does not adversely affect any of the design functions for the SF system. During normal operation, one SF pump (SFP-1A or SFP-1 B) is in operation maintaining the pool temperature below its design limit. A 140&deg;F temperature alarm is provided to alert control room personnel prior to reaching the SF pool designed temperature limit. In addition to the SF pool temperature alarm, temperature indication is provided in the control room. With no SF pumps running, the large quantity of SF pool water inventory continues to remove decay heat from the fuel assemblies. Heat is removed by conduction through the SF pool walls and evaporation off the surface of the water. Water is routinely added to the SF pools to maintain level via the Demineralized Water (OW) system. The ability to maintain SF pool level will not be affected by shutting down the running SF pumps. This alignment is acceptable for limited periods of time.
Prior to reaching the spent fuel pool design limit or at any time when system operation is desired, the system can be restarted and forced cooling reestablished. It is concluded that operation within the design temperature limits of the system will not have an adverse effect on the SF function to remove decay heat from stored fuel assemblies.
The SF system limits radioactive fission product release during a Fuel Handling Accident by providing at least 21 feet of water above the assumed failed fuel assembly. Since shutting down the SF pumps will not adversely affect the water level in the SF pool, this design function will not be affected.
The last design function of the SF system is to assure that the fuel assemblies in the SF pool do not achieve a critical state. Since the SF pool temperature will be maintained within its design limits and this change does not affect the fuel geometry, rack design, or boron concentration in the SF pools, this change does not affect this design function.
It is concluded that a license amendment is not required for this change since the SF system will be operated within its design limits, will not affect equipment used or the consequences of an accident, nor will it create a new accident.

U. S. Nuclear Regulatory Commission                                                  Attachment 2 3F0516-05                                                                            Page 4of13 ID Number:              AR 689505 Title                   Decay Heat System Abandonment Summary and Conclusions The purpose of abandoning the Decay Heat (DH) system is to facilitate SAFSTOR at CR3. Due to permanent cessation of operations CR3 is no longer licensed to store fuel in the reactor vessel or commence power operations. CR3 has been safely shutdown since September 26, 2009. Systems within the facility are currently being evaluated under the site procedure Al-9003, "System Evaluation, Categorization, and Abandonment." The DH system's primary purpose was to provide cooling to the Reactor Coolant system during shutdown, refueling, and accident scenarios. The DH system was also capable of providing cooling and makeup water to the SF Pools.
The Al-9003 process has determined that the DH system is no longer required. This Screen/Evaluation examined the impact system abandonment has on our regulatory requirements and if prior approval by the NRC is required. This Screen/Evaluation also examines necessary FSAR changes and any conforming procedure changes.
Calculation F13-0003 "CR-3 Spent Fuel Pool Time to Uncover Fuel Analysis" conservatively calculated that under a complete loss of all forced cooling and no additional make-up water it would take 23 days for the water level of the SF pools to reach 10 ft above the fuel storage racks. Furthermore the FSAR states that the SF system is not designed to be single failure proof, it is only required to be a reliable form of cooling. Because the site will still be able to provide a reliable source of SF cooling, and if an event were to occur there is more than an adequate time period to respond with mitigative actions, and none of the responses in the evaluation were answered yes, this evaluation has determined that prior NRC approval is not required for this change.

U. S. Nuclear Regulatory Commission                                                    Attachment 2 3F0516-05                                                                              Page 5of13 ID Number:              AR 693221 Title                   Structural Evaluation of Spent Fuel Pool to 212 Degrees Fahrenheit Summary and Conclusions The activity is to incorporate the results of Calculation 814-0001, "Structural Evaluation of Spent Fuel Pool Heat-up" into the CR-3 Final Safety Analysis Report. The calculation determined that the spent fuel pools and pool liner will retain structural integrity and leak tightness in the event that an extended loss of cooling were to occur and coolant inventory were to reach boiling temperature.
In regards to previously analyzed accidents the only remaining accident for CR-3 is a fuel handling accident (FHA). This analytical change cannot affect the frequency of occurrence or consequences of a FHA. This activity does not create the possibility of any new accident since it does not change any plant equipment or how it is operated.
The FSAR identifies that loss of spent fuel cooling (malfunction) could occur since the cooling function is not single failure proof, however, this activity does not increase the likelihood or consequences of a malfunction since it does not change any cooling equipment or how it is operated. Likewise there is no different result due to loss of cooling since pool integrity is maintained .
. Based on fuel pool heatup calculations following more than 4 years of decay, it can be determined that the operating core design basis limit for the fuel cladding will not be exceeded.
This activity is not changing the design basis limit for the fuel cladding.
There is no evaluation methodology for fuel pool integrity vs. temperature in the FSAR so the" calculation methodology for boiling in the fuel pools is not a change to an evaluation methodology.
The Evaluation determined that no prior NRC approval is required to implement these activities.

U. S. Nuclear Regulatory Commission                                                    Attachment 2 3F0516-05                                                                               Page 6of13 ID Number:             AR 693870 Title                   Partial Abandonment of Instrument Air And Station Air Systems Summary and Conclusions This evaluation assesses the impact from the reduction in Instrument and Station Air normal and backup compressors.           The current FSAR Section 9.10 recognizes the following automatically available air supplies:
* Three electric driven compressors
* One diesel driven air compressor
* Sixty-four bottles in the station air system as a backup to instrument air The partial abandonment of the air systems will reduce the available air supplies to:
* Two electric driven compressors Much of the equipment supported by the air systems have been or are being abandoned. The abandonment of these systems removes a majority of the previously described FSAR functions that were supported by the air systems. Additionally, the abandonment of systems has resulted in de-energizing many of the loads that contributed to ambient heating. With these loads de-energized heatup rates are reduced allowing much more time to address loss of ventilation that may be caused by a loss of air. Also, by assuming the ventilation continues to operate, the release of all of the activity due to a fuel handling accident occurs over a short duration (two hours).instead of a much longer duration if there was no ventilation operating.                 This concentration of the release ensures a conservative calculation of the offsite doses. Thus the only impact of a loss of ventilation (due to loss of air) is only to reduce the consequences of the accident by slowing the release rate.
Additionally, Calculation F13-0003 "CR-3 Spent Fuel Pool Time to Uncover Fuel Analysis,"
determined that using conservative assumptions which credit no active SSCs functions, the time for the spent fuel pools to heat up and boil the inventory off to 10 ft above the fuel storage racks would be 23 days. This duration provides ample time to effect repairs of existing cooling water loops. This calculation also assumes no forced cooling from the ventilation or spent fuel coolant systems and no actions to replace pool inventory during that time. Thus the urgency of maintaining Spent Fuel cooling and inventory makeup to protect fuel clad is greatly reduced.
FSAR Section 9.10 describes the instrument air functions in terms of achieving safe shutdown (Section 9.10.2). CR-3 is in a permanently shut-down condition, thus this consideration is no longer applicable. Current procedural guidance directs Operations to promptly restore the air compressor backup when the backup is out of service.
Abandonment of IAP-4, IAP-3A and the SA bottles does not significantly impact the probability of malfunctions. The abandonment of these defense in depth air supplies does not change any evaluated accidents. This change can be implemented without NRC approval.

U. S. Nuclear Regulatory Commission                                                  Attachment 2 3F0516-05                                                                             Page 7of13 ID Number:              AR 699191 Title                   Retirement of PT-501. Intake Canal Survey Summary and Conclusions CR3 was designed to utilize the Intake Canal as the UHS, and operability of the UHS is required while in modes 1-4 to ensure that the reactor could be adequately cooled at any time. This led to an analysis to ensure that the Intake Canal was designed and maintained appropriately to ensure that adequate flow could be provided to the RW pumps during a Minimum Tide Hurricane Blowout condition. CR3 is currently in the process of decommissioning and due to this the reactor has been permanently defueled and heat loads remaining in the Spent Fuel Pool are low. This analysis has determined that it is acceptable to remove the Minimum Tide Hurricane Blowout design basis from the FSAR and cease performance of PT-501 which ensured operability of the UHS. It is acceptable to secure the RW pumps for the short duration of any Minimum Tide Hurricane Blowout, and then resume cooling when the level in the canal has returned. The FSAR has already recognized that cooling to the SF system can be shut down for short durations due to the large thermal capacity of the water in the pool. A loss of cooling would allow for the temperature of the pool to increase. F13-0003 determined that after a loss of cooling it would take 23 days for the level in the SF Pool to decrease to within 10 ft. of the Spent Fuel and as long as the fuel was covered by water the cladding would not be challenged. It is unlikely that liquid releases will be made during a Minimum Tide Hurricane Blowout, but if the dilution flow decreases to unacceptable levels the release will be terminated.
Prior NRC approval of this activity is not required.

U. S. Nuclear Regulatory Commission                                                    Attachment 2 3F0516-05                                                                              Page 8of13 ID Number:               AR 706204 Title                    FSAR Change to Eliminate SCBA Requirements Summary and Conclusions Licensing Document Change Request (LDCR) 2014-0056 requests a change to the FSAR, Section 7.4.5. This change is to eliminate the requirement for the operators to don Self-Contained Breathing Apparatus (SCBA), upon detection of a dangerous chemical or toxic gas release. This is also a requirement currently outlined in AP-513, ''Toxic Gas". The requirement for donning the SCBA is to allow the control room operators to continue performing any required actions. Without the SCBA it is possible the control room operators would be impaired.
The control room and control room operators have traditionally been an important part of maintaining the safe operation of the plant including both non-safety-related and safety-related SSCs. Removing the requirement for donning the SCBA during a toxic gas release will certainly increase the likelihood of a control room operator not being able to respond to a transient event or malfunctioning equipment. The safety of the operator is at risk and his ability to respond to any event could be jeopardized.
However, the basic objective of Title 1O of the Code of Federal Regulations is to establish requirements directed toward protecting the health and safety of the public from the uncontrolled release of radioactivity. The protection of public health and safety is ensured through the design of physical barriers to guard against the uncontrolled release of radioactivity. At this point, the CR3 control room operators are not needed to maintain the health and safety of the public. The only potential threat to the health and safety of the public is the spent fuel in the spent fuel pool. Neither a loss of spent fuel cooling or a fuel handling accident requires immediate response from control room operators. There is no other SSC that is important to the safety of the public that needs immediate operator actions. Again, Operator response, or lack of response, does not change the likelihood of a malfunction of any SSC related to spent fuel.
The currently analyzed accidents (fuel handing accident) and transients (seismic, hurricane,
* etc.) are not changed or affected by any operator actions. If an accident were to occur there is sufficient time for the mitigation to occur without depending on main control room operators.
This change can be implemented without NRC approval.

U.S. Nuclear Regulatory Commission                                                      Attachment 2 3F0516-05                                                                               Page 9 of 13 ID Number:              AR 709569 Title                   Installation of Air Cooled Chillers Summary and Conclusions EC 93675 installs a new air cooled chiller on the control complex roof. This chiller will provide chilled water for the control complex ventilation system, and will functionally replace the existing control complex chillers (CHHE-1A/B).
EC 93676 installs two new 50% capacity air cooled chillers on the control complex roof. These chillers will provide cooling water to the spent fuel heat exchangers, and will functionally replace the SW and -RW systems for spent fuel decay heat removal.
The plant has now been shutdown for over 4 years, and the rate of decay heat generation has decreased significantly from that considered in the FSAR for operation and refueling. No freshly irradiated fuel can be added to the pools. The SF cooling system and its supporting cooling water systems have the capability to provide forced flow cooling of the stored irradiated fuel.
However the urgency of restoring the 'intended design function' of cooling the fuel is significantly reduced.
Loss of spent fuel cooling is discussed in FSAR 9.3.1. This section identifies that "pool water thermal storage capacity affords ample time for mitigative steps to be taken if all SF System cooling fails". Calculation F13-0003 "CR-3 Spent Fuel Pool Time to Uncover Fuel Analysis,"
determined that using conservative assumptions which credit no active SSCs functions, the time for the spent fuel pools to heat up and boil inventory off to 10 ft above the fuel storage racks would be 23 days. This calculation assumes no forced cooling and no actions to replace pool inventory during that time.
A malfunction of an SSC in the control complex or spent fuel cooling systems could result in a loss of cooling to the control complex or spent fuel pool. Loss of cooling to the control complex was previously evaluated in FSAR 9. 7.2.1.g. A period of 120 minutes without cooling was used as the limit before exceeding temperature limits for safety related equipment in the control complex. Since no safety related equipment currently exists in the control complex (due to decommissioning and EC 94154), there are no longer any time or temperature limits for this structure.
Since the loss of spent fuel cooling will not result in the uncontrolled release of radioactivity, this function is no longer considered Class I. Further, the spent fuel cooling system has no relevance to reactor operation and therefore does not meet Class II criteria. Therefore, the spent fuel cooling system is considered Class 111.
The requirements for Class Ill design do not include seismic qualification or tornado or tornado missile protection. This indicates that the new system has more vulnerabilities to damage; however, that does not necessarily equate to loss of function. Both earthquakes and tornados are low probability events. If a loss of function does occur, significantly more time is available to restore the design function via repair or alternate and diverse capabilities by adding inventory, and/or feed and bleed pool inventory to maintain pool temperature and level within normal ranges.
The piping systems will be configured such that the control complex chiller can be re-aligned to provide cooling to the spent fuel pool. The will provide some redundancy and allow a spent fuel chiller to be taken out of service for maintenance or repairs.
Both of the new systems will be procured and installed as non-safety related, in accordance with the system classifications established in EC 94154.
This evaluation has determined that prior NRC approval for these changes is not necessary.

U. S. Nuclear Regulatory Commission                                                     Attachment 2 3F0516-05                                                                              Page 10of13 ID Number:              AR 730617 Title                   Electrical System Changes for SAFSTOR 1 Summary and Conclusions This Evaluation is for five related activities that will result in a major reconfiguration of the AC and DC electrical distribution systems at CR-3 to reduce supply capability to match the lower power requirements during the decommissioning process.
OTO/AR 663489 will abandon the 6900V system entirely, the three Step-up Transformers, the Unit Auxiliary Transformer, the Reactor Auxiliary Transformer, the Startup Transformer, the Backup ES Transformer (BEST), and the 4160V BEST Auxiliary Bus 3.
EC 93743 will use MTSW-7 and MTTR-7 to supply power from the 12KV circuit to the 4160V Unit Buses 3A and 3B.
EC93745 will provide a cable crosstie between the A Unit 4160V bus to the B Unit 4160V bus and then crosstie from the B Unit 4160V bus to the B ES 4160V bus.
OTO/AR 650657 will abandon the A and C diesel engine generators and related support systems. It will also abandon the non-1 E station battery C and emergency feed pump battery D and their support equipment. The plant computer panel VBDP-7 will be switched to the alternate source, transformer VBTR-2E/3E, and the E inverter VBIT-1 E will be abandoned.
EC 94614 will change the power supply for DC Distribution Panels 3A, 3B, 4A, and 4B from Station Battery C to Station Batteries A and B. A cross tie circuit will be installed between the A and B Station Batteries main distribution panel using existing conduits and main distribution panel switches to facilitate abandoning Station Battery A in the future. EC 94614 Revision 4 is internally bypassing DPDP-1A Switch 1 using insulated conductors because the switch is not closing correctly. The other end of the DC cross-tie cable will still terminate to DPDP-1 B Switch
: 20. Therefore, the DPDP-1A to DPDP-1 B DC cross-tie will perform the same function and will still be "switchable" using DPDP-1 B Switch 20. Both the screen and the evaluation performed under Reg AR 715019 had indicated that the cross-tie would be switchable at both DPDP-1A and at DPDP-1 B. Although the conclusions of the Regulatory Screen and Evaluation are not affected, both are revised under this Reg AR to clarify that the cross-tie is switchable from DPDP-1 B only.
The only remaining credible accident for CR-3 during decommissioning is a Fuel Handling Accident (FHA). Changes to the electrical distribution system will not affect the reliability of fuel handling equipment or the probability of a human performance event leading to a damaged fuel assembly. The fuel handling equipment will continue to be powered by the distribution system, and the FSAR states that the fuel handling grapple designed so that when loaded with the fuel assembly, the fuel grapple cannot be opened as a result of operator error, electrical or hydraulic failure.
Per the prohibition of 10 CFR 50.82(a)(2) that CR-3 can no longer have fuel in the reactor, all irradiated fuel is stored in the spent fuel pools (SFPs). Therefore, the only design functions that must be preserved relate to providing inventory control and cooling of the SFPs.

U. S. Nuclear Regulatory Commission                                               Attachment 2 3F0516-05                                                                        Page 11 of 13 Calculation F13-0003, 'CR3 Spent Fuel Pool Time to Uncover Fuel Analysis' determined that under the most conservative conditions it will take 22.5 days for the SFPs inventory to be reduced to 10 ft. above the top of the fuel storage racks in the event of a loss of forced pool cooling. This calculation assumes no forced cooling and no actions to replace pool inventory during that time.
Calculation N13-0001, 'Public and Control Room Dose from a Fuel Handling Accident' determined that doses from a FHA to either occupants of the Control Room or receptors. at the exclusion area boundary would be well below any dose limits. No electrical equipment is credited in mitigating the consequences of the FHA.
With the plant in the decommissioning status the design function to automatically repower both ES busses is no longer required. The credible accident and transient scenarios that existed for the plant in power operation during heat-up, reactor operation, and cool down are no longer possible. Therefore, the design functions required to respond to those events are no longer required per the limitations of 10 CFR 50.82(a)(2). Repowering the "B" ES bus will occur automatically, and repowering the "A" ES bus can be accomplished manually within minutes to hours, well within the time available to restore forced SFP cooling to protect the health and safety of the public.
The Evaluation determined that no prior NRC approval is required to implement these activities.

U. S. Nuclear Regulatory Commission                                                      Attachment 2 3F0516-05                                                                               Page 12of13 ID Number:               AR 746109 Title                     Flood Level Change Summary and Conclusions The purpose of this EC is to change the design basis flood height from Elevation 121.4' to Elevation 107' for Class II I SSCs. This will help justify the elimination of required preventative maintenance activities during SAFSTOR and future decommissioning. There are no physical design changes to any structures, systems and components (SSCs).
The scope of this Screen/Evaluation includes changing design basis documents, preventative maintenance requirements, and changes to FSAR chapter 2.4.2 (with associated figures), and any applicable procedures.
The only accident that remains in the FSAR is the Fuel Handling Accident (FHA), which is initiated by dropping or impacting a fuel element during fuel handling activities. The flood level for Class I SSCs and the new proposed flood level for Class Ill SSCs are still much less than the operating deck of the spent fuel building.
The direct connections between the fuel handling equipment and a fuel assembly are the fuel grapples. The fuel grapples are designed such that a fuel assembly cannot be disengaged while the weight of the fuel assembly is suspended from the grapple. Further, loss of power will not cause the fuel grapple to disengage. Numerous interlocks exist for fuel movement and load limits exist to protect the fuel assemblies and fuel storage racks from damage.
It is also not likely that fuel handling activities will occur before, during, or immediately after any hurricane warnings, or pending flooding. Therefore, during those times of maximum anticipated flooding, no fuel handling activities are expected to take place.
Analysis/Calculation S14-0001 is the current analysis of record showing the spent fuel pool liner, spent fuel pool, and spent fuel pool support structure (a subset of the Auxiliary Building) are designed to withstand applicable natural phenomena loading. External flood height is not an applicable load condition. The change in flood design basis will not change the protection features of the spent fuel pool and pool support structure.
EC 94154 already evaluated loss of spent fuel cooling and determined that there are no active mechanical or electrical equipment that are essential to spent fuel pool cooling. Any minor internal flooding from failure of penetration seals, RW piping, etc. in the Auxiliary Building would only affect this mechanical and electrical equipment.
Changing the design basis of the flood height for Class Ill SSCs does not fundamentally change any SSC. All current SSCs that were important or now important to safety have been designed and analyzed with a higher flood elevation. The changes proposed in EC 99162 and to the FSAR, Section 2.4.2 do not change the physical design of any existing SSC. With a lower flood height, there is no flooding at the berm level and no external flood postulated for any SSC important to safety. Also with a lower flood height, no preventative maintenance activities are required to ensure the current flood protection system is functional. Other than the items subject to PM activities, the lower flood height does not change the manner in which any SSCs are operated.
This evaluation determined that prior NRC approval for this change is not necessary.

U. S. Nuclear Regulatory Commission                                                  Attachment 2 3F0516-05                                                                           Page 13of13 ID Number:              AR 1984169 Title                   Abandonment of the Seismic Instrumentation System Summary and Conclusions The activity is to abandon the Seismic Monitoring (SI) system.
10CFR Part 100 Section Vl(a)(3) states: "Required Seismic instrumentation. Suitable instrumentation shall be provided so that the seismic response of nuclear power plant features important to safety can be determined promptly to permit comparison of such response with that used as the design basis. Such a comparison is needed to decide whether the plant can continue to be operated safely and to permit such timely action as may be appropriate."
CR3 is permanently shutdown and no decision about shutting down is required. Further, CR3's Part 50 license no longer permits operation of the reactor or emplacement or storage of nuclear fuel in the reactor vessel. Consequently, no decision is required regarding restart of the reactor.
Further, the remaining safety related SSCs are associated with the Spent Fuel Pool integrity, the fuel storage racks, spent fuel assemblies, fuel transfer tubes and valving. Survival of the spent fuel pool and associated safety related SSCs during a seismic event is essentially a pass-fail event. Either the spent fuel pool leaks or it does not. Regardless, the seismic monitors do nothing to mitigate design basis accidents. They indicate a seismic event only. No other safety or important to safety SSC depend on the seismic monitoring system. In either event, existing procedures already verifies the condition of the pool. The seismic monitors are redundant at this point. The seismic instrumentation no longer serves an important function at CR3. The current location of the instruments does not support Operations in determining if a seismic event has affected the spent fuel pool. There are already procedure requirements to do this. CR3 will not restart or return to service so knowing the recorded levels of the seismic event are not critical the decision making process.
Inspections of safety related and important to safety SSCs is required by not only AP-961 but also AD-EG-ALL-1215, Post Seismic Event Shutdown Inspection & Testing. The AD procedure contains acceptance criteria with scope expansion based on results. It can be performed without knowledge of the earthquake magnitude while the magnitude is being determined.
Consequently, the SI system may be abandoned without adversely impacting the safety of CR3 spent fuel inventory or requiring prior NRC approval for implementation.}}