View source for 3F0520-01, Safety Analysis Report, Quality Assurance Program and 10 CFR 50.59 - 10 CFR 72.48 Report - May 2020

{{Adams
| number = ML20139A004
| issue date = 05/18/2020
| title = Safety Analysis Report, Quality Assurance Program and 10 CFR 50.59 - 10 CFR 72.48 Report - May 2020
| author name = Reising R
| author affiliation = Duke Energy Florida, LLC
| addressee name = 
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000302, 07201035
| license number = DPR-072
| contact person = 
| case reference number = 3F0520-01
| document type = Letter type:, Quality Assurance Program, Report, Technical
| page count = 79
}}

=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. DPR-72 10 CFR 50.71(e) 10 CFR 50.54(a)(3) 10 CFR 50.59(d)(2) 10 CFR 72.48(d)(2)
May 18, 2020 3F0520-01 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001

==Subject:==
Crystal River Unit 3 - Safety Analysis Report, Quality Assurance Program and 10 CFR 50.59 - 10 CFR 72.48 Report - May 2020

==Reference:==
: 1. Crystal River Unit 3 - Safety Analysis Report and 10 CFR 50.59 - 10 CFR 72.48 Report - May 2018 (ADAMS Accession No. ML18155A423)

==Dear Sir:==

In accordance with 10 CFR 50.71(e), Duke Energy Florida, LLC (DEF), hereby submits Revision 6 to the Crystal River Unit 3 (CR-3) Defueled Safety Analysis Report (DSAR) as Attachment 3 and the CR3 Quality Assurance Program Manual (CR3 QAPM) as Attachment 4 to this report.
DSAR text changes are indicated by revision bars on the outside right border of each page.
This revision replaces the previous revision of the DSAR in its entirety. Revision 1 of the DSAR was submitted as part of Reference 1 above.
The DSAR revision includes material which describes the organization, modifications, system abandonments and other changes to CR-3 that have been implemented subsequent to the May 24, 2018 submittal (Reference 1). As required by 10 CFR 50.71(e), a summary of changes made to the DSAR, from Revision 1 to Revision 6, is provided in Attachment 1 (18 changes).
The CR3 QAPM section was removed from the DSAR in Revision 6 and is now contained in a separate document titled CR3 QAPM. The CR3 QAPM is still included, by reference, in the DSAR, as required by 10 CFR 50.54(a)(1). A copy of the CR3 QAPM, which did not reduce the commitments, is attached as required by 50.54(a)(3) and 10 CFR 50.71(e) to be submitted every 24 months after submittal of certifications required by 10 CFR 50.82(a)(1).
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 10 CFR 72.48 for changes made to the plant and the Independent Spent Fuel Storage Installation (ISFSI). contains the 10 CFR 50.59 and 10 CFR 72.48 Report - May 2020, which includes a summary of all evaluations completed this reporting period (April 1, 2018 through April 1, 2020) with the exception of evaluations associated with changes, tests, or experiments that have not been fully implemented. In the event that multiple revisions were performed, the final 10 CFR 50.59 evaluation is being reported due to the cumulative nature of these changes.
No new regulatory commitments are made in this letter.

U. S. Nuclear Regulatory Commission                                                     Page 2 of 2 3F0520-01 If you have any questions regarding this submittal, please contact Mr. Mark Van Sicklen, Nuclear Regulatory Affairs, at (352) 501-3045.
I declare under penalty of perjury that the forgoing is true and correct. Executed on May 18, 2020 Sincerely,
'l!:::::!1liLP,esident Operations Support RR/mvs  DSAR Revision 6 Change Summary Description  10 CFR 50.59 and 72.48 Report - May 2020  DSAR Revision 6  CR3 QAPM xc:      Regional Administrator, Region I NMSS Project Manager

DUKE ENERGY FLORIDA, LLC CRYSTAL RIVER UNIT 3 LICENSE NUMBER DPR-72 DOCKET NUMBER 50-302 / 72-1035 ATTACHMENT 1 DSAR REVISION 6 CHANGE 

==SUMMARY==
DESCRIPTION

U. S. Nuclear Regulatory Commission                                                   Attachment 1 3F0520-01                                                                               Page 1 of 5 DSAR REVISION CHANGE 

==SUMMARY==
DESCRIPTION The Defueled Safety Analysis Report (DSAR) Revision change summary below reflects plant modifications, system abandonment activities, and information and analyses that constitute changes to the DSAR since the publication of Revision 1. Applicable figures and tables were included in these changes but are not individually identified in the summary.
This DSAR revision includes changes made to incorporate the following:
A. Changes incorporated into revision 2
* DSAR Change Package 2018-06: This change revised DSAR Chapter 1, Introduction &
Summary, Chapter 3, Facility Design, and Chapter 6, Conduct of Operations. The changes were made to reflect the abandonment of the B Train Emergency Diesel Generator and its auxiliaries and support systems, and the Off-site Power Transformer.
* DSAR Change Package 2018-09: This change revised DSAR Chapter 1, Introduction &
Summary, and Chapter 3, Facility Design, to reflect the (then) current status of the plant, and to remove text that was formerly retained as historical information. The changes include:
o   Removal of all discussion of accidents o   Update to the sources of electrical power o   Removal of discussion of wet fuel storage o   Removal of discussion of bowed/twisted fuel, re-caging activities, and spent fuel assembly grid straps o   Removal of unnecessary details about the fuel and assembly design o   Removal of discussions related to Class I and Class 1* Systems, Structures, and Components o   Removal of the technical discussion of Wind Load, RB Shell Method of Analysis and design information, and RB liner design information, as well as details of the Seismic Methods of Analysis for other Class 1 structures o   Removal of Section 3.5, Spent Fuel Cooling System and 3.6, Fuel Handling System o   Addition of statements that indicate the Ventilation Systems, Fire Protection System, and electrical Distribution Systems are scheduled for future abandonment o   Removal of discussion of emergency ventilation system operational requirements and flow data for the spent fuel area, and discussion of the Control Complex chiller cross-tie capability for spent fuel cooling o   Removal of discussion of design philosophy and design details (including radiation protection and fire prevention) of the main control room o   Removal of the stand-by diesel as backup power for the UHF radio system o   Removal of discussion of additional capacity for industrial air during cyclic loads in the plant B. Changes incorporated into Revision 3
* DSAR Change Package 2017-02: This change revised DSAR Chapter 3, Facility Design; to reflect implementation of the dormancy phase electrical system (EC 407372).
The dormancy phase electrical system provides electrical facilities to support dormancy phase equipment such as the dormancy ventilation system fans, the dormancy liquid

U. S. Nuclear Regulatory Commission                                                    Attachment 1 3F0520-01                                                                                Page 2 of 5 management equipment, the dormancy RCA access related support equipment, and building inspection / maintenance capabilities. This Engineering Change facilitated transition to the SAFSTOR II, dry dormancy phase of decommissioning.
* DSAR Change Package 2017-04: This change revised DSAR Chapter 3, Facility Design; to remove reference to control room annunciation for Fire Service Water Tank storage volume monitoring. EC 407270 Revision 1 modified the level monitoring to local indication and Operator rounds for Fire Service water. This modification was necessary to facilitate the upcoming transition to dormancy, when power supplies would be de-energized and the control room would become unoccupied.
* DSAR Change Package 2018-07: This change revised DSAR Chapter 1, Introduction &
Summary, Chapter 3, Facility Design, and Chapter 4, Radioactive Waste & Radiation Protection. At the time of the revision, portions of the Radiation Monitoring System remained in service, but were no longer required. Design Criterion 17, Monitoring Radioactive Release (Category B), Design Criterion18, Monitoring Fuel and Waste Storage (Category B), and Design Criterion 70, Control of Releases of Radioactivity to the Environment (Category B), were no longer applicable to CR-3 based on the state of decommissioning at the time. Permanently installed effluent monitors formerly required based on potential criticality accidents and potential activity releases due to normal operation, anticipated abnormal operating conditions, and postulated accidents were no longer required with all spent fuel in dry storage and no potential for a consequential release from CR-3. The DSAR revision changed the text for Criterion 17, Criterion 18, and Criterion 70 to reflect that status, and to indicate that future monitoring of effluent discharge and of radiation inside the facility would implemented using appropriate surveys and samples. The revision also changed the text of Chapter 4, Radioactive Waste and Radiation Protection, to reflect the associated changes to the Design Basis, the System Design/System Description, and the System Evaluation for the Radiation Monitoring System accordingly. Note: Due to the licensing history of CR-3, which predates Appendix A, the design criteria for CR-3 given in Chapter 1 of the DSAR are numbered and worded differently than Appendix A.
* DSAR Change Package 2018-08: This change further revised DSAR Chapter 1, Introduction & Summary, Chapter 3, Facility Design, and Chapter 4, Radioactive Waste
       & Radiation Protection to reflect the fully abandoned status of the Radiation Monitoring System.
* DSAR Change Package 2018-11: This change revised DSAR Chapter 1, Introduction &
Summary, and Chapter 3, Facility Design to reflect the fully abandoned status of the Demineralized Water System and the Cycle Make-up and Water Treatment System.
With no requirements remaining for Demineralized water, the DSAR is updated to remove the discussion and information for this system. Likewise, the Cycle Makeup and Water Treatment System discussion is removed, as its DSAR identified function was to provide a makeup source for the Demineralized Water System.
* DSAR Change Package 2018-13: This change revised DSAR Chapter 1, Introduction &
Summary, and Chapter 6, Conduct of Operations, modifying the QA Plan to reflect the (then) current status of the plant, and to replace PNSC with the Independent Management Assessment organization and an Independent Safety Reviewer. The changes made conform with the conditions in the NRC approved May 1, 2018 Safety Evaluation to Entergy Nuclear Operations, Inc. for Vermont Yankee Nuclear Power

U. S. Nuclear Regulatory Commission                                                    Attachment 1 3F0520-01                                                                                Page 3 of 5 Station (ADAMS ML18099A166), and were made to bring our Quality Assurance program in line with the industry of decommissioning nuclear power plants.
* DSAR Change Package 2018-16: This change revised DSAR Chapter 1, Introduction &
Summary, and Chapter 6, Conduct of Operations, modifying statements which describe the QA commitments to Regulatory Guide 1.8 and ANSI N18.1-1971. These changes aligned the requirements for personnel experience and training requirements with that appropriate for a plant in decommissioning. Based on the status of decommissioning and system abandonment, there are no safety-related SSC, and the only SSC important to safety are ISFSI-related SSC. Consequently, statements describing personnel selection, qualifications, experience, training, and certification were revised accordingly, but retained for the Radiation Protection Manager. With the cessation of power operations, permanent defueling of the reactor and completion of the transfer of nuclear fuel into sealed passive storage containers, there are no required nuclear safety related SSCs. The only remaining activities at CR3 resembling those addressed by Regulatory Guide 1.8 and ANSI N18.1- 1971 are those of Radiation Protection, which while important, are not within the applicability of Appendix B to 10 CFR Part 50. Therefore, this revision is a clarification and not a reduction in a commitment previously approved by the NRC.
C. Changes incorporated into Revision 4
* DSAR Change Package 2019-05: This change revised DSAR Chapter 1, Introduction &
Summary, and Chapter 6, Conduct of Operations, to provide editorial corrections to a previously implemented change that was implemented under change 2018-13. (see above). DSAR Change Package 2018-013 replaced the Plant Nuclear Safety Committee (PNSC) with the Independent Management Assessment organization (IMA) and Independent Safety Reviewer, based upon changes previously approved by the NRC for Vermont Yankee. When making the administrative procedure revisions to match the DSAR, and reviewing the corresponding NRC Safety Evaluation, we identified the need for a clarification when we benchmarked the implementing procedures at VY as part of our implementation of the change. The terms Independent Management Assessment (IMA) and Independent Safety Reviewer (ISR) were clarified in this change (DSAR Change Package 2019-05). This change is considered editorial since the referenced NRC SER for Vermont Yankee Nuclear Power Station (ADAMS ML18099A166) was already adopted.
D. Changes incorporated into Revision 5
* DSAR Change Package 2018-14: This change updated DSAR Chapter 3, Facility Design, to reflect the partially abandoned status of the Fire Service System as it was at the time, and as it would be going into dry dormancy, with all electrical services to the legacy CR-3 Fire System removed. The legacy system is functionally replaced by the newly installed above ground yard mains, and the new standpipes in the turbine building and auxiliary building. All legacy FS system components (fire pumps, fire suppression systems, hose stations, piping, fire detection systems, and miscellaneous support equipment) are abandoned except for approximately 17 components that Interface with FST-1A and water supply piping. (Select legacy equipment is retained available to provide isolation where necessary while maintaining access to stored Fire Service water via FST-1A.). NOTE: ISFSI-related Fire Service is not affected, and not in the scope of the change implemented by DSAR Change Package 2018-14.

U. S. Nuclear Regulatory Commission                                                  Attachment 1 3F0520-01                                                                              Page 4 of 5
* DSAR Change Package 2018-15: This change updated DSAR Chapter 3, Facility Design, and Chapter 4, Radioactive Waste & Radiation Protection based on abandonment of legacy ventilation systems and transition to the Dormancy Ventilation System. Although the remaining functions for ventilation equipment were below the threshold of importance for inclusion in the DSAR it was concluded that, though not technically required, a brief description of the Dormancy Ventilation System in the DSAR might be helpful. This DSAR revision removed text describing legacy ventilation system related information and added Dormancy Ventilation system information.
* DSAR Change Package 2019-01: This change updated DSAR Chapter 3, Facility Design, based on abandonment of the Chilled Water System and the Spent Fuel Cooling System. All of the Chapter 3 text discussing these systems was removed, as the only relevant equipment now remaining in service is those few Chilled Water components located within the Nuclear Administration Building. Those components are not discussed in the DSAR. With relocation of all spent fuel to dry storage and subsequent draining of the spent fuel pool completed, and the abandonment of the control complex completed, these two systems were no longer required to operate in the plant and have been abandoned.
* DSAR Change Package 2019-02: This change revised DSAR Chapter 1, Introduction &
Summary, and Chapter 3, Facility Design to reflect the fully abandoned status of the DC Power System as the plant transitioned to dry dormancy status. With no functions remaining, the DC Power System is completely de-energized and fully abandoned, and the DSAR is updated to remove all discussion of the DC Power System.
* DSAR Change Package 2019-03: This change updated DSAR Chapter 3, Facility Design, based on partial abandonment of the Communications System and the Commercial Telephone System. The partial abandonment reduced the footprint of the systems, removing from service those SSC's that are no longer required in dormancy, while retaining in service only those portions of the systems necessary to support the UHF Radio System (for personnel safety during periodic building entries), required ISFSI-related commercial service dedicated lines (for the ENS Phone and the State Hot Ringdown Phone), and the corporate voice-over-internet phone system (where needed for general business and personal communications purposes). All of the Chapter 3 text discussing these systems was removed.
* DSAR Change Package 2019-06: This change revised DSAR Chapter 1, Introduction &
Summary, Chapter 3, Facility Design, and Chapter 6, Conduct of Operations. The changes were made to update the DSAR based on changes to the Fire Protection Program for dry dormancy.
* DSAR Change Package 2019-07: This change revised DSAR Chapter 1, Introduction &
Summary, Chapter 2, Site & Environment, Chapter 3, Facility Design, Chapter 4, Radioactive Waste & Radiation Protection, Chapter 5, Safety Analysis, and Chapter 6, Conduct of Operations. The changes were made to update the DSAR based on system abandonments, removing several figures, discussions, and details that are no longer relevant or significantly beneficial as the plant entered dry dormancy. RG 1.184 Rev 1, NEI 98-03 (Rev 1 & draft Rev 2), and Humboldt Bay Power Plant Unit 3 Defueled Safety Analysis Report, Revision 12 were used as references for guidance in determining the appropriate level of detail for this revision of the CR-3 DSAR. This revision removes all remaining information related to the legacy plant systems, and adds a brief description of the equipment provided for convenience during dormancy. Chapter 1 introduction and

U. S. Nuclear Regulatory Commission                                                   Attachment 1 3F0520-01                                                                              Page 5 of 5 general information was revised to delete unnecessary historical information, such as the inception of the FSAR, and to delete unnecessary details about ISFSI, and to delete the statement regarding the organization of the DSAR being in conformance with historical AEC guidance. (This revision does not necessarily preserve that conformance, and conformance is not required. RG 1.184 Rev 1 Section 8.2 describes the expectations for maintaining the contents of the FSAR.). The Facility Description information contained in sections 1.2, 1.3, and 1.4 was revised to match the SAFSTOR 2
       - Dormancy state of the plant, with no legacy SSC's mechanically or electrically active, except those few supporting the Dormancy Ventilation System, Dormancy Electrical System (and connected equipment), SAFSTOR 2 Liquid Management Equipment, Dormancy Fire System, and Part 37 Security equipment. Chapter 2 is revised to remove unnecessary meteorological detail, geological detail, and site preparation/foundation bearing material detail that is no longer required. The various accident conditions, and associated plans and programs, and analyses for which the details were important during operation of CR-3 are no longer applicable. As there are no design basis accidents or events which could cause a release with consequences that approach 10 CFR 100 limits, the level of detail previously provided is not required. Chapter 3 is revised to move the Fire Protection Program to Chapter 6, Section 6.5, and to delete all else. A brief description of the equipment provided for convenience during dormancy is provided. (Dormancy ventilation, Dormancy Electrical, SAFSTOR 2 Liquid Management, Dormancy Fire Service, and Part 37 Security equipment). A brief discussion is added describing the layup condition of the reactor vessel and its internals, and the surveillance capsules and Incore detectors that are located within the vessel. Chapter 4 is revised to remove information about legacy installed Waste Disposal equipment and systems, and replace it with information pertinent to waste handling during dormancy, and to remove unnecessary details about radiation shielding that are no longer relevant with CR-3 permanently shut-down and all fuel relocated out of the plant. Chapter 5 is deleted, as there are no applicable accidents and therefore no corresponding Safety Analyses for CR-3. Chapter 6 is revised to place a short description of the current revision of the Fire Protection Program (Rev 36) at the beginning of Section 6.5, followed by the (revised)
Fire Protection Program information from Chapter 3. The Fire Protection Program mark-up moved from Chapter 3 is a mark-up provided under DTO AR 714369, EC 407270 Rev 3, LDCR 2018-0014, REG 10 Screen 2241960, and LDCR 2019-0006, REG 10 Screen 2264782. Relocating the information from Chapter 3 to Chapter 6 is editorial.
Chapter 6 is also revised to remove descriptions of activities or program elements related to Training, Procedures, and/or Records that are no longer required.
E. Changes incorporated into Revision 6 DSAR Change Package 2020-01: This change revised DSAR Chapter 1, Introduction &
Summary, Chapter 2, Site & Environment, Chapter 4, Radioactive Waste & Radiation Protection, and Chapter 5, Conduct of Operations. The changes were made to update the DSAR to reflect implementation of the NRC approved Partial Site Release, and to remove the Quality Assurance Program related text from the DSAR based on relocating the QA Plan to a separate stand-alone Quality Assurance Program Manual. (CR3 QAPM). No content changes were made to the QA Plan, only wording changes to reflect internal section references as opposed to DSAR section references, since the CR3 QAPM was made stand alone.

DUKE ENERGY FLORIDA, LLC CRYSTAL RIVER UNIT 3 LICENSE NUMBER DPR-72 DOCKET NUMBER 50-302 / 72-1035 ATTACHMENT 2 10 CFR 50.59 AND 10 CFR 72.48 REPORT - MAY 2020

U. S. Nuclear Regulatory Commission                               Attachment 2 3F0520-01                                                          Page 1 of 5 10 CFR 50.59 and 72.48 Evaluation Summaries Table of 10 CFR 50.59 Evaluations ID Number            Title AR 2212624           Reg10 50.59 Screen for CP0500 REV. 018 AR 2267032           CR-3 50.59 Screen / Reg Review of EC 414782 AR = Action Request Table of 10 CFR 72.48 Evaluations ID Number            Title AR 2267032           CR-3 50.59 Screen / Reg Review of EC 414782 AR = Action Request

U. S. Nuclear Regulatory Commission                                                     Attachment 2 3F0520-01                                                                                 Page 2 of 5 ID Number:                AR 2212624 Title                     Reg10 50.59 Screen for CP0500 REV. 018 Summary and Conclusions This change modifies language in the DSAR and CP-500 regarding radiation monitors. The radiation monitors will remain in service until removed from service by an abandonment under AI-9003. A separate 50.59 will evaluate abandonment of the radiation monitors. This change also removes the General Design Criteria that governs the requirements for radiation monitors on operating nuclear plants.
DSAR 4.1.1 states, The radiation Monitoring System (RMS) detects, indicates, alarms and records radiation levels and concentrations of radioactivity at selected locations inside of the plant to verify compliance with 10CFR20. Changing how this DSAR -
described design function, compliance with 10CFR20, is considered adverse.
Regulatory requirements important to the formulation of an operating power reactors effluents program are found in 10CFR50.34a and 10CFR36a. 10CFR50, Appendix I, and supporting guidance documents, such as NUREG-17 (operational source term), NUREG-0133 (Preparation of Radiological Effluent Technical Specification for Nuclear Power Plants), and NUREG-1301 (Offsite Dose Calculation Manual Guidance) are used to support implementation of 10CFR50.34a and 10CFR50.36a. A reading of 10CFR50, Appendix I, and some of these key supporting documents indicates they were intended for operating facilities, or to be used during the licensing phase of a proposed new operating facility. CR3 decommissioning status has reduced to the point where there are no design basis accidents in the DSAR due to all fuel being transferred to dry storage. All fuel at CR3 was transferred to a dry storage facility in January 2018. There are no noble gases remaining. The remaining source term continues to decrease and is below a level that presents a consequential release risk. Because of CR3s significantly reduced source term coupled with the fact radiation monitors are intended for operating plants that present a much greater risk of consequential radioactive releases resulting from plant operation than is presented by the current condition of CR3, radiation monitors are not needed to ensure 10CFR20 limits are not exceeded. Likewise, because of the significantly reduced source term and non-operational status of CR3, there is no reduction in operational margin as no potential for a consequential release exists. Hence, no License Amendment is required.

U. S. Nuclear Regulatory Commission                                                    Attachment 2 3F0520-01                                                                                Page 3 of 5 ID Number:               AR 2267032 Title                    CR-3 CFR 50.59 Screen / Reg Review of EC 414782 Summary and Conclusions Problem being addressed by EC 414782:
SEP-4&5 (redundant pumps for the sewage lift station south of the Technical Support Center (TSC)) are required to support SAFSTOR2 (Dormancy) staff in the Nuclear Administration Building (NAB) but the existing power feeder to the SEP-4&5 control panel is from TSC electrical distribution panel ACDP-111 which will be de-energized in Dormancy. Also, the SEP-4&5 control panel is located in the TSC which will not have ventilation, lighting, or active fire protection.
This activity will move the SEP-4&5 control panel outside near the lift station and re-power the control panel from Nuclear Support Operations Center (NSOC) electrical distribution panel ACDP-192. NSOC electrical distribution panel ACDP-192 is being selected as the new power source since there are existing non-asbestos containing cables routed from the room containing ACDP-192 to Manhole TSC-4 in an underground duct bank that can be re-tasked to power the SEP-4&5 control panel. Manhole TSC-4 is adjacent to the sewage lift station.
Activity being reviewed scope:
: 1. Electrically isolate the SEP-4&5 control panel from TSC electrical distribution panel ACDP-111 and from downstream Terminal Box ME-35.
: 2. Move the SEP-4&5 control panel outside near the sewage lift station south of the TSC.
The control panel is rated for outdoor use.
: 3. Disconnect Manhole ASB-1, ASB-2, ASB-3, TSC-3, and TSC-4 sump pump power cables in ACDP-205 (NSOC).
: 4. Re-task Manhole TSC-3 and TSC-4 sump pump power cables to power the SEP-4&5 control panel from ACDP-192 (NSOC; adjacent to ACDP-205).
: 5. Re-connect the SEP-4&5 control panel to Terminal Box ME-35.
: 6. Test the SEP-4&5 control panel.
There are no accidents listed in the DSAR. There is a Radioactive Waste Handling Event.
Communications Cable CTP503 and CTP504 have no credible mechanism for initiating a Radioactive Waste Handling Event, affecting the frequency of occurrence of a Radioactive Waste Handling Event, or affecting the consequences of a Radioactive Waste Handling Event.
Manhole ASB-1, ASB-2, ASB-3, TSC-3, and TSC-4 will not be structurally compromised by removal of power to the sump pump motors.
Due to the construction of the cables (water resistant designs), EC 414782 will not result in more than a minimal increase in the likelihood of occurrence of a malfunction of an SSC important to the IOEP previously evaluated in the DSAR. Further, a complete loss of communications is not credible due to the frequent use or monthly testing and multiple communications pathways available per the IOEP.
There are no consequences from Cable CTP503 or Cable CTP504 failing during normal operations at CR3. There are no consequences from Cable CTP503 or Cable CTP504 failing during the Radioactive Waste Handling Event since DSAR Section 4.2.3.2 does not credit any type of emergency communications with limiting the consequences of the event. Cable CTP503 and CTP504 are low energy communications circuits and have no credible mechanism for creating any type of accident. Failure of Cable CTP503 and CTP504 can only lead to the loss of the related communications functions.
No license amendment is required to implement EC 414782.

U. S. Nuclear Regulatory Commission                                                    Attachment 2 3F0520-01                                                                                Page 4 of 5 ID Number:               AR 2267032 Title                    CR-3 CFR 72.48 Screen / Reg Review of EC 414782 Summary and Conclusions NOTE: The activity being reviewed (implementing document) is an engineering change to the CR3 plant. The portion of the activity being screened that is relevant to ISFSI is the sparing of Manhole ASB-1, ASB-2, ASB-3, TSC-3, and TSC-4 sump pump power cables. There are ISFSI communications cables (CTP503 and CTP504) between NSOC and the ISFSI Security Operations Center (SOC) in Manholes ASB-1, ASB-2, ASB-3, and TSC-4 that could potentially become submerged in water in the manholes after power is removed from the sump pump motors.
Problem being addressed by EC 414782:
SEP-4&5 (redundant pumps for the sewage lift station south of the Technical Support Center (TSC)) are required to support SAFSTOR2 (Dormancy) staff in the Nuclear Administration Building (NAB) but the existing power feeder to the SEP-4&5 control panel is from TSC electrical distribution panel ACDP-111 which will be de-energized in Dormancy. Also, the SEP-4&5 control panel is located in the TSC which will not have ventilation, lighting, or active fire protection.
This activity will move the SEP-4&5 control panel outside near the lift station and re-power the control panel from Nuclear Support Operations Center (NSOC) electrical distribution panel ACDP-192. NSOC electrical distribution panel ACDP-192 is being selected as the new power source since there are existing non-asbestos containing cables routed from the room containing ACDP-192 to Manhole TSC-4 in an underground duct bank that can be re-tasked to power the SEP-4&5 control panel. Manhole TSC-4 is adjacent to the sewage lift station.
Activity being reviewed scope:
: 1. Electrically isolate the SEP-4&5 control panel from TSC electrical distribution panel ACDP-111 and from downstream Terminal Box ME-35.
: 2. Move the SEP-4&5 control panel outside near the sewage lift station south of the TSC.
The control panel is rated for outdoor use.
: 3. Disconnect Manhole ASB-1, ASB-2, ASB-3, TSC-3, and TSC-4 sump pump power cables in ACDP-205 (NSOC).
: 4. Re-task Manhole TSC-3 and TSC-4 sump pump power cables to power the SEP-4&5 control panel from ACDP-192 (NSOC; adjacent to ACDP-205).
: 5. Re-connect the SEP-4&5 control panel to Terminal Box ME-35.
: 6. Test the SEP-4&5 control panel.
There are no accidents listed in the DSAR. There is a Radioactive Waste Handling Event.
Communications Cable CTP503 and CTP504 have no credible mechanism for initiating a Radioactive Waste Handling Event, affecting the frequency of occurrence of a Radioactive Waste Handling Event, or affecting the consequences of a Radioactive Waste Handling Event.
Manhole ASB-1, ASB-2, ASB-3, TSC-3, and TSC-4 will not be structurally compromised by removal of power to the sump pump motors.
Due to the construction of the cables (water resistant designs), EC 414782 will not result in more than a minimal increase in the likelihood of occurrence of a malfunction of an SSC important to the IOEP previously evaluated in the DSAR. Further, a complete loss of communications is not credible due to the frequent use or monthly testing and multiple communications pathways available per the IOEP.

U. S. Nuclear Regulatory Commission                                            Attachment 2 3F0520-01                                                                        Page 5 of 5 There are no consequences from Cable CTP503 or Cable CTP504 failing during normal operations at CR3. There are no consequences from Cable CTP503 or Cable CTP504 failing during the Radioactive Waste Handling Event since DSAR Section 4.2.3.2 does not credit any type of emergency communications with limiting the consequences of the event. Cable CTP503 and CTP504 are low energy communications circuits and have no credible mechanism for creating any type of accident. Failure of Cable CTP503 and CTP504 can only lead to the loss of the related communications functions.
No license amendment is required to implement EC 414782.

DUKE ENERGY FLORIDA, LLC CRYSTAL RIVER UNIT 3 LICENSE NUMBER DPR-72 DOCKET NUMBER 50-302 / 72-1035 ATTACHMENT 3 DSAR REVISION 6

Defueled Safety Analysis Report
* Duke Energy Florida
* Crystal River Unit 3 This is Revision 6 of the Living DSAR.
All changes incorporated within are fully approved and D
this version should be utilized for Quality Activities.
S A
R Revision 6

Revision:         6 DEFUELED SAFETY ANALYSIS REPORT                                      LOEP LIST OF EFFECTIVE PAGES                      Page:       1 of 1 DSAR Title                                    Total Pages Revision Chapter 1 - Introduction & Summary                                                  4        6 Tables:
Table 1-1, Crystal River Unit 3 In-Plant Quality Program Functions - Deleted Rev. 6 Table 1-2, Crystal River Unit 3 Quality Program Commitments - Deleted Rev. 6 Figures:
Figure 1-1, Drawing CR3-G86-D, Sheet 1 of 1, Drawing Revision 15 - Deleted Rev. 6 Figure 1-2, Nuclear Generation Department Organization - Deleted Rev. 6 DSAR Title                                    Total Pages Revision Chapter 2 - Site & Environment                                                      8        6 No Tables Figures:
Figure 2-1, General Area Map Figure 2-2, CR3 Controlled Area Figure 2-3, Site Topography Within a 5 Mile Radius - Deleted Rev. 6 Figure 2-4, Topography Plan and Local Aerial Photograph - Deleted Rev. 6 DSAR Title                                    Total Pages Revision Chapter 3 - Auxiliary Systems                                                       2        5 No Tables or Figures DSAR Title                                    Total Pages Revision Chapter 4 - Radioactive Waste & Radiation Protection                                5        6 Tables:
Table 4-1, Assumptions for Radioactive Waste Handling Event No Figures DSAR Title                                    Total Pages Revision Chapter 5 - Conduct of Operations                                                  12        6 No Tables or Figures END - OF - LOEP

Revision:                    006 DEFUELED SAFETY ANALYSIS REPORT                                                     Chapter:                         1 INTRODUCTION & 

==SUMMARY==
Page:                    1 of 4 TABLE OF CONTENTS
: 1. INTRODUCTION AND 

==SUMMARY==
 ................................................................................................................ 2

==1.1      INTRODUCTION==
.................................................................................................................................... 2 1.1.1   GENERAL INFORMATION ........................................................................................................... 2 1.1.2   CRYSTAL RIVER UNIT 3 DSAR REVISIONS (SUBMITTED TO NRC) ............................................ 2 1.2      

==SUMMARY==
PLANT DESCRIPTION ..................................................................................................... 3 1.2.1   SITE CHARACTERISTICS ............................................................................................................. 3 1.2.2   ISFSI CHARACTERISTICS ............................................................................................................ 3 1.3      QUALITY PROGRAM ........................................................................................................................... 3 1.4      INTERACTIONS BETWEEN CRYSTAL RIVER UNIT 3 AND THE FOUR FOSSIL FIRED PLANTS ........................................................................................................................ 3 1.4.1   WELL WATER SYSTEM ................................................................................................................ 3 1.4.2   INTAKE AND DISCHARGE CANALS .......................................................................................... 3 1.4.3   FIRE SYSTEM MAKEUP ................................................................................................................ 3

==1.5      CONCLUSION==
S ...................................................................................................................................... 3

==1.6      REFERENCES==
 ......................................................................................................................................... 4

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==1.1     INTRODUCTION==

1.1.1        GENERAL INFORMATION As documented in Nuclear Regulatory Commission (NRC) to Crystal River Nuclear Plant (CR3) {{letter dated|date=March 13, 2013|text=letter dated March 13, 2013}} (ADAMS Accession No. ML13058A380), the NRC has acknowledged CR3s certification of permanent cessation of power operation and permanent removal of fuel from the reactor vessel.
All spent nuclear fuel is in dry storage in the ISFSI and all new fuel has been shipped offsite to a new owner. As such, no SSCs at CR3 meet the definition of safety related in accordance with 10 CFR 50.2.
The DSAR was developed as the principal licensing source document describing pertinent equipment, structures, systems, operational constraints and practices, accident analyses, and decommissioning activities associated with the defueled status of the Crystal River Unit 3 Nuclear Generating Plant. As such, the DSAR is intended to serve in the same role as the Final Safety Analysis Report (FSAR) during the period of power operation.
Construction of Crystal River Unit 3 (CR3) was authorized by the AEC through issue of provisional construction permit CPPR-51 on September 25, 1968 in Docket 50-302. Construction of CR3 was completed and the operating license issued December 3, 1976. Fuel was loaded in 1976. CR3 last produced power in September 2009, while shutting down for Refuel 16. During activities to replace steam generators, a portion of the containment concrete wall delaminated. While completing repairs additional delamination occurred. CR3 was officially retired on February 5, 2013.
The CR3 Nuclear Steam Supply System (NSSS) is a pressurized water reactor type. It used chemical shim for reactivity control and generated steam with a small amount of superheat in Once-Through Steam Generators (OTSG). The NSSS and nuclear fuel were supplied by Babcock & Wilcox Company (B&W).
CR3 has an Independent Spent Fuel Storage Installation (ISFSI) located on the east berm of the plant.
The ISFSI has the capacity for 40 Dry Shielded Canisters (DSCs), each holding up to 32 fuel assemblies. The ISFSI consists of the NUHOMS Reinforced Concrete Horizontal Storage Modules, each containing one 32PTH1-TYPE 2W DSC, manufactured for CR3 by Areva TransNuclear Corporation, under Certificate of Compliance 1004, Amendment Number 14. The ISFSI includes one on-site storage container (OSSC) which contains greater than class C waste. The 10 CFR 72.212 Report that documents how the site meets Part 72 has been issued as procedure ISFS-0212.
On January 22, 2019, CR3 submitted a Partial Site Release request with the NRC to reduce the licensed footprint by releasing 3854 acres of the non-impacted areas from the 4738-acre site per 10 CFR 50.83, Release of Part of a Power Reactor Facility or Site for Unrestricted Use (ADAMS Accession No. ML19022A076). As documented in NRC to CR3 {{letter dated|date=January 2, 2020|text=letter dated January 2, 2020}} (ADAMS Accession No. ML19339G509), the NRC approved the release of the non-impacted areas. The new 884-acre Site, also referred to as the Controlled Area and defined by the new Site Boundary, is reflected in DSAR, Revision 6.
1.1.2        CRYSTAL RIVER UNIT 3 DSAR REVISIONS (SUBMITTED TO NRC)
Revision 0, Submitted August 31, 2017 Revision 1, Submitted May 24, 2018

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==SUMMARY==
PLANT DESCRIPTION 1.2.1        SITE CHARACTERISTICS The 884-acre site is characterized by isolation from nearby population centers; sound foundation for structures; and favorable conditions of hydrology, geology, seismology, and meteorology.
1.2.2        ISFSI CHARACTERISTICS CR3 has 1,243 spent nuclear fuel assemblies stored at the site. The fuel assemblies are stored in dry storage on an Independent Spent Fuel Storage Installation (ISFSI) pad. The ISFSI is described in Section 1.1.1. In addition to the spent fuel assemblies, CR3 has one failed fuel basket stored within the ISFSI in a dry shielded canister with spent fuel, and one OSSC which contains greater than class C waste.
1.3     QUALITY PROGRAM The Quality Program is located in a document entitled Crystal River Quality Assurance Program Manual (QAPM).
1.4     INTERACTIONS BETWEEN CRYSTAL RIVER UNIT 3 AND THE FOUR FOSSIL FIRED PLANTS AI-1300, Engineering, Maintenance and Support Interfaces, is a CR3 document which contains descriptions of the numerous interactions between CR3 and other DEF organizations. It also defines the scope of the interfacing activities. The document is for use by organizations who perform activities which may affect the licensing/design basis of CR3 to identify those activities requiring the knowledge and participation of Nuclear Operations. A brief discussion of some of the interfaces follows:
1.4.1        WELL WATER SYSTEM Well water to Units 1, 2, and 3 is furnished from a common system. Units 4 and 5 are on separate wells. The maintenance and operation of the Units 1, 2, and 3 system is under the supervision and direction of the Fossil Plant Superintendent.
1.4.2        INTAKE AND DISCHARGE CANALS The intake and discharge canals are common between Units 1 and 2 and the nuclear unit. Maintenance of the canals is the responsibility of the Crystal River Fossil Operations.
1.4.3        FIRE SYSTEM MAKEUP Makeup to the 300,000 gallon fire service water storage tank is available from the Fossil fire protection system.

==1.5     CONCLUSION==
S On the basis of the information presented in this Defueled Safety Analysis Report and referenced material, DEF concludes that Crystal River Unit 3 was designed, constructed, and is being operated without undue risk to the health and safety of the public.

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==1.6   REFERENCES==
: 1. G/C, Inc., Structural Calculation FC 16.00.1/1A, Duke Calculation S02-0016, Revision 0, Attachment 1.
: 2. Daniel, R. C., et al., Effects of High Burnup on Zircaloy-Clad, Bulk UO2, Plate Fuel Element Samples, WAPD-263, September, 1962.
: 3. Crystal River Quality Assurance Program Manual (QAPM).

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: 2. SITE AND ENVIRONMENT.............................................................................................................................. 2 2.1      

==SUMMARY==
 ............................................................................................................................................. 2 2.2      SITE AND ADJACENT AREAS ............................................................................................................. 2 2.2.1    SITE LOCATION AND TOPOGRAPHY ........................................................................................ 2 2.2.2    SITE OWNERSHIP .......................................................................................................................... 2 2.2.3    SITE ACTIVITY .............................................................................................................................. 4 2.2.4    MAJOR TRANSPORTATION ROUTES, WATERWAYS, AND AIRPORTS .............................. 4 2.3      METEOROLOGY ................................................................................................................................... 5 2.3.1    METEOROLOGICAL INPUT ......................................................................................................... 5 2.4      HYDROLOGY ........................................................................................................................................ 5 2.4.1    CHARACTERISTICS OF STREAMS IN VICINITY OF THE SITE .............................................. 5 2.4.2    MAXIMUM HURRICANE SURGE LEVEL ................................................................................... 5

==2.5      REFERENCES==
 ......................................................................................................................................... 6 FIGURE 2-1, GENERAL AREA MAP............................................................................................................................ 7 FIGURE 2-2, CR3 CONTROLLED AREA ..................................................................................................................... 8

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==SUMMARY==

The site is located on the Gulf of Mexico, 70 miles north of Tampa, Florida as shown in Figure 2-1.
Dilution flow for liquid releases is drawn from and returned to the Gulf of Mexico using present intake and discharge canal.
The Nuclear Regulatory Commission approved and issued a Partial Site Release on January 2, 2020 reducing the originally licensed 4738-acre site, known as the Owner Controlled Area, to an 884-acre site, referred to now as the Controlled Area (Reference 2 and Reference 3). Figure 2-2 shows the historical Owner Controlled Area and the NRC approved Controlled Area.
The site region is predominantly agricultural in nature.
2.2       SITE AND ADJACENT AREAS 2.2.1          SITE LOCATION AND TOPOGRAPHY The property wholly owned and controlled by Duke Energy Florida (herein referred to as the "site") is located in the northwestern portion of Citrus County, State of Florida, and lies either wholly or partly in Sections 28, 29, 31, 32, and 33, Township 17S, Range 16E.
This site location is approximately 71/2 miles NW of Crystal River, and 70 miles N of Tampa as shown in Figure 2-1.
Crystal River Unit 3 is located at latitude 28&deg; 57' 25.87"N and longitude 82&deg; 41' 55.95" W.
Situated between the mouths of the Withlacoochee and Crystal Rivers, the site is primarily an industrial site.
Topographic features and an aerial photograph of the site is shown in Figure 2-2.
2.2.2          SITE OWNERSHIP The Crystal River site consists of 884 acres owned and controlled by Duke Energy Florida. This Controlled Area, its boundaries and the immediate adjacent areas to the plant site, is indicated in Figure 2-2.
Legal Description of the Controlled Area:
A parcel of land lying and being in Sections 28, 29, 31, 32, and 33, Township 17 South, Range 16 East, Citrus County, Florida, being more particularly described as follows:
COMMENCE at a point marking the Southwest corner of the Southwest 1/4 of Section 28, Township 17 South, Range 16 East, Citrus County, Florida; thence coincident with the West boundary of the Southwest 1/4 of said Section 28, N 00&deg;39'12" W a distance of 2647.62 feet to the POINT OF BEGINNING; thence deporting said West boundary, S 89&deg;28'02" E a distance of 5264.62 feet; thence S 00&deg;36'50" E a distance of 2883.83 feet; thence S 59&deg;19'37" W a distance of 72.41 feet; thence S 09&deg;15'51" W a distance of 296.31 feet; thence S 86&deg;56'55" E a distance of 69.64 feet; thence S 00&deg;11'38" W a distance of 63.17 feet to a point coincident with a non-tangent curve, concave Westerly, said curve having a radius of 173.55 feet, a delta angle of 11&deg;58'02", and being subtended by a chord bearing of S 05&deg;29'52" E for a distance of 36.18 feet; thence coincident with the arc of said curve for a

Revision:          006 DEFUELED SAFETY ANALYSIS REPORT                                 Chapter:             2 SITE & ENVIRONMENT                                Page:           3 of 8 distance of 36.25 feet; thence S 00&deg;29'09" W a distance of 44.95 feet to a point coincident with a tangent curve, concave Westerly, said curve having a radius of 126.61 feet, a delta angle of 30&deg;43'23", and being subtended by a chord bearing of S 15&deg;50'51" W for a distance of 67.08 feet; thence coincident with the arc of said curve for a distance of 67.89 feet; thence S 31&deg;12'32" W a distance of 39.80 feet to a point coincident with a tangent curve, concave Easterly, said curve having a radius of 154.80 feet, a delta angle of 26&deg;39'08", and being subtended by a chord bearing of S 17&deg;52'58" W for a distance of 71.36 feet; thence coincident with the arc of said curve for a distance of 72.01 feet; thence S 04&deg;33'24" W a distance of 420.60 feet to a point coincident with a line being 60.00 feet South of and parallel with the main line of a set of railroad tracks; thence coincident with said parallel line, N 89&deg;43'29" W a distance of 1450.27 feet to a point marking the intersection of a line being 60.00 feet South of and parallel with the main line of said railroad tracks with a line being 60.00 feet Southerly of and parallel with a spur line of said railroad tracks, said point also being a point coincident with a tangent curve, concave Southerly, said curve having a radius of 499.40 feet, a delta angle of 45&deg;02'15", and being subtended by a chord bearing of S 67&deg;45'24" W for a distance of 382.52 feet; thence deporting said parallel line with the main line of said railroad tracks, coincident with said line being 60.00 feet Southerly of and parallel with the spur line of said railroad tracks, also being coincident with the arc of said curve for a distance of 392.55 feet; thence continue coincident with said parallel line being 60.00 feet Southerly of said spur line for the following two (2) courses: 1) S 45&deg;14'17" W a distance of 690.75 feet to a point coincident with a tangent curve, concave Southeasterly, said curve having a radius of 1366.79 feet, a delta angle of 10&deg;55'22", and being subtended by a chord bearing of S 39&deg;46'36" W for a distance of 260.17 feet; 2) thence coincident with the arc of said curve for a distance of 260.56 feet; thence deporting said parallel line, S 34&deg;23'36" E a distance of 10.70 feet; thence S 19&deg;56'53" W a distance of 215.96 feet; thence S 73&deg;48'31" W a distance of 31.96 feet to a point coincident with the aforesaid line being 60.00 feet Southerly of and parallel with the spur line of said railroad tracks; thence coincident with said parallel line for the following two (2) courses: 1) S 15&deg;18'52" W a distance of 747.55 feet to a point coincident with a tangent curve, concave Northerly, said curve having a radius of 632.86 feet, a delta angle of 118&deg;31'38", and being subtended by a chord bearing of S 74&deg;34'42" W for a distance of 1087.92 feet; 2) thence coincident with the arc of said curve for a distance of 1309.19 feet; thence departing said parallel line, N 46&deg;09'29" W a distance of 103.99 feet; thence S 74&deg;47'29" W a distance of 1320.58 feet; thence S 84&deg;09'04" W a distance of 183.83 feet; thence S 89&deg;04'50" W a distance of 351.67 feet; thence N 41&deg;30'58" W a distance of 188.25 feet; thence N 71&deg;52'09" W a distance of 383.16 feet; thence N 06&deg;43'16" W a distance of 373.76 feet; thence N 63&deg;03'04" W a distance of 134.93 feet; thence N 01&deg;53'57" E a distance of 93.81 feet; thence N 15&deg;16'27" E a distance of 22.75 feet; thence N 80&deg;12'56" W a distance of 61.09 feet; thence N 42&deg;00'11" W a distance of 182.57 feet; thence N 42&deg;03'09" W a distance of 109.07 feet; thence N 42&deg;07'09" W a distance of 109.06 feet; thence N 42&deg;57'43" W a distance of 39.62 feet; thence N 47&deg;46'52" W a distance of 39.18 feet; thence N 52&deg;54'06" W a distance of 39.57 feet; thence N 57&deg;45'59" W a distance of 39.14 feet; thence N 69&deg;59'27" W a distance of 20.92 feet; thence S 77&deg;09'18" W a distance of 5145.53 feet; thence N 05&deg;04'29" W a distance of 430.25 feet; thence N 77&deg;33'22" E a distance of 3944.20 feet; thence N 19&deg;57'31" W a distance of 220.80 feet; thence N 19&deg;03'26" W a distance of 939.21 feet; thence N 02&deg;07'23" E a distance of 35.66 feet; thence N 88&deg;16'26" W a distance of 3639.80 feet; thence N 00&deg;00'00" W a distance of 345.49 feet; thence S 88&deg;28'53" E a distance of 1461.51 feet; thence S 85&deg;20'23" E a distance of 1461.50 feet; thence N 02&deg;45'30" E a distance of 45.13 feet; thence N 40&deg;00'37" E a distance of 132.68 feet; thence S 86&deg;17'04" E a distance of 245.38 feet; thence S 86&deg;02'02" E a distance of 607.43 feet; thence N 52&deg;55'57" E a distance of 45.67 feet; thence S 86&deg;01'46" E a distance of 122.99 feet; thence S 42&deg;53'12" E a distance of 46.23 feet; thence S 85&deg;57'35" E a distance of 230.08 feet; thence N 53&deg;42'27" E a distance of 109.55 feet; thence N 80&deg;53'44" E a distance of 41.36 feet; thence N 80&deg;30'02" E a distance of 50.82 feet; thence N 82&deg;13'46" E a distance of 30.90 feet; thence N 84&deg;47'47" E a distance of 27.24 feet; thence N 88&deg;43'56" E a distance of 39.40 feet; thence S 88&deg;37'22" E a distance of 68.16 feet; thence S 86&deg;58'29" E a distance of 78.13 feet; thence S 85&deg;58'08" E a distance of 86.72 feet; thence N 60&deg;01'50" E a distance of 23.49 feet; thence N 87&deg;48'16" E a distance of 85.49 feet; thence S 86&deg;54'36" E a distance of 65.09 feet; thence N 00&deg;59'12" E a distance of 735.00 feet; thence S 89&deg;51'16" E a distance of 1741.39 feet; thence N 01&deg;19'56" W a distance of 1775.85 feet; thence S 89&deg;55'19" W a distance of 303.10 feet; thence N 00&deg;19'41" E a distance of 917.60 feet; thence S 89&deg;28'02" E a distance of 490.19 feet to the POINT OF BEGINNING.

Revision:          006 DEFUELED SAFETY ANALYSIS REPORT                                Chapter:              2 SITE & ENVIRONMENT                                Page:           4 of 8 Containing an area of 38,489,493.36 square feet, 883.597 acres, more or less.
There are no public access roads to areas adjacent to the plant site except at the plant access road. Approximately four miles east of the plant, a dirt road crosses the site access road. The north and south site boundaries are bordered by woods and swamps and are generally inaccessible. The Crystal River is located due south of the site and is used for commercial fishing and pleasure craft. Directly west of the plant is the Gulf of Mexico, from which the Crystal River plant site historically received its condenser cooling water. Fishing and pleasure craft have unrestricted access to the Gulf waters. Company property extends to the Gulf of Mexico. Duke Energy Florida has no legal rights to any appurtenant structures which extend into the Gulf beyond the bulkhead line described previously. Small craft are prevented from entering the discharge canal by a blockade at the bulkhead line. This blockade was installed for safety concerns due to increased water turbulence caused by the mixing of reintroduced water to the canal from the helper cooling towers.
2.2.3          SITE ACTIVITY Duke Energy Florida has six fossil fuel generating units (Unit 1, Unit 2, Unit 4, Unit 5, and two Combined Cycle units) at the plant site.
The presence of industry, transportation, or operations in the vicinity of the site does not pose any potentially significant effects on the safe operation of the nuclear facility. Since CR3 has permanently shutdown and removed spent fuel from the reactor, the risks of industry, transportation, or operations in the vicinity of CR3 are significantly reduced.
2.2.4          MAJOR TRANSPORTATION ROUTES, WATERWAYS, AND AIRPORTS The only major road is U.S. Highway 19, a four-lane divided highway through Citrus County. U.S. Highway 19 links St. Petersburg with Tallahassee. A railroad spur off the Seaboard Coast Line Railroad serves the DEF site.
Boat landings are few, small, and scattered, although the Intercoastal Waterway passes within 10 miles of the site.
A canal has been constructed between the Gulf and the DEF plant site for delivery of coal by barges and intake of cooling water.
No new major highways are expected to pass through the area. However, there will be an increase in local roads as new subdivisions and mobile home courts are constructed in the southern part of the area.
The major waterways are:
: a. Crystal River Entrance Channel
: b. Cross Florida Barge Canal (only a western section has been constructed)
Presently, there is only one airfield and no known missile bases in the area. The airfield is located in a southeasterly direction about eight miles from the plant site. At present, there are no known plans to rebuild any airports within the five mile radius of the plant.
The present airfield has one turf runway, 2,666 feet in length, oriented in a north-south direction and one paved runway, 4,557 feet in length, oriented in an east-west direction. This field has runway lights, a rotating beacon, and a lighted wind-sox for night landings. The east-west runway is the one used by most small craft landing and taking off.

Revision:          006 DEFUELED SAFETY ANALYSIS REPORT                              Chapter:              2 SITE & ENVIRONMENT                            Page:            5 of 8 2.3       METEOROLOGY 2.3.1         METEOROLOGICAL INPUT There is no credible accident that can result in a radiological release exceeding EPA Protective Action Guidelines (References 6, 7 and 8). Therefore, there is no need to maintain the ability to do real time off site dose assessment and to collect real time meteorological data. Nevertheless, if a need arises, current meteorological data can be obtained from local and national weather services.
Updated annual average atmospheric dispersion and deposition factors (X/Q and D/Q) were calculated for Crystal River Unit 3 using five years of data (Reference 4) to supplement 1976 calculations, which were based on one year of data (Reference 5). These results provide a range of values to account for multiple release points, the controlled area boundary (References 2 and 3), nearby residences, and nearby work locations in the unrestricted area immediately surrounding the controlled area.
Short term atmospheric dispersion factors (X/Q) have also been calculated using the same 5-year dataset (Reference 4) used for the annual average factors. These factors were developed to aid in the assessment of short term abnormal radioactive releases. A range of values have been calculated to account for different release locations with respect to the controlled area boundary.
2.4       HYDROLOGY 2.4.1         CHARACTERISTICS OF STREAMS IN VICINITY OF THE SITE The major streams in the general vicinity of the site are the Withlacoochee River and the Crystal River.
Withlacoochee River is the major stream, having a drainage area at its entrance into the Gulf of Mexico of approximately 2,000 square miles. The discharge of the Withlacoochee due to rain runoff is augmented by a base flow of groundwater runoff and artesian spring discharges. Crystal River is much smaller than Withlacoochee River, with its major discharge consisting of artesian spring discharges.
The plant site is located approximately 3.8 miles south of the mouth of the Withlacoochee and about the same distance north of the mouth of the Crystal River. The Cross-Florida Barge Canal which intersects with the Withlacoochee River inland meets the Gulf about one mile southeast of the mouth of the Withlacoochee River and two miles northwest of the site. The average flow from the Withlacoochee drainage basin, a portion of which enters the Gulf via the Cross-Florida Barge Canal, is approximately 1,820 cfs, with a maximum and minimum flow of 9,130 cfs and 830 cfs, respectively. The average flow of the Crystal River is approximately 600 cfs. The natural stream flows in the vicinity of the plant site have a high mineral content.
2.4.2         MAXIMUM HURRICANE SURGE LEVEL For Class III SSCs, the flood level is shown on FEMA's Flood Insurance Rate Maps. For CR3, the maximum flood height is EL 107' (plant datum), as evaluated in EC 299162.

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==2.5   REFERENCES==
: 1. EC 299162, Revise Flood Design Basis
: 2. 3F0119-01, CR3 to NRC Partial Site Release submittal, including site boundaries and surveys
: 3. 3N20-001, NRC to CR3 Partial Site Release approval
: 4. Calculation N18-0003, Calculation of X/Q & D/Q for SAFSTOR
: 5. NUS Corporation Report: NUS-1753, Crystal River Unit 3 Input to Revise FSAR Incorporating Meteorological Data, January 1, 1975 - December 31, 1975
: 6. ISFSI Only Emergency Plan
: 7. License Amendment Request #322, Revision 0, Independent Spent Fuel Storage Installation (ISFSI)-Only Emergency Plan, and ISFSI-Only Emergency Action Level Bases Manual, for the CR-3 SAFSTOR Period with Spent Fuel on Site (May 26, 2016)
: 8. Safety Evaluation Input For The Crystal River Unit 3 Independent Spent Fuel Storage Installation Only Emergency Plan (CAC NO. L53129) (August 12, 2016)

Revision:   006 DEFUELED SAFETY ANALYSIS REPORT         Chapter:       2 SITE & ENVIRONMENT    Page:     7 of 8 FIGURE 2-1, GENERAL AREA MAP

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: 3. AUXILIARY SYSTEMS ..................................................................................................................................... 2

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: 3.       AUXILIARY SYSTEMS The plant systems have been drained, vented, and de-powered, and are in their SAFSTOR Dormancy (Cold and Dark) condition, awaiting the decontamination and demolition phase of decommissioning. The Reactor Building, Auxiliary Building, Intermediate Building, Control Complex, and Turbine Building are unoccupied, and will remain unoccupied in Dormancy, except for periodic inspections. The Reactor Vessel was flooded up to the core flood nozzles with the closure head in place, and with surveillance capsules and Incore Detectors contained within the vessel. 10 CFR Part 37 required access controls and monitoring are provided by the Reactor vessel and its closure head, and by secure controlled and monitored access through a single Reactor building access point.
Limited equipment has been installed/provided to facilitate building entries for inspections and associated activities when necessary. The equipment includes exhaust fans installed above the seawater room (Dormancy Ventilation),
an arrangement of sump pumps, filters, and holding tanks (SAFSTOR 2 Liquid Management), and a small on-demand AC electrical lighting and power distribution arrangement (Dormancy Electrical). The fans provide a moderate induced draft through the Reactor Building, Auxiliary Building, Intermediate Building and Seawater Room to minimize mold growth and stagnant air conditions in locations where periodic building entries/inspections will be required. The sump pumps, filters, and tanks provide the capability to manage any influx of water into the various buildings and sumps, including sampling and subsequent releases. The pressurized fire water system has been isolated from the unoccupied buildings, and a dry standpipe arrangement has been installed to facilitate providing water to the interior of the unoccupied buildings if it were to become necessary. A gravity fed ring-header is provided on the top of the berm, connected to one of the two legacy fire service water storage tanks. See Chapter 5, Section 5.5 for more details about the Fire Protection Program.
END-OF-CHAPTER

DEFUELED SAFETY ANALYSIS                                                               Revision:                     006 REPORT                                                     Chapter:                           4 RADIOACTIVE WASTE &                                                          Page:                     1 of 5 RADIATION PROTECTION TABLE OF CONTENTS
: 4. RADIOACTIVE WASTE & RADIATION PROTECTION ............................................................................ 2 4.1      RADIOACTIVE WASTE ........................................................................................................................ 2 4.1.1     LIQUID WASTE .............................................................................................................................. 2 4.1.2     SOLID WASTE ................................................................................................................................ 2 4.2      RADIATION PROTECTION .................................................................................................................. 2 4.2.1     OFF-SITE DOSE CALCULATION MANUAL ............................................................................... 2 4.2.2     RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM .......................................... 2 4.2.2.1     Summary of Estimated Doses ............................................................................................. 3 4.2.3     RADIOACTIVE WASTE HANDLING EVENT ............................................................................. 3 4.3      HEALTH PHYSICS................................................................................................................................. 3 4.3.1     PROGRAM DESCRIPTION ............................................................................................................ 3 4.3.2     TRAINING (TRAINING AND HEALTH PHYSICS) ................................................................................ 3 4.3.3     TOTAL RISK ASSESSMENT ......................................................................................................... 3 4.3.4     ACCESS CONTROL AND THE RWP (ALL ORGANIZATIONS) ....................................................... 4 4.3.5     HEALTH PHYSICS RESOURCES .................................................................................................. 4 4.3.6     RELATED MEDICAL PROGRAMS ............................................................................................... 4 TABLE 4-1, ASSUMPTIONS FOR RADIOACTIVE WASTE HANDLING EVENT.............................................................. 5

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: 4.       RADIOACTIVE WASTE & RADIATION PROTECTION 4.1      RADIOACTIVE WASTE The installed radioactive waste disposal systems that were used when the plant was operating and during transition to dormancy (SAFSTOR 2 - Cold & Dark) have been permanently removed from service. However, the capability to process radioactive waste when required is provided by other means, as described below. Administrative controls assure compliance with applicable regulations. Compliance with the methods and limits of the Offsite Dose Calculation Manual (ODCM), which implements the dose limits of 10 CFR 50, Appendix I, assures that effluent concentration limits and public dose limits are not exceeded. ODCM limits have their bases in 10 CFR 20, 10 CFR 50 Appendix I, and 40 CFR 190.
4.1.1         LIQUID WASTE In SAFSTOR 2 Dormancy, the liquid waste is collected, sampled, managed, and released using the SAFSTOR 2 Liquid Management equipment. All releases of liquid effluent to the environment are under strict administrative control. Liquid effluent releases are in the batch mode, and a numbered discharge permit is issued for each batch release. Details of the sampling and analysis criteria are given in the ODCM.
4.1.2         SOLID WASTE Solid Radioactive wastes are collected and processed on or off site, and shipped to a licensed/permitted burial site for disposal. Solid wastes are packaged in containers which conform to DOT requirements (49 CFR) for transport to a licensed disposal facility. The total curie content and major radionuclide composition by waste type are reported in the Radioactive Effluent Release Report required pursuant to 10 CFR 50.36a.
4.2      RADIATION PROTECTION 4.2.1         OFF-SITE DOSE CALCULATION MANUAL The ODCM provides the information and methodologies used to evaluate the impact of radiological liquid and gaseous effluent discharged from the plant. The ODCM is used to demonstrate that the plant complies with the requirements of 40 CFR 190, 10 CFR 20, and the dose guidelines of 10 CFR 50, Appendix I.
4.2.2         RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM A program is required to monitor the radiation and radionuclides in the environs of the plant. The program provides (1) representative measurements of radioactivity in the highest potential exposure pathways, and (2) verification of the accuracy of the effluent monitoring program and modeling of environmental pathways. The program is (1) contained in the Offsite Dose Calculation Manual (ODCM), (2) conforms to the guidance of 10 CFR 50, Appendix I, and (3) includes the following:
: a. Monitoring, sampling, analysis, and reporting of radiation and radionuclides in the environment in accordance with the methodology and parameters in the ODCM.
: b. A Land Use Census to ensure that changes in the use of areas at and beyond the site boundary are identified and that modifications to the monitoring are made if required by the results of this census, and
: c. Participation in an Interlaboratory Comparison Program to ensure that independent checks on the precision and accuracy of the measurements of radioactive materials in environmental sample matrices are performed as part of the quality assurance program for environmental monitoring.1TSA149 TSA149 This paragraph was added when Technical Specification Amendment 149 was issued. Obtain Licensing &
Regulatory Affairs concurrence before changing.

DEFUELED SAFETY ANALYSIS                             Revision:            006 REPORT                               Chapter:                4 RADIOACTIVE WASTE &                             Page:             3 of 5 RADIATION PROTECTION 4.2.2.1       Summary of Estimated Doses Estimates of the maximum whole body dose and the maximum organ dose to an individual member of the public at or beyond the site boundary (i.e., in the unrestricted area) that would be received as a result of the release of liquid and gaseous effluents, and direct radiation (e.g., ISFSI) dose estimates, can be found in the Annual Radioactive Effluent Release Reports submitted to the NRC.
4.2.3         RADIOACTIVE WASTE HANDLING EVENT A radioactive waste handling event has been postulated to be the limiting event for decommissioning activities during dormancy at CR3. The event is postulated to be the airborne dispersal of radioactive waste resin upon dropping of a high integrity container (HIC) outside of the power block. Although an airborne release is not expected to occur with a drop, or while in storage awaiting shipment due to the low flammability and reactivity of the spent resin, a release is nevertheless postulated.
The activity and isotopic mix is a composite of two resin shipments in order to maximize activity. Resin shipments made from August 2013 through May 2015 were reviewed and the isotopic distribution was obtained from the shipments with the highest activity. This created a composite maximum shipment having a total activity of approximately 265 millicuries. Release fractions were taken from 10 CFR 30, Schedule C. The maximum controlled area atmospheric dispersion factor (X/Q) is used, based on the controlled area as defined in the CR3 Partial Site Release Request.
Table 4-1 contains the design inputs for calculation of the event consequences. The results of the event analysis determined that a receptor at the closest point on the controlled area boundary would receive a dose of < 1 millirem TEDE.
4.3      HEALTH PHYSICS 4.3.1         PROGRAM DESCRIPTION Crystal River Unit 3's Radiation Protection Program resides within the Health Physics and Radiation Safety Procedures. These procedures describe the programmatic content and operating philosophy of the Radiation Protection Program. The Health Physics Section of the Radiation Protection Department is responsible for the administration of these procedures in accordance with the requirements of the Plant Operating Manual (POM).
The Radiation Protection Program is based upon a Risk versus Benefit ALARA methodology and is designed to prevent the occurrence of non-stochastic health effects and to minimize the probability of occurrence of stochastic health effects within three distinct populations: individual radiation workers, the workforce, and members of the general public. TSA149 4.3.2         TRAINING (TRAINING AND HEALTH PHYSICS)
The Radiation Protection Department is responsible for orientation and training of personnel in radiation protection principles and procedures to maintain exposures As Low As Reasonably Achievable (ALARA).
4.3.3         TOTAL RISK ASSESSMENT The health and safety of employees is of paramount importance to Duke Energy Florida (DEF). Therefore, before prescribing the use of any personal protective measures, all of the risk factors associated with the task to be performed will be evaluated and the protective measures chosen will be those that offer the best protection against TSA149 This Section was added when Technical Specification Amendment 149 was issued. Obtain Licensing &
Regulatory Affairs concurrence before changing.

DEFUELED SAFETY ANALYSIS                               Revision:          006 REPORT                                Chapter:             4 RADIOACTIVE WASTE &                              Page:           4 of 5 RADIATION PROTECTION the greatest risk present. To the extent practical, industrial, environmental and radiological risks will be eliminated in the planning phase of the work control process. Personnel monitoring devises, protective clothing, portable shielding and respiratory protection equipment are available for use when conditions warrant.
4.3.4          ACCESS CONTROL AND THE RWP (ALL ORGANIZATIONS)
Only trained/qualified individuals are granted unescorted access into a Radiation Controlled Area (RCA). The Radiation Work Permit (RWP) is the mechanism used to authorize and document the requirements for entry into an RCA. Personnel entering an RCA are required to read and understand the information presented by the RWP authorizing their entry.
A prospective assessment is performed for each category of radiation worker and is reviewed on an annual basis during the yearly Radiation Protection Program self-assessment. This prospective assessment and the requirements specified on the RWP will determine the need for and level of personnel monitoring required for the various work activities to be performed inside of the Restricted Area.
An alternative to access control to High Radiation Areas may be provided as follows:
In lieu of the "control device" or "alarm signal" required by paragraph 20.1601(a) of 10 CFR 20, a High Radiation Area in which the intensity of radiation is greater than 100 mrem/hr but less than 1000 mrem/hr shall be barricaded and conspicuously posted as a High Radiation Area and entrance thereto shall be controlled by issuance of a Radiation Work Permit and any individual or group of individuals permitted to enter such areas shall be provided with one or more of the following:
: a. A radiation monitoring device which continuously indicates the radiation dose rate in the area, or
: b. An integrating alarming dosimeter which alarms when a preset integrated dose or dose rate is received.
Entry into such areas with this alarming dosimeter may be made after the dose rate levels in the area have been established and personnel have been made knowledgeable of them, or
: c. An individual qualified in Health Physics Procedures with a radiation dose rate monitoring device, who is responsible for providing positive control over activities in the area and who performs periodic radiation surveillance at the frequency specified by the Radiation Work Permit.
This provision is addressed in Permanently Defueled Technical Specification 5.8 "High Radiation Area" and must be maintained as stated unless prior NRC approval is granted in accordance with 10 CFR 50.90 and 10 CFR 20.1601(c).
4.3.5          HEALTH PHYSICS RESOURCES The necessary manpower, instrumentation and related equipment needed to support the Radiation Protection Program as described in the appropriate procedures are provided to support the safe operation of the unit.
Instrumentation and equipment are available to sample for the various forms of radioactive materials (i.e., gaseous, liquid and solid) and provide quantitative and qualitative data as necessary. The instrumentation and equipment is periodically checked and calibrated to assure quality performance.
4.3.6          RELATED MEDICAL PROGRAMS Medical qualifications and health physics bioassay requirements must be met to enter Crystal River Unit 3's Respiratory Protection Program. Bioassay services are provided on a routine basis and are available based upon recommendations received by either Health Services or Health Physics personnel. Any employee may request bioassay services and their results at any time. personnel in the respiratory protection program must pass initial and annual physical requirements. Bioassay services (i.e., Invivo counting, urine analysis, etc.) are provided on a routine basis and available on recommendation by either Health Services, Health Physics, or the employee.

DEFUELED SAFETY ANALYSIS                          Revision:           006 REPORT                            Chapter:              4 RADIOACTIVE WASTE &                          Page:           5 of 5 RADIATION PROTECTION TABLE 4-1, ASSUMPTIONS FOR RADIOACTIVE WASTE HANDLING EVENT Input Assumption                                              Basis
: 1. Source activity:                                    1. NUS Waste Resin Shipments 13-045 and 15-020 265 milliCi of mixed fission and activation products
: 2.  /Q:            4.75e-4 sec/m3                     2. Calculation N18-0003
: 3. Breathing Rate:       3.5E-04 m3/s                  3. Regulatory Guide 1.183
: 4. Internal dose conversion factors                    4. Federal Guidance Report 11
: 5. External dose conversion factors                    5. Federal Guidance Report 12
: 6. Release Fraction: 1%, except for H-3 which is      6. 10 CFR 30.72, Schedule C release fraction 50%                                                     for mixed fission and activation products, and Tritium specific release fraction.
END-OF-CHAPTER

Revision:                     006 DEFUELED SAFETY ANALYSIS REPORT                                                     Chapter:                           5 CONDUCT OF OPERATIONS                                                              Page:                   1 of 12 TABLE OF CONTENTS
: 5. CONDUCT OF OPERATIONS .......................................................................................................................... 3 5.1      ORGANIZATION AND RESPONSIBILITY ......................................................................................... 3 5.1.1     FUNCTIONAL ORGANIZATION .................................................................................................. 3 5.1.2     OPERATING ORGANIZATION ..................................................................................................... 3 5.1.3     QUALIFICATIONS ......................................................................................................................... 3 5.2      TRAINING .............................................................................................................................................. 3 5.2.1     CONCEPT ........................................................................................................................................ 3 5.2.2     PLANT TRAINING PROGRAMS ................................................................................................... 3 5.2.2.1     General Employee Training ................................................................................................ 3 5.2.2.2     General Respiratory Training ............................................................................................. 3 5.2.2.3     Operator Training ............................................................................................................... 4 5.2.2.4     Technical Training .............................................................................................................. 4 5.2.2.5     Special Training .................................................................................................................. 4 5.2.2.6     Contractor and Off-Site Employee Training ....................................................................... 4 5.2.3     RETRAINING .................................................................................................................................. 4 5.2.3.1     Health Physics Technicians Retraining .............................................................................. 4 5.2.3.2     Maintenance Section Retraining ......................................................................................... 4 5.2.3.3     Contractor Retraining ......................................................................................................... 4 5.2.4     EXAMINATIONS ............................................................................................................................ 4 5.2.5     DOCUMENTATION........................................................................................................................ 4 5.3      PHYSICAL SECURITY (ISFSI-ONLY SECURITY PLAN)............................................................................ 5 5.4      EMERGENCY PLAN (ISFSI-ONLY EMERGENCY PLAN) ........................................................................... 5 5.5      PLANT FIRE PROTECTION PROGRAM ............................................................................................. 5 5.

==5.1     INTRODUCTION==
 ............................................................................................................................ 6 5.

==5.2     DESCRIPTION==
................................................................................................................................. 6 5.5.3     ORGANIZATION; RESPONSIBILITY, AUTHORITY AND QUALIFICATION............................................................................................................................ 7 5.5.4     ADMINISTRATIVE CONTROLS AND PROCEDURES ............................................................... 7 5.5.5     FIRE AND EMERGENCY RESPONSE ACTIVITIES.................................................................... 7 5.5.6     FIXED FIRE PROTECTION FEATURES WITH CONTROLS AND COMPENSATORY MEASURES .................................................................................................... 7 5.5.6.1     Water Supply System ......................................................................................................... 7 5.5.6.2     Fire Detection ..................................................................................................................... 8 5.5.6.3     Manual Fire Suppression Systems ...................................................................................... 8 5.5.6.4     Controls and Compensatory Measures ............................................................................... 8 5.6      PLANT PROCEDURES .......................................................................................................................... 8 5.7      RECORDS ............................................................................................................................................. 10 5.7.1     OPERATING RECORDS ............................................................................................................... 10

Revision:                     006 DEFUELED SAFETY ANALYSIS REPORT                                                     Chapter:                           5 CONDUCT OF OPERATIONS                                                              Page:                   2 of 12 TABLE OF CONTENTS 5.7.1.1    Operations Log ................................................................................................................. 10 5.7.2     ADMINISTRATIVE RECORDS ................................................................................................... 10 5.7.3     MAINTENANCE RECORDS ........................................................................................................ 10 5.7.4     HEALTH PHYSICS RECORDS .................................................................................................... 10 5.7.4.1    Personnel Exposure .......................................................................................................... 10 5.7.4.2    Radiological Surveys ........................................................................................................ 10 5.7.4.3    Survey Instrument Calibration .......................................................................................... 10 5.7.4.4    Radiological Monitoring and Waste Disposal .................................................................. 11 5.7.4.5    Instrumentation Calibration .............................................................................................. 11 5.7.4.6    Shipping, Receiving and Inventory of Radioactive Material ............................................ 11 5.7.5     OTHER RECORDS ........................................................................................................................ 11 5.7.5.1    Special Nuclear Material Inventory .................................................................................. 11 5.8      ADMINISTRATIVE CONTROL .......................................................................................................... 11 5.8.1     INDEPENDENT SAFETY REVIEWER (ISR) .............................................................................. 11 5.8.1.1    Composition...................................................................................................................... 11 5.8.1.2    Review Frequency ............................................................................................................ 11 5.8.1.3    Responsibilities ................................................................................................................. 12 5.8.1.4    Requirements .................................................................................................................... 12 5.8.1.5    Records ............................................................................................................................. 12

Revision:          006 DEFUELED SAFETY ANALYSIS REPORT                                Chapter:             5 CONDUCT OF OPERATIONS                                Page:          3 of 12
: 5.        CONDUCT OF OPERATIONS 5.1       ORGANIZATION AND RESPONSIBILITY 5.1.1         FUNCTIONAL ORGANIZATION The organization responsible for the Crystal River Unit 3 (CR3) plant is headed by the Senior Vice President, Operations Support. Additional description of the plant and corporate organization is provided in Section 1.3 5.1.2         OPERATING ORGANIZATION The CR3 organization is discussed in Section 1.3.
5.1.3         QUALIFICATIONS Crystal River Unit 3 personnel have the combination of education, skill, and experience commensurate with their level of responsibility. These qualifications provide reasonable assurance that decisions and actions during all conditions are such that the plant is maintained in a safe manner. The qualifications of plant managerial, supervisory, technical support and technician personnel meet or exceed the requirements set forth in ANSI N18.1-1971 as endorsed by Regulatory Guide 1.8, Rev. 1 (9/75) and clarified by the Crystal River Quality Assurance Program Manual (QAPM).
5.2       TRAINING 5.2.1         CONCEPT Duke Energy Florida has implemented a training program designed to indoctrinate personnel in the administrative and technical aspects of the plant. The overall conduct and administration of the training program of plant personnel is the responsibility of the General Manager Decommissioning.
5.2.2         PLANT TRAINING PROGRAMS 5.2.2.1       General Employee Training All new employees will receive Generic Plant Access Training prior to being granted unescorted access to Crystal River Unit 3. The program is designed to familiarize new employees with Crystal River Unit 3 policies and commitments in the areas of Security, Administrative Instructions, Emergency Procedures, Industrial Safety, and Compliance.
Additional training will be provided to those employees whose jobs require work in the Radiation Controlled Area (RCA) of the Crystal River Nuclear Plant. This training will familiarize personnel with radiological procedures.
5.2.2.2       General Respiratory Training All employees whose job at the Crystal River Nuclear Plant requires the use of respiratory equipment will receive Respiratory Protection Training.

Revision:          006 DEFUELED SAFETY ANALYSIS REPORT                              Chapter:               5 CONDUCT OF OPERATIONS                            Page:           4 of 12 5.2.2.3        Operator Training This training program provides background training to Crystal River Unit 3 personnel who are monitoring the safe storage of nuclear fuel.
5.2.2.4        Technical Training All craftsmen will participate in a training program. The program will provide training to craft personnel in areas pertinent to their job functions.
5.2.2.5        Special Training From time to time, special "one-time" courses may be required. One example of this type of training is the "Independent Spent Fuel Storage Installation Overview" training Crystal River Unit 3 presented to plant employees.
5.2.2.6        Contractor and Off-Site Employee Training All contractor and off-site personnel requiring unescorted access to CR3 will be required to participate in Generic Plant Access Training.
5.2.3          RETRAINING 5.2.3.1        Health Physics Technicians Retraining Retraining for Radiation Protection personnel will be conducted under the Technical Training Program (see Section 5.2.2.4).
5.2.3.2        Maintenance Section Retraining Retraining for Mechanical Maintenance Personnel, Electrical Maintenance Personnel, and Instrumentation and Control Technicians will be conducted under the Technical Training Program (see Section 5.2.2.4).
5.2.3.3        Contractor Retraining Retraining for contractor personnel will be conducted under the guidance of Section 5.2.2.6, "Contractor and Off-Site Employee Training."
5.2.4          EXAMINATIONS Various types of examinations will be administered during the training of Crystal River Unit 3 personnel.
A minimum score of 70% will be used as a cut-off for passing unless otherwise stated.
5.2.5          DOCUMENTATION All training will be documented. Training documentation will be maintained by the Technical Support Manager.

Revision:          006 DEFUELED SAFETY ANALYSIS REPORT                              Chapter:              5 CONDUCT OF OPERATIONS                              Page:         5 of 12 5.3       PHYSICAL SECURITY (ISFSI-ONLY SECURITY PLAN)
Duke Energy Florida has developed and submitted to the NRC the ISFSI-Only Security Plan. Pursuant to 10 CFR 2.390(d), the plan has been determined to contain proprietary and/or safeguards information and shall be withheld from public disclosure. For the latest revisions of the Plans, contact Crystal River Unit 3 Security or Licensing.
5.4       EMERGENCY PLAN (ISFSI-ONLY EMERGENCY PLAN)
Duke Energy Florida has developed the ISFSI Only Emergency Plan to describe the elements of an integrated preparedness program to respond to potential emergencies at CR3. In the event of an emergency at CR3, actions are required to identify and assess the nature of the emergency and bring it under control in a manner that protects the health and safety of plant personnel.
The ISFSI Only Emergency Plan describes the organization and responsibilities of Duke Energy Florida and its CR3 and Corporate staffs for implementing emergency measures. It describes interfaces with Federal, State of Florida, and Citrus County organizations, to be notified in the event of an emergency and may provide, or be requested to provide, assistance.
CR3 is licensed under the requirements of 10 CFR 50, "Domestic Licensing of Production and Utilization Facilities," and 10 CFR 72, Licensing Requirements For The Independent Storage of Spent Nuclear Fuel, High Level Radioactive Waste, and Reactor-Related Greater Than Class C Waste. Consistent with the requirements of 10 CFR 50, this Plan is based upon the requirements of 10 CFR 50, Section 50.47(b) and Appendix E, "Emergency Planning and Preparedness for Production and Utilization Facilities," as applicable to CR3 in its permanently shutdown and defueled status. Sections 5.0 through 20.0 of this Plan address the standards delineated in 10 CFR 50.47(b)(1) through (16). In addition, the Plan is also intended to meet appropriate State of Florida and U.S.
Nuclear Regulatory Commission (NRC) regulations in accordance with Duke Energy Floridas Operating License (No. DPR-72).
Because the analyses of the credible design basis events and consequences indicates there are no postulated accidents that would result in off-site dose consequences that require off-site emergency planning, emergencies are divided into two classifications: 1) Notification of an Unusual Event (Unusual Event) and 2) Alert. This classification scheme has been discussed and agreed upon with responsible off-site organizations and is compatible with the State Plan.
Duke Energy Florida is responsible for planning and implementing emergency measures within the CR3 CONTROLLED AREA. This Plan is provided to meet that responsibility. To carry out specific emergency measures discussed in the Plan, detailed implementing procedures are established and maintained. The Plan provides a listing of the implementing procedures.
In addition to the description of activities and steps that can be implemented during a potential emergency, this Plan provides a general description of the steps taken to recover from an emergency situation. It also describes the training, exercises, planning, and coordination appropriate to maintain an adequate level of emergency preparedness.
5.5       PLANT FIRE PROTECTION PROGRAM The Fire Protection Program is described fully in the Fire Protection Plan. With all spent fuel in ISFSI, the Fire Protection Program requirements for the Dormancy Phase of Decommissioning are limited to preventing or minimizing the release of radioactive materials resulting from fires involving contaminated plant SSCs or radioactive wastes.

Revision:       006 DEFUELED SAFETY ANALYSIS REPORT                                 Chapter:           5 CONDUCT OF OPERATIONS                                Page:        6 of 12 Only those areas that contain radioactive materials are required to be addressed by the Plan which includes the Auxiliary Building, the Intermediate Building, and the Reactor Building. Elements of the Fire Protection Program are applied to other buildings at Management discretion.
5.

==5.1         INTRODUCTION==

The Fire Protection Program at Crystal River Unit 3 (CR3) consists of activities and functions that are performed to minimize the probability and consequences of a postulated fire. In the event of a fire, the program and system designs ensure the requirements of 10 CFR 50.48(f) are satisfied.
The Fire Protection Program has been formulated in accordance with the following governing documents:
: a. Fire Hazard Analysis The Fire Hazard Analysis (FHA) is a study of plant designs, potential fire hazards in the plant, potential threats of these hazards occurring and a fire loading for the various fire areas. The FHA identifies the location and designations of fire areas and zones within Crystal River Unit 3. The Fire Hazard Analysis is updated periodically to reflect plant modifications which affect the document. Changes to the FHA are performed under the 10 CFR 50.59 evaluation process per 10 CFR 50.48(f)(3).
: b. Fire Protection Plan The Fire Protection Plan has been developed to describe the operational elements of the Fire Protection Program for Crystal River Unit 3 in order to assure response to a fire emergency is timely and adequate.          The Plan addresses the Fire Protection Program organization, responsibilities, authorities, administrative controls, manual suppression equipment, procedures, training, and basic design change processes. The Fire Protection Plan addresses the functionality requirements, limitations and surveillance requirements of the Fire Protection Program. The Fire Protection Plan is updated periodically, as necessary, in support of changes to the program.
Changes to the Fire Protection Plan are performed under the 10 CFR 50.59 evaluation process per 10 CFR 50.48(f)(3).
5.

==5.2         DESCRIPTION==

The Fire Protection Program uses the Defense-In-Depth concept to achieve a high degree of fire safety at CR3. The Defense-In-Depth concepts: (1) Reasonably prevent fires from occurring; (2) Rapidly detect, control, and extinguish fires that do occur and that could result in radiological hazard; and (3) Ensure that the risk of fire-induced radiological hazards to the public, environment and plant personnel is minimized.
Details of each element of the CR3 Fire Protection Program are provided in the Crystal River Unit 3 Fire Protection Plan. Elements of the fire protection program include the following:
: a. Organization; Responsibility, Authority and Qualifications
: b. Administrative Controls and Procedures
: c. Fire and Emergency Response Activities
: d. Fixed Fire Protection Features with Controls and Compensatory Measures
: e. Reducing risk of fire-induced radiological hazard to the public, environment and plant personnel.

Revision:         006 DEFUELED SAFETY ANALYSIS REPORT                                  Chapter:             5 CONDUCT OF OPERATIONS                                 Page:          7 of 12 5.5.3          ORGANIZATION; RESPONSIBILITY, AUTHORITY AND QUALIFICATION The line of authority for those positions with responsibility for the fire protection program is defined in the CR3 Fire Protection Plan with detailed responsibilities and qualifications.
5.5.4          ADMINISTRATIVE CONTROLS AND PROCEDURES Administrative controls covering CR3's Fire Protection Program are provided by the Fire Protection Plan and the Plant Operating Manual. Controls that are established include administrative restraints of:
: a. In Situ and transient combustibles
: b. Use of temporary structures
: c. Ignition sources
: d. Smoking
: e. Leak testing
: f. Design, maintenance and plant modification processes
: g. Surveillance of installed systems 5.5.5          FIRE AND EMERGENCY RESPONSE ACTIVITIES In the event of a fire incident at CR3, response actions are defined by plant procedures contained in the Plant Operating Manual (POM).
Emergency communications can be accomplished by telephone or plant radios.
Local area fire departments provide manual fire suppression capability at CR3. This assistance is defined under a Memorandum of Understanding (MOU). CR3 Pre-Fire Plans are available to support fire department operations.
5.5.6          FIXED FIRE PROTECTION FEATURES WITH CONTROLS AND COMPENSATORY MEASURES Design and administrative controls ensure that fire protection features are installed and maintained to perform their intended function.
In situ fire protection at CR3 includes (but is not limited to) a water supply system and manual fire suppression systems. These installed features provide safety of both personnel and plant property, as described in the Fire Protection Plan.
5.5.6.1        Water Supply System There is one storage tank, Fire Service Tank 1A (FST-1A), containing 277,000 gallons surveilled volume dedicated to fire service water storage. Tank level is monitored during plant rounds via a pressure indicator mounted on the southeastern side of FST-1A. Tank level make-up is available manually from the Fossil Fire Protection system or via tanker truck(s).
A new fire service yard main is installed on the North, West and South Berms, supplied by FST-1A. Private fire hydrants are placed along the yard mains in areas where fire department apparatus can access plant structures.
Isolation and sectional valves are provided along the yard mains to facilitate maintenance.

Revision:          006 DEFUELED SAFETY ANALYSIS REPORT                            Chapter:              5 CONDUCT OF OPERATIONS                             Page:          8 of 12 5.5.6.2       Fire Detection Fire detection at CR3 in dormant plant areas is by observation.
5.5.6.3       Manual Fire Suppression Systems Manual fire suppression entails the use of fire protection equipment intended for use by trained personnel. Such equipment at CR3 includes:
: a. Fire Extinguishers Portable fire extinguishers are located throughout the site, but not within dormant plant areas. The types of fire extinguishers in use are listed in the Fire Protection Plan.
: b. Standpipes The standpipe systems installed at CR3 are Class I, installed in the Auxiliary Building and Turbine Building.
: c. Fire Hydrants Eight (8) private fire hydrants, each connected to the yard main, are installed on the north, west and south sides of the berm.
5.5.6.4       Controls and Compensatory Measures The CR3 fire protection program is designed to assure that adequate levels of protection are provided at all times.
This is accomplished by establishing controls and compensatory measures through surveillance procedures, as well as requiring reporting and documenting of noncompliances when systems, barriers, and other fire protection features are determined to be non-functional.
5.6      PLANT PROCEDURES Plant procedures provide detailed written instructions designed to govern the normal and off-normal conditions under which the plant is maintained. These procedures are developed to meet the intent of Appendix A of Regulatory Guide 1.33, "Quality Assurance Requirements (Operation)," Revision 2, February 1978, and are in accordance with the operational Quality Assurance Program described in the QAPM.
Plant procedures are written, reviewed, and implemented by the plant staff in accordance with the QAPM. These procedures provide instructions for the safe operation, control, and maintenance of plant systems and equipment and furnish documentation of actions taken.
The following procedure categories are to be included in the Plant Operating Manual (POM):
: a. Administrative Instructions Administrative Instructions cover the following topics: administrative policies, plant organization and responsibilities, conduct of plant operations, conduct of maintenance, conduct of training, and housekeeping.
: b. Operating Procedures Operating Procedures are designed to provide instructions for operations and include: electrical system operation, and waste disposal.

Revision:          006 DEFUELED SAFETY ANALYSIS REPORT                                Chapter:               5 CONDUCT OF OPERATIONS                               Page:           9 of 12
: c. Emergency Plans and Implementing Procedures These procedures provide instructions and outline responsibilities of on-site and off-site personnel in the event of a site emergency. The Emergency Plan, fire protection, emergency classification, evacuation and notification, medical emergencies, and violent weather documents are included in this category.
: d. Maintenance Procedures Maintenance Procedures provide written instructions on the maintenance, repair, and replacement of equipment and components.
: e. Waste Handling, Chemistry, and Radiation Protection Procedures These procedures provide written instructions governing the surveillance, scheduling, and control of waste handling, chemistry, and radiological protection activities.
: f. Compliance Procedures Compliance Procedures provide a means to ensure plant commitments and requirements are met. Activities controlled by these procedures include:          processing of Licensing correspondence, commitment management, administration of some regulatory required programs, and reporting requirements.
: g. Surveillance Procedures Surveillance Procedures include those tests, checks, calculations, calibrations, inspections, and reports required to be performed on a periodic basis. These include all phases of plant operation and plant safety to ensure all systems/components are functional and/or available. There are no ITS/PDTS Surveillance Requirements. There are, however, Surveillance Procedures which implement Offsite Dose Calculation (ODCM)-related Radiological Technical Specifications Surveillance Requirements.
: h. Special Nuclear Materials Handling and Accountability Manual This document describes the procedures to be followed at Crystal River Unit 3 to meet the requirements of 10 CFR 70, governing the handling and accountability of special nuclear material.
: i. Security Procedures Security Procedures encompass the entire spectrum of actions, planned or implemented, to maintain ISFSI security. Included as a part of these procedures are the following areas: guard force organization, operation and responsibilities, alarm systems, personnel and materials access, communications, security equipment, contingency actions, emergency actions, loss or breach of security, and security training.
: j. Preventive Maintenance Procedures Preventive Maintenance Procedures are written to ensure proper maintenance is provided for equipment/components to prevent the unscheduled outage of such equipment due to sudden failure.
: k. Quality Control Procedures Quality Control Procedures include those procedures designed to control materials, equipment, services, drawings, documents, instructions, and special processes to support plant operations. These procedures also include the training requirements of Quality Assurance and Quality Control inspection and compliance personnel.
: l. Other procedures may be used which provide detailed written instructions for the performance of processes not governed by any of the aforementioned procedures. These procedures provide a preplanned method of conducting activities in order to reduce errors. These procedures may include instructions for temporary procedures, work controls, information technology, and other functions, and are used by the plant staff to ensure that normal operations are conducted in a safe manner.

Revision:           006 DEFUELED SAFETY ANALYSIS REPORT                                Chapter:               5 CONDUCT OF OPERATIONS                                Page:          10 of 12 5.7      RECORDS 5.7.1         OPERATING RECORDS The following records of operation will be kept. These will be preserved as required.
5.7.1.1       Operations Log The Shift Supervisor summarizes plant conditions and events during his shift in the Operations log. The Operations log shall include plant status information and should include any changes in status of the availability of systems, unusual occurrences, etc. At the end of each shift, the Shift Supervisor signs the log, signifying the report is a comprehensive, accurate summary of plant events and activities.
5.7.2         ADMINISTRATIVE RECORDS The following is the responsibility of the General Manager Decommissioning:
Investigation and reporting of abnormalities and corrective action taken including, but not limited to, the following:
: a. Personnel overexposure
: b. Loss or theft of licensed radioactive material Corrective action will be taken immediately.
5.7.3         MAINTENANCE RECORDS The Shift Supervisors are responsible for the maintenance records for plant equipment and instrumentation.
5.7.4         HEALTH PHYSICS RECORDS The Manager Radiation Protection and Chemistry maintains the following Health Physics records:
5.7.4.1       Personnel Exposure The Supervisor, Nuclear Radiation Protection maintains the following:
: a. Dosimetry records (monthly or more frequently)
: b. Radio bioassay records (as deemed necessary)
: c. Records of radiation exposure history and current exposure status (as required by 10 CFR 20) 5.7.4.2       Radiological Surveys The Supervisor, Nuclear Radiation Protection is responsible for routine radiological surveys and job-specific radiological surveys.
5.7.4.3       Survey Instrument Calibration The Supervisor, Nuclear Radiation Protection maintains all HP survey meters in accordance with 10 CFR 20.2103.

Revision:         006 DEFUELED SAFETY ANALYSIS REPORT                                 Chapter:            5 CONDUCT OF OPERATIONS                                 Page:        11 of 12 5.7.4.4       Radiological Monitoring and Waste Disposal The Manager Radiation Protection and Chemistry is responsible for the following:
: a. Liquid waste discharged
: b. Gaseous activity released
: c. Monitoring reports 5.7.4.5       Instrumentation Calibration The Manager Radiation Protection and Chemistry maintains all environmental monitors.
5.7.4.6       Shipping, Receiving and Inventory of Radioactive Material The Manager Radiation Protection and Chemistry is responsible for the following:
: a. Records in sufficient detail to satisfy the appropriate sections of 10 CFR 20, 10 CFR 70, 10 CFR 71 and 49 CFR 173.
: b. Solid radioactive waste shipped.
5.7.5         OTHER RECORDS 5.7.5.1       Special Nuclear Material Inventory The Site SNM Custodian is responsible for records in sufficient detail to satisfy the requirements of 10 CFR 70.
5.8      ADMINISTRATIVE CONTROL Administrative controls are established to ensure plant operations, maintenance tests, and emergency responses are performed in accordance with reviewed and approved procedures. The General Manager Decommissioning has the responsibility and authority to operate the plant within the limits of the administrative controls.
A review of the operating logs, charts, and other data is performed by Operations personnel.
5.8.1         INDEPENDENT SAFETY REVIEWER (ISR)
An Independent Safety Reviewer (ISR) is established to review activities in accordance with Section 5.8.1.3. The ISR recommends to the General Manager Decommissioning approval or disapproval of the activities reviewed in Section 5.8.1.3.
5.8.1.1       Composition An Independent Safety Review is performed by individual(s) designated by the General Manager Decommissioning, independent of activities assessed and who provide the appropriate level of expertise in the activities assessed.
5.8.1.2       Review Frequency ISRs are convened by the General Manager Decommissioning or his designated alternate when the need arises.

Revision:         006 DEFUELED SAFETY ANALYSIS REPORT                                 Chapter:            5 CONDUCT OF OPERATIONS                                Page:        12 of 12 5.8.1.3       Responsibilities
: a. Proposed changes to the facility as described in the DSAR. This review is to confirm that the change does not adversely affect safety and if a Permanently Defueled Technical Specification change or NRC review is required.
: b. Proposed changes to procedures as described in the DSAR and tests or experiments not described in the DSAR. This review is to confirm that the change does not adversely affect safety and if a Permanently Defueled Technical Specification change or NRC review is required.
: c. Proposed Permanently Defueled Technical Specification changes and other license amendments, except in those cases where the change is identical to a previously reviewed change.
: d. License Event Reports that are required to be submitted to the NRC. This review includes results of any investigations made and recommendations resulting from such investigations to prevent or reduce the probability of recurrence of the event.
: e. Investigation of all violations of the Technical Specifications, including the preparation and forwarding of reports covering evaluation and recommendations to prevent recurrence, to the General Manager Decommissioning.
: f. Review proposed changes to the ISFSI-Only Security Plan that have been initially determined to decrease the effectiveness of the plan.
: g. Review proposed changes to the ISFSI-Only Emergency Plan that have been initially determined to decrease the effectiveness of the plan.
: h. Review of any accidental, unplanned or uncontrolled radioactive release, including the preparation of reports covering evaluation, recommendations and disposition of the corrective action to prevent recurrence, and the forwarding of these reports to the General Manager Decommissioning.
: i. Review, prior to implementation, changes to the ODCM.
5.8.1.4       Requirements Render determinations in writing with regard to whether or not each item considered under Section 5.8.1.3 complies with the facility license and governing regulations.
5.8.1.5       Records Document the results of all ISR activities performed under the responsibility provisions of this DSAR. Copies shall be provided to the General Manager Decommissioning.
END-OF-CHAPTER

DUKE ENERGY FLORIDA, LLC CRYSTAL RIVER UNIT 3 LICENSE NUMBER DPR-72 DOCKET NUMBER 50-302 / 72-1035 ATTACHMENT 4 CR3 QAPM

CR3-QAPM Revision 0 CRYSTAL RIVER QUALITY ASSURANCE PROGRAM MANUAL (CR3-QAPM)
Revision 0 Crystal River Unit 3 (CR3) 1

CR3-QAPM Revision 0 Crystal River Nuclear Power Plant Quality Assurance Program Manual (QAPM)
Approvals signed electronically ____________________________________
Manager, Technical Support SAFSTOR                     Date 2

CR3-QAPM Revision 0 TABLE OF CONTENTS 1.0        Quality Program - Policy Statement ................................................................ 4 2.0        Introduction ..................................................................................................... 4 3.0        Organization .................................................................................................... 5 4.0        Quality Program ............................................................................................ 11 5.0        Design Control .............................................................................................. 13 6.0        Procurement Document Control .................................................................... 13 7.0        Instructions, Procedures and Drawings ......................................................... 14 8.0        Document Control ......................................................................................... 15 9.0        Control of Purchased Material, Equipment and Services .............................. 15 10.0       Identification and Control of Materials, Parts and Components..................... 16 11.0       Control of Special Processes ........................................................................ 16 12.0       Inspections .................................................................................................... 17 13.0       Test Control................................................................................................... 17 14.0       Control of Measurement and Test Equipment ............................................... 18 15.0       Handling, Storage and Shipping.................................................................... 19 16.0       Inspection, Tests and Operating Status ........................................................ 19 17.0       Nonconforming Material, Parts, or Components ........................................... 19 18.0       Corrective Action ........................................................................................... 20 19.0       Quality Assurance Records ........................................................................... 20 20.0       Audits and Independent Reviews .................................................................. 23 21.0       Program Commitment ................................................................................... 25 22.0       Quality Assurance Staff ................................................................................. 25 23.0       Glossary Of Terms ........................................................................................ 26 Table 1 - Crystal River Unit 3 In-Plant Quality Program Functions................................ 27 Table 2 - Crystal River Unit 3 Quality Program Commitments ...................................... 28 Figure 1 - CR3 Nuclear Organization ............................................................................ 30 3

CR3-QAPM Revision 0 1.0     Quality Program - Policy Statement Duke Energy Florida (DEF) maintains and operates Crystal River Unit 3 (CR3) in a manner that will ensure the health and safety of the public and workers. The facility shall be operated in compliance with the requirements of the Code of Federal Regulations, the applicable Nuclear Regulatory Commission (NRC) Facility Operating Licenses, and applicable laws and regulations of the state and local governments.
The Quality Assurance Program Manual (QAPM) includes the description in the following sub-sections and the associated implementing documents that provide for control of activities that affect the quality of SSCs classified as important-to-safety (ITS) to satisfy the requirements of 10 CFR 71 and 10 CFR 72.
The QAPM contained here-in is the top-level policy document that establishes the manner in which quality is to be achieved and presents our overall philosophy regarding achievement and assurance of quality. Implementing documents assign more detailed responsibilities and requirements and define the organizational interfaces involved in conducting activities within the scope of the QAPM. Compliance with the QAPM and implementing documents is mandatory for personnel directly or indirectly associated with implementation of the QAPM.
Responsibility for developing, implementing, and verifying execution of the QAPM is delegated to the General Manager Decommissioning and authority for developing and verifying execution of the program to the senior manager for Nuclear Oversight (NOS).
This Quality Program is revised and submitted to the NRC as required by 10 CFR 50.54(a). For later interim changes, contact the NOS Section.
2.0     Introduction With the Certification of Permanent Cessation of Power Operations DEF has permanently ceased power operations of CR3. To address this changing environment at CR3, a QAPM has been developed to support the decommissioning activities of the station and ensure continued oversight of SAFSTOR, Decommissioning and the Independent Spent Fuel Storage Installation (ISFSI). Sections 2.0, 21.0 (and Table 1),
22.0 and 23.0 describe the DEF Quality Program. With the completion of the transfer of the Spent Nuclear Fuel into the ISFSI, CR3 no longer contains any nuclear safety related structures, systems, or components (SSCs).
The QAPM satisfies the requirements of 10 CFR 50 Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Processing Plants. Since CR3 no longer has any nuclear safety related SSCs, the primary focus of the QAPM is on satisfaction of applicable requirements of 10 CFR 71, Subpart H, Quality Assurance for Packaging and Transportation of Radioactive Material, and 10 CFR 72, Subpart G, Quality Assurance for Independent Storage of Spent Nuclear Fuel and High-Level Radioactive Waste. As such, the QAPM provides an overview of the quality program controls which governs the operation and maintenance of CR3 ITS items and activities controlled by 10 CFR 72 and to the usage of transportation packages controlled by 10 CFR 71. The methods of implementation of the requirements of the QAPM are commensurate with the item's or activity's importance to safety. The applicability of the requirements of the QAPM to 4

CR3-QAPM Revision 0 other items and activities is determined on a case-by-case basis as defined within approved procedures. Additional regulatory commitments are listed within Table 2.
3.0     Organization CR3 has established the functional responsibilities and authorities of positions/organizations involved in the Quality Program. The functional departmental responsibilities described below, serves to document the relationships between organization positions. In certain instances, duties and authority to execute and audit the quality activities are delegated to other organizations. In all cases, DEF retains responsibility for the Quality Program for CR3.
Verification of conformance to established quality requirements for ITS items and activities are accomplished by those individuals or groups who do not have direct responsibility for performing or directly supervising the work being verified.
Persons and organizations performing quality assurance functions have sufficient authority and organizational freedom to identify quality problems; initiate necessary action to provide for resolution of nonconformances through designated channels; verify implementation of solutions; and, control further processing, delivery, or installation of a nonconforming item until the proper disposition of the deficiency or the unsatisfactory condition has been approved.
The following summarizes the functional responsibilities and authorities of positions/organizations involved in directing and managing the CR3 Quality Program (see Figure 1):
: 1. Duke Energy Corporation Corporate Organization The Chairman, President and Chief Executive Officer has overall responsibility for Design, Construction, and Operation of generation and transmission facilities. The Executive Vice President and Chief Operating Officer of Duke Energy reports to the Chairman, President and Chief Executive Officer and is responsible for: the Duke nuclear operating fleet; enterprise project management and construction; new plant development and construction; and decommissioning activities. The Executive Vice President and Chief Operating Officer of Duke Energy has overall authority and responsibility for the QA Program. Reporting to the Executive Vice President and Chief Operating Officer of Duke Energy is the Chief Nuclear Officer (CNO) who directs several activities including the operation of the operating nuclear sites through the Senior Vice Presidents, Nuclear Operations. Also reporting to the Chairman, President and Chief Executive Officer are Group Executives responsible for providing support to the Nuclear plants for the following: electrical transmission; electrical distribution; laboratory services; switchyard maintenance and technical support; support for the emergency response communications; Information Technology Services; document control and record management activities; and administration of the Access Authorization, Fitness for Duty, and Fatigue Rule programs. As such, the attainment of quality rests with those assigned the responsibility of performing the activity. The verification of quality is assigned to qualified personnel independent of the responsibility for performance or direct supervision of the activity. The degree of independence varies commensurate with the activity's importance to safety.
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: 2. Duke Energy Nuclear Duke Energy Nuclear is responsible for: the nuclear operating fleet; new nuclear plant development and construction; enterprise project management and construction; and decommissioning activities. These activities are directed by the Executive Vice President and Chief Operating Officer of Duke Energy.
: 3. Operations Support The Senior Vice President Operations Support reports directly to the Executive Vice President and Chief Operating Officer of Duke Energy. Operations Support is responsible for defining and executing the decommissioning strategy for CR3 to meet required regulations and commitments. The Vice President Project Management and Construction reports to the Senior Vice President Operations Support and is responsible for contracts, engineering oversight, and management related to existing plant upgrades, modifications, new plant construction and construction of the ISFSI at CR3, as requested.
: a. General Manager Decommissioning The General Manager Decommissioning reports to the Senior Vice President Operations Support who reports to the Executive Vice President and Chief Operating Officer of Duke Energy. The General Manager Decommissioning shall have onsite responsibility for overall plant safety and shall ensure acceptable performance of the staff in operating, maintaining, and providing technical support to the plant ITS items and activities. The results of Independent Management Assessments are reported to this executive.
The General Manager Decommissioning is given overall authority to staff, operate, and maintain CR3. In carrying out this assignment, the General Manager Decommissioning has the support of an organization consisting of the following:
* Operations & Maintenance Manager
* Decommissioning Technical Support Manager
* Radiation Protection & Chemistry Manager The General Manager Decommissioning is responsible for:
* The safe, economic, and environmentally sound decommissioning of the nuclear plant, and shall delegate in writing the succession during his absence.
* Planning and monitoring of all activities necessary to achieve SAFSTOR conditions/dormancy with dry fuel storage.
* The CR3 ALARA Program.
* Development, maintenance, and interpretation of the monitoring program and the radiological effluent program as delineated in the Off-Site Dose Calculation Manual (ODCM).
* Investigation and reporting of abnormalities and corrective action taken including, but not limited to, the following:
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         -   Uncontrolled release of radioactivity
         -   Personnel overexposure
         -   Loss or theft of licensed radioactive material
         -   Corrective actions will be taken immediately in accordance with the Permanently Defueled Technical Specifications (PDTS).
         -   Approving prior to implementation, each proposed test, experiment, or modification to systems or equipment that affect stored nuclear fuel.
: 1) Operations & Maintenance Manager The Operations & Maintenance Manager is responsible for:
* Operation of plant equipment required to support SAFSTOR and Spent Fuel cooling and monitoring
* Operations Support for decommissioning activities
* Implementation of the Emergency Planning Program as described in the Defueled Safety Analysis Report (DSAR) Section 5.4
* Implementation of the CR3 Fire Protection Plan
* Managing, directing, and supervising the activities of the maintenance work controls and integrated daily work schedule
* Plant Maintenance
* Project Supervision
* Construction
* Facility Services
* Directing the activities of the day-to-day scheduling
* Initial and continuing training of Operations and Maintenance personnel
: 2) Radiation Protection & Chemistry Manager The Radiation Protection & Chemistry Manager is responsible for:
* Directing the overall plant chemistry activities of CR3 to ensure that the plant is functioning within prescribed procedures in the Chemistry area.
* Interfacing with regulatory agencies and organizations to ensure an effective working relationship between plant and agency personnel and to ensure cost effective compliance.
* Establishing program controls for offsite-dose-calculation methodologies, radioactive-effluent controls and radiological environmental-monitoring activities. These controls are contained in the ODCM which shall contain the contents specified within Compliance Procedure CP0500.
* Maintaining a regular interface with other departments within the company and is responsible for developing the processes needed to manage the Chemistry and Radiation Protection Departments including planning, monitoring, and follow-up to ensure consistent high performance.
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* Directing the overall plant radiation protection activities of CR3 to ensure that the plant is functioning within prescribed procedures in the Radiation Protection area.
* Performing reviews as required by 10 CFR 20.1101(c).
* Generating, managing, and maintaining the Historical Site Assessment for use during and after the SAFSTOR period.
* Initial and continuing training of radiation protection, chemistry, and environmental personnel.
: 3) Decommissioning Technical Support Manager The Decommissioning Technical Support Manager is responsible for:
* Supporting the Emergency Planning Program as described in DSAR Section 5.4.
* Coordinating changes to procedures as necessary for changing plant conditions as described in Section 7.0.
* Providing Document Control services as described in Section 8.0.
* Facilitating site personnel Training and qualification documentation.
* Maintaining Performance Improvement as described in Sections 17.0 and 18.0.
* Supporting licensing activities to prepare, control, and support regulatory activities associated with placing the facility in SAFSTOR and eventual decommissioning.
* Updating the DSAR and PDTS.
* Serving as the primary contact for the NRC.
* Initial and continuing training of Engineering personnel.
* Responsible for strategic planning and scheduling of decommissioning related activities.
* Coordinating site support and alignment associated with conduct of decommissioning.
* Providing engineering leadership and oversight to ensure the safe, reliable, and cost efficient decommissioning of the plant.
* Performing engineering activities that maintain configuration control of the plant design basis.
* Resolving engineering related issues to support safe and efficient operation of systems required to assure safe storage of spent fuel.
* Providing engineering leadership and support for managing engineering programs.
* Developing, controlling and monitoring the Fire Protection Program which includes ensuring that the program meets all regulatory requirements imposed by the regulatory agencies as they relate to acceptable implementation, and for ensuring the program is adequately implemented.
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* Participating in the Project Management Oversight Board (PMOB), Plant Emergency Response Organization (ERO), SAFSTOR related organizations, Financial Management Steering Committee and Site programs as required.
* Identifying SSC designations throughout the period up to and including the SAFSTOR period.
* Updating required SSC designations and classifications as changes to SSCs occur.
* Revising engineering documents to reflect changing materiel condition.
* Ensuring the final plant configuration supports remediation activities following the SAFSTOR period.
* Coordinating engineering activities with the Historical Site Assessment.
* Ensuring Independent Audits are periodically performed to monitor overall performance and confirm that activities affecting quality comply with the QA Program and that the QA Program is effectively implemented.
* The Independent Audits are performed by individual(s) designated by the General Manager Decommissioning, independent of activities assessed and who provide the appropriate level of expertise in the activities assessed.
* The Independent Management Assessment results are communicated in an understandable form and in a timely fashion to a level of management having the authority to effect corrective action. In addition, these results are reported in a timely fashion to the General Manager Decommissioning.
: 4. Nuclear Oversight The senior manager for NOS reports to the CNO. The senior manager for NOS is responsible for and reports to the CNO on all matters related to the independent monitoring and auditing of activities performed by the line organizations for, or in support of Duke Nuclear plants and CR3 activities. NOS provides oversight of Nuclear Decommissioning through QA program audits. The senior manager for NOS has the authority and organizational freedom to: identify quality problems, initiate, recommend or provide solutions to quality problems through designated channels, verify the implementation of solutions to quality problems, and ensure cost and schedule do not influence decision making involving quality. This includes full access to Nuclear Decommissioning and all levels of management up to and including the Chief Executive Officer of Duke Energy Corporation.
The senior manager for NOS is delegated primary ownership of the CR3 QA program. If significant quality problems are identified, NOS personnel have the authority to stop work pending satisfactory resolution of the identified problem.
Reporting to the senior manager for NOS is Employee Concerns, which investigates concerns identified through the Employee Concerns Programs to determine their validity and initiate corrective actions as appropriate. Employee Concerns also promotes the 9

CR3-QAPM Revision 0 Safety Conscious Work Environment (SCWE) Program and is sensitive to SCWE concerns during investigations performed.
The senior manager for NOS is also responsible for scheduling periodic independent audits of NOS through the Nuclear Industry Exchange Program (NIEP) to confirm QA program compliance and validate the effectiveness of the Independent Oversight Program.
: 5. Department Interfaces Departmental interfaces are identified in appropriate procedures. Quality related activities performed by departments other than Nuclear Generation or Operations Support are identified by and conducted in accordance with approved departmental interface agreements. The following are generic descriptions of those other corporate departments and the services they provide. These generic organizations are referred to, as appropriate, within this document; however, approved departmental interface agreements establish and define the applicability of the QAPM to the services they provide:
: a. Corporate Communications Corporate Communications provides support for the nuclear site's emergency response organization.
: b. Environmental Health and Safety Environmental, Health and Safety will provide environmental and laboratory support services.
: c. Nuclear Finance Nuclear Finance provides support for the nuclear sites in the areas of decommissioning, workforce planning and development.
: d. Customer Operations Customer Operations provides electrical distribution and switchyard engineering, as well as providing electrical maintenance and testing support.
: e. Information Technology Information Technology provides a variety of services and technical support to the Duke Energy Nuclear fleet and CR3 for information technology applications and systems such as equipment databases, applications, infrastructure, and plant process information systems. They are also responsible for the development and maintenance of selected information technology services and support, including electronic document management, some of which support QA related activities.
: f. Supply Chain Supply Chain provides procurement services, storage, inventory control, and receipt inspection/testing.
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: 6. Agents and Contractors Duke may contract various activities such as engineering, procurement, and construction. These contracts will identify QAPM requirements that are applicable to the contractors and their subcontractors, consistent with the requirements of Sections 6.0 and 9.0.
: 7. NOS Internal Audits is responsible for performing internal audit activities and provides objective oversight of plant performance relating to ITS items and activities.
NOS Internal Audits reports to the senior manager for NOS.
: 8. Fleet Quality Control (QC) is responsible for plant inspections and non-destructive examinations. Fleet QC reports to the senior manager for NOS.
: 9. Independent Management Assessments 4.0     Quality Program The Quality Program complies with the requirements of 10 CFR 50, Appendix B. This program requires that all persons performing quality activities associated with CR3 comply with the program. CR3 conducts or delegates the responsibility to conduct audits of the program activities.
The Quality Program takes into account the need for special controls, processes, tests, equipment, tools and skills to obtain the required quality.
Managers/Supervisors selecting personnel shall assure that the qualifications stated in the job description meet education and experience requirements identified in the job description. The exception is the Radiation Protection Manager, who shall meet or exceed the qualifications of Regulatory Guide 1.8, September 1975 (as clarified in Table 2).
An indoctrination and training program is provided for personnel performing quality activities to assure that they are knowledgeable of the Quality Program's procedures and requirements. The indoctrination and training program includes appropriate procedures and personnel records.
Additionally, personnel responsible for performing ITS activities are instructed as to the purpose, scope and implementation of the ITS manuals, instructions, and procedures.
Personnel performing safety-related activities are trained and qualified in the activity being performed.
Personnel who carry out Health Physics, or perform quality assurance functions, may report to the appropriate onsite manager; however, they shall have sufficient organizational freedom to ensure their ability to perform their assigned functions.
Quality activities such as inspection, examination and test are done with appropriate equipment and under suitable conditions.
CR3's organizational structure and the functional responsibility assignments assure that:
: 1. Attainment of program objectives is accomplished by those who have been assigned responsibility for performing work. Activities may include interim examinations, checks, and inspections of the work by the individual performing the work.
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: 2. Verification of conformance to established program requirements is accomplished by a qualified person who does not have responsibility for performing or directly supervising the work. The method and extent of such verification shall be commensurate with the importance of the activity to plant safety and reliability.
: 3. In structuring the organization and assigning responsibility, quality assurance should be recognized as an interdisciplinary function involving many organizational components and, therefore, should not be regarded as the sole domain of a single quality assurance group. Quality assurance encompasses many functions and activities and extends to various levels in all participating organizations, from the top executive to all workers whose activities may influence quality.
The CR3 Quality Program applies to operation and maintenance of CR3 ITS items and activities controlled by 10 CFR 72 and to the usage of transportation packages controlled by 10 CFR 71 and may be applied to other equipment and activities at management discretion (e.g., equipment important for SAFSTOR, Fire Service System, and Radioactive Waste System). Augmented Quality Assurance (non-safety-related with special quality/regulatory requirements) may be applied to programs such as Fire Protection, Radioactive Material Packages. The CR3 ITS items and activities are identified in the 72.212 report.
Changes to the Quality Program which result in more stringent requirements will be entered in appropriate implementing procedures within 90 days of the Quality Program change unless otherwise specified in the requirement to change the Quality Program or unless a longer period is evaluated and accepted by the senior manager for NOS. All other Quality Program changes will be reflected in appropriate implementing procedures at their next revision.
The CR3 Quality Program meets the requirements of 10 CFR 50 Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Processing Plants. Additional regulatory commitments are listed within Table 2. The NRC is to be notified of QAPM changes in accordance with 10 CFR 50.54(a)(3) or 10 CFR 50.54(a)(4).
For activities where quality considerations are subject to interpretation, the management responsible for the activity shall also be responsible for assuring that programmatic controls are applied. This in no way negates the need for clear management controls for ITS activities. NOS personnel evaluate and verify that controls are in place and effectively implemented through inspection and audit activities.
CR3's QAPM meets the Quality Assurance Requirements of 10 CFR 71, Subpart H, Quality Assurance for Packaging and Transportation of Radioactive Material, with respect to the procurement, maintenance, repair, and use of transportation packaging.
All other activities related to the packaging (i.e., design, fabrication, assembly, testing, and modification) are satisfied by obtaining certifications from packaging suppliers that these activities were conducted in accordance with an NRC-approved Quality Assurance Program. CR3's QAPM meets the Quality Assurance Requirements of 10 CFR 72, Subpart G, Quality Assurance for Independent Storage of Spent Nuclear Fuel and High-Level Radioactive Waste, for storage of spent fuel at power reactor sites under General License.
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CR3-QAPM Revision 0 5.0     Design Control Measures shall be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions. These measures shall include provisions to assure that appropriate quality standards are specified and included in design documents and that deviations from such standards are controlled. Measures shall also be established for the selection and review for suitability of application of materials, parts, equipment, and processes that are essential to the functions of the SSCs.
Measures shall be established for the identification and control of design interfaces and for coordination among participating design organizations. These measures shall include the establishment of procedures among participating design organizations for the review, approval, release, distribution, and revision of documents involving design interfaces.
The design control measures shall provide for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program.
The verifying or checking process shall be performed by individuals or groups other than those who performed the original design, but who may be from the same organization. Where a test program is used to verify the adequacy of a specific design feature in lieu of other verifying or checking processes, it shall include suitable qualifications testing of a prototype unit under the most adverse design conditions.
Design changes, including field changes, shall be subject to design control measures commensurate with those applied to the original design and be approved by the organization that performed the original design unless the applicant designates another responsible organization.
CR3 verifies that the 10 CFR Part 72 Certificate of Compliance (CoC) holder exercises Design Control for the storage mode and that the Part 71 packaging suppliers exercise design control for their packages.
Maintenance or modifications which may affect ITS SSCs are performed in a manner that ensures quality requirements, material specifications, and inspection requirements are met. Maintenance or modifications of ITS equipment are planned and performed in accordance with written procedures, documented instructions, or drawings appropriate to the circumstances which conform to applicable codes, standards, specifications, and criteria.
6.0     Procurement Document Control Procurement documents are reviewed by qualified personnel, prior to purchase, to assure that quality and technical requirements have been specified. Individuals reviewing these procurement documents are not involved with the other phases of the procurement activity. These reviews are performed and documented in accordance with approved written procedures.
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CR3-QAPM Revision 0 CR3's Quality Program contains provisions which require that:
: 1. Procedures are established which clearly delineate the sequence of actions to be accomplished in the preparation, review, approval and control of procurement documents and which identify those positions or groups responsible for performing those functions.
: 2. Review of and concurrence with the procurement documents are performed by qualified personnel to assure that the quality requirements are stated. This review is to determine that quality requirements can be inspected and controlled, that there is adequate acceptance or rejection criteria, and that the procurement document has been prepared in accordance with DEF Quality Program procedure requirements.
: 3. Documented evidence of the review and approval of procurement documents is provided and available for verification.
: 4. Procurement documents identify those 10 CFR 50 Appendix B requirements that must be complied with by the supplier's quality program. Details regarding evaluation and selection of suppliers are stated in written procedures.
: 5. Procurement documents contain or reference, as applicable, basic technical requirements such as regulatory requirements (e.g., 10 CFR Part 21) and design bases and identify the documentation to be prepared, maintained, submitted, and made available to DEF for review and/or approval. DEF's procurement documents include provisions for control of nonconformances.
: 6. Procurement documents contain the requirements for the retention, control, and maintenance of records as appropriate.
: 7. Procurement documents contain the right of access to vendor's facilities and records for source inspection and audit by DEF.
: 8. Changes and/or revisions to a procurement document are subject to review and approval requirements at least equivalent to those for the original document.
: 9. Procurement document control may be applied to Augmented Quality Assurance Programs (e.g., FP-Q, RW-Q, 10 CFR 71-Q, 10 CFR 72-Q, or Q Class B) as directed by DEF's procedural requirements.
Additional clarification for procurement control to ensure materials obtained by CR3 conform with the Part 72 CoC or the Part 71 package requirements may be found in Regulatory Guide 7.10 as committed to in Table 2.
7.0    Instructions, Procedures and Drawings CR3's Quality Program contains requirements to assure that each of the 18 criteria within 10 CFR 50, Appendix B are delineated, accomplished, and controlled in accordance with approved written procedures.
CR3's Quality Program contains provisions which require that instructions, procedures, or drawings include appropriate quantitative (such as dimensions, tolerances, and operating limits) or qualitative (such as workmanship samples) acceptance criteria for determining that quality activities have been satisfactorily accomplished.
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CR3-QAPM Revision 0 8.0     Document Control CR3 has a document control system for documents which prescribe activities affecting quality.
CR3's Quality Program contains provisions which require that:
: 1. Measures are established to review documents, such as instructions, procedures, and drawings (and changes thereto) prior to release to assure that the quality requirements are sufficiently, clearly, and accurately stated.
: 2. Changes to documents are reviewed and approved by the same organizations that performed the original review and approval unless delegated by the appropriate DEF organization to another qualified responsible organization.
: 3. The reviewing organization(s) has access to pertinent background information upon which to base its approval and has an adequate understanding of the requirements and intent of the original document.
: 4. Approved changes are promptly included with instructions, procedures, drawings, and other appropriate documents.
: 5. Obsolete or superseded documents are controlled to prevent their use.
: 6. Documents are available at the start of the work for which they are needed.
: 7. A method for identifying the current revision of instructions, procedures, and drawings is established and implemented. This information is updated and distributed as necessary to predetermined responsible personnel.
As a minimum under this criteria, the controlled documents include:
: 1. Design specifications.
: 2. Design, manufacturing, construction, and installation drawings.
: 3. Manufacturer inspection and testing instructions.
: 4. Procurement documents.
: 5. Maintenance, repair, and modification instructions.
: 6. Test surveillance instructions.
: 7. In-service inspection instructions.
9.0     Control of Purchased Material, Equipment and Services Vendor evaluation surveys to qualify potential suppliers in accordance with approved written procedures are conducted by CR3 or through qualified contractors. Suppliers quality assurance programs are reviewed and concurred with prior to implementation of activities except for commercial-grade calibration services. When purchasing commercial-grade calibration services, verification that the suppliers accreditation meets NEI 14-05A, Revision 0, Guidelines For The Use Of Accreditation In Lieu Of Commercial Grade Surveys For Procurement Of Laboratory Calibration And Test Services, shall be performed and documented.
DEF assures that quality requirements of the purchase document have been met, using source inspection, receipt inspection or document review, as appropriate.
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CR3-QAPM Revision 0 Control of purchased material, equipment and services may be applied to Augmented Quality Assurance Programs (e.g., FP-Q, RW-Q, 10 CFR 71-Q, 10 CFR 72-Q, or Q Class B) as directed by DEF's procedural requirements.
Additional clarification for procurement control to ensure materials obtained by CR3 conform with the Part 72 CoC or the Part 71 package requirements may be found in Regulatory Guide 7.10 as committed to in Table 2.
10.0 Identification and Control of Materials, Parts and Components CR3 has established measures for the identification and control of materials, parts, and components.
CR3's Quality Program contains provisions which require that:
: 1. Procedures are established which describe identification and control of material, parts, and components, including partially fabricated assemblies.
: 2. Identification requirements are determined during the initial planning stages (i.e., during generation of specification and design drawings).
: 3. Identification is specified to the extent that the item identified can be traced to the associated documentation, such as drawings, specifications, purchase orders, manufacturing and inspection documents, and physical or chemical mill test reports.
: 4. The degree of identification is specified on the design drawing or in referenced technical documents.
: 5. Measures are provided to assure that the location and method of identification do not affect the function or quality of the item being identified.
: 6. Measures are provided for the verification of correct identification of materials, parts, and components prior to release for manufacturing, shipping, construction and installation.
11.0 Control of Special Processes Participating organizations provide written procedures for performance of special processes such as welding, heat treating, chemical cleaning, coating/painting, and non-destructive testing and include the requirements for qualification of personnel performing the work.
CR3's Quality Program contains provisions which require that:
: 1. Measures are established to assure adequate performance and control of special processes such as welding, heat treating, chemical cleaning, coating/painting and non-destructive testing.
: 2. Measures are established to assure that procedures, equipment and personnel connected with special processes are qualified in accordance with the requirements of applicable codes, standards, and specifications.
: 3. Measures are established to assure that special processes are performed by qualified personnel in accordance with approved written procedures. These procedures provide for recording evidence of verification and, if applicable, inspection and process results.
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: 4. An active file is maintained on qualification records of all special process procedures and equipment, and personnel performing special processes.
: 5. Special process procedures and the credentials of qualified personnel are regularly reviewed to assure they are of the latest revision and that personnel qualifications have not expired.
12.0      Inspections Written procedures are required for the performance of inspection. Inspections are performed by qualified personnel other than those who performed or directly supervised the work being inspected.
CR3's Quality Program contains provisions which require that:
: 1. Inspection personnel are independent from the individual or group physically performing and directly supervising the activity being inspected.
: 2. Inspection procedures, instructions and/or checklists are provided which document the date performed, by whom and/or by what equipment, the type of observation, the results, the data collected and its acceptability.
: 3. Inspection procedures or instructions are available for use prior to performing the inspection operation.
: 4. Measures are provided for qualifying the inspectors and maintaining the current status of each inspector's qualifications.
: 5. Measures are established to assure that inspection equipment is within calibration prior to performing an inspection operation.
: 6. Measures are provided for monitoring processing methods, equipment, and personnel if inspection of processed material is impossible or disadvantageous. Inspection and process monitoring are provided when control is inadequate without both.
: 7. Specific hold points are indicated in appropriate documents for mandatory witnessing or inspection beyond which work shall not proceed without the consent of DEF's designated representative.
13.0 Test Control Required tests are performed in accordance with approved written procedures to assure compliance with design documents. Testing activities are conducted during the operational phase to verify the compliance of components to design requirements.
CR3's Quality Program contains provisions which require that:
: 1. A test program is established to assure that all testing required to demonstrate that the item will perform satisfactorily in service is identified, documented, and accomplished in accordance with approved written procedures.
: 2. The test program covers the required tests, including, where appropriate, prototype qualification tests, proof tests prior to installation, preoperational tests, and operational tests.
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: 3. Written test procedures are prepared which incorporate or reference the requirements and acceptance limits contained in applicable design and procurement documents.
: 4. The written test procedures include, as appropriate, instructions for test method and identification of test prerequisites such as:
: a. calibrated instrumentation;
: b. adequate and appropriate equipment;
: c. trained, qualified, licensed and/or certified personnel;
: d. preparation, condition, and completeness of item to be tested; and
: e. suitable and controlled environmental conditions.
: 5. Test results are documented and evaluated to assure that test requirements have been satisfied.
14.0 Control of Measurement and Test Equipment DEF has established and implemented appropriate test and calibration procedures for test devices used to verify the acceptability of items within the Quality Program.
Calibration records and controls are provided for measurement and test equipment.
CR3's Quality Program contains provisions which require that:
: 1. Procedures are established which describe the calibration technique, calibration frequency, maintenance and control of measuring and test instruments, tools, gauges, fixtures, reference standards, transfer standards, and non-destructive test equipment to be used in the measurement, inspection, and monitoring of ITS SSCs.
: 2. Measurement and test equipment is uniquely identified and has traceability to the calibration test data.
: 3. Measurement and test instruments are calibrated and maintained at specified intervals, based on the required accuracy, purpose, the degree of usage, stability characteristics, and other conditions affecting the measurement.
: 4. Measurement and test equipment is calibrated on or before the designated due date or before use.
: 5. When measurement and test equipment is found to be out of calibration, an investigation is conducted and documented to determine the validity of previous inspections performed and the acceptability of those items previously inspected.
: 6. Calibrating instruments have known valid relationships to a nationally recognized standard. If no national standard exists, the basis for calibration is documented.
: 7. Facilities used for calibrating sensitive or close tolerance measurement and test equipment provide an environment that is sufficiently controlled to allow the measuring device to be evaluated and calibrated to its required accuracy.
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CR3-QAPM Revision 0 15.0 Handling, Storage and Shipping Approved written procedures and sound storage principles are used for material handling, storage, and shipping activities for plant spare parts and operating supplies.
16.0 Inspection, Tests and Operating Status Approved written procedures are used to assure the proper marking of equipment denoting its status.
CR3's Quality Program contains provisions which require that:
: 1. Measures are established and documented to identify the inspection, test, and operation status of SSCs, which provide means for assuring that required inspections and tests performed are known throughout manufacturing, installation, and operation.
: 2. Measures are established to control the use of inspection and status indicators, including the authority for application and removal of tags, markings, and labels.
: 3. Measures to preclude bypassing of required inspections, tests, and other critical operations are provided through approved written procedures.
: 4. The status of nonconforming, inoperative, or malfunctioning SSCs is clearly identified to prevent use.
17.0 Nonconforming Material, Parts, or Components Written requirements are followed by persons performing quality activities, including contractors, to identify, document, segregate, disposition and report any nonconformance, deviation or other condition adversely affecting quality.
CR3's Quality Program contains provisions which require that:
: 1. Measures and procedures are established to control the identification, documentation, segregation, review, disposition, and notification of the affected organization of nonconformances.
: 2. Documentation is provided which clearly identifies the nonconforming item, describes the nonconformance and disposition of the nonconformance, inspection requirements, and includes signature approval of the disposition.
: 3. Measures are established and documented defining the responsibility and authority for determining the disposition of nonconforming items and approving the disposition.
: 4. Nonconforming items are segregated from acceptable items (where feasible) and uniquely identified as nonconforming until properly dispositioned for use.
: 5. Acceptability of "rework" or "repair" of materials, parts, SSCs are verified by re-inspection and/or testing of the item in accordance with approved written procedures.
: 6. Nonconforming items which are dispositioned "use as is" or "repair" are formally controlled through approved procedures or design changes.
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: 7. Nonconformance reports are made part of the quality assurance records.
18.0 Corrective Action Approved written procedures are followed by persons performing quality activities, including contractors, to assure that corrective action is taken to preclude the recurrence of nonconformances, deviations or other discrepancies adversely affecting quality.
CR3's Quality Program contains provisions which require that:
: 1. Conditions adverse to quality, such as failures, deficiencies, deviations, defective material and equipment, and nonconformances are promptly identified and corrected.
: 2. Evaluation of nonconformance and determination of the need for corrective actions are in accordance with approved written procedures.
: 3. Measures are established to determine the cause of the nonconformance and institute corrective action to preclude the recurrence of those significant conditions adverse to quality.
: 4. Measures are established to follow up on corrective actions to assure proper implementation and close out of the corrective action documentation.
: 5. Measures are established to document and report to appropriate levels of management significant conditions adverse to quality, cause of the conditions, and corrective action taken.
19.0 Quality Assurance Records A system has been established and implemented for the collection, storage, and maintenance of quality assurance records as required. Quality assurance records transmitted to the quality files are done so in accordance with approved written procedures. Quality assurance records do not include vital records. Vital records are records maintained to meet regulatory or other commitments and are not required to meet the collection, storage and maintenance requirements of Quality Assurance Records.
CR3's Quality Program contains provisions which require that:
: 1. Quality assurance records are of two categories, lifetime and nonpermanent.
Nonpermanent records are required to show evidence that an activity was performed in accordance with applicable requirements but need not be retained for the life of the plant. Lifetime records are required to be maintained for the life of the plant while the particular item is installed in the plant or stored for future use.
: 2. Quality assurance records are those records that furnish documentary evidence of the quality of items and of activities affecting quality. A document is considered a quality assurance record when the document has been completed.
These records include the results of reviews, inspections, examinations, tests, audits, assessments, monitoring of work performance and material analysis, the qualification of personnel, procedures, and equipment, training records, 20

CR3-QAPM Revision 0 design drawings and subsequent modifications, specification reports, procurement       documents,      calibration     procedures    and    reports, nonconformance and corrective action reports, and other records required by documents. The records are identifiable and retrievable.
: 3. The inspection and test records contain the following:
: a. Description of the types of operation.
: b. Evidence of completion and/or verification of manufacturing inspection or test operation.
: 4. Records are stamped, dated, initialed, signed, or otherwise authenticated by authorized personnel. Measures are established to control the use of electronic signatures that prevent falsification or alteration of electronic records and provide a means to authenticate electronic signatures.
: 5. Storage facilities are constructed, located, and secured to prevent destruction and minimize deterioration or loss of records. Quality assurance records are maintained either in a vault for single copy records or as duplicate records in remote locations.
On November 30, 2016, the NRC granted an exemption request related to Record Keeping Requirements. This exemption permits the elimination of requirements to maintain records that are no longer necessary due to the permanently shut down and decommissioning status of CR3. Specifically, those QA Records are no longer required to be retained when: 1) the CR3 licensing basis requirements previously applicable to the nuclear power unit and associated SSCs are no longer effective (i.e., removed from the FSAR or Technical Specifications by the appropriate change mechanisms); or 2) for SSCs associated with safe storage of fuel in the spent fuel pools (SFPs) where spent fuel has been completely removed from the SFPs, and the associated licensing bases are no longer effective.
With regard to operating phase records only, CR3 has established a listing of those types of operating phase records that have been assigned specific retention periods.
These retention periods apply to records not covered by the above cited exemption from record keeping.
The following records shall be retained for at least five years:
: a. Records and logs of facility operation covering time intervals at each power level.
: b. Records and logs of principal maintenance activities, inspections, repair and replacement of principal items of equipment related to nuclear safety.
: c. All Reportable Events submitted to the Commission.
: d. Records of surveillance activities, inspections and calibrations required by PDTS.
: e. Records of reactor tests and experiments.
: f. Records of changes made to Operating Procedures.
: g. Records of radioactive shipments.
: h. Records of sealed source and fission detector leak tests and results.
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: i. Records of annual physical inventory of all sealed source material of record.
The following records shall be retained for the duration of the Facility Operating License:
: a. Records and drawing changes reflecting facility design modifications made to systems and equipment described in the DSAR.
: b. Records of new and irradiated fuel inventory, fuel transfers and assembly burnup histories.
: c. Records of facility radiation and contamination surveys.
: d. Records of radiation exposure for all individuals entering radiation control areas.
: e. Records of gaseous and liquid radioactive material released to the environs.
: f. Records of transient or operational cycles for facility components.
: g. Records of training and qualification for current members of the plant staff.
: h. Records of inservice inspection performed pursuant to PDTS.
: i. Records of Quality Assurance activities required by the Quality Program Description.
: j. Records of reviews performed for changes made to procedures or equipment or reviews of tests and experiments pursuant to 10 CFR 50.59.
: k. Records of meetings of the Plant Nuclear Safety Committee and the Independent Management Assessments.
: l. Records for Environmental Qualification.
: m. Records of analytical results required by the Operational Radiological Environmental Monitoring Program.
: n. Records of reviews performed for changes made to the ODCM.
: o. Records of reviews and changes to the Process Control Program. This documentation shall contain 1) sufficient information to support the change together with the appropriate analyses or evaluations justifying the change(s), and 2) a determination that the change will maintain the overall conformance of the solidified waste product to existing requirements of Federal, State, or other applicable regulations.
The CR3 program for storage of records on microfilm, dual storage or in electronic format meets the preservation requirement for the retention of QA Records.
For management of electronic records, the appropriate controls on quality are summarized as follows:
: a. The Electronic Records Management (eRM) system does not allow deletion or modification of records. (NOTE: Authorized deletion of records per the Record Retention Rules is controlled.)
: b. The eRM system provides redundancy (i.e., system backup, dual storage, etc.).
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: c. The legibility of each record is verified prior to acceptance into the eRM system.
: d. The media used by the eRM system is maintained to ensure the records are acceptably copied onto a new media before the manufacturer's certified useful life of the media is exceeded. This includes verification of the records so copied.
: e. Periodic random inspections of records are performed to verify that there has been no degradation of record quality.
: f. If the eRM system in use is to be replaced by a new system, the records stored on the old system are acceptably converted into the new system before the old system is taken out of service. This includes verification of the records so copied.
To implement those controls, CR3 uses the following:
           -  NIRMA TG 11-2011, "Authentication of Records and Media"
           -  NIRMA TG 15-2011, "Management of Electronic Records"
           -  NIRMA TG 16-2011, "Software Quality Assurance Documentation and Records"
           -  NIRMA TG 21-2011, "Required Records Protection, Disaster Recovery and Business Continuation" 20.0 Audits and Independent Reviews INDEPENDENT AUDIT The functions and activities affecting the nuclear programs at CR3 are independently audited. NOS is responsible for auditing activities that are performed by or for the Crystal River Plant. The audits are performed using the Nuclear Operating Fleet 10 CFR 50 Appendix B compliant audit program.
ORGANIZATION Personnel performing independent audit activities have no direct responsibilities in the areas being audited.
Selection of audit personnel is based on experience and/or training which establishes that their qualifications are commensurate with the complexity or special nature of the area being audited. The process for qualification of personnel to perform and lead audits is established in procedures.
Personnel performing audits shall have access to records, procedures, and personnel to gather data.
AUDIT PROCESS The independent audit process includes gathering data, analyzing data, focusing on selected issues and identifying deficiencies. The results of independent audits are communicated to management in a manner that causes action to correct deficiencies and develop action to prevent recurrence. In addition, this process should evaluate corrective measures adopted to eliminate the deficiencies identified.
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CR3-QAPM Revision 0 Planning activities identify the organizations to be evaluated, the characteristics to be focused on during the independent audit, and the applicable acceptance criteria.
Independent audit activities are selected with flexibility based on various factors.
Preparation activities may include a review of performance data, relevant documentation, previous audit data, industry experience, team member experience, and management input. These activities enable the team to focus on issues which may impact safety and reliability when analyzing data.
Audits are scheduled on the basis of the status and safety importance of the activities or processes being performed. The schedule is flexible and dynamic to allow audits to be changed depending on plant conditions, events, or issues raised by senior management.
NUCLEAR OVERSIGHT AUDIT PROGRAM Audits of facility activities shall be performed by NOS. Audits will be scheduled based on plant performance and importance to safety, but with a frequency not to exceed twenty-four months. These audits shall encompass:
* The conformance of facility operation to provisions contained within the PDTS and applicable license conditions.
* The performance, training and qualifications of the staff supporting the Crystal River Nuclear Plant.
* The results of actions taken to correct deficiencies in facility equipment, structures, systems or method of operation that affect ITS items and activities.
* The performance of activities required by the Quality Assurance Program to meet the criteria of 10 CFR 50 Appendix B, for activities performed by the staff supporting the Crystal River Nuclear Plant.
* The Radiological Environmental Monitoring Program, and the results thereof.
* The performance of activities required by the Quality Assurance Program for effluent and environment monitoring.
* The ODCM and implementing procedures.
* The Process Control Program and implementing procedures for processing and packaging of radioactive wastes.
* Audits of activities prescribed by the Code of Federal Regulations will be performed at the frequencies prescribed by the applicable regulation. These audits shall encompass:
o Emergency Preparedness [per 10 CFR 50.54(t)]
o Security [per 10 CFR 50.54(p)]
Results of audits will be provided to NOS management for review. A periodic briefing of NOS activities, along with potential issues and recommendations, shall be presented to the Senior Vice President, Operations Support.
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CR3-QAPM Revision 0 Follow-up is accomplished to assure that corrective action is taken as a result of the audits and that deficient areas are reaudited, when necessary, to verify implementation of adequate corrective actions.
INDEPENDENT REVIEW PROGRAM The Independent Safety Reviewer provides the independent review of the following:
* Proposed changes to the facility as described in the DSAR. This review is to confirm that the change does not adversely affect safety and if a PDTS change or NRC review is required.
* Proposed changes to procedures as described in the DSAR and tests or experiments not described in the DSAR. This review is to confirm that the change does not adversely affect safety and if a PDTS change or NRC review is required.
* Proposed PDTS changes and license amendments, except in those cases where the change is identical to a previously reviewed proposed change.
* Licensee Event Reports that are required to be made to the NRC. This review includes results of any investigations made and recommendations resulting from such investigations to prevent or reduce the probability of recurrence of the event.
* Any other matter related to ITS items requested by the General Manager Decommissioning or referred for review by other organizations.
See DSAR Section 5.8.1.
21.0 Program Commitment During the SAFSTOR and Decommissioning phase, DEF will comply with the Regulatory Positions of the Regulatory Guides listed in Table 2, as clarified in that Table. DEF considers and will refer to such Regulatory Positions as requirements.
When Regulatory Guides are superseded by an approved revision, that revision will not be implemented unless the DSAR is modified accordingly.
22.0 Quality Assurance Staff Persons performing quality assurance functions, as defined in 10 CFR 50 Appendix B, conduct reviews and audits of departments, suppliers and contractors that perform ITS functions in connection with CR3. These reviews and audits are performed in accordance with approved written procedures and in compliance with approved minimum requirements for audit frequency. Persons performing quality assurance functions are authorized to identify quality problems and may recommend to management that work be stopped under totally unacceptable conditions. Such persons have the organizational freedom to effectively perform the quality assurance functions.
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CR3-QAPM Revision 0 23.0 Glossary Of Terms Terms used in the CR3 Quality Program are defined below.
: 1. Quality Activity The term "quality activity" is a general term used to describe activities within the total Quality Program. The purpose of using the term "quality activity" is to reserve the words "control" and "assurance" for those specific functions of the Quality Program defined as "quality control" and "quality assurance."
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CR3-QAPM Revision 0 Table 1 - Crystal River Unit 3 In-Plant Quality Program Functions Quality Assurance         Quality Control            Work Owner's surveillance       Plant staff personnel    The work starts with and management             provide the Quality      commercial operation evaluation of the          Control function.        and continues operating staff            Included as a            throughout the life of activities through         verification of          the plant.
audit of the               compliance is the implementation of the      function identified as commitments.               the Compliance Section which assures compliance by the Quality Control activities through audits and witness action.
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CR3-QAPM Revision 0 Table 2 - Crystal River Unit 3 Quality Program Commitments This table presents the Regulatory Guides and ANSI Standards endorsed by CR3 as part of its Quality Program.
In each of the ANSI Standards, other documents (i.e., other Standards, codes, regulations, tables, or appendices) required to be included as part of the Standard are either referenced or described in a special section of the Standard. The specific applicability or acceptability of these referenced Standards, codes, regulations, tables, or appendices is either covered in other specific areas in the CR3 Quality Program Description, including this Table, or such documents are not considered as Quality Program requirements, although they may be used as guidance. Whenever a standard endorsed in Table 2 invokes ANSI N45.2, CR3 shall interpret the statement to mean a QA Program which meets the requirements of 10 CFR 50, Appendix B.
When Sections of Standards are referenced within a clarification, it is understood that CR3 shall comply with the referenced Sections as clarified.
NRC Regulatory Guide 1.8 - "Personnel Selection and Training" (Revision 1, 9/75) - Endorses ANSI N18.1-1971.
With the Certification of Permanent Cessation of Power Operations, DEF has permanently ceased power operations of CR3. CR3 follows this Regulatory Guide and associated Standard for the remaining ITS activities associated with SAFSTOR, decommissioning, and the ISFSI with the following clarifications:
: 1) CR3 often uses additional non-CR3 employees and contract personnel to augment the facility staff. These personnel may or may not report to the General Manager, Decommissioning, Crystal River Nuclear Plant. When used to perform safety-related activities, these personnel shall meet the education, training and experience requirements of ANSI N18.1-1971 for equivalent positions or else they shall meet the requirements for certification as inspection, examination or testing personnel as set forth in CR3's commitment to ANSI N45.2.6-1978.
: 2) Operators licenses, which are issued by the NRC and discussed in ANSI N18.1 1971, are no longer required based upon the NRC approval of TS Amendment 244.
With regard to section 4 of ANSI N18.1-1971, titled Qualifications: Selection and qualification of personnel is based on the established requirements of that position through CR3 hiring and Nuclear Operations policies, thereby meeting the intent of ANSI N18.1. The hiring policies are governed by the CR3 Human Resources Department.
With regard to paragraph 5.5 of ANSI N18.1-1971, titled Retraining and Replacement Training: CR3s retraining and replacement training program for the facility staff shall be maintained under the direction of the General Manager, Decommissioning, Crystal River Nuclear Plant. Subjects addressed in Section 5.5 of ANSI N18.1-1971 will be included as appropriate per the System Approach to Training in accordance with 10 CFR 50.120.
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: 3) With regard to paragraph 4.2.1 of ANSI N18.1-1971, titled Plant Manager, and paragraph 4.2.2 of ANSI N18.1-1971, titled Operations Manager: Operators licenses, which are issued by the NRC and discussed in ANSI N18.1-1971, are no longer required based upon the NRC approval of TS Amendment 244. The CR3 management structure does not require any positions to attend equivalent training.
NRC Regulatory Guide 7.10, Revision 3 (6/15), "Establishing Quality Assurance Programs for Packaging Used in the Transportation of Radioactive Material With the Certification of Permanent Cessation of Power Operations, DEF has permanently ceased power operations of CR3. CR3 uses this Regulatory Guide for QA Program content guidance as it applies to users of transportation packaging. This guidance is also considered for the use of Part 72 storage modules under a General License.
NUREG/CR-6407, "Classification of Transportation Packaging and Dry Fuel Storage System Components According to Important to Safety (2/96)"
With the Certification of Permanent Cessation of Power Operations, DEF has permanently ceased power operations of CR3. CR3 follows this NUREG for the classification of transportation packaging and dry fuel storage system components according to ITS.
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CR3-QAPM Revision 0 Figure 1 - CR3 Nuclear Organization 30}}