3F0514-08, NRC Commitment Change Report - May 2014

NRC Commitment Change Report - May 2014

ML14150A431
Person / Time
Site:	Crystal River
Issue date:	05/29/2014
From:	Wunderly B Duke Energy Florida
To:	Document Control Desk, Office of Nuclear Reactor Regulation
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ML14150A427	List: 3F0514-08, NRC Commitment Change Report - May 2014
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3F0514-08
Download: ML14150A431 (44)

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Crystal River Nuclear Plant 15760 W. Power Line Street ENERGY, Crystal River, FL 34428 Docket 50-302 Operating License No. DPR-72 SECURITY-RELATED INFORMATION Attachment 2 to this letter contains SECURITY-RELATED INFORMATION - WITHHOLD UNDER 10 CFR 2.390. Upon separation of Attachment 2, this letter is decontrolled.

Ref: 10CFR50.9 May 29, 2014 3F0514-08 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001

Subject:

Crystal River Unit 3 - NRC Commitment Change Report - May 2014

Reference:

CR-3 to NRC letter, "Crystal River Unit 3 - Certification of Permanent Cessation of Power Operations and that Fuel Has Been Permanently Removed from the Reactor," dated February 20, 2013. (ADAMS Accession No. ML13056A005)

Dear Sir:

The purpose of this letter is to provide notification of inactivations or modifications to regulatory commitments contained in previously docketed correspondence from Duke Energy Florida, Inc.,

to the NRC. The attached report contains the Crystal River Unit 3 (CR-3) Nuclear Operations Commitment System (NOCS) reference numbers, source of the original commitment, statement of the original commitment, statement of commitment inactivation or modification, if revised, and justification for the inactivation or modification. This report is being submitted in accordance with Nuclear Energy Institute (NEI) document NEI 99-04, Revision 0, "Guidelines for Managing NRC Commitment Changes," dated July 1999.

Of the one hundred (100) CR-3 regulatory commitments that were modified or inactivated between January 5, 2012 and January 5, 2014, forty-nine (49) modified or inactivated regulatory commitments meet the NEI 99-04 criteria for NRC notification. Attachment 1 contains the NRC Commitment Change Report - May 2014. Attachment 2 contains the SECURITY-RELATED INFORMATION NRC Commitment Change Report - May 2014.

Several of the regulatory commitments were modified or inactivated prior to the CR-3 10 CFR 50.82(a) certification related to decommissioning and may need to be revisited as a result of the decommissioning decision referenced above.

No new regulatory commitments are made in this letter.

If you have any questions regarding this submittal, please contact Mr. Dan Westcott, Manager, Nuclear Regulatory Affairs, at (352) 563-4796.

SECURITY-RELATED INFORMATION A 00(

Attachment 2 to this letter contains SECURITY-RELATED INFORMATION - WITHHOLD /

UNDER 10 CFR 2.390. Upon separation of Attachment 2, this letter is decontrolled. aa__

U. S. Nuclear Regulatory Commission Page 2 of 2 3F0514-08 SECURITY-RELATED INFORMATION Sincerely, Blair Wunderly Plant Manager Crystal River Nuclear Plant BW/sam NRC Commitment Change Report - May 2014 NRC Commitment Change Report - May 2014 SECURITY-RELATED INFORMATION xc: Regional Administrator, Region I NRR Project Manager SECURITY-RELATED INFORMATION to this letter contains SECURITY-RELATED INFORMATION - WITHHOLD UNDER 10 CFR 2.390. Upon separation of Attachment 2, this letter is decontrolled.

DUKE ENERGY FLORIDA, INC.

CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302 / LICENSE NUMBER DPR-72 ATTACHMENT 1 NRC COMMITMENT CHANGE REPORT - MAY 2014

U. S. Nuclear Regulatory Commission Attachment 1 3F0514-08 Page 1 of 41 Nuclear Operations Commitment System (NOCS) Number: 5945 Source Document:

Crystal River Unit 3 (CR-3) to NRC letter, 3F0785-27, dated July 23, 1985.

Original Commitment:

In the future, prospective Reactor Operator [RO] and Senior Reactor Operator [SRO]

candidates will be trained per the criteria in heat transfer, fluid flow, and thermodynamics in some combination of curricula of the Replacement Operator Training Program and the Non-Licensed Operator Training Program, which precedes it.

Reactor Operator and Senior Reactor Operator upgrade candidates will have received the required training from the combined curricula of both programs by the time they are certified for the NRC License Examination.

Instant SRO candidates will have received equivalent training based upon review of applicable class/training transcripts, additional training by participation in the Replacement Operator Training Program, and supplemental courses as necessary.

The commitment revision as stated in the previous three paragraphs is evaluated to be an improvement to both programs. The revision continues to meet the requirement of TMI [Three Mile Island] Action I.A.2.1.4 for RO and SRO training, and the SER [Safety Evaluation Report]

transmitted April 6, 1983. Current job and task analysis indicates the need for heavy heat transfer, fluid flow and thermodynamics training in the Non-Licensed Operator Training Program. The Replacement Operator Training Program will review and supplement this initial training as necessary.

The Non-Licensed Operator Training Program requires the same standards of performance and uses the same grading criteria (80/70 criteria) as the Replacement Operator Training Program.

Modify/Inactivate Commitment:

INACTIVATE: This commitment is based on the issuance of NUREG-0737, "Clarification of TMI Action Plan Requirements." At the time, the nuclear industry had no requirement to train on the items of Heat Transfer, Fluid Flow, and/or Thermodynamics. However, since the issuance of NUREG-0737, several requirements have been put into effect that now mandate this training.

Specifically, those requirements now exist in 10 CFR 55, Section 41 (a) and (b) and Section 43 (b). Additionally, 10 CFR 50.120 calls for the implementation of a Systematic Approach to Training (SAT). Per the SAT process, the training necessary for a nuclear plant operator is analyzed for the need to train.

The knowledge requirements for a person possessing, or training for, an RO or SRO license is delineated in NUREG-1 122, "Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Pressurized Water Reactors." The actual standards for testing of persons applying for a license, or testing on a continual basis, are contained in NUREG-1021, "Operator Licensing Examination Standards for Power Reactors." Institute of Nuclear Power Operations (INPO) document ACAD 10-001, "Guidelines for Initial Training and Qualification of Licensed Operators," Section 3.1.8, provides a complete listing of the training requirements associated with these topics. These documents specifically call out the need to train on Heat Transfer, Fluid Flow, and Thermodynamics.

U. S. Nuclear Regulatory Commission Attachment 1 3F0514-08 Page 2 of 41 Justification for Change:

The changes proposed to this commitment are in paper form only. The actions recommended by the source document are not being deleted, but are being brought up-to-date to the requirements of the Code of Federal Regulations. The actions of this commitment are now covered under the following documents: 10 CFR 50.120, 10 CFR 55.41, 10 CFR 55.43, NUREG-1021, NUREG-1122 and ACAD 10-001. Therefore, this change cannot negatively impact the ability of a System, Structure, or Component (SSC) to perform its safety function or negatively impact the ability of plant personnel to ensure a SSC is capable of performing its intended safety function.

U. S. Nuclear Regulatory Commission Attachment 1 3F0514-08 Page 3 of 41 Nuclear Operations Commitment System (NOCS) Number: 8664 Source Document:

CR-3 to NRC letter, 3F0580-02, dated May 2, 1980.

Original Commitment:

From Enclosure A #4: The NRC issued a Confirmatory Order for CR-3 on 4-14-80. The Order commitments are listed in Table A-4 and discussed in Enclosure B.

Table A-4 & Table A-6 cross reference: Actions which will allow the operator to cope with various combinations of loss of instrumentation and control functions. This includes training to assure positive and safe manual response by the operator in the event that competent instruments are unavailable.

From Enclosure B Item 16: The Training Department at CR-3 will work with engineering to develop a course to be presented to all licensed personnel & technicians. The course will include as a minimum:

- A review of NNI [non-nuclear instrumentation]/ICS[integrated control system], including all proposed modifications.

- A review of what indications are available to the operator during system upsets and NNI/ICS power losses.

- A review of control system interactions caused by NNI/ICS power losses.

- A review of Emergency Procedures & Abnormal Procedures necessary to shut down the plant with emphasis on how to regain manual control during NNI/ICS power losses.

- A review of all NNI/ICS event trees developed by the Nuclear Safety Task Force.

From Enclosure B Item 17: The Training Department at CR-3 will work with engineering to develop a course to be presented to all licensed operators. This course will include as a minimum:

- Training on all mods that result from the 2-26-80 incident.

- Training on how mods will affect plant response & control.

- Training on all procedure changes that occur as a result of these mods.

- The operators will also be trained on the sequence of events, ICS response to failed NNI, and how lessons learned from TMI-2 affected the transient.

From Enclosure B Item 27: The operator can isolate letdown from the control room by closure of makeup system valve MUV-49 or valves MUV-40 and MUV-41. These valves are the system containment isolation valves and are powered by Engineered Safeguard power which is independent of NNI & ICS power supplies. Operators will be advised by training and procedure that these valves are available to isolate letdown in the event of a loss of NNI or ICS power supplies.

Modify/Inactivate Commitment:

INACTIVATE: The NRC issued a Confirmatory Order for CR-3 on April 14, 1980. This commitment was as a result of the partial loss of non-nuclear instrumentation (NNI) power at CR-3 and was contained in the response to the Confirmatory Order. The actions to be taken by the commitment were. completed, the information was provided to the Operations, Engineering, and Maintenance personnel, and the courses were developed and/or modified. Actions to address future training are now governed by 10 CFR 50.120, requiring training via the

U. S. Nuclear Regulatory Commission Attachment 1 3F0514-08 Page 4 of 41 Systematic Approach to Training (SAT), along with NUREG-1 021, NUREG-1 122 and ACAD 10-001.

The knowledge requirements for a person possessing, or training for, an RO or SRO license is delineated in NUREG-1 122. The actual standards for testing of persons applying for a license, or testing on a continual basis, are contained in NUREG-1021. INPO document, ACAD 10-001, Section 3.1.5 and 3.3 gives a complete listing of the training requirements associated with these topics. These documents specifically call out the need to train on instrumentation, both theories of operation and specific failure modes that could result in erroneous operation.

Justification for Change:

The changes proposed to this commitment are in paper form only. The actions recommended by the source document are not being deleted, but are being controlled by an up-to-date document: 10 CFR 50.120. The actions of this commitment are now covered under other documents, which include 10 CFR 50.120, 10 CFR 55.41, 10 CFR 55.43, NUREG-1021, NUREG-1 122 and ACAD 10-001. Therefore, this change cannot negatively impact the ability of a SSC to perform its safety function or negatively impact the ability of plant personnel to ensure a SSC is capable of performing its intended safety function.

U. S. Nuclear Regulatory Commission Attachment 1 3F0514-08 Page 5 of 41 Nuclear Operations Commitment System (NOCS) Number: 62732 Source Document:

CR-3 to NRC letter, 3F1297-24, dated December 16, 1997.

Original Commitment: page 2 of 6 (NOTES 27477/Memo NED98-0071), response to NRC audit question 4: FPC [Florida Power Corporation] procedure AP-961, "Earthquake," is being revised to provide guidance to operators on how to cope with relay chatter subsequent to a seismic event and identification of bad actor relays.

Modify/Inactivate Commitment:

INACTIVATE: This commitment was made by CR-3 to respond to Generic Letter 87-02, "Verification of Seismic Adequacy of Mechanical and Electrical Equipment in Operating Reactors, Unresolved Safety Issue (USI) A-46." The Generic Letter was issued to implement the USI A-46 resolution which concluded that the seismic adequacy of certain equipment in operating nuclear power plants should be reviewed against seismic criteria not in use when these plants were licensed. The criteria for evaluating the concerns for USI A-46 was equipment associated with safe shutdown. The term "safe shutdown" is defined as bringing the plant to, and maintaining it in, a hot shutdown condition during the first 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> following a safe-shutdown earthquake (SSE).

Justification for Change:

The criterion for evaluating the concerns for USI A-46 was to apply the specific concern to plant equipment associated with safe shutdown. The term "safe shutdown" is defined as bringing the plant to, and maintaining it in, a hot shutdown condition during the first 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> following an SSE. CR-3 is no longer authorized to operate the reactor or emplace or retain fuel in the reactor vessel. By permanently removing fuel from the reactor, CR-3 is already in a safe shutdown condition and is not an operating reactor. Therefore, this change cannot negatively impact the ability of a SSC to perform its safety function or negatively impact the ability of plant personnel to ensure a SSC is capable of performing its intended safety function.

U. S. Nuclear Regulatory Commission Attachment 1 3F0514-08 Page 6 of 41 Nuclear Operations Commitment System (NOCS) Number: 926 Source Document:

CR-3 to NRC letter, 3F0485-07, dated April 9, 1985.

Original Commitment:

FPC administrative controls implementing overtime limits include:

A. Scheduled work is limited to the following maximum work hours (except during extended periods of shutdown for refueling, major maintenance, or major plant modifications):

1. An individual is not permitted to work more than 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> straight (excluding shift turnover time).

2. An individual is not permitted to work more than 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> in any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period, nor more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in any 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> period, nor more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in any seven (7) day period (all excluding shift turnover time).

3. A break of at least 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> is allowed between work periods (including shift turnover time).

B. In the event that special circumstances arise that require deviation from the above, such deviations are authorized at the nuclear Plant Manager level (or his designee) or above with appropriate documentation of the cause and how the following guidelines are met:

1. If personnel are required to work in excess of 16 continuous hours, their work is carefully selected to prevent unsafe operation of the plant.

2. It is preferred that personnel who are assigned to work in excess of 16 consecutive hours not be assigned to tasks which affect core reactivity.

C. The use of overtime is considered on an individual basis and not for the entire staff on a shift.

Modify/Inactivate Commitment:

INACTIVATE: This commitment was made in the CR-3 response to Generic Letter 82-12, "Nuclear Power Plant Staff Working Hours." Generic Letter 82-12 has been superseded by the regulatory requirements in 10 CFR 26 Subpart I, which constitute an obligation. The regulatory obligation is implemented by ADM-NGGC-0206, "Managing Fatigue and Working Hour Limits."

Justification for Change:

The changes proposed to this commitment are in paper form only. The actions recommended by the source document are not being deleted, but are being brought up-to-date to the requirements of the 10 CFR 26 Subpart I. Therefore, this change cannot negatively impact the ability of a SSC to perform its safety function or negatively impact the ability of plant personnel to ensure a SSC is capable of performing its intended safety function.

U. S. Nuclear Regulatory Commission Attachment 1 3F0514-08 Page 7 of 41 Nuclear Operations Commitment System (NOCS) Number: 932 Source Document:

CR-3 to NRC letter, 3F0485-07, dated April 9, 1985.

Original Commitment:

The administrative procedure controlling overtime limits is subjected to various types and levels of reviews to assure procedure changes are appropriate. These include:

1) Procedure changes are evaluated per 10 CFR 50.59 to assure the change will not lead to an unreviewed safety question.

2) The 50.59 evaluation and procedure change are then subjected to an interdepartmental review and plant review committee review and approval.

3) Proposed procedure revisions are compared to commitments to assure commitments continue to be satisfied.

4) Procedure changes and implementation are subject to periodic internal audits.

Modify/Inactivate Commitment:

INACTIVATE: This commitment was made in response to NUREG-0737, "Clarification of TMI Action Plan Requirements," Item A 1.3.1, and has been superseded by the regulatory requirements in 10 CFR 26 Subpart I, which constitute an obligation. The regulatory obligation is implemented by ADM-NGGC-0206.

Justification for Change:

The changes proposed to this commitment are in paper form only. The actions recommended by the source document are not being deleted, but are being brought up-to-date to the requirements of the 10 CFR 26 Subpart I. Therefore, this change cannot negatively impact the ability of a SSC to perform its safety function or negatively impact the ability of plant personnel to ensure a SSC is capable of performing its intended safety function.

U. S. Nuclear Regulatory Commission Attachment 1 3F0514-08 Page 8 of 41 Nuclear Operations Commitment System (NOCS) Number: 1036 Source Document:

CR-3 to NRC letter, 3F1280-08, dated December 15, 1980.

Original Commitment:

FPC commits to the limitation of overtime for nuclear power plant operators.

Modify/Inactivate Commitment:

INACTIVATE: This commitment was made in response to NUREG-0737, "Clarification of TMI Action Plan Requirements," and has been superseded by the regulatory requirements in 10 CFR 26 Subpart I, which constitute an obligation. The regulatory obligation is implemented by ADM-NGGC-0206.

Justification for Change:

The changes proposed to this commitment are in paper form only. The actions recommended by the source document are not being deleted, but are being brought up-to-date to the requirements of the 10 CFR 26 Subpart I. Therefore, this change cannot negatively impact the ability of a SSC to perform its safety function or negatively impact the ability of plant personnel to ensure a SSC is capable of performing its intended safety function.

U. S. Nuclear Regulatory Commission Attachment 1 3F0514-08 Page 9 of 41 Nuclear Operations Commitment System (NOCS) Number: 100559 Source Document:

CR-3 to NRC letter, 3F021 1-01, dated February 25, 2011.

Original Commitment:

Proposed Technical Specification limits for the DFT [Diesel Fuel Tank]-4 fuel oil supply tank level will be administratively maintained until the License Amendment is implemented.

Currently, the operating fuel tank volume is being administratively controlled under the provisions of NRC Administrative Letter 98-10, "Dispositioning of Technical Specifications that are Insufficient to Assure Plant Safety." Plant procedures have been revised to include a conservative value for fuel oil volume in DFT-4, since the current ITS [Improved Technical Specifications] 3.7.19 and SR [Surveillance Requirement] 3.7.19.1 were determined to be non-conservative.

Modify/inactivate Commitment:

INACTIVATE: This commitment was made in CR-3 to NRC letter, "Crystal River Unit 3 -

License Amendment Request (LAR) #311, Emergency Feedwater Pump Fuel Oil Volume" and is no longer necessary based upon implementation of License Amendment No. 240 (ADAMS Accession No. ML11292A041).

Justification for Change:

The administrative controls that maintained the DFT-4 tank level have been replaced by changes to ITS 3.7.19 and SR 3.7.19.1 which were implemented by License Amendment 240.

Therefore, this change cannot negatively impact the ability of a SSC to perform its safety function or negatively impact the ability of plant personnel to ensure a SSC is capable of performing its intended safety function.

U. S. Nuclear Regulatory Commission Attachment 1 3F0514-08 Page 10 of 41 Nuclear Operations Commitment System (NOCS) Number: 40434 Source Document:

CR-3 to NRC letter, 3F0289-04, dated February 10, 1989.

Original Commitment:

The FPC surveillance program for ATWS [anticipated transient without scram] will include testing of each channel from sensor input to actuated device on a 6 month interval with channel calibration, including the input sensors, to be done on a refueling interval.

Modify/Inactivate Commitment:

MODIFY: This commitment was included in the Final Design of the ATWS System that was submitted to the NRC for approval. A statement has been added specify that a 6 month interval testing is not required during NO MODE.

Justification for Change:

The ATWS System is not functional during NO MODE operations. Therefore, this change cannot negatively impact the ability of a SSC to perform its safety function or negatively impact the ability of plant personnel to ensure a SSC is capable of performing its intended safety function.

U. S. Nuclear Regulatory Commission Attachment 1 3F0514-08 Page 11 of 41 Nuclear Operations Commitment System (NOCS) Number: 100483 Source Document:

CR-3 to NRC letter, 3F0805-09, dated August 30, 2005.

Original Commitment:

The following interim compensatory measures that were adopted in response to NRC Bulletin 2003-01 will be maintained or revised to account for the modified RB sump strainer.

Accident Assessment Team (Technical Support Center and Emergency Operations Facility) guidance such as: (1) back-flushing the sump strainer (new strainer designed for gravity drain loads); (2) provision of core Decay Heat boil-off matching flow versus time curves; (3) terminating unnecessary ECCS [Emergency Core Cooling System]/CSS [Containment Spray System] pump operation; and, (4) commencement of Borated Water Storage Tank (BWST) refill following ECCS and BS pump switchover to the sump in preparation for re-establishing an injection flow path.

Modify/Inactivate Commitment:

INACTIVATE: This commitment is no longer required due to the permanent defueled condition of CR-3. The NRC to CR-3 letter dated June 5, 2013, "Crystal River Nuclear Plant - Closeout of Generic Letter 2004-02, "Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors (TAC NO. MC4678)," (ADAMS Accession No. ML13144A777), acknowledged the withdrawal of CR-3 commitments associated with Generic Letter 2004-02.

Justification for Change:

This commitment is related to accident mitigation for accidents that require recirculation of water through the containment sump. Given the permanent cessation of plant operations, these accidents are no longer credible. Therefore, this change cannot negatively impact the ability of a SSC to perform its safety function or negatively impact the ability of plant personnel to ensure a SSC is capable of performing its intended safety function.

U. S. Nuclear Regulatory Commission Attachment 1 3F0514-08 Page 12 of 41 Nuclear Operations Commitment System (NOCS) Number: 100484 Source Document:

CR-3 to NRC letter, 3F0805-09, dated August 30, 2005.

Original Commitment:

A surveillance procedure, SP-175A, will confirm that the strainer integrity is maintained prior to start-up from refueling outages (required by ITS 3.5.2.7). A specific checklist item in this procedure requires inspection to ensure there are no breaches or gaps greater than 1/16 inch in width.

Modify/Inactivate Commitment:

INACTIVATE: This commitment is no longer required due to the permanent defueled condition of CR-3. The NRC to CR-3 letter dated June 5, 2013, "Crystal River Nuclear Plant - Closeout of Generic Letter 2004-02, "Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors (TAC NO. MC4678)," (ADAMS Accession No. ML13144A777), acknowledged the withdrawal of CR-3 commitments associated with Generic Letter 2004-02.

Justification for Change:

This commitment is related to accident mitigation for accidents that require recirculation of water through the containment building sump. Given the permanent cessation of plant operations, these accidents are no longer credible. Therefore, this change cannot negatively impact the ability of a SSC to perform its safety function or negatively impact the ability of plant personnel to ensure a SSC is capable of performing its intended safety function.

U. S. Nuclear Regulatory Commission Attachment 1 3F0514-08 Page 13 of 41 Nuclear Operations Commitment System (NOCS) Number: 100505 Source Document:

CR-3 to NRC letter, 3F1207-05, dated December 18, 2007.

Original Commitment:

Crystal River Unit 3 will establish administrative limits that will increase the minimum inventory of the Borated Water Storage Tank (BWST) within the allowed Technical Specification range during the requested period of extension.

Modify/Inactivate Commitment:

INACTIVATE: This commitment is no longer required due to the permanent defueled condition of CR-3. The NRC to CR-3 letter dated June 5, 2013, "Crystal River Nuclear Plant - Closeout of Generic Letter 2004-02, "Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors (TAC NO. MC4678)," (ADAMS Accession No. ML13144A777), acknowledged the withdrawal of CR-3 commitments associated with Generic Letter 2004-02.

Justification for Change:

This commitment is related to accident mitigation for accidents that require recirculation of water through the containment building sump. Given the permanent cessation of plant operations, these accidents are no longer credible. Therefore, this change cannot negatively impact the ability of a SSC to perform its safety function or negatively impact the ability of plant personnel to ensure a SSC is capable of performing its intended safety function.

U. S. Nuclear Regulatory Commission Attachment 1 3F0514-08 Page 14 of 41 Nuclear Operations Commitment System (NOCS) Number: 100504 Source Document:

CR-3 to NRC letter, 3F1207-02, dated December 10, 2007.

Original Commitment:

Crystal River Unit 3 will proceduralize additional sump backflush methods utilizing the SFP

[Spent Fuel Pool] inventory as defense-in-depth strategies for mitigating sump screen blockage.

Modify/Inactivate Commitment:

INACTIVATE: This commitment is no longer required due to the permanent defueled condition of CR-3. The NRC to CR-3 letter dated June 5, 2013, "Crystal River Nuclear Plant - Closeout of Generic Letter 2004-02, "Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors (TAC NO. MC4678)," (ADAMS Accession No. ML13144A777), acknowledged the withdrawal of CR-3 commitments associated with Generic Letter 2004-02.

Justification for Change:

This commitment is related to accident mitigation for accidents that require recirculation of water through the containment building sump. Given the permanent cessation of plant operations, these accidents are no longer credible. Therefore, this change cannot negatively impact the ability of a SSC to perform its safety function or negatively impact the ability of plant personnel to ensure a SSC is capable of performing its intended safety function.

U. S. Nuclear Regulatory Commission Attachment 1 3F0514-08 Page 15 of 41 Nuclear Operations Commitment System (NOCS) Number: 2170 Source Document:

CR-3 to NRC letter, 3F1 183-03, dated November 4, 1983.

Original Commitment:

1.1 Post-trip review (program description and procedure):

Florida Power Corporation has in place and is maintaining a program to ensure that unplanned reactor shutdowns are analyzed and that a determination is made that the plant can be safely restarted. The following addresses programmatic elements:

The personnel performing restart analysis are the "person-on-call" and the Shift Technical Advisor (STA).

The "person-on-call" is a senior plant management individual holding a current Senior Reactor Operator license and is the designee of the Director, Nuclear Plant Operations.

The "person-on-call" is the evaluation team leader in assessing & justifying restart.

The "person-on-call" and the STA jointly ensure that the restart criteria are met.

The "person-on-call" must:

1) hold a current Senior Reactor Operator license

2) have completed Emergency Coordinator training, and;

3) be designated as the "person-on-call" by the ADNPO.

The STA must have completed Florida Power Corporation's STA training program and be designated as the STA by the (Supervisor, Operations Engineering and Support).

Note: Manager Shift Operation (MSO) is presently the person-on-call.

Modify/inactivate Commitment:

INACTIVATE: This commitment was made in the CR-3 response to Generic Letter 83-28, "Required Actions Based on Generic Implications of Salem ATWS Events," for Post-Trip Review. CR-3 committed to have programmatic elements in place to determine that the plant can be restarted safely. By permanently removing fuel from the reactor, the requirement to evaluate conditions for restart has been eliminated.

Justification for Change:

By permanently removing fuel from the reactor, the requirement to evaluate conditions for restart has been eliminated. Therefore, this change cannot negatively impact the ability of a SSC to perform its safety function or negatively impact the ability of plant personnel to ensure a SSC is capable of performing its intended safety function.

U. S. Nuclear Regulatory Commission Attachment 1 3F0514-08 Page 16 of 41 Nuclear Operations Commitment System (NOCS) Number: 40436 Source Document:

CR-3 to NRC letter, 3F0987-14, dated September 21, 1987.

Original Commitment:

Various administrative restrictions are in effect regarding maintenance of the DH [Decay Heat]

system. Prior to removing a Decay Heat train from service, the following requirements must be met:

A. no more than one DH train shall be removed from service at any one time, and B. the refueling canal is flooded or at least one steam generator is available for cooling either by forced flow or natural circulation, and C. the plant must be stable and under control within existing equipment, procedural, and personnel capability, and D. before removing the equipment from service, the redundant equipment will be determined to be operable per the surveillance program, and E. is an emergency diesel generator is to be taken out of service, no equipment in the opposite safeguards train shall be out of service, and F. this item eliminated via memo OPOO-0008.

G. additional compensating measures, such as a dedicated operator, may be considered to perform in the place of an automatic safety function if conditions warrant and approval is granted by appropriate management.

Modify/Inactivate Commitment:

INACTIVATE: Generic Letter No. 87-12, "Loss of Residual Heat Removal (RHR) While the Reactor Coolant System [RCS] is Partially Filled," was associated with the safe operations of pressurized-water reactors (PWRs) when the RCS water level is below the top of the reactor vessel with fuel in the reactor vessel. CR-3 responded to Generic Letter 87-12 with this specific commitment defining various administrative restrictions that were in place to control maintenance of the DH System. This commitment is associated with safe reactor operation during outages with fuel in the reactor vessel to ensure adequate backup methods of DH removal are available prior to removing a DH train from service.

Justification for Change:

This commitment is associated with safe operation during outages to ensure adequate methods of DH removal are available when fuel is in the reactor. CR-3 is no longer authorized to operate the reactor or emplace or retain fuel in the reactor vessel. Therefore, this change cannot negatively impact the ability of a SSC to perform its safety function or negatively impact the ability of plant personnel to ensure a SSC is capable of performing its intended safety function.

U. S. Nuclear Regulatory Commission Attachment 1 3F0514-08 Page 17 of 41 Nuclear Operations Commitment System (NOCS) Number: 100406 Source Document:

CR-3 to NRC letter, 3F0702-07, dated July 3, 2002.

Original Commitment:

The following compensatory measures will be put in place prior to online preventive maintenance to the EGDGs [Emergency Diesel Generators] that is planned to take over 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

1. CR-3 will perform procedure CP-253, "Power Operation Risk Assessment and Management," which requires both a deterministic and probabilistic evaluation of risk for the performance of all maintenance activities. This procedure uses the Level 1 PSA

[Probabilistic Safety Assessment] model to evaluate the impact of maintenance activities on core damage frequency. CR-3 will not plan any maintenance that results in "higher risk" (orange color code) during extended EDG maintenance.

2. ECCS equipment, Emergency Feedwater, Control Complex cooling and Auxiliary Feedwater (FWP-7 and MTDG-1) will be designated administratively as "protected" (no planned maintenance or discretionary equipment manipulation). [the term "discretionary equipment manipulation" is not intended to preclude manipulations required for normal operation of the plant, required surveillances or operator response to abnormal conditions.]

3. Prior to initiating a planned EDG outage, CR-3 will verify the availability of offsite power to the 230 KV switchyard and ensure that the capability to power both ES [Engineered Safeguards] busses is available from each of the two ES offsite power transformers (OPT and BEST).

4. CR-3 will not initiate an EDG extended preventive maintenance outage if adverse weather, as designated by Emergency Preparedness procedures, is anticipated.

5. No elective maintenance will be scheduled in the switchyard that would challenge the availability of offsite power to the ES busses.

6. A periodic fire watch will be established in fire areas that are considered risk significant by the IPEEE [Individual Plant Examination of External Events], affect both EDGs or have increased risk significance due to extended EDG maintenance. [The fire areas are listed in ITS Bases Table B3.8.1-1.]

7. A comprehensive walk down of redundant electrical and mechanical systems will be performed. (Ref. NCR [Nuclear Condition Report] 89327-04)

Modifyllnactivate Commitment:

INACTIVATE: LAR 257 extended the completion time for restoring an inoperable EDG to operable status from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 14 days (Reference ITS 3.8.1, Condition B.4) if alternate AC power is available. The items identified in this commitment are conditions to be implemented prior to allowing the extension of the Limiting Condition for Operation [LCO] completion time for ITS 3.8.1, AC Sources - Operating, from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 14 days for a planned extended EDG outage. ITS 3.8.1 is only applicable in MODES 1 through 4 with alternate AC (AAC) power available.

Justification for Change:

As a result of permanently removing fuel from the reactor, this commitment is no longer applicable. Therefore, this change cannot negatively impact the ability of a SSC to perform its safety function or negatively impact the ability of plant personnel to ensure a SSC is capable of performing its intended safety function.

U. S. Nuclear Regulatory Commission Attachment 1 3F0514-08 Page 18 of 41 Nuclear Operations Commitment System (NOCS) Number: 100507 Source Document:

CR-3 to NRC letter, 3F0208-01, dated February 25, 2008.

Original Commitment:

CR-3 will perform procedure CP-253, "Power Operation Risk Assessment and Management,"

which requires both a deterministic and probabilistic evaluation of risk for the performance of all maintenance activities. This procedure uses the Level 1 PSA model to evaluate the impact of maintenance activities on core damage frequency. CR-3 will not plan any maintenance that results in "higher risk" (orange color code) during an extended outage (greater than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />) of the LPI [Low Pressure Injection], BS [Building Spray], DC [Decay Heat Closed Cycle Cooling Water] or RW-DC [Decay Heat Seawater] system.

Modify/Inactivate Commitment:

INACTIVATE: This commitment was included in LAR 295 that resulted in License Amendment No. 229 (ADAMS Accession No. ML081060231) for increasing the Completion Times associated with one inoperable Low Pressure Injection (LPI) Train (Reference ITS 3.5.2), one Reactor Building Spray (RBS) Train (Reference ITS 3.6.6), one Decay Heat Closed Cycle Cooling Water (DC) Train (Reference ITS 3.7.8) and one Decay Heat Seawater (RW, or RW-DC) (Reference ITS 3.7.10) Train from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 7 days. This ITS change also eliminated the second Completion Time for RBS (ITS 3.6.6), Emergency Feedwater (EFW) (Reference ITS 3.7.5), AC Sources - Operating (Reference ITS 3.8.1) and Distribution System - Operating (Reference ITS 3.8.9). Part of the acceptance of the ITS change was that CR-3 committed to perform CR-3 site-specific procedure, CP-253, as defined by this commitment.

Justification for Change:

None of the referenced Technical Specifications are applicable during a defueled condition.

CR-3 is no longer authorized to operate the reactor or emplace or retain fuel in the reactor vessel. Therefore, this change cannot negatively impact the ability of a SSC to perform its safety function or negatively impact the ability of plant personnel to ensure a SSC is capable of performing its intended safety function.

U. S. Nuclear Regulatory Commission Attachment 1 3F0514-08 Page 19 of 41 Nuclear Operations Commitment System (NOCS) Number: 100508 Source Document:

CR-3 to NRC letter, 3F0208-01, dated February 25, 2008.

Original Commitment:

The Remote Shutdown Panel, the Appendix R Cooler and the opposite train of LPI, BS, DC, RW-DC, EFW, Auxiliary Feedwater system, Emergency Feedwater Initiation and Control System, HPI [High Pressure Injection], and their power supplies (AC and DC) will be administratively designated as "protected" (i.e., no planned maintenance or discretionary equipment manipulation) during extended (greater than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />) outage on the LPI, BS, DC or RW-DC system.

Modify/Inactivate Commitment:

INACTIVATE: This commitment was included in LAR 295 that resulted in License Amendment No. 229 for increasing the Completion Times associated with one inoperable LPI Train (Reference ITS 3.5.2), one RBS Train (Reference ITS 3.6.6), one DC Train (Reference ITS 3.7.8) and one RW, or RW-DC (Reference ITS 3.7.10) Train from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 7 days. This change also eliminated the second Completion Time for RBS (Reference ITS 3.6.6), EFW (Reference ITS 3.7.5), AC Sources - Operating (Reference ITS 3.8.1) and Distribution System -

Operating (Reference ITS 3.8.9). Part of the acceptance of the ITS change was that CR-3 committed to designate the equipment, as defined by this commitment, with "Protective" signs to increase the awareness of the protection of those components during the LCO time.

Justification for Change:

None of the referenced Technical Specifications are applicable during a defueled condition.

CR-3 is no longer authorized to operate the reactor or emplace or retain fuel in the reactor vessel. Therefore, this change cannot negatively impact the ability of a SSC to perform its safety function or negatively impact the ability of plant personnel to ensure a SSC is capable of performing its intended safety function.

U. S. Nuclear Regulatory Commission Attachment 1 3F0514-08 Page 20 of 41 Nuclear Operations Commitment System (NOCS) Number: 100509 Source Document:

CR-3 to NRC letter, 3F0208-01, dated February 25, 2008.

Original Commitment:

CR-3 will not initiate an extended preventive maintenance outage (greater than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />) on the LPI, BS, DC or RW-DC system if adverse weather, as designated by Emergency Preparedness procedures, is anticipated.

Modify/Inactivate Commitment:

INACTIVATE: This commitment was included in LAR 295 that resulted in License Amendment No. 229 for increasing the Completion Times associated with one inoperable LPI Train (Reference ITS 3.5.2), one RBS Train (Reference ITS 3.6.6), one DC Train (Reference ITS 3.7.8) and one RW, or RW-DC (Reference ITS 3.7.10) Train from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 7 days. This change also eliminated the second Completion Time for RBS (Reference ITS 3.6.6), EFW (Reference ITS 3.7.5), AC Sources - Operating (Reference ITS 3.8.1) and Distribution System -

Operating (Reference ITS 3.8.9). Part of the acceptance of the ITS change was that CR-3 committed not to initiate an extended preventive maintenance outage during adverse weather as defined by this commitment.

Justification for Change:

None of the referenced Technical Specifications are applicable during a defueled condition.

CR-3 is no longer authorized to operate the reactor or emplace or retain fuel in the reactor vessel. Therefore, this change cannot negatively impact the ability of a SSC to perform its safety function or negatively impact the ability of plant personnel to ensure a SSC is capable of performing its intended safety function.

U. S. Nuclear Regulatory Commission Attachment 1 3F0514-08 Page 21 of 41 Nuclear Operations Commitment System (NOCS) Number: 100510 Source Document:

CR-3 to NRC letter, 3F0208-01, dated February 25, 2008.

Original Commitment:

When extended maintenance (greater than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />) is performed (scheduled or emergent) on a train of the LPI or BS system, CR-3 will limit transient combustibles in, and establish a periodic fire watch in the decay heat pump vault of the opposite train, and the following rooms:

• Non-safety 4160V and 480V switchgear rooms

• Opposite train ES 4160V and ES 480V switchgear rooms

• Opposite train battery room

• Opposite train charger room

• Opposite train inverters room

• Remote Shutdown Panel room

• Relay/CRD [Control Rod Drive] room and adjoining corridor

• 'B' EFIC [Emergency Feedwater Initiation and Control] room

• Cable Spreading room Modify/Inactivate Commitment:

INACTIVATE: This commitment was included in LAR 295 that resulted in License Amendment No. 229 for increasing the Completion Times associated with one inoperable LPI Train (Reference ITS 3.5.2), one RBS Train (Reference ITS 3.6.6), one DC Train (Reference ITS 3.7.8) and one RW, or RW-DC (Reference ITS 3.7.10) Train from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 7 days. This change also eliminated the second Completion Time for RBS (Reference ITS 3.6.6), EFW (Reference ITS 3.7.5), AC Sources - Operating (Reference ITS 3.8.1) and Distribution System -

Operating (Reference ITS 3.8.9). Part of the acceptance of the ITS change was that CR-3 committed to limit the potential of a fire and increase the probability of being aware of a fire in the rooms defined by this commitment.

Justification for Change:

None of the referenced Technical Specifications are applicable during a defueled condition.

CR-3 is no longer authorized to operate the reactor or emplace or retain fuel in the reactor vessel. Therefore, this change cannot negatively impact the ability of a SSC to perform its safety function or negatively impact the ability of plant personnel to ensure a SSC is capable of performing its intended safety function.

U. S. Nuclear Regulatory Commission Attachment 1 3F0514-08 Page 22 of 41 Nuclear Operations Commitment System (NOCS) Number: 100511 Source Document:

CR-3 to NRC letter, 3F0208-01, dated February 25, 2008.

Original Commitment:

When extended maintenance (greater than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />) is performed (scheduled or emergent) on a train of the DC or RW DC system, CR-3 will limit transient combustibles in, and establish a periodic fire watch in the seawater room, and the following rooms:

• Non-safety 4160V and 480V switchgear rooms

" Opposite train ES 4160V and ES 480V switchgear rooms

" Opposite train battery room

• Opposite train charger room

• Opposite train inverters room

• Remote Shutdown Panel room

• Relay/CRD room and adjoining corridor

• 'B' EFIC room

• Cable Spreading room Modify/inactivate Commitment:

INACTIVATE: This commitment was included in LAR 295 that resulted in License Amendment No. 229 for increasing the Completion Times associated with one inoperable LPI Train (Reference ITS 3.5.2), one RBS Train (Reference ITS 3.6.6), one DC Train (Reference ITS 3.7.8) and One RW, or RW-DC (Reference ITS 3.7.10) Train from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 7 days. This change also eliminated the second Completion Time for RBS (Reference ITS 3.6.6), EFW (Reference ITS 3.7.5), AC Sources - Operating (Reference ITS 3.8.1) and Distribution System -

Operating (Reference ITS 3.8.9). Part of the acceptance of the ITS change was that CR-3 committed to limit the potential of a fire and increase the probability of being aware of a fire in the rooms defined by this commitment.

Justification for Change:

None of the referenced Technical Specifications are applicable during a defueled condition.

CR-3 is no longer authorized to operate the reactor or emplace or retain fuel in the reactor vessel. Therefore, this change cannot negatively impact the ability of a SSC to perform its safety function or negatively impact the ability of plant personnel to ensure a SSC is capable of performing its intended safety function.

U. S. Nuclear Regulatory Commission Attachment 1 3F0514-08 Page 23 of 41 Nuclear Operations Commitment System (NOCS) Number: 100512 Source Document:

CR-3 to NRC letter, 3F0208-01, dated February 25, 2008.

Original Commitment:

When extended maintenance (greater than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />) is performed (scheduled or emergent) on a train of the LPI, BS, DC or RW-DC system, CR-3 will limit transient combustibles and establish a periodic fire watch in the fire zones containing routed cables associated with the pressurizer PORV [Power-Operated Relief Valve] and PORV block valves. These rooms include:

• PORV/PORV Block Valve power supply breaker areas

• Cable Spreading room

• Relay/CRD room and adjoining corridor

• Intermediate Building 119' elevation

• Auxiliary Building 119' elevation

" 'B' ES 4160V Switchgear Room

• Remote Shutdown Room

• 'A'/'B' Battery Room Modify/inactivate Commitment:

INACTIVATE: This commitment was included in LAR 295 that resulted in License Amendment No. 229 for increasing the Completion Times associated with one inoperable LPI Train (Reference ITS 3.5.2), one RBS Train (Reference ITS 3.6.6), one DC Train (Reference ITS 3.7.8) and one RW, or RW-DC (Reference ITS 3.7.10) Train from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 7 days. This change also eliminated the second Completion Time for RBS (Reference ITS 3.6.6), EFW (Reference ITS 3.7.5), AC Sources - Operating (Reference ITS 3.8.1) and Distribution System -

Operating (Reference ITS 3.8.9). As part of the acceptance of the ITS change, CR-3 committed to limit the potential of a fire and increase the probability of being aware of a fire in the rooms defined by this commitment which was associated with operation of the PORV and PORV Block valve.

Justification for Change:

None of the referenced Technical Specifications are applicable during a defueled condition.

CR-3 is no longer authorized to operate the reactor or emplace or retain fuel in the reactor vessel. Therefore, this change cannot negatively impact the ability of a SSC to perform its safety function or negatively impact the ability of plant personnel to ensure a SSC is capable of performing its intended safety function.

U. S. Nuclear Regulatory Commission Attachment 1 3F0514-08 Page 24 of 41 Nuclear Operations Commitment System (NOCS) Number: 100513 Source Document:

CR-3 to NRC letter, 3F0208-01 dated February 25, 2008.

Original Commitment:

There shall be administrative controls to limit the maximum time allowed for any combination of conditions that result in a single contiguous occurrence of failing to meet the LCO for ITS 3.6.6, Reactor Building Spray and Containment Cooling Systems, ITS 3.7.5, Emergency Feedwater (EFW) system, ITS 3.8.1, AC Sources - Operating, and ITS 3.8.9, Distribution Systems -

Operating. These administrative controls shall ensure that the Completion Times for those Conditions are not inappropriately extended. The administrative controls will ensure that Completion Time is not extended beyond the additive Completion Times of the two Required Actions for restoration of OPERABILITY unless a risk evaluation is performed. If unit operation within an LCO will exceed the maximum Completion Time, then either the shutdown Condition within the LCO should be entered or a risk evaluation shall be performed and the risk impact managed under CP-253, "Power Operation Risk Assessment and Management."

Modify/Inactivate Commitment:

INACTIVATE: This commitment was included in LAR 295 that resulted in License Amendment No. 229 for increasing the Completion Times associated with one inoperable LPI Train (Reference ITS 3.5.2), one RBS Train (Reference ITS 3.6.6), one DC Train (Reference ITS 3.7.8) and one RW, or RW-DC (Reference ITS 3.7.10) Train from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 7 days. This change also eliminated the second Completion Time for RBS (Reference ITS 3.6.6), EFW (Reference ITS 3.7.5), AC Sources - Operating (Reference ITS 3.8.1) and Distribution System -

Operating (Reference ITS 3.8.9). Part of the acceptance of the ITS change was that CR-3 committed to limit the inappropriate implementation of the extended Completion Time as defined by this commitment.

Justification for Change:

None of the referenced Technical Specifications are applicable during a defueled condition.

CR-3 is no longer authorized to operate the reactor or emplace or retain fuel in the reactor vessel. Therefore, this change cannot negatively impact the ability of a SSC to perform its safety function or negatively impact the ability of plant personnel to ensure a SSC is capable of performing its intended safety function.

U. S. Nuclear Regulatory Commission Attachment 1 3F0514-08 Page 25 of 41 Nuclear Operations Commitment System (NOCS) Number: 100549 Source Document:

CR-3 to NRC letter, 3F1008-05, dated October 13, 2008.

Original Commitment:

Quarterly monitoring will be developed and implemented to ensure that the ECCS, DH, and BS suction and discharge piping will be maintained sufficiently full of water to ensure that the systems can reliably perform their intended functions.

The inspections will include a requirement for periodic verification (every 92 days) that the ECCS, DH, and BS piping will be maintained sufficiently full of water by a combination of Ultrasonic Testing (UT), and venting as deemed necessary, of locations identified to be potentially susceptible to gas intrusion. (high to low pressure interfaces)

Modify/inactivate Commitment:

INACTIVATE: This commitment was contained in CR-3's response to Generic Letter 2008-01, "Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal, and Containment Spray Systems," which resulted in three locations being selected for quarterly Ultrasonic Testing (UT) examinations based on those locations being identified as high-to-low pressure interfaces. Duke Energy Florida (DEF) certified to the NRC, pursuant to 10 CFR 50.82, that it had permanently ceased operation of CR-3 and permanently removed the fuel from the reactor by letter dated February 20, 2013 (ADAMS Accession No. ML13056A005).

Therefore, this commitment is no longer applicable to CR-3's permanently defueled condition.

Justification for Change:

This commitment is related to periodic monitoring of the low pressure side of three identified high-to-low pressure interfaces in the DH/LPI system: upstream of DHV-5 and DHV-6 and downstream of DHV-41. The intent was to monitor for gas formation on the low pressure side of each valve that could potentially result from seat leakage across the valve from the high pressure side. The high pressure sides were associated with normal RCS pressure and normal Core Flood Tank (CFT) pressures, with the associated gases in solution on the high pressure side being either hydrogen gas from the RCS (consequence of letdown flow through the gas space of the Makeup Tank) or nitrogen gas from the CFTs. Given the cessation of power operation at CR-3, the RCS and CFTs will never again experience normal operating pressures and the three previous locations of high-to-low pressure interface will never again exist.

Furthermore, given the permanent removal of nuclear fuel from the reactor vessel, the credible accidents that rely upon the BS, HPI, and LPI systems for Design Basis accident mitigation are no longer possible and the DH system function to maintain reactor core cooling during normal plant shutdown conditions is no longer required. Therefore, this change cannot negatively impact the ability of a SSC to perform its safety function or negatively impact the ability of plant personnel to ensure a SSC is capable of performing its intended safety function.

U. S. Nuclear Regulatory Commission Attachment 1 3F0514-08 Page 26 of 41 Nuclear Operations Commitment System (NOCS) Number: 100550 Source Document:

CR-3 to NRC letter, 3F1008-05, dated October 13, 2008.

Original Commitment:

Additionally, should any maintenance activities breach the ECCS, DH, or BS system boundary, a UT will be performed as deemed necessary to verify the respective system(s) are sufficiently full prior to return to service.

Modify/inactivate Commitment:

INACTIVATE: This commitment was contained in CR-3's response to Generic Letter 2008-01, "Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal, and Containment Spray Systems," and is related to maintaining the viability of: 1) the BS, HPI, and LPI systems for Design Basis accident mitigation; and 2) the DH system for reactor core cooling during normal plant shutdown conditions. This commitment is no longer applicable to CR-3's permanently defueled condition.

Justification for Change:

Given the permanent removal of nuclear fuel from the reactor vessel, the credible accidents that rely upon the BS, HPI, and LPI systems for Design Basis accident mitigation are no longer possible and the DH system function to maintain reactor core cooling during normal plant shutdown conditions is no longer required. Therefore, this change cannot negatively impact the ability of a SSC to perform its safety function or negatively impact the ability of plant personnel to ensure a SSC is capable of performing its intended safety function.

U. S. Nuclear Regulatory Commission Attachment 1 3F0514-08 Page 27 of 41 Nuclear Operations Commitment System (NOCS) Number: 100551 Source Document:

CR-3 to NRC letter, 3F1008-05, dated October 13, 2008.

Original Commitment:

Procedures will be enhanced to provide additional detail where needed concerning venting sequence, venting duration, dynamic venting, etc., and to UT appropriate piping locations following fill and vent as deemed necessary, to ensure piping is sufficiently full prior to return to service.

Modify/Inactivate Commitment:

INACTIVATE: This commitment was contained in CR-3's response to Generic Letter 2008-01, "Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal, and Containment Spray Systems," and is related to using appropriate fill and vent practices to maintain the viability of the BS, HPI, and LPI systems for Design Basis accident mitigation and the DH system for reactor core cooling during normal plant shutdown conditions. This commitment is no longer applicable to CR-3's permanently defueled condition.

Justification for Change:

Given the permanent removal of nuclear fuel from the reactor vessel, the credible accidents that rely upon the BS, HPI, and LPI systems for Design Basis accident mitigation are no longer possible and the DH system function to maintain reactor core cooling during normal plant shutdown conditions is no longer required. Therefore, this change cannot negatively impact the ability of a SSC to perform its safety function or negatively impact the ability of plant personnel to ensure a SSC is capable of performing its intended safety function.

U. S. Nuclear Regulatory Commission Attachment 1 3F0514-08 Page 28 of 41 Nuclear Operations Commitment System (NOCS) Number: 100452 Source Document:

CR-3 to NRC letter, 3F1203-02, dated December 9, 2003.

Original Commitment:

The Control Complex Habitability Envelope Integrity Program will include ASTM E741 (Tracer Gas) testing on a periodic basis as recommended in Regulatory Guide 1.197.

Modify/Inactivate Commitment:

INACTIVATE: This commitment was contained in CR-3's response to Generic Letter 2003-01, "Control Room Habitability," and is no longer needed.

Justification for Change:

CR-3 is permanently shut down and the credible accidents described in the CR-3 Final Safety Analysis Report (FSAR) have been reduced to the Fuel Handling Accident. CR-3 Calculation N13-0001, "Public and Control Room Dose from a Fuel Handling Accident," evaluates the drop of a spent fuel assembly onto the spent fuel racks which results in breaking all the fuel rods.

The purpose of this analysis is to determine the dose to operators in the Control Room and public dose at the Exclusion Area Boundary (EAB) and Low Population Zone. Due to the amount of decay assumed (4 years), the results of this analysis are applicable after September 26, 2013. The 30 day integrated Control Room dose is projected to be 1.3E-02 millirem Total Effective Dose Equivalent (TEDE) (Control Room isolation is not credited), which is below the 5 rem TEDE Control Room dose limit established in 10 CFR 50.67. The resulting two hour integrated dose at the EAB is projected to be 5.9E-03 millirem TEDE, which is well below the EAB limit. Control Room Habitability is no longer necessary to prevent operators from exposure to high dose concentrations. Therefore, this change cannot negatively impact the ability of a SSC to perform its safety function or negatively impact the ability of plant personnel to ensure a SSC is capable of performing its intended safety function.

U. S. Nuclear Regulatory Commission Attachment 1 3F0514-08 Page 29 of 41 Nuclear Operations Commitment System (NOCS) Number: 100408 Source Document:

CR-3 to NRC letter, 3F0803-03, dated August 8, 2003.

Original Commitment:

NRC Bulletin 2003-01: Potential Impact of Debris Blockage on Emergency Sump Recirculation at Pressurized Water Reactors.

EOP [Emergency Operating Procedure] and TSC [Technical Support Center] procedure revisions will be issued for use no later than September 5, 2003, fully implementing the interim compensatory measures.

Modify/inactivate Commitment:

INACTIVATE: This commitment was contained in CR-3's response to NRC Bulletin 2003-01 and is no longer needed.

Justification for Change:

This commitment is related to accident mitigation for those accidents that require recirculation of water through the containment building sump. Given that CR-3 has permanently ceased operations, accident mitigation is no longer required for those accidents since they are no longer credible. Therefore, this change cannot negatively impact the ability of a SSC to perform its safety function or negatively impact the ability of plant personnel to ensure a SSC is capable of performing its intended safety function.

U. S. Nuclear Regulatory Commission Attachment 1 3F0514-08 Page 30 of 41 Nuclear Operations Commitment System (NOCS) Number: 100440 Source Document:

CR-3 to NRC letter, 3F0204-05, dated February 27, 2004.

Original Commitment:

PEF [Progress Energy Florida] verified that it has contingency plans for obtaining and analyzing highly radioactive samples from the RCS, containment sump, and containment atmosphere.

The contingency plans are contained in plant procedures and implementation is complete.

Establishment and maintenance of contingency plans is considered a regulatory commitment.

Modify/Inactivate Commitment:

INACTIVATE: This commitment was included in LAR 273 and was made as a compensatory measure to support deletion of the Post-Accident Sampling System (PASS) from the CR-3 ITS.

This was completed with License Amendment No. 213 (ADAMS Accession No. ML041940505).

PASS was required by NUREG-0737, "Clarification of TMI Action Plan Requirements," Item II.B.3, to enable licensee personnel to obtain reactor coolant and containment atmosphere samples under accident conditions within one hour without incurring radiation exposure to any individual in excess of 3 rem to the whole body or 18.75 rem to the extremities. This commitment is no longer applicable to CR-3's permanently defueled condition.

Justification for Change:

This commitment is related to mitigation of those accidents that require sampling of the RCS, containment sump, and containment atmosphere for radioactive material. CR-3 has permanently ceased operations and these accidents are no longer credible. Therefore, this change cannot negatively impact the ability of a SSC to perform its safety function or negatively impact the ability of plant personnel to ensure a SSC is capable of performing its intended safety function.

U. S. Nuclear Regulatory Commission Attachment 1 3F0514-08 Page 31 of 41 Nuclear Operations Commitment System (NOCS) Number: 100441 Source Document:

CR-3 to NRC letter, 3F0204-05, dated February 27, 2004.

Original Commitment:

The capability for classifying fuel damage events at the Alert level threshold has been established for CR-3 at radioactivity levels of >300 mCi/gm 1-131 dose equivalent. This capability is described in plant procedures and implementation is complete. The capability for classifying fuel damage events is considered a regulatory commitment.

Modify/inactivate Commitment:

INACTIVATE: This commitment was included in LAR 273 and was made as a compensatory measure to support deletion of PASS from the CR-3 ITS. This was completed with License Amendment No. 213 (ADAMS Accession No. ML041940505). PASS was required by NUREG-0737, "Clarification of TMI Action Plan Requirements," Item ll.B.3, to enable licensee personnel to obtain reactor coolant and containment atmosphere samples under accident conditions within one hour without incurring radiation exposure to any individual in excess of 3 rem to the whole body or 18.75 rem to the extremities. Given the permanent cessation of CR-3 operation and the removal of nuclear fuel from the reactor vessel, it is no longer necessary to sample the RCS and containment atmosphere for radioactive iodines.

Justification for Change:

This commitment is implemented in the current revision of the CR-3 Radiological Emergency Response Plan (RERP) that requires declaration of an Alert with RCS activity greater than 300 mCi/gm 1-131 dose equivalent (RERP Rev. 32, Table 8.1, Section 5.1, Item 2). Table 8.1, Fission Product Barrier Matrix, clearly states that applicability is limited to MODES 1 - 4. CR-3 is no longer authorized to operate the reactor or emplace or retain nuclear fuel in the reactor vessel. By permanently removing fuel from the reactor, MODES 1 through 4 can never again be achieved and this required commitment is no longer applicable. The capability of declaring an Alert per the Fission Product Barrier Matrix is also no longer necessary. Therefore, this change cannot negatively impact the ability of a SSC to perform its safety function or negatively impact the ability of plant personnel to ensure a SSC is capable of performing its intended safety function.

U. S. Nuclear Regulatory Commission Attachment 1 3F0514-08 Page 32 of 41 Nuclear Operations Commitment System (NOCS) Number: 100442 Source Document:

CR-3 to NRC letter, 3F0204-05, dated February 27, 2004.

Original Commitment:

PEF verified that it has an ability to assess radioactive iodines released to offsite environs. The capability for monitoring iodines will be maintained within plant procedures. Implementation of this commitment is complete. The capability to monitor radioactive iodines is considered a regulatory commitment.

Modify/Inactivate Commitment:

INACTIVATE: This commitment was included in LAR 273 and was made as a compensatory measure to support deletion of the PASS from the CR-3 ITS. This was completed with License Amendment No. 213 (ADAMS Accession No. ML041940505). PASS was required by NUREG-0737, "Clarification of TMI Action Plan Requirements," Item ll.B.3, to enable licensee personnel to monitor for radionuclides in the reactor coolant, the containment atmosphere, and the containment sump. This commitment is no longer applicable to CR-3's permanently defueled condition.

Justification for Change:

This commitment was made to monitor radioactive iodines released to offsite environs caused by nuclear fuel damage events. With the cessation of operation at CR-3 and the permanent removal of nuclear fuel from the reactor vessel, the only remaining credible fuel damage accident sequence is a Fuel Handling Accident in the SFP. Monitoring of radioactive iodines following a Fuel Handling Accident in the SFP is not required by NUREG-0737, Item II.B.3.

Given the permanent removal of nuclear fuel from the reactor, this commitment is no longer applicable. Therefore, this change cannot negatively impact the ability of a SSC to perform its safety function or negatively impact the ability of plant personnel to ensure a SSC is capable of performing its intended safety function.

U. S. Nuclear Regulatory Commission Attachment 1 3F0514-08 Page 33 of 41 Nuclear Operations Commitment System (NOCS) Number: 4565 Source Document:

CR-3 to NRC letter, 3F0578-08, dated May 30, 1978.

Original Commitment:

Control of activities to minimize fire hazards - 1) The fire prevention work permit (FPWP) is obtained from the Maintenance Superintendent see Note. 2) The FPWP procedure requires that a person be designated as fire watch during and after the work, a person designated by the Maintenance Superintendent shall inspect the work areas for flammables and for the presence of ample portable extinguishers. 3) Combustibles which cannot be removed are shielded from flames and sparks. 4) The watch is continued during lunch and rest periods and for one half hour after completion of the work. 5) The plant operators are notified prior to start and after conclusion of work.

Cutting, welding or grinding that is to be done outside the maintenance shops is controlled using a fire prevention work permit procedure.

Note 1: Cutting, welding or grinding that is to be done during Mode 1, 2, & 3 in the cable spreading room, ES switchgear rooms, station battery rooms, CRD equipment room and control center must be evaluated by the fire protection engineer. In all other areas of the plant during Mode 1, 2, and 3 and during Modes 4, 5, and 6 in all locations of the plant, authorization may be given by the Operations Work Coordinator (OWC) on a fire prevention work permit.

Note 2; In accordance with CP-252 and REG-NGGC-0110 under NTM [Nuclear Task Management] 507299-75, the following modification to the above commitment is being made.

"Due to the permanent cessation of operations and permanent removal of fuel from the reactor vessel, CR-3 is no longer capable of reaching Modes 1, 2 or 3. The restriction on hot work activities in the cable spreading room, ES switchgear rooms, station battery rooms, CRD equipment room and control center no longer apply. All hot work permits will be approved in accordance with FIR-0003.

Control of Ignition Sources:..

1. Control of welding, cutting, grinding, and open flame work: Activities involving ignition sources shall be controlled by the use of a fire prevention work permit, which is issued under the guidelines of CP-1 18, "Fire Prevention Work Permit". A person designated by the Maintenance Supervisor shall survey the work area for flammables and for the presence of ample portable extinguishers. Combustibles which cannot be removed are shielded from flames or sparks.

This meets the intent of the following NRC guideline: (NRC to FPC 2/3/78):

2. Before issuing the permit, the responsible foreman or supervisor should physically survey the area where the work is to be performed and establish that the following precautions have been accomplished: (see Note 1).

3. All moveable combustible material below and within a 35-foot radius of the cutting, welding, grinding, or open flame work has been removed. (Per NFPA 51b, see Note 3).

4. All immovable combustible material below and within a 35-foot radius has been thoroughly protected by asbestos curtains, metal guards, or flameproof covers and fire extinguishers, hose

U. S. Nuclear Regulatory Commission Attachment 1 3F0514-08 Page 34 of 41 or other fire fighting equipment are provided at the work site. (See NFPA 51b.) (Also, asbestos curtains not in last rev. of NFPA 518). Welding, cutting, or spark producing work is prohibited in the cable spreading room, engineered safeguards switchgear rooms, battery rooms, control rod drive equipment room, and control center during operational Modes 1, 2, or 3 (power operations, startup, hot standby) unless permission is given by the Nuclear Plant Manager.

Welding, cutting, or spark-producing work conducted in these areas must be performed in accordance with CP-118. (see Notes 1 and 4.)

Administrative Controls:..The CR-3 FPP will describe the administrative controls established at CR-3 to prevent fires from work involving ignition sources.

Note 3: FPC has no specific commitment to NFPA 51b.

Note 4: CP-1 18 has been superceded by FIR-0003 and the term "fire prevention work permit" has been replaced by the term "hot work permit."

Modify/Inactivate Commitment:

MODIFY: This commitment was included in the CR-3 response to NRC guideline, "Nuclear Plant Fire Protection Functional Responsibilities, Administrative Controls and Quality Assurance." This commitment has been revised to clarify that FIR-0003 will continue to be used for hot work activities, however, the provisions for power modes will not continue to be used as a result of the permanent shutdown of CR-3.

Justification for Change:

This clarification regarding the inability of CR-3 to reach MODES 1, 2, and 3 does not negatively impact the ability of a SSC to perform its safety function or negatively impact the ability of plant personnel to ensure a SSC is capable of performing its intended safety function.

U. S. Nuclear Regulatory Commission Attachment 1 3F0514-08 Page 35 of 41 Nuclear Operations Commitment System (NOCS) Number: 6547 Source Document:

CR-3 to NRC letter, 3F0685-01, dated June 5, 1985.

Original Commitment:

Since evaluating the potential for and consequences of a refueling cavity water seal failure, refueling procedures have been revised and a new Abnormal Procedure has been implemented. The procedures provide guidance on how to respond to an uncontrolled decrease in refueling cavity water level.

Modify/inactivate Commitment:

INACTIVATE: This commitment was made in the CR-3 response to NRC Bulletin 84-03, "Refueling Cavity Water Seal." This Bulletin required an evaluation of the potential for and consequences of a refueling cavity water seal failure and the submittal of a summary report.

The evaluation was to include, in part, a consideration of the time to cladding damage without operator action and the potential effects on stored nuclear fuel and fuel in transfer. As a result of the evaluation, site refueling procedures were revised and a new abnormal procedure (AP-1080) was implemented. This commitment is no longer applicable to CR-3's permanently defueled condition.

Justification for Change:

CR-3 committed to providing procedural guidance on how to respond to an uncontrolled decrease in refueling cavity water level. The guidance provided was to safely place fuel suspended in the fuel handling bridge into the deep end of the fuel transfer canal. Given the permanent removal of nuclear fuel from the reactor, the requirement to mitigate the consequences of a loss of refueling canal level has been eliminated. Therefore, this change does not negatively impact the ability of a SSC to perform its safety function or negatively impact the ability of plant personnel to ensure a SSC is capable of performing its intended safety function.

U. S. Nuclear Regulatory Commission Attachment 1 3F0514-08 Page 36 of 41 Nuclear Operations Commitment System (NOCS) Number: 100156 Source Document:

CR-3 to NRC letter, 3F0586-19, dated May 14, 1986.

Original Commitment:

FPC will select individuals to perform STA duties who have Bachelor's degrees in a scientific or engineering discipline with at least 4 years nuclear power experience.

Update 10/27/2008: NTM AR [Action Request] 256002-12 modified this regulatory commitment to read, "Progress Energy will select individuals to perform STA duties who have Bachelor's degrees in a scientific or engineering discipline with a minimum of one year nuclear power plant experience, with at least six months experience on site."

Modify/Inactivate Commitment:

INACTIVATE: This commitment defined the qualification requirements for the STA associated with Generic Letter 86-04, "Policy Statement on Engineering Expertise on Shift," and ANSI/ANS-3.1-1981, "Selection, Qualification and Training of Personnel for Nuclear Power Plants." NUREG-0737, "Clarification of TMI Action Plan Requirements,"Section I.A.1.1, requires the STA position to be manned for operating MODES 1 through 4. This commitment is no longer applicable to CR-3's permanently defueled condition.

Justification for Change:

NUREG-0737 only requires the STA position to be manned for operating MODES 1 through 4.

CR-3 is no longer authorized to operate the reactor or place or retain fuel in the reactor vessel and will never again operate in MODES 1 through 4. Therefore, inactivating this commitment does not negatively impact the ability of plant personnel to ensure a SSC is capable of performing its intended safety function and likewise does not negatively impact the ability of a SSC to perform its safety function or the ability of plant personnel to ensure a SSC is capable of performing its intended safety function.

U. S. Nuclear Regulatory Commission Attachment 1 3F0514-08 Page 37 of 41 Nuclear Operations Commitment System (NOCS) Number: 100224 Source Document:

CR-3 to NRC letter, 3F0797-25, dated July 21, 1997.

Original Commitment:

Evaluation of drift data and the impact of the increased surveillance intervals will be completed within 180 days after restart from a refueling outage and any potential revisions to calibration intervals of setpoints will be made during this period.

Modify/Inactivate Commitment:

INACTIVATE: This commitment is associated with License Amendment No. 152 (ADAMS Accession No. ML020700585). The evaluation of drift data and any impact on increased surveillances after restart from a refueling outage is no longer applicable to CR-3's permanently defueled condition.

Justification for Change:

The original intent of this re-occurring commitment was to evaluate drift data and the impact of increased surveillance intervals within 180 days after restart from a refueling outage. Since CR-3 will no longer restart from a refueling outage, this commitment is no longer necessary and this change will not negatively impact the ability of a SSC to perform its safety function or negatively impact the ability of plant personnel to ensure a SSC is capable of performing its intended safety function.

U. S. Nuclear Regulatory Commission Attachment 1 3F0514-08 Page 38 of 41 Nuclear Operations Commitment System (NOCS) Number: 40449 Source Document:

CR-3 to NRC letter, 3F0987-20, dated September 25, 1987.

Original Commitment:

FPC utilizes in-house valve actuator training for craft and engineering personnel. The craft are trained with this program prior to performing maintenance on actuators unless under the direct supervision of trained personnel.

Modify/Inactivate Commitment:

INACTIVATE: This commitment was made in the CR-3 response to NRC Bulletin 85-03, "Motor-Operated Valve Common Mode Failures During Plant Transients Due to Improper Switch Settings." This commitment is no longer applicable in CR-3's permanently defueled condition. By permanently removing fuel from the reactor, fuel bundles are now only located in the SFP. There are no motor operated valves (MOVs) that support the operation and control of the SFP. This includes the SF Cooling System, Service Water System and the RW System.

Justification for Change:

CR-3 is no longer authorized to operate the reactor or place or retain nuclear fuel in the reactor vessel. Given the permanent removal of nuclear fuel from the reactor, the requirement associated with this commitment is no longer applicable. Therefore, this change does not negatively impact the ability of a SSC to perform its safety function or negatively impact the ability of plant personnel to ensure a SSC is capable of performing its intended safety function.

U. S. Nuclear Regulatory Commission Attachment 1 3F0514-08 Page 39 of 41 Nuclear Operations Commitment System (NOCS) Number: 96018 Source Document:

Babcock & Wilcox (B&W) Recommendation Tracking System #47-1163743-17, dated September 1, 1988.

Original Commitment:

Institute formal motor operated valve training programs for engineers, electricians and mechanics who perform maintenance on these valves, and for Reactor Operators who could benefit from an understanding of valve operations when confronted with off-normal events.

Modify/Inactivate Commitment:

INACTIVATE: The source of this commitment is a B&W Safety and Performance Improvement Program (SPIP) recommendation. The SPIP is a B&W system for providing technical and service information to nuclear customers pertaining to installation and operation of B&W supplied equipment. The SPIP is a formal program that was formed by the B&W Owner's Group (BWOG) and was implemented at affected sites under close scrutiny by the NRC. There are no MOVs where malfunction can impact fuel cooling, storage or handling in CR-3's permanently defueled condition. This includes the SF Cooling System, Service Water System and the RW System. There are no MOVs in any currently operating system where temperature or pressure conditions exist that are adverse to personnel safety.

Justification for Change:

CR-3 is no longer authorized to operate the reactor or place or retain fuel in the reactor vessel.

Given the permanent removal of nuclear fuel from the reactor, the requirement associated with this commitment is no longer applicable. Therefore, this change does not negatively impact the ability of a SSC to perform its safety function or negatively impact the ability of plant personnel to ensure a SSC is capable of performing its intended safety function.

U. S. Nuclear Regulatory Commission Attachment 1 3F0514-08 Page 40 of 41 Nuclear Operations Commitment System (NOCS) Number: 6638 Source Document:

CR-3 to NRC letter, 3F1285-09, dated December 16, 1985.

Original Commitment:

Florida Power Corporation will provide one time subject training, commensurate with assigned responsibilities within a reasonable time frame after appointment of personnel to a position for technicians and their direct supervision as follows:

1. Senior Nuclear I & C [Instrumentation and Control] Electrical Supervisor, Nuclear Electrical/I

& C Supervisors, Chief Nuclear Technical Support Technicians and Nuclear Technical Support Technicians.

2. Radiation Protection Manager, Health Physics Supervisors, Chief Health Physics Technicians, Health Physics Technicians.

3. Nuclear Chemistry Manager, Nuclear Chemistry Supervisors, Chief Nuclear Chemistry Technicians, Nuclear Chemistry Technicians.

Modify/Inactivate Commitment:

INACTIVATE: This commitment supports requirements included in NUREG-0737, "Clarification of TMI Action Plan Requirements." The training for personnel to recognize and react to core damage incidents and thereby minimize the resultant affects is no longer required as a result of CR-3's permanently defueled condition.

Justification for Change:

Given the permanent removal of nuclear fuel from the reactor, mitigation of core damage training is no longer required. SF safety training will be maintained by a program designed for current plant conditions. Therefore, this change does not negatively impact the ability of a SSC to perform its safety function or negatively impact the ability of plant personnel to ensure a SSC is capable of performing its intended safety function.

U. S. Nuclear Regulatory Commission Attachment 1 3F0514-08 Page 41 of 41 Nuclear Operations Commitment System (NOCS) Number: 969 Source Document:

CR-3 to NRC letter, 3F0684-06, dated June 8, 1984.

Original Commitment:

Maintenance crew training of proper bolting/stud practices, tools application, specifications and requirements is accomplished through classroom instruction and on the job training. Such training is documented on maintenance training records.

Modify/inactivate Commitment:

INACTIVATE: This commitment was made in the CR-3 response to NRC Bulletin 82-02, "Degradation of Threaded Fasteners in the Reactor Coolant Pressure Boundary of PWR Plants," to ensure the training of personnel on de-tensioning and re-tensioning practices, gasket installation, and other measures to eliminate RCS leakage during operations. This commitment is no longer applicable to CR3's permanently defueled condition.

Justification for Change:

Given the permanent removal of nuclear fuel from the reactor, this commitment is no longer applicable. Therefore, this change does not negatively impact the ability of a SSC to perform its safety function or negatively impact the ability of plant personnel to ensure a SSC is capable of performing its intended safety function.