3F0512-01, NRC Commitment Change Report - May 2012
| ML12137A354 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 05/14/2012 |
| From: | Hobbs T Progress Energy Florida |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| 3F0512-01 | |
| Download: ML12137A354 (12) | |
Text
M Progress Energy Crystal River Nuclear Plant Docket No. 50-302 Operating License No. DPR-72 Ref: 10 CFR 50.9 May 14, 2012 3F0512-01 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001
Subject:
Crystal River Unit 3 - NRC Commitment Change Report - May 2012
Dear Sir:
The purpose of this letter is to provide notification of changes to regulatory commitments contained in previously docketed correspondence from Florida Power Corporation (FPC), now doing business as Progress 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 the revised commitment, if revised, and justification for the change. 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 eighteen (18) CR-3 regulatory commitments that were modified or inactivated between January 5, 2010 and January 5, 2012, ten (10) modified or inactivated regulatory commitments meet the NEI 99-04 criteria for NRC notification.
No new regulatory commitments are made in this letter.
If you have any questions regarding this submittal, please contact Mr. Dan Westcott, Superintendent, Licensing and Regulatory Programs at (352) 563-4796.
Sincerely, Terry Hobbs Plant General Manager Crystal River Nuclear Plant TH/dwh Attachment xc: Regional Administrator, Region II Senior Resident Inspector NRR Project Manager Progress Energy Florida, Inc.
Crystal River Nuclear Plant 15760 W. Power Line Street J ,
Crystal River, FL 34428
FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302 / LICENSE NUMBER DPR-72 ATTACHMENT NRC COMMITMENT CHANGE REPORT - MAY 2012
U.S. Nuclear Regulatory Commission Attachment 3F0512-01 Page 1 of 10 Nuclear Operations Commitment System (NOCS) Number: 5645 Source Document:
Crystal River Unit 3 (CR-3) to NRC letter, 3F1085-05, dated October 11, 1985.
Original Commitment:
FPC will perform Preventive Maintenance procedure PM-118 on a twelve month basis in accordance with B&W Owner Group ATWS Committee recommendations. The six maintenance items requested will be included in PM-1 18.
Listed below are the six maintenance items:
- 1) Verification of breaker physical condition, including wiring insulation and termination, all retaining rings, pole bases, arc quencher, stationary and moveable contacts, and tightness of nuts and bolts.
- 2) Verification of the optimum freedom of the armature as specified in General Electric Service Advice 175-9.3S, Item #S1.
- 3) Verification of proper trip response time as specified in Service Advice 175-9.3S, Item #S6.
- 4) Lubrication of trip shaft and latch roller bearings with Mobile 28 lubricant.
- 5) Examination and cleaning of breaker enclosure.
- 6) Functional test of the breaker prior to returning it to service.
Modify/Inactivate Commitment:
Modify NOCS No. 5645 to include a onetime extension to the preventive maintenance performance cycle on the AC and DC Control Rod Drive [CRD] System Breakers, from 10/06/2011 to mid-2014 (projected time when CR-3 will enter Mode 5), due to the extended outage at CR-3 to repair the Reactor Building.
Justification for Change:
The six CRD trip breakers that are being maintained by PM-1 18 and PM-i 18A will be in an off-normal alignment due to the layup conditions at CR-3 for the duration of the extended outage.
The Control Rod Drive Control System (CRDCS) is not capable of rod withdrawal (the 68 6-pole individual rod disconnect breakers are all tagged open and the bulkhead connections are disconnected) and the DC breakers will be (and currently are) in the OPEN position. The AC breakers are currently also OPEN, but will be CLOSED when dry layup plans are implemented to support maintaining certain equipment energized. Therefore, with the CRDCS in the described alignment, Improved Technical Specification (ITS) 3.3.3, Reactor Protection System (RPS) - Reactor Trip Module (RTM), and ITS 3.3.4, Control Rod Drive Trip Devices, are not applicable (i.e., the CRD trip breakers are not required to be OPERABLE since they are not being relied upon to trip the reactor). As a result, there is minimal need for periodic breaker maintenance until the plant enters an alignment in which RTM and CRD trip devices are required to be OPERABLE (MODE 5) when any CRD trip breaker is in the closed position and the CRDCS is capable of rod withdrawal. This condition has been identified in Condition Report 486712. Corrective actions (CORRs) that require PM-1 18 and PM-1 18A be performed prior to entering MODE 5, such that the breakers are OPERABLE when their safety function is applicable, are classified as Mode Restraints.
U.S. Nuclear Regulatory Commission Attachment 3F0512-01 Page 2 of 10 Nuclear Operations Commitment System (NOCS) Number: 62177 Source Document:
CR-3 to NRC letter, 3F0889-14, dated August 25 1989.
Original Commitment:
- 1) The Crystal River intake canal shall be surveyed on a maximum 24 month interval. The entire length of the canal from the CR-3 intake structure to the point where the dredged portion of the canal intersects the natural bottom of the Gulf of Mexico (approximate length of 8 statute miles) shall be included in this surveillance program.
The survey shall be performed or supervised by a professional surveyor licensed in the state of Florida with expertise in hydrographic surveying.
- 2) At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, visually inspect the surface area of the intake canal at the CR-3 intake structure to determine if floating transportable materials are present that would impede safety related flow of water into the intake structure.
For the straight section of the canal west of the CR-3 extension:
A) Demonstrate that a continuous minimum width of 100 ft across the canal has a maximum bottom elevation 73' at all transects.
Modify/inactivate Commitment:
Modify NOCS No. 62177 for the straight section of the canal west of the CR-3 extension to add an alternative as described below:
B) If (A) cannot be satisfied in the barge slip and turning basin, demonstrate that a minimum continuous width of 90 ft has an maximum average bottom elevation of 72 ft at all transects west of the barge turning basin and that a minimum continuous width of 90 ft has an maximum average bottom elevation of 75 ft at all transects in the barge slip and turning basin.
Justification for Change:
The proposed change has been evaluated in Engineering Change 69107 and Calculation M10-0088. The intake canal acceptance criteria is based off of the original hurricane blowout analyses which evaluated the setdown effect of Probable Maximum Hurricane wind speeds on the existing canal dimensions. After the construction of CR-3, Units 1 and 2 began maintaining the canal bottom approximately 5 feet lower to allow for larger coal barges. Calculation M10-0088 evaluated the canal for the dimensions stated in this commitment change using the same methodology as the original bases and found that this condition is bounded by the original results. Since the evaluation was benchmarked to the original analyses and did not take credit for the entire operating margin, the change does not negatively impact the ability of the Raw Water Systems to perform their safety function.
U.S. Nuclear Regulatory Commission Attachment 3F0512-01 Page 3 of 10 Nuclear Operations Commitment System (NOCS) Number: 62621 Source Document:
CR-3 to NRC letter, 3F0897-09, dated August 20 1997.
Original Commitment:
Attachment B, Responses to NRC Questions, include the following commitments which will be incorporated into FPC procedures upon approval of TSCRN [Technical Specification Change Request Notice] 211:
Following future [Once-Through Steam Generator] inspections, tubes may be added to Table 1 as described in the response to Question 2A.
The criteria for confirmation of a tube as a first span IGA [Intergranular Attack] is described in the response to Questions 2B, 4, and 5.
Following future inspections, tubes may be removed from Table 1 as described in the response to Question 2C.
CR-3 will not leave tubes in service with first span IGA indications within one inch of the edge of the lower tube sheet secondary face (LTS+1) or within one inch of the first support plate lower edge (01-1.75).
The protocol to disposition pit like IGA is described in the response to Question 4 (A through K).
Modify/Inactivate Commitment:
Inactivate NOCS Commitment No. 62621 Justification for Change:
Tubes in the original OTSGs, with the specific degradation mechanism of first span IGA in the "B" OTSG, were previously approved to remain in service by the NRC. At that time approval was based on the OTSGs still being able to perform their safety function (remove design basis heat from the Reactor Coolant System) with those known indications in service. The replacement OTSGs differ from the old OTSGs in that the tube material is thermally treated 690 versus the mill-annealed Alloy 600 in the old OTSGs. The new material and design is more resistant to IGA and other corrosion mechanisms than the original OTSGs. ITS Amendment No. 234 has since removed the alternate repair criteria to keep a tube with first span IGA in service and so the ability of the OTSGs to continue performing their safety function is improved or at least maintained. OTSG tube integrity will still be maintained even after the ability to keep first span IGA tubes in service has been removed. Therefore, inactivation of NOCS 062621 will not negatively impact the ability of the OTSGs to perform their safety function or negatively impact the ability of licensee personnel to ensure the OTSG tubes are capable of performing their intended safety function for leakage and structural integrity.
U.S. Nuclear Regulatory Commission Attachment 3F0512-01 Page 4 of 10 Nuclear Operations Commitment System (NOCS) Number: 62639 Source Document:
CR-3 to NRC letter, 3F0897-05, dated August 21, 1997.
Original Commitment:
Work Requests related to the containment boundary will be reviewed by Engineering to determine IWE/IWL applicability and to conservatively classify components as Class CC or Class MC as required.
Modify/inactivate Commitment:
Inactivate NOCS Commitment No. 62639.
Justification for Change:
In a Federal Register, dated August 8, 1996 (61 FR 41303), the NRC amended its regulations (rule) to incorporate by reference the 1992 Edition and Addenda of Subsections IWE and IWL of Section Xl of the ASME Code. Subsections IWE and IWL give requirements for inservice inspection (ISI) of Class CC (concrete containments) and Class MC (metallic containments) of light-water-cooled power plants. The amended rule became effective on September 9, 1996; it requires the licensees to incorporate the new requirements into their ISI plans and to complete the first containment inspection within five years (i.e., no later than September 9, 2001). Any repair or replacement (R/R) activity to be performed on containments after the effective date of September 9, 1996, has to be carried out in accordance with the respective requirements of Subsections IWE and IWL.
The Original Commitment was an alternate requirement that was implemented for the Containment Inspection Rule, with full implementation of the rule scheduled for June 1, 1998.
Since CR-3 has fully implemented the Containment Inspection Rule, the alternate requirements are no longer necessary.
U.S. Nuclear Regulatory Commission Attachment 3F0512-01 Page 5 of 10 Nuclear Operations Commitment System (NOCS) Number: 100174 Source Document:
CR-3 to NRC letter, 3F0899-05, dated August 20, 1999.
Original Commitment:
Revise the OTSG Inservice Inspection surveillance procedure to include 20% sample inspection criteria for inspections subsequent to 11 R as discussed herein by September 30, 1999.
Modify/inactivate Commitment:
Inactivate NOCS Commitment No. 100174.
Justification for Change:
In letter 3F0899-05 (dated 8/20/99) Florida Power Corporation provided a response to an NRC Request for Additional Information regarding Technical Specification License Amendment Request (LAR) # 249, Revision 1. The response was in regard to a proposed Alternate Repair Criteria (ARC) for axial tube end crack-like (TEC) indications. The intent of the commitment identified in letter 3F0899-05 was to ensure that OTSG inspections after the 11 R outage will include both a 100% inspection of tubes with known TEC and a 20% sample of inservice tubes with no previously identified TEC.
Certain tubes with TEC indications in the original OTSGs were previously approved to remain in service by NRC License Amendment No. 188 issued on October 1, 1999. At that time, approval was based on the OTSGs still being able to perform their safety function (remove design basis heat from the RCS) because any in-service TECs would not degrade the RCS pressure boundary beyond allowed by the ITS. The replacement OTSGs installed by Engineering Change 63038 differ from the old OTSGs in that the tube material is thermally treated 690 versus the mill-annealed Alloy 600 in the old OTSGs. The new OTSG material and design is more resistant to stress corrosion cracking and other potential corrosion mechanisms than the original OTSGs. The NRC-approved ITS Amendment No. 234 has since removed the alternate repair technique which allowed TEC indications to remain in service in the old OTSGs. Also, the replacement OTSGs are not allowed to have TECs remaining in service. However, even without the ability to leave TEC indications in service, the new OTSGs will continue to perform their safety function since any tube with unacceptable degradation will have to be removed from service by plugging. OTSG tube integrity will still be maintained without the need to keep TEC indications in service. Therefore, inactivation of NOCS 100174 will not negatively impact the ability of an OTSGs to perform its' safety function or negatively impact the ability of licensee personnel to ensure the OTSG tubes are capable of performing their intended safety functions of leakage and structural integrity at normal and accident conditions.
U.S. Nuclear Regulatory Commission Attachment 3F0512-01 Page 6 of 10 Nuclear Operations Commitment System (NOCS) Number: 100286 Source Document:
CR-3 to NRC letter, 3F0601-07, dated June 28, 2001.
Original Commitment:
Following each inservice inspection of steam generator tubes, but prior to returning the CR-3 steam generators to service, FPC will verbally notify the NRC of the following: (1) Number of tubes with circumferential cracking indications inboard of the roll repair; (2) Number of tubes with circumferential cracking indications in the original roll region, including the zone adjacent to the tube-to-sheet seal weld if no re-roll is present; and, (3) Determination of the best-estimate total leakage that would result from an analysis of the limiting Large Break Loss-of-Coolant Accident (LBLOCA) based on as-found circumferential cracking in the original tube-to-tubesheet rolls, tube-to-tubesheet re-roll repairs and the zones adjacent to the seal welds.
Modifyllnactivate Commitment:
Inactivate NOCS Commitment No. 100286.
Justification for Change:
Tubes in the original OTSGs tubesheets were previously approved to remain in service with repair rolls in the upper and lower tubesheet by NRC License Amendment No. 198. At that time, approval was based on the OTSGs still being able to perform their safety function (remove design basis heat from the RCS) with those re-rolls because the new re-rolls created a new pressure boundary and essentially removed any tube end upper or lower tubesheet eddy current indications from service. The replacement OTSGs differ from the old OTSGs in that the tube material is thermally treated 690 versus the mill-annealed Alloy 600 in the old OTSGs. The new OTSG material and design is more resistant to stress corrosion cracking and other potential corrosion mechanisms than the original OTSGs. ITS Amendment No. 234 has since removed the alternate repair technique which allowed the installation of repair rolls in the new OTSGs. Even without the ability to perform repair rolls the new OTSGs will continue to perform their safety function since any tube with unacceptable degradation will have to be removed from service by plugging. OTSG tube integrity will still be maintained without the ability of repair rolls to be used to keep an otherwise good tube in service. Therefore, inactivation of NOCS 100286 will not negatively impact the ability of the OTSGs to perform their safety function or negatively impact the ability of licensee personnel to ensure the OTSG tubes are capable of performing their intended safety function for leakage and structural integrity.
U.S. Nuclear Regulatory Commission Attachment 3F0512-01 Page 7 of 10 Nuclear Operations Commitment System (NOCS) Number: 100288 Source Document:
CR-3 to NRC letter, 3F0601-07, dated June 28, 2001.
Original Commitment:
Demonstrate that the primary-to-secondary leakage following a LBLOCA, as described in Appendix A to Topical Report BAW-2374, Revision 1, is acceptable based on the as-found condition of the steam generators. This is required to demonstrate that adequate margin and defense-in-depth are maintained. For the purpose of this evaluation, "acceptable" means a best estimate of the leakage expected due to a LBLOCA where that leakage would not result in a significant increase of radionuclide release (e.g., in excess of 10 CFR 100 limits). A summary of this evaluation shall be provided to the NRC following the completion of steam generator tube inservice inspection with the report required by Improved Technical Specification 5.7.2.e.
Modify/inactivate Commitment:
Inactivate NOCS Commitment No. 100288.
Justification for Change:
Tubes in the original OTSGs tubesheets were previously approved to remain in service with repair rolls in the upper and lower tubesheet by NRC License Amendment No. 198. At that time, approval was based on the OTSGs still being able to perform their safety function (remove design basis heat from the RCS) with those re-rolls because the new re-rolls created a new pressure boundary and essentially removed any tube end upper or lower tubesheet eddy current indications from service. The replacement OTSGs differ from the old OTSGs in that the tube material is thermally treated 690 versus the mill-annealed Alloy 600 in the old OTSGs. The new OTSG material and design is more resistant to stress corrosion cracking and other potential corrosion mechanisms than the original OTSGs. NRC approved ITS Amendment No.
234 has since removed the alternate repair technique which allowed the installation of repair rolls in the old OTSGs. Also, the replacement OTSGs are not allowed to have reroll repairs performed on them. Even without the ability to perform repair rolls, the new OTSGs will continue to perform their safety function since any tube with unacceptable degradation will have to be removed from service by plugging. OTSG tube integrity will still be maintained without the ability for a repair rolls to be used to keep an otherwise good tube in service. Therefore, inactivation of NOCS 100288 will not negatively impact the ability of the OTSGs to perform their safety function or negatively impact the ability of licensee personnel to ensure the OTSG tubes are capable of performing their intended safety functions of leakage and structural integrity at normal and accident conditions.
U.S. Nuclear Regulatory Commission Attachment 3F0512-01 Page 8 of 10 Nuclear Operations Commitment System (NOCS) Number: 100348 Source Document:
CR-3 to NRC letter, 3F0601-07, dated June 28, 2001.
Original Commitment:
Demonstrate that the primary-to-secondary leakage following a LBLOCA, as described in Appendix A to Topical Report BAW-2374, Revision 1, is acceptable based on the as-found condition of the steam generators. This is required to demonstrate that adequate margin and defense-in-depth are maintained. For the purpose of this evaluation, "acceptable" means a best estimate of the leakage expected due to a LBLOCA where that leakage would not result in a significant increase of radionuclide release (e.g., in excess of 10 CFR 100 limits). A summary of this evaluation shall be provided to the NRC following the completion of steam generator tube inservice inspection with the report required by Improved Technical Specification 5.7.2.e 90 days after breaker closure following restart.
Modifyllnactivate Commitment:
Inactivate NOCS Commitment No. 100348.
Justification for Change:
Tubes in the original OTSGs tubesheets were previously approved to remain in service with repair rolls in the upper and lower tubesheet by NRC License Amendment No. 198. At that time, approval was based on the OTSGs still being able to perform their safety function (remove design basis heat from the RCS) with those re-rolls because the new re-rolls created a new pressure boundary and essentially removed any tube end upper or lower tubesheet eddy current indications from service. The replacement OTSGs differ from the old OTSGs in that the tube material is thermally treated 690 versus the mill-annealed Alloy 600 in the old OTSGs. The new OTSG material and design is more resistant to stress corrosion cracking and other potential corrosion mechanisms than the original OTSGs. NRC approved ITS Amendment No.
234 has since removed the alternate repair technique which allowed the installation of repair rolls in the old OTSGs. Also, the replacement OTSGs are not allowed to have reroll repairs performed on them. Even without the ability to perform repair rolls, the new OTSGs will continue to perform their safety function since any tube with unacceptable degradation will have to be removed from service by plugging. OTSG tube integrity will still be maintained without the ability for a repair rolls to be used to keep an otherwise good tube in service. Therefore, inactivation of NOCS 100348 will not negatively impact the ability of the OTSGs to perform their safety function or negatively impact the ability of licensee personnel to ensure the OTSG tubes are capable of performing their intended safety functions of leakage and structural integrity at normal and accident conditions.
U.S. Nuclear Regulatory Commission Attachment 3F0512-01 Page 9 of 10 Nuclear Operations Commitment System (NOCS) Number: 100402 Source Document:
CR-3 to NRC letter, 3F0601-07, dated June 28, 2001.
Original Commitment:
Following each inservice inspection of steam generator tubes, but prior to returning the CR-3 steam generators to service (prior to Mode 4), FPC will verbally notify the NRC of the following:
A. Number of tubes with circumferential cracking indications inboard of the roll repair.
B. Number of tubes with circumferential cracking indications in the original roll region, including the zone adjacent to the tube-to-tubesheet seal weld if no reroll is present.
C. Determination of the Best-Estimate Total Leakage that would result from an analysis of the limiting large break loss-of-coolant accident (LBLOCA) based on as-found circumferential cracking in the original tube-to-tubesheet rolls, tube-to-tubesheet re-roll repairs, and the zones adjacent to the seal welds."
Modify/inactivate Commitment:
Inactivate NOCS Commitment No. 100402.
Justification for Change:
Tubes in the original OTSGs tubesheets were previously approved to remain in service with repair rolls in the upper and lower tubesheet by NRC License Amendment No. 198. At that time, approval was based on the OTSGs still being able to perform their safety function (remove design basis heat from the RCS) with those re-rolls because the new re-rolls created a new pressure boundary and essentially removed any tube end upper or lower tubesheet eddy current indications from service. The replacement OTSGs differ from the old OTSGs in that the tube material is thermally treated 690 versus the mill-annealed Alloy 600 in the old OTSGs. The new OTSG material and design is more resistant to stress corrosion cracking and other potential corrosion mechanisms than the original OTSGs. NRC approved ITS Amendment No.
234 has since removed the alternate repair technique which allowed the installation of repair rolls in the old OTSGs. Also, the replacement OTSGs are not allowed to have reroll repairs performed on them. Even without the ability to perform repair rolls, the new OTSGs will continue to perform their safety function since any tube with unacceptable degradation will have to be removed from service by plugging. OTSG tube integrity will still be maintained without the ability for a repair rolls to be used to keep an otherwise good tube in service. Therefore, inactivation of NOCS 100402 will not negatively impact the ability of the OTSGs to perform their safety function or negatively impact the ability of licensee personnel to ensure the OTSG tubes are capable of performing their intended safety functions of leakage and structural integrity at normal and accident conditions.
U.S. Nuclear Regulatory Commission Attachment 3F0512-01 Page 10 of 10 Nuclear Operations Commitment System (NOCS) Number: 100403 Source Document:
CR-3 to NRC letter, 3F0601-07, dated June 28, 2001.
Original Commitment:
Demonstrate that the primary-to-secondary leakage following a LBLOCA, as described in Appendix A to Topical Report BAW-2374, Revision 1, is acceptable based on the as-found condition of the steam generators. This is required to demonstrate that adequate margin and defense-in-depth are maintained. For the purpose of this evaluation, "acceptable" means a best estimate of the leakage expected due to a LBLOCA where that leakage would not result in a significant increase of radionuclide release (e.g., in excess of 10 CFR 100 limits). A summary of this evaluation shall be provided to the NRC following the completion of steam generator tube inservice inspection with the report required by Improved Technical Specification 5.7.2.e 90 days after breaker closure following restart."
Modify/inactivate Commitment:
Inactivate NOCS Commitment No. 100403.
Justification for Change:
Tubes in the original OTSGs tubesheets were previously approved to remain in service with repair rolls in the upper and lower tubesheet by NRC License Amendment No. 198. At that time, approval was based on the OTSGs still being able to perform their safety function (remove design basis heat from the RCS) with those re-rolls because the new re-rolls created a new pressure boundary and essentially removed any tube end upper or lower tubesheet eddy current indications from service. The replacement OTSGs differ from the old OTSGs in that the tube material is thermally treated 690 versus the mill-annealed Alloy 600 in the old OTSGs. The new OTSG material and design is more resistant to stress corrosion cracking and other potential corrosion mechanisms than the original OTSGs. NRC approved ITS Amendment No.
234 has since removed the alternate repair technique which allowed the installation of repair rolls in the old OTSGs. Also, the replacement OTSGs are not allowed to have reroll repairs performed on them. Even without the ability to perform repair rolls, the new OTSGs will continue to perform their safety function since any tube with unacceptable degradation will have to be removed from service by plugging. OTSG tube integrity will still be maintained without the ability for a repair rolls to be used to keep an otherwise good tube in service. Therefore, inactivation of NOCS 100403 not negatively impact the ability of an OTSGs to perform its' safety function or negatively impact the ability of licensee personnel to ensure the OTSG tubes are capable of performing their intended safety functions of leakage and structural integrity at normal and accident conditions.