ML15218A158
| ML15218A158 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 11/21/1997 |
| From: | Labarge D NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML15218A159 | List: |
| References | |
| NUDOCS 9711280191 | |
| Download: ML15218A158 (14) | |
Text
0 1 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20556-0001 DUKE ENERGY CORPORATION DOCKET NO. 50-269 OCONEE NUCLEAR STATION. UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 227 License No. DPR-38
- 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment to the Oconee Nuclear Station, Unit 1 (the facility)
Facility Operating License No. DPR-38 filed by the Duke Energy Corporation (the licensee) dated October 20, 1997, as supplemented by letters dated November 3, 6, and 10, 1997, comply with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
9711290191 971121 PDR ADOCK 05000269 P
-2
- 2.
Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 3.B of Facility Operating License No. DPR-38 is hereby amended to read as follows:
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 227, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
- 3.
This license amendment is effective as of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION H Arbert N.BerkowDirector Project Directorate 11-2 Division of Reactor Projects - /Il Office of Nuclear Reactor Regulation
Attachment:
Technical Specification Changes Date of Issuance:
November 21, 1997
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 2055-0o1 DUKE ENERGY CORPORATION DOCKET NO. 50-270 OCONEE NUCLEAR STATION. UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 227 License No. DPR-47 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment to the Oconee Nuclear Station, Unit 2 (the facility)
Facility Operating License No. DPR-47 filed by the Duke Energy Corporation (the licensee) dated October 20, 1997, as supplemented by letters dated November 3, 6, and 10, 1997, comply with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
-2
- 2.
Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 3.B of Facility Operating License No. DPR-47 is hereby amended to read as follows:
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.
227, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
- 3.
This license amendment is effective as of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION He ert N. Berkow, Director Project Directorate 11-2 Division of Reactor Projects - 1/Il Office of Nuclear Reactor Regulation
Attachment:
Technical Specification Changes Date of Issuance:
November 21, 1997
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 205-0001 DUKE ENERGY CORPORATION DOCKET NO. 50-287 OCONEE NUCLEAR STATION. UNIT 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 224 License No. DPR-55
- 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment to the Oconee Nuclear Station, Unit 3 (the facility)
Facility Operating License No. DPR-55 filed by the Duke Energy Corporation (the licensee) dated October 20, 1997, as supplemented by letters dated November 3, 6, and 10, 1997, comply with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
-2
- 2.
Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 3.B of Facility Operating License No. DPR-55 is hereby amended to read as follows:
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.
224, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
- 3.
This license amendment is effective as of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION erbert N. Berkow, Director Project Directorate 11-2 Division of Reactor Projects - I/I Office of Nuclear Reactor Regulation
Attachment:
Technical Specification Changes Date of Issuance:
November 21, 1997
ATTACHMENT TO LICENSE AMENDMENT NO. 227 FACILITY OPERATING LICENSE NO, DPR-38 DOCKET NO. 50-269 TO LICENSE AMENDMENT NO. 227 FACILITY OPERATING LICENSE NO. DPR-47 DOCKET NO. 50-270 TO LICENSE AMENDMENT NO. 224 FACILITY OPERATING LICENSE NO. DPR-55 DOCKET NO. 50-287 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by Amendment number and contain vertical lines indicating the areas of change.
Remove Pages Insert Paaes 3.1-14 3.1-14 4.17-1 4.17-1 4.17-2 4.17-2 4.17-3 4.17-3 4.17-4 4.17-4 4.17-5 4.17-5 4.17-5a
3.1.6 Leakage Specification 3.1.6.1 If the total reactor coolant leakage rate exceeds 10 gpm, the reactor shall be shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of detection.
3.1.6.2 If unidentified reactor coolant leakage'(excluding normal evaporative losses) exceeds I gpm or if any reactor coolant leakage is evaluated as unsafe, the reactor shall be shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of detection.
3.1.6.3 If any reactor coolant leakage exists through a non-isolable fault in a RCS strength boundary (such as the reactor vessel, piping, valve body, etc., except the steam generator tubes), the reactor shall be shutdown, and cooldown to the cold shutdown condition shall be initiated within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of detection.
3.1.6.4 If the total leakage through the tubes of any one steam generator equals or exceeds 150 gallons per day, a reactor shutdown shall be initiated within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and the reactor shall be in a cold condition within the next 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
3.1.6.5 If reactor shutdown is required by Specification 3.1.6.1, 3.1.6.2 or 3.1.6.3, the rate of shutdown and the conditions of shutdown shall be determined by the safety evaluation for each case and justified in writing as soon thereafter as practicable.
3.1.6.6 Action to evaluate the safety implication of reactor coolant leakage shall be initiated within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of detection. The nature, as well as the magnitude, of the leak shall be considered in this evaluation. The safety evaluation shall assure that the exposure of offsite personnel to radiation is within the guidelines of 10 CFR 20.
3.1.6.7 If reactor shutdown is required per Specification 3.1.6.1, 3.1.6.2, 3.1.6.3 or 3.1.6.4, the reactor shall not be restarted until the leak is repaired or until the problem is otherwise corrected.
3.1.6.8 When the reactor is critical and above 2% power, two reactor coolant leak detection systems of different operating principles shall be operable, with one of the two systems sensitive to radioactivity. The systems sensitive to radioactivity may be out-of-service for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> provided two other means to detect leakage are operable.
3.1.6.9 Loss of reactor coolant through reactor coolant pump seals and system valves to connecting systems which vent to the gas vent header and from which coolant can be returned to the reactor coolant system shall not be considered as reactor coolant leakage and shall not be subject to the consideration of Specifications 3.1.6.1, 3.1.6.2, 3.1.63.
3.1.6.4, 3.1.6.5, 3.1.6.6 or 3.1.6.7 except that such losses when added to leakage shall not exceed 30 gpm.
3.1.6.10
- a.
The maximum allowable leakage for valves CF-12, CF-14, LP47 and LP-48 shall be as follows:
Oconee 1. 2, and 3 3.1-14 Amendment No.227(Unit 1) z Amendment No227(Unit 2)
Amendment No224(Unit 3)
.7 1
4.17 STEAM GENERATOR TUBING SURVEILLANCE Applicability Applies to the surveillance of tubing of each steam generator.
Objective To ensure integrity of the steam generator tubing through a defined inservice surveillance program, and to minimize exposure of personnel to radiation during performance of the surveillance program.
Specification 4.17.1 Examination Methods Inservice inspection of steam generator tubing shall include non-destructive examination by eddy-current testing or other equivalent techniques. The inspection equipment shall provide a sensitivity that will detect defects with a penetration of 20 percent or more of the minimum allowable as-manufactured tube wall thickness.
4.17.2 Acceptance Criteria The steam generator shall be considered operable after completion of the specified actions. All tubes examined exceeding the repair limit shall be repaired by sleeving or rerolling or removed from service (e.g., plugged, stabilized).
4.17.3 Selection and Testing The steam generator tube minimum sample size, inspection result classification, and the corresponding action required shall be as specified in Table 4.17.1. The inservice inspection of steam generator tubes shall be performed at the frequencies specified in Specification 4.17.4 and the inspected tubes shall be verified acceptable per Specification 4.17.5. The tubes selected for each inservice inspection shall include at least 3% of the total number of tubes in both steam generators, with one or both steam generators being inspected. The tubes selected for these inspections shall be selected on a random basis except:
- a.
The first sample inspection during each inservice inspection of each steam generator shall include:
- 1.
All tubes that previously had detectable wall penetrations (>20%) and have not been plugged or sleeve repaired in the affected area.
- 2.
At least 50% of the tubes inspected shall be in those areas where experience has indicated potential problems.
- 3.
A tube adjacent to any selected tube which does not permit passage of the eddy current probe for tube inspection.
Oconee 1, 2, and 3 4.17-1 Amendment No. 22 7(Unit 1)
Amendment No.227 (Unit 2)
Amendment No.224(Unit 3)
- b.
Tubes in the following Group(s) may be excluded from the first sample if all tubes in a Group in both OTSG are inspected. No credit will be taken for these tubes in meeting minimum sample size requirements.
(1)
Group A-i: Tubes within one, two, or three rows of the open inspection lane.
- c.
All tubes which have been repaired using the reroll process will have the new roll area inspected during the inservice inspection.
- d.
The tubes selected as the second and third samples (if required by Table 4.17-1) during each inservice inspection may be subjected to less than a full tube inspection provided:
- 1.
The tubes selected for these samples include the tubes from those areas of the tubesheet array where tubes with imperfections were previously found.
- 2.
The inspections include those portions of the tubes where imperfections were previously found.
The results of each sample inspection shall be classified into one of the following three categories:
CategorY Inspection Results C-1 Less than 5% of the total tubes inspected are degraded tubes and none of the inspected tubes are defective.
C-2 One or more tubes, but no more than 1% of the total tubes inspected are defective, or between 5% and 10% of the total tubes inspected are degraded tubes.
C-3 More than 10% of the total tubes inspected are degraded tubes or more than I % of the inspected tubes are defective.
NOTES:
(1)
In all inspections, previously degraded tubes must exhibit significant
(>10%) further wall penetrations to be included in the above percentage calculations.
(2)
Where special inspections are performed pursuant to 4.17.3.b, defective or degraded tubes found as a result of the inspection shall be included in determining the Inspection Results Category for that special inspection but need not be included in determining the Inspection Results Category for the general steam generator inspec tion, unless the mechanism of degradation is random in nature.
(3)
Where special inspections are performed pursuant to 4.17.3.c, defective or degraded tube indications found in the new roll area as a result of the inspection and any indications found in the originally rolled region of the rerolled tubes, need not be included in determining the Inspection Results Category for the general steam generator inspection.
Oconee 1, 2, and 3 4.17-2 Amendment No.2 2 7(Unit 1)
Amendment No.227(Unit 2)'
Amendment No.224(Unit 3)
4.17.4 Inspection Intervals The above required inservice inspections of steam generator tubes shall be performed at the following frequencies.
- a.
Inservice inspections shall be performed at intervals of not less than 12 nor more than 24 calen'dar months after the previous inspection. If the results of two consecutive inspections following service under all volatile treatment (AVT) conditions fall into the C-I category or if two consecutive inspections demonstrate that previously observed degradation has not continued and no additional degradation has occurred, the inspection interval may be extended to a maximum of 40 months.
- b.
If the results of the inservice inspection of a steam generator performed in accordance with Table 4.17-1 at 40 month intervals fall in Category C-3, subsequent inservice inspections shall be performed at intervals of not less than 10 months nor more than one fuel cycle after the previous inspection. The increase in inspection frequency shall apply until a subsequent inspection meets the conditions specified in 4.17.4.a and the interval can be extended to a maximum of 40 months.
C.
Additional, unscheduled inservice inspections shall be performed on each steam generator in accordance with the first sample inspection specified in Table 4.17-1 during the shutdown subsequent to any of the following conditions:
- 1.
A seismic occurrence greater than the Operating Basis Earthquake,
- 2.
A loss-of-coolant accident requiring actuation of the engineered safeguards, or
- 3.
A main steam line or feedwater line break.
- d.
After primary to secondary leakage in excess of the limits of Specification 3.1.6, an inspection of the affected steam generator will be performed in accordance with the following criteria:
- 1.
If the leaking tube is in a Group as defined in Section 4.173.b, all of the tubes in this Group in this steam generator will be inspected. If the results of this inspection fall into the C-3 category, additional inspections will be performed in the same Group in the other steam generator.
- 2.
If the leaking tube has been repaired by the reroll process and is leaking in the new roll area, all of the tubes in the steam generator that have been repaired by the reroll process will have the new roll area inspected. If the results of this inspection fall into the C-3 category, additional inspections will be performed in the new roll area in the other steam generator.
- 3.
If the leaking tube is not in a Group as defined in 4.17.4.d.1, then an inspection will be performed on the affected steam generator in accordance with Table 4.17-1 with an initial inspection sample size of 6% of the tubes in the affected steam generator.
Oconee 1, 2, and 3 4.17-3 Amendment No.227(Unit 1)
Amendment No.227(Unit 2)
Amendment No. 224(Unit 3)
4.17.5 Definitions As used in this specification:
- a.
Imperfection means an exception to the dimensions, finish or contour of a tube from that required by fabrication drawings or specifications. Eddy-current testing indications below 20%
of the nominal tube or sleeve wall thickness, if detectable, may be considered as imperfections.
- b.
Derradation means a service-induced cracking, wastage, wear or general corrosion occurring on either the inside or outside of a tube or a sleeve.
- c.
Degraded Tube means a tube or a sleeve containing imperfections 220% of the nominal wall thickness caused by degradation.
- d.
% Degradation means the percentage of the tube or sleeve wall thickness affected or removed by degradation.
- e.
Defect means an imperfection of such severity that it exceeds the repair limit. A tube or sleeve containing a defect is defective.
- f.
Repair Limit means the imperfection depth beyond which the tube shall be either removed from service by plugging or repaired by sleeving or rerolling because it may become unserviceable prior to the next inspection; it is equal to 40% of the nominal tube or sleeve wall thickness.
The Babcock and Wilcox process (or method) equivalent to the method described in report, BAW-1 823P, Revision I will be used for sleeving repairs.
The rerolling repair process will only be used to repair tubes with defects in the upper tubesheet area. The rerolling repair process will be performed only once per steam generator tube using a l inch reroll length. The new roll area must be free of degradation in order for the repair to be considered acceptable. The rerolling process used by Oconee is described in the topical report, BAW-2303P, Revision 3.
- g.
Unserviceable describes the condition of a tube if it leaks or contains a defect large enough to affect its structural integrity in the event of an Operating Basis Earthquake, a loss-of-coolant accident, or a steam line or feedwater line break as specified in Specification 4.17A.
b1.
Tube Inspection means an inspection of the steam generator tube from the point of entry completely to the point of exit. The degraded tube above the new roll area can be excluded from future periodic inspection requirements because it is no longer part of the pressure boundary once the repair roll is installed.
4.17.6 Reports
- a.
The number of tubes plugged or repaired in each steam generator shall be reported to the NRC within 30 days following the completion of the plugging or repair procedure.
Oconee 1, 2, and 3 4.17-4 Amendment No.227(Unit 1)
Amendment No.227(Unit 2)
Amendment No.224(Unit 3)
- b.
The results of the steam generator tube inservice inspection shall be reported to the NRC within 3 months following completion of the inspection. This report shall include:
- 1.
Number and extent of tubes inspected.
- 2.
Location and percent of wall-thickness penetration for each indication of a degraded tube.
- 3.
Identification of tubes plugged or repaired.
- 4.
Number of tubes repaired by rerolling and number of indications detected in the new roll area of the repaired tubes.
- c.
Results of steam generator tube inspections which fall into Category C-3 and require prompt notification of the NRC shall be reported pursuant to Specification 6.6.2.1.a prior to resumption of plant operation. The written followup of this report shall provide a description of investigations conducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence.
The program of periodic inservice inspection of steam generators provides the means to monitor the integrity of the tubing and to maintain surveillance in the event there is evidence of mechanical damage or progressive deterioration due to design, manufacturing errors, or operating conditions. Inservice in spection of the steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures may be taken.
Repair or removal from service will be required for any tube with service-induced metal loss in excess of 40% of the tube or sleeve nominal wall thickness or with a through wall crack. Additional corrective actions may be required to stabilize a circumferentially cracked tube.
The initial sample of tubes inspected in a steam generator includes tubes from three groups. First, lane tubes are inspected to assure their integrity. Second, all other inservice tubes with degradation, inspected in previous inspections, are inspected to assure tube integrity and determine degradation growth, if any.
Third, a random sample of 3% of the total number of tubes in both steam generators is inspected. The results of the latter inspection dictate the extent of further examinations.
An objective of this Specification is to provide an inspection plan which will insure, with a high degree of confidence, that no more than 30 defective tubes will remain in a steam generator after an initial C-3 category inspection.
Following an 18% random inspection (C-3 category inspection) an unaffected area is identified. The unaffected area will be logically and consistently defined based on generator design, defect location and characteristics. The criteria for accepting an area as unaffected depend on the number of defects found in the sample inspected in that area and are established such that there is a 0.05 or smaller probability of accepting the area as unaffected if it contains 30 or more defective tubes.
Experience with Babcock and Wilcox steam generators has indicated that tubes near the open inspection lane are susceptible to forms of degradation unique to that area. Therefore, tubes within one, two, or three rows of the inspection lane have been defined as a special group. If all of these tubes are inspected Oconee 1, 2, and 3 4.17-5 Amendment No.22 7(Unit 1)
Amendment No.227(Unit 2)
Amendment No.224(Unit 3)
in both steam generators, no credit will be taken for them in meeting minimum sample size requirements and the results of their inspection will not be used in classifying the results of the general inspection into C-1, C-2 or C-3 categories, unless the mechanism of tube degradation is random in nature. Random degradation mechanisms are those which based on location, steam generator design and operation, and operating experience cannot logically and consistently be shown as limited to a local area.
The affected area will be 100% inspected to assure all defective tubes therein are identified and either removed from service or repaired by sleeving. NRC concurrence in this determination is required prior to completion of the inspection.
Degraded steam generator tubes can be repaired by the installation of sleeves which span the area of degradation and serve as a replacement pressure boundary for the degraded portion of the tube, thus permitting the tube to remain in service. An additional repair method for degraded steam generator tubes consists of rerolling the tubes in the upper tube sheet to create a new roll area and pressure boundary for the tube. The rerolling method will ensure that the area of degradation will not serve as a pressure boundary, thus permitting the tube to remain in service. The degraded tube above the new roll area can be excluded from future periodic inspection requirements because it is no longer part of the pressure boundary once the repair roll is installed in the upper tube sheet.
All tubes which have been repaired using the reroll process will have the new roll area inspected during the inservice inspection. Defective or degraded tube indications found in the new roll area as a result of the inspection of the new roll and any indications found in the originally rolled region of the rerolled tube need not be included in determining the Inspection Results Category for the general steam generator inspection.
The rerolling repair process will only be used to repair tubes with defects in the upper tubesheet area.
The rerolling repair process will be performed only once per steam generator tube using a l inch reroll length. Thus, multiple applications of the rerolling process to any individual tube is not acceptable. The new roll area must be free of degradation in order for the repair to be considered acceptable. After the new roll area is initially deemed acceptable, future degradation in the new roll area will be analyzed to determine if the tube is defective and needs to be removed from service. The rerolling process used by Oconee is described in the topical report, BAW-2303P, Revision 3.
This inspection plan enables exposures to be maintained as low as reasonably achievable to the personnel involved in the inspection and assured that generator areas with significant numbers of degraded tubes are adequately inspected.
Oconee 1, 2, and 3 4.17-5a Amendment No. 227(Unit 1)
Amendment No. 227(Unit 2)
Amendment No. 224(Unit 3)