ML20070K193
| ML20070K193 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 03/04/1991 |
| From: | Matthews D Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20070K197 | List: |
| References | |
| NUDOCS 9103180337 | |
| Download: ML20070K193 (10) | |
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UNITED STATES f
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NUCLEAR REGULATORY COMMISSION j
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I DUKE POWER COMPANY NW'TH CAROLINA ELECTRIC MEMBERSHIP CORPORATION SALUDA RIVER ELECTRIC COOPERATIVE, INC.
DOCKET NO. 50-413 CATAWBA NUCLEAR STATION, UNIT 1 AMEN 0 MENT TO FACILITY OPERATING LICENSE Amendment No. 84 License No. NPF-35 1.
The Nuclear Regulatory Comission (the Comission) ilas found that:
A.-
The application for amendment to the Catawba Nuclear Station, Unit 1 (the facility) Facility Operating License No. NPF-35 filed by the Duke Power Company, acting for itself, North Carolina Electric Membership Corporation and Saluda River Electric Cooperative, Inc.
(licenseet) dated December 19, 1990, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),
and the Comissic"'s rules and regulations as set forth in 10 CFR Chapter 1 B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the
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Comission 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 Comission's regulations set forth in 10 CFR Chapter 1 D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the publics and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.
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-2 2.
Accordingly, the license is hereby ;. mended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 2.C.(2) of Facility Operating License No. NPF-35 is hereby amended to read as follows:
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 84
, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. Duke Power Company shall operate the facility in accordance with the Tect il Specifications and the Environmental Protection Plan.
3.
This license amendment is effective as of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
'd 4'M 0 vid B. Matthews, Director Project Directorate 11-3 Division of Reactor Projects-!/II Of fice of Nuclear Reactor Regulation
Attachment:
Technical Specification Changes Date of Issuances March 4. 1991
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UNITED STATES
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NUCLEAR REGULATORY COMMISSION WASHINGTON D C. 20$65 j
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DUKE POWER COMPANY NORTH CAROLINA MUNICIPAL POWER AGENCY NO. 1 PIEDMONT MUNICIPAL POWER AGENCY DOCKET NO. 50-414 CATAWBA NUCLEAR STATION, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 78 License No. NPF-52 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment to the Catawba Nuclear Station, Unit 2 (the facility) Facility Operating License No. NPF-52 filed by the Duke Power Company, acting for itself, North Carolina Munici Agency No. 1 and Piedmont Municipal Power Agency (licensees) pal Power dated December 19, 1990, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations as set forth in 10 CFR Chapter Is B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission L
C.
There is re**onable assurance (i) that the activities authorized by i
this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be 1
conducted in compliance with the Comission's regulations set forth in 10 CFR Chapter I D.
The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and i
E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.
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. 2.
Accordingly, the license is hereby amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 2.C.(2) of Ft.cility Operating License No. NPP-52 is hereby amended to read as follows:
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 78, and the Environmental Protection Plan contained in Appendix 0, both of which are attached hereto, are hereby incorporated into this license.
Duke Power Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This license amendment is effective as of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION 7
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David B.'%
Matthews, Director Project Directorate !!-3 Division of Reactor Projects-I/II Office of Nuclear Reactor Regulation
Attachment:
Technical Specification Changes Date of Issuance: March 4,1991 I
ATTACHMENT TO LICENSE AMEN 0 MENT NO. 84 FACILITY OPERATING LICENSE NO. NPF-35 DOCKET NO. 50-413 AND TO LICENSE AMEN 0 MENT NO, 78 FACILITY OPERATING LICENSE NO. nPF-62 DOCKET NO. 50-414 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 Pages 3/4 4-15 3/4 4-15 3/4 4-16 3/4 4-16 3/4 4-16a 3/4 4-16a 03/4 4-3 B 3/4 4-3 0 3/4 4-3a B 3/4 4-4 8 3/4 4-4 i
i I
REACTOR COOLANT SYSTEN SURVEILLANCE REQUIREMENTS (Continued) 4.4.5.4 Acceptance Criteria a.
As used in this specification:
1)
Imperfection means an exception to the dimensions, finish or contour of a tube or sleeve from that required by fabrication l
drawings or specifications.
Eddy-current testing indications below 20% of the nominal tube or sleeve wall thickness, if l
detectable, may be considered as imperfections; 2)
Degradation means a service-induced cracking, wastage, wear or general corrosion occurring on either inside or outside of a tube or sleeve; I
3)
Degraded Tube means a tube or sleeve containing imperfections greater than or equal to 20% of the nominal tube or sleeve wall thickness caused by degradation; 4)
% Degradation means the percentage of the tube or sleeve wall I
thickness affected or removed by degradation; 5)
Defect means an imperfection of such severity that it exceeds tiie repair limit.
A tube or sleeve containing a defect is l
defective; 6)
Rc3 air Limit means the imperfection depth at or beyond which the tuae shall be removed from service by plugging or repaired by sleeving.
It also means the imperfection depth at or beyond which a sleeved tube shall be plugged.
The repair limit is equal to 40% of the nominal tube or sleeve wall thickness.
For Unit 1, this definition does not apply to the region of the tube subject to the alternate tube plugging criteria.
If a tube is sleeved due to degradation in the F* distance, then any defects found in the tube below the sleeve will not necessi-tate plugging.
The Babcock & Wilcox process described in Topical Report BAW-2045(P)-A will be used for sleeving.
1 7)
Unserviceable describes the condition of a tube if it leaks or
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contains a defect large enough to affect its structural integ-rity in the event of an Operating Basis Earthquake, a loss-of-coolant accident, or a steam line or feedwater line break as specified in 4.4.5.3c., above; 8)
Tube Inspection means an inspection of the steam generator tube from the point of entry (hot leg side) completely around the U-bend to the top support of the cold leg; CATAWBA - UNITS 1 & 2 3/4 4-15 Amendment No. 84 (Unit 1)
Amendment No. 78 (Unit 2)
REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 9)
Preservice Inspection means an inspection of the full length of each tube in each steam generator performed by eddy current techniques prior to service to establish a baseline condition of the tubing.
This inspection shall be performed prior to initial POWER OPERATION using the equipment and techniques expected to be used during tubsequent inservice inspections.
- 10) Tube Roll Expansion is that portion of a tube which has been increased in diameter by a rolling process such that no crevice exists between the outside diameter of the tube and the tubesheet.
11)
F* Distance is the minimum length of the roll expanded portion of the tube which cannot contain any defects in order to ensure the tube does not pull out of the tubesheet.
The F* distance is 1.60 inches and is measured from the bottom of the roll expansion transition or the top of the tubesheet if the bottom of the roll expansion is above the top of the tubesheet.
Included in this distance is a safety factor of 3 plus a 0.5 inch eddy current vertical measurement uncertainty.
- 12) Alternate tube plugging criteria does not require the tube to be removed from service or repaired when the tube degradation exceeds the repair limit so long as the degradation is in l
that portion of the tube from F to the bottom of the tubesheet.
This definition does not apply to tubes with degradation (i.e., indications of cracking) in the F*
distance.
I b.
The steam generator shall be determined OPERABLE after completing the corresponding actions (plug or repair all tubes exceeding the repair limit and all tubes containing through-wall cracks) required l
by Table 4.4-2.
For Unit 1, tubes with defects below F* fall under the alternate tube plugging criteria and do not have to be plugged.
4.4.5.5 Reports Within 15 days following the coCpletion of each inservice inspection a.
s of steam generator tubes, the number of tubes repaired in each steam l
generator shall be reported to the Commission in a Special Report pursuant to Specification 6.9.2; b.
The complete results of the steam generator tube inservice inspection shall be submitted to the Commission in a Special Report pursuant to Specification 6.9.2 within 12 months following the completion of the-inspection.
This Special Report shall include:
1 1)
Number and extent of tubes inspected, CATAWBA - UNITS 1 & 2 3/4-4-16 Amendment-No. 84 (Unit 1)
Amendment No. 78 (Unit 2)
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I SURVEILLANCE REQUIREMENTS (Continued) 2)
Location and percent of wall-thickness penetration for each
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indication of an imperfection, and j
3)
Identification of tubes repaired.
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For Unit 2, results of steam generatoo tube inspections, which fall j
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into Category C-3, shall be reported in a Special Report to the Commission pursuant to Specification 6.9.2 within 30 days pd prior 1
to resumption of plant operation.
This report shall provise a description of investigations conducted to determine cause of the j
tube degradation and corrective measures taken to. prevent recurrence, d.
For Unit 1, the results of inspections for all tubes for which the alternate tube plugging criteria has been applied shall be reported to the Nuclear Regulatory Commission in accordance with 10 CFR 50.4, 4
prior to restart of the unit following the inspection.
This report r
shall include:
1)
Identification of applicable tubes, and 2) location and size of the degradation.
i i
1 4
CATAWBA - UNITS 1 & 2 3/4 4-16a Amendment No. 34 (Unit 1)
Amendment No. 78 (Unit 2)
REACTOR COOLANT SYSTEM BASES STEAM GENERATORS (Continued) generator tubes is based on a modification of Regulatory Guide 1.83, Revision 1.
Inservice inspection of steam generator tubing is essential in order to main-tain surveillance of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to design, manu-facturing errors, or inservice conditions that lead to corrosion.
Inservice inspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken.
The B&W process (or method equivalent) to the inspection method described in Topical Report BAW-2045(P)-A will be used.
Inservice inspection of steam gen-erator sleeves is also required to ensure RCS integrity.
Because the sleeves introduce changes in the wall thickness and diameter, they reduce the sensitivity of eddy current testing, therefore, special inspection methods must be used.
A method is described in Topical Report BAW-204S(P)-A with supporting validation data that demonstrates the inspectability of the sleeve and underlying tube.
As required by NRC for licensees authorized to use this repair process, Catawba com-mits to validate the adequacy of any system that is used for periodic inservice inspections of the sleeves, and will evaluate and, as deemed appropriate by Duke Power Company, implement testing methoos as better methods are developed and a
validated for commercial use.
The plant is expected to be operated in a manner such that the secondary coolant will be maintained within those chemistry limits found to result in negligible corrosion of the steam generator tubes.
If the secondary coolant I
chemistry is not maintained within these limits, localized corrosion may likely L
result in stress corrosion cracking.
The extent of cracking during plant opera-tion woula be limited by the limitation of steam generator tube leakage between the Reactor Coolant System and the Secondary Coolant System (reactor-to-secondary leakage = 500 gallons per day per steam generator).
Cracks having a reactor-to-secondary leakage less than this limit during operation will have an adequate margin of safety to withstand the loads imposed during normal operation and by postulated accidents. Operating plants have demonstrated that reactor-to-secondary leakage of 500 gallons per day per steam generator can readily be detected by radiation monitors of steam generator blowdown.
Leakage in excess of this limit will require plant shutdown and an unscheduled inspection, during which the leaking tubes will be located and repaired, j
Wastage-type defects are unlikely with proper chemistry treatment of the secondary coolant.
However, even if a defect should develop in service, it will be found auring scheduled inservice steam generator tube examinations.
Repair will be required for all tubes with imperfections exceeding the repair limit of 40% of the tube nominal well thickness.
For Unit 1, defective tubes which fall under the alternate tube plugging criteria do not have to be repaired. Defec-tive 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, allowing the tube to remain in service.
Steam generator tube inspections of operating plants have demonstrated the capability to reliably detect wastage type degradation that has penetrated 20% of the original tube wall thickness.
CATAWBA - UNITS 1 & 2 8 3/4 4-3 AmendmentNo.gUnit1 Amendment No.
Unit 2
REACTOR COOLANT SYSTEM BASES
,S_ TEAM GENERATORS (Continued)
Whenever the results of any steam generator tubing inservice inspection f all into Category C-3, these results will be reported to the Commission pur-suant to Specification 6.9.2 prior to resumption of plant operation.
Such cases will be considered by the Commission on a case-by-case basis and may result in a requirement for analysis, laboratory examinations, tests, additional eddy-current inspection, and revision of the Technical Specifications, if necessary.
If a tube is sleeved due to degradation in the F* distance, then any defects in the tube below the sleeve will remain in service without repair.
3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.6.1 LEAKAGE DETECTION SYSTEMS The Leakage Detection Systems required by this specification are provided to monitor and detect leakage from the reactor coolant pressure boundary.
These Detection Systems are consistent with the recommendations of Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection Systems,"
May 1973.
l CATAWBA - UNITS 1 & 2 B 3/4 4-3a Amendment No. 84 (Unit 1)
Amendment No.78 (Unit 2)
REACTOR COOLANT SYSTEM BASES 3/4.4.6.2 OPERATIONAL LEAKAGE l
PRESSURE BOUNDARY LEAKAGE of any magnitude is unacceptable since it may be indicative of an impending gross failure of the pressure boundary.
Therefore, the presence of any PRESSURE BOUNDARY LEAKAGE requires the unit to be promptly placed in COLD SHUTDOWN.
Industry experience has shown that while a limited amount of leakage is expected from the Reactor Coolant System, the unidentified portion of this leakage can be reduced to a threshold value of less than 1 gpm.
This thres-hold value is sufficiently low to ensure early detection of additional leakage.
The total steam generator tube leakage limit of 1 gpm for all steam generators not isolated from the Reactor Coolant System ensures that the dosage contribution from the tube leakage will be limited to a small fraction of 10 CFR Part 100 dose guideline values in the event of either a steam generator tube rupture or steam line break.
The 1 gpm limit is consistent with the assumptions used in the analysis of these accidents.
The 500 gpd leakage limit per steam generator ensures that steam generator tube integrity is maintained in the event of a main steam line rupture or under LOCA conditions.
The 10 gpm IDENTIFIED LEAKAGE limitatinn provides allowance for a limited amount of leakage from known sources whose presence will not interfere with the detection of UNIDENTIFIED LEAKAGE by the Leckage Detection Systems.
The CONTROLLED LEAKAGE limitation restricts operation when the total flow l
supplied to the reactor coolant pump seals exceeds 40 gpm with the modulating valve in the supply line fully open at a nominal Reactor Coolant System pres-sure of 2235 psig.
This limitation ensures that in the event of a LOCA, the safety injection flow will not be less than assumed in the safety analyses.
The 1 gpm leakage from any Reactor Coolant System pressure isolation valve is sufficiently low to ensure early detection of possible in-series check valve failure.
It is apparent that when pressure isolation is provided by two in-series check valves and when failure of one valve in the pair can go undetected for a substantial length of time, verification of valve integrity is required.
Since these valves are important in preventing overpressurization and rupture of the ECCS low pressure piping which could result in a LOCA that bypasses containtrent, these valves should be tested periodically to ensure low probability of gross failure.
The Surveillance Requirements for Reactor Coolant System pressure isolation valves provide added assurance of valve integrity thereby reducing the prob-ability of gross valve failure and consequent intersystem LOCA.
Leakage from the pressure isolation valve is IDENTIFIED LEAVAGE and will be considered as a portion of the allowed limit.
CATAWBA - UNIT 3 1 & 2 B 3/4 4-4 Amendment No,.84 Unit 1 Amendment No. 78 Unit 2
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