ML20118B700
| ML20118B700 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 09/25/1992 |
| From: | Matthews D Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20118B701 | List: |
| References | |
| NUDOCS 9210070007 | |
| Download: ML20118B700 (19) | |
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UNITED STATES 3
D (.ni NUCLEAR REGULATORY COMMISSION
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DUKE POWER COMPANY NORTH CAROLINA ELECTRIC MEMBERSHIP LQRPORATION SALUDA RIVLR ELECTRIC COOPERATIVE. INC, DOCKET NO. 50-413 CATAWBA NVCLEAR STATION. UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.102 License No. NPF-35 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment to the Catawba Nuclear Station, Unit I (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. (licensees) dated August 24, 1992, as supplemented September 2, 4, 17, and 23, 1992, ccmplies 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, acd (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 issuar:e of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
9210070007 920923 PDR ADOCK 05000413 P
. 2.
Accordingly, the license is hereby amended by page changes to the Technical-Specifications as indicated in the attachment to-this license 4,2cadment, and Paragraph 2.C.(2) of f acility 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.102, and the Environmental Protection Plan contained ir, Appendix 8, 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 4 j/ L d t M A ? N
- f*;sDavidB.Matthews, Director Project Directorate 11-3 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation Attachmant
Technical Specification Changes Date of Issuance:
September 25, 1992 l,.
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UNITED STATES 5
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DUKE POWER COMPANY NORTH CAROLINA MUNICIPAL POWER AGENCY N0. 1 PIEDMONT MUNICIPAL POWER AGENCY DOCKET NO. 50-414 CATAWBA NUCLEAR STATION, UNIT 2 AMEN 0 MENT TO FACIllTY OPERATING LICENSE Amendment No. 96 License No. NPF-52 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment to the Catawba Nuclear Station, Unit 2 (the facility) Facility Operating License No. NPT-52 filed by the Duke Power Company, acting for itself, North Carolina Municipal Power Agency No.1 and Piedmont Municipal Power Agency (licensees) dated August 24, IL2, as supplemented September 2, 4, 17, and 23, 1992, complies 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;
8.
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.
Accordingly, the license is hereby amended by page 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-52 is-hereby amended to read as follows:
Technical Specifications
'The Technical Specifications contained in Appendix A, as revised through Amendment No.
96, 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 Technical Specifications and the Environmental Protection Plan.
3.
This license amenament is effective as of its date of issuance.
FOR THE NUCLEAR REGULATvRY COMMISSION fcr( Wl $
- avid B.-Matthews, Director Project Directorate 11-3 Division of. Reactor Projects - 1/II Office of Nuclear Reactor Regulation
Attachment:
Technical Specification Changes Data of Issuance: September 25, 1992 1
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ATTACHMENT TO LICENSE AMUl0 MENT NO.102 FACILITY OPERATING LICENSE N0, NPF-35 DOCKET NO 50-413 AND TO LICENSE AMEN 0 MENT NO. 9f>_
FACILITY OPERATING LICENSE NO, NPF-52 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 Panes Insert Paagi 3/4 4-13 3/4 4-13 3/4 4-13a 3/4 4-14 3/4 4-14 3/4 4-15 3/4 4-15 3/4 4-15a 3/4 4-15a 3/4 4-16 3/4 4-16 3/4 4-16a 3/4.4-16a 3/4 4-16b 3/4 4-20 3/4 4-20 B 3/4 4-3 8 3/4 4-3 8 3/4 4-3a B 3/4 4-3a B 3/4 4-4 8 3/4 4-4 8 3/4 4-5 8 3/4'4-5
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REACTOR C00LAN' SYSTEM SURVFILLANCE REQUIREMENTS (Continued) 1)
All nonplugged tubes that previously had detectable wall penetrations (greater than 20%),
2)
Tubes in those areas where experience has indicated potential problems, and 3)
A tube inspection (pursuant to Specification 4.4.5.4a.8) shai',
be performed on each selected tube.
If any selected tube does not permit the passage of the eddy current prete for a tube inspection, this shall be recorded and an adjacent tube shall be selected and subjected to a tube inspection, c.
for Unit 1, In addition to the.3% sample, all tubes for which the alternate pluggic, criteria has been previously applied shall be inspected in the tow sheet region.
d.
The tubes selected as the second and third samples (if required L,y Table 4.4-2) during each inservice inspection may be subjected to i partial tube inspection provided:
1)
The tubes selected for these samples include the tubes from those areas of the tube sheet array where tubes with imperfections were previously found, and 2)
The inspections include those portions of the tubes where imperfections were previour,1y found, e.
For Unit 1, implementation of the interim steam generator tube / tube support plate elevation plugging limit requires a 100% bobbin probe inspection for all hot leg tube support plate intersections and all cold leg intersectiori down to the lowest cold leg tube support plate with outer diameter : tress corrosion cracking (00 SCC) indications.
An int.pection using the rotating panq ke coil (RPC) probe is required in order to show operability of tubes with finw like bobbin coil signal amplitudes greater than 1.0 volt but less than 2.5 volts.
For tubes that will be administratively plugged or repaired, no RPC inspection is required.
The RPC results. ire to be avaluated to I
establish that the principal indications can be characterized as OD i
SCC.
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 inspectt.d are degraded tubes and none of the inspected tubes a.e defective.
CATN#BA - UNITS 1 & 2 3/4 4 13 Amendment No.102 (Unit '.)
Amendment No. 96 (Unit 2)
SURVEILLANCE. REQUIREMENTS (Continued)
I Category Inspectic,n Results C-2 One or more tubes, but not 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 1% of the inspected tubes are defective.
Note:
In all inspections, previously degraded tubes must exhibit significant (greater than 10%) further wall penetrations to be included in the abcce precentage calculations.
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CATAWBA - UNITS 1 & 2 3/4 4-13a Amendment No.102 (Unit 1) l Amendment No. 96 (Unit 2) y-.
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REA' TOR C0'LANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)-
4.4.5.3 Inspection Frequencies - The above required inservice inspections of steam generator tubes shall be performed at the following frequencies:
a.
The first inservice inspection shall be performed after 6 Effective Full Power Months but within 24 calendar months of iritial criticality.
Subsequent inservice inspections shall be performed at intervals of not less than 12 nor more than 24 calendar months after the previous inspection.
If two consecutive instictions, not including the preservice inspection, result in Oi' inspectio results falling into the C-1 category or if two consecutive inspecticns demonstrate that previously observed degradation has not continued and no additional degradation has occurred, the inspection interval may be extended to a maximum of once per 40 months; b.
If the results of the inservice inspection of a steam generator conducted in accordance with Table 4.4-2 at 40-month intervals fall in Category C-3, the inspection frequency shall be increased to at least once par 20 months.
The increase in inspection frequency shall apply until the subsequent inspections satisfy the criteria of Specification 4.4.5.3a.; the interval may then be extended to a maximum of once per 40 months; and c.
Additional, unscheduled inservice inspections shall be performed on each 4taam generator in accordance with the first sample inspection specified in Table 4.4-2 during the shutdown subsequent to any of the following conditions:
1)
Reactor-to-secondary tubes leak (not including leaks originating from tube-to-tube sheet welds) in excess of the limits of Specification 3.4.6.2, or 2)
A seismic occurrence greater than the Operating Basis Earthquake, or 3)
A loss-of-coolant accident requiring actuation of the Engineered Safety Features, or 4)
A main steam line or feedwater line break, d.
For Unit 1, tubes in which the tube support plate elevation IPC plugging limit have been applied shall be inspected during all future refueling outages.
CATAWBA - UNITS 1 & 2 3/4 4 ~l Amendment No.102 (Unit 1)
Amendment No. 96 (Unit 2)
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SURVEILLANCE REQUIREMENTS (Continued) 4.d.5.4 Acceptance Criteria I
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 dr:iwings or specifications.
Eddy-current testing indications below 20% of the nominal tube or sleeve wall thickness, if 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; 3)
Degraded Tube means a tube or sleeve containing imperfections greater than or eq.al to 20% of the nominal tube nr sleeve wall thickness caused by degradation; 4)
% Degradation means the pp.rcentage of the tube or sleeve wall thickness affected or removed by degradation; 5)
Defect means an imperfection of such severity that it exceeds the repair limit.
A tube or sleeve containing a defect is defective; 6)
Repair Limit means the imperfection depth at or beyond which the tube 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 Flugging 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, Rev. I will be used for sleeving.
For Unit 1 also, this definition does not apply for tcbes experiencing outer diameter stress corrosion cracking confirmed by bobbin probe inspection to be within the thickness of the tube support plates.
See 4.4.5.4.a.13 for the plugging limit for use within the thickness of the tube suppnrt plate.
- CATAWBA - UNITS 1 & 2 3/4 4-15 Amendment No.102 (Unit 1)
Amendment No, 96 (Unit 2)
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, REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 7)
Unserviceable describes the condition of a tube if it leaks or 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; For Unit 1, for a tube in which the tube support plate elevation I.
interim plugging (IPC) limit has been applied, the inspection will include all the hot leg intersections and all cold leg intersections down to and including, at least, the level of the last crack indica-tion for which the interim plugging criteria limit is to be applied.
3
.D CATAWBA - UNITS 1 & 2 3/4 4-15a Amendment No.102 (Unit 1)
Amendment No. 96 (Unit 2)
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REACTOR C00LAfd SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 9)
Preservice Inspection means an inspection of the full length of each tube in each steam generator performed by eddy curr9nt techniques prior to service to establish a baseline condition of the tubing.
This insoection shall b9 performed prior to initial POWER OPERATION using the equipment and techniques expected to be used during subsequent 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 ar the top of the tubesheet if the boStom 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 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.
- 13) The Tube Support Plate Interim Plugging Criteria Limit is used for disposition of a steam generator tube f or continued service that is experiencing outer diameter initiated stress corrosion cracking confined within the thickness of the tube support plates.
For application of the tube support plate interim plugging cri-teria limit, the tub. s disposition for continued service will be based upon standaru bobbin probe signal amplitude of flaw like indications.
The plant specific guidelines used for all inspec-tions shall be amended as appropriate to accommodate the addi-tional information needed to evaluate tube support plate signals with respect to the voltage / depth parameters as specified in Specification 4.4.5.2.
Pending incorporation of the voltage verification requirement in ASME standard verifications, an ASME standard calibrated against the laboratory standard will be utilized in Catawba Unit 1 for consistent voltage normalization.
1.
A tube can remain in service if the signal amplitude of a crack indication is less than or equal to 1.0 volts, regard-less of the depth of tube wall penetration, if, as a result, the projected end of cycle distribution of crack indicationc CATAWBA - UNITS 1 & 2 3/4 4-16 Amendment No.102 (Unit 1)
I Amendment No. 96 (Unit 2) l
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REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) is verified to result in total primary to secondary leakage less than 1.0 gpm (includes operational and accident leak-age).
The basis for determining expected leak rates from the projected crack distribution is oravided in SECL-92-282.
2.
A tube can remain in service with a bobbin coil signal amplitude greater than 1.0 volt but less than 2.5 volts provided a rotating pancake coil (RPC) inspection does not detect degradation.
3.
Indications of degradation with a flaw type bobbin coil signal amplitude of equal to or greater than 2.5 volts will be plugged or repaired.
Certain tubes as identified in SECL-92-282, will be excluded from application of the Interim Plugging Criteria Limit as it has been determined that these tubes may collapse or deform following a postulated LOCA + SSE Event.
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 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 a.
Within 15 days following the completion of each inservice inspection of steam generator tubes, the number of tubes repaired in each steam 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)
Number and extent of tubes inspected, i
4 CATAWBA - UNITS 1 & 2
'/4 4-16a Amendment No.102 (Unit 1)
Amendment No, 96 (Unit 2)
REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 2)
Location and percent of wall-thickness penetration for each indication of an imperfection, and 3)
Identification of tubes repaired.
c.
For Unit 2, results of steam generator tube inspections, which fall into Category C-3, shall be reported in a Special Report to the Commission pursuant to Specification 6.9.2 within 30 days and prior j
to resumption of plant operation.
This report shall provide a description of investigations conducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence.
i d.
For Unit 1, the results of inspections fo? 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, prior to restart of the unit following the inspection.
This report shall include:
1)
Identification of applicable tubes, and 2)
Location and size of the degradation.
e.
For Unit 1, the results of inspections performed under 4.4.5.2 for all tubes in which the tube support plate elevations interim plugging limit has been applied shall be reported to the Commission following the inspection and prior to the resumption of plant operatioa.
The report shall include:
1.
Listing of applicable tubes.
2.
Location (applicable intersections per tube) and extent of degradation (voltage).
B CATAWBA - UNITS 1 & 2 3/4 4-16b Amendment No.102 (Unit 1)
Amendment No. 96 (Unit 2) 1
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REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.6.2 Reactor Coolant System leakage shall be limited to:
a.
No PRESSURE BOUNDARY LEAKAGE, L
b.
I gpm UNIDENTIFIED LEAKAGE, c.
0.4 gpm total reactor-to-secondary leakage through all steam generators and 150 gallons per day through any one steam generator, d.
10 gpm DENTIFIED LEAKAGE from the Reactor Coolant System, e.
40 gpm CONTROLLED LEAKAGE at a Reactor Coolant System pressure of 2235 1 20 psig, and f.
1 gpm leakage at a Reactor Coolant System pres 94re of 2235 1 20 psig from any Reactor Coolant Systeti Pressure Isolacion Yalve specified in Table 3.4-1.
APPLICABILITY:
MODES 1, 2, 3, and 4.
ACTION:
a.
With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
b.
With any Reactor Coolant System leakage greater than any one of the above limits, excluding PRESSURE BOUNDARY LEAKAGE and leakage from Reactor Coolant System Pressure Isolation Valves, reduce the leakage rate to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />, c.
With any Reactor Coolant System Pressure Isolation Valve leakage greater than the above limit, isolate the high pressure portion of the affected system from the low pressure portion within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by t
usa of at least two closed manual or deactivated automatic valves, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
CATAWBA - UNITS 1 & 2 3/4 4-20 Amendment No.102 (Unit 1)
Amendment No. 96 (Unit 2)
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I REACTOR COOLANT SYSTEM BASES REllrr '/ALVES (Continued) reactor coolant system pressure except for limited periods where the PORV has been isolated due to excessive seat leakage and except for limited periods where the PORV and/or block valve is closed because of testing and is fully capable of being returned to its normal alignment at any time, provided that this evolution is covered by an approved procedure.
This is a function that reduces challenges to the code safety valves for overpressurization events.
- 5) Manual control of a block valve to isolate a stuck-open PORV.
Testing of the PORVs includes the emergency N supply from the Cold Leg Accumulators.
This test 2
demonstrates that the valves in the supply line optrate satisf actorily and that the nonsafety portion of the instrument air system is not'necessary for proper PORV operation.
3/4.4.5 STEAM GENERATORS 1he Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of ;he Reactor Coolant System will be maintained.
The program for inservice inspectior, of steam 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, Rev. 1, will be used.
Inservice inspection of steam generator 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-2045(P)-A, Rev. I with sunporting validation data that demonstrates the inspectability of the sleeve ar.o underlying tube.
As required by NRC for licensees authorized to use this repair process, Catawba commits 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 Compan;, implement testing methods as better methods are developed and validated for commercial use.
I 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 chemistry is not maintained within these limits, localized corrosion may likely result in stren corrosion cracking.
The extent of cracking during plant opera-tion would be limited by the limitation of sNam generator tube leakage between the Reactor Coolant System and the Secnndary Coolant System (reactor-to-secondary leakage = 150 gallons per day per steam generator).
Cracks having a reactor-to-l secondary leakage less than this limit during operation will have an adequate margin of safety to withstand the loads imposed during normal operati;n and by postulated accidents.
Operating plants have demonstrated that reactor-to-secondary leakage of 150 gallons per day per steam generator can readily be _
l detected.
Leakage in excess of this limit will require plant shutdown and aa j
unscheduled inspection, during which the leaking tubes will be located and repaired.
CATAWBA UNITS 1 & 2 B 3/4 4-3 Amendaent No.102 (Unit 1)
Amendment No. 96 (Unit 2).
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f REACTOR COOLANT SYSTEM BASES STEAM GENERATORS (Continued)
Vastage-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 during scheduled inservice steam generator tuoe examinations.
Repair will be required for all tubes with imperfections exceeding the repair limit of i
40% of the tube nominal wall thickness.
For Unit 1, defective tubes which fall under the alternate tube plugging criteric do not have to be repairad.
Defec-tive steam generator tubes can be repaired by the installation of sleeves which span the ar?a of degradation, and serve as a replacement pressure boundary for the degraded portion of the tube, allowing the tube to remain in servica.
Steam generator tube inspections of operating plants have demonstrated the capability to reliably detect wastage type degradetion that has penetrated 20% of the original tube wall thickness.
Tubes experiencing outer diameter stress corrosion cracking within the thickness of the tube support plates are plugged or repaired by the criterion of 4.4.5.4.a.13.
Whenever the results of any steam generator tubing inservice inspection fall into Category C-3, these results will be reported to the Commission pur-suant to Specification 6.9.2 prior to resamption of plant operation.
Such cases will be considered by the Commission on a case-by-care 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* dirtance, then any defects in the tube below the sleeve will remain in service without repair.
3/4.4.6 REACTOR C00' TNT SYSTEfi LEAKAGE i
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.
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Amendment No.iO2 Unit 1)
Amendment No. 96 ((Unit 2).
CATAWBA - UNITS 1 & 2 8 3/4 4-3a i
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REACTOR COOLANT SYSTEM BASES 3/4.4.6.2 OPERATIONAL LEAKAGE 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 limits of 0.4 gpm for all steam l
generators not isolated from the Reactor Coolant System ensures that the dosage contribution from the tube leaksge 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 0.4 gpm limit is consistent with the assumptions used in the analysis of these accidents.
The 150 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 limitation provides allowance for a limited amount r r seakage from kaown sources whose presence will not interfere with the detaction af UNIDENTIFIED LEAKAGE by the Leakage Detection Systemt.
The CONTROLLED LEAKAGE limitation restricts operation when tbc total flow supplied to the reactor coolant pump seals exceeds 40 gpm with ine modulating valve in the suoply line fully open at a nominal Reactor Coolant System pres-sure of 2235 psig.
This limitation enseres 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 containment, these valves should be tested periodically to ensure low probability of gross failure.
The Surveillance Requirements for Reactor Cnolant System pressure isolation valves provide added assuraace of valve integrity thereby reducing the prob-ability of gross valve failure and consequent-intersystem LOCA.
Leakage from the pressure isolation valve is IDENTIFIED LEAKAGE and will.be considered as a portion of the allowed limit.
CATAWBA - UNITS 1 & E B 3/4 4-4 Amendment No.WL(Unit ;)
Amendment No.96 (Unit 2)
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REACTOR n'0LANT SYSTEM BASES 3/4.4.7 CHEMISTRY The limitations on Reactor Coolant System chemistry, ensure that corrosion of the Reactnr Coolant System it minimized and reduces the potential for Reactor Coolant System leakage or failure due to stress corrosion.
Maintaining the chemistry within the Steady-State Limits provides adequate carrosion protection to ensure the structural integrity of tho Reactor Coolant System over the life of the plant.
The associated effects of exceeding the oxygen, chlorice, and fluoride limits are time and temperature dependent.
Corrosion studies show that operation may be continued with contaminant concentration levels in excess of i
the Steady-State Limits, up to the Transient Limits, for the specified limited time intervals without having a significant effect on the structural integrity of the Reactor Coolant System.
The time interval permitting continued operation within the restrictions of the Transient Limits provides time for taking ccrrec-tive actions to restore the contaminant concentrations to within the Steady-State Limits.
The Surveillance Requirements provide adequate at.curance that concentrations i
iri excess of the limits will be detected in sufficient time to take corrective a c ti ';n.
3/4.4.8 s/ECIFIC ACTIVITY The limitations on the specific activity of the reactor coolant ensure that the resulting 2-houc doses at the SITT BOUNDARY will not exceed an appro-priately small fraction of Part 100 dose guideline values following a steam generator tube ruoture accident in conjunction with an assumed steady-state 4
primary-to-seconoary steam generator leakage rate of 0.4 gpm.
The values for l
the limits on specific activity represent limits based upon a parametric eval-uation by the NRC of typical site locations.
These values are conservative in that specific site parameters of the Catawbs site, such as SITE BOUNDARY location and meteorological :onditions, were not considered in this evaluation.
The ACTION statement permitting POWER OPERATION to continue for limited time periods with the reactor coolant's specific activity greater that.
1.0 microcurie / gram DOSE EQUIVALENT I-131, but within the allowable limit shown on Figure 3.4-1,-accommodates possible iodine spiking phenomenon which may occur following changes in THERMAL POWER.
A CATAWBA - UNITS 1 & 2 B 3/4 4-5 Amendment No.102 (Unit 1)
Amendment No. 96 (Unit 2)
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