ML20155G363
| ML20155G363 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 06/06/1988 |
| From: | Matthews D Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20155G367 | List: |
| References | |
| NUDOCS 8806170268 | |
| Download: ML20155G363 (14) | |
Text
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%,h UNITED STATES O
NUCLEAR REGULATORY COMMISSION j
j WASHINGTON, D, C. 20555 e
DUKE POWER COMPANY NORTH CAROLINA ELECTRIC MEftBERSHIP CORPORATION SALUDA RIVER ELECTRIC COOPERATIVE, INC.
DOCKET N0. 50-413 CATAWBA NUCLEAR STATION, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 47 License No. NPF-35 1.
The Nuclear Regulatory Comission (the Comission) has 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.,
(licensees dated October 8, 1987, as supplemented December 3, 1987, 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; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comissien; 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 Comission's regulations and all applicable requirenents have been satisfied.
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. 2.
Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachments to this license amerdment, and Paragraph 2.C.(2) of Facility Op3 rating License No. fiPF-35 is hereby amended to read as follows:
j (2) Technical Specifications l
l The Technical Specifications contained in Appendix A, as revised through Amendment No. 47 are hereby incorporated into the license.
l The licensee 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 COMillSSI0tl l
Original signed by:
Avid B. liatthews, Director Froject Directorate 11-3 Division of Reactor Projects-I/II
Attachment:
Technical Specification Changes i
Date of Issuance: June 6, 1988 1
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l t
0FFICIAL RECORD COPY
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%,.....l DUKE POWER COMPANY NORTH CAROLINA MUNICIPAL POWER AGENCY NO. 1 PIEDMONT MUNICIPAL POWER AGENCY DOCKET l10. 50-414 CATAWBA NUCLEAR STATION, UNIT 2 AMEllDHENT TO FACILITY OPERATING LICENSE Amendment No.
40 License No. NPF-52 1.
The !!uclear Regulatory Comission (the Comission) has found that:
A.
The application for auendment 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 Municipal Power Agency No. 1 and Piedmont Municipal Power Agency, (licensees) dated October 8, 1987, as supplemented December 3, 1987, 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 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 i
conducted in compliance with the Comission's regulations set fortn 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 Comission's regulations and all applicable requirements have been satisfied.
a-m
.,, 2.
Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachments to this-license amendment, and Paragraph 2.C.(2) of Facility Operating License No. NPF-52 is hereby amended'to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.
40, are hereby incorporated into the license.
The licensee 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 Original signed by:
David B. Matthews, Director Project Directorate 11-3 Division of Reactor Projects-I/II
Attachment:
Technical Specification Changes Date of Issuance:
June 6, 1988 0FFICIAL RECORD COPY LA:PJQI-3 PM:PDII-3 I
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ATTACHMENT TO LICENSE ANENDMENT N0. 47
.]
FACILITY OPERATING LICENSE NO. NPF-35 DOCKET NO. 50-413 j
AND TO LICENSE AMENDMENT NO. 40 FACILITYOPERATINGLICENSENO.NPF-g i
DOCKET NO. 50-414 i
Replace the following pages of the Appendix "A" Technical Specifications with l
the enclosed pages.
The revised pages are identified by Amendment number and contain vertical lines indicating the areas of change. The corresponding over-leaf pages are also provided to maintain document completeness.
l Amended Page Overleaf Page 3/4 4-13 3/4 4-14 3/4 4-15 3/4 4-16 3/4 4-16a(new page) 3/4 4-18 3/4 4-17 8 3/4 4-3 8 3/4 4-4 l
i l
l l
l L
REACTOR COOLANT SYS*:i.M SURVEILLANCE 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) shall be performed on each selected tube.
If any selected tube does not permic the passage of the eddy current probe 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 plugging criteria has been previously applied shall be inspected in the tubesheet region.
d.
The tubes selected as the second and third samples (if required by Table 4.4-2) during each inservice inspection may be subjected to a 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 previously found.
The results of each sample inspection shall be classified into one of the following three categories:
i 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 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.
i Note:
In all inspections, previously degraded tubes must exhibit significant (greater than 10%) further wall penetrations i
to be included in the above percentage calculations.
i CATAWBA - UNITS 1 & 2 3/4 4-13 Amendment No. 47 (Unit 1)
Amendment No. 40 ; Unit 2)
REACTOR COOLANT 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 initial 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 inspections, not including the preservice inspection, result in all inspection results falling into the C-1 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 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 intervais fall in Category C-3, the inspection frequency shall be increased to at least once per 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 steam generator in accordance with the first sample inspection specified in Table 4.4-2 durina 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 exiess 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.
CATAWBA - UNITS 1 & 2 3/4 4-14
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 from that required by fabrication drawings or specifications.
Eddy-current testing indications below 20% of the nominal tube wall thickness, if detectable, may be consicered as imperfections; j
l 2)
Degradation means a service-induced cracking, wastage, wear or general corrosion occurring on either inside or outside of a tube; 3)
Degraded Tube means a tube containing imperfections greater than or equaf to 20% of the nominal wall thickness caused by degradation; 4)
% Degradation means the percentage of the tube wall thickness affected or removed by degradation; 5)
Defect means an imperfection of such severity that it exceeds the plugging limit.
A tube containing a defect is defective; 6)
Plugging Limit means the imperfection depth at or beyond which the tube shall be removed from service and is equal to 40% of the nominal tube wall thickness.
For Unit 1, this definition does not apply to the region of the tube subject to the alter-nate tube plugging criteria.
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) b be Inspection means an inspection of the steam generator tube from the point of entry (hot leg side) completely around the U-bend *.o the top support of the cold leg; l
CATAWBA - UNITS 1 & 2 3/4 4-15 Amendment No. 47 (Unit 1)
Amendment No. 40 (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 subsequent inservice inspections.
- 10) Tube Roll Expansion is that porti n of a tube which has been increased in diameter by a rolling process such that no crevice i
exists between the outside diameter of the tube and the tubesheet.
11)
F* Distance is the minimum length of the roll expanded portion j
of the tube which cannot contain any defects in order to ensure i
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 it, a safety factor of 3 plus a 0.5 inch eddy current vertical measurement uncertainty.
j
- 12) Alternate tube plugging criteria does not require the tube to be removed from service or repaired when the tube degradation exceeds the plugging 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 j
aegradation (i.e., indications of cracking) in the F*
i
- distance, b.
The steam generator shall be determined OPERABLE after completing the corresponding actions (plug or repair all tubes exceeding the plugging limit and all tubes concaining through-wall cracks) required by Table 4.4-2.
For Unit 1, tubes with defects below F* fall under the alternate tube plugging criteria r.nd do not have to be plugged.
4.4.5.5 Reports Within 15 days following the completion of each inservice inspection a.
of steam generator tubes, the number of tubes plugged in each steam generator shall be reported to the Commission in a 3pecial 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, CATAWBA - UNITS 1 & 2 3/4 4-16 Amendment No. 47 (Unit 1)
Amendment No. 40 (Unit 2)
(
SURVEILLANCEREQUIREMENTS(Continued}
2)
Location and percent of I'-t ickness penetration for e6ch indication of an imperfect '
nd 3)
Identification of tubes p
.d.
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 to resumption of plant operation.
This report shall provide a description of investigations conducted to determine cause of the tute degradation and corrective measures taken to prevent recurrence.
d.
For Unit 1, the results of inspections for all tuoes for which the alternate tube plugging criteria has been apolied 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.
CATAWBA - UNITS 1 & 2 3/4 4-16a Amendment No. 47 (Unit 1)
I Amendment Nc. 40 (Unit 2)
Table ?.4-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE c3
?$
INSPECTED DURIMG INSERVICE INSPECTION 25 2-a EE Preservice Inspection 4o Yes Z
No. of Steam Generators per Unit Four Four
'a g
First Inservice Inspection All Two Second & Subsequent Inservice Inspections Onel One2 TABLE NOTATIONS 1.
The inservice inspection may be limited to one steam generator on a rotating senedule encompassing 3 N % of the tubes (where N is the number of steam generators in the plant) if the results of the first or previous inspections u,);
indicate that all steam generators are performing in a like manner.
Note that un;'er some circumstances, the operating conditions in one or more steam 4,
4 generators may be found to be more severe than those in other steam generators.
Under such circumstances the sample sequence shall be modified to inspect the
'd most severe conditions.
2.
Each of the other two steam generators not inspected during the first inservice inspections shall be inspected during the second and third inspections. The fourth and subsequent inspecticas shall follow the instructions described in 1 above.
l
Table 4.4-2 STEAM GENERXTOR IU8E INSPECTION O
IST SAMPLE INSPECTION 2ND SAMPLE.NSPECTION 3dD SAMPLE INSPECTION
>y Sample Size Result Action Required Result Action Required Result Action Required 4 minimum of S Tubes per S.G.
C-1 None N.A.
N.A.
N.A.
N.A.
I g
C-2 Plug defective tubes
- and inspect addi-C-1 None M.A.
N.A.
tional 25 tubes in this S.G.
g C-2 Plug defective tubes
- C-1 Mone O) and inspect additional g
45 tubes in this S. G.
C-2 Plug defective tubes' O.
Perform action for y
C-3 C-3 result of first sample Perform action for C-3 C-3 result of first N.A.
N.A.
sample C-3 inspect all tubes in this S G.,
plug All other S.G.s are defective tubes
- and inpsect 25 tubes C-1 None M.A.
N.A.
2 in each other S. G.
Some S.G.s C-2 but no additional S.G.
Perform action for C-2 N.A.
N.A.
are C-3 result of second sasq.le Notification to NitC pursuant to 550.72 (b)(2) of 10 CFR Part 50 Additional S.G. is Inspect all tubes in each C-3 S.G. and plug defective
(
O tub?s.* Notification to N.A.
N.A.
NRC pursuant to 550.72 y
3 (b)(2) of 10 CFR Part 50 Ig S=3N Where N is the number of steam generators in the unit, and n is the number of steam generators inspected during an inspection g
n
- For Unit 1, defective tubes which fall under the alternate plugging criteria do not have to be plugged.
n.
33 cLCL 99 1
tb tb sc ct n ZZ
33 ta m I
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REACTOR COOLANT SYSTEM BASES STEAM J NERATORS (Continued) genera ar 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-j 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 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 tne steam generator tubes.
If the secondary coolant chemistry is not maintained within these limits, localized corrosion may likely result in stress corrosion cracking.
The extent of cracking during plant opera-tion would 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 gallor 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 plugged.
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 during scheduled inservice steam generator tube examinations.
Plugging will be required for all tubes with imperfections exceeding the plugging limit of 40% of the tube nominal wall thickness.
For Unit 1, defective tubes which fall under tne alternate tube plugging criteria do not have to be plugged.
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.
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 resumption of plant operation.
Such j
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.
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.
CATAWBA - UNITS 1 & 2 8 3/4 4-3 Amendment No. 47 @ nit 1)
Amendment No. 40 (Unit 2)
REACTOR COOLANT SYSTEM BASES These Detection Systems are consistent with the recommendations of Regulatory Guide 1.45, "Reactor Coolant Pressure Boundary Leakage Detection Systems,"
May 1973.
3/4.4.6.2 OPERATIONAL LEAKAGE PRESSURE B0UNDARY 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 8OUNDARY LEAKAGE requires the unit to be promptly placed in COLO SHUTOOWN.
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 srall 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 limitation provides allowance for a limited amount of leakage from known sources whose presence will not interfere with the detection of UNIDENTIFIED LEAKAGE by the Leakage Detection Systems.
1 The CONTROLLED LEAKAGE limitation restricts operation when the total flow 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 containment, 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 LEAKAGE and will be considered as a portion of the allowed limit.
CATAWBA - UNITS 1 & 2 B 3/4 4-4
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