ML20081F896
| ML20081F896 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 03/17/1995 |
| From: | Bewkow H NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20081F902 | List: |
| References | |
| NUDOCS 9503220320 | |
| Download: ML20081F896 (10) | |
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UNITED STATES j
,j NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 205664 001
+ * * *,o DUKE POWER COMPANY NORTH CAROLINA ELECTRIC MEMBERSHIP CORPORATION SALUDA RIVER ELECTRIC COOPERATIVE. INC.
DOCKET NO. 50-413
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CATAWBA NUCLEAR STATION. UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 130 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 November 29, 1994, as supplemented by letters dated January 12, January 27, February 6, February 16, and March 13, 1995, complies with the standards and t
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 I; B.
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 i
and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations set forth irt 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.
9503220320 950317 PDR ADOCK 05000413 P
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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-35 is hereby amended to read as follows:
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 130, 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 amendment is effective as of its date of issuance and shall be implemented within 30 days from the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION l
H rbert N. Berkow, Director Project Directorate II-3 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation Attachment.
Technical Specification d
Changes Date of Issuance: March 17, 1995 I
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y-4 UNITED STATES
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NUCLEAR REGULATORY COMMISSION e
WASHINGTON, D.C. 20656 4 001 g
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DUKE POWER COMPANY l
NORTH CAROLINA MUNICIPAL POWER AGENCY NO. 1 PIEDMONT MUNICIPAL POWER AGENCY I
DOCKET NO. 50-414 1
CATAWBA NUCLEAR STATION. UNIT 2
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AMENDMENT TO FACILITY OPERATING LICENSE i
Amendment No. 124 License No. NPF-52 a'
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. NPF-52 filed.
by the Duke Power Company, acting for itself, North Carolink Municipal Power Agency No. I and Piedmont Municipal Power Agency
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(licensees), dated November 29, 1994, as supplemented by letters dated January 12, January 27, February 6, February 16. and March 13, 1995, 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 Commission; i
1 C.
There is reasonable assurance (1) that the activities authorized 1
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-l 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; j
and E.
The issuance of this amendment is in accordance with 10 CFR Part I
51 of the Commission's regulations and all applicable requirements 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 attachment to this license amendment, and Paragraph 2.C.(2)
NPF-52 is hereby amended to read,of Facility Operating License No.
as follows:
Technical Specifications i
The Technical Specifications contained in Appendix A, as revised through Amendment No. 124, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are-4 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 and shall l
be implemented within 30 days-from the date of issuance.
I FOR THE NUCLEAR REGULATORY COMMISSION Herbert N. Berkow, Director Project Directorate 11-3 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
I Technical Specification
}
Changes Date of Issuance:
March 17, 1995 a
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ATTACHMENT TO LICENSE AMENDMENT N0.130 FACILITY OPERATING LICENSE NO. NPF-35 DOCKET NO. 50-413 SIE TO LICENSE AMENDMENT NO. 124-FACILITY OPERATING LICENSE NO. NPF-52
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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 Paaes Insert Paaes i
VII VII 3/4 4-13 3/4 4-13 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 4-16b 3/4 A4-27 1
3/4 B4-27 3/4 4-27 B 3/4 4-5 B 3/4 4-5
- reformatted, no changes 0,
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LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE0VIREMENTS SECTION pag i
3/4.4.2 SAFETY VALVES Shutdown................................................. 3/4 4-7 0perating................................................ 3/4 4-8 3/4.4.3 PRESSURIZER..............................................
3/4 4-9 3/4.4.4 RELIEF VALVES............................................ 3/4 4-10 3/4.4.5 STEAM GENERATORS.........................................
3/4 4-12 TABLE 4.4-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING INSERVICE INSPECTION.............................. 3/4 4-17 TABLE 4.4-2 STEAM GENERATOR TUBE INSPECTION.....................
3/4 4-18 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems................................ 3/4 4-19 Operati on al Leakage...................................... 3/4 4-20 TA"LE 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES.... 3/4 4-22 3/4.4.7 CHEMISTRY................................................ 3/4 4-24 TABLE 3.4-1 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS............. 3/4 4-25 TABLE 4.4-3 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS SURVEILLANCE REQUIREMENTS............................................. 3/4 4-26 3/4.4.8 SPECIFIC ACTIVITY....................................... 3/4 4-27 l
FIGURE 3.4-1 DOSE EQUIVALENT I-131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY > 1 pCi/ gram DOSE EQUIVALENT I-131.................................... 3/4 4-29 TABLE 4.4-4 REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS P R0G RAM.................................................. 3 / 4 4-3 0 3/4.4.9 PRESSURE / TEMPERATURE LIMITS Reactor Coolant System................................... 3/4 4-32 FIGURE 3.4-2 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS APPLICABLE UP TO 16 EFPY................................. 3/4 4-33 FIGURE 3.4-3 REACTOR COOLANT SYSTEM C00LDOWN LIMITATIONS -
APPLICABLE UP TO 16 EFPY................................. 3/4 4-34 TABLE 4.4-5 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM WITH0RAWAL SCHEDULE...................................... 3/4 4-35 Pressurizer.............................................. 3/4 4-36 Overpressure Protection Systems.......................... 3/4 4-37 3/4.4.10 STRUCTURAL INTEGRITY................................ 3/4 4-39 3/4.4.11 REACTOR COOLANT SYSTEM VENTS........................ 3/4 4-40 i
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CATAWBA - UNITS 1 & 2 VII Amendment No. 130 (Unit 1)
Amendment No. 124 (Unit 2)
REACTOR COOLANT SYSTEM 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 permit 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.
j 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:
l 1)
The tubes selected for these samples. include the tubes from m
l 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.
e.
For Unit 1, implementation of the interim steam generator tube / tube support plate elevation plugging limit for Cycle 9 requires a 100%
l bobbin probe inspection for all hot leg tube support plate intersections and all cold leg intersections down to the lowest cold leg tube support plate with outer diameter stress corrosion cracking (00 SCC) indications.
The determination of tube support plate intersections having OD SCC indications shall be based on the performance of at least 20 percent random sampling of tubes inspected over their full length. An inspection using the rotat'ng pancake coil (RPC) probe is required in order to show operabi';ty of tubes with flaw like bobbin coil signal amplitudes greater tne 1.0 volt but less than 2.7 volts. The RPC results are to be evaluated to l
establish that the principal indications can be characterized as 00 SCC.
The results of each rample inspection shall be classified into one of the following three categories:
Cateaory Inspection Resulti C-1 Less than 5% of the total tubes inspected are degraded tubes and none of the inspected tubes are defective.
CATAWBA - UNITS 1 & 2 3/4 4-13 Amendment No. 230 (Unit 1)
Amendment No. 1M (Unit 2)
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-i bend to the top support of the cold leg; I
For Unit 1, for a tube in which the. tube support plate elevation
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interim plugging (IPC) limit has been applied, the inspection i
will include all the hot leg intersections and all cold leg 1
intersections down to and including, at least, the level of the last crack indication for which the interim plugging criteria limit is to be applied.
9)
Preservice Insoection means an inspection of the full length of i
each tube in each steam generator performed by addy current J
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 Exoansion 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.
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- 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 l
expansion transition or the top of the tubesheet if the bottom i
of the roll expansion is above the top of the tubesheet.
l Included in this distance is a safety factor of 3 plus a 0.5 inch eddy current vertical measurement uncertainty.
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- 12) Alternate tube cluaaina 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.
CATAWBA - UNITS 1 & 2 3/4 4-15a Amendment No.130 (Unit 1)
Amendment No.124 (Unit 2)
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I REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)
- 13) The Tube-Sunnort Plate Interim Pluaaina Criteria Limit is used I
for disposition of.a steam generator tube for 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 criteria limit, the tube's disposition for continued service will be based upon standard bobbin probe signal ampli-tude' of flaw like indications. The plant specific guidelines i
used for all inspections shall be consistent with the eddy l
current guidelines in Appendix A of WCAP-13854 as appropriate to accomodate the additional information needed to evaluate tube support plate signals with respect to the voltage parameters as specified in Specification 4.4.5.2.
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 i
less of the' depth of tube wall penetration, if, as a result, '?
the projected end of cycle distribution of crack indication,s is verified to result in total primary to secondary leakage '
less than 17.5 gpm (includes operational and accident leak 3 age).
2.
A tube can remain in service with a bobbin coil signal amplitude greater than 1.0 volt but less than 2.7 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 2.7 volts will be plugged or repaired.
1 e.
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1 CATAWBA - UNITS 1 & 2 3/4 4-16 Amendment No. 130 (Unit 1)
Amendment No. 124 (Unit 2)
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REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 4.
If as a result of leakage due to a mechanism other than 00
. SCC at the tube support plate intersection, or some other cause, an unscheduled mid-cycle inspection is performed, the following repair criteria apply instead of 4.4.5.4.a.13.2.
If bobbin voltage is within expected limits the indication-can remain in service. The expected repair limits are
, determined from the following equation:
t gy _y y,y g
V<
1+ (.2) (
U) where.
measured voltage V
V, =
voltage at beginning of cycle (BOC) time period of operation to unscheduled outage At cycle length (full operating cycle length where-CL operating cycle is the time between two scheduled steaw generator inspections) 4.5 volts for 3/4-inch tubes I
V,t Certain tubes as identified in WCAP-13494,REV.1, 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.
.5 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; i
CATS BA - UNITS 1 & z 3/4 4-16a Amendment No. 130 (Unit 1)
Amendment No. 124 (Unit 2) p.--
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REACTOR COOLANT SYSTEM SURVEILLANCE REOUIREMENTS (Continued) 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
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inspection. This Special Report shall include:
1)
Number and extent of tubes inspected, 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 I
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 j
tube degradation and corrective measures taken to prevent recurrence.
j 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, 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 implementation of the voltage-based repair criteria to tube support. plate intersections, notify the NRC staff prior to returning the steam generators to service should any of the following conditions arise:
1.
If the estimated leakage based on the actual measured end-of-cycle voltage distribution would have exceeded the leak limit (for the postulated main steam line break utilizing licensing basist tsrumptions) during the previous operating cycle.
2.
If circumferential crack like indications are detected at the tube support plate intersections.
3.
If the indications are identified that extend beyond the confines of the tube support plate.
4.
If the calculated conditional burst probability exceeds 1 X 10'2,
notify the NRC and provide an assessment of the safety.
significance of the occurrence.
CATAWBA - UNITS 1 & 2 3/4 4-16b Amendment No. 130 (Unit 1)
Amendment No.124 (Unit 2) l
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REACTOR COOLANT SYSTEM 3/4.4.8 SPECIFIC ACTIVITY
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LIMITING CONDITION FOR OPERATION 3.4.8 The specific activity of the reactor coolant shall be limited to:
a.
Less than or equal to 1 microcurie per gram DOSE EQUIVALENT I-131, and b.
Less than or equal to 100/E microcuries per gram of gross specific activity.
APPLICABILITY: MODES 1, 2, 3, 4, and 5.
l ACTION:
MODES 1, 2 and 3*:
a.
With the specific activity of the reactor coolant greater than 1 microcurie per gram DOSE EQUIVALENT I-131 for more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> during one continuous time interval or exceeding the limit line shown on Figure 3.4-1, be in at least HOT STANDBY with T,, less than 500*F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; b.
With the gross specific activity of the reactor coolant greater than 100/E microcuries per gram of gross radioactivity, be in at least HOT STANDBY with T,,less than 500*F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; and c.
The provisions of Specification 3.0.4 are not applicable.
l
- With T,, greater than or equal to 500*F.
CATAWBA - UNITS 1 & 2 3/4 4-27 Amendment No. 130 (Unit 1) l Amendment No. 124 (Unit 2)
l REACTOR' COOLANT-SYSTEM BASES 3/4.4.7 CHEMISTRY The limitations on Reactor Coolant System chemistry, ensure that corrosion of the Reactor Coolant System is 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 corrosion protection to ensure the structural integrity of the Reactor Coolant System over the life of the plant. The associated effects of exceeding the oxygen, chloride, and fluoride limits are time and temperature dependent.
Corrosion studies show that operation may be continued with contaminant concentration levels in excess of 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 corrective actions to restore the contaminant concentrations to within the Steady-State Limits.
The Surveillance Requirements provide adequate assurance that concentrations in excess of the limits will be detected in sufficient time to take corrective action.
3/4.4.8 SPECIFIC ACTIVITY The limitations on the specific activity of the reactor coolant ensure that the resulting 2-hour doses at the SITE BOUNDARY will not exceed an appropriately small fraction of Part 100 dose guideline values following a steam generator tube rupture accident in conjunction with an assumed steady-state primary-to-secondary steam generator leakage rate of 0.4 gpm. The values for the limits on specific activity represent limits based upon a parametric evaluation by the NRC of typical site locations. These values are conservative in that specific site parameters of the Catawba site, such as SITE BOUNDARY location and meteorological conditions, were not considered in this evaluation.
The ACTION statement permitting POWER OPERATION to continue for limitcd time priods with the reactor coolant's specific activity greater than i
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.
I CATAWBA - UNITS 1 & 2 B 3/4 4-5 Amendment No. 130 (Unit 1)
Amendment No. 124 (Unit 2) j