ML20080J122
| ML20080J122 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 02/17/1995 |
| From: | Berkow H Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20080J124 | List: |
| References | |
| NUDOCS 9502270112 | |
| Download: ML20080J122 (9) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION f
WASHINGTON, D.C. 20565 0001 E
p DUKE POWER COMPANY NORTH CAROLINA ELECTRIC MEMBERSHIP CORPORATION SALUDA RIVER ELECTRIC COOPERATIVE. INC.
DOCKET NO. 50-413 CATAWBA NUCLEAR STATION. UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 128 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 Membershio Corporation and Saluda River Electric Cooperative, Inc. (licensees), dated January 10, 1994, as supplemented March M and September 15, 1994, and January 5, 1995, 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 forti 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 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 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 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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9502270112 950217 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 Soecifications t
The Technical Specifications contained in Appendix A, 4s revised through Amendment No. 128, and the Environmental Protection Plan contained in Appendix B, both of which are attached tereto, are hereby incorporated into this license. Duke Power Ctmpany shall operate the facility in accordance with the Technicai 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 f
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' He bert N. Berkow, Director Project Directorate II-3 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Technical Specification Changes Date of Issuance:
February 17, 1995 r
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i p atoo y*-
4 UNITED STATES 2
E NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 206eM1001 g.....,o DUKE POWER COMPANY NORTH CAROLINA MUNICIPAL POWER AGENCY NO. 1 PIEDMONT MUNICIPAL POWER AGEN11 DOCKET NO. 50 #11 CATAWBA MUCLEAR STATION. UNIT 2
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NiENDMENT TO FACILITY OPERATING LICENSE Amendment No. 122 License No. NPF-52 1.
The Nuclear Regulatory Commission (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 Municipal Power Agency No.1 and Piedmont Municipal Power Agency (licensees), dated January 10, 1994, as supplemented March 21 and l
September 15, 1994, and January 5, 1995, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules arsd regulations as set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the i
provisions of the Act, and the rules and regulations of the Comission; C.
There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
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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-52.is hereby amended to read as follows:
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 122, 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 H rbert N. Berkow, Director Project Directorate II-3 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Technical Specification Changes Date of Issuance:
February 17, 1995
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ATTACHMENT TO LICENSE AMENDMENT NO.128 FACILITY OPERATING LICENSE NO. NPF-35 DOCKET NO. 50-413 AND TO LICENSE AMENDMENT NO.
122 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 Paaes Insert Paaes 2-B4 2-B4 B 2-6 B 2-6 3/4 2-14 3/4 2-14 8 3/4 2-4 B 3/4 2-4
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UNIT 2 TABLE 2.2.-1 k
REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS N
E FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUE 1.
Manual Reactor Trip N.A.
N.A.
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2.
Power Range, Neutron Flux a.
High Setpoint
$109% of RTP*
s110.9% of RTP*
b.
Low Setpoint
$25% of RTP*
$27.1% of RTP*
3.
Power Range, Neutron Flux, s5% of RTP* with a 56.3% of RTP* with High Positive Rate time constant a time constant 2 2 seconds 2 2 seconds 4.
Intermediate Range, Neutron Flux 525% of RTP*
$31% of RTP*
5 5
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5.
Source Range, Neutron Flux s10 cps
$1.4 x 10 cps E
6.
Overtemperature AT See Note 1 See Note 2 7.
Overpower AT See Note 3 See Note 4 8.
Pressurizer Pressure-Low 21945 psig 21938 psig***
9.
Pressurizer Pressure-High 52385 psig s2399 psig 10.
Pressurizer Water Level-High 592% of instrument span s93.8% of instrument span 11.
Reactor Coolant Flow-Low 291% of loop minimum 289.7% of loop minimum l
measured flow **
measured flow **
g
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- RTP = RATED THERMAL POWER
- Loop minimum measured flow - 96,250 gpm
- Time constants utilized in the lead-lag controller for Pressurizer Pressure-Low are 2 seconds for lead and 1 second for lag. Channel calibration shall ensure that these time constants are adjusted to these values.
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k LIMITING SAFETY SYSTEM SETTINGS 8ASES
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Pressurizer Pressure In each of the pressurizer pressure channels, there are two independent bistables, each with its own trip setting to provide for a High and Low Pres-sure trip thus limiting the pressure range in which reactor operation is per-mitted. The Low setpoint trip protects against low pressure which could lead to DN8 by tripping the reactor in the event of a loss of reactor coolant pressure.
On decreasing power the Low Setpoint trip is automatically blocked by P-7 (a power level of approximately 10% of RATED THERMAL POWER with turbine 9mpulse chamber pressure at approximately 10% of full power equivalent); and on increasing power, is automatically reinstated by P-7.
The High Setpoint trip functions in conjunction with the pressurizer relief and safety valves to protect the Reactor Coolant System against system l
overpressure.
Pressurizer Water Level The Pressurizer High Water Level trip is provided to prevent water relief through the pressurizer safety valves. On decreasing power, the Pressurizer i
High Water Level trip is automatically blocked by P-7 (a level of approxi-mately 10% of RATED THERMAL POWER with a turbine impulse chamber pressure at approximately 10% of full power equivalent); and on increasing power, is i
automatically reinstated by P-7.
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Reactor Coolantflg The Low Reactor Coolant Flow trips provide core protection and prevents DN8 by mitigating the consequences of a loss of flow resulting from the loss of one or more reactor coolant pumps.
On increasing power above P-7 (a power level of approximately 10% of RATED THERMAL POWER or a turbine impulse chamber pressure at approximately 10%
of full power equivalent), an automatic Reactor trip will occur if the flow in more than one loop drops below 91% of nominal full loop flow. Above P-8 (a l
power level of approximately 48% of RATED THERMAL POWER) an automatic Reactor trip will occur if the flow in any single loop drops below 91% of nominal full l
loop flow. Conversely, on decreasing power between P-8 and P-7 an automatic Reactor trip will occur on low reactor coolant flow in more than one loop and below P-7 the trip function is automatically blocked.
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CATAWBA - UNITS 1 & 2 B 2-6 Amendment No.128(Unit 1)
Amendment No.122(Unit 2)
POWER DISTRIBUTION LIMITS 3/4.2.5 DNB PARANETERS
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LIMITING C0lWITION FOR OPERATION 2.
Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of initially being within the region of pro-hibited operation specified on Figure 3.2-1, verify that the combination of THERMAL POWER and Reactor Coolant System total flow rate are restored to within the regions of restricted or i
permissible operation, or reduce THERMAL POWER to less than 5%
of RATED THERMAL POWER within the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
SURVEILLANCE REQUIREMENTS
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4.2.5.1 Each of the parameters of Table 3.2-1 shall be verified to be within their limits at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
4.2.5.2 The Reactor Coolant System total flow rate indicators shall be sub-jected to a CHANhEL CALIBRATION at least once per 18 months. The measurement instrumentation shall be calibrated within 7 days prior to the performance of the flow measurement.
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4.2.5.3 The Reactor Coolant System total flow rate shall be determined by measurement at least once per 18 months.
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i CATAWBA - UNITS 1 & 2 3/4 2-14 Amendment No.128(Unit 1)
Amendment No.122(Unit 2)
l POWER DISTRIBUTION LIMITS
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BASES 3/4.2.5 DNB PARAMETERS (Continued) to maintain a design limit DN8R throughout each analyzed transient. As noted on Figure 3.2-1, Reactor Coolant System flow rate and THERMAL POWER may be
" traded off" against one another (i.e., a low measured Reactor Coolant System flow rate is acceptable if the THERMAL POWER is also low) to ensure that the calculated DNBR will not be below the design DNBR value. The relationship.
defined on Figure 3.2-1 remains valid as_ long as the limits placed on the nuclear enthalpy rise hot channel factor, FaH(X,Y) in Specification 3.2.3 are maintained. The indicated T value and the indicated pressurizer pressure value correspond to analytic W limits of 594.8'F and 2205.3 psig respectively, with allowance for measurement uncertainty. When Reactor Coolant System flow rate is measured, no additional allewances are necessary prior to comparison l
with the limits of Figure 3.2-1 since a Reactor Coolant System total flow rate measurement uncertainty, greater than or equal to the value stated on Figure i
3.2-1, has been allowed for in determination of the design DNBR value.
l The measurement error for Reactor Coolant System total flow rate is based upon the performance of past precision heat balances. Sets of elbow tap i
coefficients, as determined during these heat balances, were averaged for each elbow tap to provide a single set of elbow tsp coefficients for use in calculating Reactor Coolant System flow. This set of coefficients establishes l
the calibration of the Reactor Coolant System flow rate indicators and becomes the set of elbow tap' coefficients used fo. Reactor Coolant System flow l
measurement. Potential fouling of the feedwater venturi, which might not be detected, could bias the result from these heat balances in a non-conservative manner. Therefore, a penalty of 0.1% for undetected fouling of the feedwater venturi is included in Figure 3.2-1.
Any fouling which might bias the Reactor Coolant System flow rate measurement greater than 0.1% can be detected by monitoring and trending various plant performance parameters.
If detected, action shall be taken before performing subsequent precision heat balance measurements, i.e., ef ther the effect of the fouling shall be quantified and i
compensated for in the Reactor Coolant System flow rate measurement or the venturi shall be cleaned to eliminate the fouling.
The 12-hour periodic surveillance of these parameters through instrument readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation.
Indication instrumentation measurement uncertainties are accounted for in the limits provided in Table 3.2-1.
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CATAWBA - UNITS 1 & 2 B 3/4 2-4 Amendment No.128 (Unit 1)
Amendment No.122 (Unit 2)
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