ML20080F324
| ML20080F324 | |
| Person / Time | |
|---|---|
| Site: | McGuire, Mcguire  |
| Issue date: | 01/12/1995 |
| From: | Berkow H Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20080F327 | List: |
| References | |
| NUDOCS 9501200031 | |
| Download: ML20080F324 (10) | |
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UNITED STATES s
j NUCLEAR REGULATORY COMMISSION t
WASHINGTON, D.C. 20086 4 001 49.....,o DUKE POWER COMPANY DOCKET NO. 50-369 McGUIRE NUCLEAR STATION. UNIT 1 AMENDMENT TO FACILITY OPERATING l.ICENSE Amendment No.153 License No. NPF-9 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment to the McGuire Nuclear Station, Unit 1 (the facility), Facility Operating License No. NPF-9 filed by the Duke Power Company (licensee) dated January 10, 1994, as supplemented September 15, 1994, January 5 and 10, 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 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 Comission'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.
9501200031 950112 DR ADOCK 050003 9
i 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-9 is hereby amended to read as follows:
Technical Soecifications The Technical Specifications contained in Appendix A, as revised through Amendment No.153, are hereby_ incorporated into this license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This license ame.ylment is effective as of its date of issuance and shall j
be implemented with!n 30 days from the date of issuance, i
l FOR THE NUCLEAR REGULATORY COMMISSION rbert N. Berkow, Director Project Directorate II-3 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Technical Specification Changes j
i Date of Issuance:
January 12, 1995
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t UNITED STATES yq NUCLEAR REGULATORY COMMISSION f
WASHINGTON D.C. 2066H001 DUKE POWER COMPANY DOCKET NO. 50-370 McGUIRE NUCLEAR STATION. UNIT 2 i
AMERDMENT TO FACILITY OPERATING LICENSE Amendment No. 135 1
License No. NPF-17 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment to the McGuire Nuclear Station, Unit 1 (the facility), Facility Operating License No. NPF-17 filed by the Duke Power Company (licensee) dated January 10, 1994, as supplemented September 15, 1994, January 5 and 10, 1995, complies with the standards and requirements of the Atomic Energy Act of 4354, 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 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.
f k r 2.
Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license I
amendment, and Paragraph 2.C.(2) of Facility Operating License No.
NPF-17 is hereby amended to read as follows:
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 135, are hereby incorporated into this license.
The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Pl an.
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 erbert 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:
January 12, 1995 i
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ATTACHMENT TO LICENSE AMENDMENT NO. 153 FACILITY OPERATING LICENSE NO. NPF-9 l
DOCKET NO. 50-369 AND TO LICENSE AMENDMENT NO. 135 FACILITY OPERATING LICENSE NO. NPF-17 l
DOCKET NO. 50-370 Replace the following pages of the Appendix "A" Technical Specifications with
.j the enclosed pages. The revised pages are identified by Amendment number and contain vertical lines indicating the areas of change.
l Remove Paaes Insert Paaes 2-5 2-5 B 2-6 B 2-6 3/4 2-22a 3/4 2-22a l
3/4 3-14a 3/4 3-14a 1
B 3/4 2-5 8 3/4 2-5 B 3/4 2-Sa B 3/4 2-Sa l
M TABLE 2.2-1 8
REACTOR TRIP SYSTEM INSTRUMENTATION TRIF SETPOINTS m
FUNCTIONAL UNIT TRIP SETPomT ALLOWABLE VALUES E
g
- 1. Manual Reactor Trip N.A.
N.A.
- 2. Power Range, Neutron Flux Low Setpoint - s 25% of RATED Low Setpoint - s 26% of RATED g
THERMAL POWER THERMAL POWER N
High Setpoint - s 109% of RATED HighSetpoint - s 110% of RATED THERMAL POWER THERMAL POWER
- 3. Power Range, Neutron Flux, s 5% of RATED THERMAL POWER with s 5.5% of RATED THERMAL POWER High Positive Rate a time constant 2 2 seconds with a time constant 2 2 seconds
- 4. Intermediate Range, Neutron s 25% of RATED THERMAL POWER s 30% of RATED THERMAL POWER Flux m
5 5
- 5. Source Range, Neutron Flux s 10 counts per second s 1.3 x 10 counts per second
- 6. Overtemperature AT See Note 1 See Note 3 g
- 7. Overpower AT See Note 2 See Note 4
- 8. Pressurizer Pressure--Low 2 1945 psig 2 1935 psig 55
- 9. Pressurizer Pressure--High s 2385 psig s 2395 psig 55
- 10. Pressurizer Water Level--High s 92% of instrument span s 93% of instrument span
- 11. Low Reactor Coolant Flow 2 91% of minimum measured 2 90% of minimum measured l
flow per loop
- flow per loop
- __FF 3;;
50
- Minimum measured flow is 95,500 gpm per loop.
1 m
-. m
.mm_.
~.
LIMITING SAFETY SYSTEM SETTINGS BASES Pressurizer Pressure In each of the pressure channels, there are two independent bistables, each with its own trip setting to provide for a High and Low Pressure trip thus limiting the pressure range in which reactor operation is permitted. The Low Setpoint trip protects against low pressure which could lead to DNB 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 impulse chamber pressure at approximately 10% of full power equivalent); and on increasing power, automatically reinstated by P-7.
The High Setpoint trip functions in conjunction with the pressurizei relief and safety valves to protect the Reactor Coolant System against system 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 High Water Level trip is automatically blocked by P-7 (a power level of approximately 10% of RATED THERMAL POWER with a turbine impulse chamber pressure at approximately 10% of full equivalent); and on increasing power, automatically reinstated by P-7.
Low Reactor Coolant Flow The Low Reactor Coolant Flow trips provide core protection to prevent DNB 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 i
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 the P-7 an automatic Reactor trip will occur on loss of flow in more than one loop and below P-7 the trip function is automatically blocked.
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McGUIRE - UNITS 1 & 2 B 2-6 Amendment No. 153 (Unit 1)
Amendment No. 135 (Unit 2)
5 13 POWER DISTRIBUTION LIMITS 3/4.2.5 DNB PARAMETERS SURVEILLANCE REQUIREMENTS 4.2.5.1 Each of the parameters of Table 3.2-1 shall be measured by averaging the indications (meter or computer) of the operable channels and 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 />.
r 4.2.5.2 The RCS total flow rate indicators shall be subjected to a CHANNEL CALIBRATION at least once per 18 months.
l 4.2.5.3 The RCS total flow rate shall be determined by measurement at least l
once per 18 months.
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McGUIRE - UNITS 1 AND 2 3/4 2-22a Amendment No. 153 (Unit 1)
Amendment No. 135 (Unit 2) w
TABl.E 4.3-1 (Continued)
TABLE NOTATION (11) -
The TRIP ACTUATING DEVICE OPERATIONAL TEST shall independently verify-the OPERABILITY of the undervoltage and shunt trip circuits for the Manual Reactor Trip Function.
(12) -
The TRIP ACTUATING DEVICE OPERATIONAL TEST shall independently verify i
the OPERABILITY of the undervoltage and shunt trip attachments of the Reactor Trip Breakers.
(13) -
Prior to placing breaker in service, a local manual shunt trip shall be performed.
(14) -
The automative undervoltage trip capability shall be verified
- operable, (15) -
Overtemperature setpoint, overpower setpoint, and T,, channels require an 18 month channel calibration. Calibration of the AI i
channels is required at the beginning of each cycle upon completion of the precision heat balance. RCS loop AT values shall be
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determined by precision heat balance measurements at the beginning of r
each cycle.
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McGUIRE - UNITS I and 2 3/4 3-14 a Amendment No.153 (Unit 1)
Amendment No. ~ 135 (Unit 2)
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5 POWER DISTRIBUTION LIMIIS BASES
{
3/4.2.4 OVADRANT POWER TILT RATIO The QUADRANT POWER TILT RATIO limit assures that the radial power distri-bution satisfies the design values used in the power capability analysis.
Radial power distribution measurements are made during STARTUP testing and periodically during power operation.
The limit of 1.02, at which corrective action is required provides DNB and linear heat generation rate protection with the x-y plane power tilts. The 7
peaking increase that corresponds to a QUADRANT POWER TILT RATIO of 1.02 is included in the generation of the AFD limits.
The 2-hour time allowance for operation with a tilt condition greater than 1.02 but less than-1.09 is provided to allow identification and correction of a dropped or misaligned rod.
In the event such action does not correct the tilt, the margin for uncertainty on F (X,Y,Z) is reinstated by reducing the power by o
3% from RATED THERMAL POWER for each percent of tilt in excess of 2.0%.
For purposes of monitoring QUADRANT POWER TILT RATIO when one excore detector is inoperable, the moveable incore detectors are used to confirm that the normalized symmetric power distribution is consistent with the QUADRANT POWER TILT RATIO. The incore detector monitoring is done with a full incore i
flux map or two sets of four symmetric thimbles.
3/4.2.5 DNB PARAMETERS The limits on the DNB-related parameters assure that each of the parameters are maintained within the normal steady-state envelope of operation assumed in the transient and accident analyses. The limits are consistent with the initial FSAR assumptions and have been analytically demonstrated adequate to maintain a design limit DNBR throughout each analyzed transient. As noted on Figure 3.2-1, RCS flow rate and THERMAL POWER may be " traded off" against one another (i.e., a low measured RCS flow rate is acceptable if the power level is decreased) 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, F,(X,Y), in Specification 3.2.3 are maintained. The indicated T values and tbeindicatedpressurizerpressurevaluescorrespondtoanalytica'flimitsof 592.6*F and 2220 psia respectively, with allowance for indication instrumen-tation measurement uncertainty. When RCS flow rate is measured, no additional allowances are necessary prior to comparison with the limits of Figure 3.2-1 since an RCS total flow rate measurement uncertainty, greater than or equal to the value stated on Figure 3.2-1 has been allowed for in determination of the design DNBR value.
s McGUIRE - UNITS 1 AND 2 B 3/4 2-5 Amendment No. 153 (Unit 1)
Amendment No. 135 (Unit 2)
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POWER DISTRIBUTION LIMITS BASES jf_4J. 5 DNB PARAMETERS (Continued)
The measurement error for RCS total flow rate is based upon the performance of past precision heat balances. Sets of elbow tap coefficients, as determined during these heat balances, were averaged for each elbow tap to provide a single set of elbow tap coefficients for use in calculating RCS flow. This set
-of coefficients establishes the calibration of the RCS flow rate indicators and becomes the set of elbow tap coefficients used for RCS flow 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 RCS 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., either the effect of the fouling shall be quantified and compensated for in the RCS 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.
Indica-tion instrumentation measurement uncertainties are accounted for in the limits provided in Table 3.2-1.
McGUIRE - UNITS 1 AND 2 8 3/4 2-5a Amendment No.153 (Unit 1)
Amendment No. 135 (Unit 2)
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