ML20153H344
| ML20153H344 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 09/24/1998 |
| From: | Chandu Patel NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20153H348 | List: |
| References | |
| NUDOCS 9810020014 | |
| Download: ML20153H344 (11) | |
Text
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UNITED STATES
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NUCLEAR REGULATORY COMMISSION WASHINGTON D.C. 2065Ho01 49
,o ENTERGY OPERATIONS. INC.
DOCKET NO. 50-382 WATERFORD STEAM ELECTRIC STATION. UNIT 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.145 License No. NPF-38 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Entergy Operations, Inc. (the licensee) dated September 18,1998, as superseded by letter dated September 23,1998, complies with the standards and requirements of the Atomic Energy Act of1954, as amended (the Act), and the Commiss:on's rules and regulations 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, and (ii) that such activities will be conducted in compliance with the Commission's regulations; 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.
9810020014 980924 PDR ADOCK 05000382 P
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. l 2.
Accordingly, the license is amended by changes to the Technical Specifications as l
indicated in the attachment to this license amendment, and paragraph 2.C(2) of Facility l
Operating License No. NPF-38 is hereby amended to read as follows:
(2)
Technical Soecifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 145, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in 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 Mcb4 f Chandu P. Patel, Project Manager l
Project Directorate IV-1 Division of Reactor Projects lil/IV Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: September 24, 1998 1
4 l
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ATTACHMENT TO LICENSE AMENDMENT NO.145 TO FACILITY OPERATING LICENSE NO. NPF-38 DOCKET NO. 50-382 Replace the following pages of the Appendix A Technical Specifications with the attached pages. The revised pages are identified by Amendment number and contain verticallines indicating the areas of change. The corresponding overleaf pages are also provided to maintain document cornpleteness.
- REMOVE PAGES INSERT PAGES 2-3 2-3 2-4 2-4 3/4 3-4 3/4 3-4 3/4 34a 3/4 3-12 3/4 3-12 B 2-3 B 2-3
.3
L IABLE 7 7-1
,r REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES
- 1. Manual Reactor Trip -
Not Applicable Not Apphcable
- 2. Linear Power Level-High Four Reactor Coolant Pumps 5108% of RATED THERMAL POWER 5108.76% of RATED THERMAL POWER OpeiaGng Lo uiling Power Level-High (1) s 0.257% of RATED THERMAL POWER (6) s 0.280% of RATED THERMAL POWER (6) l 3.
v i
4.
Pressurizer Pressure - High 52350 psia s 2359 psia
[
5.
Pressurizer Pressure - Low 21684 psia (2) 21649.7 psia (2) i 6.
Containment Pressure - High 517.1 psia 517.4 psia i
7.
Steam Generator Pressure - Low 2764 psia (3) 2 749.9 psia (3) 8.
Steam Generator Level-Low 2 27.4 % (4) 2 26.48% (4) t 9.
Local Power Density - High s 21.0 kW/ft (5) s 21.0 kW/ft (5)
- 10. DNBR - Low 2 1.26 (5) 2 1.26 (5)
- 11. Steam Generator Level-High 5 87.7 % (4) s 88.62 % (4) i
- 12. Reactor Paviedien System Logic Not Applicable Not App;kisbie
[
- 13. Reactor Trip Breakers Not Applicable Not Apphcable i
- 14. Core Protechon Calculators Not Applicable Not Apphcable j
- 15. CEA Calculators Not Applicable Not Apphcable l
- 16. Reactor Coolant Flow - Low 219.00 psid (7) 218.47 psid (7) f WATERFORD - UNIT 3 2-3 Amendment No.12:443.145 i
TABLE 2.21 (Continued)
REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS TABLE NOTATIONS (1)
Trip may be manually bypassed above 10 % of RATED THERMAL POWER *; bypass shall be d
automatically removed when THERMAL POWER
- is less than or equal to the reset point of the bistable. The reset point shall be within 3.0x10 % of RATED THERMAL POWER
- below the 4
bistable setpoint which is nominally 10 % of RATED THERMAL POWER *. This accounts for the d
deadband of the bistable.
(2)
Value may be decreased manually, to a minimum of 100 psia, as pressurizer pressure is reduced, provided the margin between the pressurizer pressure and this value is maintained at less than or equal to 400 psi; the setpoint shall be increased automatically as pressurizer pressure is increased until the trip setpoint is reached. Trip may be manually bypassed below 400 psia; bypass shall be automatically removed whenever pressurizer pressure is greater than or equal to 500 psia.
1 (3)
Value may be decreased manually as steam gerierator pressure is reduced, provided the margin between the steam generator pressure and tnis value is maintained at less than or equal to 200 psi; the setpoint shall be increased automatically as steam generator pressure is increased until the trip setpoint is reached.
(4)
% of the distance between steam generator upper and low level instrument nozzles.
1 (5)
As stored within the Core Protection Calculator (CPC). Calculation of the trip setpoint includes measurement, calculational and processor uncertainties, and dynamic allowances. Trip may be manually bypassed below 10d% of RATED THERMAL POWER *; bypass shall be automatically removed when THERMAL POWER
- is greater than or equal to 10"% of RATED THERMAL POWER *.
(6)
As measured by the Logarithmic Power Channels.
(7)
The setpoint may be altered to disable trip function during testing pursuant to Specification 3.10.3.
- As measured by the Logarithmic Power Channels.
WATERFORD - UNIT 3 2-4 Amendment No. +2,145
'.i
,i TABLE 3.3-1 g
REACTOR PROTECTIVE INSTRUMENTATION E
I g
MINIMIM TOTAL No.
CHANNELS CHANNELS APPLICABLE FUNCTIONAL IMIT OF CHANNELS TO TRIP OPERABLE M00ES ACTION t
{
Q 1.
Manual Reactor Trip 2 sets of 2 1 set of 2
- 2. sets of 2 1 2 1
r w
2 sets of 2 I set of 2 2 sets of 2 A3. 4*, $*
8 2.
Linear Power Level - Nigh 4
2 3
1, 2 2f, 3#
3.
Logaritissic Power Level-Nigh I
Startup and Operetlag 4
2(a)(d) 3 2**
N, N i
a.
4 2
3 3*, 4*, 5*
8 b.
Shutdown 4
0 2
3,4,5 4
l l
4.
Pressurizer Pressure - Nigh 4
2 3
1, 2 28, 38 y
3.
Pressurizer Pressure - Low 4
2(b) 3 1, 2 2f, 38 l
w 6.
Containment Pressure - High 4
2 3
1, 2 28, 38 l
7.
Steam Generator Pressure - Low 4/SG 2/SG 3/SG 1, 2 28, 38 i
8.
Steam Generator Level - Low 4/SG 2/SG 3/SG 1, 2 28, 3#
9.
Local Pswer Density - Nigh 4
2(c)(d) 3 1, 2 N 38
- 10. DNBR - Low 4
2(c)(d) 3 1, 2 28, 3#
- 11. Steam Generator Level - Nigh 4/SG 2/SG(g) 3/SG 1, 2 2#, 38 i
j
- 12. Reactor Protection System Logic 4
2 3
1 2 5
C A
m 3. 4*, $*
8 l
5
- 13. Reactor Trip Breakers 4
2(f) 4 1 2 5
in i
A i
N 3. 4*, 5*
8
- 14. Core Protection Calculators 4
2(c)(d) 3 1, 2 27 N and 7
=P
- 15. CEA Calculators 2
1 2(e) 1, 2 6 and 7
}
r
,M 16. Reactor Coolant Flow - Low 4/SG 2/SG(c) 3/SG 1, 2 2#, M l
I i
i os
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TABLE 3.31 (Continued)
TABLE NOTATION
- With the protective system trip breakers in the closed position, the CEA drive system capable of CEA withdrawal, and fuel in the reactor vessel.
- The provisions of Specification 3.0.4 are not applicable.
- Not applicable above 10d% RATED THERMAL POWER.0) l (a)
Trip may be manually bypassed above 10d% of RATED THERMAL POWER 0); bypass shall be automatically removed when THERMAL POWERU)is less than or equal to the reset point of the bistable. The reset point shall be within 3.0x104% of RATED THERMAL POWER")below the bistable setpoint which is nominally 10d% of RATED THERMAL POWER"). This accounts for the deadband of the bistable.
(b)
Trip may be manually bypassed below 400 psia; bypass shall be automatically removed whenever pressurizer pressure is greater than or equal to 500 psia.
(c)
Trip may be manually bypassed below 104% of RATED THERMAL POWERU); bypass shall be automatically removed when THERMAL POWERU)is greater than or equal to 10d% of RATED THERMAL POWERU). During testing pursuant to Special Test Exception 3.10.3, trip may be manually bypassed below 5% of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is greater than or equal to 5% of RATED THERMAL POWER.
(d)
Trip may be bypassed during testing pursuant to Special Test Exception 3.10.3.
(e)
See Special Test Exception 3.10.2.
l (f)
Each channel shall be comprised of two trip breakers; actual trip logit shall be one-out-of-l two taken twice.
l (g)
High steam generator level trip may be manually bypassed in Modes 1 and 2, at 20%
l power and below.
") As measured by the Logarithmic Power Channels.
l WATERFORD - UNIT 3 3/434 Amendment No. 14,40,100,145
TABLE 3.31 (Continued) 9 TABLE NOTATION ACTION STATEMENTS ACTION 1 -
With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 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/or open the protective system trip breakers.
ACTION 2 -
With the number of channels OPERABLE one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may continue provided the inoperable channel is placed in the bypassed or tripped condition within i hour. If the ir.apr2!c channelis bypassed, the desirability of maintaining this enannel in the bypassed condition shall be documented by the Plant Operations Review Committee in accordance with plant administrative procedures. The channel shall be returned to OPERABLE status prior to STARTUP following the next COLD SHUTDOWN.
WATERFORD - UNIT 3 3/4 34a Amendment No. 145
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TABLE 4.3-1 (Continued) 5 REACTOR PROTECTIVE INSTRUNDtrATION SW VEILLANCE REQUfREIGrTS 3
B CNANNEL MODES FOR WittM CHANNEL CHANNEL FwCTIONAL SURVEILLANCE e
5 TUNCTIONAL INIIT
_ CHECK CALIBRATION TEST
_IS REqulRED' w
- 13. Reactor Trip Breakers N.A.
N.A.
M(10),$/U(1) 1,2,3*,44,$8 j
14 Core Protection Calculators D(2,4),R(4,5) Q(9),R(s) 1, 2 r
- 15. cEA taicaisters R
g,R(s) 1, 2
- 16. Reactor Coolant Flow - Law R
Q 1, 2
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TABLE 4.3-1 (Continued)
TARLE NOTATIONS
- With the reactor trip breakers in the closed position, the CEA drive system capable of CEA withdrawal, and fuel in the reactor vessel.
- The provisions of Specification 4.0.4 are not applicable when reducing reactor power to less than 10 % of RATED THERMAL POWER *)from a reactor power level greater than d
d 10 % of RATED THERMAL POWERS). Upon reducing power below 10 % of RATED d
THERMAL POWERS), a CHANNEL FUNCTIONAL TEST shall be performed within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> if not performed during the previous 31 days. This requirement does not apply with the reactor trip breakers open.
(1)
Each startup or when required with the reactor trip breakers closed and the CEA drive system capable of rod withdrawal, if not performed in the previous 7 days.
(2)
Heat balance only (CHANNEL FUNCTIONAL TEST not included):
a.
Between 15% and 80% of RATED THERMAL POWER, compare the Linear Power Level, the CPC at AT power, and CPC nuclear power signals to the calorimetric calculation.
If any signal is within -0.5% to +10% of the calorimetric calculation, then do not calibrate except as required during initial power ascension following refueling.
If any signal is less than the calorimetric calculation by more than 0.5%, then adjust the affected signal (s) to within 0.0% to +10.0% of the calorimetric caluclation.
If any signal is greater than the calorimetric calculation by more than 10%, then adjust the affected signal (s) to within 0.0% to 10% of the calorimetric.
b.
At or above 80% of RATED THERMAL POWER, compare the Linear Power Level, the CPC A T power, and CPC nuclear power signals to the calorimetric calculation.
If any signal differs from the calorimetric calculation by an absolute difference of more than 2%, then adjust the affected signal (s) to agree with the calorimetric l
calculation within -2% to +2%.
During PHYSICS TESTS, these daily calibrations may be suspended provided these calibrations are performed upon reaching each major test power plateau and prior to proceeding to the next major test power plateau.
I l
(3)
Above 15% of RATED THERMAL POWER, verify that the linear power subchannel l
gains of the excore detectors are consistent with the values used to establish the
(
shape annealing matrix elements in the Core Protection Calculators.
(4)
Neutron detectors may be excluded from CHANNEL CALIBRATION.
(5)
After each fuelloading and prior to exceeding 70% of RATED THERMAL POWER, the incore detectors shall be used to determine or verify acceptable values for the shape annealing matrix elements used in the Core Protection Calculators.
- )As measured by the Logarithmic Power Channels.
WATERFORD - UNIT 3 3/4 3-12 Amendment No. 69;+25,145 l
BASES Manual Rametor Trio The Manual Reactor Trip is a redundant channel to the automatic protective instrumentation channels and provides manual reactor trip capability.
Linear Power Level - High The Linear Power Level-High trip provides reactor core protection against rapid reactivity excursions which might occur as the result of an ejected CEA. This trip initiates a reactor trip at a linear power level of less than or equal to 108% of RATED THERMAL POWER.
Locarithmic Power Level-High The Logarithmic Power Level-High trip is provided to protect the integrity of fuel cladding j
and the Reactor Coolant System pressure beundary in the event of an unplanned criticality from a shutdown condition. A reactor trip is initiatec by the Logarithmic Power Level-High trip at a THERMAL POWER
- level of less than or equal to 0.257% of RATED THERMAL POWER
- unless this trip is manually bypassed by the operator. The operator may manually bypass this trip when the THERMAL POWER
- level is above 10-4% of RATED THERMAL POWER *; this bypass is automatically removed when the THERMAL POWER
- level decreases to 10-4% of RATED THERMAL POWER *.
Pressurizar Pressure - High The Pressurizer Pressure - High trip, in conjunction with the pressurizer safety valves and main steam safety valves, provides Reactor Coolant System protection against overpressurization in the event of loss of load without reactor trip. This trip's setpoint is at less than or equal to 2350 psia which is below the nominal lift setting of 2500 psia for the pressurizer safety valves and its operation avoids the undesirable operation of the pressurizer safety valves.
Pressurizar Pressure - Low The Pressurizer Pressure - Low trip is provided to trip the reactor and to assist the Engineered Safety Features System in the event of a Loss of Coolant Accide11. During normal operation, this trip's setpoint is set at greater than or equal to 1684 psia. This, rip's setpoint may be manually decreased, to a minimum value of 100 psia, as pressurizer presaure is reduced during plant shutdowns, provided the margin between the pressurizer pressure and this trip's setpoint is maintained at less than or equal to 400 psi; this setpoint increases automatically as pressurizer pressure increases until the trip setpoint is reached.
- As measured by the Logarithmic Power Channels.
WATERFORD - UNIT 3 B 2-3 Amendment No. Ms,145
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