ML20087L501
| ML20087L501 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 08/21/1995 |
| From: | Wharton L NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20087L503 | List: |
| References | |
| NUDOCS 9508250310 | |
| Download: ML20087L501 (10) | |
Text
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- 4 UNITED STATES j
NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20066 4 001
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UNION ELECTRIC COMPANY CALLAWAY PLANT. UNIT 1 DOCKET NO. 50-483 AMENOMENT TO FACILITY OPERATING LICENSE Amendment No. 102 License No. NPF-30 1.
The Nuclear Regulatory Commission (the Commission) has found that:
I A.
The application for amendment filed by Union Electric Company (UE, the licensee) dated April 17, 1995, as supplemented June 30, 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 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.
2.
Accordingly, the license is amended by changes to the Technical Specifica-tions as indicated in the attachment to this license amendment, and para-graph 2.C.(2) of Facility Operating License No. NPF-30 is hereby amended to read as follows:
4 f
6 9508250310 950821
~
PDR ADOCK 05000483 P
f (2) Technical Soecifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No.102
, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into the license. UE shall operate the facility in 2
i accordance with the Technical Specifications and the Environmental Protection Plan.
l 3.
This license amendment is effective as of its date of issuance.
The Technical Specifications are to be implemented within 30 days from the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Lyn L. Raynard Wharton, Project Manager 4
Project Directorate III-3 Division of Reactor Projects - III/IV 4
Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of issuance:
August 21, 1995 4
4 1
i n
1 ATTACHMENT TO LICENSE AMENDMENT N0. 102 OPERATING LICENSE NO. NPF-30 DOCKET NO. 50-483 Revise Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages. The revised pages are identified by the captioned amendment number and contains marginal lines indicating the area of change.
MMQy1 INSERT 2-4 2-4 2-7 2-7 2-9 2-9 2-10 2-10 B 2-5 B 2-5 B 2-6 8 2-6 j
B 2-6a B 2-6a i
i
-r
TABLE 2.2-1 h
REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS E
k TOTAL SENSOR ERROR i
FUNCTIONAL UNIT All0WANCE fTA)
I 1S1 TRIP SETPOINT ALLOWA8LE VALUE E
- 1. Manual Reactor Trip N.A.
N.A.
N.A.
N.A.
N. A.
U
- 2. Power Range, Neutron Flux
~
a.
High Setpoint 7.5 4.56 0
s109% of RTP*
$112.3% of RTP*
b.
Low Setpoint 8.3 4.56 0
s25% of RTP*
s28.3% of RTP*
- 3. Power Range, Neutron Flux, 2.4 0.5 3
s4% of RTP*
$6.3% of RTP*
High Positive Rate with a time with a time constant 22 constant 22 seconds seconds
- 4. Deleted
?
- 5. Intermediate Range, 17.0 8.41 0
s25% of RTP*
$35.3% of RTP*
Neutron Flux 5
5
- 6. Source Range, Neutron Flux 17.0 10.01 0
s10 cps sl.6 x 10 cps
- 7. Overtemperature AT 9.3 6.47 1.83 See Note 1 See Note 2
+1.24***
- 8. Overpower AT 5.0 1.90 1.65 See Note 3 See Note 4 l
- 9. Pressurizer Pran ure-Low 5.0 2.21 2.0 21885 psig 21874 psig
- 10. Pressurizer Pressure-High 7.5 4.96 1.0 s2385 psig s2400 psig
_.k
- 11. Pressurizer Water Level-8.0 2.18 2.0
$92% of s93.8% of l
8 High instrument span instrument span g
- 12. Reactor Coolant Flow-Low 2.5 1.38 0.6 190% of loop 288.8% of loop minimum minimum 2
measured flow **
measured flow **
.M
'g RTP = RATED THERMAL POWER j-g Minimum Measured Flow - 95,660 gpm l
Two Allowances (temperature and pressure, respectively)
TABLE 2.2-1-(Continuedl TABLE NOTATIONS 9
r-
-g; NOTE 1: OVERTEMPERATURE AT
-E x
AT (1+r,5)
(1+T S) T 1
4 1-i
-T' 4 K (P-P')- f (AI)}
s AT, { K,- K c-2 3
i
_(l"2) 1 +r S (1+r 5) 1 +r S S
z 3
3 6
y w
Where: AT
-Measured AT; L
1Q Lead-Lag compensator on measured AT; 1+r 5 2
r,rz Time constants utilized in lead-lag compensator for AT,r 28s,7 53s; i
3 2
I Lag compensator on measured AT; 1+r S 3
4 73 Time constant utilized in the lag compensator for AT, r3 - Os; AT, Indicated AT at RATED THERMAL POWER; K,
1.15;
=
K 0.0251/*F; 2
1+r,5 The function generated by the lead-lag compensator for T dynamic 1+r 5 compensation; 3
}
Time constants utilized in the lead-lag compensator for T,,,
r 2 28s, T,73 4
4 a
i s 4s; g!
s I
T Average temperature,*F; c+
l 1
Lag compensator on measured T.,;
z 1+r S 6
.M Time constant utilized in the measured T,, lag compensator, 76-Os; 7
6 g
TABLE 2.2-1 (Continued)
TABLE NOTATIONS:fContinued) g-g NOTE 3:
OVERPOWER AT 5
AT (1+r,5) 5 AT, { K - K T - K [T
- 7"]- f (AT)}
1 7S 1
1 i
7 6
3 6
2 e
(1+T S) 1 +7 S 1 +7 S, 1 +r S 1 +7 S z
z 3
7 6
6 4
e-.
Where: AT Measured AT; Lead-Lag compensator on measured AT; J111 3
1+r S 2
Time constants utilized in lead-lag compensator for AT,r h8s,7 s3s; r,72 i
i 2
1 Lag compensator on measured AT; a
1+r S 3
t Time constant utilized in the lag compensator for AT, r3 - Os; 7
3 Indicated AT at RATED THERMAL POWER; AT o
K.
1.090; l
K3 0.02/*F for increasing average temperature and 0 for decreasing. average 8 k_
temperature; El
_r S__
The function generated by the rate-lag compensator for T, dynamic 7
1+r S compensation; 7
E Time constant utilized in the rate-lag compensator for T
, 7 2 los; 7
7 7
M
.'g
_1 Lag compensator on measured T,;
=
1+7 S 6
Time constant utilized in the measured T, lag compensator, r6-Os; 7
6 I
TABLE 2.2-1 (Continued)
TABLE NOTATIONS fContinued)
%g NOTE: 3 (Continued) l 7
K.
0.0015/*F for T > T" and K6-O for T s T";
E T
Average Temperature, *F; Indicated T at RATED THERMAL POWER (Calibration temperature for AT T"
~
y instrumentation, s 588.4*F);
Laplace transform operator, s"; and S
=
f (AI) 0 for all AI.
=
2 NOTE 4:
The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than l
7 2.4% of AT span.
o Mes
!!a
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m
l LIMITING SAFETY SYSTEM SETTINGS BASES l
Intermediate and Source Ranoe. Neutron Fim The Intermediate and Source Range, Neutron Flux trips provide core
]
protection during reactor startup to mitigate the consequences of an uncontrolled rod cluster control assembly bank withdrawal from a.
4 subcritical condition.
These trips provide redundant protection to the Low Setpoint trip of the Power Range, Neutrcn Flux chgnnels. The Source Range channels will initiate a Reactor trip at about 10 counts per second unless 1
manually blocked when P-6 becomes active.
The Intermediate Range channels will initiate a Reactor trip at a current level equivalent to approximately i
25% of RATED THERMAL POWER unless manually blocked when P-10 becomes active.
Overtemoerature AT The Overtemperature AT trip provides core protection to prevent DNB for all combinations of pressure, power, coolant temperature, and axial power distribution, provided that the transient is slow with respect to piping 4
transit delays from the core to the temperature detectors, and pressure is within the range between the Pressurizer High and Low Pressure trips.
The Setpoint is automatically varied with:
(1) coolant temperature to correct for temperature induced changes in density and heat capacity of water and i
i includes dynamic compensation for piping delays from the core to the loop temperature detectors, (2) pressurizer pressure, and (3) axial power i
i distribution. With normal axial power distribution, this Reactor Trip limit is always below the core Safety Limit as shown in Figure 2.1-1.
If axial peaks are greater than design, as indicated by the difference between j
top and bottom power range nuclear detectors, the Reactor trip is automatically reduced according to the notations in Table 2.2-1.
l Delta-T, as used in the Overtemperature and Overpower AT trips, represents the 100% RTP value as measured by the plant for each loop.
For j
the startup of a refueled core until measured at 100% Rated Thermal Power (RTP), Delta T is initially assumed at a value which is conservatively lower than the,last measured 100% RTP Delta T for each loop.
This normalizes each loop's AT trips to the actual, operating conditions existing i
at the time of measurement, thus forcing the trip to reflect the equivalent full power conditions as assumed in the accident analyses. These differences in vessel AT can arise due to several factors, the most prevalent being measured RCS loop flows greater than Minimum Measured Flow, i
and slightly asymmetric power distributions between quadrants.
While RCS loop flows are not expected to change with cycle life, radial power redistribution between quadrants may occur, resulting in small changes in loop specific vessel AT values. Accurate determination of the loop specific vessel AT value should be made when performing the Incore/Excore quarterly recalibration and under steady state conditions (i.e., power distributions not affected by Xe or other transient conditions).
l The time constants utilized in the lag compensation of measured AT, 7,
3 and measured T, r, are set in the field at 0 seconds. This setting corresponds to T$e f300 NLL cards used for lag compensation of these signals.
Safety analyses that credit Overtemperature AT for protection must account for these field adjustable lag cards as well as all other i
i CALLAWAY - UNIT 1 B 2-5 Amendment No. 28, 67, 102 i
LIMITING SAFETY SYSTEM SETTINGS BASES i
j Overtemperature AT (Continued) first order lags (i.e., the combined RTD/thermowell response time and the scoop transport delay and thermal lag). The safety analyses use a total j
first order lag of less than or equal to 6 seconds.
l Overpower AT
?
The Overpower AT trip provides assurance of fuel integrity (e.g., no
}
fuel pellet melting and less than 1% cladding strain) under all possible j
overpower conditions, limits the required range for Overtemperature AT l
trip, and provides a backup to the High Neutron Flux Trip.
The Setpoint is automatically varied with:
(1) coolant temperature to 2
correct for temperature induced changes in density and heat capacity of l
water, and (2) rate of change of temperature for dynamic compensation for piping delays from the core to the loop temperature detectors, to ensure i
that the allowable heat generation rate (kW/ft) is not exceeded. The Overpower AT trip provides protection to mitigate the consequences of i
various size steam breaks as reported in WCAP-9226, " Reactor Core Response
?
to Excessive Secondary Steam Releases."
Delta-T,, as used in the Overtemperature and Overpower AT trips, represents the 100% RTP value as measured by the plant for each loop.
For the startup of a refueled core until measured at 100% Rated Thermal Power (RTP), Delta T is initially assumed at a value which is conservatively lower than the,last measured 100% RTP Delta T for each loop.
This o
normalizes each loop's AT trips to the actual operating conditions existing at the time of measurement, thus forcing the trip to reflect the equivalent full power conditions as assumed in the accident analyses.
These differences in vessel AT can arise due to several factors, the most prevalent being measured RCS loop flows greater than Minimum Measured Flow, and slightly asymmetric power distributions between quadrants.
While RCS loop flows are not expected to change with cycle life, radial power redistribution between quadrants may occur, resulting in small changes in loop specific vessel AT values. Accurate determination of the loop specific vessel AT value should be made when performing the Incore/Excore quarterly recalibration and under steady state conditions (i.e., power distributions not affected by Xe or other transient conditions).
The time constants utilized in the lag compensation of measured AT, 7,
3 and measured T, T, are set in the field at 0 seconds.
This setting corresponds to he f300 NLL cards used for lag compensation of these signals. Safety analyses that credit Overpower AT for protection must account for these field adjustable lag cards as well as all other first order lags (i.e., the combined RTD/thermowell response time and the scoop 1
transport delay and thermal lag).
The safety analyses use a total first order lag of less than or equal to 6 seconds.
CALLAWAY - UNIT 1 B 2-6 Amendment No. g, 102
LIMITING SAFETY SYSTEM SETTINGS BASES 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 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 pressurizer 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 l
CALLAWAY - UNIT 1 B 2-6s Amendment No. 28 102
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