ML20086D804
| ML20086D804 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 11/07/1991 |
| From: | Black S Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20086D805 | List: |
| References | |
| NUDOCS 9111260264 | |
| Download: ML20086D804 (9) | |
Text
_ _ _ _ _ _ _. _ _. _ _ _ _ _ _ _ _
63320p
/
o UNITED STt.TES
!"s NUCLEAR REGULATORY COMMISSION
{ [ ',"v.
,i WASHINo ton. D. c. 20555 i
i 4
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WOLF CREEK NUCLEAR OPERATING CORPORATION WOLF CREEK GENERATING STATION DOCKET NO. 50-482 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 52 License No. NPF-42 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment to the Wolf Creek Generatint..tation (the facility) Facility Operating License No. NPF-42 filed oy the Wolf Creek Nuclear Operating Corporation (trie Corporation),
dated June 11, 1991 and supplemented by letters dated August 30, 1991, and September 20, 1991, 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, as 6 mended, 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 license 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.
I 9111260P64 911107 l
POR ADOCK 050004GP P
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.g.
2.
Accordingly, the license is amended by changes to tho Technical Specifi-cations as indicated in the attachment to this license amendment and Paragraph 2.C.(2) of Facility Operating License No. NPF-42 is hereby amended to read as follows:
2.
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.
52, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated in the license.
The Corporation shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
The ' license amendment is effective as of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
- 2. M /
'C Suzanne 4. Black, Director Project Directorate IV-2 Division of Reactor Projects - III/IV/V Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
November 7, 1991
ATTACHMENT TO LICENSE AMENDMENT NO. 52 FACILITY OPERATING LICENSE NO. NPF-42 DOCKET NO. 50-482 Revise Appendix A Technical Specifications by removing the pages identif.ed below and inserting the enclosed pages.
The revised pages are identified by amendment number and contain marginal lines indicating the area of change.
The corresponding overleaf pages are also provided to maintain document completeness.
REMOVE INSERT 2-4 2-4 2-7 2-7 2-8 2-8 2-9 2-9 2-10 2-10 3/4 3-9 3/4 3-9 3/4 3-12a 3/4 3-12a B 2-5 B 2-5
]
TABLE 2.2-1 t
P REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS
'k SENSOR g;
TOTAL ERROR FUNCTIONAL UNIT ALLOWANCE (TA)
Z (S)
TRIP SETPOINT ALLOWA8LE VALUE 1.
Manual Reactor Trip N.A.
N.A.
N.A.
N.A.
N.A.
2.
Power Range, Neutron Flux a.
High Setpoint 7.5 4.56 0
$109% of RTP*
$112.3% of RTP*
b.
Low Setpoint 8.3 4.56 0
$25% of RTP*
128.3% of RTP*
3.
Power Range, Neutron Flux, 2.4 0.5 0
<4% of RTP* with
<6.3% of RTP* with i time constant i time constant High Positive Rate 12 seconds 12 seconds 4.
Power Range, Neutron Flux, 2.4 0.5 0
<4% of RTP* with
<6.3% of RTP* with i time constant i time constant High Negative Hate 32 seconds 12 seconds 5.
Intermediate Range, 17.0 8.41 0
$25% of RTP*
$35.3% of RTP*
Neutron Flux 5
5 6.
Source *snge, Neutron Flux 17.0 10.01 0
110 cps
$1.6 x 10 cps 7.
Overtemperature AT 7.2 3.50 2.72 See Note 1 See Note 2 l
[
8.
Overpower AT 5.5 1.83 0.17 See Note 3 See Note 4 l
9.
Pressurizer Pressure-Low 3.7 0.71 2.49 11915 psig 11906 psig s
[
10.
Pressurizer Pressure-High 7.5 0.71 2.49
$2385 psig
$2400 psig a
(
11.
Pressurizer Water Level-High 8.0 2.18 1.96 192% of instrument
$93.9% of instrument o
span span b
- RTP = RATED THERMAL POWER
- Loop design flow = 93,750 gpm E
TABLE 2.2-1 (Continued) g r-TABLE NOTATIONS 9
NOTE 1:
OVERTEMPERATURE AT
[T (1 f s) - T'] + K (P - P') - f (AI)}
AT (1 f
- 5) i AT, {K1-K2
'4 3
y Ts 7
T3 Z
Measured AT; Where:
AT
=
1 Lead-lag compensator on measured AT;
=
Time constants utilized in lead-lag compensator for AT, It=8s,
=
T1, T2 T2 = 3 s; Lag compensator on measured AT;
=
1 T35 Time constant utilized in the lag compensatcr for AT, T3 = 0 s;
=
T3 Indicated AT at RATED THERMAL POWER; aT,
=
1.10; K1
=
0.0137/ F; K2
=
5
}5 The function generated by the lead-lag compensator for T,yg l
=
F.
dynamic compensation; R
Time constants utilizeri in the lead-lag compensator for Tyy9, I4 = 28 s,
=
T4, Ts 5
T3 = 4 s; Average temperature, 'f; z
T
=
1 Lag compensator on measured Tavg;
=
7 T85 Time constant utilized in the measured T lag cocpensator, To = 0 s;
=
rs avg
-. ~.,.
P
' TABLE 2.2-1 (Continued)
E TABLE NOTATIONS (Continued) n k
NOTE 1:
(Continued) x T'
588.5'F (Nominal T,y9 at RATED THERMAL POWER);
c-6 K3
.= '0.000671; Pressurizer pressure, psig;-
P
=
P' 2235 psig (Nominal RCS operating pressure);
=
S
= : Laplace transform operator, s 1; and f (AI) is a function of the indicated difference between top and bottom' detectors'of the t
power-range neutron ion chambers; with gains to be selected based on measured instrument--
response during plant STARTUP tests such that:
y (i) for qt ~9b between -27% and + 7%, f (AI) =.0, where qt and gb are percent t
RATED THERML POWER in the top and bottom halves of the core respectively, and qt[ 4b
- l is total THERMAL POWER in percent.of RATED THERMAL POWER; (ii) for each percent that the magnitude of qt exceeds -27%, tht AT Trip Setpoint 9b shall be automatically reduced by 1.57% of its value at RATED TIERMAL POWER; and (iii) for each percent that.the magnitude of qt ~9b exceeds +7%, the aT Trip Setpoint g
shall be automatically. reduced by 0.85% of its value at RATED THiRMAL POWER.
l 5
NOTE 2:
The channel's maximum Trip Setpoint shall not exceed its computed Trip.Setpoint by more than i
g 2.6% 'of AT span.
I O
r
4 TABLE 2.2-1 (Continued) 5 G
TABLE NOTATIONS (Continued) ny NOTE 3:
OVERPOWER AT T-4U 1f S -
- f (AI)}
2 IO 4
5 1
~
O T5 1
Ts5 Is T
1 + T35 o
7 3
-4 Where:
AT
= Measured AT; I
I I
lead-lag compensator on measured AT;
=
7 1
Time constants utilized in lead-lag compensator for AT, It = 8 s, T2 = 3 s; T2, 12
=
Lag compensator on measured AT;
=
1 + T35 l
Time constant utilized in the lag compensator for AT, 13 = 0 s;
=
13 Indicated AT at RATED THERMAL POWER; AT,
=
1.08; K4
=
0.02/ F for increasing average temperature and 0 for decreasing average Ks
=
temperature; The function generated by the rate-lag compensator for T,yg dynamic
=
F 1
175 compensation; R
Time constant utilized in the rate-lag compensator for T,,g, 17 = 10 s;
=
R t7 5
1 Lag compensator on measured T,yg;
=
7 3
Time constant utilized in the measured T lag compensator, is = 0 s; IS Ts
=
yg
~.
g TABLE 2.2-1 (Continued)
G TABLE NOTATIONS (Continued) 2 A
NOTE 3:
(Continued) m 0.00128/ F for T > T" and Kc = 0 for T 5 T";
Ks
=
Eq
-T
=
Average temperature, F;
T" Indicated T at RATED THERMAL POWER (Calibration temperature fc.- AT
=
avg instrumentation, 5 588.5 F);
Laplace transform operator, s 2; and S
=
f (AI)
= 0 for all A1.
2 t!OTE 4:
The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than y
3.7% of AT span.
l c$
iir a
8e W
4 TABLE 4.3.
REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS p,
TRIP.
g'
-ANALOG-ACTUATING MODES FOR
^
. CHANNEL.
DEVICE WHICH CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION
. SURVEILLANCE-g
' FUNCTIONAL UNIT
'CHECKL CALIBRATION TEST TEST LOGIC TEST IS REQUIRED
[
1.
Manual Reactor-Trip N.A.
N.A.
N.A.
R(11)
N.A.
1, 2, 3*, 4*, 5*
2.
Power Range, Neutron Flux a.
High Setpoint 5.
D(2,4)-
Q N.A.
N.A.
1, 2 M(3,4)
Q(4,6)
R(4,5)-
b.
' Low Setpoint 5
R(4)
S/U(1)
N.A.
N.A.
1##, 2-3.
Power Range, Neutron Flux,-
N.A.
R(4)
Q N. A.
N.A.
1, 2 R
High Positive Rate
=
y 4.
Power Range, Neutron Flux, N.A.
R(4)
Q N.A.
N.A.
1, 2 High Negative Rate m
5.
Intermediate Range, S
R(4,5)
S/U(1)
N.A.
I,. A.
1#N, 2 Neutron Flux.
6.
. Source Range, Neutron Flux 5
R(4,5,12)
S/U(1),Q(9)
N.A.
N.A.
2M, 3, 4, 5 7..
Overtemperature AT-S R
Q N.A.
N.A.
1, 2 l.
8.
Overpower AT S
R Q
N.A.
N.A.
1, 2 9.
Pressurizer Pressure-Low S
R.
Q N.A.
N.A.
1 10.
Pressurizer Pressure-High 5
R Q
N.A.
N.A.
1, 2
[
11.
Pressurizer Water Level-High S
R Q
N.A.
N.A.
Od 12.
Reactor Coolant Flow-Low 5
R
'Q N.A.
N.A.
1 T
M
v
-TABLE 4.3-l'(Continued) j REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS 4
n TRIP k
ANALOG
' ACTUATING-
' MODES FOR-t x
CHANNEL-DEVICE WHICH CHANNEL ' CHANNEL OPERATIONAL' OPERATIONAL ACTUATION SURVEILLANCE-i FUf!CTIONAL UNIT CHECK' CALIBRATION TEST TEST LOGIC TE5T IS REQUIRED g
[
13.
Steam Generator Water Level -
S R
Q(15)
N.A.
.N.A.
1, 2
(
. Low-Low 14.
Undervoltage - Reactor:
-N.A.
R
.N.A.
Q..
N.A.
1-
[
Coolant Pumps 15.
Underfrequency - Reactor.
N.A.
R N.A.
Q
.N.A.
I' Coolant Pumps-16.
R a.
Low Fluid Oil Pressure N.A.
R N.A.
S/U(1,10)
N.A.
1
[
b.
Turbine Stop Valve' N.A.
R:
N.A.
S/U(1,10)'
.N.A.
1 4
Closure o
17.
Safety Injection Input from N.A.
LN.A.
N.A.
R' N.A._
1, 2
'ESF 18.
Reactor Trip' System Interlocks a.
Intermediate' Range Neutron Flux, P-6 N.A.
R(4)
R N.A.
N.A.
2#..
b.
Power Range Neutron Flux, P-8 N.A.
R(4)
R N.A.
N.A.
1 c.
Power Range Neutron Flux, P-9 N. A.-
R(4)
R N.A.
N.A.
1.
4 CL 2a i
g 1
m
TABLE 4.3-1 (Continued)
TABLE NOTATIONS (13) D LETED.
(14) DELETED.
(15) The MODES specified for these channels in Table 4.3-2 are more restric-tive and, therefore, applicable.
(16) The TRIP ACTUATING DEVICE OPERATIONAL TEST shall independently verify the OPERABILITY of the undervoltage and shunt trip attachments of the Reactor Trip Breakers.
(17) local manual shunt trip prior to placing breaker in service.
(18) Automatic undervoltage trip.
WOLF CREEK - UNIT 1 3/4 3-12a Amendment No. 12, 26, f3, 52
L LIMITING SAFETY SYSTEM SETTINGS BASES Intermediate and Source Range, Neutron Flux The Intermediate and Source Range, Neutron Flux trips provide core protection during reactor startup to mitigate the consequences of an un-controlled rod clustar control assembly bank withdrawal from a subcritical condition.
These trips provide redundant protection to the Low Setpoint trip of the Power Range, Neutron Flux channels.
The Source Range channels will initiate a Reactor trip at about 105 counts per second unless manually blocked when P-6 becomes active.
The Intermediate Range channels will initiate a Reactor trip at a current level equivalent to approximately 25% of RATED THERMAL POWER unless manually blocked when P-10 becomes activc.
Overtemperature 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 transit delays from the core to the temperature detectors and pressure is l
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 includes dynamic compensation for piping delays from the core to the loop temperature detectors, (2) pressurizer pressure, and (3) axial power 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 top and bottom power range nuclear detectors, the Reactor trip is automatically reduced according to the notations in Table 2.2-1.
Overpower AT The Overpower AT Reactor trip provides assurance of fuel integrity (e.g., no fuel pellet melting and less than 1% cladding strain)_under all possible overpower conditions, limits the required range for Overtemperature t.T trip, and provides a backup to-the High Neutron Flux trip.
The Setpoint is automatically varied with: (1) coolant temperature to correct for temperature induced changes in density and heat capacity of water, and (2) rate of change of temperature for dynamic compensation for piping delays from the core to the loop temperature detectors, to ensure that the allowable heat generation rate (kW/ft) is not exceeded.
The Overpower AT trip provides protection to mitigate the consequences of various size steam breaks as reported in WCAP-9226, " Reactor Core Response to Excessive Secondary Steam-Releases."
f WOLF CREEK - UNI) 1 B 2-5 Amendment No. 52
.=
4'i t
LIMITING SAFETY SYSTEM SETTINGS B_ASES i
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 l
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 T"ERMAL POWER with turbine impulse chamber pressure at approximately 10% of full power equivalent); and i
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 1
overpressure.
1
{
Pressurizer Water Level The Pressurizer High Water level trip is provided to prevent water relief through the pressurizer safety valves.
On decreasing power the Pressur*zer 4
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 pres-8 sure at approximately 10% of full power equivalent); and on increasing power,
. automatically reinstated by P-7.
Reactor Coolant Flow I
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 more than one loop drops below 90% of nominal full loop flow.
Above P-8 (a power level of approximately 48% of RATED THERMAL POWER) an automatic Reactor trip will occur if the flow in any single loop drops below 90% of nominal full loop flow.
Conversely on decreasing power between P-8 and the 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.
WOLF CREEK - UNIT 1 B 2-6
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