ML20063M037
| ML20063M037 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 03/03/1994 |
| From: | Black S Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20063M039 | List: |
| References | |
| NUDOCS 9403080358 | |
| Download: ML20063M037 (12) | |
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li i ) "I UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555 4001 gs
,e WOLF CREEK NUCLEAR OPERATING CORPORATION WOLF CREEK GENERATING STATION DOCKET N0. 50-482 AMENDMENT TO FACILITY OPERATING LICENSE Amendment NoJ2 License No. NPF-42
- 1. The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment to the Wolf Creek Generating Station (the facility) Facility Operating License No. NPF-42 filed by the Wolf Creek Nuclear Operating Corporation (the Corporation), dated February 7, 1994, 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 amended, 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 j
in compliance with the Commission's regulations; 1
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.
9403000358 940303 i
PDR ADOCK 05000482 P
. 2.
Accordingly, the license is amended by changes to the 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 folicws:
2.
Technical Soecifications The Technical Specifications contained in Appendix A, as revised through Amendment No.72, 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 and shall be implemented within 60 days of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION k
Suzanne C. Black, Directo 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: March 3, 1994 a
l ATTACHMENT TO LICENSE AMENDMENT NO.72__
FACILITY OPERATING LICENSE NO. NPF-42 DOCKET N0. 50-482 Revise Appendix A Technical Specifications by removing the pages identified 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 2-8 2-8 2-10 2-10 3/4 2-16 3/4 2-16 B 3/4 2-3 8 3/4 2-3
g TABLE 2.2.-l REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS
?
SENSOR TOTAL ERROR FUNCTIONAL UNIT ALLOWANCE (TA) 1 (S)
TRIP SETPOINT ALLOWABLE VALUE E
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
s109% of RTP*
sil2.3% 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 0
s4% of RTP* with s6.3% of RTP* with High Positive Rate a time constant a time constant 22 seconds 22 seconds 4.
Power Range, Neutron Flux, 2.4 0.5 0
54% of RTP* with
$6.3% of RTP* with High Negative Rate a time constant a time constant 22 seconds 22 seconds m
a 5.
Intermediate Range, 17.0 8.41 0
<25% of RTP*
s35.3% of RTP*
Neutron Flux 5
5 6.
Source Range, Neutron Flux 17.0 10.01 0 s10 cps s1.6 x 10 cps
@S 7.
Overtemperature AT 7.0 4.86 1.67 See Note 1 See Note 2 l
8.
Overpower AT 4.6 2.02 0.14 See Note 3 See Note 4 l
E 9.
Pressurizer Pressure-Low 3.7 0.71 2.49 21915 psig 21906 psig 10.
Pressurizer Pressure-High 7.5 0.71 2.49
$2385 psig s2400 psig E!
11.
Pressurizer Water Level-High 8.0 2.18 1.96
$92% of instrument
$93.9% of instrument
- m span span E
h m
- -[
o RTP - RATED THERMAL POWER
- Loop design flow - 93,600 gpm N
e-
g G
TABLE 2.2-1 (Continued) n TABLF NOTATIONS h
NOTE 1:
DYERTEMPERATURE AT h
(l f Is5) $ AT, [Kt - Ka AT
[T(if 3
- 5) - T'] + K (P - P') - f (aI))
t l
M 3
g Where:
AT
=
Measured AT; f
lead-lag compensator en measured AT;
=
Time constants utilized in lead-lag compensator for AT, In = 6 s, t, in i
=
l ta = 3 s; f-f,,3 Lag compensator on measured AT;
=
.la Time constant utfifred in the lag compensator for AT, is = 2 s; is
=
AT, Indicated AT at RATED THERMAL POWER:
=
K 1.10;
=
Km 0.0137/'F;
=
I g
dynamic compensation;The function generated by the lead-lag compensa
=
Time constants utfilzed in the lead-lag cespensator for T,y, 1 T4, is a
a 4 = 16 s, is = 4 s; E
T
= Average tesperature, 'F; y
I L89 compensator on measured T,,,;
=
5 S
Time constant uttilzed in the measured T,,, lag compensator, ta = G s; ta
=
5 TABLE 2.2-1-(Continued)
G TABLE NOTATIONS (Continuedl n
Ag NOTE 1:
(Continued)
T' 5; 586.5*F (Nominal T AT RATED THERMAL POWER);
g m
=
]
K3 0.000671;
=
P Pressurizer pressure, psig
=
P' 2235 psig (Nominal RCS operating pressure);
=
S laplace transform operator, s";
=
and f (AI) is a function of the indicated difference between top and bottom detectors of the power-i range neutron ion chambers; with gains to be selected based on measured instrument response during plant STARTUP tests such that:
~
a, (1) for q, - q, between -25% and + 5%, f (AI) = 0, where q and q are percent RATED THERMAL POWER'in l
.j i
the top and bottom halves of the core respectively, an,d q, +3q,is total THERMAL POWER in percent of RATED THERMAL POWER; (ii) for each percent that the magnitude of q - q, exceeds -25%, the AT Trip Setpoint shall be automatically reduced by 1.8% of its valu,e at RATED THERMAL POWER; and y
(iii) for each percent that the magnitude of q
-q exceeds +5%, the AT Trip Setpoint shall be g
automatically reduced by 1.56% of its 'value a,t RATED THERMAL POWER.
Y:
'+
NOTE 2: The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than.1.8%
g of AT span.
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gg TABLE 2.2-1 (Continued)
G 9
TABLE NOTATIONS (Continued) h)
NOTE 3:
(Continued)
K, 0.00128/*F for T > T" and K6 = 0 for T < T";
=
T Average temperature, *F; t
T" Indicated T a
instrumentalion,t RATED THERMAL POWER (Calibration temperature for LJ
=
s 586.5 F);
l Laplace transform operator, s"; and S
f (AI)
O for all AI.
2 NOTE 4:
The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 2.6% of AT span.
l O
F an CJ
.N
~
03 h
_-,c,
POWER DISTRIBUTION LIMITS 314.2.5 DNB PARAMETERS LIMITING CONDITION FOR OPERATION ACTION:
(Continued) 4.
Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced THERMAL POWER limit required by ACTION 1.b and/or 3, above; subsequent POWER OPERATION may proceed provided that the indicated RCS total flow rate is demonstrated to be within the region of acceptable operation prior to exceeding the following THERMAL POWER levels:
A nominal 50% of RATED THERMAL POWER, a.
b.
A nominal 75% of RATED THERMAL POWER, and Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of attaining greater than or equal to 95%
c.
of RATED THERMAL POWER.
SURVEILLANCE REOUIREMENTS 4.2.5.1 The provisions of Specification 4.0.4 are not applicable to Specification 3.2.5.c.
4.2.5.2 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 />.
4.2.5.3 The RCS total flow rate indicators shall be subjected to a CHANNEL CALIBRATION at least once per 18 months.
4.2.5.4 The RCS total flow rate shall be determined by precision heat balance measurement at least once per 18 months.
Within 7 days prior to perfonning the precision heat balance, the instrumentation used for determination of steam pressure, feedwater pressure, feedwater temperature, and feedwater venturi AP in the calorimetric calculations shall be calibrated.
4.2.5.5 The feedwater venturi shall be inspected for fouling and cleaned as necessary at least once per 18 months.
WOLF CREEK - UNIT 1 3/4 2-15 Amendment No. 61
TABLE 3.2-1 DNB PARAMETERS LIMITS Four Loops in PARAMETER Operation 1.
Indicated Reactor Coolant System T,,
s590.5'F 2.
Indicated Pressurizer Pressure 22220 psig*
3.
Reactor Coolant System Flow Rate 238.4 x 10' GPM 1
T i
S
- Limit not applicable during either a THERMAL POWER ramp in excess of 5% of RATED THERMAL POWER per minute or a THERMAL POWER step in excess of 10% of RATED THERMAL POWER.
WOLF CREEK - UNIT 1 3/4 2-16 Amendment No. 61,69,72
POWER DISTRIBUTION LIMITS BASES 00ADRANT POWER TILT RATIO (Continued)
The 2-hour time allowance for operation with a tilt condition greater that. 1.02 but less than 1.09 is provided to allow identification and correction of a dropped or misaligned control 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 maximum alloweo power by 3% for each percent of tilt in excess of 1.
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 flux map or two sets of four symmetric thimbles.
The two sets of four symmetric thimbles is a unique set of eight detector locations. These locations are C-8, E-5, E-11, H-3, H-13, L-5, L-11, N-8.
3/4.2.5 DNB PARAMETERS The limits on the Reactor Coolant System T, and the pressurizer pressure assure that each of the parameters are m,aintained within the normal steady-state envelope of operation assumed in the transient and accident analyses.
The limits are consistent with the initial USAR assumptions and have been analytically demonstrated adequate to maintain a DNBR above the safety analysis limit DNBR specified in the CORE OPERATING LIMITS REPORT (COLR) throughout each analyzed transient. The indicated T value of 590.5'F and the indicated pressurizer pressure value of 222E,psig correspond to analytical limits of 593.0*F and 2205 psig respectively, with allowance for measurement uncertainty.
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.
Fuel rod bowing reduces the value of DNB ratio.
Credit is available to offset this reduction in the generic margin.
The generic margins completely offset any rod bow penalties. This is the margin between the correlation DNBR limit and the safety analysis limit DNBR.
These limits are specified in the COLR.
The applicable values of rod bow penalties are referenced in the USAR.
When RCS flow rate and Fu(X,Y), per Specification 3.2.3, are measured, no additional allowances are necessary prior to comparison with the limits in the COLR.
Measurement uncertainties of 2.5% for RCS total flow rate and 4%
for Fa(X,Y) have been allowed for in determination of the design DNBR value.
WOLF CREEK - UNIT 1 B 3/4 2-3 Amendment No. H,M,69,72
POWER DfSTRIBUTVON LIMTH BASES DNB PARAMETERS (Continued)-
The measurement uncertainty for RCS total flow rate is based upon performing a precision heat balance and using the result to calibrate the RCS flow rate indicators.
not be detected could bias the result from the precision heat balance in nonconservative manner.
Therefore, an inspection is performed of the feedwater venturi each refueling outage.
The 12-hour periodic surveillance of indicated RCS flow is sufficient to detect only flow degradation which could lead to operation outside the acceptable region of operation specified in Table 3.2-1.
This surveillance also provides adequate monitoring to detect any core crud buildup.
WOLF CREEK - UNIT 1 B 3/4 2-4 Amendment No. 61
,