ML20059M061
| ML20059M061 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 11/10/1993 |
| From: | Murley T Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20059M062 | List: |
| References | |
| NUDOCS 9311180136 | |
| Download: ML20059M061 (15) | |
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., e UNITED STATES
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j NUCLEAR REGULATORY COMMISSION g
f WASHINGTON, D.C. 20555-0001 g
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us WOLF CREEK NUCLEAR OPERATING CORPORATION WOLF CREEK GENERATING STATION DOCKET NO. 50-482 AMENDMENT TO FACILITY OPERATING LICENSE i
Amendment No. 69 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 January 5,1993, as supplemented by letter dated October 1,1993, 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 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 j
the Commission's regulations and all applicable requirements have been satisfied.
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s 9311180136 931110 PDR ADOCK 05000482 P
PDR t
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2.
Accordingly, the license is amended by changes to the Technical Specifi-cations as indicated in the attachment to this license amendment and 1
Paragraphs 2.C.(1) and 2.C.(2) of Facility Operating License No. NPF-42 are hereby amended to read as follows:*
(1) Maximum Power level The Operating Corporation is authorized to operate the facility at reactor core power levels not in excess of 3565 megawatts thermal (100% power) in accordance with the conditions specified herein and 1
in Attachment I to this license. The activities identified in Attachment I to this license shall be completed as specified. is hereby incoroorated into this license.
(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No.
69, 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 90 days of issuance.
FOR THE NUCLEAR REGULATORY COHMISSION b
Thomas E. Murley, Direc or Office of Nuclear Reactor Regulation
Attachment:
1.
Page 3 of License 2.
Changes to the Technical Specifications Date of Issuance:
November 10, 1993
- Page 3 is attached, for convenience, for the composite license to reflect this change.
Please remove page 3 of the existing license and replace with the attached oage.
B (1)
Pursuant to Section 103 of the Act and 10 CFR Part 50 " Domestic Licensing of Production and Utilization Facilities," the Operating Corporation, to pm;sess, use and operate the facility at the designated locateo in Coffey County, Kansas, in accordance with the procedures and limitations set forth in this license; (2) KG&E, KCPL and KEPC0 to possess the facility at the designated location in Coffey County, Kansas, in accordance with the procedures and limitations set forth in this license; (3) The Operating Corporation, pursuant to the Act and 10 CFR Part 70, l
to receive, possess and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Final Safety Analysis Report, as supplemented and amended.
(4) The Operating Corporation, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (5) The Operating Corporation, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required any byproduct, source or special nuclear material withoct restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (6) The Operating Corporation, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.
C.
This license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and-is subject to the additional conditions specified or incorporated below:
(1) Maximum power Level l
The Operating Corporation is authorized to operate the facility at reactor core power levels not in excess of 3565 megawatts thermal l
(100% power) in accordance with the conditions specified herein and in Attachment I to this license. The activities identified in Attachment I to this license shall be completed as specified. is hereby incorporated into this license.
Amendment No. 63,69
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ATTACHMENT TO LICENSE AMENDMENT NO. 69
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l FACILITY OPERATING LICENSE NO. NPF-42 t
DOCKET NO. 50-482 l
Revisa 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.
i The corresponding overleaf pages are also provided to maintain document completeness.
i REMOVE INSERT I
1-5 1-5 j
2-4 2-4 2-8 2-8 2-10 2-10 3/4 2-16 3/4 2-16 i
B 3/4 2-3 B 3/4 2-3 h
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2 l
I 1
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DEFINITIONS PRESSURE BOUNDARY LFAKAGE 1.22 PRES';URE BOUNDARY LEAKAGE shall be leakage (except steam generator tube leakage) through a nonisolable fault in a Reactor Coolant System component body, pipe wall, or vessel wall.
PROCESS CONTROL PROGRAM 1.23 The PROCESS CONTROL PROGRAM (PCP) shall contain the current formulas, sampling, analyses, tests, and determinations to be made to ensure that the processing and packaging of solid radioactive wastes based on demonstrated processing of' actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Parts 20, 61, and 71, State.
regulations, burial ground requirements, and other requirements governing the disposal of solid radioactive waste.
PURGE - PURGING 1.24 PURGE or PURGING shall be any controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration, or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.
00ADRANT POWER TILT RATIO j
1.25 QUADRANT POWER TILT RATIO shall be the ratio of the maximum upper excore detector calibrated output to the average of the upper excore detector t
calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater. With one excore detector inoperable, the remaining three detectors shall be used for computing the average.
RATED THERMAL POWER i
1.26 RATED THERMAL POWER shall be a total core heat transfer rate to the reactor coolant of 3565 MWt.
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REACTOR TRIP SYSTEM RESPONSE TIME 1.27 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its Trip Setpoint at the channel sensor i
until loss of stationary gripper coil voltage.
t REPORTABLE EVENT 4
1.28 A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 to 10 CFR Part 50.
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q WOLF CREEK - UNIT 1 1-5 Amendment No. 42,61, 69 1
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DEFINITIONS i
l SHUTDOWN MARGIN 1.29 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is subtritical or would be subcritical from its present condition assuming all full-length rod cluster assemblies (shutdown and control) are fully inserted except for the single rod cluster assembly of highest reactivity worth which is assumed to be fully withdrawn.
SITE BOUNDARY l.30 The SITE BOUNDARY shall be that line beyond which the land is neither l
owned, nor leased, nor otherwise controlled by the licensee.
l SLAVE RELAY TEST 1.31 A SLAVE RELAY TEST shall be the energization of each slave relay and verification of OPERABILITY of each relay. The SLAVE RELAY TEST shall include a continuity check, as a minimum, of associated testable actuation devices.
l SOURCE CHECK 1.32 A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a source of increased radioactivity.
STAGGERED TEST BASIS 1.33 A STAGGERED TEST BASIS shall consist of:
A test schedule for n systems, subsystems, train.5, or other a.
designated components obtained by dividing the specified test j
interval into n equal subintervals, and j
b.
The testing of one system, subsystem, train, or other designated component at the beginning of each subinterval.
THERMAL POWER l
1.34 THERMAL POWER shall be the total core heat transfer rate to the reactor coolant.
TRIP ACTUATING DEVICE OPERATIONAL TEST I
1.35 A TRIP ACTUATlHG DEVICE OPERATIONAL TEST shall consist of operating the l
Trip Actuating Device and verifying OPERABILITY of alarm, interlock and/or trip functions.
The TRIP ACTUATING DEVICE OPERATIONAL TEST shall include adjustment, as necessary, of the Trip Actuating Device such that it actuates at the required Setpoint within the required accuracy.
l WOLF CREEK - UNIT 1 1-6 Amendment No. E,61 s
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g TABLE 2.2-1 r-
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REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS h
SENSOR TOTAL ERROR FUNCTIONAL UNIT ALLOWANCE (TA)
Z (S)
TRIP SETPOINT ALLOWABLE VALUE 1.
Manual Reactor Trip H.A.
N.A.
N.A.
N.A.
N.A.
w N
2.
Power Range, Neutron Flux a.
High Setpoint 7.5 4.56 0
$109% of RTP*
1112.3% of RTP*
b.
Low Setpoint 8.3 4.56 0
125% 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 High Positive Rate i time constant i time constant 12 seconds 12 seconds 4.
Power Range, Neutron Flux, 2.4 0.5 0
<4% of RTP* with
<6.3% of RTP* with High Negative Rate i time constant i time constant y
4 12 seconds 12 seconds 5.
Intermediate Range, 17.0 8.41 0
125% of RTP*
135.3% of RTP*
Neutron Flux 6.
Source Range, Neutron Flux 17.0 10.01 0
110 cps
$1.6 x 10 cps 8
5 7.
Overtemperature AT 7.9 4.61 2.57 See Note 1 See Note 2 f
8.
Overpower AT 5.0 2.15 0.15 See Note 3 See Note 4
[
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 12385 psig 12400 psig l
11.
Pressurizer Water Level-High 8.0 2.18 1.96 192% of instrument 193.9% of instrument i,e -
span span E e *RTP = RATED THERMAL POWER
- Loop design flow = 93,600 gpm 5
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g TABLE 2.2-1 (Continued)
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TABLF NOTATIONS NOTE 1:
OVERTEMPERATURE AT AT ff *jf (1 f r3 ) $ AT, {K 2 h i f) [T (T 5
t-K ta5) - T'] + K (P P') - f (AI)]
3 y
- -e Where:
AT
=
Measured AT; 1
I lead-lag compensator on measured AT;
=
Time constants utilized in lead-lag compensator for AT, r: = 5 s, l
t, 12
=
i 12 = 3 s; g f
,3 Lag compensator on measured AT;
=
"4 13
=
Time constant utilized in the lag compensator for AT, is = 2 s; AT, Indicated AT at RATED THERMAL POWER;
=
Kg 1.10;
=
Kg 0.0137/*F;
=
g f
l
= The function generated by the lead-lag compensator for T,yg dynamic compensation; g
I Time constants utilized in the lead-lag compensator for T,,9, 14 = 16 s, T4, is
=
a is = 4 s; P
T
=
Average temperature, 'F; 1
,M Lag compensator on measured T,yg;
=
5 9
Time constant utilized in the measured T,,, lag compensator, is = 0 s; ts
=
_ TABLE 2.2-1 (Continued)
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TABLE NOTATIPMS (Continued) n NOTE 1:
(Continued) n i
T' S 581.2*F (Nominal T,yg at RATED THERMAL POWER);
C Ka 0.000671;
=
P
=
Pressurizer pressure, psig; P'
2235 psig (Nominal RCS operating pressure);
=
S
=
Laplace transform operator, s 1; and fg(AI) is a function of the indicated difference between top and bottom detectors of the power-range neutron ion chambers; with gains to be selected based on measured instrument response during plant STARTUP tests such that:
y (1) for qt g between -25% and + 7%, f (AI) = 0, where qt and qb are percent RATED THERMAL POWER in the top and bottom halves of the core respectively, and qt + % IS total THERMAL POWER in percent of RATED THERMAL POWER; (ii) for each percent that the magnitude of qt ~ % exceeds -25% the AT Trip Setpoint shall be automatically reduced by 1.8% of its value at RATED THERMAL POWER; and (iii) for each percent that the magnitude of qt " % exceeds +7%, the AT Trip Setpoint g
shall be automatically reduced by 1.384% of its value at RATED THERMAL POWER.
ea NOTE 2:
The channel's maximus Trip Setpoint shall not exceed its computed Trip Setpoint by more than 2.5% of AT span.
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g TABLE 2.2-1 (Continued) 9 TABLE NOTATIONS (Continued) oy NOTE 3: OVERPOWER AT sf fl f 5 $ AT, (K4 - Ks (7 T - Ks [T If5 - D ~ I (AI)I t25 AT ts 15 1
185 l
18 7
2 A
Where:
AT
=. Measured AT; f jif
= lead-lag compen.ator on measured AT; Time constants utilized in lead-lag compensator for AT, is = 6 s, is = 3 s; tt, is
=
Lag compensator on measured AT;
=
1 ts5 Time constant utilized in the lag compensator for AT, is = 2 s; ts
=
AT, Indicated AT at RATED THERMAL POWER;
=
K4 1.10;
=
0.02/'F for increasing average temperature and 0 for decreasing average Ks
=
temperature; If 5 The function generated by the rate-lag compensator for T,,, dynamic
=
compensation; f
Time constant utilized in the rate-lag compensator for T,,,, t, = 10 s;
=
t, 1 [re lag compensator on measured T,,g;-
s ts
= Time constant utilized in the measured T,,, lag compensator, is = 0 s; 3
5 TABLE 2.2-1 (Continued)
G TABLE NOTATIONS (Continued) n
[r" NOTE 3:
(Continued) n Ks 0.00128/*F for T > T" and Ks = 0 for T i T";
=
Eq T
Average temperature, 'F;
=
T"
=
Indicated T,yg at RATED THERML POWER (Calibration temperature for AT instrumentation, 5 581.2 'F);
l S
=
Laplace transform operator, s 1; and f (al) 0 for all AI.
t
=
NOTE 4:
The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than rf 2.8% of AT span.
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ta
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.D.
POWER DISTRIBUTION LIMITS 3 /4. 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.
A nominal 50% of RATED THERMAL POWER, i
b.
A nominal 75% of RATED THERMAL POWER, and c.
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 eyual to 95%
of RATED THERMAL POWER.
SURVEILLANCE RE0VIREMENTS 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 CAllBRATION 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 performing 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
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TABL E 3.2-1 bNB PARAMETERS f
LIMITS f
Four Loops in i
PARAMETER Operation 1.
Indicated Reactor Coolant System T,y
<585.0*F l
f 2.
Indicated Pressurizer Pressure 22220 psig*
l 3.
Reactor Coolant System Flcw Rate 238.4 x 10' GPM i
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- Limit not applicable during either' a THERHAL POWER ramp in excess of 5% of RATED THERMAL POWER per minute or a THERMAL POWER step in excess of 10% of I
RATED THERMAL POWER.
l WOLF CREEK - UNIT 1 3/4 2-16 Amendment No. 61,69 -
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POWER DISTRIBUTION LIMITS BASES i
OVADRANT POWER TILT RATIO (Continued) l 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 i
correction of a dropped or misaligned control rod.
In the event such ACTION l
does not correct the tilt, the margin for uncertainty on F (X,Y,Z) is o
reinstated by reducing the maximum allowed power by 3% for each percent of tilt in excess of 1.
- For purposes of monitoring QUADRANT POWER TILT RATIO when one excore I
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 1,y and the pressurizer i
pressure assure that each of the parameters are m,aintained within the normal l
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
-1 safety analysis limit DNBR specified in the CORE OPERATING LIMITS REPORT (COLR) throughout each analyzed transient. The indicated T value of 585'F l
and the indicated pressurizer pressure value of 2220 psig c,,r, respond to o
t analytical limits of 587.7*F and 2205 psig respectively, with allowance for l
[
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.
J Fuel rod bowing reduces the value of DNS 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 i
limit and the safety analysis limit DNBR. These limits are specified in the COLR.
i The applicable values of rod bow penalties are referenced in the USAR.
When RCS flow rate and F3 (X,Y), per Specification 3.2.3, are measured, no additional allowances are n,ecessary prior to comparison with the limits in I
the COLR. Measurement uncertainties of 2.5% for RCS total flow rate and 4%
for F,,(X,Y) have been allowed for in determination of the design DNBR value.
i WOLF CREEK - UNIT 1 B 3/4 2-3 Amendment No. H,M,69 i
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POWER DISTRIBUTION LIMITS BASES ONB 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.
Potential fouling of the feedwater venturi which might not be detected could bias the result from the precision heat balance in a 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 monitcring to detect any core crud buildup.
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1 WOLF CREEK - UNIT I B 3/4 2-4 Amendment No. 61 i