ML20203M605
| ML20203M605 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 04/22/1986 |
| From: | Youngblood B Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20203M608 | List: |
| References | |
| TAC-60559, NUDOCS 8605010605 | |
| Download: ML20203M605 (15) | |
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4 UNITED STATES P
1 NUCLEAR REGULATORY COMMISSION
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KANSAS GAS & ELECTRIC COMPANY KANSAS CITY POWER AND LIGHT COMPANY KANSAS ELECTRIC POWER COOPERATIVE, INC.
WOLF CREEK GENERATING STATION DOCKET N0. 50-482 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.1 License No. NPF-42 1.
The Nuclear Regulatory Comission (+'.' Comission) has found that:
A.
The application for amendment to the Wolf Creek Generating Station (the facility) Facility Operating License No. NPF-42 filed by Kansas Gas and Electric Company acting for itself and Kansas City Power and Light Company and Kansas Electric Power Cooperative, Inc., (the licensee) dated January 20, 1986, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),
and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in confomity with the application, the provisions of the Act, and the rules and regulations of the Comission; 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 Comission's regulations; D.
The issuance of this amendment will not be inimical to the comon 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 Comission's regulations and all applicable requirements have been satisfied.
8605010605 860422 PDR ADOCK 05000482 P
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2.
Accordingly, the license is amended by changes to the Technical Specifications 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:
B.
Technical Specifications The Technical Specifications contained in Appendix A, as rev! sed through Amendment No.1, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into the license.
KG&E shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION 0.
M B. J. Youngblood, Directo.-
PWR Project Directorate #4 Division of PWR Licensing-A
Attachment:
Changes to the Technical Specifications Date of Issuance:
April 22, 1986
. 2.
Accordingly, the license is amended by changes to the Technical Specifications 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:
B.
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 1, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into the license.
KG&E shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
}b B. J. Youngblood, Director PWR Project Directorate #4 Division of PWR Licensing-A
Attachment:
Changes to the Technical Specifications Date of Issuance:
April 22, 1986 l
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.o ISSUANCE OF AMENDMENT N0. 1 TO WOLF CREEK GENERATING STATION Dated April 22. 1986 ESTRIBUTION
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ATTACHMENT TO LICENSE AMENDMENT NO. 1 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 the captioned amendment number and contain marginal lines indicating the area of change. The corresponding overleaf pages are also provided to maintain document completeness.
AMENDED PAGE OVERLEAF PAGE 3/4 2-1 3/4 2-2 3/4 2-3 3/4 2-4 2-8 2-7 B 3/4 2-1 B 3/4 2-2 B 3/4 2-3 B 3/4 2-4
3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE (AFD)
LIMITING CONDITION FOR OPERATION 3.2.1 The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained within the allowed operational space defined by Figure 3.2-1.
APPLICABILITY: MODE 1 above 50 PERCENT RATED THERMAL POWER *.
ACTION:
a.
With the indicated AFD outside of the Figure 3.2-1 limits, 1.
Either restore the indicated AFD to within the Figure 3.2-1 limits within 15 minutes, or 2.
Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 30 minutes and reduce the Power Range Neutron Flux - High Trip Setpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
b.
THERMAL POWER shall not be increased above 50% of RATED THERMAL POWER i
unless the indicated AFD is within the Figure 3.2-1 limits.
i "See Special Test Exception 3.10.2.
WOLF CREEK - UNIT 1 3/4 2-1 Amendment No. I
POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENTS 4.2.1.1 The indicated AFD shall be determined to be within its limits during POWER OPERATION above 50% of RATED THERMAL F'0WER by:
l a.
Monitoring the indicated AFD for each OPERABLE excore channel:
1)
At least once per 7 days when the AFD Monitor Alarm is OPERABLE, and 2)
At least once per hour for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after restoring the AFD Monitor Alarm to OPERABLE status.
b.
Monitoring and logging the indicated AFD for each OPERABLE excore channel at least once per hour for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and at least once per 30 minutes thereafter, when the AFD Monitor Alarm is inoperable. The logged values of the indicated AFD shall be assumed to exist during the interval preceding each logging.
4.2.1.2 The indicated AFD shall be considered outside of its limits when at least two OPERABLE excore channels are indicating the AFD to be outside the limits.
3 WOLF CREEK - UNIT 1 3/4 2-2 Amendment No.1
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10 30 60 FLUX DIFFERENCE ( AI) %
FIGURE 3.2-1 WOLF CREEK UNIT 1 AX1AL FLUX DIFFERENCE LIMITS A8 A FUNCTION OF RATED THEhMAL POWE WOLF CREEK - UNIT 1 3/4 2-3 Amendment No. I t
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POWER DISTRIBUTION LIMITS 3/4.2.2 HEATFLUXHOTCHANNELFACTOR-Fg i
1 LIMITING CONDITION FOR OPERATION 3.2.2 F (Z) shall be limited by the following relationships:
q F (Z) 5 [2.32] [K(Z)] for P > 0.5, and 9
P F (Z) 5 [4.64] [K(Z)] for P 1 0.5.
9 1
Where:
i P _ THERMAL POWERRATED THERMAL POWER, and
~
K(Z) = the function obtained from Figure 3.2-2 for a given core height location.
i APPLICABILITY:
MODE 1.
ACTION:
With F (Z) exceeding its limit:
q
- ~
a.
Reduce THERMAL POWER at least 1% for each 1% F (Z) exceeds the limit j
q within 15 minutes and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; POWER OPERATION may proceed for up to a total of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; subsequent POWER OPERATION may proceed provided the Overpower AT Trip Setpoints have been reduced at least 1% for each 1% F (Z) exceeds the limit; and 9
b.
Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced limit required by ACTION a., above; THERMAL POWER may then be increased provided F (Z) is demonstrated through incore mapping to be within its limit.
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WOLF CREEK - UNIT 1 3/4 2-4
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TABLE 2.2-1 (Continued) o G
TABLE NOTATIONS k
NOTE 1:
OVERTEMPERATURE AT m
ATfl (1
Ta3) $ AT, {K I
[T (1 + 4 ) - T'] + K (P - P') - f (aI)}
-K 2 S
3 y
y 5
w Heasured AT by RTD Manifold Instrumentation; w
Where:
AT
=
I{
lead-lag compensator on measured AT;
=
ti, T2 Time constants utilized in lead-lag compensator for AT, t
=8s,
=
T2 = 3 s; yf 3
Lag compensator on measured AT;
=
4 13 Time constant utilized in the lag compensator for AT, 13 = 0 s;
=
AT, Indicated AT at RATED THERMAL POWER;
=
i 1.10; K
=
0.0137/*F; K
=
2 I{I'b The function generated by the lead-lag compensator for T
=
3 dynamic compensation; yg Time constants utilized in the lead-lag compensator for T,yg, 14 = 28 s,
=
T4, is is = 4 s; Average temperature, "F; T
=
I y
Ts3 Lag compensator on measured T,yg;
=
Time constant utilized in the measured T,yg lag compensator, is = 0 s;
=
ts
TABLE 2.2-1 (Continued)
Eg TABLE NOTATIONS (Continued)
NOTE 1:
(Continued) m 7
T' 5 588.5*F (Nominal T,yg at RATED THERMAL POWER);
E K
0.000671;
=
3
--e Pressurizer pressure, psig; s
P
=
2235 psig (Nominal RCS operating pressure);
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=^
Laplace transform operator, s 1; S
=
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:
7 (i) for qt - 9b between -27% and + 7%, f (AI) = 0, where qt and qb are percent I
2 RATED THERMAL POWER in the top and bottom halves of the core respectively, and qt+9b is total THERMAL POWER in percent of RATED THERMAL POWER; (ii) for each percent that the magnitude of qt 9b exceeds -27%, the AT Trip Setpoint shall be automatically reduced by 1.57% of its value at RATED THERMAL POWER; and (iii) for each percent that the magnitude of qt 9b exceeds +7%, the AT Trip Setpoint shall be automatically reduced by 1.05% of its value at RATED THERMAL POWER.
NOTE 2:
The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than
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3.3% of AT span.
a ita f.
3/4.2 POWER DISTRIBUTION LIMITS BASES The specifications of this section provide assurance of fuel integrity during Condition I (Normal Operation) and II (Incidents of Moderate Frequency) events by:
(a) maintaining the minimum DNBR in the core greater than or equal to 1.30 during normal operation and in short-term transients, and (b) limitina the fission gas release, fuel pellet temperature, and cladding mechanical properties to within assumed design criteria.
In addition, limiting the peal linear power density during Condition I events provides assurance that the initial conditions assumed for the LOCA analyses are met and the ECCS acceptance criteria limit of 2200*F is not exceeded.
The definitions of certain hot channel and peaking factors as used in these specifications are as follows:
F (Z)
Heat Flux Hot Channel Factor, is defined as the maximum local 0
heat flux on the surface of a fuel rod at core elevation Z divided by the average fuel rod heat flux, allowing for manufacturing tolerances on fuel pellets and rods; Fh Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio of the integral of linear power along the rod with the highest integrated power to the average rod power; and Fxy(Z)
Radial Peaking Factor, is defined as the ratio of peak power density to average power density in the horizontal plane at core elevation Z.
3/4.2.1 AXIAL FLUX DIFFERENCE (AFD) 1 The limits on AXIAL FLUX DIFFERENCE assure that the F (Z) upper bound 9
envelope of F limit times the normalized axial peaking factor is not exceeded q
during either normal operation or in the event of xenon redistribution following power changes.
Provisions for monitoring the AFD on an automatic basis are derived from the plant process computer through the AFD Monitor Alarm.
The computer deter-mines the 1-minute average of each of the OPERABLE excore detector outputs and provides an alarm message immediately if the AFD for at least 2 of 4 or 2 of 3 OPERABLE excore channels are outside the AFD limits and the THERMAL POWER is greater than 50% of RATED THERMAL POWER.
3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR, and RCS FLOWRATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR The limits on heat flux hot channel factor, RCS flowrate, and nuclear enthalpy rise hot channel factor ensure that: (1) the design limits on peak local power density and minimum DNBR are not exceeded, and (2) in the event of a WOLF CREEK - UNIT 1 B 3/4 2-1 Amendment No. 1 s
7
POWER DISTRIBUTION LIMITS BASES 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR, and RCS FLOWRATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Continued)
LOCA the peak fuel clad temperature will not exceed the 2200*F ECCS acceptance criteria limit.
Each of these is measurable but will normally only be determined periodically as specified in Specifications 4.2.2 and 4.2.3.
This periodic surveillance is sufficient to insure that the limits are maintained provided:
a.
Control rods in a single group move together with no individual rod insertion differing by more than + 12 steps, indicated, from the group demand position, b.
Control rod groups are sequenced with overlapping groups as described in Specification 3.1.3.6, i
1 WOLF CREEK - UNIT 1 B 3/4 2-2 Amendment No.
1 i
This page is deleted.
l UOLF CREEK - UNIT 1 B 3/4 2-3 Amendment No. 1 l
I POWER DISTRIBUTION LIMITS
' BASES HEAT FLUX HOT CHANNEL FACTOR, and RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Continued) c.
The control rod insertion limits of Specification 3.1.3.6 are maintained, and d.
The axial power distribution, expressed in terms of AXIAL FLUX DIFFERENCE, is maintained within the limits.
F will be maintained within its limits provided Conditions a. through H
- d. above are maintained.
As noted on Figure 3.2-3, RCS flow rate and F H
may be " traded off" against one another (i.e., a low measured RCS flow rate is acceptable if the measured F is also low) to ensure that the calcu g
lated DNBR will not be below the design DNBR value.
The relaxation of F as H
a function of THERMAL POWER allows changes in the radial power shape for all permissible rod insertion limits.
gascalculatedinSpecification3.2.3andusedinFigure3.2-3, accounts
- 's for F less than or equal to 1.49.
This value is used in the various accident 3g analyses where F influences parameters other than DNBR, e.g., peak clad tem H
perature, and thus is the maximum "as measured" value allowed.
Fuel rod bowing reduces the value of DNB ratio.
Credit is available to offset this reduction in the generic margin.
The generic margins, totaling 9.1% DNBR, completely offset any rod bow penalties. This margin includes the following:
1)
Design limit DNBR of 1.30 vs. 1.28, 2)
Grid spacing (K ) f 0.046 vs. 0.059, s
3)
Thermal Diffusion Coefficent of 0.038 vs. 0.059, 4)
DNBR Multiplier of 0.86 vs. 0.88, and 5)
Pitch Reduction.
The applicable values of rod bow penalties are referenced in the FSAR.
When an F measurement is taken, an allowance for both experimental error q
and manufacturing tolerance must be made.
An allowance of 5% is appropriate' for a full-core map taken with the Incore Detector Flux Mapping System, and a 3% allowance is appropriate for manufacturing tolerance.
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WOLF CREEK - UNIT 1 B 3/4 2-4 l
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