ML20059F488
| ML20059F488 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 10/27/1993 |
| From: | Gody A Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20059F491 | List: |
| References | |
| NUDOCS 9311040327 | |
| Download: ML20059F488 (14) | |
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WASHINGTON, D.C. 20555-0001
.....f WISCONSIN ELECTRIC POWER COMPANY DOCKET NO. 50-266 POINT BEACH NUCLEAR PLANT. UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.
142 License No. DPR-24 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The applications for amendment by Wisconsin Electric Power Company (the licensee) dated June 11, 1993, as supplemented by letter dated October 19, 1993, 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 conformity with the applications, 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 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 Comission's regulations and all applicable requirements have been satisfied.
9311040327 931027 PDR ADOCK 05000266 P
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.8 of Facility Operating License No.
DPR-24 is hereby amended to read as follows:
B.
Technical Soecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.
142, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective imediately upon issuance. The Technical Specifications are to be implemented within 20 days from the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION ffb
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Anthony T. Gody, Jr., Project Manager Project Directorate III-3 Division of Reactor Projects III/IV/V Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of issuance: October 27, 1993
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UNITED STATES NUCLEAR REGULATORY COMMISSION g..v,/
WASHINGTON, D.C. 20$n-0001 WISCONSIN ELECTRIC POWER COMPANY DOCKET NO. 50-301 POINT BEACH NUCLEAR PLANT. UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 146 License No. DPR-27 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The applications for amendment by Wisconsin Electric Power Company (the licensee) dated June 11, 1993, as supplemented by letter dated October 19, 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 applications, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (1) 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.
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Accordingly, the license is amended by changes to the Technical Specifications as. indicated in the attachment to this license amendment, and paragraph 3.B of Facility Operating License No.
DPR-27 is hereby amended to read as follows:
B.
Technical SoecificatioD1 The Technical Specifications contained in Appendices A and 8, as revised through Amendment No. 146, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications..
3.
This license amendment is effective immediately upon issuance. The Technical Specifications are to be implemented within 20 days from the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION-d b
Anthony T. Gody, Jr., Project Manager Project Directorate III-3 4
Division of Reactor. Projects III/IV/V Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of issuance: October 27, 1993
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r ATTACHMENT TO LICENSE AMENDMENT NOS.142 AND l46 TO FACILITY OPERATING LICENSE NOS. DPR-24 AND DPR-27 DOCKET NOS. 50-266 AND 50-301 Revise Appendix A Technical Specifications by removing the pages identified-below and inserting the enclosed pages. The revised pages are identified b amendment number and contain marginal lines indicating the area of change. y REMOVE INSERT
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15.2.1-1 15.2.1-1 15.2.1-2 15.2.1-2 Figure 15.2.1-1 Figure 15.2.1-1 Figure 15.2.1-2 15.2.3-2 15.2.3-2 15.2.3-3 15.2.3-3 15.2.3-6 15.2.3-6 15.2.3-7 15.2.3-7 15.3.1-19 15.3.1-19 r
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t 15.2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 15.2.1 SAFETY LIMIT, REACTOR CORE AeolicabilitY:
Applies to the limiting combinations of thermal power, reactor coolant system pressure, and coolant temperature during operation.
Obiective:
To maintain the integrity of the fuel cladding.
Specification:
1.
The combination of thermal power level, coolant pressure, and coolant temperature shall not exceed the limits shown in Figure 15.2.1-1 for Unit I and Figure 15.2.1-2 for Unit 2.
The safety limit is exceeded if the point defined by the combination of reactor coolant system average temperature and power level is at any time above the appropriate pressure line.
Basis:
The restrictions of this safety limit prevent overheating of the fuel and pos-sible cladding perforation which would result in the release of fission products to the reactor coolant. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.
Operation above the upper boundary of the nucleate boiling regime could result in excess cladding temperature because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient.
DNB is not a directly measurable parameter during operation and therefore thermal power and Reactor Coolant temperature and pressure have been related to DNB.
1 1
Unit 1 - Amendment No. #,22,86.720,142 Unit 2 - Amendment No. 27.29.90.123.146 15.2.1-1 l
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This relation has been developed to predict the DNB flux and the location of DNB for axially uniform and non-uniform heat flux distributions. The local DNB heat flux ratio, DNBR, defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat Cux, is indicative of the margin to DNB.
The DNB design basis is as follows: there must be at least a 95 percent proba-bility at a 95 percent confidence level that DNB will not occur during steady state operation, normal operational transients, and anticipated transients and is an appropriate margin to DNB for all operating conditions.
The curves of Figure 15.2.1-1 and 15.2.1-2 are applicable for a core of 14 x 14 l
OFA. The curves also apply to the reinsertion of previously-depleted 14 x 14 standard fuel assemblies into an 0FA core. The use of these assemblies is justified by a cycle-specific reload analysis. The WRB-1 correlation is used to generate these curves. Uncertainties in plant parameters and DNB correlation predictions are statistically convoluted to obtain a DNBR uncertainty factor.
This DNBR uncertainty factor establishes a value of design limit DNBR. This value of design limit DNBR is shown to be met in plant safety analyses, using values of input parameters considered at their nominal values.
1 Unit 's - Amendment No. ES,720,142 Unit 2 - Amendment No. 21,99,J22,146 15.2.1-2
Figure 15.2.1-1 REACTOR CORE SAFETY LIMITS
- POINT BEACH UNIT 1 l
660<
2400 P11A 642-
!!!O PSIA E!E<-
2000 P11A o672<
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- gjg, 1775 P11A 620<
598<
ggg.
2.
.1
.2
.5 4
.5
.5
.7
.8
.9 3.
1.8 1.2 Porte treaction er nominst Unit 1 - Amendment No. 85.J20.142
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- o-p Figure 15.2.1-2 REACTOR CORE SAFETY LIMITS POINT BEACH UNIT 2 5
I 660
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653 -
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2400 psio g,o. ;
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630 - -
c 620 - -
2000psio-g, 610 -
, N 1775 pela 600 - -
590 - -
$53.
0 0.1 0.2 0.3 CA 0.5 0.6 0.7 0.8 0.9 1
.1.1 1.2 Core Power # roc' Hon of nomina 0 j
Unit 2 - Amendment No.146 J
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Low pressurizer pressure -
21865 psig for operation at 2250 psia primary system pressure 21790 psig for operation at 2000 psia primary system pressure (4)
Overtemperature 1
AT (In S) 3 sat, (K -K (T( In S)-T )( 1+r 81"i +K (P-P')-f(AI))
S 1
i2 3
4 2
where indicated AT at rated power, "F AT o
average temperature, *F T
T' s
573.9 F (Unit 1)
T' s
570.0 F (Unit 2)
P pressurizer pressure, psig P'
2235 psig
=
K 5
1.30 i
K 0.0200
=
2 K
0.000791 3
25 sec r
i 3 sec 7
2 3
2 sec for Rosemont or equivalent RTD 7
O sec for Sostman or equivalent RTD 2 sec for Rosemont or equivalent RTD T
4 0 sec for Sostman or equivalent RTD
=
and f(AI) is an even function of the indicated difference between top and bottom detectors of the power-range nuclear ion chambers; with gains to be selected based on measured instrument response during plant startup tests, where q, and q, are the percent power in the top and bottom halves of the core respectively, and q, + q, is total core power in percent of rated power, such that:
(a) for q, - q, within -17, +5 percent, f(AI) = 0.
(b) for each percent that the magnitude of q, - q, exceeds <5 percent, the AT trip setpoint shall be automatically reduced by an equivalent of 2.0 percent of rated power.
Unit No. 1 - Amendment No. H BJ EE 99, 15.2.3-2 Unit So. 2 k N 6 h nt No. 19.99,97,723, 146
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(c) for each percent that the magnitude of q, - q, exceeds -17 percent, the AT trip setpoint'shall be automatically reduced by an equivalent of 2.0 percent of rated power.
I (5)
Overpower 47 ( 1+f 5 }
3 I
I sat,[K -K ( T S+1 ) ( 1+f,,5) T-K,[T( Iw S) - T'))
6 3 3
4 where indicated AT at rated power. 'F AT, average temperature, "F T
T' s
573.9 F (Unit 1)
T' s
570.0 F (Unit 2)
K s
1.089 of rated power 4
0.0262 for increasing T K
3 0.0 for decreasing T 0.00123 for T 2 T' K,
0.0 for T < T' 10 sec i
s 2 sec for Rosemont or equivalent RTD 7
3 0 sec for Sostman or equivalent RTD 2 sec for Rosemont or equivalent RTD 7
4 0 sec for Sostman or equivalent RTD (6)
Undervoltage - 275 percent of normal voltage (7)
Indicated reactor coolant flow per loop -
290 percent of normal indicated loop flow (8)
Reactor coolant pump motor breaker open (a)
Low frequency set point 255.0 HZ (b)
. Low voltage set point 275 percent of normal voltage.
Unit 1 - Amendment No. 3.28.85.90.94.
J20.J23.142 Unit 2 - Amendment No. 32.99.97.98.723.
F.5.2.3-3 J26.146
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l power distribution, the reactor trip limit, with allowance for errors (2), is always below the core safety limit as shown on Figure 15.2.1-1 for Unit I and Figure 15.2.1-2 for Unit 2.
If axial peaks are greater than design, as indicated by the difference between top and bottom power range nuclear detectors, the reactor trip limit is automatically reduceduur),
The overpower, overtemperature and pressurizer pressure system setpoints include the effect of reduced system pressure operation (including the effects of fuel densification). The setpoints will not exceed the core safety limits as shown in Figure 15.2.1-1 for Unit I and Figure 15.2.1-2 for Unit 2.
The overpower limit criteria is that core power be prevented from reaching a value at which fuel pellet centerline melting would occur. The reactor is prevented from reaching the overpower limit condition by action of the nuclect overpower and overpower AT trips.
I The high and low pressure reactor trips limit the pressure range in which reactor operation is permitted. The high pressurizer pressure reactor trip setting is lower than the set pressure for the safety valves (2485 psig) such that the reactor is tripped before the safety valves actuate. The low pressurizer pres-sure reactor trip trips the reactor in the unlikely event of a loss-of-coolant accident").
l The low flow reactor trip protects the core against DNB in the event of either a decreasing actual measured flow in the loops or a sudden loss of power to one or both reactor coolant pumps. The setpoint specified is consistent with the value used in the accident analysis'8)
The low loop flow signal is caused by a condi-tion of less than 90 percent flow as measured by the loop flow instrumentation.
The loss of power signal is caused by the reactor coolant pump breaker opening Unit 1 - Amendment No. 52,86,720,142 Unit 2 - Amendment No. 58,90,123.146
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as actuated by either high current, low supply voltage or low electrical fre-quency, or by a manual control switch. The significant feature of the breaker trip is the frequency setpoint, 55.0 HZ, which assures a trip signal before the pump inertia is reduced to an unacceptable value. The high pressurizer water level reactor trip protects the pressurizer safety valves against water relief.
The specified setpoint allows adequate operating instrument error (') and transient overshoot in level before the reactor trips.
The low-low steam generator water level reactor trip protects against loss of feedwater flow accidents. The specified setpoint assures that there will be sufficient water inventory in the steam generators at the time of trip to allow for starting delays for the auxiliary feedwater system.(')
Numerous reactor trips are blocked at low power where they are not required for protection and would otherwise interfere with normal plant operations. The prescribed setpoint above which these trips are unblocked assures their availability in the power range where needed. Specifications 15.2.3.2.A(1) and 15.2.3.2.C have 11% tolerance to allow for a 2% deadband of the P10 bistable which is used to set the limit of both items. The difference between the nominal and maximum allowed value (or minimum allowed value) is to account for "as measured" rack drift effects.
Sustained operation with only one pump will not be permitted above 3.5 percent power.
If a pump is lost while operating between 3.5 percent and 50 percent power, an orderly and immediate reduction in power level to below 3.5 percent is allowed. The power-to-flow ratio will be maintained equal to or less than unity, which ensures that the minimum DNB ratio increases at lower flow because the i
maximum enthalpy rise does not increase above the maximum enthalpy rise which occurs during full power and full flow operation.
References
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(7)
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FSAR 14.1.1 FSAR 14.3.1 FSAR 3.2.1 c2) FSAR, Page 14-5" (5) cs)
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FSAR 14.2.6 FSAR 7.2, 7.3 FSAR 14.1.11 15.2.3-7 Unit 1 - Amendment No. 86,9A.JD3,142 i
Unit 2 - Amendment No. 90,98,105,146
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OPERATIONAL LIMITATIONS The following DN8 related parameters shall be maintained within the limits shown during Rated Power operation:
1.
T,, shall be maintained below 578'F.
2.
Reactor Coolant System (RCS) pressurizer pressure shall be maintained:
22205 psig during operation at 2250 psia, or 21955 psig during operation at 2000 psia.
3.
Reactor Coolant System raw measured Total Flow Rate (See Basis).
a.
Unit 1 2 181,800 gpm Unit I b.
Unit 2 2 179,200 gpm Unit 2 Basis:
The reactor coolant system total flow rate for Unit 1 of 181,800 gpm is based on
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an assumed measurement uncertainty of 2.1 percent over thermal design flow (178,000 gpm). The reactor coolant system total flow rate for Unit 2 of 179,200 gpm is based on an assumed measurement uncertainty of 2.1 percent over thermal design flow (175,400 gpm). The raw measured flow is based upon the use of normalized elbow tap differential pressure which is calibrated against a precision flow calorimetric at the beginning of each cycle.
4 Unit 1 - Amendment No. {4 SJ,BE,J20, g
Unit 2 - Amendment No. AS,90,723,146 15.3.1-19
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