ML20138B966
| ML20138B966 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 11/27/1985 |
| From: | Stolz J Office of Nuclear Reactor Regulation |
| To: | Toledo Edison Co, Cleveland Electric Illuminating Co |
| Shared Package | |
| ML20138B968 | List: |
| References | |
| TAC 56794, NPF-03-A-091 NUDOCS 8512120418 | |
| Download: ML20138B966 (13) | |
Text
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'o, UNITED STATES j
NUCLEAR REGULATORY COMMISSION y
s-wAssiscTON, D, C. 20555
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TOLEDO EDISON COMPANY AND THE CLEVELAND ELECTRIC ILLUMINATING COMPANY DOCKET NO. 50-346 DAVTS-BESSE NL' CLEAR POWER STATION, UNIT NO.1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 91 License No. NPF-3 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by the Toledo Edison Company)and The Cleveland Electric illuminating Company (the licensees dated February 13, 1985, 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, the provisions of the Act, and the rules and regulations of the Comission; t
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 Commission's regulations and all applicable requirements have.
been satisfied.
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-3 is hereby amended to read as follows:
8512120418 851127 PDR ADOCK 05000346 P
. Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 91, are hereby incorporated in the license. The Toledo Edison Company shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION 4
s JonF.Stolz,Directje P
Project Directorate #6 0 vision of PWR Licensing-B
Attachment:
Changes to the Technical l
Specifications Date of issuance: November 27, 1985 t
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ATTACHMENT TO LICENSE AMENDMENT NO. 91 FACILITY OPERATING LICENSE NO. NPF-3 DOCKET NO. 50-346 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by Amendment number and contain a vertical line indicating the area of change. The corresponding overleaf pages are also provided to maintain document completeness, s
Remove Insert 2-3 2-3 2-7 2-7 3/4 2-14 3/4 2-14 B 2-1 B 2-1 B 2-8 B 2-8 e
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Figure 2.1-2 Reactor Core Safety Limit.
- RATED THERMAL PO'a'ER
- 120
(-48,112.0)
(44.112)
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- 100 49,100)
(-49,100.0)
( 48,89.1)
(44,89.1) p
- 80
( 49,77.1) ()
,)(4g,77,3)
ACCEPTABLE
- 60 OPERATION UNACCEPTABLE FOR SPECIFIED UNACCEPTABLE OPERATION RC PUMP OPERAT*CN COMBINATION 40 t
- 20 t
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f
-60
-40
-20 0
20 40 60 Axial Power Imbalance, %
PUMPS OPERATING REACTOR COOLANT FLOW, GPM 4
380.160 3
283,980 DAVIS-CESSE, UNIT 1 2-3 Amendnent No. JJ, JJ, 73, M, f)
Me 9I y,
r-T t
= - - -
Figure 2.2-1 Trip Setpoint for Flux -- aFlux/ Flow i
Curve shows trip setpoint for an approximately 25% flow reduction for three pump operation (283,980 gpm). The actual setpoint will be directly proportional to the actual flow with three pumps.
- RATED THERMAL POWER
- 120 UNACCEPTABLE OPERATION UNACCEPTABLE OPERATION
(-18.2.106.8)
(18.2.106.8)
M =1.000 M =-1.000 3
PUMP
- 100 2
LIMIT I
I
(-34.0,91.0) 34.0,91.0) i I
1 EO (18.2,I79.7)
(-18.2J79.7) l 3iPUMP
(-34.0,63.9) l LIf!T l
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34.0,63.9)
- 60 l
g 1
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ACCEPTABLE OPERATION F@
l SPECIhlEDRCPJMP COMBqNATION. I I
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- 20 l
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-60
-40
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20 40 60 Axial Power Imbalance, %
DAVIS-BESSE, UNIT 1 2-7 knendment No. 77, 75, JA, /J AI, Jtf, 9I
POWER DISTRIBUTION LIMITS DNB PARAMETERS LIMITING CONDITION FOR OPERATION 3.2.5 The following DNB related parameters shall be maintained within the limits shown on Table 3.2-1.
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a.
Reactor Coolant Hot Leg Temperature b.
Reactor Coolant Pressure c.
Reactor Coolant Flow Rate APPLICABILITY: MODE 1 ACTION:
If parameter a or b above exceeds its limit, restore the parameter to within its limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 5% 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 />.
If parameter c exceeds its limit either:
1.
Restore the parameter to within its limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, or 2.
Limit THERMAL POWER at least 21 below RATED THERMAL POWER for each 1%
parameter c is outside its limit for four pump operation within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. or limit THERMAL POWER at least 21 below 75% of RATED THERMAL POWER for each 15 parameter c is outside its limit for 3 pump operation within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.2.5.1 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 />.
4.2.5.2 The Reactor Coolant System total flow rate shall be determined
~
to be within its limit by measurement at least once per 18 months.
DAVIS-BESSE, UNIT 1 3/4 2-13 Amendment No. 64 wm
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it 1Antt 3.2-1 k
- DNS MARGIN G
a, LIM 115 m
h, Four Reactor 11eree Reactor Coolant Pumps Coolant Pumps E
Parameter Operallnq Operatine "4
Reactor Coelant Hot Leg 1 610 1 61nIII lemperature i *F g
ReactorCoolantPressure,psig.(2) 1 2062.7 t 2058.7III I3I 1 99,664 g 291',080 l
Reactor Coolant flow Rate, ppn 3
1 7%
N Applicable to tk loop with 2 Reactor Coolant Pumps Operating.
(2) Limit not applicable during either a TIERMAL POurR ramp increase in excess of 51 of
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MIES ilERNAL POWER per minute or a llERMAL POWER step increase of greater than 101 of MIED llERNAL POWER.
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I3IThese flotis include a flow rate uncertainty of 2.55, and are based 'on a minimum of 64 lumped burnable poison rod assemblies in place in the core.
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' SAFETY LIMITS 1
BASES The reactor trip envelope appears to approach the safety limits more close-ly than it actually does because the reactor trip pressures are measured at a location where the indicated pressure is about 30 psi less than core out-let pressure, providirg a more conservative margin to the safety limit.
The curves of Figure 2.1-2 are based on the more restrictive of two thermal limits and account for the effects of potential fuel densification and po-tential fuel rod bow.
1.
The 1.30 DNBR limit produced by a nuclear power peaking factor of FO" 2.56 or the combination of the radial peak, axial peak, and position of the axial peak that yields no less than a 1.30 DNBR.
2.
The combination of radial and axial peak that causes central fuel melt-ing at the hot s;:ot. The limits are 20.4 kW/f t for batches 1E, 48, and SA and 20.5 kW/ft for batenes SB, 6, and 7.
Power peaking is not a directly observable quantity and therefore limits have been established on the basis of the reactor power imbalance produced by the power peaking.
The specified flow rates for curves 1 and 2 of Figure 2.1-2 correspond to the expected minimum flow rates with four pumps and three pumps, respective-ly.
The curve of Figure 2.1-1 is the most restrictive of all possible reactor coolant pump-maximum thermal power combinations shown in BASES Figure 2.1.
The curves of BASES Figure 2.1 represent the conditions at which a minimum DNBR of 1.30 is predicted at the maximum possible thermal power for the num-ber of reactor coolant pumps in operation or the local quality at the point of minimum DNBR is equal to +22%, whichever condition is more restrictive.
These curves include the potential effects of fuel rod bow and fuel densifi-Cation.
The DNBR as calculated by the B&W-2 DNB correlation continually increases from point of minimum DNBR, so that the exit OWR is always higher. Extrap-olation of the correlation beyond its published quality range of +225 is justified on the basis of experimental data.
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8 2-2 DAVIS-BESSE, Ui:IT 1 haendment flo. 77, JE, #, 57,
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2.1 SAFETY LIMIT 3 BASES l
2.1.1 and 2.1.2 REACTOR CORE The restrictions of this safety limit prevent overheating of the fuel cladding and possible cladding perforation unich would result in the release of fission products to the reactor coolant. Overneating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime wnere the heat transfer coefficient is large and the cladding surface tamperature is sligntly aoove the coolant saturation temperature.
Ooeration abovt the unoer bouncary of the nucleate boiling regime would result in excessive claading temoeratures pecause of the onset of cecarture from nucleate boiling (DN8) and the resultant snare reduction in heat transfer coefficient. DNS is not a directly measuraole parameter during operation and therefore TriERMAL POWER and Reactor Coolant Tenper-ature and Pressure have been related to DN8 througn the 88W-2 DNS correlation. The DN8 correlation has been develcoed to predict the DN8 flux and tne location of DNS for axially unifom ano non-onifam neat flux distributiens. The local DN8 neat flux ratio. DNBR, cefinec as the ratio of the neat flux that would cause DN8 at a particular core location to tne local heat flux, is incicative of the margin to DNS.
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The minimum value of tne DNBR during steady state operation, nomal operational transients, and anticipated transients is limited to 1.30.
This value corresponds to a 95 percent probability at a 95 percent confinance level that DN8 will not occur and is chosen as an appropriate margin to DN8 for all operating conditions.
The curve presented in Figure 2.1-1 represents the conditions at wnicr.
a minisus DN8R of 1.30 is predicted for the maxisam possible thermal power 112 wnen the reactor coolant flow is 380,160 tiPM, which is 108% of I
design flow rate for four operating reactor coolant pumps. This urve is based on the following hot channel factors with potential fuel densifi-cation and fuel rod bowing effects:
N Fq = 2.56; Fg = 1.71; F"g = 1.50 The design limit power peaking factors are the most restrictive calculated at full power for the range from all control rods fully witadrawn to minism allowable control rod witharawal, and fem tne core DN8R design basis.
DAVIS-4 ESSE. UNIT 1 82-1 Amendment No..h. 91 o
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LIMmNG SAFITY SYSTEM SETTINGS SASES Contatorit Nfon enssun The Cantairunent High Pmssure Trip $atacint e 4 psig, pmvides positive assuranca that a mactor trip will occur ~in the unlikely event of a staan line failun in the contaiment vessel or a less-of.
coolant ac:1 dent, even in the absence of a AC Low Pressum trip.
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i ames-Fiaure 2J Pressura/ Temperature Limits at Maximum Allowable l
Power for Mini== DN3R l
d 2400 3 PUMP CURVE 2300 3
4 PUMP CURVE 3
2200
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5 2
2100
=
G 2000
=
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1903 1300 I
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!!O 630 510 820 630
$40 Rea:t:r Outlet Temnerature. (*F) p'.'v p s
- i.gv qpu s povEn 4
300,1.50 1125 3
283,980 89.1%
0AV;5-3I53I, C;IT 1 3 2-8 Amendment flo., X,X,4s. 9I
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