ML20080K298
| ML20080K298 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 08/30/1983 |
| From: | Stolz J Office of Nuclear Reactor Regulation |
| To: | Toledo Edison Co, Cleveland Electric Illuminating Co |
| Shared Package | |
| ML20080K301 | List: |
| References | |
| NPF-03-A-060 NUDOCS 8309290093 | |
| Download: ML20080K298 (9) | |
Text
pu%q k yi UNITED STATES y
p, NUCLEAR REGULATORY COMMISSION 3
- j WASHINGTON, D. C. 205,55
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THE TOLEDO EDISON COMPANY AND THE CLEVELAND ELECTRIC ILLUMINATING COMPANY DOCKET NO. 50-346 DAVIS-BESSE NUCLEAR POWER STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 60 License No. NPF-3 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by The Toledo Edison Company and The Cleveland Electric Illuminating Company (the licensees) dated October 14, 1982, complies with the standards and require-ments 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 application, 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 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.
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s 8309290093 830830 PDR ADOCK 05000346 l
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2 2.
Accordingly, Facility Operating License No. NPF-3 is hereby amended as indicatad belcw and by changes to the Technical Specifications as indicated in the attachment to this license amendment:
Revise paragraph 2.C.(2) to read as follows:
Technical Soecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.60, are hereby incorporated in the license.
The Toledo Edison Company snall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of the date of its issuance.
R.THE NUCLEAR REGULATORY CCMMISSION 7
/
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chrkF.Stol:, Chief Ope ting Reactors Branch #4 ision of Licensing Attacnment:
Changes to the Technical Specifications Date of issuance: August 30, 1903 s'
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ATTACHMENT TO LICENSE AMENDMENT NO. 60 FACILITY OPERATING LICENSE NO. NPF-3 DOCKET NO. 50-346 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages as indicated.
The revised pages are identified by Amendment numoer and contain vertical lines indicating the area of change.
The corresponding overleaf pages are also provided to maintain document completeness.
Paaes 3 2-6 3/44-4 8 3/4 4-la b
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i j ' LIMIT!NG SAF ti SYSTEM SETTINGS EASES 1
RC Hieh Tencerature The RC high temperature trip <618 F pr' eve'nts the reactor outlet temperatura l
from exceeding the design limits and acts is 'a backup trip for all power ex-j cursion transients.
Flux -- A Flux / Flow Tha power level trip setpoint produced by the reactor coolant system flow is based on a flux-co-flow ratio which has been established c'o accommodate flow decreasing transients from high power where protection is not provided by the high' flux / number of reactor coolant pumps on trips.
The power level trip setpoint produced by the power-to-flow ratio provides both high power level and low flow protection in the event the reactor power level increases or the reactor coolant flow rate decreases.
The power level setpoint produced by the power-to-flow racio' provides overpower DN3 protection For every flow race there is a maxi =cm par-Ifor all modes of pump operation.
ble
.missible power level, and for every power level there is a minimum permissi Examples of typical power level and low flow rate combinations lov flow rate.
for the pu=p situations of Table 2.2-1 that would result in a trip are as j
! allows:
l Tre; would occur when four reactor coolant pumps are operating if power l
l il.
flow race is 100% of full flow rate, or flow is 107.0 and reactor coolant of full flow rate and power level is 100".
race is 93.5%
Trip would occur.when three reactor coolant pumps are operating if' power is 80.0% and reactor coolant flow rate is 74.7% of full flev rate, or flow 2.
70.0% of full flev rate and power is 75%.
rate is For safety calculations the maximum calibration and instrumentation errors for the power level were used.
Full flow rate in the above two examples is defined as the flow calculated by the heat balance at 100% power.
l B 2-5 Amendment No. 36',3{,45 l.
, DAVIS-3 ESSE, UNIT 1 w. - - - --
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!LIMITING SAFETY SYSTEM Sta.INGS BASES Th9 AXIAL POWER Di3AIANCE boundaries are established in order.co prevent re-cetor thernal limits from being exceeded.
These ther=al limits are either power peaking kW/ft limits or DN3R limits.
The AXIAL POWER Di3AiiNCE reduces the power level trip produced by a flux-co-flow ratio such that the boundaries nof Figure 2.2-1 are produced.
I I
RC Pressure - Low. High, and Pressure Temperature 131e high and low trips are provided to limit the pressure range in which re-actor operation is per=1tted.
During a slow reactivity insertion startup accident frem low power or a slow rsactivity insertion from high power, the RC.high prsssure setpoint is reached b2 fore the high flux trip.setpoint.
The trip setpoint for RC high pressure, 2300 psig, has been established to maintain the system pressure below the safe-cy limit, 2750 psig, for any design transient.
The RC high,1ressure trip is backed up by the pressurizer code safety valves for RCS over pressure protec-tion, and is therefore set lower than the set pressure for these valves, s 2525 l
psig.
The RC high pressure trip also backs up the high flux trip.
Ghe RC low pressure,1983.4 as.ig, and RC oressure-temnerature (12.60 To ko-5662) psig, trip s'etpoints have been established to maintain the DN3 rat greater than or equal 'to 1.30 for those design accidents that result in a pressura reduction.
It also prevents reactor operation at pressures below the valid range of DN3 correlation li=its, protecting against DN3.
i Eigh Flux / Number of Reactor Coolant Pumos ?n In conjunction with the flux - a flux / flow trip the high flux / number of reac-tor coolant pu=ps on trip prevents the minimum core DN3R from decreasing below The il.30 by tripping the reactor due to the loss of reactor coolant punp(s).
l pump monitors also restrict the power level for the number of pumps in operation.
I Amendment No. JMT/A, 60
- j. DAVIS-BESSE, UNIT 1 B 2-6
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3.4.2 Decay Hea Re== val Syste= relief valve DH-4849 shall be 0? IRA 3.LI_.
-1:h a lif:. se::ing of.$ 330 PSIO* a=d isc'a:ic: valves DE-11 a=d DE-12 epe: a=4 cc=:: 1 pcva: :o. heir valva op2:ators re==ved.
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ACTION:
A.
'dith DH-4849 no: OPERA 3LE:
1.
E.ak.a the valve 0?IP.A31.E vi
eigh: hours; or 2.
a.
W1:
=e== c=a hour, disable the capab:7 cf be h high pressure i=j ectic'= (E?!) pu=ps== i=j a:: vatar i=:s the reacter c=cla== systa=; a=d b.
W':
=ez: eigh: hours:
1.
Disable the au:==a:1= =a=sf er of -skaup pt. p suctie=
to the berated water s = rage ta=k c= ics =akeup ra=k level; a=d 2.
Heduca _a.kaup a=k level to 4 73 i=ches a=4 reduca reac:== coola=: systa= pressura n-2 pressur'.=e lev 4.1 vi@'
- he acceptable regic= c= Fi'gures 3.4.2-a (i= vmE 4) a=d 3.4.2-b (i= Y.0DI 5).
3.
V' h DE-11 or DE-12 closed, ope = DE-21 a=d DE-23 vi-W o=a hour.
C.
Wi:h :he c=== o1 power o: re=cred fro = DE-11 a=d DE-12, r e=va the pcver :o the valve operaters a: the F.cter Cc====1 Ce=:ers vi*-
o=a hour.
ETI-" A.NCE ?2CUIR."Ih"IS 4.4.2.
Decay Esa: Re=cval Systa= relief valve DE-4849 shall be datar-
=i=ed CP'"'*M
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per the su=vailla=ca require =a=
s of Specifica:Lo= 4.0.5..
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b.
at least c=ca per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by ve=1fying ai:her:-
i 1.
1sola:1o= vaivas DE-11 a=d DE-12 epen vi:h e====1 power
.re=eved f== their valve operators; or 2.
valves DE-21 a=d DE-23 epe=.
ha : sat:1=g pressure sha.11 c== espe =d to a= hie== c==di:1==s of the valve a: =c_.'.=al cperazi=g :e=perature a=d pressura.
CATIS-3ISSE, UNI 1 3/4 4^3 Amendment No. U
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..iLn C..iLn t.. t. C.N :., R O F e..... C N 3.4.3 All pressuri:er ::de safety valves shall be CFE,A31.E wie a i
lift sa::ing Of < 2525 DSIG.*
When ne isolated, c e pr4ssuri:er I
!' e.!e::r:=atic relief valve snail have a trip se:; in: Of >
2250'PSIE ii. and an allewable value Of 3, e3n0 psici. -
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Wim :ne :rsssuri:gr ::de safety' valve ine:erable, ef ter rss :re te ir.::eratie valve :: "FE:J5LI status wi-hin 15 =inutas Or te in MCT 3E1.7.,~WN di*'.in 12 0:urs.
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s 4.4.2 F r ce :ressuri:er ::de safety valves, =ere are n: aediti:nal Surveillance Re:uirerents :her than these re:uired by 5:eci#ication 1.C.5.
.:P Oe : tssuri:er ele:: :=ati: elief valve a channel cali-bra:icn ce:k snall be ;er'er:ed.every 18 mn:hs.
- ne hf: se:::ng :ressure shall ::r-ss:end := adient ::nditiens Of
=e vaive a: nc=i'nal ::erating ta=:erature and pressure.
t
.~ Allewa:ie value f:r enannel calibrati:n check.
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A endnen: No. 23,60
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RE.:C.ux CCCLINT SYS EM
?ASES Ne pressurizer coce safety valves must be set such that the peak Reactor Coolant System pressure does not exceed 110% of design system pressure (2500 psig) or, 2750 psig.
The control rod group withdrawal accident will result in the most limiting high pressure in the RCS.
The analysis assumes RPS high pressure trip at 2300 psig and the ' code safety valves open at 2500 psig.
The tolerance on the RPS instrument accuracy is 30 psi and, it is +31 for the code safety valve settings.
The pressurizer electromatic relief valve was assumed not to open for this transient.
The resulting system peak pressure was calculated to be 2715 psig.
Therefore, the code safety valve setpoint is conserva ~
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tively set at < 2525 ps.ig which is the maximum pressure of 2500 psig
+15 for tolerance.
The pressurizer electromatic relief valve should be set such that i-t will open be fore the code safety valves are opened.
However, it should not open on any anticipated transients.
Loss of Feedwater (LOFW) v as identified as the limiting anticipated transient for RCS pressure.
T'1e analysis assumes RPS high pressure trip at 2300 psig; with 30 psi #or inst-ument errors, the resulting peak RCS pressure is calculated te te 2380 psi g.
This includes a 50 psig pressure overshoot on a LCR4 transient.
3 AVIS-BE5SE, UNIT 1 3'3/4 4-la Amencmen No 7;,60
REACTOR COOLANT SYSTEM SASES 3/4.4.4 PRESSURIZER A steam bubble in the pressurizer ensures that the RCS is not a hydraulically solid system and p capable of acconsnodating pressure surges during operation. The steam bubble also protects the pressurizar code safety valves and power operated relief valves against water relief.
The low level limit is based on providing enough water volume to prevent a reactor coolant system low pressure condition that would actuate the Reactor Protection System or the Safety Featura Ac*wation Systam.
The hign level limit is based 'on providing enougn steam volume to prevent a pressurizer nich level as a result of any transi ent.
The power operated relief valves and steam bubble function to relieve RCS pressure during all design transients. Operation of the power operated reifef valves minimi:es the undesirable opening of the spring-loaded pressuri:er code, safety valves.
3/4.a.5 STEAM GENERATORS The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be maintained. The program for inservice inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision 1.
Insarvice inspection of steam generator tubing is essential in order to maintain surveillance of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or inservice conditions that lead to corrosion.
Inservice inspection of steam generator tubing also provides a means of charactarizing the nature and cause of any tube degradation so that corrective measures can be taken.
The plant is expected to be operated in a manner such that the secondary coolant will be maintained within those chemistry limits found to result in negligible corrosion of the steam generator tubes.
If the secondary coolant chemistry is not maintained within these chemistry limits, localized ccerosion may likely result in stress corrosion crackini;.
The extant of cracking during plant operation would be limitad by the limitation of steam generator tube leakage between the primary coolant system and the secondary coolant system (primary-to-secondary leakage = 1 GPM).
Cracks having a primary-to-secondary leakage less than this limit during operation will have an adequata margin of safety to withstand the loads DAVIS-BESSE, UNIT 1 83/44-2
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