ML20006E364
| ML20006E364 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 02/05/1990 |
| From: | TOLEDO EDISON CO. |
| To: | |
| Shared Package | |
| ML20006E361 | List: |
| References | |
| NUDOCS 9002220745 | |
| Download: ML20006E364 (11) | |
Text
. _ _ _ _
Dockot Numbar 50-346
~Licensa Numb 2r NPF-3 Serial Number 1760
- Attachment.2 ,
Page 1 2.1 SAFETT LIMITS
. ,1 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 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 would result in excessive cladding temperatures 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_3eactor _
Coolmat T -n=rature =ad Jea=surt havagen related to DNR G#vs.y. d.. ksg c.Mg W _corrielad f.e' f::f i. dia . d h!;'h i L his;;f ::
,.. . 10 0,1 "" f1= :nf 2: 1 :::ir ;;i;11; ' ' ;; _..d
- ; _ ;.i f - - h;- - flu f i ;;' t ;i_;;;> leat flux rat o, DNBR, dit of the heat flux that would cause DNB at a particular core location to the local heat flux, is indicative of the marain to'DNB.
3 p-e minimum value of the DNBR during steady state operation, nor rational transients, and anticipated transients is limited to 1.3 W 24 4 The value corresponds to a 95 percent probability at a 95 percent MY confidence level'that DNB vill not cecur and is chosen as an appropriate margin to DNB for all o ra nr randi riam u .. re a, .% & c ereWa* 16:4 The curve presen n F gure 1-1 represents the conditions at which a minimum DNB ;fJ.2^ is predicted for the maximum possible thermal power 112% when the reactor coolant flow is 380,000 GPM, which is approximately 108% of design flow rate for four operating reactor coolant pumps. -(The minimum required measured flov is 389,500 GPM.)
This curve'is based on the following hot channel factors with potential fuel densification and fuel rod bowing effects:
F 0 - 2.83; F"g = 1.71; F 3 = 1.65 The design limit power peaking factors are the most restrictive calculated at full power for the range from all control rods fully withdrawn to minimum allovable control rod withdraval, and form the core DNBR design basis.
DAVIS-BESSE, UNIT 1 B 2-1 Amendment No. 11, 33, 91, 123 9002220745 900205 PDR ADOCK 05000346 F# PNV
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[' (Dockht' Number 50-346
(.
.LicInsa Number NPT-3 Serial Number 1760
'-Attachment 2 Page'2 Insert A-The B&V-2 and BVC CHF correlations have been developed _to predict DNB for
.;..a.. ~._ . axially-uniform and non-uniform heat flux distributions.' The'B&V --- - " - - "--
1 correlation applies to Mark-B fuel and the BVC correlation applies to all B&W fuel.vith zircaloy spacer grids, s,~
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Docket Numb 3r 50-346
' License Number NPF-3 Serial Number-1760-Attachment 2 Page 3 SAFETY LIMITS -
l BASES 1
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 potential fuel rod bow.
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- 1. - The DNBR limit produced by a nuclear power peaking factor of F .- . 3 or the combination of the radial peak, axial _oeakm and pbsitionoftheaxialpeakthatyieldsnolesst .T. *^ D'd. l e bu_8FJ ie it .
- 2. The combination of radial and axial jh *h-+ r="--- -e*
melting at the The limit . . . 22.0 k./ft f a bat;h 1" 2..d 20.5 kv/ft fo vLg g e..d 0 Power peaking is not a directly observable quantity and therefore limits I have been established on the basis of the reactor power imbalance l produced by the power peaking.
4 The specified flow rates for the two curves of Figure 2.1-2 correspond i to the analyzed minimum flow rates vith four pumps and three pumps, !
respectively.
The curve of Figure 2.1-1 is the most restrictive of all possible reactor coolant pump-maximum thermal power combinations shown in BASES nura (1 _ Th-_
--u s_o f _ SES Figure 2.1 represent the conditions at W J.;- ML-' ef '. 2^ is predicted at the maximum possible i
power for t r of reactor coolant pumps in 9parchste
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- he local quality at the point of minimum DN b ion is agg pgrictive.
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i DAVIS-BESSE, UNIT 1 B 2-2 Amendment No. 11,
-il 33, 45, 61, 80, 123 l
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Docket Numbar 50-346
'Licensa Numbar NPF Serial Number 1760 Attachment 2 Page 4 SAFETT LIMITS
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BASES - ~~~ ~~ ~ '
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er I.l &(B@Y-For the curve of BASES Figure 2.1, a pressure-temperature point d to the left of the curve vould result in a DNBR greater t ger
. ee a local quality at the point of minimus'DNBR than +2 gg(3pe) tha articular reactor coolant pump situation. The R cu or ' ~ - ~ -
three pump operation is less restrictive than the four pump curve.
2.1.3 REACTOR COOLANT SYSTEM PRESSURE The restriction of this Safety Limit protects the integrity of the Reactor Coolant System from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere.
The reactor pressure vessel and pressurizer are designed to Section III of tha ASME Boiler and Pressure Vessel Code which permits a maximum transient pressure of 110%, 2750 psig, of design pressure. The Reactor
-Coolant Systes piping, valves and fittings, are designed to ANSI B 31.7, t
1968 Edition, which permits a maximum transient pressure of 110%, 2750 psig, of component design pressure. The Safety Limit of 2750 psig is i
L therefore consistent with the design criteria and associated code requirements. ,
The entire Reactor Coolant System is hydrotested at 3125 psig,125%
of design pressure, to demonstrate integrity prior to initial operation.
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1 DAVIS-BESSE. UNIT 1 B 2-3 Amendment No. 11, 33, 45, 123
' Docket Numbar 50-346' m
h, Licenes Number NPF-3:
4.\ Serial Humber.1760
- i Attachment 2 '
, ggg; ;g g,g Pap 5 ,, g , . ,g Ser at no ]koj ,te 12///84 l LIMITING SAFETY SYSTEM SETTING L BASES _
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The AXIAL POWER IMBALAtCE bouncaries are estaclishec in order to crevent reac-ter therinal limits fran being exceeced. These thermal limits are either power peaking kW/ft limits or Of8R : limits. *he AXIAL POWER IMB ALANCE recuces the power level trio produced by a flux-to-flow ratio such that the bounca-ries of Figure 2.2-1 are procuced.-
RC Pressure - low. Mich. and Pressure Temeerature The high and low trios are provided to limit the pressure range in which reac-ter operation is permitted.
t During a slow reactivity insertion startuo accident from low oower or a slow reactivity insertion from high power, the RC high cressure setpoint is-gg m e % hefn tne nign flux trip setpoint. The trio satooint for RC hign press ure psig, has been established to maintain tne system pressure ee-low the sa ty limit, 2750 psig, for any design transient. The RC high cres-sure trip is backed up by the pressurizer code safety valves for RCS over pressure protection, anc is therefore set lower than tne set crassure for -
these flux trip.
valves,1 2525 psig. The RC high pressure trip also backs up the high The RC low pressure ,1983.4' psig, and RC pressure-temoerature (12.60 tout -
5662.2) psig, trip setpoi t have oeen establishec to maintain the ONB ratio greater than or ecual to for those design accicents that result in a l cressure reduction. It a s events reactor coeration at cressures below the valid range of 0?S corr i n limits st Ote .
e m iin. protec a ble4'bHB u &
Hien Flux / Number of Reactor Coo nt u.
In conjunction with tne flux - aflux/ flow trip tne nign flux /nuccer of reac-ter cool t pumos on trio prevents the minimum core Of8R from cecreasing below by trioping the reactor due to tne loss of reactor coolant oumo(s J The pumo monitors also restrict tne oower level for the numoer of pumos i / peration.
1: A= miaW* allowW 'b MB rab i
S 2-6 DAVIS-BESSE, U:!IT 1 Amendment !!o. 22, /3, C, J
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Dock 3t'Humb2r 50 346
- bicense Humber N M*3
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" "**a LIMITING'SIFETY SYSTEM SETTINGS s
. BASES Containment High Pressure
.The Containment High Pressure Trip Setpoint < 4 psig, provides positive assurance that a reactor trip will occur in the unlikely
. event of a steam line failure in the containment vessel. or a loss-of-
~
coolant accident, even in the absence of a RC 1.ow Pressure trip.
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BAVIS-BESSE, UNIT 1 B 2-7 (s
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DockstNumbar-50-346;
- Licsnso;Humbar HPF-3.
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DE D s v,q,; .u Serial Number 1760- ,7 ;g.,y .,
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, Bases Figure 2.1. Pressure / Temperature Limits at Maximum- :
Allowable Power for~ Minimum DNBR J
2300-1 i
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2200 -
(636.3,?.159.8) t
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-(633.4,2129.8)' / .
2100 -
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E ACCEPTABLE 2000 /
. OPERATION m
- (621.4.1929.8) / (625.7,1959.8) ,_l 1900 =
/ UNACCEPTABLE -i
/ ;0PERATION
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f -- - 3 Pump 1800 -
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7
-(608.2 I (614.3,1759.8) 1729.8) 1700 t i e t
- , 595 605 615 625 635 645 I
Reactor Outlet Temperature,'F '
Required Measured Flow to ensure Pumos . Flow, gom Power = Comoliance, com 4 380',000 112". . 389,500 3 283,860 90.5% 290,957 t
-DAVIS-BESSE, UNIT 1 B 2-8 Amendment No. 11, 33, 45, 91,123 4
Docket Number 50-346- '
1 License,Humber NPF-3
= Serial Humber 1760
- Attachment 2 Page 8
'13/4.2. POWER DISTRIBUTION LDi1TS :
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BASES l
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f fuel integrity during
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The specifications ?* this see n )Gnour aurorta l Condition I (notw; cyeration) and II (incidents o odorate frequency) events- )
uring normal opera- j
)~ by: (a) naintaining tne minimus DNBR in the core 2 l tion and during short ters transients. (b) maintaining the peak linear power density s 18.4 kW/f t during normal operation, and (c) maintaining the peak power density less than the limits given in the bases to specification 2.1 during short term transients. In addition, the above criteria must be met in order to meet the assumptions used for the loss-of-coolant accidents.
The power imbalance envelope defined in Figures 3.2-1 and 3.2-2 and the insertion limit curves. Figures 3.1-2 and 3.1-3 are based on LOCA analyses which have defined the maximum linear heat rate such that the caximum clad temperature vill cot exceed the Final Acceptance Criteria of 2200*T following a LOCA. Operation outside of the power imbalance envelope alone does not con-st1tute a situation that would cause the Final Acceptance Criteria to be ex-caeded should a LOCA occur. The power imbalance envelope represents the bound-ary of operation limited by the Final Acceptance Criteria only if the. control rods are at the -insertion limits,' as defined by_ Tigures 3.1-2 and 3.1-3 and if the steady-state limit QUADRAlft POWER TILT exists. Additional conservatisn'is.
introduced by application of:-
- a. Nuclear uncertainty factors.
- b. Thermal calibration uncertainty.
- c. Tual densification effects.
- d. Hot rod manuf acturing tolerance f actors.
- s. P'otencial fual r'od tiov aff acts.
l The ACTION statements which permit limited variations from- the basic require-ments are accompanied by additional restrictions which ensures that the orig-l inal criteria are met.
The ' definitions of the design limit nuclear power peaking factors as used in these specifications are as follows:
T Nuciaar heat. fluz het channel factor, is defined as the nazimum local fuel 9 rod linear power density divided by the average fuel rod linear power den-sity, assuming nominal fuel pellet and rod dimensions.
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L DAVIS-BEssE, UNIT 1 B 3/4 2-1 Amendnent No. .WA,45
' Docket Number 50-346-License Number NPF-3 ,,
Serial Humber-1760 Attachment 2 .
Page 9-
'P0kTR DISTRIBUTION LIMITS BASES N
b, ne seasurement of enthalpy rise hot channel- factor. Fg. shall be in--
creased by 5 percent to account for seasurement error.
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For Condition II events, the core is protected frou exceedin paha)ases co - specification 2.1 locally, and from going bel A6 i i !.% by automatic protection on power. AXIAL PokTR IMBALAN p tempe ure. Only conditions 1 through 3. above, are aandatory since the AXIAL P0kTR IMBALANCE is an explicit input to the reactor protection system.
The QUADRANT POWER TILT limit ' assures that the radial power distribution aat-isfies the design values used in the power capability analysia. Radial power distribution esasurements are made during startup testing and periodically dur-ing> power operation. .
The QUADRANT pohTR TILT limit at which corrective action is required provides DNB and linear heat generation rate protection with x-y plane power. tilts. In.
the event the tilt is not corrected, the margin for uncertainty on Tg is-rain-stated by reducing the power by 2 percent, for each percent of tilt in excess of the limit.
3]I.2.5. DNB PARAMETIRS The limits on the DNB related parameters assure that each of the parameters are maintained within-the normal steady state envelope of operation assumed in the transient and accident analyses. Ihe limits are consistant with the FSAR initial assumptions a e been analytically demonstrated adequate to main--
tain a minimus D -d . . M hea>> h ==-h ===1===d er=== sat..
3 reded w nc swielewd allomMe bN 8 ratio ne 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> period c surve ance o t . par t es ough instrument read-out is ufficiant to ensure that the parameters are restored within their lia-its following load changes and other expected transient operation. The 18 month periodic asasure:sent of the RCS total flow rate using delta P instruman-tation is adequate to detect flow degradation and ensure correlation of the flow indication channels .with asasured flow such that the indicated percent flow will provide sufficient verification of flow rate. on a 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> basis. .
.l DAVIS-BESSE, UNIT 1 B 3/4 2-3 Amendment No. M ,'45
D'ocket.Humber 50 346.
1Licenes Number NPF U
-Serial Number 1760
. . Attachment 2-
, Page 10.
3/4.4 REACTOR COOLANT SYSTEM BASES-
- h 3/4.4.1 ' REACTOR COOLANT LOOPS _ / b *O== alles.nelle 1U B rab w.
The plan i ..n oa o p ate v th reactor coolant loops.in operation ' amintain DNBR abov during all normal operations and anticipate iones. Vith one ter coolant pump not in operation [
in one loop . "- POWER is restricted .by the Nuclear Overpower Based .
on RCS Flov . POWER INRALAMr2 ensuring that the DNBR will be amintained abov at the maximum possible TPetal POWER for' the number of reactor coolant p .l
{
operation or the local quality at the
. point of minimus DNBR equal whiehawae in m eestrictive. l.
MM*?M5 Did In M0DE 3 when RCS pressure or temperaturiWhTgHaTr ~than the decay heat removal system's design condition (i.e. 330 psig and 350'F), a single ,
reactor coolant loop provides-sufficient heat removal capability. The '
rematader of MODE 3 as well as in MODES 4 and 5 either a single reactor coolant loop or a DBR loop will be-sufficient for decay heat removals but single failure considerations require that at least two loops be
' OPERABLE. Thus. if the reaetor coolant loops are not -0PERAELE. this specification requires two DER loops to be OPnARLE.
3 Natural circulation flow or the operation of one DER pump provides ad-
-equate flow to ensure sizing, prevent stratification and produce gradual reactivity changes during boron consentration reductions in the Reactor Coolant System. The- reactivity change rate associated with baron redue-tion vill. therefore, be within the capacity of operator reaognition and' control. -!
3/4.4.2 and 3/4.4.3 SAFETT VALVES
' The pressuriser code safety valves operate to prevent the RCS- from being pressurised aoove its safety Limit of 2750 psas.. Each safety valve 1s-designed to relieve 336,000 lbs per hour of saturated steam at the:
-valve's setpoint.
The relief capacity of a single safety valve is adequate to relieve any I
. overpressure condition which could occur during shutdown. In the event that no safety valves are OPER&BLE. an operating DER loop, connected to the RCS, provides overpressure relief capability and vill prevent RCS overpressurization. During operation. all pressuriser code safety valves east be OPER&BLR to prevent the RCS from being. pressurized above its safety limit.of 2750 psig. The combined relief capacity of all of these valves is greater than the maniana surge rate resulting from any transient.
The relief capacity of the decay heat removal system relief valve is adequate to relieve any overpressure condition which could occur during shutdown. In the event that-this relief valve is not OPnARLE. reactor coolant system pressure, pressuriser level and make up water inventory is limited and the capability of the high pressure injection system to DAVIS-BESSE. LWIT 1 B 3/4 6-l Aaenament No.33.38.57.
92,128
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Dockht Nuab2r-50-346' !
License Number NPF-3 Serial Number 1760 1
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BAV-2096,'" Davis-Besse Nuclear Power Station.
Unit 1, Cycle 7 -- Reload Report", November 1989 I
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aDockst Nu:ber 50-346 '
ly. l: .* License-Number NPF-3' LUi ' Serial Number 1760.
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Babcock & Wilcox Document Identifier 86-1177306-00, "TECO Mark-BZ FA Seismic and,LOCA Analysis" i-i
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