ML20064N181

Proposed Tech Specs,Relocating Certain cycle-specific Protective Limits & Core Operating Limits to COLR

ML20064N181
Person / Time
Site:	Davis Besse
Issue date:	03/18/1994
From:	CENTERIOR ENERGY
To:
Shared Package
ML20064N175	List: ML20064N173 ML20064N179 ML20064N181
References
2197, NUDOCS 9403290203
Download: ML20064N181 (29)

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{{#Wiki_filter:.1 Docliet Number 50-346 .. License Number NPF-3 Serial Number 2197 Attachment Page 5 INDEX .) ADMINISTRATIVE CONTROLS SECTION PAGE Meeting Frequency........................................ 6-9 Quorum................................................... 6-9 Review.................................................. 6-10 Audits................................................... 6-11 1 Authority................................................ 6-12 Records.................................................. 6-12 6.5.3 Technical Review and Control............................. 6-12 6.6 REPORTABLE EVENT ACTION....................................... 6-12a -~ SAFETY LIMIT VIOLATIO _..K.PROTEC:nVE LIM IT VIOLATiod 6-13 1 6.7 1 - + _~ - _ _ __w 6.8 PROCEDURES AND PR0 GRAMS....................................... 6-13 l 6.9 REPORTING REQUIREMENTS 6.9.1 Routine Reports.......................................... 6-14C 6.9.2 Special Reports.......................................... 6-18 i 6.10 RECORD RETENTf0N............................................. 6-18 6.11 R ADI AT ION P ROT ECTION PR0 GRAM _................................. 6-20 l 6.12 HIGH RADIATION AREA.......................................... 6-20 l 6.13 E NV I RONME NTAL QU AL I F I CAT I ON.................................. 6-21 6.14 P ROCESS CONTROL PROGRAM ( PCP)................................ 6-22 6.15 0FFSITE DOSE CALCULATION MANUAL (00CM)....................... 6-22 CAVIS-BESSE UNIT 1 xyl Amendment No. 33.L 1 170 9403290203 Y4031t3 PDR ADOCK 05000346 P PDR

c. Ddchet. Number 50-346

• - ' License ~ Number NPF-3 Seria1_ Number 2197 Attachment Page 6

.i 2.0 SATETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETT LIMITS REACTOR CORE 2.1.1 The combination of the reactor coolant core outlet pressure and outlet temperature shall not exceed the safety limit shovn in Figure 2.1-1. j APPLICABILITY: MODES 1 and 2. ACTION: Vhenever the point defined by the combination of reactor coolant core outlet pressure and outlet temperature has exceeded the safety limit, be in HOT STANDBY vithin one hour. REACTOR CORE ,,{gc, & y 2.1.2 The combination f reactor THERMAL POVER and AXIAL POVER IMBALANCE .l shall not exceed the limit shovn in N. m 2."-2 for the various 1 combinations three and four reactor coolant pump operation. il, ' C ogi. OPEK.A710(s 1

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APPLICABILITY: MODE 1. tf ACTION: Whenever the point defined by the combination of Reactor. coolant System-flow, AIIAL P IMBALANCE and THERMAL POVER has exceeded the appropriate , be in HOT STANDBY vithin one hour. tr t.e, e e. REACTOR COOLANT SYSTEM PRESSURE 2.1.3 The Reactor Coolant System pressure shall not exceed 2750 psig.- APPLICABILITY: MODES 1, ", 3, 4 and 5. ACTION: 4 MODES 1 and 2 - Vher ever the Reactor Coolant System pressure has exce eded 2750 psig, be in HOT STANDBY vith the Reactor Coolant Systes pressure vithin its limit within nne hour. MODES 3, A Vhenever the Reactor Coolant System pre:sure has and 5 exceeded 2750 psig, reduce the Reactor Coolant System pressure to within its limit within 5 minutes. ca va d Corn tu O. b c.^ I tre gireme,46 of Srecifkdirt. 07.2. 2-1 DAVIS-DESSE. UNIT 1

, Doc $et Number'50-346 I License ilumber t1PF-3 Serial Number 2197 Wl*" 'g h4 fo

1. d Attachment Page 7

// Figure 2.1-2 Reactor Core Safety Limit %RATEDTHERMALP0dR N l 120 4 PUMP LIMIT (33.0,112. (-44.0,112.0) (-49.0,100.0) -.100 (33.0,90.0) (-44.0,90.0) 3 PUMP LIMIT (47.1,87.2) - 80 (-49.0,78.0 ) ( >(47.1,65.2) -60 UNACCEPTABLE UNACCEPTABLE OPERATION ACCEPTABLE OPERATION OPERATION FOR SPECIFIED RC PUMP COMBINATION 40 - 20 t i f I I -60 -40 -20 0 20 40 60 AXIAL POWER IMBALANCE, % Reautred Measured Flow to Ensure Pumos Operating Reactor C:olant Flow, gom Compliance, gom a 380.000 389,500 J 3 293,860 290,957 4 DAVIS-EESSE, UNIT I 2-3 Arae ndmen t No. I1, 16, 33,

45. 61. 80, 91,123 1

- Dockst flumber 50-346 [iicenie Number NPF-3 f -Serial-Number 2197 Attachment THIS PAGE PROVIDED ~" FORINFORMAIl0N ONLY SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS i 2.2 LIMITING SAFETY SYSTEM SETTINGS REACTOR PROTECTION SYSTEM SETPOINTS' O l 2.2.1 The Reactor Protection System instrumentation setpoints shall be set ' consistent with the Trip Setpoint values shown in Table 2.2-1. APPLICABILITY: As shown for each channel in Table 3.3-1. ACTION: With a Tseactor Protection System instrumentation setpoint less conserv-ative than the value shown in the Allowable Values column of Table 2.2-1, declare the channel inoperable and apply the applicable ACTION statement requirement of Specification 3.3.1.1. until the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value. 1 0 AVIS-BESSE, UNIT 1 24

e>mrc w r, mro $ $ " $ S . w E

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s e=

= T' Table 2.2-1 Reactor Protection System Instrumentation Trip Setpoints S$5$ g esc M E$ f8 Functional unit Trip setpoint Allevable values 'u N =; e=o ece A 1. Manual reactor trip Not applicable. Not applicable. "?y wm 2. tilgh flux <104.94% of RATED THERHAL P011ER with (104.94I of RATED Ti!ERHAL POVER vi th four pumps operating leur pumps operatingt <80.6% of RATED THERMAL POVER vith <80.6% of RATED THERHAL POVER vith Ihree pumps operating ~ Ihree pumps operatingt 3. RC high temperature 3618'F f618'FI 4. Flux -- aflux/flo'vIII F rour pump trly setpoint not to tour pump allovable values n ] exceed the llait line of Figure exceed g p 2.2-1. For three pump operat h - ror three pump operation, see Figure /' . 2_- i 2.2-1 ) III 5. RC low pressure 11900.0 psig 11900.0 pal'g* 119 00. 0 psi g *

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6. RC high pressure 52.155 psig $2355.0 psiga $2355.0 psig** wa I)' 7. RC pressure-temperature 1(16.00 T

• F 7957.5) psig 1(16.00T
• P 7957.5) psigt out S

3. iiigh fluxgaber of RC <55;l% of RATED Tl!ERMAL POVER vith <55.1% of RATED THERHAL POVER with g p pumps on one pump operating in each loop one pump operating in each loopt ts. <0.0% of RATF.D THERHAL POVER with <0.0% of RATED THERHAL POVER vith - %g" Ivo pumps operating in one loop and two pumps operating In one loop and .~% no pumps operating In the other loop no pumps operating in the other loopf %.U <0.0% of RATED THERMAL POVER vlth no <0.0% of RATED THERNAL POWER villa no pumps operating or only one pump pumps operating or only one pump op-3 operating eratingI 9. Containment pressure h!gh 34 psig f,4 psigt Pu y % sdroi r.; 4o Pv,,,allu ste v.lue.r,,d s ecu.d % lhif r =3.sk exeuO A U4,1 n.ms44.% ." h C oM OPaunac, wteTS 1+ % C ds ofWW& L.tMG RGOF-&r -{ove ed 'titree. f mp oprs[sor-for fe w **J -th ee f P f> ) AtPOR-otus%-

IN$[?. i ". 23;'. 2 W c Table 2.2-1. (Cont'd) g oe

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T 5E=r M II) Trip may be manually bypassed when RCS pressure 31820 psig by actuating shutdown bypass provided that: E $E$ ?$ a. The high flux trip setpoint is $3I of RATED THERMAL-POVER. y" b. The shutdown bypass high pressure trip setpoint of $1820 psig is imposed. t.> e c. The shutdown bypass is removed when RCS pressure >1820 psig.

• Allovable value for CHAhWEL FUNCTIONAL TEST.
  • Allovable value for CHAhWEL CALIBRATION.

3 B Allovable value for CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION. Y. E E m es c =a - r i "1:p @ 3!ss z ll5 2 cp a m e

== c=z N E clD mac: O ~~~ M E rve CllE c:3

l Docket Number 50-346 License !! umber !!PF-3 Serial fiumber 2197 D d efr_ e d. r<f <o.- w d i l ^ " * ' ' * ' * " ' , 11 s. " pELETEh Figure 2.2-1 Trip Setpoint for Flux -- aFlux/ Flow % RATED THERMAL POWER UNACCEPTABLE UNACCEPTABLE OPERATION-OPERATION 120 Cur shows trip t (-17.0,108.0) (17.0,108.0) point for an approximately j M =+1.00 M =-2.27 3 n or re g [4 PUMP 100 l pump operation (-30.6,94.4) O LIMIT l (283.860 gpm). 8 I The actual set-1 (17.0,80. point will be 80 (-17.0.80.6) calculated by the Reactor Protection l (30.6,77.1) System and will be j 3 PJMP l j directly propor-(- 30.6,67.0 ) EXAMPLE [ tional to the 6 l l actual flow with l l three pumps. IACCEPTABLE OPE T!ON FOR (30.6,49.7) lSPECIFIED RC UMP l lCOMBINATI0' I 40 l I I I l l I I i l -20 1 I l l l l 1 I t 1 t I il i -80 -6 -40 -20 0 20 40-60 80 AXIAL POWER IMBALANCE. *. l' DAVIS-BESSE, UNIT 1 27 Amendment tio. 11 16, 33, 45, 61, 80. 91.123

Ddeket Number 50-346 a License Number NPF-3 Serial Number 2197 Attachment Page 12 2.1 SAFETY LIMITS BASES 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 Reactor Coolant Temperature and pressure have been related to DNB using critical heat flux (CHF) correlations. The local DNB heat flux ratio. DNBR, defined as the ratio of the heat flux that would cause DNS at a particular core location to the local heat flux, is indicative of the margin to DNS. The B&W-2 and BWC CHF correlations have been developed to predict DNB for axially uniform and non-uniform heat flux distributions. The B&W-2 correlation applies to Mark-B fuel and the BWC correlation applies to all B&W fuel with zircaloy spacer grids. The minimum value of the DNBR during steady state operation, nomal operational transients, and anticipated transients is limited to 1.30 (B&W-2) and 1.18 (BWC). The value corresporids to a 95 percent probability at a 95 percent confidence level that DNB will not occur and is chosen as an appropriate margin to DNB for all operating conditions. The curve presented in Figure 2.1-1 represents the conditions at which a minimum DNBR equal to or greater than the correlation limit 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 flow is 389,500 GPN). This curve is based on the klk:9fhot channe factors with potential fuel densification and fuel rod bowing ef fectsM cle.Sig _F - 2.93; F - ' " ; [7 -12 % q g 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 allowable control rod withdrawal, and fom the core DNBR design basis. DAVIS-BESSE, UNIT 1 B 2-1 Amendment No. JJ 33,91 J23.149

Docket'11 umber 50-346 61cenne Number flPF-3 i Serial flumber 2197 Attachment Pap 13 [The CORC OPCHATING LIMITS REPORT includes curves for l protective limits for AXI AL POWER IMBALANCE and for knuclear overpower based on reactor coolant system flow. A protective limit is a cycle-specific limit that ensures that a safety limit is not exceeded by requiring operation within both the cycle design ( perating) limits and the Reactor Protection System $AFETY LIMITS _ tpoints. These protective limit curves reflect BASES 1 j The curves of figuce 2.b2--crc kscd ca more restrictive of two thermal limits and account fo he effects of potential fuel densification and (a s c'euribed s N. c.oxCiorcs nea<, umrs acron.TJ~ 1 potential fuel rod bo es p 1. The ONBR limit produced by a[ nuclear power peaking factortef^ j JF - 2M or the combination of the radial peak, axial peak, g and position of the axial peak that yields no less than i i the DNBR limit. l 2. The combination of radial and axial peak that causes central h ?!.e i kit.is 20.5 ',s/f t.fer ell" - h fuel melting at the hot spo}t. z. , - ~ - n 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 the tuc-curycs cf cigurc".bfgorrespond tn jl the analyzed minimum flow rates with four pumps and three pumps, respectively. T r 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 equal to the DNBR limit is predicted at the maximum possible thermal power for the number of reactor coolant pumps in operation or the local quality at the point of minimum DNBR is equal to the corresponding DNB correlation quality limit (+22% (B&W-2) or +26% (BWC)), whichever condition is more restrictive. r T k h." ds Zoe -l' fueI d<E m du/m3 % op<<.-isy c7eIc. g HrfeJ,L +ke coze cwwrom, unirs waer. ) we r Coff OfEl-ATirJG LIM ITS /LEPN.T~ curve s he profe dwe. kh & u, Au rwee unume ~ J -fu nuele., r c e chr c c o l s n t~ sy r fr.n hlstJ l overfower b.e.s e d on d C_ onvis-r m, unn i .a A= ncent m>. JJ.n.p.sl.n. / / /.1 f.9

Doc e,t 11 umber 50-346 License llumber NPF-3 Serial flumber 2197 Attachment 11113PABE PROVIDED FORINFORMAll0NON SAFETY LIMITS BASES For the curve of BASES Figure 2.1, a pressure-temperature point above and to the left of the curve would result in a Of(BR greater than 1.30 (B&W-2) or 1.18 (BWC) and a local quality'at the point of minimum Di(BR less than +22% (B&W-2) or +26% (BWC) for that particular reactor coolant pump situation. The DNBR curve for 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 the ASME Boiler and Pressure Vessel Code which permits a maximum transient pressure of 110%, 2750 psig, of design pressure. The Reactor Coolant System piping, valves and fittings, are designed to ANSI B 31.7, 1968 Edition. which pemits a maximum transient pressure of 110%, 2750 psig, of component design pressure. The Safety Limit of 2750 psig is therefore consistent with the design criteria and associated code requirements. l The entire Reactor Coolant System is hydrotested at 3125 psig,125% of design pressure, to demonstrate integrity prior to initial operation. l DAVIS-BESSE, UNIT 1 B 2-3 Amendment fio. JJ,U,/5,J/3,149

1 Dob et Number 50-346 h . ' License !! umber flPF-3 -Serial fiumber 2197 THIS PAGE PROVDED Attachment ~ " FORINFORMATION ON

2. 2 LIMITING SAFETY SYSTEM SETTINGS BASES 2.2.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETp0!NTS The reactor protection system instrumentation trip setpoints specified in Table 2.2-1 are the values at which the reactor trips are set for each param-eter. The trip setpoints have been selected to ensure that the reactor core and reactor coolant system are prevented from exceeding their safety limits.

The shutdown bypass provides for bypassing certain functions of the reactor protection system in order to pennit control rod drive tests. Zero power PHYS-ICS TESTS and certain star ao and shutdown procedures. The purpose of the shutdown bypass high press e trip is to prevent normal operation with shut-down bypass activated. Tb high pressure trip setpoint is' lower than the normal low pressure trip segoint so that the reactor must be tripped before the bypass is initiated. The high flux trip setpoint of <5.0% prevents any significant reactor power from being produced. Sufficient natural circula-tion would be available to remove 5.0% of RATED THERMAL POWER if none of the reactor coolant pumps were operating. Manual Reactor Trip The manual reactor trip is a redundant channel to the automatic reactor protec-tion system instrumentation channels and provides manual reactor trip capabil-ity. High Flux A high flux trip at high power level (neutron flux) provides reactor core pro-tection against reactivity excursions which are too rapid to be protected by temperature and pressure protective circuitry. During normal station operation, reactor trip is initiated when the reactor power level reaches 104.94% of rated power. Due to transient overshoot, heat balance, and instrument errors, the maximum actual power at which a trip would be actuated could be ll2%, which was used in the safety analysis. 'l 1 DAVIS-BESSE, UNIT 1 8 2-4 Amendment No. /J, 61 I

DNchet tiumber 50-346 License Number NPF-3 Serial Number 2197 Attachment Page 16 LIMITING SAFETY SYSTEM SETTINGS BASES RC High Temperature The RC high temperature trip 1 61B'F prevents the reactor outlet temperature from exceeding the design limits and acts as a backup trip for all power excursion transients. Flux -- AFlux/ Flow The power level trip setpoint produced by the reactor coolant system flow is based on a flux-to-flow ratio which has een established to accommodate flow decreasing transients from high power where protection is not provided by the high flux / number of reactor coolant pumps on trips. 1 The power level trip setpoint produced by the power-to-flow ratio provides both high power level and low flow protbetion 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 ratio For provides overpower DNB protection for all modes of pump operation. every flow rate there is a maximum permissible power level, and for every power level there is a minimum pemissible low flow' rate. txamples of typical power level and low flow rate combinations for ( pump situations of Table 2.2-1 that would result in a trip ar follows: 1. Trip would occur when four reactor coolant ..ps are operating if power is 108.0% and reactor coolant f1 ate is 100% of full flow rate, or flow rate is 92.59% of f low rate and power level is ~ 100%. 2. Trip would occ en three reactor coolant pumps are operating if f % and reactor coolant flow rate is 74.7% of full flow power is rap r flow rate is 69.44% of full flow rate and power is 75%. g calculated value of 80.68%.trote that the value of 80.6% in Figure 2.2-1 For safety calculations the instrumentation errors for the power level I were used. Full flow rate 4n the don two-exag k: is defined as the l flow calculated by the heat balance at 100% power. At the time of the calibration the RCS flow will be greater than or equal to the value in Table 3.2-2. DAVIS-BESSE. UNIT 1 B 2-5 Amendment No. 16, 13.lb.fd.70, 123

7 Docket Number 50-346 Lic'ense Number UPF-3 Serial Number 2197 A t. tac hmen t Page 17 LIMITING SAFETY SYSTEM SETTINGS BASES The AXIAL POWER IMBALANCE boundaries are established in order to prevent reactor themal limits from being exceeded. These thermal limits are either power peaking kW/ft limits or DNBR limits. The AXIAL POWER IMBALANCE reduces the power level trip produced by a flux-to-flow ratio such that the boundaries of 9 rF2.24 ara nrnduced. 3 Qhc. 4-(90r*e-. IA 5t ( OAE~ OPfhATIsh I.Jbt ITS RE/0LT) RC Pressure - Low, High, and Pressure Temperature The high and low trips are provided to limit the pressure range in which reactor operation is permitted. During a slow reactivity insertion startup accident from low power or a slow reactivity insertion from high power, the RC high pressure setpoint is reached before the high flux trip setpoint. The trip setpoint for RC high pressure, 2355 psig, has been established to maintain the system pressure below the safety limit, 2750 psig, for any design transient. The RC high pressure trip is backed up by the pressurizer code safety valves for RCS over pressure protection, and is therefore set lower than the set pressure for these valves, 1 2525 psig. The RC high pressure trip also backs up the high flux trip. The RC low pressure,1900,0 psig, and RC pressure-temperature (16.00 Tout ' 7957.5) psig, trip setpoints have been established to maintain the DNB ratio greater than or equal to the minimum allowable DNB ratio for those design accidents that result in a pressure reduction. It also prevents reactor operation at pressures below the valid range of DNB, correlation limits, protecting against DNB. High Flux / Number of Reactor Coolant Pumps On ln conjunction with the flux - aflux/ flow trip the high flux / number of reactor coolant pumps on trip prevents the minimum core DNBR from decreasing below the minimum allowable DNB ratio by tripping the reactor due to the loss of reactor coolantpump(s). The pump monitors also restrict the power level for the number of pumps in operation. DAVIS-SESSE, UNIT 1 B 2-6 Amendment No. U,M,f,0,51,149

. DocNet. tiumber 50-346 License fiumber flPF-3 THIS PAGE PROVIDED Serial fiumber 2197

;r FORINFORMATION ONLY LIMITING SAFETY SYSTEM SETTINGS BASES Containment High Pressure

.The Containment High Pressure Trip Setpoint < 4 psig, provides ~ positive assurance that a reactor trip will occur fn 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 Low Pressure trip. i 4 i l i BAVIS-BESSE, UNIT 1 B 2-7

DocIetflumber 50-346

1. Ubicen'se flumber flPF-3 Serial tiumber 2197 Attachment.

Page 19 l l lPOVER DISTRIBUTION LIMITS NUCLEAR EAT TLUX HOT CHANNEL FACTOR - F 1.IMITI?:C CONDITION 70R OPERATION ON OG $ Ml$$ spcE$1id.Ir $$a Loll &** Of6L&& L.IM ITS t Etb At.~r"_ 3.2.2 F shall beA 1=ited by the f ollowing relatio s:) q 2.93-F s j Q P POWER f TED ThlERMAL POWER "" APPLTCABILITY: MODE 1 ACTION: With T exceeding its limit: q I Reduce TERMAL POWER at least 12 for each 1% Tg eiceeds the limit within and flux-a 15 minutes and"rimilarly reduce the high flux crip setpoint a. flux-flow trip setpoint within 4 hours. j Demonstrate through incore mapping that F is within its licit within 24 q hours af ter exceeding the limit or reduce THIRMAL POWER to less than 5% b. of RATID TERMAL POVER vithin the next 2 hours. f the cause of t'he out of limit condition prior to in-Identify and correct creasing THERMAL POWER above the reduced limit required by a or b, above; c. is demonstrated subsequent POWER OPERATION cay proceed provided that TQ through incere mapping to be within its limit 'at's nominal 50% of PJLTED THERMAL POWER prior to exceeding this THERMAL POWER, at a nominal 75% of RATED TERMAL POWER ' prior to exceeding this THERW.AL POWER and within 24 hours af ter attaining 95% or greater RATED THERMAL POWER. l SURVEILLANCE REOUIR_ Eyr5 shall be deter:1ned to be within its li=it by using the incere ,4.2.2.1 FO Idetectors td obtain a power distribution =ap: l I 3/4 2-5 Amendment rio. 45 DAVIS-BESSE, UNIT 1

D5ctet Number 50-346 'u Lice'nse fiumbe r flPF-3 Serial fiumber 2197 . Attachment 'Page 20 THIS PAGE PROVIDED POWER DISTRIBUTION LIMITS _ SURVEILLANCE REQUIREMENTS (Continued) Prior to initial operation above 75 percent of RATED THERFAL a. POWER after each fuel loading, and At least once per 31 Effective Full Power Days. b. The provisions of Specification 4.0.4 are not applicable. c. 4.2.2.2 The measured F of 4.2.2.1 above, shall be increased by 1.4% to account for manufacturing tolerances and further increased by 7.5". n to account for measurement uncertainty. i 'l l l 4 OAVIS-BESSE. UNIT 1 3/4 2-6

~ D'ogket Number 50-346 'o Lic'ense !! umber NPF-3 Serial Number 2197 Attachment Page 21 POWER DISTRIBUTION LIMITS NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR - F H LIMITING CONDITION FOR OPERATION _ Urf*W W l'*'ta sf au he d is tb, CoAE O/FIATMJG x L uote T.S L Ef 0 5 T 3.2.3 F shall be Ikmited by the following relationsh g 3g 1.71 O + 0. @ @ q F g1 5~ M OWER 4) ) where P = THERMAL POWER I / P < 1.0 APPLICABILITY: MODE 1. ACTION: With F"g exceeding its limit: N exceeds the Reduce THERMAL POWER at least 1% for each 1% that F limitwithin15minutesandsimilarlyreducetheHihNFluxTrip a. setpoint and Flux - AFlux - Flow Trip Setpoint within 4 hours. N Demonstrate through in-core mapping that F is within its limit within 24 hours after exceeding the limit h reduce THERMAL b. POWER to less than 5% of RATED THERMAL POWER within the next 2 hours. Identify and correct the cause of the out of limit condition c. prior to increasing THERMAL POWER above the reduced limit required by a or b, aboge; subsequent POWER OPERATION may is demonstrated through in-core proceed provided that F mappingtobewithinit$Hlimit at a nominal 50% of RATED THERMAL POWER prior to exceeding this THERMAL POWER, at a nominal 75% of RATED THERMAL POWER prior to exceeding this THERNAL POWER and within 24 hours after attaining 95% or greater RATED THERMAL POWER. DAVIS-BESSE, UNIT 1 3/4 2-7

Dodket flumber 50-346 > Liceriae ilumber ilPF-3 Serial fiumber 2197 Attachment Page 22 Tills PAGE PRDVIDED POVER DISTRIBUTION LIMITS l SURVEILLANCE REQUIREMEtTTS l 4.2.3.1 E shall be determined to be within its limit by using the incore l detectors to obtain a power distribution maps Prior to operation above 75 percent of RATED THERMAL POVER af ter each a. fuel loading, and b. At least once per 31 Effective Full Power Days. c. The provisions of Specification 4.0.4 are not applicable. 4.2.3.2 The measured F of 4.2.3.1 above, shall be increased by 5% for measurement uncertainty. i DAVIS-BESSE, UNIT 1 3/4 2-8 Amendment No.135

Do$ket Number 50-346 ' " Licetise fiumber flPF-3 Serial Number 2197 Attachment Page 23 THIS PAGE PROVIDED 3/4.2 POWER DISTRIBUTION LIMITS BASES he specifications of this section provide assurance of fuel integrity)during T events ondition I (normal operation) and II (incidents of moderate frequency C (a) maintaining the minimum DNBR in the core > the minimum allowable DNB ratio during normal operation and during short tersi transients, (b) maintaining by: 18.4 kW/f t during normal operation, and the peak linear power density 1 (c) maintaining the peak power density less than the limits given in the bases In addition, the above to specification 2.1 during short term transients. criteria must be met in order to meet the assumptions used for the loss-of-coolant accidents. The power imbalance envelope and the insertion limit curves defined in the CORE OPERATING LIMITS REPORT are based on LOCA analyses which have defined the maximum linear heat rate such that the maximum clad temperature will not exceed the Final Acceptance Criteria of 2200*F following a LOCA. Operation outside of the power imbalance envelope alone does not constitute a situation that would cause the Final Acceptance Criteria to be exceeded should a LOCA occur. The power imbalance envelope represents the boundary of operation limited by the Final Acceptance Criteria only if the control rods are at the insertion limits, as defined in the CORE OPERATING LIMITS REPORT and if the steady-state limit QUADRANT POWER TILT' exists. Additional conservatism is introduced by application of: a. Nuclear uncertainty factors. b. Thermal calibration uncertainty, c. Fuel densification effects. d. Hot rod manufacturing tolerance factors. Potential fuel rod bow effects. e. The ACTION statements which permit limited variations from the basic require-ments are accompanied by additional restrictions which ensures that the original criteria are met. The definitions of the design limit nuclear power peaking factors as used in these specifications are as follows: Nuclear heat flux hot channel factor, is defined as the maximum local Fg fuel rod linear power density divided by the average fuel rod linear power density, assuming nominal fuel pellet and rod dimensions. Amendment No. //,33,/f,1/ A,149 DAVIS-BESSE. UNIT 1 B 3/4 2-1

(Do$k,etfHumber(50 346 ' J License;11 umber NPF-3. L Serla1' Nur.iber 2197 Attachment-Page<24 POWER DISTRIBUTION LIMITS BASES Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio F g of the integral of linear power along the rod on which minimum DNBR occurs to the average rod power. It has been determined by extensive analysis of possible operating power shapes' that the design limits on nuclear power peaking and on minimum DNBR at full power are me p = id:d:s. 4 gg g Ib N 8'* E- ""d 'f'/* N^j OM" oeg nxrMG UM rr.1 1-EtonT' ) N 3 > g' e,.> n >. r '4 ~~' ~ an - Power peaking is not a directly observable quantity and therefore limits h' ve a been established on the bases of the AXIAL POWER IMBALANCE produced by the It has been determined that the above hot channel factor l'im-power peaking. its will be met provided the following conditions are maintained. Control rods in a single group move together with no individual rod in-1. se'rtiondifferingbymorethant,6.5%(indicatedposition)fromthegroup average height. Regulating rod groups are sequenced with overlapping groups as required 2. in Specification 3.1.3.6. The regulating rod insertion limits of Specification 3.1.3.6 are main- '3. tained. AXIAL POWER IMBALANCE limits are maintained. The AXIAL POWER IMBALANCE 4. is a measure of. the difference in power between the top and bottom halves of the core.- Calculations of core average axial peaking factors for many . plants and measurements from operating plants under a' variety of operat-The cor-ing conditions have been correlated with' AXIAL' POWER IMBALANCE. relation shows that the design power shape is not exceeded if the AXIAL POWER.!MBALANCE is maintained between the limits specified in Specifica-tion 3.2.1. 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 al-lowable control rod insertion and are the core DNBR design basis. Therefore, for operation at a fraction of RATED THERMAL POWER, the design limits are When usfng incore detectors to make power distribution maps' to deter-met. mineF4andF3g: Meas shall be increased by 1.4 The measurement of total peaking factor F percent'to account.for manufacturing tole 9ance,s and further increased by a. 7.5 percent to account for measurement error. DAVIS _BESSE. UNIT 1 8 3/4 2-2 Amendment No. JJ. 61

dockyt Numbe r 50-346 .: License Number NPF-3 Serial Humber 2197 Attachment Page 25 TlilS PAGE PROVIDED POWER DISTRIBUTION LIMITS BASES N The measurement of enthalpy rise hot channel factor, Fincreasedby5p b. For Condition 11 events, the core is protected from exceeding the values given in the bases to specification 2.1 locally, and from going below the minimum allowable DNB ratio by automatic protection on power, AXIAL POWER IMBALANCE Only conditions 1 through 3, above, are mandatory pressure and temperature. since the AXIAL POWER IMBALANCE is an explicit input to the reactor protection system. The QUADRANT POWER TILT limit assures that the radial power distribution satisfies the design values used in the power capability analysis. Radial power distribution measurements are made during startup testing and periodically ~ during power operation. The QUADRANT POWER TILT limit at which corrective action is required providesIn DN8 and linear heat generation rate protection with x-y plane power tilts. the event the tilt is not corrected, the margin for uncertainty on Fn is reinstated by reducing the power by 2 percent for each percent of tif t in excess of the Ifmit. 3/4.2.5 DNB PARAMETERS _ 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 limits are consistent with the FSAR the transient and accident analyses. initial assumptions and have been analytically demonstrated adequate to main-tain a minimum DNBR greater than the minimum allowable DNB ratio throughout each analyzed transient. i The 12 hour periodic surveillance of these parameters through instrument read-i nut is sufficient to ensure that the parameters are restored within their The 18 l limits following load changes and other expected transient operation. month periodic measurement of the RCS total flow rate using delta P instrumenta-tion is adequate to detect flow degradation and ensure correlation of the flow indication channels with measured flow such that the indicated percent flow will 1 provide sufficient verification of flow rate on a 12 hour basis. ) l DAVIS-BESSE. UNIT 1 B 3/4 2-3 Amendment No. 33,E 149 l i

DdEhetNumber50-346 License ilumber NPF-3 Serial flumber 2197 Attachment Page 26 THIS PAGE PRDVIDED 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 COOLANT LOOPS AND COOLANT CIRCULATION _ STARTUP AND POWER OPERATION LIMITING CONDITION FOR OPERATION 3.4.1.1 Both reactor coolant loops and both reactor coolant pumps in each loop shall be in operation. APPLICABILITY: MODES I and 2*. ACTION: With one reactor coolant pump not in operation. STARTUP and POWER a. OPERATION may be initiated and may proceed provided THERMAL POWER is restricted to less than 80.6% of RATED THERMAL POWER and within 4 hours the setpoints for the following trips have been reduced in accordance with Specification 2.2.1 for operation with three reactor coolant pumps operating: 1. High Flux 2. Flux-aFlux-Flow SURVEILLANCE REQUIREMENTS The above required reactor coolant loops shall be verified to be l 4.4.1.1.1 In operation and circulating reactor coolant at least once per 12 hours. The Reactor Protection Systen trip setpoints for the instrumenta-l 4.4.1.1.2 tion channels specified in the ACTION statement above shall be verified to be in accordance with Specification 2.2.1 for the applicable number of reactor coolant pumps operating either: Within 4 hours after switching to a three pump ccabination if a. the switch is made while operating, or b. Prior to reactor criticality if the switch is made while shut-down.

• See Special Test Exception 3.10.3.

DAVIS *BESSE, UNIT 1 3/4 4-1 Amendment No. 16.33.38 /5 $0, 123,135

Do(k,etNumber50-346 License Number NPF-3 Serial Number 2197 Attachment Y a FORINOMTiBN dei ADMINISTRATIVE CONTROLS COMPOSITION The Station Review Board (SRB) shall be composed of at least six 6.5.1.2 members of the Davis-Besse onsite management organization. The members shall 4 be as a minimum, managers or individuals reporting directly to managers from each of the following disciplines: plant operations, maintenance, planning, The members shall radiological controls, engineering, and quality assurance. Sections 4.2, 4.4, or 4.6 fu meet the requirements of ANSI N18.1-1971, applicable required experience. The SRB Chairman shall be drawn from the SRB members and designated in writing by the Plant Manager. ALTERNATES All alternate members shall be appointed in writing by the SRB 6.5.1.3 Chairman; however, no more than two alternates shall participate as voting members in SRB activities at any one time. MEETING FREQUENCY _ The SRB shall meet at least once per calendar month and as convened 6.5.1.4 by the SRB Chairman or his designee. QUORUM

6. 5.1. 5 A quorum of the SRB shall consist of the Chairman or his designee and four members including alternates.

RESPONS!BILITIES The Station Review Board shall be responsible for: 6.5.1.6 Review of plant administrative procedures and changes thereto. a. Review of the safety evaluation for 1) procedures, 2) changes b. to procedures, equipment or systems, and 3) tests or experiments completed under the provisions of 10 CFR 50.59, to verify that such actions do not constitute an unreviewed safety question, Review of proposed procedures and changes to procedures and c. equipment determined to involve an unreviewed safety question as defined in 10 CFR 50.59. DAVIS-BESSE, UNIT 1 6-6 Amendment *:o. 12,76,93,103,132,123,

1. 2, 169

Do Iset flumber 50-346 Licetise !! umber llPF-3 Serial flumber 2197 -Attachment 'Page 28 ADMINISTRATIVE CONTROLS d. Review of proposed tests or experiments determined to involve an unreviewed safety question as defined in 10 CFR 50.59. Review of reports of violations of codes, regulations, orders, e. Technical Specifications, or Operating License requirements having nuclear safety significance or reports of abnonnal degradation of systems designed to contain radioactive material. f. Review of all proposed changes to the Technical Specifications or the Operating License. g. Deleted h. Review of reports of significant operating abnormalities or devi-ations from normal and expected perfonnance of plant equipment that affect plant safety.

i. Review of the Industrial Security Plan, the Security Training and Qualification Plan, and the Security Contingency Plan, and changes thereto.

j. Review of the Davis-Besse Emergency Plan and changes thereto.

k. Review of items which may constitute potential nuclear safety hazards as identified during review of facility operations. 1. Investigations or analyses of special subjects as requested by the Company Nuclear Review Board. pgg. 4g m. Review of all REPORTABLE EVENTS. @ l* b A*f* Review of all Safety Limit Violation ReportshSection 6.7). n. f Review of any unplanned, accidental or uncontrolled radioactive o. releases, evaluation of the event, ensurance that remedial action j is identified to prevent recurrence, review of a report covering -l the evaluation and forwarding of the report to the Plant Manager 1 and to the CNRB. i p. Review of the changes to the OFFSITE DOSE CALCULATION MANUAL. q. Review of the changes to the PROCESS CONTROL PROGRAM. Review of the Annual Radiological Environmental Operating Report. c. s. Review of the Semiannual Radioactive Effluent Release Report. T. Review of the Fire Protection Program and changes thereto, i DAVIS-BESSE, UNIT 1 6-7 Amendment No. 27,26,33.32,H3. AL9/, L 74 e

_ Docket flumber 50-346 i-License llumber NPF-3 Serial !! umber 2197 Attachment Page 29 'l ADMINISTRATIVE CONTROLS SAFETY LIMIT V10LAT10t{oe Pto TEunvf Li 'T' 80LMi 4 6.7 6.7.1 The following actions shall be taken in the event a Safety Limit is violated: The facility shall be placed in at least HOT STANOBY within one hour. a. b. The Safety Limit violation shall be reported to the NRC Operations Center by telephone as soon as possible and in all cases within one hour. In addition the Vice President, Nuclear and the CNRB shall be notified within 24 hours. A Safety Limit Violation Report shall be prepared. The report shall c. be reviewed by the SRB. This report shall describe (1) applicable circumstances preceding the violation, (2) effects of the violation upon facility components, systems or structures, and (3) corrective action taken to prevent recurrence. d. The Safety Limit Violation Report shall be submitted to the Comission, the CNRB and the Vice President, Nuclear within 14 days (_A_ of the4 olAtinn_ 1

6. *7. 7

'N.SE A'1~ e% c

6. 8 PROCEDURES AND PROGRAMS 6.8.1 Written procedures shall be established, implemented and maintained covering the activities referenced below:

The applicable procedures recomended in Appendix "A" of Regulatory a. Guide 1.33, November,1972. b. Refueling operations. Surveillance and test activities of safety related equipment. c. d. Industrial Security Plan implementation. e. Davis-Besse Emergency Plan implementation. f. Fire Protection Program implementation. g. The radiological environmental monitoring program. h. The Process Control Program,

i. Of fsite Dose Calculation Manual implementation.

6.8.2 Each procedure of 6.8.1 above, and changes thereto, shall be reviewed and approved prior to implementation as set forth in 6.S.3 above. DAVIS-BESSE, UNIT 1 6-13 Amendment No. 9,12.27,86.93 JD9, 139 )

t 9 ./ ,.' Docket Number 50-346 License Number NPF-3 Serial.liumber 2197 Attachment Page 30 Insert (TS 6.7.2) The following actions shall be taken in the event the Protective 6.7.2 Limit of Specification 2.1.2 is violated: The facility shall be placed in at least HOT STANDBY vithin a. one hour. The Protective Limit violation shall be reported to the NRC b. Operations Center by telephone as soon as possible and in all. cases within one hour. In addition the Vice President, Nuclear and the CNRB shall be notified within 24 hours. .A Protective Limit Violation Report shall be prepared. The c. report shall be reviewed by the SRB. This report shall describe (1) applicable circumstances preceding the violation, (2) effects of the violation upon facility components, systems or structures, and (3) corrective action taken to prevent recurrence. The Protective Limit Violation Report shall be submitted to d. the CNRB and the Vice President, Nuclear within 14 days of the violation. =

Docket flumber 50-346 . Lice'nse Number flPF-3 Serial Number 2197 Attachment Page 31 ADMINISTRATIVE CONTROLS microcuries per gram as a function of time for the duration of the specific activity above the steady-state level; and (5) The time duration when the specific activity of the primary coolant exceeded the radiciodine limit. r2 LL bde Gm MONTHLY OPERATING REPOR g. 2.. l /24%-lor ha-lec6ow $y sfem.$cfpesh-s [

6. 9.1. 6 Routine reports of operating statistics, shutdown experience and challenges to the Pressurizer Pilot Operated Relief Valve (PORV) and the Pressurizer Code Safety Valves shall be submitted on a monthly basis to arrive no later than the 15th of each month following the calendar month covered by the report, as follows: The signed original to the Nuclear Regulatory Commission, Document Control Desk, Washington, D. C. 20555, and one copy each to the Region III Administrator and the Davis-Besse Resident Inspector.

' 3.2. L tdue lear k.4 Flux Ho+ Cha,me.1 PocW FQ CORE OPERATING LIMITS REPORT 3 qqg gg g 6.9.1.7 Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle and any remaining part of a ( reload cycle for the following: I 3.1.1.3c Negative Moderator Temperature Coefficient Limit 3.1.3.6 Regulating Rod Insertion Limits 3.1.3.7 Rod Program 3.1.3.8 Xenon Reactivity ggy j 3.1.3.9 Axial Power Shaping Rod Insertion Limits h L _ 3.2.1 AXIAL POWER IMBALANCE d"g' I'l [3. 2. 4 -QUADRANT POWER TILT V ( The analytical methods useo to determine the core operating limits addressed the individual Technical Specifications shall be those previously reviewed nd approved by the NRC, specifically: 1) BAW-10122A Rev.1. " Normal Operating Controls," May 198 2) BAW-10ll6A, " Assembly Calculations and Fitted Nuc 4 r Data," May 1977 3) BAW-10117P-A, " Babcock & Wilcox Version of User's Manual," l January 1977 4) BAW-10ll8A, " Core Calculati Techniques and Procedures," December 1979 5) BAW-10124A, " 3 - A Three-Dimensional Nodal Code for Calculating Core Reactiv and Power Distributions," August 1976 6) BAW-SA, " Verification of Three-Dimensional FLNiE Code," August \\ ) BAW-10152A, " NOODLE - A Multi-Dimensional Two-Group Reactor Simulator," June 1985 DAVIS-BESSE. UNIT I 6-16 Amendment No. 8.12,93,J04.135.M/,154

- Docketflumber50-346 a License 11 umber flPF-3 Serial !! umber 2197 i Attachment Page 32 ) Insert (TS 6.9.1.7) The analytical methods used to determine the core operating limits those addressed by the individual Technical Specifications shall be: previously reviewed and approved by the NRC, as-described in " Safety Criteria and Methodology for Acceptable Cycle BAW-10179P-A, Reload Analyses", or any other new NRC-approved analytical methods used to determine core operating limits that are not yet referenced in the applicable BAW-10179P-A revision. The applicable BAW-10179P-A revision (the approved revision at the time the reload analyses are performed) The CORE shall.be listed in the CORE OPERATING LIMITS REPORT. OPERATING LIMITS REPORT shall also list any new NRC-approved analytical methods used to determine core operating limits that are not yet referenced in the applicable BAW-10179P-A revision. j -l

Do'cke t. - Numbe r 50-3 4 6 rLicense Number flPF-3 Serial fiumber 2197 Attachment Page 33 i ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT _ (Continued) r' ) BAW-10119 " Power Peaking Nuclear Reli 77 The methodolo gram received NRC approval in the Safety Evaluation i d ry ll, 1990. The core operating limits.shall be' detennined so that all applicable limits (e.g.. fuel thermal-mechanical limits, core thennal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met. The CORE OPERATING LIMITS REPORT, including any mid-cycle revision or supple-ments thereto, shall be provided upon issuance for each reload cycle to the flRC Document Control Desk with copies to the Regional Administrator and Resident Inspector. DAVIS-BESSE, UNIT 1 6-17 Amendment No.144 .}}