ML19317F705

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Amend 11 to License NPF-3,changing Tech Specs 2.1-.2,3.1-2, 4.1-2&5.2-.3 on DNBR Margin to Reflect Increase in Observed RCS Flow Rate
ML19317F705
Person / Time
Site: Davis Besse Cleveland Electric icon.png
Issue date: 06/16/1978
From: Stolz J
Office of Nuclear Reactor Regulation
To:
Shared Package
ML19317F697 List:
References
NUDOCS 8001230598
Download: ML19317F705 (51)
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THE ICLED3 CDISO'1 CGimil MD Tile CLCVELA!ID ELECTRIC ILLITAIJATING CGIPAIY DCCKET N0. 50-346 i D.WIS-BC35C tiUCLEAR PGiER STATIGI, IkIIT NO.1 i Niillri:Dir 10 FACILITY CPEMTING LICOISE Amendment Wo. 11 License rio. NPF-3 1. The tiuclear r>cgulatory Co:anission (the Comission) has foun.i tnat: A. Tne issuance of this license a.'enament complies with tne standarus and requireunts of the /\\tomic Energy Act of 1954, as atendeo (the Act) and tha Coi;missien's rules and regulations set forth in 10 CFR Cnapter I; s. Tt'e facility vill operate in conformity with the license, as amenced, the provisiots of the Act, ano the rules and regulations of the Coomission; C. There is reasonable r surance (i) that the activities autnorizeu oy this amendn.ent can te conducted without endangering tne cealtn and safety of the pi,ulic, and (ii) that such activities vill ce conducted in compliance with the Commission's regulations; D. the issuance of this amendment will not be inimical to tne cocraon defense and ;ecurity or to +r.3 health ma safety of the puolic; anti E. The issuance ot this amenainent is in accordance with 10 CFJs Parc 51 of the Comission's ragulations and all applicable requirements have oeen sati3.ied. F 2. Accordingly, the hended Eacility Omrating License No. JPF-3 is hereby amended by changing the reennical 5cacifications as indicated in the attacninent to this license amenament. Also, the license is anended by deleting License Condition 2.C.(3)(i) to Facility Operating License tiPF-3. 8001230Nh

, 2.C.( 3) Technical Soecifications The Iechnical Specifications contained in App 2ndices A and B, as revised througn teendment No.11 are here' y o incorporated in the license. Toledo Edison Company shall operat? the facility in accordance aith the Tecnnical specifications. 3. this licensa amendnent is effective as of the <iate of issuance.

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\\ . ~ r 7%- D d b */ f ' '# ( Jghn F. '3tolz, Chief Light L'ater Reactors Branch No.1 A hivision of Project ilanagercnt Attacttnent: Changes to the rechnical Soecifications JUN l'8 1978 Date of Issuance: 9 s i 3

n + w .o D 9 ) u\\, a T'IE TOL'.IC CDISCN CGidMY yr \\nM D g \\h 'nIE_CLLVELAJD ELECTAIC ILUJAInATI.G CCP.PMY DCC/Sr SG. 50-345 . CAVI5-EC35E IGM PGJ.R STATIai, UNIT 30.1 AE;r.t'.C:lijo FACILITY CPE.uTI% LIGE".3C Amrrment..o.11 License do. UPE-3 1. The..'uclear segulatory cenussion (the Coacissicn) has Zouna that: A. F.e issuance of tais licenae a encaant c : lies with tne standard; and requirc.:ents or tne Atcaic L:.ergy Act of 1354, as a cnccJ (tne Act) anc the Ocmission's rulas and rejulations sat torta in 10 CFR Cnapter I; 3. Tne facility will ocerate in conformity witn the licence, as renced, the provisions of tne 14t, anc tne rules and regulations of tne Cemissicc; C. There is rossonacic acsurance (1) taat tne activities autnoticaJ by tais armnu..ent can ce conducted vitocut endangering tne n221:n and safety of tne nuclic, and (ii) enat sucn activities will ce ccccccted in cc@liance wit'l tne Cc:saission's regulations; 3. na: ircuance of this caerdaent will not to inimical to tne cc.ar,n cefense and security er to tne nealth and safety of the puclic; and E. The issuance of this acerxiccnt is in accercanco with 10 CF2 Part 51 of the Comisaicn's regulations and all apolicacia requirements have been satisfied. 2. Accordingly, the amended Facility Operating License No. GPF-3 is hereby amerwed by changing the Technical Joecifications as indicated in tha atte .: to this license amndment. Also, the licence is amended by deleting License Condition 2.C.(3)(i) to Pacility Operating License UPF3. ornese summame > 4 oats > , NRC 70aN 318 (N6) NROf 0240 _f us a.eovannuswr emunne orrican sete-semas

. 2.C.(3) Tecnnical Soecifications ':tc Technical Specifications ccratained in A;pendices A and D, as revised througn Amendcent No.11 are herecy incorporcted in ene license. Teleco Ecison Coocany shall operate tne facility in accordance with tne Tec. vical Specificatlor.s. 3. This license acend: rent is effective as of the date of issuance. FOR ':5fS NUCII.M ?LGJLERY CLtMISIGi Original signed by John F. Stolz Jchn F. Stolz, Chief Light Water Peactors eranch.40. 1 Divisien of Project P.anagement .\\ttacrJ.unt: C1.anges to the Technical D specifications Cate et Issuance: JUN 16198 h\\ g \\f nM SEE PREVIOUS YELLOW FOR CONCURRENCES

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ATTACHMENT TO LICENSE AMENDMENT NO. 11 FACILITY OPERATING LICENSE NO. NPF-3 00CKET 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 number and contain vertical lines indicating the area of change. The corresponding overleaf pages are also provided to maintain document completeness. Pages 2-2 2-3 2-5 2-7 2-8 B 2-1 B 2-2 4 B 2-3 B 2-8 3/4 1-16 3/4 1-26 3/4 1-28 3/41-28a,(added) 9 )- 3/4 1-29 3/4 1-29 a thru c (added) 3/4 1-30 'f 1 3/4 1-31 3 3/4 1-32 0 3/4 2-2 3/4 2-2 a (added) 3/4 2-3 3/4 2-3 a (added) 3/4 2-4 3/4 2-4 a (added) 3/4 2-12 3/4 2-14 8 3/4 1-2 B 3/4 2-1 B 3/4 2-2 5-4 l l l l i

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS REACTOR CORE 2.1.1 The combination af the reactor coolant core outlet pressure and outlet temperature shall not exceed the safety limit shown in Figure 2.1-1. APPLICABILITY: MODES 1 and 2. ACTION: Whenever the point defined by the combination of reactor coolant core outlet pressure and outlet temperature has exceeded the safety limit, be in HOT STANDBY within one hour. REACTOR CORE 1 2.1.2 The combination of reactor THERMAL POWER and AXIAL POWER IMBALANCE shall not exceed the safety limit shown in Figure 2.1-2 for the various combinations of two, three and four reactor coolant pump operation. APPLICABILITY: MODE 1. ACTION: Whenever the point defined by the combination of Reactor Coolant System flow, AXIAL POWER IMBALANCE and THERMAL POWER has exceeded the appropriate safety limit, be in HOT STANDBY within one hour. REACTOR COOLANT SYSTEM PRESSURE 2.1.3 The Reactor Coolant System pressure shall not exceed 2750 psig. APPLICABILITY: MODES 1, 2, 3, 4 and 5. ACTION: MODES 1 and 2 - Whenever the Reactor Coolant System pressure has ex-ceeded 2750 psig, be in HOT STANDBY with the Reactor Coolant System pressure within its limit within one hour. MODES 3, 4 - Whenever the Reactor Coolant System pressure has and 5 exceeded 2750 psig, reduce the Reactor Coolant System pressure to within its limit within 5 minutes. DA'/IS-BESSE, UNIT 1 2-1

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m s ~ SAFETY > LIMIT 5 AND LIMITING SAFETY SYSTEM SETTINGS 2.2 LIMITING SAFETY SYSTEM SETTINGS REACTOR PROTECTION SYSTEM SETPOINTS, 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 Reactor 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. DAVIS-BESSE, UNIT 1 2-4

TABl.E 2.2-1 aDg REACTOR PROTECTION SYSTEM INSTRlMENTATION TRIP SETPOINTS FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES NI 1. Manual Reactor Trip Not Applicable Not Applicable Eq 2. High Flux 1 105.5% of RATED TilERMAL POWER < 105.6% of RATED TiiERMAL POWEa s with four pumps operating Eith four pumps operating # 3 m < 80.7% of RATED THEIMAL POWER < 80.8% of RATED THERMAL POWER Eith three pumps operating Eith three pumps cperating# < 53.0% of RATED TilERMAL POWER with < 53.1% of RATED TilERMAL POWER with one pump operating in each loop one pump operating in each loop # 3. RC High Temperature < 619"F < 619.08*F Y 4. Flux - A Flux-Flow (1) Trip setpoint not to Allowable Values not to exceed, exceed the limit line of the limit line of Figure 2.2-2. Figure 2.2-1. III 5. RC Low Pre nure > 1985 psig > 1984.0 psiga > 1976.5 psig** 6. RC liigh Pressure 1 2355 psig 1 2356.0 psig* 1 2363.5 p J** II 7. RC Pressure-Temperature > (16.25 T F - 7873) psig > ( M.25 T F - 7873.64) psig# out ott a a., O.

i E TABLE 2.2-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION TRIP SETP0lNTS a m FUNCTIONAL UNIT TRIP SETPOINT_ ALLOWABLE VALUES U 8. liigh Flux / Number of < 55.0% of RATED TilERMAL POWER < 55.28% of RATED TilERMAL POWER Reactor Coolant Pumps On{j) sith one pump operating in each sith,one pump operating in each ~ loop loop < 0.0% of RATED TilERMAL POWER with < 0.28% of RATED TilERMAL POWER wi ) Two pump operating in one loop and two pumps operating in one loop and no pumps operating in the other loop no pump operating in the other loop, < 0.0% of RATED Tl!ERMAL POWER with < 0.28% of RATED TilERMAL POWER with no pumps operating or m.ly one pump nopumpsgperatingoronlyonepump m operating a operating 4 psig# 9. Containment Pressure liigh 5 4 psig 1 III Trip may be manually bypassed when RCS pressure 1 1820 psig by actuating Shutdown Bypass provided that: a. The liigh Flux Trip Setpoint is < 5% of RATED TilERMAL POWER The Shutdown Bypass is removed when RCS Pressure > 1 1820 psig is imposed, and The Shutdown Bypass liigh Pressure Trip Setpoint of < b. 820 psig. c. Allowable Value for CilANNEL FUNCTIONAL TEST Allowable Value for CllANNEL CAllBRATION

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m 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 cari. meter ~ ducir.9 Operation and therefore THERMAL p0WER and Reactor Coolant Temper-ature and pressure have been related to DNB through the B&W-2 DNB correlation. The DNB correlation 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 locr.i heat flux, is indicative of the margin to DNB. Th.! minimum value cf the DNBR during steady state operation, normal operational transients, and ar.ticipated transients is limited to 1.32. This value corresponds to a 95 percent probability at a 99 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 of 1.32 is predicted for the maximun possible thermal power i 112". when the reactor coolant flow is 387, 200 GFM, which is 110". of l design flow rate for four operating reactor coolant pumps. This curve is based on the following hot channel factors with potential fuel densifi-cation and fuel rod bowing effects: Fg = 2.56; F = 1.71; h=1.50 J .iH The design limit power peaking factors are the most restrictive calculated at full power for the range frem all control rods fully withdrawn to minimum allowable control rod withdrawal, and form the core DNBR design basis. DAVIS-BESSE, UNIT 1 B 2-1 Amendment No. ~l n

SAFETY LIMITS BASES The reactor trip envelope appears to approach the safety limit more closely 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 outlet pressure, providing a more conservative margin to the safety limit. The curves of Figure 2.1-2 are based on the more restrictive of two thernal limits and account for the effects of potential fuel densification and potential fuel rod bow: 1. The 1.32 DNBR limit produced by a nuclear power peaking factor of Fg = 2.56 or the combination of the radial peak, axiel peak and position of the axial peak that yields no less than a 1.32 DNBR. 2. The combination of radial and axial peak that cau s central fuel melting at the hot spot. The limit is 20.4 c:x/ft. 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, 2, and 3 of Figure 2.1-2 correspond to the expected minimum flow rates with four pumps, three pumps, and one pump in each loop, respectively. The curve of Figure 2. -1 is the most restrictive of all possible reactor coolant pump-maximum thermal power combinations shown in BASES Figure 2.1. The curve of BASES Figure 2.1 represent the conditions at which a minimum DNBR of 1.32 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 +22%, whichever condition is more restrictive. This curve includes the potential effects of fuel rod bow and fuel densification, i The DNBR as calculated by the B&W-2 DNB correlation continually increases from point of minimum DNBR, so that the exit DNBR is always higher. Extrapolation of the correlation beyond its published quality range of +22% is justified on the basis of experimental data. DAVIS-BESSE, UNIT 1 B 2-2 Amendment No. Il

2. 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 DNBR greater than 1.32 or a local quality at the point of minimum DNBR less than +22% for that particular reactor coolant pump situation. The 1.32 DNBR curve for four pump operation is more restrictive than any other reactor coolant pump situation because any pressure / temperature point above and to the left of the four pump curve will be above and to the left of the three pump and two pump curves. l 2.1.3 REACTOR CCOLANT 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 permits 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. The entire 9 actor Coolant System is hydrotested at 3125 psig,125% of design pressure, to demonstrate integrity prior to initial operation. DAVIS-BESSE, UNIT 1 B 2-3 Amendment No,11

2.2 LIMITING SAFETY SYSTEM SETTINGS BASES 2.2.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS The Reactor Protection System Instrumentation Trip Setpoint specified in Table 2.2-1 are the values at which the Reactor Trips are set for each parameter. The Trip Setpoints have been selected to ensure that the reactor core and reactor coolant system are prevented from exceeding their safety limits. Operation with a trip setpoint less conservative than its Trip Setpoint but within its specified Allowable Value is accept-able on the basis that each Allowable Value is equal to or less than the drift allowance assumed for each trip in the safety analyses. The Shutdown Bypass provides for bypassing certain functions of the Reactor Protection System in order to permit control rod drive tests, zero power PHYSICS TESTS and certain startup and shutdown procedures. The purpose of the Shutdown Bypass High Pressure trip is to prevent normal operation with Shutdown Bypass activated. This high pressure trip setpoint is lower than the normal low pressure trip setpoint 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. Tufficient natural circulation 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 Protection System instrumentation channels and provides manual reactor trip capability. Hioh Flux A High Flux trip at high power level (neutron flux) provides reactor core protection 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 105.5% of rated power. Due to calibration and instrument errors, the maximum actual power at which a trip would be actuated could be 112%, which was used in the safety analysis. DAVIS-BESSE, UNIT 1 B 2-4

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=.._.._.._rn:--. 540 $80 600 610 620 630 640 REACTOR OUTLET TEMPERATURE, *F RC Flow Power /umes Operating 387,200 GPM 112% Four Pumos j Pressure / Temperature Limits at Maximum ) 1 AlloweWe Power for Mimmum DN8R SASES Fvre 2.1 CAVIS-8 ESSE, UNIT 1 B 2-3 Amencment No.11 l

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 ~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 Low Pressure trip. 3 i i BAVIS-3 ESSE, UNIT l B 2-7

i REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 3. Verifying the boric acid addition system solution tempera-ture when it is the source of borated water. b. At least once per 24 hours by verifying the BWST temperature when it is the source of borated water and the outside air temperature is < 35'F. DAVIS-BESSE, UNIT T 3/4 1-16

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t-2_ - n---t _ _ t _-. _ _ _ _.. 3000 7000 8000 9000 10,000 11,000 12,000 13,000 14,000 CONCENTRATION OF BORIC ACID SOLUTION, ppm B Figure 3,11 Minimum Boric Acid Tank Contained Volume as Function of Stored Boric Acid Concentration-Davis-Besse 1, Cycle 1 DAVIS-BESSE, UNIT 1 3/4 1-16 Amendment NO. 11

pa-REACTIVITY CONTROL SYSTEMS SAFETY R00 INSERTION LIMIT LIMITING CONDITION FOR OPERATION 3.1.3.5 All safety rods shall be fully withdrawn. APPLICABILITY: 1* and 2*f. ACTION: With a maximum of one safety rod not fully withdrawn, except for sur-veillance testing pursuant to Specification 4.1.3.1.2, within one hour either: i a. Fully withdraw the rod or b. Declare the rod to be inoperable and apply Specification 3.1.3.1. i SURVEILLANCE REQUIREMENTS 4.1.3.5 Each safety rod shall be determined to be fully withdrawn: a. Within 15 minutes prior to withdrawal of any regulating rod during an approach to reactor criticality. b. At least once per 12 hours thereafter. 4

  • See Special Test Exception 3.10.1 and 3.10.2.
  1. Wi th K.f >- ' 0*

ef DAVIS-BESSE, UNIT 1 3/4 1-25

REACTIVITY CONTROL SYSTEMS REGULATING R00 INSERTION LIMITS LIMITING CONDITION FOR OPERATION 3.1.3.6 The regulating rod groups shall be limited in physical insertion as shown on Figures 3.1-2a and -2b and 3.1-3a, -3b, -3c, and -3d with a rod group overlap of 2515% between sequential withdrawn groups 5 and 6 for operation up to 145 + 5 EFPD, and between secuential withdrawn groups 5, 6 and 7 after 14T 1 5 EFPD.** APPLICABILITY: MODES 1* and 2*#. ACTION: With the regulating rod groups inserted beyond the above insertion limits, or with any group sequence or overlap outside the specified limits, except for surveillance testing pursuant to Specification 4.1.3.1.2, either: a. Restore the regulating groups to within the limits within 2 hours, or b. Reduce THERMAL POWER to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the rod group oositicn using the above figures within 2 hours, or c. Be in at least HOT STANDBY within 6 hours.

  • See Special Test Exceptions 3.10.1 and 3.10.2.
  1. With K,ff >,1.0
    • For operation between restart after BPRA removal and 145 + 5 EFPD, regulaving rod group 7 shall be fully inserted in the core and shall not have an overlap with group 6.

DAVIS-BESSE, UNIT l 3/4 1-26 Amendment No.11

7. REACTIVITY CONTROL SYSTEMS REGULATING ROD INSERTION LIMITS SURVEILLANCE REQUIREMENTS 4.1.3.6 The position of each regulating group shall be determined to be within the insertion, sequence and overlap limits at least once every 12 hours except when: a. The regulating rod insertion limit alarm is inoperable, then verify the groups to be within the insertion limits at least once per 4 hours; b. The control rod drive sequence alarm is inoperable, then verify the groups to be within the sequence and overlap limits at least once per 4 hours. I 1 DAVIS-BESSE, UNIT 1 3/41 27

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rj= ~. li'-T.E : r =: 0 0 20 40 ' 60 80 100 120 140 160 180 200 220 240 260 280 300 RCD INDEX (% wdl 0 25 50 75 100 t 1 I I f GROUP 7 0 25 50 75 100 I I I I I GROUP 6 0 25 50 75 100 I I f 1 1 GROUP 5 t I ( Figure 3.122 Regulating Rod Group Insertion Limits for Operation to 145:5 EFPD l (Four Pumps) r I 1 DAVIS-8 ESSE, UNIT I 3/4 1-28 Amendment No. 11 d

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) i i i I i / I I 5 r f -l I 5 i i i i i I I i i r g -H c i i i i i i;/i i / (235,50)[_.' [50 e C I i i i t i t i I C i I t I t I if I l "i I y d l t i i i If I I l 5 i i I i i f i i i i I h l I I I/ I I i I ACCEP ABLE i l ! fi I t i OPERATION I I i / t I I I l l l-l 7 (86, 15) j j i ) + i i fi t ,i I i l i -i i ../ 158. 03.1 1 t r i t i f 0 20 40 60 80 100 120 140 160 180 200 220 240 260 220 300 400 INDEX (%, ed) 0 25 50 75 100 1 I 1 1 1 GROUP 7 0 25 50 75 100 I I I 1 1 GROUP 6 0 25 50 75 100 i f I I I GROUP 5 Figure 3.1-2b Regulating Rod Group insertion Limits for Operation After 14525 EFPD (Four Pumps) CAVIS-BESSE, UNIT 1 3/4 1 28a Amendment No.11 i l

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UNACCEPTABLE OPERATION 8

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0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 ROD INCEX (%, wd) 0 25 50 75 100 I f f I I GROUP 7 o 25 50 75 100 I I I I I GROUP 6 0 25 50 75 100 1 1 I I I GROUP 5 Figure 3.1-3a Regulating Rod Group Insertion Limit for Operation to 145+ 5 EFPD (Three Pumps) DAVIS-BESSE, UNIT 1 3/4 l-29 Amendment No. Il m

100 I l 1 l I i l I I i l l l 90 l g l ] i T~~ i i 4 80 (274,77L (300, 77) '"T-g { g i g g i 70 (274, 69.5)

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5 1 1 i i i i if I I i j g i c r. t -{ j i f 1 'l ? i-.l 'i t--^ I l -iACCEPTABLE- ' d (86,12)[l ! OPERATION. j } r i i_ / I i l I i I l ...l. I 1 l - (59, 0) i l~ }- l T i j 'I ~ t i ~i ~l 0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 1 ROD INDEX (% wdl 0 25 50 75 100 1 I I I I O 25 50 75 100 GROUP 7 I I I I l GROUP 6 0 25 50 75 100 I i 1 1 t GROUP 5 Figure 3.13b Regulatmg Rod Group Insertion Limits for Operation After 145:5 EFPD (Three Pumps) DAVIS-BESSE, UNIT 1 3/4 1-29a Anencment No. 11

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mlS REGION _

g 20 =-s =hst;LE: (79. s.51 ! A 1 iACCEPTABLE I I I ,n M* I' .! [. . l.. j j OP_ERATION l l l t. P. - re. ;(55. 0) l' "I f l I 'T l 'j' l l l 0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 l ROD INDEX (% wdl 0 25 50 75 100 l l 1 I I l 1 GROUP 7 0 25 50 75 100 j I I i 1 I l GROUP 6 0 25 50 75 100 1 I I I I GROUP 5 l l Figure 3.1-3c Regulating Rod Group insertion Limits for Operation to 145+ 5 EFPD (Two Pumps) DAVIS-BESSE, UNIT 1 3/4 1-29b Amendment NO.11

e ~ 100 L. I i l l i I i 90 f i l 4 i z 80 f f f i l l 70 a 4 I l I I j. I }. .L... i 3 t I 1 i 60 5 I I I } b (274.52) (300, 52) c l 1 i l l } I I I C U UNACCEPT A8LE OPERATION ! i p' (274.47) g j -+. -. I h l l I I t Z l (252.411 . 40 ,. l -1 I l I i l I I o 2 4- - 4 g ni j-l I i i l ' l l l l I l I I I M (2M.'261 I 3 l l t l lf i i i A

  • I i

i i If I i J+T' -i -i!- r ::-{ (86. 8.5) - if i i t 1 i l .. " :l i:I [i l I i ACCEPTABLE OPERAIION i u(59. 01 V i; l 1 l ! j i I i 1 0 0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 ROD INOEX (% wd1 0 25 50 75 100 t i 1 l 1 GROUP 7 0 25 50 75 100 l 1 I I I GROUP 6 0 25 50 75 100 i f f f I GROUP 5 Figure 3.1-3d Regulating Rod Group insertion Limits for Operation After 145:5 EFPO (Two Pumps) i DAVIS-BESSE, UN:T 1 3/4 1-29c Amendment No.ll I

-~\\ REACTIVITY CONTROL SYSTEMS ROD PROGRAM LIMITING CONDITION FOR OPERATION 3.1.3.7 Each control rod (safety, regulating and APSR) shall be pro-grammed to operate in the core position and rod group specified in Figure 3.1-4. l APPLICABILITY,: MODES 1* and 2*. ACTION: With any control rod not programmed to operate as specified above, be in HOT STANDBY withia l hour. SURVEILLANCE REOUIREMENTS 4.1.3.7 a. Each control rod shall be demonstrated to be programmed to operate in the specified core position and rod group by: 1. Selection and actuation from the control room and verifi-cation of movement of the proper rod as indicated by both the absolute and relative position indicators: a) For all control rods, after the control rod drive patches are locked subsequent to test, reprogramming or maintenance within the panels. b) For specifically affected individual rods, following maintenance, test, reconnection or modification of power or instrumentation cables from the control rod drive control system to the control rod drive. 2. Verifying that each cable that has been disconnected has been properly matched and reconnected to the specified control rod drive. b. At least once each 7 days, verify that the control rod drive patch panels are locked.

  • See Special Test Exceptions 3.10.1 and 3.10.2.

DAVIS-BESSE, UNIT 1 3/4 1-30 Amendment No.ll

i A B 5 7 5 C 3 2 1 4 D 6 8 6 8 6 E 4 4 4 3 F 5 8 7 5 7 8 5 G 1 2 2 2 H 7 6 5 6 5 6 7 K 2 2 2 1 L 5 8 7 5 7 8 5 M 3 4 4 4 ) N 6 8 6 8 6 1 O-4 1 2 3 P 5 7 5 R 1 2 3 4 5 6 7 8 91011121314 15 Bank No. Rods Purpose 1 4 Safety 2 8 Safety 3 4 Safety 4 8 Safety 5 12 Regulating 6 9 Regulating 7 8 Regulating 8 8 APSR Figure 3.1-4 Control Rod Core Location and Group Assignments for Modified Cycle 1 DAVIS-BESSE, UNIT 1 3/4 1-31 Amencment No. 11 l i

9 ~ w INTENTIONALLY LEFT BLANK DAVIS-BESSE, UNIT 1 3/4 1-3E Amendment No,11

3/4.2 POWER DISTRIBUTION LIMITS AXIAL POWER IMBALANCE LIMITING CONDITION FOR OPERATION 3.2.1 AXIAL POWER IMBALANCE shall be maintained within the limits shown on Figures 3.2-1, 3.2-2 and 3.2-3. APPLICABILITY: MODE 1 above 40% of RATED THERMAL POWER.* ACTION: With AXIAL PCWER IMBALANCE exceeding the limits specified above, either: a. Restore the AXIAL POWER IMBALANCE to within its limits within 15 minutes, or b. Be in at least HOT STANDBY within 2 hours. SURVEILLANCE REQUIREMENTS 4.2.1 The AXIAL POWER IMBALANCE shall be determined to be within limits at least once every 12 hours when above 40% of RATED THERMAL POWER except when the AXIAL POWER IMBALANCE alarm is inoperable, then calculate the AXIAL POWER IMBALANCE at least once per hour. l

  • See Special Test Exception 3.10.1 DAVIS-BESSE, UNIT 1 3/4 2-1 l

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y j 18.7,18.5 o** UNACCEPTA8LE _. ACCEPTA8LE _. UNACCEPTA8Lf OPERATION 7*ERATION OPERATION O ~ ~ - ' ' ' - - - ~ ~ - - ~ ~ ~ ~ ' ~ ~ '-*- ~~~ ~ ~ ~ ~ ~ ~ - - ' ~' 40 50 -40 30 -20 10 0 10 20 30 40 50 60 AXtAL POWER IMBALANCE, % l l Figure 12-3a AXIAL POWER IMBALANCE Envelope for Operation to 145+ 5 EFPO (Two Pumps) DAVIS-BESSE, UNIT 1 3/4 2-4 Amendment No.11

e 1 80.____...___. 70 t 52, 60 a j 115.52 7 ,g12,52 m 50 -14.2. 47 T" _- 10.2.47 w g 20.7.42 7 i:118,42 _. c 4 e E UNACCEPTABLE-! Iy OPERATION f 15.7. 31 o i-g g ~ 29.2. 22.5 F E - :I 18.7.18.5 j E ACCEPTA8LE UNACCZPTA8LE 10 _ OPE R ATION _ OPERATION 0 60 50 40 30 20 10 0 10 20 30 40 50 60 AXIAL POWER IMSALANCE. % Figure 3.2-3b AXIAL POWER IMBALANCE Envelope for Operation After 145+ 5 EFPO (Two Pumps) DAVIS-BESSE, UNIT 1 3/4 2-4a Amendment No. 11

~ POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION (Continued) ACTION: (Continued) d. With the QUADRANT POWER TILT determined to exceed the Maximum Limit of Table 3.2-2, reduce THERMAL POWER to 1 15% of RATED THERMAL POWER within 2 hours. SURVEILLANCE REQUIREMENTS 4.2.4 The QUADRANT POWER TILT shall be determined to be within the limits at least once every 7 days during cesration above 15% of RATED THERMAL POWER except when the QUADRANT POWER TILT alarm is inoperable, then the QUADRANT POWER TILT shall be calculated at least once per 12 hours. DAVIS-BESSE, UNIT 1 3/4 2-11

TABLE 3.2-2 QUADRANT POWER TILT LIMITS STEADY STATE TRANSIENT MAXIMUM LIMIT LIMIT LIMIT Measurement Independent QUADRANT POWER TILT 4.92 11.07 20.0 QUADRANT POWER TILT as Measured by: Symmetrical Incore Detector System 3.40 8.90 20.0 Power Range Channels 1.96 6.96 20.0 Minimum Incore Detector System 1.92 4.40 20.0 I DAVIS-BESSE, UNIT 1 3/4 2-12 Amendment No. li

POWER DISTRIBUTION LIMITS ONS PARAMETERS LIMITING CONDITION FOR OPERATION 3.2.5 The following CNB related parameters shall be maintained within the limits shown on Table 3.2-1: a. Reactor Coolant Hot Leg Temperature. b. Reactor Coolant Pressure c. Reactor Coolant Flow Rate APPLICABILITY: MODE 1. ACTION: With any of the above parameters exceeding its limit, restore the param-eter to within its limit within 2 hours or reduce THERMAL POWER to less than 5% of RATED THERMAL PCWER within the next 4 hours. SURVEILLANCE RECUIREMENTS 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 hours. 4.2.5.2 The Reactor Ccolant Systen total ficw rate shall be determined to be within its limit by measurement at least once per 18 months. OAVIS-BESSE, UNIT 1 3/4 2-13 l \\

TABLC 3.2-1 2M DNB MARGIN f/, LIMITS I2 Four Reactor Three Reactor One Reactor Coolaint Pisups Coolant Pianps Coolant Pump c-f*, Pa rameter Operating Operating Operating in Each Loop -e III Reacwr Coolant ilot Leg < 611.1 < 611.I < 611.1 Temperature T *F ) ig Reactor Coolant Pressure, psig.( } > 2062.7 > 2058.7bI > 2091.4 I3) Reactor Coolant Flow Rate, gie > 396,880 > 297,340 > 195,760 l Mu to b Applicable to the loop with 2 Reactor Coolant Pianps Operating. 1 (2) Limit not applicable during either a TilERMAL POWER ramp increase in excess of 5% of RATED TilERMAL POWER per minute or a TilERMAL POWER step increase of greater than 10% F of RAILD TilERMAL POWER. g (3)These flows include a flow rate uricertainty of 2.51. S E b s

3/4.1 REACTIVITY CONTROL SYSTEMS BASES 3/4.1.1 BORATION CONTROL 1 3/4.1.1.1 SHUTDOWN MARGIN A sufficient SHUTDOWN MARGIN ensures that 1) the reactor can be made subcritical from all operating conditions, 2) the reactivity transients associated with postulated accident conditions are controllable within acceptable limits, and 3) the reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition. During Modes 1 and 2 tne SHUTDCWN MARGIN is known to be within limits if all control rods are OPERABLE and withdrawn to or beyond the insertion limits. SHUTDOWN MARGIN recuirements vary througnout core life as a function of fuel depletion, RCS boren concentration and RCS T The most restrictive condition occurs at ECL, with I atnoY8a.d coerating 8 temperature. The SHUTCOWN MARGIN required fl9 consistent with FSAR safety analysis assumotiens. 3 /4.1.1. 2 BORON OILUTION A minimum flew rate of at least 2S00 GPM provides adequate mixing, prevents strat1/icaticr.,nd ensures that reactivity changes will be gradual through the Reactor Coolant System in the core during boren concentration reductions in the Reactor C:olant System. A ficw rate of at least 2800 GPM will circulate an equivalent Reactor Coolant System volume of 12,110 cubic feet in approximately 30 minutes. The reactivity change rate associated with baron concentration reduction will be within the capacility for operator recognition and control. 3/4.1.1. 3 MODERATOR TEMPERATURE COEF:ICIENT The limitations on moderator temperature coefficient (MTC) are provided to ensure that the assumptions used in the accident and transic. analyses remain valid through each fuel cycle. The surveillance require-ment for measurement of the MTC each fuel cycle are adecuate to confirm the MTC value since this coefficient changes slowly due principally to the reduction in RCS boron concentration associated witn fuel burnup. The confirnation that the measured MTC value is within its limit crovides assurance that the coefficient will be maintained witnin acceotable values tnroughcut each fuel cycle. CAVIS-SESSE, UNIT 1 33/41-1

REACTIVITY CONTROL SYSTEMS BASES 1 3/4.1.4 MINIMUM TEMPERATURE FOR CRITICALITY This specification ensures that the reactor will not be made critical with the Reactor Coolant System average temperature less than 525'F, This limitation is required to ensure 1) the moderator temperatura, coeffi-cient is within its analyzed temperature range, 2) the protective instrumentation is within its normal operating range, 3) the pressurizer is capable of being in an OPERABLE status with a steam bubble, and 4) the reactor pressure vessel is above its minimum RT temperature. NOT J 2/4.1.2 BORATION SYSTEMS The boron injection system ensures that negative reactivity control is available during each made of facility operation. The components required to perfonn this function include 1) barated water sources, 2) 1 makeup or DHR pumps, 3) separate flow paths, 4) beric acid pumps, 5) associated heat tracing systems, and 6) an emergency pcwer supply from OPERABLE emergency busses. With the RCS average temoerature above 200*F, a minimum of two separate and redundant boron injection systems are provided to ensure single functional capability in the event an assumed failure renders one of the systems inoperable. Allowable out-of-service periods ensure that minor component repair or corrective action may be completed without undue risk to overall facility safety from injection system failures during the repair pericd. The baration capability of either system is sufficient to provide a SHUTDOWN MARGIN fecrn all cperating conditiens of 1.0%.1k/k after xenon decay and cooldewn to 200*F. The maximum boration capability requirement occurs at EOL from full pcwer equilibrium xencn conditions and requires the equivalent of either 7373 gallens cf cThe ppm berated vatar frem the beric acid stcra6e tanks er 52,726 gallens of 18CC ppm berated water from the berated water stcra6 tank. The requirements for a minimum contained volume of 434,650 gallons of borated water in the borated water storage tank ensures the capa-bility for borating the RCS to the desired level. The specified quantity of borated water is consistent with the ECCS requirements of Specification 3.5.4. Therefore, the larger volume of borated water is specified. With the RCS temperature below 200*F, one injection system is acceptable without single failure consiceration on the basis of the DAVIS-BESSE, UNIT I 3 3/4 1-2 Amendment No. 11

3/4.2 POWER DISTRIBUTION LIMITS BASES The specifications of this section provide assurance of fuel integ-rity during Condition I (Nonnal Operation) and II (Incidents of Moderate Frequency) events by: (a) maintaining the minimum DNBR in the core > l.32 during normal operation and during short tenn transients, (b) maintaining the peak linear power density 1 18.4 kw/ft during normal operation, and (c) maintaining the peak power density 1 20.4 kw/ft during short tenn 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, 3.2-2 and 3.2-3 and the insertion limit curves, Figures 3.1-1 and 3.1-3 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 by Figures 3.1-1 and 3.1-3 and if the steady-state limit GUADRANT POWER TILT exists. Additional conservatism is introducted by application of: a. Nuclear uncertainty factors. b. Thermal calibration uncertainty. c. Fuel densification effects. d. Hot rod manufacturing tolerance factors, e. Potential fuel rod bow effects. The ACTION statements which pennit limited variations from the basic requirements are accompanied by additional restrictions which ensures that the original criteria are met. The definitions of the design limit nuclear pcwer peaking factors as used in these specifications are as follows: F Nuclear Heat Flux Hot Channel Factor, is defined as tne maximum g local fuel rod linear power density divided by tne average fuel rod linear power density, assuming nominal fuel pellet and rod dimensions. DAVIS-BESSE, UNIT 1 B 3/4 2-1 Amendment No.11

r ^ POWER DISTRIBUTION LIMITS BASES F Nuclear Enthalpy Rise Hot Channel Factor, is defined as the g ratio 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 or possible operating power shapes that the design limits on nuclear power peaking and on minimum DNBR at full power are met, provided: 1.71 q1 2.94; F F g1 Power Peaking is not a directly observable quantity and therefore limits have been established on the bases of the AXIAL POWER IMBALANCE produced by the power peaking. It has been detennined that the above hot channel factor limits will be met provided the following conditions are maintained. 1. Control rods in a single group move together with no individual rod insertion differing by more than t 6.57, (indicated position) from the group average height. 2. Regulating rod groups are sequenced with overlapping groups as required in Specification 3.1.3.6. 3. The regulating rod insertion limits of Specification 3.1.3.6 are maintained. 4. AXIAL POWER IMBALANCE limits are maintained. The AXIAL POWER IMBALANCE 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 operating conditions have i been correlated with AXIAL POWER IMBALANCE. The correlation shows that the design power shape is not exceeded if the AXIAL POWER IMBALANCE is maintained between the limits specified in Specification 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 allowable control rod insertion and are the core DNBR design i basis. Therefore, for operatien at a fraction of RATED THERMAL POWER, the design limits are met. Whenusjngincoredetectorstomakepowerdistribu-g and F'.iH tion maps to detennine F eas The measurement of total peaking factor, F , shall be a. increased by 1.4 percent to account for makufacturing toler-ances and further increased by 7.5 parcent to account for measurement error. DAVIS-BESSE, UNIT 1 B 3/4 2-2 Amendment No.ll l i ( 1

I, l k 2' ?l j t w 3 . &l t * --+'- j - 4. , L., ' i i = [4 d t ,a -l l I ,g m ^} i: ', ' Q...;D; \\ ~ .I ,h2 '[Qfi.(, T .% ei 1 ' ' lq c 1

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a n s M qW: e Mg. /t.1P'.. 'L3 9) i ~ ^- t g ' 1el ,- Ki;:Q; ' I' k A.,%. 1 E .i f l, )}, 'g(h %y n 0 $2 -. g -.p g ~- 1 .9 as j u , LW hj. 'T 2 4 ~ ,a z+ . O h. \\., 'I &,'&M l 2,4b.....*1 ~ t s y---{ 1 g @L R - [_ Ts _,c-OAVIS - BESSE NUCLEAR POWER STATON LOW POPULATON ZONE FIGURE 5. I-2 DAVIS-BESSE, UNIT 1

5. '

~ s DESIGN FEATURES DESIGN PRESSURE AND TEMPERATURE 5.2.2 The reactor containment building is designed and shall be maintained for a maximum internal pressure of 40 psig and a temperature of 264*F. 5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The reactor core shall contain 177 fuel assemblies with each fuel assembly containing 208 fuel rods clad with Zircaloy -4. Each fuel rod shall have a nominal active fuel length of 144 inches and contain a maximum total weight of 2500 grams uranium. The initial core loading shall have a maximum enrichment of 3.0 weight percent U-235. Reload fuel shall be similar in physical design to the initial core loading and shall have a maximum enrichment of 3.5 weight percent U-235. CONTROL RODS 5.3.2 The reactor core shall contain 53 safety and regulating and 8 axial power shaping (APSR) control rods. The safety and regulating control rods shall contain a nominal 134 inches of absorber manterial. The APSR's shall contain a nominal 36 inches of absorber material at their lower ends. The nominal values of absorber material shall be 80 percent silver, 15 percent indium and 5 percent cadmium. All control rods shall be clad with stainless steel tubing. l l l i DAVIS-BESSE, UNIT 1 5-4 Aoendment No.ll .}}

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