ML19225C932
| ML19225C932 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 07/03/1979 |
| From: | Gammill W Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19225C933 | List: |
| References | |
| NUDOCS 7908030307 | |
| Download: ML19225C932 (37) | |
Text
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f U;.; TED STATES NJCLEAR REGULATORY COMMisslON
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E W ASHINGTON. D. C 20555
.7 FLORIDA PO'AER CCRPORAIIOil CITY OF ALACHUA CITY OF BUihNELL CIT ( OF GAINES7ILLE CITY OF KISSIMPEE CITY Ci LEESBURG-CITY OF NE" SMYRNA SEACH AND UTILI?TES CCWISSION. CITY CF NEW SMYRNA SEACH CITY OF GCALA ORLANCO UTILITIES COMMISSION ND CITY OF CRLANCO SESRING UTILITIES CCidil55 ION SE'iINGLE ELECTRIC COOPE?AT"lE. INC-CITY OF TALOMASSEE CCCKET NO. EO-202 CRYSTAL RIVER UNIT 3 NUCLEAR ';ENERA'NG PL;NT AMENOMENT TO FACILI? CPERATING LI:ENSE Amendment No. 19 License No. U?2-72 1.
The Nuclear Regulatcry Ccmission (the Ccmissicr) has found that:
A.
The applications fer mendment by Florida Power Corporation, et al (the licensees) datr:d February 28, 1979 and March 15,1979, as supplemented May 15, 1979, corply with the standards and require-ments of the Atamic Energy Act of 1954, as amended (the Act), and the Cor: mission's rules and regulations set forth in 10 CFR Chapter I; 3.
The fac'lity nii? cparate in ccnfornity with :.9e rolications,
-he ;mvisicns cf One Act, ir.d 7.e *u'es and regCations of
- he Cem.issien; C.
There is reascnable assurance (i) that ?.e acitvities.:tthorized
- v :nis amencment can be conductec withcut encangering the teci:-
N d.d safety cf the cublic, and (ii) 03a: such ac f vities will ca ccaducted in :0moliance 41 n the Comiss:cn N regulatiens; C.
The issanca of :nis amend: en will not ce -i'mcai c :na comen iefensa inc :e uri:y er :c :ne heal:n and sa?? y of ce oclic; anc E.
The issuance of :n:
mer.cmen; is in acc:rnnce witn 10 CF: 7:r:
51 of tne Ocmissicr.'s requia i:n: and Ali 1;;iicaole mquiremen :
n ze been satisfiec.
7908030307 si20 308
. 2.
Accordingly, the license is amendec' by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-72 is hereby amended to read as folicws:
Technical Soecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.19, are hereby incorporated in the license.
Florida Power Corporation shall operate the facility in accordance with the Technical Specifications.
3.
This l' cense amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY CCMMISSI0ti A h b f.
rCv-William P. Gannill, Acting Assistant Director for Orarating Reac+wr Projects Div.sion of Operating Reactors
Attachment:
Changes to the Technical Specification-Date of Issuance:
July 3,1979 420 309
ATTACHMENT TO LICENSE AMENCMENT NO.19 FACILITY OPERATING LICENSE NO. OPR-72 DOCKET NO. 50-302 Replace the follewing pages of the Appendix "A" Technical Specifications with the' enclosed pages. 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.
Paces 2-2 2-3 2-5 82-1 B 2-6 8 2-7 3 2-8 (added) 3/4 1-27 3/4 1-28 3/4 1-29 3/4 1-30 3/4 1-33 3/4 1-34 3/4 1-35 3/4 1-36 3/4 1-37 3/4 1-38 3/4 1-39 (added) 3/4 2-2 3/4 2-3 3/42-4 3/4 2-11 3/4 2-13 3 3/4 2-2 8 3/4 2-3 420 3:0
2.0 SAFETY LIMITS AND LDi1 TING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS REACTOR CORE 2.1.1 The combination of the reactor coolant core outlet pressure and outlet temperature shall not exceed the safety limit shown in Figure 2.1 -1.
APPLICABILITY:.10 DES 1 and 2.
ACTION:
Whenever the point defined by the canbination of reactor coolant core outlet pressure and outlet temperature has exceeded ';he safety limit, be in HOT STANCBY within one hour.
REACTOR CORE 2.1. 2 The ccmbination of reactor THERMAL POWER and AXIAL POWER IMBALANCE sb=ll not exceed the safety limit shown in Figure 2.1-2 for the various-combinations of three and four reactor coolant pump operation.
APPLICABILITY : MODE 1.
ACTION:
Whenever the point defined by the combination of Reactor C7olant System flow, AXIAL POWER IMBALANCE and THERMAL POWER has exceeded the appropriate safety limit, be in HOT STANDBY within one hour.
REACTCR COOLANT SYSTEM PRESSURE 2.1. 3 The Reactor Coolant System pressure shall not exceed 2750 psig.
APPLICABILITY : MODES 1, 2, 3, 4 and S.
ACTION:
MCDES 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.
MCDES 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.
CRYSTAL RIVER - UNIT 3 2-1 Amendment No. 17 420 311
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CUAVE AE ACTOR COOLANT FLCW (L3/HR) 1 139.86 a 106 2
104.47 x 106 Figurs 2.12 Reactor Core Safety Lim:t
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CRYSTAL RIVER - UNIT 3 2-3 Amendment No. JE; g enc 4cd al3
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SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.2 LIMITING SAFETY SYSTEM SETTINGS REACTOR PROTECTION SYSTFM SETPOINTS 2.2.1 The Reactor Protection System instrumentation setpoints shall be set consistent with the Trip Setpoint va' es shown in Table 2.2-1.
APPLICABILITY: As shown for each channel in Table 3.1-1.
ACT7 0,N,:
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 u.e 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.
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REACTOR PROTECTION SYSTEtt INSTRUMENTATION TRIP SETPOINTS FUNCTION UNIT TRIP SETPOINT ALLOWABLE VALUES 9
e 8.
Reactor Containment Vessel Ei Pressure liigh
< 4 psig 3,4 psig T4 w
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a (1) Trip may be manually bypassed when RCS prassure 5,1720 psig by actuating Shutdown Bypass provided that:
a.
The Nuclear Overpower Trip Setpoint is < 5% of Rt.TED TilERMAL POWER b.
The Shutdown Bypass RCS Pressuie - lifeli Trip Setpoint of 3,1720 psig is imposed, and
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c.
The Shuidown Bypass is removed wher. RCS Pressure > 1800 psig.
r-a l CJD V4 CN
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 c! adding perforation which would result in the release of fission proc'ucts to the reactor coolant. Overheating of 'he 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 tr.e 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 r.ucleate 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 Ter.tper-ature and Pressure have been related to DNB through the BAW-2 DNB correla-ti on. The DNB correlation has been developed to predict the DNB flux and the location of DNB for axially unifann and non-unifonn heat flux distributions. The local DNB heat flux ratio, DNBR, defined as the ratic of the heat flux that would cause DNB at a particular core location to the local heat flux, is indicative of the trargin to DNB.
The minimum value of the DNBR during steady state operation, normal operational transients, and anticipated transients is li'nited to 1.30.
This value corresponds 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 foi-all operating conditions.
The curve presented in Figure 2.1-1 represents the conditions at which a minimum DNBR of 1.30 is predicted for the maximum pcssible thennal power 112% when the reactor coolant flow is 139.86 x 106 lbs/hr, which is 106.5% of the design ricw rate for four operating reactor coolant pumos. This curve is based on the following nucleu pcwer peaking factors with potential fuel densification effects:
F')g=1.71; Ff=1.:0 F = 2.57; The design limit power peaking factors are the most restrictive calcu-lated at full power for the range from all cont.ol rods fully withdrawn to minimum allovat:le control rod withd: awal, and forin the core DNBR design basis.
1;) Ny. 76,1 9 CRYSTAL RIVER - UNIT 3 B 2-1 Amendm
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SAFETY LIMITS BASES The reactor trip envelope appears t: approach the safety limit more closely than it actually does because the reactor trip pressures are measured at a loctrian where the indicated pressure is about 30 psi less than core outlet pressure, oroviding a more conservative margin to the safety limit.
The curves of Figure 2.1-? are based a the more restrictive of two thermal limits and account for the effects of potential fuel densifica-tion and potential fuel rod bow:
1.
The 1.30 DNBR limit produced by a nuclear power peaking factor of F = 2.57 or the combination of the radial peak, axial peak and position of tne axial peak that yields no less than a 1.30 DNBR.
2.
The combination of radial and axial peak that causes centrai fuel melting at the hot soo:. The limit is 19.7 kw/ft.
Power peaking is not a directly observable quantity and therefore limtts have been established on the basis of the reactor power imbalance produced by the power peaking.
The specified flow rates for curves 1 and 2 of Figure 2.1-2 cor-respond to the expected minimum flow rates with four pumos and three pumps, respectively.
The curve of Figure 2.1-1 is the most restrictive of all possible reactor coolant pump-maximum t hrmal pcw0r ccmbinations snown in BASES Figure 2.1.
The curves of BASES Figure 2.1 represent the conditians at which a minimum t BR of 1.30 is predicted it the maximum possible th'ermal power for the numbv of reactor coulant pumps in operation or the local pality at the point of minimum DNBR is equal to 22%,
whithever condition is more restrictive.
These curves include the potential effects of fuel r bcw and fuel densification,.
The DNBR as calculated by the BAW-2 C!<B correlation continually increases from point of minimum DNBR, so that the exit DNBR is alwcys higher. Extrapolation of the correlation beycnd its published quality range of 22% is justified on the basis of experimental data.
CRYSTAL RIVER - UNIT 3 8 2-2 Amendment No. }$,17 9L.d 3W w
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LIMITING SAFETY SYSTEM SETTINGS BASES RCS Outlet Temoeratura - Hich The RCS Oct' it Temperature High trip < 619 F prevents the reretor outlet temperature from exceeding the design limits and acts as a backup trip for all power e::.cursion transients.
Nuclear Overocwer :: ed on RCS Flow and AXIAL POWER IMBALANCE The power level trip setpoint produced by the reactor coolant system flow is based on a flux-to-flow ratio which has been established to accannodate flow decreasing transients frem high power.
The power level trip setpoint produced by the power-to-flow ratio provides both high power level and low flow protection in the event the reactor power level increases or the reactor coolant flow rate decreases.
The power level setpoint produced by t:1e power-tc-flow ratio provides overpcwer DNE protection for all modes of pump operation. For every flow rate there is a maximum permissible power level, and for every power level there is a minimum permissible low flow rate. Typical power level ar.d icw flow rate canbinations for the pump situations of Table 2.2-1 are as folicws:
1.
Trip would occur when four reactor coolant pumps are operating if power is > 104.?% and reactor flow rate is 100%, or flow rate is < 95!3% and power level is 100%.
2.
Trip would occur when three reactor coolant pumps are operating if power is > 77.9% and reactor flow rate is 74.7%, or flow rate i s < 71.9% a n'd power is 75%.
For safety calculations the maximum calibration and instrumentation errors for the power level cere used.
CRYSTAL RIVER - UNIT 3 B 2-5 Amencmqdt-No. 76 17 rc V 74a
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a LIMITING SAFETY SYSTEM SETTINGS BASES The AXIAL POWER IMBALANCE boundaries are established in order to prevent reactor thermal limits from being exceeded.
These thermal limits are either pcwer peaking kw/ft lin4its or DNBR limits. The AXIAL POWER IMBALANCE reduces the power level trip produced by the flux-to-flow ratio such that the boundaries of Figure 2.2-1 are produced.
The flux-to-flow ratio redu es the power level trip and associated reactor power-reactor power-imbalance boundaries by 1.043% for a 1%
flow reduction.
RCS Pressure - Low, Hioh and Variable Low The High and Low trips are provided to limit tne pressure range in which reactor operation is permitted.
During a slow reactivity insertion startup accident fran low power or a riew reactivity insertion from high power, the RCS Pressure-High setpoint is reached before the Nuclear Overpower Trip Setpoint.
The trip setpoint for RCS Pressure-High, 2300 psig, has been established to maintain the system pressure below the safety limit, 2750 psig, for any design transient. The RC5 Pres;ure-High trio 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, 2500 psig. The RCS Pressure-High trip also backs up the Nuclear Overpower trip.
The RCS Pres' sura-Low,1800 psig, and RP.S Pressure-Variable Low, maintainY.F-5209.2)psig,TripSetpointshavebeenestablishedtehe DNB ratio (11.80 7 0
Tccidents that result in a pressure reduction.
It also prevents reactor operation at pressures belcw the valid range of DNB correlation limits, protecting against UNB.
Due to the calibration and instrumentation errors, the safety analysis used a RCS Pressure-Variable Lcw Trip Setpoint of (11.80 l'out F-5249.2) psig.
f,2g 320 CRYSTAL RIVER - UNIT 3 B 2-6 Amendment No.16,19 w
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LIMITING SAFETY SYSTEM SETTINGS BASES Reactor Containment Vessel Pressure - Hich The Reactor Containment Vessel Pressure-High 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 RCS Pressure - Low tri p.
l CRYSTAL RIVER - UNIT 3 B 2-7 Amendment'jp jJ6, 7g 7
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1600 580 600 620 640 660 gM Rasetor Outl'at Terp, 'T REACTOR COOLANT FLOW PUMPS OPERATING CURVE FLOV (15/hH POWER (%RTP)
(TYPE OF LIMIT) 139.86 x 106 (106.7%)
117.3%
4 Pumps (DNBR) 2 104.47 x 106,(79.7%)
90.5%
3 Pumps (DNBR)
PRESSURE /TEMPERATLRE LIMITS AT MAXIMUM ALLOWABLE POWER FOR MINIMUM DNBR BASES FIGURE 2 1 CRYSTAL RIVER - UNIT 3 B 2-8 Amendment Eo. 19
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CRYSTAL R WER - UNIT 3 3/4 1-27 Amendment No. I, 2, N, I 9
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GROUP 6 Figure 3.1-2 Regulating Rod Group Insertion Limits For 4 Pump Operation After 233 - 10 " /PD CRYST,'. RIVER - UNIT 3 3/4 1-25 Aner,cment No. 76,19 420 324 e
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GROUP 6 Figure 3.1-4 Regulating Rod Group Insertion Limits For 3 Purnp Operarmn After 233 : 10 EFPD CRYSTAL RIVER - UNIT 3 3/4 1-30 Amendment No. J6,;3 420 326
l REACTIVITY CONTROL SYSTEMS RCD PROGRA.'4 LIMITING CONDITION FOR OPERATION 3.1.3.7 Each control rod (safety, regulating and APSR) shall be pro-grammed to oper$te in the core ocsition and rod group specified in Figure 3.1-7.
APPLICABILITY: MODES 1* and 2*.
ACTION:
With any control rod not orogramn.e4 tc cpc. ate as specified above, be in HOT STANDBY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
SURVEILLANCE REQUIREuENTS 4.1.3.7 a.
Each cortrol rod shall be demonstrated to be programmad to operate in the specified core position and rod group by:
1.
Selection and actuation frcm 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 controi.ods, after the control rod drive patchs are locked subsequent to test, reprogrammir.;
or maintenance within the panels.
b)
For specifically affe;ted individual rods, following maintenance, test, reconnection or modification of power or instrumentation cables from the control rod drive centrol system to the control rod drive.
2.
Verifyiel 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 raten panels are locked.
" Sea Special Tes; Exceptions 3.10.1 and 3.10.2.
j 90 jfl!)f**"' "
' 8 CRYSTAL RIVER - UNIT 3 3/4 1-33 j
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GROUP NUMBER OF ROOS FUNCTION 1
8 SAFETY 2
8 SAFETY 3
12 F *.~. TY 4
9 SAFI TY 5
8 CONTROL 6
8 CONTROL 7
8 CONTROL 8
8 APSRs TOTAL 69 Figure 3.17 Control Rod Locations And Group Assagnments CRYSTAL RIVER - UNIT 3 3/4 1-34 Amendment No. 1 9 i
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CRYSTAL RIVER - UNIT 3 3/4 1-35 Amendment No. ! 9 A20 329
' ~ ~
~ ~ ~ ~
- ~ ~ ~
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REACTIVITY CONTROL SYSTEMS XENON REACTIVITY LIMITING CnNDITI0ri FOR OPERATION 3.1.3.8 THERMAL POWER shall not be increased above the power level cutof' specified in Figures 3.1-1 and 3.1-2 unless xenon reactivity is within 10 percent of the equilibrium value for RATED THERMAL POWER and is approaching stability.
APPLICABILITY: MODE 1.
ACTION:
With the requirements of the above specification not satisfied, reduce THERMAL POWER to less than or equal to the power level cutoff within 15 minutes.
SURVEILLANCE REQUIREMENTS 4.1.3.8 Xenon reactivity shall be determined to be within 10% of the equilibrium value for RATED THERMAL POWER and to be approaching stability prior to increasing THERMAL POWER above the pcwer level cutoff.
'on 330 4 L. v CRYSTAL RIVER - UNIT 3 3/4 1-36 Amendment No.'9 i
e em n
REACTIVITY CONTROL SYSTEFS AXIAL POWER SHAPING RCD INSERTION LIMITS LIMITING CONDITION FOR OPERATICN 3.1.3.9 The axial power shaoing red group shall be limited in physical insertion as shown on Figures 3.1-9 and 3.1-10.
APPLICABILITY: MODES 1 ar.d 2*.
ACTION:
With the axial power shaping rod group outside the above insertion limits, either:
a.
Restore the axial power shaping rod group to within the limits within 2 hou s, or b.
Reduce THERMAL POWER to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the rod group po-sition using the above figure within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, or c.
Be in at least HOT STANDBY within 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />..
SURVEILLANCE REOUIREMENTS 4.1.3.9 The position of the axial power shaping rod group shall be determined to be within the insertion limits at least once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
- W1:n x,ff > 1.0.
/i20 33l l CRYSTAL RIVER - UNIT 3 3/4 1-37 knencment No. 16, 3
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ROD POSITION, % WITHDRAWN Figure 3.1-10 Axial Power Shaping 9ad Group Insertion Umits After 233 :: 10 EFPD
?mu0 333 CRYSUL RIVER - UNIT 3 3/4 1-39 Amencment No.Ig
- M
=
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3/4.2 POWER DISTRIBUTION LIMITS AXIAL POWER IMBALANCE LIMITING CONDITION FOR OPERATION 3.2.1 AXIAL POWER IM3ALANCE shall be maintained within the limits shown on Figures 3.2-1 and 3.2-2.
APPLICABILITY: MODE I above 40% of RATED THERMAL POWER.*
ACTION:
With AXIAL PCWER IMBALANCE exceeding the limits specified above, either:
a.
Restore tr.a AXIAL POWER IMBALANCE to within its limits within 15 minutes, or b.
Se in at least HOT STANDBY within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
SURVEILLANCE REOUIREMENTS 4.2.1 The AXIAL POWER IMBALANCE shall be determined to be within limits in each core quadrant at least once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when above 40% of RATED THERMAL POWER except when an AXIAL POWER IMBALANCE monitor is inoperable, tnen calculate the AXIAL PCWrR IMBALANCE in each core quadrant with an inoperable monitor at least oncu per Dnur.
- See Special Test Exception 3.10.1.
CRYSTAL RIVER - UNIT 3 3/4 2-1
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Figure 3.2-1 Axial Power imbalsnce Envelope For Operation From 0 EFPD To 233 : 10 EFPO 420 335 CRYSTAL RIVER - UNIT 3 3/4 2-2 Amendment No. J, 2, 76,19 e
j POWER. % OF RATED THERMAL POWER m q._ :.. 1
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Figure 3.2-2 Axial Power Imbalance Envelope Foi Ooeration After 3 : 10 EFPD J
CRYSTAL RIVER - UNIT 3 1/4 2-3 Amendment No. 16,l9 1
330
~.
=
POWER DISTRIBUTION LI-ITS NUCLEAR HEAT FLUX HOT CHANNEL FACTOR - Fq LIMITING CONDITION FOR OPERATION 3.2.2 F shall be limited by the following relationships:
g Fq <_ 3.08 THERMAL POWER
~
where P = RATED THERMAL POWER and P < l.0.
APPLICABILITY: MODE 1.
ACTION:
With F exceeding its limit:
g a.
Reduce THERFAL POWER at least 1% for each 1% F exceeds the n
limit within 15 minutes and similarly reduce tMe Nticlear Overpower Trip Setpoint and Nuclear Overpower based en RCS Flow and AXIAL POWER IMSALANCE Trip Setpoint within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, b.
Demonstrate *.hrough in-core mapping that F is within its limit 0
within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after exceeding the limit or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
c.
Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER above the reduced limit re-quired by a or b, above; subsequent POWER OPERATION may proceed provided that Fn is demonstrated through in-cora map-ping to be within its 1Ymit 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 THERMAL POWER and within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after attaining 95% or greater RATED THERMAL POWER.
3URVEILLANCE RECUIREMENTS 4.2.2.1 F shall be determined ta be within its limit by using the incore detectors ko obtain a power distribution map:
,,7 D
G0 lCRYSTA RIVER - UNIT 3 3/4 2-4 Amendment No. 13 4
e
TABLE 3.2-2 QUADRANT POWER TILT LIMITS STEADY STATE TRANSIENT MAXIMllM LIMIT
_ LIMIT LIMIT QUADRANT POWER TILT as Measured by:
Symetrical Incore Detector System 3.46 8.96
- 20. P -
Pcwer Range Channels 1.96 6.96 20.0 Minimum Incore Detector System 1.90 4.40 20.0 420 337 CRYSTAL PIVER - UNIT 3 3/4 2-11 Amendrr. eat No. H, 1 9
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POWER DISTRIBUTION LIMITS DNB PARAMETERS LIMITING CONDITION FOR OPERATION 3.2.5 The following DNB related parameters shall be maintained within the limits shown on Table 3.2-1:
a.
Reactor Coolant Hot Leg Temperature b.
Reactor Coolant Pressure c.
Reactor Coolant Flow Rate APPLICABILITY: MODE 1.
ACTION:
With any of the abcve Jaraiaeters, exceeding its limit, restore the param-eter to within its limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduca THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.2.5.1 Each of the~ parameters of Table 3.2-1 shall be verified to be within their limits at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
4.2.5.2 The Reactor Coolant System total flow rate shall be determined to be within its limit by.neasurement at least once per 18 months.
?Ou' bb qs CRYSTAL RIVER - UNIT 3 3/4 2-12
-.-+
i TABLE 3.2-1 c'$
j "I
DNB MARGIN N
LIMITS u2 0
Four Reactor Three Reactor Coolant Pumps Coolant Pumps Parameter Operating Operating c-5 Reactor Coolant flot leg U}
Temperature, I *f
< 604.6
< 604.6 g
II III l
Reactor Coolant Pressure, psig
> 2067.6
> 2057.2 6
f Reactor Coolant Flow Rate, Ib/hr
> 139.86 x 10
> 104.47 x 10 Rs.
'?
U j
I E
l UIApplicable to !!.e loop with 2 Reactor Coolant Pianps Operating.
R m
5 I2I Limit not applicable during either a THERMAL POWER ramp increase in excess of 5% of RATED TilERMAL POWER per minute or a TilERHAL POWER step increase z
s P
of greater than 10% of RATED TilERMAL POWER.
m c3 L~
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3/4.2 p0WER DISTRIBUTION LIMITS BASES The specifications of this section provide assurant.e of fuel integrity during Condition I (Normal Operation) and II (Incidents of Moderate Frequency) events by:
(a) maintaining the minimum DNBR in the core
> l.30 during normal operation and during short term transients, (b)
_maintai...'ng the peak linear power density 1 18.0 kw/ft during r,cnnal operation, and (c) maintaining the peak power density 1 19.7 kw/ft during short term transients.
In addition, the abcve criteria must be met in order to meet the assumptions used for the loss-of-coolant accidents.
The power-imbalance envelope defined iri Figures 3.2-1 and the insertion limit curves, Figures 3.1-1, 3.1-2, 3.1-3, 3.1-4 and 3.1-9, are based on LOCA analyses which have defined the maxic:um 'inear 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 opera-tion limited by the Final Acceptance Criteria only if the control rods are at the insertion limits, as Jefined by Figures 3.1-1, 3.1-2, 3.1-3, 3.1-4, and 3.1-9, 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.
The conservative application of the above peak'1g augmentation factors compensatts for the potential peaking pena'ty due i Nel rod bow.
The ACTION statements which pennit limited variations from tne basic requirements are accompanied by additional restrictions which ensures that the original criteria are met.
The definitions of the design limit nuclear power peaking factorr, as used in these specifications are as follows:
F Nuclear Heat Flux Hot Channel Factor, is defined as the maximum 9
local fuel rod lic:ar power density divided by the average fuel rod linear pcwer density, e.ssuming nominal fuel pellet and rod dimensions.
CRYSTAL RIVER - UNIT 3 3 3/4 2-1 Amendment No.13 4 0 u0 e
POWER DISTRIBUTION LD'ITS BASES F
Nuclear Enthalpy Nse Hot Channel Factor, is defined as the H
ratio of the integral of linear power along the rod on which minimum DNBR cccurs 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 met, provided:
q 1 08; F' H 1 I*7I F
3 Power Peaking is not a directly observable quantity and therefore limits nave been established on the bases of the AXIAL POWER IMBALANCE produced by the power peaking. It has been detemined 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 + 6.5% (indicated rasition) from the group average height.
2.
Regulating rod groups are sequenced with overlapping groups as required in Specification 3.1.3.5.
3.
The regulating rod insertion limits of Specification 3.1.3.6 and the axial power shaping red insertion limits of Specification 3.1.3.9 are maintained.
4.
AXIAL POWER IMBALANCE limits are maintained, be AXIAL POWER IMBALANCE is a measure of the difference in pcwer between the top and bottom halves of the core. Calculations of core average axial peaking factors for many plants and measurements frm operating plants under a variety of operating conditions have been correlated with AX:AL POWER IMBALANCE. The corre1ation shows that the design power shape is not exceeded if the AXIAL PCWER IMBALANCE is maintained within the !imits
- C 1ures 3.2-1 and ?.2-2.
The design limit power peaking factors are the mi otrictive cal-culated at full power for the range frm all control rods fully withdrawn to minimum allowable control rod insertion and are the core CNBR design basis. Therefore, for operation at a fraction of RATED THERMAL POWER, the design limits are met. When usjng incore detectors to make power c'istribu-q and F'aH*
tion maps to deternine F
, shall be a.
The measurement of total peaking factor, F increased by 1.4 percent to account for manufacturing tolerances and further increased by 7.5 percent to account for measurement error.
CRYSTAL RIVER - UNIT 3 B 3/4 2-2 knendment No.16,13 I
G3
.b ri\\
POWER DISTRIBUTION LIMITS Paces The measurement of enthalpy rise hot channel factor, F)Yo,r. h b.
s be increased by 5 percent to account for measurement er For Condition II events, the core is. otected from exceedg 19.7 kw/ft locally, and from going below a minimum DNBR of 1.30 by matic protection on power, AXIAL POWER IMBALANCE, pressure and tempe.
_re.
Only conditions 1 through 3, above, are mandatory since the AXIA. POWER IMBALANCE is an cxplicit input to the T rsctor Protection Systa The QUADRANT POWER TILT limit assur es 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. For QUADRANT POWER TILT, the safety (measurement independent) limit for Steady State is 4.92, for Transient State is 11.07, and for the Maximum Limit is 20.0.
The QUADRANT POWER TILT limit at which corrective action is required provides CNB and linear heat generation rate protection with x-y plane power tilts. The limit was selected to provide an allowance for the uncertainty associated with the power tilt.
In the event th'a tilt is not corrected, the margin for uncertainty on F is reinstated by reducing the power by 2 percent for each percent of tilk in excess of the limit.
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 transient and ac;;ident analyses. The limits are consistent with the FSAR initial assumptions and have been analytically de!nonstrated adecuate to maintain a minimum CNBR of 1.30 throughout each analyze! ;ransient.
The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> periodic surveillance of these parameters through instru-ment readout is sufficient to ensure that the parameters are restcred within their limits following load chaages and other expe:ted transient operation. The 18 month periedic reasurement of the RCS total flow rate is 6dequate to detect flow degradation and ensure correlation of the flow indication channels with measu ed ficw such that the indicated percent flow will provide sufficient ve-ification 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.
CRYSTAL RIVER - UNIT 3 B 3/4 2-3
/
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