ML20034A073
| ML20034A073 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 04/04/1990 |
| From: | Hebdon F Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20034A071 | List: |
| References | |
| NUDOCS 9004190274 | |
| Download: ML20034A073 (10) | |
Text
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'o, UNITE] STATES
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',g NUCLEAR REGULATORY COMMISSION g
E WASHINGTON D. C. 205$$
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OMAHA PyB,LJC,,PpFER DISTRJ,CJ DOCKET NO. 50-285 FORT CAL,HOUN STATI0,N,,, UNIT NO. 1
.A.M.Epp,M,E,H,T,,Tp, f,A,CJ,LJ,T,Y, pf,E,R,A,TJ,Np,,LJ CE NS E Amendment No. 126 License No. DPR-40 1.
The Nuclear Rcgulatory Concission (the Connission) has found that:
l ti.
The application for amendment by the Omaha Public Fower District (the licensee) dated January 26, 1990, as supplernented February 6, 1990, complies with the standards and requirements of the Atornic Energy Act of 1954, as amended (the Act), and the Corciission's rules and regulations set forth in 10 CFR Chapter I; B.
The f acility will operate in conformity with the application, as amended, the provisions of the Act, and the rules and regulations of the Coomiission; C.
Thcre is reasonable assurance:
(i)that.theactivitiesauthorized Ly this arcndr.ent can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Connission's regulations; D.
The issuance of this license amendraent will not be inimical to the conon defense and security or to the health and safety of the public; and E.
The issuance of this emendn'ent is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied, i
M'j4190274 90o4o4
,s p
ADocK o3000285 FDC l
i 2
2..
Accordingly, Fecility Operating License No. DPR-40 is acended by changes to the Technical Specifications as indicated in the attachrntnt to this i
lict.nse atendment, ar.d paragraph 3.B. of Facility Operating License No.
DPR-40 is hereby arnendtd to read as follows:
B.
Tec,hnical Specifications The Technical Specifications contained in Appendix A, as rtvised through Arnendment No.126, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
The license amendinent is effective prior to the reactor going critical after the 1990 outage.
FOR THE ilVCLEAR REGULATORY C0KMISS10t; 1
%dLAQ. /A Fredcrick J. Hebden. Director Froject Directorate IV Divisicn of Reactor Projects - III, i
IV, Y and Special Prcjects Office of t;uclear Feactor Regulation Attachrent:
Changes to tk Technical Specifications Date of 1ssuar.ce: April 4, 1990 1
ATTACHMENT TO LICENSE AMEliDMENT t10 m FACILITY OPERATING LICENSE NO. DPR-40 DOCKET NO. 50-285 Revise Appendix "A" Technical Specifications as indicated below. The revised pages are identified by arr.endrcient nutiber arid contain vtrtical lines indicating the area of change.
Remove P,ajes Insert P,aj,es viii viii 1-2 1-2 2-57 2-57 2-57a 2-57a Figure 1-3 Figure 1-3 Figure 2-5 Figure 2-5 Figure 2-9 Figure 2-9 b
i.
IECHNICAL SPECIFICATIONS - FIGURES TABLE OF CONTENTS PAGE WHICH FIGURE DESCRIPTION FIGURE FOLLOWS l i
1-1 TMLP Safety Limits 4 Pump Operations..........
1-3 12 Axial Power Distribution LSSS for 4 Pump Operation...
1-3 1-3 TMLP LSSS 4 Pump Operation...............
1-3 2-1A RCS Press-Temp Limits Heatup..............
2-6 2-1B RCS Press-Temp Limits Cooldown.............
2-6 2-3 Pred*:ted Radiation induced NDTT Shift.........
2-6 2-10 Spent Fuel Pool Region 2 Storage Criteria.......
2-38 24 PDil..........................
2 53 2-5 Allowable Peak Linear Heat Rate vs Burnup.......
2-53 2-6 LCO for Excore Monitoring of LHR............
2-53 27 LCO for DNB Monitoring.................
2-53 2-8 Flux Peaking Augmentation Factors...........
2-53 T
I 2-9 FR'IXY and Core Power Limitations.........
2-53 T
1 viii Amendment No. JJE,126 i
l.0 SAFETY llMITS AND tiMITING SAFETY SYSTEM SETTINGS i
1.1 Safety Limits - Reactor Core (Continued) would cause DNB at a particular core location to the actual heat flux at t
that location, is indicative of the margin to DNB.
The minimum value of the DNBR during steady state operation, normal operational transients, and anticipated transients is limited to 1.18.
A DNBR of 1.18 corresponds to a 95% probability at a 95% confidence level that DNB will not occur, which is considered an appropriate margin to DNB for all
(
operating conditions.(1)
The curves of Figure 1-1 represent the loci of points or reactor thermal power (either neutron flux instruments or a T instruments), reactor coolant system pressure, and cold leg temperature for which the DNBR is 1.18.
The area of safe operation is below these lines.
The rcactor core safety limits are based on radial peaks limited by the CEA insertion limits in Section 2.10 and axial shapes within the axial i
power distribut1 n trip limits in Figure 1-2 and a total unrodded planar 9
radial peak (F of 1.75.
assumption tha[y )e unrodded integrated total radial peak (Fp ) isTheLS 1
I th l.70. This peaking factor is slightly higher (more conservative) than I
the maximum predicted unrodded total radial peak during core life, excluding measurement uncertainty.
Flow maldistribution effects for operation under less than full reactor coolant flow have been evaluated via model test.(2)
The flow model data established the maldistribution factors and hot channel inlet temperature for the thermal analyses that were used to establish the safe operating envelopes presented in Figure 1 1.
The reactor protective system is designed to prevent any anticipated combin6 tion of transient conditions for reactor coolant system temperature, pressure, and thermal power level l
I that would result in a DNBR of less than 1.18.(1) l References (1)
USAR, Section 3.6.6 i
(2)
USAR, Section 1.4.6 1
P l2 Amendment No. E,72,f3, f7,70,77, 71,117,126
+
I 2.0 LIMITING CONDITIONS FOR OPERATION 2.10 Reactor Core (Continued) 2.10.4 tower Distribution limits (Continued)
I (b)
If while operating under the provisions of part (a), the plant computer incore detector alarms become inoperable, o>eration may be continued for seven days from the date of tie last valid core power distributio') without reducing power provided each of the.following conditions is satisfied:
(i)
A core power distribution was obtained utilizing incore detectors within 7 days prior to the incore detector alarm outage and the measured peak linear heat rate was no greater than 90Y,of the value allowed by (1) above.
(ii) The Axial Shape Index as measured by excore detectors remains within i.05 of the value obtained at the time of the last measured incore power distribution.
(iii) Power is not increased nor has it been increased since the time of the last incore power distribution.
(c)
When the linear heat rate is continuously monitored by the excore detectors, withdraw the full length CEA's beyond the long term insertion limits of Specification 2.10.2.7 and maintain the Axial Shape Index, Y;te is exceeding its limits
, with the limits of Figure 2-6.
If the linear heat ra as determined by the Axial Shape Index, Yj, being outside the limits of Figure 2-6, where 100 percent of the allowable power represents the maximum power allowed by the following expression:
L i
14.4 xM I
where l
1.
L is the maximum allowable linear heat rate as determined from Figure 2-5 and is based on the core average burnup at the time of the latest incore power map.
2.
M is the maximum allowable fraction of rated thermal power as determined by the F T limit curveofFigure2-9whenmonitoringbyexcore detectors. M = 1 when monitoring kw/ft using incore detectors.
(i)
Restore the reactor power and Axial Shape Index, Y;,
to within the limits of Figure 2-6 within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, or-l l
2 57 Amendment No. E, Jp, 72, f},97,117,126
2.0 LIMITING CONDITIONS FOR OPERATION 2.10 Reactor Core (Continued) 2.10.4 Power Distribution Limits (Continued)
(ii)
Be in at least hot standby within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
(2)
Total Intearated Radial Peakina Factor I
R defined by FgT = Fg (1+T )
The calculated value of F distributionmapwithnonongisdeterminedfromahower shall be limited to $1.70.
F l
trippable CEA's inserted and with all full length CEA's at or above the Long Term Steady State Insertion Limit for the existing Reactor Coolant Pump combination. The azimuthal tilt, T, is the measured value of T at the time FR q
q is determined.
I With FR 21.70 within 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />s:
l T
(a)
Reduced power to bring power and F within the limits of Figure 2-9,withdrawthefulllengkhCEA'stoorbeyondtheLong Term Steady State Insertion Limits of Specification 2.10.2(7),
and fully withdraw the NTCEA's, or (b)
Be in at least hot standby.
(3)
Total Planar Radial Peakina Factor The calculated value of F T defined as F T=F (1+T shallbe85termiN5dfrok)apower shall be limited to 11.757# F distribution map with no non Nippable CEA's inserted and with all i
i full length CEA's at or above the Long Term Steady State Insertion Limit for the existing Reactor Coolant Pump combination.- This determination shall be limited to core planes between 15% and 85% of full core height inclusive and shall exclude regions influence by grid offects. The azimuthal tilt, T, is the measured value of T at q
q the time F is determined.
xy With F T 21.75 within 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />s:
l xy to within the limits of h
ReducepowertobringpowerandF{T (a) x Figure 2-9, withdraw tie full leng h CEA's to or beyond the Long Term Steady State Insertion Limits of Specification 2.10.2(7),
and fully withdraw the NTCEA's, or (b)
Be in at least hot standby.
1 2-57a Amendment No. 72, f), f7, 79, 77, 92, 199,117,126
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' 2400 psio
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540 2250 psia s
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O 520
$10 1750 psio 500 60 70 80 90 100 110 120 CORE POWER (% OF RATED POWER)
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O + l 0*
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\\R IN PF(B)
= 1.0 81100%.
=
.0088 + 1.8 50%<B<100%
= 1.4 Bs50%
i A1(Y) = -0.35294Yi + 1.08824 Yg1 25 l
0.57143Y3 + 0.875 Y >.25
=
i Thermal Morgin/L0w Pressure t.SSS Om0h0 Public P0wer District Figure 4 Pump Operation F0rt Calhoun St0 tion-Unit NO.1 1-3 Amendment No. g,77,92,
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16 i
4 i
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F s
g 15 g
UNACCEPTABLE OPERATION 1-y 14.4 KW/f7_
=
12:<
14 g
ACCEPTABLE OPERATION a
t w
s 13
)o-a a
l 12 f
O 2500 5000 7500 10000 12500 15000 CYCLE AVERAGE BURNUP (MWD /MTU)
Allowable Peak Linear Heat Rate Omaha Public Power District Figure vs. Burnup Fort Calhoun Station-Unit No.1
- 2 Amendment No. 8,20,72,47,II,
t 110 4
(LIxIT F4 LIMIT-5 3
x 5
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~
80 8
(1.86, 75)
(1.92, 75)
-=
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i M
70 I
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0 1.65 1.70 1.75 1.80 1.85 1.90 1.95 2.00 2.05 Fh AND Fj a
FhfiandCorePcWer OsanaPublicPowerDistrict figure limitations FortCalhounStation-UnitNo.1 24 Amendment No. 8,20,32,#3,47,f0,//,
92n109 E 12b
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