ML20196L340
| ML20196L340 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 04/30/1999 |
| From: | Blair J, Flannigan R, Morris M WOLF CREEK NUCLEAR OPERATING CORP. |
| To: | |
| Shared Package | |
| ML20196L339 | List: |
| References | |
| NUDOCS 9905280145 | |
| Download: ML20196L340 (14) | |
Text
W:lf Crc k G:n: rating Statinn yr Cycla 11 NUCLEAR OPERATING CORPORATION Core Operating Limits Report WOLF CREEK GENERATING STATION CYCLE 11 CORE OPERATING LIMITS REPORT i
Revision 1 April,1999 Prepared by:
M[ttnew K. Morris [
Reviewed by:
jr Approved by:
R. D. Flannigan ' "
g;f D
D 05 0 82 P
PDR Page 1 of 14
e W:lf Creek G:nerating Stati:n Cycla11 C:re Optrating Limits R: port UCLEAR OPERATING CORPORATION i
1.0 ' CORE OPERATING LIMITS REPORT i
The CORE OPERATING LIMITS REPORT (COLR) for Wolf Creek Generating Station Cycle 11 has been prepared in accordance with the requirements of Technical Specification 6.9.1.9.
The core operating limits that are included in the COLR affect the following Technical Specifications:
3.1.1.3.b Moderator Temperature Coefficient (MTC) EOL Limit j
3.1.3.5 Shutdown Rod Insertion Limit 3.1.3.6 Control Rod insertion Limit 3.2.1 Axial Flux Difference (AFD) 3.2.2 Heat Flux Hot Channel Factor-F (Z) g 3.2.3 Nuclear Enthalpy Rise Hot Channel Factor - FA 3.9.1.b Refueling Boron Concentration r*
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W:lf Creek G:n:rcting Stati:n Cycla 11 v
Csm Operz ng Limits R:p:rt
' NUCLEAR OPERATING CORPORATION 2.0 OPERATING LIMITS The cycle-specific parameter limits for the specifications listed in Section 1.0 are presented in the subsections below:
2.1 Moderator Temperature Coefficient (MTC) EOL Limit (Tech Spec 3.1.1.3.b)
The EOL MTC shall be less negative than -50 pcm/deg F.
The 300 PPM MTC Surveillance Limit is -41 pcm/deg F (all rods withdrawn, Rated Thermal Power condition).
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Walf Creek G:n: rating Stati:n Cycla 11 NUCLEAR OPERATING CORPORATION Cere Opecting Limits Rrport i
Revision 1
'2.2 Shutdown Rod Insertion Limit (Tech Spec 3.1.3.5)
The shutdown rods shall be fully withdrawn (i.e., positioned within the interval of > 222 and < 231 steps withdrawn).
2.3 Control Rod insertion Limits (Tech Spec 3.1.3.6)
The Control Bank Insertion Limits are specified in Figure 1.
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r Wsif Creek G:ntrating Stati:n Cycla 11 Ccra Opercting Limits R:psrt NUCLEAR OPERATING CORPORATION Revis,on 1 i
(FULLY WITHDR AWN) 220
[28.7 %",~ 222 ) '
( 7 8.7 % 2 2'2')'
l
~
l 200 BANK
.g.
180 I
160I g 0 % 'i d t ) '.
S l
I T
E 140 l
P BANK:
l g
- c- -
i 120 1
T H 100 4
e s
D R
A 80 N
.;D '
j BANK-.
60 40
+
4 I
20 e
[s.2%,o) 0 0
20 40 60 80 100 (F
IN S E RTE D)
RATED THERM AL POWER (Percent)
Figure 1 Rod Bank insertion Limits Versus Thermal Power-Four Loop Operation Fully Withdrawn shall be the condition where control rods are at a position within the interval of 2 222 and 5 231 steps withdrawn.
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W lf Creek G:n:r: ting Stati:n Cycla 11
~
UCLEAR OPERATING CORPORATION Cere Opercting Limits Repart Revision 1 2.4 Axial Flux Difference (AFD)
The indicated Axial Flux Difference (AFD) allowed operational space is defined by Figure 2.
l l
l l
1 l
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W lf Creek G:n r: ting Staticn CREEK Cycla 11 NUCLEAR OPERATING CORPORATION Cere Opercting Limits Raptrt Revision 1 110 I"I 100 - UNACCEPTABLE UNACCEPTABLE O P E R ATIO N O P E R A TIO N g
0 F
90 R
A T
E D 80 T
H ACCEPTABLE O PE R ATIO N R 70 M
A L
P 60 O
W E
R 50
( 28.so l (24.so) 40
-40 30 20 10 0
10 20 30 40 AXIAL FLUX DIFFERENCE (DELTA I) %
Figure 2 Axial Flux Difference Limits as a Function of Rated Thermal Power Page 7 of 14 I
W lf Creek G:nerating Stati:n CREEK Cycl 311 s
NUCLEAR OPERATING CORPORATION Ctre Oper;t g Limits R: port 2.5 Heat Flux Hot Channel Factor - Fo(Z)
F" F (Z)s #
- K(Z), for P > 0.5 g
7 E(Z)s
- K(2), frPs0.5 g
03 THERMAL POWER e,
P
=
RATED THERMAL POWER '
)
F""'
the F (Z) Limit at RATED THERMAL POWER (RTP)
=
g o
2.50, and
=
K(2) is defined in Figure 3.
Fl"' (K(Z))
for P > 0.5 F7(Z) s p
F7(Z) s. j" (K(Z))
F W(Z)
(0.5]
W(Z)
Ratio of the Fo from normal operation transients to the Fo at
=
steady state conditions Page 8 of 14
Wcif Creek G:nerating Stati:n CREEK Cyci)11 r
NUCLEAR OPERATING CORPORATION Ccre Ope g 1.imits R ptrt (Tech Spec 4.2.2.2.e.1)
With measurements indicating maximum
'F"(Z)'
over z sK(Z),
has increased since the previous determination of F7(Z),
F7(Z) shall be increased over that specified in 4.2.2.2.c by an appropriate factor, the F Penalty Factor.
g See A.ppendix A for:
1.
W(Z) 2.
F Penalty Factor g
i i
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W:lf Creek Gen: rating Statinn CREEK Cycla 11 UCLEAR OPERATING CORPORATION Core Operating Limits Report 1.2 Ny 1.0 E
OF
$0.8 e
E O
E M 0.6 h
8.
O N0.4 U
Fa' = 2.50 EO 0.2 Ei.v. tion (ft) x(z)
E o.o 1.000 s.o
- 1. coo 12.0 c.925 0.0 0
2 4
6 8
10 12 CORE HEIGHT (FT)
Figure 3 K(Z) - Normalized Peaking Factor Vs. Core Height Page 10 of 14
f I-W lf Creek G:nerating Stati:n Cycle 11 NUCLEAR OPERATING CORPORATION Core Operaung Umh Repod Revision 1
'2.6
' Nuclear Enthalpy Rise Hot Channel Factor - F5 (Tech Spec 3.2.3) l FE shall be limited by the following relationship:
F5 s Fi" 1.0 + PF,(1.0 - P)
- Where, F5"
= The F5 limit at RATED THERMAL POWER (RTP) l 1.586
=
PF,
= the power fat. tor multiplier for F5
= 0.3 P
THERMAL POWER RATED THERMAL POWER Fn = Measured values of F5 obtained by using the movable incore detectors to obtain a power distribution map. The measured values of FE shall be used since an uncertainty of 4% for incore measurement of FL has been included in the above limit.
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Wclf Creek G:n:rcting Stati:n Cycle 11 UCLEAR OPERATING CORPORATION Core Operat g Limits Report 2.7 Refueling Boron Concentration (Tech Spec 3.9.1.b)
The refueling boron concentration shall be greater than or equal to 2300 PPM.
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W:lf Creek G:n rating Stati n Cycb 11 NUCLEAR OPERATING CORPORATION Core Operating Limits Report APPENDIX A A.
Input relating to Specification 4.2.2.2.c:
W(Z) =,n(Z)** ""*"'
F
(,),,,,,
These values are issued in'a controlled report which will be provided on
- request, input relating to Specification 4.2.2.2.e.1 Cycle Burnup F7(Z) Penalty Factor 0
2.00 3900 2.00 4000 3.00 5500 3.00 5600 2.00 21000 2.00 Note: All cycle burnups outside of the above table shall use a 2% penalty factor for comoliance with 4.2.2.2.e Surveillance Requirement.
Linear interpo ation should be used for intermediate cycle bumups.
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Wrif Creek G:n: rating Stati:n y
Cycb 11 NUCLEAR OPERATING CORPORATION om O m ng mik rem Technical Specification BASES 3/4.2.5 Cycle 11 Safety Analysis DNB Limit 1.76 WRB-2 Design Limit DNBR 1.23 i
I e
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