ML20211P708
| ML20211P708 | |
| Person / Time | |
|---|---|
| Site: | Perry  |
| Issue date: | 08/31/1997 |
| From: | Reda R, Robert Williams GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20211P698 | List: |
| References | |
| J11-03060SRLR, J11-03060SRLR-R00, J11-3060SRLR, J11-3060SRLR-R, NUDOCS 9710210121 | |
| Download: ML20211P708 (20) | |
Text
_
3;. il GENuclear Energy J11-03060SRLR Revision 0 Class I August 1997 i
Supplemental Reload Licensing Report for i
PERRY N~UCLEAR POWER PLANT UNIT 1 Reload 6 Cycle 7 1
=
EEA "A8sli Ms86Lo P
9,:
GE Nuclear Energy J11-03060SRLR Revision 0 ClassI August 1997 J11 -03060SRLR, Rev. O S'ipplemental Reload Licensing Report for Perry Nuclear Power Plant Unit 1 Reload 6 Cycle 7 Approved N'-
M Approved R.J. Reda, Manager R.D. Williams Fuel and Facility Licensing Fuel Project Manager i
'ERRY l loload6 J11-03060SRLR Rev.0
(..
Important Notice Regartling Contents of This Report Please Read Carefully This report was prepared by General Electric Company (G E) solely for Cleveland Electric illuminating Company. The information contained in this report is believed by GE to be an accurate and true represLntation of the facts known, obtained or provided to GE at the time this report was prepared.
The only undertakings of GE respecting information in this document are contained in the contract between Cleveland Electric illuminating Company and G E, and nothing con-tained in this document shall be construed as changing the contract. The use of this in-formation by anyone other than CEI for any purpose other than that for which it is in-tended,is not authorized; and with respect to any such unauthorized use, neither GE nor any of the contributors to this document makes any representation or warranty (ex-pressed or implied) as to the completeness, accuracy or usefulness of the information contained in this document or that such use of such information may not infringe privately owned rights; nor do they assume any responsibility for liability or damage of any kind which may result from such use of such intormation.
Page 2
'ERRY I Jil-03060SRLR Reload 6 Rev.0 1..
Acknowledgement The engineering and reload licensing analyses, which form the technical basis of this Supplemental Reload Licensing Repon, were performed by J.E. Fawks. The Supplemental Heload Licensing Repon was prepared by J.E. Fawks. This document has been verified by D.B. Waltermire.
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PERRYI-J11-03060SRLR Reload 6 Rev.0 m
-l,.
The basis for this report is General Electric Standard Applicationfor Reactor Fuel, NEDE-24011-P-A-13, August 1996; and the U.S. Supplement, NEDE-24011-P-A-13-US, August 1996.
1.
Plante 4que Iternn fs.,
Apperdix A: Ar.alysis Conditions
?
Appendix B: Basis for Analysis ofless-of-Feedwater Heating Event AppenG C: Analyzed Operating Domain
~
Appendix D: Transient Analysis Appendix E: Power and Flow Dependent MCPR and MAPLHGR Multipliers 2.
Reload FuelIlundles 1
Cycle s
Fuel Type Loaded Number Irradiated:
GE 10-PXXB 306-10GZ2-120M-150--T (GE8x8NB-1) 4 24 GE l o-P8SXB306-i l G73-120M-150-T (GE8 x 8N B-1 )
5 192 G E l o-P8S XB 306-I I GZ3-120M-150-T (G E8x8NB-1 )
6 44 Gell-P9SUB338-10GZ-129T-146-T (Gell) 6 152 Gell-P9SUB338-120Z-120T-146-T (Gell) 6 84 Neyc l
GE12-P10SSB369-12GZ-120T-150 ', (GE12) 7 160 gel 2-F 10SSB369-14GZ-120T--150-T (GE12) 7 92 Total 748 3.
Reference Core Loading Pattern Nominal previous cycle core average exposure at end of cycle:
26397 mwd /MT
( 23947 mwd /ST)
Minimum previous cycle core average exposure at end of cycle 25995 mwd /MT from cold shutdown considerations:
( 23582 mwd /ST)
Assumed reload cycle core average expoEure at beginning of 14552 mwd /MT cycle:
( 132G mwd /ST)
Assumed reload cycle core average exposure at end of cycle:
26637 mwd /MT
( 24165 mwd /S T)
Referen.7 tore loaaing pattern:
Figure 1 Page 4
PERRY 1 Jl1-03060SRLR Re1 pad 6 Rev.0 4.
Calculated Core Effective Multiplicatior: and Control System Worth - No Volds,20oC Beginning of Cycle, kenean, Uncontrolled 1.117 Fully controlled 0.959 Strongest control rod out 0.989 R, Maximum increase in cold core reactivity with exposure into cycle,4k 0.000 5.
Standby Liquid Control System Shutdown Capability
' 9 Boron (ppm)
Shutdown Margin (Ak) s+
(at 20 C)
(at 160*C, Xenon Free) 660 0.025 6.
Heload Unique GETAB Anticipated Operational Occurrences (AOO) Analysis Initial Condition Parameters Exposure: BOC7 to EOC7 Increased core flow /Feedwater temperature 420*F Peaking Factors Fuel Bundle Bundle Initial Desir,o Lo:al Radial Axial R-Factor Power Flow MCPR (MWt)
(1000 lb/hr)
GE12 1.45 1.50 1.36 1.040 6.991 116.3 1.29 GE11 1.45 1.42 1.36 1.035 6.612 118 9 1.24 GE8::8NB-1 1.20 1.60 1.40 1.000 7.484 114.2 1.18 Exposure: HOC 7 to EOC7 Increased core flow /Feedwater temperature reduction to 250 F
~
Peaking Factors Fuel Bundle Bundle Initial Design Local Radial Axial R-Factor Power Flow MCPR C :Wt)
(1000 lb/hr)
GE12 1.45 1.62 1.35 1.040 7.533 112.8 1.26
~
gel 1 1.45 1.52 1.35 1.035 7.049 116.2 1.23 GE8x8NB-1 1.20 1.66 1.40 1.000 7.734 111.8 1.I9
{
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,.pp
P.ERRY1
/
J11-03060SRLR e,
Reload 6 Rev,0 j
7.
Selected Margin Impmvemeat Options Rechculation pump trip:
Yes Rod withdrawallimiter:
Yes Thennal powe' monitor:
Yes impicved scram time:
No (ODYN Option B)
Measured scram time:
No Exposure dependentlimits:
No Exposure points analyzed:
1(EOC) 8.
Operating Flexibility Options Single-loop operation:
Yes Load line lirnit:
No Extended loadlinelimit:
No Maximum extended loadlinelimit:
No increased con: flow throughout cycle:
Yes 4
Flow point analyzed:
105.0 %
Increased core flow at EOC:
Yes Feedwater temperature reduction throughout cycle:
Yes Temperature reduction:
170.0 F Final feedwater temperature reduction:
Yes ARTS Program:
No Maximum extended operating domain:
Yes Moisture separator reheater OOS:
No Turbine bypass system OOS:
No Safety /rellef valves OOS:
Yes AP OOS:
No Main steam isolation valves OOS:
No Page 6
,, - Rehad 6 Rev, 0 90 Core-wide AOO Analysis Results Methods usedt GEMINI; GEXL-PLUS
-l Exposure range: BOC7 to EOC7 Increased core fbw/Feedwater temperature 420 F
~
Uncorrected ACPR Event Flux Q/A GE12 GE11 GE8x8NB-1 Fig.
(%NBR)
(%NBR)
Load Reject w/o Bypass 301 109 0 20 0.15 0.09 2
Loss of 100 F Feedwater 115 115 0.12 lleating Exposure range: BOC7 to EOC7 Increased core flow /Feedwater temperature reduction to 250 F Uncorrected ACPR Event Flux Q/A GE12 3 Ell GE8x8NB-1 Fig.
(%NBR)
(%NBR)
FW Controller Failure 204 115 0.18 0.13 0.10 3
- 10. Local Rod Withdrawai Error (With Li.niting Instrument Fallure) AOO Summary The generic bounding BWR/6 rod withdrawal error analysis described in NEDE-24011-P-A-US is applied.
- 11. Cycle MCPR Values 1 In agreement with commitments to the NRC ( letter from M.A. Smith to the Document Control Desk,10CFR Part 21, Reportable Condition. Safety Limit MCPR Evaluation, May 24,1996 ) a cycle-specific Szfety Limit MCPR calculation was performed, and has been reported in both the Safe y Limit MCPR and the Operating Limit MCPR shown below. This cycle specific SLMCPR was determined using the analysis basis docu-mented in GESTAR with thc following exceptions:
' l. The reference core loading pattern in Figure I was analyzed.
- 2. The actual bundle parameters ( e.g., local peaking ) were used.
- 3. The full cycle exposure range was analyzed.
I. For single-loop operation, the MCPR operating linut is 0 01 greater than the two-loop value. The MCPR limit does not change because of channel bow. Channel bow is reflected m the monitanng of the core.
Page 7 1
PERRYI t, ', -
Reload 6 311-03060SRLR Rev.O
- 11. Cycle MCPR Values (cont)
Safety limit:
1.09 Single loop operation safety limit:1.10 Non-pressurization events:
Exposure Range: BOC7 to EOC7 Option A GE12 Gell GE8x8NB-1 Rod Withdrawal Error 1.20 Fuel Loading Error (misoriented) 1.28 1.24 1.25 Fuel Loading Error (mislocated) 1.26 Loss of 100'F Feedwater Heating 1.21 Pressurizatloa events: 2 Exposure range: BOC7 to EOC7 Increased core flow /Feedwater temperature 420 F Exposure point: EOC7 Option A GE12 Gell GE8x8NB-1 Load Reject w/o Bypass 1.30 1.26 1.19 Exposure range: BOC7 to EOC7 Increased core flow /Feedwater temperature reduction to 250 F Exposure point: EOC7 Option A GE12 Gell GE8x8NB-1 FW Controller Failure 1.28 1.24 1.20
- 12. Overpressurization Analysis Summary 3 Psi Py Plant Event (psig)
(psig)
Response
MSIV Closure (Flux Scram) 1258 1289 Figure 4
- 3. Ect'S MCPR value is 1.17.
- 3. He MSIV desure (flux scram) analysis is performed using GEMINI methods at the 102% powerlevel to account for the power level un-certainties specified in Regulatory Guide 1.49. He dome pressure is set to 1045 psig as specified in the OPle3 *)esign Guide. TDP-0087 Revision O. De analysis was performed with the 111Q,Mst setpoim safety valves operauonal.
Page 8
r PERRY 1-2 J11-03060SRLR t l, Reload 6 Rev. 0 -
j i
- 13. Loading Error Results4 l
Variable water gep misoriented bundle analysis: Yes Mislocated bundle analysis:
Yes Misoriented Fuel Bundle ACPR GE10-P8SXB306-11GZ3-120M-150-T (GE8x8Nel) 0.16 GE 10-P8SXB306-10GZ2-120M-150-T (GE8x8NB-1) 0.09 Gell-P9SUB33&-10GL12W-146-T (gel l) 0.15
~
gel l-P9SUB338-12GL12W-146-T (gel 1) 0.14
- GE12-P10SSB369-14GL12W-150-T (GCl2) 0.19 GE12-P10SSB369-12GL12W-150-T (GE 12) 0.11 Mislocated Fuel Bundle-ACPR Fuel Loading Error (mislocated) 0.17
- 14. Control Rod Drop Analysis Results This is a banked position withdrawal sequence plant, therefore, the control rod drop accident analysis is not required. NRC approvalis documented in NEDE-240ll-P-A-US.
- 15. Stability Analysis Results GE Sile380 recommendations have been included in tw operating procedures; therefor:, no stability analy-sis is required. NRC approval for deletion of a cycle-specific stability analysis is documented in NEDE-24011-P-A-US. This plant recognizes the issuance of NRC Bulletin No. 88-07, Supplement 1 Power Oscillations in Bolling Water Reactors (BWRs), and will comply with the recommendations contained therein.
- 16. Loss-of-Coolant Accident Results-LOCA method used: SAFE /REFLOOD The peak clad temperature (PCI') is $ 2184'F at all exposures; the local oxidation (fraction) is s 0.065 at all exposures. The core-wide metal water reaction is 0.25%. Tir MAPLHGR multiplier for single-loop op-eration (SLO)is 0.78 for the new fuel. The most and least limiting MAPLHGRs for the new fuel are as fol-lows.
- 4. Includes a a02 penalty due to variable water gap R-factor uncertainty.
Page 9
f-hl Jl1-03060g I
6
- 16. Loss-of-Coolant Accident Results (cont)
Bundle Type: gel 2-P10SSB369-14GZ-120T-150-T I-Average Planar Exposure MAPLIIGR(kW/ft)
PCT Oxidation (GWd/ST)
(GN55$T)
Most Limiting Least Limiting
( F)
Fraction 0.00 0.00 9.00 9.35 0.20 0.22 9.N 937 2037 0.043 1.00 1.10 9.12 9.44 2051 0.045 2.00 2.20 9.23 9.55 3.00 331 9.36 9.66 4.00 4.41 9.50 9.76
~
5.00 5.51 9.65 9.87 2091 0.049 6.00 6.61 9.77 9.98 7.00 7.72 9.89 10.10 8.00 8.82 10.01 10.21 2d2 0.053 9.00 9.92 10.14 10.33 10.00 11.02 10.26 10.45 2171 0.059 11.00 12.13 10.36 10.46 12.00 13.23 10.33 10.47 2167 0.058 13.00 14.33 10.28 10.47 2180 0.060 14.00 15.43 10.21 10 37 15.00 16.53 10.14 10.26-2119 0.050-17.00 18.74 9.97 10.12 20.00 22.05 9.72 9.91 2N8 0.N1 25.00 27.56 9.29 9.47 1951 0.031 30.00 33.07 8.86 9.02 35.00
-38.58 8.43 8.57 1809 0.018 40.00 44.09 7.99 8.14-45.00 49.60 7.52 7.66 1728 0.013 50.00
-55.12 7.04 7.20 1617 0.004 55.00 60.63 6.53 6.72 1552 0.003 58.24 64.20 6.19 6.42 58.32 64.29 6.41 60.00 66.14 6.25 1480 0.002 60.11 66.26 6.24 Page 10
PERRYl J11-03060SRLR
.. - RHoad 6 Rev.0
- 16. ' Loss-of-Coolant Accident Results (cont) l l
Bundle Type: GE12-P10SSB369-120Z-120T-150-T l
Average Planar Exposure MAPLliGR(kW/ft)
PCT Oxidation (GWd/ST)
(GWd/MT)
Most Limiting Least Limiting
('F)
Fraction 0.00 0.00 8.89 9.19 0.20 0.2'2 8.94 9.21 2019 0.040 1.00 -
1.10 9.05 9.31 2038 0.043 2.00 2.20 9.19 9.44 3.00 3.31 9.32 9.58 4.00 4.41 9.46 9.71 5.00 5.51 9.61 9.84 2086 0.047 6.00 6.61 9.75 9.98 7.00 7.72 9.90 10.12 8.00 8.L2 10.05 10.26 2137 0.654 9.00 9.92 10.19 10.40 10.00 11.02 10.33 10.53 2181 0.060 11.00 12.13 10.41 10.53 12.00 13.23 10.38 10.53 2174 0.059 13.00 14.33 10.32 10.51 2178 0.059 14.00 15.43 10.25 10.40 15.00 16.53 10.17 10.29 2123 0.051 17.00 18.74 10.00 10.15 20.00 22.05 9.74 9.93 2050 0.041 25.00 27.56 9.31 9.49 1955 0.031 30 'X) 33.07 8.88 9.04 35.00 38.58 8.45 8.58 1810 0.018 40.00 44.09 8.00 8.17 45.00 49.60 7.54 7.68 1731 0.013 50.00 55.12 7.05 7.22 1617 0.003 55.00 60.63 6.54 6.72 1551 0.003 58.22 64.18 6.20 6.42 58.30 64.27 6.41-60.00 66.14 6.25 1479 0.002 60.08 66.22 6.25 Page11
PERRYl Jil-03060SRLR RWload 6 Rev.0 me BEBs M BEMimm mBIBBBEMBEBEBsBIBim mMBEMBEMBEMMMBEMm mBEBIBEMBEBEBEBsBEBEBiBEBim
- Lm:MMMMMBEMMMMMMMis
- MMMMBsMMMMMMMMMM
- MMMMBEMMMBEMMBEMMM
- BEMMMMMMMMMMMMMM
- M M M M M M M M M M M M M M M
- MMMMMMMBEMMMMMMM f:-*iMMMMMMMMMMBEMME" i:
"MMMMMMMMMBEBEMBi*
"MMMMMMMMMMM" "MMMMMMMMM" l
- B M M M E!**
IIIIIIIIIIII 1 5 5 7 91115151719 2123 25 27 29 3153 35 57 591113 45 if il 5153 55 57 59 Fuel Type A=GE10-P8SXB306-10GZ2-120M-150-T (Cycle 4)
E=GE l l-P9SUB 338-12GZ-12(TT-146-T (Cycle 6)
B4E10-P8SXB306-llGZ3-120M-150-T (Cycle 5)
F=GE12-P10SS3369-14GZ-12(TT-150-T (Cycle 7)
C=G E 10-P8S XB306-1 I GZ3-120M-150- r (Cycle 6)
G=G E 12-P 10SS B 369-1202-120T-150-T (Cycle 7)
D=GE l 1-P9SU B 338-10GZ-12(TT-146-T (C)cle 6)
Figure 1 Reference Core Loading Pattern Page 12
'ERRYl
- leload 6
- Jl1-03060SRLR Rev.0-I-
Neutron Flux.
Vessel Press Rise (psl}
- - - Ave Surface Heat Flux
- - - - Safety Valve Flow 150.0
- Core inlet Flow 300.0 - --- Relief Valve Flow
--- Bypass Valve Flow g 100.0
%, N g 200.0 Y,%
g
~s~_
g 50,0 100.0 I-7 I
I
\\
l T
I
\\
'I I'
0,0 O.0 -
O.0 3.0 6.0 0.0 3.0 0.0 Time (sec)
Time (sec)
Level (inch-REF-SEP-SKRT)
Void eactivity Vessel Steam Flow
- - - - - Doppi r Reactivity 200.0
--- Turbine Steam Flow 1,0 Scram Raactivity Feedwater Flow
- - TotalR activity g
I 8
C g100.0 g o 0.0 h. -- - * *.. : -
e.,
g L',
s,
.g o
\\t M' 's O
\\
g l'
b i
- E g,
Q --.*- My - - - - - - - -- M-0.0
- 1.0 lj x
ul'
\\
lJ'
-100.0
- 2.0 -
L-0.0 3.0 6.0 0.C 3.0 60 Time (sec)
Time (sec)
Figure 2 Plant Response to Load Reject w/o Bypass (BOC7 to EOC7 increased core flow /Feedwater temperature 420 F)
Page 13
~
Re;RRY I Jil-03060SRLR
)E cad 6 '
Rev.0
- f,.
Neutron Flux
--- Vessel press Rise (pol)
- - Ave Surface Heat Flux
- - - - Safety Valve Flow 150.0 - - -- Core inlet Flow 125.0 - --- Relief Valvr clow e
- -- Core inlet Sutr:ooling
--- Bypass Valve Flow
- ,,_u s*** (
h' > s i
g 100.0 C e 75.0 aT
', N c
s i
l.
50.0 25.0 i
n 1
0.0
- 25.0 I
0.0 6.0 12.0 0.0 6.0 124
(
Time (sec)
Time (sec)
Level (inch-REF-SEP-SKRT)
Void Reactivity
- - - - - Vessel Steam Flow
- - Doppler Reactivity 150.0 :. -. Iutbina.Slasciflow., _
1.0 - --- Scram Reactivity
--- Feedwater Flow
--- TotalReactivity w
n
. ] 300.0
\\
u n
0.0 M y _.._.* _.. g...,
I g
\\w g
o
.g l,*
O
\\
o -
w
- l.,.
x
\\'
S...
\\
/
Sa0
/
i.-
e -1.0 t
e g
/
. j l l '. l'.g,.,.
.. N.
I..
0.0
~ 2.0 I
0.0 6.0 -
12.0 0.0 6.0 12.0 Time (sec)
Time (sec)
Figure 3 Plant Response to FW Controller Failure (BOC7 to EOC7 Incteased core flow /Feedwater imperature reduction to 250 F)
Page 14
PERRY l Reload 6
. J11-03060SRLR Rev.0
'f,'
ll'
)
Pleutron Flux Vessel Press Rise (psi)
A ve Surface Heat Flux
- - - - Safety Valve Flow 150 0 - - - - --
e inlet Flow 300.0 - --- Ralief Valve Flow
--- Bypass Valve Flow
~
a.,
1 g 100.0
/,bs
's',
200.0
=
9 N'-
E k
E i
~
~ ~~.
~,
50.0 100 0 0.0 O.0 O.0 4.0 8.0 00 4.0 8.0 Time (sec)
Time (sec)
Levet(inch-REF-SEP-SKRr)
'hid Readvity
- - - - - Vessel Steam Flow
- - - - Dop Reacti '
200 0 - --- Turbine Steam Flow 1.0 - --- Ser m Reactivity
--- Feedwater Flow
---T Reactivtty g
.2 g 100.0 g 0.0
.\\
g
. g' s y. -
g-
>i..T...-
s o
\\\\
\\.. - _..
s.
~~..
g{1.0 3'
O.0
)'
'\\ - - -%
g~
g g
E-
\\.
\\\\\\\\
l-II !
-100.0
- 2.0 0.0 4.0 8.0 0.0 4.0 8.0 Time (sec)
Time (sec)
Figure 4 Plant Response to MSIV Closure (Flux Scram)
Page 15
PERRYl-Relord 6 J11-03060SRLR Rev.0 p..
Appendix A Analysis Conditions To reflect actual plant parameters accurately, the values shown in Table A-1 were used this cycle.5 Table A-1 Increased core flow /Feedwater temperature as noted Analysis Value Parameter /Feedwater tempereture 420*F 250'F Thermal power, MWt 3579.0 3579.0 Core flow, Mlb/hr 109.2 109.2 Reactor pressure, psia 1056.0 1023.7 Inlet enthalpy, BTU /lb 528.3 508.4 Non-fuel power fraction -
0.037 0.037 Steam flow analysis. Mlb/hr 15.41 12.59
~ Dome pressure, psig 1025.0 994.6 Turbine pressure, psig 974.8 960.3 No. of Dual Mode S/R Valves ( see footnote 5 )
19 19 Relief mode lowest setpoint, psig ( see footnote 5 )
1133.0 1133.0 Safety mode lowes: setpoint, psig ( see footnote 5 )
1200.0 1200.0 4
l
- 5. There are a total of 19 va!ves; the two lowu setpoint safety /rchef valves are assumed to be out-of-service in the transient analysis. For the hts 1VFS overpressurintion analysis. 6 safety valves are assumed out-of-scresce.
Page 16
PERRYl Jl1-03060SRLR
_ R load 6 Rev. O l-Appendix B Basis for Analysis of Loss-of-Feedwater Heating Event The loss of feedwater heating event was analyzed with the 3D BWR simulator code described in NEDE-24011-P-A, which perm:ts the use of this code for this analysis.The transient plots normally reported in Section 9 are not outputs of the 3D BWR simulator code; therefore, these items are not included iri Isis document -
The transient analysis inputs normally reported in Section 6 of this document are internally calculated in the 3D BWR simulator code.
Page 17
[dlDd 6 Y
Ji 3
v Appendix C Analyzed Operating Domain The core-wide adnormal operational occurrence (AOO) analysis results reponed in Section 9 are the most limiting values over the entire allowable operating range. This range covers the folio sing operating options:
- 1. Standard 100% power / flow map;
- 2. End-of-cycle power coastdown;
- 3. MEOD with 100% power, flow range from 75% to 105% of rated; cnd
- 4. Partial feedwater heating to 320*F during the cycle with final feedwater temperature reduct:on to 250 F after All Rods Out at end of cycle.
Limiting events and conditions analyzed are based on NEDE-24011-P-A-US and the US AR 'nalytical re-suits. The Reload 6/ Cycle 7 analyse:, were performed assuming all four tarbine control valves in a full are mode of operation. This is conservative for partial are configuration.
t 5
Page 18 l
m
PERRYl Jl1-03060SRLR
.. Belo' ad 6 Rev.O v.
Appendix D Transient Analyses The turbine trip without bypass (TTNBP) analysis is a pressure increase event that is bounded by the load rejection without bypass (LRNBP) analysis.
The LRNBP is limiting at nonnal feedwater temperature and increased core flow.
l The feedwater controller failure (FWCF) is limiting at reduced feedwater temperature and increased core
[
flow.
The pressure regulator failure down scale (PRFDS)is not limiting.
Transients were not mn for the intermediate feedwater temperature cases (320 F and 310*F) because the op-erating limit would not improve for those conditions. The LRNBP and fuel loading error analysis sets the operating limit and does not change vith feedwater temperatuce, Page 19
PE RRY l Jil-03060SRLR Reload 6
'e,
Rev.0 Appendix E Power and Flow Dependent MCPR and MAPLHGR Multipliers he original MEOD offrated MCPR and MAPLHGR multipliers were confirmed to be applicable to this cycle. The MEOD original MCPRr equatious must be multiplied by the ratio of 1.09/1.07. Furthermore, for fuel GE8x8NB-1 the MCPRr equation at or below 40% rated flow needs to be multiplied by (1.0+0.0032(40-F)) where F is core flow in terms of % rated.
(
Page 20