ML20101F334

From kanterella
Jump to navigation Jump to search
Supplemental Reload Licensing Rept for Perry Nuclear Power Plant Unit 1,Reload 5,Cycle 6
ML20101F334
Person / Time
Site: Perry FirstEnergy icon.png
Issue date: 01/31/1996
From: Reda R, Robert Williams
GENERAL ELECTRIC CO.
To:
Shared Package
ML20101F312 List:
References
J11-02581SRLR, J11-02581SRLR-R00, J11-2581SRLR, J11-2581SRLR-R, NUDOCS 9603250459
Download: ML20101F334 (20)
v • d • e


Text

1 GE Nuclear Energy J11-02581SRLR Revision 0 ClassI January 1996 J11-02581SRLR, Rev. O Supplemental Reload Licensing Report for I Perry Nuclear Power Plant Unit 1 l Reload 5 Cycle 6 l

l i

l Approved 8 Approved R.J Reda, Manager R.D. Williams Fuel and Facility Licensing Fuel Project Manager i

i 9603250459 960320 PDR ADOCK 05000440 P PDR

l PERRYl n :-vao mn _n Reload 5 Rev.O Important Notice Regarding 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 representation of the facts known, obtained or provided to GE at the time this report was prepared. I l

The only undertakings of GE respecting information in this document are contained in the contract between Cleveland Electric Illuminating Company and GE, 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 j 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 i owned rights; nor do they assume any responsibility for liability or damage of any kind i which may result from such use of such information.

l Page 2

PERRYI J 1 1 -o.>.3615 KLK Reload 5 Rev.O Acknowledgement The engineering and reload licensing analyses, which fann the technical basis of this Supplemental Reload Licensing Report, were performed by J.E. Fawks. The Supplemental Reload Licensing Report was prepared

  • by J.E. Fawks. This document has been verified by M.E. Harding.

Page 3

,PERRYl Jil-02581SRLR

, Reload 5 Rev.0 Tbc basis for this repon is GeneralElectric Standard Applicationfor Reactor Fuel, NEDE-240ll-P-A-11, November 1995; and the U.S. Supplement, NEDE-24011-P-A-11-US, November 1995.

1. Plant-unique Items Appendix A: Analysis Conditions Appendix B: Basis for Analysis of Loss-of-Feedwater Heating Event Appendix C: Analyzed Operating Domain Appendix D: Transient Analysis Appendix E: Power and Flow Dependent MCPR and MAPLHGR Multipliers Appendix F: GE8x8NB-1 Rotated Bundle Analysis
2. Reload Fuel Bundles Cycle Fuel Type Loaded Number Irradiated:

GE8B-P8SQB322-7GZ-120M-150-T (GE8x8EB) 3 16 GE8B-P8SQB320-9GZ-120M-150-T (GE8x8EB) 3 24 G E 10-P8SX B 306-1 OGZ2-120M-150-T (GE8 x8NB-1) 4 136 GE10-P8SXB 306-11GZ3-120M-150-T (GE8x8NB-1) 4 68 gel 0-P8SXB 306-11GZ3-120M-150-T (GE8x8NB-1) 5 224 Hem GE 10-P8SXB 306-11GZ3-120M-150-T (GE8x8NB-1) 6 44 Gell-P9SUB338-10GZ-120T-146-T (Gell) 6 152 GE11-P9SUB338-120Z-12(TT-146-T (GE11) 6 84 Total 748

3. Reference Core Loading Pattern Nominal previous cycle core average exposure at end of cycle: 25662 mwd /MT

( 23280 mwd /ST)

Minimum previous cycle core average exposure at end of cycle 25551 mwd /MT from cold shutdown considerations: ( 23180 mwd /ST)

Assumed reload cycle core average exposure at beginning of 13084 mwd /MT cycle: ( 11870 mwd /ST)

Assumed reload cycle core average exposure at end of cycle: 25155 mwd /MT

( 22820 mwd /ST)

Reference core loading pattem: Figure 1 Page 4 l

PERRYl J i l-02581SRLR Rev.0

. '. Reload 5

4. Calculated Core Effective Multiplication and Control System Worth - No Voids,20*C Beginning of Cycle, kerrecove Uncontrolled 1.116 Fully controlled 0.952 Strongest control rod out 0.988 R Maximum increase in cold core reactivity with exposure into cycle, Ak 0.000
5. Standby Liquid Control System Shutdown Capability Boron Shutdown Margin (Ak)

(ppm) (20*C, Xenon Free) 660 0.033

6. Reload Unique GETAB Anticipated Operational Occurrences (AOO) Analysis Initial Condition Parameters Exposure: BOC6 to EOC6 Increased core flow /Feedwater temperature 420 F Peaking Factors Fuel Bundle Bundle Initial Design Local Radial Axial R-Factor Power Flow MCPR j (MWt) (1000 lb/hr)

Gell 1.45 1.34 1.57 1.035 6.275 122.2 1.23 GE8x8NB-1 1.20 1.59 1.40 1.000 7.408 114.3 1.19 GE8x8EB 1.20 1.49 1.40 1.051 6.975 118.6 1.16 Exposure: BOC6 to EOC6 Increased core flow /Feedwater temperature reduction to 250*F Peaking Factors Fuel Bundle Bundle Initial Design Local Radial Axial R-Factor Power Flow MCPR (MWt) (1000 lb/hr) gel 1 1.45 1.44 1.57 1.035 6.703 120.0 1.22 GE8x8NB-1 1.20 1.65 1.40 1.000 7.700 111.6 1.20 GE8x8EB 1.20 1.56 1.40 1.051 7.295 115.7 1.17 Page 5

PERRYl J l 1--U.bB 15KLK

, Reload 5 Rev.0

7. Selected Margin Improvement Options Recirculation pump trip: Yes Rod withdrawal limiter: Yes i

Thermal power monitor: Yes Improved scram time: No (ODYN Option B)

Measured scram time: No Exposure dependent limits: No Exposure points analyzed: 1

8. Operating Flexibility Options Single-loop operation: Yes Load line limit: No Extended load line limit: No Maximum extended load line limit: No Increased core flow throughout cycle: Yes 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 / relief valves OOS: Yes

- ADS OOS: No EOC RPT OOS: No Main steam isolation valves OOS: No Page 6

PERRY 1 Jil-02581SRLR

', Reload 5 Rev.0

9. Core-wide AOO Analysis Results Methods used: GEMINI; GEXL-PLUS Exposure range: BOC6 to EOC6 Increased core flow /Feedwater temperature 420 F Uncorrected ACPR Event Flux Q/A Gell GE8x8NB-1 GE8x8EB Fig.

(%NBR) (%NBR)

Load Reject w/o Bypass 332 112 0.15 0.12 0.09 2 Loss of 100 F Feedwater 115 115 0.12 0.12 0.12 -

Heating Exposure range: BOC6 to EOC6 Increased core flow /Feedwater temperature reduction to 250 F Uncorrected ACPR Event Flux Q/A Gell G E8x8NB-1 GE8x8EB Fig.

(%NBR) (%NBR)

FW Controller Failure 244 117 0.15 0.12 0.11 3

10. Local Rod Withdrawal Error (With Limiting Instrument Failure) AOO Summary The generic bounding BWR/6 rod withdrawal crTor analysis described in NEDE-24011-P-A-US is not ap-plied. A cycle-specific rod withdrawal analysis found the ACPR to be 0.12 based upon a one foot withdrawal, and is not bounded by the generic RWE analysis reported in the referenced report.
11. Cycle MCPR Values!

Safety limit: 1.07 Single loop operation safety limit: 1.08 Non-Pressurization events:

Exposure range: BOC6 to EOC6 Option A Gell GE8x8NB-1 GE8x8EB Rod Withdrawal Enor 1.19 1.19 1.19 Fuel Loading Error ( misoriented ) ( see App. F ) 1.22 1.23 1.21 Fuel Loading Error ( mislocated ) 1.22 1.22 1.22 Loss of 100*F Feedwater Heating 1.19 1.19 1.19 Page 7

1 i

.PERRYl J11-02581SRLR

. . Reload 5 Rev.0 l Pressurization events:2  ;

Exposure range: BOC6 to EOC6 Increased : ore flow /Feedwater temperature 420 F Exposure point: EOC6 i Option A ,

Gell . GE8x8NB-1 GE8x8EB )

Load Reject w/o Bypass 1.24 1.20 1.17 l

Exposure range: BOC6 to EOC6 Increased core flow /Feedwater temperature reduction to 250 F i Exposure point: EOC6 Option A Gell GE8x8NB-1 GE8x8EB FW Controller Failure 1.23 1.21 1.19

12. Overpressurization Analysis Summary 3 Psl Py Plant Event (psig) (psig) Response MSIV Closure (Flux Scram) 1264 1294 Figure 4
13. Loading Error Resultsd Variable water gap misoriented bundle analysis: Yes Mislocated bundle analysis: Yes ACPR Event GE11 GE8x8NB-1 GE8x8EB Fuelloading error (misoriented)(see App.F) 0.15 0.16 0.14 Fuelloading error ( mislocated) 0.15 0.15 0.15
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 approval is documented in NEDE-24011-P-A-US.
1. For single-4oop operauon. the MCPR operating limit is 0.01 greater than the two-loop value. The MCPR limit does not change because of channel bow. Channel bow is refleaed in the monitonng of the core.
2. ECCs MCPR value is 1.17.
3. The MsIV closure (flux scram) analysis is performed using GEMN1 methods at the 102% power level to account for the power level un-certainties specified in Regulatory Guide 1.49. The dome pressure is ses to 1045psig as specified in the OPI-3 Design Guide,463HA247 Revision 2. The analysis was performed with the 13 highest setpoint safety valves operauonal.
4. Delta CPR penality of 0.02 for the tihed misonented bundle has been apphed.

Page 8

PERRYl J i l-02581SRLR

, Reload 5 Rev.O i

l

15. Stability Analysis Results GE SIL-380 recommendations have been included in the operating procedures; therefore, no stability analy-sis is required. NRC approval for deletion of a cycle-specific stability analysis is documented in NEDE-240ll-P-A-US. This plant recognizes the issuance of NRC Bulletin No. 88-07, Supplement 1, Power Oscillations in Boiling WaterReactors (BWRs), and will comply with the recommendations contained therein.
16. Loss-of-Coolant Accident Results LOCA method used: SAFE /REFLOOD The peak clad temperature (PCT) is $ 2185'F at all exposures; the local oxidation (fraction) is s 0.%5 at l all exposures. The core-wide metal water reaction is 0.20%. The MAPLHGR multiplier for single-loop op-eration (SLO) is 0.78. The MAPLHG Rs for GE10-P8SXB 306-l l GZ3-120M-150-T are contained in docu-  !

ment 23A7227. ,

4 i

i Page 9

i PERRYl J i l-025b l 5KLM l Rev.0 l

. ". Reload 5 l l

l

16. Loss-of-Coolant Accident Results (cont) )

Bundle Type: gel 1-P9SUB338-10GZ-120T-146-T I

Average Planar Exposure MAPLIIGR(kW/ft) PCT Oxidation I (GWd/ST) (GWd/MT) Most Limiting Least Limiting (*F) Fraction I 0.00 0.00 18.65 11.67 - -

0.20 0.22 11.69 11.69 2179 0.065 l

1.00 1.10 11.79 11.79 2175 0.063 1 2.00 2.20 11.83 11.83 - -

i 3.00 3.31 11.86 11.86 - -

l 4.00 4.41 11.90 11.90 - -

5.00 5.51 11.93 11.93 2179 0.063 6.00 6.61 11.93 11.93 - -

7.00 7.72 11.93 11.93 - -

8.00 8.82 11.93 11.93 2180 0.063 9.00 9.92 11.92 11.92 - -

10.00 11.02 11.90 11.90 2179 0.063 12.50 13.78 11.81 11.81 2180 0.062 15.00 16.53 11.76 11.76 2185 0.064 17.50 19.29 11.72 11.72 - -

20.00 22.05 11.67 11.67 2182 0.062 25.00 27.56 11.32 11.43 2148 0.056 30.00 33.07 10.70 10.71 - -

35.00 38.58 10.02 10.08 1970 0.031 40.00 44.09 9.34 9.42 - -

45.00 49.60 8.68 8.74 1780 0.015 50.00 55.12 8.00 8.04 - -

55.00 60.63 7.25 7.35 1557 0.002 59.I7 65.22 6.62 6.72 - -

59.26 65.32 - 6.71 - -

59.73 65.84 - 6.64 - -

l Page 10

. . _ ~ - . . . _ _ _ _ ._ . _ _ , _ _ __. __ . _

PERRYl J i l-02581SRLR

, Reload 5 Rev.0

16. Loss-of-Coolant Accident Results (cont)

Bundle Type: Gell-P9SUB338-12GZ-120T-146-T Average Planar Exposure MAPLHGR(kW/ft) PCT Oxidation (GWd/ST) (GWd/MT) Most Limiting Least Limiting ( F) Fraction 0.00 0.00 11.50 11.50 - -

ii 0.20 0.22 11.53 11.53 2148 0.059 l.00 1.10 11.67 11.67 2153 0.059 2.00 2.20 11.76 11.76 - -

3.00 3.31 11.84 11.84 - -

4.00 4.41 11.93 11.93 - -

12.01 2175 0.062 l 5.00 5.51 12.01 6.00 6.61 11.99 11.99 - -

I 7.00 7.72 11.96 11.96 - -

8.00 8.82 11.94 11.94 2180 0.064 9.00 9.92 11.91 11.91 - -

10.00 11.02 11.88 11.88 2180 0.063 12.50 13.78 11.79 11.79 2180 0.%3 15.00 16.53 11.74 11.74 2182 0.064 l 17.50 19.29 11.71 11.71 - -

20.00 22.05 11.67 11.67 2184 0.064 25.00 27.56 11.31 11.42 2153 0.057 30.00 33.07 10.70 10.71 - -

35.00 38.58 10.01 10.07 1970 0.031 40.00 44.09 9.33 9.42 - -

45.00 49.60 8.67 8.73 1779 0.015 50.00 55,12 7.99 8.03 - -

55.00 60.63 7.25 7.34 1559 0.002 59.15 65.20 6.61 6.71 - -

59.17 65.22 - 6.71 - -

59.67 65.78 -

6.63 - -

Page11

, $$'a'd' 5' J l1--02581g

.8+s8888888888888888m ll m M M BsBsBsB8M M BsBEB8.

x M M M M E+8 988+88+Es+88+Es+s s+8 8sx lLess+sMMBsMBsMs+Es+EBEBa8+8 Bass

MBEMMBEMB8MBENs+sBsM8+ss+E it:MEMMBEMBENE+EMMMBBBBBs  !

ll:BsBEMMMMMBEMMMBsBsBEs+8

BsM M BEBsBsBs H BsBsBsBs M BEBs i ll:MMMMHHBEMMMMBsMMBE

':-*8MMMBEMMMMMBsBEMM85 .

*BsBsMMMMMMMBsBss+sBa*
58888888+8888888M88888+8" l1 "MBEMs+sBsBEMBEM*
    • sMBEM IIIIIIIIl 11I M Ms**

1 3 5 7 6 11 15 16 17 il 21 23 25 27 26 31 33 35 37 39 41 43 45 47 49 51 53 55 57 59 Fuel Type BN!!!INsh!!!bbbi oNI l I lh cl !! I$hlIP s 06- G 0Z y Edilt"lilitiiSEli": lit? 18;:l: ! ll$ill:"sa!!tl?8tlM:llt? lE;;l: !

Figure 1 Reference Core Loading Pattern i

Page 12

PERRYl J i l-4)2581SRLR

. -Reload 5 Rev.0 i

Neutron Flux Vessel Press Rise (psi)

- - - Ave Surface Heat Flux - - - - - Safety Valve Flow l 150.0 -

- -- Core inlet Flow 300.0 - --- Relief Valve Flow '

- - - Bypass Valve Flow i,

A -'

g 100.0 N ,y T ,'4 200.0 --

E \ 'T N s W

E E

--/-, $

100.0 50.0 -

7--,

,~ . \

g I

\

I

\ '

I \ <

' ' I d 0.0 O.0 0.0 3.0 6.0 0.0 3.0 60 l Time (sec) Time (sec) l Level (incNREF-SEP-SKRT) eactivity

- - - Vessel Steam Flow ----- Dop er Reactivity  ;

200.0 - --- Turbine Steam Flow 1.0 - - - - Scra Reactivity

--- Feedwater Flow - Total activity .-

g w ,'\ , . ,

$ \. ..

g ,.-

g 100.0 g ,s, ,,

k 0.0 ,

y l,

'% s.'

.,-.---~~,,,,.:,~~~

G ,.**

g e, i O

[ ,, *. . . O \

N l. ,' ', s

. p ha . ,' '

.2:

D 0.0 Qd-[ ----------

N. -1.0

, E h, g

\h s.

_ t oo.o I -

2.0

\\

'1 I -

0.0 3.0 6.0 0.0 3.0 6.0 Time (sec) Time (sec)

Figure 2 Plant Response to Load Reject w/o Bypass (BOC6 to EOC6 Increased core flow /Feedwater temperature 420 F)

Page 13 1

1 PERRYl J i l-02581SRLR <

. Reload 5 Rev.0 i

i t

i Neutron Flux Vessel Press Rise (psi)

- Ave Surface Heat Flux - - - - - Safety Valve Flow 150.0 - --- Core inlet Flow 125.0 - --- Relief Valve Flow

- - - Core inlet Subcooling --- Bypass Valve Flow 100 0 #'

      • ~e"~d &'i - '

75.0 -

Ib &

w Ss w C '.N g C Y 's, $

  • ~ ~ ~ * - -

50 0 -

25.0 -

i e

. _ = _ . .a ___

l 0.0 ' ' ' I

- 25.0 '

0.0 6.0 12.0 0.0 6.0 12.0 ,

Time (sec) Time (sec)

Level (inch-REF-SEP-SKRT) Void Reactivity

- - - - Vessel Steam Flow - - - - Doppler Reactivity 150.0  :-. --- _Iutbine Staa01Elow. , 1.0 - --- Scram Reactivity

--- Feedwater Flow

( --- Total Reactivity j

\ G w . [,

1 g 100.0

\ .

1o 0.0 - -- W J --,-r.--,-,---- ,

l

\

    • . ,o g

C _. -

,, ,\

  1. ll .",
  1. I

.. .a 1 15 I 50.0 -

l . .

= -1.0 -

l i: : : . . '

e { l

. 1:: .: - ~... .

1 ll'.l \

' I -20 I ' I 0.0 0.0 6.0 12.0 00 6.0 12.0 Time (sec) Time (sec)

Figure 3 Plant Response to FW Controller Failure (BOC6 to EOC6 Increased core flow /Feedwater temperature reduction to 250 F)

Page 14

I j

PERRYl J il-02581SRLR

' -Reload 5

. . . _ Rev. 0 N utron Flux Vessel Press Rise (psi)

A Surface Heat Flux - - - - - Safety Valve Flow j 150.0 - ---. Co e inlet Flow 300.0 - '--- Relief Valve Flow

-- - Bypass Valve Flow

~

/s y ',.

g 100.0 N ',

p200.0 w N. w

[ s- [

Y $

%. ?.

~ .

50.0 -

., 100.0 -

l .

0.0 '

  • I 0.0 0.0 4.0 80 0.0 4.0 8.0 Time (sec) Time (sec) i Level (inch-REF-SEP-SKRT) Void Reactivi

- - - - - Vessel Steam Flow - - -

Doppler Re i ty 200.0 - --- Turtnne Steam Flow 1.0 - - - - Scr R ctivi

--- Feedwater Flow --- o ctivity n ./

se .

\

we \ -

100.0 - 0.0 , ,

g .

g

'\ y ".

j

' . g *} , -

g - ' .

00

%.l,':q, ~'-

N

.k

- \\ '\

R. -1.0 g\

z 4

\.

l

\\\'

-300.0 . 1 -

-2.0

l '

OO 4.0 8.0 0.0 4.0 80 Time (sec) Time (sec) I t

Figure 4 Plant Response to MSIV Closure (Flux Scram)

Page 15

PERRYl J i l-02581SRLR

.

  • Reload 5 Rev.0 4

Appendix A Analysis Conditions To reflect actual plant parameters accurately, the values shown in Table A-1 were used this cycle.5 l

Table A-1 1

Increased core flow /Feedwater temperature 420 F i Parameter Analysis Value Thermal power, MWt 3579.0  ;

Core flow, Mlb/hr 109.2 I Reactor pressure, psia 1056.0  ;

Inlet enthalpy, BTU /lb 528.8 I Non-fuel power fraction 0.038 Steam flow analysis, Mlb/hr 15.41 Dome pressure, psig 1025.0 Turbine pressure, psig 974.8 g No. of Dual Mode S/R Valves ( see footnote 5 ) 19 Relief mode lowest setpoint, psig ( see footnote 5 ) 1133.0 Safety mode lowest setpoint, psig ( see footnote 5 ) 1200.0 Increased core flow /Feedwater temperature reduction to 250*F Parameter Analysis Value Thermal power, MWt 3579.0 Core flow, Mlb/hr 109.2 Reactor pressure, psia 1023.7 Inlet enthalpy, BTU /lb 508.4 Non-fuel power fraction 0.038 Steam flow analysis, Mlb/hr 12.59 l Dome pressure, psig 994.6 l Turbine pressure, psig 960.3 No. of Dual Mode S/R Valves ( see footnote 5 ) 19 Relief mode lowest setpoint, psig ( see footnote 5 ) 1133.0 Safety mode lowest setpoint, psig ( see footnote 5 ) 1200.0

5. There are a total of 19 valves, the two lowest setpoint safety /rchef valves are assurned to be out-of-seruce in the transient analysis. For the MsIVFS overpressunzation analysis. 6 safety valves are assumed out-of-semce.

Page 16

PERRYI J i i--vad i d KLn

,, . Reload 5 Rev.0 l

Appendix B Basis for Analysis of Loss-of-Feedwater Heating Event l The loss of feedwater heating event was analyzed with the 3D BWR simulator code described in i NEDE-24011-P-A, which permits 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 in this document.

The transient analysis inputs normally reported in Section 6 of this document are internally calculated in the 3D BWR simulator code.

1 I

l Page 17

- . .~ _

l .PERRYI J il-0258]SRLR )

,o - Reload 5 Rev.0 e .- -  ;

Appendix C  :

Analyzed Operating Domain i

The core-wide adnormal operational occurrence (AOO) analysis results reponed in Section 9 ere the most limiting values over the entire allowable operating range. This range covers the following operating options: ;

l. Standard 100% power / flow map; 1

i 2. End-of-<ycle power coastdown;

3. MEOD with 100% power, flow range from 75% to 105% of rated; and
4. Partial feedwater heating to 320*F during the cycle with final feedwater temperature l

, reduction to 250*F after All Rods Our at end of cycle.  ;

i I 8.niting events and conditions analyzed are based on NEDE-240ll-P-A-US and the US AR analytical re-euits. The Reload 5/ Cycle 6 analyses were performed assuming all four turbine control valves in a full arc  :

' mode of operatio,. This is conservative for panial arc configuration.  !

I i

t Page 18

.. . - . . _ . . -- _ _ - - - .-- -. . . - . - ~-.

PERRYI J i l-02581 Sr,tg

, .. Reload 5  :.ev. 0 Appendix D Transient Analyses i

i I

i The turbine trip without bypass (TfNBP) analysis is a pressure increase event that is bounded by the load

! rejection wit'aout bypass (LRNBP) analysis.

l The LRNBP is limiting at normal feedwater temperature and increased core flow.

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 run for the intermediate feedwater temperature cases (320 F and 370*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 with feedwater temperature.

Page 19

PERRYl J i l-02581SRLR

.< Rev.0

, . .' ." Reload 5 Appendix E Power and Flow Dependent MCPR and MAPLHGR Multipliers  ;

l i

The original MEOD offrated MCPR and M APLHGR multipliers, the absolute MCPR power limit below 40%

power, and the absolute MCPR flow limit above the OLMCPR were confirmed to be applicable to this cycle.

The absolute power and now dependent MCPR Gow limit must be limited to the cycle 6 OLMCPR.

l 1

i i

l 1

Page 20

Retrieved from "https://kanterella.com/w/index.php?title=ML20101F334&oldid=2775479"