ML20206A576

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Rev 0 to Supplemental Reload Licensing Submittal for Limerick Generating Station Unit 1,Reload 1
ML20206A576
Person / Time
Site: Limerick Constellation icon.png
Issue date: 02/28/1987
From: Charnley J, Lambert P, Zarbis W
GENERAL ELECTRIC CO.
To:
Shared Package
ML19292H074 List:
References
23A5801, 23A5801-R, 23A5801-R00, NUDOCS 8704080079
Download: ML20206A576 (27)
v • d • e


Text

.O 23A5801 Rsvisicn 0 Class I February 1987

O

~O SUPPLEMENTAL RELdAD LICENSING SUBMITTAL FOR O LIMERICK GENERATING STATION UNIT 1, RELOAD 1 O

Prepare P. A. Lambdrt

O f '

Verified:

O w. A. z als Approvad* - f /

!O . Charnley" load Fuel Licensing er i

i

!O NUCLEAR ENERGY BUSINESS OPERATIONS

  • GENERAL ELECTRIC COMPANY Q SAN JOSE, CALIFORNIA 95125 GENERAL h ELECTRIC 1/2 O 87o40soo79 87o4o3 PDR ADOCh 05000352 P PDR ,
C) 23A5801~ Rsv. 0 )

l IMPORTANI NOTICE REGARDING I C) CONTENTS OF THIS REPORT PLEASE READ CAREFULLY

C) This report was prepared by General Electric solely for Philadelphia Electric Company (PECO) for PECO's use with the U.S. Nuclear Regulatory Com-mission (USNRC) for amending PECO's operating license of the Limerick Gener-ating Station Unit 1. The information contained in this report is believed by lC) General Electric to be an accurate and true representation of the facts known, obtained or provided to General Electric at the time this report was prepared.

The only undertakings of.the General Electric Company respecting informa-i C) tion in this document are contained in the contract between Philadelphia Electric Company and General Electric Company for nuclear fuel and related services for the nuclear system for Limerick Generating Station Unit 1 and nothing contained in this document shall be construed as changing said con-C) tract. The use of this information~except as defined by said contract, or for any purpose other than that for which it is intended, is not authorized; and with respect to any such unauthorized use, neither General Electric nor any of the contributors to this document makes any representation or warranty

- (express or implied) as to the completeness, accuracy or usefulness of the information contained in this document or that such use of such information i 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

. C) information.

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C)

A O 23A5801 Ray. O ACKNOWLEDGMENT O

The engineering and reload licensing analyses which form the technical basis of this Supplemental Reload Licensing Submittal, were performed by M. W.

Smith and O. Raza of the Nuclear Fuel and Engineering Services Department.

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1. PLANT UNIQUE ITEMS (1.0)*

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GEMINI System of Methods Used**

Appendix A: Analysis Conditions Appendix B: Operation with Inoperable RPT

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2. RELOAD FUEL BUNDLES (1.0, 2.0, 3.3.1 AND 4.0) 4 Fuel Type Cycle Loaded Number i

O Irradiated BP8CRB163 (BP8x8R) 1 116 BP8CRB248 (BP8x8R) 1 308 .

O BP8CRB278 (BP8x8R) 1 72 i New BC320A*** (GE8x8EB)**** 2 268 lO .

4 Total 764 O

! *( ) Refers to area of discussion in " General Electric Standard Application for Reactor Fuel," NEDE-24011-P-A-8, dated May 1986. A letter "S" preceding the number refers to the U.S. Supplement, NEDE-24011-P-A-8-US, May 1986.

lO **NRC approval documented in the letter from G. C. Lainas (NRC) to J. S.

Charnley (GE),." Acceptance for Referencing of Licensing Topical Report,

NEDE-24011-P-A, 'GE Generic Licensing Reload Report,' Supplement to Amendment 11," dated March 22, 1986.

!O *** Bundle descriptions contained in the letter from J. S. Charnley (GE) to i

G. C. Lainas (NRC), dated January 29, 1987, i

! ****NRC approval documented in the letter from C. O. Thomas (NRC) to J. S.

Charnley (GE), " Acceptance for Referencing of Licensing Topical Report NEDE-24011-P-A-6, Amendment 10, ' General Electric Standard Application for

O Reactor Fuel'," dated May 28, 1985, and in the letter from H.N. Berkow (NRC) to J. S. Charnley (GE), " Acceptance for Approval of Fuel Designs

( Described in Licensing Topical Report NEDE-24011-P-A-6, Amendment 10 for Extended Burnup Operation," dated December 3, 1985.

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23A5801 Rav. O O

3. REFERENCE CCRE LOADING PATTERN (3.3.1) e e

Nominal previous cycle core average exposure at end of cycle: 8566 mwd /ST Minimum previous cycle core average exposure at end of cycle from cold shutdown considerations: 8166 mwd /ST O

I Assumed reload cycle core average exposure at end of

, cycle: 16494 mwd /ST Core loading pattern: Figure 1 0

4. CALCULATED CORE EFFECTIVE MULTIPLICATION AND CONTROL SYSTEM WORTH - NO VOIDS, 20*C (3.3.2.1.1 and 3.3.2.1.2)

Beginning of Cycle, K-effective g j Uncontrolled 1.118 Fully Controlled 0.955

Strongest Control Rod out 0.980 g R, Maximum Increase in Cold Core Reactivity with 0.010 Exposure into Cycle, Delta K
5. STANDBY LIQUID CONTROL SYSTEM SHUTDOWN CAPABILITY (3.3.2.1.3) g.

1 Shutdown Margin (aK)

EEm (20*C, Xenon Free) l 660 0.036 g:

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C) 23A5801 Rav. 0
6. RELOAD UNIQUE TRANSIENT ANALYSIS INPUT (3.3.2.1.5 AND S.2.2)

C)

(Cold Water Injection Events Only)

() Void Fraction (%) 43.3 Average Fuel Temperature (*F) 1080 Void Coefficient N/A* (d/% Rg) -8.22/-10.28 Doppler Coefficient N/A* (d/*F) -0.249/-0.237

'(3 Scram Worth N/A* ($) **

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C)
C)  ;

l i C)

  • N = Nuclear Input Data, A = Used in Transient Analysis
    • Generic exposure independent values are used as given in " General Electric {

l C) Standard Application for Reactor Fuel," NEDE-24011-P-A-8, dated May 1986. )

() 9

23A5801 Rsv. O O

7. RELOAD UNIQUE GETAB TRANSIENT ANALYSIS INITIAL CONDITION PARAMETERS (S.2.2) i Si Exposure: E0C2 (100% Flow)

Fuel Peaking Factors Bundle Power Bundle Flow Initial Design Local Radial Axial R-Factor (MWt) (1000 lb/hr) MCPR BP8x8R 1.20 1.49 1.40 1.051 6.301 105.0 1.26 8 GE8x8EB 1.20 1.50 1.40 1.051 6.330 106.9 1.26 Exposure: E0C2 Extended Load Line Limit

  • Fuel -Peaking Factors Bundle Power Bundle Flow Initial II i Design Local Radial Arial R-Factor (MWt) (1000 lb/hr) MCPR BP8x8R 1.20 1.47 1.40 1.051 6.195 90.8 1.24 GE8x8EB 1.20 l'.48 1.40 1.051 6.225 92.6 1.24 9

Exposure: Extended EOC2 with Increased Core Flow

  • Fuel Peaking Factors Bundle Power Bundle Flow Initial Design Local Radial Axial R-Factor (MWt) (1000 lb/hr) MCPR BP8x8R 1.20 1.50 1.40 1.051 6.315 110.7 1.27-

'~

GE8x8EB 1.20 1.50 1.40 1.051 6.340 112.7 1.2)

Exposure: Extended E0C2 with Increased Core Flow and Feedwater Temperature Reduction

  • g Fuel Peaking Factors Bundle Power Bundle Flow Initial Design Local Radial Axial R-Fsetor (MWt) (1000 lb/hr) MCPR BP8x8R 1.20 1.50 1.40 1.051 6.315 110.7 1.27 GE8x8EB 1.20 1.50 1.40 1.051 6.340 112.7 1.27 e O
  • See Appendix A 9

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8. SELECTED MARGIN IMPROVEMENT ~ OPTIONS (S.2.2.2) .

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Transient Recategorization: No Recirculation Pump Trip: Yes Rod Withdrawal Limiter: No

O Thermal Power Monitor: No Improved Scram Time: Yes (ODYN Option B)

Exposure Dependent Limits: No Exposure Points Analyzed: 1 O

9. OPERATING FLEXIBILITY OPTIONS (S.2.2.3)

Single-Loop Operation: Yes

O 'Aad Line Limit: No Extended Load Line Limit: Yes (See Appendix A)

Increased Core Flow: Yes (See Appendix A)

Feedwater Temperature Reduction: Yes (See Appendix A)

O Feedwater Heaters Out-of-Service: Yes (See Appendix A)

ARTS Program: No Maximum Extended Operating Domain: No

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23A5801 Rsv. 0

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10. CORE-WIDE TRANSIENT ANALYSIS RESULTS (S.2.2.1)

Flux ACPR Q/A Transient (% NBR) (% NBR) BP8x8R GE8x8EB Figure Exposure Range: All Cycle 2 Exposures (100% Flow)*

Load aejection w/o Bypass ** 437 118 0.16 0.16 2 9 Loss of 100*F Feedwater 121 119 0.15 0.14 3 Heating **

Feedwater Controller Failure 464 122 0.19 0.19 4 9

Flux ACPR Q/A Transient (% NBR) (% NBR) BP8x8R GE8x8EB Figure Exposure Range: All Cycle 2 Exposures (Extended Load Line Limit)*

O Feedwater Controller Failure 384 121 0.17 0.17 5 Flux Q/A ACPR Transient (% NBR) (% NBR) BP8x8R GE8x8EB Figure Exposure Range: All Cycle 2 Exposures with Increased Core Flow

G:

  • See Appendix A (p

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11. LOCAL ROD WITHDRAWAL ERROR (WITH LIMITING INSTRUMENT FAILURE) TRANSIENT
O (s e s.2.2.1)

Limiting Rod Pattern: Figure 8 Rod Porition ACPR MLHGR (kW/ft) s

' Rod Block (feet Reading withdrawn) BP8x8R GE8x8EB BP8x8R GE8x8EB 104 4.5 0.11 0.11 14.54 14.54 105 5.5 01.13 0.13 15.16 15.16 106 6.0 0.14 0.14 15.41 15.41 C) 107 9.5 0.19 0.19 16.14 16.14 Setpoint Selected: 107 0

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12. CYCLE MCPR VALUES (S.2.2 and S.2.5.4)

Non-Pressurization Events Exposure Range: All Cycle 2 Exposures (100% Flow)* i BP8x8R GE8x8EB Loss of 100*F Feedwater Heating 1.22 1.21 Rod Withdrawal Error 1.26 1.26 O

Pressurization Events Exposure Range: All Cycle 2 Exposures (100% Flow)*

Option A Option B G BP8x8R GE8x8EB BP8x8R GE8x8EB Loc' Rejection Without Bypass 1.28 1.28 1.24 1.24 Fee ater Controller Failure 1.30 1.30 1.28 1.27 Exposure Range: All Cycle 2 Exposures (Extended Load Line Limit)*

l l

Option A Option B BP8x8R GE8x8EB BP8x8R GE8x8EB g Feedwater Controller Failure 1.28 1.28 1.26 1.26 Exposure P,snge: All Cycle 2 Exposures with Increased Core Flow

Exposure Range: All Cycle 2 Exposures with Increased Core Flow and I Feedwater Temperature Reduction *,**

Option A Option B g BP8x8R GE8r8EB B_P8x8R GE8x8EB Feedwater Controller Failure 1.31 1.31 1.29 1.29 9

  • See Appendix A.
    • These limits bound operation with Feedwater Heaters Cat-of-Service (FH00S) throughout Cycle 2; see Appendix A.

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13. OVERPRESSURIZATION ANALYSIS

SUMMARY

(S.2.3)

!O P

al #v Trancient (psig) (psig) Plant Response MSIV Closure 1230 1265 Figure 9

!O (Flur Scram)

14. CONTROL ROD DROP ANALYSIS RESULTS (S.2.5.1)
O Banked Position Withdrawal Sequence is utilized at the Limerick Generating Station Unit 1; therefore, the Control Rod Drop Analysis (CRDA) is not required. NRC approval for deletion of a cycle specific CRDA *s documented in Amendment 9 to NEDE-24011-P-A-8-US, May 1986.

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15. STABILITY ANALYSIS RESULTS (S.2.4)

GE SIL-380 recommendations have been included in the Limerick Generating O Station Unit 1 operating procedures and/or Technical Specifications; therefore, no stability analysis is required. NRC approval for deletion of a cycle specific stability analysis is documented in Amendment 8 to NEDE-24011-P-A-8-US.

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23A5801 Rsv. O O

16. ' LOSS-OF-COOLANT ACCIDENT RESULT (S.2.5.2) 9 LOCA Method Used: SAFE /REFLOOD.

Technical Specification MAPLHGk Limits for BC320A Fuel Bundle O

MAPLHGR (kW/ft)

Exposure Most Least Oxidation (GWd/ST) Limiting

  • Limiting
  • PCT (*F)** Fraction **

0.2 11.52 11.57 1927 0.022 g 2.0 11.76 11.84 1954 0.025  !

4.0 12.07 12.21 1993 0.028 j 6.0 12.44 12.63 2033 0.031 i

?O 12.67 12.72 2041 0.032 e I 10.0 12.77 12.86 2055 0.033 12.5 12.78 12.84 2049 0.032 15.0 12.56 12.58 2011 0.028 20.0 12.04 12.06 1925 0.021 GD -

50.0 6.86 6.88 1349 0.001 An analysis consistent with NEDC-31300, " Single-Loop Operation Analysis for Limerick Generating Station Unit 1," September 1986, results in a O Single-Loop Operation MAPLHGR multiplier of 0.89 for the BC320A (GEGx8EB) fuel bundle.

MAPLHGR limits for all lattices (including the natural lattice) are O provided in NED0-31401, " Basis of MAPLHGR Technical Specifications for Limerick Unit 1," February 1987.

O 9

  • Enriched lattices only. Natural lattices are non-limiting for LOCA events.
    • Maximum lattice values.

16 9

O 23AS801 R;v. 0 o

sMMMMMMHs
!MMMMMMMMMR O
ssM M M M M M M M M M Mes
sMMMMMMMMMMMMMs
MMMMMMMMMMMMMMM o :::MMMMMMMMMMMMMMM
M M M M M M M M M M M M M M M
MMMMMMMMMMMMMMM
M E M M M M M M M M M M M M M
MMMMMMMMMMMMMMM o ':: M M M M M M M M M M M M M M M ll "MMMMMMMMMMMMM*
    • MMMMMMMMMMM**

l iMMMMMMMMMi

"MMMMMMM" lIIIIIIIIIIIII O 1 3 5 7 911131517192123252729313335373941434547495153555759 FUEL TYPE O

A = BP8CRB163 C = BP8CRB278 B = BP8CRB248 D = BC320A O Figure 1. Reference Core Loading Pattern 17 O

l 23A5801 R;v. 0 .O i

j 1

1 NEUTRCN FLUK 1 VESSEL PRE h RISE (PSI! i 2 AVE SURFACE HEAT FLUX 2 SAFETY VALVE FLOW 150.0 310.0 i

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2' Iss. 3 p -

288.0 E

5 E s .. . ~1 N N

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s. 0 0. 0 _ , , _

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0. 9 2.0 4.0 6.1 0. 0 2. 8 4.I 6.0 g'~

TIME (SECONDS 1 TIME (SECONCSI 1 LEVEL (INCH-PEF.SEP.SART) 1VOIDREACTIh!TY 2,...

2 VESSEL STEANFLOW

? S"* EPPZ'o" i..

2 OcPPLER RE ALTIVITY 2

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O-Figure 2. Plant Response to Generator ',oad Rejection Without Bypass 18 9

c 23A5801 Rev. 0

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1

'150.8

, 1 NEU RON FLUX 1 VESSEL PRESS RISEtPSI)

AVE SURFACE HEAT FLUX 2 RELIEF VALVE FLCW

O is .. . A' P 15g S2v 2 8'e'55 va'vc r'ov I

100.0

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I ?I E 5 $!!!!'!REACT!vtTY is.. . i..

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Figure 3." Plant Response to Loss of 100'F Feedwater Heating l'

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23A5801 Rev. O O-O is.. .

I hEtRON FLUE 1 VES L ESS RISE (PSI) 2 AV! $URFACE HEAT FLUX 2 $AF Y4 LVE FLOW 3 C0al: IPLET TLow 3 REL F V LVE FLOW l 138.9 aemr '"E' "D2 4 BYP/ $V VE FLOW g I99.9 ^

y e I ... . -- --== M. -

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8. 8 18.8 20.0 38.8 0. 0 10.0 28.4 30.0 t!=E (SECONOSI TIME (SECONOS)

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! LEVEL (INCH-NEF.SEP.SMRT) 1 VOIC AEACTIN TV 2 WESSEL STEANFL0w 2 DCFPLER RE IVITY i,... 2 3_"!_!1_PP7_?" -

i.e 2 Sct 55 vit e .

i....

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c A 9 y;

se.e , i.e ,

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t_ 2. .

... i s. . 2 e. . n. ... i s. . 2.. as. .

Time (SECO*Cl) f!ME (SECONOS2 g

91 Figure 4. Plant Response to Feedwater Controller Failure 20 9:

O 23A5801 Rrov. O O

e is ..

I NELIRON FLUC 1 VESS :L ESB RISEtPSI1 2 AVE SURFACE HEAT Fl.UX 2 SAFE 'Y V LVE FLOV 3 CDs l INLET  ; LOW J RELI EF V LVE FLCW 1s 0. 8 e 7= ' m_ P n_= 4 BYPA SS V VE FLOW O x

/ s .. .  ;

100.0 *M s  :  :  :  :

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Q g 5

t' 4

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8. 0 10.0 20.0 30.9 f. 0 18.8 29.4 30.0 TIPE (SECcWS] TIME (SE34SI
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I LEVEL C INCH->EF-SEP-SMRT) I V01 EACTI ' TV 2 WESSEL STEA SLOW 4 REA T

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rinE csEcc.cs: Ti E isEce =S:

O Figure 5. Plant Response to Feedwater Controller Failure (Extended Load Line Limit Analysis) 0- 21

23A5801 R;v. O O

O 150. e 1 NEERON FLUC 1 VESSEL ESS RISE (PSI) 2 AVE SURFACE HEAT FLUX 2 SAFE rY V LvE FLOW f 3 CONI' Itt.ET LOW 3 RELIEF V LVE FLOW IS O. e 2 mi tu_E' t'* 4 BYPASS V VE FLOW

( -

s ee. e A

M 100.0 'M" -

a b

  • - 50.0 5 9 2

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mu e  ; ;:= = :::  :

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l. e 18.8 20.0 3 e. 0 0. 8 10.0 20.0 30.0 IIPE (SECON051 TIME (SECCCS) 9 1 LEVEL (INCH-:EF-SEP-SKRf3 1 v0! 'EACT!k TV 2 VESSEL STEAMFLOW 2 OOP R RE IVITY is o. e 3$!De_yQ OW g,, }- gjg J 3 b

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e. e is.e 2e. e s e. 0 e. . i e. . ii. e 3e.e g f!PE (SECCWS3 TIME (SECCWSI 1

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Figure 6.

9 Plant Response to Feedwater Controller Failure (Increased Core Flow) 22 g i

. . . . - . . . . . E

. - - . - - . a - -

______d

'O 23A5801 R;v. 0

O is...

1 NEthRON FLur i VESSEL ' ESS RISE (PSI) 2 AVE' SURFACE HEAT FLUX 2 SAFE rY V LvE FLOW 3 CGHl: IEET LOW EF V LV E FLOW O is s. c c- . - - =m 3. avP REL,ISv vc reev

/ s...o A a _ N 2 180.0 "N" E

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!LZvCLCINCH.htr.SEP.SKRT) i VO! hEACT! TV 2 viSSEL STEA 4*tCW 2 DCP L ER RE IVITY 3 TufiGINE STEW LCW 3 SCP P EEACT! ITY t$ 8. 0 ' crE y arEn c_& 3.8 = tcr a. :en e vrev

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0. 8 10.0 20.8 38.8 0. 0 19.0 20.0 30.8 TIME (SECCNDS) f!ME (SEC0405)

O Figure 7. Plant Response to Feedwater Controller Failure (Increased Core Flow and Feedwater Temperature Reduction)

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23A5801 Rev. O O O

2 6 -10 14 18 22 26 30 34 38 42 46 50 54 58 e

59 38 38 55 0 51 0 10 10 0 47 0 43 38 0 8 8 0 38 39 0 0 g

35 to 0 0 10 31 0 0 0 0 27 10 0 0 10 g 23 0 0 19 38 0 8 8 0 38 15 0 g 11 0 10 10 0 7 0 3 38 38 g NOTES: 1. NIDiBER INDICATES NUMBER OF NOTCHES WITHDRAWN OUT OF 48.

BLANK IS A WITHDRAWN ROD.

2. ERROR ROD IS (26,35).

g I

L 9;

Figure 8. Limiting Rod Pattern for Rod Withdrawa. Error Analysis 24

,l

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O 23A5801 Rev. 0 4

lO l

1 NEUTRON F_UX 1 VESSEL PRESS RISE (PSI)-

2 AVE SUFFAI M'AT FLUX 2 SAFETY VALVE FLOW 150.0 330.0 h h be

>0 Nd 5 i ... . 2 ... N 4

i 0

8 g

5 W

)NN r

- se. s i .e. e I

10 1

... .. ..e _- L _ _ ..

0. 0 5.0 0. I 5.0 T!"E (SECONOS1 TIPE (SECONOS1

!O 1 LEVEL (INC4 EEF-SEP-SKRT) V010 REAC!TIVITY 2 VESSEL STEAMFLOW 00PPLER PE ACTIV!TY

2. .. . 2 gs!sp petov guigg535;;

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.= s.e . -

V f

v Em A

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=

v .i..

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W W

-t e s. s 2.8 I

'O ... s.a ... S. l f tPE (SEC08e05) TIME (SECO W S)

]

O Figure 9. Plant Response to MSIV Closure (Flux Scram) 25/26 O

10 23A580_1 Rsv. O I

APPENDIX A ANALYSIS CONDITIONS lQ To provide Limerick Generating Station Unit I with operating improve-l ments, expanded operating domain analyses were performed for extended load lO line limit (trirA) operation, for increased core flow (ICF) up to 105% of

rated flow, for final feedwater temperature reduction (FFWTR) to a temperature (at full power) of 360*F, and for operation with feedwater heaters out-of-

!- service (FH00S) to a temperature reduction of 60*F.-

!O i

Coastdown operation beyond full power operation is conservatively bounded by the MCPR operating limits given in Section 12 of this document at the I apnlicable core flow and feedwater temperature conditions and in the expanded jO operating domain.

1 i

l 100% Flow i

l The Load Rejection without Bypass event and the Feedwater Controller

]O Failure event were analyzed at 100% power and 100% flow in conformance with the NRC approved procedure for application of the GEMINI set of methods. The *

! Loss of 100*F Feedwater Heating event and the MSIV Closure-Flur Scram event lO were analyzed at 102% power and 100% flow.

Extended Load Line Limit t

O Operation in the ELLLA region was justified for Limerick Generating Sta-

! tion Unit 1 in Reference A-1. To confirm operation in the ELLLA region (expanded operating domain bounded by the Average Power Range Monitor (APRM) rod block line, the rated power line, and the rated load line) during Cycle 2,

O 2 the limiting event (Feedwater Controller Failure) was analyzed at the nominal  ;

i end-of-cycle (EOC) 2 exposure at the 100% power /87% flow point. The resulting l

' f j ACPR is reported in Section 10 of this document. No change to the operating limit MCPR is required to operate in the ELLLA region. The APRM rod block line lO (i.e. , 0.58 W + 50%, where W is recirculation drive flow in percent of rated) i i

! 27 10 1

23A5801 Rty. O O given in Reference A-1 is applicable to Limerick Generating Station Unit 1 Cycle 2 operation. Similarly, the APRM scram line is given as 0.58 W + 62%. gp Increaaed Core Flow and Final Feedwater Temperature Reduction Operation with ICF throughout the operating cycle, and with ICF and with 9-FFWTR at the end of the operating cycle, was justified for Limerick Generating Station Unit 1 in Reference A-2. For Cycle 2, the limiting event (Feedwater Controller Failure) was analyzed at ICF conditions (100% power /105% flow) to establish the operating limit MCPR for operation with ICF throughout the cycle. ID Analyses for cycle extension with ICF and FFWTR were performed at the extended end-of-cycle (EEOC)* 2 exposure point. The limiting event (Feedwater Controller Failure) was analyzed to establish the operating limit MCPR for II operation with ICF and FFWTR between EOC2 and the remainder of the cycle.

Feedwater Heaters Out-of-Service 9

Operation with FH00S throughout the operating cycle is justified in Reference A-2. For Cycle 2, the operating limit MCPR determined at ICF with i FFWTR conditions bounds operation with FH00S. Therefore, operation with FH00S, with or without ICF, is justified throughout Cycle 2, including the cycle extension to EEOC2.

1 I GD i 0-9

  • EEOC is the assumed cycle core average exposure beyond the end of full power rated operation attainable, at full power, using ICF and FFWTR.

28 (p

O 23A5801 Rsv. O REFERENCES
O A-1. NZDC-31139, " General Electric Boiling Water Reactor Extended Load Line Limit Analysis for Limerick Generating Station Unit 1, Cycle 1," April 1986.
O i A-2. NEDC-31323, " Increased Core Flow and Partial Feedwater Heating Analysis for Limerick Generating Station Unit 1, Cycle 1," October 1986.
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i 29/30

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50- 23A5801 Rsv. 0 i

,O APPENDIX B OPERATION WITH INOPERABLE RPI l The analyses reported in the body of this document have been performed on lO the basis of operable Recirculation Pump Trip (RPT). The limiting event (Feedwater Controller Failure) has been conservatively analyzed on the basis of inoperable RPT, at Increased Core Flow with Final Feedwater Temperature Reduction. The resulting ACPR is 0.26*, resulting in an operating limit lO MCPR of 1.38* (Option A) and 1.35* (Option B). Operation with inoperable RPT

.is permissible throughout Cycle 2 with Increased Core Flow and reduced feed-

) water temperature (due to either Final Feedwater Temperature Reduction or

- Feedwater Heaters Out-of-Service) using the aforementioned limits.

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j *The results and limits apply to both the BP8x8R and GE8x8E8 fuel designs. l l

i 31 l lO  :

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1

23A5801 Rsv. O O

e i s e. .

1 net fRth Flut i VESS ESS RISECPSI) 2 AVI SLN" ACE FE AT FLUX 2 SAFEYVLVE FLOW 3 C01': 14_ET 10w 3 RELIEF V LVE FLOW 150.0 'ml

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... .... .... .... ... .... .... .... j TIME (SECONOS) f!ME (SEC080$1  ;

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e 1 LEVEL (!NCH-4EF-SEP.SKRT) 1 V013 ~ ACT! TY l 2 VESSEL STEA9 FLOW 2 00F* REA Iv!TV 1s 8. 0 e I.8 e i'

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8. 9 L 4. 8 28.0 30.0 8. O t 8. 0 26. t 30. t fine isECON0f1 f!RE (SECOIS$3 h' Figure B-1. Plant Response to Feedwater Controller Failure (Increased O Core Flow and Feedwater Temperature Reduction with Inoperable RPT) 32' (FINAL) e

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