ML20086K499

Cycle 6 Reload Analysis

ML20086K499
Person / Time
Site:	Grand Gulf
Issue date:	10/31/1991
From:	Garner N, Hibbard M, Roberts C SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER
To:
Shared Package
ML19353B429	List: ML20086K407 ML20086K440 ML20086K461 ML20086K465 ML20086K474 ML20086K499 ML20086K511 ML20086K519
References
EMF-91-169, NUDOCS 9112130140
Download: ML20086K499 (36)

Text

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SIEMENS I ~1 ll EMF-91 -169 s

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Grand Gulf Unit 1 Cycle 6 Reload Analysis

=.

October 1991 I

N Siemens Nuclear Power Corporation 31

.i 9112130140 011205 ADOCK 0500g6

{,DR

SIEMENS EMF 91 169 Issue Date:

3g7 33 g GRAND GULF UNIT 1 CYCLE 6 RELOAD' ANALYSIS Prepared by Md M. J. Hibbard BWR Fuel Engineering Fuel Engineering and Ucerising 06L.Ab dp.mA'cAnWH F.

C. C. Roberts'

)

BWR Fuol Engineering Fuel Engineering and Ucensing r

,k bO~

N. L Garner BWR Fuel Engineering Fuel Engineering and Ucensing l

~~.

October 1991 f

1 Siemens Nuclear Power Corporation E g e g a,; va u 3:v eg 5 :m 3

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CUSTOWER OtSCLAIMER IWPORTANT NOT1CE REGARDING CONTENTS AND USE OF THIS ROCUMENT PGASE READ CAREFULLY 5emens Nudeer Power Corporatows wartunnes and representanons concoming the sutect maner of thse document are mose set form m the Agreement between Gemens Nunser Power Corporeson and the Cusemer pursua. t a wne W:

document e issued Accorengly, except as omernse expressly previoed m suca Agat, nemer Semens Nudeer Power Corporacon nor any person acnng on its behest makes any warranty or represenmoon, orpressed or imped. *e respect to me accuracy, comptenenese. or usefumess of the eformanon contamed m Ws document, or that tne use of any infonnacon, apparatus, memoa or process osaoned a mes cexument wel not etnnge pnvatey owned ngnts: or assumes any liabdices wie respect to the use of any informauon. apoaratus. method or process oisdosed in this document.

The clormason contmoed heren is for me som use of me Customer.

In oroer e and unparment of nghs of Semens Nucteer Power Corporanon en potenta or evensons who may be educed m the mtormoson contained m os document, the roepent by no ecceoeance of ns occurnent, agoes not to pubash or make putsc use (m me peeent use of me term) of sucn n'ormason unti so autnonzed m wneng by semena Nudeer Power Corporabon or unal after sas (6) mones fotowmg termmason or emprecon of the afocowas Agreement and any esension swoof. unW egressJy proveed b. me Agreement. No ngnts er heentes e or to any peam are vnpued by tne fumsenmg of as cucument

EMF 91 169 Page; TABLE OF CONTENTS Section P,agg

1.0 INTRODUCTION

1 2.0 FUEL MECHANICAL DESIGN ANALYSIS 4

3.0 THERMAL HYDRAUUC DESIGN ANALYSIS -

3.2 Hycraulic Characterization 5

3.2.3 Fuel Centerline Temperature.

5 3.2.5 Bypass Flow....

5 3.3 MCPR Fuel Cladding Integrity Safety Umit.

5 3.3.1 Nominal Coolant Condition in Safety Umit Monte Carly Analysi 5

3.3.2 Design Basis Radial Power Distribution.

S 3.3.3 Design Basis Local Power Distnbution 5

5 4.0 NUCLEAR DESIGN ANALYSIS.

4.1 Fuel Bundle Nuclear Design Analysis,

9 4.2 Core Nuclear Design Analysis..

8 4.2.1 Core Configuration..

8 4.2.2 Core Reactivity Characteristics 8

4.2.4 Core Hydrodynamic Stability.

9 9

5.0 ANTICIPATED OPERATIONAL OCCURRENCES..

5.1 Analysis of Plant Transients 13 5.2 Analyses For Reduced Flow Operation 13 5.3 Analyses For Reduced Power Operation 13 5.4 ASME Overpressurization Analysis....

13

. 5.5 Control Rod Withdrawal Error.

13 5.6 Fuel Loading Error.

14 5.7 Determination of Thermal Umits 14 14 6.0

. F0STULATED ACCIDENTS 6.1 Loss-Of Coolant Accident '.

20 6.1.1 Break Location Spectrum 20 6.1.2 Break Size Spectrum 20 6.1.3 MAPLHGR Analysis For SNP 8x8 and 9x9-5 Fuel 20 6.2 Control Rod Drap Accident 20 21 7.0 TECHNICAL SPECIFICATIONS......

7.1 Umiting Safety System Settings 23 7.1.1 MCPR Fuel Claoding Integrity Safety Umit 23 7.1.2 Steam Dome Pressure Safety Umit.

23 23

EMF-91 169 Pageii TABLE OF CONTENTS (Continued)

Section Pace l

7.2 Umiting Conditions For Operation.......

23 7.2.1 Average Planar Unear Heat Generation Rate fcr SNP Fuel 23 7.2.2 Minimum Cntical Power Ratio.....

24 7.2.3 Unear Heat Generation Rate For SNP Fuel 24 7.3 Surveillance Requirements 25 7.3.1 Scram insertion Time Surveillance 25 7.3.2 Stability Surveillance 25 8.0 METHODOLOGY REFERENCES.

26

9.0 REFERENCES

27

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EMF.91'-169 Page iii-

- UST OF TABLES 1

Tab!e Pace 4.1 NEUTRONIC DESIGN VALUE3 10 i

UST OF FIGURES.

Ficure Page 11 POWER / FLOW MAP USED FOR GRAND GULF UNIT 1 MEOD ANALYSIS -

3 3.1

-GRAND GULF UNIT 1 CYCLE 6 SAFETY UMIT DESIGN RADIAL HISTOGRAM 6

3.2 --

GRAND. GULF UNIT 1 CYCLE 6 SAFETY LIMIT DESIGN BASIS LOCAL POWER DISTRIBUTION........

7-4.1 '

' GRAND GULF UNIT 1 CYCLE 6 BUNDLE DESIGNS.....

11' 4.2 GRAND GULF UNIT 1, CYCLE 6 REFERENCE CORE LOADING PATTERN (OUARTER CORE, REFLECTIVE SYMMETRY)....

12 5.1 FLOW DEPENDFJ'T MCPR UMITS FOR GRAND GULF UNIT 1 CYCLE 6,

15=

5.2 POWER DEPENDENT MCPR UMITS FOR GRAND GULF UNIT 1 CYCLE 6.....

16 5.~3 FLOW DEPENDENT LHGRFAC VALUE FOR GRAND GULF UNIT 1 CYCLE 6.

17

~5'.4 POWER DEPENDENT LHGRFAC VALUE FOR GRAND' GULF UNIT 1 CYCLE 6 18 5.5 EXPOSURE DEPENDENT MCPR LIMITS FOR GRAND GULF UNIT 1 CYCLE 6 19 6.1-MAPLHGR VS AVERAGE PLANAR EXPOSURE FOR SNP BX8 AND 9X9-5 RELOAD FU EL.....,......................

22 4

EMF-91 169 P3ge1

1.0 INTRODUCTION

This report provides the results of the analyses performed by Siemens Nuclear Power Corporation (SNP)in support of the Cycle 6 reload for Grand Gulf Unit 1. This report is intended to be used in conjunction with SNP to' tal report XN NF 80-19(A). Volume 4. Revision 1, "Ar.; Mation of the ENC Methodology to '

Aekads," which describes the analyses performed in support of this reload, identifies ' a used for those analyses, and provides a generic reference list. Section num.m 1 rt as te same as corresponding section numbers in XN-NF-80-19(A). Volume 4

+ 'r*vlogy used in this report which supersedes XN-NF-80-19(A). Volume 4, (i>

Pct u nopropriate.

v The NSSS vendor performed extensive safety anyses for Grand Gulf Unit 1 in conjunction with the extension of the power / flow operating map to the Maximum Extended Operating Domain (MEOD) in Cycle 1 (Reference 1). These analyses established appropriate operating limits for MEOD operation. The initial reload of SNP fuelin Grand Gulf Unit 1 occurred in Cycle 2. In support of the initial reload of SNP fuel, extensive additional safety analyses were performed by SNP to either justify the NSSS vendor operating limits or, where necessary, to provide appropriate limits for SNP fuel using SNP methodologies (Reference 2). Subsequent SNP analyses supported an additional reload of SNP fuel in Cycle 3 (Reference 9), Cycle 4 (Reference 12), and Cycle 5 (Reference 15).

Changes from Cycle 5 to Cycle 6 for Grand Gulf Unit 1 include an additional cal cad of SNP fuel resulting in a core comprised of twice burned SNP 8x8 designs and jour SNP 9x9-5 LTAs, once burned SNP 9x9-5 fuel, and fresh SNP 9x9 5 fuel. The 9x9-5 teload fuel is mechanically, neutronically, and thermal hydraulically compatible with the co-resioont Ext and 9x9-5 fuel inserted in previous cycles. The cycle length remains 18 months and tho nominal cycle energy is 1748 GWd. A reload batch design composed of 272 assemblies with axial enriched zoning and up t04.38 w/o U235 assembly average enrichment containing axially varying Gd 0 is used to meet the cycle energy requirements. A portion of each assembly 23 contains from eight to ten Gd 0 rods. The balance of the core is composed of 240 twice burned 23 SNP 8x8 reload fuel assemblies,4 twice burned 9x9-S lead fuel assemblies, and 284 once burned SNP 9x9-5 reload fuel assemblies.

EMF 91-169 Page 2 The design and safety analyses reported in this document were based on design and operational assumptions in effect for Grand Gulf Unit 1 during Cycle 5 operation and conditions bounding Cycle 6 operation. The MCPR and MCPR, limits have been revised to reflect SNP p

calculated limits. Provision has been made in the flow dependent MCPRs for " loop manual" operation (Reference 11). Analyses were performed at EOC-30 EFPD, at EOC, and at EOC+30 EFPD providing limits for Cycle 6 that are cycle exposure dependent. The analyses also included support of the power / flow operation map for MEOD as shown in Figure 1.1. MCPR values were determined using the ANFB Cntical Power Correlation (Reference 8.9). Monitoring to the plant thermal limits presented in this report will be performed using SNP's core monitonng methodology, POWERPLEX* CMSS, in accordance with SNP's thermal limits methodology.

THERMEX (Reference 8.6).

SNP evaluated the LOCA-seismic response and operation with feedwater heaters out of service for Cycle 2 and subsequent cycles. These evaluations remain applicable for Cycle 6.

The Cycle 6 SLO analyses are performed using SNP methodology (References 5 and 8.1 tnrough 8.18). The Cycle 6 results supersede the previous cycle's results.

12 0 i 100 -

(75,100)

(105,100) u*

E j.

80 ELL Region w

f ICF sf,,

o 6

c go

/

Region O

e sO o

e 60 i

/

3 40 -

e

)

u o

U (105,42) 20 -

(34.3,2S)

(73.5,25) o O

to 20 30 40 50 60 70 80 90 100 110 12 0 t

Core Flow, Percent of Roted m

5 8i' FIGURE 1.1 POWER / FLOW MAP USED FOR GPAND GULF UNIT

• a m

I 1

l 4

EMF 91-169 Page 4 2.0 FUEL MECHANICAL DESIGN ANALYSIS Applicable Fuel Design Report:

References 3.10.

and 13 Qualification analyses provided in the references are applicable to ths Grand G SNP fuel assemblies. Minor mochanical design changes are discussed in Reference The expected power history for the fuel to be irradiated during Cycle 6 is boun design LHGR of Figure 4.1 of Reference 16 and Figure 3.1 of Reference 13.

Seismic /LOCA analysis results for Cycle 5 reponed in Appendix A of Reference 1 valid for Cycle 6.

I

' - - ~ - ' - - _ _ _ _ _ _ _. _ _ _ _. _ _ _ _ - - - - - - - - - - - - - - - - ' - ~ ' ^ - - - - - ' ~ ~ ~~'

_ _. _. _.. ~... __.

EMF-91-169 Page5' 3.0

' THERMAL HYDRAULIC DESIGN ANALYSIS 3.2 Hydraulic Characterization-3.2.3 : Fuel Centerline Temoerature Fuel Centerline Me; ting is protected by the transient LHGR limit given in References 13 and 16.

3.2.5 Svoass Flow Calculated Bypass Flow 10.6%

~

(Exclusive of Water Rod Flow at 104.2%P/108%F) 3.3

' MCPR Fuel Claddino inteority Safetv Umit See Reference 4 1.06*

1.07 "

'3.3.1 ' Nominal Coolant Condition in Safety Limit Monte Carlo Analysis Core Power

. 5074 MWt Core inlet Enthalpy.-

' 520.5 Btu /lbm Reference Pressure.

1050 psia Feedwater Temperature 420"F

.Feedwater Flow Rate 21.8 Mlbm/hr 3.3.2 Desion Basis Radial Power Distribution See Figure 3.1 -

u 3.3.3-Desion Basis Local Pcwer Distribution

See Figure 3.2 e

The 1.06 includes effects for channel bow.

For single loop operation the safety limit MCPR increases to 1.07 due to increased uncertainties associated with SLO.

l^

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. _...,.. ~..,, _ _. _.... _..... _... _., _

12 0 4

10 0 I

s 4

e 80 1

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4 I

3 1

m E

60 uo

_O E

I i

3 40 Z

g i

~

20 1

-t

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O j

O.2 0.3 0.4 0.5 0.6 0.7 O.8' O.9

1. O' 1.I 1.2

1. 3 1.4

.1.5 m

l Radial Power Peakings 5-l JS i

T'-

s m

FIGURE 3.1 GRAND GULF UNIT 1 CYCLE 6 SAFETY LIMIT DESIGN HADIAL illSTOGRAM S*'

k

.e

a EMF 91-169 Page 7 C0NTR0L R00 0

N 0.986 1.025 1.018 1.030 1.063 1.030 1.018 1.025 0.986 T

- R 1.025 0.967 1.047 0.989 0.814 0.989 1.047 0.966 1.025 0

L 1.018 1.047 1.028 0.970 0.994 0.968 1.027 1.047 1.019 R

1.030 0.989 0.970 0.897 0.000 1.050 0.970 0.990 1.031 0

D 1.063 0.814 0.994 0.000 0.000 0.000 0.999 0.814 1.064 1.030 0.989 0.968 1.050 0.000 0.889 0.982 0.993 1.032 1.018 1.047 1.027 0.970 0.999 0.982 1.035 1.051 1.020 1.025 0.966 1.047 0.990 0.814 0.993 1.051 0.967 1.027 0.986 1.025 1.019 1.031 1.064 1.032 1,020 1.027 0.987 FIGURE 3.2 GRAND GULF UNIT 1 CYCLE 6 SAFETY LIMIT DESIGN BASIS LOCAL POWER DISTRIBUTION

,4

,n a

.._-_m-,__-

e i

EMF-91-169 Page8 4.0 NUCLEAR DESIGN ANALYSIS 4,1 Fuel Bundle Nuclear Des 6cn Analysis I

Assembly Average Enrichment, w/o U235

- 3.38 ANF 1.5 H -

- 2.94 ANF 15 L Radial Enrichment Distribution See Reference 10 Axial Enrichment Distribution Figure 4.1 Burnable Poisons Figure 4.1 Location of Non-Fueled Rods See Reference 10 Neutronic Design Parameters -

Table 4.1 4.2 Core Nuclear Desian Analysis 4.2.1 Core Confiauration Figure 4.2 Core Exposure at EOC5 -

24805 mwd /MTU

- Core Exposure at BOC6 13385 mwd /MTU

- Core Exposure at EOC6 25831 mwd /MTU

Maximum Cycle 6 Ucensing Exposure Umit 26649 mwd /MTU ahy -

3.:.

t '

i

~ -,.

-...... -.

EMF-91-169 Page 9 4 2.2 Core Reactivity Charactenstics(U.(2)

BOC6 Cold K-effective All Rods Out 1.11869 BOC6 Cold K-effective, All Rods in 0.95220 BOC6 Colo K-effective, Strongest Rod Out 0.98914 Reactivity Defect /R Value(3)

.07% Delta.K/K Standby Uquid Control System Reactivity,660 PPM Cold Conditions, K-effective 0.96850 (Ulncludes calculational bias.

(2) Evaluated at nominal EOC5-818 mwd /MTU.

(3)The R-Value will be revised based on actual EOC5 conditions.

4.2.4 Core Hydrodynamic 3tability Core hydrodynamic stability is addressed by the licensee.

P I

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i e,

4

r EMF 91-169 Page 10

+

TABLE 4.1 NEUTRONIC DESIGN VALUES Fuel Assembiv (9x9-51

- Number of fuel rods 76-

' Number of inert water rods-5 Fuel rod enrichments

'See Reference 10 Fuel rod pitch, inches 0.563 Fuel assembly loading kgU ANF 1,5 H -

175.70

- ANF 1.5 L 175.57 Core Data-Number of fuel assemblies 800 Rated thermal power, MWt 3833 Rated core flow, Mlbm/hr 112.5 Core inlet subcooling, Btu /lbm 22.2 Moderator temperature, 'F 551 Channel thickness, inch 0.120 Fuel assembly pitch, inch 6.0 Sym. water gap thickness, inch 0.545 Control Rod Data

- Absorber materlaf 84C Total blade span, inch 9.804 Tota: blade support span, inch 1,55

. Blade thickness, inch -

0,328

~ Blade face to-face internal dimension, inch 0.238 Absorber rods per blade (wing) 72 (18) l.

Absorber rod outside diameter, inch 0.22 l"

Absorber rod inside diameter, inch 0.166 1-

' Absorber denshy, percent of theoretical 70 o

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L l

i s,

, _.... + -...... ~. -.,,

..--..-,,.,,,.4..--,.

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EMF 91 169 Page 11 l

8/ete Desian 8 ear Iundle Desion fo**

Ahf95*3388 9CZ*120N 150 Assos.29 B 002 120= t$0 l

@ tarleket and Cadot tat e ( AN8 1.5*)

to. EaH eh e t aM caoot iMa caws.1,$t 3 12 in ANF95 071L-woG 12 in AnF95 071L. woo 12 in AmF95 369L 9G3.0 12 in AnF95 319t.9c4.5 12 in ANF95 369L 9G4.5 ANF95 319L 9G7.0 36 in 24 in AmF95 369L 9G7.0 AmF95-378L 8G5.5

'2 in Auf 95-32BL 8G5.5 42 in 42 in Auf 95 378L 8G7.0 42 in ANF95 328L 9G7.0 6 in AmF95 071L nog 6 in AnF95 071L-moG l

l l

FIGURE 4.1 GRAND GULF UNIT 1 CYCLE 6 BUNDLE DESIGNS

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l r

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l EMF 91-169 Page 12 1

M C1 80 C1 80 31 F0 Ot F0 Ct 50 03 g

03 y

M C1 so 21 M

D1 FC 01 F0 01 90 M

EO M

M M

M l

80 D'

M 01 F0 D1 80 D1 to C1 EO 01 EO Di M

I C1 4

Di 80 C1 F0 01 80 Ci 80 M

E0 M

Di M

80 Of to 01 80 D1 so Ci r0 C1 to 01 E0 M

M 0+

to 31 80 31 80 C1 80 C1 80 M

E0 M

On 4

82 Dt 70 Of FC C1 80 01 to Ct EO Lt EO M

M 31 EQ 31 30 C1 30 C1 EQ D1 to M

E0 M

M 80 Os to C1 80 C1 FC 01 M

M EC 31 M

C1 90 C1 E0 C1-80 C1 E0 M

M_

M C1 M

C0 M

EO M

EQ M

EQ M

EO M

01 C1 M

Di EO -

01 E0 01 EO Q1 EO Di C1 C1 a2 M

M E0 M

EO M

ED M

M M

M Os M

31 D1 M

Of M

M M

M-M M

M M

M x1 n e * * ? roe

v. Cve en tweed M

M A

240 SNP 8x8 3.37 w/o U-235 (ANF 1.3)

B, '

4 SNP 9x9 3.25 w/o U 235 (ANF 1.3)

C 180 SNP 9x9 3.42 w/o U-235 (ANF-1.4)

D 104 SNP 9x9 3.42 w/o U-235 (ANF 1.4)

E 100 SNP 9x9 3.38 w/o U-235 (ANF-1.5)

F 172 SNP 9x9 2.94 w/o U-235 (ANF 1.5) f

- n FIGURE 4.2 GRAND GULF UNIT 1 CYCLE 6 REFERENCE CORE LOADING PATTERN (QUARTER CORE, REFLECTIVE SYMMETRY)

L

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t EMF 91 169 l

' Page 13 i

5.0

! ANTICIPATED OPERATIONAL OCCURRENCES -

Applicable Generic Transient Methodology Report References 5. 8.8 5.1 '

Analysis of Plant Transients Reference 4 (Applicable at rated conditions)

Transient Detta-CPR*

EOC 30 EFPO EQQ EOC+30 EfPD LRNB 0.14 0.16 0.18 LFWH

• 0.09 0.09 ^

0.09-CRWE***

0.10 0.10 0.10 FWCFNB-

.0.13 0.15 0.16

'Umiting values.

-Applicable at all conditions.

Statistically determined, Reference 6.

Exposure Dependent Umit MCPR, Figure 5.5

- Analyses For Reduced Flow Ooeration Reference 4 5.2

. MCPR, Figure 5.1

LHGRFAC, Figure 5.3 5.3 -

Analyses For Reduced Power Ooeration Reference 4

MCPR, Figure 5.2 LHGRFACp Figure 5.4

_. 5. 4 ASME Overoressurization Analysis Reference 4 Limiting Event 7

MSIV Closure -

Worst Single Failure MSIV Position Scram Trip 4

4

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.,m.

p

~

,-,-.6*.-n e,-,,

+

,,y,-r-,,-

y-viv..

.p c. --

yy 9

9

l EMF 91 169 Page 14 5.5 Control Rod Witndrawal Error Reference 6 Values of delta CPR as a function of core power level resulting from a CRWE transient were developed in Reference 6 on a generic basis for BWR/6 class of plants (including Maximum Extended Operating Domain operation). Analysis has been performed demonstrating continued applicability of the generic CRWE analysis results.

5.6 Fuel Leadino Error Reference 8.1 With Loadino Error Correctiv Leaded Core Maximum LHGR, kW/ft 12.97 11.80 Minimum MCPR*

1.21 1.31

• Determined using ANFB Critical Power Correlation.

5.7 Determination of Thermal Limits The results of tha analyses presented in Sections 5.1, 5.2, and 5.3 are used for the determination of the operating limit. Section 5.1 provides the results of analyses at rated conditions ' including the operating limit as a function of exposure in the cycle (MCPR,,

Figure 5.5). Sections 5.2 and 5.3 provide for the determination of operating limit at off-rated conditions of reduced flow and reduced power operation (MCPR,, Figure 5.1 and MCPR.p Figure 5.2). The highest value of MCPR from among the ones presented in these figures for the operating condition of the reactor is to be selected as the operating limit of interest.

4

1.6 i

1.5 l'

l.4 O

v O'

a

1. 3 U

2 1.2 1.1

^

^

^

1.0 O

10 20 30 40 50 60 70 80 90 10 0 110 12 0 Core Flow, Percent of Roted 2

u?

=1 FIGURE 5.1 FLOW DEPENDENT MCPR LIMITS FOR GRAND GULF UNIT 1 CYCLE 6

2.50 2.25 Co re Flo w > 50 7.

2.00 g

s y

Core Flow < 50 %

a 1.75 U

2 s.50 -

i.25 -

i.oo O

10 20 30 40 50 60 70 80 90 100 110 12 0 Core Power, Percent of Rated mr m?

E2 e :.

FIGURE 5.2 POWER DEPENDENT MCPR UMITS FOR GRAND GULF UNIT 1 CYCLE 6

^

s 1.1 i

i i

i i

1.0 7

J.

)>

0.9 q+

CD g8 t2-0.8 cre I

l 0.7 0.6 0.5 O

10 20 30 40 50 60 70 80 90 10 0 110 12 0 Total Core Flow, Percent of Roted q

m) 4 FIGURE 5.3 FLOW DEPENDENT L11GRFAC VALUE FOR GRAND GULF UNIT 1 CYCLE 6 GE i

r_ ;

EMF-91 169 Page 18 1.6 i

i 9s9 LNGRFAC(0) 1.4 1..

30 t.o o

.=.a o.e O,6 1

1 o

to 20 30 40 50 60 70 80 90 too tio 12 0 Core Power, Percent of Roted i,6 8:8 LHCRFAC(0) t,4 1.2 S

U I

t.0 ao=

_a o.s o.6 o..

o to 20 30 40 50 60 70 80 to too 110 12 0 Core Powv, Percent of Roted FIGURE 5.4 POWER DEPENDENT LHGRFAC VALUE FOR GRAND' GULF UNIT 1 CYCLE 6

1.5 s

1.4 i

. 4.

?

.3 g

1.25 vcr

[1.

I U

i 2

1.20 t

1.2 i

f t

t.

l i

f I

I i

BOC EOC-30 EOC EOC+30 i

m i

1 C

Core Average Exposure

l? A e-t eL 1

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FIGURE 5.5 EXPOSURE DEPENDENT MCPH LIMITS FOR GRAND GULF UNIT 1 CYCLE 6 i

e EMF 91169 Page 20 60 POSTULATED ACCIDENTS 6.1 Loss Cf Coolant Accident 6.1.1 Break location Scectrum Reference 7 6.1.2 Break Size Soectrum Reference 7 6.1.3 MAPLHGR Anaivsis For SNP ex8 and 9x9 5 Fuel References 8 and 12 Limiting Greas:

Double Ended Guillotine Pipe Break 1)

Recirculation Pump Discharge Une with 1.00 Discharge Coefficient (1.0 DEG/RD)

The spray heat transfer coefficients identified in 10CFR50 Appendix K are used for the 9x9 5 fuel in an identical manner as in the previous appiaved analysis for Grand Gulf 1 2

(Reference 15). This includes the use of 5 BTU /hr-ft,.F for all of the unheated surf aces including the five water rods.

MAPLHGR rsaults for the two reload fuel types are reported below:

Peak Local Maximum Metal Water PCT ('F)

Reaction (%)

8x8 Fuels 1691 0.3 9x9 Fuels 1713 0.5 The core wide metal water reaction is less than 0.1%.

. -. ~. -

i EMF 91 169 Page 21 The MAPLHGR limits for 8x8 and 9x9 5 are shown in Figure 61. These are bounding limits. The 949 5 limits are bounding for the LTA. The 8x8 limits are provided in Reference 8 For single loop operation, a reduction factor of 0 86 is applied to the two-loop MAPLHGR limits shown in Figure 6.1, Application of this reduction factor ensures that the PCT for a single loop operation LOCA is bounded by the two-loop LOCA analysis.

6.2 Control Aod Droo Accident Reference 8.1 Dropped Control Rod Worth, mk 11.4 4

i Doppler Coefficient. AK/K/'F 10.4 x 10 Effective Delayed Neutron Fraction 5.40 x 10'3 Four Bundle Local Peaking Factor 1.225 Maximum Deposited Fuel Rod Enthalpy, cal /g 166 l

i i-

4 4

{

16

[

a 15 i

(0,14.3) 8x8 Fuel 14 3

(2 0.14.3)

(8-13 i

i

,q (0,12.5) 9x9-5 Fuel BC 12 (2 0,12.5) i E

o 4

I ii l

3 Q.

2 30 -

l i

I 9-(55,9.0) 1 l

j a-

[

1 i

t

~

(50,7.9) 7 8

E 0

i a

10 20 1

t

]

30 40 50 Average Planor Burnup GWd/MTU 60 m

I r

m?

FICURE 6.1 MAPLHGR VS AVERAGE PLANAR EXPOSURE FOR SNP a

I 8 AND 9X9 5 RELOAD FUEL ME l

a t

EMF 91 169 Page 23 70 TECHNICAL SPECIFICATIONS 7.1 Umitina Safety System Settinos 7.1.1 MCPR Fuel Claddina Intearity Safety Limit Safety Umit MCPR 1.06' 1.07 "

7.1.2 Steam Dome Pressure Safety Umit Pressure Safety Umit 1325 psig 7.2 kimitina Conditions For Operation 7.2.1 Averace Planar Unear Heat Generation Aate for SNP Fuel The following MAPLHGR limits are consistent with 10CFR50.46 requirements. The MAPLHGR limit is not used to protect the design basis LHGR limits for the fuel types co resident in Cycle 6.

Average Planar MAPLHGR MAPLHGR Exoosure 8x8 9x9-5 0.0 GWd/MTU 14.3 kW/ft 12.5 kW/ft 20.0 14.3 "5

50,0 7.9 9.5 55.0 9.0 For single loop operation, a reduction factor of 0.86 is applied to the above two-loop MAPLHGR limits, g

The 1.06 safety limrt accounts for channel bow.

A safety limit of 1.07 is to be applied during single loop operation.

I

EMF 91 169 Page 24 j

7.2.2 Minimum Critical Power Aatio MCPR(f)

Figure 5.1 MCPR(p)

Figure 5.2

)

MCPR(e)

Figure 5 5 7.2.3 Linear Heat Generation Pate For SNP Fue,j The LHGRlimits for SNP 8x8 fuel for Grand Gulf 1 have been extended to support Cycle 6 operation. These limits, which are based on Figure 4.1 of Reference 16. are as follows:

Averaae Planar Exoosuro LHGR 0.00 GWd/MTU 16.0 kW/ft 25.40 14.1 40.00 10.0 55.00 8.0 The LHGR limits for 9x0 5 fuel, based on Figure 3.1 of Reference 13, for SNP reload fuel during Cycle 6 operation are as follows:

Averaae Planar Exposure LHGR 0.00 GWd/MTU 13.1 kW/ft 15.50 13.1 55.00 8.0 LHGRFAC, and LHGRFAC multipliers are applied directly to the Technical Specification p

LHGR limits for each fuel type at reduced power and/or flow conditions to ensure protection of the limits.

LHGRFAC Multipliers for Off-Nominal Conditions:

LHGRFAC(f)

Figure 5.3 LHGRFAC(p)

Figure 5,4

EMF 91 169 Page 25 7.3 Eurveillance Aecuirements 7.3.1 Scram insertion Time Surveillance Thermal margins are based on analyses in which scram performance was assumed consistent with the Technical Specification limits.

No additional surveillance for scram performance is required above that already being done for conformance to Technical Specifications.

7.3.2 Stability Surveillance Core stability surveillances have been addressed by the Licensae in TS 4.4.1.1.1.

mm

EMF 91169 Page 26 0.0 METHODOLOGY REFERENCES Section 8 References 8.1 through 8.18 are contained in the following report:

"Enon Nuclear Methodology for Boiling Water Reactors:

Application of the ENC Methodology to BWR Reloads." XN-NF 80-19(A), Volume 4. Re.ision 1. Euen Nuclear Company, Richland, Washington (March 1985).

Reference 8.6 is superseded by:

8.6

• Euon Nuclear Methodology for Boihng Water Reactors THERMEX: Thermal Limits Methodology Summary Description," XN-NF 8019(P)(A), Ve!ume 3.

Revision 2 (January 1987).

References 8.9 and 8.18 are superseded by:

8.9 "ANFB Critical Power Correlation," ANF 1125(P)( A), ard Supplements 1 and 2 (April 1990)

Reference 8.10 is superseded by:

8.10

" Advanced Nuclear Fuels Corporation Critical Power Methodology for Boiling Water Reactors," ANF 524(P)(A), Revision 2, and Supplements 1 and 2 (November 1990).

i

-=.

EMF 91 169 Page 27 90 REFERENCES 1.

Changes to Allow Operation with One Recirc cal Specification Domain," August 15,1986.

x ended Operating 2.

Company, Richland, WA, August 1986." Grand Gulf Un xxon Nuclear 3.

" Generic Mechanical Design for Exxon Nuclear Jet XN-NF 85-67(P)(A), Revision 1, Exxon Nuclear Company, Ric p emoer 1986.

4.

Corporation, Richland, WA, October 1991," Grand Gulf U ens Nuclear Power 5.

"COTRANSA2:

ANF-913(P)(A), Volume 1, Revision 1 and Supplem ysis,"

,,n ugust 1990.

6.

"BWR/S Generic Rod Withdrawal Error Analysis, MCPRp " X Company, Richlano, WA, May 1986, and XN NF-825(P',(A) Sup xxon Nuclear

. ctober 1986.

7.

Nuclear Company, Richland, WA, April 1986." Generic

, Exxon B.

" Grand Gulf Unit 1 LOCA Analysis," XN-NF-86-38, Exxon Nucl June 1986.

y, Richland. WA, 9.

Fuels Corporation, Aschland, WA, August 1967." Gran vanced Nuclear 10.

" Grand Gulf Unit 1 Reload ANF 1.5 Design Report, Mechanical Neutronic Design for Advanced Nuclear Fuels 9x9-5 Fuel Asse y raulic, and Advanced Nuclear Fuels Corporation, Richland, WA, July 1991

- 1080(P),

11.

" Grand Gulf Nuclear Station Unit NESDO;_88;0_Q2, MSU System Services Inc., November 1988. Revis 1

12.

Corporation, Richland, WA, November 1988," Grand Gu uclear Fuels 13.

ANF 88152(P)(A) with Amendment 1 and Supplem e oad Fuel,"

Corporation, Richland, WA, November 1990.

, Advanced Nuclear Fuels 14.

March 5,1991 (RAC:026:91), Letter, R. A. Copeland (ANF) to n

ange."

I i

r

._n,

{

EMF 91 169 l

Page 20 i

15.

" Grand Gulf Unit 1 Cycle 5 Reload Analysis," ANF 90-022. Revision 2, Advanced Nuclear Fuels Corporation, Richland, WA, August 1990.

16.

' Grand Gulf Unit 1 XN 1.3, Cycle 4 Mechanical Design Report." ANF 88183(P),

Supplement 1, Siemens Nuclear Power Corporation, Richland, WA, August 1991.

l I

f 4-A.

\\

EMF 91 169 issue Date:

10/31/?1 GRAND GULF UNIT 1 CYCLE 6 RELOAD ANALYSIS Distnbution O. C. Brown R. A. Copeland L J. Federico D. L Garber N. L. Garner D. E. Hershberger.

M. J. Hibbard R. B. Macduff J. N. Morgan R. S. Reynolds C. C. Roberts C. J. Volmer G. N. Ward A. W. Will Entergy Operations /S. L Leonard (40)

Document Control (5)

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