ML20115J619

Rev 1 to Grand Gulf Unit 1 Cycle 6 Reload Analysis

ML20115J619
Person / Time
Site:	Grand Gulf
Issue date:	07/31/1992
From:	Garner N, Hibbard M, Roberts C SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER
To:
Shared Package
ML20115J613	List: ML20115J610 ML20115J614 ML20115J619
References
EMF-91-169, EMF-91-169-R01, EMF-91-169-R1, NUDOCS 9210280271
Download: ML20115J619 (36)

Text

+2,. y 1 e SIEMENS  !

EMF 91-169 i Revision 1 ,

Grand Gulf Unit 1 Cycle 6 Reload Analysis July 1992 3

Nuclear Division i

8ho2*B8uM68lh6 P

b SIEMENS EMF 91-169 Revision 1 issue Date: 8/5/92 GRAND GULF UNIT 1 CYCLE 6 RELOAD ANALYSIS Prepared by i32ast e h

' / N L. Garner BWR Fuel Engineering Fuel Engineering and Licensing S We W.

C, C Roberts /

BWR Fuel Engineering Fuel Engineering and Licensing

%dlo MY M. J. Hibbard BWR Fuel Engineering Fuel Engineering and Licensing July 1992 i

i l

1 1

CUSTOMER DISCLAIMEQ IMPORTANT NOTICE REGARDING C0fMNTS AND USE OF THis DOCUMENT PLEASE READ CAREFULLY Siemens Power Corporation's warrention and representatkms corwmang the subhet maner of tNo document are those est forth in the Agreement between Siemene Power Corponmon and the Customer pursuant to which INe document la leeued. - Accordingly, escoopt as otherwise expressly provided in such Agreement, nether Siemens Power Corporamon nor any person acting on tie benaff makes av warranty or representation, empressed or impded, witri roepect ;

to the accuracy,i~,i-f __w , or usefulness of the information contairsed in tNo document, or that the use of any iniormation, apperatus, method or p,rocess

- m th.e document wel not infringe privately owned rights; c assumes any .r samaa witn roepect to the use of any information, apparatus, method or procesa disclosed in tNe document.

The information contained herein is foi the solo use of the Cestomer.

In order to avoid imperment of rights of Siemene Power Corporation in patente -

or inventions wnich may be biduded in the b1 formation contamed in the document, the redpient, by les acceptance of INe document, agrees not to publien -

or make putAc use (in the patent use of the term) of such information untu so-authertzed in wneng by Siemens Power Corporation or unta after six (6) monthe '

followmg terminetton or expiration of the aforesaid Agreement wid any extension thereof, unises expresely provuled in the Agreement. No rights or liconees in or to any patente are impiled by the fumseNng of thee document.

5 e o m .mnw w ,- ~ , 4

D EMF.91 169 Revision 1 Pagei TABLE OF CONTENTS Section Eggg .

1.0 I NT R O D U CTI O N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 2.0 FUEL MECHANICAL DESIGN ANALYSIS . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 3.0 THERMAL HYDRAULIC DESIGN ANALYSIS . . . . . . . . . . . . . . . . . . . ...... 5 3.2 Hydraulic Cht.racterization . . . . . . . . . . . , . . . . . . . . . . . . . . . . . . . . . 5- '

3.2.3 Fuel Centerline Temperature . . . . . . . . . . . . . . . . , . . . . . . . . .. 5

3. 2. 5 B y pa ss Flo w . . . . . . . . . . . . . . . . . . . . . . . . . . - . . . . . . . . . . . .

5 .

3.3 MCPR Fuel Cladding integrity Safety Limit . . . . . . . . . . . -. . . . . . . . . . . 5-3.3.1 Nominel Coolant Condition ire Safety Limit Monte Carlo Analysis . . ..................................... 5 3.3.2 Design Basis Radial Power Distribution . . . . , . . . . . . . . . . . . . . 5 3.3.3 Design Basis Local Power Distribution . . . . . . . . . . . . . . . . . . . . 5 4.0 NUCLEAR DESIGN ANALYSIS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8

^

4.1 Fuel Bundle Nuclear Design Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . 8-4.2 - Core Nuclear Design Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

8 4.2.1 Core Configuration . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 4.2.2 Core Reactivity Characteristics . . . . . . . . . . . . . . . . . . . . . . . . . 9 4.2.4 Core Hydrodynamic Stability . . . . . . . . . . . . . . , . . . . . . . . . . 9 5.0 ANTICIPATED OPERATIONAL OCCURRENCES . . . . . . . . . . . . . . . . . . . . . . 13  ?

5,1 Analysis of Plant Transients . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .- 13 i 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 . . . . . . . . . . . . . . . . . . . . . . . .. ..... 14' 5.6 Fuel Loading Error .....................................

14 5.7 Determination of Thermal Limite . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 2

6.0 FO STU LATED ACCIDENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20-6.1 Loss-O f-Coolant Accident . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . - 20

-6.1.1 Br eak Location Epectrum . . . . . . . . . . . . . . . . . . . . . . . . . . . . - 20 6.1.2 Break Size Spectrum . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20-6.1.3 MAPLHGR Analysis For SNP 8x8 and Gx9-5 Fuel . . . . . . . . . . . ;20 y 6.2- Control Rod Drop ~ Accident ' . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21 7.0 TECHNICA', SPECIFICATIONS . . . . . . . . . . . . . . . . . . . . . . . . . . , . . . . . . 23 7.1 Lim! ting Safety System Settings . . . . . . . . . . . . . . . . . . . . . . . . . , , , . 23 7.1.1 MCPR Fuel Cladding Integrity Safety Limit . . . . . . . . . . . . . . . . -23 7.1.2' Steam Dome Pressure Safety Limit . . . . . . . . . . . . . . . . . . . . . 23 _l g --py- n, e rE--

EMF.91 169 Revision 1 i Page il j l

TABLE OF CONTENTS (Continued)

Section EADA 7.2 Limiting Conditions For Operation . . . . . . . . . . . . . . . . . . . . . . . . . . . 23 7.2.1 Averat J Planar Linear Hett Generation Rate for SNP Fuel . . . . . 23 7.2.2 Minimum Critical Power Ratio . . . . . . . . . . . . . . . . . . . . . . . . 24 7.2.3 Linear Heat Generation Rate For SNP Fuel . . . . . . . . . . . . . . . . 24 7.3 Surveillance Requirements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 25 7.3.1 Scram insertion Time Surveillance . . . . . . . . . . . . . . . . . . . . . 25 7.3.2 St ability Surveillance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 25 G.0 METHO DO LO GY R EFERENCE S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 26 9.0 REFERE*4Cf 'i ............................................. 27

+

s EMF 91 169 Revision 1 Page ill-LIST OF TABLES Iabla Eaan 4.1 NEUTRONIC DE SIG N VALU E S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 P

LIST OF FIGUEES j Flourf Etan l t

1.1 POWER / FLOW MAP USED FOR GRAND GULF UNIT 1 Mt:OD ANALYSIS . . . . . 3 3.1 GRAND GULF UNIT 1 CYCLE 6 SAFETY LIMIT DESIGN RADIAL  :

H I ST O G R A M . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 3.2 GRAND GULF UNIT 1 CYCLE 6 SAFETY LIMIT DESIGN BASIS LOCAL POWER DISTRIBUTIO N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 4.1 ' GRAND GULF UNIT 1 CYCLE 6 BUNDLE DESIGNS . . . . . . . . . . . . . . . . . . . . 11 4.2 GRAND GULF UNIT 1, CYCLE 6 REFERENCE CORE LOADING PATTERN (QUARTER CORE, REFLECTIVE SYMMETRY) ....................... 12 5.1 FLOW DEPENDENT MCPR LIMITS FOR GRAND GULF UNIT 1 CYCLE 6 . . . . 15 5.2 POWER DEPENDENT MCPR LIMITS FOR GRAND GULF UNIT 1 CYCLE 6 . . . . 16 5.3 FLOW DEPENDENT LHGRFAC VALUE FOR GRAND SULF UNIT.1 CYCLE 6 .. 17 5.4 ' POWER DEPENDENT LHGRFAC VALUE FOR GRAND GULF UNIT 1 CYCLE 6 . 18 S.5 EXPOSURE DEPENDENT MCPR LIMITS FOR GRAND. GULF UNIT 1 CYCLE 6 . 19-6.1 MAPLHGR VS AVERAGE PLANAR EXPOSURE FOR SNP 8X8 AND 9X9 5 RELOADFUEL............................................. 22

.,; . , - , ,, . - . . . = . - .;- .,. -- , - .~---a . . .

4 EMF 91 169 Revision 1 Pageiv l An SNP investigation into the scnsitivity of FWCF event uverity to the water levelin i the steam separator has led to the conclusion that the procedure used in past SNP FWCF l analyses is non-conservative relative to the benchmark cases for the SNP COTRANSA2 -

l methodology. SNP has established a new procedure which corrects this non conservatism l l

1 and provides conformance with the approval basis for the methodology.  ;

I l- Ident:fication of the non-conservatism required SNP to evaluate the impact upon l

' t l analyses performed for the Cycle 6 licensing campaign for Grand Gulf Unit 1 as provided in l EMF 91-168 and EMF 91 169. The FWCF case with the least margin to MCPR operating f

. I limits for Grand Gulf Cycle 6 operation (104.2%P/108%F at EOC 30) has been teenalyzed -

l using the new procedure. P.asults from this reanalysis were used to assure that the MCPR l operating limit remairis valid at the most limiting condition and to establish a bounding 1 l Increase in event delta CPR to be applied to results for all other cases.

l l Revision 1 of this report is issued to effect the changes in results associated with the I revised procedure for FWCF analysis. There were no changes made to the text of the report, ,

l Changes in tabulated results from Revision 0 are indicated by revision bars in the left margin  ;

l of the report.

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EMF 91 169 Revision 1 Page 1

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 rents .:;

intended to be ust.d in conjunction with SNP topical report XN NF 8019fA), Vowe 4, Revision 1, " Application of the ENC Methodology to BWR Reloads," which describes the analyses performed in support of this reload, identifies the methodology used for those analyses, and provides a genotic reference list. S6ction numbers in this report are the same as corresponding section numbers in XN NF 8019(A), Volume 4, Revision 1. Methodology used in this report which supersedes XN NF-8019(A), Volume 4, Revision 1,is referenced as appropriate.

The NSSS vendor performed extensive sifety analyses for Grand Gulf Unit 1 In conjunction with the extension of the power / flow operating map to the Maximum Extended .

Operating Domain 04EODi m Cycle 1 (Reference 1). These analyses established approp. late operating limits for (AEOD o$wration. The initial reload of SNP fuel in Grand Gulf Unit 1 occurred in Cycle 2. In support of the initial reload of SNP fuel, extensive additinnel 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 methodoloales (Reference 2).

Subsequent SNP analyses supported an additional reload of SNP fuelin C yett ? (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 reload of SNP fuel resulting in a core comprised of twice bumed SNP 8x8 designs and four SNP 9x9 5 LTAs, once burned SNP 9x9 5 fuel, and fresh SNP 9x9 5 fuel. The 9x9 5 reload fuel is mechanically, neutronically, and thermal hydraullcally compatible with the co-resident 8x8 and 9x9 5 fuel inserted in previous cycles. The cycle length remains 18 months and the nominal cycle energy is 1748 GWd. A reload batch design composed of 272 assemblies with axial enriched zoning and up to 3.38 w/o U235 assembly average enrichment containing axially varying Gd203 si used to meet the cycle entcq requirements. A portior' of each assembly contains from eight to ten Gd23 0 rods. The caiance of the core is composed of 240 1

e l

EMF 91 169 Revision 1 Page 2 twice burned SNP 8x8 reload fuelassemblies,4 twice burned 9x9 5 lead fuel assemblies, and '

284 once burned SNP 9x9 5 reload fuel assemblies, i 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 ac3 ration. The MCPRpand MCPRg limits have been revised to ref act SNP calculated limits. Provision has been made in the flow dependent MCPRa 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 Critical Power Correlation (Reference 8.9). Monitoring to the plant thermal limits presented in this report will be performed using SNP's core monitoring methodology, POWERPLEX* CMSS, in accordance with SNP's thermallimits 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 through 8.18). The Cycle 6 results supersede the previous cycle's results.

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O 10 20 30 40 50 60 70 80 90 100 110 12 0 5 m s.

Core Flow, Percent of Roted =

, s. 7'

c. E. ?

%3" u-.

FIGURE 1.1 POWER / FLOW MAP USED FOR GRAND GULF UNIT 1 MEOD ANALYSIS

EMF 91 169 Revision 1 Page 4 2.0 FUEL MECHANICAL DESIGN ANALYSIS Applicable Fuel Design Report: References 3,10, and 13 4

Qualification analyses provided in the references are applicable to the Grand Gulf Unit 1 SNP fuel assemblies. Minoi mechanical design changes are discussed in Reference 14.

The expected power history for the fuel to be irradiated during Cycle 6 is bounded by the design LHGR of Figure 4.1 of Reference 16 and Figure 3.1 of Reference 13.

Seismic /LOCA analysis results for Cycle 5 reported in Appendix A of Reference 15 remain valid for Cycle 6.

4

, -- , -y .

EMF 91 169 Revision 1 Page 5 3.0 THERMAL HYDRAULIC DESIGN ANALYSIS 3.2 Hydraulic Characterization 3.2.3 Fuel Centerline Temnerature Fuel Centerline Melting is protected by the transient LHGR limit given in References 13 and 16, 3.2.5 Bvoans Flow Calculated Bypass Flow 10.6%

(Exclusive of Water Rod Flow at 104.2%P/108%F) 3.3 MCPR Fuel Claddino Inteority Safety Limit See Reference 4 1.06' 1.07 ' '

3.3.1 Nominal Coolant Condition in Safety Limit Monte Carlo Analysig 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 Deslan Basis Radial Power Distribution See Figure 3.1 3.3.3 Deslan Basis local Power Distribution See Figure 3 2 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.

_ ~ .

. I EMF.91 169 Revision 1 Page 6 1

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EMF 91 169 Revision 1 Page 7 C0NTR0L R0D 0

N 0.986 1.025 1.018 1.030 1.063 1.030 1.018 1.02" 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

EMF 91 169 Revision 1 Page 8 4.0 NUCLEAR DESIGN ANALYSIS 4.1 Fuel Bundle Nuclear Deslan Analysia Assembly Average Enrichment, w/o U235 3.38 ANF 1.5 H 2.94 ANF 1.5 L Radial Enrichment Distribution See Reference 10 Axlal 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 Deslan Analysis 4.2.1 Core Confiauration Figure 4.2 Core Exposure et EOC5 24805 mwd /MTU Core Exposure at BOC6 13385 mwd /MTU Core Exposure at EOC6 25831 mwd /MTU Maximum Cycle 6 Licensing Exposure Limit 26649 mwd /MTU

+=

EMF 91169 Revision 1 Page 9 4.2.2 Core Reactivity Characteris1[g3III'I2I BOC6 Cold K effective, All Rods Out 1.11869 BOC6 Culd K effective, All Rods in 0.95220 BOC6 Cold K effective, Strongest Rod Out 0.98914 Reactivity Defect /R Value(31 .07% Delta K/K Standby Liquid Control System Reactivity,660 PPM ,

Cold Conditions, K effective 0.96850 Illincludes calculational bias.

(2iEvaluated at nominni EOC5 818 mwd /MTU.

(3)The R Value will be revised based on actua! EOC5 conditions.

4.2.4 Core Hydrodynamic Slabili1Y r

Core hydrodynamic stability is addressed by the licensee.

1 P

l EMF 91 169 Revision 1 Page 1C TABLE 4.1 NEUTRONIC DESIGN VALUES l

Fuel Assembly (9r9 5)

Number of fuel rods 76 Number of inert water rods 5 Fuel rod enrichments See Reference 10 l Fuel rod pitch, inches 0.563 Fuel assembly loading, kgU ANF 1.5 H 175.70 ANF 1,6 L 175.57 Core Data Number of fuel assemblies 800 Rated thermal power, MWt 3833 Rated core flow, Mlbm/h' 112.5 Core inlet subcooling, Btu /lbm 22.2 Moderator temperature, "F 551 Channel thickness, inch 0.120 Fuel assembly pitch, inch 6.0 Syrn water gap thickness, inch 0.545 Control Rod Data Absorber material B4C Total blade span, inch 9.804 Total 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)

Absorber rod outside diameter, inch 0.22 Absorber rod inside diameter, inch 0.166 Absorber density, percent of theoretical 70

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I s

EMF 91 169 .

Revlolon 1

, Pe 911 kaufle Dae f an f or t k adla Dealan fort AltFM*33AA 9G212m 150 AmFM 296a 9a212te 150 i

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utah Enrf-  : and "if ala (AmF 1.5N1 Law inefehamnt and " " llnfa tant 1.lL1 12 in AmtM UTIL ans 12 in amp 95 071L mos 12 in Anf95 369L 963.0 12 in - Mt95 319L 904.S 12 in Amf95 369L 904.5 '

Anf95 319L 967.0 24 in Ampf5 369L 9G7.0 i

AmF95 375L 4GS.S Anf95 328L tal.5 t

. 42 in 42 in 42 in AmF95 379L 4GT.0 42 in ANF95 328L 9sT.0 1

6 In Ahf95 071L*m00 6 in - ANF95 071L mos 4 3 i

' FIGURE 4.1 ' GRAND GULF UNIT 1 CYCLE 6 BUNDLE DESIGNS t

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k EMF 91 169  !

Revision 1  :

Page 12 A2 Cl Po Cl 80 Of PO 09 P0 C1 90 Of A2 Of A2 A2 C1 70 Of AJ Ot 80 Of PO Of Pc A2 to A2 A2 A2 A2 P0 Of. A2 01 80 Of PO On 70 C1 80 Of to 01 A2 Ct A2 Of P0 C1 70 Of Po C1 P0 A2 le A2 Of A2 PO Ot P0 Ct PO Of P0 Ct P0 Ci to 09 90 (J A2 09 PO Of P0 Oi 80 C1 P0 Ct P0 A2 to A2 Of P2 PO 01 PO 01 70 Ct PO Of 70 Cl 90 0% 10 A2 A2 01 90 Of Po Ct PO 01 to 01 M A2 to A2 A2 PC Of PO Ct P0 C1 PO Of A2 A2 to Of A2 Ct P0 C1 P0 C1 P0 Cl 80 A2 #J A2 C1 A2 PO A2 80 Ai to A2 to A2 to A3 Of C1 A2 Of to - Ot to 01 M 01 to 01 C1 Ci 02 A2 A2 to At to A2 to A2 A2 A2 A2 01 A2 01 Ot A2 01 A2 A2 A2 A2 A2 A2 A2 A2 A2 XV 8* v==

A2 A2 l

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) l FIGURE 4.2 GRAND GULF UNIT 1 CYhE 6 REFERENCE CORE LOADING PATTERN (QUARTER CORE, REFLECTIVE SYMMETRY)

4 EMF 91 169 Revision 1 Page 13 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)

TransitD1 Delta CPR*

EOC 30 EFPQ EQC EOC + 30 EFPD LRNB 0.14 0.16 0.18 LFWH" 0.09 0.09 0.09 CRWE*" 0.10 0.10 0.10 l FWCFNB 0.14 ,

0.17 0.18 Limiting values.

Applicable at all conditions.

Statistically determined, Reference 6.

Exposure Dependent Limit MCPR, Figure 5.5 5.2 Analysen For Reduced Flow Ooeration Referer',-

MCPR g Figure 5.1 LHGRFACg Figure 5.3 5.3 Analvsen For. Reduced Power Onerall20 Reference 4 -

MCPR p Figure 5.2 LHGRFAC p - Figure 5.4 5.4 - ASME Overoressurization Analysis- Reference 4 Limiting Event MSIV Closure Worst Single Failure MSIV Position -

Scram Trip

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• EMF 91 169 l Revision 1 Page 14 5.5 Control Rod Withdrawal Error Reference 6 Values of delta CPR as a function of core power level resulting from a CRWE transient I were developed in Reference 6 on a generic basis for BWR/F class of plants (including Maximum Extended Operating Domain operation). Analysis has been performed demonstrating continued applicability of the generic CRWE analysis results. i 5.6 Fuel Loadina Error Reference 8.1 With Loadina Error Correctiv Lgaded Core Maximum LHGR, kW/ft 12.97 11.80 Minimum MCPR* 1.21 1.31

• Determined uring ANFB Critical Power Correlation.

5.7 Determination of Thermal Limits The results of the ant./ses 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 (MCPRg, Figure 5.1 end MCPRp ,

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.

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0 10 20 30 40 50 60 70 80 90 100 110 12 0 m

Core Flow, Percent of Rated 2.mI aib ePy e o _.

_a 3 m O, *W FIGURE 5.1 FLOW DEPEf0ENT MCPR LIMITS FOR GRAND GULF UNIT 1 CYCLE 6 e

i

c 2.50 , , , .

2.25 - -

Core Flow > nc %

2.00 - -

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1.00 ^ ' - ' ^ ' ^ ' - ' ^ ' ^ ' ^ ' ^ ' ^ ' ' ^

0 10 20 30 40 50 60 70 80 90 100 110 12 0 Core Power, Percent of Roted ,E i

.m sib o *_ y e o _.

-e 3 O e -a 40 FIGURE 5.2 POWER DEPENDENT MCPR LIMITS FOR GRAND GULF UNIT 1 CYCLE 6

, y .. _c , . . -

3 6 1.1 . .

a

1. 0 -- _

t i

0.9

- g+9> _

O v

u gp b O.8 - .

cr o

I J

0.7 - .

4 0.6 - _ .

0.5

- ' - ' ' - ' ' - ' - ' - ' - r - ' - ' -

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

mIb o E. ;

eO

.*U CD Q -.a 4D FIGURE 5.3 FLOW DEPENDENT LHGRFAC VALUE FOR GRAND GULF UNIT 1 CYCLE 6

• _ - - - - -- _ _ . _m__m_ _ _ _ _ - _ __ _ _ . - _..._..__m...-_.._m

EMF 91 169 Revision 1 Page 18 t.6 , , , , , , , , , ,

9st LNGRFAC(g) n.4 -

1.2 -

^

a O

t.0 -

o 1 5 '

0.e -

0.6 -

,,, i , , , , . . , ,

0 10 20 30 40 SO 60 70 SO 90 10 0 180 12 0 ,

Core Power, Percent of R0ted '

t.6 , , , , , , , , , , ,

Bw8 LHGRFAC(p) 1.4 -

1.2 -

3u t.D -

o 5

0.e -

0.o -

0.4 ' ' ' ' ' ' '-

0 10 20 30 40 60 60 70 s0 of 10 0 810 12 0 Core Power, Percent of Rot 0d FIGURE 5.4 POWER DEPENDENT LHGRFAC VALUE FOR GRAND GULF UNIT 1 CYCLE 6

i.5 1

1.4 - -

1.3 - -

9 v

1.25 m -

Q.

U y 1.20 1.2 - -

1.1 - -

g i I I I BOC EOC-30 EOC EOC+30 ,,,

2r=

Core Average Exposure j gi eo*

U$E FIGURE 5.5 ' EXPOSURE DEPENDENT MCPR LIMITS FOR GRAND GULF UNIT 1 CYCLE 6

i EMF 91 169 Revision 1 Page 20 6.0 POSTULATED ACCIDENTS 6.1 Loss Of Coolant Accident 6.1.1 Break Location Soectrum Reference 7 6.1.2 Break Sire Soectrum Reference 7 6.1.3 MAPLHGR Analysis For SNP 8x8 and 9x9 5 Fuel References 8 and 12 Limiting Break: Double Ended GuillWne Pipe Break in Recirculation Pump Discharge Line with 1.00 Discharge Cr' efficient (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 prev!ous approved analysis for Grand Gulf 1

, (Reference 15). This includes the use of 5 BTU /hr ft2 .oF for all of the unheated surfaces including the five water rods.

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

Peak Local Maximum Metal Water PCT (*F) Reaction (%)

ex8 Fuels 1691 0.3 9x9 Fuels 1713 0.5 The core wide metal water reaction is less than 0.1 %.

EMF 91 169 Revision 1 Page 21 The MnPLHGR limits for 8x8 and 9x9 5 are shown in Figure 6.1. These are bounding limits. The 9x9 5 limits are bounding for the LTA. The 8x8 limits are provided in Reference S. For single-loop operation, a reduction factor of 0.86 is applied to the two loop MAPLHGF 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 Rod Dron Accident Reference 8.1 Dropped Control Rod Worth, mk 11.4 Doppler Coefficient, AK/K/'F -

10.4 x 10 6 Effective Delayed Neutron Fraction 5.40 x 10 3 Four Bundle Local Peaking Factor 1.225 Maximum Deposited Fuel Rod Enthalpy, cal /g 166 4

.-,y - -

• y y r--- - w- g-. -.. , r.a,y, w-,

16 . . . . .

15 -

(0,14.3) 8x8 Fuel

~ _

(2 0,14.3) 13 -

d (0,12.5) 9xC-5 Fuel x 12 -

(2 0,12.5) -

er C

,- T. 11 -

g _

G.

2 to - _'

(55,9.0) i 0 :.~ _

8 - _

(50,7.9)

,_ . . i ' '

O 10 20 30 40 50 60 m Average Pianor Burnup GWd/MTU

?g5 Lh eeoE. :-.

U$E FIGURE S.1 MAPLHGR VS AVERI.GE PLANAR EXPOSURE FOR SNP 8X8 AND 9X9-5 RELOAD FUEL

EMF 91 169 Revision 1 Page 23 7.0 TECHNICAL EPECIFICATIONS 7.1 Limitino Safety System Settinos 7.1.1 MCPR Fuel Claddino Inteority Safety Limit Safety Limit MCPR 1.06' 1.07 '

• 7.1.2 Steam Dome Pressure Safety Limit Pressure Safety Limit 1325 psig 7.2 Limitino Conditions For Ooeration 7.2.1 Aversoe Planar Linear Heat Generation Rate 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 cm resident in Cycle 6.

Average Planar MAPLHGR MAPLHGR

, Exoosure 8x8 9x9-5 0.0 GWd/MTU 14.3 kW/it 12.5 kW/ft 20.0 14.3 12.5 50.0 7.9 9.5 55.0 - 9.0 For single-loop operation, a reduction fac'.or of 0.86 is applied to the above two-loop-MAPLHGR limits.

l The 1.06 safety limit accounts for channel bow.

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

l l

I _ . - - _ _ - . . _ ..

EMF 91 169 Revision 1 Page 24 7.2.2 Minimum Critical Power Ratio Mf MCPR(f) Figure A.1 MCPR(p) Figure 5.2 MCPR(e) Figure 5.5 7.2.3 Linear Heat Generation Rate For SNP Fuel The LHGR limits 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 Exoosure LHGR 0.00 GWd/MTU 16.0 kW/ft 25.40 14.1 40.00 10.0 55.00 8.0 9

The LHGR limits for 9x9 5 fuel, based on Figure 3.1 of Reference 13, for SNP reload fuel during Cycle 6 operation are as follows:

Averace Planar Exoosure LHGR 0.00 GWd/MTU 13.1 ' kW/ft 15.50 13.1 55.00 8.0 LHGRFAC, and LHGRFACp multipliers are applied directly to the Technical Specification 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:

l LHGRFAC(f) Figure 5.3 LHGRFAC(p) Figure 5.4 i

-J

1 EMF 91 109 Revision 1

• F' age 25 -

7.3 Surveillance Recuirements 7.3.1 Scram insertion Time Surveillance i Thermal margins are based on analyses in which scram performance was assumed consistent with the Technical Specification limits. No' additional surveillance for 6 'am performance is required above that already being donc for conferrnance to Technical Specifications-7.3.2 Stabilitvjurveillance Core stability surveillancas have been addressed by the Licensee in TS 4.4.1.1.1.

(

l

l l

EMF 91-169 Revision 1 Page 26 8.0 METHODOLOGY REFERENCES Section 8 References 81 through 8.18 are concained in the following report-

" Exxon Nuclear Methodology for Boil!. j Water Reactors: Application of the ENC Methodology to BWR Reloads," XN NF 8019(A), Volume 4, Revision 1, Exxon Nuclear Company, Richland, Washington (March 1985).

Reference 8.6 is superseded by:

8.9 " Exxon Nuclear Methodology for Boiling Water Reactors THERMEX: Thermal Limits Methodology Summary Description," XN-NF-80-19(P)(A), Volume 3, Revision 2 (January 1987).

References 8.9 and 8.18'are superseded by:

8.9 "ANFB Critical Power Correlation," ANF-1125(P)(A), and 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).

. - - ~ - -. .

EMF 91-? 69-Revision 1 Page 27

9.0 REFERENCES

1. Letter,- Lester L. Kintner (USNRC) to O. D. Kingsley, Jr. (MW.- " Technical Specification Changes to Allow Operation with Ora lecirculation I m od Extended Operating Domain," August 15,1986.

2. " Grand Gulf Unit 1 Cycle 2 Reload Analysis," XN NF 86 35, Revision 3, Exxon Nuclear Company, Richland, WA, August 1986.

3. " Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel,"

XN NF 85 67(P)(A), Revision 1, Exxon Nuclear Company, Richland, WA, September 1986.

4. " Grand Gulf Unit 1 Cycle 6 Plant Transient Analysis," EMF-91-168, Siemens Nuclear Power Corporation, Richiand, WA, October 1991.

5. - COTRANSA2: A Computer Program for Boiling Water Reactor Transient Analysis,"

ANF-913(P)(A), Volume 1, Revision 1 and Supplements 2,3, and 4, August 1990.

6. "BWR/6 Generic Rod Withdrawal Error Analysis, MCPRp," XN NF-825(A), Exxon

- Nuclear Company, Richland, WA, May 1986, and XN NF-825(P)(A), Supplement 2, October 1986.

7. ' Generic LOCA Break Spectrum Analysis for BWR/6 Plants," XN-NF 86 37(P), Exxon Nuclear Company, Richland, WA, April 1986.

8. " Grand Gulf Unit 1 LOCA Analysis," XN-NF-86-38, Exxon Nuclear Company, Richland, WA, June 1986.

9. " Grand Gulf Unit 1 Cycle 3 Reload Analysis," ANF-87-67. Revision 1, Advanced l Nuclear Fuels Corporation, Richland, WA,- August 1987.

l

10. " Grand Gulf Unit 1 Reload ANF-1.5 Design Report, Mechanical, Thermal Hydraulic, and Neutronic Design for Advanced Nuclear Fuels 9x9 5 Fuel Assemblies," ANF-91-OBO(P), _

Advanced Nuclear Fuels Corporation, Richland, WA, July 1991. -

11. " Grand Gulf Nuclear Station Unit 1 Revised Flow Dependent Thermal Limits,"

NESDO-88-OO3, MSU System Services Inc., November 1988.

12. " Grand Gulf Unit 1 Cycle 4 Reload Analysis," ANF-88-149, Advanced Nuclear Fuels Corporation, Richland, WA, November 1988.

13. " Generic Mechanical Design.for Advanced Nuclear Fuels 9x9 5 BWR Reload Fuel,"

ANF-88-152(P)(A) with Amendment 1 and Supplement 1, Advanced Nuclear Fuels Corporation, Richland, WA, November 1990.

l '

t L-

I EMF 91-169 Revision 1 Page 28

14. Letter, R. A. Copeland (ANF) to R. C. Jones (NRC), " Minor M nical Design Change," March 5,1991 (RAC:026:91).

9

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(Pf, Supplement 1, Siernens Nuclear Power Corporation, Richland, WA, August 1991.

s e

C;,

..i-EMF 91 169 :

= Revision 1 -

Issue Date:J-8/5/92

' GRAND GULF UNIT 1 CYCLE 6 RELOAD ANALYSIS Distribution O. C. Brown R. A. Copeland L. J. Federico D. L. Garber-N. L. Garner -

D. E. Hershberger M. J. Hibbard J. N. Morgan C. C. Roberts I

C. J. Volmer -

G. N. Ward-Entergy Operations /S. L. Leonard (40) ,

Document Control (5) l

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a