ML20112H484

Susquehana Unit 1 Cycle 2 Reload Analysis,Design & Safety Analyses for ENC XN-1 8x8 Reload Fuel

ML20112H484
Person / Time
Site:	Susquehanna
Issue date:	12/14/1984
From:	Collingham R, Jensen S, Krysinski T SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER
To:
Shared Package
ML17139C813	List: ML20112H484 ML20112H497 ML20112H504 ML20112H515
References
XN-NF-84-116, NUDOCS 8501170162
Download: ML20112H484 (26)

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• XN-NF-84-116 Issue Date: 12/14/84 SUSQUEHANNA UNIT 1 CYCLE 2 RELOAD ANALYSIS Design and Safety Analyses For ENC XN-1 8x8 Reload Fuel Prepared by:

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S. E. Jensen, Lead En BWR Safety Analysisgineer Concur: (( 4, .4, R. E. Collingh a , Manager n ,.-

= BWR Safety Anafysis 9/9/

Concur:

T. L. Krysigski, Manager BWR Neutron cs Concur: $ h D'N / /Y bY J. N./ Morgan, Manager CustomerSe[vicesEngineering Approve: -l Y < a tb t i1 H. B. Stout, Manager Licensing & Safety Engineering Approve: [I[ Manager IH. E. WiTTiamson, hem /2/W'd(/

Neutronics & Fuel Management Approve:

G. A. Sofer/ Manager N / c23f' Fuel Enginliering & Technical Services ERON \ UCLEA R CO V 3A\ Y, \ C.

i XN-NF 116 NUCLEAR REGULATORY COMMISSION O!SCLAIMER IMPORTANT NOTICE REGARDING CONTENTS AND USE OF THIS DOCUMENT plEASE READ CAREFULLY This technical report was derived through research and development programs sponsored by Exxon Nuclear Company, Inc. It is being submitted by Exxon Nuclear to the U.S. Nuclear Regulatory Commission as part of a technical contribution to facilitate safety analyses by licensees of the U.S. Nuclear Regulatory Commission wnich utilize Exxon Nuclear-fabricated reload fuel or other technical services provided by Exxon Nuclear for light water power reactors and it is true and correct to the best of Exxon Nuclear's knowledge, information, and belief. The i n formation contained herein may be used by the U.S. Nuclear Regulatory Commission in its review of this report , and by licensees or applicants before the U.S. Nuclear Regulatory Commission which are customers of Exxon Nuclear in their demonstration of compliance with the U.S. Nuclear Regulatory Commission's regulatinns. Without derogating from the foregoing, neither Exxon Nuclear nor any person acting on its behalf:

A. Makes any warranty, express or implied, with respect to the accuracy, completeness, or usefulness of the information contained in this document, or that the use of any information, apparatus, method, or process disclosed in this document will not infringe privately owned rights, or

8. Assumes any liabilities with respect to the use of, or for damages resulting from the use of, any information, apparatus, method, or process disclosed in this document.

- " " ' ~ _ _- . _ _ _ _ _ _ _ _ _ _ _ . __ _ _ _ . _ _ . _ _ _ _ . _ _

XN-NF 116 TABLE OF CONTENTS Section Page  !

Number Number 1

1.0 INTRODUCTION

................................................I 2.0 FUEL MECHANICAL DESIGN ANALYSIS............................ 2 3.0 THERMAL HYORAUL IC DES I GN ANALYS I S. . . . . . . . . . . . . . . . . . . . . . . . . . 3 3.2 HYDRAULIC CHARACTERIZAT10N............................ 3 3.2.5 Bypass Flow.................................... 3 3.3 MCPR FUEL CLADDING INTEGRITY SAFETY LIMIT............. 3 i

3.3.1 Coolant Thermodynamic Conditions............... 3 J 3.3.2 Design Basis Radial Power Distribution......... 3 3.3.3 Design Basis Local Power Distribution.......... 3 -

4.0 NUCLEAR DESIGN ANALYSIS.................................... 4 4.1 FUEL BUNDLE NUCLEAR DESIGN ANALYSIS................... 4 4.2 CORE NUCLEAR DESIGN ANALYSIS.......................... 4 4.2.1 Core Configuration............................. 4 .

4.2.2 Core Reactivity Characteristics................ 4 4 4.2.3 Core Hydrodynamic Stability.................... 5 )

5.0 ANTICIPATED OPERATIONAL OCCURRENCES........................ 6  :

5.1 ANALYSIS OF PLANT TRANS!ENTS AT RATED CON 0!T10NS...... 6 I 5.2 ANALYSES FOR REDUCED FLOW OPERAT10N................... 6  :

5.3 ASME OVERPRESSURIZATION ANALYSIS...................... 6 5.4 CONTROL R00 WITH0RAWAL ERR 0R.......................... 7 -

5.5 FUEL LOADING ERR 0R.................................... 7 l

5.6 DETERM INATION OF THERMAL MAR GINS.... .. .......... .. .. .. 7

XN-NF-84-116 6.0 POSTU LAT ED ACC I DE NTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 6.1 LO SS -0F -COOL ANT ACC I D E NT. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 6.1.1 Break Location Spectrum........................ 8 6.1.2 Break Size Spectrum............................ 8 6.1.3 MAPLHGR Analyses for ENC XN-1 8x8 Fuel........................................... 8 6.2 CONTROL R0D DROP ACC10ENT............................. 8 7.0 TECHNICAL SPECIFICATIONS................................... 9 7.1 LIMITING SAFETY SYSTEM SETTINGS....................... 9 7.1.1 Fuel Cladding Integrity Safety Limit........... 9 7.1.2 Steam Dome Pressure Safety Limit............... 9 7.2 LIMITING CONDITIONS FOR OPERAT10N..................... 9 7.2.1 Average Planar Linear Heat Generation Rate Limits for ENC XN-1 8x8 Fuel........................................... 9 7.2.2 Minimum Critical Power Ratio................... 9 7.2.3 Surveillance Requirements..................... 10 9.0 ADD I T IO N AL REF ER E NC E S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 APPENDICES A. SE I SM IC -LOC A E VALU AT 10N. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19

XN-NF-84-116 LIST OF TABLES Table Title Page 4.1 Neut ro ni c Des i gn Val ues . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 LIST OF FIGURES Figure Title Page 3.1 Susquehanna Unit 1 Cycle 2 Safety Limit Ra di al Powe r Hi s t og ram. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 3.2 Susquehanna Unit 1 Cycle 2 Safety Limit Loca l Pea ki n g Fact o rs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 4.1 En ri chment Di stri but i on . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . 15 4.2 Reference Core Loading Pattern........................ 16 Sol Starting Control Rod Pattern for Control Rod Wi thdrawa l Erro r Ana lys i s .. . . . . . . . . . . . . . . . . . . . . . . . . . . . 17 5.2 Low Flow MCPR Limits.................................. 18 9

1 XN-NF-84-116

~

SUSQUEHANNA UNIT 1 CYCLE 2 RELOAD ANALYSIS Design and Safety Analyses For ENC XN-18x8 Reload Fuel

1.0 INTRODUCTION

This report provides the results of the analyses performed by Exxon Nuclear Company (ENC) in support of the Cycle 2 reload for Susquehanna Unit 1, which is scheduled to commence operation in spring 1985. This rrport is intended to be used in conjunction with ENC topical report XN-NF-80-19(A), Volume 4, " Application of the ENC Methodology to BWR Reivada, which describes the analyses performed in support of this raload, identifies the methodology used for those analyses, and provides a g:neric reference list. Section numbers in this report are the same as corresponding section numbers in XN-NF-80-19(A), Volume 4 Appendix A presents a seismic-LOCA evaluation of ENC 8x8 fuel.

The Susquehanna Unit 1 Cycle 2 core will comprise a total of 764 fuel assemblies, including 192 unirradiated ENC XN-1 8x8 assemblies, and 572 previously irradiated assemblies of various 8x8 lattice configurations fabricated by General Electric Company. The reference core configuration is described in Section 4.2.

The design and safety analyses reported in this document were based on the design and operational assumptions in effect for Susquehanna Unit 1 during the previous operating cycle. Specifically, analyses were performed on the basis of the existing bases in the plant Technical Specifications, including the Cycle 1 power-flow operating map.

i 2 XN-NF-84-116 2.0 FUEL MECHANICAL DESIGN ANALYSIS l l

Applicable Fuel Design Report Ref. 9.1 The power history depicted in Figure 5.10 of Reference 9.1 bounds the expected power history of the Susquehanna Unit 1 XN-1 fuel type. The reanalysis required by the SER of Reference 9.1 accepting the mechanical design report were accomplished and reported in XN-NF-81-21 (P), Revision 2.

-Fuel Centerline Temperature Exposure at Minimum Margin Point 5000 MWD /MT Centerline Temperature at 120% Overpower 4057 0F Melting Point of Fuel 5000 0F

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~ Margin to Centerline Melting 943 0F 9

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r 3 XN-NF-84-116 3.0 THERMAL HYDRAULIC DESIGN ANALYSIS 3.2 HYDRAULIC CHARACT.ERIZATION The hydraulic compatibility of the XN fuel type with the coresident P8xBR fuel is demonstrated by the generic compatibility analysis in the generic applications document, XN-hF-80-19( A), Volume 4 3.2.5 Bypass Flow Calculated Bypass Flow Fraction 10.4 %

3.3 MCPR FUEL CLADDING INTEGRITY SAFETY LIMIT See Reference 9.3 3.3.1 Coolant Thermodynamic Conditions Core Power 3633 MW Steam Dome Pressure 1053 psia Feedwater Temperature 420 F Feedwater Flow Rate 15.6 Mlbm/hr Feedwater Enthalpy 398 BTU /lbm ,

3.3.2 Design Basis Radial Power Distribution See Figure 3.1 3.3.3 Design Basis Local Power Distribution See Figure 3.2

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4 XN-NF-84-116 4.0 NUCLEAR DESIGN ANALYSIS 1

.4.1 FUEL BUNDLE NUCLEAR DESIGN ANALYSIS I Assembly Average Enrichment 2.72 w/o i Radial Enrichment Distribution Figure 4.1 Axial Enrichment Distribution Uniform 2.81 w/o with 6 inch Natural Uranium at top.

Burnable Poisons Figure 4.1 Note: Burnable poisons are distributed uniformly over the enriched length of the designated rods. The natural uranium axial blanket sections do not contain burnable absorber material.

Non-fueled Rods Figure 4.1 Neutronic Design Parameters Table 4.1 4.2 CORE NUCLEAR DESIGN ANALYSIS 4.2.1 Core Configuration -

Figure 4.2 Core Exposure at E0C1 10,700 MWD /MT Core Exposure at B0C2 8,790 MWD /MT l Core Exposure at E0C2 14,540 MWD /MT I Note: Cycle 2 safety analyses are valid for E0C1 l exposure from -700 MWD /MT to +700 MWD /MT

! from the nominal value reported above.

4.2.2 ~ Core Reactivity Characteristics 80C2 Cold K-effective (ARO) 1.09577 BOC2 Cold K-effective, Strongest Rod Out 0. % 493 Reactivity Defect (R-Value) 1.45 % Rho SBLC Reactivity, 660 ppm, Colo Conditions, K-effective 0.94554 i

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5 XN-NF-84-116

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4.2.3 Core Hydrodynamic Stability Susquehanna Unit I has adopted a detect and suppress approach to ossuring hydrodynamic stability during plant operation. Because of differences in the I. relative pressure drop characteristics of the upper and lower tie plate designs between the ENC XN-1 fuel and the coresident G.E. Type P8x8R fuel, the ENC fuel is slightly more stable than the G.E.

fuel. The detect and suppress operating requirements implemented during Cycle 1 operation will provide assurance of stable core operation for Cycle 2.

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6 XN-NF-84-116 5.0 ANTICIPATED OPERATIONAL OCCURRENCES Generic Transient Analysis Report Ref. 9.2 5,1 ANALYSIS OF PLANT TRANSIENTS AT RATED CONDITIONS See Reference 9.3 Limiting Events: Loss of Feedwater Heating (LFWH)

Load Rejection Without Bypass (LRWB)

Feedwater Controller Failure (FWCF)

Maximun Maximum Maximum K;ximum Transient

• Power Heat Flux Pressure Delta-CPR LRWB 252% 108.9%- 1203 psia .19 .

FWCF 140% 109.3% 1191 psia .17 l LFWH 120% 113.2% 1074 psia .13

• All transients are bounding results.

Note: All transients were evaluated at E0C2 conditions, beginning with 104% of rated power and 100% of rated recirculation flow.

5.2 ANALYSES FOR REDUCED FLOW OPERATION See Reference 9.4 ,

Limiting Transient: Recirculation Flow Increase 5.3 ASME OVERPRESSURIZATION ANALYSIS See Reference 9.3 Event MSIV Closure Single Failure MSIV Position Scram Trip Maximum Pressure j 1318 psig Maximum Sensed Pressure 1303 psig

1 7 XN-NF-84-116 5.4 CONTROL R0D WITHDRAWAL ERROR (CRWE)

Starting Control Rod Pattern Figure 5.1 Rod Block Distance ENC GE Setting Withdrawn Delta-CPR Delta-CPR 105% 3.0 ft .21 .23 106%* 3.5 ft .23 .26 107% 4.0 ft .25 .28 108%* 4.5 ft .26 30

• Rod block settings of 106% or 108% may be selected for Cycle 2 5.5 FUEL LOADING ERROR Delta-CPR .14 5.6 DETERMINATION OF THERMAL MARGINS i

Event

• Delta-CPR MCPR Limit Model LRWB .19 1.25 COTRANSA FWCF .17 1.23 COTRANSA LFWH .13 1.19 PTSBWR3 CRWE .23 1.29 at 106% RBM XTGBWR

.26 1.32 at 108% RBM XTGBWR

• Events are results of bounding analyses.

MCPR Operating Limits at Rated Conditions Fuel Type MCPR Operating Limit 106% RBM ENC XN-18x8 1.29 GE P8x8R 1.32 108% RBM ENC XN-1 8x8 1.32 GE P8x8R 1.36 MCPR Onerating Limits at Off-Rated Conditions Figure 5.2

-. _ _ _ _ _ _ _ _ _ _ _ _ _ _ - . _ _ _ _ _ _. _ J

l 8 XN-NF-84-116 I I

60 3 POSTULATED ACCIDENTS 6.1 LOSS-0F-COOLANT ACCIDENT 6.1.1 Break Location Spectrum Ref. 9.5 6.1.2 Break Size Spectrum Ref. 9.5 6.1.3 MAPLHGR Analyses for ENC XN-1 Fuel Ref. 9.6 Limiting Break: Double-Ended Guillotine Break Recirculation Pump Discharge Line 0.4 Break Coefficient Bundle Average Exposure MAPLHGR Peak Cladding Peak Local MWD /MT kW/ft Temperature Oxidation 0 13.0 2074 F 1.9%

5,000 13.0 2093 F. 2.0%

10,000 13.0 2116 F 2.1%

15,000 13.0 2136 F 2.2%

19,000 13.0 2147 F 2.3%

25,000 11.5 1977 F 1.6%

30,000- 10.4 1846 F 1.0%

35,000 10.4 1852 F 1.2%

6.2. CONTROL R0D DROP ACCIDENT See Ref. 8.1 Dropped Control Rod Worth 8.9 mk Doppler Coefficient -9.5x(10**-6) 1/k dk/dT Effective Delayed Neutron Fraction .0050 Four-Bundle Local Peaking Factor 1.28 Deposited Enthalpy 153 cal /gm

9 XN-NF-84-116 7.0 TECHNICAL SPECIFICATIONS 7.1 LIMITING SAFETY SYSTEM SETTINGS 7.1.1 MCPR Fuel Cladding Integrity Safety Limit MCPR Safety Limit 1.06 7.1.2 Steam Dome Pressure Safety Limit Pressure Safety Limit 1325 psig 7.2 LIMITING CONDITIONS FOR OPERATION 7.2.1 Average Planar Linear Heat Generation Rate Limits for ENC XN-18x8 Fuel Bundle Average MAPLHGR Exposure Limit (MWD /MT) ,(kW/ft) v io.u 5,000 13.0 10,000 13.0 15,000 13.0 19,000 13.0 25,000 11.5 30,000 10.4 35,000 10.4 7.2.2 Minimum Critical Power Ratio Fuel Type MCPR ENC XN-18x8 1.?9 at RBM 106%

ENC XN-18x8 1.32 at .RBM 108%

GE P8x8R 1.32 at RBM 106%

GE P8x8R 1.36 at RBM 108%

i 10 XN-NF-84-116 7.2.3 Surveillance Requirements The ENC reload safety analyses were per' ormed assuming current Technical Specification values for scram speed and scram delay time. No additiGnal surveillance requirements are necessary as a result of the ENC reload safety analysis for Susquehanna Unit 1 Cycle 2.

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11 XN-NF-84-116 9.0 ADDITIONAL REFERENCES 9.1 S.F. Gaines, " Generic Mechanical Design for Exxon Nuclear det Pump BWR Reload Fuel," XN-NF-81-21(A), Revision 1, Exxon Nuclear Co.,

Inc., Richland, WA 99352 (January 1982).

9.2 R.H. Kelley, " Exxon Nuclear Plant Transient Methodology for Boiling Water Reactors," XN-NF-79-71(P), Revision 2, Exxon Nuclear Co., Inc.,

Richland, WA 99352 (November 1981).

9.3 T.H. Keheley, "Susquehanna Unit 1 Cycle 2 Plant Transient Analysis,"

XN-NF-84-118, Exxon Nuclear Co., Inc., Richland, WA 99352 (November 1984).

9.4 T.H. Keheley, "Susquehanna Unit 1 Cycle 2 Plant Transient Analysis, Reduced Flow MCPR Analysis," XN-NF-84-118 Supplement 1, Exxon Nuclear Co., Inc., Richland, WA 99352 (December 1984).,

9.5 J.E. Krajicek, " Generic LOCA Break Spectrum Analysis Jet-Pump BWR 3/4 With Modified Low Pressure Coolant Injection Logic," XN-NF-84-117, Exxon Nuclear Co., Inc., Richland, WA 99352 (December 1984).

9.6 0.J. Braun, "Susquehanna Unit 1 LOCA-ECCS Analysis, MAPLHGR Results,"

XN-NF-84-119, Exxon Nuclear Co., Inc., Richland, WA 99352 (December 1984).

5

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4 12 XN-NF-84-116 Table 4.1 Neutronic Design Values Fuel Pellet Reference 9.1 Fuel Rod Reference 9.1 1 4 ~ Fuel Assembly Reference 9.1 1

Core Data Number of fuel assemblies 764 l Rated thermal power, MW 3293 Rated core flow, Mlbm/hr 100 Core inlet subcooling, BTV/lbm 24.0 Moderator temperature, F 548.8 Channel thickness, inch . .080 Fuel assembly pitch, inch 6.00 Wide water gap thickness, inch .562 Narrow water gap thickness, inch .562 .

C@ntrol Ro'd Data Absorber material B4C Total blade span . inch 9.75 Total blade support span, inch 1.58 Blade thickness, inch 0.260 Blade face-to-face internal dimension, inch 0.200 Absorber rods per blade .

76 Absorber rod outside diameter, inch 0.188

, Absorber rod inside diameter, inch 0.138 Absorber density, % of theoretical 70.0 4

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Figura 3.2 Susquehanna Unit 1 Cycle 2 Safety Limit Local Peaking Factors

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2.48 W/0 U235 HM RODS (16) ---

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2.48 W/0 U235 + 2.00 W/0 GD203 W RODS ( 2) --.

INERT WATER R0D Figure 4.1 Susquehanna-1 Cycle 2 (XN-1) Enriched Zone Enrichment Distribution

16 XN-NF-84-116

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B  : B : B  : B : B : B :B :B  : C : B  : A :

B  : C : B  : C : B :C :B :C :B  : A :

B  : B : C : B  : C : B :B :B  : A : A :

l

B : C : B : C :B :C :B  : A :

A : A : A : A : A : A : A :

Fuel Number of Type Assemblies Description -

A 140 GE 8x8 Type II 1.76 w/o U-235 8 432 GE 8x8 Type III 2.19 w/o 0-235 C 192 XN-1 8x8 2.72 w/o U-235 Figure 4.2 Susquehanna Unit 1 Cycle 2 Reference Loading Pattern By Fuel Type (One Quarter of Symmetrical Core Loading) l

17 XN-NF-84-116 1

I 02 06 10 14 18 22 26 30 34 38 42 46 50 54 58 59 -- -- -- -- -- -- -- 59 55 -- -- 00 -- 36 -- 00 -- -- 55 51 -- -- -- -- -- -- -- -- -- -- -- 51 47 -- -- 00 -- 00 -- 00 -- 00 -- 00 -- -- 47 43 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 43 '

39 -- 00 -- 00 --

10 -- 12 -- 10 -- 00 -- 00 -- 39 35 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 35 31 -- 36 -- 00 -- 12 -- 00* -- 12 -- 00 -- 36 -- 31 27 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 27 23 -- 00 -- 00 -- 10 -- 12 -- 10 -- 00 -- 00 -- 23 19 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- 19 15 -- -- 00 -- 00 -- 00 -- 00 -- 00 -- -- 15 11 -- -- -- -- -- -- -- -- -- -- -- 11 07 -- -- 00 -- 36 -- 00 -- -- 07 03 -- -- -- -- -- -- -- 03 02 06 10 14 18 22 26 30 34 38 42 46 50 54 58

• Control Rod Being Witharawn Rod Position in Notches Withdrawn Full in = 00 Full out = --

Figure 5.1 Susquehanna Unit 1 Cycle 2 Control Rod Withdrawal' Analysis Initial Control Rod Pattern

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APPENDIX A SEISMIC-LOCA EVALUATION The structural response of the ENC XN-1 8x8 fuel is the same as the structural response of the GE 8x8R fuel it replaces in the Susquehanna Unit 1 core. Therefore, the seismic-LOCA structural response evaluation performed in support of the initial core remains applicable and continues to provida assurance that control blade insertion will not be inhibited following the occurrence of the design basis seismic-LOCA event.

The physical and geometric properties of the ENC XN-18x8 and the GE 8x8R fuel types which are important to the dynamic response of the fuel are summarized in Table A1. The close agreement between the important parameters for the two fuel types indicates that the structural response would be very similar for both fuel types.

Similarity in the natural frequencies of the two fuel types is further assured by the stiffness of the fuel assembly channel box. Both fuel types use the same fuel assembly channel box, and the' channel box dominates the overall dynamic response of the incore fuel. ENC calculations show that approximately 97% of the stiffness of a fuel assembly is attributable to the stiffness of the channel box. For this reason, the dynamic structural response of the reload core is essentially that of the initial core, and the original seismic-LOCA analysis remains applicable. Deformation of the channel to the point that control blade insertion is inhibited is not predicted to occur.

Table Al Comparison of Physical and Geometric Characteristics for ENC and GE 8x8 Assemblies Property Fuel Type E NC GE Fuel Rod Pitch, inch .641 .640 Number of Spacers 7 7 Assembly Weight, lbs 596 600 Assembly Length, inch 176.05 176.16 Fuel Rod Diameter, inch 484 483 Cladding Thickness, inch .035 032

20 XN-NF-84-116 Issue Date 12/14/84 SUSQUEHANNA UNIT 1 CYCLE 2 RELOAD ANALYSIS Design and Safety Analyses For ENC XN-18x8 Reload Fuel Distribution D.J. Braun J.C. Chandler R.E. Collingham i' S.F. Gaines R.G. Grummer S.E. Jensen T.H. Keheley J.E. Krajicek T.L. Krysinski J.N. Morgan L.A. Nielson H.G. Shaw/PP&L (60)

G.A. Sofer R.B. Stout H.E. Williamson Document Control (5) 9 la m p nas i mili- iiis i