ML20076G974

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Cycle 9 Reload Analysis
ML20076G974
Person / Time
Site: Dresden Constellation icon.png
Issue date: 08/01/1983
From: Chandler J, Stout R, Williamson H
SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER
To:
Shared Package
ML17194B696 List:
References
XN-NF-83-47, NUDOCS 8309010216
Download: ML20076G974 (32)
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Text

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xx.nr.83.o DRESDEN UNIT 3 CYCLE 9

RELOAD ANALYSIS

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EfgON NUCLEAR COMPANY,Inc.

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1 8309010216 B30825 PDRADOCK05000g P

XN-NF-83-47 Issue Date: 8/1/83 DRESDEN UNIT 3 CYCLE 9 RELOAD ANALYSIS Mechanical, Thermal Hydraulic, and Nuclear Design Analyses i

Prepared by:

J. C. Chandler 7!8 U Reload Fuel Licensing v : ~/ ,/b

.)"; o p f(

l Approve:

R~. B. Stout,' Manager I} (TTus 4 s'3

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Licensing & Safety Engineering Approve: f-_! f[ _ 7 4 // 3

~H. E. Williamson, Manager Neutronics & Fuel Management

?h-4 U; Approve: I /S ', 'l'.

G. A. Soferf Manager Fuel Engineering & Technical Services 9f EXff)N NUCLEAR COMPANY,Inc.

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l NUCLEAR REGULATORY COMMISSION DISCLAIMER IMPORTANT NOTICS 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 sub-mitted by Exxon Nuclear to the USNRC as part of a technical contri-bution to facilitate safety analyses by licensees of the USNRC which utilize Exxon Nuclear fabricated reload fuel or other technical services provkfed by Exxon Nuclear for liaht water power reactors and it is true and correct to the best of Exxon Nuclear's knowledge, information, \

and belief. The information contained herein may be used by the USNRC in its review of this report, and by licensees or applicants before the USNRC which' are customers of Exxon Nuclear in their demonstration of comoliance with the USNRC's regulations.

Without derogating from the foregoing, neither Exxon Nuclear nor any person acting on its behalf: t A. Makes any warranty, express or implied, with respect to the accuracy, completeness, or usefulness of the infor-mation contained in this document, or that the use of any information, apparatus, method, or process disclosed in this document will not infnnge privately owned rights; or B. Assumes any liabilities with respect to the use of, or for darrages resultmg from the use of, any information, ap-paratus, method, or process disclosed in this document.

XN. NF- F00, 766 ,

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I 1 XN-NF-83-47 TABLE OF CONTENTS e Section Page

1.0 INTRODUCTION

....................................... 1 2.0 FUEL MECHANICAL DESIGN ANALYSIS .................... 1 3.0 THERMAL HYDRAULIC DESIGN ANALYSIS .................. 2 4.0 NUCLEAR DESIGN ANALYSIS ............................ 2 5.0 ANTICIPATED OPERATIONAL OCCURRENCES ................ 4 6.0 POSTULATED ACCIDENTS ............................... 5 7.0 TECHNICAL SPECIFICATIONS ........................... 6 9.0 ADDITIONAL REFERENCES .............................. 8 APPENDIX A - Surveillance Requirements ....................... A-1 APPENDIX B - ASEA-ATOM Developmental Control Blades .......... B-1

l ii XN-NF-83-47 LIST OF TABLES l Table No. Page 4.1 Dresden 3 Reload Batch XN-2 Neutronic Design Values ..................................... 9 5.1 Determination of Thermal Margins .................. 11 LIST OF FIGURES Figure No. Page 3.1 Dresden 3 Cycle 9 Safety Limit Radial Power Histogram ................................... 12

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3.2 Dresden 3 Cycle 9 Safety L imit Local Peaking .... .. 13 l 4.1 rarichment Distribution for Fuel Type XN-2 14

) 8x8 (Enricned Lattice 3.02 w/o U-235) . . . . . . . . . . . . .

4.2 Dresden Unit 3 Cycle 9 Reference Loading Pattern (One Quarter of Symmetrical Core Loading) .......................................... 15 4.3 Decay Ratio vs Reactor Power ...................... 16 5.1 Starting Control Rod Pattern for Control Rod Withdrawal Analysis ........................... 17 5.3a MCPR for Automatic Flow Control (AFC) ............. 18 5.3b MCPR for All Conditions ........................... 19 l

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f f 1 XN-NF-83-47

1.0 INTRODUCTION

This report presents the results of analyses performed by Exxon Nuclear Company (ENC) in support of the Cycle 9 (XN-2) reload for Dresden Unit 3, which is scheduled to commence operation near the end of 1983.

The Cycle 9 core will comprise 184 unirradiated Type XN-2 reload fuel assemblies f abricated by ENC, 224 once-irradiated Type XN-1 fuel assemblies, and 316 General Electric 8x8 assemblies. The ENC-fabricated assemblies are as described in XN-NF-81-21 (Reference 9.1). The core configuration is described in Section 4.0 of this report.

s Cjcle 9 operation of Dresden Unit 3 will include the use of eight developmental control blades f abricated by ASEA-ATOM. These blades are described in Appendix B.

This report is intended to be used i,n conjunction with XN-NF-80-19, Volume 4, " Application of the ENC Methodology to BWR Reloads," which describes the analyses which were performed in generation of the results reported in this document.

2.0 FUEL MECHANICAL DESIGN ANALYSIS Applicable Fuel Design Report Reference 9.1 f The power history depicted in Figure 5.10 of Reference 9.1 bounds the expected power history of the Dresden 3 Type XN-2 fuel.

The XN-2 fuel design is as described i

in Reference 9.1 except for the pellet axial height in most of the fuel rods, which is slightly shorter than the generic design for improved resistance to pellet-cladding interaction.

2 XN-NF-83-47 Fuel Centerline Temperature Exposure at Minimum Margin Point 21,200 MWD /MT Centerline Temperature at 120% Overpower 46070F Melting Point of Fuel 49000F Margin to Centerline Melting 2930F 3.0 THERMAL HYDRAULIC DESIGN ANALYSIS 3.2 Hydraulic Characterization Reference 9.7 The Dresden 3 Type XN-2 fuel is identical to the Dresden 3 Type XN-1 fuel in its hydraulic characteristics 3.2.5 Calculated Bypass Flow Fraction 10.8% (E0C) f 3.3 MCPR Fuel Cladding Integrity Safety Limit Reference 9.3 3.3.1 Coolant Thermodynamic Condition '

Core Rated Thermal Power 2527 MWt Core Inlet Flow Rate 98 x 106 lbm/hr Steam Dome Pressure 1020 psia Feedwater Temperature 3200F 3.3.2 Design Basis Radial Power Distribution Figure 3.1 3.3 3 Design Basis Local Power Distribution Figure 3.2 4.0 NUCLEAR DESIGN ANALYSIS 4.1 Fuel Bundle Nuclear Design Analysis for Fuel Type XN-2 8x8 Assembly Average Enrichment 2.83%

Radial Enrichment Distribution Figure 4.1 l

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3 XN-NF-83-47 Axial Enrichment Distribution Uniform 3.02% with 6" Natural Uranium Ends Burnable Poisons Figure 4.1 Non-Fueled Rods Figure 4.1 Neutronic Design Parameters Table 4.1 Maximum Lattice Km 1.224 4.2 Core Nuclear Design Analysis 4.2.1 Core Configuration Figure 4.2 Ccre Exposure at E0C8(1), MWD /MT 21,626/21,130 Core Exposure at 80C9, MWD /MT 14,082 Core Exposure at E0C9, MWD /MT 21,000 4.2.2 Core Reactivity Characteristics BOC9 Cold K-effective, All Rods Out 1.104 BOC9 Cold K-effective,

All Rods In .949 BOC9 Cold K-effective, Strongest Rod Out .985 Technical Specification R-Value .0004(2)

SBLC Reactivity, 700F, 600 ppm .942 4.2.4 Stability Analysis Reactor Core Stability Figure 4.3 Maximum Decay Ratio Value 0.33 Based on results of most recent pressure drop tests (1) Nominal value/Value Used in Shutdown Reactivity Calculations.

(2) Accounts for B4C Settling in Control Rod Tubes (Maximum K-effective with strongest rod withdrawn occurs at 80C9).

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I 4 XN-NF-83-47

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5.0 ANTICIPATED OPERATIONAL OCCURRENCE _SS Applicable Generic Transient Analysis Report Reference 9.2 5.1 Analysis of Plant Transients at Rated Conditions Reference 9.3 Limiting Transients:(1)

Generator Load Rejection Without Bypass (LRWB)

Loss of Feedwater Heating (LFWH)

Feedwater Controller Failure - Maximum Demand (FWCF) 5.2 Analyses for Reduced Flow Operation Reference 9.4 Limiting Transient: Recirculation Flow Increase 5.3 ASME Overpressurization Analysis Reference 9.3 Event MSIV Closure Single Failure MSIV Position Scraa Trip Maximum Pressure 1347.6 psig Maximum Sensed Pressure 1323.1 psig 5.4 Control Rod Withdrawal Error (CRWE)

Starting Control Rod Pattern for Analysis Figure 5.1 Rod Distance Block Setting Withdrawn ACPR 106 4.5 ft. .11 107 4.5 .11 108 5.5 .14 5.5 .14 109(2) 110 6.5 .16 5.5 Fuel Loading Error ACPR 0.19 (1) Results of Limiting Transient Evaluations reported in Section 5.6.  ;

(2) Rod Block setting of 110% selected for Cycle 9 operation. l l

5 XN-NF-83-47 5.6 Determination of Thermal Margins Table 5.1 MCPR Operating Limits at Rated Conditions Fuel Type MCPR Operating Limit ENC 8x8 1.30 GE 8x8, 8x8R 1.30 MCPR Operating Limits at Off-Rated Conditions Automatic Flow Control Figure 5.3a All Conditions Figure 5.3b Limits established by recirculation flow increase transient 6.0 POSTULATED ACCIDENTS 6.1 LOSS OF COOLANT ACCIDENT 6.1.1 Break Location Spectrum Reference 9.5 a 6.1.2 Break Size Spectrum Reference 9.5

' 6.1.3 MAPLHGR Analyses for ENC XN-1 and XN-2 fuel Reference 9.6 Limiting Break: Double-Ended Guillotine Break Recirculation Pump Suction Line 1 1.0 Break Coefficient Assembly Peak Average Local Peak Clad Burnup MAPLHGR MWR Temperature (MWD /MT) (kW/ft) (%) (OF)

0. 13.0 .8 1879 10,000 13.0 1.0 1942 15,000 13.0 1.7 2123 18,000 12.85 1.9 2159 20,000 12.6 1.5 2074

( 25,000 11.95 1.2 2011

( 30,000 11.2 4.5 2153 35,000 10.45 1.4 1808

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6 XN-NF-83-47 6.2 CONTROL R0D DROP ACCIDENT (1) See XN-NF-80-19, Vol. 1 4 Dropped Control Rod Worth 0.0062 Doppler Coefficient (7730F) -10.18 x 10-6 1/K AK/AT, (OF)-1 Effective Delayed Neutron Fraction 0.0055 Four Bundle Local Peaking Factor 1.232 Maximum Deposited Fuel Rod Enthalpy 85 7.0 TECHNICAL SPECIFICATIONS 7.1 LIMITING SAFETY SYSTEM SETTINGS 7.1.1 MCPR Fuel Cladding Integrity Safety Limit All Fuel Types 1.05 f

, 7.1.2 Steam Dome Pressure Safety Limit Pressure Safety Limit 1345 psig 7.2 LIMITING CONDITIONS FOR OPERATION 7.2.1 Average Planar Linear Heat Generation Rate for ENC XN-1 and XN-2 8X8 fuel Bundle Average Exposure MAPLHGR (MWD /MT) (kW/ft) 0 13.0 10,000 13.0 15,000 13.0 18,000 12.85 20,000 12.6 25,000 11.95 30,000 11.2 35,000 10.45 (1) Reported for limiting G.E. control blade. See Appendix B for evaluation i of ASEA-ATOM development control blade. 1

L 7 XN-NF-83-47 7.2.2 Minimum Critical Power Ratio Fuel Type MCPR ENC XN-1, XN-2 1.30

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GE 8x8, 8x8R 1.30 Reduced Flow MCPR Limits Automatic Flow Control Figure 5.3a All Conditions Figure 5.3b 7.2.3 Surveillance Requirements See Appendix A.

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f 8 XN-NF-83-47 9.0 ADDITIONAL REFERENCES 9.1 S. F. Gaines, " Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel," XN-NF-81-21(A), Revision 1 (January

{ 1982).

9.2 R. H. Kelley, " Exxon Nuclear Plant Transient Methodology for Boiling Water Reactors," XN-NF-79-71, Revision 2 (November 1981).

9.3 R. H. Kelley, "Dresden Unit 3 Cycle 9 Plant Transient Analysis,"

XN-NF-83-58 (July 1983).

9.4 R. H. Kelley, "Dresden Unit 3 Analyses for Reduced Flow Operation," XN-NF-81-84 (December 1981).

9.5 J. E. Krajicek, " Generic Jet Pump BWR/3 LOCA Analysis Using the ENC EXEM Evaluation Model," XN-NF-81-71(A) (October 1981).

9.6 J. E. Krajicek, "Dresden Unit 3 LOCA Analysis Using the ENC EXEM Evaluation Model; MAPLHGR Results," XN-NF-81-75(P) (November 1981) and Supplement 1 (July 1983).

9.7 J. C. Chandler, "Dresden Unit 3 Cycle 8 Reload Analysis," XN-NF 81-76, Revision 1 (December 1981).

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9 XN-NF-83-47

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Table 4.1 Dresden 3 Reload Batch XN-2 Neutronic Design Values Fuel Pellet Reference 9.1 Fuel Rod Reference 9.1 Fuel Assembly Reference 9.1 Fuel Assembly Loading, KgUO2 196.5 Fuel Assembly Lc6 ding, KgU 173.2 Core Data Number of fuel assemblies 724 Rated thermal power, MW 2527 Rated core flow, 106 lbm/hr 98.0 Core inlet subcooling, BTU /lbm 24.6 Moderator temperature, cF 546 Channel thickness, inch 0.080 Channel inside face-to-face dimension, inch 5.278 Fuel assembly pitch, inch 6.0

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Wide water gap thickness, inch 0.750 l Narrow water gap thickness, inch 0.374 Control Rod Data (1)

Absorber material BC4 Total blade span, inch 9.750 Total central support span, inch 1.562 Blade thickness, inch 0.3120 (1) Applies to G.E. fabricated control blades. Refer to Appendix B for description of developmental contorl blades fabricated by ASEA-ATOM.

10 XN-NF-83-47 l

Table 4.1 Design 3 Reload Batch XN-2 Neutronic Design Values (Cont.)

Blade face-to-face internal dimension, inch U.200 Absorber rods per blade 84 Absorber rod outside diameter, inch 0.188 Absorber rod inside diameter, inch 0.138 Absorber density, % of theoretical 70 i

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Table 5.1 Determination of Thermal Margins Indicated Maximum Maximum Maximum MCPR Event Model Exposure Power Flow Heat Flux Power Pressure Limit (2)

LRWB COTRANSA EOC9 100% 100% 112.5%(3) 300%(3) 1273(3) psig 1.30(4)

FWCF COTRANSA EOC9 100% 100% 115.9% 260% 1196 1.26 LFWH PTSBWR3 E0C9 100% 100% 118.5% 120% 1039 1.21 CRWE(1) XTGBWR BOC9 100% 100% - - -

1.21 C

i (1) Rod Block setting of 110% selected for Cycle 9 operation.

(2) Indicated limits applicable to both ENC 8x8 fuel and G.E. 8x8 fuel.

(3) Nominal case results; all other events bounding case results.

(4) Statistically determined value; all other events bounding value.

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WI DE Figure 3.2 Dresden-3 Cycle 9 Safety Limit Local Peaking Note 1: Enrichment distribution is explained in Figure 4.1.

14 XN-NF-83-47

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WIDE LL ---

1.35 W/O U235 f L ---

2.00 W/O U235 ML ---

2.37 W/O U235 M ---

3.51 W/O U235 H ---

3.76 W/O U235 ML* ---

2.37 W/O U235 + 3.50 W/O GD203 W --- INERT WATER ROD FIGURE 4.1 ENRICHMENT DISTRIBUTION FOR FUEL TYPE XN-2 8x8 (Enriched Lattice 3.02 w/o U-235)

15 XN-NF-83-47 l C2 G0 F1 C2 C2 G0 F1 C2 C2 GO F1 C2 C2 F1 E3 G0 C2 G0 F1 F1 F1 G0 F1 F1 F1 GO F1 GO D3 E3 F1 G0 F1 G0 C2 GO C2 G0 C2 G0 F1 GO F1 F1 E3 C2 F1 G0 C2 C2 F1 F1 C2 C2 F1 G0 C2 G0 C2 A3 C2 F1 C2 C2 C2 G0 C2 C2 C2 GO F1 GO F1 E3 E3 G0 F1 GO F1 GO F1 G0 F1 GO F1 G0 C2 E3 A3 F1 G0 C2 F1 C2 G0 C2 GO F1 GO F1 F1 E3 C2 F1 G0 C2 C2 F1 G0 C2 C2 F1 F1 F1 A3 C2 F1 C2 C2 C2 G0 F1 C2 C2 G0 C2 C2 E3 G0 F1 GO F1 GO F1 GO F1 G0 C2 E3 B3 F1 G0 F1 - GO F1 G0 F1 F1 C2 E3 B3 C2 F1 G0 C2 G0 C2 F1 F1 C2 B3 i C2 G0 F1 GO F1 E3 E3 A3 l E3 F1 D3 F1 C2 E3 A3 E3 E3 E3 A3 E3 X = Fuel Type -

Y = Cycles irradiated Fuel Number of Type Assemblies Description A 24 GE 8x8 2.50 w/o U-235 (reinserted assemblies)

B 12 GE 8x8 2.62 w/o U-235 (reinserted assemblies) f C 200 GE 8x8R 2.65 w/o U-235 0 8 GE 8x8 2.50 w/o U-235 E 72 GE 8x8 2.62 w/o U-235 F 224 XN-18x8 2.69 w/o U-235 G 184 XN-2 8x8 2.83 w/o U-235 Figure 4.2 Dresden Unit 3 Cycle 9 Reference loading Pattern By Fuel Type (One Quarter of Symetrical Core Loading) 1 l

16 XN-NF-83-47 1.0 -----------------------

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A-1 XN-NF-83-47

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l APPENDIX A SURVEILLANCE REQUIREMENTS l

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I A-2 XN-NF-83-47 APPENDIX A - SURVEILLANCE REQUIREMENTS The thermal margin (MCPR) requirements associated with the generator load rejection transient without bypass to the condenser (LRWB) are based on a statistical combination of uncertainties in calculated parameters and measured plant performance in the area of control rod drive performance. The Plant Technical Specifications require that control rod drive performance be monitored on an individual rod basis at regular intervals. This Appendix provides for modification of MCPR operating limits if the measured control rod drive performance falls outside the statistical basis used in the thermal margin calculation.

For a mean control rod insertion time to 90% insertion of 2.58 seconds or s less, the MCPR operating limits established by the statistical evaluation of the LRWB transient are valid. For a mean 90% insertion time corresponding to the Technical Specification limit of 3.50 seconds, an additional thermal margin conservatist of 0.05 is required. Between those two values, the MCPR operating limit should be determined by the following formula:

i MCPRs= MCPRa + 0.054T - 0.140 where:

MCPR S = Operating Limit MCPR adjusted for observed scram time statistical behavior; MCPRa= Operating Limit MCPR obtained from cycle analysis; and T = Statistical mean of observed scram insertion times to the 90% insertion point.

R.

B-1 XN-NF-83-47

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APPENDIX B ASEA-ATOM DEVELOPMENTAL CONTROL BLADES 1

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B-2 XN-NF-83-47 L

B.1 PROGRAM DESCRIPTION Cycle 9 operation for the Dresden Unit 3 nuclear plant will utilize a total of eight (8) ASEA-ATOM (AA) control blades. Four of these blades will contain a single control zone wherein the primary control material will be B 4C. The remaining four blades will contain twc control zones: a zone equivalent to that of the first set of AA blades and a second zone containing hafnium as the primary control material. This second zone is located at the top six inches of the control blade.

The fuel management strategy projected for Cycle 9 operation is the

} Single Rod Sequence (SRS) mode. The core loading consists of grouping higher exposed fuel assemblies around a single centrol blade withdrawal sequence. It is only these control blades which are scheduled for insertion and used for core control during the cycle. In order to acquire operational experience, the locations of the AA control blades were selected such that the AA control blades constitute part of the single withdrawal sequence. Consequently, the AA control blades will reside in relatively low reactivity (low control rod worth) positions within the core. The locations of the AA control blades are shown in Figure B.1 B.2 LICENSING ANALYSIS On an equal fuel assembly basis, the reactivity worth of both types of AA control blades is estimated to be approximately 9% higher than anticipated for the current control blade worths (AA Report #TR-BR-82-98). Hence, the establishment of the Cycle 9 operating limits has included an evaluation of the impact of the AA control blades upon these limits. A description cf the assumptions and analyses performed in support of this program is provided,

B-3 XN-NF-83-47 S.2.1 Plant Transient Analysis The analyses associated with establishing reactor operating limits as a result of anticipated plant transients utilized a scram reactivity worth associated with the core comprised of all control blades of the current design. As a consequence, the scram reactivity worth does not take credit for the higher worth associated with the AA control blades and envelopes the anticipated scram reactivity associated with the core comprised of the current control olades and the eight AA blades. Hence, the reactor operating limits reported for Cycle 9 are conservative with respect to the utilization of the AA control blades.

B.2.2 Loss-of-Coolant Accident Analysis The analysis of the postulated loss-of-coolant accident utilizes a scram reactivity worth curve which conservatively envelopes the anticipated scram curve expected for Cycle 9. This curve is subsequently used in deternining the operating limits (MAPLHGR) for Cycle 9 and as such conserva-tively envelopes the scram reactivity curve expected with the utilization of the AA control blades.

B.2.3 Control Rod Withdrawal Error Analysis The analysis of an inadvertent withdrawal of a control rod (CRWE) was performed utilizing the approved methodology (l) and explicitly modeled the presence of the eight AA control blades. Hence, the limiting control rod pattern was selected including any effect due to the presence of the AA control blades. As was previously indicated, the withdrawn control rod was determined not to be an AA control rod. Hence, the results of Section 5.4 report the limiting CRWE for Cycle 9. This result is primarily due to locating the AA blade in a low reactivity position within the core. Any change in the

(

B-4 XN-NF-83-47' location of the AA control blades for reactor operation in cycles subsequent

-to Cycle 9 will be reported and analyzed for those reactor cycles.

B.2.4 Control Rod Drop Accident Analysis The approved methodology for the analysis of the dropped control rod (l) correlates the fuel rod deposited enthalpy to four independent parameters; control rod worth, Doppler coefficient, delayed neutron fraction, and four bundle local peaking. Of these four, only the control rod worth is affected by the utilization of the AA cantrol blades. The values of the four independent parameters for Cycle 9 and the highest worth AA control blades are:

Dropped Rod Worth .0032 f

Doppler Coefficient -10.18x10-6{A},(op)-1

+ Delayed Neutron Fraction 0.0055 Four Bundle local Peaking 1.116 The resulting fuel rod deposited enthalpy is 36 cal /gm, which is well f within the current 280 cal /gm limit.

B.2.5 Fuel Bundle Loading Error Analysis The use of AA developmental control blades has no impact on the analysis

-of fuel bundle mislocation and misorientation errors.

{

(1) Safety Evaluation for the Exxon Nuclear Company Topical Report: " Exxon Nuclear Methodology for Boiling Water Reactors - Neutror,ics Methods for Design and Analysis," [XN-NF-80-19(P)], Volume 1, dated May 1980,

, Supplements 1 and 2 dated April 1981.

e B-5 XN-NF-83-47

)

i I i i l i l (

A B C 0 F G H J K L M N P R (

ASEA-ATOM Two Zone Blade Z

ASEA-ATOM Single Zone Slade 1

General Electric Blades Figure B.1 Control Blade locations for Dresden Unit 3

XN-NF-83-47 c

Issue Date: 8/1/83 DRESDEN UNIT 3 CYCLE 9 RELOAD ANALYSIS Distribution JC Chandler RE Collingham GC Cooke SL Garrett SE Jensen WV Kayser RH Kelley TL Krysinski JL Maryott GF Owsley RB Stout RI Wescott HE Williamson CECO /LC O'Malley (60)

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