ML17199U412

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Cycle 11 Reload Analysis
ML17199U412
Person / Time
Site: Dresden Constellation icon.png
Issue date: 09/30/1987
From: Jason White
SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER
To:
References
ANF-87-097, ANF-87-97, NUDOCS 8803150337
Download: ML17199U412 (47)
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Text

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ANF-87-097 ADVANCED NUCLEAR FUELS CORPORATION I *

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DRESDEN UNIT 3 CYCLE 11 RELOAD ANALYSIS SEPTEMBER 1987 AN AFFILIATE OF KRAFTWERK UNION

(;f}KWU

ERRATA SHEET FOR THE DRESDEN UNIT 3 CYCLE 11 RELOAD ANALYSIS REPORT, ANF-87*097 Please-make the following corrections on Page 6:

BOC *cold k*eff, all rods out BOC cold k-eff, all rods in BOC cold k-eff, strongest rod out

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Number Was J.1071 0.9573

. 0.9844.

Number Should Be 1.1l03 0.9605 0.9876 RAPIFAX ND.

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ADVANCED NUCLEAR FUELS CORPORATION Prepared By:

DRESDEN UNIT 3 CYCLE 11 RELOAD ANALYSIS J. A. White BWR Safety Analysis ANF-87-097 Issue Date: 9/30/87 Licensing and Safety Engineering Fuel Engineering and Technical Services AN AFFILIATE OF KRAFTWERK UNION

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CUSTOMER DISCLAIMER IMPORTANT NOTICE REGARDING CONTENTS AND USE OF THIS DOCUMENT Pl.EASE READ CAREFULLY Advanced Nuclear Fuels Corporation's wanantles and representations con-cerning the subject matter of this document are those set forth in the Agreement between Advanced Nuclear Fuels Corporation and the Customer pursuant to which this document is issued. Accordingly, except as otherwise expressly pro-vided in such Agreement, neither Advanced Nuclear Fuels Corporation nor any person acting on its behalf makes any warranty or representation. expressed 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 dlsctosed in this document will not infringe privately owned rights; or assumes any liabilitles with respect to the use of any information, ap-paratus, method or process disclosed in this document.

The information contained herein is for the sole use of Customer.

In order to avoid impairment of rights of Advanced Nuclear Fuels Corporation in patantS or.inventions which may be included in the Information contained in this d!leument. the *recipient, by its acceptance of this document, agrees not to publish or make public use (in the patent use of the term) of such information until*

so authorized in writing by Advanced Nuclear Fuels Corporation or until after six (6) months following termination or expiration of the aforesaid Agreement and any extension thereof, unless otherwise expressly provided in the Agreement. No rights or licenses in or.to any patents are implied by the furnishing of this docu-ment.

XN-NF*F00,765 (1/87)

- i -

ANF-87-097 TABLE OF CONTENTS Section Page

1. 0 INTRODUCTION........ ;.. *.. *.......... -...............................

2.0 FUEL MECHANICAL DESIGN ANALYSIS...................................

3.0 THERMAL HYDRAULIC DESIGN ANALYSIS.................................

3. 2 Hydraulic Characterization.........................................
3. 2.1 Hydraulic Compatibility.............................. *:::*............ *
  • 3.2.3 Fuel Centerline Temperature.......................... ~::...........

3.2.5 Bypass Flow........................................................

3.3 MCPR Fuel Cladding Integrity Safety Umit. ~..................... *.'..

3. 3. 1. Coo 1 ant Thermodynamic Condition....................................

3.3.2 Design Basis Radial Power Distribution..................,......... -

3.3.3 Design Basis Local Power Distribution........ ;....................

  • 4.0 NUCLEAR DESIGN ANALYSIS...........................................

4.1 4.2

4. 2.1 4.2.2 4.2.4 5.0 5.1 5.2 5.3 5.4 5.5 5.6 5.7 6.0

.6.1 Fuel Bundle N~clear Design Analysis...........* ******:****........

Core Nuclear Design Analysis......................................

Core Configuration.......................... '......................

Core Reactivity Characteristics...................................

Core Hydrodynamic Stabi 1 i ty.......................................

ANTICIPATED OPERATIONAL OCCURRENCES...............................

Analysis Of Plant Transients At Rated Conditions..................

Analysis For Reduced Flow Operation................... ~***********

Analysis For Reduced Power Operation................... :*********;

ASME Overpressurization Analysis..................................

Control Rod Withdrawal Error......................................

Fuel Misleading Error.............................................

Determination Of Thermal Margins..................................

POSTULATED ACCIDENTS..............................................

Loss-Of-Coolant Accident..........................................

1 2

3 3

3 3

3 3

3 4

4 5

  • 5 5

5 6

6 7

7 7

7 7

8 8

8 10 10

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  • Section 6.1.1 6.1. 2 6.1.3 6.2 7.0 7.1 7.1.1 7.1.2 7.2 7.2.l 7.2.2 7.2.3 7.3 7.3.1 8.0 9.0

- ii -

TABLE OF CONTENTS (Continued)

ANF-87-0 Page Break Location Spectrum........................................... 10 Break Size Spectrum................. *.......... -............... -.....

  • 10 MAPLHGR Ana 1 yses..................................................

10 Control Rod Drop Accident............ 1........................... 11 TECHNICAL SPECIFICATIONS;.............. ~............ ~............. 12 Limiting Safety System Settings................................... 12 MCPR Fuel Cladding Integrity Safety Limit ***.*.................... 12 Steam Dome Pressure Safety Limit........*......................... 12 Limiting Conditions For Operation................................. 12 Average Planar Linear Heat Generation Rate........................ 1 Minimum Critical Power Ratio....................................... 1 Linear Heat Generation Rate....................................... 13 Survei 11 ance Requi remenfs..................... *.................... 13 Scram Insertion Time Surveillance........ :..... ~.................. 13 METHODOLOGY

REFERENCES:

15 ADDITIONAL REFERENCES.~*-...........................................

16 APPENDIX A - SINGLE LOOP OPERATION A.l ANTICIPATED OPERATIONAL OCCURRENCES........................ ~...... A:-1 A.2 POSTULATED ACCIDENTS................. :............................ A-3 REFERENCES.-.*..................................... *................ A-4

- iii ANF-87-097 LIST OF TABLES Table Page 4.1 Neutronic Design Values........................................... 23 Figure 3.1 3.2

  • ' e!:!

3.5 3.6 4.1 4.2 LIST OF FIGURES Hydraulic Demand Curves For Dresden.... _,_........................ '..

17 Design Basis Radial Power Distribution............................ 18 Design Basis Local Power Distribution For D3 9x9 3.35-9Gd4.0...... 19' Design Basis Local Power Distribution For D3 9x9 3.35-8Gd4.0...... 20 Design Basis Local Power Distribution F_g_r D3 8x8 3.02-6Gd3.0...... 21 Design Basis Local Power Distribution For D3 8x8 2.87-5Gd3.0...... 22 Enrichment Distribution.For Dresden Unit 3 Reload Batch XN-4L 9D3. 35-9Gd4. 0................ :.................................. ;::.

24 Enrichment Distribution For Dresden Unit 3 Reload Batch XN-4H 9D3. 35-9Gd4. 5................... *.... -.................................

25 4.3 Enrichment Distribution For Dresden Unit 3 Reload Batch XN-4 9D3. 35-9Gd3. 0..................................................... 26 4.4 Dresden Unit 3 Cycle 11 Reference Loading Map By Fuel Type (One Quarter Of Symmetric Core Loading)...................... 27 4.5 Decay Ratio vs. Reactor Power For Dresden 3 Cycle 11.............. 28 5.1 Starting Control Rod Pattern For Control Rod Withdrawal Analysis..

29 5.2 Reduced Flow MCPR Limit - All Conditions.......................... 30 5.3 Reduced Flow MCPR Limit - Automatic Flow Control (8x8 Fuel)....... 31 5.4 Reduced Flow MCPR Limit - Automatic Flow Control (9x9 Fuel)....... 32 7.1 Reduced Flow MCPR Technical Specification Limit - All Conditions..

33 7.2 Reduced Flow MCPR Technical Specification Limit - Automattc Flow Control (8x8 Fuel)........................................... 34 7.3 Reduced Flow MCPR Technical Specification Limit - Automatic Flow Control (9x9 Fuel)........................................... 35

1 ANF-87-097

1.0 INTRODUCTION

This report provides the results of the analysis performed by Advanced Nuclear Fuels (ANF) in support of the Cycle 11 reload for Dresden Unit 3.

This report is intended to be used in conjunction with the ANF topical report XN-NF 19(P), Volume 4, Revision 1, "Application of the ENC Methodology to BWR

Reloads, 11 which describes the analyses performed in support of this reload, identifies the methodology used for those analyses, and provides a generic reference list.

Section numbers in this report are the same as corresponding section numbers in XN-NF-80-19(P)(A), Volume 4, Revision 1.

The Dresden Unit 3 Cycle 11 core is to comprise a total of 724 fuel assemblies, including 168 unirradiated ANF XN-4 9x9 assemblies, 176 irradiated ANF 9x9 assemblies, and 380 irradiated ANF 8x8 assemblies.

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 Dresden Unit 3 during the previous operating cycle.

2 ANF-87-097 2.0 FUEL MECHANICAL DESIGN ANALYSIS Applicable ANF Fuel Design Report:

Reference 9.1 To assure that the expected power history for both the 8x8 and 9x9 fuel to be irradiated during Cycle 11 of Dresden Unit 3 is bounded by the assumed power history in the fuel mechanical design analysis, LHGR operating 1 imits have been specified.

In

addition, LH_GR operating limits for Anticipated Operational Occurrences have been specified.

3 ANF-87-097 3.0 THERMAL HYDRAULIC DESIGN ANALYSIS 3.2 Hydraulic Characterization 3.2.1 Hydraulic Compatibility Component hydraulic resistances for the constituent fuel types inithe Dresden 3 Cycle 11 core have been determined in single phase flow tests of full scale assemblies.

Figure 3.1 i 11 ustrates the hydraulic demand curves 1 for ANF 9x9 fuel and ANF 8x8 fuel in the Dresden Unit 3 core.

The similar hydraulic performance indicates adequate compatibility for all the co-residence fuel in the Dresden core.

3.2.3 Fuel Centerline Temperature 3.2.5 Applicable Generic Report Bypass Flow Calculated Bypass Flow Fraction at 100% power/108% flow 3.3 MCPR Fuel Cladding Integrity Safety Limit Safety Limit MCPR 1.05 3.3.1 Coolant Thermodynamic Condition Rated Thermal Power Feedwater Flowrate (at SLMCPR)

Steam Dome Pressure (at SLMCPR)

Feedwater Temperature Reference 9.1 10.5%

2527 MWT 12.4 Mlb/hr 1020 psia 340°F

4 ANF-87-0.

3.3.2 Design Basis Radial Power Distributfon See Figure 3.2 3.3.3 Design Basis Local Power Distribution See Figures 3.3 through 3.6

'* 5 ANF-87-097 4.0 NUCLEAR DESIGN ANALYSIS 4.1 Fuel Bundle Nuclear Design Analysis Assembly Average Enrichment Radial Enrichment Distribution ANF XN-4L 9x9 ANF XN-4H 9x9 Axial Enrichment Distribution Axial Burnable Poison Distribution 9 rods at 4.0 w/o Gd203 9 rods at 4.5 w/o Gd203 3.13%

Figure 4.1 Figure 4.2 Uniform 3.35% with 6" natural top blanket Figure 4.1 Figure 4.2 9 rods at 3.0 w/o Gd203 Figure 4.3 Note:

The burnable poison is reduced (3.0 w/o Gd203) in the top six inches and bottom twelve inches of the enriched length of the designated rods.

The central zone contains either 4.0 w/o Gd203 or 4.5 w,/o Gd203.

The natural urania axial blanket sections do not contain burnable absorber material.

Non-Fueled Rods Neutronics Design Parameters Maximum Lattice K-infinity Figures 4.1 and 4.2 Table 4.1 1.235 4.2 Core Nuclear Design Analysis 4.2.1 Core Configuration Core. Exposure at fOClO, MWd/MTU Nominal Value Shutdown Reactivity Calculations Core Exposure at BOCll, MWd/MTU Core Exposure at EOCll, MWd/MTU Figure 4.4 21,625 21,113 14,548 23,429

ANF-87-o*

Note:

Cycle 11 safety analyses are_valid for EOC10 exposure from

-513 MWd/MTU to +513 MWd/MTU from the nominal value reported above.

4.2.2 Core Reactivity Characteristics 4.2.4 BOC Cold K-eff, All Rods Out BOC Cold K-eff, All Rods In BOC Cold K-eff, Strongest Rod Out Reactivity Defect (R-Value)*

Standby Liquid Control System Reactivity, Cold Conditions, 600 ppm Core Hydrodynamic Stability*

Maximum Decay Ratio Value Natural Recirculation at 100%

Flow Control Line

-L 19H /. 110J

{).9§73- ().Of60f 0.984'4J O,t:j'/7t fl, fl,

0. 0017 fiP.';-Pr'fll ftl~(j,(

0.9381

/'fu 3/4/j9 Figure 4.5 0.35

  • Includes 0.0004 to account for B4C settling in control rod tubes.

7 ANF-87-097 5.0 ANTICIPATED OPERATIONAL OCCURRENCES 5.1 Event LRWB

.FWCF LFWH 5.2

. 5.3 Applicable Generic Transtent Analysis Report Reference 9.2 Analysis Of Plant Transients Reference 9.3

Load Rejection Without Bypass (LRWB)

Feedwater Controller Failure (FWCF)

Loss of Feedwater Heating (LFWH)

Maximum Maximum Maximum Delta Power Flow*

Heat Flux Power Pressure CPR**

Model 100%

108%

115%

319%

1229 psig

.33 COT RAN SA 100%

108%'

117%

241%

1199 psig

.23 COTRANSA 100%

108%

119%

120%

1049 psig

.19 PTSBWR3

  • Bounding for 100% flow.
    • Delta-CPR results for most limiting fuel type from bounding analysis.

Analysis For Reduced Flow Operation Reference 9.3 Limiting Transient: Recirculation Flow Increase Transient (RFIT)

Analysis For Reduced Power Operation Reference 9.3 Limiting Transient:

Load Rejection Without Bypass (LRWB) 5.4-ASME Overpressurization Analvsis Limiting Event Worst Single Failure Maximum Pressure Maximum Steam Dome Pressure MSIV Closure Direct-Scram 1324 psig 1297 psig

8 5.5 Control Rod Withdrawal Error Starting Control Pattern for Analysis 100% Flow*

Rod Block Distance Setting Withdrawn

'105%

5.0 ft 106%

5.0 107%

5.5 108%

6.5 109%

7.5 l10%

8.5

  • More limiting than 108% flow.

5.6 Fuel Misloading Error Maximum LHGR Minimum MCPR Maximum Delta CPR 5.7 Determination Of Thermal Margins Summary of Thermal Margin Requirements LRWB FWCF LFWH CRWE 100%

100%

100%

100%

108%

108%

108%

100%

Delta CPR 9x9 8x8 0.20 0.20 0.22 0.21 0.26 0.24 0.28 0.26 0.29 0.28 0.31 0.30 Delta CPR*

0.23/0.26 0.20/0.23 0.18/0.19 0.30/0.31

  • Values for 8x8/9x9 fuels, respectively.
    • MCPR limits for 110% rod block monitor setting.

ANF-87-0

. *.Figure 5.1

  • 13.8 kW/ft
1. 53 0.19 MCPR Limit*

1.28/1.31

1. 25/1. 28
1. 23/ 1. 24 1.35/1.3.

9 MCPR Operating Limits at Rated tonditions*

Fuel Type 9x9 Fuel 8x8 Fuel MCPR Limit 1.36 1.35

  • MCPR limits for 110% rod block monitor setting.

MCPR Operating Limits at Off Rated Conditions Reduced Flow MCPR Limits Manual and Automatic Flow Control Automatic Flow Control - 8x8

- 9x9 ANF-87-097 Figure 5.2 Figure 5.:3

.Figure 5.~4

10 ANF-87-097 6.0 POSTULATED ACCIDENTS 6.1 Loss-Of-Coolant Accident 6.1.1 Break Location Spectrum Reference 9.5 6.1.2 Break Size Spectrum Reference 9.5 6.1.3 MAPLHGR Analyses Reference 9.6 The MAPLHGR limits of Reference 9.6 are valid for the Dresden 3 9x9 (XN-3 and XN-4) fuels for Cycle 11 operation.

Limiting Break:

Double ended guillotine pipe break Recirculation pump suction line 1.0 Discharge Coefficient Bundle Average Peak Clad Peak Local Exposure MAPLHGR Temperature MWR (MWd/MTU)

(kw/ft)

( o F)

(%)

0 11.40 2006 2.2 5000

11. 75 2045 2.4 10000 11.40 1893

.9 15000 10.55 1805

.6 20000 9.70 1710

.4 25000 8.85 1623

.3 30000 8.00 1529

.2 35000 7.15 1421

.1 40000 6.30 1309

. 1

11 6.2 Control Rod Drop Accident Dropped Control Rod Worth*

Doppler Coefficient, 1/k dk/dl Effective Delayed Neutron fraction Four-Bundle Local Peaking Factor Maximum Deposited Fuel Rod Enthalpy, cal/gm ANF-87-0

- 0.0119

-11.1 x io-6(°F)-l 0.0051 1.40

. 187

12 ANF-87-097 7.0 TECHNICAL SPECIFICATIONS 7.1 Limiting Safety System Settings 7.1.1 MCPR Fuel Cladding Integrity Safety Limit MCPR Safety Limit (All Fuels) 1.05 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 Limits for 8x8 and 9x9 Fuels Bundle Average 9x9 8x8 Exposure MAPLHGR MAPLHGR (MWd/MTU)

CkW/ftl (kW/ft) 0 11.40 13.00 5000

11. 75 13.00 10000 11..40 13.00 15000 10.55 13.00 18000 12.85 20000 9.70 12.60 25000 8.85 11.95 30000 8.00
11. 20 35000 7.15 10.45 40000 6.30

13 7.2.2 Minimum Critical Power Ratio 7.2.3 7.3

).3.1 Rated Conditions MCPR Limits Fuel Type 9x9 Fuel 8x8 Fuel Off-Rated Conditions MCPR Limits Manual and Automatic Flow Control Automatic Flow Control - 8x8 Fuel

- 9x9 Fuel linear Heat Generation Rate 8x8 Fuel Planar Average Exposure LHGR MWd/MTU kW/ft 0

16.0 25.4 14.1 42.0 9.3 Surveillance Requirements Scram Insertion Time Surveillance 1.36 1.35 Figure 7.1 Figure 7.2 Figure 7.3 ANF-87-0 Figures 3.1 and 3.2 of Reference 9.1 9x9 Fuel Planar Average Exposure LHGR MWd/MTU kW/ft 0

14.5 5.0 14.5 25.2 10.8 48.0 7.2 Individual control rod drive insertion times shall be monitored in accordan.

with existing Technical Specification requirements.

If the average inserti time to the 90% insertion point for all the control rod drives in the core,

14 ANF-87-097 based on the most recent observation for each drive, exceeds 2.62 seconds, the MCPR operating limit for each fuel type in the core shall be increased by an amount determined by:

OLMCPR = MCPRs + (Tav - 2.62)

  • MCPRt, where OLMCPR is the MCPR operating limit, MCPRs is the Technical Specification MCPR operating limit based on compliance with the statistical assumptions, Tav is the average control rod insertion time to 90%, and MCPRt is 0.074 for 8x8 fuel and 0.080 for 9x9 fuel.

This surveillance requirement does not supersede the Control Rod* Drive operability requirement or the scram insertion time requirements specified elsewhere.

15 ANF-87-097 8.0 METHODOLOGY REFERENCE~

See XN-NF-80-*19(P)(A),

Volume 4,

Revision 1 for a

complete bibliography.

--~--

16 ANF-87-097 9.0 ADDITIONAL REFERENCES 9.1 "Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel,"

XN-NF-85-67(P)(A),

Revision 1,

Exxon Nuclear

Company, Richland, Washington (September 1986).

9.2 "Exxon Nuclear Plant Transient Methodology for Boiling Water Reactors,"

XN-NF-79-7l(P),

Revision 2, including Supplements 1 through 3(P)(A),

Exxon Nuclear Company, Richland, Washington (November 1981).

9.3 "Dresden Unit 3 Cycle 11 Plant Transient Analysis," ANF-87-096, Advanced Nuclear Fuels Corporation, Richland, Washington (August 1987).

9.4 "Dresden Unit 3 LOCA Analysis Using the EXEM/BWR Evaluation Model," XN-NF-81-75, Exxon Nuclear Company, Richland, Washington.

9.5 "Generic Jet Pump BWR/3 LOCA Analysis Using the EXEM Evaluation Mode.l,".

XN-NF-81-7l(A),

Exxon Nuclear Company,

Richland, Washington (August 1981).

6 "Dresden Unit 3 LOCA-ECCS Analysis MAPLHGR Results for ENC 9x9 Fuel," XN-NF-85-63, Exxon Nuclear Company, Richland, Washington (September 1985).

I. I I ' l, I

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1.2 1.4 1.6 RADIAL POWER PEAKING Figure 3.2 Design Basis Radial Power Distribution I

1. 8 co

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19 ANF-87.

LL 0.97 LL 0.92 l

1.00 L

0.97 ML 1.05 ML 1.05 L

0.96 L

0.98 LL 0.90 LL 0.92 l

0.98 Ml 1.02 M

1.11 1.08 Ml*

0.93 M

1.10..

"L*

0.99 L

0.95 L

1.00 ML J.02 ML*

0.94 M

1.02 H

1.06 H

1.06 M

1.00 1.05 ML 0.99

~-----~--~~----------------------------------------------------

L 0.97 1.11 1.02 H

1.04 H

1.01 H

0.99 H

1.01

.... fll*

0.85 :... "

.1.05

~--~--------------------------------------------------------------

ML M

H H

W H

H H

H

~-,'.::~ __ '. __ '.::~ __ '. __ '.::: __ ~ __ '.::'. __ ~ __ :::: __ : __ ::: __ '. __ ::~~--: __ '.::~ __ '. __ '.:'.It ML I.OS L

0.96 ML*

0.93 M

1.10 H

1.06 M

LOO H

0.99 H

1.01

.. H 0.98.

H f).98 w

0.00 H

0.98 H

0.98 H

1.00

~

1.01 Ml*

0.83 H

'I.09 1.03

~---------------------------------------------~--------~----------

l 0.98 ML*

0.99 M

I.OS ML*

0.8S H

1.02 H

1.01 ML*

0.83 H

1.08 Ml 0.95

~-----------------------------------------------------------

LL 0.90 L

0.9S ML 0.99 1.05.

H 1.10 H

1.09 1.03 "1.

0.95..

l 0.91..

*----~------------------------------------------

Figure 3.3 Design Basis Local Power Distribution For D3 9x9 3.35-9Gd4.0

)*",;

'"-4'"-..

20 ANF-87-097

  • * * * * * * * * * ***** * *** ~ ~ *-* * * * ****** * *

~------------------------------------

LL 0.97 LL 0.93 l

0.97 l

0.96 ML 1.02 ML 1.01 l

0.95 l

0.96 LL 0.91 LL 0.93 l

0.97 l

0.96 l

0.96 Ml*

0.99 1.08 Ml*

0.99 ML 0.97 1.03.

M 1.0$

M 1.03 H

1.05 M

1.07 H

1.07 H

1.03 ML*

0.94 H

1.06 H

1.02 M

1.07 M

1.0.1 H

1.03 ML 0.98 1.04 Ml*

0.88

'l 0.94 ML 0.97

  • . M
    • . 1. 04

~--~*----------------------------------~---------~--------------------------------

ML 1.02 f!L 1.01 l

0.95 M

1.07 ML*

0.94 M

1.07 H

1.07

.. H 1.06 M

1.01 H

1.03 H

l.02 H

.1.03 w

0.00 H

1.01 H

1.00 H

1.01 w

0.00 H

1.00 H

J.00 H

1.00 H

1.01 H

J.03 ff J.03 H

1.08 H

1.08 M

1.03

~---~-------------------------~-------------------~--------------------------

L 0.96 ML 0.98 1.04 ML*

0.88 H

1.03 H

1.03 ML*

0.87 H

J.07 ML 0.96 LL 0.91 L

0.94 ML 0.97 1.04 H

1.08..

H 1.08 M

1.03..

ML 0.96 L

0.92

*-~~------------------------------------------------------

Figure 3.4 Design Basis Local Power Distribution For D3 9x9 3.35-8Gd4.0

21 ANF-87.

LL 1.04 l*

1.02 ML 1.01 Ml 1.00 Ml 0.99 ML 1.00

  • Ml 1.01 l

1.01

---~~-~----~-------------------------------------------------------------

L 1.02...

Ml 1.00 ML*

0.95 1.05 1.04 1.04 ML*

0.94 Ml 0.99 Ml 1.01 Ml 1 *.00 ML*

0.95 1.05 M

1.04 H

1.03 H

1.03 H

1.00 H

1.02 H

1.00 H

1.02 H

1.00 1.02 H

1.01 Ml 0.96 1.03

~-~~-------------------------------------------------------

ML 0.99 M

1.04 H

. - 1.02

.H 1.00 w

0.00 H

0_99*

H 1.00 1.02 ML 1.00 ML 1.01 M

1.04 ML*

0.94 H

1.02 1.02 H

1.00 H

1.01..

H 0.99 H

1.00..

H 0.99 ML*

o.88 ML*

0.88 1.01.

1.02 Ml 0:95

---~-~----~---~---~-~----~---------~-~-----------------------------------

L -

1.01 ML 0.99 ML 0.96 1.02 1.02 Ml 0.95 Ml 0.98

~~----------------------------------------------------------

Figure 3.5 Design Basis Local Power Distribution For D3 8x8 3.02-6Gd3.0

11-:--r-

      • -~

22 ANF-87-097 LL 1.05 L

1.02 L

1.02 ML**

0.97 Ml 1.01 Ml 0.97 ML 1.00 1.04 Ml 0.99 M

1.03 Ml 1.00 M

1.04 tll 1.01 "L*

0.95....

L 1.02 ML 0.99

~----------~

ML 1.01 ML 0.97 M

1.03 H

1.02 H

1.01 H

1.02 1.02 ML_

0.96

~---------------------------------------------------------

. ML 1.00

.1.04 H

1. 02 -
  • H 1.00 H

0.99 H

0.99 H

1.01 1.02 ML 0.99 M

  • l.03 H

1.01 H

0.99 w

0.00 H

0.99 H

1.00 M*

1.02

. ML 1.00 1.04 H

.1.02 H

0.99 H

0.99 H

0.99 Ml*

0.88.

-M

  • 1.02 ML 1.01 l
  • 1.02 ML*

.0.95 Ml 0.99 M

1.02 ML 0.96 H

1.01 1.02 H

1.00 M

1..02 ML*

0.88 M

1.02.

1.01 ML 0.96 Ml 0.96 ML 0.98

.- _____ _. ____ -------------------- ---~ --~----- --~ ----

Figure 3.6 Design Basis local Power Distribution For D3 8x8 2.87-5Gd3.0

23 TABLE 4.1 NEUTRONIC DESIGN VALUES Number of fuel assembl'i es Rated thermal power, MW Rated core flow, Mlbm/hr Core inlet subcooling, Btu/lbm Moderator temperature, F Channel thickness, inch Fuel assembly pitch, inch Wide water 9ap thickness, inch Narrow water tap thickness, ~nch Control Rod Data

  • Absorber material Total blade span, inch Total blade support span, inch Blade thickness, inch Blade face-to-face internal dimension, Absorber rods per blade
  • Absorber rod outstde diameter, inch*

Absorber rod inside diameter, inch Absorber density, % of theoretical

  • Based on actual operating experience.

inch ANF-87-0.

724 2527 98.0 22.7*

546 0.080 6.0 0.750 0.374 B4C, Hf 9.750 1.562 0.3120 0.200 84 0.188 0.138 70

24 ANF-87-097 LL LL L

L ML ML L

L LL LL L

ML M

M : ML*

  • M : ML*

.:

  • L L

ML M~* :

M H

H M

M :

ML L

M M

H H

H H

ML*

M

ML M

H H

W

  • H H

H H

ML ML*

H H

H w

H H

H L

M M

H H

H H

ML*

M L

ML*

M :

ML*. :

H H

ML*

H ML LL L

ML M

H H

M :

ML L

LL Rods ( 5) 1.50 W/O U235 L Rods (12) 2.20 W/O U235 ML Rods (10) 2.84 W/O U235 M Rods (16) 3.72 W/O U235 H Rods (27) 4.34 W/O U235 ML*

Ro_ds ( 9) 2.84-W/O U235 + 4.00 W/O GD203 w Rods ( 2)

Inert Water Rod Figure 4.1 Enrichment Distribution For Dresden U~it 3 Reload Batch XN-4L 903.35-9Gd4.0

.. 1 **

25 ANF-87-0.,

~-----~~--~----------------~-----------~--------------

LL LL
L L

ML ML L

L LL LL L

ML

M
M
ML*
M
ML*

L.

w

. L ML

ML*
M *
H *
H
M
M ML

~-~------~--------------------------~~------

. L

M
M
H H

H H

ML*
M ML
M H

H W : H H

. H
H

;.. ---....,_-___ -------------------------------------------- ~ -----*- ------

ML

ML*
H
H H

w H

H

H

~------------------------------------------------------------

L

M
M H
H
H
H
ML*

M

~=--------------------------------------------------

L

ML*
M
ML*

H

H
ML*
H ML LL
. - L ML
M H
H
M Ml L*

LL Rods ( 5) 1.50 W/O U235 L

Rods (12) 2 :20 W/O U235 ML Rods (10) 2.84 W/O U235 M Rods (16) 3.72 W/O U235 H

Rods (27) 4.34 W/O U235 ML*

Rods ( 9) 2.84 W/0 U235 + 4.50 W/O GD203 w Rods ( 2)

Inert Water Rod Figure 4*2 Enrichment Distribution For Dresden Unit 3 Reload Batch XN-4 903.35-9Gd4.5

26 ANF-87-097

~-------------------------.-----

.. LL LL L

L.

ML

L LL

~-~------~~---~~---~--------------~---------------------J LL L

.. ML

M M
ML*

. M*

. ML*

L L

ML

ML*
. M
H
H
M M

ML

~-----------------~----------------------

L

M M

H H

H

H
ML*
M

~

ML

M H
  • H W

H

H
H ML
ML*

H

H
H W : H
H
H

~---~-----------------------~------~--------------------------------

L

M
M H
H H
ML*
M

~-------------------------------r-~~-------------------------------------

L

ML*
M
ML*
H.
H
ML*

H ML

::------------~--~-----------~------------

LL L

ML M.

H H

M ML L -

--~----------------------------~-----------------------------------------

LL Rods ( 5) 1.50 W/O U235 L

Rods (12) 2.20 W/O U235 ML Rods (10) 2.84 W/O U235 M

Rods (16) 3.72 W/O U235 H

Rods (27) 4.34 W/0 U235 ML*

Rods ( 9) 2.84 W/O U235 + 3.00 W/O GD203 w Rods ( 2)

Inert Water Rod Figure 4.3*

Enrichment Distribution For Dresden Unit 3.Reload Batch XN-4 9D3.35-9Gd3.0

27 82 82 DO 82 82 82 DO 82 82 82 82 DO Cl DO 82 DO Cl DO Cl DO DO Cl A3 Cl DO Cl. A3 Cl DO.

82 82 DO Cl 82 82 DO Cl 82 82 DO 82 82 DO 82 82 82 DO 82 82 Cl 82 DO Cl DO 82 DO Cl DO Cl DO DO Cl A3 Cl DO*

Cl 82 Cl DO Cl 82 DO Cl 82 82 DO Cl 82 82 DO 82 Cl DO 82 82 Cl DO 82 82 Cl 82 DO

'82 DO Cl DO Cl DO Cl A3 DO Cl DO 82 DO Cl A3 Cl Cl A3 82 DO Cl DO Cl Cl A3 A3 A3 A3 82 Cl 82 Cl A3 A3 A3 A3 A3 A3 A3 A3 A3 A3 A3 A3 A3. A3 A3 AJ lv:-1 X = Fuel Type

~

Y = Cycles Irradiated Fue 1-.

Number.

~ Assemblies A

196

. B 184 c

176 D

168 Description XN-1 8x8 2.69 w/o U-235 XN-2 8x8 2.83 w/o U-235 XN-3 9x93.13 w/o U-235 XN-4 9x9 3.13 w/o U-235 DO 82 82 ANF-87-0'9 A3 A3 Cl DO Cl A3 A3 DO Cl 82 A3 A3 82 DO Cl A3 A3 DO Cl A3 A3 A3 Cl Cl A3 A3 A3 A3 A3 Cl A3 A3 Cl A3 A3 A3 A3 I

A3 Figure 4.4 Dresden Unit 3 Cycle 11 Reference Loading Map By Fuel Type (One Quarter Of Symmetric Core Loading)

0 N

0 -

I-c:x:

a::

c:x:

u LLJ a

28 ANF.-87-097 1.0 0.8 0.6 0.4 0.2 0.0 0

... Natural Circulation 100% Rod line 20 40 60 80 PERCENT POWER

.':1 100 Figure 4.5 Decay Ratio vs. Reactor Power For Dresden 3 Cycle 11

/

59 55 51 47 43 39 35 31 27 23 19 15

. 11 7

3*

29 ANF-87-097 2

6 10 14 18 22 26 30 34 38 42 46 50 54 58 59 34 55 40 08 08 40 51 10 36 36 10 47 40 00 06 06 00 40 43 20 20 39 08 16 04 04 16 08 35 34 40 34 31 08 16 04 04 16 24 27 20 23 40 00 06 28 00* - -

40 19 10 36 15 40 08 08 4,0 11 34 7

3 2

6 10 14 18. 22 26 30 34 38 42 46 50 54 58 Note: *Control Rod Being Withdrawn, Rod Positions in Notches, Full *in= 0, Full out= Blank or 48 Figure 5.1

.starting Control Rod pattern for Control Rod Withdrawal Analysis

E-t

~

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~

...:I Cl::

~

u

!1

~

0

...:I rz..

Q rz1 u

Q rz1 Cl::

1.7

,..... ~

8X8 FUELS 1.6

9X9 FUELS 1.5 1.4 1.3 1.2 1.1 1.0....... ---------------------------------------------------------

20 30 40 50 80 70.

80 90..

100 110 120 TOTAL CORE RECIRCULATING FLOW (% RATED, 98 MLB/HR)

Figure 5.2 Reduced Flow MCPR Limit - All Conditions w 0.

)>

z,,

I 00 I

0 l.O

2.0-----------------------------------------------------.......

MCPR OPERATING LIMIT = 1.28 1.9

llCPR OPERATING LIMIT = 1.32 E-4

                      • MCPR OPERATING LIMIT = 1.36

~

1.8

....:3

~

Pot 1.7

\\

u

\\.

~

1.6

\\..

\\..

0

....:3 w

~

1.5 ca rz1 u

1.4 ca rz1

~

1.3 1.2... -__________________________

~

~

~

~

~

~

~

oo

~ m w

TOTAL CORE RECIRCULATING FLOW (% RATED, 98 MLB/HR)

Figure 5.3 Reduced Flow MCPR Limit - Automatic Flow Control (8x8 Fuel)

~

~

!1

~

~

~.

p..*

u

~.

~

0

....:i

~

Cl f'zl u

~

Cl.

rzl

~

2.0-------------------------------------------------------.

1.9 t.8 1.7 1.6 1.5 1.4 1.3

--- :UCPR OPERATING LIM IT -

1.31

MCPR OPERATING LIMIT = 1.35

  • :-* * * * * * * *

',,~~

1.2...... --------------------------------4 20 30 40 50 60 70 80 90 100 110 120 TOTAL CORE RECIRCULATING FLOW (% RATED, 98 MLB/HR)

Figure 5.4 Reduced Flow MCPR Limit - Automatic Flow Control (9x9 Fuel) w N

)>

z

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CX>

I 0

l.O

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~

~

  • ~

u

~

~

0

~

IZ4 c:l rz1 u p

c:l rzl

~

1.7 8X8 FUELS 1.8

9X9 FUELS

\\

\\

\\

\\

1.5

\\

\\

\\

\\

\\ ' '

1.4 1.3 1.2 1.1 1.0-....--...------,....--------._.--------...... -

,20 30 40 50 60 70 80 90 100 110 120 TOTAL CORE RECIRCULATING FLOW (% RATED, 98 MLB/HR)

Figure 7.1 Reduced Flow MCPR Techni 1 Specification Limit - All Conditions w

w

)::>

z "Tl I

CX>

-....J I

0

E-4

!1

~

~

~

u

!1

~

0

~

~

Cl r:r::I u

Cl

  • ~

. 2~0---------------------------------------------------------.

~.9 1.8 1 *. 7 1.6 1.5 1.4 1.3

--- MCPR OPERATING LIMIT = 1.28

MCPR OPERATING LIMIT = 1.32

  • * * * * * * * * *
  • \\ *.

\\ *.

\\.

\\ *.

\\ *.

\\.. ' *..

\\

1.2_,_ ____________________________ _..,.

.20 30 40 50 60 70 80 90 100 110 120 TOTAL CORE RECIRCULATING FLOW {% RATED, 98 MLB/HR)

Figure 7.2 Reduced Flow MCPR Technical Specification Limit - Automatic Flow Control (8x8 Fuel) l>

z,,

I 00 I

0

\\.0

E-t

i!1 t"'.:

i..

2.0-P""----------------------------------------------.

--- MCPR OPERATING LIM IT = 1.31 1.9

MCPR OPERATING LIMIT = 1.35

  • * * * * * * * * *

....:i. 1.8 1."I 1.6 1.5 1.4 1.3

\\..

\\.. '.

\\ *.

'....,~ *.....

~

1.2-+---..----..--...... --....... --...---....... -..... --..... --.... ---.

20 30 40 50 60 "10 80 90 100 110 120 TOTAL CORE RECIRCULATING FLOW (% RATED, 98 MLB/HR)

Figure 7.3 Reduced Flow MCPR Techni~pecification Limit - Automatic Flow Control (9x9 Fuel)

~

w Ui

)>

z I

00 I

0

ANF-87-097 APPENDIX A SINGLE LOOP OPERATION This Appendix provides limits and justification of those limits for Single loop Operati6n (SLO).

A. l ANTICIPATED OPERATIONAL OCCURRENCES Reference A. 1 The NSSS *supplier has provided analyses which demonstrate the safe.ty-of plant operation with a single recirculation loop out of service. for an e~tended period of time.

Theie analyses restrict the overall operation of the plant to lower bundle power levels and lower nodal power levels than are allowed when both reci rcul at ion* systems are in operation.

The physical. interdependence between core power and recirculation flow rate inherently limits the core to less. than rated power.

Because the ANF fuel was designed to be compatible with the co-resident fuel in thermal hydraulic, nuclear, and mechanical design performance, and because the ANF methodology has given results which are consistent with those of previous analyses for normal two-loop operation, the analyses performed by the NSSS supplier for single loop operation are also applicable to single loop operation with fuel and analyses provided by ANF.

For single loop operation, the NSSS vendor found that an increase of 0.01 in the MCPR safety limit was needed to account for the increased flow measurement uncertainties and increased tip uncertainties associated with single pump operation.

ANF has evaluated the effects of the increased fl ow measurement

.uncertainties on the safety limit MCPR and found that the NSSS vendor determined increase in the allowed safety limit MCPR is alS-o -applicable to ANF fuel during single loop operation.

Thus, increasing the safety limit MCPR by 0.01 for single loop operation (1.06) with ANF. fuel is sufficiently conservative to al so bound the increased fl ow measurement uncertainties for single loop operation.

A-2 ANF It is conservative to use the reduced flow two-loop operating MCPR limit or full flow MCPR operating limit plus.01 (whichever is greatest) for single loop operations.

These-limits conservatively bound all transients' from SLO conditions.

The reduced flow MCPR limit.fs to protect against boiling transition during flow excursions to maximum two-pump flow; excursions to such high flows are not possible during *single loop one-pump operation.

Thus, conservatively maintaining this two-loop limit assures that there is even more thermal margin under single loop conditions than under two-loop full power/full flow conditions.

Reactor operation within the limitations which assured adequate stability for the previous cycle will continue to assure adequate stability for Cycle 11.

The stability analyses reported in Figure 4.3 of the main body of this report cover the operating region of Single Loop Operation; the calculated decA ratios are within the acceptable values.

~

-'.{

ANF-87-097 A.2 POSTULATED ACCIDENTS Reference A.2

  • ANF performed LOCA analyses from.single loop conditions and deter.mi.ned the appropr*iate SLO MAPLHGR multiplier of.91 to ANF MAPLHGR's.(Section 7.2) for both the 8x8 and 9x9 fuels.

All calculations were performed with the NRC approved* EXEM/BWR ECCS Evaluation Model according to Appendix K of 10 CFR 50.

I

'*f.

A-4 ANF-87-0 REFERENCES A.I "Dresden Unit 3 Cycle 11 Plant Transient Analysis," ANF-87-096, Advanced Nuclear Fuels Corporation, Richland, Washington (August 1987).

A.2 "LOCA-ECCS Analysis for Dresden.Units During Single Loop Operation with ANF Fuel," ANF-87-111, Advanced Nuclear Fuels Corporation, Richland, Washington (August 1987).

.'" l i

DRESDEN UNIT 3 CYCLE 11 RELOAD ANALYSIS Distribution D. A. Adkisson D. J. Braun

0. C. Brown R. E. Collingham T. P. Currie L. J. Federico S. E. Jensen T. H. Keheley T. L. Krysinski J. L. Maryott J. N. Morgan D. F. Richey D. R. Swope C. J. Volmer R. I. Wescott J. A. White H. E. Williamson CECo/J. M. Ross (60)

Document Control (5)

ANF-87-097 Issue Date: 9/30/87

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