ML20153G305

From kanterella
Jump to navigation Jump to search
Cycle 10 Reload Analysis,Design & Safety Analyses for ENC XN-3 9x9 Reload Fuel
ML20153G305
Person / Time
Site: Dresden Constellation icon.png
Issue date: 09/12/1985
From: Chandler J
SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER
To:
Shared Package
ML17199F643 List:
References
XN-NF-85-57, NUDOCS 8602280226
Download: ML20153G305 (42)
v • d • e


Text

- - -

XN-NF-85-57 DRESDEN UNIT 3 CYCLE 10 RELOAD ANALYSIS Design and Safety Analyses For ENC XN -3 9x9 Reload Fuel SEPTEMBER 1985 RICHLAND,WA 99352 ERON NUCLEAR COMPANY,INC.

p22:88n 88888a.,

PDR

e ~ s ERRATA NO. 1 XN NF-85-57 Dresden Unit 3, Cycle 10 Reload Analysis _

issued document,

1. Make the following pen-and ink changes in the -

XN-NF 85-57:

change

a. Page 3, Section 3.2.3, " Fuel Centerline Temperature;"

"153%" to "124%" in the line " Centerline Temperature at 153%

Power." (Correction of typographical error)

2. Post this sheet inside the front cover of the corrected report.

l l

l f

Xft-NF-85-57 Issue Date: 9/12/85 DREdDEN Uti!T 3 CYCLE 10 RELOAD AtlALYSIS Design and Safety Analyses For ENC XN-3 9x9 Reload Fuel Prepare: (- # v d .-- 9, / h #i" J C. Chandler, Senior Engineer Reload Licensing Concur:

3 .

o ingham_ tanager f 4 ((

BWR Safety An sis f ', *,' '

n a < _ - _

Approve: hbd v

.. Stout,flanager

/2 L MIi Licensing and Safety Engineering Concur: PAV N T.L. KFisTrisk,~ Manager d 9 /I N @

BWR Neutronics Approve: M h k [g P 9AkII H.E. Will(amson, Han_(ger fleutronics and Fuel Management Approve:

$Cl.l}[ zskoe ffSV

. R itter, Hg/jtnager j k h $ G Fuel Engineering and Technical Services

)

I

XN-NF-85-57 NUCLEAR REGULATORY COMMISSION DISCLAIMER IMPORTANT NOTICE REGARDING CONTENTS AND USE OF THIS DOCUMENT PLEASE READ CAREFULLY This technical report was derived through research and development programs spon,ored 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 which 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 information 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 custarters of Exxon Nuclear in their demonstration of compliance with the U.S. Nuclear Regulatory Commission's regulations.

Without derogating from the foregoing, neither Exxon Nuclear nor any person acting on its behalf:

A. Makes any warranty, express or inalied, 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 B. Assumes any liabilities with respect to ;he use of, or for damages resulting from the use of, any information, apparatus, method, or process disclosed in this document.

s i XN-NF-85-57 TABLE OF CONTENTS Section Page

{ Number Number 1.0 IllTRODUCT!0!1..................................................I 2.0 FUEL MECHANICAL DESIGN ANALYSIS.............................. 2 3.0 THERMAL HYORAULIC DESIGN ANALYSIS............................ 2 3.2 HYORAUL IC CHARACTERIZAT !0N. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 3.2.1 Hydraulic Compatibility.................................... 2 3.2.3 Fuel Centerline Temperature................................ 3 3.2.5 BypassF10w................................................3 3.3 MCPR FUEL CLADDING INTEGRITY SAFETY LIMIT.................... 3 3.3.1 Coolant Thermodynamic Condition............................ 3 3.3.2 Design Basis Radial Power Distribution..................... 3 3.3.3 Design Basis Local Power Distribution...................... 3 4.0 NUC L E AR DE S IGfl Afl ALY S I S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 4.1 FUEL BUNOLE NUCLEAR DESIGN ANALYSIS.......................... 4 4.2 CORE flVC L EAR DE S IGN ANALYS I S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 4.2.1 Core Configuration......................................... 4 4.2.2 Core Reactivity Characteristics............................ 4 4.2.4 Core Hydrodynamic Stability................................ 5 5.0 AflTICIPATED OPERAT10flAL 0CCURREllCES.......................... 5 5.1 AllALYSIS OF PLAfli TRAflSIENTS AT RATED CONDIT I0flS. . . . . . . . . . . . . 5 5.2 ANALYSES FOR REDUCED FLOW OPERAT!0fl. . . . . . . . . . . . . . . . . . . . . . . . . . 6 5.4 ASME OVERPRESSURIZATION AflALYSIS............................. 6

11 XN NF-85-57 5.5 CONTROL ROD WITHDRAWAL ERR 0R................................. 6 5.6 FUEL MISLOADING ERR 0R........................................ 6 5.7 DETERMINATION OF THERMAL MARGINS............................. 6 6.0 POSTULATED ACCIDENTS ........................................ 7 6.1 LOSS-0F COOLANT ACCIDENT..................................... 7 6.1.1 Break Location Spectrum.................................... 7 6.1.2 Break Size Spectrum........................................ 7 6.1.3 MAPLHGR Analyses for ENC XN-3 9x9 Fuel....................................................... 7 6.2 CONTROL R0D DRO P ACC I D EN T . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 7.0 TECHNICAL SPECIFICATIONS..................................... 8

{

7.1 LIMITING SAFETY SYSTEM SETTINGS............ ................. 8 7.1.1 Fuel Cladding Integrity Safety Limit....................... 8 7.1.2 Steam Dome Pressure Safety Limit............ .............. 8 7.2 LIMITING CONDITIONS FOR OPERAT!0N............................ 8 7.2.1 Average Planar Linear Heat Generation Rate Limits for ENC XN-3 9x9 Fue1....................................................... 8 7.2.2 Minimum Critical Power Ratio............................... 9 7.3 SURVEILLANCE REQUIREMENTS.................................... 9 7.3.1 Scram Insertion Time Surveillance.......................... 9 7.3.3 Procedural Controls....................................... 10 8.0 METh000 LOGY REFERENCES...................................... 10 9.0 ADD I Il0NA L R E F E REllC E S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11

mei 111 XN-NF-85-57 APPENDICES A. S I N G L E L OO P O P E RA T 10N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 8 B. REFORMULATION OF MCPR LIMITS................................ 31 C. RECIRCULATION PIPE REPLACEMENT PROGRAM...................... 33 l

l

iv XN-NF 85-57 LIST OF TABLES Table Title PELe 4.1 Neutronic Design Va1ues................................. 22 LIST OF FIGURES F3ure Title Pyte 3.1 Hydraulic Ocmand Curves for Dresden Unit 3 Cycle 10 Core........................................... 12 3.2 0(sign Basis Radial Power Distribution.................. 13 3.3 Design Basis Local Power Distribution f o r ENC 9x 9 Fuel Type s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 3.4 Design Basis local Power Olstribution f o r L N C X fi 2 8 x 8 f u e l . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15 3.5 Design Basis local Power Distribution f o r E NC Xfl 1 8 x 8 F ue l . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16 3.6 Design Basis local Power Distribution f o r G . E . 8 x 8 f u e l Ty p e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17 4.1 Enrichment Di stritation for XN 3 9x9 f uel . . . . . . . . . . . . . . .18 4.2 Enrichment Distribution for Xfi 3A 9x9 fuel.............. 19 4.3 Reference Core loading Pattern.......................... 20 4.4 Core Hydrodynamic Stability....................... ..... 21 5.1 Starting Control Rod Pattern for Control Rod W i t hd rawa l E rro r An al y s i s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 3 5.2 Reduced Flow MCPR for All Conditions.................... 24 5.3 Reduced Flow MCPR for Automatic Flow Control............ 25 7.1 Reduced Flow MCPR Technical Specification limit (All Conditions)........................................ 26 7.2 Reduced Flow MCPR Technical Specification Limit (Automatic Flow Control)................................ 27

1 XN NF-85-57 DRESDEN UNIT 3 CYCLE 10 RELOAD ANALYSIS Design and Safety Analyses for ENC XN 3 9x9 Reload Fuel

1.0 INTRODUCTION

This report provides the results of the analyses performed by Exxon Nuclear Company (ENC) in support of the Cycle 10 reload for Dresden Unit 3, which is scheduled to commence operation in Spring 1986. This report is intended to be used in conjunction with ENC topical report XN-NF 80 19(P), Volume 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 generic reference list.

Section numbers in this report are the same as corresponding section numbers in XN NF 8019(P),, Volume 4 Revision 1.

The Dresden Unit 3 Cycle 10 core will comprise a total of 724 fuel assemblics, including 176 untrradiated ENC XN 3 9x9 assemblies. 400 previously irradiated ENC fabricated 8x8 assemblies, and 140 previously l

Irradiated Type P8x8R assemblies fabricated by General Electric. No G.E. f abricated 7x7, 8x8, or 8x0R fuel assemblies are to be irradiated in the Dresden Unit 3 Cycle 13 core. The reference core configuration is described in Section 4.2.

2 Xfi-f4F-85-57 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. Additional information and the results of design studies covering the development of 9x9 fuel assemblies for BWR reloads are contained in Reference 9.8.

2.0 FUEL MECHAfitCAL DESIGN ANALYSIS Applicable Fuel Design Report: Reference 9.1 The expected power history for the 9x9 fuel to be irradiated during Cycle 10 of Dresden Unit 3 is bounded by the assumed power history in the fuel mechanical design analysis. The mechanical qualification of Eilt 8x8 fuel as presented in Referente 9.4 remains valid for Cycle 10 operation of Dresden Unit 3.

3.0 THERMAL HYDRAULIC _DESIGfi AtlALYSIS 3.2 HYORAULIC CHARACTER!ZAil0N 3.2.1 Hydraulic Compatibility Component hydraulic resistances for the constituent fuel types in the Dresdon Unit 3 Cycle 10 core have been determined in single phase flow tests of full scale assemblies. The compatibility of the Efic Xfi 1 and Xfi 2 0x0 fuel types with representative G.E. Ox0 fuel has been demonstrated generically in Xfi flF 80 19(P), Volomo 4. Figure 3.1 tilustrates the 1

3 XN-NF-85-57 hydraulic demand curves for ENC and G.E. 8x8 fuel and ENC 9x9 fuel in the Dresden Unit 3 core. The XN-3 9x9 fuel performance falls between that of the ENC 8x8 fuel and that of the G.E. 8x8 fuel, indicating adequate compatibility for coresidence in the Dresden core.

3.2.3 Fuel Centerline Temperature Exposure at Minimum Margin Point 5000 MWD /MT an q.

Centerline Temperature at 15P. Power 4115 F Melting Point of Fuel 5050 F Margin to Centerline Melting 935 F 3.2.5 Bypass Flow Calculated Bypass Flow Fraction 10.07.

3.3 MCPR FUEL CLADDING INTEGRITY SAFETY LIMIT j 3.3.1 Coolant Thermodynamic Condition Rated Thermal Power 2527 MWt feedwater Flowrate (at SLMCPR) 12.6 Mlbm/hr Steam Dome Pressure (at SLMCPR) 1015 psia feedwater Temperature 345 F 3.3.2 Design Basis Radial Power Olstribution See Figure 3.2 3.3.3 Design Basis Local Power Distribution See Figures 3.3 through 3.6

+

4 XN-NF-85-57 4.0 NUCLEAR DESIGN ANALYSIS 4.1 FUEL BaNDLE NUCLEAR DESIGN ANALYSIS Assembly Average Enrichment 3.13%

Radial Enrichment Distribution ENC XN-3 9x9 Figure 4.1-ENC XN-3A 9x9 Figure 4.2 Axial Enrichment Distribution Uniform 3.35%

with 6" natural urania ends Burnable Poisons Figs. 4.1, 4.2 Non-Fueled Rods Figs. 4.1, 4.2 Neutronic Design Parameters Table 4.1 Maximum Littice k infinite 1.235 4.2 CORE NUCLEAR DESIGN ANALYSIS 4.2.1 Core Configuration Figure 4.3 Core Exposure at E0C9, MWD /MT Nominal Value 19,754 Shutdown Reactiv'ty Calculations 19,524 Core Exposure at 80C10, MWD /MT 13.026 Cori Exposure at E0C10, MWD /MT 21,665 4.2.2 Core Reactivity Characteristics 80 Cold K effective, All Rods Out 1.1095 BCC Cold K-effective, All Rods in 0.9594 B]C Cold K-effective Strongest Rod Out 0.9874

5 XN-NF-85-57 Technical Specification R-Value 0.0022 Includes 0.0004 to account for B4C settling in control rod tubes Standby Liquid Control System Reactivity, Cold Conditions, 600 ppm 0.9420 4.2.4 Core Hydrodynamic Stability Figure 4.4 Maximum Oecay Ratio Values 100% Flow Control Line 0.33 5.0 ANTICIPATED OPERATIONAL OCCURRENCES Applicable Generic Transient Analysis Report Reference 9.2 5.1 ANALYSIS OF PLANT TRANSIENTS AT RATED CONDITIONS Reference 9.3 Limiting Transient (s): Load Rejection Without Bypass (LRWB)

Feedwater Controller Failure (FWCF)

Loss of Feedwater Heating (LFWH)

Maximum Maximum Maximum Delta-Event Power Flow Heat Flux Power Pressure CPR* Model LRWB 100% 100% 109.5% 275% 1273 psia 0.28 COTRANSA FWCF 100% 100% 112.1% 185% 1191 psia 0.20 COTRANSA LFWH 100% 100% 120.0% 120% 1064 psia 0.20 PTSBWR3

  • Delta-CPR results for most limiting fuel type 1
(

q.

6 XN-NF-85-57:

5.2.' ANALYSES FOR REDUCED FLOW OPERATION Reference 9.3 Limiting Transient (s): Recirculation Flow Increase (RFIT) 5.4 ASME OVERPRESSURIZATION ANALYSIS Reference 9.3

, Limiting Event Containment Isolation Worst Single Failure Direct Scram Maximum Pressure 1341 psia Maximum Steam Dome Pressure 1316 psia 5.5 CONTROL R00 WITHDRAWAL ERROR Starting Control Rod Pattern for Analysis Figure 5.1 Rod Block Reading Distance Withdrawn Delta-CPR 105% 3.5 ft 0.07/0.07*'

106% 4.0 0.10/0.10 '

107% 4.5 0.11/0.11 i

108% 5.0 0.13/0.13 l 109% 5.5 0.14/0.14 110%** 6.0 0.15/0.15

  • Delta-CPR values for 8x8/9x9 fuel
    • Rod Block Monitor setting selected for Cycle 10 operation 5.6 FUEL MISLOADING ERROR Maximum LHGR 12.8 kW/ft Minimum MCPR 1.577

. Maximum Delta-CPR 0.19 5.7 DETERMINATION OF THERMAL MARGINS' a.

Summary of Thermal Margin Requirements

.f

7 XN-NF-85-57 Event Power Flow Delta-CPR* MCPR Limit

  • LRWB 100% 100% 0.24/0.28 1.29/1.33 LFWH 100% 100% 0.20/0.20 1.25/1.25

.FWCF 100% 100% 0.17/0.20 1.22/1.25 CRWE 100% 100% 0.15/0.15 1.20/1.20

  • Limits for 8x8 fuel /9x9 fuel MCPR Operating Limits at Rated Conditions Fuel Type MCPR Limit 9x9 Fuel 1.33 8x8 Fuel 1.29 i

MCPR Operating Limits at Off-Rated Conditions Reduced Flow MCPR Limits Manual and Automatic Flow Control Figure 5.2 Automatic Flow Control Figure 5.3 6.0 POSTULATED ACCIDENTS 6.1 LOSS-0F-COOLANT ACCIDENT 6.1.1 Break Location Spectrum Reference 9.6 6.1.2 Break Size Spectrum Reference 9.6 6.1.3 MAPLHGR Analyses.for XN-3, XN-3A 9x9 Fuel Reference 9.7 See Reference 9.5 for MAPLHGR analyses for XN-1 and XN-2 fuel. Existing MAPLHGR limits for these fuel types remain .

valid for Cycle 10 operation.

Limiting Break: Double-ended guillotine pipe break Recirculation pump suction line

' 1.0 Discharge Coefficient

8 XN-NF-85-57 Bundle Average Peak Clad Peak Local >

Temperature 4 Exposure MAPLHGR MWR 5 0 11.40 2006 F 2.20%

5,000 11.75 2045 F 2.44%

10,000 11.40 1893 F- 0.91%

15,000 10.55 1805 F 0.63%

20,000 9.70 1710 F- 0.44%

4 25,000 8.85 1623 F 0.29% ,

i 30,000 8.00 1529 F 0.18%

35,000 7.15 1421'F 0.12%

,! 40,000 6.30 1309 F 0.08%

6.2 CONTROL R00 DROP ACCIDENT Reference 8.1 ,

Dropped Control Rod Worth 0.0074 Doppler Coefficient,1/k dk/dT -10.5x(10)-6 Effective Delayed Neutron Fraction 0.0052 Four-Bundle Local Peaking Factor 1.278 l Maximum Deposited Fuel Rod Enthalpy, cal /gm 109 1

4 7.0 TECHNICAL SPECIFICATIONS 7.1 LIMITING SAFETY SYSTEM SETTINGS 7.1.1 MCPR Fuel Cladding Integrity Safety Limit 3 MCPR Safety Limit 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 XN-3 & XN-3A 9x9 Fuel f

,. ,. -. e,. -. - . , ,

+

8 L

4 9 ' XN-N' F-85-57 i

!- Bundle ~ Average Exposure MAPLHGR I O 11.40 kW/ft

5,000 11.75
10,000. 11.40 15,000 10.55 20,000 9.70 25,000 8.85 i 30,000 8.00 35,000 7.15 40,000 6.30 1

7.2.2 Minimum Critical Power Ratio Rated Conditions MCPR Limits .

Fuel Type Limit. *

9x9 Fuel 1.33 1 6x8 Fuel 1.29 l

i Off-Rated Conditions MCPR Limits Reduced Flow MCPR Limits Manual and Automatic Flow Control Figure 7.1 Autcmatic Flow Control Figure 7.2 7.3 SURVEILLANCE REQUIREMENTS 7.3.1 Scram Insertion Time Surveillance j Individual control rod drive insertion times shall be monitored in accordance with existing Technical Specification requirements. If the i average insertion time'to the 90% insertion point for all the control rod

10 -

XN-NF-85-57

! drives in .the core, based on the most recent' observation for each' drive, exceeds 2.73 seconds, the MCPR operating limit for each fuel type in' the 1- core shall be increased by an amount determined by:

OLMCPR = MCPRs + ( Tav - 2.73 )

  • 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 0.130 for 8x8 fuel and 0.143 for 9x9 fuel.

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

t 7.3.3 Procedural Controls Procedural controls on fuel rod local linear heat generation rate shall be established such that the power distribution assumptions of the mechanical design analysis remain applicable.

8.0 METHODOLOGY REFERENCES See XN-NF-80-19, Volume 4 for complete bibliography.

i

J 11 XN-NF-85-57 9.0 ADDITIONAL REFERENCES 9.1 " Generic Mechanical Design for Exxon Nuclear det Pump BWR Reload Fuel," XN-NF-85-67(P), Exxon Nuclear Company, Richland, Washington (July 1985).

9.2 " Exxon Nuclear Plant Transient Methodology for Boiling- Water-Reactors," XN-NF-79-71(P), Revision 2, Exxon Nuclear Company, Richland, Washington (November 1981).

9.3 "Dresden Unit 3 Cycle 10 Plant Transient Analysis," XN-NF-85-62, Exxon Nuclear Company, Richland, Washington (September 1985).

9.4 " Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel," XN-NF-81-21(A'. Revision 1, Exxon Nuclear Company, Richland, Washington (January 1982).

9.5 "Dresden Unit 3 LOCA Analysis Using the EXEM/BWR Evaluation Model,"

XN-NF-81-75, Exxon Nuclear Company, Richland, Washington 9.6 " Generic Jet Pump BWR3 LOCA Analysis Using the EXEM Evaluation Model,"

XN-NF-81-71(A), Exxon Nuclear Company, Richland, Washington (August 1981).

9.7 Dresden Unit 3 LOCA-ECCS Analysis MAPLHGR Results for ENC 9x9 Fuel , "

i XN-NF-85-63, Exxon Nuclear Company, Richland, Washington (September 1985).

9.8 " Demonstration of 9x9 Assemblies for BWRs," EPRI NP-1580-5, Electric Power Research Institute, Palo Alto, California (May 1984).

1 1

I n

12 XN-NF-85-57

. . m,.

eM e C C L

- 3 k* - u m

C c

e u

-n- x U

cM m e Z O I U I  :

5 -8** NaJ e c

l C j M x- x s-a

% e L

w e c

c

-m m (n~

-c cs z

2 g l O >

-E L1 v h_

>, Ie a

c 5 m r c

- t:

-m h.

c.n -

C @

L

=

e __m e.

- c

-E g

.e u.

2 3 6 6 . . ' * .-

W w , g C e g g

  • . N. . . . .

- - - e e e c 35d83AW j0 NOIIDWWj '93M0g Iigggggy :_,7.r.ei.t._.

1

' . = w,, _ ,_

l 13 XN-NF-85-57 -l I

.e z

o

=

g D F CD j

- , =

s oc I T H S (n l l -

~ "

O I

I

i 5u x s QC U l I

6f x ~

a  :

O Q- l 5 o

a e

J i

l

=

o E3=

C i

~

I U s Q o 'a I r s c I

.u - '

b I

  • Om*

~

c a w e (n -

5 o E C ,

CD l I

z i n

6 G g Ln ta O , , , , , , a o

S S S R R 2 S3I19W3SS9 JO 839 WON

14 XN-NF-85-57

LL  : L  : ML  : M  : H  : H  : M  : ML  : L  :
0.88 : 0.93 : 0.98 : 1.07 : 1.12 : 1.12 : 1.06 : 0.97
0.92 :
L  : ML  : M  : ML*  : H  : H  : E*  : H  : ML  :
0.97 : 0.99 : 1.04 : 0.85 : 1.03  : 1.03 : 0.85 : 1.12~ : 0.97 :
L  : M  : M  : H  : N  : H  : H  : ML*  : M  :
0.94 : 1.09 : 0.99  : 1.01  : 0.99  :

1.00 : 1.02 : 0.85

1.06 : .

t . . . . . . . .

1

Pt.  : 98.*  : H  : H  : H  : W  : H  : H  : H  :

1.04 : 0.92 :

1.05 : 0.99 : 1.00 : 0.00 : 1.00 : 1.03 : 1.12 :

. ML .

. M .

. H .

. H .

. W .

. H .

. H . H . H .

. 1.04 .

. 1.07 .

. 1.05 .

. 1.01. C.. Pn. .

. 1.03 .

. 0.99 .

. 1.03 .

. 1.12 .

i, . . . . . . . . .

M .

M . . w .

H .

H . p... . .

u. . .

. L . . . H . . . . . .

. 0.9C. .

. 1.10 .

. 1.09. . .

1.0.t .

i . e. .

0.9*, .

. 1.01 .

. 0..Q. .

. 1.C7 .

. p... , . p.. .

. p. . y. . p...

. L . . . M

~

.. h.

~

. 0.9:- . -

1.C'. -

. 0. *. .*

. 1. e. ~.

t . ^. .

. 1.02 . O.9: .

. 1.04 -

. C . c..' .-

LL -

L  : P'  : M  : * .

M;*  : M  : ML  : L  :

. 0.e~~s C . 3. :. . 1. O '.

. 1. '. .' - ' . . " . . 0 . 0. . . 1.n: w. . O . c. .

. C . :. -

LL  : LL  : L  : L
w. M.  : L  : L  : LL -
0.9*  : 0.59 : C.95  : 0.95  : '. . G 1.0*  : 0.9"  : 0.97 -

0.!!  :

Figure 3.3 Design Basis Local Power Distribution ENC XN-3 and XN-3A 9x9 Fuel

1' r

I i

15 XN-NF-85-57

L : ML : ML : M : M : M  : ML : ML :
0.99 : 0.99 : 0.94 : 1.06 : 1.05 : 1.07 : 0.94 : 0.98 :
E  : ML'  : M  : H  : H  : ML*
  • M : ML : -
1.01 : 0.93 : 1.04 : 1.02 : 1.01 : 0.83 : 1.04 : 0.94 :  !
:  :  :  :  :  :  :  : j

________________________________________________________ j

ML  : M  : H  : E  : H  : E : ML'  : M :
0.99 : 1.07 : 1.02 : 0.98 : 0.99 : 1.00 0.83 : 1.07 :

l

ML : M : H  : E  : W : E : E  : M  :

i  : 0.98 : 1.06 : 1.01 : 0.98 : 0.00 : 0.99 : 1.01 : 1.05 :

I

ML  : M : H  : E  : E  : H : H  : M  :
0.99 : 1.07 : 1.03 : 0.93 : 0.98 : 0.98 : 1.02 : 1.06 :

4

,  : E  : E  : M  : E  : H  : E : M  : ML  :

1  : 1.01 : 0.93 : 1.05 : 1.03 : 1.01 : 1.02 : 1.04 : 0.94 :

L : ML'  : E  : M  : M  : M : ML*
0.99 : 0.99 : 0.93 : 1.07 : 1.06 : 1.07 : 0.93 0.99 :

w _________________________________________________________

I :  :  :  :  :  :  :  :  :

D : LL : L : E  : E  : ML : E  : ML  : L :

E : 0.95 : 0.99 : 1.01 : 0.99 : 0.98 : 0.99 : 1.01 : 0.99 :

WI DE i

l Figure 3.4 Desion Basis Local Power Distribution

, for ENC XN-2 Sx2 Fuel I t

l l

,- - ~ , , . - , . -

16 XN-NF-85-57 L- ML ML 2 'M- M- *- ML ML

- --L.99 : V.87 ::, D.95 : 1.C5 : 1 04 : 1.05 : D.*5  : 0.98
*L  : ML* : M
  • H  : H  : 'L. : M  : ML :
- - - - .ao :  : o4  : 1 03 : 1.C3  : 1.03  : 0.28  : 1 04  : C.95 :
'ML : M  : H  : H  : H
  • H  : ML* : M- :
r.97 : 1.05
1 -

04  : 1.01  : 1.01  : 1.02  : 0.A8 : 1.C5 :

ML  : ' P.
  • H  : H  : W  : H  : N  : "  :
C.97 7 1.04 :. 1.C3
1.01 .. . . . .
C.00 : 1.01 : 1 03 : 1.C4 "L *  : H  : H  : H  : H  : H  : P  :
,.97  : 1 . . 4 : 1. *t 3 : 1. * : 1.t1 : 1.31 : 1.73 : 1.05 :

I vt  : ML  : > H  : a  : H  : M  : PL  :

. .. . o. A . . a. 4

. .t. s

. 1 . .- ,. .

. ....i

. 1. a.. 4 1.r3 .. e. . a. t.

L  : ML. : "L  : M  : "  : M  : PL. : PL  :
.*! : 0.95 : J.94 : 1.:s : 1.*4 : 1.05 : C.94 : 0.97 :
LL  : L  : ML  : PL  : "L  : "L  : ML  : L  :

. .=.=

. ,.eu- 3.96. .

. r. . a 3 . . 2. 7 n.37 .

. o.39 .- C . = =.

I I

i Figure 3.5 Design Basis '.ocal Power Distribution for ENC XN-1 Sx3 Fuel l l

1 I

4

17 XN.NF-85-57

1.  : ML  : MH  : MH  : MH  : MH  : M  : ML  :

l

0.97
0. 95 : 1.T  : 1. 01  : 1.00  : 1.01  : 0. 96  : 0. 95 :

l l .........................................................................

I  : E  : E  : **  : H  : MM*  : H  : MH  : M  :

0.96  : 1.02 : 0.99 : 1.06  : 0.96  : 1.05  : 0.99 : 0.96 :
K.  : Mri  : H  : H  : H  : H  : H  : Md  :
0.95 : 1.01 : 1.07  : 1.05  : 1.03 : 1.03 : 1.05 : 1.01  :
ML  : MH  : H  : W  : H  : H  : **  : MH  :
0.95  : 1.01  : 1.07  : 0.00  : 1.05  : 1.03  : 0. 96  : 1.00 :
M.  : **  : *  : Ki -

W  : H  : H  : K4  :

0.95  : 1.01  : 0.99  : 0. 99  : 0.0C  : 1.05  : 1. 06 : 1. 01  :
M. . M-  : M-  : K- . F  : H  : ** : Ki  :
0. 96  : 1.02 : 1.00  : 0.93 : 1.C'  : 1.07  : 0.99 : 1.02 :

......................................................................... 1 i

L  : M. M.  : e M-  : MH  : Me  : M;  :
0. 9E  : 0.94 . 1. 02 . 1. C-. . 1.C.  : 1. 01  : 1. 02  : 0.95  :

W .........................................................................

4 . . . . . l D -

L '.  : L  : M. -

K. v.  : K.  : K.  : L .

l E

C o.

a

c. % .
e. 9t. .

. C . :. :. r. . :. :.

0. c. :, .

. 0. 9:. 0.5' -

h ,. O -L Figure 3.6 Design Basis L G.E. cXc n

. r uel:ypti .c:a1 Ocwer Distribution for

18 XN-NF-85-57 r . . . . . . .

LL  : L  : ML  : M  : H  : H : M  : ML  : L  :

i ...__... __.................._______.......___._______.....________......

i . . . .

L  : ML  : M  : ML*  : H  : H : ML*  : H  : ML  :
L  : 'M  : M  : H  : H  : H : H  : ML*  : M  :
ML*  :  : H  : H  : W : H  : H  : H  :
ML H
:  : H  : W  : H  : H  : H  : H-  :
ML M H 4
L  : M  : M  : H  : H  : H  : H  : ML*  : M  :

,  : L  : ML*  : ML  : M  : H  : H  : M  : M  : ML  :

LL  : L  : ML*  : M  : M  : ML*  : M  : ML  : L  :

w _______......_...........................................................

D  : LL  : LL  : L  : L  : ML  : ML  : L  : L  : LL  :

4e WIDE

i j LL RODS ( 5) --- 1.50 W/0 U235 1 L RODS (12) ---

2.20 W/0 U235 j' ML RODS (11) ---

2.84 W/0 U235 M RODS (16) --- 3.72 W/0 U235 i

H RODS (27) --- 4.34 W/0 U235 ML* RODS ( 8) ---

2.84 W/0 U235 + 4.00 W/0 GG203 W RODS ( 2) ---

INERT WATER R00 1

FIGURE 4.1 ENRICHMENT DISTRIBUTION FOR 903.35- 8G4.0 (XN-3) 1 J

I i

a r l

19 XN-NF-85-57

.~

LL  : L  : ML  : M  : H  : H  : N  : ML  : L  :

W 4

-  : L  : ML*  : M  : ML*  : H  : H  : ML*  : H : ML  :

a. . . . . .

r .................___...... .... ___..................___........____.....

L  : M  : M  : H  : H  : H  : H  : ML*  : M  :

4 . . . .

ML  : ML*  : H  : H  : H  : W  : H  : H : H  :

.i . . . . . .

ML  : M  : H  : H  : W  : H-  : H  : H : H  :

l . . . . . .

L  : M  : M  : H  : H  : H  : H  : ML*  : M  :

j  : L  : ML  : ML*  : M  : H  : H  : M  : M  : ML  :

4 . . . .

l  : LL  : L  : ML  : M  : M  : ML*  : M  : ML*  : -L  :

i. . . . . . . . .

w ___..........__......___...................... __...__ ..................

. I .

D  : LL  : LL  : L  : L  : ML  : ML  : L  : L  : LL  :

4 . . . . . . . .

E . . . . . . . . .

1 1 WIDE i

j LL RODS ( 5) ---

1.50 W/0 U235 L RODS (12) ---

2.20 W/0 U235 ML R0DS (10) ---

2.84 W/0 U235 i M ROOS (16) ---

3.72 W/0 U235 l H RODS (27) --- 4.34 W/0 U235 i ML* RODS ( 9) ---

2.84 W/0 U235 + 4.00 W/0 GD203 i W RODS ( 2) ---

INERT WATER R0D I

i

,1 FIGURE 4.2 ENRICHMENT DISTRIBUTION FOR 903.35- 9G4.0 (XN-3A) t ,

4

, . - ,. -,,_c~

_ - ~ , , ___ _ , , . , , _ _ - _ . , - , _ , . , , . , - . . . , . . _-. , , , . , , . _ . -

20 XN-NF-85-57 B2 B2 DO B2 B2 B2 DO B2 B2 B2 DO B2 82 C1 A3 B2 DO C1 DO B2 DO C1 DO B2 DO C1 00 C1 C1 A3 DO Cl B2 C1 00 C1 B2 C1 DO B2 00 B2' C1 A3 A3 B2 D0 C1 B2 B2 DO C1 B2 B2 DO C1 DO C1 A3 A3 B2 B2 DO B2 B2 B2 00 B2 B2 B2 DO C1 C1 A3 A3 B2 00 C1 00 B2 00 C1 DO B2 DO B2 C1 A3 A3

, DO C1 B2 C1 DO C1 B2 C1 DO C1 00 C1 A3 B2 00 C1 B2 B2 00 C1 B2 B2 DO C1 A3 A3

) B2 B2 DO B2 B2 B2 DO B2 B2 C1 C1 A3 A3 B2 DO B2 00 B2 DO C1 D0 C1 B2 A3 A3 DO C1 00 C1 00 B2 00 C1 C1 A3 A3 B2 30 B2 DO C1 C1 C1 A3 A3 A3 E2 C1 C1 C1 C1 A3 A3 A3 A2 1

C1 :1 A3 A3 A3 A3 A3 A3 A3 A3 A3

] ,

i: Fuel Ty:e y = Cy:ies Irra:icte:

fus' N;-ter of Ty: e Asse-tiies Descri;;ien A 140 GE B ER 2.65 w/ L-235

! E 224 XN1 5 C 2.69 m/o U-235 XN? S S 2.83 w/o U-235 C 184 D 176 Xh3 9 9 3.13 w/o 0-235 figure 4. 3 Dresden U.it 3 Cycle 10 Reference loading Pattern by Fuel Type (One Quarter of Smtrical Core Loading)  ;

i

l i

.~

21 XN-NF-85-57 1.0 _ ___ _._ __ ___ _ .

0.8 -

e R~ 0.6 .

M d Natural e Circulation a:

E 0.4 .

v w

c 100% Rod Line 0.2 -

0.0 ' ' ' '

0 2: 40 6: 80 100 PERCENT PC G Ficre 4.4 Decay Ratic vs. Reactor Power for Dresder 3 Cy:le 10 i

l 1

22 XN-NF-85-57

, Table 4.1 Neutronic Design Values Fuel Pellet Reference 9.1 Fuel Rod Reference 9.1 Fuel Assembly ~ Reference 9.11 l Core Data Number of fuel assemblies 724 Rated thermal power, MW 2527 Rated core flow, M1bm/hr 98.0 Core inlet subcooling, BTU /lbm 24.6 Moderator temperature, F 546 Channel thickness, inch 0.080 Fuel assembly pitch, inch 6.0 Wide water gap thickness, inch 0.750 Narrow water gap thickness, inch 0.374 Control Rod Data Absorber material B4C, Hf Total blade span, inch 9.750 Total blade support span, inch 1.562 Blade thickness, inch 0.3120 Blade face-to-face internal dimension, inch 0.200 Absorber rods per blade 84 s Absorber rod outside diameter, inch 0.188 Absorber rod inside diameter, inch 0.138 Absorber density, % of theoretical 70-4 I

23 XN-NF-85-57 cc 40 40 40  :

cc

'. 0 10 .

51 -

32 10

, 6 10 32 -,

47 32 20l 32 20 di "--

32 10 0 20 0 10 32 ..,

39 40 32 32 40 35 10 0 18 14 18 0 10 31 40 32 32 40 40 32 32 40 27 10 0 18 14 18 0 10 ,

23 40 l32 32 40 19 I

32 10 0 20 0* 10 32  :,

15 20 32 32 , 20 i l '. '-- '

32 10 6 10 32 -*

07 i- 10 10 J i

C3 '-  :

40 40 40 -

C2 C5 10 14 13 22 25 30 34 35 42 45 5: 54 53 i

Note:

  • Control rod being withdrawn, rod positions in notches, full in = 0, full out = blani or 48 Figure 5.1 5 tarting Control Rod Pattern for Control Rod Withdrawal Analysis J

i

  • e

.m, - , , , , ---- - , - - - , - , - , , - -

, - - - ,-----,-,-,,,-,7 - - - - s ---- -,.--~ ~ ---,,--e

2 E k a Y' O i

0

0 1

s

) n

0 R i o

9 H t

/ i B d L n M o C

8 9 l

0 ,

l 8 A D

E -

T A t R i

% i m

( L e0 7

W R O P L C F M G w N o I l

0 T F 6 A L d U e C c R u I

d C e E R 0  : 0 R x 5 8 E R 2 O

C 5 L er A

0 T u O g 4 T i F

9 x

9 0

3

, . 0 6, 5, .

3, 1, 2

4 2. 0 1 1 1 1 1 1 1

,E3 Ep&5 EM

1.8 . -

17r --

9x9 t 1.6 .,

E a

E 8x8 1.5 . . $

=

o E

E 1.4 . .

1.3 5

1.2 - - - - *

  • k sa 70 3D 40 SD 60 70 80 90 1DO p TOTAL CORE RECIRCULATING FLOW (% RATED, 98 MLB/HR)

Figure 5.3 Reduced Flow MCPR Limit - Automatice Flow Control

- - _- _- . _ . ~. . - _ _ . _ _ _ . - - - . - . - _ - . _ . . _ _ - . -- - . . . . - - _ . . - _ - _ - ..

s l

1.6 .

4 8x8

1. 5, ,

9x9 r

i U l.4 ,

a 1

25 t2 85 1. 3, E$

E M

1. 2, .

? -

) 1.1 .

E.

1.0  ;  ;  :  :  :  :  :  : y

.i 20 30 40 50 60 70 80 90 100 .

, m

.i m

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

Figure 7.1 Reduced Flow MCPR Tech. Spec. Limit - All Conditions

l.8 - -

177 9x9 t- 1.6 ..

Ea

$ 8x8

$2 1.5 ..

O 85 E

E 1.4 . .

I3 5

$i 1.2 - - - - - - - -

En 20 3D 40 SD 60 70 80 90 100 T 10 TOTAL CORE RECIRCULATING FLOW (% RATED, 98 MLB/HR)

Figure 7.2 Reduced Flow MCPR Tech. Spec. Limit -

Automatic Flow Control

i 28 XN NF-85-57 i

l I APPENDIX A ,

I SINGLE LOOP OPERATION r

i d

$ The NSSS supplier has provided analyses which demonstrate the safety of -

! plant operation with a single recirculation loop out of service for an ,

1

^

extended period of time. These analyses restrict the overall operation of the plant to lower bundle power levels and lower nodal power levels than are I allowed when both recirculation systems are in operation. The physical 4

interdependence between core power and recirculation flow rate inherently L

i limits the core to less than rated power. Because the ENC fuel was designed 4

I to be compatible with the coresident fuel in thermal hydraulic, nuclear, and

! mechanical design performance, and because the ENC methodology has given results which are consistent with those of the previous analyses for normal j two-loop operation, the analyses performed by the NSSS supplier for single 1

I loop operation are also applicable to single loop operation with fuel and

} analyses provided by ENC.

i j

i i With a-single recirculation loop in operation, the NSSS supplier's analyses support continued operation with an increase of 0.01 in the MCPR safety

! limit. Because of the similarity between the ENC XN-3 9x9 fuel type and the -

i other fuel types making up the remainder of the core, and because of the  ;

i j similarity in the magnitude of the uncertainties which determine the MCPR safety limit, this small increase in the safety limit value is also

)

f I

i .

29 XN-NF-85-57 appropriate for operation with ENC fuel and analyses. For Cycle 10 4

operation with both recirculation loops in operation, the MCPR safety limit is 1.05, which is the same value as was used for the previous cycle. For Cycle 10 operation with a single recirculation loop in service, the MCPR safety limit is 1.06, which is also the same value as was used for the previous cycle.

The consequences of core-wide transients at the reduced power and flow conditions necessitated by single loop operation are bounded by the consequences of these events at rated conditions. The additional conservatism imposed by the reduced flow MCPR operating limits specified in the main body of this report assures that the MCPR safety limit will not be violated during anticipated operational occurrences with a single recirculation loop in service. No modification to the delta CPR defining

' the rated conditions MCPR operating limit is required, and the reduced flow MCPR limit curve remains conservatively applicable during single loop operation. Because the reduced flow MCPR limit curves are based on equipment performance which physically cannot happen during single loop operation, the added conservatism present in the curves compensates for the penalties associated with increased uncertainties in the MCPR safety limit and control rod drive performance. The reduced flow MCPR limit curves are applicable without modification during single loop operation.

The stability characteristics of the Cycle 10 core are equivalent to or better than those of the previous cycle core. Reactor operation within the limitations which assured adequate stability for the previous cycle will

30 XN-NF-85-57 continue to assure adequate stability for Cycle 10. In addition, the stability analyses reported in Section 4.2.4 of the main body of this report cover the operating region of Single loop Operation. The calculated decay ratio is 0.53, which is within the appropriate acceptance criterion for Single Loop Operation.

The NSSS supplier justified MAPLHGR reduction factors for resident fuel types 8x8, 8x8R, and P8x8R during single loop operation. Because the ENC 8x8 fuel is very similar to the P8x8R fuel in design and operational characteristics, the MAPLHGR reduction facters defined by P8x8R fuel are applicable to the ENC 8x8 fuel for single loop operation. The LOCA analyses performed in support of the Cycles 8 and 10 reloads demonstrate a similarity in ECCS performance for cores fueled with 8x8 and 9x9 fuel, with the 9x9 fuel exhibiting a consistently lower value of Peak Cladding Temperature (PCT). Because thc LOCA performance of the various fuel types is similar and because the 9x9 fuel exhibits a consistently lower PCT, the MAPLHGR reduction factors for 8x8 fuel are conservatively applicable to 9x9 fuel.

l

[

l l

l l

31 XN-NF-85-57 APPENDIX B FORMULATION OF MCPR LIMITS The original Exxoa Nuclear analyses perform.ed in support of operation of .

Dresden Unit 3 as reported in XN-NF-81-76 defined thermal margin requirements in terms of a minimum critical power ratio (MCPR) as formulated in XN-NF-524, Revision 0. This MCPR formulation contained a flow iteration during the calculation of the critical heat flux associated with any given fuel operating state. During the NRC Staff review of XN-NF-524 and the Dresden license amendment application, it was determined that differences between the proposed ENC formulation of MCPR and the existing formulation in the Dresden design basis were unnecessarily confusing. Cycle 8 and 9 operation of Dresden Unit 3 and Cycle 9 and 10 operation of Dresden Unit 2 were authorized on the basis of this flow-iterative formulation of MCPR, but the topical report, XN-NF-524, was revised (Revision 1) to incorporate a constant flow MCPR formulation consistent with the previous design basis.

Revision 1 of XN-NF-524 was approved generically for BWR applications, and Dresden continued to operate with the flow-iterative MCPR under the provisions of 10 CFR 50.59(a)(1).

The introduction of the ENC XN-3 9x9 fuel necessitated a revision to the .

plant Technical Specifications in order to specify operating limits for the new fuel type. At this time, new, constant flow MCPR operating limits are

~,_

$4V 32 XN-NF-85-57

- i

, /

incore fuel types consistent with the approved. ENC provided for all methodology in XN-NF-524(A), Revision 1. All MCPR limits reported in this document are quoted on a constant flow basis.

+

i -

l /

l l '

l f

I l

I -

D l /.

2

33 XN-NF-85-57 APPENDIX C RECIRCULATION PIPE REPLACEMENT PROGRAM Following the completion of Cycle 9 operation, the Dresden Unit 3 plant was altered to replace ' cracked stainless steel piping in the recirculation systems. Although the original plant configuration was followed ' closely, there were some differences in dimensions between the original piping and the replacement piping.

Exxon Nuclear has evaluated the geometric differences between the original and replacement piping. The contained volumes and piping spool-configurations are sufficiently similar that the results.of earlier analyses performed by Exxon Nuclear based on the original plant configuration 'are l

L valid within the reporting accuracy of the design basis documents.

l l

l l

i L

XN-NF-85-57 Issue Date: 9/12/85 t

i Distribution G.J. Busselman J.C. Chandler R.E. Collingham T.P. Currie S.E. Jensen s

T.H. Keheley T.L. Krysinski J.L. Maryott J.N. Morgan T.W. Patten G.L. Ritter J.M. Ross G.A. Sofer D.R. Swope R.I. Wescott H.E. Williamson CECO /JM Ross (60)

Document Control (5) sh

- .-

Retrieved from "https://kanterella.com/w/index.php?title=ML20153G305&oldid=2958178"