ML021540365

Stations Meeting Handouts to Discuss Topical Report DPC-NE-1005P, Nuclear Design Methodology Using CASMO-4/SIMULATE-3 MOX, Submittal by Duke

ML021540365
Person / Time
Site:	Mcguire, Catawba, McGuire
Issue date:	05/23/2002
From:	Chandu Patel NRC/NRR/DLPM
To:
References
DPC-NE-1005P
Download: ML021540365 (33)

Text

NRC FORM 658 U.S. NUCLEAR REGULATORY COMMISSION (9-1999)

TRANSMITTAL OF MEETING HANDOUT MATERIALS FOR IMMEDIATE PLACEMENT IN THE PUBLIC DOMAIN This form is to be filled out (typed or hand-printed) by the person who announced the meeting (i.e., the person who issued the meeting notice). The completed form, and the attached copy of meeting handout materials, will be sent to the Document Control Desk on the same day of the meeting; under no circumstances will this be done later than the working day after the meeting.

Do not include proprietary materials.

DATE OF MEETING 2 02.

The attached document(s), which was/were handed out in this meeting, is/are to be placed 5-123 /

in the public domain as soon as possible. The minutes of the meeting will be issued in the near future. Following are administrative details regarding this meeting:

Docket Number(s) 5D-4-31.

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Distribution of this form and attachments:

Docket File/Central File PUBLIC NRC FORM 658 (9-1999)

PRINTED ON RECYCLED PAPER This form was designed using InForms

Duke Power - Nuclear Regulatory Commission Meeting Topical Report DPC-NE-1005P Nuclear Design Methodology Using CASMO-4/SIMULATE-3 MOX Rockville, MD May 23, 2002

Agenda Duke - NRC Meeting Nuclear Design Methodology Using CASTMO-4/SIMULATE-3 MOX Thursday, May 23, 2002

Introductions

Mixed Oxide (MOX) Fuel Project Status and Plans Duke Nuclear Analysis Methodologies CMS - Core Management System (Analytical Models)

Lunch Qualification of Nuclear Analysis Methodologies Power Reactor Benchmark Analyses Fuel Pin Power Distribution Benchmark Analyses Statistically Combined Power Distribution Uncertainty Factors Dynamic Rod Worth Measurement NRC Questions Adjourn Discussion Lead Duke (Nesbit)

Duke (Nesbit)

Studsvik (Smith)

All Duke (Eller)

Duke (Eller)

Duke (Naugle)

Duke (Eller)

Duke (Thomas)

NRC

2 Mixed Oxide (MOX) Fuel Project Status and Plans Meeting with Nuclear Regulatory Commission

- DPC-NE-1005P Review S. P. Nesbit Duke Power hDuke May 23, 2002 rdPower.

Plutonium Disposition Program

• Goal: To dispose of surplus weapons plutonium January 2000 Department of Energy (DOE) Record of Decision September 2000 U.S.-Russian Federation Plutonium Disposition Agreement

• Initial Approaches Fabrication into mixed oxide (MOX) fuel and use in existing light water reactors Immobilization in vitrified high-level radioactive waste 1PONWr.

A Dk EpC-,j

MOX Fuel Project

"* MOX Fuel Fabrication 12/00: DCS submitted Environmental Report to NRC 2/01: DCS submitted Construction Authorization Request to NRC

"* MOX Fuel Qualification

"* MOX Fuel Irradiation

"° Fresh MOX Fuel Transportation and Packaging

"° Project Management rdPower.

Plutonium Disposition Program Changes Recently announced changes from the Department of Energy Termination of immobilization portion of program Design changes to the MOX Fuel Fabrication Facility to accommodate a wider variation of feed material One year delay in the provision of batch quantities of MOX fuel from the MOX Fuel Fabrication Facility A.4,,,C,

4

MOX Fuel Qualification and Irradiation

"* Maximize use of European experience base Research programs Established manufacturing process Reactor irradiation experience

"* Proven fuel assembly design

"* Confirmatory lead assembly program

"* NRC reactor operating license amendments in accordance with 10 CFR 50.90 Pk~uk6

MOX Fuel-Related Submittals

"* July 2000: DCS Fuel Qualification Plan provided to NRC for information

• August 2000: Framatome COPERNIC Topical Report (MOX applications)

• April 2001: DCS MOX Fuel Qualification Plan revised and provided to NRC for information

• August 2001: Duke Power Nuclear Analysis Topical Report (MOX and LEU applications)

PkDuke dPower.

A Dýý Cý7 MOX Fuel-Related Submittals (cont.)

"* September 2001: Duke Power Thermal-Hydraulic Statistical Core Design Topical Report, Appendix E (advanced Mk-BW fuel assembly design, to be used for MOX fuel)

"* April 2002: Framatome Advanced Mark-BW Fuel Assembly Design Topical Report

"* April 2002: Framatome MOX Fuel Design Topical Report P~h Duep.8

MOX Fuel Lead Assembly Program

"* Original approach - fabricate two MOX fuel lead assemblies at Los Alamos National Laboratory (LANL) and begin use in McGuire Nuclear Station in fall 2003

"* LANL fabrication activities terminated May 2000

"* Alternatives under consideration Fabrication at existing European MOX fuel fabrication facilities Start irradiation -2004 Fabrication at Savannah River MOX Fuel Fabrication Facility, when constructed and licensed Start irradiation -2008 Pk Duk MOX Fuel Qualification and Irradiation Plans

• 2002?: Submit MOX Fuel Lead Assembly License Amendment Request (Duke Power)

• 2003?: Submit Updated Fuel Qualification Plan (DCS)

• 2003: Submit MOX Fuel Safety Analysis Topical Report (Duke Power)

Duke oPower.

A Dý*!

10

MOX Fuel Qualification and Irradiation Plans (cont.)

0 December 2003: Submit License Amendment Requests for Batch Utilization of MOX Fuel at McGuire and Catawba (Duke Power)

• 2004: Submit MOX Fuel LOCA Topical Report (Framatome)

• 2004?: Begin MOX fuel lead assembly irradiation

• YOWer.

ADý&.C..",11

2 Duke Nuclear Analysis Methodologies Meeting with Nuclear Regulatory Commission

- DPC-NE-1005P Review S. P. Nesbit Duke Power Pkiike May 23, 2002 Pofwer.

Duke Power Fuel Management

"* Purchasing uranium, conversion, enrichment, and fabrication

"* Core design and analysis

"* Fuel mechanical design and analysis

"* Fuel thermal-hydraulic analysis

"* Safety analysis

"* Criticality analysis

"* Spent fuel management a*Por.

4 Duke Power Reload Analyses

• Full-scope reload analyses (except loss of coolant accident analyses) for the seven Oconee, McGuire, and Catawba units

• 1982: Startup of first Oconee core with Duke loading pattern and safety analysis

• 1991: Startup of first McGuire/Catawba core with Duke loading pattern and safety analysis Duke 0,Power.

A N6'&

C.,173 Selected Duke Topical Reports

• 1981: NFS-1001 approved Oconee Steady-state nuclear analyses EPRI-CELL, PDQ07, and EPRI-NODE 0 1985: DPC-NF-2010 approved McGuire/Catawba Steady-state nuclear analyses EPRI-CELL, CASMO-2, PDQ07, and EPRI-NODE a Pwer.

Selected Duke Topical Reports (cont.)

• 1992: DPC-NE-1004 approved Oconee, McGuire, and Catawba Steady-state nuclear analyses CASMO-3 and SIMULATE-3P

• 2000: DPC-NE-2012 approved McGuire/Catawba Dynamic Rod Worth Measurement applications CASMO-3, SIMULATE-3P, S3K

ýaf Duke tPower.

5 Impetus for DPC-NE-1005P

"* Implementation of CASMO4 lattice code Improved methodology General benefits for all fuel types Consistency with Oconee (topical report submittal planned for 2002)

Methods transition planned for late 2002 Analyses supporting 2004 reloads at McGuire and Catawba

"* Demonstration of MOX fuel analysis capability Lead assembly cores (2004?)

Batch cores (2008?)

PkDU16

DPC-NE-1005P - Overall Approach

"* Compare CASMO4 and SIMULATE-3 MOX calculations to applicable plant and experimental data Power reactor benchmarks Critical experiment benchmarks

"* Quantify uncertainty factors for MOX and LEU fuel applications at McGuire and Catawba

"* Same fundamental approach as used in previously approved Duke nuclear analysis topical reports Overview of Presentations

"* Analytical Models (Topical Report Section 2)

"* Nuclear Analysis Methodology Qualification

"* Power Reactor Benchmark Analyses (TR Section 3)

"* Fuel Pin Power Distribution Benchmark Analyses (TR Section 4)

"* Statistically Combined Power Distribution Uncertainty Factors (TR Section 5)

"* Dynamic Rod Worth Measurement (TR Section 6)

~PkOWW 6,Powe8

1 CMS - Core Management System Studsvik Scandpower, Inc.

Kord S. Smith Vice-President of Technical Development Studsvik Scandpower, Inc.

504 Shoup Ave., Suite 201 Idaho Falls, ID 83402 (208) 522-1060 kord@west.soa.com Studsvik Scandpower Organization Products: Computer codes for nuclear power plant in-core fuel management.

Staff:

More than 40 nuclear engineers, with offices worldwide.

Studsvik Scandpower AB Sweden Studsvik Scandpower Studsvik Scandpower USA Norway Studsvik Scandpower Studsvik Scandpower Germany Switzerland Studsvik Japan Japan Studsvik-Scandpower May 23, 2002 2

Studsvik Scandpower CMS CASMO-4 Lattice Physics Code CMS-LINK Linking Code XIMAGE SIMULATE-3 GARDEI Fuel management Steady-State Analysis On-line Monito SIMULATE-3R SIMULATE-3K Real Time Simulator Transient Analysis Studsvik"Scandpower May 23, 2002 3

CMS Has Wide Range of Applications

• 11 Countries

• 55 Companies

- Nuclear Utilities

- Nuclear Fuel Vendors

- Regulatory Agencies

- National Laboratories

- Universities

• Applied to more than 70 BWRs and 90 PWRs

• Applied to more than 2000 BWR and PWR Cycles Studsvik-Scandpower May 23, 2002 4

2 ring

3 CASMO-4 Lattice Physics Code

Purpose:

Analyze the detailed behavior of a fuel bundle over its lifetime

- Treat fuel, burnable absorbers, control rods, instruments, water gaps

- Provide bundle data for downstream core analysis codes Neutronic Data Library:

Basic data library is NJOY-generated (70 groups from 0 -10 MeV) using mostly ENDF/B-IV, with some JEF-1, and JEF-2.1 data.

- Numerous temperatures and background cross sections used to treat resonance self-shielding (Doppler broadening)

- Contains more than 100 materials commonly used in LWRs Studsvik-Scandpower May 23, 2002 5

4 CASMO-4 Overcomes Limitations Inherent In Current-Coupling-Collision-Probability Methods Studsvik"Scandpower Method of Characteristics (MOC)

The Characteristics Method The equation being solved is the,olison to the characteristics form of the Boltzoan transtport equation.

ot0

.".

• nexp{-Zs) 1-exp(-Zasl Much more angular detail:

Echphy "l

region todivided into multiple fiat source regions.

"* Azimuthal: 32-128 angles

"* Polar: 2-5 angles

"* Ray spacing: 0.1 cm Parallel tracks (noteroous zimuthal/polar angles) are stperimlosedon the global problem.n Intersections of source non torface and tracks dne points for each of the unkoown angular fluxnes.

Ougongangul~arttx along anyray are known from incoming angular

'e, fluxescrois sections, e

sourcesand track lengths.

/

Studsvik'Scandpower 7

5 CASMO-4 Benchmarking of k-eff vs. Criticals L-library (ENDF/B-[V)

B&W (LEU PWR, cold) 1.00050 KRITZ-3 (LEU PWR, cold and hot) 0.99699 KRITZ-4 (LEU BWR, cold and hot) 0.99900 KRITZ-3 (MOX, cold and hot) 0.99912 VIP (MOX, cold) 0.99973 ENDF/B-VI library B&W (LEU PWR, cold) 1.00301 KRITZ-3 (LEU PWR, cold and hot) 0.99701 KRITZ-4 (LEU BWR, cold and hot) 0.99990 KRITZ-3 (MOX, cold and hot) 0.99803 VIP (MOX, cold) 0.99530 JEF-2.2 Library B&W (LEU PWR, cold) 1.00227 KRITZ-3 (LEU PWR, cold and hot) 0.99779 KRITZ-4 (LEU BWR, cold and hot) 0.99945 KRITZ-3 (MOX, cold and hot) 0.99785 VIP (MOX, cold) 0.99863 Studsvik-Scandpower May 23, 2002 9

CASMO-4 Library Status

• L-library is currently used by all customers for production analysis L-library has little bias in reactivity between LEU and MOX ENDF/B-VI Library has -400 pcm bias between LEU and MOX JEF-2.2 library has about -150 pcm bias between LEU and MOX Pin power distribution accuracy is insensitive to library At present, Studsvik Scandpower recommends customers to continue to use the L-library When new libraries are fully tested and ready, JEF-2.2 will probably move into production sooner than ENDF/B-VI (at Studsvik Scandpower)

Studsvik'Scandpower May 23, 2002 10

6 CASMO-4 Case Matrix For Each Unique Lattice

" Depletion Cases (History Cases):

Several coolant temperatures and/or void Several boron concentrations Several fuel temperatures

" Branch Cases:

Many coolant temperatures Many boron concentrations Many fuel temperatures All control rod types

" Reflector Cases:

Radial baffle/reflector Top and bottom axial reflectors Studsvik-Scandpower May 23, 2002 11 CASMO-4 Data Produced

"* Macroscopic cross sections (two groups)

"* Microscopic cross sections and yields for fission products (Xe, Sm)

"* Discontinuity factors (treats bundle heterogeneities)

"* In-core detector constants

"* Pin-by-pin power distributions (two groups)

"* Bundle-averaged isotopics vs. depletion

" Kinetics data:

- delayed neutron yields and decay constants in 6 groups Neutron velocities Studsvik-Scandpower May 23, 2002 12

7 CMS-LINK Linking Code Reads all relevant CASMO-4 output data for each fuel type Collects fuel descriptors and geometrical data Computes 1 -D, 2-D, and 3-D tabular data tables for each variable Cross sections Detector data Pin-by-pin data Etc.

Creates separate tables for history and instantaneous parameters Coolant density or void Fuel temperature Control rod Etc.

Spline fits data to guarantee accuracy of downstream linear interpolation Creates binary data library for SIMULATE-3 and StMULATE-3K Studsvik'Scandpower May 23, 2002 13 SIMULATE-3 MOX

"* SIMULATE-3 core simulator first introduced in 1985

"* Used for steady-state core analysis: reload core design, safety parameter generation, RPS limit generation, and operational plant support

"* Full two-group advanced nodal code:

I or 4 nodes per assembly Explicit reflectors (no albedos)

Explicit tracking of I, Xe, Pm, Sm Discontinuity factors to treat bundle heterogeneities Quartic polynomial spatial representation of intra-nodal flux distributions Quadratic transverse leakage treatment Quadratic intra-nodal bumup gradient modeling Spectral history treatment of bundle interface spectrum interactions Pin power reconstruction Studsvik-Scandpower May 23, 2002 14

8 SIMULATE-3 MOX Enhancements

"* First MOX analysis applications in 1989

"* Goal: achieve MOX/LEU mixed-core accuracy comparable to that of LEU cores

"* Continuous development throughout the 1990's

"* Currently used in Germany, Switzerland, UK, Japan, and the U.S.

" Enhancements relative to original SIMULATE-3:

Analytic (sinh, cosh) intra-nodal thermal fluxes replaced quartic polynomials Corner point flux interpolation model improved Spatial re-homogenization of cross sections to treat global flux gradients Two-group pin power form functions replaced total power form functions Instantaneous spectral effects on 2-group cross sections modeled at interfaces P-3 transport effects modeled at bundle interfaces Studsvik"Scandpower May 23, 2002 15 Verification/Validation of SIMULATE-3 MOX vs. CASMO-4

• Direct 11/4-core calculations with CASMO-4

• Same detail as lattice physics computation

• Explicit isotopic depletion for all nuclides

• Reference cases for SIMULATE-3 MOX - permit testing of all SIMULATE-3 MOX modeling approximations:

"* Verify/Improve Nodal Approximations

"* Investigate errors in Pin Power Predictions

"* Investigate Complicated Depletion Effects

"* Investigate Detector Modeling Approximations

"* Study 2-D Baffle/Reflector Effects

"* Perform Moveable/Detector Analysis Studsvik-Scandpower May 23, 2002 16

04o Cc, 04 C0 co M

El 1

Ell i R R

10 SIMULATE-3K MOX Features

"* Transient version of SIMULATE-3 MOX

"* Permits time-dependent boundary conditions for:

Boron concentration Core inlet coolant temperature/flow Control rods positions System pressure

" Features:

Spatial neutronics model is identical to steady-state SIMULATE-3 MOX All neutronic data taken from standard CMS-LINK library Fully-implicit temporal differencing of frequency-transformed diffusion equation Analytic solution of delayed neutron precursor equations (6 groups)

Spontaneous fission/alpha-n neutron sources modeled User-specified or automatic time-step selection Studsvik-Scandpower May 23, 2002 19 SIMULATE-3K/SIMULATE-3 Differences

"* Fuel temperatures are computed using an explicit fuel pin conduction model

"* Coolant densities are computed using an explicit channel hydraulic model

"* At HZP, pin conduction and channel hydraulic differences have zero effect on computations

"* All MOX enhancements are identical in S3 and S3K Studsvik-Scandpower May 23, 2002 20

SIMULATE-3K Applications 11

" PWRs:

- Reactivity excursions (ejected bank/rod)

- Dynamic rod worth measurements (DRWM)

- Dropped rod transients

- Steam line break analysis

- Boron dilution accidents

" BWRs:

- Stability analysis

- SCRAM reactivity curve generation

- Reactivity excursions (dropped/stuck rod)

- Operational transients (e.g., pump trips, turbine trips)

Studsvik'Scandpower May 23, 2002 21

CMS MOX Usage

"* BNFL:

Reload core design for MOX fuel sales support Analysis of MOX loaded cores for fuel contracts Analysis of spent/recycled fuel for reprocessing facility support

"* Japan Analysis of European MOX-loaded cores for licensing activities Core design/support for MOX fuel introduction at TEPCO (BWR)

Core design/support for MOX fuel introduction at Kansai (PWR)

"* Germany:

Core design and support for MOX-loaded Gundremingen (BWR)

Licensing authority (TUV) verification analysis

"* Switzerland:

Core design/support for Beznau Units I and 11 (PWR)

On-line core monitoring (GARDEL) for Beznau Unit I

"* U.S.:

Duke Power Studsvik"Scandpower May 23, 2002 24 12

13 KKB I (Beznau Unit 1) Boron Results by Cycle 50 25 4 0 20

= 30 15 E 0

<x20 10 o 0

5 C

c 10 a) 0F

=o0 1 0ý Avv 24 25 26 27 28 Cycle Number Studsvik"Scandpower May 23, 2002 25 KKB2 Boron Results by Cycle 20 2O

• 5 0"

22 23 24 25 26 Cycle Number Studsvik"Scandpower May 23, 2002 26

14 KKB 1 2D Reaction Rates

]RMS 1.33%

3 S40 2.5 30 2

4 S30 S

-1.5

=

x 20 1

10 0.5 U

0 24 25 26 27 Studsvik'Scandpower May 23, 2002 27 KKB 1 3D Reaction Rates IM 2s.36%I 50 Non-equilibrium points 240 4c

' 20 2Z S10 o0 0*

24 25 26 27 2S

(\\cr e MaLy 2,0 Studsvik'Scandpower May 23, 2002 28

15 KKB2 2D Reaction Rates

• 4JRMSl1.13%

3 2.5 2

"1.5 4

S2 0

0 0.5

• 0 22 23,25 Studsvik"Scandpower May 23, 2002 29 KKB2 3D Reaction Rates RM s2.48%I 4

!5

.5-1 4 02

22 S2 0

00 22 23 24 25 26 Studsvik-Scandpower May 23, 2002 30

16 Summary

" CASMO-4 and SIMULATE-3 MOX have already been used for core design and analysis in MOX-fueled PWRs and BWRs.

" Accuracy in MOX-fueled cores is comparable to that obtained in LEU-fueled cores.

" CASMO-4/SIMULATE-3 MOX can be applied with confidence for Duke Power's upcoming MOX applications.

Studsvik Scandpower remains committed to continued development of models and codes for applications in LEU-and MOX-fueled LWRs.

Studsvik-Scandpower May 23, 2002 31

2 Qualification of Nuclear Analysis Methodologies Meeting with Nuclear Regulatory Commission DPC-NE-1005P Review Jim Eller Duke Power May 23, 2002 SDuke A

Porvwc-Goal Define a modeling technique which has:

"* Acceptable accuracy

"* Reliable performance

"* Direct and understandable approach

"* Builds on existing experience base

"* Effective use of human and computer resources PkDuke

PWR Measurements

"* BOC startup tests at HZP Critical soluble boron concentration Control rod bank worth Isothermal temperature coefficient

"* At power critical soluble boron letdown

"* At power core power distribution measurement Pk Duke4

Measurement of Core Power Distribution

"* Moveable incore fission chamber

"* Travels up central instrument tube of fuel assembly

"* Approximately 1/3 of all fuel assemblies instrumented

"* Measured electrical signal is proportional to flux level in center of fuel assembly

"* Flux level measured in radial center of fuel assembly is related to average assembly power Pk Duke OPower.

Core Design Methodology

" Requires a conservative verification of multiple fuel pin performance criteria

" Precision of core model pin by pin power distribution prediction must be known

" Measured pin by pin power distribution data is not available from power reactor operation 4VP6~6

Laboratory Experiments

" Some experiments measure power distribution in critical arrays of fuel pins

" Useful experiments utilize materials and lattice arrangements that are similar to PWR fuel

" Analytic models of experimental geometries allow comparison of predicted and measured pin power distributions Ph Duke Summary

" DPC-NE-1005 seeks to extend and improve currently licensed reload core design methodology

" Goal is to define a core modeling technique that is accurate, consistent, and efficient

" Benchmark approach is dictated by available measurements ADýý c

8