ENHANCEMENT OF COMPUTER SYSTEMS FOR CANDU REACTOR PHYSICS SIMULATIONS

Transcription

1 ENHANCEMENT OF COMPUTER SYSTEMS FOR CANDU REACTOR PHYSICS SIMULATIONS E. Varin, M. Dahmani, W. Shen, B. Phelps, A. Zkiek, E-L. Pelletier, T. Sissaoui Candu Energy Inc. WORKSHOP ON ADVANCED CODE SUITE FOR DESIGN, SAFETY ANALYSIS AND OPERATION OF HEAVY WATER REACTORS October 2-5, 2012

2 Outline Physics codes Physics codes in the CANDU Industry Current status Functionality : Why improvements are needed Effort to enhance physics codes Development New Tool MINER Coupling Strategy Assessment of GANLIB and DONJON Development Approach Conclusion 1

3 Physics codes in the Candu Industry Physics codes for Reactor Analysis Design Operation support Safety analysis Static and time-average model Kinetics and FP transient Control Interface with TH Deterministic Codes 2

4 Physics codes in the Candu Industry General Modelling Approach Three steps methodology Lattice calculation Supercell calculation Reactor full-core calculation Generally sequential calculations 3

5 Physics codes in the Candu Industry D. Altiparmakov 2011 Unrestricted 4 Illimité

6 Physics code in the Candu Industry Current Status Design Applications WIMS-AECL DRAGON RFSP (SIMULATE, TIME-AV) MCNP SCALE package (TSUNAMI) Safety Analysis RFSP (CERBERUS, CERBRRS) Operational Analysis RFSP (SIMULATE) HQSIMEX SORO FP transient software 5

7 Main features of the lattice code WIMS-AECL History: WIMS-D distributed to AECL in 1971, developed by AECL since Library: 89-group libraries based on ENDF/B versions V, VI and VII Geometry: 2-D, single-cell with fuel cluster, multiple cells with fuel cluster, sector, off-center sagged model Method of Solution: 2-D Collision Probability (CP) method in full energygroup structure Resonance Treatment: Advanced self-shielding capabilities (capability to represent distributed self-shielding effects) Programming Features: Modular design, Re-written with FORTRAN90/95 Other Features: Critical buckling search Depletion calculation Simple input cards for production run 6

8 Main features of the supercell code DRAGON History: Developed by Ecole Polytechnique de Montréal since 1989 Library: multi-group libraries based on ENDF/B versions V, VI, VII or JEF Geometry: Capabilities to define any complicated 2-D and 3-D cluster geometry, multiple cells with cluster Method of Solution: 2-D/3-D CP method and Method of Characteristics (MOC) Resonance Treatment: advanced self-shielding capabilities including subgroup method Programming Features: Modular design, FORTRAN77, GANLIB, 64-bit capability Other Features: Comprehensive SPH technique Critical buckling search Visualization of 2-D geometries Flexibility (users can implement some computational scheme in CLE-2000, without having to recompile the code) 7

9 Main features of the core-calculation code RFSP Problem Solved: Two-group static and kinetics neutron-diffusion equations for the eigenvalue and fixed-source problems, fluxmapping equations Geometry: 3-D, Cartesian geometry Method of Solution: Finite-difference method Cross-Section Model: micro-depletion method, simple-cell method (SCM), multicell methodology Other Features: Xenon modelling Reactor Control capability Use thermal-hydraulics data from code, such as CATHENA 8

10 Reactor Core Theory Homogenized regions : Fuel Bundle compositions and properties ( )

11 Physics code in the Candu Industry Inter-dependency and Cross Section model PPV and history-based calculations WIMS-AECL Macroscopic XS WIMS Utilities SCM Micro Depletion Multicell correction 10

12 SCM Method Moderator Gap CT Fuel PT Coolant 11

13 SCM Method WIMS-AECL Main Fuel Table for SCM Simple-Cell Calculations Core Flux Calculations 3 Iterations 12

14 Micro-Depletion Method State-of-the-art method and proven technology (developed and used in LWR industry several years ago, SIMULATE-3, ANC) Cell macroscopic XS and the microscopic XS for tracked nuclides are interpolated as function of irradiation and state parameters. The burnup equations are solved for each fuel region according to its local conditions, to obtain the dependence of the nuclide concentrations on the burnup history: Changes in the selected nuclides are explicitly calculated The balance of the nuclides are implicitly treated Applicable to any fuel types 13

15 Calculation of Lattice-Cell Properties (L ) ( L 0 ) x is the actual lattice-cell cross section in diffusion is the reference lattice-cell cross section x x ( L) x ( L 0 ) l l x i m N i m N i 0, m i x, m l is the correction to the reference lattice-cell cross x section when a single state parameter l is perturbed, i N is the actual number density of nuclide m over m region i, i N is the reference number density of nuclide m over 0,m region i, and ( x, m, Tf ) is the microscopic cross section of nuclide m over region i. 14

16 MultiCell Correction het _ x factor single x cell x single cell x het_ factor x Periodic BC at this surface 2 Lattice Pitches Y Void BC at this surface Periodic BC at this surface Reflective BC at this surface X 15

17 Current Status of Physics Codes There are sources of inaccuracy in the current deterministic methods: The first category due to the approximations in transport calculations: Geometry Resonance Self-Shielding treatment Leakage treatment Transport solvers The second category due to the approximations in the core calculations: Energy group Cross Section Model Homogenization Process Fuel/Reflector interface Diffusion approximation 16

18 Current Status of Physics Codes Current three steps methodology Lattice calculation Supercell calculation Reactor core calculation Two step methodology Lattice calculation including the control elements Reactor core calculation Continuous Energy Formalism Even Parity Transport Equation Numerical Methods Stochaistic Monte Carlo Characteristics Integrodifferential Sn Diffusion Deterministic Methods Pn Multigroup Formalism SPn Integral Collision Probability One step methodology Full-core transport calculation Finite Difference Nodal Finite Elements W. Shen

19 Effort to develop new Physics toolset 18

20 Effort to develop new Physics toolset Towards the Next generation toolset Development New diffusion code MINER Use nodal methods Test Database (HDF) Coupling Strategy RFSP - CATHENA Assessment of GANLIB and DONJON 19

21 Development of New Tool Development of New Tool : MINER Current RFSP limitations: 2 group formalism Mesh centered finite difference method The MINER code has been developed to solve the multi-group diffusion equation, using both the finite-difference formulation and the nodal method formulation: Solves NG-group diffusion problems, where NG may be any integer; Couples with RFSP to solve full-core problems based on the RFSP input file; Works for 1D, 2D, and 3D Cartesian problems; Uses the finite-difference method or the nodal method (GNEM); Solves problems with or without ADFs (Assembly Discontinuity Factors) in both the finitedifference method and the nodal method; Written with FORTRAN-90/95; and Uses the data management based on HDF5 (Hierarchical Data Format). 20

22 MINER benchmarking Standard Test Cases 21

23 MINER benchmarking CANDU 22x22 benchmark Use RFSP geometry Use RFSP cross section 22

24 MINER benchmarking CANDU 22x22 8x8 Mesh FDM Compared with GNEM (% of power differences) A B C D E F G H J K L M N O P Q R S

25 MINER: Lesson Learned Testing: New Database format HDF Not conclusive for diffusion solvers Nodal approach GNEM method is adequate Multigroup approach Tested for static and kinetics solution 24

26 Physics in Safety Analysis Kinetics Calculations Based on Thermal-hydraulics data Test of Internal Coupling with RFSP and CATHENA F. Doria CNS Candu Safety Course Unrestricted 25 Illimité

27 Coupling Calculation Schemes Script Calculation Scheme Based on reading & writing input/restart files every coupling interval Results in large computation time Generate CATHENA input CATHENA Read Input/Restart Files Compute Solution Write Restart and Output Files Generate RFSP inputs RFSP Read Input/Restart Files Compute Solution Write *.daf and Output Files Direct Coupling Calculation Scheme Based on exchanging data between the two codes during the simulation More efficient CATHENA Read Input/Restart Files Compute Solution PVM RFSP Read Input/Restart Files Compute Solution 26

28 Verification Transient calculations: Bulk Reactor Power *CERBRRS module, coupling interval = 0.5 s Reactor Inlet Header Temperature *CERBERUS module, coupling interval = 0.1 s Simulation Time (min): Reduced by a factor of 2 with direct coupling Computer speed has a more significant impact *CERBERUS/CATHENA *CERBRRS/CATHENA PVM scripts PVM scripts Computer 1 (PENTIUM 4, 3.0 GHz, 2.5 GB) Computer 2 (Intel Core Duo E8400, 3.0 GHz, 2.0 GB) Protected - Sensitive Protégé - Délicat

29 Lesson Learned from Coupling Testing: Parallel environment Efficient Transient are driven by Thermal-hydraulics phenomena Generate a lot of outputs. 28

30 Assessment of DONJON DONJON is a multi-group, 3-D reactor core physics code It works in conjunction with DRAGON, TRIVAC codes UTILIB TRIVAC GANLIB DRAGON DONJON 29

31 Description of GANLIB GANLIB is a library linked to the software application (API): CLE-2000: computer language that uses a specific syntax to create the input files. Allows the creation of the computational scheme. LCM objects: data structures used to transfer between the modules of the software application and towards the multi-physics application

32 Main features of the code DONJON Problem Solved: Multigroup static and kinetics neutron-diffusion equations for the eigenvalue and fixed-source problems, multi-group low-order transport equations (SPn) Geometry: 2-D and 3-D, Cartesian geometry, Hexagonal geometry Method of Solution: Finite-difference method, finite-element, low-order transport method for the diffusion theory Cross-Section Model: Multi-parameter database, Macro-depletion method, History-based method such as direct use of DRAGON to perform lattice calculations inside DONJON Programming Features: Modular design, FORTRAN77, GANLIB, 64-bit capability Other Features: Flexibility (users can implement some computational scheme in CLE-2000, without having to recompile the code) Reactor Control capability Coupled to a thermal-hydraulics code, such as CATHENA 31

33 Assessment of DONJON Diffusion and low-order transport solvers Geometry Capability (Cartesian and Hexagonal) Modeling PWR reactor Core tracking along with the refuelling Transient calculations Optimization of the fuel management Transport Equation Numerical Methods Stochaistic Deterministic Methods Continuous Energy Formalism Monte Carlo Multigroup Formalism Integrodifferential Integral Even Parity Characteristics Sn Pn Collision Probability Diffusion SPn Finite Difference Nodal Finite Elements W. Shen

34 Assessment of DONJON (cont.) Two categories of solvers: Diffusion: Spatial discretization: Finite Difference Methods or Finite Element Methods Scattering approximation: Diffusion (P 1 ) Low-order transport solvers (SP n ): Spatial discretization: Finite Element Methods Scattering approximation: any order (n>1) Results using CANDU model are OK 33

35 Assessment of DONJON (cont.) 3-D Hexagonal geometry: ZED-2 28-Element UO 2 experiment has been modelled 34

36 Relative Value Assessment of DONJON (cont.) 3-D Hexagonal geometry: Flux distribution (cooled Case) Cu Foil Activity PANTHER DonJon Distance From Core Centre (cm) 35

37 DONJON Capability Core tracking using DRAGON/DONJON: A coupling between DRAGON and DONJON has been developed to simulate a period of core tracking from the Gentilly-2 station Performed with MPI coupling 36

38 Lesson Learned from DONJON Assessment DONJON is capable of modeling CANDU reactors in normal operating conditions as well as simulating transients DONJON is a flexible code that can be directly coupled with the lattice transport code DRAGON and/or with any thermal-hydraulics code. DONJON has the capability to work in the parallel computing environment GANLIB is an adequate repository GANLIB is not a relational database DONJON is currently composed of two many modules DONJON needs enhanced cross section models, such as micro-depletion 37

39 Towards New Physics Codes 38

40 Steps towards New Physics Codes Choose DONJON/GANLIB and DRAGON as basis Development plan for reactor physics code Database and coupling strategy 39

41 Development Methodology RFSP-NG Backbone GANLIB UTILIB Define functionalities DONJON 3-4 RFSP MINER Define and build the Sequences Line-by-line verification of the modules Build Modules The functionalities needed from RFSP and MINER will be written as modules compatible with other modules coming from DONJON Incorporate Modules and Sequences Define new Data structures 40 Prototype: RFSP-NG

42 Development Methodology GEOMETRIC INFORMATION MATERIAL INFORMATION Separate the Main Functions GEOMETRY MATERIAL INFORMATION MATHEMATICAL SOLVER (STATIC/KINETICS/HARMONICS/ADJOINT /SPn/Fixed Source) OUTPUT INFORMATION (EDITION) MATHEMATICAL SOLVER OUTPUT INFORMATION Use Ganlib Data Structures as link between Functions 41

43 Development Methodology : MATERIAL Macroscopic XS Function of (L, B) Macroscopic Incremental XS Function of (L, B) Microscopic XS Function of (L, B) Isotopic Concentration Depletion Information Local Parameters Burnup distribution B Reconstructed Flux Information from Transport Calculations Information from Core Calculations MATERIAL Input information Output information Modules Macroscopic XS For Fuel Information Macroscopic XS For Reflector Information Macroscopic Incremental XS For Device Information 42

44 Development Methodology : Sequences GEOM MATERIAL SYSTEM Control Functionality Burnup Functionality Sequences Grouping of Functions Iteration strategy Encapsulation for users EXAMPLES TIME-AVER STEADY TRANSIENT SOLVER OUTPUT Refuelling Selection Functionality 43 STEADY

45 Steps towards New Physics Codes Database and coupling strategy NGToolset-API NGToolset-API RFSP-API CLE-2000 LCM UTILIB GANLIB 44

46 Summary and Future Work Summary: Review of the current codes Present assessment & development performed Path Forward for reactor codes Based on DONJON and GANLIB Based on WIMS-AECL and Micro-Depletion Based on RFSP Propose a step approach to ensure success and to allow flexibility 45

47 Thank You