Application and development of tools for HTR neutronics and thermal hydraulics analysis at IKE
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1 Application and development of tools for HTR neutronics and thermal hydraulics analysis at IKE W. Bernnat, M. Buck, N. BenSaid, K. Hossain, M. Mesina IKE, University of Stuttgart The PBMR-4 Core Design - 2nd Workshop - OECD/NEA, Issy les Moulineaux January 26
2 General Objectives: Development of an Integrated Tool for the Analysis of Stationary and Dynamic Behavior of HTRs thermal flux, x Height, cm -5 5 Neutronics reactor design (stationary and transient) Radius, cm 25 3 In-Core Thermalhydraulics and fuel thermal behavior. Modeling of components and dynamics of Power Conversion Unit (PCU). The PBMR-4 Core Design - 2nd Workshop - OECD/NEA, Issy les Moulineaux January 26 2
3 HTR applications Neutronics stationary Neutronics transient Thermal hydraulics PCU simulation The PBMR-4 Core Design - 2nd Workshop - OECD/NEA, Issy les Moulineaux January 26 3
4 Codes available ZIRKUS, KIND THERMIX/KONVEK ANISN/DORT/TORT (1D-3D S N -transport) Monte Carlo : MCNP(X)/KENO NJOY (Cross section preparation) RSYST (IKE), general reactor physics applications MICROX-2 RESMOD (IKE), multicell spectralcode FLOWNEX (M-Tech), system code The PBMR-4 Core Design - 2nd Workshop - OECD/NEA, Issy les Moulineaux January 26 4
5 ZIRKUS Modular system with funcional modules and data base system Restricted to pebble bed HTR types Flexibility high, but rather complex Several adaptations to PBMR like systems (annular core) were made together with simplifications of input structure The PBMR-4 Core Design - 2nd Workshop - OECD/NEA, Issy les Moulineaux January 26 5
6 Main modular ZIRKUS sequence for stationary neutronics - thermal hydraulics calculations KUGEL WQRFL NIVERM NEWA MICROX MAGRU HBLOCK VORNEK Definition of spherical fuel element Cross-sections for reflectors (temperature, density, impurity) Number densities initialisation/shuffling Dancoff factors Cross section resonance treatment and spectral calculation (microscopic XS for spectral zones) Macroscopic cross sections for burnup zones 2D diffusion calculation, variable group numbers, Xe distribution Power density calculation The PBMR-4 Core Design - 2nd Workshop - OECD/NEA, Issy les Moulineaux January 26 6
7 Main modular ZIRKUS sequence for stationary neutronics - thermal hydraulics calculations ZBUCK S B U R N N Z W NZW AUS LDTHERMI T H E R M IX TZIRK DIFK Bucklings for spectral zones derived fro m d iffu sio n calcu latio n D ep letio n calcu latio n fo r all b u rn u p zones D ec a y h eat c alcu latio n Interface decay heat data Interface ZIRKUS-THERMIX T h erm al h yd rau lic c alcu latio n Interface THERMIX-ZIRKUS Diffusion constants for cavity region The PBMR-4 Core Design - 2nd Workshop - OECD/NEA, Issy les Moulineaux January 26 7
8 ZIRKUS-features first core transition core equilibrium core flow pattern of pebbles reload strategy fuel/moderator elements 2 D representation for burn up 3 D stationary calculations (burnup distribution 2D) with Monte-Carlo codes MCNP and/or KENO temperature coefficients water ingress xenon reactivity flexibility, connection to other codes possible The PBMR-4 Core Design - 2nd Workshop - OECD/NEA, Issy les Moulineaux January 26 8
9 Typical ZIRKUS/THERMIX applications Universität Stuttgart Crosssection Database MICROX/ MCNP FIRST-CORE TRANSITION CORE EQUILIBRIUM CORE ZIRKUS Database Calculation of Reactivity Coefficients. Database for Transients Monte Carlo, S N -Transport Thermal- LOFC hydraulic DLOCA analysis The PBMR-4 Core Design - 2nd Workshop - OECD/NEA, Issy les Moulineaux January 26 9
10 ZIRKUS options SINGLE MODULE INPUTS SEQUENCE OF MODULES EXECUTION OF SINGLE MODULES AND SEQUENCES ARCHIVING OF CALCULATIONAL RESULTS RESTART-OPTION INTERFACE TO THERMAL HYDRAULICS INTERFACE TO TRANSPORT PROGRAMS INTERFACE TO TRANSIENT CODES KIND; RZ KIND The PBMR-4 Core Design - 2nd Workshop - OECD/NEA, Issy les Moulineaux January 26 1
11 ZIRKUS modular chain Universität Stuttgart KUGEL, DIFK WQRFL ZBUCK NIVERM SBURN NEWA NZW MICROX NZWAUS MAGRU LDTHERMI HBLOCK THERMIX VORNEK TZIRK The PBMR-4 Core Design - 2nd Workshop - OECD/NEA, Issy les Moulineaux January 26 11
12 Simulation platform Universität Stuttgart The PBMR-4 Core Design - 2nd Workshop - OECD/NEA, Issy les Moulineaux January 26 12
13 Cross section data base and spectral code Data evaluations: JEFF 3.1 MICROX-2 spectral code RESMOD spectral code MCNP(X) Monte Carlo Thermal neutron scattering laws for graphite The PBMR-4 Core Design - 2nd Workshop - OECD/NEA, Issy les Moulineaux January 26 13
14 MICROX-2 for detailed calculation of spectra in fast, resonance and thermal range FDTAPE 92 groups fast range Geometry of cell, temperature Dancoff factor, buckling Zonewise Nuclide composition Z I GARTAPE Pointwise Resolved resonan-ce range MICROX-2 R K GGTAPE 11 groups thermal range Condensed microscopic cross sections U S The PBMR-4 Core Design - 2nd Workshop - OECD/NEA, Issy les Moulineaux January 26 14
15 Present MICROX-2 Library based on JEFF Nuclides Temperatures: 3; 6; 8; 9; 11; 14; 18; 24 K 1-H-1 34-Se Ag Ce Eu Th Am B-1 36-Kr Ag-11m 58-Ce Eu Pa Am B Rb Cd Ce Eu Np Am-242m 6-C- 38-Sr-9 48-Cd Ce Eu Np Am N-14 4-Zr Cd Pr Eu Np Am O-16 4-Zr-95 5-Sn Pr Eu Np Am-244m 14-Si Nb Te-129m 6-Nd Gd Np Cm Si Nb Te Nd Gd U Cm Si-3 42-Mo-1 53-I Nd Gd U Cm Cr-5 42-Mo I Nd Gd U Cm Cr Mo I Nd Gd U Cm Cr Tc Xe Nd Gd U Cm Cr Ru Xe Nd Gd U Cm Mn Ru Xe Nd Tb Pu Cm Fe Ru Xe Pm Tb Pu Cm Fe Ru Xe Pm Dy Pu Cm Fe Ru Cs Pm-148m 66-Dy Pu Cm Fe Rh Cs Pm Dy Pu Ni Rh Cs Pm Dy Pu Ni-6 46-Pd Cs Sm Dy Pu Ni Pd La Sm Ho Ni Pd Sm Hf Ni Pd Sm Hf Pd Sm Hf Sm Hf Sm Hf Sm-154 Universität Stuttgart The PBMR-4 Core Design - 2nd Workshop - OECD/NEA, Issy les Moulineaux January 26 15
16 Thermal neutron scattering on graphite Frequency distributions of graphite Universität Stuttgart 4 35 Rho(Omega) in 1/eV GA ORNL NCSU Omega in ev The PBMR-4 Core Design - 2nd Workshop - OECD/NEA, Issy les Moulineaux January 26 16
17 Comparison of measured and calculated spectra EPhi(E) (arbitrary units) Experiment GAC-5A 274 K GAC-5a : GA model GAC-5a : ORNL model GAC-5a : NCSU model Experiment GAC-5F 6 K GAC-5f : GA model GAC-5F : ORNL model Neutron Energy (ev) The PBMR-4 Core Design - 2nd Workshop - OECD/NEA, Issy les Moulineaux January 26 17
18 Transient calculations ZKIND and RZKIND codes Both are neutron kinetic/dynamic codes for the pebble bed core primarily developed for the HTR-MODUL reactor. The Tasks are: calculation of time dependent power distributions and reactivity effects temperature distributions The PBMR-4 Core Design - 2nd Workshop - OECD/NEA, Issy les Moulineaux January 26 18
19 Thermal-hydraulic Models Different models to simulate the heat production and temperature distribution in fuel Macro Model Macro and Micro Model enhanced Micro Model (2D RZKIND) (1D ZKIND) (1D PKIND) The PBMR-4 Core Design - 2nd Workshop - OECD/NEA, Issy les Moulineaux January 26 19
20 Macro Model: Macro and Micro model of Heat conduction In RZKIND at the time a homogeneous model is implemented. The Fuel spheres are subdivided into several shells but the fuel temperature is assumed to be identical to the graphite temperature in the corresponding shell. The fuel temperature calculated for reactivity feedback is averaged over all shells containing fuel (except outer shell). The moderator temperature is the average of graphite temperatures of all shells. The PBMR-4 Core Design - 2nd Workshop - OECD/NEA, Issy les Moulineaux January 26 2
21 Thermal-hydraulic Models used in KIND codes Macrosystem Microsystem <T> τ =T - <T> r K r C rm M r M r S Shell Matrix Kernel Share of Matrix Graphite Coating <T> : average Temperature over the fine structure τ : Temperature increase in the particle The PBMR-4 Core Design - 2nd Workshop - OECD/NEA, Issy les Moulineaux January 26 21
22 Data Flow to RZKIND ZIRKUS ZIRKUS Geometry Cross Section data Data Library RZKIND OUTPUT FILE User s Input File KONTR Condensation of cross sections ZIRKUS-Geometry RZKIND-Geometry ZKIND VPF Polynomial fit of cross sections RZKIND Data-Library The PBMR-4 Core Design - 2nd Workshop - OECD/NEA, Issy les Moulineaux January 26 22
23 Features of the KIND Codes Simulate Inlet temperature disturbances Mass flow disturbances Changes in power Reactivity disturbances Control rod movement or SAS insertion Xenon effects external reactivity effects The PBMR-4 Core Design - 2nd Workshop - OECD/NEA, Issy les Moulineaux January 26 23
24 Relative Power 2, 1,8 1,6 1,4 1,2 1,,8,6,4,2 Example: Control rod withdrawal (2D RZ-KIND calculation for HTR Modul) (a) (b) (c) control rod 1.Scram (a) 2.Scram (b) 2.Scram, control rods withdraw (c) (a) first scram works at 12% neutron flux (b) second scram works if average He coolant outlet Temperature exceeds limit (c) neither 1. Scram nor 2. scram works control rods withdraw to the maximum upper end position (-1cm), Time (s) The PBMR-4 Core Design - 2nd Workshop - OECD/NEA, Issy les Moulineaux January 26 24
25 Influence of detailed fuel temperature model (control rod velocity: 1 cm/s) Relative Power [-] Withdrawal of all rods with 1 cm/s RZKIND (bue) calculation Particle Modell of ZKIND (blue) 2D RZ-KIND calculation (red) 1D ZKIND particle model calculation Extreme differences between the two models Time [s] The PBMR-4 Core Design - 2nd Workshop - OECD/NEA, Issy les Moulineaux January 26 25
26 Temperature [ C] Fast transient calculation UO 2 Temperature Fuel Coat.1 Coat.2 Matrix-max Matrix-average FE_Shell average Matrix Temperature Fuel Element Shell ZKIND (Particle model) - 1 cm/s Time [s] The PBMR-4 Core Design - 2nd Workshop - OECD/NEA, Issy les Moulineaux January 26 26
27 General Objectives : Thermal hydraulics Reliable and flexible integrated tool for analysis of static and dynamic behavior of HTRs: Emphasis on detailed description of in-vessel behavior, especially coupling of thermal-hydraulics and neutronics Existing tools as starting point for further development Present code THERMIX The PBMR-4 Core Design - 2nd Workshop - OECD/NEA, Issy les Moulineaux January 26 27
28 THERMIX discretisation Examples of analyses and applications HTR with annular core [PBMR] Design of PBMR with annular core and compact central column Thermal power: 4 MW System pressure: 85 bar Inlet temperature: 5 C Outlet temperature: 9 C Initial steady state temperature distribution central grahite column annular core The PBMR-4 Core Design - 2nd Workshop - OECD/NEA, Issy les Moulineaux January 26 28
29 Examples of analyses and applications Results for Annular Core Reactor LOFC with depressurization Solid temperature development LOFC without depressurization: Solid temperature development Gas temperature and velocity (annular core only) The PBMR-4 Core Design - 2nd Workshop - OECD/NEA, Issy les Moulineaux January 26 29
30 Further developments 3D calculations in ZIRKUS Improved treatment of streaming in cavities for diffusion calculations New 2D/3D thermal hydraulics module Extensions of the space time kinetics modules more energy groups more flexible mesh grid coupling to new thermal hydraulics code detailed heat transfer model for coated particles embedded into graphite matrix for 2D version The PBMR-4 Core Design - 2nd Workshop - OECD/NEA, Issy les Moulineaux January 26 3
31 Benchmark PBMR 4 Neutronics/Thermal hydraulics ZIRKUS and DORT model (exercise 1) THERMIX and KONVEK model (exercise 2) The PBMR-4 Core Design - 2nd Workshop - OECD/NEA, Issy les Moulineaux January 26 31
32 Mesh and media assignment : Steady State Exercise 1 73,6 8,55 92, ,5 24,5 211, ,6 26, ,5 292,5 Mesh subdivision ,6 8,6 92, ,6 6,95 11,5 7, ,95 11,5 6,95 13,6 18, ,4 12, The PBMR-4 Core Design - 2nd Workshop - OECD/NEA, Issy les Moulineaux January 26 32
33 Compositions for Benchmark Model (THERMIX-Konvek) The PBMR-4 Core Design - 2nd Workshop - OECD/NEA, Issy les Moulineaux January 26 33
34 Results neutronics exercise 1 Calculations performed with the Diffusion code of ZIRKUS (2D) Cavities treated via direction dependent diffusion coefficients For comparisons: DORT 2D S N -calculations P - transport corrected Cavities treated as vacuum Value ZIRKUS/Diffusion DORT-S16 DORT-S12 DORT-S8 DORT-S6 DORT-S4 DORT-S2 k-eff Maximum Power density (W/cm3) Maximum fast flux (n/cm2/s) 2.5E E E E E E E+14 Maximum thermal flux (n/cm2/s) 3.55E E E E E E E+14 Leakage from core (% per lost neutron) Leakage from domain (% per lost neutron) The PBMR-4 Core Design - 2nd Workshop - OECD/NEA, Issy les Moulineaux January 26 34
35 Fast flux density PBMR-4 exercise 1 DORT S16 7 x fast flux, Bottom Top Height, cm Radius, cm The PBMR-4 Core Design - 2nd Workshop - OECD/NEA, Issy les Moulineaux January
36 Thermal flux density PBMR-4 exercise 1 DORT- S x thermal flux, Bottom Top Height, cm Radius, cm The PBMR-4 Core Design - 2nd Workshop - OECD/NEA, Issy les Moulineaux January
37 Relative difference DORT S8 to S16 reference solution (fast group) 2 1 diff fast flux s8-s16, Height, cm Radius, cm The PBMR-4 Core Design - 2nd Workshop - OECD/NEA, Issy les Moulineaux January
38 Relative difference DORT S8 to S16 reference solution (thermal group) 2 diff thermal flux s8-s16, Height, cm Radius, cm The PBMR-4 Core Design - 2nd Workshop - OECD/NEA, Issy les Moulineaux January 26 38
39 Relative difference DORT S2 to S16 reference solution (fast group) 2 15 diff fast flux s2-s16, Height, cm Radius, cm The PBMR-4 Core Design - 2nd Workshop - OECD/NEA, Issy les Moulineaux January
40 Relative difference DORT S2 to S16 reference solution (thermal group) 1 diff thermal flux s2-s16, Height, cm Radius, cm The PBMR-4 Core Design - 2nd Workshop - OECD/NEA, Issy les Moulineaux January
41 Fission neutron source distribution (DORT) S-16 calculation for exercise x fission source distribution, TOP Height, cm Radius, cm The PBMR-4 Core Design - 2nd Workshop - OECD/NEA, Issy les Moulineaux January 26 41
42 Relative difference in fission neutron source distribution between S-8 and S-16 calculation diff source distribution s8-s16, TOP Height, cm Radius, cm The PBMR-4 Core Design - 2nd Workshop - OECD/NEA, Issy les Moulineaux January 26 42
43 Relative difference in fission neutron source distribution between S-2 and S-16 calculation 4 3 diff source distribution s2-s16, TOP Height, cm Radius, cm The PBMR-4 Core Design - 2nd Workshop - OECD/NEA, Issy les Moulineaux January 26 43
44 Exercise 2 Heat flow to surface cooling system 974,5 kw! Maximum fuel temperature 996,7 C! Inlet helium temperature ( C) 5 Outlet helium temperature ( C) Inlet Pressure (MPa) Outlet Pressure (MPa) 9 Total pressure drop (kpa) - Inlet to outlet Pebble bed pressure drop (kpa) - Top / bottom fuel 278 Average fuel temperature ( C) Average moderator temperature ( C) Average helium temperature in core ( C) The PBMR-4 Core Design - 2nd Workshop - OECD/NEA, Issy les Moulineaux January 26 44
45 Pebble surface temperature distributiion exercise Radius [m] , , ,5 The PBMR-4 Core Design - 2nd Workshop - OECD/NEA, Issy les Moulineaux January ,5 Height [m] Surface Temperature [ C ]
46 Flow field exercise 2 The PBMR-4 Core Design - 2nd Workshop - OECD/NEA, Issy les Moulineaux January 26 46
47 Remarks exercise 1 k eff strongly dependent from diffusion constants in cavity over pebble bed core for diffusion method k eff about 1% lower for S N (DORT) method, depending from SN order S N order from influence of fluxes at outer boundaries and cavities For S N calculations the absorber cross sections should be adopted transport correction for P calculation adequate? Treatment of cavities well defined for S N calculations The PBMR-4 Core Design - 2nd Workshop - OECD/NEA, Issy les Moulineaux January 26 47
48 Remarks exercise 2 Spatial discretisation could be refined fuel, moderator and reflector temperatures seem to be adequate for steady state cases The PBMR-4 Core Design - 2nd Workshop - OECD/NEA, Issy les Moulineaux January 26 48
49 Remarks exercise 3 A reproduction of the reference cross section data was not possible Coupled calculations with interpolated cross sections led to inconsistent results Some interpolated data should be compared The PBMR-4 Core Design - 2nd Workshop - OECD/NEA, Issy les Moulineaux January 26 49
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