Special HTGR topics: Modelling, Tools and Uncertainty analysis Frederik Reitsma Nuclear Power Technology Development Section Department of Nuclear Energy Joint IAEA-ICTP Workshop on the Physics and Technology of Innovative High Temperature Nuclear Energy Systems
Required Analysis for Design
Engineering Analysis Synergy ----------------------------------- CFD to SAG NEA to CFD Detailed Component Reactor Power Temperatures Computational Profile Nuclear Structural Fluid Engineering Analysis Dynamics Analysis (SAG) (CFD) (NEA) CFD to NEA CFD to SAG Reactor Flow Distribution Fluid/Structure Interaction and Temperatures CFD to TFAG Detailed Flow TFAG to CFD Distributions Cycle Flow TFAG to/from NEA Conditions Detailed Flow Distributions and Neutronic 94 136 92 145 MT 1803 93 95 96 Annular Flow Path Data 98 1804 1812 138 H Core 1817 Thermal Fluid P RPV Core sturcture graphite 139 Barrel CHT CHT CHT CHT CHT 1819 RCCS Reactor Core 1 1805 1814 1806 91 1807 146 1808 1813 1815 1818 RD 147 148 78 1816 Analysis M PT MT CHT 1809 140 Reactor Outlet Reactor_Inlet 3 6 137 2 1810 1802 1811 1801 90 H PHT (TFAG) Control rod bypass flow RD 142 141 RD 1703 1704 1705 1706 1707 4 M HTR 2010 – Confidentia 3 l / Prague, Oct 18-20,
Nuclear Engineering Analysis Reactor Shielding and Activation Core Neutronics / Thermal Hydraulics Analyze Shielding and Activation of core structures and Analyze Neutronic Design for surrounding feedback to both engineering and safety Monte Carlo Analysis (MCNP) or simplified transport analysis Steady-State (VSOP, MCNP) (MicroShield) and Transient (TINTE) Analysis Input to engineering on core Input to engineering on core component activities for component temperatures, power maintenance / decommissioning profiles etc. Input to safety for worker dose Input to safety on maximum fuel temperatures, control rod worth Dust Generation and etc. Activation Graphite and metallic dust Fission Product Releases generation in the core and fuel handling system and activation Determine Fission Product of the dust Releases for both normal operation and accident Fuel Source Term scenarios Neutron and Photon source Using Diffusion Theory from spent and used fuel (GETTER, NOBLEG) Input to the development of the Input to the rest of the source used and spent fuel tanks and term analysis chain HTR 2010 – Prague, Oct waste handling 4 18-20, 2010
Nuclear Engineering Analysis Plant Shielding and Activation Helium Pressure Boundary Source Term Analyze Shielding and Activation of components Analyze the source term within the helium pressure boundary Monte Carlo Analysis (MCNP) during normal operation and or simplified transport analysis releases in accident scenarios (MicroShield) New code under development Input to engineering on (DAMD) component activities for maintenance and Considers behaviour of dust and decommissioning fission products in the system Input to safety for worker dose Building Retention Public Dose Analyze the behaviour of the Analyze the expected public building for accident scenarios dose in accident scenarios Accident analysis code Input to the building design and (ASTEC) safety Input to the building design and safety HTR 2010 – Prague, Oct 18-20, 2010 5
Mechanical Engineering Analysis Computational Fluid Dynamics ----------------------------------- o Responsible for the Detailed Component and Sub-System Analysis of the PBMR Plant o Expert Use of Commercial and In-House Tools ▪ ANSYS Fluent ▪ Star-CD/Star CCM+ ▪ In-House Customisation o Analysis Functions ▪ 3D Simulation of Complex Phenomena ▪ Provide Insight to System Behaviour ▪ Component Optimisation ▪ Input to Safety Analysis HTR 2010 – Prague, Oct 18-20, 2010 6
Mechanical Engineering Analysis Computational Fluid Dynamics ----------------------------------- Predicts leak flows used in all reactor codes (Flownex, TINTE, VSOP) Bypass Core flow Leakage HTR 2010 – Prague, Oct 18-20, 2010 7
Mechanical Engineering Analysis Structural Analysis ----------------------------------- o Responsible for the Structural Analysis of the PBMR Plant o Expert Use of Commercial Tools ▪ Nastran/Patran ▪ Marc/Mentat ▪ PFC 3D ▪ Dytran o Analysis Functions ▪ Modelling Complex Structural Systems ▪ Interpreting Load Information with Respect to ASME Codes ▪ Structural Verification ▪ Input to Safety Analysis See Paper 158 Geometric layout optimisation of graphite reflector components by Christiaan 8 HTR 2010 – Prague, Oct 18-20, 2010 Erasmus & Michael Hindley
Mechanical Engineering Analysis Structural Analysis ----------------------------------- Sphere Flow Analysis HTR 2010 – Prague, Oct 18-20, 2010 9
Nuclear Eng Analysis: Nuclear and Mechanical Engineering VSOP Analysis Software TINTE / MGT RELAP Analysis Software Verification and Validation NOBLEG V&V plans (and many other supporting GETTER documents) RADAX Mech Eng Analysis: Procedures accepted by NNR DAMD ANSYS MCNP Software Engineering focus areas: Star-CD ASTEC Analysis software development and Fluent implementation of associated quality and SCALE MSC V&V processes GENII Flownex Legacy analysis software reverse AMBER RELAP engineering MicroShield iSight Legacy software maintenance NJOY Matlab Methods and software development ATILLA EFD.Lab DIREKT PFC3D + 3DEC Spectra Risk Spectrum MONK HRA Calculator Open Foam Fuelnet Salome vulaSHAKA HTR 2010 – Prague, Oct 18-20, 2010 10
Uncertainty Analysis
Introduction • Traditionally conservative analyses are used for nuclear power plant safety and licensing analyses • Reliable and high fidelity codes and models allows the use of best- estimate plus uncertainty analysis (as replacement) • Methodologies to determine the uncertainties in a consistent way have been developed and applied for LWRs • We need to determine if the same methods can be used for other reactor technologies, where the sources and magnitude of uncertainties are different. • Also, since very limited experimental results are available for HTGRs some approaches used in the LWR methodology cannot be followed • The evaluation of uncertainties in high temperature gas-cooled reactors analysis are being investigated within the IAEA coordinated research project AMNT Berlin 2017 17 May 2017 12
Purpose: • To determine the uncertainty in HTGR calculations at all stages of coupled reactor physics, thermal-hydraulics and depletion calculations • Follow the approach of the OECD / NEA UAM LWRs; and aims to: – establish and utilize a benchmark for uncertainty analysis in best- estimate coupled HTGR modelling and analysis – use as a basis a series of well defined problems with complete sets of input specifications. – subdivide the coupled system calculation into several steps, each of which can contribute to the total uncertainty – identify input, output, and assumptions for each step. – the resulting uncertainty in each step will be calculated (including propagating from previous steps). – where possible have reference results and/or experimental results to be used AMNT Berlin 2017 17 May 2017 13
Why do we expect different results ? • Compared to LWRs: – Different materials – Graphite moderator (vs water moderated) – Higher operating temperatures ~ Average 900K – 1260K maximum – Higher enrichments ~8.5% - 15.5% – Higher burnup ~ 100,000 MWd/te • Modelling aspects – Resonance treatment of coated particle fuel (double heterogeneity) – Stochastic nature of fuel (particles; pebbles; movement) – Much harder neutron spectrum (than PWRs) due to high temperatures and use of graphite moderator – long mfp of neutrons exclude the use of simple assembly calculations • spectrum calculations need special treatment / in-line / mini-core • need multi group core analysis (4, 13, 22 groups) • mostly can ’ t use the traditional PWR calculation chain of assembly calculations -> 2-group parameterized library -> core simulator • more difficult to propagate uncertainties step-wise – need different approach AMNT Berlin 2017 17 May 2017 14
CRP on HTGR Uncertainty in Analysis • Objective: – To contribute new knowledge towards improving the fidelity of calculation models in the design and safety analysis of high temperature gas-cooled reactors by fully accounting for all sources of uncertainty in calculations. Sensitivity and Uncertainty Analyses Uncertainty Uncertainties of Tool A Input Parameters Propagation Tool C Tool B Uncertainty Data Physical Data A Set of Method A Problems Technological Data Method C Model Approximations Method B 1 AMNT Berlin 2017 17 May 2017 15 5
Reference designs Prismatic Design: The MHTGR (an earlier General Atomics 350MWth design considered for NGNP) was adopted as the main prismatic reference design. 120 o Symmetry Line Permanent Core Barrel Reflector (2020 Coolant Channel RPV (SA-533B) (Alloy 800H) Graphite) Neutronic Boundary Pebble Bed Reactor Design: The HTR-Module-based design, upgraded to 250MWth will be the reference design with some Fuel Block (H-451 Replaceable Graphite) simplifications introduced. Reflector Block Fuel Block with (H-451 Graphite) RSC Hole (H-451 Replaceable Reflector Graphite) Block with CR Hole Outside Air (H-451 Graphite) OECD / NEA UAM-10+ 16
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