Computational and Experimental Benchmarking for Transient Fuel Testing: Neutronics Tasks

Transcription

1 Computational and Experimental Benchmarking for Transient Fuel Testing: Neutronics Tasks T. Downar W. Martin University of Michigan C. Lee Argonne National Laboratory November 19, 2015

2 Objective of Neutronics Task A comprehensive evaluation of existing TREAT Facility neutronics data using the next generation reactor core neutronics codes. This will be performed in accordance with established guidelines per the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP). Objective 1 will yield a fully characterized reactor core with dynamic input and feedback from the U.S. Nuclear Regulatory Commission (NRC) (via advisory board member participation) which may be utilized to support the safety case for the TREAT Facility restart.

3 Neutronics Tasks Neutronics Benchmarks A comprehensive neutronics benchmarking analysis will be conducted using PROTEUS (DoE NEAMS code), PARCS/AGREE (U.S. NRC code) and Open MC (Monte Carlo code). An IRPhEP will result from this comprehensive benchmarking analysis. [Led by the University of Michigan] Steady State Two steady state condition benchmarking tests will be selected and studied. Transient Two transient condition benchmarking problems will be selected and studied.

4 Some Potential TREAT Modeling Issues Boron was introduced to TREAT fuel during manufacturing process, causing ~7.6 ppm impurity (vs 1 2 ppm in reactor grade graphite) Probably non uniform Exact impurity content unknown Carbon in the fuel is partially graphitized There is a lack of data on precise conditions of some experiments (e.g., core temperatures) Some of the TREAT Facility reference documents have conflicting values for precise compositions and dimensions Calculated eigenvalue is very sensitive to geometry of the axial and radial graphite reflector regions (e.g. Outgas tubes, measurement holes, bolt holes, etc. must be modeled) The C 12 thermal capture xsec in ENDFVII.0 is ~20% low

5 Steady State Subtasks Task # Task Title Sub Task Owner 1. Neutronics BenchmarkTask Lead T. Downar, UM 1.1 Steady State (SS) Survey candidate problems T. Downar, UM Preliminary SS modeling of candidate problems T. Downar, UM Down select to two problems for benchmark evaluation T. Downar, UM SS modeling with deterministic U.S. NRC codes PARCS/AGREE T. Downar, UM SS modeling with deterministic NEAMS code PROTEUS C. Lee, ANL SS modeling with Monte Carlo code OPENMC K. Sun, MIT Comparison of experimental data & model results T. Downar, UM Benchmark level evaluation of selected problems T. Downar, UM Evaluation of uncertainties in selected problems T. Downar, UM Preparation of IRPhEP documentation T. Downar, UM Submission of SS benchmark for peer review T. Downar, UM

6 Transient Subtasks 1.2 Transient (TR) Survey available TREAT TR data for benchmark problem T. Downar, UM Preliminary TR modeling of candidate problems T. Downar, UM Down select to two problems for benchmark evaluation T. Downar, UM Perform TR modeling with deterministic U.S. NRC codes PARCS/AGREE T. Downar, UM Perform TR modeling with deterministic NEAMS code PROTEUS C. Lee, ANL Perform TR modeling with Monte Carlo code OPENMC W. Martin, UM Benchmark level evaluation of selected problems T. Downar, UM Evaluation of uncertainties in selected problems T. Downar, UM Preparation of IRPhE Documentation T. Downar, UM Submission of TR benchmark for peer review T. Downar, UM

7 Approaches for Solving the Full Core TREAT Reactor Problem Stochastic Methods Deterministic Methods Monte Carlo Procedure Differential Transport Equation (Boltzman Equation) P N Theory B N Theory S N Theory Diffusion Theory

8 U.S. NRC Coupled Code System Cross Section Generation Lattice Code: SCALE / SERPENT GENPMAXS Neutron Flux Solver Neutron Flux Solver: PARCS Σ Φ Temperature/Fluid Calculation T/H code: TRACE AGREE Cross Section Library (PMAX) Depletion

9 PARCS: Parallel Advanced Reactor Core Simulator U.S. NRC version 3.2 T. Downar Y. Xu V. Seker A. Ward January, 2015

10 PARCS Flux Solution Kernels Geometry Type Kernel Name Solution Method Energy Treatment Angle Treatment CMFD FD 2G Diffusion Cartesian 3D ANM nodal 2G Diffusion FMFD FD MG SP3 NEMMG nodal MG SP 3 Hexagonal 3D Cylindrical 3D CMFD FD 2G Diffusion TPEN nodal MG Diffusion CMFD FD 2G Diffusion FMFD FD MG Diffusion/ SP 3 CMFD = Coarse Mesh Finite Difference NEM = Nodal Expansion Method ANM = Analytic Nodal Method MG = Multigroup FMFD = Fine Mesh Finite Difference

11 Application of PARCS to SPERT RIA Tests

12 PARCS SPERT Results E+02 Test 43 Power Test Initial Power (MW) Pressure Inlet Temp Excursion Rod Worth ($) 43 ~50 W Atm 78 F 4 (300 K) Power (MW) E E E E E+02 PARCS SPERT 70 ~50 W 1500 psi 251 F 4 (394 K) ~50 W 1500 psi 500 F 4 (533 K) E E Time (sec) psi 504 F 4 (533 K) Test 43 Reactivity psi 502 F 4 (533 K) Raymond C. Wang, Yunlin Xu, Nathanael Hudson, Thomas J. Downar, Validation of the U.S. NRC Coupled Code System TRITON/TRACE/PARCS Using the Special Power Excursion Reactor Test III, Nuclear Technology, Volume 183 / Number 3 / 2013 / Pages Reactivity ($) Time (sec) PARCS SPERT

13 PARCS SPERT Results (HZP, HFP) Power (MW) E E E E E E+00 Reactivity ($) Hot Zero Power Test 60 Power Time (sec) Test 60 Reactivity Time (sec) PARCS SPERT PARCS SPERT Power (MW) 7.50E E E E E E E E E E E E E E E E+00 Reactivity ($) Hot Full Power Test 86 Power Time (sec) Test 86 Reactivity Time (sec) PARCS SPERT PARCS SPERT

14 NRC Code System for the NGNP (High Temperature Gas Reactors)

15 AGREE: Macro Scale Heat Conduction Prismatic Basic mesh is one equilateral triangular prism of the hexagonal block This triangle is linked to the neighboring block through parallel conduction and radiation resistances across the gap 15 Cartesian Basic mesh is one rectangular fuel block This block is linked to the neighboring block through parallel conduction and radiation resistances across the gap

16 AGREE Solution Algorithm Semi Implicit Method for Pressure Linked Equations (SIMPLE) Guess the Initial Parameters T,u,v,p Solve the Momentum Equations u*,v* Solve the Pressure Correction Equation p, u, v Correct the Pressure and Velocity p=p+ p p u=(1 u )u + u (u*+u ), v Solve the Energy Equation T* T=(1 T )T + T T* no Converged? yes Update Guessed Parameters Advance Time, Initial guessed parameters at new time are taken from the previous time 16

17 Applications of PARCS/AGREE: HTTR and OECD /NEA MTHGR 350 Benchmark Model Outside Air RPV Permanent Reflector (2020 Graphite) Core Barrel (Alloy 800H) Coolant Channel RPV (SA 533B) Neutronic Boundary Neutronic Boundary Core Restraint Element (Alloy 800H) Upper Reflector Block Replaceable Reflector Block Helium Gap Upper Plenum Thermal Protection Structure (UPTPS) (Alloy 800H) Upper Plenum Metallic Plenum Element (Alloy 800H) Coolant Channel Core Barrel Fuel Block Replaceable Reflector Block with CR Hole Repl. Central Reflector Support Block (2020 Grph) Fuel Block with RSC Hole Bottom Reflector Block (H 451 Grph) Bottom Transition Reflector Block (H 451 Grph) Replaceable Reflector Block (H 451 Graphite) Replaceable Reflector Block with CR Hole (H 451 Graphite) Outside Air Fuel Block with RSC Hole (H 451 Graphite) Fuel Block (H 451 Graphite) Outlet Plenum Metallic Core Support Structure (MCSS) (Alloy 800H) Fluid Inlet Fluid Outlet Flow Distribution Block (2020 Grph) Post Block (2020 Grph) Ceramic Tile (Ceraform 1000) Insulation (Kaowool) 17

18 Task 1.1 (Steady State) Schedule Task # Task Title Sub Task Owner 1. Neutronics Benchmark Task Lead T. Downar, UM 1.1 Steady State (SS) Survey candidate problems T. Downar, UM Preliminary SS modeling of candidate problems T. Downar, UM Down select to two problems for benchmark evaluation T. Downar, UM SS modeling with deterministic U.S. NRC codes PARCS/AGREE T. Downar, UM SS modeling with deterministic NEAMS code PROTEUS C. Lee, ANL SS modeling with Monte Carlo code OPENMC K. Sun, MIT Comparison of experimental data & model results T. Downar, UM Benchmark level evaluation of selected problems T. Downar, UM Evaluation of uncertainties in selected problems T. Downar, UM Preparation of IRPhEP documentation T. Downar, UM Submission of SS benchmark for peer review T. Downar, UM

19 Task 1.2 (Transient) Schedule Transient (TR) Survey available TREAT TR data for benchmark problem Preliminary TR modeling of candidate problems Down select to two problems for benchmark evaluation Perform TR modeling with deterministic U.S. NRC codes PARCS/AGREE Perform TR modeling with deterministic NEAMS code PROTEUS Perform TR modeling with Monte Carlo code OPENMC Benchmark level evaluation of selected problems Evaluation of uncertainties in selected problems Preparation of IRPhE Documentation Submission of TR benchmark for peer review T. Downar, UM T. Downar, UM T. Downar, UM T. Downar, UM C. Lee, ANL W. Martin, UM T. Downar, UM T. Downar, UM T. Downar, UM T. Downar, UM

20 IRPhEP Benchmark Experiment Documentation Section 1 Description of the Experiment A detailed description of the experiments and all relevant experimental data will be provided in the appropriate subsections of section 1. Section 2 Evaluation of Experimental Data Missing data or weaknesses and inconsistencies in published data will be discussed and resolved in the appropriate subsections of section 2. The effects of uncertainties in parameter data on the measurement results will be discussed and quantified. Codes and modelling methods used for calculations of the effects will be specified and the use of data with large uncertainties or data that require assumptions on the part of the evaluator will be justified. Section 3 Benchmark Specifications Benchmark specifications will be provided which will include all the data necessary to construct calculational models that best represent the experiment. The benchmark model specifications will retain as much detail as necessary to model all important aspects of the actual experiment. Section 4 Calculated Results Calculated results obtained with the benchmark model specification data given in Section 3 will be tabulated in this section. These will be regarded as sample calculation and methodologies used for the sample calculations and any other recommendations for the calculations will be described. Section 5 References / Appendices Appendix A will provide a description of the options, cross section data, and an input listing for the codes used in the calculations of the results given in Section 4.

21 Preliminary Results of Applicatioin of SERPENT and PARCS to TREAT Cores Minimum Critical Core M8CAL Minimum Critical Core Methods Serpent Monte Carlo PARCS Deterministic M8CAL

22 Single Fuel Assembly Model from Batman Report 1 SERPENT Model Top Down

23 Control Rod Model from Batman Report 1 SERPENT Model Top Down

24 Minimum Critical Core SERPENT calculation with 100 inactive/ 500 active 1E5 neutron population per cycle at 300K Axial and Radial Vacuum Boundary Conditions SERPENT keff: SERPENT Model Axial View

25 M8CAL Core SERPENT calculation with 100 inactive/ 500 active 1E5 neutron population per cycle at 300K SERPENT keff: SERPENT Model Axial View

26 ENDFVII vs ENDFVII.1 ENDF7 ENDF7.1* Diff (pcm) M8CAL / / K eff calculated using SERPENT with 100 inactive/ 500 active neutron population per cycle. *Only carbon cross section is from 7.1.

27 PARCS Single Assembly Discretization 12 sections in the fuel (10.19cm)* Generate Cross Sections for axial region Vacuum Axial Boundary Conditions Reflective Radial Boundary Conditions Four Reflector Regions Two Graphite Reflector (~60cm) Two Zirconium Spacer (~0.5cm) Orange = Fuel Yellow = Reflector Zirconium Spacer *Mean Free Path ~4cm

28 ENDFVII vs ENDFVII.1 ENDF7 ENDF7.1* Diff (pcm) M8CAL Single Assembly K eff calculated using SERPENT with 100 inactive/ 500 active neutron population per cycle. *Only carbon cross section is from 7.1.

29 Single Assembly Power Shape 1.20E E E E 01 Serpent Keff: PARCS without ZDF: PARCS with ZDF: PARCS with ZDF power shape: on the way 8.00E 01 SERPENT 7.00E 01 PARCS No ZDF 6.00E

30 Steady State Subtasks Task # Task Title Sub Task Owner 1. Neutronics BenchmarkTask Lead T. Downar, UM 1.1 Steady State (SS) Survey candidate problems T. Downar, UM Preliminary SS modeling of candidate problems T. Downar, UM Down select to two problems for benchmark evaluation T. Downar, UM SS modeling with deterministic U.S. NRC codes PARCS/AGREE T. Downar, UM SS modeling with deterministic NEAMS code PROTEUS C. Lee, ANL SS modeling with Monte Carlo code OPENMC K. Sun, MIT Comparison of experimental data & model results T. Downar, UM Benchmark level evaluation of selected problems T. Downar, UM Evaluation of uncertainties in selected problems T. Downar, UM Preparation of IRPhEP documentation T. Downar, UM Submission of SS benchmark for peer review T. Downar, UM

31 Uncertainty Analysis Current simulation capabilities are utilized to propagate the effects of known uncertainties in measured data and assess their impact upon the derived models and benchmark specifications. Where measurement uncertainties are unavailable, reasonable estimates are derived using available data and best engineering judgment. The principal tool for our uncertainty analysis will be the ORNL TSUNAMI code which is part of the SCALE code package. The TSUNAMI code stands for Tools for Sensitivity and UNcertainty Analysis Methodology Implementation and was developed at ORNL over the last several decades The U.S. NRC has been funding ORNL and the University of Michigan this past year to apply TSUNAMI to the OECD UAM Benchmark

32 Uncertainty Analysis 14 Unique lattice models 8 fuel lattices Fuel lattices: 2G XS for transport, absorption, kappa-fission, nufission, UQ performed based on stochastic sampling of nuclear data with SCALE/Sampler 500 sets of stochastic samples total 7000 Polaris lattice calculations 500 PARCS core calculations