Consortium for the Advanced Simulation of Light Water Reactors

Current Monte Carlo neutron transport applications use continuous energy cross section data to provide the statistical foundation for particle trajectories. This "classical" algorithm requires storage and random access of very large data structures. Recently, Forget et al.[1] reported on a fundamentally new approach, based on multipole expansions, that distills cross section data down to a more abstract mathematical format. Their formulation greatly reduces memory storage and improves data locality at the cost of also increasing floating point computation. In the present study we determine the hardware performance parameters, including power usage, of the multipole algorithm relative to the classical continuous energy algorithm. This study is done to guage the suitability of both algorithms for use on next-generation high performance computing platforms.

The purpose of this work is to provide improved VUQ capability for CTF. We will initially focus on uncertainty quantification, but we will keep the larger scope of SQA, verification, validation, calibration, and optimization in mind while we work on uncertainty quantification. This is a living document that will be amended as the work is completed. Note that any reference to the CTF software is based on the code version of 12/17/13.

The milestone for the operational reactor depletion demonstrates a cycle depletion simulation of Watt's Bar Unit 1 Cycle 1.The simulation was performed using the VERA-CS components MPACT and COBRA-TF.The simulation was performed on the Titan and Eos machines at the Oak Ridge Leadership Computing Facility.The simulation results include the critical boron concentration, the pin resolved power, temperature, and fluid density distributions, and detector responses as a function of time.These results are presented and the critical boron concentration is compared with measured data.The comparison is reasonable given the modeling approximation used at this point.Several suggestions, both near and longer term, are given for areas of future work which include improving the methods, data, models, and computational performance.For those solution results which are not compared to measurement, e.g.power and temperature distributions, their behavior coincides with analysts' expectations.An important next step in the validation of VERA-CS will be the comparison of the computational results to the measured flux data.To satisfy this L1 milestone, numerous capabilities were successfully added as a part of supporting L2 and L3 milestones.These include: full core depletion, generation of a new cross section library, coupling of neutronics and thermal hydraulic feedback, and detector response modeling.In addition to the operational reactor depletion, several 2-D core parametric studies were performed to assess the new simulation capability and to develop guidelines on model approximations and discretizations.Some of the conclusions from these studies are:The simulation of B-10 depletion in the coolant is necessary to accurately predict critical boron concentrations If the end of cycle state is of primary interest in the simulation, then significant approximations may be made in the cycle history.The ability to make use of this approximation without a notable loss in accuracy will considerably accelerate multi-cycle analysis.The new 47-group cross section library with a transport correction of the P 0 scattering can provide solutions very efficiently and which are nearly as accurate as solutions generated with explicit P 2 scattering.Simplified modeling of the equilibrium xenon concentration is well predicted by the existing depletion solver, and only in certain circumstances (e.g.large and "instantaneous" changes to rod position or T/H conditions) would an explicit physical model need to be included.Overall the milestone is considered a success as the objectives were met and several new and important results were produced from this work.The results and conclusions of this milestone will continue to guide future work and become the basis for addressing the CASL Challenge Problems.

The objectives of this milestone work are to assess the improved capabilities of the multi-scale thermal-hydraulic (T/H) codes and models developed by the Consortium for Advanced Simulation of Light Water Reactors (CASL) for Departure from Nucleate Boiling (DNB) Challenge Problem (CP), in accordance with the DNB CP implementation plan.The multi-scale models can range from fine mesh Computational Fluid Dynamics (CFD) simulation of flow field surrounding a fuel rod to a full core modeling of a Pressurized Water Reactor (PWR).The assessments are performed based on the CASL subchannel code COBRA-TF (CTF) and the CFD code Hydra-TH.Significant improvements have been achieved on the CASL multi-scale T/H code and modeling capabilities based on the CTF subchannel code and the Hydra-TH CFD code in the past year for DNB CP applications.The improvements are reflected in the new transient modeling for the Reactivity Insertion Accident (RIA) DNB predictions and full assembly modeling using the CTF code and its processor, as well as the rod bundle modeling for single-phase flow and heat transfer simulations using the Hydra-TH code, the pre-processor for mesh generation and the post-processor for data visualization.The following code capabilities have been demonstrated in the assessment:-The modeling and simulation of the TK experiments demonstrated that CTF is able to simulate a fast transient with a large power pulse.-The CTF preprocessor was found to be very helpful in greatly reducing the model creation effort and minimizing human error for large model setup such as a subchannel model for the entire 17x17 fuel assembly.A reasonably fast CTF execution time can be achieved with the Krylov solver with the large model.-The modeling and simulation of the 3x3 subchannel geometry under single phase flow conditions have been successfully completed using the Hydra-TH CFD code including mesh generation and result visualization.The Hydra-TH results indicate the similar capabilities of the subchannel single phase fluid solutions as compared to the STAR-CCM+ results.The code and model assessments also indicated additional improvements needed for the planned DNB CP applications.Recommendations on code-specific improvements are listed in Section 6 of the report.

This milestone demonstrates integration of the MOOSE-based Peregrine fuel performance code into the CASL Virtual Environment for Reactor Analysis (VERA).Specifically targeted at VERA's core simulator functionality (VERA-CS), the existing VERA subchannel-neutronics capability based on COBRA-TF (CTF) and Insilico was extended to use the axisymmetric 2D R-Z fuel rod modeling functionality of PEREGINE to replace the simple fuel rod model in CTF.This three-way coupling involved defining and implementing data interfaces, software infrastructure development to support the requirements of each code, and numerical methods to implement the coupled physics.The result is a new VERA capability known as Tiamat.Development of Tiamat required extensions to the VERA build system to support a "meta build" that uses the MOOSE/libMesh/Peregrine native build system, development of a new driver layer designed to support long-term CASL code integration requirements, and implementation of data transfers for the coupled applications using the DataTransferKit (DTK).Peregrine is now compiled and tested under the VERA continuous integration server.Tiamat, a new advanced simulation tool consisting of three coupled codes -CTF for multiphase subchannel flow, Insilico for neutronics and Peregrine for fuels performance -has been developed.Tiamat was used to simulate a 17x17 assembly model for Watts Bar Unit 1 Cycle 1, which exceeds the goals of this milestone.