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DOE OSTI2015

Significance of Dynamic and Transient Analysis in the Design and Operation of Hybrid Energy Systems

Energy systems were historically designed and operated with a specific energy conversion objective, while managing loads and resources. In the recent years, the increased utilization of non-dispatchable renewable sources such as wind and solar has played a role in power quality and the reliability of power systems. In order to mitigate the risk associated with the non-dispatchable resources an integrated approach, such as Hybrid Energy Systems (HES), has to be taken, integrating the loads and resource management between the traditional thermal power plants and the non-dispatchable resources. As our electric energy becomes more diverse in its generation resources, the HES with its operational control system, its real-time view and its dynamic decisions making will become an essential part of the integrated energy systems and improve the overall grid reliability. The operational constraints of the energy sources on both the thermal power plants and the non-dispatchable resources in HES, plays a vital role in the planning and design stage. It is an established fact that the choice of energy source depends on the available natural resources and possible infrastructure. A critical component of decision-making depends on the complementary nature and controllability of the energy sources to supply the load demands with high reliability. Controllability of complex HES to achieve desired performance and flexibility is implemented via coordinated control systems while simultaneously generating electricity and other useful products such as useful heat or hydrogen. These systems are based on instrumentation, signal processing, control theory, and engineering system design. The entire HES along with the control systems are characterized by widely varying time constants. Hence, for a well-coordinated control and operation, we propose physics based modeling of the subsystems to assist in a dynamic and transient analysis. Dynamic and transient analysis in real and non-real time domain of the potential sources is capable of providing satisfactory results at the feasibility analysis stage. A simplified HES consisting of complementary energy sources is used to demonstrate the proposed approach. Suitable real time simulations to perform dynamic and transient analysis will be performed to verify results from the non-real time analysis. The role of such analyses, especially the real time dynamic and transient analysis for manual controls is elaborated. It is also discussed that faster than real time simulation will pave the future for controller rooms and hence manual control procedures. Multiple tools such as Real Time Simulator, Matlab®, Simulink®, PSCAD®, MOOSE, etc. are expected to be used for the HES analysis.

24 POWER TRANSMISSION AND DISTRIBUTION↗

DOE OSTI2015

Assessment of Pellet-Clad Interaction Indicators in Watts Bar Unit 1 using the VERA Framework

The Consortium for Advanced Simulation of Light Water Reactors (CASL) aims to provide high-fidelity, multiphysics simulations of light water reactors (LWRs) by coupling a variety of codes within the Virtual Environment for Reactor Analysis (VERA). One of the primary goals of CASL is to predict local cladding failure through pellet-clad interaction (PCI). This capability is currently being pursued through several different approaches, such as with Tiamat, which is a simulation tool within VERA that more tightly couples the MPACT neutron transport solver, the CTF thermal hydraulics solver, and the MOOSE-based Bison-CASL fuel performance code. However, the process in this paper focuses on running fuel performance calculations with Bison-CASL to predict PCI using the multicycle output data from coupled neutron transport/thermal hydraulics simulations. In recent work within CASL, Watts Bar Unit 1 has been simulated over 12 cycles using the VERA core simulator capability based on MPACT and CTF. Using the output from these simulations, Bison-CASL results can be obtained without rerunning all 12 cycles, while providing some insight into PCI indicators. Multi-cycle Bison-CASL results are presented and compared against results from the FRAPCON fuel performance code. There are several quantities of interest in considering PCI and subsequent fuel rod failures, such as the clad hoop stress and maximum centerline fuel temperature, particularly as a function of time. Bison-CASL performs single-rod simulations using representative power and temperature distributions, providing high-resolution results for these and a number of other quantities. This will assist in identifying fuels rods as potential failure locations for use in further analyses.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2015

LIGHT WATER REACTOR ACCIDENT TOLERANT FUELS IRRADIATION TESTING

The purpose of Accident Tolerant Fuels (ATF) experiments is to test novel fuel and cladding concepts designed to replace the current zirconium alloy uranium dioxide (UO2) fuel system. The objective of this Research and Development (R&D) is to develop novel ATF concepts that will be able to withstand loss of active cooling in the reactor core for a considerably longer time period than the current fuel system while maintaining or improving the fuel performance during normal operations, operational transients, design basis, and beyond design basis events. It was necessary to design, analyze, and fabricate drop-in capsules to meet the requirements for testing under prototypic LWR temperatures in Idaho National Laboratory's Advanced Test Reactor (ATR). Three industry led teams and one DOE team from Oak Ridge National Laboratory provided fuel rodlet samples for their new concepts for ATR insertion in 2015. As-built projected temperature calculations were performed on the ATF capsules using the BISON fuel performance code. BISON is an application of INL’s Multi-physics Object Oriented Simulation Environment (MOOSE), which is a massively parallel finite element based framework used to solve systems of fully coupled nonlinear partial differential equations. Both 2D and 3D models were set up to examine cladding and fuel performance.

Accident Tolerant Fuels↗

DOE OSTI2015

MODELING OF THE GROUNDWATER TRANSPORT AROUND A DEEP BOREHOLE NUCLEAR WASTE REPOSITORY

The concept of disposal of high-level nuclear waste in deep boreholes drilled into crystalline bedrock is gaining renewed interest and consideration as a viable mined repository alternative. A large amount of work on conceptual borehole design and preliminary performance assessment has been performed by researchers at MIT, Sandia National Laboratories, SKB (Sweden), and others. Much of this work relied on analytical derivations or, in a few cases, on weakly coupled models of heat, water, and radionuclide transport in the rock. Detailed numerical models are necessary to account for the large heterogeneity of properties (e.g., permeability and salinity vs. depth, diffusion coefficients, etc.) that would be observed at potential borehole disposal sites. A derivation of the FALCON code (Fracturing And Liquid CONvection) was used for the thermal-hydrologic modeling. This code solves the transport equations in porous media in a fully coupled way. The application leverages the flexibility and strengths of the MOOSE framework, developed by Idaho National Laboratory. The current version simulates heat, fluid, and chemical species transport in a fully coupled way allowing the rigorous evaluation of candidate repository site performance. This paper mostly focuses on the modeling of a deep borehole repository under realistic conditions, including modeling of a finite array of boreholes surrounded by undisturbed rock. The decay heat generated by the canisters diffuses into the host rock. Water heating can potentially lead to convection on the scale of thousands of years after the emplacement of the fuel. This convection is tightly coupled to the transport of the dissolved salt, which can suppress convection and reduce the release of the radioactive materials to the aquifer. The purpose of this work has been to evaluate the importance of the borehole array spacing and find the conditions under which convective transport can be ruled out as a radionuclide transport mechanism. Preliminary results show that modeling of the borehole array, including the surrounding rock, predicts convective flow in the system with physical velocities of the order of 10-5 km/yr over 105 years. This results in an escape length on the order of kilometers, which is comparable to the repository depth. However, a correct account of the salinity effects reduces convection velocity and escape length of the radionuclides from the repository.

Modeling↗

DOE OSTI2015

Fuel Performance Calculations for FeCrAl Cladding in BWRs

This study expands upon previous neutronics analyses of the reactivity impact of alternate cladding concepts in boiling water reactor (BWR) cores and directs focus toward contrasting fuel performance characteristics of FeCrAl cladding against those of traditional Zircaloy. Using neutronics results from a modern version of the 3D nodal simulator NESTLE, linear power histories were generated and supplied to the BISON-CASL code for fuel performance evaluations. BISON-CASL (formerly Peregrine) expands on material libraries implemented in the BISON fuel performance code and the MOOSE framework by providing proprietary material data. By creating material libraries for Zircaloy and FeCrAl cladding, the thermomechanical behavior of the fuel rod (e.g., strains, centerline fuel temperature, and time to gap closure) were investigated and contrasted.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2014

Physics Integration KErnels (PIKE)

Pike is a software library for coupling and solving multiphysics applications. It provides basic interfaces and utilities for performing code-to-code coupling. It provides simple "black-box" Picard iteration methods for solving the coupled system of equations including Jacobi and Gauss-Seidel solvers. Pike was developed originally to couple neutronics and thermal fluids codes to simulate a light water nuclear reactor for the Consortium for Simulation of Light-water Reactors (CASL) DOE Energy Innovation Hub. The Pike library contains no physics and just provides interfaces and utilities for coupling codes. It will be released open source under a BSD license as part of the Trilinos solver framework (trilinos.org) which is also BSD. This code provides capabilities similar to other open source multiphysics coupling libraries such as LIME, AMP, and MOOSE.

Pawlowski, Roger↗

DOE OSTI2016

Determination of Experimental Fuel Rod Parameters using 3D Modelling of PCMI with MPS Defect

An in-reactor experiment is being designed in order to validate the pellet-cladding mechanical interaction (PCMI) behavior of the BISON fuel performance code. The experimental parameters for the test rod being placed in the Halden Research Reactor are being determined using BISON simulations. The 3D model includes a missing pellet surface (MPS) defect to generate large local cladding deformations, which should be measureable after typical burnup times. The BISON fuel performance code is being developed at Idaho National Laboratory (INL) and is built on the Multiphysics Object-Oriented Simulation Environment (MOOSE) framework. BISON supports both 2D and 3D finite elements and solves the fully coupled equations for solid mechanics, heat conduction and species diffusion. A number of fuel performance effects are included using models for swelling, densification, creep, relocation and fission gas production & release. In addition, the mechanical and thermal contact between the fuel and cladding is explicitly modelled using a master-slave based contact algorithm. In order to accurately predict PCMI effects, the BISON code includes the relevant physics involved and provides a scalable and robust solution procedure. The depth of the proposed MPS defect is being varied in the BISON model to establish an optimum value for the experiment. The experiment will be interrupted approximately every 6 months to measure cladding radial deformation and provide data to validate BISON. The complete rodlet (~20 discrete pellets) is being simulated using a 180° half symmetry 3D model with MPS defects at two axial locations. In addition, annular pellets will be used at the top and bottom of the pellet stack to allow thermocouples within the rod to measure the fuel centerline temperature. Simulation results will be presented to illustrate the expected PCMI behavior and support the chosen experimental design parameters.

BISON↗

DOE OSTI2016

Analysis of Experimental Fuel Rod Parameters using 3D Modelling of PCMI with MPS Defect

BISON↗

DOE OSTI2016

STEADY STATE MODELING OF THE MINIMUM CRITICAL CORE OF THE TRANSIENT REACTOR TEST FACILITY

With the advent of next generation reactor systems and new fuel designs, the U.S. Department of Energy (DOE) has identified the need for the resumption of transient testing of nuclear fuels. The DOE has decided that the Transient Reactor Test Facility (TREAT) at Idaho National Laboratory (INL) is best suited for future testing. TREAT is a thermal neutron spectrum, air-cooled, nuclear test facility that is designed to test nuclear fuels in transient scenarios. These specific scenarios range from simple temperature transients to full fuel melt accidents. DOE has expressed a desire to develop a simulation capability that will accurately model the experiments before they are irradiated at the facility. It is the aim for this capability to have an emphasis on effective and safe operation while minimizing experimental time and cost. The multi physics platform MOOSE has been selected as the framework for this project. The goals for this work are to identify the fundamental neutronics properties of TREAT and to develop an accurate steady state model for future multiphysics transient simulations. In order to minimize computational cost, the effect of spatial homogenization and angular discretization are investigated. It was found that significant anisotropy is present in TREAT assemblies and to capture this effect, explicit modeling of cooling channels and inter-element gaps is necessary. For this modeling scheme, single element calculations at 293 K gave power distributions with a root mean square difference of 0.076% from those of reference SERPENT calculations. The minimum critical core configuration with identical gap and channel treatment at 293 K resulted in a root mean square, total core, radial power distribution 2.423% different than those of reference SERPENT solutions.

Critical↗

DOE OSTI2016

Least-Squares PN Formulation of the Transport Equation Using Self-Adjoint-Angular-Flux Consistent Boundary Conditions.

In this paper, we study the Least-Squares (LS) PN form of the transport equation compatible with voids in the context of Continuous Finite Element Methods (CFEM).We first deriveweakly imposed boundary conditions which make the LS weak formulation equivalent to the Self-Adjoint Angular Flux (SAAF) variational formulation with a void treatment, in the particular case of constant cross-sections and a uniform mesh. We then implement this method in Rattlesnake with the Multiphysics Object Oriented Simulation Environment (MOOSE) framework using a spherical harmonics (PN) expansion to discretize in angle. We test our implementation using the Method of Manufactured Solutions (MMS) and find the expected convergence behavior both in angle and space. Lastly, we investigate the impact of the global non-conservation of LS by comparing the method with SAAF on a heterogeneous test problem.

Least-Squares↗

DOE OSTI2016

Least-Squares PN Formulation of the Transport Equation Using Self-Adjoint-Angular-Flux Consistent Boundary Conditions

In this paper, we study the Least-Squares (LS) PN form of the transport equation compatible with voids [1] in the context of Continuous Finite Element Methods (CFEM).We first deriveweakly imposed boundary conditions which make the LS weak formulation equivalent to the Self-Adjoint Angular Flux (SAAF) variational formulation with a void treatment [2], in the particular case of constant cross-sections and a uniform mesh. We then implement this method in Rattlesnake with the Multiphysics Object Oriented Simulation Environment (MOOSE) framework [3] using a spherical harmonics (PN) expansion to discretize in angle. We test our implementation using the Method of Manufactured Solutions (MMS) and find the expected convergence behavior both in angle and space. Lastly, we investigate the impact of the global non-conservation of LS by comparing the method with SAAF on a heterogeneous test problem.

97 MATHEMATICS AND COMPUTING↗

DOE OSTI2016

Hybrid PN-SN Calculations with SAAF for the Multiscale Transport Capability in Rattlesnake

Two interface conditions, the Lagrange multiplier method and the upwinding method, for hybrid \pn-\sn calculations is proposed for the self-adjoint angular flux (SAAF) formulation of the transport equation using the continuous finite element method (FEM) for spatial discretization. These interface conditions are implemented in Rattlesnake, the radiation transport application built on MOOSE, for the on-going multiscale transport simulation effort at INL. For smoothing the solution at the interface for the Lagrange multiplier method, a method based on \sn Lagrange interpolation on the sphere is proposed. Numerical results indicate that the interface conditions give the expected convergence.

97 MATHEMATICS AND COMPUTING↗

DOE OSTI2017

Heat Source Characterization in a TREAT Fuel Particle Using Coupled Neutronics-BCMC Calculations

This work presents a multi-physics, multi-scale approach to modeling the Transient Test Reactor (TREAT) currently prepared for restart at the Idaho National Laboratory. TREAT fuel is made up of microscopic fuel grains (r ˜ 20µm) dispersed in a graphite matrix. The novelty of this work is in coupling a binary collision Monte-Carlo (BCMC) model to the Finite Element based code Moose for solving a microsopic heat-conduction problem whose driving source is provided by the BCMC model tracking fission fragment energy deposition. This microscopic model is driven by a transient, engineering scale neutronics model coupled to an adiabatic heating model. The macroscopic model provides local power densities and neutron energy spectra to the microscpic model. Currently, no feedback from the microscopic to the macroscopic model is considered. TREAT transient 15 is used to exemplify the capabilities of the multi-physics, multi-scale model, and it is found that the average fuel grain temperature differs from the average graphite temperature by 80 K despite the low-power transient. The large temperature difference has strong implications on the Doppler feedback a potential LEU TREAT core would see, and it underpins the need for multi-physics, multi-scale modeling of a TREAT LEU core.

36 MATERIALS SCIENCE↗

DOE OSTI2015

Reactor Pressure Vessel Integrity Assessments with the Grizzly Aging Simulation Code

Grizzly is a simulation tool being developed at Idaho National Laboratory (INL) as part of the US Department of Energy’s Light Water Reactor Sustainability program to provide improved safety assessments of systems, components, and structures in nuclear power plants subjected to age-related degradation. Its goal is to provide an improved scientific basis for decisions surrounding license renewal, which would permit operation of commercial nuclear power plants beyond 60 years. Grizzly is based on INL’s MOOSE framework, which enables multiphysics simulations in a parallel computing environment. It will address a wide variety of aging issues in nuclear power plant systems, components, and structures, modelling both the aging processes and the ability of age-degraded components to perform safely. The reactor pressure vessel (RPV) was chosen as the initial application for Grizzly. Grizzly solves tightly coupled equations of heat conduction and solid mechanics to simulate the global response of the RPV to accident conditions, and uses submodels to represent regions with pre-existing flaws. Domain integrals are used to calculate stress intensity factors on those flaws. A physically based empirical model is used to evaluate material embrittlement, and is used to evaluate whether crack growth would occur. Grizzly can represent the RPV in 2D or 3D, allowing it to evaluate effects that require higher dimensionality models to capture. Work is underway to use lower length scale models of material evolution to inform engineering models of embrittlement. This paper demonstrates an application of Grizzly to RPV failure assessment, and summarizes on-going work.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2017

A SUMMARY OF THE THEORETICAL BASIS OF THE RELAP-7 NUCLEAR REACTOR SYSTEM SAFETY ANALYSIS CODE

The RELAP-7 code is the next generation nuclear reactor system safety analysis code being developed at the Idaho National Laboratory (INL)[1]. The code is based on INL’s scientific software development framework, MOOSE (Multi-physics Object Oriented Simulation Environment)[2]. The primary basis of the RELAP-7 governing theory includes 7-equation two-phase flow, reactor core heat transfer, and reactor kinetics models. This paper will focus upon the compressible, fully nonequilibrium two-phase flow model.

7-equation two-phase model↗

DOE OSTI2017

EFFECT OF INCLINED WAVES ON DEEPLY EMBEDDED NUCLEAR FACILITIES

Several conceptual designs for advanced reactors show that most if not all of the structure is embedded in the soil. Consequently, analysis techniques that have been verified and validated for the analysis of light water reactors on surface foundations may not be directly applicable to deeply embedded reactors. Specifically two effects that may change the in-structure response (ISR) of embedded facilities are: (i) the effect of the kinematic interaction becomes more prominent as the embedment depth increases; (ii) non-vertically propagating waves, which are usually not considered for surface-founded facilities, may be more important for deeply embedded facilities. A preliminary study is presented in this article to understand the response of embedded nuclear facilities to non-vertically propagating seismic waves. The inclined wave field is generated from source-to-site simulations using representative earthquake faults with different dip angles. The embedment depth of the nuclear containment facility is also varied to obtain a better understanding of the effect of embedment. This analysis of the soil-structure ensemble is conducted using the application called Multi-hazard Analysis for STOchastic time-DOmaiN phenomena (MASTODON), developed on top of Idaho National Lab’s open source MOOSE finite element framework. The Domain Reduction Method implemented in MASTODON to incorporate the seismic excitation within the region of interest is verified and then used to reduce the computational effort in analyzing this problem. Preliminary analysis using linear soil and structure material properties and tied contact between the soil and structure indicate that embedding the nuclear facility reduces significantly the response of the structure in all the earthquake scenarios considered.

42 ENGINEERING↗

DOE OSTI2018

A Pebble Tracking Transport Algorithm for Pebble Bed Reactor Analysis

Pebble bed reactor (PBR) designs pose unique challenges to modeling and simulation. The challenges originate from the multiple levels of heterogeneity partly due to random packing in and flow through the core of fuel pebbles. This paper proposes a novel pebble tracking transport (PTT) algorithm to address some of the challenges facing PBR analysis. Each individual pebble is tracked and modeled on an unstructured mesh of which the nodes are aligned with the pebble centroids. The neutron transport equation is solved on this pebble tracking mesh with pre-assembled elemental matrices. These elemental matrices are independent of the cross sections and the streaming directions, thus can be assembled before the calculation without introducing too much memory overhead especially when combined with spatial domain decomposition. Numerical results show that these matrices can be assembled efficiently. PTT is implemented in Rattlesnake, the INL MOOSE-based multigroup radiation transport application. Both the angular and spatial convergence are investigated and the results indicate that low order angular and spatial resolution is typically sufficient for analysis. A complete work flow for PBR analysis with PTT including, pebble packing, meshing, cross section generation, solution of the transport equation, postprocessing, etc. is demonstrated with a simplified reactor with 41,048 pebbles.

97 MATHEMATICS AND COMPUTING↗

DOE OSTI2018

SELF-LIMITING TRANSIENT PULSE SIMULATION METHOD EXHIBITING TIME LAG PHENOMENON USING MAMMOTH

Coupling of multi-physics and multi-scale allows the exploration of phenomenon not previously studied with high fidelity. The heat diffusion time lag in TREAT-like fuel requires both multi-physics and multi-scale coupling to model the full transient behavior. The reactor physics code MAMMOTH, which has the ability to couple the heat conduction equation and the neutron diffusion equation, simulated a self-limiting transient pulse meeting these requirements. The MOOSE MultiApp system allowed the coupling of many micro-scale simulations with a single macro-scale simulation. The micro-scale geometry was a 0.0044 cm diameter UO2 fuel grain surrounded by a graphite moderator. The macro-scale geometry was a homogeneous fuel with the same isotope ratio of the micro-scale material surrounded by a graphite reflector. Two temperature-dependent two energy group cross-section libraries were computed using Serpent 2 for use in MAMMOTH. Both the homogeneous unit cell and heterogeneous unit cell cross-section library had infinite reactor boundary conditions. The macro-scale simulation had finite boundary conditions. By varying the feedback temperature, macro or micro, given to the cross-sections, the effect of the time lag was explored using a Picard coupling method. The time lag increases the peak power density by 9 % to 10 % and the energy deposited by 8 % to 9 %. The effect of differing grain and moderator temperature at the micro-scale was also explored showing differences in time lag behavior. Future research will improve the fidelity of the method through representative macro-scale geometries, temperature-dependent cross-sections, and greater energy group resolution.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗