Mark D. DeHart

Validation of SCALE-4 for Burnup Credit Applications

In the past, criticality analysis of pressurized water reactor (PWR) fuel stored in racks and casks has assumed that the fuel is fresh with the maximum allowable initial enrichment. If credit is allowed for fuel burnup in the design of casks that are used in the transport of spent light water reactor fuel to a repository, the increase in payload can lead to a significant reduction in the cost of transport and a potential reduction in the risk to the public. A portion of the work has been performed at Oak Ridge National Laboratory (ORNL) in support of the U.S. Department of Energy (DOE) efforts to demonstrate a validation approach of criticality safety methods to be used in burnup credit cask design. The date, the SCALE code system developed at ORNL has been the primary computational tool used by DOE to investigate technical issues related to burnup credit. The SCALE code package is a well-established code system that has been widely used in away from reactor applications. Criticality safety analyses are performed via the criticality safety analysis sequences (CSAS) and spent-fuel characterization via the shielding analysis sequence (QSAS) and spent-fuel characterization via the shielding analysis sequence (SAS2H). The SCALE 27-group burnup library containing ENDF/B-IV (actinides) and ENDF/B-V (fission products) data has been used for all calculations. The American National Standards Institute/American Nuclear Society (ANSI/ANS)-8.1 criticality safety standard requires validation and benchmarking of the calculational methods used in evaluating criticality safety limits for applications outside reactors of correlation against critical experiments that are applicable. Numerous critical experiments for fresh PWR-type fuel in storage and transport configurations exist and can be used as part of a validation database. However, there are no critical experiments with burned PWR-type fuel in storage and transport configurations. As an alternative, commercial reactors offer an excellent source of measured critical configurations. Part of the work that has been performed to date to validate the SCALE-4 code system for burnup credit applications using measured critical configurations includes: 1. fresh fuel critical experiments having geometric and nuclear characteristics similar to PWR spent fuel in storage and transport configurations 2. commercial PWR hot-zero-power and hot-full-power reactor critical configurations. The ability to closely predict reactor critical conditions is important in the validation of a methodology for spent-fuel applications because input data are determined based on relatively little detail of reactor core operation. Such limited information is expected to be representative of data available when burnup credit calculations are being performed in the determination of optimum cask loadings. The results reported demonstrate the ability of the ORNL SCALE-4 methodology to predict a value of k eff very close to the known value of 1.0, both for fresh fuel criticals and for the more complex reactor criticals. Beyond these results, additional work in the determination of biases and uncertainties is necessary prior to use in burnup credit applications

A flexible nonlinear diffusion acceleration method for the SN transport equations discretized with discontinuous finite elements

Abstract This work presents a flexible nonlinear diffusion acceleration (NDA) method that discretizes both the S N transport equation and the diffusion equation using the discontinuous finite element method (DFEM). The method is flexible in that the diffusion equation can be discretized on a coarser mesh with the only restriction that it is nested within the transport mesh and the FEM shape function orders of the two equations can be different. The consistency of the transport and diffusion solutions at convergence is defined by using a projection operator mapping the transport into the diffusion FEM space. The diffusion weak form is based on the modified incomplete interior penalty (MIP) diffusion DFEM discretization that is extended by volumetric drift, interior face, and boundary closure terms. In contrast to commonly used coarse mesh finite difference (CMFD) methods, the presented NDA method uses a full FEM discretized diffusion equation for acceleration. Suitable projection and prolongation operators arise naturally from the FEM framework. Via Fourier analysis and numerical experiments for a one-group, fixed source problem the following properties of the NDA method are established for structured quadrilateral meshes: (1) the presented method is unconditionally stable and effective in the presence of mild material heterogeneities if the same mesh and identical shape functions either of the bilinear or biquadratic type are used, (2) the NDA method remains unconditionally stable in the presence of strong heterogeneities, (3) the NDA method with bilinear elements extends the range of effectiveness and stability by a factor of two when compared to CMFD if a coarser diffusion mesh is selected. In addition, the method is tested for solving the C5G7 multigroup, eigenvalue problem using coarse and fine mesh acceleration. While NDA does not offer an advantage over CMFD for fine mesh acceleration, it reduces the iteration count required for convergence by almost a factor of two in the case of coarse mesh acceleration.

Reactor Physics Scoping and Characterization Study on Implementation of TRIGA Fuel in the Advanced Test Reactor

The Advanced Test Reactor (ATR) is conducting scoping studies for the conversion of its fuel from a highly enriched uranium (HEU) composition to a low-enriched uranium (LEU) composition, through the Reduced Enrichment for Research and Test Reactors Program, within the Global Threat Reduction Initiative. These studies have considered a wide variety of LEU plate-type fuels to replace the current HEU fuel. Continuing to investigate potential alternatives to the present HEU fuel form, this study presents a preliminary reactor physics scoping and feasibility analysis of TRIGA fuel within the current ATR fuel element envelope and compares it to the functional requirements delineated by the Naval Reactors Program, which includes >4.8×1014 fissions/s·g−1 of 235U in test positions, a fast–to–thermal neutron flux ratio that has a <5% deviation from its current value, a desired steady cycle power within the corner lobes, and an operational cycle length of 56 days at 120 MW. Other design parameters outside those put forth by the Naval Reactors Program that are investigated herein include axial and radial power profiles, effective delayed neutron fraction, and mean neutron generation time. The result of this study demonstrates potential promise for implementation of TRIGA fuel in the ATR from a reactor physics perspective; discussion of observations and limitations are provided herein.

Evaluation of the Enhanced LEU Fuel (ELF) Design for Conversion of the Advanced Test Reactor to a Low-Enrichment Fuel Cycle

Abstract A conceptual low-enrichment uranium (LEU) fuel design has been developed for the Advanced Test Reactor (ATR) at Idaho National Laboratory. The ATR is currently fueled with a high-enrichment fuel but is slated to be converted to LEU under programs led by the National Nuclear Security Administration of the U.S. Department of Energy. A conceptual LEU fuel design, the Enhanced LEU Fuel (ELF), has been developed assuming power peaking control through the use of variable fuel meat thicknesses and no use of burnable poison. In initial work, this design was shown to satisfy performance requirements for ATR operation. Following these design calculations, a safety analysis process was initiated to demonstrate that the ELF design would successfully meet safety limits for postulated accident conditions. Those calculations, performed using RELAP5 and ATR-SINDA, require physics analysis to provide spatial power distributions and kinetics parameters for various core operations configurations. This article describes the findings of the physics analysis and provides predictions for the behavior of a LEU-fueled version of ATR, and compares these to calculations of the performance of the current high-enrichment uranium fuel.

RELAP-7 Development Updates

RELAP-7 is a nuclear systems safety analysis code being developed at the Idaho National Laboratory, and is the next generation tool in the RELAP reactor safety/systems analysis application series. RELAP-7 development began in 2011 to support the Risk Informed Safety Margins Characterization (RISMC) Pathway of the Light Water Reactor Sustainability (LWRS) program. The overall design goal of RELAP-7 is to take advantage of the previous thirty years of advancements in computer architecture, software design, numerical methods, and physical models in order to provide capabilities needed for the RISMC methodology and to support nuclear power safety analysis. The code is being developed based on Idaho National Laboratory’s modern scientific software development framework – MOOSE (the Multi-Physics Object-Oriented Simulation Environment). The initial development goal of the RELAP-7 approach focused primarily on the development of an implicit algorithm capable of strong (nonlinear) coupling of the dependent hydrodynamic variables contained in the 1-D/2-D flow models with the various 0-D system reactor components that compose various boiling water reactor (BWR) and pressurized water reactor nuclear power plants (NPPs). During Fiscal Year (FY) 2015, the RELAP-7 code has been further improved with expanded capability to support boiling water reactor (BWR) and pressurized water reactor NPPs analysis.morexa0» The accumulator model has been developed. The code has also been coupled with other MOOSE-based applications such as neutronics code RattleSnake and fuel performance code BISON to perform multiphysics analysis. A major design requirement for the implicit algorithm in RELAP-7 is that it is capable of second-order discretization accuracy in both space and time, which eliminates the traditional first-order approximation errors. The second-order temporal is achieved by a second-order backward temporal difference, and the one-dimensional second-order accurate spatial discretization is achieved with the Galerkin approximation of Lagrange finite elements. During FY-2015, we have done numerical verification work to verify that the RELAP-7 code indeed achieves 2nd-order accuracy in both time and space for single phase models at the system level.«xa0less