SCALE/DB-Public
This is a collection of public data produced by AMPX or curated from quality sources that the SCALE code system relies on for user problem input.
Brown, Jesse [Oak Ridge National Laboratory (ORNL)↗
Engineering topics
Goluoglu, Sedat
Publications and source records attributed to Goluoglu, Sedat.
SCALE
SCALE v6.3.1 is a beta version built from existing SCALE code. This code provides modeling and simulation of nuclear reactor physics, criticality safety, radiation shielding, and associated spent fuel characterization for nuclear facilities and nuclear fuel transportation and storage package designs (please see attachment for more details).
Ade, Brian↗
CAAS-3S Radiation Testing for Y-12 and UPF with Godiva-IV
Goals: 1) Qualify the CAAS-3S system to a mixed-field radiation dose and dose rate, and 2) Extend the Y-12 Shielding MCNP Validation to rad-si. This paper represents the efforts of Y-12 and UPF personnel, and their interpretation of the test results.
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
CAAS-3S Radiation Testing for Y-12 and UPF with Godiva-IV
The Y-12 National Security Complex and the Uranium Processing Facility (UPF) selected the Mirion CAAS-3S as the Criticality Accident Alarm System for UPF and for Y-12 facilities replacing their legacy CAAS as part of efforts to extend the facility lifespans. As part of this process, the CAAS-3S system was exposed to a high radiation dose and dose rate during reactor testing with the Godiva-IV fast burst reactor. The reactor testing was designed around preliminary analyses that determined Y-12 and UPF requirements, and simulations of the radiation field within the reactor facility were used to determine reactor operating parameters, CAAS equipment locations, and the design of a neutron shield wall. This paper presents the design and results of the testing, and discusses how the test results were interpreted by criticality safety engineers at Y-12 and UPF.
46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
Development of Reactor Physics Benchmark Evaluations for Power Burst Facility Experiments
The PBF is an open vessel with a closed loop water cooling system that can handle up to 28MW of heat generation during steady state operations. This project details the facility characteristics relevant to performing simulation benchmarks for the intent of inclusion in the International Reactor Physics Experiment Evaluation Project (IRPhEP).
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Architecture Improvements and Retesting of the Microscopic Depletion Implementation in the MAMMOTH Reactor Physics Application
Nuclear fuel experiments conducted at the Transient Reactor Test Facility (TREAT) will be analyzed with the MAMMOTH reactor physics application, currently under development at Idaho National Laboratory. MAMMOTH natively couples with the BISON, RELAP-7, and Rattlesnake applications within the MOOSE framework. The implemented methods allow for simulating the irradiation of fuel in a nuclear reactor from the beginning of life in a nuclear reactor until it is placed in TREAT for fuel testing using the same finite element mesh throughout the analysis retaining a very high level of resolution and fidelity. The calculation of the isotopic distribution in fuel requires the solution to the decay and transmutation equations coupled to the neutron transport equation. This report focuses on three significant improvements to the MAMMOTH micro-depletion capability. First, the depletion architecture within MAMMOTH is modified to improve the flexibility and speed of burnup calculations. The results of various benchmark tests demonstrate that the refactoring of the previously implemented capability does not introduce any regression. The second improvement involves the expansion of the data functionality. MAMMOTH can now read the decay data files used by the Oak Ridge Isotope Generation (ORIGEN) code, thus enabling access to data of more than 1,600 isotopes. The third improvement is the capability to perform depletion at a constant power level which was verified via a comparison to ORIGEN.
73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
Implementation of Depletion Architecture in the MAMMOTH Reactor Physics Application
Goal is to add functionality to MAMMOTH to perform depletion calculations. The purpose of this is to allow the use of the same mesh for fuel elements depleted in commercial reactors or the Advanced Test Reactor (ATR) before being placed in the Transient Reactor Test Facility (TREAT) for further experimentation.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
On the effects of pre- and post-transient rod positions for TREAT temperature-limited transient powers
Not Available
Goluoglu, Sedat↗
Enhanced geometric capabilities for the transient analysis code T-ReX and its application to simulating TREAT experiments
Here, advances in computational architecture have prompted a resurgence in the simulation of reactor transients from first principles. Most codes are unable to simulate transient events with complex models, and require numerous approximations. The code T-ReX (Transient-Reactor eXperiment simulator), an extensive update to TDKENO, has been developed as a transient analysis tool with few geometric limitations and minimal theoretical approximations. T-ReX achieves this by employing the Improved Quasi-Static (IQS) method to solve the time dependent Boltzmann transport equation with explicit representation of delayed neutrons. The primary change in T-ReX relative to TDKENO is the incorporation of a modified version of the Monte Carlo code KENO-VI to calculate the flux shape and model the geometry of a problem. Using KENO-VI to model systems allows exact representation of the geometry. The changes to T-ReX are verified by comparison of solutions to computational benchmark problems found with a previous version of TDKENO that made use of KENO V.a, and several other codes with time-dependent capabilities. In addition, a three-dimensional model of the Transient Reactor Test Facility (TREAT) core at the Idaho National Laboratory (INL) is constructed with KENO-VI, and used to validate T ReX. T-ReX produces results that agree with benchmark problems, and are in better agreement with TREAT experimental data than TDKENO.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Effects of Boron and Graphite Uncertainty in Fuel for TREAT Simulations
As a result of the accident at Fukushima, the United States has made a concerted effort to develop fuel with enhanced accident tolerance for the current fleet of reactors. In 2012 Congress laid out a schedule for the development of accident tolerant fuel (ATF), the goal of which is to test in a commercial reactor by 2022. Idaho National Lab's (INL) Transient Reactor Test Facility (TREAT) is one of the locations that will be performing ATF testing and is scheduled for restart in 2018. TREAT was built to subject fuel to conditions ranging from mild transients to severe reactor accidents. However, typically a number of pre-test calibrations are required before evaluation of fuel can begin, which is costly and time consuming. Advanced modeling techniques and current computational capacity make full core TREAT simulations possible, with the goal of such simulations to understand the pre-test core and minimize the number of required calibrations. But, in order to simulate TREAT with a high degree of precision the reactor materials and geometry must also be modeled with a high degree of precision. This paper examines how uncertainty in the reported values of boron and graphite have an effect on simulations of TREAT.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Optimization of a Monte Carlo Model of the Transient Reactor Test Facility
The ultimate goal of modeling and simulation is to obtain reasonable answers to problems that don't have representations which can be easily evaluated while minimizing the amount of computational resources. With the advances during the last twenty years of large scale computing centers, researchers have had the ability to create a multitude of tools to minimize the number of approximations necessary when modeling a system. The tremendous power of these centers requires the user to possess an immense amount of knowledge to optimize the models for accuracy and efficiency. Research reactors require that the neutron flux be well characterized so that the flux in a given location or the power delivered to the target in an experiment center can be accurately determined.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Effects of Boron and Graphite Uncertainty in Fuel for TREAT Simulations
Advanced modeling techniques and current computational capacity make full core TREAT simulations possible, with the goal of such simulations to understand the pre-test core and minimize the number of required calibrations. But, in order to simulate TREAT with a high degree of precision the reactor materials and geometry must also be modeled with a high degree of precision. This paper examines how uncertainty in the reported values of boron and graphite have an effect on simulations of TREAT.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Optimization of a Monte Carlo Model of the Transient Reactor Test Facility
The ultimate goal of modeling and simulation is to obtain reasonable answers to problems that don’t have representations which can be easily evaluated while minimizing the amount of computational resources. With the advances during the last twenty years of large scale computing centers, researchers have had the ability to create a multitude of tools to minimize the number of approximations necessary when modeling a system. The tremendous power of these centers requires the user to possess an immense amount of knowledge to optimize the models for accuracy and efficiency.This paper seeks to evaluate the KENO model of TREAT to optimize calculational efforts.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Preliminary Results for the OECD/NEA Time Dependent Benchmark using Rattlesnake, Rattlesnake-IQS and TDKENO
One goal of the MAMMOTH M&S project is to validate the analysis capabilities within MAMMOTH. Historical data has shown limited value for validation of full three-dimensional (3D) multi-physics methods. Initial analysis considered the TREAT startup minimum critical core and one of the startup transient tests. At present, validation is focusing on measurements taken during the M8CAL test calibration series. These exercises will valuable in preliminary assessment of the ability of MAMMOTH to perform coupled multi-physics calculations; calculations performed to date are being used to validate the neutron transport solver Rattlesnake\cite{Rattlesnake} and the fuels performance code BISON. Other validation projects outside of TREAT are available for single-physics benchmarking. Because the transient solution capability of Rattlesnake is one of the key attributes that makes it unique for TREAT transient simulations, validation of the transient solution of Rattlesnake using other time dependent kinetics benchmarks has considerable value. The Nuclear Energy Agency (NEA) of the Organization for Economic Cooperation and Development (OECD) has recently developed a computational benchmark for transient simulations. This benchmark considered both two-dimensional (2D) and 3D configurations for a total number of 26 different transients. All are negative reactivity insertions, typically returning to the critical state after some time.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
A Parametric Study of the Adjoint Flux Calculation for TDKENO
TDKENO is a time-dependent transport code designed to simulate reactor transients. One application is to study experiments performed at the Transient Reactor Test Facility (TREAT) at the Idaho National Laboratory (INL). TDKENO solves the time-dependent, three-dimensional Boltzmann transport equation with explicit representation of delayed neutrons. Instead of directly integrating this equation, the Improved Quasi Static (IQS) method is employed. This involves factoring the neutron flux into two components - a rapidly varying (in time) amplitude equation and a slowly varying shape equation. These are solved separately on different time scales. The shape equation is solved using the Monte Carlo transport code KENO, from Oak Ridge National Laboratory's SCALE code package. Using the Monte Carlo method to solve the shape equation is computationally intensive, but the operation is only performed when needed - namely when the reactor system is undergoing a significant change. The amplitude equation is solved deterministically. This operation contains the time dependence and can be calculated relatively quickly, resulting in computational savings. TDKENO begins by solving the adjoint form of the transport equation for use as a weighting function. It uses this adjoint flux for multiple aspects of the simulation, including the calculation of all of the point kinetics parameters . The flux shape is found by solving a modified form of the transport equation in 'forward' mode. This paper examines the effect the adjoint weighting function has on the solution. (authors)
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Parallel Performance of the Time Dependent Transport Code TDKENO with Applications to TREAT Simulations
Renewed interest in high fidelity simulation of excursion events has prompted the improvement of codes that solve the time-dependent Boltzmann transport equation. Often these codes make approximations to the transport equation in order to achieve results in a reasonable amount of time. One method, the improved quasi-static (IQS), makes few approximations compared to adiabatic, diffusion, and point kinetics. This is the methodology employed by the code TDKENO. One downside to this approach is a portion of the calculation (flux shape) is done with a Monte Carlo code and must be called upon several times. To minimize run time the flux shape calculation is solved with a modified version of KENO-VI from Scale 6.2. This version of KENO runs in parallel across hundreds of nodes, thus improving performance. The parallel capabilities of KENO allows TDKENO to simulate complex transient experiments with a large number of histories. These improvements are applied to the simulation of TREAT calibration experiments. (authors)
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Verification of the Time-dependent Transport Code TDKENO with the Monte Carlo Code KENO-VI
In this paper we focus on the computer code TDKENO that employs a hybrid methodology for solving the time dependent, three-dimensional Boltzmann transport equation. Originally, TDKENO used the Monte Carlo Transport code KENO V.a from SCALE to calculate flux shapes and fixed sources. KENO V.a is quick to compute these quantities but is limited to simple geometries. For this work TDKENO has been modified to support the Monte Carlo Transport code KENO-VI in addition to KENO V.a. KENO-VI allows generalized geometry input not possible in KENO V.a inputs. In order to verify the changes we ran TDKENO using KENO V.a and KENO-VI inputs then compared the results. Three different inputs were analyzed, the 16a1 benchmark,a simple TREAT model, and a transient model (2855) from the M8 calibration tests performed at INL.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Motivation for Using Generalized Geometry in the Time Dependent Transport Code TDKENO
We are proposing to use the code, TDKENO, to model TREAT. TDKENO solves the time dependent, three dimensional Boltzmann transport equation with explicit representation of delayed neutrons. Instead of directly integrating this equation, the neutron flux is factored into two components – a rapidly varying amplitude equation and a slowly varying shape equation and each is solved separately on different time scales. The shape equation is solved using the 3D Monte Carlo transport code KENO, from Oak Ridge National Laboratory’s SCALE code package. Using the Monte Carlo method to solve the shape equation is still computationally intensive, but the operation is only performed when needed. The amplitude equation is solved deterministically and frequently, so the solution gives an accurate time-dependent solution without having to repeatedly We have modified TDKENO to incorporate KENO-VI so that we may accurately represent the geometries within TREAT. This paper explains the motivation behind using generalized geometry, and provides the results of our modifications. TDKENO uses the Improved Quasi-Static method to accomplish this. In this method, the neutron flux is factored into two components. One component is a purely time-dependent and rapidly varying amplitude function, which is solved deterministically and very frequently (small time steps). The other is a slowly varying flux shape function that weakly depends on time and is only solved when needed (significantly larger time steps).
IQS↗