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A Complex-Geometry Validation Experiment for Advanced Neutron Transport Codes

Description: The Idaho National Laboratory (INL) has initiated a focused effort to upgrade legacy computational reactor physics software tools and protocols used for support of core fuel management and experiment management in the Advanced Test Reactor (ATR) and its companion critical facility (ATRC) at the INL.. This will be accomplished through the introduction of modern high-fidelity computational software and protocols, with appropriate new Verification and Validation (V&V) protocols, over the next 12-18 months. Stochastic and deterministic transport theory based reactor physics codes and nuclear data packages that support this effort include MCNP5[1], SCALE/KENO6[2], HELIOS[3], SCALE/NEWT[2], and ATTILA[4]. Furthermore, a capability for sensitivity analysis and uncertainty quantification based on the TSUNAMI[5] system has also been implemented. Finally, we are also evaluating the Serpent[6] and MC21[7] codes, as additional verification tools in the near term as well as for possible applications to full three-dimensional Monte Carlo based fuel management modeling in the longer term. On the experimental side, several new benchmark-quality code validation measurements based on neutron activation spectrometry have been conducted using the ATRC. Results for the first four experiments, focused on neutron spectrum measurements within the Northwest Large In-Pile Tube (NW LIPT) and in the core fuel elements surrounding the NW LIPT and the diametrically opposite Southeast IPT have been reported [8,9]. A fifth, very recent, experiment focused on detailed measurements of the element-to-element core power distribution is summarized here and examples of the use of the measured data for validation of corresponding MCNP5, HELIOS, NEWT, and Serpent computational models using modern least-square adjustment methods are provided.
Date: November 1, 2013
Creator: Nigg, David W.; LaPorta, Anthony W.; Nielsen, Joseph W.; Parry, James; DeHart, Mark D.; Bays, Samuel E. et al.
Partner: UNT Libraries Government Documents Department

Physics: A New Reactor Physics Analysis Toolkit

Description: In the last year INL has internally pursued the development of a new reactor analysis tool: PHISICS. The software is built in a modular approach to simplify the independent development of modules by different teams and future maintenance. Most of the modules at the time of this summary are still under development (time dependent transport driver, depletion, cross section I/O and interpolation, generalized perturbation theory), while the transport solver INSTANT (Intelligent Nodal and Semi-structured Treatment for Advanced Neutron Transport) has already been widely used1, 2, 3, 4. For this reason we will focus mainly on the presentation of the transport solver INSTANT
Date: June 1, 2011
Creator: Rabiti, C.; Wang, Y.; Palmiotti, G.; Hiruta, H.; Cogliati, J. & Alfonsi, A.
Partner: UNT Libraries Government Documents Department

MRDAP User/Developer Documentation

Description: The Multi-Reactor Design and Analysis Platform (MRDAP) is designed to simplify the creation, transfer and processing of data between computational codes. MRDAP accomplishes these objectives with three parts: 1. allows each integrated code, through a plugin, to specify the required input for execution and the required output needed, 2. creates an interface for execution and data transfer, 3. enables the creation of Graphical User Interfaces (GUI) to assist with input preparation and data visualization. Ultimately, the main motivation of this work is to enable analysts (who perform reactor physics calculations routinely), by providing a tool that increases efficiency and minimizes the potential for errors or failed executions.
Date: September 1, 2009
Creator: Cogliati, Joshua & Milvich, Michael
Partner: UNT Libraries Government Documents Department

Deterministic methods for time-dependent stochastic neutron transport

Description: A numerical method is presented for solving the time-dependent survival probability equation in general (lD/2D/3D) geometries using the multi group SNmethod. Although this equation was first formulated by Bell in the early 1960's, it has only been applied to stationary systems (for other than idealized point models) until recently, and detailed descriptions of numerical solution techniques are lacking in the literature. This paper presents such a description and applies it to a dynamic system representative of a figurative criticality accident scenario.
Date: January 1, 2009
Creator: Baker, Randal S
Partner: UNT Libraries Government Documents Department

Verification of Multiphysics software: Space and time convergence studies for nonlinearly coupled applications

Description: High-fidelity modeling of nuclear reactors requires the solution of a nonlinear coupled multi-physics stiff problem with widely varying time and length scales that need to be resolved correctly. A numerical method that converges the implicit nonlinear terms to a small tolerance is often referred to as nonlinearly consistent (or tightly coupled). This nonlinear consistency is still lacking in the vast majority of coupling techniques today. We present a tightly coupled multiphysics framework that tackles this issue and present code-verification and convergence analyses in space and time for several models of nonlinear coupled physics.
Date: May 1, 2009
Creator: Ragusa, Jean C.; Mahadevan, Vijay & Mousseau, Vincent A.
Partner: UNT Libraries Government Documents Department

A Parallel Multigrid Method for Neutronics Applications

Description: The multigrid method has been shown to be the most effective general method for solving the multi-dimensional diffusion equation encountered in neutronics. This being the method of choice, we develop a strategy for implementing the multigrid method on computers of massively parallel architecture. This leads us to strategies for parallelizing the relaxation, contraction (interpolation), and prolongation operators involved in the method. We then compare the efficiency of our parallel multigrid with other parallel methods for solving the diffusion equation on selected problems encountered in reactor physics.
Date: January 1, 2001
Creator: Alcouffe, Raymond E.
Partner: UNT Libraries Government Documents Department

The Covariance and Biocovariance of the Stochartic Neutron Field

Description: The use of stochastic neutron field theory (neutron noise) for the measurement of reactor physics parameters goes back to the early work of Serber, Feynmann, and Orndoff. Since then, a large variety of methods and applications has been developed. In the majority of these methods, some form of modified one-point reactor kinetics was used for the interpretation of the measurements. In fact, the high level of sophistication of the instrumentation used was not matched by the theory. In 1965, Bell developed a general theory of the stochastic neutron field, and in 1987, Munoz-Cobo et al enlarged this treatment to include the effect of the detectors in the neutron field. In both instances, the complexity of the theoretical results were beyond the computing capabilities then available thus, the mismatch between experimental and theoretical methods remained in existence because the powerful Monte-Carlo methods then at work, were only applicable to static neutron fields. This problem was eliminated by the development of a time-dependent Monte-Carlo code specially written by T. E. Valentine for the analysis of stochastic measurements that gave them relevance to the results of the general theory. The purpose of this work is to illustrate the derivation of observables of the stochastic neutron filed from its general treatment.
Date: January 1, 1998
Creator: Perez, R.B.
Partner: UNT Libraries Government Documents Department

The use of active learning strategies in the instruction of Reactor Physics concepts

Description: Each of the Active Learning strategies employed to teach Reactor Physics material has been or promises to be instructionally successful. The Cooperative Group strategy has demonstrated a statistically significant increase in student performance on the unit exam in teaching conceptually difficult, transport and diffusion theory material. However, this result was achieved at the expense of a modest increase in class time. The Tutorial CBI programs have enabled learning equally as well as classroom lectures without the direct intervention of an instructor. Thus, the Tutorials have been successful as homework assignments, releasing classroom time for other instruction. However, the time required for development of these tools was large, on the order of two hundred hours per hour of instruction. The initial introduction of the Case-Based strategy was roughly as effective as the traditional classroom instruction. Case-Based learning could well, after important modifications, perform better than traditional instruction. A larger percentage of the students prefer active learning strategies than prefer traditional lecture presentations. Student preferences for the active strategies were particularly strong when they believed that the strategies helped them learn the material better than they would have by using a lecture format. In some cases, students also preferred the active strategies because they were different from traditional instruction, a change of pace. Some students preferred lectures to CBI instruction, primarily because the CBI did not afford them the opportunity to question the instructor during the presentation.
Date: January 1, 2000
Creator: Robinson, Michael A.
Partner: UNT Libraries Government Documents Department

Interlaboratory Neutron Flux Spectral Measurement Program

Description: This report details an interlaboratory cooperative study undertaken on neutron flux spectral measurements, calculations, and data correlation for reactor irradiation and physics study by various sites. The Testing portion of this study was the irradiation of a wide variety of flux monitors in a special snout capsule holder. The attached sketch shows the capsule holder design, facility design, and the capsule position in relation to adjacent process tubes. A summary of the irradiations performed is provided.
Date: March 7, 1966
Creator: DeMers, A.E.
Partner: UNT Libraries Government Documents Department

DESIGN STUDY FOR A LOW-ENRICHED URANIUM CORE FOR THE HIGH FLUX ISOTOPE REACTOR, ANNUAL REPORT FOR FY 2010

Description: This report documents progress made during FY 2010 in studies of converting the High Flux Isotope Reactor (HFIR) from high enriched uranium (HEU) fuel to low enriched uranium (LEU) fuel. Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum alloy. With axial and radial grading of the fuel foil and an increase in reactor power to 100 MW, calculations indicate that the HFIR can be operated with LEU fuel with no degradation in performance to users from the current level. Studies are reported of support to a thermal hydraulic test loop design, the implementation of finite element, thermal hydraulic analysis capability, and infrastructure tasks at HFIR to upgrade the facility for operation at 100 MW. A discussion of difficulties with preparing a fuel specification for the uranium-molybdenum alloy is provided. Continuing development in the definition of the fuel fabrication process is described.
Date: February 2011
Creator: Cook, David Howard; Freels, James D.; Ilas, Germina; Jolly, Brian C.; Miller, James Henry; Primm, R. Trent, III et al.
Partner: UNT Libraries Government Documents Department

Reactor Physics and Criticality Benchmark Evaluations for Advanced Nuclear Fuel - Final Technical Report

Description: The nuclear industry interest in advanced fuel and reactor design often drives towards fuel with uranium enrichments greater than 5 wt% 235U. Unfortunately, little data exists, in the form of reactor physics and criticality benchmarks, for uranium enrichments ranging between 5 and 10 wt% 235U. The primary purpose of this project is to provide benchmarks for fuel similar to what may be required for advanced light water reactors (LWRs). These experiments will ultimately provide additional information for application to the criticality-safety bases for commercial fuel facilities handling greater than 5 wt% 235U fuel.
Date: September 11, 2008
Creator: Anderson, William; Tulenko, James; Rearden, Bradley & Harms, Gary
Partner: UNT Libraries Government Documents Department

DECOMMISSIONING THE PHYSICS LABORATORY, BUILDING 777-10A, AT THE SAVANNAH RIVER SITE (SRS)

Description: SRS recently completed a four-year mission to decommission {approx}250 excess facilities. As part of that effort, SRS decommissioned a 48,000 ft{sup 2} laboratory that housed four low-power test reactors, formerly used by SRS to determine reactor physics. This paper describes and reviews the decommissioning, with a focus on component segmentation and handling (i.e. hazardous material removal, demolition, and waste handling). The paper is intended to be a resource for engineers, planners, and project managers who face similar decommissioning challenges.
Date: January 17, 2007
Creator: Musall, J & Cathy Sizemore, C
Partner: UNT Libraries Government Documents Department

PARTISN results for the C5G7 MOX benchmark problems

Description: In early 2001 the Nuclear Energy Agency solicited participants for a proposed new benchmark. The benchmark, known as C5G7 MOX, is intended to be a basis to measure current transport code abilities in the treatment of reactor core problems without spatial homogenization. We have participated with the code transport code PARTISN. PARTISN (PARallel TIme Dependent SN), PARTISN solves the linear Boltzmann transport equation in static and time dependent forms on one, two and three dimensional orthogonal grids using the deterministic (SN) method. A variety of spatial discritization methods are incorporated into PARTISN, however all calculations performed here used the diamond difference approach, coupled with a volume fraction method for non-Cartesian problem geometries. Acceleration of the source iterations is accomplished with diffusion synthetic acceleration (DSA).
Date: January 1, 2002
Creator: Dahl, J. A. (Jon A.) & Alcouffe, Raymond E.
Partner: UNT Libraries Government Documents Department

METHODS FOR MODELING THE PACKING OF FUEL ELEMENTS IN PEBBLE BED REACTORS

Description: Two methods for the modeling of the packing of pebbles in the pebble bed reactors are presented and compared. The first method is based on random generation of potential centers for the pebbles, followed by rejection of points that are not compatible with the geometric constraint of no (or limited) pebbles overlap. The second method models the actual physical packing process, accounting for the dynamic of pebbles as they are dropped onto the pebble bed and as they settle therein. A simplification in the latter model is the assumption of a starting point with very dilute packing followed by settling. The results from the two models are compared and the properties of the second model and the dependence of its results on many of the modeling parameters are presented. The first model (with no overlap allowed) has been implemented into a code to compute Dancoff factors. The second model will soon be implemented into that same code and will also be used to model flow of pebbles in a reactor and core densification in the simulation of earthquakes. Both methods reproduce experimental values well, with the latter displaying a high level of fidelity.
Date: September 1, 2005
Creator: Ougouag, Abderrafi M.; Cogliati, Joshua J. & Kloosterman, Jan-Leen
Partner: UNT Libraries Government Documents Department