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A coarse-mesh nodal method-diffusive-mesh finite difference method

Description: Modern nodal methods have been successfully used for conventional light water reactor core analyses where the homogenized, node average cross sections (XSs) and the flux discontinuity factors (DFs) based on equivalence theory can reliably predict core behavior. For other types of cores and other geometries characterized by tightly-coupled, heterogeneous core configurations, the intranodal flux shapes obtained from a homogenized nodal problem may not accurately portray steep flux gradients near fuel assembly interfaces or various reactivity control elements. This may require extreme values of DFs (either very large, very small, or even negative) to achieve a desired solution accuracy. Extreme values of DFs, however, can disrupt the convergence of the iterative methods used to solve for the node average fluxes, and can lead to a difficulty in interpolating adjacent DF values. Several attempts to remedy the problem have been made, but nothing has been satisfactory. A new coarse-mesh nodal scheme called the Diffusive-Mesh Finite Difference (DMFD) technique, as contrasted with the coarse-mesh finite difference (CMFD) technique, has been developed to resolve this problem. This new technique and the development of a few-group, multidimensional kinetics computer program are described in this paper.
Date: May 1, 1994
Creator: Joo, H. & Nichols, W. R.
Partner: UNT Libraries Government Documents Department

Dynamic and static error analyses of neutron radiography testing

Description: Neutron radiography systems are being used for real-time visualization of the dynamic behavior as well as time-averaged measurements of spatial vapor fraction distributions for two phase fluids. The data in the form of video images are typically recorded on videotape at 30 frames per second. Image analysis of he video pictures is used to extract time-dependent or time-averaged data. The determination of the average vapor fraction requires averaging of the logarithm of time-dependent intensity measurements of the neutron beam (gray scale distribution of the image) that passes through the fluid. This could be significantly different than averaging the intensity of the transmitted beam and then taking the logarithm of that term. This difference is termed the dynamic error (error in the time-averaged vapor fractions due to the inherent time-dependence of the measured data) and is separate from the static error (statistical sampling uncertainty). Detailed analyses of both sources of errors are discussed.
Date: March 1, 1999
Creator: Joo, H. & Glickstein, S.S.
Partner: UNT Libraries Government Documents Department

Development of a scattering probability method for accurate vapor fraction measurements by neutron radiography. Revision 1

Description: Recent test results indicated drawbacks associated with the simple exponential attenuation method (SEAM) as currently applied to neutron radiography measurements to determine vapor fractions in a hydrogenous two-phase flow in a metallic conduit. The scattering component of the neutron beam intensity exiting the flow system is not adequately accounted for by SEAM, and this leads to inaccurate results. To properly account for the scattering effect, a neutron scattering probability method (SPM) is developed. The method applies a neutron-hydrogen scattering kernel to scattered thermal neutrons that leave the incident beam in narrow conduits but eventually show up elsewhere in the measurements. The SPM has been tested with known vapor (void) distributions within an acrylic disk and a water/vapor channel. The vapor (void) fractions deduced by SPM are in good agreement with the known exact values. Details of the scattering correction method and the test results are discussed.
Date: November 1, 1998
Creator: Joo, H. & Glickstein, S.S.
Partner: UNT Libraries Government Documents Department

Detailed analyses of dynamic and static errors in neutron radiography testing

Description: Neutron radiography systems are being used for real-time visualization of the dynamic behavior as well as time-averaged measurements of spatial vapor fraction distributions for two phase fluids. The extraction of quantitative data on vapor-liquid flow fields is a significant advance in the methodology of fundamental two-phase flow experimentation. The data in the form of video images are typically recorded on videotape at 30 frames per second. Image analysis of the video pictures is used to extract time-dependent or time-averaged data. The determination of the average vapor fraction requires averaging of the logarithm of time-dependent intensity measurements of the neutron beam (gray scale distribution of the image) that passes through the fluid. This could be significantly different than averaging the intensity of the transmitted beam and then taking the logarithm of that term. This is termed the dynamic error (error in the time-averaged vapor fractions due t the inherent time-dependence of the measured data) and is separate from the static error (statistical sampling uncertainty). The results provide insight into the characteristics of these errors and help to quantify achievable bounds on the limits of these errors. The static error was determined by the uncertainties of measured beam intensities. It was found that the maximum static error increases as liquid thickness increases and can be reduced by increasing the neutron source strength. The dynamic error increased with large fluctuations in the local vapor fractions and with increasing liquid thickness. Detailed analyses of both sources of errors are discussed.
Date: January 1, 1999
Creator: Joo, H. & Glickstein, S.S.
Partner: UNT Libraries Government Documents Department

Void fraction measurements using neutron radiography

Description: Real-time neutron radiography is being evaluated for studying the dynamic behavior of two phase flow and for measuring void fraction in vertical and inclined water ducts. This technique provides a unique means of visualizing the behavior of fluid flow inside thick metal enclosures. To simulate vapor conditions encountered in a fluid flow duct, an air-water flow system was constructed. Air was injected into the bottom of the duct at flow rates up to 0.47 I/s (1 cfm). The water flow rate was varied between 0--3.78 I/m (0--1 gpm). The experiments were performed at the Pennsylvania State University nuclear reactor facility using a real-time neutron radiography camera. With a thermal neutron flux on the order of 10{sup 6}n/cm{sup 2}/s directed through the thin duct dimension, the dynamic behavior of the air bubbles was clearly visible through 5 cm (2 in.) thick aluminum support plates placed on both sides of the duct wall. Image analysis techniques were employed to extract void fractions from the data which was recorded on videotape. This consisted of time averaging 256 video frames and measuring the gray level distribution throughout the region. The distribution of the measured void fraction across the duct was determined for various air/water mixtures. Details of the results of experiments for a variety of air and water flow conditions are presented.
Date: December 31, 1992
Creator: Glickstein, S. S.; Vance, W. H. & Joo, H.
Partner: UNT Libraries Government Documents Department

Void fraction measurements of acrylic discs via neutron radiography

Description: Simulation experiments have been initiated to verify that neutron radiography methods can accurately measure void fractions under various operating conditions in a steam-water flow channel. Recent neutron radiography experiments measured void fractions in an air-water channel at atmospheric pressure and room temperature conditions. Because steam-water densities at atmospheric pressure and low temperature vary significantly from those at high pressure and high temperature, questions have been raised as to the ability of the neutron radiographic technique to deduce vapor fractions under various steam-water operating conditions. In response to this concern, the macroscopic neutron cross sections presented by two extreme steam-water conditions were simulated using acrylic discs containing various size holes. Acrylic, which contains hydrogen as a major constituent closely resembles the properties of water as seen by a thermal neutron beam. Through holes appear to the neutrons as steam voids at atmospheric pressure. The effect of increased vapor density, (and neutron macroscopic cross section), which would occur at high pressure, was simulated by not drilling the holes completely through the discs. The shallow hole plus the remaining material was made to simulate a through hole filled with a vapor density on the order of 14% of water. The measured void fractions deduced from digitally imaged neutron radiographs are in good agreement with the expected values for the two cases studied.
Date: May 1, 1994
Creator: Glickstein, S. S.; Joo, H.; Vance, W. H. & Murphy, J. H.
Partner: UNT Libraries Government Documents Department

Interpreting neutron radiographs via computer simulation

Description: Computer simulation of the neutron radiographic process has been performed using Monte Carlo techniques. The results were compared to extensive experimental studies aimed at interpreting neutron radiographs of a special gage that contained varying amounts of water. The gage is intended to be used as a calibration device for measuring the water content in an enclosure under varying experimental conditions. Edge effects between regions of changing water thicknesses complicated the calibration procedure. Computer simulation of the experimental gage provided considerable insight into understanding and interpreting the radiographs. The results of computer analysis of the neutron radiographic testing of the gage were in excellent agreement with densitometer scans for most of the cases that were evaluate. The computer simulation indicated that the importance of edge effects, beam divergence, position of the film, source neutron energy and effects of water temperature can be accurately and effectively studied using Monte Carlo computer simulation of the neutron radiographic experiment This type of analysis can significantly help in the interpretation of neutron radiographs as well as in the design of experimental systems that use neutron radiography as a measuring tool. Detailed results of the experimental tests and analyses are provided.
Date: May 1, 1992
Creator: Glickstein, S. S.; Joo, H. & Vance, W. H.
Partner: UNT Libraries Government Documents Department

Vapor fraction measurements in a steam-water tube at up to 15 bar using neutron radiography techniques

Description: Real time neutron radiography has been used to study the dynamic behavior of two phase flow and measure the time averaged vapor fraction in a heated metal tube containing boiling steam water operating at up to 15 bar pressure. The neutron radiographic technique is non-intrusive and requires no special transparent window region. This is the first time this technique has been used in an electrically heated pressurized flow loop. This unique experimental method offers the opportunity to observe and record on videotape, flow patterns and transient behavior of two phase flow inside opaque containers without disturbing the environment. In this study the test sections consisted of stainless steel tubes with a 1.27 cm outer diameter and wall thicknesses of 0.084 cm and 0.124 cm. The experiments were carried out at the Pennsylvania State University 1 megawatt TRIGA reactor facility utilizing a Precise Optics neutron radiography camera. The inlet water temperature to the test section was varied between 120 to 170 C and the flow rate set to 2.3 l/min. T{sub sat} 200 C at these conditions. The tube was resistivity heated by passing high currents ({approximately}1,000 A) through the stainless steel wall. Scattering due to water in the 1 cm tube is significant and Monte Carlo calculations simulating the experiment were made to correct for this effect on the vapor fraction measurement. Details of the experimental technique, methods for analyzing the data and the results of the experiments are discussed.
Date: February 1, 1998
Creator: Glickstein, S.S.; Murphy, J.H. & Joo, H.
Partner: UNT Libraries Government Documents Department

The MC21 Monte Carlo Transport Code

Description: MC21 is a new Monte Carlo neutron and photon transport code currently under joint development at the Knolls Atomic Power Laboratory and the Bettis Atomic Power Laboratory. MC21 is the Monte Carlo transport kernel of the broader Common Monte Carlo Design Tool (CMCDT), which is also currently under development. The vision for CMCDT is to provide an automated, computer-aided modeling and post-processing environment integrated with a Monte Carlo solver that is optimized for reactor analysis. CMCDT represents a strategy to push the Monte Carlo method beyond its traditional role as a benchmarking tool or ''tool of last resort'' and into a dominant design role. This paper describes various aspects of the code, including the neutron physics and nuclear data treatments, the geometry representation, and the tally and depletion capabilities.
Date: January 9, 2007
Creator: Sutton TM, Donovan TJ, Trumbull TH, Dobreff PS, Caro E, Griesheimer DP, Tyburski LJ, Carpenter DC, Joo H
Partner: UNT Libraries Government Documents Department

Implementation of nodal equivalence parameters in DIF3D-VARIANT for core analysis of prismatic Very High Temperature Reactor (VHTR).

Description: The VARIANT module of the DIF3D code has been upgraded to utilize surface-dependent discontinuity factors. The performance of the new capability is verified using two-dimensional core cases with control rods in reflector and fuel blocks. Cross sections for VHTR components were generated using the DRAGON and HELIOS codes. For rodded block cross sections, the DRAGON calculations used a single-block model or the multi-block models combined with MCNP4C flux solutions, whereas the HELIOS calculations utilized multi-block models. Results from core calculations indicate that multiplication factor, block power, and control rod worth are significantly improved by using surface-dependent discontinuity factors.
Date: March 15, 2007
Creator: Lee, C. H.; Joo, H. K.; Yang, W. S. & Taiwo, T. A.
Partner: UNT Libraries Government Documents Department

Application of a generalized interface module to the coupling of PARCS with both RELAP5 and TRAC-M

Description: In an effort to more easily assess various combinations of 3-D neutronic/thermal-hydraulic codes, the USNRC has sponsored the development of a generalized interface module for the coupling of any thermal-hydraulics code to any spatial kinetics code. In this design, the thermal-hydraulics, general interface, and spatial kinetics codes function independently and utilize the Parallel Virtual Machine (PVM) software to manage inter-process communication. Using this interface, the USNRC version of the 3D neutron kinetics code, PARCS, has been coupled to the USNRC system analysis codes RELAP5 and TRAC-M. RELAP5/PARCS assessment results are presented for an OECD/NEA main steam line break benchmark problem. The assessment of TRAC-M/PARCS has only recently been initiated; nonetheless, the capabilities of the coupled code are presented for the OECD/NEA main steam line break benchmark problem.
Date: April 1, 1999
Creator: Barber, D.A.; Wang, W.; Miller, R.M.; Downar, T.J.; Joo, H.G.; Mousseau, V.A. et al.
Partner: UNT Libraries Government Documents Department

A generalized interface module for the coupling of spatial kinetics and thermal-hydraulics codes

Description: A generalized interface module has been developed for the coupling of any thermal-hydraulics code to any spatial kinetics code. The coupling scheme was designed and implemented with emphasis placed on maximizing flexibility while minimizing modifications to the respective codes. In this design, the thermal-hydraulics, general interface, and spatial kinetics codes function independently and utilize the Parallel Virtual Machine software to manage cross-process communication. Using this interface, the USNRC version of the 3D neutron kinetics code, PARCX, has been coupled to the USNRC system analysis codes RELAP5 and TRAC-M. RELAP5/PARCS assessment results are presented for two NEACRP rod ejection benchmark problems and an NEA/OECD main steam line break benchmark problem. The assessment of TRAC-M/PARCS has only recently been initiated, nonetheless, the capabilities of the coupled code are presented for a typical PWR system/core model.
Date: March 1, 1999
Creator: Barber, D.A.; Miller, R.M.; Joo, H.G.; Downar, T.J.; Wang, W.; Mousseau, V.A. et al.
Partner: UNT Libraries Government Documents Department

Coupled 3D reactor kinetics and thermal-hydraulic code development activities at the U.S. Nuclear Regulatory Commission

Description: The USNRC version of the 3D neutron kinetics code, Purdue Advanced Reactor Core Simulator (PARCS), has been coupled to the USNRC thermal-hydraulic (T/H) codes RELAP5 and the consolidated TRAC (merger of TRAC-BF1 and TRAC-PF1). These coupled codes may be used to audit license safety analysis submittals where 3D spatial kinetics and thermal-hydraulic effects are important. The coupling scheme was designed and implemented with emphasis placed on maximizing flexibility while minimizing modifications to the respective codes. In this design, the T/H and neutronic codes function independently and utilize the Parallel Virtual Machine software to communicate with each other through code specific Data Mapping Routines, and a General Interface. RELAP5/PARCS validation results are presented for two NEACRP rod ejection benchmark problems. The validation of TRAC-M/PARCS has only recently been initiated, nonetheless, the capabilities of the coupled code are presented for a typical PWR system/core model.
Date: September 27, 1999
Creator: Barber, D.; Miller, R.M.; Joo, H.; Downar, T.; Wang, W. & Mousseau, V.
Partner: UNT Libraries Government Documents Department