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PARET code and the analysis of the SPERT I transients

Description: The PARET code has been adapted for the testing of methods and models and for subsequent use in the analysis of transient behavior in research reactors. Comparisons with the experimental results from the SPERT-I transients are provided. The code has also been applied to the analysis of the IAEA 10 MW benchmark cores for protected and unprotected transients. The present version of the PARET code provides a convenient means of assessing the various models and correlations proposed for use in the analysis of research reactor behavior. For comparison with experiments the SPERT-I cores B-24/32, B-12/64, and D-12/25 were chosen. The results of this comparison of the PARET code with the SPERT I cores are generally quite favorable. The agreement with the B-24/32 core is particularly good. This core might be considered a more representative core for research reactors than either the B-12/64 or D-12/25 cores.
Date: January 1, 1982
Creator: Woodruff, W.L.
Partner: UNT Libraries Government Documents Department

Evaluation and selection of hot channel (peaking) factors for research reactor applications

Description: A proposed method for selecting and applying hot channel factors is presented along with some justification for these selections. The method is illustrated by example, and the sensitivity to some of the choices is examined. The uncertainty in the heat transfer coefficient is a major contributor to the reduction in thermal-hydraulic safety margins. The uncertainty introduced by the heterogeneity in the fuel is another important contributor and an area where more information may be useful in reducing this uncertainty.
Date: January 1, 1987
Creator: Woodruff, W.L.
Partner: UNT Libraries Government Documents Department

Benchmark revisited: the accuracy of finite difference methods and the potential for the use of nodal methods in the analyses of research and test reactors

Description: Nodal methods are tested for use in the neutronics analysis of research reactors. The results are compared to finite difference methods in both cost and accuracy. Transport effects and nodal transport are also considered. The IAEA benchmark reactor has been used as a base for this assessment. The nodal method was found to be very effective in reducing computation costs, and the accuracy of the solution is often higher than that attained with finite difference. In fact, in order to have a finite difference solution of comparable accuracy, the costs can be prohibitive. The usual choice of mesh structure may produce significant inaccuracies in the solution, and transport effects in many cases may be quite large. The use of transport nodal methods for some applications may be justified and far less expensive than Monte Carlo or Sn transport methods.
Date: January 1, 1985
Creator: Woodruff, W.L.
Partner: UNT Libraries Government Documents Department

Upgrades to the WIMS-ANL code.

Description: The dusty old source code in WIMS-D4M has been completely rewritten to conform more closely with current FORTRAN coding practices. The revised code contains many improvements in appearance, error checking and in control of the output. The output is now tabulated to fit the typical 80 column window or terminal screen. The Segev method for resonance integral interpolation is now an option. Most of the dimension limitations have been removed and replaced with variable dimensions within a compile-time fixed container. The library is no longer restricted to the 69 energy group structure, and two new libraries have been generated for use with the code. The new libraries are both based on ENDF/B-VI data with one having the original 69 energy group structure and the second with a 172 group structure. The common source code can be used with PCs using both Windows 95 and NT, with a Linux based operating system and with UNIX based workstations. Comparisons of this version of the code to earlier evaluations with ENDF/B-V are provided, as well as, comparisons with the new libraries.
Date: October 14, 1998
Creator: Woodruff, W. L.
Partner: UNT Libraries Government Documents Department

A comparison of the RELAP5/MOD3 and PARET/ANL codes with the experimental transient data from the SPERT-IV D-12/25 series.

Description: The results from the RELAP5/MOD3 and PARET/ANL codes are compared with the SPERT-IV series of experimental reactivity insertion transients. The PARET/ANL code provides conservative estimates of SPERT-IV experimental data for the midrange transients and for the more severe transients. The PARET results are similar to the results obtained earlier for the SPERT-I D-12/25 series of experiments. The RELAP5/MOD3 code (including the developmental version 3.2.1.2) gives results comparable to PARET for some midrange transients, but seriously diverges from the experimental data when significant boiling is present. Based on the results of this study, the use of the RELAP5 code for research reactor applications should be limited to transients that do not generate substantial boiling and voids. We hope to be able to resolve these differences in further work with the NRC staff and its contractors. The RELAP5 code would be a more useful tool for the analyses research reactor transients with the addition of suitable correlations for low pressures and plate type geometry.
Date: January 16, 1998
Creator: Woodruff, W. L.
Partner: UNT Libraries Government Documents Department

Thermal-hydraulic aspects of flow inversion in a research reactor

Description: PARET, a neutronics and thermal-hydraulics computer code, has been modified to account for natural convection in a reactor core. The code was then used to analyze the flow inversion that occurs in a reactor with heat removal by forced convection in the downward direction after a pump failure. Typical results are shown for a number of parameters. Research reactors normally operating much above ten MW are predicted to experience nucleate boiling in the event of a flow inversion. Comparison with experimental results from the Belgian BR2 reactor indicated general agreement although nucleate boiling that was analytically predicted was not noted in the BR2 data.
Date: November 1, 1986
Creator: Smith, R.S. & Woodruff, W.L.
Partner: UNT Libraries Government Documents Department

Additional capabilities and benchmarking with the SPERT transients for heavy water application of the PARET code

Description: The capabilities of the PARET code have been expanded to include the ability to compute steady-state and transient results for heavy water reactors. A comparison is provided between PARET and the SPERT II series of transients. Another significant improvement in the code is the addition of a restart capability. The current capabilities of the code are summarized. 7 refs., 12 figs., 3 tabs.
Date: January 1, 1989
Creator: Woodruff, W.L.
Partner: UNT Libraries Government Documents Department

Applications and results for the supercell option of the WIMS-D4M code

Description: The Supercell option of the WIMS-D4M code is used with a model for the Advanced Neutron Source design to illustrate the capability, and the results are compared with Monte Carlo. The capability is also used to successfully model Russian designed fuel assemblies with concentric tubes. The capability to model homogenized and resonance corrected fuel and to properly treat secondary regions containing resonance materials is particularly useful. The Supercell option is well suited to modeling non-lattice regions, such as, reflector, control and/or experimental regions of research reactors.
Date: December 31, 1995
Creator: Woodruff, W.L. & Costescu, C.I.
Partner: UNT Libraries Government Documents Department

Neutronic analysis of the JMTR with LEU fuel and burnable poison

Description: The results of neutronics calculations are presented for the JMTR equilibrium core with LEU silicide fuel, boron and cadmium burnable poisons in the sideplates, and a cycle length of 24 days instead of 11 days with the current HEU fuel. The data indicate that several options are feasible provided that silicide fuels with high uranium densities are successfully demonstrated and licensed. 2 refs., 10 figs., 5 tabs.
Date: January 1, 1984
Creator: Nagaoka, Yoshiharu; Oyamada, Rokuro; Matos, J.E. & Woodruff, W.L.
Partner: UNT Libraries Government Documents Department

Radiological consequence analysis with HEU and LEU fuels

Description: A model for estimating the radiological consequences from a hypothetical accident in HEU and LEU fueled research and test reactors is presented. Simple hand calculations based on fission product yield table inventories and non-site specific dispersion data may be adequate in many cases. However, more detailed inventories and site specific data on meteorological conditions and release rates and heights can result in substantial reductions in the dose estimates. LEU fuel gives essentially the same doses as HEU fuel. The plutonium buildup in the LEU fuel does not significantly increase the radiological consequences. The dose to the thyroid is the limiting dose. 10 references, 3 figures, 7 tables.
Date: January 1, 1984
Creator: Woodruff, W.L.; Warinner, D.K. & Matos, J.E.
Partner: UNT Libraries Government Documents Department

Comparison of safety parameters and transient behavior of a generic 10 MW reactor with HEU and LEU fuels

Description: Key safety parameters are compared for equilibrium cores of the IAEA generic 10 MW reactor with HEU and LEU fuels. These parameters include kinetics parameters, reactivity feedback coefficients, control rod worths, power peaking factors, and shutdown margins. Reactivity insertion and loss-of-flow transients are compared. Results indicate that HEU and LEU cores will behave in a very similar manner.
Date: January 1, 1983
Creator: Matos, J.E.; Freese, K.E. & Woodruff, W.L.
Partner: UNT Libraries Government Documents Department

Analysis of the KUCA MEU experiments using the ANL code system

Description: This paper provides some preliminary results on the analysis of the KUCA critical experiments using the ANL code system. Since this system was employed in the earlier neutronics calculations for the KUHFR, it is important to assess its capabilities for the KUHFR. The KUHFR has a unique core configuration which is difficult to model precisely with current diffusion theory codes. This paper also provides some results from a finite-element diffusion code (2D-FEM-KUR), which was developed in a cooperative research program between KURRI and JAERI. This code provides the capability for mockup of a complex core configuration as the KUHFR. Using the same group constants generated by the EPRI-CELL code, the results of the 2D-FEM-KUR code are compared with the finite difference diffusion code (DIF3D(2D) which is mainly employed in this analysis.
Date: January 1, 1982
Creator: Shiroya, S.; Hayashi, M.; Kanda, K.; Shibata, T.; Woodruff, W.L. & Matos, J.E.
Partner: UNT Libraries Government Documents Department

WIMS-D4M user manual

Description: The Winfrith Improved Multigroup Scheme (WIMS) code has been used extensively throughout the world for power and research reactor lattice physics analysis. There are many WIMS versions currently in use. The D4 version selected by the RERTR program was originally developed in 1980). It was chosen for the accurate lattice physics capability and an unrestricted distribution privilege. The code and its 69-group library tape 166259 generated in Winfrith were obtained from the Oak Ridge National Laboratory Radiation Shielding Information Center (RSIC) in 1992. Since that time the RERTR program has added three important features. The first was the capability to generate up to 20 broad-group bumup-dependent macroscopic or microscopic ISOTXS cross sections for each composition of the unit cell, a new ENDF/B-V based nuclear data library, and a new Supercell option. As a result of these modifications and other minor ones, the code is now named WIMS-D4M. A supplementary reference guide can be obtained from the RSIC that contains detailed explanations of all user options, library contents, along with several sample problems. Primary applications of WIMS for research reactor modeling do not require an extensive knowledge of all WIMS user options. This user guide is primarily addressed to the needs of the research reactor community although the code can be used for most thermal reactor lattices. The guide is written based on the experience of the RERTR staff with WIMS-D4M and will discuss only the most needed options for research reactor analyses.
Date: July 1, 1995
Creator: Deen, J.R.; Woodruff, W.L. & Costescu, C.I.
Partner: UNT Libraries Government Documents Department

Analyses for conversion of the Georgia Tech Research Reactor from HEU to LEU fuel

Description: This document presents information concerning: analyses for conversion of the Georgia Tech Research Reactor from HEU to LEU; changes to technical specifications mandated by the conversion of the GTRR to low enrichment fuel; changes in the Safety Analysis Report mandated by the conversion of the GTRR to low enrichment fuel; and copies of all changed pages of the SAR and the technical specifications.
Date: September 1, 1992
Creator: Matos, J.E.; Mo, S.C. & Woodruff, W.L.
Partner: UNT Libraries Government Documents Department

Analyses for conversion of the Georgia Tech Research Reactor from HEU to LEU fuel

Description: The 5 MW Georgia Tech Research Reactor (GTRR) is a heterogeneous, heavy water moderated and cooled reactor, fueled with highly-enriched uranium aluminum alloy fuel plates. The GTRR is required to convert to low enrichment (LEU) fuel in accordance with USNRC policy. Results of design and safety analyses performed by the RERTR Program at the Argonne National Laboratory for LEU conversion of the GTRR are summarized. Only those parameters which could change as a result of replacing the fuel are addressed. The performance of the reactor and all safety margins with LEU fuel are expected to be about the same as those with the current HEU fuel.
Date: January 1, 1992
Creator: Matos, J.E.; Mo, S.C. & Woodruff, W.L.
Partner: UNT Libraries Government Documents Department

A comparison of the PARET/ANL and RELAP5/MOD3 codes for the analysis of IAEA benchmark transients

Description: The PARET/ANL and RELAP5/MOD3 codes are used to analyze the series of benchmark transients specified for the IAEA Research Reactor Core Conversion Guidebook (IAEA-TECDOC-643, Vol. 3). The computed results for these loss-of-flow and reactivity insertion transients with scram are in excellent agreement and agree well with the earlier results reported in the guidebook. Attempts to also compare RELAP5/MOD3 with the SPERT series of experiments are in progress.
Date: December 31, 1996
Creator: Woodruff, W.L.; Hanan, N.A.; Smith, R.S. & Matos, J.E.
Partner: UNT Libraries Government Documents Department

A comparison of WIMS-D4 and WIMS-D4m generated cross-section data with Monte Carlo

Description: Cross-section and related data generated by a modified version of the WIMS-D4 code for both plate and rod type research reactor fuel are compared with Monte Carlo data from the VIM code. The modifications include the introduction of a capability for generating broad group microscopic data and to write selected microscopic cross-sections to an ISOTXS file format. The original WIMS-D4 library with H in ZrH, and [sup 166]Er and [sup 167]Er added gives processed microscopic cross-section data that agree well with VIM ENDF/B-V based data for both plate and TRIGA cells. Additional improvements are in progress including the capability to generate an ENDF/B-V based library.
Date: January 1, 1992
Creator: Woodruff, W.L.; Deen, J.R. (Argonne National Lab., IL (United States)) & Costescu, C.I. (Illinois Univ., Urbana, IL (United States))
Partner: UNT Libraries Government Documents Department

Status report on conversion of the Georgia Tech Research Reactor to low enrichment fuel

Description: The 5 MW Georgia Tech Research Reactor (GTRR) is a heterogeneous, heavy water moderated and cooled reactor, fueled with highly-enriched uranium aluminum alloy fuel plates. The GTRR is required to convert to low enrichment (LEU) fuel in accordance with USNRC policy. The US Department of Energy is funding a program to compare reactor performance with high and low enrichment fuels. The goals of the program are: (1) to amend the SAR and the Technical Specifications of the GTRR so that LEU U{sub 3}Si{sub 2}-Al dispersion fuel plates can replace the current HEU U-Al alloy fuel, and (2) to optimize the LEU core such that maximum value neutron beams can be extracted for possible neutron capture therapy application. This paper presents a status report on the LEU conversion effort.
Date: January 1, 1991
Creator: Karam, R.A. (Georgia Inst. of Tech., Atlanta, GA (United States)); Matos, J.E.; Mo, S.C. & Woodruff, W.L. (Argonne National Lab., IL (United States))
Partner: UNT Libraries Government Documents Department