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Computer aided failure analysis. Final report

Description: A computer aided failure analysis (CAFA) system was developed to troubleshoot defects in electronic assemblies. Through a question and answer procedure, the system provides step-by-step corrections to guide a troubleshooter to the fault location. A diagnostic logic routine has been established for one product and the software necessary to store and implement the routine has been developed. The TSO terminal has been installed and the completed system is functional. A visual aid catalog has been developed for the current CAFA routine.
Date: December 1, 1978
Creator: Smith, R.S.
Partner: UNT Libraries Government Documents Department

Analysis of the effect of transverse power distribution in an involute fuel plate with and without oxide film formation.

Description: Existing thermal hydraulics computer codes can account for variations in power and temperature in the axial and thickness directions but variations across the width of the plate cannot be accounted for. In the case of fuel plates in an annular core this can lead to significant errors which are accentuated by the presence of an oxide layer that builds up on the aluminum cladding with burnup. This paper uses a three dimensional SINDA model to account for the transverse variations in power. The effect of oxide thickness on these differences is studied in detail. Power distribution and fuel conductivity are also considered. The lower temperatures predicted with the SINDA model result in a greater margin to clad and fuel damage.
Date: October 27, 1998
Creator: Smith, R. S.
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

An alternative LEU design for the FRM-II

Description: The Alternative LEU Design for the FRM-II proposed by the RERTR Program at Argonne National Laboratory (ANL) has a compact core consisting of a single fuel element that uses LEU silicide fuel with a uranium density of 4.5 g/cm[sup 3] and has a power level of 32 MW. Both the HEU design by the Technical University of Munich (TUM) and the alternative LEU design by ANL have the same fuel lifetime (50 days) and the same neutron flux performance (8 x 10[sup 14] n/cm[sup 2]/s in the reflector). LEU silicide fuel with 4.5 g/cm[sup 3] has been thoroughly tested and is fully-qualified, licensable, and available now for use in a high flux reactor such as the FRM-II. Computer models for the HEU and LEU designs have been exchanged between TUM and ANL and discrepancies have been resolved. The following issues are addressed: qualification of HEU and LEU silicide fuels, stability of the fuel plates, gamma heating in the heavy water reflector, a hypothetical accident involving the configuration of the reflector, a loss of primary coolant flow transient due to an interrupted power supply, the radiological consequences of larger fission product and plutonium inventories in the LEU core, and cost and schedule. Calculations were also done to address the possibility that new high density LEU fuels could be developed that would allow conversion of the TUM HEU design to LEU fuel. Based on the excellent results for the Alternative LEU Design that were obtained in these analyses, the RERTR Program concludes that all of the major technical issues regarding use of LEU fuel instead of HEU fuel in the FRM-II have been successfully resolved and that it is definitely feasible to use LEU fuel in the FRM-II without compromising the safety or performance of the facility.
Date: February 1, 1997
Creator: Hanan, N.A.; Mo, S.C.; Smith, R.S. & Matos, J.E.
Partner: UNT Libraries Government Documents Department

An alternative LEU design for the FRM-II

Description: The Alternative LEU Design for the FRM-II proposed by the RERTR Program at Argonne National Laboratory (ANL) has a compact core consisting of a single fuel element that uses LEU silicide fuel with a uranium density of 4.5 g/cm{sup 3} and has a power level of 32 MW. Both the HEU design by the Technical University of Munich (TUM) and the alternative LEU design by ANL have the same fuel lifetime (50 days) and the same neutron flux performance. LEU silicide fuel with 4.5 g/cm{sup 3} has been thoroughly tested and is fully-qualified, licensable, and available now for use in a high flux reactor such as the FRM-II. The following issues raised by TUM were addressed in Ref. 1: qualification of HEU and LEU silicide fuels, gamma heating in the heavy water reflector, radiological consequences of larger fission product and plutonium inventories in the LEU core, and cost and schedule. The conclusions of these analyses are summarized below. This paper addresses three additional safety issues that were raised by TUM in Ref. 2: stability of the involute fuel plates, a hypothetical accident involving the configuration of the reflector, and a loss of primary coolant flow transient due to an interrupted power supply. Based on the excellent results for the Alternative LEU Design that were obtained in these analyses, the RERTR Program concludes that all of the major technical issues regarding use of LEU fuel instead of HEU fuel in the FRM-II have been successfully resolved and that it is definitely feasible to use LEU fuel in the FRM-II without compromising the safety or performance of the facility.
Date: December 1, 1996
Creator: Hanan, N.A.; Mo, S.C.; Smith, R.S. & Matos, J.E.
Partner: UNT Libraries Government Documents Department

A discriminator with a current-sum multiplicity output for the PHENIX multiplicity vertex detector

Description: A current output multiplicity discriminator for use in the front-end electronics (FEE) of the Multiplicity Vertex Detector (MVD) for the PHENIX detector at RHIC has been fabricated in the a 1.2-{micro} CMOS, n-well process. The discriminator is capable of triggering on input signals ranging from 0.25 MIP to 5 MIP. Frequency response of the discriminator is such that the circuit is capable of generating an output for every bunch crossing (105 ns) of the RHIC collider. Channel-to-channel threshold matching was adjustable to {+-} 4 mV. One channel of multiplicity discriminator occupied an area of 85 {micro} x 630 {micro} and consumed 515 {micro}W from a single 5-V supply. Details of the design and results from prototype device testing are presented.
Date: December 1, 1996
Creator: Smith, R.S.; Kennedy, E.J. & Jackson, R.G.
Partner: UNT Libraries Government Documents Department

Alterative LEU designs for the FRM-II with power levels of 20-22 MW.

Description: Alternative LEU Designs for the FRM-II have been developed by the RERTR Program at Argonne National Laboratory (ANL) at the request of an FRM-II Expert Group established by the German Federal Government in January 1999 to evaluate the options for using LEU fuel instead of HEU fuel in cores with power levels of 20 MW. The ANL designs would use the same building structure and maintain as many of the HEU design features as practical. The range of potential LEU fuels was expanded from previous studies to include already-tested silicide fuels with uranium densities up to 6.7 g/cm{sup 3} and the new U-Mo fuels that show excellent prospects for achieving uranium densities in the 8-9 g/cm{sup 3} range. For each of the LEU cores; the design parameters were chosen to match the 50 day cycle length of the HEU core and to maximize the thermal neutron flux in the Cold Neutron Source and beam tubes. The studies concluded that an LEU core with a diameter of about 29 cm instead of 24 cm in HEU design and operating at a power level of 20 MW would have thermal neutron fluxes that are 0.85 times that of the HEU design at the center of the Cold Neutron Source. With a potential future upgrade to a power of 22 MW, this ratio would increase to 0.93.
Date: September 27, 1999
Creator: Hanan, N. A.; Smith, R. S. & Matos, J. E.
Partner: UNT Libraries Government Documents Department

Mineral resources of the North Algodones Dunes Wilderness Study Area (CDCA-360), Imperial County, California

Description: This report presents the results of a mineral survey of the North Algodones Dunes Wilderness Study Area (CDCA-360), California Desert Conservation Area, Imperial County, California. The potential for undiscovered base and precious metals, and sand and gravel within the North Algodones Dunes Wilderness Study Area is low. The study area has a moderate potential for geothermal energy. One small sand-free area between the Coachella Canal and the west edge of the dune field would probably be the only feasible exploration site for geothermal energy. The study area has a moderate to high potential for the occurrence of undiscovered gas/condensate within the underlying rocks. 21 refs.
Date: January 1, 1984
Creator: Smith, R.S.U.; Yeend, W.; Dohrenwend, J.C. & Gese, D.D.
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

Neutronic safety and transient analyses for potential LEU conversion of the IR-8 research reactor.

Description: Kinetic parameters, isothermal reactivity feedback coefficients and three transients for the IR-8 research reactor cores loaded with either HEU(90%), HEU(36%), or LEU (19.75%) fuel assemblies (FA) were calculated using three dimensional diffusion theory flux solutions, RELAP5/MOD3.2 and PARET. The prompt neutron generation time and effective delayed neutron fractions were calculated for fresh and beginning-of-equilibrium-cycle cores. Isothermal reactivity feedback coefficients were calculated for changes in coolant density, coolant temperature and fuel temperature in fresh and equilibrium cores. These kinetic parameters and reactivity coefficients were used in transient analysis models to predict power histories, and peak fuel, clad and coolant temperatures. The transients modeled were a rapid and slow loss-of-flow, a slow reactivity insertion, and a fast reactivity insertion.
Date: September 27, 1999
Creator: Deen, J. R.; Hanan, N. A.; Smith, R. S.; Matos, J. E.; Egorenkov, P. M. & Nasonov, V. A.
Partner: UNT Libraries Government Documents Department

Post-radiation memory correction using differential subtraction for Phenix

Description: In colliders such as RHIC, the radiation levels are well below those of colliders such as LHC. The problem is that there can be enough radiation at the inner detector (Multiplicity-Vertex Detector or MVD) to significantly affect a low-priced, nonradiation-hard CMOS process. If the radiation affects the entire analog memory in a uniform fashion, then a real-time correction should be able to be performed to correct any changes seen in the memory and also the induced correlated noise from detector pickup thus precluding the need for a more expensive rad-hard process. This paper will present testing on memories fabricated in a `soft` process and exposed to ionizing radiation. We used a single pipeline as a reference to be subtracted in a cell-by-cell basis from each pipe during read out and investigated the spatial effects of using different pipes for the reference. Use of this method reduced the noise which was common to all pipes (common-mode noise) and thus reduced both common-mode input noise and pattern noise generated from address lines being exercised on the AMU. The correlation across the memories (6-, 8-, and 16-channel AMUs fabricated in the Orbit 1.2{mu} CMOS process) vs. radiation dose was found to be quite good. Both pre-and post-radiation results are presented on systems designed for PHENIX and WA98 at CERN as well as measured results on the minimization of the effects of injected systematic noise.
Date: June 1, 1995
Creator: Britton, C.L. Jr.; Wintenberg, A.L.; Womac, M.; Kennedy, E.J.; Smith, R.S.; Young, G.R. et al.
Partner: UNT Libraries Government Documents Department

A Radiation-Hard Analog Memory In The AVLSI-RA Process

Description: A radiation hardened analog memory for an Interpolating Pad Camber has been designed at Oak Ridge National Laboratory and fabricated by Harris Semiconductor in the AVLSI-RA CMOS process. The goal was to develop a rad-hard analog pipeline that would deliver approximately 9-bit performance, a readout settling time of 500ns following read enable, an input and output dynamic range of +/-2.25V, a corrected rms pedestal of approximately 5mV or less, and a power dissipation of less than 10mW/channel. The pre- and post-radiation measurements to 5MRad are presented.
Date: December 31, 1995
Creator: Britton, C.L. Jr.; Wintenberg, A.L.; Read, K.F.; Simpson, M.L.; Young, G.R.; Clonts, L.G., Kennedy, E.J., Smith, R.S., Swann, B.K. et al.
Partner: UNT Libraries Government Documents Department