Search Results

Advanced search parameters have been applied.
open access

850/sup 0/C VHTR plant technical description

Description: This report describes the conceptual design of an 842-MW(t) process heat very high temperature reactor (VHTR) plant having a core outlet temperature of 850/sup 0/C (1562/sup 0/F). The reactor is a variation of the high-temperature gas-cooled reactor (HTGR) power plant concept. The report includes a description of the nuclear heat source (NHS) and of the balance of reactor plant (BORP) requirements. The design of the associated chemical process plant is not covered in this report. The reactor de… more
Date: June 1, 1980
Partner: UNT Libraries Government Documents Department
open access

Severe accident core heatup transients in modular high temperature gas-cooled reactors without operating Reactor Cavity Cooling Systems

Description: The ultimate decay heat removal system for the current Modular High Temperature Gas-Cooled reactors is a completely passive natural convection air cooling loop. This paper considers an extremely remote accident scenario, where even this passive system fails, and heat rejection is only via a layer of thermal insulation to the reactor silo structure and the surrounding soil. The results show that even in this case the peak fuel temperatures remain well within safe limits. However, vessel and conc… more
Date: January 1, 1988
Creator: Kroeger, P. G.
Partner: UNT Libraries Government Documents Department
open access

Characterization of solids in the Three Mile Island Unit 2 Reactor defueling water: Addendum

Description: Shortly after ORNL/TM-10362 was issued, it was discovered that a series of 31 figures had been inadvertently omitted. These figures, which consist of scanning electron microscope (SEM) photographs and energy-dispersive X-ray fluorescence scans, provide significant information about the results obtained in the tests performed with water sample W3. This Addendum includes these figures. Details of and comments on the SEM photographs may be found in ORNL/TM-10362.
Date: March 1, 1988
Creator: Campbell, D.O.
Partner: UNT Libraries Government Documents Department
open access

An Evaluation of liquid metal leak detection methods for the Clinch River Breeder Reactor Plant

Description: This report documents an independent review and evaluation of sodium leak detection methods described in the Clinch River Breeder Reactor Preliminary Safety Analysis Report. Only information in publicly available documents was used in making the assessments.
Date: December 1, 1977
Creator: Morris, C. J. & Doctor, S. R.
Partner: UNT Libraries Government Documents Department
open access

Acoustic measurements of the boiling stability tests on THORS sodium loop. [LMFBR]

Description: Acoustic data of boiling stability tests on the THORS (Thermal-Hydraulic Out-of-Reactor Safety) facility were obtained using three sodium-immersible high temperature microphones. The data was analyzed in both the time and frequency domains and provides the following information: (1) the acoustic signal due to sodium boiling was clearly observed; (2) the signal level and the repetition rate of boiling pulses are directly proportional to the applied heat flux; (3) a typical boiling pulse consists… more
Date: January 1, 1977
Creator: Sheen, S.H.; Bobis, J.P. & Carey, W.M.
Partner: UNT Libraries Government Documents Department
open access

LOFT system structural response during subcooled blowdown

Description: The Loss-of-Fluid Test (LOFT) facility is a highly instrumented, pressurized water reactor test system designed to be representative of large pressurized water reactors (LPWRs) for the simulation of loss-of-coolant accidents (LOCAs). Detailed structural analysis and appropriate instrumentation (accelerometers and strain gages) on the LOFT system provided information for evaluation of the structural response of the LOFT facility for loss-of-coolant experiment (LOCE) induced loads. In general, th… more
Date: January 1, 1978
Creator: Martinell, J. S.
Partner: UNT Libraries Government Documents Department
open access

Deposition and removal of radioactive isotopes from LMFBR components

Description: The development of an analytical model to describe the production, transport and eventual removal of radioactive materials in the primary sodium of LMFBR's is a continuing Sodium Technology activity sponsored by the Department of Energy. This paper describes studies directed toward obtaining an understanding of the deposition from sodium of fuel cladding activated corrosion products onto stainless steel alloys and the effect of their diffusion into the base metal on the process required to deco… more
Date: January 1, 1980
Creator: Hill, E. F.; Lutton, J. M. & Maffei, H. P.
Partner: UNT Libraries Government Documents Department
open access

Rapid quenching of molten lithium-aluminum jets in water under loss-of-control-rod-cooling conditions

Description: A series of fifteen tests were performed to investigate the thermal interactions between molten LiAl control rod material and water under conditions prototypic of the loss-of-control-rod-cooling (LCRC) accident scenario. The experimental parameters such as melt mass, stream diameter, melt temperature and flowrate, water depth and water temperature were controlled or varied to agree with analytically determined conditions, thus insuring prototypicality of the experiments and applicability of the… more
Date: January 1, 1992
Creator: Greene, G. A.; Finfrock, C. C.; Schwarz, C. E.; Allison, D. K. & Hyder, M. L.
Partner: UNT Libraries Government Documents Department
open access

Once-through steam generator (OTSG) materials and water chemistry. [PWR]

Description: Materials and water chemistry research results associated with the development of the Oconee-1 Reactor steam generator are presented. A summary of water chemistry data acquired during preoperational testing and power operation to date is also included. These data confirm the operational practicality of the nuclear once-through concept using volatile water treatment and high purity condensate demineralized feedwater.
Date: January 1, 1974
Creator: Pocock, F.J. & Levstek, D.F.
Partner: UNT Libraries Government Documents Department
open access

FLOWTRAN benchmarking with onset of flow instability data from 1988 Columbia University single-tube OFI experiment

Description: Benchmarking FLOWTRAN, Version 16.2, with an Onset of Significant Voiding (OSV) criterion against measured Onset of Flow Instability (OFI) data from the 1988--89 Columbia University downflow tests has shown that FLOWTRAN with OSV is a conservative OFI predictor. Calculated limiting flow rates based on the Savannah River Site (SRS) OSV criterion were always higher than the measured flow rates at OFI. This work supplements recent FLOWTRAN benchmarking against 1963 downflow tests at Columbia Unive… more
Date: June 1, 1990
Creator: Chen, K.; Paul, P. K. & Barbour, K. L.
Partner: UNT Libraries Government Documents Department
open access

Radiochemical analysis of the first plateout probe from the Fort St. Vrain high-temperature gas-cooled reactor

Description: This report presents the analysis of radioactive elements on the first plateout probe from the Fort St. Vrain high-temperature gas-cooled reactor. The plateout probe is a device which samples the primary coolant for condensible fission products. Circuit inventories of individual radionuclides are estimated from the probe analysis. The analysis shows that the radioactive contamination in the primary circuit is remarkable low, with activation product concentrations much greater than that of fissi… more
Date: June 1, 1982
Creator: Burnette, R. D.
Partner: UNT Libraries Government Documents Department
open access

Passive safety and the advanced liquid metal reactors

Description: Advanced Liquid Metal Reactors being developed today in the USA are designed to make maximum use of passive safety features. Much of the LMR safety work at Argonne National Laboratory is concerned with demonstrating, both theoretically and experimentally, the effectiveness of the passive safety features. The characteristics that contribute to passive safety are discussed, with particular emphasis on decay heat removal systems, together with examples of Argonne's theoretical and experimental pro… more
Date: January 1, 1988
Creator: Hill, D.J.; Pedersen, D.R. & Marchaterre, J.F.
Partner: UNT Libraries Government Documents Department
open access

CRBRP sodium circulating pump design evaluation

Description: The following topics are discussed: (1) primary sodium pump design concept; (2) pump level control system; (3) resolution of design problems in stress analysis, dynamics analysis, and mechanical design; (4) model testing; (5) planned performance tests; and (6) fabrication status. (DG)
Date: December 1, 1977
Creator: Marrujo, F.; Cook, M.; Manners, L. & Cothran, H.
Partner: UNT Libraries Government Documents Department
open access

Comparison of TRAC and RELAP5 reactor system calculations for a DEGB LOCA in K-14. 1

Description: A comparison of TRAC and RELAP5 predictions of steady-state and DEGB LOCA results (FI phase) for K-14.1 has been made. Both codes had been previously benchmarked against 1985 L Reactor AC Flow data and were under configuration control. The purpose of the code-to-code comparison is to provide insight on the transient uncertainty in TRAC plenum and tank bottom plenum pressures. The comparisons focus on LOCA results between 0.5 and 2.0 s, which is the primary period of interest for Flow Instabilit… more
Date: September 1, 1990
Creator: Griggs, D.P. (Westinghouse Savannah River Co., Aiken, SC (United States)) & Liebmann, M.L. (Wais and Associates (United States))
Partner: UNT Libraries Government Documents Department
open access

Silicon mass transfer in sodium loops and the resulting/thermal hydraulic effects. [LMFBR]

Description: The element silicon in the surface of new, 300 series stainless steel has been shown to rapidly dissolve in sodium above 525/sup 0/C. It deposits in slightly cooler regions as a crystalline compound with sodium and oxygen. In tests, the deposits have caused increases in hydraulic friction factor (hence, increased pressure loss) of up to 300% at Reynolds Numbers of 14/sup 4/ to 10/sup 5/.Also, they have contributed to local losses of heat transfer rate to 1/10 the original value, at a Reynolds N… more
Date: February 1, 1980
Creator: Yunker, W.H.
Partner: UNT Libraries Government Documents Department
open access

Computing the effect of plastic deformation of piping on pressure transient propagation. [LMFBR]

Description: The computer program PTA-1 performs pressure-transient analysis of large piping networks using the one-dimensional method of characteristics applied to a fluid-hammer formulation. The effect of elastic-plastic deformation of piping on pulse propagation is included in the computation. Each pipe is modeled as a series of rings, neglecting axial effects, bending moments, and inertia. The fluid wave speed is a function of pipe deformation and, consequently, of position and time. Comparison with exi… more
Date: January 1, 1977
Creator: Youngdahl, C.K. & Kot, C.A.
Partner: UNT Libraries Government Documents Department
open access

Integral fast reactor safety features

Description: The Integral Fast Reactor (IFR) is an advanced liquid-metal-cooled reactor concept being developed at Argonne National Laboratory. The two major goals of the IFR development effort are improved economics and enhanced safety. In addition to liquid metal cooling, the principal design features that distinguish the IFR are: (1) a pool-type primary system, (2) an advanced ternary alloy metallic fuel, and (3) an integral fuel cycle with on-site fuel reprocessing and fabrication. This paper focuses on… more
Date: January 1, 1988
Creator: Cahalan, J. E.; Kramer, J. M.; Marchaterre, J. F.; Mueller, C. J.; Pedersen, D. R.; Sevy, R. H. et al.
Partner: UNT Libraries Government Documents Department
open access

Reactor thermal-hydraulic FY 1986 status report for the multimegawatt Space Nuclear Power Program

Description: PNL's 1986 activities can be divided into three basic areas: code assessment, correlation assessment and experimental activities. The ultimate goal of all these activities is developing computer codes and verifying their use to perform the thermal-hydraulic analysis and design of the reactor core and plenum of the various proposed concepts. To perform this task as assessment is made of existing computer codes, models, correlations, and microgravity experimental data.
Date: October 1, 1986
Creator: Krotiuk, W.J. & Antoniak, Z.I.
Partner: UNT Libraries Government Documents Department
open access

Calculation of Savannah River K Reactor Mark-22 assembly LOCA/ECS power limits

Description: This paper summarizes the results of TRAC-PF1/MOD3 calculations of Mark-22 fuel assembly of loss-of-coolant accident/emergency cooling system (LOCA/ECS) power limits for the Savannah River Site (SRS) K Reactor. This effort was part of a larger effort undertaken by the Los Alamos National Laboratory for the US Department of Energy to perform confirmatory power limits calculations for the SRS K Reactor. A method using a detailed three-dimensional (3D) TRAC model of the Mark-22 fuel assembly was d… more
Date: January 1, 1992
Creator: Fischer, S. R.; Farman, R. F. & Birdsell, S. A.
Partner: UNT Libraries Government Documents Department
open access

US GCFR demonstration plant design

Description: A general description of the US GCFR demonstration plant conceptual design is given to provide a context for more detailed papers to follow. The parameters selected for use in the design are presented and the basis for parameter selection is discussed. Nuclear steam supply system (NSSS) and balance of plant (BOP) component arrangements and systems are briefly discussed.
Date: May 1, 1980
Creator: Hunt, P. S. & Snyder, H. J.
Partner: UNT Libraries Government Documents Department
open access

Summary data for U. S. commercial nuclear power plants in the United States

Description: A compilation of data is presented for all United States commercial nuclear power plants for which a construction permit application was made through the Nuclear Regulatory Commission. The data are compiled in four separate tables with cross-referencing indexes: Table 1--General Data; Table 2--Reactor Data; Table 3--Site Data, and Table 4--Circulating-Water System Data. The power plants are listed in numerical order by docket number in all four tables.
Date: March 20, 1978
Creator: Heddleson, F.A.
Partner: UNT Libraries Government Documents Department
open access

Measurement of stochastic flow-induced power fluctuations at LOFT

Description: Investigation of power fluctuations at LOFT began after peak to peak power fluctuations of 1.5% of full power were observed prior to Loss of Coolant Experiment L2-3. The power fluctuations are caused by fluctuations in primary coolant flow. The fluctuations in coolant flow result in fluctuations in moderator temperature, which through the moderator coefficient of reactivity, cause the power fluctuations. This conclusion is supported experimentally by the high coherence between power and flow, a… more
Date: January 1, 1980
Creator: Cannon, J.W. & Clemo, T.M.
Partner: UNT Libraries Government Documents Department
Back to Top of Screen