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The Effectiveness of Spray Cooling

Description: Abstract: "A possible method of cooling a liquid-fuel reactor is by spraying liquid metal through the liquid fuel, and then circulating the liquid metal through a heat exchanger. To evaluate the effectiveness of this cooling method, a few simple experiments were made with mercury sprayed through water. On the basis of the results, it was concluded that this method was intrinsically a low-power-density method, which could not find application except where a low fissionable-material inventory was the dominating requirement in a low-power reactor. Even there, it is thought that a boiling homogeneous reactor might be superior. The results are reported, in spite of their probably lack of value in the reactor program, simply to make the record complete."
Date: October 1, 1953
Creator: Dayton, R. W.; Allen, C. M. & Miller, N. E.
Partner: UNT Libraries Government Documents Department

Design for a Fast Flux Liquid Metal Loop in the Advanced Test Reactor

Description: The purpose of this document is to present criteria for the Titles I and II design of the Fast Flux Loop Facility which is to be a part of the Advanced Test Reactor. This facility is to be used for irradiation of fast reactor assemblies under simulated fast reactor operating conditions.
Date: July 1962
Creator: Babcock & Wilcox Company. Atomic Energy Division.
Partner: UNT Libraries Government Documents Department

Multichannel Clad-Relocation Model for Fast-Reactor Loss-of-Flow Accidents

Description: During an unprotected undercooling accident in a liquid-metal fast breeder reactor, the motion and relocation of the molten cladding can he important because of its potentially significant effect on reactivity, blockage formation, and subsequent fuel motion. The present study analyzes the clad relocation problem1 based on a multichanneI film-flow model. The important aspects considered in the analysis are the nonuniform transverse ci1ad-melting pattern and diversions of sodium vapor flow within a subassembly. It has been shown that the motion of molten clad and subsequent blockage formations can be significantly influenced by this interconnected channel effect. Several sample cmi cu lat ions have been made in order to demonstrate these points.
Date: August 1979
Creator: Ishii, M.; Chen, W. L. & Grolmes, M. A.
Partner: UNT Libraries Government Documents Department

Liquid Metal Fuel Reactor: Four-Inch Utility Test Loop: Design, Construction, Operation, and Experimental Results

Description: Report issued by the Brookhaven National Laboratory discussing fuel reactor test loops. The "design, construction, operation, and experimental results" (p. 1) are presented. This report includes tables, illustrations, and photographs.
Date: July 14, 1960
Creator: Hoffman, K. C.; Isler, R. J.; Scarlett, C. H. & Schoener, G. A.
Partner: UNT Libraries Government Documents Department

Irradiation data for the MFA-1 and MFA-2 tests during FFTF cycles 10A and 10B

Description: This report provides key information on the irradiation environment of the MONJU fuel test MFA-1 and MFA-2 in the Fast Flux Test Facility (FFTF) during operating cycles 10A and 10B.This information includes the fission powers, neutron fluxes, sodium temperatures and sodium flow rates in MFA-1, MFA-2, and adjacent assemblies. It also includes MFA-1 and MFA-2 compositions as a function of exposure during cycles 10A and 10B. The work was performed at the request of Power Reactor and Nuclear Fuels (PNC) of Japan.
Date: October 30, 1996
Creator: Nelson, J.V.
Partner: UNT Libraries Government Documents Department

Status of LMFBR Reheat in Western Europe--1972: Report of the United States of America LMFBR Sodium Reheat Team Visit to France , Germany (FRG), Netherlands, and United Kingdom, May 22-June 6, 1972

Description: From summary: This report describes the trips to Western European countries designing and constructing LMFBR demonstration plants for the purpose of determining the directions being taken by these countries to provide acceptable quality steam to commercially available steam turbines.
Date: March 1973
Creator: Weber, Max
Partner: UNT Libraries Government Documents Department

An Overview of Pool-Type LMFBRs : General Characteristics

Description: This report describes the results of a study conducted by a "Pool Study Group" organized at ANL in mid-1975 to examine the present state of the air of design of pool-type LMFBRs. The study concentrated on examination of various design options used to date in the principle pool-type projects and design studies in this country and abroad, including the Phenix and Super-Phenix reactors (France), PFR and CFR (U.K.), RN-600 (U.S.S.R.) and EBR-II (U.S.A.). The objective of the report is to provide a step toward better understanding of the pool-type system and of the advantages and disadvantages of the various possible approaches to its design.
Date: May 1976
Creator: Amorosi, A.; Hutter, E.; Marciniak, T. J.; Monson, H. O.; Seidensticker, R. W. & Simmons, W. R.
Partner: UNT Libraries Government Documents Department

Explosive Interaction of Molten UO2 and Liquid Sodium

Description: The interim report presented describes a continuation of the work reported in ANL-7890, Interaction of Sodium with Molten Uranium dioxide and Stainless Steel Using a Dropping Mode of Contact. In the current study, sodium was injected into a pool of molten uranium dioxide. The experiment consistently produced vapor explosions, both with the injection nozzle above and beneath the surface of the uranium dioxide. Although the efficiency of the conversion of thermal to mechanical energy was small (due in part to very conservative data analysis and an inefficient geometry), the results did demonstrate that there is no intrinsic reason why reactor materials cannot produce a vapor explosion.
Date: March 1976
Creator: Armstrong, D. R.; Goldfuss, G. T. & Gebner, R. H.
Partner: UNT Libraries Government Documents Department

Tritium and Hydrogen Transport in LMFBR Systems: EBR-II, CRBR, and FFTF

Description: A tritium and hydrogen transport model has been employed to simulate concentration profiles, tritium losses to auxiliary containment systems, and cold trap burdens for EBR-II, CRBR, and FFTF. Experimental data from EBR-II were found to correlate well with calculated tritium and hydrogen profiles. A major change relative to previous transport models, namely, the inhibiting effect of oxide coatings on tritium permeation through reactor structural surfaces, has been incorporated into the current model. Tritium release rates to auxiliary systems where oxide barrier effects were included were predicted to be approximately two orders of magnitude lower than those for the reference case where structural surfaces were assumed to be totally oxide-free. Tritium releases during operation of large LMFBRs are expected to present essentially no hazard to the environment.
Date: September 1978
Creator: Renner, T. A. & McPheeters, C. C.
Partner: UNT Libraries Government Documents Department

Cladding Failure by Local Plastic Instability

Description: Cladding failure is one of the major considerations in analysis of fuel-pin behavior during hypothetical accident transients since time, location, and nature of failure govern the early post-failure material motion and reactivity feedback. Out-of-pile thermal transient tests of both irradiated and unirradiated fast-reactor cladding show that local plastic instability, or bulging, often precedes rupture and that the extent of local instability limits the initial rip length. To investigate the details of bulge formation and growth, a perturbation analysis of the equations governing large deformation of a cylindrical shell has been developed, resulting in a set of linear differential equations for the bulge geometry. These equations have been solved along with appropriate constitutive equations and various constraints on the ends of the cladding. Sources for bulge formation that have been considered include initial geometric imperfections and thermal perturbations due to either eccentric fuel pellets or non-symmetric cooling. Of these, only the first is relevant to out-of-pile burst tests. Here it has been found that the most likely imperfection that will grow unstably to failure leads to a bulge around half the circumference with an axial length 1.1 times the deformed diameter. This is in general agreement with burst-test results. For the case of in-reactor fuel pins, it has been found that thermal perturbations can significantly affect local instability, particularly if the deformation process is thermally activated with a high activation energy.
Date: December 1977
Creator: Kramer, J. M. & Deitrich, L. W.
Partner: UNT Libraries Government Documents Department

Transient Energy Transfer by Conduction and Radiation for a Sudden Contact Between Molten UO₂ and Sodium

Description: The transient energy transfer following a sudden contact between molten uranium dioxide and sodium has been investigated, taking into consideration both conduction and internal thermal radiation in uranium dioxide. Analytical expressions for the contact-interface temperature valid for small times are derived. Illustrative calculations indicate that on a time scale relevant to fuel-coolant interactions, internal radiation of molten uranium dioxide should have an insignificant effect on the contact-interface temperature between molten uranium dioxide and sodium. It thus appears that for the purpose of assessing the potential for an explosive fuel-coolant interaction, the contact-interface temperature may be adequately determined based on consideration of pure conduction.
Date: 1978?
Creator: Cho, D. H. & Chan, S. H.
Partner: UNT Libraries Government Documents Department