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in subjects: "Sodium Cooled Reactors"

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Dynamic analysis of the EBR-II plant with the DSNP-ND simulation package

Development and initial testing have been completed for a special version of the DSNP simulation language software package, DSNP-ND, which simulates EBR-II based on the models of the primary system components used in the NATDEMO program. Preliminary results are included for a comparison with an original NATDEMO prediction for the SHRT-45 transient. 7 refs.

Date: January 1, 1990

Creator: Madell, J.T. (System Simulation and Analyses, Inc., Western Springs, IL (USA)); Dean, E.M. (Argonne National Lab., Idaho Falls, ID (USA)); Mohr, D. & Hung, T.C. (Argonne National Lab., IL (USA))

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Experimental Breeder Reactor II (EBR-II): Instrumentation for core surveillance

EBR-II has operated for 25 years in support of several major programs. During this time period, several of the original, non-replaceable, flow sensors, RDT sensors and thermocouples have failed in the primary system. This has led to the development of new sensors and the use of calculated values using computer models of the plant. It is important for the next generation of LMR reactors to minimize or eliminate the use of non-replaceable sensors. EBR-II is perhaps the best modeled reactor in the… more

Date: January 1, 1989

Creator: Christensen, L.J.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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Low-order dynamic modeling of the Experimental Breeder Reactor II

This report describes the development of a low-order, linear model of the Experimental Breeder Reactor II (EBR-II), including the primary system, intermediate heat exchanger, and steam generator subsystems. The linear model is developed to represent full-power steady state dynamics for low-level perturbations. Transient simulations are performed using model building and simulation capabilities of the computer software Matrix{sub x}. The inherently safe characteristics of the EBR-II are verified… more

Date: July 1, 1990

Creator: Berkan, R. C. (Tennessee Univ., Knoxville, TN (USA). Dept. of Nuclear Engineering); Upadhyaya, B. R. & Kisner, R. A. (Oak Ridge National Lab., TN (USA))

Item Type: Report

Partner: UNT Libraries Government Documents Department

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Calculation of temperature coefficients of reactivity for EBR-II kinetic analyses

Temperature coefficients of reactivity for use in coupled neutronics-thermohydraulics kinetics codes, as for example the EROS code used for Experimental Breeder Reactor-II (EBR-II) kinetic analyses, are both loading and problem-modeling sensitive. To enable appropriate temperature coefficients to be calculated for differing loading configurations and differing subassembly groupings in the kinetics analyses, an addition ((TEMCO) has been made to the EBRPOCO code. EBRPOCO calculates detailed axia… more

Date: January 1, 1987

Creator: Meneghetti, D. & Kucera, D.A.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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The influence of helium on mechanical properties of model austenitic alloys, determined using sup 59 Ni isotopic tailoring and fast reactor irradiation

The objective of this effort is to study the separate and synergistic effects of helium and other important variables on the evolution of microstructure and macroscopic properties during irradiation of structural metals. The alloys employed in this study were nominally Fe-15Cr-25Ni, Fe-15Cr-25Ni-0.04P and Fe-15Cr-45Ni (wt %) in both the cold worked and annealed conditions. Tensile testing and microscopy continue on specimens removed from the first, second and third discharges of the {sup 59}Ni … more

Date: November 1, 1990

Creator: Hamilton, M. L.; Garner, F. A. (Pacific Northwest Lab., Richland, WA (USA)) & Oliver, B. M. (Rockwell International Corp., Canoga Park, CA (USA))

Item Type: Report

Partner: UNT Libraries Government Documents Department

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Effect of elevated-temperature neutron irradiation on fracture toughness of ceramics

Single-crystal forms of MgAl/sub 2/O/sub 4/, Y/sub 3//Al/sub 5/O/sub 12/, and Al/sub 2/O/sub 3/ were irradiated in the EBR-II fast fission reactor to a fluence of 0.3 x 10/sup 26/ n/m/sup 2/ at 1015/sup 0/K, and to /sup -/1 to 2 x 10/sup 26/ n/m/sup 2/ at 925 and 1100/sup 0/K. Fracture toughness was subsequently measured at room temperature by an indentation technique, and radiation-induced defect aggregates were characterized by transmission electron microscopy. A slight increase in toughness … more

Date: December 3, 1984

Creator: Clinard, F. W., Jr.; Hurley, G. F.; Youngman, R. A. & Hobbs, L. W.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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A decade of radiological and shielding experience at the Fast Flux Test Facility

The Fast Flux Test Facility (FFTF) has operated for almost a decade after first going critical during February 1980. Based on about 2,000 effective full-power days of operation, it is concluded that radiological safety can be achieved in large liquid metal-cooled fast reactors. The collective dose equivalents received by operating personnel are significantly lower than those received at commercial light water reactors. No major contamination problems have been encountered in operating and maint… more

Date: November 1, 1990

Creator: Bunch, W. L.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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Application of three-dimensional transport code to the analysis of the neutron streaming experiment

This paper summarized the calculational results of neutron streaming through a Clinch River Breeder Reactor (CRBR) Prototype coolant pipe chaseway. Particular emphasis is placed on results at bends in the chaseway. Calculations were performed with three three-dimensional codes: the discrete ordinates radiation transport code TORT and Monte Carlo radiation transport code MORSE, which were developed by Oak Ridge National Laboratory (ORNL), and the discrete ordinates code ENSEMBLE, which was devel… more

Date: January 1, 1990

Creator: Chatani, K. & Slater, C. O.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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Inherently safe shutdown of a liquid metal reactor upon a loss of intermediate cooling

Two unprotected (i.e., no scram or plant protection system action) loss-of-heat sink transients are scheduled to be performed on the Experimental Breeder Reactor-II in the Spring of 1986. One is to be initiated from full power (about 60 MW) and the other from half power. The loss-of-heat sink in each test is to be accomplished by essentially stopping the secondary-loop sodium coolant flow in about 20 s. Pretest predictions for the two tests, provided herein, show the reactor to passively shut d… more

Date: January 1, 1986

Creator: Feldman, E.E. & Mohr, D.

Item Type: Report

Partner: UNT Libraries Government Documents Department

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Experimental and calculational results of an irradiation experiment in EBR-II

During the period from 1967 to 1970, several actinide samples were irradiated by ANL in the fast reactor EBR-II for approximately 10/sup 23/ nvt. In 1978, ORNL began an experimental determination of the exposure products in the irradiated samples, and to date, 12 of the samples have already been analyzed. In a previous paper preliminary experimental results for /sup 239/Pu sample were compared with the exposure products predicted by calculations. In this paper, the comparison is extended to sam… more

Date: January 1, 1981

Creator: Gilai, D.; Cooper, J. H.; Laing, W. R.; Raman, S.; Stelson, P. H.; Walker, R. L. et al.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Analysis of carbon transport in the EBR-II and FFTF primary sodium systems

An analysis of the carburization-decarburization behavior of austenitic stainless steels in the primary heat-transport systems of the EBR-II and FFTF has been made that is based upon a kinetic model for the diffusion process and the surface area of steel in contact with flowing sodium at various temperatures in the two systems. The analysis was performed for operating conditions that result in sodium outlet temperatures of 474 and 566/sup 0/C in the FFTF and 470/sup 0/C in the EBR-II. If there … more

Date: January 1, 1976

Creator: Snyder, R. B.; Natesan, K. & Kassner, T. F.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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Effect of preinjected helium on the response of V-20Ti pressurized tubes to neutron irradiation

Vanadium-205 titanium tubes, pressurized to stresses of 34 and 39 MPa, were irradiated in the Experimental Breeder Reactor (EBR-II) at 700/sup 0/C to a displacement damage level of 22 dpa. Sections of the tubes were injected with 15 appM He prior to irradiation to determine ethe effect of helium on the microstructural and creep response of this alloy to irradiation. It was found that helium promoted cavity formation, primarily within existing precipitates, but total swelling remained low. Heliu… more

Date: January 1, 1986

Creator: Vitek, J. M.; Braski, D. N. & Horak, J. A.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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LMFBR in-core instrumentation and in-vessel surveillance systems. Second quarterly report, April--June 1972.

No Description Available.

Date: January 1, 1972

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Performance testing of refractory alloy-clad fuel elements for space reactors

Two fast reactor irradiation tests, SP-1 and SP-2, provide a unique and self-consistent data set with which to evaluate the technical feasibility of potential fuel systems for the SP-100 space reactor. Fuel pins fabricated with leading cladding candidates (Nb-1Zr, PWC-11, and Mo-13Re) and fuel forms (UN and UO/sub 2/) are operated at temperatures typical of those expected in the SP-100 design. The first US fast reactor irradiated, refractory alloy clad fuel pins, from the SP-1 test, reached 1 a… more

Date: January 1, 1985

Creator: Dutt, D.S.; Cox, C.M.; Karnesky, R.A. & Millhollen, M.K.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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Breeder Spent Fuel Handling Program multipurpose cask design basis document

The Breeder Spent Fuel Handling (BSFH) Program multipurpose cask Design Basis Document defines the performance requirements essential to the development of a legal weight truck cask to transport FFTF spent fuel from reactor to a reprocessing facility and the resultant High Level Waste (HLW) to a repository. 1 ref.

Date: September 1, 1985

Creator: Duckett, A. J. & Sorenson, K. B.

Item Type: Report

Partner: UNT Libraries Government Documents Department

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The search for advanced remote technology in fast reactor reprocessing

Research and development in fast reactor reprocessing has been under way about 20 years in several countries throughout the world. During the past decade in France and the United Kingdom, active development programs have been carried out in breeder reprocessing. Actual fuels from their demonstration reactors have been reprocessed in small-scale facilities. Early US work in breeder reprocessing was carried out at the EBR-II facilities with the early metal fuels, and interest has renewed recently… more

Date: January 1, 1990

Creator: Burch, W.D.; Herndon, J.N. & Stradley, J.G.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Recent work on sodium-cooled reactor purification systems

Fast Flux Test Facility (FFTF) cold traps, including the Reactor Primary Cold Trap, the Reactor Secondary Cold Trap and the Closed Loop Cold Trap were fabricated and their thermal performances measured on sodium test systems. The cold traps all met reactor requirements after necessary economizer modifications. Mercury tests were performed to study the effect of woven wire packing on the flow of a liquid metal adjacent to a heated vertical wall as occurs in the cold trap crystallizer. The mesh w… more

Date: February 1, 1978

Creator: Bloom, G.R. & McPheeters, C.C.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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Mechanical properties of highly irradiated 20 percent cold-worked Type 316 stainless steel

Recent experiments have extended the mechanical properties data base of 20 percent cold worked 316 stainless steel to a fluence of 1.3 x 10/sup 23/ n/cm/sup 2/ (E > 0.1 MeV), the goal exposure of the Fast Flux Test Facility. Both uniaxial and biaxial tests were conducted on speciments of developmental cladding which were irradiated in the Experimental Breeder Reactor-II at temperatures ranging from 370/sup 0/C to 650/sup 0/C. Uniaxial tensile tests were conducted at strain rates ranging from 10… more

Date: April 5, 1982

Creator: Hamilton, M. L.; Cannon, N. S. & Johnson, G. D.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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Decontamination and disposition of facilities program

The Program Plan represents the top level guidance document for stating and performing the objectives of the Decontamination and Disposition (D and D) Program. The program objectives, historical data, and program description are presented, along with the top level planning. The planning includes a statement of the funding basis and the corresponding first and second level activity network schedules. The management plan is included which describes the procedures to be used for cost and schedule … more

Date: January 23, 1975

Creator: Graves, A.W.

Item Type: Report

Partner: UNT Libraries Government Documents Department

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Operator decision aid for breached fuel operation in liquid metal cooled nuclear reactors

The purpose of this paper is to report the development of an expert system that provides continuous assessment of the safety significance and technical specification conformance of Delayed Neutron (DN) signals during breached fuel operation. The completed expert system has been parallelized on an innovative distributed-memory network-computing system that enables the computationally intensive kernel of the expert system to run in parallel on a group of low-cost Unix workstations. 1 ref.

Date: January 1, 1991

Creator: Gross, Kenny C.; Hawkins, Robert E. & Nickless, W. K.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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Material control in nuclear fuel fabrication facilities. Addendum to Part I: fuel descriptions and fabrication processes

The addendum contains additional information on existing and planned nuclear fuel fabrication facilities in the world, and descriptions of a selected number of facilities including a conceptual facility which replaces the one initially presented and has been used as a basis for the discussions of material accountability and control. 8 figures, 4 tables.

Date: October 1, 1978

Creator: Borgonovi, Giancarlo M.; McCartin, Timothy J.; Miller, Charles L. & Nguyen, Tien

Item Type: Report

Partner: UNT Libraries Government Documents Department

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A measurement-based method for predicting margins and uncertainties for unprotected accidents in the Integral Fast Reactor concept

A measurement-based method for predicting the response of an LMR core to unprotected accidents has been developed. The method processes plant measurements taken at normal operation to generate a stochastic model for the core dynamics. This model can be used to predict three sigma confidence intervals for the core temperature and power response. Preliminary numerical simulations performed for EBR-2 appear promising. 6 refs., 2 figs.

Date: January 1, 1990

Creator: Vilim, R. B.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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Actinide recycle potential in the IFR (Integral Fast Reactor)

Rising concern about the greenhouse effect reinforces the need to reexamine the question of a next-generation reactor concept that can contribute significantly toward substitution for fossil-based energy generation. Even with only the nuclear capacity on-line today, world-wide reasonably assured uranium resources would last for only about 50 years. If nuclear is to make a significant contribution, breeding is a fundamental requirement. Uranium resources can then be extended by two orders of mag… more

Date: January 1, 1989

Creator: Chang, Y. I.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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CRBRP reactor system shield design and impact of new materials irradiation damage data

Structural materials radiation damage data have received considerable attention in the design of LMFBR permanent reactor structures. A recent FFTF/CRBRP structural materials irradiation experiment, conducted as part of the national program plan for irradiation effects on mechanical properties of LMFBR out-of-core structural materials, has found engineering application in setting energy-dependent neutron exposure limits of irradiated structures. This paper describes the impact of the new irradia… more

Date: January 1, 1980

Creator: Wrights, G.N.

Item Type: Article

Partner: UNT Libraries Government Documents Department