27 research outputs found
Neutronic performance of high-density LEU fuels in water-moderated and water-reflected research reactors
At the Reduced Enrichment for Research and Test Reactors (RERTR) meeting in September 1994, Durand reported that the maximum uranium loading attainable with U{sub 3}Si{sub 2} fuel is about 6.0 g U/cm{sup 3}. The French Commissariat a l`Energie Atomique (CEA) plan to perform irradiation tests with 5 plates at this loading. Compagnie pour L`Etude et La Realisation de Combustibles Atomiques (CERCA) has also fabricated a few uranium nitride (UN) plates with a uranium density in the fuel meat of 7.0 g/cm{sup 3} and found that UN is compatible with the aluminum matrix at temperatures below 500 C. High density dispersion fuels proposed for development include U-Zr(4 wt%)-Nb(2 wt%), U-Mo(5 wt%), and U-Mo(9 wt%). The purpose of this note is to examine the relative neutronic behavior of these high density fuels in a typical light water-reflected and water-moderated MTR-type research reactor. The results show that a dispersion of the U-Zr-Nb alloy has the most favorable neutronic properties and offers the potential for uranium densities greater than 8.0 g/cm{sup 3}. On the other hand, UN is the least reactive fuel because of the relatively large {sup 14}N(n,p) cross section. For a fixed value of k{sub eff}, the required {sup 235}U loading per fuel element is least for the U-Zr-Nb fuel and steadily increases for the U-Mo(5%), U-Mo(9%), and UN fuels. Because of volume fraction limitations, the UO{sub 2} dispersions are only useful for uranium densities below 5.0 g/cm{sup 3}. In this density range, however, UO{sub 2} is more reactive than U{sub 3}Si{sub 2}
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Evaluation of differential shim rod worth measurements in the Oak Ridge Research Reactor
Reasonable agreement between calculated and measured differential shim rod worths in the Oak Ridge Research Reactor (ORR) has been achieved by taking into account the combined effects of negative reactivity contributions from changing fuel-moderator temperatures and of delayed photoneutrons. A method has been developed for extracting the asymptotic period from the shape of the initial portion of the measured time-dependent neutron flux profile following a positive reactivity insertion. In this region of the curve temperature-related reactivity feedback effects are negligibly small. Results obtained by applying this technique to differential shim rod worth measurements made in a wide variety of ORR cores are presented
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Blackness coefficients, effective diffusion parameters, and control rod worths for thermal reactors
Simple diffusion theory cannot be used to evaluate control rod worths in thermal reactors because of the strongly absorbing character of the control material. However, good results can be obtained from a diffusion calculation by representing the absorber slab by means of a suitable pair of internal boundary conditions, ..cap alpha.. and ..beta.., which are ratios of neutron flux to neutron current. Methods for calculating ..cap alpha.. and ..beta.. in the P/sub 1/, P/sub 3/, and P/sub 5/ approximations, with and without scattering, are presented. By appropriately weighting the fine-group blackness coefficients, broad group values, and , are obtained. The technique is applied to the calculation of control rod worths of Cd, Ag-In-Cd, and Hf control elements. Results are found to compare very favorably with detailed Monte Carlo calculations. For control elements whose geometry does not permit a thin slab treatment, other methods are needed for determining the effective diffusion parameters. One such method is briefly discussed and applied to the calculation of control rod worths in the Ford Nuclear Reactor at the University of Michigan. Calculated and measured worths are found to be in good agreement
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Blackness coefficients, effective diffusion parameters, and control rod worths for thermal reactors - methods
Simple diffusion theory cannot be used to evaluate control rod worths in thermal neutron reactors because of the strongly absorbing character of the control material. However, reliable control rod worths can be obtained within the framework of diffusion theory if the control material is characterized by a set of mesh-dependent effective diffusion parameters. For thin slab absorbers the effective diffusion parameters can be expressed as functions of a suitably-defined pair of blackness coefficients. Methods for calculating these blackness coefficients in the P/sub 1/, P/sub 3/, and P/sub 5/ approximations, with and without scattering, are presented. For control elements whose geometry does not permit a thin slab treatment, other methods are needed for determining the effective diffusion parameters. One such method, based on reaction rate ratios, is discussed
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Transition from HEU to LEU fuel in Romania's 14-MW TRIGA reactor
The 14-MW TRIGA steady state reactor (SSR) located in Pitesti, Romania, first went critical in the fall of 1979. Initially, the core configuration for full power operation used 29 fuel clusters each containing a 5 {times} 5 square array of HEU (10 wt%) -- ZrH -- Er (2.8 wt%) fuel-moderator rods (1.295 cm o.d.) clad in Incology. With a total inventory of 35 HEU fuel clusters, burnup considerations required a gradual expansion of the core from 29 to 32 and finally to 35 clusters before the reactor was shut down because of insufficient excess reactivity. At this time each of the original 29 fuel clusters had an overage {sup 235}U burnup in the range from 50 to 62%. Because of the US policy regarding the export of highly enriched uranium, fresh HEU TRIGA replacement fuel is not available. After a number of safety-related measurements, the SSR is expected to resume full power operation in the near future using a mixed core containing five LEU TRIGA clusters of the same geometry as the original fuel but with fuel-moderator rods containing 45 wt% U (19.7% {sup 235}U enrichment) and 1.1 wt% Er. Rods for 14 additional LEU fuel clusters will be fabricated by General Atomics. In support of the SSR mixed core operation numerous neutronic calculations have been performed. This paper presents some of the results of those calculations
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Comparison of calculated quantities with measured quantities for the LEU-fueled Ford Nuclear Reactor
The Ford Nuclear Reactor (FNR) went critical on December 8, 1981 with 23 LEU fuel elements. Five of these 23 elements were fabricated by CERCA and the others by NUKEM. Since that time a substantial data base of experimental results for LEU cores has been accumulated by the University of Michigan FNR staff. This paper compares some of the experimental data with analytical calculations based, for the most part, on three-dimensional diffusion theory. The critical configuration, control rod worths, axial rhodium reaction rate profiles and thermal flux distributions have been calculated and compared with measurements
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The ORR Whole-Core LEU Fuel Demonstration
The ORR Whole-Core LEU Fuel Demonstration, conducted as part of the US Reduced Enrichment Research and Test Reactor Program, has been successfully completed. Using commercially-fabricated U{sub 3}Si{sub 2}-Al 20%-enriched fuel elements (4.8 g U/cc) and fuel followers (3.5 g U/cc), the 30-MW Oak Ridge Research Reactor was safely converted from an all-HEU core, through a series of HEU/LEU mixed transition cores, to an all-LEU core. There were no fuel element failures and average discharge burnups were measured to be as high as 50% for the standard elements and 75% for the fuel followers. Experimental results for burnup-dependent critical configurations, cycle-averaged fuel element powers, and fuel-element-averaged {sup 235}U burnups validated predictions based on three-dimensional depletion calculations. Calculated values for plutonium production and isotopic mass ratios as functions of {sup 235}U burnup support the corresponding measured quantities. In general, calculations for reaction rate distributions, control rod worths, prompt neutron decay constants, and isothermal temperature coefficients were found to agree with corresponding measured values. Experimentally determined critical configurations for fresh HEU and LEU cores radially reflected with water and with beryllium are well-predicted by both Monte Carlo and diffusion calculations. 17 refs
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Calibrations of a tritium extraction facility
A tritium extraction facility has been built for the purpose of measuring the absolute tritium concentration in neutron-irradiated lithium metal samples. Two independent calibration procedures have been used to determine what fraction, if any, of tritium is lost during the extraction process. The first procedure compares independently measured /sup 4/He and /sup 3/H concentrations from the /sup 6/Li(n,..cap alpha..)T reaction. The second procedure compared measured /sup 6/Li(n,..cap alpha..)T//sup 197/Au (n,..gamma..)/sup 198/Au thermal neutron reaction rate ratios with those obtained from Monte Carlo calculations using well-known cross sections. Both calibration methods show that within experimental errors (approx. 1.5%) no tritium is lost during the extraction process
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Analytical analyses of startup measurements associated with the first use of LEU fuel in Romania's 14-MW TRIGA reactor
The 14-MW TRIGA steady state reactor (SSR) is located in Pitesti, Romania. Beginning with an HEU core (10 wt% U), the reactor first went critical in November 1979 but was shut down ten years later because of insufficient excess reactivity. Last November the Institute for Nuclear Research (INR), which operates the SSR, received from the ANL RERTR program a shipment of 125 LEU pins fabricated by General Atomics and of the same geometry as the original fuel but with an enrichment of 19.7% 235U and a loading of 45 wt% U. Using 100 of these pins, four LEU clusters, each containing a 5 x 5 square array of fuel rods, were assembled. These four LEU clusters replaced the four most highly burned HEU elements in the SSR. The reactor resumed operations last February with a 35-element mixed HEU/LEU core configuration. In preparation for full power operation of the SSR with this mixed HEU/LEU core, a number of measurements were made. These included control rod calibrations, excess reactivity determinations, worths of experiment facilities, reaction rate distributions, and themocouple measurements of fuel temperatures as a function of reactor power. This paper deals with a comparison of some of these measured reactor parameters with corresponding analytical calculations