Dimensional Behavior of the Experimental Gas-cooled Reactor Fuel Element at Elevated Temperatures PDF Download
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Author: W. R. Martin
Publisher:
ISBN:
Category : Nuclear reactions
Languages : en
Pages : 56
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Author: W. R. Martin
Publisher:
ISBN:
Category : Nuclear reactions
Languages : en
Pages : 56
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Book Description
Author:
Publisher:
ISBN:
Category :
Languages : en
Pages :
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The Experimental Gas-Cooled Reactor (EGCR) fuel assemblies consist of a cluster of seven fuel elements contained in a graphite sleeve. Each element is composed of hollow cylindrical UO2 pellets encapsulated in a type 304 stainless steel tube. The dimensional behavior of the fuel element was determined in an apparatus which simulated the thermal conditions predicted for the EGCR. Particular emphasis was placed on determining the relationship between the fuel temperature and axial expansion, the radial expansion characteristics of the fuel, the effect of cladding and fuel interaction on heating and subsequent cooling, the effect of rapid temperature excursions on the degradation of the fuel, and accumulative effects in the fuel element due to thermal cycling. An element that contains a radial gap between the cladding and the fuel pellet was found to respond to thermal cycling in the same manner that the individual components would react if subjected to the same thermal conditions and tested separately. Both the axial and radial expansion of the fuel pellet are very nearly a function of the maximum central temperature. The axial expansion of the fuel pellet column can be reduced appreciably at elevated temperatures by "dishing" the ends of the pellets. The pellets fracture radially and circumferentially upon heating, but redistribution of the fuel does not occur. If no radial gap exists between the fuel and the cladding, the expansion characteristics of the element during thermal cycling are a function of the fuel temperature, cladding temperature, and the external pressure exerted on the ele ment by the coolant stream. Thermal cycling may introduce plastic axial strains intc the cladding depending upon the details of the temperature cycle and the pressure conditions. (auth).
Author:
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Category :
Languages : en
Pages :
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Research and development progress specifically directed toward the construction of a 40-Mw(e) prototype power plant employing a high-temperature, gas-cooled, graphitemoderated reactor known as the HTGR is reported. Irradiation of element III-B in the in-pile loop continued satisfactorily. The element has generated a total of l36.3 Mw-hr of fission heat. The gross activity in the purge stream increased slightly to about 350 mu C/cm/sup 3/. By taking larger gas samples than were previously taken, a value of 0.02 VC/cm/sup 3/ was obtained for the gross activity of the primary loop. Element III-A, which was removed from the loop after generating 133 Mw-hr of fission heat, was disassembled and examined. No fuel-compact damage of any type was visible. Determination of the distribution of fission products in the element is under way, Fissionproduct- release data for in-pile-loop element III-A were calculated. During the 133 Mw- hr of operation, the release fraction increased by approximately one order of magnitude. Also calculated were the xenon and krypton release data for the first 100 Mw-hr of III-B operation. The release rate for the longer-lived isotopes increased bv about a factor of 10 and that of the shorter-lived isotopes by about a factor of 100. A test was run in which the in-pileloop purge flow, was stopped. The primariy-loop activity level rose sharply during the first hour, increased at a slower rate for the next 11 hr, and then appeared to level off. When purge flow was resumed, the gross activity in the primary loop was cleaned up with a half life of about 2.2 hr. An attempt was made to identify Cs/sup 137/ and Ba/ sup 140/ plateout in portions of the in-pile loop. A very small amount of cesium (less than a monolayer) was found, but no barium could be detected. The validity of two basic assumptions made in the one-dimensional burnup code FEVER was investigated. As a result of extensive lifetime studies and power-distribution and temperaturecoefficient calculations, the initial fuel loading for the Peach Bottom core was specified. A series of control-rodworth calculations and a recalculation of the postulated rod-fall accident were made for this loading. The test of the prototype control rod and drive was satisfactorily coNonempleted during the quarter. During the course of the test the drive completed 590,641 starts and stops, 5,756 scrams, and more than 2.6 million inches of random regulating motion in helium at reactor temperatures. These totals far exceed the expected life requirements of the system. Preparations are being made for testing the prototype emergency shutdown rod and drive. The apparatus for the barium permeation experiment with a full-diameter sleeve was completed, and preliminary calibration runs were started. Following these runs, the system will be operated until an equilibrium distribution of barium is reached. At that time, a series of corings will be made on all of the compacts and the sleeve to evaluate the overall barium and strontium distribution. Other experiments on barium behavior, including permeation experiments with reducedscale fuel elements and exVeriments on the vaporization, sorption, and diffusion of barium, were continued, and the data are being analyzed. Measurements were made to compare the room-temperature back diffusion of argon, krypton, and xenon through a sample of sleeve graphite against a helium pressure difference. The results show that the difference between the effective back-diffusion coefficients of krypton and xenon seems to increase with increasing helium pressure difference across the sleeve. The argon and krypton back-diffusion data at an average pressure of 3 atm are essentially the same, A FORTRAN code was written to recalculate the retention of neutron poison material and fission products in the core as well as their condensation on and revaporization from the upper reflector following a complete loss-of-coolant-circulation accident. (auth).
Author:
Publisher:
ISBN:
Category : Energy development
Languages : en
Pages : 236
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Author:
Publisher:
ISBN:
Category : Nuclear energy
Languages : en
Pages : 236
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Author:
Publisher:
ISBN:
Category : Nuclear energy
Languages : en
Pages : 700
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Author: R. E. Dahl
Publisher:
ISBN:
Category : Gas cooled reactors
Languages : en
Pages : 94
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Author: Oak Ridge National Laboratory
Publisher:
ISBN:
Category : Government publications
Languages : en
Pages : 98
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Author: U.S. Atomic Energy Commission
Publisher:
ISBN:
Category : Light water reactors
Languages : en
Pages : 180
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Author: W. R. Martin
Publisher:
ISBN:
Category : Nuclear fuel elements
Languages : en
Pages : 28
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