"PIN-CUSHION" IRRADIATION TESTS OF URANIUM AND ITS ZIRCONIUM ALLOYS. Metallurgy Program 6.1.1

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Published: 1956

Total Pages:

ISBN-13:

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ABS>An irradiation test of 1/16 in. diameter x 3/16 in. long U/sup 235/ and U/sup 235/-Zr alloy specimens, in the range 2 to 20 wt.% Zr, is described. The small pins were mounted in heat transfer blocks, or ''cushions, '' in such a way that suriace roughening and length changes could be observed, recorded, and compared for five stabilizing heat treatments up to exposure levels of approximately 1% total atom burnup. Addition of 2% Zr was found to be very beneficial, but the behavior of the alloys underwent distinct changes in going from 2 to 4 wt.% Zr. Material, both as-cast and subsequently isothermally treated, changed from minus to plus elongation, whereas wrought materials, either gamma-slow cooled or isothermally treated, changed from plus to minus behavior. Data-quenched wrought material varied unpredictably. Addition of Zr up to 20% showed further improvement in stability for all treatments. Factors which may have contributed to this behavior are discussed. (auth).

"Pin-cushion" Irradiation of Cast Uranium-plutonium Alloy Specimens

Author: S. H. Paine

Publisher:

Published: 1958

Total Pages: 36

ISBN-13:

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The irradiation behavior of small pin specimens of chill-cast U-10 wt.% Pu and U-15 wt.% Pu specimens projecting from heat transfer cushions is described. The specimens were coated with a thin layer of carbonyl-deposited nickel. The 15 wt.% Pu pins showed excellent dimensional stability at 1/3 wt.% total burnup. The 10 wt.% Pu specimens were somewhat inferior, failure of one of them indicating that the improvement in performance over unalloyed uranium may be due to the restraint imposed by the cladding.

Irradiation of Extrusion-clad Uranium-2 W/o Zirconium Alloy for EBR-1, Mark III
Irradiation of Extrusion-clad Uranium-2 W/o Zirconium Alloy for EBR-1, Mark III

Author: J. H. Kittel

Publisher:

Published: 1959

Total Pages: 40

ISBN-13:

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The fuel material specified for the Mark III core of EBR-I was uranium-2 wt. % zirconium alloy coextruded with Zircaloy-2 cladding. From previous work on swaged or rolled uranium-2 wt% zirconium alloy, it was anticipated that the extruded alloy would be dimensionally unstable under irradiation unless stabilized by suitable heat treatment. In order to determine an effective heat treatment, irradiation studies were made on both clad and unclad extruded uranium-2 wt.% zirconium alloy specimens at irradiation temperature estimated at 200 to 750 deg C. The irradiation specimens included material with three different heat treatments, selected on the basis of previous studies, and material transient melted in its cladding. For unclad specimens, it was found that the irradiation temperature strongly influenced the various irradiation growth rates resulting from different heat treatments. Growth rates of the clad specimens were relatively insensitive to either irradiation temperature or prior heat treatment. An exception was the transient-melted material, which shortened under irradiation. The cladding had only limited ability to restrain the swelling rates of specimens irradiated at the more elevated temperatures. Clad transient-melted material was found to be most resistant to high-temperature swelling under irradiation. The results of the present study combined with observations in earlier investigations resulted in a recommendation that the reference heat treatment for the core consist of gamma solution at 800 deg C followed by isothermal transformation at 690 deg C.

Advances in High Temperature Gas Cooled Reactor Fuel Technology
Advances in High Temperature Gas Cooled Reactor Fuel Technology

Author: International Atomic Energy Agency

Publisher:

Published: 2012-06

Total Pages: 639

ISBN-13: 9789201253101

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This publication reports on the results of a coordinated research project on advances in high temperature gas cooled reactor (HTGR) fuel technology and describes the findings of research activities on coated particle developments. These comprise two specific benchmark exercises with the application of HTGR fuel performance and fission product release codes, which helped compare the quality and validity of the computer models against experimental data. The project participants also examined techniques for fuel characterization and advanced quality assessment/quality control. The key exercise included a round-robin experimental study on the measurements of fuel kernel and particle coating properties of recent Korean, South African and US coated particle productions applying the respective qualification measures of each participating Member State. The summary report documents the results and conclusions achieved by the project and underlines the added value to contemporary knowledge on HTGR fuel.

Challenges for Coolants in Fast Neutron Spectrum Systems
Challenges for Coolants in Fast Neutron Spectrum Systems

Author: International Atomic Energy Agency

Publisher:

Published: 2020-07-30

Total Pages: 268

ISBN-13: 9789201078209

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This publication evaluates the different coolant options considered for nuclear applications with a fast neutron spectrum (i.e. fusion, fission and accelerators), compiles the latest information in the field and identifies research needs.