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IRRADIATION BEHAVIOR OF HIGH-BURNUP URANIUM--PLUTONIUM ALLOY PROTOTYPE FUEL ELEMENTS.

IRRADIATION BEHAVIOR OF HIGH-BURNUP URANIUM--PLUTONIUM ALLOY PROTOTYPE FUEL ELEMENTS. PDF Author:
Publisher:
ISBN:
Category :
Languages : en
Pages :

Book Description


IRRADIATION BEHAVIOR OF HIGH-BURNUP URANIUM--PLUTONIUM ALLOY PROTOTYPE FUEL ELEMENTS.

IRRADIATION BEHAVIOR OF HIGH-BURNUP URANIUM--PLUTONIUM ALLOY PROTOTYPE FUEL ELEMENTS. PDF Author:
Publisher:
ISBN:
Category :
Languages : en
Pages :

Book Description


The Irradiation Behavior of High-burnup Uranium-plutonium Alloy Prototype Fuels Elements

The Irradiation Behavior of High-burnup Uranium-plutonium Alloy Prototype Fuels Elements PDF Author: W. N. Beck
Publisher:
ISBN:
Category :
Languages : en
Pages : 47

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Nuclear Science Abstracts

Nuclear Science Abstracts PDF Author:
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Category : Nuclear energy
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Irradiation Behavior of Uranium Carbide Fuels

Irradiation Behavior of Uranium Carbide Fuels PDF Author: D. I. Sinizer
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Category : Nuclear fuels
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Scientific and Technical Aerospace Reports PDF Author:
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Nuclear Science Abstracts PDF Author:
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Category : Nuclear energy
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Plutonium Research Program PDF Author: U.S. Atomic Energy Commission. Plutonium Research Coordinating Committee
Publisher:
ISBN:
Category : Plutonium
Languages : en
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The Effects of Irradiation on Uranium-plutonium-fissium Fuel Alloys

The Effects of Irradiation on Uranium-plutonium-fissium Fuel Alloys PDF Author: J. A. Horak
Publisher:
ISBN:
Category : Alloys
Languages : en
Pages : 40

Book Description
A total of 35 specimens of U-Pu-fissium alloy and 2 specimens of U-10 wt% Pu-5 wt% Mo alloy were irradiated as a part of the fuel-alloy development program for fast breeder reactors at Argonne National Laboratory. Total atom burnups ranged from 1.0 to 1.8% at maximum fuel temperatures ranging from 230 to 470 deg C. Emphasis was placed on the EBR-II Core-III reference fuel material, which is an injection-cast, U-20 wt% Pu-10 wt% fissium alloy. It was found that this material begins to swell catastrophically at irradiation temperatures above 370 deg C. The ability of the fuel to resist swelling did not appear to vary appreciably with minor changes in zirconium or fissium content. Decreasing the Pu to 10 wt%, however, significantly improved the swelling behavior of the alloy. Both pour-cast and thermally cycled material and pour-cast, extruded, and thermally cycled material appeared to be more stable under irradiation than injection-cast material. Under comparable irradiation conditions, the specimens of U-20 wt% Pu- 5 wt% Mo alloy were less dimensionally stable than the U-Pu-fissium alloys investigated.

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Publisher:
ISBN:
Category : Nuclear engineering
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