Author:
Publisher:
ISBN:
Category :
Languages : en
Pages :
Book Description
IRRADIATION BEHAVIOR OF HIGH-BURNUP URANIUM--PLUTONIUM ALLOY PROTOTYPE FUEL ELEMENTS.
The Irradiation Behavior of High-burnup Uranium-plutonium Alloy Prototype Fuels Elements
Nuclear Science Abstracts
Irradiation Behavior of Uranium Carbide Fuels
Author: D. I. Sinizer
Publisher:
ISBN:
Category : Nuclear fuels
Languages : en
Pages : 52
Book Description
Publisher:
ISBN:
Category : Nuclear fuels
Languages : en
Pages : 52
Book Description
Scientific and Technical Aerospace Reports
Nuclear Science Abstracts
Plutonium Research Program
Author: U.S. Atomic Energy Commission. Plutonium Research Coordinating Committee
Publisher:
ISBN:
Category : Plutonium
Languages : en
Pages : 104
Book Description
Publisher:
ISBN:
Category : Plutonium
Languages : en
Pages : 104
Book Description
The Effects of Irradiation on Uranium-plutonium-fissium Fuel Alloys
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.
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.
Transactions of the American Nuclear Society
Author: American Nuclear Society
Publisher:
ISBN:
Category : Nuclear engineering
Languages : en
Pages : 662
Book Description
Publisher:
ISBN:
Category : Nuclear engineering
Languages : en
Pages : 662
Book Description