Author: V. E. HazelPublisher:
ISBN:
Category :
Languages : en
Pages : 0
Book Description
POST-IRRADIATION EXAMINATION OF ZIRCALOY-2- AND INCOLOY-800-CLAD FUEL RODS IRRADIATED TO 7000 MWD/TU IN THE CONSUMERS BIG ROCK POINT REACTOR.
Author: V. E. HazelPOST-IRRADIATION EXAMINATION OF ZIRCALOY-2- AND INCOLOY-800-CLAD FUEL RODS IRRADIATED TO 7000 MWd/TU IN THE CONSUMERS BIG ROCK POINT REACTOR.
Author: Scientific and Technical Aerospace Reports
Author: Nuclear Science Abstracts
Author: Fuel Element Experience in Nuclear Power Reactors
Author: Massoud T. SimnadPublisher: Gordon & Breach Publishing Group
ISBN:
Category : Technology & Engineering
Languages : en
Pages : 646
Book Description
Corrosion Abstracts
Author: Publisher:
ISBN:
Category : Corrosion and anti-corrosives
Languages : en
Pages : 728
Book Description
Final Report
Author: J. W. WeberPostirradiation Examination Results for the Irradiation Effects Test Series IE-ST-2, Rod IE-002. [PWR].
![Postirradiation Examination Results for the Irradiation Effects Test Series IE-ST-2, Rod IE-002. [PWR]. PDF](https://books.google.com/books/content/images/frontcover/Txz2wQEACAAJ?fife=w80-h100)
Author: Publisher:
ISBN:
Category :
Languages : en
Pages :
Book Description
A postirradiation examination was conducted on a zircaloy-clad, UO2-fueled, pressurized water reactor (PWR) type rod which had been tested in the Power Burst Facility as part of the Irradiation Effects Test Series of the Thermal Fuels Behavior Program. The fuel rod, previously irradiated to a burnup of 15,800 MWd/t was subjected to a power ramp from 28 to 55 kW/m peak power at an average ramp rate of 4 kW/m/min. Posttest fuel restructuring and relocation, fission product redistribution, and fuel rod cladding deformation were evaluated and analyzed.
Final Report
Author: Publisher:
ISBN:
Category :
Languages : en
Pages : 4
Book Description
One of the early candidate fuel elements for N Reactor use was the coextruded, Zircaloy-2-clad seven-rod cluster. As part of the program of evaluating the seven-rod cluster geometry, three-rod cluster fuel elements, two and three feet long, were irradiated. These long cluster fuel elements were irradiated to determine the distortion (or sag) which might occur at the center of the unsupported length during irradiation. Two three-rod clusters made up of 0.630 inch diameter rods, containing natural uranium cores were irradiated in KER Loop 3. The rods of one cluster were three feet long; the rods of the other were two feet long. The three-feet long rods were supported at their ends and at their midlengths, the two-feet long rods only at their ends. During the irradiation, the maximum core temperature was 435 C. The fuel elements were discharged from the loop after they had reached an exposure of 1400 MWD/T. Following the discharge, the fuel elements were visually examined in the KE view pit. No sag was observed in any of the rods. The test demonstrated that two- and three-feet long rod cluster fuel elements can be irradiated without appreciable sag occurring in the rods.
Embrittlement of Zircaloy-Clad Fuel Rods Irradiated Under Film Boiling Conditions
Author: RR. HobbinsPublisher:
ISBN:
Category : Cladding
Languages : en
Pages : 14
Book Description
Pressurized water reactor type fuel rods are being irradiated under postulated accident conditions in the Power Burst Facility at the Idaho National Engineering Laboratory as part of the Thermal Fuels Behavior Program. In these tests, film boiling was achieved by establishing a mismatch reactor power and coolant flow, while maintaining the coolant pressure at 15 MPa.