Atomic Energy of Canada Limited

About: Atomic Energy of Canada Limited is a company organization based out in Ottawa, Ontario, Canada. It is known for research contribution in the topics: Neutron & Zirconium alloy. The organization has 4845 authors who have published 4826 publications receiving 102951 citations.

Electrolytic hydrogen generation using CANDU nuclear reactors

Abstract: Hydrogen generation on a large scale will be required to power the emerging hydrogen economy without emitting large amounts of carbon dioxide into the atmosphere. All large-scale hydrogen production facilities currently in operation use processes such as steam-methane reforming or partial oxidation of heavy hydrocarbons, which emit large amounts of carbon dioxide. Atomic Energy of Canada Ltd. (AECL) is studying several process options for large-scale production of hydrogen without emitting carbon dioxide. One of these options is to produce hydrogen via steam electrolysis, where the heat and electricity are produced by nuclear reactors. This paper investigates the technical issues and economics of hydrogen production by integrating a steam electrolysis system with the Advanced CANDU Reactor ACR-1000. The overall thermal-to-hydrogen efficiency of the combined ACR-1000 and steam electrolysis process is 33-34% compared to about 27% for conventional water electrolysis. The heat recovered from the steam electrolysis loop is used to provide heat at appropriate locations in the electricity generation cycle.

Evolution of Microstructure in Zirconium Alloys During Irradiation

Abstract: X-ray diffraction (XRD) and transmission electron microscopy (TEM) have been used to characterize microstructural and microchemical changes produced by neutron irradiation in zirconium and zirconium alloys. Zircaloy-2, Zircaloy-4, and Zr-2.5Nb alloys with differing metallurgical states have been analyzed after irradiation for neutron fluences up to 25 {times} 10{sup 25} n.m{sup {minus}2} for a range of temperatures between 330 and 580 K. Irradiation modifies the dislocation structure through nucleation and growth of dislocation loops and, for cold-worked materials in particular, climb of existing network dislocations. in general, the a-type dislocation structure tends to saturate at low fluences. The c-component dislocation structure, however, may evolve over long periods of irradiation. The phase structure is also modified by irradiation. The common alloying/impurity elements, Fe, Cr, and Ni, are relatively insoluble in the {alpha}-phase during irradiation irrespective of the state of the phase initially containing these elements, i.e., metastable {beta}-phase or stable intermetallic precipitate. The stable intermetallic particles may undergo structural changes dependent on their composition and the temperature. For the metastable dual-phase {alpha}/{beta}-alloys (Zr-2.5Nb alloy), the {beta}-phase structure is modified during irradiation, but the change is complex, being a combination of thermal decomposition and radiation-induced mixing.

Upper limits on the mixing of heavy neutrinos in the beta decay of 63Ni.

Abstract: The beta spectrum of $^{63}\mathrm{Ni}$ was studied with a \ensuremath{\pi} \ensuremath{\surd}2 beta spectrometer, a multistrip source, and multidetector array. No evidence was found for heavy neutrinos with masses ranging from 4 to 40 keV. The upper limit, including numerical evaluation of systematic errors, for the mixing of a 17 keV neutrino was set at 0.3% (90% confidence level). The beta spectrum endpoint was found to be 66.946\ifmmode\pm\else\textpm\fi{}0.020 keV, about 1 keV higher than the value derived from the 1983 atomic mass table, 65.92\ifmmode\pm\else\textpm\fi{}0.15 keV.

Hydrogen Absorption and the Lifetime Performance of Titanium Nuclear Waste Containers

Abstract: For the conditions expected in a Canadian nuclear waste vault, the two corrosion processes most likely to lead to the failure of titanium waste containers are crevice corrosion and/or hydrogen induced cracking (HIC). In this report the processes likely to lead to hydrogen absorption by titanium alloys (Grades-2, -12, -16), and hence to render them susceptible to HIC are discussed. The possible paths to container failure via a combination of crevice corrosion, general passive corrosion and HIC are described and a criterion for container failure by HIC is defined. The modelling procedures developed to predict the consequences of hydrogen absorption are described, and the experimental methods used to measure required modelling parameters are discussed and the results summarized. The predictions of these models show that for alloys on which crevice corrosion either does not occur or is limited in extent (Ti-16 and Ti-12), container lifetimes of >10 a are achievable. As a result only those containers assumed to fail prematurely due to the presence of manufacturing defects need be included in assessment calculations.