Published: Jan 1967
| ||Format||Pages||Price|| |
|PDF (456K)||31||$25||  ADD TO CART|
|Complete Source PDF (15M)||31||$148||  ADD TO CART|
Irradiation experiments performed to define behavior trends of materials for use as reactor structural components are frequently accomplished in reactor environments dissimilar to those expected during actual service. In order to accurately assess the damage produced in experimental environments for application of results to operating reactor cases, a damaging-neutron exposure criterion must be established which will account for the significantly damaging portion of the incident neutron spectra of both reactor environments. Several such exposure criteria have been evaluated through use of the results of metallurgical tests of reference steel specimens after irradiation in light- and heavy-water as well as graphite moderated reactor environments. The radiation-induced transition temperature or nil ductility transition (NDT) temperature increases of the several steels involved are presented versus neutrons per square centimeter determined by each of the following techniques: (1) assumption of a fission spectrum, extrapolation of activation data induced at a high Mev threshold to 1 Mev and reporting exposure > 1 Mev, and (2) calculated spectra utilized to determine activation cross section for exposures above energy limits 1, 0.5, and 0.183 Mev. The differences observed by this analysis are intercompared and are discussed in relation to absolute magnitude as well as in terms of engineering significance. The benefits which can be attained by applying these criteria to data relating directly to a pressurized, light water power reactor are indicated. Conclusions are drawn concerning the value of each neutron-exposure criterion as it affects radiation damage research results and their application.
neutrons, embrittlement, nuclear reactors, pressure vessels, steels, radiation effects, nuclear radiation
Serpan, C. Z.
Research chemist, Metallurgy Div., U.S. Naval Research Laboratory, Washington, D. C.
Steele, L. E.
Reactor Development and Technology Div., U.S. Atomic Energy Commission, Germantown, Md.
Paper ID: STP41339S