DAMAGING NEUTRON EXPOSURE CRITERIA FOR EVALUATING THE EMBRITTLEMENT OF REACTOR PRESSURE VESSEL STEELS IN DIFFERENT NEUTRON SPECTRA.
Abstract
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 moderated reactor environments as well as in graphite moderated reactor environments. The radiation-induced transition temperature or nil-ductility transition (NDT) temperature increases of the several steels involved are presented versus n/cm2 determined by each of the following techniques: (a) assumption of a fission spectrum, extrapolation of activation data induced at a high Mev threshold to 1 Mev, and reporting exposure > 1 Mev, and (b) calculation of spectra used to determine activation cross section for exposures above energy limits of 1, 0.5, and 0.183 Mev. The differences observed by this analysis were intercompared in relation to absolute magnitude as well as in terms of engineering significance. By applying these criteria to data relating directly to a pressurized light water power reactor, benefits to the lifetime of the reactor can be realized. The results of this study to date indicate that data relating to the properties of steels irradiated in or near the core of pressurized light water moderated reactors can be confidently intercompared for engineering applications assuming a fission spectrum and accounting for neutrons of energies >1 Mev. On the other hand, calculated spectra and average cross section adjustment to an energy limit as low as >0.183 Mev must be applied to data from highly moderated reactors. (Author)
Document Details
- Document Type
- Technical Report
- Publication Date
- Jul 28, 1966
- Accession Number
- AD0639748
Entities
People
- Charles Z. Serpan Jr.
- Lendell E. Steele
Organizations
- United States Naval Research Laboratory