Gas retention in irradiated beryllium
Abstract
Helium (an inert gas) with low solubility in beryllium is trapped in irradiated beryllium at low temperatures (<100{degree}C) while the tritium generated may have some mobility and be released. The subject of tritium retention in irradiated beryllium within fusion reactor blankets is of considerable interest in their conceptual design. Results from experiments on three sets of irradiated beryllium specimens are examined in this paper. The beryllium specimens were irradiated at abut 75{degree}C in capsules to protect them from the cooling water. One set of samples was irradiated to {approximately}3 {times} 10{sup 22} n/cm{sup 2} (E > 1 MeV). In these samples the calculated helium generated was {approximately} 14,000 appm. They are described in terms of swelling, annealing, microstructure, and helium bubble behavior (size, density and mobility). A second sample was irradiated to {approximately}5 {times} 10{sup 22} n/cm{sup 2} (E > 1 MeV). In that one the calculated helium and tritium generated were {approximately}24,000 appm He and {approximately}3720 appm, and tritium content was examined in a dissolution experiment. Most of the tritium was released as gas to the glovebox indicating the generated tritium was retained in the helium bubbles. In a third set of experiments a specimen was examined by annealingmore »
- Authors:
-
- (EG and G Idaho, Inc., Idaho Falls, ID (USA))
- (Sandia National Labs., Albuquerque, NM (USA))
- Publication Date:
- Research Org.:
- EG and G Idaho, Inc., Idaho Falls, ID (USA)
- Sponsoring Org.:
- DOE/ER
- OSTI Identifier:
- 6133872
- Report Number(s):
- EGG-FSP-9125
ON: DE91006206; TRN: 91-001931
- DOE Contract Number:
- AC07-76ID01570
- Resource Type:
- Technical Report
- Country of Publication:
- United States
- Language:
- English
- Subject:
- 70 PLASMA PHYSICS AND FUSION TECHNOLOGY; 36 MATERIALS SCIENCE; BERYLLIUM; SWELLING; HELIUM; SOLUBILITY; TRITIUM; ANNEALING; BREEDING BLANKETS; BUBBLES; IRRADIATION; THERMONUCLEAR REACTOR MATERIALS; ALKALINE EARTH METALS; BETA DECAY RADIOISOTOPES; BETA-MINUS DECAY RADIOISOTOPES; ELEMENTS; FLUIDS; GASES; HEAT TREATMENTS; HYDROGEN ISOTOPES; ISOTOPES; LIGHT NUCLEI; MATERIALS; METALS; NONMETALS; NUCLEI; ODD-EVEN NUCLEI; RADIOISOTOPES; RARE GASES; REACTOR COMPONENTS; YEARS LIVING RADIOISOTOPES; 700209* - Fusion Power Plant Technology- Component Development & Materials Testing; 360106 - Metals & Alloys- Radiation Effects
Citation Formats
Beeston, J.M., Miller, L.G., Longhurst, G.R., and Causey, R.A. Gas retention in irradiated beryllium. United States: N. p., 1990.
Web. doi:10.2172/6133872.
Beeston, J.M., Miller, L.G., Longhurst, G.R., & Causey, R.A. Gas retention in irradiated beryllium. United States. doi:10.2172/6133872.
Beeston, J.M., Miller, L.G., Longhurst, G.R., and Causey, R.A. Fri .
"Gas retention in irradiated beryllium". United States.
doi:10.2172/6133872. https://www.osti.gov/servlets/purl/6133872.
@article{osti_6133872,
title = {Gas retention in irradiated beryllium},
author = {Beeston, J.M. and Miller, L.G. and Longhurst, G.R. and Causey, R.A.},
abstractNote = {Helium (an inert gas) with low solubility in beryllium is trapped in irradiated beryllium at low temperatures (<100{degree}C) while the tritium generated may have some mobility and be released. The subject of tritium retention in irradiated beryllium within fusion reactor blankets is of considerable interest in their conceptual design. Results from experiments on three sets of irradiated beryllium specimens are examined in this paper. The beryllium specimens were irradiated at abut 75{degree}C in capsules to protect them from the cooling water. One set of samples was irradiated to {approximately}3 {times} 10{sup 22} n/cm{sup 2} (E > 1 MeV). In these samples the calculated helium generated was {approximately} 14,000 appm. They are described in terms of swelling, annealing, microstructure, and helium bubble behavior (size, density and mobility). A second sample was irradiated to {approximately}5 {times} 10{sup 22} n/cm{sup 2} (E > 1 MeV). In that one the calculated helium and tritium generated were {approximately}24,000 appm He and {approximately}3720 appm, and tritium content was examined in a dissolution experiment. Most of the tritium was released as gas to the glovebox indicating the generated tritium was retained in the helium bubbles. In a third set of experiments a specimen was examined by annealing at a succession of temperatures to more than 600{degree}C for tritium release. In the temperature range of 300--500{degree}C little release (0.01--0.4%) occurred, but there was a massive release at just over 600{degree}C. Theories of swelling appear to adequately describe bubble behavior with breakaway release occurring at high helium contents and at large bubble diameters. 8 refs., 6 figs.},
doi = {10.2172/6133872},
journal = {},
number = ,
volume = ,
place = {United States},
year = {Fri Jun 01 00:00:00 EDT 1990},
month = {Fri Jun 01 00:00:00 EDT 1990}
}
-
Fission gas retention and axial expansion of irradiated metallic fuel
Direct electrical heating and infrared heating techniques were used in out-of-reactor experiments on irradiated metallic fuel. The results indicate that accelerated expansion can occur during thermal transients and that the accelerated expansion is driven by retained fission gases. The results also demonstrate that gas retention and, hence, expansion behavior is a function of axial position within the pin. 12 refs., 12 figs.