Development of very-high-density low-enriched-uranium fuels

Nuclear Engineering and Design

Volumn 178, Issue 1, 1997, Pages 119-126

  Snelgrove, J L ^a   Hofman, G L ^a   Meyer, M K ^a   Trybus, C L ^a   Wiencek, T C ^a  

^a ARGONNE NATIONAL LABORATORY   (United States)

Author keywords

[No Author keywords available]

Indexed keywords

EID: 0000671651     PISSN: 00295493     EISSN: None     Source Type: Journal    
DOI: 10.1016/S0029-5493(97)00217-3     Document Type: Article

References (17)

1. Beghi, G., 1968. Gamma Phase Uranium-Molybdenum Fuel Alloys, Euratom Report, EUR 5053e.
2. 2 by ion or neutron irradiation. J. Nucl. Mater. 230, 158-163.
3. 2O-reflected research reac-tors. Proc. 19th Int. Meet. on Reduced Enrichment for Research and Test Reactors. Seoul, Korea.
4. Cabane, G., Donzé, G., 1959. Stabilisation de la phase γ dans les alliages ternaires à base d'uranium-molybdene. J. Nucl. Mater. 4, 363-373.
5. Copeland, G.L., 1995. Fuel qualification plan for the ANS reactor. Oak Ridge National Laboratory Report, ORNL/ M-4324.
6. Durand, J.P., Fanjas, Y., 1994. LEU fuel development at CERCA: status as of October 1993. Proc. 16th Int. Meet. on Reduced Enrichment for Research and Test Reactors. Japan Atomic Energy Research Institute Report, Oarai, Japan, JAERI-M 94-02, pp. 71-78.
7. Durand, J.P., Laudamy, P., Richter, K., 1994. Preliminary developments on MTR plates with uranium nitride. Proc. Int. Meet. on Reduced Enrichment for Research and Test Reactors, 18-23 September. Argonne National Laboratory Report, ANL/RERTR/TM-20, Williamsburg, VA.
8. Graber, M.J., Zelezny, W.F., Moen, R.A., 1996. In: Francis, W.C. (Ed.), Annual Progress Report on Reactor Fuels and Materials Development for FY 1965. USAEC Report, IDO-17154, February, p. 11.
9. 6Fe-Al dispersion fuel elements. J. Nucl. Mater. 150, 238-243.
10. Hofman, G.L., Neimark, L.A., 1987. Prospects for stable high-density dispersion fuels. Proc. X Int. Meet. on Reduced Enrichment for Research and Test Reactors. Buenos Aires, Argentina.
11. Hofman, G.L., Rest, J., Snelgrove, J.L., 1996a. Irradiation behavior of uranium oxide-aluminum dispersion fuel. Proc. 19th Int. Meet. on Reduced Enrichment for Research and Test Reactors. Seoul, Korea.
12. 2 interdiffusion and its implication for the performance of highly loaded fuel operating at higher temperatures and fission rates. Proc. 19th Int. Meet. on Reduced Enrichment for Research and Test Reactors. Seoul, Korea.
13. Kim, K.H., Lee, D.B., Kim, C.K., Kuk, I.H., 1996. Development of high loading U-Mo alloy fuel by centrifugal atomization. Proc. 19th Int. Meet. on Reduced Enrichment for Research and Test Reactors. Seoul, Korea.
14. Kramer, D., Johnston, W.V., 1963. Post irradiation annealing of uranium-molybdenum ternary alloys. J. Nucl. Mater. 9 (2), 213-215.
15. Rest, J., 1995. The DART dispersion analysis tool: a mechanistic model for predicting fission-product-induced swelling of aluminum dispersion fuels. Argonne National Laboratory Report, ANL-95/36.
16. Van Thyne, R.J., McPherson, D.J., 1957. Transformation kinetics of uranium-niobium and ternary uranium-molybdenum-base alloys. Trans. ASM 49, 576-597.
17. Yakel, H.L., 1974. Review of X-ray diffraction studies in uranium alloys. In: Burke, J.J., Colling, D.A., Gorum, A.E., Greenspan, J. (Eds.), Physical Metallurgy of Uranium Alloys. Proc. 3rd Army Materials Technology Conference. Brook Hill Publishing, Chestnut, MA, pp. 259-308.