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Solid Micro-Beads Hetero-Structure Fuel for Ultra-high Temperature Applications

Published online by Cambridge University Press:  26 February 2011

Liviu Popa-Simil  and
Gabriel Vasilescu
Liviu Popa-Simil
Affiliation:
lpopas@ieee.org, LAVM Inc., R&D, 3213_C Walnut St., Los Alamos, NM, 87544-2092, United States
Gabriel Vasilescu
Affiliation:
vasilescu@rogers.ca, LAVM Inc., Los Alamos, NM, 87544, United States
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Abstract

Higher conversion efficiencies require high operation temperatures that are difficult to obtain due to the actual thermo-physical properties of the nuclear fuels. The initial intrinsic thermal conductivity of the actual fuel pellets, mainly ceramics like structures (oxides, nitrides, carbides, MOX, beads) is low. The center of the pellet is near melting temperature while the cladding operation temperature has to be low. The fission products deposition and burnup effects are further dimming the thermal conductivity. More the cooling agent's chemical reactivity increases with temperature is another main reasons of keeping the operation temperature low. The usage of a hetero-structure of solid fuel soaked into a drain fluid is increasing the thermal conductivity. Properly shaped beads structure drives to the possibility of preventing most of the fission products of being stored inside the fuel lattice deteriorating its properties, being drained outside the nuclear reactor. This changes inspired from the nature, makes the nuclear reactor resembling with a plant having self cleaning and curing properties while operating at higher temperatures. Immersing the fuel into liquid metal higher operation temperatures are allowed due to increase in thermal conductivity by 3 to 20 times. Making the fuel beads shorter than the range of fission products, their trajectories end in the drain liquid that is tolerant to nuclear recoils damage. Due to better thermal conductivity the temperature field differences inside the nuclear reactor becomes smaller, allowing the operating temperature to rise significantly without safety concerns. There is possible to continuously remove the fission products by smoothly circulating the drain liquid. The low flow is needed to give time to short lives fission products to decay inside the reactor highly shielded volume. There are several fissionable materials and drain liquids matching which assure high operation temperatures, and allow He cooling, and high temperatures gas turbines cycles. The conversion efficiency might be higher than 70% depending on the chosen actinide / drain-liquid / cooling-liquid combination. The new concept on fission products continuous release and separation minimizes the waste and the total radioactivity stored inside the reactor to few weeks integrated operation amount remaining constant over the time. That makes the fuel's remnant radioactivity lower by a factor >100 than the actual reactors level. The fuel reactivity might be controlled by poisoning and transmutation or by assuring specific reactive geometries which to allow a ultrahigh burnup without the need of over-criticality loads. The fuel's deformability opens the way for interesting applications. The advantages of micro structured nuclear fuel are higher thermal conductivity, fission products removal, appropriate reactivity, higher efficiencies and longer fuel life.

Keywords

actinidethermal conductivitymicrostructure
Type
Research Article
Information
MRS Online Proceedings Library (OPL) , Volume 981: Symposium JJ – Structural and Refractory Materials for Fusion and Fission Technologies , 2006 , 0981-JJ07-10
Copyright
Copyright © Materials Research Society 2007

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References

REFERENCES

1. Marquis, J., Novel Integrated Reactor/Power Conversion System. NERI, 1999. 99–0198(1): p. 2.Google Scholar
2. POPA-SIMIL, L., Report on Nuclear Fuel's Thermal Conductivity Modeling and Measurement - Urania. LANL Reports, 2004. LA-UR-04-5194(1): p. 1–45.Google Scholar
3. FINK, J.K., Thermal Conductivity and Thermal Diffusivity of Solid UO2. International Nuclear Safety Center (INSC), 2004. Preliminary Recommendation.(anl.gov): p. 28.Google Scholar
4. LANNING D. D., C.L.P., ,, FRAPCON-3: Modifications to Fuel Rod Material Properties and Performance Models for High-Burnup Application. Nuclear Regulatory Commission, NUREG/CR-6534 U.S., 1997. 1((PNNL-11513)): p. 1–124.Google Scholar
5. Yang, J.H., K.W.K., KUN W.S., et al.,, Fabrication and Thermal Conductivity of (Th,U)O2 Pellets. Nuclear Technology, 2004. 147(1): p. 113–119.Google Scholar
6. INOUE, M., Thermal conductivity of uranium-plutonium oxide fuel for fast reactors. Journal of Nuclear Materials, 2000. 282((2/3): p. 186–95.Google Scholar
7. Minato, K., Takano, A.M. , M., Arai, Nakajima, Y , K., Itoh, A., Ogawa, T.,,Fabrication of nitride fuels for transmutation of minor actinides. Journal of Nuclear Materials, 2003. 320: p. 1–2.Google Scholar
8. Spriggs, G. D., Criticality of uranyl-nitrate solution containing boric acid. Transactions of the American Nuclear Society, 1985. 50: p. 304–5.Google Scholar
9. VERALL, R. A., KRSTIC, V.M.D. , V. D.,, Silicon carbide as an inert-matrix for a thermal reactor fuel. Journal of Nuclear Materials, 1999. 274: p. 54.Google Scholar
10. DZIADOSZ, D.,Saglam, A.T.N. , M., Sapyata, J.J.,, Weapons-Grade Plutonium-Thorium PWR Assembly Design and Core Safety Analysis. Nuclear Technology, 2004.147(1):p. 69–83.Google Scholar
11. BURAKOV, B., A.E., Immobilization of Excess Weapons Plutonium in Russia: A Review of LLNL Contract Work. Proc. Meet. For Coordination and Review of Work,, 2001. UCRL-ID-143846: p. 229–234.Google Scholar
12. Degueldre, C., Lee, A.T. , Y. W.,, Thermal conductivity of zirconia based inert matrix fuel: use and abuse of the formal models for testing new experimental data. Journal of Nuclear Materials, 2003. 319: p. 6–14.Google Scholar
13. MOGENSEN, M.,Walker, P.J.H. , C. T.,, Behaviour of fission gas in the rim region of high burn-up UO{sub 2} fuel pellets with particular reference to results from an XRF investigation. Journal of Nuclear Materials, 1999. 264: p. 1–2.Google Scholar
14. White, R. J., The development of grain-face porosity in irradiated oxide fuel. Journal of Nuclear Materials, 2004. 325(1): p. 61–77.Google Scholar
15. Maucec, M., P.A., Glumac, B., Jeraj, R.,, Criticality safety study of NPP Krsko spent fuel pool with Monte Carlo computer code MCNP4A. 1994.Google Scholar
16. BOWMAN, C. D., V.F., Accelerator-driven thermal spectrum system for weapons plutonium destruction. Transactions of the American Nuclear Society, 1993. 69: p. 102–3.Google Scholar
17. LOAIZA, D.J., S.W., Criticality Data for Spherical 235U, 239Pu, and 237Np Systems Reflector-Moderated by Low Capturing-Moderator Materials. Nuclear Technology, 2004. 146(2): p. 143–154.Google Scholar
18. RIMPAULT, G., The ERANOS Code and Data System for Reactor Neutronics Analysis. Proc. Int. Conf. on the Physics of Fuel Cycles and Advanced Nuclear Systems, PHYSOR 2002, Seoul, Korea, 2002.Google Scholar
19. ZIGLER, J.F., SRIM. IBM-web, 2003.Google Scholar
20. VALENIUS, J., Transmutation Course. University Web, 2004. Thermal conductivity.Google Scholar