STUDY ON APPLICATION OF HAFNIUM HYDRIDE CONTROL RODS TO FAST REACTORS

Transcription

1 STUDY ON APPLICATION OF HAFNIUM HYDRIDE CONTROL RODS TO FAST REACTORS Konashi K. 1, Iwasaki T. 2, Terai T. 3, Yamawaki M. 4, Kurosaki K. 5, Itoh K. 6 1 Tohoku University, Oarai, Ibaraki, Japan 2 Tohoku University, Sendai, Miyagi, Japan 3 The University of Tokyo, Tokyo, Japan 4 Tokai University, Kanagawa, Japan 5 Osaka University, Osaka, Japan 6 Nuclear Development Corporation, Tokaimura, Ibaraki, Japan 1. Introduction Application of hafnium hydride (HfH x ) to control rods of fast reactors (FRs) is discussed in this paper focusing on superb characteristics of HfH x in comparison with the conventional control rod material boron carbide (B 4 C). Zirconium hydride is widely employed as a neutron moderator in nuclear reactors due to the fact that the fast neutrons in nuclear reactors are efficiently moderated in the metal-hydride. The hydride fuel of U-Zr hydride developed by General Atomics (GA) has been in use for more than 40 years in many TRIGA reactors around the world both in constant power and pulsed power operating conditions 1,2. Recently, a new type of hydride fuel has been studied for transmutation of nuclear wastes 3,4. Succeeding those development works we propose to apply the hafnium hydride control rod concept to FRs. 2. Application of hafnium hydride (HfH x ) to FR core absorber 2.1 Comparison of B 4 C and HfH x control rods The B 4 C has been mainly used for control and shut down material of fast reactors. But the life-time of B 4 C control rods is restricted by Pellet-Cladding Mechanical Interaction (PCMI) failure due to the He gas swelling of B 4 C pellet which is caused by the following nuclear reaction. B 10 (n,α) Li 7 (1) In order to prolong the control rod life time we propose to use HfH x as absorber material for FRs by the reasons that He gas is not generated in nuclear reaction of HfH x, and that HfH x can be expected to absorb neutron for more than 40 years owing to the fact that Hf 178 and Hf 179 which are generated by neutron captures of Hf 177 and Hf 178 respectively, also have large neutron capture cross sections which are expressed by the reaction (2). Hf 177 (n,γ) Hf 178 (n,γ) Hf 179 (n,γ) Hf 180 (2) The concept to use HfH x as control material is illustrated in Figure HfH 1.0 Long-life absorber Bonded He Control rod worth ( % ρ) B 10 80% enriched B C B 10 40% enriched B C natural B C Operation time ( years ) Figure 1 Schematic draw of HfH x application concept to FRs Figure 2 Decreases in worth of HfH 1.0 and B 4 C during reactor operation

2 As Hf has supreme characteristics of good absorption cross section for thermal neutrons, of excellent mechanical properties and of an extreme corrosion resistance, it has been used as control rods for LWRs. Although it is well known that the cross sections of neutron capture of Hf are small in FRs in comparison with LWRs, our recent work 5 showed that fast neutrons generated in the region of driver fuel assembly are moderated in the region of Hf-hydride assembly and that the Hf-control rod can efficiently absorb neutrons in the fast reactor core. Our calculation result of control rod worth variations in an FR is shown in Figure2. Although the worth of B 4 C decreases rapidly, the worth of HfH 1.0 keeps high value. 2.2 Material properties of HfH x The HfH x disks (Figure 3) were fabricated 5 and their physical properties were examined. The thermal conductivity of HfH x measured by Tsuchiya et al. 6 is shown in Figure 4. The thermal conductivity of ZrH x and TiH x measured by Yamanaka et al. 7,8 are shown in Figure 5. It can be seen that their conductivities are almost regulated by metal (Hf, Zr and Ti) conductivity. The HfH x disks are going to be irradiated in Japanese experimental fast reactor Joyo to examine their integrity under the irradiation. Figure 3 HfH 1.0 disk 5 Figure 4 Thermal conductivity of HfH x 6 Figure5 Thermal conductivity of ZrH x and TiH x 7,8 2.3 HfH x control rod behaviors Preliminary design of an HfH x control rod assembly Preliminary design of an HfH x control rod assembly was performed in consideration to suit a large scale liquid metal cooled fast reactor (LMFR), the feasibility study of which is being undertaken in Japan 9. The typical specifications of the core are follows. Thermal output: 3570 MWt Core inlet/outlet temperature: 550/395 deg C Fuel life time: 6 year Fuel fast neutron dose: 4.6 x n/cm 2 (E>0.1MeV)

3 Small size rods of HfH x were selected in our HfH x control rod assembly design to keep the rods temperature to be low under core irradiations in order not to cause dissociation of HfH x and not to cause hydride transfer through the claddings during some temperature rise accidents. The cross section of a control rod assembly which is composed of 55 control rods is shown in Figure 6. The rod cladding outer/inner diameter is 17.9/15.5 mm respectively. In order to restrict excessive temperature rise and to get high Hf inventory, the linear heat rate is depressed to be 80 W/cm and structural specifications of the absorber pellet outer diameter: 15.2 mm and rod pitch (triangular array) : 20.7 mm are derived from the parametrical survey to get high absorber volume ratio Control rod Guide tube Shroud tube Figure 6 Cross section of HfH x control rod assembly HfH x control rod Based on the above design of an HfH x control rod, the temperature distribution in it was analyzed. The radial temperature profiles of the HfH x pellet and the TRIGA type hydride fuel are shown in Figure 7. The maximum temperature of the HfH x pellet is lower than that of the TRIGA type pellet. The thermodynamic data of figure 8 show that the equilibrium H 2 -pressure of HfH 1.0 is also lower than that of ZrH 1.6. These facts assure the stability of the HfH x pellet under the operating conditions because the TRIGA fuel stabilities have been fully confirmed through many operating experiences. 700 Temperature (degree C) HfH 1.0 (Linear power=80w/cm) U-ZrH 1.6 (Linear power=450w/cm) 300 Pellet Gap Cladding Radial position (r/r0) Figure 8 Equilibrium hydrogen pressure of HfH 1.0 and TRIGA fuel Figure 7 Temperature radial profiles of Hfhydride control rod and TRIGA type hydride fuel

4 2.3.3 HfH x pellet Under irradiation conditions with radial temperature distributions, two type stresses are generated in HfH x pellets. One is thermal stress and the other is hydrogen stress. The hydrogen stress is caused by hydrogen redistribution, namely the density variation of the hydride with H/Hf ; x. For instance the tensile thermal stress caused by the radial temperature difference of 27.7 deg C in an HfH x pellet (Figure 7) amounts to 5.0 kg/mm 2 at the surface. Due to the lack of HfH x physical property data such as the diffusion coefficient, the heat of transport and the fracture stress, the hydrogen stress is hard to predict and it is also hard to definitely assure the integrity of HfH x pellet. But we may be allowed to presume its soundness under irradiation because the thermal stress is much lower than that of the TRIGA fuel and because the thermal stress is canceled by the hydrogen stress which appears in the contrary manner against thermal stress. Namely the thermal stress is tensile and the hydrogen stress is compressive at the pellet surface because the surface lattice parameter is larger than the center. The dependency on hydrogen content of the lattice parameter is shown in Figure The hydrogen is transported from center to surface side along the temperature gradient. (a) (b) Figure 9 Lattice parameter of (a) ZrH x and (b) TiH x HfH x rod cladding The ferritic-martensitic stainless steel is the most preferable alloy for an HfH x control rod in order to avoid swelling problem of claddings whose life-times are expected as long as fuel life-time of six years. The creep rupture strength of candidate alloys for FRs which have been developed by the Japan Atomic Energy Agency (JAEA) is shown in Figure Although the modified austenitic steel base alloy which is named PNC316 has higher strength than the ferritic-martensitic stainless steel named PNC-FMS, we can select the PNC-FMS as a candidate alloy for the HfH x control rod cladding because we can neglect inner pressure increase and because it has supreme swelling resistance and well compatibility with sodium. Those efficient characteristics of PNC-FMS have been verified by JAEA.

5 Figure 10 Comparison of creep rupture curves of PNC-FMS cladding with other FBR ferritic /martensitic and austenitic steels Discussion The irradiation experience of HfH 1.55 for FR control use was only reported by Risovany et al. 12. The hafnium hydride specimens were irradiated in Russian fast reactor BOR-60 for 15,840 hours under conditions of K argon and of fast neutron dose 2.5 x m -2. After the irradiation, all of the specimens retained their shape and integrity, and no hydrogen was released from the hafnium hydride specimens. This test result assures our hydride control rod design which is described above. The Russian irradiation test showed occurrence of radial cracks from centers to surfaces of the specimen and showed the linear dimensions increase by 2.2 %. We might imagine that the dimensional increase is due to generation of some cracks and not due to swelling, but the mechanism should be carefully examined hereafter. 4. Conclusion Study on application of hafnium hydride (HfH x ) to control rods of fast reactors (FRs) has been performed focusing on superb characteristics of HfH x in comparison with the conventional control rod material boron carbide (B 4 C). A preliminary configuration of a control rod assembly was designed to suit the Japanese large scale LMFR whose feasibility is now being studied by JAEA, and it is assured that a long life (more than 6 years) HfH x control rods can be designed from the view point of their in-core behaviors. But this study remains at first stage, therefore in order to realize HfH x control rods for FRs, following studies should be carried out in succession. Accumulate HfH x physical property data and select appropriate hydrogen content (H/Hf : x) for control rod material in FRs. Perform irradiation tests on HfH x control rods to ascertain their behaviors. Check the compatibility of HfH x pellet with sodium under in-core environment. Check the possibility of hydrogen transport through cladding under in-core environment including temperature rise accidents. Perform FR core design with HfH x control rods, and perform structural design of control rod. Perform safety analysis of FRs with HfH x control rods. 5. References [1] Simnad M.T., "The U-ZrH x alloy: Its properties and use in TRIGA fuel", Journal of Nucl. Eng. Design, Vol.64, 1981, pp

6 [2] Olander D.R. and Marowen Ng, Hydride fuel behavior in LWRs, Journal of Nucl. Mater., Vol.346 (2-3), 2005, pp [3] Konashi K. et al., Development of actinide-hydride target for transmutation of nuclear waste, Proc. International Conference on Back-End of the Fuel Cycle from Research to Solutions, GLOBAL2001, Paris France, September 9/13, [4] Konashi K. and Yamawaki M., The development of thorium hydride fuel, Characterization and quality control of nuclear fuels edited by Ganguly C. and Jayaraj R.N., Allied Publishers Pvt. Ltd., 2004, pp [5] Konashi K. et al., Development of advanced control rod of hydride hafnium for fast reactor, Proc. ICAPP 06, Reno, USA, [6] Tsuchiya et al., to be presented at international symposium on metal-hydrogen systems, Oct. 2006, Hawaii. [7] Yamanaka S. et al., Thermal properties of zirconium hydride, Journal of Nucl. Mater., Vol. 294, 2001, pp [8] Ito M. et al., Electrical and thermal properties of titanium hydrides, Journal of Alloys and Compounds, Vol. 420, 2006, pp [9] Kotake S., Current status of the feasibility study on commercialized fast reactor cycle systems and rector core performance of the promising fast reactors, Proc. INES-1, Tokyo, Japan, [10] Ito M. et al., Effect of electronegativity on the mechanical properties of metal hydrides with a fluorite structure, Journal of Alloys and Compounds, in press. [11] Shikakura et al., Development of high-strength ferritic/martensitic steel for FBR core materials, Journal of Japanese Energy Society of Japan, Vol.33[12], 1991, pp (in Japanese). [12] Risovancy V.D. et al., Hafnium in nuclear engineering, ISBN: