In-reactor investigation. of the composite UO 2 -BeO fuel: background, results and perspectives

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1 In-reactor inestigation Absrtract NUCM2016_0074 Reference P3.05 Introduction of the composite UO 2 -BeO fuel: background, results and perspecties M. A. McGrath 1 B. Yu. Volko 1 Y. Russin 2 1- Institute for Energy Technology (IFE), OECD Halden Reactor Project (Norway) 2 -JSC UMP (Kazakhstan); International concern in the area of safe and reliable nuclear fuel utilisation has drien inestigations oriented towards the deelopment of fuels and claddings with enhanced tolerance to accident conditions. One of the options to enhance safety is an increase in fuel thermal conductiity, which leads to a reduction of fuel temperature and hence fission gas release for the same power leel. The stored energy, which is also an issue for the first phase of design basis accidents, is also reduced by enhanced fuel thermal conductiity. The composite uranium-beryllium oxide fuel (UO 2 -BeO) is one of the promising fuel types in this context. The out-of-pile data obtained by some researchers for UO 2 fuel within a BeO matrix hae shown that small amounts of BeO substantially enhance the fuel thermal performance and may improe the irradiation properties of the fuel pellets produced from this composite [1,2]. A similar fuel type has been proposed which incorporates BeO particles in a UO 2 matrix, which facilitates fuel production technology and also enhances fuel conductiity. This fuel type has been produced by UMP (in Kazakhstan) with 3wt % BeO for a special in-reactor experiment that was performed in the Halden reactor (HBWR) in [3]. In-pile measurements, including fuel temperature, fuel elongation and gas pressure in a fuel rod with UO 2 -BeO fuel hae been performed and analyzed in comparison with standard UO 2 fuel up to a burnup of 30 MWd/kg oxide. The thermal conductiity model which has been deeloped in Halden for irradiated UO 2 fuel [4] is proposed to be modified for UO 2 -BeO fuels tested in HBWR. In order to inestigate the UO 2 -BeO fuel irradiation performance at a higher leel of fuel burnup, and to consider perspecties for this fuel to be adopted for NPPs the future, additional experiments are planned in the HBWR.

2 1. Properties of BeO BeO has some adantageous properties compared to UO 2 such as enhanced thermal conductiity and a lower thermal neutron absorption rate. In addition, in nuclear reactions with high-energy neutrons (~1.9 MeV), beryllium is a neutron multiplier, releasing more neutrons than it absorbs. BeO has good compatibility with uranium dioxide during sintering giing the possibility to produce a stable cold fuel with potentially low fission gas release or increased heat generation rate in comparison with standard UO 2 fuel. Some thermal and physical properties of BeO hae been collected from different sources [5,6,7] and are compared with those of other oxides and heat-resistance materials in Table 1 and Figure 1. The data illustrate some potential adantages of using BeO as an additie or matrix compound for standard oxide nuclear fuel. Howeer, there are some disadantages in comparison to UO 2, for example the toxicity of Be isotopes requires careful handling of BeO and the low solubility of BeO in UO 2 means only composite fuel types can be produced. Moreoer the amount of BeO additie needs to be compensated by an increase of fissile material density. Table 1. Properties of BeO compared to other fuel addities or compound materials Oxide Density [g/cm3] Thermal Cond. [W/(mK)] Melting temperature [K] Thermal neutron absorption [barn] BeO UO MgO CaO Al2O ZrO ThO SiC Figure 1. Comparison of BeO conductiity with other materials American Berilla Inc: Mo W

3 2. Out of pile inestigation of UO 2 -BeO composite fuel BeO is chosen as an additie due to its high thermal conductiity and compatibility with UO 2 fuel. Most of the studies of UO 2 -BeO composite hae been performed with different types of non-irradiated fuels [1,2,8,9]. Two types of composite were inestigated, first with BeO matrix (so called continuous phase surrounded UO 2 granules) and second with BeO addities in the form of particles (so called discontinuous phase) with the aim of determining the effectie fuel thermal conductiity and the effect of these different fuel microstructures - shown in Figures 2 and 3 - on thermal performance of the composite fuel. Wt/m К 46,0 44,0 16,0 14, «pure» UO 2 2 0,5% BeO 3 3% BeO 4 5% BeO 5 30% BeO Figure 2. Composite UO 2 fuel in BeO matrix (continuous phase) Calculated and measured out-of-pile performance of composite UO 2 -BeO fuel: Better conductiity of fuel with BeO matrix rather than with BeO particle addities [8,9] possible neutron multiplication effect of Be in the fuel on reactor core performance [10]. reduction of stored energy, gas pressure and cladding strain under DB LOCA scenario [11] 12,0 10,0 8,0 6,0 4, Figure 4. Thermal conductiity of UO 2 and UO 2 - BeO fuel pellets as a function of temperature for different wt% of BeO in the fuel* К Figure 3. Composite UO 2 fuel with BeO particles (addities of discontinuous phase) 10 mm *UO 2 -BeO fuel pellets with BeO additie in the form of particles was deeloped, produced and inestigated in the laboratories of UMP (Kazakhstan) [5] where the amount of BeO in UO 2 -BeO fuel was also optimized (3-5 wt%). Some of the data obtained are presented in Figure 4 and gien in Table 2 [3]. Table 2. Thermal conductiity of UO 2 -BeO fuels* T, К Thermal conductiity,w/m К 1.0 wt % BeO 1.3 wt % BeO 1.5 wt% BeO 5.0 wt% BeO

4 Volume fraction of pores, % 3. Irradiation performance of composite UO2-BeO fuel The multipurpose irradiation test with UO 2 -BeO (3 wt%) composite fuel supplied by JSC «UMP» (Kazakhstan) was performed in the Halden reactor (HBWR) together with fie other types of fuel including large grain UO 2 fuel also produced by «UMP». This UO 2 fuel acted as a reference for comparing the thermal and FGR behaiour of the UO 2 -BeO fuel under irradiation. Some initial data from these two fuel types are gien in Table 3 [3,12]. The test was performed under HBWR saturated coolant conditions (235 C and 28 bar) which are adequate for a fuel performance study. The rods were identically instrumented with fuel center thermocouple (TFs) and fuel elongation detector (EF) at the upper end, and pressure transducer (PF) fixed to the lower end of the rod. The instrumentation allowed the basic irradiation properties such as fuel densification, swelling, thermal performance and gas pressure deelopment under irradiation to be estimated. 3.1 Densification and swelling compared to UO 2 fuel The fuel densification at BOL and swelling with burnup, which are important parameters for nuclear fuel, were determined from the fuel elongation measurements normalised to the same conditions. Figure 5. Fuel olume change of UO 2 -BeO s UO 2 fuel The data show that densification of UO 2 -BeO fuel (~0.9 ol%) is slightly larger than for UO 2 fuel (5 ol%) but that the swelling, at about 1.0% / 10 MWd/kg oxide is similar to that for UO 2 fuel. The fuel porosity and pore size distributions shown in Figure 6 may hae some effect on the fuel densification. 50,0 40,0 30,0 20,0 10,0 0,0 Table 3. Test fuel rod data Dopant Oxide density [g/cm 3 ] UO 2 -BeO fuel 3 wt% BeO (~9.7 ol%) UO 2 fuel None Enrichment [%] Grain size [µm] 9 45 Fuel pellets design IDF/ODF/IDC/ODC [mm] 1.8/9.13/9.30/10.75 Fill pressure at RT [bar] 10 Instrumentation 1,0 2,0 3,0 4,0 5,0 6,0 7,0 8,0 9,0 10,0 11,0 12,0 13,0 14,0 Pores size, micron Fuel thermocouple (TF) Pressure transducer (PF) Fuel stack elongation (EF) Figure 6. UO 2 -BeO fuel pore size distribution and ceramography compared to reference UO 2 fuel

5 3.2 Thermal performance and gas pressure deelopment in test rods with UO 2 -BeO fuel and UO 2 fuel under irradiation The thermal performance of the UO 2 -BeO fuel under irradiation is one of the most important parts of the test performed in the HBWR. The fuel was irradiated to a burnup of about 30 MWd/kg oxide with the two rods (UO 2 -BeO rod 3 and UO 2 rod 4) irradiated at an LHR in the range kw/m (the UO 2 -BeO fuel rod LHR was about 1-2 kw/m lower than for the UO 2 fuel rod). A difference between the fuel centre temperature was measured in these two rods; in the range o C depending on the heat generation rate. The power history and measured fuel centre temperatures are shown in Figure 7. In order to study the fuel thermal performance under irradiation, measured fuel temperatures were normalised to the same power and plotted against burnup as shown in Figure 8. The rise of the fuel temperature at BOL was partly related to fuel densification whereas the subsequent temperature reduction mostly reflects fuel-clad gap closure due to fuel swelling (see Figure 5). This thermal behaiour was ery similar for both fuel types but with lower fuel temperature in the UO 2 -BeO fuel. Power up-rates were performed seeral times during the test in order to promote FGR at different burnups, and the measured gas pressures were normalised to zero power and RT in order for the onset of FGR as a function of burnup to be obsered see Figure 9. Alongside the rod pressures, peak (solid) fuel pellet temperatures, deried from the measured (central hole) fuel pellet temperatures, are plotted enabling the FGR thermal threshold to be determined. The data indicated a FGR threshold in the UO 2 fuel of about 1250 C whereas no FGR was indicated in the UO 2 -BeO fuel which maintained a stable gas pressure oer a similar heat rating to the UO 2 fuel due to its lower center temperature. Figure 7. Power history and fuel temperature measured in UO 2 -BeO and reference UO 2 rods as a function of aerage burnup Figure 8. Power history and fuel temperature normalised to 25 kw/m in UO 2 -BeO and reference UO 2 rods s aerage burnup Figure 9. Peak fuel temperature and measured rod gas pressure normalised to 25 C in UO 2 - BeO and reference UO 2 rods s burnup

6 Relatie conductiity Fuel conductiity, W/mK Fuel centeline temperature, o C 3.3 Thermal conductiity correlation for UO 2 -BeO fuel irradiated in the HBWR Thermal conductiity of the UO 2 -BeO fuel tested in the HBWR was ealuated by means of the Halden correlation for UO 2 fuel [4] modified with the data obtained by S. Ishimoto [8] for the two types of fuel: BeO addities (discontinuous/dispersed) and BeO matrix (continuous) as shown in Figure 10. The correlation (eq. 1) includes a technological factor gien in Table 4 for the two fuel types. Halden conductiity correlation modified for UO 2 -BeO fuel: λ UO2-BeO = 1 + KBeO FV BeO Bu Bu T e T, W/mK (1) where: Bu is the fuel burnup in MWd/kg oxide, fuel temperature T in o C, K BeO and FV BeO are the technological factor and olume % fraction of BeO in UO 2 fuel, respectiely Table 4. Fuel type UO 2 with BeO addities BeO- addities BeO-matrix Vol % BeO UO 2 in BeO matrix KBeO Figure 10. Relatie conductiity change of the two types of UO 2 -BeO fuel s ol.% BeO in UO 2 fuel [1] The comparison of the conductiity correlations for UO 2 and different types of UO 2 -BeO fuel with 9.7ol % of BeO including modified correlation from FRAPTRAN [11] are shown in Figure 11 as a function of fuel temperature. The Halden conductiity correlation was implemented in the FTEMP3 code in order to compare the fuel performance calculations and the fuel centre temperatures measured in the UO 2 -BeO fuel during the test. The data has a good agreement as shown in Figure 12. UO2 at BOL: UO2-BeO(9.7ol%)-3wt% corr (1) - addities UO2-BeO (9.7ol%)-3wt%, corr FRAPTRAN UO2-BeO(9.7ol%)-3wt% corr (1) matrix Fuel local temperature, o C Figure 11. Comparison of the conductiity correlations used for UO 2 -BeO s UO 2 fuel behaiour analysis FTEMP3 Calculation for UO 2 -BeO fuel FTEMP3 Calculation for UO 2 fuel Measured temperatures in UO 2 -BeO fuel Measured temperature in UO 2 undoped fuel Fuel temperature is calculated using conductiity correlation (1) for fuel UO 2 and UO2-BeO 3%wt (~9.7 ol%) LHR, W/cm Figure 12. Comparison of the fuel temperatures measured in UO 2 and UO 2 - BeO fuels and calculated by FTEMP3 code

7 4. Discussion and future work The out-of-pile inestigation of UO 2 -BeO fuel showed: The high conductiity BeO compound enhances the conductiity of the composite UO 2 -BeO fuel by small amounts of content. Fuel can be produced as a composite fuel due to the low solubility of BeO in UO 2. The effectie conductiity of UO 2 -BeO fuel is dependent not only on the amount of introduced BeO in UO 2 but also on fuel microstructure. The published data suggested that UO 2 integrated into a BeO matrix has a higher conductiity than fuel produced with BeO addities in a UO 2 matrix. Howeer, it was obsered that this difference diminishes with temperature. Neutronic calculations show that the presence of Be isotopes in the fuel may produce a neutron multiplication effect on the reactor core performance [10]. Analysis of accidents shows some positie effect of the enhanced conductiity of UO 2 -BeO fuel due to a reduction of stored energy, gas pressure and cladding strain under DB LOCA scenario [11 ]. The in-pile UO 2 -BeO fuel inestigation in the Halden reactor: The main goal was to obtain the first in-pile data on UO 2 -BeO fuel under irradiation and study the effect of the irradiation on the fuel behaiour, in particular fuel mechanical stability and fuel composite conductiity. The fuel was produced and supplied by JSC UMP (Kazakhstan) with 3.0 wt% BeO (~9.7 ol%) in iew of additie particles. The fuel was irradiated for about 700 days to a burnup of about 30 MWd/kg oxide with the rod instrumented with fuel thermocouple, fuel elongation detector and gas pressure transducer. In-pile data indicated that the dimensional stability of the UO 2 -BeO fuel was similar to UO 2 fuel irradiated in the same assembly. The fuel densification was ealuated to about 0.9 ol. % which is slightly larger than for standard UO 2 fuel (0.5 ol %) whereas the fuel swelling was ealuated at around 1% per 10 MWd/kg oxide for both fuels. Rod gas pressure measurements demonstrated the absence of FGR in UO 2 -BeO fuel at similar LHR to UO 2 fuel due to lower temperature. Fuel temperature measurements confirmed the higher UO 2 -BeO fuel conductiity in-pile in accordance with out-of-pile data. The thermal conductiity correlation deeloped in Halden for UO 2 fuel was tentatiely modified for the UO 2 -BeO composite fuel types. Future works The irradiation of the fuel is planned to continue with the aim of studying fuel thermal conductiity degradation under irradiation. The next important task for the UO 2 -BeO fuel re-irradiation will be to determine the FGR thermal threshold. PIE will be performed after the fuel irradiation to check for degradation of the microstructure of the composite fuel. It is suspected that the microstructure of the BeO matrix will be degraded under irradiation at high temperature. It would be interesting to compare the in-pile behaiour of composite UO 2 fuel with BeO matrix against UO 2 fuel with BeO addities for example in a test in the Halden reactor

8 References 1) R. Latta, S.T. Reankar, A.A. Solomon, Modeling and Measurement of Thermal Properties of Ceramic Composite Fuel for Light Water Reactors, Heat Transfer Engineering, 29(4), , ) Kein McCoy, Claude Mays, Enhanced thermal conductiity oxide nuclear fuels by co-sintering with BeO: II. Fuel performance and neutronics J. Nuclear Materials, 375 (2008), ) Y. Russin, Y. Shahorosto, M. McGrath, A. Gagarin, B. Volko, A. Kuchkoski, Innoatie fuel deelopment in ULBA (Kazakhstan) and testing in the Halden Reactor, WRFPM, TopFuel 2011, Chengdu, China, Sept ) Wiesenack W., Terberg T., Thermal performance of high burnup fuel in-pile temperature data and analysis International Topical Meeting on LWR Fuel Performance. USA, Park City, Utah, April 10 3, 2000 p ) W.D. Kingery, H.K. Bowen, D. R. Uhlmann, Introduction to Ceramics, John Wiley & Sons, New York, ) D-Joo Kim, Y. W. Rhee, J-H. Kim et al., Fabrication of microcell UO2-Mo pellets with enhanced thermal Conductiity, J. of Nuclear Materials 462, (2015), ) G.P. Akishin, S.K. Turnae, V. Ya. Vaispapir, et.al., Thermal conductiity of Berillium Oxide Cearamic, J. Refractories and Industrial Ceramics, Vol. 50, No 6, ) S. Ishimoto, M. Hirai, K. Ito, Y. Korei, Thermal Conductiity of UO2-BeO Pellets, J. Nuclear Science and Technology, March Vol. 33, No.2, p ) D.S.Li, H.Garmestani, J.Schwartz, Modelling thermal conductiity in UO2 with BeO additions as a function of microstructure, J. Nuclear Materials 392 (2009) ) A.A. Koalishin, V.N. Proselko, V.D. Sidorenko, Y. Stogo, On the possibility of using Uranium-Be Oxide Fuel in VVER reactor, ISSN , Physics of Atomic Nuclei, 2014, Vol 77, No 14, pp ) D. Chandramouli, S. T. Reankar, Deelopment of Thermal Models and Analysis of UO2-BeO Fuel during a LOCA, Int. J. of Nuclear Energy V. 2014, Article ID ) B. Volko, T. Terberg, M. McGrath, Experimental inestigations of addities on irradiation Performance of Oxide Fuel, WRFPM, TopFuel 2014, Sendai, Japan, Sep