On the Effect of MOX Fuel Conductivity in Predicting Melting in FR Fresh Fuel by Means of TRANSURANUS Code

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1 On the Effect of MOX Fuel Conductivity in redicting Melting in FR Fresh Fuel by Means of TRANSURANUS Code Ahmed ALY, Christophe DEMAZIERE Chalmers University of Technology, Department of Applied hysics, Division of Nuclear Engineering, SE , Gothenburg, Sweden ABSTRACT Davide ROZZIA University of isa Largo Lucio Lazzarino , isa (I), Italy daviderozzia@libero.it Alessandro DEL NEVO ENEA CR-Brasimone Località Brasimone Camugnano (BO), Italy alessandro.delnevo@enea.it The capability of the fuel to operate at high power without melting is important to Fast Reactors. Reactor design limits normally require that there be a low probability of fuel melting during steady-state operation, including overpower conditions. This requirement has a direct effect on the steady-state power limit of the fuel pin and hence on the reactor power. The development of computational tools that are able to capture the occurrence of high temperature phenomena and mechanisms is thus an important step in reducing the margin of conservatisms and increasing the reactor efficiency. Among the experiments that were conducted for this purpose, HEDL - experiment has been selected and simulated using TRANSURANUS code to exploit its capability to capture the inception of MOX fuel melting. The experiment included 8 fresh pins with cladding Outside Diameters (OD) 5.84 mm, and 8 fresh 6.35 mm OD pins. It was performed during 71 to investigate the effect of the initial fuel-to-cladding diametric gap size (from to 0.25 mm) on the linear heat rate needed to initiate incipient melting at beginning-oflife. All 5.84 mm OD pins with fuel-to-cladding gaps equal to or less than 0.14 mm had no fuel melting. The remaining 5.84 mm OD pins and all the 6.35 mm OD pins experienced partial fuel melting. This work reported hereafter consists of two parts. The main objective of the first one is to assess the capability of TRANSURANUS to predict the measured melting heights of the tested rods and its implications on the thermal conductivity correlations implemented in the code. The second part included modifications that targeted the high temperature thermal conductivity term in two TRANSURANUS correlations. The modifications were incorporated into the code that was recompiled

2 INTRODUCTION The usage of MOX fuel is a strategic option for a more sustainable use of nuclear reactors for power production. MOX fuel is currently well-developed as a fuel for LWRs. On the long run, the development of GEN-IV FRs is a necessity if a better use of the fuel resources is targeted.[1] MOX fuel should be designed and modelled properly to be used in those kinds of reactors. There are some differences between the behaviour of MOX fuel for LWRs and FRs. Those differences are due to the differences in the design features between them. In particular, the stronger neutron flux of FRs and the higher design operation temperature are among those features that affect the in-reactor fuel behaviour. Various phenomena and parameters such as FGR, melting temperature, fuel fragment relocation, fuel swelling, and thermal conductivity of MOX fuel are affected both by the temperature field and the higher neutron flux.[2] While the behaviour of MOX for LWRs is well modelled and documented, its performance in FRs needs to be addressed thoroughly. The main focus of the work reported hereafter is on the thermal conductivity of FR MOX fuel. The thermal conductivity of FR MOX is generally lower than the conductivity of LWR MOX.[3] This draws some uncertainties about the modelling of thermal conductivity for FRs. A model that runs in an adequate manner for LWR MOX may not perform properly for FR MOX. Therefore, a separate assessment of thermal conductivity models for both types of reactors is required. In this paper, the thermal conductivity of FR MOX is considered by means of TRANSURANUS[4] code. The code was used to simulate the HEDL -[5] experiment. Several correlations implemented in the code were assessed for their ability to predict the thermal conductivity of fresh MOX fuel rods irradiated in HEDL - experiment. The experiment was conducted in EBR-II during 71 to investigate the melting propensity of 25% uo2 enriched fresh fuel whose design was typical of the Fast Flux Test Facility (FFTF). Several open literature thermal conductivity correlations were studied as well.[6] The details of this study are not discussed in this paper. The correlation by Baron-Hervè 95[7] was chosen to provide some basis for modelling in TRANSURANUS. Accordingly, two of the correlations implemented in TRANSURANUS were modified. The modifications targeted the high temperature thermal conductivity terms. The new versions of the correlations were then tested using TRANSURANUS for their ability to accurately simulate the fuel melting of the HEDL - rods. 2 DESCRITION OF THE EXERIMENT The HEDL -[5] test was designed to provide integral power-to-melt data on 25% uo2 enriched fuel. The fuel pin fabrication parameters simulate the FFTF/FBR design conditions. The experiment was performed in EBR-II to investigate the effects of initial fuelto-cladding gap sizes, from to mm, on the linear-heat-rate needed to cause incipient fuel melting at beginning-of-life. - was an encapsulated -pin subassembly, it consisted of 8 fresh pins with cladding outside diameters 5.84 mm, and 8 fresh 6.35 mm OD pins. The fresh pins were clad with 316 stainless steel (20% cold worked). The experiment consisted of: Slow power increase up to selected conditioning level held for 1 h. Rapid power increase up to a designed power level held for 10 minutes and scram. The EBR-II power history during HEDL - is given in Figure 1. Figure 2 reports the axial power profile of the pins. A summary of the design parameters and operating conditions is presented in Table 1. At the end of the experiment, metallographic examinations were

3 804.3 performed to determine the axial extent of fuel melting, the central void, the gap width and the radial extent of the columnar grain region. Figure 1: EBR-II power history.[5] Figure 2: HEDL -, power profile.[5] Table 1: HEDL - design data. Rod N Rod Id. 2 3R 5 6 7R R 25R 26R 27R Gap μm %TD Clad OD mm X X X X X X X X X X X X X X X X 5.84 X X X X X X X X 6.35 X X X X X X X X Fuel 25% UO 2-75% UO 2 Cladding 316 stainless steel (20% cold worked) Filling gas 98% He at 1 bar O/M 1.96 Active length 343 mm Na inlet temp. 371 C Max LHR kw/m Melting yes yes no no yes yes yes yes yes yes yes yes yes yes no Yes 3 DEVELOMENT OF THE INUT DECK The activity is performed using TRANSURANUS code, version v1m1j12, with the deterministic option, steady state thermal and mechanical analysis.[4] The models selected are generally the standard ones except fuel conductivity that has been subjected to comparative analyses. The choice of the relocation model mainly influences the gap width during the irradiation. The fuel relocation model is taken as the modified FRACON-3 (REL8). The fuel melting temperature was measured and reported in the database: 2760 C. Accordingly, the standard melting criterion (that depends on uo2 content, O/M ratio and burn-up), is deactivated and the exact value is introduced in the source of the code. Only the active part of the fuel is accounted for in the simulation. It is divided into 17 axial sections that were obtained from Figure 2. The nominal design values are used if available. The fuel average grain size and the gas plenum length are not reported in the database. They were instead assumed to be, respectively, 22 μm (average grain diameter) and 300 mm (close to the active length).

4 804.4 The boundary conditions implemented for the analysis are: linear heat rate at 17 axial position, sodium bulk temperature histories (axially variable), heat transfer coefficient at cladding outer surface (two values are given in the database), and coolant pressure (0.1 Ma). The rate of power increase during the final ramp is not given, it has been fixed at 500 kw/ (m*h), according to typical power ramp tests.[8] 4 ASSESSMENT OF HEDL - BY TRANSURANUS CODE In this section, four MOX thermal conductivity correlations implemented in TRANSURANUS are analyzed, as follows: CND-31 is the TU standard MOX correlation fitted to ITU data.[9] It accounts for the influence of the local temperature, the local burn-up, and the local porosity. CND-32 is the original MOX correlation of Carbajo et al.[10] It accounts for the local temperature, the local burn-up, the local porosity, the O/M, the dissolved and precipitated Fs and the radiation effects. CND-33 is the original MOX correlation of Lanning and Beyer.[11] It accounts for the local temperature, the local burn-up, the local porosity and the O/M. CND-34 is the original correlation of Wiesenack.[12] It is derived from the UO2 correlation reduced by a constant factor of These correlations are used in the code to model MOX fuel regardless of the modelled reactor type. They were mainly validated for LWR MOX. Their usage for FBR MOX is not fully validated. LWR MOX fuel contains lower amounts of plutonium and is operated at lower power and temperature than FBR MOX. There is a limiting difference between the thermal conductivities of low (<15%wt u) and high plutonium MOX fuel.[3] There is also uncertainty about the ability of the correlations to model FBR MOX at the higher temperature level reached in HEDL - experiment (close and beyond melting temperature). Figure 3: Rod---2, gap width as predicted by REL-8 model compared to experimental measurement. Figure 4: Rod---27R, gap width as predicted by REL-8 model compared to experimental measurement. The prediction of the fuel melting at beginning of life mainly depends on the correct simulation of the gap width and on the simulation of the fuel temperature radial profile. The TU standard REL08 is used to capture the gap width. REL8 is the modified FRACON-3 model. It accounts only for the tangential strain due to relocation depending on the as fabricated gap, the burn-up, the linear heat rate. It does not apply when gap is closed.[4] The

5 804.5 model is able to predict the gap width in a good manner in most of the irradiated rods. This is illustrated in Figure 3 and Figure 4 for two representative rods. Therefore, the predicted melting heights by TU will reflect on the ability of the assessed thermal conductivity correlations to predict the conductivity of the fuel. Figure 5 depicts the measured melting heights for the sixteen irradiated rods in the HEDL - experiment. The manufacturing parameters and power levels of those rods are summarized in Table 1. Along with the measured melting heights, the upper and lower limits of melting extents predicted by the studied correlations are represented in the figure. It can be observed that correlations CND-31, CND-32 and CND-33 over-predicts the melting heights inside all of the studied rods. CND-33 predicted the closest value to the measured melting heights. CND-32 predicted severe melting of the rods. It can be concluded that the correlations under-predict the thermal conductivity of the studied rods. The results obtained using CND-34 correlation are in contrast to results from the other correlations. CND-34 resulted in a general under prediction of the melting heights inside the rods that suffered incipient melting. This is an evidence that the correlation over predicts the thermal conductivity. However, CND-34 was able to simulate the rods that did not experience melting in a good manner. Using this correlation, the code did not predict melting in rods - -6 and For rod --5, the correlation predicted a relatively low melting height. Figure 5: Axial elevation of melting, experimentally and as predicted by TU. 5 AMENDMENTS TO CND-31 AND CND-33 Among the correlations considered, the Baron-Hervè 95 model was selected to form a good basis for possible model improvements in the high temperature range. This correlation[7] accounts for local temperature, O/M ratio, and the local u content. The high temperature term is based on Delette and Charles work.[13] The correlation was compared with the studied TU correlations. This was done by capturing the radial distribution of certain fuel rod (--2) conditions predicted by TU (i.e O/M ratio, fuel density, u-content, and temperature) at one instant of time almost before the reactor SCRAM. The obtained radial conditions of the rod were fed to the BH-95 correlation to obtain thermal conductivities along the radius of the pellet at peak position. The obtained thermal conductivities were then plotted against the temperature range inside the fuel rod in Figure 6. At high temperatures, BH-95 correlation predicts higher thermal conductivity than CND-31, CND-32 and CND-33 (correlations that over-predicted melting) but lower than CND-34 (under-predicted melting).

6 804.6 As a result, a particular focus was put on the high temperature thermal conductivity term of this correlation. In Figure 7, the lattice thermal conductivity of BH-95 is lower than those of CND-31 and CND-33. Its ambipolar electronic thermal conductivity term increases more rapidly than the ambipolar terms of CND-31 and CND-33 leading to the overall higher thermal conductivity predicted by the BH-95 correlation. The high temperature ambipolar thermal conductivity term implemented in BH-95 correlation was used as a substitute to the ambipolar terms in CND-31 and CND-33. CND-31 was chosen because it is the standard correlation in TU. CND-33 was chosen because it predicted the melting height closest to the experimental measurements. CND-32 was neglected because it largely over-predicted the melting heights. CND-34 was disregarded because it is an approximate correlation based on UO2 not MOX fuel. The code was then recompiled to generate a modified version. The HEDL - rods were simulated again using the modified CND-31 and CND-33. The results of the new simulations were compared with the results obtained from the original correlations to assess the effect of the newly implemented term. Figure 6: Radial profile of thermal conductivity for rod --2 as a function the radial temperature profile inside it. Figure 7: Lattice and electronic conductivities as a function of temperature at 92%TD, 25%u content for BH-95, and O/M ratio of In Figure 8, the effect of the modifications of the targeted correlations on the incipient melting is plotted along with the original code prediction. It can be noted that the modified correlations resulted in lower melting heights than those predicted by the original correlations. In general, the melting heights by the modified correlations are comparable to the experimentally measured heights. For the rods --3R and --24R, although the predicted melting heights did not decrease, the code still predicted lower melting fractions. The modified code was able to predict the non-melting inside one of the three rods that did not experience partial melting (--6). The melting heights predicted for the other two rods were lower than those predicted by the original code. The predicted melting fraction inside rod --33 reached a maximum of 0.2%. The maximum temperature did not exceed the melting temperature of the fuel.

7 804.7 Figure 8: Axial elevation of melting as measured experimentally and as predicted by TU original and modified correlations. In general the modifications reduced the level of conservatism inside the code when the issue of incipient melting of the fuel is addressed, as presented in Table 2. The melting height inside each rod is normalized to the experimentally measured height. The documented values show significant enhancement in the code prediction of melting. The coloured cells represent melting heights within ±5% (in green) and ±5-10% (in brown). The blue cells represent the rods that did neither melt during the experiment nor the simulations. Generally, if the melting heights are not around the ±5.5% uncertainty, the prediction tends to be conservative but on a lower level than the original code. For each correlation, the code failed to be conservative for only one rod outside of the 5.5% margin. Rod ID Table 2: Summary of the predicted melting heights by the original/modified correlations normalized by the experimental measurements for each rod. *Green (±5%), Brown (±5-10%), Blue (no melting predicted) CND-31 CND-33 CND-31 CND-33 CND-31 MOD CND-33 MOD Rod ID CND-31 MOD CND-33 MOD R R R M M M M 26R M NM M NM 27R R M M M M CONCLUSIONS The capabilities of TRANSURANUS code in simulating the inception of MOX fuel melting have been assessed against HEDL - experiment. It includes 16 fresh FBR rods irradiated in EBR-II to investigate the effects of initial fuel-to-cladding gap sizes, from to mm, on the linear-heat-rate needed to cause incipient fuel melting at beginning-oflife. From the results obtained from the performed simulations, the following conclusions can be drawn: The fuel conductivity correlations directly impact the fuel temperature. Four options have been tested. Three correlations over-predict the axial extent of fuel melting

8 804.8 being therefore conservative. Lanning and Beyer correlation is closer to the experiment. The Wiesenack correlation under-predicts the axial extension of fuel melting. However, it is the only one that simulates correctly the un-melted rods. The high temperature thermal conductivity implemented in BH-95 correlation was implemented in CND-31 and CND-33 correlations. This modification resulted in a generally better prediction of the thermal conductivity of both correlations. The code is less conservative than the original one but not under predicting the melting heights. The code was able to simulate correctly one of the three un-melted rods. The lack of O/M in CND-31 did not seem to lead to modified performance from CND-33 where it is included. This is due to the high temperature nature of HEDL - experiment. The high temperature conductivity term is the governing factor of thermal conductivity at this level of temperature. Modeling of densification and restructuring and assessment of design tolerances such as O/M ratio, gap initial size and pellet density are not considered in this paper. However, they are expected to impact the results and deserve further investigations. REFERENCES [1] IAEA Status and Advances in MOX Fuel Technology, Technical report number 415, Vienna, Austria, [2] A.R. Massih, Models for MOX fuel behavior, a selective review, SKI Report 2006:10, Stockholm, Sweden, [3] C. Duriez, J.-. Alessandri, T. Gervais, and Y. hilipponneau, Thermal conductivity of hypostoichiometric low u content (U,u)O2-x mixed oxide, Journal of Nuclear Materials 277 ( ), [4] K. Lassmann, A. Schubert,. Van Uffelen, J. van de Laar, TRANSURANUS Handbook version v1m1j12, JRC, ITU [5] R.B. Baker, Integral heat rate to incipient melting in UO 2 -UO 2 FR fuel, Hanford Engineering Development Laboratory, HEDL-TME 77-23UC-79b, U.S., 78. [6] A. Aly, Modelling thermal dependent phenomena of MOX fuel with focus on thermal conductivity, Master s thesis, Chalmers University of Technology, Sweden [7] D. Baron, Fuel thermal conductivity: A review of the modeling available for UO2, (U- Gd)O2 and MOX fuel, Thermal performance of high burn-up LWR fuel, seminar proceedings, Cadarache, France, 3-6 March, 98. [8] S. Djurle, et al., The Super-Ramp roject, Final report of the Super-Ramp project, STIR-32, Studsvik AB Atomenergi, Studsvik, Sweden, 84. [9] A. Schubert, et al. resent Status of the MOX Version of the TU Code, EHG Meeting on High Burn-up Fuel erformance, Sandefjord, Norway, 9-14 May, [10] J.J. Carbajo, G.L. Yoder, S.G. opov, V.K. Ivanov, A review of the thermo-physical properties of MOX and UO2 fuels, Journal of Nuclear Materials 299 (181-8), 2001.

9 804.9 [11] D.D. Lanning, C.E. Beyer, roposed FRACON-3 MOX fuel thermal conductivity model compared to Halden fuel temperature data, EHG Meeting on High Burn-up Fuel erformance, Storefjell, Gol, Norway, 8-13 September, [12] W. Wiesenack, Assessment of UO2 Conductivity Degradation Based on In-pile Temperature Data, roceedings of the 97 International Topical Meeting on LWR Fuel erformance, ortland, Oregon, March 2-6, 97. [13] G. Delette, M. Charles, Thermal Conductivity of Fully Dense Unirradiated UO2: A New Formulation from Experimental Results Between 100 C and 2500 C, and Associated Fundamental roperties, IAEA TCM on Water Reactor Fuel Element Modelling at High Burnup and its Experimental Support, Windermere, UK, -23 September, 94. ABBREVIATIONS BH-95 Baron-Hervè 95 EBR-II Experimental Breeder Reactor II FBR Fast Breeder Reactor FFTF Fast Flux Test Facility FGR Fission Gas Release FR Fast Reactor F Fission roducts GEN-IV Generation-IV HEDL Hanford Engineering Development Laboratory ITU LHR LWR MOX OD O/M TD TU Institute of Trans-Uranium elements Linear Heat Rate Light Water Reactor Mixed Oxide fuel Outer Diameter Oxygen to Metal ratio Theoretical Density TRANSURANUS code