Numerical analysis of the venturi flowmeter in the liquid lead-bismuth eutectic circuit after long-term operation

The liquid Lead-bismuth eutectic is used as the coolant for Gen-IV reactor concepts. However, due to its strong corrosive and high operating temperature, it is difficult to accurately measure the flow rate in long-term operating conditions. Venturi flowmeter is a simple structured flowmeter, which plays a very important role in the flow measurement of high-temperature liquid metals, especially since the existing flowmeters are difficult to be competent. It has the advantages of easy maintenance and stable operation. Therefore, it is necessary to study the operating conditions of the venturi flowmeter under high-temperature conditions. This work performs a series of simulations of the fluid-solid interaction between the flow liquid metal and venturi flowmeter with COMSOL software, including the dimensional sensitivity analysis of the venturi flowmeter to explore the most suitable structure and parameters for liquid heavy metal, the sensitivity analysis of the geometric parameters of the venturi tube on the varying conditions. It shows that when the contraction angle of the venturi flowmeter is 33◦, the diffusion angle is 13◦, the diameter of the throat is 8 mm, and the temperature of the lead-bismuth eutectic is 733.15 K, it is most suitable for the measurement in the lead-bismuth circuit

AC Loss Reduction in REBCO Coated Conductors using the Hexagonal Arrangement Cabling Method

High-temperature superconducting (HTS) rare-earth barium copper oxide (REBCO) coated conductors with outstanding critical current density under high fields can help realize a high-field path toward magnetic-confinement fusion. REBCO cabling methods such as conductor on round core (CORC) cables, twisted stacked tape conductor (TSTC) cables, and Rutherford cables are based on the cable-in-conduit conductor (CICC) developed for low-temperature superconducting (LTS) Nb3Sn and Nb-Ti conductors. However, the REBCO coated conductor has challenges in achieving current transposition by twisting due to its ceramic-like mechanical behavior. In addition, it is more sensitive to external perpendicular magnetic fields with its rectangular cross-section than metallic LTS superconductors. In order to solve these issues, a hexagonal arrangement REBCO cabling method with the inherent advantage of mechanical protection and inductance balance is proposed in this paper. The electromagnetic behavior of REBCO coated conductors in the cable is evaluated using H-formulation and T-A formulation-based finite element methods. Results show that AC losses can be reduced using the hexagonal arrangement method compared with non-twisted cables and TSTC cables, which makes it a potentially helpful cabling method for ultra-high-field large-scale applications with high-level inductance balance requirements, especially the central solenoid coils of thermonuclear fusion reactors

ICONE14-89253 PREDICTIONS OF CRITICAL HEAT FLUX IN ANNULAR PIPES WITH TRACE V4.160 CODE

ABSTRACT This paper presents the assessment of TRACE (version v4.160) against the Critical Heat Flux (CHF) experiments in annular tubes performed at the Royal Institute of Technology (KTH) in Stockholm, Sweden. The experimental database includes data for coolant mass fluxes between 250 and 2500 kg/m 2 s and inlet subcoolings of 10 and 40 K at a pressure of 70 bar. The work presented in this paper supplements the calculations of single round tube experiments carried out earlier and provides a broader scope of validated geometries. In addition to the Biasi and CISE-GE CHF correlations available in the code, a number of experimental points at low flow conditions are available for the annular geometry experiments, which also permitted the assessment of the Biasi/Zuber CHF correlation used in TRACE v4.160 for low flow conditions. Experiments with different axial power distribution were simulated and the effects of the axial power profile and the coolant inlet subcooling on the TRACE predictions were investigated. The results of this work show that the Biasi/Zuber correlation provides good estimation of the CHF at 70 bar, and, for the same conditions, the simulation of the annular experiments resulted in the calculation of lower CHF values compared to single-tube experiments. The analysis of the performance of the standard TRACE CHF correlations shows that the CISE-GE correlation yields critical qualities (quality at CHF) closer to the experimental values at 70 bar than the Biasi correlation for annular flow conditions. Regarding the power profile, the results of the TRACE calculations seem to be very sensitive to its shape, since, depending on the profile, different accuracies in the predictions were noted while other system conditions remained constant. The inlet coolant subcooling was also an important factor in the accuracy of TRACE CHF predictions. Thus, an increase in the inlet subcooling led to a clear improvement in the estimation of the critical quality with both Biasi and CISE-GE correlations. To complement the work, three additional CHF correlations were implemented in TRACE v4.160, namely the Bowring, Tong W-3 and Levitan-Lantsman CHF models, in order to assess the applicability of these correlations to simulate the CHF in annular tubes. The improvement of CHF predictions for low coolant mass flows (up to 1500 kg/m 2 s) is noted when applying Bowring CHF correlation. However, the increase in the inlet subcooling increases the error in predicted critical quality with the Bowring correlation. The LevitanLantsman and Tong-W-3 correlations provide results similar to the Biasi model. Therefore, the most correct CHF predictions among the investigated correlations were obtained using CISE-GE model in the standard TRAC v4.160 code

Numerical analysis of the venturi flowmeter in the liquid lead-bismuth eutectic circuit after long-term operation

The liquid Lead-bismuth eutectic is used as the coolant for Gen-IV reactor concepts. However, due to its strong corrosive and high operating temperature, it is difficult to accurately measure the flow rate in long-term operating conditions. Venturi flowmeter is a simple structured flowmeter, which plays a very important role in the flow measurement of high-temperature liquid metals, especially since the existing flowmeters are difficult to be competent. It has the advantages of easy maintenance and stable operation. Therefore, it is necessary to study the operating conditions of the venturi flowmeter under high-temperature conditions. This work performs a series of simulations of the fluid-solid interaction between the flow liquid metal and venturi flowmeter with COMSOL software, including the dimensional sensitivity analysis of the venturi flowmeter to explore the most suitable structure and parameters for liquid heavy metal, the sensitivity analysis of the geometric parameters of the venturi tube on the varying conditions. It shows that when the contraction angle of the venturi flowmeter is 33°, the diffusion angle is 13°, the diameter of the throat is 8 mm, and the temperature of the lead-bismuth eutectic is 733.15 K, it is most suitable for the measurement in the lead-bismuth circuit

Two-way coupling between the reactor dynamics code DYN3D and the fuel performance code TRANSURANUS at assembly level

In the last two decades the reactor dynamics code DYN3D was coupled to thermal hydraulics system codes, a sub-channel code and CFD codes. These earlier developed code systems allow modeling of the thermal hydraulics phenomena occurring during reactor transients and accidents in more detail. Nevertheless, none of these code systems include a sufficiently sophisticated fuel behavior model, which is able i.e. to take into account the fission gas behavior during normal operation, off-normal conditions and transients. Furthermore, no two-way coupling to a fuel performance code has so far been reported in the open literature for calculating a full core with detailed and well validated fuel behavior models. A new two-way coupling approach between DYN3D and the fuel performance code TRANSURANUS is presented. In the coupling, DYN3D provides the time-dependent rod power and thermal hydraulics conditions to TRANSURANUS, which in turn transfers parameters like fuel temperature and cladding temperature back to DYN3D. The main part of the development is a so-called general TRANSURANUS coupling interface that is applicable for linking of any other reactor dynamics codes, thermal hydraulics system codes and sub-channel codes to TRANSURANUS. Beside its generality, other features of this interface are the application at either fuel assembly or fuel rod level, one-way or two-way coupling, automatic switch from steady to transient conditions in TRANSURANUS (including update of the material properties etc.), writing of all TRANSURANUS output files and possibility of manual pre- and post-calculations with TRANSURANUS in standalone mode. The TRANSURANUS code can be used in combination with this coupling interface in various scenarios: different fuel compositions in the reactor types BWR, PWR, VVER, HWR and FBR, considering time scales from milliseconds (i.e. RIA) over seconds/ minutes (i.e. LOCA) to years (i.e. normal operation) and thence different reactor states. Results of DYN3D-TRANSURANUS are shown for a control rod ejection transient in a German PWR. In particular, it appears that for all burn-up levels the two-way coupling approach systematically calculates higher maximum values for the node fuel enthalpy (max. difference of 46 J/g) and node centerline fuel temperature (max. difference of 181 K), compared to DYN3D standalone in best estimate calculations. These differences can be completely explained by the more detailed TRANSURANUS modeling of fuel thermal conductivity, radial power density profile and heat transfer in the gap. As known from fuel performance codes, the modeling of the heat transfer in the gap is sensitive and causes also larger differences in case of low burn-up. No convergence problems occurred for DYN3D-TRANSURANUS. The coupled code system can improve the assessment of safety criteria, at a reasonable computational cost with a CPU time of less than seven hours without parallelization.JRC.E.3-Materials researc

Getting the details of fuel rod simulation in reactor safety analysis right: performance of the code coupling DYN3D-TRANSURANUS for RIA

Over the last decades the importance of fuel-specific processes for safety analysis has grown, due to an increase of discharge burn-up for more efficient use of nuclear fuels. This results in a high burn-up structure (HBS) characterized by a decrease of the original grain size to ~0.1 - 0.2 μm, high concentration of pores with a diameter ~0.5 - 1 μm and depletion of fission gas from the matrix. It is known that the HBS can have an important effect on the fuel behavior during design basis accidents (DBA). For example, pellet cladding mechanical interaction (PCMI) can be observed under reactivity initiated accident (RIA) conditions, which can lead to fuel rod failure and in later stages to interaction between fuel and coolant. To analyze the behavior of high burn-up fuel in detail several experimental projects had begun, e.g. High Burnup Rim Project and the OECD/NEA Cabri Water Loop Project. Nevertheless, most of the reactor dynamics codes, thermal hydraulics system codes and sub-channel thermal hydraulics codes still include a simplified fuel behavior model. Thence licensing calculations concerning fuel rod performance are so far done in a conservative manner. However, today multi-physics code systems are more and more state-of-the-art thanks to hard- and software. For advanced safety analysis, two general trends can be observed in the field of fuel behavior modeling. On the one side, fuel performance codes are still being improved for a more accurate simulation of design basis accident (DBA) conditions. On the other side, the benefit and potential is being analyzed resulting from replacement of simplified fuel behavior models in neutronics, thermal hydraulics and CFD codes by a two-way coupling approach to a fuel performance code. Nevertheless, no real two-way coupling to a fuel performance code has so far been reported in the open literature for calculating a full core with detailed and well validated fuel behavior correlations. Thence in 2012 several German institutions wanted to couple their codes to the fuel performance code TRANSURANUS [16], which is used in EU by research organizations, safety authorities and industry. GRS has considered a coupling to its thermal hydraulics system code ATHLET, KIT presented a first external coupling approach for its sub-channel code KTF, and HZDR started a development for coupling TRANSURANUS to its reactor dynamics code DYN3D.JRC.E.3-Materials researc

Microhardness and local properties of high burnup UO2 fuel

This work presents Vickers microhardness indentation and Young's modulus results obtained from light water reactor (LWR) UO2 fuel at burnup ≥ 60 GWd/tHM. The hardness profiles are compared with porosity distribution profiles measured by Scanning Electron Microscopy image analysis, with particular focus on the pellet rim, where the high burnup structure (HBS) is present. The formation of HBS and porosity build-up leads to a noticeable change of apparent microhardness in the rim zone of the fuel pellet, primarily related to porosity increase. The microhardness was also measured on a longitudinal section. At each radial position, no significant difference was observed between the microhardness at the mid-pellet and at end of the pellet, close to pellet-pellet interface. The relationship between the local values of the microhardness and the Young's modulus is also evaluated. The current data provides the basis for an assessment of microhardness and other mechanical properties in relation with local fuel structure and conditions and supports the development of predictive models in fuel performance codes like TRANSURANUS.JRC.E.2-Safety of Irradiated Nuclear Material

Investigation of Feedback on Neutron Kinetics and Thermal Hydraulics from Detailed Online Fuel Behavior Modelling during a Boron Dilution Transient in a PWR with the Two-way Coupled Code System DYN3D-TRANSURANUS

Recently the reactor dynamics code DYN3D (including an internal fuel behavior model) was coupled to the fuel performance code TRANSURANUS at assembly level. The coupled code system applies the new general TRANSURANUS coupling interface, hence it can be used for one-way or two-way coupling. In the coupling, DYN3D provides process time, time-dependent rod power and thermal hydraulics conditions to TRANSURANUS, which in case of the two-way coupling approach replaces completely the internal DYN3D fuel behavior model and transfers parameters like radial fuel temperature distribution and cladding temperature back to DYN3D. For the first time results of the coupled code system are presented for a post-crisis heat transfer. The corresponding heat transfer regime is mostly film boiling, where the cladding temperature can rise up to several hundreds of degrees. The simulated boron dilution transient assumed an injection of a 36 m3 slug of under-borated coolant into a German PWR core initiated from a sub-critical reactor state (extreme RIA conditions). The feedback from detailed fuel behavior modeling was found negligible on the neutron kinetics and thermal hydraulics during the first power rise. In a later phase of the transient, the node injected energy can differ up to 22 J/kg, even for nodes without film boiling. Furthermore the thermal hydraulics can be affected strongly even in fresh fuel assemblies, where film boiling appeared in one node in the two-way approach in spite of no onset of film boiling in the one-way approach. For nodes with film boiling in both coupling approaches the two-way approach determined always higher maximum node average fuel enthalpies by about 100 J/g and higher maximum node clad surface temperatures by about 230 °C for a fresh fuel assembly. Since the numerical performance of DYN3D-TRANSURANUS was fast and stable for these extreme transient conditions, it is therefore concluded that the coupled code system can improve the assessment of safety criteria, at a reasonable computational cost.JRC.E.3-Materials researc

Post Irradiation Examination of Fast Reactor Metallic Fuel for Minor Actinides Transmutation

Advanced nuclear reactors and closed nuclear fuel cycles are important option to achieve sustainable nuclear energy supplies to satisfy future demands while reducing the long-term radiotoxicity of high level waste. Spent fuel reprocessing and the subsequent recycling of U and Pu as fuel and transmutation of Minor Actinides (MA) Np, Am, Cm in fast reactors are necessary steps to gain this goal. The METAPHIX programme is a collaboration between the Central Research Institute of Electric Power Industry (CRIEPI, Japan) and the Institute for Transuranium Elements (ITU, a Joint Research Centre of the European Commission) with the support of the Commissariat à l´Energie Atomique et aux Energies Alternatives (CEA, France) devoted to study the safety and effectiveness of a closed nuclear fuel cycle based on MA separation and irradiation in metallic fuel using fast reactor. In this frame, three assemblies containing nine Na-bonded experimental pins of metal alloy fuel prepared at ITU were loaded in the Phénix reactor in 2003 and irradiated at three different burn-up, 2.5 at.%, 7 at.% and 10 at%. Table I reports the average compositions of four different metallic alloys ingots fabricated: UPuZr used as reference, UPuZr + 2 or 5 wt.% MA, and UPuZr + MA + Rare Earths (RE). RE were added to reproduce the output of a pyrometallurgical reprocessing facility. PIE is performed in ITU to characterize the metallic fuel behavior in pile, considering properties ranging from the macroscopic morphology of the fuel matrix to the microanalysis of phase and elemental redistribution / segregation. This paper presents some highlights of the results obtained from the low and medium burn-up fuel examination.JRC.E.2-Safety of Irradiated Nuclear Material