Lead-bismuth Eutectic Coolant Research Articles

Po-Containing Molecules in Fusion and Fission Reactors.

Fission and fusion reactors can only play a role in the future energy landscape if they are inherently safe by design. For some reactor concepts, a major remaining issue is the undesired production of radiotoxic 210Po. To filter out the volatile Po species, information on their molecular composition is needed. An experimental characterization is very challenging due to the large required amount of radioactive Po. An alternative quantum chemistry approach was taken to predict the temperature-dependent stability of relevant diatomic Po-containing molecules. Experimental data on lighter analogue molecules was used to establish a well-founded methodology. The relative occurrence of the Po species was estimated in the cover gas of (i) the lead-bismuth eutectic coolant in the accelerator-driven MYRRHA fission reactor and (ii) the Pb-Li eutectic tritium breeder in the DEMO fusion reactor. In both systems, Po is found to occur mainly as PbPo molecules and atomic Po.

Sub-channel analysis for Pb-Bi-cooled direct contact boiling water fast reactor

Pb-Bi-cooled direct contact boiling water fast reactor (PBWFR) featured with a direct-contact heat exchanger between lead-bismuth eutectic coolant and water could significantly simplify the primary system and enhance the natural circulation capability, meeting the potential needs for small modular reactor design. It is of great importance to conduct thermal-hydraulic analysis of the PBWFR core in detail. In this paper, a self-developed SUB-channel AnalysiS code SUBAS is adopted to study the thermal hydraulic characteristics of the PBWFR core. The fidelity and the reliability of the code have been preliminarily benchmarked. With SUBAS, the space grid is studied to figure out its impact on the temperature and flow distributions in each sub-channel. Besides, the application of space grids would increase the pressure drops and decreases the cross flow between adjacent sub-channels. To study the transient performance of the PBWFR core, the power transient and the inlet blockage accident are calculated by SUBAS. The results of the power transient show the cross-flow effect would be weakened in the sub-channel which has higher coolant temperature and larger mass flow rate. For the inlet blockage accident, the results indicate the influence of the small area blockage is relatively weak on the overall performance of the assembly but is significant on the local parameters. With consideration of time and space, the blockage influence only exists in a certain area. This research may provide contribution to the design of PBWFR.

3D numerical study of LBE-cooled fuel assembly in MYRRHA using SIMMER-IV code

The present paper is based on the work carried out in the framework of the European FP7 project MAXSIMA, in which MYRRHA safety studies are performed. MYRRHA is a pool-type 100MWth system with MOX fuel designed to operate both in ADS and critical modes. It uses lead-bismuth eutectic (LBE) as primary coolant. The MOX fuel has almost the same density as the LBE coolant. In case of pin failure fuel pellets may break into chunks and particles carried by the coolant upwards and redistributed in the reactor pool. The transmutation group at IKET/KIT mainly with the numerical analysis tool is involved for studying severe accidents for MYRRHA reactor design. The highlight of the current work is that 3D simulations with explicit modelling on the gaps between fuel assemblies and 3D macroscopic pin bundle models are performed for the first time using a reactor safety analysis code, SIMMER-IV, with 3D geometry. In this paper, the numerical analyses are conducted for a single fuel assembly blockage and 19 pin-rods on basis of an LBE coolant experiment. The 3D analysis has been applied with both scales namely fuel assembly scale and pin bundle scale. For the fuel assembly scale, the evaluation of a single fuel assembly blockage using non-axisymmetric geometry configuration in the subcritical mode is addressed. For the pin bundle scale, the 3D pin bundle simulations show a good agreement from the experiment conducted at KALLA liquid metal laboratory. Note that this is the only code applied by now for blockage analyses after pin failure in MYRRHA. The current work has formed a solid basis for the safety analysis for MYRRHA in the future.

Power maximization method for land-transportable fully passive lead–bismuth cooled small modular reactor systems

Although current pressurized water reactors (PWRs) have significantly contributed to global energy supply, PWR technology has not been considered a trustworthy energy solution owing to its problems of spent nuclear fuels (SNFs), nuclear safety, and nuclear economy. In order to overcome these problems, a lead–bismuth eutectic (LBE) fully passive cooling small modular reactor (SMR) system is suggested. This technology can not only provide the solution for the problems of SNFs through the transmutation feature of the LBE coolant, but also strengthen safety and economy through the concept of natural circulation cooling SMRs. It is necessary to maximize the advantages, namely safety and economy, of this type of nuclear power plants for broader applications in the future. Accordingly, the objective of this study is to maximize the reactor core power while satisfying the limitations of shipping size, materials endurance, and criticality of a long-burning core as well as safety under beyond design basis events. To achieve these objectives, the design limitations of natural circulating LBE-cooling SMRs are derived. Then, the power maximization method is developed based on obtaining the design limitations. The results of this study are expected to contribute to the effectiveness of the reactor design stage by providing insights to designers, as well as by formulating methods for the power maximization of other types of SMRs.

Developing Procedures for Fracture Toughness Tests in Lead-bismuth Eutectic

T91 ferritic-martensitic steel is one of the candidate materials to future fast reactors where lead-bismuth eutectic (LBE) is chosen as the coolant. The interaction of the steel with LBE coolant should be investigated by means of mechanical tests and detailed microstructural analysis. Quantitative analysis requires fracture toughness tests in LBE. However tests in LBE are not easily applied with conventional methods due to the conductive nature of LBE. In addition, there are no standards or guidelines for fracture toughness tests in LBE. In this work fracture toughness tests in inert environment and LBE are compared and technical problems encountered during tests in LBE are discussed.

CFD analysis of the MYRRHA primary cooling system

MYRRHA is a pool type reactor system, with all the systems immersed in the lead-bismuth eutectic coolant. Due to corrosion and solidification concerns, there is a strong incentive to know the flow patterns inside the primary cooling system. Since the flow inside the vessel is typically three-dimensional, it is simulated by the computational fluid dynamics software FLUENT. For these simulations, the Renormalization group k−∈ turbulence model has been chosen. This turbulence model is used in combination with physical simplifications and a porous medium approach to simulate the reactor behavior. Nevertheless, these simplifications show to lead to a rather good representation of the design parameters.The simulations for nominal conditions show that, even with a high temperature margin derived from the ASCHLIM benchmark simulations, no corrosion or solidification temperature limits are exceeded. Nevertheless, some design features are found that might need some reconsideration. In addition to that, an extrapolation was made to a steady state, natural convection scoping simulation.

Neutronic model of a mirror based fusion–fission hybrid for the incineration of the transuranic elements from spent nuclear fuel and energy amplification

The Georgia Institute of Technology has developed several design concepts of tokamak based fusion–fission hybrids for the incineration of the transuranic elements of spent nuclear fuel from Light-Water-Reactors. The present paper presents a model of a mirror hybrid. Concerning its main operation parameters it is in several aspects analogous to the first tokamak based version of a “fusion transmutation of waste reactor”. It was designed for a criticality k eff ≤ 0.95 in normal operation state. Results of neutron transport calculations carried out with the MCNP5 code and with the JEFF-3.1 nuclear data library show that the hybrid generates a fission power of 3 GW th requiring a fusion power between 35 and 75 MW, has a tritium breeding ratio per cycle of TBR cycle = 1.9 and a first wall lifetime of 12–16 cycles of 311 effective full power days. Its total energy amplification factor was roughly estimated at 2.1. Special calculations showed that the blanket remains in a deep subcritical state in case of accidents causing partial or total voiding of the lead–bismuth eutectic coolant. Aiming at the reduction of the required fusion power, a near-term hybrid option was identified which is operated at higher criticality k eff ≤ 0.97 and produces less fission power of 1.5 GW th. Its main performance parameters turn out substantially better.

Diffusional stability of ferritic–martensitic steel composite for service in advanced lead–bismuth cooled nuclear reactors

AbstractTo combat liquid metal corrosion from lead–bismuth eutectic coolant in the fuel cladding and coolant piping of generation IV fast fission reactors, a composite that employs a Fe–Cr–Si steel layer weld clad on a structural layer of alloy T91 (Fe–9Cr–1Mo) is being developed. Diffusion of Si away from the cladding during service can compromise corrosion resistance, whereas carbon redistribution will affect mechanical properties and phase stability. Diffusion of silicon and carbon in a manufactured sample of the composite has been investigated both experimentally and by modelling. The diffusion coefficient for silicon in the T91/Fe–12Cr–2Si system was found to be 1·36 × 10−12 cm2 s−1 at 650°C, implying that concentration in the clad layer should not drop below 1·25%Si at a distance of 51 μm from the interface after 50 years' service at 650°C. The rapid carbon diffusion observed should be helpful during processing and operation. The two layers will have similar properties once carbon diffuses into the ...

A pressure-tube Advanced Burner Test Reactor concept

A new steady-state fast neutron test reactor has been conceptually designed. This paper presents a new concept for a CANDU-based fast test reactor that is horizontal in orientation, with individual pressure/Calandria tubes (PT/CT) running the entire length of the scattering-medium tank (Calandria) filled with Lead-Bismuth-Eutectic (LBE). This approach for a test reactor will provide more flexibility in refueling, sample removal, and ability to completely reconfigure the core to meet different users’ requirements. Full core neutronic analysis showed a small hexagonal reactor, LBE-cooled, trans-uranics (TRU)-67Zr fuel with HT-9 cladding and structures, with a core power of 100 MW th produced a fast flux (>0.1 MeV) of 1.4 × 10 15 n/cm 2 s averaged over the whole length of six irradiation channels with a total testing volume of more than 77 l. Loading of TRU from legacy UO 2 spent fuel allowed core continuous operation for 180 effective full power days with a net fissile-Pu burning rate of ∼6.4%. Since high neutron fluence impact on PT/CT might be an issue of concern for this design, oxide dispersion strengthened (ODS) ferritic steel was used as PT/CT material without any impact on core neutronic behavior. An innovative shutdown/control system which consisted of the six outermost fuel channels was proven to be effective in shutting the core down when flooded with boric acid as a neutron absorber. The new shutdown/control system has the advantage of causing the minimum perturbation of the axial flux shape when the control channels are partially flooded with boric acid. A preliminary thermal-hydraulic analysis of LBE produced acceptable pumping power and clad temperature. Voiding the core from its LBE coolant resulted in a positive reactivity insertion which is typical of most fast reactors. Hypothetical accidents of draining the Calandria and/or the external LBE reflector tank of the LBE core resulted in negative reactivity insertion which shut the reactor down.

Development of polonium surface contamination measure in lead-bismuth eutectic coolant

Polonium contamination is one of the issues to be solved for use of lead-bismuth eutectic in a nuclear system. It is important to develop an effective method to remove polonium surface contamination. Some experiments were performed to remove polonium surface contamination by baking method. The results showed that characteristics are different between quartz glass contamination removal and stainless steel contamination removal. A method to estimate polonium density distribution in metal useful for investigation of polonium surface contamination was developed.

Corrosion/precipitation in non-isothermal and multi-modular LBE loop systems

Precise simulations of all hydrodynamic conditions encountered in practical lead–bismuth eutectic (LBE) coolant loop systems by use of flowing conditions in the laboratory are difficult and expensive, if not impossible. Therefore it is important and necessary to develop corrosion models to predict corrosion behaviors at the design stage of practical LBE coolant systems and to properly interpret and apply experimental results from test loops. In the present study, we extended a kinetic corrosion model for a simple LBE loop to a model for a loop with multiple modules to include effects of geometry variations. The model is applied to an ideal loop with contractions and expansions and a test loop named ‘DELTA’ set up in the Los Alamos National Laboratory. Analyses show the combined effects of the axial geometry variations and the axial temperature profile on the corrosion/precipitation distribution in the entire loop. The present study illustrates systematically different dependence of corrosion behaviors on the hydraulic factors in an open pipe flow, a simple loop flow and a multi-modular loop flow.

Optimization of Height-to-Diameter Ratio for an Accelerator-Driven System

The height-to-diameter (H/D) ratio of a lead-bismuth eutectic (LBE)-cooled accelerator-driven system (ADS) has been evaluated in terms of the neutron multiplication, the coolant void worth, and the coolant velocity. For a model ADS, an optimization of the H/D ratio is performed with a Monte Carlo code both for the effective multiplication factor keff and for the multiplication of the external neutrons. In the optimization, ten cases of H/D values have been analyzed for a homogeneous fuel blanket. Also, the dependency of the optimal H/D ratio on the target/buffer is addressed. The Monte Carlo simulations show that the optimal H/D configuration of the ADS core is quite different for the two important measures, and a high H/D ratio can provide a significantly higher source multiplication than the traditional pancake core. Furthermore, various core analyses including depletion calculations are conducted for three selected heterogeneous cores with different H/D ratios, which are a small H/D value (pancake type), a medium H/D value, and a high H/D value, respectively. Void reactivity coefficients of the LBE coolant are evaluated and compared for the three designs to quantify the effects of the H/D ratio. Additionally, a thermal-hydraulic analysis has been performed to derive a maximum allowable core height subject to the LBE velocity limit due to its corrosion and erosion characteristics. It is shown that the practically optimal H/D ratio for source multiplication is tightly constrained by the maximum allowable LBE velocity, depending on the core design parameters.