Molten Salt Nuclear Reactor

Molten Salt Reactors (MSR) with the fuel dissolved in the liquid salt and fluoride-salt-cooled High-temperature Reactors (FHR) have many research themes in common. This paper gives an overview of the international R&D efforts on these reactor types carried out in the framework of Generation-IV. Many countries worldwide contribute to this reactor technology, among which the European Union, France, Japan, Russia and the USA, and for the past few years China and India have also contributed. In general, the international R&D focuses on three main lines of research. The USA focuses on the FHR, which will be a nearer-term application of liquid salt as a reactor coolant, while China also focuses on solid fuel reactors as a precursor to molten salt reactors with liquid fuel and a thermal neutron spectrum. The EU, France and Russia are focusing on the development of a fast spectrum molten salt reactor capable of either breeding or transmutation of actinides from spent nuclear fuel. Future research topics focus on liquid salt technology and materials behavior, the fuel and fuel cycle chemistry and modeling, and the numerical simulation and safety design aspects of the reactor. MSR development attracts more and more attention every year, because it is generally considered as most sustainable of the six Generation-IV designs with intrinsic safety features. Continuing joint efforts are needed to advance common molten salt reactor technologies.

In this book, a multi-physics approach to the modeling and analysis of nuclear reactor core behavior is presented, and applied to study the dynamics of Molten Salt Reactors (MSR). The Multi-Physics Modeling (MPM) approach is implemented in a unified simulation environment, which results able to catch the synergy between the different phenomena involved in the
reactor core behavior, whose modeling would otherwise require either the adoption of existing simulation tools with drastic modifications of their structure and possible loss of significant information or the development of purpose-made numerical codes for the specific analyzed situation.
The book is organized into four chapters, which are conceived so as to be read independently of each other. Here, the rationale of the R&D activities performed at the Politecnico di Milano by the authors, and giving rise to the book contents, is briefly outlined.
In the last years, MSRs have been the subject of a growing and renewed interest from the scientific community in the framework of the Generation IV International Forum. Due to the availability of an advanced conceptual design, the Molten Salt Breeder Reactor (MSBR) developed at the Oak Ridge
National Laboratory during the 1960s is usually considered in the literature as reference system for benchmark analyses and validation purposes. Coherently, the MSBR has been chosen as reference configuration for the analyses presented in the book. In such kind of Circulating Fuel Reactor (CFR), like the other MSRs, the coupling between neutronics and thermo-hydrodynamics is a key issue. This feature cannot be neglected in order to perform an adequate description of the reactor dynamic behavior, which shows peculiar aspects with respect to solid-fuelled conventional nuclear power plants. Relevant differences in terms of safety, fuel cycle and technology also distinguish the MSRs from the other nuclear reactors (Chapter 1).
The developed MPM methodology and the adopted models for neutronics and thermo-hydrodynamics required a deep investigation for what concerns the assessment and the extension of the simulation environment, represented in the specific case by the finite element COMSOL Multiphysics® software (chosen thanks to its flexible and modular numerical structure), but the same methodology can be applied to other multi-physics platforms. As far as thermo-hydrodynamics is concerned, a generalized approach was developed (Chapter 2) and exploited for the assessment of COMSOL simulations
(Chapter 3), making use of a dedicated finite volume computational fluid dynamics code (FLUENT®), in order to better appreciate the differences in numerical approaches to turbulence. The generalized approach was built in order to carefully take into account the molten salt mixture specificities (i.e., a fuel that operates also as coolant), the reactor core power conditions and the heat transfer in graphite. In this context, a Nusselt number correlation was developed, which takes into account the effect of internal heat generation on
fluid heat transfer characteristics. As far as neutronics is concerned, a module for the "reactor physics" was built in the COMSOL environment of simulation, and allowed to extend the potentialities of this computing platform. Numerical results were assessed by means of: (i) a code-to-code comparison with
dedicated neutron transport codes, in the case of static fuel; (ii) a comparison with simplified neutron kinetics models, representative of the zero-power dynamics, in the case of circulating fuel.
After the assessment of the COMSOL capabilities to cope with the adopted models for neutronics and thermo-hydrodynamics, the MPM approach was applied to study the dynamic behavior of a single-channel representative of the average conditions of the MSBR core (Chapter 4). For this case study, the MPM approach resulted particularly suitable because of the strong nonlinear coupling between the fuel motion and neutron dynamics, which requires a careful description of the time-space distribution of the physical quantities. Several different transients were analyzed, such as those driven by: reactivity variations due to control rod movements; fuel mass flow rate variations due to the change of the primary pump working conditions; presence of periodic perturbations, due to local precipitation of fissile solid compounds within the molten salt mixture. The analyses gave significant information on the MSBR dynamic behavior and highlighted the several advantages and potentialities offered by the proposed MPM approach (for instance, the "modularity", namely the possibility to include other physical phenomena and couplings).
These potentialities are of more general interest in the prospect of studying the design configuration, the dynamics and the control strategy of next generation MSRs, as well as of other nuclear reactor types.

The first efforts in the development of the molten salt reactor technology were carried out in the sixties by the Oak Ridge National Laboratory and culminated with the design of the thermal-spectrum Molten Salt Breeder Reactor (MSBR). Only recently, the attention has been focused on fast-spectrum configurations , such as the Molten Salt Fast Reactor (MSFR) proposed in the framework of the Euratom EVOL (Evaluation and Viability of Liquid Fuel Fast Reactor System) Project, thanks to their favourable characteristics in terms of sustainability, waste minimization and improved safety. As a matter of fact, the MSFR has been recognized as a long term alternative to solid-fuelled fast neutron systems and has been identified as Gen-IV reference MSR configuration. From the dynamic behaviour point of view, the main feature that characterises this kind of systems is the presence of a liquid fuel that circulates in the primary circuit acting simultaneously as coolant. This feature leads to a complex and highly coupled behaviour, which requires a careful investigation, due to some peculiarities like the drift of delayed neutron precursors along the primary circuit. Although considerable studies have been carried out for the analysis of the graphite-moderated MSRs, the adoption of a fast spectrum configuration without graphite in the core is expected to notably modify the dynamic characteristics of the system, thus requiring further investigation. This work proposes an approach to the dynamics and stability analysis of molten salt reactors. In particular, the well-developed methods of the theory of linear systems are applied to the analysis of two case studies, namely: the MSBR and the recently proposed MSFR. This analysis is intended to provide a basic understanding of the inherent stability properties and of the dynamic characteristics of such kind of nuclear reactors, highlighting the main peculiarities of the new design compared with the more familiar graphite-moderated concept.

Intisari— One Fluid-Molten Salt Reactor (OF-MSR) adalah salah satu dari sekian konsep yang dicanangkan untuk reaktor nuklir Generasi IV. Umumnya, reaktor ini menggunakan U-233 sebagai bahan bakar fisil start-up. Karena U-233 tidak tersedia di alam, maka reaktor ini harus menggunakan bahan bakar fisil lain untuk start-up, salah satunya U-235. Berdasarkan alasan tersebut, telah dilakukan analisis secara netronik terhadap desain teras One Fluid-Molten Salt Reactor berbasis bahan bakar LiF-U 238 F4-U 235 F4-ThF4 dengan kondisi bahan bakar baru. Dilakukan variasi pada pengayaan bahan bakar (12%; 15% dan 18%), perbandingan komposisi molar Thorium dan Uranium pada bahan bakar (70%:30%; 75%:25%; dan 80%:20%), dan jari-jari kanal bahan bakar (2 cm sampai 5 cm, dengan selisih 0,5 cm). Aspek netronik yang dianalisis yaitu faktor multiplikasi efektif (keff), rasio konversi (CR), serta aspek keselamatan yaitu koefisien reaktivitas suhu (αT) dan koefisien reaktivitas void (αϑ). Kalkulasi parameter netronik dilakukan dengan program Monte Carlo N-Particle Extended. Dari hasil penelitian, didapatkan bahwa desain optimum untuk One Fluid-Molten Salt Reactor adalah pada pengayaan 12%, perbandingan komposisi molar Th-U 70%:30%, dan jari-jari saluran core 2,5 cm. Nilai keff dan CR desain ini masing-masing 1,00253 ± 0,00065 dan 0,881. Dengan kata lain, desain ini belum bisa beroperasi sebagai reaktor pembiak pada kondisi awal. Sementara, koefisien reaktivitas suhu dan void masing-masing sebesar-0,00255 $/K dan-0,2131 $/% void, yang menunjukkan bahwa desain ini memenuhi syarat keselamatan melekat.

This paper presents a multi-physics modelling (MPM) approach developed for the study of the dynamics of the Molten Salt Reactor (MSR), which has been reconsidered as one of the future nuclear power plants in the framework of the Generation IV International Forum for its several potentialities. The proposed multi-physics modelling is aimed at the description of the coupling between heat transfer, fluid dynamics and neutronics characteristics in a typical MSR core channel, taking into account the spatial effects of the most relevant physical quantities. In particular, as far as molten salt thermo-hydrodynamics is concerned, Navier–Stokes equations are used with the turbulence treatment according to the RANS (Reynolds Averaged Navier–Stokes) scheme, while the heat transfer is taken into account through the energy balance equations for the fuel salt and the graphite. As far as neutronics is concerned, the two-group diffusion theory is adopted, where the group constants (computed by means of the neutron transport code NEWT of SCALE 5.1) are included into the model in order to describe the neutron flux and the delayed neutron precursor distributions, the system time constants, and the temperature feedback effects of both graphite and fuel salt. The developed MPM approach is implemented in the unified simulation environment offered by COMSOL Multiphysics Ò , and is applied to study the behaviour of the system in steady-state conditions and under several transients (i.e., reactivity insertion due to control rod movements, fuel mass flow rate variations due to the change of the pump working conditions, presence of periodic perturbations), pointing out some advantages offered with respect to the conventional approaches employed in literature for the MSRs.

Hightemperature absorption spectroscopy is used to study the effect of temperature in the range 550–750°C and the cation composition of a solvent salt on the spectroscopic characteristics and stability of [MoCl3]6- complex ions in melts based on alkali metal chlorides (LiCl; CsCl; LiCl–KCl, NaCl–CsCl,
KCl⎯CsCl, and NaCl–KCl–CsCl eutectic mixtures; and equimolar NaCl–KCl mixture).

The aim of this paper is the extension of a multiphysics OpenFOAM solver for the analysis of the Molten Salt Fast Reactor (MSFR), developed in previous works (Cervi et al., 2017, 2018). In particular, the neutronics sub-solver is improved by implementing a new module based on the SP3 approximation of the neutron transport equation. The new module is successfully tested against a Monte Carlo model of the MSFR, in order to assess its correct implementation. Then, a neutronics analysis of the MSFR is carried out on a simplified axial-symmetric model of the reactor. Particular focus is devoted to the analysis of the MSFR helium bubbling system and its effect on reactivity. The presence of bubbles inside the reactor is handled with a two-fluid thermal-hydraulics module, previously implemented into the solver. The void reactivity coefficient is evaluated on the basis of the bubble spatial distribution calculated by the multiphysics solver. Then, the results are compared to simulations carried out with uniform bubble distributions, highlighting significant differences between the two approaches. The
outcomes of this work constitute a step forward in the multiphysics analysis of the Molten Salt Fast Reactor and represent a useful starting point for the optimization of the MSFR helium bubbling system, as well as for the development of appropriate control strategies.

Chloro-species of molybdenum in oxidation states (III), (IV) and (V) were studied in a range of alkali chloride melts employing high temperature electronic absorption spectroscopy. The reaction rates of disproportionation of MoCl 6 3complex ions in various melts between 550 and 750 o C were determined. Mo(IV) species MoCl 6 2were generated and thermodynamic properties of Mo(IV) chloride in NaCl-CsCl eutectic melt estimated. Absorption spectra of molybdenum(V) MoCl 6 were measured in a chloride melt for the first time. Speciation and behavior of molybdenum is required for designing processes for molten salt reprocessing of spent nuclear fuel (SNF).

Riset terhadap reaktor generasi IV telah dilakukan berbagai negara hingga saat ini, untuk meningkatkan kualitas dan mengatasi persoalan-persoalan reaktor yang ada pada masa sekarang. Salah satunya adalah riset terhadap reaktor maju Molten Salt Reactor (MSR) yang menggunakan bahan bakar yang dilarutkan dalam garam cair. MSR diklaim memiliki prospek yang bagus untuk digunakan sebagai reaktor generasi IV. MSR menggunakan konsep Full-Passive Safety System dan Inherent Safe untuk menjamin keselamatan reaktor. Mengaplikasikan siklus bahan bakar Thorium, reaktor ini bersifat thermal breeder, yang diperlukan untuk mengurangi penggunaan Uranium-235 dan menjamin sustainabilitas energi. Disamping itu, MSR bekerja pada suhu operasi tinggi, sehingga memungkinkan utilisasi panas keluaran untuk berbagai keperluan termal selain listrik.

Molten Salt Reactors (MSR) with the fuel dissolved in the liquid salt and fluoride-salt-cooled Hightemperature Reactors (FHR) have many research themes in common. This paper gives an overview of the international R&D efforts on these reactor types carried out in the framework of Generation-IV. Many countries worldwide contribute to this reactor technology, among which the European Union, France, Japan, Russia and the USA, and for the past few years China and India have also contributed. In general, the international R&D focuses on three main lines of research. The USA focuses on the FHR, which will be a nearer-term application of liquid salt as a reactor coolant, while China also focuses on solid fuel reactors as a precursor to molten salt reactors with liquid fuel and a thermal neutron spectrum. The EU, France and Russia are focusing on the development of a fast spectrum molten salt reactor capable of either breeding or transmutation of actinides from spent nuclear fuel. Future research topics focus on liquid salt technology and materials behavior, the fuel and fuel cycle chemistry and modeling, and the numerical simulation and safety design aspects of the reactor. MSR development attracts more and more attention every year, because it is generally considered as most sustainable of the six Generation-IV designs with intrinsic safety features. Continuing joint efforts are needed to advance common molten salt reactor technologies.

In the frame of a major re-evaluation of the molten salt reactor (MSR), we have developed a new concept called Thorium Molten Salt Reactor (TMSR), particularly well suited to fulfill the criteria chosen by the Generation IV forum. This reactor may be operated in simplified and safe conditions in the Th/ 233 U fuel cycle with fluoride salts. Amongst all TMSR configurations, many studies have highlighted the configurations with no moderator in the core as simple and very promising. Since 233 U does not exist on earth and is not being produced today, we aim at designing a critical MSR able to burn the Plutonium and the Minor Actinides produced in today's reactors, and consequently to convert this Plutonium into 233 U. Thus, the current fuel cycle can be closed thanks to TMSRs started with transuranic elements on a Thorium base, i.e. started in the Th/Pu fuel cycle, similarly to fast neutron reactors operated in the U/Pu fuel cycle. We analyze the characteristics of these reactor configurations, in terms of fissile matter inventory, salt reprocessing, waste production and burning, and finally deployment capabilities. Using a simple kinetic-point model, we analyze the reactor's behaviour as the total reactivity margins are introduced in the core. We thus confirm, beyond the classical advantages of molten salt reactors, the satisfactory behaviour of the TMSR and the excellent level of deterministic safety which can be achieved in such reactors. We then illustrate how the reactor can be driven with no control rod, either by controlling the extracted power or by monitoring the operating temperature. Finally we stress the hardiness and the flexibility of this TMSR concept, allowing it to be adjustable without loosing its advantages in the event of any technological stumbling block.

Molten Salt Reactor (MSR) designs are frequently accompanied by a blanket salt. This way the irradiation of the outer reactor wall will be strongly reduced. On the other hand, the barrier between the core and blanket will undergo higher irradiation and it will be necessary to replace it several times during the reactor lifetime. Furthermore, this blanket salt will also have a positive impact on neutron economy by improving the breeding performance. In this paper a blanket of a generic two fluid molten salt reactor utilizing fast thorium-uranium cycle was investigated. This was done by tracking the evolution of uranium, neptunium and plutonium isotopes with burnup, which was then influenced by removal of uranium from the blanket. A significant reduction in the production of minor actinides was observed. The uranium vector removed from the core was then investigated for proliferation resistance, using NUREC proliferation resistance metric and comparison with other weapon designs. The evaluation concluded that while the presence of U-232 increases radiological hazard associated with this uranium, thereby erecting a radiological barrier, it cannot be treated as "self-protecting" based on IAEA and NRC standards, requiring 1 Sv/h at 1m dose rate. Moreover ideas on how an interested party could reduce this radiological hazard were discussed.

In this work, the PSI FAST code system is extended for the modeling and simulation of Molten Salt Reactor (MSR) dynamics. The thermal-hydraulic code TRACE has been provided with a module for 1-D Delayed Neutron Precursor (DNP) balance and decay heat modeling in fluid fuel, and with built-in MSR materials. The reactor power is determined by means of a Point-Kinetics approach that makes use of power-weighted values of temperature and DNP distributions in the core. To validate the module, models for the comparison with experiments performed in  the Molten Salt Reactor Experiment (MSRE) have been developed. The MSRE was a graphite moderated reactor built and operated in the sixties at Oak Ridge National Laboratory (ORNL). The neutronic characteristics of the reactor have been determined by means of the Monte Carlo code SERPENT. Two models of the MSRE plant with different detailing have been set up in order to determine the importance of the plant components. In particular, different descriptions of the external cooling loop have been tested and compared. Some significant transients have been considered for the module assessment, by comparison with available experimental data from ORNL, both in time and frequency domain.

The unique design features of the molten salt fast reactor (MSFR) should enable higher coolant temperatures than in conventional water reactors, with a significant improvement in the achievable thermodynamic performance. The use of a molten salt as both fuel and coolant, however, poses several advanced heat transfer challenges, such as the design of innovative heat exchangers and energy conversion systems. In this work, we address a preliminary but quantitative analysis of the energy conversion system for the MSFR, based on reference design data from the SAMOFAR H2020-EURATOM project. We consider three main technologies, i.e., the supercritical steam cycle, the closed helium cycle and the helium/steam combined cycle. Preliminary design results are presented for each technology, based on a simplified modelling approach. The considered cycles show promising efficiency improvements, with the best performance being proven by the supercritical steam cycle. The analysis also highlights th...

The Molten Salt Fast Reactor (MSFR) has been recognised as a promising reactor concept and has been adopted as reference circulating-fuel MSR configuration in the framework of the Gen-IV International Forum. Up to now, a number of dynamic studies have been performed to investigate the transient response of the system and its stability, but no control-oriented analyses have been carried out yet. This work focuses on a preliminary evaluation of the MSFR controllability issues, assuming a decentralized feedback scheme of control. For this purpose, a simplified model of the reactor primary circuit has been set up adopting a point-kinetics description and a one-dimensional description of neutronics and thermal–hydraulics, respectively. First, the effect of all the possible input variables on different sets of output variables is investigated by means of the Relative Gain Array (RGA) analysis, providing a brief overview of the best input/output pairings with no restrictions on the number of control variables that can be used. Some control schemes are then selected and implemented in the model of the primary circuit. The transient responses are analysed and compared, considering different control strategies. This work is thought to provide a first insight into some MSFR controllability issues as well as a preliminary evaluation of possible control options.

In the last years, there has been a rapid growth of research and development activities on high temperature molten salts for nuclear and non-nuclear applications. The study of their heat transfer characteristics is a key issue in the current development of the molten salt reactor (MSR) that is one of the innovative nuclear reactors proposed by the Generation IV International Forum. The MSR is a sort of circulating fuel reactor (CFR), which adopts a molten halide salt mixture containing the fissile material and playing the distinctive role of both fuel and coolant. In a thermal-neutron-spectrum MSR, the reactor core is composed by a graphite matrix (neutron moderator), through which the liquid nuclear fuel flows and leads to a strong and intrinsic coupling between thermo-hydrodynamics and neutronics. This peculiar feature requires a suitable and qualified multi-physics simulation environment for a proper description of the system (fuel/coolant +graphite) behaviour. With reference to such complex and non-linear system, the present work is intended to give two different but linked contributions, in the perspective of a multi-physics modelling able to accurately describe the synergy of the involved different physical phenomena (e.g., by means of COMSOL Multiphysics s , 2008 software). (I) On one hand, this work is aimed at establishing a useful validation framework for the assessment of computational fluid-dynamics (CFD) analyses of liquids with internal heat generation. For this purpose, an analytic approach for both fluid velocity and temperature fields in a circular pipe surrounded by a solid region has been developed. It is an extension of the approach elaborated for pipe flow in (Di Marcello, V., Cammi, A., Luzzi, L., 2010. A generalized approach to heat transfer in pipe flow with internal heat generation. Chemical Engineering Science 65, 1301 À 1310), in order to take into account the heat conduction in the solid domain (represented by the graphite matrix in the specific case of interest). (II) On the other hand, the paper presents a preliminary investigation of the thermal-hydraulic behaviour occurring in a typical graphite-moderated MSR core channel. The developed analytic approach has been conveniently applied to this case study: (i) by testing the capabilities of COMSOL Multiphysics s (2008) to evaluate the heat transfer characteristics and the hydrodynamic behaviour of such system; and (ii) by investigating the applicability of correlations for the Nusselt number to fluids with internal heat generation. For a deeper insight into the numerical solutions provided by COMSOL, a code-to-code comparison has been also carried out, adopting a dedicated CFD finite volume software (i.e., FLUENT s). As a result, a satisfactory agreement has been found between the analytic solution and the numerical computations provided by COMSOL and FLUENT for a wide range of Reynolds numbers. As concerns the Nusselt number evaluation, the correlations usually adopted for molten salts have been proved as unsuitable for the analysed system, whereas the correlation advanced in the above mentioned paper (that explicitly takes account of the internal heat source) appears more appropriate. The results of the present work are thought to be useful because they provide: (i) a ''generalized'' analytic approach to the heat transfer that is applicable in a more general context; (ii) an insight into the heat transfer characteristics of the considered graphite-moderated MSR core channel; and (iii) a preliminary assessment of computations that is essential in view of the adoption of multi-physics tools (like COMSOL) for more complex and representative simulations of the dynamic behaviour of the molten salt reactor, and more in general of other kinds of CFRs.

Molten salt reactors, in the configuration presented here and called Thorium Molten Salt Reactor (TMSR), are particularly well suited to fulfil the criteria defined by the Generation IV forum, and may be operated in simplified and safe conditions in the Th/ 233 U fuel cycle with fluoride salts. The characteristics of the nonmoderated TMSR based on a fast neutron spectrum are detailed in this paper: we aimed at designing an optimised TMSR with the simplest configuration. Using a simple kinetic-point model, we analyze the reactor’s transient as the total reactivity margins are introduced in the core. We thus confirm, beyond the classical advantages of molten salt reactors, the satisfactory behaviour of the TMSR in terms of safety and the excellent level of stability which can be achieved in such reactors.

Fusion-fission hybrid systems (FFH) represent a coupling between a fusion device and a subcritical fission reactor driven by neutrons produced by fusion reactions. This kind of systems, in principle, can be useful for different purposes, for example, as energy amplifier or as nuclear waste burner. In this work, the preliminary characteristics and potentialities of a FFH based on a tokamak device characterized by high magnetic field (B > 9 T) and high-density plasma (n > 1014 cm–3) have been evaluated. During the years high magnetic field compact tokamaks have been designed, built and operated. Thanks to their characteristics, such as compactness, high field and high density plasma, these devices can produce intense neutron fluxes, and therefore are good candidates to be incorporated in FFH while operating in a sub-ignited regime. An additional advantage is that their design is based on relatively simple existing technology. In this work, a coupling between a tokamak (operating in DD-mode) and a subcritical molten salt fission blanket has been proposed. Molten salt reactor could be adapted for this purpose and could help the hybrid system to increase its energy balance. A model with a molten salt fission blanket (keff = 0.92, P = 95 MW) instead of the lithium one, has been considered and studied in terms of neutronic evaluations.
Presented preliminary numerical calculations based on a neutron Monte-Carlo code confirm the potentialities of the system.

Molten salt reactors in the configurations presented here, called Molten Salt Fast Reactors (MSFR), have been selected for further studies by the Generation IV International Forum. These reactors may be operated in simplified and safe conditions in the Th/ 233 U fuel cycle with fluoride salts. We present here the concept, before focusing on a possible optimization in term of minimization of the initial fissile inventory. Our studies demonstrate that an inventory of 233 U lower than 4 metric tons per GWe may be easily reached, and bring to light the limitations of the concept due to the irradiation damages to the structural materials and to the capacities of the heat exchangers. We conclude that these two issues will have to be studied in depth to allow a realistic evaluation of the global possibilities of such a reactor.

In this research, the simulation of passive cooling system in PCMSR’s waste container has been done by using conduction, free convection heat transfer calculation. Simulation was first assumed in transient condition when the fission product liquid salt 7LiF-BeF2 just entered the waste container. Then, heat transfer analysis was done in steady state condition.

Under these conditions, the fluid temperature was kept below its boiling point of 1600 K to avoid the potential of severe accidents. Thus cooling was carried out by radially transferring heat to the surface of the waste container, where a channel of boiling water as a heat exchanger media was located. Simulation was conducted for heat generation values at 12667 kW/m3. Low-Re model is used in FLUENT to predict free-convection behaviour.

The simulation results using the FLUENT software showed that cooling process byfree convection heat transfer to the surface of waste container was capable to reduce fission product decay heatof 12667kW∙m-3. The liquid waste temperature reached 1350 K in steady state condition.The grid with total elements of 640000 is used. Heat transfer coefficient value is 7509 W∙m-2∙K-1 with 633 kW∙m-2 heat flux.