Eulerian Two-Phase Flow Modeling of Steam Direct Contact Condensation for the Fukushima Accident Investigation

Steam condensation is characterized by a relatively large interfacial region between gas and liquid which, in computational fluid dynamic (CFD) analyses, allows the creation of a discretized domain whose average cell size is larger than the interface itself. For this reason generally one fluid model with interface tracking (e.g. volume of fluid method, VOF) is employed for its solution in CFD, since the solution of the interface requires a reasonable amount of cells, reducing the modeling efforts. However, for some particular condensation applications, requiring the computation of long transients or the steam ejected through a large number of holes, one-fluid model becomes computationally too expensive for providing engineering information, and a two-fluid model (i.e. Eulerian two-phase flow) is preferable. Eulerian two-phase flow requires the introduction of closure terms representing the interactions between the two fluids in particular, in the condensation case, drag and heat transfer. Both terms involve the description of the interaction area whose definition is different from the typical one adopted in the boiling analyses. In the present work a simple but effective formulation for the interaction area is given based on the volume fraction gradient and then applied to a validation test case of steam bubbling in various subcooling conditions. It has been shown that this method gives realistic values of bubble detachment time, bubble penetration for the cases of interest in the nuclear application and in the particular application to the Fukushima Daiichi accident.

CFD Analysis of Single Phase Counter-Current Buoyancy Driven Flows and its Applications to Passive Reactor Design

Increasingly, nuclear plants rely on natural circulation, for both fault conditions and / or normal power removal. Prediction of such buoyancy-driven flows is needed. However, their complex nature leads to 3D effects in ‘wide’ geometries, making prediction impossible with system codes. Even in slender “pipe-like” geometries countercurrent flow of hot and cold fluid makes a one-dimensional simulation totally misleading. However, simply moving to a three-dimensional CFD treatment is not sufficient. The strong anisotropy of the turbulence and the coexistence of various flow regimes make the choice of an appropriate turbulence model difficult. Countercurrent flow in a pipe might occur when the “natural” buoyant flow was of hot fluid up the pipe, but a feature such as a local heat-sink (an un-insulated valve in the pipe, perhaps) acts as a source of cold fluid, which attempts to flow down the pipe as a counter-current flow. On a different scale, counter current flow such as this would occur for example inside the secondary containment. This countercurrent flow problem captures the complexities of most buoyant flows, and this provides a challenging model problem. In this paper, we describe the design and preliminary analysis of an experimental rig being built to study this. Initial CFD and experimental results are presented.

Decommissioning of Obsolete Nuclear Facilities in the UJV Rez, a. s.

More than 55 years of activities in the company UJV Rez, a. s. (Nuclear Research Institute Rez a.s. in the past) which is a leading institution in all areas of nuclear R&D in the Czech Republic and had a dominant position in the nuclear program since it was established (1955), there are several obsolete nuclear facilities that shall be decommissioned. The total amount of radioactive waste (RAW) resulting from decommissioning for the next processing will be ∼ 1500 m3 and the expected amount RAW for releasing into the environment is 240 tons after the decontamination. For the RAW processing several decontamination methods such as high press water jetting, chemical treating in ultrasonic bath, dry ice blasting and abrasive blasting were performed. Decommissioning started in 2003 and will be finished in 2016. This decommissioning of nuclear facilities in UJV is the only ongoing decommissioning project in the Czech Republic.

The Super Equivalence Method in Monte Carlo Based Homogenization

The general equivalence theory (GET) and the superhomogenization method (SPH) are widely used for equivalence in the standard two-step reactor physics calculation. GET has behaved well in light water reactor calculation via nodal reactor analysis methods. The SPH was brought up again lately to satisfy the need of accurate pin-by-pin core calculations. However, both of the classical methods have their limitations. The super equivalence method (SPE) is proposed in the paper as an attempt to preserve the surface current, the reaction rates and the reactivity. It enhances the good property of the SPH method through reaction rates based normalization. The concept of pin discontinuity factors are utilized to preserve the surface current, which is the basic idea in the GET technique. However, the pin discontinuity factors are merged into the homogenized cross sections and diffusion coefficients, thus no additional homogenization parameters are needed in the succedent reactor core calculation. The eigenvalue preservation is performed after the reaction rate and surface current have been preserved, resulting in reduced errors of reactivity. The SPE has been implemented into the Monte Carlo method based homogenization code MCMC, as part of RMC Program, under developed in Tsinghua University. The C5G7 benchmark problem have been carried out to test the SPE. The results show that the SPE method not only suits for the equivalence in Monte Carlo based homogenization but also provides improved accuracy compared to the traditional GET or SPH method.

Scenario Analysis on the Benefits of Multi-National Cooperation for the Development of a Common Nuclear Energy System Based on PWR and LFR Fleets

Financial aspects, environmental concerns and non-favorable public opinion are strongly conditioning the deployment of new Nuclear Energy Systems across Europe. Nevertheless, new possibilities are emerging to render competitive electricity from Nuclear Power Plants (NPPs) owing to two factors: the first one, which is the fast growth of High Voltage lines interconnecting the European countries’ national electrical grids, this process being triggered by huge increase of the installed intermittent renewable electricity sources (Wind and PV); and the second one, determined by the carbon-free constraints imposed on the base load electricity generation. The countries that due to public opinion pressure can’t build new NPPs on their territory may find it profitable to produce base load nuclear electricity abroad, even at long distances, in order to comply with the European dispositions on the limitation of the CO2 emissions. In this study the benefits from operating at multinational level with the deployment of a fleet of PWRs and subsequently, at a proper time, the one of Lead Fast Reactors (LFRs) are analyzed. The analysis performed involves Italy (a country with a current moratorium on nuclear power on spite that its biggest utility operates NPPs abroad), and the countries from South East and Central East Europe potentially looking for introduction or expansion of their nuclear power programmes. According to the predicted evolution of their Gross Domestic Product (GDP) a forecast of the electricity consumption evolution for the present century is derived with the assumption that a certain fraction of it will be covered by nuclear electricity. In this context, evaluated are material balances for the front and the back end of nuclear fuel cycle associated with the installed nuclear capacity. A key element of the analysis is the particular type of LFR assumed in the scenario, characterized by having a fuel cycle where only fission products and the reprocessing losses are sent for disposition and natural or depleted uranium is added to fuel in each reprocessing cycle. Such LFR could be referred to as “adiabatic reactor”. Owing to introduction of such reactors a substantive reduction in uranium consumption and final disposal requirements can be achieved. Finally, the impacts of the LFR and the economy of scale in nuclear fuel cycle on the Levelized Cost of Electricity (LCOE) are being evaluated, for scaling up from a national to a multinational dimension, illustrating the benefits potentially achievable through cooperation among countries.

3D Flow Field Analysis for a Traditional II+ PWR Containment Under Normal Condition

Flow field analysis is a foundation to many thermal-dynamic phenomena in the nuclear containment. There are several ventilation systems under normal condition to assure the proper environment for staff and facilities, and the two main ventilation systems are the Reactor Pit Ventilation system (EVC) and the Containment Continuous Ventilation system (EVR). The fans of the ventilation systems recycle the air in the containment to various rooms, and cooling coils are cooled by the Nuclear Island Chilled Water system (DEG). In this paper, we simulated the 3D flow field in a Chinese traditional generation II+ PWR reactor containment under normal conditions with a commercial CFD software ANSYS FLUENT. According the actual geometry data of the reactor plant, we built the geometry model, including the EVC and EVR system, and they were checked by plant experts to assure authenticity. Proper maximum mesh sizes were set for different parts at the ICEM CFD, and the grid number was about ten millions. We used the fan model in the ANSYS FLUENT to simulate fans in the ventilations. The calculated values of flow rate in ventilation systems were in good agreement with the design values.

CFD Validation of Void Distribution in a Rod Bundle With Spacer

As CFD (Computational Fluid Dynamics) continues to grow in the nuclear industry the need for validation is essential to this growth. The focus of this paper is to highlight the validation of AREVA’s 2-Phase CFD models in a rod bundle with spacer. The experimental work is based on void distribution measurements using a 64×64 wire mesh sensor at the Paul Scherrer Institute adiabatic test loop fuel rod bundle (SUBFLOW). The 2-Phase models are based on the Eulerian multiphase framework where the interaction models are further developed using user routines in a commercial CFD tool to achieve stability and accuracy in a rod bundle configuration with a spacer. The 2-Phase CFD models coincide with the SUBFLOW experiments in showing the effect of void collecting in the center of the sub-channels slightly downstream of the spacer and further downstream the smaller bubbles migrate toward the rod surfaces whereas the larger bubbles stay in the sub-channels. Good agreement is achieved between CFD and void distribution experiments. The multi-phase CFD method was used by AREVA to improve the performance of the new products GAIA and ATRIUM™ 11.

Development of Glass Melter Technology for HLLW Vitrification in Japan

This paper describes the development of glass melter technology, primarily the liquid fed joule-heated ceramic melter process (LFCM) for the vitrificaton of high-level radioactive liquid waste (HLLW) since 1977 in Japan. In 2013 the active test at the vitrification facility (K-facility) in Rokkasho commercial reprocessing plant was successfully completed for the final acceptance test. During this period many activities on LFCM process development have been carried out in the engineering scale or the full-scale inactive cold tests including the radioactive laboratory scale hot tests. In particular, the design of melter bottom structure and the operating method should be optimized in order to avoid the operational problems caused by accumulation of noble metals (Ru, Rh, Pd), electro-conducive deposits on the melter bottom. Through the operation of inactive and active test facilities in Tokai, the design basis for the Tokai Vitrification Facility (TVF) has been provided. The hot operation of the TVF was started in 1995 to demonstrate the LFCM process including the performance of the melter off-gas clean-up system etc. The TVF has provided the basis of the process design and the operation method for the K-facility melter in Rokkasho. In case of commercial scale vitrification, the glass production rate of the melter should be several times larger than that of the TVF. The K-facility full-scale inactive mock-up melter (KMOC) has been planned to confirm the influence of scale-up factors and the difference between Tokai and Rokkasho wastes. Through the testing operation of the KMOC, which was initially started in 2000, it has been found that the stable formation of a cold cap on a molten glass surface is fundamentally important to avoid the excessive precipitation of noble metals and the yellow phase formation. The active test of the K-facility has been proceeding under the same conditions as the KMOC, and was successfully completed in May, 2013. The advanced glass melter development programs have also commenced from 2009 to ensure a more robust and noble metals are compatible with the LFCM system and also to provide a higher processing rate. The second K-facility full-scale inactive mock-up melter (K2MOC) has been installed in the vitrification technology development facility (X-14) at Rokkasho. Its testing operation has commenced from November, 2013.

A Double Porosity Model to Describe Both Permeability Change and Dissolution Processes

Cement is a practical material for constructing the geological disposal system of radioactive wastes. However, such materials alter groundwater up to 13 in pH around the repository, changing the permeability of natural barrier. So far, the authors have examined the relation of permeability change with dissolution process by flowing a high pH solution (NaOH, 0.1 mM) into a bed packed with amorphous silica particles. Here, the particle diameters were adjusted to a size fraction of 74 to 149 μm by sieving. Its specific surface area was estimated as 350 m2/g by the BET method using nitrogen gas. The experimental results showed that the permeability did not immediately change although the soluble silicic acid continuously flowed out of the packed bed. This study proposes a new mathematical model considering the diffusion and dissolution processes in the inner pore of the particle. This model assumed that each packed particle (74 to 149μm in diameter) consists of the sphere-shaped aggregation of smaller particles (20 nm in diameter). OH− ions diffuse into the pore between such small particles, and simultaneously consumed by the reaction with small particles. The radius of the each packed particle (sphere-shaped aggregation of small particles) was defined by the length from the center of the aggregation to the region where the small particles still remains. Since the outer small particles more easily dissolve than inner small particles because of diffusion process of OH− ions, each packed particle gradually shrinks. The fundamental equations consist of a simple diffusion equation of spherical coordinates of OH− ions considering the reaction term, which is linked by the equation to describe the size change of small particles with time. Here, this model also considered a change (time and space) of the diffusion oefficient caused by the change of the porosity between small particles. Besides, the change of over-all permeability of the packed bed was evaluated by Kozeny-Carman equation and the calculated radii of packed particles. The dissolution rate constant already reported was used. The calculated result was able to well describe the experimental result, though there was no fitting parameter in the comparison with the experiment results. While the flow paths of underground cannot be simply simulated by a packed bed, this approach suggested that the dynamic behavior of permeability in a natural barrier depends also on non-uniformity of dissolution processes in inner pores (secondary pores) of minerals.

Time-Dependent, Three Dimensional Nodal Transport Code Development Based on Unstructured Mesh

A space-time nodal transport code, DAISY, was developed to evaluate dynamic neutron behavior in innovative nuclear system. The steady transport process is based on an arbitrary triangles-z mesh nodal method which can treat complicated geometry configuration with enough precision and acceptable calculated quantity. This code employs the improved quasi-static method for neutron kinetics with a predictor-corrector scheme to improve computational efficiency. The direct method and the point approximation for neutron kinetics are also implemented into DAISY to evaluate the precision and efficiency of this predictor-corrector scheme. This code was verified by several transient benchmarks. It shows that the predictor-corrector scheme in DAISY can greatly reduce the computational time with enough precision.