Liquid metal fast reactor using sodium as a coolant typically utilize a tightly packed triangular lattice of fuel pins enclosed in a hexagonal duct. During the reactor operation, partial or total flow blockage of coolant channels, may occur at different spatial locations within the fuel assembly, due to potential isolated or combined causes, including collection and accumulation of debris, and cladding deformation. Previous studies have shown that the flow characteristics within the wire-wrapped fuel assembly, i.e., including the interior and exterior subchannels, are very complicated and strongly influenced the flow and heat transfer phenomena between the coolant fluid and fuel rods. It is important to understand and characterize the effects of channel blockage to the flow mixing characteristics in the exterior sub-channels (or bypass channels) of the wire-wrapped fuel bundle. Texas A&M University has conducted isothermal flow experiments in a wire-wrapped 61-pin hexagonal fuel bundle to support the research on advanced nuclear fuel development sponsored by the US Department of Energy (DOE). The experimental facility employs matched-index-of-refraction (MIR) techniques and laser diagnostic velocity measurement techniques. In this article, we present the time-resolved particle image velocime-try (TR-PIV) measurements to characterize turbulent flow characteristics in the exterior subchannels of the wire-wrapped fuel bundle, under the presence of a localized total blockage of one of the exterior subchannels. From the obtained TR-PIV velocity fields, turbulent flow characteristics including mean velocity, root-mean-square fluctuating velocity, and Reynold stress, are computed and presented. In addition, spectral analysis to the turbulent velocity fields is performed to investigate the characteristic flow frequencies associated to different flow conditions with the presence of the blockage. Finally, proper orthogonal decomposition (POD) analysis is performed to the velocity snapshots to reveal the dominant flow structures that play important roles in the flow dynamics and heat transfers of the fuel bundle.

An experimental campaign was performed on a non-uniformly heated 19-pins wire-spaced fuel pin bundle simulator, cooled by Heavy Liquid Metal and installed in the NACIE-UP (NAtural CIrculation Experiment-UPgrade) facility located at the ENEA Brasimone Research Center (Italy). The experimental tests concerned mass flow rate transition of the primary coolant from forced to natural circulation, with fuel pin bundle simulator characterized by non-uniform power distribution. The main objective of the performed experimental campaign was to perform integral system and local thermal-hydraulic analysis, in particular to investigate the flow in different flow regimes and specifically the transition from forced to natural circulation flow and, more specifically, analyze the behavior of the 19-pins wire-spaced fuel pin simulator (FPS) during such transient. Indeed, the performed test were characterized by non-uniform heating of the bundle (i.e. just some pins switched on), so the effects of this non-uniformity on the local temperatures and on the overall system behavior was evaluated. A deep investigation on the local temperature distribution was performed thanks to the accurate instrumentation provided in the bundle (67 thermocouples). For instance, in some cases, the wall temperatures relative to pins switched off remained below the relative sub-channel temperature, depending on the heating distribution. The obtained experimental data provided useful information for the characterization of the bundle and the computation of the heat transfer coefficient. Moreover, the collected system data can be helpful for STH codes validation, whereas the local fuel bundle data, especially the ones from dissymmetric tests can be useful for the qualification and benchmarking of CFD codes and coupled STH/CFD methods for HLM systems.

Nuclear fuel cladding is an integral part of nuclear reactors and choosing the proper material is imperative to the design of a reactor. In this paper, the neutronic properties and the fuel performance of a 17 × 17 Westinghouse Pressurized Water Reactor (PWR) assembly using ceramic matrix composites (CMC) Silicone Carbide (SiC) as a cladding material is investigated. The material analysis is compared against traditional Zircaloy-4 cladding used in a 17 × 17 Westinghouse PWR assembly. The codes used in the analysis are the Michigan Parallel Characteristics based Transport (MPACT) code coupled with CTF, the North Carolina State University version of the Coolant Boiling in Rod Arrays Two Fluids (COBRA-TF) code, and the fuel performance code BISON as well as the uncertainty analysis code DAKOTA. Additionally, annular geometry for the fuel pellet is modeled to assess its merit compared to ordinary CMC SiC or traditional Zircaloy-4 claddings. It is found that on the neutronics side, the CMC SiC shows lower achievable U-235 enrichments required to reach the same burnup and effective neutron multiplication factor as Zircalloy-4 claddings. These results are an advantage that can be seen in the economic cost analysis done and additionally from the reactor operation point of view. Also, it is found that the different criteria of safe operation of Westinghouse PWR assemblies like the plenum pressure, the fuel-cladding contact pressure, the peak fuel temperature, and the fission gas release criteria are all achieved with CMC SiC with some criteria having larger design margins than of the Zircaloy-4 cladding. Furthermore, a critical heat flux (CHF) study shows that CMC SiC has even larger thermal margins than the ordinary Zircaloy-4 cladding, resulting in a more profitable fuel cycle due to the greater amount of power that the fuel pins can be operated at. An uncertainty quantification for the CHF Ratio (CHFR) is also done to assess the largest magnitudes of importance that affect the CHFR calculated.

In the context of the studies on GEN. IV/ADS nuclear systems, the correct evaluations of the temperature distribution in the fuel pin bundle is of central interest. In particular, the use of lead or lead-bismuth eutectic (LBE) as coolant for the new generation fast reactors is one of the most promising choices. Due to the high density and high conductivity of lead or LBE, a detailed analysis of the thermo-fluid dynamic behavior of the heavy liquid metal (HLM) inside the sub-channels of a fuel rod bundle is necessary in order to support the Front-End Engineering Design (FEED) of GEN. IV/ADS prototypes and demonstrators. In this frame, the synergy between numerical analysis by CFD and data coming from large experimental facilities seems to be crucial to assess the feasibility of the components. At ENEA-Brasimone R.C., large experimental facilities exist to study HLM free, forced and mixed convection in loops and pools: e.g. NACIE-UP is a large scale LBE loop for mixed convection experiments. The MYRRHA-19 like Fuel Pin Bundle Simulator installed in the NACIE-UP facility allows to make non-uniform and dissymmetric tests with only a few pins heated. This technical feature of the FPS is very interesting for CFD validation and this kind of data tests in HLM fuel bundles are not so common in the literature. In the present paper, a post-test validation is made by a detailed CFD model of the test section. Experimental data, statistically treated by the error propagation theory, are briefly presented and a preliminary comparison with CFD results using different models/turbulent Prandtl numbers are shown. Three monitored section at different levels are compared both for wall and bulk temperatures. This post-test comparison with this experimental configuration is unique and represents a further step towards the validation of the CFD models and methods in fuel bundle geometries cooled by HLM.

The concept of multi-dimensional heterogeneous resonance integral tables is proposed. The new type of resonance integral is designed for different fuel pins appearing in one lattice with two extra dimension of optical radius and number density ratio in the fuel. Numerical results show that this treatment improves the accuracy of embedded self-shielding method on irregular lattices.

The efficient solution of the neutron diffusion equation for large scale whole core calculations is of paramount importance; especially if the detailed pin-level power distribution and reaction rates are required. For heterogeneous whole core calculations finite element based techniques have been one approach to modelling the detailed pin geometry in whole core calculations. A new approach, pioneered in the last few years, is isogeometric analysis (IGA) methods which enable the exact geometry of fuel pins to be modelled. In order to efficiently solve elliptic partial differential equations (PDEs), such as the neutron diffusion equation, typically multi-grid or multi-level iterative solution techniques are used such as the algebraic multi-grid method. However, using IGA methods it is possible to develop true geometric multi-grid techniques which are potentially much more efficient than the standard algebraic multi-grid methods. In this paper we explore the use of IGA methods to develop a scalable, multi-level, iterative algorithm which is then used to solve the neutron diffusion equation over several geometries. This multilevel solution algorithm utilises a single patch multi-grid framework suggested by Hofreither and Takacs that takes advantage of the tensor product construction to provide scalability with respect to spatial, and polynomial refinement. Furthermore, a two-level balancing Neumann-Neumann solver is used to extend the solver to multiple patches in a scalable way. It is seen that the number of iterations required depends on the mapping between the unit square and the physical geometry, as well as the material coefficients.

This study recaps the experimental effort to characterize a transverse plane flow field in a tightly-packed, 61-pin, wire-wrapped, hexagonal fuel bundle prototypical for a sodium fast reactor. The motive was to produce high spatiotemporal experimental data for computational fluid dynamics turbulence model validation. The matched-index-of-refraction and laser-based optical measurement techniques were utilized on an isothermal experimental flow facility. Measurements were performed on a transverse plane perpendicular to the axial flow. Fluid flow in the three types of subchannels (corner, exterior, and interior) were quantified. All measurements have been performed at a bundle-averaged Reynolds number of approximately 10,400. Results include flow statistics such as ensemble-averaged velocity, root-mean-square fluctuating velocity, and Reynolds stress. Of interest was the flow behavior around the restriction caused by the wire spacer and hexagonal duct wall, where recirculation regions formed. Regions of maximum and minimum momentum transfer coincided with regions of maximum and minimum fluctuations. These regions highlight locations of maximum and minimum cooling of the fuel pins. The experimental data will be used to benchmark computational fluid dynamics simulations of the sodium fast reactor fuel bundle using Reynolds-averaged Navier Stokes and large-eddy simulation turbulence modeling methods.

In sodium-cooled fast reactors, the core is not arranged in its most reactive configuration. In this case, when the fuel melts to form a molten pool, the recriticality may occur by positive reactivity insertion due to core compaction. To prevent such recriticality, special devices of the fuel subassembly structure for discharging the molten fuel from the core region, have been investigated by the Japan Atomic Energy Agency (JAEA). On the other hand, the inherent feature of core geometry and the neutron characteristics may provide the similar effect to prevent such recriticality. The purpose of this study is to design the core specification its deformation in CDA causes negative feedback to subcritical condition, without any fuel discharge device. The convex shaped core has the longer fuel length in the inner-core region and the shorter fuel in the outer-core region. Therefore, the core geometry as intact status has a lower neutron leakage effect. When the fuel melts in CDA, the core height is compacted and negative reactivity insertion is expected during molten pool formation. The convex shaped core is based on the large-scale cylindrical homogeneous core (3,600 MWth, 4.95m in core diameter, and 0.75m in core height). The calculation showed that the compaction of cylindrical core leads to a reactivity gain, whereas the convex shaped core results in negative reactivity effect. In this geometry, both inner-core and outer-core are divided into two regions. Furthermore, we introduced the smaller diameter pin for inner-core and keep uniform Pu enrichment for all regions. The smaller diameter pins in high importance region are effective for flat-distribution. Through pin diameter survey, we confirmed the advantages of smaller diameter pin, such as reducing pressure loss of core coolant and decreasing the height of molten pool.

Experimental investigations of thermal hydraulic properties of coolant flow in experimental models of fuel assemblies (FA) were performed. FA models with 7, 19 and 37 pins in hexagonal arrangement were used as test sections. Friction loss coefficients of smooth rod bundles are presented and compared. Hydraulic resistance of spacing grids with different geometry is discussed. Comparison of obtained data with correlations of Rehme and Idelchik is carried out. Averaged heat transfer dependences on distance from spacer grid top end are presented and compared with different dependences. Experimental results for coolant velocity distribution in subchannels of experimental models were obtained using both Pitot pipe technique and laser Doppler annemometry (LDA). It was shown that a strong disturbance of the flow just behind the spacer grid is present. The flow stabilization occurs at relatively small distances from the spacer elements.

Boron Neutron Capture Therapy (BNCT) is a kind of the targeted therapy with two element. It can kill the cancer cells while the effect on normal cells is very small, and it is suitable for the treatment of the various stage cancer so it will be the ideal radiotherapy for cancer treatment in the future. And Commercial Miniature Neutron Source Reactor (C-MNSR) was designed and constructed by CIAE, which is used for Neutron Activation Analysis (NAA), Training and teaching. The reactor with thermal power 27kW is an under-moderated reactor with pool-tank type, U-AL alloy with High Enriched Uranium (HEU) as fuel, light water as coolant and moderator, and metal beryllium as reflector. The fission heat produced by the reactor is removed by the natural circulation. Design C-MNSR with a epi-thermal neutron beam for BNCT is studied while the conversion from HEU to LEU (Low Enrichment Uranium) ( 235 U percent≤20%) is carried on. As it has the advantages of MNSR safety, economy, easy operation and its application, and it can improve the epi-thermal neutron flux density and meet the requirements of BNCT. The fuel cage of C-MNSR with size of φ230×248mm in the reactor core, there are ton rows of 355lattices are concentrically arranged, the central lattice is reserved for central control rod, and four tie rods are uniformly arranged at the eighth row which link the upper and lower grid plates, the rest 350 fuel lattices are for fuel pins or dummies. The diameter of the fuel meat is 4.3mm, the height is 230mm, with Uranium enrichment is 17%; the diameter of the fuel element is 5.5mm, the height is 248mm. The frame design of the epithermal neutron beam is: Fluental material used as neutron moderation layer with its thickness is 50cm and its density is 2.85g/cm3; Cd with thickness of 0.1cm used as thermal neutron absorption layer, Lead with thickness of 10cm used as gamma ray shielding layer. And the neutron collimator parts is a composition of graphite, Cd and polythene with boron. The total length of the beam is 114.5cm, and the distance from the exit of the beam to the core is 130cm. The results show that the epithermal neutron flux density at the exit is 1.58 × 10 9 n·cm-2·s-1 at full power of 27kW. and the fast neutron density at the exit is 5.45 × 10 7 n · cm-2 · s-1 at full power. Fast neutron dose contamination (Df/ φepi) is 2.88 × 10 −11 Gy · cm 2 · n −1 and gamma dose contamination (Dγ/φepi) 2.18× 10 −14 Gy·cm 2 ·n −1 .

The guide tube (GT) representation in the reactor-internal system model has been crude and inaccurate in the past because the lack of experimental data on a) the rotary stiffness of the hold down bolt joint and b) the rotary stiffness of the support pins. In addition, the slotted enclosure and the complex interior in the continuous section further complicated the modeling accuracy. In this paper, the static and dynamic characteristic of the domestically manufactured CAP1000 guide tubes are studied both experimentally and analytically. Using the stiffness data obtained from the tests and combined with the enclosure stiffness calculated from the detailed 3D finite element models, i.e., one for the intermediate section and one for the continuous section, a simplified model of the GT was established. Using the simplified model, the dynamic characteristics (natural frequency and modal shapes) were compared to the dynamic test results in the aerial condition. It can be concluded that the simplified guide tube model is believed to be the most accurate one, by far, to be used in reactor-internal system analyses.

The hydrogen emission of zirconium hydride at high temperature is a challenging issue for many researchers. The hydrogen emission content of zirconium hydride pins should be evaluated to confirm the application feasibility. The comparison of theory analysis and experiment data indicated Richardson’s Law could offer a conservative result for calculating the hydrogen emission content of zirconium hydride pins at high temperature. Furthermore, the methods of preventing hydrogen loss should be developed for the purpose of extending the work temperature or time. The results showed a ZrO 2 layer prepared for zirconium hydride could not prevent hydrogen loss after exposure at 923K in an inert environment and ZrO 2 transformed into Zr 3 O gradually due to the opposite movement of hydrogen and oxygen. Finally, a further improvement to prevent hydrogen loss was developed. The zirconium hydride with a ZrO 2 layer in the cladding of He+CO 2 exhibited no significant reduction of hydrogen content. It is helpful to prevent the hydrogen loss by increasing the oxygen potential on the outside of ZrO 2 layer.

In the design study of an advanced loop-type sodium-cooled fast reactor in Japan Atomic Energy Agency, a specific fuel assembly (FA) named FAIDUS (Fuel Assembly with Inner DUct Structure) has been adopted as one of the measures to enhance safety of the reactor during the core disruptive accident. Thermal-hydraulics evaluations in FAIDUS under various operation conditions are required to confirm its design feasibility. In this study, thermal-hydraulics in FAIDUS are investigated by using a subchannel analysis code ASFRE, which is applicable to a wire-wrapped fuel pin bundle with a distributed resistance model and a simplified turbulence mixing model. At first, the distributed resistance model was validated by comparison of pressure drop coefficients with experimental data obtained in water experiments with simulated FAs under the condition of wide-range Reynolds number. And then, the turbulence mixing model was validated by comparison of temperature distribution in the pin bundle with experimental data obtained in sodium experiments with simulated FAs. After the applicability of ASFRE to FAs was confirmed through these validations, thermal-hydraulic analyses of a FA with 271 fuel pins without the inner duct and a FAIDUS with 255 fuel pins were conducted. The obtained results indicate that no significant asymmetric temperature distribution occurs in a FAIDUS as a FA without an inner duct. In addition, the temperature distribution of FAIDUS with 255 fuel pins under the low flow rate condition tended to be the same as that of a FA with 271 fuel pins due to the local flow acceleration and the flow redistribution caused by the buoyancy force.

The amount of spent fuel and high-level waste already available, and which will be produced by the future NPPs operation, calls for the evaluation of any possible technological solution that could minimize the burden of their disposal: reduction of Minor Actinide (MA) content, in addition to the radiotoxicity and radioactivity, and of the generated thermal load (decay heat). In this context, R&D efforts currently focus on the development of methodologies and technical solutions for Partitioning and Transmutation. MAs and long-lived fission products are in fact the main contributors to the long-term radiotoxicity of spent nuclear fuel, and their transmutation to short-lived fission products, in fast spectrum nuclear reactors, in transmuters or in Accelerator Driven Systems (ADS), by neutron irradiation of dedicated fuels/targets, is a promising and widely investigated option. In order to provide substantial input for the safety assessment of innovative nuclear fuels dedicated to MA transmutation, several irradiation tests are being carried out. In some options, the investigated fuels/targets are uranium free, or of low uranium content, to improve the transmutation performance and contain high concentrations of MA and plutonium compounds. Two molybdenum based CER-MET fuels, called ITU-5 & ITU-6, were prepared at JRC Karlsruhe for the irradiation experiment FUTURIX-FTA (FUel for Transmutation of transURanium elements in phenIX/Fortes Teneurs en Actinide). The experiment performed from 2007 to 2009 in the Phénix reactor, France, in cooperation with CEA. The experiment ended after 235 equivalent full power days (EFPD) at a Linear Heat Rate of circa 130 W/cm and reached burn-ups of 18 %FIHMA and 13 %FIHMA, respectively. Afterwards, the pins were transported to the Hot Cells of JRC Karlsruhe for Post Irradiation Examination. After a short summary describing the fuel preparation and irradiation conditions of the FUTURIX FTA irradiation experiment, the paper will give an overview on the current status and further planning of the Post Irradiation Examinations of ITU-5 & ITU-6 at JRC Karlsruhe. Finally, the results of the characterisations will be discussed and conclusions on the irradiation performance will be drawn. The results of this experiment will help to increase the knowledge and understanding of the irradiation behaviour of metal based transmutation targets and the qualification and validation of models developed to predict fuel safety performance.

This paper presents an assessment results for the developed RANS (Reynolds Averaged Navier-Stokes simulation) based CFD (Computational Fluid Dynamics) methodology applicable to real scale 217-pin wire wrapped fuel assembly of the KAERI (Korea Atomic Energy Research Institute) PGSFR (Prototype Gen-IV Sodium-cooled Fast Reactor). Complicated and vortical flow phenomena in the wire-wrapped fuel bundles were captured by a shear stress transport (SST) turbulence model, and by a vortex structure identification technique based on the critical point theory. The CFD results show good agreement with the JAEA experiment with the 127-pin wire-wrapped fuel assembly. The JAEA experiment study was implemented using water for validating pressure drop formulas in ASFRE code. The edge vortex structures are longitudinally developed, and have a higher axial velocity than corner vortex structures and wakes nearby pins and wires. The wire spacers locally induce a tangential flow by up to about 16 % of the axial velocity. The tangential flow in the corner and edge sub-channels is much stronger than that in the interior subchannels. The large-scale edge vortex structures have higher turbulence intensity and lower vorticity than the small-scale wakes. The corner vortex structures have lower turbulence intensity and vorticity than the small-scale wakes. The driving forces in the X-, Y-, and Z-directions are not only dependent on the axial velocity, but also significantly dependent on the angular position between the wire-spacer and rod, and the relative position between the wire-spacer and duct wall.

In the context of GEN-IV heavy liquid metal-cooled reactors safety studies, the flow blockage in a fuel sub-assembly is considered one of the main issues to be addressed and one of the most important and realistic accident for Lead Fast Reactors (LFR) fuel assembly. The blockage in a fast reactor Fuel Assembly (FA) may have serious effects on the safety of the plant leading to the FA damaging or melting. The temperature of the coolant leaving the FA is considered an important indicator of the health of the FA (i.e. the effective heat removal) and is usually monitored via a dedicated, safety-related system (e.g. thermocouple). The external or internal blockage of the FA may impair the correct cooling of the fuel pins, be the root cause of anomalous heating of the cladding and of the wrapper and potentially impact also fuel pins not directly located above or around the blocked area. In order to model the temperature and velocity field inside a wrapped FA under unblocked and blocked conditions, detailed experimental campaign as well as 3D thermal hydraulic analyses of the FA is required. The present paper is focused on the CFD pre-test analysis and design of the new experimental facility ‘Blocked’ Fuel Pin bundle Simulator (BFPS) that will be installed into the NACIE-UP (NAtural CIrculation Experiment-UPgrade) facility located at the ENEA Brasimone Research Center (Italy). The BFPS test section will be installed into the NACIE-UP loop facility aiming to carry out suitable experiments to fully investigate different flow blockage regimes in a 19 fuel pin bundle providing experimental data in support of the development of the ALFRED (Advanced Lead-cooled Fast Reactor European Demonstrator) LFR DEMO. In particular, the ‘Blocked’ Fuel Pin bundle Simulator (BFPS) cooled by lead bismuth eutectic (LBE), was conceived with a thermal power of about 250 kW and a uniform wall heat flux up to 0.7 MW/m 2 , relevant values for a LFR. It consists of 19 electrical pins placed on a hexagonal lattice with a pitch to diameter ratio of 1.4 and a diameter of 10 mm. The geometrical domain of the fuel pin bundle simulator was designed to reproduce the geometrical features of ALFRED, e.g. the external wrapper in the active region and the spacer grids. Pre-tests calculations were carried out by applying accurate boundary conditions; the conjugate heat transfer in the clad is also considered. The numerical simulation test matrix covered the envisioned experimental range in terms of mass flow rate; the wall heat flux was imposed in order to have a fixed temperature difference across the BFPS in unblocked conditions. The blockages investigated are internal blockages of different extensions and in different locations (central subchannel blockage, corner sub-channel blockage, edge subchannel blockage, one sector blockage, two sector blockage). High resolution RANS simulations were carried out adopting the ANSYS CFX V15 commercial code with the laminar sublayer resolved by the mesh resolution. The loci of the peak temperatures and their width as predicted by the CFD simulations are used for determining the location of the pin bundle instrumentation. The CFD pre-test analysis allowed also investigating the temperature distribution in the clad to operate the test section safely.

Related to the high conversion type BWR, the experimental study on the CHF were performed with a forced convection type water flow loop. In the test section of the loop, three heater pins as a rod bundle are installed, and water flows around the heater pins in the triangular flow channel at atomospheric pressure. The ratio of distance between two heater pins and the diameter of the heater pin ( i.e. , P/D) was the constant as 1.18 in all the experiments. This experiment conducted at the constant mass flux of 435 kg/m 2 s. The wire spacers, which were made of stainless steel wire coated by Teflon tube, were wound on the heater pins. The pitches of the wire spacer wound on the surface of the heater pins were 50 mm and 100mm. The water temperature at the inlet of the test section was controlled in the range between 60 and 90 °C by the pre-heater. The parameters in the current work were the pitch of the wire spacer and the quality at the location of the burnout (i.e., local quality). The results indicated that the CHF obtained in the test section with the wire spacer was larger than that without the wire spacer. The wire spacer must promote the CHF value. The smaller pitch of the wire spacer resulted to the smaller CHF value.

The issue of the temperature measurement in a nuclear reactor is an important element to ensure safe operation of the nuclear power plant. To prevent damages and radioactive releases the fuel in the reactor must be continuously cooled. The coolant temperature field of the VVER-440 reactor is measured with thermocouples installed at the outlet part of fuel assemblies. Since the power output of the fuel pins is not equal, a non-uniform temperature field at the inlet of the fuel assembly head is formed. Next, the temperature field is subsequently mixed by passing through the assembly head, which contains some constructional elements helping to mixing coolant flow. This mixing is not perfect and due to the effect described above, the signal on the thermocouple can be affected. This phenomenon was introduced in 2009 Atomic Energy Research (AER) Symposium in Bulgaria. 7 international institutions participated with the main goal to explain the mixing character of the coolant and to compare results. For further study of this phenomenon the new detailed computational model of the upper part of the fuel assembly was created and subsequently on the ANSYS FLUENT CFD code verified. The main output of these simulations is study of the coolant temperature distribution on the thermocouple. Computational model, based on the source geometry given by AER symposium, was created in preprocessor GAMBIT 2.4.6. Model contains over 13 million hexahedral cells. Thermohydraulic simulations where performed in ANSYS FLUENT v14.5 and results were compared with data from AER Symposium. There were considered two cases with different pins power outputs. With compare to AER symposium results the achieved resultant temperature on the thermocouple position for both cases indicate comparable accuracy. Furthermore, some flow fluctuations in the assembly head area where found.

This paper presents recent advances in the modeling of two-phase boiling flow and critical heat flux that have been implemented in the Extended Boiling Framework (EBF) [1, 2, 3]. The EBF code was developed as a customized module built on the foundation of the commercial Computational Fluid Dynamics (CFD) code STAR-CD, which provides general two-phase flow modeling capabilities, for the detailed analysis of the two-phase flow and heat transfer phenomena that occur in Boiling Water Reactor (BWR) fuel assemblies. These phenomena include coolant phase changes and multiple flow regimes that directly influence the coolant interaction with the fuel pins and, ultimately, the reactor performance. An effort to expand the EBF two-phase models and to explore their applicability to other CFD codes is currently underway. The paper presents results of recent CFD analyses of Critical Heat Flux (CHF) experiments that have measured the axial distribution of wall temperature in two-phase upward flow in a vertical channel with a heated wall. The experiments were designed to produce the onset of CHF in the upper half of the heated channel. The simulated axial distribution of wall temperature is compared with experimental data, illustrating the ability of the extended EBF model to capture the onset of CHF for a wide range of thermal-hydraulic conditions relevant for BWRs. The paper concludes with a discussion of results and plans for future work.

In control-oriented simulators, the neutronics is usually modeled by implementing the point-wise kinetics. In the framework of the study of the control strategy for innovative reactor concepts, such a simplified description is less effective since prevents the possibility of exploiting the capabilities of advanced control schemes. In order to overcome these limitations, a spatial neutronics model based on Modal Method has been considered. This approach allows separating the spatial and time dependence of the neutron flux, which can be represented as sum of the eigenfunctions of the multi-group neutron diffusion equation weighted by time-dependent coefficients. In this way, the system dynamic behavior is reduced to the study of these time-dependent coefficients and can be represented by a set of Ordinary Differential Equations (ODEs), reducing the computational burden. In this paper, a test case involving three fuel pins of an innovative Lead-cooled Fast Reactor has been set up and investigated. Once obtained the eigenfunctions, the set of ODEs for studying the time-dependent coefficients has been implemented in the MATLAB environment [1]. Finally, in order to assess the performance of the developed model, the outcomes have been compared with the results obtained from the neutron diffusion partial differential equation, achieving a satisfactory agreement.