A simplified technique for shakedown limit load determination of a large square plate with a small central hole under cyclic biaxial loading

A simplified technique for determining the shakedown limit load of a structure was previously developed and successfully applied to benchmark shakedown problems involving uniaxial states of stress ( [Abdalla et al., 2007a], [Abdalla et al., 2007b] and [Abdalla et al., 2007c]). In this paper, the simplified technique is further developed to handle cyclic biaxial loading resulting in multi-axial states of stress within the large square plate with a small central hole problem. Two material models are adopted namely: an elastic-linear strain hardening material model obeying Ziegler's linear kinematic hardening (KH) rule and an elastic-perfectly-plastic (EPP) material model. The simplified technique utilizes the finite element (FE) method and employs small displacement formulation to determine the shakedown limit load without performing lengthy time consuming full elastic-plastic cyclic loading FE simulations or conventional iterative elastic techniques. The simplified technique is utilized to generate the shakedown domain for the plate problem subjected to cyclic biaxial tension along its edges. The outcomes of the simplified technique showed very good correlation with the results of analytical solutions as well as full elastic-plastic cyclic loading FE simulations. Material hardening showed no effect on the shakedown domain of the plate in comparison to employing EPP-material.     

Characterization of minor actinide mixed oxide fuel

The global nuclear energy partnership (GNEP) was created in order for ‘fuel-cycle supplier’ nations to provide assured supplies of nuclear fuel to ‘fuel-cycle customer’ nations. The customer nations would utilize the fuel for electricity generation and subsequently return it to the supplier nation after it is spent. This spent fuel would then be reprocessed by the supplier nation in order to recycle the actinide constituents, mainly uranium and plutonium, in advanced nuclear power reactors, and thus reduce waste volumes [1] and [2]. The International Atomic Energy Agency would control the nuclear materials. One of the thrust areas for the GNEP program is the development of these actinide bearing fuels for transmutation in a fast reactor.     

A second order turbulence model based on a Reynolds stress approach for two-phase boiling flow and application to fuel assembly analysis

High-thermal performance PWR (pressurized water reactor) spacer grids require both low-pressure loss and high critical heat flux (CHF) properties. Numerical investigations on the effect of angles and position of mixing vanes and to understand in more details the main physical phenomena (wall boiling, entrainment of bubbles in the wakes, recondensation) are required.In the field of fuel assembly analysis or design by means of CFD codes, the overwhelming majority of the studies are carried out using two-equation Eddy Viscosity Models (EVM), especially the standard K-? model, while the use of Reynolds Stress Transport Models (RSTM) remains exceptional.The simulation of swirling flow generated by the mixing vanes plays an important role for the prediction of the CHF for the fuel assemblies. For this reason, according to [14] and [Mimouni et al., 2009b], rotation effects and RSTM model are more specifically addressed in the paper.Before comparing performance of EVM and RSTM models on fuel assembly geometry, we performed calculations with simpler geometries, the DEBORA case and the ASU-annular channel case. ASU-annular channel case has already been addressed in [14] and [Mimouni et al., 2009b].Then, a geometry closer to actual fuel assemblies is considered. It consists of a rectangular test section in which a 2 × 2 rod bundle equipped with a simple spacer grid with mixing vanes is inserted. The influence of the turbulence model on target variables linked to CHF limitation will be discussed. Moreover, the sensitivity to the mesh refinement will be particularly examined. The study of this case is a further step towards the modelling of the two-phase boiling flow in real-life grids and rod bundles.     

Measurement of L X-ray fluorescence cross-sections for elements with 45 ? Z ? 50 using synchrotron radiation at 8 keV

The L shell fluorescence cross-sections of the elements in range 45 ? Z ? 50 have been determined at 8 keV using Synchrotron radiation. The individual L X-ray photons, Ll, Lα, Lβ_I, Lβ_II, Lγ_I and Lγ_II produced in the target were measured with high resolution Si(Li) detector. The experimental set-up provided a low background by using linearly polarized monoenergetic photon beam, improving the signal-to-noise ratio. The experimental cross-sections obtained in this work were compared with available experimental data from Scofield [1] and [2] Krause [3] and [4] and Scofield and Puri et al. [5] and [6].These experimental values closely agree with the theoretical values calculated using Scofield and Krause data, except for the case of Lγ, where values measured of this work are slighter higher.     

Verification and validation considerations regarding the qualification of numerical schemes for LES for dilution problems

Since a long time, the Large Eddy Simulation (LES) concept is considered as a very promising candidate for advanced thermal hydraulic modeling in Nuclear Reactor Safety. This paper shows how LES is successfully applied in an industrial framework to free shear flows at high Reynolds numbers and to the associated transport of scalars (e.g. boron). Extensive verification and validation attempts towards this objective have already been performed for the Trio_U code ( [H?hne et al., 2006], [Bieder et al., 2007] and [Bieder and Graffard, 2007]). In the first part, this paper presents a short overview of what has been done for predicting the boron concentration at the core inlet under accident conditions. These calculations are then related to the demands of Best Practice Guidelines (BPG), which have been discussed by Mahaffy et al. (2007). It is shown, that high quality LES simulations for free sheer flows can be performed on tetrahedral meshes, what significantly simplifies the mesh generation procedure in topologically complex geometries. Guidelines specifically devoted to the LES framework are proposed to analyse the capability of numerical schemes to treat correctly the scalar transport.     

An experimental investigation on the quenching of a hot vertical heater by water injection at high flow rate

An experimental investigation on rewetting has been carried out by injecting water from the top of a hot vertical heater. Tests have been performed with varied range of experimental conditions (200-500 °C surface temperatures, constant water flow rates 5.77-30.98 g s⁻¹). Effect of several coolant injection systems on the hydrodynamics of rewetting has been studied. It is observed that for a particular range of flow rate and initial wall temperature (21.58 g s⁻¹, 300 °C) a circumferentially symmetric wet front is observed for the region closer to the coolant injection point even while using sub-cooled water. Rewetting velocity has been calculated from the temperature transients measured during the experiment and was found to vary within 1.0-20.0 cm s⁻¹. Two different rewetting models ( [Sahu et al., 2006] and [Sahu et al., 2008a]) have been used to compare the present experimental data and the comparison is found to be fairly good in both the cases. It has been observed that the flow rate varies linearly with effective Biot number (M) and varies inversely with magnitude of precursory cooling (N) in the present investigation.     

Experimental results on the coolability of a debris bed with multidimensional cooling effects

Within the reactor safety research, the removal of decay heat from a debris bed (formed from corium and residual water) is of great importance. In order to investigate experimentally the long term coolability of debris beds, the scaled test facility “DEBRIS” (Fig. 1) has been built at IKE. A large number of experiments had been carried out to investigate the coolability limits for different bed configurations ( [Rashid et al., 2008], [Groll et al., 2008] and [0055]). Analyses based on one-dimensional configurations underestimate the coolability in realistic multidimensional configurations, where lateral water access and water inflow via bottom regions are favoured. Following the experiments with top- and bottom-flooding flow conditions this paper presents experimental results of boiling and dryout tests at different system pressures based on top- and bottom-flooding via a down comer configuration.A down comer with an internal diameter of 10 mm has been installed at the centre of the debris bed. The debris bed is built up in a cylindrical crucible with an inner diameter of 125 mm. The bed of height 640 mm is composed of polydispersed particles with particle diameters 2, 3 and 6 mm. Since the long term coolability of such particle bed is limited by the availability of coolant inside the bed and not by heat transfer limitations from the particles to the coolant, the bottom inflow of water improves the coolability of the debris bed and an increase of the dryout heat flux can be observed. With increasing system pressure, the coolability limits are enhanced (increased dryout heat flux).     

Experimental determination of photofission cross-sections of Th using electron accelerator

Photofission cross-section of ²³²Th was measured using Bremsstrahlung radiation energy 7.4–9.2 MeV with energy step of 0.3 MeV by employing Lexan polycarbonate film as Solid State Nuclear Track Detectors (SSNTD). The photon intensity from the Microtron accelerator facility was estimated to be 10¹⁰ photons/s at a distance of 15 cm from the Bremsstrahlung converter using EGS-4 code (Nelson et al., 1985). Photofission cross-sections were evaluated using fission fragment angular distribution measurements. The present experimental results were compared with EMPIRE-2.19 (Herman et al., 2005) code prediction of RIPL-1 and RIPL-2 ( and ) and a new analytical formula (Gupta and Saxena, 2005) for fission barrier.     

Interface temperature between solid and liquid corium in severe accident situations: A comprehensive study of characteristic time delay needed for reaching liquidus temperature

T_liquidus was proposed as the interface temperature (Seiler and Froment, 2000), for various severe accident situations for thermalhydraulic steady state. This proposal was made on the basis of the analysis of solidification front stability in thermalhydraulic steady state for volumetrically heated corium pools and was extended to reactor transients with slow solidification rates that are controlled by the long-term decrease in residual power. The conclusions were corroborated by prototypic corium and variable solidification rates obtained by experimental approaches ( [Dauvois et al., 2000] and [Journeau et al., 2003]) for corium containing small amounts of silica or none at all. When the concentration in silica increases (approximately above 10 wt%), it was concluded from the experiments that a plane-front situation could not be obtained.The present work offers a theoretical approach to the maximum time delay that is necessary for mass transfer and full phase-segregation in volumetrically heated liquid pools bounded by a crust. It is concluded that full segregation is obtained for in-vessel situations within time delays that are shorter or of the same order of magnitude as the characteristic time for the corium pool to form and evolve to a quasi-steady-state situation. The characteristic time delay for mass transfer associated with simulant material experiments is also determined. Phase segregation can also be obtained for corium-concrete interaction, provided that the silica content is less than approximately 10 wt%. However in the latter case, more complex phenomena occur at the interface due to the interaction with sparging gas (such as porous medium formation) which requires a different model approach.     

Local damage to reinforced concrete structures caused by impact of aircraft engine missiles Part 1. Test program, method and results

Structural damage induced by an aircraft crashing into a reinforced concrete structure includes local damage caused by the deformable engines, and global damage caused by the entire aircraft. Local damage to the target may consist of spalling of concrete from its front face together with missile penetration into it, scabbing of concrete from its rear face, and perforation of missile through it. Until now, local damage to concrete structures has been mainly evaluated by rigid missile impact tests. Past research work regarding local damage caused by impact of deformable missiles has been limited. This paper presents the results of a series of impact tests of small-, intermediate-, and full-scale engine models into reinforced concrete panels. The purpose of the tests was to determine the local damage to a reinforced concrete structure caused by the impact of a deformable aircraft engine.     

Formation of helium induced nanostructure ‘fuzz’ on various tungsten grades

The response of a variety of W material grades to nanostructure ‘fuzz’ formation is explored. W targets are exposed to He or D₂-0.2He plasmas in PISCES-B at 900-1320 K to below sputter threshold He⁺ ions of energy 25-60 eV for up to 2.2 × 10⁴ s. SEM and XPS reveal nanoscopic reorganization of the W surface to a layer of ‘fuzz’ of porosity ∼90% as determined by a ‘fuzz’ removal/weight loss method. The variability of ‘fuzz’ growth is examined at 1120 K for 1 h durations: SR, SC and doped W grades - La₂O₃ (1% wt.), Re (5% and 10% wt.), and TiC (1.5% wt.) developed 2-3 μm thick ‘fuzz’ layers, while a VPS grade developed a layer 4 μm thick. An RC grade revealed additional ‘fuzz’ at deep (>100 μm) grain boundaries. However, heat treatment up to 1900 K produced reintegration of ‘fuzz’ with the bulk and He release at ∼1000 K and ∼1400-1800 K due to depopulation from vacancy complexes.     

Ideal mechanical properties of vanadium by a first-principles computational tensile test

We employ first-principles total energy method based on the density functional theory with the generalized gradient approximation to investigate the ideal tensile strengths of a bcc vanadium (V) single crystal systemically. The ideal tensile strengths are calculated to be 19.1, 32.8 and 31.0 GPa for bcc V in the [1 0 0], [1 1 0] and [1 1 1] directions, respectively. We show that the [0 0 1] direction is the weakest direction due to the occurrence of structure transition at the lower strain and the [1 1 0] direction is strongest because of the stronger interaction of atoms between the (1 1 0) planes in comparison with the (0 0 1) and (1 1 1) planes. We derive the Young’s modulus formula V single crystal in different tensile directions and give detailed analysis. According to the elastic constants of V single crystal, we have estimated some mechanical quantities of polycrystalline V, which are the bulk modulus of B, the shear modulus of G, Young’s modulus of E and the Poisson’s ratio of ν. The results might provide a useful reference for V as a candidate structural material in the fusion Tokamak.     

Experimental investigation on critical heat flux in horizontal channel for low pressure low flow (LPLF) condition

Studies reported in the past on critical heat flux (CHF) are mostly limited to vertical flow, large channel diameter, high pressure and high mass flux. Only few investigations are reported in the literature for horizontal flow CHF especially under low pressure and low flow conditions. Hence, predictive methods of CHF for horizontal flow are scarce. There is a need for understanding CHF in horizontal flow under low pressure and low flow conditions because they are commonly encountered in nuclear reactor fuel channels of pressurized heavy water reactor (PHWR) under loss of coolant accidental (LOCA) conditions. The present work investigates CHF of horizontal flow for low flow rates (mass flux of 100–400 kg/m² s) at nearly atmospheric pressure conditions. Parameters covered in this study are diameter (5.5 mm, 7.5 mm and 9.5 mm), length (0.45 m and 0.8 m) and a inlet temperature of 32 °C. The first occurrence of ‘red hot’ spot on the test section is considered as the onset of critical heat flux condition in the present work. Experimental results obtained are compared with Groeneveld et al. (2007) look up table data for vertical flow after applying correction factor given by Wong et al. (1990). The deviation of experimental CHF data from those predicted using Groeneveld et al. (2007) look up table and Wong et al. (1990) correction factor is more than 50%.     

Simulation of hydrogen mitigation in catalytic recombiner. Part-II: Formulation of a CFD model

This paper aims at formulation of a model compatible with CFD code to simulate hydrogen distribution and mitigation using a Passive Catalytic Recombiner in the Nuclear power plant containments. The catalytic recombiner is much smaller in size compared to the containment compartments. In order to fully resolve the recombination processes during the containment simulations, it requires the geometric details of the recombiner to be modelled and a very fine mesh size inside the recombiner channels. This component when integrated with containment mixing calculations would result in a large number of mesh elements which may take large computational times to solve the problem. This paper describes a method to resolve this simulation difficulty. In this exercise, the catalytic recombiner alone was first modelled in detail using the best suited option to describe the reaction rate ( [Prabhudharwadkar et al., 2005] and [Prabhudharwadkar et al., 2011]). A detailed parametric study was conducted, from which correlations for the heat of reaction (hence the rate of reaction) and the heat transfer coefficient were obtained. These correlations were then used to model the recombiner channels as single computational cells providing necessary volumetric sources/sinks to the energy and species transport equations. This avoids full resolution of these channels, thereby allowing larger mesh size in the recombiners. The above mentioned method was successfully validated using both steady state and transient test problems and the results indicate very satisfactory modelling of the component.     

Conceptual design of the FAST load assembly

Fusion advanced studies torus (FAST) is a proposal for a satellite facility which can contribute the rapid exploitation of ITER and prepare ITER and DEMO regimes of operation, as well as exploiting innovative DEMO technology. FAST is a compact (R₀ = 1.82 m, a = 0.64 m, triangularity δ = 0.4) machine able to investigate non-linear dynamics effects of alpha particle behaviours in burning plasmas [1], [2] and [5]. The project is based on a dominant 30 MW of ion cyclotron resonance heating (ICRH), 6 MW of lower hybrid (LH) and 4 MW of electron cyclotron resonance heating (ECRH). FAST operates at a wide range [3] and [4] of parameters, e.g., in high performance H-mode (B_T up to 8.5 T; I_P up to 8 MA) as well as in advanced Tokamak operation (I_P = 3 MA), and full non-inductive current scenario (I_P = 2 MA). Helium gas at 30 K is used for cooling the resistive copper magnets [6]. That allows for a pulse duration up to 170 s. To limit the TF magnet ripple ferromagnetic insert have been introduced inside the vacuum vessel (VV). Ports have been designed to also accommodate up to 10 MW of negative neutral beam injection (NNBI). Tungsten (W) or liquid lithium (L-Li) have been chosen as the divertor plates material, and argon or neon as the injected impurities to mitigate the thermal loads.     

Attenuation of ambient dose equivalent from neutrons by thick concrete, cast iron and composite shields for high energy proton, He, Ca and U ions on Cu targets for shielding design

Data on neutron dose attenuation by thick concrete, cast iron, and cast iron plus concrete composite shields for heavy ions and protons having high energies (200-1000 MeV/u) are necessary for shielding designs of high-powered heavy ion accelerator facilities. Neutron production source terms, shield material attenuation lengths, and neutron dose rate reduction effectiveness of the bulk shielding in the angular range from 0° to 125° were determined by the Particle and Heavy Ion Transport Code (PHITS) for beams of 300 and 550 MeV/u ⁴⁸Ca ions, 200 and 400 MeV/u ²³⁸U ions, 800 MeV/u ³He and 1 GeV protons. Calculated results of interaction lengths of concrete and cast iron were also compared with similar work performed by Agosteo et al., and to experimental and other calculated data on interaction lengths. The agreement can be regarded as acceptable.     

Development and overall testing of the traditional MTC estimation based on noise diagnostics at all four units of the Paks NPP

Moderator Temperature Coefficient (MTC) of reactivity is an important parameter of nuclear reactors and its value should be in a specific range during the reactor operation. Determining MTC with traditional methods is performed through deviating the reactor state from the normal operation. Noise diagnostics has the potential to provide an MTC estimation method during normal operation without changing the reactor state.This paper overviews the developments made on the traditional noise based estimators and the steps of their application in VVER-440 type reactors. Modifications were made to consider the effect of propagating perturbations through geometry and coolant velocity and to compensate the frequency response of the slow thermocouples. Frequency dependence of H₁ was calculated in high resolution which involved the revision of the method applied for reading the MTC value earlier (see the details in [Kiss et al., 2010a] and [Kiss et al., 2010b]).Investigations were performed at all four units of the Paks NPP to verify the dependence of the results on specific core features (e.g. on core loading patterns) and instrumentation of the individual units. For a better comparison of the results, a common frequency range of the evaluation was chosen for all units, thus this range was made narrower and its weight went closer to the frequency range affected by power feedback. Absolute values of the estimated MTC are smaller than the reference value due to the effect of feedback and low coherence. During the investigated fuel cycles estimations were quite close to the reference for two units (the deviation was around 10%), while the agreement was only moderate for the other two units (here the deviation was around 40%).     

Synthesis and characterization of psyllium-NVP based drug delivery system through radiation crosslinking polymerization

In order to develop the hydrogels meant for the drug delivery, we have prepared psyllium-N-vinylpyrrolidone (NVP) based hydrogels by radiation induced crosslinking. Polymers were characterized with SEMs, FTIR and swelling studies. Swelling of the hydrogels was studied as a function of monomer concentration, total radiation dose, temperature, pH and [NaCl] of the swelling medium. The swelling kinetics of the hydrogels and release dynamics of anticancer model drug (5-fluorouracil) from the hydrogels have been carried out for the evaluation of swelling and drug release mechanism. It has been observed that diffusion exponent ‘n’ have 0.8, 0.9, 0.8 and gel characteristics constant ‘k’ have 9.22 × 10⁻³, 2.06 × 10⁻³, 11.72 × 10⁻³ values for the release of drug from the drug loaded hydrogels in distilled water, pH 2.2 buffer and pH 7.4 buffer, respectively. The present study shows that the release of drug from the hydrogels occurred through Non-Fickian diffusion mechanism.