The Pebble Bed Modular Reactor is a particular design of pebble bed reactor under development by South African company PBMR Ltd since 1994. The project entails the construction of a demonstration power plant at Koeberg near Cape Town and a fuel plant at Pelindaba near Pretoria. Wikipedia.
Roberts D.E.,South African Council for Scientific and Industrial Research |
Du Plessis A.,South African Council for Scientific and Industrial Research |
Du Plessis A.,Stellenbosch University |
Steyn J.,South African Council for Scientific and Industrial Research |
And 4 more authors.
Spectrochimica Acta - Part B Atomic Spectroscopy | Year: 2010
The detection of metallic silver on Chemical Vapour Deposited (CVD) grown silicon carbide and in Pebble Bed Modular Reactor (PBMR) supplied tri-structural isotropic (TRISO) coated particles (with 500 μm diameter zirconium oxide surrogate kernel) has been studied with femtosecond Laser Induced Breakdown Spectroscopy (femto-LIBS). The SiC layer of the TRISO coated particle is the main barrier to metallic and gaseous fission products of which 110mAg is of particular interest for direct cycle high temperature reactors. This work is a feasibility study for diagnosing and profiling silver transport through the silicon carbide layer of fuel particles for a high temperature gas reactor in out-of-reactor experimentation. The zirconium oxide is a surrogate for the enriched uranium oxide fuel. The conclusion reached in this study was that femto-LIBS can achieve good surface spatial resolution and good depth resolution for studies of silver in experimental coated particles. The LIBS technique also offers a good alternative for a remote analytical technique. © 2010 Elsevier B.V. All rights reserved.
Stoker C.C.,Pbmr Pty Ltd. |
Olivier L.D.,Independent Nuclear Consultants |
Stassen E.,Pbmr Pty Ltd. |
Reitsma F.,Pbmr Pty Ltd. |
Van Der Merwe J.J.,Pbmr Pty Ltd.
Nuclear Engineering and Design | Year: 2010
The determination of radionuclide source terms is vital for any reactor design and licensing safety evaluation. This paper provides an overview of the PBMR analysis tools with specific focus on the modelling of mobile and deposited radionuclide source terms within the pressure boundary of a typical pebble-bed high temperature reactor (HTR). The main focus is on the Dust and Activity Migration and Distribution (DAMD) software code system that models the activation, migration and time-dependent distribution of dust and atomic particles in an HTR such as the AVR and PBMR. Since DAMD provides a time-dependent systems integrated model of HTR designs, most of the obvious physical phenomena relevant to source terms are at play. These include the neutron flux, activation cross-sections, radioactive decay, dust production rates, dust impurity levels, dust filter capabilities, dust particle size distributions, thermal-hydraulic parameters influencing the migration and distribution of particles throughout the main power system and subsystems, and helium coolant leakage and make-up rates. At this stage the DAMD calibration and validation is mainly based on the operational data, experiments and measurements made during 21 years of operating life of the AVR. The comparisons of the DAMD results with various AVR measurements provide confidence in the use of DAMD for the PBMR design and safety evaluations. In addition, sensitivity analyses are performed with DAMD to determine the bounding system parameters that drive the migration and distribution of radionuclides. The use of DAMD to evaluate design configurations, e.g. the effect of the introduction and placement of filters on the radionuclide distribution, is also shown. In conclusion, the importance of a systems modelling approach for radionuclide transport and distribution within the pressure boundary of a typical HTR system, is demonstrated. Since the DAMD code system is calibrated and validated against the AVR measurements it can be concluded that the radionuclide source term phenomena in the AVR, resulting in the measured AVR contamination levels, is taken into account in the design and safety evaluation of the PBMR. © 2010 Elsevier B.V. All rights reserved All rights reserved.
Van Rooyen I.J.,South African Council for Scientific and Industrial Research |
Van Rooyen I.J.,Nelson Mandela Metropolitan University |
Van Rooyen I.J.,Pbmr Pty Ltd. |
Van Rooyen I.J.,Idaho National Laboratory |
And 6 more authors.
Nuclear Engineering and Design | Year: 2012
The integrity and property behavior of the SiC layer of the Tri-isotropic (TRISO) coated particle (CP) for high temperature reactors (HTR) are very important as the SiC layer is the main barrier for gaseous and metallic fission product release. This study describes the work done on un-irradiated SiC samples prepared with varying phosphorus levels to simulate the presence of phosphorus due to transmutation. 30Si transmutes to phosphorous ( 31P) and other transmutation products during irradiation, which may affect the integrity of the SiC layer. The P-doping levels of the SiC samples used in this study cover the range from 1.1 × 10 15 to 1.2 × 10 19 atom/cm 3 and are therefore relevant to the PBMR operating conditions. Annealing from 1000 °C to 2100 °C was performed to study the possible changes in nanostructures and various properties due to temperature. Characterization results by X-ray diffraction (XRD), secondary ion mass spectrometry (SIMS), scanning electron microscopy (SEM), transmission electron microscopy (TEM) and high resolution transmission electron microscopy (HRTEM), are reported in this article. As grain boundary diffusion is identified as a possible mechanism by which 110mAg, one of the fission activation products, might be released through intact SiC layer, grain size measurements is also included in this study. Temperature is evidently one of the factors/parameters amongst others known to influence the grain size of SiC and therefore it is important to investigate the effect of high temperature annealing on the SiC grain size. The ASTM E112 method as well as electron back scatter diffraction (EBSD) was used to determine the grain size of various commercial SiC samples and the SiC layer in experimental PBMR Coated Particles (CPs) after annealing at temperatures ranging from 1600 °C to 2100 °C. The HRTEM micrograph of the decomposition of SiC at 2100 °C are shown and discussed. Nanotubes were not identified during the TEM and HRTEM analysis although graphitic structures were identified. The preliminary conclusion reached is that the P-content at these experimental levels (1.1 × 10 15 to 1.2 × 10 19 atom/cm 3) does not have a significant influence on the nanostructure of SiC at high temperatures without irradiation. © 2011 Elsevier B.V.
Friedland E.,University of Pretoria |
Van Der Berg N.G.,University of Pretoria |
Malherbe J.B.,University of Pretoria |
Hancke J.J.,Pbmr Pty Ltd. |
And 3 more authors.
Journal of Nuclear Materials | Year: 2011
Transport of silver and iodine through polycrystalline SiC layers produced by PBMR (Pty) Ltd. for cladding of TRISO fuel kernels was investigated using Rutherford backscattering analysis and electron microscopy. Fluences of 2 × 1016 Ag+ cm-2 and 1 × 10 16 I+ cm-2 were implanted at room temperature, 350 °C and 600 °C with an energy of 360 keV, producing an atomic density of approximately 1.5% at the projected ranges of about 100 nm. The broadening of the implantation profiles and the loss of diffusors through the front surface during vacuum annealing at temperatures up to 1400 °C was determined. The results for room temperature implantations point to completely different transport mechanisms for silver and iodine in highly disordered silicon carbide. However, similar results are obtained for high temperature implantations, although iodine transport is much stronger influenced by lattice defects than is the case for silver. For both diffusors transport in well annealed samples can be described by Fickian grain boundary diffusion with no abnormal loss through the surface as would be expected from the presence of nano-pores and/or micro-cracks. At 1100 °C diffusion coefficients for silver and iodine are below our detection limit of 10-21 m2 s-1, while they increase into the 10-20 m2 s-1 range at 1300 °C. © 2011 Elsevier B.V. All rights reserved.
van Niekerk S.I.,Pbmr Pty Ltd. |
van Niekerk S.I.,University of Pretoria |
Steyn H.,University of Pretoria
South African Journal of Industrial Engineering | Year: 2011
The case of a nuclear engineering project was investigated to establish the relevant success criteria for the development of complex, high-technology systems. The project was first categorised according to an existing scheme, and the Delphi method was used to determine the criteria for project success that apply to this specific case. A framework of project success dimensions was extended to include criteria that are of specific importance for the project under consideration. While project efficiency (delivery on time and within budget) obviously still needs to be controlled, the results provide empirical evidence for the notion that, for 'super high tech' projects, this is relatively less important. The relative importance of the dimensions of success was also evaluated and presented on a timeline stretching from project execution to 10 years after project completion. This provided empirical evidence for certain concepts in the literature.
Fontanet J.,CIEMAT |
Herranz L.,CIEMAT |
Ramlakan A.,Pbmr Pty Ltd.
International Congress on Advances in Nuclear Power Plants 2010, ICAPP 2010 | Year: 2010
A break in the Helium Pressure Boundary (HPB) in a HTGR will result in the primary depressurization and the transport, by the helium, of circulating and resuspended particles into the confinement building. Uncertainties associated with the characterization of these aerosols could potentially have a substantial impact on its behavior in the building and, consequently, in the source term leaked to the environment. Therefore, sensitivity studies in aerosol source entering the confinement are of utmost relevance in the safety analysis performed in HTGRs. This paper analyses the effect of the break location, inlet mass flow rate and size distribution to the aerosol mass fraction leaked outside the confinement. In order to do so, three HPB break size accidents (small, large and very large) have been modeled with the ASTEC vl.3 code. The results have highlighted that the variable that has the most significant effect in the external release is the rate at which aerosol enters the confinement. The effect of other variables depends on the scenario. For small break, the particle size has a moderate effect whereas it is negligible in large breaks. Moreover, the break location plays a significant role for large breaks although the net released mass is hardly affected. Finally, the released mass fraction is shown to be correlated with the in-confinement residence time by a simple expression.
Stempniewicz M.M.,Nuclear Research and Consultancy Group |
Goede P.,Pbmr Pty Ltd.
Nuclear Engineering and Design | Year: 2016
This paper describes the work performed to find the sorption coefficients that represent well the available experimental data for cesium, iodine, and silver on dust particles. The purpose of this work is to generate a set of coefficients that may be recommended for computer code users. The following data was correlated:. •I-131 on AVR dust.•Ag-110m on AVR dust.•Cs-134 and Cs-137 on AVR dust. The results are summarized as follows:. •The available data can be correlated. The data scatter is about 4 orders of magnitude. Therefore the coefficients of the Langmuir isotherms vary by 4 orders of magnitude.•Sorption rates are higher at low temperatures and lower at high temperatures. This tendency has been observed in the data compiled at Oak Ridge. It is therefore surmised that the highest value of the sorption coefficients are appropriate for the low temperatures and the lowest value of the sorption coefficients are appropriate for the high temperatures. The recommended sorption coefficients are presented in this paper.•The present set of coefficients is very rough and should be a subject for future verification against experimental data. © 2015 Elsevier B.V.
Boer B.,Technical University of Delft |
Lathouwers D.,Technical University of Delft |
Kloosterman J.L.,Technical University of Delft |
Van Der Hagen T.H.J.J.,Technical University of Delft |
Strydom G.,Pbmr Pty Ltd.
Nuclear Technology | Year: 2010
The DALTON-THERMIX code system has been developed for safety analysis and core optimization of pebble-bed reactors. The code system consists of the new three-dimensional diffusion code DALTON, which is coupled to the existing thermal-hydraulic code THERMLX. These codes are linked to a database procedure for the generation of neutron cross sections using SCALE-5. The behavior of pebble-bed reactors during a loss of forced cooling (LOFC) transient is of particular interest since the absence of forced cooling could lead to a significant increase of the temperature of the coated particle fuel. Therefore, the reactor power may be constrained during normal operation to limit the temperature. For validation purposes, calculation results of normal operation, an LOFC transient, and a control rod withdrawal transient without SCRAM have been compared with experimental data obtained in the High Ternperature Reactor-10 (HTR-10). The code system has been applied to the 400-MW(thermal) pebble bed modular reactor (PBMR) design, including the analysis of three different LOFC transients. Theses results are verified by a comparison with the results of the existing TINTE code system. It was found that the code system is capable of modeling both small (HTR-10) and large (PBMR) pebble-bed reactors and therefore provides a flexible tool for safety analysis and core optimization of future reactor designs. The analyses of the LOFC transients show that the peak fuel temperature is only slightly elevated (less than +100°C) as compared to its nominal value in the HTR-10 but reaches a maximum value of 1648°C during the depressurized LOFC case of the PBMR benchmark, which is significantly higher than the peak fuel temperature (976°C) during normal operation.
Hindley M.P.,Pbmr Pty Ltd. |
Erasmus C.,Pbmr Pty Ltd.
7th South African Conference on Computational and Applied Mechanics, SACAM 2010 | Year: 2010
A probabilistic analyses methodology for the evaluation of graphite reflector components is proposed. The assessment of the integrity of the graphite reflector is done using Finite Element Analysis (FEA) which allows the modelling of the changes in material properties as well as the subsequent deformation and stress state experienced by each component of the reflector. The calculated stress is used in a probabilistic assessment of the integrity of each of the reflector components. Finally, expected statistical variances in the material properties are accounted for in the FEA simulation. Graphite is a quasi-brittle material that experiences changes in mechanical material properties when subjected to high doses of neutron irradiation. A number of campaigns have been launched to characterise the changes in graphite properties due to exposure to temperature and irradiation. These efforts however, depend on limited and costly experiments. Material models calculated to fit the experimental data are then used in the structural assessments. Expected statistical variances of these material model parameters have to be incorporated into structural assessments in order to determine the impact they can have on the integrity of the component. The design of the PBMR reflector consists of various interfacing graphite geometries, subjected to different fast neutron damage and extreme thermal loading. The loading combined with the extremely non-linear material behaviour (as the material properties of graphite change with temperature and neutron fluence) gives rise to complex stress states within the reflector blocks. At some point, the stress state will reach a critical level and cracks will initiate within the blocks. The presence of cracks presents a useful point to define the end of the part's life. The life of these graphite reflector parts in a Pebble Bed Modular Reactor (PBMR) core determines the service life of the Core Structures. As the replacement of the Core Structures' components will be costly and time consuming, it is important that the components of the Core Structures be designed for the best life possible as part of the conceptual design of the Pebble Bed Modular Reactor (PBMR). Structural assessments of reflector components are done with FEA models which incorporate the material property changes. A method that can be used for calculating the probability of failure of graphite parts in the nuclear core design has been developed by Hindley, et al.. The stresses obtained from individual FEA are used to calculate the Probability of Failure (PoF) for each analysis. The influence of this is first explained for the best fit material model. The effect of variances of the material model parameters is investigated by simulations using various material model parameter sets, each with unique statistical perturbations and comparing the results obtained from such non-linear analyses. The final aim of the graphite analyses is to assess the life that can be expected from the replaceable reflector components. © SACAM.