Prismatic Core Research Articles

New knapping methods in the Howiesons Poort at Sibudu (KwaZulu-Natal, South Africa)

The lithic technology study of layer Grey Sand at Sibudu reveals a large number of cores on flakes. Varying knapping methods of core reduction are presented here. Most of the core reduction techniques can be attributed to bladelet or small flake production. Also, we point to a new type of blade production, from prismatic cores, revealed by the study of debitage. These innovative knapping methods demonstrate the technological variability associated with Howiesons Poort lithic assemblages.

Thermal analysis of prismatic gas-cooled reactor core under coolant channel blockage accidents

Limited results of studies on coolant channel blockage accidents in a prismatic-gas cooled reactor core are available, although coolant channel blockage of the fuel block in a prismatic gas-cooled reactor has been regarded as a highly important phenomenon. In this study, therefore, intensive and extensive thermal analyses were carried out to enhance the understanding and knowledge about the thermal performance of multi-hole type prismatic fuel blocks under coolant channel blockage accidents. A series of thermo-fluid calculations were performed using the CORONA code. In contrast to existing works, a full core model was used to consider an accurate coolant flow redistribution after the blockage in this work. The results of the CORONA calculations show that the single channel blockage accidents of PMR200 are not a safety concern regardless of the location. However, the maximum fuel temperature exceeds the safety limit in the case of more than ∼10 channel blockage at the same region of the fuel column. It was also found that the selection of the computational domain is crucial. A full core analysis is required for accurate results in the case of severe blockage accidents.

CFD Analysis and Assessment for Cross-Flow Phenomena in VHTR Prismatic Core

The very high temperature gas-cooled reactor (VHTR) is a uranium-fueled, graphite-moderated, and helium-cooled reactor envisioned as one of the promising future nuclear reactor concepts due to its high efficiency, safety, and variety of applications. In this reactor concept, core bypass flow and cross flow are currently considered as key issues because of their inherent uncertainties and complication of prediction. Recently, the computational fluid dynamics (CFD) method has received a great deal of attention as a method for understanding the flow behavior in the VHTR core, including the bypass and the cross flow. However, validation of this method has not been sufficiently conducted yet based on real experimental data. For this reason, prediction capability of the CFD method was validated in this study by comparing the predictions with the existing multi-hole experimental data obtained by Groehn as a part of the core thermal-hydraulics design study for the NHDD PMR-200. A two-stacked fuel block with wedge-shaped cross gap was simulated for computational domain as the experimental setup. The flow loss coefficients and the velocity distributions of the cross flow from the experiment were compared to the CFD predictions extensively. As a result, good agreements between the CFD predictions and the experimental results were observed, confirming prediction capability of the CFD method for the complicated cross flow in the VHTR core. Furthermore, the velocity distributions and pressure distribution in the cross gap were investigated to identify the characteristics of the cross flow.

Computational fluid dynamics analysis of core bypass flow and crossflow in a prismatic very high temperature gas-cooled nuclear reactor based on a two-layer block model

The very high temperature gas-cooled nuclear reactor (VHTR) has been designated as one of the promising reactors that will serve for the Next Generation (Generation IV) Nuclear Plant. For a prismatic VHTR core, the bypass flow and crossflow phenomena are important design considerations. To investigate the coolant distribution in the reactor core based on the two-layer block facility built at Texas A&M University, a three-dimensional steady-state CFD analysis was performed using the commercial code STAR-CCM+ v6.04. Results from this work serve as a guideline and validating source for the related experiments. A grid independence study was conducted to quantify related errors in the simulations. The simulation results show that the bypass flow fraction was not a strong function of the Reynolds number. The presence of the crossflow gap had a significant effect on the distribution of the coolant in the core. Uniform and wedge-shape crossflow gaps were studied. It was found that a significant secondary flow in the crossflow gap region moved from the bypass flow gap toward coolant holes, which resulted in up to a 28% reduction of the coolant mass flow rate in the bypass flow gap.

Core design and safety analyses of 600 MWt, 950 °C high temperature gas-cooled reactor

The conceptual core design study of high temperature gas-cooled reactor (HTGR) is performed. The major specifications are 600MW thermal output, 950°C outlet coolant temperature, prismatic core type, enriched uranium fuel. The decay heat in the core can be removed with only passive measures, for example, natural convection reactor cavity cooling system (RCCS), even if any electricity is not supplied (station blackout). The transient thermal analysis of the depressurization accident in the case the primary coolant decreases to the atmosphere pressure shows that the fuels and the reactor pressure vessel temperatures are kept under their safety limit criteria. The fission product release, Ag-110m and Cs-137 from the fuels under the normal operation is small as to make maintenance of devices in the primary cooling system, such as a gas turbine, without remote maintenance. The HTGRs can achieve the advanced safety features based on their inherent passive safety characteristics.

Thermal Characteristics of Sandwich Plates with Prismatic Cores

In this paper are analyzed thermal characteristics of sandwich plates with prismatic core in order to optimize their geometry. Open structure of sandwich plates makes them convenient for cooling: the fluid flows through the open channels in the core and takes away the heat which can act on one or both surfaces of the plate. The construction of sandwich plates with prismatic core is also convenient for application where good heat conduction and high carrying capacity are requested simultaneously. If one bears in mind this double application, the task is to perform the optimization of the sandwich plate's geometrical parameters in order to apply various combinations of structural and thermal loads. Results presented in this paper provide possibility for large geometrical flexibility in combining thermal and structural characteristics optimization.

Study on Influence of Turbulence Model Selection on Prediction of Flow Distribution and Hot Spot Fuel Temperature in Prismatic HTGR Cores

CFD analyses were performed for a cross-flow experiment in prismatic HTGR core fuel blocks to investigate the effect of turbulence models on the prediction of the core flow distribution. A transition model was considered in addition to a fully turbulent model because the flow in the coolant channels of the fuel block is turbulent whereas that in the bypass gap between the blocks is laminar. Additional analyses were carried out to investigate how the variation of the coolant channel flow dependence on the turbulence model selection affected the hot spot fuel temperature. The results showed that the gamma-theta transition model led to a better prediction than the others because it reasonably simulated the flow region where the bypass and cross gaps intersected. The result for the hot spot fuel temperature, however, showed that the temperature change caused by the turbulence model selection is so small that it can be absorbed in a margin of the engineering design. Therefore, the fully turbulent model can be used as an efficient alternative to the transition model in the engineering calculation of an HTGR core flow.

VALIDATION OF NUMERICAL METHODS TO CALCULATE BYPASS FLOW IN A PRISMATIC GAS-COOLED REACTOR CORE

For thermo-fluid and safety analyses of a High Temperature Gas-cooled Reactor (HTGR), intensive efforts are in progress in the developments of the GAMMA+ code of Korea Atomic Energy Research Institute (KAERI) and the AGREE code of the University of Michigan (U of M). One of the important requirements for GAMMA+ and AGREE is an accurate modeling capability of a bypass flow in a prismatic core. Recently, a series of air experiments were performed at Seoul National University (SNU) in order to understand bypass flow behavior and generate an experimental database for the validation of computer codes. The main objective of the present work is to validate the GAMMA+ and AGREE codes using the experimental data published by SNU. The numerical results of the two codes were compared with the measured data. A good agreement was found between the calculations and the measurement. It was concluded that GAMMA+ and AGREE can reliably simulate the bypass flow behavior in a prismatic core.

Numerical and experimental investigation on labyrinth seal mechanism for bypass flow reduction in prismatic VHTR core

Core bypass flow in block type very high temperature reactor (VHTR) occurs due to the inevitable gaps between the hexagonal core blocks for the block installation and refueling. Since the core bypass flow affects the reactor safety and efficiency, it should be minimized to enhance the core thermal margin. In this regard, the core bypass flow reduction method applying the labyrinth seal mechanism was developed and optimized by using the single-objective shape optimization method. Response surface approximation (RSA) method was adopted as the optimization method. Side wall of the replaceable reflector block was redesigned and response surface approximate model was adopted to optimize the shape of the reflector wall. Computational fluid dynamics (CFD) analyses were carried out not only to assess the limitation of existing method of bypass flow reduction, but also to optimize the design of a newly developed reduction method. The experiment with Seoul National University (SNU) multi-block experimental facility was performed to demonstrate the performance of the reduction method. It was found that the effect of the existing bypass flow reduction method by sealing the bypass gap exit was restricted nearby the lower region of the core. However, the flow resistance factor of the bypass gap increased from 18.04 to 26.24 by the optimized reduction method. The results of the performance test showed that the bypass flow distribution was reduced throughout the entire core regions. The bypass flow ratios at the inlet and the outlet were reduced by 36.19% and 14.66%, respectively. In conclusion, it is expected that the core thermal margin and efficiency of VHTR reactor cooling can be enhanced by applying the developed bypass flow reduction method.

The Gravettian calvaria from Mollet III cave (Serinyà, Northeastern Iberian Peninsula)

Traditionally, fossil directors have been key elements for the characterization of chrono-cultural series over time. Current research on lithic technology demonstrate extensive and growing variability within technocomplexes, particularly in the Upper Paleolithic, making more complex the previously established chrono-cultural organization.In this article we analyze the lithic technology of level 497C from Cova Gran de Santa Linya, dated to 26.220 ± 220 BP (31260–30580 cal BP). Chronometric data indicates the assemblage is located between MIS 3 and MIS 2, likely placing it within the Gravettian tradition. The level 497C constitutes one of the few archaeological sites in NE Iberia for this period and allows identifying technical traits in core reduction methods, stone tools manufacture and site composition.Reduction strategies through prismatic cores were used to obtain distinct sequences of blades and bladelets. The assemblage of retouched artifacts is mainly composed by “fond commun”, pieces with abrupt retouch, and burins (mostly on oblique truncation). Backed tools, which were common during this period, are notably rare, as are Gravette points and Noailles burins, classic indicators that are not represented here. Nevertheless, a group of partially retouched macro-flakes forms a little-known category of instruments that can provide remarkable information regarding the technical behaviors of human groups.The technological features seen in level 497C may be consistent with the Gravettian technocomplex. However, we suggest the integration of technical-organizational indicators to assist the contextualization of the diversity of responses and technological decisions identified in technocomplexes. In this sense, the characteristics of this assemblage offer keys for evaluating archaeological sequences without fossil directors, thereby expanding our understanding of technological variability in this period in Iberia.

Technological change in Iberomaurusian culture: The case of Tamar Hat, Rassel and Columnata lithic assemblages (Algeria)

This paper highlights the results of a technological analysis carried out for the first time on nine lithic assemblages belonging to three Iberomaurusian sites in Algeria (Tamar Hat rockshelter, Rassel cave and Columnata open-air site). The purpose of this paper is to identify the technical system adopted for the production of lithic artifacts in each of the studied assemblages in order to search for similarities and discrepancies in the technical behaviours between coastal and inland Iberomaurusian populations. This requires recognition of the reduction core strategies and transformation of the produced blanks, as well as research into debitage methods and techniques employed. In all three sites, local raw materials were preferred and the lithic production was primarily geared towards obtaining relatively standardized lamellar blanks with different debitage schemes. In lower occupations at Tamar Hat, the bladelets were mainly produced by implementing a single and complex chaîne opératoire of lamellar production which integrated variable blanks according to numerous and different schemes. The dominant scheme was an elaborate debitage oriented to the production of short and narrow bladelet blanks from reduced prismatic and sub-pyramidal cores, while by-products were exploited as bladelet cores to produce micro-bladelets and burin spalls. A change in the core reduction strategies appeared in the upper occupations of Tamar Hat where elongated bladelets were produced according to a common simple debitage widespread in the latest occupations. The implementation of a single chaîne opératoire is preserved in the lower layer of Rassel, although with a less elaborate roughing out processing, as for the upper occupations of Tamar Hat. In contrast, at Columnata three independent chaînes opératoires were implemented to produce robust blanks, which represents a stark contrast with previous methods known in the other sites. The technological analysis has provided strong arguments for a different know-how in technical behaviours between Iberomaurusian populations living in coastal rockshelters and those in hinterlands open-air sites. Thus, the different geographical areas seem to cover variable economic entities which would suggest a new adaptation of the same populations to different ecological niches.

Constitutive modeling and finite element procedure development for stress analysis of prismatic high temperature gas cooled reactor graphite core components

High temperature gas cooled reactors, such as prismatic and pebble bed reactors, are increasingly becoming popular because of their inherent safety, high temperature process heat output, and high efficiency in nuclear power generation. In prismatic reactors, hexagonal graphite bricks are used as reflectors and fuel bricks. In the reactor environment, graphite bricks experience high temperature and neutron dose. This leads to dimensional changes (swelling and or shrinkage) of these bricks. Irradiation dimensional changes may affect the structural integrity of the individual bricks as well as of the overall core. The present paper presents a generic procedure for stress analysis of prismatic core graphite components using graphite reflector as an example. The procedure is demonstrated through commercially available ABAQUS finite element software using the option of user material subroutine (UMAT). This paper considers General Atomics Gas Turbine-Modular Helium Reactor (GT-MHR) as a bench mark design to perform the time integrated stress analysis of a typical reflector brick considering realistic geometry, flux distribution and realistic irradiation material properties of transversely isotropic H-451 grade graphite.

Design and homogenization of metal sandwich tubes with prismatic cores

Hollow cylindrical tubes with a prismatic sandwich lining designed to replace the solid cross-sections are studied in this paper. The sections are divided by a number of revolving periodic unit cells and three topologies of unit cells (Square, Triangle and Kagome) are proposed. Some types of multiple-topology designed materials are also studied. The feasibility and accuracy of a homogenization method for obtaining the equivalent parameters are investigated. As the curved elements of a unit cell are represented by straight elements in the method and the ratios of the lengths of the curved elements to the lengths of the straight elements vary with the changing number of unit cells, some errors may be introduced. The frequencies of the first five modes and responses of the complete and equivalent models under an internal static pressure and an internal step pressure are compared for investigating the scope of applications of the method. The lower bounds and upper bounds of the number of Square, Triangular and Kagome cells in the sections are obtained. It is shown that treating the multiple-topology designed materials as a separate-layer structure is more accurate than treating the structure as a whole.

Thermal–hydraulics analyses for 1/6 prismatic VHTR core and fuel element with and without bypass flow

Presented are the results of thermal–hydraulics analyses for a Very High Temperature Reactor (VHTR) 1/6 core and hexagonal fuel element, with and without helium-coolant bypass flow in interstitial gaps and in the control rod channels. Owing to the complexity and massive size of the VHTR, a full core analysis requires extensive and parallelized computation capabilities and a long time (weeks to months) to complete. These demanding requirements are mostly due to the 3-D computational fluid dynamics (CFD) simulation of the helium flow in the 10-m long coolant channels in the reactor core. Results demonstrate the effectiveness of coupling a 1-D helium flow in the channels together with a recently developed and validated convective heat transfer correlation, to a 3-D heat conduction in the graphite and fuel compacts in the core fuel elements. This simplified thermal–hydraulics analysis methodology for VHTR markedly reduces computation time and memory requirements, while maintaining accurate results. The helium bypass flow in interstitial gaps between fuel elements in the VHTR core provides additional cooling of the edge regions, but increases the temperature near the center of the prismatic fuel elements. Results also show that helium “bleed” flow through the control rod channels minimally affects the temperature distribution within the fuel elements. The heat generation in the corner burnable poison rods in the VHTR core fuel elements affects the temperature only in the close vicinity of the rods.

A multi-scale homogenization and reconstruction approach for solid material temperature calculations in prismatic high temperature reactor cores

Traditional full-core heat transfer analysis of high temperature reactors uses effective coarse mesh parameters that are typically derived from a priori analysis and/or simplified analytical models that approximate the subscale temperature response. Consequently, different assumptions are made on each spatial scale, potentially yielding inconsistent solution methods with large associated uncertainties. In contrast, homogenized cross-sections used in full-core neutronics analysis are obtained using consistent homogenization techniques applied in conjunction with unit cell calculations. This approach has been proven both efficient and accurate. It is therefore surprising that formal homogenization techniques are rarely used in heat transfer analysis of nuclear reactors. In this work we take advantage of distinct unit cells that can be identified on each spatial scale in the MHTGR reactor core with the view to develop a consistent and accurate methodology for constructing hierarchical coarse mesh models for solid heat conduction in this reactor type. Three techniques have been used: formal multi-scale expansion homogenization is applied to obtain effective unit cell thermodynamic parameters; coarse mesh temperature discontinuities are defined to ensure continuity of the fine-scale temperatures at interfaces; and reduced order models for the time-dependent temperature response of the unit cells are obtained using proper orthogonal decomposition applied to detailed unit cell simulation results. The result is an efficient method, aimed toward unstructured CFD frameworks, that accurately captures coarse mesh temperatures with the capability of fully reconstructing the fine scale solution at any hierarchical level. The advantages of this method are illustrated for a small prismatic HTGR core using a cascaded solution approach. Starting at the finest scale, the TRISO coated particles, high resolution unit cell calculations are performed in a hierarchical fashion to build up a library of homogenized coarse mesh parameters and reduced order models, which are then used for the full-core heat conduction analysis. We demonstrate the accuracy and efficiency of the method by comparing results for a typical HTR power excursion transient against detailed reference solutions.

Numerical Simulation and Turbulent Convection Heat Transfer Correlation for Coolant Channels in a Very-High-Temperature Reactor

A turbulent convection correlation is developed for use as a surrogate of full three-dimensional (3-D) computational fluid dynamics numerical calculations for helium coolant flow in channels of a prismatic core high-temperature gas-cooled reactor or very-high-temperature reactor. It is developed based on the results of a 3-D thermal-hydraulics numerical analysis of a prismatic core hexagonal fuel module with a central flow channel. The analysis accounts for the changes in material properties with temperature, including the nuclear graphite, helium coolant, and the constituent materials of the fuel compacts, as well as the chopped-cosine axial fission power profile in the reactor core. The heated flow channel is 1.5875 cm in diameter and 8.0 m high. In addition, the flow channel extends 1.2 m and 0.8 m into unheated top and bottom graphite reflector blocks, for a total length of 10 m. Results show that entrance mixing increases the local turbulent heat transfer coefficient, but its effect diminishes after a distance into the heated section equal to 25 channel diameters. The developed correlation is within ±2% of the 3-D numerical results for both uniform and chopped-cosine axial power profiles. It is comparable to those reported in the literature based on experimental measurements of the local heat transfer coefficient for gas flows in uniformly heated tubes, including the entrance flow mixing section.

Preliminary reactor physics assessment of the HTR module with 14% enriched UCO fuel

Preliminary reactor physics assessment of the HTR module with 14% enriched UCO fuel

Comparative modular analysis of two complex sulfosalt structures: sterryite, Cu(Ag,Cu)3Pb19(Sb,As)22(As–As)S56, and parasterryite, Ag4Pb20(Sb,As)24S58

The crystal structures of two very close, but distinct complex minerals of the lead sulfosalt group have been solved: sterryite, Cu(Ag,Cu)(3)Pb(19)(Sb,As)(22)(As-As)S(56), and parasterryite, Ag(4)Pb(20)(Sb,As)(24)S(58). They are analyzed and compared according to modular analysis. The fundamental building block is a complex column centred on a Pb(6)S(12) triangular prismatic core, with two additional long and short arms. The main chemical and topological differences relate to the short arm, which induces a relative a/4 shift (~2 Å along the elongation parameter) of the constitutive rod layers, as illustrated by distinct cell settings within the same space group (P2(1)/n and P2(1)/c, respectively). Selection of the shortest (i.e. strongest) (Sb,As)-S bonds permitted to enhance the polymeric organization of (Sb,As) atoms with triangular pyramidal coordination. These two quasi-homeotypic structures are expanded derivatives of owyheeite, Ag(3)Pb(10)Sb(11)S(28). The hierarchy of organization levels from zero- to three-dimensional entities is subordinated to building operators, which appear as the driving force for the construction of such complex structures. Minor cations (Ag, Cu) or the As-As pair in sterryite secure the final locking, which favours the formation of one or the other compound.

Effects of graphite surface roughness on bypass flow computations for an HTGR

Bypass flow in a prismatic high temperature gas reactor (HTGR) occurs between graphite blocks as they sit side by side in the core. Bypass flow is not intentionally designed to occur in the reactor, but is present because of tolerances in manufacture, imperfect installation and expansion and shrinkage of the blocks from heating and irradiation. It is desired to increase the knowledge of the effects of such flow; it has been suggested that it may be as much as 20% of the total helium coolant flow [INL Report 2007, INL/EXT-07-13289]. Computational fluid dynamic (CFD) simulations can provide estimates of the scale and impacts of bypass flow. Previous CFD calculations have examined the effects of bypass gap width, level and distribution of heat generation and effects of shrinkage. The present contribution examines the effects of graphite surface roughness on the bypass flow for different relative roughness factors for three gap widths. Such calculations should be validated using specific bypass flow measurements. While such experiments are currently underway for the specific reference prismatic HTGR design for the next generation nuclear plant (NGNP) program of the U.S. Dept. of Energy, the data are not yet available. To enhance confidence in the present calculations, wall shear stress and heat transfer results for several turbulence models and their associated wall treatments are first compared for steady flow in a single tube that is representative of a coolant channel in the prismatic HTGR core. The results are compared to published correlations for wall shear stress and Nusselt number in turbulent pipe flow. Turbulence models that perform well are then used to make steady bypass flow calculations in a symmetric one-twelfth sector of a prismatic block that includes bypass flow. The comparison of shear stress and Nusselt number results with published correlations constitutes a partial validation of the CFD model. Results indicate that increasing surface roughness increases the maximum fuel and helium temperatures as do increases in gap width. However, maximum coolant temperature variation due to increased gap width is not changed by surface roughness.

The effects of crossflow gap and axial bypass gap distribution on the flow characteristics in prismatic VHTR core

In prismatic VHTR design, the occurrence of the bypass gap between the core blocks is inevitable for installation and refueling. Since the core bypass flow affect both the reactor efficiency and core thermal margin, it is important to evaluate the bypass flow distribution accurately. The gap size varies with the reactor fuel cycle as the dimension of graphite block varies due to the fast-neutron induced shrinkage and thermal expansion. In particular, the dimensional change of graphite block is different from each location in the core, since the neutron fluence of the blocks is different locally. In addition, fuel and reflector blocks are rearranged by axial shuffling for refueling, so that bypass gap distribution is not axially uniform. Therefore, the consideration of axial profile of the bypass gap is important for accurate evaluation of bypass flow but most of the previous studies assumed uniform bypass gap size. In present study, the multi-block air test experiment was carried out to evaluate the bypass flow distribution with the consideration of axially varying bypass gap profile and the crossflow effect. Axial bypass gap profile of 6mm, 2mm, 4mm, 2mm from top to bottom layer and the crossflow gap of 2mm were tested in the experiment. In addition, comparative computational fluid dynamics (CFDs) analysis for the multi-block experiment was carried out to supplement the experimental limit and to analyze the local flow characteristics in detail. The results suggest that the CFD code, CFX-12 has a sufficient capability to analyze the VHTR core and the prismatic VHTR core should be modeled as the flow network system. The experimental data and CFD simulation results could be used to validate the thermal–hydraulics system code for VHTR system.