bluuebunny/crossref_metadata_2025 huggingface dataset:
Idaho National Laboratory Experimental Program to Measure the Flow Phenomena in a Scaled Model of a Prismatic Gas-Cooled Reactor Lower Plenum for Validation of CFD Codes
Authors: McIlroy Hugh M. , McEligot Donald M. , Pink Robert J.
Published: July 2009

The experimental program that is being conducted at the Matched Index-of-Refraction (MIR) Flow Facility at Idaho National Laboratory (INL) to obtain benchmark data on measurements of flow phenomena in a scaled model of a prismatic gas-cooled reactor lower plenum using 3-D Particle Image Velocimetry (PIV) is presented. A description of the scaling analysis, experimental facility, 3-D PIV system, measurement uncertainties and analysis, experimental procedures and samples of the data sets that have been obtained are included. Samples of the data set that will be presented include the mean velocity field in an approximately 1:7 scale model of a region of the lower plenum of a typical prismatic gas-cooled reactor (GCR) similar to a General Atomics Gas-Turbine-Modular Helium Reactor (GTMHR) design. This experiment has been selected as the first Standard Problem endorsed by the Generation IV International Forum. The flow in the lower plenum consists of multiple jets injected into a confined cross flow — with obstructions. The model consists of a row of full circular posts along its centerline with half-posts on the two parallel walls to approximate flow scaled to that expected from the staggered parallel rows of posts in the reactor design. The model is fabricated from clear, fused quartz to match the refractive-index of the mineral oil working fluid. The benefit of the MIR technique is that it permits high-quality measurements to be obtained without locating intrusive transducers that disturb the flow field and without distortion of the optical paths. An advantage of the INL MIR system is its large size which allows obtaining improved spatial and temporal resolution compared to similar facilities at smaller scales. Results concentrate on the region of the lower plenum near its far reflector wall (away from the outlet duct). Inlet jet Reynolds numbers (based on the jet diameter and the time-mean average flow rate) are approximately 4,300 and 12,400. The measurements reveal developing, non-uniform flow in the inlet jets and complicated flow patterns in the model lower plenum. Data include three-dimensional vector plots, data displays along the coordinate planes (slices) and charts that describe the component flows at specific regions in the model. Information on inlet velocity profiles is also presented.
Using an Effective Viscosity to Account for the Effect of the Walls on the Pressure Drop Through an Annular Packed Bed
Authors: du Toit C. G. , van der Walt A. J. K. , van der Merwe J.
Published: July 2009

The question arose whether the correlation for the pressure drop prescribed for cylindrical pebble bed reactors by the Nuclear Safety Commission (KTA) of Germany could still be applied to the proposed annular configuration of the Pebble Bed Modular Reactor (PBMR) currently being developed in South Africa. An approach is described which uses the extended Brinkman equation for fully developed flow together with the original KTA correlation, to account for the resistance of the pebbles, and an effective viscosity, to account for the effect of the walls. A cylindrical packed bed with the same hydraulic diameter as the annular core was first of all considered. The pressure drops for various Reynolds numbers were calculated using a correlation which accounts for the effect of the wall. The formulation of the correlation for an infinite bed was then used along with the Brinkman equation to determine the appropriate values of the effective viscosity to give the same pressure drops. It was then assumed that the effective viscosities obtained in this way could be applied to the annular configuration of the PBMR. The pressure drop through the annular core was then calculated for various Reynolds numbers employing the effective viscosities in the extended Brinkman equation. It was found that the friction coefficients that could be derived from these pressure drops were in good agreement with the friction coefficients obtained from physical experiments performed on a scale model of the PBMR annular core. It was therefore concluded that the strategy followed could be used with the necessary care to predict the pressure drop through the annular core of the PBMR.
Preliminary Design Analysis of Hot Gas Ducts for a Nuclear Hydrogen System of 200 MWt
Authors: Song Kee-Nam , Kim Yong-Wan
Published: July 2009

Very High Temperature Gas Cooled Reactor (VHTR) has been selected as a high energy heat source for a nuclear hydrogen generation. The VHTR can produce hydrogen from heat and water by using a thermo-chemical process or from heat, water, and natural gas by steam reformer technology. Korea Atomic Energy Research Institute (KAERI) is in the process of carrying out a nuclear hydrogen system by considering the indirect cycle gas cooled reactors that produce heat at temperatures in the order of 950°. The nuclear hydrogen system is planning to produce hydrogen by using nuclear energy and a thermo-chemical process. Helium gas is the choice for the coolant of the nuclear hydrogen system since it is an inert gas, with no affinity to a chemical or nuclear activity; therefore a radioactivity transport in the primary circuit of the nuclear hydrogen system is minimal under a normal operation. Moreover, its gaseous nature avoids problems related to a phase change and water-metal reactions and therefore improves its safety. A coaxial double-tube hot gas duct (HGD) is a key component connecting the reactor pressure vessel and the intermediate heat exchanger (IHX) for the nuclear hydrogen system. In this study, a preliminary design analysis for the primary and secondary HGDs of the nuclear hydrogen system was carried out. These preliminary design activities include a preliminary decision on the geometric dimensions, a preliminary strength evaluation and an appropriate material selection. A preliminary decision on the geometric dimensions of the HGDs was undertaken based on three engineering concepts, such as a constant flow velocity model (CFV model), a constant flow rate model (CFR model), a constant hydraulic head model (CHH model), and also based on a heat balanced model (HB model). We compared the geometric dimensions and their preliminary strength evaluation results from the various models.
A Synthesis on the HTR Scenario Studies at CEA
Authors: Boucher Lionel , Greneche Dominique
Published: July 2009

From 2002 to 2007, an important program study on HTR has been carried out by CEA and AREVA NC under the joint CEA–AREVA NC project “prospective studies on the management of Plutonium and the back end of the cycle.”
Research and Development on Hydrogen Production by Thermochemical Water-Splitting IS Process
Authors: Terada Atsuhiko , Iwatsuki Jin , Takegami Hirokaki , Kubo Shinji , Sakaba Nariaki , Onuki Kaoru
Published: July 2009

A project using the High Temperature Engineering Test Reactor (HTTR) is being proceeded at the Japan Atomic Energy Agency (JAEA), which aims to develop technologies of nuclear production of hydrogen to meet its massive demand in future. The HTTR with thermal power of 30MW is the Japanese high-temperature gas-cooled reactor (HTGR), built and operated at the site of the Oarai Research & Development Center of JAEA. Under the HTTR project, JAEA has been conducting research and development on thermochemical IS process aiming to reach its final goal of demonstrating the nuclear hydrogen production by an HTTR-IS system. The IS process is a chemical process composed of three reactions, i.e. hydrogen iodide decomposition, sulfuric acid decomposition, and production of hydrogen iodide and sulfuric acid. JAEA completed one-week continuous hydrogen production using a glass-made bench-scale test apparatus equipped with an automatic control system in 2004, where the hydrogen production rate was about 30 NL/hr. Based on the successful test, present activity focuses on the development of components to be used in the corrosive process environments and, also, of technologies to realize high thermal efficiency of hydrogen production. This paper describes present status of the activities such as conceptual design of a test plant driven by sensible heat of helium gas heated by an electric heater simulating the HTTR, and component tests for service in the corrosive environments of high temperature sulfuric acid. Also, R&D results are described on glass lining technologies and a reliability evaluation of components made of SiC ceramics, which are promising candidates for application to, e.g. the sulfuric acid decomposer and the hydriodic acid distillation column.
Investigation of Printed Circuit Heat Exchangers for VHTRs
Authors: Mylavarapu Sai K. , Sun Xiaodong , Figley Justin , Needler Noah J. , Christensen Richard N.
Published: July 2009

Very High Temperature Reactors (VHTRs) require high temperature (900–950 °C), high integrity, and high efficiency heat exchangers during normal and off-normal conditions. A class of compact heat exchangers, namely, the Printed Circuit Heat Exchangers (PCHEs), made of high temperature materials, found to have the above characteristics, are being increasingly pursued for heavy duty applications. A high-temperature helium experimental test facility, primarily aimed at investigating the heat transfer and pressure drop characteristics of the PCHEs, was designed and is being built at the Ohio State University. The test facility was designed for a maximum operating temperature and pressure of 900 °C and 3 MPa, respectively. Owing to the high operating conditions, a detailed investigation on various high temperature materials was carried out to aid in the design of the test facility and the heat exchangers. The study showed that alloy 617 is the leading candidate material for high temperature heat exchangers. Two PCHEs, each having 10 hot and 10 cold plates with 12 channels in each plate, are currently being fabricated from alloy 617 plates and will be tested once the test facility is constructed. To supplement the experiments, computational fluid dynamics modeling of a simplified PCHE model is being performed and the results for three flow rate cases of 15, 40, and 90 kg/h and a system pressure of 3 MPa are discussed. In summary, this paper focuses on the study of the high-temperature materials, the design of the helium test facility, the design and fabrication of the PCHEs, and the computational modeling of a simplified PCHE model.
Analysis of the HTGR Proliferation Resistance
Authors: Petrunin Vitaly V. , Kodochigov Nikolay G. , Sukharev Yury P. , Osipov Sergey L. , Marova Elena V.
Published: July 2009

Increased interest in development of nuclear power engineering, first of all in non-nuclear countries, puts an emphasis on the designing of small and medium nuclear power plants and determines the growth in nuclear technology export from countries with advanced nuclear industries. It accentuates the issue of reduction of the nuclear material proliferation risk, which was repeatedly raised on the national an international levels (materials of INPRO, GNEP, IAEA).
RAPHAEL: Achievements Within the VHTR Materials and Components Programmes
Authors: Buckthorpe Derek , Breuil Eric
Published: July 2009

HTR projects have been launched within the European Union Framework Programmes (FP’s) to consolidate and advance HTR and VHTR technology within Europe. This paper reviews the main achievements arising out of the work in the area of materials and component development. The programme to date addresses material and qualification requirements for the reactor pressure vessel, high temperature resistant alloys and technological development aspects of the power circuit components, material property needs and issues for the graphite core and requirements for Codes and Standards. The experimental programme includes irradiation and feature testing, tests on reduced scale mock-ups and bearings, corrosion, modelling and analysis issues. For the 6th Framework activities which are current, the main European research focus on VHTR is through the RAPHAEL-IP. Results and main conclusions from the work are reported, also a summary of the status of the test work and recommendations for future actions. This programme of work provides important results for the International Generation IV VHTR Materials and Components Research and Development programme as part of the EURATOM contribution.
High Temperature Interaction Between UO2 and Carbon: Application to TRISO Particles for Very High Temperature Reactors
Authors: Gosse´ S. , Gue´neau C. , Boinski F. , Alpettaz T. , Chatain S. , Chatillon C.
Published: July 2009

For Very High Temperature Reactors (V-HTR), the study of the Uranium-Carbon-Oxygen system is of major importance to predict the high temperature behaviour of the TRISO fuel particle. Firstly, the high level operating temperature of the fuel materials in normal and accidental conditions requires studying the possible chemical interaction between the UO2 fuel kernel and the surrounding structural materials (C, SiC) that could damage the particle. The formation of the gaseous carbon oxides at the fuel (UO2)-buffer (C) interface that leads to the build up of the internal pressure in the particle has to be predicted. Secondly, the U-C-O ternary system is also involved in the fabrication process of “UCO” kernels made of a mixture of UO2 and UC2. For the fabrication of such mixture of uranium oxide and carbide, the phase diagram and thermodynamic properties of the U-C-O system are necessary to investigate in order to perform adequate heat treatments. For both reasons, a new study of the U-C-O ternary system has been undertaken. Firstly, some thermodynamic calculations (equilibrium CO(g) and CO2(g) pressures, phase diagrams) were performed using the thermodynamic FUELBASE database dedicated to generation IV fuels [1]. The results allow representing the different phase equilibria involving carbide and oxicarbide phases at high temperature. They also show the high level of CO(g) and CO2(g) equilibrium pressures above the UO2±x fuel in equilibrium with carbon which could lead to the failure of the particle for high oxygen stoichiometry of the uranium dioxide. In a second step, the partial pressures of CO(g) and CO2(g) resulting from the UO2/C interaction have been measured by high temperature mass spectrometry. Two types of samples were used (i) pellets made of a mixture of UO2 and C powders or (ii) UO2 kernels disseminated in a carbon bed. The kinetic measurements of the release of CO(g) and CO2(g) lead to measured pressures that are lower than the equilibrium pressures predicted from thermodynamic calculations. This discrepancy can be explained by limitations due to distinct kinetic mechanisms. Rates of CO(g) formation have been established taking into account the oxygen stoichiometry of uranium oxide and temperature. The major gaseous product is always CO(g) which release significantly starts at 1473 K. The influence of the different geometries is shown. The limitative kinetic step can be an interface or a diffusion process as a function of the type of sample. These results underline the up most importance of kinetic factors for studying the UO2 / C interaction to determine realistic CO(g) pressure levels inside a TRISO particle or to improve the fabrication process of the “UCO” kernels.
Development of the Code GOLT for Performance Evaluation of Coated-Particles Fuel
Authors: Golubev Igor , Kadarmetov Igor
Published: July 2009

The coated particles (CP) performance computer code GOLT (Russian abbreviation of Gas-Cooled Fuel) is under development at the A. A. Bochvar All-Russia Research Institute of Inorganic Materials. The main goal of the code is supporting development of fuel for the Gas-Turbine Modular Helium Reactor (GT-MHR). The first version GOLT-v1 has capable to calculate temperature distribution along particle radius, fuel kernel swelling, development of internal pressure under coating due to formation of gaseous fission products and CO, development of stresses and deformation in each coating layer. For TRISO-type particles special probabilistic failure model was developed. According to the failure model integrated probability of silicon carbide failure depends on probability of each dense pyrocarbon layer failure. Probabilistic version GOLT-v2 takes into account possibility of gap formation between buffer and inner dense pyrocarbon layer or between kernel and buffer that influences on maximal fuel temperature and stresses distribution in coating. More detail model of buffer performance at irradiation was developed and included in the code. List of probable coating failure mechanisms was extended. The ability of coating failure due to Kernel-Coating Mechanical Interaction (KCMI) as well as model of failure due to kernel migration was added. Thermo-dynamical code ASTRA is used in some tasks as supporting tool for calculating internal pressure and chemical interaction between SiC coating and fission products and CO. The version GOLT-v3 has accumulated all capabilities of previous versions and included Monte-Carlo analysis for estimation of fraction of failed particles with account of statistical dispersion of structural, materials and operating parameters. In the paper short description of capabilities of last versions of the code is presented. Main attention is putted to results of development version GOLT-v2a for evaluation fuel performance during accidents.
A Second-Order Accurate Convection Heat Transfer Discretisation Scheme for a Systems Simulation Reactor Model
Authors: van Antwerpen H. J. , Greyvenstein G. P.
Published: July 2009

Nuclear reactor behaviour is determined by the relation between the temperature and the effective neutron reaction cross-sections of the materials in the reactor core. In order to accurately calculate the material temperatures, heat transfer to the reactor coolant is of critical importance. To increase the accuracy of heat transfer calculation, this paper presents a second-order accurate convection heat transfer discretisation for application in two or more dimensions in a finite-volume discretisation. This discretisation is compared to a first-order accurate scheme by means of a grid dependence study on representative reactor geometry. To obtain a certain level of accuracy, the use of second-order accurate convection discretisation reduced the solution time by up to 60 percent, as the second-order accurate scheme required fewer grid points than the first-order accurate scheme. This is of particular importance for systems simulation reactor models, as they use the minimum number of grid points and have to run at the highest possible simulation speed. This study was done using the systems simulation code Flownex.
Fission Product Release From HTGR Fuel Under Core Heatup Accident Conditions
Authors: Verfondern Karl , Nabielek Heinz
Published: July 2009

Various countries engaged in the development and fabrication of modern fuel for the High Temperature Gas-Cooled Reactor (HTGR) have initiated activities of modeling the fuel and fission product release behavior with the aim of predicting the fuel performance under operating and accidental conditions of future HTGRs. Within the IAEA directed Coordinated Research Project CRP6 on “Advances in HTGR Fuel Technology Development” active since 2002, the 13 participating Member States have agreed upon benchmark studies on fuel performance during normal operation and under accident conditions. While the former has been completed in the meantime, the focus is now on the extension of the national code developments to become applicable to core heatup accident conditions. These activities are supported by the fact that core heatup simulation experiments have been resumed recently providing new, highly valuable data. Work on accident performance will be — similar to the normal operation benchmark — consisting of three essential parts comprising both code verification that establishes the correspondence of code work with the underlying physical, chemical and mathematical laws, and code validation that establishes reasonable agreement with the existing experimental data base, but including also predictive calculations for future heating tests and/or reactor concepts. The paper will describe the cases to be studied and the calculational results obtained with the German computer model FRESCO. Among the benchmark cases in consideration are tests which were most recently conducted in the new heating facility KUEFA. Therefore this study will also re-open the discussion and analysis of both the validity of diffusion models and the transport data of the principal fission product species in the HTGR fuel materials as essential input data for the codes.
Simulation of Performance of Centrifugal Circulators With Vaneless Diffuser for GCR Applications
Authors: Tauveron N.
Published: July 2009

In the frame of the international forum GenIV, CEA has selected various innovative concepts of Gas cooled Nuclear Reactor. Among them, an indirect-cycle gas reactor is under consideration. Thermal hydraulic performances are a key issue for the design. For transient conditions and decay heat removal situations, the thermal hydraulic performance must remain as high as possible. In this context, all the transient situations, the incidental and accidental scenarii must be evaluated by a validated system code able to correctly describe, in particular, the thermalhydraulics of the whole plant. As concepts use a helium compressor to maintain the flow in the core, a special emphasis must be laid on compressor modelling. Centrifugal circulators with a vaneless diffuser have significant properties in term of simplicity, cost, ability to operate over a wide range of conditions. The objective of this paper is to present a dedicated description of centrifugal compressor, based on a one dimensional approach. This type of model requires various correlations as input data. The present contribution consists in establishing and validating the numerical simulations (including different sets of correlations) by comparison with representative experimental data. The results obtained show a qualitatively correct behaviour of the model compared to open literature cases of the gas turbine aircraft community and helium circulators of High Temperature Gas Reactors. Further work on modelling and validation are nevertheless needed to have a better confidence in the simulation predictions.
Challenges and Opportunities in Providing a Digital Protection System for the PBMR
Authors: Marais Johan , Ridolfo Charles F.
Published: July 2009

The Republic of South Africa is currently developing the Pebble Bed Modular Reactor (PBMR); an advanced, fourth-generation reactor that incorporates inherent safety features, which require no human intervention and which provide an unprecedented level of nuclear safety. In addition to electrical power generation, the reactor is uniquely suited for a variety of non-traditional nuclear applications including oil sands extraction, desalination, and hydrogen production. A state-of-the-art digital Protection System for the PBMR is currently being developed in conjunction with Westinghouse Electric Company (WEC). The Protection System provides for: • reactor shutdown using two different reactor trip methodologies (dropping of the control rods and insertion of Small Absorber Spheres (SASs) which are composed of boron carbide); • post-event monitoring; and • manual reactor shutdown, which is independent of software-based systems. The reactor shutdown and post-event instrumentation monitoring components of the Protection System are being implemented utilizing the WEC ‘Common Q’ platform, which is comprised of ‘commercially dedicated’ Programmable Logic Controllers (PLCs), colour-graphic Flat Panel Displays (FPDs) with integral touch screens, and high-speed data communication links. High reliability and availability are achieved through component redundancy, continuous automatic self-testing which is run online in a background mode, and implementation of a multi-channel system design which is tolerant to failures. The Protection System is also designed to support periodic surveillance testing through a suite of built-in computer-aided test facilities that are accessible via an FPD interface. These allow various system surveillance requirements to be readily performed in a convenient and systematic manner. This paper discusses the following topics with regard to the PBMR Protection System: development strategy, functional requirements, selection of applicable Codes and Standards, key design specifications, architectural configuration, design and implementation challenges, and unique opportunities that are provided by this type of Protection System.
Density Change of an Oxidized Nuclear Graphite by Acoustic Microscopy and Image Processing
Authors: Chi Se-Hwan , Contescu Cristian I. , Burchell Timothy D.
Published: July 2009

The strong correlation between the density and the physical and, mechanical properties of graphite suggests that the method of nondestructive density evaluation could be developed into a characterization technique of great value for the overall improvement of safety of graphite moderator reactors. In this study, the oxidation-induced density changes in nuclear graphite for VHTR were determined by a conventional destructive bulk density measurement method (BM), and by a new non-destructive method based on acoustic microscopy and image processing (AM). The results were compared in order to validate the applicability of the latter method. For a direct comparison of the results from both measurements, two specimens were prepared from a cylindrical graphite sample (1 inch diameter and 1 inch height, oxidized to 10% weight loss at 973 K in air for 5 hours). The specimens were used for characterization by BM and AM methods, respectively. The results show that, even with a large standard deviation of the AM, the density changing trend from both methods appeared the same. This observation may be attributed to the fact that AM images reflect characteristic density changes of the graphite sample through the acoustic impedance changes. This study demonstrates the possibility of using AM as a nondestructive technique for the evaluation of density changes in graphite when a database is prepared through a systematic series of experiments.
Maintaining a Technology-Neutral Approach to Hydrogen Production Process Development Through Conceptual Design of the Next Generation Nuclear Plant
Authors: Patterson Michael W. , Park Charles V.
Published: July 2009

The Energy Policy Act of 2005 (EPAct) charges the Department of Energy (DOE) with developing and demonstrating the technical and economic feasibility of using high temperature gas-cooled reactor (HTGR) technology for the production of electricity and/or hydrogen. The design, construction and demonstration of this technology in an HTGR proto-type reactor are termed the Next Generation Nuclear Plant (NGNP) Project. Currently, parallel development of three hydrogen production processes will continue until a single process technology is recommended for final demonstration in the NGNP — a technology neutral approach. This analysis compares the technology neutral approach to acceleration of the hydrogen process downselection at the completion of the NGNP conceptual design to improve integration of the hydrogen process development and NGNP Project schedule. The accelerated schedule activities are based on completing evaluations and achieving technology readiness levels (TRLs) identified in NGNP systems engineering and technology roadmaps. The cost impact of accelerating the schedule and risk reduction strategies was also evaluated. The NGNP Project intends to design and construct a component test facility (CTF) to support testing and demonstration of HTGR technologies, including those for hydrogen production. The demonstrations will support scheduled design and licensing activities, leading to subsequent construction and operation of the NGNP. Demonstrations in the CTF are expected to start about two years earlier than similarly scaled hydrogen demonstrations planned in the technology neutral baseline. The schedule evaluation assumed that hydrogen process testing would be performed in the CTF and synchronized the progression of hydrogen process development with CTF availability.
Diffusion of Silver in Single Crystalline 6H-SiC
Authors: Hlatshwayo T. T. , van der Berg N. G. , Friedland E. , Malherbe J. B. , Chakraborty P.
Published: July 2009

In a modern high-temperature nuclear reactor, safety is achieved by encapsulating the fuel elements by CVD-layers of pyrolytic carbon and silicon carbide (SiC) to prevent the fission products release. Some studies have raised doubts on the effectiveness of SiC layer as a diffusion barrier to fission fragments due to 110mAg released from the coated particle at high temperatures ranging from 1500°C to 1600°C [1].
High Temperature Behavior of Candidate VHTR Heat Exchanger Alloys
Authors: Wright Richard , Simpson Joel , Wertsching Alan , Swank W. David
Published: July 2009

Several nickel based solid solution alloys are under consideration for application in heat exchangers for very high temperature gas cooled reactors. The principal candidates being considered for this application by the Next Generation Nuclear Plant (NGNP) project are Inconel 617 and Haynes 230. While both of these alloys have an attractive combination of creep strength, fabricability, and oxidation resistance a good deal remains to be determined about their environmental resistance in the expected NGNP helium chemistry and their long term response to thermal aging. A series of experiments has been carried out in a He loop with controlled impurity chemistries within the range expected for the NGNP. The influence of oxygen partial pressure and carbon activity on the microstructure and mechanical properties of Alloys 617 and 230 has been characterized. A relatively simple phenomenological model of the environmental interaction for these alloys has been developed.
An Efficient Hybrid Sulfur Process Using PEM Electrolysis With a Bayonet Decomposition Reactor
Authors: Gorensek Maximilian B. , Summers William A. , Lahoda Edward J. , Bolthrunis Charles O. , Greyvenstein Renee
Published: July 2009

The Hybrid Sulfur (HyS) Process is being developed to produce hydrogen by water-splitting using heat from advanced nuclear reactors. It has the potential for high efficiency and competitive hydrogen production cost, and has been demonstrated at a laboratory scale.
Autonomous Control Strategies for VHTR-Based Systems for Hydrogen Production
Authors: Tsvetkov Pavel V. , Alajo Ayodeji B. , Ames David E.
Published: July 2009

This paper is focused on feasible autonomous control strategies for Generation IV Very High Temperature Reactors (VHTR)-based systems for hydrogen production. Various burnable poison distributions and fuel compositions are considered. In particular, utilization of TRUs in VHTRs is explored as the core self-stabilization approach. Both direct cycle and indirect cycle energy conversion approaches are discussed. It is assumed that small-scale VHTRs may be considered for international deployment as grid-appropriate variable-scale self-contained systems addressing emerging demands for hydrogen. A Monte Carlo-deterministic analysis methodology has been implemented for coupled design studies of VHTRs with TRUs using the ORNL SCALE 5.1 code system. The developed modeling approach provides an exact-geometry 3D representation of the VHTR core details properly capturing VHTR physics. The discussed studies are being performed within the scope of the U.S. DOE NERI project on utilization of higher actinides (TRUs and partitioned MAs) as a fuel component for extended-life VHTR configurations.
Parametric Assessments of High-Pressure and Low-Pressure Conduction Cooldown Events for the VHTR
Authors: Tsuji Nobumasa , Nakano Masaaki , Richards Matt
Published: July 2009

The Very High Temperature Reactor (VHTR) has been selected by the U.S. as the Generation IV technology for the Next Generation Nuclear Plant (NGNP), and both the U.S. and Japan have been developing VHTR concepts based on a prismatic, block-type core design. For these VHTR concepts, the primary coolant (helium) inlet temperature is expected to be in the range 490°C to 590°C and the outlet temperature is expected to be in the range 850°C to 950°C. Passive safety is one of the fundamental requirements for the VHTR, and the VHTR is designed to be passively safe even during Loss of Coolant Accidents (LOCAs) and Loss of Flow Accidents (LOFAs). For the VHTR, these two transient events are referred to as a Low-Pressure Conduction Cooldown (LPCC) and High-Pressure Conduction Cooldown (HPCC), respectively. During both events, the decay heat is conducted through the graphite to the vessel. The heat is transferred from the vessel by thermal radiation and natural convection to a passive Reactor Cavity Cooling System (RCCS). In this paper, we describe parametric studies of LPCC and HPCC events using a 30-degree sector, 3-dimensional ANSYS model of the VHTR, which includes a detailed radiation exchange model between the RPV and RCCS.
On Eliminating ‘Hot Streaking’ and Stratification in the VHTR Lower Plenum Using Helicoid Inserts
Authors: Rodriguez Sal B. , El-Genk Mohamed S.
Published: July 2009

The helium-cooled high temperature reactor (HTR) and Very High Temperature Reactor (VHTR) designs operate at exit temperatures ranging from 873 – 923 K and 1,123 – 1,223 K, respectively. The high exit temperatures and the low heat capacity of helium require operating at a high flow velocity (> 70 m/s). The high temperature and flow velocity of the helium jets exiting the coolant channels in the prismatic reactor’s lower plate in the lower plenum (LP) cause “hot spots” (“hot streaking”) and stratification. To minimize or eliminate hot streaking and enhance mixing, this work investigated using static, quadruple helicoid inserts at the exit of the coolant channels. The helicoid inserts introduce radial and azimuthal momentum flow components, which with the extensive entrainment and mixing of the surrounding gas, significantly reduce the impingement onto the lower plate, thereby minimizing hot streaking and stratification in the LP. The present analysis was conducted using FUEGO, Sandia National Laboratories’ 3D, finite element, incompressible, reactive flow, massively-parallel code with state-of-the art turbulence models. We used helium at 1,273 K and the dynamic Smagorinsky turbulence model.
Development of Local Heat Transfer and Pressure Drop Models for Pebble Bed High Temperature Gas-Cooled Reactor Cores
Authors: McLaughlin Brian , Worsley Matthew , Stainsby Richard , Grief Andrew , Dennier Ana , MacIntosh Shawn , vanHeerden Eugene
Published: July 2009

This paper describes pressure drop and heat transfer coefficient predictions for a typical coolant flow within the core of a pebble bed reactor (PBR) by examining a representative group of pebbles remote from the reflector region. The three-dimensional steady state flow and heat transfer predictions utilized in this work are obtained from a computational fluid dynamics (CFD) model created in the commercial software ANSYS FLUENT™. This work utilizes three RANS turbulence models and the Chilton-Colburn analogy for heat transfer. A methodology is included in this paper for creating a quality unstructured mesh with prismatic surface layers on a random arrangement of touching pebbles. The results of the model are validated by comparing them with the correlations of the German KTA rules for a PBR.
Advanced High Temperature Gas-Cooled Reactor Systems
Authors: Kato Yasuyoshi
Published: July 2009

Three systems have been proposed for advanced high temperature gas-cooled reactors (HTGRs): a supercritical carbon dioxide (S-CO2) gas turbine power conversion system; a new MicroChannel Heat Exchanger (MCHE); and a once-through-then-out (OTTO) refueling scheme with burnable poison (BP) loading. An S-CO2 gas turbine cycle attains higher cycle efficiency than a He gas turbine cycle due to reduced compression work around the critical point of CO2. Considering temperature lowering at the turbine inlet by 30°C through the intermediate heat exchange, the S-CO2 indirect cycle achieves efficiency of 53.8% at turbine inlet temperature of 820°C and turbine inlet pressure of 20 MPa. This cycle efficiency value is higher by 4.5% than that (49.3%) of a He direct cycle at turbine inlet temperature of 850°C and 7 MPa. A new MCHE has been proposed as intermediate heat exchangers between the primary cooling He loop and the secondary S-CO2 gas turbine power conversion system; and recuperators of the S-CO2 gas turbine power conversion system. This MCHE has discontinuous “S”-shape fins providing flow channels with near sine curves. Its pressure drop is one-sixth reference to the conventional MCHE with zigzag flow channel configuration while the same high heat transfer performance inherits. The pressure drop reduction is ascribed to suppression of recirculation flows and eddies that appears around bend corners of zigzag flow channels in the conventional MCHE. An optimal BP loading in an OTTO refueling scheme eliminates the drawback of its excessively high axial power peaking factor, reducing the power peaking factor from 4.44 to about 1.7; and inheriting advantages over the multi-pass scheme because of the lack of fuel handling and integrity checking systems; and reloading. Because of the power peaking factor reduction, the maximum fuel temperatures are lower than the maximum permissible values of 1250°C for normal operation and 1600°C during a depressurization accident.
Reprocessing Nuclear Fuel Containing Polymers
Authors: Olivier Fiquet , Raynal Boivinet , Pierre Trabuc
Published: July 2009

Organic radiolysis generates enough hydrogen gas to question the safety of radioactive fuel transportation and long-term storage. A safety analysis points out the absolute necessity to get rid of all organic substances in nuclear fuel long-term storage. In the past decades, R and D activities have been producing quantities of rod fuel samples embedded in polymer resins for characterization purposes. Until recently, resin has not been removed from samples and today large sample quantities have to be reprocessed. The “STAR” nuclear facility at CEA Cadarache in France devoted to used fuel stabilization and conditioning, recently decided to implement in the hot cell a particular process to achieve the safety requirements. In order to define a versatile process, efficient for any kind of polymer, thermal treatment has been chosen over a chemical or mechanical process. The definition of this particular thermal treatment must take into account; the hot cell environment, the nuclear safety rules and the behavior of resins. A prototypic furnace has been built for study purposes and thermal cycle validation. Today, the thermal cycle has been defined in two phases as follow: First phase: pyrolysis is used to transform resin into residues and gases. A post gas treatment will be added to the furnace for total gas oxidation. Second phase: Air thermal treatment will achieve the complete residue oxidation and guarantee a hydrogen free product. The final equipment will be available in 2009 for testing and validation cycles with a radioactive free simulator before it is to be implemented in the hot cell in 2010.
Use of Sub-Modeling Techniques to Calculate Peak Stresses in Bolt Head-to-Shank Fillet Radius
Authors: Smith Richard E. , Speicher Stephen J.
Published: July 2009

There is an ever-increasing use of three-dimensional finite element models in the field of structural analysis to simulate structural response of complex geometries. Although these models are effective in simulating gross structural behavior, they are oftentimes not able to include sufficient detail to simulate small structural details where stress concentrations can occur. To overcome this limitation, sub-models can be used to calculate stresses in areas of peak stress. This paper discusses the process involved in calculating peak stresses in bolt head-to-shank interfaces using sub-modeling methods.
Experimental Study of Flow Induced Vibration on Heat Transfer Helical Tube Bundle
Authors: Gao Lixia , Yang Jie , Jiang Zilong , Ma Jianzhong , Li Song
Published: July 2009

In this paper, the Flow Induced Vibration (FIV) tests of heat transfer helical tube bundle have been carried out. The results show that the more large flow velocity in tube becomes, the more high pressure drop of bundle is. The dynamic buckling of bundle does not occur during steady-state operation. The fatigue evaluation has been done according to ASME code and the structure is founded to satisfy the ASME code requirements.
A Study on the Convective Mass Transfer of Nitrogen to Water for a Gas Pressurizing System
Authors: Seo Kyoungwoo , Kim Young-In , Seo Jae-Kwang , Lee Doo-Jeong
Published: July 2009

Mass transfer due to a concentration difference of nitrogen can occur in a nuclear system. Our research work seeks to analyze and understand the mass transfer phenomena of nitrogen in water under the condition of a natural convection using the commercially available CFD computer model, FLUENT 6.3. The maximum solubility was employed to express the boundary condition at an interface between the nitrogen and water. First, the case that nitrogen was transferred to water by only a diffusion was simulated to verify the application of the UDS (User defined scalar) model in FLUENT 6.3 for a mass transfer. Diffusion equation, which was described as a PDE (Partial Differential Equation) with non-homogeneous boundary conditions, was solved and the solved results of the PDE showed a good agreement with those of the FLUENT simulation in the same condition. The same cylinder geometry with that of the diffusion case was used to estimate the convective mass transfer. By the natural convection caused by the thermal boundary condition, the mass transfer of nitrogen had a convection effect. The result of FLUENT 6.3 to compute the convective mass transfer showed that the nitrogen was transferred simultaneously in the entire region by the convection effect and it took about several hours until the mole fraction of nitrogen in the water side reached 50% of the maximum saturated value. The averaged mass transfer coefficient was calculated and compared with the results obtained from the heat and mass transfer analogy. The calculated coefficients showed the lower value than those obtained from the various correlations. When the steam mass transfer toward the gas side was negligible, the pressure drop of the gas side due to the reduced nitrogen caused by a mass transfer was computed using the ideal gas law and the Custom Field Function model in the FLUENT 6.3.
Local and Area-Averaged Flow Structure of Air-Water Two-Phase Flow in a Vertical Annulus
Authors: Ozar Basar , Jeong Jae Jun , Dixit Abhinav , Julia´ Jose Enrique , Hibiki Takashi , Ishii Mamoru
Published: July 2009

The flow structure of gas-liquid two-phase flow has been investigated in a vertical annulus channel. The annulus consisted of a geometry where the inner diameter was 19.1 mm and the outer diameter was 38.1 mm. The total height of the test section was 4.37 m. Experiments were conducted for nineteen inlet flow conditions. These flow conditions covered bubbly, cap-slug, and churn-turbulent flows. The local flow parameters, such as void fraction, interfacial area concentration, and bubble interface velocity, were measured at nine radial positions within the gap of the annulus at z/Dh = 230 of the test section. Radial distributions of these parameters were interpreted in terms of turbulent velocity profile, lift and wall forces. In addition, the local measurements were used to calculate distribution parameter, C0 in drift-flux model, and area averaged interfacial area concentration. Ishii’s (1977) model was modified and a new correlation of C0 was proposed based on the experimentally obtained C0 values. The area-averaged interfacial area concentration (IAC) values were compared with the most widely used models (Ishii and Mishima, 1980; Spore et al., 1983; Hibiki and Ishii, 2002). The advantages and drawbacks of these models were highlighted.
Static Instability in Supercritical Parallel-Channel Systems
Authors: Chatoorgoon Vijay
Published: July 2009

An analytical study of static instability in parallel channels at supercritical pressure is conducted. Until now, primarily the density-wave type has been investigated and reported. This paper derives an analytical expression for the static instability in parallel channels, which lends useful insight. This topic that would be of immense value to the new reactor designs that aim to use supercritical light water on the primary side. The finding is that static instability would unlikely be a problem in a supercritical pressure reactor because of the low inlet core temperature that it would be encountered.
Severe Fuel Damage Experiments With Advanced Cladding Materials to be Performed in the QUENCH Facility (QUENCH-ACM)
Authors: Sepold L. , Große M. , Steinbru¨ck M. , Stuckert J.
Published: July 2009

The QUENCH out-of-pile experiments are part of the Severe Fuel Damage (SFD) program at the Karlsruhe Research Center. They are to investigate the hydrogen source term that results from reflooding an uncovered core of a Light-Water Reactor (LWR) with emergency cooling water. In the QUENCH experimental program Zircaloy-4 was used as standard-type material for rod cladding and grid spacer. Up to the end of 2007, 12 QUENCH experiments have been performed with this type of cladding; two test bundles contained B4C and one AgInCd absorber. One experiment (QUENCH-12) was conducted with Zr1%Nb cladding (VVER-type). Due to the niobium-bearing cladding, the VVER-type test QUENCH-12 could be regarded as a precursor for the upcoming program “QUENCH-ACM” with advanced cladding materials, i.e. M5, Duplex, ZIRLO, to be tested under SFD or BDBA (beyond design basis accident) conditions. These materials were developed for longer operation times in nuclear power reactors and extended burnup. They are optimized regarding their corrosion behavior under operational conditions and were also tested for LOCA (loss of coolant accident) and RIA (reactivity-initiated accident) conditions by the manufacturers. However, there are only very limited data available on the behavior of the new alloys in the SFD/BDBA temperature range, i.e. above 1500 K. The QUENCH-ACM test series has been defined with three experiments, i.e. QUENCH-14 through QUENCH-16. As in the Zircaloy-4 experiments, fuel is represented by ZrO2 pellets. Also, the test section instrumentation will be as usual with thermocouples attached to the cladding, shroud, and cooling jacket at elevations between −50 mm and 1350 mm. The QUENCH-ACM test series is scheduled to be performed in the period of 2008–2010. Test matrix and test bundle arrangements are presented in this paper.
Integrated System CFD Modelling of the Flow Distribution Within a Pebble Bed Modular Reactor
Authors: Huang T. H.
Published: July 2009

The complex hydrodynamic phenomena of a Pebble Bed Modular Reactor (PBMR) are partly attributed to interconnecting flow paths. Modelling the hydrodynamic phenomena provides insight to the reactor over a wide range of operating conditions. The Systems Computational Fluid Dynamics (SCFD) approach combines the advantage of conventional Computational Fluid Dynamics (CFD) modelling in simulating the detailed flow interaction to the more robust solver applied in Flownex thermal-hydraulics software. Although this approach is inherently geometry-dependent, its application in a large network or complex model has seen a reduction in computational speed compared to the conventional CFD modelling approach, without compromising the integrity of the result. This expedites the design-analysis cycle and allows testing of a combination of design changes.