bluuebunny/crossref_metadata_2025 huggingface dataset:
X-Ray Radiography: An Alternative Method for Determining HTR Fuel Particle Coating Layer Thickness and Density
Authors: Banchet Julien , Tisseur David , Guillermier Pierre
Published: July 2009

The HTR TRISO particle consists of a fissile kernel and surrounding layers, whose density and thickness are among the key fuel parameters. Destructive methods i.e. the sink float method and image analysis of ceramography, were developed in the past and are still used to evaluate these particle parameters. Although exhibiting great accuracy, these methods generate effluents and wastes, and are extremely cost/time consuming. In the framework of the AREVA NP HTR R&D program, the development of nondestructive evaluation methods as alternatives to destructive methods is carried out and aims at a new HTR Fuel QC strategy. In this scope, an innovative method was developed to automatically measure particle layer density and thickness from X-Ray Phase Contrast Imaging (PCI). First tested at the European Synchroton Radiation Facility (ESRF), this method was then applied to a custom built industrial demonstrator. Comparisons between the density and thickness values obtained by the developed method and their corresponding values obtained with destructive methods justify progressing to the validation phase. Particle samples were selected among the particle batches that were characterized by destructive methods. Layer density and thickness were determined by the X-Ray based technique on the industrial demonstrator as well as at the ESRF. Correlation levels obtained from this benchmark demonstrated that both parameters can be confidently measured by the developed method. Additionally, it is important to stress that this technique provides the opportunity to directly determine buffer density on finite particles as opposed to the sink float method. Thanks to its accuracy, its rapidity and its absence of waste generation, it is planned to implement the X-Ray thickness and density measurement method on the French lab scale fuel line. It was also decided to enter the characterization work package of the IAEA Coordinated Research Project 6 in order to benchmark the AREVA NP method with foreign techniques and materials.
Air Ingress Phenomena in a Depressurization Accident of the Very-High-Temperature Reactor
Authors: Takeda Tetsuaki
Published: July 2009

The inherent properties of the Very-High-Temperature Reactor (VHTR) facilitate the design of the VHTR with high degree of passive safe performances, compared to other type of reactors. However, it is still not clear if the VHTR can maintain a passive safe function during the primary-pipe rupture accident, or what would be a design criterion to guarantee the VHTR with the high degree of passive safe performances during the accident. The primary-pipe rupture accident is one of the most common of accidents related to the basic design regarding the VHTR, which has a potential to cause the destruction of the reactor core by oxidizing in-core graphite structures and to release fission products by oxidizing graphite fuel elements. It is a guillotine type rupture of the double coaxial pipe at the nozzle part connecting to the side or bottom of the reactor pressure vessel, which is a peculiar accident for the VHTR. This study is to investigate the air ingress phenomena and to develop the passive safe technology for the prevention of air ingress and of graphite corrosion. The present paper describes the influences of a localized natural circulation in parallel channels onto the air ingress process during the primary-pipe rupture accident of the VHTR.
Development of Plant Dynamics Analysis Code for HTGR-Steam Cycle
Authors: Mizokami Yorikata , Saguchi Tetsuo , Minatsuki Isao , Petit Dominique
Published: July 2009

A control and operation method of a High Temperature Gas Cooled Reactor (HTGR) power plant with Steam Cycle has been investigated, and an adequacy of the method has been examined by plant dynamics analyses. In this plant, the steam cycle system with regenerative and reheat cycle is employed, where the heat generated in the reactor is transferred to the secondary coolant (water and steam) through the steam generator (SG), which has the steam turbine and generator, is installed in the secondary system. The reactor and steam cycle system control methods were evaluated. Analytical results through system modeling concluded that the reactor power changes could be effectively achieved by utilizing primary system helium flow rate control, and overall plant system responsiveness and control could be achieved by employing feed water flow control system and turbine governor control system. The plant dynamics analysis code ASURA, which was developed by MHI, can simulate the major components and systems of this plant. It was used for this examination. In order to adequate of the control and operation method, some cases of plant dynamics analysis by ASURA was conducted in this study.
Modelling Silver: Evaluation of German Experience
Authors: van der Merwe Hanno , Olivier Dirk
Published: July 2009

For direct cycle gas cooled high temperature reactor designs, operating conditions may be limited as a result of excessive maintenance dose rates caused by the Ag-110m source term on the turbine. It is therefore important to accurately predict silver release from fuel during reactor operation. Traditionally diffusion models were used to derive transport parameters from limited irradiation testing of fuel materials and components. Best estimates for all applicable German fuel irradiation tests with defendable uncertainty ranges were never derived. However, diffusion theory and current parameters cannot account for all irradiation and heat-up test results, and for some tests, it appears unacceptably conservative. Other transport mechanisms have been suggested, and alternative calculation models are being considered. In this paper the applicable German irradiation test results are evaluated with a classic diffusion model as well as an alternative model called the Molecular Vapour transport Release (MVR) model. New transport models and parameters for silver in fuel materials are suggested and compared.
Results of Neutronic Benchmark Analysis for a High Temperature Reactor of the GT-MHR Type
Authors: Boyarinov Viktor F. , Bryzgalov Vladimir I. , Davidenko Vladimir D. , Fomichenko Peter A. , Glushkov Evgeny S. , Gomin Evgeny A. , Gurevich Mikhail I. , Kodochigov Nikolay G. , Marova Elena V. , Mitenkova Elena F. , Novikov Nikolay V. , Osipov Sergey L. , Sukharev Yury P. , Tsibulsky Viktor F. , Yudkevich Mikhail S.
Published: July 2009

The paper presents a description of benchmark cases, achieved results, analysis of possible reasons of differences of calculation results obtained by various neutronic codes. The comparative analysis is presented showing the benchmark–results obtained with reference and design codes by Russian specialists (WIMS-D, JAR-HTGR, UNK, MCU, MCNP5-MONTEBURNS1.0-ORIGEN2.0), by French specialists (APOLLO2, TRIPOLI4 codes), and by Korean specialists (HELIOS, MASTER, MCNP5 codes). The analysis of possible reasons for deviations was carried out, which was aimed at the decrease of uncertainties in calculated characteristics. This additional investigation was conducted with the use of 2D models of a fuel assembly cell and a reactor plane section.
Providing Thermal-Hydraulic Boundary Conditions to the Reactor Code TINTE Through a Flownex-TINTE Coupling
Authors: Marais Deon , Greyvenstein Gideon P.
Published: July 2009

TINTE is a well established reactor analysis code which models the transient behaviour of pebble bed reactor cores but it does not include the capabilities to model a power conversion unit (PCU). This raises the issue that TINTE cannot model full system transients. One way to overcome this problem is to supply TINTE with time-dependant thermal-hydraulic boundary conditions which are obtained from PCU simulations. This study investigates a method to provide boundary conditions for the nuclear code TINTE during full system transients. This was accomplished by creating a high level interface between the systems CFD code Flownex and TINTE. An indirect coupling method is explored whereby characteristics of the PCU are matched to characteristics of the nuclear core. This method eliminates the need to iterate between the two codes. A number of transients are simulated using the coupled code and then compared against stand-alone Flownex simulations. The coupling method introduces relatively small errors when reproducing mass flow, temperature and pressure in steady state analysis, but become more pronounced when dealing with fast thermal-hydraulic transients. Decreasing the maximum time step length of TINTE reduces this problem, but increases the computational time.
Microbial Treatment of Irradiated Graphite for Separation of Radioisotope 14C From Bulk Graphite 12C
Authors: Dunzik-Gougar Mary Lou , van Ravenswaay Francis , Kuczynski Leszek , Slabber Johan M. F.
Published: July 2009

The Pebble Bed Modular Reactor is being developed in South Africa. Important for PBMR implementation is a viable strategy for waste management. Irradiated graphite from fuel and structural components is too voluminous for practical treatment with traditional higher level waste methods and too radioactive to recycle. To clean the graphite of radionuclides, a two-step process is being pursued: (1) non-carbon radionuclides (activation products, fission products and actinides) are removed on an elemental basis by a chemical or microbial process. (2) 14C requires separation at an isotopic level, which would be impractical with established methods (gaseous diffusion or centrifuge). PBMR is investigating a method of isotope separation using biofractionation. Preliminary experiments indicate that microorganisms do separate radioactive 14C from stable 12C. An aqueous slurry of 14C-spiked, powdered graphite was “fed” to the microbes for 15–18 hours. The microbes initially contained only background levels of 14C, i.e. orders of magnitude less than the slurry. In post-experiment analyses, a sample of the microbes was found to contain approximately twice the amount of 14C present in the bulk slurry material. Experiments are underway to further quantify and verify these results, which indicate distinct microbial processing mechanisms for 14C and 12C. The most current results will be presented.
Systematic Qualification of the Coupled Neutron Transport and Thermal-Hydraulics Code DORT-TD/THERMIX
Authors: Tyobeka Bismark , Pautz Andreas , Ivanov Kostadin
Published: July 2009

In order to present credible results in nuclear design and safety analysis, computer codes must adhere to stringent qualification procedures imposed by nuclear licensing authorities. Such procedures form the basis for a quality assured verification and validation process. This is particularly true for advanced nuclear systems of Generation IV type, where little licensing experience exists as well as little or no plant data is available. Qualification of nuclear design and analysis codes can be achieved in various ways, namely: comparison of results from a code with results from another code i.e. code to code benchmarking; comparison of results from a given code with experimental results, i.e. code to experiment benchmarking; comparison of results from a given code with operational plant data; and finally, comparison of the results of a given code with known analytical solutions. In this paper, a systematic qualification of the coupled neutron transport and thermal hydraulics code DORT-TD/THERMIX is presented. As part of developing this coupled code to the level where it can be used as an independent tool by both designers of pebble-bed High-Temperature Gas-cooled Reactors (HTGRs) and regulators, an effort has been made to verify the coupling scheme as well as the validity of application for this code package. At these initial stages a code to code comparison has been adopted as the qualification method of choice. This is done for both steady-state and transient benchmark problems, ranging from simplified to detailed models. As shown in the results section, all benchmarks have been successfully recalculated and generally show good to very good agreement with the “reference” solutions.
Euratom Research and Related Strategy in the Field of Nuclear Systems and Its Support to V/HTR as Part of the Contribution to GIF
Authors: Manolatos Panagiotis , Van Goethem Georges , Deffrennes Marc , Hugon Michel , Webster Simon
Published: July 2009

Energy is at the top of the political agenda. Energy technology, including nuclear, has a role to play in addressing the associated challenges. The Euratom research programme in the area of nuclear fission is contributing to the development of innovative reactor concepts that are more sustainable, intrinsically safe and proliferation resistant, economic and capable of co-generating both electricity and process heat. The need for such Generation-IV reactors is also stressed in the EU’s recently published Strategic Energy Technology Plan, and the development of these advanced systems is the focus of the Sustainable Nuclear Energy Technology Platform established in September 2007, which is at the cornerstone of the EU’s strategy in nuclear R&D and brings together all the key European nuclear research and industrial players in this field. The paper provides a summary of the present situation and prospects in these areas, with an emphasis on V/HTR technology.
Development of the CAPP Code Based on the Finite Element Method for the Analysis of VHTR Cores
Authors: Lee Hyun Chul , Jo Chang Keun , Noh Jae Man
Published: July 2009

In this study, we developed a neutron diffusion equation solver based on the finite element method for CAPP code. Three types of triangular finite elements and five types of rectangular depending on the order of the shape functions were implemented for 2-D application. Ten types of triangular prismatic finite elements and seventeen types of rectangular prismatic finite elements were also implemented for 3-D application. Two types of polynomial mapping from the master finite element to a real finite element were adopted for flexibility in dealing with complex geometry. They are linear mapping and iso-parametric mapping. In linear mapping, only the vertex nodes are used as the mapping points. In iso-parametric mapping, all the nodal points in the finite element are used as the mapping points, which enables the real finite elements to have curved surfaces. For the treatment of spatial dependency of cross-sections in the finite elements, three types of polynomial expansion of the cross-sections in the finite elements were implemented. They are constant, linear, and iso-parametric cross-section expansions. The power method with the Wielandt acceleration technique was adopted as the outer iteration algorithm. The BiCGSTAB algorithm with the ILU (Incomplete LU) decomposition preconditioner was used as the linear equation solver in the inner iteration. The neutron diffusion equation solver developed in this study was verified against two well known benchmark problems, IAEA PWR benchmark problem and OECD/NEA PBMR400 benchmark problem. Results of numerical tests showed that the solution converged to the reference solution as the finite elements are refined and as the order of the finite elements increases. Numerical tests also showed that the higher order finite element method is much efficient than lower order finite element method or finite difference method.
PBMR Spent Fuel Bulk Dry Storage Heat Removal
Authors: de Wet G. J. , Dent C.
Published: July 2009

A low decay heat (implying Spent Fuel (SF) pebbles older than 8–9 years) bulk dry storage section is proposed to supplement a 12-tank wet storage section. Decay heat removal by passive means must be guaranteed, taking into account the fact that dry storage vessels are under ground and inside the building footprint. Cooling takes place when ambient air (drawn downwards from ground level) passes on the outside of the 6 tanks’ vessel containment (and gamma shielding), which is in a separate room inside the building, but outside PBMR building confinement and open to atmosphere. Access for loading / unloading of SF pebbles is only from the top of a tank, which is inside PBMR building confinement. No radioactive substances can therefore leak into atmosphere, as vessel design will take into account corrosion allowance. In this paper, it is shown (using CFD (Computational Fluid Dynamics) modelling and analytical analyses) that natural convection and draught induced flow combine to remove decay heat in a self-sustaining process. Decay heat is the energy source, which powers the draught inducing capability of the dry storage modular cell system: the more decay heat, the bigger the drive to expel heated air through a higher outlet and entrain cool ambient air from ground level to the bottom of the modular cell.
Spectral Zones Determination for Pebble Bed Reactor Cores Using Diffusion Theory Spectral Indices
Authors: Mphahlele Ramatsemela , Ougouag Abderrafi M. , Ivanov Kostadin N. , Gougar Hans D.
Published: July 2009

A practical methodology is developed for the determination of spectral zones in Pebble Bed Reactors (PBR). The methodology involves the use of spectral indices based on few-group diffusion theory whole core calculations. In this work a spectral zone is defined as made up of a number of nodes whose characteristics are collectively similar and that are assigned the same few-group diffusion constants. Therefore, spectral indices that reflect the physical behaviors of interest can be used to characterize said behaviors within each zone and thus to identify and distinguish the spectral zones. Several plausible spectral indices have been investigated in this work. Special emphasis and focus was placed on the trend or behavior of the spectral index at various positions along the radial and axial dimensions in the core. The ratio of group-wise surface currents to total surface fluxes, has been used to successfully identify spectral zone boundaries. A plot of the absolute value of this ratio versus position in the reactor exhibits a series of minima and maxima points. These extrema correlate with regions of significant spectral changes, and therefore are identified as plausible spectral zone boundaries.
Perspectives for the French R&D Program for High and Very High Temperature Reactors
Authors: Yvon Pascal , Hittner Dominique , Delbecq Jean-Michel
Published: July 2009

A R&D programme has been launched addressing the needs of the development of an indirect cycle flexible modular HTR operating at 850°C for electricity generation and/or heat production for industrial processes. In the frame of this program, several significant technical challenges required to demonstrate the viability and performance of the system have been successfully addressed. Design and safety analysis needed the development of computational tools, therefore reactor physics, and thermo-fluid dynamics codes have been developed and are now in the process of being validated in the frame of international code-to-code and code to experiment benchmarks. Most importantly, the performance of the HTR/VHTR fuel identified as TRISO-coated particle must prove to be excellent in operating as well as accidental conditions. A manufacturing and quality control process has been developed and now fuel qualification based on irradiation and heating safety tests is being prepared on the basis of irradiation programs in France and in the frame of the GENERATION IV International Forum (GIF) as well as the development of fuel behaviour models including performance data, failure particle prediction and long-term integrity of the coating. Material and component technologies have been investigated in normal and accident conditions for V/HTR objectives. Significant progress has been made for vessel structures and reactor core structural elements. Major challenges still lie ahead for plate type compact intermediate heat exchangers, especially at temperatures above 850°C, but an alternative solution with helical tubes is also being developed. In order to demonstrate that materials have adequate performance over long service life under impure helium environment and constraints, the research programme focuses on microstructural and mechanical property data, long-term irradiation behaviour, corrosion, modelling and codification of design rules as well as qualification of components in representative helium test loops. The potential of this type of reactor for higher performances in terms of fuel burn-up and temperature (VHTR objective) has been explored, in particular for application to hydrogen production. The major research axes on hydrogen production technologies include the development and optimization of high temperature electrolysis and thermo-chemical water splitting processes such as sulphur/iodine or hybrid sulphur. Alternative thermo-chemical hydrogen generation processes operating at lower temperatures are also investigated. This paper addresses the R&D work performed since 2001 and the future work anticipated until 2012, where decisions about a demonstrator could be made at a European level within the Sustainable Nuclear Energy Technological Platform (SNE-TP). This program is strongly connected to the Euratom Framework Programmes as well as to GIF.
The Backend of the Fuel Cycle of HTR/VHTR Reactors
Authors: Grambow B. , Abdelouas A. , Guittonneau F. , Vandenborre J. , Fachinger J. , von Lensa W. , Bros P. , Roudil D. , Perko J. , Marivoet J. , Sneyers A. , Millington D. , Cellier F.
Published: July 2009

For various countries, the direct disposal of high level nuclear fuel wastes is a key option for the backend of the fuel cycle. For HTR/VHTR reactors this is assumed for the introductory phase of this reactor system. However, closed fuel cycles or a separation of spent coated-particles from the graphite moderator and specific treatment, conditioning and disposal of these waste streams are also possible. In the European Community project “RAPHAEL”, fuel waste performance is going to be studied in depth, including post-irradiation fuel characterization, analysis of the stability and failure mechanism of coatings and of fuel kernels and overall performance of waste packages with compact fuel and/or only with fuel particles in geological disposal environments. Different confinement matrices for separated fuel particles (vitrification, SiC, ZrO2) have been adapted to limit release of radionuclides into groundwater at low temperatures over geological time spans. The investigations are limited to Low-Enriched Uranium (LEU) fuel with uranium oxide and uranium oxycarbide kernels that will allow higher burn-up, but may be more susceptible to leaching.
A HR-SEM Investigation of the Microstructure of PBMR Test Fuel Particles
Authors: van der Berg N. G. , Malherbe J. B. , Botha A. J.
Published: July 2009

There is currently renewed interest in high temperature nuclear fission power reactors. The Pebble Bed Modular Reactor (PBMR) is one of several high temperature gas-cooled reactors being investigated by researchers. The South African design of the PBMR is based on the original German design, with the fuel particles (called TRISO particles) being small multilayer spheres.
Dry Cooling Towers for GT-MHR
Authors: Baxi C. B.
Published: July 2009

Due to problems with the availability and the price of water, and the concerns relating to adverse environmental effects of wet cooling systems, the need for water conserving cooling systems has been increasing. Presently, dry cooling accounts for over 30,000 MWe of capacity in more than 30 countries. GT-MHR is specially suited for use of dry cooling due to 1) high efficiency, 2) high heat rejection temperatures and 3) large temperature difference between the turbine inlet and heat rejection temperatures. Higher efficiency means the amount of energy rejected to the cooling per MWe is less. The majority of heat is rejected in precooler and intercooler at helium temperature of more than 100 °C. This results in higher temperature difference for heat rejection. Also due to large temperature difference between the turbine inlet and heat rejection temperatures, changes in ambient temperature have a smaller effect on overall thermal efficiency. Preliminary evaluation shows that pure dry cooling is economical for GT-MHR for water cost of more than 0.8$/m3 and power cost of 3.5 c/kWh. A combination of dry and wet cooling can reduce large percentage of the water use without affecting the efficiency.
Influence of Temperature on the Micro- and Nanostructures of Experimental PBMR TRISO Coated Particles: A Comparative Study
Authors: van Rooyen Isabel J. , Neethling Jan H. , Mahlangu Johannes
Published: July 2009

The PBMR fuel consists of TRISO Coated Particles (CPs) in a graphite matrix. The three layer system, IPyC-SiC-OPyC, forms the primary barrier to fission product release, with the SiC layer acting as the main pressure boundary of the particle. The containment of fission products inside the CPs is however a function of the operating temperature and microstructure of the SiC layer. During accident conditions, the CPs will reach higher temperatures than normal operating conditions. The Fuel Design department of PBMR is therefore investigating various characteristics of the SiC layer, especially nano characteristics at variant conditions. The integrity of the interface between the SiC and Inner PyC layers is also important for fission product retention and therefore interesting TEM images of this region of the experimental PBMR TRISO particles will be shown. Transmission electron microscope (TEM) images of the microstructure of TRISO coated particles of three different experimental batches after annealing will be discussed. Particles annealed at 1980°C for 1 hour revealed that the inner PyC layer debonded from the SiC layer. Changes observed in the diffraction rings are evidence that the PyC structure is becoming organized or anisotropic. The SiC layer, on the other hand, did not show any changes as a result of annealing. Only the cubic 3C-SiC phase was observed for a limited number of grains analyzed. The nano hardness and elasticity measurements of the three test batches were done using a CSM Nano Hardness Tester. These results are compared to indicate possible differences between the 1 hour and 5 hour annealing time as a function of annealing temperature from 1000°C to 1980°C. The variation of hardness and elasticity as a function of temperature for the three experimental batches are identified and discussed. This preliminary TEM investigation and nano hardness measurements have contributed new knowledge about the effect of high temperature annealing on the microstructure of TRISO CPs produced by PBMR.
South African Safety Assessment Framework for the Pebble Bed Modular Reactor
Authors: Joubert Jean , Kohtz Norbert , Coe Ian
Published: July 2009

It is planned to construct a first of a kind Pebble Bed Modular Reactor (PBMR) in South Africa. A need has been recognized to accompany the licensing process for the PBMR with independent safety assessments to ensure that the safety case submitted by the applicant complies with the licensing requirements of the NNR. At the HTR 2006 Conference, the framework and major challenges on safety assessment that the South African National Nuclear Regulator (NNR) faces in developing and applying appropriate strategies and tools were presented. This paper discusses the current status of the various NNR assessment activities and describes how this will be considered in the NNR Final Report on the PBMR Safety Case. The traditional safety assessment process has been adapted to take into account the developmental nature of the project. By performing safety assessments, the designer and applicant must ensure that the design as proposed for construction and as-built meets the safety requirements defined by the regulatory framework. The regulator performs independent safety assessments, including independent analyses in areas deemed safety significant and potentially safety significant. The developmental nature of the project also led to the identification of a series of regulatory assessment activities preceding the formal assessment of the safety case. Besides an assessment of the resolution of Key Licensing Issues which have been defined in an early stage of the project and are discussed in /1/, these activities comprise the participation in an SAR Early Intervention Process, the execution of a regulatory HAZOP and the development of a regulatory assessment specification for the formal assessment of the safety case. This paper briefly describes these activities and their current status. During the last two years, significant progress was made with the development or adjustment of tools for the independent analysis by the regulator of the steady state core design, of the transient neutronic/thermal hydraulic behaviour of the reactor, of fission product release from the fuel elements, and of activation of fuel matrix and graphite impurities.
Development of a Sulfur Dioxide Depolarized Electrolyzer for Hydrogen Production Using the Hybrid Sulfur Thermochemical Process
Authors: Summers William A. , Steimke John L. , Hobbs David T. , Colon-Mercado Hector R. , Gorensek Maximilian B.
Published: July 2009

The Hybrid Sulfur Process is a leading candidate among the thermochemical cycles being developed to use heat from advanced nuclear reactors to produce hydrogen via watersplitting. It has the potential for high efficiency, competitive cost of hydrogen, and it has been demonstrated at a laboratory scale to confirm performance characteristics. The major developmental issues with the HyS Process involve the design and performance of a sulfur dioxide depolarized electrolyzer, the key component for conducting the electrochemical step in the process. This paper will discuss the development program and current status for the SDE being conducted at the Savannah River National Laboratory.
Post Irradiation Testing of High Temperature Reactor Spherical Fuel Elements Under Accident Conditions
Authors: Freis D. , Bottomley D. , Ejton J. , de Weerd W. , Kostecka H. , Toscano E. H.
Published: July 2009

A new furnace for accident condition testing of spherical High Temperature Reactor (HTR) fuel elements has been installed and is now operating in the Hot Cells of the Institute for Transuranium Elements (ITU) Karlsruhe. The recent apparatus was constructed on the basis of a former development by Forschungszentrum Ju¨lich (FzJ) [Schenk 1988] where it was named Ku¨FA, the German acronym for cold finger apparatus. In a preceding publication [Toscano 2004] the general concept and details of the device were described. The present paper reports on the first operation under hot conditions, the calibration of the fission gas measurement and of the efficiency of the cold finger, which is used to plate out solid fission products. Finally the results of fission product release and analysis of two heating tests on two fuel elements from the HFR K6 irradiation experiment [Nabielek 1993] are presented and discussed.
PBMR Desalination Options: An Economic Study
Authors: de Bruyn Riaan , Hannink Ryan , Kuhr Reiner , Van Ravenswaay Jan P. , Zervos Nick , Bhagat Kishna
Published: July 2009

The Pebble Bed Modular Reactor (PBMR), under development in South Africa, is an advanced helium-cooled graphite moderated high-temperature gas-cooled nuclear reactor. The heat output of the PBMR is primarily suited for process applications or power generation. In addition, various desalination technologies can be coupled to the PBMR to further improve the overall efficiency and economics, where suitable site opportunities exist. Several desalination application concepts were evaluated for both a cogeneration configuration as well as a waste heat utilization configuration. These options were evaluated to compare the relative economics of the different concepts and to determine the feasibility of each configuration. The cogeneration desalination configuration included multiple PBMR units producing steam for a power cycle, using a back-pressure steam turbine generator exhausting into different thermal desalination technologies. These technologies include Multi-Effect Distillation (MED), Multi-Effect Distillation with Thermal Vapor Compression (MED-TVC) as well as Multi-Stage Flash (MSF) with all making use of extraction steam from backpressure turbines. These configurations are optimized to maximize gross revenue from combined power and desalinated water sales using representative economic assumptions with a sensitivity analysis to observe the impact of varying power and water costs. Increasing turbine back pressure results in a loss of power output but a gain in water production. The desalination systems are compared as incremental investments. A standard MED process with minimal effects appears most attractive, although results are very sensitive with regards to projected power and water values. The waste heat utilization desalination configuration is based on the current 165 MWe PBMR Demonstration Power Plant (DPP) to be built for the South African utility Eskom. This demonstration plant is proposed at the Koeberg Nuclear site and utilizes a direct, single shaft recuperative Brayton Cycle with helium as working fluid. The Brayton Cycle uses a pre-cooler and inter-cooler to cool the helium before entering the low-pressure compressor (LPC) and the high-pressure compressor (HPC) respectively. The pre-cooler and intercooler rejects 218 MWt of waste heat at 73°C and 63°C, respectively. This waste heat is ideally suited for some low temperature desalination processes and can be used without negative impact on the power output and efficiency of the nuclear power plant. These low temperature processes include Low Temperature Multi-Effect Distillation (LT-MED) as well Reverse Osmosis (RO) with pre-heated water. The relative economics of these design concepts are compared as add-ons to the PBMR-DPP and the results include a net present value (NPV) study for both technologies. From this study it can be concluded that both RO as well LT-MED provide water at reasonable production rates, although a final study recommendation would be based on site and area specific requirements.
HTGR Component Test Facility (CTF) Feasibility Study for NGNP
Authors: van Ravenswaay Jan P. , Holtzhausen Jacques , van der Merwe Jaco , Olivier Kobus , du Bruyn Riaan , Haasbroek Andries , Fox Marius
Published: July 2009

The Next Generation Nuclear Plant (NGNP) Project is a US-based initiative led by Idaho National Laboratories to demonstrate the viability of using High Temperature Gas-Cooled Reactor (HTGR) technology for the production of high temperature steam and/or heat for applications such as heavy oil recovery, process steam/cogeneration and hydrogen production. A key part of the NGNP Project is the development of a Component Test Facility (CTF) that will support the development of high temperature gas thermal-hydraulic technologies as applied in heat transport and heat transfer applications in HTGRs. These applications include, but are not limited to, primary and secondary coolants, direct cycle power conversion, co-generation, intermediate, secondary and tertiary heat transfer, demonstration of processes requiring high temperatures as well as testing of NGNP specific control, maintenance and inspection philosophies and techniques. The feasibility of the envisioned CTF as a development and testing platform for components and systems in support of the NGNP was evaluated. For components and systems to be integrated into the NGNP full scale or at least representative size tests need to be conducted at NGNP representative conditions, with regards to pressure, flow rate and temperature. Typical components to be tested in the CTF include heat exchangers, steam generators, circulators, valves and gas piping. The Design Data Needs (DDNs), Technology Readiness Levels (TRLs) as well as Design Readiness Levels (DRLs) prepared in the pre-conceptual design of the NGNP Project and the NGNP lifecycle requirements were used as inputs to establish the CTF Functional and Operating Requirements (F&ORs). The existing South African PBMR test facilities were evaluated to determine their current applicability or possible modifications to meet the F&ORs of the CTF. Three concepts were proposed and initial energy balances and layouts were developed. This paper will present the results of this CTF study and the ongoing efforts to establish the CTF.
Preliminary Development of a Unified Viscoplastic Constitutive Model for Alloy 617 With Special Reference to Long Term Creep Behavior
Authors: Sham T.-L. , Walker Kevin P.
Published: July 2009

The expected service life of the Next Generation Nuclear Plant is 60 years. Structural analyses of the Intermediate Heat Exchanger (IHX) will require the development of unified viscoplastic constitutive models that address the material behavior of Alloy 617, a construction material of choice, over a wide range of strain rates. Many unified constitutive models employ a yield stress state variable which is used to account for cyclic hardening and softening of the material. For low stress values below the yield stress state variable these constitutive models predict that no inelastic deformation takes place which is contrary to experimental results. The ability to model creep deformation at low stresses for the IHX application is very important as the IHX operational stresses are restricted to very small values due to the low creep strengths at elevated temperatures and long design lifetime. This paper presents some preliminary work in modeling the unified viscoplastic constitutive behavior of Alloy 617 which accounts for the long term, low stress, creep behavior and the hysteretic behavior of the material at elevated temperatures. The preliminary model is presented in one-dimensional form for ease of understanding, but the intent of the present work is to produce a three-dimensional model suitable for inclusion in the user subroutines UMAT and USERPL of the ABAQUS and ANSYS nonlinear finite element codes. Further experiments and constitutive modeling efforts are planned to model the material behavior of Alloy 617 in more detail.
Public Acceptance of HTGR Technology
Authors: Hannink Ryan , Kuhr Reiner , Morris Tony
Published: July 2009

Nuclear energy projects continue to evoke strong emotional responses from the general public throughout the world. High Temperature Gas-Cooled Reactor (HTGR) technology offers improved safety and performance characteristics that should enhance public acceptance but is burdened with demonstrating a different set of safety principles. This paper summarizes key issues impacting public acceptance and discusses the importance of openly engaging the public in the early stages of new HTGR projects. The public gets information about new technologies through schools and universities, news and entertainment media, the internet, and other forms of information exchange. Development of open public forums, access to information in understandable formats, participation of universities in preparing and distributing educational materials, and other measures will be needed to support widespread public confidence in the improved safety and performance characteristics of HTGR technology. This confidence will become more important as real projects evolve and participants from outside the nuclear industry begin to evaluate the real and perceived risks, including potential impacts on public relations, branding, and shareholder value when projects are announced. Public acceptance and support will rely on an informed understanding of the issues and benefits associated with HTGR technology. Major issues of public concern include nuclear safety, avoidance of greenhouse gas emissions, depletion of natural gas resources, energy security, nuclear waste management, local employment and economic development, energy prices, and nuclear proliferation. Universities, the media, private industry, government entities, and other organizations will all have roles that impact public acceptance, which will likely play a critical role in the future markets, siting, and permitting of HTGR projects.
VHTR-Based Systems for Autonomous Co-Generation Applications
Authors: Tsvetkov Pavel V. , Ames David E. , Alajo Ayodeji B. , Lewis Tom G.
Published: July 2009

As highly efficient advanced nuclear systems, Generation IV Very High Temperature Reactors (VHTR) can be considered in a variety of configurations for electricity generation and process heat applications. Simultaneous delivery of electricity, low-temperature process heat (for potable water production, district heating, etc.) and high temperature process heat (for hydrogen production, etc.) by a single cogeneration system offers unique deployment options as “all-in-one” power stations. This paper is focused on the VHTR-based systems for autonomous co-generation applications. The analysis is 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. It accounts for system performance characteristics including VHTR physics features, control options and energy conversion efficiencies. Utilization of TRUs in VHTRs is explored to stabilize in-core fuel compositions (core self-stabilization) leading to extended single-batch OTTO (Once-Through-Then-Out) modes of operation without intermediate refueling.
“Used Fuel” Vectors and Waste Minimization Strategies for VHTRs Operating Without Refueling
Authors: Tsvetkov Pavel V. , Lewis Tom G. , Alajo Ayodeji B. , Ames David E.
Published: July 2009

Generation IV Very High Temperature Reactors (VHTRs) are well-known for their flexibility with respect to feasible fuel cycle options. In this paper, the LEU- and TRU-fueled VHTR configurations are analyzed accounting for their capabilities to attain an extended single-batch OTTO (Once-Through-Then – Out) mode of operation without intermediate refueling. The requirement of waste minimization is imposed as one of the design constraints defining possible system configurations and deployment strategies. The resulting “used fuel” vectors are examined considering anticipated disposal options as well as viability of fuel reprocessing. 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 presented analysis is focused on prismatic block core concepts for VHTRs. It is 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.
Radioactive Waste Management Plan for PMBR (Pty) Ltd Fuel Plant
Authors: Makgae Mosidi E.
Published: July 2009

The Pebble Bed Modular Reactor (Pty) Ltd Fuel Plant (PFP) radioactive waste management plan caters for waste from generation, processing through storage and possible disposal. Generally, the amount of waste that will be generated from the PFP is Low and Intermediate Level Waste. The waste management plan outlines all waste streams and the management options for each stream. It also discusses how the Plant has been designed to ensure radioactive waste minimisation through recycling, recovery, reuse, treatment before considering disposal. Compliance to the proposed plan will ensure compliance with national legislative requirements and international good practice. The national and the overall waste management objective is to ensure that all PFP wastes are managed appropriately by capitalising on processes that minimise, reduce, recover and recycle without exposing employees, the public and the environment to unmitigated impacts. Both International Atomic Energy Agency (IAEA) and Department of Minerals and Energy (DME) principles act as a guide in the development of the strategy in order to ensure international best practice, legal compliance and ensuring that the impact of waste on employees, environment and the public is as low as reasonably achievable. The radioactive waste classification system stipulated in the Radioactive Waste Management Policy and Strategy 2005 will play an important role in classifying radioactive waste and ensuring that effective management is implemented for all waste streams be it gaseous, liquid or solid waste.
The Component Test Facility: A National User Facility for Testing of High Temperature Gas-Cooled Reactor (HTGR) Components and Systems
Authors: Balls Vondell J. , Duncan David S. , Austad Stephanie L.
Published: July 2009

The Next Generation Nuclear Plant (NGNP) and other High-Temperature Gas-cooled Reactor (HTGR) Projects require research, development, design, construction, and operation of a nuclear plant intended for both high-efficiency electricity production and high-temperature industrial applications, including hydrogen production. During the life cycle stages of an HTGR, plant systems, structures and components (SSCs) will be developed to support this reactor technology. To mitigate technical, schedule, and project risk associated with development of these SSCs, a large-scale test facility is required to support design verification and qualification prior to operational implementation. As a full-scale helium test facility, the Component Test facility (CTF) will provide prototype testing and qualification of heat transfer system components (e.g., Intermediate Heat Exchanger, valves, hot gas ducts), reactor internals, and hydrogen generation processing. It will perform confirmation tests for large-scale effects, validate component performance requirements, perform transient effects tests, and provide production demonstration of hydrogen and other high-temperature applications. Sponsored wholly or in part by the U.S. Department of Energy, the CTF will support NGNP and will also act as a National User Facility to support worldwide development of High-Temperature Gas-cooled Reactor technologies.
A Methodology to Examine the Long-Term Energy, Environmental and Economic Benefits of Advanced Nuclear Technologies
Authors: Bhatt Vatsal , Friley Paul , Lee John , Reisman Ann
Published: July 2009

Nuclear energy is on the verge of a possible nuclear renaissance, driven by the simultaneous growing global demand for energy and a growing awareness of the need to combat climate change by lowering atmospheric carbon emissions. Widespread deployment of advanced nuclear technologies will require nuclear energy to be competitive in the energy market. Given the long lead-time and high cost of building capital-intensive nuclear facilities, it is important to perform analyses up-front to gain insight into what combinations of economic, environmental, technical and policy conditions will be required for nuclear to play a significant future role. This paper describes an analytical approach that can be used to define those conditions where nuclear energy could contribute. The methodology, using the MARKAL model, is both rich in technical detail and yet conceptually transparent. It is flexible, easy to modify for the needs of a particular analysis and can vary parameters of interest to address uncertainties in data and in future conditions. It can also examine the impacts of present or potential future government policies on the ultimate deployment of nuclear technologies over time. The example results provided in the paper illustrate some issues of interest to the nuclear and energy communities that can be addressed using MARKAL.