bluuebunny/crossref_metadata_2025 huggingface dataset:
Ceramic Composites for Near Term Reactor Application
Authors: Snead Lance L. , Katoh Yutai , Windes William E. , Shinavski Robert J. , Burchell Timothy D.
Published: July 2009

Currently, two composites types are being developed for incore application: carbon fiber carbon composite (CFC), and silicon carbide fiber composite (SiC/SiC.) Irradiation effects studies have been carried out over the past few decades yielding radiation-tolerant CFC’s and a composite of SiC/SiC with no apparent degradation in mechanical properties to very high neutron exposure. While CFC’s can be engineered with significantly higher thermal conductivity, and a slight advantage in manufacturability than SiC/SiC, they do have a neutron irradiation-limited lifetime. The SiC composite, while possessing lower thermal conductivity (especially following irradiation), appears to have mechanical properties insensitive to irradiation. Both materials are currently being produced to sizes much larger than that considered for nuclear application. In addition to materials aspects, results of programs focusing on practical aspects of deploying composites for near-term reactors will be discussed. In particular, significant progress has been made in the fabrication, testing, and qualification of composite gas-cooled reactor control rod sheaths and the ASTM standardization required for eventual qualification.
The PBMR Fuel Plant: Proven Technology in Advanced Safety Environment
Authors: Bra¨hler Georg , Froschauer K. , Welbers P. , Boyes D.
Published: July 2009

The PBMR Fuel Plant (PFP), to be constructed at the Pelindaba site near Johannesburg will fuel the first South African Pebble Bed Modular Reactor. The qualification of the PBMR fuel shall be based on past experience with fuel which was produced in the German NUKEM/HOBEG plant and irradiated in the German AVR reactor. Accordingly, the PFP must produce the same fuel as the German plant did, and consequently, the design of the PFP has in essence to be a copy of the NUKEM/HOBEG plant. As a reminder this plant had been operated in accordance with the German regulatory rules which were defined in the years 1970/80. Since then, the requirements with regard to radiological protection, criticality safety and emission control have been significantly tightened, and of course the PFP must be designed in accordance with the most advanced international norms and standards. The implications which follow from these two potentially conflicting requirements, as defined above, are highlighted, and technical solutions are presented. Hence, the change from administrative criticality safety control to technical control, i.e. the application of safe geometry as far as possible, and the introduction of technical solutions for the remaining safe mass regime will be described. A lot of equipment in the Kernel area and in the recycling areas needed to be redesigned in safe geometry. The sensitive processes for Kernel Calcining, for the Coating and the Overcoating remain under safe mass regime, but the safety against criticality is completely independent from staff activities and based on technical measures. A new concept for safe storage of large volumes of Uranium-containing liquids has been developed. Also, the change from relatively open handling of Uranium to the application of containment enclosures wherever release of radioactivity into the room atmosphere is possible, will be addressed. This change required redesign of all process steps requiring the handling of dry Uranium oxides and uncoated Kernels. Finally, the introduction of processes for the near-total recycling of Uranium and chemicals, as well as for decontamination and purification of liquid and gaseous effluents will be presented. These processes were not available from NUKEM/HOBEG fuel facility and needed to be developed now, also following the above mentioned requirements, with respect to criticality safety and radiological protection.
Idaho National Laboratory Experimental Research in High Temperature Electrolysis for Hydrogen and Syngas Production
Authors: Stoots Carl M. , O’Brien James E. , Herring J. Stephen , Condie Keith G. , Hartvigsen Joseph J.
Published: July 2009

The Idaho National Laboratory (Idaho Falls, Idaho, USA), in collaboration with Ceramatec, Inc. (Salt Lake City, Utah, USA), is actively researching the application of solid oxide fuel cell technology as electrolyzers for large scale hydrogen and syngas production. This technology relies upon electricity and high temperature heat to chemically reduce a steam or steam / CO2 feedstock. Single button cell tests, multi-cell stack, as well as multi-stack testing has been conducted. Stack testing used 10 × 10 cm cells (8 × 8 cm active area) supplied by Ceramatec and ranged from 10 cell short stacks to 240 cell modules. Tests were conducted either in a bench-scale test apparatus or in a newly developed 5 kW Integrated Laboratory Scale (ILS) test facility. Gas composition, operating voltage, and operating temperature were varied during testing. The tests were heavily instrumented, and outlet gas compositions were monitored with a gas chromatograph. The ILS facility is currently being expanded to ∼15 kW testing capacity (H2 production rate based upon lower heating value).
A Study on an Evaluation of Indirect Cycle Plant Systems for HTGR
Authors: Minatsuki Isao , Oyama Sunao , Mizokami Yorikata , Ballot Bernard
Published: July 2009

In the world now, several types of indirect system concept have been investigated for the High Temperature Gas cooled Reactor power plant (HTGR). From a point of optimization of HTGR, it is important to investigate and to compare their power conversion systems from a technical and an economical view point. In the first step of this study, an indirect steam cycle (ID-SC), an indirect gas turbine cycle (ID-GT), an indirect gas turbine combined cycle (ID-CCGT) and a direct gas turbine cycle (D-GT) has been chosen as the systems to be compared. The followings are chosen items for comparison analysis: a) Plant efficiency; b) Amount of commodities (which can estimate capital cost); c) Flexibility of reactor core design; d) Technical issues to be developed; e) Compatibility with hydrogen production system, etc. And for the second step, as the system optimization study among the selected system, sensitiveness to plant efficiency by changing the inlet and the outlet temperature of reactor core has been investigated from an economical and plant efficiency point of view.
HTR Fuel Integrity With Electromagnetic, Vision and Radiographic Nondestructive Evaluation Methods
Authors: Guillermier Pierre , Banchet Julien , Tisseur David , Hermosilla Lara Se´bastien , Bivert Marc , Piriou Marc
Published: July 2009

In order to ensure HTR fuel qualification, as well as reactor safety, particles need to satisfy a set of specifications including particle integrity. To achieve this goal, AREVA NP has been engaged for several years in a R&D program aiming at the development of innovative industrial non destructive evaluation methods for HTR fuel as alternatives to destructive methods. After investigating a number of potential techniques, development has been focused on vision and eddy currents, both aiming at crack detection. High resolution Phase Contrast X-Ray imaging was also studied for structural defects characterization. For all these techniques, besides the development of HTR fuel dedicated control methods, equipment and probes were specifically designed, tested and optimized thanks to experiments conducted on real and artificial flaws, yielding for some of the methods to potential industrialization and quality control performed over 100% of the fuel production.
New Methods for HTR Fuel Waste Management
Authors: Guittonneau Fabrice , Abdelouas Abdesselam , Grambow Bernd , Dialinas Manoe¨l , Cellier Franc¸ois
Published: July 2009

Considering the need to reduce waste production and greenhouse emissions by still keeping high energy efficiency, various 4th generation nuclear energy systems have been proposed. As far as graphite moderated reactors are concerned, one of the key issues is the large volumes of irradiated graphite encountered (1770 m3 for fuel elements and 840 m3 for reflector elements during the lifetime (60 years) of a single reactor module [1]). With the objective to reduce volume of waste in the HTR concept, it is very important to be able to separate the fuel from low level activity graphite. This requires to separate TRISO particles from the graphite matrix with the sine qua non condition to not break TRISO particles in case of future embedding of particles in a matrix for disposal. According to National Regulatory Systems, in case of limited graphite waste production or of short duration HTR projects (e.g. in Germany), direct disposal without separation is acceptable. Nevertheless, in case of large scale deployment of HTR technology, such approach is not economical and sustainable. Previous attempts in graphite management (furnace, fluidised bed and laser incinerations and encapsulation matrices) dealt with graphite matrix only. These are the reasons why we studied the management of irradiated compact-type fuel element. We simulated the presence of fuel in the particles by using ZrO2 kernels. Compacts with ZrO2 TRISO particles were manufactured by AREVA NP. Two original methods have been studied. First, we tested high pressure jet to erode graphite and clean TRISO particles. Best erosion rate reached about 0.18 kg/h for a single nose ending. Examination of treated graphite showed a mixture of undamaged TRISO particles, particles that have lost the outer pyrolytic carbon layer and ZrO2 kernels. Secondly, we studied the thermal shock method by immerging successively graphite into liquid nitrogen and hot water to cause fracturing of the compact. This produced particles and graphite fragments with diameter ranging from several centimetres to less than 500 μm. This relatively simple and economic method may potentially be considered as a pretreatment step and be coupled with other method(s) before reprocessing and recycling for example.
Fission-Product Behaviour During Irradiation of TRISO-Coated Particles in the HFREU1bis Experiment
Authors: de Groot Sander , Dubourg Roland , Bakker Klaas , Kissane Martin , Barrachin Marc
Published: July 2009

The irradiation experiment HFR-EU1bis, coordinated by the European Joint Research Centre – Institute for Energy, was performed in the High Flux Reator (HFR) at Petten to test five spherical HTR fuel pebbles of former German production with TRISO coated particles in conditions beyond the specifications of current HTR reactor designs (central temperature of 1250°C). In this paper, the behaviour of the fission products (FPs) and kernel micro-structure evolution during the test are investigated. While FP behaviour is a key issue for potential source term evaluation it also determines the evolution of the oxygen potential in the oxide kernel which in turn is important for formation of carbon oxides (amoeba effect and pressurization). Fission-gas release from the kernel can induce additional mechanical loading and finally some FPs (Ag, Cs, Sr) might alter the mechanical integrity of the coatings. This study is based on postirradiation examinations (ceramography + EPMA) performed both on UO2 kernels and on coatings. Significant evolutions of the kernel as a function of temperature are shown (grain structure, porosity, size of metallic inclusions). The quality of the ceramography results allows characteristics of the intergranular bubbles in the kernel (and estimation of swelling) to be determined. Remarkable results considering FP release from the kernel have been observed and will be presented. Examples are the significant release of Cs out of the kernel as well as Pd, whereas Zr remains trapped. Mo and Ru are mainly incorporated in metallic precipitates. These observations are interpreted and mechanisms for FP and micro-structural evolutions are proposed. These results are coupled to the results of calculations performed with the mechanistic code MFPR (Module for Fission Product Release) and the thermodynamic database MEPHISTA (Multiphase Equilibria in Fuels via Standard Thermodynamic Analysis). The effect of high flux rate and high temperature on fission gas behaviour, grain size evolution and kernel swelling are discussed. In addition, solid-FP behaviour (Cs, Mo, Zr, Ba, Sr) is discussed in connection with the evolution of kernel oxygen potential and evolution of the pressure of carbon oxides. The paper intends to be exemplary on how the combination of post-irradiation examination results and fuel modelling increases fundamentally the understanding of HTR fuel behaviour.
RAPHAEL-FT & Generation IV PYCASSO-I Irradiation
Authors: de Groot Sander , Basini Virginie , Guillermier Pierre , Bakker Klaas , Sawa Kazuhiro , Lee Young-Woo , Escleine Jean-Michel , Perez Marc , Ueta Shohei , Kim Bong-Goo
Published: July 2009

Within the Raphael (V)HTR 6th framework EU-program, the PYCASSO experiments have been devised to investigate coating behaviour under irradiation. Samples have been included from CEA (France), JAEA (Japan) and KAERI (Republic of Korea), which makes this irradiation a real Generation IV effort. The experiment is a separate effect test, where the influence of fuel (coating corrosion or micro structural change due to fission products), thermal gradients, and variation in coating microstructure and dimensions have been minimized by the use of dummy kernels (Al2O3 and ZrO2), high conductivity particle holder material combined with low energy production of the kernels, and strict (fabrication) quality control and selection procedures respectively. The purpose of the experiment is threefold for the partners involved: - for CEA to determine the behaviour of pyrocarbon under irradiation, especially the interaction of pyrocarbon swelling and creep with SiC coating layers. The results will be used to validate and improve HTR fuel performance modelling. - for JAEA to investigate the behaviour of ZrC coatings, which have been successfully manufactured, but require post-irradiation investigation and characterization. - for KAERI to determine the influence of fabrication of pyrocarbon layers with different densities on the behaviour under irradiation. The paper will go into more detail on the goals to be achieved by the different partners. The PYCASSO-I irradiation is performed in the High Flux Reactor (HFR) in Petten, The Netherlands. The experiment accommodates temperature regions of 900, 1000 and 1100°C, and contains 76 separate particle sample holders. The PYCASSO-I irradiation is a completely new design and will be described in detail, including the route from the concept definition via feasibility studies, fabrication and assembly, up to the irradiation, which took only 1, 5 year. At the time of the conference, the PYCASSO-I irradiation will be finished and a full evaluation of the irradiation will be presented. Additionally, the future post irradiation examination planned for the PYCASSO-I samples and the details of the PYCASSO-II irradiation will be outlined.
IHX Design and Development for the Next Generation Nuclear Plant
Authors: Penfield Scott R. , Greyvenstein Renee , Rittenhouse Phillip L. , Nash James
Published: July 2009

This paper summarizes the results of a conceptual design study addressing the design and technology development requirements for a high-temperature intermediate heat exchanger (IHX) for the Next Generation Nuclear Plant (NGNP). Results of the study confirmed the incentives for compact heat exchangers and suggested new IHX configurations that provide for maintainability at the heat transfer module level. Scoping analyses provided encouragement that IHX life would not be limited by creep or fatigue effects, given the PBMR NGNP Heat Transport System architecture and operating conditions. However, corrosion rates implied by existing data are troubling for thin sections, and improved characterization of environmental effects was identified as a high priority for technology development.
Pebble Bed Modular Reactor Fuel Enrichment Discrimination Using Delayed Neutrons
Authors: Skoda R. , Rataj J. , Uher J.
Published: July 2009

The Pebble Bed Modular Reactor (PBMR) is a helium-cooled, graphite-moderated high temperature nuclear power reactor which utilise fuel in form of spheres that are randomly loaded and continuously circulated through the core until they reach their prescribed end-of-life burn-up limit. When the reactor is started up for the first time, the lower-enriched start-up fuel is used, mixed with graphite spheres, to bring the core to criticality. As the core criticality is established and the start-up fuel is burned-in, the graphite spheres are progressively removed and replaced with more start-up fuel. Once it becomes necessary for maintaining power output, the higher enriched equilibrium fuel is introduced to the reactor and the start-up fuel is removed. During the initial run of the reactor it is important to discriminate between the irradiated startup fuel and the irradiated equilibrium fuel to ensure that only the equilibrium fuel is returned to the reactor. There is therefore a need for an on-line enrichment discrimination device that can discriminate between irradiated start-up fuel spheres and irradiated equilibrium fuel spheres. The device must also not be confused by the presence of any remaining graphite spheres. Due to it’s on-line nature the device must accomplish the discrimination within tight time limits. Theoretical calculations and experiments show that Fuel Enrichment Discrimination based on delayed neutrons detection is possible. The paper presents calculations and experiments showing viability of the method.
HTR Pebble Fuel Burnup Experimental Benchmark
Authors: van Heek Aliki , Charpin Florence , van der Marck Steven , Wolters Jorrit , Trakas Christos , Aguiar Luis , Bomboni Eleonora , Cerullo Nicola , Lomonaco Guglielmo , Freis Daniel , Frybort Jan
Published: July 2009

The HTR pebble fuel experiment HFR EU1bis was irradiated in the High Flux Reactor, Petten, The Netherlands, in 2004 and 2005. It consisted of five fuel pebbles from the German HTR program (GLE4 type, UO2 fuel, 16.75% enrichment) and six minisamples (UO2 fuel, 9.75% enrichment). Its instrumentation included three flux monitor sets. The experiment was loaded in a REFA-170 rig, surrounded by a strongly moderating filler element. The central fuel temperature was held at 1250°C during the irradiation. In the framework of the European RAPHAEL project, Post Irradiation Examination (PIE) has been done at NRG in Petten, The Netherlands and at JRC ITU in Karlsruhe, Germany. In Petten, flux monitor analysis has been done, whereas in Karlsruhe, a quantitative evaluation of γ-emitters was used to make a burn-up determination. A benchmark description based on this experiment has been written by NRG. Until now, five RAPHAEL project participants have modeled the experiment, each with their own neutronics code system. Participating codes are three versions of MONTEBURNS (MCNP with ORIGEN), MURE/MCNP and OCTOPUS (MCNP with FISPACT). The pebble burnup and isotopic inventories (Bq/gram initial HM) of selected fission products and actinides in the fuel pebble samples are both calculated and determined by gamma spectrometry, mass spectrometry and ion chromatography by JRC-ITU. Additionally, two participants calculated the flux monitor activities that were measured by NRG. A burnup measurement of 11.0 % FIMA by gamma spectrometry could be confirmed by calculation. Differences between the various modeling approaches and the experimental burn-up determination will be discussed.
Chromium Activity Measurements in Nickel Based Alloys for Very High Temperature Reactors: Inconel 617, Haynes 230 and Model Alloys
Authors: Gosse´ S. , Alpettaz T. , Chatain S. , Gue´neau C. , Rouillard F. , Cabet C.
Published: July 2009

The alloys Haynes 230 and Inconel 617 are potential candidates for the intermediate heat exchangers (IHX) of (V)-HTR reactors. The behaviour under corrosion of these alloys by the (V)-HTR coolant (impure helium) is an important selection criterion because it defines the service life of these components. At high temperature, the Haynes 230 is likely to develop a chromium oxide on the surface. This layer protects from the exchanges with the surrounding medium and thus confers certain passivity on metal. At very high temperature, the initial microstructure made up of austenitic grains and coarse intra and intergranular M6C carbide grains rich in W will evolve. The M6C carbides remain and some M23C6 richer in Cr appear. Then, carbon can reduce the protective oxide layer. Then, the alloy loses its protective coating and can corrode quickly. Experimental investigations were performed on these nickel based alloys under an impure helium flow [1]. To predict the surface reactivity of chromium under impure helium, it is necessary to determine its chemical activity in a temperature range close to the operating conditions of the heat exchangers (T ≈ 1273 K). For that, high temperature mass spectrometry measurements coupled to multiple effusion Knudsen cells are carried out on several samples: Haynes 230, Inconel 617 and model alloys 1178, 1181, 1201. This coupling makes it possible thermodynamic equilibrium to be obtained between the vapour phase and the condensed phase of the sample. The measurement of the chromium ionic intensity (I) of the molecular beam resulting from a cell containing an alloy provides the values of partial pressure according to the temperature. This value is compared to that of the pure substance (Cr) at the same temperature. These calculations provide thermodynamic data characteristic of the chromium behaviour in these alloys. These activity results call into question those previously measured by Hilpert [2], largely used in the literature.
Development of Materials for a High Temperature Gas Cooled Reactor in Korea
Authors: Park Ji Yeon , Kim Dae Whan , Kim Dong Jin , Kim Eoun Sun , Kim Hong Pyo , Kim Sung Ho , Kim Weon-Ju , Kim Woo Gon , Ryu Woo Seog , Kim Young Do
Published: July 2009

The very high temperature reactor (VHTR) materials R&D started as an objective of the development of key technologies for a nuclear hydrogen production funded by the Korea government in 2006. We are performing materials R&D for 5 components: a reactor pressure vessel (RPV), a reactor/process intermediate heat exchanger (IHX) and hot gas duct, a control rods component, a reflector and support structures in the core region and reactor materials for the sulfur-iodine (SI) process. The scopes of our works are focused on a material screening/selection and qualification, codifications of a high temperature structural design to a very high temperature region, and to support the licensing of a system design, material characterizations and database establishments. Current target materials are modified 9Cr-1Mo, alloy 617, graphite, ceramic fiber reinforced composite, Fe-Si and SiC.
Corrosion Issues of HTR Structural Metallic Materials
Authors: Cabet Celine , Rouillard Fabien , Girardin Gouenou , Blat Martine
Published: July 2009

Cooling helium of HTRs is expected to contain a low level of impurities: oxidizing gasses and carbon-bearing species. Reference structural materials for pipes and heat exchangers are chromia-former nickel base alloys — typically alloys 617 and 230 — and, as is generally the case in any high temperature process, their long term corrosion resistance relies on the growth of a surface chromium-oxide that can act as a barrier against corrosive species. This implies that the HTR environment must allow for oxidation of these alloys to occur, while it remains not too oxidizing against in-core graphite. First, studies on the surface reactivity under various impure helium containing low partial pressures of H2, H2O, CO and CH4 show that alloys 617 and 230 oxidize in many atmospheres from intermediate temperatures up to 890–970°C, depending on the exact gas composition. However when heated above a critical temperature, the surface oxide becomes unstable: it was demonstrated that at the scale/alloy interface the surface oxide interacts with the carbon from the material. These investigations have established an environmental area that promotes oxidation. When expose in oxidizing HTR helium, alloys 617 and 230 actually develop a sustainable surface scale over thousands of hours. On the other hand if the scale is destabilized by reaction with the carbon, the oxide is not protective anymore and the alloy surface interacts with gaseous impurities. In the case of CH4-containg atmospheres, this causes rapid carburization in the form of precipitation of coarse carbides on the surface and in the bulk. Carburization was shown to induce an extensive embrittlement of the alloys. In CH4-free helium mixtures, alloys decarburize with a global loss of carbon and dissolution of the pre-existing carbides. As carbides take part to the alloy strengthening at high temperature, it is expected that decarburization impacts the creep properties. Carburization and decarburization degrade rapidly the alloy properties and thus result in an unacceptably high risk on the material integrity at high temperature. Therefore, the purification system shall control the gas composition in order to make this unique helium atmosphere compatible with the in-core graphite as well as with structural materials. This paper reviews the data on the corrosion behavior of structural material in HTR and draws some conclusion on appropriate helium chemistry regarding the material compatibility at high temperature.
Turbulent Scalar Flux Models for the Mixing of Hot Gases in the Swirling Flow of High-Temperature Reactor Lower Plenums
Authors: von Lavante Dominik , Laurien Eckart
Published: July 2009

With recent progress in high-temperature pebble-bed reactor programs research focus has started to include more ancillary engineering issues. One very important aspect for the realisability is the mixing of hot and colder helium in the reactor lower plenum. Under nominal operating conditions, depending on core design, the temperature of hot gas leaving the core can locally differ up to 210° C. Due to material limitations, these temperature differences have to be reduced to at least ±15° C. Several reduced-size air experiments have been performed on this problem, but their applicability to modern commercially sized reactors is not certain. With the rise in computing power CFD simulations can be performed in addition, but advanced turbulence modeling is necessary due to the highly swirling and turbulent nature of this flow. The presented work uses the geometry of the German HTR-Modul which consists of an annular mixing channel and radially arranged ribs. Using the commercial CFD code ANSYS CFX, we have made detailed analyses of the complex 3D vortical flow phenomena within this geometry. Several momentum transport turbulence models, e.g. the classical k-e model, advanced two-equation models and Reynolds-Stress Models were compared with respect to their accuracy for this particular flow. In addition, the full set of turbulent scalar flux transport equations was implemented for modeling the three components of turbulent transport of enthalpy seperately and were compared with the standard turbulent Prandtl number approach. As expected from previous work in related fields of turbulence modeling, the differences in predicting the mixing performance between models were significant. Only the full Reynolds-Stress model coupled with the scalar flux equations was able to reproduce the experimentally observed reduction of mixing efficiency with increasing Reynolds number. The correct scaling of mixing efficiencies demonstrates that the utilized turbulence models are able to reproduce the physics of the underlying flow. Hence they could be employed for the scaling and optimization of the lower plenum geometry. The results also showed that the original geometry used for the HTR-Modul is insufficient to provide adequate mixing, and that hence a not sufficiently mixed coolant for future reactor designs might be an issue. Based on this work, an optimization for future lower plenum geometries has become feasible.
Neutronic Characterization of the SAFARI-1 Material Testing Reactor
Authors: Makgopa B. M. , Belal M. , Strydom W. J.
Published: July 2009

This work presents a neutronic analysis of the core in the South African Fundamental Atomic Research Installation (SAFARI-1) for future Pebble Bed Modular Reactor (PBMR) fuel irradiation experiments. Monte Carlo simulation of the core with and without the rig has been performed. The results show a negligibly small reactivity worth of the rig, which is expected, due to the small amount of heavy metal loading in the pebble and the low fuel enrichment. This effect will be further investigated when the rig is extended to include more than one fuel pebble. Results further show perturbations in the neutron and photon flux as well as the power distribution in core position B6. A 50% thermal neutron flux depression is observed in position B6 due to the insertion of the rig. A 60% increase in axial photon heating values is also observed in position B6. The neutron and photon flux and power distributions in the other incore irradiation positions (D6 and F6) are slightly affected by the insertion of this rig. Fluxes and power distributions in positions D6 and F6 will be studied in detail when they are loaded with isotope production rigs.
A Technique to Simulate a Tube Break in a High-Pressure Gas/Cooling Water Heat Exchanger
Authors: van Antwerpen H. J. , Mulder E. J.
Published: July 2009

The gas cycles of most High Temperature Gas-Cooled Reactors (HTR’s) reject heat to water at some stage. In the helium/water heat exchangers of HTR’s with direct Brayton cycles, the helium is usually at a much higher pressure than the water. If the pressure boundary between the helium and the water fails inside the heat exchanger, the effect on the rest of the water system has to be established in order to do a proper system design. This can be done most efficiently by using a system simulation code, however, very few system simulation codes has the capability to do gas/liquid interface tracking as required for this problem. This study describes a calculation method with which a gas/liquid heat exchanger tube rupture can be calculated in a simulation code without interface tracking. The course of events after tube rupture is described and appropriate calculation models derived. A mathematical model for a pressure relief valve (PRV) was also created. The calculation models were implemented in the system simulation software Flownex and used to study a tube rupture on a 5000kPa helium/water heat exchanger. The assembled calculation network solved stable and within reasonable time. The simulation provided insight into the course of events following the tube break. It was shown that the acceleration of water out of the helium cooler, by choked-flow helium, caused the main pressure pulses during the event. The maximum pressure in the water loop occurs on the opposite side of the helium cooler due to constructive interference of the initial pressure wave with itself. It was also shown that by changing only pipe lengths, the system could become prone to severe oscillations after a tube rupture event.
XRD and SEM/EDS Study of the Phase Distribution in the UO2 and Carbon Reactional System at High Temperature
Authors: Boinski F. , Dugne O. , Gosse´ S. , Brackx E. , Gue´neau C.
Published: July 2009

Several fissile kernels are considered for the TRISO particles of (Very) – High Temperature Reactors (V)-HTR. Considering uranium as the fissile nucleus, the basic chemical composition of the TRISO fuel is always made of UO2 which can be added by some UC2 carbide. The high level operating temperature of this fuel implies to determine the products formed by the interactions between these two previous compounds. Some UO2 kernels embedded in black carbon were heated in the 1250–1400°C temperature range in order to determine the kinetics of gaseous species formation [1] and the relative stability of the oxide and carbide phases. After High Temperature Mass Spectrometry (HTMS) experiments, the products formed during the interaction between uranium oxide (UO2) and carbon powders were characterized by various global and punctual analysis methods. The XRD diagram showed the presence of UO2 and UC phases. The contrasts of density observed by SEM in the Quadrant Back-Scattering Detector (QBSD) mode also allowed to highlight both oxide and carbide phase distributions within the TRISO kernels. During SEM observations, some particles showed particular profiles resulting from “non-uniform” reactional mechanisms as already described by Lindemer [2]. In other singular cases, the interaction between UO2 and carbon led to the formation of the UC phase in the middle of the kernel, the UO2 phase remaining at the outside part. Complementary EDS analyses confirmed the results on both oxide and carbide phases. By considering the interference energies between the K-ray of carbon and the N-ray of uranium, the study of the ray intensities consolidated the contrast distributions observed in SEM. Thanks to these results, some assumptions are also advanced concerning the dissolution of oxygen in the UC crystalline structure.
Proposal for an International Experimental Pebble Bed Reactor
Authors: Serfontein Dawid , Mulder Eben , Teuchert Eberhard
Published: July 2009

HTRs, both prismatic block fuelled and pebble fuelled, feature a number of uniquely beneficial characteristics that will be discussed in this paper. In this paper the construction of an international experimental pebble bed reactor is proposed, possible experiments suggested and an invitation extended to interested partners for co-operation in the project. Experimental verification by nuclear regulators in order to facilitate licensing and the development of a new generation of reactors create a strong need for such a reactor. Suggested experiments include: • Optimized incineration of waste Pu in a pebble bed reactor: The capability to incineration pure reactor grade plutonium by means of ultra high burn-up in pebble bed reactors will be presented at this conference in the track on fuel and fuel cycles. This will enable incineration of the global stockpile of separated reactor grade Pu within a relatively short time span; • Testing of fuel sphere geometries, aimed at improving neutron moderation and a decrease in fuel temperatures; • Th/Pu fuel cycles: Previous HTR programs demonstrated the viability of a Th-232 fuel-cycle, using highly enriched uranium (HEU) as driver material. However, considerations favoring proliferation resistance limit the enrichment level of uranium in commercial reactors to 20%, thereby lowering the isotopic efficiency. Therefore, Pu driver material should be developed to replace the HEU component. Instead of deploying a (Th, Pu)O2 fuel concept, the proposal is to use the unique capability offered by pebble bed reactors in deploying separate Th- and Pu-containing pebbles, which can be cycled differently; • Testing of carbon-fiber-carbon (CFC) structures for in-core or near-core applications, such as guide tubes for reserve shutdown systems, thus creating the possibility to safely shutdown reactors with increased diameter; • Development of very high temperature reactor components for process heat applications; • Advanced decay heat removal systems e.g. design specific air flow channels, or heat pipe designs external to the reactor pressure vessel; • Development of a plutonium fuelled peaking reactor with the proposed duel cycle; • A radial coolant flow pattern with increased power output; • Testing of carbon-fiber-carbon (CFC) core barrel applications. The design will facilitate ease of licensing by sacrificing performance in favor of safety and employing redundant defense-in-depth safety systems. Speedy licensing is therefore expected. The economic model will be based on a commercial expedition of the agreed experimental value to collaborating participants. Target costs will be minimized by exploiting known technology only and by utilizing off-the-shelf components as far as possible.
South African Licensing Framework for the Pebble Bed Modular Reactor
Authors: Bester P. , Hill T.
Published: July 2009

The electrical utility in South Africa (Eskom) plan to construct a first of a kind Pebble Bed Modular Reactor (PBMR). It has been recognized that there is a need to adapt the licensing process for the PBMR to ensure that credible and effective licensing process be developed and implemented for this technology. This paper will briefly outline the regulatory framework within South Africa, explain the licensing process adopted and present the challenges that the South African National Nuclear Regulator (NNR) was facing in developing and implementing the licensing process and how these are being addressed. The paper will discuss the update of the regulatory framework and the gaps identified in terms of regulatory requirements needed for such a project. The scope of the regulatory assessment for the licensing of the PBMR is based on the licensing requirements and criteria defined by the NNR in regulatory documents that expand on the current legislative requirements. In addition guidance is provided on selected issues in regulatory guidance documents and position papers. The requirements comprise, besides the general requirements to respect good engineering practice and the ALARA and defense-in-depth principle, specific risk criteria and radiation dose limits. These are categorized for normal operation and operational occurrences as well as for design basis events and beyond design basis events for workers and the public. Additional requirements and recommendations are stipulated by the NNR on safety important areas like quality and safety management, qualification of the nuclear fuel and the core structures, core design, verification and validation of computer codes, source term analysis and others. Selected NNR Position Papers have been developed to elaborate and provide further clarification on NNR requirements. For preparation of the PBMR safety case so-called Key Licensing Issues have been defined and agreed with the applicant. Discussions relating to these Key Licensing Issues allow important nuclear safety aspects identified for the PBMR demonstration plant to be clarified in advance of the safety case submittal.
Condition Monitoring of High Temperature Components With Sub-Sized Samples
Authors: Hoffelner Wolfgang , Pouchon Manuel , Samaras Maria , Froideval Annick , Chen Jiachao
Published: July 2009

Advanced nuclear plants are designed for long-term operation in quite demanding environments. Limited operation experience with the materials used in such plants necessitate a reliable assessment of damage and residual life of components. Non-destructive condition monitoring of damage is difficult, if not impossible for many materials. Periodic investigation of small samples taken from well defined locations in the plant could provide an attractive tool for damage assessments. This paper will discuss possibilities of using very small samples taken from plant locations for complementary condition monitoring. Techniques such as micro/nano-indentation, micropillar compression, micro bending, small punch and thin strip testing can be used for the determination of local mechanical properties. Advanced preparation techniques such as focused ion beam (FIB) allow the preparation of samples from these small volumes for micro-structural analyses with transmission electron microscope (TEM) and advanced X-ray synchrotron techniques. Modeling techniques (e.g. dislocation dynamics DD) can provide a quantitative link between microstructure and mechanical properties. Using examples from ferritic oxide dispersion strengthened materials the DD approach is highlighted to understand component life assessments.
NGNP Process Heat Utilization: Liquid Metal Phase Change Heat Exchanger
Authors: Sabharwall Piyush , Patterson Mike , Utgikar Vivek , Gunnerson Fred
Published: July 2009

One key long-standing issue that must be overcome to fully realize the successful growth of nuclear power is to determine other benefits of nuclear energy apart from meeting the electricity demands. The Next Generation Nuclear Plant (NGNP) will most likely be producing electricity and heat for the production of hydrogen and/or oil retrieval from oil sands and oil shale to help in our national pursuit of energy independence. For nuclear process heat to be utilized, intermediate heat exchange is required to transfer heat from the NGNP to the hydrogen plant or oil recovery field in the most efficient way possible. Development of nuclear reactor-process heat technology has intensified the interest in liquid metals as heat transfer media because of their ideal transport properties. Liquid metal heat exchangers are not new in practical applications. An important rationale for considering liquid metals as the working fluid is because of the higher convective heat transfer coefficient. This explains the interest in liquid metals as coolant for intermediate heat exchange from NGNP. The production of electric power at higher efficiency via the Brayton Cycle, and hydrogen production, requires both heat at higher temperatures and high effectiveness compact heat exchangers to transfer heat to either the power or process cycle. Compact heat exchangers maximize the heat transfer surface area per volume of heat exchanger; this has the benefit of reducing heat exchanger size and heat losses. High temperature IHX design requirements are governed in part by the allowable temperature drop between the outlet of NGNP and inlet of the process heat facility. In order to improve the characteristics of heat transfer, liquid metal phase change heat exchangers may be more effective and efficient. This paper explores the overall heat transfer characteristics and pressure drop of the phase change heat exchanger with Na as the heat exchanger coolant. In order to design a very efficient and effective heat exchanger one must optimize the design such that we have a high heat transfer and a lower pressure drop, but there is always a tradeoff between them. Based on NGNP operational parameters, a heat exchanger analysis with the sodium phase change is presented to show that the heat exchanger has the potential for highly effective heat transfer, within a small volume at reasonable cost.
Fission Product Monitoring of TRISO Coated Fuel for the Advanced Gas Reactor-1 Experiment
Authors: Scates Dawn M. , Hartwell John K. , Walter John B. , Drigert Mark W. , Harp Jason M.
Published: July 2009

The US Department of Energy has embarked on a series of tests of TRISO-coated particle reactor fuel intended for use in the Very High Temperature Reactor (VHTR) as part of the Advanced Gas Reactor (AGR) program. The AGR-1 TRISO fuel experiment, currently underway, is the first in a series of eight fuel tests planned for irradiation in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The AGR-1 experiment reached a peak compact averaged burn up of 9% FIMA with no known TRISO fuel particle failures in March 2008. The burnup goal for the majority of the fuel compacts is to have a compact averaged burnup greater than 18% FIMA and a minimum compact averaged burnup of 14% FIMA. At the INL the TRISO fuel in the AGR-1 experiment is closely monitored while it is being irradiated in the ATR. The effluent monitoring system used for the AGR-1 fuel is the Fission Product Monitoring System (FPMS). The FPMS is a valuable tool that provides near real-time data indicative of the AGR-1 test fuel performance and incorporates both high-purity germanium (HPGe) gamma-ray spectrometers and sodium iodide [NaI(Tl)] scintillation detector-based gross radiation monitors. To quantify the fuel performance, release-to-birth ratios (R/B’s) of radioactive fission gases are computed. The gamma-ray spectra acquired by the AGR-1 FPMS are analyzed and used to determine the released activities of specific fission gases, while a dedicated detector provides near-real time count rate information. Isotopic build up and depletion calculations provide the associated isotopic birth rates. This paper highlights the features of the FPMS, encompassing the equipment, methods and measures that enable the calculation of the release-to-birth ratios. Some preliminary results from the AGR-1 experiment are also presented.
The Pebble Bed Modular Reactor (PBMR) as a Source of High Quality Process Heat for Sustainable Oil Sands Expansion
Authors: Morris Antony , Kuhr Reiner , McKinnell James
Published: July 2009

Bitumen extraction, processing and upgrading consume large quantities of natural gas for production of steam, hot water and hydrogen. Massive expansion of bitumen production is planned in response to increasing energy demands, higher oil prices, and the desire for energy security. The Pebble Bed Modular Reactor (“PBMR”) in its Process Heat configuration supports applications that are likely to compete in a cost effective and environmentally sustainable way with natural gas fired boilers for the production of steam for bitumen recovery. Technology development work to produce hydrogen using this nuclear technology is also underway. The PBMR has the benefit of size, passive nuclear safety characteristics (encompassing Generation IV safety principles), high reliability, high temperature process heat (750–950°C) in a modular design suited to the oil sands industry.
Use of High-Temperature Reactor Heat in Refineries, Underground Refining, and Biorefineries for Liquid Fuels Production
Authors: Forsberg Charles
Published: July 2009

The production of liquid fuels (gasoline, diesel, and jet fuel) from hydrocarbons or biomass is energy intensive. For example, the thermal energy input into U.S. refineries is approximately equal to the thermal energy output of the nation’s nuclear power plants. The yield of liquid fuels per barrel of oil or ton of biomass can be increased if nuclear energy provides the thermal heat necessary for conversion of such feedstocks into liquid fuels. This allows the hydrocarbons and biomass that would have been burnt for the production of heat to be used as additional feedstocks for production of additional liquid fuels. Simultaneously, the carbon dioxide emissions from production facilities are reduced. The use of heat from high-temperature reactors would increase liquid fuels production by 10 to 30% per ton of hydrocarbon or biomass feedstock with corresponding reductions in greenhouse gas releases. The maximum temperature of heat to be supplied is generally less than 700°C to avoid thermal decomposition of the hydrocarbons or biomass. The temperature requirements, heat requirements, and the ultimate market size for these different applications of high-temperature heat are described.
Investigation of the Impact of Temperature on Hydrogen Production Cost From Advanced Water Splitting Technologies
Authors: Bolthrunis Charles O. , Allen Daniel , Goff Karl , Summers William , Lahoda Edward
Published: July 2009

One of the key technology challenges in the development of water splitting technologies is the requirement for high temperature process heat. High-Temperature Gas-Cooled Reactors (HTGRs) can supply this heat, but challenges multiply as the reactor outlet temperature, and therefore the maximum process temperature rises. A reasonable implementation strategy for applying HTGRs to these technologies would be to begin with a reactor outlet and a maximum process temperature that is achievable with today’s technology and increase those temperatures in stages as improved technology emerges. This paper investigates what those temperatures should be in the first commercial demonstration by examining the effect of these temperatures on the cost of production of hydrogen. Parameters investigated include the fundamental thermodynamic limits of each technology, reaction kinetics, materials of construction cost, process complexity, component expected life, and availability. Based on this study, comparisons are made between the leading water splitting technologies and the advantages and disadvantages of each are explained.