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)O 2 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.

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.

Since the 1980s, the ASME Code has made numerous improvements in elevated-temperature structural integrity technology. These advances have been incorporated into Section II, Section VIII, Code Cases, and particularly Subsection NH of Section III of the Code, “Components in Elevated Temperature Service.” The current need for designs for very high temperature and for Gen IV systems requires the extension of operating temperatures from about 1400°F (760°C) to about 1742°F (950°C) where creep effects limit structural integrity, safe allowable operating conditions, and design life. Materials that are more creep and corrosive resistant are needed for these higher operating temperatures. Material models are required for cyclic design analyses. Allowable strains, creep fatigue and creep rupture interaction evaluation methods are needed to provide assurance of structural integrity for such very high temperature applications. Current ASME Section III design criteria for lower operating temperature reactors are intended to prevent through-wall cracking and leaking and corresponding criteria are needed for high temperature reactors. Subsection NH of Section III was originally developed to provide structural design criteria and limits for elevated-temperature design of Liquid-Metal Fast Breeder Reactor (LMFBR) systems and some gas-cooled systems. The U.S. Nuclear Regulatory Commission (NRC) and its Advisory Committee for Reactor Safeguards (ACRS) reviewed the design limits and procedures in the process of reviewing the Clinch River Breeder Reactor (CRBR) for a construction permit in the late 1970s and early 1980s, and identified issues that needed resolution. In the years since then, the NRC, DOE and various contractors have evaluated the applicability of the ASME Code and Code Cases to high-temperature reactor designs such as the VHTGRs, and identified issues that need to be resolved to provide a regulatory basis for licensing. The design lifetime of Gen IV Reactors is expected to be 60 years. Additional materials including Alloy 617 and Hastelloy X need to be fully characterized. Environmental degradation effects, especially impure helium and those noted herein, need to be adequately considered. Since cyclic finite element creep analyses will be used to quantify creep rupture, creep fatigue, creep ratcheting and strain accumulations, creep behavior models and constitutive relations are needed for cyclic creep loading. Such strain- and time-hardening models must account for the interaction between the time-independent and time-dependent material response. This paper describes the evolving structural integrity evaluation approach for high temperature reactors. Evaluation methods are discussed, including simplified analysis methods, detailed analyses of localized areas, and validation needs. Regulatory issues including weldment cracking, notch weakening, creep fatigue/creep rupture damage interactions, and materials property representations for cyclic creep behavior are also covered.

Energy security and greenhouse gas reductions are thought to be two of the most urgent priorities for sustaining and improving the human condition in the future. Few places pit the two goals so directly in opposition to one another as the Alberta oil sands. Here, Canadian natural gas is burned in massive quantities to extract oil from one of North America’s largest native sources of carbon-intensive heavy oil. This conflict need not continue, however; non-emitting nuclear energy can replace natural gas as a fuel source in an economical and more environmentally sound way. This would allow for the continued extraction of transportation fuels without greenhouse gas emissions, while freeing up the natural gas supply for hydrogen feedstock and other valuable applications. Bitumen production in Alberta has expanded dramatically in the past five years as the price of oil has risen to record levels. This paper explores the feasibility and economics of using nuclear energy to power future oil sands production and upgrading activities, and puts forth several nuclear energy application scenarios for providing steam and electricity to in-situ and surface mining operations. This review includes the Enhanced CANDU 6, the Advanced CANDU Reactor (ACR) and the Pebble Bed Modular Reactor (PBMR). Based on reasonable projections of available cost information, nuclear energy used for steam production is expected to be less expensive than steam produced by natural gas at current natural gas prices and under $7/MMBtu (CAD). For electricity production, nuclear becomes competitive with natural gas plants at natural gas prices of $10–13/MMBtu (CAD). Costs of constructing nuclear plants in Alberta are affected by higher local labor costs, which this paper took into account in making these estimates. Although more definitive analysis of construction costs and project economics will be required to confirm these findings, there appears to be sufficient merit in the potential economics to support further study. A single 500MWth PBMR reactor is able to supply high-pressure steam for a 40,000 to 60,000 bpd Steam Assisted Gravity Drainage (SAGD) plant, whereas the CANDU and ACR reactors are unable to produce sufficient steam pressures to be practical in that application. The CANDU, ACR and PBMR reactors have potential for supplying heat and electricity for surface mining operations. The primary environmental benefit of nuclear energy in this application is to reduce CO 2 emissions by up to 3.1 million metric tons per year for each 100,000 barrel per day (bpd) bitumen production SAGD facility, or 2.0 million metric tons per year for the replacement of 700MWe of grid electricity with a nuclear power plant. Should carbon emissions be priced, the economic advantages of nuclear energy would be dramatically improved such that with a $50/ton CO 2 e at the releases expected for typical projects using natural gas, breakeven gas prices for nuclear drop to less than $3.50/MMBtu, well below the current natural gas price of $10/MMBtu for SADG steam production.

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.

This paper describes the experimental validation of a proposed method that uses a small amount of helium injection to prevent the onset of natural circulation in high temperature gas reactors (HTGR) following a depressurized loss of coolant accident. If this technique can be shown to work, air ingress accidents can be mitigated. A study by Dr. Xing L. Yan et al. (2008) developed an analytical estimate for the minimum injection rate (MIR) of helium required to prevent natural circulation. Yan’s study used a benchmarked CFD model of a prismatic core reactor to show that this method of helium injection would impede natural circulation. The current study involved the design and construction of an experimental apparatus in conjunction with a CFD model to validate Yan’s method. Based on the computational model, a physical experimental model was built and tested to simulate the main coolant pipe rupture of a Pebble Bed Reactor (PBR), a specific type of HTGR. The experimental apparatus consisted of a five foot tall, 2 inch diameter, copper U-tube placed atop a 55-gallon barrel to reduce sensor noise from outside air movement. Hot and cold legs were simulated to reflect the typical natural circulation conditions expected in reactor systems. FLUENT was used to predict the diffusion and circulation phases. Several experimental trials were run with and without helium injection. Results showed that with minimal helium injection, the onset of natural circulation was prevented which suggests that such a method may be useful in the design of high temperature gas reactors to mitigate air ingress accidents.

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.

The ASME Committee on Nuclear Risk Management (CNRM) has established a working group to pursue the development of a PRA standard that can be used for advanced non-LWR plants. The applications of such PRAs include the performance of PRAs to support licensing and design decisions, and to meet NRC requirements for Design Certifications and Construction and Operating Licenses. The purpose of this paper is to summarize the significant progress that has been made to date in developing a new PRA standard for non-LWRs from the personal point of view of the working group chairman.

In the last years considerable efforts have been made at the Institute for Transuranium Elements (ITU) in order to reestablish European knowledge and ability in safety testing of irradiated high temperature reactor (HTR) Fuel Elements. In the framework of the 6th European framework programme a cold finger apparatus (Ku¨FA) furnace, formerly installed at FZ-Ju¨lich (FzJ), has been installed in a hot cell at ITU [Freis 2008] in order to test fission product release under high temperature and non-oxidising conditions. Several analytical methods (e.g. Gamma-spectrometry, mass-spectrometry) have been applied in order to analyse different isotopes released during Ku¨FA tests. After the heating tests, examinations of the fuel elements were performed including scanning electron microscopy (SEM) and micro-hardness testing of coated particles. Individual coated particles were object of heating tests in a Knudsen cell with a coupled mass spectrometer measuring all released species. In order to cover more accident scenarios, a second furnace for oxidising-conditions (air- or water-ingress) was constructed and installed in a cold lab. Furthermore a disintegration apparatus, based on anodic oxidation, was constructed and fuel elements were dissolved obtaining thousands of individual coated particles for further examination. A fully automated irradiated microsphere gamma analyzer (IMGA) is under construction and will be used, in particular, to identify and sort out failed particles.

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

Some recent studies of material response have identified an issue that crosses over and blurs the boundary between ASME Boiler and Pressure Vessel Code Section III Subsection NB and Subsection NH. For very long design lives, the effects of creep show up at lower and lower temperature as the design life increases. Although true for the temperature at which the allowable stress is governed by creep properties, the effect is more apparent, e.g. creep effects show up sooner, at local structural discontinuities and peak thermal stress locations. This is because creep is a function of time, temperature and stress and the higher the localized stress, the lower in temperature creep begins to cause damage. If the threshold is below the Subsection NB to NH temperature boundary, 700°F for ferritic steels and 800°F for austenitic materials, then this potential failure mode will not be considered. Unfortunately, there is no experience base with very long lives at temperatures close to but under the Subsection NB to NH boundary to draw upon. This issue is of particular interest in the application of Subsection NB rules of construction to some High Temperature Gas Reactor (HTGR) concepts. The purpose of this paper is, thus, twofold; one part is about statistical treatment and extrapolation of sparse data for a specific material of interest, A533B; the other part is about how these results could impact current design procedures in Subsection NB.