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.

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.

The irradiation experiment HFR-EU1 in the HFR Petten is currently being conducted by the European Commission’s Joint Research Centre – Institute for Energy (JRC-IE). The irradiation targets are 5 spherical High Temperature Reactor (HTR) fuel pebbles, 2 of INET production and 3 of former German production. Both types are made of TRISO coated particles and are tested for their potential for very high temperature performance and high burn-up. The irradiation started on 29 September 2006 and, by 24 February 2008, had accumulated 12 reactor cycles totaling 332.8 efpd and a calculated maximum burn-up of 8.9% FIMA (INET) and 11.2% FIMA (AVR). The objective of the HFR-EU1 test is to irradiate 5 HTR fuel pebbles at conditions beyond the characteristics of current HTR reactor designs with pebble bed cores, e.g. HTR-Modul, HTR-10 and PMBR. This should demonstrate that pebble bed HTRs are capable of enhanced performance in terms of sustainability (further increased power conversion efficiency, improved fuel use) and thus reduced waste production. The surface temperature of all pebbles was held constant during the irradiation, with the exception of HFR downtime and power transients. HFR-EU1 should demonstrate the feasibility of low coated particle failure fractions under normal operating conditions and more specifically: • high fuel surface temperature of 900°C (INET) and 950°C (AVR); • very high burn-up of 17% FIMA (INET) and 20% FIMA (AVR) which is significantly higher than the license limit of the HTR-Modul (approx. 8% FIMA); it will be explained in this paper why this objective had to be somewhat reduced due to excessive irradiation time requirements and technological difficulties. This paper provides the irradiation history of the experiment performed so far including data on fission gas release.

The Next Generation Nuclear Plant (NGNP) will most likely produce electricity and process heat, with both being considered for hydrogen production. To capture nuclear process heat, and transport it to a distant industrial facility requires a high temperature system of heat exchangers, pumps and/or compressors. The heat transfer system is particularly challenging not only due to the elevated temperatures (up to ∼ 1300K) and industrial scale power transport (≥50 MW), but also due to a potentially large separation distance between the nuclear and industrial plants (100+m) dictated by safety and licensing mandates. The work reported here is the preliminary analysis of two-phase thermosyphon heat transfer performance with alkali metals. A thermosyphon is a device for transporting heat from one point to another with quite extraordinary properties. In contrast to single-phased forced convective heat transfer via ‘pumping a fluid’, a thermosyphon (also called a wickless heat pipe) transfers heat through the vaporization / condensing process. The condensate is further returned to the hot source by gravity, i.e. without any requirement of pumps or compressors. With this mode of heat transfer, the thermosyphon has the capability to transport heat at high rates over appreciable distances, virtually isothermally and without any requirement for external pumping devices. Two-phase heat transfer by a thermosyphon has the advantage of high enthalpy transport that includes the sensible heat of the liquid, the latent heat of vaporization, and vapor superheat. In contrast, single-phase forced convection transports only the sensible heat of the fluid. Additionally, vapor-phase velocities within a thermosyphon are much greater than single-phase liquid velocities within a forced convective loop. Thermosyphon performance can be limited by the sonic limit (choking) of vapor flow and/or by condensate entrainment. Proper thermosyphon requires analysis of both.

US and international applications for large onsite cogeneration (steam and power) systems are emerging as a near term market for the PBMR. The South African PBMR demonstration project applies a high temperature (900°C) Brayton cycle for high efficiency power generation. In addition, a number of new applications are being investigated using an intermediate temperature range (700–750°C) with a simplified heat supply system design. This intermediate helium delivery temperature supports conventional steam Rankine cycle designs at higher efficiencies than obtained from water type reactor systems. These designs can be adapted for cogeneration of steam, similar to the design of gas turbine cogeneration plants that supply steam and power at many industrial sites. This temperature range allows use of conventional or readily qualifiable materials and equipment, avoiding some cost premiums associated with more difficult operating conditions. As gas prices and CO 2 values increase, the potential value of a small nuclear reactor with advanced safety characteristics increases dramatically. Because of its smaller scale, the 400–500MW t PBMR offers the economic advantages of onsite thermal integration (steam, hot water and desalination coproduction) and of providing onsite power at cost versus at retail industrial rates avoiding transmission and distribution costs. Advanced safety characteristics of the PBMR support the location of plants adjacent to steam users, district energy systems, desalination plants, and other large commercial and industrial facilities. Additional benefits include price stability, long term security of energy supply and substantial CO 2 reductions. Target markets include existing sites using gas fired boilers and cogeneration units, new projects such as refinery and petrochemical expansions, and coal-to-liquids projects where steam and power represent major burdens on fuel use and CO 2 emissions. Lead times associated with the nuclear licensing process may support early applications in the 2018–2020 timeframe. This paper summarizes the design options likely to shape these emerging steam and cogeneration applications.

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.

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.

Loss of primary coolant flow test is under planning by using the High Temperature engineering Test Reactor (HTTR). In this test, all the gas circulators are tripped and the position of all control rods keeps its initial one. The new calculation model was developed to perform the preliminary analysis for the test. This model is so improved that an equivalent fuel channel model based on one point kinetics code and a whole reactor model based on two-dimensional thermal analysis code are coupled to simulate the reactor performance during the loss of coolant flow. Both calculation codes were used in the safety evaluation of the HTTR licensing. The improved calculation model was validated by comparison between the calculated result and the experimental one obtained from the coolant flow reduction test in the HTTR. The loss of primary coolant flow test simulates the depressurization accident and the data obtained from the test is useful for the validation and improvement of the calculation code applied to the safety analysis in the future HTGR such as Very High Temperature Reactor which is selected as one of candidates of the generation IV reactor system.

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.

Due to its high operating temperature (up to 850°C with present technologies, possibly higher in the longer term), and its power range (a few hundred MW), the modular HTR could address a larger scope of industrial process heat needs than other present nuclear systems. Even if HTR can contribute to competitive electricity generation, this potential for industrial heat applications is the main incentive for developing this type of reactor, as it could open to nuclear energy a large non-electricity market. However several issues must be addressed and solved successfully for HTR to actually enter the market of industrial process heat: 1) as an absolute prerequisite, to develop a strategic alliance of nuclear industry and R&D with process heat user industries. 2) to solve some key technical issues, as for instance the design of a reactor and of a coupling system flexible enough to reconcile a single reactor design with multiple applications and versatile requirements for the heat source, and the development of special adaptations of the application processes or even of new processes to fit with the assets and constraints of HTR heat supply, 3) to solve critical industrial issues such as economic competitiveness, availability and 4) to address the licensing issues raised by the conjunction of nuclear and industrial risks. In line with IAEA initiatives for supporting non-electric applications of nuclear energy and with the orientations of the SET-Plan of the European Commission, the (European) HTR Technology Network (HTR-TN) proposes a new project, together with industrial process heat user partners, to provide a first impetus to the strategic alliance between nuclear and non-nuclear industries. End user requirements will be expressed systematically on the basis of inputs from industrial partners on various types of process heat applications. These requirements will be confronted with the capabilities of the HTR heat source, in order to point out possible discrepancies and issues, to assess the feasibility of different coupling schemes and to identify development needs. Partners from nuclear regulatory organisations will also address the feasibility of licensing such coupling schemes. The issues they will raise will be taken into consideration for defining coupling design bases and identifying R&D needs. A detailed roadmap for designing an industrial demonstrator of a HTR coupled with process heat applications will be inferred from this analysis, as well as R&D actions required for supporting the development of the reactor, of the coupling system and of possible adaptations or innovations in industrial processes.

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 Germany two HTR nuclear power plants had been built and operated, the AVR-15 and the THTR-300. Also various projects for different purposes in a large power range had been developed. The AVR-15, an experimental reactor with a power output of 15 MWel was operated for more than 20 years with excellent results. The THTR-300 was designed as a prototype demonstration plant with 300 MWel and should be the technological basis for the entire future reactor line. The THTR-300 was prematurely shut down and decommissioned because of political reasons. But because of the accompanying comprehensive R&D program and the operation time of about 5 years, the technology was proved and essential operational results were gained. The AVR steam generator was installed above the reactor core. The six THTR heat exchangers were arranged circularly around the reactor core. Both heat exchanger systems have been operated successfully and furthermore acted as a residual heat removal system. The technology knowledge and experience gained on these existing HTR plants is still available at Westinghouse Electric Germany GmbH since Westinghouse is one of the legal successors of the former German HTR companies. As a follow-up project of THTR, the HTR-500 was developed and designed up to the manufacturing stage. For this plant additionally to the 8 steam generators, two residual heat removal heat exchangers were foreseen. These were to be installed in a ring around the reactor core. All these HTRs were designed for the generation of electricity using a steam cycle. Extensive research work has also been done for advanced applications of HTR technology e.g. using a direct cycle within the HHT project or generating process heat within the framework of the PNP project. Because of the critical attitude of the German government to the nuclear power in the past 20 years in Germany there was only a very limited interest in the further development of the HTR technology. As a consequence of the German decision, at the beginning of the 90s, to phase out nuclear power completely, research and funding of further development of HTR reactor design was also cut down. Today’s HTR reactor designs, such as the PBMR in South Africa, use a direct cycle with a gas turbine. This technology is also based on the THTR technology and PBMR is a licensed party. For the HTR-PM in China and the future oil sand projects powered by HTR’s in Canada and Siberia however the use of steam generators is required. Westinghouse and Dresden University cooperate in the field of steam generator technology for HTR reactors. The existing know-how for HTR is based on a huge pool of knowledge gained by the past German HTR projects mentioned above and consists especially of the design methodology, the mechanical layout and material issues for helium heated steam generators. The project team consists of experienced specialists who have worked on HTR projects in the past and of young graduate engineers. Main goal of the project is to analyze the existing know-how and to adjust it to the state of the art. As a first step, the existing design and its methodology is being analyzed and the different points of improvement are identified. The final step of the program is the description of a new methodology which fulfills the severe requirements of the customer and all of the actual licensing conditions. One of the reasons why this project has been launched is that the requirements of life expectancy for HTR components increase and the material limits will be reached, especially at high temperatures. This implies that the design of helix heat exchangers has to allow inservice inspections; this was not a requirement for the previous THTR design. Methodologies for in-service inspections already had been developed, but they are not sufficient for today’s tube lengths and have to be adapted. Another example, based on operating experience, is using reheaters to increase the efficiency is not recommended today. Using supercritical steam conditions to increase the efficiency should be investigated instead. In general, the economic benefit has to be balanced against the additional costs resulting from better material and more complex manufacturing.

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.

This paper addresses that major changes in the safety approach, for instance the increased use of Probabilistic Risk Assessment (PRA), have been made. All commercial reactors in operation today belong to the Generations II and III. Generation IV International Forum (GIF) has launched several programs aimed at developing the next generation of nuclear energy systems. Part of the research effort is focused on new reactor concepts, such as the Very High Temperature Reactor (VHTR), currently developing in Korea. In parallel to the design process of VHTR currently underway, regulatory approach is moving forward to define new licensing rules. So, Korea Institute of Nuclear Safety (KINS) is defining, as a goal to risk-inform, the regulation and developing the regulatory framework and licensing process more efficient, predictable, and stable. However, the licensing of NPPs has focused until now on Light Water Reactors (LWRs) and has not incorporated systematically insights and benefits from PRA. In the meantime, USNRC and IAEA have recently drafted a risk-informed regulation and technology-neutral framework (TNF) for new plant licensing along with the innovative Gen-IV system design. KINS also expects that advanced NPPs will show enhanced margins of safety, and that advanced reactor designs will comply with the national safety goal policy statement. In order to meet these expectations, PRA tools are currently being considered by KINS; among them are frequency-consequence (F-C) curves, which plot the frequency of having Consequence. This paper discusses the role and the usefulness of such curves in risk-informing the licensing process in Korea, and shows that the use of F-C curve allows the implementation of both structural and rational Defence-In-Depth (DID). This paper focuses on F-C curves as means to assess the licensing basis events (LBEs) from the regulatory viewpoint on the innovative small and medium reactor (SMR) sized VHTR deployment in Korea. The principle underlying the F-C curve is that event frequency and dose are inversely related, i.e., the higher the dose consequences, the lower is the allowed event frequency.

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.