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 / CO 2 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 (H 2 production rate based upon lower heating value).

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

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.

The Next Generation Nuclear Power/Advanced Gas Reactor (NGNP/AGR) Fuel Development and Qualification Program included the design, installation, and testing of a 6-inch diameter nuclear fuel particle coater to demonstrate quality TRISO fuel production on a small industrial scale. Scale-up from the laboratory-scale coater faced challenges associated with an increase in the kernel charge mass, kernel diameter, and a redesign of the gas distributor to achieve adequate fluidization throughout the deposition of the four TRISO coating layers. TRISO coatings are applied at very high temperatures in atmospheres of dense particulate clouds, corrosive gases, and hydrogen concentrations over 45% by volume. The severe environment, stringent product and process requirements, and the fragility of partially-formed coatings limit the insertion of probes or instruments into the coater vessel during operation. Pressure instrumentation were installed on the gas inlet line and exhaust line of the 6-inch coater to monitor the bed differential pressure and internal pressure fluctuations emanating from the fuel bed as a result of bed and gas “bubble” movement. These instruments are external to the particle bed and provide a glimpse into the dynamics of fuel particle bed during the coating process and data that could be used to help ascertain the adequacy of fluidization and, potentially, the dominant fluidization regimes. Pressure fluctuation and differential pressure data are not presently useful as process control instruments, but data suggest a link between the pressure signal structure and some measurable product attributes that could be exploited to get an early estimate of the attribute values.

A major element of the Next Generation Nuclear Plant (NGNP)/Advanced Gas Reactor (AGR) Fuel Development and Qualification Program is developing fuel fabrication processes to produce high quality uranium-containing fuel kernels, TRISO-coated particles and fuel compacts needed for planned irradiation tests. The goals of the program also include developing the fabrication technology to mass produce this fuel at low cost. Kernels for the first AGR test, AGR-1, consisted of uranium oxycarbide (UCO) microspheres that were produced by an internal gelation process followed by high temperature steps to convert the UO 3 + C “green” microspheres to UO 2 + UC x . The high temperature steps also densified the kernels. Babcock and Wilcox (B&W) fabricated UCO kernels in their Lynchburg facility for the AGR-1 irradiation experiment, which went into the Advanced Test Reactor (ATR) at Idaho National Laboratory in December 2006. An evaluation of the kernel process prior and after these kernels were produced led to several recommendations to improve the fabrication process. These recommendations included testing alternative methods of dispersing carbon during broth preparation, evaluating the method of broth mixing, optimizing the broth chemistry, optimizing sintering conditions, and demonstrating fabrication of larger diameter UCO kernels needed for the second AGR irradiation test, AGR-2. Based on these recommendations and requirements, a test program was defined and performed. Certain portions of the test program were performed by Oak Ridge National Laboratory (ORNL), while tests at larger scale were performed by B&W. The tests at B&W have demonstrated improvements in both kernel properties and process operation. Changes in the form of carbon black used and the method of mixing the carbon prior to forming kernels led to improvements in the phase distribution in the sintered kernels, greater consistency in kernel properties, a reduction in forming run time, and simplifications to the forming process. Process parameter variation tests in both forming and sintering steps led to an increased understanding of the acceptable ranges for process parameters and additional reduction in required operating times. Another result of this test program was to double the kernel production rate. Following the development tests, approximately 40 kg of natural uranium UCO kernels have been produced for use in coater scale up tests, and approximately 10 kg of low enriched uranium UCO kernels for use in the AGR-2 experiment.

For direct cycle gas cooled high temperature reactor designs, operating conditions may be limited as a result of excessive maintenance dose rates caused by the Ag-110m source term on the turbine. It is therefore important to accurately predict silver release from fuel during reactor operation. Traditionally diffusion models were used to derive transport parameters from limited irradiation testing of fuel materials and components. Best estimates for all applicable German fuel irradiation tests with defendable uncertainty ranges were never derived. However, diffusion theory and current parameters cannot account for all irradiation and heat-up test results, and for some tests, it appears unacceptably conservative. Other transport mechanisms have been suggested, and alternative calculation models are being considered. In this paper the applicable German irradiation test results are evaluated with a classic diffusion model as well as an alternative model called the Molecular Vapour transport Release (MVR) model. New transport models and parameters for silver in fuel materials are suggested and compared.

The evaluation of fission gas release from spherical fuel during irradiation testing is critical to understand expected fuel performance under real reactor conditions. Online measurements of Krypton and Xenon fission products explain coated particle performance and contributions from graphitic matrix materials used in fuel manufacture and irradiation rig materials. Methods that are being developed to accurately evaluate fission gas release are described here together with examples of evaluations performed on irradiation tests HFR-K5, -K6 and EU1bis.

For the Pebble Bed Modular Reactor (PBMR) Demonstration Power Plant (DPP) several neutron flux measurements are made, both within the Reactor Pressure Vessel (RPV) and outside the RPV. The measurements within the RPV are performed by the Core Structures Instrumentation (CSI) system. While those outside the RPV are performed by the Nuclear Instrumentation System (NIS). The PBMR has a long annular core with a relative low power density, requiring flux monitoring over the full 11 M of the active core region. The core structures instrumentation measures the neutron flux in the graphite reflector. Two measurement techniques are used; Fission Chamber based channels with high sensitivity for initial fuel load and low power testing and SPND channels for measurements at full and near full power operation. The CSI flux monitoring supports data acquisition for design Verification and Validation (V&V), and the data will also be used for the characterization of the NIS for normal reactor start-ups and low power operation. The CSI flux measurement channels are only required for the first few years of operation; the sensors are not replaceable. The Nuclear Instrumentation System is an ex core system that includes the Post Event Instrumentation. Due to the long length of the PBMR core, the flux is measured at several axial positions. This is a fission chamber based system; full advantage is taken of all the operating modes for fission chambers (pulse counting, mean square voltage (MSV), and linear current). The CSI flux monitoring channels have many technical and integration challenges. The environment where the sensors and their associated signal cables are required to operate is extremely harsh; temperature and radiation levels are very high. The selection and protection of the fission chambers warranted special attention. The selection criteria for sensors and cables takes cognizance of the fact that the assemblies are built in during the assembly of the reactor internal structures, and that they are not replaceable. This paper describes the challenges in the development of the monitoring systems for the measurement of neutron flux both within the RPV and the ex core region. The selection of detector configuration and the associated signal processing will be discussed. The use of only analogue signal processing techniques will also be elaborated on.

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

In order to support the Next Generation Nuclear Plant (NGNP) Program 2018 deployment schedule, the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program must reduce the AGR fuel irradiation testing time in the Advanced Test Reactor (ATR) from approximately 2 1/2 calendar years to 1 1/2 calendar years. The AGR fuel irradiation testing requirements are: (a) burn-up of at least 14% FIMA; (b) Fast neutron fluence (E > 0.18 MeV) – maximum < 5.1 × 10 25 n/m 2 ; (c) limit of fission power density is 350 W/cc; and (d) irradiation time < 1 1/2 calendar years. The accelerated testing could be accomplished by utilizing the ATR North East flux trap (NEFT) position, which can provide more control of the thermal neutron flux rate than the ATR B-10 position currently being used for the AGR-1 fuel testing, which is regulated to achieve the fuel temperature and burn-up rate requirements. In addition, the Fast (E > 1.0 MeV) to Thermal (E < 0.625 eV) neutron flux ratio (F/T) for the NEFT is much harder (higher) than the F/T ratio for the B-10 position. Thus, an appropriate configuration of Beryllium (Be) and water will need to be determined in order to soften (lower) the F/T ratio to the desired value. The proposed AGR 7-position fuel test configuration in the NEFT will utilize a graphite holder consisting of six fuel specimen positions arranged around the perimeter of the graphite holder with a seventh fuel specimen position in the center of the holder. To soften the neutron spectrum in the fuel compacts, the water volume in the outer water annulus can be increased. To reduce the compact power density, a hafnium filter could be incorporated around the graphite holder. After several trials, a hafnium filter with a thickness of 0.008 inches appeared to adequately reduce the power density to achieve the fuel testing requirements. It was also determined that the chosen beryllium-tube and water annulus configuration would adequately soften the neutron spectrum to achieve the fuel testing requirement. This neutronics study is based upon typical ATR cycle operation of 50 effective full power days (EFPD) per cycle for seven proposed irradiation cycles, and a NE lobe power of the 14 MW. The MCWO-calculated fuel compact power density, burnup (% FIMA), and fast neutron fluence (E > 0.18 MeV) results indicate that the average fuel compact burnup and fast neutron fluence reach 14.79% FIMA and 4.16 × 10 25 n/m 2 , respectively. The fuel compact peak burnup reached 16.68% FIMA with corresponding fast neutron fluence for that fuel compact of 5.06 × 10 25 n/m 2 , which satisfied the fuel testing requirements. It is therefore concluded that accelerating the AGR fuel testing using the proposed AGR 7-position fuel test configuration in the NEFT is very feasible.

HTR projects have been launched within the European Union Framework Programmes (FP’s) to consolidate and advance HTR and VHTR technology within Europe. This paper reviews the main achievements arising out of the work in the area of materials and component development. The programme to date addresses material and qualification requirements for the reactor pressure vessel, high temperature resistant alloys and technological development aspects of the power circuit components, material property needs and issues for the graphite core and requirements for Codes and Standards. The experimental programme includes irradiation and feature testing, tests on reduced scale mock-ups and bearings, corrosion, modelling and analysis issues. For the 6th Framework activities which are current, the main European research focus on VHTR is through the RAPHAEL-IP. Results and main conclusions from the work are reported, also a summary of the status of the test work and recommendations for future actions. This programme of work provides important results for the International Generation IV VHTR Materials and Components Research and Development programme as part of the EURATOM contribution.

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 purpose of this paper is to present the results of a study to establish strategies for the reliability and integrity management (RIM) of passive metallic components for the PBMR. The RIM strategies investigated include design elements, leak detection and testing approaches, and non-destructive examinations. Specific combinations of strategies are determined to be necessary and sufficient to achieve target reliability goals for passive components. This study recommends a basis for the RIM program for the PBMR Demonstration Power Plant (DPP) and provides guidance for the development by the American Society of Mechanical Engineers (ASME) of RIM requirements for Modular High Temperature Gas-Cooled Reactors (MHRs).

Sandia National Laboratories (SNL), General Atomics Corporation (GA) and the French Commissariat a` l’Energie Atomique (CEA) have been conducting laboratory-scale experiments to investigate the thermochemical production of hydrogen using the Sulfur-Iodine (S-I) process. This project is being conducted as an International Nuclear Energy Research Initiative (INERI) project supported by the CEA and US DOE Nuclear Hydrogen Initiative. In the S-I process, 1) H 2 SO 4 is catalytically decomposed at high temperature to produce SO 2 , O 2 and H 2 O. 2) The SO 2 is reacted with H 2 O and I 2 to produce HI and H 2 SO 4 . The H 2 SO 4 is returned to the acid decomposer. 3) The HI is decomposed to H 2 and I 2 . The I 2 is returned to the HI production process. Each participant in this work is developing one of the three primary reaction sections. SNL is responsible for the H 2 SO 4 decomposition section, CEA, the primary HI production section and General Atomics, the HI decomposition section. The objective of initial testing of the S-I laboratory-scale experiment was to establish the capability for integrated operations and demonstrate H 2 production from the S-I cycle. The first phase of these objectives was achieved with the successful integrated operation of the SNL acid decomposition and CEA Bunsen reactor sections and the subsequent generation of H 2 in the GA HI decomposition section. This is the first time the S-I cycle has been realized using engineering materials and operated at prototypic temperature and pressure to produce hydrogen.

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.

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

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.