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Materials and Components

2008;():1-10. doi:10.1115/HTR2008-58050.

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.

Commentary by Dr. Valentin Fuster
2008;():11-28. doi:10.1115/HTR2008-58122.

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.

Commentary by Dr. Valentin Fuster
2008;():29-34. doi:10.1115/HTR2008-58130.

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.

Commentary by Dr. Valentin Fuster
2008;():35-42. doi:10.1115/HTR2008-58146.

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.

Commentary by Dr. Valentin Fuster
2008;():43-47. doi:10.1115/HTR2008-58149.

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.

Commentary by Dr. Valentin Fuster
2008;():49-56. doi:10.1115/HTR2008-58152.

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.

Topics: Corrosion
Commentary by Dr. Valentin Fuster
2008;():57-68. doi:10.1115/HTR2008-58166.

In the frame of the international forum GenIV, CEA has selected various innovative concepts of Gas cooled Nuclear Reactor. Among them, an indirect-cycle gas reactor is under consideration. Thermal hydraulic performances are a key issue for the design. For transient conditions and decay heat removal situations, the thermal hydraulic performance must remain as high as possible. In this context, all the transient situations, the incidental and accidental scenarii must be evaluated by a validated system code able to correctly describe, in particular, the thermalhydraulics of the whole plant. As concepts use a helium compressor to maintain the flow in the core, a special emphasis must be laid on compressor modelling. Centrifugal circulators with a vaneless diffuser have significant properties in term of simplicity, cost, ability to operate over a wide range of conditions. The objective of this paper is to present a dedicated description of centrifugal compressor, based on a one dimensional approach. This type of model requires various correlations as input data. The present contribution consists in establishing and validating the numerical simulations (including different sets of correlations) by comparison with representative experimental data. The results obtained show a qualitatively correct behaviour of the model compared to open literature cases of the gas turbine aircraft community and helium circulators of High Temperature Gas Reactors. Further work on modelling and validation are nevertheless needed to have a better confidence in the simulation predictions.

Commentary by Dr. Valentin Fuster
2008;():69-74. doi:10.1115/HTR2008-58195.

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.

Commentary by Dr. Valentin Fuster
2008;():75-79. doi:10.1115/HTR2008-58200.

Several nickel based solid solution alloys are under consideration for application in heat exchangers for very high temperature gas cooled reactors. The principal candidates being considered for this application by the Next Generation Nuclear Plant (NGNP) project are Inconel 617 and Haynes 230. While both of these alloys have an attractive combination of creep strength, fabricability, and oxidation resistance a good deal remains to be determined about their environmental resistance in the expected NGNP helium chemistry and their long term response to thermal aging. A series of experiments has been carried out in a He loop with controlled impurity chemistries within the range expected for the NGNP. The influence of oxygen partial pressure and carbon activity on the microstructure and mechanical properties of Alloys 617 and 230 has been characterized. A relatively simple phenomenological model of the environmental interaction for these alloys has been developed.

Commentary by Dr. Valentin Fuster
2008;():81-89. doi:10.1115/HTR2008-58215.

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

Commentary by Dr. Valentin Fuster
2008;():91-96. doi:10.1115/HTR2008-58269.

The effect of oxygen partial pressure on fatigue and SCC growth rates in alloy 617 has been studied using both static and fatigue loading @ 650°C over the oxygen partial pressure range 10−19 –10−2 atm. Tests were conducted at either constant stress intensity factor, K, for static conditions or constant ΔK in fatigue. Oxygen concentration was measured on both the inlet and outlet as well as in-situ with a probe located directly at the specimen surface. For fatigue loading the crack path was observed to be transgranular but crystallographic with a decreasing growth rate as the oxygen concentration decreased. However, for static loading the crack path shifted to intergranular and exhibited an increasing crack growth rate with decreasing oxygen concentration.

Commentary by Dr. Valentin Fuster
2008;():97-105. doi:10.1115/HTR2008-58281.

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.

Topics: Creep , Temperature , Design
Commentary by Dr. Valentin Fuster
2008;():107-108. doi:10.1115/HTR2008-58284.

The unique combination of physical properties inherent to graphite makes it an attractive material for use as a moderator in high-temperature nuclear reactors (HTR’s). High-temperature physical properties of three nuclear grade graphites manufactured by GrafTech International Holdings Inc. (GrafTech) (PCEA, PCIB-SFG, and PPEA) have been determined experimentally and are presented here. Tensile strength, Young’s modulus, thermal conductivity, specific resistance, and coefficient of thermal expansion (CTE) data are collected at temperatures from 25 °C to as high as 2000 °C and are found to be consistent with classical graphite behavior.

Commentary by Dr. Valentin Fuster

Safety and Licensing

2008;():109-125. doi:10.1115/HTR2008-58030.

Over its 1968–1988 life, PSCo relicensed the Fort St. Vrain (FSV) High-temperature Gas Reactor (HTGR) for light water reactor (LWR) technology requirements. Estimates of the financial losses associated with the plant range from $500 million to $2 billion in 1980 dollars. Colorado ratepayers, the shareholders of Gulf General Atomics and its corporate successors — General Atomics, GA Technologies or just GA and Public Service Company of Colorado (PSCo) bore these losses. Two critical plant issues required solution for the plant’s economic success — (1) the high-cost of 93% enriched uranium fuel and (2) low unit availability. While fuel costs were beyond utility control, low availability was controllable, yet remained unresolved. Commercially isolated for twenty years, PSCo shut the plant down in 1988. Economic success of future HTGRs depends upon avoiding similar complications. This paper examines the issues that made FSV uneconomic, including those fundamental to HTGR technology and others attributable to the utility operator and its culture. Knowing the history of FSV and HTGR design, designers should anticipate reasonable challenges. Preparations will help manage future HTGR risks, costs, and assure operating success. Regulators and industry can assure more effective, economic operations in the next round of HTGR designs.

Commentary by Dr. Valentin Fuster
2008;():127-133. doi:10.1115/HTR2008-58036.

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).

Commentary by Dr. Valentin Fuster
2008;():135-142. doi:10.1115/HTR2008-58037.

This paper discusses the reliability and integrity management (RIM) strategies that have been applied in the design of the PBMR passive metallic components for the helium pressure boundary (HPB) to meet reliability targets and to evaluate what combination of strategies are needed to meet the targets. The strategies considered include deterministic design strategies to reduce or eliminate the potential for specific damage mechanisms, use of an on-line leak monitoring system and associated design provisions that provide a high degree of leak detection reliability, and periodic non-destructive examinations combined with repair and replacement strategies to reduce the probability that degradation would lead to pipe ruptures. The PBMR RIM program for passive metallic piping components uses a leak-before-break philosophy. A Markov model developed for use in LWR risk-informed inservice inspection evaluations was applied to investigate the impact of alternative RIM strategies and plant age assumptions on the pipe rupture frequencies as a function of rupture size. Some key results of this investigation are presented in this paper.

Commentary by Dr. Valentin Fuster
2008;():143-146. doi:10.1115/HTR2008-58038.

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.

Commentary by Dr. Valentin Fuster
2008;():147-155. doi:10.1115/HTR2008-58051.

A scoping study on a confinement approach to a PBMR type plant was performed to quantify its performance in terms of retention and thermal-hydraulics in case of a postulated breach in the Helium Pressure Boundary. This paper introduces a simulation strategy for accident analyses of a PBMR confinement. Based on two accident sequences characterized by Helium Pressure Boundary breaches, a small (AOO) and a large (DBA) break, the estimates from the ASTEC and CONTAIN codes are presented and compared to check their capabilities and consistency. The results obtained indicate that both codes predict very similar thermal-hydraulic responses of the confinement both in magnitude and timing. As for the aerosol behaviour, both codes predict that most of inventory coming into the confinement is eventually depleted on the walls and only about 1% of the aerosol dust is released to environment. The cross-comparison of codes states that largest differences are in the inter-compartmental flows and the in-compartment gas composition. Concerning the capabilities of the codes, CONTAIN has shown to be more robust in dealing with the injection of large aerosol mass flow and with a very small injection of fission products. ASTEC however has an ad-hoc model for aerosol retention in filters.

Topics: Pressure , Helium
Commentary by Dr. Valentin Fuster
2008;():157-165. doi:10.1115/HTR2008-58058.

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.

Commentary by Dr. Valentin Fuster
2008;():167-174. doi:10.1115/HTR2008-58150.

The possibility of fuel and graphite degradation due to chemical attack is a perennial issue for HTR’s. For the direct cycle used in the PBMR design, only air ingress is a problem that merits serious attention. Initially, and as reported at a previous conference, investigation of the problem was tackled by assuming worst case conditions for a break at the core outlet pipe to determine what the grace time would be, before counter measures need be taken. The current work identified worst case break positions, quantifying air ingress rates, assuming a Guillotine break. These calculations include first order corrosion reactions in the bottom reflector and the core. Taking the worst possible large break location and the maximum initial air ingress as a determinant, a period of 24 hours was determined to be sufficient to prevent both serious fuel and core structure degradation. The acceptability of the extent of corrosion will be determined by the Safety Analysis Report (SAR), which is under preparation. However, it was realized that a more realistic specification and analysis of the problem was required to enable design decisions to be made, and a more detailed model of the break and the Main Power System (MPS) cavities was developed. This includes the maximum movement of large piping postulating a Double Ended Guillotine Break (DEGB) at worst possible locations. Further calculations on the improved model are described that investigate the influence of various pipe separations i.e. 50 mm and 500 mm at the turbine inlet. A strong correlation between the opening size and total core corrosion rate was confirmed. The simulation also established an approximate duration for air to be expelled to stop further ingress and the volume flow requirements for the inert gas system using helium or nitrogen.

Topics: Corrosion , Cycles
Commentary by Dr. Valentin Fuster
2008;():175-184. doi:10.1115/HTR2008-58160.

Various countries engaged in the development and fabrication of modern fuel for the High Temperature Gas-Cooled Reactor (HTGR) have initiated activities of modeling the fuel and fission product release behavior with the aim of predicting the fuel performance under operating and accidental conditions of future HTGRs. Within the IAEA directed Coordinated Research Project CRP6 on “Advances in HTGR Fuel Technology Development” active since 2002, the 13 participating Member States have agreed upon benchmark studies on fuel performance during normal operation and under accident conditions. While the former has been completed in the meantime, the focus is now on the extension of the national code developments to become applicable to core heatup accident conditions. These activities are supported by the fact that core heatup simulation experiments have been resumed recently providing new, highly valuable data. Work on accident performance will be — similar to the normal operation benchmark — consisting of three essential parts comprising both code verification that establishes the correspondence of code work with the underlying physical, chemical and mathematical laws, and code validation that establishes reasonable agreement with the existing experimental data base, but including also predictive calculations for future heating tests and/or reactor concepts. The paper will describe the cases to be studied and the calculational results obtained with the German computer model FRESCO. Among the benchmark cases in consideration are tests which were most recently conducted in the new heating facility KUEFA. Therefore this study will also re-open the discussion and analysis of both the validity of diffusion models and the transport data of the principal fission product species in the HTGR fuel materials as essential input data for the codes.

Commentary by Dr. Valentin Fuster
2008;():185-191. doi:10.1115/HTR2008-58185.

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.

Topics: Licensing
Commentary by Dr. Valentin Fuster
2008;():193-203. doi:10.1115/HTR2008-58192.

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.

Topics: Safety
Commentary by Dr. Valentin Fuster
2008;():205-212. doi:10.1115/HTR2008-58218.

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

Commentary by Dr. Valentin Fuster
2008;():213-222. doi:10.1115/HTR2008-58230.

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.

Topics: Helium
Commentary by Dr. Valentin Fuster
2008;():223-231. doi:10.1115/HTR2008-58299.

Depressurized loss of coolant accident (DLOCA) is one of the most important design basis accidents for high temperature gas-cooled reactors. Analysis of the reactor characteristic behavior during DLOCA can provide useful reference to the physics, thermo-hydraulic and structure designs of the reactor core. In this paper, according to the preliminary design of the 250MW Pebble-bed Modular High Temperature Gas-cooled Reactor (HTR-PM), three cases of DLOCA: a instantaneous depressurization along with a flow coastdown and scram at zero time, a main pipe with a diameter of 65mm rupture, and a instrument pipe with a diameter of 10mm broken, are studied by the help of two different kinds of software THERMIX and TINTE. The key parameters of different cases including reactor power, temperature distribution of the core and pressure vessel, and the decay power removal by the passive residual heat remove system (RHRS) are compared in detail. Some uncertainties, such as residual heat calculation, power distribution, heat conductivity of fuel element, etc., are analyzed in order to evaluate the safety margin of the maximum fuel temperature during DLOCA. The calculating results show that, the decay heat in the DLOCA can be removed from the reactor core solely by means of physical processes in a passive way, so that the temperature limits of fuel and components are still obeyed. It also illustrates that the HTR-PM can reach 250MW reactor power per unit and still can keep the inherent safety.

Commentary by Dr. Valentin Fuster
2008;():233-240. doi:10.1115/HTR2008-58307.

A modular gas-cooled reactor design with a thermal output of 600MWt and a core exit temperature of 950° C has been designed by the Korea Atomic Energy Research Institute based on the GT-MHR reactor concept which adopts a prismatic core. A sensitivity study on the transient plant behavior during a postulated depressurized LOFC accident concurrent with the failure of the RCCS was performed. In the transient analysis, the GAMMA+ code which can handle multi-dimensional, multi-component problem was used. The RCCS is a passive system which is very reliable and supplies a significant heat removal mechanism during abnormal conditions in a GCR. To investigate the safety characteristics of a GCR under the one of the worst accidental scenarios, a simultaneous failure of the RCCS with a depressurized LOFC was assumed. The thermal behavior of the reactor system was analyzed in various conditions. It is found that the maximum temperature of the reactor fuel compact could exceed 1600° C at about 50 hours at the condition of a depressurized LOFC with a failure of the RCCS. A problem with the structural integrity of the reactor pressure vessel could also be a critical factor. The insulation of a reactor cavity wall serves as a dominant obstacle against a heat transfer from the reactor vessel to the surrounding ground when the RCCS fails to operate. Without insulation material on the reactor cavity wall, the gradients of the increasing rate of the maximum temperature diminish and the peak values decrease. The maximum temperatures of the fuel compact and the reactor vessel are less sensitive to the concrete and surrounding soil properties, those are the thermal conductivity and volumetric heat capacity, when the insulation material is used. The uncertainties in the properties of the concrete and the surrounding soil become significant without an insulation material in the cavity. To improve the safety of a modular GCR, more effective and feasible heat removal mechanism need to be devised based on the comprehensions on the heat transfer characteristics.

Commentary by Dr. Valentin Fuster
2008;():241-253. doi:10.1115/HTR2008-58317.

Simulation of some fluid phenomena associated with Generation IV reactors requires the capability of modeling mixing in two- or three-dimensional flow. At the same time, the flow condition of interest is often transient and depends upon boundary conditions dictated by the system behavior as a whole. Computational fluid dynamics (CFD) is an ideal tool for simulating mixing and three-dimensional flow in system components, whereas a system analysis tool is ideal for modeling the entire system. This paper presents the reasoning which has led to coupled CFD and systems analysis code software to analyze the behavior of advanced reactor fluid system behavior. In addition, the kinds of scenarios where this capability is important are identified. The important role of a coupled CFD/systems analysis code tool in the overall calculation scheme for a Very High Temperature Reactor is described. The manner in which coupled systems analysis and CFD codes will be used to evaluate the mixing behavior in a plenum for transient boundary conditions is described. The calculation methodology forms the basis for future coupled calculations that will examine the behavior of such systems at a spectrum of conditions, including transient accident conditions, that define the operational and accident envelope of the subject system. The methodology and analysis techniques demonstrated herein are a key technology that in part forms the backbone of the advanced techniques employed in the evaluation of advanced designs and their operational characteristics for the Generation IV advanced reactor systems.

Commentary by Dr. Valentin Fuster
2008;():255-263. doi:10.1115/HTR2008-58318.

The modular Pebble Bed Advanced High Temperature Reactor (PB-AHTR), with a nominal power output of 900 MWth, is the most recent US Berkeley design for a liquid fluoride salt cooled reactor. Due to the high volumetric heat capacity of the primary coolant, the PB-AHTR operates with a high power density core with a similar average coolant temperature as in modular helium reactors. The reactivity control system for the PB-AHTR uses a novel buoyantly-driven shutdown rod system that can be actively or passively activated during reactor transients. In addition to a traditional active insertion mechanism, the new shutdown rod system is designed to also operate passively, fulfilling the role of a reserve shutdown system. The physical response of the shutdown rod was simulated both computationally and experimentally, using scaling arguments where applicable, with an emphasis on key phenomena identified by a preliminary PIRT study. This paper discusses preliminary results from this effort.

Topics: Design
Commentary by Dr. Valentin Fuster
2008;():265-274. doi:10.1115/HTR2008-58336.

The AVR pebble bed reactor (46 MWth) was operated 1967–1988 at coolant outlet temperatures up to 990°C. Also because of a lack of other experience the AVR operation is a basis for future HTRs. This paper deals with insufficiently published unresolved safety problems of AVR and of pebble bed HTRs. The AVR primary circuit is heavily contaminated with dust bound and mobile metallic fission products (Sr-90, Cs-137) which create problems in current dismantling. The evaluation of fission product deposition experiments indicates that the end of life contamination reached several percent of a single core inventory. A re-evaluation of the AVR contamination is performed in order to quantify consequences for future HTRs: The AVR contamination was mainly caused by inadmissible high core temperatures, and not — as presumed in the past — by inadequate fuel quality only. The high AVR core temperatures were detected not earlier than one year before final AVR shut-down, because a pebble bed core cannot be equipped with instruments. The maximum core temperatures were more than 200 K higher than precalculated. Further, azimuthal temperature differences at the active core margin were observed, as unpredictable hot gas currents with temperatures > 1100°C. Despite of remarkable effort these problems are not yet understood. Having the black box character of the AVR core in mind it remains uncertain whether convincing explanations can be found without major experimental R&D. After detection of the inadmissible core temperatures, the AVR hot gas temperatures were strongly reduced for safety reasons. Metallic fission products diffuse in fuel kernel, coatings and graphite and their break through takes place in long term normal operation, if fission product specific temperature limits are exceeded. This is an unresolved weak point of HTRs in contrast to other reactors and is particularly problematic in pebble bed systems with their large dust content. Another disadvantage, responsible for the pronounced AVR contamination, lies in the fact that activity released from fuel elements is distributed in HTRs all over the coolant circuit surfaces and on graphitic dust and accumulates there. Consequences of AVR experience on future reactors are discussed. As long as pebble bed intrinsic reasons for the high AVR temperatures cannot be excluded they have to be conservatively considered in operation and design basis accidents. For an HTR of 400 MWth, 900°C hot gas temperature, modern fuel and 32 fpy the contaminations are expected to approach at least the same order as in AVR end of life. This creates major problems in design basis accidents, for maintenance and dismantling. Application of German dose criteria on advanced pebble bed reactors leads to the conclusion that a pebble bed HTR needs a gas tight containment even if inadmissible high temperatures as observed in AVR are not considered. However, a gas tight containment does not diminish the consequences of the primary circuit contamination on maintenance and dismantling. Thus complementary measures are discussed. A reduction of demands on future reactors (hot gas temperatures, fuel burn-up) is one option; another one is an elaborate R&D program for solution of unresolved problems related to operation and design basis accidents. These problems are listed in the paper.

Topics: Safety
Commentary by Dr. Valentin Fuster

Economics

2008;():275-285. doi:10.1115/HTR2008-58157.

To measure the value of a technology investment under uncertainty with standard techniques like net present value (NPV) or return on investment (ROI) will often uncover the difficulty to present convincing business case. Projected cash flows are inefficient or the discount rate chosen to compensate for the risk is so high, that it is disagreeable to the investor’s requirements. Decision making and feasibility studies have to look beyond traditional analysis to reveal the strategic value of a technology investment. Here, a Real Option Analysis (ROA) offers a powerful alternative to standard discounted cash-flow (DCF) methodology by risk-adjusting the cash flow along the decision path rather than risk adjusting the discount rate. Within the GEN IV initiative attention is brought not only towards better sustainability, but also to broader industrial application and improved financing. Especially the HTR design is full of strategic optionalities: The high temperature output facilitates penetration into other non-electricity energy markets like industrial process heat applications and the hydrogen market. The flexibility to switch output in markets with multi-source uncertainties reduces downside risk and creates an additional value of over 50% with regard to the Net Present Value without flexibility. The supplement value of deploying a modular (V)HTR design adds over 100% to the project value using real option evaluation tools. Focus of this paper was to quantify the strategic value that comes along a) with the modular design; a design that offers managerial flexibility adapting a step-by-step investment strategy to the actual market demand and b) with the option to switch between two modes of operation, namely electricity and hydrogen production. We will demonstrate that the effect of uncertain electricity prices can be dampened down with a modular HTR design. By using a real option approach, we view the project as a series of compound options — each option depending on the exercise of those that preceded it. At each end of the design phase, the viability will be reviewed conditional on the operating spread at each time step. We quantify the value of being able to wait with the investment into a next block until market conditions are favourable and to be able to abandon one block if market conditions are disapproving. To derive the intrinsic value of this multi block HTR design, it will be compared with a reference investment of a full commitment light water reactor without any managerial flexibility. In another case, we raise the question to what extent product output diversification is a suitable strategy to cope with long term market uncertainty in electricity price. What is the value of a multi-potent technology that is able to produce output for energy markets others than the electricity market? To investigate this, we concentrate on The Netherlands, a country with an intense industrial demand in electricity and hydrogen.

Commentary by Dr. Valentin Fuster
2008;():287-292. doi:10.1115/HTR2008-58212.

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

Commentary by Dr. Valentin Fuster
2008;():293-301. doi:10.1115/HTR2008-58251.

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

Commentary by Dr. Valentin Fuster

Experimental Programs and Benchmarking

2008;():303-310. doi:10.1115/HTR2008-58015.

A power-generating unit with the high-temperature helium reactor (GT-MHR) has a turbomachine (TM) that is intended for both conversion of coolant thermal energy into electric power in the direct gas-turbine cycle, and provision of helium circulation in the primary circuit. The vertically oriented TM is placed in the central area of the power conversion unit (PCU). TM consists of a turbocompressor (TC) and a generator. Their rotors are joined with a diaphragm coupling and supported by electro-magnetic bearings (EMB). The complexity and novelty of the task of the full electromagnetic suspension system development requires thorough stepwise experimental work, from small-scale physical models to full-scale specimen. On this purpose, the following is planned within the framework of the GT-MHR Project: investigations of the “flexible” rotor small-scale mockup with electro-magnetic bearings (“Minimockup” test facility); tests of the radial EMB; tests of the position sensors; tests of the TM rotor scale model; tests of the TM catcher bearings (CB) friction pairs; tests of the CB mockups; tests of EMB and CB pilot samples and investigation of the full-scale electromagnetic suspension system as a part of full-scale turbocompressor tests. The rotor scale model (RSM) tests aim at investigation of dynamics of rotor supported by electromagnetic bearings to validate GT-MHR turbomachine serviceability. Like the full-scale turbomachine rotor, the RSM consist of two parts: the generator rotor model and the turbocompressor rotor model that are joined with a coupling. Both flexible and rigid coupling options are tested. Each rotor is supported by one axial and two radial EMBs. The rotor is arranged vertically. The RSM rotor length is 10.54 m, and mass is 1171 kg. The designs of physical model elements, namely of the turbine, compressors, generator and exciter, are simplified and performed with account of rigid characteristics, which are identical to those of the full-scale turbomachine elements.

Commentary by Dr. Valentin Fuster
2008;():311-318. doi:10.1115/HTR2008-58043.

The experimental program that is being conducted at the Matched Index-of-Refraction (MIR) Flow Facility at Idaho National Laboratory (INL) to obtain benchmark data on measurements of flow phenomena in a scaled model of a prismatic gas-cooled reactor lower plenum using 3-D Particle Image Velocimetry (PIV) is presented. A description of the scaling analysis, experimental facility, 3-D PIV system, measurement uncertainties and analysis, experimental procedures and samples of the data sets that have been obtained are included. Samples of the data set that will be presented include the mean velocity field in an approximately 1:7 scale model of a region of the lower plenum of a typical prismatic gas-cooled reactor (GCR) similar to a General Atomics Gas-Turbine-Modular Helium Reactor (GTMHR) design. This experiment has been selected as the first Standard Problem endorsed by the Generation IV International Forum. The flow in the lower plenum consists of multiple jets injected into a confined cross flow — with obstructions. The model consists of a row of full circular posts along its centerline with half-posts on the two parallel walls to approximate flow scaled to that expected from the staggered parallel rows of posts in the reactor design. The model is fabricated from clear, fused quartz to match the refractive-index of the mineral oil working fluid. The benefit of the MIR technique is that it permits high-quality measurements to be obtained without locating intrusive transducers that disturb the flow field and without distortion of the optical paths. An advantage of the INL MIR system is its large size which allows obtaining improved spatial and temporal resolution compared to similar facilities at smaller scales. Results concentrate on the region of the lower plenum near its far reflector wall (away from the outlet duct). Inlet jet Reynolds numbers (based on the jet diameter and the time-mean average flow rate) are approximately 4,300 and 12,400. The measurements reveal developing, non-uniform flow in the inlet jets and complicated flow patterns in the model lower plenum. Data include three-dimensional vector plots, data displays along the coordinate planes (slices) and charts that describe the component flows at specific regions in the model. Information on inlet velocity profiles is also presented.

Commentary by Dr. Valentin Fuster
2008;():319-322. doi:10.1115/HTR2008-58063.

The nuclear core of High Temperature Gas Reactor (HTGR) with pebble bed type has been investigated intensively due to its benefits in management, but its complicated flow geometry requested the reliable analytical method. Recent studies have been made using the three dimensional computational methods but they need to be evaluated with the experimental data. Due to the complicated and narrow flow channel, the intrusive methods of flow measurement are not proper in the study. In the present study, we developed a wind tunnel for the pebble bed geometry in the structure of Face Centered Cubic (FCC) and measure the flow field using the Particle Image Velocimetry (PIV) directly. Due to the limitation of the image harnessing speed and accessibility of the light for particle identification, the system is scaled up to reduce the mean flow velocity by keeping the same Reynolds number of the HTGR. The velocity fields are successfully determined to identify the stagnation points suspected to produce hot spots on the surface of the pebble. It is expected that the present data is useful to evaluate the three dimensional Computational Fluid Dynamics (CFD) analysis. Furthermore, It would provide an insight of experimental method if the present results are compared by those of scaled down and liquid medium.

Commentary by Dr. Valentin Fuster
2008;():323-328. doi:10.1115/HTR2008-58098.

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 × 1025 n/m2 ; (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 × 1025 n/m2 , respectively. The fuel compact peak burnup reached 16.68% FIMA with corresponding fast neutron fluence for that fuel compact of 5.06 × 1025 n/m2 , 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.

Commentary by Dr. Valentin Fuster
2008;():329-335. doi:10.1115/HTR2008-58107.

The paper presents a description of benchmark cases, achieved results, analysis of possible reasons of differences of calculation results obtained by various neutronic codes. The comparative analysis is presented showing the benchmark–results obtained with reference and design codes by Russian specialists (WIMS-D, JAR-HTGR, UNK, MCU, MCNP5-MONTEBURNS1.0-ORIGEN2.0), by French specialists (APOLLO2, TRIPOLI4 codes), and by Korean specialists (HELIOS, MASTER, MCNP5 codes). The analysis of possible reasons for deviations was carried out, which was aimed at the decrease of uncertainties in calculated characteristics. This additional investigation was conducted with the use of 2D models of a fuel assembly cell and a reactor plane section.

Commentary by Dr. Valentin Fuster
2008;():337-346. doi:10.1115/HTR2008-58127.

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.

Commentary by Dr. Valentin Fuster
2008;():347-353. doi:10.1115/HTR2008-58134.

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.

Topics: Fuels
Commentary by Dr. Valentin Fuster
2008;():355-361. doi:10.1115/HTR2008-58155.

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.

Commentary by Dr. Valentin Fuster
2008;():363-371. doi:10.1115/HTR2008-58164.

Within the framework of the ANTARES program, AREVA NP, EDF and the CEA have launched a joint R&D program on metallic materials for VHTR. Reference alloys for circuit and Intermediate Heat eXchanger (IHX) are nickel-based with about 22%wt. of chromium. Compatibility with the HTR primary helium appears to be a determining property for the material selection and qualification. The coolant is actually polluted by a low level of impurities that can interact with metals at high temperature. Oxidation, carburization and/or decarburization occur, in relation to atmosphere characteristics, temperature and alloy chemical composition. As these corrosion effects can notably influence the mechanical properties, they often are determining to the component service life. Since the corrosion behavior is highly sensitive to environmental conditions, material studies require dedicated facilities that shall allow for a strict control of the environment throughout the entire specimen exposure. AREVA NP, CEA, and EDF have developed experimental loops respectively under the names the Chemistry Loop, CORINTH and CORALLINE, ESTEREL; these high temperature helium flow systems are equipped with high accuracy hygrometers and gas analyzers. A benchmark was defined to cross-validate the lab devices and procedures. It is composed of two tests. The joint protocol sets the operating parameters in terms of material, specimen preparation, temperature and heating program, gas pressure and flow rate, time, gas composition. Corrosion is assessed by mass change associated to observations and analyses of the corroded coupons considering the surface scales (nature, morphology and thickness), the internal oxidation (nature, distribution and depth) and the possible carburization/decarburization (type and depth). For benchmark test 1, AREVA NP, CEA, and EDF produced similar results in terms of operation of the tests as well as about the corrosion criteria. On the other hand, benchmark test 2 showed a difference in the residual water vapor level between loops that was shown to strongly influence the specimen behavior. Discrepancies in the alloy corrosion were studied regarding gas flow rates and effective oxygen potential in helium. As a consequence, the experimental tools and procedures have been upgraded. French laboratories have now efficient corrosion facilities and methods at their disposal to study and qualify the corrosion behavior of structural materials in HTR environment.

Topics: Corrosion
Commentary by Dr. Valentin Fuster
2008;():373-380. doi:10.1115/HTR2008-58206.

The Very High Temperature Reactor (VHTR) is the leading candidate for the reactor component of the Next Generation Nuclear Plant (NGNP). This is because the VHTR demonstrates great potential in improving safety characteristics, being economically competitive, providing a high degree of proliferation resistance, and producing high outlet temperatures for efficient electricity generation and/or other high temperature applications, most notably hydrogen production. In addition, different fuel types can be utilized by VHTRs, depending on operational goals. In this case, the recovery and utilization of the valuable energy left in LWR fuel in order to create ultra long life single batch cores by taking advantage of the properties of TRU fuels. This paper documents the initial process in the study of TRU fueled VHTRs, which concentrates on the verification and validation of the developed whole-core 3D VHTR models. Many of the codes used for VHTR analysis were developed without a full appreciation of the importance of randomness in particle distribution. With this in mind, the SCALE code system was chosen as the computational tool for the study. It provides the opportunity of utilizing SCALE versions 5.0 and 5.1, making it possible to compare and analyze different techniques accounting for the double heterogeneity effects associated with VHTRs. Startup physics results for Japan’s High Temperature Test Reactor (HTTR) were used for experiment-to-code benchmarking. MCNP calculations were employed for code-to-code benchmarking. Results and analysis are included in this paper.

Commentary by Dr. Valentin Fuster
2008;():381-390. doi:10.1115/HTR2008-58214.

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.

Commentary by Dr. Valentin Fuster
2008;():391-404. doi:10.1115/HTR2008-58258.

Since the last decade, Tractebel Engineering has been involved in several consecutive projects in the field of High Temperature Gas Reactor (HTGR). The objectives of the present project called RAPHAEL (www.raphael-project.org ) is to provide R&D results in order to consolidate available data on generic V/HTR technologies and to develop innovative solutions to further contribute to the improvement of HTR performances. One of the objectives of the RAPHAEL Sub-project Safety is to qualify tools for performing safety analyses and supporting the safety approach and demonstration. One of the work packages concerns the validation of the existing thermal-hydraulic system codes capabilities needed to perform transient analysis in V/HTR. This validation is carried out by benchmarking against experimental data and by comparing simulation results given by several codes. The current paper presents the work performed at Tractebel Engineering on the simulation of the HE-FUS3 experimental loop — ENEA facility, Brasimone (Italy) — with the MELCOR v.1.8.6 code. The HE-FUS3 loop contains a wide range of components characteristic of a V/HTR like compressor, pipes, diffusers, valves, heaters and heat exchangers. Even if the loop characteristics/configuration is not prototypical of a V/HTR design, the loop is useful to assess the objectives identified by the Project, i.e. helium operating fluid, design pressure and temperature set respectively at 10.5 MPa and 530 °C. The experimental data of the HE-FUS3 loop made available for the benchmark are a set of steady state tests for the thermal-hydraulic characterization of the loop and two transient tests — Loss Of Flow Accidents (LOFA). Moreover, to assess the characteristics of the compressor, data have also been provided from a compressor test campaign. From the code-to-experiment comparison the ability of MELCOR v.1.8.6 to reproduce the experimental results is judged.

Commentary by Dr. Valentin Fuster
2008;():405-412. doi:10.1115/HTR2008-58282.

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.

Commentary by Dr. Valentin Fuster
2008;():413-419. doi:10.1115/HTR2008-58330.

In the Advanced Gas Cooled Pebble Bed Reactors for nuclear power generation, the fuel is spherical coated particles. The energy transfer phenomenon requires detailed understanding of the flow and temperature fields around the spherical fuel pebbles. Detailed information of the complex flow structure within the bed is needed. Generally, for computing the flow through a packed bed reactor or column, the porous media approach is usually used with lumped parameters for hydrodynamic calculations and heat transfer. While this approach can be reasonable for calculating integral flow quantities, it may not provide all the detailed information of the heat transfer and complex flow structure within the bed. The present experimental study presents the full velocity field using particle image velocity technique (PTV) in a conjunction with matched refractive index fluid with the pebbles to achieve optical access. Velocity field measurements are presented delineating the complex flow structure.

Commentary by Dr. Valentin Fuster
2008;():421-426. doi:10.1115/HTR2008-58331.

In this investigation Particle Image Velocimetry technique was implemented to a matched refractive index facility which was placed in a rectangular channel of L:1016 mm×W:76.2 mm×H:76.2 mm. Water was pumped into either one or both of the inlet jets which were entering the channel’s top wall with several different Reynolds numbers. The instantaneous and time-resolved velocity fields were successfully obtained from which several flow characteristics such as vorticity, turbulence instabilities and Reynolds stresses can be calculated.

Topics: Jets , Fuel rods
Commentary by Dr. Valentin Fuster

Industrial Process Heat Applications

2008;():427-433. doi:10.1115/HTR2008-58216.

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.

Commentary by Dr. Valentin Fuster
2008;():435-443. doi:10.1115/HTR2008-58259.

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.

Topics: Heat
Commentary by Dr. Valentin Fuster

Application to Production of Hydrogen

2008;():445-453. doi:10.1115/HTR2008-58007.

The program for hydrogen production with high temperature nuclear heat has been launched in Korea since 2004. Iodine sulfur (IS) process is one of the promising processes for a hydrogen production because it does not generate a carbon dioxide and massive hydrogen production may be possible. However, the highly corrosive environment of the process is barrier to the application in the industry. Therefore, corrosion behaviors of various materials were evaluated in sulfuric acid to select appropriate materials compatible with the IS process. The materials used in this work were Ni base alloys, Fe-Si alloys, Ta, Au, Pt, Zr, SiC and so on. The test environments were boiling 50wt% sulfuric acid without/with HI as an impurity and 98wt% sulfuric acid. The surface morphologies and cross sectional areas of the corroded materials were examined by using SEM equipped with EDS. From the results of the weight loss and potentiodynamic experiments, it was found that a Si enriched oxide is attributable to a corrosion resistance for materials including Si in boiling 98wt% sulfuric acid. Moreover the passive Si enriched film thickness increased with the immersion time leading to an enhancement of the corrosion resistance. Corrosion behaviours of the material tested are discussed in terms of the chemical composition of the materials, a corrosion morphology and the surface layer’s composition.

Topics: Corrosion
Commentary by Dr. Valentin Fuster
2008;():455-458. doi:10.1115/HTR2008-58009.

The key interface component between the reactor and chemical systems for the sulfuric acid based processes to make hydrogen is the sulfuric acid decomposition reactor. The materials issues for the decomposition reactor are severe since sulfuric acid must be heated, vaporized and decomposed. SiC has been identified and proven by others to be an acceptable material. However, SiC has a significant design issue when it must be interfaced with metals for connection to the remainder of the process. Westinghouse has developed a design utilizing SiC for the high temperature portions of the reactor that are in contact with the sulfuric acid and polymeric coated steel for low temperature portions. This design is expected to have a reasonable cost for an operating lifetime of 20 years. It can be readily maintained in the field, and is transportable by truck (maximum OD is 4.5 meters). This paper summarizes the detailed engineering design of the Westinghouse Decomposition Reactor and the decomposition reactor’s capital cost.

Topics: Design , Hydrogen , Sulfur
Commentary by Dr. Valentin Fuster
2008;():459-463. doi:10.1115/HTR2008-58011.

In order to establish the optimal start-up method and to understand the dynamic behavior of the SI process coupled to a VHTR through an Intermediate Heat Exchanger (IHX), the development of a dynamic simulation code is necessary. Perturbation of the flow rate or temperature in input streams may result in various transient states. An understanding of the dynamic behavior due to these factors is very important to establish a safe operation method for a hydrogen production plant including a VHTR. Based on the mass and energy balance sheets of an electrodialysis-embedded SI process, proposed by KAERI, equivalent to a 200 MWth VHTR, not only the establishment of a thermal pathway (draft) between the SI process and the VHTR system but also the equipment sizing required in the SI process was carried out. A dynamic simulation code for the SI –2nd Skid (sulfuric acid concentration and decomposition part) was prepared at first for each chemical reactor. The reliability of the computer code has been confirmed by the convergence value at a steady state. This confirmation has been performed for the primary and secondary sulfuric acid distillation columns, the sulfuric acid vaporizer, the sulfuric acid decomposer, and the sulfur trioxide decomposer, respectively. An integrated computer code with a visualization function has been prepared by coupling each proven computer code, according to the thermal pathway. The dynamic behaviors of the integrated the SI-2nd Skid according to several anticipated scenarios were evaluated and the dominant and mild factors are discussed.

Commentary by Dr. Valentin Fuster
2008;():465-472. doi:10.1115/HTR2008-58022.

The utilization of alternate sources of energy is becoming more important due to the constantly growing world-wide demand for energy. The production of hydrogen via the Hybrid Sulphur process is a possible alternative that may contribute to alleviating the pressure on energy resources. The current field of interest is to investigate the operation of the sulphuric acid decomposition reactor operating at pressure ranges between 8 and 9 MPa. The reduction of SO3 to SO2 is, however, favoured at low pressures while maintaining high operating temperatures. Considering this, the need to investigate the possibility of operating at lower operating pressures is important in striving for higher process efficiencies. The proposed decomposition reactor is a multi-stage reactor system operated adiabatically with inter-stage heating in order to simplify the reactor design and improve the over-all conversion and efficiency of the process. At a pressure of 8–9 MPa and temperature of 900°C, the maximum conversion of SO3 to SO2 that can be achieved is between 48% and 54%. The proposed multi-stage reactor system has 5 packed bed (catalyst) reactor stages with 4 intermediate heat exchangers, and by lowering the operating pressure to 3 kPa, a maximum conversion of 72% could be achieved. The viability of the HyS process mainly depends on the performance of the SO3 decomposition reactor.

Commentary by Dr. Valentin Fuster
2008;():473-479. doi:10.1115/HTR2008-58023.

The world’s energy consumption is increasing constantly due to the growing population of the world. The increasing energy consumption has a negative effect on the fossil fuel reserves of the world. Hydrogen has the potential to provide energy for all our needs by making use of fossil fuel such as natural gas and nuclear-based electricity. Hydrogen can be produced by reforming methane with carbon dioxide as the oxidizing agent. Hydrogen can be produced in a Plasma-arc reforming unit making use of the heat energy generated by a 500 MWt Pebble Bed Modular Reactor (PBMR). The reaction in the unit takes place stoichiometrically in the absence of a catalyst. Steam can be added to the feed stream together with the Carbon Dioxide, which make it possible to control the H2 /CO ratio in the synthesis gas between 1/1 and 3/1. This ratio of H2 /CO in the synthesis gas is suitable to be used as feed gas to almost any chemical and petrochemical process. To increase the hydrogen production further, the Water-Gas Shift Reaction can be applied. A techno-economic analysis was performed on the non-catalytic plasma-arc reforming process. The capital cost of the plant is estimated at $463 million for the production of 1132 million Nm3 /year of hydrogen. The production cost of hydrogen is in the order of $12.81 per GJ depending on the natural gas cost and the price of electricity.

Commentary by Dr. Valentin Fuster
2008;():481-490. doi:10.1115/HTR2008-58044.

One of the key technology requirements to achieve the nuclear hydrogen demonstration is the establishment of control scheme which harmonizes the reactor operation with chemical plant operation. This study focused on developing the control scheme to be considered in the HTTR-IS nuclear hydrogen production system, which are in case of (a) abnormal shut-down and (b) restart-up of the IS process. The key parameters and equipments have been determined and the control operations are simulated. The simulation results show that the impact of abnormal events initiated in the IS process on the reactor operation can be effectively minimized by the rapid initiation of diverter valves using model-based fault diagnosis method. Furthermore, the thermal shock to components in the IS process can be prevented in case of restart-up operation of the IS process by controlling the helium gas flow-rate of primary and secondary cooling system. It is confirmed by the analysis that the control scheme developed enable to maintain the reactor operation normally under all conditions and supply heat from nuclear reactor to the IS process hydrogen production system flexibly.

Commentary by Dr. Valentin Fuster
2008;():491-495. doi:10.1115/HTR2008-58084.

The copper-chlorine (Cu-Cl) thermochemical cycle uses both heat and electricity to carry out a series of chemical and electrochemical reactions with the net reaction being the splitting of water into hydrogen and oxygen. The process forms a closed loop with all intermediate chemicals being recycled. All of the chemical and electrochemical reactions can be carried out at temperatures that do not exceed about 530°C. Thus, the heat requirement of this process can be satisfied by intermediate temperature nuclear reactors such as the Super Critical Water Reactor (SCWR) developed in Canada by Atomic Energy of Canada Limited (AECL). AECL is particularly interested in developing the electrochemical reactions that comprise the Cu-Cl cycle. There are two variations on the Cu-Cl cycle. In the original cycle copper metal is produced electrochemically by the disproportionation of cuprous chloride (CuCl), which is dissolved in hydrochloric acid (HCl) electrolyte. It is expected that this reaction will be carried out at a temperature that is below 100°C. Hydrogen gas is then produced by a chemical reaction that takes place between the copper metal and gaseous HCl at a temperature of 430–475°C. It was recognized by AECL that these two reaction steps could be replaced by a single electrochemical reaction that generates hydrogen directly. It is expected that this step will also be carried out at a temperature below 100°C. In this process, referred to as the CuCl/HCl electrolysis step, hydrogen gas is produced at the cathode of an electrochemical cell by the reduction of protons that are supplied by aqueous 6 M HCl while cupric chloride (CuCl2 ) is produced at the anode by the oxidation of CuCl, which is dissolved in 6 M HCl. The CuCl2 that is formed is recycled and is used in a reaction with steam at 400°C to produce a copper oxychloride. This reaction is common to both versions of the Cu-Cl cycle. It is the purpose of this paper to present electrochemical results from both half-cell and single-cell studies carried out to verify and understand the CuCl/HCl electrolysis step. Half-cell electrochemical data is presented that demonstrates the practicality of the electrode reactions. Electrochemical data is presented to show that the CuCl/HCl electrolysis step can be carried out in a single-cell. In both the half-cell and single-cell experiments platinum electrocatalysts are used to carry out the desired reactions.

Commentary by Dr. Valentin Fuster
2008;():497-508. doi:10.1115/HTR2008-58086.

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).

Commentary by Dr. Valentin Fuster
2008;():509-516. doi:10.1115/HTR2008-58088.

A project using the High Temperature Engineering Test Reactor (HTTR) is being proceeded at the Japan Atomic Energy Agency (JAEA), which aims to develop technologies of nuclear production of hydrogen to meet its massive demand in future. The HTTR with thermal power of 30MW is the Japanese high-temperature gas-cooled reactor (HTGR), built and operated at the site of the Oarai Research & Development Center of JAEA. Under the HTTR project, JAEA has been conducting research and development on thermochemical IS process aiming to reach its final goal of demonstrating the nuclear hydrogen production by an HTTR-IS system. The IS process is a chemical process composed of three reactions, i.e. hydrogen iodide decomposition, sulfuric acid decomposition, and production of hydrogen iodide and sulfuric acid. JAEA completed one-week continuous hydrogen production using a glass-made bench-scale test apparatus equipped with an automatic control system in 2004, where the hydrogen production rate was about 30 NL/hr. Based on the successful test, present activity focuses on the development of components to be used in the corrosive process environments and, also, of technologies to realize high thermal efficiency of hydrogen production. This paper describes present status of the activities such as conceptual design of a test plant driven by sensible heat of helium gas heated by an electric heater simulating the HTTR, and component tests for service in the corrosive environments of high temperature sulfuric acid. Also, R&D results are described on glass lining technologies and a reliability evaluation of components made of SiC ceramics, which are promising candidates for application to, e.g. the sulfuric acid decomposer and the hydriodic acid distillation column.

Commentary by Dr. Valentin Fuster
2008;():517-526. doi:10.1115/HTR2008-58191.

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.

Commentary by Dr. Valentin Fuster
2008;():527-529. doi:10.1115/HTR2008-58196.

The Hybrid Sulfur Process is a leading candidate among the thermochemical cycles being developed to use heat from advanced nuclear reactors to produce hydrogen via watersplitting. It has the potential for high efficiency, competitive cost of hydrogen, and it has been demonstrated at a laboratory scale to confirm performance characteristics. The major developmental issues with the HyS Process involve the design and performance of a sulfur dioxide depolarized electrolyzer, the key component for conducting the electrochemical step in the process. This paper will discuss the development program and current status for the SDE being conducted at the Savannah River National Laboratory.

Commentary by Dr. Valentin Fuster
2008;():531-532. doi:10.1115/HTR2008-58207.

The Hybrid Sulfur (HyS) Process is being developed to produce hydrogen by water-splitting using heat from advanced nuclear reactors. It has the potential for high efficiency and competitive hydrogen production cost, and has been demonstrated at a laboratory scale.

Commentary by Dr. Valentin Fuster
2008;():533-540. doi:10.1115/HTR2008-58223.

High-temperature reactors are a potential low-carbon source of high-temperature heat for chemical plants—including hydrogen production plants and refineries. Unlike electricity, high temperature heat can only be transported limited distances; thus, the reactor and chemical plants will be close to each other. A critical issue is to understand potential safety challenges to the reactor from the associated chemical plant events to assure nuclear plant safety. The U.S. Nuclear Regulatory Commission (NRC) recently sponsored a Phenomena Identification and Ranking Table (PIRT) exercise to identify potential safety-related physical phenomena for high-temperature reactors coupled to a hydrogen production or similar chemical plant. The ranking process determines what types of chemical plant transients and accidents could present the greatest risks to the nuclear plant and thus the priorities for safety assessments. The assessment yielded four major observations. Because the safety philosophy for most chemical plants (dilution) is different than the safety philosophy for nuclear power plants (containment), this difference must be recognized and understood when considering safety challenges to a nuclear reactor from coupled chemical plants or refineries. Accidental releases of hydrogen from a hydrogen production facility are unlikely to be a major hazard for the nuclear plant assuming some minimum separation distances. Many chemical plants under accident conditions can produce heavy ground-hugging gases such as oxygen, corrosive gases, and toxic gases that can have major off-site consequences because of the ease of transport from the chemical plant to off-site locations. Oxygen presents a special concern because most proposed nuclear hydrogen processes convert water into hydrogen and oxygen; thus, oxygen is the primary byproduct. These types of potential accidents must be carefully accessed. Last, the potential consequences of the failure of the intermediate heat transport loop that moves heat from the reactor to the chemical plant must be carefully assessed.

Commentary by Dr. Valentin Fuster
2008;():541-549. doi:10.1115/HTR2008-58225.

Sandia National Laboratories (SNL), General Atomics Corporation (GA) and the French Commissariat à 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) H2 SO4 is catalytically decomposed at high temperature to produce SO2 , O2 and H2 O. 2) The SO2 is reacted with H2 O and I2 to produce HI and H2 SO4 . The H2 SO4 is returned to the acid decomposer. 3) The HI is decomposed to H2 and I2 . The I2 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 H2 SO4 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 H2 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 H2 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.

Commentary by Dr. Valentin Fuster
2008;():551-557. doi:10.1115/HTR2008-58245.

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.

Commentary by Dr. Valentin Fuster
2008;():559-570. doi:10.1115/HTR2008-58305.

The Sulfur-Iodine thermochemical cycle offers a promising approach to the high efficiency production of hydrogen from nuclear power. Several SI cycles have been proposed by several research group. General Atomic (GA) studied I2 separation by extractive distillation using H3 PO4 . RWTH introduced the concept of reactive distillation. In this process, HIx stream coming from the Bunsen reaction is fed to the column. And HIx is distillated and decomposed at the same time to obtain hydrogen. Korea Institute of Energy Research (KIER) and Japan Atomic Energy Agency (JAEA) concentrate HIx using electro-dialysis cell and concentrated HIx is fed to the column to produce HI vapor, which is decomposed to produce hydrogen. HI was separated from HIx solution by an extractive distillation using H3 PO4 . However, a large amount of electric energy was required to recycle H3 PO4 . Most of SI processes have difficulties producing hydrogen because it has excess iodine in HI decomposition Section. SI cycle with electrodialysis cell uses membrane reactor to separate H2 and HIx. The current state of the membrane technology is not compatible with the process needs. This study examined several cases of flowsheets to overcome the problems mentioned above. The flowsheets were revised by adding the iodine separator and excluding membrane reactor. The thermal efficiency of SI process was analyzed using the revised flowsheet.

Commentary by Dr. Valentin Fuster

Application to Synfuels Production (BTL, CTL, KTL)

2008;():571-577. doi:10.1115/HTR2008-58085.

A system analysis has been performed to assess the efficiency and carbon utilization of a nuclear-assisted coal gasification process. The nuclear reactor is a high-temperature helium-cooled reactor that is used primarily to provide power for hydrogen production via high-temperature electrolysis. The supplemental hydrogen is mixed with the outlet stream from an oxygen-blown coal gasifier to produce a hydrogen-rich gas mixture, allowing most of the carbon dioxide to be converted into carbon monoxide, with enough excess hydrogen to produce a syngas product stream with a hydrogen/carbon monoxide molar ratio of about 2:1. Oxygen for the gasifier is also provided by the high-temperature electrolysis process. Results of the analysis predict 90.5% carbon utilization with a syngas production efficiency (defined as the ratio of the heating value of the produced syngas to the sum of the heating value of the coal plus the high-temperature reactor heat input) of 64.4% at a gasifier temperature of 1866 K for the high-moisture-content lignite coal considered. Usage of lower moisture coals such as bituminous can yield carbon utilization approaching 100% and 70% syngas production efficiency.

Commentary by Dr. Valentin Fuster
2008;():579-588. doi:10.1115/HTR2008-58132.

The rapidly increasing cost of petroleum products and uncertainty of long-term supply have prompted the U.S. military to aggressively pursue production of alternative fuels (synfuels) such as coal-to-liquids (CTL). U.S. Air Force is particularly active in this effort while the entire military is involved in simultaneously developing fuel specifications for alternative fuels that enable a single fuel for the entire battle space; all ground vehicles, aircraft and fuel cells. By limiting its focus on coal, tar sands and oil shale resources, the military risks violating federal law which requires the use of synfuels that have lifecycle greenhouse gas emissions less than or equal to emissions from conventional petroleum fuels. A climate-friendly option would use a high temperature nuclear reactor to split water. The hydrogen (H2 ) would be used in the reverse water gas shift (RWGS) to react with carbon dioxide (CO2 ) to produce carbon monoxide (CO) and water. The oxygen (O2 ) would be fed into a supercritical (SC) coal furnace. The flue gas CO2 emissions would be stripped of impurities before reacting with H2 in a RWGS process. Resultant carbon monoxide (CO) is fed, with additional H2 , (extra H2 needed to adjust the stoichiometry: 2 moles H2 to one mole CO) into a conventional Fischer-Tropsch synthesis (FTS) to produce a heavy wax which is cracked and isomerized and refined to Jet Propulsion 8 (JP-8) and Jet Propulsion 5 (JP-5) fuels. The entire process offers valuable carbon-offsets and multiple products that contribute to lower synfuel costs and to comply with the federal limitation imposed on synfuel purchases. While the entire process is not commercially available, component parts are being researched; their physical and chemical properties understood and some are state-of-the-art technologies. An international consortium should complete physical, chemical and economic flow sheets to determine the feasibility of this concept that, if pursued, has broad applications to military and civilian aviation fleets and freight-hauling diesel engines.

Commentary by Dr. Valentin Fuster
2008;():589-596. doi:10.1115/HTR2008-58226.

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.

Commentary by Dr. Valentin Fuster
2008;():597-611. doi:10.1115/HTR2008-58239.

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 CO2 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 CO2 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.

Commentary by Dr. Valentin Fuster

Waste Management

2008;():613-618. doi:10.1115/HTR2008-58025.

The molded block fuel element (FE) also called monolith is a molded body, consisting of a substantially isotropic highly crystalline graphite matrix, fuel regions within the same matrix and cooling channels. The fuel regions contain the fuel in the form of coated particles which are well bonded to the remaining graphite matrix, so that both parts of the block form a monolithic structure. The monolith meets the requirements for the very high temperature reactors attaining helium outlet temperatures above 1000°C. To fabricate the molded blocks FE demonstration plant was erected and put into operation. The equipment worked without malfunction. The produced block FEs meet the specifications of GA machined block FEs. All specimens and block segments irradiated at temperature up to 1600°C and max. fast fluence E > 0, 1 MeV of 11×1021 n/cm2 show perfect behaviour without any damage.

Commentary by Dr. Valentin Fuster
2008;():619-622. doi:10.1115/HTR2008-58026.

Following the fabrication technique originally developed for HTR-molded block fuel elements a process was introduced to fabricate nuclear graphite with the aim to attain the improved irradiation stability above 3 × 1022 n/cm2 , E > 0,1 MeV and to increase corrosion resistance. Nuclear highly crystalline natural graphite is used. A phenol formaldehyde resin with additives of silicon or zirconium oxide powder serves as binder. The mixture thus obtained is isostatically consolidated into spheres and spheres are crushed to granules from which the 0.3 – 3 mm fraction is obtained. The granulate is hot molded into graphite bodies. The green bodies are heated to about 800 °C to carbonise the resin and subsequently annealed at 1900°C in vacuum. The key feature of the proposed process is based on the chemical affinity of binder coke with the structure obtained by carbonisation of green bodies. Consequently it reacts selectively in situ with the added SiO2 , or ZrO2 to carbides in vacuum at 1900°C. Silicon carbides and zirconium carbides are characterised by high mechanical strength and very good resistance to corrosion. The properties of reactor graphite, such as density, mechanical properties and in particular stability to fast neutron irradiation are considerably improved.

Topics: Graphite , Silicon , Zirconium
Commentary by Dr. Valentin Fuster
2008;():623-630. doi:10.1115/HTR2008-58047.

South Africa is planning to expand its nuclear power generating capacity by deploying a number of pressurized-water reactors and pebble-bed modular reactors. It can be expected that this program will impact on the current and planned spent fuel and radioactive waste management systems in South Africa. This paper proposes an approach to develop a strategy for the management of PBMR spent fuel that would contribute to the optimization of the overall national radwaste management system. The approach is expected to provide a conceptual spent fuel management strategy and will also highlight areas that need to be further developed, thus providing guidance for basic technology development.

Commentary by Dr. Valentin Fuster
2008;():631-638. doi:10.1115/HTR2008-58119.

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

Commentary by Dr. Valentin Fuster
2008;():639-648. doi:10.1115/HTR2008-58170.

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

Commentary by Dr. Valentin Fuster
2008;():649-657. doi:10.1115/HTR2008-58177.

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

Commentary by Dr. Valentin Fuster
2008;():659-668. doi:10.1115/HTR2008-58233.

Generation IV Very High Temperature Reactors (VHTRs) are well-known for their flexibility with respect to feasible fuel cycle options. In this paper, the LEU- and TRU-fueled VHTR configurations are analyzed accounting for their capabilities to attain an extended single-batch OTTO (Once-Through-Then – Out) mode of operation without intermediate refueling. The requirement of waste minimization is imposed as one of the design constraints defining possible system configurations and deployment strategies. The resulting “used fuel” vectors are examined considering anticipated disposal options as well as viability of fuel reprocessing. A Monte Carlo-deterministic analysis methodology has been implemented for coupled design studies of VHTRs with TRUs using the ORNL SCALE 5.1 code system. The developed modeling approach provides an exact-geometry 3D representation of the VHTR core details properly capturing VHTR physics. The presented analysis is focused on prismatic block core concepts for VHTRs. It is being performed within the scope of the U.S. DOE NERI project on utilization of higher actinides (TRUs and partitioned MAs) as a fuel component for extended-life VHTR configurations.

Commentary by Dr. Valentin Fuster
2008;():669-675. doi:10.1115/HTR2008-58247.

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

Commentary by Dr. Valentin Fuster
2008;():677-682. doi:10.1115/HTR2008-58280.

A new European Project has been launched in April 2008 under the 7th EURATOM Framework Programme (FP7-211333), with a duration of four years, addressing the ‘Treatment and Disposal of Irradiated Graphite and other Carbonaceous Waste (CARBOWASTE)’. The objective of this project is the development of best practices in the retrieval, treatment and disposal of irradiated graphite & carbonaceous waste-like structural material e.g. non-graphitised carbon bricks and fuel coatings (pyrocarbon, silicon carbide). It addresses both legacy waste as well as waste from future generations of graphite-based nuclear fuel. After defining the various targets for an integrated waste management, comprehensive analysis of the key stages from in-reactor storage to final disposal will then be undertaken with regard to the most economic, environmental and sustainable options. This will be supported by a characterisation programme to localize the contamination in the microstructure of the irradiated graphite and so more to better understand their origin and the release mechanisms during treatment and disposal. It has been discovered that a significant part of the contamination (including 14 C) can be removed by thermal, chemical or even microbiological treatment. The feasibility of the associated processes will be experimentally investigated to determine and optimise the decontamination factors. Reuse of the purified material will also be addressed to close the ‘Graphite Cycle’ for future graphite moderated reactors. The disposal behaviour of graphite and carbonaceous wastes and the improvement of suitable waste packages will be another focus of the programme. The CARBOWASTE project is of major importance for the deployment of HTR as each HTR module generates (during a 60 years operational lifetime) about 5,000 to 10,000 metric tonnes of contaminated graphite containing some Peta-Becquerel of radiocarbon. It is strongly recommended to take decommissioning and waste management issues of graphite-moderated reactors already into account when designing new HTR concepts.

Topics: Graphite
Commentary by Dr. Valentin Fuster

Poster Presentations

2008;():683-689. doi:10.1115/HTR2008-58040.

A fuel performance analysis code for a very high temperature gas-cooled reactor (VHTR) COPA (Co ated Pa rticle) is being developed at the Korea Atomic Energy Research Institute (KAERI). The COPA code consists of nine modules: BURN, TEMTR, TEMPEB, TEMBL, MECH, FAIL, FPREL, ABAQ, and MPRO. The BURN determines neutron flux and fluence at a location of a reactor core, and then calculates a fuel burnup, a fission rate per volume and a fission product inventory throughout a fuel particle and a fuel element. The TEMTR, TEMPEB and TEMBL calculate the temperature distributions in a coated fuel particle, a pebble and a fuel block by using a one-dimensional finite difference method, respectively. The MECH performs mechanical calculations on a coated fuel particle by using a finite element method. The FAIL performs probabilistic calculations to estimate the failure probabilities of the coating layers during an experiment or a reactor operation. The FPREL estimates the migrations of gaseous and metallic fission products through a fuel particle and a fuel element by using a one-dimensional finite difference method. The ABAQ performs the analysis of the crack and debonding in a coated fuel particle. The MPRO calculates the material properties of the kernel, low-density pyrocarbon, high-density pyrocarbon, silicon carbide, matrix graphite, and structural graphite. Each module is used to produce input data for other modules or is inserted into other modules. The COPA code is one of the computer codes taking part in the IAEA-CRP-6 benchmarking program. The stresses and failure fractions calculated by the COPA-MECH and COPA-FAIL showed good agreements with the results by the other countries’ codes. In order to establish a good database of the related material properties, KAERI is participating in an international irradiation experiment, is planning its own irradiation and post-irradiation experiments, and will perform ab-initio calculations on the fuel materials.

Topics: Fuels
Commentary by Dr. Valentin Fuster
2008;():691-698. doi:10.1115/HTR2008-58053.

An experimental system for in situ high temperature measurements of spectral emissivity of VHTR materials has been designed and constructed. The design consists of a cylindrical block of silicon carbide with several machined cavities for placement of test samples, as well as a black body cavity. The block is placed inside a furnace for heating to temperatures up to 1000°C. A shutter system allows for selective exposure of any given test sample for emissivity measurements. An optical periscope guides the thermal radiation from the sample to a Fourier Transform Infra Red (FTIR) spectrometer which is used for real-time measurements of spectral emissivity over a wavelength range of 0.8μm to 10μm. To specifically address the needs of VHTR applications, the system has been designed for studies with VHTR grade helium environments and air transients. Inlet and outlet gas compositions are measured using a gas chromatograph, which in conjunction with ex situ analysis of the samples by electron microscopy and x-ray diffraction will allow for the correlation of surface corrosion of the materials and their spectral emissivities under different operating and accident conditions.

Commentary by Dr. Valentin Fuster
2008;():699-703. doi:10.1115/HTR2008-58056.

Very High Temperature Gas Cooled Reactor (VHTR) has been selected as a high energy heat source for a nuclear hydrogen generation. The VHTR can produce hydrogen from heat and water by using a thermo-chemical process or from heat, water, and natural gas by steam reformer technology. Korea Atomic Energy Research Institute (KAERI) is in the process of carrying out a nuclear hydrogen system by considering the indirect cycle gas cooled reactors that produce heat at temperatures in the order of 950°. The nuclear hydrogen system is planning to produce hydrogen by using nuclear energy and a thermo-chemical process. Helium gas is the choice for the coolant of the nuclear hydrogen system since it is an inert gas, with no affinity to a chemical or nuclear activity; therefore a radioactivity transport in the primary circuit of the nuclear hydrogen system is minimal under a normal operation. Moreover, its gaseous nature avoids problems related to a phase change and water-metal reactions and therefore improves its safety. A coaxial double-tube hot gas duct (HGD) is a key component connecting the reactor pressure vessel and the intermediate heat exchanger (IHX) for the nuclear hydrogen system. In this study, a preliminary design analysis for the primary and secondary HGDs of the nuclear hydrogen system was carried out. These preliminary design activities include a preliminary decision on the geometric dimensions, a preliminary strength evaluation and an appropriate material selection. A preliminary decision on the geometric dimensions of the HGDs was undertaken based on three engineering concepts, such as a constant flow velocity model (CFV model), a constant flow rate model (CFR model), a constant hydraulic head model (CHH model), and also based on a heat balanced model (HB model). We compared the geometric dimensions and their preliminary strength evaluation results from the various models.

Topics: Design , Ducts , Hydrogen
Commentary by Dr. Valentin Fuster
2008;():705-708. doi:10.1115/HTR2008-58092.

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.

Commentary by Dr. Valentin Fuster
2008;():709-713. doi:10.1115/HTR2008-58112.

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.

Commentary by Dr. Valentin Fuster
2008;():715-723. doi:10.1115/HTR2008-58131.

Understanding of the fission product transport in TRISO fuel particles is fundamental to improving the safety and performance of high temperature gas cooled reactors. Previous experiments showing silver release from TRISO fuel have focused on release measurements and not direct observation of the fission product transport. The possible diffusion of Ag via a grain boundary diffusion mechanism is being examined. By characterizing the SiC grain boundary structure according the coincidence site lattice scheme and detecting diffusion along specific grain boundaries, insight into the relationship between SiC microstructure and Ag release may be obtained. In addition computer modeling is being used to investigate the diffusion of silver through SiC. We employ a multi-scale approach based on ab initio techniques, molecular dynamics, and continuum rate equations in order to establish relationships between complex microstructures and diffusion rates. Initial work has begun on transport through bulk SiC and on building realistic models of grain boundaries in SiC.

Commentary by Dr. Valentin Fuster
2008;():725-732. doi:10.1115/HTR2008-58142.

As a preliminary study of cost estimates for nuclear hydrogen systems, the hydrogen production costs of the nuclear energy sources benchmarking GT-MHR and PBMR are estimated in the necessary input data on a Korean specific basis. G4-ECONS was appropriately modified to calculate the cost for hydrogen production of SI process with VHTR as a thermal energy source rather than the LUEC. The estimated costs presented in this paper show that hydrogen production by the VHTR could be competitive with current techniques of hydrogen production from fossil fuels if CO2 capture and sequestration is required. Nuclear production of hydrogen would allow large-scale production of hydrogen at economic prices while avoiding the release of CO2 . Nuclear production of hydrogen could thus become the enabling technology for the hydrogen economy. The major factors that would affect the cost of hydrogen were also discussed.

Commentary by Dr. Valentin Fuster
2008;():733-739. doi:10.1115/HTR2008-58147.

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 M6 C carbide grains rich in W will evolve. The M6 C carbides remain and some M23 C6 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.

Commentary by Dr. Valentin Fuster
2008;():741-747. doi:10.1115/HTR2008-58156.

The coated particles (CP) performance computer code GOLT (Russian abbreviation of Gas-Cooled Fuel) is under development at the A. A. Bochvar All-Russia Research Institute of Inorganic Materials. The main goal of the code is supporting development of fuel for the Gas-Turbine Modular Helium Reactor (GT-MHR). The first version GOLT-v1 has capable to calculate temperature distribution along particle radius, fuel kernel swelling, development of internal pressure under coating due to formation of gaseous fission products and CO, development of stresses and deformation in each coating layer. For TRISO-type particles special probabilistic failure model was developed. According to the failure model integrated probability of silicon carbide failure depends on probability of each dense pyrocarbon layer failure. Probabilistic version GOLT-v2 takes into account possibility of gap formation between buffer and inner dense pyrocarbon layer or between kernel and buffer that influences on maximal fuel temperature and stresses distribution in coating. More detail model of buffer performance at irradiation was developed and included in the code. List of probable coating failure mechanisms was extended. The ability of coating failure due to Kernel-Coating Mechanical Interaction (KCMI) as well as model of failure due to kernel migration was added. Thermo-dynamical code ASTRA is used in some tasks as supporting tool for calculating internal pressure and chemical interaction between SiC coating and fission products and CO. The version GOLT-v3 has accumulated all capabilities of previous versions and included Monte-Carlo analysis for estimation of fraction of failed particles with account of statistical dispersion of structural, materials and operating parameters. In the paper short description of capabilities of last versions of the code is presented. Main attention is putted to results of development version GOLT-v2a for evaluation fuel performance during accidents.

Commentary by Dr. Valentin Fuster
2008;():749-754. doi:10.1115/HTR2008-58167.

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.

Commentary by Dr. Valentin Fuster
2008;():755-759. doi:10.1115/HTR2008-58180.

The strong correlation between the density and the physical and, mechanical properties of graphite suggests that the method of nondestructive density evaluation could be developed into a characterization technique of great value for the overall improvement of safety of graphite moderator reactors. In this study, the oxidation-induced density changes in nuclear graphite for VHTR were determined by a conventional destructive bulk density measurement method (BM), and by a new non-destructive method based on acoustic microscopy and image processing (AM). The results were compared in order to validate the applicability of the latter method. For a direct comparison of the results from both measurements, two specimens were prepared from a cylindrical graphite sample (1 inch diameter and 1 inch height, oxidized to 10% weight loss at 973 K in air for 5 hours). The specimens were used for characterization by BM and AM methods, respectively. The results show that, even with a large standard deviation of the AM, the density changing trend from both methods appeared the same. This observation may be attributed to the fact that AM images reflect characteristic density changes of the graphite sample through the acoustic impedance changes. This study demonstrates the possibility of using AM as a nondestructive technique for the evaluation of density changes in graphite when a database is prepared through a systematic series of experiments.

Commentary by Dr. Valentin Fuster
2008;():761-762. doi:10.1115/HTR2008-58181.

There is currently renewed interest in high temperature nuclear fission power reactors. The Pebble Bed Modular Reactor (PBMR) is one of several high temperature gas-cooled reactors being investigated by researchers. The South African design of the PBMR is based on the original German design, with the fuel particles (called TRISO particles) being small multilayer spheres.

Commentary by Dr. Valentin Fuster
2008;():763-764. doi:10.1115/HTR2008-58198.

In a modern high-temperature nuclear reactor, safety is achieved by encapsulating the fuel elements by CVD-layers of pyrolytic carbon and silicon carbide (SiC) to prevent the fission products release. Some studies have raised doubts on the effectiveness of SiC layer as a diffusion barrier to fission fragments due to 110m Ag released from the coated particle at high temperatures ranging from 1500°C to 1600°C [1].

Commentary by Dr. Valentin Fuster
2008;():765-772. doi:10.1115/HTR2008-58309.

The high-temperature gas-cooled reactor technology is the only nuclear technology capable of achieving coolant temperatures as high as 950 °C and at the same time ensuring safe and efficient production of both electricity and hydrogen. OKBM and GA started independent research in this area in the 1990s. In 1995, OKBM in cooperation with GA started development of the GT-MHR design which combines a safe modular reactor and a power conversion unit based on the high-efficiency Brayton cycle. The power conversion unit in the GT-MHR design has integral configuration, with vertical arrangement of the turbomachine consisting of a synchronous generator and a turbocompressor. Active electromagnetic bearings are used as supports. In order to select optimal technical solutions, the effect of the following factors on the design was considered: vertical or horizontal arrangement, submerged or remote generator with oil bearings, and different turbomachine rotor speeds. Application of electromagnetic bearings and diaphragm coupling between the rotors, integral arrangement of the turbomachine inside the power conversion system vessel, and use of helium as coolant required performance of comprehensive analyses and experiments. For this purpose, the helium turbomachine technology demonstration program was developed and is currently being implemented. This technology demonstration program aims at validating the quantitative and qualitative characteristics of such turbomachine components as electromagnetic and catcher bearings, control system, computer codes, generator, diaphragm coupling, turbocompressor, etc. At the concluding stage of the technology demonstration program, a full-scale turbocompressor model will be tested at a helium test facility. The present paper lists the main parameters of the GT-MHR turbomachine and describes the status of experimental validation of its components.

Commentary by Dr. Valentin Fuster
2008;():773-776. doi:10.1115/HTR2008-58329.

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 (KüFA) furnace, formerly installed at FZ-Jülich (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 KüFA 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.

Commentary by Dr. Valentin Fuster

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