Helium Cooled Pebble Bed Blanket Research Articles

Preliminary accident analysis of the loss of vacuum in vacuum vessel for the European DEMO HCPB blanket concept

Design basis accidents are investigated continuously for the European DEMO reactor accompanying its development. One selected postulated initial event (PIE) is a loss of vacuum (LOVA) in vacuum vessel (VV) with large ingress of air induced by rupture in a VV penetration. It has been investigated for the helium cooled pebble bed (HCPB) blanket concept according to the DEMO baseline 2017. The associated primary heat transfer system (PHTS) and the related systems in the tokamak building, from the VV to the PHTS vault and galleries, are considered for the investigation. The LOVA is postulated to occur at a port seal of the electron cyclotron equatorial port plug on the side of the closure plate with (i) a small leak of 1.0 × 10−3 m2, or (ii) a large break size of 1.0 × 10−2 m2. Air ingress from one port cell into the VV leads to the VV pressurization and the fusion power termination followed by an unmitigated plasma disruption. A loss of off-site power for 32 h is assumed to coincide with the disruption. An in-vessel loss of coolant accident (LOCA) is considered as a consequence if the affected first wall (FW) reaches the defined temperature of 1000 °C. The radioactive inventories in the VV (tritium, W-dust) can mobilize towards the VVPSS, the affected systems in the building and the environment due to pressurization, venting and leak conditions. MELCOR 1.8.6 for fusion is applied for this deterministic safety analysis. The resulting releases of radioactivity to the environment are then provided for dose calculation using the computer systems UFOTRI and COSYMA. Outcomes of this LOVA analysis are critically discussed: the transient evolutions of different cases are compared; hydrogen production is detected in case of aggravating FW failure; the source terms (tritium, W-dust) are transported to the connected systems; and the dose results from the environmental releases are provided.

Analysis of the nuclear loads on Chromium monoblock divertor target for DEMO

The Plasma Facing Components (PFCs) of DEMO divertor play a fundamental role for the heat removal and particle exhaust functions, during fusion plasma. This implies a very harsh operation environment characterized by the combination of complex loading and stress conditions, intense particle bombardment, high heat flux deposition and a significant neutron irradiation. In the framework of the European DEMO divertor project (WPDIV), several target design concepts alternative to the baseline ITER-like technology are under evaluation, such as Chromium block design. It consists of chromium block, tungsten armor tile and copper alloy cooling pipe, and it is very attractive from an activation reduction, waste management and safety point of view. Therefore, the assessment of the most impacting nuclear loads for the PFC design is a pivotal aspect for DEMO divertor R&D program and this study is fully devoted to the comprehension of the spatial distributions of the main nuclear quantities (radiation damage, He-production and nuclear heating density) on the overall divertor targets. Neutronics analyses have been performed with MCNP5 Monte Carlo code using both DEMO MCNP Water Cooled Lithium Lead (WCLL) and Helium Cooled Pebble Bed (HCPB) blanket models, with a fully heterogeneous representation of the divertor cassette and Chromium PFCs geometry. The results of this analysis show that nuclear heating density and radiation damage distributions are higher with WCLL blanket than HCPB, except for He-production which seems not significantly affected by the blanket configuration. For each nuclear quantity, the peak values are achieved in correspondence of the baffle region and especially on the external units of the targets, as expected, being this region the most exposed to irradiation.

Shutdown dose rates in-cryostat outside the EU-DEMO vacuum vessel

Future fusion reactors using deuterium–tritium fuel will exhibit high fluences of high-energy neutrons inside and around the reactor vacuum vessel (VV). As well as causing material damage, fusion neutrons will activate materials, the decay of which leads to radiation fields in and around the reactor after shutdown. Gamma-ray emission from activated materials is a particular radiological hazard during periods of reactor shutdown. This must be accounted for in the design of the reactor shielding to ensure that risks are reduced as low as reasonably achievable.Recent neutronics work has evaluated the shutdown dose rates (SDDRs) in the EU DEMOnstration power plant (DEMO) around the ports and throughout the cryostat, incorporating prospective shielding improvements to the VV and ports. Prior to the proposed shielding design improvements, calculations for the model including the helium-cooled pebble bed (HCPB) blanket showed that radiation leakage through the blanket and VV leads to biological-equivalent SDDRs (following 12 days’ decay) above 103μSv/h throughout the cryostat, ignoring additional contribution from radiation streaming through the port openings. Inclusion of the proposed VV changes reduces this dose rate to below 100 μSv/h. The work finds an approximate order-of-magnitude reduction in SDDR throughout the cryostat when all proposed shielding improvements are applied, leading to dose rates in the cryostat in the range of hundreds to thousands of μSv/h for the full model. The work shows that to further reduce dose rates inside the cryostat, improving the shielding performance of the ports is required, with particular emphasis on the lower port and the equatorial electron–cyclotron launcher which currently dominate the dose rates.

Recent studies to the impact of a ceramic breeder environment on the mechanical properties of EUROFER97 under operating conditions

The reduced activation ferritic-martensitic steel EUROFER is a prime candidate for fusion structural applications. In the advanced helium-cooled pebble bed blanket concept of DEMO, this steel is in direct contact with Li-ceramics pebbles and experiences cyclic thermomechanical loading under operating conditions. In this work, the effect of ceramic breeder environment on both the microstructure and particularly the fatigue properties of the RAFM steel was investigated. For this purpose, EUROFER samples were embedded in ceramic pebbles and annealed at 550 °C under a purge gas atmosphere for durations ranging 8–128 days. Microstructural studies show that annealing at the given conditions leads to the formation of a complex corrosion multilayer consisting of ferrite and iron oxide in an outer layer, as well as chromium oxide in both an intermediate and an inner layers. For annealing durations up to 32 days, the thickness of the oxide layer is continuously increasing. However, for more prolonged annealing, the thickness of the corrosion layer reaches saturation at about 20 µm. Similar to the growth behavior of the oxide layer, the strain-controlled LCF tests showed a remarkable reduction of the LCF lifetime with increasing in the annealing duration up to 32 days. However, beyond an annealing duration of 64 days, an additional decrease of lifetime is not observed. When performing additional experiments (annealing with ceramic pebbles in vacuum and annealing in purge gas without pebbles) it is found that corrosion is mainly caused by purge gas impurities, namely oxygen and water. Furthermore, the decrease in fatigue lifetime is related to the degradation of surface quality, causing an accelerated progress of damage. Finally, Charpy impact tests combined with carrier gas hot extraction investigations revealed that hydrogen embrittlement of the samples, annealed in a breeder environment containing 0.1 vol.% H2, can be excluded.

Effect of HIP at 1000–1200 °C on microstructure and properties of extruded Be-Ti composites

Solid titanium beryllide blocks will be used for neutron multiplication in the helium-cooled pebble bed (HCPB) blanket concept of EU DEMO. A combination of hot extrusion of Be-Ti powders and subsequent hot isostatic pressing (HIP) of the obtained Be-Ti composites has been proposed for manufacturing such blocks. This work is devoted to the study of the effect of HIP at 1000–1200 °C on the structure and properties of Be-Ti composites in order to optimize the HIP parameters. The HIP at 1000–1200 °C resulted in an almost single-phase titanium beryllide (TiBe12) with small amounts of Be and other phases, which gradually dissolve with an increase in the HIP temperature. Such a treatment at 1000 and 1100 °C provides a very fine-grained microstructure of TiBe12 with an average grain size of 0.3 and 0.6 μm, respectively. The resulting titanium beryllide is characterized by high microhardness of 1350–1480HV0.1 depending on the HIP temperature. According to the nanoindentation tests of the Be-Ti composite after HIP at 1100 °C, the indentation modulus of TiBe12 can be estimated as 295GPa. The fracture toughness of the TiBe12 was determined as 1.5–1.7 MPa·m1/2. The temperature of 1100 °C was chosen as optimal for the HIP of Be-Ti composites after hot extrusion. The titanium beryllide obtained in this way was used to manufacture a reduced size mockup of Ø20mm × 18 mm. The mockup has no visible surface defects and can be used for further experiments.

Post-Test Numerical Analysis of a Helium-Cooled Breeding Blanket First Wall under LOFA Conditions with the MELCOR Fusion Code

The validation of numerical tools employed in the analysis of incidental transients in a fusion reactor is a topic of main concern. KIT is taking part in this task providing both experimental data and by performing numerical analysis in support of the main codes used for the safety analyses of the Helium Cooled Pebble Bed (HCPB) blanket concept. In recent years, an experimental campaign has been performed in the KIT-HELOKA facility to investigate the behavior of a First Wall Mock-Up (FWMU) under Loss Of Flow Accident (LOFA) conditions. The aim of the experimental campaign was twofold: to check the expected DEMO thermal-hydraulics conditions during normal and off-normal conditions and to provide robust data for code validation. The present work is part of these validation efforts, and it deals with the analysis of the LOFA experimental campaign with the system code MELCOR 1.8.6 for fusion. A best-estimate methodology has been used in support of this analysis to ease the distinction between user’s assumptions and code limitations. The numerical analyses are here described together with their goals, achievements, and lesson learnt.

Activation Analysis and Evaluation of Radionuclide Inventory Decay Heat for EU DEMO Vacuum Vessel Components

Activation inventory data were assessed for different European DEMOnstration Fusion Power Reactor (EU DEMO) vacuum vessel (VV) segments in order to evaluate the activity, dose rates, and decay heat of the structural materials and the impact of heterogeneity as a part of the Safety and Environment project of the EUROfusion/Power Plant Physics Technology program. Activity is a good indicator of the condition of irradiated material while decay heat and dose rate estimates are crucial for power plant operation and public safety. In addition, the VV structures were analyzed in terms of different breeder blanket configurations that were implemented in the model. Helium-cooled pebble bed and water-cooled lithium-lead breeder blanket modules were considered. These modules directly shield the VV from neutron irradiation; therefore, their design has a great influence on subsequent structure activation.

Analysis of supercritical carbon dioxide Brayton cycles for a helium-cooled pebble bed blanket DEMO-like fusion power plant

Nuclear fusion is expected to be a clean and almost-unlimited power source in the near future. The first net power demonstration plant (DEMO) is planned to start operation in 2050. The supercritical carbon dioxide (S-CO2) Brayton cycle is an excellent candidate for integration with a fusion power plant, such as DEMO, because of its high efficiency at intermediate temperatures and low interaction of coolant with tritium.This work analyses a set of S-CO2 Brayton cycle layouts for its integration in a DEMO-like fusion power plant, considering the specific requirements and heat availability characteristics. A framework has been developed to integrate the PROCESS code and the numerical solver EES to study the thermal and economic aspects of integrating the different S-CO2 cycle layouts. In total, 14 layouts have been studied and grouped into a more conservative (DEMO1, pulsed operation) and more advanced (DEMO2, steady-state operation) fusion reactors. The PROCESS code has been used to obtain the DEMO 2018 Baseline, which defines the available power from each heat source and their boundary conditions. This code has also been used to assess the cost of the optimal layout. Thermal storage has been added to the DEMO1 scenario to avoid standby times that could negatively affect the cycle equipment lifetime and efficiency. Besides, these boundary conditions have been extended to account for possible technical improvements by the time of its construction in the DEMO2 scenario.A sensitivity analysis of the most characteristic parameters of the cycles shows a strong dependence on the turbine inlet temperature for all layouts, which is constrained by the reactor material limits. The cycle efficiency (electric power produced before consumptions non-related to the cycle) has been selected as the figure of merit for the optimisation. The results show a 38% cycle efficiency for DEMO1 and 56% for DEMO2 scenarios. These efficiencies drop to 20% and 38% values, respectively, when the reactor and cooling loop power consumptions are considered. These values are obtained for current fusion reactor conceptual designs. The economic analysis shows the economic viability of DEMO2 scenarios.

Exploratory tritium breeding performance study on a water cooled lead ceramic breeder blanket for EU DEMO using Serpent-2

• Explored the tritium breeding performance of WLCB Breeder blanket using Serpent-2. • Analyzed the sensitivity of TBR to materials in first wall and breeder zone. • Analyzed the effect of first wall and breeder zone structure on TBR. • Compared the effect of Lead-alloy and Lead on TBR. Innovative Water cooled Lead Ceramic Breeder (WLCB) blanket concepts are being developed at Karlsruhe Institute of Technology (KIT) to explore alternative options for the European demonstration fusion power plant (DEMO). Compared to the Helium Cooled Pebble Bed blanket (HCPB, which is one of the two driver blanket concepts of the European DEMO), Lead/ Lead-alloy is used as neutron multiplier instead of Be/Be-alloy and pressurized subcooled water is used as coolant instead of Helium in the WLCB. The tritium self-sufficiency is the vital function that a Breeding Blanket has to achieve. The absorption of neutrons by water leads to a decrease in the number of neutrons, so the requirement to improve the tritium breeding capacity of the water-cooled blankets is particularly prominent. The Monte Carlo neutron transport code Serpent-2 developed by VTT in Finland has already been benchmarked to be applicable to neutronics calculations in the fusion reactors in a previous study. In this paper, an exploratory TBR study on Water cooled Lead Ceramic Breeder Blanket is presented.

A Water cooled Lead Ceramic Breeder blanket for European DEMO

The development of a demonstration fusion power plant (DEMO) is widely considered a crucial step beyond ITER leading toward fusion electricity. In a viable DEMO plant two major functions have to be achieved, namely to breed tritium for fuel self-sufficiency and to extract high-grade heat to allow an efficient electricity production. The breeding blanket is the key component to ensure these two functions. The solid breeder blanket is considered in the design of several ITER Test blankets and DEMO breeding blankets. In Europe, the helium cooled solid breeder blanket, so-called Helium Cooled Pebble Bed (HCPB) blanket, is one of the two driver blanket candidates selected for the European DEMO. In spite of the considerable progress and achievements, several potential issues inherent to the choice of the neutron multiplier and coolant remain with this concept. To overcome these challenges, research activities on an innovative concept (a Water cooled Lead Ceramic Breeder - WLCB), in which Lead/Lead-alloy is used as neutron multiplier, ceramic pebbles as tritium breeder, pressurized sub-cooled water as coolant, are on-going. This concept avoids the potential safety issue (exothermic hydrogen-producing Be-water reaction or PbLi-water reaction in case of water leakage) when Beryllium/Beryllide is used as neutron multiplier or lithium lead is used tritium breeder in a water cooled blanket concept. The neutronics, thermal hydraulics and mechanical performance figures of the above-mentioned innovative blanket concept are presented. This work concludes with possible future work and outlook towards the development of innovative alternative blanket concepts for the European demonstration fusion power plants.

Comparative activation analyses for the HCPB breeding blanket in DEMO

The Helium Cooled Pebble Bed (HCPB) breeding blanket is one out of two concepts considered as driver blanket for the European DEMOnstration fusion power plant. The latest design, developed within the Breeding Blanket project of EUROfusion’s Power Plant Physics and Technology (PPPT) programme, assumes a fission-like breeder pin arrangement with breeder ceramics and Be12Ti as neutron multiplier.For the assessment of safety and waste management issues related to the use of the HCPB blanket in DEMO, a thorough activation analysis is required. Such an analysis should provide the total activity inventory and decay heat as well as the breakdown into components and materials as resulting from the irradiation over the assumed blanket lifetime in DEMO. An accurate calculation of the activity inventories requires the application of a suitable computational approach coupling the neutron transport simulation and the activation calculation with fully heterogeneous 3D geometry model of the DEMO.Precise materials specifications, including impurities and tramp materials, are essential prerequisites for adequate activation calculations. For the materials used in the HCPB blanket, the material compositions are provided with the material specifications set up for PPPT neutronics and activation calculations. These include the Eurofer steel, the breeder ceramics, tungsten and the beryllium in the Be12Ti used as neutron multiplier.Two difference tritium breeding material compositions of activation analyses were performed using each Li4SiO4+Li2TiO3 the latest material and previous material definition Li4SiO4. And beryllium titanate, the activation calculations were performed using the impurities specifications as provided by different potential beryllium suppliers. The most essential here is a uranium content in the ore used for the Be12Ti metallurgical engineering. This can permit a justification of the potential environmental risks associated with the application of the various materials options. The effect of the different Lithium compositions on the activity, both on the short- and long-term time scale, is evaluated in this work in the context of the radioactive waste management issue for DEMO.

BENCHMARKING OF THE SERPENT 2 MONTE CARLO CODE FOR FUSION NEUTRONICS APPLICATIONS

Analyses of radiation fields resulting from a deuterium-tritium (DT) plasma in fusion devices is a critical input to the design and validation of many aspects of the reactor design, including, shielding, material lifetime and remote maintenance requirements/scheduling. Neutronics studies, which perform in-depth analysis are typically performed using radiation transport codes such as MCNP, TRIPOLI, Serpent, FLUKA and OpenMC. The Serpent 2 Monte-Carlo code, developed by VTT in Finland, is the focus of this work which seeks to benchmark the code for fusion applications. The application of Serpent 2 in fusion specific analysis requires validation of the codes performance in an energy range, and a geometrical description, which significantly differs to conventional nuclear fission analysis, for which the code was originally developed. A Serpent model of the Frascati Neutron Generator (FNG) Helium Cooled Pebble Bed (HCPB) mock up experiment has been prepared and the calculated results compared against experimental data, as well as the reference Monte Carlo code MCNP. The analysis is extended to a model of DEMO with HCPB blanket concept. For this model, the flux, nuclear heating, tritium production and DPA are calculated, all of which are integral nuclear responses in fusion reactor analysis. In general, a very good agreement is demonstrated for both of the benchmarks, with any discrepancies pinpointed to different physics models implemented.

Development and characterization of advanced neutron multiplier materials

Neutron multiplier materials are essential for self-sufficient tritium production and are closing the fuel cycle of fusion reactors. Until now, a concept of pebble bed consisting of interchanging layers of beryllium and lithium ceramic pebbles was considered for the Helium-Cooled Pebble Bed (HCPB) tritium-breeding module of the first experimental fusion reactor ITER as well as for the next demonstration fusion reactor DEMO. However, this concept depends on the availability of large amounts of pure beryllium pebbles and is also limited by its material properties like for example the tritium accumulation under irradiation.The results of tritium retention and analytical microstructural studies of beryllium pebbles obtained within the framework of the HIDOBE irradiation campaign suggest that a significant fraction of generated tritium (up to 100% below 500°C) is trapped within helium bubbles. Being negligible in the ITER tritium-breeding module (TBM), the total accumulated tritium inventory imposes severe safety issues and exceeds acceptable limits for the DEMO blanket. Therefore, advanced neutron multiplier materials such as beryllides have to be well characterized for their applicability in the HCPB blanket of DEMO and beyond. The usage of an advanced material with lower volumetric swelling, lower tritium retention, increased irradiation and chemical resistance as well as with higher melting temperature allows to switch from the pebble bed concept to a solid hexagonal block-based one. For the fabrication of titanium beryllide samples both the semi-industrial fabrication route utilizing the hot extrusion of rods and the industrial approach using vacuum hot pressing of a Be-Ti powder mixture were explored. In this contribution, we discuss reasons for the transition from pure beryllium to beryllides, respective changes of the HCPB blanket design and a successful demonstration of the feasibility of a beryllide block fabrication by an industrial method.

Tritium release and retention in beryllium and titanium beryllide after neutron irradiation up to damage doses of 23-38 dpa

• For the first time, beryllium and titanium beryllide as massive pellets were irradiated in a nuclear reactor at temperatures of 710−1040 K up to production of 430–653 appm tritium and 4144–5992 appm helium, respectively. • The tritium retention in irradiated Be-7at.%Ti pellets is much lower than that in Be pellets, for example, 1 and 0.5 % to 48 and 34 % at 940 and 1040 K, respectively. • Open channels along grain and sub-grain boundaries are formed under neutron irradiation in beryllium, essentially contributing to the enhanced tritium release. Titanium beryllide is considered as an advanced neutron multiplier in the Helium Cooled Pebble Bed (HCPB) blanket of DEMO. Neutron irradiation of titanium beryllide together with beryllium as a reference material in material testing nuclear reactors can give essential data for the DEMO blanket design. Be-7at.%Ti (Be-7Ti) as well as Be were irradiated in the HFR, Petten, the Netherlands, at four temperatures of 710, 800, 940, 1040 K up to 23, 31, 36, 38 dpa, respectively. The post-irradiation examination (PIE) included thermal-programmed desorption (TPD) and Vickers hardness tests as well as microstructure study by optical metallography. Be and Be-7Ti pellets maintained their integrity after irradiation. Microstructure of Be-7Ti contains two phases, mainly, TiBe 12 , and also small amount of Be. Under irradiation, gas bubbles were formed in Be samples as well as in Be-phase in Be-7Ti samples. These bubbles contain helium and tritium produced in Be under irradiation. TPD tests showed a much lower tritium retention in Be-7Ti than in Be for all four irradiation temperatures. Vickers hardness of TiBe 12 phase is much higher than that of Be-phase. According to the obtained data, Be-7Ti could be considered more preferred than Be as a neutron multiplier material in future fusion reactors due to the enhanced radiation damage resistance.

Design of Two Experimental Mock-Ups as Proof-of-Concept and Validation Test Rigs for the Enhanced EU DEMO HCPB Blanket

Within the framework of EUROfusion activities, the helium cooled pebble bed (HCPB) breeding blanket is under development in Karlsruhe Institute of Technology (KIT). The enhanced HCPB, using a fissionlike fuel breeder pin assembly configuration, is proposed as the near-term breeding blanket for the European Union DEMOnstration power plant (EU DEMO). The helium gas with a pressure of 8 MPa is used as the coolant, EUROFER is used as the structural material, advanced ceramic breeder pebbles are used as the tritium breeder, and Be12Ti is used as the neutron multiplier material. In contrast to the former HCPB cooling plate configuration, the fuel breeder pin assemblies greatly reduce the pressure drop, reducing the circulating power to the level where state-of-the-art helium turbomachinery can be used. However, because of the reduced coolant velocity in the breeder zone, the heat transfer performance is compromised, especially in the annular channel of the fuel breeder pins. An increased surface roughness is therefore proposed as a heat transfer augmentation technique for the fuel breeder pins. Although heat transfer augmentation using artificial roughness is common, it is relatively novel for small annular gaps with moderate velocity as the ones in the fuel breeder pins. Currently, a dedicated correlation for the small annular rough-wall channel is available to predict the Nu number. This correlation is wished here to be benchmarked and validated experimentally. Therefore, an experimental investigation on a fuel breeder pin mock-up (mock-up 1) is planned. Additionally, based on computational fluid dynamics calculation, unsteady, nonuniform flow patterns were found at the return flow after the jet impingement in the first-wall region. Another upscaled mock-up (mock-up 2) to investigate the nonuniform flow patterns of the return flow in the annular channel of the fuel breeder pin is planned. The dedicated experimental campaigns are foreseen at the Helium Loop Karlsruhe (HELOKA) in KIT as validation and proof-of-concept test rigs for this enhanced pin design. In this paper, the motivation and the preliminary design of these two mock-ups of the enhanced EU DEMO HCPB blanket are shown, together with the plan for the foreseen experiments.

Advancements in the Helium-Cooled Pebble Bed Breeding Blanket for the EU DEMO: Holistic Design Approach and Lessons Learned

The helium-cooled pebble bed (HCPB) blanket is one of the two concepts proposed as a driver blanket for the European Union Demonstration Fusion Power Reactor (EU DEMO). In contrast to past conceptual design studies, in the frame of the current Power Plant Physics and Technology of the EUROfusion Consortium, the ongoing EU DEMO preconceptual design activities have adopted a holistic and integrated (i.e., systems engineering) design approach. As a consequence of this new approach, many interfaces and requirements have been identified, some of them driving the design of the blankets. This paper shows the advancements in the HCPB breeding blanket and describes the lessons learned after implementing the new approach. This new set of requirements has led us to reconsider fundamental aspects of the HCPB blanket design, especially in the way of how the heat is extracted from the blanket. Among others, the requirement to achieve a mature balance of plant (BOP) system plays a central role as a key design driver and has forced us to reduce pressure drops in the breeding blanket. In this regard, the blanket has been redesigned, leading to an enhanced concept based on single-module segments with a hexagonal matrix of fuel-breeder pins. Both the fuel-breeder pins and the first wall (FW) are equipped with turbulence promoters (augmented wall roughness in the fuel-breeder pins and V-ribs in the FW), following a similar idea as in the past MAGNOX, Advanced Gas Reactor (AGR), and Gas Cooled Reactor (GCR) programs in fission. This has led to minimizing the pressure drops while maximizing the heat transfer. Also, the blanket outlet temperature has been extended to 520°C, following the same principle as in Generation IV’s GCRs of maximizing the temperature difference across the core to minimize the reactor mass flow rate and thus the circulating power. All these features have led to a remarkably low plant circulating power (80 to 90 MW) and the required power per helium blower (5 to 6 MW), which potentially solves the key long-standing problem of the BOP technology readiness level for an ≈2.4-GW(thermal) helium-cooled DEMO reactor.

Discrete Element Analysis of Heat Transfer in the Breeder Beds of the European Solid Breeder Blanket Concept

In this work, an in-house thermal–Discrete Element Method (DEM) code, recently developed at Karlsruhe Institute of Technology to evaluate the heat transfer in ceramic packed pebble beds, was applied to study the thermal behavior of the breeder beds of the European solid breeder blanket concept. The breeder zone of the helium-cooled pebble bed (HCPB) blanket for the Demonstration (DEMO) reactor was considered as the reference model implementing the same materials, applying the related neutronic heating, and simulating the relevant bed thicknesses. The code was used to evaluate the temperature profile generated by the neutronic heating in the thickness of the breeder bed. A column cutout of packed pebbles bounded by upper and bottom walls, representing the cooling plates of the HCPB, was considered as a representative geometry to carry out the work. The implemented three-dimensional network model evaluates the heat transfer inside packed beds through chains of thermal resistances describing the thermal contacts (particle-particle and particle-wall) occurring in the assembly. Besides thermal transport through the mechanical contact area, thermal transport through the surrounding gas phase is accounted for including the Smoluchowski effect. Sensitivity studies revealed the influence of the operational conditions and the parameters that mainly affect the temperature profile in the bed.

Preliminary RAMI assessment for HCPB and HCLL ITER test blanket module ancillary systems

A RAMI (Reliability, Availability, Maintainability and Inspectability) assessment performed on the ITER Test Blanket Module (TBM) ancillary systems is presented. The assessment is aimed at evaluating design that may jeopardize the achievement of the overall 75% availability requirements for the TBM plant breakdown structure. The ancillary systems of the European Test Blanket Modules to be operated in ITER that have been analyzed in this assessment are the helium cooling systems (HCSs) of both the Helium Cooled Lithium Lead (HCLL) and Helium Cooled Pebble Bed (HCPB), the lithium-lead loop of the HCLL TBM and the Tritium Extraction System (TES) of the HCPB TBM. RAMI performance was assessed by means of reliability block diagrams (RBDs) over a foreseen mission of 20 years, with operating cycles reflecting ITER schedule. In particular, a set of events leading to unavailability of the systems was initially defined by means of a failure mode and effect analysis. Then RBDs were implemented according to reliability-wise integration of the components included in such systems. Different RBDs were defined depending on the component being judged to impact normal operation or start-up operation mode. Systems analysis was performed exploiting a modular approach in order to elicit the relative contribution to overall availability so to support possible design improvement by highlighting critical sub-systems. In particular the unavailability of cooling, tritium extraction and trapping functions were assessed separately for HCLL and HCPB concepts. Finally the integrated impact of all ancillary systems on overall tokamak operational availability was estimated and the assessed systems, considering the current preliminary level of design development, appear able to meet the expected performance.

Validation of Multi-Physics Integrated Procedure for the HCPB Breeding Blanket

The wide range of requirements and constraints involved in the design of nuclear components for fusion reactors makes the development of multi-physics analysis procedures of utmost importance. In the framework of the European DEMO project, the Karlsruhe Institute of Technology (KIT) is dedicating several efforts to the development of a multi-physics analysis tool allowing the characterization of breeding blanket design points which are consistent from the neutronic, thermal-hydraulic and thermal-mechanical points of view. In particular, a procedure developed at KIT is characterized by the implementation of analysis software only. A preliminary step for the validation of such a procedure has been accomplished using a dedicated model of the DEMO Helium Cooled Pebble Bed Blanket 4th outboard module. A global model representative of nuclear irradiation in DEMO and two local models have been set up. Nuclear power deposition and the spatial distribution of its volumetric density have been calculated using Monte Carlo N-Particle transport code for the aforementioned models and compared in order to validate the procedure set up. The outcomes of this comparative study are herein presented and critically discussed.

Nuclear analyses of solid breeder blanket options for DEMO: Status, challenges and outlook

This work presents results of the neutronic analyses for an innovative Helium Cooled Pebble Bed (HCPB) blanket concept for DEMO. Comprehensive 3D Monte Carlo particle transport simulations were performed with the MCNP5 code making use of a highly detailed HCPB DEMO torus sector model. The advanced HCPB Single Module Segmentation (SMS) blanket was consequently optimized to accomplish out the design requirements. A completely new Molten Lead Ceramic Breeder (MLCB) SMS blanket utilizing a solid breeder and a Pb neutron multiplier together was also investigated and proved to be an alternative option for DEMO design.