Blanket System Research Articles

Comparative studies for two different orientations of pebble bed in an HCCB blanket

The Indian Test Blanket Module (TBM) program in ITER is one of the major steps in its fusion reactor program towards DEMO and the future fusion power reactor vision. Research and development (R&D) is focused on two types of breeding blanket concepts: lead–lithium ceramic breeder (LLCB) and helium-cooled ceramic breeder (HCCB) blanket systems for the DEMO reactor. As part of the ITER-TBM program, the LLCB concept will be tested in one-half of ITER port no. 2, whose materials and technologies will be tested during ITER operation. The HCCB concept is a variant of the solid breeder blanket, which is presently part of our domestic R&D program for DEMO relevant technology development. In the HCCB concept Li2TiO3 and beryllium are used as the tritium breeder and neutron multiplier, respectively, in the form of a packed bed having edge-on configuration with reduced activation ferritic martensitic steel as the structural material. In this paper two design schemes, mainly two different orientations of pebble beds, are discussed. In the current concept (case-1), the ceramic breeder beds are kept horizontal in the toroidal–radial direction. Due to gravity, the pebbles may settle down at the bottom and create a finite gap between the pebbles and the top cooling plate, which will affect the heat transfer between them. In the alternate design concept (case-2), the pebble bed is vertically (poloidal–radial) orientated where the side plates act as cooling plates instead of top and bottom plates. These two design variants are analyzed analytically and 2D thermal-hydraulic simulation studies are carried out with ANSYS, using the heat loads obtained from neutronic calculations. Based on the analysis the performance is compared and details of the thermal and radiative heat transfer studies are also discussed in this paper.

Nuclear Heating and Radiation Damage in the NBI Region of the ITER Vacuum Vessel

The ITER vacuum vessel (VV) is a double walled toroidal shaped stainless steel structure divided into nine 40 degree sectors. In the design process for the ITER blanket system (which provides shielding for the VV), determining integrated nuclear heating loads on the VV is important for cooling system sizing and determining localized nuclear heating on the VV is important for assessing thermal stress loads. Further, determining radiation damage, displacements per atom (dpa) on the VV, is important in meeting pressure vessel limits. Near the neutral beam injection (NBI) region of the VV (both sector 2 and sector 3), there are significant gaps and cut-outs in the blanket system to accommodate the 3 heating neutral beam (HNB) ports and the diagnostic neutral beam (DNB) port. These features lead to higher localized radiation loads. Previous analysis indicated high nuclear heating and dpa in the NBI region. The CAD based DAG-MCNP5 transport code was used to perform neutronics calculations in detailed, updated CAD models of the NBI region. For this work, a 40 degree model of sector 2 (which includes the HNB1 port, the DNB port, and the HNB2 port) was analyzed. Three design options were investigated which add shielding in the DNB port region by using port liners. Mesh tally maps of both nuclear heating and dpa are provided for the VV in the BM13-16 region. Peak dpa values ranged from 0.41–0.65 dpa. Two of the 3 design options investigated had peak dpa values near the DNB port within the ITER dpa limit of 0.5 dpa. Peak nuclear heating results ranged from 1.7 W/cm3 to 2.0 W/cm3. The mesh tally maps of nuclear heating have been provided to the ITER Organization for subsequent finite element engineering analysis. Preliminary analysis has shown the thermal stress levels are unacceptable with the added shielding. The results of this work are being used by the ITER Blanket and Tokamak Integration groups to assess the current design and modify blanket module (BM) design where needed if radiation loads are excessive.

RAMI comparison of two China test blanket modules

Tritium breeding blanket module is one of the most important parts of fusion reactor. One of the main goals of ITER is to test and validate design concepts of tritium breeding blankets. China has proposed two test blanket module (TBM) concepts, including Helium cooled ceramic breeder test blanket module system (HCCB TBS) and the dual functional lithium-lead test blanket module system (DFLL TBS), which will be tested in one of the three equatorial ports of ITER. A detailed comparative study on the design schemes of these two TBSs concepts was performed in this paper. In this study, the RAMI (Reliability, Availability, Maintainability, Inspectability) approach was applied on those two test blanket systems. The modeling and analyzing process were done by the Reliability and Probabilistic Safety Assessment Software System RiskA. In addition, comparative study in the phases of functions, capabilities and efficiencies of tritium breeding, heat removal and coolant purification is also performed. It is found that HCCB TBS is a better choice currently to fulfill the engineering demand of ITER construction for its lower difficulty of technical implementation. However, DFLL TBS possesses a high performance prospect from the viewpoint of long-term ITER development due to its inherent advantages compared to HCCB TBS. Discussions of priorities for future research needed for these two TBS are also proposed in the paper.

Recent R&D Results on Fusion Nuclear Technology for ITER and DEMO Reactor in Japan

Several key components, such as superconducting coils, remote handling equipment, heating systems, have been designed and manufactured by JADA (Japan Domestic Agency). These activities have been carried out in accordance with the agreed schedule; in collaboration with the ITER organization and other domestic agencies. As a significant technical program using ITER, to design and to manufacture the TBS (Test blanket system), some R&D and design activities have also been conducted in Japan. Under the IFERC (the International Fusion Energy Research Center) projects of BA (Broader Approach) activities, design and R&D activities on fusion DEMO reactor have been carried out. For the DEMO R&D activity, five basic R&D subjects for a DEMO blanket system have been selected, and been studies under close collaborations between EU and JA: structure materials (RAFM steels and SiC/SiC composites), functional materials (tritium breeders and neutron multipliers), and tritium technology. From 2007, the above projects produced a set of fruitful results. A series of advanced technologies for the DEMO blanket system has also been carried out by Universities in Japan. Some significant basic R&D studies have also been carried out under US-JA collaborative program.

MHD thermohydraulics analysis and supporting R&D for DCLL blanket in the FNSF

A Fusion Nuclear Science Facility (FNSF) has been recognized in the fusion community as a necessary facility to resolve the critical technology issues of in-vessel components prior to the construction of a DEMO reactor (Abdou et al., 1996) [1]. Among these components, development of a reliable, low-cost and safe blanket system that provides self-sufficient tritium breeding and efficient conversion of the extracted fusion energy to electricity, while meeting all material, design and configuration limitations is among the most important but still challenging goals. In the recent FNSF study in the US (Kesel et al., 2015) [2], a Dual-Coolant Lead-Lithium (DCLL) blanket has been selected as the main breeding blanket concept. This paper summarizes the most important details of the proposed DCLL blanket design, presents the MHD thermohydraulic analysis for the PbLi flows in the blanket conduits and introduces supporting R&D studies, which are presently ongoing at UCLA. We also discuss the required pre-FNSF R&D in the area of MHD Thermofluids to support the further work on the DCLL blanket design & analysis and its integration into the fusion facility.

Thermal Design Investigation for a Flinabe Blanket System

The present study evaluates the thermal design of a blanket system using Flinabe in order to facilitate further discussions on its applicability as a self-cooled liquid blanket system. Molecular dynamics simulations were performed to evaluate the Prandtl number of mixtures in five compositions (LiF–NaF–BeF2 = 31–31–38, 36–27–37, 42–22–36, 49–16–35, and 67–0–33). Thermofluid analysis was carried out to estimate the temperature margin and pressure drop per unit length in a simple geometry model of the blanket system. The Prandtl number of Flinabe is above 100 at 400°C. The present study reveals that Flinabe remarkably relaxes the design conditions compared to Flibe as a coolant owing to its low melting point. In contrast, the pressure drop per unit length of Flinabe is higher than that of Flibe because the viscosity exponentially increases at low temperature. The temperature margin is quite dependent on the heat load on the first wall. If the pressure drop per unit length is around 1.0 MPa/m, the heat load value must be approximately below 0.7 MW/m2.

Efficient 2,4-dichloronitrobenzene removal in the coupled BES-UASB reactor: Effect of external voltage mode

In this study, bioelectrochemical-upflow anaerobic sludge blanket (BES-UASB) system was developed for treatment of 2,4-dichloronitrobenzen (DClNB) containing wastewater to investigate the effect of external voltage with different supplying modes. Results showed that 2,4-dichloroaniline (DClAN) was under detection limit in R1 (applied with intermittent voltage) and R2 (applied with continuous voltage) when the DClNB loading increased from 25 to 100gm3d−1 (hydraulic retention time (HRT) was decreased from 24 to 6h) while sudden accumulation of DClAN (1.7mgL−1) was observed in R0 (control). Dechlorination efficiency (DE) decreased to 32.7%, 45.0% and 45.3% in R0, R1 and R2 when HRT was further shortened to 4h. Microbial community analysis indicated the significant enrichment of dechlorination-related species in R1 and R2 compared with R0, e.g., Dehalobacter and Dehalococcoides. In summary, the BES-UASB system with intermittent voltage is an alternative process for efficient treatment of DClNB containing wastewater, and the energy input was reduced markedly.

The Cooling Performance Improvement of First Wall Using He/CO2 Binary Mixtures Gas for CN HCCB TBM

The Helium (He) gas Cooled Ceramic Breeder (HCCB) test blanket module (TBM) is the primary option of the Chinese TBM program. In order to enhance the cooling ability of the first wall (FW) of the HCCB TBM and reduce the circulation power, a binary mixtures gas of He gas and additive CO2 deserves to be another option for the coolant of the blanket, based on high temperature gas-cooled reactors which are a generation-IV fission reactor concept, when it is reported that forced convective heat transfer can be enhanced by means of binary mixing with unreactive gas (e.g., CO2, molecular weight 44). This technique can significantly enhance the plant’s overall efficiency and reduce the cost of electricity. In order to evaluate the cooling performance of the He/CO2 binary mixtures gas and its circulation power in the FW of the HCCB TBM, a three-dimensional computational fluid dynamics (CFD) numerical simulation, combined experimental research method is applied. The results reveal that under the condition of the cooling requirements of the FW (e. g., maximum temperature, radial temperature gradient) similar to the pure He gas, the flow velocity and circulation power of the He/CO2 binary mixtures gas (mole fraction 0.4) are reduced by 70% and 87%, respectively. It implies that the thermal efficiency of a He-cooled blanket system can be fairly enhanced by means of this technique. In the near future experiment plan, it should be tested to validate the correlative cooling scheme of the HCCB TBM, in which the pure He gas and He/CO2 binary mixtures gas are used as coolant, respectively, at our High-Pressure Helium-Cooled Loop facility. The CFD numerical results will be selected as the reference for the experiments. A new approach may be provided for cooling the high heat flux components of a fusion reactor.

Accident analysis on LOCA in HCCR-TBS towards CCWS-1

Korean Helium Cooled Ceramic Reflector (HCCR) Test Blanket System (TBS) will be operated at elevated temperature with high pressure helium coolant during normal operation in ITER. One of the main ancillary systems of HCCR-TBS is Helium Cooling System (HCS) which play an important role to extract heat from HCCR Test Blanket Module (TBM) by the helium coolant to keep the operational temperature and the extracted heat is finally transferred to ITER CCWS-1 (Component Cooling Water System) by a Printed Circuit Heat Exchanger (PCHE) in the HCS. In such circumstances if Loss Of Coolant Accident (LOCA) occurs in the PCHE, the high pressure helium coolant in the primary side goes into the lower pressure water in the secondary side thus pressurizing CCWS-1. In addition, since the helium coolant contains tritium due to permeation from the TBM, tritium migrates into CCWS-1, a non-nuclear system. In this paper, accident analysis for LOCA in the heat exchanger is presented. For the analysis, GAMMA-FR code which has been developed for fusion applications was used. Main components in the HCS and CCWS-1 were modelled as volume and junctions. The accident analysis was performed for the reference case with ten channels rupture and sensitivity study was also performed by changing the crack size. The results show that pressure and tritium requirement of CCWS-1 can be met in spite of LOCA in the heat exchanger of the HCCR-TBS HCS.

THALLIUM: An experimental facility for simulation of HCLL In-box LOCA and validation of RELAP5-3D system code

The experimental facility THALLIUM (Test HAmmer in Lead LithIUM) was designed to reproduce the hydraulic transient in the HCLL TBS (Helium Cooled Lithium Lead Test Blanket System) breeder loop following the rupture of a Cooling or a Stiffening Plate, i.e. In-box LOCA (Loss Of Coolant Accident). In addition, the facility will be used to validate the simulations of the pressure wave propagation with RELAP5-3D system code. THALLIUM reproduces the geometry of the LLE (Lead Lithium Eutectic) loop of the HCLL TBS and it was installed at the ENEA Brasimone Research Centre. In this paper, the facility is presented in detail, emphasizing the similarities and the differences with the reference HCLL TBS design. Particular attention has been paid to the TBM mock-up and to the TBS Storage/Recirculation Tank, which are the components that required the biggest simplifications among the key components of the facility. An analytical study of the pressure wave was carried out, with the aim to highlight a possible source of inaccuracy that might be displayed in the future simulations with the code. The results of the analytical study show that the fluid-structure interactions will decrease the speed of sound in the LLE. Instead, RELAP5-3D does not consider the fluid-structure interactions and, therefore, a difference between the experimental and the simulation data could arise from the different wave speed. Finally, the RELAP5-3D nodalization and the results of a pre-test simulation are presented. These results proved to be useful for the design of the facility

Characterization of functional materials for liquid blanket systems by cathodoluminescence measurement

Applicability of cathodouminescence (CL) measurements in development of functional materials for advanced liquid blanket systems was investigated by using a conventional scanning microprobe and CCD spectrometer. Firstly, CL spectra of lithium compounds for tritium fuel breeding, oxide, nitride and carbide functional materials for advanced blanket systems, fluoride materials for coolants, etc. were obtained to construct a database. Relations between changes in the spectra and the crystallinities of specimens were examined for ZrO2, Y2O3 and TiO2 coatings baked at different temperatures. The specimens with lower crystallinities showed broader luminescence peak at longer wavelengths. Changes in CL spectra by irradiation damages were also examined for Y2O3 and SiC by ion beam irradiations. Since luminescence intensity decreased significantly for the specimens with low crystallinities, an SEM system which can control the electron beam current more than 2 orders while keeping the spatial resolution of <1μm is suitable for quick acquisition of CL spectra in the development of the functional materials. Once a database of CL spectra for candidate materials and their changes by crystallinities and irradiation damages is constructed, the CL measurement would be a reliable tool also in inspection and monitoring of materials during reactor construction and operation.

Analysis and experimental justification of electrical connector for ITER blanket module

In ITER blanket system, electrical connectors (“E–straps”, ES) are used to form a low impedance electrical path from shield blocks (SB) to the vacuum vessel (VV). Main functions of ES is providing current from SB to VV. ES shall withstand electromagnetic (EM) loads and Joule heating resulted from electrical current with magnitude up to 137kA during 300ms, accommodate cyclic relative displacements SB-to-VV and operate in harsh environment: neutron radiation, thermal cycling and high vacuum.Manufacturing of the ES from one solid blank by electrical discharge machining (EDM) has been developed and proposed by NIKIET instead of the assembling from separate parts. This paper presents design and analyses efforts applied on development of ES made of solid blank. Patterns of temperature, EM loads and stress-strain state, accumulated fatigue damages have been found with numerical simulations for all specified loading conditions. Strength assessment shows that current design of ES meets the applicable design code: structural design criteria for in-vessel components (SDC-IC). Results of cyclic mechanical tests at temperature are presented in this paper.

Formation of Cr2O3 layers on coolant duct materials for suppression of hydrogen permeation

Oxidation control of candidate duct materials for liquid cooled blanket systems (i.e. SS316, SS316L, Inconel 600 and Hastelloy X) was studied to form a single composition Cr2O3 layer on the surfaces for (i) making a stable and robust underlayer for ceramic coating fabrication by the metal organic decomposition (MOD) method and (ii) suppression of hydrogen permeation by the Cr2O3 layer itself. In heat treatments of the bare plates at 700°C in air of atmospheric pressure and air of reduced pressure of 5Pa, oxidized Fe and/or Ni diffused to the outermost surfaces of the oxide layers and a single composition layer of Cr2O3 could not obtained on the surfaces. However, the Cr2O3 surface layers could be formed by a heat treatment of chrome plated plates at 700°C in air of atmospheric pressure. MOD coatings were successfully fabricated on the stable Cr2O3 surface layers. The preliminary evaluation of deuterium permeation through the Cr2O3 covered plates showed that the permeation reduction factors (PRFs) of the Cr2O3 layers would be 1–2 orders.

Study on Equilibrium Compositions of HTO/H2 Mixture in Purge Gas Flow in Tritium Breeding Blanket

The equilibrium chemical compositions of purge gas at the outlet of the Test Blanket Module in Helium Cooled Ceramic Reflector (HCCR) Test Blanket System are studied. Mole fractions of H, T, O, H2, HT, T2, H2O, HTO, and T2O in the equilibrium state are calculated by a Gibbs free energy minimization method starting from the initial state of H2/HTO mixture. The standard Gibbs free energy for tritium species obtained by the density functional theory is used in the calculations. The tritium recovery rates in the form of HT, T2, HTO and T2O are estimated from the equilibrium chemical compositions obtained in the calculations. The effects of H2 concentration in the purge gas on the tritium recovery process are also investigated by parametric study with variations of purge gas flow rate and volume ratio of hydrogen to helium in the purge gas.

Measurements of rates of cooling of a manikin insulated with different mountain rescue casualty bags

BackgroundAccidental hypothermia is common in those who sustain injuries in remote environments. This is unpleasant and associated with adverse effects on subsequent patient outcomes. To minimise further heat loss, a range of insulating systems are available to mountain rescue teams although the most effective and cost-efficient have yet to be determined.MethodsUnder ambient, still, dry, air conditions, a thermal manikin was filled with water at a temperature of 42 °C and then placed into a given insulation system. Water temperature was then continuously observed via an in-dwelling temperature sensor linked to a PROPAQ 100 series monitor and recorded every 10 min for 130 min. This method was repeated for each insulating package.ResultsThe vacuum mattress/Pertex©/fibrepile blanket system, either on its own or coupled with the Wiggy bag, was the most efficient with water temperatures only decreasing by 3.2 °C over 130 min. This was followed by the heavy-weight casualty bags without the vacuum mattress/Pertex©/fibrepile blanket system, decreasing by 4.2–4.3 °C. With the Blizzard bag, a decline in water temperature of 5.4 °C was seen over the study duration while a decrease of 9.5 °C was noted when the plastic survival bag was employed.ConclusionsUnder the still-air conditions of the study, the vacuum mattress/Pertex©/fibrepile blanket was seen to offer comparable insulation effectiveness compared to be both heavy-weight casualty bags. In turn, these three systems appeared more efficient at insulating the manikin than the Blizzard bag or plastic survival bag.

Enhanced treatment of Fischer-Tropsch wastewater using up-flow anaerobic sludge blanket system coupled with micro-electrolysis cell: A pilot scale study

The coupling of micro-electrolysis cell (MEC) with an up-flow anaerobic sludge blanket (UASB) system in pilot scale was established for enhanced treatment of Fischer-Tropsch (F-T) wastewater. The lowest influent pH (4.99±0.10) and reduced alkali addition were accomplished under the assistance of anaerobic effluent recycling of 200% (stage 5). Simultaneously, the optimum COD removal efficiency (93.5±1.6%) and methane production (2.01±0.13m3/m3·d) at the lower hydraulic retention time (HRT) were achieved in this stage. In addition, the dissolved iron from MEC could significantly increase the protein content of tightly bound extracellular polymeric substances (TB-EPS), which was beneficial to formation of stable granules. Furthermore, the high-throughput 16S rRNA gene pyrosequencing in this study further confirmed that Geobacter species could utilize iron oxides particles as electron conduit to perform the direct interspecies electron transfer (DIET) with Methanothrix, finally facilitating the syntrophic degradation of propionic acid and butyric acid and contributing completely methane production.

ITER blanket manifold final design and validation

The Final Design Review for the Blanket Manifold (MA) was successfully held in December 2015. A concerted effort has been necessary to finalise the multi-pipe design, verification by analysis and practical validation to address challenging design requirements, and installation/maintenance processes.The MA provides cooling water to the Blanket System with a nominal inlet pressure and temperature of 4.0MPa and 70°C, respectively. The return circuits differ in temperatures by up to 40°C, which induces shear stress in neighbouring fixed pipe supports. The pipe to support interface has to accommodate conflicting requirements. It has to provide good thermal conductance in order to remove the neutronic heating while electrically insulated to mitigate excessive electromagnetic loads. The thermal expansion due to the average increase of outlet water temperature is accommodated by compliant support legs.Customised design features for both initial installation and subsequent remote maintenance have been introduced to accommodate the as built VV mounting points. This paper provides details as to how the design has mitigated the conflicting requirements, analysis verification, practical validation of design choices and the achievable build tolerances.

Safety and Environment aspects of Tokamak- type Fusion Power Reactor- An Overview

Naturally occurring thermonuclear fusion reaction (of light atoms to form a heavier nucleus) in the sun and every star in the universe, releases incredible amounts of energy. Demonstrating the controlled and sustained reaction of deuterium-tritium plasma should enable the development of fusion as an energy source here on Earth. The promising fusion power reactors could be operated on the deuterium-tritium fuel cycle with fuel self-sufficiency. The potential impact of fusion power on the environment and the possible risks associated with operating large-scale fusion power plants is being studied by different countries. The results show that fusion can be a very safe and sustainable energy source. A fusion power plant possesses not only intrinsic advantages with respect to safety compared to other sources of energy, but also a negligible long term impact on the environment provided certain precautions are taken in its design. One of the important considerations is in the selection of low activation structural materials for reactor vessel. Selection of the materials for first wall and breeding blanket components is also important from safety issues. It is possible to fully benefit from the advantages of fusion energy if safety and environmental concerns are taken into account when considering the conceptual studies of a reactor design. The significant safety hazards are due to the tritium inventory and energetic neutron fluence induced activity in the reactor vessel, first wall components, blanket system etc. The potential of release of radioactivity under operational and accident conditions needs attention while designing the fusion reactor. Appropriate safety analysis for the quantification of the risk shall be done following different methods such as FFMEA (Functional Failure Modes and Effects Analysis) and HAZOP (Hazards and operability). Level of safety and safety classification such as nuclear safety and non-nuclear safety is very important for the FPR (Fusion Power Reactor). This paper describes an overview of safety and environmental merits of fusion power reactor, issues and design considerations and need for R&D on safety and environmental aspects of Tokamak type fusion reactor.

Nuclear Analyses of Indian LLCB Test Blanket System in ITER

Heading towards the Nuclear Fusion Reactor Program, India is developing Lead Lithium Ceramic Breeder (LLCB) tritium breeding blanket for its future fusion Reactor. A mock-up of the LLCB blanket is proposed to be tested in ITER equatorial port no.2, to ensure the overall performance of blanket in reactor relevant nuclear fusion environment. Nuclear analyses play an important role in LLCB Test Blanket System design & development. It is required for tritium breeding estimation, thermal-hydraulic design, coolants process design, radioactive waste management, equipment maintenance & replacement strategies and nuclear safety. The nuclear behaviour of LLCB test blanket module in ITER is predicated in terms of nuclear responses such as tritium production, nuclear heating, neutron fluxes and radiation damages. Radiation shielding capability of LLCB TBS inside and outside bio-shield was also assessed to fulfill ITER shielding requirements. In order to supports the rad-waste and safety assessment, nuclear activation analyses were carried out and radioactivity data were generated for LLCB TBS components. Nuclear analyses of LLCB TBS are performed using ITER recommended nuclear analyses codes (i.e. MCNP, EASY), nuclear cross section data libraries (i.e. FENDL 2.1, EAF) and neutronic model (ITER C-lite v.l). The paper describes a comprehensive nuclear performance of LLCB TBS in ITER.

Qualification of MELCOR and RELAP5 models for EU HCPB TBS accident analyses

The F4E, Amec Foster Wheeler and INL comprehensive methodology for fusion breeding blanket accident analysis, published last year, consists of several phases. Following the selection of the reference accident scenarios, development of detailed accident analysis specifications and assessment of the analysis codes, the Test Blanket Systems (TBS) models are constructed using the selected codes and modelling approaches. The models are then qualified according to a test matrix by comparison with Test Blanket Module (TBM) finite element design analyses, code-to-code comparisons of TBS normal operation and transient cases and sensitivity studies of accident scenarios. The test cases executed gradually move from qualification of separate systems models to complete TBS models and from the simulation of steady state and normal plasma pulses operation to consideration of power excursion, operational transients and accident events. Finally, both qualified models are used to analyse a selected accident scenario of 32h loss-of-offsite power with sensitivity studies dedicated to the uncertainty evaluation. This step completes the qualification process. The qualification results obtained for the EU Helium-Cooled Pebble-Bed (HCPB) TBS models are reported in the paper.