With the US rejoining ITER, the US chamber technology community has resumed participation in discussion in the ITER Test Blanket Working Group (TBWG) and has proposed to develop, in collaboration with other parties, liquid and solid... more
With the US rejoining ITER, the US chamber technology community has resumed participation in discussion in the ITER Test Blanket Working Group (TBWG) and has proposed to develop, in collaboration with other parties, liquid and solid breeder blanket concepts to be tested in ITER. Presently, the US focus on the liquid breeder option is the dual coolant heliumcooled reduced activation ferritic steel structure with self-cooled Pb-17Li breeder (DCLL) that uses flow channel insert (FCI) as the MHD and thermal insulator. When projected for a reference tokamak power reactor design, it has the potential for a gross thermal efficiency of > 40%. The US is planning for an independent test blanket module (TBM) that will occupy half an ITER test port with corresponding supporting ancillary equipment. An initial design, testing strategy and corresponding test plan have been completed for the DCLL concept. The DCLL TBM conceptual design for the integrated testing phase, including the choice of configuration, relevant design analyses, ancillary equipment, testing strategy and corresponding test plan, have been prepared for the transition into the preliminary design phase.
The conceptual design of the ancillary systems of the European Test Blanket Modules (TBMs), namely the PbLi circuit, the Helium Cooling Systems (HCSs), the Coolant Purification Systems and Tritium Extraction Systems (TESs), is proceeding... more
The conceptual design of the ancillary systems of the European Test Blanket Modules (TBMs), namely the PbLi circuit, the Helium Cooling Systems (HCSs), the Coolant Purification Systems and Tritium Extraction Systems (TESs), is proceeding as per time schedule adopted by the European Domestic Agency for ITER (Fusion for Energy (F4E)). A general description of these systems, based on the present baseline, is given in this paper.
Recent research results obtained in Europe, Japan, China and the USA on reduced activation ferritic/martensitic (RAFM) steels are reviewed. The present status of different RAFM steel products (plate, powder HIPped steel, many types of... more
Recent research results obtained in Europe, Japan, China and the USA on reduced activation ferritic/martensitic (RAFM) steels are reviewed. The present status of different RAFM steel products (plate, powder HIPped steel, many types of fusion and diffusion welds, unirradiated and irradiated states) is sufficient to present a strong case for the use of the steels in ITER test blanket modules. For application in DEMO, more research is needed, including the use of the International Fusion Materials Irradiation Facility (IFMIF) in order to quantify the effects of large amounts of transmutation products, such as helium and hydrogen.
An accurate evaluation of the toroidal field ripple in ITER has been carried out by finite element models including the presence of the 18 TF coils and a set of ferromagnetic inserts that aim to lower the field ripple well below 1%. As... more
An accurate evaluation of the toroidal field ripple in ITER has been carried out by finite element models including the presence of the 18 TF coils and a set of ferromagnetic inserts that aim to lower the field ripple well below 1%. As shown, a set of ad hoc distributed plates made of AISI 430 stainless steel (Bsat = μ0Msat ∼1.5 T) and located at the outboard plasma region side in between the vessel shells can reduce the peak ripple at the plasma boundary to ∼0.4% at full toroidal field (i.e., BTF ∼5.3 T at R = 6.2 m). Better compensation can be achieved by adopting higher magnetic saturation materials (e.g., EUROFER). The Test Blanket Modules pair modeled here and made of EUROFER (Bsat ∼1.8 T at 300 K) introduces a large perturbation to the field ripple up to ∼1.1% at full field.
As far as cooling is concerned, air conditioner is considered as most preferred choice nowadays. However, because of its expensiveness and impairment to environment, it becomes prominent to redesign and develop an equipment to overcome... more
As far as cooling is concerned, air conditioner is considered as most preferred choice nowadays. However, because of its expensiveness and impairment to environment, it becomes prominent to redesign and develop an equipment to overcome such flaws. This article aims inexpensive, reliable and eco-friendly way to dealing with hot summer. Peltier module consisted thermoelectric blanket could be a viable option for this. Operating principle is based on peltier effect. Direct application of electric current to peltier module, evolved heat at one junction and absorbed at other junction. As a results temperature difference is created on both side. Efforts have been made to use this phenomenon in blanket. Thermoelectric blanket consists of 16 pieces of peltier modules which absorbs heat from inner side of blanket and dissipate it into outside environment via heat sink. This project focuses on providing comfort at minimal cost.
The HCPB (Helium Cooled Pebble Bed) and HCLL (Helium Cooled Lithium Lead) Test Blanket Modules (TBMs), developed in EU to be tested in ITER, adopt helium at 80 bar as primary coolant.
Testing of breeding blanket modules (TBMs) is one of the ITER goals foreseen from the very beginning of the ITER Project. Six half port TBMs and associated systems are expected to be tested simultaneously in three available Test Ports.... more
Testing of breeding blanket modules (TBMs) is one of the ITER goals foreseen from the very beginning of the ITER Project. Six half port TBMs and associated systems are expected to be tested simultaneously in three available Test Ports. This paper presents an initial assessment of the TBM and ITER interface requirements. Four areas of interface were identified. The first area is the port cell interface area, including components like the port plug frame, backside shield, dummy TBM and corresponding tools needed for the TBM maintenance and replacement. The second area is the hot cell, including the needed additional hardware for the service of TBMs, additional remote handling tools, and additional building space needed for the maintenance of the TBM ancillary equipment and the corresponding testing utilities and tools. The third area is the tokamak cooling water system (TCWS) with the need to accommodate six TBM heat transfer systems, each with a footprint of 57 m 2 . The fourth area of interface is the tritium plant. In all these areas modifications in the current ITER design are needed to accommodate the TMB testing. These changes must be incorporated in the new ITER baseline design which is now under preparation. The latest experiments on JET revealed unexpectedly high sensitivity of plasmas in H-mode of confinement to ripples of the magnetic field. The ferromagnetic test modules can create additional ripples. This new issue of interface between ITER and TBMs is also addressed.
Fusion nuclear technology (FNT) research in the United States encompasses many activities and requires expertise and capabilities in many different disciplines. The US Enabling Technology program is divided into several task areas, with... more
Fusion nuclear technology (FNT) research in the United States encompasses many activities and requires expertise and capabilities in many different disciplines. The US Enabling Technology program is divided into several task areas, with aspects of magnet fusion energy (MFE) fusion nuclear technology being addressed mainly in the Plasma Chamber, Neutronics, Safety, Materials, Tritium and Plasma Facing Component Programs. These various programs work together to address key FNT topics, including support for the ITER basic machine and the ITER Test Blanket Module, support for domestic plasma experiments, and development of DEMO relevant material and technological systems for blankets, shields, and plasma facing components. In addition, two inertial fusion energy (IFE) research programs conducting FNT-related research for IFE are also described. While it is difficult to describe all these activities in adequate detail, this paper gives an overview of critical FNT activities. (N.B. Morley). ponents, systems and technologies of the plasma chamber that are required to contain, shield, extract energy from, and breed tritium fuel for the thermonuclear fusion plasma. FNT advances will be needed both for near-term magnetic (MFE) and inertial (IFE) fusion energy experiments, and ultimately for MFE and IFE energy-producing power reactors. An incomplete list of FNT components and systems includes: 0920-3796/$ -see front matter
Neutronics and nuclear data have an essential role in the development programme leading to the realization of a fusion power plant. A well-qualified nuclear database and validated computational tools are required in the nuclear design of... more
Neutronics and nuclear data have an essential role in the development programme leading to the realization of a fusion power plant. A well-qualified nuclear database and validated computational tools are required in the nuclear design of fusion devices for reliable neutronics and activation calculations with the associated uncertainties. The EU is conducting a large effort on neutronics and nuclear data
Reduced activation ferritic/martensitic steels (RAFMs) are recognized as the primary candidate structural materials for fusion blanket systems. The RAFM F82H was developed in Japan with emphasis on hightemperature properties and... more
Reduced activation ferritic/martensitic steels (RAFMs) are recognized as the primary candidate structural materials for fusion blanket systems. The RAFM F82H was developed in Japan with emphasis on hightemperature properties and weldability. Extensive irradiation studies have conducted on F82H, and it has the most extensive available database of irradiated and unirradiated properties of all RAFMs. The objective of this paper is to review the R&D status of F82H and to identify the key technical issues for the fabrication of an ITER test blanket module (TBM) suggested from the recent research achievements in Japan. This work clarified that the primary issues with F82H involve welding techniques and the mechanical properties of weld joints. This is the result of the distinctive nature of the joint caused by the phase transformation that occurs in the weld joint during cooling, and its impact on the design of a TBM will be discussed.
The losses of high-energy particles from the plasma depend on the toroidal field (TF) ripple in Tokomak machine. TBM (test blanket module), using RAFM (reduced activation ferritic/martensitic) steels as structure material, impacts on TF... more
The losses of high-energy particles from the plasma depend on the toroidal field (TF) ripple in Tokomak machine. TBM (test blanket module), using RAFM (reduced activation ferritic/martensitic) steels as structure material, impacts on TF ripple in International Thermonuclear Experimental Reactor (ITER). The aim in this paper was to investigate the impact of TBM on TF ripple in ITER. It was analyzed based on ANSYS code and the Chinese DFLL (Dual Function Lithium Lead)-TBM as instances of analysis. The results indicated the TF ripple was still beyond the acceptable level of ITER (ı TF < 0.3%) while considering several kinds of configurations (different masses, different dimensions, and different distances to plasma) of the DFLL-TBM. The correction coil might be one way to further reduce the effect on ripple of TF, and the ferromagnetic inserts under TF coil need to continue optimized.
The activities of testing and qualification of the two European Test Blanket Modules (TBMs) to be installed in ITER as well as their main ancillary circuits are of outstanding importance in view of the experimental activities planned... more
The activities of testing and qualification of the two European Test Blanket Modules (TBMs) to be installed in ITER as well as their main ancillary circuits are of outstanding importance in view of the experimental activities planned during ITER lifetime. Having in mind this purpose the EBBTF has been designed and installed at ENEA Brasimone Research Centre in order to meet the important requirements coming directly from the scientific communities to tests and qualify blanket components and technologies in view of ITER and DEMO reactors. ENEA is responsible for the activities of design, construction and installation of the system at ENEA Brasimone and for all of them different national companies have been involved. Essentially, EBBTF consists of the upgraded HeFus3 helium loop coupled with a new lead lithium loop named IELLLO (Integrated European Lead Lithium LOop). The two loops working together will be able to test the HCLL and HCPB mock-up, up to 1:1 scale. The device and its original design solution are described in this paper.
Materials design limits derived so far from the data generated in Europe for the reduced activation ferritic/martensitic (RAFM) steel type Eurofer are presented. These data address the short-term needs of the ITER Test Blanket Modules and... more
Materials design limits derived so far from the data generated in Europe for the reduced activation ferritic/martensitic (RAFM) steel type Eurofer are presented. These data address the short-term needs of the ITER Test Blanket Modules and a DEMOnstration fusion reactor. Products tested include plates, bars, tubes, TIG and EB welds, as well as powder consolidated blocks and solid-solid HIP joints. Effects of thermal ageing and low dose neutron irradiation are also included. Results are sorted and screened according to design code requirements before being introduced in reference databases. From the physical properties databases, variations of magnetic properties, modulus of elasticity, density, thermal conductivity, thermal diffusivity, specific heat, mean and instantaneous linear coefficients of thermal expansion versus temperature are derived. From the tensile and creep properties databases design allowable stresses are derived. From the instrumented Charpy impact and fracture toughness databases, ductile to brittle transition temperature, toughness and behavior of materials in different fracture modes are evaluated. From the fatigue database, total strain range versus number of cycles to failure curves are plotted and used to derive fatigue design curves. Cyclic curves are also derived and compared with monotonic hardening curves. Finally, irradiated and aged materials data are compared to ensure that the safety margins incorporated in unirradiated design limits are not exceeded.
Neutron benchmark experiments carried out at 14 MeV neutron generators on ITER relevant materials and components have provided validation of nuclear data used in shielding and activation calculations for ITER, as well as an assessment of... more
Neutron benchmark experiments carried out at 14 MeV neutron generators on ITER relevant materials and components have provided validation of nuclear data used in shielding and activation calculations for ITER, as well as an assessment of the related uncertainties. Further validation activity is still needed which could be performed in ITER, especially for activation cross sections, for dose rate calculations and for tritium production rates. In particular, concerning the nuclear performance of breeder blankets to be tested in ITER, the paper discusses the extensive preparatory work, to be carried out in a coordinated way among the participating Parties, required to finalize the design of neutronics Test Blanket Modules (TBMs), so that these tests can provide a complete and comparable, therefore useful, picture of the different concepts. Neutronics experiments being carried out at neutron generators on TBM mock-ups are providing important results on the quality of relevant neutron cross sections and for the preparation of the measurement techniques.
The Helium-Cooled Lithium-Lead and the Helium-Cooled Pebble Bed are the two breeding blankets concepts for the DEMO reactor which have been selected by EU to be tested in ITER in the framework of the Test Blanket Module projects. They... more
The Helium-Cooled Lithium-Lead and the Helium-Cooled Pebble Bed are the two breeding blankets concepts for the DEMO reactor which have been selected by EU to be tested in ITER in the framework of the Test Blanket Module projects. They both use a 9%CrWVTa Reduced Activation Ferritic-Martensitic steel, called EUROFER, as structural material and helium as coolant. This paper gives an overview of the status of the EUROFER qualification program and discusses the future needs for design criteria requirements and fabrication validation.
An analysis is carried out on the three-dimensional modeling and computation of the magnetic field in ITER. The commercial finite element code ANSYS-EM is employed for this study. In particular, an emphasis is put on the analysis of the... more
An analysis is carried out on the three-dimensional modeling and computation of the magnetic field in ITER. The commercial finite element code ANSYS-EM is employed for this study. In particular, an emphasis is put on the analysis of the characteristics of non-axisymmetric magnetic fields produced by ferromagnetic materials, including ferromagnetic inserts (FIs) and helium cooled solid breeder test blanket modules (TBMs). It is found that the ITER design requirement for toroidal field ripple is violated by the presence of TBMs, even in the presence of FIs. Calculations of TBM-produced error fields also show that TBM produces a significant error field at q = 2 surface exceeding the ITER design requirement. Discussions are made of the potential implication of the TBM-produced non-axisymmetric fields on plasma performance and the design of a TBM emulation system.
This paper presents the status of the design and of the development programme of the two test blanket systems (TBSs) based on the blanket concepts supported by the EU, namely the helium cooled lithium lead (HCLL) and helium cooled pebble... more
This paper presents the status of the design and of the development programme of the two test blanket systems (TBSs) based on the blanket concepts supported by the EU, namely the helium cooled lithium lead (HCLL) and helium cooled pebble bed (HCPB) concepts.
In support of the breeder blanket development program, the EU is conducting a dedicated neutronics R&D effort to provide the basis for the design of nuclear tests to be performed in ITER on the Test Blanket Modules (TBMs). It includes the... more
In support of the breeder blanket development program, the EU is conducting a dedicated neutronics R&D effort to provide the basis for the design of nuclear tests to be performed in ITER on the Test Blanket Modules (TBMs). It includes the development of computational tools comprising both Monte-Carlo and deterministic transport, sensitivity and uncertainty codes, the generation of high quality neutron crosssection and covariance data libraries. These are validated experimentally in view of their application in the ITER TBM and the DEMO design. To this purpose, two neutronics experiments have been carried out on mock-ups of both the Helium Cooled Pebble Bed (HCPB) and the Helium Cooled Lithium Lead (HCLL) variants of ITER TBMs, at 14-MeV neutron sources. Redundant experimental techniques have been used to measure the resulting tritium production rate and the neutron and gamma ray spectra which are needed to predict the blanket shielding performance, nuclear power production and all nuclear loads. The comparison of experiment and corresponding calculation is obtained with the associated uncertainty margin based on experimental as well as calculational uncertainties. At the same time, suitable nuclear measuring techniques for TBMs in ITER, in particular for the tritium production, are being developed, optimised, and tested in the mock-up experiments.
One of the most challenging issues for testing the different Test Blanket Modules (TBM) concepts in ITER is the demonstration of the ability to correctly and efficiently manage the bred tritium.
This paper gives an overview of the most recent developments for the Helium-Cooled Lithium Lead Test Blanket Modules (HCLL-TBM) in terms of TBM design, related analyses, fabrication developments and safety features. It also addresses the... more
This paper gives an overview of the most recent developments for the Helium-Cooled Lithium Lead Test Blanket Modules (HCLL-TBM) in terms of TBM design, related analyses, fabrication developments and safety features. It also addresses the issues concerning the interfaces of the HCLL-TBM system with ITER and the corresponding proposals of its integration in the ITER machine and buildings. Beside the overview of the progresses realized in several domains of this project, the paper finally outlines the remaining R&D necessary for the main unsolved issues to cope with an installation of the HCLL-TBM system for day one of ITER operation.
Within the framework of the R&D activities promoted by European Fusion Development Agreement on the helium-cooled pebble bed test blanket module to be irradiated in ITER, ENEA Brasimone and the Department of Nuclear Engineering of the... more
Within the framework of the R&D activities promoted by European Fusion Development Agreement on the helium-cooled pebble bed test blanket module to be irradiated in ITER, ENEA Brasimone and the Department of Nuclear Engineering of the University of Palermo performed intense research activities on the modelling of the thermo-mechanical behaviour of both beryllium and lithiated ceramics pebble beds, which are envisaged to be used, respectively, as neutron multiplier and tritium breeder. In particular, at the DIN a thermo mechanical constitutive model was developed for both lithiated ceramics and beryllium pebble beds and it was successfully implemented on a commercial finite element code to analyze the experimental results of the ENEA test campaigns on TAZZA, HELICHETTA and HELICA mock-ups.
One of the ENEA contributions in the activity planned for the European Fusion Technology Programme for 1997-1998 is the design and testing of helium cooled pebble bed blanket (HCPB) mock-ups for the DEMO fusion reactor. The reference... more
One of the ENEA contributions in the activity planned for the European Fusion Technology Programme for 1997-1998 is the design and testing of helium cooled pebble bed blanket (HCPB) mock-ups for the DEMO fusion reactor. The reference medium scale mock-up, called HEXCALIBER (HE-FUS3 experimental cassette of lithium beryllium pebble beds), will reproduce a portion of the HCPB module with the L 4 SiO 4 ceramic breeder and the beryllium neutron multiplier, both in the form of pebble beds heated by flat heaters. The feasibility of the final design and the manufacturing solutions will be first evaluated by testing a reduced portion of the mock-up called HELICA (HE-FUS3 lithium cassette) filled with L 4 SiO 4 breeder pebbles. Aimed at defining the thermal map of the mock-up and the related stresses, some preliminary FEM theoretical calculations were carried-out by ANSYS code. This paper presents the design of two mock-ups, the preliminary calculations, the raw material behaviours and the qualification of the manufacturing procedures. The experimental tests will be carried-out on the HE-FUS3 facility at ENEA Brasimone.
Mock-ups of DEMO breeding blankets, called Test Blanket Modules (TBMs), inserted and tested in ITER in dedicated equatorial ports directly facing the plasma, are expected to provide the first experimental answers on the necessary... more
Mock-ups of DEMO breeding blankets, called Test Blanket Modules (TBMs), inserted and tested in ITER in dedicated equatorial ports directly facing the plasma, are expected to provide the first experimental answers on the necessary performance of the corresponding DEMO breeding blankets. Several DEMO breeding blanket designs have been studied and assessed in the last 20 years. At present, after considering various coolant and breeder combinations, all the TBM concepts proposed by the seven ITER Parties use Reduced-Activation Ferritic/Martensitic (RAFM) steel as the structural material. In order to perform valuable tests in ITER, the TBMs are expected to use the same structural material as corresponding DEMO blankets. However, due to the fact that this family of steels is ferromagnetic, their presence in the ITER vacuum vessel will create perturbations of the ITER magnetic fields that could reduce the quality of the plasma confinement during H-mode. As a consequence, a legitimate question has been raised on the necessity of using RAFM steel for TBMs structural material in ITER. By giving a short description of the main TBM testing objectives in ITER and assessing the consequences of not using such a material, this paper gives a comprehensive answer to this question. According to the working group author of the study, the use of RAFM steel as structural material for TBM is judged mandatory.
a b s t r a c t 3D Monte Carlo (MC) transport codes are of first importance for the assessment of breeding blankets neutronic performances. This article supported by the EFDA Goal Oriented Training Program Eurobreed presents the... more
a b s t r a c t 3D Monte Carlo (MC) transport codes are of first importance for the assessment of breeding blankets neutronic performances. This article supported by the EFDA Goal Oriented Training Program Eurobreed presents the difference in results between the CEA MC code TRIPOLI-4 and MCNP on two fusion neutronics benchmarks, assessing therefore TRIPOLI-4 calculation capabilities on shielding and tritium production rate (TPR). The first selected benchmark, assessing the shielding capability, is the Frascati neutron generator (FNG) ITER bulk shield experiment whereas the second benchmark, assessing the TPR calculation, is the preliminary design of the FNG helium cooled lithium-lead (HCLL) test blanket module (TBM) mock-up. To ensure the consistency of the geometry description, MCAM tool is used for automatic TRIPOLI − MCNP geometry conversions and check. A good coherence between TRIPOLI-4 and MCNP for neutron flux, reaction rates and TPR calculations is obtained. Moreover, it appears that MCAM performs fast, automatic and appropriate TRIPOLI − MCNP geometry conversions and finally that the tabulated FNG neutron source model from KIT is appropriate for TRIPOLI-4 calculations.
A testing strategy and corresponding test plan have been presented for the two proposed US candidate breeder blankets: (1) a helium-cooled solid breeder concept with ferritic steel structure and Be neutron multiplier, but without a fully... more
A testing strategy and corresponding test plan have been presented for the two proposed US candidate breeder blankets: (1) a helium-cooled solid breeder concept with ferritic steel structure and Be neutron multiplier, but without a fully independent TBM and (2) a dual-coolant ...
• Breeding benchmark experiment on LLCB TBM in ITER was performed. • Nuclear responses measured are TPR and reaction rate of 115 In(n, n) 115m In reaction. • Measured responses are compared with calculations by MCNP and FENDL 2.1 library.... more
• Breeding benchmark experiment on LLCB TBM in ITER was performed. • Nuclear responses measured are TPR and reaction rate of 115 In(n, n) 115m In reaction. • Measured responses are compared with calculations by MCNP and FENDL 2.1 library. • TPR measurements agree with calculations in the estimated error bar. • Measured 115 In(n, n) 115m In reaction rates are underestimated by the calculations. a b s t r a c t Experiments were conducted with breeding blanket mock-up consisting of two layers of breeder material lithium titanate pebbles and three layers of pure lead as neutron multiplier. The radial dimensions of breeder, neutron multiplier and structural material layers are similar to the current design of the Indian Lead–Lithium cooled Ceramic Breeder (LLCB) blanket. The mock-up assembly was irradiated with 14 MeV neutrons from DT neutron generator. The local tritium production rates (TPR) from 6 Li and 7 Li in breeder layers were measured with the help of two different compositions of Li isotopes (60.69% 6 Li and 7.54% 6 Li) in Li 2 CO 3. Tritium production in the multiplication layers were also measured with above mentioned two types of pellets to compare the experimental tritium production with calculations. TPR from 6 Li at one location in the breeder layer was also measured by direct online measurement of tritons from 6 Li(n, t) 4 He reaction using silicon surface barrier detector and 6 Li to triton converter. Additional verification of neutron spectra (E n > 0.35 MeV) in the mock-up zones were obtained by measuring 115 In(n, n) 115m In reaction rate and comparing it with calculated values in all five layers of mock-up. All the measured nuclear responses were compared with transport calculations using code MCNP with FENDL2.1 and FENDL3.0 cross-section libraries. The average C/E ratio for tritium production in enriched Li 2 CO 3 pellets was 1.11 in first breeder zone and 1.09 in second breeder zone with uncertainty 8.3% at 1 level. The experimental details and results are discussed in this paper.
Under the US Fusion Nuclear Science and Technology program, we selected the Dual Coolant Lead Lithium (DCLL) concept as our primary Test Blanket Module (TBM) for testing in ITER. The DCLL blanket concept has the potential to be a... more
Under the US Fusion Nuclear Science and Technology program, we selected the Dual Coolant Lead Lithium (DCLL) concept as our primary Test Blanket Module (TBM) for testing in ITER. The DCLL blanket concept has the potential to be a high-performance DEMO blanket design with a projected thermal efficiency of >40%. Reduced activation ferritic/martensitic (RAF/M) steel is used as the structural material. Helium is used to cool the first wall and blanket structure, and the self-cooled Pb-17Li breeder is circulated for power conversion and for tritium extraction. A SiC-based flow channel insert (FCI) is used as an electrical insulator for magnetohydrodynamic pressure drop reduction from the circulating Pb-17Li and as a thermal insulator to separate the high-temperature Pb-17Li (~650°C to 700°C) from the RAF/M structure, which has a corrosion temperature limit of ~480°C. The RAF/M material must also operate at temperatures above 350°C but less than 550°C. We are continuing the development of the mechanical design and performing neutronics, structural and thermal hydraulics analyses of the DCLL TBM module. Prototypical FCI structures were fabricated and further attention was paid to MHD effects and the design of the inboard blanket for DEMO. We are also making progress on related R&D needs to address key areas. This paper is a summary report on the progress and results of recent DCLL TBM development activities.
In support of the breeder blanket development program, the EU is conducting a dedicated neutronics R&D effort to provide the basis for the design of nuclear tests to be performed in ITER on the Test Blanket Modules (TBMs). It includes the... more
In support of the breeder blanket development program, the EU is conducting a dedicated neutronics R&D effort to provide the basis for the design of nuclear tests to be performed in ITER on the Test Blanket Modules (TBMs). It includes the development of computational tools comprising both Monte-Carlo and deterministic transport, sensitivity and uncertainty codes, the generation of high quality neutron crosssection and covariance data libraries. These are validated experimentally in view of their application in the ITER TBM and the DEMO design. To this purpose, two neutronics experiments have been carried out on mock-ups of both the Helium Cooled Pebble Bed (HCPB) and the Helium Cooled Lithium Lead (HCLL) variants of ITER TBMs, at 14-MeV neutron sources. Redundant experimental techniques have been used to measure the resulting tritium production rate and the neutron and gamma ray spectra which are needed to predict the blanket shielding performance, nuclear power production and all nuclear loads. The comparison of experiment and corresponding calculation is obtained with the associated uncertainty margin based on experimental as well as calculational uncertainties. At the same time, suitable nuclear measuring techniques for TBMs in ITER, in particular for the tritium production, are being developed, optimised, and tested in the mock-up experiments.
Basic concepts and the performance of DEMO for an early realization have been investigated with a limited extension of its plasma physics and technology from the second phase of the International Thermonuclear Experimental Reactor (ITER)... more
Basic concepts and the performance of DEMO for an early realization have been investigated with a limited extension of its plasma physics and technology from the second phase of the International Thermonuclear Experimental Reactor (ITER) operation (EPP phase). With the same plasma size as that of ITER, net electric power up to 600 MW is possible with β N > 4.0, H > 1.0 and a divertor heat load of H div < 15 MW/m 2 . Through a consideration of the requirements for a DEMO-relevant blanket concept, Korea has proposed a He cooled molten lithium (HCML) blanket as an ITER TBM. It uses He as a coolant and Li is used as a tritium breeder by considering its potential advantages. Low activation Ferritic Steel (FS) is used as a structural material and two layers of graphite are inserted as a reflector in the breeder zone to increase the tritium breeding ratio (TBR) and the shielding performances. The design and the performance of the KO HCML test blanket module (TBM) are being modified in terms of its He coolant efficiency and its optimized path with a performance analysis; with a 3D Monte Carlo analysis (MCCARD code) for the neutronics; with the CFD code (CFX-10) for the thermal-hydraulics; with ANSYS-10 for the thermo-mechanical analysis.
The US is proposing a prototype of a dual coolant liquid lead-lithium DEMO blanket concept for testing in the International Thermonuclear Experimental Reactor (ITER) as an ITER test blanket module (TBM). Because safety considerations are... more
The US is proposing a prototype of a dual coolant liquid lead-lithium DEMO blanket concept for testing in the International Thermonuclear Experimental Reactor (ITER) as an ITER test blanket module (TBM). Because safety considerations are an integral part of the design process to ensure that this TBM does not adversely impact the safety of ITER, a safety assessment has been conducted for this TBM and its ancillary systems as requested by the ITER project. Four events were selected by the ITER international team (IT) to address specific reactor safety concerns, such as vaccum vessel (VV) pressurization, confinement building pressure build-up, TBM decay heat removal capability, tritium and activation products release from the TBM system and hydrogen and heat production from chemical reactions. This paper summarizes the results of this safety assessment conducted with the MELCOR computer code.
The US is proposing a test blanket module (TBM) to be placed in half of the three dedicated test ports of ITER. The TBM is based on the dual coolant lithium lead (DCLL) blanket concept. Conventional ferritic steel (F82H) is used as the... more
The US is proposing a test blanket module (TBM) to be placed in half of the three dedicated test ports of ITER. The TBM is based on the dual coolant lithium lead (DCLL) blanket concept. Conventional ferritic steel (F82H) is used as the structure of the first wall (FW), the two breeder channels, the back plate, the inlet/out piping, and the shield plug. Two separate cooling circuits are employed: helium is used to cool the FW and blanket structure while the Pb-17Li is used as a coolant and breeder mainly in the two breeder channels. SiC flow channel inserts (FCI) are used to thermally and electrically isolate the flowing Pb-17Li from the relatively lowtemperature structure. A 2 mm thick beryllium layer is used as a plasma facing material on the FW area (1.25 m 2 ) subjected to 0.78 MW/m 2 neutron wall load. In this paper, we present results pertaining to the radioactive inventory and decay heat levels at shutdown and at several post-irradiation times following the pulsed operation scheme of ITER.
A testing strategy and corresponding test plan have been presented for the two proposed US candidate breeder blankets: (1) a helium-cooled solid breeder concept with ferritic steel structure and Be neutron multiplier, but without a fully... more
A testing strategy and corresponding test plan have been presented for the two proposed US candidate breeder blankets: (1) a helium-cooled solid breeder concept with ferritic steel structure and Be neutron multiplier, but without a fully independent TBM and (2) a dual-coolant helium-cooled ferritic steel structure with self-cooled LiPb breeding zone that uses a flow channel insert as MHD and thermal insulator. Example test module designs and configuration choices for each line of ITER TBM are shown and discussed in the paper. In addition, near-term R&D items for decision-making on testing of both solid breeder and dual-coolant PbLi liquid breeder blanket concepts in ITER are identified.
Testing blanket concepts in the integrated fusion environment is one of the principal objectives of ITER. Blanket test modules will be inserted in ITER from Day 1 of its operation and will provide the first experimental data on the... more
Testing blanket concepts in the integrated fusion environment is one of the principal objectives of ITER. Blanket test modules will be inserted in ITER from Day 1 of its operation and will provide the first experimental data on the feasibility of the D-T cycle for fusion. With the US rejoining ITER, the US community has decided to have strong participation
The EU Breeding Blanket Programme aims the testing of two blankets concept in ITER in form of Test Blanket Modules. In the equatorial port #16 the two EU TBMs-a solid and a liquid blanket conceptwill be exposed to the plasma and the... more
The EU Breeding Blanket Programme aims the testing of two blankets concept in ITER in form of Test Blanket Modules. In the equatorial port #16 the two EU TBMs-a solid and a liquid blanket conceptwill be exposed to the plasma and the complex system of their auxiliary systems dedicated to heat and Tritium removal will be integrated in the surrounding ITER buildings. The development of the conceptual design of the EU TBM System is the main objective of the Grant F4E-2008-GRT-09 contract launched by F4E and assigned to a European Consortium. This paper presents an overview of the results after about 20 months of activities: namely, the design of the main subsystems of the EU TBSs and a concept of integration in ITER.
a b s t r a c t 3D Monte Carlo (MC) transport codes are of first importance for the assessment of breeding blankets neutronic performances. This article supported by the EFDA Goal Oriented Training Program Eurobreed presents the... more
a b s t r a c t 3D Monte Carlo (MC) transport codes are of first importance for the assessment of breeding blankets neutronic performances. This article supported by the EFDA Goal Oriented Training Program Eurobreed presents the difference in results between the CEA MC code TRIPOLI-4 and MCNP on two fusion neutronics benchmarks, assessing therefore TRIPOLI-4 calculation capabilities on shielding and tritium production rate (TPR). The first selected benchmark, assessing the shielding capability, is the Frascati neutron generator (FNG) ITER bulk shield experiment whereas the second benchmark, assessing the TPR calculation, is the preliminary design of the FNG helium cooled lithium-lead (HCLL) test blanket module (TBM) mock-up. To ensure the consistency of the geometry description, MCAM tool is used for automatic TRIPOLI − MCNP geometry conversions and check. A good coherence between TRIPOLI-4 and MCNP for neutron flux, reaction rates and TPR calculations is obtained. Moreover, it appears that MCAM performs fast, automatic and appropriate TRIPOLI − MCNP geometry conversions and finally that the tabulated FNG neutron source model from KIT is appropriate for TRIPOLI-4 calculations.
In support of the ITER Test Blanket Module (TBM) program and coordinated by the Test Blanket Working Group, ITER party members have been focusing on the liquid metal blanket design concepts, most of which have been extensively explored.... more
In support of the ITER Test Blanket Module (TBM) program and coordinated by the Test Blanket Working Group, ITER party members have been focusing on the liquid metal blanket design concepts, most of which have been extensively explored. For the demonstration power reactor (DEMO) design, we will have to accommodate the neutron wall loading and first wall heat flux, breed and extract adequate tritium for the D-T fuel cycle and achieve high coolant outlet temperature for high power conversion efficiency. Most proposed liquid metal TBMs have the potential of achieving similar DEMO goals and requirements. Furthermore, all liquid metal TBMs are to satisfy ITER safety requirements and to be operated and tested within ITER operation scenarios. For the development of liquid metal TBM concepts, many R&D elements are common to a few designs such as the areas of Reduced Activation Ferritic/Martensitic Steel (RAFM, also abbreviated as FS in the following) or V-alloy fabrication, thermal fluid MHD, FS/PbLi, FS/Li and V-alloy/Li compatibility, irradiation effects on different materials, tritium extraction, and SiC flow channel insert (FCI) development, etc. With a well-coordinated ITER TBM program, different parties' R&D activities can supplement and complement each other via collaborations. This paper will present respective designs and R&D programs from seven ITER parties.
Europe has developed two reference tritium breeder blankets concepts that will be tested in ITER under the form of Test Blanket Modules: (i) the Helium-Cooled Lithium-Lead which uses the liquid Pb-15.7Li as both breeder and neutron... more
Europe has developed two reference tritium breeder blankets concepts that will be tested in ITER under the form of Test Blanket Modules: (i) the Helium-Cooled Lithium-Lead which uses the liquid Pb-15.7Li as both breeder and neutron multiplier, (ii) the Helium-Cooled Pebble-Bed with lithiated ceramic pebbles as breeder and beryllium pebbles as neutron multiplier.
Large-scale finite element modeling (FEM) of the US Dual Coolant Lead Lithium ITER Test Blanket Module including damage evolution is under development. A comprehensive rate-theory based radiation damage creep deformation code was... more
Large-scale finite element modeling (FEM) of the US Dual Coolant Lead Lithium ITER Test Blanket Module including damage evolution is under development. A comprehensive rate-theory based radiation damage creep deformation code was integrated with the ABACUS FEM code. The advantage of this approach is that time-dependent in-reactor deformations and radiation damage can now be directly coupled with 'material properties' of FEM analyses. The coupled FEM-Creep damage model successfully simulated the simultaneous microstructure and stress evolution in small tensile test-bar structures. Applying the integrated Creep/FEM code to large structures is still computationally prohibitive. Instead, for thermo-structural analysis of the DCLL TBM structure the integrated FEM-creep damage model was used to develop true stress-strain behavior of F82H ferritic steel. Based on this integrated damage evolution-FEM approach it is proposed to use large-scale FEM analysis to identify and isolate critical stress areas for follow up analysis using detailed and fully integrated creep-FEM approach.
In 1996, the European Community started the development of a water-cooled Pb-17Li blanket test module for ITER. First tests are currently scheduled to start with the beginning of the Basic Performance Phase prior to d-t operation. The... more
In 1996, the European Community started the development of a water-cooled Pb-17Li blanket test module for ITER. First tests are currently scheduled to start with the beginning of the Basic Performance Phase prior to d-t operation. The test module is designed to be representative for a DEMO breeding blanket and relies on the liquid alloy Pb-17Li as both tritium breeder and neutron multiplier material, and water at PWR pressure and temperature as coolant. The structural material is martensitic steel. The straight, box-like structure of this blanket confines a pool of liquid Pb-17Li which is slowly circulated for ex-situ tritium extraction and lithium adjustment. The box and the Pb-17Li pool are separately cooled, the former with toroido-radial tubes, the latter with a bundle of double-walled U-tubes, equally made of martensitic steel and equipped with a permeation barrier. This paper presents the latest design and three manufacturing schemes with different degrees of technology. Advanced techniques such as solid or powder HIP are proposed to provide design flexibility. With a 3D neutronics analysis, the power and tritium generation were determined.
In 1996, the European Community started the development of a water-cooled Pb-17Li blanket test module for ITER. First tests are currently scheduled to start with the beginning of the Basic Performance Phase prior to d-t operation. The... more
In 1996, the European Community started the development of a water-cooled Pb-17Li blanket test module for ITER. First tests are currently scheduled to start with the beginning of the Basic Performance Phase prior to d-t operation. The test module is designed to be representative for a DEMO breeding blanket and relies on the liquid alloy Pb-17Li as both tritium breeder and neutron multiplier material, and water at PWR pressure and temperature as coolant. The structural material is martensitic steel. The straight, box-like structure of this blanket confines a pool of liquid Pb-17Li which is slowly circulated for ex-situ tritium extraction and lithium adjustment. The box and the Pb-17Li pool are separately cooled, the former with toroido-radial tubes, the latter with a bundle of double-walled U-tubes, equally made of martensitic steel and equipped with a permeation barrier. This paper presents the latest design and three manufacturing schemes with different degrees of technology. Advanced techniques such as solid or powder HIP are proposed to provide design flexibility. With a 3D neutronics analysis, the power and tritium generation were determined.
An attractive blanket concept for the fusion reactor is the dual coolant Pb-17Li liquid (DCLL) breeder design. Reduced activation ferritic steel (RAFS) is used as the structural material. Helium is used to cool the first wall and blanket... more
An attractive blanket concept for the fusion reactor is the dual coolant Pb-17Li liquid (DCLL) breeder design. Reduced activation ferritic steel (RAFS) is used as the structural material. Helium is used to cool the first wall and blanket structure, and the self-cooled breeder Pb-17Li is circulated for power conversion and for tritium breeding. A SiC f /SiC composite insert is used as the magnetohydrodynamic (MHD) insulation to reduce the impact from the MHD pressure drop of the circulating Pb-17Li and as the thermal insulator to separate the high temperature Pb-17Li from the helium cooled RAFS structure. For the reference tokamak power reactor design, this blanket concept has the potential of satisfying the design limits of RAFS while allowing the feasibility of having a high Pb-17Li outlet temperature of 700 • C. We have identified critical issues for the concept, some of which include the first wall design, the assessment of MHD effects with the SiC-composite flow coolant insert, and the extraction and control of the bred tritium from the Pb-17Li breeder. R&D programs have been proposed to address these issues. At the same time we have proposed a test plan for the DCLL ITER-Test Blanket Module program.
An attractive blanket concept for the fusion reactor is the dual coolant Pb-17Li liquid (DCLL) breeder design. Reduced activation ferritic steel (RAFS) is used as the structural material. Helium is used to cool the first wall and blanket... more
An attractive blanket concept for the fusion reactor is the dual coolant Pb-17Li liquid (DCLL) breeder design. Reduced activation ferritic steel (RAFS) is used as the structural material. Helium is used to cool the first wall and blanket structure, and the selfcooled breeder Pb-17Li is circulated for power conversion and for tritium breeding. A SiC f /SiC composite insert is used as the magnetohydrodynamic (MHD) insulation to reduce the impact from the MHD pressure drop of the circulating Pb-17Li and as the thermal insulator to separate the high temperature Pb-17Li from the helium cooled RAFS structure. have proposed a test plan for the DCLL ITER-Test Blanket Module program.
An attractive blanket concept for the fusion reactor is the dual coolant Pb-17Li liquid (DCLL) breeder design. Reduced activation ferritic steel (RAFS) is used as the structural material. Helium is used to cool the first wall and blanket... more
An attractive blanket concept for the fusion reactor is the dual coolant Pb-17Li liquid (DCLL) breeder design. Reduced activation ferritic steel (RAFS) is used as the structural material. Helium is used to cool the first wall and blanket structure, and the selfcooled breeder Pb-17Li is circulated for power conversion and for tritium breeding. A SiC f /SiC composite insert is used as the magnetohydrodynamic (MHD) insulation to reduce the impact from the MHD pressure drop of the circulating Pb-17Li and as the thermal insulator to separate the high temperature Pb-17Li from the helium cooled RAFS structure. have proposed a test plan for the DCLL ITER-Test Blanket Module program.
A neutronics experiment performed at the Frascati Neutron Generator (FNG) on a mock-up of the Helium-Cooled Pebble Bed (HCPB) breeder test blanket module (TBM) has been analysed on the basis of neutron transport, sensitivity and... more
A neutronics experiment performed at the Frascati Neutron Generator (FNG) on a mock-up of the Helium-Cooled Pebble Bed (HCPB) breeder test blanket module (TBM) has been analysed on the basis of neutron transport, sensitivity and uncertainty calculations using both deterministic and probabilistic computational methods. The calculations revealed a slight but systematic underestimation of the measured tritium activities by 5 to
With the US rejoining ITER, the US chamber technology community has resumed participation in discussion in the ITER Test Blanket Working Group (TBWG) and has proposed to develop, in collaboration with other parties, liquid and solid... more
With the US rejoining ITER, the US chamber technology community has resumed participation in discussion in the ITER Test Blanket Working Group (TBWG) and has proposed to develop, in collaboration with other parties, liquid and solid breeder blanket concepts to be tested in ITER. Presently, the US focus on the liquid breeder option is the dual coolant heliumcooled reduced activation ferritic steel structure with self-cooled Pb-17Li breeder (DCLL) that uses flow channel insert (FCI) as the MHD and thermal insulator. When projected for a reference tokamak power reactor design, it has the potential for a gross thermal efficiency of > 40%. The US is planning for an independent test blanket module (TBM) that will occupy half an ITER test port with corresponding supporting ancillary equipment. An initial design, testing strategy and corresponding test plan have been completed for the DCLL concept. The DCLL TBM conceptual design for the integrated testing phase, including the choice of configuration, relevant design analyses, ancillary equipment, testing strategy and corresponding test plan, have been prepared for the transition into the preliminary design phase.
With the US rejoining ITER, the US chamber technology community has resumed participation in discussion in the ITER Test Blanket Working Group (TBWG) and has proposed to develop, in collaboration with other parties, liquid and solid... more
With the US rejoining ITER, the US chamber technology community has resumed participation in discussion in the ITER Test Blanket Working Group (TBWG) and has proposed to develop, in collaboration with other parties, liquid and solid breeder blanket concepts to be tested in ITER. Presently, the US focus on the liquid breeder option is the dual coolant heliumcooled reduced activation ferritic steel structure with self-cooled Pb-17Li breeder (DCLL) that uses flow channel insert (FCI) as the MHD and thermal insulator. When projected for a reference tokamak power reactor design, it has the potential for a gross thermal efficiency of > 40%. The US is planning for an independent test blanket module (TBM) that will occupy half an ITER test port with corresponding supporting ancillary equipment. An initial design, testing strategy and corresponding test plan have been completed for the DCLL concept. The DCLL TBM conceptual design for the integrated testing phase, including the choice of configuration, relevant design analyses, ancillary equipment, testing strategy and corresponding test plan, have been prepared for the transition into the preliminary design phase.
The helium cooled pebble bed (HCPB) Test blanket module (TBM) for the DEMO Reactor foresees the utilization of lithiate ceramics as breeder in form of pebble beds. The pebbles are organized in several layers alternatively stacked among... more
The helium cooled pebble bed (HCPB) Test blanket module (TBM) for the DEMO Reactor foresees the utilization of lithiate ceramics as breeder in form of pebble beds. The pebbles are organized in several layers alternatively stacked among couples of cooling plates (CP). ENEA has launched an experimental programme for the out-of-pile thermomechanical testing of mock-ups simulating a portion of the HCPB-TBM. The programme foresees the fabrication and testing of different mock-ups, to be tested in the HE-FUS3 facility at ENEA Brasimone. The paper describes the HELICHETTA III campaign carried-out in 2003. In particular, the test section layout, the pebble filling procedure, the experimental setup and the results of the relevant thermo-mechanical test are herewith presented.
