Helium Cooled Lithium Lead Test Blanket Modules Research Articles

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.

Tritium Extraction From HCLL/WCLL/DCLL PbLi BBs of DEMO and HCLL TBS of ITER

The tritium balance in DEMO reactor is a key factor for the success of the production of energy from the thermonuclear fusion reactor. Three of the four breeder blankets (BBs) concepts candidates for DEMO used the eutectic Pb–16Li enriched at 90% in 6Li as breeder, the water-cooled lithium–lead, dual-coolant lithium–lead, and helium-cooled lithium–lead (HCLL) BBs; therefore, the design and characterization of the tritium extraction and removal system (TER) from PbLi with high efficiency are a critical issue in the European Roadmap. In ITER research reactor, a PbLi BB concept will be qualified, the HCLL BB. An HCLL test blanket module, PbLi loop, instrumentations, and auxiliary systems will be characterized with the support of European infrastructures. However, the tritium extraction unit from PbLi (TEU) selected and designed for ITER is based on the gas–liquid contactor technology, a reliable technology but with less than 40% efficiency. Instead, the candidates TER technologies for the lead–lithium loops of DEMO BBs, the permeator against vacuum (PAV) and vacuum sieve tray (VST), will not qualified in ITER, because these systems will not be fully mature by the start of the reactor. PAV and VST can theoretically achieve efficiency above 80%. The present works aim to analyze the technologies candidates for ITER and DEMO reactors, describe and compare TEU and TER design for each concept of BB and the integration of TER in DEMO tokamak building taking into account two design requirements: self-sufficient sustainable of fusion nuclear reactor and safety requirements.

Assessment of HCLL-TBM optimum welding sequence scenario to minimize welding distortions

The ITER HCLL-TBM (Helium Cooled Lithium Lead Test Blanket Module) box assembly development implies the welding development of the following components: the Box and the Stiffening Grid (SG) made of vertical and horizontal Stiffening Plates (noted respectively v-SP and h-SP). This multi-chamber box structure in EUROFER97 steel is made of plates cooled by multiple meandering channels where circulates pressurized helium. For the assembly of these components, characterized by numerous multipass welds, Gas Tungsten Arc Welding (GTAW) is envisaged as reference process. Moreover, the TBM has large dimensions and thin plates which makes it very sensitive to welding distortions and is problematic regarding the assembly feasibility and compliance with geometric tolerances. This paper presents the numerical simulation and experimental work performed to optimize the v-SP to box assembly sequence, which is the most critical assembly regarding distortions, in order to minimize welding distortions.One of the technical lock of this study is high calculation times needed for this large component which implies to set up a simplified welding simulation method. The study is composed of three main phases: an experimental-numerical study of a T-joint fillet mock-up GTAW used to develop the preliminary welding procedure and to validate a simplified simulation method, a numerical optimization of the v-SP to box welding sequence via the simplified method, and the experimental application of the optimized v-SP to Box welding sequence on the TBM mock-up. The calculation and comparison of four different v-SP to box welding sequences allowed to identify the best sequence regarding welding distortions and to apply it experimentally.

Feasibility analysis of vacuum sieve tray for tritium extraction in the HCLL test blanket system

This paper describes the quantitative analysis for the design of a tritium extraction system that uses liquid PbLi droplets in vacuum (Vacuum Sieve Tray, VST), for application to the ITER helium-cooled lithium lead (HCLL) test blanket system (TBS). The parametric dependences of tritium extraction efficiency from the main geometrical features such as initial droplet velocity, nozzle head height, nozzle diameter, and flow rate are discussed. With nozzle diameters between 0.4 and 0.6mm, extraction efficiency is estimated from 0.77 to 0.96 at the falling height of 0.5m, with flow rate between 0.2 and 1.0kg/s. The device has a height of 1.6m, within the external dimensions of the HCLL Test Blanket Module (TBM), and no additional pumping power is required. The attained results are considered attractive not only for ITER, but also in view of the application of the VST concept as a candidate tritium extraction system for the European Union's demonstration fusion reactor (DEMO). The extraction efficiency of a single droplet column, which is the basis of the design analysis presented, has been validated experimentally with hydrogen. However, further experiments are required on an integrated system with size relevant to the proposed HCLL-TBS design to validate system-level effects, particularly regarding the desorption process in an array of multiple droplets.

Development of the Helium Cooled Lithium Lead blanket for DEMO

The Helium Cooled Lithium Lead (HCLL) blanket is one of the candidate European blanket concepts selected for the DEMOnstration fusion power plant that should follow ITER. In a fusion power plant, the blanket is one of the key components because of its impact on the plant performance, availability, safety and economics. In 2012, the European Fusion Development Agreement (EFDA) agency issued new specifications for DEMO: this paper describes the work performed to adapt the previous 2007 HCLL-DEMO blanket design to those specifications.A new segmentation has been defined assuming straight surfaces for all blanket modules. Following the Multi Module Segment (MMS) option, all modules are attached to a common back supporting structure which also serves as manifold for Helium and PbLi distribution. A detailed CAD design of the central outboard module has been defined. Thermo-hydraulic and thermo-mechanical analyses on of the First Wall and Breeder Zone have been carried out. For the attachment of the modules to the common backplate, a new solution based on the use of Tie Rods, derived from the design of the corresponding HCLL Test Blanket Module for ITER, has been proposed. This paper also identifies the priorities for further development of the HCLL blanket design.

Determination of the 6Li Content in the LiPb of a Neutronics Mock-Up of the European HCLL TBM

A consortium of several European laboratories has performed neutronics experiments with a representative mock-up of the European helium-cooled lithium-lead (HCLL) test blanket module (TBM) irradiated with DT neutrons from intense neutron generators. The aim of these experiments was to provide experimental data for checking nuclear data and calculational tools for the prediction accuracy of important parameters such as the tritium production rate and neutron and gamma-ray flux spectra. The mock-up consisted of bricks of solid LiPb arranged in layers separated by Eurofer sheets. The 6Li concentration in the LiPb determines the slow neutron flux distribution in the mock-up, and an accurate knowledge of this value is of paramount importance for the analysis of these neutronics experiments. The analysis of the tritium production rate experiments revealed discrepancies between the real 6Li concentration and the one specified by the manufacturer of the LiPb (natural Li composition). Here we report on the investigation of the 6Li concentration in the LiPb with several experimental techniques: 1) time-of-arrival neutron spectra measured inside the mock-up irradiated with short pulses of 14-MeV neutrons from a DT neutron generator, 2) transmission measurements on LiPb bricks with moderated neutrons from an AmBe source to check for differences between bricks, and 3) mass spectroscopic methods on small samples taken from selected LiPb bricks. We found that the 6Li concentration varies only very little between the bricks. The weight fraction of lithium in the LiPb was 0.61 wt% as quoted by the manufacturer, but the 6Li abundance was half of the natural value in lithium, 3.8 at% instead of 7.5 at%.

HCLL TBM design status and development

The Helium Cooled Lithium Lead Test Blanket Module (HCLL-TBM) is one of the two European TBMs, representative of DEMO blanket technology, that shall be ready for installation in ITER starting from the first day of operation. It is developed by the CEA in the framework of the activities of the Test Blanket Module Consortium of Associates and under contract with the European Joint Undertaking Fusion For Energy (F4E). A series of 4 different HCLL-TBMs will be installed into one of the ITER equatorial ports during the experimental campaign. This paper aims at presenting the development process and design status of HCLL-TBM, with particular focus on the so-called “integral-TBM (IN-TBM)”, which will include all the main features of the corresponding DEMO blanket module. The design of such a complex component already poses considerable issues and is complicated by the peculiarities of the ITER environment (high temperature, radiation damage, pulsed plasma operation), which are not always addressed in present nuclear design codes. After recalling the main features of the HCLL IN-TBM, supporting design analyses are presented and manufacturing techniques are discussed.

Numerical benchmarks TRIPOLI − MCNP with use of MCAM on FNG ITER bulk shield and FNG HCLL TBM mock-up experiments

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.

Studies on additional electrical heating of the HCLL TBM breeder zone

Among the test blanket modules (TBMs) to be tested in the successive ITER phases, European design activities have focused on the electromagnetic (EM) TBM to be tested in ITER first phase (H-H), and on the integral (IN) TBM to be tested in the full duty D-T phase. The scope of the present contribution is to report activities performed on the helium cooled lithium lead (HCLL) TBM, in its EM version for the H-H phase. This phase is characterized for the TBM by relevant D-T phase magnetic field, surface heat flux and disruption induced loads, and the lack of nuclear volumetric heating. On one hand, H-H phase gives the opportunity of experiments and use of instrumentation which will not be possible in the irradiated environment of D-T phase. On the other hand, H-H phase specificity is a temperature level in the breeder zone much lower than that in the D-T phase. Main EU objectives assigned to TBM testing in H-H phase for design validation and for the build-up of a reliability database can be fulfilled at this temperature level, which is not the case for the scientific experimental program (MHD, corrosion and tritium permeation). Hence, partial heating of the TBM is foreseen. A technical solution for an electrically heated configuration of the breeder zone has been investigated. Finite element (FE) analyses were performed to assess the thermal behaviour of the TBM for the non-heated and heated cases.

Measurement and analysis of neutron flux spectra in a neutronics mock-up of the HCLL test blanket module

Fast neutron and gamma-ray flux spectra and time-of-arrival spectra of slow neutrons have been measured in a neutronics mock-up of the European Helium-Cooled Lithium–Lead Test Blanket Module with the aim to validate nuclear cross-section data. The mock-up was irradiated with fusion peak neutrons from the DT neutron generator of the Technical University of Dresden. A well characterized cylindrical NE-213 scintillator was inserted into two positions in the LiPb/EUROFER assembly. Pulse height spectra from neutrons and gamma-rays were recorded from the NE-213 output. The spectra were then unfolded with experimentally obtained response matrices of the NE-213 detector. Time-of-arrival spectra of slow neutrons were measured with a 3He counter placed in the mock-up, and the neutron generator was operated in pulsed mode. Monte Carlo calculations using the MCNP code and nuclear cross-section data from the JEFF-3.1.1 and FENDL-2.1 libraries were performed and the results are compared with the experimental results. A good agreement of measurement and calculation was found with some deviations in certain energy intervals.

Neutronics experiments on HCPB and HCLL TBM mock-ups in preparation of nuclear measurements in ITER

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 cross-section 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. In particular, the present paper summarizes the preliminary results of the latest of such experiments, i.e. the one conducted on a mock-up of the HCLL blanket. It describes also the dedicated neutronics R&D activities, including the efforts to specify the neutronics tests and objectives in TBMs in ITER, considering the various design concepts.

Requirements and proposals for control and monitoring measurements of the HCLL TBM

Test Blanket Modules (TBM) operation shall not impact the ITER machine safety/availability. The monitoring of all the features relevant for the control and safe operation of the systems will therefore be indispensable. Under this perspective, the control and monitoring parameters to be measured in the Helium Cooled Lithium Lead (HCLL) TBM system are summarized and tabulated in this paper. Relevant sensors’ range and operating conditions are indicated. Their environmental and geometrical constraints in the TBM and its systems are reported. The most suitable sensors and/or measuring techniques are then suggested. Required characteristics are checked against off-the-shelf ones, when available, and needed development for their adaptation to the TBM environment is identified.

Transient thermo-hydraulical/thermo-mechanical analysis of the HCLL-TBM for ITER

This paper concerns the design calculations and performance evaluation of the helium cooled lithium lead-test blanket module (HCLL-TBM) for ITER. Previous works have shown the importance of integrated multi-physics approaches for enabling correct TBM design and specifications. Particularly, this work presents the results of a transient thermo-fluid/thermal-stress analysis in which the thermo-hydraulical and the thermo-mechanical calculations are sequentially and recursively coupled. The above procedure has been used to study the thermo-mechanical behaviour of the HCLL-TBM taking into account the transient loading conditions due to the pulsed plasma operation in ITER. New finite elements (FE) models have been specifically developed for this task: all calculations have been carried out in three dimensions with representative CAD models, including detailed descriptions of all relevant TBM components. Temperature evolutions in the coolant, structural and breeder materials have been calculated. The results of the FE calculations have been analyzed with respect to thermo-hydraulics and thermo-mechanical stresses in order to verify the compliance with the established design criteria (maximum temperature in the structural material and RCC-MR rules with SDC-IC complements).

Thermal–mechanical and thermal–hydraulic integrated study of the Helium-Cooled Lithium Lead Test Blanket Module

The Helium-Cooled Lithium Lead Test Blanket Module (HCLL-TBM) is one of the two TBM to be installed in an ITER equatorial port since day 1 of operation, with the specific aim to investigate the main concept functionalities and issues such as high efficiency helium cooling, resistance to thermo-mechanical stresses, manufacturing techniques, as well as tritium transport, magneto-hydrodynamics effects and corrosion. In particular, in order to show a DEMO-relevant thermo-mechanical and thermal–hydraulic behavior, the HCLL-TBM has to meet several requirements especially as far as its coolant thermofluid-dynamic conditions and its thermal–mechanical field are concerned. The present paper is focused on the assessment of the HCLL-TBM thermal–mechanical performances under both nominal and accidental load conditions, by adopting a computational approach based on the Finite Element Method. A realistic 3D finite element model of the whole HCLL-TBM, in the horizontal first wall design has been set up, consisting of about 597,000 elements and 767,000 nodes. In particular, since the thermal fields of both the module and the coolant are strictly coupled, the helium flow domain has been modeled too and a thermal contact model has been set up to properly simulate the convective heat transfer between the structure wall and the coolant. Pure conductive heat transfer has been assumed within the Pb–Li eutectic alloy of the breeder units. The volumetric density of the nuclear deposited power, recently calculated at Department of Nuclear Engineering of the University of Palermo by the MCNP 4C code, has been applied as distributed thermal load in order to assess the potential influence on the module thermo-mechanical performances of the markedly non-uniform poloidal and toroidal distributions that have been predicted within the Segment Box. Different loading scenarios have been considered as to the heat flux onto the module First Wall. Steady state and transient thermal–mechanical analyses have been carried out and the detailed spatial distributions of the thermal and mechanical fields obtained as to the considered loading scenarios are reported, together with a critical analysis intended to verify their compliance with the agreed design criteria and the DEMO relevance requirements.

Study of the helium-cooled lithium lead test blanket module nuclear behaviour under irradiation in ITER

Abstract The present paper deals with the detailed investigation of the helium-cooled lithium lead test blanket module (HCLL-TBM) nuclear behaviour under irradiation in ITER, carried out at the Department of Nuclear Engineering of the University of Palermo adopting a numerical approach based on the Monte Carlo method. A realistic 3D heterogeneous model of the HCLL-TBM was set-up and inserted into an ITER 3D semi-heterogeneous model that realistically simulates the reactor lay-out up to the cryostat. A Gaussian-shaped neutron source was adopted for the calculations. The main features of the HCLL-TBM nuclear response were assessed, paying a particular attention to the neutronic and photonic deposited power, the tritium production rate and the spatial distribution of their volumetric densities. Structural material irradiation damage was also investigated through the evaluation of displacement per atom and helium and hydrogen production rates.

A new cooling scheme for the HCLL TBM

The Helium Cooled Lithium Lead (HCLL) blanket is one of the two blanket concepts selected by the European Union to be tested in ITER. It is based on the use of Eurofer as structural material, helium as coolant and eutectic lithium–lead as breeder/neutron multiplier material. The design of the corresponding Test Blanket Module (TBM) for ITER has undergone several revisions in the last years. This paper presents an alternative cooling scheme for the HCLL-TBM, where the First Wall (FW) is cooled by vertical (poloidal) instead of horizontal (toroidal) channels. New Finite Element models have been developed and thermal and thermo-hydraulical analyses of the new design have been performed. Results show that the new cooling scheme presents several advantages with respect to the previous one: (i) the total number of cooling channels in the FW can be reduced; (ii) the overall pressure drops in one cooling channel are lower; (iii) the temperature profile in the breeding zone is more uniform.

The HCLL Test Blanket Module system: Present reference design, system integration in ITER and R&D needs

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.

Tritium Extraction Systems for the European HCLL/HCPB TBMs

One of the most challenging issues for the TBM (Test Blanket Module) testing campaign foreseen in ITER is the operation of TES (Tritium Extraction Systems). This is essential not only to prove the ability to manage correctly the bred tritium but also to validate and qualify the neutronic codes for the prediction of tritium production in view of their use in future fusion plants. Two are the European candidates to be tested in ITER: the HCPB (Helium Cooled Pebble Bed) TBM and the HCLL (Helium Cooled Lithium Lead) TBM. For both these TBM concepts the following points have been addressed in this work:a) the gas stream to be processed by TESb) the TES process flow diagramc) a first assessment of the required space

Design approach for the main ITER test blanket modules for the EU helium cooled lithium–lead blankets

One of the two blanket concepts developed by the EU for testing in ITER is the helium cooled lithium–lead (HCLL) concept. Four types of HCLL test blanket modules are defined in a way that they will provide a progressive qualification of the HCLL blanket line up to an integral demonstration in the final TBM. In this paper the latest design of the final TBM is presented. The most recent developments consist in geometrical optimisations and modifications needed to have a realistic manufacturing sequence, in defining the attachments to the port frame, and in assessing the integration of the testing instrumentation required for each mock-up. The status of studies on sub-components fabrication is presented. Preliminary proposals for the design of the initial TBM are also discussed.

HCLL TBM for ITER-design studies

Within the scope of the design and qualification of a helium-cooled lithium–lead (HCLL) blanket module for DEMO, test blanket modules (TBM), have to be designed and inserted in ITER for testing. Several different TBMs will be tested in order to achieve the different testing objectives, according to the ITER operating scenario. This paper aims at presenting the design status of the so-called “integral-TBM (IN-TBM)”, which will include all main features of the corresponding DEMO blanket module, such as design architectures, Eurofer structural material, functional parameters, and all relevant manufacturing processes.