Helium-cooled Lithium Lead Test Blanket Research Articles

Numerical Simulations with RELAP5-3D and RELAP5/mod3.3 of the Second Experimental Campaign on In-Box LOCA Transients for HCLL TBS

In-box LOCA was identified as one of the worst accidental scenarios for the HCLL TBS (Helium Cooled Lithium-Lead Test Blanket System). Aiming to experimentally analyze the consequences of this transient, ENEA designed and built THALLIUM (Test HAmmer in Lead LIthiUM), a facility that reproduces the LiPb loop of the HCLL TBS. Two experimental campaigns were carried out by simulating the rupture of a stiffening plate and the related helium injection in the LiPb loop. The obtained experimental data were used to check the capabilities of RELAP5 system code to reproduce the pressure wave propagation that follows this accident. The first simulations were made with RELAP5-3D using LBE (Lead–Bismuth Eutectic) as a system fluid, as the thermophysical properties of LiPb are tabulated only up to a maximum value of 40 bar in this version of the code. Then, LiPb properties were implemented in RELAP5/mod3.3, after selecting the proper correlations from a literature review. This work summarizes the numerical simulations of the second experimental campaign, which was simulated with both versions of the code. The simulations highlight that the code is able to accurately reproduce the experimental results and that RELAP5-3D is slightly more precise than RELAP5/mod3.3 in predicting the pressure trends.

Numerical simulations with RELAP5-3D of the first experimental campaign on In-box LOCA transient for HCLL TBS

• Results of RELAP5-3D simulations of an experimental campaign in a mock-up of the HCLL TBS LiPb loop. • RELAP5-3D is able to accurately reproduce the pressure wave propagation in the facility. • Two main general differences between numerical and experimental results have been highlighted. • Average discrepancy of about 3.5 bar between experimental data and numerical results on 5 tests. • The code accurately predicts the first pressure peak, the most relevant input for the designers. In this work the capabilities of RELAP5-3D are tested against the experimental data gathered in the first experimental campaign on THALLIUM (Test HAmmer in Lead LIthiUM). THALLIUM, a LiPb facility, was built in 2015 at ENEA R.C. Brasimone with the aim to study the In-box LOCA by reproducing the LiPb loop of the HCLL TBS and the pipe forest at 1:1 scale. HCLL TBS (Helium Cooled Lithium-Lead Test Blanket System) was one of the Breeding Blanket concepts selected for ITER. The ITER classification system identified the In-box LOCA as a critical accidental transient for this type of TBS. For this reason, codes able to reproduce the consequences of this transient are needed to support the design of the LiPb loop. The simulations with RELAP5-3D used LBE (Lead-Bismuth Eutectic) as system fluid, as the thermophysical properties of LiPb are tabulated only up to a maximum value of 40 bar. The results of the simulations suggest that RELAP5-3D is able to reproduce the pressure wave propagation in the facility, with an average discrepancy of about 3.5 bar (i.e. an average percentage discrepancy of about 9.7 %), but also highlight two general discrepancies between computations and experiments.

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.

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.

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.

Sensitivity and uncertainty analyses of the HCLL mock-up experiment

Within the European Fusion Technology Programme dedicated computational methods, tools and data have been developed and validated for sensitivity and uncertainty analyses of fusion neutronics experiments. The present paper is devoted to this kind of analyses on the recent neutronics experiment on a mock-up of the Helium-Cooled Lithium Lead Test Blanket Module for ITER at the Frascati neutron generator. They comprise both probabilistic and deterministic methodologies for the assessment of uncertainties of nuclear responses due to nuclear data uncertainties and their sensitivities to the involved reaction cross-section data. We have used MCNP and MCSEN codes in the Monte Carlo approach and DORT and SUSD3D in the deterministic approach for transport and sensitivity calculations, respectively. In both cases JEFF-3.1 and FENDL-2.1 libraries for the transport data and mainly ENDF/B-VI.8 and SCALE6.0 libraries for the relevant covariance data have been used. With a few exceptions, the two different methodological approaches were shown to provide consistent results. A total nuclear data related uncertainty in the range of 1–2% (1σ confidence level) was assessed for the tritium production in the HCLL mock-up experiment.

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.

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.

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.

On the influence of the supporting frame on the nuclear response of the Helium-Cooled Lithium Lead Test Blanket Module for ITER

Within the European Fusion Technology Programme, very intense research activities have been promoted on the Helium-Cooled Lithium Lead (HCLL) breeding blanket concept with the specific aim of manufacturing a Test Blanket Module (TBM) to be irradiated in ITER. HCLL-TBM is foreseen to be located in an ITER equatorial port, being housed inside a proper steel-supporting frame. In particular, since that frame has been designed to provide two cavities separated by a dividing plate and HCLL-TBM is foreseen to fill just one of them, its nuclear response could vary accordingly to the filling status of the other one, unless the dividing plate is thick enough to isolate the components housed in the two cavities, avoiding any significant nuclear interaction among them. At the Department of Nuclear Engineering of the University of Palermo a research campaign has been launched to investigate the potential impact on the HCLL-TBM nuclear response of either the dividing plate thickness or the filling status of the frame cavities. A computational approach based on the Monte Carlo method has been followed and a detailed parametric study has been carried out considering thicknesses of the dividing plate ranging from 0 to the reference value (20 cm) and assuming, for each case, the presence of either void, or a water-cooled steel plug, or a HCLL-TBM or a Helium-Cooled Pebble Bed (HCPB) TBM in the other frame cavity. The analyses have been carried out by means of the MCNP-4C code, running a large number of histories (2 × 10 7) so that the results obtained are affected by statistical uncertainties lower than 1%. The results obtained are herewith reported and critically discussed.

On the nuclear response of the helium-cooled lithium lead test blanket module in ITER

The helium-cooled lithium lead (HCLL) concept has been recently selected as one of the two European reference designs foreseen for the breeding blanket of a demonstration fusion reactor. In particular, within the framework of the research and development activities on this blanket line, an HCLL test blanket module (TBM) has to be designed and manufactured to be implemented in ITER. At the Department of Nuclear Engineering (DIN) of the University of Palermo, a research campaign has been carried out to investigate the nuclear response of HCLL-TBM inside ITER by a numerical approach based on the Monte Carlo method. A realistic 3D heterogeneous model of HCLL-TBM has been set-up and inserted into an ITER 3D semi-heterogeneous one that simulates realistically the reactor lay-out up to the cryostat. A Gaussian-shaped neutron source has been adopted for the calculations. The main features of the HCLL-TBM nuclear response have been determined, paying particular attention to the deposited power and the tritium production rate together with the spatial distribution of their volumetric densities. The radiation damage of the structural material has also been investigated through the evaluation of displacement per atom and helium and hydrogen production rate.