Helium Cooled Research Articles

Status of EU DEMO heat transport and power conversion systems

DEMO in Europe is considered to be the nearest-term reactor design to follow ITER and capable of demonstrating production of electricity, operating with a closed fuel-cycle and to be a facilitating machine between ITER and a commercial reactor. The aim of this paper is to show the design progress of the complex system’s “chain” devoted to the extraction of the plasma generated pulsed thermal power and its conversion into electricity delivered to the grid, including the Primary Heat Transport System (PHTS), the Power Conversion System (PCS) and the Intermediate Heat Transport System (IHTS) – provided by an Energy Storage System (ESS) – in between PHTS and PCS, which is introduced for the scope of smoothing the generated pulsed plasma power removed by PHTS transmitted to PCS for a more continuous conversion/production of electricity and to guaranty plant lifetime. This will be done with reference to two breeding blanket concepts: the Helium Cooled Pebble Bed (HCPB) Breeding Blanket (BB) and the Water Cooled Lithium Lead (WCLL) BB.The PHTS design criteria and the system design finalization through a descriptive 3D CAD model which shows the system inside the tokamak building are presented. The design and operational challenges as well as the on-going development plans to investigate possible design simplifications (e.g. the direct coupling of Breading Blanket PHTS to PCS) and improved technology readiness actions potentially leading to better plant reliability and cost minimisation are briefly described.

A methodology for thermo-mechanical assessment of in-box LOCA events on fusion blankets and its application to EU DEMO HCPB breeding blanket

Abstract Within the framework of EUROfusion research activities, the Karlsruhe Institute of Technology (KIT) is developing the Helium Cooled Pebble Bed (HCPB) breeding blanket (BB) for European demonstration fusion power plant (DEMO), with the support of BME and Wigner-RCP (Hungary) and CIEMAT (Spain). During the development of the BB concept in the past years, the in-box Loss Of Coolant Accident (LOCA) has been identified as a main design driver for BB. During this event, a leakage of pressurized coolant within the blanket box happens (in-box LOCA), pressurizing the blanket module and endangering its structural integrity. This accident has not been investigated in a detailed approach yet. In this paper, a more precise methodology is introduced taking into account the temperature evolution of the structure for assessing the structural integrity of BB under in-box LOCA and has been applied to the HCPB BB. The methodology and results are presented and critically discussed.

Status of the EU DEMO HCLL breeding blanket design development

In the framework of the European “HORIZON 2020” innovation and research programme, the EUROfusion Consortium develops a design of a fusion power demonstrator (DEMO). One of the key components in the fusion reactor is the Breeding Blanket (BB) surrounding the plasma, ensuring tritium self-sufficiency, heat removal for conversion into electricity, and neutron shielding. CEA-Saclay, with the support of Wigner-RCP and Centrum výzkumu Řež, is in charge of the development of one of the four BB concepts investigated in Europe for DEMO: the Helium Cooled Lithium Lead (HCLL) BB. The rationales of the HCLL are the use of Eurofer as structural material, eutectic liquid lithium-lead (PbLi) as tritium breeder and neutron multiplier, and helium gas as coolant.This paper shows the basic description of the DEMO HCLL BB concept and its design evolution during the past years, from a design based on the ITER Test Blanket Module (TBM) concept to a more advanced design called “Advanced-Plus” concept. This new HCLL BB concept that has been designed in order to improve Tritium Breeding Ratio (TBR) and shielding performances is presented. This new reference HCLL BB design has been analyzed and show very promising nuclear performances. Nevertheless, the “Optimized Conservative” concept, based on ITER TBM, is still considered as a robust back-up solution since structural improvements are still necessary on the “Advanced-Plus” concept. Moreover, a new Back Supporting Structure (BSS) is presented in this paper, designed to support the BB modules, with the aim to reduce pressure drops and thermal stresses.

Nuclear analysis of the HCLL “Advanced-Plus” breeding blanket

In the frame of the European R&D effort to develop a fusion demonstrator (DEMO) a new concept of Helium Cooled Lithium Lead breeding blanket, called “Advanced-Plus” design, is studied to improve the tritium breeding. The aim of this paper is to present the nuclear analysis of this new HCLL design carried out with the TRIPOLI-4® Monte Carlo code and the Joint European Fission and Fusion nuclear data version 3.2 (JEFF-3.2). Nuclear quantities such as: neutron flux, tritium breeding ratio, nuclear heating, displacement damage and helium production are reported. Sensitivity analysis is also carried out to investigate the impact, in tritium breeding, of the next DEMO baseline with reduced outboard thickness and the advantage, in tritium production point of view, of the single module segment against the current multi-module segment option.

Conceptual design of HCCR-TBM diagnostics

The design of diagnostics system for the Helium Cooled Ceramic Reflector (HCCR) Test Blanket System (TBS) has now progressed from generic approaches to the conceptual design integration phases. It is proposed to monitor the temperature, strain, magnetic field strength and neutron flux in selected positions of the corresponding Test Blanket Module (TBM). The temperature can be measured by thermocouple sensors to be integrated for the first wall and breeding zones of TBM. The strain will be monitored by appropriate gauges on the first wall of TBM. Several magnetic pick-up coils could be installed to measure magnetic field strength inside the TBM structure. For the neutron flux monitoring, pairs of Neutron Activation System (NAS) probes and Micro-Fission Chambers (MFCs) would be installed in the TBM structure. In this paper, a brief description on the proposed diagnostics system as well as the key components is presented, and the integration challenges of HCCR TBS diagnostics system are outlined.

Investigation on the “advanced-plus” Helium Cooled Lithium Lead Breeding Blanket design concept for TBR enhancement regarding thermal and mechanical behavior

In the framework of the European “HORIZON 2020” research program, the EUROfusion Consortium develops a design of a fusion demonstrator (DEMO). The Breeding Blanket (BB) directly surrounding the plasma is a major component ensuring tritium self-sufficiency, shielding against neutrons from D-T plasmas and heat extraction for electricity conversion. CEA-Saclay, with the support of Wigner-RCP and IPP-CR is responsible of developing the Helium Cooled Lithium Lead (HCLL) BB concept. In order to enhance Tritium Breeding Ratio (TBR), an alternative HCLL BB design, called “Advanced-Plus” concept, has been investigated, with the aim to reduce the amount of steel, while ensuring good thermal and mechanical behavior. The thermal behaviors of the horizontal Stiffening Plates (hSPs) with various channels designs in terms of dimensions and layout have been parsed on simplified FE models for verifying and optimizing the HCLL Breeding Zone (BZ) before analyzing a HCLL BB ¼ model. Moreover, sensitive analyses have been carried out for testing various hSPs pitches, alternative Helium hydraulic schemes inside the hSPs as well as the detachable FW option in applying a zero Heat Flux (HF) on the FW. Furthermore, mechanical calculations have also been performed to assess the behavior of the module in faulted condition. The results obtained with the Cast3M-qualified-FEM code are presented and discussed.

Neutronics Analysis of Helium Cooled Ceramic Breeder Blanket with S-shaped Lithium Zone and Cooling Plate for CFETR

China fusion engineering test reactor (CFETR) is a tokamak fusion experimental device under design to bridge the research and development gaps between international thermonuclear experimental reactor and demonstration. Helium cooled ceramic breeder (HCCB) blanket is the candidate blanket concepts for CFETR. A blanket with S-shaped lithium zone and cooling pipes reduces the space of helium manifold and is beneficial to tritium breeding. The neutronics analyses of the HCCB blanket have been performed with the Monte Carlo code monte carlo N particle transport code and nuclear cross section data from the fusion evaluated nuclear data library-3.1b data library. The tritium breeding ratio without port is 1.22 and meets the requirements of tritium self-sufficiency, as well as, the nuclear heating power of 214 MW is achieved and the global energy multiplication factor is 1.33. The shielding performance was shown to meet the design limit. The aim of the analysis is to provide the reference for the design and optimization of CFETR blanket system.

Analysis and optimization of the manifolds of the HCCB blanket for CFETR

ABSTRACT Thermal-hydraulic performance is a challenging issue in helium-cooled ceramic breeder (HCCB) blanket design due to the extremely complicated working environment and the strict limits of materials temperature. The heat loads deposited on the HCCB blanket comprises of severe surface heat flux from plasma and the volumetric nuclear heat from neutron irradiation, which can be exhausted by the built-in cooling channels of the components. High pressure helium with 8 MPa, distributed from the coolant manifolds is employed as coolant in the blanket. The design and optimization of the manifolds configuration was performed to guarantee the accurate flow control of helium coolant. The flow distribution in the coolant manifolds was investigated based on the structural improvement of manifolds aiming at overall uniform mass flow rates and better flow streamline distribution without obvious vortexes. The peak temperature of different functional materials in the blanket under normal operating condition is below allowable material limits. It is found that the components in the current blanket module could be cooled effectively under the intense thermal loads due to the updated design and optimization analysis of manifolds.

Design optimization and analysis of CN HCCB TBM-set

After Conceptual Design (CD) Approval by ITER orgnization, Chinese Helium Cooled Ceramic Breeder (HCCB) Test Blanket Module (TBM) design has been updated during the preliminary design phase, in which the manufacturability was the main concerned issue. The main components design such as the back manifold zone, the rib&cap, and the breeding zone were updated through the neutronic, thermal, structural and hydraulic analyses with MCNP and ANSYS. The flow diagram of the He coolant and purge gas in TBM was optimized, which can satisfy the heat remove need of TBM. Different materials’ temperature and stresses were well below the limit. The design of the associated TBM shield was also updated in order to simplify production and assembly.

BB LOCA analysis for the reference design of the EU DEMO HCPB blanket concept

Following DEMO configuration in 2015 the design of the Helium Cooled Pebble Bed (HCPB) blanket concept has been significantly improved. For the reference design HCPB2015V3 and the associated primary heat transfer system (PHTS) the breeding blanket (BB) loss of coolant accident (LOCA) has been investigated as the design basis accident (DBA). System code MELCOR186 for fusion is used for the LOCA simulation. The LOCA can take place in the breeder zone (BZ) or in the first wall (FW) of the BB, and it is assumed to be in one affected equatorial blanket module (BM) on the outboard (OB). Different scenarios due to break sizes and plasma shutdown conditions are taken into account. He inventory during normal operation is determined as steady state results. Transient results of LOCA are evaluated for the mass flow rate, pressure, helium temperature, structure temperature and mass. Finally, issues encountered in this analysis are discussed.

Optimization of the pressure drops in the helium in-module manifolds of the EU DEMO “Optimized Conservative” HCLL Breeding Blanket

In Europe the technology to develop DEMO is being supported by HORIZON 2020 research programme under EUROfusion Consortium. One of the key components that DEMO relies on is a working Breeding Blanket (BB). There are currently four Breeding Blanket concepts being developed in Europe, whose primary purpose is to convert kinetic energy of the neutrons into heat and to transfer this converted heat to a coolant, breed tritium to make plant self-sufficient of fuel and to shield magnets from radiation. This paper describes helium distribution system in the “Optimized Conservative” concept of Helium Cooled Lithium Lead (HCLL) BB, according to the HCLL 2014 Design Description Document. This contains three manifolds to distribute the helium in to several parts of the BB namely the First Wall, Cooling Plates, horizontal Stiffening Plates and the vertical Stiffening Plates. The work mainly focuses on optimization study and reduction of the pressure drops in the helium manifolds. The same is achieved by changing several geometrical parameters to keep the pressure drops under the limit. To achieve the results ANSYS-FLUENT commercial Computational fluid dynamic code is used. The final results include the modified geometry which gives lowest pressure drops for the given mass flow rate.

On a Full Three-Dimensional Thermal Structural and Hydraulic Finite Element Model of the JT-60SA Toroidal Field Coils

The JT-60SA toroidal field coils (TFC) are currently being manufactured in Europe, and their assembly is progressing at full speed in QST, Naka. As part of their final acceptance, the coils are tested at nominal current in a dedicated facility of CEA, Saclay. To help defining the cool down strategy for the testing of the TFCs, and also to anticipate their behavior during the cooling down of the JT-60SA tokamak, a fully three-dimensional finite element model of a TFC has been created, which includes finite elements with thermal, mechanical, and hydraulic formulations, to solve simultaneously the convective heat exchange between the helium coolant and the structures, the conductive heat exchange in the structures themselves, and the radiation heat exchange between the TFC and its surroundings. The model has been extensively used for predicting the behavior of the TFC during cool down and warm up, and the results have been benchmarked against the experimental evidence collected during the cold tests carried out in Europe. Several kind of analyses were carried out, from simple cool down simulations, to more sophisticated simulations of fast transient events, like the dynamic quench test which all TFCs have undergone. The model is built using the ANSYS commercial finite element code and is readily up-scalable to any size of similar magnets. This paper describes the rationale behind the definition of the model, and the results obtained in preparation of the first cold tests, and in view of the operation of the tokamak. The associated modeling challenges are reviewed and critically discussed.

Thermal Characteristics of High Temperature Naturally Circulating Helium Cooling Loop

Abstract The paper deals with the process properties in terms of the heat transfer, i.e. the thermal performance of the thermal-process units within a helium loop intended for the testing of the decay heat removal (DHR) from the model of the gas-cooled fast reactor (GFR). The system is characterised by a natural circulation of helium, as a coolant, and assume the steady operating conditions of the circulation. The helium loop consists of four main components: the model of the gas-cooled fast reactor, the model of the heat exchanger for the decay heat removal, hot piping branch and cold piping branch. Using the thermal calculations, the thermal performance of the heat exchanger model and the thermal performance of the gas-cooled fast reactor model are determined. The calculations have been done for several defined operating conditions which correspond to the different helium flow rates within the system.

Simulations and pre-experiment uncertainty analyses for a hydrogen diffusion experiment using a “two side purged membrane” setup

An experiment termed “Q-PETE” representative for the situation in the Helium Cooled Pebble Bed (HCPB) breeding zone and suitable for the validation of relevant tritium transport codes has been devised. In a temperature controlled setup a hydrogen loaded gas is directed over a steel membrane (several millimeter thickness) into which it can permeate. On the other side of the membrane a sweep gas flow collects the permeated hydrogen and transports it to a quadrupole mass spectrometer for quantitative time resolved detection. This paper introduces methods to handle time resolved permeation experiments with sweep gas, considering the residence time distribution of the sweep gas. The methods are used to quantitatively simulate the planned experiments and expected signal distortion effects. Additionally, an uncertainty model is applied to find optimized process- and geometry parameters for the experiments.

Non-linear assessment of critical failure modes in the first wall of the European TBM

The assessment of the different components of the Helium Cooled Pebble Bed (HCPB) and Helium Cooled Lithium Lead (HCLL) Test Blanket Modules (TBMs) based on elastic FE analyses has been performed considering main failure modes and respective structural design criteria. This study has been carried out for the operation at full power and for the so-called Approach 2 cooling that is set up at reduced heat extraction capability for compliance with temperature targets to be relevant to DEMOnstration power plant blanket system. Consequently this situation is the worst condition in terms of cooling performance as a result it reveals that many locations in the First Wall (FW) become critical in the sense that certain criteria are not fulfilled, particularly those for immediate plastic flow localization, progressive deformation (ratcheting) and creep-fatigue interaction. Of course one can better reach code compliance by improving the cooling performance (so called Approach 1) but it will degrade the DEMO relevancy.The critical failure modes are re-assessed performing non-linear analysis using an elasto-viscoplatic model, particularly developed for RAFM steels such as the structural material EUROFER97 by Aktaa & Schmitt, and considering the appropriate design criteria of the inelastic route of RCC-MRx and SDC-IC. Despite the conservatism in these criteria, the results show all-clear signal with respect to immediate plastic flow localization and at least much smaller breaches of the limits with respect to the other failure modes. Reporting details and results of the non-linear failure assessments demonstrates their potential and capability in supporting specific and efficient improvements to the design of highly loaded components. The conservatism in the considered design criteria and ways for its reduction are discussed in addition.

Design and analysis of manifolds for Indian HCCB blanket module

The Indian LLCB blanket concept is one of the Indian DEMO relevant test blanket module, to be tested in ITER as a part of the TBM program. Helium-Cooled Ceramic Breeder (HCCB) is an alternative blanket concept for its DEMO, which consists of lithium titanate (Li2TiO3) as ceramic breeder (CB) material in the form of packed pebble beds and beryllium as the neutron multiplier. Indian TBM program in ITER is one of the major steps in Indian fusion reactor program towards DEMO. It establishes the synergy among research institutes and industries for carrying out the R&D activities in the critical areas like design of tritium breeding blankets relevant to DEMO. In this paper, the HCCB blanket based on ITER like fusion reactor has been proposed and analysed. The design of manifolds is an important part in a breeding blanket design since it determines how helium is getting distributed in to different circuits for cooling different zones and components in blanket module. The flow scheme for HCCB blanket has been optimized based on its flow parameters. The inlet helium coolant of 300C comes out at 360C from FW and is then distributed into breeding unit zones, grid plates, top plate and bottom plate in parallel configuration. Hydraulic analysis of different circuits using ANSYS CFX has been performed to estimate the flow distribution from manifolds, velocities and pressure drop in different circuits. Heat transfer coefficients have also been evaluated using the obtained velocities in different cooling channels. The detailed design and analyses of the flow schemes in different circuits of blanket module and the purge helium flow have been discussed in this report. The thermo-mechanical design and analysis at using ANSYS mechanical has also been performed at design and operating conditions for p type and s type damages including fatigue and discussed in this report. This study will further help in the design of DEMO blanket module.

Neutronics analyses for a stellarator power reactor based on the HELIAS concept

A first neutronics analysis of the Helical-Axis Advanced Stellarator (HELIAS) power reactor is conducted in this work. It is based on Monte Carlo (MC) particle transport simulations with the Direct Accelerated Geometry Monte Carlo (DAGMC) approach which enables particle tracking directly on the CAD geometry. A suitable geometry model of the HELIAS reactor is developed, including a rough model of a breeder blanket based on the Helium Cooled Pebble Bed (HCPB) breeder blanket concept.The resulting model allows to perform first neutronic calculations providing a 2D map of the neutron wall loading, a 3D distribution of the neutron flux, and a rough assessment of the tritium breeding capability. It is concluded that the applied methodology, making use of MC particle transport simulations based on the DAGMC approach, is suitable for performing nuclear analyses for the HELIAS power reactor.

Structural assessment of the HCPB breeding blanket segments in the EU DEMO reactor under normal operation and a central plasma disruption

The aim of the work reported in this paper is to evaluate the structural integrity of the Helium Cooled Pebble Bed (HCPB) breeding blanket segments in the EU DEMO reactor under the boundary conditions given by the current version of the attachment system, which connects the segments to the vacuum vessel. A 3D model of a DEMO sector has been developed using the FE software ANSYS and thermo-mechanical analyses are performed considering thermal and mechanical loads under normal operation as well as additional EM loads during a central plasma disruption. The results of the structural analysis are presented and the stresses on the back supporting structure of the breeding blanket segments are evaluated taking into account the RCC-MRx code. The design of back supporting structure fulfills the required criteria for the normal as well as the considered off-normal operation conditions. With respect to the iterative development process, these results represent an important input for the further optimization of the design of the attachment system.

Numerical study on the purge gas flow and heat transfer characteristics in helium cooled solid breeder blanket of CFETR

The helium cooled solid breeder blanket module was proposed and designed to achieve the goal of self-sustainable tritium breeding in China Fusion Engineering Test Reactor (CFETR), which was regarded as the transition from ITER to DEMO fusion reactors. A series of preliminary analyses, including thermal hydraulics, neutronics, structural mechanics and materials, have been performed for the blanket module design. However, most literatures focus on the thermal hydraulic behaviors of high pressure helium coolant flowing in cooling channels, while the thermal hydraulic characteristics of low pressure purge gas flowing slowly in pebble beds of tritium breeding zones has not been investigated in detail. In this work, the three dimensional model of typical helium cooled solid breeder blanket module was established employing Computational Fluid Dynamics (CFD) method. The corrected Ergun equation was adopted to simulate the coolant flow features in the porous medium. The temperature and flow field distributions in the First Wall (FW), neutron multiplier zones and tritium breeding zones were achieved. The most critical parts in the blanket module during normal operation condition were illustrated. Besides, the influences of purge gas inlet velocity and tritium breeding pebble-bed ball diameter on the whole blanket module thermal hydraulic features were revealed. This work could contribute to the design and optimization of helium cooled solid breeder blanket in CFETR.

Tritium technologies and transport modelling: main outcomes from the European TBM Project

The tritium technologies and transport modelling play an essential role for the achievement of the TBM Programme scientific objectives. Tritium technologies deployed in TBS (Test Blanket System) can be divided in processing systems and measurement devices. In the current design of the European HCLL (Helium Cooled Lithium Lead) and HCPB (Helium Cooled Pebble Bed) TBS, tritium extraction from flowing Pb-16Li and tritium concentration in He are the most relevant tritium technologies under development. The optimization and integration of real time tritium concentration measurement devices in TBS is a complex task needed to keep under control the tritium concentration in the TBSs and to support their correct operation. For HCLL and HCPB-TBS application, the tritium concentration measurement devices have to operate in flowing Pb-16Li and in He in a wide range of operating conditions.In parallel with the activities on tritium technologies, an important effort has been spent in the last few years in developing tritium transport modelling capabilities, which is one of the most relevant objectives of the European TBM Programme.This paper first of all provides an overview on the most relevant and updated R&D and design results achieved in the above mentioned areas within the European TBM Project. Recent results on tritium inventory and permeation rate in HCLL-TBS, predicted by a 1-D dynamic system simulation code developed within the Ecosimpro platform, are presented and discussed in the part dealing with the TBS tritium transport modelling development.Considerations on DEMO relevancy are also given, particularly for the tritium processing systems as the change of scale from TBM to a breeding blanket for DEMO may impact their feasibility and performance.