Conceptual design of the blanket mechanical attachment for the helium-cooled lithium-lead reactor

The conceptual design of a new type of fusion reactor based on the helium-cooled lithium-lead (HCLL) blanket has been performed within the European Power Plant Conceptual Studies. As part of this activity, a new attachment system suitable for the HCLL blanket modules had to be developed. This attachment is composed of two parts. The first one is the connection between module and the first part of a shield, called high temperature shield, which operates at a temperature around 500 °C, close to that of the blanket module. This connection must be made at the lateral walls, in order to avoid openings through the first wall and breeding zone thus avoiding complex design and fabrication issues of the module. The second connection is the one between the high temperature shield and a second shield called low temperature shield, which has a temperature during reactor operation around 150 °C. The design of this connection is complex because it must allow the large differential thermal expansion (up to 30 mm) between the two components. Design proposals for both connections are presented, together with the results of finite element mechanical analyses which demonstrate the feasibility to support the blanket and shield modules during normal and accidental operation conditions.     

Engineering studies for integration of the test blanket module (TBM) systems inside an ITER equatorial port plug

The European test blanket module (EU-TBM), first prototype of the breeding blanket concepts under development for the future DEMO power plant to produce the tritium, will be developed to be tested in three equatorial ports of ITER dedicated to this. The CEA Cadarache under the contract of Association EURATOM/CEA and in close relation with Association EURATOM/HAS works on the integration of the EU-TBM inside ITER tokamak.The installation of the TBM into the vacuum vessel is made with the help of a port plug, constituted with two components: the Shield module and the Port-Plug frame. The Shield module provides the neutron shielding inside the Port-Plug frame, which maintains in cantilever position the TBM and its shield module and closes the vacuum vessel port.This paper will describe the EU-TBM design and integration activities on the cooled shield module and on its interface with the TBM component. A particular attention, in term of thermal and mechanical studies, is dedicated to the design of the shield and test blanket module attachment, and also to the shield design and its internal cooling system.     

The technological features of manufacturing process for RF CHC TBM

The development of manufacturing technology for the ceramic helium-cooled test blanket module (CHC TBM) is performed in the framework of the concept for RF Federal government program to master the fusion nuclear energy and as a part of the development of DEMO blanket technology.The main technical approach to the development of CHC TBM manufacturing technology is to provide the “combined” analogy with design decisions of DEMO blanket structural elements.The manufacturing technology of CHC TBM structural elements (first wall cramp, load-bearing back cramp, tritium-breeding element and attachment system) has been proposed during the period of 2004-2007. The design details of TBM structural elements and critical issues of manufacturing technology development are also presented in this paper.     

Progress in engineering design of Indian LLCB TBM set for testing in ITER

The Indian Lead–Lithium Ceramic Breeder (LLCB) Test Blanket Module (TBM) is the Indian DEMO relevant blanket module, as a part of the TBM program in ITER. The LLCB TBM will be tested from the first phase of ITER operation in one-half of an ITER port no. 2. LLCB TBM-set consists of LLCB TBM module and shield block, which are attached with the help of attachment systems. This LLCB TBM set is inserted in a water-cooled stainless steel frame called ‘TBM frame’, which also provides the separation between the neighboring TBM-sets (Chinese TBM set) in port no. 2. In LLCB TBM, high-pressure helium gas is used to cool the first wall (FW) structure and lead–lithium eutectic (Pb–Li) flowing separately around the ceramic breeder (CB) pebble bed to cool the TBM internals which are heated due to the volumetric neutron heating during plasma operation. Low-pressure helium is purged inside the CB zones to extract the bred tritium. Thermal-structural analyses have been performed independently on LLCB TBM and shield block for TBM set using ANSYS. This paper will also describe the performance analysis of individual components of LLCB TBM set and their different configurations to optimize their performances.     

Experimental investigation on the possible techniques of pebbles packing for the HCPB Test Blanket Module

The Helium Cooled Pebble Bed Test Blanket Module (TBM) features a structural box that consists of the first wall, two caps and a stiffening grid. Inside the stiffening grid the breeding units (BUs), consisting of the beryllium and lithium ceramic pebble beds and cooling plates, are accommodated. The BUs are closed by the BU back plates and several structural plates of the manifold system as well as the TBM back plate consequently the BUs may not be accessed directly after the assembly of the TBM box; however, access is possible through dedicated penetrations in the TBM caps. According to the current manufacturing strategy, the assembly of the TBM structural sub-components is based on several welding processes which require post-welding heat treatments (PWHT) at temperatures which exceed the temperature limit of the beryllium pebbles. For that reason the beryllium pebble beds will be packed after the TBM box is assembled and heat treated. The packing of the BUs will be performed using a small-diameter (5 mm) tube that will be inserted into some penetrations in the TBM caps. It is expected that the lithium ceramic pebbles can withstand the high temperatures of the PWHT (this assumption needs to be verified) therefore the current strategy is to pack the ceramic pebble beds during the TBM box assembly. This study experimentally demonstrates the packing procedures for the beryllium beds using a full-scale Plexiglas mock-up as well as the optimization of the packing process by dedicated measures such as vibrating and tilting of the mock-up. In addition the impacts of the experimental results on the TBM design are summarized and the paper is concluded by proposing a packing strategy that can be used to achieve a packing factor of 63.6%.     

A neutronic experiment to support the design of an Indian TBM shield module for ITER

A shield module is associated with an Indian Test Blanket Module (TBM) in ITER to limit the radiation doses in port inter-space areas. The shield module is made of stainless steel plates and water channels. It is identified as an important component for radiation protection because of its radiation exposure control functionality. The radiation protection classification leads to more assurance of the component design. In order to validate and verify the design of the shield module, a neutronic laboratory-scale experiment is designed and executed. The experiment is planned by considering the irradiation under a neutron source of 14 MeV and yields of 10 10 ns ⁻¹. The reference neutron spectrum of the ITER TBM shield module has been achieved through optimization of the neutron source spectrum by a combination of steel and lead materials. The neutron spectrum and flux are measured using a multiple foil activation technique and neutron dose-rate meter LB 6411 (He-3 proton recoil counter with polyethylene), respectively. The neutronic design simulation is assessed using MCNP5 and FENDL 2.1 cross-section data. The paper covers neutronic design, irradiation and the outcome of the experiment in detail.     

Status of ITER TBM port plug conceptual design and analyses

The test blanket module port plug (TBM PP) consists of a TBM frame and two TBM-sets. However, at any time of the ITER operation, a TBM set can be replaced by a dummy TBM. The frame provides a standardized interface with the vacuum vessel (VV)/port structure and provides thermal isolation from the shield blanket. As one of the plasma-facing components, it shall withstand heat loads while at the same time provide adequate neutron shielding for the VV and magnet coils. The frame design shall provide a stable engineering solution to hold TBM-sets and also provide a mean for rapid remote handling replacement and refurbishment. This paper presents main design features of the conceptual design of TBM PP with two dummy TBMs. Also analysis results are summarized to evaluate shielding, hydraulic, and thermal and structural performances of the TBM PP design.     

Mechanical testing of a FW panel attachment system for ITER

An objective of experiments and finite element simulations was to check the stiffness, the strength and the fatigue resistance of the attachment of the First Wall panels onto a shield block of blanket modules according to the ITER 2001 design. The panel has a poloidal key at the rear side (in so-called option A with the rear access bolting) and it is attached by means of special studs located on a key-way in the shield block. Special device for a test of stud tensile pre-load relaxation during a thermal cycling was developed. True-to-scale panels, the shield block mock-up and simplified studs were fabricated and the assembly was loaded alternatively by radial moment, poloidal force or poloidal moment simulating the loading during off-normal plasma operations. Thermal cycling led to an acceptable stud pre-load relaxation. Mechanical cycling caused neither the pre-load relaxation nor the loss of the contact in the key-way nor a damage of the attachment system. The combination of poloidal moment and radial force during vertical displacement events (VDEs) seems to be a most dangerous case because it could lead to the loss of the key–key-way contact.     

Structural analysis of ITER TBM Frame and Dummy TBM

One of the main engineering performance goals of ITER is to test and validate design concepts of tritium breeding blankets. To accomplish these goals, three ITER equatorial ports are dedicated to the test of Test Blanket Modules (TBMs) that are mock-ups of tritium breeding blankets. These TBMs, associated with appropriate shield blocks, will also provide the same thermal and nuclear shielding as the main blanket. The main function of TBM Port Plug (PP) is to accommodate TBMs and provide a standardized interface with the vacuum vessel (VV)/port structure.The feasibility of the design concept of the Frame including two Dummy TBMs has been investigated by proposing design improvements of the reference design through an extensive set of thermal, electromagnetic (EM) and stress analyses. As well, the related static strength was evaluated in accordance with the structural design criteria for ITER in-vessel components (SDC-IC). This paper outlines the engineering aspects of the ITER TBM Frame and Dummy TBM and focuses on the feasibility of the present design by structural assessment.     

Status of ITER blanket attachment design and related R&D

Main function of the ITER blanket system [1], [2], [3] is to shield the vacuum vessel (VV) from nuclear radiation and thermal energy coming from the plasma. Blanket system consists of discrete blanket modules (BM). Each BM is composed of a first wall panel and a shield block (SB). The shield block is attached to the VV by means of four flexible supports and three or four shear keys, through key pads. All listed supports do have parts with ceramic electro-insulating coatings necessary to exclude the largest loops of eddy currents and restrict EM loads. Electrical connection of each SB to the VV is through two elastic electrical straps. Cooling water is supplied to each BM by one coaxial water connector. This paper summarizes the recent evolution of the blanket attachment system toward design solutions compatible with design loads and numbers of load cycles, and providing sufficient reliability and durability. This evolution was done in a frame of pre-defined external interfaces. The ongoing supporting R&D is also briefly described.     

Neutronic design of Indian LLCB TBM

India is developing lead lithium cooled ceramic breeder (LLCB) TBM to be tested in ITER. Liquid lead lithium along with lithium titanate has been adopted as basic material in Indian TBM for neutron multiplication and tritium breeding. RAFMS is used as the structural material and the first wall is cooled by helium. Li-6 enrichment is taken as 60 and 90% in lithium titanate and lead lithium, respectively. The LLCB TBM design is under progress and two design variants are being considered viz. plate design and tube design. In plate design the lead lithium and lithium titanate zones are arranged alternatively and are parallel to the first wall of TBM. In tube design circular tubes of RAFMS are assumed parallel to first wall and lead lithium flows inside the tubes or outside the tubes and lithium titanate is placed accordingly. For the neutronic design of the LLCB TBM, a detailed 3D neutronic model with “look alike” LLCB TBM in equatorial port in ITER has been constructed. A 3D neutron source has been used for the D-T neutrons emitted by plasma. Neutronic study is carried out using Monte Carlo transport code with FENDL-2.1 library with the following objectives: (1) to examine the profiles of heating and tritium production rates in the LLCB TBM, both in the radial and toroidal direction, in order to identify locations where neutronics measurements can be best performed with least perturbation from the surroundings, (2) to provide both local and integrated values for nuclear heating rates required for subsequent thermo-mechanical analysis, and (3) to compare the tritium production capabilities of two variants of the geometries. This paper will present the main findings from this neutronic study.     

Design and setup of a testing device to investigate a reduced sized attachment system mock up for the ITER EU HCPB-TBM under different mechanical loading conditions

In order to determine the forces acting on the EU-Helium Cooled Pebble Bed Test Blanket Module (HCPB-TBM) during operation, a measurement system is developed. Therefore, two force reconstruction (FR) methods using measured strain signals are selected that are suitable for the application to the TBM. The first one, the augmented Kalman filter is a combined deterministic-stochastic approach. A second FR method based on the concept of a model predictive controller is proposed in this paper, which uses an optimization algorithm. In order to test the selected methods a testing device has been built which can be used to apply different force excitations on a reduced sized TBM mock up and measure the resulting strain signals of 16 strain gages. A simple tube mock up has been designed and manufactured to test and calibrate the FR algorithms. In addition, a second TBM mock up with attachment system is described. Finally, first results of the FR of a worst-case test case from simulated strain data of the simple tube mock up are presented.     

Investigation into the in-box LOCA consequence and structural integrity of the KO HCCR TBM in ITER

Korea has developed a Helium Cooled Ceramic Reflector (HCCR) Test Blanket Module (TBM) and its auxiliary system in ITER. In parallel with its design, safety analysis has performed including accident analysis with the selected reference accidents. Among them, the effect of in-box LOCA to the structural integrity of the TBM was investigated. From the transient analysis of the GAMMA-FR on the in-box LOCA, it is found that the pressure of the internal TBM can be increased up to 8 MPa with the same pressure of the operating coolant through the Tritium Extraction System (TES) and He purge lines in the TBM. Structural analysis with ANSYS code for TBM was performed with this condition and it is confirmed that the TBM can endure and it does not affect the ITER machine by the failure.     

Design development and analytical assessment of LLCB TBM in Russian Federation during 2012–2013

The activity on the design, analysis, and R&D for the test blanket module (TBM) with lead–lithium (LL) eutectic coolant and ceramic breeder (CB) was performed in the Russian Federation (RF) according to the technical program of cooperation between the leading research institutes of India (“leader” of the LLCB TBM concept) and RF (“partner”). During the period of 2012–2013, the joint efforts of the RF and Indian specialists were focused on the development of the TBM's basic design with an optimal set of parameters (in particular for testing on both H-H and H-D operation phases of International Thermonuclear Experimental Reactor (ITER) machine). This article briefly describes the results of the TBM design and analysis that have been obtained by the RF specialists (“NIKIET” and D.V. Efremov Institute) in support of the LLCB concept (both DEMO blanket and TBM itself). The main directions of this activity in RF institutes were as follows:

• –development of the TBM design taking into account the ability to manufacture the TBM elements (load-bearing casing, tritium-breeding zone, and attachment system);
• –thermal analysis (in both stationary and transient approaches) of TBM design options (four variations of helium and eutectic flowing directions);
• –structural analysis of TBM design elements for Inductive I operation mode; and
• –recommendations (based upon the results of comparative analysis) on the reference design to be used on further stages of concept development.

The critical issues and further plans on the development of LLCB TBM and corresponding DEMO blanket in the RF are also presented in this article.     

Approach to the lifetime assessment of the bayonet back plate for IFMIF target

The IFMIF facility is aimed at the production of high flux (10¹⁸ n/m²/s) of 14 MeV neutrons to test the candidate Fusion materials under significant neutron damage, up to 50 dpa/year. The conceptual configuration of the IFMIF target, based on the bayonet back plate (BP), has been developed in the past years by several authors. The appropriate engineering design of the back plate, to be developed in the EVEDA (Engineering Validation and Engineering Design Activities) phase, would require a very high level of knowledge on the materials behaviour under irradiation, that will be acquired only after some years of IFMIF experimental activities. For this reason the back plate, which is primarily invested by the highest IFMIF neutron flux, has to be considered a sacrificial component. In spite of its systematic replacement, the engineering design has to be optimised and the lifetime analysis has to be made carefully, in order to credibly estimate the expected replacement frequency. Since the replacement time interval must be conservatively shorter than the back plate lifetime and, at each replacement, the facility has to be stopped for, at least, one week and subjected to risky and uncomfortable operations, it is necessary to perform a trustworthy analysis of the lifetime. To this purpose the various interconnections between the main damaging causes are discussed in order to evidence the most plausible reasons of back plate malfunctioning. Due to the lack of knowledge in some fields and the early stage of design, the analysis is only semi-quantitative. The analysis, which accounts for erosion/corrosion, hydraulic stability, neutron damage and thermo-mechanical stress as the main damaging causes, evidences also the research areas which deserve foremost attention during the EVEDA phase. The considered malfunctions are: lithium boiling, burning/piercing of the back plate, non-sufficient neutron flux, brittle rupture of the back plate, creep rupture, loss of tightness of the back plate sealing.     

Dynamic analysis of AP1000 shield building for various elevations and shapes of air intakes considering FSI effects subjected to seismic loading

The shield building of AP1000 was designed to protect the steel containment vessel of nuclear power plants. When an accident releases mass energy to containment, natural circulation of air outside containment cools steel containment vessel by air intake and water drains by gravity to enhance cooling with evaporation. However, the air intake in the original design located around the upper corner of shield building may not be the optimal position of shield building. In the previous study, the influence of various elevations and shapes of air intake on natural frequency considering fluid-structure effects under different water levels has been performed. In the present study, three elevations and two shapes (rectangle and circle) of air intakes with 71.3, 64.75 and 58.21 m are established and expressed as location I, II and III, respectively. The influences of various elevations and shapes of air intake on the structural response and stress distribution of shield building considering fluid-structure effects under seismic loading are also performed to identify the optimal design for stress analysis to improve the passive cooling system for AP1000 and CAP1400 (in China) in the future. The results of structural analyses indicated that the von Mises stress of both rectangular and circular air intakes at the lower location were greater than that of the higher location, and the stress for circular air intake was less than that of rectangular air intake under seismic loading. In addition, the simulation result also indicated that an optimal elevation of air intake should be implemented around the location II of shield building with circular shape, and the original design of air intake located around the upper corner of shield building may not be the optimal arrangement.     

LOCA analysis for Korean helium cooled solid breeder TBM

One of the major ITER goals is test blanket module (TBM) program which is for the demonstration of the breeding capability that would lead to tritium self-sufficiency in a reactor and the extraction of high-grade heat suitable for electricity generation under the ITER fusion environment. While the engineering design of Korean helium cooled solid breeder (HCSB) TBM and its ancillary systems has been performed, a safety assessment on different possible accident scenarios should be carried out for the purpose of licensing. In this paper, accident analyses for several loss of coolant accident (LOCA) cases were performed in order to assess safety aspects of the TBM design using RELAP5/MOD3.2. Since the TBM forms a loop with helium cooling system (HCS) which is one of ancillary systems required for removing heat deposited in the TBM by neutron wall loading and surface heat flux from plasma, it is necessary to model the complete loop for accident analysis. In this study, the helium passage including the TBM and HCS was nodalized for each accident scenario. The TBM and HCS components were modeled as the associated heat structures provided by RELAP5 to include heat transfer across solid boundaries. Based on computational results it was found that current design of the TBM is robust from the safety point of view.     

Analysis of the reference accidental sequence for safety assessment of LLCB TBM system

Safety analysis of the reference accidental sequence has been carried out for Lead Lithium cooled Ceramic Breeder (LLCB) Test Blanket Module (TBM) system; India's prototype of DEMO blanket concept for testing in International Thermonuclear Experimental Reactor (ITER). The accidental event analyzed starts with a Postulated Initiating Event (PIE) of ex-vessel loss of first wall helium coolant due to guillotine rupture of coolant pipe with simultaneous assumed failure of plasma shutdown system. Three different variants of the sequences analyzed include simultaneous additional failures of TBM and ITER first wall, failure of TBM box resulting in to spilling of lead lithium liquid metal in to vacuum vessel and reactor trip on Loss of Coolant Accident (LOCA) signal from TBM system. The analysis address specific reactor safety concerns, such as pressurization of confinement buildings, vacuum vessel pressurization, release of activated products and tritium during these accidental events and hydrogen production from chemical reactions between lead–lithium liquid metal and beryllium with water. An in-house customized computer code is developed and through these deterministic safety analyses the prescribed safety limits are shown to be well within limits for Indian LLCB-TBM design and it also meets overall safety goal for ITER. This paper reports transient analysis results of the safety assessment.     

Development of remote handling technology of liquid lithium target and replaceable back plate with lip seal in IFMIF-EVEDA

The IFMIF is an accelerator-based intense neutron source for testing candidate fusion materials. Intense neutrons equivalent to neutron irradiation damage of about 50 dPa/y are emitted inside the Li flow through a back plate. Around the back plate, a lip seal made of 316 L is welded by laser-welding system for replacement by remote handling. The back plate will be designed for replacement at least every year. According to material tests of the lip seal weld joint (316 L/316 L) at room temperature, significant deterioration was not observed. Further investigation of the welding process of the lip seal such as a welding direction and a welding joint shape is in progress. Remote handling procedure of the back plate is examined. At first, three lip seal joints of connection piping will be cut by the laser cutting/welding device and then the target assembly with the back plate will be moved to a hot cell. The back plate lip seal will be cut by the laser arm in the hot cell. After machining and Li cleaning of the lip seal, a new back plate will be welded and moved to test cell/target room.