Fusion Safety Research and Development in JAERI

This paper describes the current status and future plans of the fusion safety research and development regarding to the developments of the dust removal system and safety analysis code and the thermofluid experiments in the Japan Atomic Energy Research Institute (JAERI) for a fusion experimental reactor. The containment of the radioactive material is the key to achieve fusion safety. In the event of accidents, the source terms need to be evaluated with sufficient accuracy. Therefore, in JAERI, the dust characterization have been investigated and the dust removal system using electric force has been developed and tested. A safety analysis code including both thermal and plasma transient analyses under the various event sequences has been developed. Moreover, the preliminary experiments of thermofluid transients in the vacuum vessel such as Ingress of Coolant Event (ICE) and Loss of Vacuum Event (LOVA) have been started and the experimental results using preliminary LOVA/ICE apparatus during 1995–1996 are summarized in this paper.     

Dust tracking techniques applied to the STARDUST facility: First results

An important issue related to future nuclear fusion reactors fueled with deuterium and tritium is the creation of large amounts of dust due to several mechanisms (disruptions, ELMs and VDEs). The dust size expected in nuclear fusion experiments (such as ITER) is in the order of microns (between 0.1 and 1000 μm). Almost the total amount of this dust remains in the vacuum vessel (VV). This radiological dust can re-suspend in case of LOVA (loss of vacuum accident) and these phenomena can cause explosions and serious damages to the health of the operators and to the integrity of the device. The authors have developed a facility, STARDUST, in order to reproduce the thermo fluid-dynamic conditions comparable to those expected inside the VV of the next generation of experiments such as ITER in case of LOVA. The dust used inside the STARDUST facility presents particle sizes and physical characteristics comparable with those that created inside the VV of nuclear fusion experiments. In this facility an experimental campaign has been conducted with the purpose of tracking the dust re-suspended at low pressurization rates (comparable to those expected in case of LOVA in ITER and suggested by the General Safety and Security Report ITER-GSSR) using a fast camera with a frame rate from 1000 to 10,000 images per second. The velocity fields of the mobilized dust are derived from the imaging of a two-dimensional slice of the flow illuminated by optically adapted laser beam. The aim of this work is to demonstrate the possibility of dust tracking by means of image processing with the objective of determining the velocity field values of dust re-suspended during a LOVA.     

Exchange Flow Characteristics in a Tokamak Vacuum Vessel of Fusion Reactor Under the Loss-of-Vacuum Conditions

When a Tokamak vacuum vessel of fusion reactor is broken, buoyancy-driven exchange flows will take place through breaches after the inside pressure of the vacuum vessel (VV) becomes equal to the outside pressure. The exchange flow may bring a mixture of activated dusts and tritium from the inside of the VV to the outside through the breaches. Moreover, the exchange flow may remove decay heat from the plasma-facing components. A preliminary LOVA (Loss Of VAcuum event) apparatus was constructed to investigate quantitative heat transfer characteristics of the exchange flows through the breaches under the LOVA conditions. The results of this study, the relationship between Froude numbers and breach locations in the VV was determined and empirical correlations for the average Froude numbers were derived.     

ITER articulated inspection arm (AIA): R&d progress on vacuum and temperature technology for remote handling

This paper is part of the remote handling (RH) activities for the future fusion reactor ITER. The aim of the R&D program performed under the European Fusion Development Agreement (EFDA) work program is to demonstrate the feasibility of close inspection tasks such as viewing or leak testing of the Divertor cassettes and the Vacuum Vessel (VV) first wall of ITER.It is assumed that a long reach, limited payload carrier penetrates the ITER chamber through the openings evenly distributed around the machine such as In-Vessel Viewing System (IVVS) access or through upper port plugs.To perform an intervention a short time after plasma shut down, the operation of the robot should be realised under ITER conditioning i.e. under high vacuum and temperature conditions (120 °C).The feasibility analysis drove the design of the so-called articulated inspection arm (AIA) which is a 8.2 m long robot made of five modules with a 11 actuated joints kinematics. A single module prototype was designed in detail and manufactured to be tested under ITER realistic conditions at CEA-Cadarache test facility.As well as demonstrating the potential for the application of an AIA type device in ITER, this program is also dedicated to explore the necessary robotic technologies required to ITER's IVVS deployment system.This paper presents the whole AIA robot concept, the first results of the test campaign on the prototype vacuum and temperature demonstrator module.     

Safety Analysis of ITER Test Blanket Module for Water Cooled Blanket with Pebble Bed Breeder

Performance test of test blanket modules in the fusion environment using the International Thermonuclear Experimental Reactor (ITER) is one of the most important mile-stone for the development of the breeding blanket of the fusion power plant. In the design of test blanket modules in the ITER, it is very important to show that test modules do not cause additional safety concern to the ITER. This work has been performed for the evaluation of the preliminary safety of the test blanket module of a water cooled solid blanket, which is the primary candidate of the breeding blanket in Japan currently. Major issues of the evaluation were, establishment of post-accident cooling in the test blanket module, hydrogen gas generation by Be/steam reaction, and pressure increase and spilled water amount by the event of coolant leakage. The analyses results showed that, suppression tank system is necessary to accommodate the over-pressure by the coolant water after pipe break in the box of the test module. Coolant water pipe break of the first wall of the test blanket module will result relatively small impact to the ITER safety because of the small inventory of the coolant water of the test module and large volume of the vacuum vessel of the ITER. However, it was clarified that the water cooled blanket with beryllium pebble as the multiplier will have the potential hazard of the hydrogen formation. Further investigation to maintain the safety on this aspect is required.     

Design optimization of first wall and breeder unit module size for the Indian HCCB blanket module

The Indian test blanket module(TBM) program in ITER is one of the major steps in the Indian fusion reactor program for carrying out the RD activities in the critical areas like design of tritium breeding blankets relevant to future Indian fusion devices(ITER relevant and DEMO).The Indian Lead–Lithium Cooled Ceramic Breeder(LLCB) blanket concept is one of the Indian DEMO relevant TBM,to be tested in ITER as a part of the TBM program.Helium-Cooled Ceramic Breeder(HCCB) is an alternative blanket concept that consists of lithium titanate(Li_2TiO_3) as ceramic breeder(CB) material in the form of packed pebble beds and beryllium as the neutron multiplier.Specifically,attentions are given to the optimization of first wall coolant channel design and size of breeder unit module considering coolant pressure and thermal loads for the proposed Indian HCCB blanket based on ITER relevant TBM and loading conditions.These analyses will help proceeding further in designing blankets for loads relevant to the future fusion device.     

Preliminary thermal-hydraulic and safety analysis of China DFLL-TBM system

China has proposed the dual-functional lithium-lead (DFLL) tritium breeding blanket concept for testing in ITER as a test blanket module (TBM), to demonstrate the technologies of tritium self-sufficiency, high-grade heat extraction and efficient electricity production which are needed for DEMO and fusion power plant. Safety assessment of the TBM and its auxiliary system should be conducted to deal with ITER safety issues directly caused by the TBM system failure during the design process. In this work, three potential initial events (PIEs) – in-vessel loss of helium (He) coolant and ex-vessel loss of He coolant and loss of flow without scram (LOFWS) – were analyzed for the TBM system with a modified version of the RELAP5/MOD3 code containing liquid lithium-lead eutectic (LiPb). The code also comprised an empirical expression for MHD pressure drop relevant to three-dimensional (3D) effect, the Lubarsky–Kaufman convective heat transfer correlation for LiPb flow and the Gnielinski convective heat transfer correlation for He flow. Since both LiPb and He serve as TBM coolants, the LiPb and He ancillary cooling systems were modeled to investigate the thermal-hydraulic characteristic of the TBM system and its influence on ITER safety under those accident conditions. The TBM components and the coolants flow within the TBM were simulated with one-dimensional heat structures and their associated hydrodynamic components. ITER enclosures including vacuum vessel (VV), port cell and TCWS vault were also covered in the model for accident analyses. Through this best estimate approach, the calculation indicated that the current design of DFLL-TBM and its auxiliary system meets the thermal-hydraulic and safety requirements from ITER.     

RAMI Analysis Program Design and Research for CFETR (Chinese Fusion Engineering Testing Reactor) Tokamak Machine

Chinese Fusion Engineering Testing Reactor (CFETR) is a test reactor which shall be constructed by National Integration Design Group for Magnetic Confinement Fusion Reactor of China with an ambitious scientific and technological goal. The reactor has the equivalent scale compared with ITER, but has the complementary function to ITER. CFETR is a demonstration of long pulse or steady-state operation with duty cycle time not less than 0.3–0.5 and the full cycle of tritium self-sustained with TBR not less than 1.2. At the same time it will be exploring options for DEMO blanket and divertor with an easy changeable core by remote handling way. To be able to reach its scientific and technological objectives, as one of technical risks control methods, RAMI analysis need to be done during the hold lifetime of CFETR, from conception design to decommissioning. Base on stating of CFETR lifetime and preliminary operational programme, the RAMI analysis program and process are designed and discussed, it consists of five major steps: (1) functional analysis are performed, (2) calculating reliability block diagrams, (3) analyzing failure mode, effects and criticality analysis, (4) risk mitigation actions are taken to ensure every system is compatibility with RAMI objectives, (5) All the RAMI analysis are integrated as the final RAMI analysis reports to be reviewed in the system final design review. Along with the elements of the analysis the vacuum vessel (VV) system was performed to provide as examples, detailed showing how the CFETR RAMI analysis is carried out. CFETR RAMI analysis guidelines were designed and established, after constantly revised and improved these analysis criteria and programs will become the basis standards for CFETR RAMI analysis. Preliminary RAMI analysis of CFETR VV system was obtained, which will be updated with the VV system design progresses.     

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.     

Safety and environmental aspects of organic coolants for fusion facilities

Organic coolants, such as OS-84, offer unique advantages for fusion reactor applications. These advantages are with respect to both reactor operation and safety. The key operational advantage is a coolant that can provide high temperature (350–400°C) at modest pressure (2–4 MPa). These temperatures are needed for conditioning the plasma-facing components and, in reactors, for achieving high thermodynamic conversion efficiencies (>40%). The key safety advantage of organic coolants is the low vapor pressure, which significantly reduces the containment pressurization transient (relative to water) following a loss of coolant event. Also, from an occupational dose viewpoint, organic coolants significantly reduce corrosion and erosion inside the cooling system and consequently reduce the quantity of activation products deposited in cooling system equipment. On the negative side, organic coolants undergo both pyrolytic and radiolytic decomposition, and are flammable. While the decomposition rate can be minimized by coolant system design (by reducing coolant inventories exposed to neutron flux and to high temperatures), decomposition products are formed and these degrade the coolant properties. Both heavy compounds and light gases are produced from the decomposition process, and both must be removed to maintain adequate coolant properties. As these hydrocarbons may become tritiated by permeation, or activated through impurities, their disposal could create an environmental concern. Because of this potential waste disposal problem, consideration has been given to the recycling of both the light and heavy products, thereby reducing the quantity of waste to be disposed. Preliminary assessments made for various fusion reactor designs, including ITER, suggest that it is feasible to use organic coolants for several applications. These applications range from first wall and blanket coolant (the most demanding with respect to decomposition), to shield and vacuum vessel cooling, to an intermediate cooling loop removing heat from a liquid metal loop and transferring it to a steam generator or heat exchanger.     

Radioactivity Confinement Studies Within the SEAL Program

In the framework of the European SEAL program, investigations have been performed with the aim of optimizing the second confinement function and plant layout with respect to normal operation as well as abnormal operation, including accident conditions. This has been done for two conceptual fusion reactor designs: one using water as the coolant and the other using helium. The starting point of these investigations was the SEAFP project design. For the water-cooled reactor design the studies were focused on design options such as pressure suppression spray system, pressure suppression pool with closed containment or with venting to gravel bed filter and stack, and separate expansion volume optionally operated with a vacuum and equipped with spray system. Similar analyses were performed for the helium-cooled reactor design. The analyses were focused on design options comprising a single, large confinement volume or a vent duct connected to an expansion volume operated at vacuum in comparison with the SEAFP Model 1. The thermal-hydraulic analyses performed with the MELCOR code provide an integrated assessment of the cooling loop and confinement system dynamics.     

Preliminary thermal hydraulic safety analysis of water-cooled ceramic breeder blanket for CFETR

The Chinese fusion engineering test reactor (CFETR) was expected to bridge from the international thermonuclear experimental reactor (ITER) to the demonstration fusion reactor (DEMO). The water-cooled ceramic breeder (WCCB) blanket is one of the blanket candidates for CFETR. In this paper, preliminary thermal hydraulic safety analyses have been carried out using the system safety analysis code RELAP5 originally developed for light water fission reactors. The pulse operation and three typical loss of coolant accidents (LOCAs), namely, in-vessel LOCA, in-box LOCA, and ex-vessel LOCA, were simulated based on steady-state initialization. Simulation results show that important thermal hydraulic parameters, such as pressure and temperature can meet the design criterion which preliminarily verifies the feasibility of the WCCB blanket from the safety point of view.     

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.     

Lubricant coating of dowel for the ITER vacuum vessel gravity support

The ITER vacuum vessel gravity supports located in the lower level shall sustain loads in radial, toroidal and vertical directions. The hinge type VVGS consists of two hinges, upper and lower blocks and dowels. In order to develop the design concept and verify the structural integrity of the hinge system, the design analysis has been performed in detail. Inclination of 15° for the hinge based supporting system was introduced to provide centering force to make stable equilibrium state of the vacuum vessel. Due to this inclination the hinges are rotated by the radial expansion of the VV during operation and baking, respectively. If a dowel is seized in the hinge, the supporting system can be highly stressed due to the restrained displacement in the seized dowel. Therefore, solid lubricant coatings were suggested on dowels in order to avoid seizing in the sliding area. In this work, several sets of coupons were made with different coating materials to investigate the effect according to the selection of coating material. Also, a test facility was designed to cover the ITER relevant loading and boundary conditions, e.g. vacuum condition, temperature, contact pressure, cycles, etc. From those test results, the optimized coating method was found to avoid seizure of dowel in the ITER VVGS.     

Design modification and optimization of the ITER cooling water system

ITER (Latin for “the way”), the largest fusion experimental reactor in the world, is designed to demonstrate the technological feasibility of nuclear fusion energy conversion, at plant scale, from high temperature deuterium-tritium plasma using the Tokamak magnetic confinement arrangement.ITER will have a large vacuum vessel that hosts the plasma facing components. These components include the blanket and the divertor that will operate at temperatures, heat loads, and neutron flux higher than those reached in a nuclear fission power plant reactor.One of the main critical issues of the ITER reactor is the design of the cooling water system to transfer the heat generated in the plasma to the in-vessel components and the heat loads from the auxiliary systems to the environment.This paper describes the current ITER cooling water system and recent design modifications and optimizations.     

Design progress on ITER port plug test facility

To achieve the overall ITER machine availability target, the availability of diagnostics and heating port plugs shall be as high as 99.5%. To fulfill these requirements, it is mandatory to test the port plugs at operating temperature before installation on the machine and after refurbishment.The ITER port plug test facility (PPTF) provides the possibility to test upper and equatorial port plugs before installation on the machine. The port plug test facility is composed of several test stands. These test stands are first used in the domestic agencies and on the ITER Organization site to test the port plugs at the end of manufacturing. Two of these stands are installed later in the ITER hot cell facility to test the port plugs after refurbishment. The port plugs to be tested are the Ion Cyclotron (IC) heating and current drive antennas, Electron Cyclotron (EC) heating and current drive launchers, diagnostics and test blanket modules port plugs.Test stands shall be capable to perform environmental and functional tests. The test stands are composed of one vacuum tank (3.3 m in diameter, 5.6 m long) and the associated heating, vacuum and control systems. The vacuum tank shall achieve an ultimate pressure of 1 × 10^?5 Pa at 100 °C containing a port plug. The heating system shall provide water at 240 °C and 4.4 MPa to heat up the port plugs. Openings are provided on the back of the vacuum tank to insert probes for the functional tests.This paper describes the tests to be performed on the port plugs and the conceptual design of the port plug test facility. The configuration of the standalone test stands and the integration in the hot cell facility are presented.     

RAMI Analysis of HCCB TBS for ITER

ITER is the first worldwide international experimental nuclear fusion facility, which aims to prove the physics and technological basis for future fusion power plants. As main stages of ITER technical risk control, the reliability, availability, maintainability and inspectability (RAMI) approach should be applied to all ITER components during their design phase to reduce potential technical risks. Test blanket modules play a key role in ITER. Helium cooled ceramic breeder (HCCB) TBM is one of TBM concepts which were proposed by China. HCCB TBM and its ancillary system are called HCCB test blanket system (TBS). The RAMI analysis was performed on the conceptual design of the ITER HCCB TBS in this paper. A functional breakdown was prepared in a bottom-up approach, resulting in the system being divided into 3 main functions, 1 support function, 14 sub-functions and 50 basic functions. These functions were described using the IDEF0 method. Reliability block diagrams were prepared to estimate the reliability and availability of each function under the stipulated operating conditions. The inherent availability of the HCCB TBS expected after implementation of mitigation actions was calculated to be 94.69 % over 2 years. A failure modes, effects and criticality analysis was performed with criticality charts highlighting the risk level of the different failure modes with regard to their probability of occurrence and their effects on the availability.     

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.     

Electromagnetic and Structural Analysis on Vacuum Vessel for CFETR During Plasma Major Disruption

China Fusion Engineering Test Reactor (CFETR) is a superconducting magnet tokamak and its goal is to achieve the magnetic confinement fusion. The electromagnetic (EM) transients cause mechanical forces, which represent one of the most vital loads for tokamak vacuum vessel (VV). This paper is focused on calculational methods and results for the EM loads on the simplified but practical model of CFETR VV with respect to plasma major disruption scenarios as a reference of the design and analysis. Commercial finite element method software, ANSYS, was employed to evaluate the eddy current on the VV module with the 22.5 ° sector model for major conducting structure of the tokamak including double-walled VV, T-shape rib, and three ports. The plasma current is damping as exponential function 36 ms corresponding to the current simulating in ITER outputs, which are one of major sources of EM loads on VV components. As the results of calculating the eddy currents and EM forces, stress and deformation on CFETR VV can be obtained, which is useful for the structural design of VV.