Demonstration Reactor Research Articles

The Feasibility of Application of the Existing IVVS Concept to CFETR

The Chinese Fusion Engineering Test Reactor (CFETR) is the next-generation fusion device of China for realization of fusion energy. Its mission aims to bridge the gaps between the International Thermonuclear Experimental Reactor and the demonstration reactor. To perform the maintenance work on the CFETR, remote handling systems shall be equipped on the device. To inspect and identify the status and position of the in-vessel components of CFETR, a remotely operated in-vessel viewing system (IVVS) shall be used. Such an IVVS prototype with articulated arm and viewing system has been built in China. Its dimension bases on the size of Experimental Advanced Superconducting Tokamak (EAST) due to similar shape of the reactor and the limited project budget. To assemble and store the IVVS around the Tokamak, a cylindrical vacuum vessel with foundation support inside is designed and checked. To isolate the vacuum of the vessel from the vacuum of Tokamak, an ultrahigh vacuum gate valve with a bellow is installed between them. To maintain the vacuum degree of the Tokamak after its connection with the vessel, a set of vacuum unit with molecular pump is configured for the vessel. Before integration testing on the EAST, some performance tests of the IVVS are carried out on the remote handling test facility and in other laboratories. The test results show that it is feasible to apply such an IVVS concept to the CFETR.

Three Confinement Systems—Spherical Tokamak, Standard Tokamak, and Stellarator: A Comparison of Key Component Cost Elements

Since the 1950s “next-step” fusion devices and power plant studies have been developed for a number of magnetic confinement systems but an open question remains: can a magnetic fusion device be simplified to the point where it will be cost competitive and operate with high availability? Concept designs based on the ARIES advanced tokamak, spherical tokamak (ST), and the quasi-axisymmetric (QAS) stellarator option have progressed in recent years through a series of Princeton Plasma Physics Laboratory (PPPL) studies with an underlying intent to improve the engineering feasibility of each, giving special attention to concepts that simplify the device configuration and improve maintenance features. For the ST option, design details centered on a 3-m Fusion Nuclear Science Facility that evolved to incorporate vertical maintenance, high temperature superconductor magnets, a small inboard duel coolant lead lithium blanket, and a liquid metal divertor. In collaboration with the Korean fusion demonstration reactor (K-DEMO) and CFETR concept study teams the tokamak design has evolved to increase plasma component access within a vertical maintenance approach using enlarged toroidal field coils incorporating a low- and high-field Nb3Sn winding pack that provide a peak field of 16 T. A recent PPPL stellarator study focused on simplifying the stellarator winding topology to improve access to in-vessel components; combining coil optimization with winding surfaces that incorporated geometry constraints specified by engineering. This paper centered on a 1000-MW power plant design with a tokamak like vertical maintenance scheme that allows access to remove large internal blanket sectors. Results of three confinement studies (PPPL developed ST, K-DEMO, and QAS stellarator) will be presented to highlight concepts that simplify each device configuration and improved their maintenance features. Scaling each option to a common 1000-MW net electric power plant mission allows comparisons to be made of key cost elements such as the size of major core components, sizing of the reactor hall, or external facilities needed to handle and store activated in-vessel components.

Review of particle-in-cell modeling for the extraction region of large negative hydrogen ion sources for fusion.

Particle-in-cell (PIC) codes are used since the early 1960s for calculating self-consistently the motion of charged particles in plasmas, taking into account external electric and magnetic fields as well as the fields created by the particles itself. Due to the used very small time steps (in the order of the inverse plasma frequency) and mesh size, the computational requirements can be very high and they drastically increase with increasing plasma density and size of the calculation domain. Thus, usually small computational domains and/or reduced dimensionality are used. In the last years, the available central processing unit (CPU) power strongly increased. Together with a massive parallelization of the codes, it is now possible to describe in 3D the extraction of charged particles from a plasma, using calculation domains with an edge length of several centimeters, consisting of one extraction aperture, the plasma in direct vicinity of the aperture, and a part of the extraction system. Large negative hydrogen or deuterium ion sources are essential parts of the neutral beam injection (NBI) system in future fusion devices like the international fusion experiment ITER and the demonstration reactor (DEMO). For ITER NBI RF driven sources with a source area of 0.9 × 1.9 m2 and 1280 extraction apertures will be used. The extraction of negative ions is accompanied by the co-extraction of electrons which are deflected onto an electron dump. Typically, the maximum negative extracted ion current is limited by the amount and the temporal instability of the co-extracted electrons, especially for operation in deuterium. Different PIC codes are available for the extraction region of large driven negative ion sources for fusion. Additionally, some effort is ongoing in developing codes that describe in a simplified manner (coarser mesh or reduced dimensionality) the plasma of the whole ion source. The presentation first gives a brief overview of the current status of the ion source development for ITER NBI and of the PIC method. Different PIC codes for the extraction region are introduced as well as the coupling to codes describing the whole source (PIC codes or fluid codes). Presented and discussed are different physical and numerical aspects of applying PIC codes to negative hydrogen ion sources for fusion as well as selected code results. The main focus of future calculations will be the meniscus formation and identifying measures for reducing the co-extracted electrons, in particular for deuterium operation. The recent results of the 3D PIC code ONIX (calculation domain: one extraction aperture and its vicinity) for the ITER prototype source (1/8 size of the ITER NBI source) are presented.

Shutdown Dose Rate Calculation for the Preliminary Concept of K-DEMO Equatorial Port Area

This paper presents a shutdown dose rate (SDR) calculation to estimate a newly developed preliminary concept of the equatorial port model for the Korean fusion demonstration reactor (K-DEMO). The equatorial port model was designed to provide effective radiation shielding abilities adopting the port shielding block and labyrinth structure, which was integrated into the modified K-DEMO neutronic analysis model by the Monte Carlo automatic modeling system. A rigorous two-step method was applied to couple transport (MCNP) and activation (FISPACT) codes according to three steps of calculation procedures. The results of neutron flux and decay gamma transport calculation for the equatorial port area have been indicated in this paper. The SDR calculation result near the equatorial port area reveals that a dose level in the equatorial port interspace was below $100~\mu $ Sv/h after 12 days of the shutdown. Shielding blocks and two labyrinth structures prove that they effectively attenuate neutrons on the current concept of the equatorial port model.

Current Progress of Tritium Fuel Cycle Technology for CFETR

Tritium fuel cycle technology is crucial for the successful development of China fusion engineering test reactor (CFETR) which aims to bridge the gap between a fusion experimental reactor (ITER) and a demonstration reactor. One of the major missions of CFETR is to realize tritium fuel self-sufficiency with a tritium breeding ratio over 1.0. In order to attain tritium self-sufficiency as well as ensure safe handling of tritium, it is important to establish a closed tritium fuel cycle. Therefore, a tritium plant which consists of various tritium processing systems and tritium safety systems is indispensable for the successful operation of CFETR. Since the mid-2000s, when China joined the ITER project, a series of R&D activities on tritium fuel cycle technology have been carried out. The current progress in terms of conceptual design of tritium plant, development of sub-system technologies and tritium related materials is presented in this paper.

Status of CEA Magnet Design Tools and Applications to EU DEMO PF and CS Magnets

In the framework of the design activities conducted in the EU for dimensioning the future fusion demonstration reactor (DEMO), extensive analyses were carried out within the framework of the EUROfusion consortium, with the objective of defining the design of the DEMO magnets system. To this aim, CEA has developed ad-hoc predimensioning tools and associated methods in order to size the different magnets: toroidal field coils, central solenoid (CS) and poloidal field (PF) coils. Once predimensioned, the magnet concepts undergo a detailed evaluation procedure with more complex tools and methods. As these detailed analyses are time consuming, the whole design process benefits from an accurate and robust predimensioning process. The predimensioning tools described herein address various aspects driving the operational limits of the magnets while energized in the tokamak, related to electromagnetic, thermal, and mechanic phenomena. In this paper, we expose the latest developments in the predimensioning tools and the methods employed for obtaining a rapid and reliable macroscopic semi-analytical representation of the magnets. The assessment of the predimensioning tools on reference configurations (e.g., ITER) is also described. The application of these design tools on the DEMO configuration issued by EUROfusion is presented, and the resulting magnet design proposals are summarized: First, the PF system, composed of six coils, with one PF selected for study here. Second, the CS, with a modular geometry (five modules) and pancake winding. The approach here assesses the resilience of the CS to a fast-transient (breakdown). The outcomes of a sensitivity study on parameters/criteria are discussed, and extended to some tentative recommendations on the design approach for the DEMO PF and CS magnets. For the PF, a design is obtained for the PF5 coil following a dimensioning methodology consistent with the ITER PF system design with the same level of definition of the plasma operating scenario. As for the CS, it is stressed that consideration of the internal field (from shielding) substantially affects temperature margin results.

Thermohydraulic Analyses on CEA Concept of TF and CS Coils for EU-DEMO

In the framework of the European fusion program for energy, EUROfusion funds the studies for the future fusion power demonstrator reactor DEMOnstration fusion reactor (DEMO). Commissariat a l'energie atomique et aux energies alternatives (CEA) is involved in the conceptual design of the superconducting conductors for the toroidal field (TF) and central solenoid (CS) magnets. The CEA design proposal corresponds to wind and react Nb3Sn pancake-wound coils using cable-in-conduit conductors cooled at about 4.5 K by forced flow of supercritical helium. This paper presents the latest thermal-hydraulic analyses performed on both TF and CS conductors. Two TF conductor designs have been analyzed, with nominal currents of 111 kA and 88 kA, respectively. The analyses were performed in normal (burn) and off-normal (quench) conditions. Burn simulations focused on the central and lateral pancakes with heat load corresponding to neutron heating. The central pancake is the most critical one regarding the magnetic field, while the lateral one receives a more important heat load from the casing. The influence of case cooling on temperature margin (ΔTma) was analyzed by means of a dedicated 2D Cast3M model. The ΔTma sensitivity to driving parameters such as conductor friction factor correlations was analyzed. The impact of interturn thermal coupling on temperature margin was also assessed. Quench studies were performed on the TF conductor, considering a quench initiated either on the innermost turn or at the middle of hydraulic length. Regarding the CS conductor, burn scenario was led on the latest design featuring a nominal current of 53.7 kA, focusing on the impact of ac losses and of dwell duration.

Study on melting and solidification behaviors of tungsten loaded by high heat flux for divertor in tokamak fusion reactor

Tungsten is a candidate as a surface material of divertor in Tokamak fusion reactor. A pulsed high heat flux emitted in plasma disruption and Type I Edge Localized Mode (ELM) is considered to melt the tungsten surface repeatedly. The repetitive phase change can transmute surface characteristics of divertor and then reduce the thermal resistance. For the safety design and the development acceleration of fusion demonstration reactor, investigation and understanding of surface melting and solidification procedure are important. In previous studies, experiments have been conducted by Ueda et al. and the surface after the laser irradiation had the geometry as a crater with a center cone. However, the formation mechanism of the crater was not fully understood. In this paper, two measurement methods are studied. One is an in-situ observation method of tungsten and molybdenum surface during the laser irradiation using high speed video camera, and another is a 3-dimensional measurement of tungsten surface shape after the laser irradiation experiments using a stereoscopic photography technique. In the former, it was confirmed that a molten pool had wavy surface pattern until solidification, and the applicability of the stereoscopic photography technique for the in-situ observation during the laser irradiation was indicated in the latter.

Advanced high-temperature superconducting magnet for fusion reactors: Segment fabrication and joint technique

Superconducting magnets may potentially become larger and more complex for use in future demonstration and commercial fusion reactors. For such situations, the innovative design concept of remountable (or demountable) high-temperature superconducting (HTS) magnets has been proposed for both tokamak and helical reactors that uses the HTS material features of high thermal stability and low cryogenic power. In Japan, the derived concept of joint-winding of HTS coils has been proposed, in which the coils are wound by connecting short conductor segments. This paper first briefly summarizes the design proposal history and current research and development (R&D) status of the important technologies for joint design, and then reports recent R&D progress in mechanical lap joints with inserted indium foils. Based on the joint resistance as a function of temperature, the applied magnetic field and the numbers of layers and rows of HTS tapes in the conductor, the joint resistance of a 100-kA-class HTS conductor joint can be reduced to one third of its initial value by low-temperature heat treatment. Mechanical behavior of these joints indicates that stress relaxation of indium does not affect the fabrication time and that mechanical joints containing indium are preferable to soldered joints under high electromagnetic forces.

Pre-test analysis of protected loss of primary pump transients in CIRCE-HERO facility

In the frame of LEADER project (Lead-cooled European Advanced Demonstration Reactor), a new configuration of the steam generator for ALFRED (Advanced Lead Fast Reactor European Demonstrator) was proposed. The new concept is a super-heated steam generator, double wall bayonet tube type with leakage monitoring [1]. In order to support the new steam generator concept, in the framework of Horizon 2020 SESAME project (thermal hydraulics Simulations and Experiments for the Safety Assessment of MEtal cooled reactors), the ENEA CIRCE pool facility will be refurbished to host the HERO (Heavy liquid mEtal pRessurized water cOoled tubes) test section to investigate a bundle of seven full scale bayonet tubes in ALFRED-like thermal hydraulics conditions. The aim of this work is to verify thermo-fluid dynamic performance of HERO during the transition from nominal to natural circulation condition. The simulations have been performed with RELAP5-3D© by using the validated geometrical model of the previous CIRCE-ICE test section [2], in which the preceding heat exchanger has been replaced by the new bayonet bundle model. Several calculations have been carried out to identify thermal hydraulics performance in different steady state conditions. The previous calculations represent the starting points of transient tests aimed at investigating the operation in natural circulation. The transient tests consist of the protected loss of primary pump, obtained by reducing feed-water mass flow to simulate the activation of DHR (Decay Heat Removal) system, and of the loss of DHR function in hot conditions, where feed-water mass flow rate is absent. According to simulations, in nominal conditions, HERO bayonet bundle offers excellent thermal hydraulic behavior and, moreover, it allows the operation in natural circulation.

Nuclear analysis of structural damage and nuclear heating on enhanced K-DEMO divertor model

This paper addresses nuclear analysis on the Korean fusion demonstration reactor (K-DEMO) divertor to estimate the overall trend of nuclear heating values and displacement damages. The K-DEMO divertor model was created and converted by the CAD (Pro-Engineer™) and Monte Carlo automatic modeling programs as a 22.5° sector of the tokamak. The Monte Carlo neutron photon transport and ADVANTG codes were used in this calculation with the FENDL-2.1 nuclear data library. The calculation results indicate that the highest values appeared on the upper outboard target (OT) area, which means the OT is exposed to the highest radiation conditions among the three plasma-facing parts (inboard, central and outboard) in the divertor. Especially, much lower nuclear heating values and displacement damages are indicated on the lower part of the OT area than others. These are important results contributing to thermal-hydraulic and thermo-mechanical analyses on the divertor and also it is expected that the copper alloy materials may be partially used as a heat sink only at the lower part of the OT instead of the reduced activation ferritic-martensitic steel due to copper alloy’s high thermal conductivity.

Development of a systematic, self-consistent algorithm for the K-DEMO steady-state operation scenario

An optimum plasma pressure/current density profile and corresponding heating/current drive (H/CD) determination scheme is newly developed by integrating equilibrium, stability, confinement, and H/CD, self-consistently subject to maximize the fusion gain for Korean fusion demonstration reactor (K-DEMO) steady-state operation scenarios. The integrated plasma modeling package, FASTRAN/IPS, is adopted for the integrated numerical apparatus. The target pressure profile with a pedestal structure is investigated by varying its peaking, pedestal height and width as a first step. Formation of stable equilibria is evaluated by solving the Grad–Shafranov equation and checking linear MHD stability. For the case of potentially stable equilibrium, required external heating distribution is calculated by considering both power balance and external current drive alignment to reproduce the pressure profile of the stable equilibrium. Electron/ion temperature and poloidal flux evolutions are solved with the derived heating configuration to find a steady-state scenario and achieve self-consistent plasma profiles. A self-consistent target steady-state pressure and current profile parameters are proposed through designed systematic algorithm with fusion power PF = 2070 MW, fusion gain Q = 19.7, and normalized beta βN = 2.8 at toroidal field BT = 7.4 T and plasma current IP = 15.5 MA. Feasibility of fusion power PF = 3000 MW operation is also explored with enhanced density and temperature limit assumption.

Development of determination method of 93Mo content in metal waste generated at the Japan Power Demonstration Reactor

A separation method of Mo from Nb, Zr, and the matrix elements of rubble waste was modified to determine the content of 93Mo in metal waste. A separation scheme to treat 1 g of metal waste was established by optimizing the amount of ascorbic acid, the rinsing solution, and repeating of the procedure. A thin-layer source was prepared using direct drop deposition and evaporation to measure 93Mo content. Finally, 93Mo content in the metal waste generated at the Japan Power Demonstration Reactor was analyzed using the developed method.

Thermohydraulic Assessment for the Modified Concept of the K-DEMO Divertor Target

A pressurized water cooling divertor target applying the tungsten monoblock type has been primarily considered in the Korean fusion demonstration reactor (K-DEMO). The target peak heat flux locally concentrated around the striking point was set to 10 MW/m2 in K-DEMO divertor system. In a previous study [Im et al., IEEE Trans. Plasma Sci., Vol. 44, p. 2493 (2016)] the thermomechanical analyses for a high heat flux unit of K-DEMO divertor target applying reduced activation ferritic martensitic (RAFM) steel as heat sink material were carried out to verify the thermal and mechanical stabilities. The results of the thermomechanical analyses showed that the stabilities of the divertor target design applying the derived design parameters were close to the allowable limits, since the thickness of RAFM coolant tube was too thin due to the low thermal conductivity of RAFM steel. The aim of this study is to propose the structurally modified divertor concept switching the flowing path of coolant from poloidal direction to toroidal direction. By changing the flow direction, the design and material could be independently selected by the local intensity of the heat flux. The CuCrZr and RAFM steel were employed to the peak heat flux region and the non-peak heat flux region as a heat sink material, respectively. The effects of the modified concept were assessed by performing thermohydraulic analyses. The result showed that the modified concept more efficiently dissipated the heat flux compared to the conventional concept.

Hydrogen Isotope Retention and Permeation in Neutron-Irradiated Tungsten and Tungsten Alloys Under PHENIX Collaboration

Irradiation effects on heat-load and heat removal, thermo-mechanical properties, and tritium behavior in neutron-irradiated tungsten and tungsten alloy are being investigated under US-Japan PHENIX (Plasma facing components evaluation by tritium Plasma, HEat and Neutron Irradiation eXperiments) collaboration (2013–2018) to demonstrate feasibility and safety of helium-cooled divertor concept for a fusion demonstration (DEMO) and future fusion reactors. The PHENIX Task 3 is aimed at improved understanding of irradiation response on tritium retention and permeation in tungsten and tungsten alloys under divertor-relevant high-flux plasma for a Fusion Nuclear Science Facility (FNSF) and DEMO. This paper describes the challenge in elucidating tritium behavior in neutron-irradiated plasma facing components (PFCs), the PHENIX plans for neutron-irradiation and post irradiation examination, and progress in tritium behavior in neutron-irradiated tungsten.

Application of Prompt Self-Powered Neutron Detectors to the Lead-Cooled Fast Reactor Demonstrator ALFRED: Validation of the Monte Carlo Model for Selected SPNDs

The advanced lead fast reactor European demonstrator (ALFRED) is a European research initiative into the framework of the Generation IV International Forum facilities. ALFRED is a scaled down reactor compared to the industrial prototype European lead fast reactor proposed in lead-cooled European advanced demonstration reactor. It has a relatively low power (125 MWe) with a compact design to reduce the cost but maintaining its representativeness and it is cooled by pure lead. One of the open issues is linked to the neutron flux in-core monitoring system because of the harshness of the environment the detectors should be installed in, due to high temperatures, and the neutron-gamma radiation field levels. Monte Carlo simulation is a possible way of facing the problem, reproducing into a virtual world the reactor core, the surrounding environment and radiation interactions. In previous works, neutron spectra and gamma doses at possible detectors' locations in ALFRED were retrieved, with consideration on the applicability of each suitable device currently available. Fission chambers (FCs) were found to be exploited at reactor start-up and intermediate power range. Prompt self-powered neutron detectors (SPNDs) seemed to be the best solution to monitor the reactor full power, becoming the main research target: their effective applicability on field has to be demonstrated. SPND applications do not include reactor control purposes usually. Moreover, their irradiation experience involved thermal and epithermal neutron spectra monitoring, mainly. The lack of data when SPNDs sense fast neutron fluxes in terms of prompt-response pushed the authors to deepen the study in such direction. The work herein shows the mathematical approach based on Monte Carlo simulation of SPNDs by the Monte Carlo N-particle eXtended code (MCNPX), so as to study the capability of the code in reproducing real devices' signals while experimented on field. Such a verification turned out to be the preliminary stage for studying new concepts for SPNDs, in terms of sensitive materials and geometries, envisaging the possibility for designing, prototyping, and testing new devices in suitable fast neutron-flux facilities.

Investigation of wetting property between liquid lead lithium alloy and several structural materials for Chinese DEMO reactor

The dual-cooled lead lithium (PbLi) blanket is considered as one of the main options for the Chinese demonstration reactor (DEMO). Liquid PbLi alloy is used as the breeder material and coolant. Reduced activation ferritic/martensitic (RAFM) steel, stainless steel and the silicon carbide ceramic matrix composite (SiCf) are selected as the substrate materials for different use. To investigate the wetting property and inter-facial interactions of PbLi/RAFM steel, PbLi/SS316L, PbLi/SiC and PbLi/SiCf couples, in this paper, the special vacuum experimental device is built, and the ‘dispensed droplet’ modification for the classic sessile droplet technique is made. Contact angles are measured between the liquid PbLi and the various candidate materials at blanket working temperature from 260 to 480 °C. X-ray photoelectron spectroscopy (XPS) is used to characterize the surface components of PbLi droplets and substrate materials, in order to study the element trans-port and corrosion mechanism. Results show that SiC composite (SiCf) and SiC ceramic show poor wetting properties with the liquid PbLi alloy. Surface roughness and testing temperature only provide tiny improvements on the wetting property below 480 °C. RAFM steel performs better wetting properties and corrosion residence when contacted with molten PbLi, while SS316L shows low corrosion residence above 420 °C for the decomposition of protective surface film mainly consisted of chromic sesquioxide. The results could provide meaningful compatibility database of liquid PbLi alloy and valuable reference in engineering design of candidate structural and functional materials for future fusion blanket.

Status and improvement of CLAM for nuclear application

A program for China low activation martensitic steel (CLAM) development has been underway since 2001 to satisfy the material requirements of the test blanket module (TBM) for ITER, China fusion engineering test reactor and China fusion demonstration reactor. It has been undertaken by the Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences under wide domestic and international collaborations. Extensive work and efforts are being devoted to the R&D of CLAM, such as mechanical property evaluation before and after neutron irradiation, fabrication of scaled TBM by welding and additive manufacturing, improvement of its irradiation resistance as well as high temperature properties by precipitate strengthening to achieve its final successful application in fusion systems. The status and improvement of CLAM are introduced in this paper.

A Summary of the Department of Energy’s Advanced Demonstration and Test Reactor Options Study

An assessment of advanced reactor technology options was conducted to provide a sound comparative technical context for future decisions by the U.S. Department of Energy (DOE) concerning these technologies. Strategic objectives were established that span a wide variety of important missions, and advanced reactor technology needs were identified based on recent DOE and international studies. A broad team of stakeholders from industry, academia, and government was assembled to develop a comprehensive set of goals, criteria, and metrics to evaluate advanced irradiation test and demonstration reactor concepts. Point designs of a select number of concepts were commissioned to provide a deeper technical basis for evaluation. The technology options were compared on the bases of technical readiness and the ability to meet the different strategic objectives. Using the study’s evaluation criteria and metrics, an independent group of experts from industry, universities, and national laboratories scored each of the point designs. Pathways to deployment for concepts of varying technical maturities were estimated for the different demonstration systems with regard to cost, schedule, and possible licensing approaches. This study also presents the trade-offs that exist among the different irradiation test reactor options in terms of the ability to conduct irradiations in support of advanced reactor research and development and to serve potential secondary missions.The main findings of the study indicate the following: (1) for industrial process heat supply, a high-temperature gas-cooled reactor is the best choice because of the high outlet temperature of the reactor and its strong passive and inherent safety characteristics; (2) for resource utilization and waste management, a sodium-cooled fast reactor (SFR) is best because of the use of a fast flux to destroy actinides; (3) to realize the advantages of a promising but less-mature technology, a fluoride salt-cooled high-temperature reactor and a lead-cooled fast reactor fare about the same; (4) for fulfilling the needs of a materials test reactor, a SFR is considered best because of its ability to produce high fast flux, incorporate test loops, and provide additional large volumes for testing.

Analysis of hydrogen and dust explosion after vacuum vessel rupture: Preliminary safety analysis of Korean fusion demonstration reactor using MELCOR

Hydrogen explosion is one of the most dangerous accident phenomena in both nuclear fusion and fission reactor. In the typical fusion reactor, tritium and deuterium are fuel elements for fusion reaction, and they exist in condensed state around the cryo-pump in the normal operation condition. In the accident situation, increasing heat transfer system temperature can mobilize lots of tritium in the vacuum vessel. And then, coolant can react with high-temperature structures generating hydrogen gas inside vacuum vessel. In addition, air ingress into plasma terminates fusion reaction producing activated dust on the first wall plasma facing surfaces. Through these series of process, mobilized tritium and dust aerosol can explode in the vacuum vessel and damage the containment building integrity, making a sort of source term leakage to environment through various system volumes. In this paper, preliminary safety analysis for hydrogen and dust explosion accident for assumed design of Korean fusion demonstration reactor is conducted. An MELCOR system analysis code is used to simulate this accident and parametric analysis to investigate effect on the accident scale and safety systems like detritiation system such as heating, ventilation, and air conditioning isolation system and pressure suppression system. As a result, aerosol distribution of each control volume and its release route are estimated, and final aerosol release to the environment is compared with release guideline for ITER. The amount of mobilized aerosol in demonstration reactor is evaluated conservatively using ITER normal operation condition.