Divertor Cryopump Research Articles

Design and development of the EAST plug-in cryopump for long-pulse high-performance operation

At present, the in-vessel cryopump of the EAST divertor demonstrates inadequate pumping performance during long-pulse high confinement mode (H-mode) plasma operations, and the disassembly and maintenance of this cryopump pose significant challenges. In response to these issues and considering the constraints of the installation space of the cryopump, an EAST plug-in cryopump have been developed based on cryoadsorption and cryocondensation principles. The plan is to install four plug-in cryopumps at different vertical ports of the EAST vacuum vessel. These cryopumps will operate in a staggered manner with a specific time interval to achieve a quasi-continuous pumping system. This paper provides a systematic discussion of the design, simulation analysis, and experimental studies of the plug-in cryopump to validate its feasibility for the EAST application. In conclusion, each individual cryopump exhibits an average pumping speed of approximately 9255 L/s for hydrogen, 7001 L/s for deuterium, and 5290 L/s for helium. The pumping capacity and regeneration time also meet the requirements of EAST experiments. Our ultimate goal is to achieve the long-pulse high-performance plasma operation in EAST with the support of the plug-in cryopumps, while also laying the foundation for the development of divertor cryopumps for future fusion reactors.

Preparing the operation of Wendelstein 7-X in the steady-state regime

The Wendelstein 7-X stellarator (W7-X) started operation in December 2015 and has performed three experimental campaigns up to November 2018. Since then the device has been completed to the planned configuration (most importantly with predominantly actively cooled first wall, including the High Heat Flux divertor and a divertor cryo pump) to be able to perform stationary plasma programs. After this completion pFhase, at the end of 2021, the commissioning has been performed and scientific operation has started in the fall of 2022.

Optimization of the snowflake divertor for power and particle exhaust on NSTX–U

Abstract In this paper, simple analytical modeling and numerical simulations performed with the multi-fluid edge transport code UEDGE are used to identify optimal snowflake divertor (SFD) configurations for heat flux mitigation and sufficient cryopumping performance on the National Spherical Torus eXperiment Upgrade (NSTX–U). A model is presented that describes the partitioning of sheath-limited SOL power and particle exhaust in the SFD as a result of diffusive transport to multiple activated strike points. The model is validated against UEDGE predictions and used to analyze a database of 70 SFD-minus equilibria. The optimal location for the entrance to a divertor cryopumping system on NSTX–U is computed for enabling sufficient pumping performance with acceptable power loading in a variety of SFD-minus configurations. UEDGE simulations of one promising equilibrium from the database indicate that a significant redistribution of power to the divertor legs occurs as a result of neutral particle removal near one of the SFD-minus strike points in the outboard scrape-off layer. It is concluded that pump placement at the optimal location is advantageous as the large number of compatible equilibria reduces the precision required of real-time SFD configuration control systems and enables acceptable divertor solutions even if UEDGE-predicted power redistribution slightly reduces the achievable pumping performance.

Summary of thermal analyses to determine the refrigeration power for the JT-60SA helium refrigerator

JT-60SA is an upgrade of JT-60U using exclusively superconducting coils to confine the plasma with a typical flat top duration up to 100 s in purely inductive mode. In order to define the helium refrigerator requirements for the superconducting coil system, their thermal protection and the cryopumps, heat loads at different temperature levels have been defined. Because of plasma pulse operation, transient heat pulses appear at the 4 K level for the superconducting coils and the cold structures as well as at the divertor cryopumps of JT-60SA.Since the heat loads were calculated at an earlier stage of the project, the loads of each component design had a certain amount of uncertainty. In JT-60SA, an attempt was made to define reasonable contingencies for the heat load by applying Monte Carlo simulation and put the 99% confidence value of the heat load as requirement for the refrigerator. Exergy analysis of heat loads at different temperatures and the helium mass flows of the JT-60SA resulted in an equivalent refrigeration power of about 9 kW at 4.5 K. This paper summarizes all the thermal analyses performed to determine the refrigeration power of the JT-60SA helium refrigerator.

Monitoring and control of the magnet system of JT-60SA

The construction of the full-superconducting tokamak JT-60 Super Advanced (JT-60SA) is in progress under the JA-EU broader approach agreement. The Cryogenic System of JT-60SA has five circuits to cool the Poloidal field coils, Toroidal field coils, thermal shields, high temperature superconducting current leads (HTS-CLs) and feeders, and divertor cryopumps. The physical conditions such as pressures, temperatures and mass flows of the magnet cooling loops are monitored by a Magnet Controller The magnet controller also performs local control of the helium flows. A particular abnormal situation is a quench or a fast discharge of a magnet which would require many hours for recovery to normal operation. In order to avoid fast discharges as far as possible, the MC uses “safety interlocks” to supervise the magnet coils and the HTS CL. This safety interlocks are called slow discharge. In this paper, the slow discharge thresholds of the HTS-CLs are derived from simulations of the cryogenic loops using the thermal analysis programs FLOWER, HEATER, and SUPERMAGNET.

Assessment of the JT-60SA divertor cryopump performance

The cryogenic pumping system is one of the key components that define the operational window of JT-60SA. It is envisaged to utilize condensation pumps that work at 3.8 K, and proper thermal shields cooled with water and 80 K gaseous helium are necessary to allow having them installed at short distance to the torus in order to exploit high pumping speeds. In this paper, one complete divertor cryopump unit in a 40° sector is studied by Monte Carlo simulation. The pumping speeds of one cryopump unit for D2 were found to be between 14.8 m3/s and 18.1 m3/s depending on the assumptions taken. In addition, the heat loads to the cryopump condensation pipes during steady state nominal operation, as well as the heat loads and transient behavior after massive gas injection events are assessed. The present design can fulfil the requirement of pumping, cooling and regeneration.

Performance of the JT-60SA cryogenic system under pulsed heat loads during acceptance tests

The JT-60SA cryogenic system a superconducting tokamak currently under assembly at Naka, Japan. After one year of commissioning, the acceptance tests were successfully completed in October 2016 in close collaboration with Air Liquide Advanced Technologies (ALaT), the French atomic and alternative energies commission (CEA), Fusion for Energy (F4E) and the Quantum Radiological Science and Technology (QST). The cryogenic system has several cryogenic users at various temperatures: the superconducting magnets at 4.4 K, the current leads at 50 K, the thermal shields at 80 K and the divertor cryo-pumps at 3.7 K. The cryogenic system has an equivalent refrigeration power of about 9.5 kW at 4.5 K, with peak loads caused by the nuclear heating, the eddy currents in the structures and the AC losses in the magnets during cyclic plasma operation. The main results of the acceptance tests will be reported, with emphasis on the management of the challenging pulsed load operation using a liquid helium volume of 7 m3 as a thermal damper.

Fuel recycling and natural density in EAST H-mode discharges

The natural density (plasma density without gas puffing) of an ELMy H-mode discharge is strongly affected by fuel recycling and plasma parameters. In the EAST 2012 campaign fuel recycling and natural density were investigated in H-mode discharges. The results show that natural density is decreased gradually with high frequency ELMs, and increased with low frequency ELMs or in ELM-free phase. Plasma density is increased gradually when H-mode discharge is in the phase of alternating ELM and ELM-free. A statistical investigation and scaling of natural density with various plasma normalized parameters shows the natural density is affected by plasma current, heating power, magnetic configuration, first wall temperature, and divertor cryopump status. The influence of plasma heating power, total number of injected particles, and accumulated fuel retention is very weak, probably due to the intensive lithium conditionings. The results indicate that the natural density could be partially controlled by the combination of different parameters.

Multi‐Fluid Transport Modeling of NSTX Upgrade Standard and Snowflake Divertor Configurations

AbstractIn NSTX Upgrade (NSTX‐U), power exhaust is expected to challenge available material and heat removal technologies, which are limited to ∼10 MW/m2. Preliminary analysis of heat flux mitigation strategies, including the snowflake divertor configuration and radiative divertor operation, is performed with the multi‐fluid edge transport code, UEDGE. Divertor recycling between 95% and 99%, and power exhaust between 7 and 9 MW are explored. Compared to the standard divertor, the particular snowflake divertor studied here strongly reduces outer target heat flux, but tends to have higher (>10 MW/m2) inner target heat flux. This suggests that snowflake divertor optimization should be pursued in future work. Both neon and argon impurity seeding scenarios mitigate high heat fluxes in both standard and snowflake configurations, but argon can cause radiative collapse of the core plasma at only 3% concentration. Divertor cryopumping is also considered, and modeling shows ∼50% more particle removal with the snowflake divertor. (© 2014 WILEY‐VCH Verlag GmbH & Co. KGaA, Weinheim)

Deposition of 13C tracer in the JET MkII-HD divertor

Migration of 13C has been investigated at JET by injecting 13C-labelled methane at the outer divertor base at the end of the 2009 campaign. The 13C deposition profiles on carbon fibre composite divertor tiles were measured by secondary ion mass spectrometry and Rutherford backscattering techniques. 13C was mainly deposited near the puffing location on the outer divertor base tiles. High amounts of 13C were also found at the outer vertical target: at the bottom of the lower and at the top of the upper plates. Thirty-three percent of puffed 13CH4 was instantly pumped out by the divertor cryopump, which is close to the pump duct entrance. Global 13C transport in the torus was modelled by the EDGE2D/EIRENE and DIVIMP codes, and local 13C migration in the vicinity of the injection location by the ERO code. The DIVIMP and EDGE2D simulations show strong prompt deposition of 13C directly adjacent to the injection point as well as in the far scrape-off layer (SOL) along both the inner and outer divertor targets. In addition, the measured 13C deposition along the outer divertor wall tiles is qualitatively reproduced. However, EDGE2D and DIVIMP do not predict any deposition along the divertor surfaces facing the private plasma on the inner floor tile and inboard of the outer strike point on tile 5. The ERO calculations also indicate that most of the deposition occurs close to the injection location on the vertical face of the LBSRP tile and the horizontal part of tile 6.

Global Evaluation of Prompt Dose Rates in ITER Using Hybrid Monte Carlo/Deterministic Techniques

The hybrid Monte Carlo (MC)/deterministic techniques - Consistent Adjoint Driven Importance Sampling (CADIS) and Forward Weighted CADIS (FW-CADIS) - enable the full 3-D modeling of very large and complicated geometries. The ability of performing global MC calculations for nuclear parameters throughout the entire ITER reactor was demonstrated. The 2 m biological shield (bioshield) reduces the total prompt operational dose by six orders of magnitude. The divertor cryo-pump port results in a peaking factor of 120 in the prompt operational dose rate behind the bioshield of ITER. The equatorial port, plugged by 2 m of shielding, increases the prompt dose rate behind the bioshield by a factor of 47. The peak values of the prompt dose rates at the back surface of the bioshield were 240 μSv/hr and 94 μSv/hr corresponding to the regions behind the divertor cryo-pump port and the equatorial port, respectively.

I-mode: an H-mode energy confinement regime with L-mode particle transport in Alcator C-Mod

An improved energy confinement regime, I-mode, is studied in Alcator C-Mod, a compact high-field divertor tokamak using ion cyclotron range of frequencies (ICRFs) auxiliary heating. I-mode features an edge energy transport barrier without an accompanying particle barrier, leading to several performance benefits. H-mode energy confinement is obtained without core impurity accumulation, resulting in reduced impurity radiation with a high-Z metal wall and ICRF heating. I-mode has a stationary temperature pedestal with edge localized modes typically absent, while plasma density is controlled using divertor cryopumping. I-mode is a confinement regime that appears distinct from both L-mode and H-mode, combining the most favourable elements of both. The I-mode regime is investigated predominately with ion ∇B drift away from the active X-point. The transition from L-mode to I-mode is primarily identified by the formation of a high temperature edge pedestal, while the edge density profile remains nearly identical to L-mode. Laser blowoff injection shows that I-mode core impurity confinement times are nearly identical with those in L-mode, despite the enhanced energy confinement. In addition, a weakly coherent edge MHD mode is apparent at high frequency ∼100–300 kHz which appears to increase particle transport in the edge. The I-mode regime has been obtained over a wide parameter space (B T = 3–6 T, I p = 0.7–1.3 MA, q 95 = 2.5–5). In general, the I-mode exhibits the strongest edge temperature pedestal (T ped) and normalized energy confinement (H 98 > 1) at low q 95 (<3.5) and high heating power (P heat > 4 MW). I-mode significantly expands the operational space of edge localized mode (ELM)-free, stationary pedestals in C-Mod to T ped ∼ 1 keV and low collisionality , as compared with EDA H-mode with T ped < 0.6 keV, . The I-mode global energy confinement has a relatively weak degradation with heating power; leading to increasing H 98 with heating power.

EAST in-vessel components design

Divertor and other in-vessel components are very important parts of EAST superconducting Tokamak. The primary purpose of these components is to protect the vacuum vessel, RF system and diagnostic components from plasma particles and heat loads. Other function of in-vessel components is additional to particles and heat loads handling. The divertor is designed to provide particles exhaust into the divertor cryopump, provide recycling control and impurity control. Passive plates stabilize plasma vertical instability. In-vessel coils are used for plasma instability control, error field correct, RWM and ELMs suppress. Heat loads and electric–magnetic forces are quite complicate for in-vessel components. To make the design of in-vessel components, we consider their proper geometry, structure strength, electro–magnetic characteristic, thermal conductive characteristic, particles exhaust characteristic and so on.

Fuel Retention over a Full Day of Experiments in JET

A series of 39 consecutive and repetitive discharges (Ip = 2MA, BT = 2.4T, <ne> = 3.8 × 1019m-3, gas rate ~1.5 × 1022 Ds-1 and with 2.8 MW of ICRH over a duration of 11s) has been performed in JET for a full day in order to study the particle retention behaviour as a function of the wall inventory and the global balance for a significant number of discharges associated to a high gas injection. Since the active pumping was achieved using the divertor cryopump only, its regeneration has allowed a direct calibration of the value of the pumped particle flux to be used in the particle balance analysis during the plasma operations for the “DOC-L” configuration. Taking into account the outgased flux between the discharges, the resulting wall inventory over the full day of operation is zero. During, the 11 sec of the ICRH power, about 8 % of the particles injected are retained in the machine equilibrated by a particle recovery between of 8% of the quantity injected. This shows that the gas released between pulses has been overestimated in previous JET gas balance analysis and that the particles trapped in the machine are localised in areas which are outgasing between the discharges.

Helium exhaust experiments on JET with Type I ELMs in H-mode and with Type III ELMs in ITB discharges

An analysis of helium exhaust experiments on JET in the MkII-GB divertor configuration is presented. Helium is pumped by applying an argon frost layer on the divertor cryo pump. Measurement of the helium retention time, , is performed in two ways: by the introduction of helium in gas puffs and measurement of the subsequent decay time constant of the helium content, ; and by helium beam injection and measurement of the helium replacement time, . In ELMy H-mode, with plasma configuration optimized for pumping, is achieved, where is the thermal energy replacement time. For quasi-steady internal transport barrier (ITB) discharges, the achieved is significantly lower. The achieved helium recycling coefficient, confirmed by an independent measurement to be Reff ≈ 0.91, is the same in both scenarios. None of the discharges are dominated by core confinement. The difference in is instead due to the confinement properties of the edge plasma, which is characterized by Type I ELMs for the H-mode discharges studied, and Type III ELMs for the quasi-steady ITB discharges. This difference is quantified by an independent measurement of the ratio of the helium replacement time with a helium edge source to the energy confinement time.

Advanced tokamak research in DIII-D

Advanced tokamak (AT) research in DIII-D seeks to provide a scientific basis for steady-state high performance operation in future devices. These regimes require high toroidal beta to maximize fusion output and high poloidal beta to maximize the self-driven bootstrap current. Achieving these conditions requires integrated, simultaneous control of the current and pressure profiles and active magnetohydrodynamic stability control. The building blocks for AT operation are in hand. Resistive wall mode stabilization by plasma rotation and active feedback with non-axisymmetric coils allows routine operation above the no-wall beta limit. Neoclassical tearing modes are stabilized by active feedback control of localized electron cyclotron current drive (ECCD). Plasma shaping and profile control provide further improvements. Under these conditions, bootstrap supplies most of the current. Steady-state operation requires replacing the remaining inductively driven current, mostly located near the half radius, with non-inductive external sources. In DIII-D this current is provided by ECCD, and nearly stationary AT discharges have been sustained with little remaining inductive current. Fast wave current drive is being developed to control the central magnetic shear. Density control, with divertor cryopumps, of AT discharges with ELMing H-mode edges facilitates high current drive efficiency at reactor relevant collisionalities. An advanced plasma control system allows integrated control of these elements. Close coupling between modelling and experiment is key to understanding the separate elements, their complex nonlinear interactions, and their integration into self-consistent high performance scenarios. This approach has resulted in fully non-inductively driven plasmas with βN ≤ 3.5 and βT ≤ 3.6% sustained for up to 1 s, which is approximately equal to one current relaxation time. Progress in this area, and its implications for next-step devices, will be illustrated by results of these and other recent experiment and simulation efforts.

High performance advanced tokamak regimes in DIII-D for next-step experiments

Advanced Tokamak (AT) research in DIII-D [K. H. Burrell for the DIII-D Team, in Proceedings of the 19th Fusion Energy Conference, Lyon, France, 2002 (International Atomic Energy Agency, Vienna, 2002) published on CD-ROM] seeks to provide a scientific basis for steady-state high performance operation in future devices. These regimes require high toroidal beta to maximize fusion output and poloidal beta to maximize the self-driven bootstrap current. Achieving these conditions requires integrated, simultaneous control of the current and pressure profiles, and active magnetohydrodynamic stability control. The building blocks for AT operation are in hand. Resistive wall mode stabilization via plasma rotation and active feedback with nonaxisymmetric coils allows routine operation above the no-wall beta limit. Neoclassical tearing modes are stabilized by active feedback control of localized electron cyclotron current drive (ECCD). Plasma shaping and profile control provide further improvements. Under these conditions, bootstrap supplies most of the current. Steady-state operation requires replacing the remaining Ohmic current, mostly located near the half radius, with noninductive external sources. In DIII-D this current is provided by ECCD, and nearly stationary AT discharges have been sustained with little remaining Ohmic current. Fast wave current drive is being developed to control the central magnetic shear. Density control, with divertor cryopumps, of AT discharges with edge localized moding H-mode edges facilitates high current drive efficiency at reactor relevant collisionalities. A sophisticated plasma control system allows integrated control of these elements. Close coupling between modeling and experiment is key to understanding the separate elements, their complex nonlinear interactions, and their integration into self-consistent high performance scenarios. Progress on this development, and its implications for next-step devices, will be illustrated by results of recent experiment and simulation efforts.

Cryotechnology for Wendelstein 7-X

For the underlying task to cool the Wendelstein 7-X (W7-X) superconducting coil conductors and the divertor cryo-pumps, an extended cryo-system is required. Besides the magnet coils and cryo-pumps proper, the coil housings and supports, the thermal shields of the cryostat and cryo-pumps, and the current leads (CL) have to be cooled. This is achieved by a refrigeration plant including coolant transfer lines and distribution boxes, and a widespread pipework within the cryostat. All these components are finally defined, and purchase actions are under way.

Performance of high triangularity plasmas as the volume of the secondary divertor is varied in DIII-D

This paper examines the sensitivity of edge pedestal and divertor performance parameters to changes in the volume of a secondary divertor that tends to form inside the vacuum chamber in high triangularity plasma shapes. The sensitivity was examined by varying the vertical distance of the secondary X-point from the target plate, Z S X , while holding the primary X-point height fixed. The ion ∇ B drift was in the direction of the primary divertor. Discharges with and without active primary divertor cryo-pumping were examined. The effective rate of rise of the core density at the L–H transition increased 80% as Z S X was reduced. At high density achieved by gas injection, the core line-averaged density at the H–L back transition decreased by 30% as Z S X was reduced. Both of these results indicate that performance may be affected when core plasma screening of neutrals in the secondary divertor is reduced as Z S X decreases. The peak heat flux in the secondary divertor decreased as Z S X was reduced due to increased flux expansion until a threshold was reached at Z S X ∼3 cm where it remained constant. This indicated that the high recycling character of the plasma above the target in the secondary divertor was lost when the length of the outer divertor leg was reduced below a critical value.

Radiative divertor and scrape-off layer experiments in open and baffled divertors on DIII-D

Recent progress towards an increased understanding of the physical processes in the divertor and scrape-off layer (SOL) plasmas in DIII-D has been made possible by a combination of new diagnostics, improved computational models and changes in divertor geometry. The work focused primarily on ELMing H mode discharges. The physics of partially detached divertor plasmas, with divertor heat flux reduction by divertor radiation enhancement using D2 puffing, was studied intwo dimensions, and a model of the heat and particle transport was developed that includes conduction, convection, ionization, recombination and flows. Plasma and impurity particle flows were measured with Mach probes and spectroscopy and compared with the UEDGE model. The model now includes self-consistent calculations of carbon impurities. Impurity radiation was increased in the divertor and SOL with `puff and pump' techniques using SOL D2 puffing, divertor cryopumping and argon puffing. The important physical processes in plasma-wall interactions were examined with a DiMES (divertor material evaluation system) probe, plasma characterization near the divertor plate and the REDEP code. Experiments comparing single null plasma operation in baffled and open divertors demonstrated a change in the edge plasma profiles. These results are consistent with a reduction in the core ionization source calculated with UEDGE. Divertor particle control in ELMing H mode with pumping and baffling resulted in a reduction in H mode core densities to ne/nGr ≈ 0.25 (with nGr the Greenwald density). Divertor particle exhaust and heat flux were studied as the plasma shape was varied from a lower single null to a balanced double null, and finally to an upper single null.