In-vessel Control Research Articles

In-vessel dust and tritium control strategy in ITER

Abstract A baseline strategy for dust and tritium-inventory control and recovery in ITER has been established and preparations are underway for its implementation. Limits on dust and tritium-inventory are an integral part of the ITER safety case and are fixed at 1 kg for tritium, 1000 kg for mobilisable dust and 11 kg (beryllium)/76 kg (tungsten) for dust on hot surfaces. Maximum average T-retention rates of ∼1 g/shot are estimated for baseline inductive operation at QDT = 10, suggesting that the in-vessel T-retention could reach the administrative limit of 640 g in as little as ∼2 months of operation. Baking is expected to remove a significant fraction of the T co-deposited on the divertor targets. Despite large uncertainties, dust quantities are expected to remain well below safety limits over the divertor cassette lifetime. In situ aspiration during divertor cassette exchange is foreseen as the main dust removal technique.

ELM control experiments in the KSTAR device

The fourth KSTAR campaign in 2011 concentrated on active edge-localized mode (ELM) control by various methods such as non-axisymmetric magnetic perturbations, supersonic molecular beam injection (SMBI), vertical jogs of the plasma column and edge electron heating. The segmented in-vessel control coil (IVCC) system is capable of applying n ⩽ 2 perturbed field with different phasing among top, middle and bottom coils. Application of an n = 1 perturbed field showed a desirable ELM suppression result. Fast vertical jogs of the plasma column achieved ELM pace-making and ELMs locked to 50 Hz vertical jogs were observed with a high probability of phase locking. A newly installed SMBI system was used for ELM control and the state of mitigated ELMs was sustained by the optimized repetitive SMBI pulse for a few tens of ELM periods. A change in ELM behaviour was seen due to edge electron heating although the effect of ECH launch needs supplementary analyses. The ECEI images of suppressed/mitigated ELM states showed apparent differences when compared with natural ELMy states. Further analyses are ongoing to explain the observed ELM control results.

Analysis of edge magnetic field line structure in ITER due to in-vessel ELM control coils

In this work we evaluated the ITER ELM coils design based on two metrics: the Chirikov vacuum magnetic island overlap parameter, and the vacuum Field Line Loss Fraction. The study was performed for a range of current amplitudes for three different n=4 waveforms: square, cosine and sine. The results indicated that ITER ELM coils are designed with a high level of flexibility to accommodate different operation scenarios (H-mode and Steady State) with different values of q95 and q-profiles. The magnetic island overlap analysis showed that ITER ELM coils are capable of matching the DIII-D I-coil spectrum. The Field Line Loss analysis showed that edge vacuum stochastization might be achieved that is similar or greater than in DIII-D. Fault analysis of the coils indicated that ITER ELM coils are robust and show good characteristics even with 11% of dead coils.

Key Features in the Operation of KSTAR

The Korea Superconducting Tokamak Advanced Research (KSTAR) device is aimed at advanced tokamak (AT) research. Three years have passed since it achieved its first plasma in 2008. Because it is a superconducting machine and is working toward AT research, it has unique features in terms of the machine engineering and operation. The toroidal field (TF) magnet coils are made of Nb <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">3</sub> Sn, which provide high TFs up to 3.5 T, and have been fully tested. The poloidal field (PF) magnet coils, consisting of both Nb <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">3</sub> Sn and NbTi, which have a maximum current of 25 kA in their design, were tested up to 15 kA. A thermal hydraulic analysis is being conducted for PF magnet coil operation. All plasma-facing components (PFCs) are equipped with water cooled graphite tiles and have the capability of being baked up to 350°C. A startup scenario, which considered both the effect of the ferromagnetic material in the cable in conduit conductor jacket of the magnet coils as well as a nonferromagnetic up-down asymmetry in the cryostat structure, was developed and demonstrated its effectiveness by the last two year's reliable operations. Passive stabilizers and in-vessel control coils (IVCCs) are key components to realize AT operation in KSTAR. The segmented IVCC coils were connected to form circular coils for internal vertical control in 2010, and diverted plasmas with high elongation (κ~1.8, δ>;0.6) were achieved. A neutral beam injection (NBI) system was developed aiming at 2 MW, 300 s per ion source which meets the long-pulse requirement of KSTAR. An NBI ion source with a power of 1.7 MW at 100 kV has been commissioned for 10 s. Finally, ELMy H-modes were successfully produced with 1.3-MW NBI power at a plasma current of 0.6 MA in the 2010 campaign. The first H-mode discharge (#4200) in KSTAR was achieved one year earlier than officially planned and was done at B <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">T</sub> = 2 T with I <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">p</sub> = 0.6 MA in a well-balanced double null configuration after boronization on the PFC. Successful operations in the early days of KSTAR including H-mode experiments revealed the capability of advanced and steady-state operation which is essential for the international thermonuclear experimental reactor (ITER) and future fusion reactors.

SST-1 Magnet System Refurbishment: An Update

The Magnet System of the Steady State Superconducting Tokamak (SST-1) has been completely refurbished under the SST-1 Mission. Since Jan 2009, a wide spectrum of refurbishment has been undertaken which, includes developing reliable designs and processes leading to the fabrication of leak tight low DC resistances in SST-1 magnet winding packs, equipping each of the sixteen SST-1 Toroidal Field (TF) magnets with a supercritical helium cooled bubble type thermal shields and testing each of the prepared TF magnets under representative conditions in cold with nominal currents along with manifolds and isolators in near representative conditions. Each of the sixteen SST-1 TF magnets has been tested fully and successfully in a dedicated test stand in nineteen campaigns during June 10, 2010 and was concluded on Jan 24, 2011. These campaigns ensured that all the sixteen TF magnets could be charged to their nominal currents of 10000 A in either two-phase or supercritical cooling conditions with leak-tight inter-double pancake resistances being in the range of 150 pico ohms to 1200 pico ohms. The supercritical helium cooled thermal shields welded in the inner bore of the TF magnets have also performed as per the design specifications. Subsequently, the assemblies of the SST-1 TF magnets and the Poloidal Field (PF) magnets in SST-1 machine shell have begun. The SST-1 TF magnets are being assembled in pairs (known as octants) together with a SST-1 vacuum vessel module, sector of 80 K bubble type thermal shields and a pair of outer-inter-coil-structures. The octant assemblies have been completed. The nine superconducting Poloidal Field (PF) magnets will shortly be assembled being supported from the TF cases. The resistive central solenoid magnets, compensating coils and the newly designed in-vessel radial control coils will be subsequently assembled. The assembled magnets inside the cryostat are expected to be cooled down starting from Jan 2012 when a detailed engineering validation of the magnet system would be undertaken. The detailed SST-1 magnet system refurbishment including some of the first-of-its-kind test results are discussed in this paper

KSTAR equilibrium operating space and projected stabilization at high normalized beta

Along with an expanded evaluation of the equilibrium operating space of the Korea Superconducting Tokamak Advanced Research, KSTAR, experimental equilibria of the most recent plasma discharges were reconstructed using the EFIT code. In near-circular plasmas created in 2009, equilibria reached a stored energy of 54 kJ with a maximum plasma current of 0.34 MA. Highly shaped plasmas with near double-null configuration in 2010 achieved H-mode with clear edge localized mode (ELM) activity, and transiently reached a stored energy of up to 257 kJ, elongation of 1.96 and normalized beta of 1.3. The plasma current reached 0.7 MA. Projecting active and passive stabilization of global MHD instabilities for operation above the ideal no-wall beta limit using the designed control hardware was also considered. Kinetic modification of the ideal MHD n = 1 stability criterion was computed by the MISK code on KSTAR theoretical equilibria with a plasma current of 2 MA, internal inductance of 0.7 and normalized beta of 4.0 with simple density, temperature and rotation profiles. The steep edge pressure gradient of this equilibrium resulted in the need for significant plasma toroidal rotation to allow thermal particle kinetic resonances to stabilize the resistive wall mode (RWM). The impact of various materials and electrical connections of the passive stabilizing plates on RWM growth rates was analysed, and copper plates reduced the RWM passive growth rate by a factor of 15 compared with stainless steel plates at a normalized beta of 4.4. Computations of active RWM control using the VALEN code showed that the n = 1 mode can be stabilized at normalized beta near the ideal wall limit via control fields produced by the midplane in-vessel control coils (IVCCs) with as low as 0.83 kW control power using ideal control system assumptions. The ELM mitigation potential of the IVCC, examined by evaluating the vacuum island overlap created by resonant magnetic perturbations, was analysed using the TRIP3D code. Using a combination of all IVCCs with dominant n = 2 field and upper/lower coils in an even parity configuration, a Chirikov parameter near unity at normalized poloidal flux 0.83, an empirically determined condition for ELM mitigation in DIII-D, was generated in theoretical high-beta equilibria. Chirikov profile optimization was addressed in terms of coil parity and safety factor profile.

Development of KSTAR in-vessel components and heating systems

In-vessel components of the Korea Superconducting Tokamak Advanced Research (KSTAR) were developed for 2010 campaign to provide a crucial circumstance for achieving the strongly shaped and diverted plasma. Moreover, the in-vessel components such as limiter, divertor, passive stabilizer, in-vessel control coil (IVCC) system demonstrated good performances satisfying the original design concepts. In addition to the plasma facing components and the IVCC, in-vessel cryo-pump (IVCP) system was also installed to leverage divertor operation. Besides the in-vessel components, there have been substantial progresses in development of the heating and current drive system. The KSTAR heating and current drive system includes all kinds of the major heating systems such as neutral beam injection (NBI), ion cyclotron range of frequency (ICRF), electron cyclotron resonance heating and current drive (ECH and ECCD), lower hybrid current drive (LHCD) systems. As an initial stage for full equipment of the heating systems to total power of 26 MW, several key systems such as 1st NBI (called NBI-1), ICRF, and ECH-assisted startup system successfully demonstrated their excellent feasibilities in the design and performances for dedication to the 2010 campaign.

Fabrication and installation of KSTAR in-vessel control coils

The in-vessel control coil (IVCC) system, which has been designed for dedication of various active feedback plasma control functions, successfully fabricated and installed in the vacuum vessel of the Korea Superconducting Tokamak Advanced Research (KSTAR). The IVCC system consists of sixteen segmented coils that were independently fabricated outside the vacuum vessel and installed without any inside welding or brazing joints. The segmented coil system has several advantages such as eliminating possibility of cooling water leakage at the welded or brazed joints, simplification in fabrication and installation, and easy repair and maintenance of the coil system. Each segment contains eight oxygen-free high conductive coppers, which are grouped to four pairs, called as sections. Consequently, a segmented coil forms four sections for position control, field error correction (FEC), and resistive wall mode (RWM) control in accordance with electrical connection outside the cryostat. The eight conductors (or four sections) with internal coolant holes were enclosed in a rectangular welded jacket made of stainless steel 316LN and electrically insulated from the conductors by epoxy/glass composite layers. This coil system was commissioned up to 5 kA (30 kA-turns) for 5 s to achieve tentative use for the fast vertical plasma position control in the 2010 campaign of the KSTAR. This paper describes the several remarkable results in the fabrication and installation of the IVCC as well as commissioning results.

Equilibrium and global MHD stability study of KSTAR high beta plasmas under passive and active mode control

The Korea Superconducting Tokamak Advanced Research, KSTAR, is designed to operate a steady-state, high beta plasma while retaining global magnetohydrodynamic (MHD) stability to establish the scientific and technological basis of an economically attractive fusion reactor. An equilibrium model is established for stability analysis of KSTAR. Reconstructions were performed for the experimental start-up scenario and experimental first plasma operation using the EFIT code. The VALEN code was used to determine the vacuum vessel current distribution. Theoretical high beta equilibria spanning the expected operational range are computed for various profiles including generic L-mode and DIII-D experimental H-mode pressure profiles. Ideal MHD stability calculations of toroidal mode number of unity using the DCON code shows a factor of 2 improvement in the wall-stabilized plasma beta limit at moderate to low plasma internal inductance. The planned stabilization system in KSTAR comprises passive stabilizing plates and actively cooled in-vessel control coils (IVCCs) designed for non-axisymmetric field error correction and stabilization of slow timescale MHD modes including resistive wall modes (RWMs). VALEN analysis using standard proportional gain shows that active stabilization near the ideal wall limit can be reached with feedback using the midplane segment of the IVCC. The RMS power required for control using both white noise and noise taken from NSTX active stabilization experiments is computed for beta near the ideal wall limit. Advanced state-space control algorithms yield a factor of 2 power reduction assuming white noise while remaining robust with respect to variations in plasma beta.

Design features of the KSTAR in-vessel control coils

In-vessel control coils (IVCCs) are to be used for the fast plasma position control, field error correction (FEC), and resistive wall mode (RWM) stabilization for the Korea Superconducting Tokamak Advanced Research (KSTAR) device. The IVCC system comprises 16 segments to be unified into a single set to achieve following remarkable engineering advantages; (1) enhancement of the coil system reliability with no welding or brazing works inside the vacuum vessel, (2) simplification in fabrication and installation owing to coils being fabricated outside the vacuum vessel and installed after device assembly, and (3) easy repair and maintenance of the coil system. Each segment is designed in 8 turns coil of 32 mm × 15 mm rectangular oxygen free high conductive copper with a 7 mm diameter internal coolant hole. The conductors are enclosed in 2 mm thick Inconel 625 rectangular welded vacuum jacket with epoxy/glass insulation. Structural analyses were implemented to evaluate structural safety against electromagnetic loads acting on the IVCC for the various operation scenarios using finite element analysis. This paper describes the design features and structural analysis results of the KSTAR in-vessel control coils.

Safety design of Pb–Bi-cooled direct contact boiling water fast reactor (PBWFR)

In Pb–Bi-cooled direct contact boiling water small fast reactor (PBWFR), steam is generated by direct contact of feedwater with primary Pb–Bi coolant above the core, and Pb–Bi coolant is circulated by steam lift pump in chimneys. Safety design has been developed to show safety features of PBWFR. Negative void reactivity is inserted even if whole of the core and upper plenum are voided hypothetically by steam intrusion from above. The control rod ejection due to coolant pressure is prevented using in-vessel type control rod driving mechanism. At coolant leak from reactor vessel and feedwater pipes, Pb–Bi coolant level in the reactor vessel required for decay heat removal is kept using closed guard vessel. Dual pipes for feedwater are employed to avoid leak of water. Although there is no concern of loss of flow accident due to primary pump trip, feedwater pump trip initiates loss of coolant flow (LOF). Injection of high pressure water slows down the flow coast down of feedwater at the LOF event. The unprotected loss of flow and heat sink (ATWS) has been evaluated, which shows that the fuel temperatures are kept lower than the safety limits.

유전알고리듬을 이용한 비균일 하중을 받는 구조물의 지지위치 최적화 연구

It is important to determine supporting locations for structural stability when a structure is loaded with non-uniform load or supporting locations as well as the number of the supporting structures are restricted by the problem of space. Moreover, the supporting location optimization of complex structure in real world is fTequently faced with discontinuous design space. Therefore, the traditional optimization methods based on derivative are not suitable Whereas, Genetic Algorithm (GA) based on stochastic search technique is a very robust and general method. The KSTAR in-vessel control coil installed in vacuum vessel is loaded with nonelectro-magnetic load and supporting locations are restricted by the problem of space. This paper shows the supporting location optimization for structural stability of the in-vessel control coil. Optimization has been performed by means of a developed program. It consists of a Finite Element Analysis interfaced with a Genetic Algorithm. In addition, this paper presents an algorithm to find an optimum solution in discontinuous space using continuous design variables.

Design, fabrication, installation and testing of in-vessel control coils for DIII-D

Since 1995, DIII-D has performed correction of magnetic field imperfections using a set of six external picture frame coils located on the vessel mid-plane. Recently, these coils have also demonstrated significant benefits when used for feedback of the resistive wall mode, an instability that limits the plasma performance at high beta. Modelling has shown that substantial performance improvements can be achieved by installing new coils inside the vessel and expanding the poloidal coverage above and below the mid-plane. Two prototype internal coils were installed in 2001 and have been tested successfully. Installation of a set of twelve internal coils and magnetic sensors in the DIII-D tokamak is to be completed in December 2002. The design requirement for the new coil system was to maximize the magnetic field at the plasma edge, operate with a frequency range of dc to 1000 Hz, and fit behind the existing graphite wall tiles. The coil design adopted and installed is a water-cooled hollow copper conductor insulated with polyamide and housed inside a stainless steel tube that forms a vacuum boundary. The coil is rigidly mounted to the inside of the vacuum vessel. The primary challenge in the design of these coils was in joining of both the copper conductor and the stainless tube without overheating the polyamide insulator.

In-vessel Type Control Rod Drive Mechanism Using Magnetic Force Latching for a Very Small Reactor

A highly reliable control rod drive mechanism driven by an electric motor installed inside the reactor vessel (INV- CRDM) for a very small reactor has been designed. The INV-CRDM contributes to the compactness and simplicity of the reactor system, and can eliminate the possibility of a rod ejection accident. In the design, a new type of latch mechanism using an electromagnetic force to directly connect both of the shafts, one of which was the motor driven shaft and the other the control rod driving shaft, was applied so as to make the INV-CRDM very compact. The cable supplying current remained stationary, even when both of the shafts was moving. The required functions of the latch mechanism are to maintain an adequate latching force for the control rod shaft to move within a stroke of 370 mm, and to release the shafts in a shorter time than 0.2 s after a scram signal is received. A functional test with a model that approximately simulated the design was conducted to test the latching force and de-latching at room temperature. The test showed that the latching force increased with the current of the magnet coil, as did the de-latch time. The post-test analysis with a finite element analysis code revealed that the clearance between the two shafts greatly affected the latching force. With the same analysis method, the design analysis of the latch mechanism at a high temperature condition of 300°C was conducted, and it was confirmed that the latch mechanism contained enough latching force.

Magnetics control simulations for KSTAR plasmas

Simulation studies on magnetics control for the proposed Korea Superconducting Tokamak Advanced Research (KSTAR) plasmas are presented. Here, the magnetics control includes the fast time scale (∼10 ms) plasma position control, the slow time scale (∼1 s) plasma shape and current control, and the control of resistive wall modes (RWM). The plasma position control is realized by using two pairs of inner control coils inside KSTAR vacuum vessel. The plasma shape and current control is achieved by using superconducting poloidal field (PF) coils in KSTAR. The voltage limitations in PF coil power supplies are taken into account in the simulation. A set of 12 in-vessel control coils, which are mounted on the inner wall of KSTAR vacuum vessel, are used for the control of RWMs. The required feedback currents in in-vessel control coils for the stabilization of RWMs are estimated.

In-vessel Type Control Rod Drive Mechanism Using Magnetic Force Latching for a Very Small Reactor.

In-vessel Type Control Rod Drive Mechanism Using Magnetic Force Latching for a Very Small Reactor.

Development of In-vessel Type Control Rod Drive Mechanism for Marine Reactor

A highly reliable control rod drive mechanism (CRDM) installed inside the reactor vessel has developed for use of an advanced marine reactor. This CRDM contributes to compactness and simplicity of the reactor system, and it can eliminate the possibility of a rod ejection accident. The CRDM works in the high temperature and high pressure water—310°C and 12 MPa, the same atmosphere as the primary loop. Driving force is produced by a synchronous motor with the rotor of a permanent magnet, which has been developed. An innovative latch mechanism using separable ball nuts can latch the driving shaft connecting the control rod and de-latch it for scram. The rod position detector using a magnetostrictive wire type sensor on the principle of Wiedeman effect has been developed, accuracy of which is verified to have a detecting error within 1.2 mm. Ball bearings for thrust and radial supports in rotation have been developed to be capable of working under the high temperature water for a long period. Under condition of high temperature water, the performance tests on latching, de-latching, holding and vertical movement, and durability test were conducted to verify the design conditions. This CRDM can be applied to not only the marine reactor, but also to light water reactors, PWR and BWR.

Development of In-vessel Type Control Rod Drive Mechanism for Marine Reactor.

A highly reliable control rod drive mechanism (CRDM) installed inside the reactor vessel has developed for use of an advanced marine reactor. This CRDM contributes to compactness and simplicity of the reactor system, and it can eliminate the possibility of a rod ejection accident. The CRDM works in the high temperature and high pressure water—310°C and 12 MPa, the same atmosphere as the primary loop. Driving force is produced by a synchronous motor with the rotor of a permanent magnet, which has been developed. An innovative latch mechanism using separable ball nuts can latch the driving shaft connecting the control rod and de-latch it for scram. The rod position detector using a magnetostrictive wire type sensor on the principle of Wiedeman effect has been developed, accuracy of which is verified to have a detecting error within 1.2 mm. Ball bearings for thrust and radial supports in rotation have been developed to be capable of working under the high temperature water for a long period. Under condition of high temperature water, the performance tests on latching, de-latching, holding and vertical movement, and durability test were conducted to verify the design conditions. This CRDM can be applied to not only the marine reactor, but also to light water reactors, PWR and BWR.

The Effects of Process Control Strategies on Composting Rate and Odor Emission

Poor compost quality and odor emission are often significant problems in the composting industry. Composting process control can potentially help reduce both of these problems. In spite of the recent development of a number of process control strategies, very few direct comparisons have been made between these, particularly in terms of compost quality and odor emission. To help address this need, a series of experiments were conducted to evaluate the effects of several in-vessel process control strategies on organic matter conversion, nitrogen transformation and pH, and odor emission. The strategies focussed on aeration control. Fixed aeration, temperature feedback, oxygen feedback, and combined temperature/oxygen feedback algorithms were tested. A modified algorithm called linear temperature feedback was also developed and tested. Results showed that the compost temperature profiles were quite similar for the various feedback control algorithms, whereas fixed rate aeration led to significantly higher temperature, as expected. Compost properties such as C:N ratio and organic matter loss were also similar between process control methods. However, oxygen content was maintained more consistently using oxygen feedback or linear temperature feedback algorithms. Linear temperature feedback is preferable to oxygen feedback in that it does not require oxygen sensors to operate. Mass emission rates of odorous gas (methyl mercaptan and dimethyl sulfide) were typically found to increase with higher aeration rates, such as those used to limit temperature, though the gas concentration was lower. For maximum retention of nitrogen, adequate supply of readily biodegradable carbon in the feedstock is vital.

A Concept of Passive Safety, Pressurized Water Reactor System with Inherent Matching Nature of Core Heat Generation and Heat Removal

The reduction of manpower in operation and maintenance by simplification of the system are essential to improve the safety and the economy of future light water reactors. At the Japan Atomic Energy Research Institute (JAERI), a concept of a simplified passive safety reactor system JPSR was developed for this purpose and in the concept minimization of developing work and conservation of scale-up capability in design were considered. The inherent matching nature of core heat generation and heat removal rate is introduced by the core with high reactivity coefficient for moderator density and low reactivity coefficient for fuel temperature (Doppler effect) and once-through steam generators (SGs). This nature makes the nuclear steam supply system physically-slave for the steam and energy conversion system by controlling feed water mass flow rate. The nature can be obtained by eliminating chemical shim and adopting in-vessel control rod drive mechanism (CRDM) units and a low power density core. In order to simplify the system, a large pressurizer, canned pumps, passive residual heat removal systems with air coolers as a final heat sink and passive coolant injection system are adopted and the functions of volume and boron concentration control and seal water supply are eliminated from the chemical and volume control system (CVCS). The emergency diesel generators and auxiliary component cooling system of “safety class” for transferring heat to sea water as a final heat sink in emergency are also eliminated. All of systems are built in the containment except for the air coolers of the passive residual heat removal system. The analysis of the system revealed that the primary coolant expansion in 100% load reduction in 60 s can be mitigated in the pressurizer without actuating the pressure relief valves and the pressure in 50% load change in 30 s does not exceed the maximum allowable pressure in accidental conditions in regardless of pressure regulation.