Sunil Pak

Diagnostic neutron activation system for KSTAR

The Korea Superconducting Tokamak Advanced Research (KSTAR) device has begun deuterium plasma operation and increased neutron generation from D-D fusion reaction in the plasma is expected as the heating power of the plasma increases. A neutron activation system utilizing the pneumatic transfer of encapsulated metal samples has been implemented to monitor the fusion neutron source strength and the total fusion power from the KSTAR plasma. The prototype pneumatic transfer system for the ITER neutron activation system was slightly modified to be used for KSTAR, and a High Purity Germanium (HPGe) detection system was used to count gamma photons from the activated samples. The Monte Carlo code MCNP and the inventory code FISPACT were also used for the calculations to evaluate the total number of neutrons emitted from the D-D fusion reactions in the KSTAR plasma. The analysis of data from an activation measurement obtained from the 2011 KSTAR campaign shows the neutron flux at the irradiation station was 2.2 × 108 cm−2s−1, and the total neutron yield was 4.7 × 1013 n/s for a typical NBI-heated, H-mode KSTAR plasma shot.

Operational Radioactivity Evaluation of ITER Diagnostic Neutron Activation System

At the International Thermonuclear Experimental Reactor (ITER), a neutron activation system is foreseen to measure neutron fluence on the first wall and to evaluate total fusion power from ITER plasma. This system utilizes the method of counting gamma radiation from the metal sample irradiated by the neutron flux at the irradiation end located near the plasma. Evaluation of the operational radioactivity of the sample, which has been induced during ITER operation, is necessary in order to determine sample material, mass, and the appropriate irradiation time. Neutronic calculations are performed to evaluate operational activity of the sample. Activation of the sample is calculated according to the irradiation time for various sample materials such as Si, Al, Ti, Fe, Nb, and Cu and by assuming that the samples are irradiated at the midplane inboard region where the neutron flux is one of the strongest.

Dynamic Amplification Factor of the ITER Diagnostic Upper Port Plug

The diagnostic upper port plug in ITER is a long metal box cantilevered to the vacuum vessel port with 42 × M52 studs and nuts. The plug structure has a heavy payload at the front, such as the diagnostic first wall and the diagnostic shield module to protect the diagnostic components from plasma and neutron fluxes. This kind of structural configuration is susceptible to a resonance with the transient external load. For the upper port plug, the design-driving load is electromagnetic (EM) forces due to plasma disruptions. In this paper, the dynamic amplification factor (DAF) of the structure is calculated for such EM loads. The bolted joint at the back flange of the plug structure is also considered together with the port extension of the vacuum vessel and its influence on the dynamic behavior is investigated. The analysis results show that the bolted joint reduces the DAF as well as the natural frequency of the structure.

ITER perspective on fusion reactor diagnostics - A spectroscopic view

The ITER tokamak requires diagnostics that on the one hand have a high sensitivity, high spatial and temporal resolution and a high dynamic range, while on the other hand are robust enough to survive in a harsh environment.In recent years significant progress has been made in addressing critical challenges to the development of spectroscopic (but also other) diagnostics. This contribution presents an overview of recent achievements in 4 topical areas:• First mirror protection and cleaning• Nuclear confinement• Radiation mitigation strategy for optical and electronic components• Calibration strategies

Factors affecting the measurement accuracy of ITER neutron activation system

One of the main purposes of the ITER2 neutron activation system (NAS) is to evaluate the total neutron production rate from all over the plasma. The measurement accuracy depends on the position and profile of the plasma and the material in front of the irradiation end. It is required to minimize the amount of material and its density variation across the field of view between the plasma and the irradiation end. Due to the radiation and thermal environment of the ITER in-vessel, however, the measurement from ITER NAS cannot avoid the strong influence from in-vessel materials such as the diagnostic first wall, blanket modules, and divertor cassettes, those are located near the irradiation ends. In order to improve the reliability of the measurement in such environment, special cutouts in the diagnostic first wall are introduced near the irradiation end structures located in the port plugs. The effect of the materials and the position and profile of the neutron source in the plasma are evaluated for these irradiation locations, as well as the ones under the divertor cassettes and between blanket modules, by the neutron transport calculation. Calculation results show that simultaneous measurements at upper port and divertor location can provide highly accurate results even without a position or profile correction from other diagnostics.

Electromagnetic analysis of ITER diagnostic port plugs and diagnostic components during plasma events

Eddy current induced electromagnetic (EM) loads during plasma events are the design driver for ITER port plug (PP) structure, diagnostic first wall (DFW), shield module (DSM) and diagnostic system supported by the PP structure. Generic models using commercial software OPERA, MAXWELL and ANSYS are developed and benchmarks are performed for global EM analysis to obtain port-specific design driving EM loads. The 20 degree vessel sector models of the upper and equatorial PP structure take the same ITER TF, CS and PF coil and plasma current (15 MA baseline plasma scenario) as input, then solve for eddy current induced on all passive structural components for various DINA disruption cases. The worst load case can be exponential decay or linear decay of plasma current depending on the dimension and location of diagnostic components inside the PPs. Static and transient magnetic fields from generic models are mapped onto an excel datasheet to establish component design load specification. Three levels of modeling effort are suggested. The global model analysis is used to validate the impact of component design to the PP global system response as well as to study its component dynamic effect as a result of the disruption loads on the PP structure. The local sub-model analysis can be used to extract more accurate EM loads on diagnostics and the excel datasheet of static and transient field maps are used as the initial design load specification for in-port components.

Integration of diagnostics on ITER

Diagnostics play a very important role in the modern Tokamak where optimum performance is essential. To achieve this, the device must be equipped with reliable and robust sensors and instrumentation that allow the operation envelope to be fully explored. Development of these diagnostics to maintain this reliability is necessary. Further to the development, the systems must be integrated in a way that maintains their performance while simultaneously satisfying the key requirements needed for safety and tokamak operation. ITER will have 50 diagnostics; almost all of which are utilized primarily for the real-time operation of the tokamak. While there is still much work to do, to date, significant progress has been made in the development of these systems. The work load for the developments is shared across all the ITER partners. This paper focuses on the challenges for the integration of the systems.