A preliminary assessment of beryllium dust oxidation during a wet bypass accident in a fusion reactor

A beryllium dust oxidation model has been developed at the Idaho National Laboratory (INL) by the Fusion Safety Program (FSP) for the MELCOR safety computer code. The purpose of this model is to investigate hydrogen production from beryllium dust layers on hot surfaces inside a fusion reactor vacuum vessel (VV) during in-vessel loss-of-cooling accidents (LOCAs). This beryllium dust oxidation model accounts for the diffusion of steam into a beryllium dust layer, the oxidation of the dust particles inside this layer based on the beryllium–steam oxidation equations developed at the INL, and the effective thermal conductivity of this beryllium dust layer. This paper details this oxidation model and presents the results of the application of this model to a wet bypass accident scenario in the ITER device.     

Electromagnetic load calculation of the ITER machine using a single finite element model including narrow slits of the in-vessel components

We evaluate electromagnetic (EM) loads on the main systems of the ITER machine using a single finite element model. The 20° sector of the full ITER machine includes the main in-vessel components as well as the vacuum vessel. Narrow slits of the in-vessel components are effectively modeled by using the element splitting method without significant increase of computation memory and time as well as without sacrificing the accuracy. Furthermore, the halo current is taken into account at the same time together with the plasma current. To apply both currents concurrently, dedicated conversion codes are utilized to transfer the plasma simulation results by DINA to the electromagnetic analysis by ANSYS-EMAG used here. The electromagnetic loads on the ITER machine are calculated for various disruption scenarios. Investigation on the analysis results is made to find the worst plasma disruption case and the design-driving load component for each system as well as to compare load contribution from eddy and halo currents. The effect of the narrow slits on load reduction is also examined.     

Decay Heat Removal by Natural Circulation of Vacuum Vessel Coolant for ITER

The decay heat-driven temperature transients of the in-vessel components following a postulated loss of all in-vessel cooling have been calculated. The resulting time-dependent heat load to the vacuum vessel is due to radiation from the backplate and convection of postulated steam between backplate and vacuum vessel. It is shown, that even for a failure of all in-vessel cooling and total loss of power, the ITER design can rely on passive decay heat removal by natural circulation in one of the two existing cooling loops of the vacuum vessel. A mathematical model describes the transient operating conditions and shows that the temperature established by natural circulation does not exceed 200°C at the maximum shut down heat load to the vacuum vessel. Therefore, no additional emergency cooling system is required if the existing heat exchanger is designed for natural circulation and a bypass is used during normal operation to maintain operation temperature.     

Fusion Power: Looking to the Future

Presentations that were made at a Fusion Power Associates symposium, Fusion Power: Looking to the Future, are summarized. The topics included overview and personal perspectives, status of ITER, stellarators, inertial confinement and innovative concepts. Also included is a summary of work on laser fusion at Osaka University.     

Safety Analysis Results for ITER In-Vessel Loss of Coolant Events

A number of postulated in-vessel loss of coolant accidents (LOCAs) associated with the first wall and baffle cooling systems of the ITER detailed design have been analyzed for the ITER Non-site Specific Safety Report (NSSR-1). A range of break sizes from one first wall tube break (1.57 × 10^–4 m²) to damage to all in-vessel components (0.6 m² break) have been examined. These events span the ITER event classification from likely events to extremely unlikely events. In addition, in-vessel pipe breaks in combination with bypass of the two confinement barriers through a generic penetration have been examined. In all cases, when the vacuum vessel pressure suppression system is activated, most of the radioactive inventory is carried to the suppression pool where it remains for the duration of the event. Releases in these events are well within ITER release limits.     

Safety and Environmental Activities for ITER

This paper will summarize highlights of the safety approach and discuss the ITER EDA safety activities. The ITER safety approach is driven by three major objectives: (1) Enhancement or improvement of fusion's intrinsic safety characteristics to the maximum extent feasible, which includes a minimization of the dependence on dedicated safety systems; (2) Selection of conservative design parameters and development of a robust design to accommodate uncertainties in plasma physics as well as the lack of operational experience and data; and (3) Integration of engineered mitigation systems to enhance the safety assurance against potentially hazardous inventories in the device by deploying well-established nuclear safety approaches and methodologies tailored as appropriate for ITER.     

Monte-Carlo Calculation of the γ-Ray Dose Distribution Inside a Wall of a Building and in Air

The Monte Carlo method is used to calculate the coefficients used to convert from the measured (using retrospective luminescence dosimetry) -ray dose inside the wall of a building to the dose in air. The -ray dose distributions in the wall and in air were calculated simultaneously. The calculations were performed for a -ray sources located in air, on the surface of soil, and in a 0–5 cm thick top layer of soil. The -ray spectra in air and inside a wall are calculated and it is shown that the spectrum depends strongly on the type of source – the average energy of the spectrum is 0.25 MeV for a source located on a soil surface, 0.39 MeV for a source located inside the soil, and 0.53 MeV for a source located in air.     

Activation γ-Spectrometric Method of Determining the Mass of Samples of Nuclear Materials

The possibility of using activation -spectrometry to determine the mass content of nuclear materials in matter is investigated. Irradiation of samples for a short time with moderated neutrons from a ~10⁷ sec^–1 Pu–Be source is used to induce 1436 keV -ray emission from ¹³⁸Cs. These -rays are suitable for measurements; the mass of the nuclear materials is determined from the intensity of the radiation. Three series of experiments are performed with sets of samples consisting of uranium and uranium dioxide with different mass and degrees of enrichment.Experiments showed that the error in determining the mass of uranium samples can reach 1–3% with 30–60 min irradiation and the same measurement duration.Special experiments were performed to investigate the influence of the experimental geometry and the self-absorption of the rays in the sample, which limit the possibility of -spectrometric measurements on samples of nuclear materials.The activation -spectrometric method can be used for analyzing metallic uranium samples, powder samples, samples of fuel micropellets and uranium hexafluoride, and plutonium samples.     

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

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

Possibilities for long pulse ignited tokamak experiments using resistive magnets

Tokamak designs based on copper and copper alloy magnets could be used for the investigation of the physics issues associated with long pulse (>40 sec) ignited operation during the next phase of fusion research. The engineering characteristics of designs with magnets that use copper or beryllium copper alloys are presented. Active cooling of the magnets with either liquid nitrogen or water is considered. Inertial cooling is also discussed. The physics performance of the designs is calculated and compared to the performance of the Physics Phase ITER design, and the BPX tokamak.MIT Plasma Fusion Center.     

Analysis of an ex-vessel break in the ITER divertor cooling loop

In the present work the integrated ECART code, developed for severe accident analysis in LWRs, is applied on the analysis of a large ex-vessel break in the divertor cooling loop of the international thermonuclear experimental reactor (ITER). A comparison of the ECART results with those obtained by Studsvik Nuclear AB (S), utilizing the MELCOR code, was also performed in the general framework of the quality assurance program for the ITER accident analyses. This comparison gives a good agreement in the results, both for thermal-hydraulics and the environmental radioactive releases. Mainly these analyses, from the point of view of the ITER safety, confirm that the accidental overpressure inside the vacuum vessel and the Tokamak cooling water system (TWCS) Vault is always well below the design limits and that the radioactive releases are adequately confined below the ITER guidelines.     

Preliminary safety analysis of ex-vessel LOCA for the European HCPB TBM system

Ex-vessel loss of coolant accident caused by a double-ended pipe break of the helium coolant system inside port cell is considered as one of the most critical accident for the European Helium Cooled Pebble Beds Test Blanket Module (HCPB TBM) system. The resulting rapid helium blow-down causes an immediate block of the TBM cooling, which requires a prompt plasma shutdown. Even after the plasma shutdown the temperature can increase over the design limit and the accident sequence can lead up to a break of the TBM box protection after the failure of different protection systems. Thus air ingresses in the vacuum vessel from the damaged TBM system and steam from the surrounding ITER blanket and divertor structures. The evaluation of this sequence is very important for the definition of the correct protection strategy of the system. To consider all these different events a methodology has been developed in KIT combining different codes for a complete analysis of the accident. In particular, this paper shows an application of MELCOR code to model beryllium–steam reaction in a particular accidental sequence for the long term cooling.     

Constraints on very-high-β, edge-stabilized tandem-mirror reactors

There are several tandem-mirror schemes which propose a very high and edge stabilization for the center-cell plasma ( being the ratio of the plasma pressure to the vacuum magnetic-field pressure). While the exact criteria for the edge stabilization are uncertain, it is possible to analyze the option space in which a very-high- mirror reactor would operate. The primary physics constraints on such a reactor are the energy balance at ignition, the buildup of He⁴ ash and the hot-particle( ^hot ), and the need for adiabatic conservation of the hot-particle gyro-orbits in the axial field gradients at the center-cell ends. There are also engineering constraints on the allowable wall loading and plant size. In this paper, a wall-stabilized tandem-mirror reactor is analyzed and is found to be an attractive device requiring low center-cell vacuum fields (of the order of 2 to 3 tesla). A primary requirement is that the plasma edge have a thermal conductivity near classical values.     

Possible Pathways for Pursuing Burning Plasma Physics and Fusion Energy Development

This report has been prepared in response to a request from the U.S. Department of Energy's (DOE) Office of Fusion Energy Sciences to consider possible alternatives on reduced cost options for next-step devices. A central focus of next-step devices is the study of burning plasmas, which explore the impact of substantial fusion energy production via the deuterium-tritium reaction.An important part of the U.S. Fusion Energy Sciences Program is its participation in the International Thermonuclear Experimental Reactor (ITER) program. Taking into account the international situation and U.S. domestic issues, the ITER process is exploring reduced-cost options to the present ITER device. A Special Working Group, reporting to the ITER Council, has been formed to explore these issues on behalf of the ITER Parties, i.e., the European Union, Russian Federation, Japan, and the United States. This report and its related activities will aid the United States in the international process.This report is the result of a broad-based U.S. community effort to discuss, debate, and work together on the crucial issues involved in considering next-step options. The main content of this report is based on three potential pathways identified at a broadly attended community Forum for Next-Step Fusion Experiments (University of Wisconsin, Madison, April 1998) organized principally by the University Fusion Association and by the work of the ITER Steering Committee—US (ISCUS) on reduced cost ITER options. The Madison Workshop was followed by a smaller Workshop on Next-Step Options (University of California, San Diego, June 1998) to focus on preparing this report. A broadly-announced Website was established to facilitate access to documents related to this process.     

Diagnostics carried by a light multipurpose deployer for vacuum vessel interventions

ITER will greatly rely on remote-handling operations to accomplish its scientific missions. Robotic systems will also be required to operate inside vacuum vessels in order to limit or replace human access, to intervene quickly between experimental sessions for in-vessel inspections and measurements, and to preserve the machine conditioning and thus improve machine availability. In this prospect, a multipurpose carrier prototype called Articulated Inspection Arm (AIA) was developed by CEA laboratories within the European work program. With an embedded camera, it successfully demonstrated close inspection feasibility inside Tore Supra tokamak. The AIA robot was designed for mini-invasive operations with interchangeable diagnostics to be plugged at its head. This covers various applications for the safety, the operation and the scientific mission (in-vessel inspection, plasma diagnostics calibrations or inner components analysis and treatments). This paper presents recent analysis and results obtain with diagnostics developed by CEA for in-vessel remote-handling intervention.     

Results of simulations of stability and quench behavior for pulsed tests of the ENEA 12T Nb3Sn CICC coil

In the ENEA Frascati Laboratory a facility is being assembled to test the ENEA Nb₃Sn CICC coil in pulsed regimes. The characteristics of the coil (dimensions, cable-in-conduit conductor, strand designed for use in variable field) are such to make these tests of primary importance to predict the behaviour of the ITER (International Thermo-nuclear Experimental Reactor) Central Solenoid Model Coil, which will be built in the next two years. In particular, the stability and quench behaviour of the coil will be tested and compared to the predictions of the thermo-hydraulic analysis code SARUMAN. Other important parameters will be the ramp rate limination, the limiting current and the conductor losses. Several testing scenarios (ramp up and discharge) are described and the present status of testing programme definition is given, together with the associated analyses.     

Evaluation of the Environmental Gamma-ray Dose Rate by Skyshine Analysis During the Maintenance of an Activated TFC in ITER

Gamma-ray exposure dose rates at the ITER site boundary were estimated for the cases of removal of a failed activated Toroidal Field (TF) coil from the torus and removal of a failed activated TF coil together with a sector of the activated Vacuum Vessel (VV). Skyshine analyses were performed using the two-dimensional SN radiation transport code, DOT3.5. The exposure gamma-ray dose rates on the ground at the site boundary (presently assumed to be 1 km from the ITER building), were calculated to be 1.1 and 84 Sv/year for removal of the TF coil without and with a VV sector, respectively. The dose rate level for the latter case is close to the tentative radiation limit of 100 Sv/year so an additional 14 cm of concrete is required in the ITER building roof to satisfy the criterion for a safety factor often for the site boundary dose rate.     

Simulation of VDE under intervention of vertical stability control and vertical electromagnetic force on the ITER vacuum vessel

Vertical displacement events (VDEs) and disruptions usually take place under intervention of vertical stability (VS) control and the vertical electromagnetic force induced on vacuum vessels is potentially influenced. This paper presents assessment of the force that arises from the VS control in ITER VDEs using a numerical simulation code DINA. The focus is on a possible malfunctioning of the ex-vessel VS control circuit: radial magnetic field is unintentionally applied to the direction of enhancing the vertical displacement further. Since this type of failure usually causes the largest forces (or halo currents) observed in the present experiments, this situation must be properly accommodated in the design of the ITER vacuum vessel. DINA analysis shows that although the ex-vessel VS control modifies radial field, it does not affect plasma motion and current quench behavior including halo current generation because the vacuum vessel shields the field created by the ex-vessel coils. Nevertheless, the VS control modifies the force on the vessel by directly acting on the eddy current carried by the conducting structures of the vessel. Although the worst case was explored in a range of plasma inductance and pattern of VS control in combination with the in-vessel VS control circuit, the result confirmed that the force is still within the design margin.