Ingress Of Coolant Event Research Articles

BEPU analysis of an Upgraded ICE facility test by TRACE/DAKOTA coupling

The Ingress of Coolant Event (ICE) in the plasma chamber is one of the safety issues in fusion nuclear plants. The best estimate thermal-hydraulic system codes adopted to perform deterministic safety analysis should be validated against the phenomena typical of accidental transients in fusion installations. TRACE (TRAC/RELAP Advanced Computational Engine), best estimate thermal-hydraulic system code developed by USNRC, has been adopted to simulate an ICE. The calculated results have been compared to the experimental data obtained in one test performed in the upgraded Integrated ICE facility at JAERI. In this updated configuration the pressure suppression system is connected to the top of the plasma chamber instead of the bottom of the vacuum vessel. The facility nodalization has been developed in the SNAP environment/architecture. To qualify the code and the nodalization, an accuracy evaluation has been performed both from a qualitative and quantitative point of view. Then, considering the presence of some uncertainties in the input-deck development, an uncertainty analysis has been carried out. The probabilistic method to propagate the input uncertainties has been selected and the analysis has been carried out with the DAKOTA toolkit coupled with TRACE code in SNAP. In the uncertainty analysis, some relevant statistical parameters have been considered to characterize the dispersion of the results and the correlation between the uncertain input parameters selected and the PC pressure chosen as figure of merit.

Flash-Boiling Characterization During Ingress of Coolant Event for Dust Issue in ITER

Abstract Dust resuspension inside the vacuum vessel (VV) is one of the safety issues of the fusion reactor ITER. Plasma interaction with the plasma facing components (PFC) leads to their erosion, generating dust. One of the accident scenarios leading to dust resuspension is the ingress of coolant event (ICE) where a leak of the coolant pipes inside the VV conducts to injection and flash atomization of the cooling water. The metallic dust, produced by the erosion, is then oxidized by water throughout an exothermic reaction that produces hydrogen leading to a loss of confinement risk due to hydrogen and dust combustion. The steam flow, produced by the flash atomization of the liquid leaking from the breach, is considered to be the main source of dust resuspension. Therefore, experimentations about the two-phase flow generated by the flashing liquid jet are important to identify the main physical phenomena involved in the aerosol particles resuspension for ITER-like conditions that impose in particular low pressure level. Flash-boiling experiments were conducted under primary vacuum conditions. We studied the behavior and the structure of the flow resulting from superheated water injection into low pressure environment. Using shadowgraphy and particle image velocity (PIV), qualitative information and quantitative measurements on the two-phase flow that develops for different superheat conditions were gathered. The measured spray lateral spreading and droplets velocity are shown to increase with the superheat level. The use of a transparent nozzle also confirmed the strong coupling between the external structure of the atomized spray and the two-phase flow that develops upstream of the coolant circuit breach.

Analysis of nodalization strategies to model a suppression tank for fusion plants with TRACE code

The Ingress of Coolant Event (ICE) in the plasma chamber is one of the main safety issues in nuclear fusion plants. The ICE is caused by the rupture of the coolant tubes installed in the plasma facing components; the coolant ingress causes a pressure rise in the plasma chamber and vacuum vessel, normally under high-vacuum conditions. To mitigate the system pressurization leading to mechanical structure failure, a pressure suppression system is installed. Safety analyses of the hypothetical challenging accidental scenarios can be conducted by deterministic models that need to be validated against experimental data characterizing the target phenomena. The paper presents a study of different nodalization strategies for modeling a suppression tank for fusion plants, using the best estimate thermal-hydraulic system code TRACE (TRAC/RELAP Advanced Computational Engine). The TRACE code is developed by USNRC to perform safety analyses for light water fission reactors. Both mono-dimensional and three-dimensional approaches were adopted to model the suppression tank. The experimental data from the JAERI Integrated ICE facility (scaling factor of 1/1600 with respect to the ITER-FEAT design) was used to benchmark different nodalization options. In addition to the qualitative accuracy evaluation, the Fast Fourier Transform Based Method (FFTBM) was applied to quantify the accuracy of the code results for different nodalizations.

ASTEC - RAVEN coupling for uncertainty analysis of an ingress of coolant event in fusion plants

The integrated ICE (Ingress-of-Coolant Event) facility, scaled 1/1600 with respect to the ITER-FEAT design, was built at JAERI with the aim of reproducing the phenomenology occurring in an ICE accident. An ICE occurs when a rupture in the coolant pipes causes the pressurized coolant to enter into the Plasma Chamber, which is held under high vacuum condition. A suppression system is used to mitigate the overpressurization and to prevent mechanical damages to the structures. The CPA module of the ASTEC severe accident code (Study carried out with ASTEC V2, IRSN all rights reserved, [2020]), has been adopted for the modelling and the simulation of a test conducted in the ICE facility. The experimental results of the main thermal-hydraulic parameters have been compared to the code results to characterize the ASTEC capability to predict the phenomenology of a low-pressure two-phase flow transient occurring in a fusion reactor. By coupling the ASTEC code with the uncertainty tool RAVEN, developed by Idaho National Laboratory, an uncertainty analysis has been conducted on the transient. The aim of the present activity is to investigate the dispersion and the sensitivity of the code response to the variation of selected uncertain input parameters, which could influence the simulation of an ICE. The activity also provides a first application of uncertainty analysis through the RAVEN-ASTEC coupling.

Ingress of Coolant Event simulation with TRACE code with accuracy evaluation and coupled DAKOTA Uncertainty Analysis

Among the Postulated Initiating Events in nuclear fusion plants, the Ingress of Coolant Event (ICE) in the Plasma Chamber is one of the main safety issues. In the present paper, the best estimate thermal-hydraulic system code TRACE, developed by USNRC, has been adopted to study the ICE, and it has been qualified based on experimental results obtained in the Integrated ICE facility at JAERI. A nodalization has been developed in the SNAP environment/architecture, using also the TRACE 3D Vessel component where multidimensional phenomena could occur. The accuracy of the code calculation has been assessed both from a qualitative and quantitative point of view. In addition, an Uncertainty Analysis (UA), with the probabilistic method to propagate the input uncertainties, has been performed to characterize the dispersion of the results. The analysis has been carried out with the DAKOTA toolkit coupled with TRACE code in the SNAP environment/architecture. Results show the adequacy of the 3D nodalization and the capability of the code to follow the transient evolution also at a very low pressure. Response correlations have been computed to characterize the correlation between the selected uncertain input parameters and the Plasma Chamber pressure.

Remote handling strategy and prototype tooling of the ITER vacuum vessel pressure suppression system bleed line valve assembly and rupture disk assembly

As part of the development work performed under the ITER Robotic Test Facility (IRTF) program, development and substantiation of the maintenance strategy for the Vacuum Vessel Pressure Suppression System (VVPSS) must be investigated. The VVPSS is a key safety aspect of the vacuum vessel of ITER. It ensures that after a coolant leak, and subsequent pressure event, the vessel is protected from the expanding gasses by controlling and venting them appropriately. After an Ingress of Coolant Event (ICE), or during maintenance, the bleed line valve and rupture disk assemblies will require removal and replacement. During removal the confinement function of vessel must be retained. The remote handling of these components while maintaining the first confinement boundary is challenging due to the access restrictions around the pipe flange and environmental conditions in the area. To understand and develop the remote handling strategy for ITER, prototype tooling a mock-up environment has been created. This allows development and confidence in the proposed maintenance strategy leading to the final solution for use at ITER. In this paper we will discuss the prototype tests, the strategies and challenges which need to be overcome for the ITER application, and design considerations that must be taken to future designs to achieve a feasible solution for ITER.

Analysis of the Ingress of Coolant Event tests performed in the upgraded ICE facility aimed at the ECART code validation

The activities on the validation of the ECART code against the eight Ingress of Coolant Event (ICE) tests performed in the upgraded ICE facility are discussed. These analyses have been carried out to extend the ECART validation to incidental sequences related to future fusion plants, specifically - in the case of ICE - versus Loss Of Coolant Accident (LOCA) in volumes with near vacuum conditions. The upgraded ICE facility consists of a boiler, injecting water at high pressure inside a low-pressure tank simulating the Plasma Chamber (PC). This PC is in turn connected to the Vacuum Vessel (VV) through the divertor, and to the pressure Suppression (ST) by means of several relief pipes. Finally, the VV is connected with the Drain Tank (DT). Eight tests were performed investigating different numbers of relief pipes, different initial PC and VV temperatures, and different injected water mass flow rates, pressures and temperatures. The ECART results show an overall good agreement with the experimental data, confirming that ECART is also a valuable tool for the safety analysis in future fusion plants, as already pointed out in previous works.

Analysis of Ingress of Coolant Accident to the Vacuum Vessel of the W7-X Fusion Experimental Facility

An event of water coolant ingress into the vacuum vessel (VV) is one of the most important events leading to severe consequences in nuclear fusion reactors. The ingress of coolant to the VV could appear due to coolant pipe rupture of in-vessel components. Any damage of in-vessel components could lead to water ingress and may lead to pressure increase and possible damage of the VV. Therefore, it is important to understand thermohydraulic processes in the VV during the ingress of coolant event (ICE) to prevent overpressurization of the VV. This technical note updates the developed Wendelstein 7-X (W7-X) model in accordance with the experience gained from the modeling of ICE experiments. Calculation results using the updated model are compared with the results obtained using an older model and the results of other researchers. The calculation results of the updated W7-X model show a much smaller pressure increase rate in the VV compared to the old model. In order to find the maximal area of partial break, which increases pressure in the VV but does not reach burst disk activation pressure (no steam release from the VV to the environment), the best-estimate approach is provided. The results of the analysis reveal that partial break using the updated W7-X model could be much bigger than what was considered before.

Integrated assessment of thermal hydraulic processes in W7-X fusion experimental facility

Energy received from the nuclear fusion reaction is one of the most promising options for generating large amounts of carbon-free energy in the future. However, physical and technical problems existing in this technology are complicated. Several experimental nuclear fusion devices around the world have already been constructed, and several are under construction. However, the processes in the cooling system of the in-vessel components, vacuum vessel and pressure increase protection system of nuclear fusion devices are not widely studied. The largest amount of radioactive materials is concentrated in the vacuum vessel of the fusion device. Vacuum vessel is designed for the vacuum conditions inside the vessel. Rupture of the in-vessel components of the cooling system pipe may lead to a sharp pressure increase and possible damage of the vacuum vessel. To prevent the overpressure, the pressure increase protection system should be designed and implemented. Therefore, systematic and detailed experimental and numerical studies, regarding the thermal-hydraulic processes in cooling system, vacuum vessel and pressure increase protection system, are important and relevant.In this article, the numerical investigation of thermal-hydraulic processes in cooling systems of in-vessel components, vacuum vessels and pressure increase protection system of fusion devices is presented. Using the experience gained from the modelling of “Ingress of Coolant Event” experimental facilities, the numerical model of Wendelstein 7-X (W7-X) experimental fusion device was developed. The integrated analysis of the thermal-hydraulic processes occurring in the cooling system of in-vessel components, vacuum vessel and pressure increase protection system of W7-X device in the case of loss of coolant accident was performed using RELAP5 computer code. The results of analysis showed that the design of W7-X facility prevents the failure of vacuum vessel.

二流体モデルに基づくテストブランケットモジュール内水侵入事象解析

二流体モデルに基づくテストブランケットモジュール内水侵入事象解析

Real-time measurement of hydrogen generation level during beryllium dust oxidation by steam depending on the dust arrangement geometry

The paper concentrates on the safety issues in the International Thermonuclear Experimental Reactor (ITER) and describes the experiment on the measurement of hydrogen generation rate in case of Ingress of Coolant Event (ICE)—leak inside the vacuum vessel during interaction between water and beryllium (Be) dust. The ICE situation in ITER was simulated in a facility; the active spectroscopy was used to define the hydrogen content by the dynamics of oxidant concentration at a sampling frequency up to 10 Hz. Hydrogen release in time at temperatures of 500–900 °C is investigated, and different versions of dust arrangement are considered, i.e. on the surface and in a slot between armoring tiles at different initial density. The obtained results are compared with the known experiments.

Numerical study on direct-contact condensation of vapor in cold water

Numerical analyses were carried out to predict quantitatively the condensation characteristics between the water and vapor inside a suppression tank in the International Thermonuclear Experimental Reactor (ITER). The suppression tank is one of the ITER safety devices. The pressure rise in an ITER vacuum vessel during an ingress-of-coolant event (ICE) is suppressed by the effect of condensation of vapor in the suppression tank. The water–vapor two-phase flow characteristics with condensation under the ITER pressure conditions were clarified numerically. Moreover, flow visualization experiments were carried out to understand the condensation behavior under low pressure and cold water. The predicted flow configurations agreed well with the experimental results.

Comparison of athena/ relap results against ice experimental data

In order to demonstrate the adequacy of the International Thermonuclear Experimental Reactor design from a safety stand point as well as investigating the behavior of two-phase flow phenomena during an ingress of coolant event, an integrated ICE test facility was constructed in Japan. The data generated from the ICE facility offers a valuable means to validate computer codes such as athena/ relap5, which is one of the codes used at the Idaho National Engineering And Environmental Laboratory (INEEL) to evaluate the safety of various fusion reactor concepts. In this paper we compared numerical results generated by the athena code with corresponding test data from the ICE facility. Overall we found good agreement between the test data and the predicted results.

Fusion safety codes: international modeling with MELCOR and ATHENA–INTRA

For a number of years, the world fusion safety community has been involved in benchmarking their safety analyses codes against experiment data to support regulatory approval of a next step fusion device. This paper discusses the benchmarking of two prominent fusion safety thermal-hydraulic computer codes. The MELCOR code was developed in the US for fission severe accident safety analyses and has been modified for fusion safety analyses. The ATHENA code is a multifluid version of the US-developed RELAP5 code that is also widely used for fusion safety analyses. The ENEA Fusion Division uses ATHENA in conjunction with the INTRA code for its safety analyses. The INTRA code was developed in Germany and predicts containment building pressures, temperatures and fluid flow. ENEA employs the French-developed ISAS system to couple ATHENA and INTRA. This paper provides a brief introduction of the MELCOR and ATHENA–INTRA codes and presents their modeling results for the following breaches of a water cooling line into the tokamak vacuum vessel, (1) water injection in vacuum with condensation of the flashed steam and (2) water injection in vacuum. The modeling predictions are compared with experimental results from the ingress-of-coolant event (ICE) facility in Japan. It is observed that the codes’ predictions exhibit good agreement with the experiment data, which suggests that the codes include the proper physical mechanisms associated with the chosen accident scenarios. It is thereby concluded that either of the codes can be used to model the defined transient.

Entrainment behavior of activated dusts in the fusion reactor under LOVA

The safety analysis for of loss-of-vacuum accident (LOVA) as well as ingress-of-coolant event (ICE) is among the last task left in ITER R&D. On the assumption of LOVA after occurring ICE, it is corollary that activated dusts are under the wet condition. Transport behavior of in-vessel activated dusts under wet condition is not known well in comparison with dry case. In this study, the wetted particles transport behavior is investigated experimentally by using a vacuum vessel that serves as a steam generator. The relations among the relative humidity, the entrainment of particles in the exhaust gas flow and the adhesion rate of dust particles on the confined wall have been cleared. The entrainment ratio decreases as the relative humidity increases and increases as the initial pressure difference increases.

Numerical Analysis on Ingress of Coolant Event in Vacuum Vessel of Fusion Reactor Using Modified TRAC-BF1

The semi three-dimensional numerical analysis model of the FDR-ITER to simulate the in-vessel LOCA or the ICE was constructed using the modified TRAC-BF1 code in which the vacuum vessel, the pressure suppression tank and the relief pipe header are modeled using one VESSEL component. Analytical results obtained from the modified TRAC-BF1 code were compared with those from the MELCOR code concerning pressure in the vacuum vessel when a cooling pipe of 0.6 m2 flow area breaks in it. The mass flow rate of the injected water into the vacuum vessel and the initial wall temperature in the modified TRAC-BF1 input data were the same as those calculated in the MELCOR analysis. The maximum pressure in the vacuum vessel obtained from the modified TRAC-BF1 code is 10% higher than that from the MELCOR code, but it is still below the design pressure 0.5 MPa of the FDR-ITER vacuum vessel. The semi three-dimensional analysis using the modified TRAC-BF1 obtained that the maximum delay in opening time among the four rupture-disks is within 3 ms. The maximum pressure in the vacuum vessel does not exceed the design pressure, even if one of rupture-disks is not opened.

Analysis of Pressure Rise in an ITER-Like Fusion Reactor During In-Vessel LOCA by A Modified TRAC-PF1 Code

Damage of cooling tubes of plasma facing components (PFCs) results in water discharge into a vacuum vessel (W) of a fusion reactor. Flashing in vacuum, water pool boiling and impingement-jet on a surface of the PFC are the main heat transfer phenomena responsible for steam production that causes a rapid pressurization of the W. This is called an in-vessel loss-of-coolant accident (LOCA) event or ingress-of-coolant event (ICE). The ICE event is one of the most severe accidents in the fusion reactors.The integrated ICE test facility was constructed to demonstrate the safety design approach of International Thermonuclear Experimental Reactor (ITER) and obtain validation data for the ITER safety analysis codes. Then, an experimental study was performed using the integrated ICE test facility and at the same time the code validation study with the TRAC code was carried out. The pressure rise characteristics in the current ITER machine during the ICE event were analyzed numerically using the verified TRAC-PF1 code and the effects of the relief pipe diameter and suppression tank volume regarding to the pressure rise due to the ICE events were clarified quantitatively from the present analytical results.

減圧下での水蒸発と凝縮に関する実験と解析

When cooling tubes installed in plasma-facing components (PFCs) of a fusion reactor are damaged by some accidents, water may be discharged into a plasma chamber (PC) and vacuum vessel (VV) of the fusion reactor. In such a case, the discharged water will impinge on a surface of PFCs at high temperature and evaporate and then the VV may be broken due to pressurization by steam. This event is called an ingress-of-coolant event (ICE). The water-vapor two-phase flow behavior in the PC and VV during the ICE was investigated using an integrated ICE test facility. Furthermore, the condensation effect due to vapor inside a suppression tank which is installed to reduce the pressurization by the ICE was observed visually. In addition, the thermal-hydraulics during the ICE were validated numerically by the TRAC-PF1 code and analytical results on the pressure rise were in good agreement with the experimental results.

Numerical Analysis on Ingress of Coolant Event in Vacuum Vessel of Fusion Reactor Using Modified TRAC-BF1.

The semi three-dimensional numerical analysis model of the FDR-ITER to simulate the in-vessel LOCA or the ICE was constructed using the modified TRAC-BF1 code in which the vacuum vessel, the pressure suppression tank and the relief pipe header are modeled using one VESSEL component. Analytical results obtained from the modified TRAC-BF1 code were compared with those from the MELCOR code concerning pressure in the vacuum vessel when a cooling pipe of 0.6 m2 flow area breaks in it. The mass flow rate of the injected water into the vacuum vessel and the initial wall temperature in the modified TRAC-BF1 input data were the same as those calculated in the MELCOR analysis. The maximum pressure in the vacuum vessel obtained from the modified TRAC-BF1 code is 10% higher than that from the MELCOR code, but it is still below the design pressure 0.5 MPa of the FDR-ITER vacuum vessel. The semi three-dimensional analysis using the modified TRAC-BF1 obtained that the maximum delay in opening time among the four rupture-disks is within 3 ms. The maximum pressure in the vacuum vessel does not exceed the design pressure, even if one of rupture-disks is not opened.

Thermal hydraulic characteristics during ingress of coolant and loss of vacuum events in fusion reactors

The thermal hydraulic characteristics in the vacuum vessel (VV) of a fusion reactor under an ingress of coolant event (ICE) and a loss of vacuum event (LOVA) were investigated quantitatively using preliminary experimental apparatuses. In the ICE experiments, pressure rise characteristics in the VV were clarified for experimental parameters of the wall temperature and water temperature and for cases with and without a blowdown tank. In addition, the functional performance of a blowdown tank with and without a water cooling system was examined and it was confirmed that the blowdown tank with a water cooling system is effective for suppressing the pressure rise during the ICE.In the LOVA experiments, the saturation time in the VV from vacuum to atmosphere was investigated for various breach sizes and it was found that the saturation time is in inverse proportion to the breach size. In addition, the characteristics of exchange flow through breaches were clarified for the different breach positions on the VV. It was proven from the experimental results that the exchange flow became a counter-current flow when the breach was positioned on the top of the VV and a stratified flow when it was formed on the side wall of the VV, and that the exchange flowunder the stratified flow condition was smoother than that of counter-current flow. On the basis of these results, the severest breach condition in ITER was changed from the top-break case to the side-break case.To predict with high accuracy the thermal hydraulic characteristics during ICEs and LOVAs under ITER conditions, a large scale test facility will be necessary. The current conceptual design of the combined ICE-LOVA test facility with a scaling factor of 1/1000 in comparison with the ITER volume is presented.