Pressure Suppression System Research Articles

Experimental qualification of the ITER pressure suppression system by means of a large scale facility: assessment of similitude analysis

The paper deals with the assessment of the pressure suppression system of the international thermonuclear experimental reactor (ITER). An extensive experimental research programme performed in a small-scale test facility analysed the steam direct condensation at sub-atmospheric pressure (about 400 tests). A similitude analysis was elaborated for extrapolating the results to the ITER full scale system. The derived scale laws were applied to analyse the performances of the pressure suppression system in order to manage a large loss of coolant accident (LOCA) scenario postulated in the vacuum vessel of ITER. Experimental tests, performed in a large-scale test facility built at the University of Pisa (geometrical scale about 1 and 1/10 power scale), analysing the first part of the large LOCA event, assessed the scale laws emphasizing some discrepancies determined in the extrapolation of reduced scale results to the actual system. Introducing a justified correction factor, the presented similitude analysis assured a suitable and reliable temperature and pressure increase prediction of full-scale systems, even in scaled configuration affected by relevant thermal stratification.

Concept of operation of the remote handling system for the ITER vacuum vessel pressure suppression system

A concept of the remote handling (RH) tools to maintain the components of the ITER Vacuum Vessel Pressure Suppression System (VVPSS) is developing. These RH tools are designed to remove and replace the rupture disk assembly and the bleed line valve assembly, to clean the relief lines, and to perform any associated operations to maintain them. The RH tools work together with the Neutral Beam Remote Handling System (NBRHS). The main functions of the RH tools are disconnection and connection of the bolted flanges, keeping the static confinement during the RH operations, limiting the spread of contamination, and vacuum or water jet cleaning of the contamination inside the relief line. The RH interfaces in the components are defined during the design process of the RH tools in collaboration with the VVPSS component designers. An RH sequence study is being carried out to demonstrate the feasibility of the maintenance operations within the constrained NB cell environment, and the capability of the NBRHS.

Hydrogen explosion mitigation in DEMO vacuum vessel pressure suppression system using passive recombiners

An important issue for the Fusion Reactors is the hydrogen explosion hazard assessment. For this reason, in the Work Package of Safety Analyses and Environment (WPSAE) of the EUROFusion consortium, a task was established to identify the potential for hydrogen production in DEMO vessel in accident situations and to investigate the possible solutions which can prevent the risk of hydrogen and dust explosion. The hydrogen in the Vacuum Vessel (VV) can lead to combustion progressing in deflagration and detonation. Besides, the tungsten dust could enhance the effects of H2 reaction as demonstrated by experiments, however, in this first set of accidents, only the H2 generated by the chemical reaction is accounted for.For these reasons, the solutions to avoid or limit the H2 production and/or to control safely the risk of H2 accumulation and explosion need to be screened starting from the solutions adopted in the fission technology, and later on, adapted and assessed in the Pressure Suppression System (PSS) foreseen in DEMO. One promising idea seems to install passive catalytic recombiners in the PSS.The aim of this paper is firstly to set a range of operating parameters for the PAR functioning and intervention suitable in DEMO PSS configuration. Secondly to verify if the present criteria and parameters work efficiently in mitigating the H2 accumulation a PSS in accidental conditions. To test the theoretical effectiveness of the PAR intervention, two reference accidents have been analyzed: an in-VV Loss Of Coolant Accident (LOCA) in the frame of the Design Basis Accident (DBA) and a Loss Of Flow Accident (LOFA) without active plasma shutdown in the context of the Beyond Design Basis Accidents (BDBA).The results seem promising if the performances of the PAR will be confirmed in sub atmospheric and saturated conditions.

Full scale test facility for operability verification of tokamak pressure suppression system at sub-atmospheric conditions

The reliable and safe operation of tokamak fusion reactors is mainly based on radioactive contaminants confinement, achieved preserving the integrity of robust barriers, among which the Vacuum Vessel (VV) performs an essential role. The VV protection against postulated over pressurization events is guaranteed by the safety functions implemented in the Pressure Suppression System (PSS), in which steam from VV is discharged and Direct Contact Condensation (DCC) occurs at sub-atmospheric conditions. Aiming to verify the VVPSS performance of ITER machine (i.e. steam condensation efficiency at different DCC regimes with/without non-condensable), in steady state and transient conditions relevant for fusion reactor postulated events, and contribute to fill the knowledge gap regarding sub-atmospheric condensation, an extensive experimental research activity is ongoing with the Large and Full Scale Test Facility (LSTF and FSTF): designed, assembled, instrumented and commissioned at Laboratory B. Guerrini at DICI University of Pisa. This facility is mainly composed of a Steam Generator (up to 500 g/s, about 1.8 MW), three different steam lines for wide mass flowrate regulation, two air lines for mixed condensation characterization, Experimental Test Tank (92 m3, where steam condensation is investigated adopting two different spargers, scaled and full scale with 100 and 1000 holes, respectively), Auxiliary Tank (10 m3 for plant conditioning), Vacuum Pump, Chiller and Water Storage Tank (15 m3). Two Steam Storage Tanks (15 m3 each) allow to accumulate and provide the foreseen steam peak flowrate value (of 5 kg/s) in the postulated accident. FSTF is highly instrumented (81 thermo-resistances, 28 thermocouples, 35 pressure transmitters, 18 strain gauges and 4 vortex flow meters) and fully remotely controlled (64 on-off and control valves) by a dedicated Data Acquisition and Control System (DACS). Six video cameras and 4 led lights assure steam injection video-recording, for detailed characterization of condensation regimes. Preliminary results are shown for 50 g/s of steam injected through the scaled sparger into water pool at 30 °C with cover gas at 0.1 bar absolute, providing a condensation efficiency of about 100% and verifying instable chugging condensation regime.

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.

Large scale experimental facility for performance assessment of the vacuum vessel pressure suppression system of ITER

The nuclear fusion reactor ITER (International Thermonuclear Experimental Reactor) foresees a Pressure Suppression System (PSS) in order to manage a Loss of Coolant Accident (LOCA) in the Vacuum Vessel (VV). A LOCA scenario is postulated to occur in the cooling systems of plasma facing components. The Vacuum Vessel Pressure Suppression System (VVPSS) has a key safety function because a large internal pressure in the VV could lead to a breach of the primary confinement barrier.Steam condensation would occur in ITER at sub-atmospheric pressure, differently than the applications in nuclear fission reactors. Steam direct contact condensation (DCC) in sub-atmospheric conditions have never been implemented before, therefore an experimental assessment of the VVPSS is necessary to simulate this phenomenon in ITER's thermal-hydraulic conditions.Under the financial support of ITER Organization, an experimental program was performed at the Department of Civil and Industrial Engineering (DICI) of the University of Pisa in the past 3 years, using a small-scale experimental rig (1/22 scale factor).In order to assess the scale laws elaborated by means of the experimental results obtained in this small scale facility, a Large Scale facility was built at University of Pisa. This facility is almost a full scale mock-up of ITER's VVPSS from the geometric point of view (1/1.09 scale factor) and it is 1/10 scale factor from the steam mass flow rate point of view.This paper illustrates the results of the first test campaign. A 1/10 scale sparger system SPS-A (having a diameter equal to DN125 and 100 holes) was used with a maximum steam mass flow rate of 0.5 kg/s. This experimental configuration represents real operating conditions on the basis of the results of elaborated similitude analysis. The preliminary analysis of the experimental results confirms the agreement in terms of condensation regimes between the two reduced scale (1/22 and 1/10 scale) as foreseen by the elaborated scale laws.

Direct steam condensation at sub-atmospheric pressure: experimental tests and similitude analysis

This paper deals with experimental and theoretical analyses of the steam direct condensation at sub-atmospheric pressure. These operation conditions occur in the Vacuum Vessel Pressure Suppression System (VVPSS) of ITER Nuclear Fusion Reactor. This safety system permits to manage the Ingress of Coolant Event (ICE cat. IV) accident condition, postulated to occur in the ITER Vacuum Vessel (VV) as a consequence of the rupture of the shielding blanket cooling piping or the rupture of the divertor cassette piping.Experimental tests of steam direct condensation at sub-atmospheric pressures in a 1:22 scale facility were performed at the laboratory Guerrini of DICI-University of Pisa. Presently a full-scale facility is being building at DICI. This paper illustrates the similitude analysis elaborated to scale up the experimental results obtained in the reduced scale facility. The main variables which influence the steam direct condensation were analyzed and the relative scale laws determined. CFD numerical simulations of direct steam condensation in the actual geometry and in a vessel of equal volume permitted to verify some of the assumptions on which the scale laws are based. Scaling studies were first performed in order to calculate the transient of steam mass flow rate occurring during the ICE IV event in the scale apparatus, starting from the thermal hydraulic studies performed at ITER and successfully compared with the experimental results carried out in scaled apparatus. This confirmed the suitability of scale laws and in the same time the capability of the VVPSS to condense the injected steam at sub-atmospheric pressure, matching the safety goal to reduce the system pressurization.

DEMO Divertor preliminary safety assessment

European DEMOnstration plant is currently finalizing a pre-conceptual design phase. A carefully selected set of safety analyses are contextually being performed on DEMO systems in order to assess its integrated performance and capability to achieve expected targets while keeping in a safe operation domain. A key in-vessel system is the Divertor, in charge of exhausting a major part of the plasma ions thermal power in a region far from plasma core to control plasma pollution. To accomplish this mission a modular design approach was adopted based on two main sub-systems: the divertor cassette and plasma facing units (PFUs). Each cassette supports two vertical plasma facing units, one inboard and one outboard, intercepting the main part of the particle loads from the plasma. An assembly of 3 of such divertor cassettes and related PFUs is then foreseen for each of DEMO 16 sectors, hence resulting into 48 cassettes. Two independent heat transfer systems for cassettes and plasma facing units respectively provide heat sink function and connect the Divertor to DEMO Balance of Plant.An overview of the preliminary safety analyses performed on such system is here proposed. At first, a safety-perspective overview of current design is proposed. In particular, moving from the divertor system functions, specific component failure modes and related possible consequences, a set of relevant postulated initiating events are identified.Then, aiming at investigating whether accident consequences remain within accepted safe domain, two accidents from this set were selected to be analysed deterministically by means of MELCOR code. In particular, an in-vessel Loss of Coolant Accident initiated by a large rupture of the divertor PFU cooling pipes and a Loss of Flow Accident (LOFA) initiated by heat transfer system (HTS) pump seizure have been analysed. The nodalization includes divertor PFU and related HTS among other in-vessel components and safety devices (e.g., Vacuum Vessel Pressure Suppression System) devoted to Vacuum Vessel pressurization transient control or Fast Plasma Shutdown System. Parametric analyses were also performed to investigate possible impact of the design solutions (e.g., implementation of isolation valves in Divertor loop) on the accident consequences.

VVER-440/V213 long-term containment pressurization during severe accident

The contribution deals with the performance of a new containment heat removal system that is intended to be installed in near future at Dukovany NPP in Czech Republic. This NPP is equipped with 4 units of VVER-440/V213 reactors. During the last two decades these units were systematically backfitted with new safety systems in order to cope with consequences of severe accidents. These safety systems were designed especially for the 4th level of defence in depth and thus they are – as much as possible – independent from the original plant systems. Namely, installation of a new depressurisation line is currently under preparation in order to avoid high-pressure core melting scenarios. In-vessel corium retention via external reactor vessel cooling (“reactor cavity flooding”) was adopted as a cornerstone strategy of severe accident management for all VVER-440/V213 reactors operated in central Europe. Regarding the hydrogen issue, large capacity passive autocatalytic recombiners were installed in containment. The prevention of containment pressurisation in the case of loss of ultimate heat sink represents the most important point of remaining safety issues. At first, special features of VVER-440/V213 containment equipped with bubbler condenser pressure suppression system are briefly described in presented paper. Than the mechanism of containment pressurisation due to steaming as a consequence of in-vessel retention and possible solutions for heat removal from the containment are discussed. The final design chosen at Dukovany NPP is based on the heat exchanger and turbo pump located inside the containment that are cooled and driven by external cooling water. External cooling circuit will be connected during the accident via special inlet/outlet nozzles installed at containment boundary. Main advantage of such containment cooling system is full independence on other plant systems (including essential service water), the fact that radioactive coolant is not taken beyond the containment boundary and all active components of the system (mobile pump, isolation valves) are located outside the containment and, thus, are easily accessible. The performance of this containment heat removal system during severe accidents were analysed using the ASTEC code and the most important results are presented. From the obtained results it follows that there is sufficient time margin for assembling the external cooling loop. Thus, the containment cooling can be assured long before the containment ultimate pressure is reached or before the coolant level in containment sump drops below critical level, when the external reactor cooling can be lost. After start of the heat removal system, the system is capable to prevent further containment pressurisation, increase the coolant level in containment sump and decrease the pressure to atmospheric or even sub-atmospheric level.

Safety cases for design-basis accidents in LWRs featuring passive systems

This paper presents results from a series of integral tests performed at Framatome’s INKA test facility in Karlstein (Germany) which simulates a KERENA boiling water reactor (BWR). The scope of the test series was on the behaviour of and interaction between the different passive systems and components under the conditions of extended loss of alternating power (ELAP). These SBO-like conditions were aggravated in three out of four tests by parallel LOCA (Loss of Coolant Accident). The scenarios of all four tests fully correspond to Design Basic Conditions (DBC). They were: main steam line break, feed water line break, reactor pressure vessel (RPV) bottom leak and station blackout (SBO, non-LOCA).In the tests, the passive systems integrated in KERENA and INKA, respectively, have fulfilled their design functions fully satisfactorily and as follows:The Passive Pressure Pulse Transmitter (PPPT) triggered the RPV depressurization without delay. The Emergency Condenser (EC) system removed decay heat along with stored energy from the RPV to the containment. The Containment Cooling Condenser (CCC) system forwarded said power to a heat sink outside of the containment. The passive containment pressure suppression system kept the containment pressure within the design range, partially displacing surplus thermal energy from the drywell to the wetwell, in particular in the early phases after occurrence of LOCA. The passive core flooding system replenished the coolant inventory of the RPV thereby ensuring water levels in the RPV which are fully sufficient for core cooling.Moreover, the systems have cooperated as anticipated by the designers, quietly and without perturbing each other.Hence the test results, which are reported and discussed more in detail within this paper, soundly confirm the underlying design and its passive features.Said tests were carried out as a part of the joint research project EASY (Evidence of Design Basis Accidents Mitigation solely with passive safety Systems), the overarching objective of which was the development and validation of the code system AC2 of GRS (Gesellschaft für Anlagen- und Reaktorsicherheit gGmbH).

Pressure suppression system influence on vacuum vessel thermal-hydraulics and on source term mobilization during a multiple first Wall – Blanket pipe break

Being the Vacuum Vessel Pressure Suppression System (VVPSS) one of the most important passive safety systems to be foreseen in DEMO plant, design and integration challenges have to be faced to ensure that best performance within safety requirements are always achieved. In this framework, parametric safety analyses have been performed to support VVPSS design activities; in particular to determine the minimum flow area required by the suppression system pipework to limit the vacuum vessel pressure below the limit imposed as a requirement by design. The selected Postulated Initiating Event (PIE) is a double-ended guillotine break in the Primary Heat Transfer System (PHTS) feeding pipe of the Breeding Zone (BZ) during plasma activity. Coolant is discharged inside the VV upper port volume and an unmitigated disruption occurs, causing further breaks in the first wall (FW) cooling channels. Considering that limiters could be introduced in the future design of the EU DEMO reactor to prevent damages to plasma-facing components, the same parametric study has been performed considering limiters accident mitigation effects. Because discharge flow area toward the suppression pool could also affect the source terms mobilization, the transport of radioactive products (e.g. tritium, tungsten dust and activated corrosion products) has been simulated with the fusion version of the MELCOR code (ver. 1.8.6).

Computational Fluid Dynamics Thermal Analysis of ITER Pressure Suppression Tanks

Abstract The aim of the paper is to present the results of the investigation of the thermal conditions (temperature distribution, heat losses) in the support system of the vapor suppression tank (VST) of the vacuum vessel pressure suppression system (VVPSS), a safety important system of ITER fusion reactor, protecting the vacuum vessel (VV) against overpressures. The VVPSS includes four VSTs of identical volume and installed as two stacked assemblies. The study focuses on the optimization of the design of the thermal insulation at the bottom of the VSTs, interfacing with the basement and also on the identification of the thermal loads at the interface between the tank support and the tank pressure boundary. A computational fluid dynamics (CFD) analysis of the VST has been performed for four different insulation configurations and considering both steady-state and transient loads following accidental conditions. The results of the analysis are used to provide recommendation on the optimum configuration of the thermal insulation. Measures for minimization of the thermal gradient in the critical area of the joint between the tank hemispherical head and support skirt to limit the thermal fatigue on the welds are also suggested.

3D transient CFD simulation of an in-vessel loss-of-coolant accident in the EU DEMO fusion reactor

Accidental transients in fusion reactors, such as the in-vessel loss-of-coolant accident (LOCA) considered here, are generally analyzed using system-level codes. However, because of their lumped nature, such tools cannot predict local pressure and temperature values, which are instead of interest to assess the integrity of the vacuum vessel (VV) structures. It is then fundamental to prove that the system-level tools’ predictions are at least conservative. In this work, we analyze the helium flow inside the VV following a LOCA in the helium-cooled blanket of the EU DEMO, using a 3D transient computational fluid-dynamics (CFD) model implemented in the commercial STAR-CCM+ code. In view of the large pressure ratio, a hypersonic flow regime (Ma > 5) develops, with the formation of shock fronts requiring a high time and space resolution. The model is applied to compute the evolution of the pressure distribution inside the VV and then to compare it with the information provided by the system-level GETTHEM model. The predictions by the two models of the intervention time of the VV pressure suppression system are compared, showing that the 0D model is conservative.

Scaling analysis of the pressure suppression containment test facility for the small pressurized water reactor

The small PWR has been paid more and more attention due to its diversity of application and flexibility in the site selection. However, the large core power density, the small containment space and the rapid accident progress characteristics make it difficult to control the containment pressure like the traditional PWR during the LOCA. The pressure suppression system has been used by the BWR since the early design, which is a suitable technique that can be applied to the small PWR. Since the configuration and operating conditions are different from the BWR, the pressure suppression system should be redesigned for the small PWR. Conducting the experiments on the scale down test facility is a good choice to reproduce the prototypical phenomena in the test facility, which is both economical and reasonable. A systematic scaling method referring to the H2TS method was proposed to determine the geometrical and thermohydraulic parameters of the pressure suppression containment response test facility for the small PWR conceptual design. The containment and the pressure suppression system related thermohydraulic phenomena were analyzed with top-down and bottom-up scaling methods. A set of the scaling criteria were obtained, through which the main parameters of the test facility can be determined.

Analysis of the effects of primary heat transfer system isolation valves in case of in-vessel loss-of-coolant accidents in the EU DEMO

Abstract As DEMO is the first European device planned to produce electricity from fusion, the volume of its Primary Heat Transfer Systems (PHTS) will be consistently larger if compared to present or next-generation tokamaks such as ITER. The consequences of an in-vessel Loss-Of-Coolant Accident (LOCA) would then be more important, and within the EUROfusion Consortium different possible mitigation measures are being investigated. Among these, the introduction of Isolation Valves (IsoVs) on the main cooling loops of the Breeding Blanket is being considered, in view of the many benefits they would introduce, not only in case of accidents, but also e.g. during the maintenance of the in-vessel components. Fast-closing IsoVs on the PHTS would help in relaxing not only the requirements of the VV pressure suppression system (VVPSS) design, but also those related to the expansion volumes that shall accommodate the contaminated coolant discharged from the PHTS after a LOCA. In the present work, the GETTHEM code, the system-level thermal-hydraulic model developed for the EU DEMO at Politecnico di Torino, is used to assess the beneficial effects of the introduction of the IsoVs. The effects of the actuation time of the IsoVs and of their location are parametrically investigated, considering both water and helium as PHTS coolants, with particular reference to the reduction of the in-vessel space-averaged pressure and of the suppression system size.

Direct condensation of steam in a water tank at sub-atmospheric pressures

In the nuclear fusion reactor, ITER, the Loss of Coolant Accident (LOCA) in the Vacuum Vessel has to be managed with pressure suppression systems working at sub-atmospheric pressure. The operating conditions differ considerably from those experienced in the fission nuclear power plants such as BWR. The direct condensation at sub-atmospheric conditions is not sufficiently known, therefore, the effectiveness of systems operating at these particular conditions have to be investigated experimentally.A research program is being carried out at the University of Pisa, funded by ITER, in order to study the steam direct condensation for nuclear fusion reactor conditions. For this purpose, an experimental test facility was designed and built and an extended experimental program was performed. Video cameras were used to visualize the steam condensation at different mass flow rates.This paper deals with the elaboration of images of the steam jet flowing from a hole in the water. The steam condensation regimes depend on three governing parameters: downstream exit pressure, water temperature and steam mass flow rate per hole. Moreover, the condensation regimes are characterized by different shapes of steam jet.The image analysis permitted to determine the heat transfer coefficient in the stable condensation regime at sub-atmospheric conditions. The results obtained are compared with those correspondent at steam condensation at atmospheric pressure, emphasizing the great importance of the downstream exit pressure and the subcooling on the steam condensation.

Numerical analysis of sub-atmospheric steam condensation in suppression tank with SIMMER IV code

Abstract One of the key safety components for nuclear fusion plants is the suppression tank, which is designed to protect the Vacuum Vessel (VV) against accidental pressurization events, e.g. Loss Of Coolant Accident (LOCA). In this framework the attention is focused on the Vacuum Vessel Pressure Suppression System (VVPSS), made of water tanks in which the pure steam, or eventually mixed with incondensable gases, is injected and consequently the overpressure is dumped profiting of Direct Contact Condensation (DCC). The design constraints of fusion reactor dictate that the pressure resulting (long-term) from any accidental or baking condition should be always kept lower than 0.15 MPa. The study of the phenomena evolving during DCC in LOCA conditions is the major novelty, especially in consideration of the lack of similar studies in the available literature. In this context, a wide series of experimental tests was carried out at Pisa University (UNIPI), Department of Civil and Industrial Engineering (DICI), in a Small Scale Test Facility (SSTF), designed and instrumented for investigating DCC at sub-atmospheric pressure, by varying water pool temperature, pressure and steam mass flow rate. The adoption and assessment of suitable numerical codes, to reliably simulate such a cutting-edge multiphase multicomponent scenario, have a crucial role for contributing to the phenomena understanding and for possible safety analysis of full-scale components. On this basis, a preliminary evaluation of the cartesian three-dimensional SIMMER IV code capabilities in simulating DCC at sub-atmospheric conditions was carried out, taking as reference one UNIPI test. SIMMER IV code was able to set up precise initial low-pressure boundary conditions and simulate superheated steam condensation in subcooled water pool, with condensation efficiency comparable to the experimental one. Moreover, SIMMER IV code predicted a longitudinal steam plume dimension and injected steam velocity consistent with experimental data.

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.

Improved design and experimental research on cryostat pressure suppression system of EAST

As one of the most important safety systems for EAST, the cryostat pressure suppression system (CPSS) is designed to protect the cryostat of EAST under overpressure condition, and ensure the safety and the tightness during EAST operation. The structure and workflow of CPSS are presented detailedly in this paper. The rupture disc subassembly and pressure monitoring module, which are key components for CPSS, are upgraded aiming at the potential safety issues. An experimental test facility was built to verify the reliability of improved CPSS. And the experimental results indicate that the new system meets the design and operation requirements.

ITER Tokamak Cooling Water System Design Status

The tokamak cooling water system (TCWS) is the primary cooling system of the ITER tokamak machine, providing cooling water to the vacuum vessel and in-vacuum vessel components. In addition, it provides water and gas baking to its clients as well as drying them prior to maintenance activities. It is a Safety Important System that is subject to the French Order on Nuclear Pressure Equipment. The TCWS design has been modified significantly since the preliminary design. Such changes required the approval of the French Nuclear Safety Authority (ASN). The first main modification was to relocate the main equipment of the cooling loop for in-vessel components from level L4 to level L3 of the Tokamak Building in order to improve the overall building shielding capacities, and the second dealt with the ability of the TCWS to significantly reduce the mass and energy released in case of a loss-of-coolant accident (LOCA). This second improvement was required to allow a strong improvement of the vacuum vessel pressure suppression system (VVPSS), i.e., the system that prevents overpressure in the vacuum vessel in case of a LOCA. The green light for the two aforementioned modifications of the TCWS (together with the new VVPSS configuration) has been recently obtained from the ASN. Most recently, the three major subsystems of the TCWS needed for first plasma operations (to start the end of 2025) have passed the final design review process. The final design has been approved and procurement/manufacturing has started. In addition, the final design of all piping not functionally required for first plasma operation but nevertheless needed due to installation constraints has been reviewed and approved as well. A few kilometers of nuclear-grade piping have been delivered to the ITER Organization (IO) together with the main drain tanks of the system. All the other components are expected to be received on site starting May 2021. This paper gives a detailed description and status of the main TCWS subsystems needed for first plasma either because they are functionally required or because of installation constraints. The major modifications highlighted above as well as all the improvements accomplished in the final design are also detailed, together with a status on procurement and construction activities.