Haut PDF In service inspection and repair of the sodium cooled ASTRID reactor prototype

In service inspection and repair of the sodium cooled ASTRID reactor prototype

In service inspection and repair of the sodium cooled ASTRID reactor prototype

The ‘integrated readiness level’ is also discussed in this paper with respect to access within the reactor block, fluids, positioning and maintenance aspects. I. INTRODUCTION Within the framework of the future sodium-cooled fast reactor prototype called ASTRID, France has launched a large R&D program 1 on in-service inspection and repair (ISI&R) which has been identified as a difficult task to perform 2 (as sodium coolant is opaque, hot and highly chemically reactive) on the basis of experience feedback (French Phenix and Superphenix SFRs, as well as also foreign power plants). ISI&R is thus considered to be a major issue to be taken into account in order to improve the reactor’s safety (as inspection gives information on the actual reactor structure health), to consolidate its availability and to protect its associated investment.
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In Service Inspection and Repair of Sodium cooled ASTRID Prototype

In Service Inspection and Repair of Sodium cooled ASTRID Prototype

The next ASTRID prototype will benefit from these studies, as ISI&R aim at participating to plant safety, at consolidating the availability and at protecting the associated investment. At the end of the ASTRID pre-conceptual design phase (2011-2012, general options were chosen and then the conceptual phase (2013-2015) allows now to continue the improvements of the ISI&R tools: they mainly focus on the inspection of reactor block structures (immersed in sodium at about 200°C), but sodium-water (steam generator) and sodium-gas (compact) heat exchangers for power conversion system are also looked at. Using available design and specifications for potential defects to be detected, experiments are being performed using acoustic techniques, in parallel with simulations mainly done with the patented CIVA code.
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Status of the astrid sodium fast reactor project from conceptualdesign to basic design phase

Status of the astrid sodium fast reactor project from conceptualdesign to basic design phase

ESPN: Equipement Sous Pression Nucléaire (French Nuclear Rule) FBR: Fast Breeder Reactor GEN IV: Fourth Generation Reactor I&C: Instrumentation & Control ISI&R: In-Service Inspection & Repair JAEA: Japan Atomic Energy Agency MHI: MITSUBISHI Heavy Industry MFBR: Mitsubishi FBR Systems PCS: Power Conversion System P2C Confirmation Configuration Phase R&D: Research and Development

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ASTRID reactor : Design overview and main innovative options for Basic Design

ASTRID reactor : Design overview and main innovative options for Basic Design

Key Words: ASTRID - innovation - overview 1. Introduction Fast neutron reactors have a large potential as sustainable energy source. In particular, Sodium Fast Reactors (SFR) with a closed fuel cycle and potentialities for managing radioactive waste, allow improved use of natural resources and minimization of high level waste. Among the fast reactor systems, the sodium cooled reactor has the most comprehensive technological basis as result of the experience gained from decades of worldwide operation of several experimental, prototype and commercial size reactors.
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Simultaneous measurements of velocity and temperature by non-intrusive optical methods in a complex geometry : Application to the upper plenum of the sodium cooled reactor ASTRID

Simultaneous measurements of velocity and temperature by non-intrusive optical methods in a complex geometry : Application to the upper plenum of the sodium cooled reactor ASTRID

6.4 Analysis and comparison with literature 6.4.1 Experiment 1: Single negative buoyant jet at Re= 2640 and F r d = -21.2 The Reynolds number in this case is 2640 and the discharge Froude number is negative of value 21.2. The uid ow is a turbulent fountain since a dense uid is ejected into a lower density environment. This fountain starts as two symmetri- cal vortex rings and then continues to ow into the uid of the test section. The starting jet decelerates due both to the entrainment of surrounding uid and to the negative buoyancy force. So it passes by three main stages. It penetrates to its maximum height in the surrounding medium, then decreases and nally uctuates around a mean value of the maximum height of penetration. There is an interaction between the second and the third phase because of the upward and downward ows which restrict the rise of the jet again. This reduces im- mediately the initial penetration height to a smaller steady value (established ow). The dierent phases are shown in section 6.3.1. Since the Reynolds num- ber is low, the ow is expected to be laminar to transitional. Concerning the comparison with experiments found in literature, the maximum height reached by a round fountain at low Reynolds follows the scaling shown in equation (37) and that at high Reynolds is given by equation (38). However, the range of Reynolds number was not well precised and the values of n and p varied from one study to another. Table 13 summarizes the literature review and the values of n and p found in each to compare it with the value we calculate later.
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Experimental and numerical activities in support of the design of astrid sodium-gas heat exchanger

Experimental and numerical activities in support of the design of astrid sodium-gas heat exchanger

francescovitillo@hotmail.com ; xavier.jeanningros@cea.fr ; lionel.cachon@cea.fr ; chiara.galati@cea.fr ; paul.olympio@cea.fr ; sylvain.madeleine@cea.fr ; ABSTRACT In the framework of the development of the ASTRID Sodium-cooled Fast Reactor prototype, the CEA is studying the technical feasibility of adopting a Brayton power conversion cycle to eliminate the sodium- water interaction hazard. The sodium-gas (i.e. nitrogen) heat exchanger is the critical component to be designed, especially considering the fact that such a component has never been designed before for an operating nuclear power plant. Compact heat exchanger technologies are crucial to have reasonable dimensions of this component. The CEA is working on several design possibilities, especially in terms of heat transfer pattern and inlet/outlet header geometry, to find the optimal configuration. This paper aims to describe the experimental and numerical activities related to these topics. In particular, for the heat transfer patterns, traditional Printed Circuit Heat Exchangers (PCHE) as well as an innovative PCHE geometry are studied: Laser Doppler and Particle Image Velocimetry facilities are described, together with a Validation of Heat Exchange in GAS (VHEGAS) facility, exploited to acquire a wide database to be used to validate the numerical model. The CFD model validation is detailed and a first set of heat transfer and pressure drop correlations is obtained. The comparison between traditional and innovative PCHE geometries is then shown, to demonstrate that the innovative PCHE is potentially more compact than traditional PCHEs. Regarding the inlet/outlet headers, the adopted calculation methodology is described. First characterizing maldistribution in large channel bundle and secondly adopting a porous media approach to be able to correctly represent the physical phenomena in a reasonably large computational domain.
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In service inspection and repair developments for SFRs

In service inspection and repair developments for SFRs

amplitude Further under-sodium viewing tests were performed with the specific positioning device for US sensors (see Figure 28) and aimed at introducing C- scan images acquired thanks to this four degrees of freedom robot arm able to carry and precisely position high temperature ultrasonic transducers (TUSHT) under 200°C sodium. In the sodium pot, several mock- ups are positioned with different objectives: Imaging, NDT in ASTRID representative structures, sub- assembly identification and telemetry through screens. Regarding under-sodium imaging, the VISION mock-up contains engraved letters and grooves, simulating open fissures, a small triangle with sharp edges and a portion of piping. It is initially a set of images obtained by targeting this mock-up that is reconstituted and compared with those obtained in water.
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Study of isolation valve for sodium fast reactor

Study of isolation valve for sodium fast reactor

Key Words: Valve, Sodium, Isolation 1. Introduction The ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration) project is a technological integration prototype intended for the industrial-scale safety and operation demonstration of 4th generation sodium-cooled generators. This integrated design reactor includes an intermediate sodium circuit comprised of 4 secondary loops. Each DN 700 diameter loop consists of an intermediate exchanger built into the reactor tank, a sodium/gas exchanger and an electromagnetic pump. To reinforce containment, an isolation valve is placed on each of these secondary loops, along with on each hot and cold leg. The nature of the selected coolant, the operating conditions and associated dimensions, along with the operating requirements associated with these components, imply an in-depth study of the technology and materials used and, in more general terms, of the overall operation of this specific valve.
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large electro-magnetic pump conceptual design for the astrid sodium-cooled fast reactor

large electro-magnetic pump conceptual design for the astrid sodium-cooled fast reactor

Key words: Electro-Magnetic Pump, ASTRID, Sodium Fast Reactor 1. Introduction SFR is one of the 4 th –generation Sodium-cooled fast reactor (GENIV) concepts selected to secure the nuclear fuel resources and to manage radioactive waste. In the June 2006, French Government submitted CEA to design studies of ASTRID prototype as a part of sustainable management of radioactive materials and wastes [1] in collaboration with industrial partners. ASTRID will be an industrial prototype with improving safety, operability and robustness against external hazards compared with previous SFRs for aim at a GENIV safety and operation.
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Development of electro-magnetic pump for the astrid sodium-cooled fast reactor

Development of electro-magnetic pump for the astrid sodium-cooled fast reactor

Saint-Paul lez Durance Cedex, France Tel: +81-45-770-2410, Fax: +81-45-770-2317, Email of main author: tetu.suzuki@toshiba.co.jp Abstract – In the framework of the SFR (Sodium Fast Reactor) prototype called ASTRID (Advance Sodium Technological Reactor for Industrial Demonstration), the large capacity Electro- Magnetic Pumps (EMP) as main circulating pumps on the intermediate sodium circuits is applied instead of mechanical pumps by CEA. The use of EMP has several decisive technological merits compared with mechanical pump in the reactor design, operation and maintenance. Nevertheless, some theoretical and technological developments have to be carried out in order to validate the design tools which take Magneto Hydro Dynamic (MHD) phenomena into account and the applicability of the EMP to the steady state and transient operating conditions of ASTRID. For their developments, a collaboration agreement between the CEA and TOSHIBA Corporation came into force to carry out a joint work program on the EMP for ASTRID design and development. CEA carried out the theoretical analysis, and the EMP experimental model is constructed by CEA to support these theoretical developments. This model consists of a middle-size annular EMP for the liquid metal sodium. The various testing program using this model has been started in 2016.And, TOSHIBA carried out the examination of design specification for ASTRID, an electromagnetic design, a structural design and various analyses. The structure design has been examined the placement of the sodium boundary and the withstand pressure, etc. And, if the thicknesses of the structure increase for withstand pressure, the pump efficiency falls because the loss of the electromagnetic force increases. Therefore the balance of withstand pressure and the efficiency has been considered by an electromagnetism design. This paper describes the design studies and experimental activities for the EMP development within the framework of the CEA- TOSHIBA collaborations.
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Inspection specifications leading to extended ASTRID Design rules

Inspection specifications leading to extended ASTRID Design rules

Abstract. CEA initiated a study in 2008 to improve the design rules of fast reactors with French utilities (EDF), French designers (AREVA) and non-destructive examination (NDE) specialists (Aix Marseille University), focusing on the specific issue of in-service inspection (ISI). Thus, at the end of 2012, the RCC-MRx specifications for NDEs was enlarged, orienting design and manufacturing choices and rules to account for future in service inspection. Due to the complexity of the links between design, materials, access, inspection techniques and tools, these rules cannot be considered as strict instructions, but rather as leading to fruitful dialogue between designers and inspectors. The links between in-service inspection and manufacturing processes and specifications are now being explored in further detail. This article describes the approach and R&D program in support of this specific work. This initiative should lead to better connections and compromise between design work, material specifications and in-service inspection, called RC-CND rules (Design rules taking into account NDE requirements).
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Numerical and Experimental Validation of the Prototype of A Bio-Inspired Piping Inspection Robot

Numerical and Experimental Validation of the Prototype of A Bio-Inspired Piping Inspection Robot

80% 1.8 0.46 10800 (Clockwise) 4.1. Estimation of Currents and Forces of the Motors From Experiments To attain static phases, the threshold limits T U and T L were set as 1.5 V for the clamping/central retraction phases and 0.42 V for the de-clamping/central elongation phases. These values were set higher than the results of static force models because the weight of umbilicus and electromechanical factors were not considered in the numerical analysis. As the central actuator encounters lesser static forces as they have no masses attached to them, the same threshold limits were retained to attain larger displacements. The threshold values are also essential while operating under real-time conditions, where the exact diameter of pipelines are sometimes unknown. Lower values would have been more acceptable but the consumption of voltage by the motor at its start disrupts the calculation of generated force. This is due to the velocity profile used. In addition, the time of conversion by the ADC to the BB black is either quicker or lags behind. Thus, at the start of experiments, higher starting torque/current was usually observed. In such scenarios, the threshold values played an essential role for effective functioning of the motors and to cease them at the right moment. When threshold values were reached, the torque reached its stall point, which stopped the motor and the PWM was set to idle phase. The digital voltage values generated by the ADC of the BB were extracted and the results were plotted with the help of MATLAB. With the voltage values obtained during each step of the experiment, the current induced on the actuators was estimated by the equation:
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Neutron flux monitoring with in-vessel fission chambers to detect an inadvertent control rod withdrawal in a sodium-cooled fast reactor

Neutron flux monitoring with in-vessel fission chambers to detect an inadvertent control rod withdrawal in a sodium-cooled fast reactor

The high neutron flux inside the core can deplete the fissile layer of fission chambers quickly. This harsh environment and compact design of SFR cores does not permit an in-core detector installation, however, in-vessel installation outside the core is feasible. The current possible ex-core regions for detector installation are: the lower part of the above-core structure (ACS), the region beyond the radial neutron shield and the core support structure (CSS) [1]. One of the main drawbacks associated with ACS is that the detector signal is cut off during fuel handling operations as the above core structure moves along with the rotatable plugs. This explains why SFRs need a combination of detectors at different locations for a consolidated monitoring system. In this paper, the lateral installation of fission chambers is studied. The de- tectors installed in such radial locations might need a neutron channel in the radial neutron shield assemblies to allow for a sufficiently intense neutron flux to reach the fission chambers. Such arrangement has previously been used in Superphenix as well. The neutron channel design is discussed in more detail in Section 3.2. Throughout this paper, fission chambers and detectors are used synonymously.
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Analysis of the Gas Entrainment by Vortex in the Upper Plenum of the ASTRID Reactor Mock-up

Analysis of the Gas Entrainment by Vortex in the Upper Plenum of the ASTRID Reactor Mock-up

The 4th generation sodium fast neutron reactor ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration) is developed by the CEA and several industrial partners. The design is in progress especially for the vessel and the equipment. From the state of art of the former sodium reactors PHENIX and SUPER-PHENIX and the EFR program (European Fast Reactor) the global and the local thermal-hydraulic issues have been listed. But the results of those previous projects are not sufficient for the current ASTRID development. New needs for experimental means were identified, especially for both validation of numerical codes and specific studies. In this way, a thermal-hydraulic loop, the PLATEAU facility, was developed and built at the CEA Cadarache research center in 2012. Since experiments with sodium are complex to carry out due to the harmfulness of this liquid metal, the tests are led in models using water as simulant fluid. Different mock-ups can be connected to this loop to study the different issues at various reactor conditions. The first model connected to the PLATEAU facility is MICAS, mock-up of the hot plenum. This one is dedicated to study the flow regime, both for code validation and engineering design development. It was designed at 1/6 scale and was built in transparent polymer to carry out optical measurements as laser velocimetry and fast imaging.
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Durability of concrete structures: prevention, evaluation, inspection, repair and prediction

Durability of concrete structures: prevention, evaluation, inspection, repair and prediction

Vous avez des questions? Nous pouvons vous aider. Pour communiquer directement avec un auteur, consultez la première page de la revue dans laquelle son article a été publié afin de trouver ses coordonnées. Si vous n’arrivez pas à les repérer, communiquez avec nous à PublicationsArchive-ArchivesPublications@nrc-cnrc.gc.ca. Questions? Contact the NRC Publications Archive team at

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Application of the technology neutral framework to sodium cooled fast reactors

Application of the technology neutral framework to sodium cooled fast reactors

Although core disruptive accidents (CDAs) were not considered as part of the design basis for the CRBR, a large amount of regulatory attention was given to these a[r]

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Flow analysis in the upper plenum of the micas model in support of the astrid reactor program

Flow analysis in the upper plenum of the micas model in support of the astrid reactor program

temperature, to the various models. The first mockup, MICAS, commissioned in 2015, represents the upper plenum of the ASTRID reactor at a 1/6 scale. It aims at studying the main different thermohydraulic issues such as the gas entrainment at the free surface [4], the flow around the Intermediate Heat Exchangers (IHX), the behavior of the jet going out from the core, and the flow crossing the Above Core Structure (ACS). During the past projects in the 90s, various models were built to study those issues. A lot of experiments were carried out in the 1/8 scale mockup COLCHIX4 [5] regarding the free surface state, the IHXs supply, and the flow crossing the ACS. The velocity fields around the IHX were measured using propellant sensor probes. This technique, wildly used in the hydrogeological domain to measure the river flow, is very accurate, but intrusive. Regarding the flow in the ACS, it was measured by analyzing the pressure losses inside the sheath tubes hanging the control rods and the instrumentation of the core. The flow versus the pressure losses was calibrated in a dedicated setup prior to the COLCHIX experiments. This method is quite heavy to implement because it requires specific tests to build a pressure losses versus flow correlation. Spaccapaniccia and al. [6] presented some Particle Imaging Velocity (PIV)
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MHD interaction in an Electromagnetic Pump for high flow rate loop of ASTRID Sodium Fast Reactor secondary circuit -performances

MHD interaction in an Electromagnetic Pump for high flow rate loop of ASTRID Sodium Fast Reactor secondary circuit -performances

average electromagnetic force density Conclusions We have analysed the performances of a large size electromagnetic pump, the so-called ASTRID pump. First was determinate the optimal frequency for supply currents by the use of uncouple computation between electromagnetism and fluid flow. Then, needed current density was found to reach the wished operating point. In a second time, a strong coupling computation was done between electromagnetism and fluid flow. It has showed that bloc pumping hypothesis lowered the overall performances of the pump in comparison to real flow computation.
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A comprehensive numerical modelling of a fission chamber to be operated over a wide dynamics in the vessel of a sodium-cooled reactor

A comprehensive numerical modelling of a fission chamber to be operated over a wide dynamics in the vessel of a sodium-cooled reactor

space, r a the anode radius, the mean current is then: I = 2πhq e r a n e (r a )v e (r a ) (2) A correction is then applied to account for the columnar recombination [ 24 ]. The saturation curves that are obtained for CFHT1.5 and CFHT1.0, with pressures of 1, 2 and 4 bar, are shown on Fig. 1 . A fission chamber is usually operated with near the saturation point, defined as the inflection point of the saturation curve. It corresponds to the situation in which all the created charges are eventually collected at the electrodes. The saturation point is found to be identical for P and BCC locations. The saturation plateau, within which the bias voltage can vary without a significant change of the mean current, is comfortably large in all cases. For instance, for the CFHT1.5 at 1 bar, the saturation point is given by ∆V sat = 450 V , but the current differs from I(∆V sat ) by less than 5% for 130 ≤ ∆V ≤ 760.
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Optimization of a sodium-cooled fast reactor operation with a gas power conversion system during a loss of off-site power

Optimization of a sodium-cooled fast reactor operation with a gas power conversion system during a loss of off-site power

The main issue following the reactor scram is to remove the decay heat. A standard solution is to rely- on passive DRACS like RRB systems described in the section 1.2 and modeled in the section 2.2. In this paper, an alternative procedure (preliminary studied, Bertrand et al., 2016b) is proposed and substitutes the RRB with the gas PCS to remove the decay heat. After the IE and the first consequences, several actions are required to keep an efficient heat removal with the water heat sink via the gas PCS and the secondary circuit. As for the standard procedure, 30 seconds after the IE, emergency electrical supplies are available to ensure a primary mass flow rate of 25% of its nominal value during the whole transient. Contrarily to the procedure with RRB, 30 seconds after the IE, the secondary pumps are powered by backup systems and the secondary flow rate is maintained at 25% of its nominal value during the whole transient. The feasibility study and the sizing of the emergency supplies associated to the primary and secondary pumps were realized for the European Fast Reactor (EFR, 1999). Few sec- onds after the IE, a backup water flow rate is provided as well in the coolers to ensure the heat removal. Furthermore, the water flow rate regulates the gas inlet temperatures of the two compressors to the setpoint of 27 °C, in order to avoid damages on the compressors and to keep an efficient heat sink. The power provided by emergency supply limits the maximal water flow rate to 40% of its nominal value. For this procedure, the PCS is operated in two different ways. First, it cools the reactor down to reach the cold shutdown state and then, it maintains this safe state. In the first step, the gas flow rate must be strong enough to remove the decay heat and the energy previously accumulated in the sodium; once the cold shutdown state is reached, only the decay heat must be removed to maintain the reactor in a safe state. For this purpose, a regulation of the TM rotation speed, by the BP1 valve, ensures a gas flow. The target of this regulation is time dependent and described by the two parameters (Figure 3):
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