To comply with the GEN IV objectives, the design of the 600 MWe Advanced Sodium Technological Reactor for Industrial Demonstration (ASTRID) includes enhanced safety compared to current generation II or III reactors. One of the most important safety aspects is the ability to shut down the reactor in any condition.

In the early stage of the ASTRID conceptual design phase (2013), it was decided to support a robust safety demonstration by complementing ASTRID low void fraction (CFV) design core natural behavior with two types of Complementary Safety Devices dedicated to core damage prevention that would passively shutdown the reactor. The first type is based on the exploitation of the Curie point of the electromagnetic delatch system of one of the diversified reactor shutdown systems, to address unprotected loss of heat sink transients. The second type is a hydraulically suspended absorber rod subassembly, called RBH, dedicated to unprotected loss of flow (ULOF) transients.

Concepts of the latter type are innovative for a French sodium-cooled fast reactor. Their operating principle is simple. Under normal operation, the absorber rod is hydraulically suspended above the core by the upward flow of the sodium coolant. Should an ULOF event and the associated drop in flow rate occur, this upward force would become insufficient, thus allowing the absorber material insertion into the active core region by gravity.

Based on the ASTRID functional specifications and safety requirements, the structured process of value engineering that led to the selected options/designs of ASTRID RBH concepts is presented in this paper. Several RBH subassembly concepts specifically designed by CEA are under investigation whose design features and main parameters involved in their operation are described. Their current study/development status as well as related experimental and simulation support are finally discussed with regard to how they could match the ASTRID framework and future milestones.

INTRODUCTION/BACKGROUND I.

Late 2015, the three-years conceptual design phase (AVP2) for the 600 MWe Advanced Sodium Technological Reactor for Industrial Demonstration (ASTRID) prototype drew to an end. During this phase, main reference innovative technical options to improve the safety level with progress made in SFR-specific fields and to thus comply with the GEN IV objectives were provided with greater consistency and confirmed with a view to a consistent preliminary design. [START_REF] Abonneau | The ASTRID Project: Status and Prospects Towards the Conceptual Design Phase[END_REF] Designed with a defence-in-depth approach based on redundancy, diversity and independence, ASTRID has two distinct, diversified and independent fast-acting automatic reactor shutdown systems. Each shutdown system consists of sensors, logic circuit, drive mechanisms and mobile neutron absorber rods in stationary wrappers. The rod and wrapper form the absorber subassemblies distributed in the core. In ASTRID innovative reactivity control architecture, these systems are respectively called RBC (control and shutdown device) and RBD (diverse control and shutdown device). Both RBC and RBD systems are dedicated to the core reactivity management (power regulation, compensation for the reactivity change with time and normal or emergency shutdown) during the cycle.

Complementary safety devices dedicated to core damage prevention that would shut down the reactor passively have been implemented in ASTRID low void fraction (CFV) design core. [START_REF] Vénard | The ASTRID core at the midterm of the conceptual design phase (AVP2)[END_REF] They aim at complementing its natural behavior in case of some unprotected (i.e. with complete failure of all automatic shutdown systems) transients in order to meet temperature criteria on coolant, core and primary circuit structures within a robust margin and safety demonstration. Since the reactor shall be shut down passively, the actuation of these complementary safety devices has to be triggered directly by the sole physical effect induced by the accidental transient. ASTRID functional specifications and safety requirements for such complementary safety devices have been detailed in a previous paper, [START_REF] Guénot-Delahaie | Conceptual designs of complementary safety devices for ASTRID: from selection method to selected options[END_REF] as well as the structured process of value engineering. As a result of this process, two types of Complementary Safety Devices have been selected as the concepts the most capable to cope with both unprotected loss of flow (ULOF) and unprotected loss of heat sink (ULOHS) transients. The first type is based on the exploitation of the Curie point of the electromagnetic delatch system of ASTRID RBD shutdown system to address ULOHS transients. The second type is a hydraulically suspended absorber rod subassembly, called RBH, dedicated to ULOF transients. Both constitute the reference choice made for ASTRID reactor. [START_REF] Guénot-Delahaie | Conceptual designs of complementary safety devices for ASTRID: from selection method to selected options[END_REF] In this paper:

• the outlines of the value engineering process that led to the selected options/designs of ASTRID RBH concept are reminded in chapter II; • design features of several under investigation variants of RBH concept specifically designed by CEA and main parameters involved in their operation are described in chapter III; • their current studies/development status to reach appropriate Technology Readiness Levels [START_REF] Rodriguez | Qualification program of the ASTRID SFR project: Definition, Methodology and associated Risk Evaluation & Management[END_REF] as well as related experimental and simulation support are presented in chapter IV and V and discussed with regard to how they could match ASTRID framework and future milestones [START_REF] Rouault | ASTRID, The SFR GENIV Technology Demonstrator Project: Where are We, Where Do We Stand For[END_REF] .

REMINDER ON SELECTED OPTIONS AND II. DESIGNS

State of the art II.A.

CEA patented an active/passive safety system in 1973, where the absorber rod is at once held by an electromagnet and hydraulically suspended. [START_REF][END_REF] Other CEA studies were performed in the 80s related to a pure hydraulically self-resettable system. The hydraulically suspended concept was also explored late in the 70s in the United States. [START_REF] Dupen | Self-Actuated Shutdown System for a Commercial Size LMFBR[END_REF] The most actual emphasis is being given to this concept in Russia. [START_REF] Yu | Development of passive safety devices for sodium-cooled fast reactor[END_REF] It has in particular been chosen for BN-800 and may be implemented in BN-1200 sodium-cooled fast reactors. [START_REF] Vasilyev | BN-1200 reactor power unit design development[END_REF] This concept is innovative for a French sodium fast reactor.

Results of the value engineering II.B.

II.B.1. RBH concept selected for ASTRID

The RBH concept/device specifically designed by CEA consists of a mobile absorber rod in a stationary hexagonal wrapper tube almost identical to the fuel subassembly wrapper; the mobile rod is entirely contained within the wrapper tube at any time. It operates with a sodium flow path reduced section called "working zone" that creates an important pressure drop giving rise to hydrodynamic force intended to support totally the mobile rod above the core under the upward flow of the sodium coolant associated to normal operation. Should a ULOF event and the associated drop in flow rate occur, this upward force would become insufficient, making the rod drop hydraulically actuated and allowing the absorber material to get inserted by gravity into the active core region. A mechanism via an electromagnetic grip is developed that aims only at actively cocking/resetting the absorber rod before divergence. During normal operation, the absorber rod would be disconnected from the grip. This set of design options allows developing a system that is:

• completely independent of and distinct from the automatic shutdown systems; • diversified with respect to the automatic shutdown systems; this allows to fight against common mode failures of triggering and/or insertion. Indeed, the accidents which might affect all automatic shutdown systems could result from the failure of sensors or logic circuit, or from the mechanical inability to insert the absorber rods into the core, consecutive to rod and/or drive mechanism jamming or malfunction. Thus: o the way of triggering (passive instead of active) is a first source of diversification; o since RBH is engineered to have the head of its mobile rod permanently housed in the subassembly, it is diversified with respect to RBC in terms of insertion (i.e. it could ensure reactor shutdown in case of significant deformation of the reactor block that would be likely to block the RBC mobile rods in their wrappers); o since the RBH rod would be disconnected from the electromagnetic grip of the mechanism during normal operation, it is diversified with respect to RBD in terms of triggering (i.e. any self-welding between the rod head and the grip would be avoided); • as far as possible independent of the slab and of the main control room:

o any risk of re-connection under power between the rod and the grip, whether intentional or not, that could lead to the inhibition of passive actuation has to be eliminated by design. Management of operator interpretation fault and/or failure in mechanism grip state and/or failure of control system is the key point; o with a view to that, all the following measures may be contemplated: once RBH cocking is effective, the grip has to be put at such a distant parked position from the rod head that reconnection would not be possible even in case of electromagnet power supply; the mechanism kinematic line might include a component that would be supplied solely during reactor start-up phases (with pin-out outside the main control room) the grip parked position might be continuously monitored and an alarm could be triggered in case this position would not be respected anymore; the electromagnet power supply control knob might be deported outside the main control room; • in situ resettable after actuation, which allows in situ testability and rod worth measurement capability;

• not likely to introduce additional risks: o because the working zone is not active as the rod is in its lower position (during handling situations or after spurious drop during operation), any unexpected rod ascent from its lower position and subsequent reactivity insertion is avoided; o during operation, the wrapper ascent under the rod upthrust might be counteracted thanks to the mechanism; • of no particular impact on reactor operation recovery in case of spurious actuation, which would be detectable anyway; • not particularly significant in the core design 2 (as shown on Fig. 1) and the reactor integration because: o the slab design can accommodate the penetrations dedicated to the cocking mechanism of the limited number (i.e. 3) of required RBH in order to comply with requirement of about 1000 pcm to be inserted into the core after actuation of all RBH (with natural boron carbide B 4 C as absorber material), o it has a very low impact on the core's fuel volume fraction (less than one per cent); its impact on neutronics performance would be negligible; As such, this system abides satisfactorily by ASTRID specifications including functional specifications, safety and performance requirements to allow a safe state in the core to be achievable at temperatures compatible with the thermal criteria for the boiling point of sodium and the resistance of the structures, with a view to the short-and long-term management of the reactor.

II.B.2. Abandoned RBH concepts

Concepts based upon a self-resettable rod, on the one hand, and concepts involving a drive mechanism with a Curie point (electro)magnetic grip that would support the absorber rod in conjunction with hydrodynamic force, on the other hand, have been set aside. As regards the former:

• potential additional risks are deemed to be associated to uncertainty on the rod position (uncontrollable in all situations): o spurious ascent of the rod would not be detectable during handling phase; o under operation, reactivity insertion might occur in case a decrease in coolant flow rate would be followed by a sharp increase, inducing the rod ascent. As regards the latter:

• fine regulation of (electro)magnetic force that would equilibrate hydrodynamic force seems inconceivable on the whole lifespan: o with a permanent magnet whose magnetic properties would be sensitive to neutron flux, o based on operating experience from Phénix French reactor, the (electro)magnetic lifting force may decrease (as was the case within the backup control rod installed on Phénix; the cause of this decrease is not fully confirmed); • passive actuation could anyway be inhibited, intentionally or not, through increase of the magnetic force.

DESIGN FEATURES OF ASTRID RBH III.

III.A.1. RBH subassembly

The RBH subassembly whose stationary hexagonal wrapper tube is almost identical to the fuel subassembly wrapper is supported on the reactor grid plate. In the upper part of the rod, a shaft links the connecting surface to the electromagnet to the rod body which houses a bundle of vented absorber pins. In the lower part, the rod body is linked to the rod spike. In the rod body, the absorber pins are arranged on two circular arrays around a central pin as a bundle, covered by a cylindrical wrapper. They are hung from a collar attached to the rod shaft along with a centering grid at their lower end. B4C boron carbide pellets are stacked and shrouded in the absorber pin cladding tubes.

The working zone is developed thanks to the cooperation between a female piece (called "virole") supported by the stationary wrapper and a solid surface component (male piece) of the mobile rod. When both pieces are placed beside each other, the resulting reduced clearance (of about a few millimeters) awards the expected important pressure drop giving rise to the rod hydrodynamic support capability under sodium nominal flow rate. When both pieces are not facing each other anymore, the working zone is disabled (this configuration is exploited either to initiate the rod drop in the core or to avoid any expected ascent of the rod from its lower position in the subassembly). Here, as illustrated on Fig. 2, it has been chosen to develop the working zone under the fissile core, the male piece (called "pion") being supported on some given height by the lower end of the rod spike. Apart the "pion", the rod spike consists of a central shaft bearing some radial stiffeners. [START_REF]French patent by CEA[END_REF] Among the parameters involved in the hydrodynamic force phenomenon are the geometrical characteristics of the working zone ("pion" diameter and shape; working zone length; clearance between "pion" and "virole"; surface roughness).

Sodium enters at the foot of the subassembly, runs through the working zone, the absorber pin bundle and the annular gap between the mobile rod and stationary wrapper, and exits through the top of the cylindrical wrapper.

The rod movement inside the stationary wrapper during cocking or drop is guided/restrained at two levels at least: along the length of the rod spike (the working zone "virole" acting as a guide bush) and along the entire length of the pin bundle wrapper (guidance and/or angular movement restriction by means of pads located in the upper part of the subassembly). The parts of the wrapper in contact with the mobile rod ("virole" and pads) may have a hardfacing surface to ensure the durability of guidance throughout the life of the subassembly.

Once actuated, the RBH rod falls freely into the reactor core until it reaches a stop supported by the stationary wrapper. However, in the last part of the rod stroke, rod velocity is smoothly decreased using a passive device called "hydraulic dashpot" in which sodium is allowed to flow from one chamber to another through narrow clearances. The associated damping action provides some shock absorption; the final impact at the end of the stroke may however challenge the mechanical integrity of the rod and sleeve structures, especially the connections between their respective components. Two main types of dashpot design are currently under study for ASTRID RBH S/As:

• the first one (named _1) consists of an annular flange at the bottom of the rod body which penetrates in an annular groove provided from the upper face of the working zone "virole". Due to its radial design to be included in a given internal across flats dimension, this type of dashpot is a limitation to the development of a "virole" with high diameter; however the whole axial space between the altitudes of the fissile core bottom and the top grid plate can be made the most of to develop long "pion"; • the second one (named _2) consists of the rod spike lower end interacting with a component supported by the subassembly spike. This type of design allows the development of a "virole" with high diameter; however the axial space available to develop "pion" is reduced due to the axial dimension of the above-mentioned component.

All in all, two main variants of RBH conceptual designs are currently under study at CEA:

• the RBH_1 variant, as partially illustrated on Fig. 2, with a thin and long "pion" (i.e. the "pion" length is maximized), • the RBH_2 variant with a thick and short "pion (i.e. the "pion" diameter is maximized). In both cases, the sodium apparent mass of the rod is similar. In particular, the RBH_1 variant allows the decoupling of the lift and dashpot functions by differentiating the associated surfaces. This is highly suitable to avoid that the surfaces implied in the lift function be distorted/damaged because of the shock in the dashpot at the end of each rod drop stroke. [START_REF]French patent by CEA[END_REF] In particular, the RBH_1 variant presents some concave component (cf. its annular flange which is furthermore under neutron flux during normal operation) whose impact in terms of rod vibrations and/or hydraulic instabilities needs to be checked.

III.A.2. RBH cocking mechanism

A cocking mechanism with an in-sodium electromagnetic grip (as is the case for RBD drive mechanism) has been chosen as reference option, a mechanical grip being the alternative.

The main advantages are: • a simplified mechanism since no mechanical transmission to the grip is necessary, • a better protection against potential blockage of the rod in its wrapper (the disconnection would be easier), • an easier instrumentation configuration to detect the rod presence if need be.

Some technical difficulties identified with the RBD drive mechanism in-sodium electromagnet would be less sensitive:

• no safety function is attributed to the RBH electromagnet whose development is thus easier ; • it can be made the most of the development actions of the RBD electromagnet; • RBH ambient conditions are less stringent than RBD ones as regards the electromagnet.

On the contrary, the potential behavior of the electromagnet in normal operation with the rod disconnected needs to be investigated as it both:

• would be placed at the core outlet in a sodium flow mixing zone, • could be a ferromagnetic particle trap.

DEVELOPMENT STATUS AND ONGOING

IV. STUDIES

Final choices between RBH variants will be guided by:

• hydraulics performance: coolant actuation flow rate, hydrodynamic force, drop time; • management of other performance: neutronics efficiency, absorber rod cooling, reliability of rod insertion, cavitation, dashpot ability, subassembly integration… Actually, all these topics are intrinsically related to each other and need a strongly iterative management. The hydraulics topic is multi-parametric (cf. geometrical characteristics of the working zone, sodium flow rate value, rod mass) and cannot be decoupled from:

• the mechanical topic that aims at ensuring on the one hand the correct cocking of the rod, on the other hand the correct insertion of the rod in the core after actuation (especially the clearance between "pion" and "virole" is here involved that dictates in return the flow rate value in the RBH subassembly), • the damping topic (from hydraulics and mechanics point of view) that aims at ensuring the integrity of the components implied in the rod drop and shock (especially the design of the dashpot has consequences on the geometrical characteristics of the working zone via the axial and radial subassembly integration and on the pressure drop compilation in the RBH subassembly), • the safety topic that aims at ensuring no selfresettability capability to the rod, • the subassembly integration topic according to the limitation of the length of all ASTRID core subassemblies, • the within RBH subassembly pressure drop compilation topic that encompasses both: o the search for a maximization of the working zone pressure drop that must however remain consistent with the core pressure drop, o the compliance, via the flow rate value available for the lift function (deduced from the RBH subassembly pressure drop compilation), with other core thermal hydraulics criteria (for instance in terms of what part of the core flow rate RBH subassemblies could withdraw) or other hydraulics criteria (such as subassembly upthrust, cavitation, vibrations and thermal stripping), • the operational conditions of ASTRID reactor.

Hereafter are some focuses on ongoing related studies performed with a view to find an RBH operating point that would abide by all these topics and criteria. This operating point needs to be supported by water-loop experimental results to confirm/bring into question preliminary design features, to provide more information and greater consistency and to optimize the design. Related R&D is presented in chapter V along with the features of the RBH mock-up to be built.

Hydraulics calculations IV.A.

As regards the hydrodynamic force (assuming appropriate coolant actuation flow rate):

• it has been calculated through a one-dimensional modelling based upon Euler-Bernoulli theorems associated to the RBH subassembly pressure drop compilation. The rod apparent mass, the core pressure drop, the coolant actuation flow rate and the « pion » diameter are the first order parameters; • CFD calculations are ongoing.

As regards drop:

• the rod drop kinetics has been studied based on a simplified analytical modelling on a preliminary RBH design. Drop time obtained complied with requirements; • this analysis needs to be deepened as regards RBH_1 and RBH_2 variants. Following approaches will be used: o first, a classical approach based on CFD static calculations for various positions of the rod along its drop stroke to be integrated in a more global model to reproduce the free drop kinetics (before arrival in the dashpot); o then an innovative approach based on smooth particle hydrodynamics featuring the flow, to treat the drop kinetics in the dashpot (through a two-dimensional axisymmetric modelling of it). Optimization of the dashpot design is also expected.

Other hydraulics topics IV.B.

As regards cavitation and subassembly upthrust (N.B. both are treated through the same one-dimensional modelling as the one used to determine the lift force on the rod):

• the risk of cavitation may be significant at the working zone level where sodium flow mean velocity values can be high. If operating conditions in this zone can be computed through some operating Thoma criterion, only an experimental determination of the cavitation Thoma criterion is possible, the final comparison between these criteria showing whether cavitation can be expected or not. • the risk of subassembly upthrust, which may be high at nominal power and flow rate, can be managed by weighing down the subassembly stationary wrapper, by hydraulically locking it in the grid and by correctly designing the above the grid internal part of the subassembly spike.

Absorber rod cooling IV.C.

The objective is to determine the flow rate that is necessary to cool the absorber pins of the RBH bundle in such a way that their mechanical integrity would be guaranteed. This analysis is performed using a design tool based on a subchannel model of subassemblies and is based on some thermal criteria. Nominal operation scenarios and incidental scenarios involving the untimely dropping of a rod during nominal reactor operation are being considered. The aims for the chosen criteria are to minimize carburization of the cladding, to prevent meltdown of the absorber material and to keep the pellet/clad sodium bond from reaching the boiling point. A flow rate of a few kg/s per RBH subassembly enables compliance with all criteria in all scenarios. Finally the RBH subassembly flow rate is dictated less by the rod cooling requirements than by the rod lift needs.

Reliability of rod insertion IV.D.

Even if the rod is disconnected from its mechanism apart during cocking, problems regarding reliability of rod insertion might occur:

• the potential guide levels being quite distant from one another with regard to their functional reduced clearances, the risk of blockage/slowing down of the rod drop has to be investigated, taking account of the possibilities of misalignment between the rod and the stationary wrapper (for instance due to the manufacturing/mounting tolerances of the structures or the irradiation-induced subassembly bending). A preliminary simplified analysis needs to be revisited in light of RBH_1 and RBH_2 characteristics;

• the risk of blockage by bracing and/or stick-slip has also to be taken into account because of the rod body weight deported away from the working zone at the beginning of the rod drop. A preliminary analysis of bracing showed that no blockage had to be feared and that an increase of the radial clearance at the upper guide level would even be beneficial.

Subassembly integration IV.E.

Functional criteria and constraints to be complied with within the subassembly integration process are related to:

• the core reactivity management, that dictates the relative positioning between the absorber and fissile columns, • the rod stroke management, that dictates the positioning of stops in the wrapper and the guide provisions, • the management of differential deformations due to multiplicity of structural materials involved, that dictates the expansion provisions, • the keeping of functional guide clearances, that dictates the relative positioning between the guide levels and the fissile core and/or the selection of materials.

A calculation tool has been developed to perform the optimization of the subassembly integration in a multiparameters and multi-criteria framework (apart from the above-mentioned functional criteria and constraints, input data include especially the target subassembly length, the elevation of the fissile core imposed by ASTRID fuel subassembly design and the relevant material and design features of the components and their connections). Uncertainties (cf. manufacturing/mounting tolerances, material properties…) will be included in an upgraded version of this tool.

EXPERIMENTAL INVESTIGATIONS OF RBH V.

MOCK-UP HYDROMECHANICAL FEATURES Objectives

V.A.
It has been explained above how much experimental support is needed to understand the hydro-mechanical behavior of RBH concept depending on the numerous parameters and design features identified and to confirm its feasibility and robustness as regards the actuation technology, whose sensitivity to environmental variations (flow rate, deflections, vibrations…) has especially to be assessed. This is the purpose of out-of-pile loop tests scheduled in CEA Cadarache BACCARA facility [START_REF] Gastaldi | Experimental platforms in support of the ASTRID program: existing and planned facilities[END_REF] on a RBH mock-up whose main design options and development phases have been chosen in order that it be an analytical support to the RBH design and allow the qualification with sodium/water similitude conditions of the RBH hydromechanical behavior to reach Technology Readiness The BACCARA facility (see Fig. 3) is part of the GISEH platform. [START_REF] Rodriguez | Qualification program of the ASTRID SFR project: Definition, Methodology and associated Risk Evaluation & Management[END_REF][START_REF] Gastaldi | Experimental platforms in support of the ASTRID program: existing and planned facilities[END_REF] It has recently been refurbished with a view to RBH mock-up tests. Two configurations of the RBH mock-up are contemplated that correspond respectively to two types of tests, static and dynamic, to be performed chronologically:

• in the first step, mainly dummy stationary wrapper and rod are included, that build some working zone as illustrated on Fig. 4; the modular design aims at testing various geometrical configurations of the working zone; the rod mass is representative; • in the second step, with a view to test the whole operation scheme (cocking and release of the rod by the mechanism; lift; drop) of the RBH, some cocking grip and dashpot features will be added to the RBH mock-up.

Test program and schedule V.C.

As regards static tests:

• some aim at characterizing the effect of the "pion" diameter and of the clearance between the "pion" and the "virole" on the rod lift and at following the evolution of the required flow rate; • some others aim at optimizing the rod lift by investigating on the one hand the effects of the surface roughness and the working zone length on the required flow rate and on the other hand the effect of the "pion" form on the rod lift (via the regular and singular pressure drops it will induce in the working zone); • some others aim finally at characterizing the effect of some misalignment between "pion" and "virole" on the rod lift. First tests should be done in the second half of 2016.

As regards dynamic tests, to be performed on a reference configuration (rod mass, "pion" diameter and form, nominal flow rate) of the RBH mock-up that will be chosen on the basis of the static test results:

• rod cocking, increase of flow rate to the nominal value and rod release will first be simulated; • the rod behavior (with the rod either in natural position or submitted to some misalignment) will then be monitored while decreasing the flow rate until triggering of the rod drop.

Check and measurements of the mock-up behaviour/movements are to be performed with the help of following sensors:

• laser Doppler vibrometers aim at measuring the working zone output flow velocity; • an accelerometer (plus a window) located at the working zone output aim at detecting cavitation acoustically and visually; • several pressure sensors are foreseen distributed between the working zone and the rest of the test section; • force sensors aim at measuring the hydrodynamic force.

CONCLUSIONS VI.

In this paper, the design activities of the ASTRID RBH complementary safety device dedicated to ULOF transient management have been described. Several variants of RBH concept specifically designed by CEA are under study whose design features need experimental and simulation support to be confirmed or revised and optimized. Scheduled out-of-pile water loop tests are part of the global qualification plan towards licensing of ASTRID start-up core.

Most relevant issues to be addressed in parallel or thereafter in a R&D program:

• Try to benefit from international hydraulically suspended rod design/test feedback experience.

• Hydraulics calculations vs test results

o BACCARA test results will constitute a validation database for the hydraulics modelling.

THE

  INNOVATIVE RBH COMPLEMENTARY SAFETY DEVICE FOR ASTRID TO ADDRESS UNPROTECTED LOSS OF FLOW TRANSIENTS: FROM DESIGN TO QUALIFICATION I. Guénot-Delahaie, D. Lorenzo, B. Valentin, M. Zabiégo, V. Soukphouangkham, F. Biscarrat, T. Lambert, M. Phélip French Alternative Energies and Atomic Energy Commission (CEA), DEN/Cadarache/DEC F-13108 St-Paul-lez-Durance, FRANCE Tel: +33 4 42 25 75 73, Fax: +33 4 42 25 70 42, Email: isabelle.guenot-delahaie@cea.fr

Fig. 1 .

 1 Fig.1. Configuration of ASTRID core including RBH S/As (fuel S/As in yellow and red; RBC and RBD S/As in blue and black; RBH S/As in pink ; dummy S/As in white and grey)• provided studies and R&D confirm the feasibility and the robustness of this system (it is the purpose of chapters IV and V to report the progress made in this field), likely to comply with hydraulics performance requirements such as an actuation coolant flow rate of 45 % of the flow rate nominal value and a response/drop time (accounting for the duration between the time at which the coolant flow rate in the subassembly reaches the actuation value and the time at which absorber rods are fully inserted in the core) in the 1 to 30 s range (as suggested by some preliminary parametric thermal hydraulics calculations performed with the CATHARE system code 10 , which have shown that the response time is not a first-order parameter in the management of the ULOF transient).Actually, the value of the actuation coolant flow rate of 45 % of the flow rate nominal value has been chosen to satisfy at the same time: o the need for a hydrodynamic force high enough to support the rod above the core under the upward flow of the sodium coolant associated to normal operation condition, o the need for compatibility with all operational conditions and reactor availability factor that dictates following requirements:not to actuate before automatic reactor scram, should this works; not to interfere with any normal mode of operation including start-up and divergence, partial power operation (the associated minimum flow rate contemplated equals 55 % of the flow rate nominal value), shutdown and

Fig. 2 .

 2 Fig.2. Illustration of ASTRID RBH_1 S/A variant [Rod upper position on the left (with working zone activated) and lower position on the right (with working zone deactivated) Working zone "virole" in turquoise Working zone "pion" in checkered turquoise/magenta Rod spike in magenta Dashpot components in royal blue Rod body in green Upper guidance level illustrated by red triangles]

  Level 4-5 in 2016-2017 (TRL 4 = analytical validation of lift phenomenon; TRL 5 = almost analytical validation of the complete hydro-mechanical behavior in a representative environment). BACCARA facility and RBH mock-up V.B.

Fig. 3 .

 3 Fig.3. Sketch and picture of the BACCARA facility

ACKNOWLEDGMENTS

The authors would like to acknowledge the teams involved in the RBH design and development at CEA, AREVA and EDF.


I Guénot-Delahaie
email: isabelle.guenot-delahaie@cea.fr

D Lorenzo

M.-S X Jeanningros

J.-C Chenaud

B Garnier

J.-M Valentin

G Escleine

Avakian

Conceptual Designs Of Complementary Safety Devices For Astrid: From Selection Method To Selected Options

To comply with the GEN IV objectives set for it, the 600 MWe Advanced Sodium Technological Reactor for Industrial Demonstration (ASTRID) promises enhanced safety. The ability to shut the reactor down in any condition is one of the most important safety aspects.

At the end of ASTRID preconceptual design phase (end of 2012), the need for additional safety devices that would complement ASTRID low void fraction (CFV) design core natural behavior in case of unprotected loss of flow (ULOF) and loss of heat sink (ULOHS) transients emerged from the transient studies in order to meet temperature criteria on coolant, core and primary circuit structures.

The process from the selection method, on the basis of ASTRID functional specifications and safety requirements, to the selected options/designs, on the basis of itemization of these specifications/requirements and value engineering, of ASTRID Complementary Safety Devices (DCS) is presented. In this paper, only DCS dedicated to core melting Prevention (DCS-P) that would passively shutdown the reactor, i.e. whose actuation would be triggered by the sole physical effects induced by the transient, are dealt with regardless of whether they would depend on ASTRID first two distinct and independent fast-acting reactor shutdown systems (RBC and RBD) or not.

Several concepts of such DCS-P were considered among which SEPIA (abbreviation for SEntinal for Passive Insertion of Antireactivity) concept of absorber subassembly specifically designed by CEA, (DCS-P)-H concept of hydraulically suspended absorber rod subassembly and (DCS-P)-RBD_Curie concept based on the exploitation of the Curie point of the RBD electromagnetic delatch system; they are listed and discussed with regard to whether and how they could effectively match each type of ULOF and ULOHS transient (some concepts that were abandoned are discussed as well). Concepts which are the most promising in ASTRID framework, resulting from the best way their combination could deal with all considered transients, are finally described; these will be further investigated and developed in the near future.

I. INTRODUCTION/BACKGROUND

End of 2012, the two-years preconceptual design phase for the 600 MWe Advanced Sodium Technological Reactor for Industrial Demonstration (ASTRID) prototype was drawing to an end. During this phase, innovative technical options to improve the safety level with progress made in SFR-specific fields and to thus comply with the GEN IV objectives were considered; main options have been chosen as the reference options for the conceptual design studies intended to provide them with greater consistency. [START_REF] Le Coz | The ASTRID project: status and future prospects[END_REF] Designed with a defence-in-depth approach based on redundancy, diversity and independence, ASTRID has two distinct, diversified and independent fast-acting automatic reactor shutdown systems. Each shutdown system consists of sensors, logic circuit, drive mechanisms and mobile neutron absorber rods in stationary wrappers. The rod and wrapper form the absorber subassemblies distributed in the core.

In the reference design of the reactivity control architecture of the ASTRID reactor, defined at the end of the preconceptual design phase, these systems are respectively called RBC and RBD. For simplicity's sake, RBx may later on refer also to the individual element which solely comprises the absorber subassembly (RBx-AS) and its drive mechanism (RBx-BK) housed in the control plug, which is a part of the reactor's top shield.

As a major chosen design option, disconnection in case of scram between the RBD rod and its drive mechanism would occur via an in-sodium electromagnet (EM) that does not extend beyond the absorber subassembly lifting head. This constitutes a diversification against common mode failure of insertion of RBC rods into the subassemblies that makes it possible to guarantee safe core shutdown in case of significant deformation of the reactor block that would be likely to block the RBC mobile rods in their wrappers (disconnection of the RBC rods takes place at the level of the slab). [START_REF] Guénot-Delahaie | State of the art of the conceptual designs for ASTRID control and shutdown rods[END_REF] In ASTRID innovative aforementioned architecture, both RBC and RBD systems are dedicated to power regulation, compensation for the reactivity change during the lifetime and normal or emergency shutdown. Therefore, the RBD rod is partially inserted into the core during operation and is under neutron flux as well as the in-sodium EM.

II. NEED FOR COMPLEMENTARY SAFETY DEVICES

At the end of ASTRID preconceptual design phase, the need for additional safety devices that would complement ASTRID low void fraction (CFV) design core natural behavior in case of some unprotected (i.e. with complete failure of all automatic shutdown systems) transients emerged from the transient studies in order to meet temperature criteria on coolant, core and primary circuit structures.

As far as severe accident prevention is concerned, following representative events of transients have been studied:

• Unprotected Loss of Flow (ULOF) sequences: o total loss of power supply resulting in the trip of both primary and secondary coolant pumps and steam generators drying out; o trip of all the primary pumps while the secondary coolant pumps remain operational to remove power; o failure of a pump -diagrid connection (flow losses are faster in this scenario than in the others); • a sequence of Unprotected Loss of Heat Sink (ULOHS): the secondary pumps are tripped and the steam generators dry out.

Management of Unprotected Transient of Power (UTOP) sequences is beyond the scope of this paper.

The accidents which might affect all automatic shutdown systems could result from the failure of sensors or logic circuit, or from the mechanical inability to insert the absorber rods into the core (consecutive to rod and/or drive mechanism jamming).

The studies showed that a CFV-type core can provide a positive safety margin in relation to the boiling point. However, the integration of uncertainties showed that this margin is insufficient vis-à-vis a robust demonstration showing that sodium does not boil. [START_REF] Chenaud | Status the ASTRID core at the end of the pre-conceptual design phase 1[END_REF] With a view to ensure a robust margin and safety demonstration by resorting to an additional safety measure to cover these uncertainties and to achieve conditions which are compatible with the resistance of the structures, a selection study of potential complementary safety devices (dubbed DCS into French) has been launched.

In this paper, only DCS dedicated to core melting Prevention (DCS-P) are dealt with; mitigation of severe accidents is beyond its scope.

III. SELECTION METHOD

DCS-P specifications and safety requirements have been defined through an iterative approach between all people involved in related studies (designers, neutronics, thermal hydraulics, safety). They are reported in chapter III.A.

Feasibility and preconceptual design studies have been performed to ensure possibilities complying with these specifications exist. The subsequent value engineering process has been supported by preliminary preconceptual design studies performed in parallel; the concepts and design features (actuation principles and devices, geometry of systems, integration into subassemblies, absorber materials, absorber elements insertion into core) considered were thus directly:

• those that complied with most specifications,

• compatible with ASTRID reactor configuration (especially its design and operating conditions), • deemed properly supported on a technological basis along with CEA expertise and knowledge of sodium fast reactor state of the art: o in the field of absorber subassemblies design (for example, as regards compatibility of functional zones with consequences of structural material behavior under neutron flux), o as regards optimization of performances by decoupling them, o as regards robustness and actuation capability.

III.A. ASTRID DCS-P functional specifications and safety requirements

The directives for ASTRID are to develop DCS-P with functional specifications and safety requirements as follows:

• As abovementioned, to be specifically engineered for prevention of severe accidents; in the very unlikely case of a whole core accident, DCS-P might provide its mitigation with favorable but on no account worsening effects. • To be effective against ULOF and/or ULOHS and to cope with the short and long term (24 hours) management of such unprotected transients.

There may be a global interest to develop systems that are efficient on the largest range of transients and not dedicated to specific transients. • To be a negative reactivity insertion system as diverse as possible compared to RBC and RBD systems to fight against common mode failure. Diversification in terms of triggering is at least required.

• Preferably not to depend on the automatic shutdown systems to ensure that even if these fail, a safe state in the core will occur at temperatures compatible with the thermal criteria for the fuel melting margin, the boiling point of sodium and the resistance of the structures; should independence not be achieved, the reliability of the automatic shutdown systems should not be decreased by DCS-P. • To insert negative reactivity passively, insofar as the failure of the two automatic shutdown systems can be linked to the instrumentation and control system, i.e. DCS-P actuation would be triggered directly by the sole, whether unique or multiple, physical effect induced by the transient. • To be independent from the slab.

Independence may be considered in terms of triggering and insertion. • To have the capability of ensuring safe core shutdown in case of significant deformation of the reactor block that would be likely to block the RBC mobile rods in their wrappers N.B. Since RBD subassemblies have been designed accordingly to the very same requirement included in their specifications, it is therefore deemed negotiable in the case of DCS-P. In this way, independence from the slab in terms of insertion would not be strictly required. • To be easy to reset after actuation in order to provide: o in situ testability; o neutronic weighing capability.

This very restricting requirement means that devices that cannot be reloaded are deemed not convenient even if a demonstration of their actuation capability/reliability based on experimental testing and probabilistic considerations could be provided. Furthermore, this requirement may be inconsistent with a complete independence from the slab requirement, should the latter be strictly required.

• Spurious actuation must be detectable and shall not introduce additional risks. • Not to actuate before automatic reactor scram, should this works. • Not to interfere with any normal mode of operation (including start-up, shutdown and partial power operation) and handling. This in particular dictates the actuation value of sensitive parameter.

• As regards performances:

o To provide sufficient irreversible negative reactivity. o To be sufficiently fast acting on looking at postulated faulted conditions. o To have a limited impact on core performances and on integration in the reactor. o Not to penalize and/or increase the complexity of reactor operation (in terms of risk of spurious actuation and its impact on operation recovery). o To have a limited impact on fuel cycle and waste (in terms of management of new and irradiated fuel).

III.B. Value engineering process

On looking at these specifications and requirements, advantages and disadvantages of each concept among those considered as further on mentioned will first be highlighted along with its description in chapter IV.

Then the result of the global ranking process to qualitatively evaluate the performance characteristics of each system with respect to some synthetic categories underlying the specifications and requirements listed above or complementing them will be presented in chapter V.

IV. TOWARDS SELECTED OPTIONS AND DESIGNS

This feasibility study has amply taken advantage of worldwide past studies and sometimes development of passively activated safety shutdown devices for sodiumcooled fast reactors initiated mainly in the 90' and recently synthesized in an overview of existing concepts. [START_REF] Burgazzi | Analysis of solutions for passively activated safety shutdown devices for SFR[END_REF] Technical possibilities "available on the shelf" have been privileged among which concepts developed at CEA within French R&D programme and considered in ASTRID preconceptual design phase.

IV.A. Concepts considered

At the beginning of our search for passive safety features/systems 1) which would insert absorber material preferably by the gravity force, 2) where the delatch would be achieved by the coolant temperature rise and/or the coolant flow rate decrease under the considered transients, the following concepts were considered as candidates:

• hydraulic actuated systems in response to a drop in flow rate; • heat-actuated system in response to an increase in temperature: o the thermal expansion device, which inserts absorber elements using the differential expansion of structural pieces, o the Curie point electromagnet using the temperature sensing alloy, which will lose magnetism at a predefined temperature, o the melting point device, which drops absorber elements when a low melting point material melts.

IV.A.1. Abandoned Concepts

Concepts based upon melting point device, on the one hand, and concepts involving liquid absorber or absorber particles injected in the primary circuit, on the other hand, have been rapidly moved aside.

As regards the former:

• since non resettable after actuation, they present a problematic risk of spurious triggering at handling temperatures (no locking capability unlike mechanical devices); • moreover, ensuring suitable melting conditions is deemed a difficult task, especially because function of each device cannot be confirmed before operation. As regards the latter, their spurious injection would in particular not be compatible with investment protection.

Anyway, these concepts do not comply with in situ testability and neutronic weighing capability requirements.

IV.A.2. SEPIA Concept

SEPIA is an abbreviation for SEntinal for Passive Insertion of Antireactivity.

The SEPIA concept/device specifically designed by CEA is adapted to respond to an increase in temperature. Its actuation is triggered by the differential expansion of two cylindrical shells under temperature transient caused by ULOF and/or ULOHS. [START_REF] Guénot-Delahaie | State of the art of the conceptual designs for ASTRID control and shutdown rods[END_REF] SEPIA consists of a capsule inserted into the centre of a fuel subassembly pin bundle. The differential expansion of the two shells, which are located in the upper part of the capsule, causes several locking fingers on a column of spherical absorber pellets to rotate and thus the absorber elements to fall inside the capsule.

This set of design options allows developing a system that is:

• completely independent and distinct from the automatic shutdown systems; • highly responsive to temperature transients, since the actuating system benefits directly from the fuel subassembly flow rate; • robust from the operational point of view, because the mechanical principle for the actuating system allows out-of-reactor qualification, a locking capability to avoid any risk of spurious triggering at handling temperatures, the development of sufficiently wide movement for the required actuating precision, and the possibility to use mechanical means to ensure reliable actuation.

The system remains operational for conditions in which the subassemblies would be significantly deformed;

• not particularly significant in the core design because: o it requires no absorber drive mechanism and therefore no slab penetration, o it has a very low impact on the core's fuel volume fraction (less than one per cent), o the design concept for the capsule-carrying fuel subassembly is qualified (DCC Phénix) and it is also required for irradiation experiments in ASTRID.

Studies have shown that the thermocouples placed over the subassemblies enable the detection and localisation of the absorber column insertion in a SEPIA subassembly.

The main disadvantage is that as currently designed SEPIA can be used only one time and can therefore not be tested in situ and neutronic weighed. This disadvantage might be overcome by designing some sort of dedicated reset mechanism provided that compatibility with space available on the slab can be ensured.

IV.A.3. (DCS-P)-H Concept

This concept consists of a mobile absorber rod in a stationary hexagonal wrapper tube identical to the one of fuel subassemblies. At the normal operation condition, the absorber rod is hydraulically suspended above the core by the upward flow of the sodium coolant. Should a LOF event and the associated drop in flow rate occur, this upward force could become insufficient, making the rod drop hydraulically actuated and allowing the absorber material to insert by gravity into the active core region.

CEA patented an active/passive safety system in 1973, where the absorber rod is at once held by an electromagnet and hydraulically suspended. 5 Other CEA studies were performed in the 80s related to a pure self-resettable hydraulic system. The hydraulic suspension concept was also explored late in the 70s in the United States. [START_REF] Dupen | Self-Actuated Shutdown System for a Commercial Size LMFBR[END_REF] Thing worthy of notice, both French and American systems were engineered to be entirely self-contained within the reactor core control subassemblies, as is highly desirable for ASTRID.

The most actual emphasis is being given to this concept in Russia. [START_REF] Yu | Development of passive safety devices for sodium-cooled fast reactor[END_REF] It has in particular been chosen for BN800 and might be implemented in BN1200 sodiumcooled fast reactors.

The analysis of these different options points out that: • a drive mechanism via its grip may be either a help for cocking/resetting the absorber rod and/or supporting it in its upper position or of no use (i.e. cocking may be passive using hydrodynamic force), • such systems operate with a reduced section called working zone that creates an important pressure drop giving rise to hydrodynamic force intended to support totally or partially the mobile rod under nominal flow rate.

The main advantages of this concept are: Because the working zone is not active as the rod is in its lower position, the probability of spurious ascent of the rod is quite negligible during handling phase. During operation, the wrapper ascent under the rod upthrust might be counteracted thanks to the mechanism. o Limited impact on operation since spurious drop at nominal power would be detectable.

• In case of a mechanism with a Curie point magnetic grip: o possibility to address ULOHS and ULOF at once.

The main disadvantages of this concept are: • Since the system requires a small clearance between the mobile rod and the wrapper to create the working zone, it seems difficult to address the capability of ensuring safe core shutdown in case of significant core distorsion. This requirement is nevertheless not strictly required.

• In case of self-resettable rod:

o potential additional risks are deemed to be associated to uncertainty on the rod position (uncontrollable in all situations): spurious ascent of the rod would not be detectable during handling phase; under operation, reactivity insertion might occur in case a decrease in coolant flow rate would be followed by a sharp increase, inducing the rod ascent; should a reactor scram threshold be exceeded, it seems not desirable to rely on automatic shutdown systems to mitigate the consequences of a passive DCS failure. • In case of mechanism and a grip (whatever kind, mechanical or magnetic or electromagnetic), these could be blocked, intentionally or not (through reconnexion of the grip and/or increase of magnetic force). Management of operator interpretation fault and/or failure in mechanism grip state and/or failure of control system is the key point; it lies on the mechanism designers to eliminate by design the risk of inhibition of passive actuation. • In case of a mechanism with a Curie point magnetic grip: o Fine regulation of (electro)magnetic force that would equilibrate hydrodynamic force seems inconceivable on the whole lifespan: with a permanent magnet whose magnetic properties would be sensitive to neutron flux, based on operating experience from Phénix French reactor, the EM lifting force may decrease (as was the case within the backup control rod installed on Phénix; the cause of this decrease is not completely confirmed). o Passive actuation could anyway be inhibited, intentionally or not, through increase of the magnetic force.

IV.A.4. Passive Safety Features on RBD

Additionally to its function in the reactivity control system, RBD could fulfill the purpose of a passive complementary safety device using the Curie point characteristics of suitable material in its in-sodium electromagnetic delatch system: since the magnetic force is abruptly lost when the alloy is heated up to its Curie point in case of core outlet temperature increase under ULOHS, the device detaches the absorbing part of the rod.

Japan sodium-cooled fast reactor (JSFR) has incorporated such passive shutdown mechanism namely the self-actuated shutdown system (SASS) into its backup shutdown system. [START_REF] Nakanishi | Development of advanced loop-type fast reactor in Japan -(5): adoption of selfactuated shutdown system to JSFR[END_REF] JSFR SASS has obviously achieved a 7 Technology Readiness Level (TRL). Unlike JSFR SASS however, ASTRID RBD delatch system would be housed in the subassembly and partially inserted in the core as aforementioned.

The main characteristics of (DCS-P)-RBD_Curie concept are as follows:

• It is not physically independent from automatic shutdown systems; • The independence from the slab in terms of passive triggering is complied with insofar as the latter cannot be inhibited (even if the intensity in the EM were to be increased intentionally or not the lifting force would sharply decrease above the Curie point temperature); • The delatch mechanism is used both for passive and active operations; the safety function can thus be easily demonstrated and verified during operation; • RBD is designed to eliminate mechanical rod jamming from cause of this reactor shutdown system's failure. • No modification to RBD reliability is expected.

Enhancement of thermal expansion through devices like CREED studied during EFR project would be another way of disconnecting the absorber rod from the electromagnet of its drive mechanism. [START_REF] Favet | Third shutdown level for EFR project[END_REF] The so-called (DCS-P)-RBD_CREED would have the same first two characteristic as (DCS-P)-RBD_Curie. However, by adding a device on RBD system, this would be complexified and there is concern about its reliability not being modified.

(DCS-P)-RBD_Curie concept is therefore preferred to (DCS-P)-RBD_CREED. Anyway, the feasibility and operability of an electromagnetic delatch system partially inserted in the core is to be demonstrated (in terms of neutron flux influence on magnetic properties and magnetic force).

IV.B. Impact of (DCS-P) on transients

Preliminary calculations were performed with the CATHARE system code aimed at defining following specifications in order to orientate design studies:

• Case of (DCS-P)-H o Global negative reactivity worth of (DCS-P)-H system in the 500 to 1000 pcm range

V. VALUE ENGINEERING

V.A. Preliminary discussions

A preliminary value engineering led to retain (DCS-P)-H_MdB_armement as the best (DCS-P)-H concept for ASTRID for following reasons:

• it is deemed preferable to develop a mechanism (from which the French abbreviation MdB as a suffix) to avoid any unexpected rod rising from its lower position (during handling situations or after spurious drop during operation) and subsequent reactivity insertion; consequently, this mechanism aims only at cocking (dubbed MdB_armement into French), • other safety specifications are complied with at best, • preliminary design calculations have shown suitable performances can be achieved, provided following points be deemed acceptable (as not strictly required):

• the independence from the slab is not strictly complied with, • the ability to be actuated under multiple types of beyond-the-design accidents is not achieved.

A gradual assessment of DCS-P has been performed, firstly with regard to individual transients and secondly considering how DCS-P combination could cope with all transients notified in the specifications.

V.B. To address ULOF

At this stage, (DCS-P)-H_MdB_armement and SEPIA are involved in the comparative analysis with regard to compliance with specifications reported in Table I. The odds are in favour of (DCS-P)-H_MdB_armement. Even if dedicated drive mechanisms a might be thought of to give SEPIA the possibility to be tested and neutronic weighted, the required number of SEPIA in the core (~10) would induced too crowded a slab (in terms of penetrations and of upper face).

V.C. To address ULOHS

Even if operability/reliability of (DCS-P)-RBD_Curie is to be demonstrated, it is a better solution than SEPIA with main concerns as listed in Table I.

V.D. To address ULOF and ULOHS

Two main solutions emerged: 1. (DCS-P)-H_MdB_armement against ULOF + (DCS-P)-RBD_Curie against ULOHS o provided that mainly independence from automatic shutdown systems is not strictly required (as regards ULOHS).

a Such mechanisms would probably be more simple than traditional drive mechanisms, even if measurement capability as for any fuel subassembly could not be avoided; this gives in fact rise to a first difficulty since thermocouples leave restricted space.

VI. CONCEPTS SELECTED FOR ASTRID

As the result of the value engineering process performed in ASTRID framework, (DCS-P)-H_MdB_armement and RBD_Curie are deemed the most promising concepts whose combination could the best way cope with all considered transients, especially insofar as actuation mechanisms associated to two different sensitive physical quantities (temperature and flow rate) are involved; they thus constitute the reference choice made for ASTRID reactor.

In order to comply with requirement of about 1000 pcm to be inserted into the core after actuation of all (DCS-P)-H_MdB_armement (with natural boron carbide B 4 C as absorber material), the impact on core and reactor design would be limited to 3 locations on the grid, 3 slab penetrations and 3 mechanisms. The impact on the core's fuel volume fraction would be less than one percent; the impact on neutronic performances would be negligible. As a first proposal to be validated, the configuration of ASTRID core including (DCS-P)-H is shown on Fig. 1.

VII. ONGOING R&D AND STUDIES

Several (DCS-P)-H_MdB_armement conceptual designs are currently under study at CEA with differences related to:

• the kind of grip (mechanical, magnetic ?),

• the working zone characteristics (position, height).

Assuming compatibility with all operational conditions and safety specifications is checked, final choices will be guided by:

• Hydraulic performances: coolant actuation flow rate, hydraulic force, drop time. N.B. As regards response/drop time (accounting for the duration between the time at which the coolant flow rate in the subassembly reaches the actuation value and the time at which absorber rods are fully inserted in the core), preliminary parametric thermohydraulics calculations performed with the CATHARE system code have shown that it is not a first-order parameter; the [5 s; 30 s] time bracket is deemed a suitable target.

• Management of other performances: neutronic efficiency, rod cooling, reliability of rod insertion, cavitation, dashpot ability, subassembly integration…

Most relevant issues to be addressed (in a R&D programme):

• (DCS-P)-H_MdB_armement:

o Hydraulic tests to be launched during conceptual design phase; TRL objective: 4 in 2014 and 5 in 2015. o Hydraulics calculations vs test results

• (DCS-P)-RBD_Curie:

o Try to benefit from JSFR SASS design/test feedback experience.

VIII. CONCLUSIONS

In this paper, ASTRID need for Complementary Safety Devices in support of a robust safety demonstration of CFV core damage prevention in case of ULOF and ULOHS transients has been explained. Corresponding specifications and safety requirements have been defined and candidate concepts the most compatible with ASTRID reactor configuration and planning considered.

As a result of a structured process of value engineering, (DCS-P)-H_MdB_armement and (DCS-P)-RBD_Curie have been selected as the concepts the most capable to cope together with ULOF and ULOHS transients. Studies to confirm and optimize preliminary design features need to be pursued along with R&D actions to be launched in the near future.

  (assuming unitary worth of about 200 pcm) o Response time in the 5 to 30 s range o Actuation coolant flow rate of 40 % of the flow rate nominal value. • Case of SEPIA o Global negative reactivity worth of SEPIA in the 500 to 950 pcm range (assuming unitary worth in the 40 to 80 pcm range) o Response time around 5 s o Actuation temperature around 650°C. • Case of (DCS-P)-RBD_Curie o Actuation temperature around 650°C.

Fig. 1

 1 Fig.1 Configuration of ASTRID core including (DCS-P)-H (fuel S/As in yellow and red; RBC and RBD S/As in black and blue; (DCS-P)-H in pink)

TABLE I

 I Compliance with specifications -Comparative analysis of DCS-P against ULOF events: (DCS-P)-H_MdB_armement vs. SEPIA

	Specifications				SEPIA	(DCS-P)-H_MdB_armement
	Safety Requirements Independence	from	automatic	OK	Independence from slab to be
		shutdown systems			studied by mechanism designers
		Independence from slab
		Detectability
		Response time
	Robust	No additional risks		OK	OK
	demonstration	Capability of in situ test and	NO unless dedicated mechanism is	OK
		neutronic weighing		developed
	Impact on core performances and reactor integration		less than 1% on the core's fuel	around 1% on the core's fuel
					volume fraction	volume fraction
						3 slab penetrations
	Impact of spurious actuation (save due to any device's own	Spurious actuated S/A must be get	No particular impact
	failure) on: (1) reactor operation recovery; (2) fuel cycle and	out of core
	waste
	Technology Readiness Level TRL			currently 3	currently 3
						achievable: 4 in 2014 ; 5 in 2015
						(7 according to Russia experience
						feedback)
	Synthesis

of main disadvantages/concerns and ongoing studies to confirm feasibility

	No capability of in situ test and	Independence from slab to be
	neutronic weighing	studied by mechanism designers
	Spurious actuated S/A must be get
	out of core	Out-of-pile loop tests of actuation
	Risk of non simultaneous actuation	technology (to assess sensitivity to
	to be assessed	environmental variations: flow
	Out-of-pile loop tests of actuation	rate,	deflections,	vibrations,
	technology	impurities…)

ACKNOWLEDGMENTS

The authors would like to acknowledge R. Lavastre, P. Marsault, C. Vénard, P. Sciora, D. Blanchet, B. Fontaine from CEA, who contributed to this work.


I Guénot-Delahaie

D Lorenzo

B Valentin

J.-M Escleine

T Helfer

State of the art of the conceptual designs for ASTRID control and shutdown rods

A critical look at the conceptual designs of control and shutdown rods and absorber elements, along with the lessons learnt from the operation of French fast reactors (Phénix and Super-Phénix especially) and the associated irradiation tests, has yielded improved and even innovative absorber assembly design concepts which are presented in this paper. To comply with the GEN IV objectives set for the 600 MWe Advanced Sodium Technological Reactor for Industrial Demonstration (ASTRID), these design concepts have been researched with a view to improved economy/sustainability and enhanced safety.

The two main measures undertaken to achieve economy, among many others, have been to reduce the absorber subassembly dimensions and boron carbide enrichment, as well as to extend the residence time. To achieve enhanced safety, measures could include improved components and/or structural materials and guidance surface coatings/hard-facings in active shutdown systems. As part of these measures, a new kind of absorber assembly has also been designed -called SEPIA 1 -pertaining to safety devices for the passive insertion of negative reactivity in the core.

Preliminary thermal-hydraulic and structural mechanical analyses have been carried out with the CADET and LICOS project codes to show their feasibility. Further detailed analyses need to be carried out to achieve optimum dimensions that comply with the RAMSES II design rules.

The paper discusses the basis of the conceptual designs, giving due consideration to emerging design concepts, analysis backups and further R&D required for design qualification.

Preliminary discussions

In 2010, the CEA started the first phase of conceptual design for the 600 MWe Advanced Sodium Technological Reactor for Industrial Demonstration (ASTRID) prototype to define innovative technical options.

Designed with a defence-in-depth approach based on redundancy, diversity and independence, ASTRID has two distinct and independent fast-acting reactor shutdown systems. Each shutdown system consists of sensors, logic circuit, drive mechanisms and mobile neutron absorber rods in stationary wrappers. The rod and wrapper form the absorber subassemblies distributed in the core. The absorber subassembly of the first system dedicated to power regulation, compensation for the reactivity change during the lifetime and normal or emergency shutdown is called RBC subassembly, and that of the second system dedicated to emergency shutdown, RBS subassembly. The respective drive mechanisms, RBC-BK and RBS-BK, are housed in the control plug, which is a part of the reactor's top shield.

With a view to a more robust design and safety demonstration, additional core safety features/systems, called SEPIA, are designed as potential options based on passive insertion of negative reactivity equivalent to a 3 rd shutdown system, allowing the return to a safe state in case of a loss of coolant accident without scram [START_REF] Gauché | Generation IV reactors and the ASTRID prototype: lessons from the Fukushima accident[END_REF].

Major design options

Itemising ASTRID specifications

The following main improvement objectives and related topics for subsequent investigation stemming from ASTRID specifications motivated the structure of RBC, RBS and SEPIA subassembly design activities: [START_REF] Gauché | Generation IV reactors and the ASTRID prototype: lessons from the Fukushima accident[END_REF] enhanced safety (a) lessons learned from previous fast-reactors operations (b) structural materials and guidance surface coatings/hard-facings [START_REF] Munoz | Recovery of materials from PHENIX to support the qualification of ASTRID design options[END_REF] projected cost savings (a) reducing absorber subassembly length (b) extending residence time: it is economically desirable to match the residence cycle and, if possible, life span of the absorber elements with those for the fuel cycles (4 runs of 360 or 400 effective full power days, or EFPD, each) (c) reducing the total number of absorber subassemblies (d) keeping the enriched absorber material requirement as low as possible

Value engineering

As regards RBC and RBS subassemblies within first phase of ASTRID conceptual design, it was decided to use the same design principles as for the French Phénix, Super-Phénix and EFR systems:

 alignment of all (fuel and absorber) subassembly heads in the core,  retention of the mobile absorber rod inside the wrapper and maintaining its gravity actuation;  for the RBS subassemblies, disconnection in case of scram, via an in-sodium electromagnet that does not extend beyond the absorber subassembly lifting head. This constitutes a diversification against common mode failure of insertion of control rods into the subassemblies that makes it possible to guarantee safe core shutdown in case of significant deformation of the reactor block that would be likely to block the RBC mobile rods in their wrappers (disconnection of the RBC rods takes place at the level of the slab).

With reducing the height of the subassemblies as an objective, the following options for the absorber subassemblies were chosen:

 RBC subassembly with lower guide bush and olive-shaped rod spike linked to an absorber drive mechanism with a bayonet-type gripper;

 RBS subassembly with a dashpot located in the lower part and an unsegmented absorber bundle linked to an electromagnetic gripper.

The current standard generic absorber element for the French RNR reactor type has been renewed as the reference option for the start-up core. This absorber element consists of a vented pin with a permanent sodium bond between the stainless steel cladding and the absorber column, which is equipped with a confining shroud (to prevent migration of the absorber fragments). The main absorber rods consist of bundles of such pins, which may or may not be enclosed within a cylindrical pin bundle wrapper depending on the cooling requirements. Boron carbide (B 4 C), which may or may not be enriched in 10 B (the neutron-capturing isotope of the neutronabsorbing boron element) is currently the only neutron-absorbing material and is not currently being reassessed for the first cores. Most of the R&D and qualification requirements mentioned further on stem from the operating issues inherent to this type of design and/or absorber material. However, both the search for more durable standard absorber elements and the specific requirements for SEPIA system operational modes fully justify requirements that tend to break with current feedback on design/absorber type. Alternative absorber materials like HfB 2 and TiB 2 , whether or not they are enriched in 10 B, and even Hf are all being considered for the longer term, along with a single-or double-donut shaped absorber column design.

As for the structural materials, the first ASTRID cores will be based on a choice of known and already tested materials, some of which have even been tested in Phénix cores, e.g. EM10 for the hexagonal tube material and AIM1 for the cladding. All this helps to increase the reliability of rod insertion due to a better understanding and integration of irradiation swelling.

Design features -Description

RBC subassembly

The RBC subassembly consists of a mobile control rod in a stationary hexagonal wrapper tube identical to the fuel subassembly wrapper, as shown in Fig. 1. Its length is 4.5m and it is supported on grid plate. Nineteen absorber pins housed in the mobile RBC rod are held by guide rails and hang freely from the top. They are arranged in triangular pitch as a bundle, covered by a cylindrical wrapper. B4C boron carbide pellets are stacked and shrouded in the absorber pin clad tubes. Sodium enters at the foot of the subassembly, passes through the absorber pin bundle and through the annular gap between the mobile RBC rod and stationary wrapper, and exits through the top of the cylindrical wrapper. RBC rod movement inside the stationary wrapper is guided at two levels: at the lower end of the rod spike (mobile guidance along a guide bush located over the spike) and along the entire length of the pin bundle wrapper (stationary guidance by means of pads located in the upper part of the subassembly). The parts of the wrapper in contact with the mobile rod (guide bush and pads) have a hard-faced surface to ensure the durability of guidance throughout the life of the subassembly.

FIG. 1. RBC subassembly (red part representing the position of fissile core; orange arrow showing sodium flow entry)

A bayonet-type gripper at the bottom end of RBC-BK may be used to hold/release the head of the mobile RBC rod, as illustrated in Fig. 2.

FIG. 2. How the RBC rod could be connected to a bayonet-type gripper

(courtesy of AREVA-NP)

RBS subassembly

The RBS subassembly consists of a mobile rod in a stationary hexagonal wrapper tube identical to the fuel subassembly wrapper, as shown in Fig. 3. The rod's mechanical structure is maintained by a central anchor shaft linking the connecting surface to the electromagnet (in the upper part of the rod) to the dashpot piston (in the lower part). A collar is attached to the anchor shaft, from which the pins are hung along with a centring grid at the lower end of the pins. Unlike for the RBC rods, the objective is not to optimize pin bundle cooling by channelling most of the sodium into a pin bundle wrapper, inasmuch as the absorber material here is not introduced into the core during operation. When the rod has dropped, shock absorption is activated by the piston, which enters a cylinder attached to the wrapper. A wide diametrical gap is engineered between the wrapper and the rod along the entire length of the rod's stroke to ensure rod insertion even in conditions where the wrapper is significantly deformed (bending and/or crushing in the across-flat area).

FIG. 3. RBS subassembly (red part representing the position of fissile core; orange arrow showing sodium flow entry)

SEPIA subassembly

The safety directives for ASTRID are to develop a diversified negative reactivity insertion system that is independent from the two shutdown systems to ensure that if these fail, the hot zero power state in the core will occur at temperatures compatible with the thermal criteria for the fuel melting margin, the boiling point of sodium and the resistance of the structures. To the extent that the failure of the first two systems can be linked to the instrumentation and control system, a requirement is that the third system can be actuated directly in response to a physical phenomenon. A number of different concepts are under study, including a hydraulic actuation system (in response to a drop in flow rate) and a heat-actuated system (in response to an increase in temperature). The research on this last concept is more advanced; actuation is triggered by the differential expansion of two cylindrical shells under temperature transient caused by the loss of the primary flow and/or the secondary flow which is not protected due to the failure of the first two systems. This device, called SEPIA, is illustrated in Fig. 4.

SEPIA consists of a capsule inserted into the centre of a fuel subassembly pin bundle. The differential expansion of the two shells, which are located in the upper part of the capsule, causes several locking fingers on a column of spherical absorber pellets to rotate and thus the absorber elements to fall inside the capsule. This set of design options allows developing a system that is:

 completely independent and distinct from the first two shutdown systems;  highly responsive to temperature transients, since the actuating system benefits directly from the fuel subassembly flow rate;

 robust from the operational point of view, because the mechanical principle for the actuating system allows out-of-reactor qualification, a locking capability to avoid any risk of untimely triggering at handling temperatures, the development of sufficiently wide movement for the required actuating precision, and the possibility to use mechanical means to ensure reliable actuation. The system is also far enough away from the neutron flux that it is unaffected by irradiation swelling, and remains operational for conditions in which the subassemblies would be significantly deformed due to the mechanical decoupling of the capsule and the central wrapper, to the form of the absorber elements (column of spherical pellets) or to the wide diametral gap created between the absorber elements and the capsule;

 not particularly significant in the core design because: a) it requires no absorber drive mechanism and therefore no slab penetration, b) it has a very low impact on the core's fuel volume fraction (less than one per cent), c) the design concept for the capsule-carrying fuel subassembly is qualified (DCC Phénix) and it is also required for irradiation experiments in ASTRID.

Studies have shown that the thermocouples placed over the subassemblies enable the detection and localisation of the absorber column insertion in a SEPIA subassembly. An ultrasonic detection system is also being investigated.

The main points that remain to investigate in terms of feasibility practically all have to do with qualification, inpile tests for absorber elements and out-of-pile loop tests for the actuation technology.

FIG. 4. Illustration of the SEPIA subassembly

Examples of studies carried out

Absorber element design methodology and topics/specific points to be investigated to match the required residence time

The generic design process aims to guarantee throughout the life of an absorber element both the integrity of the cladding in compliance with the RAMSES II design rules and with respect to the propagation of any internal cracks initiated by corrosion of the cladding material in connection with absorber cladding chemical interaction (ACCI), and the absence of: 1) melting of the absorber or of the shroud, 2) absorber cladding mechanical interaction (ACMI) or 3) interaction between absorber elements. Support calculations have been performed using the LICOS project code [START_REF] Michel | Simulation of the Pellet Cladding Interaction phenomenon with the PLEIADES fuel performance software environment[END_REF].

The issues relating to compliance with meltdown prevention criteria must be distinguished from life span issues. French feedback on these issues mostly concerns boron carbide (B 4 C). Melting temperature and integral thermal conductivity under radiation are the parameters that, together with the geometry of the basic absorber entity (pellet, sphere, etc.), determine whether or not there is compliance with the meltdown prevention criterion. Currently, the most restrictive factor concerning B 4 C is clearly the excessively conservative formalism describing the changes to its thermal conductivity under fast neutron flux.

The volumetric swelling of the absorber entity under irradiation and internal corrosion of the cladding material by the absorber material have been identified as the phenomena most likely to reduce the absorber element's life span. Independently, their kinetics determine the instant that ACMI occurs and the time it takes to reach the maximum acceptable corroded thickness relative to the cladding mechanical strength. Together, these phenomena determine the life span over and above which the loss of shroud ductility becomes incompatible with its containment function.

Due to the heavy production of helium when boron materials are subjected to neutron flux, the swelling kinetics are especially dependent on the changes in the capture density rate for 10 B. The absence of ACMI will then depend on not exceeding a certain capture density threshold. In the absence of a confining shroud, the appearance of ACMI is subject to the random behaviour of any fragments in the joint and to compression caused by macro-/micro-fracturing of the absorber entity. By holding fragments in place and preventing compression, the shroud significantly slows down ACMI. Achieving long service lives such as those sought for ASTRID and the Generation IV sodium-cooled fast reactors is subject to the reliability and performance of this component.

Strict compliance with the B 4 C and shroud meltdown prevention criteria as regards conventional absorber elements has never led to a technological impasse. With respect to the ASTRID project, compliance presents less of a design optimisation problem than it used to, now that 10 B enrichment has been scaled down. A life span of at least two runs out of the four required for ASTRID is an accessible objective to start with, by renewing the components and materials that have been validated in the Phénix and with the added reliability of the shroud and the use of "carbothermal"-generated B 4 C instead of the "magnesiothermal"-generated B 4 C [START_REF] Kryger | Absorber elements. The nuclear fuel of pressurized water reactors and fast reactors -design and behaviour[END_REF] that is the sole basis for the knowledge gleaned from Phénix feedback. Even if in this respect the presence of a shroud is supposed to cancel out the B 4 C fabrication differences, it still is necessary to update a behavioural comparison of the two types of product, including for the corrosive potential. Due to the neutron volume requirements for SEPIA absorber entities (solid spheres), the current conservatism of the thermal conductivity law for B 4 C turns out to be incompatible with compliance with the meltdown prevention requirement for the absorber. This obstacle alone fully justifies experimental efforts to do away with or at least to better define this conservative approach and find alternative absorber materials that are significantly more conductive and/or heat-resistant than boron carbide. Another major research and development challenge, practically as important as the nature of the absorber material, is to perfect a suitable shroud for SEPIA absorber entities.

Preliminary thermohydraulic analysis

This analysis is done using CADET [START_REF] Valentin | The thermal-hydraulics of a pin bundle with an helical wire wrap spacer. Modelling and qualification for a new subassembly concept[END_REF] and LICOS design tools, based on the thermal criteria that apply to the above-mentioned design process. Nominal operation scenarios and incidental scenarios involving the untimely dropping of a rod during nominal reactor operation are being considered.

The aims for the chosen criteria are to minimize carburisation of the cladding, to prevent meltdown of the absorber material and keep the sodium joint from reaching the boiling point.

RBC and RBS subassemblies

In the case of the RBC and RBS subassemblies, a flow rate of a few kg/s per subassembly enables compliance with all criteria in every scenario.

For both, the cavitation and fly-off risks still remain to be studied. Even so, the low flow rates required to cool these subassemblies makes it possible to significantly limit these two risks, so the feasibility studies reveal nothing that would disqualify the chosen designs. However, before launching a full thermohydraulic design for these two subassemblies, more thorough studies do need to be done.

SEPIA subassembly

A specific thermohydraulic feature of this subassembly is the column of spherical absorber pellets that are introduced into a device called the capsule, which is placed within a fuel pin bundle from which two rows of pins have been removed. The capsule has no internal flow rate for cooling the absorber spheres, so these are cooled by conduction and then by thermal coupling with the bundle coolant medium. The CADET code is used to do the design analysis to integrate the fuel pin bundle and the transverse thermocouples, while the LICOS code serves to assess the internal temperature of the designed absorber sphere. Two absorber materials are analysed: B 4 C and hafnium diboride HfB 2 . The limit value of thermal conductivity after irradiation is assumed for each one. The two above-mentioned scenarios are studied.

The design analysis shows that whatever the scenario, the fissile pin bundle is always correctly cooled with a flow rate of about 36 kg/s, even without flow inside the central wrapper. The most severe design scenario is the one featuring an untimely drop of the column of spherical absorber pellets with the reactor continuing nominal operation. Therefore, given the current knowledge on B 4 C (especially on how its thermal conductivity changes under irradiation):

 for B 4 C: it is not possible from the thermohydraulic viewpoint to ensure the meltdown prevention criterion, even with a flow inside the central wrapper or by increasing the flow rate in the fissile bundle;

 only by replacing B 4 C with HfB 2 can the meltdown be avoided.

So, the results do show, however, that the feasibility of the SEPIA subassembly is linked more to the chosen absorber material solution (B 4 C versus HfB 2 ) than to its thermohydraulic feasibility, which remains to be refined and validated but so far does not give reason to abandon this design concept.

Subassembly spikes with additional discriminatory feature

The goal is to keep the core sub-critical under all plant conditions, including design basis conditions and design extension conditions. In order to meet this goal, the normal operating reactivity will be defined so that the sum of all credible reactivity insertions during normal operation is insufficient to cause criticality. During refuelling, absorber rods will be used to reduce the reactivity still further, to account for the additional reactivity associated with possible handling errors.

Additional protection against handling error during refuelling will be provided by including discriminatory features (currently under study) on the subassembly spikes to prevent substitution errors, e.g. the wrong type of subassembly from being completely lowered into the wrong position.

Outlook

R&D and qualification requirements to support design options

In 2012, the different requirements for R&D and qualification of the absorber subassemblies (including the materials) for the CFV core (low sodium void fraction) were analysed, which made it possible to identify the more structurally significant ones with respect to the ASTRID plan. The analysis was based on lessons learnt and on the identification of lacunae in the state of the knowledge and the experimental databases.

Given the current knowledge, as we progress with the ASTRID specifications with the perspective of at least matching the life span of the absorber elements and subassemblies to that of an ASTRID reactor fuel core, it emerges that the main qualification issues concern:

 for absorber element R&D: shroud behaviour, structural corrosion (cladding and/or shroud), and absorber thermal behaviour, with the intent to use the most durable materials possible as a common thread,  for absorber subassembly R&D: the behaviour of hard-faced parts that are implicated in guidance problems.

The qualification issues relating to SEPIA were discussed above.

Post-irradiation experiments on already irradiated objects

Generally speaking, the qualification of absorber elements and subassemblies will broadly rely on the postirradiation examinations of irradiation experiments carried out in the Phénix reactor. The objects of interest for the ASTRID core and the testing requirements were identified in connection with the "Phénix Treasures" project [START_REF] Munoz | Recovery of materials from PHENIX to support the qualification of ASTRID design options[END_REF].

R&D actions on shroud behaviour and cladding/shroud corrosion

The shroud and corrosion behaviour studies must be linked. Before start-up, these studies will consist of sodium loop tests, specific mechanical tests of corroded specimens and analytical irradiation experiments. As for the conventional absorber elements, they must first consolidate or confirm (relative to the "carbothermal" B 4 C) what can already be considered as acquired (half-life span) and then reach at least the intended life span of the conventional absorber elements by assessing other types of shroud and/or cladding materials (coatings included) and other design options. If no integral or even representative irradiation occurs prior to the start-up of ASTRID, the scenario will aim for two cycles for the initial loading of ASTRID, based on what we already know. These two cycles would be subject to an appropriate monitoring plan and efforts to introduce, as soon as possible, some precursor materials with the potential for longer life spans (cf. the Superphenix approach).

R&D actions on absorber materials

Three main areas for research and development on absorber materials were identified:

 Consolidating and advancing the knowledge on boron carbide (B 4 C) to improve phenomenological modelling and formalisation of its irradiation behaviour. The first step is to make the best use of the most relevant irradiated absorber entities from the "Phénix Treasures" project, in particular those fabricated from "carbothermal" B 4 C.

 Studying the changes in the properties and behaviour of HfB 2 and TiB 2 under irradiation. These diborides have been identified among other boron materials as potentially and globally the most capable of meeting the operating requirements for SEPIA systems. They also show the right potential to be appropriate longterm substitutes for the use of B 4 C in general. A preliminary R&D step is needed to confirm whether or not they are capable of maintaining under neutron flux the initial properties they were chosen for, i.e. a high and distinctly metallic thermal conductivity and high refractoriness.

 Hafnium metal, which is a robust and durable alternative under rapid flux to boron materials as long as its low intrinsic absorbing power does not constitute an unacceptable handicap. It also has excellent mitigation potential, which is an advantage. The main question is whether it is capable of remaining under rapid flux for long periods (ageing) without unacceptable loss of its geometric integrity.

Irradiation

With respect to absorber qualification, investigations will be launched to carry out:

 irradiation of specific materials to qualify the coatings/hard-facings for guidance parts that are exposed to neutron flux,  irradiation of absorber elements, using design and flux conditions that are as representative as possible of the start-up core in support of 1) validating the overall design-fabrication-service load process, 2) qualifying "carbothermic" B 4 C and 3) demonstrating the shroud life span.

ABBREVIATIONS

ACCI

  Absorber-cladding chemical interaction ACMI Absorber-cladding mechanical interaction AIM1 Austenitic improved material 1 ASTRID Advanced sodium technological reactor for industrial demonstration EFPD Effective full power days EM10 Martensitic alloy RAMSES II Design rules for the fast breeder core structures embrittled by irradiation RBC-BK Drive mechanism associated with RBC subassembly RBS-BK Drive mechanism associated with RBS subassembly SEPIA Sentinal for passive insertion of antireactivity

This is an abbreviation for "SEntinal for Passive Insertion of Antireactivity".

ACKNOWLEDGEMENTS

The authors wish to thank AREVA-NP teams involved in absorber assembly components and drive mechanisms design and both AREVA-NP and EDF for their financial support.