In this paper we investigate the possibility of using a set of in-vessel, ex-core, fission chambers to detect the spatial changes in the power profile that would result from an IRW. A preliminary core design of a Generation IV, 1500 MWth sodium-cooledfastreactor (SFR) designed at CEA  is used for our study. We use the Monte Carlo based code SERPENT2  in criticality mode for the neutron transport calculations. The paper is organized as follows: We review NFMS objectives and limitations in Section 2. This is followed by reactor core description, choice of detectors and their location in Section 3.1 to 3.3. We present the calculation route and the computational tools used in Section 3.4. We show the results from the criticality calculations in Section 4 followed by a discussion in Section 5 and conclusion in Section 6.
Sodium-cooledFastReactor (SFR), gas Power Conversion System (PCS), Loss Of Off-site Power (LOOP), Multiobjective Optimisation Problem (MOP).
The French Commission for Atomic Energy and Alternative Energy (CEA) in collaboration with its industrial partners develops Sodium-cooledFast Reactors (SFR) as industrial-scale demonstrators mainly guided by safety and operability objectives. In this paper, a SFR reactor associated to a nitrogen closed Brayton cycle for the Power Conversion System (PCS) is considered. In incidental and accidental conditions, the operation of the reactor must be defined to keep it under control and to fulfil safety requirements. This paper is dedicated to an alternative procedure to control a Loss Of Off-site Power (LOOP). Usually, in case of LOOP, the SFR standard procedure relies on passive Decay Heat Removal (DHR) systems to cool down the primary circuit. In this paper, an alternative solution substitutes the latter by the gas Power Conversion System (PCS). This aims at reducing the delay to reach the cold shutdown state while fulfiling safety criteria dealing with thermal stress issues. The operating of the gas PCS required three regulations:
FRAMATOME, DTIML, 10 rue Juliette Récamier, 69006 Lyon, France
Received: 11 February 2019 / Received in ﬁnal form: 20 May 2019 / Accepted: 2 August 2019
Abstract. The feedback produced by operating Sodium-cooledFast Reactors (SFRs) has shown the importance of material tribological properties. Where galling or adhesive wear cannot be allowed, hardfacing alloys, known to be galling-resistant coatings, are usually applied on rubbing surfaces. The most used coating is the cobalt-base alloy named Stellite 6 ® because of its outstanding friction and wear behaviour. Nevertheless, cobalt is an element which activates in the reactor leading to complex management of safety during reactor maintenance and mainly decommissioning. As a consequence, a collaborative work between CEA, EDF and FRAMATOME has been launched for selecting promising cobalt-free hardfacing alloys for the 600 MWe Sodium-cooledFast breeder reactor project named ASTRID. Several nickel-base alloys have been selected from literature review then deposited by Plasma Transferred Arc or Laser Cladding on 17Cr austenitic stainless steel 316L(N) according to RCC-MRx Code (AFCEN Code). Among the numerous properties required for qualifying their use as hardfacing alloys in SFR, good corrosion behaviour and good friction and wear behaviour in sodium are essential. First results on these properties are shown in this article. Firstly, the corrosion behaviour of all coatings was evaluated through exposure tests in puri ﬁed sodium for 5000 h at 400 °C. All coatings showed an acceptable corrosion behaviour in sodium. Finally, the friction and wear properties of one alloy candidate, NiCrBSi alloy, were studied in sodium in a dedicated designed facility. The in ﬂuence of the oxygen concentration in sodium on the friction and wear properties was evaluated.
The French Commission for Atomic Energy and Alternative Energy (CEA) in collaboration with its indus- trial partners develops Sodium-cooledFast Reactors (SFR) as industrial-scale demonstrators mainly guided by safety and operability objectives. In this paper, a SFR reactor associated to a nitrogen closed Brayton cycle for the Power Conversion System (PCS) is considered. In incidental and accidental con- ditions, the operation of reactor must be defined to keep it under control and to fulfil safety requirements. This paper is dedicated to an alternative procedure to control a Loss Of Off-site Power (LOOP). Usually, in case of LOOP, the SFR standard procedure relies on passive Decay Heat Removal (DHR) systems to cool down the primary circuit. In this paper, an alternative solution substitutes the latter by the gas Power Conversion System (PCS). The operation of the gas PCS required three regulations:
Key words: Electro-Magnetic Pump, ASTRID, SodiumFastReactor
1. Introduction SFR is one of the 4 th –generation Sodium-cooledfastreactor (GENIV) concepts selected to secure the nuclear fuel resources and to manage radioactive waste. In the June 2006, French Government submitted CEA to design studies of ASTRID prototype as a part of sustainable management of radioactive materials and wastes  in collaboration with industrial partners. ASTRID will be an industrial prototype with improving safety, operability and robustness against external hazards compared with previous SFRs for aim at a GENIV safety and operation.
4 CNRS,IN2P3,LPSC, F-74019 Annecy-le-Vieux Cedex, France
I. I NTRODUCTION
A new generation of nuclear reactors is investigated with the criteria of sustainability, enhanced safety, economics, and proliferation resistance. Among different designs of GEN IV reactors, Sodium-cooledFast Reactors (SFR) are chosen as reference systems due to the most extensive industrial experience and operational feedback available for this type. Under the research and development of SFR reactors, the domain of severe accident is addressed with high priority in the context of improved safety requirements. In the French frame of SFR safety research, oriented mainly around ASTRID reactor, an innovative severe accident mitigation architecture is being investigated. In this paper, the safety study approach and the mitigation strategy is introduced.
The core design of SodiumcooledFast Reactors uses sub-assembly ducts to support fuel elements. The positions of the core assemblies and their mechanical interaction have to be controlled owing to their effect on the core safety and performance, Briggs et al. (2014). The main operational constraints are related to core reactivity control, fuel handling and shutdown system reliability. As far as core assembly duct bowing due to thermal expansion, irradiation swelling and duct-to-duct interaction occurs during the reactor lifetime, a core restraint system is needed. In France, since Phénix and Super-Phénix, the core design relies on natural core restraint, Bernard (1979).
II. Brief History of SodiumFastReactor Develop- ment in France
The very ﬁrst tests conducted by the CEA using liquid metals date back to 1953. More than half a century later, the CEA has signiﬁcantly progressed in the ﬁeld of sodium- cooledfastreactor (SFR) technology. Such progress is reﬂected in the design, construction and operation of three fast breeder reactors: the experimental reactor (see Fig. 1): RAPSODIE; the prototype reactor: PHENIX; the commer- cial-size prototype reactor: SUPERPHENIX; and the Euro- pean project integrating feedback from operating fast breed- er reactor plants in Europe — EFR, European fastreactor. Since 2007, an important program has been launched by the three partners CEA, AREVA, and EDF in order to develop an innovative SFR concept. The purpose is to reach the construction of a prototype, an advanced sodium technolog- ical reactor for industrial demonstration (ASTRID), by 2020. 4) The whole SFR French program and development can be synthesized in one picture (see Fig. 1). In addition, Table 1 provides the most relevant news from the past ten years.
Saint-Paul lez Durance Cedex, France
Tel: +81-45-770-2410, Fax: +81-45-770-2317, Email of main author: firstname.lastname@example.org
Abstract – In the framework of the SFR (SodiumFastReactor) prototype called ASTRID
(Advance Sodium Technological Reactor for Industrial Demonstration), the large capacity Electro- Magnetic Pumps (EMP) as main circulating pumps on the intermediate sodium circuits is applied instead of mechanical pumps by CEA. The use of EMP has several decisive technological merits compared with mechanical pump in the reactor design, operation and maintenance. Nevertheless, some theoretical and technological developments have to be carried out in order to validate the design tools which take Magneto Hydro Dynamic (MHD) phenomena into account and the applicability of the EMP to the steady state and transient operating conditions of ASTRID. For their developments, a collaboration agreement between the CEA and TOSHIBA Corporation came into force to carry out a joint work program on the EMP for ASTRID design and development. CEA carried out the theoretical analysis, and the EMP experimental model is constructed by CEA to support these theoretical developments. This model consists of a middle-size annular EMP for the liquid metal sodium. The various testing program using this model has been started in 2016.And, TOSHIBA carried out the examination of design specification for ASTRID, an electromagnetic design, a structural design and various analyses. The structure design has been examined the placement of the sodium boundary and the withstand pressure, etc. And, if the thicknesses of the structure increase for withstand pressure, the pump efficiency falls because the loss of the electromagnetic force increases. Therefore the balance of withstand pressure and the efficiency has been considered by an electromagnetism design. This paper describes the design studies and experimental activities for the EMP development within the framework of the CEA- TOSHIBA collaborations.
To start, the attractive thermodynamical performance of a sCO 2 condensing cycle is reported in section
2 for SFRs conditions.
2. THERMODYNAMIC PERFORMANCE FOR AN SFR APPLICATION
In CYCLOP (CYCLe OPtimisation) , the CEA/DEN tool for power conversion cycle modelling, a cycle is represented by a set of fluid loops built from different components (heat source and sink, turbo-machines, heat exchangers) which are described by macroscopic parameters such as efficiency, pressure and heat losses if not adiabatic, as well as mass flow rate. Components are connected by points where thermodynamic states (temperature, pressure) are stored. This tool solves automatically the mass and energy balances for all components of the cycle from a minimum set of input data, allowing all cycle parameters to be quickly modified and optimised using the well-known deterministic Nelder-Mead algorithm . CYCLOP has been validated on Rankine steam cycles coupled to French Pressurized Water Reactors (such as CRUAS) and the French SFR Superphenix. It has also been extensively benchmarked in the frame of R&D programs, from the helium Brayton cycle applied to a Gas-cooledFastReactor concept  to the sCO 2 cycle for an SFR . In this code, the
KEYWORDS: refueling, fuel handling systems, availability, pantograph, external storage, flask
CEA, AREVA, and EDF have an extensive experience and signiﬁcant expertise in sodium-cooledfast reactors over the past 40 years of R&D and feedback experiments. 1) Some improvements are needed on the SFR to meet the GEN IV goals, and in particular the reduction of investment and operating costs: the fuel handling system (FHS) can be considered as an essential step in the reactor design. The reactor refuelling system provides the means of transporting, storing, and handling reactor core subassemblies. The sys- tem consists of the facilities and equipment needed to ac- complish the scheduled refuelling operations. The choice of a FHS impacts directly on the general design of the reactor vessel (primary vessel, storage, and ﬁnal cooling before going to reprocessing), its construction cost, and availability factor. The fuel handling design must take into account various issues particularly operating strategies such as core design and management and core conﬁguration. Moreover, the FHS will have to cope with safety assessment: a perma- nent cooling strategy to prevent fuel breakage, plus the
5.1.Mitigation or practical elimination of the core melt accident ?
To demonstrate that whole core melt cannot occur would be an interesting way if one could rely on its physical impossibility or a demonstration of practical elimination. This would make it possible to get rid of certain mitigation devices or even to simplify the overall design. It will be noted that this is a route that is tempted by some SFR designers. However, in the current state of knowledge, it seems difficult to present a reactor project without consideration of fourth level of defense in depth. At the present stage, the design objective is to maintain a high level of core melt prevention complemented by a mitigation capacity of the core melt accident.
2. Scope of Work
Before deining the several routes chosen in the past and that could be investigated for the future, a review of the diferent options has been carried out using the fastreactor database and recent technological development in SFR design. he considered options concern fuel handling systems (under rotating plugs), transfer assemblies options between reactor vessel and external storage, and also, in the particular case of fuel handling through gas corridor, fuel handling in the EVST. he work performed is a characterisation of solutions, a performance review, and an analysis of the main advantages and drawbacks of the options compared to a so-called Starting Reference Solution (SRS) based upon well-known French SFR options or some option already envisaged in French project, that is, EFR reactor [ 14 ]. he main features of the SRS are described below.
 B. Faure, P. Archier, J.-F. Vidal, and L. Buiron. “A 2D/1D Algorithm for Effective Cross- Section Generation in FastReactor Neutronic Transport Calculations.” Nuclear Science and Engineering, pp. 1–12 (2018).
 P. Gauth´e and P. Sciora. “Sensitivity studies of SFR unprotected transients with global neu- tronic feedback coefficients.” In Int. Conf. on Fast Reactors and Related Fuel Cycles (2017).  D. Wade and Y. Chang. “The integral fastreactor concept: physics of operation and safety.”
4. CONCLUSION AND PERSPECTIVES
In this paper, our new reference sub-assembly to core calculation scheme for SodiumFastReactor, called APOLLO3-SFR, has been presented. Each one of those two steps has been detailed and several elements of validation against continuous energy Monte Carlo have been shown. For the sub-assembly step, the results obtained with AP3-SFR scheme are very close (few dozens of pcm) to the TRIPOLI- 4
CEA is currently working on a new prototype of SodiumcooledFastReactor, ASTRID, which must demonstrate strong safety level. Spacer pads, which are small bosses stamped through faces of the hexagonal duct, play a central role in this framework. Indeed, this component maintains proper spacing of fuel subassemblies. This spacing impact significantly the core reactivity, which could be decreasing due to a hypothetical core-flowering phenomenon then, be increasing during the compaction phase. Therefore, the stiffness of spacer pads needs to be well characterized and improved in order to prevent from unsafe reactivity insertion due to radial core compaction. For this purpose, a finite element model is used in order to reproduce the stamping process and obtain a representative geometry of the bosses. Moreover, the mechanical response of this model is successfully compared to experimental results of duct crushing at nominal temperature condition. Finally, this model is used in order to achieve an optimization of the design of the spacer pads for ASTRID subassemblies. This study shows that stiffness of this component can significantly be improved only by modifying its geometry.
However, given the complexity of acoustic signals it is unlikely that detailed leak noise knowledge can be easily credited when designing a system for a new plant. Even the normal background noise of a new plant will not be known beforehand and will be depending both on the operating point of the reactor and other loud systems running nearby. This implies that the detection system will need to be able to recognize several different noise types. To the requirements cited above, we therefore propose the addition of two desirable system properties:
Fission chambers are used to monitor the neutron flux in a nuclear reactor, and therefore give an indication of the reactor power. While the replacement of argon by xenon led to a marked reduction in partial discharge activity, it is important to demonstrate that a xenon filled chamber would also perform adequately for the measurement of neutron flux in a reactor. Therefore, experiments were made using two chambers with the same geometry ( Fig. 3 ), and with the same filling pressure but with different fill gases (argon and xenon), and placed inside the ISIS [ 11 ] nuclear reactor.
Key Words: Valve, Sodium, Isolation
The ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration) project is a technological integration prototype intended for the industrial-scale safety and operation demonstration of 4th generation sodium-cooled generators. This integrated design reactor includes an intermediate sodium circuit comprised of 4 secondary loops. Each DN 700 diameter loop consists of an intermediate exchanger built into the reactor tank, a sodium/gas exchanger and an electromagnetic pump. To reinforce containment, an isolation valve is placed on each of these secondary loops, along with on each hot and cold leg. The nature of the selected coolant, the operating conditions and associated dimensions, along with the operating requirements associated with these components, imply an in-depth study of the technology and materials used and, in more general terms, of the overall operation of this specific valve.
The ‘integrated readiness level’ is also discussed in this paper with respect to access within the reactor block, fluids, positioning and maintenance aspects.
Within the framework of the future sodium-cooledfastreactor prototype called ASTRID, France has launched a large R&D program 1 on in-service inspection and repair (ISI&R) which has been identified as a difficult task to perform 2 (as sodium coolant is opaque, hot and highly chemically reactive) on the basis of experience feedback (French Phenix and Superphenix SFRs, as well as also foreign power plants). ISI&R is thus considered to be a major issue to be taken into account in order to improve the reactor’s safety (as inspection gives information on the actual reactor structure health), to consolidate its availability and to protect its associated investment.