core in case of variations in physical parameters characteristic of the cooling of the core such as flow rate reduction or temperature increase.
5.3.Decay heat removal
The total loss of the decay heat removal function is a situation that must be practically eliminated. The primary natural convection of the primary circuit and decay heat removal systems is a strong characteristic of the liquid metal reactors that shall be developed. A set of redundant, diversified systems, including diversified localization, and combining active and passive systems, shall be implemented in order to make negligible the risk of common mode failure. Thermal transfer from the main vessel is an interesting solution, but options using radiative heat transfers have limitations. Nevertheless the decay heat decreasing with time, these options can ensure the decay heat removal function a certain time after the reactor shutdown. It is also important to maintain a sufficient primary sodium inventory to ensure decay heat is removal under all circumstances, particularly in the event of vessels’ leakage. In case of a core melt accident, sufficient cooling means must be maintained to ensure post- accidental cooling of corium over the long term.
The French Atomic Energy and Alternative Energies Commission (CEA), EDF and AREVA French partners run a coordinated research program on Generation IV (Gen-IV) Sodium-cooledFast Reactors (SFRs) within the frame- work of the French Act, dated June 28th 2006, since 2007. Safety and reliability are essential needs for development and operation of Gen-IV SFRs. Develop- ment of a neutron flux monitoring system (NFMS) for the French sodium- cooledfastreactor is one of the R&D areas identified within this program. There is a need to confidently detect incidents in their earliest possible stages . Traditionally, SFRs use a combination of high temperature fission cham- bers and proportional counters at various locations outside the core. For the purpose of the French Gen-IV program, fission chambers have been identified as the best choice for monitoring the core in all states of reactor operations .
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 GENIV 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
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 GENIV 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.
During reactor start up, the PD signal count may potentially be on the same order of magnitude as the neutron signal count [ 4 ], which increases the uncertainty of the reactor power measurement. Fission chambers may also be used for the detection of fuel cladding failures which is based on very precise neutron count in the sodium coolant circulation lines. Reduced reliability of the neutron count from the HTFCs due to partial discharge activity thus represents a security risk for nuclear reactor operation.
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:
- It is playing an essential role in core surveillance and core monitoring.
- It is the subject of significant level of temperature (core outlet mean temperature 550°C with some extrema at 575°C) and thermomechanical solicitations. As a consequence the justification of its lifetime by thermomechanical calculations is a critical technical point that remains today challenging. At the same time even if this components is designed to be replaceable it has to be designed for the whole reactor lifetime (60 years) and its replacement must be envisaged only in exceptional situation to avoid losing the whole reactor investment.
7.2 Key results
Activities carried out in the frames of the VINCO project allowed to strengthen the links between the partners, establish running cooperation, especially in the ﬁeld of simulation capabilities in participating institutions, initiate common educational and training actions and exchange the practices of experimental works in hot cell laboratories. Financial and legal framework analysis in V4 countries carried out within the project helped to identify the possible international cooperation schemes in V4 countries. Mutual learning and exchange of scienti ﬁc staff between the laboratories took place, mainly in form of benchmark learning exercises on both, the neutronic and the thermo- hydraulic analyses and were devoted to the development of input models as well as the ef ﬁcient use of various calculation tools utilized by different users. Several joint events were organized, such as School, workshops and exchange visits. An important part of the project was related to educational issues. Database of (nuclear) Educational Resources has been prepared and a brochure on Generation-IV technology prepared and printed. Finally, communication campaigns were organized to provide the information about nuclear technology for a broader public and establish contact with decision makers in the V4 Region.
In this context, the sCO 2 cycle is viewed as a potential candidate for advanced power conversion
systems since it is claimed to avoid most of the problems of the Rankine steam and other Brayton gas cycles while retaining many of their advantages : i) high efficiency due to the low compression work required in the reduced compressibility region near the critical point; ii) compact turbomachinery resulting from the high density and low heat capacity of the working fluid; iii) simpler system layout than a Rankine steam cycle; iv) lower performance sensitivity to the system pressure losses than a Brayton (eg. nitrogen) cycle. The past decade has seen a growing interest in the sCO 2 cycle through a
Due to its low fissile content, Pu from spent MOX fuels is sometimes regarded as not recyclable in LWR. Based on the existing French nuclear infrastructure (La Hague reprocessing plant and MELOX MOX manufacturing plant), AREVA and the CEA have evaluated the conditions of Pu multirecycling in a 100% LWR fleet. As France is currently supporting a FastReactor prototype project, scenario studies have also been conducted to evaluate the contribution of a 600 MWe SFR in the LWR fleet.
CEA – CEA Paris-Saclay, Gif-sur-Yvette Cedex, France
Received: 12 May 2020 / Received in ﬁnal form: 31 July 2020 / Accepted: 18 August 2020
Abstract. From 2010 to 2019, the French Alternative Energies and Atomic Commission (CEA) associated with industrial partners realized the Basic Design of a prototype SodiumFastReactor. This project was called ASTRID (ASTRID for Advanced Sodium Technological Reactor for Industrial Demonstration). ASTRID design studies were ﬁnanced through governmental funds until the end of the basic design. These funds covered also the design studies for the core manufacturing workshop, the refurbishment or construction of large test loops. One year before the term of this Basic Design phase (in 2018), industrial partners, CEA and the French State conducted a review of fast neutrons reactors and fuel cycle strategy. The review which is now translated into the Multiannual Energy Program concluded that the perspective of industrial deployment of Fast Reactors is more distant. Yet it has been concluded to keep this option open, requiring to maintain competences, and to progress on technological barriers and further develop know-how. The strategy for complete closure of nuclear fuel cycle is maintained as a long-term sustainability objective (in the second half of the 21st century). Therefore, as a direct consequence of this decision, the ASTRID project stopped at the end of 2019 at its Basic Design phase. Quickly the question raised on the Knowledge Management (KM) and Know-How capitalization of the huge amount of studies and results realized during ten years (around 23 000 technical documents). Moreover the challenge was to realize this KM process in less than one year, before the ASTRID project team deﬁnitive split. The paper is presenting an innovative KM methodology which has been created and speciﬁcally performed on the ASTRID project. It is based on a series of interviews and video recordings, all transformed into some New KM tools called “MOOK” (MOOK for Management of Organized Online Knowledge). All these MOOKs considered as “data rich contents” are then inter-connected and linked by the ASTRID Product Breakdown Structure to some fundamental documents, for a comprehensive and quick mapping of the project. They ﬁnally form an ef ﬁcient KM tool recorded in a PLM Software (PLM for Product Lifecycle Management). Thus the ASTRID project team has realized a high level and easy-to-use “GPS” (Global Positioning System) tool to keep the ASTRID history, context, knowledge and know-how for years. This KM methodology can be easily adapted to other nuclear projects and needs.
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.
2.3.Reference rotatable seals layout and RST operational principles
The innovative layout of seals proposed by the CEA-TECHNETICS Sealing Laboratory to replace the liquid-metal seals is based on a combination of an inflatable and a massive dynamic elastomer seals. Both seals, mounted side-by-side, are working in the axial direction providing independent functions. In this design the dynamic sealing of the reactor vessel cover gas during the up/down vertical movements of the RP is ensure by the inflatable seal, while the massive double lips-seals (inverted-π shape) once compressed at the end of the translation ensure the sealing during the RP rotation. The RST reproduces the RP cinematic, starting with pressurization of the inflatable seal followed by a vertical movement of the seals holder plate up to a position where the rotatable seals is compressed by the upper plate. After deflation of the seal the upper plate can be set in rotation. The enclosed volume V E (Fig. 3) delimited by the inflatable and rotatable seals and the free-surface of the liquid in the casing can be pressurized at a value corresponding to the standard pressure of the vessel gas cover.
Carey ( 1979 ) had investigated the factors influencing background noise of a reactor. Data were collected over 17 months period and then analysed to determine the linear average spectra, auto and cross power spectra. Acoustic sound power is estimated and compared with results. It was shown that flow velocity influences acoustic power level and noise field can be considered as weakly stationary. Hayashi et al. ( 1996 ) developed a twice squaring method for real time sodium boiling detection. In their method, signal to noise ratio is enhanced by non-linear amplification of a band limited signal. Band-pass frequency is selected from PSD graphs, focussing on pulsive nature of boiling signal. It consists of five steps: band-pass filtering, squaring, another band-pass filtering and squaring and integration. A low pass filter is then applied to obtain the feature signal. The threshold for boiling detection is later calculated from the mean and the standard deviation of the feature signal in non-boiling conditions. This approach is successful if the mean and the standard deviation are always the same for signals. But if not, it sets up the problem of choosing an adequate value for the detection threshold.
In order to reach the strict safety criteria expected for nuclear reactors, high fidelity simulation tools are required in all the disciplines involved in the reactor physics analysis (neutronics, system thermo-hydraulics and mechanics). When it comes to model the dynamic behavior of a system in incidental or accidental situations, bridges need to be built at the interfaces between the codes. Despite all physical phenomena are tightly connected, the complexity of the problem (non-linearity, range of time scales, number of unknowns. . . ) leads to choose pragmatic approaches in which the coupling information is collapsed into a set of lumped parameters. In particular, neutronic calcula- tion methodologies are often based on point kinetics (PK) models in which the lumped parameters are pre-computed prior to the transient simulation using a “best-estimate” critical neutron transport calculation.