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Schyns M. - One of the best experts on this subject based on the ideXlab platform.
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In-Pile Qualification of the Fast-Neutron-Detection-System
EDP Sciences, 2018Co-Authors: Fourmentel D., Villard J-f., Destouches C., Geslot B., Vermeeren L., Schyns M.Abstract:In order to improve measurement techniques for neutron flux assessment, a unique system for online measurement of fast neutron flux has been developed and recently qualified in-pile by the French Alternative Energies and Atomic Energy Commission (CEA) in cooperation with the Belgian Nuclear Research Centre (SCK•ECEN). The Fast-Neutron-Detection-System (FNDS) has been designed to monitor accurately high-energy neutrons flux (E > 1 MeV) in typical Material Testing Reactor conditions, where overall neutron flux level can be as high as 1015 n.cm-2.s-1 and is generally dominated by thermal neutrons. Moreover, the neutron flux is coupled with a high gamma flux of typically a few 1015 γ.cm-2.s-1, which can be highly disturbing for the online measurement of neutron fluxes. The patented FNDS system is based on two detectors, including a miniature fission chamber with a special fissile material presenting an energy threshold near 1 MeV, which can be 242Pu for MTR conditions. Fission chambers are operated in Campbelling mode for an efficient gamma rejection. FNDS also includes a specific software that processes measurements to compensate online the fissile material depletion and to adjust the sensitivity of the detectors, in order to produce a precise evaluation of both thermal and fast neutron flux even after long term irradiation. FNDS has been validated through a two-step experimental program. A first set of tests was performed at BR2 reactor operated by SCK•CEN in Belgium. Then a second test was recently completed at ISIS reactor operated by CEA in France. FNDS proved its ability to measure online the fast neutron flux with an overall accuracy better than 5%
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In-pile qualification of the fast-neutron-detection-system
'EDP Sciences', 2017Co-Authors: Barbot L., Fourmentel D., Villard J-f., Destouches C., Geslot B., Vermeeren L., Schyns M.Abstract:Cet article est issu de la conférence "ANIMMA 2017" (Advancements in Nuclear Instrumentation Measurements Methods and their Applications) tenue à Liège du 19 au 23 juin 2017International audienceIn order to ensure the quality and the relevance of irradiation programs in the future Jules Horowitz Reactor (JHR), the French Alternative Energies and Atomic Energy Commission (CEA) has significantly increased its Research and development effort in the field of in-pile instrumentation during the last decade. Major progresses have thus been achieved in the capability to perform accurate in-pile measurements using reliable and updated techniques. A significant part of this effort have been conducted in the framework of the Joint Instrumentation Laboratory between the CEA and the Belgian Nuclear Research Centre (SCKCEN).In order to improve measurement techniques for neutron flux assessment, a unique system for online measurement of fast neutron flux has been developed and recently qualified in-pile. The Fast-Neutron-Detection-System (FNDS) has been designed to monitor accurately high-energy neutrons flux (E > 1 MeV) in typical Material Testing Reactor conditions, where overall neutron flux level can be as high as 10$^{15}$ n.cm$^{-2}$.s$^{-1}$ and is generally dominated by thermal neutrons. Moreover, the neutron flux is accompanied by a high gamma flux of typically a few 10$^{15}$$\gamma$cm$^{-2}$.s$^{-1}$, which can be highly disturbing for the online measurement of neutron fluxes. The patented FNDS system is based on two detectors allowing the simultaneous detection of both thermal and fast neutron flux. Thermal neutrons can be measured using a Self-Powered-Neutron-Detector (SPND) or a $^{235}$U miniature fission chamber, while fast neutron detection requires a miniature fission chamber with a special fissile material presenting an appropriate energy threshold, which can be $^{242}$Pu for MTR conditions. Fission chambers are operated in Campbelling mode for an efficient gamma rejection. FNDS also includes a specific software that processes measurements to compensate online the fissile material depletion and to adjust the sensitivity of the detectors, in order to produce a precise evaluation of both thermal and fast neutron flux even after long term irradiation. FNDS has been validated through a two-step experimental program. A first set of tests was performed at BR2 reactor operated by SCKCEN in Belgium. Two FNDS prototypes were operated in-pile during nearly 1000 hours. These tests exhibited the consistency of the measurement of thermal to fast neutron flux ratio with MCNP calculations, as well as the right compensation of fissile material depletion. Then a second test was recently completed at ISIS reactor operated by CEA in France. For this irradiation, FNDS signal was compared to reference thermal and fast neutron flux measurements using activation dosimeters analyzed under COFRAC Quality Certification. During this latter test, FNDS proved its ability to measure online the fast neutron flux with an overall accuracy better than 5%.This paper describes the innovative features of FNDS and discusses the results of its final in-pile qualification. FNDS is now operational and is assumed to be the first and unique acquisition system able to provide an online measurement of the fast neutron flux in MTR conditions. This system will of course be used to perform spectral neutron characterization of JHR channels, but it may also be implemented in future irradiation experiments, for a better and real-time evaluation of the fast neutron flux received by material and fuel samples
Francesco Belloni - One of the best experts on this subject based on the ideXlab platform.
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coupled system thermal hydraulic cfd analysis of a protected loss of flow transient in the myrrha reactor
Annals of Nuclear Energy, 2018Co-Authors: Antonio Toti, Jan Vierendeels, Francesco BelloniAbstract:Abstract Within the MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) project, the Belgian Nuclear Research Centre SCK•CEN is developing and designing a flexible irradiation facility, configured as an accelerator driven system (ADS) with a MOX fueled reactor core, able to operate in both critical and sub-critical modes. The system design features a compact pool-type primary cooling system operating with molten Lead-Bismuth Eutectic (LBE). With regard to the thermal-hydraulic design and safety assessment of the installation, a major challenge is represented by the complex coolant flow field characterizing the large open regions of the primary vessel, namely the cold and hot plenum, with the presence of pronounced three-dimensional phenomena that may impact the evolution of accidental transients such as loss of flow events. In order to have a realistic representation of such effects, a coupled system thermal-hydraulic/CFD model of the MYRRHA reactor is developed and presented in this paper. The proposed multi-scale methodology, which couples the 1D system code RELAP5-3D to the CFD code FLUENT, is based on domain decomposition and a novel implicit numerical scheme is developed. The coupled reactor-scale computational model is applied in this work to the analysis of a postulated protected loss of flow (PLOF) accident, and preliminary validated against RELAP5-3D stand-alone solution data. The results of the analysis were found in agreement, demonstrating the capability of the tool to perform integral simulation taking into account 3D flows and local phenomena.
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extension and application on a pool type test facility of a system thermal hydraulic cfd coupling method for transient flow analyses
Nuclear Engineering and Design, 2018Co-Authors: Antonio Toti, Jan Vierendeels, Francesco BelloniAbstract:Abstract The development of multi-scale modeling capabilities through the use of coupled system thermal-hydraulic and CFD codes is addressed in this paper, which presents the extension and application of a partitioned, domain decomposition computational method coupling the 1D code RELAP5-3D and the CFD code FLUENT. An implicit coupling scheme for the solution of the flow field, based on a Quasi-Newton algorithm, is developed and tested on a case with multiple coupling interfaces. Moreover, modeling capabilities of the tool are enhanced through the implementation of thermal coupling to compute conjugate heat transfer phenomena. In this work, a coupled model has been developed for the analysis of the pool-type test facility E-SCAPE, currently under commissioning at the Belgian Nuclear Research Centre SCK•CEN. The installation represents a thermal-hydraulic scaled model of the MYRRHA reactor, with an electrical core simulator, cooled by Lead-Bismuth Eutectic (LBE). The coupling tool is applied on the analysis of a total loss of flow (LOF) transient, involving the complex transition from forced to natural circulation flow in the primary system. The simulation results, compared against stand-alone system thermal-hydraulics (STH) simulation data, highlighted that the transient behavior of a pool-type system in natural circulation is characterized by complex multi-dimensional flow and temperature fields, difficult to predict by 1D STH codes. The study also confirmed the capability of the developed tool to predict the impact of such phenomena on the system behavior, and to capture the development of thermal stratification in a plenum at low flow condition. The planned experimental tests will be used for the validation of the tool, in the view of its use for design and licensing activities.
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improved numerical algorithm and experimental validation of a system thermal hydraulic cfd coupling method for multi scale transient simulations of pool type reactors
Annals of Nuclear Energy, 2017Co-Authors: Antonio Toti, Jan Vierendeels, Francesco BelloniAbstract:Abstract The paper describes the development and validation of a coupling methodology between the best-estimate system thermal-hydraulic code RELAP5-3D and the CFD code FLUENT, conceived for high fidelity plant-scale safety analyses of pool-type reactors. The computational tool is developed to assess the impact of three-dimensional phenomena occurring in accidental transients such as loss of flow (LOF) in the Research reactor MYRRHA, currently in the design phase at the Belgian Nuclear Research Centre, SCK•CEN. A partitioned, implicit domain decomposition coupling algorithm is implemented, in which the coupled domains exchange thermal-hydraulics variables at coupling boundary interfaces. Numerical stability and interface convergence rates are improved by a novel interface Quasi-Newton algorithm, which is compared in this paper with previously tested numerical schemes. The developed computational method has been assessed for validation purposes against the experiment performed at the test facility TALL-3D, operated by the Royal Institute of Technology (KTH) in Sweden. This paper details the results of the simulation of a loss of forced convection test, showing the capability of the developed methodology to predict transients influenced by local three-dimensional phenomena.
Antonio Toti - One of the best experts on this subject based on the ideXlab platform.
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coupled system thermal hydraulic cfd analysis of a protected loss of flow transient in the myrrha reactor
Annals of Nuclear Energy, 2018Co-Authors: Antonio Toti, Jan Vierendeels, Francesco BelloniAbstract:Abstract Within the MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) project, the Belgian Nuclear Research Centre SCK•CEN is developing and designing a flexible irradiation facility, configured as an accelerator driven system (ADS) with a MOX fueled reactor core, able to operate in both critical and sub-critical modes. The system design features a compact pool-type primary cooling system operating with molten Lead-Bismuth Eutectic (LBE). With regard to the thermal-hydraulic design and safety assessment of the installation, a major challenge is represented by the complex coolant flow field characterizing the large open regions of the primary vessel, namely the cold and hot plenum, with the presence of pronounced three-dimensional phenomena that may impact the evolution of accidental transients such as loss of flow events. In order to have a realistic representation of such effects, a coupled system thermal-hydraulic/CFD model of the MYRRHA reactor is developed and presented in this paper. The proposed multi-scale methodology, which couples the 1D system code RELAP5-3D to the CFD code FLUENT, is based on domain decomposition and a novel implicit numerical scheme is developed. The coupled reactor-scale computational model is applied in this work to the analysis of a postulated protected loss of flow (PLOF) accident, and preliminary validated against RELAP5-3D stand-alone solution data. The results of the analysis were found in agreement, demonstrating the capability of the tool to perform integral simulation taking into account 3D flows and local phenomena.
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extension and application on a pool type test facility of a system thermal hydraulic cfd coupling method for transient flow analyses
Nuclear Engineering and Design, 2018Co-Authors: Antonio Toti, Jan Vierendeels, Francesco BelloniAbstract:Abstract The development of multi-scale modeling capabilities through the use of coupled system thermal-hydraulic and CFD codes is addressed in this paper, which presents the extension and application of a partitioned, domain decomposition computational method coupling the 1D code RELAP5-3D and the CFD code FLUENT. An implicit coupling scheme for the solution of the flow field, based on a Quasi-Newton algorithm, is developed and tested on a case with multiple coupling interfaces. Moreover, modeling capabilities of the tool are enhanced through the implementation of thermal coupling to compute conjugate heat transfer phenomena. In this work, a coupled model has been developed for the analysis of the pool-type test facility E-SCAPE, currently under commissioning at the Belgian Nuclear Research Centre SCK•CEN. The installation represents a thermal-hydraulic scaled model of the MYRRHA reactor, with an electrical core simulator, cooled by Lead-Bismuth Eutectic (LBE). The coupling tool is applied on the analysis of a total loss of flow (LOF) transient, involving the complex transition from forced to natural circulation flow in the primary system. The simulation results, compared against stand-alone system thermal-hydraulics (STH) simulation data, highlighted that the transient behavior of a pool-type system in natural circulation is characterized by complex multi-dimensional flow and temperature fields, difficult to predict by 1D STH codes. The study also confirmed the capability of the developed tool to predict the impact of such phenomena on the system behavior, and to capture the development of thermal stratification in a plenum at low flow condition. The planned experimental tests will be used for the validation of the tool, in the view of its use for design and licensing activities.
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improved numerical algorithm and experimental validation of a system thermal hydraulic cfd coupling method for multi scale transient simulations of pool type reactors
Annals of Nuclear Energy, 2017Co-Authors: Antonio Toti, Jan Vierendeels, Francesco BelloniAbstract:Abstract The paper describes the development and validation of a coupling methodology between the best-estimate system thermal-hydraulic code RELAP5-3D and the CFD code FLUENT, conceived for high fidelity plant-scale safety analyses of pool-type reactors. The computational tool is developed to assess the impact of three-dimensional phenomena occurring in accidental transients such as loss of flow (LOF) in the Research reactor MYRRHA, currently in the design phase at the Belgian Nuclear Research Centre, SCK•CEN. A partitioned, implicit domain decomposition coupling algorithm is implemented, in which the coupled domains exchange thermal-hydraulics variables at coupling boundary interfaces. Numerical stability and interface convergence rates are improved by a novel interface Quasi-Newton algorithm, which is compared in this paper with previously tested numerical schemes. The developed computational method has been assessed for validation purposes against the experiment performed at the test facility TALL-3D, operated by the Royal Institute of Technology (KTH) in Sweden. This paper details the results of the simulation of a loss of forced convection test, showing the capability of the developed methodology to predict transients influenced by local three-dimensional phenomena.
Jan Vierendeels - One of the best experts on this subject based on the ideXlab platform.
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coupled system thermal hydraulic cfd analysis of a protected loss of flow transient in the myrrha reactor
Annals of Nuclear Energy, 2018Co-Authors: Antonio Toti, Jan Vierendeels, Francesco BelloniAbstract:Abstract Within the MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) project, the Belgian Nuclear Research Centre SCK•CEN is developing and designing a flexible irradiation facility, configured as an accelerator driven system (ADS) with a MOX fueled reactor core, able to operate in both critical and sub-critical modes. The system design features a compact pool-type primary cooling system operating with molten Lead-Bismuth Eutectic (LBE). With regard to the thermal-hydraulic design and safety assessment of the installation, a major challenge is represented by the complex coolant flow field characterizing the large open regions of the primary vessel, namely the cold and hot plenum, with the presence of pronounced three-dimensional phenomena that may impact the evolution of accidental transients such as loss of flow events. In order to have a realistic representation of such effects, a coupled system thermal-hydraulic/CFD model of the MYRRHA reactor is developed and presented in this paper. The proposed multi-scale methodology, which couples the 1D system code RELAP5-3D to the CFD code FLUENT, is based on domain decomposition and a novel implicit numerical scheme is developed. The coupled reactor-scale computational model is applied in this work to the analysis of a postulated protected loss of flow (PLOF) accident, and preliminary validated against RELAP5-3D stand-alone solution data. The results of the analysis were found in agreement, demonstrating the capability of the tool to perform integral simulation taking into account 3D flows and local phenomena.
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extension and application on a pool type test facility of a system thermal hydraulic cfd coupling method for transient flow analyses
Nuclear Engineering and Design, 2018Co-Authors: Antonio Toti, Jan Vierendeels, Francesco BelloniAbstract:Abstract The development of multi-scale modeling capabilities through the use of coupled system thermal-hydraulic and CFD codes is addressed in this paper, which presents the extension and application of a partitioned, domain decomposition computational method coupling the 1D code RELAP5-3D and the CFD code FLUENT. An implicit coupling scheme for the solution of the flow field, based on a Quasi-Newton algorithm, is developed and tested on a case with multiple coupling interfaces. Moreover, modeling capabilities of the tool are enhanced through the implementation of thermal coupling to compute conjugate heat transfer phenomena. In this work, a coupled model has been developed for the analysis of the pool-type test facility E-SCAPE, currently under commissioning at the Belgian Nuclear Research Centre SCK•CEN. The installation represents a thermal-hydraulic scaled model of the MYRRHA reactor, with an electrical core simulator, cooled by Lead-Bismuth Eutectic (LBE). The coupling tool is applied on the analysis of a total loss of flow (LOF) transient, involving the complex transition from forced to natural circulation flow in the primary system. The simulation results, compared against stand-alone system thermal-hydraulics (STH) simulation data, highlighted that the transient behavior of a pool-type system in natural circulation is characterized by complex multi-dimensional flow and temperature fields, difficult to predict by 1D STH codes. The study also confirmed the capability of the developed tool to predict the impact of such phenomena on the system behavior, and to capture the development of thermal stratification in a plenum at low flow condition. The planned experimental tests will be used for the validation of the tool, in the view of its use for design and licensing activities.
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improved numerical algorithm and experimental validation of a system thermal hydraulic cfd coupling method for multi scale transient simulations of pool type reactors
Annals of Nuclear Energy, 2017Co-Authors: Antonio Toti, Jan Vierendeels, Francesco BelloniAbstract:Abstract The paper describes the development and validation of a coupling methodology between the best-estimate system thermal-hydraulic code RELAP5-3D and the CFD code FLUENT, conceived for high fidelity plant-scale safety analyses of pool-type reactors. The computational tool is developed to assess the impact of three-dimensional phenomena occurring in accidental transients such as loss of flow (LOF) in the Research reactor MYRRHA, currently in the design phase at the Belgian Nuclear Research Centre, SCK•CEN. A partitioned, implicit domain decomposition coupling algorithm is implemented, in which the coupled domains exchange thermal-hydraulics variables at coupling boundary interfaces. Numerical stability and interface convergence rates are improved by a novel interface Quasi-Newton algorithm, which is compared in this paper with previously tested numerical schemes. The developed computational method has been assessed for validation purposes against the experiment performed at the test facility TALL-3D, operated by the Royal Institute of Technology (KTH) in Sweden. This paper details the results of the simulation of a loss of forced convection test, showing the capability of the developed methodology to predict transients influenced by local three-dimensional phenomena.
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benchmark exercise for fluid flow simulations in a liquid metal fast reactor fuel assembly
Nuclear Engineering and Design, 2016Co-Authors: Elia Merzari, K Van Tichelen, Jan Vierendeels, Paul Fischer, Haomin Yuan, S Keijers, J De Ridder, Joris Degroote, H Doolaard, V R GopalaAbstract:Abstract As part of a U.S. Department of Energy International Nuclear Energy Research Initiative (I-NERI), Argonne National Laboratory (Argonne) is collaborating with the Dutch Nuclear Research and consultancy Group (NRG), the Belgian Nuclear Research Centre (SCK·CEN), and Ghent University (UGent) in Belgium to perform and compare a series of fuel-pin-bundle calculations representative of a fast reactor core. A wire-wrapped fuel bundle is a complex configuration for which little data is available for verification and validation of new simulation tools. UGent and NRG performed their simulations with commercially available computational fluid dynamics (CFD) codes. The high-fidelity Argonne large-eddy simulations were performed with Nek5000, used for CFD in the Simulation-based High-efficiency Advanced Reactor Prototyping (SHARP) suite. SHARP is a versatile tool that is being developed to model the core of a wide variety of reactor types under various scenarios. It is intended both to serve as a surrogate for physical experiments and to provide insight into experimental results. Comparison of the results obtained by the different participants with the reference Nek5000 results shows good agreement, especially for the cross-flow data. The comparison also helps highlight issues with current modeling approaches. The results of the study will be valuable in the design and licensing process of MYRRHA, a flexible fast Research reactor under design at SCK·CEN that features wire-wrapped fuel bundles cooled by lead-bismuth eutectic.
Fourmentel D. - One of the best experts on this subject based on the ideXlab platform.
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In-Pile Qualification of the Fast-Neutron-Detection-System
EDP Sciences, 2018Co-Authors: Fourmentel D., Villard J-f., Destouches C., Geslot B., Vermeeren L., Schyns M.Abstract:In order to improve measurement techniques for neutron flux assessment, a unique system for online measurement of fast neutron flux has been developed and recently qualified in-pile by the French Alternative Energies and Atomic Energy Commission (CEA) in cooperation with the Belgian Nuclear Research Centre (SCK•ECEN). The Fast-Neutron-Detection-System (FNDS) has been designed to monitor accurately high-energy neutrons flux (E > 1 MeV) in typical Material Testing Reactor conditions, where overall neutron flux level can be as high as 1015 n.cm-2.s-1 and is generally dominated by thermal neutrons. Moreover, the neutron flux is coupled with a high gamma flux of typically a few 1015 γ.cm-2.s-1, which can be highly disturbing for the online measurement of neutron fluxes. The patented FNDS system is based on two detectors, including a miniature fission chamber with a special fissile material presenting an energy threshold near 1 MeV, which can be 242Pu for MTR conditions. Fission chambers are operated in Campbelling mode for an efficient gamma rejection. FNDS also includes a specific software that processes measurements to compensate online the fissile material depletion and to adjust the sensitivity of the detectors, in order to produce a precise evaluation of both thermal and fast neutron flux even after long term irradiation. FNDS has been validated through a two-step experimental program. A first set of tests was performed at BR2 reactor operated by SCK•CEN in Belgium. Then a second test was recently completed at ISIS reactor operated by CEA in France. FNDS proved its ability to measure online the fast neutron flux with an overall accuracy better than 5%
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In-pile qualification of the fast-neutron-detection-system
'EDP Sciences', 2017Co-Authors: Barbot L., Fourmentel D., Villard J-f., Destouches C., Geslot B., Vermeeren L., Schyns M.Abstract:Cet article est issu de la conférence "ANIMMA 2017" (Advancements in Nuclear Instrumentation Measurements Methods and their Applications) tenue à Liège du 19 au 23 juin 2017International audienceIn order to ensure the quality and the relevance of irradiation programs in the future Jules Horowitz Reactor (JHR), the French Alternative Energies and Atomic Energy Commission (CEA) has significantly increased its Research and development effort in the field of in-pile instrumentation during the last decade. Major progresses have thus been achieved in the capability to perform accurate in-pile measurements using reliable and updated techniques. A significant part of this effort have been conducted in the framework of the Joint Instrumentation Laboratory between the CEA and the Belgian Nuclear Research Centre (SCKCEN).In order to improve measurement techniques for neutron flux assessment, a unique system for online measurement of fast neutron flux has been developed and recently qualified in-pile. The Fast-Neutron-Detection-System (FNDS) has been designed to monitor accurately high-energy neutrons flux (E > 1 MeV) in typical Material Testing Reactor conditions, where overall neutron flux level can be as high as 10$^{15}$ n.cm$^{-2}$.s$^{-1}$ and is generally dominated by thermal neutrons. Moreover, the neutron flux is accompanied by a high gamma flux of typically a few 10$^{15}$$\gamma$cm$^{-2}$.s$^{-1}$, which can be highly disturbing for the online measurement of neutron fluxes. The patented FNDS system is based on two detectors allowing the simultaneous detection of both thermal and fast neutron flux. Thermal neutrons can be measured using a Self-Powered-Neutron-Detector (SPND) or a $^{235}$U miniature fission chamber, while fast neutron detection requires a miniature fission chamber with a special fissile material presenting an appropriate energy threshold, which can be $^{242}$Pu for MTR conditions. Fission chambers are operated in Campbelling mode for an efficient gamma rejection. FNDS also includes a specific software that processes measurements to compensate online the fissile material depletion and to adjust the sensitivity of the detectors, in order to produce a precise evaluation of both thermal and fast neutron flux even after long term irradiation. FNDS has been validated through a two-step experimental program. A first set of tests was performed at BR2 reactor operated by SCKCEN in Belgium. Two FNDS prototypes were operated in-pile during nearly 1000 hours. These tests exhibited the consistency of the measurement of thermal to fast neutron flux ratio with MCNP calculations, as well as the right compensation of fissile material depletion. Then a second test was recently completed at ISIS reactor operated by CEA in France. For this irradiation, FNDS signal was compared to reference thermal and fast neutron flux measurements using activation dosimeters analyzed under COFRAC Quality Certification. During this latter test, FNDS proved its ability to measure online the fast neutron flux with an overall accuracy better than 5%.This paper describes the innovative features of FNDS and discusses the results of its final in-pile qualification. FNDS is now operational and is assumed to be the first and unique acquisition system able to provide an online measurement of the fast neutron flux in MTR conditions. This system will of course be used to perform spectral neutron characterization of JHR channels, but it may also be implemented in future irradiation experiments, for a better and real-time evaluation of the fast neutron flux received by material and fuel samples
