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International audience; This article presents a procedure for optimizing the charge comparison method (CCM) used for pulse shape discrimination (PSD). Without prior knowledge of the signals or the readout system, our procedure automatically optimizes the integration periods maximizing the discrimination ability of the radiation detector. This procedure is innovative in its adaptability and automation without being complicated to implement on a standard computer. Another advantage of this approach is the possibility to use it even if the operation of the readout system and the recording process of the signal is not fully known. Therefore, it enables all detection systems generating signals whose temporal evolution depends on the origin to optimize the integration periods of the CCM. Our procedure is based on verifying that two criteria are met in terms of the number of components and the correlation of Gaussian fits made on the distribution of the tail-to-total integral resulting from the CCM. We tested the procedure for different application cases. First, the optimization of the integration periods of the CCM was performed for the discrimination between fast neutrons and gamma rays with a plastic scintillator and a silicon photomultiplier (SiPM) in the energy range [250 keVee; 4.5 MeVee]. The integration periods, from the laboratory's experience with photomultiplier tubes (PMTs) and plastic scintillators, gave a Figure of Merit (FoM) of 0.58 corresponding to a rejection ratio (RR) of 8.6%. The procedure improved the FoM up to 0.88 corresponding to a RR of 1.9%. We also applied the procedure to the discrimination between beta and gamma rays with a PMT and a phoswich organic detector and to the discrimination between signals collected from neutrons or partial discharges within a fission chamber.

A new fission chamber was developed and built for prompt-fission-neutron spectra (PFNS) measurements in neutron-induced fission of actinides, and is described in Taieb et al. (2016). It allowed us to minimize incident- and outgoing-neutron scattering, and optimize both time resolution and alpha to fission fragment discrimination. The fission chamber was validated and used for 238U and 252Cf PFNS measurements. In order to measure PFNS of samples with a natural alpha activity of the order of 10 MBq, further improvements to the detector were done and are presented in this work. Tests of the improved fission chamber lead to a fission fragment detection efficiency of 91% for a 14.41 MBq 240Pu deposit, and 99% for a 247 kBq 242Pu deposit. The fission chamber was successfully used for 239Pu PFNS measurements at the Weapons Neutron Research facility of the Los Alamos Neutron Science Center (WNR@LANSCE).

The availability of neutron fields with a high neutron flux, suitable for irradiation testing of nuclear instrumentation detectors relevant for applications in nuclear facilities such as material testing reactors (MTRs), nuclear power reactors and future fusion reactors is becoming increasingly limited. Over the last several years there has been increased interest in the experimental capabilities of the 250 kW Jožef Stefan Institute (JSI) TRIGA research reactor for such applications, however, the maximal achievable neutron flux in steady-state operation mode falls short of MTR-relevant conditions. The JSI TRIGA reactor can also operate in pulse mode, with a maximal achievable peak power of approximately 1 GW, for a duration of a few ms. A collaboration project between the JSI and the French Atomic and Alternative Energy Commission (CEA) was initiated to investigate absolute neutron flux measurements at very high neutron flux levels in reactor pulse operation. Such measurements will be made possible by special CEA-developed miniature fission chambers and modern data acquisition systems, supported by the JSI TRIGA instrumentation and activation dosimetry. Additionally, measurements of the intensity of Cherenkov light are proposed and being investigated as an alternative experimental method. This paper presents the preparatory activities for an exhaustive experimental campaign, which were carried out in 2019-2020, consisting of test measurements with not fully appropriate fission chambers, activation dosimetry and silicon photomultipliers (SiPMs) The presented results provide useful and promising experimental indications relevant for the design of the experimental campaign.

The work presented in this thesis deals with the characterization of gamma field inside a nuclear reactor by experiments and computational modelling. In the first part of the thesis an outline of the nuclear with neutrons and neutron transport. A description of high energy photon and electron reactions and importance of their coupling for accurate calculations of energy deposition. Particle transport equations are presented with emphasis on deriving adjoint operators used for variance reduction of Monte Carlo particle transport codes. Characterization of gamma radiation field using Monte Carlo transport codes only takes into account prompt gamma generation from fission, inelastic scattering and prompt (n,gamma) reactions. Previous evaluations suggest a roughly 30 % underestimation compared to measurements. A JSIR2S code package for delayed radiation field calculations has been developed and validated by numerous experiments. Characterization of neutron and prompt gamma radiation field inside the JSI TRIGA reactor core irradiation facilities was performed using the kerma approximation. The computational model was later expanded and the criticality source term translated to a fixed source for calculations of variance reduction parameters. The methodology has been validated by experiments, showing good agreement for neutrons, while underestimating the gamma field due to neglecting delayed radiation field. Several experimental campaigns were performed at JSI TRIGA reactor using fission and ionization chamber and Thermoluminescent dosimeters. An experimental procedure for estimation of the delayed gamma fraction was developed. Validation of the JSIR2S was performed on the above mentioned measurements, showing agreement within the uncertainty. use case on using the JSIR2S for calibration of semiconductor detectors in the JSI TRIGA reactor after reactor shut-down is described. The JSIR2S code package is also applied to shut-down dose rate calculations in fusion problems showing good agreement with experiments and similar two-step and single-step methodology codes for delayed radiation field characterization. Doktorsko delo obravnava eksperimentalno in računsko karakterizacijo polja žarkov gama v jedrskih reaktorjih. V prvem delu predstavimo jedrske reakcije z nevtroni, njihov transport in transport nastalih visoko energijskih fotonov in elektronov ter izpostavili pomembnost njihovega prelivanja za izračune deponirane energije. Predstavimo tudi splošno transportno enačbo s poudarkom na adjungiranih operatorjih, ki se uporabljajo za redukcijo variance v preračunih transporta delcev z metodo Monte Carlo. Pri uporabi metode za Monte Carlo transprot delcev se navadno upoštevajo samo promptne reakcije produkcije žarkov gama kot so fisija, neelastično sipanje in promptne reakcije (n,gamma). Meritve kažejo na 30 % podcenitev izračunov, kar pripisujemo zakasnelemu sevanju. Razvili smo programski paket JSIR2S za izračune polja zakasnelega sevanja. Paket smo validirali na vrsti eksperimentov na reaktorju IJS TRIGA ter na nekaterih fuzijskih eksperimentih za izračun hitrosti doze po zaustavitvi. Začetno karakterizacija polja nevtronov in promptnih žarkov gama v obsevalnih mestih sredice reaktorja IJS TRIGA smo izvedli v približku kerme. Razširjen računski model smo uporabili za izračune sevalnega polja zunaj biološkega ščita reaktorja. Nevtronski izvor smo iz preračunov kritičnosti prevedli na fiksni izvor, ki smo ga uporabili za določanje parametrov redukcije variance. Za polje nevtronov dobimo ujemanje znotraj negotovosti, polje žarkov gama pa podcenimo ker ne upoštevamo polja zakasnelega sevanja. Na reaktorju IJS TRIGA smo izvedli več eksperimentov z meritvami s fisijskimi in ionizacijskimi celicami ter termoluminescentnimi dozimetri. Razvili smo proceduro za eksperimentalno določanje prispevka polja zakasnelih žarkov gama. Pridobljene rezultate smo uporabili za validacijo JSIR2S, kjer dobimo ujemanje znotraj negotovosti. JSIR2S smo uporabili tudi za kalibracijo polprevodniških dozimetrov v polju žarkov gama po zaustavitvi reaktorja IJS TRIGA. JSIR2S smo aplicirali tudi na fuzijske probleme za izračune hitrosti doze po zaustavitvi reaktorja, ki kažejo na dobro ujemanje z eksperimenti oziroma z preračuni s podobnimi programskimi orodji z uporabo eno ali dvo-koračne metode.

International audience; Fuel debris have a distinct neutron signature that can be detected to locate the said debris in a damaged nuclear power plant. Neutron measurement in a damaged PCV environment is however submitted to severe deployments constraints, including a high-dose-rate gamma background and limited available space. The study was therefore oriented towards small fission chambers (FC), with U-235-enriched active substrates. To investigate the expected performance of the FC in various irradiation conditions, a numerical model of the detector head was built. We describe the elaboration and experimental calibration of the numerical model and the Monte Carlo study of the fission rate inside U-235 coatings per generated neutron. The evaluation of a representative calibration coefficient then allowed us to carry out a multi-parameter performance study of a FC underwater, aiming at computing an explicit response function linking, on the one hand, the activity and spatial distribution of neutron emitters in a water container, with, one the other hand, the expected count rates measured by a fission chamber as a function of its radial and axial position inside the water volume. The FC underwater behavior was subsequently corroborated by a measurement campaign on a FC response, set at different positions inside a water drum, as a function of its axial and radial distance to a Cf-252 neutron source attached near the center of the container. We finally present an approach in which fuel debris localization is defined as an Inverse Problem, solvable with a Maximum-Likelihood Expectation Maximization (ML-EM) iterative algorithm. The projector matrix is built by capitalization on the results of the previously consolidated numerical studies. The ML-EM was tested on simulated data sets with a varying number of active voxels. Our first results indicate that, for a thermal neutron flux in the order of 10 n.cm−2.s−1 at the detector, originating voxels are identified with a spatial resolution in the radial plane in the order of 10 to 100 cm2.

As the modernization of the nuclear instrumentation system progresses, research reactors have adopted digital wide-range neutron power measurement (DWRNPM) systems. These systems typically monitor the neutron flux across a range of over 10 decades. Because neutron detectors only measure the local neutron flux at their position, the local neutron flux must be converted to total reactor power through calibration, which involves mapping the local neutron flux level to a reference reactor power. Conventionally, the neutron power range is divided into smaller subranges because the neutron detector signal characteristics and the reference reactor power estimation methods are different for each subrange. Therefore, many factors should be considered when preparing the calibration procedure for DWRNPM channels. The main purpose of this work is to serve as a reference for performing the calibration of DWRNPM systems in research reactors. This work provides a comprehensive overview of the calibration of DWRNPM channels by describing the configuration of the DWRNPM system and by summarizing the theories of operation and the reference power estimation methods with their associated calibration procedure. The calibration procedure was actually performed during the commissioning of an open-pool type research reactor, and the results and experience are documented herein.

A new measurement is presented of the 6 Li(n, α )t cross section from 245 keV to 10 MeV using a 252Cf fission chamber with 6 LiI(Eu) and Cs 2 LiYCl 6 :Ce (CLYC) scintillators which act as both target and detector. Neutron energies are determined from the time of flight (TOF) method using the signals from spontaneous fission and reaction product recoil. Simulations of neutron downscatter in the crystals and fission chamber bring 6 Li(n, α )t cross section values measured with the 6 LiI(Eu) into agreement with previous experiments and evaluations, except for two resonances at 4.2 and 6.5 MeV introduced by ENDF/B-VII.1. Suspected neutron transport modeling issues cause the cross section values obtained with CLYC to be discrepant above 2 MeV.

International audience; This work aims at assessing the performance of a 235U enriched fission chamber in order to localize fuel debris, prior to dismantling operations, in a flooded primary containment vessel of a damaged nuclear reactor such as Fukushima Daiichi. Based on both a comprehensive scan of the environment and the detection of neutrons emitted by the melted core, fuel debris can be localized. In this paper, we carry out a simulation study using the MCNP6 code to investigate fission chamber response in the frame of fuel debris localization measurements in a damaged nuclear reactor. The CFUF34 fission chamber (manufactured by PHOTONIS) and the primary containment vessel of Fukushima Daiichi Unit 1 were chosen to conduct this work. Impact of different parameters were investigated with MCNP6, such as: neutron energy, water temperature, fission chamber position (altitude, lateral shift, and rotation), and sensitivity loss due to sediments potentially covering fuel debris. In summary, we show that fuel debris should be sought by their thermal neutron signature at a distance of a few centimeters and that potential rotational movements of the fission chamber up to 60° have a limited impact on signals measured. We also show that sensitivity loss due to sediments potentially covering fuel debris has been evaluated on the order of a factor 10 considering a 30 cm-thick sediment layer. On the other hand, experiments were performed to assess the impact of a strong gamma dose rate on fission chamber measurements. These irradiation trials involved a CFUE32 fission chamber (also manufactured by PHOTONIS) available in our laboratory and three different irradiation means: an X-ray tube, an 192Ir source, and a linear electron accelerator. These experiments enable to draw the conclusion that the fission chamber is not impacted by the gamma dose rate up to 104 Gy h−1, which is in good agreement with specifications provided by the manufacturer (PHOTONIS). In addition, no performance degradation was observed after an integrated gamma dose of 2200 Gy on the fission chamber in a 10 min irradiation. However, when the fission chamber is irradiated by gamma dose rates above 104 Gy h−1 (upper limit of the operating domain specified by PHOTONIS), a significant gamma background is observed. Nevertheless, as the gamma dose rates at Fukushima Daiichi should not exceed 103 Gy h−1, fission chamber measurements performed towards fuel debris localization in the primary containment vessels of the units would not be affected by the severe gamma-ray irradiation.

Fission product yields (FPY) are one of the most fundamental quantities that can be measured for a fissioning nucleus and are important for basic and applied nuclear physics. Recent measurements using mono-energetic and pulsed neutron beams generated using Triangle Universities Nuclear Laboratory’s tandem accelerator and employing a dual fission chamber setup have produced self-consistent, high-precision data critical for testing fission models for the neutron-induced fission of 235,238U and 239Pu between neutron energies of 0.5 to 15.0 MeV. These data have elucidated a low-energy dependence of FPY for several fission products using irradiations of varying lengths and neutron energies. This paper will discuss new measurements just beginning utilizing a RApid Belt-driven Irradiated Target Transfer System (RABITTS) to measure shorterlived fission products and the time dependence of fission yields, expanding the measurements from cumulative towards independent fission yields. The uniqueness of these FPY data and the impact on the development of fission theory will be discussed.

International audience; MCNPX in its latest version is able to simulate the transportation of Fission Fragments. It opens the door to Fission Chamber simulations. Such simulations are not straightforward and comparisons with experimental spectra often failed. A procedure is described in the present paper to perform such simulation and to process the result to obtain realistic spectra. Simulated spectra are compared with experimental ones in various conditions to validate the method and to present its limitations.

The CEA MONACO v2 a multichannel acquisition system dedicated to neutron and gamma measurements. It is unique as it integrates all the following features in one module: automatic generation of saturation curves, automatic generation of pulse discrimination curves, detector pulse characterization using the embedded oscilloscope module, pulse mode acquisitions in count rate or pulse height analysis, fluctuation mode and current mode acquisitions. Sensors are plugged to a single connector and the implemented operating modes run constantly in parallel. Firsly designed for on line local neutron and gamma measurements with gaseous detectors in reactor experiments, the MONACO v2 system will also be available for self-powered detectors thanks to its wide current mode working range. After two years of development, CEA teams tested two MONACO v2 prototypes in the Slovenian TRIGA Mark II research reactor in 2018, using CEA miniature ion chambers and SPNDs. The system is now ready for industrialization to be available on the nuclear instrumentation market.

Prompt γ rays from the spontaneous fission of 252Cf were measured with cerium-doped LaBr3 detectors. The average prompt fission γ-ray multiplicity, M γ , was determined as a function of fragment mass number (A) and total kinetic energy (TKE). High-statistics data, obtained from three detectors of different size at different angles relative to the fission chamber, confirms unequivocally a saw-tooth like shape of the γ multiplicity as a function of fragment mass, in contrast to previously published data. In addition, the TKE-dependence of M γ was determined experimentally and compared with recent data.

A wide-range neutron flux measurement instrument is developed herein for monitoring the total neutron emission rate and yield of the Large Helical Device (LHD) during deuterium experiments implemented from March 2017 in the National Institute for Fusion Science (NIFS), Japan. The instrument is designed for and installed on the Neutron Flux Monitoring (NFM) system, which measures the counting rate using a 235U Fission Chamber. By combining the pulse counting and Campbell methods, the Digital Signal Processing Unit (DSPU) realized a wide dynamic range of over six orders of magnitude from 1 × 103 counts/s (cps) to 5 × 109 cps. This study explains and discusses how the instrument is developed, including topics from the predevelopment activities to the verification test at the Kyoto University Critical Assembly (KUCA). Experimental results in the LHD using the finished products suggest that the NFM system works well during deuterium experiments.

The BEAVRS benchmark was proposed in 2013 to serve as a non-proprietary benchmark based on measured reactor data to validate high-fidelity reactor simulation methods. In its third version release, the benchmark now includes an open source repository to provide users all source files related to BEAVRS. This version will also contain complete documentation of the uncertainty quantification work conducted. As part of this process, this paper elaborates on a systematic approach to quantifying the uncertainty that arises from using radial tilt-corrected data to compare BEAVRS data to simulated data. These uncertainty metrics are combined with model fitting uncertainties from fitting simulation model trends to BEAVRS data in order to compute overall time-dependent uncertainty of fission reaction rates. It is found that axially-integrated fission chamber time-dependent 95% uncertainty values are 1.8% and 1.9% for Cycle 1 and Cycle 2 respectively, which match well with results using traditional methods for uncertainty quantification.

International audience; Xenon high temperature fission chamber, designed for sodium-cooled fast reactors, unlike the argon filled fission chambers, can operate at temperatures greater than 500◦C without partial-discharges and discriminate neutrons and partial-discharges at temperatures up to 650◦C.

During the operation of the Superphenix and Phenix reactors, an aberrant electrical signal was detected from the fission chambers used for neutron flux monitoring. This signal, thought to be due to partial electrical discharge (PD) is similar to the signal resulting from neutron interactions, and is generated in fission chambers at temperatures above 400 °C. This paper reports work on the characterization and localization of the source of this electrical signal in a High Temperature Fission Chamber (HTFC). The relation between the shape of the PD signal and various parameters (nature and pressure of the chamber filling gas, electrode gap distance, and fission chamber geometry) are first described. Next, experiments designed to identify the location within the chambers where the PD are being generated are presented. After verification and refinement of the results of these localization studies, it should be possible to propose changes to the fission chamber in order to reduce or eliminate the PD signal.

Neutrons and photons are characteristically emitted during the nuclear fission process when a deformed, neutron-rich nucleus divides into two fragments that then deexcite. During deexcitation, neutrons are emitted first, followed by photons; this process gives rise to correlated emissions. Few data exist on event-by-event neutron-photon correlation. In this work, $^{252}\mathrm{Cf}(\mathrm{s}\mathrm{f})$ neutron and photon correlations were measured with an array of 45 liquid organic scintillation detectors and a fission chamber. The measured correlations are compared with MCPNX-PoliMi simulations using the built-in model and two event-by-event fission models, CGMF and FREYA, which predict correlations in prompt emissions from fission. Experimental results suggest weak neutron-photon competition during fragment deexcitation.

Neutron sensors capable of real-time measurement of neutrons in high-flux environments are necessary for tests aimed at demonstrating the performance of experimental nuclear reactor fuels and materials in material test reactors (MTRs). In-core Micro-Pocket Fission Detectors (MPFDs) have been studied at Kansas State University for many years. Previous MPFD prototypes were successfully built and tested with promising results. Efforts are now underway to develop advanced MPFDs with radiation-resistant, high-temperature materials capable of withstanding irradiation test conditions in high performance material and test reactors. Stackable MPFDs have been designed, built, and successfully demonstrated as in-core neutron sensors. Advances in the electrodeposition and measurement of neutron reactive material, along with refinements to composition optimization simulations, have enhanced the capabilities of contemporary MPFDs.

Within the bilateral project between the CEA Cadarache and the Jožef Stefan Institute (JSI) a wide variety of measurements using multiple fission chambers simultaneously inside the reactor core were performed. The fission rate axial profiles were measured at different positions in the reactor core and at different control rod configurations. A relative comparison of the calculated fission rates using the MCNP code and the measured fission rates was performed. In general the agreement between the measurements and calculations is good, with the deviations within the uncertainties. For better observation and understanding of the neutron flux redistribution due to the control rod movement, the neutron flux and fission rate had been calculated through the entire reactor core for different control rod configurations. The detector position with minimum signal variations due to the regulating and compensating control rod movement during normal operation was determined. The minimum variation is optimal in case we want to reliably determine the reactor power without influence of the regulating and compensating control rod positions.

VENUS-F is a fast zero power reactor with 30 wt% U fuel and Pb/Bi as a coolant simulator. Depending on the experimental configuration, various neutron spectra (fast, epithermal, and thermal islands) are present. This paper gives a review of the nuclear instrumentation that is applied for reactor control and in a large variety of physics experiments. Activation foils and fission chambers are used to measure spatial neutron flux profiles, spectrum indices, reactivity effects (with positive period and compensation method or the MSM method) and kinetic parameters (with the Rossi-alpha method). Fission chamber calibrations are performed in the standard irradiation fields of the BR1 reactor (prompt fission neutron spectrum and Maxwellian thermal neutron spectrum).