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• This paper proposed a novel automatic calibration method which can automatically deduct various baseline of Campbelling mode and obtain neutron yield in real time. • The automatic calibration method was verified with the experimental data on EAST. It is suitable for different experiment conditions, and the Relative Root Mean Square Error values of automatic calibration results were obviously lower than that of offline fixed calibration results. To satisfy the wide-range requirement for the fission chamber-based neutron flux monitoring (NFM) system on experimental advanced superconducting tokamak (EAST), the pulse-counting and Campbell modes were combined to expand the measurement range. However, to obtain the neutron yields, the results of the Campbell mode should first be converted into neutron count rates via linear calibration. During the plasma experiment, the non-neutron backgrounds vary with respect to different experimental conditions, which leads to a baseline dynamic change in the Campbell mode. In future research, NFM system need provide real-time feedback parameters regarding the operational status of the fusion device. Therefore, an automatic calibration method was developed to automatically deduct various baselines of the Campbell mode and obtain the neutron yields in real time. This method continuously updates the average value of each sub-interval (multiple sub-intervals are set in the calibration region) using a linear superposition averaging method. In addition, to optimize the dual-mode switching threshold, ensure the calibration region and define the upper limit of Campbell mode, a simulated neutron signal generator was designed and applied. The switching threshold of the dual modes was set to 400 kcps, the calibration region of the Campbell mode was selected as 50–500 kcps, and the upper limit of the equivalent count rate in the Campbell mode reached to 10 8 cps with a relative error of<4%. Finally, the automatic calibration method was tested using experimental data from the NFM system on EAST. The automatic calibration results basically overlapped with the reference results and superior to the offline fixed calibration results. Under different experimental conditions, the relative root mean square error (RRMSE) values of the automatic calibration results were lower than those of the offline fixed calibration results. Therefore, the proposed automatic calibration method is accurate and reliable, and can satisfy the real-time measurement requirements of wide-range NFM system. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

The Large Helical Device (LHD) plans to start deuterium experiments in March 2017. The expected performances of the neutron flux monitor and the vertical neutron camera have been investigated by neutron Monte Carlo simulations using the MCNP-6 Monte-Carlo neutronics code and the cross-section library FENDL-3.0. Three neutron flux monitors using 235 U fission chambers are located on the center axis of LHD and at two toroidal locations on the horizontal plane outside the cryostat. The detection efficiency for the neutron emission in the plasma has been evaluated. This is almost insensitive to the neutron source profile and the plasma position. The vertical neutron camera consists of 11 channels of stilbene detectors with a parallel collimator embedded in the concrete floor slab. The response function of the stilbene detector has been evaluated with PHITS code. Based on the response function, the count rate of the stilbene detector is estimated to be ∼6 × 10 4 cps for maximum neutron yield shots. Also, the spatial resolution is estimated to be 70 mm, which is smaller than the pitch of adjoining channels. [ABSTRACT FROM AUTHOR]

• Systematic overestimation discovered in previously reported fission chamber neutron sensitivity data. • Tallying neutron flux from Maxwellian distribution source in MCNP was identified as the source of error. • New procedure to properly simulate neutron moderation was implemented. • Corrected results are reported. During the course of the work carried out for the estimation of the sensitivity of 6Li loaded fiber detectors (Borella, 2022), it was realized that the MCNP simulations carried out with the approach described in (Borella et al., 2013) were affected by a systematic bias. In this paper, we explain the source of the systematic bias and report the new results for (Borella et al., 2013). [ABSTRACT FROM AUTHOR]

Via thermal cycling and vibration tests, this study aims to demonstrate that the in-vessel components of the microfission chamber (MFC) system can withstand the extreme International Thermonuclear Experimental Reactor (ITER) environment. In thermal cycle tests, the signal cable of the device was bent into a smaller radius and it was given more bends than those in its actual configuration within ITER. A faster rate of temperature change than that under the typical ITER baking scenario was then imposed on in-vessel components. For the vibration tests, strong 10 G vibrational accelerations with frequencies ranging from 30 Hz to 2000 Hz were imposed to the detector and the connector of the in-vessel components to simulate various types of electromagnetic events. Soundness verification tests of the in-vessel components conducted after thermal cycling and vibration testing indicated that problems related to the signal transmission cable functioning were not found. Thus, it was demonstrated that the in-vessel components of the MFC can withstand the extreme environment within ITER. [ABSTRACT FROM AUTHOR]

Applying Higher Order Campbelling methods in neutron flux monitoring with fission chambers is advantageous due to their capabilities to suppress the impact of unwanted noises and signal contributions (such as gamma radiation). This work aims to verify through experimental results that the basic assumptions behind the Higher Order Campelling methods are valid in critical reactors. The experiments, reported in this work, were performed at the MINERVE reactor in Cadarache. It is shown that the calibration of a fission chamber and the associated electronic system is possible in higher order mode. With the use of unbiased cumulant estimators and with digital processing, it is shown that over a wide count rate range, accurate count rate estimation can be achieved based on signal samples of a few ms , which is a significant progress compared to similar experimental results in the literature. The difference between the count rate estimated by pulse counting and by the Higher Order Campelling is less than 4%. The work also investigates the possibility of monitoring transient events. For this purpose, a control rod drop event was followed in Higher Order Campbelling mode. [ABSTRACT FROM AUTHOR]

New fission chambers dedicated to Prompt Fission Neutron Spectra measurements with the time-of-flight technique have been developed. The actinide mass embedded in the chamber was maximized, while the alpha-fission discrimination and the time resolution were optimized. Moreover, to reduce the neutron background and spectra distortions, neutron scattering with the materials were minimized by the choice of material and structure. These chambers were then tested and validated during tests and in-beam experiments. [ABSTRACT FROM AUTHOR]

Fission chambers have become one of the main devices for the measurement of neutron fluxes in nuclear facilities; including fission reactors, future fusion ones, spallation sources, etc. The main goal of a fission chamber is to estimate the neutron flux inside the facility, as well as instantaneous changes in the irradiation conditions. A Monte Carlo Fission Chamber Designer (MCFCD) has been developed in order to assist engineers in the complete design cycle of the fission chambers. So far MCFCD focuses on the most important neutron reactions taking place in a thermal nuclear reactor. A theoretical model describing the most important outcomes in fission chambers design has been developed, including the expected electrical signals (current intensity and drop in potential) and, current-polarization voltage characteristics (sensitivity and saturation plateau); the saturation plateau is the zone of the saturation curve where the output current is proportional to fission rate; fission chambers work in this region. Data provided by MCFCD are in good agreement with measurements available. [ABSTRACT FROM AUTHOR]

This paper investigates, through numerical simulations, the performance of a signal analysis method by which a high temperature fission chamber can be used over a wide range of count rates. Results reported in a previous paper (Elter et al., 2015 [1] ) indicated that the traditional Campbell method and the pulse mode cannot provide a sufficient overlap at medium count rates. Hence the use of the so-called Higher Order Campbell (HOC) methods is proposed and their performance is investigated. It is shown that the HOC methods can guarantee the linearity (i.e. correctness) of the neutron flux estimation over a wide count rate, even during transient conditions. The capabilities of these methods for suppressing parasitic noise (originating from various sources) are verified. [ABSTRACT FROM AUTHOR]

Preparation of thin U- and Th-coated 0.3 mm diameter Pt working electrodes by the cyclic potential sweep method is described. Uranyl- and thorium hydroxide layers were electrodeposited from ethanol solutions containing 0.02 M natural uranyl and 0.02 M natural thorium nitrate, each with 3.6 M ammonium nitrate. The cell for electrodeposition was specially developed in order to accommodate the small working electrodes for this research by including a working electrode probe, 3-D translation stage, and microscope. The source material deposition was analyzed using digital microscopy and scanning electron microscopy, and confirmed using x-ray fluorescence measurements. The appropriate potential range for electrodeposition was determined to be −0.62 V to −0.64 V for a 0.3 mm diameter Pt working electrode placed 1 cm from the counter electrode. Smooth, uniform deposition was observed near the central region of the working electrode, while surface cracking and crystalline formations were found near the edge of the working electrode. The final procedure for sample substrate preparation, electrolytic solution preparation and electrodeposition are described. [ABSTRACT FROM AUTHOR]

• The moderator of the NFM#01 system has been determined according to neutronic analysis results. • NFM system is capable of measuring a wide range of total neutron yield and fusion power. • Measurement performance of NFM system can meet ITER requirements well based on current physical design. The ITER Neutron Flux Monitor (NFM) system consists of four subsystems located at ITER Equatorial Ports 1#, 7#, 8#, and 17#. NFM system carries out the measurement of total neutron yield and fusion power by employing 235U fission chambers. The total neutron yield to be measured spans a range of 7 orders of magnitude — from 1014 to 3.2 × 1020 n/s. Suitable moderator and detectors were selected to cover such a wide range of neutron yield and meet the requirements of time resolution and accuracy. Moderators and detectors' sensitivities of NFM system are reasonably designed. The detectors' sensitivities of the NFM#01 system under different moderators have been calculated with MCNP code, and graphite serves as the optimal moderator. The detectors' sensitivities of the other three NFM subsystems have also been simulated simultaneously. Based on the required sensitivities of detector, the appropriate mass of 235U was selected for each fission chamber. According to the analysis of the measurement range of each NFM subsystem, it is finally confirmed that the measurement performance of the NFM system can meet the requirements well. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

For efficient utilization of research reactors, such as TRIGA Mark II reactor in Ljubljana, it is important to know neutron flux distribution in the reactor as accurately as possible. The focus of this study is on the neutron flux redistributions due to control rod movements. For analyzing neutron flux redistributions, Monte Carlo calculations of fission rate distributions with the JSI TRIGA reactor model at different control rod configurations have been performed. Sensitivity of the detector response due to control rod movement have been studied. Optimal radial and axial positions of the detector have been determined. Measurements of the axial neutron flux distribution using the CEA manufactured fission chambers have been performed. The experiments at different control rod positions were conducted and compared with the MCNP calculations for a fixed detector axial position. In the future, simultaneous on-line measurements with multiple fission chambers will be performed inside the reactor core for a more accurate on-line power monitoring system. [ABSTRACT FROM AUTHOR]

The detectors of the neutron flux monitoring system of the foreseen French GEN-IV sodium-cooled fast reactor (SFR) will be high temperature fission chambers placed in the reactor vessel in the vicinity of the core. The operation of a fission chamber over a wide-range neutron flux will be feasible provided that the overlap of the applicability of its pulse and Campbelling operational modes is ensured. This paper addresses the question of the linearity of these two modes and it also presents our recent efforts to develop a specific code for the simulation of fission chamber pulse trains. Our developed simulation code is described and its overall verification is shown. An extensive quantitative investigation was performed to explore the applicability limits of these two standard modes. It was found that for short pulses the overlap between the pulse and Campbelling modes can be guaranteed if the standard deviation of the background noise is not higher than 5% of the pulse amplitude. It was also shown that the Campbelling mode is sensitive to parasitic noise, while the performance of the pulse mode is affected by the stochastic amplitude distributions. [ABSTRACT FROM AUTHOR]

CEA developed fission chambers and ionization chambers were utilized at the JSI TRIGA reactor to measure neutron and gamma fields. The measured axial fission rate distributions in the reactor core are generally in good agreement with the calculated values using the Monte Carlo model of the reactor thus verifying both the computational model and the fission chambers. In future, multiple absolutely calibrated fission chambers could be used for more accurate online reactor thermal power monitoring. [ABSTRACT FROM AUTHOR]

Highlights: [•] Accurate fission chamber models require accurate design information. [•] Fissile mass amount, active layer thickness and structural materials determine the detector sensitivity. [•] Fission products in fission chambers can be modeled and transported in a Monte Carlo code such as MCNPX. [•] Fission products transport plays a key role when determining a fission chamber sensitivity. [•] The model is in good agreement with experimental data. [Copyright &y& Elsevier]

Abstract: Nuclear heating measurements in Material Testing Reactors (MTR) are crucial for the design of the experimental devices and the prediction of the temperature of the hosted samples. Nuclear heating in MTR materials (except fuel) is mainly due to the energy deposition by the photon flux. Therefore, the photon flux is a key input parameter for the computer codes which simulate nuclear heating and temperature reached by samples/devices under irradiation. In the Jules Horowitz MTR under construction at the CEA Cadarache, the maximal expected nuclear heating levels will be about 15 to 18Wg−1 and it will be necessary to assess this parameter with the best accuracy. An experiment was performed at the OSIRIS reactor to combine neutron flux, photon flux and nuclear heating measurements to improve the knowledge of the nuclear heating in MTR. There are few appropriate sensors for selective measurement of the photon flux in MTR even if studies and developments are ongoing. An experiment, called CARMEN-1, was conducted at the OSIRIS MTR and we used in particular a gas ionization chamber based on miniature fission chamber design to measure the photon flux. In this paper, we detail Monte-Carlo simulations to analyze the photon fluxes with ionization chamber measurements and we compare the photon flux calculations to the nuclear heating measurements. These results show a good accordance between photon flux measurements and nuclear heating measurement and allow improving the knowledge of these parameters. [Copyright &y& Elsevier]

Abstract: Neutronic analysis is performed for in situ calibration of the microfission chamber (MFC), which is the in-vessel neutron-flux monitor at the International Thermonuclear Experimental Reactor (ITER). We present the design of the transfer system for a neutron generator, which consists of two toroidal rings and a neutron-generator holder, and estimate the effect of the system on MFC detection efficiency through neutronic analysis with the Monte Carlo N-particle (MCNP) code. The result indicates that the designed transfer system does not affect MFC detection efficiency. In situ calibrations by the point-by-point method and by the rotation method are also simulated and compared by neutronic analysis. The results indicate that the rotation method is appropriate for full calibration because the calibration time is shorter (all neutron-flux monitors can be calibrated simultaneously). However, the rotation method makes it difficult to compare the results with neutronic analysis, so the point-by-point method should be performed prior to full calibration to check the accuracy of the MCNP model. [Copyright &y& Elsevier]

Highlights: [•] We model void in a small lead-cooled fast reactor. [•] We examine the changes neutron-spectrum in various scenarios. [•] We use a combination of two fission-chambers to detect the void. [•] The detector-response is dependent on the position of the detectors. [•] The use of fission chambers to detect voiding appears as a promising monitoring technique. [Copyright &y& Elsevier]

Abstract: In this paper, we address the question of how the signal of a uranium fission chamber is sensitive to the thickness of its fissile coating. We first review the electrodeposition technique in order to status on the maximum surface density reasonably achievable. Then, we assess the effect of the neutron flux perturbation on the possible loss of detection efficiency. The most important part of this work is the assessment how the energy spectrum of a fission chamber is affected by the so-called self-absorption of the fission products in the fissile coating. We present a Monte Carlo code we developed in order to achieve that goal. That numerical tool allowed us to reveal the linear relation between the sensitivity loss and the coating thickness, with a slope of 8%/(mg/cm2). Numerical results are discussed with respect to experimental data and a need of new experiments is expressed. Finally, we showed the frequent assumption of the straight trajectory of fission products is not valid for a small fraction of those particles which have a kinetic energy less than 20keV/amu. [Copyright &y& Elsevier]

Abstract: Knowledge of prompt fission γ-ray emission has been of major interest in reactor physics for a few years. Since very few experimental spectra were ever published until now, new measurements would be also valuable to improve our understanding of the fission process. An experimental method is currently being developed to measure the prompt fission γ-ray spectrum from some tens keV up to 10 MeV at least. The mean multiplicity and total energy could be deduced. In this method, the γ-rays are measured with a bismuth germanate (BGO) detector which has the advantage to present a high P/T ratio and a high efficiency compared to other γ-ray detectors. The prompt fission neutrons are rejected by the time of flight technique between the BGO detector and a fission trigger given by a fission chamber or a scintillating active target. Energy and efficiency calibration of the BGO detector were carried out up to 10.76 MeV by means of the 27Al(p,γ) reaction. First prompt fission γ-ray spectrum measurements performed for the spontaneous fission of 252Cf and for 1.7 and 15.6 MeV neutron-induced fission of 238U at the CEA, DAM, DIF Van de Graaff accelerator, will be presented. [Copyright &y& Elsevier]

Abstract: Using two 238U samples placed in a gridded ionization chamber and a parallel-plate fission chamber, fluence of monoenergetic fast neutrons was determined. Four runs of measurements were performed. Analysis showed that although the neutron fluences for the two 238U samples differ by 20–33 times in the present work, the fluences at the position of the sample in the gridded ionization chamber determined by the two ways are in agreement within experimental uncertainties. [Copyright &y& Elsevier]

Abstract: The engineering design of fission chambers as on-line radiation detectors for IFMIF is being performed in the framework of the IFMIF-EVEDA works. In this paper the results of the experiments performed in the BR2 reactor during the phase-2 of the foreseen validation activities are addressed. Two detectors have been tested in a mixed neutron-gamma field with high neutron fluence and gamma absorbed dose rates, comparable with the expected values in the HFTM in IFMIF. Since the neutron spectra in all BR2 channels are dominated by the thermal neutron component, the detectors have been surrounded by a cylindrical gadolinium screen to cut the thermal neutron component, in order to get a more representative test for IFMIF conditions. The integrated gamma absorbed dose was about 4×1010 Gy and the fast neutron fluence (E >0.1MeV) 4×1020 n/cm2. The fission chambers were calibrated in three BR2 channels with different neutron-to-gamma ratio, and the long-term evolution of the signals was studied and compared with theoretical calculations. [Copyright &y& Elsevier]

Abstract: A comprehensive study is performed in order to assess the pertinence of fission chambers coated with different fissile materials for high neutron flux detection. Three neutron scenarios are proposed to study the fast component of a high neutron flux: (i) high neutron flux with a significant thermal contribution such as BR2, (ii) DEMO magnetic fusion reactor, and (iii) IFMIF high flux test module. In this study, the inventory code ACAB is used to analyze the following questions: (i) impact of different deposits in fission chambers; (ii) effect of the irradiation time/burn-up on the concentration; (iii) impact of activation cross-section uncertainties on the composition of the deposit for all the range of burn-up/irradiation neutron fluences of interest. The complete set of nuclear data (decay, fission yield, activation cross-sections, and uncertainties) provided in the EAF2007 data library are used for this evaluation. [Copyright &y& Elsevier]

The signals of fission chambers are traditionally evaluated with Campbelling methods at medium count rates. Lately there has been a growing interest in the extension of Campbelling methods to cover a wider range of count rates. These methods are applied to measure the neutron flux in the stationary state of the reactor. However, there has not been any attempt to generalize these techniques for transient reactor states. This short note is devoted to a discussion of this question. It is shown through analytic and numerical calculations that for practical reasons the traditional, stationary Campbelling methods can be applied for transient scenarios as well. [ABSTRACT FROM AUTHOR]

Abstract: An ionization chamber with fast timing properties was built at CERN for measuring fission cross-sections of minor actinides at the n_TOF neutron beam. The design of this new chamber and of the front-end electronics was optimized to match the innovative features of the n_TOF facility, in particular the high instantaneous neutron flux and low background. For the most radioactive isotopes, a special version of the chamber, designed according to the ISO2919 standards, was built in order to comply with the radioprotection requirements at CERN. The detector and front-end electronics are here described, together with the simulated and measured response to fission fragments and α-particles. The performances of the chamber during the first measurement campaign at n_TOF are presented, focusing in particular on the fast time response, the good background rejection capability, low-background and high detection efficiency. [Copyright &y& Elsevier]

Abstract: Micro-pocket fission detectors (MPFD) have been fabricated and tested as in-core flux monitors in the 250kW TRIGA nuclear reactor at Kansas State University. The prototype devices have been coated with a natural uranyl-nitrate to provide a neutron reactive coating. The devices are composed of alumina substrates sealed together to form a miniature gas pocket 3mm in diameter and 1mm wide. The devices are radiation hard and can operate in pulse mode in a neutron flux exceeding 1012 cm−2 s−1. Placed in the central thimble of the reactor core, the MPFDs have shown count rate linearity from low to high power. Dead time losses become apparent at power levels exceeding 100kW, yet are still low enough to allow for pulse mode operation. [Copyright &y& Elsevier]

To reduce the contribution of series white noise, a discrete FET input voltage preamplifier for the high-sensitivity fission chamber detector was realized on a 25 ∗ 25 mm2 PCB by using the operational amplifier as the gain and driver stage. The rise time reaches a constant of 12 ns when the capacitance exceeds 400 pF. The slope of equivalent noise charge is 10 electrons/pF in a shaping system of 50 ns time constant. Using this preamplifier, the threshold performances of fission chamber under neutron source irradiation a neutron source were verified, moreover the charge collection time of detector was observed. [ABSTRACT FROM AUTHOR]

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). [ABSTRACT FROM AUTHOR]

This report describes a testing device capable of emulating neutron fission chamber signals over their whole operating range. The device assists in the design and testing of neutron flux monitoring equipment including reactimeters, spectrometers, period meters and protection channels. Because of its hardware architecture, testing of neutron flux monitoring equipment can be done both in a laboratory environment and in-situ. Main characteristics of the device allow one to test the equipment over a wide range of count rates. A flexible mathematical algorithm, that is used to calculate the signal model of fission chambers, gives the opportunity for user to set the pulse shape and the charge spectrum such as those in a real fission chamber. It is also possible to set dynamic signals, for example, such as an exponentially varying count rate or a neutron point reactor kinetics signal with a configurable number of delayed neutron groups and constants. The device was verified by testing of the International Thermonuclear Experimental Reactor (ITER) neutron flux divertor monitor data acquisition system. [ABSTRACT FROM AUTHOR]

A fast preamplifier was built with discrete JFET and operational amplifier for high-sensitivity fission chamber detector. Noise characteristics of preamplifier and detector were researched through measuring equivalent noise charge, processing the preamplifier output and analyzing pulse height. Its rise time is 56 ns with total input capacitance of 810 pF which includes preamplifier, detector and 5-meter-long cable. Considering the rise time and noise characteristics, the most suitable shaping time is 0. 1 μ s. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

This work aims at assessing the performance of a235U 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. • Investigation of fission chamber response in the frame of fuel debris localization measurements by MCNP6 simulation study. • Fuel debris sought underwater by their thermal neutron signature at a distance of a few centimeters. • Sensitivity loss on the order of a factor 10 considering a 30 cm-thick sediment layer on the top of fuel debris. • Experiments to assess the impact of strong gamma dose rates on fission chamber measurements. • No impact of gamma dose rates up to 104 Gy h−1. Beyond, the fission chamber registers a significant gamma background. [ABSTRACT FROM AUTHOR]

An updated method of preparing thin, uranium-coated, 0.25-mm diameter platinum electrodes using the cyclic potential sweep method is described. Sixty-four samples were prepared using a 0.2 M UO 2 (NO 3) 2 solution with 19.695% enriched uranium. Improvements were made to the sample preparation, electrodeposition, and analysis methods, producing smoother samples with more uniform surface features than prior work. Alpha-particle spectrometry was used to determine the mass of uranium deposited on each sample. The average mass of uranium deposited on the metallic electrode from 100 potential sweep cycles was 0.987 × 10−7 g with a standard deviation of 0.860 × 10−7 g. Removing the outliers reduces the average deposited mass to 0.741 × 10−7 g with a standard deviation of 0.410 × 10−7 g. A smooth, uniform deposition was observed over the entire electrode after deposition and no cracking or crystalline formations were observed. This improved method of electrodeposition remains an attractive method of depositing fissile material onto small metallic substrates for the fabrication of Micro-Pocket Fission Detectors. [ABSTRACT FROM AUTHOR]