Fission Chamber Research Articles

Measurement and analysis of the neutron-induced total cross-sections of 209Bi from 0.3 eV to 20 MeV on the Back-n at CSNS* *Supported by the National Natural Science Foundation of China (12375296), the Key Laboratory of Nuclear Data Foundation (JCKY2022201C153), the National Key Research and Development Plan (2022YFA1603303), and the Natural Science Foundation of Hunan Province of China (2021JJ40444, 2020RC3054)

The neutron-induced total cross-section of 209Bi is crucial for the physical design and safety assessment of lead-based fast reactors, and the quality of experimental data should be improved for evaluation and application. A recent experiment was conducted on the back-streaming white neutron beamline (Back-n) at the China Spallation Neutron Source (CSNS) using the neutron total cross-section spectrometer (NTOX). The neutron energy was determined using a fast multi-cell fission chamber and the time-of-flight technique. Two high-purity bismuth samples, 6 mm and 20 mm in thickness, were chosen for neutron transmission measurements and comparisons. The neutron total cross-sections of 209Bi, ranging from 0.3 eV to 20 MeV, were derived considering neutron flight time determination, flight path calibration, and background subtraction. A comparison of the experimental results with the data in the ENDF/B-VIII.0 library showed fair agreement, and the point-wise cross-sections were found to be consistent with existing experimental data. Special attention was given to the determination of resonance parameters, which were analyzed using the R -matrix code SAMMY and Bayesian method in the 0.5 keV to 20 keV energy range. The extracted resonance parameters were compared to previously reported results and evaluated data. This study is recognized as the first one where the neutron total cross-section of bismuth across such a broad energy spectrum is measured in a single measurement or experiment, and it provides valuable data for the assessment of related reaction information for evaluated libraries and the advancement of lead-bismuth-based nuclear systems.

TRU oxide sample reactivity worths measured in the FCA-IX assemblies with systematically changed neutron energy spectra

ABSTRACT A series of integral experiments was conducted to evaluate the fission and the capture cross-sections of transuranic (TRU) nuclides at the fast critical facility FCA of the Japan Atomic Energy Agency (JAEA). The experiments were carried out using seven uranium-fueled assemblies of the FCA. The neutron energy spectra of the core regions were adjusted so as to change from an intermediate neutron spectrum to a fast neutron spectrum on an assembly-by-assembly basis. The integral data measured with these experimental configurations provide some neutron energy characteristics: 1) fission rate ratios (FRRs) of 237Np, 238Pu, 242Pu, 241Am, 243Am, and 244Cm relative to 239Pu by using absolutely calibrated fission chambers, 2) small sample reactivity worths (SRWs) of 237Np, 238Pu, 240Pu, 241Am, and 243Am where oxide powders of around 15 to 20 grams were used, 3) criticalities, and 4) spectral indices such as fission rate ratios of 238U relative to 235U. In this paper, details of the SRW measurements are reported, and the latest Japanese Evaluated Nuclear Data Library JENDL-5 is tested by using the integral data obtained in systematically varied neutron energy spectra.

Estimating the neutron yield in a deuterium plasma with the JET neutron camera

The JET neutron camera is a well-established detector system at JET, which has 19 sightlines each equipped with a liquid scintillator. The system measures a 2D profile of the neutron emission from the plasma. A first principle physics method is used to estimate the DD neutron yield that is based on JET neutron camera measurements and is independent of other neutron measurements. This paper details the data reduction techniques, models of the neutron camera, simulations of neutron transport, and detector responses used to this end. The estimate uses a simple parameterized model of the neutron emission profile. The method makes use of the JET neutron camera’s upgraded data acquisition system. It also accounts for neutron scattering near the detectors and transmission through the collimator. These components together contribute to 9% of the detected neutron rate above a 0.5 MeVee energy threshold. Despite the simplicity of the neutron emission profile model, the DD neutron yield estimate falls on average within 10% agreement with a corresponding estimate from the JET fission chambers. The method can be improved by considering more advanced neutron emission profiles. It can also be expanded to estimate the DT neutron yield with the same methodology.

A pile noise experiment in the reference core of the nuclear research reactor LR-0

For safety reasons, it is important to simulate precisely the kinetic behavior of light water nuclear reactors (LWR) in nominal or incidental situations. In the point kinetic approximation, a few integral parameters allow describing the reactor’s behavior. It is of interest to measure those parameters, such as the delayed neutron fraction and delayed neutron abundances, in low power research reactors loaded with benchmarked fuel configurations to compare with simulations results. In this context, Commissariat à l’Énergie Atomique et aux Énergies Alternatives (CEA) and Centrum Výzkumu Řež (CVŘ) have been collaborating towards measuring the delayed neutron fraction of LR-0, a VVER-type research reactor operated by CVŘ. This paper presents the first estimation of the delayed neutron fraction of the reactor loaded with six fuel assemblies, arranged in a small hexagonal lattice with a center cavity accessible for instrumentation. This configuration was chosen because it was very close to the so-called reference configuration with seven fuel assemblies, the one that was benchmarked in the NEA IRPheP Handbook. The measurement was performed in July 2021 using the CEA-developed acquisition system SPECTRON dedicated to pile noise experiments. Pile noise refer to a set of techniques which focus on the correlations induced by the fission process in the signals of neutron detectors. The experimental campaign was designed to meet two objectives: first to provide an estimation of the delayed neutron fraction; second to study the impact of the location of the neutron detectors on the quality of the results. Two fission chambers were used, one was fixed and located in the core center and the second was placed in various locations in the reflector. Several configurations were tested with the mobile detector progressively moved up to 20 cm from the closest fuel element. The reactor fission rate was obtained thanks to gamma spectrometry on metal foils (gold, nickel and iron). Simulations with TRIPOLI-4® and JEFF33 libraries were performed to convert the saturated activities into fission rates. The delayed neutron fraction was estimated at 782 ± 10 pcm and the generation time was measured at 39.1 ± 0.5 μs. The uncertainty balance is as follows: 2 % associated to the Diven factor and 1.5 % associated to the core’s integral fission rate and 0.5% for the statistical uncertainty. The reproducibility of the experiments was found very good, with a negligible dispersion amongst experiments (0.2 %). These results are valuable complements to the characterization of LR-0 reference neutron benchmark field.

Recoil Proton Telescopes and Parallel Plate Avalanche Counters for the 235U(n,f) cross section measurement relative to H(n,n)H between 10 and 450 MeV neutron energy

With the aim of measuring the 235U(n,f) cross section at the n_TOF facility at CERN over a wide neutron energy range, a detection system consisting of two fission detectors and three detectors for neutron flux determination was realized. The neutron flux detectors are Recoil Proton Telescopes (RPTs), based on scintillators and solid state detectors, conceived to detect recoil protons from the neutron-proton elastic scattering reaction. This system, along with a fission chamber and an array of parallel plate avalanche counters for fission event detection, was installed for the measurement at the n_TOF facility in 2018, at CERN.An overview of the performances of two RPTs — especially developed for this measurement — and of theparallel plate avalanche counters are described in this article. In particular, the characterizationin terms of detection efficiency by Monte Carlo simulations and response to neutron beam, the studyof the background, dead time correction and characterization of the samples, are reported. Theresults of the present investigation show that the performances of these detectors are suitable foraccurate measurements of fission reaction cross sections in the range from 10 to 450 MeV.

Progress of Gamma Ray Irradiation Experiments on ITER Diagnostics from JADA

The Japan domestic agency (JADA) is responsible for procuring five diagnostic systems for ITER: the microfission chamber (MFC), poloidal polarimeter (PoPola), edge Thomson scattering (ETS), divertor impurity monitor (DIM), and divertor infrared thermography (IRTh) systems. Components of these systems that will be installed in high radiation zones must undergo radiation resistance tests to ensure reliability. The JADA diagnostic group has been conducting gamma ray irradiation experiments on diagnostic components at QST since 2018. The MFC is a neutron diagnostic system that uses uranium fission chambers, the mineral-insulated cables of which were evaluated for corrosion resistance. In-situ observations of the MFC preamplifier have been launched. The PoPola provides the plasma current profile by detecting the polarizations of far-infrared laser beams at their inlets and outlets. Irradiation tests confirmed the durability of PoPola piezo actuators. The ETS is a laser-aided diagnostic to measure electron temperature and density in plasmas from scattered spectra. Laser-induced damages to the optical elements caused by irradiation were investigated. The DIM is a spectroscopic system having a wavelength range of 200 nm to 1000 nm. The effects of irradiation on optical devices, metal mirrors, and radiation-resistant optical fibers were investigated. The IRTh is an infrared thermography system to observe the surface temperature of the divertor. The optical elements and electrical devices of the IRTh have undergone irradiation experiments. The progress of the gamma ray irradiation experiments on the ITER diagnostic systems from JADA are reported.

Nuclear design of a shielded cabinet for electronics: The ITER radial neutron camera case study

The Radial Neutron Camera (RNC) is a diagnostic system located in ITER Equatorial Port #1 providing several spatial and time-resolved parameters for the fusion power estimation, plasma control and physics studies. The RNC measures the uncollided 14 MeV and 2.5 MeV neutrons from deuterium-tritium (DT) and deuterium-deuterium (DD) fusion reactions through an array of neutron flux detectors located in collimated Lines of Sight. Signals from RNC detectors (fission chambers, single Crystal Diamonds and scintillators) need preamplification because of their low amplitude. These preamplifiers have to be as close as possible to the detectors in order to minimize signal degradation and must be protected against fast and thermal neutrons, gamma radiation and electromagnetic fields. The solution adopted is to host the preamplifiers in a shielded cabinet located in a dedicated area of the Port Cell, behind the Bioshield Plug. The overall design of the cabinet must ensure the necessary magnetic, thermal and nuclear shielding and, at the same, satisfy weight and allocated volume constraints and maintain its structural integrity. The present paper describes the nuclear design of the shielded cabinet, performed by means of 3D particle transport calculations (MCNP), taking into account the radiation streaming through the Bioshield penetrations and the cross-talk effect from the neighboring Lower and Upper Ports. We present the assessment of its nuclear shielding performances and analyze the compliancy with the alert thresholds for commercial electronics in terms of neutron flux and cumulated ionizing dose.

Neutronic simulations of in-vessel neutron calibrations for ITER neutron diagnostics by using simplified ITER model

The absolute calibration of the detection efficiency for the measurement of the total neutron yield in the whole plasma is one of the most important issues in neutron diagnostics. In ITER, a compact deuterium-tritium (D-T) neutron generator will be used as a neutron source for the neutron calibration prior to the D-T plasma experiments, where the neutron source will be moved in the vacuum vessel. Neutronic simulations of neutron diagnostic calibrations have been carried out for the strategy and scheduling of the calibration experiments. We have established a simplified 360° ITER model for the simulation, which includes neutron flux monitors (NFMs) in an equatorial port (EQ), micro fission chambers, diverter neutron flux monitors, and a neutron activation system. At first, angular neutron spectra of the compact D-T neutron generator have been evaluated by MCNP calculations. We evaluate the discrepancies among the detection efficiencies to be obtained by the neutron calibration experiment, those by idealistic D-T ring source, and actual detection efficiencies for the plasma neutron source. Point efficiency measurement using the compact D-T neutron generator facing NFM in EQ#1 is an effective method for the NFM calibration.

Study of Reactor Core Loading Monitoring at the GUINEVERE Facility

The potential use of a pulsed neutron source (PNS) to measure reactivity during nuclear fuel loading as a means to prevent core loading errors has been studied at the GUINEVERE facility. This facility couples the deuteron accelerator GENEPI-3C to the fast neutron subcritical reactor VENUS-F at the Belgian Nuclear Research Center SCK·CEN. The 14-MeV neutrons are produced in the reactor core center via fusion reactions. PNS experiments were performed in five reactor configurations corresponding to the different loading steps of VENUS-F. The evolution of the neutron flux during these PNS experiments was measured by several 235U fission chambers in various positions in the inner and outer reflector and analyzed using the area-ratio method. The results show that, despite strong spatial effects, a strong correlation between the reactivity values given by the area-ratio method and some reference reactivity values remains throughout the reactor unloading. Monte Carlo simulations were first validated by comparison with the data and then used to investigate the sensitivity of the method to a core loading error. First results show that some loading errors could be experimentally detected using a PNS.

GRAPhEME: Performances, achievements (@EC-JRC/GELINA) and future (@GANIL/SPIRAL2/NFS)

GRAPhEME is a γ-spectrometer developed by CNRS/IPHC Strasbourg (France), in collaboration with EC-JRC Geel (Belgium) and IFIN-HH Bucharest (Romania). With its 6 High Purity Planar Germanium detectors and one fission chamber, GRAPhEME, installed at the EC-JRC GELINA facility, was optimized for measurements of accurate (n, xnγ) cross sections on actinides. The experimental methodology is based on the prompt γ-ray spectroscopy coupled to time-of-flight measurements. In this paper, we present an overview of fifteen years of experiments with GRAPhEME at EC-JRC GELINA facility, illustrated by main achievements to highlight the performances reached by our spectrometer. Beyond the experimental work, a close collaboration with theoreticians has emerged allowing the use of the data produced with GRAPhEME to test and constraint nuclear reaction codes like TALYS, CoH and EMPIRE. In a near future, GRAPhEME will be available to start measurement campaigns at the new neutron beam facility SPIRAL2/NFS. There, studies of (n, 2n) and (n, 3n) reactions will be possible and will complete the work done at EC-JRC GELINA on (n, n ) reactions. Despite the amount of cross section data provided by GRAPhEME up to now, the prompt γ-ray spectroscopy method presents some weaknesses that our collaboration tries to overcome. This goes through new calculation schemes based on theoretical modeling constrained on experimental data to infer the total (n, xn) cross section, new instrument to measure conversion electrons but also by being proactive in dissemination activities to make the nuclear structure community aware of our needs about new accurate nuclear structure information on actinides.

Measurement of delayed neutrons in a thermal nuclear reactor by means of a long run pile noise experiment in sub-critical state

The transfer function of a zero-power thermal reactor was successfully measured thanks to neutron noise techniques from 1 mHz to 160 Hz. During a month-long experimental campaign, the fluctuations of the neutron population in critical and subcritical configurations of the core were acquired using excore fission chambers and analysed through Cross-Power Spectral Density (CPSD) methodology. Firstly, the reactor’s kinetic parameters, i.e. prompt decay constant, effective delayed neutron fraction and generation time, were obtained at critical state. It required calibrating the reactor’s power, which was done by metal foil activation and measurement of the 235U fission rate. Secondly, these parameters were used to estimate the groups’ abundances of delayed neutrons from the CPSD measured in a sub-critical state. Fitting data with a point kinetic model was done with Bayesian inference - CONRAD and Stan programs were used. A very good agreement was found between experimental abundances and the ones computed with TRIPOLI-4 Monte-Carlo transport code and JEFF3.1.1 nuclear data library. Uncertainties on prior abundances between 6 % to 101 %, held mainly by nuclear data, were lowered down to 4 % to 54 %, depending on the delayed group.

Measurement of the delayed-neutron yield in the thermal neutron induced fission of 239Pu

This article presents an experimental effort to provide high-quality data to improve the evaluation of the 239Pu delayed neutron yield in the thermal energy range. The set-up is composed of a long counter with sixteen 3He tubes, a fast shutter system to produce irradiation cycles with short rising/falling times, and a miniature fission chamber containing 114μg of 239Pu. The whole system was installed in the PF1B experimental zone of the Institut Laue-Langevin, which provides a cold neutron beam. The repetition of irradiation/decay cycles enables to saturate the delayed neutron precursors and to measure their yield through the observed activity, shortly after the beam-stop. The innovation of our measurement technique relies on the clear distinction between prompt and delayed neutron counting, thanks to boron absorbers, without the necessity to move the sample. In such a way, it is possible to normalize the counting of delayed neutron emission to the one of total neutron emission, based on the well-known value of the prompt neutron multiplicity. The present work provides a delayed neutron yield value of vd = 0.642(5)%. The latter is in 1σ agreement with the IAEA recommendation of 0.628(38)%, with a strongly reduced uncertainty thanks to our normalization technique.

The past and the future of the GAINS spectrometer @ GELINA

GAINS (Gamma Array for Inelastic Neutron Scattering), currently installed at the GELINA (Geel Electron Linear Accelerator) neutron source of the EC-JRC (European Commission-Joint Research Centre), is one of the best instruments available worldwide for measuring high resolution, low uncertainty neutron inelastic cross-section data relevant for both fundamental research and nuclear physics applications (like Generation IV facilities). It features 12 HPGe detectors coupled with a 235U fission chamber for neutron flux monitoring. The paper presents a short history of the GAINS development over the years and reviews some of the main experimental results together with plans for future experiments.

Regenerative fission chambers with low variation of sensitivity for high fluence incore measurements

Fission chambers are widely used for online neutron flux monitoring in nuclear reactors. However, the depletion of the fissile deposit of the fission chamber with irradiation limits the life duration of the detector when used in a permanent incore location within the vessel of a power reactor. The addition of a fertile material has been considered, at the cost of yielding a non-monotonous evolution of the sensitivity which could be difficult to distinguish from a genuine flux evolution.In this paper, our goal is to find, for a given neutron flux, a feasible mixture of fissile and fertile isotopes to obtain the more stable sensitivity over a high fluence. We devise a methodology based on the DARWIN suite with no simplification of the evolution chains of isotopes and applicable to any neutron flux. All the isotopes that are routinely used in fission chambers are investigated. It is illustrated with a flux that is representative of the one of a Pressurized Water Reactor: we obtain a mixture of 235U−232Th the sensitivity of which is stable up to 3% over four years at full power corresponding to a fluence about 3⋅1022n/cm2. The practical realization of such a mixture for a fission chamber is described. We discuss the robustness of the stability performance with respect to nuclear data, initial isotopic composition and prior knowledge of the neutron flux. The spectral index, which quantifies the amount of thermalized neutrons, needs to be assessed with a precision about 8%. As a result, such a stable and long-living regenerative fission chamber has to be specifically conceived for a dedicated reactor and measurement location.

Neutronic analysis and measurement performance assessment of ITER neutron flux monitor system

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.

Fission chamber data acquisition system for neutron flux measurements on the Mega-Amp Spherical Tokamak Upgrade.

Neutron flux measurements are important in fusion devices for both safety requirements and physics studies. A new system has been built for the Mega-Amp Spherical Tokamak Upgrade (MAST Upgrade) that provides neutron count, DC, and Campbell mode measurements for a 1 µs period at 1MHz. The acquisition system uses a Red Pitaya board to sample current from two fission chambers mounted on the side of the MAST-U vessel. The system-on-chip design of the Zynq-7020 on the Red Pitaya also allows a web server implementation using Flask for data retrieval and diagnostic configuration over the MAST Upgrade network.

Simultaneous measurement of energy spectrum and fluence of neutrons using a diamond detector

Due to the excellent radiation hardness and high–temperature endurance, diamond detectors are suitable for intense neutron measurements and promising for neutron diagnostics of scientific fusion devices. In the present work, simultaneous measurement of energy spectrum and fluence of neutrons using a diamond detector was realized for the first time. The absolute response matrix of the diamond detector was simulated based on detailed analysis of the nuclear reactions and the proper selection of nuclear reaction data. Neutron energy spectra as well as neutron fluences for 5.0, 5.5, 8.5, 9.5 and 10.5 MeV neutrons from d–d reaction were measured using the diamond detector based on the absolute response matrix. The measured neutron energy spectra and neutron fluences are reasonable compared with those detected using a EJ-309 liquid scintillator and a 238U fission chamber, respectively, which verifies the reliability of the present work. Furthermore, the energy spectrum and fluence of a 14.2 MeV d–t neutron source were also measured using the diamond detector. The present work demonstrates the ability of simultaneous measurement of energy spectrum and fluence as well as for both d–d and d–t neutrons using a diamond detector, which is of great significance for neutron diagnostics of scientific fusion devices.

Fusion neutron emissivity tomography for ITER Vertical Neutron Camera

ITER Vertical Neutron Camera (VNC) is a diagnostic intended to measure time- and space-resolved neutron emission profile in the poloidal plasma section. It is composed of the two fan-shaped collimating structures: lower (5 collimators) and upper (6 collimators). Every collimator contains 2 fission chambers and 2 diamond detectors.We developed an algorithm of simultaneous reconstruction of DD and DT neutron emissivity profiles from measurements of VNC neutron flux monitors. The algorithm is based on maximum likelihood principle. We consider operation of diamond detectors as threshold counters. By setting deposited energy threshold equal to 3.5 MeV diamond detectors become DT neutron flux monitors. Relying on the different energetic sensitivities of VNC neutron flux monitors (fission chambers and diamond detectors) it is possible to distinguish fluxes of DD and DT neutrons.We evaluated sensitivity of neutron emissivity reconstruction to the statistical errors of measurements, errors of contribution matrix, and errors of magnetic surface shapes. The algorithm has been demonstrated to be robust against errors of the input data.

Experimental Verification of Proposed Internal Remote Monitoring System for Spent Nuclear Fuel Dry Cask Storage Applications

It has become a common practice to store sufficiently cooled spent nuclear fuel (SNF) assemblies in interim storage dry casks with passive cooling. These dry casks require nuclear safeguards monitoring because they contain plutonium. Past studies on dry cask modeling and simulations have shown that a remote monitoring system (RMS) situated inside the dry cask could continually monitor and detect the removal of even a single SNF assembly from the cask. This conceptual RMS design was tested by conducting laboratory-scale experiments using small-size 252Cf neutron sources. These small-size sources were surrounded by neutron-reflecting materials in the experiments to mimic the SNF assemblies as a surface neutron source to the fission chamber detectors of the RMS. Experimental and simulation results showed that the removal or diversion of even a single neutron source is detectable within 4 min with a probability of detection greater than 80%.

Neutron yield measurement system of HL-2A tokamak

This research presents the development of HL-2A neutron yield measurement which includes 235U fission chamber and BF3 and 3He proportional counters. Equivalent noise formula of the radiation detection signal amplification system was derived to guide the development of the signal amplification system. Then all detectors were calibrated in situ by using the 252Cf neutron source. The neutron yield of the HL-2A during neutral beam heating was analyzed. These results indicate that the developed neutron flux diagnostic system can obtain neutron yield results under various experimental conditions of the HL-2A tokamak, and can provide information on neutron yield.