MINERVE Reactor Research Articles

A spatially auto-corrected reactimeter algorithm

Reactimeter algorithms enable real-time evaluation of the reactivity of a near-critical reactor from the signal of a neutron detector. They only depend on a few parameters defined by the inverse point kinetic equations, and which can be calculated by most neutron transport codes. However, the neutron population measured locally by a detector may not always lead to a good estimate of the reactivity. For example, this would be the case for an in-core detector placed just next to a moving control rod. In such special cases – or when measurement precision is an issue – costly simulations are needed to provide correction factors that account for variations in the spatial distribution of the neutron flux. This paper aims to provide an auto-corrected reactimeter algorithm (ACRA) in the special cases where rapid perturbations are applied to a core close to criticality. Such perturbations were successfully performed in the MINERVE reactor using a dedicated control rod to alternatively insert and withdraw a small sample of material under study. ACRA has been validated against 3D Monte Carlo neutron transport calculations using the signals of four in-core detectors located just next to the moving sample.

Experimental Estimation of the Kinetic Parameters of MINERVE Zero Power Reactor

Kinetic neutron parameters are of fundamental importance in the field of nuclear reactor dynamics and control. Moreover, the precursor yield fraction and the neutron generation time for a given nuclear reactor are dependent on the properties of the reactor. Thus, in-pile experiments, such as oscillation experiments and noise experiments, are commonly conducted to measure those values. In this work, a method for determining the kinetic parameters of a reactor along with their covariance data from in-pile experiments is presented. It is performed by combining values of the reactor’s response function obtained from both oscillation and noise experiments over a wide range of frequencies. The method is carried out for the MINERVE zero power reactor (ZPR) using a reanalysis of both oscillation and noise experiments that were conducted in the MINERVE reactor in 2013 and 2014. Moreover, various advantages and disadvantages of performing multiple in-pile experiments and combining their results in order to obtain a single set of kinetic parameters along with their covariance data are considered. Some suggestions for the design of such in-pile experiments are also discussed.

Am-241 thermal neutron capture cross section and neutron capture resonance integral from reactor activation and oscillation measurements

In order to validate the neutron capture cross section in 241Am, comparisons of calculations with activation experiments at the TRIGA reactor at JSI and pile oscillation measurements at the MINERVE reactor at CEA Cadarache were done. The results suggest that the evaluation in the new test library JEFF-4T1 on average provide an improvement of the C/E results compared to the existing nuclear data library releases. However, due to some discrepancies, additional experiments at JSI as well as further development of the computational model of the JSI TRIGA reactor are planned to improve the accuracy and reliability of the neutron activation analysis and the nuclear data inferred from them.

Neutron reactor dosimetry monitoring by optical, nanostructural, and morphological changes of NiO thin films

The effect of mixed neutron-gamma (n/γ) irradiation on microstructural, morphological, and optical properties of nickel oxide (NiO) thin films has been studied using UV–visible spectrophotometry, photoluminescence (PL) spectroscopy, X-ray diffraction (XRD), and scanning electron microscopy (SEM). (NiO) thin films were irradiated in a Minerve reactor core at a doses range between 100 and 1000 Gy. The fast neutron dose, mainly due to elastic neutron scattering interactions in the range of 1–5 MeV, represents around 70% of the total irradiation dose. The results obtained show that the irradiation of the NiO thin films lead to significant changes in the optical properties, in the structure and in the morphology of the materials. The optical analysis shows visible range activation, which resulted in an increase in the absorbance of the band, centered around 645 nm. Otherwise, the integrated area of the band presents a linear dose response from 85 to 115 at dose ranging from 0 to 500 Gy. Furthermore, the band gap energy of the samples was linearly decreased from 3.52 eV to 3.44 eV by increasing the dose from 100 to 500 Gy, reaching 3.62 eV at 1000 Gy of irradiation n/γ dose. Urbach energy exhibited the same linear behavior, hence increasing from 372 to 400 meV. The variability of the integrated area, band gap energy and Urbach energy did not exceed 3% during 25 days. XRD analysis demonstrates that the crystallite size and the peak intensity for each preferred orientation planes (111) change with neutron-gamma radiation dose. The lattice parameter of preferential orientation (111) plane from rises linearly from 4.171 to 4.199 Å ± 0.0021 Å. Its variability is low, not exceeding 2%. SEM picture of irradiated NiO thin films shows the pore formation whose distribution increases with a dose level up to 500 Gy. Overall, the obtained results suggest that NiO thin films could considered as new alternative mean in neutron dosimetry application in nuclear reactor.

Neutron resonance transmission analysis of cylindrical samples used for reactivity worth measurements

A characterisation of cylindrical samples by Neutron Resonance Transmission Analysis (NRTA) at the GELINA facility of JRC Geel (Belgium) is presented. The samples were designed and produced for reactivity worth measurements in the MINERVE reactor of CEA Cadarache (France). NRTA was applied to determine the nuclide composition of UO2, Al2O3 and liquid samples that were doped with silver. The volume number densities of 238U, 107Ag and 109Ag obtained by NRTA are within 2 % fully consistent with the values that are quoted by the manufacturer. In addition, the NRTA data reveal a tungsten contamination which is not reported by the provider. It is shown that such a contamination contributes by up to 5.7 % to the reactivity worth.

Evaluation of ENDF/B-VI.8 and ENDF/B-VII.0 Nuclear Data Libraries using OSMOSE Samples of MINERVE Reactor

The evaluation of nuclear data libraries is an important tool to develop nuclear technology. To achieve this goal, OSMOSE program utilized samples that are composed of separated actinides which contain isotopes of americium, plutonium, uranium, and thorium. By inserting those samples in the core configuration of MINERVE reactor, they can be used for improvement of nuclear cross-section databases by cross checking their calculated reactivity worth per each sample using several nuclear data library versions and experimental measurements. The present study investigated the effect of the used nuclear data library on the calculated reactivity of MINERVE reactor with several OSMOSE samples employed in R1-UO2, R1-UO2, and R1-MOX core configurations. The study performed a comparison between the calculated effective multiplication factor using MCNPX code with ENDF/BVI.8 and ENDF/BVII.0 nuclear data libraries. Moreover, the relative difference between the calculated reactivity worth by ENDF/BVI.8 and ENDF/BVII.0 was examined and discussed using available experimental results of the relevant core.

Neutron resonance transmission analysis of cylindrical samples used for reactivity worth measurements

A characterization of cylindrical samples by neutron resonance transmission analysis (NRTA) at the GELINA facility of JRC Geel (Belgium) is presented. The samples were designed and produced for reactivity worth measurements in the MINERVE reactor of CEA Cadarache (France). NRTA was applied to determine the nuclide composition of UO2, Al2O3 and liquid samples that were doped with silver. The volume number densities of 238U, 107Ag and 109Ag obtained by NRTA are within 2% fully consistent with the values that are quoted by the manufacturer. In addition, the NRTA data reveal a tungsten contamination which is not reported by the provider. It is shown that such a contamination contributes by up to 5.7% to the reactivity worth.

Innovative Hybrid Pile Oscillator Technique in the Minerve Reactor: Open Loop Versus Closed Loop

Pile oscillator techniques are powerful methods to measure small reactivity worth of isotopes of interest for nuclear data improvement. This kind of experiment has long been implemented in the Minerve experimental reactor, operated by CEA Cadarache, Saint-Paul-les-Durance, France. A hybrid technique mixing reactivity worth and flux variations surrounding the oscillator is presented here. It was made possible after developing high-sensitivity miniature fission chambers introduced next to the irradiation channel. A test campaign, called MAESTRO-SL, took place in 2015. Its objective was to assess the feasibility of the hybrid method and investigate the separation of mixed neutron effects, such as fission and capture or scattering and capture. The experimental results are presented and discussed in this paper, which focus on comparing two measurement setups, one using a power control system (closed-loop technique) and another one in which reactor power is free to drift (open-loop technique). First, it is demonstrated that the open loop is equivalent to the closed loop, and both techniques bring similar results. It is shown that measuring the flux change around oscillated samples provides valuable information on partial neutron cross sections. Second, uncertainty management and methods’ reproducibility are discussed. As expected, both techniques can achieve similar performances. In this paper, uncertainty as low as 0.007 pcm for samples’ reactivity worth and $1.2 \times 10^{-4}$ for local flux change were achieved.

Innovative hybrid pile oscillator technique in the Minerve reactor: open loop vs. closed loop

Pile oscillator techniques are powerful methods to measure small reactivity worth of isotopes of interest for nuclear data improvement. This kind of experiments has long been implemented in the Mineve experimental reactor, operated by CEA Cadarache. A hybrid technique, mixing reactivity worth estimation and measurement of small changes around test samples is presented here. It was made possible after the development of high sensitivity miniature fission chambers introduced next to the irradiation channel. A test campaign, called MAESTRO-SL, took place in 2015. Its objective was to assess the feasibility of the hybrid method and investigate the possibility to separate mixed neutron effects, such as fission/capture or scattering/capture. Experimental results are presented and discussed in this paper, which focus on comparing two measurements setups, one using a power control system (closed loop) and another one where the power is free to drift (open loop). First, it is demonstrated that open loop is equivalent to closed loop. Uncertainty management and methods reproducibility are discussed. Second, results show that measuring the flux depression around oscillated samples provides valuable information regarding partial neutron cross sections. The technique is found to be very sensitive to the capture cross section at the expense of scattering, making it very useful to measure small capture effects of highly scattering samples.

Dead time corrections on the Feynman-Y curve using the backward extrapolation method

ABSTRACT Dead time losses in neutron detection, caused by both the detector and the electronics dead time, is a highly nonlinear effect, known to create high biasing in physical experiments as the power grows over a certain threshold. Analytic modeling of the dead time losses is a highly complicated task due to the different nature of the dead time in the different components of the monitoring system (paralyzing vs. non-paralyzing), and the stochastic nature of the fission chains. The most basic analytic models for a paralyzing dead time correction assume a non-correlated source, resulting in an exponential model for the dead time correction. While this model is often used and very useful for correcting the average count rate in low count rates, it is totally impractical in noise experiments and the so-called Feynman-α experiments. In the present study, a new technique is introduced for dead time corrections, based on backward extrapolation of the losses, created by imposing increasing artificial dead time on the data, back to zero. The method is implemented on neutron noise measurements carried out in the MINERVE reactor, demonstrating high accuracy in restoring the corrected values of the Feynman-Y variance-to-mean-ratio.

A hybrid pile oscillator experiment in the Minerve reactor

Pile oscillator techniques are powerful methods to measure small reactivity worth of material of interest for nuclear data improvement. In the Minerve reactor, operated by CEA Cadarache, experiments are being held using a reactivity oscillator method based on the measurement of the global reactivity perturbation associated to samples motion in the core center. The technique has recently been extended by introducing new miniature fission chambers close to the oscillation location, in order to measure local neutron flux perturbations. New detectors and associated acquisition systems were developed for that purpose by CEA. First results are presented and discussed in this paper. Data processing techniques are presented in details which demonstrate that global and local perturbations can be unraveled. Experiments also demonstrate that measuring the local flux depression around oscillated samples indeed provide valuable information regarding partial cross sections. The preliminary results are found to be very sensitive to capture cross section at the expense of scattering cross section, making the technique very useful to measure small capture effects of highly scattering samples and improve nuclear data knowledge, as foreseen in the future CEA’s ZEPHYR critical facility to be built in Cadarache around 2026.

Experimental validation of analytic formulas for the statistical uncertainty in the Feynman-[formula omitted] method

Noise experiments, and the Feynman-α method in particular, are considered a standard in-pile experiment, often used in zero power reactors in a sub critical configuration to estimate the reactivity level. In the Feynman-α method, the variance to mean ratio of event triggered neutron count is fitted on the so called Feynman-Y function, resulting with an experimental estimation of the α decay mode.In a recent study, analytic formulas for the expected statistical uncertainty in the Feynman-α method were derived, using the backward stochastic transport equation. The outline of the present study is to experimentally validate these analytic formulas. The formulas were implemented on 4 different signals, obtained from the MINERVE reactor at CEA Cadarache. For each of the four measurements, the signal was truncated in increasing measurement lengths (ranging from 1 to 60min), in order to observe the functional behavior of the statistical error through time. Then, the measured statistical error was compared with the analytic estimation.Results indicate a very high correspondence between the expected statistical error and the measured one.

Effects of Am and Cm distributions on neutronic parameters of MINERVE reactor

Minor actinide recycling and transmutation are important stages in the advanced nuclear systems design such as Generation IV reactors that offers a high level of safety, waste reduction, and minimizes its environmental and radiological impacts. In the present work, two preliminary model configurations are used to examine the effects of adding two long-lived isotopes of minor actinides (243Am and 247Cm) in MINERVE reactor reference core (R1-UO2) that represents a LWR loaded with UO2 fuel rods. The effects are studied in two models of uniform and nonuniform minor actinides distributions, and include the numerical evaluation of neutronic parameters such as neutron spectrum, flux distribution and fission rate, using MCNPX transport code. The analysis of and comparison between the uniform and nonuniform models have been discussed to evaluate which of either is expected to be favored for minor actinide transmutation and energy production.

Americium-241 integral radiative capture cross section in over-moderated neutron spectrum from pile oscillator measurements in the Minerve reactor

An experimental program, called AMSTRAMGRAM, was recently conducted in the Minerve low power reactor operated by CEA Cadarache within the frame of the CHANDA initiative (Solving CHAllenges in Nuclear Data). Its aim was to measure the integral capture cross section of 241 Am in the thermal domain. Motivation of this work is driven by large differences in this actinide thermal point reported by major nuclear data libraries. The AMSTRAMGRAM experiment, that made use of well characterized EC-JRC americium samples, was based on the oscillation technique commonly implemented in the Minerve reactor. First results are presented and discussed in this article. A preliminary calculation scheme was used to compare measured and calculated results. It is shown that this work confirms a bias previously observed with JEFF-3.1.1 (C/E-1 = −10.5 ± 2%). On the opposite, the experiment is in close agreement with 241 Am thermal point reported in JEFF-3.2 (C/E-1 = 0.5 ± 2%).

Kinetic Parameter Measurements in the MINERVE Reactor

In the framework of an international collaboration, teams of the PSI and CEA research institutes measure the critical decay constant ( $\alpha _{0}=\beta /\Lambda $ ), delayed neutron fraction ( $\beta $ ) and generation time ( $\Lambda $ ) of the Minerve reactor using the Feynman- $\alpha $ , Power Spectral Density and Rossi- $\alpha $ neutron noise measurement techniques. These measurements contribute to the experimental database of kinetic parameters used to improve nuclear data files and validate modern methods in Monte Carlo codes. Minerve is a zero-power pool reactor composed of a central experimental test lattice surrounded by a large aluminum buffer and four high-enriched driver regions. Measurements are performed in three slightly subcritical configurations (−2 $\text {c} \!\! | $ to −30 $\text {c} \!\! | $ ) using two high-efficiency 235U fission chambers in the driver regions. Measurement of $\alpha _{0}$ and $\beta $ obtained by the two institutes and with the different techniques are consistent for the configurations envisaged. Slight increases of the $\beta $ values are observed with the subcriticality level. Best estimate values are obtained with the Cross-Power Spectral Density technique at −2 $\text {c} \!\! | $ , and are worth: $\beta =716.9\pm 9.0$ pcm, $\alpha _{0}=79.0\pm 0.6\,\,\text {s}^{-1}$ and $\Lambda =90.7\pm 1.4\,\,\mu \text {s}$ . The kinetic parameters are predicted with MCNP5-v1.6 and TRIPOLI4.9 and the JEFF-3.1/3.1.1 and ENDF/B-VII.1 nuclear data libraries. The predictions for $\beta $ and $\alpha _{0}$ overestimate the experimental results by 3-5% and 10-12%, respectively; that for $\Lambda $ underestimate the experimental result by 6-7%. The discrepancies are suspected to come from the driven system nature of Minerve and the location of the detectors in the driver regions, which prevent accounting for the full reactor.

Design of a high order Campbelling mode measurement system using open source hardware

This paper reviews a new real-time measurement instrument dedicated for online neutron monitoring with fission chambers in nuclear reactors. The instrument implements the higher order Campbelling methods and self-monitoring capabilities on an open source development board. The board includes an CPU/FPGA System on a Chip.The feasibility of the measurement instrument was tested both in laboratory with a signal generator and in the Minerve reactor. It is shown that the instrument provides reliable and robust count rate estimation over a wide reactor power range based on the third order statistics of the fission chamber signal. In addition, the system is able to identify whether the measured count rate change is due to the malfunction of the detector or due to the change in the neutron flux. The applied self-monitoring method is based on the spectral properties of the fission chamber signal. During the experimental verification, the considered malfunction was the change of the polarization voltage.

Performance of Higher Order Campbell methods, Part II: calibration and experimental application

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.

Experimental study of columnar recombination in fission chambers

In this paper, we present experimental saturation curves of a small gap miniature fission chamber obtained in the MINERVE reactor. The chamber is filled with argon at various pressures, and the fissile material can be coated on the anode, cathode, or both. For analyzing the recombination regime, we consider a model of columnar recombination and discuss its applicability to our chamber. By applying this model to the data, it is possible to estimate the ratio between the recombination coefficient k and an effective column radius b, appearing in the model, to be k/b=(2.5±0.9)×10−6m2/s for argon. From these results, a routine measurement of the recombination regime is proposed in order to detect gas leakage. This online diagnosis would be beneficial in terms of lifetime and reliability of the neutron instrumentation of nuclear reactors.

Analysis of safety parameters for the MINERVE reactor

Abstract The present study aims to show the effect of different cross section libraries on MINERVE reactor safety parameters. The MCNP5 calculation model of the MINERVE reactor facility is used to determine core and safety parameters such as axial and radial fission rate distributions, control rod worth and spectral indices. Different neutron spectra were achieved by changing the experimental lattice within the MINERVE reactor. MINERVE provides a large experimental basis for the improvement of the cross section databases. The current study calculates these parameters by MCNP5 code using three different cross section libraries (the continuous energy cross sections of the ENDFB-VI, T16-2003, and ENDF/B-VII.1 libraries). Neutronic calculations were performed for R1UO2, R1MOX, R2UO2, and R2MOX core configurations, representative of a LWR loaded with UO2, mixed oxide matrix, over moderated UO2 and over moderated MOX, respectively. The study aims to determine the most suitable cross section library to be used with MCNP5 reactor calculation code for light water reactor fuel lattice. The MCNP5 results were compared with the experimental results.

First burnup credit application including actinides and fission products for transport and storage cask by using French experiments

The burnup credit (BUC) methodology for a transport and storage cask application, including actinides and fission products, is implemented at AREVA TN using the French BUC calculation route for pressurized water reactor (PWR) UO2 used fuel. The methodology is based on the connection of the French depletion code DARWIN2 and the French criticality safety package CRISTAL V1. The BUC methodology includes the experimental validation of the computation codes dedicated to the calculation of the used fuel inventory calculations. Indeed, the results of the comparison calculation–experiment (C−E)/E allow to determine either a set of isotopic correction factors (ICFs) for the BUC nuclides considered in the criticality calculation or keff-penalty terms directly used for the definition of the keff-acceptance criterion for the criticality assessment of the transport and storage cask. These ICFs or keff-penalty terms are one of the key of the BUC method to guarantee the conservativeness of the fuel reactivity in safety-criticality calculations using BUC approach. A French BUC program has been developed at CEA/Cadarache in the framework of the CEA–AREVA collaboration in order to validate fuel inventory calculations. This program involves two kinds of experiments: chemical analyses and microprobe measurements of PWR irradiated fuel pins (French PIE program) on one hand, and reactivity worth measurements of the BUC nuclides in the MINERVE reactor on the other hand. This paper highlights, through a first industrial AREVA TN's application of the BUC method, including fission products, that the French PIE program and reactivity worth measurements in MINERVE reactor are suitable for the implementation of BUC in transport and storage cask applications loaded with PWR UO2 used fuels assemblies.