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The DISCOMS project (Distributed Sensing for Corium Monitoring and Safety) aimed at providing innovative solutions not requiring local electrical power supplies, for remote monitoring of a severe nuclear accident. The solutions are based on both long length SPNDs (Self Powered Neutron Detectors) and on distributed OFSs (Optical Fibre Sensors) capable to detect the onset of a severe accident, the corium pouring on the containment building concrete basemat, and its interaction with the concrete floor under the reactor vessel, until it spreads in the core catcher (EPR case). This paper mainly focuses on these last three detection targets achievable with distributed OFSs. It is based on the results of a Molten Core & Concrete Interaction (MCCI) experiment, namely VULCANO, held in June 2018 with a concrete crucible equipped with overall ~ 180 m long optical fibre sensing cables. This small scale experiment (50 kg of prototypical corium) has demonstrated the ability of distributed OFSs to remotely provide useful data during the MCCI run: i) temperature profiles images up to about 580°C (single wavelength Raman DTS reflectometer) until cooling down to room temperature, ii) high spatial-resolution frequency shifts profiles, due to combined (non-selective) strain and temperature influences (Rayleigh OFDR and Brillouin reflectometers), and iii) cables lengths ablated by the corium on sections weakened by the temperature (Raman DTS, Rayleigh OFDR, telecom and photon counting reflectometers). [ABSTRACT FROM AUTHOR]

Up to now, predicting accurately the Fission Gas Release (FGR) from high burn up UO2 and/or MOX (Mixed Oxide) fuels at off-normal conditions, such as power transient, reactivity-initiated accident (RIA) and loss-of-coolant accident (LOCA), is still a significant and very challenging task. For this purpose, different R&D programs have been carried out in France, as well as in other countries. This has been done with a specific emphasis on mechanisms which promote the FGR under accidental conditions. These studies can be performed thanks to dedicated integral experiments conducted in-pile (i.e. in Materials Testing Reactor) with the corresponding cost and constraints, or at the laboratory scale with annealing tests which allow to be representative of specific parameters (thermal history for instance). During these annealing tests under well-known conditions (temperature, atmosphere), both the absolute level and the time dependence of the released gases should be monitored, together with the corresponding fuel micro-structural changes, since experimental knowledge of fission gas release alone is not efficient enough. This approach requires more and more accurate on-line measurements. This corresponds to the driving force of the present work. In this contribution, we will present our progress in developing an experimental platform that can submit nuclear fuel and cladding samples to annealing tests involving very high temperatures (up to 2500°C) and very fast temperature ramp (up to thousands of °C/s) with controlled thermal gradients and temporal dynamics. This new platform implements innovative instrumentation, such as optical diagnostics to measure fuel fragmentation kinetics and infrared pyrometry for temperature monitoring. This experiment is based on a high-power laser (1.5kW) coupled to an experimental chamber with controlled atmosphere (Ar, N2, or vacuum) and specific optical components. Based on the spatial beam profile and temporal power function of the laser, it is possible which such a system to produce complex spatio-temporal temperature gradients, relevant for addressing different research needs. It provides access to extreme conditions that are very difficult to reach with other means. Particularly, one of main objectives of this work is to investigate conditions of Reactivity Initiated Accident (RIA). The first experiments performed on inactive materials, non-irradiated uranium dioxide, is presented in order to highlight the capabilities of this technique. [ABSTRACT FROM AUTHOR]

In two earlier papers [1], [2] we investigated the possibility of extracting traditional multiplicity count rates from the cumulants of fission chamber signals in current mode. It was shown that if all neutrons emitted from the sample simultaneously are also detected simultaneously, the multiplicity rates can be retrieved from the first three cumulants of the currents of up to three detectors, but the method breaks down if the detections of neutrons of common origin take place with a time delay spread wider than the pulse shape. To remedy these shortcomings, in this work we extended the theory to two- and three-point distributions (correlations). It was found thatthe integrals of suitably chosen two- and three-point moments with respect to the time differences become independent of the probability density of the time delays of detections. With this procedure, the multiplicity rates can be retrieved from the detector currents for arbitrary time delay distributions. To demonstrate the practical applicability of the proposed method, a measurement setup was designed and built. The statistics (shape and amplitude distribution) of the detector pulse were investigated as important parameters of the theoretical model. Simulations were performed to estimate the expected value of the multiplicity rates in the built setup. Measurements were performed and two types of moments (the mean and the covariance function) of the recorded detector signals were calculated. Values of singles rates were successfully recovered. [ABSTRACT FROM AUTHOR]

An advanced neutron detection system for highly localized measurements in nuclear reactor cores was developed and tested in the Laboratory for Reactor Physics and System Behaviour (LRS) at the École polytechnique fédérale de Lausanne (EPFL), Switzerland, in close collaboration with the Detector group of the Laboratory for Particle Physics (LTP) at the Paul Scherrer Institute (PSI), Switzerland. The miniature-size detector is based on the coupling of a ZnS:6LiF scintillator/converter screen of 1 mm2 and 0.2 mm thickness with a 10-m optical fiber, the latter being connected to a silicon photomultiplier (SiPM). In this development version, the output signal is processed via analog read-out electronics. The present work documents the characterization of a detection system prototype in the mixedradiation fields o f t he C ARROUSEL f acility a nd i ts t esting in the CROCUS zero-power reactor operated at LRS. The fibercoupled scintillator shows a linear response with the reactor power increase up to 6.5 W (i.e. around 108 cm-2s-1 total neutron flux), with a s ubsequent l oss o f l inearity d ue t o e lectronic dead time of the analog system. Nevertheless, the detector shows excellent neutron counting capabilities whether compared to other localized detection systems available at LRS, e.g. miniature fission chambers and an sCVD diamond detector. [ABSTRACT FROM AUTHOR]

Neutron detection systems utilize statistical alarm techniques where a measured false alarm rate (FAR) can vary drastically from the FAR predicted by a theoretical model. The ability to set an alarm threshold that results in a practically controlled FAR is crucial to characterize detector sensitivity with both accuracy and precision. A generalized and automated method is presented to statistically evaluate FAR performance by assuming that the FAR itself is not deterministic, but a normal stochastic process over a specific parameter to be corrected that will hereafter be referred to as the correction. In this manner, a specific correction results in not only a point estimate of FAR, but also a confidence interval. The central objective is focused exclusively on characterization assuming that experiments are executed in a tightly controlled environment so that an accurate comparison is enabled across detectors. Once a correction is calculated, the estimated FAR is only assumed accurate in a similar environment for sensitivity evaluation. Initially, the calculated correction factor was used to compare FARs across various distributions including normal, corrected normal, Poisson, and a simplified normal distribution. Later verification data sets were used to empirically demonstrate the rate of containment of measured confidence coefficients using two detectors of different technology. A second application uses the correction method to improve the signal-to-noise ratio metric to agree more with dynamic sensitivity results. Finally, a third application studies the effect of altering the duration of background acquisition on FAR performance. [ABSTRACT FROM AUTHOR]

Each year States report to IAEA loss, theft or out of regulatory control radioactive materials. Most incidents are minor, but material is potentially available for criminal acts. Measures to reduce the radiological and nuclear threat are many-faceted. An important component is the ability to detect illicit transport of radioactive material. Spectroscopic Radiation Portal Monitors are deployed around the world to detect illegal radioactive material traffic. A combination of experimental data collected during testing campaigns and simulations is a good way to study the performance of Spectroscopic Radiation Portal Monitors in realistic conditions. The paper presents a process to evaluate the performances of a portal, based on a combination of experimental data and MCNP simulations to calculate the detection probability and the false alarm rate. IRSN developed platforms for testing Spectroscopic Radiation Portal Monitors for pedestrian control. Experimental data were collected from an available commercial Spectroscopic Radiation Portal Monitor, tested in the framework of the Illicit Trafficking Radiation Assessment Program phase II Round Robin Test. Many scenarios were tested with different sources using realistic setups and many experimental data were collected. The tested equipment was then simulated with MCNP only based on the data provided in the user manual and the standards found in the industry. To get a realistic idea of the uncertainty, all the variables inherent in the measurement were considered, their relative contributions were identified and quantified, then propagated to predict an overall uncertainty. The combination of experimental data, numerical simulations and uncertainty evaluation showed good agreement with experimental assays. The results were used to test the sensitivity of a Spectroscopic Radiation Portal Monitor to special nuclear materials for different alarm thresholds. This process applied to different scenarios according to defined targets should help in the selection of operating characteristics of the portal. [ABSTRACT FROM AUTHOR]

6LiF:ZnS(Ag) micro-particle neutron detectors are a promising technology to further improve neutron detection capabilities for a variety of applications. Specifically, we have been investigating 6LiF micro-particles coated with ZnS(Ag) to increase the neutron detection efficiency, light production, and light collection efficiency when compared to the existing powder-based technology (EJ-426 from Eljen Technology). Extensive radiation and light transport simulations with single micro-particles have been performed to find the optimal 6LiF diameter and ZnS(Ag) coating thickness. Full-scale multi-particle simulations also have been performed to determine the optimal pitch (particle-to-particle distance) and detector thickness. Randomizations of 6LiF radius, ZnS(Ag) coating thickness, position of particles, as well as shape of particles and partial coating have been performed to account for possible manufacturing imperfections. EJ-426 sheets have been modeled for reference purposes by defining spherical grains of 6LiF and ZnS(Ag) and compared against experiments. The simulation results show that the coated micro-particles should dramatically increase the neutron detection efficiency, light production, and light collection efficiency when compared to the existing EJ-426 technology. [ABSTRACT FROM AUTHOR]

Activated cooling water in nuclear facilities can present a significant radiation source around primary cooling system causing radiation damage to electrical components, increasing doses to personnel and in the case of fusion facilities additional heating to superconducting coils. As there are only few sources of gamma rays with energies in the range of 6 MeV and 7 MeV an irradiation system using activated cooling water as the source of energetic gamma rays is proposed at the Jožef Stefan Institute (JSI) TRIGA Mark II research reactor. Two different conceptual designs, one utilizing central irradiation channel and one utilizing radial piercing port for water activation, are presented and analysed in the paper. Despite an order of magnitude higher water activation in central channel compared to radial piercing port the 16N decay rate in the irradiation facility is comparable between both design (order of 108 decays per second) due to longer transient time from central channel to irradiation facility. In the irradiation facility the expected biological dose rates due to the 16N decay rate are in order of several mSv/h. From the results he conceptual design utilizing the radial piercing port currently presents the best option for the irradiation facility due to the simpler design of the irradiation loop, already present shielding of the loop and comparable number of 16N decay rates to central channel. [ABSTRACT FROM AUTHOR]

Increased sensitivity to nuclear safety and security issues has prompted public entities and private institutions to maximize their capability to rapidly assess risks and intervene in the case of accident or threat. Quick intervention and response are achieved through nuclear measurements via airborne, land, and underwater systems. A network of cohesive, well-integrated and easy deployable radiation monitoring systems combined with real-time analysis of data is essential to facilitate and enhance the decision-making process during these most critical moments enhancing the quality of the management plan. The presented radiation monitoring systems can be integrated in several form factors which depend mainly on operational needs and internal battery for autonomous operation. Compact ARM based computers is embedded, which can store large amount of data in their non-volatile memory, run automatic data analysis and trigger alarms in case of exceeding radiation levels. All the systems can communicate with redundant interfaces in failover configuration and upload the acquired environmental information in a central database. The same monitoring systems can alert the emergency response personnel on the field as well, through wireless connection to common tablets or cellphone or SMS, guaranteeing a prompt response in case an illicit transportation of radiological or nuclear material is detected. [ABSTRACT FROM AUTHOR]

The design and configuration of a multi-layered imaging system with the ability to detect thermal neutrons, fast neutrons and gamma rays has been developed and its efficacy demonstrated. The work presented here numerically determines the systems efficiency and spatial resolution, using 252Cf and 137Cs as a case study. The novelty of this detection system lies in the use of small form factor detectors in a three-layer design, which utilises neutron elastic scattering and Compton scattering simultaneously. The current configuration consists of 10 mm thick natural lithium glass (GS10) scintillator integrated with a 20 mm thick plastic scintillator (EJ-204) in the first layer, a 15 mm thick lithium glass (GS10) scintillator in the second and a 30 mm thick CsI(Tl) scintillator forming the final layer. Each of these layers is backed with an 8 x 8 silicon photomultiplier diode (SiPM) array. The overall size of the imaging system is 27 mm x 27 mm x 135 mm. MCNPv6.1 and Geant4-10.04 were alternatively used to optimise the overall configuration and to investigate detection modalities. Results show promising performance with high precision source localisation and characterization abilities. Measurements were virtually obtained of two gamma-ray sources within steel enclosures at angles of 15°, 30° and 50° separation in order to test spatial resolution ability of the system. With the current active size of the system and the 8x8 SiPM configuration, the results estimate the spatial resolution to be close to 30°. The ability of the system to characterise and identify sources based on the type and energy of the radiation emitted, has been investigated and results show that for all radiation types the system can identify the source energy within the energy range of typical reported sources in literature. [ABSTRACT FROM AUTHOR]

The deep geological repository system provides long-term protection against the undesirable effects of ionizing radiation on the population and the environment. An important part of the long-term safety strategy is development of a monitoring program that collects information about the behaviour of the deep geological repository throughout its whole lifetime. A simplified model of the disposal system, geosphere, and biosphere was developed using the GoldSim simulation tool to demonstrate the behaviour of the hypothetical deep geological repository located in crystalline rocks. Also an initial model of the reference biosphere was created based on the scenario of an agricultural habitation (normal evolution scenario) and it was developed based on the recommendations provided in the BIOMASS methodology. After a significant period of time, disposal containers will be degraded and evolution changes in the repository system will occur. Several important parameters appear in the annual effective dose calculation for an individual from critical exposure group within the reference biosphere model. One of them are, for example, distribution coefficients and so-called translocation factors that define the transported rate of released radionuclides into the environment. This paper provides a view into the selected part of the deep geological repository through the data obtained by monitoring during the selected period of time. Simulations describing changes in the repository system. The aim of this contribution is to evaluate the impact of selected changes on the annual effective dose for an adult individual from a critical exposure group while it is assumed, that the respondent consumes contaminated crops and animal products. This model includes various biosphere components and multiple exposure pathways such as inhalation, ingestion and external exposure. [ABSTRACT FROM AUTHOR]

In the context of nuclear security, uranium ore concentrates (UOCs) play an important role: they are traded in large quantities and this makes their use "out of regulatory control" a possible scenario. Once an incident of illicit trafficking o f n uclear m aterial is detected, an understanding of its origin and production process is required; this implies the necessity to use analytical techniques able to measure characteristic parameters (e.g. physical, chemical, isotopic characteristics of the nuclear materials) which are referred to, in the field o f t he n uclear f orensics, a s signatures. The present study investigates the potential of image texture analysis (i.e. the angle measure technique), combined with the spectrophotometric determination of colours for the evaluation of the origin of several UOCs. The use of different multivariate statistical techniques allows the categorization of about 80 different samples into a few groups of UOCs powders, which makes this approach a promising method complementing the already established methods in nuclear forensics. [ABSTRACT FROM AUTHOR]

A decommissioning of nuclear fuel cycle facilities is inseparable from the problems of radioactive waste disposal. One of these problems is the categorization of a waste according to the content of beta- and alpha-emitters. Beta-emitters can be identified by existing technologies; however, the trouble arises when detecting alpha-emitting elements, primarily the long-lived members of the actinium chain with the specific activity of kBq/kg when they are spread inside a structural material. The report considers an application of an active neutron method-a differential die-away technology for reliable control of small quantities of FM. The essence of this method consists in sounding the interrogated item by pulsed thermal neutrons and recording the induced fission neutrons. The ratio of the number of fission neutrons to the number of source neutrons gives the normalized number of fission neutrons that is linked to the FM mass in the interrogated object. The work presents the scheme and principle of operation of an experimental device, as well as the results of measurement of concrete structures that contain internal traces of fissile materials. Analysis of the results shows that the proposed method allows the detection of ~ 6 mg of fissile material per kg of concrete with possible localization (cartogram) of the contaminated area. [ABSTRACT FROM AUTHOR]

Limitations of fast neutron radiography include low detection efficiency and poor spatial resolution. D-T neutron radiography is one compact fast neutron radiography method. Based on D-T associated alpha particle method and coded source imaging method, we indicate one new method to improve resolution of D-T neutron radiography. This method could get distribution of D-T neutrons by detecting alpha particles. Without real coded mask, the D-T radiography structure is considered as coded source imaging of fast neutrons. With reconstruction method, the real object could be reconstructed from projections. One prospect setup of D-T associated alpha neutron source has been carried out with Monte-Carlo simulation. The projection images of two different situations are collected and reconstruction results show that it's possible to improve image quality of D-T neutron radiography. [ABSTRACT FROM AUTHOR]

In this research, a combined fast-neutron/γ-raybackscatter imaging technique is described. The aim of this work is to understand corrosion defects in pipelines by measuring differences in the scattered radiation flux, generated when different steel thicknesses are irradiated by a neutron and γ-ray focused beam. A californium-252 radiation source is used to produce fast neutron and γ rays, exploiting its spontaneousfission. This mixed radiation field is collimated and directed towards the steel samples. Backscattered neutrons and γ rays aremeasured as a function of the steel thickness using 4 liquid organic scintillation detectors linked to a real-time, pulse-shape discrimination system, which separates and retains the neutron and γ-ray event data. In this paper, we describe how, using asingle radiation source and detection system, it is possible to perform and combine two complementary imaging modalities. This research is validated by an MCNP6 computer simulation study. The backscatter imaging system developed for this research and the experimental results of the measurements carried out using the National Physical Laboratory neutron low-scatter facility are also presented in this paper. [ABSTRACT FROM AUTHOR]

As part of its R&xD activities in the fields of radioactivewaste drum storage and homeland security, the NuclearMeasurement Laboratory of CEA Cadarache has started studiesrelated to the detection of induced delayed fission gamma rays asa signature of U/Pu presence either in radioactive wastes or incargo containers and luggage. The study described in the presentpaper explores the feasibility of detecting fission delayed gammarays of nuclear materials interrogated by a pulsed neutrongenerator. For this purpose, Monte Carlo simulations have beenperformed with ACT, the MNCP6 Activation Control Card.Simulated results have been compared with experimental data tovalidate the numerical model. Samples of uranium andplutonium have been irradiated for 2 hours with a pulsed D-Tneutron generator delivering 14 MeV neutrons with an averageemission of 8.107 n/s, which are thermalised in a graphite cellcalled REGAIN. At the end of irradiation, activated nuclearmaterials were placed in a low-background, high-resolutiongamma spectroscopy station in order to detect delayed gammarays emitted by fission products. Anomalies have been observedin the calculated time decay curve of fission delayed gamma rayswith MCNP6 ACT card, but the time behavior is correct for non-fission activated materials like aluminum or copper. On the otherhand, the number of counts recorded in the main simulatedgamma ray lines from activated nuclear material fission productsis consistent with the experimental results, thus validating thesimulation scheme in view of further studies on thecharacterization of radioactive waste drums or special nuclearmaterial detection in cargo containers. [ABSTRACT FROM AUTHOR]

Alpha emitters are usually identified and quantified by alpha spectrometry measurements in a vacuum chamber performed in laboratory environments. This study shows that transuranic elements can be distinguished under ambient conditions using a grid collimator. The aim of this work was to use numerical simulations with the MCNP6 code to design a grid with a resolution high enough to differentiate the same radionuclide combinations as alpha spectrometry in a vacuum chamber, namely 239Pu + 240Pu, 241Am + 238Pu and 244Cm. Results show that a compromise is required to obtain the best performances in terms of energy resolution and detection efficiency, leading to the choice of two hexagonal grid collimators. The first has a collimation height of 0.5 cm and an apothem of 1 mm. Laboratory tests on electrodeposited sources show that the target radionuclides can be identified without prior deconvolution, with an energy resolution of about 70 keV and a detection efficiency of 0.74% at incident energies of 5–6 MeV. The second grid has the same collimation height but a coarser mesh with an apothem of 2 mm. In this case, the alpha peaks are still distinguishable, but with a lower resolution of 125 keV. The detection efficiency is three times higher however. [ABSTRACT FROM AUTHOR]

Methods for the non-destructive assay of nuclear waste drums are of great importance to the nuclear waste management community, especially where loss in continuity of knowledge about the content of drums happened or chemical processes altering the contents of the drums may occur. Muon scattering tomography has been shown to be a promising technique for the non-destructive assay of nuclear waste drums in a safe way. By measuring tracks of muons entering and leaving the probed sample and extracting scattering angles from the tracks, it is possible to draw conclusions about the contents of the sample and its spatial arrangement. Within the CHANCE project, a newly built large-scale mobile detector system for scanning and imaging the contents of nuclear waste drums using atmospheric muons is currently undergoing commissioning. [ABSTRACT FROM AUTHOR]

Compton imaging is one of the main methods to localize radioactive hotspots, which emit high-energy gamma-ray photons, above 200 keV. Most of the Compton imaging systems are composed by at least two detection layers or one 3D position sensitive detector. In this study, we demonstrate the application of a new miniature pixelated single plane detector to Compton imaging. In this configuration, we do not have the information on interaction depth but we successfully test its ability to perform Compton localization by means of comparing different Compton reconstruction algorithms applied to real data measured with our single plane detection system. [ABSTRACT FROM AUTHOR]

This work proposes a solution to identify the number of sources of radiation, as well as their respective intensities and locations based on data acquired by Global Positioning System (GPS) receivers and affordable radiological sensors, such as Geiger-M¨uller counters (GMC). An optimization algorithm is required to minimize the estimation error in terms of location, intensity and number of sources of radiation given all the intensity measurements acquired in different locations, taking into account the sensors' models, background radiation intensity values and noise. Experimental results were achieved in a laboratory with controlled sources of radiation. The solution was also tested with real data gathered by a GMC connected to a mobile phone with a software application developed by the authors to synchronize the sensor readings with GPS data. The sensor and the mobile phone are attached to a quadcopter flying over the scenario with sources of radiation. [ABSTRACT FROM AUTHOR]

In 2003, the work on environmental rehabilitation of the former coastal technical base of the Russian Navy in Gremikha town, with technical and financial support of the State Atomic Energy Corporation "Rosatom" and the French Alternative Energies and Atomic Energy Commission had begun. In 2017, the task of searching, characterization and packaging in transport containers for export to the long-term storage of high-level waste containing 152Eu was set as part of this program. These wastes were presented as fragments of reactor control and safety system rods of first generation nuclear submarine reactors and were placed in temporary storage containers mixed with other radioactive waste and construction garbage on the territory of the Gremikha branch of the North-West Center "SevRAO". To accomplish the task, the spectrometric system "Gamma-Pioneer" was developed and made at National Research Center "Kurchatov Institute". The system consist of a collimated spectrometric gamma-ray CdZnTe detector 500 mm3 volume, a dosimeter for continuous measurement of ambient equivalent dose rate at the instrument location and a color video camera with zoom lens. The device is placed in a protective frame and installed as an attachment implements on robot Brokk-90. Communication between the measuring system and the control computer is carried out using USB extender and 100 meters cable. The video camera is connected to the control computer via Ethernet. To search for the most active sources of gamma radiation, an equivalent dose rate calculation program using the results of measured instrumental spectra was developed. The characterization of the extracted fragments was carried out by activity evaluation of the main radionuclides polluting the samples under investigation. The paper describes design and operating principles of the system, as well as methods for the equivalent dose rate calculation and the activity evaluation of the fragments of reactor control and safety system rods based on spectral measurements obtained using the developed radiometric system [ABSTRACT FROM AUTHOR]

The EU-CHANCE project aims at the issue of the characterization of conditioned radioactive waste (CRW) and one objective of CHANCE is to focus on: Calorimetry as a comprising non-destructive technique to reduce uncertainties on the inventory of radwaste containing shielded and hidden material difficult to be measured by other means. A MCNP6-based numerical study comprising the particle flux out of a 200L mock-up drum in a Large Volume Calorimeter (LVC) currently manufactured by KEP Nuclear (France) will be presented and discussed. For the analyses, the particle flux and energy deposition in each layer of the calorimeter were determined. The results yield that a significant fraction of the radiation would leave the system and not contribute to the measurable heat deposition. The expected energy deposition is obtained and cumulated for each layer over the whole energy range revealing the fraction of particles actually escaping the LVC calorimeter. While this escape fraction needs and can be determined, the LVC is a very suitable apparatus for the anticipated experiments on large and heterogeneous waste drums that possibly contain deeply buried beta-emitters (e.g. Sr/Y-90) or shielded alpha-sources hidden inside the drum with a significant level of gamma and neutron radiation background radiation. The high-energy part of this gamma and neutron flux may even reach the reference chamber of the calorimeter and deposit some energy there, compromising the calibration and may cause a double-bias. [ABSTRACT FROM AUTHOR]

A set of In-Pile Sections (IPS) has been irradiated in the BR2 reactor at SCK•CEN in Belgium during the 1970's and 1980's. The primary goal of the IPS was to replicate the thermo-hydraulic loop of a sodium-cooled fast reactor in order to study severe accident scenarios. The top part of the IPS contained the sodium-cooled loop whereas the lower part contained the fuel element. Due to the experimental conditions, the rupture of the fuel pins contained in the IPS occurred and fuel fragments may have been deposited in the rest of the IPS loop. The part of the IPS containing the fuel pins has been cut from the rest of the IPS and underwent post-irradiation examinations at specialized EU laboratories, while the top parts remained stored at SCK•CEN. To prepare for future transport, dismantling and conditioning, a reliable estimation of the total fissile content in the stored parts of the IPS is indispensable. In this framework, two IPS were measured with a Canberra WM3400 neutron coincidence counter with customized electronics. The measurements of the IPS were challenging due their length (roughly 6 m) and intense gamma-ray radiation background. For each IPS an axial scan was carried out with a series of short measurements (600-700 s each) recording the Totals rate and Reals rate. Based on the results of the axial scans, measurements with longer measurement time were conducted for the axial positions with the larger values of Reals rates. A system of equations was then established to quantify the 240Pu content in the different sections of the IPS from the Reals rates in each measurement position and account for cross-talk between the neutron emission associated to the different sections. A set of Monte Carlo simulations was carried out to estimate the probability to record a Real count in the detector due to spontaneous fission events occurring in a given section of the IPS. The 240Pu content in each section of the IPS was calculated by combining the measured Reals rates and the detection probabilities calculated with the simulations. The total fissile content in the IPS was then determined with scaling factors based on burnup calculations for the irradiated fuel assemblies in the IPS. The results indicate that both IPS measured with the neutron coincidence counters have a fissile content lower than the limit for transport. It is expected that the envisaged segmentation of the IPS in shorter sections required to fit into 200L drums will provide an additional safety margin on this limit. [ABSTRACT FROM AUTHOR]

A neutron spectrometry campaign was carried out in the AMANDE accelerator control room. These measurements had several objectives, one of which was to verify the possibility of determining, with the HERMEIS Bonner sphere system, very low fluence and ambient dose equivalent rates. These measurements were also expected to provide comparison values with calculations, performed with MCNPX 2.6.0, used for modeling the whole facility and to verify the radiological zoning implemented. Neutrons of 3.3 MeV then 15 MeV were produced in the experimental hall of AMANDE and measurements were made in the control room, behind a 40 cm thick concrete wall. The ambient dose equivalent rates derived from the measured spectra are respectively of the order of 0.2 and 10 μSv/h, which are in agreement with a LB6411 surveymeter data also involved. [ABSTRACT FROM AUTHOR]

In the framework of the radioactive waste drum characterization using neutron coincidence counting, the Nuclear Measurement Laboratory of CEA Cadarache is studying plastic scintillators as an alternative to ideal but costly 3He gas proportional counters. Plastic scintillators are at least 5 times cheaper for the same detection efficiency, and in addition, they detect fast neutrons about three orders of magnitude faster than 3He detectors. However, they are sensitive to gamma rays, which implies the necessity to identify precisely gamma background sources that may affect the useful signal. This paper presents a detailed analysis of the gamma-ray spectrum of a radioactive waste drum containing glove box filters contaminated by plutonium dioxide. Gamma emissions accompanying inelastic scattering (n,n') and (α,n) reactions that can lead to neutron-gamma coincidences parasitizing useful coincidences from plutonium spontaneous fissions are identified. Some of these parasitic gamma rays having energies up to several MeV, we plan to reject high-energy scintillator pulses with an electronics rejection threshold above 1 MeV, which should preserve the major part of useful fission neutron pulses. [ABSTRACT FROM AUTHOR]

The present article presents the mechanical characterization of the fuel rods oscillator developed for the purposes of the COLIBRI experimental program in CROCUS. COLIBRI aims at investigating the radiation noise related to fuel vibrations. The main motivation is the increased amplitudes in the neutron noise distributions recorded in ex- and in-core detectors that have been observed in recent years in Siemens pre-Konvoi type of pressurized water reactors. Several potential explanations have been put forward, but no definitive conclusions could yet be drawn. Among others, changes in fuel assembly or pin vibration patterns, due to recent modifications of assembly structural designs, were pointed out as a possible cause. Computational dynamic tools are currently developed within the Horizon 2020 European project CORTEX, to help with understanding the additional noise amplitude. The COLIBRI program is used for their validation. An in-core device was designed, tested, and licensed between 2015 and 2019 for fuel rods oscillation in CROCUS, in successive steps from out-of-pile tests with dummy fuel rods to critical in-core tests. The characterization of its mechanical behavior is presented, in air and in water, and as a function of the load, for safety and experimental purposes. The device allows simultaneously oscillating up to 18 fuel rods. The maximum oscillation amplitude is 5 mm, while the maximum allowed frequency is 2 Hz, i.e. in the frequency range in which the induced neutron flux fluctuations are most pronounced in nuclear power plants. [ABSTRACT FROM AUTHOR]

Several non-destructive assays techniques have been developed for the measurement of fissile materials in the fields of dismantling, decommissioning, nuclear security, and nuclear safeguards. Among these techniques, neutron coincidence counting is based on the detection of time-correlated neutrons from induced and spontaneous fissions. 3He Tubes have been the primary choice for neutron coincidence counting due to their high detection efficiency, rather low sensitivity to gamma-rays and proven field reliability. This paper covers the implementation of a new electronic setup to a Canberra WM3400 neutron coincidence counter. First we describe the properties of the used detectors, with focus on the characteristics of the default electronics and highlight its limitations such as the high input capacitance, short shaping time and the necessity for selected tubes. We then propose the new electronic setup to overcome these limitations. This setup includes a dedicated preamp for every tube , the possibility to adjust for gain differences between the tubes and a better optimised shaping time for 3He detectors. We carried out measurements with the two electronic systems to compare their performances in terms of gamma-ray sensitivity, efficiency and die-away time. The gamma ray sensitivity was measured with calibrated 137Cs and a 60Co sources at the Laboratory for Nuclear Calibration of the Belgian Nuclear Research Centre with dose rates between 10 μSv/h and 50 mSv/h. Measurements with a 252Cf source were used to determine the die-away time of the system and the total measurement efficiency for the considered geometry. The measurements showed that, with the default electronics, neutron count-rates are already affected by gamma radiation at a dose rate of 10÷30 μSv/h. On the other hand the neutron coincidence counter equipped with the new electronics proved to be insensitive to gamma-radiation up to a dose rate of at least 20 mSv/h. The high-voltage set with the new electronics is lower than in the case of the default electronics and is within the range recommended by the tubes manufacturer. The die-away time was not affected by the used electronics. A reduction of about 20% in the neutron detection efficiency due to the used discriminator threshold was observed. [ABSTRACT FROM AUTHOR]

Gamma logging for uranium exploration are currently based on total counting with Geiger Müller gas detectors or NaI (TI) scintillators. However, the total count rate interpretation in terms of uranium concentration may be impaired in case of roll fronts, when the radioactive equilibrium of the natural 238U radioactive chain is modified by differential leaching of uranium and its daughter radioisotopes of thorium, radium, radon, etc. Indeed, in case of secular equilibrium, more than 95 % of gamma rays emitted by uranium ores come from 214Pb and 214Bi isotopes, which are in the back-end of 238U chain. Consequently, these last might produce an intense gamma signal even when uranium is not present, or with a much smaller activity, in the ore. Therefore, gamma spectroscopy measurements of core samples are performed in surface with high-resolution hyper-pure germanium HPGe detectors to directly characterize uranium activity from the 1001 keV gamma ray of 234mPa, which is in the beginning of 238U chain. However, due to the low intensity of this gamma ray, i.e. 0.84 %, acquisitions of several hours are needed. In view to characterize uranium concentration within a few minutes, we propose here a method using both the 92 keV gamma ray of 234Th and the 98.4 keV uranium X-ray. This last is due to uranium self-induced fluorescence caused by gamma radiations of 214Pb and 214Bi, which create a significant Compton scattering continuum acting as a fluorescence source and resulting in the emission of uranium fluorescence X-rays. The comparison of the uranium activity obtained with the 92 keV and 98.4 keV lines allows detecting a uranium heterogeneity in the ore. Indeed, in case of uranium nugget, the 92 keV line leads to underestimated uranium concentration due to gamma self-absorption, but on the contrary the 98.4 keV line leads to an overestimation because of increased fluorescence. In order to test this new approach, several tens of uranium ore samples have been measured with a handheld HPGe FALCON 5000 detector. [ABSTRACT FROM AUTHOR]

The knowledge of the fissile material mass is a key challenge to enhance radioactive waste management and to ensure a high level of safety in nuclear industry. Data is analyzed according to the principles of the neutron measurement techniques. As proportional counters filled with 3He gas display high neutron detection efficiency and a good gamma-ray discrimination, they are the reference detector for passive neutron coincidence counting. A charge preamplifier or a current amplifier, depending on applications, collects the electric pulse produced by neutron interaction in the 3He gas and a threshold discriminator produces a logic pulse used for neutron counting. This paper describes the performance assessment of different commercially available electronics from Mirion Technologies, Precision Data Technology (PDT), Mesytec, as well as MONACO electronics originally developed by CEA LIST for fission chamber measurements in experimental reactors. Comparative passive neutron measurements are carried out with these electronics at CEA/DEN Nuclear Measurement Laboratory in Cadarache. Overall, PDT and Mesytec electronics show similar detection efficiency as the ACH-NA98 charge amplifier, which is commonly used in our laboratory for such applications. However, MONACO electronics have a lower detection efficiency, similar to Mirion 7820 current amplifier used in specific high-count rate applications. An optimisation of MONACO settings would probably be necessary to adapt to 3He counters instead of fission chambers. [ABSTRACT FROM AUTHOR]

A pulse experiment model was validated in order to support future pulse experimental campaigns. All pulse experiments data was collected and are publicly available at http://trigapulse.ijs.si/. A comparison of the measured pulse physical parameters (maximal power, total released energy and full width at half maximum) and theoretical predictions (Fuchs-Hansen and the Nordheim-Fuchs models) was made. [ABSTRACT FROM AUTHOR]

Measurement of a cylindrical tungsten target reactivity worth has been performed on the light water zero-power reactor of VENUS-II at China Institute of Atomic Energy (CIAE) in order to verify the neutron evaluated data related to the engineering design of Chinese initiative Accelerator Driven Systems (CiADS). The reactivity worth of the tungsten target was measured and processed as -1.234±0.114mk by a period method. The experimental result was compared with the simulation ones calculated by MCNP with five different libraries, i.e., ENDF/B-VII.0, ENDF/B-VII.1, JENDL-4.0, CENDL-3.1 and JEFF-3.2. By comparing the results of experiment and simulation, the simulated results from ENDF/B-VII.0, JENDL-4.0 and JEFF-3.2 are higher than the experimental result, however that from CENDL-3.1 is lower. The result from ENDF/B-VII.1 library shows better agreement with the experiment one and the relative deviation is less than 2%. Through the analysis of the differences of the results, non-tungsten elements cross sections in the ENDF/B-VII.1 mainly affect the tungsten radiation capture and elastic scattering reaction rates in the energy range of 10-9-10-7 MeV, which results in a better simulated tungsten target reactivity worth value. Therefore, it is recommended that the tungsten target reactivity worth should be calculated with the ENDF/B-VII.1. [ABSTRACT FROM AUTHOR]

The measurements of physical parameters of the TRIGA reactor and Nuclear power plant Krško (NEK) reactor cores have been in the past performed on hand of the neutron flux signal obtained from uncompensated ionization cells and by employment of the a digital meter of reactivity (DMR). At the TRIGA reactor only one ionization cell is currently used for flux measurements. During the insertion of one control rod the neutron flux distribution is significantly altered affecting the flux measurements of inserting different control rods. The problem is presently solved by assigning a correction factor to each control rod what introduces an additional uncertainty. In the present paper the implementation of four fission cells for reactivity measurements is presented. In this way determining the correct gamma background and its subtraction, performed by DMR algorithms, becomes less important as previously by using ionization chambers. The larger number of detectors also reduces the flux redistribution effects on the signal during individual control rod movements. [ABSTRACT FROM AUTHOR]

The reactor operator is an inseparable part of reactor safety and his actions influence the course of the reactor transients. Understanding the operator's effect on reactor control thus contributes to the knowledge of the safe operation of nuclear reactors. Complementary to studies performed on simulators, or utilizing standard reactor or plant operational data, the dedicated experiments at research reactors can help to improve the insight into human aspects of reactor control. The potential of research reactors to study the human operator aspects of reactor control is broad; the related activities thus may also help to increase the utilization of research reactor facilities. The paper summarizes the considerations that had to be assessed during the implementation of such experiments at the VR-1 research reactor from the viewpoint of the reactor facility. Further, it recapitulates the activities that have been performed in this field at the VR-1 reactor in the collaboration of the Czech Technical University in Prague with the Defence Academy of the United Kingdom in the previous years. [ABSTRACT FROM AUTHOR]

The present article gives an overview of the first experimental campaigns carried out in the AKR-2 and CROCUS reactors within the framework of the Horizon 2020 European project CORTEX. CORTEX aims at developing innovative core monitoring techniques that allow detecting anomalies in nuclear reactors, e.g. excessive vibrations of core internals. The technique will be mainly based on using the fluctuations in neutron flux, i.e. noise analysis. The project will result in a deepened understanding of the physical processes involved. This will allow utilities to detect operational problems at a very early stage, and to take proper actions before such problems have any adverse effect on plant safety and reliability. The purpose of the experimental campaigns in the AKR-2 and CROCUS reactors is to produce noise-specific experimental data for the validation of the neutron noise computational models developed within this framework. The first campaigns at both facilities consisted in measurements at reference static states, and with the addition of mechanical perturbations. In the AKR-2 reactor, perturbations were induced by two devices: a rotating absorber and a vibrating absorber, both sets in experimental channels close to the core. In CROCUS, the project benefited from the COLIBRI experimental program: 18 periphery fuel rods were oscillated at a maximum of ±2 mm around their central position in the Hz range. The present article documents the experimental setups and measurements for each facility and perturbation type. [ABSTRACT FROM AUTHOR]

The present article describes the preliminary validation study of simulated in-core and reflector n eutron spectra in preparation of oncoming experimental programs in the zeropower reactor CROCUS at EPFL. For this purpose, a set of activation foils were irradiated at three characteristic positions in the CROCUS reactor, and the subsequent activities were analyzed via γ spectrometry. The experimental setup was then modeled with the Monte Carlo neutron transport code Serpent2 and associated with an analysis tool to include the effect of the reactor power history during experiments. The comparison of calculated and measured reaction rates (C/E) indicates a general consistency (at 2σ) between calculated and measured spectra. However, offsets of C/E values were observed in (n, γ) reactions, up to 18% for 115In and 8% for 63Cu dosimeters. This could be caused by an unexpected isotopic composition, uncertainties in nuclear data, or the spectrometry analysis. In addition, a 100-groups spectrum unfolding was performed using the experimentally determined reaction rates and the Serpent2 spectra as the prior knowledge. The unfolded spectra were mainly adjusted in the thermal and fast ranges, while few modifications w ere m ade i n t he e pithermal r egion d ue t o the low contribution of epithermal neutrons in activation processes. Moreover, within energy groups where the capture reactions show resonant behavior, flux depletion (up to 38% as compared to the prior spectra) is observed due to the absence of self-shielding effect in the unfolding process. For this purpose, an unfolding method based on energy groups weighting is developed and tested. [ABSTRACT FROM AUTHOR]

In the present article, we detail the method used to experimentally determine the power of the CROCUS zero-power reactor, and to subsequently calibrate its ex-core monitor fission chambers. Knowledge of the reactor power is a mandatory quantity for a safe operation. Furthermore, most experimental research programs rely on absolute fission rates in design and interpretation – for instance, tally normalization of reaction rate studies in dosimetry, or normalization of power spectral density in neutron noise measurements. The minimization of associated uncertainties is only achieved by an accurate power determination method. The main experiment consists in the irradiation, and therefore, the activation of several axially distributed Au-197 foils in the central axis of the core, which activities are measured with a High-Purity Germanium (HPGe) gamma spectrometer. The effective cross sections are determined by MCNP and Serpent Monte Carlo simulations. We quantify the reaction rate of each gold foil, and derive the corresponding fission rate in the reactor. The variance weighted average over the distributed foils then provides a calibration factor for the count rates measured in the fission chambers during the irradiation. We detail the calibration process with minimization of respective uncertainties arising from each sub-step, from power control after reactivity insertion, to the calibration of the HPGe gamma spectrometer. Biases arising from different nuclear data choices are also discussed. [ABSTRACT FROM AUTHOR]

Gamma rays in nuclear reactors, arising either from fission or decay processes, significantly contribute to the heating and dose of the reactor components. Zero power research reactors offer the possibility to measure gamma rays in a purely neutronic environment, allowing for validation experiments of computed spectra, dose estimates, reactor noise and prompt to delayed gamma ratios. This data then contributes to models, code validation and photo atomic nuclear data evaluation. In order to contribute to aforementioned experimental data, gamma detection capabilities are being added to the CROCUS reactor facility. The CROCUS reactor is a two-zone, uranium-fueled light water moderated facility operated by the Laboratory for Reactor Physics and Systems Behaviour (LRS) at the Swiss Federal Institute of Technology Lausanne (EPFL). With a maximum power of 100W, it is a zero power reactor used for teaching and research, most recently for intrinsic and induced neutron noise studies. For future gamma detection applications in the CROCUS reactor, an array of four detectors - two large 5"x10" Bismuth Germanate (BGO) and two smaller Cerium Bromide (CeBr3) scintillators - was acquired. The BGO detectors are to be arbitrarily positioned in the core reflector and out of the vessel for measurements at arbitrary distances. The CeBr3 detectors on the other hand are small enough to be set in the guide tubes of the control rods for in-core measurements. We present a study of the neutron and gamma flux in the core and reflector using the MCNP 6.2 and Serpent 2 Monte Carlo codes for coupled neutron and photon transport criticality calculations. More specifically, we investigate and compare predicted spectra as well as reactivity worth of different envisioned experimental setups. We further predict pulse height spectra as well as doses to the crystals with and without cadmium shielding to estimate allowable reactor powers with respect to detector radiation hardness. The results serve as basis for calibration and aid in the design and regulatory approval of the experiments. [ABSTRACT FROM AUTHOR]

This paper describes the neutron irradiation tests of 7 THERMOCOAX Self-Powered Neutron detectors (SPNDs) in the BR1 reactor at SCK•CEN. The SPNDs were fabricated according to the same specifications, but from different fabrication batches. The SPND signals were recorded during stepwise power-up of the reactor, proving the linearity of the SPND response within a wide thermal neutron flux range: from as low as 0.8·109 n/(cm2) to 2.6·1011 n/(cm2s). Intercomparison of the SPND signals shows a very small spread, confirming the repeatability of the THERMOCOAX fabrication processes. The overall neutron sensitivities of the seven SPNDs agreed within a 1% margin. The experimental data were analyzed in terms of prompt and various delayed responses. Prompt contributions to the signal are due to external gamma induced processes and to processes involving gamma rays emitted instantaneously upon neutron capture. The main contribution in a rhodium SPND is due to activation of the rhodium emitter and beta emission during decay of the activated rhodium and leads to a delayed response with a characteristic time of a few minutes. Activation and subsequent beta decay in other materials present in the SPND lead to additional minor delayed signal contributions. The partial SPND sensitivities due to all these processes were calculated using an MCNPX based model and were compared with experimental sensitivities based on the recorded data. The results were in fair agreement; for the overall SPND neutron sensitivity an agreement within a 1% margin was achieved. [ABSTRACT FROM AUTHOR]

The results of experiments performed with a Ø25 x 25 mm solution-grown stilbene crystal in mono-energetic neutron fields in the 80-to-230 keV energy range are presented. The goal of the measurements, performed at the AMANDE facility, was to explore the capabilities of this organic scintillator to measure neutrons at the lowest possible energy with good pulse shape discrimination (PSD). The time of flight (TOF) technique was used in order to help with the neutron-gamma discrimination. The data are collected via a programmable digital acquisition (DAQ) system CAEN DT7530 using the software CoMPASS with the charge comparison method (CCM). The data are analysed using post-processing codes developed in the ROOT environment. The results show that the stilbene detector has discrimination capabilities for energies as low as 80 keV. [ABSTRACT FROM AUTHOR]

The Advanced Gas Reactor-5/6/7 (AGR-5/6/7) experiment is the fourth and final experiment in the AGR experiment series and will serve as the formal fuel qualification test for the TRISO fuels under development by the U.S. Department of Energy. Certain locations in this experiment reach temperatures higher than any of the previous AGR tests, up to 1500°C. Such extreme temperatures create unique challenges for thermocouple-based temperature measurements. High-temperature platinum-rhodium thermocouples (Types S, R, and B)and tungsten-rhenium thermocouples (Type C) suffer rapiddecalibration due to transmutation of the thermoelements fromneutron absorption. For lower temperature applications, previousexperience with Type K thermocouples in nuclear reactors haveshown that they are affected by neutron irradiation only to alimited extent. Similarly, Type N thermocouples, which are morestable than Type K at high temperatures, are only slightly affectedby neutron fluence. Until recently, the use of these nickel-basedthermocouples was limited when the temperature exceeds 1050°Cdue to drift related to phenomena other than nuclear irradiation.Recognizing the limitations of existing thermometery to measuresuch high temperatures, the sponsor of the AGR-5/6/7 experimentsupported a development and testing program for thermocouplescapable of low drift operation at temperatures above 1100°C. High Temperature Irradiation Resistant Thermocouples (HTIR-TCs)based on molybdenum/niobium thermoelements have been underdevelopment at Idaho National Laboratory (INL) since circa 2004. A step change in accuracy and long-term stability of thisthermocouple type has been achieved as part of the AGR-5/6/7thermometry development program. Additionally, long-termtesting (9000+ hrs) at 1250°C of the Type N thermocouplesutilizing a customized sheath developed at the University ofCambridge has been completed with low drift results. Both theimproved HTIR and the Cambridge Type N thermocouple typeshave been incorporated into the AGR-5/6/7 test, which beganirradiation in February 2018 in INL's Advanced [ABSTRACT FROM AUTHOR]

This paper presents some results of the development of two new scintillation detectors usable for the detection and spectrometry of neutron and photon components of the mixed field. It is a series of liquid organic scintillators containing carbon, hydrogen, nitrogen and oxygen. Elements of individual scintillator designs are arranged in various chemical bonds defining physical and chemical properties of the scintillators. These scintillators have advantageous useful properties. They are non-hygroscopic, not poisonous and are not sensitive to contact with the outside atmosphere and can be stored in conventional laboratory containers. [ABSTRACT FROM AUTHOR]

In this article we present a first part of the results obtained during an irradiation campaign conducted at the Jozef Stefan Institute to observe the behaviour of piezoelectric materials under gamma and neutron flux. Specific instrumentation has been developed and has enabled the monitoring throughout the irradiation of several materials such as lead zirconate titanate (PZT) or modified Bismuth Titanate (BiT) in either massive or thick film form. Various parameters such as resonance frequency, electromechanical coupling coefficient, electrical capacitance, dielectric losses were measured as a function of the flow and dose received. The results obtained confirm that the samples work up to doses of 10 18 n°/cm2 and that the behaviour of the samples varies according to their composition and form. [ABSTRACT FROM AUTHOR]

The nuclear radiation energy deposition rate is a key value for the thermal design of experiments, on materials and nuclear fuels, carried out in experimental channels of nuclear research reactors. Studies are led for two kinds of sensor currently dedicated to quantifying this value and corresponding to calorimeter. Development of new sensors but also improvement of their calibration and their associated interpretation methods are necessary. These aims are possible by many ways such as numerical simulations of sensor, characterizations under laboratory conditions and experimental campaign under irradiation conditions. The calibration step under non-irradiation conditions represents a crucial phase. This phase requires the development of specific benches. The present paper focuses on a new thermal-transient bench and its use to perform calibration of a polish single-cell calorimeter. The new bench is detailed. First studies of the influence of external conditions (temperature, velocity) on the calorimeter sensitivity are presented and discussed. [ABSTRACT FROM AUTHOR]

Saclay CEA/IRFU is working for the delivery of five Non-Invasive Profile Monitors in the frame of the in-kind contribution agreement signed with the European Spallation Source. Neutrons will be produced by spallation reactions of 2 GeV proton beam impinging on a Tungsten target. To accelerate protons a powerful linear accelerator of 5MW is under construction. Diagnostic devices are mandatory tools for the tuning and protection of the machine. The non-invasive profile monitors provide a measurement of the beam profile in transverse directions to the beam propagation. This project raises several physical and technical challenges including low signal detection of ions or electrons, profile distortions induced by the beam Space Charge effect and non-uniformities of electric field. Simulation and model of the critical aspects of the detector have been performed in order to prove the performance and the feasibility of the detector. A series of prototypes has been built with different readout types, and tested in real conditions at the 3MeV proton accelerator IPHI. All of them show some advantages and drawbacks revealed by the tests in real beam conditions. In this paper we present the results of the tests for the various configuration readout systems to agree with the model and simulation of the detector. In concluding remarks, we will discuss the performance of the prototypes and point out the camerabased one to be the more suitable for the final design. [ABSTRACT FROM AUTHOR]

This paper concerns experimental and numerical works on a new differential calorimeter called CALORRE and validated recently under irradiation conditions in MARIA reactor at low nuclear absorbed dose rate level. Works focus on a specific configuration of CALORRE which was designed especially for the measurement of high nuclear energy deposition rates inside Material Testing Reactors. Due to the high level, a new calibration system was fabricated in order to determine the response of the new configuration under laboratory conditions for a very wide range of electrical power never applied in the literature. The response of the new configuration can be considered linear for this very wide range of electrical power. An analytical calculation shows the contribution of each heat transfer in specific zones. 3D thermal simulations performed by means of COMSOL Multiphysics under irradiation conditions give the predictions of the calorimeter behavior under real conditions (up to 20W.g−1). [ABSTRACT FROM AUTHOR]

Mechanical loading originating due to Pellet- Cladding Interaction (PCI) in PWR nuclear fuel rod constitutes a first order importance phenomenon when attempting at increasing the flexibility o f n uclear r eactors o peration t o f ollow grid demand. The need to improve the understanding of such complex multi-physics phenomena motivates the realization of particular irradiation sequences carried out in highly instrumented devices located in Material Testing Reactors. Among this kind, ADELINE irradiation device is being developed and will be installed in the future Jules Horowitz Reactor. In the perspective of designing the adapted measurement methodology to detect and characterize PCI phenomenology during irradiation, we present in this paper the experimental bench and its associated measurement program, designed to investigate eventual effects of PCI on the non-linear, flow i nduced v ibrations o fuclear fuel r od. A nalytical and numerical models cannot predictively describe the system due to the complexity of phenomena thus the IMPIGRITIA experimental set-up has been developed to reproduce the mechanical interaction between the pellet and the clad at low pressure, room temperature and out of neutron flux. T he d esigned test bench presents a clamped free single short rod, centred in the test section by mean of four centring elements. Different rod configurations are implemented and localized closure of the gap is remotely realized by means of a dilatation system. Laser Doppler Vibrometry is used to measure the transversal displacement of the sample rod in three different conditions: in air, in stagnant water and under turbulent axial flow r ate. T he experimental program and expected results are presented and discussed [ABSTRACT FROM AUTHOR]

Decay heat is the thermal power released by radioactive decays of unstable isotopes after the nuclear reactor shutdown, and delayed fission reactions. It constitutes a key parameter for the nuclear reactor safety and the nuclear fuel cycle; for this reason, design codes have to be qualified by comparison with experimental measurements. The CEA's package DARWIN2.3 has been qualified for the calculation of PWR decay heat with two integral measurements: the MERCI experience and the CLAB laboratory's experiments; performed respectively on the following cooling time intervals: 40 min – 40 days and 12 years – 25 years. A lack of validation in the first hour of cooling time requires to consider large margins on the calculated decay heat value. As a result, delays in core unloading, intervention of human operators and safety systems dimensioning may occur. The PRESTO experiment, under conception at CEA, deals with a decay heat measurement between 1 and 40 minutes of cooling time for a PWR fuel sample irradiated in the Jules Horowitz Reactor (JHR). A previous thermal study showed that measurements could be sensitive to the decay heat 1 minute after the beginning of the cooling time. Now, a more precise estimation of power sources was performed with the Monte-Carlo code TRIPOLI. In this framework, four different device configurations were considered. Our results show that the irradiation power is not enough elevated in configurations where a tungsten shield is present. [ABSTRACT FROM AUTHOR]

In order to determine the capacity of the cold neutron source refrigerator of HANARO, the nuclear heating rate at CN vertical hole is measured by using the heat-flow calorimetric method and confirmed by the calculation. The heating rate measurement device of HANARO was composed of a calorimeter sensor, an air containing aluminum sleeve for fitting the sensor to the CN hole, aluminum weight and a lead wire assembly. The calorimeter sensor consists of a cylindrical Al sample and container, two thermocouples and the electric heater for the calibration of the calorimeter. The sample is separated by an air gap from the Al container surrounded by an air containing Al sleeve. After installation of the calorimeter at a measurement position of HANARO, the heat transfer inside the calorimeter was simulated by the electric heating for the sample. The nuclear heating rates at the CN hole were determined at three reactor powers of 1, 4 and 8 MW by using the calibration curve and the temperature measurements at each reactor power. The measured nuclear heating rate per unit mass of Al sample at 8 MW reactor power is 0.143 W/g and it is equivalent to the 0.494 W/g at 30 MW. The nuclear heating rate was calculated by using the MCNP code. The calculation model for the whole facility including the reactor core and the reflector tank were established. In the calculation procedure, the heat generations by various radiations were evaluated with considering the prompt, delayed and activation effects. The measured heating rate was reasonably well supported by the calculation using the cold neutron facility design code. It will be very useful for the moderator cell of cold neutron source of HANARO. [ABSTRACT FROM AUTHOR]

A gamma thermometer suitable for very high gamma heating levels (up to 20 W/g) has been designed and modelled by means of detailed finite element calculations. Based on a sensitivity analysis, the predicted accuracy of this gamma thermometer is better than 5 %. A novel miniaturized gamma thermometer is proposed in which a single thermocouple is used as the gamma absorption element, allowing a reduction of the sensor diameter down to 3 mm. Monte Carlo calculations (by MCNP) have been performed to assess the relative contribution of neutrons to the nuclear heating in a gamma thermometer. Calculations have been performed for gamma thermometers with an inner body made of various materials, such as stainless steel, tungsten, molybdenum and rhodium. By using gamma thermometers made of different materials, it will be possible to deduce the nuclear heating rates in these materials and also to separate out the neutron and gamma heating contributions. The Monte Carlo calculations show that nuclear heating of rhodium is mainly due to neutrons, converting the rhodium gamma thermometer effectively in a neutron thermometer. The sensitivities of the gamma thermometers with W, Mo or Rh as heated materials have been modelled by finite element calculations. It is found that both the Mo and the Rh based sensor have a very linear response up to a nuclear heating of 20 W/g. [ABSTRACT FROM AUTHOR]