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At least 199 records · Page 11
All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

Fissile material detection using neutron time-correlations from photofission

The detection of special nuclear materials (SNM) in commercial cargoes is a major objective in the field of nuclear security. In this work we investigate the use of two-neutron time-correlations from photo-fission using the Prompt Neutrons from Photofission (PNPF) detectors in Passport Systems Inc.’s (PSI) Shielded Nuclear Alarm Resolution (SNAR) platform for the purpose of detecting ~5 kg quantities of fissionable materials in seconds. The goal of this effort was to extend the secondary scan mode of this system to differentiate fissile materials, such as highly enriched uranium, from fissionable materials, such as low enriched and depleted uranium (LEU and DU). Experiments were performed using a variety of material samples, and data were analyzed using the variance-over-mean technique referred to as Y 2F or Feynman-α. Results were compared to computational models to improve our ability to predict system performance for distinguishing fissile materials. Simulations were then combined with empirical formulas to generate receiver operating characteristics (ROC) curves for a variety of shielding scenarios. We show that a 10 second screening with a 200 μA 9 MeV X-ray beam is sufficient to differentiate kilogram quantities of HEU from DU in various shielding scenarios in a standard cargo container.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2016

Crystal growth of LiIn 1–x Ga x Se 2 crystals

Lithium containing chalcogenide single crystals have become very promising materials for photonics and radiation detection. Detection applications include nuclear nonproliferation, neutron science, and stellar investigations for the search of life. Synthesis and single crystal growth methods for lithium containing chalcogenide, specifically LiIn 1-x Ga x Se 2 , single crystals are discussed. This study elucidates the possibility of improving neutron detection by reducing the indium capture contribution; with the incorporation of the lithium-6 isotope, gallium substitution may overcome the neutron detection efficiency limitation of 6 LiInSe 2 due to appreciable neutron capture by the indium-115 isotope. As a figure of merit, the ternary parent compounds 6 LiInSe 2 and 6 LiGaSe 2 were included in this study. Quality crystals can be obtained utilizing the vertical Bridgman method to produce quaternary compounds with tunable optical properties. Here, quaternary crystals of varying quality depending on the gallium concentration, approximately 5 x 5 x 2 mm 3 or larger in volume, were harvested, analyzed and revealed tunable absorption characteristics between 2.8-3.4 eV.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Boron-Based Neutron Scintillator Screens for Neutron Imaging

In digital neutron imaging, the neutron scintillator screen is a limiting factor of spatial resolution and neutron capture efficiency and must be improved to enhance the capabilities of digital neutron imaging systems. Commonly used neutron scintillators are based on 6 LiF and gadolinium oxysulfide neutron converters. This work explores boron-based neutron scintillators because 10 B has a neutron absorption cross-section four times greater than 6Li, less energetic daughter products than Gd and 6 Li, and lower γ-ray sensitivity than Gd. These factors all suggest that, although borated neutron scintillators may not produce as much light as 6 Li-based screens, they may offer improved neutron statistics and spatial resolution. This work conducts a parametric study to determine the effects of various boron neutron converters, scintillator and converter particle sizes, converter-to-scintillator mix ratio, substrate materials, and sensor construction on image quality. The best performing boron-based scintillator screens demonstrated an improvement in neutron detection efficiency when compared with a common 6 LiF/ZnS scintillator, with a 125% increase in thermal neutron detection efficiency and 67% increase in epithermal neutron detection efficiency. The spatial resolution of high-resolution borated scintillators was measured, and the neutron tomography of a test object was successfully performed using some of the boron-based screens that exhibited the highest spatial resolution. For some applications, boron-based scintillators can be utilized to increase the performance of a digital neutron imaging system by reducing acquisition times and improving neutron statistics.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2022

Review of Direct Neutron Conversion and Detection Processes

Commercial and laboratory neutron detection systems use indirect neutron response of materials like pressurized helium-3 (via nuclear reaction 3He(n, p)3H) to measure and count neutrons emanating from a source. Recently a host of semiconductors, especially a ternary semiconductor of lithium indium diselenide (6LiInSe2) and a quaternary alloy of enriched lithium-6, indium, phosphorous, and selenium (6LiInP2Se6), have shown promising neutron counting possibilities by directly converting neutrons into charge-carrying elements within the body of the semiconductor. These semiconductors have high thermal neutron capture cross sections, suitable energy bandgaps (~2.0 electron volts) for room-temperature operations, and a favorable electronic band structure for efficient electron charge transport. The article examines the semiconductor properties of these compounds in terms of their neutron counting capabilities and possible ways to extract neutron energy information from them. Lithium-6 and boron-10 (with thermal neutron absorption cross sections of 938 ± 6 and 3855 ± 26 barns, respectively) produce charged particles to be measured via indirect neutron interactions. The efficiency of indirect conversion neutron detectors is limited because of the inefficiencies in conversion mechanism. In case of direct conversion, the neutrons create charged particles in a single material for neutron capture and charge collection, increasing detection efficiency. Unlike 3He proportional counters, which provide no neutron energy information, the semiconductors can be used as neutron energy spectrometer. Fully resolved neutron energy by 6LiInP2Se6 from a plutonium-beryllium source has been reported in the literature. We will discuss the influence of these multilayered semiconductors’ crystallographic structures and growth techniques on neutron energy determination.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2021

Investigations of radiation damaged arranged scintillating particle composites

Composites consisting of periodically arranged scintillating particles within an acrylic (poly methyl methacrylate) matrix are excellent templates for the gamma-insensitive detection of fast neutrons. In order to study the impact of radiation damage on acrylic-based scintillating particle composites, we exposed acrylic samples to different doses of fast neutrons up to 119.7 kGy and investigated the change in optical characteristics using optical absorption spectroscopy in the 200–800 nm wavelength range. The experimental results are compared to coupled MCNP6 and optical ray-tracing (FRED) simulations, and qualitative agreement is found. Neutron-induced radiation damage in the acrylic matrix manifests as a red-shift in the ultra-violet absorption edge of the acrylic and a corresponding decrease in the light propagation efficiency, which leads to detector performance degradation. We show for the first time that the effects of radiation damage can be mitigated by the use of wavelength-shifting coatings. We predict that composites with wavelength-shifting coatings can enable neutron detectors with high tolerance against neutron-induced radiation damage, a property that is particularly desired for neutron detection in high radiation environments.

36 MATERIALS SCIENCE↗
DOE OSTI2013

Annual Report FY2013-- A Kinematically Complete, Interdisciplinary, and Co-Institutional Measurement of the 19F(α,n) Cross-section for Nuclear Safeguards Science

The goal of this proposal is to enable neutron detection for precision Non-Destructive Assays (NDAs) of actinide-fluoride samples. Neutrons are continuously generated from a UFx matrix in a container or sample as a result of the interaction of alpha particles from uranium-decay α particles with fluorine nuclei in the matrix. Neutrons from 19F(α,n)22Na were once considered a poorly characterized background for assays of UFx samples via 238U spontaneous fission neutron detection [SMI2010B]. However, the yield of decay-α-driven neutrons is critical for 234,235U LEU and HEU assays, as it can used to determine both the total amount of uranium and the enrichment [BER2010]. This approach can be extremely valuable in a variety of safeguard applications, such as cylinder monitoring in underground uranium storage facilities, nuclear criticality safety studies, nuclear materials accounting, and other nonproliferation applications. The success of neutron-based assays critically depends on an accurate knowledge of the cross section of the (α,n) reaction that generates the neutrons. The 40% uncertainty in the 19F(α,n)22Na cross section currently limits the precision of such assays, and has been identified as a key factor in preventing accurate enrichment determinations [CRO2003]. The need for higher quality cross section data for (α,n) reactions has been a recurring conclusion in reviews of the nuclear data needs to support safeguards. The overarching goal of this project is to enable neutron detection to be used for precision Non- Destructive Assays (NDAs) of actinide-fluoride samples. This will significantly advance safeguards verification at existing declared facilities, nuclear materials accounting, process control, nuclear criticality safety monitoring, and a variety of other nonproliferation applications. To reach this goal, Idaho National Laboratory (INL), in partnership with Oak Ridge National Laboratory (ORNL), Rutgers University (RU), and the University of Notre Dame (UND), will focus on three specific items: (1) making a precision (better than 10 %) determination of the absolute cross section of the 19F(α,n)22Na reaction as a function of energy; (2) determining the spectrum of neutrons and γ-rays emitted from 19F(α,n)22Na over an energy range pertinent to NDA; and (3) performing simulations with this new cross section to extract the neutron yield (neutrons/gram/second) and resulting neutron- and gamma ray-spectra when α particles interact with fluorine nuclei in actinide samples, to aid in the design and reduce uncertainty of future NDA measurements and simulations.

VANDLE, HPGe gamma-ray detector, liquid neutron de↗
DOE OSTI2018

A high-efficiency, low-Ĉerenkov Micro-Layered Fast-Neutron Detector for the $\mathrm{TREAT}$ hodoscope

Here this paper presents an enabling technology for high-efficiency fast-neutron detection for the Transient Reactor Test (TREAT) Facility hodoscope. The hodoscope is currently outfitted with 25.4-mm long Hornyak buttons for fast-neutron detection. However, they require improvements, as they suffer from low detection efficiency, Ĉerenkov radiation contaminating the signal, and non-linearity of detector response at large transients. To address these issues, a layered ZnS:Ag device has been constructed. The layered design provides various advantages: it incorporates greater scintillation volume, Ĉerenkov background is significantly reduced by eliminating the need for additional light guides, and the detection process is simplified by achieving gamma-ray rejection using only pulse-height discrimination. The result is a greater number of fast neutron counts with reduced background noise contaminating the signal. On being irradiated with a 252Cf source (whose neutron energy spectrum closely mimics a reactor fission spectrum), a 25.4-mm long layered device was found to exhibit an intrinsic fast-neutron detection efficiency of 8.1% for neutrons with an average energy of 2 MeV, while a 43-mm long device allows for a greater number of fast neutron to be detected, increasing the efficiency to 9.1%. Further lengthening of the device up to 20 cm has been numerically predicted to exhibit even higher efficiencies. The detectors can be mass-produced easily and inexpensively.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2015

Performance of an RPM based on Gd-lined plastic scintillator for neutron and gamma detection [ANIMMA--2015-IO-372]

A Radiation Portal Monitor (RPM) was developed by the Istituto Nazionale di Fisica Nucleare (INFN) and Ansaldo Nucleare (ANN) within the FP7 SCINTILLA European project. The system was designed to detect both gamma and neutron radiation with a single technology. It is conceived to monitor vehicle and cargo containers in transits across borders or ports, to find radioactive elements and to avoid illegal trafficking of strategic nuclear materials. The system is based on a {sup 3}He-free neutron detection technology using plastic scintillators coupled to Gadolinium to detect and discriminate gamma from neutron signals. During the 3 years of the SCINTILLA project the construction and test of the first two prototypes drove the definition of the final layout of a full RPM system consisting of two twin pillars as a portal for vehicle and cargo container scan. A custom System Control Software (SCS) manages the electronics of the RPM, the ancillary devices and the data analysis. The combination of the detector layout and of the software functionalities enables both to distinguish neutrons and gammas and to identify the energy range of a detected gamma source. The system was initially characterized via static tests with gamma and neutron sources in the INFN laboratory. These measurements were used to calibrate the detector, evaluate the response of the single pillars as well as of the full system, and optimize the RPM configuration and discrimination algorithm. During this phase, specific tests were performed to study the stability over time of the system, monitoring the measured the neutron and gamma count rates over periods of several weeks. The results allow us to demonstrate the reliability and robustness of the RPM. In a second time the RPM performance was studied via dynamic tests performed during the SCINTILLA test and benchmark campaigns. These measurements took place in the JRC ITRAP+10 facility at Ispra (Varese-Italy). The laboratory is equipped with an experimental set-up for dynamic tests of multiple systems according to international standards. The performed measurements utilized radioactive sources with activities selected according to ANSI and IEC standards to test the detector alarm performances in terms of gamma and neutron response, sensitivity to high gamma fields, sensitivity to moderated neutron sources as well as false alarm rates (FAR). In addition, the RPM was tested in challenging configurations exceeding the requirements set by international standards to determine the real limits of the system. The results obtained during these campaigns demonstrated that the system detection efficiency is not only compliant to international standards for its category, but often exceeds them, demonstrating the validity of the chosen technology and of the implemented layout. The positive performance also showed the effectiveness of the SCS and of its functionalities. To further demonstrate the system capabilities, a test in a real-life environment of the RPM is planned to happen in a near future by installing the detectors in a seaport. In this presentation I will give an overview of the RPM characteristics, of its performances as determined in the test campaign mentioned above and of future plans, to demonstrate how this technology can be an effective choice for the realization of {sup 3}He-free RPM detectors. (authors)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2013

Multiple source associated particle imaging for simultaneous capture of multiple projections

Disclosed herein are representative embodiments of methods, apparatus, and systems for performing neutron radiography. For example, in one exemplary method, an object is interrogated with a plurality of neutrons. The plurality of neutrons includes a first portion of neutrons generated from a first neutron source and a second portion of neutrons generated from a second neutron source. Further, at least some of the first portion and the second portion are generated during a same time period. In the exemplary method, one or more neutrons from the first portion and one or more neutrons from the second portion are detected, and an image of the object is generated based at least in part on the detected neutrons from the first portion and the detected neutrons from the second portion.

47 OTHER INSTRUMENTATION↗
DOE OSTI2026

Scintillator Manufacture

This paper will be a literature review of composite scintillators, and their applications in neutron detection. Due to the shortage of 3 He, the traditional medium for neutron detection, new materials need to be explored for this purpose. Composite scintillating particle materials can serve as a solution to replace 3 He, offering faster response times, and all-solid-state form factors for diversified applications in the field. This project will include manufacturing Europium (Eu) and Sodium (Na) doped LiCaAlF 6 (LiCAF) scintillating particles for neutron detection.

36 MATERIALS SCIENCE↗
DOE OSTI2012

Alternatives to Helium-3 for Neutron Multiplicity Counters

Multiplicity counters are an example of a neutron detection system affected by the shortage of 3He. Efforts are underway to identify potential 3He neutron detector replacements for use in multiplicity counters. Boron-10 and 6Li based systems are two of the options being explored as near-term 3He alternatives for neutron multiplicity counters. Simulations of BF3, 10B-lined proportional counters and 6Li/ZnS(Ag) sheets in various configurations have been performed with the Monte Carlo particle transport code MCNPX and compared to simulations of existing 3He counters. The system performances are being compared using figure of merits that are the square or cube of the total system efficiency divided by the die-away time. Design considerations include developing a system with enough neutron detection material to achieve the highest possible efficiency, while simultaneously minimizing system size. Adequate moderation is required to thermalize the incident neutrons for increased counting efficiency but as the system size increases so will the die-away time (average lifetime of neutrons in the counter). The optimal moderator configuration is one for which the increase in neutron detection efficiency is not defeated by an increase in die-away time. Thus, the entire system performance must be evaluated with every configuration change. The simulation results will be validated against a bench-top demonstrator unit design based on the system identified through simulations as having the highest performance potential. Presented here are the simulation results with various configurations of BF3 filled proportional counters, 10B lined proportional counters and 6Li/ZnS(Ag) sheets, and preliminary bench-top measurements.

Alternatives to Helium-3 for Neutron Multiplicity ↗
DOE OSTI2012

ALTERNATIVES TO HELIUM-3 FOR NEUTRON MULTIPLICITY DETECTORS

Collaboration between the Pacific Northwest National Laboratory (PNNL) and the Los Alamos National Laboratory (LANL) is underway to evaluate neutron detection technologies that might replace the high-pressure helium (3He) tubes currently used in neutron multiplicity counter for safeguards applications. The current stockpile of 3He is diminishing and alternatives are needed for a variety of neutron detection applications including multiplicity counters. The first phase of this investigation uses a series of Monte Carlo calculations to simulate the performance of an existing neutron multiplicity counter configuration by replacing the 3He tubes in a model for that counter with candidate alternative technologies. These alternative technologies are initially placed in approximately the same configuration as the 3He tubes to establish a reference level of performance against the 3He-based system. After these reference-level results are established, the configurations of the alternative models will be further modified for performance optimization. The 3He model for these simulations is the one used by LANL to develop and benchmark the Epithermal Neutron Multiplicity Counter (ENMC) detector, as documented by H.O. Menlove, et al. in the 2004 LANL report LA-14088. The alternative technologies being evaluated are the boron-tri-fluoride-filled proportional tubes, boron-lined tubes, and lithium coated materials previously tested as possible replacements in portal monitor screening applications, as documented by R.T. Kouzes, et al. in the 2010 PNNL report PNNL-72544 and NIM A 623 (2010) 1035–1045. The models and methods used for these comparative calculations will be described and preliminary results shown

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Device and method for the location and identification of a radiation source

A handheld device for the location and identification of a radiation source, including: a radiation transparent housing; a radiation locator device disposed within the radiation transparent housing operable for determining the location of the radiation source, wherein the radiation locator device includes a plurality of gamma detection crystals arranged in a geometric pattern and separated by a gamma shielding material, a plurality of detectors coupled to the plurality of gamma detection crystals, and a processor module coupled to the plurality of detectors; one or more of a neutron detection crystal and a gamma spectroscopy crystal disposed within the radiation transparent housing adjacent to the radiation locator device; and one or more detectors coupled to the one or more of the neutron detection crystal and the gamma spectroscopy crystal and the processor module; wherein the one or more of the neutron detection crystal and the gamma spectroscopy crystal, the one or more detectors, and the processor module are collectively operable for identifying the radiation source.

Vencelj, Matjaz↗
DOE OSTI2022

High-fidelity photoneutron detection via neutron activation analysis

To interdict illicit radioactive sources and prevent nuclear terror, governments have deployed portal monitor systems at ports of entry. Such portal monitors rely upon the detection of spontaneously-emitted radiation signatures, however, detection scenarios become more challenging for highly enriched uranium, which emits relatively few spontaneous radiation signatures. Photon active interrogation techniques can augment current cargo inspection scenarios and improve detection capabilities for illicit nuclear material. A high-energy photon interrogation source can induce photofission in nuclear material, producing strong radiation signatures for detection. Prompt fission signals are of specific interest because prompt neutrons are greater than 99% of the fission signal. When using detectors capable of spectroscopy, e.g., pulse shape discrimination-capable organic scintillators, the intense interrogating source often will cause pulse pile-up, degrading particle classification. In this work, we demonstrate photoneutron detection via neutron activation of aluminum and iron. The experiments rely upon nuclear reactions that can only be induced by neutrons, and are robust against pile-up, thereby providing high-confidence neutron rates. In this work, we apply neutron activation analysis for the detection of high-energy neutrons produced by high-energy photon irradiation; the depleted uranium sample was irradiated with a 9-MeV bremsstrahlung beam. We find that the irradiated depleted uranium produced an aluminum activation flux of 1.77 × 10 5 n cm –2 s –1 and an iron-56 activation flux of 1.05 × 10 5 n cm –2 s –1 . Furthermore, the ability to detect radioactivity induced by neutrons created in (, n) and photofission reactions demonstrates a new approach for photon active interrogation to identify nuclear materials.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2015

Design and testing of a combined neutron and gamma assay system for measuring fissile material in fuel and waste items - 15340

A combined neutron and gamma ray assay system for the measurement of fissile material in both fuel and waste items has been designed, modelled, developed and tested. The purpose of the system is to measure and characterize a wide range of fissile materials including uranium, plutonium, americium and mixed oxide fuel arising from a variety of processes and facilities across the United Kingdom. Dounreay Site Restoration Limited at Dounreay currently stores these items form the United Kingdom Nuclear Decommissioning Authority's un-irradiated fuel inventory. This inventory comprises historic materials from ex-United Kingdom Atomic Energy Authority nuclear fuel projects and experiments together with fast reactor fuel materials and some commercial fuel items as delivered to Dounreay for recovery. The combined assay system has been designed and constructed to support the operation of a fuel processing facility, which will be operated at Dounreay to treat these historic legacy fuel materials, making them suitable for long term storage. Appropriate characterization of the fissile material is required for safety, criticality control and nuclear materials safeguards purposes prior to final disposition. The system consists of both gamma ray and neutron measurement components. The results of a neutron measurement are combined with the gamma ray isotopic ratio measurement results to generate the assay result for the fissile material being characterized. The gamma ray component of the system employs a safeguards quality high purity Germanium detector with electro-mechanical cooling and the measurements of fissile material containers are made in a shielded chamber with a tin and copper graded lining. Containers of fissile material are rotated during the gamma ray measurement and a variable automatic steel collimator aperture mechanism is employed to adjust the measured count rate to control detector dead time. Gamma ray spectra are analysed using the plutonium and uranium isotopic ratio code PC/FRAM. The gamma ray system can also be configured to perform a direct assay of small quantities of uranium bearing material. For this assay measurement a series of cadmium filters can be positioned in front of the detector to reduce the contribution from 59 keV gamma rays from Am-241. In addition to gamma ray isotopic measurements, both active and passive neutron assay is performed by the system. The neutron-measuring component consists of a thick walled cylindrical polyethylene chamber, which provides neutron moderation for 30 one-inch diameter He-3 detector tubes located in re-entrant channels in the polyethylene moderator. Two removable polyethylene-inserts incorporating lower plug assemblies are provided, one with and one without a cadmium lining. Also two top plug units are provided, again one with cadmium and one without a cadmium liner. The Neutron Subsystem can be operated both as a passive and as an active neutron coincidence counter. In both fast and thermal active neutron mode, the neutron component employs two americium-beryllium (Am-Be) alpha-n neutron sources, one positioned in the top and the other in the bottom plug unit. The neutron sources are removed when the counter is operated in passive mode. Passive neutron detection efficiency and active neutron performance has been modelled using the MCNP Monte Carlo computer simulation code as part of the design process. In the absence of fissile samples, operation of the gamma ray component of the system has been tested initially using calibrated Eu-152 and Ba-133 gamma ray sources and the use of archived spectra from fissile material measurements. Initial testing of the neutron system has employed a calibrated Cf-252 spontaneous fission neutron source. Neutron testing has been confirmed using further MCNP simulations. The results of test measurements with the system are presented and compared with the results of MCNP measurement simulations. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2014

Evaluation of two-stage system for neutron measurement aiming at increase in count rate at Japan Atomic Energy Agency-Fusion Neutronics Source

In order to increase the count rate capability of a neutron detection system as a whole, we propose a multi-stage neutron detection system. Experiments to test the effectiveness of this concept were carried out on Fusion Neutronics Source. Comparing four configurations of alignment, it was found that the influence of an anterior stage on a posterior stage was negligible for the pulse height distribution. The two-stage system using 25 mm thickness scintillator was about 1.65 times the count rate capability of a single detector system for d-D neutrons and was about 1.8 times the count rate capability for d-T neutrons. The results suggested that the concept of a multi-stage detection system will work in practice.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2015

Excess wing in glass-forming glycerol and LiCl-glycerol mixtures detected by neutron scattering

The relaxational dynamics in glass-forming glycerol and glycerol mixed with LiCl is investigated using different neutron scattering techniques. The performed neutron spin echo experiments, which extend up to relatively long relaxation time scales of the order of 10 ns, should allow for the detection of contributions from the so-called excess wing. This phenomenon, whose microscopic origin is controversially discussed, arises in a variety of glass formers and, until now, was almost exclusively investigated by dielectric spectroscopy and light scattering. In conclusion, we show here that the relaxational process causing the excess wing can also be detected by neutron scattering, which directly couples to density fluctuations.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2015

Education and training in the field of nuclear instrumentation and measurement: CEA/INSTN (National Institute for Nuclear Sciences and Technologies) strategy to improve and develop new pedagogical tools and methods

Part of the French Alternative Energies and Atomic Energy Commission (CEA), the National Institute for Nuclear Science and Technology (INSTN) is a higher education institution whose mission is to provide students and professionals a high level of scientific and technological qualification in all disciplines related to nuclear energy applications. In this frame, INSTN carries out education and training (E and T) programs in nuclear instrumentation and radioprotection. Its strategy has always been to complete theoretical courses by training courses and laboratory works carried out on an extensive range of training tools that includes a large panel of nuclear instrumentation as well as software applications. Since its creation in 1956, the INSTN has conducted both education and vocational programs on ionizing radiation detection. An extensive range of techniques have commonly been used during practical works with students and employees of companies who need to get the knowledge and specialization in this field. Today, the INSTN is mainly equipped with usual detectors and electronics in large numbers in order to be able to accommodate up to 48 trainees at the same time in two classrooms, with only two trainees for one workstation in order to optimize their learning. In the field of the neutron detection systems, the INSTN has strongly developed its offer taking advantage of the use of research reactors, such as ISIS reactor (700 kW) at Saclay. The implementation of neutron detection systems specific to the courses offers a unique way of observing and analysing the signal coming from neutron detectors, as well as learning how to set the parameters of the detection system in real conditions. Providing the trainees with an extensive overview of each part of the neutron monitoring instrumentation apply to a nuclear reactor, hands-on measurements on the ISIS reactor play a major role in ensuring a practical and comprehensive understanding of the neutron detection system and their integration in the safety system of nuclear reactors. It also gives a solid background for the follow up and the development of the neutron detection systems. Another field of activity of the INSTN is the development of new teaching tools using software applications. We describe hereafter two applications that have recently been developed by the institute, O.S.I.R.I.S and DOSIMEX. The INSTN and the company OREKA have developed an innovative teaching tool named O.S.I.R.I.S. The tool is built on a virtual 3D environment in which users operate in a totally free way. The action is located in a pressurised water reactor in the environment of a steam generator building. The users are students or professionals who want to learn and train on concrete situations on how to protect the workers against radiation. Through the use of this serious game, users must: establish a predictive dose evaluation, implement the principles of radiation protection, perform the dose result of the operation, analyse the gap between the predictive collective dose and the collective dose achieved, and decide about the lessons to provide feedback. Taking into account the increasing constraints related to radiation protection, it is essential to use computer codes to evaluate the dose rates generated by sources of ionizing radiation. Many powerful codes exist but are often relatively complex to implement and are dedicated to experts. In addition, these codes are often 'black boxes' that does not allow the user to understand the underlying physics. The objective of DOSIMEX codes developed by the INSTN is to give, by the means of a VBA code with a simple interface, the ability to easily conduct a wide range of radiological exposure situations: calculation of gamma, beta, neutron equivalent dose rate, calculation of internal contamination and atmospheric transfer.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗