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DOE OSTI2023

P2M Simulation Exercise on Past Fuel Melting Irradiation Experiments

This paper presents the results of the Power To Melt and Maneuverability (P2M) Simulation Exercise on past fuel melting irradiation experiments, organized within the Organisation for Economic Co-operation and Development/Nuclear Energy Agency Framework for IrraDiation ExperimentS (FIDES) framework by the Core Group (CEA, EDF, and SCK·CEN) and open to all FIDES members. The exercise consisted in simulating two past power ramps where fuel melting was detected: (1) the xM3 staircase power transient [ramp terminal level (RTL) 70 kW·m -1 , average burnup 27 GWd·tU -1 ], carried out in 2005 in the R2 reactor at Studsvik (Sweden), where the rodlet maintained its integrity, and (2) the HBC4 fast power transient (RTL 66 kW·m -1 , average burnup 48 GWd·tU -1 ), carried out in 1987 in the BR2 reactor at SCK·CEN (Belgium), where the cladding failed during the experiment. The exercise was joined by 13 organizations from 9 countries using 11 different fuel performance codes. In this paper, the main results of the Simulation Exercise are presented and compared to available postirradiation examinations (PIE) or on-line measurements during the power ramps (fuel and clad diameters, rod elongation, pellet-clad gap, and fission gas release). Since the focus of the Simulation Exercise is on fuel melting assessment, determination of the boundary between melted/nonmelted fuel and the consequent definition of a melting radius from PIE are first discussed. During the HBC4 ramp, fuel melting was predicted by most of the codes despite differences in the melting models. Higher discrepancies were observed for the xM3 rod that can be attributed partly to power uncertainty and partly to the limited capability of the models to describe partial melting of the fuel during this ramp. Finally, possible code developments to improve simulation results are presented.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Physics and technology in the ion-cyclotron range of frequency on Tore Supra and TITAN test facility: implication for ITER

To support the design of an ITER ion-cyclotron range of frequency heating (ICRH) system and to mitigate risks of operation in ITER, CEA has initiated an ambitious Research & Development program accompanied by experiments on Tore Supra or test-bed facility together with a significant modelling effort. The paper summarizes the recent results in the following areas: Comprehensive characterization (experiments and modelling) of a new Faraday screen concept tested on the Tore Supra antenna. A new model is developed for calculating the ICRH sheath rectification at the antenna vicinity. The model is applied to calculate the local heat flux on Tore Supra and ITER ICRH antennas. Full-wave modelling of ITER ICRH heating and current drive scenarios with the EVE code. With 20 MW of power, a current of 400 kA could be driven on axis in the DT scenario. Comparison between DT and DT(3He) scenario is given for heating and current drive efficiencies. First operation of CW test-bed facility, TITAN, designed for ITER ICRH components testing and could host up to a quarter of an ITER antenna. R&D of high permittivity materials to improve load of test facilities to better simulate ITER plasma antenna loading conditions.

Litaudon, X↗
DOE OSTI2011

Prompt fission neutron spectra from fission induced by 1 to 8 MeV neutrons on {sup 235}U and {sup 239}Pu using the double time-of-flight technique

Prompt fission neutron spectra from {sup 235}U and {sup 239}Pu were measured for incident neutron energies from 1 to 200 MeV at the Weapons Neutron Research facility (WNR) of the Los Alamos Neutron Science Center, and the experimental data were analyzed with the Los Alamos model for the incident neutron energies of 1-8 MeV. A CEA multiple-foil fission chamber containing deposits of 100 mg {sup 235}U and 90 mg {sup 239}Pu detected fission events. Outgoing neutrons were detected by the Fast Neutron-Induced {gamma}-Ray Observer array of 20 liquid organic scintillators. A double time-of-flight technique was used to deduce the neutron incident energies from the spallation target and the outgoing energies from the fission chamber. These data were used for testing the Los Alamos model, and the total kinetic energy parameters were optimized to obtain a best fit to the data. The prompt fission neutron spectra were also compared with the Evaluated Nuclear Data File (ENDF/B-VII.0). We calculate average energies from both experimental and calculated fission neutron spectra.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2011

Fibre optic extensometer for high radiation and high temperature nuclear applications

In the framework of the Joint Instrumentation Laboratory (LCI), gathering resources from SCK-CEN (Belgium) and CEA (France), we are developing an optical sensor in order to accurately measure radiation-induced elongation of material placed in the core of a Material Testing Reactor (MTR). This extensometer displays common advantages of Fibre Optic (FO) sensors: high resolution, easy remote sensing and multiplexing, and also compact size which is of particular interest for in pile experiments with little room available. In addition, light weight reduces gamma heating hence limiting the thermal effect. In accordance with the specifications, the sensor has preferably two fixing points defining a gauge length of 10 to 15 mm (as a minimum). The diameter is less than 2 mm. Intense gamma and neutron irradiation as well as high temperatures are the most difficult environment conditions to withstand. Reactor radiation produces huge losses in common optical fibre. The losses can be limited by selecting the fibres (mainly with high purity silica core), the wavelength range (800-1200 nm), and a measurement based on interferometry (largely insensitive to losses in the fibre thanks to the wavelength encoding of the useful signal). Heavy neutron - mainly - and gamma flux such as in MTR, also produce compaction of silica, resulting in a significant drift and preventing the use of commercial FO sensors in such environment. Knowing this issue we revised the basic scheme of Extrinsic Fabry Perot Interferometer (EFPI) in order to limit the effects of compaction. A first sensor prototype fixed on a stainless steel support was tested in the Smirnof test facility in the BR2 MTR in Mol (Belgium). The support was subject to a constant mechanical and thermal stress, and his dimensions were not supposed to vary. This test showed a very low drift of the revisited EPFI design under high irradiation field in comparison with a commercial EFPI. This result has to be confirmed with second generation sensors with an increased robustness. The other difficulty to face is high temperature. Fibre optics with metal coating allows safe operation under temperatures up to 400 deg.C and even higher. But differential dilatation between silica and typical metallic material produces differential elongation in the range of 0.5 10{sup -2} i.e 5000{mu}{epsilon} for an increase in temperature of 300-400 deg. C. Such large elongation has to be considered carefully in the sensor design and its fixing on the sample. Laboratory tests with a second generation of adapted sensors demonstrated their good accuracy: less than 1% error over a displacement of 100{mu}m (< 1 {mu}m error), at room temperature. The displacement measured can reach 250{mu}m (20000{mu}{epsilon} with a gauge length of 12.5 mm) or more. We are currently implementing metal-coated fibre and we are preparing the next in pile irradiation on the BR2 reactor scheduled for mid-2011. Other applications of the sensor can be considered. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2011

The key-role of instrumentation for the new generation of research reactors

Experimental reactors have been indispensable since the beginning of the use of nuclear energy to support many important fields of industry and research: safety, lifetime management and operation optimisation of nuclear power plants, development of new types of reactors with improved resources and fuel cycle management, medical applications, material development for fusion... Over the last decade, modifications of the operational needs and the ageing of the nuclear facilities have led to several closures and time is coming for new key European Experimental Reactors (EER) within a European and International Framework. Projects like MYRRHA and JHR are underway to define and implement a new consistent EER policy: - Meeting industry and public needs, keeping a high level of scientific expertise; - With a limited number of EER, specified within a rational compromise between specialisation, complementarities and back-up capacities; - To be put into effective operation in this or the next decade. These new projects will give to the scientific community high performances allowing innovative fields of R and D. A new generation of instrumentation to address new phenomena and that allows better on-line investigation of some key physical parameters is necessary to achieve these challenges. One initiative to progress in this direction is the Joint Instrumentation Laboratory between CEA and SCK.CEN which has already given significant results and patents. Major scientific challenges to achieve in the field of instrumentation for this new generation of European Research Reactors have to be investigated and are described in this paper as well as a short description of the JHR and MYRRHA reactors that will be serving as flexible irradiation facilities for testing them. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2011

Irradiation tests in BR2 of miniature fission chambers in pulse, Campbelling and current mode

The FNDS system ('Fast Neutron Detection System') for the on-line in-pile detection of the fast neutron flux in the presence of a significant thermal neutron flux and a high gamma dose rate is being developed in the framework of the SCK.CEN-CEA Laboratoire Commun. The system has been patented in 2008. The system consists of a miniature Pu-242 fission chamber as main detector, complemented by a U-235 fission chamber or a rhodium Self-Powered Neutron Detector (SPND) for thermal neutron flux monitoring and a dedicated acquisition system that also takes care of the processing of the signals from both detectors to extract fast neutron flux data. This paper describes a FNDS qualification experiment in the SCK.CEN BR2 reactor, with experimental results on a large set of fission chambers in current and Campbelling mode. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2011

Combined analysis of neutron and photon flux measurements for the Jules Horowitz reactor core mapping

We study the combined analysis of nuclear measurements to improve the knowledge of the irradiation conditions in the experimental locations of the future Jules Horowitz Reactor (JHR). The goal of the present work is to measure more accurately neutron flux, photon flux and nuclear heating in the reactor. In a Material Testing Reactor (MTR), nuclear heating is a crucial parameter to design the experimental devices to be irradiated in harsh nuclear conditions. This parameter drives the temperature of the devices and of the samples. The numerical codes can predict this parameter but in-situ measurements are necessary to reach the expected accuracy. For this reason, one objective of the IN-CORE program [1] is to study the combined measurements of neutron and photon flux and their cross advanced interpretation. It should be reminded that both neutron and photon sensors are not totally selective as their signals are due to neutron and photon interactions. We intend to measure the neutron flux by three different kinds of sensors (Uranium Fission chamber, Plutonium Fission chamber and Self Powered Neutron Detector), the photon flux by two different sensors (Ionization chamber and Self Powered Gamma Detector) and the nuclear heating by two different ones (Differential calorimeter and Gamma Thermometer). For the same parameter, we expect that the use of different kinds of sensors will allow a better estimation of the aimed parameter by mixing different spectrum responses and different neutron and gamma contributions. An experimental test called CARMEN-1 is scheduled in OSIRIS reactor (CEA Saclay - France) at the end of 2011, with the goal to map irradiation locations in the reactor reflector to get a first validation of the analysis model. This article focuses on the sensor selection for CARMEN-1 experiment and to the way to link neutron and photon flux measurements in view to reduce their uncertainties but also to better assess the neutron and photon contributions to nuclear heating. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2011

An innovative acoustic sensor for first in-pile fission gas release determination - REMORA 3 experiment

A fuel rod has been instrumented with a new design of an acoustic resonator used to measure in a non destructive way the internal rod plenum gas mixture composition. This ultrasonic sensor has demonstrated its ability to operate in pile during REMORA 3 irradiation experiment carried out in the OSIRIS Material Testing Reactor (CEA Saclay, France). Due to very severe experimental conditions such as temperature rising up to 150 deg.C and especially, high thermal fluence level up to 3.5 10{sup 19} n.cm{sup 2}, the initial sensor gas speed of sound efficiency measurement was strongly reduced due to the irradiation effects on the piezo-ceramic properties. Nevertheless, by adding a differential signal processing method to the initial data analysis procedure validated before irradiation, the gas resonance peaks were successfully extracted from the output signal. From these data, the molar fractions variations of helium and fission gas were measured from an adapted Virial state equation. Thus, with this sensor, the kinetics of gas release inside fuel rods could be deduced from the in-pile measurements and specific calculations. These data will also give information about nuclear reaction effect on piezo-ceramics sensor under high neutron and gamma flux. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2011

Intracavitary in vivo dosimetry based on multichannel fiber-coupled radioluminescence and optically stimulated luminescence of Al{sub 2}O{sub 3}:C

Fiber Optic Dosimetric Catheters (FODCs) composed of chains of alumina crystals are investigated by the CEA LIST within the French ANR-INTRADOSE Project in the purpose of intracavitary in vivo dosimetry (IVD) during Brachytherapy (BT) with iridium sources and Intensity-Modulated Radiation Therapy (IMRT) with linear accelerators. A dedicated process involving PMMA fibers, cast altogether forming hexagonal bundle, is demonstrated. Optically Stimulated Luminescence (OSL) signals are recorded on-line after irradiation and absorbed doses are compared to planned dose distribution. Real-time dose measurements may also be performed by recording the Radioluminescence (RL), spontaneously emitted by the crystals during irradiation. In this case, a correction method is implemented to correct for stem effect influence (Cerenkov and scintillation generated within the fibers). For BT, the dual-fiber subtraction method is used (using a reference fiber) whereas the time discrimination method is used for IMRT. The experimental dose distribution leads to an underestimation of the source-sensor distance presumably due to energy dependence of the alumina crystal at low photon energy. At the time being, Monte-Carlo modeling of the FODC is performed with the aim to estimate this energy dependence and finally correct for it. Finally, metrological and preclinical validations are still running at Centre Leon Berard (Lyon (France)) in the purpose of checking the compliance of the FODC prototypes with treatment specifications and medical constraints. (authors)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2015

Reproducible Data Processing Research for the CABRI R.I.A. experiments Acoustic Emission signal analysis

The CABRI facility is an experimental nuclear reactor of the French Atomic Energy Commission (CEA) designed to study the behaviour of fuel rods at high burnup under Reactivity Initiated Accident (R.I.A.) conditions such as the scenario of a control rod ejection. During the experimental phase, the behaviour of the fuel element generates acoustic waves which can be detected by two microphones placed upstream and downstream from the test device. Studies carried out on the last fourteen tests showed the interest in carrying out temporal and spectral analyses on these signals by showing the existence of signatures which can be correlated with physical phenomena. We want presently to return to this rich data in order to have a new point of view by applying modern signal processing methods. Such an antecedent works resumption leads to some difficulties. Although all the raw data are accessible in the form of text files, analyses and graphics representations were not clear in reproducing from the former studies since the people who were in charge of the original work have left the laboratory and it is not easy when time passes, even with our own work, to be able to remember the steps of data manipulations and the exact setup. Thus we decided to consolidate the availability of the data and its manipulation in order to provide a robust data processing workflow to the experimentalists before doing any further investigations. To tackle this issue of strong links between data, treatments and the generation of documents, we adopted a Reproducible Research paradigm. We shall first present the tools chosen in our laboratory to implement this workflow and, then we shall describe the global perception carried out to continue the study of the Acoustic Emission signals recorded by the two microphones during the last fourteen CABRI R.I.A. tests. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

First Industrial Tests of a Matrix Monitor Correction for the Differential Die-away Technique of Historical Waste Drums

The fissile mass in radioactive waste drums filled with compacted metallic residues (spent fuel hulls and nozzles) produced at AREVA NC La Hague reprocessing plant is measured by neutron interrogation with the Differential Die-away measurement Technique (DDT). In the next years, old hulls and nozzles mixed with Ion-Exchange Resins will be measured. The ion-exchange resins increase neutron moderation in the matrix, compared to the waste measured in the current process. In this context, the Nuclear Measurement Laboratory (LMN) of CEA Cadarache has studied a matrix effect correction method, based on a drum monitor, namely a 3He proportional counter located inside the measurement cavity. After feasibility studies performed with LMN's PROMETHEE 6 laboratory measurement cell and with MCNPX simulations, this paper presents first experimental tests performed on the industrial ACC (hulls and nozzles compaction facility) measurement system. A calculation vs. experiment benchmark has been carried out by performing dedicated calibration measurements with a representative drum and {sup 235}U samples. The comparison between calculation and experiment shows a satisfactory agreement for the drum monitor. The final objective of this work is to confirm the reliability of the modeling approach and the industrial feasibility of the method, which will be implemented on the industrial station for the measurement of historical wastes. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Passive neutron coincidence counting with plastic scintillators for the characterization of radioactive waste drums

The quantification of radioactive material is essential in the fields of safeguards, criticality control of nuclear processes, dismantling of nuclear facilities and components, or radioactive waste characterization. The Nuclear Measurement Laboratory (LMN) of CEA is involved in the development of time-correlated neutron detection techniques using plastic scintillators. Usually, 3He proportional counters are used for passive neutron coincidence counting owing to their high thermal neutron capture efficiency and gamma insensitivity. However, the global {sup 3}He shortage in the past few years has made these detectors extremely expensive. In addition, contrary to {sup 3}He counters for which a few tens of microseconds are needed to thermalize fast neutrons, in view to maximize the {sup 3}He(n,p){sup 3}H capture cross section, plastic scintillators are based on elastic scattering and therefore the light signal is formed within a few nanoseconds, correlated pulses being detected within a few dozen- or hundred nanoseconds. This time span reflects fission particles time of flight, which allows reducing accordingly the duration of the coincidence gate and thus the rate of random coincidences, which may totally blind fission coincidences when using {sup 3}He counters in case of a high (α,n) reaction rate. However, plastic scintillators are very sensitive to gamma rays, requiring the use of a thick metallic shield to reduce the corresponding background. Cross talk between detectors is also a major issue, which consists on the detection of one particle by several detectors due to elastic or inelastic scattering, leading to true but undesired coincidences. Data analysis algorithms are tested to minimize cross-talk in simultaneously activated detectors. The distinction between useful fission coincidences and the correlated background due to cross-talk, (α,n) and induced (n,2n) or (n,n'γ) reactions, is achieved by measuring 3-fold coincidences. The performances of a passive neutron coincidence counting system for radioactive waste drums using plastic scintillators have been studied using the Monte Carlo radiation transport code MCNPX-PoliMi v2.0 coupled to data processing algorithms developed with ROOT data analysis software. In addition to the correlated background, accidental coincidences are taken into account in the simulation by randomly merging pulses from different calculations with fission and (α,n) sources. (authors)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2015

Measurements of SNAC2 area dosimeters placed in different configurations around the PROSPERO reactor and comparison with TRIPOLI-4 calculations

In the context of criticality accident alarm system tests, several experiments were carried out in 2013 on the PROSPERO reactor to study the response to neutron and gamma of different devices and dosimeters, particularly on the SNAC2 dosimeter. This article presents the results of this criticality dosimeter in different configurations, and compares the experimental measurements with the results of calculation performed with the TRIPOLI-4 Monte-Carlo Neutral Particles transport code. PROSPERO is a metallic critical assembly managed by the Criticality, Neutron Science and Measurement Department located at the French CEA Research Center of Valduc. The core, surrounded by a reflector of depleted uranium, is composed of 2 horizontal cylindrical blocks made of a highly enriched uranium alloy which can be placed in contact, and of 4 depleted uranium control rods which allow the reactor to be driven. This reactor, placed in a cell 10 m x 8 m x 6 m high, with 1.4-meter-thick concrete walls, is used as a fast neutron spectrum source and is operated at stable power level in delayed critical state, which can vary from 3 mW to 3 kW. PROSPERO is extensively used for electronic hardening or to study the effect of the neutrons on various materials. The SNAC2 criticality dosimeter is a zone dosimeter allowing the off line measurement of criticality accident neutron doses. This dosimeter consists of the pile up of seven activation foils embedded into a 23 mm diameter x 21 mm height cadmium container. The activation measurement of each foil, using a gamma spectroscopy technique, gives information about the neutron reaction rates. The SNAC2 software allows the spectrum unfolding from these values, taking into account the hypothesis of a particular spectrum shape, in three components: a Maxwell spectrum component for the thermal range, a 1/E component for the epithermal range, and a Watt spectrum component for the high energy range. Moreover, from the neutron spectrum, the SNAC software can calculate the neutron fluence integrated by the dosimeter and the neutron dose. During the 3 weeks measurement campaign many radioprotection devices were used. To modify the spectrum seen by these devices, several shields of various thicknesses made of concrete or polyethylene, with or without cadmium covers, were placed in the PROSPERO cell. These devices allow the study of criticality accident spectra in several environments: from metal to pseudo liquid. The fluxes measured by the SNAC2 devices were compared with TRIPOLI-4 calculations. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Scintilla European project, the successful research results

The Scintilla FP7 project is ended in December 2014, the fruitful results of 3 years development and tests will be presented. SCINTILLA offers the capacity to finding a reliable alternative to Helium-3 based detection systems since the gas which is predominantly used in nuclear safeguards and security applications has now become very expensive, rare and nearly unavailable. SCINTILLA benchmarks results are based on international standards. Radiation Portal tests were carried out at the Joint Research Centre (JRC) in Ispra (Italy). The scintilla project addresses few mains issues. The first is to develop neutron detectors for Radiation Portal Monitor (RPM) and the second is the need of new wearable integrated solutions for Spectrometric Personal Radiation Monitor (SPRM). The partners which provide technical systems of the scintilla project are INFN-ANSALDO, CEA, SYMETRICA and SAPHYMO. For RPM, the objective is to find reliable alternatives to Helium-3 historical neutron detector and provide technical solutions which cope with tests for reliable mobile and cost effective. For Spectrometric Personal Radiation Monitor (SPRM), SCINTILLA is innovating in technology areas that offer complementary capabilities for detecting and identifying gamma, Two CZT (Cadmium Zinc Telluride) addressing contexts of used by first responder technologies, one is a wearable detector and the second is a gamma camera complemented by advanced image processing technologies. (authors)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2015

Compaction in optical fibres and fibre Bragg gratings under nuclear reactor high neutron and gamma fluence

In the framework of the development by CEA and SCK.CEN of a Fabry Perot Sensor (FPS) able to measure dimensional changes in Material Testing Reactor (MTR), the first goal of the SAKE 1 (Smirnof extention - Additional Key-tests on Elongation of glass fibres) irradiation was to measure the linear compaction of single mode fibres under high fast neutron fluence. Indeed, the compaction of the fibre which forms one side of the Fabry Perot cavity, may in particular cause a noticeable measurement error. An accurate quantification of this effect is then required to predict the radiation-induced drift and optimize the sensor design. To achieve this, an innovative approach was used. Approximately seventy uncoated fibre tips (length: 30 to 50 mm) have been prepared from several different fibre samples and were installed in the SCK.CEN BR2 reactor (Mol Belgium). After 22 days of irradiation a total fast (E > 1 MeV) fluence of 3 to 5x10{sup 19} n{sub fast}/cm{sup 2}, depending on the sample location, was accumulated. The temperature during irradiation was 291 deg. C, which is not far from the condition of the intended FPS use. A precise measurement of each fibre tip length was made before the irradiation and compared to the post irradiation measurement highlighting a decrease of the fibres' length corresponding to about 0.25% of linear compaction. The amplitude of the changes is independent of the capsule, which could mean that the compaction effect saturates even at the lowest considered fluence. In the prospect of performing distributed temperature measurement in MTR, several fibre Bragg gratings written using a femtosecond laser have been also irradiated. All the gratings were written in radiation hardened fibres, and underwent an additional treatment with a procedure enhancing their resistance to ionizing radiations. A special mounting made it possible to test the reflection and the transmission of the gratings on fibre samples cut down to 30 to 50 mm. The comparison of measurements made before and after the irradiation, at the same temperature, allowed us to measure the loss in reflectivity as well as the Bragg wavelength drift. The results are quite promising for some of the investigated gratings. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Materials characterisation with the associated particle technique

Since the last past years, the Nuclear Measurement Laboratory of CEA Cadarache, France, together with Partners from European and National projects, has been studying the application of fast neutron interrogation with the Associated Particle Technique for material identification in different areas of homeland and maritime security, and for the characterisation of the materials constituting radioactive waste. Fast 14 MeV neutrons are produced from the H-3(H-2,n)alpha fusion reaction in a sealed tube neutron generator embedding an alpha detector. The alpha particle is used to tag neutron direction and emission time, thus allowing the electronic selection of neutron-induced gamma spectra in the voxels of interest. Gamma rays emitted by tagged neutron interactions on the present nuclei (C, O, N, Fe, Al, Si, Cl, etc.) are recorded with spectroscopic detectors and analysed to determine elemental proportions, thus allowing material identification. Investigations have been conducted for the detection of explosives, illicit drugs and contraband materials in cargo containers, for the inspection of objects lying on the sea floor suspected to contain explosives like mines, bombs, torpedoes, etc., for the recognition of an improvised chemical device, and for material identification in radioactive waste packages. Recently the detection of special nuclear materials is being investigated using time correlation analysis between induced fission particles, instead of gamma-ray spectroscopy. The paper presents an overview of these studies and last results. (authors)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2019

Magnetic and Mechanical Design of a 15-T Large Aperture Dipole Magnet for Cable Testing

© 2019 IEEE. A large aperture Nb 3 Sn dipole is proposed to replace the magnet assembly of EDIPO, which was irreversibly damaged in 2016. The goal is to generate a background field of 15 T at 4.2 K in a clear aperture of approximately 100×150 mm 2 and over a uniform length of 1000 mm in order to test superconducting cables for both fusion and high-energy physics applications. The magnet features a block-type coil design wound with wide Rutherford cable (two alternative coil cross sections are considered) and supported by a mechanical structure based on keys-and-bladders technology. In the end regions, the coils tilt up (flare) through a hard-way bend of the cables to provide room for the test well, following a layout already adopted in the LBNL HD2 and CERN-CEA FRESCA2 magnets. The two considered coil design alternatives aim at minimizing the mechanical stress in the coil windings. One coil pack design makes the use of two double pancake coils per pole, whereas the other alternative features three double pancakes per pole. Both design options are presented focusing on the results of numerical computations carried out with finite-element models to investigate peak stresses in the coils during room-temperature pre-loading, cool down, and powering.

Sarasola, Xabier (ORCID:0000000324306939)↗
DOE OSTI2017

Benchmarking of Improved DPAC Transient Deflagration Analysis Code

The deflagration pressure analysis code (DPAC) has been upgraded for use in modeling hydrogen deflagration transients. The upgraded code is benchmarked using data from vented hydrogen deflagration tests conducted at the HYDRO-SC Test Facility at the University of Pisa. DPAC originally was written to calculate peak pressures for deflagrations in radioactive waste storage tanks and process facilities at the Savannah River Site. Upgrades include the addition of a laminar flame speed correlation for hydrogen deflagrations and a mechanistic model for turbulent flame propagation, incorporation of inertial effects during venting, and inclusion of the effect of water vapor condensation on vessel walls. In addition, DPAC has been coupled with chemical equilibrium with applications (CEA), a NASA combustion chemistry code. The deflagration tests are modeled as end-to-end deflagrations. As a result, the improved DPAC code successfully predicts both the peak pressures during the deflagration tests and the times at which the pressure peaks.

42 ENGINEERING↗