NE213 Detector Research Articles

Measurements of the neutron spectra from the p+Be neutron generator of the NPI CAS

The Nuclear Physics Institute of the Czech Academy of Sciences (NPI CAS) operates several neutron sources, including a generator of a continuous spectrum of neutrons up to 33 MeV. Accelerated protons from the U-120M cyclotron are directed to an 8 mm thick beryllium target to produce the neutrons.In this work, the neutron spectrum at distances of 3–8 m was directly measured with NE213 and stilbene scintillator detectors in the forward, diagonal and perpendicular directions. At a distance of 50 cm from the Be target, the neutron spectrum was validated with a gold activation detector. From previous work, the neutron spectra at close distances were characterized for use in high-intensity irradiation. Both the newly measured spectra at large distances and the previously measured spectrum were simulated using the Monte Carlo code, MCNP6.2, with ENDF7PROT and JENDL-4.0/HE nuclear data libraries. The comparison of the simulated and experimentally measured spectra revealed several limitations of both libraries.

A functional form for liquid scintillator pulse shapes

Digitization of detector signals enables analysis of the original waveform to extract timing, particle identification, and energy deposition information. Here we present the use of analytical functions based on sigmoids to model and fit such pulse shapes from liquid organic scintillators, though the method should also be applicable to other detector systems. Neutron and gamma interactions in NE213 detectors were digitized from the phototube anode and fit using a sigmoid-based function. The acuity of the fit in extracting timing information and performing neutron-gamma pulse-shape discrimination are presented and discussed.

Neutron flux measurements on a mock-up of a storage cask for high-level nuclear waste using 2.5 MeV neutrons

To store and dispose spent nuclear fuel, shielding casks are employed to reduce the emitted radiation. To evaluate the exposure of employees handling such casks, Monte Carlo radiation transport codes can be employed. Nevertheless, to assess the reliability of these codes and nuclear data, experimental checks are required. In this study, a neutron generator (NG) producing neutrons of 2.5 MeV was employed to simulate neutrons produced in spent nuclear fuel. Different configurations of shielding layers of steel and polyethylene were positioned between the target of the NG and a NE-213 detector. The results of the measurements of neutron and γ radiation and the corresponding simulations with the code MCNP6 are presented. Details of the experimental set-up as well as neutron and photon flux spectra are provided as reference points for such NG investigations with shielding structures.

New physics model in GEANT4 for the simulation of neutron interactions with organic scintillation detectors

The accurate determination of the response function of organic scintillation neutron detectors complements their experimental characterization. Monte Carlo simulations with GEANT4 can reduce the effort and cost implied, especially for complex detection systems for which the characterization is more challenging. Previous studies have reported on the inaccuracy of GEANT4 in the calculation of the neutron response of organic scintillation detectors above 6 MeV, due to an incomplete description of the neutron-induced alpha production reactions on carbon. We have improved GEANT4 in this direction by incorporating models and data from NRESP, an excellent Monte Carlo simulation tool developed at the Physikalisch-Technische Bundesanstalt (PTB), Germany, for the specific purpose of calculating the neutron response function of organic scintillation detectors. The results have been verified against simulations with NRESP and validated against Time-Of-Flight measurements with an NE213 detector at PTB. This work has potential applications beyond organic scintillation detectors, to other types of detectors where reactions induced by fast neutrons on carbon require an accurate description.

Generation of the neutron response function of an NE213 scintillator for fusion applications

In this work we present a method to evaluate the neutron response function of an NE213 liquid scintillator. This method is particularly useful when the proton light yield function of the detector has not been measured, since it is based on a proton light yield function taken from literature, MCNPX simulations, measurements of gamma-rays from a calibration source and measurements of neutrons from fusion experiments with ohmic plasmas. The inclusion of the latter improves the description of the proton light yield function in the energy range of interest (around 2.46 MeV).We apply this method to an NE213 detector installed at JET, inside the radiation shielding of the magnetic proton recoil (MPRu) spectrometer, and present the results from the calibration along with some examples of application of the response function to perform neutron emission spectroscopy (NES) of fusion plasmas.We also investigate how the choice of the proton light yield function affects the NES analysis, finding that the result does not change significantly. This points to the fact that the method for the evaluation of the neutron response function is robust and gives reliable results.

Neutron spectrum unfolding using artificial neural network and modified least square method

In the present paper, neutron spectrum is reconstructed using the Artificial Neural Network (ANN) and Modified Least Square (MLSQR) methods. The detector's response (pulse height distribution) as a required data for unfolding of energy spectrum is calculated using the developed MCNPX-ESUT computational code (MCNPX-Energy engineering of Sharif University of Technology). Unlike the usual methods that apply inversion procedures to unfold the energy spectrum from the Fredholm integral equation, the MLSQR method uses the direct procedure. Since liquid organic scintillators like NE-213 are well suited and routinely used for spectrometry of neutron sources, the neutron pulse height distribution is simulated/measured in the NE-213 detector. The response matrix is calculated using the MCNPX-ESUT computational code through the simulation of NE-213 detector's response to monoenergetic neutron sources. For known neutron pulse height distribution, the energy spectrum of the neutron source is unfolded using the MLSQR method. In the developed multilayer perception neural network for reconstruction of the energy spectrum of the neutron source, there is no need for formation of the response matrix. The multilayer perception neural network is developed based on logsig, tansig and purelin transfer functions. The developed artificial neural network consists of two hidden layers of type hyperbolic tangent sigmoid transfer function and a linear transfer function in the output layer. The motivation of applying the ANN method may be explained by the fact that no matrix inversion is needed for energy spectrum unfolding. The simulated neutron pulse height distributions in each light bin due to randomly generated neutron spectrum are considered as the input data of ANN. Also, the randomly generated energy spectra are considered as the output data of the ANN. Energy spectrum of the neutron source is identified with high accuracy using both MLSQR and ANN methods. The results obtained from MLSQR and ANN methods for 252Cf and 241Am-9Be source are validated against the ISO spectrum. The unfolded neutron energy spectra from both MLSQR and ANN methods show a good agreement with the actual spectrum of 252Cf and 241Am-9Be source.

Development of MCNPX-ESUT computer code for simulation of neutron/gamma pulse height distribution

In this paper, the development of the MCNPX-ESUT (MCNPX-Energy Engineering of Sharif University of Technology) computer code for simulation of neutron/gamma pulse height distribution is reported. Since liquid organic scintillators like NE-213 are well suited and routinely used for spectrometry in mixed neutron/gamma fields, this type of detectors is selected for simulation in the present study. The proposed algorithm for simulation includes four main steps. The first step is the modeling of the neutron/gamma particle transport and their interactions with the materials in the environment and detector volume. In the second step, the number of scintillation photons due to charged particles such as electrons, alphas, protons and carbon nuclei in the scintillator material is calculated. In the third step, the transport of scintillation photons in the scintillator and lightguide is simulated. Finally, the resolution corresponding to the experiment is considered in the last step of the simulation.Unlike the similar computer codes like SCINFUL, NRESP7 and PHRESP, the developed computer code is applicable to both neutron and gamma sources. Hence, the discrimination of neutron and gamma in the mixed fields may be performed using the MCNPX-ESUT computer code. The main feature of MCNPX-ESUT computer code is that the neutron/gamma pulse height simulation may be performed without needing any sort of post processing.In the present study, the pulse height distributions due to a monoenergetic neutron/gamma source in NE-213 detector using MCNPX-ESUT computer code is simulated. The simulated neutron pulse height distributions are validated through comparing with experimental data (Gohil et al. Nuclear Instruments and Methods in Physics Research Section A: Accelerators, Spectrometers, Detectors and Associated Equipment, 664 (2012) 304–309.) and the results obtained from similar computer codes like SCINFUL, NRESP7 and Geant4. The simulated gamma pulse height distribution for a 137Cs source is also compared with the experimental data.

Measurement of Gamma Energy Distributions and Multiplicities Using STEFF

Prompt gamma-ray energy distributions and multiplicities released during thermally induced fission of 235U have been measured using STEFF (SpecTrometer for Exotic Fission Fragments). Thermal neutrons are provided by the high-flux reactor at the ILL, Grenoble. STEFF is a unique 2E2v device that uses a coincidence timing method to measure the emission of prompt gamma rays as a function of the fragment mass and energy. Following scission, the fission fragments contain excitation energy that is released via prompt neutron and gamma-ray emission as the fragment decays to the ground state. STEFF contains an array of 11 NaI detectors surrounding the uranium target providing a 6.8% photopeak detection efficiency for gamma rays released within 1 ns of the scission time. STEFF also consists of 7 NE213 detectors, which detect the emission of prompt neutrons, the release of which is associated with reduction of fragment energy and, to a lesser extent, fragment spin. This experiment acts as a direct response to the NEA high priority demand which requires more accurate knowledge of heating caused by gamma emission in the next generation of nuclear reactors.

Comparison of deuterated and normal liquid scintillators for fast-neutron detection

Liquid organic scintillators have long been used for fast-neutron detection due to their fast response and the ability to use pulse-shape discrimination (PSD) to distinguish between neutron and γ-ray interactions. Deuterated liquid organic scintillators have, in addition to the aforementioned properties, structure on their pulse-height spectra due to the different characteristics of n–d vs. n–p scattering.We report on the direct comparison of two small-volume liquid scintillator detectors with exactly the same geometries (11.43cm diameter and 2.54cm thickness); the first based on NE213 produced by Nuclear Enterprise, and the second based on EJ-315 produced by Eljen Technologies, the latter being a deuterated scintillator. A variety of photomultiplier tubes (PMTs) were used, but the quality of the data was determined to be independent of the PMT type.It is shown that, while the light output of the deuterated detector is lower than for the conventional non-deuterated detector, it is possible to detect 60keV neutrons. The PSD capabilities of the deuterated detector match and surpass those of the NE213 detector. Its efficiency is comparable to that of the NE213 detector at neutron energies above 2MeV; below that energy its relative efficiency suffers from the significantly lower n–d scattering cross-section. For very low neutron energies (<200keV), the relative efficiency of the deuterated detector increases again, due to the lower noise level for our particular detectors, showing the importance of low photomultiplier tube noise for the detection of low-energy neutrons.

Reactor power measurement by gamma and neutron radiation in Heavy Water Zero Power Reactor (HWZPR)

Both gamma rays and neutrons are produced and distributed in different nuclear reactions within the Heavy Water Zero Power Reactor (HWZPR) core. The intensities of distributed gamma rays and neutrons from the reactor core are proportional to the reactor power. The NE213 organic scintillation detector can detect both gamma and neutron simultaneously. In this article, a new method has been used for calibration of the reactor power by measurement of gamma rays and neutrons with the 2″×2″ dimensions NE213 detector and suitable nuclear electronic equipments outside of core. The reactor calibrated power by this technique is on-line, prompt, and independent of safety and control rods and fuel configuration of the reactor with a good efficiency. The detector, when used 7–10m from the core center, has necessary efficiency for full-power range measurement (100W) and in case of accident occurring and sudden increase of the power up to 170W.

Methodology for the neutron time of flight measurement of 120-GeV proton-induced reactions on a thick copper target

A methodology for the time-of-flight measurement of the neutron energy spectrum for a high-energy proton-beam-induced reaction was established at the Fermilab Test Beam Facility of the Fermi National Accelerator Laboratory. The 120-GeV proton beam with 3×105protons/spill was prepared for event-by-event counting of incident protons and emitted neutrons for time-of-flight energy determination. An NE213 organic liquid scintillator (12.7cm in diameter by 12.7cm in length) was employed with a veto plastic scintillator and a pulse-shape discrimination technique to identify neutrons. Raw waveforms of NE213, veto and beam detectors were recorded to discriminate the effects of multi-proton beam events by considering different time windows. The neutron energy spectrum ranging from 10 to 800MeV was obtained for a 60-cm-long copper target at 90° with respect to the beam axis. The obtained spectrum was consistent with that deduced employing the conventional unfolding technique as well as that obtained in a 40-GeV/c thin-target experiment.

Neutron beam preparation with Am–Be source for analysis of biological samples with PGNAA method

Material analysis with prompt gamma neutron activation analysis (PGNAA) requires a proper geometrical arrangement for equipments in laboratory. Application of PGNAA in analysis of biological samples, due to small size of sample, needs attention to the dimension of neutron beam. In our work, neutron source has been made of 241Am–Be type. Activity of 241Am was 20 Ci which lead to neutron source strength of 4.4 × 107 neutrons per second. Water has been considered as the basic shielding material for the neutron source. The effect of various concentration of boric acid in the reduction of intensity of fast and thermal components of the neutron beam and gamma ray has been investigated. Gamma ray is produced by (α, n) reaction in Am–Be source (4.483 MeV), neutron capture by hydrogen (2.224 MeV), and neutron capture by boron (0.483 MeV). Various types of neutron and gamma ray dosimeters have been employed including BF3 and NE-213 detectors to detect fast and thermal neutrons. BGO scintillation detector has been used for gamma ray spectroscopy. It is shown that the gamma and neutron radiation dose due to direct beam is of the same magnitude as the dose due to radiation scattered in the laboratory ambient. It is concluded that 14 kg boric acid dissolved in 1,000 kg water is the optimum solution to surround the neutron source. The experimental results have been compared with Monte Carlo simulation.

Measurement and analysis of neutron flux spectra in a neutronics mock-up of the HCLL test blanket module

Fast neutron and gamma-ray flux spectra and time-of-arrival spectra of slow neutrons have been measured in a neutronics mock-up of the European Helium-Cooled Lithium–Lead Test Blanket Module with the aim to validate nuclear cross-section data. The mock-up was irradiated with fusion peak neutrons from the DT neutron generator of the Technical University of Dresden. A well characterized cylindrical NE-213 scintillator was inserted into two positions in the LiPb/EUROFER assembly. Pulse height spectra from neutrons and gamma-rays were recorded from the NE-213 output. The spectra were then unfolded with experimentally obtained response matrices of the NE-213 detector. Time-of-arrival spectra of slow neutrons were measured with a 3He counter placed in the mock-up, and the neutron generator was operated in pulsed mode. Monte Carlo calculations using the MCNP code and nuclear cross-section data from the JEFF-3.1.1 and FENDL-2.1 libraries were performed and the results are compared with the experimental results. A good agreement of measurement and calculation was found with some deviations in certain energy intervals.

Nuclear Data Benchmarking for Lead by LiAl/Pb Fusion Blanket Mock-Up Experiment

For the International Thermonuclear Experimental Reactor (ITER), the European Union intends to introduce a test blanket module using liquid lithium lead. In the present study, a preliminary experiment was conducted using a LiAl/Pb assembly, which has a neutronics performance similar to lithium lead. The neutron spectrum at the inside of the assembly was measured with an NE-213 detector. We found that the spectra calculated with MCNP5 and JENDL-3.3 underestimate the 14-MeV peak by no less than 30%, while ENDF/B-VII.0 and JEFF-3.1 gave good agreement. The neutron nuclear data of lead stored in JENDL-3.3 may have some problems.

High Count Rate Neutron Spectrometry With Liquid Scintillation Detectors

Liquid scintillation detectors are widely used in nuclear/high-energy physics and nuclear fusion for spectral measurements in mixed radiation fields due to their compactness, fast response and neutron/gamma discrimination capabilities. The use of response functions evaluated for the specific system and of appropriate methods of data analysis allows such systems to be used as broadband spectrometers for photons and neutrons. System stability and ability to reach high throughput count rates are key challenges for several applications (e.g., neutron spectrometry for nuclear fusion devices), but standard analog electronics limits the operation of liquid scintillation neutron spectrometers to low count rates ( ~ 3 ldr 10 <sup xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">4</sup> s <sup xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">-1</sup> ) . The count rate capabilities of a liquid scintillation neutron spectrometer (NE213 detector) from the Physikalisch-Technische Bundesanstalt (PTB) has been extended up to ~ 4.2 ldr 10 <sup xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">5</sup> s <sup xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">-1</sup> , by coupling it to a digital acquisition system developed at ENEA-Frascati. Measurements have been carried out at PTB using gamma sources and accelerator-produced 2.5 MeV and 14 MeV neutrons. For 14 MeV neutron measurements, digital pulse height spectra (PHS) obtained at high count rates have been compared to PHS recorded with standard analog electronics. The results show that stable PHS (within 1%) can be obtained at high count rate despite the high sensitivity of the gain of photomultiplier tubes to count rate variations.

Blanket mock-up experiment with a LiAlPb assembly irradiated with 14 MeV neutrons in preparation of the HCLL-TBM neutronics experiment

Irradiation experiments with a LiAlPb assembly have been performed using the DT neutron generator of Technical University of Dresden. The main goal was to qualify LiF thermoluminescence detectors (TLDs) for tritium breeding experiments and to test a recently improved NE-213 spectrometer, both for application in a neutronics mock-up experiment of the European HCLL-TBM at ENEA Frascati and Technical University of Dresden (TUD) planned for 2008. The LiF-TLD were set up inside the LiAlPb assembly and irradiated with the DT neutron generator of TUD in continuous mode. A good agreement with TPR calculated with MCNP was found. Fast neutron spectra inside the assembly have been obtained with the NE-213 detector. The flux of the thermal and epithermal neutrons inside the assembly, which is important for the tritium production rate from Li 6 , was measured with a He 3 counter and the DT neutron generator operated in pulsed mode.

Modeling, analysis and prediction of neutron emission spectra from acoustic cavitation bubble fusion experiments

Self-nucleated and external neutron nucleated acoustic (bubble fusion) cavitation experiments have been modeled and analyzed for neutron spectral characteristics at the detector locations for all separate successful published bubble fusion studies. Our predictive approach was first calibrated and validated against the measured neutron spectrum emitted from a spontaneous fission source ( 252Cf), from a Pu–Be source and from an accelerator-based monoenergetic 14.1 MeV neutrons, respectively. Three-dimensional Monte-Carlo neutron transport calculations of 2.45 MeV neutrons from imploding bubbles were conducted, using the well-known MCNP5 transport code, for the published original experimental studies of Taleyarkhan et al. [Taleyarkhan, et al., 2002. Science 295, 1868; Taleyarkhan, et al., 2004. Phys. Rev. E 69, 036109; Taleyarkhan, et al., 2006a. PRL 96, 034301; Taleyarkhan, et al., 2006b. PRL 97, 149404] as also the successful confirmation studies of Xu et al. [Xu, Y., et al., 2005. Nuclear Eng. Des. 235, 1317–1324], Forringer et al. [Forringer, E., et al., 2006a. Transaction on American Nuclear Society Conference, vol. 95, Albuquerque, NM, USA, November 15, 2006, p. 736; Forringer, E., et al., 2006b. Proceedings of the International Conference on Fusion Energy, Albuquerque, NM, USA, November 14, 2006] and Bugg [Bugg, W., 2006. Report on Activities on June 2006 Visit, Report to Purdue University, June 9, 2006]. NE-213 liquid scintillation (LS) detector response was calculated using the SCINFUL code. These were cross-checked using a separate independent approach involving weighting and convoluting MCNP5 predictions with published experimentally measured NE-213 detector neutron response curves for monoenergetic neutrons at various energies. The impact of neutron pulse-pileup during bubble fusion was verified and estimated with pulsed neutron generator based experiments and first-principle calculations. Results of modeling-cum-experimentation were found to be consistent with published experimentally observed neutron spectra for 2.45 MeV neutron emissions during acoustic cavitation (bubble) fusion experimental conditions with and without ice-pack (thermal) shielding. Calculated neutron spectra with the inclusion of ice-pack shielding are consistent with the published spectra from experiments of Taleyarkhan et al. [Taleyarkhan, et al., 2006a. PRL 96, 034301] and Xu et al. [Xu, Y., et al., 2005. Nuclear Eng. Des. 235, 1317–1324] where ice-pack shielding was present, whereas without ice-pack shielding the calculated neutron spectrum is consistent with the experimentally observed neutron spectra of Taleyarkhan et al. [Taleyarkhan, et al., 2002. Science 295, 1868; Taleyarkhan, et al., 2004. Phys. Rev. E 69, 036109] and Forringer et al. [Forringer, E., et al., 2006a. Transaction on American Nuclear Society Conference, vol. 95, Albuquerque, NM, USA, November 15, 2006, p. 736; Forringer, E., et al., 2006b. Proceedings of the International Conference on Fusion Energy, Albuquerque, NM, USA, November 14, 2006] and also that from GEANT computer code [Agostinelli, S., et al., 2003. Nuclear Instrum. Methods Phys. Res. A 506, 250–303] predictions [Naranjo, B., 2006. PRL 97 (October), 149403] in which ice shielding was also absent. The results of this archive confirm for the record that the confusion and controversies caused from past reports [Reich, E., 2006. Nature (March) 060306. news@nature.com; Naranjo, B., 2006. PRL, 97 (October) 149403] have resulted from their neglect of important details of bubble fusion experiments. Results from this paper demonstrate that ice-pack shielding between the detector and the fusion neutron source, gamma photon leakage and neutron pulse-pileup due to picosecond duration neutron pulse emission effects play important roles in affecting the spectra of neutrons from acoustic inertial confinement thermonuclear fusion experiments.

A method for digital processing of pile-up events in organic scintillators

Pile-up events in radiation detectors are those in which two or more pulses partly or completely overlap. In standard analog electronics pile-up events are rejected using a pile-up inspector. When digital acquisition techniques are used, the recorded waveforms of pile-ups can be elaborated and the contributing single pulses reconstructed. A method for the off-line digital processing of pile-ups from liquid organic scintillators (NE213) is proposed: pile-ups in which the overlapping pulses are separated more than 15 ns can be reduced to single pulses and then correctly identified as neutrons (n) or gammas (γ). An analysis of the errors introduced by the method is carried out. The method has been applied to data acquired in the mixed n–γ field of JET deuterium plasma discharges from the NE213 detector in the central line of sight of the neutron profile monitor and the results are described.

Simulation of a NE213 γ-response function using an improved multi-parameter minimization procedure

The γ-response functions of a 15 cm diameter×6 cm thick NE213 detector were recently measured with standard 22Na, 60Co and 241Am γ-sources in order to calibrate the scintillation counter and were simulated in the present work with the MCNPX2.5e Monte-Carlo code. The MINUIT2 package, developed at C.E.R.N., was used to minimize the χ 2 between measured and calculated pulse height distributions in order to derive the experimental resolution parameters. A C ++ code was written to simulate various resolution broadening and background shapes. The results are in good agreement with the experimental response function except in the low-energy range (below 100 keV) which could be improved by a better description of the detector environment.

06/00089 Importance of the MUSE experiments foremerging ADS concepts from the nuclear data viewpoint: Plaschy, M. et al. Annals of Nuclear Energy, 2005, 32, (8), 843–856

Fusion integral benchmark experiments have been carried out for copper and tungsten at FNS of JAERI. Leakage neutron spectra were measured with an NE213 detector by the TOF method to validate evaluated nuclear data files of JENDL-3.2, JENDL-fusion file, ENDF/B-VI, FENDL/E-1.0 and FENDL/E-2.0. From the result of comparison between experiments and calculations by MCNP-4B, below 1 MeV disagreement was observed for copper and it became remarkable with increase of the thickness. It was therefore, pointed out that there might be a problem in the evaluated (n, 2n) neutron spectrum for all the nuclear data libraries. For tungsten, the calculated result with JENDL-fusion file and FENDL/E-2.0 overpredicted the experimental values around 3–7 MeV. Near 10 MeV, underestimation of ENDF/B-VI was remarkably pronounced. For the whole energy range, JENDL-3.2 was preferable. However, there was a serious problem left that discrepancy between JENDL-fusion file and JENDL-3.2 was very large.