Neutron diffraction measurements at the INES diffractometer using a neutron radiative capture based counting technique

The global shortage of ³He gas is an issue to be addressed in neutron detection. In the context of the research and development activity related to the replacement of ³He for neutron counting systems, neutron diffraction measurements performed on the INES beam line at the ISIS pulsed spallation neutron source are presented. For these measurements two different neutron counting devices have been used: a 20 bar pressure squashed ³He tube and a Yttrium-Aluminum-Perovskite scintillation detector. The scintillation detector was coupled to a cadmium sheet that registers the prompt radiative capture gamma rays generated by the (n,γ) nuclear reactions occurring in cadmium. The assessment of the scintillator based counting system was done by performing a Rietveld refinement analysis on the diffraction pattern from an ancient Japanese blade and comparing the results with those obtained by a ³He tube placed at the same angular position. The results obtained demonstrate the considerable potential of the proposed counting approach based on the radiative capture gamma rays at spallation neutron sources.     

Conversion factors from counts to chemical ratios for the EURITRACK tagged neutron inspection system

The EURopean Illicit TRAfficking Countermeasures Kit (EURITRACK) uses 14 MeV neutrons produced by the ³H(d,n)⁴H fusion reaction to detect explosives and narcotics in cargo containers. Reactions induced by fast neutrons produce gamma rays, which are detected in coincidence with the associated alpha particle to determine the neutron direction. In addition, the neutron path length is obtained from a time-of-flight measurement, thus allowing the origin of the gamma rays inside the container to be determined. Information concerning the chemical composition of the target material is obtained from the analysis of the energy spectrum. The carbon, oxygen, and nitrogen relative count contributions must be converted to chemical proportions to distinguish illicit and benign organic materials. An extensive set of conversion factors based on Monte Carlo numerical simulations has been calculated, taking into account neutron slowing down and photon attenuation in the cargo materials. An experimental validation of the method is presented by comparing the measured chemical fractions of known materials, in the form of bare samples or hidden in a cargo container, to their real chemical composition. Examples of application to real cargo containers are also reported, as well as simulated data with explosives and illicit drugs.     

Detection unit optimization of a neutron searching detector using Monte Carlo Simulations

The NSD is a portable Neutron Searching Detector developed at Rotem Industries Ltd. with a high efficiency for counting fast and thermal neutrons employing improved gamma rejection. The NSD detection-unit consists of two ³He detectors installed within a polypropylene moderator. The latest international standards for detection of illicit trafficking of radioactive materials require high sensitivity, relatively small dimensions, and light mass. In order for it to meet these standards, the NSD detection-unit was optimized using Monte Carlo N-Particle transport code (MCNP). The moderator mass and dimensions were reduced without deterioration, even improving the instrument's sensitivity. The purposed moderator improvements covered in this paper work well for traditional hand-held neutron search detectors based on ³He tubes as well as for new neutron detection technologies due to the severe worldwide shortage of ³He.Three geometrical moderator configurations were examined using the MCNP code—a rectangular box, a circular cylinder, and an elliptical base cylinder. The optimization results showed that both the rectangular box moderator and the elliptical base cylinder moderators achieve the appropriate sensitivity required by the standards with about 30% reduced mass. A prototype was fabricated with the rectangular box moderator configuration, and its response was successfully validated by comparing empirical measurements against the results of the MCNP code.Performance examination of the optimal detection unit prototype was made regarding the latest international standards. The results showed a 17% improvement in detection limit for radioactive materials along with a 14% to 17% increased neutron detection response, while keeping the false alarm rate below the required threshold, and maintaining a 26% mass reduction.     

Gamma-ray detection via D(γ, n)H neutron counting

The photodisintegration of deuterium (Q = −2.22 MeV) in combination with a neutron counter has been used to detect high-energy γ-rays. The γ-ray source is surrounded by a tank filled with heavy water and the emitted neutrons are counted in 4π geometry with neutron detectors embedded in a graphite moderator. The absolute detection efficiency of a small test system (1.8 1 D₂O) has been determined for E_γ = 2.6–8.1 MeV and found to agree well with calculations (e.g., 8.0 × 10⁻⁵ at E_γ = 4.4 MeV). It is proposed that the system should be useful for the study of capture reactions induced on highly radioactive targets. Improvements in detection efficiency as well as limitations in data analyses are discussed.     

Microchannel plate special nuclear materials sensor

Nova Scientific Inc., is developing for the Domestic Nuclear Detection Office (DNDO SBIR #HSHQDC-08-C-00190), a solid-state, high-efficiency neutron detection alternative to ³He gas tubes, using neutron-sensitive microchannel plates (MCPs) containing ¹⁰B and/or Gd. This work directly supports DNDO development of technologies designed to detect and interdict nuclear weapons or illicit nuclear materials. Neutron-sensitized MCPs have been shown theoretically and more recently experimentally, to be capable of thermal neutron detection efficiencies equivalent to ³He gas tubes. Although typical solid-state neutron detectors typically have an intrinsic gamma sensitivity orders of magnitude higher than that of ³He gas detectors, we dramatically reduce gamma sensitivity by combining a novel electronic coincidence rejection scheme, employing a separate but enveloping gamma scintillator. This has already resulted in a measured gamma rejection ratio equal to a small ³He tube, without in principle sacrificing neutron detection efficiency. Ongoing improvements to the MCP performance as well as the coincidence counting geometry will be described. Repeated testing and validation with a ²⁵²Cf source has been underway throughout the Phase II SBIR program, with ongoing comparisons to a small commercial ³He gas tube. Finally, further component improvements and efforts toward integration maturity are underway, with the goal of establishing functional prototypes for SNM field testing.     

Monte Carlo simulation of pulse pile-up effect in gamma spectrum of a PGNAA system

We have applied a pile-up Monte Carlo simulation code on gamma spectrum of a prompt gamma neutron activation analysis (PGNAA) system. The code has been run in nonparalyzable mode for a specific geometry of a PGNAA system with ²⁴¹Am-⁹Be source and NaI(Tl) detector to obtain the distortion due to “pile-up” in the pulse height of gamma spectrum. The results show that the main background in the nitrogen region of interest (ROI) is due to two pile-ups. We have also evaluated the variation of count rate and total photon sampling over the Monte Carlo spectra. At high count rates, not only the nitrogen ROI but also carbon ROI, and hydrogen peak are disturbed strongly. Comparison between the results of simulations and the experimental spectra has shown a good agreement. The code could be used for other source setups and different gamma detection systems.     

An airport cargo inspection system based on X-ray and thermal neutron analysis (TNA)

A cargo inspection system incorporating a high-resolution X-ray imaging system with a material-specific detection system based on Ancore Corporation's patented thermal neutron analysis (TNA) technology can detect bulk quantities of explosives and drugs concealed in trucks or cargo containers. The TNA process utilises a 252Cf neutron source surrounded by a moderator. The neutron interactions with the inspected object result in strong and unique gamma-ray signals from nitrogen, which is a key ingredient in modern high explosives, and from chlorinated drugs. The TNA computer analyses the gamma-ray signals and automatically determines the presence of explosives or drugs. The radiation source terms and shielding design of the facility are described. For the X-ray generator, the primary beam, leakage radiation, and scattered primary and leakage radiation were considered. For the TNA, the primary neutrons and tunnel scattered neutrons as well as the neutron-capture gamma rays were considered.     

In situ prompt gamma-ray activation analysis of water pollutants using a shallow 252Cf-HPGe probe

A shallow ²⁵²Cf-HPGe probe used for in situ prompt γ-ray activation of water pollutants is described. A 2.7 μg ²⁵²Cf neutron source and a 10% HPGe detector are inserted into a waterproof stainless steel probe, which is designed to be submerged and recovered in field operation. A laboratory test is performed to obtain the neutron flux distribution and prompt γ-ray contribution to the HPGe detector counts from around the submerged probe. The concentrations of toxic cadmium and chlorine in water are determined in the prompt γ-ray spectrum. The detection limit of industrial pollutants and some improvements of the current design are discussed.     

Neutron detection gamma ray sensitivity criteria

The shortage of ³He has triggered the search for effective alternative neutron detection technologies for national security and safeguards applications. Any new detection technology must satisfy two basic criteria: (1) it must meet a neutron detection efficiency requirement, and (2) it must be insensitive to gamma-ray interference at a prescribed level, while still meeting the neutron detection requirement. It is the purpose of this paper to define measureable gamma ray sensitivity criteria for neutron detectors. Quantitative requirements are specified for: intrinsic gamma ray detection efficiency and gamma ray absolute rejection. The gamma absolute rejection ratio for neutrons (GARRn) is defined, and it is proposed that the requirement for neutron detection be 0.9<GARRn<1.1 at a 10 mR/h exposure rate. An example of the results from a ³He based neutron detector is provided showing that this technology can meet the stated requirements. Results from tests of some alternative technologies are also reported.     

Commissioning of the NPDGamma Detector Array: Counting Statistics in Current Mode Operation and Parity Violation in the Capture of Cold Neutrons on B4C and 27Al

The NPDGamma γ-ray detector has been built to measure, with high accuracy, the size of the small parity-violating asymmetry in the angular distribution of gamma rays from the capture of polarized cold neutrons by protons. The high cold neutron flux at the Los Alamos Neutron Scattering Center (LANSCE) spallation neutron source and control of systematic errors require the use of current mode detection with vacuum photodiodes and low-noise solid-state preamplifiers. We show that the detector array operates at counting statistics and that the asymmetries due to B₄C and ²⁷Al are zero to with- in 2 × 10⁻⁶ and 7 × 10⁻⁷, respectively. Boron and aluminum are used throughout the experiment. The results presented here are preliminary.     

A CdTe detector with a Gd converter for thermal neutron detection

A CdTe detector with a Gd converter has been developed and investigated as a neutron detector for neutron imaging. The fabricated Gd/CdTe detector with the 25 μm thick Gd was designed on the basis of simulation results of thermal neutron detection efficiency and spatial resolution. The energy resolution of the Gd/CdTe detector is less than 4 keV, which is enough to discriminate neutron capture gamma rays from background gamma emission. The Gd/CdTe detector shows the detection of neutron capture gamma ray emission in the ¹⁵⁵Gd(n, γ)¹⁵⁶Gd, ¹⁵⁷Gd(n, γ)¹⁵⁸Gd and ¹¹³Cd(n, γ)¹¹⁴Cd reactions and characteristic X-ray emissions due to conversion-electrons generated inside the Gd film. The observed efficient thermal neutron detection with the Gd/CdTe detector shows its promise in neutron radiography application.     

Gamma sensitivity of single-crystal CVD diamond neutron detectors

We have studied the gamma sensitivity of single-crystal CVD diamond neutron detectors using a ²⁵²Cf neutron source placed in a moderator. It has been shown that a major contribution to the count rate of the detectors is made by the gamma rays from the source. We have compared the count rates of a detector with a ¹⁰B boron isotope-based slow-neutron converter and without it. With allowance for the theoretically calculated detection efficiency, the difference between the count rates is consistent with the fraction of slow neutrons measured using a scintillation detector.     

Neutron threshold activation detectors (TAD) for the detection of fissions

Prompt fission neutrons are one of the strongest signatures of the fission process. Depending on the fission inducing radiation, their average number ranges from 2.5 to 4 neutrons per fission. They are more energetic and abundant, by about 2 orders of magnitude, than the delayed neutrons (≈3 vs. ≈0.01) that are commonly used as indicators for the presence of fissionable materials.The detection of fission prompt neutrons, however, has to be done in the presence of extremely intense probing radiation that stimulated them. During irradiation, the fission stimulation radiation, X-rays or neutrons, overwhelms the neutron detectors and temporarily incapacitate them. Consequently, by the time the detectors recover from the source radiation, fission prompt neutrons are no longer emitted. In order to measure the prompt fission signatures under these circumstances, special measures are usually taken with the detectors such as heavy shielding with collimation, use of inefficient geometries, high pulse height bias and gamma-neutron separation via pulse-shape discrimination with an appropriate organic scintillator. These attempts to shield the detector from the flash of radiation result in a major loss of sensitivity. It can lead to a complete inability to detect the fission prompt neutrons.In order to overcome the blinding induced background from the source radiation, the detection of prompt fission neutrons needs to occur long after the fission event and after the detector has fully recovered from the source overload. A new approach to achieve this is to detect the delayed activation induced by the fission neutrons. The approach demonstrates a good sensitivity in adverse overload situations (gamma and neutron “flash”) where fission prompt neutrons could normally not be detected.The new approach achieves the required temporal separation between the detection of prompt neutrons and the detector overload by the neutron activation of the detector material. The technique, called Threshold Activation Detection (TAD), is to utilize appropriate substances that can be selectively activated by the fission neutrons and not by the source radiation and then measure the radioactively decaying activation products (typically beta and gamma rays) well after the source pulse. The activation material should possess certain properties: a suitable half-life of the order of seconds; an energy threshold below which the numerous source neutrons will not activate it (e.g., 3 MeV); easily detectable activation products (typically >1 MeV beta and gamma rays) and have a usable cross-section for the selected reaction. Ideally the substance would be a part of the scintillator.There are several good material candidates for the TAD, including fluorine, which is a major constituent of available scintillators such as BaF₂, CaF₂ and hydrogen free liquid fluorocarbon. Thus the fluorine activation products, in particular the beta particles, can be measured with a very high efficiency in the detector.The principles, applications and experimental results obtained with the fluorine based TAD are discussed.     

Dose measuring system for boron neutron capture therapy

Dose measuring systems for boron neutron capture therapy (BNCT) of brain tumors are presented. The systems are a real-time monitoring system, an integral measuring system and a ¹⁰B concentration measuring system. The real-time monitoring with a small PN junction silicon detector made it possible to simultaneously measure the thermal neutron flux and the gamma dose rate in a patient during neutron therapy. Another monitoring of dose equivalents of thermal neutrons and gamma rays was performed with a BGO scintillation detector connected to an optical fiber. The accurate neutron fluence and gamma dose were determined with the integral measurements of the foil activation method and thermoluminescent dosimeters (TLDs) after irradiation. Kerma doses of thermal neutrons and gamma-rays were also measured with the TLD at the same time. Preliminary measurements of ¹⁰B concentration in tissue and blood of a patient were carried out by prompt gamma-ray spectroscopy.     

Detecting fissionable materials in a variety of shielding matrices via delayed gamma and neutron photofission signatures—Part 2: Experimental results

Successful detection of fissionable material contained in a variety of matrices was demonstrated by photon active interrogation of fissionable and inert target materials. Samples were irradiated with pulsed 15 MeV photons generated by a LINAC and tungsten electron/photon converter, operating at 15 Hz. Matrix materials included air (no matrix), wood, water, and lead. A unique dual mode gamma/neutron detector was used to acquire data from both fission product gamma and fission product neutron emission. Neutron emission was recorded by detecting the 478 keV capture gamma from the ¹⁰B (n,α)⁷Li reaction, generating a photopeak in the recorded gamma spectrum. Two signatures were found to correctly differentiate between the fissionable target (²³⁸U) and inert targets (lead, steel, air, and beryllium), with substantial differences in delayed gamma and neutron signatures for fissionable and inert materials in all cases. The signatures are simple to compute and are not significantly affected by system variations or interferences expected during cargo scanning.     

He and BF3 neutron detector pressure effect and model comparison

Radiation detection systems for homeland security applications must possess the capability of detecting both gamma rays and neutrons. The radiation portal monitor systems that are currently deployed use a plastic scintillator for detecting gamma rays and ³He gas-filled proportional counters for detecting neutrons. Proportional counters filled with ³He are the preferred neutron detectors for use in radiation portal monitor systems because ³He has a large neutron cross-section, is relatively insensitive to gamma-rays, is neither toxic nor corrosive, can withstand extreme environments, and can be operated at a lower voltage than some of the alternative proportional counters. The amount of ³He required for homeland security and science applications has depleted the world supply and there is no longer enough available to fill the demand. Thus, alternative neutron detectors are being explored.Two possible temporary solutions that could be utilized while a more permanent solution is being identified are reducing the ³He pressure in the proportional counters and using boron trifluoride gas-filled proportional counters. Reducing the amount of ³He required in each of the proportional counters would decrease the rate at which ³He is being used; not enough to solve the shortage, but perhaps enough to increase the amount of time available to find a working replacement. Boron trifluoride is not appropriate for all situations as these detectors are less sensitive than ³He, boron trifluoride gas is corrosive, and a much higher voltage is required than what is used with ³He detectors. Measurements of the neutron detection efficiency of ³He and boron trifluoride as a function of tube pressure were made. The experimental results were also used to validate models of the radiation portal monitor systems.     

Gamma ray measurements during deuterium and 3He discharges on TFTR

Gamma ray count rates and energy spectra have been measured in TFTR deuterium plasmas during ohmic heating and during injection of deuterium neutral beams for total neutron source strengths up to 6 × 10¹⁵ neutrons per second. The gamma ray measurements for the deuterium plasmas are in general agreement with predictions obtained using simplified transport models. The 16.6 MeV fusion gamma ray from the direct capture reaction D(³He, γ)⁵Li was observed during deuterium neutral beam injection into ³He plasmas for beam powers up to 7 MW. The measured yield of the 16.6 MeV gamma ray is consistent with the predicted yield. The observation of this capture gamma ray establishes the spectroscopy of the fusion gamma rays from the D-³He reactions as a viable diagnostic of total fusion reaction rates and benchmarks the modeling for extension of the technique to D-T plasmas.     

Neutron fluence rate measurement using prompt gamma rays

A gamma ray spectrometer, with a 3(') ? X 3(') NaI(Tl) detector, with a moderator sphere has been utilised to measure the neutron fluence rate, with this value the H(10) was estimated. When a neutron is captured by the hydrogen-based moderator, a 2.22 MeV prompt gamma ray is produced. In a multichannel analyser the net area under the 2.22 MeV photopeak is proportional to the total neutron fluence rate. The features of this system were determined by a Monte Carlo study that includes 3-, 5- and 10-inches diameter, water and polyethylene moderators and a (239)Pu-Be source. The prompt gamma response was extended to monoenergetic neutron sources. To verify the response, a (239)Pu-Be source in combination with a 10(') polyethylene sphere having a gamma-ray spectrometer with NaI(Tl) was utilised to estimate the neutron fluence rate and the H(10). These results were compared with neutron fluence rate and H(10) obtained using a Bonner sphere spectrometer and with the H(10) measured using a neutron remmeter.     

Evaluation of gamma and neutron irradiation effects on the properties of mica film capacitors

We present an investigation of gamma and neutron radiation effects on mica film capacitors from an electrical point of view. We have studied quantitatively the effects of gamma and neutron irradiation on mica film capacitors of thickness, 20 and 40 μm (0.7874 and 1.5748 mil) with two different areas, 01 and 04 cm². The capacitance has been measured at room temperature in the frequency range 100 Hz-10 MHz. Negligible change in the capacitance due to high gamma dose of⁶⁰Co, 15 kGy at dose rate 0.25 kGy/h, has been observed. However, appreciable change in the capacitance has been observed due to low doses of fast neutrons (cumulative dose, 115 cGy) with flux ∼ 9.925 x 10⁷ neutrons/cm² h from²⁵²Cf neutron source of fluence, 2.5 × 10⁷ neutrons/s. We have also observed that the impact of gamma and neutron irradiation is more at frequencies higher than 10 kHz. These results show that the mica capacitors do not show any radiation response below 10 kHz. The study shows the radiation response of mica film capacitors to gamma and fast neutron radiations. Mica capacitors show low gamma radiation response in comparison to fast neutron radiation, because a total dose of kGy order has been given by gamma source and only few cGy dose has been given by fast neutron source.     

Hot wire chemical vapour deposition (HWCVD) of boron carbide thin films from ortho-carborane for neutron detection application

Detection of neutrons is possible if suitable converters such as Li, LiF or ¹⁰B in the form of thin films are used along with the semiconductor device. The use of boron (¹⁰B) in some host matrix as a neutron detector is attractive due to its large neutron capture cross-section. Boron carbide (BC) films are deposited on silicon substrates by HWCVD technique using solid ortho-carborane (o-C₂B₁₀H₁₂) precursor with argon as carrier gas. The films contain ¹⁰B required for neutron detection as confirmed by the Secondary Ion Mass Spectroscopy. Variations in its structure as well as the chemical bonding configurations using Fourier Transform Infra-Red, Raman and X-ray diffraction spectroscopy have been studied.