252Cf Neutron Source Research Articles

DEVELOPMENT OF AN EXTENDED-RANGE BONNER SPHERE SPECTROMETER FOR CHINA INITIATIVE ACCELERATOR-DRIVEN SYSTEM.

To measure the extended-range neutron spectra and calibrate the extended-range neutron dosemeters of the China initiative Accelerator-Driven System (CiADS), an Extended-range Bonner Sphere Spectrometer (EBSS) has been developed. The EBSS was designed based on the PHITS codes, investigating various combinations of materials and diameters of the neutron moderators and the neutron multipliers for extended-range neutrons. Finally, seven polyethylene-only spheres and seven extended-range spheres were selected and subsequently built. The neutron multipliers of the extended-range spheres embedded concentric shells of lead, copper and tungsten. The response functions of the EBSS were analyzed and experimentally validated. It was subsequently tested with 252Cf neutron source and cosmic ray neutron source. The results demonstrate that the EBSS is capable of accurately measuring neutron spectra.

Measuring and unfolding fast neutron spectra using solution-grown trans-stilbene scintillation detector

We propose an overall procedure for measuring and unfolding fast neutron spectra using a trans-stilbene scintillation detector. Detector characterization was described, including the information on energy calibration, detector resolution, and nonproportionality response. The digital charge comparison method was used for the investigation of neutron-gamma Pulse Shape Discrimination (PSD). A pair of values of 600 ns pulse width and 24 ns delay time was found as the optimized conditions for PSD. A fitting technique was introduced to increase the trans-stilbene Proton Response Function (PRF) by 28% based on comparison of the simulated and experimental electron-equivalent distributions by the Cf-252 source. The detector response matrix was constructed by Monte-Carlo simulation and the spectrum unfolding was implemented using the iterative Bayesian method. The unfolding of simulated and measured spectra of Cf-252 and AmBe neutron sources indicates reliable, stable and no-bias results. The unfolding technique was also validated by the measured cosmic-ray induced neutron flux. Our approach is promising for fast neutron detection and spectroscopy.

Study on the Possibility of Neutron Gamma Discrimination in GAGG Crystal

The scintillators capable of neutron detection are being the research hotspot recently. In this manuscript, we investigated the possibility of cerium-doped gadolinium aluminium gallium garnet (GAGG: Ce) crystal to achieve neutron gamma discrimination through pulse shape discrimination (PSD) method. PSD is a method based on the principle that the waveforms of scintillation light differ for different particles. The experiments show that it’s hard for single GAGG crystal to distinguish the neutron from the gamma-ray. Therefore, a composite crystal made of gamma-sensitive crystal and GAGG crystal was furtherly developed. Using the 252Cf neutron source, the composite is proven to be capable of discriminating the neutrons from the gamma-ray and the corresponding FOM is 3.33.

Measurement of prompt gamma production from neutron capture on manganese

Prompt gamma radiation is an important part of radiation fields with neutrons. Its good description is vital for precise predictions in void swelling issues because it is a contributor in gamma heating effect. The energy of prompt gamma rays is unique for each nuclide, thus can be used as its signature. Prompt gamma activation analysis is a technique using these gammas in nuclide identification. Despite the prompt gamma importance, current nuclear data libraries describe its origins very inaccurately. Measurement of gamma leakage spectrum of manganese bath with 252Cf neutron source was performed using well-defined HPGe detector with high-density polyethylene shielding and stilbene scintillator. Stilbene measured gamma/neutron spectrum was separated by a Pulse Shape Discrimination method. MCNP6.2 simulation of the experiment was performed with ENDF/B-VIII.0 and JEFF-3.3 libraries for neutrons (and gamma production from neutron interactions) and MCPLIB04 library for photons. Analysis of results revealed that ENDF/B-VIII.0 and JEFF-3.3 description of prompt gamma radiation production is significantly discrepant. INDEN updated data for gamma production on 55Mn was tested as well and show significant improvement compared to original ENDF/B-VIII.0 data. Nevertheless, significant discrepancies up to 50% were still observed. An alternative approach using prompt gamma production tables provided by IAEA was tested as well. Its results are basically in agreement with the calculation using INDEN updated 55Mn data.

Optimization study of chlorine detection sensitivity in concrete based on prompt gamma analysis using 252Cf neutron source

Chloride attack is a serious problem that decreases the durability of concrete structures of bridges and highways. A compact neutron salt meter with a252Cf neutron source and germanium (Ge) gamma-ray detector based on prompt gamma-ray neutron activation analysis (PGNAA) has been proposed to determine the chlorine concentration in concrete structures. The Optimization of the dimensions of its components, such as polyethylene (PE) moderator, graphite reflector, and lead shield, as well as the positions of the 252Cf source and the Ge detector has been performed to make it highly sensitive for the detection of gamma-rays of chlorine, in addition to lightweight and small volume for in situ use. The results demonstrated that gain factors of 2.5 and 2.2 were obtained for gamma-ray intensity of chlorine and chlorine-to-hydrogen ratio (CHR), respectively, whereas the weight and volume became 19.1% and 23.4%, respectively, compared with the reference setup. The effectiveness of optimization was confirmed by preliminary experiments.

Experimental Verification of Proposed Internal Remote Monitoring System for Spent Nuclear Fuel Dry Cask Storage Applications

It has become a common practice to store sufficiently cooled spent nuclear fuel (SNF) assemblies in interim storage dry casks with passive cooling. These dry casks require nuclear safeguards monitoring because they contain plutonium. Past studies on dry cask modeling and simulations have shown that a remote monitoring system (RMS) situated inside the dry cask could continually monitor and detect the removal of even a single SNF assembly from the cask. This conceptual RMS design was tested by conducting laboratory-scale experiments using small-size 252Cf neutron sources. These small-size sources were surrounded by neutron-reflecting materials in the experiments to mimic the SNF assemblies as a surface neutron source to the fission chamber detectors of the RMS. Experimental and simulation results showed that the removal or diversion of even a single neutron source is detectable within 4 min with a probability of detection greater than 80%.

A backing detector for order-keV neutrons

We have designed and tested a large-area (0.15 m2) neutron detector based on neutron capture on 6Li. The neutron detector design has been optimized for the purpose of tagging the scattering angle of keV-scale neutrons. These neutron detectors would be employed to calibrate the low-energy (<100 eV) nuclear recoil in detectors for dark matter and coherent elastic neutrino nucleus scattering (CEνNS). We describe the design, construction, and characterization of a prototype. The prototype is designed to have a tagging efficiency of ∼25% at the relevant O(keV) neutron energies, and with a mean capture time of ∼17μs. The prototype was characterized using a 252Cf neutron source and agreement with the simulation was observed within a few percent level.

Neutron yield measurement system of HL-2A tokamak

This research presents the development of HL-2A neutron yield measurement which includes 235U fission chamber and BF3 and 3He proportional counters. Equivalent noise formula of the radiation detection signal amplification system was derived to guide the development of the signal amplification system. Then all detectors were calibrated in situ by using the 252Cf neutron source. The neutron yield of the HL-2A during neutral beam heating was analyzed. These results indicate that the developed neutron flux diagnostic system can obtain neutron yield results under various experimental conditions of the HL-2A tokamak, and can provide information on neutron yield.

Prompt gamma-ray methods for industrial process evaluation: A simulation study

Abstract Radioisotope applications in industrial process inspection and evaluation using gamma-ray emitters provide otherwise unavailable information. Offering alternative gamma-ray sources can support the technology by complementing sources’ availability and radiation safety. This work proposes to replace gamma-ray from radioisotopes with prompt gamma-ray from the interaction of neutrons with stable isotopes injected into the industrial process or with the structural material of the industrial process equipment. Monte Carlo N-Particle Transport Code (MCNP5) was used to simulate the irradiation of two-phase flow pipes by 252Cf neutron source. Two simulations were run for each pipe, with and without mixing the liquid phase with the stable isotope 157Gd. The detected gamma-ray spectra were analysed, and images of the two phases inside the pipes were produced. The images were compared to images obtained from simulations of gamma transmission measurement using 60Co. Furthermore, results for prompt gamma computed tomography (CT) were presented and discussed. The studies’ outcomes indicate the potential of prompt gamma-ray to carry out the sealed sources applications of gamma transmission measurements and imaging.

Effect of Using Different Data Libraries and Simulation Codes on the Calculation of Spectra and Operational Quantities for the D2O-252Cf Source at PTB

The neutron reference field produced by a heavy-water moderated 252Cf source is used at PTB for calibrating neutron-measuring devices. Knowledge of the precise neutron spectrum is very important for the investigation of operational radiation protection quantities. Recently, new fission spectrum data of 252Cf has been proposed based on the latest nuclear data library version. At PTB, earlier calculations for the D2O moderated 252Cf neutron source were carried out more than 20 years ago, thus updated and more detailed calculations are required. In this paper, a detailed simulation model of the PTB moderated 252Cf source assembly has been prepared and investigated using new spectral data and two different Monte Carlo transport codes MCNP6.1 and PHITS3.22, with ENDF/BVIII.0, ENDF/B-VII.1, ENDF/B-V, and ENDL85 evaluated nuclear data libraries. The results show that the evaluated nuclear data libraries influence the calculated operational quantities by (3-5) %. The dosimetric quantities calculated with the PHITS code and the ENDF/B-VII.1 data library agree well with the MCNP6 results.

Simulation and prediction of the attenuation behaviour of the KNN–LMN–based lead-free ceramics by FLUKA code and artificial neural network (ANN)–based algorithm

ABSTRACT The significance and novelty of the present work are the preparation of the non-lead ceramic by the general formula of (1-x) K0.5Na0.5NbO3-xLa Mn0.5Ni0.5O3 (KNN–LMN) with different x (0<x<0.295) (mol) to examine the shielding qualities of KNN–LMN non-lead ceramic via FLUKA code. The obtained results via FLUKA code are used as input data for training the artificial neural network algorithm and prediction. For this purpose, the radio isotopic 252Cf neutron source is simulated by the HI-PROBE, RADDECAY and DCYSCORE cards using the FLUKA code. As a result, the neutron–gamma photon shielding ability of the KNN–LMN lead-free ceramics exposed to the 252Cf neutron source is estimated and predicted. Findings show that by increasing the concentration of the x in (1-x) K0.5Na0.5NbO3-xLa Mn0.5Ni0.5O3 lead-free ceramics results in an ascending trend in density. In addition, the increment of the x rate (x refers to the concentration of La Mn0.5Ni0.5O3 in KNN–LMN non-lead ceramics) causes an increase in the value of the neutron attenuation parameter (), and a strong relationship is monitored between and density. Moreover, descending order of is reported for half-value layer values against gamma photon. From the attained results, it can be concluded that increaisng the rate of x results in the better shielding proficiency in terms of neutron and gamma photon for chosen KNN–LMN–based lead-free ceramics.

Degradation of silicon detectors under long-term irradiation by 252Cf fission products

The response function of the recoil nuclei in detectors designed for detection of neutrinos or dark matter particles can be determined only through usage of a neutron source with a known energy spectrum. A possible solution for a compact neutron calibration source is a combination of a 252Cf neutron source and a semiconductor detector that detects fission fragments, and thus records the neutron emission moment. This work is devoted to the degradation study of the operating parameters for silicon semiconductor detectors irradiated by fission fragments of the nuclide of 252 Cf. Two types of Si detectors were under investigations - silicon-lithium Si(Li) p-i-n detectors and silicon surface barrier detectors. As a result of the measurements, the maximum permissible radiation doses for the correct operation of both types of detectors and the relation of the received radiation dose to the spectroscopic characteristics of the detectors were determined.

An extensive study on the neutron-gamma shielding and mass stopping power of (70-x) CRT–30K2O–xBaO glass system for 252Cf neutron source

ABSTRACT The present study focuses on the charged-uncharged particles shielding performance of the addition of a mixed type of cathode ray tube (CRT) in a glass system that is irradiated by the 252Cf neutron source via the MCNPX simulation and analytical calculations, as well as Phy-X: PSD and SRIM software. The CRT waste glass is inserted into the glass system with (70-x) CRT–30K2O–xBaO general formula for x = 0, 10, 20 mol% that produces CG1, CG2, and CG3 glass shielding materials. Using Watt Fission Distribution (WFD) and Doppler Effect (DE) the neutron-gamma photon spectra were extracted for shielded (in the presence of the glass materials) and unshielded (in air) cases. Some calculated attenuation parameters related to the neutron deduced that CG1 is the best neutron attenuator among the selected glass samples. Moreover, by increasing the density of the glass from CG1 to CG3, the ascending trend is observed for the linear attenuation coefficient (LAC, cm−1) of the studied glass, and the best shielding competence is monitored for CG3. Furthermore, two sharp peaks are found in Zeff graphs which may be due to K-edge absorption of Ba and Pb elements and by decreasing the Pb element from CG1 to CG3 the second peak gradually becomes smooth. In addition, Mass Stopping Power/ Projected Ranges of the proton (H1) and alpha particles (He+2) are also estimated by SRIM code and findings show that CG1 can better stop proton and alpha particles in comparison with the other chosen glass structures.

Gamma-blindness & neutron detection efficiency assessments with centrifugally tensioned metastable fluid detectors in low-scatter environment

A range of tests and associated analyses were performed in a state-of-art low scatter environment to demonstrate and validate the performance of the (Medium) Centrifugally Tensioned Metastable Fluid Detector (MCTMFD) for gamma blind detection of fast and epithermal neutrons for a variety of configurations, using Cf-252 and Am-Li isotope neutron sources. All measurements were performed in the Low Scatter Room (LSR) of Radiological Exposures & Metrology Laboratory at Pacific Northwest National Laboratory (PNNL). Experiments were performed to evaluate intrinsic neutron detection efficiency, ɛint n, the intrinsic gamma-neutron detection efficiency, ɛintγn, and the gamma absolute rejection ratio for neutrons (GARRn) as a function of tensioned metastable state negative pressure, Pneg, varying from 2.5 bar towards 9 bar. The specific embodiment of the MCTMFD utilized for the experiments at PNNL included a 16 cm 3 sensitive volume configured for either fast neutron detection using Decafluoropentane (DFP) as the detection medium, or fast and thermal neutron detection (using DFP with a chemical mixture composed of natural boron). Experiments were performed with Cf-252 neutron sources ranging in intensity from 1 × 104 n/s to 3.45 × 105 n/s, an Am-Li neutron source with intensity of 2.5 × 104 n/s, used in both a bare configuration and in combination with a 30 cm OD moderating sphere of D2O, with and without a 0.5 mm Cd shell, and intense Cs-137 and Co-60 gamma sources offering dose rates ranging from 27 mR/h to 53 R/h. It was found that the MCTMFDs remained gamma-blind (ɛintγn < 10−11) even under 53 R/h gamma background; the associated intrinsic fast neutron detection efficiency (ɛint n) varied from ∼0.2% for Pneg ∼2.5 bar, ∼10% for Pneg ∼5.5 bar and, ∼30% for Pneg ∼9 bar for a Cf-252 fission spectrum. In comparison, for the softer (Am-Li) spectrum, the ɛint n varied from ∼1% for Pneg ∼5.5 bar and approached the ɛint n values for Pneg states above 7 bar. The same MCTMFD when using a borated detection fluid with 20% B-10, offered a ×10–20 improvement in ɛint n for Pneg states ranging from 4.5–5 bar, for detecting epithermal neutrons from the Am-Li source (placed within a 30 cm OD, D2O moderator with and without a 0.5 mm Cd shield).

Conceptual study of salt-meter with 252Cf neutron source for on-site inspection of bridge structure

We studied the feasibility of a portable salt-meter incorporating a prompt gamma neutron activation analysis with a californium-252 neutron source to meet urgent demands for non-destructive methods without pre-processing in large bulk materials such as concrete structures. This technique is aiming at the chloride ion concentration in concrete structures with a depth profile from the surface to steel bar. From the portability point of view, a californium-252 neutron source is adopted and we have performed a preliminary experiment and simulations with a simple geometry. In this paper, we describe the performances of salt detection sensitivities by applying the prompt gamma neutron activation analysis.

Validation of the neutron lead transport for fusion applications

Lead is an important material, both for fusion or fission reactors. The cross sections of natural lead should be validated because lead is a main component of lithium-lead modules suggested for fusion power plants and it directly affects the crucial variable, tritium breeding ratio. The presented study discusses a validation of the lead transport libraries by dint of the activation of carefully selected activation samples. The high emission standard 252Cf neutron source was used as a neutron source for the presented validation experiment. In the irradiation setup, the samples were placed behind 5 and 10 cm of the lead material. Samples were measured using a gamma spectrometry to infer the reaction rate and compared with MCNP6 calculations using ENDF/B-VIII.0 lead cross sections. The experiment used validated IRDFF-II dosimetric reactions to validate lead cross sections, namely 197Au(n, 2n)196Au, 58Ni(n,p)58Co, 93Nb(n, 2n)92mNb, 115In(n,n')115mIn, 115In(n,γ)116mIn, 197Au(n,γ)198Au and 63Cu(n,γ)64Cu reactions. The threshold reactions agree reasonably with calculations; however, the experimental data suggests a higher thermal neutron flux behind lead bricks. The paper also suggests 252Cf isotropic source as a valuable tool for validation of some cross-sections important for fusion applications, i.e. reactions on structural materials, e.g. Cu, Pb, etc.

Evaluation of CLYC-6 and CLYC-7 scintillators for detection of nuclear materials

The Cs2LiYCl6 (CLYC) scintillator is proven to be suitable for dual γ-ray/neutron detection owing to its distinct response to both radiation types. This feature makes this material attractive for applications in nuclear security and safeguards. Two 3.8 cm diameter CLYC crystals, one enriched with 95% 6Li, and the second one depleted of 6Li and with >99% 7Li, were characterized with conventional gamma-ray sources as well as with (α, n) and a bare and heavy-water-moderated 252Cf neutron source. The suitability of the two scintillators for detection and characterization of special fissionable materials was evaluated experimentally using uranium and plutonium samples with different isotopic concentrations. Whereas the two detectors have comparable performances for gamma rays, 6Li-enriched CLYC shows better response for neutron gross count, because of its sensitivity to thermal neutrons. A simple method based on neutron/gamma ratios within certain energy regions has been demonstrated to be able to distinguish between different neutron emitting sources, showing very promising results for the development of more sophisticated identification algorithms based on neutron/gamma signatures. The results reported here suggest the 6Li-enriched one as a better candidate for nuclear security applications.

A Monte-Carlo study on the fluorescent nuclear track detector (FNTD) response to fast neutrons: Which information can be obtained by single layer and 3D track reconstruction analyses?

Fluorescence Nuclear Track Detectors (FNTDs) are part of a new technology developed for particle detection and applicable to personal neutron dosimetry. The objective of this study is to simulate the FNTD fast neutron response to: (i) assess and understand the performance of the existing neutron dosimeter design (Landauer Inc.) and its associated single layer track-spots analysis; and (ii) evaluate the potential information that can be obtained by the analysis of the 3D reconstructed recoil proton trajectories. To achieve that, a FLUKA Monte Carlo (MC) model of the current FNTD design was developed and the FNTD response was investigated for mono-energetic neutrons and the 252Cf and 241AmBe neutron sources. The investigation of the recoil proton densities behind the different converters showed that the single layer analysis and dose calculation algorithm, based on the comparison and subtraction of the track densities behind the different converters, works properly only up to neutron energies ~13 MeV. Above this neutron energies, recoil protons generated in the detector housing (PE) have a range larger than the thickness of the PTFE and 6Li-enriched glass, reaching the FTND and, therefore, adding to the signal in these detection regions and influencing the secondary electron discrimination and the energy determination algorithm. MC simulations show that the FNTD 3D reconstructed recoil proton tracks can provide estimates of the irradiation angles and average neutron energy. The results show that the angle or displacement (dX/dZ or dY/dZ) distributions of the recoil proton tracks can be used to obtain information on irradiation angle; the angle with the detector's normal (polar angle), the most important because of its influences on the FTND sensitivity, can be determined in laboratory and for irradiation angles < 60° with an 4° uncertainty already for doses > 4.5 mSv in the case of a214AmBe neutron irradiation. The neutron field mean energy can also be determined for normal irradiation by analysing the depth distribution of the recoil proton tracks already for a minimum of 150 tracks, or 2.5 mSv for 241AmBe, assuming a scanned area is ~2.0 mm2. Therefore, the present study contributes to understanding the performance of the current FNTD design and analysis for neutron dosimetry and investigates a new detector evaluation approach to gain additional information on the irradiation conditions.

Production of actinium-225 from a (n,p) reaction: Feasibility and pre-design studies

Abstract Actinium-225 is used in nuclear medicine for the treatment of malignant tumours. It can be applied to produce Bi-213 in a reusable generator or can be used alone as an agent for radiation therapy, in particular for targeted alpha therapy. However, the availability of Ac-225 for worldwide use, particularly in low- and middle-income countries, is limited. We present a feasibility study employing GATE, an open-source Monte Carlo simulation toolkit, on the production of Ac-225 from a neutron generator. This work suggests that a design consisting of three concentric cylinders, the innermost a Cf-252 neutron source, the middle nickel cylinder acting as a proton-producing target and the outer cylinder a RaCl2 target may provide a feasible design outline for an Ac-225 generator.

Real-time source localisation by passive, fast-neutron time-of-flight with organic scintillators for facility-installed applications

Fast neutron time-of-flight (ToF) has been used to characterise the location of a source of a mixed radiation field. Two EJ-309 organic scintillators and a fast, digital, data acquisition system have been used in a variety of positions to identify the location of a 252Cf neutron source inside a steel, water-filled tank. A methodology for extracting the distance between the neutron source and the neutron detector has been developed and verified with MCNP simulations. A reconstruction algorithm using the ToF data has been developed. The location of the neutron source has been estimated on this basis to be within 20 cm of its known location with a spatial resolution of ±7.8 cm. p-values extracted from the null test hypothesis have been estimated to be 0.975 and 0.996 for experimental and simulation data, respectively. By correctly identifying the location of the source, the potential for the system to discern between scattered and unscattered neutrons is demonstrated.