Spontaneous Fission Source Research Articles

Use of a Deuterium-Deuterium Neutron Generator for the Add-A-Source Waste Matrix Correction Technique

Neutron generators offer an alternative to radioactive sources in the application of active techniques for nondestructive assay of nuclear materials. The Add-A-Source technique has enabled improvements in the ability to accurately measure waste bearing nuclear materials by measuring and correcting for the non-nuclear matrix effect. The technique has historically been performed with a Cf-252 spontaneous fission source. In this work we have demonstrated the use of a Deuterium-deuterium neutron generator in place of the Cf-252 source. A DD generator enables operational advantages and allows the simplification of the instrument design, while offering equivalent corrective performance as a Cf-252 source.

Development of a method for measuring the effective fraction of delayed neutrons using a 240Pu spontaneous fission source

Development of a method for measuring the effective fraction of delayed neutrons using a 240Pu spontaneous fission source

Multiplicity counting using organic scintillators to distinguish neutron sources: An advanced teaching laboratory

In this advanced instructional laboratory, students explore complex detection systems and nondestructive assay techniques used in the field of nuclear physics. After setting up and calibrating a neutron detection system, students carry out timing and energy deposition analyses of radiation signals. Through the timing of prompt fission neutron signals, multiplicity counting is used to carry out a special nuclear material (SNM) nondestructive assay. Our experimental setup is comprised of eight trans-stilbene organic scintillation detectors in a well-counter configuration, and measurements are taken on a spontaneous fission source as well as two (α,n) sources. By comparing each source's measured multiplicity distribution, the resulting measurements of the (α,n) sources can be distinguished from that of the spontaneous fission source. Such comparisons prevent the spoofing, i.e., intentional imitation, of a fission source by an (α,n) neutron source. This instructional laboratory is designed for nuclear engineering and physics students interested in organic scintillators, neutron sources, and nonproliferation radiation measurement techniques.

Modelling of Tarapur BWR lattice and core to estimate the required number of SB-BE startup neutron sources

During the refuelling and criticality of Boiling Water Reactors (BWR), fission counters are used as Source Range neutron Monitors (SRMs) to get the count rates. The study shows that the neutron counts due to spontaneous fission sources are insufficient to meet the SRM's limitations of 3 counts per second (cps). In this work, TAPS fuel assembly is modelled using the lattice transport code DRAGON5, and lattice parameters are generated as a function of fuel depletion. These parameters have been further used to model the reload-2 of the BWR core. 3D diffusion theory code KINFIN is used to simulate the BWR core with external neutron sources to estimate the neutron flux at three SRM locations. Thermal fluxes are then used to compute the neutron count rates using the sensitivity of SRMs for different combinations of five Sb-Be neutron sources.

Instrumentation for correlated prompt [formula omitted] emission studies in coincidence with fission fragments

Recent theoretical and experimental results have brought renewed interest and focus on the topic of fission fragment angular momentum. Measurements of neutrons and γ rays in coincidence with fission fragments remain the most valuable tool in the exploration of fission physics. To achieve these scientific goals, we have developed a system that combines a state-of-the-art fission fragment detector and n−γ radiation detectors. A new twin Frisch-gridded ionization chamber has been designed and constructed for use with a spontaneous fission source and an array of forty trans-stilbene organic scintillators (FS-3) at Argonne National Laboratory. The new ionization chamber design we present in this work aims at minimizing particle attenuation in the chamber walls, and provides a compact apparatus that can be fit inside existing experimental systems. The ionization chamber is capable of measuring fission fragment masses and kinetic energies, whereas the FS-3 provides neutron and γ-ray multiplicities and spectra. The details of both detector assemblies are presented along with the first experimental results of this setup. Planned event-by-event analysis and future experiments are briefly discussed.

The FFIS spectrometer for determination of fission fragment mass distribution with the energy–velocity method

The FFIS spectrometer (Fission Fragment Identification Spectrometer) for determining independent fission yields has been developed based on the E–v method. By directly measuring the time-of-flight (TOF) and kinetic energy of fission fragments on their flight path, the post-neutron-emission mass yield distributions can be obtained. With the aim of getting the highest possible mass resolution, an axial grid ionization chamber and a TOF measurement system based on microchannel plate (MCP) detectors were designed. The ionization chamber with a silicon nitride entrance window of 100 nm thickness was tested and calibrated with a 252Cf spontaneous fission source. The intrinsic energy resolution for 1.27 MeV/u 63Cu particles is measured to be 450 keV (FWHM). The time resolution was determined with two MCP timing detectors to be 157 ps (FWHM) for 241Am radiation source, implying that the intrinsic time resolution is about 110 ps (FWHM) for a single timing detector. Thermal-neutron-induced fission of 235U was studied at the In-hospital Neutron Irradiator (IHNI). An event-by-event energy loss addback technique was developed based on Geant4 calculation to minimize the correction error. The first results of mass yield distributions from 235U(nth,f) are reported.

Special Nuclear Material Identification Through One-Minute Measurement with a New Backpack Radiation Device in Real Scenario Conditions

The constant concerns in global nuclear safety, aimed at deterring and combating the illicit trafficking of Material Out of Regulatory Control (MORC) and its possible use in criminal acts has raised the necessity of new detection solutions with higher efficiency and resolution to provide a high level of accuracy in the report to the authorities. Today’s radioactive isotope identifiers perform gamma spectroscopy identification and, sometimes, neutron counting. This paper presents the results of measurements performed in real-scenario conditions with a new type of portable radioactive isotope identifier for the detection and identification of both gamma and neutron sources. Its singular features are: 1) the capability to identify sources through the detection of neutrons, discriminating spontaneous fission sources (Cf-252), α-n sources (Am/Be, Am/Li) and nuclear material containing mix of isotopes of plutonium or uranium 2) the capability to make cross correlation between gamma and neutron measurements to achieve a higher level of accuracy in the identification of SNM that emits both neutrons and characteristics gammas. The test results are compared with international standards. The device exceeds the standard performance by triggering a neutron alarm for a 20.000 n/s Cf-252 source at a five times greater distance than the ANSI N42.34 one.

Application of an added-sinusoid, signal-multiplexing scheme to a compact, multiplexed neutron scatter camera

The development of compact and portable neutron imagers could be transformative for nuclear nonproliferation and security. Of the currently feasible neutron imaging concepts, the neutron scatter camera is the one that can be made most compact and versatile. Beneficially, this design also inherently measures the energy of incident neutrons on an event-by-event basis. Single volume and quasi-single volume neutron scatters have been proposed to fully maximize the potential of neutron scatter cameras and enable truly man-portable neutron imagers. Such imagers will likely have a very large number of readout channels to maximize the efficiency, so signal multiplexing to reduce readout channels would significantly reduce costs and size and facilitate wide-spread deployment. In this work a multiplexing method based on adding well-defined sinusoids to detector signal is applied to a quasi-single volume neutron scatter camera comprised of 6-mm × 6-mm × 60-mm EJ-299-34 plastic scintillators coupled to arrays of SensL 60035 C-series silicon photomultipliers to create a multiplexed neutron imager (MiNI). Such straightforward multiplexing enables the use of more detector elements for a given number of available signal readouts. The necessary calibrations and system characterizations are carried out to assess the multiplexing method’s performance and to understand the behavior of the prototype MiNI. Its ability to successfully image a neutron source is demonstrated with a 252Cf spontaneous fission source placed in various locations around the MiNI. Finally, neutron spectroscopy is also performed, and the feasibility and benefits of multiplexing a quasi-single volume neutron scatter camera are demonstrated.

The influence of physical parameters on the in-situ metal carbonyl complex formation studied with the Fast On-line Reaction Apparatus (FORA)

Abstract The Fast On-line Reaction Apparatus (FORA) was used to investigate the influence of various reaction parameters onto the formation and transport of metal carbonyl complexes (MCCs) under single-atom chemistry conditions. FORA is based on a 252Cf-source producing short-lived Mo, Tc, Ru and Rh isotopes. Those are recoiling from the spontaneous fission source into a reaction chamber flushed with a gas-mixture containing CO. Upon contact with CO, fission products form volatile MCCs which are further transported by the gas stream to the detection setup, consisting of a charcoal trap mounted in front of a HPGe γ-detector. Depending on the reaction conditions, MCCs are formed and transported with different efficiencies. Using this setup, the impact of varying physical parameters like gas flow, gas pressure, kinetic energy of fission products upon entering the reaction chamber and temperature of the reaction chamber on the formation and transport yields of MCCs was investigated. Using a setup similar to FORA called Miss Piggy, various gas mixtures of CO with a selection of noble gases, as well as N2 and H2, were investigated with respect to their effect onto MCC formation and transport. Based on this measurements, optimized reaction conditions to maximize the synthesis and transport of MCCs are suggested. Explanations for the observed results supported by simulations are suggested as well.

Investigations on neutron attenuation properties of poly-boron materials using Am-Be and 252Cf sources neutron spectra

The neutron attenuation characteristics of polyboron materials was studied using Am-Be (α,n) and 252Cf (Spontaneous Fission) sources. Neutron Attenuation Factor (NAF) is obtained for poly boron materials with addition of different concentrations of boron by neutron transmission experiments. Neutron transport calculations are carried by simulating the transmission of neutron from Am-Be and 252Cf sources through polyboron materials using Monte Carlo based simulations. The experimental results are fairly in good agreement with transport calculations. The results showed that Neutron Attenuation Factor (NAF) mainly depend on the incident neutron energy and boron number density present in HDPE shielding material. The effect of boron in HDPE on neutron shielding is only marginal (2%) because of low value of neutron cross section at fast energy region. However, in the low neutron energy region, it varies with wt % of boron and 20% boron as additive in HPDE is an optimum value for neutron attenuation. The details of experimental observations and theoretical investigations are discussed in this work.

Simulations of the source-range detector response for the fuel-loading process of the AP1000 cores

In this paper, the ex-core source-range detector response was simulated for AP1000 cores in its fuel-loading process by using a Monte Carlo code. Firstly, the current sensitivity coefficients of the ex-core source-range detectors have been calibrated by utilizing the experiments with single fuel assembly and the primary source loaded in the core. Secondly, both relative and absolute calibration methods have been applied and analyzed for the sensitivity coefficients calibration. Thirdly, the simulation values for the ex-core source-range response with fully loaded fuel have been validated by using the actual measurement of the first AP1000 cores in the world. Fourthly, the contribution of the 238U spontaneous fission to the ex-core source-range detector response has been studied. Through the numerical simulation results, the following observations can be obtained: 1) The relative calibration method is better than the absolute calibration method, especially when the ex-core source-range response is small. 2) Compared with the measurements, the relative deviation between the simulated and the measured values is within 14% for the two AP1000 cores at Sanmen Nuclear Power Plant. 3) The intensity of 238U spontaneous fission source accounts for 15% of the total source strength in the AP1000 core Unit 1, but the maximum change of response is only 0.02 cps, which is still insignificant for the ex-core detector response simulation in PWR.

Measurement of TOF of fast neutrons with 238U target

We developed a Dual-PPACs detector for fast neutron measurements that consists of two sets of PPAC: conventional PPAC and fission PPAC. A238U(U3O8) coating is placed in the fission PPAC’s anode, which is used as the neutrons conversion layer. An experiment was performed to measure neutron time-of-flight (TOF) in which 252Cf spontaneous fission source was used. An excellent time resolution of 164ps has been observed at 6 mbar in isobutene gas. With the excellent time resolution of Dual-PPACs detector, exact neutron energy can be extracted from the timing measurement. The experimental detection efficiency was 1.9 × 10−7, consistent with the efficiency of 2.5 × 10−7 given by a Geant4 simulation. Ultimately, the results show that the Dual-PPACs detector is a suitable candidate for measuring fast neutrons in the future CiADS system.

Preliminary study of prompt gamma neutron activation analysis for Melamine

Prompt gamma neutron activation (PGNA) technique can be applied for detecting explosives in hidden places by using spontaneous fission source 252Cf source or neutron generator that emits thermal and cold neutrons. Some of emitted neutrons will be captured by nuclei and results in production of prompt gamma rays. Explosive materials can be identified by observing prompt gamma energy spectrum using high purity germanium (HPGe) detector. Gamma energy peaks indicate the presence of corresponding explosive nuclei in scanned materials. This paper describes the experiments conducted to investigate the possibility of explosive detection by using PGNA with 241Am/Be neutron source and limitations that come with it.

Comprehensive Second-Order Adjoint Sensitivity Analysis Methodology (2nd-ASAM) Applied to a Subcritical Experimental Reactor Physics Benchmark: IV. Effects of Imprecisely Known Source Parameters

By applying the Second-Order Adjoint Sensitivity Analysis Methodology (2nd-ASAM) to the polyethylene-reflected plutonium (PERP) benchmark, this work presents results for the first- and second-order sensitivities of this benchmark’s leakage response with respect to the spontaneous fission source parameters. The numerical results obtained for these sensitivities indicate that the 1st-order relative sensitivity of the leakage response to the source parameters for the two fissionable isotopes in the benchmark are all positive, signifying that an increase in the source parameters will cause an increase in the total neutron leakage from the PERP sphere. The 1st- and 2nd-order relative sensitivities with respect to the source parameters for 239Pu are very small (10−4 or less). In contradistinction, the 1st-order and several 2nd-order relative sensitivities of the leakage response with respect to the source parameters of 240Pu are large. Large values (e.g., greater than 1.0) are also displayed by several mixed 2nd-order relative sensitivities of the leakage response with respect to parameters involving the source and: (i) the total cross sections; (ii) the average neutrons per fission; and (iii) the isotopic number densities. On the other hand, the values of the mixed 2nd-order relative sensitivities with respect to parameters involving the source and: (iv) the scattering cross sections; and (v) and the fission cross sections are smaller than 1.0. It is also shown that the effects of the 1st- and 2nd-order sensitivities of the PERP benchmark’s leakage response with respect to the benchmark’s source parameters on the moments (expected value, variance and skewness) of the PERP benchmark’s leakage response distribution are negligibly smaller than the corresponding effects (on the response distribution) stemming from uncertainties in the total, fission and scattering cross sections.

Comparison of the single and double count using MCNP6 and ONMS Geant4 software

The quantification and detection of fissile material are valuable for safeguards, nuclear security, and fuel management. Non-Destructive Assay (NDA) (i.e. using neutron and gamma measurements) provide reliable means to detect and estimate the fissile content (i.e. mass) in a vast range of material in different forms found across various facilities in the nuclear fuel cycle. Neutron measurements are used either for total counting (i.e. pure metal fissile samples) or coincidences/multiplicity counting (i.e. impure fissile samples). These various flavors of neutron measurement are routinely used to quantify fissile material mass and inventory lists. Monte Carlo codes have evolved considerably and are used routinely for modeling and simulation of diverse physics. MCNP6 (Monte Carlo N Particle) and ONMS-Geant4 (Open Neutron Multiplicity Software) are used in this work to generate single and double count (i.e. based on the accidental and real plus accidental histograms) for two fresh fuel sources with the assumption that the neutron emission process is solely due to spontaneous fission. Previous work has shown that for these fresh fuel sources, about 93% of the neutrons emitted is due to spontaneous fission. ONMS was designed to provide an open-source platform that will facilitate researchers in the nuclear community to perform neutron counting simulations. ONMS has been shown to give good results for detectors with high efficiencies, strong (α, n) sources, strong spontaneous fission sources and sources baring plutonium. Detailed model simulations result from ONMS and MCNP6 are compared for a weak californium source and uranium-bearing fresh fuel assembly using a detector set up with relatively low efficiency. The detector response is characterized to compare the single and double counts obtained from both code packages. For the weak californium source, the neutron triplets are also evaluated from both code packages.

A hybrid detector telescope for the simultaneous detection of a wide range of charged particles from Z = 1 to fission fragments

Abstract A hybrid detector telescope has been developed for simultaneous detection of a wide range of charged particles from Z = 1 to fission fragments (FFs). The telescope consists of two detection mediums, namely, gas-ionization chamber and CsI(Tl) crystals, enabling three stages of particle identification (PI). The gas-ionization chamber is segmented longitudinally into two parts in order to separate heavier FFs from other reaction products. The characteristics of the ionization chamber were obtained using 252Cf spontaneous fission source. The PI capabilities of the telescope has been demonstrated through in-beam measurements in 12 , 13 C + 232Th fission reactions. The potential advantage of such a hybrid detector telescope to measure the yield of light charged particles which are emitted along the direction of fission fragments, has been established. The details of the design aspects, its characteristic performance, and the results from in-beam measurements are presented in this article.

Measured neutron light-output response for trans-stilbene and small-molecule organic glass scintillators

Abstract The neutron light-output response from quasi-monoenergetic neutrons was measured for a O5.08 × 5.08 cm trans-stilbene and a O5.08 × 5.08 cm small-molecule organic glass scintillator . Quasi-monoenergetic neutrons were isolated from a time-of-flight measurement of a Cf-252 spontaneous fission source with a flight distance of 200 cm. Two different methods of waveform analysis were implemented, where the neutron light output response proportional to the detected pulse height distribution (PHD) and pulse integral distribution (PID) were extracted for both types of scintillators. The light output response proportional to the pulse integral was determined by integrating the digitized waveform with an integration window length of 150 ns, which contained > 90% and > 95% of the scintillation light for an averaged neutron and photon waveform above 0.5 MeVee. The extracted light-output data were fitted with a semi-empirical function based on the Birks’ formula. The results show that the small-molecule organic glass scintillator produced more light than the trans-stilbene scintillator for a range of neutron energies of 0.79 ± 0.04 MeV to 3.65 ± 0.38 MeV. The fitted semi-empirical function was used in MCNPX-PoliMi with MPPost to simulate the detector response from an independent measurement of a Cf-252 spontaneous fission source to test the fidelity of the extracted light output response functions. The simulated and measured total neutron count rate agreed to within ± 3 % and ± 1% for the trans-stilbene and small-molecule organic glass scintillators. Due to the different light-output response of the two scintillators, the intrinsic neutron detection efficiency was calculated for equal observable ranges in light output and neutron-equivalent energy units. The intrinsic neutron detection efficiency in the observable light-output range of 0.06 MeVee to 2.4 MeVee was calculated to be 28.66 ± 1.43 % and 34.66 ± 1.73 % for pulse height and pulse integral analysis, respectively. For the same observable light-output range, the intrinsic efficiency of small-molecule organic glass was calculated to be 32.54 ± 1.63 % and 37.39 ± 1.87 % for pulse height and pulse integral analysis, respectively. When observing equal light-output ranges, the small-molecule organic glass was 11.92 ± 6.23% and 7.88 ± 7.35% more efficient than the trans-stilbene using pulse height and pulse integral analysis, respectively. When observing equal neutron-equivalent energy ranges, the trans”–stilbene scintillator was 8.4 ± 6.46% and 11.95 ± 6.23% more efficient than the small-molecule organic glass for pulse height and pulse integral analysis, respectively.

Research on the Artificial Neural Network Unfolding Method for the Water-Pumping-Injection Multi-Homocentric Sphere Neutron Spectrometer

This work aimed to study an algorithm for the unfolding spectrum acquired by using a water-pumping-injection multi-homocentric sphere neutron spectrometer. The readings of the water-pumping-injection multi-homocentric sphere neutron spectrometer under 32 different neutron sources, including 241Am-Be spontaneous, 238Pu Watt, and 232Th Maxwell fission, were obtained by using the FLUKA Monte Carlo code and were input as the neural network training values for constructing the method of back-propagation artificial neural network neutron spectrum unfolding (MBANSU). Two neutron readings that were Monte Carlo experimentally measured from 252Cf spontaneous and 245Cm Maxwell fission sources were used as the testing values for MBANSU. The calculated neutron energy spectrum was in good agreement with the ENDF energy spectrum, and the information entropy of the result spectrum approximated that of the ENDF spectrum. Results indicate that the established MBANSU algorithm has potential applications for unfolding spectra collected by using the water-pumping-injection multi-homocentric sphere neutron spectrometer.

Reactions of fission products from a 252Cf source with NO and mixtures of NO and CO in an inert gas

Abstract Fission products recoiling from a 252Cf spontaneous fission source were stopped in various mixtures of inert gases containing CO and NO. For the elements of the transisition metal series Mo, Tc, Ru, and Rh previous observations of pure carbonyl complexes were reproduced. However, no formation of volatile mixed nitrosyl-carbonyl complexes or pure nitrosyl complexes for these elements have been observed. Instead, efficient production of volatile nitrosyl compounds for single iodine atoms, presumably nitrosyl iodide, NOI, was detected. This observation is of interest as potential transport path for iodine in nuclear accident scenarios and as a model for radiochemistry with the recently discovered heaviest halogen tennessine (Z=117).

Computational investigation of arranged scintillating particle composites for fast neutron detection

Composite materials have the potential to play an important role in enabling high-performance and cost-effective neutron detectors for fundamental science, global security, and dosimetry applications. This study presents neutron and gamma transport simulation results on composites consisting of 6Li-containing GS20 scintillator glass particles arranged in an organic matrix. These composites achieve high sensitivity for neutrons while suppressing signals from gamma rays. The simulations successfully identified optimal composite parameters such as particle size, geometry, and inter-particle pitch. This information is key to the subsequent fabrication and scale-up of the desired GS20-in-acrylic composite. The simulation results presented in this paper demonstrate that arranged composites can provide high neutron detection efficiency with gamma misidentification values of <1 in 106 gamma rays, assuming irradiation with a 252Cf spontaneous fission source.