Fission Detectors Research Articles

Estimation and Measurement of Alpha Decay Pulses in Fission Detectors and Their Practical Application for Verifying Detector Health

Estimation and Measurement of Alpha Decay Pulses in Fission Detectors and Their Practical Application for Verifying Detector Health

Recoil Proton Telescopes and Parallel Plate Avalanche Counters for the 235U(n,f) cross section measurement relative to H(n,n)H between 10 and 450 MeV neutron energy

With the aim of measuring the 235U(n,f) cross section at the n_TOF facility at CERN over a wide neutron energy range, a detection system consisting of two fission detectors and three detectors for neutron flux determination was realized. The neutron flux detectors are Recoil Proton Telescopes (RPTs), based on scintillators and solid state detectors, conceived to detect recoil protons from the neutron-proton elastic scattering reaction. This system, along with a fission chamber and an array of parallel plate avalanche counters for fission event detection, was installed for the measurement at the n_TOF facility in 2018, at CERN.An overview of the performances of two RPTs — especially developed for this measurement — and of theparallel plate avalanche counters are described in this article. In particular, the characterizationin terms of detection efficiency by Monte Carlo simulations and response to neutron beam, the studyof the background, dead time correction and characterization of the samples, are reported. Theresults of the present investigation show that the performances of these detectors are suitable foraccurate measurements of fission reaction cross sections in the range from 10 to 450 MeV.

233U(n,γ) measurements at LANSCE

Uranium-233 plays an important role in the Th-U fuel cycle, which has been proposed as an alternative to the U-Pu fuel cycle due to its reduced amount of transuranium elements. The available experimental 233U(n,γ) cross section data in the literature are scarce, [1–3]. In 2008, the 233U(n,γ) cross section was investigated at LANL using the DANCE detector combined with a PPAC, however the statistics in the keV regime were inadequate for a reliable extraction of the cross section at 100 keV. An accurate measurement of the 233U(n,γ) cross section is required by the NCSP to complete the neutron-induced cross section data; a new evaluation reported the need of 233U capture data. The challenge in this measurement lies in the difficulty of measuring capture cross section data due to the competing capture and fission channels. Fission reactions are around one order of magnitude more likely than capture for 233U. The accuracy in the capture cross section measurement relies on the discrimination between the γ-rays produced in capture and fission reactions, for which an experimental setup combining capture and fission detectors is needed. Following this requirement, a new measurement has been performed at LANSCE combining the γ-ray array DANCE with the neutron detector NEUANCE to identify fission and neutron-capture events. This measurement will provide results of the 233U capture-to-fission ratio in the Resolved and Unresolved Resonance regions.

Simulated Performance of the Micro-Pocket Fission Detector in the Advanced Test Reactor Critical Facility

The Micro-Pocket Fission Detector (MPFD) is a small-form-factor real-time fission chamber. MPFD performance has been simulated in the Advanced Test Reactor Critical Facility located at Idaho National Laboratory. The neutron and gamma-ray flux profiles and magnitudes were simulated using MCNP in the near-core B-8 irradiation position. These simulations were performed at 69 discrete axial locations inside the B-8 position 55 for three separate orientations of the nearby hafnium outer shim control cylinders and at a power level of 700 W(thermal). The resulting neutron and gamma-ray flux values were used to determine the MPFD response for various fissile masses and detector gas pressures. The optimal gas-operating pressure was determined to be between 30 and 60 psig. The required fissile layer mass was determined to be between 0.5 to 1.0 µg of 235U. Additionally, the gamma ray to fission fragment interaction rate was determined to be 1.43 × 103 with average energy deposition for gamma rays and fission fragments in 30 psig argon gas to be 1 keV and 3.5 MeV, respectively.

Comparison of Results from Recent NNSA and CEA Measurements of the 239Pu(n, f) Prompt Fission Neutron Spectrum

The National Nuclear Security Administration (NNSA)/DP French Alternative Energies and Atomic Energy Commission (CEA)/DAM agreement on cooperation on fundamental science is a U.S.-French collaborative effort to combine intellectual and experimental resources and further the relevant nuclear science. Recently, both the NNSA and CEA experimental teams performed high-statistics measurements of the 239Pu(n, f) prompt fission neutron spectrum (PFNS) at the Los Alamos Neutron Science Center, both of which were recently published in the journal Physical Review C. These separate measurements used the same experimental area and a common neutron detector array, but differ in many aspects, including background assessments, data acquisition systems and philosophies, fission detectors, and PFNS extraction techniques. Hence, some aspects of the experimental methods and associated uncertainties are highly correlated while others are independent. The results from both measurements broke new ground for PFNS measurements given their higher accuracy and more detailed study of corrections necessary for the measured quantity compared to existing literature measurements, and both will significantly impact PFNS nuclear data evaluations for the foreseeable future. The focus of this work is to document a comparison of the results from these distinct measurements in terms of the acquired data, the PFNS results, and the measured average PFNS energies. While systematic differences between the PFNS results are present on the 1–3% level, the acquired data relative to each respective measurement at low incident neutron energies are in remarkable agreement, as are the conclusions regarding the magnitude and position of features in the PFNS relating to second-chance fission, third-chance fission, and pre-equilibrium neutron emission.

Evaluation and validation of external neutron source strength of operating Fast Breeder Test Reactor

Fast Breeder Test Reactor is one of the high flux operating test facility in India. The study aims to find out the source strength of the neutron source subassembly during its residence time in the reactor core and its contribution to the observed count rates in the startup neutronic channels which are ex-core 235U coated fission detectors. The entire power history during various fuel irradiation campaigns was considered in the calculation of neutron source strength. The study is experimentally validated by shifting the external neutron source subassembly to an elevated position in the subcritical core. The experiment is devised in such a way that inherent neutron source strength is not amplified and the count rates obtained in the detectors will be solely from the external neutron source. The count rates observed during the experiment at the ex-core detectors are compared with the simulated count rates during the lifted position of the source subassembly and found to be in good agreement.

A compact fission detector for fission-tagging neutron capture experiments with radioactive fissile isotopes

In the measurement of neutron capture cross-sections of fissile isotopes, the fission channel is a source of background which can be removed efficiently using the so-called fission-tagging or fission-veto technique. For this purpose a new compact and fast fission chamber has been developed. The design criteria and technical description of the chamber are given within the context of a measurement of the 233U(n, γ) cross-section at the n_TOF facility at CERN, where it was coupled to the n_TOF Total Absorption Calorimeter. For this measurement the fission detector was optimized for time resolution, minimization of material in the neutron beam and for alpha-fission discrimination. The performance of the fission chamber and its application as a fission tagging detector are discussed.

Neutron spectrum unfolding of the multiple activation foils based on sparse representation

An unfolding method based on sparse representation is proposed and applied to unfolding problem of multiple activation foils measurement. The Online Dictionary Learning (ODL) algorithm is applied for sparse representation. The algorithm LARS-EN is used for reconstruction. The Akaike’s Information Criterion (AIC) is applied to select the best coefficient path. Nuclear reactor of Qinshan phase II nuclear power plant (NPP) is considered as the study object and training samples are generated by Monte Carlo code MCNP. By using the unfolding method proposed in this paper, two sets of neutron spectrums are unfolded successfully with acceptable accuracies. The first set contains 4 neutron spectrums of different shapes; and the second set contains 4 neutron spectrums at different positions of the radiation surveillance capsule. Moreover, neutron spectrums of the radiation surveillance capsule can also be unfolded with acceptable accuracies with fewer equations while the two fission detectors 238U and 237Np are removed.

Electrodeposition of low-enriched uranium onto small platinum electrodes

An updated method of preparing thin, uranium-coated, 0.25-mm diameter platinum electrodes using the cyclic potential sweep method is described. Sixty-four samples were prepared using a 0.2 M UO 2 (NO 3 ) 2 solution with 19.695% enriched uranium. Improvements were made to the sample preparation, electrodeposition , and analysis methods, producing smoother samples with more uniform surface features than prior work. Alpha-particle spectrometry was used to determine the mass of uranium deposited on each sample. The average mass of uranium deposited on the metallic electrode from 100 potential sweep cycles was 0.987 × 10 −7 g with a standard deviation of 0.860 × 10 −7 g. Removing the outliers reduces the average deposited mass to 0.741 × 10 −7 g with a standard deviation of 0.410 × 10 −7 g. A smooth, uniform deposition was observed over the entire electrode after deposition and no cracking or crystalline formations were observed. This improved method of electrodeposition remains an attractive method of depositing fissile material onto small metallic substrates for the fabrication of Micro-Pocket Fission Detectors.

Preliminary results on the 233U capture cross section and alpha ratio measured at n_TOF (CERN) with the fission tagging technique

233U is of key importance among the fissile nuclei in the Th-U fuel cycle. A particularity of 233U is its small neutron capture cross-section, which is on average about one order of magnitude lower than the fission cross-section. The accuracy in the measurement of the 233U capture cross-section depends crucially on an efficient capture-fission discrimination, thus a combined set-up of fission and γ-detectors is needed. A measurement of the 233U capture cross-section and capture-to-fission ratio was performed at the CERN n_TOF facility. The Total Absorption Calorimeter (TAC) of n_TOF was employed as γ-detector coupled with a novel compact ionization chamber as fission detector. A brief description of the experimental set-up will be given, and essential parts of the analysis procedure as well as the preliminary response of the set-up to capture are presented and discussed.

Position sensitive twin ionization chamber for nuclear fission investigations

In this article, we report the recent achievements in the design of a back-to-back twin ionization chamber (TIC) for fission fragment (FF) mass, kinetic energy, and orientation. Correlated FF kinetic energies, their masses, and the angle of the fission axes in 3D Cartesian coordinates can be determined from analysis of the heights and shapes of the pulses induced by the FFs on the anodes of the TIC. These anodes were designed to consist of isolated [Formula: see text]-shaped strips connected to nodes of a chain filter made of serially-connected two-port networks. The double charge division method was implemented by digitizing four waveforms at the endpoints of the chain filters. It was shown that the fission fragments emission point on the target plane may be determined using the measured data. A position-sensitive neutron-induced fission detector for neutron imaging applications with both thermal and low energy neutrons was found as another possible implementation of the designed TIC. Preliminary measurements with thermal neutron-induced fission were done with RC chain filters and the results are reported here.

Fabrication and Testing of a Modular Micro-Pocket Fission Detector Instrumentation System for Test Nuclear Reactors

Advancements in nuclear reactor core modeling and computational capability have encouraged further development of in-core neutron sensors. Measurement of the neutron-flux distribution within the reactor core provides a more complete understanding of the operating conditions in the reactor than typical ex-core sensors. Micro-Pocket Fission Detectors have been developed and tested previously but have been limited to single-node operation and have utilized highly specialized designs. The development of a widely deployable, multi-node Micro-Pocket Fission Detector assembly will enhance nuclear research capabilities. A modular, four-node Micro-Pocket Fission Detector array was designed, fabricated, and tested at Kansas State University. The array was constructed from materials that do not significantly perturb the neutron flux in the reactor core. All four sensor nodes were equally spaced axially in the array to span the fuel-region of the reactor core. The array was filled with neon gas, serving as an ionization medium in the small cavities of the Micro-Pocket Fission Detectors. The modular design of the instrument facilitates the testing and deployment of numerous sensor arrays. The unified design drastically improved device ruggedness and simplified construction from previous designs. Five 8-mm penetrations in the upper grid plate of the Kansas State University TRIGA Mk. II research nuclear reactor were utilized to deploy the array between fuel elements in the core. The Micro-Pocket Fission Detector array was coupled to an electronic support system which has been specially developed to support pulse-mode operation. The Micro-Pocket Fission Detector array composed of four sensors was used to monitor local neutron flux at a constant reactor power of 100 kWth at different axial locations simultaneously. The array was positioned at five different radial locations within the core to emulate the deployment of multiple arrays and develop a 2-dimensional measurement of neutron flux in the reactor core.

On Use of Multi-Chambered Fission Detectors for In-Core, Neutron Spectroscopy

Presented is a short, computational study on the potential use of multichambered fission detectors for in-core, neutron spectroscopy. Motivated by the development of very small fission chambers at CEA in France and at Kansas State University in the U.S., it was assumed in this preliminary analysis that devices can be made small enough to avoid flux perturbations and that uncertainties related to measurements can be ignored. It was hypothesized that a sufficient number of chambers with unique reactants can act as a real-time, foilactivation experiment. An unfolding scheme based on maximizing (Shannon) entropy was used to produce a flux spectrum from detector signals that requires no prior information. To test the method, integral, detector responses were generated for singleisotope detectors of various Th, U, Np, Pu, Am, and Cs isotopes using a simplified, pressurized-water reactor spectrum and fluxweighted, microscopic, fission cross sections, in the WIMS-69 multigroup format. An unfolded spectrum was found from subsets of these responses that had a maximum entropy while reproducing the responses considered and summing to one (that is, they were normalized). Several nuclide subsets were studied, and, as expected, the results indicate inclusion of more nuclides leads to better spectra but with diminishing improvements, with the best-case spectrum having an average, relative, group-wise error of approximately 51%. Furthermore, spectra found from minimum-norm and Tihkonov-regularization inversion were of lower quality than the maximum entropy solutions. Finally, the addition of thermal-neutron filters (here, Cd and Gd) provided substantial improvement over unshielded responses alone. The results, as a whole, suggest that in-core, neutron spectroscopy is at least marginally feasible.

Multiplicity counting from fission detector signals with time delay effects

In recent work, we have developed the theory of using the first three auto- and joint central moments of the currents of up to three fission chambers to extract the singles, doubles and triples count rates of traditional multiplicity counting (Pázsit and Pál, 2016; Pázsit et al., 2016). The objective is to elaborate a method for determining the fissile mass, neutron multiplication, and (α, n) neutron emission rate of an unknown assembly of fissile material from the statistics of the fission chamber signals, analogous to the traditional multiplicity counting methods with detectors in the pulse mode. Such a method would be an alternative to He-3 detector systems, which would be free from the dead time problems that would be encountered in high counting rate applications, for example the assay of spent nuclear fuel.A significant restriction of our previous work was that all neutrons born in a source event (spontaneous fission) were assumed to be detected simultaneously, which is not fulfilled in reality. In the present work, this restriction is eliminated, by assuming an independent, identically distributed random time delay for all neutrons arising from one source event. Expressions are derived for the same auto- and joint central moments of the detector current(s) as in the previous case, expressed with the singles, doubles, and triples (S, D and T) count rates. It is shown that if the time-dispersion of neutron detections is of the same order of magnitude as the detector pulse width, as they typically are in measurements of fast neutrons, the multiplicity rates can still be extracted from the moments of the detector current, although with more involved calibration factors. The presented formulae, and hence also the performance of the proposed method, are tested by both analytical models of the time delay as well as with numerical simulations. Methods are suggested also for the modification of the method for large time delay effects (for thermalised neutrons).

Fabrication and testing of a 5-node micro-pocket fission detector array for real-time, spatial, iron-wire port neutron-flux monitoring

Micro-Pocket Fission Detectors (MPFDs) can be used to monitor a neutron flux within the core of a nuclear reactor. When assembled into an array, MPFDs can be used to simultaneously monitor the neutron flux at multiple locations within the reactor. One permutation of the MPFD is designed to be deployed into an iron wire flux port, typical for most research and commercial reactors. A single 5-node MPFD array was fabricated at deployed into a dry flux port within the central thimble of the Kansas State University TRIGA Mk. II research nuclear reactor. The sensitivity of each MPFD node at operating voltages of +50V, +100V, and +200V was observed and an optimal operating voltage of +100V was identified. No cross-talk was observed between instrument channels during real-time neutron flux measurement. Linear power tracking and real-time response to reactor power were both observed between 10kWth and 100kWth reactor power.

The Coincident Fission Fragment Detector (CFFD)

A Parallel Plate Avalanche Counter (PPAC) based fission detector system, called the Coincident Fission Fragment Detector (CFFD), has been developed for the ReA3 re-accelerator facility of the National Superconducting Cyclotron Laboratory (NSCL). Binary reaction kinematics are reconstructed based on position and time-of-flight measurements of fission fragments. Large area PPACs provide 1ns level time resolution and mm level position resolution. The detectors allow measurements of fission product angular and mass distributions of heavy-ion induced fusion reactions. The 30cm by 40cm active area of each PPAC provides large solid angle coverage well suited for measurements of low intensity rare-isotope beams (RIBs).

Fabrication and testing of a 4-node micro-pocket fission detector array for the Kansas State University TRIGA Mk. II research nuclear reactor

Advancements in nuclear reactor core modeling and computational capability have encouraged further development of in-core neutron sensors. Micro-Pocket Fission Detectors (MPFDs) have been fabricated and tested previously, but successful testing of these prior detectors was limited to single-node operation with specialized designs. Described in this work is a modular, four-node MPFD array fabricated and tested at Kansas State University (KSU). The four sensor nodes were equally spaced to span the length of the fuel-region of the KSU TRIGA Mk. II research nuclear reactor core. The encapsulated array was filled with argon gas, serving as an ionization medium in the small cavities of the MPFDs. The unified design improved device ruggedness and simplified construction over previous designs. A 0.315-in. (8-mm) penetration in the upper grid plate of the KSU TRIGA Mk. II research nuclear reactor was used to deploy the array between fuel elements in the core. The MPFD array was coupled to an electronic support system which has been developed to support pulse-mode operation. Neutron-induced pulses were observed on all four sensor channels. Stable device operation was confirmed by testing under steady-state reactor conditions. Each of the four sensors in the array responded to changes in reactor power between 10 kWth and full power (750 kWth). Reactor power transients were observed in real-time including positive transients with periods of 5, 15, and 30s. Finally, manual reactor power oscillations were observed in real-time.

MCNP6 simulated performance of Micro-Pocket Fission Detectors (MPFDs) in the Transient REActor Test (TREAT) facility

Micro-Pocket Fission Detectors (MPFDs) are under development for in-core neutron flux measurements at the Transient REActor Test facility (TREAT) and in other experiments at Idaho National Laboratory (INL). The sensitivity of MPFDs to the energy dependent neutron flux at TREAT has been determined for 0.0300-μm thick active material coatings of 242Pu, 232Th, natural uranium, and 93% enriched 235U. Self-shielding effects in the active material of the MPFD was also confirmed to be negligible. Finally, fission fragment energy deposition was found to be in conformance with previously reported results.

Measurement of the neutron capture cross section of the fissile isotope 235U with the CERN n_TOF total absorption calorimeter and a fission tagging based on micromegas detectors

Current and future nuclear technologies require more accurate nuclear data on (n,γ) cross sections and the α-ratios of fissile isotopes. Their measurement presents several difficulties, mainly related to the strong fission γ-ray background competing with the weaker γ-ray cascades used as the experimental signature of the (n,γ) process. A specific setup was used at the CERN n TOF facility in 2012 for the measurement of the (n,γ) cross section and α-ratios of fissile isotopes and used for the case of the 235U isotope. The setup consists of a set of micromegas fission detectors surrounding the 235U samples all placed inside a segmented BaF2 Total Absorption Calorimeter.

A compact multi-plate fission chamber for the simultaneous measurement of 233U capture and fission cross-sections

233 U plays the essential role of fissile nucleus in the Th-U fuel cycle. A particularity of 233 U is its small neutron capture cross-section which is about one order of magnitude lower than the fission cross-section on average. Therefore, the accuracy in the measurement of the 233 U capture cross-section essentially relies on efficient capture-fission discrimination thus a combined setup of fission and γ-detectors is needed. At CERN n_TOF the Total Absorption Calorimeter (TAC) coupled with compact fission detectors is used. Previously used MicroMegas (MGAS) detectors showed significant γ-background issues above 100 eV coming from the copper mesh. A new measurement campaign of the 233 U capture cross-section and alpha ratio is planned at the CERN n_TOF facility. For this measurement, a novel cylindrical multi ionization cell chamber was developed in order to provide a compact solution for 14 active targets read out by 8 anodes. Due to the high specific activity of 233 U fast timing properties are required and achieved with the use of customized electronics and the very fast ionizing gas CF4 together with a high electric field strength. This paper describes the new fission chamber and the results of the first tests with neutrons at GELINA proving that it is suitable for the 233 U measurement.