Fission Counter Research Articles

Fission counter array for pulse-mode measurements of high-flux and high-energy neutrons

This manuscript describes a neutron counting system based on cylindrical fission counters that can monitor neutron activity for high-energy neutron flux above 10 MeV under electrically noisy environments with intense gamma rays. Miniature fission counters with depleted uranium as sensitive material and modular electronics were built for digital signal processing and high-countrate operation. The counters are 9.5 mm in diameter and 71.1 mm in active length. The author presents the results of Monte Carlo simulations of the fission-counter response for selected neutron sources and energies based on ENDF7.1, JENDL-5, and TENDL-2021 nuclear data libraries from 1 meV to 200 MeV. For a white neutron beam (E‾ = 16.36 MeV) that irradiates the front face of a counter, the intrinsic efficiency is evaluated to be (2.24 ± 0.02) × 10−5 counts/n, while the efficiency of the counter in the array appears to increase by at most 6.7%.

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.

A spurious-count effect observed in Rossi-α and Feynman-α analysis using fission counter for Kyoto University Reactor

ABSTRACT For a subcritical state of Kyoto University Reactor (KUR), Rossi-α and Feynman-α analyses were carried out using a water-surrounded fission counter assigned as reactor startup channel, where the Rossi-α analysis had a long time-interval range from 0.5 µs to 2 ms and the Feynman-α analysis had a wide gate-width range from 1 µs to 5 ms. The result in a shorter time range than several microseconds shows clearly a rise in a positive correlation amplitude resulting from spurious count, which is a well-known characteristic of fission counter. A time constant of the generating process of the spurious count and a probability of occurrence of a spurious count subsequently to a neutron count can be determined from these analyses. The time constant is around 1 µs, and the occurrence probability is around 0.1%, which is one digit smaller than that evaluated from a previous Feynman-α analysis using a graphite-surrounded fission counter. The considerable difference between the present and the previous occurrence probability indicates that the probability must strongly depend on surroundings of fission counter.

Fission product yield measurements using monoenergetic photon beams

Background: High-accuracy and self-consistent fission product yield (FPY) data are needed to advance microscopic/macroscopic descriptions of the nuclear fission process, to improve the predictive power of phenomenological models, and for applications in nuclear energy, nuclear forensics, and homeland security.Purpose: In a collaboration between the Triangle Universities Nuclear Laboratory (TUNL), Los Alamos National Laboratory (LANL), and Lawrence Livermore National Laboratory (LLNL), the dependence of a number of cumulative FPYs on the incoming neutron energy has been measured and unexpected energy dependencies of certain fission products have been reported [M. E. Gooden, Nucl. Data Sheets 131, 319 (2016)]. To investigate whether this observation is unique to neutron-induced fission, a program has been initiated to measure FPYs in photon-induced fission.Method: The photon-induced FPYs were measured by a combination of fission counting using a specially designed dual-fission chamber and $\ensuremath{\gamma}$-ray counting. The measurements were carried out with a monoenergetic photon beam at the $\mathrm{HI}\ensuremath{\gamma}\mathrm{S}$ facility. Gamma-ray counting of the activated targets was performed with well-shielded high-purity germanium (HPGe) detectors over a period of two months after irradiation to properly identify the decay history of fission products.Results: We report on our photofission product yield measurements on $^{235}\mathrm{U}$, $^{238}\mathrm{U}$, and $^{239}\mathrm{Pu}$ using a monoenergetic photon beam of ${E}_{\ensuremath{\gamma}}=13\phantom{\rule{0.28em}{0ex}}\mathrm{MeV}$. More than 40 fission products were uniquely identified, and their yield values were computed. The use of the fission chamber with post-activation measurements has provided absolute fission product yield data with minimal uncertainties.Conclusion: The photon-induced cumulative fission product yields of $^{235}\mathrm{U}$, $^{238}\mathrm{U}$, and $^{239}\mathrm{Pu}$ are compared with previous photon- and neutron-induced fission measurements. In the near future data will be obtained at lower and higher photon energies.

Sensitivity analysis of source intensity and time bin size for the Rossi-alpha method in a numerical reactor model

Until now, various studies focusing on subcriticality measurements have been conducted using the Rossi-alpha method. However, no guidelines have been provided for the source intensity and time bin size of the Rossi-alpha analysis. In this study, sensitivity analyses were performed to determine an optimized source intensity and time bin size for a numerical reactor model. Using the MCNP6 Monte Carlo code, fission count signals in the fuel region were tallied 50 times with different random seeds in real-time mode. These signals were applied to the Rossi-alpha method as detector signals. The estimated keff values were compared to the reference keff values generated by the MCNP6 criticality calculation.As a result of the sensitivity test for the neutron source intensity, the Rossi-alpha method has errors of 200 pcm or less when the source intensity is greater than 105 s−1. It is shown that the standard deviations of 50 analyses are converged when the source intensity is greater than 105 s−1. At the same time, as a result of the sensitivity test for the time bin size, the Rossi-alpha method shows errors less than 200 pcm when the time bin size is smaller than 10−5 s. It is shown that the standard deviations of 50 analyses are converged when the time bins are smaller than 10−5 s. For the numerical reactor model in this study, the threshold time bin size was set to 10−5 s. However, it generally depends on the neutron generation time of the problem. Thus, it is recommended to set the neutron source intensity to 105 s−1 or higher and the time bins to the time of the neutron generation or less when the Rossi-alpha method is applied to subcriticality measurements in nuclear reactor cores.

Energy dependence of fission product yields from 235U, 238U, and 239Pu with monoenergetic neutrons between thermal and 14.8 MeV

Under a joint collaboration between TUNL-LANL-LLNL, a set of absolute fission product yield measurements has been performed. The energy dependence of a number of cumulative fission product yields (FPY) have been measured using quasi-monoenergetic neutron beams for three actinide targets, 235 U, 238 U and 239 Pu, between 0.5 and 14.8 MeV. The FPYs were measured by a combination of fission counting using specially designed dual-fission chambers and γ-ray counting. Each dual-fission chamber is a back-to-back ionization chamber encasing an activation target in the center with thin deposits of the same target isotope in each chamber. This method allows for the direct measurement of the total number of fissions in the activation target with no reference to the fission cross-section, thus reducing uncertainties. γ-ray counting of the activation target was performed on well-shielded HPGe detectors over a period of two months post irradiation to properly identify fission products. Reported are absolute cumulative fission product yields for incident neutron energies of 0.5, 1.37, 2.4, 3.6, 4.6, 5.5, 7.5, 8.9 and 14.8 MeV. Preliminary results from thermal irradiations at the MIT research reactor will also be presented and compared to present data and evaluations. This work was performed under the auspices of the U.S. Department of Energy by Los Alamos National Security, LLC under contract DE-AC52-06NA25396, Lawrence Livermore National Laboratory under contract DE-AC52-07NA27344 and by Duke University and Triangle Universities Nuclear Laboratory through NNSA Stewardship Science Academic Alliance grant No. DE-FG52-09NA29465, DE-FG52-09NA29448 and Office of Nuclear Physics Grant No. DE-FG02-97ER41033.

Fuel element burnup measurements for the equilibrium LEU silicide RSG GAS (MPR-30) core under a new fuel management strategy

After the equilibrium LEU silicide core of RSG GAS was achieved, there was a strong need to validate the new fuel management strategy by measuring burnup of fuel elements comprising the core. Since the regulatory body had a great concern on the safety limit of the silicide fuel element burnup, amongst the 35 burnt fuel elements we selected 22 fuel elements with high burnup classes i.e. from 20 to 53% loss of U-235 (declared values) for the present measurements. The burnup measurement method was based on a linear relationship between reactivity and burnup where the measurements were conducted under subcritical conditions using two fission counters of the reactor startup channel. The measurement results were compared with the declared burnup evaluated by an in-house in-core fuel management code, BATAN-FUEL. A good agreement between the measured burnup values and the calculated ones was found within 8% uncertainties. Possible major sources of differences were identified, i.e. large statistical errors (i.e. low fission counters’ count rates), variation of initial U-235 loading per fuel element and accuracy of control rod indicators. The measured burnup of the 22 fuel elements provided the confirmation of the core burnup distribution planned for the equilibrium LEU silicide core under the new fuel management strategy.

Stability improvement of noise analysis method in the case of random noise contamination for subcriticality measurements

The performances of five noise analysis methods (Rossi-alpha, cross-correlation, Feynman-alpha, Feynman difference, and power spectral density) have been evaluated for subcriticality measurement of nuclear reactors, and two new approaches have been proposed to improve the stability of keff estimations: (1) introduction of multiple detector signals and (2) new formulation of the Rossi-alpha method considering random noise contaminations. Modified thermal Godiva problem has been designed for the evaluations.The analysis was performed with fission count signals from the core of the Godiva model on three cases according to the level of random noise. The five noise analysis methods showed high accuracy of the estimated keff when there was no contamination. In the case of low level contamination, they showed stable results with acceptable accuracy. However, in the case of high level contamination, although the Rossi-alpha results still showed high accuracy, the stability of the five noise analysis methods was significantly decreased. In order to reduce the instability for estimating the keff, two novel approaches were proposed: one is to design eight-detector model problem, and the other is to derive the advanced Rossi-alpha equation with consideration of random noise. In the eight-detector problem, one signal from the core is equally divided by three planes passing through the center of the model geometry and perpendicular to one another. It describes the situation using multiple detector signals to reduce the adverse impact of random noise contamination. On the other hand, the advanced Rossi-alpha formulation was used as the fitting curve equation for the evaluations. As a result, it was confirmed that the new approaches improved the stability of keff estimations.

Measurement and analysis of fission rates in a spherical mockup of uranium and polyethylene

Measurements of the reaction rate distribution were carried out using two kinds of Plate Micro Fission Chamber (PMFC). The first is a depleted uranium chamber and the second an enriched uranium chamber. The material in the depleted uranium chamber is strictly the same as the material in the uranium assembly. With the equation solution to conduct the isotope contribution correction, the fission rate of 238U and 235U were obtained from the fission rate of depleted uranium and enriched uranium. Then, the fission count of 238U and 235U in an individual uranium shell was obtained. In this work, MCNP5 and continuous energy cross sections ENDF/BV.0 were used for the analysis of fission rate distribution and fission count. The calculated results were compared with the experimental ones. The calculation of fission rate of DU and EU were found to agree with the measured ones within 10% except at the positions in polyethylene region and the two positions near the outer surface. Because the fission chamber was not considered in the calculation of the fission counts of 238U and 235U, the calculated results did not agree well with the experimental ones.

Advantages and challenges of automated apatite fission track counting

Fission track thermochronometry data are often a core element of modern tectonic and denudation studies. Soon after the development of the fission track method, interest emerged for the development of automated counting procedures to replace the time consuming labor of counting fission tracks under the microscope. Automated track counting became feasible in recent years with increasing improvements in computer software and hardware, and the development of microscopes and stage systems. One such example examined in this study is the commercial automated fission track counting procedure from Autoscan Systems Pty. and the University of Melbourne thermochronology group. We conducted experiments that are designed to reliably and consistently test the ability of this fully automated counting system to recognize fission tracks in apatite and a muscovite external detector. Fission tracks were analyzed in a series of samples with a step-wise increase in sample complexity. In addition to the automated counting procedure, the same samples were also analyzed using conventional counting procedures. We found for all samples that the fully automated fission track counting procedure using the Autoscan System yields a larger scatter in the measured fission-track densities compared to conventional (manual) track counting. This scatter typically resulted from the false identification of tracks, regardless of the image filtering procedure used or sample coating. Large differences between track densities analyzed with the automated counting persisted between different grains analyzed in one sample as well as between different samples. As a result of these differences a manual correction of the fully automated fission track counts is necessary for each individual surface area and grain counted. This manual correction procedure significantly increases the time required to analyze a sample with the automated counting procedure compared to the conventional approach.

The use of polyimide foils to prevent contamination from self-sputtering of 252Cf deposits in high-accuracy fission counting

It is demonstrated that a thin polyimide foil can be employed to prevent contamination from the self-sputtering of a 252Cf source under vacuum, with small energy loss of the emitted fission fragments, with very small effect on the efficiency of counting the fission fragments, and with a long lifetime of the plastic foils.

Doing Time: Apatite Fission Track Analysis in Undergraduate Geoscience Courses

Geologic time is a central concept in Earth science teaching and research, but undergraduate students are seldom involved in doing time: measuring ‘ages’ of rocks or minerals. Integration of apatite fission track (AFT) analysis in geoscience classes is an accessible way to involve students in determining dates for geologic events. AFT analysis is an active-learning, inquiry-based technique in which students can visualize and understand concepts such as radioactive decay and its relationship to geologic ages, and the method and its applications can be a basis for cross-discipline instruction within a physical sciences curriculum. Fission-track analysis exercises can be incorporated into a variety of physical science classes or can be run as a separate workshop on geologic time or tectonics. We present a teaching module for incorporation of AFT analysis into general physical geology or historical geology lecture and/or laboratory. This exercise involves counting fission tracks using images from prepared samples (apatite grains already separated from a rock) with known uranium content. The data collected by students can then be used to demonstrate a variety of concepts, from radioactive decay to time-temperature paths and rates of geologic processes. With this module, undergraduates learn problem-solving techniques and experience a hands-on, quantitative approach to geologic time, rates of geologic processes, and the scientific method in general.

The design, construction and performance of a variable collimator for epithermal neutron capture therapy beams

A patient collimator for the fission converter based epithermal neutron beam (FCB) at the Massachusetts Institute of Technology Research Reactor (MITR-II) was built for clinical trials of boron neutron capture therapy (BNCT). A design was optimized by Monte Carlo simulations of the entire beam line and incorporates a modular construction for easy modifications in the future. The device was formed in-house by casting a mixture of lead spheres (7.6 mm diameter) in epoxy resin loaded with either 140 mg cm−3 of boron carbide or 210 mg cm−3 of lithium fluoride (95% enriched in 6Li). The cone shaped collimator allows easy field placement anywhere on the patient and is equipped with a laser indicator of central axis, beam's eye view optics and circular apertures of 80, 100, 120 and 160 mm diameter. Beam profiles and the collateral dose in a half-body phantom were measured for the 160 mm field using fission counters, activation foils as well as tissue equivalent (A-150) and graphite walled ionization chambers. Leakage radiation through the collimator contributes less than 10% to the total collateral dose up to 0.15 m beyond the edge of the aperture and becomes relatively more prominent with lateral displacement. The measured whole body dose equivalent of 24 ± 2 mSv per Gy of therapeutic dose is comparable to doses received during conventional therapy and is due principally (60–80%) to thermal neutron capture reactions with boron. These findings, together with the dose distributions for the primary beam, demonstrate the suitability of this patient collimator for BNCT.

Fission of Pb nuclei induced by 0.5, 1.0, 1.5, 3.7 and [formula omitted] protons in the volume of massive U/Pb and Pb targets

Experiments with relativistic protons accelerated at the Synchrophasotron LHE, Dubna, with energies of 0.5, 1.0, 1.5, 3.7 and 7.5 GeV hitting massive targets of (nat. U)/Pb and Pb were carried out using SSNTD during the years 1996–1999. The beam profiles and intensities of both primary particles and fast secondary neutrons were measured inside the massive cylinder blocks of Pb and U by counting fission fragment tracks due to the induced fission of Pb nuclei. The beam diameter typically increases by 20–30% at the depth 10 and 20 cm . With increasing the energy of protons the number of secondary neutrons increases with the depth of the target. Further studies on beam profile measurements inside the massive heavy metal targets are discussed.

Experimental investigation on secondary-count effect in Feynman-α measurement by fission counter

The Feynman-α (variance-to-mean ratio) measurements were carried out by using fission counters, which contain fissile material of around 2 g, to investigate experimentally the secondary-count effect caused by fission events in the counter. The secondary-count effect was observed when the fission counter was covered with graphite, whereas it was not observed when the bare counter was used and when the counter was surrounded by light-water. Therefore, it was concluded that the secondary-count effect depends not only on fission events in the counter but also on the surrounding material of the counter. A formula to correct the secondary-count effect was derived, and it was applied to obtain the inherent decay constant α. The validity of this formula demonstrated through the comparison with the results obtained by using 3He and BF 3 counters which are free from the secondary-count effect.

Measurement of reaction rate distributions in a plastic phantom irradiated by 40- and 65-MeV quasi-monoenergetic neutrons

Reaction rate distributions in a plastic phantom were measured with solid state nuclear track detectors and a fission counter for 40- and 65-MeV quasi-monoenergetic neutrons generated by the 7Li(p,n) reactions with 43- and 68-MeV protons at AVF cyclotron of Japan Atomic Energy Research Institute. Measured distributions were compared with calculated ones.

Uranium-233 fission detectors for neutron flux measurement in reactors

Fission counters have been developed indigenously, using 233U as a sensing material for reactor startup applications. The first model has a counting sensitivity of 0.003 cps/nv in the pulse mode, 1×10−15 A/nv in the dc mode, and 6.4×10−29 A2 Hz/nv in the Campbell mode. A more sensitive counter with about 75 mg of 233U showed an average sensitivity of 0.02 cps/nv. A proposal for 233U in-core miniature dc fission chambers is discussed. A design for high sensitivity (0.5 cps/nv) U-233 pulse counter with transmission line-type readout is also proposed.

Spent-Fuel Assay Performance and Monte Carlo Analysis of the Rensselaer Slowing-Down-Time Spectrometer

The slowing-down-time method for the nondestructive assay of light water reactor (LWR) spent fuel is under development at Rensselaer Polytechnic Institute. A series of assay measurements of an LWR fuel assembly replica were carried out at the Rensselaer lead slowing-down-time spectrometer facility by using [sup 238]U and [sup 232]Th threshold fission detectors and [sup 235]U and [sup 239]Pu probe chambers. An assay model relating the assay signal and the signals of the probe chambers to the unknown masses of the fissile isotopes in the fuel assembly was developed. The probe chamber data were used to provide individual fission counting spectra of [sup 235]U and [sup 239]Pu inside the fuel assembly and to simulate spent-fuel assay signals. The fissile isotopic contents of the fuel were determined to better than 1%. Monte Carlo analyses were performed to simulate the experimental measurements, determine certain parameters of the assay system, and investigate the effect of the fuel assembly and hydrogen impurities on the performance of the system. The broadened resolution of the system caused by the presence of the fuel was still found to be sufficient for the accurate and separate assay of the uranium and plutonium fissiles in spent fuel.

Treatment Planning of Boron Neutron Capture Therapy: Measurements and Calculations

A clinical boron neutron capture therapy (BNCT) facility, employing a high intensity epithermal neutron beam, has recently become available at the High Flux Reactor (HFR) in Petten. Another epithermal beam has been constructed at the Low Flux Reactor (LFR) to test the behaviour of a mixed epithermal neutron and photon field in water phantoms. The neutron and photon fluence distributions have been measured with a BF3 counter and a 235U fission counter and with a GM counter and a Mg(Ar) ionisation chamber, respectively. Absolute determination of the neutron fluence has been performed with foil stacks and of the gamma ray dose with TLD chips. For the design of a treatment planning system for BNCT, the use of a Monte Carlo technique has been investigated. The outcome of these calculations fits the measurements within 1%.

Treatment Planning of Boron Neutron Capture Therapy: Measurements and Calculations

A clinical boron neutron capture therapy (BNCT) facility, employing a high intensity epithermal neutron beam, has recently become available at the High Flux Reactor (HFR) in Petten. Another epithermal beam has been constructed at the Low Flux Reactor (LFR) to test the behaviour of a mixed epithermal neutron and photon field in water phantoms. The neutron and photon fluence distributions have been measured with a BF3 counter and a 235U fission counter and with a GM counter and a Mg(Ar) ionisation chamber, respectively. Absolute determination of the neutron fluence has been performed with foil stacks and of the gamma ray dose with TLD chips. For the design of a treatment planning system for BNCT, the use of a Monte Carlo technique has been investigated. The outcome of these calculations fits the measurements within 1%.