Fission Ionization Chamber Research Articles

Multi-section fission ionization chamber for measurement of [formula omitted] reaction in fission tagging method

The 239Pu(n,γ) reaction cross section is very important for operation of both thermal and fast reactors, when loaded with MOX fuels. According to the NEA/OECD High Priority Request List the precision of cross section data for this reaction should be improved. The cross section of (n,f) reaction is much higher compared to (n,γ) for this isotope. In such conditions the fission tagging technique could be applied to identify the fission background. In the past, this technique was successfully used for capture measurements at the n_TOF facility at CERN. The multi-section fission ionization chamber was constructed and used in the combination with Total Absorption Calorimeter (TAC) for detecting gamma rays for the precise measurement of 239Pu(n,γ) reaction cross section at the n_TOF facility.

Progress in measurements of fission cross sections and total cross sections at CSNS Back-n white neutron source

The neutron-induced cross sections are of great significance for the design of nuclear devices and advanced reactors for the nuclear energy production. At CSNS Back-n white neutron source, new measurements of the fission cross sections and total cross sections are performed with two sets of Day-one spectrometers based on the multi-cell fast fission ionization chamber (FIC). The neutron-induced 236,238U fission cross sections relative to 235U from the fission threshold energy to 200 MeV were measured by using the TOF method and the Fast Ionization Chamber Spectrometer for Fission Cross Section Measurement (FIXM) in the single/double bunch mode of Back-n. The experimental uncertainties are analyzed in detail, and the results from the two modes are consistent. The measured 236,238U/235U fission cross section ratios are compared with previous experiments and evaluations. The 236,238U(n,f) cross sections are obtained based on the standard 235U fission cross section. The neutron total cross sections of carbon and aluminum in the energy region from 1 eV to 20 MeV have been measured by using the TOF method and transmission method based on the Neutron Total Cross Section Spectrometer (NTOX) in the double bunch mode. The total cross section results after unfolding are in good agreement with the previous measurements as well as the broadening of the ENDF/B-Ⅷ.0 evaluation with Gaussian function within the experimental uncertainty. The present results provide the experimental data for further measurements, relevant evaluations and the design of nuclear system.

Measurement of the Pu239(n,f) cross section from 4 keV to 100 MeV using the white neutron source at the CSNS Back-n facility

The neutron-induced fission cross section of $^{239}\mathrm{Pu}$ was measured relative to $^{235}\mathrm{U}(n,f)$ at the back-streaming white neutron beam line (Back-n) of the China Spallation Neutron Source (CSNS). A multicell fast fission ionization chamber was used to perform the measurement. The reliability of the measurement was verified by the high consistency of $^{235}\mathrm{U}$'s resonances between the measurement and the evaluation data. The $^{239}\mathrm{Pu}$ fission cross sections from $4\phantom{\rule{0.16em}{0ex}}\mathrm{keV}$ to $100\phantom{\rule{0.16em}{0ex}}\mathrm{MeV}$ are obtained with 1.7--5.8 % uncertainty when unfolding uncertainties are excluded. The total uncertainties including the unfolding errors, which reflect the effect of the double-bunch unfolding method, are 2.6--15 % from $10\phantom{\rule{0.16em}{0ex}}\mathrm{keV}$ to $100\phantom{\rule{0.16em}{0ex}}\mathrm{MeV}$.

Optimization of the Radiator Composition of a "Non-Burning" Fission Ionization Chamber

The paper is devoted to the analysis and optimization of the composition of the fission ionization chamber radiator, consisting of a mixture of two and three nuclides, to ensure the constant sensitivity of the chamber over a long period of time.

ON THE POSSIBILITY TO DETECT ACTS OF CLUSTER DECAY OF ATOMIC NUCLEI BY THE METHODS OF NOISE DIAGNOSTIC

Cluster decay, as a special type of radioactive decay, up to date, is widely investigated. Nevertheless, until now, this activity is restricted: from one side by the possibilities of theoretical analysis, where some success is obtained for light nuclei only; and from another side only by experiments for nuclei defragmentation in searching for fission on magic nuclei. However, standard methods of radiometry and statistical analysis have not yet been applied. Such possibility can be realized by searching for acts of cluster decay as a rare event on an array of alpha-decay acts of actinides, which are recorded by industrial ionization fission chambers. The scheme of the experiment is discussed, which consists in the registration of every act of alpha-decay, against the background of which it is possible to detect the presence of nuclear clusters based on isotopes of 12,14С, 20О, and others, which can be formed in the decay of 234,235U. The requirements for electronics and the background conditions for such an experiment are discussed in detail.

РАСЧЕТ ПЛОТНОСТИ ПОТОКА ТЕПЛОВЫХ НЕЙТРОНОВ В РАДИАЦИОННОЙ ЗАЩИТЕ ПО ПРОГРАММЕ FRIGATE С ПРИМЕНЕНИЕМ МЕТОДА СУБМОДЕЛИРОВАНИЯ

Using the example of calculating the thermal neutron flux density in an ionization chamber, where a methodically correct solution is required to be obtained only within a fragment of the computational model, and particle transport from the source to the studied region can be carried out in a more rough approximation, the use of DSN-DSN schemes discrete ordinate method on grids with a completely heterogeneous structure in combination with the submodeling method. The paper presents the results of calculating the thermal neutron flux density and the number of fission reactions in the sensitive part of an ionization chamber located in a concrete radiation protection shaft using the FRIGATE program (DSN method using the submodeling method) and cross-validation with the TDMCC program (Monte-Carlo method). The work continued the study, aimed at verifying the method of submodeling by spatial, angular, and energy variables, recommended for use in radiation protection calculations by deterministic codes. Cross-verification was performed using the FRIGATE DSN program and the TDMCC Monte Carlo code using the example of calculating the thermal neutron flux density and the number of fission reactions in the sensitive part of the fission ionization chamber. Verification results showed good agreement between the considered functionals (constant methodological discrepancy within 25 %). The obtained results confirm the effectiveness of using DSN programs in combination with irregular grids and the submodeling method in radiation protection calculations.

Measurement of the U-238/U-235 fission cross section ratio at CSNS – Back-n WNS

Based on the China Spallation Neutron Source (CSNS) – Back-streaming white neutron source (Back-n WNS), the measurement of the U-238/U-235 fission cross section ratio in 1–20 MeV neutron energy region has been carried out by a multi-cell fast fission ionization chamber using the time-of-flight method. The measured resonance peaks of the 235U(n, f) reaction are in good agreement with those of the ENDF/B-8.0 database. The measured six groups of the U-238/U-235 fission cross section ratios in 1–20 MeV neutron energy region agree with those of the ENDF/B-8.0 database in trend, and the average discrepancies between experiment and database are from 1.9% to 2.6%. The relative experimental uncertainties are 2.3%–3.6% (1.4–20 MeV), and the measured data agree with those of database within experimental uncertainties at most energy points. The measured U-238/U-235 fission cross section ratio could provide information and data support in relevant nuclear data evaluation.

234U(n,f) cross section with the FIC detector at CERN (n TOF)

The analysis technique applied to the data and the reaction cross section 234U(n,f) from the FIC (Fission Ionization Chamber) detector at the n TOF facility is pre- sented here. A comparison of the measured neutron induced fission cross section of 234U nucleus is given with the available data from the bibliography. The measure- ments took place at the installation of CERN in Geneva. The detector was placed in front of the neutron beam for the determination of the neutron induced fission cross section of various isotopes of the Th cycle. For the data acquisition, several flash Analog to Digital Converter (fADC) channels were used. This facilitated the detailed off-line analysis of data since all information was stored in the computer. The automation of the process was required because the high amount of stored data. The data analysis aimed at the discrimination of fission events. For this end we had to deal with three main issues: i) The subtraction of the background, ii) the fitting of the pulses and iii) the automation of the process.

Toward short-lived and energy-dependent fission product yields from neutron-induced fission

Fission product yields (FPYs) are an important source of information that are used for basic and applied physics. They are essential observables to address questions relevant to nucleosynthesis in the cosmos that created the elements from iron to uranium, for example, in energy generating processes from fission recycling in binary neutron star mergers; resolving the reactor neutrino anomaly; decay heat release in nuclear reactors; and many national security applications. While new applications will require accurate energy-dependent FPY data over a broad set of incident neutron energies, the current evaluated FPY data files contain only three energy points: thermal, fast, and 14-MeV incident energies.Recent measurements using mono-energetic and pulsed neutron beams at the Triangle Universities Nuclear Laboratory (TUNL) tandem accelerator and employing a dual fission ionization chambers setup have produced self-consistent, high-precision data critical for testing fission models for the neutron-induced fission of the major actinide nuclei. This paper will present new campaign just beginning utilizing a RApid Belt-driven Irradiated Target Transfer System (RABITTS) to measure shorter-lived fission products and the time dependence of fission yields, expanding the measurements from cumulative towards independent fission yields.

A Neutron Testing Facility to Study the Radiation Hardness of the Electronic Components at the SC-1000 Synchrocyclotron of National Research Center Kurchatov Institute—Petersburg Nuclear Physics Institute

The article presents a description of the ISNP/GNEIS testing facility with a neutron spectrum that reproduces the spectrum of atmospheric neutron radiation. The facility was developed at the Petersburg Nuclear Physics Institute (PNPI) of the National Research Center Kurchatov Institute in collaboration with the Institute of Space Device Engineering, a branch of JSC (Joint Stock Company) United Rocket and Space Corporation. The spallation neutron source of the facility is based on the 1-GeV SC-1000 proton synchrocyclotron at the PNPI. The internal neutron-production lead target produces 10-ns pulses of neutrons with a repetition rate of 45–50 Hz and an average neutron intensity of 3 × 1014 n/s (in the 4σ solid angle). In the irradiation area located at a distance of 36 m from the neutron source, a high-quality collimated neutron beam with a broad energy spectrum of 1–1000 MeV and a neutron flux of 4 × 105 n/(cm2 s) allows conducting accelerated single-event soft error testing of electronic components. In the course of the irradiation, the neutron energy spectrum and intensity and the spatial profile of the beam are controlled using a fission ionization chamber (beam monitor) and a position-sensitive multiwire proportional counter (beam profile meter). The data acquisition system of the ISNP/GNEIS facility utilizes 250 MS/s 12 bit CAEN waveform digitizers for the processing of signals from the monitor and profile meter by the neutron time-of-flight technique. In the report, parameters of the ISNP/GNEIS testing facility are discussed in comparison with analogous world-class facilities, as well as requirements and recommendations of the standards used in this field.

A Method for the Evaluation of the Charge Collection Time and the Mean Charge in the Pulses of Ionization Fission Chamber

We present a method for the evaluation of the charge collection time in an ionization fission chamber with flat electrodes. The mean charge and the time of its collection are determined according to the parameters of electric signals recorded at the output of the measuring system. We use a model of formation of current pulse in the detector and a mathematical model of measuring circuit. The procedure of finding the charge collection time and the mean charge per pulse is approved in experiments with KNT-34-type ionization fission chambers. Various possibilities of practical application of the proposed procedure are considered.

Fast-neutron-induced fission cross section of Pu242 measured at the neutron time-of-flight facility nELBE

The fast-neutron-induced fission cross section of $^{242}\mathrm{Pu}$ was measured at the neutron time-of-flight facility $n$ELBE. A parallel-plate fission ionization chamber with novel, homogeneous, large-area $^{242}\mathrm{Pu}$ deposits on Si-wafer backings was used to determine this quantity relative to the IAEA neutron cross-section standard $^{235}\mathrm{U}(n,f)$ in the energy range of 0.5 to 10 MeV. The number of target nuclei was determined from the measured spontaneous fission rate of $^{242}\mathrm{Pu}$. This helps to reduce the influence of the fission fragment detection efficiency on the cross section. Neutron transport simulations performed with geant4, mcnp6, and fluka2011 are used to correct the cross-section data for neutron scattering. In the reported energy range the systematic uncertainty is below 2.7% and on average the statistical uncertainty is 4.9%. The determined results show an agreement within 0.67(16)% to recently published data and a good accordance to current evaluated data sets.

Методика определения времени сбора заряда и среднего заряда в импульсе ионизационной камеры деления

Методика определения времени сбора заряда и среднего заряда в импульсе ионизационной камеры деления

The method for determining the charge collection time and the mean charge in the pulse of the ionization fission chamber

The method for determining the charge collection time and the mean charge in the pulse of the ionization fission chamber

Small Thorium Fission Chambers for Recording Fast Neutrons in Reactor Installations

The results of development work on experimental samples of small fission ionization chambers for recording fast neutrons (En ~ 1–20 MeV) in a reactor core are presented. A technological solution for obtaining a radiator based on ThO2 as a result of nonstationary low-temperature consolidation of an oxide coating from a solution of thorium carboxylate was developed and realized. Samples of ThO2-based radiators on the surfaces of multichannel electrodes of diameter 3 and 5 mm and coatings based on the isotope 238U (in the form U3O8 based on uranium with 235U content 0.00002%) were fabricated and small fission chambers based on them were assembled. Sample radiators were bench tested for stability during thermal cycling (15 cycles, heating to 750°C) and under vibration loads (to 200 Hz). Physico-mechanical and reactor tests of the fission chambers were conducted and the discrimination characteristics of samples and their sensitivity to heat and fast-neutron fluxes were determined. A comparative analysis of the conditions of effective application of small 238U- and 232Th-based fission chambers in long-time operation as well as in single neutron measurements is made.

A multi-layered fast ionization chamber prototype for fission cross section measurements

Aiming at measuring the fission cross sections of various actinides in a wide neutron energy range at the newly built China Spallation Neutron Source (CSNS)—White Neutron Source (WNS), a dedicated multi-layered fast fission ionization chamber prototype was constructed. The physical design of the prototype and the front-end electronics which are optimized to the features of CSNS are here described. The performance of the prototype with uranium targets at a D-T neutron source is presented, focusing in particular on the good ability in fission fragment discrimination, high detection efficiency of 95.5% and time resolution of 20.3 ns.

MODELING OF CHARGE TRANSPORT IN THE ACTIVE VOLUME OF THE IONIZATION FISSION CHAMBER IN CURRENT MODE

Charge transport in the active volume of the cylindrical ionization fission chamber (FC) in the current mode has been studied. The model is based on the continuity equations for ions and electrons, as well as on Poisson’s equation for the electric field. The source for the continuity equations is calculated taking into account a correct distribution of the initial ionization density in the active volume of the chamber and the charge, mass, and energy dependent distributions of the fission fragments. The distributions of the ion and electron density and electric fields inside the active volume are found for two types of chambers – miniature chambers and “large” chambers with the consideration of the space charge. The correct algorithm for calculation of the beginning of the plateau of the current-voltage characteristic – the minimum voltage on the FC to provide the stationary operation of the chamber is given. It is shown that the often used condition of the electric field absence at the anode E (ra ) = 0 to determine this value is incorrect, since it leads to complex values of the electric field inside the chamber active volume. Neglecting the processes of ion diffusion and recombination, the sensitivity and output current of the chamber in the stationary mode are calculated. Calculations have been carried out for miniature and “large” chambers. It has been shown that the use of the approximation for the generation density of ion pairs by the fission fragment along its track to be constant, often used in practice for “large” chambers, leads to significant errors when estimating the densities of ions, electrons and electric fields inside the FC; at that, the sensitivity may differ by an order of magnitude. Исследован перенос зарядов в активном объеме цилиндрической ионизационной камеры деления (ИКД), работающей в токовом режиме. В основу модели положены уравнения непрерывности для ионов и электронов и уравнение Пуассона для электрического поля. Источник для уравнений непрерывности рассчитан с учетом корректного распределения начальной плотности ионизации в активном объеме камеры и разброса фрагментов деления по их заряду, массе и энергии. Найдены распределения плотности ионов, электронов и электрического поля внутри активного объема для двух типов камер – миниатюрных и «больших» с учетом пространственного заряда. Дан правильный алгоритм для расчета начала плато вольтамперной характеристики – минимального напряжения на ИКД для обеспечения стационарной работы камеры. Показано, что часто используемое условие отсутствия электрического поля на аноде E(ra ) = 0 для определения этой величины является некорректным, поскольку приводит к комплексным значениям электрического поля внутри активного объема камеры. Пренебрегая процессами диффузии и рекомбинации ионов, рассчитаны чувствительность и выходной ток камеры при ее работе в стационарном режиме. Расчеты проведены для миниатюрных и «больших» камер. Показано, что часто применяемое на практике использование приближения постоянства плотности генерации пар ионов осколком деления вдоль его трека для «больших» камер приводит к существенным ошибкам в оценке плотностей ионов, электронов и электрических полей внутри ИКД, при этом чувствительности могут отличаться на порядок.

Determination of the fast-neutron-induced fission cross-section of 242Pu at nELBE

The fast-neutron-induced fission cross section of 242Pu was determined in the energy range of 0.5 MeV to 10MeV at the neutron time-of-flight facility nELBE. Using a parallel-plate fission ionization chamber this quantity was measured relative to 235U(n,f). The number of target nuclei was thereby calculated by means of measuring the spontaneous fission rate of 242Pu. An MCNP 6 neutron transport simulation was used to correct the relative cross section for neutron scattering. The determined results are in good agreement with current experimental and evaluated data sets.

Fission Ionization Chamber as Reference Source in Neutron Flux Analysis

The paper considers the statistics of intervals between neutron registration by the detecting system based on fission ionization chamber KHT31-1, statistics of intervals between noise signal impulses in absence of neutrons and possibility to distinguish neutron flows from different sources. It is shown that spurious noise signal from alpha decay of 234U in a radiator of the ionization chamber KHT31-1 can be used as a reference signal in measuring neutron fluxes from different sources. The analysis of long-term continuous sets of measurements is taken as a basis of mathematical treatment.

Measurement of 232Th(n,5n γ) cross sections from 29 MeV to 42 MeV

The excitation function of the reaction 232 Th(n, 5nγ)228 Th from 29 to 42 MeV has been measured for the first time at the quasi-monoenergetic neutron beam of the UCL cyclotron CYCLONE employing the 7Li(p,n) source reaction. Taking advantage of the good energy resolution of the planar High-Purity Germanium (HPGe) detectors, prompt γ-ray spectroscopy was used to detect the γ-rays resulting from the decay of excited states of nuclei created by the (n,xn) reactions. The neutron beam was characterized by a combination of time of flight measurements carried out using a liquid scintillation detector and a 238U fission ionization chamber. Fluence measurements were performed using a proton recoil telescope. The results are compared with TALYS-1.4 code calculations.