Delayed Neutron Spectra Research Articles

Development of transient Monte Carlo in a fissile system with β-delayed emission from individual precursors using modified open source code OpenMC(TD)

In deterministic and Monte Carlo transport codes, β-delayed emission is included using a group structure where all of the precursors are grouped together in 6 groups or families, but given the increase in computational power, nowadays there is no reason to keep this structure. Furthermore, there have been recent efforts to compile and evaluate all the available β-delayed neutron emission data and to measure new and improved data on individual precursors. In order to be able to perform a transient Monte Carlo simulation, data from individual precursors needs to be implemented in a transport code. This work is the first step towards the development of a tool to explore the effect of individual precursors in a fissile system. In concrete, individual precursor data is included by expanding the capabilities of the open source Monte Carlo code OpenMC. In the modified code – named Time Dependent OpenMC or OpenMC(TD)– time dependency related to β-delayed neutron emission was handled by using forced decay of precursors and combing of the particle population. The data for continuous energy neutron cross-sections was taken from JEFF-3.1.1 library. Regarding the data needed to include the individual precursors, cumulative yields were taken from JEFF-3.1.1 and delayed neutron emission probabilities and delayed neutron spectra were taken from ENDF-B/VIII.0. OpenMC(TD) was tested in a monoenergetic system, an energy dependent unmoderated system where the precursors were taken individually or in a group structure, and in a light-water moderated energy dependent system, using 6-groups, 50 and 40 individual precursors. Neutron flux as a function of time was obtained for each of the systems studied. These results show the potential of OpenMC(TD) as a tool to study the impact of individual precursor data on fissile systems, thus motivating further research to simulate more complex fissile systems.

ФИЛЬТР КАЛЬМАНА В ЗАДАЧЕ ОЦЕНКИ 8-ГРУППОВЫХ СПЕКТРОВ ЗАПАЗДЫВАЮЩИХ НЕЙТРОНОВ ПРИ ДЕЛЕНИИ 235U ТЕПЛОВЫМИ НЕЙТРОНАМИ

In the present work the estimation of the 8-group delayed neutron energy spectra emitted in the fission of 235U by thermal neutrons have been done. Aggregate delayed neutron spectra that were measured with high resolution in the different time intervals after the thermal neutron irradiation of the 235U sample presented in the IAEA reference database for beta-delayed neutron emission have been used for the estimation process. Due to the lack of the uncertainties for the spectra presented in this database and the necessity of this data in the process of estimation of delayed neutron group spectra at first the estimation of uncertainties of these spectra have been done. In the estimation process the following components of uncertainty have been considered: statistical; the component due to the neutron background; component due to the distinctive features of neutron spectra restoration process measured using spectrometer filled by helium-3 and also the component appearing due to introduction of the correction on the distortion of the neutron spectra as a result of passing through the lead sheilding. Estimation of spectra have been made using the technique based on the Kalman filter that allows to take into account the uncertainties of the input data. Estimation results are shown in the numerical and graphic form.

Development of a Reference Database for Beta-Delayed Neutron Emission

Beta-delayed neutron emission is important for nuclear structure and astrophysics as well as for reactor applications. Significant advances in nuclear experimental techniques in the past two decades have led to a wealth of new measurements that remain to be incorporated in the databases.We report on a coordinated effort to compile and evaluate all the available β-delayed neutron emission data. The different measurement techniques have been assessed and the data have been compared with semi-microscopic and microscopic-macroscopic models. The new microscopic database has been tested against aggregate total delayed neutron yields, time-dependent group parameters in 6-and 8-group re-presentation, and aggregate delayed neutron spectra. New recommendations of macroscopic delayed-neutron data for fissile materials of interest to applications are also presented. The new Reference Database for Beta-Delayed Neutron Emission Data is available online at: http://www-nds.iaea.org/beta-delayed-neutron/database.html.

High resolution measurements of time-dependent integral delayed neutron spectra from thermal neutron induced fission of 235U

Composite delayed energy spectra from epithermal neutron induced fission of 235U have been measured for twelve delay-time intervals extending from 0.12 to 150s. These data provide a comprehensive set of composite spectra with enhanced sensitivity to neutrons emitted in less than 1s following fission. Extensive comparisons between these measurements and the composite spectra obtained by summation technique are presented and discussed herein.

Calculation of Beta Decay Half-Lives and Delayed Neutron Branching Ratio of Fission Fragments with Skyrme-QRPA

Nuclear β -decay and delayed neutron (DN) emission is important for the r -process nucleosynthesis after the freeze-out, and stable and safe operation of nuclear reactors. Even though radioactive beam facilities have enabled us to measure β -decay and branching ratio of neutron-rich nuclei apart from the stability line in the nuclear chart, there are still a lot of nuclei which one cannot investigate experimentally. In particular, information on DN is rather scarce than that of T 1/2 . To predict T 1/2 and the branching ratios of DN for next JENDL decay data, we have developed a method which comprises the quasiparticle-random-phase-approximation (QRPA) and the Hauser-Feshbach statistical model (HFSM). In this work, we calculate fission fragments with T 1/2 ≤ 50 sec. We obtain the rms deviation from experimental half-life of 3:71. Although the result is still worse than GT2 which has been adopted in JENDL decay data, DN spectra are newly calculated. We also discuss further subjects to be done in future for improving the present approach and making next generation of JENDL decay data.

The influence of the neutron source spectrum on the infinite homogeneous reactor in subcritical condition

An analysis of the influence of the neutron source spectrum on some key parameters of a reactor system in subcritical condition is presented. The analysis is conducted on the infinite homogeneous reactor, a kind of approach useful to have a quick understanding of nuclear reactor phenomena. The main differences between the spectrum in critical and subcritical condition are highlighted. Among the results of the analysis we present the impact on the multiplication factor of neglecting the delayed neutron spectrum. The homogenization error on few-group cross sections due to using critical spectra in the standard homogenization procedures is also discussed.

Spatial kinetics in fast reactors

The analysis of the solution to the spatial nonstationary equation of neutron transport is presented by the example of a fast reactor. Experiments in spatial kinetics conducted recently at the complex of critical assemblies (fast physical stand) and computations of their data using the TIMER code (for solving the nonstationary equation in multidimensional diffusion approximation for direct and inverse problems of reactor kinetics) have shown that kinetics of fast reactors substantially differs from kinetics of thermal reactors. The difference is connected with influence of the delayed neutron spectrum on rates of the process in a fast reactor.

Uncertainty evaluation of calculated and measured kinetics parameters of Tehran Research Reactor

Effective delayed neutron fraction β eff and neutron generation time Λ are important factors in reactor physics calculation and transient analysis. In the first stage of this research, these kinetics parameters have been calculated for two states of Tehran Research Reactor (TRR), i.e. cold (fuel, clad and coolant temperature 20 °C) and hot (fuel, clad and coolant temperature 65, 49 and 44 °C, respectively) states using MTR_PC computer code. The ratio of ( β eff ) i /( β eff ) core plays an important role in reactivity accident analysis codes. This parameter and its contribution to effective delayed neutron fraction from each nucleus have been calculated in cold and hot reactor states. Uncertainty of effective delayed neutron fraction is evaluated in terms of following four quantities; basic delayed neutron constants, delayed neutron spectra, energy dependence of delayed neutron yield ( ν d ) and fission cross-section of 235U and 238U. In the second stage, these parameters have been measured with an experimental method based on Inhour equation. The calculated and measured values are in good agreement. Relative Percent Errors (RPEs) are 2.8% for β eff and 5.7% for Λ in the cold state.

Sources: A code for calculating (alpha, n), spontaneous fission, and delayed neutron sources and spectra

SOURCES is a computer code that determines neutron production rates and spectra from (alpha, n) reactions, spontaneous fission, and delayed neutron emission due to the decay of radionuclides in homogeneous media, interface problems, and three-region interface problems. The code is also capable of calculating the neutron production rates due to (alpha, n) reactions induced by a monoenergetic beam of alpha particles incident on a slab of target material. The (alpha, n) spectra are calculated using an assumed isotropic angular distribution in the center-of-mass system with a library of 107 nuclide decay alpha-particle spectra, 24 sets of measured and/or evaluated (alpha, n) cross sections and product nuclide level branching fractions, and functional alpha particle stopping cross sections for Z < 106. Spontaneous fission sources and spectra are calculated with evaluated half-life, spontaneous fission branching, and Watt spectrum parameters for 44 actinides. The delayed neutron spectra are taken from an evaluated library of 105 precursors. The code outputs the magnitude and spectra of the resultant neutron sources. It also provides an analysis of the contributions to that source by each nuclide in the problem.

Calculation of delayed-neutron energy spectra in a quasiparticle random-phase approximation–Hauser-Feshbach model

Theoretical $\ensuremath{\beta}$-delayed-neutron spectra are calculated based on the Quasiparticle Random-Phase Approximation (QRPA) and the Hauser-Feshbach statistical model. Neutron emissions from an excited daughter nucleus after $\ensuremath{\beta}$ decay to the granddaughter residual are more accurately calculated than in previous evaluations, including all the microscopic nuclear structure information, such as a Gamow-Teller strength distribution and discrete states in the granddaughter. The calculated delayed-neutron spectra agree reasonably well with those evaluations in the ENDF decay library, which are based on experimental data. The model was adopted to generate the delayed-neutron spectra for all 271 precursors.

SOURCES: a code for calculating (α,n), spontaneous fission, and delayed neutron sources and spectra

SOURCES is a computer code that determines neutron production rates and spectra from (alpha,n) reactions, spontaneous fission and delayed neutron emission owing to the decay of radionuclides in homogeneous media, interface problems and three-region interface problems. The code is also capable of calculating the neutron production rates due to (alpha,n) reactions induced by a monoenergetic beam of alpha particles incident on a slab of target material. The (alpha,n) spectra are calculated using an assumed isotropic angular distribution in the centre-of-mass system with a library of 107 nuclide decay alpha-particle spectra, 24 sets of measured and/or evaluated (alpha,n) cross sections and product nuclide level branching fractions, and functional alpha particle stopping cross sections for Z < 106. Spontaneous fission sources and spectra are calculated with evaluated half-life, spontaneous fission branching and Watt spectrum parameters for 44 actinides. The delayed neutron spectra are taken from an evaluated library of 105 precursors. The code outputs the magnitude and spectra of the resultant neutron sources. It also provides an analysis of the contributions to that source by each nuclide in the problem.

Beta-delayed neutron decay of 33Na

Beta-delayed neutron decay of 33Na has been studied using the on-line mass separator ISOLDE. The delayed neutron spectra were measured by time-of-flight technique using fast scintillators. Two main neutron groups at 800(60) and 1020(80) keV were assigned to the 33Na decay, showing evidence for strong feeding of states at about 4 MeV in 33Mg. By simultaneous β–γ–n counting the delayed neutron emission probabilities P 1n=47(6)% and P 2n=13(3)% were determined. The half-life value for 33Na, T 1/2=8.0(3) ms , was measured by three different techniques, one employing identifying gamma transitions and two employing beta and neutron counting.

Calculation of the delayed neutron effectiveness factor using ratios of k-eigenvalues

In this work, we derive from the transport equation the effectiveness factor used in the β eff calculation and show that, to a first approximation, it is equal to the ratio of two k-eigenvalues: one obtained using the total neutron spectrum, and the other obtained using only the delayed neutron spectrum. This method, which we shall refer to as the k-ratio method, has been tested using several different types of reactor systems. In this report we present the results for the simplest of these systems — the Godiva-I assembly at the Los Alamos Critical Experiments Facility (LACEF). Although the method is not exact, it does yield results that are remarkably close to the traditional adjoint-weighting method.

A Point-Kinetics Approach to Sensitivity Study of Fast Reactor Dynamic Behavior and Delayed Neutron Spectra

The method of point reactor kinetics in conjunction with the new concepts of delayed spectrum factor and beta growth factor is used to calculate the sensitivity of the dynamic behavior of a fast br...

Search for an Energy Dependence in235U Delayed Neutron Spectra

A sensitive search for a dependence of composite (aggregate) delayed neutron spectra on the energy of neutrons inducing fission of /sup 235/U was performed for eight nearly contiguous delay time intervals between 0.17 to 85.5 s. The experimental arrangement combined a helium jet and tape transfer system with a beta-neutron time-of-flight spectrometer. Thermal and fast neutron measurements were performed successively. Results are compared with the spectra derived from individual precursor data complemented by theoretical estimates of missing spectra.

The analysis of delayed neutron energy spectra recorded with 3He ionization chambers

Details of the neutron detection efficiency of 3He ionization chambers have been clarified. The usefulness of pulse rise-time analysis has been explored for the example of the 17N delayed neutron spectrum. Because of obvious disadvantages, this method has been rejected, however, in favour of a complete response correction that removes all others than 3He(n,p) 3H events with great precision without disturbing the efficiency function. Peak fitting analysis is applied to structured delayed neutron spectra, and its physical significance is discussed.

Beta-delayed triton emission in the decay of 8He

A β-delayed triton branch with an intensity of (0.9 ± 0.1)% has been observed in the decay of 8He. The triton energy spectrum was measured with a specially designed ΔE − E telescope. These data as well as an earlier measured β-delayed neutron spectrum cannot be explained from the known levels in 8Li, but it turns out that the main features of both the delayed triton and the delayed neutron spectra from 8He could be understood if one assumes a new state in 8Li at about 8.8 MeV excitation energy.

Search for energy dependence among composite delayed neutron spectra of U-235

Abstract A sensitive technique is described to search for a dependence of delayed neutron spectra on fast vs. thermal fission of U-235. The results of such a search are given.

Delayed neutron spectra and intensities from evaluated precursor data

Recently evaluated and model estimated precursor data are used with preliminary ENDF/B-VI fission yields to calculate total and time-group anti nu/sub d/ and spectra. 8 refs.

Composite delayed-neutron spectra from U-235

Abstract Delayed-neutron spectra from thermal neutron-induced fission of U-235 have been measured over the neutron energy range, .01–2.0 MeV, for delay times following fission ranging from 0.17–85.5 s. A helium jet system was used for the rapid transport of fission products to a low-background area, where the spectra were determined from beta-neutron correlations using the neutron time-of-flight method.