No full-text available
To read the full-text of this research,
you can request a copy directly from the authors.
Measurement of the 239Pu(n,f)/235U(n,f) Cross-Section Ratio with the NIFFTE fission Time Projection Chamber
Abstract
The ²³⁹Pu(n,f)/²³⁵U(n,f) cross-section ratio has been measured with the fission Time Projection Chamber (fissionTPC) from 100 keV to 100 MeV. The fissionTPC provides three-dimensional reconstruction of fission-fragment ionization profiles, allowing for a precise quantification of measurement uncertainties. The measurement was performed at the Los Alamos Neutron Science Center which provides a pulsed white source of neutrons. The data are recommended to be used as a cross-section ratio shape. A discussion of the status of the absolute normalization and comparisons to ENDF evaluations and previous measurements is included.
... Finally, a 2% normalization difference between the measured 239 Pu(n,f)/ 235 U(n,f) absolute cross-section ratio using a Time Projection Chamber (TPC) detector to the Neutron Standard was recently discussed [15] 1 . Observed differences may be due to deficiencies in the Neutron Standards input data, in the Neutron Standard evaluation methodology or in both. ...
... 2. additional-input: New GMApy fit based on updated uncertainty quantification of 239 Pu data [17][18][19] and new TPC experimental shape data [15,20,21]. [11]. ...
... Case 2 in Figs.1-3 represents the advanced uncertainty quantification of 239 Pu(n,f) GMA input database [17][18][19] and the addition of new TPC experimental data with very low uncertainties [15,20,21] used as shape data (labelled additional-input and represented by the blue line). The impact of these changes on 235 U(n,f) cross sections is a reduction of up to 0.2% in the evaluated fission cross sections, on 238 U(n,f) cross sections -a reduction of 0.25% is observed, and an increase of up to 0.35% on 239 Pu evaluated fission cross sections. ...
A new evaluation of spectrum averaged cross sections (SACS) of ²³⁵U, ²³⁸U, and ²³⁹Pu measured in the ²⁵²Cf(sf) reference neutron field is presented and found to be consistent with the original Mannhart SACS evaluation in IRDF-2002. The comprehensive vetted experimental database that includes SACS ratio and absolute SACS measurements of major actinides is being used to update the SACS database employed as input of the new GMApy code to derive the Neutron Standards. An update of the current neutron standards based on Time Projection Chamber (TPC) shape data, a new comprehensive uncertainty quantification, and revised SACS experimental database is proposed which result in a 0.7% increase of the evaluated ²³⁹Pu(n,f)/²³⁵U(n,f) crosssection ratio in the 1–5 MeV energy region. The increase is due to a 0.3% reduction of the standard ²³⁵U(n,f) cross section and a 0.4% increase of the ²³⁹Pu(n,f) reference cross section in the 1–5 MeV energy region. Those changes are well within estimated USU fission cross-section uncertainties of 1.2%, but are relevant for the evaluated mean values.
... This new information addresses the questions mentioned above on fission and thus better constrains nuclear data. The measurements decisively inform the evaluations either by covering for the first time for broad incident and outoing neutron energy ranges the 239 Pu PFNS with high precision [22][23][24], or by utilizing novel methods or detectors to obtain high-precision (n,f) cross sections and ] p [25][26][27]. We also use the post-scission fission event generator CGMF [28]. It enables, like others of its kind [29][30][31], constraining nuclear data by consistently predicting several fission quantities (distributions in mass, charge and total kinetic energy, PFNS, ] p , etc.), and, thus, eliminating unphysical combinations between them. ...
... New measurements of the shape of the 239 Pu(n,f)/ 235 U(n,f) cross section ratio were performed by the NIFFTE TPC (Neutron Induced Fission Fragment Tracking Experiment Time Projection Chamber) collaboration to address concerns regarding the spread in the previous data relative to the Neutron Data Standard evaluation uncertainty [25]. The bulk of past measurements were undertaken using fission chambers which could be subject to several potential sources of systematic uncertainty [10,11,45]. ...
... With 3D track reconstruction, the fissionTPC benefits from a direct measurement of several quantities needed to adequately correct cross section ratios, including beam and target uniformity and detector efficiency, eliminating the need to make certain assumptions or auxiliary measurements. A detailed and carefully documented analysis was conducted to understand all the relevant uncertainty sources and validate the analysis results [25]. In short, the fissionTPC produced a precision shape measurement that is largely uncorrelated with previous measurements and has some different systematic uncertainties. ...
In the last decade, there has been a renaissance of fission research resulting in new high-precision experiments and advanced fission modeling. For instance, the Chi-Nu and CEA teams supplied, for the first time, the ²³⁹ Pu prompt fission neutron spectrum (PFNS) for broad ranges of incident and outgoing neutron energies. The CEA team also measured ²³⁹ Pu average prompt neutron multiplicities, ν ̄ p , with lower statistical uncertainties and a technique significantly different than the one used in the past. The NIFFTE collaboration provided ²³⁹ Pu( n ,f)/ ²³⁵ U( n ,f) cross section shape ratios with uncertainties below 1% utilizing a novel detector type. Advanced fission event generators were developed, among them CGMF, FIFRELIN, FREYA, and GEF, which calculate post-scission fission observables in a correlated manner. These new experimental data and more consistent fission models change the evaluated PFNS, ν ̄ p , and ( n ,f) cross sections, some only modestly, compared to ENDF/B-VIII.0. In turn, the individual new nuclear data distinctly change simulated effective neutron multiplication factors of fast critical assemblies, but their combined impact is small, while affecting the prediction of LLNL pulsed sphere neutron leakage spectra and reaction rates only within experimental uncertainties. Also, the parameters obtained from fitting to ν ̄ p reproduce various post-scission fission observables within the uncertainties of experimental data. This indicates that new differential experiments and consistent fission modeling reduce compensating errors present in ENDF/B-VIII.0.
... The results of the absolute ratio of 239 Pu(n,f) to 235 U(n,f) cross section measurements are shown in Figure 11. The data were taken from [16]. ...
... The constant bias at about 2% is clearly visible in the energy range 0.2-15 MeV and has an even larger spread above 15 MeV. The authors of [16] provided a very detailed and deep analysis of the modeling of their experiment (experimental details, data reduction procedure, uncertainties of different parameters and so on) but at present could not explain the existing bias, which can be treated as SDF. From our point of view, it is premature to assign the SDF to this measurement, especially taking into account that the reaction rate ratios measured in clean benchmarks show better consistency with the NIFFTE results. ...
... 239 Pu(n,f)/ 235 U(n,f) cross section ratio measured with a fission TPC[16] in comparison with the data[17][18][19] and the results of the GMA fit (solid line) with fission TPC data added to the standard (old) database. ...
Each experiment provides new information about the value of some physical quantity. However, not only measured values but also the uncertainties assigned to them are an important part of the results. The metrological guides provide recommendations for the presentation of the uncertainties of the measurement results: statistics and systematic components of the uncertainties should be explained, estimated, and presented separately as the results of the measurements. The experimental set-ups, the models of experiments for the derivation of physical values from primary measured quantities, are the product of human activity, making it a rather subjective field. The Systematic Distortion Factor (SDF) may exist in any experiment. It leads to the bias of the measured value from an unknown “true” value. The SDF appears as a real physical effect if it is not removed with additional measurements or analysis. For a set of measured data with the best evaluated true value, their differences beyond their uncertainties can be explained by the presence of Unrecognized Source of Uncertainties (USU) in these data. We can link the presence of USU in the data with the presence of SDF in the results of measurements. The paper demonstrates the existence of SDF in Prompt Fission Neutron Spectra (PFNS) measurements, measurements of fission cross sections, and measurements of Maxwellian spectrum averaged neutron capture cross sections for astrophysical applications. The paper discusses introducing and accounting for the USU in the data evaluation in cases when SDF cannot be eliminated. As an example, the model case of 238U(n,f)/235U(n,f) cross section ratio evaluation is demonstrated.
... The original 239 Pu(n,f)/ 235 U(n,f) cross section ratio measurements by Snyder et al. [22] made at LANSCE by the NIFFTE collaboration are higher than the standards evaluation by about 2%. They agree in shape with the standards evaluation. ...
... An analysis by Neudecker [25] was done to compare two GMA evaluations differing only by the addition of the 238 U(n,f)/ 235 U(n,f) [21] and 239 Pu(n,f)/ 235 U(n,f) [22] data sets to one evaluation. Except for changes below 1.5 MeV for 238 U(n,f) at the 0.5% level, the only significant changes were observed above 10 MeV for the 239 Pu(n,f) cross section and 239 Pu(n,f)/ 235 U(n,f) cross section ratio which were as large as 2% lower. ...
An effort is now underway to produce a new evaluation of the neutron standards. It is important to maintain experimental programs to increase the quality and extend the database for the neutron cross section standards in order to improve evaluations of them that will be used to convert cross section measurements made relative to those standards. Measurements have been made for most of the standard cross sections since the last evaluation of the standards. The improved database includes the cross sections for the H(n,n), 6Li(n,t), 10B(n,αγ), 10B(n,α), C(n,n), Au(n,γ), 235U(n,f) and 238U(n,f) standard reactions and ratios among them. The database also includes the 238U(n,γ) and 239Pu(n,f) cross sectionsin addition to the standard cross sections. Those data were included since there are many ratio measurements of those cross sections with the standards and absolute data are available for them.
... Many research teams have been working on fission-TPC design and neutron beam experiments, aiming at more accurate fission cross section measurement. At Los Alamos, a fission-TPC program called NIFFTE for precise measurement of fission cross section and angular distribution has been carried out, and the cross-section ratios between different isotopes and the angular anisotropy of the fragments have been measured [2,[8][9][10][11]. In order to measure the heavy ions in nuclear fission and astrophysics applications, a TPC framework project called FIDIAS as a collaboration between CEA-Irfu (France) and NCRS-Demokritos (Greece) has been built [12,13]. ...
... The Neutron Induced Fission Fragment Tracking Experiment (NIFFTE) collaboration built a dual-chamber time projection chamber, the fission-TPC, 9 to make precision fission cross section measurements, and it has provided several new fission cross section measurements to date. [10][11][12][13] Recent work has been focused on extracting FPY data from these measurements as well. 14 The aim of this work was to find a novel method to determine the fission product associated with a specific energy-loss profile (Bragg curve) extracted from fission-TPC data. ...
A fission time projection chamber (fission-TPC) was developed to provide precise neutron-induced fission measurements for several major actinides. As fission fragments lose energy in one of the gas volumes of the fission-TPC, energy loss information is captured and may be used to determine fission product yields as the stopping power of an ion is dependent on the atomic number. The work presented here demonstrates the ability to apply machine learning techniques for Bragg curve classification. A set of one million energy loss curves for 24 different fission-fragment elements was generated using common stopping power software. A ResNet architecture optimized for 1D data was used to train, test, and validate a model for light and heavy fission fragments using the simulated data. The resultant classification accuracy for the light and heavy fragments indicates that this could be a viable method for elemental classification of data from the fission-TPC.
The paper presents the results of critical, correlation, and spectral precision experiments carried out at the FKBN‑2 assembly machine using a cylindrical multiplying system made of low-background α‑phase plutonium. In the experiments, the critical state of the multiplying system was determined, the characteristics of non-stationary prompt neutron processes were investigated, and the number of nuclear reactions in the detectors was measured at a controlled power level of the multiplying system. After constructing a benchmark model for the multiplying system, a computational simulation of integral experiments was carried out. Based on the experimental results, the average lifetime of prompt neutrons and the effective fraction of delayed neutrons for the multiplying system were determined. The obtained results can be used to verify the codes of neutron-physics and refine neutron constants.
The $^{232}$Th neutron-induced fission cross section was evaluated from 500 keV to 200 MeV. The experimental $^{232}$Th fission cross sections and their ratios to the $^{235,238}$U fission cross sections in the EXFOR library were reviewed and analysed by the least-squares method. The newly published $^{232}$Th/$^{235}$U fission cross section ratios from the time-of-flight measurements at the CERN n_TOF and CSNS Back-n facilities were compiled in EXFOR. Additional simultaneous evaluation was performed by including the experimental $^{233,238}$U and $^{239,240,241}$Pu fission cross sections and their ratios. The new evaluation provides the $^{232}$Th fission cross section systematically lower than the JENDL-5 cross section. The reduction is 4% in the plateau region between 2 and 6 MeV and more significant in the subthreshold fission region. The present evaluation reduces the $^{232}$Th fission cross section averaged over the $^{252}$Cf spontaneous fission neutron spectrum from the JENDL-5 evaluation by 4%, which is closer to the other general purpose libraries but underestimates Grundl et al's measurement by 11%.
Simultaneous evaluation of 233,235,238U and 239,240,241Pu fission cross sections for fast neutrons up to 200 MeV was performed for the JENDL-5 library. Experimental covariances were estimated for each experimental dataset of cross sections or cross section ratios extracted from the EXFOR library, and they were stored in an experimental database dedicated to the new evaluation. The cross sections were expressed by Schmittroth’s roof functions, and the values on defined incident energy grids were adjusted by the least-squares method to reproduce the experimental cross sections and their ratios. The newly evaluated cross sections were validated using the spectrum averaged cross sections measured in the ²⁵²Cf spontaneous fission neutron standard field. The evaluation adopted by the JENDL-5 library was further updated by addition of two datasets and deletion of one dataset.
Fission cross sections have played a particularly important role in the history of fission and its applications. Much fundamental research has been done on this topic, and various phenomenological models devised to calculate fission barrier heights and energy-dependent cross sections. The problem remains daunting though, and the sensitivity of the calculated cross sections on estimates of fission barrier heights and level densities remains high. A fully quantitative and predictive model for fission cross sections, starting from ab initio calculations of the potential energy surface along the fission path, with dynamical and thermal properties accounted for, remains elusive. However, significant theoretical and experimental advances have been made, and are reported in this chapter. Section 1.1 starts with a brief review of the main historical milestones in our understanding of fission cross sections, some terminology and definitions, and an overview of fission cross section properties. Section 1.2 focuses on experimental techniques and detectors used in the measurement of fission cross sections. Section 1.3 introduces the formal R-matrix theory of fission cross sections, and its generalization to a double-humped fission barrier. Finally, Sect. 1.4 describes codes and techniques used in the evaluation of fission cross sections in modern libraries of nuclear data.
The nuclear fission process is not only a fascinating topic for modern science, it also plays a crucial role in a host of important technologies and applications. The safe, reliable and low-carbon production of electricity by nuclear reactors is an important component in a modern energy portfolio. Section 4.1 briefly reviews how a nuclear reactor operates and discusses the nuclear fission processes at play across the entire nuclear fuel cycle. The development of nuclear reactors preceded by a few years only the first atomic explosion in the desert of New Mexico, USA. Safeguarding existing stockpile of nuclear weapons and monitoring nuclear activities across the globe is the topic of Sect. 4.2. Scientific activities enabled by the development of underground nuclear testing, including the discovery of several elements, are also described. Nuclear fission may play a crucial role in the formation of the heaviest elements in the cosmos. In the environments of cataclysmic astrophysical events of supernovae or compact object mergers, continual rapid neutron capture and beta decays may synthesize the actinides. Section 4.3 provides a brief overview of the importance of fission in the r process and highlights the potential insights that can be gained from future observations.
Most evaluated elemental fission product yield distributions are not experimentally measured. Instead, the majority of evaluated distributions are based on analytic expressions of the Zp-model for relevant cumulative yields. Here we report independent elemental fission product yield distributions of a ²³⁵U target for incident neutron energies ranging from 0.11 MeV through 92.4 MeV. Atomic numbers are calculated by an approach that combines a 2E analysis with a stopping force analysis method, developed within this paper. These analyses are applied to more than 6.1 × 10⁶ fission fragment ionization tracks captured within the NIFFTE (Neutron Induced Fission Fragment Tracking Experiment) collaboration fission time projection chamber (fissionTPC). A 3-Z resolution was obtained with the fissionTPC spatial and energy resolutions. Tabulated results are presented for the atomic yield and experimentally derived Zp values as a function of pre-neutron-emission fragment masses for the complete range of incident neutron energies. The stopping-force-derived Zp values tend to support the unchanged charge distribution theory within uncertainty.
We describe the new ENDF/B-VIII.0 evaluated nuclear reaction data library. ENDF/B-VIII.0 fully incorporates the new IAEA standards, includes improved thermal neutron scattering data and uses new evaluated data from the CIELO project for neutron reactions on ¹H, ¹⁶O, ⁵⁶Fe, ²³⁵U, ²³⁸U and ²³⁹Pu described in companion papers in the present issue of Nuclear Data Sheets. The evaluations benefit from recent experimental data obtained in the U.S. and Europe, and improvements in theory and simulation. Notable advances include updated evaluated data for light nuclei, structural materials, actinides, fission energy release, prompt fission neutron and γ-ray spectra, thermal neutron scattering data, and charged-particle reactions. Integral validation testing is shown for a wide range of criticality, reaction rate, and neutron transmission benchmarks. In general, integral validation performance of the library is improved relative to the previous ENDF/B-VII.1 library.
With the need for improving existing nuclear data evaluations, (e.g., ENDF/B-VIII.0 and JEFF-3.3 releases) the first step was to evaluate the standards for use in such a library. This new standards evaluation made use of improved experimental data and some developments in the methodology of analysis and evaluation. In addition to the work on the traditional standards, this work produced the extension of some energy ranges and includes new reactions that are called reference cross sections. Since the effort extends beyond the traditional standards, it is called the neutron data standards evaluation. This international effort has produced new evaluations of the following cross section standards: the H(n,n), ⁶Li(n,t), ¹⁰B(n,α), ¹⁰B(n,α1γ), natC(n,n), Au(n,γ), ²³⁵U(n,f) and ²³⁸U(n,f). Also in the evaluation process the ²³⁸U(n,γ) and ²³⁹Pu(n,f) cross sections that are not standards were evaluated. Evaluations were also obtained for data that are not traditional standards: the Maxwellian spectrum averaged cross section for the Au(n,γ) cross section at 30 keV; reference cross sections for prompt γ-ray production in fast neutron-induced reactions; reference cross sections for very high energy fission cross sections; the ²⁵²Cf spontaneous fission neutron spectrum and the ²³⁵U prompt fission neutron spectrum induced by thermal incident neutrons; and the thermal neutron constants. The data and covariance matrices of the uncertainties were obtained directly from the evaluation procedure.
The normalized $^{238}$U(n,f)/$^{235}$U(n,f) cross section ratio has been measured using the NIFFTE fission Time Projection Chamber from the reaction threshold to $30$~MeV. The fissionTPC is a two-volume MICROMEGAS time projection chamber that allows for full three-dimensional reconstruction of fission-fragment ionization profiles from neutron-induced fission. The measurement was performed at the Los Alamos Neutron Science Center, where the neutron energy is determined from neutron time-of-flight. The $^{238}$U(n,f)/$^{235}$U(n,f) ratio reported here is the first cross section measurement made with the fissionTPC, and will provide new experimental data for evaluation of the $^{238}$U(n,f) cross section, an important standard used in neutron-flux measurements. Use of a development target in this work prevented the determination of an absolute normalization, to be addressed in future measurements. Instead, the measured cross section ratio has been normalized to ENDF/B-VIII.$\beta$5 at 14.5 MeV.
The MICROMEGAS (MICRO-MEsh GAseous Structure) charge amplification structure has found wide use in many detection applications, especially as a gain stage for the charge readout of Time Projection Chambers (TPCs). Here we report on the behavior of a MICROMEGAS TPC when operated in a high-energy (up to 800 MeV) neutron beam. It is found that neutron-induced reactions can cause discharges in some drift gas mixtures that are stable in the absence of the neutron beam. The discharges result from recoil ions close to the MICROMEGAS that deposit high specific ionization density and have a limited diffusion time. For a binary drift gas, increasing the percentage of the molecular component (quench gas) relative to the noble component and operating at lower pressures generally improves stability.
Fission fragment angular distributions can provide insights into the quantum mechanical state of the transition nucleus at the saddle point and are therefore an important constraint on fission theory. Available anisotropy data is sparse, especially at neutron energies above 5 MeV. For the first time, a three-dimensional tracking detector is employed to study fragment emission angles and provide a direct measurement of angular anisotropy. The Neutron Induced Fission Fragment Tracking Experiment (NIFFTE) collaboration has deployed the fission time projection chamber (fissionTPC) to measure nuclear data with unprecedented precision. The fission fragment anisotropy of $^{235}$U has been measured over a wide range of incident neutron energies from 180 keV to 200 MeV. Sources of systematic uncertainty are evaluated and the results compared with existing measurements.
We prepare high quality actinide targets for studies of neutron-induced and charged-particle-induced fission. I report on our efforts to measure fragment energy loss in the target backings and to diagnose the “crud” problem frequently found in 248Cm and 252Cf sources and targets. I discuss the preparation of multi-isotopic targets for the FissionTPC and our efforts to measure the pointing resolution of this device. The issues of target uniformity, chemical composition and radiation stability of the targets are discussed along with problems of high/low specific activity regions in a single target.
Specific uncertainty components of counting at a defined solid angle are discussed. It is potentially an extremely accurate technique for primary standardisation of activity of alpha emitters and low-energy x-ray emitters. Owing to its reproducibility, it is very well suited for half-life measurements. Considered sources of uncertainty are 1) source–detector geometry, 2) solid-angle calculation, 3) energy loss and self-absorption, 4) scattering, 5) detection efficiency. Other sources of uncertainty, such as source weighing, counting, dead time and decay data are common to other standardisation methods. Statistical uncertainty propagation formulas are presented for the solid angle subtended by a circular detector to radioactive sources. Computer simulations were performed to investigate aspects of particle scattering.
Fission cross section-ratios of 233U, 238U, 237Np, 239Pu and 232Th relative to 235U have been measured in the energy range of incoming neutrons from 1 MeV to 200 MeV using multiplate ionization chamber and time-of-flight technique at the neutron spectrometer GNEIS in Gatchina. The experimental data obtained are presented in comparison with the data of other author and theoretical calculations.
Fission cross section ratios of 233U, 238U, 232Th, 237Np, natural Pb and 209Bi to 235U have been measured in a wide energy range of incident neutrons from 1 MeV to 200 MeV using a time-of-flight technique at the neutron spectrometer GNEIS based on the 1 GeV proton synchrocyclotron of PNPI. For actinide targets, the threshold cross section method and evaluated data below 14 MeV were used for normalization of the shape measurement data, while the evaluated and recommended fission cross sections of 235U were used to convert the ratio data to absolute fission cross sections. For Pb and Bi targets, an absolute normalization of the measured cross section ratios has been done using the thickness of the targets and detection efficiencies.
A new stable version ("production version") v5.28.00 of ROOT [1] has been published [2]. It features several major improvements in many areas, most noteworthy data storage performance as well as statistics and graphics features. Some of these improvements have already been predicted in the original publication Antcheva et al. (2009) [3]. This version will be maintained for at least 6 months; new minor revisions ("patch releases") will be published [4] to solve problems reported with this version.
The Los Alamos Neutron Science Center, or LANSCE, uses the first truly high-current medium-energy proton linear accelerator, which operated originally at a beam power of 1 MW for medium-energy nuclear physics. Today LANSCE continues operation as one of the most versatile accelerator-based user facilities in the world. During eight months of annual operation, scientists from around the world work at LANSCE to execute an extraordinarily broad program of defense and civilian research. Several areas operate simultaneously. The Lujan Neutron Scattering Center (Lujan Center) is a moderated spallation source (meV to keV), the Weapons Neutron Research Facility (WNR) is a bare spallation neutron source (keV to 800 MeV), and a new ultra-cold neutron source will be operational in 2005. These sources give LANSCE the ability to produce and use neutrons with energies that range over 14 orders of magnitude. LANSCE also supplies beam to WNR and two other areas for applications requiring protons. In a proton radiography (pRad) area, a sequence of narrow proton pulses is transmitted through shocked materials and imaged to study dynamic properties. In 2005, LANSCE began operating a facility that uses 100-MeV protons to produce medical radioisotopes. To sustain a vigorous program beyond this decade, LANSCE has embarked on a project to refurbish key elements of the facility and to plan capabilities beyond those that presently exist.
Geant4 is a toolkit for simulating the passage of particles through matter. It includes a complete range of functionality including tracking, geometry, physics models and hits. The physics processes offered cover a comprehensive range, including electromagnetic, hadronic and optical processes, a large set of long-lived particles, materials and elements, over a wide energy range starting, in some cases, from 250 eV and extending in others to the TeV energy range. It has been designed and constructed to expose the physics models utilised, to handle complex geometries, and to enable its easy adaptation for optimal use in different sets of applications. The toolkit is the result of a worldwide collaboration of physicists and software engineers. It has been created exploiting software engineering and object-oriented technology and implemented in the C++ programming language. It has been used in applications in particle physics, nuclear physics, accelerator design, space engineering and medical physics.
The Neutron Induced Fission Fragment Tracking Experiment (NIFFTE) collaboration has performed measurements with a fission time projection chamber to study the fission process by reconstructing full three-dimensional tracks of fission fragments and other ionizing radiation. The amount of linear momentum imparted to the fissioning nucleus by the incident neutron can be inferred by measuring the opening angle between the fission fragments. Using this measured linear momentum, fission fragment angular distributions can be converted to the center-of-mass frame for anisotropy measurements. Angular anisotropy is an important experimental observable for understanding the quantum mechanical state of the fissioning nucleus and vital to determining detection efficiency for cross section measurements. Neutron linear momentum transfer to fissioning U235, U238, and Pu239 and fission fragment angular anisotropy of U235 and U238 as a function of neutron energies in the range 130 keV–250 MeV are presented.
Templates of uncertainties expected in specific measurement types were recently developed. One aim of these templates is to help evaluators in identifying (1) missing or suspiciously low uncertainties and (2) missing correlations between uncertainties of the same and different experiments, when estimating covariances for experimental data employed in their evaluations. These templates also provide realistic estimates of standard deviations and correlations for a particular uncertainty source and measurement type that can be used by evaluators in situations where they are not supplied by the experimenters. This information allows for a more comprehensive uncertainty analysis across all measurements considered in an evaluation and, thus, more realistic evaluated covariances. Here, we extend a template that is applicable to uncertainties expected in neutron-induced fission, (n,f), cross-section measurements. It is applied to improving covariances of ²³⁹Pu(n,f) cross-section measurements in the database underlying the Neutron Data Standards evaluations. This particular example was chosen since this evaluation is primarily based on experimental information. Also, some uncertainties of individual ²³⁹Pu(n,f) cross-section experiments in this database were suspected to be underestimated. The evaluated uncertainties obtained after updating the covariances in the database by means of the template indeed do increase compared to their original values. Even more importantly, the evaluated mean values change noticeably. These modified cross sections impact application calculations significantly, as is demonstrated by employing them in simulations of the effective neutron multiplication factor for a few selected critical assemblies. However, this updated evaluated ²³⁹Pu(n,f) cross section should not be interpreted as the final one that should replace values of the current Neutron Data Standards project. Evaluations for the Neutron Data Standards of the ²³⁹Pu(n,f) cross section must be linked to many other observables included in the associated database, most notably to cross sections for ²³⁵U(n,f), but also to those for ¹⁰B(n,α), ⁶Li(n,t), ²³⁸U(n,f), and ²³⁸U(n,γ), because of included measurements of the ²³⁹Pu(n,f) cross section that appear as ratios to these reactions. Some of these other reactions are correlated to further observables in the database. Hence, updating uncertainties of data sets of any of these observables can potentially impact the ²³⁹Pu(n,f) cross section. Uncertainties for all measurements of these linked physical observables have to be updated before a comprehensive evaluation of the ²³⁹Pu(n,f) cross section and its corresponding uncertainties can be provided.
Evaluated nuclear data uncertainties reported in the literature or archived in data libraries are often perceived as unrealistic, most often because they are thought to be too small. The impact of this issue in applied nuclear science has been discussed widely in recent years. Commonly suggested causes are: poor estimates of specific error components, neglect of uncertainty correlations, and overlooked known error sources. However, instances have been reported where very careful, objective assessments of all known error sources have been made with realistic error magnitudes and correlations provided, yet the resulting evaluated uncertainties still appear to be inconsistent with observed scatter of predicted mean values. These discrepancies might be attributed to significant unrecognized sources of uncertainty (USU) that limit the accuracy to which these physical quantities can be determined.
The objective of our work has been to develop qualitative and quantitative procedures for revealing and including USU estimates in nuclear data evaluations involving experimental input data. This paper identifies several specific clues that can be explored by evaluators in identifying the existence of USU. It then describes numerical procedures we have introduced to generate quantitative estimates of USU magnitudes. Key requirements for these procedures to be viable are that sufficient numbers of data points be available, for statistical reasons, and that additional supporting information about the measurements be provided by the experimenters. Several realistic examples are described here to illustrate these procedures and demonstrate their outcomes and limitations. Our work strongly supports the view that USU is an important issue in nuclear data evaluation, with significant consequences for applications, and that this topic warrants further investigation by the nuclear science community.
FREYA (Fission Reaction Event Yield Algorithm) is a fission event generator which models complete fission events. As such, it automatically includes fluctuations as well as correlations between observables, resulting from conservation of energy and momentum. The purpose of this paper is to present the main differences between FREYA versions 1.0 and 2.0.2 : additional fissionable isotopes, angular momentum conservation, Giant Dipole Resonance form factor for the statistical emission of photons, improved treatment of fission photon emission using RIPL database, and dependence on the incident neutron direction. FREYA 2.0.2 has been integrated into the LLNL Fission Library 2.0.2, which has itself been integrated into MCNP6.2, TRIPOLI-4.10, and can be called from Geant4.10.
New version program summary
Program title: FREYA 2.0.2
Program Files doi: ” http://dx.doi.org/10.17632/2mssy7r3gt.1”
Licensing provisions: BSD 3-clause
Programming language: Fortran 90, C++
Journal Reference of previous version: J. M Verbeke, J. Randrup, R. Vogt, “Fission Reaction Yield Algorithm FREYA for Event-by-Event Simulation of Fission,” Comp. Phys. Comm. 191, pp. 178–202; doi:10.1016/j.cpc.2015.02.002 (2015).
Does the new version supersede the previous version? Yes.
Reasons for the new version: New physics and more fissionable isotopes.
Summary of revisions: Additional fissionable isotopes, angular momentum conservation, Giant Dipole Resonance form factor for the statistical emission of photons, improved treatment of fission photon emission using RIPL database, and dependence on the incident neutron direction.
Nature of problem: Modeling of fission events.
Solution method: Simulation of complete fission events, production of secondary fission fragments, fission neutrons and photons.
Restrictions: Restricted to spontaneous fission of 238U, 238Pu, 240Pu, 242Pu, 244Cm, 252Cf; neutron-induced fission of 233U, 235U, 238U, 239Pu, 241Pu, for incident neutron energies less than 20 MeV.
The ratios of the neutron fission cross sections σf(²⁴⁰Pu)/σf(²³⁹Pu), σf(²⁴⁰Pu)/σf(²³⁵U), and σf(²³⁹Pu)/σf(²³⁵U) have been measured simultaneously with a multiplate ionization fission chamber using the Oak Ridge Electron Linear Accelerator as a neutron source over the neutron energy range from 5 keV to 20 MeV. The ²⁴⁰Pu ratio data are in overall agreement with ENDF/B-V with exceptions in relatively narrow neutron energy regions. Below 150 keV and from 10 to 20 MeV, the present ²³⁹Pu/²³⁵U fission ratios indicate significant discrepancies when compared to ENDF/B-V. These ratios are important for thermal and fast reactor applications.
Measurements of the absolute neutron fission cross sections of ²³⁵U, ²³⁸U, and ²³⁹Pu have been made at 13.9 and 14.6 MeV with a double 4π ionization chamber. The associated particle method with the time-of-flight technique was used. Our final values of σnf(²³⁵U), σnf(²³⁸U), σnf(²³⁹Pu), σnf(²³⁸U)/σnf(²³⁵U), and σnf(²³⁹Pu)/σnf(²³⁵U) are compared to previous data.
The ratio of the fission cross sections of ²³⁹Pu to that of ²³⁵U has been measured at 30 keV and from 0.5 to 5.4 MeV. A value of 0.796 ± 0.020 has been obtained at 30 keV. None of the structure reported recently by Savin et al. could be found in the MeV range in the present measurements.
The fission cross-section ratios for isotopic targets of 240Pu, 242Pu, and 244Pu relative to 235U are measured for neutron energies from 0.5 to 400 MeV and for 239Pu relative to 235U for energies from 0.85 to 62 MeV. A multiple-plate gas ionization detector was used to measure simultaneously the fission rate for each of the isotopic targets. The neutron energies were determined by the time-of-flight technique on a 20-m flight path at the Weapons Neutron Research white neutron source at the Los Alamos National Laboratory Neutron Science Center. Uncertainties are < 4% for energies < 50 MeV. This measurement provides the capability to resolve discrepancies among previous measurements for these isotopes over this energy range and are the first measurements for most of these isotopes for energies >30 MeV. The results are compared with previous measurements and to ENDF/B-VI.
This paper presents fission cross-section ratios for **2**4**1Am and **2**4**3Am relative to **2**3**5U in the neutron energy range from 0. 2 to 30 Mev. The fission cross-section ratios **2**4**1Am:**2**3**5U and **2**4**3Am:**2**3**5U were measured as a function of neutron energy using ionization fission chambers, the threshold cross-section method, and the time-of-flight technique at the Lawrence Livermore National Laboratory 100-Mev electron linear accelerator. This paper presents only the highlights of the experiment and those details that were unique to the americium samples.
The 239Pu and 241Pu neutron-induced fission cross sections have been measured from subthermal energies to 200 MeV. These measurements are part of a campaign to measure fission cross sections with high precision in support of advanced fast reactor technology. Plutonium-241 is the most active target measured in this program to date, with a half-life of 14.4 yr. The results for 239Pu are in good agreement with previous experiments and add new information to the limited knowledge on the fission cross section above 30 MeV. Discrepancies of up to 30% between the evaluations and the experimental data for 241Pu are found in the fast region, which is of particular importance for fast spectrum reactor technology, and a reevaluation of the fission cross section for this isotope is recommended.
From nuclear materials accountability to detection of special nuclear material, SNM, the need for better modeling of fission has grown over the past decades. Current radiation transport codes compute average quantities with great accuracy and performance, but performance and averaging come at the price of limited interaction-by-interaction modeling. For fission applications, these codes often lack the capability of modeling interactions exactly: energy is not conserved, energies of emitted particles are uncorrelated, prompt fission neutron and photon multiplicities are uncorrelated. Many modern applications require more exclusive quantities than averages, such as the fluctuations in certain observables (e.g. the neutron multiplicity) and correlations between neutrons and photons. The new computational model, FREYA (Fission Reaction Event Yield Algorithm), aims to meet this need by modeling complete fission events. Thus it automatically includes fluctuations as well as correlations resulting from conservation of energy and momentum. FREYA has been integrated into the LLNL Fission Library, and will soon be part of MCNPX2.7.0, MCNP6, TRIPOLI-4.9, and Geant4.10.
MCNP6 is simply and accurately described as the merger of MCNP5 and MCNPX capabilities, but it is
much more than the sum of those two computer codes. MCNP6 is the result of five years of effort by the MCNP5
and MCNPX code development teams. These groups of people, residing in Los Alamos National Laboratory’s
(LANL) X Computational Physics Division, Monte Carlo Codes Group (XCP-3), and Decision Applications Division, Radiation Transport and Applications Team (D-5), respectively, have combined their code development efforts to produce the next evolution of MCNP.While maintenance and bug fixes will continue for MCNP5 1.60 and MCNPX 2.7.0 for upcoming years, new code development capabilities only will be developed and released in MCNP6. In fact, the initial release of MCNP6 contains 16 new features not previously found in either code. These new features include the abilities to import unstructured mesh geometries from the finite element code Abaqus, to transport photons down to 1.0 eV, to transport electrons down to 10.0 eV, to model complete atomic relaxation emissions, and to generate or read mesh geometries for use with the LANL discrete ordinates code Partisn. The first release of MCNP6, MCNP6 Beta 2, is now available through the Radiation Safety Information Computational Center, and the first production release is expected in calendar year 2012. High confidence in the MCNP6 code is based on its performance with the verification and validation test suites, comparisons to its predecessor codes, the regression test suite, its code development process, and the underlying high-quality nuclear and atomic databases.
Neutron-induced fission cross sections have been measured for several isotopes of uranium and plutonium at the Los Alamos Neutron Science Center (LANSCE) over a wide range of incident neutron energies. The total uncertainties in these measurements are in the range 3–5% above 100 keV of incident neutron energy, which results from uncertainties in the target, neutron source, and detector system. The individual sources of uncertainties are assumed to be uncorrelated, however correlation in the cross section across neutron energy bins are considered. The quantification of the uncertainty contributions will be described here.
The fission Time Projection Chamber (fissionTPC) is a compact (15 cm
diameter) two-chamber MICROMEGAS TPC designed to make precision cross section
measurements of neutron-induced fission. The actinide targets are placed on the
central cathode and irradiated with a neutron beam that passes axially through
the TPC inducing fission in the target. The 4$\pi$ acceptance for fission
fragments and complete charged particle track reconstruction are powerful
features of the fissionTPC which will be used to measure fission cross sections
and examine the associated systematic errors. This paper provides a detailed
description of the design requirements, the design solutions, and the initial
performance of the fissionTPC.
Progress in experimental high‐energy physics is limited in practice by two complementary aspects: the types of beam particles available with useful intensities and energies, and the characteristics of the detection techniques available for measuring needed information about collisions of interest and their subsequent reaction products. Most impressively, advances in accelerator design over the last three decades have led to an increase in beam energies of nearly three orders of magnitude, and the advent of colliding‐beam machines has brought a comparable increase to the center‐of‐mass energy available. The diversity of useful beam species has now grown to include essentially all known particles with lifetimes greater than 10 −11 seconds . By combining the particle‐identification function in the same detector volume as the tracking and momentum‐measurement functions, this device achieves substantial reduction in overall size.
Medium to large channel count detectors are usually faced with a few unattractive options for data acquisition (DAQ). Small to medium-sized TPC experiments, for example, are too small to justify the expense and development time of application specific integrated circuits (ASIC). Commercial rack mounted electronics are too bulky and expensive for large channel counts. The combination of commercial high-speed high-density FPGAs, ADCs, and small discrete components provides another option that scales to tens of thousands of channels and is only slightly larger than ASICs using off-the-shelf components. A working example of this alternative solution is presented.
This paper has been written to provide experimental nuclear data researchers and data compilers with practical guidance on dealing with experimental nuclear reaction data uncertainties. It outlines some of the properties of random variables as well as principles of data uncertainty estimation, and illustrates them by means of simple examples which are relevant to the field of nuclear data. Emphasis is placed on the importance of generating mathematical models (or algorithms) that can adequately represent individual experiments for the purpose of estimating uncertainties in their results. Several types of uncertainties typically encountered in nuclear data experiments are discussed. The requirements and procedures for reporting information on measurement uncertainties for neutron reaction data, so that they will be useful in practical applications, are addressed. Consideration is given to the challenges and opportunities offered by reports, conference proceedings, journal articles, and computer libraries as vehicles for reporting and documenting numerical experimental data. Finally, contemporary formats used to compile reported experimental covariance data in the widely used library EXFOR are discussed, and several samples of EXFOR files are presented to demonstrate their use.
Recent interest from data users on applications that utilize the uncertainties of evaluated nuclear reaction data has stimulated the data evaluation community to focus on producing covariance data to a far greater extent than ever before. Although some uncertainty information has been available in the ENDF/B libraries since the 1970ʼs, this content has been fairly limited in scope, the quality quite variable, and the use of covariance data confined to only a few application areas. Today, covariance data are more widely and extensively utilized than ever before in neutron dosimetry, in advanced fission reactor design studies, in nuclear criticality safety assessments, in national security applications, and even in certain fusion energy applications. The main problem that now faces the ENDF/B evaluator community is that of providing covariances that are adequate both in quantity and quality to meet the requirements of contemporary nuclear data users in a timely manner. In broad terms, the approach pursued during the past several years has been to purge any legacy covariance information contained in ENDF/B-VI.8 that was judged to be subpar, to include in ENDF/B-VII.0 (released in 2006) only those covariance data deemed then to be of reasonable quality for contemporary applications, and to subsequently devote as much effort as the available time and resources allowed to producing additional covariance data of suitable scope and quality for inclusion in ENDF/B-VII.1. Considerable attention has also been devoted during the five years since the release of ENDF/B-VII.0 to examining and improving the methods used to produce covariance data from thermal energies up to the highest energies addressed in the ENDF/B library, to processing these data in a robust fashion so that they can be utilized readily in contemporary nuclear applications, and to developing convenient covariance data visualization capabilities. Other papers included in this issue discuss in considerable detail various aspects of the data producer communityʼs efforts to improve the evaluation methods and to add covariance content to the ENDF/B library. The present paper offers just a brief glimpse of these activities by drawing material from covariance papers presented at meetings, workshops and international conferences during the past five years. Highlighted are: advances in methods for producing and processing covariance data, recently developed covariance visualization capabilities, and the development and implementation of quality assurance (QA) requirements that should be satisfied for covariance data to be included in ENDF/B-VII.1.
The ENDF/B-VII.1 library is our latest recommended evaluated nuclear data file for use in nuclear science and technology applications, and incorporates advances made in the five years since the release of ENDF/B-VII.0. These advances focus on neutron cross sections, covariances, fission product yields and decay data, and represent work by the US Cross Section Evaluation Working Group (CSEWG) in nuclear data evaluation that utilizes developments in nuclear theory, modeling, simulation, and experiment.
Actinide samples were characterized in an interlaboratory comparison between IRMM and NIST, including alpha-particle counting at defined low solid angle and counting in a 2π proportional gas counter. For this comparison, nine 233UF4 samples with high uniformity in the layer thickness were prepared at IRMM by deposition under vacuum. Polished silicon wafers were used as source substrates, and these were rotated during the deposition using a planetary rotation system.The estimated uncertainties1All uncertainties in this paper correspond to one standard deviation.1 for the defined low solid-angle methods were about 0.1% at both NIST and IRMM. The agreement of reported α-particle emission rates in the energy range 2.5–5.09MeV was better than or equal to 0.02% for the defined solid-angle methods. When comparing total α-particle emission rates over the larger energy range 0–9MeV (which includes all emissions from the daughter nuclides and the impurities), the agreement of the defined solid-angle methods was better than or equal to 0.05%.The 2π proportional gas counter results were about 0.75% higher than those of the defined solid-angle methods. A re-examination of the correction for scattering from the polished silicon wafer substrates was undertaken to explain this discrepancy. Preliminary results of the recalculated 2π proportional gas counter results are also given.
A twin ionization chamber for fission fragment detection is described. The detector permits measurement of the two fission fragment kinetic energies in an advantageous 2×2π geometry with an energy resolution of <0.5 MeV. The fission fragment emission angle θ with respect to the symmetry axis of the chamber is measured with a resolution in cos θ of <0.05. The fission fragment nuclear charge distributions can be determined and a timing signal can be extracted which allows a determination of the instant of fission with a time jitter of <0.7 ns. A pulse pileup rejection technique was developed which reduces pulse pileup by more than a factor 30. The electronic treatment of the chamber pulses and the data handling procedures including several of the necessary corrections are described in detail.
A method is given for calculating the inefficiency of a fission chamber because of the loss of fragments in the foil. The method includes the effects of the anisotropy in the fragment angular distribution because of both angular and linear momentum given to the nucleus by the incident neutron. A numerical example for a 239Pu chamber shows that the linear-momentum effect is larger than the angular-momentum effect, but less than 1% at neutron energies below 15 MeV.
Until recently there has been very little cross section data for neutron-induced fission in the intermediate energy region, primarily because no suitable neutron source has existed. At Los Alamos, the WNR target-4 facility provides a high-intensity source of neutrons nearly ideal for fission measurements extending from a fraction of a MeV to several hundred MeV. This paper summarizes the status of fission cross section data in the intermediate energy range (En > 30 MeV) and presents our fission cross section data for {sup 235}U and {sup 238}U compared to intranuclear cascade and statistical model predictions.
The fission cross-section ratios of /sup 233/U to /sup 235/U and /sup 239/Pu to /sup 235/U were measured over the neutron energy range from 1 keV to 30 MeV at the Lawrence Livermore Laboratory 100-MeV Linac. Ionization fission chambers and the time-of-flight technique were used to take data simultaneously over the entire energy range. This provided accurate determinations of the shape versus neutron energy of the ratios. Two independent methods were used to determine the average value of each ratio in the interval from 1.75 to 4.0 MeV for the purpose of normalization. Over the 1-keV to 30-MeV range, the total uncertainties for the /sup 233/U-to-/sup 235/U data range from 2 to 4%; the /sup 239/Pu-to-/sup 235/U data uncertainties range from 1 to 4%.
Capabilities of spectroscopic ion beam analysis (IBA) techniques that are available in ion microprobe facilities can be greatly improved by the use of digital pulse processing. We report here development of a digital multi parameter data acquisition system suitable for IBA imaging applications. Input signals from charge sensitive preamplifier are conditioned by using a simple circuit and digitized with fast ADCs. The digitally converted signals are processed in real time using FPGA. Implementation of several components of the system is presented.
SRIM is a software package concerning the Stopping and Range of Ions in Matter. Since its introduction in 1985, major upgrades are made about every six years. Currently, more than 700 scientific citations are made to SRIM every year. For SRIM-2010, the following major improvements have been made: (1) About 2800 new experimental stopping powers were added to the database, increasing it to over 28,000 stopping values. (2) Improved corrections were made for the stopping of ions in compounds. (3) New heavy ion stopping calculations have led to significant improvements on SRIM stopping accuracy. (4) A self-contained SRIM module has been included to allow SRIM stopping and range values to be controlled and read by other software applications. (5) Individual interatomic potentials have been included for all ion/atom collisions, and these potentials are now included in the SRIM package. A full catalog of stopping power plots can be downloaded at www.SRIM.org. Over 500 plots show the accuracy of the stopping and ranges produced by SRIM along with 27,000 experimental data points. References to the citations which reported the experimental data are included.
Fission cross sections of a number of fissile isotopes have been measured relative to the fission cross section of 235U to an accuracy of approximately ± 2 % at neutron energies of 1·0, 2·25, 5·4 and 14·1 MeV. Combining these ratios with the known values of the fission cross section of 235U leads to fission cross sections having an estimated uncertainty of ± 3·5 % and which are mostly in agreement with other recent measurements.
The construction of a neutron flux monitor that can measure absolute neutron intensities in the neutron energy range from below 1 MeV to over 500 MeV is described. The detector consists of an ionization chamber with several thin deposits of fissionable material. The ionization chamber is thin enough that it does not significantly affect the neutron beam and may be left in the neutron flight path during experimental measurements to continuously monitor the beam flux. The use of this monitor at the continuous-energy spallation neutron source at the WNR target area at LAMPF is described.
We describe a novel structure for a gaseous detector that is under development at Saclay. It consists of a two-stage parallel-plate avalanche chamber of small amplification gap (100 microm) combined with a conversion-drift space. It allows a fast removal of positive ionsproduced during the avalanche development. Fast signals (3/4 1 ns) are obtained duirng the collection of the electron avalanche on the anode microstrip plane. The induced positive ion signal has a rise time of 100 ns. The fast evacuation of positive ions combined with the high granularity of the detector provide a high rate capability. Gas gains of up to $10^5$ have been achieved.
Cross sections ratios of 92-U-235, 94-Pu-239, 94-Pu-240 fission with fast neutrons
- Savin
The vapor deposition of high specific activity actinides
- Silveira
An image synthesizer
- Perlin
Neutron induced fission cross section ratios for 232Th, 235,236U, 237Np, and 239Pu from 1 to 400 MeV
- Lisowski