Fission Cross Sections Of 239Pu Research Articles

Evaluating 239Pu(n,f) cross sections via machine learning using experimental data, covariances, and measurement features

The neutron-induced 239Pu fission cross section, 239Pu(n,f), is evaluated from 1–20 MeV using experimental data and associated covariances while also considering information on the measurement, termed features here. For instance, methods to determine the background, sample backing material, or impurities in the sample, are explicitly taken into account in the evaluation process. To this end, outliers in the experimental data are identified with a modified version of the Hybrid Robust Support Vector Machine. In a second step, two machine learning methods (logistic regression with elastic net regularization and random forest regression with SHAP feature importance metric) are used to highlight measurement features that are common among many of the outlying data points. Based on this analysis, penalty uncertainties are added to the experimental covariances of outlying data points that have outlier measurement features and are put through the generalized-least-squares evaluation. The resulting evaluated mean values and covariances differ distinctly from those data evaluated without the penalty uncertainties. These results highlight that certain measurement features should be more closely examined.

Covariance generation for the prompt neutron multiplicity of 239Pu including the (n,γf) process

Fission cross section of 239Pu can be seen as a sum of the “immediate" fission and “two-step" (n,γf) reactions. In the Resolved Resonance Range of the reaction cross sections, the contribution of the (n,γf) process has an impact on the determination of the partial widths magnitude involved in the Reich-Moore approximation of the R-matrix theory. The present work aims to investigate this impact by using the CONRAD code and the partial width Γγf for the (n,γf) reaction calculated by Lynn et al. [1]. A special attention will be paid to the covariance matrix obtained on νp.

Comparison of ENDF/B-VIII.0 and ENDF/B-VII.1 in criticality, depletion benchmark, and uncertainty analyses by McCARD

The McCARD criticality, depletion, and S/U analyses with the newly published evaluated nuclear data file (ENDF) library ENDF/B-VIII.0 are conducted and their results are compared with those using the previous version of ENDF library, ENDF/B-VII.1. First of all, to verify and validate the newly generated ENDF/B-VIII.0 continuous energy cross section library in the ACE format, 60 cases from International Criticality Safety Benchmark Problems (ICSBEP) and VENUS-1 criticality benchmark analyses are performed using the McCARD code. Overall, the prediction performance of criticality is retained except for the high enriched uranium (HEU) fast and plutonium-based (PU) thermal benchmarks. The large difference in keff for the HEU fast and PU thermal benchmarks comes from the increase of the 238U capture cross sections in the fast energy region and the decrease of 239Pu fission cross section in the thermal energy region, respectively. In the results for the VERA depletion benchmarks, the large differences at the end of the burnup can be attributed to the difference of the one-group 235U absorption cross section. It leads to the increases of the difference in kinf between ENDF/B-VIII.0 and ENDF/B-VII.1 as the burnup proceeds. In the ENDF/B-VIII.0 evaluated nuclear data library, there have been considerable changes in the covariance data. The changes have a strong correlation with the decreases in the total uncertainty of multiplication factor for Godiva and the TMI-1 pin problem.

Study of the (n,γf) process on 239Pu

In the neutron resonance range, fission cross section of 239Pu can be seen as a sum of the immediate (n,f) and the two-step (n,γf) fission reactions. In that case, five channel widths should be considered for a proper evaluation, those are: two opened fission channels for Jπ = 0+, one opened fission channel for Jπ = 1+ and two J-dependent for the (n,γf) reaction. The sizeable contribution of the (n,γf) process should have an impact on the determination of the capture and fission widths involved in the Reich-Moore approximation of the R-matrix theory. The present work aims to investigate this impact by using the CONRAD code and the Γγf available from literature. Prompt neutron multiplicity (νp) has been also reproduced including the contributions of the (n,γf) process.

Nuclear data sensitivity and uncertainty analysis of effective neutron multiplication factor in various MYRRHA core configurations

A sensitivity and uncertainty analysis was carried out to estimate the uncertainty in the neutron multiplication factor keff and to identify the most important nuclear data for neutron induced reactions for criticality calculations of the latest MYRRHA designs. Sensitivity profiles, i.e. sensitivity to the nuclear data as a function of incoming neutron energy, were derived for both a critical and sub-critical core. They were calculated using codes that are based on different methodologies including stochastic and deterministic calculations (i.e. SCALE, MCNP and XSUN). The neutron induced nuclear data sensitivity analysis outlined the following quantities to be of special importance for the MYRRHA reactor concept: 239Pu(n,γ) both in resonance and fast energy region, (n,f) fast, χ and ν fast; 238U(n,n′) fast, (n,γ) resonance and fast, (n,n) resonance and fast; 240Pu ν fast; 238Pu(n,f) both resonance and fast; 56Fe(n,γ) both resonance and fast. Differences of less than 4% between codes were obtained for these quantities, with few exceptions (238Pu(n,f), 238U(n,n) and 56Fe(n,γ) reactions). Nuclear data covariance matrices of different libraries (SCALE-6, COMMARA-2 and JENDL-4.0m) were used to derive the uncertainty in keff based on the calculated sensitivities. This study reveals that the largest contributions to keff uncertainty result from the uncertainty in the average prompt neutron fission multiplicity of 239Pu, in the 238U inelastic scattering cross section and 239Pu fission cross section, using the covariances from SCALE-6, COMMARA-2 and JENDL-4.0m, respectively.

Fission Product Yields for 14 MeV Neutrons on 235U, 238U and 239Pu

We report cumulative fission product yields (FPY) measured at Los Alamos for 14 MeV neutrons on 235U, 238U and 239Pu. The results are from historical measurements made in the 1950s–1970s, not previously available in the peer reviewed literature, although an early version of the data was reported in the Ford and Norris review. The results are compared with other measurements and with the ENDF/B-VI England and Rider evaluation. Compared to the Laurec (CEA) data and to ENDF/B-VI evaluation, good agreement is seen for 235U and 238U, but our FPYs are generally higher for 239Pu. The reason for the higher plutonium FPYs compared to earlier Los Alamos assessments reported by Ford and Norris is that we update the measured values to use modern nuclear data, and in particular the 14 MeV 239Pu fission cross section is now known to be 15–20% lower than the value assumed in the 1950s, and therefore our assessed number of fissions in the plutonium sample is correspondingly lower. Our results are in excellent agreement with absolute FPY measurements by Nethaway (1971), although Nethaway later renormalized his data down by 9% having hypothesized that he had a normalization error. The new ENDF/B-VII.1 14 MeV FPY evaluation is in good agreement with our data.

Reactivity Effects of Differences between JEFF-3.1 and ENDF/B-VI.8 in Analysis of Six MASURCA Cores of the R-Z Program

Reactivity contributions of differences between JEFF-3.1 and ENDF/B-VI.8 were analyzed for six early MASURCA cores of the R-Z program using ERANOS 2.1. These cores were designed such that their neutron spectra would emulate that of an oxide-fueled sodium-cooled fast reactor, some containing enriched uranium and others containing depleted uranium and plutonium. Effects of modeling assumptions and solution methods both in ECCO lattice calculations and in BISTRO Sn flux solutions were first evaluated using JEFF-3.1 cross-section libraries. Comparisons were made between calculated and measured values for reactivity and several spectral indices. Reactivity effects of differences between JEFF-3.1 and ENDF/B-VI.8 were also quantified using perturbation theory analysis. The most important nuclide with respect to reactivity differences between cross-section libraries was 23Na, primarily a result of differences in the angular dependence of elastic scattering, which is more forward peaked in ENDF/B-VI.8 than in JEFF-3.1. Differences in 23Na inelastic scattering cross sections between libraries also generated significant differences in reactivity, more due to the differences in magnitude of the cross sections than to the angular dependence. The nuclide 238U was also found to be important with regard to reactivity differences between the two libraries mostly due to a large effect of inelastic scattering differences and two smaller effects of elastic scattering and fission cross sections. In the cores that contained plutonium, 239Pu fission cross-section differences contributed significantly to the reactivity differences between libraries.

239Pu prompt fission neutron spectra

Calculations of 239Pu(n, F) prompt fission neutron spectra have been performed for neutron energy up to 20 MeV. The exclusive spectra of pre-fission neutron reactions (n, xnf) were calculated on the basis of the Hauser-Feshbach model simultaneously with the cross sections of (n, F) and (n, 2n) reactions. The spectra of neutrons emitted by fission fragments were approximated by a sum of two Watt distributions. The components of the prompt fission neutron spectra due to pre-fission neutrons are manifested in the prompt fission neutron spectra and the average neutron energy. A correlation is established between this effect in the contribution of emissive fission (n, xnf) in the fission cross-section of 239Pu(n, F) and 235U(n, F). It is shown that the 239Pu(n, F) prompt fission neutron spectra used in applied calculations do not correspond to the experimental differential data and the systematic regularities in the spectra and their average energy found for the most carefully studied nuclei 235,238U and 232Th.

Feasability of using a graphite slowing-down-time spectrometer in the nondestructive assay of nuclear materials

A slowing-down-time spectrometer (SDTS), constructed for the study of nondestructive assay of fissile nuclear materials, is in its early stages of operation at the Nuclear Engineering Teaching Laboratory of the University of Texas at Austin. The spectrometer is made of a 101×105×122 cm 3 graphite rectangular parallelepiped and is based on injecting pulses of 14 MeV neutrons into the pile. The neutron source is a Texas Nuclear Corporation neutron generator that produces neutrons via the D–T reaction. Measurements and calculations have been conducted to study the time behavior of the neutrons and the assay capabilities of this spectrometer. A 3He detector covered with 0.8 mm of Cd was used to perform neutron die-away measurements in the graphite. A 4He detector was used to perform an assay of a fuel pin containing 13% by mass 239Pu. The calculations were made using the MCNP4B code, and a realistic 3-D mock-up of the experiment. Good agreement is found to exist between the measured time spectra and the ones predicted from the calculations. This included the experimental observation of a predicted rapid fall-off in the die-away spectrum due to the existence of the cadmium cover. Therefore, the time–energy coupling that is expected to hold in a SDTS is experimentally verified. In addition, the calculations and the measurements indicate that the interrogation of fissile materials is possible using a graphite SDTS. This is clear in the measured and calculated time signatures for the 239Pu sample, which include structure that reflects the two resonance groups and the 0.3 eV resonance peak observed in the 239Pu fission cross section.

Measurements of the Neutron Fission Cross Sections of235U (En= 0.01 eV to 30 keV) and239Pu (En= 0.01 to 60 eV)

Measurements were made of the energy dependence of the 235U neutron fission cross section over the energy range from 0.01 eV to 30 keV and of the 239Pu fission cross section over the range from 0.0...

Multilevel parametrization of the total cross section and fission cross section of239Pu in the resonance region of neutron energies

Multilevel parametrization of the total cross section and fission cross section of239Pu in the resonance region of neutron energies

Measurements of the ratio between the capture and fission cross sections of239Pu

Measurements of the ratio between the capture and fission cross sections of239Pu

Application of distribution-free statistics to the structural analysis of slow neutron cross section and resonance parameter data

The discovery of narrow intermediate structure and its explanation in terms of shape isomerism has encouraged the search for structure in the energy dependence of low-energy neutron cross-section and resonance parameter data. It is shown how the Wald and Wolfowitz distribution-free statistic can be applied to test for the energy-dependent structure of resonance parameter data in the presence of the well-known fluctuations which these parameters undergo from resonance to resonance. The statistic can also be used on a sequence of serial correlation coefficients and thereby test the assumed energy dependence of cross-section data. Illustrative examples refer to the neutron widths of 232Th, the fission and neutron widths of 234U and the serial correlation coefficients derived from the fission cross sections of 239Pu and 235U.

Detecteur de fission a basse temperature (77°K)

We describe a fission detector which has been developed for measurements with high resolution of the neutron induced fission cross section of very radioactive fissile nuclei. In this apparatus, detection is accomplished by observing with photomultipliers the scintillations produced by the fission fragments in a mixture of helium and nitrogen. The fissile material is cooled to a temperature of 77°K to reduce Doppler broadening. In order to obtain good statistical accuracy compatible with high resolution, the detector is designed to contain a large quantity of fissile material (e.g. 1 g of 239Pu). The first application of this detector was the measurement of the fission cross section of 239Pu by the time-of-flight method using the Saclay 45 MeV linear accelerator as a pulse neutron source with a nominal resolution of 1 ns/m. The results obtained are presented and compared with previous measurements.

Multilevel Parameters for the 239Pu Fission Cross Section from 40 to 100 eV

Multilevel Parameters for the ²³⁹Pu Fission Cross Section from 40 to 100 eV

A correlation analysis of the slow-neutron induced fission cross section of 239Pu

A correlation analysis, of the type suggested by Egelstaff, has been made on the results of high-resolution measurement of the fission cross section of 239Pu. Highly significant correlations are observed and interpreted as being due to fission occurring through levels, with a spacing of 460 ± 80 eV, built on the second minimum in the fission potential as predicted by Strutinsky.