Resolved Resonance Region Research Articles

Early applications of the R-matrix SAMMY code for charged-particle induced reactions and related covariances

The SAMMY code system is mainly used in nuclear data evaluations for incident neutrons in the resolved resonance region (RRR), however, built-in capabilities also allow the code to describe the resonance structure produced by other incident particles, including charged particles. (α,n) data provide fundamental information that underpins nuclear modeling and simulation software, such as ORIGEN and SOURCES4C, used for the analysis of neutron emission and definition of source emission processes. The goal of this work is to carry out evaluations of charged-particle-induced reaction cross sections in the RRR. The SAMMY code was recently used in this regard to generate a Reich-Moore parameterization of the available 17,18 O(α,n) experimental cross sections in order to estimate the uncertainty in the neutron generation rates for uranium oxide fuel types. This paper provides a brief description of the SAMMY evaluation procedure for the treatment of 17,18 O(α,n) reaction cross sections. The results are used to generate neutron source rates for a plutonium oxide matrix.

Measurements of the capture cross sections of natural silver in the resonance range with the time of flight technique

Neutron capture cross section measurements have been performed at the time-of-flight facility GELINA of the EC-JRC-Geel. Prompt gamma rays, originating from a natural silver sample, were detected by a pair of C6 D6 liquid scintillation detectors. The total energy detection principle in combination with the pulse height weighting technique has been used. In this contribution the experimental details together with the data reduction process are described. In addition, first results of calculations with REFIT are presented to verify the quality of recommended cross section data in the resolved resonance region.

The 236U neutron capture cross-section measured at the n_TOF CERN facility

The 236U isotope plays an important role in nuclear systems, both for future and currently operating ones. The actual knowledge of the capture reaction of this isotope is satisfactory in the thermal region, but it is considered insufficient for Fast Reactor and ADS applications. For this reason the 236U(n, gamma ) reaction cross section has been measured for the first time in the whole energy region from thermal energy up to 1MeV at the n TOF facility with two different detection systems: an array of C6D6 detectors, employing the total energy deposited method, and a 4p total absorption calorimeter (TAC), made of 40 BaF2 crystals. The two n TOF data sets agree with each other within the statistical uncertainty in the Resolved Resonance Region up to 800 eV, while sizable differences (up to 20%) are found relative to the current evaluated data libraries. Moreover two new resonances have been found in the n TOF data. In the Unresolved Resonance Region up to 200 keV, the n TOF results show a reasonable agreement with previous measurements and evaluated data.

Recovery of the Point Pseudoresonance Structure of Cross Sections in the Region of Unresolved Resonances

A model is proposed for recovering the point cross sections in the region of unresolved resonances. The approach consists in using probability tables P i (σ) to construct the effective cross sections σ ′ (E) on some choice of energy intervals [E i , E i+1]. In constructing the effective cross sections the densities P i (σ′) and P i (σ) are preserved. This makes it possible to preserve on the intervals [E i , E i+1] characteristics of the neutron–matter interactions such as the average cross section, average free path, transmission function, and others. The use of pseudoresonances in the field of unresolved resonances is not distinguished from the use of ordinary point cross sections in the region of resolved resonances. The proposed approach does not require modifications of the library formats and modeling schemes for the neutron trajectories in Monte Carlo programs. The calculations of two problems showed good agreement between the calculations of the change ΔK eff using probability tables and the MCNP code and calculations using the PRIZMA code, in which pseudoresonances were used.

Evaluation of neutron nuclear data on platinum isotopes

ABSTRACTNeutron nuclear data on 10 isotopes of platinum have been evaluated for the next version of Japanese Evaluated Nuclear Data Library general-purpose file in the energy region from 10−5 eV to 20 MeV. Resolved resonance parameters of naturally occurring isotopes were taken from a compilation work, while unresolved resonance parameters were obtained by fitting to the total and capture cross sections calculated from nuclear models. A statistical model code was applied to evaluate cross sections above the resolved resonance region. Compound, pre-equilibrium and direct-reaction processes were considered for cross-section calculation. Coupled-channel optical model parameters were employed for the interaction between neutrons and nuclei. Giant-dipole and pygmy resonance parameters for E1 γ-ray transition from platinum isotopes were determined so as to reproduce measured γ-ray spectrum. The present results reproduce experimental data very well. The evaluated data are compiled into Evaluated Nuclear Data File formatted data files.

Resonance Parameter Adjustment Based on Integral Experiments

This project seeks to allow coupling of differential and integral data evaluation in a continuous-energy framework and to use the generalized linear least-squares (GLLS) methodology in the TSURFER module of the SCALE code package to update the parameters of a resolved resonance region evaluation. Recognizing that the GLLS methodology in TSURFER is identical to the mathematical description of a Bayesian update in SAMMY, the SAMINT code was created to use the mathematical machinery of SAMMY to update resolved resonance parameters based on integral data. Traditionally, SAMMY used differential experimental data to adjust nuclear data parameters. Integral experimental data, such as in the International Criticality Safety Benchmark Experiments Project, remain a tool for validation of completed nuclear data evaluations. SAMINT extracts information from integral benchmarks to aid the nuclear data evaluation process. Later, integral data can be used to resolve any remaining ambiguity between differential data sets, highlight troublesome energy regions, determine key nuclear data parameters for integral benchmark calculations, and improve the nuclear data covariance matrix evaluation. SAMINT is not intended to bias nuclear data toward specific integral experiments but should be used to supplement the evaluation of differential experimental data. Using GLLS ensures proper weight is given to the differential data.

Evaluation of neutron nuclear data on mercury isotopes

ABSTRACTNeutron nuclear data on 10 isotopes of mercury have been evaluated for the next version of Japanese evaluated nuclear data library (JENDL) general-purpose file in the energy region from 10−5 eV to 20 MeV. Resolved resonance parameters (RRPs) for 200, 202Hg were supplemented with the data which had not been considered in the previous library. Unresolved resonance parameters (URPs) were obtained by fitting to the total and capture cross sections calculated from nuclear models. A statistical model code was applied to evaluate cross sections above the resolved resonance region. Coupled-channel optical model parameters were employed for the interaction between neutrons and nuclei except 203Hg. Compound, pre-equilibrium, and direct-reaction processes were considered for cross-section calculation in the high energy region. The present results reproduce experimental data very well. The evaluated data are compiled into evaluated nuclear data file (ENDF) formatted data files.

Evaluation of neutron nuclear data on tantalum isotopes

Neutron nuclear data on four isotopes of tantalum have been evaluated for the next version of Japanese Evaluated Nuclear Data Library general-purpose file in the energy region from 10−5 eV to 20 MeV. Unresolved resonance parameters were obtained by fitting to the total and capture cross sections calculated from nuclear models, while resolved resonance parameters were selected from experimental data. A statistical model code was applied to evaluate cross sections above the resolved resonance region. Compound, pre-equilibrium and direct-reaction processes were considered for cross-section calculation. Coupled-channel optical-model parameters were employed for the interaction between neutrons and nuclei. Giant-dipole and pygmy resonance parameters for E1 γ-ray transition from tantalum isotopes were determined so as to reproduce measured γ-ray spectrum for 181Ta. The present results reproduce experimental data very well. The evaluated data are compiled into Evaluated Nuclear Data File (ENDF) formatted data files.

Evaluation of neutron nuclear data on iodine isotopes

Neutron nuclear data on six isotopes of iodine have been evaluated for the next version of Japanese Evaluated Nuclear Data Library (JENDL) general-purpose file in the energy region from 10−5 eV to 20 MeV. Unresolved resonance parameters were obtained by fitting to the total and capture cross sections calculated from nuclear models, while resolved resonance parameters remain unchanged from JENDL-4.0. A statistical model code was applied to evaluate the cross sections above the resolved resonance region. Coupled-channel optical model parameters were employed for the interaction between neutrons and nuclei. Compound, pre-equilibrium and direct-reaction processes were considered for cross-section calculation in high-energy region. Giant-dipole resonance parameters for γ-ray transition from iodine isotopes were determined so as to reproduce measured γ-ray spectrum for 127I. The present results reproduce experimental data very well. The evaluated data are compiled into ENDF-formatted data files.

Experimental neutron capture data of 58Ni from the CERN n_TOF facility

The neutron capture cross section of 58 Ni was measured at the neutron time of flight facility n_TOF at CERN, from 27 meV to 400 keV neutron energy. Special care has been taken to identify all the possible sources of background, with the so-called neutron background obtained for the first time using high-precision GEANT4 simulations. The energy range up to 122 keV was treated as the resolved resonance region, where 51 resonances were identified and analyzed by a multilevel R -matrix code SAMMY. Above 122 keV the code SESH was used in analyzing the unresolved resonance region of the capture yield. Maxwellian averaged cross sections were calculated in the temperature range of kT = 5 – 100 keV, and their astrophysical implications were investigated.

On the Methodology to Calculate the Covariance of Estimated Resonance Parameters

Principles to determine resonance parameters and their covariance from experimental data are discussed. Different methods to propagate the covariance of experimental parameters are compared. A full Bayesian statistical analysis reveals that the level to which the initial uncertainty of the experimental parameters propagates, strongly depends on the experimental conditions. For high precision data the initial uncertainties of experimental parameters, like a normalization factor, has almost no impact on the covariance of the parameters in case of thick sample measurements and conventional uncertainty propagation or full Bayesian analysis. The covariances derived from a full Bayesian analysis and least-squares fit are derived under the condition that the model describing the experimental observables is perfect. When the quality of the model can not be verified a more conservative method based on a renormalization of the covariance matrix is recommended to propagate fully the uncertainty of experimental systematic effects. Finally, neutron resonance transmission analysis is proposed as an accurate method to validate evaluated data libraries in the resolved resonance region.

Measurement and analysis of theAm243neutron capture cross section at the n_TOF facility at CERN

Background:The design of new nuclear reactors and transmutation devices requires to reduce the present neutron cross section uncertainties of minor actinides. Purpose: Reduce the $^{243}$Am(n,$\gamma$) cross section uncertainty. Method: The $^{243}$Am(n,$\gamma$) cross section has been measured at the n_TOF facility at CERN with a BaF$_{2}$ Total Absorption Calorimeter, in the energy range between 0.7 eV and 2.5 keV. Results: The $^{243}$Am(n,$\gamma$) cross section has been successfully measured in the mentioned energy range. The resolved resonance region has been extended from 250 eV up to 400 eV. In the unresolved resonance region our results are compatible with one of the two incompatible capture data sets available below 2.5 keV. The data available in EXFOR and in the literature has been used to perform a simple analysis above 2.5 keV. Conclusions: The results of this measurement contribute to reduce the $^{243}$Am(n,$\gamma$) cross section uncertainty and suggest that this cross section is underestimated up to 25% in the neutron energy range between 50 eV and a few keV in the present evaluated data libraries.

Evaluation of neutron nuclear data on tellurium isotopes

Neutron nuclear data on 13 Te isotopes have been evaluated for the next release of Japanese Evaluated Nuclear Data Library (JENDL) general purpose file in the energy region from 10− 5 eV to 20 MeV. Thermal capture cross sections of 120, 121m, 127m, 129m, 131m, 132Te were determined from the latest measurements or from a simplified formula, although the resolved resonance parameters, which describe the low-energy behavior of cross sections, remain unchanged from JENDL-4.0 for 122, 123, 124, 125, 126, 128, 130Te. A statistical model code was applied to evaluate the cross sections above the resolved resonance region. Coupled-channel optical model parameters were employed for the interaction between neutrons and nuclei. Compound, pre-equilibrium, and direct-reaction processes were considered for cross-section calculation in high-energy region. The present results reproduce experimental data very well, and are found to be much better than the JENDL-4.0 data. The evaluated data are compiled into evaluated nuclear data file (ENDF)-formatted data files.

Neutron elastic scattering angular distribution in the resolved and unresolved resonance regions

We derive a simple relationship between observed total cross sections and elastic scattering angular distributions for neutron-induced reactions in the fast energy range by combining resonance theory and the optical model (OM). This relationship enables us to estimate the anisotropy in the scattering angular distribution when experimental total cross-section data are available. We apply this method to the angular distributions of 58Ni and 56Fe and compare with the evaluated values which are based on the experimental data. We also explore the method with 90Zr for which the multi-level Breit-Wigner resonance parameters are given.

Working Party on International Nuclear Data Evaluation Cooperation (WPEC)

The OECD Nuclear Energy Agency (NEA) organizes cooperation between the major nuclear data evaluation projects in the world. The NEA Working Party on International Nuclear Data Evaluation Cooperation (WPEC) was established to promote the exchange of information on nuclear data evaluation, measurement, nuclear model calculation, validation, and related topics, and to provide a framework for cooperative activities between the participating projects. The working party assesses nuclear data improvement needs and addresses these needs by initiating joint activities in the framework of dedicated WPEC subgroups. Studies recently completed comprise a number of works related to nuclear data covariance and associated processing issues, as well as more specific studies related to the resonance parameter representation in the unresolved resonance region, the gamma production from fission product capture reactions, the 235U capture cross section, the EXFOR database, and the improvement of nuclear data for advanced reactor systems. Ongoing activities focus on the evaluation of 239Pu in the resonance region, scattering angular distribution in the fast energy range, and reporting/usage of experimental data for evaluation in the resolved resonance region. New activities include two subgroups on improved fission product yield evaluation methodologies and on modern nuclear database structures. Future activities under discussion include a pilot project for a Collaborative International Evaluated Library Organization (CIELO) and methods to provide feedback from nuclear and covariance data adjustment for improvement of nuclear data. In addition to the above mentioned short-term task-oriented subgroups, WPEC also hosts a longer-term subgroup charged with reviewing and compiling the most important nuclear data requirements in a high priority request list (HPRL).

Measurement of the 54,57Fe(n,γ) Cross Section in the Resolved Resonance Region at CERN n_TOF

54Fe and 57Fe are stable iron isotopes, which play an important role in the nucleosynthesis of the slow neutron capture process (s process). In addition, these nuclei are present in many structural materials, and therefore, the knowledge of their neutron capture cross sections is of importance for reactor design studies. This contribution summarizes the results of the (n, γ) cross sections of these two isotopes in the resolved resonance region. The experiment was carried out at the CERN n_TOF facility using the Pulse-Height Weighting Technique in combination with an experimental set-up optmized for reducing neutron induced backgrounds.

Neutron Capture Cross Section of 239Pu

The Detector for Advanced Neutron Capture Experiments (DANCE) has been used to measure the 239Pu(n,γ) cross section from 10 eV to the keV region. Three experimental run conditions were used to characterize the prompt fission γ-ray spectrum across the entire energy regime, measure the cross section in the resolved resonance region, and obtain necessary count rate well into the keV region. The preliminary cross sections are in good agreement with current evaluations from 10 eV to 80 keV.

High-precision Measurement of the 238U(n,γ) Cross Section with the Total Absorption Calorimeter (TAC) at n_TOF, CERN

The neutron capture cross section of 238U is fundamental to the design and operation of current reactors and future fast nuclear reactors, and thus must be measured to a high level of accuracy. An experiment has been performed at the CERN n_TOF facility using a 4π Total Absorption Calorimeter (TAC) to measure the capture cross section in the resolved resonance region between 1 eV and 25 keV. A preliminary analysis of the TAC data is presented with particular emphasis to the experimental background in this energy region of interest.

238U Neutron Capture Cross Section Measurements at the GELINA Facility

A set of neutron capture experiments based on the time-of-flight technique were performed in order to determine the 238U capture cross section in the unresolved resonance region. The GELINA facility of the Institute for Reference Materials and Measurements (IRMM) served as the neutron source. A pair of C6D6 liquid scintillators was used to register the prompt gamma rays emerging from the uranium sample. The analysis of the experimental data is based on the total energy principle applied in combination with the pulse height weighting technique. The experimental details along with the analysis process are described. The first results in the resolved resonance region are presented and their validity provide a solid base to extend the analysis and extract the average cross section in the keV region.

Neutron Cross-Section Covariances in the Resolved Resonance Region for 55Mn

Recent evaluations of neutron cross-section covariances in the resolved resonance region reveal the need for further research in this area. Major issues include declining uncertainties in multigroup representations and the proper treatment of scattering radius uncertainty. To address these issues, the present work develops a formalism and computer code based on a multilevel Breit-Wigner formula, extending the previous work based on the kernel approximation, using resonance parameter uncertainties from the Atlas of Neutron Resonances. Analytical expressions derived for average cross-section uncertainties in the arbitrary energy bin along with their sensitivities provide a fundamental tool for determining the cross-section uncertainties. The role of resonance-resonance and resonance-potential scattering correlations is studied. As a test case, we apply this approach to estimate (n,γ) and (n,el) covariances for the structural material 55Mn and compare the results with those from the previous kernel approximation.