Neutron Cross Section Standards Research Articles

Database work for the new cross section standards evaluation

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.

Work on the neutron standards since the completion of the standards evaluation

An evaluation of the neutron standards that included the neutron cross section standards was recently completed. 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. In this paper work will be reported on new or continuing efforts on the cross section standards including that on the 30 keV Maxwellian averaged cross section for the 197Au(n,γ) reaction that is now a standard and is used in neutron capture cross section measurements as a standard for reactions important for astrophysics.

Verification of R-matrix calculations for charged-particle reactions in the resolved resonance region for the 7Be system

R-matrix theory is used to describe nuclear reactions in the resolved resonance region. It uses information on bound states and low energy resonances to accurately parametrize cross sections on the resonances as well as the non-resonant background. Since the seminal work of Lane and Thomas (1958), the approach has been widely used to analyze experimental cross-section data in a broad range of fields spanning nuclear reaction dynamics, nuclear astrophysics, ion beam analysis and their applications. Different R-matrix codes have been developed and used in these different applications with very little communication among the developers or practitioners on the capabilities, achievements or limitations of the codes. A limited comparison among three R-matrix codes on neutron-induced reactions was performed by the International Atomic Energy Agency (IAEA) International Evaluation of Neutron Cross Section Standards project (2007). Since then, significant progress has been made in their implementation of the R-matrix algorithms, and R-matrix codes have enhanced capabilities. In this paper we present, for the first time, the results of a comprehensive effort to verify the most widely used R-matrix codes in the various fields of nuclear science and applications: AMUR, AZURE2, CONRAD, EDA, FRESCO, GECCCOS, and SAMMY. In addition to the description of the capabilities of the codes and their specifications, we discuss the results of a joint exercise which was coordinated by the International Atomic Energy Agency. The aim of the exercise was to compare calculations of charged-particle reaction cross sections for the light composite system 7Be. The calculations were performed by the codes using identical input R-matrix parameters and other specifications and were limited to charged-particle channels.

New fit of thermal neutron constants (TNC) for233,235U,239,241Pu and252Cf(sf): Microscopic vs. maxwellian data

An IAEA project to update the Neutron Standards is near completion. Traditionally, the Thermal Neutron Constants (TNC) evaluated data by Axton for thermal-neutron scattering, capture and fission on four fissile nuclei and the total nu-bar of 252 Cf(sf) are used as input in the combined least-square fit with neutron cross section standards. The evaluation by Axton (1986) was based on a least-square fit of both thermal-spectrum averaged cross sections (Maxwellian data) and microscopic cross sections at 2200 m/s. There is a second Axton evaluation based exclusively on measured microscopic cross sections at 2200 m/s (excluding Maxwellian data). Both evaluations disagree within quoted uncertainties for fission and capture cross sections and total multiplicities of uranium isotopes. There are two factors, which may lead to such difference: Westcott g-factors with estimated 0.2% uncertainties used in the Axton's fit, and deviation of the thermal spectra from Maxwellian shape. To exclude or mitigate the impact of these factors, a new combined GMA fit of standards was undertaken with Axton's TNC evaluation based on 2200 m/s data used as a prior. New microscopic data at the thermal point, available since 1986, were added to the combined fit. Additionally, an independent evaluation of TNC was undertaken using CONRAD code. Both GMA and CONRAD results are consistent within quoted uncertainties. New evaluation shows a small increase of fission and capture thermal cross sections, and a corresponding decrease in evaluated thermal nubar for uranium isotopes and 239 Pu.

Toward a New Evaluation of Neutron Standards

Measurements related to neutron cross section standards and certain prompt neutron fission spectra are being evaluated. In addition to the standard cross sections, investigations of reference data that are not as well known as the standards are being considered. Procedures and codes for performing this work are discussed. A number of libraries will use the results of this standards evaluation for new versions of their libraries. Most of these data have applications in neutron dosimetry.

Recent Work Leading Towards a New Evaluation of the Neutron Standards

A new version of the ENDF/B library has been planned. The first step in producing this new library is evaluating the neutron standards. An evaluation is now underway with support from a Data Development Project of the IAEA. In addition to the neutron cross section standards, new evaluations are being done for prompt fission neutron spectra and a number of reference data. Efforts have been made to handle uncertainties in a proper way in these evaluations.

The neutron cross section standards, evaluations and applications

The measurement of a neutron cross section can be simplified if it is measured relative to a neutron cross section standard. Then it is not necessary to measure the neutron fluence. The standards in effect provide determinations of the neutron fluence. These standards are not defined standards but instead working standards. As such it is important to make measurements to improve the quality of the standards. Also better techniques for evaluating the standards must be promoted. The history of the evaluation process for the standards is discussed from the first rather primitive one to the latest very sophisticated evaluation that was an international cooperative effort. An overview of the main detectors used for applications of the standards is given.

International Evaluation of Neutron Cross Section Standards

Neutron cross section standards are the basis for the determination of most neutron cross sections. They are used for both measurements and evaluations of neutron cross sections. Not many cross sections can be obtained absolutely – most cross sections are measured relative to the cross section standards and converted using evaluations of the standards. The previous complete evaluation of the neutron cross section standards was finished in 1987 and disseminated as the NEANDC/INDC and ENDF/B-VI standards. R-matrix model fits for the light elements and non-model least-squares fits for all the cross sections in the evaluation were the basis of the combined fits for all of the data. Some important reactions and constants are not standards, but they assist greatly in the determination of the standard cross sections and reduce their uncertainties – these data were also included in the combined fits. The largest experimental database used in the evaluation was prepared by Poenitz and included about 400 sets of experimental data with covariance matrices of uncertainties that account for all cross-energy, cross-reaction and cross-material correlations. For the evaluation GMA, a least-squares code developed by Poenitz, was used to fit all types of cross sections (absolute and shape), their ratios, spectrum-averaged cross sections and thermal constants in one full analysis. But, the uncertainties derived in this manner, and especially those obtained in the R-matrix model fits, have been judged to be too low and unrealistic. These uncertainties were substantially increased prior to their release in the recommended data files of 1987. Modified percentage uncertainties were reassigned by the United States Cross Section Evaluation Working Group's Standards Subcommittee for a wide range of energies, and no covariance (or correlation) matrices were supplied at that time. The need to re-evaluate the cross section standards is based on the appearance of a significant amount of precise experimental data and improved developments in the methodology of analysis and evaluation. Initial efforts to produce a new evaluation were made by the United States Cross Section Evaluation Working Group which formed a Task Force. It was realized that international cooperation would be needed to produce the evaluation. The Working Party on International Evaluation Cooperation of the Nuclear Energy Agency Nuclear Science Committee formed a Subgroup, and the International Atomic Energy Agency formed a Coordinated Research Project (CRP). These groups worked cooperatively to improve the evaluation process. The major effort in producing the evaluation was through the CRP. The evaluations of the neutron cross section standards were finalized in October 2005. Previous difficulties experienced with a data evaluation problem known as “Peelle's Pertinent Puzzle” create biases in the fit of correlated data, and they have been addressed to reduce this phenomenon. The new evaluations of the cross section standards also include covariance matrices of the uncertainties that contain fully justifiable values. The product of this international effort has been adopted as the neutron standards for ENDF/B-VII.0.

Covariances Obtained from an Evaluation of the Neutron Cross Section Standards

New measurements and an improved evaluation process were used to obtain a new evaluation of the neutron cross section standards. Efforts were made to include as much information as possible on the components of the data uncertainties that were then used to obtain the covariance matrices for the experimental data. Evaluations were produced from this process for the 6Li(n,t), 10B(n,α), 10B(n, α 1 γ ), 197Au(n,γ), 235U(n,f), and 238U(n,f) standard cross sections as well as the non-standard 6Li(n,n), 10B(n,n), 238U(n,γ) and 239Pu(n,f) cross sections. There is a general increase in the cross sections for most of the new evaluations, by as much as about 5%, compared with the ENDF/B-VI results. Covariance data were obtained for the 6Li(n,t), 6Li(n,n), 10B(n,α), 10B(n, α 1 γ ), 10B(n,n), 197Au(n,γ), 235U(n,f), 238U(n,f), 238U(n,γ) and 239Pu(n,f) reactions. Also an independent R-Matrix evaluation was produced for the H(n,n) standard cross-section, however, covariance data are not available for this reaction. The evaluations were used in the new ENDF/B-VII library.

An International Evaluation of the Neutron Cross Section Standards

Abstract Work is reported here on the process and present results of an international evaluation of the neutron cross section standards. The evaluations include the H(n,n), 3He(n,p), 6Li(n,t), 10B(n,α), 10B(n,α1γ), 197Au(n,γ), 235U(n,f), and 238U(n,f) standard reactions as well as the 238U(n,γ) and 239Pu(n,f) reactions. This evaluation was performed to include new experiments on the standards that have been made since the ENDF/B-VI evaluation was completed and to improve the evaluation process. Evaluations have been completed for the 6Li(n,t), 197Au(n,γ) and 238U(n,γ) cross sections. Also below 20 MeV the H(n,n), 235U(n,f), 238U(n,f) and 239Pu(n,f) cross sections are completed. Many of the cross sections being evaluated are used in neutron dosimetry for fluence determination. The general trend observed for the evaluations is an increase in the cross sections for most of the reactions from fractions of a percent to several percent compared with the ENDF/B-VI results.

A New International Evaluation of the Neutron Cross Section Standards

The measurements of most neutron cross sections are made relative to neutron cross section standards. The last complete evaluation of the standards, which was for ENDF/B-VI, took place almost 15 years ago. The cross section standards produced for the ENDF/B-VI evaluation were for the H(n, n), 3He(n, p), 6Li(n, t), 10B(n, α), 10B(n, α1γ), C(n, n), Au(n, γ) and 235U(n, f) reactions. Some additional important reactions including the 238U(n, f) cross section, which is used as a standard cross section in the NEANDC/INDC Nuclear Standards File, were evaluated. These standards were accepted internationally to ensure that all evaluation projects were using the same set of standards. Significant experimental activity on the standards has occurred since that evaluation was completed which will cause changes in a new evaluation of the standards. International efforts are now underway to update the previous work by including standards measurements made since that evaluation was completed and to improve the evaluation process. Due to the need for high energy standards, the energy region will be extended to about 200 MeV for some of the cross section standards. Work on the standards evaluation has begun with an investigation of the completed experiments to determine uncertainties and correlations within and among data sets. Further work could lead to an improved evaluation procedure compared with that used in ENDF/B-VI. The status of the work on this evaluation will be discussed.

The neutron cross section standards evaluations for ENDF/B-VI

Abstract The neutron cross section standards are now being evaluated as the initial phase in the development of the new ENDF/B-VI file. These standards evaluations are following a somewhat different process compared with that used for earlier versions of ENDF. The primary effort is concentrated on a simultaneous evaluation using a generalized least squares program, R-matrix evaluations, and a procedure for combining the results of these evaluations. The ENDF/B-VI standards evaluation procedure is outlined, and preliminary simultaneous evaluation and R-matrix results are presented.