Nuclear Data Evaluation Research Articles

Adaptive Monte Carlo for nuclear data evaluation

An adaptive Monte Carlo method for nuclear data evaluation is presented. A fast evaluation method based on the linearization of the nuclear model guides the adaptation of the sampling distribution towards the posterior distribution. The method is suited for parallel computation and provides detailed uncertainty information about nuclear model parameters. Especially, the posterior distribution of the model parameters is not restricted to be multivariate normal. The method is demonstrated in an evaluation of the 181 Ta total cross section for incident neutrons. Future applications are as an efficient sampling scheme in the Total Monte Carlo method, and the restriction of parameter uncertainties in nuclear models by both differential and integral data.

Examples of Use of SINBAD Database for Nuclear Data and Code Validation

The SINBAD database currently contains compilations and evaluations of over 100 shielding benchmark experiments. The SINBAD database is widely used for code and data validation. Materials covered include: Air, N. O, H2 O, Al, Be, Cu, graphite, concrete, Fe, stainless steel, Pb, Li, Ni, Nb, SiC, Na, W, V and mixtures thereof. Over 40 organisations from 14 countries and 2 international organisations have contributed data and work in support of SINBAD. Examples of the use of the database in the scope of different international projects, such as the Working Party on Evaluation Cooperation of the OECD and the European Fusion Programme demonstrate the merit and possible usage of the database for the validation of modern nuclear data evaluations and new computer codes.

Benchmarking and validation activities within JEFF project

The challenge for any nuclear data evaluation project is to periodically release a revised, fully consistent and complete library, with all needed data and covariances, and ensure that it is robust and reliable for a variety of applications. Within an evaluation effort, benchmarking activities play an important role in validating proposed libraries. The Joint Evaluated Fission and Fusion (JEFF) Project aims to provide such a nuclear data library, and thus, requires a coherent and efficient benchmarking process. The aim of this paper is to present the activities carried out by the new JEFF Benchmarking and Validation Working Group, and to describe the role of the NEA Data Bank in this context. The paper will also review the status of preliminary benchmarking for the next JEFF-3.3 candidate cross-section files.

Nuclear data evaluation of long-lived fission products: Microscopic vs. phenomenological optical potentials

Neutron-nucleus cross sections calculated by macroscopic potentials are compared with a microscopic one to study the performance for long-lived fission products. The macroscopic potentials show a good agreement with the microscopic one at higher energies, where neutron experimental data are scarce. Besides it, analyses of differential elastic cross sections at low energies also suggest that the macroscopic potentials are still effective and applicable enough for the long-lived fission products.

BetaShape: A new code for improved analytical calculations of beta spectra

The new code BetaShape has been developed in order to improve the nuclear data related to beta decays. An analytical model was considered, except for the relativistic electron wave functions, for ensuring fast calculations. Output quantities are mean energies, log ft values and beta and neutrino spectra for single and multiple transitions. The uncertainties from the input parameters, read from an ENSDF file, are propagated. A database of experimental shape factors is included. A comparison over the entire ENSDF database with the standard code currently used in nuclear data evaluations shows consistent results for the vast majority of the transitions and highlights the improvements that can be expected with the use of BetaShape.

Application of the SAMINT methodology to the new cross section evaluations of63Cu and65Cu∗

The SAMINT methodology allows coupling of differential and integral data evaluations in a continuous-energy framework. Prior to development of the SAMINT code, integral experimental data such as in the International Criticality Safety Benchmark Experiments Project remained a tool for validation of completed nuclear data evaluations. Now, SAMINT extracts information from integral benchmarks in the form of calculated sensitivity coefficients by Monte Carlo codes such as CE TSUNAMI-3D or MCNP6 and combines it with the results of experimental cross section measurements to produce an updated cross section evaluation utilizing information from both sets of data. The use of the generalized linear least squares methodology ensures that proper weight is given to both the differential and integral data. SAMINT is not intended to bias nuclear data toward specific integral experiments, but it should be used to supplement evaluation of differential experimental data. This work demonstrates the application of the SAMINT methodology to the new Oak Ridge National Laboratory (ORNL) evaluations of the resonance parameters for two isotopes of copper: 63 Cu and 65 Cu.

Evaluation of the235U resonance parameters to fit the standard recommended values

A great deal of effort has been dedicated to the revision of the standard values in connection with the neutron interaction for some actinides. While standard data compilation are available for decades nuclear data evaluations included in existing nuclear data libraries (ENDF, JEFF, JENDL, etc.) do not follow the standard recommended values. Indeed, the majority of evaluations for major actinides do not conform to the standards whatsoever. In particular, for the n + 235 U interaction the only value in agreement with the standard is the thermal fission cross section. A resonance re-evaluation of the n + 235 U interaction has been performed to address the issues regarding standard values in the energy range from 10−5 eV to 2250 eV. Recently, 235 U fission cross-section measurements have been performed at the CERN Neutron Time-of-Flight facility (TOF), known as n_TOF, in the energy range from 0.7 eV to 10 keV. The data were normalized according to the recommended standard of the fission integral in the energy range 7.8 eV to 11 eV. As a result, the n_TOF averaged fission cross sections above 100 eV are in good agreement with the standard recommended values. The n_TOF data were included in the 235 U resonance analysis that was performed with the code SAMMY. In addition to the average standard values related to the fission cross section, standard thermal values for fission, capture, and elastic cross sections were also included in the evaluation. This paper presents the procedure used for re-evaluating the 235 U resonance parameters including the recommended standard values as well as new cross section measurements.

CENDL project, the chinese evaluated nuclear data library

The status of Chinese Evaluated Nuclear Data Library (CENDL) and the relevant CENDL project are introduced in this paper. Recently, a new version CENDL-3.2b0 was being prepared on the basis of the previous CENDL-3.1. The data in the light and actinide nuclide regions are updated from CENDL-3.1, and the new evaluations and calculations are performed mainly around structure and fission product nuclide regions. Covariance was also evaluated for structure and actinide nuclides. At the same time, the methodologies are systematically developed to fulfil the requirements of evaluations for CENDL-3.2b0. The updated nuclear reaction models for light and middle-heavy nuclides, non-model dependent nuclear data evaluation, covariance evaluation approaches, systematics, and integral validation system of nuclear data are incorporated in present CENDL project. The future developments are also planned.

Validation of nuclear data using historical critical assembly measurements

Predictive neutronics simulations depend on accurate nuclear data. We use integral experiments of near critical assemblies as one of the key validation experiments. The accuracy of these measurements is such that they are more constraining to a nuclear data evaluation than the underlying differential data. But the solutions found in calibrating to these benchmarks are degenerate and other data are necessary to discern which is correct. Recently, we have begun to use activation data for this purpose. This paper describes many of the early Los Alamos experiments that are used in this process, with particular attention given to some of the systematic errors that affect these measurements.

Research and development for accuracy improvement of neutron nuclear data on minor actinides

To improve accuracy of neutron nuclear data on minor actinides, a Japanese nuclear data project entitled “Research and development for Accuracy Improvement of neutron nuclear data on Minor ACtinides (AIMAC)” has been implemented. Several independent measurement techniques were developed for improving measurement precision at J-PARC/MLF/ANNRI and KURRI/LINAC facilities. Effectiveness of combining the independent techniques has been demonstrated for identifying bias effects and improving accuracy, especially in characterization of samples used for nuclear data measurements. Capture cross sections and/or total cross sections have been measured for Am-241, Am-243, Np-237, Tc-99, Gd-155, and Gd-157. Systematic nuclear data evaluation has also been performed by taking into account the identified bias effect. Highlights of the AIMAC project are outlined.

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.

Nuclear data for fusion technology – the European approach

The European approach for the development of nuclear data for fusion technology applications is presented. Related R&D activities are conducted by the Consortium on Nuclear Data Development and Analysis for Fusion to satisfy the nuclear data needs of the major projects including ITER, the Early Neutron Source (ENS) and DEMO. Recent achievements are presented in the area of nuclear data evaluations, benchmarking and validation, nuclear model improvements, and uncertainty assessments.

On the use of the generalized SPRT method in the equivalent hard sphere approximation for nuclear data evaluation

A consistent description of the neutron cross sections from thermal energy up to the MeV region is challenging. One of the first steps consists in optimizing the optical model parameters using average resonance parameters, such as the neutron strength functions. They can be derived from a statistical analysis of the resolved resonance parameters, or calculated with the generalized form of the SPRT method by using scattering matrix elements provided by optical model calculations. One of the difficulties is to establish the contributions of the direct and compound nucleus reactions. This problem was solved by using a slightly modified average R-Matrix formula with an equivalent hard sphere radius deduced from the phase shift originating from the potential. The performances of the proposed formalism are illustrated with results obtained for the 238 U+n nuclear systems.

Methodology and issues of integral experiments selection for nuclear data validation

Nuclear data validation involves a large suite of Integral Experiments (IEs) for criticality, reactor physics and dosimetry applications. [1] Often benchmarks are taken from international Handbooks. [2, 3] Depending on the application, IEs have different degrees of usefulness in validation, and usually the use of a single benchmark is not advised; indeed, it may lead to erroneous interpretation and results. [1] This work aims at quantifying the importance of benchmarks used in application dependent cross section validation. The approach is based on well-known General Linear Least Squared Method (GLLSM) extended to establish biases and uncertainties for given cross sections (within a given energy interval). The statistical treatment results in a vector of weighting factors for the integral benchmarks. These factors characterize the value added by a benchmark for nuclear data validation for the given application. The methodology is illustrated by one example, selecting benchmarks for 239 Pu cross section validation. The studies were performed in the framework of Subgroup 39 (Methods and approaches to provide feedback from nuclear and covariance data adjustment for improvement of nuclear data files) established at the Working Party on International Nuclear Data Evaluation Cooperation (WPEC) of the Nuclear Science Committee under the Nuclear Energy Agency (NEA/OECD).

Application of TALYS code for Calculation of Fission Cross Section and Fission Yield of Several Heavy Nuclides

Nuclear data evaluation for fission cross section and fission yield had been performed by many investigators using different models of approximation theoretically. These models are encapsulated and implemented into computer codes to perform more robust nuclear reaction data calculations. TALYS is one of most successful nuclear reaction codes that used to determine fission cross section and fission yield. In this paper, TALYS code was used to calculate some fission reaction including Am-241 (n,f), Th-232 (n,f), and U-235 (n,f). These calculations are performed using different set of reaction mechanism and optical model parameter adjustment, such as fission barrier parameter, level density parameter, transmission mechanism, and so on. Reaction mechanism and parameter adjustment are selected based on reaction characteristics to obtain more accurate and reasonable result. The accuracy of calculation result are heavily depend on the reaction mechanism selection and parameter adjustment. All obtained results have been compared with ENDF nuclear data library.

A modified Generalized Least Squares method for large scale nuclear data evaluation

Nuclear data evaluation aims to provide estimates and uncertainties in the form of covariance matrices of cross sections and related quantities. Many practitioners use the Generalized Least Squares (GLS) formulas to combine experimental data and results of model calculations in order to determine reliable estimates and covariance matrices. A prerequisite to apply the GLS formulas is the construction of a prior covariance matrix for the observables from a set of model calculations. Modern nuclear model codes are able to provide predictions for a large number of observables. However, the inclusion of all observables may lead to a prior covariance matrix of intractable size. Therefore, we introduce mathematically equivalent versions of the GLS formulas to avoid the construction of the prior covariance matrix. Experimental data can be incrementally incorporated into the evaluation process, hence there is no upper limit on their amount. We demonstrate the modified GLS method in a tentative evaluation involving about three million observables using the code TALYS. The revised scheme is well suited as building block of a database application providing evaluated nuclear data. Updating with new experimental data is feasible and users can query estimates and correlations of arbitrary subsets of the observables stored in the database.

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.

Cross Section Calculations of (n,2n) and (n,p) Nuclear Reactions on Germanium Isotopes at 14–15 MeV

Neutron incident reaction cross sections of Germanium isotopes (70,72,74,76Ge) were investigated for the (n,2n) and (n,p) reactions around 14–15 MeV. Cross section calculations have been presented for 70Ge(n,2n)69Ge, 72Ge(n,2n)71Ge, 74Ge(n,2n)73Ge, 76Ge(n,2n)75Ge, 70Ge(n,p)70Ga, 72Ge(n,p)72Ga, 74Ge(n,p)74Ga, and 76Ge(n,p)76Ga reactions. Theoretical calculations were performed with four different computer codes: ALICE/ASH for the Geometry Dependent Hybrid model, TALYS 1.6 for two component Exciton model, EMPIRE 3.2 Malta for Exciton model and PCROSS for Full Exciton model with the incident neutron energy up to 20 MeV. The (n,2n) and (n,p) reaction cross section calculations were compared with empirical formulas derived by several researchers and compared with the experimental data obtained from EXFOR database as well as with evaluated Nuclear data files (ENDF/B-VII.1: USA 2014). Results show good agreement between the theoretical calculations having a major importance in nuclear data evaluation calculations and the experimental data from literature.

Evaluation of nuclear reaction cross section data for the production of 87Y and 88Y via proton, deuteron and alpha-particle induced transmutations

Proton, deuteron and alpha-particle induced reactions on 87,88Sr, natZr and 85Rb targets were evaluated for the production of 87,88Y. The literature data were compared with nuclear model calculations using the codes ALICE-IPPE, TALYS 1.6 and EMPIRE 3.2. The evaluated cross sections were generated; therefrom thick target yields of 87,88Y were calculated. Analysis of radio-yttrium impurities and yield showed that the 87Sr(p, n)87Y and 88Sr(p, n)88Y reactions are the best routes for the production of 87Y and 88Y respectively. The calculated yield for the 87Sr(p, n)87Y reaction is 104 MBq/μAh in the energy range of 14→2.7MeV. Similarly, the calculated yield for the 88Sr(p, n)88Y reaction is 3.2 MBq/μAh in the energy range of 15→7MeV.

Extended optical model for fission

A comprehensive formalism to calculate fission cross sections based on the extension of the optical model for fission is presented. It can be used for description of nuclear reactions on actinides featuring multi-humped fission barriers with partial absorption in the wells and direct transmission through discrete and continuum fission channels. The formalism describes the gross fluctuations observed in the fission probability due to vibrational resonances, and can be easily implemented in existing statistical reaction model codes. The extended optical model for fission is applied for neutron induced fission cross-section calculations on $^{234,235,238}\mathrm{U}$ and $^{239}\mathrm{Pu}$ targets. A triple-humped fission barrier is used for $^{234,235}\mathrm{U}(n,f)$, while a double-humped fission barrier is used for $^{238}\mathrm{U}(n,f)$ and $^{239}\mathrm{Pu}(n,f)$ reactions as predicted by theoretical barrier calculations. The impact of partial damping of class-II/III states, and of direct transmission through discrete and continuum fission channels, is shown to be critical for a proper description of the measured fission cross sections for $^{234,235,238}\mathrm{U}(n,f)$ reactions. The $^{239}\mathrm{Pu}(n,f)$ reaction can be calculated in the complete damping approximation. Calculated cross sections for $^{235,238}\mathrm{U}(n,f)$ and $^{239}\mathrm{Pu}(n,f)$ reactions agree within 3% with the corresponding cross sections derived within the Neutron Standards least-squares fit of available experimental data. The extended optical model for fission can be used for both theoretical fission studies and nuclear data evaluation.