Nuclear Data Evaluation Research Articles

A computational EXFOR database

The EXFOR library is a useful resource for many people in the field of nuclear physics. In particular, the experimental data in the EXFOR library serves as a starting point for nuclear data evaluations. There is an ongoing discussion about how to make evaluations more transparent and reproducible. One important ingredient may be convenient programmatic access to the data in the EXFOR library from high-level languages. To this end, the complete EXFOR library can be converted to a MongoDB database. This database can be conveniently searched and accessed from a wide variety of programming languages, such as C++, Python, Java, Matlab, and R. This contribution provides some details about the successful conversion of the EXFOR library to a MongoDB database and shows simple usage examples to underline its merits. All codes required for the conversion have been made available online and are open-source. In addition, a Dockerfile has been created to facilitate the installation process.

On the impact of nuclear data uncertainties on LWR neutron dosimetry assessments

In this work, an overview on the relevance of the nuclear data (ND) uncertainties with respect to the Light Water Reactors (LWR) neutron dosimetry is presented. The paper summarizes results of several studies realized at the LRT laboratory of the Paul Scherrer Institute over the past decade. The studies were done using the base LRT calculation methodology for dosimetry assessments, which involves the neutron source distribution representation, obtained based on validated CASMO/SIMULATE core follow calculation models, and the subsequent neutron transport simulations with the MCNP® software. The methodology was validated using as reference data results of numerous measurement programs fulfilled at Swiss NPPs. Namely, the following experimental programs are considered in the given overview: PWR “gradient probes” and BWR fast neutron fluence (FNF) monitors post irradiation examination (PIE). For the both cases, assessments of the nuclear data related uncertainties were performed. When appropriate, a cross-verification of the deterministic and stochastic based uncertainty propagation techniques is provided. Furthermore, the observations on which particular neutron induced reactions contribute dominantly to the overall ND-related uncertainties are demonstrated. The presented results should help with assessing the overall impact of the various nuclear data uncertainties with respect to dosimetry applications and provide relevant feedback to the nuclear data evaluators.

Use of shielding integral benchmark archive and database for nucleardata validation

Shielding benchmarks were extensively used for the validation and improvement of nuclear data since many years. Recent evaluations however mostly rely on the validation using critical benchmarks, which can introduce biases and compensation effects. A new Working Party on International Nuclear Data Evaluation Co-operation Subgroup 47 (WPEC SG47) entitled "Use of Shielding Integral Benchmark Archive and Database for Nuclear Data Validation" was started in June 2019 with the objective to promote further development and use of Shielding Integral Benchmark Archive and Database (SINBAD) and thus contribute to the diversification of the nuclear data validation practice by including more extensively other types of integral measurements, such as shielding benchmarks, in the validation and evaluation procedure. Use of shielding benchmarks is expected to provide a wider-scope test of the performance of the evaluated nuclear data and would ultimately contribute to a production of more general-purpose cross-section evaluations.

Overview of the OECD-NEA Working Party on International Nuclear Data Evaluation Cooperation (WPEC)

The OECD Nuclear Energy Agency (NEA) Working Party on International Nuclear Data Evaluation Cooperation (WPEC) was established in 1989 to facilitate collaboration in nuclear data activities. Over its thirty year history, different Subgroups have been created to address topics in nearly every aspect of nuclear data, including: experimental measurements, evaluation, validation, model development, quality assurance of databases and the development of software tools.WPEC has recently completed activities on fission yield evaluation, the general nuclear database structure (GNDS) to replace the ENDF-6 format, methods to provide feedback to evaluation, studies of specific capture cross sections, new methods in thermal scattering kernel evaluation and the Collaborative International Evaluated Library Organisation (CIELO) Pilot Project. Ongoing activities in GNDS application programming interface (API) development, methods for covariance evaluation and quality assurance in nuclear data validation using the International Criticality Safety Benchmark Evaluation Project (ICSBEP) database are complemented by the work of two Expert Groups that oversee the High-Priority Request List (HPRL) for Nuclear Data and the continuous development of the GNDS. New activities on the use of integral experiments for nuclear data validation and adjustment, as well as the use of the Shielding Integral Benchmark Archive and Database (SINBAD) for validation have begun and will be coordinated alongside future Subgroups.After three decades we will review the status of WPEC, how it integrates other collections and activities organised by the NEA and how it dovetails with the initiatives of the IAEA and other bodies to effectively coordinate international activities in nuclear data.

New perspectives in neutron reaction cross-section evaluation using consistent multichannel modeling methodology: Application to 17O*

The traditional methodology of nuclear data evaluation is showing its limitations in reducing significantly the uncertainties in neutron cross sections below their current level. This suggests that a new approach should be considered. This work aims at establishing that a major qualitative improvement is possible by changing the reference framework historically used for evaluating nuclear model data. The central idea is to move from the restrictive framework of the incident neutron and target nucleus to the more general framework of the excited compound-system. Such a change, which implies the simultaneous modeling of all the reactions leading to the same compound-system, opens up the possibility of direct comparisons between nuclear model parameters, whether those are derived for reactor physics applications, astrophysics or basic nuclear spectroscopy studies. This would have the double advantage of bringing together evaluation activities performed separately, and of pooling experimental databases and basic theoretical nuclear parameter files. A consistent multichannel modeling methodology using the TORA module of the CONRAD code is demonstrated across the evaluation of differential and angle-integrated neutron cross sections of 16O by fitting simultaneously incident-neutron direct kinematic reactions and incident-alpha inverse kinematic reactions without converting alpha data into the neutron laboratory system. The modeling is fulfilled within the Reich-Moore formalism and an unique set of fitted resonance parameters related to the 17O* compound-system.

Fission fragments observables measured at the LOHENGRIN spectrometer

Nuclear fission yields are key data for reactor studies, such as spent fuel inventory or decay heat, and for understanding fission process. Despite a significant effort allocated to measure fission yields during the last decades, the recent evaluated libraries still need improvements in particular in the reduction of the uncertainties. Moreover, some discrepancies between these libraries must be explained. Additional measurements provide complementary information and estimations of experimental correlations, and new kinds of measurements enable to test the models used during the nuclear data evaluation process. A common effort by the CEA, the LPSC and the ILL aims at tackling these issues by providing precise measurements of isotopic and isobaric fission yields with the related variance-covariance matrices. Additionally, the experimental program involves a large range of observables requested by the evaluations, such as kinetic energy dependency of isotopic yields and odd-even effect in order to test the sharing of total excitation energy and the spin generation mechanism. Another example is the complete range of isotopic distribution per mass that allows the determination of the charge polarization, which has to be consistent for complementary masses (pre-neutron emission). For instance, this information is the key observable for the evaluation of isotopic yields. Finally, ionic charge distributions are indirect measurements of nanosecond isomeric ratios as a probe of the nuclear de-excitation path in the (E*, J, π) representation. Measurements for thermal neutron induced fission of 241 Pu have been carried out at the ILL in Grenoble, using the LOHENGRIN mass spectrometer. Methods, results and comparison to models calculations will be presented corresponding to a status on fission fragments observables reachable with this facility.

Impact of nuclear data evaluations on data assimilation for an LFR

The Lead-cooled Fast Reactor is one of the three technologies selected by the Sustainable Nuclear Energy Technology Platform that can meet future European energy needs. The main drawbacks for the industrial deployment of LFR are the lack of operational experience and the impact of uncertainties. In nuclear reactor design the uncertainties mainly come from material properties, fabrication tolerances, operation conditions, simulation tools and nuclear data. The uncertainty in nuclear data is one of the most important sources of uncertainty in reactor physics simulations. Furthermore, it is known that the uncertainties in reactor criti-cality safety parameters are severely dependent on the nuclear data library used to estimate them. However, the impact of using different evaluations while performing data assimilation to constraint the uncertainties in the criticality parameters has not been properly assessed yet. In this work, a data assimilation for the main isotopes contributing to the uncertainty in keff of the ALFRED lead-cooled fast reactor has been performed with the SUMMON system using JEFF-3.3, ENDF/B-VIII.0 and JENDL-4.0u2 state-of-the-art nuclear data libraries, together with critical mass experiments from the International Criticality Safety Benchmark Evaluation Project that are representative of ALFRED, in order to assess the impact of using different evaluations for data assimilation.

Research on nuclear data evaluation and development ofdata library

Nuclear data describe the basic properties of nuclei and the nuclear interaction properties between particles and nuclei. It is the basic data for the development of nuclear science, and an important bridge connecting the fundamental research of nuclear physics, nuclear engineering, and nuclear technology applications. Nuclear data are also an important branch of nuclear physics research. This paper briefly introduces the connotation, types and applications of nuclear data, as well as the experimental nuclear database, the evaluation process of nuclear data and the evaluated nuclear data libraries. Then the development history of nuclear data evaluation and data library in China are introduced, as well as the activities of the International Nuclear Data Organization and the prospects of nuclear data research were reviewed. Finally, some key issues are discussed, such as what need to be paid attention to in the nuclear data evaluation and the future development of nuclear data evluation in China.

Our Future Nuclear Data Needs

A well-established knowledge of nuclear phenomena including fission, reaction cross sections, and structure/decay properties is critical for applications ranging from the design of new reactors to nonproliferation to the production of radioisotopes for the diagnosis and treatment of illness. However, the lack of a well-quantified, predictive theoretical capability means that most nuclear observables must be measured directly and used to calibrate empirical models, which in turn provide the data needed for these applications. In many cases, either there is a lack of data needed to guide the models or the results of the different measurements are discrepant, leading to the development of evaluation methodologies to provide recommended values and uncertainties. In this review, we describe the nuclear data evaluation process and the international community that carries it out. We then discuss new measurements and improved theory and/or modeling needed to address future challenges in applied nuclear science.

TDCR measurements to determine the half-life of 55Fe

Several measurements with aliquots from a 55Fe solution were carried out using two custom-built TDCR counters. The time between the initial and the last measurements is 2304 d, i.e. a length of more than two half-lives. The data were analysed to determine the half-life of 55Fe, which was found to be (1006.2 ± 1.0) d. The determined half-life is longer than the current values from nuclear data evaluations, but it is in good agreement with recent results obtained by Pommé et al. (2019).

Nuclear cross-section adjustment based on the 2-D/1-D transport code KYADJ

The theoretical basis for the nuclear cross-section adjustment in the 2-D method of characteristics (MOC) and 1-D SN coupling sensitivity and uncertainty analysis code KYADJ is developed. The energy-dependent sensitivity coefficients are produced by the sensitivity analysis using the 3-D fine flux and adjoint flux based on the first-order perturbation theory. Not only can the sensitivity coefficients be used to estimate the uncertainty of the effective multiplication factor caused by the measurement uncertainties of nuclear cross-section data, but also they can be implemented in the cross-section adjustment according to the Bayes’ theorem and the generalized linear-least-squares method (GLLSM). The cross-section adjustment aims to promote nuclear data evaluations, code verifications, and data validations. In this paper, the adjustment method is verified by the light water reactor (LWR) benchmarks: the MOX pin-cell and the PB-2 assembly cases from the UAM benchmarks. The virtual experimental value is assumed to be the target value of the cross-section adjustment. The results show that the GLLSM can reduce the differences between the calculated and experimental (C/E) values remarkably. All the relative variations of cross-sections for the adjustment are given according to the information of the (C/E) value, the sensitivity coefficients, and the prior uncertainties. The posteriori uncertainties are given, which demonstrates the decrease of the uncertainties.

Development of CASMO5 for VVER-1000/1200 analysis and preliminary validation using critical experiments

Abstract Studsvik has recently extended the CASMO5 advanced lattice physics code for the analysis of VVER 1000 and 1200 reactors. These extensions form the basis of CASMO5-VVER, which is primarily intended to compute homogenized nodal data for SIMULATE5-VVER. CASMO5-VVER leverages the latest nuclear data and numerical methods to VVER analyses. The current CASMO5 data library, based on the ENDF/B-VII.1 nuclear data evaluation, features a 586 energy group structure and nuclear data for hundreds of unique nuclides. Resonance self-shielding, based on the Equivalence Theory and an Optimal Two-Term Rational (OTTR) method, has been extended to support hexagonal geometry. The solution to the two-dimensional transport equation is based on the new Linear Source (LS) approximation for the Method of Characteristics (MOC). The acceleration of the MOC solution is attained through the implementation of a Coarse-Mesh Nonlinear Diffusion (CMND) acceleration. Results from preliminary validation against various critical experiments are presented in this work.

Genetic algorithm for nuclear data evaluation applied to subcritical neutron multiplication inference benchmark experiments

An optimization algorithm has been developed for the first time for application to International Criticality Safety Benchmark Evaluation Project (ICSBEP) subcritical neutron multiplication inference benchmark experiments. The optimization algorithm is a genetic algorithm for nuclear data evaluation adjustments, specifically applied to subcritical benchmark measurements. The algorithm has been tested and yields improvement in (C-E)/E values of subcritical benchmark observables of interest. In this work, the genetic algorithm is applied to improvement of fission neutron multiplicity distribution parameters using several subcritical neutron multiplication inference benchmarks; specifically a series of reflected 4.5 kg α-phase spherical plutonium benchmarks. The algorithm results suggest changing the mean (ν¯) and standard deviation (σ) of the number of neutrons emitted by 240Pu in spontaneous fission from 2.1510 to 2.1460 and from 1.1510 to 1.1395, respectively. In addition, the standard deviation of the number of neutrons emitted by 239Pu in induced fission should remain unchanged at 1.1400. These changes are all within 1 standard deviation.

Nuclear data and applications at the nuclear safety and radioprotection institute: Analysis, evaluation and application

The Institut de Radioprotection et de Sûreté Nucléaire (IRSN) is the French public service expert in nuclear and radiation risks with activities that cover all the related scientific and technical issues. IRSN provides expertise in support of radiation protection and nuclear safety risks. A great deal of research work carried out at IRSN concerns the improvement of computational tools and data used on a daily basis by nuclear safety assessment experts. Particularly in this paper the approach used to improve data quality and its uncertainties is presented. Computation tools and nuclear data evaluation methodology are discussed.

Infological models of the ENDF-format nuclear data

Issues involved in the infologic modeling of the ENDF-format nuclear data libraries for the purpose of converting ENDF files into a relational database have been considered. The transfer to a relational format will make it possible to use standard readily available tools for nuclear data processing which simplify the conversion and operation of this data array. Infological models have been described using formulas of the “Entity (List of Attributes)” type. The proposed infological formulas are based on the physical nature of data and theoretical relations. This eliminates the need for a special notation to be introduced to describe the structure and the content of data, which, in turn, facilitates the use of relational formats in codes and solution of nuclear data evaluation problems. The concept of nuclear informatics has been formulated based on relational DBMS technologies as one of the tools for solving the “big data” problem in modern science and technology. The organizational and technological grounds for the transfer of ENDF libraries to a relational format are presented. Requirements to the nuclear data presentation formats supported by relational DBMS are listed. Peculiarities of the infological model construction, conditioned by the hierarchical nature of nuclear data, are identified. The sequence for the ENDF metadata saving is presented, which can be useful for the verification and validation (testing of the structural and syntactical validity and operability) of both source data and the procedures for the conversion to a relational format. Formulas of infological models are presented for the cross sections file, the secondary neutron energy distributions file, and the nuclear reaction product energy-angle distributions file. A complete array of infological models for ENDF libraries and the generation modules of respective relational tables are available on a public website.

A review of nuclear data for the k₀-based neutron activation analysis

Nuclear data for the k0-based neutron activation analysis (k0-NAA) including: k0-factors – combination of nuclear constants (atomic mass, isotopic abundance, gamma-ray yield and thermal neutron capture cross section); Q0 – the ratio of the resonance integral to thermal neutron capture cross section; and Er – effective resonance energy, along with the related nuclear data are presented and evaluated in the presentation. Accuracy and capability of k0-NAA depends considerably on the reliability of the above mentioned nuclear data. In general, the evaluation of nuclear data is essential and necessary in the field of nuclear science and technology, and this work is conducted by the nuclear data centers on the world as well as the research institutions where the nuclear data are usedfor R&D. Therefore, the evaluation of the nuclear data used in k0-NAA should also be performed so that some of data may no longer be appropriate should be redetermined. The evaluation of nuclear data in k0-NAA would contribute to the improvement of accuracy and reliability of the method, moreover, it would also contribute to the establishment a nuclear database in Vietnam in the future.

Tantalum, Titanium, and Zirconium Neutron Total Cross-Section Measurements from 0.4 to 25 MeV

The electron linear accelerator housed in the Gaerttner Linear Accelerator Center at Rensselaer Polytechnic Institute was used to generate a pulsed neutron source to measure the neutron total cross section of tantalum, titanium, and zirconium from 0.4 to 25 MeV. Neutron transmission measurements were made using the time-of-flight method with neutron flight paths of approximately 100 and 250 m. The long flight paths combined with narrow neutron pulse widths, fast detector responses, fast electronics, and data collection system provide good energy resolution for the measurements. A high signal-to-background ratio through much of the energy range combined with low statistical errors resulted in low uncertainties on cross sections.The results are presented and compared with the major nuclear data evaluations. Each measurement identifies regions where the neutron total cross sections could be reevaluated. The total cross-section measurements presented here can help nuclear data evaluators improve neutron total cross-section data in future evaluations.

Evaluation of the neutronic Superphenix start-up commissioning tests with TRIPOLI4

The Superphénix reactor was a Sodium-cooled Fast Reactor (SFR) developed under a European framework that intended to provide a general design of a commercial power plant. During the start-up of the reactor, six different batches containing dummy assemblies allowed to verify the main safety criteria at zero-power conditions tests at 180 °C. The present paper presents a stochastic analysis of two start-up cores by comparing the experimental results with the performed calculations with TRIPOLI-4 and the JEFF 3.1.1 nuclear data evaluations. The main and back-up control rod evaluations, the critical mass, as well as the handling error experiment are examined. In addition, two particular irradiation tests are studied to determine the flux distribution of the core and finally, the assessment of a subcriticality analysis is performed to verify the validity of the former MSM factors. In general, the obtained results show good agreement with the experiment, having a general reactivity bias with the performed calculations, the flux distribution is also in great agreement with the tests with small discrepancies and the MSM evaluation permitted to observe better consistency with experimental data. The obtained results allowed to validate our model and will serve as reference calculation for further research.

Formulation of Model Defects Suitable for the Resonance Regime

A method to account for model deficiencies in nuclear data evaluations in the resonance regime is proposed. The method follows the ideas of Schnabel and coworkers and relies on Gaussian processes with a novel problemadapted ansatz for the covariance matrix of model uncertainties extending the formalism to the energy region of resonances. The method was used to evaluate a set of schematic but realistic neutron reaction data generated by an R-matrix code and a well defined model defect. Using the extended ansatz for model defects the Bayesian evaluation successfully recovered the built-in model defect in size and structure thus demonstrating the applicability of the method.

Trends on Major Actinides from an Integral Data Assimilation

Nuclear data evaluations on major actinides can be improved by integral data assimilation. Appropriate integral measurements with reliable experimental techniques have been selected such as ICSBEP, IRPhE and MASURCA critical masses, PROFIL irradiation experiments and the FCAIX experimental programme (critical masses and spectral indices). Highly reliable analyses are possible with the use of as-built geometries calculated with the TRIPOLI4 Monte Carlo code. The C/E values have been used in an integral data assimilation solving the Bayes equation. The trends on the JEFF3.1.1235U capture cross section are quite consistent with recent differential measurements. Assimilation results suggest up to a 2.5% decrease for238U capture from 3 keV to 60 keV, and a 4-5% decrease for238U inelastic in the plateau region. For this energy range, uncertainties are respectively reduced from 3-4 to 1-2% and from 6-9% to 2-2.5% for 238U capture and238U inelastic. Results on239Pu fission cross sections are included in posterior uncertainties. The increase trends on239Pu capture cross-section is of around 3% in the [2 keV-100 keV] energy range. For240Pu capture cross section, the increase is of around 4% in the [3 keV-100 keV] energy range and goes in the same direction as a recent evaluation.