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At least 361 records · Page 20
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Nuclear Data Sheets for A=242

Here, detailed level schemes, decay schemes, band structure, and the experimental data on which they are based are presented for all nuclei with mass number A=242. The experimental data are evaluated and adopted values are given for level and radiation properties. For references on theory refer to the NSR file at the web site given below under Cutoff Date:. This evaluation replaces the A=242 publication by Y. A. Akovali in Nuclear Data Sheets 96, 177 (2002) (2002Ak06).

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2020

Capitalizing on nuclear data libraries' comprehensiveness to obtain solar r-process abundances

The recent observation of neutron stars merger by the LIGO collaboration and the measurements of the event’s electromagnetic spectrum as a function of time for different wavelengths, have altered profoundly our understanding of the r-process site, as well as considerably energized nuclear astrophysics research efforts. R -process abundances are a key element in r-process simulations, as a successful calculation must account for these abundances in the final debris of a stellar cataclysmic event. In this letter, mankind’s complete knowledge of neutron cross sections obtained in the last80 years, as encapsulated in the latest release of the Evaluated Nuclear Data File (ENDF/B) library, is used to obtain solar r-process abundances in a novel way. Here, ENDF/B cross sections has been successfully used for decades in nuclear power and defense applications and are now used to obtain r-process abundances in a fully traceable, documented and unbiased way.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2014

Nuclear Data for Medical Applications – Recent Developments and Future Requirements

Cancer treatment represents a major economic and medical issue because of the extensive incidence of the disease worldwide, with a particularly large rate of increase to be found in developing countries. Over the previous twenty years, the International Atomic Energy Agency (IAEA) has dedicated a series of investigations towards identifying and quantifying the production routes and decay characteristics of radioisotopes judged to be of existing and emerging importance in nuclear medicine and radiotherapy. Both the recommendations formulated during the course of these technical debates and the results of recently completed and on-going IAEA coordinated research projects are focused towards the successful evolution of improvements in nuclear data for medical applications throughout the early 21{sup st} century.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2015

Subspace-based Inverse Uncertainty Quantification for Nuclear Data Assessment

Safety analysis and design optimization depend on the accurate prediction of various reactor attributes. Predictions can be enhanced by reducing the uncertainty associated with the attributes of interest. An inverse problem can be defined and solved to assess the sources of uncertainty, and experimental effort can be subsequently directed to further improve the uncertainty associated with these sources. In this work a subspace-based algorithm for inverse sensitivity/uncertainty quantification (IS/UQ) has been developed to enable analysts account for all sources of nuclear data uncertainties in support of target accuracy assessment-type analysis. An approximate analytical solution of the optimization problem is used to guide the search for the dominant uncertainty subspace. By limiting the search to a subspace, the degrees of freedom available for the optimization search are significantly reduced. A quarter PWR fuel assembly is modeled and the accuracy of the multiplication factor and the fission reaction rate are used as reactor attributes whose uncertainties are to be reduced. Numerical experiments are used to demonstrate the computational efficiency of the proposed algorithm. Our ongoing work is focusing on extending the proposed algorithm to account for various forms of feedback, e.g., thermal-hydraulics and depletion effects.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Chlorine Worth Study Nuclear Data

The Chlorine Worth Study (CWS) was a series of experiments that took place at the National Criticality Experiments Research Center (NCERC), operated by Los Alamos National Laboratory (LANL). The focus of the experiments was to develop new integral benchmarks for the International Criticality Safety Benchmark Evaluation Project (ICSBEP) Handbook with high sensitivity to chlorine in the thermal neutron energy region and which match sensitivities of aqueous chloride operations at LANL. This work discusses the experiment and how the experimental and simulated results using different nuclear data libraries compare to each other.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

NCSP Nuclear Data Program [Slides]

This talk covers wide-ranging information on the Nuclear Criticality Safety Program (NCSP) Data Program. The Lecture includes slides on the history and mission of the NCSP along with charts and graphs on budgets from 2005 onwards. This lecture additionally mentions NCSP critical assemblies, differential experiments, and data measurements and evaluation work. The talk concludes with a gaze into NCSP benefits and successes.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2015

Sensitivities and correlations of critical experiments due to uncertainties of system parameters and nuclear data - 14416

With increasing demand on accuracy of criticality calculations, correlations between benchmark experiments due to system parameter uncertainties become of interest in bias determination. The impact of different interpretations of system parameter uncertainties on the calculation of covariances and correlations of k{sub eff} is discussed in this article. Following the suggestion of the OECD/NEA EGUACSA Benchmark phase IV [1], 21 experiments from the ICSBEP international Handbook of Evaluated Criticality Safety Benchmarks [2] are used for this analysis (LEU-COMP-THERM series 07, exp. 01-04 and 39, exp. 01-17). The experiments share certain assembly and fuel features and were carried out in the same lab. Therefore, correlations in the calculation of k{sub eff} appear. Correlation coefficients between experiments due to nuclear data uncertainties are calculated with TSUNAMI [3] and show high values for most experiments. For the correlation analysis due to system parameter uncertainties mainly two scenarios are considered: In scenario A all uncertain parameters are varied mutually for all fuel rods, representing a systematic deviation of the actual experimental setup from the design value. This analysis is performed for all experiments listed above. In scenario E certain parameters like the outer cladding diameter are varied for each fuel rod individually, representing a random scattering of the varied parameter. Scenarios B, C and D represent intermediate scenarios between A and E, which are discussed shortly. Uncertainty and sensitivity calculations of k{sub eff} on the uncertain experimental parameters based on Monte Carlo sampling are presented and discussed. The calculations were performed with SUnCISTT (Sensitivities and Uncertainties in Criticality Inventory and Source Term Tool [4]) and SCALE 6.1.2 [3]. For the 21 EGUACSA proposed experiments the outer cladding diameter is the leading effect for the uncertainty of k{sub eff} for scenario A. In contrast, for scenario E the impact of the cladding diameter on k{sub eff} cancels out due to rod wise variations leaving the fuel diameter as the leading effect. The correlation analysis based on system parameter uncertainties indicates a significant dependence of the correlation coefficients on the interpretation of the uncertainties. It shows high values for scenario A, if the neutronic features are similar (EALF, H/U). For scenario E the overall correlations are significantly smaller and variations between experiments decrease. The impact of the determined correlations on bias calculations of application cases will be discussed by the authors in the accompanying paper 'Impact of correlated data in validation procedures'. (authors)

42 ENGINEERING↗
DOE OSTI2016

Uncertainty due to Nuclear Data for an MTR Fuel Assembly

For a research reactor, and in particular for the High Flux Reactor (HFR) in Petten, every cycle has a distinct core loading. This regards both the fuel assemblies loading and mainly the loading of the experiment rigs. These experiments can be of the most diverse types, comprising strong absorbing materials as well as fissile isotopes. This poses some challenges at every cycle because of the possible influence on the power profile and neutron flux distribution, which are of great importance for the isotopic production facilities loaded at every cycle in the case of the HFR. The isotope production facilities require predictable and preferably as constant as possible irradiation conditions (neutron flux levels, mainly). Another point of interest is the cycle length, which has to be guaranteed beforehand by taking into account in the assessment analysis the expected on-power core loading changes. These changes are caused by regular loading/unloading of some of the isotope production rigs. For optimization reasons, there is also a tendency to increase the cycle lengths and at the same time to minimize the costs associated to the loading of fresh fuel and control rod assemblies. To justify a further increase in cycle length, a quantification of the uncertainties on the reactivity curve is required, so that the adopted design margins can be decreased in a controlled and responsible manner. The uncertainties in the reactivity curve can be associated to several sources, ranging from uncertainties in engineering and operational parameters to uncertainties in basic nuclear data. This paper includes the quantification of uncertainty on the reactivity curve associated to a typical HFR fuel assembly as result of uncertainties in the cross section data of the main fuel, structural and coolant isotopes. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Validation of nuclear data based on the ASPIS experiments from the SINBAD database - 25428

SINBAD, the Shielding Integral Benchmark Archive and Database, includes the experimental data relative to integral benchmark experiments relevant for radiation shielding applications. The ASPIS experiments, included in the Reactor Shielding section of SINBAD, provide neutron activation foils and spectra measurements performed before/between shielding slabs of Graphite, Sodium, Iron, Steels, Water and combinations of them. Benchmark models for the measurement simulations with the MCNP code are usually provided in SINBAD. Nonetheless, a user-friendly tool is presented, which would be useful for an independent production of MCNP benchmark models. It is based on GPL C++ code called CombLayer. The reports issued for SINBAD on the quality assessment of the ASPIS experiments are also described. The discussion on the quality of the ASPIS experiments for their use in nuclear data evaluations had a positive outcome, except for two experiments out of ten. (authors)

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2014

Nuclear Data Sheets for A = 54

The 2005 evaluation for A=54 (2006Hu08) has been updated. Detailed experimental nuclear structure data and decay data for all nuclei with mass chain A=54 are presented in this current evaluation. The new data and information are given in following datasets: {sup 54}Ca: {sup 9}Be({sup 76}Ge, X) and Be({sup 55}Sc, p), ({sup 56}Ti, 2p) {sup 54}Sc: {sup 54}Ca β{sup −} decay and {sup 54}Sc IT decay {sup 54}Ti: {sup 54}Sc β{sup −} decay {sup 54}Cr: {sup 54}Mn ε decay, {sup 56}Fe(μ, nupng), and {sup 238}U({sup 64}Ni, Xγ) {sup 54}Mn: {sup 51}V({sup 20}Ne, Xγ), {sup 55}Mn(p, pn), and {sup 56}Fe(μ, nu2ng) {sup 54}Fe: {sup 58}Ni α decay, {sup 9}Be({sup 55}Co, Xγ), {sup 54}Fe(e, e'), {sup 54}Fe(p, p'), and Coulomb Excitation {sup 54}Co: {sup 54}Ni ε decay, {sup 55}Cu εp decay, {sup 28}Si({sup 32}S, αpnγ), {sup 54}Fe(p,n), and {sup 54}Fe({sup 3}He, t) {sup 54}Ni: {sup 54}Ni IT decay, {sup 55}Zn εp decay, {sup 9}Be({sup 55}Ni, Xγ), and {sup 24}Mg({sup 32}S, Xγ) {sup 54}Zn: Ni({sup 58}Ni, X)

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2022

On the Uranium and Plutonium Nuclear Data Evaluations [Slides]

This presentation provides overviews of uraniums, 233 U and 235 U, plutonium research including motivations, current status in ENDF/B-VIII.0, the inclusion of sub-thermal data, and the inclusion of LANL ratio capture-to-fission data. Additionally, the preliminary fit of Mosby’s data as reported, fluctuating neutron multiplicities, uncertainty in evaluated libraries, and experimental effects are also presented. The presentation concludes by providing a summary of plans for the U and Pu evaluation work.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2017

Neutron nuclear data measurements for criticality safety

To support the US Department of Energy Nuclear Criticality Safety Program, neutron-induced cross section experiments were performed at the Geel Electron Linear Accelerator of the Joint Research Center Site Geel, European Union. Neutron capture and transmission measurements were carried out using metallic natural cerium and vanadium samples. Together with existing data, the measured data will be used for a new evaluation and will be submitted with covariances to the ENDF/B nuclear data library.

Guber, Klaus H.↗
DOE OSTI2015

Bayesian Monte Carlo Method for Nuclear Data Evaluation

A Bayesian Monte Carlo method is outlined which allows a systematic evaluation of nuclear reactions using TALYS. The result will be either an EXFOR-weighted covariance matrix or a collection of random files, each accompanied by an experiment based weight.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗