Nuclear data for proton beam radiotherapy

The need for a program of work focused on the nuclear data evaluation of chargedparticle reactions has arisen recently due to their increasing use in cancer therapy. This project, as part of that program, has as its main goal the selection and comparison of nuclear data for nuclear reactions induced by protons at low to intermediate energies (E < 250 MeV). The methodology of selection was based on the data base EXFOR and the compilations of radionuclide production cross sections of N. Sobolevsky. For the purpose of comparison and evaluation, theoretical calculations with the reaction model code EMPIRE-II are being used. © 2009 American Institute of Physics.

PanConverter - a tool for data conversion and recalculation, written to convert special data collections to standard import format used by PANGAEA

Supported file formats: - CrossRef XML file(s) - TRiDaS (Tree Ring Data Standard, http://www.tridas.org). Example: hdl:10013/epic.42747.d001 - IMMA (International Maritime Meteorological Archive). Used by the project CLIWOC (García-Herrera et al. 2007, http://doi.pangaea.de/10.1594/PANGAEA.743343) - NOAA IOAS (International Ocean Atlas Series). Example: hdl:10013/epic.42747.d008 - SOCAT (Surface Ocean CO2 Atlas, Bakker et al. 2014, http://doi.pangaea.de/10.1594/PANGAEA.811776) - CHUAN (Comprehensive Historical Upper-Air Network, Stickler et al. 2013, http://doi.pangaea.de/10.1594/PANGAEA.821222). Example: hdl:10013/epic.42747.d003 - Thermosalinograph (TSG) data. Format developed by Gerd Rohardt. Example: hdl:10013/epic.42747.d002 - Columus GPS Data Logger V-900 format to KML or GPX. Example: hdl:10013/epic.42747.d006

Impact of nuclear data uncertainties in the Phase-1B Benchmark

Accurate control over the spent nuclear fuel content is essential for its safe and optimized transportation, storage and management. Consequently, the reactivity of spent fuel and its isotopic content must be accurately determined. Nowadays, to predict isotopic evolution throughout irradiation and decay periods is not a problem thanks to the development of powerful codes and methodologies. In order to have a realistic confidence level in the prediction of spent fuel isotopic content, it is desirable to determine how uncertainties in the basic nuclear data affect isotopic prediction calculations by quantifying their associated uncertainties

Propagation of Nuclear Data Uncertainties in Transmutation Calculations Using ACAB Code

The assessment of the accuracy of parameters related to the reactor core performance (e.g., ke) and f el cycle (e.g., isotopic evolution/transmutation) due to the uncertainties in the basic nuclear data (ND) is a critical issue. Different error propagation techniques (adjoint/forward sensitivity analysis procedures and/or Monte Carlo technique) can be used to address by computational simulation the systematic propagation of uncertainties on the final parameters. To perform this uncertainty assessment, the ENDF covariance les (variance/correlation in energy and cross- reactions-isotopes correlations) are required. In this paper, we assess the impact of ND uncertainties on the isotopic prediction for a conceptual design of a modular European Facility for Industrial Transmutation (EFIT) for a discharge burnup of 150 GWd/tHM. The complete set of uncertainty data for cross sections (EAF2007/UN, SCALE6.0/COVA-44G), radioactive decay and fission yield data (JEFF-3.1.1) are processed and used in ACAB code.

Importance of Nuclear Data Uncertainties in Criticality Calculations

To study the propagation of the uncertainty from basic data across different scale and physics phenomena -> through complex coupled multi-physics and multi-scale simulations

Impact of Nuclear Data Uncertainties in the Phase-1B Benchmark

Accurate control over the spent nuclear fuel content is essential for its safe and optimized transportation, storage and management. Consequently, the reactivity of spent fuel and its isotopic content must be accurately determined.

Propagation of nuclear data uncertainties in fuel cycle calculations using MONTE-CARLO Technique

The uncertainty propagation in fuel cycle calculations due to Nuclear Data (ND) is a important important issue for : issue for : • Present fuel cycles (e.g. high burnup fuel programme) • New fuel cycles designs (e.g. fast breeder reactors and ADS) Different error propagation techniques can be used: • Sensitivity analysis • Response Response Surface Method Surface Method • Monte Carlo technique Then, p p , , in this paper, it is assessed the imp y pact of ND uncertainties on the decay heat and radiotoxicity in two applications: • Fission Pulse Decay ( y Heat calculation (FPDH) • Conceptual design of European Facility for Industrial Transmutation (EFIT)

Isotopic uncertainty assessment due to nuclear data uncertainties in high-burnup samples.

The accurate prediction of the spent nuclear fuel content is essential for its safe and optimized transportation, storage and management. This isotopic evolution can be predicted using powerful codes and methodologies throughout irradiation as well as cooling time periods. However, in order to have a realistic confidence level in the prediction of spent fuel isotopic content, it is desirable to determine how uncertainties affect isotopic prediction calculations by quantifying their associated uncertainties.

Importance of nuclear data uncertainties in criticality calculations

The aim of this paper is to study the importance of nuclear data uncertainties in the prediction of the uncertainties in keff for LWR (Light Water Reactor) unit-cells. The first part of this work is focused on the comparison of different sensitivity/uncertainty propagation methodologies based on TSUNAMI and MCNP codes; this study is undertaken for a fresh-fuel at different operational conditions. The second part of this work studies the burnup effect where the indirect contribution due to the uncertainty of the isotopic evolution is also analyzed.

Propagation of U235-238, Pu239 and STL Nuclear Data Uncertainties for PWR core analysis

An uncertainty propagation methodology based on Monte Carlo method is applied to PWR nuclear design analysis to assess the impact of nuclear data uncertainties in 235,238 U, 239 Pu and Scattering Thermal Library for Hydrogen in water. This uncertainty analysis is compared with the design and acceptance criteria to assure the adequacy of bounding estimates in safety margins.

Testing JEFF-3.1.1 and ENDF/B-VII.1 Decay and Fission Yield Nuclear Data Libraries with Fission Pulse Neutron Emission and Decay Heat Experiments

The aim of this work is to test the present status of Evaluated Nuclear Decay and Fission Yield Data Libraries to predict decay heat and delayed neutron emission rate, average neutron energy and neutron delayed spectra after a neutron fission pulse. Calculations are performed with JEFF-3.1.1 and ENDF/B-VII.1, and these are compared with experimental values. An uncertainty propagation assessment of the current nuclear data uncertainties is performed.

Nuclear Data Uncertainty Propagation to Reactivity Coefficients of a Sodium Fast Reactor

The assessment of the uncertainty levels on the design and safety parameters for the innovative European Sodium Fast Reactor (ESFR) is mandatory. Some of these relevant safety quantities are the Doppler and void reactivity coefficients, whose uncertainties are quantified. Besides, the nuclear reaction data where an improvement will certainly benefit the design accuracy are identified. This work has been performed with the SCALE 6.1 codes suite and its multigroups cross sections library based on ENDF/B-VII.0 evaluation.

Impact of Nuclear Data Uncertainties on Advanced Fuel Cycles and Their Irradiated Fuel - A comparison between Libraries

The uncertainties on the isotopic composition throughout the burnup due to the nuclear data uncertainties are analysed. The different sources of uncertainties: decay data, fission yield and cross sections; are propagated individually, and their effect assessed. Two applications are studied: EFIT (an ADS-like reactor) and ESFR (Sodium Fast Reactor). The impact of the uncertainties on cross sections provided by the EAF-2010, SCALE6.1 and COMMARA-2.0 libraries are compared. These Uncertainty Quantification (UQ) studies have been carried out with a Monte Carlo sampling approach implemented in the depletion/activation code ACAB. Such implementation has been improved to overcome depletion/activation problems with variations of the neutron spectrum.

Towards a complete propagation of nuclear data uncertainties Fission yields and decay data

Propagation of nuclear data uncertainties to calculated values is interesting for design purposes and libraries evaluation. XSUSA, developed at GRS, propagates cross section uncertainties to nuclear calculations. In depletion simulations, fission yields and decay data are also involved and suppose a possible source of uncertainty that must be taken into account. We have developed tools to generate varied fission yields and decay libraries and to propagate uncertainties trough depletion in order to complete the XSUSA uncertainty assessment capabilities. A simple test to probe the methodology is defined and discussed.

Propagation of nuclear data uncertainties for PWR core analysis

An uncertainty propagation methodology based on the Monte Carlo method is applied to PWR nuclear design analysis to assess the impact of nuclear data uncertainties. The importance of the nuclear data uncertainties for 235,238 U, 239 Pu, and the thermal scattering library for hydrogen in water is analyzed. This uncertainty analysis is compared with the design and acceptance criteria to assure the adequacy of bounding estimates in safety margins.