Presentation and discussion of the UAM/exercise I-1b: "Pin-Cell Burn-Up Benchmark" with the hybrid method

The aim of this work is to present the Exercise I-1b “pin-cell burn-up benchmark” proposed in the framework of OECD LWR UAM. Its objective is to address the uncertainty due to the basic nuclear data as well as the impact of processing the nuclear and covariance data in a pin-cell depletion calculation. Four different sensitivity/uncertainty propagation methodologies participate in this benchmark (GRS, NRG, UPM, and SNU&KAERI). The paper describes the main features of the UPM model (hybrid method) compared with other methodologies. The requested output provided by UPM is presented, and it is discussed regarding the results of other methodologies

Updated Co-58 evaluation for background capture reaction (MT102) (JEF/DOC-1367)

The neutron capture (n,gamma) cross-section for 27-Co-58 theoretically presents a single resonance for 9 eV. However, after plotting the processed library, a discontinuity is made clear as the cross section plummets down to cero in a small range of energy where the peak of the resonance would be expected

Propagation of Statistical and Nuclear Data Uncertainties in Monte-Carlo Burn-up Calculations

Two methodologies to propagate the uncertainties on the nuclide inventory in combined Monte Carlo-spectrum and burn-up calculations are presented, based on sensitivity/uncertainty and random sampling techniques (uncertainty Monte Carlo method). Both enable the assessment of the impact of uncertainties in the nuclear data as well as uncertainties due to the statistical nature of the Monte Carlo neutron transport calculation. The methodologies are implemented in our MCNP–ACAB system, which combines the neutron transport code MCNP-4C and the inventory code ACAB. A high burn-up benchmark problem is used to test the MCNP–ACAB performance in inventory predictions, with no uncertainties. A good agreement is found with the results of other participants. This benchmark problem is also used to assess the impact of nuclear data uncertainties and statistical flux errors in high burn-up applications. A detailed calculation is performed to evaluate the effect of cross-section uncertainties in the inventory prediction, taking into account the temporal evolution of the neutron flux level and spectrum. Very large uncertainties are found at the unusually high burn-up of this exercise (800 MWd/kgHM). To compare the impact of the statistical errors in the calculated flux with respect to the cross uncertainties, a simplified problem is considered, taking a constant neutron flux level and spectrum. It is shown that, provided that the flux statistical deviations in the Monte Carlo transport calculation do not exceed a given value, the effect of the flux errors in the calculated isotopic inventory are negligible (even at very high burn-up) compared to the effect of the large cross-section uncertainties available at present in the data files

Processing and validation of JEFF3-3.1.2 Cross-section Library into Various Formats: ACE, PENDF, GENDF, MATXSR and BOXER.

Following the processing and validation of JEFF-3.1 performed in 2006 and presented in ND2007, and as a consequence of the latest updated of this library (JEFF-3.1.2) in February 2012, a new processing and validation of JEFF-3.1.2 cross section library is presented in this paper. The processed library in ACE format at ten different temperatures was generated with NJOY-99.364 nuclear data processing system. In addition, NJOY-99 inputs are provided to generate PENDF, GENDF, MATXSR and BOXER formats. The library has undergone strict QA procedures, being compared with other available libraries (e.g. ENDF/B-VII.1) and processing codes as PREPRO-2000 codes. A set of 119 criticality benchmark experiments taken from ICSBEP-2010 has been used for validation purposes

Processing of the JEFF-3.1.2 Cross Section Library into various formats (ACE, PENDF, GENDF, MATXSR and BOXER) for testing purposes

1. Objectives and planning 1.1 Processing JEFF-3.1.2 in ACE format 1.2 Processing JEFF-3.1.2 to JANIS and BOXER format 1.3 Changes in NJOY99.364 1.4 Updates in JEFF-3.1.2 1.5 Processing TENDL-201

Progress on spent fuel data compilations for PWRs

A “Collaborative Agreement” involving the collective participation of our students in their last year of our “Nuclear Engineering Master Degree Programme” for: “the review and capturing of selected spent fuel isotopic assay data sets to be included in the new SFCOMPO database

Efectos generalizados en las secciones eficaces y factores de discontinuidad para análisis avanzado de núcleos de agua a presión

El proyecto de investigación se ha propuesto como objetivo central el desarrollo y cualificación de pruebas de principio computacionales y de validación por contratación con las medidas en reactores, de métodos de simulación computacional, detallada bidimensional, con tratamiento preciso de las heterogeneidades reales de los núcleos de agua a presión. Las actividades de desarrollo novedoso incluyen: 1. Estudio de los efectos de realimentaciones cruzadas en las secciones eficaces macroscópicas y microscópicas en dos grupos de energía (rápido y térmico) de las condiciones locales instantáneas e históricas, durante el quemado previo, de las variables más relevantes: densidad y temperatura del moderador, temperatura del combustible, material estructural (rejillas), control y absorbentes consumibles. 2. Métodos y procedimientos para el cálculo de las dependencias funcionales y algoritmos efectivos de interpolación o ajuste funcional para la inclusión de las realimentaciones en el cálculo tridimensional del núcleo. 3. Implantación de cálculo de elemento combustible y de núcleo completo, destinado al análisis detallado de las distribuciones locales de potencia, junto a las variables de análisis general del cálculo de reactore

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

A Comparison of Sensitivity/Uncertainty Methodologies for the Tritium Production in the HFTM/IFMIF Specimen Cells and measurements in Tritium activity in HCLL TBM mock-up LiPb

The prediction of the tritium production is required for handling procedures of samples, safety&maintenance and licensing of the International Fusion Materials Irradiation Facility (IFMIF)