Desarrollo de códigos neutrónicos propios para la simulación del núcleo de reactores PWR

La simulación de la física del núcleo de los reactores nucleares por su complejidad requiere del uso de computadores y del software adecuado, y su evolución es ir hacía métodos y modelos de los llamados best-estimate, con el objeto de aumentar la disponibilidad de la central manteniendo los márgenes de seguridad. Para ello el Departamento de Ingeniería Nuclear (UPM), ha desarrollado el Sistema SEANAP en uso en varias centrales nucleares españolas, que realiza la simulación en 3D y con detalle de barrita combustible del quemado nominal y real del núcleo del reactor, hace el seguimiento en línea de la operación, y ayuda a la planificación óptima de las maniobras operacionales

Accident Management Actions In An Upper-Head Small-Break Loss-Of-Coolant Accident With High-Pressure Safety Injection Failed

Since the Three Mile Island accident, an important focus of pressurized water reactor (PWR) transient analyses has been a small-break loss-of-coolant accident (SBLOCA). In 2002, the discovery of thinning of the vessel head wall at the Davis Besse nuclear power plant reactor indicated the possibility of an SBLOCA in the upper head of the reactor vessel as a result of circumferential cracking of a control rod drive mechanism penetration nozzle - which has cast even greater importance on the study of SBLOCAs. Several experimental tests have been performed at the Large Scale Test Facility to simulate the behavior of a PWR during an upper-head SBLOCA. The last of these tests, Organisation for Economic Co-operation and Development Nuclear Energy Agency Rig of Safety Assessment (OECD/NEA ROSA) Test 6.1, was performed in 2005. This test was simulated with the TRACE 5.0 code, and good agreement with the experimental results was obtained. Additionally, a broad analysis of an upper-head SBLOCA with high-pressure safety injection failed in a Westinghouse PWR was performed taking into account different accident management actions and conditions in order to check their suitability. This issue has been analyzed also in the framework of the OECD/NEA ROSA project and the Code Applications and Maintenance Program (CAMP). The main conclusion is that the current emergency operating procedures for Westinghouse reactor design are adequate for these kinds of sequences, and they do not need to be modified.

Amorphization kinectics under swift heavy ion irradiation: a cumulative overlapping-track approach

A simple illustrative physical model is presented to describe the kinetics of damage and amorphization by swiftheavyions (SHI) in LiNbO3. The model considers that every ion impact generates initially a defective region (halo) and a full amorphous core whose relative size depends on the electronic stopping power. Below a given stopping power threshold only a halo is generated. For increasing fluences the amorphized area grows monotonically via overlapping of a fixed number N of halos. In spite of its simplicity the model, which provides analytical solutions, describes many relevant features of the kinetic behaviour. In particular, it predicts approximate Avrami curves with parameters depending on stopping power in qualitative accordance with experiment that turn into Poisson laws well above the threshold value

Potential common radiation problems for components and diagnostics in future magnetic and inertial confinement fusion devices

This work aims at identifying commonpotentialproblems that futurefusiondevices will encounter for both magnetic and inertialconfinement approaches in order to promote joint efforts and to avoid duplication of research. Firstly, a comparison of radiation environments found in both fusion reaction chambers will be presented. Then, wall materials, optical components, cables and electronics will be discussed, pointing to possible future areas of common research. Finally, a brief discussion of experimental techniques available to simulate the radiation effect on materials is included

The role of spatial and temporal radiation deposition in inertial fusion chambers: the case of HiPER¿

The first wall armour for the reactor chamber of HiPER will have to face short energy pulses of 5 to 20 MJ mostly in the form of x-rays and charged particles at a repetition rate of 5–10 Hz. Armour material and chamber dimensions have to be chosen to avoid/minimize damage to the chamber, ensuring the proper functioning of the facility during its planned lifetime. The maximum energy fluence that the armour can withstand without risk of failure, is determined by temporal and spatial deposition of the radiation energy inside the material. In this paper, simulations on the thermal effect of the radiation–armour interaction are carried out with an increasing definition of the temporal and spatial deposition of energy to prove their influence on the final results. These calculations will lead us to present the first values of the thermo-mechanical behaviour of the tungsten armour designed for the HiPER project under a shock ignition target of 48 MJ. The results will show that only the crossing of the plasticity limit in the first few micrometres might be a threat after thousands of shots for the survivability of the armour.

Depth dependent lattice disorder and strain in Mn-implanted and post-annealed InAs thin films

The lattice order degree and the strain in as-grown, Mn-implanted and post-implantedannealedInAsthinfilms were investigated with depth resolution by means of Rutherford backscattering spectrometry in channeling conditions (RBS/C). Three main crystallographic axes were analyzed for both In and As sublattices. The behaviour of the induced defects was evaluated in two regions with different native defects: the interface and the surface. The results show that Mn implantation and post-implantation annealing are anisotropic processes, affecting in a different way the In and As sublattices. The mechanisms influencing the enhancement and deterioration of the crystal quality during the implantation are discussed in relation to the as-grown defects and the segregation of the elements

Molecular Dynamics Simulations of Lead and Lithium in Liquid Phase

Pb17Li is today a reference breeder material in diverse fusion R&D programs worldwide. Extracting dynamic and structural properties of liquid LiPb mixtures via molecular dynamics simulations, represent a crucial step for multiscale modeling efforts in order to understand the suitability of this compound for future Nuclear Fusion technologies. At present a Li-Pb cross potential is not available in the literature. Here we present our first results on the validation of two semi-empirical potentials for Li and Pb in liquid phase. Our results represent the establishment of a solid base as a previous but crucial step to implement a LiPb cross potential. Structural and thermodynamical analyses confirm that the implemented potentials for Li and Pb are realistic to simulate both elements in the liquid phase.

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).

A Comparison of different Uncertainty Activation Cross-Section Data Libraries and collapsed values for different neutron spectra: ADS, FISSION and FUSION

PART I:Cross-section uncertainties under differentneutron spectra. PART II: Processing uncertainty libraries

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.

Uncertainty assessment methodologies applied to Tritium production in fusion lithium breeding blankets

- Need of Tritium production - Neutronic objectives - The Frascati experiment - Measurements of Tritium activity

UPM Results BWR-PWR Exercise I-1 by MCNP5 code

The perturbation approach relies in principle on a unique “NJOY + MCNP5 + SUSD3D”calculation. The inputs are the geometry MCNP5 input file and an ENDF file containing covariances.

Burnup Uncertainty Analysis for PWR-HFP Exercise I-1

Burn-up credit analyses are based on depletion calculations that provide an accurate prediction of spent fuel isotopic contents, followed by criticality calculations to assess keff

Isotopic prediction calculation methodologies: application to Vandellos-II Reactor cycles 7-11

Determining as accurate as possible spent nuclear fuel isotopic content is gaining importance due to its safety and economic implications. Since nowadays higher burn ups are achievable through increasing initial enrichments, more efficient burn up strategies within the reactor cores and the extension of the irradiation periods, establishing and improving computation methodologies is mandatory in order to carry out reliable criticality and isotopic prediction calculations. Several codes (WIMSD5, SERPENT 1.1.7, SCALE 6.0, MONTEBURNS 2.0 and MCNP-ACAB) and methodologies are tested here and compared to consolidated benchmarks (OECD/NEA pin cell moderated with light water) with the purpose of validating them and reviewing the state of the isotopic prediction capabilities. These preliminary comparisons will suggest what can be generally expected of these codes when applied to real problems. In the present paper, SCALE 6.0 and MONTEBURNS 2.0 are used to model the same reported geometries, material compositions and burn up history of the Spanish Van de llós II reactor cycles 7-11 and to reproduce measured isotopies after irradiation and decay times. We analyze comparisons between measurements and each code results for several grades of geometrical modelization detail, using different libraries and cross-section treatment methodologies. The power and flux normalization method implemented in MONTEBURNS 2.0 is discussed and a new normalization strategy is developed to deal with the selected and similar problems, further options are included to reproduce temperature distributions of the materials within the fuel assemblies and it is introduced a new code to automate series of simulations and manage material information between them. In order to have a realistic confidence level in the prediction of spent fuel isotopic content, we have estimated uncertainties using our MCNP-ACAB system. This depletion code, which combines the neutron transport code MCNP and the inventory code ACAB, propagates the uncertainties in the nuclide inventory assessing the potential impact of uncertainties in the basic nuclear data: cross-section, decay data and fission yields

Isotopic prediction simulations applied to high burnup samples irradiated in VANDELLÓS-II Reactor core

Isotopic content assessment has a paramount importance for safety and storage reasons. During the latest years, a great variety of codes have been developed to perform transport and decay calculations, but only those that couple both in an iterative manner achieve an accurate prediction of the final isotopic content of irradiated fuels. Needless to say, them all are supposed to pass the test of the comparison of their predictions against the corresponding experimental measures.