Quantum characterization of bipartite Gaussian states

Gaussian bipartite states are basic tools for the realization of quantum information protocols with continuous variables. Their complete characterization is obtained by the reconstruction of the corresponding covariance matrix. Here we describe in details and experimentally demonstrate a robust and reliable method to fully characterize bipartite optical Gaussian states by means of a single homodyne detector. We have successfully applied our method to the bipartite states generated by a sub-threshold type-II optical parametric oscillator which produces a pair of thermal cross-polarized entangled CW frequency degenerate beams. The method provide a reliable reconstruction of the covariance matrix and allows to retrieve all the physical information about the state under investigation. These includes observable quantities, as energy and squeezing, as well as non observable ones as purity, entropy and entanglement. Our procedure also includes advanced tests for Gaussianity of the state and, overall, represents a powerful tool to study bipartite Gaussian state from the generation stage to the detection one

Methodology for Pressurized Thermal Shock Analysis in Nuclear Power Plant

The relevance of the fracture mechanics in the technology of the nuclear power plant is mainly connected to the risk of a catastrophic brittle rupture of the reactor pressure vessel. There are no feasible countermeasures that can mitigate the effects of such an event that impair the capability to maintain the core covered even in the case of properly functioning of the emergency systems. The origin of the problem is related to the aggressive environment in which the vessel operates for long term (e.g. more than 40 years), characterized by high neutron flux during normal operation. Over time, the vessel steel becomes progressively more brittle in the region adjacent to the core. If a vessel had a preexisting flaw of critical size and certain severe system transients occurred, this flaw could propagate rapidly through the vessel, resulting in a through-wall crack. The severe transients that can lead the nuclear power plant in such conditions, known as Pressurized Thermal Shock (PTS), are characterized by rapid cooling (i.e., thermal shock) of the a part of the internal reactor pressure vessel surface that may be combined with repressurization can create locally a sudden increase of the stresses inside the vessel wall and lead to the suddenly growth of the flaw inside the vessel thickness. Based on the long operational experience from nuclear power plants equipped with reactor pressure vessel all over the world, it is possible to conclude that the simultaneous occurrence of critical-size flaws, embrittled vessel, and a severe PTS transient is a very low probability event. Moreover, additional studies performed at utilities and regulatory authorities levels have shown that the RPV can operate well beyond the original design life (40 years) because of the large safety margin adopted in the design phase. A better understanding and knowledge of the materials behavior, improvement in simulating in a more realistic way the plant systems and operational characteristics and a better evaluation of the loads on the RPV wall during the PTS scenarios, have shown that the analysis performed during the 80’s were overly conservative, based on the tools and knowledge available at that time. Nowadays the use of best estimate approach in the analyses, combined with tools for the uncertainty evaluation is taking more consideration to reduce the safety margins, even from the regulatory point of view. The US NRC has started the process to revise the technical base of the PTS analysis for a more risk-informed oriented approach. This change has the aim to remove the un-quantified conservatisms in all the steps of the PTS analysis, from the selection of the transients, the adopted codes and the criteria for conducting the analysis itself thus allow a more realistic prediction. This change will not affect the safety, because beside the operational experience, several analysis performed by thermal hydraulic, fracture mechanics and Probabilistic Safety Assessment (PSA) point of view, have shown that the reactor fleet has little probability of exceeding the limits on the frequency of reactor vessel failure established from NRC guidelines on core damage frequency and large early release frequency through the period of license extension. These calculations demonstrate that, even through the period of license extension, the likelihood of vessel failure attributable to PTS is extremely low (≈10-8/year) for all domestic pressurized water reactors. Different analytical approaches have been developed for the evaluation of the safety margin for the brittle crack propagation in the rector pressure vessel under PTS conditions. Due to the different disciplines involved in the analysis: thermal-hydraulics, structural mechanics and fracture mechanics, different specialized computer codes are adopted for solving single part of the problem. The aims of this chapter is to present all the steps of a typical PTS analysis base on the methodology developed at University of Pisa with discussion and example calculation results for each tool adopted and their use, based on a more realistic best estimate approach. This methodology starts with the analysis of the selected scenario by mean a System Thermal-Hydraulic (SYS-TH) code such as RELAP5 [2][3], RELAP5-3D [1], CATHARE2 [4][6], etc. for the analysis of the global behavior of the plant and for the evaluation of the primary side pressure and fluid temperature at the down-comer inlet. For a more deep investigation of the cooling load on the rector pressure vessel internal surface at small scale, a Computational Fluid Dynamics (CFD) code is used. The calculated temperature profile in the down-comer region is transferred to a Finite Element (FE) structural mechanics code for the evaluation of the stresses inside the RPV wall. The stresses induced by the pressure in the primary side are also evaluated. The stress intensity factor at crack tip is evaluated by mean the weight function method based on a simple integration of the stresses along the crack border multiplied by the weight function. The values obtained are compared with the critical stress intensity factor typical of the reactor pressure vessel base material for the evaluation of the safety margin

An Integrated Software Platform for Best Estimate Safety Analyses of Nuclear Power Plants

Nuclear power plant safety is granted through the demonstration that regulatory acceptance criteria are fulfilled by the provided (calculated) analyses of the NPP performances and sufficient safety margins are respected during normal operation, anticipated transients and postulated accident conditions. Safety margins are very hard to determine in absolute terms, numerical calculations are used to assess their values. Over the last 30 years an extensive effort has been carried out aiming to improve the knowledge of the nuclear power plant behaviour under transient scenarios. The development of Best Estimate (BE) computer codes are the direct consequence of these noteworthy efforts. The availability of more sophisticated and specialized computer codes gives the analyst the possibility to perform very detailed analysis in all the fields involved in the safety of a NPP: thermal-hydraulics, CFD, 3D neutron kinetics etc. The possibility to create a software environment where a multidisciplinary problem can be solved adopting different specialized codes able to exchange data among them is a fruitful approach to the problem aiming to improve the results. The computational tools, adopted in best-estimate approach in licensing, include a) the best estimate computer codes; b) the nodalizations together with the procedures for the development and the qualification; c) the uncertainty methodology. The Nuclear Research Group of San Piero a Grado of the University of Pisa has developed a software platform with 15 interacting computer codes. Such platform covers the reactor simulation multidisciplinary problem from generation of neutron cross-sections, through system thermal-hydraulic analyses, up to detailed structural and fuel mechanics studies and it embeds software procedures for automatized data transfer between codes. Together with methodological procedures for nodalizations development and qualification the platform leads to a great decrease of the human induced error in the results. The developed platform has been tested and successfully applied to perform the safety analyses required by the Chapter 15 of the Final Safety Analysis Report of the CNA-2 nuclear power plant in Argentina

Identification of Limiting Case Between DBA and SBDBA (CL Break Area Sensitivity): A New Model for the Boron Injection System

Atucha-2 is a Siemens-designed PHWR reactor under construction in the Republic of Argentina. Its geometrical complexity and (e.g., oblique Control Rods, Positive Void coefficient) required a developed and validated complex three dimensional (3D) neutron kinetics (NK) coupled thermal hydraulic (TH) model. Reactor shut-down is obtained by oblique CRs and, during accidental conditions, by an emergency shut-down system (JDJ) injecting a highly concentrated boron solution (boron clouds) in the moderator tank, the boron clouds reconstruction is obtained using a CFD (CFX) code calculation. A complete LBLOCA calculation implies the application of the RELAP5-3D© system code. Within the framework of the third Agreement “NA-SA – University of Pisa” a new RELAP5-3D control system for the boron injection system was developed and implemented in the validated coupled RELAP5-3D/NESTLE model of the Atucha 2 NPP. The aim of this activity is to find out the limiting case (maximum break area size) for the Peak Cladding Temperature for LOCAs under fixed boundary conditions

Full characterization of Gaussian bipartite entangled states by a single homodyne detector

We present the full experimental reconstruction of Gaussian entangled states generated by a type--II optical parametric oscillator (OPO) below threshold. Our scheme provides the entire covariance matrix using a single homodyne detector and allows for the complete characterization of bipartite Gaussian states, including the evaluation of purity, entanglement and nonclassical photon correlations, without a priori assumptions on the state under investigation. Our results show that single homodyne schemes are convenient and robust setups for the full characterization of OPO signals and represent a tool for quantum technology based on continuous variable entanglement.Comment: 4 pages, 3 figures, slightly longer version of published PR

Comparative study between cold leg and hot leg safety injection during SBLOCA in a 4-loop PWR NPP

This article presents a comparison between two operation modes for the emergency core cooling system during a Small Break Loss of Coolant Accident (SBLOCA) in the cold leg of 4-loop PWR Westinghouse design nuclear power plant. In the first mode, the cold leg safety injection is used to mitigate the consequences of the accident and in the second mode the hot leg safety injection is used. The best estimate light water reactor transient analysis system code RELAP5 Mod3.3 was used in calculations. The plant nodalization consists of two loops; the first one represents the broken loop and the second one represents the other three intact loops. The results show that, in the cold leg safety injection the primary pressure decreases with time and remains higher than the secondary pressure for a period of time (~ 500 sec) during whichthe steam generators remains as a heat sink for the primary side, the accumulators start late and functioning on remaining transient time, and a repeatable loop seal clearing and refill occurs. During the hot leg safety injection the primary pressure decreases rapidly but remains higher than the secondary pressure for a longer period of time (~ 600 sec), the accumulators start early and functioning on a part of the transient time before they are totally discharged, and there is no repeatable loop seal clearing and refill. In the two modes the maximum clad surface temperature does not violate the safety limit

Application of Best Estimate Plus Uncertainty (BEPU) Methodology in a Final Safety Analysis Report (FSAR) of a Generic Plant

The licensing process of a nuclear power plant is motivated by the need to protect humans and the environment from ionizing radiation and, at the same time, sets out the basis for the design and determining the acceptability of the plant. An important part of the licensing process is the realization of accident analysis related to the design basis, which should be documented in the Final Safety Analysis Report (FSAR). There are different options on accidents calculation area by combining the use of computer codes and data entry for licensing purposes. One is the Best Estimate Plus Uncertainty (BEPU), which considers realistic input data and associated uncertainties. Applications of BEPU approaches in licensing procedures were initiated in the 2000s, first to analysis of Loss of Coolant Accident (LOCA), and then to the accident analysis as a whole, documented in Chapter 15 of the FSAR. This work has as main objective the implementation of BEPU methodology in all analyses contained in FSAR, through the homogenization of the analytical techniques and identification of key disciplines and key topics in the licensing process

Bianchi Type I Cosmology in N=2, D=5 Supergravity

The dynamics and evolution of Bianchi type I space-times is considered in the framework of the four-dimensional truncation of a reduced theory obtained from the N=2,D=5 supergravity. The general solution of the gravitational field equations can be represented in an exact parametric form. All solutions have a singular behavior at the initial/final moment, except when the space-time geometry reduces to the isotropic flat case. Generically the obtained cosmological models describe an anisotropic, expanding or collapsing, singular Universe with a non-inflationary evolution for all times.Comment: revised version to appear in PR

An effective method to estimate multidimensional Gaussian states

A simple and efficient method for characterization of multidimensional Gaussian states is suggested and experimentally demonstrated. Our scheme shows analogies with tomography of finite dimensional quantum states, with the covariance matrix playing the role of the density matrix and homodyne detection providing Stern-Gerlach-like projections. The major difference stems from a different character of relevant noises: while the statistics of Stern-Gerlach-like measurements is governed by binomial statistics, the detection of quadrature variances correspond to chi-square statistics. For Gaussian and near Gaussian states the suggested method provides, compared to standard tomography techniques, more stable and reliable reconstructions. In addition, by putting together reconstruction methods for Gaussian and arbitrary states, we obtain a tool to detect the non-Gaussian character of optical signals.Comment: 8 pages, 5 fis, accepted for publication on PR

AUTOMATIC ANALYSIS OF SEISMIC DATA BY USING NEURAL NETWORKS: APPLICATIONS TO ITALIAN VOLCANOES.

The availability of the new computing techniques allows to perform advanced analysis in near real time, improving the seismological monitoring systems, which can extract more significant information from the raw data in a really short time. However, the correct identification of the events remains a critical aspect for the reliability of near real time automatic analysis. We approach this problem by using Neural Networks (NN) for discriminating among the seismic signals recorded in the Neapolitan volcanic area (Vesuvius, Phlegraean Fields). The proposed neural techniques have been also applied to other sets of seismic data recorded in Stromboli volcano. The obtained results are very encouraging, giving 100% of correct classification for some transient signals recorded at Vesuvius and allowing the clustering of the large dataset of VLP events recorded at Stromboli volcano