Development and application of a thermo-hydraulic model supporting the design and the deterministic safety analysis in a sodium cooled fast reactor

Il CP-ESFR è un progetto integrato di cooperazione europeo sui reattori a sodio SFR realizzato sotto il programma quadro EURATOM 7, che unisce il contributo di venticinque partner europei. Il CP-ESFR ha l'ambizione di contribuire all'istituzione di una "solida base scientifica e tecnica per il reattore veloce refrigerato a sodio, al fine di accelerare gli sviluppi pratici per la gestione sicura dei rifiuti radioattivi a lunga vita, per migliorare le prestazioni di sicurezza, l'efficienza delle risorse e il costo-efficacia di energia nucleare al fine di garantire un sistema solido e socialmente accettabile di protezione della popolazione e dell'ambiente contro gli effetti delle radiazioni ionizzanti. " La presente tesi di laurea è un contributo allo sviluppo di modelli e metodi, basati sull’uso di codici termo-idraulici di sistema, per l’ analisi di sicurezza di reattori di IV Generazione refrigerati a metallo liquido. L'attività è stata svolta nell'ambito del progetto FP-7 PELGRIMM ed in sinergia con l’Accordo di Programma MSE-ENEA(PAR-2013). Il progetto FP7 PELGRIMM ha come obbiettivo lo sviluppo di combustibili contenenti attinidi minori 1. attraverso lo studio di due diverse forme: pellet (oggetto della presente tesi) e spherepac 2. valutandone l’impatto sul progetto del reattore CP-ESFR. La tesi propone lo sviluppo di un modello termoidraulico di sistema dei circuiti primario e intermedio del reattore con il codice RELAP5-3D© (INL, US). Tale codice, qualificato per il licenziamento dei reattori nucleari ad acqua, è stato utilizzato per valutare come variano i parametri del core del reattore rilevanti per la sicurezza (es. temperatura di camicia e di centro combustibile, temperatura del fluido refrigerante, etc.), quando il combustibile venga impiegato per “bruciare” gli attinidi minori (isotopi radioattivi a lunga vita contenuti nelle scorie nucleari). Questo ha comportato, una fase di training sul codice, sui suoi modelli e sulle sue capacità. Successivamente, lo sviluppo della nodalizzazione dell’impianto CP-ESFR, la sua qualifica, e l’analisi dei risultati ottenuti al variare della configurazione del core, del bruciamento e del tipo di combustibile impiegato (i.e. diverso arricchimento di attinidi minori). Il testo è suddiviso in sei sezioni. La prima fornisce un’introduzione allo sviluppo tecnologico dei reattori veloci, evidenzia l’ambito in cui è stata svolta questa tesi e ne definisce obbiettivi e struttura. Nella seconda sezione, viene descritto l’impianto del CP-ESFR con attenzione alla configurazione del nocciolo e al sistema primario. La terza sezione introduce il codice di sistema termico-idraulico utilizzato per le analisi e il modello sviluppato per riprodurre l’impianto. Nella sezione quattro vengono descritti: i test e le verifiche effettuate per valutare le prestazioni del modello, la qualifica della nodalizzazione, i principali modelli e le correlazioni più rilevanti per la simulazione e le configurazioni del core considerate per l’analisi dei risultati. I risultati ottenuti relativamente ai parametri di sicurezza del nocciolo in condizioni di normale funzionamento e per un transitorio selezionato sono descritti nella quinta sezione. Infine, sono riportate le conclusioni dell’attività.

Qualitative and quantitative formal model-based safety analysis : push the safety button

Magdeburg, Univ., Fak. für Informatik, Diss., 2011

Lipid changes in HIV-patients switching to the coformulated single tablet FTC/RPV/TDF (Eviplera®). Efficacy and safety analysis. GeSida Study 8114.

INTRODUCTION Rilpivirine (RPV) has a better lipid profile than efavirenz (EFV) in naïve patients (1). Switching to RPV may be convenient for many patients, while maintaining a good immunovirological control (2). The aim of this study was to analyze lipid changes in HIV-patients at 24 weeks after switching to Eviplera® (emtricitabine/RPV/tenofovir disoproxil fumarate [FTC/RPV/TDF]). MATERIALS AND METHODS Retrospective, multicentre study of a cohort of asymptomatic HIV-patients who switched from a regimen based on 2 nucleoside reverse transcriptase inhibitors (NRTI)+protease inhibitor (PI)/non nucleoside reverse transcriptase inhibitor (NNRTI) or ritonavir boosted PI monotherapy to Eviplera® during February-December, 2013; all had undetectable HIV viral load for ≥3 months prior to switching. Patients with previous failures on antiretroviral therapy (ART) including TDF and/or FTC/3TC, with genotype tests showing resistance to components of Eviplera®, or who had changed the third drug of the ART during the study period were excluded. Changes in lipid profile and cardiovascular risk (CVR), and efficacy and safety at 24 weeks were analyzed. RESULTS Among 305 patients included in the study, 298 were analyzed (7 cases were excluded due to lack of data). Men 81.2%, mean age 44.5 years, 75.8% of HIV sexually transmitted. 233 (78.2%) patients switched from a regimen based on 2 NRTI+NNRTI (90.5% EFV/FTC/TDF). The most frequent reasons for switching were central nervous system (CNS) adverse events (31.0%), convenience (27.6%) and metabolic disorders (23.2%). At this time, 293 patients have reached 24 weeks: 281 (95.9%) have continued Eviplera®, 6 stopped it (3 adverse events, 2 virologic failures, 1 discontinuation) and 6 have been lost to follow up. Lipid profiles of 283 cases were available at 24 weeks and mean (mg/dL) baseline vs 24 weeks are: total cholesterol (193 vs 169; p=0.0001), HDL-c (49 vs 45; p=0.0001), LDL-c (114 vs 103; p=0.001), tryglycerides (158 vs 115; p=0.0001), total cholesterol to HDL-c ratio (4.2 vs 4.1; p=0.3). CVR decreased (8.7 vs 7.5%; p= 0.0001). CD4 counts were similar to baseline (653 vs 674 cells/µL; p=0.08), and 274 (96.8%) patients maintained viral suppression. CONCLUSIONS At 24 weeks after switching to Eviplera®, lipid profile and CVR improved while maintaining a good immunovirological control. Most subjects switched to Eviplera® from a regimen based on NNRTI, mainly EFV/FTC/TDF. CNS adverse events, convenience and metabolic disorders were the most frequent reasons for switching.

A phase 2 randomized trial of radiotherapy (RT) plus panitumumab compared to chemoradiotherapy in patients with unresected, locally advanced squamous cell carcinoma of the head and neck (SCCHN): interim pooled safety analysis

Background: Panitumumab (pmab), a fully human monoclonal antibody against the epidermal growth factor receptor (EGFR), is indicated as monotherapy for treatment of metastatic colorectal cancer. This ongoing study is designed to assess the efficacy and safety of pmab in combination with radiotherapy (PRT) compared to chemoradiotherapy (CRT) as initial treatment of unresected, locally advanced SCCHN (ClinicalTrials.gov Identifier: NCT00547157). Methods: This is a phase 2, open-label, randomized, multicenter study. Eligible patients (pts) were randomized 2:3 to receive cisplatin 100 mg/m2 on days 1 and 22 of RT or pmab 9.0 mg/kg on days 1, 22, and 43. Accelerated RT (70 to 72 Gy − delivered over 6 to 6.5 weeks) was planned for all pts and was delivered either by intensity-modulated radiation therapy (IMRT) modality or by three-dimensional conformal (3D-CRT) modality. The primary endpoint is local-regional control (LRC) rate at 2 years. Key secondary endpoints include PFS, OS, and safety. An external, independent data monitoring committee conducts planned safety and efficacy reviews during the course of the trial. Results: Pooled data from this planned interim safety analysis includes the first 52 of the 150 planned pts; 44 (84.6%) are male; median (range) age is 57 (33−77) years; ECOG PS 0: 65%, PS 1: 35%; 20 (39%) pts received IMRT, and 32 (61%) pts received 3D-CRT. Fifty (96%) pts completed RT, and 50 pts received RT per protocol without a major deviation. The median (range) total RT dose administered was 72 (64−74) Gy. The most common grade _ 3 adverse events graded using the CTCAE version 3.0 are shown (Table). Conclusions: After the interim safety analysis, CONCERT-2 continues per protocol. Study enrollment is estimated to be completed by October 2009.

Traffic Safety Analysis for Local Agencies, 2012

Although many larger Iowa cities have staff traffic engineers who have a dedicated interest in safety, smaller jurisdictions do not. Rural agencies and small communities must rely on consultants, if available, or local staff to identify locations with a high number of crashes and to devise mitigating measures. However, smaller agencies in Iowa have other available options to receive assistance in obtaining and interpreting crash data. These options are addressed in this manual. Many proposed road improvements or alternatives can be evaluated using methods that do not require in-depth engineering analysis. The Iowa Department of Transportation (DOT) supported developing this manual to provide a tool that assists communities and rural agencies in identifying and analyzing local roadway-related traffic safety concerns. In the past, a limited number of traffic safety professionals had access to adequate tools and training to evaluate potential safety problems quickly and efficiently and select possible solutions. Present-day programs and information are much more conducive to the widespread dissemination of crash data, mapping, data comparison, and alternative selections and comparisons. Information is available and in formats that do not require specialized training to understand and use. This manual describes several methods for reviewing crash data at a given location, identifying possible contributing causes, selecting countermeasures, and conducting economic analyses for the proposed mitigation. The Federal Highway Administration (FHWA) has also developed other analysis tools, which are described in the manual. This manual can also serve as a reference for traffic engineers and other analysts.

Safety Analysis of Low-Volume Rural Roads in Iowa, 2010

Iowa features an extensive surface transportation system, with more than 110,000 miles of roadway, most of which is under the jurisdiction of local agencies. Given that Iowa is a lower-population state, most of this mileage is located in rural areas that exhibit low traffic volumes of less than 400 vehicles per day. However, these low-volume rural roads also account for about half of all recorded traffic crashes in Iowa, including a high percentage of fatal and major injury crashes. This study was undertaken to examine these crashes, identify major contributing causes, and develop low-cost strategies for reducing the incidence of these crashes. Iowa’s extensive crash and roadway system databases were utilized to obtain needed data. Using descriptive statistics, a test of proportions, and crash modeling, various classes of rural secondary roads were compared to similar state of Iowa controlled roads in crash frequency, severity, density, and rate for numerous selected factors that could contribute to crashes. The results of this study allowed the drawing of conclusions as to common contributing factors for crashes on low-volume rural roads, both paved and unpaved. Due to identified higher crash statistics, particular interest was drawn to unpaved rural roads with traffic volumes greater than 100 vehicles per day. Recommendations for addressing these crashes with low-cost mitigation are also included. Because of the isolated nature of traffic crashes on low-volume roads, a systemic or mass action approach to safety mitigation was recommended for an identified subset of the entire system. In addition, future development of a reliable crash prediction model is described.

Validation Studies of Thermal-hydraulic Code for Safety Analysis of Nuclear Power Plants

This thesis gives an overview of the validation process for thermal hydraulic system codes and it presents in more detail the assessment and validation of the French code CATHARE for VVER calculations. Three assessment cases are presented: loop seal clearing, core reflooding and flow in a horizontal steam generator. The experience gained during these assessment and validation calculations has been used to analyze the behavior of the horizontal steam generator and the natural circulation in the geometry of the Loviisa nuclear power plant. The cases presented are not exhaustive, but they give a good overview of the work performed by the personnel of Lappeenranta University of Technology (LUT). Large part of the work has been performed in co-operation with the CATHARE-team in Grenoble, France. The design of a Russian type pressurized water reactor, VVER, differs from that of a Western-type PWR. Most of thermal-hydraulic system codes are validated only for the Western-type PWRs. Thus, the codes should be assessed and validated also for VVER design in order to establish any weaknesses in the models. This information is needed before codes can be used for the safety analysis. Theresults of the assessment and validation calculations presented here show that the CATHARE code can be used also for the thermal-hydraulic safety studies for VVER type plants. However, some areas have been indicated which need to be reassessed after further experimental data become available. These areas are mostly connected to the horizontal stem generators, like condensation and phase separation in primary side tubes. The work presented in this thesis covers a large numberof the phenomena included in the CSNI code validation matrices for small and intermediate leaks and for transients. Also some of the phenomena included in the matrix for large break LOCAs are covered. The matrices for code validation for VVER applications should be used when future experimental programs are planned for code validation.

Cardiovascular adverse events during adjuvant endocrine therapy for early breast cancer using letrozole or tamoxifen: safety analysis of BIG 1-98 trial

PURPOSE: Previous analyses of adjuvant studies of aromatase inhibitors versus tamoxifen, including the Breast International Group (BIG) 1-98 study, have suggested a small numerical excess of cardiac adverse events (AEs) on aromatase inhibitors, a reduction in the incidence of hypercholesterolemia on tamoxifen, and significantly higher incidence of thromboembolic AEs on tamoxifen. The purpose of the present study is to provide detailed updated information on these AEs in BIG 1-98. PATIENTS AND METHODS: Eight thousand twenty-eight postmenopausal women with receptor-positive early breast cancer were randomly assigned (double-blind) between March 1998 and May 2003 to receive 5 years of adjuvant endocrine therapy with letrozole, tamoxifen, or a sequence of these agents. Seven thousand nine hundred sixty-three patients who actually received therapy are included in this safety analysis, which focuses on cardiovascular events. AE recording ceased 30 days after therapy completion (or after switch on the sequential arms). RESULTS: Baseline comorbidities were balanced. At a median follow-up time of 30.1 months, we observed similar overall incidence of cardiac AEs (letrozole, 4.8%; tamoxifen, 4.7%), more grade 3 to 5 cardiac AEs on letrozole (letrozole, 2.4%; tamoxifen, 1.4%; P = .001)--an excess only partially attributable to prior hypercholesterolemia--and more overall (tamoxifen, 3.9%; letrozole, 1.7%; P < .001) and grade 3 to 5 thromboembolic AEs on tamoxifen (tamoxifen, 2.3%; letrozole, 0.9%; P < .001). There was no significant difference between tamoxifen and letrozole in incidence of hypertension or cerebrovascular events. CONCLUSION: The present safety analysis, limited to cardiovascular AEs in BIG 1-98, documents a low overall incidence of cardiovascular AEs, which differed between treatment arms.

Efficient algorithms for dynamic probabilistic safety assessment. An application to convex damage domain.

The design of nuclear power plant has to follow a number of regulations aimed at limiting the risks inherent in this type of installation. The goal is to prevent and to limit the consequences of any possible incident that might threaten the public or the environment. To verify that the safety requirements are met a safety assessment process is followed. Safety analysis is as key component of a safety assessment, which incorporates both probabilistic and deterministic approaches. The deterministic approach attempts to ensure that the various situations, and in particular accidents, that are considered to be plausible, have been taken into account, and that the monitoring systems and engineered safety and safeguard systems will be capable of ensuring the safety goals. On the other hand, probabilistic safety analysis tries to demonstrate that the safety requirements are met for potential accidents both within and beyond the design basis, thus identifying vulnerabilities not necessarily accessible through deterministic safety analysis alone. Probabilistic safety assessment (PSA) methodology is widely used in the nuclear industry and is especially effective in comprehensive assessment of the measures needed to prevent accidents with small probability but severe consequences. Still, the trend towards a risk informed regulation (RIR) demanded a more extended use of risk assessment techniques with a significant need to further extend PSA’s scope and quality. Here is where the theory of stimulated dynamics (TSD) intervenes, as it is the mathematical foundation of the integrated safety assessment (ISA) methodology developed by the CSN(Consejo de Seguridad Nuclear) branch of Modelling and Simulation (MOSI). Such methodology attempts to extend classical PSA including accident dynamic analysis, an assessment of the damage associated to the transients and a computation of the damage frequency. The application of this ISA methodology requires a computational framework called SCAIS (Simulation Code System for Integrated Safety Assessment). SCAIS provides accident dynamic analysis support through simulation of nuclear accident sequences and operating procedures. Furthermore, it includes probabilistic quantification of fault trees and sequences; and integration and statistic treatment of risk metrics. SCAIS comprehensively implies an intensive use of code coupling techniques to join typical thermal hydraulic analysis, severe accident and probability calculation codes. The integration of accident simulation in the risk assessment process and thus requiring the use of complex nuclear plant models is what makes it so powerful, yet at the cost of an enormous increase in complexity. As the complexity of the process is primarily focused on such accident simulation codes, the question of whether it is possible to reduce the number of required simulation arises, which will be the focus of the present work. This document presents the work done on the investigation of more efficient techniques applied to the process of risk assessment inside the mentioned ISA methodology. Therefore such techniques will have the primary goal of decreasing the number of simulation needed for an adequate estimation of the damage probability. As the methodology and tools are relatively recent, there is not much work done inside this line of investigation, making it a quite difficult but necessary task, and because of time limitations the scope of the work had to be reduced. Therefore, some assumptions were made to work in simplified scenarios best suited for an initial approximation to the problem. The following section tries to explain in detail the process followed to design and test the developed techniques. Then, the next section introduces the general concepts and formulae of the TSD theory which are at the core of the risk assessment process. Afterwards a description of the simulation framework requirements and design is given. Followed by an introduction to the developed techniques, giving full detail of its mathematical background and its procedures. Later, the test case used is described and result from the application of the techniques is shown. Finally the conclusions are presented and future lines of work are exposed.

Core physics and safety analysis of Generation-IV Sodium Fast Reactors using existing and newly developed computational tools

El futuro de la energía nuclear de fisión dependerá, entre otros factores, de la capacidad que las nuevas tecnologías demuestren para solventar los principales retos a largo plazo que se plantean. Los principales retos se pueden resumir en los siguientes aspectos: la capacidad de proporcionar una solución final, segura y fiable a los residuos radiactivos; así como dar solución a la limitación de recursos naturales necesarios para alimentar los reactores nucleares; y por último, una mejora robusta en la seguridad de las centrales que en definitiva evite cualquier daño potencial tanto en la población como en el medio ambiente como consecuencia de cualquier escenario imaginable o más allá de lo imaginable. Siguiendo estas motivaciones, la Generación IV de reactores nucleares surge con el compromiso de proporcionar electricidad de forma sostenible, segura, económica y evitando la proliferación de material fisible. Entre los sistemas conceptuales que se consideran para la Gen IV, los reactores rápidos destacan por su capacidad potencial de transmutar actínidos a la vez que permiten una utilización óptima de los recursos naturales. Entre los refrigerantes que se plantean, el sodio parece una de las soluciones más prometedoras. Como consecuencia, esta tesis surgió dentro del marco del proyecto europeo CP-ESFR con el principal objetivo de evaluar la física de núcleo y seguridad de los reactores rápidos refrigerados por sodio, al tiempo que se desarrollaron herramientas apropiadas para dichos análisis. Efectivamente, en una primera parte de la tesis, se abarca el estudio de la física del núcleo de un reactor rápido representativo, incluyendo el análisis detallado de la capacidad de transmutar actínidos minoritarios. Como resultado de dichos análisis, se publicó un artículo en la revista Annals of Nuclear Energy [96]. Por otra parte, a través de un análisis de un hipotético escenario nuclear español, se evalúo la disponibilidad de recursos naturales necesarios en el caso particular de España para alimentar una flota específica de reactores rápidos, siguiendo varios escenarios de demanda, y teniendo en cuenta la capacidad de reproducción de plutonio que tienen estos sistemas. Como resultado de este trabajo también surgió una publicación en otra revista científica de prestigio internacional como es Energy Conversion and Management [97]. Con objeto de realizar esos y otros análisis, se desarrollaron diversos modelos del núcleo del ESFR siguiendo varias configuraciones, y para diferentes códigos. Por otro lado, con objeto de poder realizar análisis de seguridad de reactores rápidos, son necesarias herramientas multidimensionales de alta fidelidad específicas para reactores rápidos. Dichas herramientas deben integrar fenómenos relacionados con la neutrónica y con la termo-hidráulica, entre otros, mediante una aproximación multi-física. Siguiendo este objetivo, se evalúo el código de difusión neutrónica ANDES para su aplicación a reactores rápidos. ANDES es un código de resolución nodal que se encuentra implementado dentro del sistema COBAYA3 y está basado en el método ACMFD. Por lo tanto, el método ACMFD fue sometido a una revisión en profundidad para evaluar su aptitud para la aplicación a reactores rápidos. Durante ese proceso, se identificaron determinadas limitaciones que se discutirán a lo largo de este trabajo, junto con los desarrollos que se han elaborado e implementado para la resolución de dichas dificultades. Por otra parte, se desarrolló satisfactoriamente el acomplamiento del código neutrónico ANDES con un código termo-hidráulico de subcanales llamado SUBCHANFLOW, desarrollado recientemente en el KIT. Como conclusión de esta parte, todos los desarrollos implementados son evaluados y verificados. En paralelo con esos desarrollos, se calcularon para el núcleo del ESFR las secciones eficaces en multigrupos homogeneizadas a nivel nodal, así como otros parámetros neutrónicos, mediante los códigos ERANOS, primero, y SERPENT, después. Dichos parámetros se utilizaron más adelante para realizar cálculos estacionarios con ANDES. Además, como consecuencia de la contribución de la UPM al paquete de seguridad del proyecto CP-ESFR, se calcularon mediante el código SERPENT los parámetros de cinética puntual que se necesitan introducir en los típicos códigos termo-hidráulicos de planta, para estudios de seguridad. En concreto, dichos parámetros sirvieron para el análisis del impacto que tienen los actínidos minoritarios en el comportamiento de transitorios. Concluyendo, la tesis presenta una aproximación sistemática y multidisciplinar aplicada al análisis de seguridad y comportamiento neutrónico de los reactores rápidos de sodio de la Gen-IV, usando herramientas de cálculo existentes y recién desarrolladas ad' hoc para tal aplicación. Se ha empleado una cantidad importante de tiempo en identificar limitaciones de los métodos nodales analíticos en su aplicación en multigrupos a reactores rápidos, y se proponen interesantes soluciones para abordarlas. ABSTRACT The future of nuclear reactors will depend, among other aspects, on the capability to solve the long-term challenges linked to this technology. These are the capability to provide a definite, safe and reliable solution to the nuclear wastes; the limitation of natural resources, needed to fuel the reactors; and last but not least, the improved safety, which would avoid any potential damage on the public and or environment as a consequence of any imaginable and beyond imaginable circumstance. Following these motivations, the IV Generation of nuclear reactors arises, with the aim to provide sustainable, safe, economic and proliferationresistant electricity. Among the systems considered for the Gen IV, fast reactors have a representative role thanks to their potential capacity to transmute actinides together with the optimal usage of natural resources, being the sodium fast reactors the most promising concept. As a consequence, this thesis was born in the framework of the CP-ESFR project with the generic aim of evaluating the core physics and safety of sodium fast reactors, as well as the development of the approppriated tools to perform such analyses. Indeed, in a first part of this thesis work, the main core physics of the representative sodium fast reactor are assessed, including a detailed analysis of the capability to transmute minor actinides. A part of the results obtained have been published in Annals of Nuclear Energy [96]. Moreover, by means of the analysis of a hypothetical Spanish nuclear scenario, the availability of natural resources required to deploy an specific fleet of fast reactor is assessed, taking into account the breeding properties of such systems. This work also led to a publication in Energy Conversion and Management [97]. In order to perform those and other analyses, several models of the ESFR core were created for different codes. On the other hand, in order to perform safety studies of sodium fast reactors, high fidelity multidimensional analysis tools for sodium fast reactors are required. Such tools should integrate neutronic and thermal-hydraulic phenomena in a multi-physics approach. Following this motivation, the neutron diffusion code ANDES is assessed for sodium fast reactor applications. ANDES is the nodal solver implemented inside the multigroup pin-by-pin diffusion COBAYA3 code, and is based on the analytical method ACMFD. Thus, the ACMFD was verified for SFR applications and while doing so, some limitations were encountered, which are discussed through this work. In order to solve those, some new developments are proposed and implemented in ANDES. Moreover, the code was satisfactorily coupled with the thermal-hydraulic code SUBCHANFLOW, recently developed at KIT. Finally, the different implementations are verified. In addition to those developments, the node homogenized multigroup cross sections and other neutron parameters were obtained for the ESFR core using ERANOS and SERPENT codes, and employed afterwards by ANDES to perform steady state calculations. Moreover, as a result of the UPM contribution to the safety package of the CP-ESFR project, the point kinetic parameters required by the typical plant thermal-hydraulic codes were computed for the ESFR core using SERPENT, which final aim was the assessment of the impact of minor actinides in transient behaviour. All in all, the thesis provides a systematic and multi-purpose approach applied to the assessment of safety and performance parameters of Generation-IV SFR, using existing and newly developed analytical tools. An important amount of time was employed in identifying the limitations that the analytical nodal diffusion methods present when applied to fast reactors following a multigroup approach, and interesting solutions are proposed in order to overcome them.

Analysis of the operator action and the single failure criteria in a SGTR sequence using best estimate assumptions with TRACE 5.0

Steam Generator Tube Rupture (SGTR) sequences in Pressurized Water Reactors are known to be one of the most demanding transients for the operating crew. SGTR are a special kind of transient as they could lead to radiological releases without core damage or containment failure, as they can constitute a direct path from the reactor coolant system to the environment. The first methodology used to perform the Deterministic Safety Analysis (DSA) of a SGTR did not credit the operator action for the first 30 min of the transient, assuming that the operating crew was able to stop the primary to secondary leakage within that period of time. However, the different real SGTR accident cases happened in the USA and over the world demonstrated that the operators usually take more than 30 min to stop the leakage in actual sequences. Some methodologies were raised to overcome that fact, considering operator actions from the beginning of the transient, as it is done in Probabilistic Safety Analysis. This paper presents the results of comparing different assumptions regarding the single failure criteria and the operator action taken from the most common methodologies included in the different Deterministic Safety Analysis. One single failure criteria that has not been analysed previously in the literature is proposed and analysed in this paper too. The comparison is done with a PWR Westinghouse three loop model in TRACE code (Almaraz NPP) with best estimate assumptions but including deterministic hypothesis such as single failure criteria or loss of offsite power. The behaviour of the reactor is quite diverse depending on the different assumptions made regarding the operator actions. On the other hand, although there are high conservatisms included in the hypothesis, as the single failure criteria, all the results are quite far from the regulatory limits. In addition, some improvements to the Emergency Operating Procedures to minimize the offsite release from the damaged SG in case of a SGTR are outlined taking into account the offsite dose sensitivity results.

Sight distance for traffic safety analysis using GIS

Sight distance is of major importance for road safety either when designing new roads or analysing the alignment of existing roads. It is essential that available sight distance in roads is long enough for emergency stops or overtaking manoeuvres. Also, it is vital for engineers/researchers that the tools used for that analysis are both powerful and intuitive. Based on ArcGIS, the application to be presented not only performs an exhaustive sight distance calculation, but allows an accurate analysis of 3D alignment, using all new tools, from a Digital Elevation Model and vehicle trajectory. The software has been successfully utilised to analyse several two-lane rural roads in Spain. In addition, the software produces thematic maps representing sight distance in which supplementary information about crashes, traffic flow, speed or design consistency could be included, allowing traffic safety studies.

An evaluation of accident surrogates for safety analysis of rural highways. Volume 1 - executive summary. Final report.

Federal Highway Administration, Traffic Safety Research Division, McLean, Va.

An evaluation of accident surrogates for safety analysis of rural highways. Volume 2 - technical report. Final report.

Federal Highway Administration, Traffic Safety Research Division, McLean, Va.