RBMK Safety Analysis in Accidents Initiated by Partial Ruptures of the Circulation Circuit

2002 ◽  
Author(s):  
Anatoly I. Dostov ◽  
Alexander Ja. Kramerov
Keyword(s):  
Safety Evaluation ◽  
Safety Analysis ◽  
International Code ◽  
Fuel Rod ◽  
Current State ◽  
Circulation Circuit ◽  
Acceptability Criterion

The paper gives an analysis of the current state of the RBMK safety evaluation in accidents initiated by partial ruptures of the delivery part of the circulating loop. It appears from this analysis that applicability and uncertainty of the international code RELAP for RBMK safety analysis could not be determined up to the present. At the same time it is shown in the paper that fuel rod cladding temperature can reach the acceptability criterion in the accidents. As a result it has been concluded that bases of the next stage of the RBMK safety analysis would be creation of a code oriented to the special features of a reactor RBMK.

Using claims prediction model for road safety evaluation

2001 ◽  
Vol 28 (5) ◽  
pp. 804-812 ◽  
Author(s):  
Paul de Leur ◽  
Tarek Sayed
Keyword(s):  
Prediction Model ◽  
Road Safety ◽  
Claims Data ◽  
Safety Evaluation ◽  
Safety Analysis ◽  
Insurance Claims ◽  
Insurance Claim ◽  
Auto Insurance ◽  
Improvement Programs ◽  
Traffic Collision

Road safety analysis is typically undertaken using traffic collision data. However, the collision data often suffer from quality and reliability problems. These problems can inhibit the ability of road safety engineers to evaluate and analyze road safety performance. An alternate source of data that characterize the events of a traffic collision is the records that become available from an auto insurance claim. In settling an auto insurance claim, a claim adjuster must make an assessment and determination of the circumstances of the event, recording important contributing factors that led to the crash occurrence. As such, there is an opportunity to access and use the claims data in road safety engineering analysis. This paper presents the results of an initial attempt to use auto insurance claims records in road safety evaluation by developing and applying a claim prediction model. The prediction model will provide an estimate of the number of auto insurance claims that can be expected at signalized intersections in the Vancouver area of British Columbia, Canada. A discussion of the usefulness and application of the claim prediction model will be provided together with a recommendation on how the claims data could be utilized in the future.Key words: road safety improvement programs, auto insurance claims, road safety analysis, prediction models.

scholarly journals Probabilistic Safety Analysis of High Speed and Conventional Lines Using Bayesian Networks

2016 ◽  
Author(s):  
Zacarias Grande Andrade ◽  
Enrique Castillo Ron ◽  
Alan O'Connor ◽  
Maria Nogal
Keyword(s):  
Bayesian Network ◽  
High Speed ◽  
Human Error ◽  
Driver Behavior ◽  
Safety Evaluation ◽  
Safety Analysis ◽  
Human Errors ◽  
Maintenance Costs ◽  
Railway Line ◽  
Probabilistic Safety

A Bayesian network approach is presented for probabilistic safety analysis (PSA) of railway lines. The idea consists of identifying and reproducing all the elements that the train encounters when circulating along a railway line, such as light and speed limit signals, tunnel or viaduct entries or exits, cuttings and embankments, acoustic sounds received in the cabin, curves, switches, etc. In addition, since the human error is very relevant for safety evaluation, the automatic train protection (ATP) systems and the driver behavior and its time evolution are modelled and taken into account to determine the probabilities of human errors. The nodes of the Bayesian network, their links and the associated probability tables are automatically constructed based on the line data that need to be carefully given. The conditional probability tables are reproduced by closed formulas, which facilitate the modelling and the sensitivity analysis. A sorted list of the most dangerous elements in the line is obtained, which permits making decisions about the line safety and programming maintenance operations in order to optimize them and reduce the maintenance costs substantially. The proposed methodology is illustrated by its application to several cases that include real lines such as the Palencia-Santander and the Dublin-Belfast lines.DOI: http://dx.doi.org/10.4995/CIT2016.2016.3428

Evaluating the Safety Effectiveness of Advance Downgrade Warning Signs in Preventing Downgrade Truck Crashes using a Propensity Scores Framework

2019 ◽  
Vol 2673 (7) ◽  
pp. 673-683 ◽  
Author(s):  
Milhan Moomen ◽  
Mahdi Rezapour ◽  
Mustaffa N. Raja ◽  
Shaun S. Wulff ◽  
Khaled Ksaibati
Keyword(s):  
Risk Ratio ◽  
Propensity Scores ◽  
Safety Evaluation ◽  
Warning Signs ◽  
Quality Information ◽  
Bootstrap Confidence Interval ◽  
Analysis Techniques ◽  
Current State ◽  
Safety Effectiveness ◽  
Matched Data

Truck crashes on steep downgrades exact devastating tolls on lives and property. An important intervention to reduce the frequency of truck crashes on downgrades has been to present quality information to drivers about upcoming downgrades through warning signs. The use of advance warning signs on downgrades to prevent truck crashes is the current state of practice for most highway agencies; thus, it is critical to assess the safety effectiveness of such warning signs. Though several studies have evaluated the efficacies of warning signs, the safety effectiveness of advance downgrade signs is less clear. This study uses a propensity scores framework to assess the effectiveness of current downgrade warning signs by matching treated and untreated downgrade segments in Wyoming. The advantage of propensity score matching over other traditional analysis techniques is that it reduces bias in safety evaluation by mimicking randomization and accounts for confounding factors. Using the risk ratio computed from the matched data, the study found that advance downgrade warning signs are effective in reducing downgrade truck crashes. The results indicate that truck crash risks on downgrades without advance warning signs are an estimated 15% higher than downgrades with advance downgrade signs. The 90% bootstrap confidence interval for the risk ratio was found to be from 1.04 to 1.53.

scholarly journals Safety Analysis of Signal-Coordinated Arterials with Multiple Heterogeneities

2021 ◽  
Vol 2021 ◽  
pp. 1-13
Author(s):  
Yingfei Fan ◽  
Guopeng Zhang ◽  
Zhixuan Jia ◽  
Minjie Jin
Keyword(s):  
Temporal Change ◽  
Evaluation Model ◽  
Safety Evaluation ◽  
Safety Analysis ◽  
Traffic Data ◽  
Modeling Framework ◽  
Crash Frequency ◽  
Signal Coordination ◽  
Ann Arbor ◽  
Structural Heterogeneities

In the corresponding research available, the safety impact remains controversial in implementing signal coordination on arterials, which calls for an in-depth exploration with the appropriate statistical methods. Based on the traffic data from Ann Arbor City (Michigan, USA), the paper proposes a safety evaluation model considering the multiple heterogeneities. In terms of arterials with the coordinated signalization, modeling results imply that (1) the multivariate heterogeneity shows the strongest interaction on crash frequency, followed by the spatiotemporal and structural heterogeneities, and (2) the spatial variation is unrelated to the temporal change among crashes in the denoted traffic analysis zones (TAZs). In an attempt to alleviate the coupled crash risks along the coordinated arterials, the study emphasizes the necessity of dividing the subcontrol traffic areas in real time according to the correlative degree of crash distribution. Meanwhile, the modeling framework with multiple heterogeneities can be applied for the safety analysis of other urban roads.

scholarly journals Status and Assessment Method of Nuclear Safety Analysis Software in China

2021 ◽  
Vol 9 ◽  
Author(s):  
Xinli Gao ◽  
Jianping Jing ◽  
Xiangzhen Han ◽  
Bin Jia ◽  
Xinlu Tian ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Safety Evaluation ◽  
Evaluation Process ◽  
Safety Analysis ◽  
Assessment Method ◽  
Nuclear Safety ◽  
Analysis Software ◽  
Software Safety ◽  
Engineering Software ◽  
Software Applications

In recent years, China’s nuclear power industry has enjoyed a good momentum of development, and related companies have also developed many nuclear analysis software applications. However, as the National Nuclear Safety Administration (NNSA, Chinese nuclear regulatory institution) did not approve any software before 2018, all these software applications were not evaluated formally, so they have not yet been used in reactor safety analysis. In order to solve this problem, in 2018, the National Nuclear Safety Administration started to carry out an engineering applicability evaluation for software developed by Chinese companies. After several years of review, as the first approved Chinese domestic software, core physics analysis software PCM developed by the China General Nuclear Power Group officially passed the software safety evaluation of the China Nuclear Safety Administration. This study will present the basic situation of the development of China’s nuclear power engineering software and introduce the framework, methods, procedures, requirements, and other aspects of China’s software safety evaluation work. The evaluation process and evaluation key issues of PCM software will also be illustrated.

Evaluation of the Conservatism of NPP Safety Analysis Dose Calculations as Typical for Licensing Purposes

2010 ◽  
Author(s):  
Nikolaus Arnold ◽  
Nikolaus Mu¨llner ◽  
Francesco D’Auria ◽  
Oscar Mazzantini
Keyword(s):  
Safety Evaluation ◽  
Weather Conditions ◽  
Computer Code ◽  
Safety Analysis ◽  
Primary System ◽  
The Public ◽  
Best Estimate ◽  
Plant Personnel ◽  
Order Of Magnitude ◽  
Dispersion Calculation

Providing a reliable upper limit of radiological consequences to the plant personnel and the general public is typically the aim of a safety evaluation for anticipated operational occurrences or design basis accidents, as presented in a safety analysis report. A typical tool for dispersion calculation and dose evaluation is MACCS2. In the present analysis four types of calculations are presented: a first calculation, typical for licensing analysis, with the MACCS2 computer code. In a second step conservative assumptions e.g. ground release even if a stack release would be realistic, are dropped. In a third step calculation two is repeated with RODOS, a code (online decision making tool) used to predict the radiological consequences of an accidental release of activity. The step three calculation still contains all the conservative assumptions that are built in the MACCS2 code. In a last step these assumptions are removed, and a “best estimate” calculation on the dose to the public is performed. The whole analysis (step one to four) is repeated for different source terms (noble gases only, tritium dominated, primary system water …) and for different weather conditions. Two main conclusions can be drawn. The first by comparing step two (MACCS2) and step three (RODOS). Here the boundary conditions of the calculations are set to be as similar to each other as possible. The paper shows that despite the fact that MACCS2 uses a Gaussian plume model, while RODOS uses a puff model for dispersion calculation, doses of the same order of magnitude are calculated. For the second conclusion the step one (MACCS2, conservative) and step four (RODOS, best estimate) calculations are compared, it is shown that although the margin of conservatism varies considerably from case to case, the results differ at least one order of magnitude.

Safety Evaluation of the HTTR-IS Nuclear Hydrogen Production System

2010 ◽  
Author(s):  
Hiroyuki Sato ◽  
Hirofumi Ohashi ◽  
Yujiro Tazawa ◽  
Nariaki Sakaba ◽  
Yukio Tachibana
Keyword(s):  
Hydrogen Production ◽  
Production System ◽  
Evaluation Method ◽  
Cooling System ◽  
Safety Evaluation ◽  
Safety Analysis ◽  
Fuel Temperature ◽  
System Pressure ◽  
Production Of Hydrogen ◽  
The Government

Establishment of a safety evaluation method is one of the key issues for the nuclear hydrogen production demonstration since fundamental differences in the safety philosophy between nuclear plants and chemical plants exist. In the present study, a practical safety evaluation method, which enables to design, construct and operate hydrogen production plants under conventional chemical plant standards, is proposed. An event identification for the HTTR-IS nuclear hydrogen production system is conducted in order to select abnormal events which would change the scenario and quantitative results of the evaluation items from the existing HTTR safety evaluation. In addition, a safety analysis is performed for the identified events. The results of safety analysis for the indentified five Anticipated Operational Occurrences (AOOs) and three ACciDents (ACDs) show that evaluation items such as a primary cooling system pressure, temperatures of heat transfer tubes at pressure boundary, etc., do not exceed the acceptance criteria during the scenario. In addition, the increase of peak fuel temperature is small in the most severe case, and therefore the reactor core was not damaged and cooled sufficiently. These results will contribute to the safety review from the government and demonstration of the nuclear production of hydrogen.

Methodology for the Analysis of Fuel Behavior During LOCA and RIA for Licensing Purposes

2012 ◽  
Author(s):  
Martina Adorni ◽  
Alessandro Del Nevo ◽  
Francesco D’Auria
Keyword(s):  
Hydraulic System ◽  
Nuclear Power ◽  
Computer Code ◽  
Safety Analysis ◽  
Fuel Rod ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Fuel Pin ◽  
Computer Codes ◽  
Flow Blockage

Licensing requirements vary by country in terms of their scope, range of applicability and numerical values and may imply the use of system thermal hydraulic computer codes. Depending on the specific event scenario and on the purpose of the analysis, it might be required the availability of calculation methods that are not implemented in the standard system thermal hydraulic codes, as for burst temperature, burst strain and flow blockage calculations. This may imply the use of a dedicated fuel rod thermo-mechanical computer code, which can be coupled with thermal-hydraulic system and neutron kinetic codes to be used for the safety analysis. This paper describes the development and the application of a methodology for the analysis of the Large Break Loss of Coolant Accident (LB-LOCA) scenario in Atucha-2 Nuclear Power Plant (NPP), focusing on the procedure adopted for the use of the fuel rod thermo-mechanical code and its application for the safety analysis (Chapter 15 Final Safety Analysis Report, FSAR). The methodology implies the application of best estimate thermal-hydraulic, neutron physics and fuel pin performance computer codes, with the objective to verify the compliance with the specific acceptance criteria. The fuel pin performance code is applied with the main objective to evaluate the extent of cladding failures during the transient. A strong effort has been performed in order to enhance the fuel behaviour code capabilities and to improve the reliability of the code results.

TREQ: Fuel Rod Performance Code for Demanding PWR Plants

2009 ◽  
Author(s):  
I. Pastor ◽  
N. Doncel
Keyword(s):  
Fuel Cycle ◽  
Safety Evaluation ◽  
Mechanical Analysis ◽  
Physical Models ◽  
Principal Characteristics ◽  
Strategy Of Development ◽  
Fuel Rod ◽  
Safety Margins ◽  
Wide Range ◽  
Thermal Mechanical Analysis

The search for improvements in nuclear fuel cycle economics results in increasing demands for fuel operation, including higher discharged burnup, and plant manoeuvrability. To achieve these objectives without a reduction of safety margins, new fuel designs and materials that enable enhanced performance capabilities have been developed like the dopped pellets products or the new cladding materials. Besides, fuel rod performance code are incorporating the experience gained in these irradiation conditions. In these regards, the European Fuel Group (EFG) has developed the TREQ code. TREQ is a fuel rod thermal-mechanical code used for the design and safety evaluation of fuel rods for PWR reactors. TREQ benefits from more than 30 year experience coming from the PAD code. The TREQ development was started by Westinghouse, BNFL and ENUSA, the EFG and it is primarily applied to the thermal-mechanical analysis of rods operating at steady-state or slow transient operations as well as to provide fuel inputs, for other codes used in the accident analyses and it has been validated over a wide range of irradiation conditions. Thus, TREQ Fuel Rod Performance Code meets customers’ needs taking advantage of the performance of new products. The code comprises physical models which are developed on the basis of both theoretical considerations and experimental or in-pile data. The physical models of the code apply to Zircaloy-4 and ZIRLO cladding rods containing uranium and gadolinia doped pellets, fabricated according to ENUSA/Westinghouse specification. TREQ code has been licensed in Belgium up to a rod average burnup over 70 MWd/kgU for application on fuel rod design for PWR reactors. Besides, the code has been used to support several LOCA analyses in Belgium and other European countries. In Spain, TREQ was used in the licensing process of the high rod burnup programs. At present, EFG participates in several Research and Development (R&D) programs. The data at extended rod burnup, higher power and higher fuel densities, from theses programs will be incorporated to update models or reduce uncertainties. This paper presents the principal characteristics of the code, the main mechanical models and the strategy of development for the future.

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