The SCORPIO-VVER Core Monitoring and Surveillance System for VVER Type of Reactors

2014 ◽  
Author(s):  
Jozef Molnar ◽  
Radim Vocka
Keyword(s):  
Surveillance System ◽  
Nuclear Power ◽  
Power Plants ◽  
Slovak Republic ◽  
Technical Specification ◽  
Training Simulator ◽  
History Of ◽  
Monitoring And Surveillance ◽  
Technical Specifications ◽  
European Nuclear

The SCORPIO-VVER core monitoring and surveillance system has proved since the first installation at Dukovany NPP in 1999 to be a valuable tool for the reactor operators and reactor physicists. It is now installed on four units of Dukovany NPP (EDU, Czech Republic), on two units of Bohunice NPP (EBO, Slovak Republic) replacing the original Russian VK3 system and on the full scale plant training simulator at the Centre for training and education of the reactor operators and reactor physicist in Trnava (Slovak Republic). By both Czech and Slovak nuclear regulatory bodies the system was licensed as a Technical Specification Surveillance tool. Since it’s first installation, the development of SCORPIO-VVER system continues along with the changes in VVER reactors operation. The system is being adapted according the utility needs and several notable improvements in physical modules of the system were introduced. The most significant changes were done in support of the latest optimized Gd bearing fuel assemblies, improvements in the area of core design (neutron physics, core thermal hydraulics and fuel thermal mechanics), adaptation of the system to up-rated unit conditions (uprated power up to 107%), in design and methodology of the limit and technical specifications checking and improvements in the predictive part of the system. After the currently finished upgrades the SCORPIO-VVER is still in focus of Central European nuclear power plants with the roadmap of upgrades and modifications up to 2016. This paper shortly describes the system’s main functions, the history of implementation at the VVER-440 type of reactors and deals with the system’s future upgrades and plans to meet the latest requirements of efficient and safety NPP operation.

The reactor research and test facility

2020 ◽  
pp. 69-79
Author(s):  
Andrey S. KIRILLOV ◽  
Aleksandr P. PYSHKO ◽  
Andrey A. ROMANENKO ◽  
Valery I. YARYGIN
Keyword(s):  
Measurement System ◽  
Nuclear Power ◽  
Power Plants ◽  
Test Facility ◽  
Physical Parameters ◽  
Pumping System ◽  
Current State ◽  
History Of ◽  
Special Measurement ◽  
Space Nuclear Power

The paper describes an overview of the history of development and the current state of JSC “SSC RF-IPPE” reactor research and test facility designed for assembly, research and full-scale life energy tests of space nuclear power plants with a thermionic reactor. The leading specialists involved in development and operation of this facility are represented. The most significant technological interfaces and upgrade operations carried out in the recent years are discussed. The authors consider the use of an oil-free pumping system as part of this facility during degassing and life testing. Proposed are up-to-date engineering solutions for development of the automated special measurement system designed to record NPP performance, including volt-ampere characteristics together with thermophysical and nuclear physical parameters of a ground prototype of the space nuclear power plant. Key words: reactor research and test facility, thermionic reactor, life energy tests, oil-free pumping system, automated special measurement system, volt-ampere characteristics.

scholarly journals Plant Performance History of an Innovative Gate Valve in Critical Service Applications

2017 ◽  
Author(s):  
M. S. Kalsi ◽  
Patricio Alvarez ◽  
Thomas White ◽  
Micheal Green
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Gate Valve ◽  
Plant Performance ◽  
Flow Loop ◽  
Performance History ◽  
Pressurized Water ◽  
Maintenance Costs ◽  
History Of ◽  
Water Reactors

A previous paper [1] describes the key features of an innovative gate valve design that was developed to overcome seat leakage problems, high maintenance costs as well as issues identified in the Nuclear Regulatory Commission (NRC) Generic Letters 89-10, 95-07 and 96-05 with conventional gate valves [2,3,4]. The earlier paper was published within a year after the new design valves were installed at the Pilgrim Nuclear Plant — the plant that took the initiative to form a teaming arrangement as described in [1] which facilitated this innovative development. The current paper documents the successful performance history of 22 years at the Pilgrim plant, as well as performance history at several other nuclear power plants where these valves have been installed for many years in containment isolation service that requires operation under pipe rupture conditions and require tight shut-off in both Pressurized Water Reactors (PWRs) and Boiling Water Reactors (BWRs). The performance history of the new valve has shown to provide significant performance advantage by eliminating the chronic leakage problems and high maintenance costs in these critical service applications. This paper includes a summary of the design, analysis and separate effects testing described in detail in the earlier paper. Flow loop testing was performed on these valves under normal plant operation, various thermal binding and pressure locking scenarios, and accident/pipe rupture conditions. The valve was designed, analyzed and tested to satisfy the requirements of ANSI B16.41 [9]; it also satisfies the requirements of ASME QME 1-2012 [10]. The results of the long-term performance history including any degradation observed and its root cause are summarized in the paper. Paper published with permission.

Review of the Configuration Risk Management Methodologies

2021 ◽  
Author(s):  
Yuhang Zhang ◽  
Zhijian Zhang ◽  
He Wang ◽  
Lixuan Zhang ◽  
Dabin Sun
Keyword(s):  
Risk Management ◽  
Real Time ◽  
Nuclear Power ◽  
Safety Assessment ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Probabilistic Safety Assessment ◽  
Technical Specifications ◽  
Risk Monitor ◽  
Probabilistic Safety

Abstract To ensure nuclear safety and prevent or mitigate the consequences of accidents, many safety systems have been set up in nuclear power plants to limit the consequences of accidents. Even though technical specifications based on deterministic safety analysis are applied to avoid serious accidents, they are too poor to handle multi-device managements compared with configuration risk management which computes risks in nuclear power plants based on probabilistic safety assessment according to on-going configurations. In general, there are two methodologies employed in configuration risk management: living probabilistic safety assessment (LPSA) and risk monitor (RM). And average reliability databases during a time of interest are employed in living probabilistic safety assessment, which may be naturally applied to make long-term or regular management projects. While transient risk databases are involved in risk monitor to measure transient risks in nuclear power plants, which may be more appropriate to monitor the real-time risks in nuclear power plants and provide scientific real-time suggestions to operators compared with living probabilistic safety assessment. And this paper concentrates on the applications and developments of living probabilistic safety assessment and risk monitor which are the mainly foundation of the configuration risk management to manage nuclear power plants within safe threshold and avoid serious accidents.

RELIABILITY BASED OPTIMIZATION OF TECHNICAL SPECIFICATION OF FRONTLINE SYSTEMS OF NUCLEAR POWER PLANTS USING MULTI-OBJECTIVE APPROACH

2012 ◽  
Vol 19 (01) ◽  
pp. 1250002
Author(s):  
G. SRINIVAS ◽  
A. K. VERMA ◽  
A. SRIVIDYA ◽  
SANJAY KUMAR KHATTRI
Keyword(s):  
Genetic Algorithm ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Technical Specification ◽  
Multi Objective Optimization ◽  
Potential Core ◽  
Multi Objective ◽  
Genetic Algorithm Approach ◽  
Technical Specifications

Technical Specifications define the limiting conditions of operation, maintenance and surveillance test requirements for the various Nuclear Power plant systems in order to meet the safety requirements to fulfill regulatory criteria. These specifications impact even the economics of the plant. The regulatory approach addresses only the safety criteria, while the plant operators would like to balance the cost criteria too. The attempt to optimize both the conflicting requirements presents a case to use Multi-objective optimization. Evolutionary algorithms (EAs) mimic natural evolutionary principles to constitute search and optimization procedures. Genetic algorithms are a particular class of EA's that use techniques inspired by evolutionary biology such as inheritance, mutation, natural selection and recombination (or cross-over). In this paper we have used the plant insights obtained through a detailed Probabilistic Safety Assessment with the Genetic Algorithm approach for Multi-objective optimization of Surveillance test intervals. The optimization of Technical Specifications of three front line systems is performed using the Genetic Algorithm Approach. The selection of these systems is based on their importance to the mitigation of possible accident sequences which are significant to potential core damage of the nuclear power plant.

Environmental Assisted Fatigue and EDF 900 MWe PWRs Fleet: Towards an Exemption of Environmental Effects Consideration for Secondary Circuit Components

2018 ◽  
Author(s):  
Sam Cuvilliez ◽  
Gaëlle Léopold ◽  
Thomas Métais
Keyword(s):  
Environmental Effects ◽  
Nuclear Power ◽  
Power Plants ◽  
Fatigue Loading ◽  
Degradation Process ◽  
Fatigue Curve ◽  
Primary Circuit ◽  
Secondary Circuit ◽  
Technical Specifications ◽  
Circuit Components

Environmentally Assisted Fatigue (EAF) is receiving nowadays an increased level of attention for existing Nuclear Power Plants (NPPs) as utilities are now working to extend their life. In the wake of numerous experimental fatigue tests carried out in air and also in a PWR environment, the French RCC-M code [1] has recently been amended (in its 2016 edition) with two Rules in Probatory Phase (RPP), equivalent to ASME code-cases, “RPP-2” and “RPP-3” [2] [3]. RPP-2 consists of an update of the design fatigue curve in air for stainless steels (SSs) and nickel-based alloys, and is also associated with RPP-3 which provides guidelines for incorporating the environmental penalty “Fen” factor in fatigue usage factor calculations. Alongside this codification effort, an EAF screening has recently been carried out within EDF DT [4] on various areas of the primary circuit of the 900 MWe plants of the EDF fleet. This screening led to the identification of a list of 35 “sentinel locations” which are defined as areas most prone to EAF degradation process. These locations will be subjected to detailed EAF analysis in the stress report calculations (according to the above-mentioned RCC-M code cases) for the fourth decennial inspection of the 900 MWe (VD4 900 MWe) power plants. The potential impact of EAF on the secondary circuit components is another question to address in anticipation of the VD4 900 MWe, as they may be considered as class 1 or class 2 equipment for RCC-M application according to the equipment specification. This paper presents the approach proposed by EDF towards an exemption of environmental effects consideration for secondary circuit components. The argument is first based on a review of experimental campaigns led in Japan and France (respectively on fatigue test specimens and at the component scale) which indicate a Dissolved Oxygen (DO) content threshold below which environmental effects are almost inexistent. The (conservative) value of 40 ppb has been selected consistently with NUREG/CR-6909 revision 0 [5]. The second part of the argument is built, on the one hand, on the analysis of the EDF Technical Specifications for Operation (STE) which narrows the scope of the study only to unit outages, and, on the other hand, on the analysis of 5 years of operations of all 900 MWe plants of the EDF fleet (equivalent to 170 reactor-years). It has been shown that the DO content rarely exceeded the 40 ppb threshold in the secondary coolant, and that in this case, the considered locations were not submitted to any fatigue loading.

scholarly journals Contributions of the ORNL Piping Program to Nuclear Piping Design Codes and Standards

1977 ◽  
Vol 99 (1) ◽  
pp. 224-230
Author(s):  
S. E. Moore
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Manufacturing Industry ◽  
Stress Index ◽  
Paper Briefly ◽  
Technical Literature ◽  
Index Method ◽  
Engineering Data ◽  
History Of ◽  
Future Work

Since 1967, ORNL Piping Program has been engaged in providing information for the development of stress indices to be used in the analysis of piping components for nuclear power plants. This effort has surveyed the piping manufacturing industry and analyzed that industry’s products; has combed the technical literature for pertinent engineering data; has performed theoretical and experimental analysis of nuclear piping components; and has defined, tested, and improved indices for the stress-index method of analysis for piping components. This paper briefly reviews the history of piping-analysis standards; outlines the philosophy of the stress-index method of analysis; and explains some of the specific contributions made by the ORNL program to the Codes and Standards. Current and future work is also noted.

An Aging Simulation Model for the Emergency Diesel Generators of Angra 2 by Supplementary Variables

2008 ◽  
Author(s):  
Iona´ Maghali Santos de Oliveira ◽  
Paulo Fernando Ferreira Frutuoso e Melo ◽  
Marcos Oliveira de Pinho
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Point Of View ◽  
Failure Rates ◽  
Aging Effects ◽  
Supplementary Variables ◽  
Technical Specifications ◽  
Diesel Generators ◽  
Markovian Approach ◽  
Useful Lifetime

The extension of the useful lifetime of nuclear power plants is an issue of great importance and concern. From the reliability point of view, this problem requires the consideration of time-dependent failure rates and possible failure dependencies. This analysis has been typically performed through a Markovian approach. To illustrate this point, we have developed a computerized reliability analysis of the emergency diesel generators (EDGs) of a four “loop” PWR plant, considering the hypothesis of aging and perfect repair by using Supplementary Variables to cast the initially Nonmarkovian model into a Markovian one. In order to perform such analysis and to simulate aging effects, a nuclear plant has been taken for reference, which has been commercially operating for only six years. Failure rates were obtained from similar EDGs of another plant, already under aging, while repair data were taken from its technical specifications. Discontinuous repair rates were considered in order to improve maintenance strategies. Several ages were attributed to these equipments, allowing the calculus of the failure probability as well as their availability according to each regarded age. In this sense, the EDGs behavior as to aging can be obtained in detail and decisions concerning maintenance and useful lifetime extension can be made on a stronger basis. To get the desired results in terms of reliability figures and due to the discontinuous repair rates that had to be taken into account, a new numerical method that uses a part of the analytical solution, called Euler Iterative + Characteristic, has been developed in order to solve the differential equations systems, making the solving of the system faster and more efficient.

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