Plant Safety

2021 ◽  
pp. 227-268
Author(s):  
Frank R. Spellman ◽  
Lorilee Medders ◽  
Paul Fuller ◽  
Gordon Graham
Keyword(s):  
Plant Safety

scholarly journals Aircraft Impact Effects on an Aged NPP

Materials ◽  
2021 ◽  
Vol 14 (4) ◽  
pp. 816
Author(s):  
Rosa Lo Frano
Keyword(s):  
Nuclear Power ◽  
Military Aircraft ◽  
Residual Resistance ◽  
Plant Safety ◽  
Need To Evaluate ◽  
Target Wall ◽  
Impact Area ◽  
Wall Perforation ◽  
The Impact ◽  
Loss Of Strength

The impact of an aircraft is widely known to be one of the worst events that can occur during the operation of a plant (classified for this reason as beyond design). This can become much more catastrophic and lead to the loss of strength of/collapse of the structures when it occurs in the presence of ageing (degradation and alteration) materials. Therefore, since the performance of all plant components may be affected by ageing, there is a need to evaluate the effect that aged components have on system performance and plant safety. This study addresses the numerical simulation of an aged Nuclear Power Plant (NPP) subjected to a military aircraft impact. The effects of impact velocity, direction, and location were investigated together with the more unfavorable conditions to be expected for the plant. The modelling method was also validated based on the results obtained from the experiments of Sugano et al., 1993. Non-linear analyses by means of finite element (FE) MARC code allowed us to simulate the performance of the reinforced concrete containment building and its impact on plant availability and reliability. The results showed that ageing increases a plant’s propensity to suffer damage. The damage at the impact area was confirmed to be dependent on the type of aircraft involved and the target wall thickness. The greater the degradation of the materials, the lower the residual resistance capacity, and the greater the risk of wall perforation.

Effect of Creep on the Residual Stresses in Tube-to-Tubesheet Joints

2010 ◽  
Author(s):  
Nor Eddine Laghzale ◽  
Abdel-Hakim Bouzid
Keyword(s):  
Contact Pressure ◽  
Analytical Model ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Environmental Safety ◽  
Maintenance Cost ◽  
Joint Interface ◽  
Plant Safety ◽  
Material Creep ◽  
Nuclear Power Plant Safety

Steam generators are the subject of major concern in nuclear power plant safety. Within these generators, in addition to the structural integrity, the gross tightness barrier, which separates the primary and secondary circuits, is primarily ensured by the presence of a residual contact pressure at the tube-to-tubesheet joint interface. Any leakage is unacceptable, and its consequences are very heavy in terms of the human and environmental safety as well as maintenance cost. Some studies have been conducted to understand the main reasons for such a failure. However, no analytical model able to predict the attenuation of the residual contact pressure under the effect of material creep relaxation behavior. The development of a simple analytical model able to predict the change of the residual contact pressure as a function of time is laid out in this paper. The results from the analytical model are checked and compared with those of finite elements.

scholarly journals The SPES3 Facility for Testing an Integral Layout SMR: BDBE Simulation Analysis

2011 ◽  
Author(s):  
Roberta Ferri ◽  
Andrea Achilli ◽  
Cinzia Congiu ◽  
Gustavo Cattadori ◽  
Fosco Bianchi ◽  
...  
Keyword(s):  
Simulation Analysis ◽  
Nuclear Fission ◽  
Sensitivity Analyses ◽  
Medium Size ◽  
Design Calculation ◽  
Pressurized Water ◽  
Pressure Equalization ◽  
Plant Safety ◽  
Design Basis ◽  
Safety Features

The SPES3 facility is being built at the SIET laboratories, in the frame of an R&D program on Nuclear Fission, led by ENEA and funded by the Italian Ministry of Economic Development. The facility is based on the IRIS reactor design, an advanced medium size, integral layout, pressurized water reactor, based on the proven technology of PWR with an innovative configuration and safety features suitable to cope with Loss of Coolant Accidents through a dynamic coupling of the primary and containment systems. SPES3 is suitable to test the plant response to postulated Design and Beyond Design Basis Events, providing experimental data for code validation and plant safety analysis. It reproduces the primary, secondary and containment systems of the reactor with 1:100 volume scale, full elevation, prototypical fluid and thermal-hydraulic conditions. A design-calculation feedback process, based on the comparison between IRIS and SPES3 simulations, performed respectively by FER, with GOTHIC and RELAP5 coupled codes, and by SIET, with RELAP5 code, led to reduce the differences in the two plants behaviour, versus a 2-inch equivalent DVI line DEG break, considered the most challenging LOCA for the IRIS plant. Once available the final design of SPES3, further calculations were performed to investigate Beyond Design Basis Events, where the intervention of the Passive Containment Condenser is fundamental for the accident recovery. Sensitivity analyses showed the importance of the PCC actuation time, to limit the containment pressure, to reach an early pressure equalization between the primary and containment systems and to allow passive water transfer from the containment to the RPV, enhanced by the ADS Stage-II opening.

Criticality analysis for safety-critical software in nuclear power plant distributed control system

Kerntechnik ◽  
2021 ◽  
Vol 86 (5) ◽  
pp. 343-352
Author(s):  
J. Cui ◽  
Y. Cai ◽  
Y. Wu
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Application Process ◽  
Software Module ◽  
Critical System ◽  
Safety Critical ◽  
Plant Safety ◽  
Safety Integrity Level ◽  
Criticality Analysis

Abstract Software criticality analysis examines the degree of contribution that each individual failure mode of a software component has on the reliability of software. Higher safety integrity levels are assigned to software modules whose failures cause an unacceptable impact on the operation of the system, and these levels require the implementation of more rigorous software quality assurance measures as defined in IEEE Std 1012 and in the customer’s system requirements specification. In this paper, a novel software criticality analysis method is proposed, the results of which can be used to guide the development of newly developed software and the procurement of Commercial-Off-The-Shelf (COTS) software. The software structure is first analyzed and the software is divided into modules according to their functions. Then the criticality levels of software components are preliminarily classified by means of a safety criticality preliminary analysis tree, followed by their verification through the software hazard and operability analysis (HAZOP). Finally, the target Safety Integrity Level (SIL) of each software module is determined based on its criticality level and the overall safety objective (i. e., SIL) of the system it resides in. As an example, this proposed method is applied to a nuclear power plant safety-critical system to demonstrate the detail application process and to verify the feasibility of the method. Compared with the existing software criticality analysis methods, this method has better operability and verifiability, and can be utilized as a technical guidance for the software criticality analysis of nuclear power plant digital control systems.

Design of the Communication Independence for ACPR1000 Nuclear Power Plant Digital Safety I&C System

2017 ◽  
Author(s):  
Sun Na ◽  
Shi Gui-lian ◽  
Xie Yi-qin ◽  
Li Gang ◽  
Jiang Guo-jin
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Communication Networks ◽  
Nuclear Power ◽  
Protection System ◽  
Plant Safety ◽  
Safety And Reliability ◽  
Digital Protection ◽  
Architecture And Design ◽  
Reactor Protection System

Communication independence is one of the key criteria of digital safety I&C system design. This paper mainly analyzes the requirements for communication independence in safety regulations and standards, and then introduces the architecture and design features, including communication failure processing measures, of communication networks of ACPR1000 nuclear power plant safety digital protection system based on FirmSys platform developed by CTEC. The communication design meets the regulations requirements and effectively improves the safety and reliability of the system, and it is successfully applied in reactor protection system (RPS) of Yang Jiang nuclear power plant unit 5&6. In addition this design can provide reference for communication designs of other NPPs and industries.

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