Seismic Hazard and Damages—Avoiding Disaster Through Simulation, Experiment, and Experience

1999 ◽  
Vol 121 (1) ◽  
pp. 30-36 ◽  
Author(s):  
H. Shibata
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Mechanical Engineering ◽  
Design Procedure ◽  
Shaking Table ◽  
Test Facility ◽  
Design Stage ◽  
Engineering Structures ◽  
Aseismic Design ◽  
Piping Systems

This paper deals with the role of proving tests and a large shaking test facility for equipment and piping systems in conjunction with the development of aseismic design in the field of mechanical engineering, especially for nuclear power plants in Japan. To avoid seismic disaster and damage of equipment and piping systems as well as liquid storages, we had to differentiate the seismic design procedure in mechanical engineering from that for building and civil engineering structures. For this process, the dynamic analysis in this field is more significant than for other fields. The author has been trying to develop aseismic design since the design stage of the first nuclear power plant in 1958 based on his experience as a specialist of mechanical vibration. In the early 1970s, shaking tables were developed for this purpose in Japan. The largest one in Japan is a 1000-ton 2-D table. After the 1995 Kobe earthquake, we have been developing a new 1200-ton 3-D shaking table. In the paper, the author discusses the necessity of such a facility and presents a new concept of a numerical shaking table.

Analysis of ASME Class 3 Buried HDPE Piping Systems Related to Code Case N-755-1

2013 ◽  
Author(s):  
Young Seok Kim ◽  
Jung Kwang Yoon ◽  
Young Ho Kim
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Design Procedure ◽  
Piping System ◽  
Service Water ◽  
Hdpe Pipe ◽  
Division 1 ◽  
Piping Systems ◽  
Asme Code ◽  
Essential Service

This paper proposes an analysis method for Section III, Division 1, Class 3 buried High Density polyethylene (HDPE) piping system in the nuclear power plants (NPP). Although HDPE pipe would yield at high temperature (limited to 140°F), it may be suitable for the areas prone to earthquakes; owing to its comparable ductility and flexibility. Thus, the buried HDPE piping may be applicable for the safety related Essential Service Water (ESW) system in the NPPs. Despite some limitations to buried HDPE piping, the piping could be designed based on ASME Code Case [1]. Generally, codes and standards including ASME Code Case [1] do not provide load combinations for the design of both buried steel piping and HDPE piping. Meanwhile, EPRI Report [4] provides load combinations including thermal expansion effects and seismic loads with detailed seismic criteria for polyethylene pipe. In this paper, load cases and load combinations for buried HDPE piping are suggested for implementation of reference documents and a buried HDPE piping system is analyzed referring to EPRI Report [4] to evaluate stress, force, and moment using a piping stress analysis program. Additionally, this paper will recommend the design procedure in accordance with ASME Code Case [1] using an example of buried HDPE piping analysis. An investigation of soil spring coefficients and the design considerations for hydrostatic tests are suggested for the enhanced analysis of buried HDPE piping.

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 Simulation of MASPn Experiments in MISTRA Test Facility with COCOSYS Code

2008 ◽  
Vol 2008 ◽  
pp. 1-7 ◽  
Author(s):  
Mantas Povilaitis ◽  
Egidijus Urbonavičius
Keyword(s):  
Boundary Conditions ◽  
Nuclear Power ◽  
Power Plants ◽  
Volume Fraction ◽  
Characteristic Length ◽  
Initial Conditions ◽  
Test Facility ◽  
Work Package ◽  
Acceptable Agreement

An issue of the stratified atmospheres in the containments of nuclear power plants is still unresolved; different experiments are performed in the test facilities like TOSQAN and MISTRA. MASPn experiments belong to the spray benchmark, initiated in the containment atmosphere mixing work package of the SARNET network. The benchmark consisted of MASP0, MASP1 and MASP2 experiments. Only the measured depressurisation rates during MASPn were available for the comparison with calculations. When the analysis was performed, the boundary conditions were not clearly defined therefore most of the attention was concentrated on MASP0 simulation in order to develop the nodalisation scheme and define the initial and boundary conditions. After achieving acceptable agreement with measured depressurisation rate, simulations of MASP1 and MASP2 experiments were performed to check the influence of sprays. The paper presents developed nodalisation scheme of MISTRA for the COCOSYS code and the results of analyses. In the performed analyses, several parameters were considered: initial conditions, loss coefficient of the junctions, initial gradients of temperature and steam volume fraction, and characteristic length of structures. Parametric analysis shows that in the simulation the heat losses through the external walls behind the lower condenser installed in the MISTRA facility determine the long-term depressurisation rate.

Operating Reliability Assessment of FPGA-Based NPP I&C Systems: Approach, Technique and Implementation

2017 ◽  
Author(s):  
Eugene Babeshko ◽  
Ievgenii Bakhmach ◽  
Vyacheslav Kharchenko ◽  
Eugene Ruchkov ◽  
Oleksandr Siora
Keyword(s):  
Integrated Circuits ◽  
Nuclear Power ◽  
Power Plants ◽  
Failure Modes ◽  
New Technologies ◽  
Systems Approach ◽  
Reliability Assessment ◽  
Design Stage ◽  
Fpga Design ◽  
Operating Reliability

Operating reliability assessment of instrumentation and control systems (I&Cs) is always one of the most important activities, especially for critical domains like nuclear power plants (NPPs). Intensive use of relatively new technologies like field programmable gate arrays (FPGAs) in I&C which appear in upgrades and in newly built NPPs makes task to develop and validate advanced operating reliability assessment methods that consider specific technology features very topical. Increased integration densities make the reliability of integrated circuits the most crucial point in modern NPP I&Cs. Moreover, FPGAs differ in some significant ways from other integrated circuits: they are shipped as blanks and are very dependent on design configured into them. Furthermore, FPGA design could be changed during planned NPP outage for different reasons. Considering all possible failure modes of FPGA-based NPP I&C at design stage is a quite challenging task. Therefore, operating reliability assessment is one of the most preferable ways to perform comprehensive analysis of FPGA-based NPP I&Cs. This paper summarizes our experience on operating reliability analysis of FPGA based NPP I&Cs.

Nuclear Power Plant Fires and Explosions: Part IV — Water Hammer Explosion Mechanisms

2017 ◽  
Author(s):  
Robert A. Leishear
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Reactor ◽  
Water Hammer ◽  
Compression Process ◽  
Flammable Gas ◽  
Piping Systems ◽  
Fukushima Daiichi ◽  
Accident Conditions

Water hammers, or fluid transients, compress flammable gasses to their autognition temperatures in piping systems to cause fires or explosions. While this statement may be true for many industrial systems, the focus of this research are reactor coolant water systems (RCW) in nuclear power plants, which generate flammable gasses during normal operations and during accident conditions, such as loss of coolant accidents (LOCA’s) or reactor meltdowns. When combustion occurs, the gas will either burn (deflagrate) or explode, depending on the system geometry and the quantity of the flammable gas and oxygen. If there is sufficient oxygen inside the pipe during the compression process, an explosion can ignite immediately. If there is insufficient oxygen to initiate combustion inside the pipe, the flammable gas can only ignite if released to air, an oxygen rich environment. This presentation considers the fundamentals of gas compression and causes of ignition in nuclear reactor systems. In addition to these ignition mechanisms, specific applications are briefly considered. Those applications include a hydrogen fire following the Three Mile Island meltdown, hydrogen explosions following Fukushima Daiichi explosions, and on-going fires and explosions in U.S nuclear power plants. Novel conclusions are presented here as follows. 1. A hydrogen fire was ignited by water hammer at Three Mile Island. 2. Hydrogen explosions were ignited by water hammer at Fukushima Daiichi. 3. Piping damages in U.S. commercial nuclear reactor systems have occurred since reactors were first built. These damages were not caused by water hammer alone, but were caused by water hammer compression of flammable hydrogen and resultant deflagration or detonation inside of the piping.

Sensitivity Analyses for a PWR SCC Case Using the Probabilistic Loss-of-Coolant-Accident (Pro-LOCA) Software

2016 ◽  
Author(s):  
Bruce A. Young ◽  
Sang-Min Lee ◽  
Paul M. Scott
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
The United States ◽  
Design Criterion ◽  
Mitigation Strategies ◽  
Sensitivity Analyses ◽  
Base Case ◽  
Piping System ◽  
Loss Of Coolant Accident ◽  
Piping Systems

As a means of demonstrating compliance with the United States Code of Federal Regulations 10CFR50 Appendix A, General Design Criterion 4 (GDC-4) requirement that primary piping systems for nuclear power plants exhibit an extremely low probability of rupture, probabilistic fracture mechanics (PFM) software has become increasingly popular. One of these PFM codes for nuclear piping is Pro-LOCA which has been under development over the last decade. Currently, Pro-LOCA is being enhanced under an international cooperative program entitled PARTRIDGE-II (Probabilistic Analysis as a Regulatory Tool for Risk-Informed Decision GuidancE - Phase II). This paper focuses on the use of a pre-defined set of base-case inputs along with prescribed variation in some of those inputs to determine a comparative set of sensitivity analyses results. The benchmarking case was a circumferential Primary Water Stress Corrosion Crack (PWSCC) in a typical PWR primary piping system. The effects of normal operating loads, temperature, leak detection, inspection frequency and quality, and mitigation strategies on the rupture probability were studied. The results of this study will be compared to the results of other PFM codes using the same base-case and variations in inputs. This study was conducted using Pro-LOCA version 4.1.9.

Effects on Pipe Vibrations of Cavitation in an Orifice and in Globe-Style Valves

2006 ◽  
Author(s):  
Se´bastien Caillaud ◽  
Rene´-Jean Gibert ◽  
Pierre Moussou ◽  
Joe¨l Cohen ◽  
Fabien Millet
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Low Frequency ◽  
Flow Regimes ◽  
Piping System ◽  
Low Frequencies ◽  
Choked Flow ◽  
Single Hole ◽  
Piping Systems ◽  
Large Amplitude Vibrations

A piping system of French nuclear power plants displays large amplitude vibrations in particular flow regimes. These troubles are attributed to cavitation generated by single-hole orifices in depressurized flow regimes. Real scale experiments on high pressure test rigs and on-site tests are then conducted to explain the observed phenomenon and to find a solution to reduce pipe vibrations. The first objective of the present paper is to analyze cavitation-induced vibrations in the single-hole orifice. It is then shown that the orifice operates in choked flow with supercavitation, which is characterized by a large unstable vapor pocket. One way to reduce pipe vibrations consists in suppressing the orifices and in modifying the control valves. Three technologies involving a standard trim and anti-cavitation trims are tested. The second objective of the paper is to analyze cavitation-induced vibrations in globe-style valves. Cavitating valves operate in choked flow as the orifice. Nevertheless, no vapor pocket appears inside the pipe and no unstable phenomenon is observed. The comparison with an anti-cavitation solution shows that cavitation reduction has no impact on low frequency excitation. The effect of cavitation reduction on pipe vibrations, which involve essentially low frequencies, is then limited and the first solution, which is the standard globe-style valve installed on-site, leads to acceptable pipe vibrations. Finally, this case study may have consequences on the design of piping systems. First, cavitation in orifices must be limited. Choked flow in orifices may lead to supercavitation, which is here a damaging and unstable phenomenon. The second conclusion is that the reduction of cavitation in globe-style valve in choked flow does not reduce pipe vibrations. The issue is then to limit cavitation erosion of valve trims.

scholarly journals Anti-Earthquake Design of Industrial Facilities

2006 ◽  
Vol 1 (2) ◽  
pp. 190-200
Author(s):  
Heki Shibata ◽  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
The United States ◽  
Potential Hazard ◽  
Oil Refineries ◽  
Probability Of Occurrence ◽  
Aseismic Design ◽  
Low Probability

In Japan two sets of guidelines pertaining to modern aseismic design are being prepared. One is the guideline for the aseismic design of petrochemical plants and oil refineries, and the other is the code of aseismic design of nuclear power plants. The International Atomic Energy Agency also established its own guideline very recently. Several other countries also provide their own codes or guidelines. Among these, the regulatory guides of the United States are well known and quoted often; however some of them seem to be too sophisticated, for example, the three dimensional input problem. The reason for this is that the requirement of safety for a nuclear power plant is so severe that all events which have even a very low probability of occurrence should be considered. Therefore, if the results of theoretical study indicate an event which may occur even in very low probability, then from the viewpoint of conservatism, the designer must consider that event in this design. Although for the design of a nuclear power plant this might be partly true, the author feels that the probability of occurrence of the event should be evaluated in relation to the potential hazard of the design object. As well as this, he believes that proper understanding of the event in relation to the actual record of failures during past destructive earthquakes should be taken into consideration.

scholarly journals A modern data measurement system to study and test thermionic heat to electricity converters

2018 ◽  
Vol 4 (3) ◽  
pp. 179-183
Author(s):  
Andrey Kirillov ◽  
Valeriy Yarygin
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Test System ◽  
Temperature Fields ◽  
Test Facility ◽  
Measurement Systems ◽  
Data Measurement ◽  
Functional Capabilities ◽  
Thermionic Nuclear Power ◽  
A Current

Studies and tests are conducted to determine the performance of thermionic nuclear power plants (TNPP) a stage in which is pre-irradiation testing of laboratory thermionic converters (TIC) with flat and cylindrically shaped electrodes using test facilities fitted with automated data measurement systems (DMS). The TIC volt-ampere characteristics (VAC) are measured in the DMS jointly with the measured test section and experimental test facility temperature fields. The structure and the characteristics of a DMS based on products from ICP DAS Co., Ltd are presented. A developed VAC measurement program providing the operator with a convenient graphic interface and enabling adjustment of the measurement parameters has been considered. The VAC recording errors in the process of measurements have been determined using TIC simulators. The error in the VAC diffusion portion on a simulator (with a current of less than 3 A) is not more than 1%. Thanks to the use of modern components, the developed DMS offers extended functional capabilities for measuring the thermocouple signals in an experimental electrophysical test facility. The DMS structure provides for the convenience of scaling (through a larger number of measuring channels) and makes it possible to add modules from other manufacturers. The experience of operating this DMS will be used to develop the DMS for an in-pile test system designed for similar functions.

Seismic Fragility of Piping Nozzles in Nuclear Power Plants: A Case for Updating the Current State-of-Practice

2021 ◽  
Author(s):  
Abhinav Gupta ◽  
Ankit Dubey ◽  
Sunggook Cho
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Seismic Analysis ◽  
Nuclear Industry ◽  
Life Extension ◽  
Seismic Loads ◽  
Thermal Loads ◽  
Loss Of Coolant Accident ◽  
Piping Systems ◽  
Recent Developments

Abstract Nuclear industry spends enormous time and resources on designing and managing piping nozzles in a plant. Nozzle locations are considered as a potential location for possible failure that can lead to loss of coolant accident. Industry spends enormous time in condition monitoring and margin management at nozzle locations. Margins against seismic loads play a significant role in the overall margin management. Available margins against thermal loads are highly dependent upon seismic margins. In recent years, significant international collaboration has been undertaken to study the seismic margin in piping systems and nozzles through experimental and analytical studies. It has been observed that piping nozzles are highly overdesigned and the margins against seismic loads are quite high. While this brings a perspective of sufficient safety, such excessively high margins compete with available margins against thermal loads particularly during the life extension and subsequent license renewal studies being conducted by many plants around the world. This paper focuses on identifying and illustrating two key reasons that lead to excessively conservative estimates of nozzle fragilities. First, it compares fragilities based on conventional seismic analysis that ignores piping-equipment-structure interaction on nozzle fragility with the corresponding assessment by considering such interactions. Then, it presents a case that the uncertainties considered in various parameters for calculating nozzle fragility are excessively high. The paper identifies a need to study the various uncertainties in order to achieve a more realistic quantification based on recent developments in our understanding of the seismic behavior of piping systems.

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