Experiments and Analytical Simulation Work on an Innovative Steam-Injector-Driven Passive Core Injection Cooling System

2007 ◽  
Author(s):  
Michitsugu Mori ◽  
Tadashi Narabayashi ◽  
Shuichi Ohmori ◽  
Fumitoshi Watanabe
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
High Pressures ◽  
Cooling System ◽  
Steam Pressure ◽  
Injection System ◽  
The Core ◽  
Coolant Injection ◽  
Core Cooling

A Steam Injector (SI) is a simple, compact, passive pump which also functions as a high-performance direct-contact compact heater. We are developing this innovative concept by applying the SI system to core injection systems in Emergency Core Cooling Systems (ECCS) to further improve the safety of nuclear power plants. Passive ECCS in nuclear power plants would be inherently very safe and would prevent serious accidents by keeping the core covered with water (Severe Accident-Free Concept). The Passive Core Coolant Injection System driven by a high-efficiency SI is one that, in an accident such as a loss of coolant accident (LOCA), attains a higher discharge pressure than the supply steam pressure used to inject water into the reactor by operating the SI using water stored in the pool as the water supply source and steam contained in the reactor as the source of pressurization energy. The passive SI equipment would replace large, rotating machines such as pumps and motors, so eliminating the possibility of such equipment failing. In this Si-driven Passive Core Coolant Injection System (SI-PCIS), redundancy will be provided to ensure that the water and steam supply valves to the SI open reliably, and when the valves open, the SI will automatically start to inject water into the core to keep the core covered with water. The SI used in SI-PCIS works for a range of steam pressure conditions, from atmosphere pressure through to high pressures, as confirmed by analytical simulations which were done based on comprehensive experimental data obtained using reduced scale SI. We did further simulations and evaluations of the core cooling and coolant injection performance of SI-PCIS in BWR using RETRAN-3D code, developed using EPRI and other utilities, for the case of small LOCA. Reactors equipped with passive safety systems — the gravity-driven core cooling/injection system (GDCS) and depressurization valves (DPV) — would inevitably end up having large LOCA, even if they are initially small LOCA, as depressurization valves are forcibly opened in order to inject coolant from the GDCS pool to the GDCS water head at up to ∼0.2MPa. On the other hand, our simulation demonstrated that SI-PCIS could prevent large LOCA occurring in reactors by having by coolant discharged into the core in the event of small LOCA or when DPV unexpectedly open.

Experiments and Analytical Simulation on Steam Injector Driven Passive Core Injection System for Innovative-Simplified Nuclear Power Plant

2008 ◽  
Author(s):  
Shuichi Ohmori ◽  
Tadashi Narabayashi ◽  
Michitsugu Mori ◽  
Fumitoshi Watanabe
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
High Pressures ◽  
High Efficiency ◽  
Cooling System ◽  
Steam Pressure ◽  
Injection System ◽  
The Core ◽  
Core Cooling

A steam injector (SI) is a simple, compact and passive pump and also acts as a high-performance direct-contact compact heater. We are developing an innovative idea by applying SI system for core injection system in emergency core cooling systems (ECCS) to further improve the safety of nuclear power plants. The passive core injection system (PCIS) driven by high-efficiency SI is a system that, in an accident such as a LOCA (loss of coolant accident), attains discharge pressure higher than the supply steam pressure to inject water into the reactor by operating the SI, by supplying water from a pool in a containment vessel and the steam from a reactor pressure vessel (RPV). The SI, passive equipment, is used to replace large rotating machines such as pumps and motors, eliminating the failure probabilities of such active equipment. When the water and steam supply valves open, the SI-driven PCIS (SI-PCIS) will automatically start to inject water into the core to keep the core covered with water. The SI-PCIS works for the range of steam pressure conditions from atmosphere pressure through high pressures, in which the analytical simulations of SI were carried out based on the plenty amount of experimental data using reduced scale SI. We further simulated and evaluated the core cooling and water injection performance of SI-PCIS in BWR using RETRAN-3D code for the case of small LOCA. A reactor, such as ESBWR, equipped with the passive safety system by gravity-driven cooling system (GDCS) and the depressurization valves (DPVs) should be inevitable to lead to large LOCA even for the case of small LOCA by forcibly opening the DPVs to inject water from the GDCS pool due to that the GDCS water head is up to ∼0.2MPa. On the contrary, our simulation exhibited that SI-PCIS could save the reactors from leading to large LOCA by discharge of the water into a core for the cases of small LOCA or DPV unexpectedly open. In addition, we conducted the analytical simulations of SI, which grew in size for the actual nuclear power plant. A part of this report are fruits of research which is carried out by Tokyo Electric Power Company (TEPCO), Toshiba corporation, and seven universities in Japan, funded from the Ministry of Economy, Trade and Industry (METI) of Japan as the national public research-funded program.

Emergency Core Cooling System Sump Strainer Fibre Bypass Testing and Measurement

2016 ◽  
Author(s):  
Qingwu Cheng ◽  
Harry Adams ◽  
Metin Yetisir
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Cooling System ◽  
Reactor Core ◽  
Regulatory Requirements ◽  
Pressurized Water ◽  
Load Pressure ◽  
Analysis Techniques ◽  
Core Cooling

The potential of losing post-Loss Of Coolant Accident (LOCA) recirculation capability due to debris blockage of Emergency Core Cooling (ECC) strainers resulted in early replacements of ECC strainers in most nuclear power plants. To validate the performance of ECC strainers, extensive testing representing plant conditions is required. Such testing programs include thin-bed and full debris load pressure drop tests, fibre bypass tests and chemical effects tests. Multiple testing loops and state-of-the-art analysis techniques have provided in-depth understanding of sump strainer performance and the effect of chemical precipitation on debris bed head loss. ECC strainers typically use perforated plates as filtering surfaces with 1.6 to 2.5 mm holes and 35 to 40% open area, allowing some particulates and fibres to pass through the strainer filtering surfaces. Recently, the bypassed fibrous debris has been identified as a potential safety concern due to its possible deposition in the reactor core and blocking of flow into fuel assemblies. In some cases, the amount of fibre that is specified as allowed to enter a reactor core is only 15 g per fuel assembly for pressurized water reactors. Characterization and quantification of bypassed fibre debris for nuclear power plants are needed to address regulatory requirements. Testing methodology and analysis techniques to address regulatory requirements and concerns are presented in this paper. In particular, a newly developed technique is presented to address debris bypass quantification.

scholarly journals Ensembles of climate change models for risk assessment of nuclear power plants

2017 ◽  
Vol 232 (2) ◽  
pp. 185-200
Author(s):  
Matteo Vagnoli ◽  
Francesco Di Maio ◽  
Enrico Zio
Keyword(s):  
Climate Change ◽  
Risk Assessment ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Climate Models ◽  
Cooling System ◽  
Future Projections ◽  
Change Models ◽  
Temperature And Pressure

Climate change affects technical systems, structures and infrastructures, changing the environmental context for which systems, structures and infrastructure were originally designed. In order to prevent any risk growth beyond acceptable levels, the climate change effects must be accounted for into risk assessment models. Climate models can provide future climate data, such as air temperature and pressure. However, the reliability of climate models is a major concern due to the uncertainty in the temperature and pressure future projections. In this work, we consider five climate change models (individually unable to accurately provide historical recorded temperatures and, thus, also future projections) and ensemble their projections for integration in a probabilistic safety assessment, conditional on climate projections. As case study, we consider the passive containment cooling system of two AP1000 nuclear power plants. Results provided by the different ensembles are compared. Finally, a risk-based classification approach is performed to identify critical future temperatures, which may lead to passive containment cooling system risks beyond acceptable levels.

scholarly journals HEALTH IN THE SYSTEM OF VALUES OF THE POPULATION OF THE SURVEILLANCE ZONES NUCLEAR POWER PLANTSS

2021 ◽  
Vol 26 ◽  
pp. 219-234
Author(s):  
V. Prylypko ◽  
◽  
Yu. Ozerova ◽  
I. Bondarenko ◽  
M. Morozova ◽  
...  
Keyword(s):  
Family Life ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Sampling Error ◽  
Close Correlation ◽  
The State ◽  
The Core ◽  
Terminal Values ◽  
Happy Family

Objective: to determine the place of health in the system of values of the population of the surveillance zone (SZ) of nuclear power plants (NPPs) and its importance in the perception of emergency risks (ER). Materials and methods. To determine the place of health in the value system, a survey of the able-bodied population of satellite cities of Rivne (RNPP) and South Ukrainian (SUNPP) nuclear power plants was conducted using nonrepetitive sampling, where the sampling error does not exceed 7,0 %. The motivational and behavioral component that determined health in the individual hierarchy of values of the subject according to the questionnaire Berezovskaya R. A. was studied. Statistical and mathematical methods were used in the research process. Results. The array of respondents was conditionally divided into 4 groups according to their attitude to human health. And the group where a person’s life position is focused exclusively on health is the most common – 77,0 %. Group IV, which wants to live without limiting itself, is 8,1 %. The component integrity of values-goals and valuesmeans among the urban population of the SZ of both nuclear power plants is the same: the main goal in life is health, happy family life, and as a means – perseverance, diligence and health. Goal values in groups I and IV have some differences: in the first group of respondents the main goal in life is health, and in the fourth, where a person’s life guidelines exclude any restrictions – a happy family life. Values for these populations have some differences, but in both groups health appears to be the main means to an end. There is a close correlation between the core of terminal values and the average indicators of the state of concern about the risk of emergencies. Conclusions. Identified hierarchy of values: a group of stable dominant values; average status values; group of least significant values. The values of the highest status among the values-goals are – health, happy family life and interesting work. Most respondents plan to achieve them through values such as «health», «perseverance and hard work». There is a close correlation between the core of terminal values and the average indicators of the state of concern about the risk of emergencies. Key words: health, values, population, NPP surveillance zone, perception of emergency risks.

Environmental Considerations for Preparing Permit Applications of New Nuclear Power Plants Located in Greenfield Sites

2010 ◽  
Author(s):  
Ping K. Wan ◽  
Desmond W. Chan ◽  
Alice C. Carson
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Fossil Fuels ◽  
Cooling System ◽  
The United States ◽  
Attractive Alternative ◽  
Socioeconomic Impacts ◽  
Power Generators ◽  
Available Technology

Nuclear power generation has become an increasingly attractive alternative in the United States (U.S.) power market due to several factors: growing demand for electric power, increasing global competition for fossil fuels, concern over greenhouse gas emissions and their potential impact on climate change, and the desire for energy independence. Assuring the protection of people and the environment are of paramount concern to nuclear power generators and regulators as we move towards a possible nuclear renaissance. Thus, sound engineering design is of utmost important and potential environmental and safety concerns must be carefully evaluated and disposition during permitting of the new nuclear power plants. Areas to be considered in order to alleviate these concerns include the following: • Site meteorology and dispersion conditions of the area; • Evaluation of radiological consequence during normal plant operation and emergency conditions; • Water availability for plant cooling system; • Evaluation of potential land use, water use, ecological and socioeconomic impacts of the proposed action. This paper focuses on site suitability evaluation for greenfield sites through site characterization, examination of challenges/constraints in deployment of available technology/plant systems, and mapping of permitting compliance strategy. Case studies related to selection of plant systems based on the environmental site conditions, preferred compliance plan, and public acceptance, are included.

Investigation of the Fission Product Release From Molten Pools Under Oxidizing Conditions With the Code RELOS

2006 ◽  
Author(s):  
Ingo D. Kleinhietpaß ◽  
Hermann Unger ◽  
Hermann-Josef Wagner ◽  
Marco K. Koch
Keyword(s):  
Phase Boundary ◽  
Molten Pool ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Fission Products ◽  
Energy Systems ◽  
The Core ◽  
Gas Atmosphere ◽  
Upper Pool

With the purpose of modeling and calculating the core behavior during severe accidents in nuclear power plants system codes are under development worldwide. Modeling of radionuclide release and transport in the case of beyond design basis accidents is an integrated feature of the deterministic safety analysis of nuclear power plants. Following a hypothetical, uncontrolled temperature escalation in the core of light water reactors, significant parts of the core structures may degrade and melt down under formation of molten pools, leading to an accumulation of large amounts of radioactive materials. The possible release of radionuclides from the molten pool provides a potential contribution to the aerosol source term in the late phase of core degradation accidents. The relevance of the amount of transferred oxygen from the gas atmosphere into the molten pool on the specification of a radionuclide and its release depends strongly on the initial oxygen inventory. Particularly for a low oxygen potential in the melt as it is the case for stratification when a metallic phase forms the upper layer and, respectively, when the oxidation has proceeded so far so that zirconium was completely oxidized, a significant influence of atmospheric oxygen on the specification and the release of some radionuclides has to be anticipated. The code RELOS (Release of Low Volatile Fission Products from Molten Surfaces) is under development at the Department of Energy Systems and Energy Economics (formerly Department of Nuclear and New Energy Systems) of the Ruhr-University Bochum. It is based on a mechanistic model to describe the diffusive and convective transport of fission products from the surface of a molten pool into a cooler gas atmosphere. This paper presents the code RELOS, i. e. the features and abilities of the latest code version V2.3 and the new model improvements of V2.4 and the calculated results evaluating the implemented models which deal with the oxygen transfer from the liquid side of the phase boundary to the bulk of the melt by diffusion or by taking into account natural convection. Both models help to estimate the amount of oxygen entering into the liquid upper pool volume and being available for the oxidation reaction. For both models the metallic, the oxidic and a mixture phase can be taken into account when defining the composition of the upper pool volume. The influence of crust formation, i. e. the decrease of the liquid pool surface area is taken care of because it yields the relevant amount of fission products released into the atmosphere. The difference of the partial density between the gas side of the phase boundary and the bulk of the gas phase is the driving force of mass transport.

A Study on Thermal Stress Evaluation of Safety Injection System for NPPs Considering Leakage Effect

2011 ◽  
Author(s):  
Hee-Dong Sung ◽  
Sun-Hye Kim ◽  
Ik-Joong Kim ◽  
Young-Jin Kim ◽  
Jeong-Soon Park ◽  
...  
Keyword(s):  
Thermal Stress ◽  
Thermal Stratification ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Injection System ◽  
Stress Evaluation ◽  
Branch Line ◽  
Leakage Model ◽  
Cfd Analyses

Several piping failures caused by thermal stratification have been reported in some nuclear power plants since the early 1980s. However, this kind of thermal effect was not considered when the old vintage nuclear power plants were designed. Thermal stratification is usually generated by turbulent penetration from the RCS to branch line or leakage through damaged part of valve in branch line. In this paper, using the CFD analysis, characteristics of thermal stratification in a safety injection system of PWR plant were investigated and thermal stress evaluation was also conducted. First, CFD analyses were carried out on in-leakage model and out-leakage model according to operating condition. The case of out-leakage, the thermal stratification based on temperature distribution was generated a little at the rear of 1st valve. In contrast, significant thermal stratification was generated in front of 1st valve in in-leakage model because the effect of rapid flow velocity from RCS.

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