scholarly journals Modelling of Passive Heat Removal Systems: A Review with Reference to the Framatome KERENA BWR Reactor: Part I

Energies ◽  
2019 ◽  
Vol 13 (1) ◽  
pp. 35 ◽  
Author(s):  
Amirhosein Moonesi Shabestary ◽  
Frances Viereckl ◽  
Yu Zhang ◽  
Rene Manthey ◽  
Dirk Lucas ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Gravitational Force ◽  
Heat Removal ◽  
Dimensional System ◽  
Thermal Gradients ◽  
Two Phase ◽  
Passive Safety Systems ◽  
External Power ◽  
Safety Systems

Passive safety systems are an important feature of currently designed and constructed nuclear power plants. They operate independent of external power supply and manual interventions and are solely driven by thermal gradients and gravitational force. This brings up new needs for performance and reliably assessment. This paper provides a review on fundamental approaches to model and analyze the performance of passive heat removal systems exemplified for the passive heat removal chain of the KERENA boiling water reactor concept developed by Framatome. We discuss modelling concepts for one-dimensional system codes such as ATHLET, RELAP and TRACE and furthermore for computational fluid dynamics codes. Part I deals with numerical and experimental methods for modelling of condensation inside the emergency condensers and on the containment cooling condenser while part II deals with boiling and two-phase flow instabilities.

scholarly journals Modelling of Passive Heat Removal Systems: A Review with Reference to the Framatome BWR Reactor KERENA: Part II

Energies ◽  
2019 ◽  
Vol 13 (1) ◽  
pp. 109 ◽  
Author(s):  
René Manthey ◽  
Frances Viereckl ◽  
Amirhosein Moonesi Shabestary ◽  
Yu Zhang ◽  
Wei Ding ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Gravitational Force ◽  
Heat Removal ◽  
Dimensional System ◽  
Thermal Gradients ◽  
Two Phase ◽  
Passive Safety Systems ◽  
External Power ◽  
Safety Systems

Passive safety systems are an important feature of currently designed and constructed nuclear power plants. They operate independent of external power supply and manual interventions and are solely driven by thermal gradients and gravitational force. This brings up new needs for performance and reliably assessment. This paper provides a review on fundamental approaches to model and analyze the performance of passive heat removal systems exemplified for the passive heat removal chain of the KERENA boiling water reactor concept developed by Framatome. We discuss modeling concepts for one-dimensional system codes such as ATHLET, RELAP and TRACE and furthermore for computational fluid dynamics codes. Part I dealt with numerical and experimental methods for modeling of condensation inside the emergency condenser and on the containment cooling condenser. This second part deals with boiling and two-phase flow instabilities.

Super Critical Carbon Dioxide Brayton Cycle Based Heat Removal System in Nuclear Power Plants

2018 ◽  
Author(s):  
Wei Shuhong ◽  
Zheng Hua
Keyword(s):  
Carbon Dioxide ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Heat Removal ◽  
Brayton Cycle ◽  
Spent Fuel ◽  
Passive Safety Systems ◽  
Safety Systems ◽  
Critical Carbon Dioxide

Heat removal from the core and spent fuel is one of the fundamental safety functions. Mobile equipment for heat removal from the core and spent fuel is required after Fukushima accident, but there are various constraints for modification of current operating nuclear power plants, such as layout, especially when new equipment are needed inside the containment. New reactor designs emphasize passive safety systems, but most passive safety systems rely on large pool and the heat removal duration depends on water volume. Super critical carbon dioxide brayton cycle can work as a heat engine by itself without external power supply or water supply, and supply surplus electricity due to the difference between expansion work and compression work. Also, super critical carbon dioxide brayton cycle is small, can be designed as a modular, mobile system and has little effect to system configuration or layout of current operating nuclear power plants. Super critical carbon dioxide brayton cycle is a good choice for self-propelling or passive heat removal for nuclear power plant modifications or new reactor designs without difficult modification of system configuration or layout. Super critical carbon dioxide Brayton cycle based heat removal system in nuclear power plants is designed and its technical feasibility is analyzed.

Passive Safety Systems of Advanced Nuclear Power Plant: AP1000

2010 ◽  
Author(s):  
Guohua Yan ◽  
Chen Ye
Keyword(s):  
Nuclear Power ◽  
Human Error ◽  
Cooling System ◽  
Heat Removal ◽  
Injection System ◽  
Passive Safety ◽  
Human Errors ◽  
Power Sources ◽  
Passive Safety Systems ◽  
Safety Systems

In the entire history of commercial nuclear power so far, only two major accidents leading to damage of reactor core have taken place. One is Three Mile Island (TMT) accident (1979), which is caused by a series of human error, and the other is Chernobyl accident (1986), which is due to the combined reason of design defects and human errors. After TMI and Chernobyl accidents, in order to reduce manpower in operation and maintenance and influence of human errors on reactor safety, consideration is given to utilization of passive safety systems. According to the IAEA definition, passive safety systems are based on natural forces, such as convection and gravity, and stored energy, making safety functions less dependent on active systems and operators’ action. Recently, the technology of passive safety has been adopted in many reactor designs, such as AP1000, developed by Westinghouse and EP1000 developed by European vendor, and so on. AP1000 as the first so-called Generation III+ has received the final design approval from US NRC in September 2004, and now being under construction in Sanmen, China. In this paper, the major passive safety systems of AP1000, including passive safety injection system, automatic depressurization system passive residual heat removal system and passive containment cooling system, are described and their responses to a break loss-of-coolant accident (LOCA) are given. Just due to these passive systems’ adoption, the nuclear plant can be able to require no operator action and offsite or onsite AC power sources for at least 72h when one accident occurs, and the core melt and large release frequencies are significantly below the requirement of operating plants and the NRC safety goals.

scholarly journals Onshore Nuclear Power Plant Concept with Enhanced Passive Safety System

2019 ◽  
Vol 4 (6) ◽  
pp. 155-159
Author(s):  
A.H.M. Iftekharul Ferdous ◽  
T. H. M Sumon Rashid ◽  
Md Asaduzzaman Shobug ◽  
Tanveer Ahmed ◽  
Nitol Kumar Dutta
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Water Source ◽  
Passive Safety ◽  
Electric Power Supply ◽  
Passive Safety Systems ◽  
Passive Safety System ◽  
Safety Systems

Bangladesh is a developing country and it’s increasing economy can be maintained by providing sufficient amount of electric power supply. Therefore government is initiating Rooppur nuclear power project is one of them which is needed to be sited beside a vast amount of water source, lowest populated area and away from the locality to reduce the damage caused by any nuclear accidents. In this thesis paper we have shown that, the the dangers of residing errors of Rooppur nuclear power plant and give a proposal to go for onshore nuclear power plant in Bangladesh with two proposed designs of passive safety systems PSS-I & PSS-II. These systems will give safety to the power plants in the case of plant blackout during accidents.

AP1000™ Breaking the Dependence on Site Characteristics: Electrical Grid and Cooling Water

2009 ◽  
Author(s):  
Richard G. Anderson ◽  
Terry L. Schulz ◽  
Daniel T. McLaughlin
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Cooling Water ◽  
Passive Safety ◽  
Small Contribution ◽  
Passive Safety Systems ◽  
Design Basis ◽  
Site Characteristics ◽  
Diesel Generators ◽  
Safety Systems

The AP1000™ has been developed to use passive safety systems to address design basis and beyond design basis accidents to minimize impact on calculated plant Core Damage Frequency (CDF) and Large Release Frequency (LRF). Passive safety cases are supplemented with accident mitigation strategy using highly reliable non-safety grade systems. AP1000™ non-safety systems that resemble the safety systems of conventional plants are designed to mitigate accidents, when available. In addition, a number of mitigation schemes make use of non-safety systems together with select passive system features. This design approach has resulted in a small dependence on site characteristics and has minimized their contribution to CDF and LRF. Conventional nuclear power plants are designed to use motor driven safety grade equipment to address design basis accidents. Safety grade diesel generators are used to provide power to safety grade equipment in the event of a loss of offsite power. The contribution to CDF and LRF from Loss of Offsite Power (LOSP) or loss of cooling water events is significantly lower for the AP1000™ than that for conventional plants. The AP1000™ uses passive safety systems in conjunction with non-safety systems to reduce plant CDF and LRF sensitivity to site specific characteristics. Non-safety systems used for accident mitigation make use of high grade commercial components and are provided power from non-safety grade diesel generators upon LOSP. These components are well maintained to increase system reliability and to increase availability for accident mitigation. The effect of the non-safety systems is further enhanced by partial use of features of the passive safety systems for accident mitigation. In this way, events with initiating event frequencies often driven by site characteristics are mitigated with small contribution to CDF or LRF, often without the need to activate any, or some of the passive plant features.

scholarly journals Comparison Analysis of Selected Nuclear Power Plants Supplied With Helium from High-Temperature Gas-Cooled Reactor

2018 ◽  
Vol 25 (s1) ◽  
pp. 204-210 ◽  
Author(s):  
Natalia Szewczuk-Krypa ◽  
Marta Drosińska-Komor ◽  
Jerzy Głuch ◽  
Łukasz Breńkacz
Keyword(s):  
High Temperature ◽  
Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Social Aspect ◽  
Passive Safety Systems ◽  
System A ◽  
Safety Systems ◽  
High Temperature Gas

Abstract The article presents results of efficiency calculations for two 560 MW nuclear cycles with high-temperature gas-cooled reactor (HTGR). An assumption was made that systems of this type can be used in so-called marine nuclear power plants. The first analysed system is the nuclear steam power plant. For the steam cycle, the efficiency calculations were performed with the code DIAGAR, which is dedicated for analysing this type of systems. The other system is the power plant with gas turbine, in which the combustion chamber has been replaced with the HTGR. For this system, a number of calculations were also performed to assess its efficiency. Moreover, the article names factors in favour of floating nuclear power plants with HTGRs, which, due to passive safety systems, are exposed to much smaller risk of breakdown than other types of reactors which were in common use in the past. Along with safety aspects, it is also economic and social aspect which make the use of this type of systems advisable.

Seismic Re-Evaluation of the Tarapur Atomic Power Plants 1 and 2

2004 ◽  
Author(s):  
S. M. Ingole ◽  
B. Santosh Kumar ◽  
Sushil Gupta ◽  
U. P. Singh ◽  
K. Giridhar ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Support Systems ◽  
Free Field ◽  
Heat Removal ◽  
Boiling Water Reactors ◽  
Seismic Evaluation ◽  
Heavy Water Reactor ◽  
Water Reactors ◽  
Safety Systems

Two Boiling Water Reactors (BWR) of 210 MWe each at Tarapur Atomic Power Station, Units-1&2 (TAPS-1&2) were commissioned in the year 1969. The safety related civil structures at TAPS had been designed for a seismic coefficient of 0.2g and other structures for 0.1g. The work of seismic re-evaluation of the TAPS-1&2 has been taken up in the year 2002. As two new Pressurized Heavy Water Reactor (PHWR) plants of 540 MWe each, Tarapur Atomic Power Project Units-3&4 (TAPP-3&4), are coming up in the vicinity of TAPS-1&2, detailed geological and seismological studies of the area around TAPS-1&2 are available. The same free-field ground motion as generated for TAPP-3&4 has been used for TAPS-1&2. The seismic re-evaluation of the plant has been performed as per the procedure given in IAEA, Safety Reports Series entitled “Seismic Evaluation of Existing Nuclear Power Plants”, and meeting the various codes & standards, viz., ASME, ASCE, IEEE standards etc. The Safety Systems (SS) and Safety Support Systems (SSS) are qualified by adopting detailed analysis and testing methods. The equipment in the SS and SSS have been qualified by conducting a walkdown as per the procedure given in Generic Implementation Procedure, Dept. of Energy (GIP–DOE), USA. The safety systems include the systems required for safe shutdown of the plant, one chain of decay heat removal and containment of activity. The safety support systems viz., Electrical, Instrumentation & Control and systems other than SS & SSS have been qualified by limited analysis, testing and mostly by following the procedure of walkdown. The paper brings out the details of the work accomplished during seismic re-evaluation of the two units of BWR at Tarapur.

scholarly journals A study into the modes of the VVER-1000 RCP starting in an earlier inoperative loop

2019 ◽  
Vol 5 (4) ◽  
pp. 305-311
Author(s):  
Iliya Ye. Bragin ◽  
Vladimir I. Belozerov
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Primary Circuit ◽  
Test Procedures ◽  
Reactor Plant ◽  
Hydraulic Transients ◽  
Two Phase ◽  
Safety Systems ◽  
Advanced Computer ◽  
Plant Power

To simulate the mode of the RCP starting in an earlier inoperative loop, KORSAR/GP, a code supporting coupled numerical modeling of neutronic and thermal-hydraulic transients in a VVER reactor plant in operating and emergency conditions, was chosen as the computational tool. Studying these modes using thermal-hydraulic codes makes it possible to analyze the course of transients and certain emergency processes without using commercial test procedures, which contributes to laying the groundwork for addressing the issues involved in ensuring the reliability, operating safety and efficiency of nuclear power plants. Increased requirements to the safety of NPPs identify the need for avoiding excessive conservatism in the analysis based on which requirements to safety systems are formulated, as well as for enhancing the knowledge of the regularities of thermal-hydraulic transients based on advanced computer programs (or codes) designed for improved computational analysis of non-stationary thermal hydraulics in the water-cooled reactor circulation circuits in emergency and transient modes relying on inhomogeneous non-equilibrium mathematical models of two-phase flows and on a detailed description of the physical transient regularities. The purpose of the study is to analyze computationally the starting of a VVER-1000 RCP in an earlier inoperative loop at different reactor plant power values. To do this, one requires to develop the VVER-1000 reactor primary circuit computational pattern to model the transient taking place as one RCP is started, to conduct a further analysis, and to compare the key monitored reactor coolant and core parameters (power, temperature, flow rate, etc.).

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