INVESTIGATION OF THE DEPOSITION OF INSOLUBLE AEROSOLS ON THE HEAT EXCHANGE SURFACES OF THE SYSTEM OF PASSIVE HEAT REMOVAL FROM THE CONTAINMENT DURING ACCIDENTS AT NUCLEAR POWER PLANTS

2019 ◽  
Vol 83 ◽  
pp. 24-33 ◽  
Author(s):  
A. A. Fiskov ◽  
◽  
V. V. Bezlepkin ◽  
S. E. Semashko. ◽  
V. M. Pogrebenkov ◽  
...  
Keyword(s):  
Heat Exchange ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Heat Removal

scholarly journals Seismic safety evaluation during site selection for the nuclear power plants in Bangladesh

2018 ◽  
Vol 4 (4) ◽  
pp. 251-256 ◽  
Author(s):  
Sergey Shcheklein ◽  
Ismail Hossain ◽  
Mohammad Akbar ◽  
Vladimir Velkin
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Spent Nuclear Fuel ◽  
Heat Removal ◽  
Good Time ◽  
Diagnostics System ◽  
Seismic Impacts

Bangladesh lies in a tectonically active zone. Earlier geological studies show that Bangladesh and its adjoining areas are exposed to a threat of severe earthquakes. Earthquakes may have disastrous consequences for a densely populated country. This dictates the need for a detailed analysis of the situation prior to the construction of nuclear power plant as required by the IAEA standards. This study reveals the correlation between seismic acceleration and potential damage. Procedures are presented for investigating the seismic hazard within the future NPP construction area. It has been shown that the obtained values of the earthquake’s peak ground acceleration are at the level below the design basis earthquake (DBE) level and will not lead to nuclear power plant malfunctions. For the most severe among the recorded and closely located earthquake centers (Madhupur) the intensity of seismic impacts on the nuclear power plant site does not exceed eight points on the MSK-64 scale. The existing predictions as to the possibility of a super-earthquake with magnitude in excess of nine points on the Richter scale to take place on the territory of the country indicate the necessity to develop an additional efficient seismic diagnostics system and to switch nuclear power plants in good time to passive heat removal mode as stipulated by the WWER 3+ design. A conclusion is made that accounting for the predicted seismic impacts in excess of the historically recorded levels should be achieved by the establishment of an additional efficient seismic diagnostics system and by timely switching the nuclear power plants to passive heat removal mode with reliable isolation of the reactor core and spent nuclear fuel pools.

In-Calandria Retention of Corium in PHWR: Experimental Investigation and Remaining Issues

2017 ◽  
Vol 3 (2) ◽  
Author(s):  
Sumit V. Prasad ◽  
A. K. Nayak
Keyword(s):  
Experimental Investigation ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Heat Removal ◽  
Public Domain ◽  
Severe Accident ◽  
Fukushima Accident ◽  
Decay Heat ◽  
Critical Issues

After the Fukushima accident, the public has expressed concern regarding the safety of nuclear power plants. This accident has strengthened the necessity for further improvement of safety in the design of existing and future nuclear power plants. Pressurized heavy water reactors (PHWRs) have a high level of defense-in-depth (DiD) philosophy to achieve the safety goal. It is necessary for designers to demonstrate the capability of decay heat removal and integrity of containment in a PHWR reactor for prolonged station blackout to avoid any release of radioactivity in public domain. As the design of PHWRs is distinct, its calandria vessel (CV) and vault cooling water offer passive heat sinks for such accident scenarios and submerged calandria vessel offers inherent in-calandria retention (ICR) features. Study shows that, in case of severe accident in PHWR, ICR is the only option to contain the corium inside the calandria vessel by cooling it from outside using the calandria vault water to avoid the release of radioactivity to public domain. There are critical issues on ICR of corium that have to be resolved for successful demonstration of ICR strategy and regulatory acceptance. This paper tries to investigate some of the critical issues of ICR of corium. The present study focuses on experimental investigation of the coolability of molten corium with and without simulated decay heat and thermal behavior of calandria vessel performed in scaled facilities of an Indian PHWR.

Nonmetallic Inclusions in 08Kh18N10T Steel as the Cause of Initiation of Defects in Heat-Exchange Tubes of Steam Generators of Nuclear Power Plants

2019 ◽  
Vol 55 (2) ◽  
pp. 152-159
Author(s):  
V. М. Voevodyn ◽  
А. S. Mytrofanov ◽  
S. V. Hozhenko ◽  
R. L. Vasylenko ◽  
E. О. Krainyuk ◽  
...  
Keyword(s):  
Heat Exchange ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nonmetallic Inclusions ◽  
Steam Generators ◽  
08Kh18n10t Steel

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.

scholarly journals Urgent Problems in Thermal Physics of Design-Basis and Severe Accidents at Nuclear Power Plants

2016 ◽  
pp. 32-36
Author(s):  
I. Sharaievskii ◽  
N. Fialko ◽  
A. Nosovsky ◽  
L. Zimin ◽  
G. Sharaievskii
Keyword(s):  
Nuclear Fuel ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Heat Removal ◽  
Thermal Physics ◽  
Severe Accidents ◽  
Design Basis

The paper considers thermal and physical aspects in dynamics of design-basis and severe accidents at water-cooled nuclear power reactors with nuclear fuel damage. The most promising concepts of the corium confinement in the damaged reactor were analyzed. The main objective was to define the main areas to search and implement methods of efficient and safe heat removal in progression of emergencies.

scholarly journals Heat transfer intensification in emergency cooling heat exchanger and dry cooling towers on nuclear power plant using air-water mist flow

2019 ◽  
Vol 5 (4) ◽  
pp. 281-287
Author(s):  
Akram H. Abed ◽  
Sergey E. Shcheklein ◽  
Valery M. Pakhaluev
Keyword(s):  
Heat Transfer ◽  
Heat Exchanger ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Heat Removal ◽  
Ambient Air ◽  
Air Cooling ◽  
Heat Transfer Intensification ◽  
Dry Cooling

Advanced nuclear power plants are equipped with passive emergency heat removal systems (PEHRS) for removing the decay heat from reactor equipment in accidents accompanied by primary circuit leakage to the final heat absorber (ambient air). Herein, the intensity of heat dissipation to air from the outer surface of the heat exchanger achieved by buoyancy induced natural convection is extremely low, which need to a large heat exchanger surface area, apply different types of heat transfer intensification including (grooves, ribs and extended surfaces, positioning at higher altitudes, etc.). The intensity of heat removal is also strongly dependent on the ambient air temperature (disposable temperature head). Construction of nuclear power plants in countries with high ambient temperatures (Iran, Bangladesh, Egypt, Saudi Arabia, and others) which are characterized by a high level of ambient temperature imposes additional requirements on the increase of the heat exchange surfaces. The experimental investigation results of heat transfer intensification by a low energy ultrasonic which supply a fine liquid droplet (size ~3 µm) in the cooling air are presented in the present paper. In such case, the heat transfer between the surface and cooling flow involves the following three physical effects: convection, conductive heat transfer, and evaporation of water droplets. The last two effects weakly depend on the ambient air temperature and provide an active heat removal in any situation. The investigation was performed using a high-precision calorimeter with a controlled rate of heat supply (between 7800 and 12831 W/m2) imitating heated surface within the range of Reynolds numbers from 2500 to 55000 and liquid (water) flow rates from 23.39 to 111.68 kg·m-2·h-1. The studies demonstrated that the presence of finely dispersed water results in a significant increase in heat transfer compared with the case of using purely air-cooling. With a fixed heat flux, the energy efficiency increases with increasing water concentration, reaching the values over 600 W·m-2·C-1 at 111.68 kg·m-2·h-1, which is 2.8 times higher than for air cooling. With further development of research in order to clarify the optimal areas of intensification, it is possible to use this technology to intensify heat transfer to the air in dry cooling towers of nuclear power plants and thermal power plants used in hot and extreme continental climates.

Recent Reactor Head Drop Analysis for Complying NUREG-0612 Appendix A Guidelines

2008 ◽  
Author(s):  
Phuong H. Hoang ◽  
Mohammad Amin ◽  
Chee W. Mak
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Heat Removal ◽  
Federal Regulations ◽  
Load Drop ◽  
Safe Handling ◽  
Decay Heat ◽  
Parameters Analysis ◽  
Heavy Loads

The USNRC staff provided guidelines in NUREG-0612 for the control of heavy loads to meet the requirements of the Code of Federal Regulations (10 CFR) Sections 50.59 and 50.71(e) relate to the safe handling of heavy loads and load drop analyses to assure fuel assembly integrity and to permit continued decay heat removal in nuclear power plants. The USNRC staff considered plants that had installed single-failure-proof cranes or had completed load drop analyses conforming with the guidelines of Appendix A to NUREG-0612 would remain in conformance with the safe handling guidelines for lifting reactor head during the plant refueling operation. Reactor head drop analyses have been performed recently for a number of US nuclear power plants and have been reviewed by the USNRC staff for complying with the guidelines of Appendix A to NUREG-0612. This paper provides an overview of the methodology, parameters, analysis results and acceptance criteria used in these recent works.

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