Creation of Automatic System for Monitoring, Forecasting and Control of Condition of Lead-Bismuth (Lead) Coolant and Surfaces of Circuits of Nuclear Power Plants

2009 ◽  
Author(s):  
P. N. Martynov ◽  
R. Sh. Askhadullin ◽  
A. A. Simakov ◽  
A. Yu. Chaban’ ◽  
M. E. Chernov ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Reactor Core ◽  
Heat Removal ◽  
Severe Accident ◽  
Medium Size ◽  
Safety Level ◽  
Nitride Fuel ◽  
Active Safety Systems ◽  
Accident Conditions

Lead-bismuth coolant is preferable for the medium size reactors, since, in contrast to the sodium coolant, it does not interact with water and air, it is radiation resistant, insignificantly activated and it is not combustible [1]. Combination of natural properties of lead-based coolants, mono-nitride fuel, fast reactor neutronics and design approaches used for the reactor core and heat removal system brings SVBR 75/100 NPP [2] to achieve a new safety level and assures its stability without operation of active safety systems even under severe accident conditions. Analysis of possible sequences of the events even under conditions of such severe accidents as addition of total excess reactivity or all pumps trip accompanied by safety system failure leads to the conclusion on that power unit with SVBR 75/100 reactor plant (RP) has high safety level.

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.

scholarly journals Analysis of Loss of Offsite Power Events at China’s Nuclear Power Plants

2018 ◽  
Vol 10 (8) ◽  
pp. 2680 ◽  
Author(s):  
Feng Jiao ◽  
Shanshan Ding ◽  
Juan Li ◽  
Lixin Zheng ◽  
Qinghua Zhang ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Reactor Core ◽  
Electric Power System ◽  
Lessons Learned ◽  
Normal Operation ◽  
Emergency Power ◽  
Station Blackout ◽  
Accident Conditions

The function of the electric power system of nuclear power plants (NPPs) is to provide safe and reliable electricity for the equipment both in normal operation and accident conditions, and to provide emergency power for nuclear safety-related systems to maintain the safety of NPPs. Station blackout (SBO) occurs when loss of offsite power (LOOP) happens concurrently with unavailability of the onsite emergency alternating current (ac) power. LOOP is a precursor of SBO which rarely occurs but contributes significantly to reactor core damage frequency (CDF). Collecting and analyzing all LOOP events in NPPs of China from 1993 to 2017, this paper summarizes the common features of the LOOP events, and identifies the weaknesses and lessons learned from these events. Conclusions and experience feedback suggestions are put forward for improving the reliability of the offsite power supply of NPPs in China.

Experimental Investigation Into Buckling Failure of Slender Metal Plates in Severe Accident Conditions

2014 ◽  
Author(s):  
Byeongnam Jo ◽  
Wataru Sagawa ◽  
Koji Okamoto
Keyword(s):  
High Speed ◽  
Nuclear Power ◽  
Power Plants ◽  
Failure Load ◽  
Severe Accident ◽  
Steel Columns ◽  
Wide Range ◽  
Buckling Failure ◽  
The Difference ◽  
Accident Conditions

Buckling failure load of stainless steel columns under compressive stress was experimentally measured in severe accident conditions, which addresses the accidents in Fukushima Daiichi nuclear power plants. Firstly, buckling failure load defined as load which causes failure of the column (plastic collapse) was measured in a wide range of temperatures from 25 °C up to 1200 °C. The load values measured in this study were compared to numerical estimations by eigenvalue simulations (for an ideal column) and by nonlinear simulations (for a column with initial bending). Two different methods for measurement of the buckling failure load were employed to examine the effect of thermal history on buckling failure. Different load values were obtained from two methods in high temperature conditions over 800 °C. The difference in the buckling failure load between two methods increased with temperature, which was explained by the effect of creep at high temperatures. Moreover, the influence of asymmetric temperature profiles along a plate column was also explored with regard to the failure mode and the buckling failure load. In present study, all of the buckling processes were visualized by a high speed camera.

Research on Emergency System of Injection Estimation of Reactor Core Recovery and Reactor Core Decay Heat Removal

2017 ◽  
Author(s):  
Liao Feiye ◽  
Jiang Pingting ◽  
Liu Wang ◽  
He Dongyu
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Human Error ◽  
Reactor Core ◽  
Heat Removal ◽  
Lessons Learned ◽  
Fukushima Accident ◽  
Decay Heat ◽  
Core Recovery ◽  
Emergency System

One of the lessons learned from Fukushima accident is that the existing procedures used in Nuclear Power Plants (NPPs) are not executed effectively and quickly enough after such an extended accident, for the accident is complex and people are too nervous in such a situation. Thus, emergency system that helps to raise diagnosis efficiency is necessary. In the paper, a quick diagnosis system on injection estimation of reactor core recovery and decay heat removal injection estimation is developed to meet the urgent needs and strengthen requirements for the training and application among utilities and nuclear regulators. The system will assist regulators to quickly know whether the currently flow will probably recover the reactor core, or whether the current injection capacity is sufficient to quench and recover the reactor core, directly after input present parameters into the system. In the system, Matlab method is used, and intuitive insights are considered, which is propitious to give immediate graphical interface and reduce possibility of human error.

scholarly journals Evaluation of Heat Removal from RBMK-1500 Core Using Control Rods Cooling Circuit

2008 ◽  
Vol 2008 ◽  
pp. 1-8
Author(s):  
A. Kaliatka ◽  
E. Uspuras ◽  
M. Vaisnoras
Keyword(s):  
Nuclear Power ◽  
Cooling System ◽  
Reactor Core ◽  
Heat Removal ◽  
Thermal Inertia ◽  
Severe Accident ◽  
Cooling Circuit ◽  
High Thermal Inertia ◽  
The Moment ◽  
Control Rods

The Ignalina nuclear power plant is a twin unit with two RBMK-1500, graphite moderated, boiling water, multichannel reactors. After the decision was made to decommission the Ignalina NPP, Unit 1 was shut down on December 31, 2004, and Unit 2 is to be operated until the end of 2009. Despite of this fact, severe accident management guidelines for RBMK-1500 reactor at Ignalina NPP are prepared. In case of beyond design basis accidents, it can occur that no water sources are available at the moment for heat removal from fuel channels. Specificity of RBMK reactor is such that the channels with control rods are cooled with water supplied by the system totally independent from the reactor cooling system. Therefore, the heat removal from RBMK-1500 reactor core using circuit for cooling of rods in control and protection system can be used as nonregular mean for reactor cooldown in case of BDBA. The heat from fuel channels, where heat is generated, through graphite bricks is transferred in radial direction to cooled CPS channels. This article presents the analysis of possibility to remove heat from reactor core in case of large LOCA by employing CPS channels cooling circuit. The analysis was performed for Ignalina NPP with RBMK-1500 reactor using RELAP5-3D and RELAP5 codes. Results of the analysis have shown that, in spite of high thermal inertia of graphite, this heat removal from CPS channels allows to slow down effectively the core heat-up process.

Innovative Practice of Severe Accidents Measures in CAP1400

2017 ◽  
Author(s):  
Likai Fang ◽  
Xin Liu ◽  
Guobao Shi
Keyword(s):  
Nuclear Power ◽  
Heat Removal ◽  
Severe Accident ◽  
Mitigation Measures ◽  
Spent Fuel ◽  
Fukushima Accident ◽  
Power Sources ◽  
Severe Accidents ◽  
Accident Management ◽  
Accident Conditions

CAP1400 is GenIII passive PWR, which was developed based on Chinese 40 years of experience in nuclear power R&D, construction&operation, as well as introduction and assimilation of AP1000. Severe accidents prevention and mitigation measures were systematically considered during the design and analysis. In order to accommodate high power and further improve the safety of the plant, also considering feedback from Fukushima accident, some innovative measures and design requirements were also applied. Based on the probabilistic&deterministic analysis and engineering judgment, considerable severe accidents scenarios were considered. Both severe accidents initiated at power and shutdown condition were analyzed. Insights were also obtained to decide the challenge to the plant. All known severe accidents phenomena and their treatment were considered in the design. In vessel retention (IVR) was applied as one of the severe accident mitigation measures. To improve the margin of IVR success and verify the heat removal capability through reactor pressure vessel, both design innovative measures and experiments were used. The melt pool behavior and corium pool configuration were also studied by using CFD code and thermodynamic code. Hydrogen risk was mitigated by installation of hydrogen igniters, which were comprised of two serials, and were powered by multiple power sources. To further improve the safety, six extra hydrogen passive recombiners were also added in the containment. Hydrogen risk was analyzed both inside containment and outside containment considering leakage effect. Other severe accident phenomena were also considered by designed or analyzed to show the containment robustness to accommodate it. As one of the Fukushima accident feedback, full scope severe accident management guideline were developed by considering both power condition and shutdown condition, accident management for spent fuel pool was also considered. As the basis of accident management during severe accidents, survivability of equipments and instruments that are necessary in severe accident were assessed and will be further tested and/or analyzed. Such tests will consider severe accident conditions arised from hydrogen combustion.

Nuclear Power Plant Fires and Explosions: Part III — Hamaoka Piping Explosion

2017 ◽  
Author(s):  
Robert A. Leishear
Keyword(s):  
Nuclear Power ◽  
Noble Metals ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Water Hammer ◽  
Preventive Action ◽  
Hydrogen Ignition ◽  
Accident Conditions

An explosion that burst a steel pipe like a paper fire cracker at the Hamaoka Nuclear Power Station, Unit-1 is investigated in this paper, which is one of a series of papers investigating fires and explosions in nuclear power plants. The accumulation of flammable hydrogen and oxygen due to radiolysis has long been recognized as a potential problem in nuclear reactors, where radiolysis is the process that decomposes water into hydrogen and oxygen by radiation exposure in the reactor core. Hydrogen ignition and explosion has long been considered the cause of this Hamaoka piping explosion, but the cause of ignition was considered to be a minor fluid transient, or water hammer, that ignited flammable gasses in the piping, which was made possible by the presence of catalytic noble metals inside the piping. The theory presented here is that a much larger pressure surge occurred due to water hammer during operations. In fact, calculations presented here serve as proof of principle that this explosion mechanism may be present in many operating nuclear power plants. Chubu Electric, the operator of the Hamaoka plant, took appropriate actions to prevent this type of explosion in their plants in the future. In fact, this accident indicates one potential preventive action from explosions for other operating plants. Ensure that a system high point is available, where mixed hydrogen and oxygen may be removed during routine operations and during off-normal accident conditions, such as nuclear reactor meltdowns and loss of coolant accidents.

scholarly journals Analysis of criticality of melt during severe accidents in reactor vessel

2018 ◽  
pp. 3-10
Author(s):  
Yu. Kovbasenko ◽  
Yevgen Bilodid
Keyword(s):  
Boric Acid ◽  
Nuclear Power ◽  
Power Plants ◽  
Reactor Core ◽  
Cooling Water ◽  
Nuclear Fission ◽  
Reactor Vessel ◽  
Severe Accident ◽  
Emergency Situations ◽  
The Core

The article investigates the possibility of a self-sustaining chain nuclear fission reaction during the development of a severe accident in the core at nuclear power plants with reactors WWER-1000 of Ukraine. Some models for calculating a criticality at different stages of the severe accident in the reactor VVER-1000 vessel were developed and calculations of multiplication properties of fuel containing masses were performed. The severe accident in the VVER-1000 core approximately divided into seven major stages: the intact reactor core, beginning of cladding damage (swelling), cladding melting and flowing down to the support grid, melting of constructional materials, homogenization of the materials at the bottom of the reactor vessel, stratification of corium at the bottom of the reactor vessel, the exit of the corium from the reactor shaft. It was shown that at the beginning of an accident, if fuel rods geometry is maintained, criticality might appear even if the emergency protection rods is triggered. With further development of the accident, the melt of fuel and structural materials will be deeply subcritical if water cannot penetrate into the pores or voids of the melt. In the case of the formation of pores or voids in the melt and the ingress of water into them, a recriticality may arise. A compensating measure is the addition of a boric acid solution to a cooling water with a certain concentration. According to the results of the computation analysis, a reactor core loaded with TVSA fuel (Russian production) requires a higher concentration of boric acid in water to compensate the multiplication properties of the fuel system in emergency situations compared to the core loaded with TVS-WR fuel (manufactured by Westinghouse), i.e. TVS-WR fuel is safer from the criticality point of view.

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