China
Africa
Explore chapters and articles related to this topic
Heat Transfer, Thermal Hydraulic, and Safety Analysis
Published in Kenneth D. Kok, Nuclear Engineering Handbook, 2016
While the reactor protection system is designed to prevent accidents from happening, the emergency core cooling system (ECCS) is designed to respond to those accidents if they do happen. The ECCS is a set of interrelated safety systems that are designed to protect the fuel within the reactor pressure vessel, which is referred to as the reactor core, from overheating. These systems accomplish this by maintaining reactor pressure vessel (RPV) cooling water level, or if that is impossible, by directly flooding the core with coolant. Under normal conditions, heat is removed from a nuclear reactor by condensing steam after it passes through the turbine. In a BWR, condensed steam (water) is fed back into the reactor. In a PWR, it is fed back through the heat exchanger. In both cases, this keeps the reactor core at a constant temperature. During an accident, the condenser is not used, so alternate methods of cooling are required to prevent damage to the nuclear fuel. These systems allow the plant to respond to a variety of accident conditions, and additionally introduce redundancy so that the plant can be shut down even with one or more subsystem failures.
Nuclear Fission Reactor
Published in C. K. Gupta, Materials in Nuclear Energy Applications, 1989
When a reactor is shut down, heat continues to be generated in the fuel by the radioactive decay of the fission products. In the absence of an adequate supply of cooling water, the fuel elements may overheat and be damaged and the fission products would then be released. The purpose of ECCS is to supply cooling water in the event of partial or complete loss of normal coolant flow. The ECCS consists of four separate subsystems and, as shown in Figure 14, they are the high pressure core spray (HPCS), the automatic-depressurization system (ADS), the low-pressure core spray (LPCS) system, and the low-pressure coolant injection (LPCI) system. If the coolant flow loss is small, the system pressure would drop to a moderate extent. An appropriate signal would then actuate the HPCS in which the pump obtains water from the condensate storage tank (later, from the containment suppression pool). The injection piping enters the vessel near the top of the shroud and feeds spargers designed to spray water over the core into the fuel assemblies. In the event of the HPCS failing to maintain the reactor water level, the ADS relief valves operate discharging steam into the pressure suppression pool as referred to later. The resultant reduction in primary circuit pressure would then operate the LPCS and LPCI, the two low-pressure emergency cooling systems. The LPCS involves a pump drawing water from the suppression pool and discharging it from a spray sparger in the top of the reactor vessel above the core. The LPCI involves pumps drawing water from the same source and delivering it into the core, mainly through the jet pump recirculation system.
Nuclear reactors and their fuel cycles
Published in R.J. Pentreath, Nuclear Power, Man and the Environment, 2019
In the USA the development of commercial power reactors arose from the design of reactors required to power submarines. Nuclear power is an ideal form of power for submarines because, in contrast to the burning of fossil fuels, it does not require oxygen. A compact design was derived (figure 4.3.) consisting of a single fuel element assembly of up to 200 zircaloy fuel ‘pins’, each 3.5 m long, immersed in a large steel pressure vessel containing ordinary ‘light’ water. They are therefore known as pressurized water reactors, or PWRs. The light water serves both as moderator and coolant, but because it has a higher neutron-absorbing capacity than heavy water, it is necessary to increase the percentage of 235U in the core. The fuel therefore consists of uranium dioxide in which the fraction of 235U has been enriched from 0.7% to between 2 and 4%. The pressure vessel contains the reactor core, control rods which pass through the lid, and the light water under pressure. Typical operating pressures are about 13.8 to 17.2 MPa (2000 to 2500 psi) so that the water attains a temperature of around 270° C without boiling. The water passes in a closed circuit to a heat exchanger, the circuit including a pressurizer to maintain the pressure by either heating or cooling an appropriate quantity of the water. In order to refuel the reactor it is necessary to shut it down completely, remove the lid, and replace an appropriate portion (about-half) of the fuel pin assembly. This takes place at intervals of 12 to 18 months. An obvious potential danger in this type of reactor is the possible rupture of the cooling system tubing, there being no inherent way of preventing the reactor from overheating if the coolant was suddenly lost. Their design therefore incorporates a number of emergency core-cooling systems (ECCS) and the entire reactor is housed within a pressure containment building, usually double-walled.
Severe Accident Phenomena: A Comparison Among the NuScale SMR, Other Advanced LWR Designs, and Operating LWRs
Published in Nuclear Technology, 2020
Scott J. Weber, Etienne M. Mullin
The NuScale Power Module (NPM) is a small, passively cooled pressurized water reactor (PWR) in which the primary coolant circulates about the RCS via natural circulation, without the need for primary coolant pumps. The RCS is housed entirely in the reactor pressure vessel (RPV), which integrates the core region, pressurizer, and steam generators, obviating the need for large-diameter piping carrying primary coolant ex-vessel. The RPV is enclosed by the steel containment vessel (CNV), which is maintained at a near vacuum during normal operation and is almost entirely submerged in the reactor pool. Each NPM is capable of producing up to 60 MW(electric), and there can be up to 12 NPMs installed in a single plant. In response to a transient, the decay heat removal system (DHRS) actuates to depressurize and cool the NPM via passive heat exchangers located external to the CNV in the reactor pool. In response to a loss-of-coolant event or complete loss of power, a set of five emergency core cooling system (ECCS) valves, which function to circulate coolant between the RPV and CNV and reject heat to the ultimate heat sink, which is the reactor pool, opens. Figure 1 illustrates the natural circulation of primary coolant in the RPV during DHRS operation as well as the two-phase natural circulation between the RPV and CNV during ECCS operation.
Comparison of Multilayer Perceptron and Long Short-Term Memory for Plant Parameter Trend Prediction
Published in Nuclear Technology, 2020
Junyong Bae, Jeeyea Ahn, Seung Jun Lee
In emergency situations, some important safety-related systems such as the reactor protection system and the engineered safety features actuation system work automatically. For example, a reactor trip signal and corresponding actuation signals are automatically generated for the engineered safety features, such as safety injection, containment isolation, and auxiliary feedwater systems. However, as these systems do not automatically perform all actions necessary to cope with emergency situations, operator actions are still necessary. In addition, different strategies are required depending on the type of accident, and thus human operators play an important role in maintaining plant safety. The TMI accident shows how the delayed detection of human errors worsens plant integrity in an emergency situation. In 1979, TMI Unit 2, which was a Babcock & Wilcox two-loop plant, experienced a LOCA due to a pressurizer relief valve that was stuck open. When the pressure decreased below 75% of normal pressure, the emergency core cooling system (ECCS) pumps automatically actuated to inject coolant into the primary system. However, the operators deactivated one ECCS pump and throttled back the ECCS flow because they were not aware that the reactor had experienced the LOCA due to incorrect information about the pressurizer relief valve. Consequently, coolant continued to leak, which ultimately led to core damage.7
Multiscaled Experimental Investigations of Corrosion and Precipitation Processes After Loss-of-Coolant Accidents in Pressurized Water Reactors
Published in Nuclear Technology, 2019
Stefan Renger, Sören Alt, Ulrike Gocht, Wolfgang Kästner, André Seeliger, Holger Kryk, Ulrich Harm
During a loss-of-coolant accident (LOCA) in a pressurized water reactor (PWR), the outflow leaking coolant is accumulated in the containment sump. Following the injection of the accumulators and from the emergency core cooling system (ECCS) tanks, the coolant recirculates in the long-term phase of a LOCA from the sump to the reactor and the leak driven by the low pressure ECCS pumps to remove decay heat from the core. In comparison to other reactor designs, German PWRs are not equipped with containment spray systems. No alkaline substances are added during LOCA. The pH (25°C) value of the coolant remains in the neutral to slightly acidic range, the latter due to the presence of boric acid.1 During the flow from the leakage to the sump suction tubes in the containment, coolant gets in contact with containment installations. Parts of these installations are made of hot-dip galvanized steel, e.g., gratings, flight of stairs, inspection platforms, room divider, and support grids of sump strainers (Fig. 1). Vertical cross sections of German PWR containments can be found in Ref. 2.