{"id":32336,"date":"2022-05-18T05:43:28","date_gmt":"2022-05-18T05:43:28","guid":{"rendered":"https:\/\/sitepourvtc.com\/?page_id=32336"},"modified":"2023-09-26T10:16:10","modified_gmt":"2023-09-26T10:16:10","slug":"safety-systems","status":"publish","type":"page","link":"https:\/\/sitepourvtc.com\/nuclear-power\/reactor-physics\/nuclear-safety\/safety-systems\/","title":{"rendered":"Safety Systems"},"content":{"rendered":"
Most nuclear power plants introduce a \u2018defense-in-depth<\/b>\u2018 approach to achieve maximum safety, and this approach is constituted of multiple safety systems supplementing the natural features of the reactor core. Level 3 and level 4 usually rely on various safety systems<\/strong>, structures, and components. Engineered safety features<\/strong> and protection systems<\/strong> are provided to prevent evolution towards severe accidents and confine radioactive materials within the containment system. The measures at this level aim to prevent core damage in particular. Design and operating procedures are also aimed at maintaining the effectiveness of the barriers, especially the containment. For example, the emergency core cooling system (ECCS) is provided to mitigate the consequences of a loss-of-coolant accident (LOCA), even though the first level of defense makes such an occurrence highly unlikely.<\/p>\n In the regulatory arena, the term \u201csafety-related\u201d applies to systems, structures, components, procedures, and controls (of a facility or process) that are relied upon to remain functional during and following design-basis events. Safety-related systems, structures, and components have three characteristics. They ensure:<\/p>\n A containment isolation valve is safety-related, for example, because isolating the reactor coolant lines confines radioactivity to the containment building and performs the functions defined above. It helps keep radioactivity away from the public.<\/p>\n An emergency diesel generator is safety-related because providing backup power to safety-related equipment ensures the capability to shut down the reactor and maintain it safely.<\/p>\n The IEEE created its term for \u201csafety-related electric equipment,\u201d which is \u201cClass 1E.\u201d In IEEE 308 it gives the definition of Class 1E as follows:<\/p>\n The safety classification of the electric equipment and systems that are essential to emergency reactor shutdown, containment isolation, reactor core cooling, and containment and reactor heat removal or that are otherwise essential in preventing the significant release of radioactive material to the environment.<\/p>\n See also: IAEA Safety Standards, Safety Classification of Structures, Systems, and Components in Nuclear Power Plants. Specific Safety Guide No. SSG-30, ISBN 978\u201392 \u20130\u2013115413\u20132. Vienna, 2014.<\/p>\n Nowadays, the most common nuclear reactors (PWRs and BWRs) rely mostly on active safety systems. Active in the sense that they involve electrical or mechanical operation on command systems (e.g., high-pressure water pumps). But the trend is to introduce more passive design features.<\/p>\n Passive nuclear safety<\/strong> is a design approach that is more or less in use in nuclear power plants. Passive safety systems are designed to accomplish safety functions without any active intervention on the part of the operator or electrical\/electronic feedback to bring the reactor to a safe shutdown state in the event of a particular type of emergency (usually overheating resulting from a loss of coolant or loss of coolant flow). These systems take advantage of natural forces or phenomena such as gravity, pressure differences, or natural heat convection.<\/p>\n The primary design objective of the advanced passive technology is to provide greatly simplified nuclear plant designs that meet or exceed the latest regulatory requirements and safety goals while being economically competitive with other systems.<\/p>\n Passive safety systems include: passive safety injection, passive residual heat removal, and passive containment cooling. These systems have been designed to meet the NRC single-failure and other recent criteria.<\/p>\n More recently, however, new reactor designs are making more extensive use of passive safety features for a variety of purposes, for instance, for core cooling during transients, design basis accidents or even severe accidents, or for containment cooling, with the claim that passive systems are highly reliable and reduce the cost associated with the installation and maintenance of systems requiring multiple trains of equipment requiring expensive pumps, motors, and other equipment as well as redundant safety class power supplies.<\/p>\n The Reactor Protection System, RPS, is one of the safety systems and provides the following functions:<\/p>\n The RPS automatically initiates a rapid reactor shutdown (scram) by inserting control rods to preserve the integrity of the fuel cladding and reactor coolant pressure boundary. Also, the overall purpose of the reactor protection system is to prevent the release of radioactivity into the environment. The initiation of a reactor trip by the RPS prevents the core from operating in a condition that could cause damage to the core.<\/p>\n The protection system normally uses 2\/3 or 2\/4 logic. A 2\/3 logic means that a trip occurs when at least 2 signals out of 3 indicate a trip condition.<\/p>\n Reactor trip signals provided by the system are usually as follows:<\/p>\n Due to its importance to safety, the RPS is designed, constructed, and tested to the highest standards. These include requirements for the ability to withstand single failures and still provide full protection, for the independence of separate trains, and for testability to insure continued reliability.<\/p>\n The main purpose of the engineered safety features is to prevent or limit the escape of radioactivity to the environment in cases of a highly unlikely transient or accident that is too severe to be managed by the reactor protection system alone. A protective action is generated when a sufficient number and combinations of these set-points have been met or exceeded.<\/p>\n The ESF functions include:<\/p>\n The purpose of the Emergency Core Cooling Systems (ECCS) aims to provide core cooling under loss-of-coolant accident (LOCA) conditions to limit fuel cladding damage. The ECCS limits the fuel cladding temperature below the limit so that the core will remain intact and in place, with its essential heat transfer geometry preserved. The Code of Federal Regulations, CFR, requires the ECCS to be designed so that after any LOCA, the reactor core remains in a geometrical configuration amenable to cooling. The basic criteria are limiting fuel cladding temperature and oxidation to minimize clad fragmentation and the hydrogen generation from clad oxidation to protect the containment.<\/p>\n The ECCS usually consists of redundant high-pressure systems (e.g.,3×100%) and redundant low-pressure systems (e.g.,3×100%).<\/p>\n See also: Decay Heat Removal<\/a><\/p>\n The containment building is primarily designed to prevent or mitigate the uncontrolled release of radioactive material to the environment in operational states and accident conditions. Therefore it is considered the fourth and final barrier<\/strong> in the Defence-in-depth <\/strong>strategy.<\/p>\n While containment plays a crucial role in Design Basis Accidents or Design Extension conditions, it is \u201conly\u201d designed to condense steam<\/strong> from primary coolant and to contain it inside<\/strong> the building.<\/p>\n In case of Design Basis Accidents such as the Large Break Loss of Coolant Accident (LBLOCA), the pressure increase is usually significant, and active containment systems (pressure-suppression systems<\/strong>) must be available to maintain the integrity (to keep the pressure and temperature under certain limits) of the containment building.<\/p>\n Pressure-suppression systems are critical to safety and greatly affect the size of containment. Suppression refers to condensing the steam after a major break has released it from the cooling system. There are many designs of suppression systems around the world.<\/p>\n Most of Pressurized Water Reactors (PWRs)<\/a> containments use two-stage pressure-suppression systems:<\/p>\n When pressure increases inside the containment indicated, the containment spray system is automatically started, and the pumps (usually with 3\u00d7100% redundancy) take suction from the tank (refueling water storage tank can also be used) and pump the water into spray nozzles located in the upper part of the containment. The water droplets, being cooler than the steam, will remove heat from the steam, which will cause the steam to condense. This will cause a reduction in the pressure of the building and will also reduce the temperature of the containment atmosphere. The spray system usually contains extra chemical additives dissolved in the tank to enhance the removal of particular radionuclides from the containment atmosphere. Especially radioiodine, which is of particular importance, can be effectively bonded to potassium hydroxide or sodium hydroxide.<\/p>\n Most Boiling Water Reactors (BWR)<\/a> containments use pressure-suppression pools to maintain the integrity of the containment building. The major containment designs are Mark I, Mark II, and Mark III. The Mark I and Mark II containments consist of two main parts:<\/p>\n Water spray systems are usually installed in both the drywell and the wetwell. The Mark III design consists of primary containment and a drywell.<\/p>\n Containment buildings and containment pressure-suppression systems vary widely depending on certain reactor designs. In some cases, unique technologies can be installed. For example, the containment building of Loviisa NPP uses two ice condensers as the pressure-suppression system.<\/p>\n Hydrogen mitigation<\/strong> in water-cooled power reactors is an important area of study in the field of the safety of nuclear reactors<\/a>. Hydrogen and oxygen can be generated during the normal operation of a power reactor primarily due to the radiolysis of the water in the core<\/a>.<\/p>\n During accidents, hydrogen and oxygen can also be generated as a consequence of:<\/p>\n During DBAs (Design Basis Accidents),<\/strong> such as the large break loss of coolant accident, the production of hydrogen (metal\u2013water reactions in the core) is limited at low values by the operation of the emergency core cooling systems. Also, the radiolysis of the water in the core is a relatively slow process. Therefore, from the DBA\u2019s point of view, the hydrogen hazards can be eliminated by maintaining the local hydrogen concentration below its flammability limit (4% of volume). This requirement can be ensured by mixing devices or thermal hydrogen recombiners<\/strong>.<\/p>\nSafety-related<\/h2>\n
\n
Class 1E<\/h2>\n
Active and Passive Nuclear Safety<\/h2>\n
Reactor Protection System<\/h2>\n
\n
\n
\n
Engineered Safety Features<\/h2>\n
\n
Emergency Core Cooling System – ECCS<\/h2>\n
\n
Containment Systems<\/h2>\n
\n
\n
\n
Hydrogen Mitigation in Water Cooled Power Reactors<\/h3>\n
\n