Pressurized water reactors power plant primary system

These design objectives were carried over to the work on the power reactor PIUS, basically a pressurized water reactor (PWR) in which the primary system has been rearranged in order to accomplish an efficient protection of the reactor core and the nuclear fuel by means of thermal-hydraulic characteristics, in combination with inherent and passive features, without reliance on operator intervention or proper functioning of any mechanical or electrical equipment. Together with wide operating margins, this should make the plant design and its function, in normal operation as well as in transient and accident situations, much more easily understood and with less requirements on the capabilities and qualifications of the operators. 

By the end of 1994, 92 BWR nuclear power plants with a total electrical capacity of about 79 GWe were in operation in the Western countries and Japan an additional 5 plants with about 5.6 GWe were under construction at this time. Within the borders of the former Soviet Union a particular type of BWR had been built, the so-called RBMK reactor 16 plants of this type with about 17 GWe were operating by the middle of 1993. The characteristic feature of the BWR design - in contrast to the closed, one-phase PWR design - is heat removal from the reactor core by boiling water, i. e. by a mixture of water and steam. As a consequence of this difference in design, the behavior of many radionuclides in the BWR primary system during plant operation differs considerably from that in the primary circuit of a pressurized water reactor. 

The CAREM nuclear power plant design is based on a light water integrated reactor. The whole primary system (core, steam generators, primary coolant and steam dome) is contained in a single pressure vessel. Fig. ni-1. 

Because the HTR-10 test reactor is designed on the inherent safety philosophy, safety classifications of systems and components departure from the way it is done for water cooled power reactors For example, primary pressure boundary is defined to the first isolation valve Steam generator tubes are classified as Class II component Diesel generators are not required to be as highly qualified as those used for large water cooled power reactors, since no systems or components with large power demand would require an immediate start of the diesel engines at a plant black-out accident... 

The HPCS system can operate independently of normal auxiliary AC power, plant service air, or the emergency cooling water system. Operation of the system is automatically initiated from independent redundant signals indicating low reactor vessel water level or high pressure in the primary containment. The system also provides for remote-manual startup, operation, and shutdown. A testable check valve in the discharge line prevents backflow from the reactor pressure vessel when the reactor vessel pressure exceeds the HPCS system pressure such as may occur during initial activation of the system. A low flow bypass system is placed into operation until pump head exceeds the nuclear system pressure and permits flow into the reactor vessel. 

The key words of small and modular make the small modular reactors (SMRs) different than other reactors. Small denotes the reactor s decreased power size. Modular denotes (1) the primary coolant system (such as the reactor (RX) component in a light water SMR) enveloped by a pressure boundary and (2) modular construction of components. Modular design requires compact architecture that is built in facility. For instance, the term of modular for a light water SMR (LW-SMR) is used for the RX, since it covers the reactor core and primary coolant system so that the overall power of a power plant can easily be increased by increasing the modular units. 

In 1989, the steam-water power reactor concept was presented by Alekseev and colleagues working in the former USSR [16]. The use of steam-water mixture for the reactor cooling is a key feature of the concept. There are two versions of the steam-water mixture preparation and distribution system. In one, the steam is supplied externally by steam blowers to the RPV and it mixes with feedwater in the special nozzle mixers set at the fuel assembly inlet. In the other, the steam is circulated in the RPV by jet pumps. The steam-water mixture is prepared in the jet pumps. The diagram of the steam-water power reactor is shown in Fig. B.24 [3]. There is no description on the feasibility of steam-water mixture generation. The plant system is indirect cycle. The primary pressure is 16.0 MPa. The core inlet and outlet temperatures are 347 and 360°C, respectively. The core inlet quality is 40%. The average void fraction of the core is estimated to be 93%. The core average coolant density is estimated to be 0.14 g/cm. It should be pointed out that the technical and safety problems will be similar to those of the steam cooled FBR. 

The second heat removal system is an independent cooling system (ICS), which includes, besides a part of primary and secondary circuit equipment, a loop separator-cooling condenser with natural circulation. Via this loop the heat is removed to the intermediate circuit water. This system ensures independent (from the turbine generator systems) reactor cooling and independent reactor plant operation at a constant power level up to 6 % N om at the nominal steam pressure. In case of total RI de-energizig the system ensures cooling of the reactor over several days. Connection/disconnection of ICS is realized with no operator action and without using external power supply systems. 

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