Pressurized water reactor primary systems

The BWR operates at constant pressure and maintains constant steam pressure similar to most fossil-fueled hollers. The BWR primary system operates at pressure about onc-half that of a pressurized water reactor primary system, while producing steam of equal pressure and quality. 

USNRC (1997a) Assessment of pressurized water reactor primary system leaks , NUREG/CR - 6582, INEEL/ EXT-97-01068. 

Pressurized Water Reactor. The PWR contains three coolant systems the primary system, which removes heat from the reactor... 

Or to use a design in which the core of a conventional pressurised water reactor (PWR) is enclosed within a vessel of boronated water that will flood the core if pressure is lost there is no barrier between the core and the pool of water, which in case pressure in the primary system is lost will shut the reactor down and continue to remove heat from the core by natural circulation. It is calculated that in an accident situation, replenishing of cooling fluid can be done at weekly intervals (in contrast to hours or less required for current light water reactor designs) (Harmerz, 1983 Klueh, 1986). 

As motioned in Chapter 19, the name implies that a pressurized water reactor is cooled by hot high pressure water, either H2O (PWR, VVER) or DjO (PHWR). In the PWR and VVER types the coolant is also us as moderator whereas a separate D2O containing moderator tank is normally used in the PHWR type. These power reactor types have several things in common primary — secondary coolant circuits separated by heat exchangers (steam generators), a pressurizer to adjust primary system pressure and often diemical shim control for adjustment of the excess reactivity with fresh fuel. 

Minimum pressurization-temperature (MPT) curves specify the temperature and pressure limitations for reactor plant operation. They are based on reactor vessel and head stress limitations and the need to preclude reactor vessel and head brittle fracture. Figure 4 shows some pressure-temperature operating curves for a pressurized water reactor (PWR) Primary Coolant System (PCS). 

SMART (System-integrated Modular Advanced ReacTor) is an advanced integral PWR(Pressurized Water Reactor) tihat produces 330MWt at fiill power. Major primary components are housed within a single pressure vessel. New, advanced and innovative features are incorporated in the design to provide the reactor with significant enhancements in safety, reliability, performance, and operability. Major design and safety characteristics of SMART can be summarized as follows ... 

A high pressure injection system (HPI) with three pumps for the injection of borated water in the reactor. In emergency operation, which is automatically activated by low pressure of the primary system or by high pressure in the containment building, two pumps activate. Analyses show that only one pump is necessary to prevent core damage in cases of small breaks in the cooling system. 

PWR reactor Nuclear reactor where the core power is transported by pressurized water which circulates in a system of primary circuits. The production occurs within a set of Heat Exchangers (Steam Generators), using the thermal energy contained in primary water (PWR = Pressurized Water Reactor). 

In a nuclear power station, there are several ion-exchange systems for water clean-up. In a typical BWR there are purification circuits after the steam condenser, in a small stream from the reactor vessel and for the water in the fuel storage pool. In a pressurized water reactor (PWR), part of the coolant in the primary circuit is withdrawn, cooled, and pumped through an ion-exchange filter. As indicated by the word filter, these ion exchangers are also used to remove particles from the liquids. They often consist of mixed cation and anion exchangers that are very finely ground bead sizes of 400 mesh are common. 

High temperature and pressure steam in the Pressmized Water Reactor s (PWR) system are below the temperature and pressure of the primary coolant, on another word, to have better efficiency, we are obliged to increase the temperature and pressure of the primary system. This is the origin of the problem, where to handle complication come up, as a result of reactor operating conditions, the system becomes sophisticated, and in the same time, safety and safety related items increase and the choice of material used becomes more and more difficult. 

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. 

For comparison, it may be outlined that the primary system of a conventional pressurized water reactor (PWR) has a pressure vessel, two to four steam generators, a pressurizer, two to four pumps and all the corresponding connection pipes. 

It is a back-up of the first shutdown system. It is capable of causing reactor shutdown by injecting borated water into the primary circuit. Two tanks filled with borated water are connected to the primary circuit during normal operation, with the same operative pressure as the primary system. When the Boron injection system is triggered by the protection system, a valve opens and borated water, driven by gravity, floods the primary circuit. The system has two redundant trains, with two redundant triggering valves in each train. 

This report examines the severe accident sequences and radionuclide source terms at the Sizewell pressurised water reactor with a piestressed concrete containment, the Konvoi pressurized water reactor with a steel primary contaimnent, the European Pressurised water Reactor (EPR) and a boiling water reactor with a Mark 2 containment. The report concludes that the key accident sequences for European plant designs are transient events and small loss-of-coolant accidents, loss of cooling during shutdown, and containment bypass sequences. The most important chemical and transport phenomena are found to be revaporisation of volatile radionuclides from the reactor coolant system, iodine chemistry, and release paths through the plant. Additional research is recommended on release of fission products from the fuel, release of fission products from the reactor coolant system, ehemistry of iodine, and transport of radionuclide through plants. 

Figure 1.1. Four-loop primary coolant system of a 1300 MWe pressurized water reactor a) Reactor pressure vessel b) Steam generator c) Reactor coolant pump d) Pressurizer e) Pressurizer relief tank (Meyer, 1991)...

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. 

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