Pressurized water reactors reactor coolant pressurizer

The nuclear steam supply system (NSSS) of the QP300 consists of a pressurized water reactor, reactor coolant system (RCS) and associated auxiliary systems. The NSSS has retained the general design features of current PWR plant design. 

Fig. 1. Pressurized water reactor (PWR) coolant system having U-tube steam generators typical of the 3—4 loops in nuclear power plants. PWR plants having once-through steam generators contain two reactor coolant pump-steam generator loops. CVCS = chemical and volume-control system.

Fig. 21. Schematic of a pressurized-water-loop reactor coolant system.

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)...

The different reactivity control systems in a nuclear power plant allow keeping at any time the control of the nuclear fission reactions in the core power steering, safe reactor shutdown, wear compensation of the fuel. They are also part of the neutron protection of the out-of-core components. These systems can take various forms gas (such as helium 3 in some experimental reactors), liquid (soluble boron in pressurized water reactor (PWR) coolant to balance the reactivity evolution of the reactor), and most of the time solid (Table 15.1). In a reactor, they are most often combined [e.g., in PWR with Ag-In-Cd (AIC) plus boron carbide control rods and with boron present both as soluble boron and as boron carbide]. In all cases those materials incorporate neutron-absorbing nuclides, unlike the fuel which is a medium generally multiplier... 

Most nuclear reactors use a heat exchanger to transfer heat from a primary coolant loop through the reactor core to a secondary loop that suppHes steam (qv) to a turbine (see HeaT-EXCHANGETECHNOLOGy). The pressurized water reactor is the most common example. The boiling water reactor, however, generates steam in the core. 

Eig. 3. Schematic of a pressurized water reactor system. Eission heat is extracted by the lightwater coolant. The steam drives the turbine-generator. 

The key feature of the pressurized water reactor is that the reactor vessel is maintained above the saturation pressure for water and thus the coolant-moderator does not bod. At a vessel pressure of 15.5 MPa (2250 psia), high water temperatures averaging above 300°C can be achieved, leading to acceptable thermal efficiencies of approximately 0.33. 

FIGURE 17.25 A schematic representation of one type of nuclear reactor in which water acts as a moderator for the nuclear reaction. In this pressurized water reactor (PWR), the coolant is water under pressure. The fission reactions produce heat, which hoi Is water in the steam generator the resulting steam turns the turbines that generate electricity. 

The BWR reqnires substantially lower primary coolant flow through the core than pressurized water reactors. The core flow of a BWR is the sum of the feedwater flow and the recirculation flow, which is typical... 

Ion-exchange membranes also show some promise in the solution of waste problems, inter alia the treatment of spent pickle liquors by electrodialysis is discussed (80). A very high degree of deionization can be achieved with ion-selective membranes. P. Cohen investigated the electrodialysis of simulated pressurized water reactor coolant (34). The specific resistance of the treated water was as low as 0.5 to 3,0 Mi cm. 

Pressurized Water Reactor (PWR) A type of nuclear power reactor that uses ordinary water as both the coolant and the neutron moderator. The heat produced is transferred to a secondary coolant which is subsequently boiled to produce steam for power generation. 

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

The actual fuel elements in pressurized water reactors consist of individual fuel rods and control rod tubes mounted in a self-supporting construction of spacers fitted with a top and feet. Fuel elements for boiling water reactors, by comparison, have no control rod tubes, the fuel element zirkaloy claddings being used to guide the control rods and the coolant. 

When H20 is used as coolant, the same material serves also as moderator, so ihat the reactor structure can be simplified. Figure 1.8 is a schematic diagram of a pressurized-water reactor, in which the coolant and moderator consist of liquid water whose pressure of 150 bar (2200 Ib/in ) is so high that it remains liquid at the highest temperature, around 300°C (572°F), to which it is heated in the reactor. The main difference in principle from Fig. 1.7 is... 

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. 

Two-circuit reactor plant (RP) with a vessel-type pressurized water reactor is used for floating power unit of nuclear head and power station. Basic RP components reactor, steam generators and primary coolant pumps are incorporated by pressure nozzles in a compact steam-generating block. KLT-40C RP characteristics are given in Table 1. 

In the beginning of the 1950s nearly at the same time the USA and USSR launched the development of the nuclear power installations (NPI) for nuclear submarines (NS). In both countries the work was carried out for two types of NPIs with pressurized water reactors and reactors cooled by liquid metal coolant (LMC). 

The coolant does not boil if die primary circuit loses its tightness, and has die property to retain iodine, as a rule, its radionuclides represent the major factor of radiation danger just after die accident as well as the other fission products (inert gases are exception) and actinides. This reduces sharply a scale of radiation consequences of that accident in comparison with pressurized water reactors. 

The primary coolant is a substance that transports the heat generated by the nuclear chain reaction away from the reactor core. In a pressurized water reactor, which is the most common commercial reactor design, water acts as both the moderator and the primary coolant. 

Criterion 28 - Reactivity limits. The reactivity control systems shall be designed with appropriate limits on the potential amount and rate of reactivity increase to assure that the effects of postulated reactivity accidents can neither (1) result in damage to the reactor coolant pressure boundary greater than limited local 5uelding nor (2) sufficiently disturb the core, its support structures or other reactor pressure vessel internals to impair significantly the capability to cool the core. These postulated reactivity accidents shall include consideration of rod ejection (imless prevented by positive means), rod dropout, steam line rupture, changes in reactor coolant temperatme and pressure, and cold water addition. 

In the 1970s, there was a series of unanticipated operational events that occurred in commercial operating pressurized water reactors (PWRs) in the USA (NRC, 2012).These events resulted in pressures and temperatures in the RPV that were outside the P-TUmits specified for normal operation. The conditions associated with these unanticipated events could be placed into two categories. Rrst, there were approximately 30 transient events where the pressure in the RPV exceeded the allowable pressure at relatively low temperature. These events were isothermal pressure transients that generally occurred at temperatures below approximately 93 C (200°F) during reactor start-up. In many instances, the transient pressures were several times the allowable pressure. Typically, the transients occurred while the reactor coolant system was filled with water and were a result of operators failing to follow appropriate procedures to control and prevent... 

Overpressure protection for the reactor coolant pressure boundary is provided by four spring-loaded ASME Code safety valves connected to the top of the pressurizer. These valves discharge to the in-containment refueling water storage tank, where the steam is released under water to be condensed and cooled. If the steam discharge exceeds the capacity of the in-containment refueling water storage tank, it is vented to the containment atmosphere. 

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