Pressurized water reactor , general

We illustrate the general principles of thermal reactors by a short description of the two most inqx)itant power reactor types the pressurized water reactor (PWR) and the boiling water reactor (BWR). They are further discussed in Chapter 20. 

The pressurized water reactor is generally preferred for propulsion purposes (military surface vessels and submarines), partly because it can react faster on changes in power demand than many other types of thermal reactors. 

WAPD-MRP-107, "Pressurized Water Reactor (PWR) Project Technical Progress Report October 24, 1963-January 23, 1964," Westinghouse Atomic Power Div. Griggs, B., Maffei, H. P., and Shannon, D. W., HW-67818, "Multiple Rate Transitions in the Aqueous Corrosion of Zircaloy," Hanford Laboratories, General Electric Company, December 20, 1960. 

The control effectiveness of such alloys in water-moderated reactors can approach that of hafnium and is the control material commonly used in pressurized-water reactors. The alloys (generally 80% silver, 15% indium, 5% cadmium) can be readily fabricated and have adequate strength at water-reactor temperatures. The control material is enclosed in a stainless steel tube to protect it from corrosion by the high-temperature water. 

PWRs and BWRs were earlier being developed for naval and civilian purposes in the industrially developed countries. Whereas in Russia naval pressurized water reactors have been known since longer time to be utilized in icebreakers and container ships. Earlier this year, presentations of Russian representatives were explaining these technologies for stationary (inland) as well as floating power plants of PWR type to the BATAN and general audiences in Indonesia. 

In this chapter, after giving an overview of the embrittlement of Western pressurized water reactor (PWR) reactor pressure vessel (RPV) beltline materials, together with the characteristics of PWR RPVs, such as their general specification, core region materials and the effects of variables on embrittlement, the surveillance database obtained from US, French and Japanese nuclear power plants (NPPs) and those from other countries is presented based on open literature. The surveillance program of each country is also briefiy described. Trends of surveillance data which will be obtained in the near future are described. The possibility of new data from reconstituted and miniature specimen techniques is described. 

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... 

A typical pressure vessel of a pressurized water reactor is shown in Fig. 57.11. The height of the pressure vessel for a typical power station of 1,300 MWe generally is 12 m. The thickness of the cylindrical wall amounts to 25 cm. The tank is designed to withstand pressures of 175 bar and a temperature of 350° C. Typical parameters of a PWR are listed in Table 57.10. 

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. 

The purpose of this chapter is to provide a general insight into the manufacture of fuels used in nuclear reactors. The primary focus will be on uranium dioxide (UO2) fuels for light water reactors (LWRs), including both the pressurized water reactor (PWR) and the boiling water reactor (BWR). Many of the details relating to the fuel for these reactors are also presented in Sections 1.2 and 1.3 of this handbook. Some of the information from those chapters will be repeated for clarity. 

The primary consequence of burnup is a drop in /c-effective as the fuel bums out and fission products are built up. This drop is compensated by the build-up of new fissile isotopes (notably Pu-239 from U-238 neutron absorption in uranium-fueled reactors). Generally, boiling water reactors and pressurized water reactors replace the fuel in stages, with fresh fuel assemblies replacing the most burned-out assemblies at scheduled shutdowns with nonreplaced assemblies often moved (shuffled) to new positions to optimize the reactor operating characteristics. 

Neeb, K. H., Schuster, E. Origin and general behaviour of increased release of fission iodine from defective fuel rods upon shutdown of a pressurized water reactor. Siemens Forsch. u. Entwickl. Berichte 8, 92-97 (1979)... 

Contamination buiidup in pressurized water reactors 4.4.3.1 General aspects... 

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... 

Some industrial examples (Table 2) demonstrate these differences between gas-cooled and liquid-cooled nuclear reactors. The General Atomics gas-cooled HTGR (the GT-MHR) has a AT across the reactor core of 359 C, while the British Advanced Gas-Cooled Reactor (Hinkley Point B) has a AT of 355 C. Liquid-cooled reactors t5q)ically have much-smaller temperature increases across the reactor core. The Point Beach pressurized-water reactor (PWR) has a AT across the reactor core of 20 C, while the French liquid-metal fast reactor (Super Phenix) has a AT of 150 C. A liquid-cooled reactor can deliver all of its heat with small temperature differences (20 to 150 C) between (1) the hottest temperatures in the reactor coolant, piping, and heat exchangers and (2) the maximum temperature of the chemical reagents in the H2 production facility. 

Other pressurized water reactors (PWRs) have experienced reactor trip breaker failures, both before and after the February 1983 Salem 1 events. None of them however, involved an ATWS event. The reactor trip breaker failures prior to the February 1983 events at Salem 1 had been the subject of several actions taken since 1971 by the AEC/NRC, Westinghouse, and General Electric. 

The first volume of the series is intended to be as generally applicable as possible to all reactor types. The specific features of individual reactor types are taken into account in the subsequent reports in the series. The reactor types to be covered include pressurized water reactors (PWRs), boiling water reactors (BWRs), pressurized heavy water reactors (PHWRs) or more specifically PHWRs of Canadian design known as CANDU, and RBMKs. Ihe present report is devoted to specific guidance for PHWRs. 

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