China
Africa
Explore chapters and articles related to this topic
Burnup, Depletion, and Temperature Feedback
Published in Robert E. Masterson, Introduction to Nuclear Reactor Physics, 2017
During the normal operation of a thermonuclear reactor, some of the neutrons that are produced in the core can escape from the core and eventually strike the pressure vessel wall. Reactor pressure vessels are made from low-alloy carbon steels that are specifically designed to keep the pressure vessels from becoming brittle. (The composition of these pressure vessels was discussed in Chapter 1.) Previously, we learned that these neutrons can consist of a mix of fast and thermal neutrons and that the pressure vessel is approximately 9 in. (20 cm) thick. Each of these neutrons can escape from the core in different amounts that are proportional to the fast and thermal non-leakage probabilities PFNL and PTNL which were introduced to the reader in Chapter 10. The total neutron leakage rate L is therefore given by L=1−PFNLPTNL where L can be a number between 3% and 4% (0.03–0.04) for a commercial power reactor. The neutrons that escape the core inflict a considerable amount of damage to the pressure vessel wall (by dislodging atoms in the carbon–steel matrix from which the pressure vessel is made). In addition, they can penetrate quite far into the pressure vessel if their energy is high enough.
High-Temperature Gas-Cooled Thermal Reactors
Published in Kenneth D. Kok, Nuclear Engineering Handbook, 2016
The PB-1 active core is a cylinder, 2.8 m high, containing 804 fuel elements, 36 control rods, and 19 shutdown rods (Figure 5.2) (Melese and Katz 1984). The fuel elements, 89 mm in diameter, are vertically oriented in a closely packed triangular array with helium flowing up between the elements. The bottom and top graphite reflector sections are an integral part of the fuel element, which has a total height of 3.66 m including the fuel element end fittings. The side reflector, ~60 cm thick, consists of an inner ring of hexagonal graphite elements surrounded by a segmented graphite ring, with a 4 m outer diameter. Helium coolant at 345°C enters the reactor vessel from the outer annuli in the concentric ducts in each of the two loops. It cools the vessel walls and the reflector before flowing up through the core and leaving through the inner concentric ducts at 725°C. The steel reactor pressure vessel, 4.2 m in diameter and 11 m high, is designed for 385°C and 3.1 MPa (31 atm), the actual helium pressure being 2.4 MPa.
A Historical Perspective of Materials Related Structural Integrity Issues in the Nuclear Industry
Published in Peter Hirsch, David Lidbury, Fracture, Plastic Flow and Structural Integrity, 2019
Irradiation embrittlement has been successfully tackled, by and large, by significant progress in modelling and measurement using both accelerated irradiation tests, in-plant surveillance samples, and, more recently, component samples. New embrittlement phenomena have been detected at an early stage and with one or two notable exceptions, it has been possible to maintain ageing reactor pressure vessels in-service, in some cases beyond 40 years. Measurement technology has also advanced such that small materials samples can be extracted from sites of greatest risk for direct measurement of component embrittlement. This has been carried out on Magnox, AGR and light water reactors. Modelling is confirmed by such validation and accurate future trends can then be made.
Investigation on Residual Stress in SA508/Inconel Metal/CF8A Dissimilar Welded Joint for Nuclear Steam Generator Safe End Using Different Processes
Published in Nuclear Technology, 2023
Zhifang Gao, Lei Zhao, Yongdian Han
In nuclear power plants, the light water reactor pressure vessel has complicated structures. A schematic of the reactor system is shown in Fig. 1a, and the steam generator connecting to the channel head and the primary circuit is illustrated in Fig. 1b. The outlet nozzle is usually manufactured with dissimilar metal welding (DMW) to join ferritic carbon steel and austenitic stainless steel pipelines. Generally, nickel-based superalloys are used as the filler metal in welding of ferritic steel and austenitic stainless steel components. The superalloy specifically delays diffusion of carbon in carbon steel through the filler metal. Dissimilar materials joining will minimize the weight and cost of the component. Despite the necessity of DMW, it could cause the component fail prior to the design lifetime,[1–5] which is 300 000 h under the operational conditions of 350°C and 17 MPa. The types and locations of prior failures in DMW joints are still challenges that need to be evaluated because of the complexity of the manufacturing process. Hence, the structural integrity of DMW is a very important consideration to assure the safety and reliability of in-service components.[6–8]
Development and Validation of Thermal-Mechanical Creep Failure Module for Reactor Pressure Vessel Lower Head
Published in Nuclear Science and Engineering, 2023
Hao Yang, Bin Zhang, Pengcheng Gao, Runze Zhai, Jianqiang Shan
In a severe accident causing the meltdown of the pressurized water reactor (PWR) core, considerable melt may be relocated to the bottom of the reactor pressure vessel (RPV) lower head.1 A large amount of melt falling into the lower plenum will generate huge heat and heat the lower head. Such a severe accident of core melting may lead to the rupture of the lower head, resulting in the release of high-level radioactive materials into the containment and even the environment. The in-vessel retention2 (IVR) strategy, which was originated from the improved design of the VVER-440, the second generation reactor in Loviisa, Finland3 was designed to deal with the core melting accident and avoid the release of large amounts of radiation into the environment. Subsequently, the IVR strategy was applied to many newly designed reactors, such as the Westinghouse AP1000 (Ref. 4), South Korea’s APR1400 (Ref. 5), and China’s advanced PWR CAP1400 and HPR1000 (Ref. 6). The RPV lower head is a crucial component to realize the IVR strategy. It is of great significance to reasonably evaluate the mechanical and creep behavior of the lower head in a high-temperature environment.
Verification of Uncertainty Distributions for Thermal-Hydraulic Mixing Parameters Used in KWU-MIX for Pressurized Thermal Shock
Published in Nuclear Technology, 2022
The ability to extend the originally intended operating lifetime of a pressurized water reactor (PWR) depends in part on the ability of the reactor pressure vessel (RPV) to withstand pressurized thermal shock (PTS). One of the several events that lead to PTS (see Ref. 1) is a small-break loss-of-coolant accident (SBLOCA). If the break size is sufficiently large, then the corresponding pressure reduction permits the emergency core cooling (ECC) system to inject cold water into the cold leg. If the break is sufficiently small, then the water level in the downcomer remains above the top of the cold leg. When the flow of coolant through the main coolant pump is low, incomplete mixing occurs between the injected ECC water and the water from the main coolant pump. In this case, a stream of cool water flows along the bottom of the cold leg into the downcomer. There, the cool water forms a downward-flowing buoyant plume surrounded by the hot water in the downcomer.