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Energy and Environment
Published in T.M. Aggarwal, Environmental Control in Thermal Power Plants, 2021
In the nuclear plant field, steam generator refers to a specific type of large heat exchanger used in a pressurized water reactor (PWR) to thermally connect the primary (reactor plant) and secondary (steam plant) systems, which generates steam. In a nuclear reactor called a boiling water reactor (BWR), water is boiled to generate steam directly in the reactor itself and there are no units called steam generators.
Nuclear Boilers
Published in Charles F. Bowman, Seth N. Bowman, Thermal Engineering of Nuclear Power Stations, 2020
Charles F. Bowman, Seth N. Bowman
A BWR is a light water reactor in which the feedwater coming from the highest pressure feedwater heater passes directly over the fuel rods in the reactor to capture the heat generated by the fission reaction taking place in the reactor. In so doing, the feedwater is heated to the saturation temperature at the reactor’s operating pressure (nominally about 1,000 lbf/in2A (6,895 kPa)) and steam is boiled off through steam separators and dryers at the top of the reactor that wring out a portion of the moisture entrained in the steam. The result is main steam with a quality of approximately 99.75% (0.25% moisture) passing to the HP turbine. Although proven to be highly reliable, one significant disadvantage of a BWR is the fact that the steam, condensate, and feedwater passing through the turbine cycle are everywhere radioactive and thus require special care in performing operation and maintenance activities in the turbine building.
Nuclear Power
Published in Viorel Badescu, George Cristian Lazaroiu, Linda Barelli, POWER ENGINEERING Advances and Challenges, 2018
Both PWRs and boiling water reactors (BWRs) are thermal light-water reactors (LWRs) which use enriched uranium as fuel and water as both coolant and neutron moderator. The major difference between these two type of reactors is that while in a PWR the primary circuit is under high pressure and steam is produced in the steam generator of a secondary circuit, in a BWR the water boils while passing through the reactor core, producing steam at the reactor vessel output line. While the direct production of steam simplifies the overall plant scheme due to the lack of a secondary circuit, it introduces the additional problem of induced radioactivity of the turbine due to the activated steam. A typical BWR pressure vessel is shown in Fig. 7.
Modeling of distribution parameters for upward steam-water boiling flows in subchannels of a vertical rod bundle
Published in Journal of Nuclear Science and Technology, 2023
Tetsuhiro Ozaki, Takashi Hibiki
Boiling water two-phase flow has been utilized in engineering facilities as a heat transfer system because of its excellent heat transfer characteristics. In LWRs, the heat generated from fissile materials such as uranium is effectively transferred to the coolant, and the generated steam rotates a turbine to extract electrical energy. In a BWR, the coolant in the fuel bundle becomes a steam-water two-phase flow, and the steam, after passing through a steam-water separator, directly drives the turbine. The plant configuration of a BWR is simpler than that of a PWR, which receives heat from the core indirectly by a secondary cooling system. Hundreds of fuel bundles in the BWR core form independent channels by channel boxes. In these channels, the flow distribution in the core is adjusted by orifices at the inlet to appropriately cool individual fuel elements by removing generated heat. The fuel burnup proceeds by nuclear fission caused by moderated neutrons. The behaviors of neutrons, i.e. the neutron moderation characteristics, strongly depend on the coolant condition. In general, the lower the void fraction, the more effectively the neutrons are moderated and the more easily the fission occurs, resulting in higher heat power and higher burnup. Thus, the coolant conditions and the nuclear characteristics of the fuel are closely related, and understanding the thermal-hydraulic characteristics in the core is indispensable for core management, such as fuel arrangement in the core, operation planning, and stability evaluation.
Machine-Learning Analysis of Moisture Carryover in Boiling Water Reactors
Published in Nuclear Technology, 2019
Haoyu Wang, Andrew Longman, J. Thomas Gruenwald, James Tusar, Richard Vilim
The boiling water reactor (BWR) operates on a saturated steam cycle relying on mechanical equipment to remove liquid moisture from a two-phase mixture that exits the core. Excess moisture carryover (MCO) can lead to both intergranular stress corrosion cracking of turbine blades and an increase in exposure levels of maintenance personal to 60Co that is carried over in the liquid droplets that leave the steam dryers.1 While the BWR in the United States was originally designed to achieve MCO levels that are acceptable from a plant economic performance perspective, changes in recent years in the way these plants are operated and in the reactor core design efficiencies have led to increased MCO levels.1 Specifically, new fuel management strategies and the use of increased core flow (>100%) to extend the cycle aimed at minimizing fuel costs have led to a greater fraction of liquid reaching the turbine as a consequence of the increased load on the steam separators and steam dryers.
Ten years of Fukushima Dai-Ichi post-accident research on the degradation phenomenology of the BWR core components
Published in Journal of Nuclear Science and Technology, 2022
Anton Pshenichnikov, Hiroki Shibata, Takuya Yamashita, Yuji Nagae, Masaki Kurata
The experimental work is still ongoing with the hope of successful and cost-effective decommissioning of the 1F. But not only this, new experimental data will promote future safe and sustainable energy production. The part of the gained expertise will be surely transferred to improve the safety of the next generation of reactors. For now, the BWR type of reactor remain the second most significant type of electric power sources for many years until the advanced reactors of the Generation IV will replace them. The work on bridging the gaps in the accident progression knowledge is indispensable. The facility upgrade and new tests regarding the melt progression near the core plate and the lower head failure tests with aerosol analysis are planned in the future.