Nuclear reactors coolant flow

There are a number of situations where countercurrent two-phase flow can exist in nuclear reactor coolant channels. For example, during emergency core cooling of the BWR rod bundles at low flow has steam and water countercurrent flow. The water flow rate can continue for certain ranges of water and steam flow rates. However, the relative velocity between the steam and water creates waves on the liquid surface for large gas velocities. And as the steam velocity increases the waves reach the channel walls and block the downward flow of the water. This transition point is called flooding or countercurrent flow limit. Further increase in steam velocity leads another transition where water is carried upward and thus flow reversal occurs. The transitions are associated with large pressure drop in the pipe. 

The accidental depressurisation transient is analysed by using LOFTRAN. For reactor coolant system depressurisation analyses that include a primary coolant flow coast down caused by a consequential loss of offsite power, a combination of three computer codes is used to perform the DNBR analyses. First LOFTRAN is used to perform the plant system transient. FACTRAN is then used to calculate the core heat flux based on nuclear power and reactor coolant flow from LOFTRAN. Finally, VIPRE-01 is used to calculate the DNBR using heat flux from FACTRAN and flow from LOFTRAN. 

An "anticipated transient" is an event that is expected to occur one or more times during the life of a nuclear power plant. There are a number of anticipated transients, some quite trivial and others that are more significant in terms of the demands imposed on plant equipment. Anticipated transients include such events as a loss of electrical load that leads to closing of the turbine stop valves, a load increase such as opening of a condenser bypass valve, a loss of feedwater flow, and a loss of reactor coolant flow. 

Transient two-phase flow in rod bundles. In analyzing transient two-phase flows in rod bundles, such as the case resulting from a postulated loss-of-coolant or flow accident in a nuclear reactor, Ishii and Chawla (1978) developed a multi-... 

Solverg, K. O., and P. Bakstad, 1967, A Model for the Dynamics of Nuclear Reactors with Boiling Coolant with a New Approach to the Vapor Generation, Proc. Symp. on Two-Phase Flow Dynamics at Eindhoven, EURATOM Rep. (6)... 

One is the secondary- coolant reduction test by partial secondary loss of coolant flow in order to simulate the load change of the nuclear heat utilization system. This test will demonstrate that the both of negative reactivity feedback effect and the reactor power control system brings the reactor power safely to a stable level without a reactor scram, and that the temperature transient of the reactor core is slow in a decrease of the secondary coolant flow rate. The test will be perfonned at a rated operation and parallel-loaded operation mode. The maximum reactor power during the test will limit within 30 MW (100%). In this test, the rotation rate of the secondary helium circulator will be changed to simulate a temperature transient of the heat utilisation system in addition to cutting off the reactor-inlet temperature control system. This test will be performed under anticipated transients without reactor scram (ATWS). 

One of the attractive features of the fast reactor is its hard neutron spectrum. To expand this feature, a metallic fuel core is employed in the 4S. However, it is more difficult to reduce void reactivity for a core with a harder spectrum. It is very important to design the void reactivity to be negative in order to prevent a severe nuclear accident in the event of sudden loss of coolant, sudden loss of coolant flow or a large gas bubble entrainment in the core. 

The rods 106 and 108 are cooled by any suitable coolant such as molten bismuth or a sodium potassium alloy 5q circulated as indicated by the arrows in FIG. 8 by inlet and outlet conduits 109 and 111. The inlet conduit is connected to the outer tank 104 and the coolant flows downwardly therethrough into the tank 102 through ports at the bottom thereof and upwardly therethrough into the 55 outlet conduit 111. Thus, the coolant absorbs the heat of the nuclear fission chain reaction and the energy in the form of heat carried from the reactor by the coolant may be utilized for power or other purposes, if desired. 

In the period of 1998-99, two sets of experiments focused on problems of rapid decrease of concentration of boric acid in reactor coolant at nuclear reactor core inlet were performed at the University of Maryland, US, under the auspices of OECD. The situation, when there is an inadvertent supply of boron-deficient water into the reactor vessel, could lead to a rapid (very probably local) increase of reactor core power in reactor, operated at nominal power, or to a start of fission reaction in shut-down reactor (secondary criticality). In the above mentioned experiments the transport of boron-deficient coolant through reactor downcomer and lower plenum was simulated by flow of cold water into a model of reactor vessel. These experiments were selected as the International Standard Problem ISP-43 and organisations, involved in thermal — hydraulic calculations of nuclear reactors, were invited to participate in their computer simulation. Altogether 10 groups took part in this problem with various CFD codes. The participants obtained only data on geometry of the experimental facility, and initial and boundary conditions. 

Deposition of corrosion products in a circulating AT liquid metal system is important for three reasons. First, the degradation of heat transfer perfonmance of heat exchangers must be predicted. Second, radiation exposure limits for maintenance in certain areas of nuclear reactor systems that transport and deposit radioactive species must be controlled. Third, the tendency for all deposits to become detached by thermal shock or flow perturbations must be known since there is concern that these types of debris could block critical coolant channels. It is therefore valuable, when possible, to monitor reactions involving deposition as well as dissolution. 

The Fixed Bed Nuclear Reactor (FBNR) concept assumes the use pressurized water reactor (PWR) technology, but incorporates hi temperature gas cooled reactor (HTGR) type fuel and the concept of a suspended fixed bed core. Spherical fuel elements are fixed in the suspended core by the flow of water coolant. Any accident signal will cut off the power to the coolant pump causing a stop in the flow. This would make the fuel elements fall out of the reactor core, driven by gravity, and enter a passively cooled fuel chamber where they would reside in a subcritical condition. The Fixed Bed Nuclear Reactor (FBNR) is a simplified version of the fluidized bed nuclear reactor concept [XII-1 to XII-9]. In the FBNR, spherical fuel elements are in a fixed position in the core therefore, there is no concern about the consequences of multiple collisions between them, an issue that may be raised about the fluidized bed concept. Relatively little work has been done for the fixed bed nuclear reactor so far, but the experiences gained from the development of a fluidized bed reactor can facilitate the development of the FBNR. 

Technology to maintain certain oxygen regime of lead-bismuth coolant to eliminate corrosion and erosion of stainless steel claddings in coolant flow. This technology is available in the Russian Federation, which has an 80-year operation experience with small lead-bismuth cooled reactors for nuclear submarines. 

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