Optimization of designing the core fuel loading pattern in a VVER-1000 nuclear power reactor using the genetic algorithm

The Pebble Bed Modular Reactor (PBMR) is a helium-cooled, graphite-moderated high temperature nuclear power reactor which utilise fuel in form of spheres that are randomly loaded and continuously circulated through the core until they reach their prescribed end-of-life burn-up limit. When the reactor is started up for the first time, the lower-enriched start-up fuel is used, mixed with graphite spheres, to bring the core to criticality. As the core criticality is established and the start-up fuel is burned-in, the graphite spheres are progressively removed and replaced with more start-up fuel. Once it becomes necessary for maintaining power output, the higher enriched equilibrium fuel is introduced to the reactor and the start-up fuel is removed. During the initial run of the reactor it is important to discriminate between the irradiated startup fuel and the irradiated equilibrium fuel to ensure that only the equilibrium fuel is returned to the reactor. There is therefore a need for an on-line enrichment discrimination device that can discriminate between irradiated start-up fuel spheres and irradiated equilibrium fuel spheres. The device must also not be confused by the presence of any remaining graphite spheres. Due to it’s on-line nature the device must accomplish the discrimination within tight time limits. Theoretical calculations and experiments show that Fuel Enrichment Discrimination based on delayed neutrons detection is possible. The paper presents calculations and experiments showing viability of the method.

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SANICRO 71

Alloy Digest ◽

10.31399/asm.ad.ni0108 ◽

1965 ◽

Vol 14 (12) ◽

Keyword(s):

Nuclear Power ◽

Elevated Temperatures ◽

Power Reactor ◽

Base Alloy ◽

Heat Treating ◽

Good Resistance ◽

Nuclear Power Reactor ◽

Temperature Performance ◽

Resistance To Stress ◽

Nickel Base

Abstract Sanicro 71 is a nickel-base alloy having good resistance to stress-corrosion, oxidation and creep at elevated temperatures. It is recommended for nuclear power reactor heat exchanger tubes, aircraft turbojet engines and for equipment in the textile, plastic, and chemical industries. This datasheet provides information on composition, physical properties, hardness, elasticity, and tensile properties. It also includes information on high temperature performance and corrosion resistance as well as forming, heat treating, machining, joining, and surface treatment. Filing Code: Ni-108. Producer or source: Sandvik.

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Visualization of Supercritical Carbondioxide Pseudo-Boiling Inside a Rectangular Channel

10.1115/imece2001/htd-24412 ◽

2001 ◽

Author(s):

K. Sakurai ◽

H. S. Ko ◽

K. Okamoto ◽

H. Madarame

Keyword(s):

Forced Convection ◽

Supercritical Fluids ◽

Nuclear Power ◽

Power Reactor ◽

Rectangular Channel ◽

Visual Observation ◽

Next Generation ◽

Nuclear Power Reactor

Abstract The characteristics of the supercritical fluids should be precisely investigated for the next generation nuclear power reactor, i.e., Super-critical water Cooled Reactor (SCR). There are few experiments for visual observation especially in forced convection, because of the difficulty of the experiment.

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