1. Steam Generation 1.1. General Characteristics 1.2. Circulation in Boiling Water Reactors 1.3. Steam Generation in CANDU Reactors 1.4. Steam Generators 1.5. Steam Generator for PWR and CANDU Systems 1.6. Heat Transfer 2. Steam Generator Operation 2.1. Swelling and Shrinking 2.1.1. Steady State Swelling and Shrinking 2.1.2. Transient Swelling and Shrinking 2.2. Level Control 2.3. Carryover 2.4...

1. Reactor Kinetics 1.

The paper discussed the development of a simulator for the Programmable CVD Reactor system using the modular simulation approach. The design of the Programmable Reactor lends itself very naturally to a modular approach because of the segmented showerhead design while each segment is a fairly complex component, with individually controllable reactant gas supply and residual gas exhaust, the desi...

In Korea, the technology of a load fo nuclear power plant has been being developed. automatic controller which is able to control temperature and axial power distribution was automatic controller is functionally divided identification algorithm and a model predictive con former transforms the nuclear reactor status into numerically. And the latter uses them and ge manipulated values such as two...

The Next Generation Nuclear Plant (NGNP) Project and the Nuclear Hydrogen Initiative (NHI) are U.S. Department of Energy programs that are responsible for the research, development, and testing of high-temperature gas-cooled nuclear reactor technologies and thermally driven hydrogen production methods. The two programs are related, in that the nuclear reactor will provide high-temperature heat,...

Dynamic probabilistic risk assessment (DPRA) methodologies couple system simulator codes (e.g., RELAP, MELCOR) with simulation controller codes (e.g., RAVEN, ADAPT). Whereas system simulator codes model system dynamics deterministically, simulation controller codes introduce both deterministic (e.g., system control logic, operating procedures) and stochastic (e.g., component failures, parameter...

REACTOR is an expert system under development at EG&G Idaho, Inc., that will assist operators in the diagnosis and treatment of nuclear reactor accidents. This paper covers the background of the nuclear industry and why expert system technology may prove valuable in the reactor control room. Some of the basic features of the REACTOR system are discussed, and future plans for validation and eval...

Studsvik’s approach to spent nuclear fuel analyses combines isotopic concentrations, fluxes, and cross-sections, calculated by the CASMO5 neutron transport depletion code, with irradiation history data from SIMULATE5 reactor core simulator tabulated decay data. These sources are used processed SNF code compute characteristics. Recent advances in system, including cross-sections ENDF/B-VIII.R0, ...