Advanced Engineering Tools for Structural Analysis of Advanced Power Plants Application to the GE ESBWR Design
- General Electric, 75 Curtner Ave, San Jose CA 95125 (United States)
- Empresarios Agrupados, c/ Magallanes, 3, 28015 Madrid (Spain)
Experience in the design of nuclear reactors for power generation shows that the plant structures and buildings involved are one of the major contributors to plant capital investment. Consequently, the design of theses elements must be optimised if cost reductions in future reactors are to be achieved. The benefits of using the 'Best Estimate Approach' are well known in the area of core and systems design. This consists in developing accurate models of a plant's phenomenology and behaviour, minimising the margins. Different safety margins have been applied in the past when performing structural analyses. Three of these margins can be identified: - increasing the value of the load by a factor that depends on the load frequency; - decreasing the resistance of the structure's resistance, and - safety margins introduced through two step analysis. The first two type of margins are established in the applicable codes in order to provide design safety margins. The third one derives from limitations in tools which, in the past, did not allow obtaining an accurate model in which both the dynamic and static loads could be evaluated simultaneously. Nowadays, improvements in hardware and software have eliminated the need for two-step calculations in structural analysis (dynamic plus static), allowing the creation one-through finite element models in which all loads, both dynamic and static, are combined without the determination of the equivalent static loads from the dynamic loads. This paper summarizes how these models and methods have been applied to optimize the Reactor Building structural design of the General Electric (GE) ESBWR Passive Plant. The work has focused on three areas: - the design of the Gravity Driven Cooling System (GDCS) Pools as pressure boundary between the Drywell and the Wet-well; - the evaluation of the thickness of the Reactor Building foundation slab, and - the global structural evaluation of the Reactor Building.
- Research Organization:
- American Nuclear Society, 555 North Kensington Avenue, La Grange Park, IL 60526 (United States)
- OSTI ID:
- 21167960
- Resource Relation:
- Conference: ICAPP'02: 2002 International congress on advances in nuclear power plants, Hollywood, FL (United States), 9-13 Jun 2002; Other Information: Country of input: France
- Country of Publication:
- United States
- Language:
- English
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