Accid. Anal. for Nuclear Powerplants

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It is not expected that the regulatory body will always conduct a complete set of independent analyses for every submittal within the licensing process. However, the development of its in-house capacity for accident analysis gives the regulatory body greater capability in its own decision making process as well as in communication with the licensee. As regulatory bodies are typically not legally required to perform their own quantitative assessments, they can select some spot check calculations to verify the consistency of the accident analyses submitted.

Some uncertainties related to model structures are difficult to quantify formally and may be impossible to eliminate. The use of steady state developed and qualified correlations for transient calculations may also introduce uncertainties which cannot be estimated. The analyst needs to attempt to reduce the effects of such uncertainties upon the predictions (see also Section 7) but must be aware that they cannot be totally eliminated. Uncertainties in plant data are important and are usually available from historical records of actual operating plants.

1. TYPES OF COMPUTER CODES For anticipated transients and DBAs, these codes can be organized by the component or system being analysed and in general can be characterized into the following six categories: (a) (b) (c) (d) (e) (f) Reactor physics codes; Fuel behaviour codes; Thermohydraulic codes, including system codes, subchannel codes, porous media codes and computational fluid dynamics (CFD) codes; Containment analysis codes, possibly also with features for the transport of radioactive materials; Atmospheric dispersion and dose codes; Structural analysis codes.

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