COBRA-IV: the model and the method
The objective of this report is to present the mathematical basis of the COBRA-IV computer program (Wheeler et al., 1976) being developed by Battelle, Pacific Northwest Laboratory. The COBRA-IV code is an extended version of the COBRA-IIIC subchannel analysis code that computes the flow and enthalpy distributions in nuclear fuel rod bundles and cores for both steady state and transient conditions (Rowe, 1973).
- Publication Date:
- OSTI Identifier:
- Report Number(s):
- DOE Contract Number:
- Resource Type:
- Technical Report
- Research Org:
- Pacific Northwest Lab., Richland, WA (USA)
- Country of Publication:
- United States
- 22 GENERAL STUDIES OF NUCLEAR REACTORS; 42 ENGINEERING; HEAT TRANSFER; C CODES; HYDRAULICS; NUCLEAR POWER PLANTS; ROD BUNDLES; FUEL RODS; MATHEMATICAL MODELS; STEADY-STATE CONDITIONS; TRANSIENTS; COMPUTER CODES; ENERGY TRANSFER; FLUID MECHANICS; FUEL ELEMENTS; MECHANICS; NUCLEAR FACILITIES; POWER PLANTS; REACTOR COMPONENTS; THERMAL POWER PLANTS 220200* -- Nuclear Reactor Technology-- Components & Accessories; 420400 -- Engineering-- Heat Transfer & Fluid Flow
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