Modifications to the MELCOR code for application in fusion accident analyses

• Published in: Fusion engineering and design Vol. 51; pp. 555 - 563
• Main Authors: Merrill, B.J, Moore, R.L, Polkinghorne, S.T, Petti, D.A
• Format: Journal Article Conference Proceeding
• Language: English
• Published: Amsterdam Elsevier B.V 01-11-2000; New York, NY Elsevier Science
• Subjects: Accident prevention; Applied sciences; Controled nuclear fusion plants; Energy; Energy. Thermal use of fuels; Engineering design activity; Exact sciences and technology; Experimental reactors; Installations for energy generation and conversion: thermal and electrical energy; International thermonuclear experimental reactor; Loss of coolant accidents; Loss-of-cooling accidents; Loss-of-vacuum accidents; MELCOR; safety engineering; Vacuum technology; Loss-of-vacuum accidents; International thermonuclear experimental reactor; Loss-of-cooling accidents; MELCOR; Engineering design activity; Nuclear safety; Tokamak; Nuclear fusion reactor; LOCA; Data analysis; Database; Computation code; Safety; Modeling; LOVA; Heat transfer; Nuclear reactor
• ISSN: 0920-3796; 1873-7196
• DOI: 10.1016/S0920-3796(00)00220-9

For the past several years, the Fusion Safety Program at the Idaho National Engineering and Environmental Laboratory (INEEL) has modified the MELCOR code in order to assess safety issues associated with loss-of-cooling accidents (LOCAs) and loss-of-vacuum accidents (LOVAs) in the international thermonuclear experimental reactor (ITER) engineering design activity (EDA). MELCOR is a thermal hydraulics computer code developed by the Sandia National Laboratory for analyzing severe accidents in fission power plants. This paper describes these modifications and the role they played in LOVA and LOCA analyses performed for the non-site specific safety report (NSSR) for the ITER EDA.