Bücher zum Thema „Nuclear reaction codes“

Neighbour, Gareth B. Securing the safe performance of graphite reactor cores. Cambridge, UK: RSC Pub., 2010.

Boer, Brian. Optimized core design and fuel management of a pebble-bed type nuclear reactor. Amsterdam: IOS Press, 2008.

Ross, Kyle. MELCOR best practices as applied in the State-of-the-Art Reactor Consequence Analyses (SOARCA) Project. Washington, DC: U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, 2014.

Turnbull, J. Anthony. Review of nuclear fuel experimental data: Fuel behaviour data available from IFE-OCDE Halden Project for development and validation of computer codes. Paris: Nuclear Energy Agency, Organisation for Economic Co-operation and Development, 1995.

Bilanovic, Z. Neutron-photon energy deposition in CANDU reactor fuel channels: A comparison of modelling techniques using ANISN and MCNP computer codes. Chalk River, Ont: System Chemistry and Corrosion Branch, Chalk River Laboratories, 1994.

Centre, Bhabha Atomic Research, Hrsg. Operational reactor physics analysis codes (ORPAC). Mumbai: Bhabha Atomic Research Centre, 2007.

M, Blann, und OECD Nuclear Energy Agency, Hrsg. International code comparison for intermediate energy nuclear data = Comparaison internationale de codes pour le calcul de données nucléaires aux énergies intermédiaires. Paris: Nuclear Energy Agency, Organisation for Economic Co-operation and Development, 1994.

Management of ageing in graphite reactor cores. Cambridge: RSC Publishing, 2007.

B, Murfin W., Johnson Jay D, U.S. Nuclear Regulatory Commission. Division of Safety Issue Resolution., Sandia National Laboratories, Technadyne Engineering Consultants und Science Applications International Corporation, Hrsg. XSOR codes users manual. Washington, DC: Division of Safety Issue Resolution, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1993.

Ma, Chang Chun. Study of interfacial condensation in a nuclear reactor core makeup tank. 1993.

Management of Ageing Processes in Graphite Reactor Cores (Special Publication) (Special Publications). Royal Society of Chemistry, 2007.

Organization for Economic Co-operation and Development und NEA. In-Core Instrumentation and Reactor Core Assessment: Proceedings of a. OECD (Organisation for Economic Co-Operation & Dev, 1997.

Use of Computational Fluid Dynamics Codes for Safety Analysis of Nuclear Reactor Systems (Iaea Tecdoc Series). International Atomic Energy Agency, 2004.

Kaya, Sadi. COBRA-OSU: A fast running computer code for coupled kinetic-thermal hydraulic analysis of nuclear reactor cores. 1986.

Kaya, Sadi. COBRA-OSU: A fast running computer code for coupled kinetic-thermal hydraulic analysis of nuclear reactor cores. 1986.

NEA Nuclear Science Committee., International Atomic Energy Agency und Nihon Genshiryoku Kenkyūjo, Hrsg. In-core instrumentation and reactor core assessment: Proceedings of a specialist meeting, Mito-shi, Japan, 16-17 October 1996. Paris: Nuclear Energy Agency, Organisation for Economic Co-operation and Development, 1997.

Intermediate energy nuclear data: Models and codes : proceedings of a specialists' meeting, Issy-Les-Moulineaux, France, 30 May-1 June 1994. Paris: Organisation for Economic Co-operation and Development, 1994.

development, Organisation for economic co-operation and, und Nuclear Energy Agency. Intermediate Energy Nuclear Data: Models and Codes : Proceedings of a Specialists' Meeting Issy-Les-Moulineaux (France 30 May-1 June 1994). Organization for Economic, 1994.

OECD/NEA-CSNI international standard problem ISP36: CORA-W2 experiment on severe fuel damage for a Russsian type PWR : comparison report. Issy-les-Moulineaux, France: Committee on the Safety of Nuclear Installations, OECD Nuclear Energy Agency, 1996.

Cheung, F. B. Natural Circulation Phenomena in Nuclear Reactor Systems: Presented at 1994 International Mechanical Engineering Congress and Exposition, Chicago, Ill (Fact). American Society of Mechanical Engineers, 1994.

B, Cheung F., McAssey E. V, American Society of Mechanical Engineers. Heat Transfer Division. und International Mechanical Engineering Congress and Exposition (1994 : Chicago, Ill.), Hrsg. Natural circulation phenomena in nuclear reactor systems: Presented at 1994 International Mechanical Engineering Congress and Exposition, Chicago, Illinois, November 6-11, 1994. New York: American Society of Mechanical Engineers, 1994.

G, Chen, und United States. National Aeronautics and Space Administration., Hrsg. A computational fluid dynamic and heat transfer model for gaseous core and gas cooled space power and propulsion reactors. [Washington, D.C.]: National Aeronautics and Space Administration, 1996.