COBRA IIIC:  digital computer program for steady state and transient thermal-hydraulic analysis of rod bundle nuclear fuel elements

Rowe, D.S.

doi:10.2172/4480166

Cited by 88 publications

(41 citation statements)

References 1 publication

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“…Similar nuclear data are provided by the Japanese Evaluated Nuclear Data Library (JENDL) (Shibata et al, 2011). Among deterministic methods, material temperatures can be calculated with a thermal hydraulics code like COBRA (Rowe, 1973) or PLTEMP (Olson and Kalimullah, 2010). COBRA is a subchannel code for calculating heat transfer in Light Water Reactor (LWR) cores in rod bundle geometries and PLTEMP is a code for steady-state heat transfer calculations in cores having plate or tube type fuel elements (e.g.…”

Section: Reactor Modelsmentioning

confidence: 99%

Integration of facility modeling capabilities for nuclear nonproliferation analysis

García¹,

Burr²,

Coles³

et al. 2012

Progress in Nuclear Energy

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Developing automated methods for data collection and analysis that can facilitate nuclear nonproliferation assessment is an important research area with significant consequences for the effective global deployment of nuclear energy. Facility modeling that can integrate and interpret observations collected from monitored facilities in order to ascertain their functional details will be a critical element of these methods. Although improvements are continually sought, existing facility modeling tools can characterize all aspects of reactor operations and the majority of nuclear fuel cycle processing steps, and include algorithms for data processing and interpretation. Assessing nonproliferation status is challenging because observations can come from many sources, including local and remote sensors that monitor facility operations, as well as open sources that provide specific business information about the monitored facilities, and can be of many different types. Although many current facility models are capable of analyzing large amounts of information, they have not been integrated in an analyst-friendly manner. This paper addresses some of these facility modeling capabilities and illustrates how they could be integrated and utilized for nonproliferation analysis. The inverse problem of inferring facility conditions based on collected observations is described, along with a proposed architecture and computer framework for utilizing facility modeling tools. After considering a representative sampling of key facility modeling capabilities, the proposed integration framework is illustrated with several examples. Highlights: We consider integration of facility modeling into a tool for proliferation analysis.  We describe the inverse problem of inferring the facility based on observables.  We propose architecture for integrating facility models into a suitable framework.  We illustrate the proposed integration framework with several examples.

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Section: Reactor Modelsmentioning

confidence: 99%

Integration of facility modeling capabilities for nuclear nonproliferation analysis

García¹,

Burr²,

Coles³

et al. 2012

Progress in Nuclear Energy

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show abstract

“…O r i g i n a l development o f t h e COBRA computer code through t h e COBRA-IIIC version (Rowe 1973) was sponsored by t h e Atomic Energy Commission (AEC) .…”

Section: V -I Code (Wheeler E T a L 1976) I S An Extended Versionmentioning

confidence: 99%

COBRA-IV PC: A personal computer version of COBRA-IV-I for thermal-hydraulic analysis of rod bundle nuclear fuel elements and cores

Webb¹

1988

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“…The first axial point shall be 0.0 and the last axial point shall be 1.0. The formats [1][2][3][4][5] for these cards are 8F10.5. Section III.…”

Section: /'T 18 \mentioning

confidence: 99%

“…On the next page, it will print out the average [1][2][3][4][5][6][7] i. Sperature inside the. duct, the total heat generation of the assembly, the enthalpy increase inside the duct, and the enthalpy increase in the gap of the assembly, It 07-inf;s out the radial temperature distribution of the assembly'.…”

Section: Outputmentioning

confidence: 99%

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Prediction of the coolant temperature field in a breeder reactor including interassembly heat transfer

Chen

Todreas

1975

Nuclear Engineering and Design

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COBRA IIIC: digital computer program for steady state and transient thermal-hydraulic analysis of rod bundle nuclear fuel elements

Cited by 88 publications

References 1 publication

Integration of facility modeling capabilities for nuclear nonproliferation analysis

Integration of facility modeling capabilities for nuclear nonproliferation analysis

COBRA-IV PC: A personal computer version of COBRA-IV-I for thermal-hydraulic analysis of rod bundle nuclear fuel elements and cores

Prediction of the coolant temperature field in a breeder reactor including interassembly heat transfer

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