Experiment data report for Semiscale Mod-1 Test S-02-4 (blowdown heat transfer test)

Crapo, H.S.; Jensen, M.F.; Sackett, K.E.

doi:10.2172/4101003

Cited by 6 publications

(2 citation statements)

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“…Cladding temperature response of Rod FS -Tests S-02-9A and S-02-9. ------= 5. 4--1---r,--1 -1--r f Tes, s-(2'F 'i < , 1'1 -rt-IT ...In---tr-i--i-lI---1--4 1----.tIll --4---i--'-.---.--..…”

mentioning

confidence: 99%

Core thermal response during Semiscale Mod-1 blowdown heat transfer tests. [PWR]

Larson¹

1976

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The author wishes to express his thanks to the personnel of the Fluids Laboratory Division for their efforts in conducting the tests and gathering the data, to the personnel of the' 'l'est integration and Coordination Scction for vcr~f~cation and documcntiltion of the data, to R. F. Farman and R. W. Shumway for technical support, to D. M. Snider for supplying core heat transfer calculations, to G. W. Johnsen and J. K. Meier for supplying E L A P 4 calculations, to K. A. Dietz for editing and iillprovillg the quality of the overall report, and to L. A. Nelson and L. K. Keller for typing the various drafts.

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“…Cladding temperature response of Rod FS -Tests S-02-9A and S-02-9. ------= 5. 4--1---r,--1 -1--r f Tes, s-(2'F 'i < , 1'1 -rt-IT ...In---tr-i--i-lI---1--4 1----.tIll --4---i--'-.---.--..…”

mentioning

confidence: 99%

Core thermal response during Semiscale Mod-1 blowdown heat transfer tests. [PWR]

Larson¹

1976

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“…The~oco~pl~ locations in se:misc}Jle•Mod-1 heated core 8 6, COBRA-IV-I single-channel model of Semiscale Mod~ I core 7 .…”

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confidence: 99%

Analysis of the thermal-hydraulic behavior resulting in early critical heat flux and evaluation of CHF correlations for the semiscale core. [PWR; LOFT]

Snider¹

1977

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The expert consultation by C. McMonagle on the use of COBRA-IV-I and the initial COBRA-IV-I calculations by S. C. Resch are greatly appreciated. The performing of Series 2 te:sts h.y personnel of the Fluids Laboratory is acknowledged. Editing of this report by K. Dietz and help with figures by M.Tyacke is appreciated. The Loss-of-Fluid Test (LOFT) correlation used in this study was from R. C. Gottula of the LOFT Program .. ii .,.

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Transient boiling heat transfer from initial steady state caused by rapid depressurization

Fukuda

Shiotsu

Hata

et al. 1994

Nuclear Engineering and Design

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Experiment data report for Semiscale Mod-1 Test S-02-4 (blowdown heat transfer test)

Cited by 6 publications

References 0 publications

Core thermal response during Semiscale Mod-1 blowdown heat transfer tests. [PWR]

Core thermal response during Semiscale Mod-1 blowdown heat transfer tests. [PWR]

Analysis of the thermal-hydraulic behavior resulting in early critical heat flux and evaluation of CHF correlations for the semiscale core. [PWR; LOFT]

Transient boiling heat transfer from initial steady state caused by rapid depressurization

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