CORTAP: a coupled neutron kinetics-heat transfer digital computer program for the dynamic simulation of the high temperature gas cooled reactor core

Cleveland, J.C.

doi:10.2172/7315582

Cited by 4 publications

(6 citation statements)

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“…Because many of these codes were developed for the analysis of systems pre-dating the current modular HTGR designs being considered, the flexibility and adaptability each code becomes a major consideration. [10] MINET [11] ORTAP [12] RATSAM [13] TAP [14] THERMIX [15] CORTAP [16] COBRA-IIIc [17] NAKOGAS [18] ORECA [19] THERMIX Core CHAP [10] Developed by Los Alamos, this code is for the transient analysis of the 1) Fort St. Vrain Plant or the 2) 3000 MWt unit designed a few years back. It was also used for the analysis of the 2240 MWt design considered recently, as this design had similarities to the 3000 MWt unit already factored into CHAP.…”

Section: )mentioning

confidence: 99%

“…ORTAP [12] Developed by Oak Ridge, ORTAP is a combined form of the CORTAP [16], ORECA [19], BLAST [20], and ORTURB [27] codes. The inclusion of CORTAP and ORECA indicates the reactor representation is for the prismatic core.…”

Section: )mentioning

confidence: 99%

“…CORTAP [16] Developed by Oak Ridge, this code provides a single fuel channel representation of a prismatic core HTGR reactor, and includes point kinetics. It does not appear to hold great potential for the analysis of a pebble-bed reactor.…”

Section: )mentioning

confidence: 99%

“…For the pebble-bed system, the choices remain THERMIX and MINET. [10] MINET [11] ORTAP [12] RATSAM [13] TAP [14] THERMIX [15] CORTAP [16] COBRA-IIIc [17] NAKOGAS [18] ORECA [19] THERMIX Core In order to assure the necessary computer code library to handle the three types of transients considered in this subsection, i.e., loss of heat sink, loss of coolant, and reactivity, we recommend the following actions:…”

Section: Transients With Scrammentioning

confidence: 99%

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Assessment of modelling needs for safety analysis of current HTGR concepts

Kroeger¹,

Tuyle²

1985

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In view of the recent shift in emphasis of the DOE/lndustry HTGR development efforts to smaller modular designs it became necessary to review the modelling needs and the codes available to assess the safety performance of these new designs. This report provides a final assessment of the most urgent modelling needs, comparing these to the tools available, and outlining the most significant areas where further modelling is required. Plans to implement the required work are presented.

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Section: )mentioning

confidence: 99%

Section: )mentioning

confidence: 99%

Section: )mentioning

confidence: 99%

Section: Transients With Scrammentioning

confidence: 99%

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Assessment of modelling needs for safety analysis of current HTGR concepts

Kroeger¹,

Tuyle²

1985

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“…•The same data 7 were used by GA for validation of BLOOST, with the result that BLOOST predicts an undershoot very similar to the CORTAP calculations.…”

Section: Comparisons Of Cortapmentioning

confidence: 99%

High-temperature gas-cooled reactor safety studies for the Division of Reactor Safety Research. Quarterly progress report, January 1-March 31, 1980

Ball¹,

Cleveland²,

Conklin³

et al. 1980

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Modular High-Temperature Gas-Cooled Reactor Short-Term Thermal Response to Flow and Reactivity Transients

Cleveland

1990

Nuclear Technology

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The research repomd here has been conducted at the Oak Ridge National Laboratory for the Nuclear Regulatory Commission's Division of Regulatory Applications of the Office of Nuclear Regulatory Research. Tbe short-tenn thermal response of the Modular High-Temperature Gas-Cooled Reactor (MHTGR) is analyzed for a range of flow and reactivity transients. These transients include loss of forced circulation without scram, spurious withdrawal of a control rod group, moisture ingress, control rod and control rod group ejections, and a rapid core cooling event. For each event analyzed, an event description, a discussion of the analysis approach and assumptions, and results are presented. When possible, results of these analyses are compared with those presented by the designers in the MHTGR Preliinary Safety Information Document and in the MHTGR Probabilistic Risk Assessment. The importance of inherent safety features is illusrrated, and conclusions are presented regarding the safety performance of the MHTGR. Recommendations are made for a more indeptb examination of MHTGR response for some of the analyzed transients. The coupled heat transfer-neutron kinetics model is described in detail in Appendix A.

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CORTAP: a coupled neutron kinetics-heat transfer digital computer program for the dynamic simulation of the high temperature gas cooled reactor core

Abstract: NOTICE: This document contains information oi a prelim inary nature. It Is subject to revision or correction and therefore does not represent a final report.

Cited by 4 publications

References 1 publication

Assessment of modelling needs for safety analysis of current HTGR concepts

Assessment of modelling needs for safety analysis of current HTGR concepts

High-temperature gas-cooled reactor safety studies for the Division of Reactor Safety Research. Quarterly progress report, January 1-March 31, 1980

Modular High-Temperature Gas-Cooled Reactor Short-Term Thermal Response to Flow and Reactivity Transients

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