Critical Heat Flux in TRIGA-Fueled Reactors Cooled by Natural Convection

Avery, Michael; Yang, Jun; Anderson, Mark; Corradini, Michael L.; Feldman, E.E.; Dunn, F.E.; Matos, J.E.

doi:10.13182/nse11-69

Cited by 9 publications

(4 citation statements)

References 10 publications

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“…It was shown that the critical heat flux (CHF) correlations included in RELAP5 result in a non-conservative estimate of power at which CHF occurs in TRIGA reactors with respect to the traditional Bernath CHF correlation. Until further comparisons with respect to the experimental data presented in Refs [26,27] are performed, the level of conservatism with respect to measurements remains undetermined. It was also shown that for low-pressure system, the boiling model implemented in RELAP5/MOD3.3 is more accurate than the one in RELAP5/MOD3.2 and RELAP5-3D.…”

Section: Discussionmentioning

confidence: 99%

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Summary of RELAP5 Assessments Performed in Relation to Conversion Analysis of Research Reactors

Dionne¹,

Dunn²,

Feldman³

et al. 2014

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Section: Discussionmentioning

confidence: 99%

“…The applicability of a various CHF correlations was further studied experimentally and analytically by the authors of Ref. [26]. They performed an experiment studying low-pressure, natural convection CHF with simulated (electrically heated) full fuel pins under TRIGA reactor conditions.…”

Section: Analysis Of Critical Heat Flux In Trigamentioning

confidence: 99%

Summary of RELAP5 Assessments Performed in Relation to Conversion Analysis of Research Reactors

Dionne¹,

Dunn²,

Feldman³

et al. 2014

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“…In some tests, several thermocouples are welded along the wall of the tube near the outlet, and they make the monitoring of the temperature. When one abrupt increase of the temperature in some thermocouple happens, the power supply is turned off [4]. This method has several inconveniences: (i) the speed of the detection depends on the thermocouple proximity of the point where the "burn-out" happens;…”

Section: Temperature Measuringmentioning

confidence: 99%

Detection of the Departure from Nucleate Boiling in Nuclear Fuel Rod Simulators

Mesquita¹,

Rodrigues²

2013

International Journal of Nuclear Energy

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In the thermal hydraulic experiments to determin parameters of heat transfer where fuel rod simulators are heated by electric current, the preservation of the simulators is essential when the heat flux goes to the critical point. One of the most important limits in the design of cooling water reactors is the condition in which the heat transfer coefficient by boiling in the core deteriorates itself. The heat flux just before deterioration is denominated critical heat flux (CHF). At this time, the small increase in heat flux or in the refrigerant inlet temperature at the core, or the small decrease in the inlet flux of cooling, results in changes in the heat transfer mechanism. This causes increases in the surface temperature of the fuel elements causing failures at the fuel (burnout). This paper describes the experiments conducted to detect critical heat flux in nuclear fuel element simulators carried out in the thermal-hydraulic laboratory of Nuclear Technology Development Centre (CDTN). It is concluded that the use of displacement transducer is the most efficient technique for detecting critical heat flux in nuclear simulators heated by electric current in open pool.

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“…TRIGA reactors that rely on natural convection for primary flow are among the most common research reactors in the US. Since these reactors can operate with subcooled nucleate boiling during normal operation, the margin to CHF can be a limiting safety design criterion and a limitation for power production [13,14].…”

Section: Introduction *mentioning

confidence: 99%

Bandung Triga 2000 Reactor Power Analysis as a Function of the Number of Fuel Elements and the Power Peaking Factor

et al. 2020

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The reactivity value of the Bandung TRIGA 2000 reactor core has decreased over time, so the power generated by the reactor is also getting smaller, despite the control rod position is fully withdrawn. Therefore, it is necessary to reshuffle and refuel the fuel element to increase the excess reactivity by considering the safety parameters, such as axial and radial power peaking factors, DNBR, dTsat, and temperature on the cladding and in the center of the fuel element. The analyzed reactor safety parameters are the number of fuel elements, which varied at 105, 110, and 115 elements, as well as power peaking factor, which varied at 1.55, 1.65, 1.75, 1.85, and 1.95. The calculations were done using MCNP and COOLOD-N2 programs. If DNBR ≈ 1.3 is determined as the safety limit for the operation of the Bandung TRIGA 2000 reactor, at PPF 1.95 (105, 110, and 115 fuel elements), it can be considered to operate the reactor at the power of 600-700 kW. However, at PPF of 1.75 (105, 110, and 115 fuel elements), the reactor can be operated at the power of 700-800 kW, and at PPF of 1.55 (105, 110, and 115 fuel elements), the reactor can be considered for operation at the power of 800-900 kW. The results of these calculations can be used for consideration in determining the operating limits of the Bandung TRIGA 2000 reactor.Keywords: TRIGA 2000, fuel element, power peaking factor, DNBR, boiling

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Critical Heat Flux in TRIGA-Fueled Reactors Cooled by Natural Convection

Cited by 9 publications

References 10 publications

Summary of RELAP5 Assessments Performed in Relation to Conversion Analysis of Research Reactors

Summary of RELAP5 Assessments Performed in Relation to Conversion Analysis of Research Reactors

Detection of the Departure from Nucleate Boiling in Nuclear Fuel Rod Simulators

Bandung Triga 2000 Reactor Power Analysis as a Function of the Number of Fuel Elements and the Power Peaking Factor

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