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In water‐cooled nuclear reactors, the maximum power which can be extracted from the core is limited by critical heat flux (CHF). CHF in the high‐quality region is known as dryout. In advanced nuclear reactors, the coolant flow occurs solely by virtue of natural circulation; however, instabilities may occur during off‐normal operations. This may lead to premature dryout due to lower coolant flow rates seen by the heater during such oscillations. This paper describes the experimental investigation on the effect of flow oscillations on the CHF with the time period of 120 s, which is observed typically in the large‐scale natural circulation system. Based on observations made with respect to temperature transient, the continuous dryout is preceded by the transient dryout for higher flow oscillations. But as flow fluctuation decreases, the transient dryout phenomenon is found to disappear. The applicability of the look‐up table to predict CHF under oscillatory flow conditions using suitable correction factors (CFs) for premature dryout has been evaluated. CFs for the CHF under oscillations suggested by previous authors have been compared. The maximum possible degradation in CHF value suggested by previous authors has been found to agree with the present experimental data. Percentage fluctuation in heat transfer coefficient (HTC) at fully developed annular flow conditions has been evaluated, and it is found that fluctuation in HTC is in phase with the fluctuation in flow. [ABSTRACT FROM AUTHOR]

Boiling—a process widely used for its good heat transferability—is limited by a phenomenon known as critical heat flux (CHF). Our experiments revealed a new CHF mechanism that is different from previously believed theories; we refer to it as "sonneting CHF." At CHF, the flow pattern changes from bubbly flow to slug/churn flow and then to an unusual reverse annular flow, leading to a significant rise in the heater surface temperature. The reverse annular flow, however, does not sustain but breaks down into a chaotic flow pattern, resulting in unprecedented quenching of the heater surface. The flow pattern shortly reverts back to bubbly flow; this entire process repeats for a few cycles, where the heater surface temperature rises and falls with amplitudes increasing in each cycle until the heater trips. The maximum removal-surface heat flux is significantly higher than the CHF. This new understanding will enable flexible and innovative boiling systems for several energy applications. [ABSTRACT FROM AUTHOR]

Flow boiling experiments were conducted in an annular channel at very low flow rates and atmospheric pressure varying the inlet subcooling from 20 to 40. critical heat flux occurred during the oscillatory flow pattern transition from slug to churn regime. Comparison of low-pressure and low-flow critical heat flux correlations with the present data has shown that these correlations are not applicable for prediction of critical heat flux under oscillatory flow pattern transition from slug to churn. Power transient experiments were also performed at low-pressure and low-flow conditions. Flow regimes and the sequence of events leading to critical heat flux were visually observed using the high-speed camera. [ABSTRACT FROM AUTHOR]

The visualization experiments for measuring departure diameter and frequency of vapor bubbles in the saturated nucleate boiling of distilled water have been carried out at atmospheric pressure. The experiments were carried out for an isolated single cylindrical tube and that for a tube bundle under pool and cross flow boiling conditions. The test setup was a stainless steel cuboidal tank of dimensions 400 mm ×  200 mm ×  180 mm provided with a 5  ×  3 staggered tube bundle with an equilateral triangular pitch of 1.95. The departure diameter and frequency were measured using a high-speed camera to understand the parametric behavior of the bubble departure characteristics in a tube bundle under pool and cross flow condition. The range of operating conditions was: (1) heat flux q ″ = 8 − 28 k W / m 2 , (2) wall superheat Δ T s a t = 7 − 22 K , (3) inlet mass flux G = 120 − 303.5 kg / m 2 s . The departure diameter and frequency were found to increase with an increase in the heat flux, but decrease with an increase in the mass flow rate. The local departure diameter and frequency were observed to increase as the neighboring tubes in the bundle were successively heated. Empirical correlations have been developed for departure diameter and frequency for saturated nucleate boiling under cross flow conditions incorporating the effect of the tube bundle. The correlations were found to predict experimental data within acceptable limit. [ABSTRACT FROM AUTHOR]

In subcooled boiling flows beyond a certain heat flux, heat transfer is hampered due to a phenomenon known as Departure from Nucleate Boiling (DNB). Conducting DNB experiments at one-to-one nuclear reactor operating conditions is highly challenging and expensive. Another alternative approach is to use Look-up table data. However, its applicability is limited due to its dependence on rod bundle correction factors. In the present investigation, a state-of-the-art Eulerian-Eulerian two-fluid model coupled with an extended heat flux partitioning model is used to predict DNB in tubes and rod bundles with square and hexagonal lattices (relevant to Pressurized Water Reactors). In this approach, bubble departure characteristics are modeled using semi-mechanistic models based on force balance analysis. The predicted DNB values are compared with experimental and Look-up table data and found out to be within 1.8% to 20%. [ABSTRACT FROM AUTHOR]