Sensitivity Analysis of the Thermal Diffusion Coefficient Effect on the Departure From Nucleate Boiling Ratio With the VIPRE Code

[+] Author and Article Information
Rosario Delgado-Tardáguila, Marisol Corisco, Antonio Espejo, Daniel Navarro

ENUSA Industrias,
Avanzadas S.A., c/Santiago Rusiñol, 12,
Madrid, 28040, Spain

Javier Riverola

ENUSA Industrias,
Avanzadas S.A., c/Santiago Rusiñol, 12,
Madrid, 28040, Spain
e-mail: rdt@enusa.es

Manuscript received July 27, 2018; final manuscript received December 18, 2018; published online March 15, 2019. Assoc. Editor: Ignacio Gómez.

ASME J of Nuclear Rad Sci 5(2), 020909 (Mar 15, 2019) (7 pages) Paper No: NERS-18-1057; doi: 10.1115/1.4042358 History: Received July 27, 2018; Revised December 18, 2018

One of the limiting conditions during operation of a pressurized water reactor (PWR) is cladding integrity in class I (normal operations) or class II (most frequent). Cladding integrity is limited typically by the departure from the nucleate boiling (DNB), which criterion ensures an appropriate core cooling. Adequate heat transfer between the fuel cladding and reactor coolant is achieved by preventing DNB that is avoided if the local heat flux is lower than the critical heat flux (CHF). The DNB is estimated thanks to thermal-hydraulic (TH) design codes, as the VIPRE-W code that predicts the fluid behavior based on the geometry of the problem, the fuel rods and the fluid properties among others. One of the parameters that influences the DNB estimation is the thermal diffusion coefficient (TDC), which depends on the fuel design and is affected by the grid spacing. As a matter of fact, the TDC enters into the DNB calculation for thermal mixing between subchannels and in some special cases like the most primitive fuel designs, as a factor within the DNB correlation. Nevertheless, although the TDC is a variable, the TH design codes used for the DNB prediction consider the TDC as a constant. This investigation is founded on a new numerical program developed to explore the effect of the TDC on the DNB. In addition to this, variables as the effect of the turbulent momentum factor (FTM) and the correlation effect has been explored too. The most direct outcome of this research is the substantial extension of the existing studies of VIPRE-W TH code. The results show that TDC has an effect on the DNB dominated by the radial power distribution. The departure from nucleate boiling ratio (DNBR) increases up to 1.2% when TDC is a variable under normal operation radial shapes. For the design radial distribution, this effect is vanished but observable for values under 0.02 with an exponential increase of the DNBR with respect to the TDC. From this moment on, the energy exchanged between subchannels is negligible due to the flatness shape of the radial enthalpy distribution.

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Fig. 1

Fluid elements in analysis within VIPRE [5]

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Fig. 2

Forced convective flow regimes for vertical fuel rod within coolant flow shroud (with cosine power profile) [7]

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Fig. 3

Modified advance European fuel assembly (ENUSA)

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Fig. 4

The DNBR as a function of the FTM

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Fig. 5

The DNBR as a function of the TDC. First stage: design radial power distribution.

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Fig. 6

Real radial power distribution in a 17 × 17 fuel assembly

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Fig. 7

Thermal diffusion coefficient effect on the DNBR for radial distributions. Second stage: normal operation radial power distribution.



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