| 1. |
- Galushin, Sergey, et al.
(författare)
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Analysis of the effect of vessel failure and melt release on risk of containment failure due to ex-vessel steam explosion in nordic boiling water reactor using roaam1 framework
- 2020
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Ingår i: Journal of Nuclear Engineering and Radiation Science. - : ASME International. - 2332-8983 .- 2332-8975. ; 6:4
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Tidskriftsartikel (refereegranskat)abstract
- Nordic boiling water reactor (BWR) design employs ex-vessel debris coolability in a deep pool of water as a severe accident management (SAM) strategy. Depending on melt release conditions from the vessel and core-melt coolant interactions, containment integrity can be threatened by: (i) formation of noncoolable debris bed or (ii) energetic steam explosion. Melt is released from the vessel affect ex-vessel phenomena and is recognized as the major source of uncertainty. The risk-oriented accident analysis methodology (ROAAM ) is used for quantification of the risk of containment failure in Nordic BWR where melt ejection mode surrogate model (MEM SM) provides initial conditions for the analysis of debris agglomeration and ex-vessel steam explosion which determine the respective loads on the containment. Melt ejection SM is based on the system analysis code methods for estimation of leakages and consequences of releases (computer code) (MELCOR). Modeling of vessel failure and melt release from the vessel in MELCOR is based on parametric models, allowing a user to select different assumptions that effectively control lower head (LH) behavior and melt release. The work addresses the effect of epistemic uncertain parameters and modeling assumptions in MEM SM on the containment loads due to ex-vessel steam explosion in Nordic BWR. Sensitivity and uncertainty analysis performed to identify the most influential parameters and uncertainty in the risk of containment failure due to ex-vessel steam explosion. The results of the analysis provide valuable insights regarding the effect of MELCOR models, modeling parameters, and sensitivity coefficients on melt release conditions and predictions of ex-vessel steam explosion loads on the containment structures.
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| 2. |
- Gradecka, M., et al.
(författare)
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Computational fluid dynamics investigation of supercritical water flow and heat transfer in a rod bundle with grid spacers
- 2016
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Ingår i: Journal of Nuclear Engineering and Radiation Science. - : American Society of Mechanical Engineers (ASME). - 2332-8983 .- 2332-8975. ; 2:3
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Tidskriftsartikel (refereegranskat)abstract
- This paper presents a steady-state computational fluid dynamics approach to supercritical water flow and heat transfer in a rod bundle with grid spacers. The current model was developed using the ANSYS Workbench 15.0 software (CFX solver) and was first applied to supercritical water flow and heat transfer in circular tubes. The predicted wall temperature was in good agreement with the measured data. Next, a similar approach was used to investigate three-dimensional (3D) vertical upward flow of water at supercritical pressure of about 25 MPa in a rod bundle with grid spacers. This work aimed at understanding thermo- and hydrodynamic behavior of fluid flow in a complex geometry at specified boundary conditions. The modeled geometry consisted of a 1.5-m heated section in the rod bundle, a 0.2-m nonheated inlet section, and five grid spacers. The computational mesh was prepared using two cell types. The sections of the rods with spacers were meshed using tetrahedral cells due to the complex geometry of the spacer, whereas sections without spacers were meshed with hexahedral cells resulting in a total of 28 million cells. Three different sets of experimental conditions were investigated in this study: a nonheated case and two heated cases. The nonheated case, A1, is calculated to extract the pressure drop across the rod bundle. For cases B1 and B2, a heat flux is applied on the surface of the rods causing a rise in fluid temperature along the bundle. While the temperature of the fluid increases along with the flow, heat deterioration effects can be present near the heated surface. Outputs from both B cases are temperatures at preselected locations on the rods surfaces.
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| 3. |
- Rohde, M., et al.
(författare)
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A blind, numerical benchmark study on supercritical water heat transfer experiments in a 7-rod bundle
- 2016
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Ingår i: Journal of Nuclear Engineering and Radiation Science. - : American Society of Mechanical Engineers (ASME). - 2332-8983 .- 2332-8975. ; 2:2
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Tidskriftsartikel (refereegranskat)abstract
- Heat transfer in supercritical water reactors (SCWRs) shows a complex behavior, especially when the temperatures of the water are near the pseudocritical value. For example, a significant deterioration of heat transfer may occur, resulting in unacceptably high cladding temperatures. The underlying physics and thermodynamics behind this behavior are not well understood yet. To assist the worldwide development in SCWRs, it is therefore of paramount importance to assess the limits and capabilities of currently available models, despite the fact that most of these models were not meant to describe supercritical heat transfer (SCHT). For this reason, the Gen-IV International Forum initiated the present blind, numerical benchmark, primarily aiming to show the predictive ability of currently available models when applied to a real-life application with flow conditions that resemble those of an SCWR. This paper describes the outcomes of ten independent numerical investigations and their comparison with wall temperatures measured at different positions in a 7-rod bundle with spacer grids in a supercritical water test facility at JAEA. The wall temperatures were not known beforehand to guarantee the blindness of the study. A number of models have been used, ranging from a one-dimensional (1-D) analytical approach with heat transfer correlations to a RANS simulation with the SST turbulence model on a mesh consisting of 62 million cells. None of the numerical simulations accurately predicted the wall temperature for the test case in which deterioration of heat transfer occurred. Furthermore, the predictive capabilities of the subchannel analysis were found to be comparable to those of more laborious approaches. It has been concluded that predictions of SCHT in rod bundles with the help of currently available numerical tools and models should be treated with caution.
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