ASTEC application to in-vessel corium retention

Tarabelli, D.; Ratel, G.; Pélisson, R.; Guillard, G.; Barnák, M.; Matejovič, Peter

doi:10.1016/j.nucengdes.2009.02.021

Cited by 12 publications

(3 citation statements)

References 8 publications

Supporting

Mentioning

Contrasting

Order By: Relevance

“…In most cases, the numerical evaluation of in-vessel retention through external reactor vessel cooling was carried out by using integral codes for severe accident analyses. Tarabelli et al applied the ASTEC code to simulate the in-vessel retention with the improved modeling for thermal behavior of the core melt in the lower head [15]. The in-vessel module to describe the corium behavior (DIVA) and the ex-vessel module to model the external reactor vessel cooling (CESAR) were coupled in this study.…”

Section: Introductionmentioning

confidence: 99%

Numerical Evaluation of Coolability Limits of External Reactor Vessel Cooling Using an Improved Thermal-Hydraulic System Analysis Code

Lee

Park

2023

International Journal of Energy Research

View full text Add to dashboard Cite

The external reactor vessel cooling provides a means to stabilize the molten core inside the reactor vessel during the severe accident in nuclear power plants. For a credible evaluation of the chances of success of in-vessel melt retention, one requires high-fidelity predictions on the natural circulation flow around the reactor vessel and, in particular, the corresponding critical heat flux (CHF) on a hemispherical lower head. In this study, the coolability limit of external reactor vessel cooling in the Korean APR1400 reactor was investigated numerically by improving a thermal-hydraulic system analysis code, MARS-KS1.5. Since such system analysis codes did not incorporate a CHF model for a downward-facing hemisphere, three CHF correlations were newly implemented in the MARS-KS code, two of them taking into account the effect of the mass flux. A transient analysis on the CHF and surface heat flux on the lower head was carried out while the surrounding cooling water was heated to create the two-phase natural circulation flow. The heat transfer mode on the heated surface was used as a measure of evaluating the success of the strategy, and the simulation results were compared with those obtained via the default CHF model in the code. The sensitivity analysis on the heat load from the core melt was also performed.

show abstract

Section: Introductionmentioning

confidence: 99%

Numerical Evaluation of Coolability Limits of External Reactor Vessel Cooling Using an Improved Thermal-Hydraulic System Analysis Code

Lee

Park

2023

International Journal of Energy Research

View full text Add to dashboard Cite

show abstract

“…It was found that in the cases of large break loss of coolant accident (LBLOCA) and station black out, water ingression and steam venting through the insulator are critical factors for the effective cooling via boiling heat removal at the outer surface of the reactor pressure vessel. To simulate the in-vessel retention of VVER-440/V213, model improvements and adaptations of the ASTEC (Accident Source Term Evaluation Code) were employed by the CEA (French Alternative Energies and Atomic Energy Commission -Commissariatà l'énergie atomique et auxénergies alternatives) in France [2]. The external cooling of the reactor wall was simulated by applying imposed coolant temperature and heat transfer coefficient, and the obtained results were found to be in good agreement with available predictions from other codes, particularly the shape and depth of ablation and the maximum heat flux in case of a thick metallic layer.…”

Section: Introductionmentioning

confidence: 99%

Numerical study of in-vessel retention under the gallium–water external reactor vessel cooling system using MARS-LMR

Kang

Park

Kim

et al. 2015

Journal of Nuclear Science and Technology

View full text Add to dashboard Cite

To confirm the feasibility of the gallium-water IVR-ERVCS (in-vessel retention-external reactor vessel cooling system), this paper focuses on the numerical simulation of severe accidents in APR 1400 using MARS-LMR (multidimensional analysis of reactor safety-liquid metal reactor). To analyze the galliumcooled systems, the properties of liquid gallium were added to the MARS-LMR code used in our previous work. In this system, the generated decay heat is transferred to liquid gallium through the reactor pressure vessel and then removed from the water pool as a heat sink. The numerical analyses results show that the temperature range of the liquid gallium is much lower than its boiling point and confirm the natural convection. Sensitivity studies were also performed by changing several parameters such as the initial temperature of gallium and water pool inventory and their results indicated that the working time of the gallium-water IVR-ERVCS depends on the inventory of the water pool. Because liquid gallium in this system does not have a phase change, unlike water, the gallium-water IVR-ERVCS can provide stable and reliable cooling capability. To solve the limitation due to critical heat flux in IVR-ERVCS and to ensure the sufficient thermal margin, it is confirmed that the gallium-water IVR-ERVCS can be a successful severe accident mitigation strategy in nuclear power plants.

show abstract

“…Although the conclusion on failure criteria was only related to the thermal margin, the methodology and data of this project were applied to design an in-vessel retention (IVR) management scheme of VVER plants [4]. ASTEC code [5] and IVRASA code [6] were adopted for the IVR simulation related to the CHF on the outer wall of the reactor vessel. Enhancement of the CHF estimation for additional thermal margin in the IVR-ERVC strategy was carried out through two-dimensional curved test section experiments [7] and the thermal load was compared with the maximum heat removal rate on the outer wall [8].…”

Section: Introductionmentioning

confidence: 99%

Structural assessment of reactor pressure vessel under multi-layered corium formation conditions

Kim

Chang

2015

Nuclear Engineering and Technology

View full text Add to dashboard Cite

a b s t r a c tExternal reactor vessel cooling (ERVC) for in-vessel retention (IVR) has been considered one of the most useful strategies to mitigate severe accidents. However, reliability of this common idea is weakened because many studies were focused on critical heat flux whereas there were diverse uncertainties in structural behaviors as well as thermalehydraulic phenomena. In the present study, several key factors related to molten corium behaviors and thermal characteristics were examined under multi-layered corium formation conditions. Thereafter, systematic finite element analyses and subsequent damage evaluation with varying parameters were performed on a representative reactor pressure vessel (RPV) to figure out the possibility of high temperature induced failures. From the sensitivity analyses, it was proven that the reactor cavity should be flooded up to the top of the metal layer at least for successful accomplishment of the IVR-ERVC strategy. The thermal flux due to corium formation and the relocation time were also identified as crucial parameters. Moreover, three-layered corium formation conditions led to higher maximum von Mises stress values and consequently shorter creep rupture times as well as higher damage factors of the RPV than those obtained from two-layered conditions.

show abstract

ASTEC application to in-vessel corium retention

Cited by 12 publications

References 8 publications

Numerical Evaluation of Coolability Limits of External Reactor Vessel Cooling Using an Improved Thermal-Hydraulic System Analysis Code

Numerical Evaluation of Coolability Limits of External Reactor Vessel Cooling Using an Improved Thermal-Hydraulic System Analysis Code

Numerical study of in-vessel retention under the gallium–water external reactor vessel cooling system using MARS-LMR

Structural assessment of reactor pressure vessel under multi-layered corium formation conditions

Contact Info

Product

Resources

About