Heat Transfer, Thermal Hydraulic, and Safety Analysis

2016 ◽  
pp. 721-814
Author(s):  
Shripad T. Revankar
Keyword(s):  
Heat Transfer ◽  
Safety Analysis

scholarly journals Assessment of CHF and Post-CHF Heat Transfer Models for High-Pressure Condition

2021 ◽  
Vol 9 ◽  
Author(s):  
Meiqi Song ◽  
Xiaojing Liu
Keyword(s):  
Heat Transfer ◽  
High Pressure ◽  
Literature Review ◽  
Prediction Models ◽  
Safety Analysis ◽  
Pressure Condition ◽  
Predictive Capability ◽  
The World ◽  
Transfer Models

Supercritical heat transfer systems may undergo trans-critical procedures and work at subcritical conditions during startup, shutdown, or some accidents. However, well-validated heat transfer models for the high-pressure condition (P/Pc>0.7) are still missing. In the present work, with exhaustive literature review, extensive experimental databanks of CHF and post-dryout heat transfer under high-pressure condition are established, respectively. Existing prediction models for the high-pressure condition are also summarized from all over the world. Thereby, with the aid of the high-pressure experimental databank, prediction models get evaluated. It has been demonstrated that CHF correlation developed by Song et al. shows good predictive capability. Post-dryout heat transfer could get well predicted by the Song correlation. These recommended prediction models could be implemented to upgrade safety analysis codes for simulation of trans-critical transients.

Design of Condensation Heat Transfer Experiment to Evaluate Scaling Distortion in Small Modular Reactor Safety Analysis

2021 ◽  
Vol 7 (3) ◽  
Author(s):  
Palash K. Bhowmik ◽  
J. P. Schlegel ◽  
V. Kalra ◽  
C. Mills ◽  
S. Usman
Keyword(s):  
Heat Transfer ◽  
Test Data ◽  
Cooling System ◽  
Safety Analysis ◽  
Nuclear Industry ◽  
Condensation Heat Transfer ◽  
Saturated Steam ◽  
Test Facility ◽  
Working Fluid ◽  
Different Diameters

Abstract Designing a novel scaled modular test facility as a part of an experiment for condensation heat transfer (CHT) in small modular reactors (SMRs) is the main focus of this study. This facility will provide data to evaluate models' scalability for predicting heat transfer in the passive containment cooling system (PCCS) of SMR. The nuclear industry recognizes SMRs as future candidates for clean, economic, and safe energy generation. However, licensing requires proper evaluation of the safety systems such as PCCS. The knowledge gap from the literature review showed a lack of high-resolution experimental data for scaling of PCCS and validation of computational fluid dynamics tools. In addition, the presently available test data are inconsistent due to unscaled geometric and varying physics conditions. These inconsistencies lead to inadequate test data benchmarking. To fill this research gap, this study developed three scaled (different diameters) condensing test sections with annular cooling for scale testing and analysis. This facility considered saturated steam as the working fluid with noncondensable gases like nitrogen and helium in different mass fractions. This facility also used a precooler unit for inlet steam conditioning and a postcooler unit for condensate cooling. The high fidelity sensors, instruments, and data acquisition systems are installed and calibrated. Finally, facility safety analysis and shakedown tests are performed.

Development of New MHI LOCA Analysis Code: MCOBRA/RELAP5-GOTHIC

2016 ◽  
Author(s):  
Masaaki Katayama ◽  
Tetsuya Teramae ◽  
Masatsugu Mizokami ◽  
Shinya Kosaka
Keyword(s):  
Heat Transfer ◽  
Current Flow ◽  
Safety Analysis ◽  
Code System ◽  
Steam Generators ◽  
Code Model ◽  
Integral Effect ◽  
Counter Current ◽  
Relap5 Code ◽  
Containment Integrity

Mitsubishi Heavy Industries, Ltd. (MHI) has developed the safety analysis code system, MCOBRA/RELAP5-GOTHIC, for large break LOCA, small break LOCA and containment integrity analyses of PWR. The code system contains three codes, MCOBRA code, M-RELAP5 code and GOTHIC code. MHI implemented some models and correlations to MCOBRA and M-RELAP5 code to improve prediction of important phenomena of LOCA event. MHI conducted a lot of analyses for separate effect tests and integral effect tests to confirm the applicability of the MHI’s safety analysis code to LOCA key phenomena, like core heat transfer, DNB, entrainment, break flow, counter-current flow limitation (CCFL), core froth level, and so on. The code model uncertainties are quantified to calculate core heat transfer break flow rate, loop seal phenomena; heat transfer in steam generators, and so on. The results show that the code system can predict the key phenomena well and can calculate LOCA event scenario. Consequently, it is confirmed that the new code system is applicable to the LOCA licensing analysis.

The Development of Severe Accident Codes at IRSN and Their Application to Support the Safety Assessment of EPR

2006 ◽  
Author(s):  
Cataldo Caroli ◽  
Alexandre Bleyer ◽  
Ahmed Bentaib ◽  
Patrick Chatelard ◽  
Michel Cranga ◽  
...  
Keyword(s):  
Heat Transfer ◽  
Safety Assessment ◽  
Safety Analysis ◽  
Severe Accident ◽  
Navier Stokes ◽  
Complete Analysis ◽  
Primary System ◽  
Hydrogen Distribution ◽  
Pressurized Water ◽  
Fast Running

IRSN uses a two-tier approach for development of codes analysing the course of a hypothetical severe accident (SA) in a Pressurized Water Reactor (PWR): on one hand, the integral code ASTEC, jointly developed by IRSN and GRS, for fast-running and complete analysis of a sequence; on the other hand, detailed codes for best-estimate analysis of some phenomena such as ICARE/CATHARE, MC3D (for steam explosion), CROCO and TONUS. They have been extensively used to support the level 2 Probabilistic Safety Assessment of the 900 MWe PWR and, in general, for the safety analysis of the French PWR. In particular the codes ICARE/CATHARE, CROCO, MEDICIS (module of ASTEC) and TONUS are used to support the safety assessment of the European Pressurized Reactor (EPR). The ICARE/CATHARE code system has been developed for the detailed evaluation of SA consequences in a PWR primary system. It is composed of the coupling of the core degradation IRSN code ICARE2 and of the thermalhydraulics French code CATHARE2. The CFD code CROCO describes the corium flow in the spreading compartment. Heat transfer to the surrounding atmosphere and to the basemat, leading to the possible formation of an upper and lower crust, basemat ablation and gas sparging through the flow are modelled. CROCO has been validated against a wide experimental basis, including the CORINE, KATS and VULCANO programs. MEDICIS simulates MCCI (Molten-Corium-Concrete-Interaction) using a lumped-parameter approach. Its models are being continuously improved through the interpretation of most MCCI experiments (OECD-CCI, ACE…). The TONUS code has been developed by IRSN in collaboration with CEA for the analysis of the hydrogen risk (both distribution and combustion) in the reactor containment. The analyses carried out to support the EPR safety assessment are based on a CFD formulation. At this purpose a low-Mach number multi-component Navier-Stokes solver is used to analyse the hydrogen distribution. Presence of air, steam and hydrogen is considered as well as turbulence, condensation and heat transfer in the containment walls. Passive autocatalytic recombiners are also modelled. Hydrogen combustion is afterwards analysed solving the compressible Euler equations coupled with combustion models. Examples of on-going applications of these codes to the EPR safety analysis are presented to illustrate their potentialities.

Advanced Modeling of Multicomponent Vaporization/Condensation Phenomena for a Reactor Safety Analysis Code SIMMER-III

2002 ◽  
Author(s):  
Koji Morita ◽  
Tatsuya Matsumoto ◽  
Ryo Akasaka ◽  
Kenji Fukuda ◽  
Tohru Suzuki ◽  
...  
Keyword(s):  
Heat Transfer ◽  
Phase Transition ◽  
Reactor Core ◽  
Safety Analysis ◽  
Mass Diffusion ◽  
Short Time Scale ◽  
Reactor Safety ◽  
Noncondensable Gases ◽  
Diffusion Limited ◽  
Multicomponent Flows

It is believed that the numerical simulation of thermal-hydraulic phenomena of multiphase, multicomponent flows in a reactor core is essential to investigate core disruptive accidents (CDAs) of liquid-metal fast reactors. A new multicomponent vaporization/condensation (V/C) model was developed to provide a generalized model for a fast reactor safety analysis code SIMMER-III, which analyzes relatively short-time-scale phenomena relevant to accident sequences of CDAs. The model characterizes the V/C process associated with phase transition through heat-transfer and mass-diffusion limited models to follow the time evolution of the rector core under CDA conditions. The heat-transfer limited model describes the nonequilibrium phase-transition processes occurring at interfaces, while the mass-diffusion limited model is employed to represent effects of noncondensable gases and multicomponent mixture on V/C processes. Verification of the model and method employed in the multicomponent V/C model of SIMMER-III was performed successfully by analyzing two series of condensation experiments.

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