Experimental and numerical investigations of natural circulation phenomena in passive safety systems for decay heat removal in large pools

2010 ◽  
Vol 240 (10) ◽  
pp. 3170-3177 ◽  
Author(s):  
Eckhard Krepper ◽  
Matthias Beyer
Keyword(s):  
Natural Circulation ◽  
Heat Removal ◽  
Passive Safety ◽  
Numerical Investigations ◽  
Decay Heat ◽  
Passive Safety Systems ◽  
Safety Systems

Investigation of Thermohydraulic Models for Condensation and Boiling in Passive Safety Systems

2017 ◽  
Author(s):  
Frances Viereckl ◽  
R. Manthey ◽  
C. Schuster ◽  
A. Hurtado
Keyword(s):  
Heat Transfer ◽  
Natural Circulation ◽  
Heat Removal ◽  
Operating Conditions ◽  
Test Facility ◽  
Passive Safety ◽  
Condensation Process ◽  
Passive Safety Systems ◽  
Safety Systems ◽  
The Heat Transfer Coefficient

Passive safety systems represent one field of research concerning the safety-related enhancement of nuclear power plants. Passive safety systems can ensure the safe removal of decay heat without an input of electrical or mechanical energy for commissioning or operation. The heat removal chain is guaranteed on the basis of the physical principles condensation, heat conduction, boiling and natural circulation. The thermal hydraulic processes in passive safety systems disagree with the plant-specific thermal hydraulics because of different operating conditions. Since the established system codes are validated for the plant-specific conditions, the operational behavior of passive safety systems is currently not sufficiently predictable. On this account, the German Federal Ministry of Education and Research initiated the joint project PANAS to investigate the decay heat removal by passive safety systems on the basis of experimental analyses, modelling and validation. Object is the heat removal chain in advanced boiling water reactors consisting of emergency condensers (EC; heat transfer from reactor core to core flooding pools) and containment cooling condensers (CCC; heat transfer from the containment to the shielding/storage pool). At Technische Universität Dresden, the test facility GENEVA was constructed for the experimental investigation of the operational behavior of the CCC. GENEVA models the CCC concerning the original thermal hydraulic conditions of the heat source and heat sink as well as the tube geometry for the heat transfer. In this way, the comparability of the thermal hydraulic phenomena is given. Previous experiments focused on the stability analysis of the natural circulation in the test facility. The focus of PANAS is on the condensation process of saturated steam at the outside of the slightly inclined tubes and the convection respectively boiling of both a stable and an unstable two-phase flow inside these tubes. For a detailed analysis, condensation rates at the outside as well as the flow structure inside have to be investigated experimentally. Therefore, the instrumentation in the heat transfer section of GENEVA is considerably enhanced. This enhancement comprises an optical measuring system for the film thickness or droplet size of the condensate, a tipping scale for the condensate mass flow, void probes for the steam void fraction and more than 100 thermocouples outside and inside the tubes for temperature profiles in axial, radial and azimuthal direction. By reference to these parameters, it is possible to examine the thermal hydraulic models for the heat transfer. The paper outlines the available models in system codes regarding condensation and boiling concerning the operating conditions of the CCC. Since dropwise condensation could be observed in previous experiments and the condensation models in system codes focus on film condensation, the review is extended beyond native models. A sensitivity analysis of the reviewed models regarding condensation shows huge differences concerning the value of the heat transfer coefficient. Furthermore, the courses of the condensation models present different dependencies regarding the heat transfer coefficient and the wall temperature. Due to this, the necessity of the experimental investigation and later the revision of the condensation models in system codes is confirmed. The comparison of the reviewed models with first experimental results outlines the tendency for the numerical description of the condensation process. Based on the investigation and validation of models concerning the heat transfer processes in the CCC, the operational behavior will be accurately predictable by established system codes, which enhances the safety investigation and the licensing. Although the conception of this investigation is founded on the CCC, the adapted models will be able to characterize the heat transfer processes boiling and condensation for saturation conditions at a relatively low pressure (maximum 4 bar) and for natural convection in general.

Study of Decay Heat Removal Approach for Advanced Passive PWR During Station Blackout

2014 ◽  
Author(s):  
Jie Zou ◽  
Lili Tong ◽  
Xuewu Cao
Keyword(s):  
Cooling System ◽  
Heat Removal ◽  
Safety System ◽  
Injection System ◽  
Passive Safety ◽  
System Failure ◽  
Decay Heat ◽  
Passive Safety Systems ◽  
Station Blackout ◽  
Safety Systems

After Fukushima accident, decay heat removal in station blackout (SBO) accident is concerned for different NPP design. Advanced passive PWR relies on passive systems to cool reactor core and containment, such as the passive residual heat removal system (PRHR), passive injection system and passive containment cooling system (PCCS). Passive safety systems are considered more reliable than traditional active safety system under accident condition. However, in long-term SBO situation, possible failure of passive safety systems is noticed as active valves are needed in system actuation. Moreover, probability safety analysis results of advanced passive PWR show that system failure is possible without external event. Given different passive safety system failure assumptions, response of reactor coolant system and containment of advanced passive PWR is calculated in SBO accident, the integrity of core, reactor pressure vessel and containment is assessed, and decay heat removal approach is studied. The results show that containment failure is predicted with the failure of PCCS and PRHR, reactor vessel failure together with containment failure is predicted with the failure of PCCS, passive injection system and PRHR. Advices to deal with the risk of advanced passive PWR in SBO are given based on the study.

Passive Safety Systems of Advanced Nuclear Power Plant: AP1000

2010 ◽  
Author(s):  
Guohua Yan ◽  
Chen Ye
Keyword(s):  
Nuclear Power ◽  
Human Error ◽  
Cooling System ◽  
Heat Removal ◽  
Injection System ◽  
Passive Safety ◽  
Human Errors ◽  
Power Sources ◽  
Passive Safety Systems ◽  
Safety Systems

In the entire history of commercial nuclear power so far, only two major accidents leading to damage of reactor core have taken place. One is Three Mile Island (TMT) accident (1979), which is caused by a series of human error, and the other is Chernobyl accident (1986), which is due to the combined reason of design defects and human errors. After TMI and Chernobyl accidents, in order to reduce manpower in operation and maintenance and influence of human errors on reactor safety, consideration is given to utilization of passive safety systems. According to the IAEA definition, passive safety systems are based on natural forces, such as convection and gravity, and stored energy, making safety functions less dependent on active systems and operators’ action. Recently, the technology of passive safety has been adopted in many reactor designs, such as AP1000, developed by Westinghouse and EP1000 developed by European vendor, and so on. AP1000 as the first so-called Generation III+ has received the final design approval from US NRC in September 2004, and now being under construction in Sanmen, China. In this paper, the major passive safety systems of AP1000, including passive safety injection system, automatic depressurization system passive residual heat removal system and passive containment cooling system, are described and their responses to a break loss-of-coolant accident (LOCA) are given. Just due to these passive systems’ adoption, the nuclear plant can be able to require no operator action and offsite or onsite AC power sources for at least 72h when one accident occurs, and the core melt and large release frequencies are significantly below the requirement of operating plants and the NRC safety goals.

Modeling and Experiments of a Passive Decay Heat Removal System for Advanced Nuclear Reactors

2017 ◽  
Author(s):  
Andrea Bersano ◽  
Mario De Salve ◽  
Cristina Bertani ◽  
Nicolò Falcone ◽  
Bruno Panella
Keyword(s):  
Scaling Laws ◽  
Nuclear Reactors ◽  
Natural Circulation ◽  
Heat Removal ◽  
Experimental Results ◽  
Passive Systems ◽  
Decay Heat ◽  
Safety Systems ◽  
Heat Removal System ◽  
Removal System

Within the field of research and development of innovative nuclear reactors, in particular Generation IV reactors and Small Modular Reactors (SMR), the design and the improvement of safety systems play a crucial role. Among all the safety systems high attention is dedicated to passive systems that do not need external energy to operate, with a very high reliability also in the case of station blackout, and which are largely used in evolutionary technology reactors. The aim of this work is the experimental and numerical analysis of a passive system that operates in natural circulation in order to study the mechanism and the efficiency of heat removal. The final goal is the development of a methodology that can be used to study this class of systems and to assess the thermal-hydraulic code RELAP5 for these specific applications. Starting from a commercial size system, which is the decay heat removal system of the experimental lead cooled reactor ALFRED, an experimental facility has been designed, built and tested with the aim of studying natural circulation in passive systems for nuclear applications. The facility has been simulated and optimized using the thermal-hydraulic code RELAP5-3D. During the experimental tests, temperatures and pressures are measured and the experimental results are compared with the ones predicted by the code. The results show that the system operates effectively, removing the given thermal power. The code can predict well the experimental results but high attention must be dedicated to the modeling of components where non-condensable gases are present (condenser pool and surrounding ambient). This facility will be also used to validate the scaling laws among systems that operate in natural circulation.

Experimental Investigation of Non-Condensable Gases Effect on Novovoronezh NPP-2 Steam Generator Condensation Power Under the Condition of Passive Safety Systems Operation

2009 ◽  
Author(s):  
I. I. Kopytov ◽  
S. G. Kalyakin ◽  
V. M. Berkovich ◽  
A. V. Morozov ◽  
O. V. Remizov
Keyword(s):  
Steam Generator ◽  
Flow Rate ◽  
Computational Simulation ◽  
Heat Removal ◽  
Test Facility ◽  
Passive Safety ◽  
Primary Circuit ◽  
Water Radiolysis ◽  
Passive Safety Systems ◽  
Safety Systems

The design substantiation of the heat removal efficiency from Novovoronezh NPP-2 (NPP-2006 project with VVER-1200 reactor) reactor core in the event of primary circuit leaks and operation of passive safety systems only (among these are the systems of hydroaccumulators of the 1st and 2nd stages and passive heat removal system) has been performed based on computational simulation of the related processes in the reactor and containment. The computational simulation has been performed with regard to the detrimental effect of non-condensable gases on steam generator (SG) condensation power. Nitrogen arriving at the circuit with the actuation of hydroaccumulators of the 1st stage and products of water radiolysis are the main sources of non-condensable gases in the primary circuit. The feature of Novovoronezh NPP-2 passive safety systems operation is that during the course of emptying of the 2nd stage hydroaccumulators system (HA-2) the gas-steam mixture spontaneously flows out from SG cold headers into the volume of HA-2 tanks. The flow rate of gas-steam mixture during the operation of HA-2 system is equal to the volumetric water discharge from hydroaccumulators. The calculations carried out by different integral thermal hydraulic codes revealed that this volume flow rate of gas-steam mixture from SG to the HA-2 system would suffice to eliminate the “poisoning” of SG piping and to maintain necessary condensation power. In support of the calculation results, the experiments were carried out at the HA2M-SG test facility constructed at IPPE. The test facility incorporates a VVER steam generator model of volumetric-power scale of 1:46. Steam to the HA2M-SG test facility is supplied fed from the IPPE heat power plant. Gas addition to steam coming to the SG model is added from high pressure gas cylinders. Nitrogen and helium are used in the experiments for simulating hydrogen. The report presents the basic results of experimental investigations aimed at the evaluation of SG condensation power under the inflow of gas-steam mix with different gases concentration to the tube bundle, both under the simulation of gas-steam mixture outflow from SG cold header to the HA-2 system and without outflow. As a result of the research performed at the HA2M-SG test facility, it has been substantiated experimentally that in the event of an emergency leak steam generators have condensation power sufficient for effective heat removal from the reactor provided by PHR system.

Experimental and Numerical Investigations of Temperature Stratification Phenomena in Passive Safety Systems for Decay Heat Removal

2009 ◽  
Author(s):  
Eckhard Krepper ◽  
Matthias Beyer
Keyword(s):  
Nuclear Power ◽  
Natural Circulation ◽  
Heat Removal ◽  
Temperature Stratification ◽  
Small Scale ◽  
Passive Systems ◽  
Nuclear Power Reactor ◽  
Cfd Simulations ◽  
Decay Heat ◽  
Passive Safety Systems

Modern concepts of nuclear power reactor systems are equipped with passive systems for decay heat removal. Examples are the pool of the emergency condenser (BWR-1000) or the pool of the ESBWR. These systems operate without active influence from outside. The questions arise: How reliable are the based physical mechanisms? Are they understood completely? Are actual models able to describe the phenomena? In different passive systems the energy is transferred by natural circulation into large pools which are considered as infinite heat sink. The paper deals with experiments and with CFD simulations to investigate the capability of actual CFD codes to describe these phenomena. In the FZ Dresden-Rossendorf at the facility TOPFLOW heating-up tests of an emergency condenser were performed. During these tests also the temperature courses on the secondary side of the pool were recorded. The data recording comprises periods starting from single phase liquid until steam on the secondary pool side was found. During these experiments temperature stratification phenomena were observed, which were found in earlier small scale tests. In the paper also these small scale experiments are described. A detailed CFD analysis of these experiments was performed. An explanation of the observed phenomena on the basis of the small scale tests and the CFD simulations is presented.

Loss-of-Flow-Accidents (LOFA) Study for 100 MW IPWR

2014 ◽  
Author(s):  
Shasha Yin ◽  
Liang Gao ◽  
Wenxi Tian ◽  
Yapei Zhang ◽  
Suizheng Qiu ◽  
...  
Keyword(s):  
Steady State ◽  
Heat Removal ◽  
System Model ◽  
Passive Safety ◽  
Loop Model ◽  
Primary Coolant ◽  
Passive Safety Systems ◽  
Safety Systems ◽  
Coolant System ◽  
Residual Heat

The inherent system safety of the 100 MW integral pressurized water reactor (IPWR) can be improved by placing the core, the efficient once-through steam generators and the main coolant pumps in the reactor pressure vessel, and omitting some large pipes and valves in the primary coolant system which can prevent the occurrence of large break loss of coolant accident and reduce the possibility of core melt accident. The application of the passive safety systems simplifies the structures of IPWR and improves the economy of the reactor. In case of accidents, the primary coolant system establishes natural circulation to take the core decay heat away by passive safety systems using gravity and other natural driving forces, thereby enhancing the safety and reliability of the system IPWR. It’s of great significance to establish reasonable and correctable models, including the primary coolant system model, the second loop model and passive core cooling system model, to study thermal-hydraulic phenomena under steady state, transient state and accident conditions. Based on transient safety analysis program RELAP5/MOD3.4, 100 MW IPWR system was simulated. A series of models of reactor coolant system and passive safety systems were established. The main system models are composed of primary coolant system model, part of second loop model, passive safety injection system model and passive residual heat removal system model. The primary coolant system model includes core, lower plenum, downcomer, region of steam generators, upper plenum, riser, pressurizer, and surge line; the second loop model includes the main feed water line, the steam line, and steam generator tubes; passive safety injection system model includes core makeup tank, accumulator, automatic depressurization system, direct vessel injection line; and passive residual heat removal system model includes passive residual heat removal heat exchanger in containment refueling water storage tank. Based on the established models, the steady state was debugged with the RELAP5 input card. Steady state calculation was performed and the results agree well with designed values, which verifies the validity of the model and the input card. Using the steady state results as initial conditions, transient calculation was performed. Typical accidents (loss of main water accident) were calculated, which were relieved by auxiliary feedwater system (AFWS) and passive residual heat removal system (PRHR SYSTEM). The results, obtained from AFWS and PRHR SYSTEM, were contrasted and process of accident and thermal-hydraulic phenomena were analyzed according to transient calculation results. The transient calculation results showed that the integral PWR system and the passive safety system model can provide a reference for IPWR transient safety analysis.

SIRIO: An Experimental Facility for a New Heat Removal System Passively Controlled by Non-Condensable Gases

2018 ◽  
Author(s):  
R. Marinari ◽  
M. Tarantino ◽  
F. S. Nitti ◽  
A. Alemberti ◽  
M. Caramello ◽  
...  
Keyword(s):  
Heat Removal ◽  
Passive Safety ◽  
Primary System ◽  
Experimental Facility ◽  
The Past ◽  
Passive Safety Systems ◽  
Design Activities ◽  
Safety Systems ◽  
Heat Removal System ◽  
Removal System

Heat removal systems are of major importance for both present and future nuclear power plants as they belong to the set of systems devoted to ensure the integrity of the reactor core and to avoid core damage. The past experience and lessons learned on this topic suggest to adopt passive safety systems which can perform the safety function independently from operators’ actions and external energy sources, ensuring long term reactor cooling. Application of these systems to innovative reactor concepts such as (heavy) liquid metal reactors poses a problem related to the characteristic properties of the coolant: as the final heat sink of passive safety systems is often the external environment, the liquid metal will eventually undergo a phase change and solidify at the end of a complex dynamic process. The solidification of the coolant may have important effects on the transient behavior if it happens at an early stage of an accident, as the main flow path of the heat exchanger can be blocked by the coolant freezing while the decay heat in the core is still sufficiently high and need to be efficiently removed. An innovative passive safety system has been proposed for the decay heat removal system of ALFRED reactor (DEMO LFR, Gen.IV) where the issue of early coolant freezing is prevented. The innovation has been object of a patent and the system is potentially able to avoid solidification by reducing the amount of heat removed from the primary system by means of non-condensable gases passively injected into the water/steam mixture, which induce heat transfer degradation. Several numerical studies have been performed during the past years, but a complete validation of the operating principle requires an experimental assessment and characterization. To this aim the SIRIO experimental facility, scaled on the DHR of ALFRED, has been conceived. Several design activities have been performed so far for the development of the facility, such as scaling analysis on the basis of ALFRED DHR to determine the facility size, numerical simulations by means of RELAP5-3D to determine whether the facility is able to reproduce the expected physical phenomena and numerical simulations by means of Ansys CFX to investigate the performance of a heating system simulating the primary system of ALFRED based on a molten salt annulus. The present paper describes the design activities performed and provides insights on the methodologies adopted, as well as the current status of the design of the SIRIO facility.

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