Development of Passive Cooling System Models for the Assessment of the Long Term Cooling Capability in the Liquid Metal Reactor

2004 ◽  
Author(s):  
W. P. Chang ◽  
K. S. Ha ◽  
H. Y. Jeong ◽  
S. Heo ◽  
Y. B. Lee
Keyword(s):  
Cooling System ◽  
Heat Removal ◽  
Computer Code ◽  
Passive Safety ◽  
Primary Concern ◽  
Metallic Fuel ◽  
Passive Safety Systems ◽  
Operator Action ◽  
Safety Systems

This study has been carried out to assess the decay heat removal capability of the passive safety systems adopted in a conceptual design of the 600 MW(e), sodium cooled, metallic fuel loaded KALIMER. The applicability of the PVCS, which used to be the only passive safety system for KALIMER-150, is limited to a reactor capacity of 1,000 MW(t) or less. Another passive loop, PDRC, is conceptualized in order to overcome the limit as the KALIMER capacity scales up from the current 150 MW(e) to 600 MW(e). The safety analysis computer code, SSC-K, currently used for KALIMER is not capable of simulating such passive systems. With this concern, the PVCS and PDRC models are developed and they are coupled with the SSC-K for a long-term cooling assessment. The present paper thus presents the analysis results of the ULOHS using these models along with their brief introductions. The primary concern of the analyses is focused on the inherent safety as well as the system’s integrity for 72 hours without any operator action during the event.

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.

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.

iB1350: A Generation III.7 Reactor After the Fukushima Daiichi Accident

2016 ◽  
Author(s):  
Takashi Sato ◽  
Keiji Matsumoto ◽  
Kenji Hosomi ◽  
Keisuke Taguchi
Keyword(s):  
Cooling System ◽  
Boiling Water ◽  
Passive Safety ◽  
Active Safety ◽  
Water Reactor ◽  
Airplane Crash ◽  
Passive Safety Systems ◽  
Active Safety Systems ◽  
Fukushima Daiichi ◽  
Safety Systems

iB1350 stands for an innovative, intelligent and inexpensive boiling water reactor 1350. It is the first Generation III.7 reactor after the Fukushima Daiichi accident. It has incorporated lessons learned from the Fukushima Daiichi accident and Western European Nuclear Regulation Association safety objectives. It has innovative safety to cope with devastating natural disasters including a giant earthquake, a large tsunami and a monster hurricane. The iB1350 can survive passively such devastation and a very prolonged station blackout without any support from the outside of a site up to 7 days even preventing core melt. It, however, is based on the well-established proven Advance Boiling Water Reactor (ABWR) design. The nuclear steam supply system is exactly the same as that of the current ABWR. As for safety design it has a double cylinder reinforced concrete containment vessel (Mark W containment) and an in-depth hybrid safety system (IDHS). The Mark W containment has double fission product confinement barriers and the in-containment filtered venting system (IFVS) that enable passively no emergency evacuation outside the immediate vicinity of the plant for a severe accident (SA). It has a large volume to hold hydrogen, a core catcher, a passive flooding system and an innovative passive containment cooling system (iPCCS) establishing passively practical elimination of containment failure even in a long term. The IDHS consists of 4 division active safety systems for a design basis accident, 2 division active safety systems for a SA and built-in passive safety systems (BiPSS) consisting of an isolation condenser (IC) and the iPCCS for a SA. The IC/PCCS pools have enough capacity for 7-day grace period. The IC/PCCS heat exchangers, core and spent fuel pool are enclosed inside the containment vessel (CV) building and protected against a large airplane crash. The iB1350 can survive a large airplane crash only by the CV building and the built-in passive safety systems therein. The dome of the CV building consists of a single wall made of steel and concrete composite. This single dome structure facilitates a short-term construction period and cost saving. The CV diameter is smaller than that of most PWR resulting in a smaller R/B. Each active safety division includes only one emergency core cooling system (ECCS) pump and one emergency diesel generator (EDG). Therefore, a single failure of the EDG never causes multiple failures of ECCS pumps in a safety division. The iB1350 is based on the proven ABWR technology and ready for construction. No new technology is incorporated but design concept and philosophy are initiative and innovative.

AP1000® Passive Cooling Containment Analysis of a Double-Ended LBLOCA With a 3D Gothic Model

2018 ◽  
Author(s):  
Samanta Estevez-Albuja ◽  
Gonzalo Jimenez ◽  
Kevin Fernández-Cosials ◽  
César Queral ◽  
Zuriñe Goñi
Keyword(s):  
Cooling System ◽  
Heat Removal ◽  
Water Tank ◽  
Local Pressure ◽  
Water Storage Tank ◽  
Passive Safety Systems ◽  
Containment Model ◽  
Loss Of Coolant Accident ◽  
Ac Power ◽  
Safety Systems

In order to enhance Generation II reactors safety, Generation III+ reactors have adopted passive mechanisms for their safety systems. In particular, the AP1000® reactor uses these mechanisms to evacuate heat from the containment by means of the Passive Containment Cooling System (PCS). The PCS uses the environment atmosphere as the ultimate heat sink without the need of AC power to work properly during normal or accidental conditions. To evaluate its performance, the AP1000 PCS has been usually modeled with a Lumped Parameters (LP) approach, coupled with another LP model of the steel containment vessel to simulate the accidental sequences within the containment building. However, a 3D simulation, feasible and motivated by the current computational capabilities, may be able to produce more detailed and accurate results. In this paper, the development and verification of an integral AP1000® 3D GOTHIC containment model, taking into account the shield building, is briefly presented. The model includes all compartments inside the metallic containment liner and the external shield building. Passive safety systems, such as the In-containment Refueling Water Storage Tank (IRWST) with the Passive Residual Heat Removal (PRHR) heat exchanger and the Automatic Depressurization System (ADS), as well as the PCS, are included in the model. The model is tested against a cold leg Double Ended Guillotine Break Large Break Loss of Coolant Accident (DEGB LBLOCA) sequence, taking as a conservative assumption that the PCS water tank is not available during the sequence. The results show a pressure and temperature increase in the containment in consonance with the current literature, but providing a greater detail of the local pressure and temperature in all compartments.

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.

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.

Preliminary Safety Analysis of a Compact Small Reactor: Feedwater Line Break and Small-Break LOCA Accident Analysis

2017 ◽  
Author(s):  
Linsen Li ◽  
Feng Shen ◽  
Mian Xing ◽  
Zhan Liu ◽  
Zhanfei Qi
Keyword(s):  
Conceptual Design ◽  
Cooling System ◽  
Passive Safety ◽  
Accident Analysis ◽  
Primary System ◽  
Primary Coolant ◽  
Pressurized Water ◽  
Passive Safety Systems ◽  
Core Cooling ◽  
Safety Systems

A small Pressurized Water Reactor (PWR) with compact primary system and passive safety feature, which is called Compact Small Reactor (CSR), is under pre-conceptual design and development. For the purpose of preliminary assessment of the primary coolant system and capability evaluation of the passive safety system, a detailed thermal-hydraulic (T-H) system model of the CSR was developed. Several design-basis accidents, including feedwater line break, double ended direct vessel injection line break (one of the small-break Loss Of Coolant Accidents, LOCA) and etc, are selected and simulated so as to evaluate and further optimize the design of passive safety systems, especially the passive core cooling system. The results of preliminary accident analysis show that the passive safety systems are basically capable of mitigating the accidents and protecting the reactor core from severe damage. Further research will be focused on the optimization of pre-conceptual design of the thermal-hydraulic system and the passive core cooling system.

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.

Sign in / Sign up

Export Citation Format

Share Document