An Integrated Approach for Monitoring Gas Accumulation in Safety Related Systems

2009 ◽  
Author(s):  
L. Ike Ezekoye ◽  
Ryan D. Griffin ◽  
William M. Turkowski ◽  
Gregory R. Williams
Keyword(s):  
Nuclear Power ◽  
Heat Removal ◽  
Integrated Approach ◽  
Water Hammer ◽  
Relief Valve ◽  
Pump Design ◽  
Gas Accumulation ◽  
Decay Heat ◽  
Core Cooling ◽  
The U.S

Gas intrusion into safety related systems, such as the Emergency Core Cooling System (ECCS), Decay Heat Removal (DHR) and Containment Spray (CS) system in nuclear power plants is undesirable and has led to pump binding and damaging water hammer events. Furthermore, total or momentary loss of hydraulic performance in safety related pumps has occurred, which has led to pump damage rendering the pumps unable to perform their safety functions or reduce the pump discharge pressure and flow capacity to the point that the system cannot perform its design function. Extreme cases of water hammer can result in physical damage to system piping, components and supports, and possible relief valve lifting events with consequential loss of inventory. The U.S. Nuclear Regulatory Commission (NRC) issued Generic Letter GL-2008-01, “Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal, and Containment Spray Systems,” requiring U.S. utilities to demonstrate that suitable design, operation and testing measures are in place to maintain licensing commitments. GL-2008-01 outlines a number of actions that are detailed in nature, such as establishing pump void tolerance limits, establishing limits on pump suction void fractions, assuring adequate system venting capability, identifying all possible sources of gas intrusion, preventing vortex formation in tanks and determining acceptable limits of gas in system discharge piping. Regarding one of these issues, GL-2008-01 indicates that the amount of gas that can be ingested without significant impact on pump operability and reliability is not well established and is known to depend on pump design, gas dispersion and flow rate. Each U.S. nuclear power plant licensee is required to evaluate their ECCS, DHR and CS system design, operation and test procedures to assure that gas intrusion is minimized and monitored in order to maintain system operability and compliance with the requirements of the U.S. code of federal regulations 10 CFR 50 Appendix B. Typically, gases get into the safety related systems through a number of mechanisms, such as maintenance, gas desorption, vortex activities, cavitation, etc. This paper discusses the sources of gas into safety related systems and the challenges associated with management of gas voids in these systems. A number of technologies exist that can detect the gas that accumulates in the safety related piping. These technologies are discussed and an integrated approach for monitoring gas accumulation in safety related pipes is presented. Issues such as methods to get rid of gases and venting periodicity are discussed. Industry efforts to address the management of gases in these systems are also presented.

A Simple and Reliable System for Detecting and Measuring Gas Voids in Safety-Related Fluid Systems

2009 ◽  
Author(s):  
Herb Estrada ◽  
Ernest M. Hauser
Keyword(s):  
Nuclear Power ◽  
Detection System ◽  
Void Formation ◽  
Heat Removal ◽  
Critical Systems ◽  
Multiple Systems ◽  
Gas Accumulation ◽  
Void Detection ◽  
Fluid Systems ◽  
Core Cooling

Over the past 20 years there have a number of instances where nuclear power plant operators have discovered gas voids—typically air but occasionally other gases such as undissolved hydrogen—in fluid systems whose function is important to reactor safety. These systems have included emergency core cooling systems, decay heat removal systems, and containment spray systems. The amount of gas has, in some cases, been sufficient to call into question the operability of the systems; had they been needed. The automatic initiation of a system with a gas void present may lead to gas binding of its pumps, or destructive water hammer. The sources of the gas have been various and not readily controlled. The need for licensees to manage gas accumulation has been formally identified in an NRC generic letter (1). The letter points out a need for continuous monitoring, to detect and quantify gas voids in these systems, thereby to ensure their availability in accordance with design basis requirements. The letter further notes that periodic functional tests of the critical systems will not provide the required assurance of operability; if a periodic test finds a system’s functionality questionable because of gas accumulation, the question of how long its operability has been compromised is unanswered. The system described in this paper—the Linewatch Gas Void Detection System—addresses these issues definitively. It provides the means to detect the onset of void formation in any one of multiple pipes in multiple systems on a continuous basis and, following void formation, the means to quantify the amount of these voids, again continuously.

Fatigue Evaluation for Thermally Stratified DVI Piping

2012 ◽  
Author(s):  
Hwan Ho Lee ◽  
Joon Ho Lee ◽  
Dong Jae Lee ◽  
Seok Hwan Hur ◽  
Il Kwun Nam ◽  
...  
Keyword(s):  
Temperature Distribution ◽  
Thermal Stratification ◽  
Nuclear Power ◽  
Heat Removal ◽  
Local Stress ◽  
Piping System ◽  
Hydraulic Analysis ◽  
Fatigue Evaluation ◽  
Core Cooling ◽  
Thermal Hydraulic Analysis

A numerical analysis has been performed to estimate the effect of thermal stratification in the safety injection piping system. The Direct Vessel Injection (DVI) system is used to perform the functions of Emergency Core Cooling and Residual Heat Removal for an APR1400 nuclear power plant (Korea’s Advanced Power Reactor 1400 MW-Class). The thermal stratification is anticipated in the horizontally routed piping between the DVI nozzle of the reactor vessel and the first isolation valve. Non-axisymmetric temperature distribution across the pipe diameter induced by the thermal stratification leads to differential thermal growth of the piping causing the global bending stress and local stress. Thermal hydraulic analysis has been performed to determine the temperature distribution in the DVI piping due to the thermal stratification. Piping stress analysis has also been carried out to evaluate the integrity of the DVI piping using the thermal hydraulic analysis results. This paper provides a methodology for calculating the global bending stresses and local stresses induced by the thermal stratification in the DVI piping and for performing fatigue evaluation based on Subsection NB-3600 of ASME Section III.

scholarly journals Effect of Coolant Inventories and Parallel Loop Interconnections on the Natural Circulation in Various Heat Transport Systems of a Nuclear Power Plant during Station Blackout

2008 ◽  
Vol 2008 ◽  
pp. 1-11 ◽  
Author(s):  
Avinash J. Gaikwad ◽  
P. K. Vijayan ◽  
Sharad Bhartya ◽  
Kannan Iyer ◽  
Rajesh Kumar ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Heat Transport ◽  
Nuclear Power ◽  
Natural Circulation ◽  
Heat Removal ◽  
Transport Systems ◽  
Critical Examination ◽  
Decay Heat ◽  
Station Blackout

Provision of passive means to reactor core decay heat removal enhances the nuclear power plant (NPP) safety and availability. In the earlier Indian pressurised heavy water reactors (IPHWRs), like the 220 MWe and the 540 MWe, crash cooldown from the steam generators (SGs) is resorted to mitigate consequences of station blackout (SBO). In the 700 MWe PHWR currently being designed an additional passive decay heat removal (PDHR) system is also incorporated to condense the steam generated in the boilers during a SBO. The sustainability of natural circulation in the various heat transport systems (i.e., primary heat transport (PHT), SGs, and PDHRs) under station blackout depends on the corresponding system's coolant inventories and the coolant circuit configurations (i.e., parallel paths and interconnections). On the primary side, the interconnection between the two primary loops plays an important role to sustain the natural circulation heat removal. On the secondary side, the steam lines interconnections and the initial inventory in the SGs prior to cooldown, that is, hooking up of the PDHRs are very important. This paper attempts to open up discussions on the concept and the core issues associated with passive systems which can provide continued heat sink during such accident scenarios. The discussions would include the criteria for design, and performance of such concepts already implemented and proposes schemes to be implemented in the proposed 700 MWe IPHWR. The designer feedbacks generated, and critical examination of performance analysis results for the added passive system to the existing generation II & III reactors will help ascertaining that these safety systems/inventories in fact perform in sustaining decay heat removal and augmenting safety.

The Effect of IRWST Simulation Method on the Analysis of Small Break LOCA Using RELAP5

2017 ◽  
Author(s):  
Caihong Xu ◽  
Guobao Shi ◽  
Kemei Cao ◽  
Xiaoyu Cai ◽  
Zhanfei Qi
Keyword(s):  
Nuclear Power ◽  
Heat Removal ◽  
Simulation Method ◽  
Horizontal Direction ◽  
Water Storage Tank ◽  
Initiation Phase ◽  
Injection Flow ◽  
Loss Of Coolant Accident ◽  
Perfect Mixing ◽  
Core Cooling

The In-containment Refueling Water Storage Tank (IRWST) provides low-pressure safety injection flow for passive CAP1400 Nuclear Power Plant (NPP) during Loss-Of-Coolant-Accident (LOCA) and subsequent Long Term Core Cooling (LTCC). The Passive Residual Heat Removal Heat Exchanger (PRHR HX) and the spargers of Automatic Depressurization System (ADS) stage 1∼3 are submerged in the IRWST. During small break LOCA, heat and mass are delivered through PRHR HX and ADS spargers to IRWST, and IRWST is heated up before its safety injection. However, numerical and experimental investigation has shown that IRWST is not perfect mixing, and thermal stratification exists. During ADS-4/IRWST initiation phase, the temperature of IRWST injection flow is of great importance, and is affected greatly by IRWST simulation method when modeling with system code like RELAP5. In this paper, two different IRWST simulation methods where one use multi channels in horizontal direction while the other use only one, are analyzed for CAP1400 SBLOCA with RE-LAP5, and their effects are compared. Finally, the better method which uses only one channel in horizontal direction is recommended.

Setup of the Supercritical CO2 Test Facility “SCARLETT” for Basic Experimental Investigations of a Compact Heat Exchanger for an Innovative Decay Heat Removal System

2018 ◽  
Vol 4 (3) ◽  
Author(s):  
Wolfgang Flaig ◽  
Rainer Mertz ◽  
Joerg Starflinger
Keyword(s):  
Heat Exchanger ◽  
Supercritical Co2 ◽  
Nuclear Power ◽  
Power Plants ◽  
Heat Removal ◽  
Test Facility ◽  
Experimental Investigations ◽  
Decay Heat ◽  
Heat Removal System ◽  
Removal System

Supercritical fluids show great potential as future coolants for nuclear reactors, thermal power, and solar power plants. Compared to the subcritical condition, supercritical fluids show advantages in heat transfer due to thermodynamic properties near the critical point. A specific field of interest is an innovative decay heat removal system for nuclear power plants, which is based on a turbine-compressor system with supercritical CO2 as the working fluid. In case of a severe accident, this system converts the decay heat into excess electricity and low-temperature waste heat, which can be emitted to the ambient air. To guarantee the retrofitting of this decay heat removal system into existing nuclear power plants, the heat exchanger (HE) needs to be as compact and efficient as possible. Therefore, a diffusion-bonded plate heat exchanger (DBHE) with mini channels was developed and manufactured. This DBHE was tested to gain data of the transferable heat power and the pressure loss. A multipurpose facility has been built at Institut für Kernenergetik und Energiesysteme (IKE) for various experimental investigations on supercritical CO2, which is in operation now. It consists of a closed loop where the CO2 is compressed to supercritical state and delivered to a test section in which the experiments are run. The test facility is designed to carry out experimental investigations with CO2 mass flows up to 0.111 kg/s, pressures up to 12 MPa, and temperatures up to 150 °C. This paper describes the development and setup of the facility as well as the first experimental investigation.

In-Calandria Retention of Corium in PHWR: Experimental Investigation and Remaining Issues

2017 ◽  
Vol 3 (2) ◽  
Author(s):  
Sumit V. Prasad ◽  
A. K. Nayak
Keyword(s):  
Experimental Investigation ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Heat Removal ◽  
Public Domain ◽  
Severe Accident ◽  
Fukushima Accident ◽  
Decay Heat ◽  
Critical Issues

After the Fukushima accident, the public has expressed concern regarding the safety of nuclear power plants. This accident has strengthened the necessity for further improvement of safety in the design of existing and future nuclear power plants. Pressurized heavy water reactors (PHWRs) have a high level of defense-in-depth (DiD) philosophy to achieve the safety goal. It is necessary for designers to demonstrate the capability of decay heat removal and integrity of containment in a PHWR reactor for prolonged station blackout to avoid any release of radioactivity in public domain. As the design of PHWRs is distinct, its calandria vessel (CV) and vault cooling water offer passive heat sinks for such accident scenarios and submerged calandria vessel offers inherent in-calandria retention (ICR) features. Study shows that, in case of severe accident in PHWR, ICR is the only option to contain the corium inside the calandria vessel by cooling it from outside using the calandria vault water to avoid the release of radioactivity to public domain. There are critical issues on ICR of corium that have to be resolved for successful demonstration of ICR strategy and regulatory acceptance. This paper tries to investigate some of the critical issues of ICR of corium. The present study focuses on experimental investigation of the coolability of molten corium with and without simulated decay heat and thermal behavior of calandria vessel performed in scaled facilities of an Indian PHWR.

Conceptual Design and Feasibility Study for FAil-Safe Simple Economical SMR (FASES) System

2014 ◽  
Author(s):  
Ho Sik Kim ◽  
Hee Cheon No
Keyword(s):  
Conceptual Design ◽  
Feasibility Study ◽  
Cooling System ◽  
Heat Removal ◽  
Operating Conditions ◽  
Design Concepts ◽  
Decay Heat ◽  
Design Requirements ◽  
Core Cooling ◽  
Fail Safe

Because of high marketability of SMR, although many countries are trying to develop SMR, the SMR market is not formed yet. For early dominance of SMR market, the SMR system should be fail-safe, simple and economical. In order to develop FAil-safe Simple Economical SMR (FASES) system we applied the design characteristics of HTGRs into water-cooled reactors. In this study, we performed conceptual design and feasibility study for the FASES system. The feasibility study is focused on a thermal-hydraulic aspect in normal operating conditions and several accident conditions. The key design characteristic of the FASES system obtained from design concepts is to have sufficient decay heat removal capability even in the accidents involving extended station blackout (SBO), failure of passive decay heat removal system (PDHRS), and failure of emergency core cooling system (ECCS). Based on the design concepts, we could define several thermal-hydraulic design requirements. Then, we performed thermal-hydraulic analysis for feasibility study and proposed the specific design of the FASES system satisfying several design requirements.

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.

The Analysis of AP1000 Beyond Design Basis DEDVI Accident

2013 ◽  
Author(s):  
Sheng Zhu
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Cooling System ◽  
Heat Removal ◽  
The Core ◽  
Design Basis ◽  
Injection Line ◽  
Design Basis Accident ◽  
Core Cooling

Double ended break of direct vessel injection line (DEDVI) is the most typical small-break lost of coolant accident (LOCA) in AP 1000 nuclear power plant. This study simulated the DEDVI (without actuation of automatic depressurization system 1–3 stage valves, accumulators and passive residual heat removal heat exchanger) beyond design basis accident (BDBA) to validate the safety capability of AP1000 under such conditions. The results show that the core will be uncovered for about 863 seconds and then recovered by water after gravity injection from IRWST into the pressure vessel. The peak cladding temperature (PCT) goes up to 838.08°C, much lower than the limiting value 1204°C. This study confirms that in the DEDVI beyond design basis accident, the passive core cooling system (PXS) can effectually cool the core and preserve it integrate, and ensure the safety of AP 1000 nuclear power plant.

Sign in / Sign up

Export Citation Format

Share Document