External Hazard Coinciding With Small Break LOCA—Thermohydraulic Calculation With System Code ATHLET

2020 ◽  
Vol 6 (2) ◽  
Author(s):  
Kai Kosowski ◽  
Marcus Seidl
Keyword(s):  
Diesel Engines ◽  
Nuclear Power ◽  
Heat Removal ◽  
Primary Circuit ◽  
External Event ◽  
Design Basis ◽  
Loss Of Coolant Accident ◽  
License Application ◽  
Thermohydraulic Calculation ◽  
Core Cooling

Abstract The safety behaviors of a nuclear power plant (NPP) after an external hazard-initiated event, as well as after a small break (SB) loss of coolant accident (LOCA), are already well known as part of the analyses made for standard license application. The coincidence of both events leads to a beyond-design basis consideration. Such a combination of both event categories is investigated by means of the thermohydraulic system code ATHLET. The scenario assumes an external event with a LOCA caused by induced vibrations on a small pipe attached to the primary circuit, although all pipes are designed to withstand the loads created by such an external event. Furthermore, in the context of both robustness and enveloping analyses, both a loss of offsite power (LOOP) and an unavailability of the emergency diesel power supply are postulated. The NPP in the scenario considered only has access to the passive accumulators and to systems supplied via the safeguard emergency diesel engines (second quartet of emergency diesel engines), which are housed in the bunkered emergency feed building. The dedicated type of external event itself is not in focus, but rather the thermohydraulic behavior of the NPP is considered. Apart from the model's assumptions, the accident sequence is explained in detail. The remaining systems for emergency core cooling are capable of handling the LOCA under such demanding boundary conditions. Long-term cooling can be ensured. Furthermore, heat removal out of the core is always sufficient. Eventually, all safety protection objectives have been complied for this beyond-design basis scenario.

External Hazard Coinciding With Small Break LOCA Thermohydraulic Calculation With System Code ATHLET

2018 ◽  
Author(s):  
Kai Kosowski ◽  
Marcus Seidl
Keyword(s):  
Boundary Conditions ◽  
Diesel Engines ◽  
Power Supply ◽  
Heat Removal ◽  
Design Feature ◽  
Primary Circuit ◽  
External Event ◽  
Grid Connection ◽  
Design Basis ◽  
Loss Of Coolant Accident

The safety behaviors of a NPP after an external hazard-initiated event, as well as after a small break LOCA are already well known as part of the analyses done for standard licensing. In this paper, a combination of both event categories is investigated by means of the thermohydraulic system code ATHLET (analysis of the thermohydraulics of leaks and transients) developed by the Gesellschaft für Anlagen- und Reaktorsicherheit (GRS, Germany). The scenario assumes an external event with a loss of coolant accident caused by induced vibrations on a small pipe attached to the primary circuit, although all pipes are designed to withstand the loads created by such an external event. The coincidence of both events leads to a beyond-design basis consideration. Furthermore, in the context of both robustness and enveloping analyses, the unavailability of both the internal power supply via the external grid connection and the emergency diesel power supply due to the external impact is postulated. For emergency cooling, the NPP in the scenario considered has only access to the passive accumulators and to systems supplied via the safeguard emergency diesel engines (2nd quartet of emergency diesel engines) — a design feature of the Siemens KWU type PWRs —, which are housed in the bunkered emergency feed building and operate with lower voltage levels. The scenario is exemplarily modeled and simulated with ATHLET. The manner of the external event itself is not in focus, but rather the thermohydraulic behavior of the NPP is considered in the reported analysis. Beside the model assumptions and the boundary conditions, the accident sequence is explained in detail. It turns out that the remaining systems for emergency cooling are able to handle the LOCA under such demanding boundary conditions. Most importantly, the primary pressure can be lowered below the zero flow pump head of the pool cooling pumps. The long-term cooling can thereby be ensured. Furthermore, the heat removal out of the core is always sufficient. Overall it can be concluded that all safety protection objectives have been complied for this beyond-design basis scenario.

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.

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.

Modeling of Thermal Hydraulics Aspects of Combined Top and Bottom Water Reflood Experiment PARAMETER-SF2 Using SOCRAT 2.1 Code

2008 ◽  
Author(s):  
Alexander D. Vasiliev
Keyword(s):  
Experimental Data ◽  
Flow Rate ◽  
Nuclear Power ◽  
Cooling System ◽  
Thermal Hydraulics ◽  
Two Phase ◽  
Vver Fuel ◽  
Design Basis ◽  
Loss Of Coolant Accident ◽  
Core Cooling

The PARAMETER-SF2 test conditions simulated a severe LOCA (Loss of Coolant Accident) nuclear power plant sequence in which the overheated up to 1700÷2300K core would be reflooded from the top and the bottom in occasion of ECCS (Emergency Core Cooling System) recovery. The test was successfully conducted at the NPO “LUTCH”, Podolsk, Russia, in April 3, 2007 and was the second of two experiments to be performed in the frame of ISTC 3194 Project. PARAMETER facility of NPO “LUTCH”, Podolsk, is designed for studies of the VVER fuel assemblies behavior under conditions simulating design basis, beyond design basis and severe accidents. After the maximum cladding temperature of 1750K was reached in the bundle during PARAMETER-SF2 test, the top flooding (flow rate 40g/s) was begun and later approximately in 30 s the bottom flooding (flow rate 100g/s) was initiated. Two-phase (water and steam) flow determined the fuel assembly cooling conditions. The thermal hydraulic and SFD (Severe Fuel Damage) best estimate numerical complex SOCRAT 2.1 was used for the calculation of PARAMETER-SF2 experiment. Thermal hydraulics in PARAMETER-SF2 experiment played very important role and its adequate modeling is important for the thermal analysis. The results obtained by the complex SOCRAT 2.1 were compared with experimental data concerning different aspects of thermal hydraulics behavior including convective and radiative heat transfer in the bundle and the CCFL (counter-current flooding limitation) phenomenon during the reflood. The temperature experimental data were found to be in a good agreement with calculated results. It is indicative of the adequacy of modeling the complicated thermo-hydraulic behavior in the PARAMETER-SF2 test.

Radiation Dose Evaluation of Typical Design Basis Accident for Advanced PWR in China

2021 ◽  
Author(s):  
Haiying Chen ◽  
Shaowei Wang ◽  
Xinlu Tian ◽  
Fudong Liu
Keyword(s):  
Radiation Dose ◽  
Effective Dose ◽  
Half Life ◽  
Nuclear Power ◽  
Outer Boundary ◽  
Design Basis ◽  
Loss Of Coolant Accident ◽  
Design Basis Accident ◽  
Typical Design ◽  
Total Effective Dose

Abstract The loss of coolant accident (LOCA) is one of the typical design basis accidents for nuclear power plant. Radionuclides leak to the environment and cause harm to the public in LOCA. Accurate evaluation of radioactivity and radiation dose in accident is crucial. The radioactivity and radiation dose model in LOCA were established, and used to analyze the radiological consequence at exclusion area boundary (EAB) and the outer boundary of low population zone (LPZ) for Hualong 1. The results indicated that the long half-life nuclides, such as 131I, 133I, 135I, 85Kr, 131mXe, 133mXe and 133Xe, released to environment continuously, while the short half-life nuclides, such as 132I, 134I, 83mKr, 85mKr, 87Kr, 88Kr, 135mXe and 138Xe, no longer released to environment after a few hours in LOCA. 133Xe may release the largest radioactivity to environment, more than 1015Bq. Inhalation dose was the major contribution to the total effective dose. The total effective dose and thyroid dose of Hualong 1 at EAB and the outer boundary of LPZ fully met the requirements of Chinese GB6249.

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 Thermal Hydraulic Analysis of a Nuclear Reactor due to Loss of Coolant Accident with and without Emergency Core Cooling System

2020 ◽  
Vol 01 (02) ◽  
pp. 53-60
Author(s):  
Pronob Deb Nath ◽  
Kazi Mostafijur Rahman ◽  
Md. Abdullah Al Bari
Keyword(s):  
Nuclear Reactor ◽  
Cooling System ◽  
Computational Study ◽  
Reactor Core ◽  
Primary Circuit ◽  
Feed Water ◽  
Pressurized Water ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Core Cooling

This paper evaluates the thermal hydraulic behavior of a pressurized water reactor (PWR) when subjected to the event of Loss of Coolant Accident (LOCA) in any channel surrounding the core. The accidental break in a nuclear reactor may occur to circulation pipe in the main coolant system in a form of small fracture or equivalent double-ended rupture of largest pipe connected to primary circuit line resulting potential threat to other systems, causing pressure difference between internal parts, unwanted core shut down, explosion and radioactivity release into environment. In this computational study, LOCA for generation III+ VVER-1200 reactor has been carried out for arbitrary break at cold leg section with and without Emergency Core Cooling System (ECCS). PCTRAN, a thermal hydraulic model-based software developed using real data and computational approach incorporating reactor physics and control system was employed in this study. The software enables to test the consequences related to reactor core operations by monitoring different operating variables in the system control bar. Two types of analysis were performed -500% area break at cold leg pipe due to small break LOCA caused by malfunction of the system with and without availability of ECCS. Thermal hydraulic parameters like, coolant dynamics, heat transfer, reactor pressure, critical heat flux, temperature distribution in different sections of reactor core have also been investigated in the simulation. The flow in the reactor cooling system, steam generators steam with feed-water flow, coolant steam flow through leak level of water in different section, power distribution in core and turbine were plotted to analyze their behavior during the operations. The simulation showed that, LOCA with unavailability of Emergency Core Cooling System (ECCS) resulted in core meltdown and release of radioactivity after a specific time.

Modeling of Thermal Hydraulics Features of Top Water Reflood Experiment PARAMETER-SF3 Using SOCRAT/V2 Code

2009 ◽  
Author(s):  
Arcadii E. Kisselev ◽  
Valerii F. Strizhov ◽  
Alexander D. Vasiliev ◽  
Vladimir I. Nalivayev ◽  
Nikolay Ya. Parshin
Keyword(s):  
Experimental Data ◽  
Nuclear Power ◽  
Cooling System ◽  
Severe Accident ◽  
Numerical Code ◽  
Thermal Hydraulics ◽  
Vver Fuel ◽  
Design Basis ◽  
Best Estimate ◽  
Core Cooling

The PARAMETER-SF3 test conditions simulated a severe LOCA (Loss of Coolant Accident) nuclear power plant sequence in which the overheated up to 1700÷2300K core would be reflooded from the top and the bottom in occasion of ECCS (Emergency Core Cooling System) recovery. The test was successfully conducted at the NPO “LUTCH”, Podolsk, Russia, in October 31, 2008, and was the third of four experiments of series PARAMETER-SF. PARAMETER facility of NPO “LUTCH”, Podolsk, is designed for studies of the VVER fuel assemblies behavior under conditions simulating design basis, beyond design basis and severe accidents. The test bundle was made up of 19 fuel rod simulators with a length of approximately 3.12 m (heated rod simulators) and 2.92 m (unheated rod simulator). Heating was carried out electrically using 4-mm-diameter tantalum heating elements installed in the center of the rods and surrounded by annular UO2 pellets. The rod cladding was identical to that used in VVERs: Zr1%Nb, 9.13 mm outside diameter, 0.7 mm wall thickness. After the maximum cladding temperature of about 1900K was reached in the bundle during PARAMETER-SF3 test, the top flooding was initiated. The thermal hydraulic and SFD (Severe Fuel Damage) best estimate numerical complex SOCRAT/V2 was used for the calculation of PARAMETER-SF3 experiment. The counter-current flow limitation (CCFL) model was implemented to best estimate numerical code SOCRAT/V2 developed for modeling thermal hydraulics and severe accident phenomena in a reactor. Thermal hydraulics in PARAMETER-SF3 experiment played very important role and its adequate modeling is important for the thermal analysis. The results obtained by the complex SOCRAT/V2 were compared with experimental data concerning different aspects of thermal hydraulics behavior including the CCFL phenomenon during the reflood. The temperature experimental data were found to be in a good agreement with calculated results. It is indicative of the adequacy of modeling the complicated thermo-hydraulic behavior in the PARAMETER-SF3 test.

Test and Evaluation of a Filtering System for Retention of Fibres in a Coolant Flow After Loss-of-Coolant Accident

2012 ◽  
Author(s):  
T. Gocht ◽  
W. Kästner ◽  
A. Kratzsch ◽  
M. Strasser
Keyword(s):  
Cooling System ◽  
Reactor Core ◽  
Heat Removal ◽  
Test Facility ◽  
Experimental Investigations ◽  
Time Interval ◽  
Insulation Material ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Core Cooling

In case of an accident the safe heat removal from the reactor core with the installed emergency core cooling system (ECCS) is one of the main features in reactor safety. During a loss-of-coolant accident (LOCA) the release of insulation material fragments in the reactor containment can lead to malfunctions of ECCS. Therefore, the retention of particles by strainers or filtering systems in the ECCS is one of the major tasks. The aim of the presented experimental investigations was the evaluation of a filtering system for the retention of fiber-shaped particles in a fluid flow. The filtering system consists of a filter case with a special lamellar filter unit. The tests were carried out at a test facility with filtering units of different mesh sizes. Insulation material (mineral rock wool) was fragmented to fiber-shaped particles. To simulate the distribution of particle concentration at real plants with large volumes the material was divided into single portions and introduced into the loop with a defined time interval. Material was transported to the filter by the fluid and agglomerated there. The assessment of functionality of the filtering system was made by differential pressure between inlet and outlet of the filtering system and by mass of penetrated particles. It can be concluded that for the tested filtering system no penetration of insulation particles occurred.

Application of Thermal Hydraulic Code SOCRAT/V2 to Top Water Reflood Experiment PARAMETER-SF3

2009 ◽  
Author(s):  
Alexander D. Vasiliev
Keyword(s):  
Experimental Data ◽  
Nuclear Power ◽  
Cooling System ◽  
Severe Accident ◽  
Numerical Code ◽  
Thermal Hydraulics ◽  
Vver Fuel ◽  
Design Basis ◽  
Best Estimate ◽  
Core Cooling

The PARAMETER-SF3 test conditions simulated a severe LOCA (Loss of Coolant Accident) nuclear power plant sequence in which the overheated up to 1700–2300K core would be reflooded from the top and the bottom in occasion of ECCS (Emergency Core Cooling System) recovery. The test was successfully conducted at the NPO “LUTCH”, Podolsk, Russia, in October 31, 2008, and was the third of four experiments of series PARAMETER-SF. PARAMETER facility of NPO “LUTCH”, Podolsk, is designed for studies of the VVER fuel assemblies behavior under conditions simulating design basis, beyond design basis and severe accidents. The test bundle was made up of 19 fuel rod simulators with a length of approximately 3.12 m (heated rod simulators) and 2.92 m (unheated rod simulator). Heating was carried out electrically using 4-mm-diameter tantalum heating elements installed in the center of the rods and surrounded by annular UO2 pellets. The rod cladding was identical to that used in VVERs: Zr1%Nb, 9.13 mm outside diameter, 0.7 mm wall thickness. After the maximum cladding temperature of about 1900K was reached in the bundle during PARAMETER-SF3 test, the top flooding was initiated. The thermal hydraulic and SFD (Severe Fuel Damage) best estimate numerical complex SOCRAT/V2 was used for the calculation of PARAMETER-SF3 experiment. The counter-current flow limitation (CCFL) model was implemented to best estimate numerical code SOCRAT/V2 developed for modeling thermal hydraulics and severe accident phenomena in a reactor. Thermal hydraulics in PARAMETER-SF3 experiment played very important role and its adequate modeling is important for the thermal analysis. The results obtained by the complex SOCRAT/V2 were compared with experimental data concerning different aspects of thermal hydraulics behavior including the CCFL phenomenon during the reflood. The temperature experimental data were found to be in a good agreement with calculated results. It is indicative of the adequacy of modeling the complicated thermo-hydraulic behavior in the PARAMETER-SF3 test.

Sign in / Sign up

Export Citation Format

Share Document