Reactor Core Cooling Performance of a Passive Endothermic Reaction Cooling System During Design and Non-Design Basis Accidents

2018 ◽  
Author(s):  
Nathan R. Murray ◽  
Mitchell E. Sailsbery ◽  
Samuel E. Bischoff ◽  
Paul R. Wilding ◽  
Matthew J. Memmott
Keyword(s):  
Cooling System ◽  
Reactor Core ◽  
Passive Safety ◽  
Coolant Temperature ◽  
Outlet Temperature ◽  
Endothermic Reaction ◽  
Plant Safety ◽  
Water Reactors ◽  
Core Cooling ◽  
High Level

A passive endothermic reaction cooling system (PERCS) is proposed to provide reactor core cooling during a station blackout (SBO). During a SBO, a PWR in which PERCS has been installed has a peak reactor core outlet temperature remains below 640 K (692.3°F) for 30 days, which is well below the nominal accident core outlet temperature during a SBO. During a LOCA, LOFA, and LOHSA, installation of a PERCS has no significant impact on safety performance. It should be noted that the PERCS will represent a minimal heat source (unless the PERCS is very large) during DBAs as emergency systems lower the coolant temperature below the PERCS temperature. A typical PWR with an installed PERCS is modeled using RELAP5-3D. The results of the model demonstrate the high level of passive safety afforded by the PERCS which contributes to the mitigation of SBO consequences without adversely affecting nuclear plant safety during a LOCA, LOHSA, or LOFA. Future work in validating the PERCS as a method of passive safety for existing light water reactors is underway, including the refining the physical design, determining better kinetic and thermodynamic properties for MgCO3, updating the PERCS model, and using a more robust PWR plant model.

Improvement of Temperature Evaluation Model of Biological Shielding Concrete for HTTR Test Simulating LOFC With VCS Inactive

2013 ◽  
Author(s):  
Shoji Takada ◽  
Shunki Yanagi ◽  
Kazuhiko Iigaki ◽  
Masanori Shinohara ◽  
Daisuke Tochio ◽  
...  
Keyword(s):  
Heat Transfer ◽  
Natural Convection ◽  
Thermal Radiation ◽  
Cooling System ◽  
Evaluation Model ◽  
Reactor Core ◽  
Reactor Power ◽  
Coolant Temperature ◽  
Water Cooling ◽  
Reactor Outlet

HTTR is a helium gas cooled graphite-moderated HTGR with the rated power 30 MWt and the maximum reactor outlet coolant temperature 950°C. The vessel cooling system (VCS), which is composed of thermal reflector plates, cooling panel composed of fins connected between adjacent water cooling tubes, removes decay heat from reactor core by heat transfer of thermal radiation, conduction and natural convection in case of loss of forced cooling (LOFC). The metallic supports are embedded in the biological shielding concrete to support the fins of VCS. To verify the inherent safety features of HTGR, the LOFC test is planned by using HTTR with the VCS inactive from an initial reactor power of 9 MWt under the condition of LOFC while the reactor shut-down system disabled. In this test, the temperature distribution in the biological shielding concrete is prospected locally higher around the support because of thermal conduction in the support. A 2-dimensional symmetrical model was improved to simulate the heat transfer to the concrete through the VCS support in addition to the heat transfer thermal radiation and natural convection. The model simulated the water cooling tubes setting horizontally at the same pitch with actual configuration. The numerical results were verified in comparison with the measured data acquired from the test, in which the RPV was heated up to around 110 °C without nuclear heating with the VCS inactive, to show that the temperature is locally high but kept sufficiently low around the support in the concrete due to sufficient thermal conductivity to the cold temperature region.

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.

Loss of Core Cooling Test With One Cooling Line Inactive in Vessel Cooling System of High-Temperature Engineering Test Reactor

2017 ◽  
Vol 3 (4) ◽  
Author(s):  
Yusuke Fujiwara ◽  
Takahiro Nemoto ◽  
Daisuke Tochio ◽  
Masanori Shinohara ◽  
Masato Ono ◽  
...  
Keyword(s):  
Control System ◽  
High Temperature ◽  
Cooling System ◽  
Thermal Power ◽  
Reactor Power ◽  
Coolant Temperature ◽  
Water Cooling ◽  
Test Reactor ◽  
Core Cooling ◽  
Cooling Function

In the high-temperature engineering test reactor (HTTR), the vessel cooling system (VCS) which is arranged around the reactor pressure vessel (RPV) removes residual heat and decay heat from the reactor core when the forced core cooling is lost. The test of loss of forced cooling (LOFC) when one of two cooling lines in VCS lost its cooling function was carried out to simulate the partial loss of cooling function from the surface of RPV using the HTTR at the reactor thermal power of 9 MW, under the condition that the reactor power control system and the reactor inlet coolant temperature control system were isolated, and three helium gas circulators (HGCs) in the primary cooling system (PCS) were stopped. The test results showed that the reactor power immediately decreased to almost zero, which is caused by negative feedback effect of reactivity, and became stable as soon as HGCs were stopped. On the other hand, the temperature changes of permanent reflector block, RPV, and the biological shielding concrete were quite slow during the test. The temperature decrease of RPV was several degrees during the test. The numerical result showed a good agreement with the test result of temperature rise of biological shielding concrete around 1 °C by the numerical method that uses a calibrated thermal resistance by using the measured temperatures of RPV and the air outside of biological shielding concrete. The temperature increase of water cooling tube panel of VCS was calculated to be about 15 °C which is sufficiently small in the view point of property protection. It was confirmed that the sufficient cooling capacity of VCS can be maintained even in case that one of two water cooling lines of VCS loses its function.

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.

Effect of Coolant Water Temperature of ECCS on Failure Probability of RPV

2019 ◽  
Author(s):  
Jinya Katsuyama ◽  
Koichi Masaki ◽  
Kai Lu ◽  
Tadashi Watanabe ◽  
Yinsheng Li
Keyword(s):  
Water Temperature ◽  
Thermal Shock ◽  
Failure Probability ◽  
Structural Integrity ◽  
Cooling System ◽  
Pressure Vessels ◽  
Coolant Temperature ◽  
Pressurized Water ◽  
Coolant Water ◽  
Core Cooling

Abstract For reactor pressure vessels (RPVs) of pressurized water reactors, temperature of the coolant water in the emergency core cooling system (ECCS) may influence the structural integrity of the RPV during pressurized thermal shock (PTS) events. By focusing on a mitigation measure to raise the coolant water temperature of ECCS for aged RPVs to reduce the effect of thermal shock due to PTS events, we performed thermal hydraulic analyses and probabilistic fracture mechanics analyses by using RELAP5 and PASCAL4, respectively. The analysis results show that the failure probability of RPV decreased dramatically when the coolant temperature in accumulator as well as in the high- and low-pressure injection systems (HPI/LPI) was increased, although the increase in coolant temperature in the HPI/LPI only did not lead to a decrease in the failure probability.

ACR-1000™ Level 1 PSA Summary

2009 ◽  
Author(s):  
U. Menon ◽  
S. Aprodu ◽  
R. Bettig ◽  
R. Henderson ◽  
T. Ha ◽  
...  
Keyword(s):  
Cooling System ◽  
Safety Margin ◽  
Water System ◽  
Operation Performance ◽  
Light Water ◽  
Level 1 ◽  
Enriched Uranium ◽  
Core Cooling ◽  
Four Quadrant ◽  
High Level

The ACR-1000™ developed by Atomic Energy of Canada Limited (AECL) is a 1200 MWe - pressure tube type, light-water-cooled and heavy-water-moderated reactor, which has evolved from the well-established CANDU™ line of reactors. It retains the basic proven CANDU design features while incorporating innovations and state-of-the-art technologies to ensure fully competitive safety, operation, performance and economics. The major innovation in the ACR-1000 is the use of slightly enriched uranium fuel and light water coolant. ACR-1000 is a four-quadrant design (for easier maintenance and improved reliability). There are five safety systems in the ACR-1000; (i) two independent, diverse and fast acting shutdown systems (SDS1 and SDS2), which are physically and functionally independent from each other and from the reactor regulating system; (ii) Emergency core cooling system; (iii) Emergency Feedwater system; and (iv) Containment system, which includes a strong steel-lined containment structure. In addition the Reserve water system provides feedwater to the heat transport system, steam generators, moderator and shield cooling system for beyond design basis accidents. The Level 1 Probabilistic Safety Assessment (PSA) is conducted in support of the design phase of the ACR-1000. The purpose of Level 1 PSA is to identify whether the ACR-1000 design targets and the regulatory safety goal for severe core damage frequency (SCDF) are met with adequate margin and provide design feedback. An interim Level 1 PSA was conducted for internal at-power events. Interim assessments were conducted for shutdown state, internal fire and flood at-power events. An interim seismic margin assessment was conducted for the seismic events. The Level 1 PSA results show that the ACR™ design targets and safety goal for SCDF are met with significant safety margin. Based on the ACR-1000 Level 1 PSA, the accident behaviours of the ACR-1000 are well understood and their consequences can be predicted with a high-level of confidence. It also provides sufficient assurance that the release based regulatory safety goals are achievable for ACR-1000. The Level 1 PSA results also signify a robust design that provides a strong foundation for the ACR-1000 design. The paper summarizes the Level 1 PSA program, methodology followed, the results obtained, and insights gained during the development of the ACR-1000 Level 1 PSA.

Preliminary Large Break LOCA and SGTR Accident Analysis of the Compact Small Reactor

2017 ◽  
Author(s):  
Mian Xing ◽  
Linsen Li ◽  
Feng Shen ◽  
Xiao Hu ◽  
Zhan Liu ◽  
...  
Keyword(s):  
Cooling System ◽  
Thermal Power ◽  
Reactor Core ◽  
Safety System ◽  
Passive Safety ◽  
Primary System ◽  
The Core ◽  
Passive Safety System ◽  
Safety Margins ◽  
Guideline Values

This paper gives a brief introduction of the Compact Small Reactor (CSR). It is a simplified two-loop reactor with thermal power of 660MW and with compact primary system and passive safety feature. Preliminary safety analysis of the CSR is conducted to evaluate and further optimize the design of passive safety system, especially the passive core cooling system. Large Break Loss Of Coolant Accident (LBLOCA) and Steam Generator Tube Rupture (SGTR) are selected as two reference accidental scenarios. Each scenario is modeled and computed by RELAP5/MOD3.4. For the LBLOCA analysis, a guillotine break happens in the cold leg of the loop containing the core makeup tanks balance lines. The results show certain safety margins from the guideline values, and the passive safety system could supply enough cooling of the core. For the SGTR analysis, the results show the robustness of the design from the safety perspective. It is concluded that the safety systems are capable of mitigating the accidents and protecting the reactor core from severe damage.

Correction of Zero-Drift of Differential Pressure Flow Meters of ACME Facility

2014 ◽  
Author(s):  
Mingtao Cui ◽  
Tao Zhang
Keyword(s):  
Nuclear Power ◽  
Large Scale ◽  
Cooling System ◽  
Differential Pressure ◽  
Zero Drift ◽  
Test Facility ◽  
Passive Safety ◽  
Pressure Flow ◽  
Flow Meters ◽  
Core Cooling

ACME facility (Advanced Core-cooling Mechanism Experiment) is a large-scale test facility used to validate the performance of passive core-cooling system under SBLOCA (Small Break Lost of Coolant Accident) for the CAP1400, an upgraded passive safety nuclear power plant of AP1000. To simulate the features of passive safety system properly, DELTABAR, a kind of differential pressure flow meter, is designed to measure different mass flow of ACME. Because of the low pressure loss of DELTABAR, Zero-Drift problem of differential pressure flow meters in ACME is amplified, and some of the measured values are distorted seriously. To minimize the influence of Zero-Drift, analysis on zero-drift phenomenon is made, and a compensation method is proposed. The method is applying to PBL flow meters, and the result shows that the method is applicable.

Test Analysis of CAP1400 5cm Break Accident Based on the ACME Test Facility

2016 ◽  
Author(s):  
Sheng Zhu
Keyword(s):  
System Design ◽  
Nuclear Power ◽  
Reactor Core ◽  
Safety System ◽  
Test Facility ◽  
Passive Safety ◽  
Pressurized Water ◽  
Power Simulation ◽  
Passive Safety System ◽  
Core Cooling

CAP1400 is a large pressurized water reactor based on the passive safety conception. An ACME (Advanced Core-cooling Mechanism Experiment) facility has been designed and constructed in order to validate that the CAP1400 system design is acceptable to mitigate the loss of coolant accident (LOCA). The ACME test facility is an isotonic pressure, 1/3-scale height and 1/54.32-scale power simulation of the prototype CAP1400 nuclear power plant. It contains the main-loop system, passive safety system, secondary steam system and auxiliary system etc. The all of ACME test matrix including 5 kinds 21 cases .In this paper, the test results and the Realp5 prediction of the cold leg 5cm break accident of CAP1400 are compared and analyzed to briefly evaluate the ACME capability. Furthermore, 3 different types of 5cm cold leg break test cases are presented, and the transient process, system responses and key parameters tendency are analyzed based on the test. The results indicate that the passive safety system design can successfully combine to provide a continuous removal of core decay heat and the reactor core remains to be covered with considerable margin for the 3 different 5cm cold leg break accidents.

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