Simulations of the Pressurizer Safety Valve Stuck Open Accident for the Development of the Accident Sequences in the Low Power/Shutdown PSA

2004 ◽  
Author(s):  
Ho-Gon Lim ◽  
Jin-Hee Park ◽  
Seung-Chul Jang ◽  
Tae-Woon Kim
Keyword(s):  
Low Power ◽  
Nuclear Power ◽  
Safety Valve ◽  
Cooling System ◽  
Rapid Cool ◽  
Primary System ◽  
Core Damage ◽  
System Pressure ◽  
Operator Action ◽  
Accident Sequences

We have simulated the pressurizer safety valve (PSV) stuck open accident in a Korea Standard Nuclear power Plant (KSNP). The purpose of the simulation is 1) to investigate the mitigation function available for the important accident sequence having considerable core damage frequencies and 2) to support the determination of the accident sequences with relevant success criteria for the Low Power Shutdown (LP&S) Probabilistic Safety Assessment (PSA). The analysis showed that the PSV stuck open accident in a KSNP has both the characteristics of a small and medium LOCA in the sense that the primary system pressure decreases slowly but the break flow is sufficient enough to uncover the core in the early stages of the accident. we found that, in the accident sequence of a high pressure safety injection (HPSI) failure, core damage could occur earlier before reaching the actuation set-pressure of the safety injection tank (SIT) provided that no operator action is considered. We also performed the simulation of a rapid cool-down by a steam generator for this accident sequence to investigate the feasibility of the SIT injection before core damage occurs. From these simulations, if the operators start the rapid cool-down operation within 15 minutes, it was shown that SIT could be injected and core damage could be prevented if the subsequent Low Pressure Safety Injection (LPSI) and Shutdown Cooling System (SCS) is successfully operated.

Modification to Shield Tank Overpressure Protection in a CANDU Reactor for Beyond Design Basis Event

2019 ◽  
Author(s):  
Michael Huang ◽  
Khurram Khan ◽  
Ali Etedali-Zadeh ◽  
Jefferson Tse ◽  
Bing Li
Keyword(s):  
Nuclear Power ◽  
Cooling System ◽  
Reactor Core ◽  
Lower Probability ◽  
Damage State ◽  
Core Damage ◽  
Design Basis ◽  
System Pressure ◽  
Severe Event ◽  
Large Heat

Abstract The Shield Tank and End Shield Cooling System in the CANDU reactor contains a large volume of light water surrounding the Calandria and circulates water to remove heat that arises from the reactor core and Moderator. In a beyond design basis event that results in a severe event, progression in the absence of mitigating cooling actions could result in a large heat load being transferred to the water inside the shield tank from the calandria wall causing shield tank failure due to over pressurization. Following the 2011 events at Fukushima Daiichi Nuclear Power Plant, the adequacy of system pressure relief was assessed against severe events. Emergency mitigating equipment tie-ins for water make-up will likely limit the core damage state and prevent the need to protect the shield tank. However, Shield Tank Overpressure Protection (STOP) has been installed against severe event conditions pursuant to the CANDU defense-in-depth safety philosophy. A larger open vent line has been installed at some CANDU units on the top of the shield tank outside containment. This design routes the vent piping high enough to preclude any venting under any operational configuration and discharges back into the containment through an existing spare penetration. Vent piping is designed as Nuclear Class 2 in accordance with ASME BPVC Section III. Assessment of stresses in the modification piping was also completed for BDBEs including for a lower probability seismic event, steam venting and corresponding higher pressure and temperature conditions.

scholarly journals Evaluation of Passive Safety Injection System Performance Under Large Break LOCA for Qinshan PWR

2021 ◽  
Vol 9 ◽  
Author(s):  
Xuesong Wang ◽  
Lin Sun ◽  
Meiru Liu ◽  
Genglei Xia
Keyword(s):  
System Performance ◽  
Nuclear Power ◽  
Cold Water ◽  
Injection System ◽  
Passive Safety ◽  
Primary System ◽  
System Pressure ◽  
Pressure Medium ◽  
Safety Parameters ◽  
Core Cooling

In this work, a brand new passive safety injection system has been designed for the ocean-based Qinshan Phase I nuclear power plant to update and replace the traditional active ones. The passive safety injection system is made up of high pressure, medium pressure, lower pressure safety injection system, and a two-stage automatic depressurization system. To evaluate the safety injection system performance, double-ended cold leg large break LOCA has been analyzed by best-estimated safety analysis RELAP5 code. The main operation and safety parameters such as primary system pressure, safe injection mass flow rates, core water level, and peak cladding temperature have been presented. The results conclude that the safety injection system can act as similar to that of the AP1000, which can assure sufficient core cooling and keep the reactor covered by the cold water under the most severe LBLOCA condition.

The extended living probabilistic safety assessment

2019 ◽  
Vol 234 (1) ◽  
pp. 183-192
Author(s):  
Marko Čepin
Keyword(s):  
Low Power ◽  
Nuclear Power ◽  
Safety Assessment ◽  
Risk Level ◽  
Probabilistic Safety Assessment ◽  
Core Damage ◽  
Full Power ◽  
Power Operation ◽  
Plant Shutdown ◽  
Probabilistic Safety

The term living probabilistic safety assessment was defined soon after the initial probabilistic safety assessments were implemented. The objective of this article is to present the extended living probabilistic safety assessment and its applications considering realistic nuclear power plant models, including the low power and shutdown plant operating states. One of the key objectives is to compare the suitability of conventional and additional risk measures, core damage frequency and conditional core damage frequency, respectively. The methods are presented considering all states of the plant from the full power operation to the low power and shutdown states. The example models of the nuclear power plants and the results of the living probabilistic safety assessment of the plant operating states are discussed. The results show that the risk of low power and shutdown states is generally smaller than the risk of full power operation, but the low power and shutdown plant operating states differ significantly among each other regarding the risk level. The deficiency of living probabilistic safety assessment applied to the plant shutdown states is connected with significantly increased human effort for the analyses, with a significantly greater amount of results and with increased uncertainty of some parameters due to the larger dynamics of actions in the plant shutdown versus the full power operation states. The benefit of the living probabilistic safety assessment applied to the plant low power and shutdown states lays in consideration of all states and potential identification of risk significant states and directions for possible safety improvements.

Investigation of Hydrodynamic Loads Associated With Pyrotechnic Valve Actuation

2012 ◽  
Author(s):  
Christopher E. Henry ◽  
Jaehyok Lim ◽  
Basar Ozar
Keyword(s):  
Nuclear Power ◽  
Cooling System ◽  
High Reliability ◽  
Injection System ◽  
Vertical Force ◽  
Hydrodynamic Loads ◽  
System Pressure ◽  
Parallel Configuration ◽  
Core Cooling ◽  
Valve Actuation

Pyrotechnic-actuated valves are utilized for various applications requiring remote actuation with high reliability. One such application is passive safety injection (SI) within the emergency core cooling system (ECCS) within the Generation III+ advanced commercial nuclear power plant designs. The pyrotechnic (explosive) actuation within the valve internals, which opens the valve for water flow, creates a vertical force that must be supported by the surrounding piping restraints. This is a well-known phenomenon that is accommodated in the design. However, there exists also a subsequent, lesser-known axial (horizontal) force that must be accommodated also. A RELAP5/MOD3.3 (patch03) code [1] model for the pyrotechnic valve and the broader injection system was configured to analyze the extent of this water hammer. Typically, the pyrotechnic actuation occurs at relatively low reactor coolant system pressure since the injection itself will eventually be a passive gravity-driven feed. However, even at this low actuation pressure, the RELAP5 analysis demonstrates that the hydrodynamic loads can be substantial. Furthermore, the analysis shows that staggered actuation of a two-valve parallel configuration can exacerbate and magnify the load, compared to a single valve actuation.

Relief Valve Impact Analysis

2020 ◽  
Author(s):  
Alton Reich
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Impact Analysis ◽  
High Pressures ◽  
Primary System ◽  
Relief Valve ◽  
Piston Velocity ◽  
Threaded Connection ◽  
System Pressure ◽  
The Impact

Abstract In nuclear power plants power actuated pressure relief valves serve several purposes. They act as safety valves and open automatically in response to unusually high pressures in the primary system. They also act as power operated valves and are used to relieve steam in response to automatic or manually initiated control signals. These valves are required to lift completely over a short duration from the time that they receive an actuation signal, or the system pressure exceeds the set point. This short lift time results in the valve disk moving at high velocities, and can result in high impact forces on the piston and stem when the valve fully opens. In order to evaluate and improve the performance of a two-stage power actuated relief valve, an analysis was performed to calculate the impact force on the main disk piston when it opened and the resulting stresses. The analysis was based on the main disk piston velocity measured during valve testing. Of particular interest were the stresses in the threaded connection between the stem and the main disk piston.

Low-Power and Shut-Down Condition Medium-Break Loss-of-Coolant Accident Success Criterion Analysis for a Typical Three-Loop Nuclear Power Plant

2016 ◽  
Vol 2 (4) ◽  
Author(s):  
Jing Sun ◽  
Changjiang Yang
Keyword(s):  
Low Power ◽  
Nuclear Power ◽  
Power Plants ◽  
Time Window ◽  
Injection System ◽  
Success Criterion ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Action Time ◽  
Operator Action

Safety equipment demands that the success criterion of useful equipment, operator-action time window, and the damage state of the reactor core can be defined by thermal-hydraulic (T-H) analysis, which makes a basic critical contribution to probabilistic safety assessment (PSA). PSA has been widely used in the safety evaluation and assessment of nuclear power plants (NPPs). A loss-of-coolant accident (LOCA) cannot be controlled without timely safety intervention. Low-power and shut-down (LPSD) conditions of NPPs can be divided into several plant operating states (POSs) in PSA analysis. After the Fukushima nuclear accident, the topic of station black-out (SBO) has drawn widespread concern. However, some LPSD conditions, which result in severe consequences like SBO, have not drawn widespread attention and are thus analyzed and discussed herein. This paper analyzes a medium-break LOCA (MBLOCA) under a certain LPSD condition for a typical three-loop NPP. A simplified method of simulating and selecting operator-action time of MBLOCA for PSA is developed. The proposed method calculates the time windows for both manually opening the high head safety injection system (HHSI) and secondary depressurizing of the system to keep the core undamaged, which could support building PSA model and human reliability analysis.

Human reliability in non-destructive inspections of nuclear power plant components: modeling and analysis

2019 ◽  
Vol 7 (2B) ◽  
Author(s):  
Vanderley Vasconcelos ◽  
Wellington Antonio Soares ◽  
Raissa Oliveira Marques ◽  
Silvério Ferreira Silva Jr ◽  
Amanda Laureano Raso
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Human Factors ◽  
Nuclear Power ◽  
Power Plants ◽  
Failure Modes ◽  
Cooling System ◽  
Human Reliability ◽  
Non Destructive ◽  
Plant Components

Non-destructive inspection (NDI) is one of the key elements in ensuring quality of engineering systems and their safe use. This inspection is a very complex task, during which the inspectors have to rely on their sensory, perceptual, cognitive, and motor skills. It requires high vigilance once it is often carried out on large components, over a long period of time, and in hostile environments and restriction of workplace. A successful NDI requires careful planning, choice of appropriate NDI methods and inspection procedures, as well as qualified and trained inspection personnel. A failure of NDI to detect critical defects in safety-related components of nuclear power plants, for instance, may lead to catastrophic consequences for workers, public and environment. Therefore, ensuring that NDI is reliable and capable of detecting all critical defects is of utmost importance. Despite increased use of automation in NDI, human inspectors, and thus human factors, still play an important role in NDI reliability. Human reliability is the probability of humans conducting specific tasks with satisfactory performance. Many techniques are suitable for modeling and analyzing human reliability in NDI of nuclear power plant components, such as FMEA (Failure Modes and Effects Analysis) and THERP (Technique for Human Error Rate Prediction). An example by using qualitative and quantitative assessesments with these two techniques to improve typical NDI of pipe segments of a core cooling system of a nuclear power plant, through acting on human factors issues, is presented.

scholarly journals Patch-Type Vibration Visualization (PVV) Sensor System Based on Triboelectric Effect

Sensors ◽  
2021 ◽  
Vol 21 (12) ◽  
pp. 3976
Author(s):  
Sun Jin Kim ◽  
Myeong-Lok Seol ◽  
Byun-Young Chung ◽  
Dae-Sic Jang ◽  
Jonghwan Kim ◽  
...  
Keyword(s):  
Signal Processing ◽  
Low Power ◽  
Nuclear Power ◽  
Power Plants ◽  
Computing System ◽  
Sensor System ◽  
Wireless Sensor ◽  
Sensor Systems ◽  
Patch Type ◽  
Triboelectric Effect

Self-powered wireless sensor systems have emerged as an important topic for condition monitoring in nuclear power plants. However, commercial wireless sensor systems still cannot be fully self-sustainable due to the high power consumption caused by excessive signal processing in a mini-electronic computing system. In this sense, it is essential not only to integrate the sensor system with energy-harvesting devices but also to develop simple data processing methods for low power schemes. In this paper, we report a patch-type vibration visualization (PVV) sensor system based on the triboelectric effect and a visualization technique for self-sustainable operation. The PVV sensor system composed of a polyethylene terephthalate (PET)/Al/LCD screen directly converts the triboelectric signal into an informative black pattern on the LCD screen without excessive signal processing, enabling extremely low power operation. In addition, a proposed image processing method reconverts the black patterns to frequency and acceleration values through a remote-control camera. With these simple signal-to-pattern conversion and pattern-to-data reconversion techniques, a vibration visualization sensor network has successfully been demonstrated.

scholarly journals A Method to Avoid Underestimated Risks in Seismic SUPSA and MUPSA for Nuclear Power Plants Caused by Partitioning Events

Energies ◽  
2021 ◽  
Vol 14 (8) ◽  
pp. 2150
Author(s):  
Woo Sik Jung
Keyword(s):  
Exact Solutions ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Fault Tree ◽  
Ground Acceleration ◽  
Special Operations ◽  
Core Damage ◽  
Full Power ◽  
Conditional Events

Seismic probabilistic safety assessment (PSA) models for nuclear power plants (NPPs) have many non-rare events whose failure probabilities are proportional to the seismic ground acceleration. It has been widely accepted that minimal cut sets (MCSs) that are calculated from the seismic PSA fault tree should be converted into exact solutions, such as binary decision diagrams (BDDs), and that the accurate seismic core damage frequency (CDF) should be calculated from the exact solutions. If the seismic CDF is calculated directly from seismic MCSs, it is drastically overestimated. Seismic single-unit PSA (SUPSA) models have random failures of alternating operation systems that are combined with seismic failures of components and structures. Similarly, seismic multi-unit PSA (MUPSA) models have failures of NPPs that undergo alternating operations between full power and low power and shutdown (LPSD). Their failures for alternating operations are modeled using fraction or partitioning events in seismic SUPSA and MUPSA fault trees. Since partitioning events for one system are mutually exclusive, their combinations should be excluded in exact solutions. However, it is difficult to eliminate the combinations of mutually exclusive events without modifying PSA tools for generating MCSs from a fault tree and converting MCSs into exact solutions. If the combinations of mutually exclusive events are not deleted, seismic CDF is underestimated. To avoid CDF underestimation in seismic SUPSAs and MUPSAs, this paper introduces a process of converting partitioning events into conditional events, and conditional events are then inserted explicitly inside a fault tree. With this conversion, accurate CDF can be calculated without modifying PSA tools. That is, this process does not require any other special operations or tools. It is strongly recommended that the method in this paper be employed for avoiding CDF underestimation in seismic SUPSAs and MUPSAs.

LES and RANS Assessment of Rib Cooled Channel Related to SGT-800 Combustor Liner

2011 ◽  
Author(s):  
Daniel Lo¨rstad
Keyword(s):  
Pressure Drop ◽  
Cooling System ◽  
Large Temperature ◽  
Internal Cooling ◽  
Low Frequencies ◽  
System Pressure ◽  
Model Size ◽  
Periodic Boundaries ◽  
Combustor Liner ◽  
Simulation Time

The main parts of the annular combustor liner walls of the Siemens gas turbine SGT-800 are convectively cooled using rib turbulated cooling. Due to the serial system of cooling and combustion air there is a potential of further reduction of total combustor pressure drop by improvements of the cooling system. Apart from the rib cooling, also the cooling channel bypass entrance is related to a significant part of the total cooling system pressure drop. In this study, an investigation is performed for a rib cooled channel which is related to the considered combustor liner and where empirical correlations are available in order to evaluate the methodology used. The study includes an assessment of the Reynolds Averaged Navier-Stokes (RANS) and Large Eddy Simulation (LES) models available within commercial Computational Fluid Dynamics (CFD) codes and includes also an investigation of model size when using periodic boundaries for LES simulations. It is well known that a small geometrical distance in the direction of the periodic boundaries may have a strong effect on the flow field but is often neglected in practice in order to speed up LES calculations. Here the effect is assessed in order to show what size is required for accurate results, both for time averaged and transient results. In addition too small domains may be affected by spurious low frequencies originating from the periodic boundaries requiring additional simulation time for time converged statistics, but also the averages may be significantly affected. In addition the simulation period for time converged statistics is evaluated in order to show that larger model size in the periodic direction does not necessarily require longer practical simulation time, due to the fact that larger volumes may be used for the combined time and space averaging. The aim is to obtain practical guidelines for LES calculations for internal cooling flows. Then the study is extended step by step to investigate the importance due to high Reynolds number, variable fluid properties and large temperature gradients in order to cover the ranges and specifics required for SGT-800 engine conditions.

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