Design Characteristics of Ulchin Nuclear Power Units 5&6

2002 ◽  
Author(s):  
Deucksoo Lee ◽  
Dong-Su Kim ◽  
Young-Taik Lee ◽  
O-Keol Kwon ◽  
Jung-Cha Kim
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
System Design ◽  
Nuclear Power ◽  
Safety Analysis ◽  
Reactor Coolant ◽  
The Core ◽  
Core Damage ◽  
Actuation System ◽  
Design Characteristics

Ulchin nuclear power plant units 5&6 (UCN 5&6), which started excavation on January 1999, are two loop pressurized water reactors (PWR) with the capacity of 1000 MWe, and planned to start commercial operation on June, 2004 and June, 2005, respectively. The reactor coolant system of the UCN 5&6 consist of a reactor vessel, internals, and two steam generators, four reactor coolant pumps, a pressurizer and primary piping. Based on the system design of the first Korean Standard Nuclear Power Plant (KSNP), UCN 5&6 is designed to provide improvements in safety, reliability and cost by applying both advanced proven technology and experiences gained from the construction and operation of the previous KSNP plants. The result of the preliminary probabilistic safety assessment study for UCN 5&6 shows that the core damage frequency is lowered significantly. Several design improvement items have been adopted to the system design and contributed to lower the core damage frequency value. Among the design improvements, digital PPS and digital ESFAS are the key to the UCN 5&6 design. Furthermore, digitization of the Plant Protection System (PPS) and Engineered Safety Feature Actuation System (ESFAS) for the PWR is the first case in the PWR construction history. The Korean regulatory body reviewed the design concept of the digital PPS and digital ESFAS, and evaluated to be acceptable for the plant safety. Also, in-depth review on the detail design of the digital PPS/ESFAS and the special evaluation/audit for the software design process are underway to secure the software quality. The safety of the UCN 5&6 design has been evaluated through a two-year review on the preliminary safety analysis report. As a result, the construction permit was issued on May 17, 1999 by the government. In this paper, design characteristics of UCN 5&6 are discussed focussed on design improvements comparing with KSNR. And, some of the safety analysis results are presented as well as licensing status.

The Three Mile Island (TMI) Nuclear Accident: Revisited

1985 ◽  
Vol 1 (S1) ◽  
pp. 401-404
Author(s):  
Donald Reid
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Test Tube ◽  
Human Errors ◽  
The Core ◽  
Core Damage ◽  
Short Period ◽  
Water Pumps ◽  
The Universe

At 0400 hours on Wednesday, March 28, 1979, an extremely small and initially thought unimportant malfunction occurred at the nuclear power plant at Three Mile Island (TMI). Within a short period of time, that malfunction would turn into an event of momentous impact with repercussions felt over most of the world. The events of that malfunction would cause TMI to be labelled as the worst commercial nuclear incident in history and transform it into the nuclear test tube of the universe. What really happened at Three Mile Island? Thirty-six seconds after 0400 hours, several water pumps stopped functioning in the unit 2 nuclear power plant. In the minutes, hours and days that followed, a series of events—compounded by equipment failure, inappropriate procedures and human errors—escalated into the worst crisis yet experienced by the nation's nuclear power industry. This resulted in the loss of reactor coolant, overheating of the core, damage to the fuel (but probably no melting) and release outside the plant of radioactive gases. Hydrogen has was formed, primarily by the reaction between the zirconium casing that holds the radioactive fuel and steam. There, however, was no danger of the bubble inside the reactor vessel exploding, because of the absence of oxygen within the reactor.

Study on the Application of CDAG Methodology in EPR Nuclear Power Plant

2017 ◽  
Author(s):  
Yin Yuhao ◽  
Huang Yichao ◽  
Zhao Feng
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Fission Product ◽  
Damage Assessment ◽  
Nuclear Power ◽  
Severe Accident ◽  
Core Damage ◽  
Calculation Results ◽  
Design Characteristics ◽  
European Nuclear

The Westinghouse Owners Group Core Damage Assessment Guidance (CDAG), which has been authorized by the NRC staffs, is now used by licensee emergency response organization staff for estimating the extent of core damage that may have occurred during an accident at a Westinghouse nuclear power plant. On the other hand, EPR is a 3rd generation nuclear power plant, which applies the advanced European nuclear power technology. This paper introduced Core Damage Assessment Guidance methodology in detail. The CDAG methodology is then attempted to apply to the EPR nuclear power plant. Detailed calculations have been performed for the setpoints of containment radiation monitors (CRM) and core exit thermocouples (CETs) with EPR design characteristics, which are the two main methods for estimation core damage amount. This paper also focuses the discussion on the reasons of difference of core damage estimating results between CRM method and CETs method; based on the discussion, several advices are provided when the two methods show a reasonable discrepancy in conclusions. Several conclusions can be made from the discussions in this article that 1)the Westinghouse Owners Group CDAG methodology proved to be reasonable when applied to EPR power plant for core damage assessment under severe accident; 2) the CDAG methodology which reflect the latest understanding of fission product behavior, is very simple and timely for core damage assessment based on NPP (nuclear power plant) real-time parameters; 3) conservative calculation results of setpoints on CRM and CETs based on EPR design show a reasonable trend and range; 4) it is concluded that several factors such as the releasing way, RCS fission product retention, fuel burnups might have great impact on the estimating results, when the results from two main indications (CRM and CETs) show an unexpected response.

Effects on Core Damage Frequency to Sensitivity of Instrumentation and Control Reliability

2009 ◽  
Author(s):  
John T. Kitzmiller ◽  
Richard G. Anderson ◽  
Laura L. Genutis ◽  
Dennis M. Popp
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Sensitivity Study ◽  
Sensitivity Analyses ◽  
Common Cause ◽  
Core Damage ◽  
Actuation System ◽  
And Control ◽  
The Impact

Core Damage Frequency (CDF) and Large Release Frequency (LRF) can be greatly affected by the application of Instrumentation and Control (I&C) within the control and safety systems utilized in the operations of a nuclear power plant. It is important that the modeling of these systems be an accurate representation of the I&C design to capture the effects of the I&C on the operations of the nuclear power plant as far as control and safety that are needed for the effective and safe operation of power generation. Various sensitivities can be performed on I&C modeling in the AP1000 Probabilistic Risk Assessment (PRA) to determine the impact on CDF. These sensitivities can be in the areas of base reliability values (probabilities), common cause failures (CCF) of software as well as common cause (CC) factors of computer modules. Insights can be found as to the effects of these CCFs by varying the modeled probabilities in the most conservative direction to determine how the model will respond. The model response, i.e., CDF and cutset importance orders will provide insights as to how sensitive I&C modeling is to CCF and how sensitive the rest of the PRA model is to I&C. Increasing failure probability of the Protection and Safety Monitoring System (PMS) and Plant Control System (PLS), and Diverse Actuation System (DAS) to explore the minimum reliability that would support favorable CDF and LRF values, as well as, the effect of total failure of the PMS (no credit taken for PMS in core damage sequences) and PLS (no credit taken for PLS in core damage sequences) on CDF values are explored. Sensitivity analyses show that the CDF increases if no credit is taken for operator actions. For AP1000, this sensitivity study indicated a decrease in dependence on operator actions over conventional nuclear plants. This conclusion is likely a result of the increased reliability of the PMS to automatically actuate the given systems and components. For this reason, the most important system is PMS, in the case where no credit is taken for PMS in core damage sequences.

Study on Shutdown Fire PRA for Nuclear Power Plant

2018 ◽  
Author(s):  
Meiru Liu ◽  
Qingnan Zhao ◽  
Wei Deng ◽  
Jinyan Du ◽  
Lin Sun
Keyword(s):  
Risk Assessment ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
East Coast ◽  
Fire Risk ◽  
The Core ◽  
Core Damage ◽  
Risk Result

Fire Probabilistic Risk Assessment (PRA) is one of the main methods of fire safety analysis for nuclear power plants (NPPs). At present, the fire PRA under the at-power condition has been widely studied, while the research on the low power and shutdown condition (LPSD) is quite limited. Therefore, in this paper, a second generation NPP on the east coast of China is taken as the research target, and the analysis methods are based on the latest LPSD fire PRA theory in report NUREG/CR-7114. This paper studies the initiating events and ignition frequencies of fire PRA considering the real conditions in LPSD, and established LPSD Fire PRA model, finally obtained the quantitative risk result of the core damage caused by the fire According to the results of this LPSD fire PRA, the fire risk-significant sources and fire risk weakness are found out and the improvement suggestions have been promoted.

Transient Validation of Areva’s New Arcadia™ Code System

2014 ◽  
Author(s):  
Sebastian Kuch ◽  
Mario Leberig ◽  
Richard Brock ◽  
Florian Reiterer ◽  
Michael Riedmann ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Leading Edge ◽  
Safety Analysis ◽  
Thermal Hydraulics ◽  
Code System ◽  
The Core ◽  
Wide Range ◽  
Fully Coupled

AREVA has developed a new leading edge code suite to meet the challenges arising from increasing expectations in nuclear power plant availability and fuel performance while satisfying stricter safety requirements. ARCADIA™ [1] is an advanced 3D coupled neutronics/thermal-hydraulics/thermal-mechanics code system for Light Water Reactor (LWR) fuel assembly and core design calculations as well as safety analysis, using a new software architecture allowing for nodal and pin-by-pin calculation capability. ARCADIA™ was licensed by the US Nuclear Regulatory Commission (NRC) for applications for PWR UO2 cores in 2013. It is on the way to be licensed in other countries for AREVA customers. ARCADIA™ contains the steady-state and transient core-simulator ARTEMIS™ [2] for core design and coupled transient safety analysis. ARTEMIS™ can be used in a coupled mode with S-RELAP5 and CATHARE 2 to allow fully coupled transient analysis, combining the sophisticated 3D core model of ARTEMIS™ with the proven system thermal-hydraulics of S-RELAP5 and CATHARE 2 including a detailed simulation of the Instrumentation and Control (I&C). This allows simulating complex transients affecting the core as well as the primary and secondary side including I&C signals and responses. For the validation of ARTEMIS™ a comprehensive set of validation cases was selected, including international benchmarks and measurements covering various classes of transients. These cases include a ‘Load Rejection to station service’ event at a German 1300 MWe plant, where a wide range of system and core parameters was measured that allow the validation of the fully coupled code system. Another validation case is a nodal recalculation of the core behavior during the pump shaft break transient that occurred in the Gösgen nuclear power plant in 1985 [3]. The paper will provide representative example results for the abovementioned validation cases.

Research and Software Implementation of Seismic Probabilistic Safety Analysis Quantification Method

2017 ◽  
Author(s):  
Wei Gao ◽  
Guofeng Tang ◽  
Jingyu Zhang ◽  
Qinfang Zhang
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Fragility Curves ◽  
Safety Analysis ◽  
Fukushima Accident ◽  
Binary Decision ◽  
Plant Level ◽  
Accurate Quantification ◽  
Probabilistic Safety

Seismic risk of nuclear power plant has drawn increasing attention after Fukushima accident. An intensive study has been carried out in this paper, including sampling of component and structure fragility based on Monte Carlo method, fragility analysis on system or plant level, convolution of seismic hazard curves and fragility curves. To derive more accurate quantification results, the binary decision diagram (BDD) algorithm was introduced into the quantification process, which effectively reduces the deficiency of the conventional method on coping with large probability events and negated logic. Seismic Probabilistic Safety Analysis (PSA/PRA) quantification software was developed based on algorithms discussed in this paper. Tests and application has been made for this software with a specific nuclear power plant seismic PSA model. The results show that this software is effective on seismic PSA quantification.

Study for Application of Safety Soft Control and Information System in Highly-Integrated Control Rooms of CPR1000 NPP

2013 ◽  
Author(s):  
Xuegang Zhang ◽  
Wei Liu ◽  
Hai Chang ◽  
Jianbo Wen ◽  
Yiqian Wu ◽  
...  
Keyword(s):  
Information System ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
System Design ◽  
Nuclear Power ◽  
Integrated Control ◽  
Continuous Control ◽  
Design Basis ◽  
Control Rooms ◽  
Soft Control

For most of the newly built nuclear power plants, the computerized main control rooms (MCR) are adopted. The soft control, the typical feature of computerized Human-Interface System (HIS) in the computerized main control room and mediated by software rather than by direct physical connections, is comprised of safety and non-safety control interface which provides the operators with manual control for component-level, and allows both continuous control of plant process and discrete control of components in nuclear power plant. The safety soft control and information system (SSCIS) is used to give the safety commands to and check the immediate response of the safety process. This paper describes the application of the system design basis, functionality, communication, operation faceplate and system modes for SSCIS which is firstly introduced in CPR1000 nuclear power plant. The design criteria and basic design features of SSCIS is developed to be as the design basis of the design implementation. The ISG-04 ‘Highly-Integrated Control Rooms-Communications issues (HICRc)’ provides acceptable methods for addressing SSCIS communications in digital I&C system design. The NUREG0700 ‘Human-System Interface Design Review Guidelines’ is applied as reference for human factor engineering requirement in the SSCIS design. And the SSCIS design has also fully considered the possible customer usual practice.

scholarly journals Neutronic safety analysis of proposed reactor technologies for Ghana’s nuclear power plant using the MCNP code

2019 ◽  
Vol 34 (3) ◽  
pp. 238-242
Author(s):  
Rex Abrefah ◽  
Prince Atsu ◽  
Robert Sogbadji
Keyword(s):  
High Pressure ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Nuclear Reactor ◽  
Atomic Energy Commission ◽  
Clean Energy ◽  
Safety Analysis ◽  
Operating Conditions ◽  
The Government

In pursuance of sufficient, stable and clean energy to solve the ever-looming power crisis in Ghana, the Nuclear Power Institute of the Ghana Atomic Energy Commission has on the agenda to advise the government on the nuclear power to include in the country's energy mix. After consideration of several proposed nuclear reactor technologies, the Nuclear Power Institute considered a high pressure reactor or vodo-vodyanoi energetichesky reactor as the nuclear power technologies for Ghana's first nuclear power plant. As part of technology assessments, neutronic safety parameters of both reactors are investigated. The MCNP neutronic code was employed as a computational tool to analyze the reactivity temperature coefficients, moderator void coefficient, criticality and neutron behavior at various operating conditions. The high pressure reactor which is still under construction and theoretical safety analysis, showed good inherent safety features which are comparable to the already existing European pressurized reactor technology.

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