Multi-Pressure Surface Condenser Performance Evaluation: A Case Study

2004 ◽  
Author(s):  
Komandur S. Sunder Raj
Keyword(s):  
Power Plant ◽  
Nuclear Power ◽  
Performance Modeling ◽  
Plant Design ◽  
Turbine Generator ◽  
Modeling Tools ◽  
Pressure Surface ◽  
Design Basis ◽  
Predicting Performance

Surface condensers for power plant applications are generally specified and designed following turbine-condenser optimization studies. The turbine manufacturer provides turbine-generator performance data (thermal kit) at the very outset of plant design when the condenser is usually a black box and not much is known about its design. The turbine-generator guarantee would then be based on a specified condenser pressure that may or may not be attainable once the condenser is actually specified and designed. The condenser pressure used for the turbine performance guarantee might assume a single-pressure condenser while the actual design might be a multi-pressure condenser. In order to properly predict and monitor the performance and conduct diagnostics on a multi-pressure condenser, it is important to understand the design basis and develop an accurate model using performance modeling tools. The paper presents a multi-pressure condenser case study for a 600 Mwe nuclear power plant. The paper discusses the design basis used, interface between the turbine and condenser, use of a performance modeling tool for predicting performance, determining capacity losses attributable to the condenser and conducting diagnostics.

scholarly journals Assessment of Adequate Margin to Liquefaction for Nuclear Power Plants

2018 ◽  
Vol 2018 ◽  
pp. 1-7 ◽  
Author(s):  
Tamás János Katona
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Ground Acceleration ◽  
Design Basis ◽  
Essential Question ◽  
Radioactive Release ◽  
Adequate Margin

Design of nuclear power plant shall provide an adequate margin to protect items ultimately necessary to prevent an early large radioactive release in the case of earthquakes exceeding those considered in the design. An essential question is how large the margin should be to be accepted as adequate. In the practice, depending on the country regulation, a plant margin of at least 1.4 or 1.67 times the design basis peak ground acceleration is required to be demonstrated. The catastrophe at the Fukushima Daiichi Nuclear Power Plant revealed the fundamental experience that the plants designed in compliance with nuclear standards can survive the effects of the vibratory ground motion due to disastrous earthquake but may fail due to effects of phenomena accompanying or generated by the earthquakes. Liquefaction is one of those secondary effects of beyond-design basis earthquakes that should be investigated for NPPs at soil sites. However, the question has not been investigated up to now, whether a “margin earthquake”, vibratory effects of which the plant can withstand thanks to design margin, will not induce liquefaction at soil sites and will not result in loss of safety functions. In the paper, a procedure is proposed for calculation of the probability and margin to liquefaction. Use of the procedure is demonstrated on a case study with realistic site-plant parameters. Criteria for probability for screening and acceptable probabilistic margin to liquefaction are proposed. The possible building settlement due to margin earthquake is also assessed.

Performance Considerations in Replacement of Low-Pressure Turbine Rotors for Nuclear Power Plants: A Case Study

2005 ◽  
Author(s):  
Komandur S. Sunder Raj
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Low Pressure ◽  
Turbine Rotor ◽  
Modeling Tools ◽  
Low Pressure Turbine ◽  
Turbine Rotors ◽  
The Impact

In recent years, the nuclear power industry has witnessed profound changes in terms of renewal of operating licenses and power uprates. Renewal of operating licenses for an additional 20 years beyond the original licensed period of 40 years entails several considerations relating to aging management, performance, reliability, availability, and maintainability. Power uprates range from a low of up to 2% due to improved techniques in feedwater flow measurement to a high of up to 20% for extended power uprates. Since the limitations of power uprates are generally encountered in design of the turbine cycle, the impact upon the performance, reliability, availability and maintainability of the equipment and components in the turbine cycle may vary from low or moderate to significant. Several nuclear power plant owners have already replaced the low-pressure turbine rotors of their nuclear units with improved designs to mitigate blade failures and forced outages due to stress corrosion cracking, to reduce inspection intervals and maintenance, achieve higher output due to improved efficiency, etc. Others are either embarking upon or planning similar initiatives to confront aging, performance, availability, reliability and maintainability concerns stemming from renewal of operating licenses as well as the need to accommodate higher pressures and flows accompanying the proposed power uprates. Typically, the original low-pressure turbine designs utilizing built-up rotors with shrunk-on disks are being replaced with monoblock rotors with fully integral disks, couplings, blading and shrouds. The last stage blading is also longer resulting in a larger annulus area. Since these replacement programs involve significant expenditures, several factors need to be considered in order to ensure that the objectives of the rotor replacement programs are met. Using a case study, this paper examines the various considerations involved in replacing the low-pressure turbine rotors for a nuclear power plant. Design, performance and test considerations that need to be addressed before and after the low-pressure turbine rotors are replaced are discussed. The use of performance modeling tools in evaluating performance gains from low-pressure turbine rotor replacements is reviewed. Finally, the paper provides recommendations for ensuring that the objectives of a low-pressure turbine rotor replacement program are met.

scholarly journals Radiation risk assessment for nuclear power plant staff in different emergency situations

2021 ◽  
Vol 66 (3) ◽  
pp. 378-384
Author(s):  
V. I. Orlovskaya ◽  
A. G. Trifonov
Keyword(s):  
Risk Assessment ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Radiation Risk ◽  
Internal Exposure ◽  
Plant Design ◽  
Emergency Situations ◽  
Design Basis ◽  
Design Basis Accident

The paper presents the results of radiation risk assessment for the staff of a nuclear power plant design during design basis accident (spent nuclear fuel assembly falling on fuel in reactor core or storage pool during refueling operations) and a beyond design basis accident (large leakage of the primary coolant with failure of the active part of the emergency cooling system and complete blackout for 24 h). The assessment is based on state-of-the-art radiation risk models from the United Nations Scientific Committee on the Effects of Atomic Radiation (UNSCEAR) and the International Commission on Radiological Protection (ICRP). The calculation of risk indicators for occupational exposure of NPP staff in emergency situations was carried out on the basis of data obtained using a computational module created in the COMSOL 5.6 multiphysics software, doses from a radioactive cloud and internal exposure due to inhalation for such radionuclides as 134Cs, 137Cs, 131I, 133I, 90Sr. A feature of this approach is the detailed consideration of the NPP industrial site infrastructure, which allows obtaining a more accurate assessment of the radionuclide air distribution and fallout.

Current status of design basis earthquake level for nuclear power plant sites in several countries

2021 ◽  
Author(s):  
Hoseon Choi ◽  
Seung Gyu Hyun
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
South Korea ◽  
Seismic Activity ◽  
Nuclear Power ◽  
Current Status ◽  
Seismic Safety ◽  
Design Basis ◽  
Design Earthquake ◽  
Design Construction

<p>According to strict criteria step by step for site selection, design, construction and operation, the seismic safety of nuclear power plant (NPP) sites in South Korea are secured by considering design basis earthquake (DBE) level capable of withstanding the maximum ground motions that can occur on the site. Therefore, it is intended to summarize DBE level and its evaluation details for NPP sites in several countries.</p><p>Similar but different terms are used for DBE from country to country, i.e. safe shutdown earthquake (SSE), design earthquake (DE), SL2, Ss, and maximum calculated earthquake (MCE). They may differ when applied to actual seismic design process, and only refer to approximate comparisons. This script used DBE as a representative term, and DBE level was based on horizontal values.</p><p>The DBE level of NPP sites depends on seismic activity of the area. Japan and Western United States, where earthquakes occur more frequently than South Korea, have high DBE values. The DBE level of NPP sites in South Korea has been confirmed to be similar or higher compared to that of Central and Eastern Unites Sates and Europe, which have similar seismic activity.</p>

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.

Verification of Plastic Interactive Systems

i-com ◽  
2015 ◽  
Vol 14 (3) ◽  
Author(s):  
Raquel Oliveira ◽  
Sophie Dupuy-Chessa ◽  
Gaëlle Calvary
Keyword(s):  
Power Plant ◽  
User Interfaces ◽  
Nuclear Power ◽  
Interactive Systems ◽  
System Model ◽  
Equivalence Checking ◽  
The Past ◽  
Different Levels ◽  
Formal Techniques

AbstractInteractive systems have largely evolved over the past years. Nowadays, different users can interact with systems on different devices and in different environments. The user interfaces (UIs) are expected to cope with such variety. Plastic UIs have the capacity to adapt to changes in their context of use while preserving usability. Such capability enhances UIs, however, it adds complexity on them. We propose an approach to verifying interactive systems considering this adaptation capability of the UIs. The approach applies two formal techniques: model checking, to the verification of properties over the system model, and equivalence checking, to compare different versions of a UI, thereby identifying different levels of UI equivalence. We apply the approach to a case study in the nuclear power plant domain in which several UI are analyzed, properties are verified, and the level of equivalence between them is demonstrated.

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