scholarly journals Analyses of the Load Following Capabilities of Brayton Helium Gas Turbine Cycles for Generation IV Nuclear Power Plants

2017 ◽  
Vol 3 (4) ◽  
Author(s):  
A. Gad-Briggs ◽  
P. Pilidis ◽  
T. Nikolaidis
Keyword(s):  
Power Output ◽  
Nuclear Power ◽  
Power Plants ◽  
Pressure Control ◽  
Reactor Power ◽  
Critical Parameters ◽  
Power Performance ◽  
Load Following ◽  
Study Results ◽  
Cycle Efficiency

The control system for generation IV nuclear power plant (NPP) design must ensure load variation when changes to critical parameters affect grid demand, plant efficiency, and component integrity. The objective of this study is to assess the load following capabilities of cycles when inventory pressure control is utilized. Cycles of interest are simple cycle recuperated (SCR), intercooled cycle recuperated (ICR), and intercooled cycle without recuperation (IC). First, part power performance of the IC is compared to results of the SCR and ICR. Subsequently, the load following capabilities are assessed when the cycle inlet temperatures are varied. This was carried out using a tool designed for this study. Results show that the IC takes ∼2.7% longer than the ICR to reduce the power output to 50% when operating in design point (DP) for similar valve flows, which correlates to the volumetric increase for the IC inventory storage tank. However, the ability of the IC to match the ICR's load following capabilities is severely hindered because the IC is most susceptible to temperature variation. Furthermore, the IC takes longer than the SCR and ICR to regulate the reactor power by a factor of 51 but this is severely reduced, when regulating NPP power output. However, the IC is the only cycle that does not compromise reactor integrity and cycle efficiency when regulating the power. The analyses intend to aid the development of cycles specifically gas-cooled fast reactors (GFRs) and very high temperature reactors (VHTRs), where helium is the coolant.

A Framework and Model for Assessing the Design Point Performance, Off-Design Point Performance, Control, Economics and Risks of Brayton Helium Gas Turbine Cycles for Generation IV Nuclear Power Plants

2018 ◽  
Author(s):  
Arnold Gad-Briggs ◽  
Pericles Pilidis ◽  
Theoklis Nikolaidis
Keyword(s):  
Decision Making ◽  
Gas Turbine ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Decision Making Process ◽  
Design Point ◽  
Load Following ◽  
Gen Iv ◽  
Cycle Efficiency

A framework – NuTERA (Nuclear Techno-Economic and Risk Assessment) has been developed to set out the requirements for evaluating Generation IV (Gen IV) Nuclear Power Plants (NPPs) at the design conceptual stage. The purpose of the framework is to provide guidelines for future tools that are required to support the decision-making process on the choice of Gen IV concepts and cycle configurations. In this paper, the underpinning of the framework has been demonstrated to enable the creation of an analyses tool, which evaluates the design of an NPP that utilises helium closed Brayton gas turbine cycles. The tool at the broad spectrum focuses on the component and cycle design, Design Point (DP) and Off-Design Point (ODP) performance, part power and load following operations. Specifically, the design model has been created to provide functionalities that look at the in-depth sensitivities of the design factors and operation that affect the efficiency of an NPP such as temperature and pressure ratios, inlet cycle temperatures, component efficiencies, pressure losses. The ODP performance capabilities include newly derived component maps for the reactor, intercooler and recuperator for long term Off-Design (OD) operation. With regard to short term OD, which is typically driven by changes in ambient conditions, the ability to analyse the cycle load following capabilities are possible. An economic model has also been created, which calculates the component costs and the baseline economic evaluation. An incorporated risk model quantifies the performance, operational, financial and design impact risks. However, the tool is able to optimise the NPP cycle configuration based on the best economics using the Levelised Unit Electricity Cost (LUEC) as a measure. The tool has been used to demonstrate a typical decision-making process on 2 Gen IV helium closed gas turbine cycles, which apply to the Gas-cooled Fast Reactors (GFRs) and Very-High Temperature Reactors (VHTRs). The cycles are the Simple Cycle Recuperator (SCR) and Intercooled Cycle Recuperator (ICR). The tool was able to derive the most efficient cycle configurations for the ICR (53% cycle efficiency) and SCR (50% cycle efficiency). Based on these efficiency figures, the baseline LUEC ($/MWh) for the year 2020 is $62.13 for the ICR and $61.84 for the SCR. However, the inclusion of the cost of contingencies due to risks and the subsequent economic optimisation resulted in a cost of $69.70 and $69.80 for the ICR and SCR respectively.

scholarly journals A Review of the Turbine Cooling Fraction for Very High Turbine Entry Temperature Helium Gas Turbine Cycles for Generation IV Reactor Power Plants

2017 ◽  
Vol 3 (2) ◽  
Author(s):  
A. Gad-Briggs ◽  
P. Pilidis ◽  
T. Nikolaidis
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Reactor Power ◽  
Simple Cycle ◽  
Cycle Performance ◽  
Optimum Pressure ◽  
Turbine Cooling ◽  
Performance Deterioration ◽  
Cycle Efficiency ◽  
Very High

The potential for high turbine entry temperature (TETs) turbines for nuclear power plants (NPPs) requires improved materials and sophisticated cooling. Cooling is critical for maintaining mechanical integrity of the turbine for temperatures >1000 °C. Increasing TET is one of the solutions for improving efficiency after cycle optimum pressure ratios have been achieved but cooling as a percentage of mass flow will have to increase, resulting in cycle efficiency penalties. To limit this effect, it is necessary to know the maximum allowable blade metal temperature to ensure that the minimum cooling fraction is used. The main objective of this study is to analyze the thermal efficiencies of four cycles in the 300–700 MW class for generation IV NPPs, using two different turbines with optimum cooling for TETs between 950 and 1200 °C. The cycles analyzed are simple cycle (SC), simple cycle recuperated (SCR), intercooled cycle (IC), and intercooled cycle recuperated (ICR). Although results showed that deterioration of cycle performance is lower when using improved turbine material, the justification to use optimum cooling improves the cycle significantly when a recuperator is used. Furthermore, optimized cooling flow and the introduction of an intercooler improve cycle efficiency by >3%, which is >1% more than previous studies. Finally, the study highlights the potential of cycle performance beyond 1200 °C for IC. This is based on the IC showing the least performance deterioration. The analyses intend to aid development of cycles for deployment in gas-cooled fast reactors (GFRs) and very high-temperature reactors (VHTRs).

Steam Generator Replacement Project at TVA’s Sequoyah Unit 1 Nuclear Power Plant

2004 ◽  
Author(s):  
Myron R. Anderson
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Power Plants ◽  
Reactor Power ◽  
Original Design ◽  
Pressurized Water ◽  
Repair Work ◽  
Building Roof ◽  
Tennessee Valley ◽  
Generator Replacement

Pressurized Water Reactor Power Plants have at times required that large components be replaced (steam generators weighing 750,000 lbs) which have necessitated performing first time modifications to the plant that were unintended during the original design. The steam generator replacement project at Tennessee Valley Authority (TVA’s) Sequoyah Nuclear Power Station necessitated (1) two large temporary openings (21’×45’) in the plant’s Shield Building roof (2’ thick concrete) by hydro-blasting to allow the removal of the old generators and installation of the new, (2) removal and repair of the concrete steam generator enclosure roofs (20’ diameter, 3’ thick) which were removed by wire saw cutting and (3) the seismic qualification of; the design and construction of an extensive ring foundation for; the use of one of the world largest cranes to remove these components through the roof. This removal and replacement process had to be performed in an expeditious manner to minimize the amount of time the plant is shutdown so the plant could return to providing power to the grid. This paper will address some of the many technical and construction considerations required to perform this demolition and repair work safely, efficiently and in a short as possible duration.

Efficient Intensity Measures for Probabilistic Seismic Response Analysis of Anchored Above-Ground Liquid Steel Storage Tanks

2016 ◽  
Author(s):  
Hoang Nam Phan ◽  
Fabrizio Paolacci
Keyword(s):  
Ground Motion ◽  
Earthquake Engineering ◽  
Nuclear Power ◽  
Power Plants ◽  
Seismic Analysis ◽  
Time History ◽  
Response Analysis ◽  
Storage Tanks ◽  
Intensity Measures ◽  
Study Results

Liquid storage tanks are vital lifeline structures and have been widely used in industries and nuclear power plants. In performance-based earthquake engineering, the assessment of probabilistic seismic risk of structural components at a site is significantly affected by the choice of ground motion intensity measures (IMs). However, at present there is no specific widely accepted procedure to evaluate the efficiency of IMs used in assessing the seismic performance of steel storage tanks. The study presented herein concerns the probabilistic seismic analysis of anchored above-ground steel storage tanks subjected to several sets of ground motion records. The engineering demand parameters for the analysis are the compressive meridional stress in the tank wall and the sloshing wave height of the liquid free surface. The efficiency and sufficiency of each alternative IM are quantified by results of time history analyses for the structural response and a proper regression analysis. According to the comparative study results, this paper proposes the most efficient and sufficient IMs with respect to the above demand parameters for a portfolio of anchored steel storage tanks.

RETRAN Application of Turbine Trip and Load Rejection of Startup Test Analysis for Lungmen ABWR

2009 ◽  
Author(s):  
Chen-Lin Li ◽  
Chiung-Wen Tsai ◽  
Chunkuan Shih ◽  
Jong-Rong Wang ◽  
Su-Chin Chung
Keyword(s):  
Control System ◽  
Water Level ◽  
Control Systems ◽  
Nuclear Power ◽  
Pressure Control ◽  
Reactor Power ◽  
Level Control ◽  
Bypass Flow ◽  
Control Rod ◽  
Recirculation Flow

This study used RETRAN program to analyze the turbine trip and load rejection transients of Taiwan Power Company Lungmen Nuclear Power Plant’s startup test at 100% power and 100% core flow operating condition. This model includes thermal flow control volumes and junctions, control systems, thermal hydraulic models, safety systems, and 1D kinetics model. In Lungmen RETRAN model, four steam lines are simulated as one line. There are four simulated control systems: pressure control system, water level control system, feedwater control system, and speed control system for reactor internal pumps. The turbine trip event, at above 40% power, triggers the fast open of the bypass valves. Upon the turbine trip, the turbine stop valves close. To minimize steam bypassed to the main condenser, recirculation flow is automatically runback and a SCRRI (selected control rod run in) is initiated to reduce the reactor power. The load rejection event causes the fast opening of the bypass valves. Steam bypass will sufficiently control the pressure, because of their 110% bypass capacity. A SCRRI and RIP runback are also initiated to reduce the reactor power. This study also investigated the sensitivity analysis of turbine bypass flow, runback rate of RIPS and SCRRI to observe how they affect fuel surface heat flux, neutron flux and water level, etc. The results show that turbine bypass flow has larger impacts on dome pressure than RIPS runback rate and SCRRI. This study also indicates that test criteria in turbine trip and load rejection transients are met and Lungmen RETRAN model is performing well and applicable for Lungmen startup test predictions and analyses.

Canadian Nuclear Safety Commission: Readiness to Regulate SMRs in Canada

2011 ◽  
Author(s):  
S. Herstead ◽  
M. de Vos ◽  
S. Cook
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Safety Analysis ◽  
Reactor Power ◽  
Nuclear Safety ◽  
Regulatory Framework ◽  
Nuclear Industry ◽  
Plant Design ◽  
Regulatory Documents

The success of any new build project is reliant upon all stakeholders — applicants, vendors, contractors and regulatory agencies — being ready to do their part. Over the past several years, the Canadian Nuclear Safety Commission (CNSC) has been working to ensure that it has the appropriate regulatory framework and internal processes in place for the timely and efficient licensing of all types of reactor, regardless of size. This effort has resulted in several new regulatory documents and internal processes including pre-project vendor design reviews. The CNSC’s general nuclear safety objective requires that nuclear facilities be designed and operated in a manner that will protect the health, safety and security of persons and the environment from unreasonable risk, and to implement Canada’s international commitments on the peaceful use of nuclear energy. To achieve this objective, the regulatory approach strikes a balance between pure performance-based regulation and prescriptive-based regulation. By utilizing this approach, CNSC seeks to ensure a regulatory environment exists that encourages innovation within the nuclear industry without compromising the high standards necessary for safety. The CNSC is applying a technology neutral approach as part of its continuing work to update its regulatory framework and achieve clarity of its requirements. A reactor power threshold of approximately 200 MW(th) has been chosen to distinguish between large and small reactors. It is recognized that some Small Modular Reactors (SMRs) will be larger than 200 MW(th), so a graded approach to achieving safety is still possible even though Nuclear Power Plant design and safety requirements will apply. Design requirements for large reactors are established through two main regulatory documents. These are RD-337 Design for New Nuclear Power Plants, and RD-310 Safety Analysis for Nuclear Power Plants. For reactors below 200 MW(th), the CNSC allows additional flexibility in the use of a graded approach to achieving safety in two new regulatory documents: RD-367 Design of Small Reactors and RD-308 Deterministic Safety Analysis for Small Reactors. The CNSC offers a pre-licensing vendor design review as an optional service for reactor facility designs. This review process is intended to provide early identification and resolution of potential regulatory or technical issues in the design process, particularly those that could result in significant changes to the design or analysis. The process aims to increase regulatory certainty and ultimately contribute to public safety. This paper outlines the CNSC’s expectations for applicant and vendor readiness and discusses the process for pre-licensing reviews which allows vendors and applicants to understand their readiness for licensing.

Supercritical Water-Cooled Nuclear Reactors (SCWRs): Current and Future Concepts — Steam Cycle Options

2008 ◽  
Author(s):  
R. B. Duffey ◽  
I. Pioro ◽  
X. Zhou ◽  
U. Zirn ◽  
S. Kuran ◽  
...  
Keyword(s):  
Steam Turbine ◽  
Supercritical Water ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Hydrogen Generation ◽  
Electrical Energy ◽  
Reactor Outlet ◽  
Cycle Efficiency ◽  
Direct Cycle

One of the six Generation IV nuclear reactor concepts is a SuperCritical Water-cooled nuclear Reactor (SCWR), which is currently under development. The main objectives for developing and utilizing SCWRs are to increase the thermal efficiency of Nuclear Power Plants (NPPs), to decrease electrical energy costs, and possibility for co-generation, including hydrogen generation. Atomic Energy of Canada Limited (AECL) and Research and Development Institute of Power Engineering (RDIPE or NIKIET in Russian abbreviations) are currently developing pressure-tube SCWR concepts. The targeted steam parameters at the reactor outlet are approximately 25 MPa and 625°C. This paper presents a survey on modern SuperCritical (SC) steam turbine technology and a study on potential steam cycles for the SCWR plants. The survey reveals that by the time the Gen IV SCWRs are market-ready, the required steam turbine technology will be well proven. Three potential steam cycles in an SCWR plant are presented: a dual-cycle with steam reheat, a direct cycle with steam reheat, and a direct cycle with a Moisture Separator and Reheater (MSR). System thermal-performance simulations have been performed to determine the overall cycle efficiency of the proposed cycles. The results show that the direct cycle with steam reheat has the highest efficiency. The direct cycle with MSR is an alternative option, which will simplify the reactor design at the penalty of a slightly lower cycle efficiency.

scholarly journals Analyses of the Control System Strategies and Methodology for Part Power Control of the Simple and Intercooled Recuperated Brayton Helium Gas Turbine Cycles for Generation IV Nuclear Power Plants

2017 ◽  
Vol 3 (4) ◽  
Author(s):  
A. Gad-Briggs ◽  
P. Pilidis ◽  
T. Nikolaidis
Keyword(s):  
Control System ◽  
Inventory Control ◽  
Nuclear Power ◽  
Power Plants ◽  
Control Method ◽  
Control Strategies ◽  
Limiting Factors ◽  
Cycle Performance ◽  
Study Results ◽  
And Performance

An important requirement for Generation IV Nuclear Power Plant (NPP) design is the control system, which enables part power operability. The choices of control system methods must ensure variation of load without severe drawbacks on cycle performance. The objective of this study is to assess the control of the NPP under part power operations. The cycles of interest are the simple cycle recuperated (SCR) and the intercooled cycle recuperated (ICR). Control strategies are proposed for NPPs but the focus is on the strategies that result in part power operation using the inventory control method. First, results explaining the performance and load limiting factors of the inventory control method are documented; subsequently, the transient part power performances are also documented. The load versus efficiency curves were also derived from varying the load to understand the efficiency penalties. This is carried out using a modeling and performance simulation tool designed for this study. Results show that the ICR takes ∼102% longer than the SCR to reduce the load to 50% in design point (DP) performance conditions for similar valve flows, which correlates with the volumetric increase for the ICR inventory tank. The efficiency penalties are comparable for both cycles at 50% part power, whereby a 22% drop in cycle efficiency was observed and indicates limiting time at very low part power. The analyses intend to aid the development of cycles for Generation IV NPPs specifically gas cooled fast reactors (GFRs) and very high-temperature reactors (VHTRs), where helium is the coolant.

Grid-Connected Nuclear Power Plant Key Issues Summary

2014 ◽  
Vol 521 ◽  
pp. 530-535
Author(s):  
Meng Wang ◽  
Jian Ding ◽  
Tian Tang ◽  
Zhang Sui Lin ◽  
Zhen Da Hu ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Large Scale ◽  
Power Modeling ◽  
Modeling Simulation ◽  
Load Following ◽  
Key Issues ◽  
The Impact

The current situation of nuclear power plants at home and abroad is described, and the impact of large-scale nuclear power accessing to the grid is analyzed, specifically in the aspects of nuclear power modeling, simulation, load following, reliability, fault diagnosis, etc. Nuclear power accessing to the grid will bring a series of problems, the causes of each problem, the main solutions and future development directions are summarized.

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