Comparison of Nuclear Power Plants With Closed-Cycle Helium Turbine and With Steam Turbine Cycle for Combined Power and Steam Generation

1973 ◽  
Vol 95 (1) ◽  
pp. 11-18 ◽  
Author(s):  
K. Bammert ◽  
R. Buende
Keyword(s):  
Steam Turbine ◽  
Nuclear Power ◽  
Power Plants ◽  
Closed Cycle ◽  
Steam Generation ◽  
Single Cycle ◽  
Cycle System ◽  
System A ◽  
Secondary Steam

The heat of a helium-cooled reactor can be used for combined power and steam generation either in a closed-cycle helium turbine system, the so-called single-cycle system, or in a two-cycle system which consists of a helium cycle and a secondary steam turbine cycle. The optimum data for the two systems are determined within the same range of general parameters—electric power output and quantity and quality of the steam produced—as functions of the special parameters of each particular cycle system. A method of comparing different power plant systems is shown. With this method it is possible to determine those ranges in which the efficiencies achieved with one system are higher than those obtained with the other. It is described in which way the dividing line between such ranges depends on the special parameters of the cycles. The comparison shows that the single-cycle system offers advantages.

Nuclear Power Plants With Closed-Cycle Helium Turbine for Industrial Energy Supply

1971 ◽  
Vol 93 (1) ◽  
pp. 156-161 ◽  
Author(s):  
K. Bammert ◽  
E. Boehm ◽  
R. Buende
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Expansion Ratio ◽  
Inlet Temperature ◽  
Closed Cycle ◽  
Steam Generation ◽  
Regenerative Heat ◽  
Cycle Number ◽  
Industrial Energy ◽  
Gas Turbine Cycle

All the possible arrangements of closed-cycle helium turbines for combined power and steam generation can be reduced to three variations of one single arrangement. A method for the optimum layout of this arrangement is described. Optimum data are determined for the expansion ratio and the regenerative heat exchange as functions of general parameters—electric power output and quantity and quality of the steam produced—and as functions of special parameters of the gas turbine cycle—number of intercoolers, cooling and feedwater temperatures, turbine inlet temperature and pressure drop—to obtain the best possible efficiencies.

Architecting “Maintenance Program” for Piping and Pipe Support

2013 ◽  
Author(s):  
Koichi Tsumori ◽  
Yoshizumi Fukuhara ◽  
Hiroyuki Terunuma ◽  
Koji Yamamoto ◽  
Satoshi Momiyama
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Industry ◽  
Maintenance Management ◽  
3D Cad ◽  
Cad System ◽  
Nuclear Disaster ◽  
Safety Regulations ◽  
Maintenance Program

A new inspection standard that enhanced quality of operating /maintenance management of the nuclear power plant was introduced in 2009. After the Fukushima Daiichi nuclear disaster (Mar. 11th 2011), the situation surrounding the nuclear industry has dramatically changed, and the requirement for maintenance management of nuclear power plants is pushed for more stringent nuclear safety regulations. The new inspection standard requires enhancing equipment maintenance. It is necessary to enhance maintenance of not only equipment but also piping and pipe support. In this paper, we built the methodology for enhancing maintenance plan by rationalizing and visualizing of piping and pipe support based on the “Maintenance Program” in cooperating with 3D-CAD system.

The New Generation of High-Pressure Heaters for Steam Turbine Units at Nuclear Power Plants

2021 ◽  
Vol 68 (4) ◽  
pp. 285-294
Author(s):  
V. M. Zorin ◽  
A. S. Shamarokov ◽  
S. B. Pustovalov
Keyword(s):  
High Pressure ◽  
Steam Turbine ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
New Generation

Development of Long Rotating Blade Applied in the New Generation Nuclear Steam Turbine

2012 ◽  
Author(s):  
Kai Cheng ◽  
Zeying Peng ◽  
Gongyi Wang ◽  
Xiaoming Wu ◽  
Deqi Yu
Keyword(s):  
Steam Turbine ◽  
Nuclear Power ◽  
Power Plants ◽  
Structure Design ◽  
Good Reliability ◽  
Pressurized Water ◽  
Height Range ◽  
Water Reactor ◽  
Rotating Blade ◽  
Last Stage

In order to meet the high economic requirement of the 3rd generation Pressurized Water Reactor (PWR) or Boiling Water Reactor (BWR) applied in currently developing nuclear power plants, a series of half-speed extra-long last stage rotating blades with 26 ∼ 30 m2 nominal exhaust annular area is proposed, which covers a blade-height range from 1600 mm to 1900 mm. It is well known that developing an extra long blade is a tough job involving some special coordinated sub-process. This paper is dedicated to describe the progress of creating a long rotating blade for a large scaled steam turbine involved in the 3rd generation nuclear power project. At first the strategy of how to determine the appropriate height for the last-stage-rotating-blade for the steam turbine is provided. Then the quasi-3D flow field quick design method for the last three stages in LP casing is discussed as well as the airfoil optimization method. Furthermore a sophisticated blade structure design and analyzing system for a long blade is introduced to obtain the detail dimension of the blade focusing on the good reliability during the service period. Thus, except for CAD and experiment process, the whole pre-design phase of the extra-long turbine blade is presented which is regarded as an assurance of the operation efficiency and reliability.

ANALYSIS OF THE RADIOLOGICAL SAFETY CONTROL LEVEL VERSUS THE EXPOSURE OF RADIATION WORKERS IN SOUTH KOREA FROM 2008–17

2018 ◽  
Vol 184 (1) ◽  
pp. 98-108
Author(s):  
Sang-Tae Kim ◽  
Jaeryong Yoo
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Radiation Safety ◽  
Safety Management ◽  
Control Level ◽  
Radioactive Isotopes ◽  
Safety Control ◽  
Level Versus ◽  
Radiation Workers

Abstract In this study, the radiation exposure of workers at workplaces registered and licensed between 2008 and 2017 for the production/sale/use of radioactive isotopes (RI) and radioactive generators (RG) was analysed to evaluate the quality of radiation safety management controls in use. The number of facilities using RIs increased by ~26% from 2008 to 2017 whereas the number of facilities using RGs increased by ~166% over the same period. There were 33 029 radiation workers in all fields in 2008, and the number increased by ~32% to 43 467 by 2017. However, the collective effective dose of radiation received by workers decreased in all industries except for those working in nuclear power plants. In other words, the quality of radiation safety management improved over that same time period due to the systematic, continuous introduction of safety mechanisms by the regulatory authority.

scholarly journals Efficiency of Using the Auxiliary Boiler for Peak Power Generation by a Steam Turbine of NPP

2021 ◽  
Vol 64 (1) ◽  
pp. 78-90
Author(s):  
V. A. Khrustalev ◽  
M. V. Garievskii
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Steam Turbine ◽  
Nuclear Power ◽  
Power Plants ◽  
Normal Operation ◽  
Heat Output ◽  
Generation Process ◽  
System Effect ◽  
Industrial Enterprises

The article presents the technique of an estimation of efficiency of use of potential heat output of an auxiliary boiler (AB) to improve electric capacity and manoeuvrability of a steam turbine unit of a power unit of a nuclear power plant (NPP) equipped with a water-cooled water-moderated power reactor (WWER). An analysis of the technical characteristics of the AB of Balakovo NPP (of Saratov oblast) was carried out and hydrocarbon deposits near the NPP were determined. It is shown that in WWER nuclear power plants in Russia, auxiliary boilers are mainly used only until the normal operation after start-up whereas auxiliary boiler equipment is maintained in cold standby mode and does not participate in the generation process at power plants. The results of research aimed to improve the systems of regulation and power management of power units; general principles of increasing the efficiency of production, transmission and distribution of electric energy, as well as the issues of attracting the potential of energy technology sources of industrial enterprises to provide load schedules have been analyzed. The possibility of using the power complex NPP and the AB as a single object of regulation is substantiated. The authors’ priority scheme-parametric developments on the possibility of using the thermal power of the auxiliary boilers to increase the power of the steam turbine of a nuclear power plant unit equipped with WWER reactors unit during peak periods, as well as the enthalpy balance method for calculating heat flows, were applied. The surface area of the additional heater of the regeneration “deaerator – high pressure heaters” system and its cost were calculated. On the basis of calculations, it was shown that the additional power that can be obtained in the steam turbine of the NPP with a capacity of 1200 MW due to the use of heat of the modernized auxiliary boiler in the additional heat exchanger is 40.5 MW. The additional costs for the implementation of the heat recovery scheme of the auxiliary boiler at different prices for gas fuel and the resulting system effect were estimated in an enlarged way. Calculations have shown the acceptability of the payback period of the proposed modernization.

Digital Instrumentation and Control Systems Upgrades in Current Generation Nuclear Power Plants

2010 ◽  
Author(s):  
Steven A. Arndt
Keyword(s):  
United States ◽  
Cyber Security ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Plant Protection ◽  
The United States ◽  
Regulatory Compliance ◽  
Digital Systems

Over the past 20 years, the nuclear power industry in the United States (U.S.) has been slowly replacing old, obsolete, and difficult-to-maintain analog technology for its nuclear power plant protection, control, and instrumentation systems with digital systems. The advantages of digital technology, including more accurate and stable measurements and the ability to improve diagnostics capability and system reliability, have led to an ever increasing move to complete these upgrades. Because of the difficulties with establishing digital systems safety based on analysis or tests, the safety demonstration for these systems relies heavily on establishing the quality of the design and development of the hardware and software. In the United States, the U.S. Nuclear Regulatory Commission (NRC) has established detailed guidelines for establishing and documenting an appropriate safety demonstration for digital systems in NUREG-0800, “Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition,” Chapter 7, “Instrumentation and Controls,” Revision 5, issued March 2007 [1], and in a number of regulatory guides and interim staff guidance documents. However, despite the fact that the United States has a well-defined review process, a number of significant challenges associated with the design, licensing, and implementation of upgrades to digital systems for U.S. plants have emerged. Among these challenges have been problems with the quality of the systems and the supporting software verification and validation (V&V) processes, challenges with determining the optimum balance between the enhanced capabilities for the new systems and the desire to maintain system simplicity, challenges with cyber security, and challenges with developing the information needed to support the review of new systems for regulatory compliance.

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.

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