A Method and Analysis for Improving Thermal Efficiency of Nuclear Power Plant

2013 ◽  
Vol 448-453 ◽  
pp. 1928-1933
Author(s):  
Xian Ling Li ◽  
Yang Liu ◽  
Feng Jie Zhu ◽  
Shi Wei Yao
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Flow Rate ◽  
Thermal Efficiency ◽  
Nuclear Power ◽  
Operation Mode ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Increase Flow Rate ◽  
Steam Parameter

In order to improve thermal efficiency of nuclear power plant with a pressurized water reactor, this paper analyzed the reason why nuclear power plant runs with a low thermal efficiency, and then pointed out that the conventional operation mode results in a low thermal efficiency for middle and high load regions. A new method was proposed to improve its thermal efficiency by operating nuclear power plant under the condition with a high and constant steam parameter. Two problems encountered by the proposed method were investigated, and the results indicate that it is preferential way to increase flow rate of primary side coolant to boost steam parameter of the rated condition, and it is feasible to maintain steam parameter constant by regulating flow rate of primary side coolant continuously.

scholarly journals Concentration of Radiocarbon and Its Chemical Forms in Gaseous Effluents, Environmental Air, Nuclear Waste and Primary Water of a Pressurized Water Reactor Power Plant in Hungary

Radiocarbon ◽  
1995 ◽  
Vol 37 (2) ◽  
pp. 497-504 ◽  
Author(s):  
Mihály Veres ◽  
Ede Hertelendi ◽  
György Uchrin ◽  
Eszter Csaba ◽  
István Barnabás ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Discharge Rate ◽  
Chemical Forms ◽  
Pressurized Water Reactor ◽  
Primary Coolant ◽  
Pressurized Water ◽  
Water Reactor ◽  
Environmental Air

We measured airborne releases of 14C from the Paks Pressurized Water Reactor (PWR) Nuclear Power Plant (NPP). Two continuous stack samplers collect 14C in 14CO2 and 14CnHm chemical forms. 14C activities were measured using two techniques; environmental air samples of lower activities were analyzed by proportional counting, stack samples were measured by liquid scintillation counting. 14C concentration of air in the stack varies between 80 and 200 Bqm−3. The average normalized yearly discharge rates for 1988–1993 were 0.74 TBqGW−1ey−1 for hydrocarbons and 0.06 TBqGW−1ey−1 for CO2. The discharge rate from Paks Nuclear Power Plant is about four times higher than the mean discharge value of a typical Western European PWR NPP. The higher 14C production may be apportioned to the higher level of nitrogen impurities in the primary coolant. Monitoring the long-term average excess from the NPP gave D14C = 3.5‰ for CO2 and D14C = 20‰ for hydrocarbons. We determined 14C activity concentration in the primary coolant to be ca. 4 kBq liter−1. The 14C activity concentrations of spent mixed bed ion exchange resins vary between 1.2 and 5.3 MBqkg−1 dry weight.

Structural Reliability Evaluation on the Pressurized Water Reactor Pressure Vessel Under Pressurized Thermal Shock Events

2014 ◽  
Author(s):  
Hsoung-Wei Chou ◽  
Chin-Cheng Huang
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Thermal Shock ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Reactor Pressure Vessel ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Pressurized Thermal Shock ◽  
Reactor Pressure

The failure probability of the pressurized water reactor pressure vessel for a domestic nuclear power plant in Taiwan has been evaluated according to the technical basis of the USNRC’s new pressurized thermal shock (PTS) screening criteria. The ORNL’s FAVOR code and the PNNL’s flaw models are employed to perform the probabilistic fracture mechanics analysis based on the plant specific parameters of the domestic reactor pressure vessel. Meanwhile, the PTS thermal hydraulic and the probabilistic risk assessment data analyzed from a similar nuclear power plant in the United States for establishing the new PTS rule are applied as the loading condition. Besides, an RT-based regression formula derived by the USNRC is also utilized to verify the through-wall cracking frequencies. It is found that the through-wall cracking of the analyzed reactor pressure vessel only occurs during the PTS events resulted from the stuck-open primary safety relief valves that later reclose, but with only an insignificant failure risk. The results indicate that the Taiwan domestic PWR reactor pressure vessel has sufficient structural margin for the PTS attack until either the end-of-license or for the proposed extended operation.

scholarly journals Radioactive Source Specification of Bushehr’s VVER-1000 Spent Fuels

2016 ◽  
Vol 2016 ◽  
pp. 1-4 ◽  
Author(s):  
Mahdi Rezaeian ◽  
Jamshid Kamali
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Initial Step ◽  
Total Activity ◽  
Radioactive Source ◽  
Pressurized Water Reactor ◽  
Spent Fuel ◽  
Pressurized Water

Due to high radioactivity and significant content of medium- and long-lived radionuclides, different operations with spent nuclear fuels (e.g., handling, transportation, and storage) shall be accompanied by suitable radiation protections. On the other hand, determination of radioactive source specification is the initial step for any radiation protection design. In this study, radioactive source specification of the spent fuels of Bushehr nuclear power plant, which is a VVER-1000 type pressurized water reactor, was determined. For the depletion and decay calculations, ORIGEN code was utilized. The results are presented for burnups of 30 to 49 GWd/MTHM and different cooling times up to 100 years. According to these results, total activity of a spent fuel assembly with initial enrichment of 3.92%, burnup of 49 GWd/MTHM, and cooling time of 3 years is 1.92 × 1016 Bq. The results can be utilized specifically in transportation/storage cask design for spent fuel management of Bushehr nuclear power plant.

scholarly journals Study of the Rostov nuclear power plant efficiency operation under increased vacuum value

2020 ◽  
Vol 329 ◽  
pp. 03049
Author(s):  
Aleksey Babushkin ◽  
Sergey Skubienko ◽  
Ludmila Kinash
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Water Temperature ◽  
Nuclear Power ◽  
Cooling Water ◽  
Climatic Conditions ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Cooling Water Temperature ◽  
Turbine Condenser

In this study, the influence of the cooling water temperature on the thermal efficiency of a conceptual pressurized-water reactor nuclear- power plant is studied. The change in the cooling water temperature can be experienced due to the seasonal changes in climatic conditions at plant site. The article presents the results of technical and economic parameters study of nuclear power unit’s operation under increased vacuum value. Investigated seasonal variations of cooling water temperature, cooling water temperature influence on the vacuum temperature in the turbine condenser, and changing the basic technical and economic performance of nuclear power station. The mathematical model of calculation the nuclear power plant operation for a 1000 MW power unit was developed.

Analysis of RCS Depressurization Effects on Containment Temperature for Qinshan Phase II Nuclear Power Plant

2014 ◽  
Author(s):  
Liu Lili ◽  
Zhang Ming ◽  
Deng Jian
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Hydrogen Generation ◽  
Hydrogen Combustion ◽  
Severe Accident ◽  
Generation Rate ◽  
Pressurized Water Reactor ◽  
Gas Temperature ◽  
Pressurized Water

A severe accident code was applied for modeling of a typical pressurized water reactor (PWR) nuclear power plant, and the effects of RCS depressurization on the gas temperature of the relief tank cell in the containment during a station blackout (SBO) induced accident was analyzed. The sensitivity calculation indicated that the hydrogen generation rate obviously increased due to RCS depressurization in a critical stage. The results show that RCS depressurization can play an important role in hydrogen generation rate and total accumulation, and the temperature of the containment atmosphere is highly influenced by hydrogen combustion. High temperature induced by hydrogen combustion may degrade the equipment and instruments capabilities. Based on this analysis, a feasible strategy of RCS depressurization for mitigating the accident consequence is provided for developing the capacity of the SBO treatment of Qinshan Phase Nuclear Power Plant (QSP-II NPP).

Study on the Small Pressurized Water Reactor Based on Fully Ceramic Microencapsulated Fuel

2021 ◽  
Author(s):  
Jin Feng Huang
Keyword(s):  
Thermal Conductivity ◽  
Silicon Carbide ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Carbide Matrix ◽  
Silicon Carbide Matrix

Abstract After Fukushima nuclear power plant disaster, the efforts to overcome these defects of PWRs were carried out, such as replacing the cladding and fuel materials. One of these feasible efforts is using Fully Ceramic Microencapsulated (FCM) fuel replacement traditional UO2 pellets fuel into PWR. The FCM fuels are composed of Tri-structural-isotropic (TRISO) particles embedded in silicon carbide matrix. The TRISO fuel can hold its containment integrity and without fission production releases under the design temperature limit of 1600 °C. Furthermore, the silicon carbide matrix will benefit the thermal conductivity, radiation damage resistance, environmental stability and proliferation resistance. Consequently, the safety of the reactor could be significantly improved with FCM fuel instead of the conventional UO2 pellet fuel in PWR. We also analyzed the temperature distribution for the FCM fuel compared the traditional UO2 pellets, the calculation indicated that the centerline temperature is lower than UO2 pellets due to FCM higher thermal conductivity. The calculation demonstrated that, utilizing FCM replacement of conventional UO2 fuel rod is feasible and more security in a small pressurized water reactor. In this paper, a small pressurized water reactor utilizing FCM fuel is considered. A 17 × 17 fuel assemblies with Zircalloy cladding was applied in conceptual design through a preliminary neutronics and thermal hydraulics analysis. The reactor physics is accomplished, including the refuel cycle length, the effective multiplication factor, power distribution analysis being discussed. The Soluble Boron Free (SBF) concepts are introduced in small PWR, as a result, it makes the nuclear power plant more simpler and economical. FCM fuel loading has a very high excess reactivity at the beginning of reactor core life, and it is important to flat reactivity curve during operation, or to minimize excess reactivity during the core life. Consequently, conventional burnable poison configurations were introduced to suppress excess reactivity control at beginning of cycle.

Comparison and Reference Between RSE-M and ASME Section XI Regulations on In-Service Inspection for China Nuclear Power Plant

2017 ◽  
Author(s):  
Sun Haitao ◽  
Zhang Qinghua ◽  
Jia Panpan ◽  
Ling Ligong ◽  
Wang Chen ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Reactor Power ◽  
Mechanical Equipment ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Scope Of Application ◽  
Specific Implementation ◽  
Service Inspection

RSE-M and ASME Section XI regulations are currently recognized as NPP ISI regulations which are most extensively used, most detailed in content and most mature in technology in the world. In China, RSE-M or ASME section XI regulations are used to guide the ISI program preparation, ISI activities implementation and management. There are many differences between RSE-M and ASME Section XI regulations, such as application of regulation, scope of application, main frame and content, basic requirements of ISI, acceptance and evaluation of ISI results, repairing and change. At the same time, some technical clauses can be used for reference by each other, such as qualification and inspection items. By comparison of RSE-M and ASME Section XI regulations for in-service inspection rules of China pressurized water reactor power plant, the technical differences between the two rules are analyzed. Combined with application and engineer practices for in-service inspection of China nuclear power plant, some technical terms used for mutual reference are summarized to provide assistance for establishing the in-service inspection program and specific implementation. During ISI activity, applicable regulation should be chosen according to the requirements of design in consideration of inspection items, examination methods, defect acceptance criteria and evaluation. Meanwhile, implementation of RSE-M and ASME Section XI regulations should be combined with the design features, experience feedback and aging management of mechanical equipment, and draw lessons from mature technical clauses of other regulations.

Anti-Ratcheting Design Software for Primary Auxiliary Pipeline in Nuclear Power Plant

2010 ◽  
Vol 118-120 ◽  
pp. 126-130
Author(s):  
Zhe Guang Zhang ◽  
Xu Chen ◽  
Fei Xue ◽  
Wei Wei Yu
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Display Module ◽  
Ansys Simulation ◽  
Asme Code ◽  
Computing Module ◽  
The Relationship

In this study, based on ASME Code Section III, France RCC-MR, ANSYS simulation and experimental data, ratcheting and ratcheting boundary are obtained to determine allowed loading of primary auxiliary pipeline in nuclear power plant. A software of pressurized water reactor auxiliary pipeline anti-ratcheting design is compiled. The calculation method to determine the boundary of ratcheting effect is proposed in order to provide a basis for pipe structural design. This work focuses on the work of the establishment of the structure of software system. The structure mainly composed by three program modules, i.e. the database module, data computing module and data display module. The function of each module and the relationship between the modules are explained. An example to determine ratcheting boundary is presented in the paper.

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