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.

scholarly journals The Impact of Climate Changes on the Thermal Performance of a Proposed Pressurized Water Reactor: Nuclear-Power Plant

2014 ◽  
Vol 2014 ◽  
pp. 1-7 ◽  
Author(s):  
Said M. A. Ibrahim ◽  
Mohamed M. A. Ibrahim ◽  
Sami. I. Attia
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Thermal Performance ◽  
Nuclear Power ◽  
Cooling Water ◽  
Plant Performance ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
The Impact

This paper presents a methodology for studying the impact of the cooling water temperature on the thermal performance of a proposed pressurized water reactor nuclear power plant (PWR NPP) through the thermodynamic analysis based on the thermodynamic laws to gain some new aspects into the plant performance. The main findings of this study are that an increase of one degree Celsius in temperature of the coolant extracted from environment is forecasted to decrease by 0.39293 and 0.16% in the power output and the thermal efficiency of the nuclear-power plant considered, respectively.

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.

The Simulation Research of Noncondensable Gas to Condensation in Secondary Side Condenser of Floating Nuclear Power Plant Based on RELAP5

2018 ◽  
Author(s):  
Si-wei Yan ◽  
Chun-mei Li ◽  
Tie-bo Liang ◽  
Jing Zhao ◽  
Cheng-ming Hao ◽  
...  
Keyword(s):  
Heat Transfer ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Cooling Water ◽  
Gas Content ◽  
Simulation Research ◽  
Cooling Water Temperature ◽  
Noncondensable Gas ◽  
Secondary Side

Similar to conventional nuclear power plant, condensate water subcooling is a common problem in secondary coolant of floating nuclear power plant, which is caused by many reasons. In this article, RELAP5 is used to simulate the phenomenon of condensate water subcooling caused by noncondensable gas. The influence of noncondensable gas to condenser pressure, subcooling temperature, heat transfer rate, terminal temperature difference, cooling water temperature rise is presented. The results obtained through this study have shown that the model with non-condensable gas in steam can simulate condensate water subcooling, and reveal the discipline of condenser heat transfer characteristics as a function of noncondensable gas content.

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).

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