scholarly journals Radiocarbon Impact on a Nearby Tree of a Light-Water VVER-Type Nuclear Power Plant, Paks, Hungary

Radiocarbon ◽  
2013 ◽  
Vol 55 (2) ◽  
pp. 826-832 ◽  
Author(s):  
R Janovics ◽  
Z Kern ◽  
D Güttler ◽  
L Wacker ◽  
I Barnabás ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Tree Ring ◽  
Tree Rings ◽  
Nuclear Power ◽  
Pressurized Water Reactor ◽  
Concentration Data ◽  
Pressurized Water ◽  
The Past ◽  
Background Area

Tree-ring series were collected for radiocarbon analyses from the vicinity of Paks nuclear power plant (NPP) and a background area (Dunaföldvár) for a 10-yr period (2000–2009). Samples of holocellulose were prepared from the wood and converted to graphite for accelerator mass spectrometry (AMS) 14C measurement using the MICADAS at ETH Zürich. The 14C concentration data from these tree rings was compared to the background tree rings for each year. The global decreasing trend of atmospheric 14C activity concentration was observed in the annual tree rings both in the background area and in the area of the NPP. As an average of the past 10 yr, the excess 14C emitted by the pressurized-water reactor (PWR) NPP to the atmosphere shows only a slight systematic excess (∼6′) 14C in the annual rings. The highest 14C excess was 13′ (in 2006); however, years with the same 14C level as the background were quite frequent in the tree-ring series.

scholarly journals Radiocarbon Releases At the Krško Nuclear Power Plant

Radiocarbon ◽  
1997 ◽  
Vol 39 (3) ◽  
pp. 285-292 ◽  
Author(s):  
Barbara Vokal ◽  
Ivan Kobal
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Tree Rings ◽  
Indoor Air ◽  
Nuclear Power ◽  
Monitoring Program ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Radioactivity Monitoring

Since 1991, radiocarbon analyses of exhaust air have been part of the regular radioactivity monitoring program at the Krško Nuclear Power Plant (NPP), a Westinghouse 632 MWe pressurized water reactor (PWR). Activity of CO2 and hydrocarbons has been identified; the former contributes ca. 43%. A normalized release of total 14C of 0.219 TBqGWe−1a−1 was obtained. Indoor air 14C concentrations in selected rooms inside the plant have generally been <5 Bq m−3, although rare peaks of >1000 Bq m−3 may be reached. Tree rings have shown slight enhanced 14C activity due to the operation of the plant.

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.

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