A Next-Generation Reactor

1998 ◽  
Vol 120 (08) ◽  
pp. 68-71 ◽  
Author(s):  
Michael Valenti
Keyword(s):  
Power Plant ◽  
Power Efficiency ◽  
Nuclear Power ◽  
Nuclear Technology ◽  
Reactor Safety ◽  
Pressurized Water Reactor ◽  
Historical Reconstruction ◽  
Pressurized Water ◽  
European Standards ◽  
Safety Systems

This article highlights that Electricité de France’s (EDF) N4 nuclear technology has increased European standards of power, efficiency, and safety beyond previous limits. EDF, the Paris-based French national utility, has developed and is operating the N4 reactor, capable of generating 1450 megawatts, at its power plant in Chooz. The N4 was designed to put public safety concerns to rest while providing more power than the previous generation of 1,300-megawatt EDF reactors. This installation represents the next step in French, European, and possibly the world’s nuclear power. Chooz A began operations in 1967 as the first pressurized water reactor (PWR) in France, originally based on the 185-megawatt synchronized PWR Yankee Rowe plant in Massachusetts. Typical reactor safety systems analyze a problem after it occurs. Such a procedure involves painstaking historical, reconstruction that is time-consuming, often difficult to interpret, and less reliable as time passes.

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.

Application of Risk-Informed Methods to Improve Plant Operation at Kozloduy Nuclear Power Plant

2009 ◽  
Author(s):  
Emil Kichev ◽  
Ivan Ivanov ◽  
Kaliopa Mancheva ◽  
Yasen Petrov ◽  
Vesselina Vladimirova ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Level ◽  
Full Range ◽  
Economic Benefits ◽  
Pressurized Water ◽  
Safety Systems ◽  
The Impact ◽  
Service Inspection

Refueling outages at the Kozloduy Nuclear Power Plant (KNPP) Units 5 and 6 are used to perform annual repairs and preventive maintenance activities, piping inspections, and test activities. A refueling outage at KNPP typically requires 60 days and occurs on an annual basis. Testing of safety systems at the KNPP Units 5 and 6 is an extensive exercise that results in multiple actuations of all components during each test and a relatively high number of component actuations each year. This results in equipment wear out issues that can lead to considerable component replacement and/or refurbishment. Numerous piping in-service inspections are conducted in locations where there has been no industry or plant-specific indications or failures, leading to unnecessary personnel exposure. KNPP is interested in using risk-informed (RI) approaches to reduce refueling outage length, piping inspections, testing, and exposure. KNPP is a four-loop Voda-Vodyanoi Energetichesky Reaktor (VVER) with a power level of 1000 MWe. Safety systems consist of three trains. The KNPP at-power probabilistic safety assessment (PSA) model includes internal and external events. It addresses the full range of events leading to core damage frequency (CDF) and includes a simplified level 2 model leading to large early release frequency (LERF). The RI approach, as defined in the U.S. Nuclear Regulatory Commission’s (NRC’s) risk-informed (RI) Regulatory Guides (RGs) 1.174, 1.177, and 1.178, was used in this program. The specific approach used for risk-informed in-service inspection (RI-ISI) is based on the Pressurized Water Reactor Owner’s Group methodology. The overall approach for each of the three applications used a multi-step process which included the following: identification of systems to address; identification of alternatives to current maintenance, inspection, and testing practices; a risk assessment of the proposed alternatives; an assessment of the impact of the changes on deterministic considerations; identification of monitoring requirements; and an assessment of the economic benefits. The RI-ISI program also considered the impact of the changes on personnel exposure. The overall approach made extensive use of data assessments, reliability methods, and risk assessments. The results demonstrated that the proposed changes in maintenance, in-service inspection, and testing programs have a small impact on risk, based on CDF and LERF. In addition, the proposed changes provide significant benefits in terms of reduced outage time, in-service inspections, testing requirements, and personnel exposure. The economic analysis demonstrated that changes to the maintenance program provide the largest benefit followed by the changes to the in-service inspection program and then the changes to the testing program.

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.

Dynamic fault tree analysis of auxiliary feedwater system in a pressurized water reactor

Kerntechnik ◽  
2021 ◽  
Vol 86 (2) ◽  
pp. 164-172
Author(s):  
R. A. Fahmy ◽  
R. I. Gomaa
Keyword(s):  
Power Plant ◽  
System Reliability ◽  
Nuclear Power ◽  
Fault Tree ◽  
Fault Tree Analysis ◽  
Cost Effective ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Tree Analysis ◽  
Dynamic Fault Tree

Abstract The safe and secure designs of any nuclear power plant together with its cost-effective operation without accidents are leading the future of nuclear energy. As a result, the Reliability, Availability, Maintainability, and Safety analysis of NPP systems is the main concern for the nuclear industry. But the ability to assure that the safety-related system, structure, and components could meet the safety functions in different events to prevent the reactor core damage requires new reliability analysis methods and techniques. The Fault Tree Analysis (FTA) is one of the most widely used logic and probabilistic techniques in system reliability assessment nowadays. The Dynamic fault tree technique extends the conventional static fault tree (SFT) by considering the time requirements to model and evaluate the nuclear power plant safety systems. Thus this paper focuses on developing a new Dynamic Fault Tree for the Auxiliary Feed-water System (AFWS) in a pressurized water reactor. The proposed dynamic model achieves a more realistic and accurate representation of the AFWS safety analysis by illustrating the complex failure mechanisms including interrelated dependencies and Common Cause Failure (CCF). A Simulation tool is used to simulate the proposed dynamic fault tree model of the AFWS for the quantitative analysis. The more realistic results are useful to establish reliability cantered maintenance program in which the maintenance requirements are determined based on the achievement of system reliability goals in the most cost-effective manner.

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