Benchmark Calculations of Pressurizer Model for Maanshan Nuclear Power Plant Using TRACE Code

2008 ◽  
Author(s):  
Yi-Hsiang Cheng ◽  
Chunkuan Shih ◽  
Jong-Rong Wang ◽  
Hao-Tzu Lin
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Primary Coolant ◽  
Pressurized Water ◽  
Start Up ◽  
Simulation Results ◽  
Net Load ◽  
Coolant System

Pressurizer plays an important role in controlling the pressure of the primary coolant system in pressurized water reactor (PWR) power plants. An accurate modelling of the pressurizer is needed to determine the pressure histories of the primary coolant system, and thus to successfully simulate overall PWR power plant behavior during transients. The purpose of this study is to develop a pressurizer model, and to assess its pressure transients using the TRACE code version 5.0. The benchmark of the pressurizer model was performed by comparing the simulation results with those from the tests at the Maanshan nuclear power plant. Four start-up tests of the Maanshan nuclear power plant are collected and simulated: 1) turbine trip test from 100% power; 2) large-load reduction at 100% power; 3) net-load trip at 100% power; and 4) net-load trip at 50% power. The simulation results are in reasonable agreement with the start-up tests, and thus the pressurizer model built in this study is successfully verified and validated.

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.

scholarly journals Application of reverse osmosis at NPP and verification of the process for primary coolant treatment in temelín nuclear power

2021 ◽  
Vol 7 (2) ◽  
pp. 1-7
Author(s):  
Skala M. ◽  
Kůs P. ◽  
Kotowski J. ◽  
Kořenková H.
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Boric Acid ◽  
Reverse Osmosis ◽  
Nuclear Power ◽  
Power Plants ◽  
Pure Water ◽  
Feed Solution ◽  
Primary Coolant ◽  
Purification System

Drained primary coolant from nuclear power plants containing boric acid is currently treated in the system of evaporators and by ion exchangers. Reverse osmosis as an alternative process to evaporator was investigated. Using reverse osmosis, the feed primary coolant is separated into two output streams: retentate and permeate. Retentate stream consists of concentrated boric acid solution together with other components, while permeate stream consists of purified water. In the first phase ofthe project the reverse osmosis modules from several manufactures were tested on a batch laboratory apparatus. Certain modifications to the pH of the feed solution were needed to enable the tested membranes to concentrate the H3BO3 in the retentate stream, separate from the pure water in the permeate stream. Furthermore, the separation capability for other compounds present in primary coolant such as K, Li or NH3 were evaluated. In the final phase of the project the pilot-plant unit of reverse osmosis was tested in nuclear power plant Temelín. It was installed in the Special Purification System SVO-6 for the regeneration of boric acid. The aim of the tests performed in Temelín nuclear power plant was to verify possible use of reverse osmosis for the treatment of primary coolant.

scholarly journals Аnalysis Catalytic Hydrogen Recombiner Capacity Calculation Taking into Account Conditions Inside Sealed Enclosure of Containment Safety System of Nuclear Power Plants with Water-Water Energetic Reactor

2021 ◽  
Vol 64 (2) ◽  
pp. 178-186
Author(s):  
V. V. Sorokin
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Gas Exchange ◽  
Nuclear Power ◽  
Power Plants ◽  
Reference Conditions ◽  
Exchange Capacity ◽  
Primary Coolant ◽  
Radioactive Substances ◽  
Catalytic Hydrogen

Localizing safety systems are provided to contain radioactive substances in an accident and attenuate ionizing radiation at a modern nuclear power plant. Together with radioactive substances, hydrogen is also retained, which is formed during the decomposition of the primary coolant. The accumulation of hydrogen in the presence of oxygen from the atmosphere in the accident localization zone carries the danger of the formation of flammable and explosive concentrations of these components. Nuclear power plant (NPP) deigns with water-water energetic reactor (WWER) provides for a hydrogen removal system including passive catalytic hydrogen recombiners. The device capacity  is confirmed experimentally under reference conditions (lean air-hydrogen mixture, pressure and temperature close to normal, no interference with gas exchange). Capacity is an important safety parameter. In the event of an accident, conditions inside the ealed enclosure of the localizing system of NPP with WWER can  differ from the reference  ones and affect the capacity.  On the basis of calculations, the operation of recombiners with lack of  oxygen  and with hindered  gas exchange has been investigated in the paper. The decrease in capacity with lack of oxygen reaches 50 %, which is mainly  caused by an increase in underburning. Compared to the reference conditions, the effect is more pronounced in the event of an accident – 60–70 %. The hindered gas exchange is modeled by a decrease in the height of recombiner traction channel. This case can be reduced to the placement of the device in cramped conditions and the effect of the atmosphere speed inside the enclosure. Regardless of the hydrogen concentration, the operating characteristic of the device remains linear, with a two-fold decrease in height leads to a decrease in capacity by 20 %. The results can be used to substantiate the safety of NPPs with WWER and to review on the safety subtantiation of power units.

scholarly journals Validation and Evaluation of COSINE Software Package Based on the Operation Data of Third Generation Passive Pressurized Water Reactor

2021 ◽  
Author(s):  
Weibin Zhang ◽  
Chenglin Zhu ◽  
Qiao Zhang ◽  
Linlin Xu ◽  
Guoping Quan
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Software Development ◽  
Nuclear Power ◽  
Power Plants ◽  
Hot Spot ◽  
Third Generation ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor

According to the historical experience of international nuclear power software development and the requirements of relevant guidelines at home and abroad, a large number of experiments and theoretical work must be carried out to verify and confirm the empirical formulas, models and calculation methods used in the software and evaluate the models related to safety evaluation in order to make the software be applied to the design and analysis of nuclear power plants. Validation and evaluation is the most important key link in the process of nuclear power software development, which is heavy workload and difficult, and needs a lot of actual power plant operation data. This paper proposed a research on the validation and evaluation of the COSINE software package’s calculation capability and accuracy based on the operation data of the third generation passive PWR (Pressurized Water Reactor) AP1000. The comparison results between the operation limit parameters of the nuclear power plant including critical boron concentration, heat pipe factor of nuclear enthalpy rise, heat flux hot spot factor and AO (Axial Offset) showed that the data calculated by COSINE met the running requirements of the nuclear power plant, and the calculation accuracy keeps also in a good way.

Aging Evaluation for the Extension of Qualified Life of Nuclear Power Plant Equipment Through Modulated Poisson Processes

2009 ◽  
Author(s):  
Pedro L. C. Saldanha ◽  
P. F. Frutuoso e Melo
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Point Processes ◽  
Nuclear Regulatory Commission ◽  
Repairable Systems ◽  
Pressurized Water ◽  
Trade Offs ◽  
Service Water

This paper presents an application of the Modulated Point Process (MPPP) to the service water pumps of a typical pressurized water reactor, as a model for the rate of occurrence of failures (ROCOF) of a repairable system, in order to decide for an extension of the qualified life in the context of a nuclear power plant license renewal. The analysis was carried considering some field data spanning a 2,300 calendar day period, which is approximately equivalent to four burn-up cycles and refueling periods. The reliability is estimated, and maintenance strategies are discussed. As a conclusion, the MPPP is adequate for modeling the rate of occurrence of failures that are time dependent, and can be used where aging mechanisms are present in the operation of repairable systems. This means that equipment characteristics that are important may be inserted into the model and the results can help make decisions in the context of maintenance programs, as is the case with the maintenance rule concept that has been proposed by the Nuclear Regulatory Commission and is about to be implemented in nuclear power plants in Brazil, for instance. Trade-offs on the difficulty with data acquisition for applying point processes are discussed.

Protection of Nuclear Power Plants Against Severe Winds: Impact of the Local Building Configuration on the Wind Speed to Consider

2020 ◽  
Author(s):  
Arièle Défossez ◽  
Eric Dupont ◽  
Laurence Grammosenis ◽  
Hervé Cordier ◽  
Tiphaine Le Morvan
Keyword(s):  
Wind Speed ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Flow Behavior ◽  
Severe Weather ◽  
Wind Rose ◽  
Pressurized Water ◽  
Weather Events

Abstract Over the years, power plants have been hit by numerous severe weather events (storm, flood, heat wave...). EDF (Electricity of France) and ASN (Nuclear Safety Authority) want to assess the future impact of severe weather events on the power plants. Furthermore, recent research on storms estimates more accurate wind speed return values than before. For this reason, the severe wind value is an important parameter to quantify on a NPP (Nuclear Power Plant) site, in order to verify if the protection measures are sufficient or, if necessary, to design adequate protection. To cope with those objectives, wind flow behavior around a PWR (Pressurized Water Reactor) nuclear power plant is studied. The goal of this work is to check that there is no exceeding local wind speed relative to the wind entering the site. The severe winds are characterized locally near the buildings in terms of location and amplitude. Different kind of topology for the nuclear power plant sites are studied in the project: near a cliff, in a plain or in a basin. In our study, the CFD (Computational Fluid Dynamics) open source tool Code_Saturne developed at EDF-R&D is used to simulate the wind over a French PWR site located in nearly flat terrain in a plain. The 3D mesh includes buildings of the site. Several wind directions corresponding to the prevailing winds are studied. Two wind speeds corresponding to wind speed return values are studied (eg: the inlet wind speed is 25 m/s at 10 meter high for a return period of 50 years). Furthermore, several locations selected near buildings are studied carefully. Swirling flows have been viewed between buildings. Analysis of the results shows that the wind speed near the buildings does not exceed the wind speed at the entrance of the domain for the three directions studied except near the cooling towers and above buildings. However, this result should not be generalized to other PWR sites due to the specificities of each site such as relief, buildings position, buildings size, roughness, wind rose... This methodology could be applied at other nuclear power plant sites.

scholarly journals On Singular Perturbation of Neutron Point Kinetics in the Dynamic Model of a PWR Nuclear Power Plant

Sci ◽  
2020 ◽  
Vol 2 (2) ◽  
pp. 30
Author(s):  
Xiangyi Chen ◽  
Asok Ray
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Singular Perturbation ◽  
Nuclear Power ◽  
Short Communication ◽  
Primary Coolant ◽  
Variable Step ◽  
Perturbed Model ◽  
Point Kinetics ◽  
Coolant System

This short communication makes use of the principle of singular perturbation to approximate the ordinary differential equation (ODE) of prompt neutron (in the point kinetics model) as an algebraic equation. This approximation is shown to yield a large gain in computational efficiency without compromising any significant accuracy in the numerical simulation of primary coolant system dynamics in a PWR nuclear power plant. The approximate (i.e., singularly perturbed) model has been validated with a numerical solution of the original set of neutron point-kinetic and thermal–hydraulic equations. Both models use variable-step Runge–Kutta numerical integration.

scholarly journals On Singular Perturbation of Neutron Point Kinetics in the Dynamic Model of a PWR Nuclear Power Plant

Sci ◽  
2020 ◽  
Vol 2 (2) ◽  
pp. 36
Author(s):  
Xiangyi Chen ◽  
Asok Ray
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Singular Perturbation ◽  
Nuclear Power ◽  
Short Communication ◽  
Primary Coolant ◽  
Variable Step ◽  
Perturbed Model ◽  
Point Kinetics ◽  
Coolant System

This short communication makes use of the principle of singular perturbation to approximate the ordinary differential equation (ODE) of prompt neutron (in the point kinetics model) as an algebraic equation. This approximation is shown to yield a large gain in computational efficiency without compromising any significant accuracy in the numerical simulation of primary coolant system dynamics in a PWR nuclear power plant. The approximate (i.e., singularly perturbed) model has been validated with a numerical solution of the original set of neutron point-kinetic and thermal–hydraulic equations. Both models use variable-step Runge–Kutta numerical integration.

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