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.

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.

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.

Purification of a primary coolant in a nuclear power plant using a magnetic filter ? electrodeionization hybrid separation system

2005 ◽  
Vol 262 (3) ◽  
pp. 725-732 ◽  
Author(s):  
Jung-Hoon Song ◽  
Min-Chul Song ◽  
Kyeong-Ho Yeon ◽  
Jung-Bae Kim ◽  
Kun-Jai Lee ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Separation System ◽  
Primary Coolant ◽  
Magnetic Filter

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.

Study of Station Blackout Accident Process in Taishan Nuclear Power Plant

2013 ◽  
Author(s):  
Yu Li ◽  
Huiyong Zhang ◽  
Yehong Liao ◽  
Jiming Lin ◽  
Dekui Zhan
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Heat Removal ◽  
Final State ◽  
Primary Coolant ◽  
Station Blackout ◽  
Full Power ◽  
Diesel Generators

According to the design features of the nuclear power plant in Taishan, a station blackout (SBO) at full power is a complex sequence, induced by the loss of offsite power (LOOP) combined with the loss of the emergency diesel generators (EDGs). This paper shall be performed the deterministic safety analysis of SBO without PSA study, and it mainly analyzes an overheating event of primary coolant resulting in the temporary unavailability of the steam generators (SGs) feeding systems during the station blackout. The analysis of this accident is carried out using the CATHARE2 thermal hydraulic code. The analysis pointed out that the final state can be reached after the operator performs the main manual actions such as startup of the SBO diesel generators from the main control room; startup of the two ASG pumps of trains 1 and 4; Local opening of the ASG header downstream the ASG pumps to enable the two power supplied ASG pumps to feed all the four SGs; initiation of a manual cooldown via the VDA in order to ensure the long term protection of the RCP pumps seals with respect to the thermal and mechanical loads. It corresponds to the achievement of a stable heat removal conditions by the emergency feedwater system (ASG) and the atmospheric steam dump system (VDA).

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