Swelling of VVER-1000 Core Baffle: Numerical Modeling and Direct Measurement of its Geometrical Dimensions

2017 ◽  
Author(s):  
Andrii Oryniak ◽  
Igor Orynyak
Keyword(s):  
Numerical Modeling ◽  
Power Plants ◽  
Reactor Core ◽  
Calculated Data ◽  
Vertical Direction ◽  
Vertical Axis ◽  
Hot Water ◽  
Mathematical Treatment ◽  
Brittle Strength ◽  
Steel 08Kh18n10t

Traditionally, the brittle strength evaluation of reactor pressure vessel was the central issue in lifetime assessment of Ukrainian nuclear power plants (NPPs). The problem of swelling of the reactor core baffle only recently got due attention from the side of operator. Here the most efforts were given on numerical modeling of austenitic steel 08Kh18N10T swelling and its effect on induced stresses in core baffle and distortion of its geometry. The calculation shows that essential changing of core baffle dimensions is expected after 35–40 years of operation. Eventually this can lead to the contact with the core barrel. Yet, these predictions contain the big number of uncertainties related to the input data used in analysis: fluence distribution; temperature variation due to heat release induced by neutron and gamma radiation; thermal-hydraulic boundary condition between the baffle and coolant; and, especially, the adopted law of swelling in dependence with above factors as well as mechanical stresses. So, the second task was to measure the real geometry of baffle after 27 years of operation, to determine its change and compare these results with the numerically calculated data with accounting for the design tolerances. Thus, the spatial measurement system (SMS) equipped with ultrasonic gages was designed. It contains the central vertical beam which can move in vertical direction and rotate. To the lower end of the beam four horizontal levels are attached, which are equipped with device resistant to the hot water and radiation. The gages are used to measure the shortest distances to the edges of baffle. Two types of results were obtained. The first one are the measurements in the different horizontal planes obtained by rotation the SMS around the vertical axis with angular steps equal to 1 degree. These results were difficult to handle with and required a special mathematical treatment due to the possible shift of the centre of measurement. The second set of measurements was performed by moving the SMS in vertical direction. These data demonstrate the change of distance with the height. The results clearly show that problem of swelling do exists, and, in general, the measured patterns of the distortions along the vertical and angular coordinates correspond to numerically obtained results. Further work on baffle integrity is however needed.

ANALYSIS OF THERMAL DISPERSION IN SELAT SEMBILAN WATERS, NORTH SUMATRA

2018 ◽  
Vol 7 (3) ◽  
Author(s):  
Amiral Aziz
Keyword(s):  
Numerical Modeling ◽  
Power Plants ◽  
Sea Water ◽  
Cooling System ◽  
Water System ◽  
Cooling Water ◽  
Hot Water ◽  
Thermal Dispersion ◽  
Coal Fire ◽  
Steam Power Plants

Pangkalan Susu Coal Fire Steam Power Plants (CFSPP) are planned to build in Tanjung Pasir Village, Langkat Regency taking sea water as cooling condenser of power plants, and through it back into the sea. To explore the possibilities of re-entry of the circulation of hot water to the intake, it is necessary to study the termal dispersion within the framework of those plans. In this study, numerical modeling to determine termal distribution pattern that comes out from CFSPP outlet. From the study it could be concluded that with the given intake inlet – discharge outfall design configuration and under the worst scenario, cooling water system of Pangkalan Susu unit 3 & 4 is safe i.e. re-circulation of warmcooling water will not happen. The Pangkalan Susu unit 3 & 4 will consume intact sea water (30.5oC). However, the Pangkalan Susu unit 1 & 2 will be influenced by the warm cooling system that may decrease its cooling system efficiency since its inlet is covered by shattered sea water (31.2oC – 32.2oC).keywords : thermal dispersion, power plant. numerical modeling

Flow and Temperature Fields Measurement Inside Rod Bundle by the Combined Use of PIV and LIF Technique

2018 ◽  
Author(s):  
Xing Li ◽  
Sichao Tan ◽  
Zhengpeng Mi ◽  
Peiyao Qi ◽  
Yunlong Huang
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Great Influence ◽  
Reactor Core ◽  
Transfer Characteristic ◽  
Index Of Refraction ◽  
Temperature Fields ◽  
Spacer Grid ◽  
Rod Bundle ◽  
Combined Use

Thermal hydraulic research of reactor core is important in nuclear energy applications, the flow and heat transfer characteristics of coolant in reactor fuel assembly has a great influence on the performance and safety of nuclear power plants. Particle image velocimetry (PIV) and Laser induced fluorescence (LIF) are the instantaneous, non-intrusive, whole-field fluid mechanics measuring method. In this study, the simultaneous measurement of flow field and temperature field for a rod bundle was conducted using PIV and LIF technique. A facility system, utilizing the matching index of refraction approach, has been designed and constructed for the measurement of velocity and temperature in the rod bundle. In order for further study on complex heat and mass transfer characteristic of rod bundle, the single-phase experiments on the heating conditions are performed. One of unique characteristics of the velocity and temperature distribution downstream the spacer grid was obtained. The experimental results show that the combined use of PIV and LIF technique is applied to the measurement of multi-physical field in a rod bundle is feasible, the measuring characteristics of non-intrusive ensured accuracy of whole field data. The whole field experimental data obtained in rod bundle benefits the design of spacer grid geometry.

Expectations for Inservice Testing Programs at New Nuclear Power Plants

2017 ◽  
Author(s):  
Thomas G. Scarbrough
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Reactor Core ◽  
Nuclear Regulatory Commission ◽  
Lessons Learned ◽  
Mechanical Equipment ◽  
American Society ◽  
Regulatory Commission ◽  
Safety Systems

In a series of Commission papers, the U.S. Nuclear Regulatory Commission (NRC) described its policy for inservice testing (IST) programs to be developed and implemented at nuclear power plants licensed under 10 CFR Part 52. This paper discusses the expectations for IST programs based on those Commission policy papers as applied in the NRC staff review of combined license (COL) applications for new reactors. For example, the design and qualification of pumps, valves, and dynamic restraints through implementation of American Society of Mechanical Engineers (ASME) Standard QME-1-2007, “Qualification of Active Mechanical Equipment Used in Nuclear Power Plants,” as accepted in NRC Regulatory Guide (RG) 1.100 (Revision 3), “Seismic Qualification of Electrical and Active Mechanical Equipment and Functional Qualification of Active Mechanical Equipment for Nuclear Power Plants,” will enable IST activities to assess the operational readiness of those components to perform their intended functions. ASME has updated the Operation and Maintenance of Nuclear Power Plants (OM Code) to improve the IST provisions for pumps, valves, and dynamic restraints that are incorporated by reference in the NRC regulations with applicable conditions. In addition, lessons learned from performance experience and testing of motor-operated valves (MOVs) will be implemented as part of the IST programs together with application of those lessons learned to other power-operated valves (POVs). Licensee programs for the Regulatory Treatment of Non-Safety Systems (RTNSS) will be implemented for components in active nonsafety-related systems that are the first line of defense in new reactors that rely on passive systems to provide reactor core and containment cooling in the event of a plant transient. This paper also discusses the overlapping testing provisions specified in ASME Standard QME-1-2007; plant-specific inspections, tests, analyses, and acceptance criteria; the applicable ASME OM Code as incorporated by reference in the NRC regulations; specific license conditions; and Initial Test Programs as described in the final safety analysis report and applicable RGs. Paper published with permission.

The analysis of emergency situations related to fires at nuclear power plants

2021 ◽  
Vol 30 (5) ◽  
pp. 66-75
Author(s):  
S. A. Titov ◽  
N. M. Barbin ◽  
A. M. Kobelev
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Emergency Situations ◽  
The Usa ◽  
Large Fires ◽  
Short Circuits ◽  
Nuclear Reactor Core

Introduction. The article provides a system and statistical analysis of emergency situations associated with fires at nuclear power plants (NPPs) in various countries of the world for the period from 1955 to 2019. The countries, where fires occurred at nuclear power plants, were identified (the USA, Great Britain, Switzerland, the USSR, Germany, Spain, Japan, Russia, India and France). Facilities, exposed to fires, are identified; causes of fires are indicated. The types of reactors where accidents and incidents, accompanied by large fires, have been determined.The analysis of major emergency situations at nuclear power plants accompanied by large fires. During the period from 1955 to 2019, 27 large fires were registered at nuclear power plants in 10 countries. The largest number of major fires was registered in 1984 (three fires), all of them occurred in the USSR. Most frequently, emergency situations occurred at transformers and cable channels — 40 %, nuclear reactor core — 15 %, reactor turbine — 11 %, reactor vessel — 7 %, steam pipeline systems, cooling towers — 7 %. The main causes of fires were technical malfunctions — 33 %, fires caused by the personnel — 30 %, fires due to short circuits — 18 %, due to natural disasters (natural conditions) — 15 % and unknown reasons — 4 %. A greater number of fires were registered at RBMK — 6, VVER — 5, BWR — 3, and PWR — 3 reactors.Conclusions. Having analyzed accidents, involving large fires at nuclear power plants during the period from 1955 to 2019, we come to the conclusion that the largest number of large fires was registered in the USSR. Nonetheless, to ensure safety at all stages of the life cycle of a nuclear power plant, it is necessary to apply such measures that would prevent the occurrence of severe fires and ensure the protection of personnel and the general public from the effects of a radiation accident.

Numerical modeling and optimization of thermal stratification in solar hot water storage tanks for domestic applications: CFD study

Solar Energy ◽  
2017 ◽  
Vol 157 ◽  
pp. 441-455 ◽  
Author(s):  
T. Bouhal ◽  
S. Fertahi ◽  
Y. Agrouaz ◽  
T. El Rhafiki ◽  
T. Kousksou ◽  
...  
Keyword(s):  
Numerical Modeling ◽  
Thermal Stratification ◽  
Water Storage ◽  
Hot Water ◽  
Storage Tanks ◽  
Modeling And Optimization

scholarly journals Assessment of water hammer effects on boiling water nuclear reactor core dynamics

2007 ◽  
Vol 22 (1) ◽  
pp. 18-33 ◽  
Author(s):  
Anis Bousbia-Salah
Keyword(s):  
Pressure Wave ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Water Hammer ◽  
Pressure Wave Amplitude ◽  
Sensitivity Studies ◽  
Rapid Transients ◽  
Complex Phenomena

Complex phenomena, as water hammer transients, occurring in nuclear power plants are still not very well investigated by the current best estimate computational tools. Within this frame work, a rapid positive reactivity addition into the core generated by a water hammer transient is considered. The numerical simulation of such phenomena was carried out using the coupled RELAP5/PARCS code. An over all data comparison shows good agreement between the calculated and measured core pressure wave trends. However, the predicted power response during the excursion phase did not correctly match the experimental tendency. Because of this, sensitivity studies have been carried out in order to identify the most influential parameters that govern the dynamics of the power excursion. After investigating the pressure wave amplitude and the void feed back responses, it was found that the disagreement between the calculated and measured data occurs mainly due to the RELAP5 low void condensation rate which seems to be questionable during rapid transients. .

scholarly journals A Project To Estimate The Power Capacity Of Geothermal Power Plants Based On The Enthalpy Of Geothermal Fluid And Plant Design

2021 ◽  
Author(s):  
Obumneme Oken
Keyword(s):  
Power Plant ◽  
Geothermal Energy ◽  
Electricity Generation ◽  
Power Plants ◽  
Hot Water ◽  
Power Capacity ◽  
Geothermal Fluid ◽  
Direct Heating ◽  
Single Flash ◽  
Geothermal Power

Nigeria has some surface phenomena that indicate the presence of viable geothermal energy. None of these locations have been explored extensively to determine the feasibility of sustainable geothermal energy development for electricity generation or direct heating. In this context, the present study aims to provide insight into the energy potential of such development based on the enthalpy estimation of geothermal reservoirs. This particular project was conducted to determine the amount of energy that can be gotten from a geothermal reservoir for electricity generation and direct heating based on the estimated enthalpy of the geothermal fluid. The process route chosen for this project is the single-flash geothermal power plant because of the temperature (180℃) and unique property of the geothermal fluid (a mixture of hot water and steam that exists as a liquid under high pressure). The Ikogosi warm spring in Ekiti State, Nigeria was chosen as the site location for this power plant. To support food security efforts in Africa, this project proposes the cascading of a hot water stream from the flash tank to serve direct heat purposes in agriculture for food preservation, before re-injection to the reservoir. The flowrate of the geothermal fluid to the flash separator was chosen as 3125 tonnes/hr. The power output from a single well using a single flash geothermal plant was evaluated to be 11.3 MW*. This result was obtained by applying basic thermodynamic principles, including material balance, energy balance, and enthalpy calculations. This particular project is a prelude to a robust model that will accurately determine the power capacity of geothermal power plants based on the enthalpy of fluid and different plant designs.

scholarly journals Analysis of numerical studies on the thermal-hydraulic and neutronic thermal-hydraulic stability of supercritical water reactors

2021 ◽  
Vol 7 (4) ◽  
pp. 311-318
Author(s):  
Artavazd M. Sujyan ◽  
Viktor I. Deev ◽  
Vladimir S. Kharitonov
Keyword(s):  
Supercritical Water ◽  
Nuclear Power ◽  
Power Plants ◽  
Negative Impact ◽  
Reactor Core ◽  
Resistance Coefficient ◽  
Coolant Flow ◽  
Water Reactor ◽  
Hydraulic Stability ◽  
Supercritical Water Reactor

The paper presents a review of modern studies on the potential types of coolant flow instabilities in the supercritical water reactor core. These instabilities have a negative impact on the operational safety of nuclear power plants. Despite the impressive number of computational works devoted to this topic, there still remain unresolved problems. The main disadvantages of the models are associated with the use of one simulated channel instead of a system of two or more parallel channels, the lack consideration for neutronic feedbacks, and the problem of choosing the design ratios for the heat transfer coefficient and hydraulic resistance coefficient under conditions of supercritical water flow. For this reason, it was decided to conduct an analysis that will make it possible to highlight the indicated problems and, on their basis, to formulate general requirements for a model of a nuclear reactor with a light-water supercritical pressure coolant. Consideration is also given to the features of the coolant flow stability in the supercritical water reactor core. In conclusion, the authors note the importance of further computational work using complex models of neutronic thermal-hydraulic stability built on the basis of modern achievements in the field of neutron physics and thermal physics.

scholarly journals Experimental Study on Effect of Water Spray Characteristics on Performance of Cooling Tower

2020 ◽  
Vol 170 ◽  
pp. 01009
Author(s):  
Akshay S. Dhurandhar ◽  
Amarsingh B. Kanase-Patil
Keyword(s):  
Power Plants ◽  
Cooling Tower ◽  
Thermal Power ◽  
Evaporative Cooling ◽  
Hot Water ◽  
Spray Angle ◽  
Counter Flow ◽  
Steam Power ◽  
To Come ◽  
Spiral Type

Cooling tower is an indispensable part, used as a direct contact type heat exchanger mainly for evaporative cooling. Cooling tower generally dissipates, remove heat from thermal power plants. In an induced draft cooling tower of counter flow, used for a mini-steam power plant, hot water enters at the top, while the air is introduced at the bottom and exits at the top, air is allowed to come in contact with falling water droplets, causing evaporative cooling. A possibility of desired change with different spray angle, patterns, is tried and analysed. On findings, best suited spray nozzle angle resulted is 90°, and amongst three spray patterns, full cone, hollow cone and spiral type nozzle; full cone nozzle of 90° spray angle helps achieving efficiency up to 82%. The range increases successively from 9.8°C to 15.5°C for FC nozzle, in approach to WBT; the desirable fall of 3.56°C is attained with effectiveness of 81.63%.

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