MWe Nuclear Power Plant Research Articles

Far-Field Influences Shadow the Effects of a Nuclear Power Plant’s Discharges in a Semi-Enclosed Bay

The sustainable development of society depends on the reliable supply of electricity while keeping impacts on the environment to a minimum. A 951 MWe nuclear power plant in the semi-enclosed Nanwan Bay at the southern tip of Taiwan began operating in May 1984. Part of the bay is in Kenting National Park, which is known for its coral reefs and abundant marine life; thus, thermal pollution from the cooling water discharge is a great concern. Fortunately, the bay opens south to face the Luzon Strait, where the world’s strongest internal tides are generated. Because the bay is deep enough, internal waves bring up cold deep water and reduce the surface temperature by as much as 10 °C for a few hours every day. These internal waves and topographically generated upwelling also bring nutrients to the euphotic layer from the depths, but the upwelled waters quickly leave the bay along with the cooling water. As a result, a thermal plume with a temperature of 1 °C or higher than the ambient temperature only covers 1 km. By way of comparison, El Niño—Southern Oscillation- or Pacific Decadal Oscillation-related interannual variations in temperature are as high as 5 °C. The rapid turnover of the upwelled waters also helps to prevent heat released by the power plant from accumulating and diminishes the thermal stress, thus sustaining corals and other marine life forms. Typhoons, even hundreds of kilometers away, could also induce the upwelling of cold subsurface water. Consecutive typhoons have been observed to reduce the water surface temperature by up to 10 °C for two weeks or longer. Furthermore, the currents are such that the thermal plume flows out of the bay most of the time. All of these factors make the surface waters in the bay about 0.5 °C cooler than the waters outside of the bay, despite the operation of a nearby nuclear power plant.

Long-term prediction of safety parameters with uncertainty estimation in emergency situations at nuclear power plants

The correct situation awareness (SA) of operators is important for managing nuclear power plants (NPPs), particularly in accident-related situations. Among the three levels of SA suggested by Ensley, Level 3 SA (i.e., projection of the future status of the situation) is challenging because of the complexity of NPPs as well as the uncertainty of accidents. Hence, several prediction methods using artificial intelligence techniques have been proposed to assist operators in accident prediction. However, these methods only predict short-term plant status (e.g., the status after a few minutes) and do not provide information regarding the uncertainty associated with the prediction. This paper proposes an algorithm that can predict the multivariate and long-term behavior of plant parameters for 2 h with 120 steps and provide the uncertainty of the prediction. The algorithm applies bidirectional long short-term memory and an attention mechanism, which enable the algorithm to predict the precise long-term trends of the parameters with high prediction accuracy. A conditional variational autoencoder was used to provide uncertainty information about the network prediction. The algorithm was trained, optimized, and validated using a compact nuclear simulator for a Westinghouse 900 MWe NPP.

Control Rod Modeling and Worth Calculation for a Typical 1100 MWe Nuclear Power Plant Using WIMS/D4 and CITATION

A typical 1100 MWe pressurized water reactor (PWR) is a second unit installed at the coastal site of Pakistan. In this paper, verification analysis of reactivity control worth by means of rod cluster control assemblies (RCCAs) for startup and operational conditions of this typical nuclear power plant (CNPP) has been performed. Neutronics analysis of fresh core is carried out at beginning of life (BOL) to determine the effect of grey and black control rod clusters on the core reactivity for startup and operating conditions. The combination of WIMS/D4 and CITATION computer codes equipped with JENDL-3.3 data library is used for the first time for core physics calculations of neutronic safety parameters. The differential and integral worth of control banks is derived from the computed results. The effect of control bank clusters on core radial power distribution is studied precisely. Radial power distribution in the core is evaluated for numerous configurations of control banks fully inserted and withdrawn. The accuracy of computed results is validated against the reference values of Nuclear Design Report (NDR) of 1100 MWe typical CNPP. It has been observed that WIMS-D4/CITATION shows its capability to effectively calculate the reactor physics parameters.

Development of a diagnostic algorithm for abnormal situations using long short-term memory and variational autoencoder

It is recognized that an abnormal situation diagnosis is a challenging task for nuclear power plant (NPP) operators because of the excessive information and high workload in such situations. To help operators, several studies have proposed operator support systems using artificial intelligence techniques. However, those methods could neither assess anonymous cases as an unknown situation nor confirm whether its outputs are reliable or not. In this study, an algorithm that can confirm the diagnosis results and determine unknown situations using long short-term memory (LSTM) and variational autoencoder (VAE) is proposed. LSTM was adopted as the primary network for diagnosing abnormal situations. Meanwhile, VAE-based assistance networks were added to the algorithm to ensure that the credibility of the diagnosis is estimated via the anomaly score-based negative log-likelihood. The algorithm was tested and implemented using the compact nuclear simulator for the Westinghouse 900 MWe NPP. Furthermore, the simulation also considered noise-added data.

Adaptive overcurrent protection scheme suitable for station blackout power supply of nuclear power plant operated through an integrated microgrid

Inclusion of distributed energy resources based integrated microgrid (IMG) into emergency station blackout (SBO) (EM-SBO) power system of nuclear power plant (NPP) has complicated its existing protection scheme due to varying system fault level with bidirectional flow of power. During SBO in NPP, it is of topmost priority to safeguard the reactor through EM-SBO supply. However, if any severe fault occurs, it may lead to complete collapse of the safety power supply in the NPP. So, availability of this tiny power system is most important as compared to the tripping of reactor safety of the NPP. In this paper, a 460 MWe NPP station has been considered to implement IMG into its EM-SBO power system and focuses on its overcurrent protection coordination scheme as a case study. The combined fuse and numerical relays inter-linked with digital logic-based adaptive overcurrent protection (AOP) scheme has been proposed after load flow analysis and calculating short circuit fault level through ETAP software. The sensitivity, the selectivity and the speed of the protection scheme have been evaluated in different network configurations like on-grid, semi-islanding and islanding mode of operation of IMG. Finally, it is proposed that the protection schemes vary from case-to-case basis leading to necessity of communication among the relays. So, an advanced AOP scheme with digital logic linked through peer-to-peer Hi-speed communication (IEC 61,850 confirmed GOOSE technology) has been proposed to resolve the protection issue of the modern NPPs operated through IMG.

A Study on Thermal-Hydraulics Characteristics for Designing a Shell and Tube Conderser for a 1200 MWe Nuclear Power Plant

The study explores the thermal-hydraulics parameters of a condenser of a nuclear power plant with 1200MWe net electric output and 37% thermal efficiency using empirical correlations of pressure drop and heat transfer coefficient both for the tube and shell sides. Considering a two-phase fluid system, a shell and tube condenser with coolant water on the tube side and condensing steam on the shell side has been selected. For designing a condenser with a thermal load of 2060MWth, the input temperature data of cold fluid inlet and outlet temperatures are taken as 29.4ºC and 40ºC while the condensation temperature is taken as 53.97oC. Transverse, two-pass condenser with 4 shell tanks has been considered in this study and the length of each shell tank is taken as 14m. Based on these input data, this work finds heat transfer area, logarithmic mean temperature difference (LMTD), and convection heat transfer coefficient inside the tubes as 549536m2, 18.74°C, and 2869.85W/m2.ºC respectively for 20mm tube outer diameter. Hydrodynamic parameters relating to the friction factors and pressure drops on tube side are found as 0.031 and 14.86kPa respectively. Similar design data have been generated for varying coolant inlet temperatures and tube inner diameters. Results reveal that velocity of flow inside the tubes as well as the number of tubes in a bundle decrease with the increase in tube diameter. Finally, the thermal-hydraulic data may be used to design a large scale commercial condenser to be applicable for a large scale nuclear plant since limited design data are available in the literature.
 Journal of Bangladesh Academy of Sciences, Vol. 43, No. 2, 181-189, 2019

Application of integrated microgrid for strengthening the station blackout power supply in nuclear power plant

This paper analyses the emergency station blackout (EM-SBO) power supply arrangement of nuclear power plants (NPPs) pre and post Fukushima nuclear disaster at Japan. The entire world has realized that, enhancement of existing level of safety power supply is necessary for safe shutdown of reactor. In line with above, this paper is proposing inclusion of renewable energy based micro grid (MG) concept into the NPPs. A typical 500 MWe NPP has been considered for conceptual analysis of MG by integrating with EM-SBO bus as case study. The proposed MG concept has been analyzed through load flow and transient stability study by using accelerated Gauss-Seidel method in ETAP software (version-16.0.0). Based on the analysis reported, rate of change of frequency for the islanded MG is very high. To overcome this issue, relay based logic is suggested to infer that, by introducing integrated MG concept, the EM-SBO power supply may be strengthened.

Fluid-elastic instability evaluation for reactor vessel internals with structural interaction

Provided fluid velocity reaches a specific threshold value, the vibrational amplitude of the corresponding structure increases rapidly and becomes unstable. Even though nuclear industrial concerns have been focused primarily on heat exchangers, steam generators and pipes, evaluation of flow-induced vibration is also required for safety-related components such as reactor vessel internals (RVIs). For this matter, the fluid-elastic instability of RVIs in a typical 1000 MWe nuclear power plant (NPP) was examined by enhanced methods taking into account numerical capability and uncertainties. First, computational fluid dynamics (CFD) analyses of the whole geometry were conducted to assess the flow characteristics with and without consideration of fuel assembly under various NPP conditions. Based on the resulting data, upper guide structure (UGS) was selected as the most critical among five RVI subcomponents. Second, detailed modal analyses of the UGS and attached multiple tubes were performed using both added mass and fluid-structure interaction models to find out natural frequencies and corresponding mode shapes, and to investigate effect of a flooded structure. Third, fluid-elastic instability was evaluated based upon the results from the CFD and modal analyses, and compared with a stability diagram. Thereby, it was confirmed that the RVIs considered in this study had sufficient safety margin against the fluid-elastic instability under representative conditions. Moreover, the calculated stability ratios increased when either the simple model was employed without consideration of the fuel assembly as the current common practice or fluid-structure interaction effect was incorporated.

ENERGI NUKLIR SEBAGAI SOLUSI UNTUK MENGHAMBAT PEMANASAN GLOBAL

Global warming is the increase in the average temperature of the earth surface, atmosphere and oceans.The global warming in recent years has been international issues. The issues come to the surfacebecause global warming has the very big impact to the world and the lives of animal, plant and human,such as world climate change. The main cause of global warming is the combustion of fossil fuel suchas coal, oil and natural gas, that released carbon dioxide and other gases to atmosphere as greenhousegases. One of alternative to retard this global warming is by replacing fossil fuel with utilization of nuclearenergy for power plant. As a comparison, a 1,000 MWe nuclear power plant as a substitute for coal fi redpower plant at the same capacity, will reduce 6,000,000 tons of CO2 gas emission per year. Consideringenvironmental aspect, the nuclear power plant is not emitting CO2 gas, so that the use of nuclear powerplant can retard the global warming. Considering economic aspect, based on operational experienceof nuclear power plants in advanced countries, it is shown that cost of generating electricity of nuclearpower plants is more competitive than fossil fuel power plant. Considering safety aspect, nuclear powerplant operating in the world, have passed by a technological test. They have also an excellent operationreliability and a very good safety system.

Dynamic characteristics assessment of reactor vessel internals with fluid-structure interaction

Abstract Improvement of numerical analysis methods has been required to solve complicated phenomena that occur in nuclear facilities. Particularly, fluid-structure interaction (FSI) behavior should be resolved for accurate design and evaluation of complex reactor vessel internals (RVIs) submerged in coolant. In this study, the FSI effect on dynamic characteristics of RVIs in a typical 1,000 MWe nuclear power plant was investigated. Modal analyses of an integrated assembly were conducted by employing the fluid-structure (F-S) model as well as the traditional added-mass model. Subsequently, structural analyses were carried out using design response spectra combined with modal analysis data. Analysis results from the F-S model led to reductions of both frequency and Tresca stress compared to those values obtained using the added-mass model. Validation of the analysis method with the FSI model was also performed, from which the interface between the upper guide structure plate and the core shroud assembly lug was defined as the critical location of the typical RVIs, while all the relevant stress intensities satisfied the acceptance criteria.

Analysis on containment depressurization under severe accidents for a Chinese 1000 MWe NPP

Containment depressurization has been implemented for many nuclear power plants (NPPs) to mitigate the risk of containment overpressurization induced by steam and gases released in LOCA accidents or generated in molten core concrete interaction (MCCI) during severe accidents. Two accident sequences of large break loss of coolant accident (LB-LOCA) and station blackout (SBO) are selected to evaluate the effectiveness of the containment venting strategy for a Chinese 1000 MWe NPP, including the containment pressure behaviors, which are analyzed with the integral safety analyses code for the selected sequences. Different open/close pressures for the venting system are also investigated to evaluate CsI mass fraction released to the environment for different cases with filtered venting or without filtered venting. The analytical results show that when the containment sprays can't be initiated, the depressurization strategy by using the Containment Filtered Venting System (CFVS) can prevent the containment failure and reduce the amount of CsI released to the environment, and if CFVS is closed at higher pressure, the operation interval is smaller and the radioactive released to the environment is less, and if CFVS open pressure is increased, the radioactive released to the environment can be delayed. Considering the risk of high pressure core melt sequence, RCS depressurization makes the CFVS to be initiated 7 h earlier than the base case to initiate the containment venting due to more coolant flowing into the containment.

The challenges of introducing nuclear power in the Gulf Cooperation Council states

Nuclear power is a mature industry with an incredible record of safety and reliability, without the emission of greenhouse gases, and one that is becoming economically attractive for private investment. The current interest in nuclear power for meeting future electricity and seawater desalination needs in the Gulf Corporation Council (GCC) states and other countries in the Middle East is prudent, logical and timely. An achievable goal for the GCC states would be to secure 30% of their future needs of electricity and of process heat for industrial applications and seawater desalination from nuclear power by 2030. This is equivalent to completing the construction of two 1500 MWe nuclear power plants each year starting in 2016. Accomplishing this goal requires a multifaceted approach to addressing many challenges that include: 1) stimulating private investments in nuclear energy; 2) establishing specialised higher education and vocational training programmes; 3) maintaining close cooperation with the International Atomic Energy Agency (IAEA) and countries with advanced nuclear capabilities; 4) establishing a viable heavy industry; 5) collaborating with other countries in the Middle East on the development of a common electric grid; 6) investing in the mining and exploration of uranium resources; 7) identifying and licensing suitable sites for construction of nuclear plants; 8) establishing a regulatory and safety board that has a government oversight and an effective technical and RD 9) considering standardisation versus diversification in nuclear reactor types.

Blade Stress of the Reactor Coolant Pump of 300 MWe Nuclear Power Plant in China Based on Fluid-solid Coupling Method

In order to compute and analyze the blade stress in reactor coolant pump of a 300 MWe nuclear power plant in China,from the viewpoint of safety performance,the fluid-solid coupling technique is adopted to solve the fluid and solid coupling equation.Theoretical analysis shows that the blade stresses mainly include tensile stress caused by centrifugal force,bending and torsional stresses induced by flow-field pressure,and thermal stress brought by temperature field.From the analysis results,it is concluded that the max equivalent(Von-Mises) stress exists in fixed supports and the blade stress does not have strict periodicity,and the disordered distributions of blade stress lies in the complex internal flow.The differences between impeller and guide vane are proved by comparison of theory and simulation and the main factors are explained.Lastly the intensity examination proves that the intensity of the reactor coolant pump meets the requirement of ASME.This could be used for improving aerofoil design and ensuring performance and intensity.

On the introduction of nuclear power in Middle East countries: Promise, strategies, vision and challenges

Middle eastern countries have been experiencing economical growth at an annual rate of more than 5% and some of the highest population growth rates in the world. Many (>13) have recently expressed interest in nuclear power for meeting future demands of electricity and fresh water. A challenging, but possible goal, particularly for the states of the Gulf Cooperation Council (GCC) and for the countries in North Africa, would be to secure 30% of future needs of electricity and process heat for industrial applications and seawater desalination from nuclear power by 2030. For the GCC states alone, this would be equivalent to building two new, 1500 MWe nuclear power plants each year starting in 2016. Some of the challenges and important consideration include: (a) stimulating private investments, (b) establishing education and training programs that are among the world’s best; (c) maintaining close cooperation with the International Atomic Energy Agency (IAEA) and countries with advanced nuclear technology and fuel cycle capabilities to ensure safety and compliance at all levels, (c) establishing viable indigence heavy industry and international alliances on all aspects of the nuclear fuel cycle; (d) coordinating with other countries in the region on the development of a common electric grid; (e) investing in uranium mining and exploration to secure future resources, (f) identifying and licensing suitable sites for future construction of nuclear plants; (g) establishing a regulatory and safety board with government oversight and sound regulations that would make it possible to build and operate new nuclear power plants within 40–50 months, (h) investing in high technology and R&D infrastructure, and (i) investigating the options of standardization versus diversification in the nuclear reactor types.

Fire PSA of the French PWR 900 MWe series

Abstract The PSA approach described in part 1 is also used to assess the core damage frequency due to a fire in the control room of a 900 MWe Nuclear Power plant. The specifics of the control room include the cable failure modes due to the fire (spurious order in particular), two different kinds of fire detection systems in the control room and the use of controls and instrumentation of the remote shutdown panel to reach the safe shutdown state if the control room has to be evacuated. This paper will focus on the detailed study of critical compartments and on the results and insights of this study. The purpose of this paper is also to present the methodology used to assess the core damage frequency due to a fire in the control room of a French Nuclear Power Plant, some qualitative results and the main insights further to the discussions with the utility.

0490 A NOVEL ACUTE ADMISSIONS AND ASSESSMENT SERVICE FOR TEENAGERS WITH FIRST EPISODE PSYCHOSIS

As our industrial and infrastructure systems (refineries, power plants, pipelines, etc.) age, there is a considerable economic incentive to avoid unscheduled outages and to extend operation beyond the design lifetime. The avoidance of unscheduled outages is of particular interest, because the failure of even a minor component can result in the complete shutdown of a facility. For example, the unscheduled shutdown of a 1000 MWe nuclear power plant due to the failure of a valve may cost the operator between $1 million and $3 million per day, depending upon the cost of replacement power and other factors. However, if component failures could be accurately predicted, maintenance could be performed during scheduled outages, the cost of which has already been built into the price of the product, thereby minimizing the economic impact of the failure. With regard to life extension, the successful extension of operation beyond the design life translates into enhanced profits and the avoidance of costly licensing and environmental impact assessments associated with the development and construction of a new facility. However, in this case as well, the key to successful operation is the ability to avoid unscheduled downtime and hence maintain continuity of production. However, eventually, the frequency and severity of unscheduled outages render continued operation uneconomic and, at that point, replacement of the facility becomes necessary.Corrosion is a major cause of component failure, and hence the occurrence of unscheduled downtime, in complex industrial systems. In particular, the various forms of localized corrosion, including pitting corrosion, crevice corrosion, stress corrosion cracking (SCC), and corrosion fatigue (to name the common forms), are particularly deleterious, because they frequently occur without any outward sign of accumulating damage and because they often result in sudden and catastrophic failures. Thus, the development of effective general and especially localized corrosion damage prediction technologies is essential for the successful avoidance of unscheduled downtime and for the successful implementation of life extension strategies.

Assessment of integrity of components in piping of 500 MWe PHWR: using R-6 method

The primary heat transport system of 500 MWe Indian PHWR comprises of straight pipes, elbows and headers. A study was conducted to qualify piping system for leak before break. R-6 method was used to assess integrity of system for leakage size crack (LSC), the margins on crack initiation load and unstable crack growth loads. Option 2 (material specific failure assessment diagram), Category 3 (ductile tearing) analysis was used for straight pipes, elbows and header. In order to enhance the confidence in the analytical results, detailed sensitivity analysis was also performed. For sensitivity analysis, variation in material properties, LSC was considered. The effect of variation in temperature on material properties was also considered. Tensile and fracture properties used for base and weld material data were generated from pipe material obtained from 220 MWe Nuclear Power Plant, under construction.

Water requirements of an inland nuclear power station engineering and economic aspects

Water requirements for construction and operation of a 950 MWe nuclear power plant are estimated. For the bulk of water needed, i.e. condenser cooling, quantities assessed are 12.5 × 10 6 m 3/yr for sea water and 35 × 10 6 m 3/yr for fresh water. Breakeven water costs are calculated below which use of wet cooling towers is more economical than dry cooling towers. Conclusions are that desalinated water costs (70¢/m 3) are lower than breakeven costs ($1.42/m 3) i.e. it is likely to be more economical to desalt water using some form of wet, rather than dry, cooling towers. However it is even more economical to use sea water makeup without desalination but with proper design and choice of materials for cooling towers.

Conceptual design and cost study for a dual-purpose nuclear-electric reverse osmosis seawater conversion plant

This paper presents the results of a study to develop a conceptual design, and cost estimate for a 25 million gallon per day seawater reverse osmosis desalting plant operating at both Caribbean and Persian Gulf sites. The plants operate in conjunction with a 1000 MWe nuclear power plant. Four seawater membrane manufacturers were supplied with feedwater analyses and a simplified cost estimating procedure so that they could recommend an optimum membrane system. From this information, plant designs and cost estimates were developed. For both sites, a two-stage system was selected for the conceptual cost estimate. The product water cost was determined, based- upon 1978 construction costs, for both the Caribbean, and Persian Gulf sites. In addition, areas of potential cost reduction were discussed.

The bushehr 200,000 M 3/D desalination plant

The present fresh water supply in the South of Iran is fully committed, and the demands of the two 1200 MWe Nuclear Power Plants complex at Bushehr makes the construction and operation of 2 × 100,000 m 3/d desalination plants feasible. The procedure for inviting tenders together with a description of the proposed plant is given here. The method of coupling the desalination units to the Nuclear Power Plants and provision of safety features are discussed. The paper concludes with a cost analysis of the whole project.