Nuclear Power Plant Research Papers (original) (raw)

In summer 1993 we measured the transfer of (239/240)Pu to milk from herbage from a pasture located 5 km from the Chernobyl nuclear power plant. In one treatment cows were allowed to graze freely on the pasture. In a second treatment, cows were fed herbage collected from the pasture in stalls. The milk transfer coefficient; F(m) did not vary significantly between treatments and the mean value of 7.5x10(-6)d l(-1) was higher than previously reported values. Despite many values of F(m) for Pu in the literature we identified few relevant original data sets. Transfer coefficient values for Pu are only appropriate when used in conjunction with a specified time period or an appropriate model which allows for the biological half-life. We recommend for screening purposes an F(m) value of 1x10(-5)d l(-1) for Pu, with an order of magnitude lower value being appropriate for cows which are only exposed for one grazing season.

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- Geography, Ukraine, Biological Sciences, Seasonality
- by Sakil Kim
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- Computer Science, Situation awareness, Human Factors, Crew Resource Management

The simulation of nuclear power plant accident conditions requires three-dimensional (3D) modeling of the reactor core to ensure a realistic description of physical phenomena. The operational flexibility of Light Water Reactor (LWR) plants can be improved by utilizing accurate 3D coupled neutronics/thermal-hydraulics calculations for safety margins evaluations. There are certain requirements to the coupling of thermal-hydraulic system codes and neutron-kinetics codes that ought to be considered. The objective of these requirements is to provide accurate solutions in a reasonable amount of CPU time in coupled simulations of detailed operational transient and accident scenarios. These requirements are met by the development and implementation of six basic components of the coupling methodologies: ways of coupling (internal or external coupling); coupling approach (integration algorithm or parallel processing); spatial mesh overlays; coupled time-step algorithms; coupling numerics (explicit, semi-implicit and implicit schemes); and coupled convergence schemes. These principles of the coupled simulations are discussed in details along with the scientific issues associated with the development of appropriate neutron cross-section libraries for coupled code transient modeling. The current trends in LWR nuclear power generation and regulation as well as the design of next generation LWR reactor concepts along with the continuing computer technology progress stimulate further development of these coupled code systems. These efforts have been focused towards extending the analysis capabilities as well as refining the scale and level of detail of the coupling. This article analyses the coupled phenomena and modeling challenges on both global (assembly-wise) and local (pin-wise) levels. The issues related to the consistent qualification of coupled code systems as well as their application to different types of LWR transients are presented. Finally, the advances in numerical and computation techniques for coupled code simulations are summarized with outlining remaining challenges.

- by Kostadin Ivanov
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- Kinetics, Parallel Processing, Level Of Detail (LOD), Nuclear power

In order to help nuclear power plant operator reduce his cognitive load and increase his available time to maintain the plant operating in a safe condition, transient identification systems have been devised to help operators identify possible plant transients and take fast and right corrective actions in due time. In the design of classification systems for identification of nuclear power plants transients, several artificial intelligence techniques, involving expert systems, neuro-fuzzy and genetic algorithms have been used. In this work we explore the ability of the Particle Swarm Optimization algorithm (PSO) as a tool for optimizing a distance-based discrimination transient classification method, giving also an innovative solution for searching the best set of prototypes for identification of transients. The Particle Swarm Optimization algorithm was successfully applied to the optimization of a nuclear power plant transient identification problem. Comparing the PSO to similar methods found in literature it has shown better results.

- by Roberto Schirru
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- Genetic Algorithm, Cognitive Load, Nuclear power, Artificial Intelligent

This study examined to what extent nuclear risk perceptions, organizational commitment (OC), and appraisals of management are associated with each other among nuclear power plant personnel. The sample consisted of 428 nuclear power plant workers who completed a questionnaire at their workplace. Perceived nuclear risk and OC were most closely related to the appraisals of the top management of the organization. As the trust in and satisfaction with the top management increased, perceived nuclear safety and acceptance of the organizational goals and values heightened. This result is discussed in the context of industrial safety management.

- by Raija Kalimo
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- Public Opinion, Organizational Commitment, Multidisciplinary, Risk Analysis
- by Erik Berglund
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- Information Systems, Computer Science, Distributed Computing, Technology
- by Robert Broadwater
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- Controller Design, Control Systems, Power Plant, Control system
- by A. Qualls
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- Level Of Detail (LOD), System Design, Communication System, Nuclear power

The results of post-irradiation examinations of a pressure tube of fuel channel No. G-12 of KANUPP have been described. A detailed study was made in Canada by AECL. A parallel investigation on its seven rings of about 50 mm length each was also carried out at PINSTECH. Visual inspection showed normal oxidation effects. Gamma spectrometry showed the presence of 95Zr and 95Nb. Microstructural study revealed the characteristic alpha plus a transformed beta phase structure.

- by rafaqat ali
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- Materials Engineering, Nuclear Materials, Nuclear, Microstructures

Fish communities and habitat structures were evaluated by underwater visual censuses a rocky location impacted by thermal discharge (I) and at two control locations, one in a Sargassum bed (C1) and the other in a rocky shore with higher structural complexity (C2). Habitat indicators and fish communities exhibited significant differences between the impacted and control locations, with the impacted one

- by Francisco Gerson Araujo
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- Brazil, Biodiversity, Marine And Environmental Pollution, Water Pollution

The experiment LACE LA4 on thermal-hydraulics and aerosol behavior in a nuclear power plant containment, which was performed in the LACE experimental facility, was simulated with the ASTEC CPA module of the severe accident computer code ASTEC V1.2. The specific purpose of the work was to assess the capability of the module (code) to simulate thermal-hydraulic conditions and aerosol behavior

- by Ivo Kljenak
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- Boundary Condition, Severe Accident, Nuclear Power Plant, light water reactor

An evaluation of human factors in a new nuclear power plant was conducted prior to the beginning of any business operations. After the task analysis and observation of training, two stages of interviews were carried out with the operators in the Fourth Nuclear Power Plant (NPP4). The main concerns identified were problems resulting from the operating interface of the display and controls in the main control room, usability of procedures, and the layout of the main control room.The latent human errors and suggestions were listed, and the top three problems were analyzed. The operators indicated that the alarm design issues and the critical problem of the operating mode with the VDU were worth further study in order to provide suggestions for a new interface design for future power plants.

- by Sheue-ling Hwang
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- Human Factors, Interface Design, Task analysis, Interdisciplinary Engineering

The Human System Simulation Laboratory (HSSL) at the Idaho National Laboratory is one of few facilities of its kind that allows human factors researchers to evaluate various aspects of human performance and human system interaction for proposed reactor designs and upgrades. A basic system architecture, physical configuration and simulation capability were established to enable human factors researchers to support multiple, simultaneous simulations and also different power plant technologies. Although still evolving in terms of its technical and functional architecture, the HSSL is already proving its worth in supporting current and future nuclear industry needs for light water reactor sustainability and small modular reactors. The evolution of the HSSL is focused on continual physical and functional refinement to make it a fully equipped, reconfigurable facility where advanced research, testing and validation studies can be conducted on a wider range of reactor technologies. This requires the implementation of additional plant models to produce empirical research data on human performance with emerging human-system interaction technologies. Additional beneficiaries of this information include system designers and HRA practitioners. To ensure that results of control room crew studies will be generalizable to the existing and evolving fleet of US reactors, future expansion of the HSSL may also include other SMR plant models, plant-specific simulators and a generic plant model aligned to the current generation of pressurized water reactors (PWRs) and future advanced reactor designs. Collaboration with industry partners is also proving to be a vital component of the facility as this helps to establish a formal basis for current and future human performance experiments to support nuclear industry objectives. A long-range Program Plan has been developed for the HSSL to ensure that the facility will support not only the Department of Energy's Light Water Reactor Sustainability Program, but also to provide human factors guidance for all future developments of the nuclear industry.

- by Jacques Hugo
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- Human Factors, New Technology, Human Performance, System Simulation

Belgium started its nuclear programme quite early. The first installations were constructed in the fifties, and presently, more than 55 % of the Belgian electricity production is provided by nuclear power plants. After 30 years of nuclear experience, Belgium started decommissioning of nuclear facilities in the eighties with two main projects: the BR3-PWR plant and the Eurochemic reprocessing plant. The BR3-decommissioning project is carried out at the Belgian Nuclear Research Centre, while the decommissioning of the former Eurochemic reprocessing plant is managed and operated by Belgoprocess n.v., which is also operating the centralised waste treatment facilities and the interim storage for Belgian radioactive waste. Some fundamental principles have to be considered for the management of materials resulting from the decommissioning of nuclear installations, equipment and/or components, mainly based on the guidelines of the “IAEA-Safety Fundamentals. The Principles of Radioactive Was...

- by Lucien Teunckens
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- Engineering, Industrial Automation, Waste Management, Nuclear reactor

In tasks requiring sustained attention, human alertness varies on a minute time scale. This can have serious consequences in occupations ranging from air tra c control to monitoring of nuclear power plants. Changes in the electroencephalographic (EEG) power spectrum accompany these uctuations in the level of alertness, as assessed by measuring simultaneous changes in EEG and performance on an auditory monitoring task. By combining power spectrum estimation, principal component analysis and arti cial neural networks, we show that continuous, accurate, noninvasive, and near real-time estimation of an operator's global level of alertness is feasible using EEG measures recorded from as few as two central scalp sites. This demonstration could lead to a practical system for noninvasive monitoring of the cognitive state of human operators in attention-critical settings.

A new procedure for probabilistic seismic risk assessment of nuclear power plants (NPPs) is proposed. This procedure modifies the current procedures using tools developed recently for performance-based earthquake engineering of buildings. The proposed procedure uses (a) response-based fragility curves to represent the capacity of structural and nonstructural components of NPPs, (b) nonlinear response-history analysis to characterize the demands on those components, and (c) Monte Carlo simulations to determine the damage state of the components. The use of response-rather than ground-motion-based fragility curves enables the curves to be independent of seismic hazard and closely related to component capacity. The use of Monte Carlo procedure enables the correlation in the responses of components to be directly included in the risk assessment. An example of the methodology is presented in a companion paper to demonstrate its use and provide the technical basis for aspects of the methodology.► A new procedure is proposed for seismic probabilistic risk assessment of NPPs. ► It uses response-based fragility, response-history analysis, Monte Carlo simulation. ► An example for the proposed procedure is presented in a companion paper.

- by Andrew S Whittaker
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- Seismic Hazard, Monte Carlo Simulation, Monte Carlo, Risk assessment

Most structures and equipment used in nuclear power plant and process plant, such as reactor internals, fuel rods, steam generator tubes bundles, and process heat exchanger tube bundles, are subjected to flow-induced vibrations (FIV). Costly plant shutdowns have been the source of motivation for continuing studies on cross-flow-induced vibration in these structures. Damping has been the target of various research attempts related to FIV in tube bundles. A recent research attempt has shown the usefulness of a phenomenon termed as 'thermal damping'. The current paper focuses on the modeling and analysis of thermal damping in tube bundles subjected to cross-flow. It is expected that the present attempt will help in establishing improved design guidelines with respect to damping in tube bundles.

- by Zaffar Khan
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- Engineering, Materials Science, Mechanics, Engineering Design

Earthquakes are of great concerns to nuclear engineers. The consequences of earthquakes on nuclear reactors may even be the "shutdown" of nuclear power plant. Rather than increasing the structure's stiffness, one way to reduce the earthquake damages is the use of anti-vibration mechanisms. This paper deals with numerical evaluation of the efficiency of anti-vibration mechanisms applied to typical frame structures under earthquake. The most common vibration isolators and controllers are shortly described. To show the effectiveness of some anti-vibration mechanisms, an application example is presented. The building structure is modeled by finite elements, an anti-vibration mechanism is placed at the building base with special finite element, and an artificial earthquake equivalent to El Centro is generated and applied at the building base. The behavior of the frame, with and without anti-vibration mechanisms, is compared. The results in terms of floor absolute and relative displacements and shear forces at the base are reported.

- by Luciano Mendes Bezerra
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- Computer Science, Vibration Control, Finite Element, Vibration İsolation

This paper presents the development and application of a methodology to estimate, using conventional deterministic lattice/core analysis methods, the fast neutron fluence at the tips of control rods inserted during operation in PWR reactors. The developed methodology is based on a 3-D Nodal Diffusion/2-D Lattice Transport multi-step calculation scheme, in which the operating history of the nuclear environment around the tip is tracked in 3-D core follow analyses and used thereafter to provide boundary conditions for 2-D transport calculations to compute the fast neutron flux. In subsequent steps, radial and axial correction factors, both based on fast flux results from the 3-D core simulator, are applied to the calculated 2-D transport flux in order to take into account radial leakage effects as well as axial flux gradients around the tip. The fluence is finally estimated through a time-integration of the corrected 2-D transport fast flux. The developed methodology has been applied to estimate the fluence for a total of 15 control rods, over 21 operating cycles of a Swiss nuclear power plant.

An algorithm is presented for the automated analysis of rotating probe multifrequency eddy current data obtained from nuclear power plant steam generator tubes (SGT). The algorithm consists of four steps, namely, a preprocessing stage for conditioning the data, a decision tree based feature extraction stage for identifying relevant features for analysis, a neural network based classification stage for identifying signals from various defect types and benign structures, and finally a blind deconvolution based characterization stage for accurately estimating the size and orientation of the detected defects. This algorithm is optimized to maximize the probability of detection (POD), while keeping the number of false alarms (PFA) at a minimum. Initial results presented in this paper look very promising and demonstrate the effectiveness of the proposed algorithm.

- by Robi Polikar
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- Neural Network, Feature Extraction, Decision Tree, Eddy Current

In-core temperature measurements are pivotal in maintaining nuclear reactors in a safe state of operation. Thermocouples serve as the liaison in ensuring this safe state. The realization of the thermocouple's full potential is hindered by the fact that thermocouples cannot be situated in areas with high radiation fields. Radiation has the potential of generating voltages in the thermocouple wires, hence producing an error in the temperature transmitter output. In this paper, a mathematical model is developed to quantify the effect that radiation from the Canada Deuterium Uranium (CANDU) Nuclear Power Plants (NPPs) has on the thermocouple temperature reading. Subsequently, a method to offset the effect of radiation on the thermocouple is proposed. Simulation is performed to verify the effectiveness of the proposed system.

- by fawaz ali
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- Nuclear power, Deuterium, Nuclear reactor, Temperature measurement
- by haydar bayar
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- Engineering, Sustainable Development, Alternative Energy, Nuclear Energy

Reducing the unavailability of safety systems at nuclear power plants, by utilizing the probabilistic safety assessment (PSA) methodology, is one of the prime goals in the nuclear industry. In that sense, optimization of test and maintenance activities, which are defined within the technical specifications, represents quite popular and interesting domain. Obtaining optimal test and maintenance schedule is of great significance for improving system availability and performance as well as plant availability in general.

- by Marko Čepin and +1
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- Chemical Engineering, Monte Carlo Simulation, Genetic Algorithm, Case Study

On-line maintenance (OLM) represents the term, which includes testing and maintenance that is performed when the main generator of the nuclear power plant is connected to the grid. OLM on one side helps to decrease the number of activities, which would be performed during the scheduled outage, but on the other side it may contribute to a different level of risk, if the activity is performed when the plant is operating. If the risk of OLM during the power operation is much larger than the risk of similar activity performed during the shutdown the OLM may not be the desired strategy. Additionally, if the risk of OLM during the power operation is not larger than the risk of similar activity performed during the shutdown, the OLM would be the preferred strategy.

- by Marko Čepin
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- Decision Making, Power Plant, Interdisciplinary Engineering, Risk Evaluation

This paper presents recent Canadian advances in nuclear-based production of hydrogen by electrolysis and the thermochemical copper-chlorine (Cu-Cl) cycle. This includes individual process and reactor developments within the Cu-Cl cycle, thermochemical properties, advanced materials, controls, safety, reliability, economic analysis of electrolysis at off-peak hours, and integrating hydrogen plants with Canada's nuclear power plants.

- by Peter Tremaine and +1
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- Engineering, Copper, Economic analysis, Hydrogen Energy

This paper illustrates the work carried out by EDF within the framework of ISP 48 post-test analysis of NUPEC/NRC 1:4-scale model of a prestressed pressure containment vessel of a nuclear power plant [Hessheimer, M.F., Klamerus, E.W., Rightly, G.S., Lambert, L.D., Dameron, R.A., 2003. Overpressurization test of a 1:4-scale prestressed concrete containment vessel model. NUREG/CR-6810, SAND2003-0840P. Sandia National Laboratories, Albuquerque, NM]. EDF as a participant of the International Standard Problem no. 48 [Mathet, E., Hessheimer, M., Ali, S., Tegeler, B., 2005.

- by Alexis Courtois
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- Finite Element, prestressed Concrete, Research and Development, Lessons Learned

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 through an energy analysis based on the first law of thermodynamics to gain some new insights into the plant performance. The change in the cooling water temperature can be experienced due to the seasonal changes in climatic conditions at plant site. It can also come into the question of design processes for the plant site selection. In the analysis, it is considered that the condenser vacuum varies with the temperature of cooling water extracted from environment into the condenser. The main findings of the paper is that the impact of 18C increase in temperature of the coolant extracted from environment is predicted to yield a decrease of 0.45and0.45 and 0.45and0.12% in the power output and the thermal efficiency of the pressurized-water reactor nuclear-power plant considered, respectively.

- by Oguz Salim Sogut
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- Mechanical Engineering, Chemical Engineering, Energy, Water Temperature

A new classification method, for isolating steam generator tube defects in nuclear power plants using Eddy Current Test (ECT) signals, has been developed. The method uses Self-Organizing maps (SOM) with different data signatures to identify and classify these defects. A multiple inference system is proposed which evaluates different extracted characteristic SOMs to infer the defect type. Wavelet zero-crossing representation, a linear predictive coding (LPC), and other basic signal representations, such as magnitude and phase, are used to construct characteristic vectors that combine one or more of these features. These vectors are evaluated for their ability to classify tube defects and the ones with the best performance are used in the multiple inference system. The effectiveness of the method is demonstrated by applications of the characteristic maps to ECT data from various cases of tube defects in pressurized water reactor plant steam generators. The developed algorithm enables real-time applications such as fast tube defects classification systems and visualization of ECT signal feature prototypes, which may improve the speed of time-critical decision making during power plant maintenance outages.

- by eduardo L L cabral
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- Distributed Computing, Decision Making, Power Plant, Computer Software

Contemporary trend is to use simulation for understanding or training for a physical process. Nuclear simulators are complex systems to model. Control logic is an important part of simulation, and it is useful for turning various equipment on/off, interlocks handling and alarm triggering. This paper reports development of software for automating various chores required by the designers of Full Scope Training Simulator of CHUSHNUPP-2 (FSTS C-2). Designer may arrange components in a special manner as required by control logic information in the system manuals. The software provides online simulation of circuits made, and when design is complete, the file may be saved in the format required by simulation software, which does not have designer facility. The software is developed using latest software techniques and vector graphics. Vector graphics enable printing as drawing at desired scale. Existing control logic circuits developed for FSTS C-1 have been tested and the results have been found accurate.

- by Muhammad Irfan
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- Complex System, Computer Control, Nuclear Power Plant, Simulation Software

The use of an optical fiber as a distributed sensor for detecting, locating, and (with suitable signal processing) classifying intruders is proposed. Phase changes resulting from either the pressure of the intruder on the ground immediately above the buried fiber or from seismic disturbances in the vicinity are sensed by a phase-sensitive optical time-domain reflectometer (φ−OTDR). Light pulses from a cw laser with a narrow (kHz range) instantaneous linewidth and low (MHz/min range) frequency drift are injected into one end of the single mode fiber, and the backscattered light is monitored with a photodetector. Results of analyses and experimental studies to establish the feasibility of the concept are described. Simulations predict a range of 10 km with 35 m range resolution and 30 km with 90 m range resolution. Experiments indicate adequate (several π-rad) phase changes are produced by intruders on foot for burial depths in the 20 -40 cm range in sand and in clay soils. A phase perturbation in a fiber has been detected and located in a laboratory demonstration of the φ-OTDR using an Er:fiber laser as the light source. This technology could in a cost-effective manner provide enhanced perimeter security for nuclear power plants, electrical power distribution centers, storage facilities for fuel and volatile chemicals, communication hubs, airports, government offices, military bases, embassies, and national borders.

- by Juan Juarez
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- Signal Processing, Fiber Laser, Chemical Communication, Nuclear power

The aim of this paper is to examine the cognitive processes through which operators make decisions when dealing with microincidents during their actual work, and to determine whether they use a naturalistic or normative decision making strategy. That is, do they try to recognize a microincident as familiar and base decisions on pattern recognition, tacit knowledge, or condition-action rules (naturalistic), or do they need to concurrently compare and contrast options, before selecting the best possible according standard operating procedures (normative)? The method employed for data collection was a cognitive task analysis (CTA) based on operators' activities. The main finding of this research was that decision making is primarily based on naturalistic strategies. These findings contrast the normative behavior prescribed by the organization's work design and their standards of competency for training and evaluation operators work. Relevance to industry: This study presents a situated method to describe how sharp end operators make decisions during microincidents that occurs in normal operation, emphasizing how the sociotechnical environment affects their cognitive strategies, which is one of the basic steps for an organization that wants to enhance the safety culture. r

- by Mario Vidal
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- Tacit Knowledge, Decision Making, Naturalistic Decision Making, Pattern Recognition

The modular high-temperature gas-cooled nuclear reactor (MHTGR) is seen as one of the best candidates for the next generation of nuclear power plants. China began to research the MHTGR technology at the end of the 1970s, and a 10 MWth pebble-bed high-temperature reactor HTR-10 has been built. On the basis of the design and operation of the HTR-10, the high-temperature gas-cooled reactor pebble-bed module (HTR-PM) project is proposed. One of the main differences between the HTR-PM and HTR-10 is that the ratio of height to diameter corresponding to the core of the HTR-PM is much larger than that of the HTR-10. Therefore it is not proper to use the point kinetics based model for control system design and verification. Motivated by this, a nodal neutron kinetics model for the HTR-PM is derived, and the corresponding nodal thermal-hydraulic model is also established. This newly developed nodal model can reflect not only the total or average information but also the distribution information such as the power-distribution as well. Numerical simulation results show that the static precision of the new core model is satisfactory, and the trend of the transient responses is consistent with physical rules.

- by Zhe Dong
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- Kinetics, Numerical Simulation, Nuclear reactor, Control System Design

The accident at the Chernobyl nuclear power plant was the worst industrial accident of the last century that involved radiation. The unprecedented release of multiple different radioisotopes led to radioactive contamination of large areas surrounding the accident site. The exposure of the residents of these areas was varied and therefore the consequences for health and radioecology could not be reliably estimated quickly. Even though some studies have now been ongoing for 25 years and have provided a better understanding of the situation, these are yet neither complete nor comprehensive enough to determine the long-term risk. A true assessment can only be provided after following the observed population for their natural lifespan. Here we review the technical aspects of the accident and provide relevant information on radioactive releases that resulted in exposure of this large population to radiation. A number of different groups of people were exposed to radiation: workers involved in the initial clean-up response, and members of the general population who were either evacuated from the settlements in the Chernobyl nuclear power plant vicinity shortly after the accident, or continued to live in the affected territories of Belarus, Russia and Ukraine. Through domestic efforts and extensive international co-operation, essential information on radiation dose and health status for this population has been collected. This has permitted the identification of high-risk groups and the use of more specialised means of collecting information, diagnosis, treatment and follow-up. Because radiation-associated thyroid cancer is one of the major health consequences of the Chernobyl accident, a particular emphasis is placed on this malignancy. The initial epidemiological studies are reviewed, as are the most significant studies and/or aid programmes in the three affected countries.

- by Yuri Demidchik
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- Ukraine, Thyroid Cancer, Clinical oncology, Health Status

One of the fundamental challenges in studying cognitive systems in context is how to generalize beyond the specific work setting studied. How does one progress beyond the details of a single case to derive a more general understanding of... more

- by Catherine Burns
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- Psychology, Cognitive Science, Human Factors, Process Control

An optimization system based on Genetic Algorithms (GAs), in combination with expert knowledge coded in heuristics rules, was developed for the design of optimized boiling water reactor (BWR) fuel loading patterns. The system was coded in a computer program named Loading Pattern Optimization System based on Genetic Algorithms, in which the optimization code uses GAs to select candidate solutions, and the core simulator code CM-PRESTO to evaluate them. A multi-objective function was built to maximize the cycle energy length while satisfying power and reactivity constraints used as BWR design parameters. Heuristic rules were applied to satisfy standard fuel management recommendations as the Control Cell Core and Low Leakage loading strategies, and octant symmetry. To test the system performance, an optimized cycle was designed and compared against an actual operating cycle of Laguna Verde Nuclear Power Plant, Unit I.

During normal operation of PWRs, routine fuel rods failures result in release of radioactive fission products (RFPs) in the primary coolant of PWRs. In this work, a stochastic model has been developed for simulation of failure time sequences and release rates for the estimation of fission product activity in primary coolant of a typical PWR under power perturbations. In the first part, a stochastic approach is developed, based on generation of fuel failure event sequences by sampling the time dependent intensity functions. Then a three-stage model based deterministic methodology of the FPCART code has been extended to include failure sequences and random release rates in a computer code FPCART-ST, which uses state-of-the-art LEOPARD and ODMUG codes as its subroutines. The value of the 131 I activity in primary coolant predicted by FPCART-ST code has been found in good agreement with the corresponding values measured at ANGRA-1 nuclear power plant. The predictions of FPCART-ST code with constant release option have also been found to have good agreement with corresponding experimental values for time dependent 135 I, 135 Xe and 89 Kr concentrations in primary coolant measured during EDITHMOX-1 experiments.

- by muhammad iqbal
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- Materials Engineering, Nuclear Materials, Stochastic Simulation, Time Dependent

Due to the large power supply in the energy market since 1960s, the nuclear power planets have been consistently constructed throughout the world in order to maintain and supply sufficient fundamental power generation. Up to now, most of the planets have been operated to a point where the spent fuel pool has reached its design capacity volume. To prevent the plant from shutdown due to the spent fuel pool exceeding the design capacity, the dry cask storage can provides a solution for both the spent fuel pool capacity and the mid-term storage method for the spent fuel bundles at nuclear power planet.

- by Kuen Ting
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- Mechanical Engineering, Modeling, Risk assessment, Quantitative analysis

It is clear that probabilistic risk assessment or probabilistic safety assessment is embedded in the safety culture of the nuclear power industry, worldwide. Risk assessment applications are in transition in the sense that the regulatory apparatus is not yet in place, at least in the United States of America (USA), to fully support a risk-based licensing process. There is progress on the regulatory front, but it is tedious and not without its frustrations. Currently, the strategy in the USA is a “risk-informed” approach that tends to be “business as usual”, but while you're at it, “do a risk assessment”. The result is added burden (and costs) at a time of increased competition in the power field as a result of deregulation. There is hope in that some steps are finally being taken to modify the regulations to allow risk assessment to be more of a part of the licensing process. For example, the regulations having to do with maintenance, plant changes, and technical specifications have been modified to allow insights from risk assessment to be part of the basis for licensing amendments. On the matter of standards there is strong support that is scientifically based and addresses such issues as health effects and environmental impacts. There appears to be less support for standards on such matters as definition of terms, methodology, and data requirements.

- by Eko Susanto
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- Engineering, Risk assessment, Safety Culture, Nuclear power

One of the most complex cases for assessing the nuclear power plant safety is the evaluation of the response of the plant to an earthquake and calculation of the core damage frequency related with this. Plant level fragilities are convolved with the seismic hazard curves to obtain a set of doublets for the plant damage state. The standard methodology of the description of randomness and epistemic uncertainty of the fragility is based on the use of lognormal distribution. In the practice, because of large number and variety of types of components, variety of failure modes, further simplification is needed in spite of simplicity of the mathematic description of the fragility and its uncertainty. Sophisticated modeling and screening methods have to be applied for plant fragility development requiring enormous experience. Several practical assumptions utilized in the seismic PSA showing certain analogy with interval type description of uncertainties. In the paper an attempt is made for outlining some new options for nuclear power plant seismic fragility development based on the interval and p-box concept. The possibility for derivation of conditional probability of failure for cumulative absolute velocity is also highlighted.

- by Tamas Janos Katona
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- Nuclear Engineering, Uncertainty, Nuclear Power Plant

We investigated levels of DNA damage in blood cells of barn swallows (Hirundo rustica) inhabiting the Chernobyl region to evaluate whether chronic exposure to low-level radioactive contamination continues to induce genetic damage in free-living populations of animals. Blood samples were obtained from barn swallows collected at sites with different background levels of radiation, including a relatively uncontaminated area. The extent of DNA damage was evaluated using the alkaline (pH = 12.1) comet assay, a robust and sensitive electrophoresis-based technique widely employed in research ranging from biomonitoring to clinical studies. We found that levels of DNA damage, as indexed by the extent of DNA migration, were increased in barn swallows living in areas surrounding Chernobyl when compared to swallows sampled at low-level sites. The results we obtained are consistent with previous findings on this same species, which showed that swallows breeding in areas heavily contaminated with radionuclides have increased mutation rates, higher oxidative stress and incidence of morphological aberrations and tumors. Overall, these results indicate that chronic exposure to radioactive contaminants, even 20 years after the accident at the Chernobyl nuclear power plant, continues to induce DNA damage in cells of free-living animals.

- by Michael Wyatt and +1
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- Genetics, Geography, Biomarkers, Oxidative Stress

ÖZET Radyoaktif ati klar, genellikle radyoaktif elementlerin kullani ldiği ti p, nükleer sanayi, nükleer santraller ve nükleer silah denemeleriyle çevreye verilirler. Radyoaktif elementler parçalanabilme özelliğ inden dolayi , ati ldi klariortamda da işi maya devam ederek çevreye zarar verirler. Bu nedenle radyoaktif ati klari n oluş umunu azaltmak, ortaya çi kan ati klari n da çevreye zarar vermeyecek ş ekilde

- by HY Kaptan
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- Nuclear Power Plant, Radioactive waste

Experiments with real and simulated radioactive cementitious wasteforms were set up to compare the leaching behaviour of cementitious wasteforms containing nuclear power plant operational waste in field and laboratory test conditions. Experiments revealed that the average annual 137 Cs leach rate in deionised water was about thirty-five times greater compared with the measured average value for the 1st year of the field test. Cumulative leached fraction of 137 Cs for 1st year (3.74%) was close to values reported in literature for similar laboratory experiments in deionised water, however more than two orders of magnitude higher than the 1st year leached fraction of 137 Cs in the repository test (0.01%). Therefore, to compare field and laboratory test results, a scaling factor is required in order to account for surface to volume factor difference, multiplied by a temperature factor and a leach rate decrease coefficient related to the ground water composition.

- by Michael Ojovan
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- Engineering, Ground Water, Laboratory experiment, Hazardous Materials

Nowadays Robot play a vital role in all the activities in human life including industrial needs. There is a definite trend in the manufacture of robotic arms toward more dexterous devices, more degrees of-Freedom, and capabilities beyond the human arm. The ultimate objective is to save human lives in addition to increasing productivity and quality of high technology work environments. The objective of this project is to design, analysis of a Generic articulated robot Arm. This project deals with the modeling of a special class of single-link articulated inspection arms of robot. These arms consist of flexible massless structures having some masses concentrated at certain points of hollow sections at the beam. Some aspects of the articulated Robot that are anticipated as useful are its small cross section and its projected ability to change elevation and maneuver over obstacle require large joint torque to weight ratios for joint actuation. A knuckle joint actions actuation scheme is described and its implementation is detailed in this project. The parts of the (AIA) arm are analyzed for deflection and stress concentration under loading conditions in different angles.

- by IJTET Journal
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- Nuclear Power Plant, Articulated Robot, Prismatic Joint

A nuclear energy plant must implement the fundamental safety functions (FSF) to ensure its safe operation. This paper demonstrates how a functional failure identification (FFI) method identifies the FSF, amongst other mitigation functions, as part of the systems engineering process (SEP). This method supports the functional failure modes and effects definition that is part of the functional analysis sub-process as per IEEE Std 1220 (SANS 26702:2008). The following techniques are integrate d, namely: Functional Flow Block Diagram (FFBD), Failure Modes and Effects Analysis (FMEA), Fishbone diagram, Master Logic Diagram (MLD) and Fault/Failure Tree Analysis (FTA), for achieving the outcomes for a nuclear energy plant. An unmitigated functional a rchitecture should be used for the FMEA, but this approach is counter intuitive and difficul t to follow. A method is proposed that is more intuitive for engineers to follow and still re ach the results for an unmitigated functional architecture.

- by Louwrence Erasmus
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- Functional Analysis, Nuclear Energy, Nuclear Power Plant

Correct prediction of water hammer transients is of paramount importance for the safe operation of the plant. Therefore, verification of computer codes capability to simulate water hammer type transients is a very important issue at performing safety analyses for nuclear power plants. Verification of RELAP5/MOD3.3 code capability to simulate water hammer type transients employing the experimental investigations is presented. Experience gained from benchmarking analyses has been used at development of the detail RELAP5 code RBMK-1500 model for simulation of water hammer effects in reactor main circulation circuit. In RBMK-type reactors the water hammers can occur in cases of rapid check valve operation. The performed analysis using RELAP5 code RBMK-1500 model has shown that in general the maximum values of the pressure pulses due to water hammer do not exceed the permissible loads on the pipelines.

- by E. Uspuras
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- Performance Analysis, Water Hammer, Interdisciplinary Engineering, Nuclear Power Plant

The paper summarizes studies of the distributions of Sr and Cs in the water and sediments of the Black Sea carried out during a 10-year period following the 1986 accident at the Chernobyl Nuclear Power Plant. Its goal is to assess the temporal evolution of radionuclide inventories and balances and to evaluate the mixing of water masses and the sedimentation processes using man-made radionuclides as tracers. Using mathematical models and field data, mixing time-scales of 5, 16 and 24 years have been estimated, respectively, for the water layers of depths 0-50, 0-100 and 0-200 m. For the Central Basin the ventilation time of the lower halocline is estimated at 15-25 years. Cs has been used to date shelf and deep-basin sediments, providing the history of chemical and radioactive pollution and of eutrophication during the past 50 years.

Teollisuuden Voima Oy (TVO) operates two nuclear power plant units in Finland and has started to build a third one. The current nuclear power units have to continuously maintain and update existing instrumentation and control systems (I&C). Each new device will have to be classified and qualified according to its safety requirements. Using modern technology means in practice that more and more components have programmable features. The reliability of such components has proved to be difficult to demonstrate because of the nature of flaws in the software. Standards and rules given by authorities set the acceptance criteria for the components used in the safety systems of nuclear power plants. As a result of this trend, there is a clear need for an integrated and effective method to qualify software-intensive I&C systems in nuclear power plant units. The integration has three major areas: (i) definition and harmonization of requirements for software-intensive systems at different safety classes, (ii) integration of several approaches such as Software Process Improvement and Capability dEtermination (SPICE) and Failure Mode, Effects and Criticality Analysis method (FMECA) to improve confidence in qualification and (iii) integration of the system acquisition and qualification processes to improve the total effectiveness of the acquisition, delivery and deployment processes. The integrated qualification method is called the TVO SoftWare Evaluation Procedure (SWEP). It consists of a detailed qualification process and related methods for safety category B and C (IEC 61226) and Finnish safety class 3 qualifications. TVO will use the TVO SWEP method to evaluate suppliers and the conformance of their products/systems against requirements. It has been used in several cases, and it seems to save a lot of qualification resources compared to traditional methods.

- by Risto Nevalainen
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- Computer Software, Nuclear Power Plant