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The article focuses on the interest of several countries to develop nuclear power. These countries include Egypt, Gulf Cooperative Council (GCC) states (Bahrain, Kuwait, Oman, Qatar, Saudi Arabia, and United Arab Emirates), and Turkey. However, these plans of several countries to develop nuclear programs have been described as peaceful in nature and focused on electricity production. International Atomic Energy Agency (IAEA) states that, launching a nuclear power program is a major task that requires careful planning and preparation. IAEA is the one who supports and regulates nuclear power operations around the world. Details on the attempts of developing nuclear power, as well as the nuclear activities and facilities of countries including Turkey, Egypt and Saudi Arabia are also offered.

The article focuses on the new generation of nuclear power plants regarded to be safer than previous designs. These plants are adjusted to meet the regulatory requirements set by the Nuclear Regulatory Commission (NRC) in the U.S. The new plant designs utilize the passive safety systems which rely on natural forces to provide emergency water in case of a loss of coolant. Reactor designers employ a defense-in-depth strategy to prevent releases of radiation into the environment. However, NRC implies that the current plants which is likely to be built in the U.S. will not provide major advances in safety and security.

Highlights • Structural equation modelling was engaged to test the reliability and validity of the relationship. • There are significant direct path of safety climate factors and risk estimate. • Priority to safety was the most important factor that affects the relationship. • There is no significant direct path of risk control measures and risk estimate. • There is no significant direct path of decision making attitude and risk estimate. Abstract Safety culture practices tend to strongly influence workers to control possible risk, make decision, and have a true perception on the risk despite the aim of encouraging safe work environment and reducing the numbers of human errors. The present study assessed the perceptions of the workers on nine safety culture practices as well as the individual risk estimate on the chemical and radiation hazard. In this study, a questionnaire survey was conducted among the workers of nuclear and radiation facilities in Malaysia. In this study, structural equation modelling (SEM) was utilised to analyse the relationship between the variables. The reliability and validity of the measured item was found to be acceptable. Meanwhile, the significant direct path of safety climate factors managed to be shown together with risk control measures, decision making attitude, and individual risk estimate. However, risk control measure and decision-making attitude did not mediate the relationship between safety climate factors and the risk estimate. On the other hand, priority to safety was identified as the most important factor that affects risk estimate, risk control measure, and decision-making attitude. This study also emphasized the requirement of communicative information which can be performed through information sharing, dissemination, and questioning mechanism. Overall, these findings manage to facilitate radiation stakeholders in developing the framework of safety culture practices as well as strengthening the safety management system. [ABSTRACT FROM AUTHOR]

Discusses investigations into how the U.S. Energy Department handles safety for its workers. Dept. of Energy submission of magnetic tapes to Dr. Alice Stewart, its chief critic; Reforms recommended by Energy Secretary James Watkins; Details of health dangers to nuclear workers; Policies; Role of Thomas Mancuso as workplace hygienist who examined conditions at Hanford.

Abstract To evaluate the environmental impact of 14C discharged to the atmosphere from the Tokai nuclear facility during normal operational periods, cores were extracted from an 80-yr-old pine tree that grew close to the Tokai Nuclear Power Plants (Tokai NPP and Tokai-2 NPP) and the Tokai Reprocessing Plant (Tokai RP), and the 14C levels in the annual rings were measured using accelerator mass spectrometry (AMS). The 14C activity ranged from 232 to 403 Bq kg−1 C in the tree rings formed during the period 1959–2013. The activities in the periods 1959–1965 and 2011–2013 represent the background levels, and the activities recorded between 1966 and 2010 are significantly elevated compared with these background levels, indicating uptake by the tree of 14C discharged from the Tokai nuclear facility. The excess 14C activity values of 1.1–70 Bq kg−1 C for the period 1966–2010 are positively correlated with the estimated total annual 14C discharges of 0.019–2.0 TBq from the Tokai NPP, Tokai-2 NPP, and Tokai RP (R2 = 0.72; n = 40). This activity of <70 Bq kg−1 C corresponds to an additional annual effective dose to the local population, via the food ingestion pathway, of less than 3.9 µSv yr−1. These values are negligible compared with the average natural background radioactive dose in the world. [ABSTRACT FROM AUTHOR]

Abstract In order to reduce the radiation exposure of the staff in radioactive environment of nuclear facilities, the probabilistic roadmap method is applied to optimize the walking path, which is on the basis of the map model, collision detection technology and dose calculation method. The proposed method can make a route in a very short time. The advantages of the path-planning method are reflected in two simulation experiments. From the first experiment, you can see the necessity and significance of path-planning in radioactive environment. To be specific, the cumulative dose is reduced greatly. In the other experiment, the computing speed of the proposed method is striking. Compared with the traditional A* algorithm, the reaction of this proposed method is almost instantaneous. Therefore, the proposed method has a good effect on path-planning in the radioactive environment of nuclear facilities. Highlights • Probabilistic roadmap method is applied to optimize walking path in radioactive environment. • The proposed method can make a route in a very short time. • The advantages of the path-planning method are reflected in two simulation experiments. [ABSTRACT FROM AUTHOR]

In this paper, we extensively review the status and projection of the Filtered Containment Venting Systems (FCVSs) in the Nuclear Power Plants (NPPs) in Asia. As an important strategy to mitigate the Severe Accidents (SAs) in NPPs, FCVS is designed to prevent the containment overpressure failure by providing containment venting through a filtration system. The 2011 Fukushima Daiichi nuclear accident brought a negative impact to the public faith on the nuclear safety, the accident has led regulatory authorities and policy makers to re-evaluate safety criteria and reinforce safety supervision. As a result, the mitigation of SAs regains attention of nuclear community. Many investigation and research works were initialized or resumed, particularly about the enhancement of FCVS on robustness and filtration efficiency. Presently, Asia is a region where nuclear energy grow fast. To ensure nuclear safety, the FCVS is being taken into account in the SA mitigation strategy by regulatory authorities. Chinese nuclear safety administration requires NPPs to have adequate and reliable containment venting system. Thus FCVS is either equipped or about to be equipped on most reactors operating or under construction in China. Nuclear companies and research institutions are conducting research and engineering works about FCVS design and function. In South Korea, a government funded project was carried out during 2013–2017, aimed to design a domestic FCVS technology. Confirmatory tests have been performed to evaluate the function of designed FCVS. In Japan, Tokyo Electric Power Company started installing FCVS on the NPPs since 2013. Related research work is performing in Japanese universities and institutions. Taiwan’s Atomic Energy Council also requests utility companies to reexamine the present emergency operation procedures and study the strategies such as the addition of required facilities for SA prevention and mitigation in each NPP. It is noticed that India is also investigating the options to mitigate the SAs, including FCVS. [ABSTRACT FROM AUTHOR]

The implementation of seismic base isolation can substantially reduce horizontal seismic demands on structures, systems, and components (SSCs) in a nuclear facility, potentially providing significant benefits in terms of increased safety (smaller seismic risk) and reduced capital construction cost. Although increased safety of SSCs has been demonstrated previously, the possible reduction in their capital cost has not been explored. To quantify the reduction in risk enabled by isolation, nonlinear response-history analysis of a conventionally founded and a base-isolated model of a generic nuclear facility (GNF) is performed at the sites of the Idaho National Laboratory and the Los Alamos National Laboratory: sites of moderate and high seismic hazard, respectively. Seismic probabilistic risk assessment is performed to compute the mean annual frequency of unacceptable performance. The seismic risk is reduced by 7 to 8 orders of magnitude by the implementation of isolation. The costs of addressing seismic loadings are estimated for the GNF in both the conventionally founded and base-isolated GNF. The possible reductions in the required seismic ruggedness and in the cost of SSCs in the isolated GNF are quantified at both sites. A reduction in cost enabled by isolation is possible at nearly all sites of nuclear facilities in the United States, with the greatest benefit at sites of high seismic hazard, such as LANL. Two risk-calculation procedures are used in the assessment: a simplified method based on Boolean mathematics and a rigorous method based on Monte Carlo analysis. The simplified procedure, which is suitable for implementation with preliminary design calculations, produces accurate estimates of risk unless the mean annual frequencies of unacceptable performance are very small, measured here as smaller than 10 −10 . The sensitivity of the calculated risk in the conventionally founded GNF, to the choice of anchor period for the seismic hazard curve, is investigated and found to be insignificant over the range considered: 0 to 0.10 s. [ABSTRACT FROM AUTHOR]

A key decision for nuclear facilities is evaluating the need for an update of an existing seismic hazard analysis in light of new data and information that has become available since the time that analysis was completed. We introduce the newly developed risk-informed Seismic Hazard Periodic Reevaluation Methodology (referred to as the SHPRM) and present how a Senior Seismic Hazard Analysis Committee (SSHAC) Level 1 probabilistic seismic hazard analysis (PSHA) was performed in an implementation of this new methodology. The SHPRM offers a defensible and documented approach that considers both the changes in seismic hazard and engineering-based risk information of an existing nuclear facility to assess the need for an update of an existing PSHA. The SHPRM has seven evaluation criteria that are employed at specific analysis, decision, and comparison points in a graded approach, based on the nuclear facility's seismic design category. The SHPRM is implemented using a SSHAC Level 1 study performed for the Idaho National Laboratory, USA. The implementation focuses on the SHPRM evaluation criteria that are based directly on the seismic hazard, and the inputs to these criteria are derived from the SSHAC Level 1 PHSA. To illustrate outcomes of the SHPRM that do not lead to the need for an update and those that do, the example implementations of the SHPRM are performed for nuclear facilities that have target performance goals expressed as the mean annual frequency of unacceptable performance at 1 × 10 −4 , 4 × 10 −5 and 1 × 10 −5 . [ABSTRACT FROM AUTHOR]

Vulnerability assessment of Structures Systems and Components (SSCs) of nuclear facilities for earthquake ground motion exceeding the design basis ground motion has become a key issue to ensure safety in case of highly improbable but possible extreme earthquake event beyond design basis. In order to assess seismic safety of SSCs located at different floors of the building, floor response spectra (FRS) for such beyond design basis seismic demand, taking into account possible nonlinear behavior of structure, is required. Currently used methods (e.g. IAEA Safety Series 28) do not capture softening of structure at increased demand. A performance based approach would be ideal under such situations. FRS for higher seismic demand may be developed through nonlinear time history analysis of the structure. However, this approach is complicated and cumbersome for structural configurations such as that of NPPs and requires deep insight of solution algorithm and large resources for numerical solution. In this work a simplified methodology for FRS generation for higher level of seismic demand using linear time history analysis of an equivalent linear structural model is proposed. Nonlinear behavior of structure and associated damping are estimated from Nonlinear Static Pushover Analysis. Through this approach realistic estimation of stiffness degradation of the structure and increase in damping at different levels of seismic demands can be made. Subsequently, linear time history analysis is performed on degraded structure, accounting for stiffness reduction and hysteretic damping to obtain the floor time histories which are then converted to FRS. Application of this method is shown through case studies. [ABSTRACT FROM AUTHOR]

Steel-plate composite (SC) walls used in the containment internal structures of safety-related nuclear facilities intersect each other at corners of compartments to form wall-to-wall corner or L-joints. This paper focuses on joint shear behavior and strength of wall-to-wall corner L-joints, the understanding of which is imperative for the design of wall-to-wall joints irrespective of the failure mode. This paper presents the results of a large-scale test conducted on an SC wall-to-wall corner or L-joint. The test specimen was designed to undergo joint shear failure prior to flexural yielding of the connected SC walls. The test specimen did not include any tie bars in the joint shear region, which would represent the lower bound case for these joints. The specimen was subjected to cyclic loading until failure, and the test results included the joint shear force-displacement and joint shear force-shear strain responses of the specimen along with observations of concrete cracking and crushing in the joint region. The results and observations were evaluated to establish the joint shear strength and governing failure mode. The experimentally measured joint shear strength was compared with the ACI 349-06 code equation for calculating the joint shear strength of comparable reinforced concrete beam-to-column joints. A detailed nonlinear finite element model was developed to gain additional insight into the observed behavior. The model was benchmarked using experimental results. The experimental and analytical results were compared with ACI 349-06 code provisions for calculating the joint shear strength of reinforced concrete beam-to-column joints. The results from experimental and analytical investigations indicate that the ACI 349-06 code equation could be used for estimating the joint shear strength of steel-plate composite wall-to-wall L-joints with γ of 8. [ABSTRACT FROM AUTHOR]

The present study concerns the three-dimensional CFD analysis of hydrogen explosions in APR1400 containment. Initial conditions for combustion simulations are obtained by a cost-effective coupling with a lumped parameter code for preceding mixture distribution analysis. Three severe accident scenarios are examined: Small-Break Loss-Of-Coolant Accident (SB-LOCA) and Station Black-Out (SBO) with activated as well as deactivated three-way valve. Only in the last case, thermal ignition of the hydrogen-air-steam mixture could be triggered. Flame propagation in the containment and particularly in the In-containment Refueling Water Storage Tank (IRWST) is investigated in detail. Additional generic scenarios demonstrate the methods ability to reproduce strong Flame Acceleration (FA) and even the hazardous Deflagration-to-Detonation Transition (DDT). [ABSTRACT FROM AUTHOR]

The radiological characterization of contaminated elements (walls, grounds, objects) from nuclear facilities often suffers from too few measurements. In order to determine risk prediction bounds on the level of contamination, some classic statistical methods may therefore be unsuitable, as they rely upon strong assumptions (e.g., that the underlying distribution is Gaussian) which cannot be verified. Considering that a set of measurements or their average value come from a Gaussian distribution can sometimes lead to erroneous conclusions, possibly not sufficiently conservative. This paper presents several alternative statistical approaches which are based on much weaker hypotheses than the Gaussian one, which result from general probabilistic inequalities and order-statistic based formulas. Given a data sample, these inequalities make it possible to derive prediction intervals for a random variable which can be directly interpreted as probabilistic risk bounds. For the sake of validation, they are first applied to simulated data generated from several known theoretical distributions. Then, the proposed methods are applied to two data sets obtained from real radiological contamination measurements. [ABSTRACT FROM AUTHOR]

In today’s world there is more than one point of view and different debates about how to determine and what constitutes the concept of nuclear security. A narrow attitude, which we can not agree with, is that it is an area that deals with issues of the model of physical security of nuclear materials. A broader interpretation of nuclear safety advocates a more comprehensive view of the activities that support the more ‘aims and content of nuclear safety’. This multidisciplinary approach is necessary for effective and sustainable nuclear security framework, both at the national and supranational levels. Generally speaking, nuclear safety can be defined as a set of measures to ensure the conditions for the operation of the nuclear facility to prevent nuclear accidents or mitigate their consequences and reduce the radiation risk to the prescribed level. It is well known that nuclear safety cannot be treated outside of the context of responsibility for carrying out nuclear activities and nuclear damage caused by a nuclear accident, and thus the insurance to cover liability for nuclear damage. [ABSTRACT FROM AUTHOR]

The article discusses the impact of the accident occurred at the Three Mile Island Nuclear Power Plant in Pennsylvania. The accident presented the U.S. nuclear power industry with some serious problems. The accident which occurred in March 1979 shook the Americans' confidence in nuclear safety. The accident shifted the burden of proof to those who claim that nuclear power is safe. It brought into focus the ignored reactor safety problems. It brought reactor design also under close regulatory review because it was the design and structure of the reactor that contributed to the accident. The accident necessitated corrective measures to check deficiencies in nuclear plant design and operator procedures.

This paper is intended to suggest a system for evaluation on the hazards from human errors during decommissioning of nuclear facilities. The system was developed under virtual networked environment. The innovative features are real-time changing direction of workers in a scenario and real-time measuring personal exposure dose and collective exposure dose. The system will be expected to be utilized as a training tool for improving familiarization of a workplace and for preventing workers from accidents. [ABSTRACT FROM AUTHOR]

The factors supporting the safety of atomic energy are discussed. It is underscored that all factors considered in the generally accepted nuclear-safety strategy must be fully taken into account and implemented in order to achieve this goal. [ABSTRACT FROM AUTHOR]

Potential exposures are those that may occur as a result of unanticipated operational performance or accidents. Potential exposure situations are probabilistic in nature because they depend on uncertain events such as equipment failure, operator errors or external initiators beyond the control of the operator. Consequently, there may exist a range of possible radiological impacts that need to be considered. In this paper a Level 3 Probabilistic Safety Assessment (PSA) for a hypothetical scenario relevant to Innovative Nuclear Energy Systems (INS) was conducted using computer code MACCS (MELCOR Accident Consequence Code Systems). The acceptability of an INS was analyzed taking into account the general requirement that relocation or evacuation measures must not be necessary beyond the site boundary. In addition, deterministic modeling of the accident consequences for the critical meteorological conditions was carried out using the JRODOS decision support system (Real-time On-line Decision Support system for off-site emergency management in Europe). The approach used for dose and risk assessment from potential exposure of accidental releases and their comparison with acceptance criteria are presented. The methodology described can be used as input to the licensing procedure and engineering design considerations to help satisfy relevant health and environmental impact criteria for fission or fusion nuclear installations. [ABSTRACT FROM AUTHOR]

During the commissioning of Nuclear Power Plant Atucha II, a number of tests are required with the objective of demonstrating the reliability and safety of equipment and facilities. Throughout those tests, a series of measurements were taken in order to determine the stress states of different components of the primary and secondary systems, with the objective of verifying the calculation models used during the design stage. In this work we present, in particular, results of the strain measurements in pipes of the primary system connected to the steam generator during ramps of temperature and pressure. We describe the instrumentation used, the details of the measurement point, the acquisition system, and the entire process applied to the signals, including filters and corrections. Finally, we discuss some of the results obtained. With those results, the calculation models were verified correctly. [ABSTRACT FROM AUTHOR]

Inspection and monitoring of key components of nuclear power plant reactors is an essential activity for understanding the current health of the power plant and ensuring that they continue to remain safe to operate. As the power plants age, and the components degrade from their initial start-of-life conditions, the requirement for more and more detailed inspection and monitoring information increases. Deployment of new monitoring and inspection equipment on existing operational plant is complex and expensive, as the effect of introducing new sensing and imaging equipment to the existing operational functions needs to be fully understood. Where existing sources of data can be leveraged, the need for new equipment development and installation can be offset by the development of advanced data processing techniques. This paper introduces a novel technique for creating full 360° panoramic images of the inside surface of fuel channels from in-core inspection footage. Through the development of this technique, a number of technical challenges associated with the constraints of using existing equipment have been addressed. These include: the inability to calibrate the camera specifically for image stitching; dealing with additional data not relevant to the panorama construction; dealing with noisy images; and generalising the approach to work with two different capture devices deployed at seven different Advanced Gas Cooled Reactor nuclear power plants. The resulting data processing system is currently under formal assessment with a view to replacing the existing manual assembly of in-core defect montages. Deployment of the system will result in significant time savings on the critical outage path for the plant operator and will allow improved visualization of the surface of the inside of fuel channels, far beyond that which can be gained from manually analysing the raw video footage as is done at present. [ABSTRACT FROM AUTHOR]

Potential problems of the MBIR multipurpose, research, fast reactor, currently under construction, in the validation of new nuclear-power projects with fast reactors and a closed nuclear fuel cycle are discussed. The basic characteristics and experimental possibilities of the reactor are described. Considerable attention is given to the primary directions of research on nuclear safety, reactor and structural materials science, and NFC closure. Proposals for the development of an MBIR-based international research center are given. [ABSTRACT FROM AUTHOR]

A methodology for evaluating risks at nuclear facilities is developed in this work. A series of measures is drawn from the analysis of factors that determine risks. Five measures are created to evaluate risks at nuclear facilities. These include the legal and institutional framework, material control, physical protection system effectiveness, human resources, and consequences. Evaluation attributes are developed for each measure and specific values are given in order to calculate the risk value quantitatively. Questionnaires are drawn up on whether or not a state has properly established a legal and regulatory framework (based on international standards). These questionnaires can be a useful measure for comparing the status of the physical protection regime between two countries. Analyzing an insider threat is not an easy task and no methodology has been developed for this purpose. In this study, attributes that could quantitatively evaluate an insider threat, in the case of an unauthorized removal of nuclear materials, are developed by adopting the Nuclear Material Accounting & Control (NMAC) system. The effectiveness of a physical protection system, P ( E ), could be analyzed by calculating the probability of interruption, P ( I ), and the probability of neutralization, P ( N ). In this study, the Tool for Evaluating Security System (TESS) code developed by KINAC is used to calculate P ( I ) and P ( N ). Consequence is an important measure used to analyze risks at nuclear facilities. This measure comprises radiological, economic, and social damage. Social and economic damages are difficult to evaluate. Therefore, radiation levels and theft of nuclear materials that could be quantified are adopted as attributes for analyzing the consequences. Awareness of the nuclear security culture and physical protection resources such as staffing, capabilities, and cost required to provide PP should be considered when evaluating risks. In this study, these attributes are included in the measure of human resources. Human resources include such factors as trustworthiness, degree of nuclear security culture awareness, and frequency of psychiatric testing of employees. A case study performed on hypothetical facilities demonstrates that the developed methodology could be used to analyze innovative nuclear systems as well as existing facilities. [ABSTRACT FROM AUTHOR]

The technical requirements of a system that continually monitors the condition of the safety barriers of nuclear power plants are substantiated. It is shown for the example of the Rostov NPP that power units with VVER which are operating, under construction, or in the design stage can be equipped with a measuring system for monitoring radiation leaks from the steam generators. This improves safety (by not allowing emergency situations) and increases the installed capacity utilization factor (by eliminating unjustified shutdowns of power units) of operating power plants. [ABSTRACT FROM AUTHOR]

The study of electrical fires is a major concern for fire safety in the industry and more particularly for fire safety in nuclear facilities. To investigate this topic, IRSN conducted a large number of real-scale experiments involving open-door electrical cabinets burning firstly under a calorimetric hood and then inside a mechanically-ventilated compartment. The main challenges are to determine accurately the heat release rate of such a complex fire source in a vitiated atmosphere and to provide an experimental database for validating properly the combustible modelling, taking into account the oxygen depletion in an enclosure. After providing a detailed description of the fire scenarios and of the experimental apparatus, this paper focuses on the characteristic stages of the cabinet fire development, essentially based on the heat release rate time evolution of the fire. The effects of the confinement, of the outlet branch location, of the ventilation management and of the fire barrier on the fire source were then investigated. The reproducibility of electrical cabinet fires is also studied. A new model for complex fire source (applied in this study for open-door electrical cabinet fires) was then developed. This model was introduced in the zone code SYLVIA and the major features of the compartment fire experiments, such as characteristic heat release rate with effect of oxygen depletion and over-pressure peak were then calculated with a rather good agreement for this complex fire source (i.e. electrical cabinet). [ABSTRACT FROM AUTHOR]

Chemical and metallurgical operations involving plutonium and other nuclear materials account for most activities performed at the Los Alamos National Laboratory's Plutonium Facility (PF-4). The presence of large quantities of fissile materials in numerous forms at PF-4 makes it necessary to maintain an active criticality safety program. The LANL Nuclear Criticality Safety (NCS) Program provides guidance to enable efficient operations while ensuring prevention of criticality accidents in the handling, storing, processing and transportation of fissionable material at PF-4. In order to achieve and sustain lower criticality safety control infraction (CSCI) rates, PF-4 operations are continuously improved, through the use of Lean Manufacturing and Six Sigma (LSS) business practices. Employing LSS, statistically significant variations (trends) can be identified in PF-4 CSCI reports. In this study, trends have been identified in the NCS Program using the NCS Database. An output metric has been developed that measures ADPSM Management progress toward meeting its NCS objectives and goals. Using a Pareto Chart, the primary CSCI attributes have been determined in order of those requiring the most management support. Data generated from analysis of CSCI data help identify and reduce number of corresponding attributes. In-field monitoring of CSCI's contribute to an organization's scientific and technological excellence by providing information that can be used to improve criticality safety operation safety. This increases technical knowledge and augments operational safety. [ABSTRACT FROM AUTHOR]