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To contribute to the future updating on the Japanese safety criteria for pellet/cladding mechanical interaction (PCMI) failure of light water reactor fuels under reactivity-initiated accide...

Underground water-sealed gas storage caverns have become the primary method for strategic storage of LPG. Previous studies of excavation blasting effects on large-scale underground water-sealed gas...

Contractor selection is one of the major concerns of industry managers such as those in the oil industry. The objective of this study was to determine a contractor selection pattern for oil and gas industries in a safety approach. Assessment of contractors based on specific criteria and ultimately selecting an eligible contractor preserves the organizational resources. Due to the safety risks involved in the oil industry, one of the major criteria of contractor selection considered by managers today is safety. The results indicated that the most important safety criterion of contractor selection was safety records and safety investments. This represented the industry's risks and the impact of safety training and investment on the performance of other sectors and the overall organization. The output of this model could be useful in the safety risk assessment process in the oil industry and other industries.

We completed a novel assessment to determine the unquantified and uncredited safety margins (i.e., the difference between the licensing basis and as-loaded calculations) available in as-loaded spent nuclear fuel (SNF) casks. This assessment was performed as part of a broader effort to assess issues and uncertainties related to the continued safety of casks during extended storage and transportability following extended storage periods. Detailed analyses crediting the actual as-loaded cask inventory were performed for each of the casks at three decommissioned pressurized water reactor (PWR) sites to determine their characteristics relative to regulatory safety criteria for criticality, thermal, and shielding performance. These detailed analyses were performed in an automated fashion by employing a comprehensive and integrated data and analysis tool—Used Nuclear Fuel-Storage, Transportation & Disposal Analysis Resource and Data System (UNF-STD calculated decay heat margins ranged from 4 to almost 22 kW (as of 2014); and significant uncredited transportation dose rate margins were also observed. The results demonstrate that, at least for the casks analyzed here, significant uncredited safety margins are available that could potentially be used to compensate for SNF assembly and canister structural performance related uncertainties associatedmore » with long-term storage and subsequent transportation. The results also suggest that these inherent margins associated with how casks are loaded could support future changes in cask licensing to directly or indirectly credit the margins. Work continues to quantify the uncredited safety margins in the SNF casks loaded at other nuclear reactor sites.« less

The design of road tunnels is an essential infrastructure component in the highway system. The study implements reasonable criteria for obtaining vertical curve length based on sight distance requirements of road tunnels versus open roadways. The infrastructure impact of sight distance tunnel characteristics on the design of crest and sag vertical curve radii is based on safety and driving comfort criteria. Both results of crest and sag vertical curves show significant reductions of vertical radii in road tunnels compared to open roadways, that is, percentage reduction range of 35% to 71% depending on the design speed, curve type, and tunnel pavement characteristics. The driving comfort criterion generally overrides in the lower range of design speeds whereas the safety criterion overrides in the upper range of design speeds: 100– to 120 km/h for moist and dry tunnel sag vertical curve, 90 to 120 km/h for dry tunnel crest vertical curve, and 80 to 120 km/h for moist tunnel crest vertical curve.Thi...

The purpose of this study was to explore kindergarten teachers’ views and roles regarding outdoor play environments in Omani kindergartens. Thirty kindergarten teachers from 15 private kindergartens were observed and interviewed. The results indicated that teachers recognize the importance of outdoor play in children’s development and learning. Although teachers are satisfied with the safety conditions of the outdoor play environments, they feel they need to know more about how to maintain safety criteria in this environment. Teachers most frequently took on the role of supervising and guiding children’s behaviour or setting up materials for play, whereas providing verbal comments and participating in play were less common. Omani outdoor play environments were found to face specific difficulties, such as the hot climate, a shortage of materials and equipment, a lack of natural surroundings and the short length of time spent outdoors. The article concludes with policy implications and pedagogical recommend...

The International Atomic Energy Agency (IAEA) fundamental safety objective is to protect people and the environment from harmful effects of ionizing radiation. Therefore, a severe accident consequence assessment has to be able to include all quantifiable consequences on people and the environment. Our previous studies on estimation of cost per severe accident succeeded in quantifying aforementioned consequences. However, the estimation requires enormous quantity of data, time and human resources, thus it may be inappropriate at the reactor design approval stage. Finnish government uses “100 TBq cesium 137 release into environment”, which was proved to generate limited health effects, as one of the reactor design criteria for accident consequences. In this study, we perform an evaluation of annual dose from the 100 TBq cesium 137 release and confirm limited health effects. We form the environmental impact index based on insights from our previous studies and used it to assess consequences to the environmen...

In general, road authorities recognise the importance of research studies to support the development of maintenance programmes for surface characteristics to be incorporated into pavement management systems. For this purpose, the establishment of threshold values is an essential issue, but the literature concerning this topic is very scarce. The main objective of this paper was to present a methodology for the establishment of threshold values based on safety criteria and concerning skid resistance and macrotexture, represented by International Friction Index, coefficient of friction and texture depth. The methodology consists of two phases: evaluation of the influence of the pavement surface properties on road accidents and establishment of threshold values for each road environment (RE). The analysis of the influence of these parameters on accident occurrence is based on the definition of compound RE, a more appropriate technique for accident modelling. Minimum and safety values of the parameters are re...

The transportation safety case for transporting spent nuclear fuel is a requirement for licensing. It has both qualitative and semiquantitative aspects. The qualitative aspects include transportation regulations, radiation dose limits, role of the transportation package in transportation, transportation package certification process, training, emergency response, the performance of the transportation package in accidents and the evaluation of past transportation accidents. The quantitative aspects support the qualitative descriptions. Radiation doses accrued by members of the public and by workers are calculated using the code RADTRAN. Dose from both routine, incident free highway transportation and from highway transportation accidents are part of the safety case and will be compared with both background doses and the regulatory safety criteria. The radiation doses from routine transportation are calculated for the following: the maximally exposed member of the public, doses to vehicle escorts an...

This article presents a summary description of the main safety criteria and requirements adopted in the ‘Guideline for Load and Resistance Assessment of Existing European Railway Bridges – advices on the use of advanced methods’ (SB-LRA, 2007. Prepared by Sustainable Bridges – a project within EU FP6. Available from: www.sustainablebridges.net) developed within the European project Sustainable Bridges – Assessment for Future Traffic Demands and Longer Lives. It shows selected safety formats recommended for the assessment of existing bridges, starting from the most simple, for the member level assessment, and ending with the most advanced, for the probabilistic non-linear assessment of a bridge system. The article also presents simplified methods for the probabilistic non-linear assessment of bridge systems. Furthermore, it shows probabilistic models for the resistance of selected types of bridge main carrying members that may be used in the simplified probabilistic assessment of existing railway bridges.

The development of society is still marked by the need for lighter and stronger structures. The materials that respond best to these needs are composite materials. Designing composite materials is difficult as it involves designing the geometry and their composition. Traditionally, the design tasks have been based on approximate methods; the possibility for creating composite materials is almost unlimited, characterization by testing is very expensive and it is difficult to apply the results to other contexts. This article proposes a variable neighbourhood search-based model for the design of symmetric laminated composites, a general encoding for the design of composites, an evaluation function that has taken into consideration cost and safety criteria in design, the neighbourhood structures and a set of local search operators. The proposed model has been applied to different real-world problems and the results have been compared with other well-known design methods.

In the safety analysis of nuclear power plants, large break loss-of-coolant accidents (LB LOCAs) continue to be one of the major issues. In this study, the latest version of the French best-estimate computer code CATHARE2 V2.5_1 Mod6.1 was used in order to predict the thermal-hydraulic phenomena in the VVER-1000 reactor during LB LOCA. This type of reactor is in operation and under construction in several countries including Russia, Ukraine, Bulgaria, Czech Republic, China, and India. The paper first presents the CATHARE modeling of a VVER-1000 reactor, including the core, the vessel, the primary and secondary circuits with the pressurizer, the main circulation pumps, the horizontal steam generators, and the steam lines. The emergency core cooling system (ECCS) is presented also with hydroaccumulators and high- and low-pressure safety injection pumps. The break, located in the cold leg and close to the reactor, is represented by the RUPTURE module of CATHARE, which modelizes a double-ended guillotine type of break. In the current calculations, the bottom-up and top-down reflooding play a very important role, compared to some western pressurized water reactors, because of the ECCS injection into the downcomer and upper plenum of the VVER-1000. That is why special attention is paid to the validation of the CATHARE reflooding model, based on the REWET-II experiment devoted to VVER. In CATHARE, a special two-dimensional reflooding module has been developed, which takes into account the radial and axial conduction in the vicinity of the quench front. The assessment based on the REWET-II facility shows the ability of the code to predict the progression of two simultaneous quench fronts. Then, the paper presents the results of the LB LOCA analysis for VVER-1000 (reference case and sensitivity study), describing the blowdown, refill, and reflood phases of the accident. Problems related to the countercurrent flow limitation (CCFL) phenomena at the core outlet are also considered. The results of the calculations show reasonable prediction of the basic parameters important to the safety of the plant, such as quench front progression, fuel-cladding temperatures, break flows, pressure behavior, etc. As expected, the temperatures-despite the conservative character of the modeling-remain below the safety criteria. It should be noted that the calculated fuel wall temperatures are very sensitive to the proper selection of the reflooding parameters and CCFL and need skillful choice of the reflooding conditions.

LOCA-simulated experiments were performed with MDA, ZIRLO™, M5®, NDA, and Zircaloy-2 cladding specimens with local burn-ups ranging from 66 to 76 MWd/kg. Short test rods fabricated with the cladding specimens were heated, isothermally oxidized at 1,459 to 1,480K in steam flow, and finally quenched in flooding water. Rod rupture and subsequent double-sided oxidation of the cladding were also simulated in the experiments. Neither split-fracture nor fragmentation occurred during the quench in the cladding specimens which were oxidized to about 18–27% of the metallic thickness. Accordingly, the fracture boundary, a most important safety issue, is not reduced significantly by the high burn-up and use of the new alloys within the examined scope, although it may be somewhat reduced with pre-hydriding during the reactor operation as observed in unirradiated specimens.

The theory of uniform hyperbolicity is used to obtain sets of forcing functions for which the response of a dynamical system remains in a prescribed 'safe' region. The resulting safetycrit eria are asymptoticallyoptim al as the size of the safe region goes to zero.

In Korea, an evolutionary reactor called the Korean Next Generation Reactor (KNGR) is being developed. Safety and Regulatory Requirements and Guidance (SRRG) are also being developed for the regula...

The in-containment water storage system (IWSS) is a newly adopted system in the design of the Korean Next Generation Reactor (KNGR). It consists of the in-containment refueling water storage tank, holdup volume tank, and cavity flooding system (CFS). The IWSS has the function of steam condensation and heat sink for the steam release from the pressurizer and provides cooling water to the safety injection system and containment spray system in an accident condition and to the CFS in a severe accident condition. With the progress of the KNGR design, the Korea Institute of Nuclear Safety has been developing Safety and Regulatory Requirements and Guidances for safety review of the KNGR. In this paper, regarding the IWSS of the KNGR, the major contents of the General Safety Criteria, Specific Safety Requirements, Safety Regulatory Guides, and Safety Review Procedures were introduced, and the safety review items that have to be reviewed in-depth from the regulatory viewpoint were also identified.

Test runs have been carried out in Velim, Czech Republic, to analyse derailment conditions at higher speeds up to 160 kph. In the research program ‘DYSAF’ a test running gear was developed which enabled to test derailment of a wheelset in guiding and unloaded conditions. With the aid of a lateral force actuator static and dynamic Y/Q derailment situations were measured for different velocities and angle of attack. In this paper a comparison of test and simulation results of the multi body program MEDYNA is presented.Identifying the parameters of nominal wheel loads of the test running gear and friction coefficients in tread and flange simulation results show a very good and promising agreement for the compared signals. Only higher frequency terms are lacking. To include them dynamic properties of track, hydraulic device and wheel roundness have to be improved.Derailment simulation with a MBS tool is a possibility to analyse derailment cases and to develop safety criteria against derailment of rail...

The effectiveness of intentional depressurization ofa pressurized water reactor primary system as a means to maintain core cooling during a small-break loss-of-coolant accident (SBLOCA) was studied. The investigation was based on experiments conducted at the Rig of Safety Assessment-V (ROSA- V) Large Scale Test Facility (LSTF) and RELAP5/MOD3 code calculations performed for LSTF geometry, together with single lumped-volume model calculations-all simulating hypothetical total failure of the high-pressure-injection system. For cold-leg breaks ≥2.5% of the leg cross-sectional area, experimental and analytical results have shown that the break discharge depressurizes the primary system to the accumulator (ACC) and low-pressure-injection (LPI) system injection pressures, and thus the core cladding temperature would be maintained below ∼1000 K. For break areas ≤1.0%, on the other hand, additional depressurization means are needed to initiate the ACC injection before the core is overheated. RELAP5/ MOD3 calculations have shown that steam venting through the pressurizer power-operated relief valves would be effective in depressurizing the primary system to the ACC and LPI pressures. However, for break areas

A Delphic study investigating the perceived relationship between 20 contractor selection criteria (CSC) currently in use and project success factors (PSFs) in terms of time, cost and quality is described involving a sample of eight experienced construction personnel, including two validators. A consensus of the likely impact of each criterion on time, cost and quality is established in terms of pessimistic, average and optimistic values, which are then converted into expected means and variances via the PERT approach. The ten most and ten least important CSCs are identified and examined for differences and similarities between PSFs. The results show that past failures, financial status, financial stability, credit ratings, experience, ability, management personnel and management knowledge are perceived to be the dominant CSCs affecting all three PSFs, with safety criteria (safety, experience modification rate, OSHA incidence rate, management safety accountability) and the length of time in business being ...

The topic of food safety in the U.S. frequently is the lead story in the news media and continues to be an important issue for biological scientists of all disciplines. The emotional and sensational aspects of food safety often cloud the facts as reported by journalists. Because we increasingly depend on foreign countries for our food supply, and because of the commonly held belief in our society that foreign suppliers are not required to meet the safety criteria of U.S. regulatory agencies, it is important for agriculturalists with international expertise to offer objective views on the topic of food safety. This report contains the observations and views of one such scientist.

The design assessment concerning the mechanical behaviour of transport and storage casks for radioactive material to fulfil nuclear safety criteria has to be based on two essential considerations: (1) Effective analysis of the stress–strain state of the cask components under both normal operational and test conditions including hypothetical accident scenarios with suitable accepted methods. (2) Economic estimation of the required properties and the structural state of the cask components with sufficient exactness. In an overview of the codes which are available at GNS/GNB for cask impact strength analyses (ANSYS, ADINA, VDI Codes), procedures and aspects of benchmarking and validation of calculation codes are described. The results of experimental full size cask drop test programs (CASTOR, POLLUX) and corresponding pre-test calculational analyses show the suitability of the codes used. The influence of dynamic effects on the mechanical properties of material (ductile cast iron, wood) has been inv...

In this paper the results are presented of a national survey on playground safety in The Netherlands. Thirteen trained inspectors each inspected 51 playgrounds in their own region. In total, 663 playgrounds with 7150 pieces of playground equipment were subjected to an inspection. The following safety criteria were inspected: entanglement, surfacing material, installation and maintenance aspects. The results show that one out of every three pieces of playground equipment fail on one or more criteria. The number of pieces of equipment that fail in playgrounds operated by local authorities or recreational companies is higher than that in playgrounds operated by voluntary associations. The smaller the playground the higher the number of failures per piece of equipment The time has come for experts to provide the operators with information which will improve the safety level. More funding should be provided for the local authorities to prevent the occurrence of accidents.

In Germany, heat-generating radioactive waste is intended to be disposed of in the Gorleben repository. With regard to planning and construction of this facility as well as its operational and postoperational phase, safety issues must be taken into account. Emphasis is on the preliminary planning for the Gorleben project and on radiation protection and safety criteria regarding the postclosure phase of this planned repository.

With most of the current probabilistic safety criteria (PSCs), it is difficult to select a unanimously acceptable single-point safety goal and to evaluate the uncertainties in the results of a probabilistic safety assessment (PSA). A new PSC is proposed, in which the distribution of the public's safety goals (DPSG) is used as a benchmark for evaluating the results of a PSA rather than a single-point safety goal. With this approach, the DPSG and the uncertainties in the results of a PSA can be handled properly so as to give a clear answer of 'how much of the public feels a nuclear reactor is safe' to the question of 'How safe is safe enough ' The proposed PSC is compared with the current PSCs and the expected utility model. If the actual DPSG is unavailable or difficult to obtain directly, a lognormal distribution is recommended as an appropriate DPSG for core melt frequencies in terms of maximizing entropy and minimizing total social cost. The proposed DPSG and PSC for core melt frequency are applied to the results of NUREG-1150.

ETHEL, a new research facility located at the JRC-Ispra, is envisaged to be licensed for 37 PBq (1 MCi or 100 g) of tritium. A central service of laboratory operations is the overall control of tritium within the facility. The intended management of tritium has been developed on the basis of ETHEL's general safety criteria and the requirements of typical experiments with regard to quantity, quality and frequency of tritium distributions. This paper describes the method of supplying tritium to the laboratory, its accountancy and long-term storage, its distribution to individual experiments and, finally, its retrieval from experiments or the laboratory's tritium recovery system.

The European Tritium Handling Experimental Laboratory is committed to investigate the propagation modes of tritium in materials, components, equipment and process loops. As for other radiochemical facilities, a number of safety criteria and principles have been introduced into the design, construction and intended operation of the facility with the aim of optimizing the protection of both the workers and the population. This paper highlights a variety of safety aspects applied to the laboratory: confinement barriers, radiological protection, waste handling and, finally, precautions against internal and external events.

This paper reports on the present schedule of JET which includes an experimental campaign with D-T plasmas at the end of the Project programme. A dedicated facility, the Active Gas Handling System, (AGHS), has been designed and is being commissioned to process the torus exhaust streams and to recycle tritium and deuterium. The AGHS is expected to process a maximum throughput of 30 g tritium daily and total tritium inventory will not exceed 90 g. The design is subject to a comprehensive safety analysis which must show that stringent safety criteria are met. In parallel to the AGHS installation, the JET torus and its auxiliary systems are being analyzed for compliance with the same safety criteria. Modification are being implemented where required. JET will be the first experimental facility where the tritium fusion fuel processing cycle will be closed (albeit without breeding) and hence important experience and experimental data are expected to be gained for the next generation of fusion devices.

The Rig of Safety Assessment (ROSA) III facility is a volumetrically scaled (1/424) boiling water reactor (BWR) system with an electrically heated core designed for integral loss-of-coolant accident (LOCA) and emergency-core-cooling-system (ECCS) tests. Experimental results obtained so far confirm that the severest single failure assumption in ECCS is the high-pressure core spray system failure even in a large-break LOCA in a BWR. The measured peak cladding temperature was well below the present safety criterion of 1473 K, even with the single failure assumption in ECCS, and the effectiveness of ECCS for core cooling during a double-ended-break LOCA has been confirmed. The overall agreement between the results calculated by the RELAP4/MOD6/U4/J3 computer code and the experimental results is good. The similarity between the ROSA-III test and a BWR LOCA has been confirmed through the comparison of calculated results for the ROSA-III facility and a BWR system.

The Mirror Advanced Reactor Study (MARS) was a 2-yr study of designs for a commercial tandem mirror fusion reactor. Two of the goals of the MARS program were to exploit the full potential of fusion for occupational and environmental safety during operation and maintenance and to realize safe long-term disposal of radioactive wastes. Two numerical ratings devised to characterize materials with respect to impacts of induced radioactivity on these two goals are described. The ratings devised for these purposes are the remote maintenance rating and the waste disposal rating. The MARS reactor designers had these ratings available and used them as guidelines in making configuration and materials choices. Significant differences in meeting these goals were identified, depending on the materials chosen. The final MARS design fully utilized the technology available today and, in large measure, achieved the stated goals of the program.

(1985). A Review of: “Safety of dams: flood and earthquake criteria” Committee on safety criteria for dams for the National Research Council of the United States National Academy Press, Washington, D.C., USA, 1985, 276 pp, £18.50. Civil Engineering Systems: Vol. 2, No. 4, pp. 232-233.