Survey and evaluation of inherent safety characteristics and passive safety systems for use in probabilistic safety analyses

International Nuclear Information System (INIS)

Wetzel, N.; Scharfe, A.

1998-01-01

The present report examines the possibilities and limits of a probabilistic safety analysis to evaluate passive safety systems and inherent safety characteristics. The inherent safety characteristics are based on physical principles, that together with the safety system lead to no damage. A probabilistic evaluation of the inherent safety characteristic is not made. An inventory of passive safety systems of accomplished nuclear power plant types in the Federal Republic of Germany was drawn up. The evaluation of the passive safety system in the analysis of the accomplished nuclear power plant types was examined. The analysis showed that the passive manner of working was always assumed to be successful. A probabilistic evaluation was not performed. The unavailability of the passive safety system was determined by the failure of active components which are necessary in order to activate the passive safety system. To evaluate the passive safety features in new concepts of nuclear power plants the AP600 from Westinghouse, the SBWR from General Electric and the SWR 600 from Siemens, were selected. Under these three reactor concepts, the SWR 600 is specially attractive because the safety features need no energy sources and instrumentation in this concept. First approaches for the assessment of the reliability of passively operating systems are summarized. Generally it can be established that the core melt frequency for the passive concepts AP600 and SBWR is advantageous in comparison to the probabilistic objectives from the European Pressurized Water Reactor (EPR). Under the passive concepts is the SWR 600 particularly interesting. In this concept the passive systems need no energy sources and instrumentation, and has active operational systems and active safety equipment. Siemens argues that with this concept the frequency of a core melt will be two orders of magnitude lower than for the conventional reactors. (orig.) [de

Dynamics and inherent safety features of small modular high temperature gas-cooled reactors

International Nuclear Information System (INIS)

Harrington, R.M.; Ball, S.J.; Cleveland, J.C.

1986-01-01

Investigations were made at Oak Ridge National Laboratory to characterize the dynamics and inherent safety features of various modular high temperature gas-cooled reactor (HTGR) designs. This work was sponsored by the US Nuclear Regulatory Commission's HTGR Safety Research program. The US Department of Energy (DOE) and the Gas Cooled Reactor Associates (GCRA) have sponsored studies of several modular HTGR concepts, each having it own unique advantageous economic and inherent safety features. The DOE design team has recently choses a 350-MW(t) annular core with prismatic, graphite matrix fuel for its reference plant. The various safety features of this plant and of the pebble-bed core designs similar to those currently being developed and operated in the Federal Republic of Germany (FRG) are described. A varity of postulated accident sequences involving combinations of loss of forced circulation of the helium primary coolant, loss of primary coolant pressurization, and loss of normal and backup heat sinks were studied and are discussed. Results demonstrate that each concept can withstand an uncontrolled heatup accident without reaching excessive peak fuel temperatures. Comparisons of calculated and measured response for a loss of forced circulation test on the FRG reactor, AVR, are also presented. 10 refs

Developing a model for hospital inherent safety assessment: Conceptualization and validation.

Science.gov (United States)

Yari, Saeed; Akbari, Hesam; Gholami Fesharaki, Mohammad; Khosravizadeh, Omid; Ghasemi, Mohammad; Barsam, Yalda; Akbari, Hamed

2018-01-01

Paying attention to the safety of hospitals, as the most crucial institute for providing medical and health services wherein a bundle of facilities, equipment, and human resource exist, is of significant importance. The present research aims at developing a model for assessing hospitals' safety based on principles of inherent safety design. Face validity (30 experts), content validity (20 experts), construct validity (268 examples), convergent validity, and divergent validity have been employed to validate the prepared questionnaire; and the items analysis, the Cronbach's alpha test, ICC test (to measure reliability of the test), composite reliability coefficient have been used to measure primary reliability. The relationship between variables and factors has been confirmed at 0.05 significance level by conducting confirmatory factor analysis (CFA) and structural equations modeling (SEM) technique with the use of Smart-PLS. R-square and load factors values, which were higher than 0.67 and 0.300 respectively, indicated the strong fit. Moderation (0.970), simplification (0.959), substitution (0.943), and minimization (0.5008) have had the most weights in determining the inherent safety of hospital respectively. Moderation, simplification, and substitution, among the other dimensions, have more weight on the inherent safety, while minimization has the less weight, which could be due do its definition as to minimize the risk.

Integral fast reactor concept inherent safety features

International Nuclear Information System (INIS)

Marchaterre, J.F.; Sevy, R.H.; Cahalan, J.E.

1987-01-01

The Integral Fast Reactor (IFR) is an innovative liquid-metal-cooled reactor concept being developed at Argonne National Laboratory. The two major goals of the IFT development effort are improved economics and enhanced safety. The design features that together fulfill these goals are: 1) a liquid metal (sodium) coolant, 2) a pool-type reactor primary system configuration, 3) an advanced ternary alloy metallic fuel, and 4) an integral fuel cycle. This paper reviews the design features that contribute to the safety margins inherent to the IFR concept. Special emphasis is placed on the ability of the IFR design to accommodate anticipated transients without scram (ATWS)

Integral Fast Reactor concept inherent safety features

International Nuclear Information System (INIS)

Marchaterre, J.F.; Sevy, R.H.; Cahalan, J.E.

1986-01-01

The Integral Fast Reactor (IFR) is an innovative liquid-metal-cooled reactor concept being developed at Argonne National Laboratory. The two major goals of the IFR development effort are improved economics and enhanced safety. The design features that together fulfill these goals are: (1) a liquid metal (sodium) coolant, (2) a pool-type reactor primary system configuration, (3) an advanced ternary alloy metallic fuel, and (4) an integral fuel cycle. This paper reviews the design features that contribute to the safety margins inherent to the IFR concept. Special emphasis is placed on the ability of the IFR design to accommodate anticipated transients without scram (ATWS)

Role of FFTF in assessing structural feedbacks and inherent safety of LMR's

International Nuclear Information System (INIS)

Padilla, A.; Omberg, R.P.; O'Dell, L.D.; Harris, R.A.; Nguyen, D.H.; Waltar, A.E.

1985-03-01

The possibility of developing reactor designs with inherent safety characteristics sufficient to provide ''walk away'' safety is receiving additional emphasis in the LMR program. A key element in this effort is the recognition that LMR's possess safety characteristics above and beyond those employed in past safety review processes. Some of these additional safety characteristics are due to reactivity feedback effects caused by small structural movements during hypothetical severe design transients. The effect of these characteristics upon the behavior of the FFTF under such transients has been assessed and is discussed in this paper. The paper also presents a preliminary test matrix which might allow experimental verification of the structural reactivity feedback effects. Such experimental verification should be very useful to innovative designers seeking to optimize inherent safety. 8 refs., 1 fig., 2 tabs

Inherently safe reactors

International Nuclear Information System (INIS)

Maartensson, Anders

1992-01-01

A rethinking of nuclear reactor safety has created proposals for new designs based on inherent and passive safety principles. Diverging interpretations of these concepts can be found. This article reviews the key features of proposed advanced power reactors. An evaluation is made of the degree of inherent safety for four different designs: the AP-600, the PIUS, the MHTGR and the PRISM. The inherent hazards of today's most common reactor principles are used as reference for the evaluation. It is concluded that claims for the new designs being inherently, naturally or passively safe are not substantiated by experience. (author)

Technology which led to the westinghouse inherently safe liquid metal reactor

International Nuclear Information System (INIS)

Schmidt, J.E.; Coffield, R.D.; Doncals, R.A.; Kalinowski, J.E.; Markley, R.A.

1985-01-01

The Fast Flux Test Facility (FFTF) and the Clinch River Breeder Reactor programs resulted in an understanding of liquid metal reactor behavior that is being used to design inherent safety capability into liquid metal reactors. Technological advances give the same beneficial operating characteristics of conventional liquid metal reactors, however, the addition of inherently safe design features precludes the initiation of hypothetical core disruptive accidents. These innovative features permit inherent safety capability to be demonstrated with more than adequate margins. Also, the variety of inherent safety features provides the designers with options in selecting inherent design features for a specific reactor application

Inherent Safety Feature of Hybrid Low Power Research Reactor during Reactivity Induced Accident

Energy Technology Data Exchange (ETDEWEB)

Kim, DongHyun; Yum, Soo Been; Hong, Sung Teak; Lim, In-Cheol [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

Hybrid low power research reactor(H-LPRR) is the new design concept of low power research reactor for critical facility as well as education and training. In the case of typical low power research reactor, the purposes of utilization are the experiments for education of nuclear engineering students, Neutron Activation Analysis(NAA) and radio-isotope production for research purpose. H-LPRR is a light-water cooled and moderated research reactor that uses rod-type LEU UO{sub 2} fuels same as those for commercial power plants. The maximum core thermal power is 70kW and, the core is placed in the bottom of open pool. There are 1 control rod and 2 shutdown rods in the core. It is designed to cool the core by natural convection, retain negative feedback coefficient for entire fuel periods and operate for 20 years without refueling. Inherent safety in H-LPRR is achieved by passive design features such as negative temperature feedback coefficient and core cooling by natural convection during normal and emergency conditions. The purpose of this study is to find out that the inherent safety characteristics of H-LPRR is able to control the power and protect the reactor from the RIA(Reactivity induced accident). RIA analysis was performed to investigate the inherent safety feature of H-LPRR. As a result, it was found that the reactor controls its power without fuel damage in the event and that the reactor remains safe states inherently. Therefore, it is believed that high degree of safety inheres in H-LPRR.

Inherent safety, ethics and human error.

Science.gov (United States)

Papadaki, Maria

2008-02-11

stated. The reason this article is presented here is that I believe that often, complex accidents, similarly to insignificant ones, often demonstrate an attitude which can be characterized as "inherently unsafe". I take the view that the enormous human potential and the human ability to minimize accidents needs to become a focal point towards inherent safety. Restricting ourselves to human limitations and how we could "treat" or prevent humans from not making accidents needs to be re-addressed. The purpose of this presentation is to highlight observations and provoke a discussion on how we could possibly improve the understanding of safety related issues. I do not intent to reject or criticize existing methodologies. (The entire presentation is strongly influenced by Trevor Kletz's work although our views are often different.).

The minimum attention plant inherent safety through LWR simplification

International Nuclear Information System (INIS)

Turk, R.S.; Matzie, R.A.

1987-01-01

The Minimum Attention Plant (MAP) is a unique small LWR that achieves greater inherent safety, improved operability, and reduced costs through design simplification. The MAP is a self-pressurized, indirect-cycle light water reactor with full natural circulation primary coolant flow and multiple once-through steam generators located within the reactor vessel. A fundamental tenent of the MAP design is its complete reliance on existing LWR technology. This reliance on conventional technology provides an extensive experience base which gives confidence in judging the safety and performance aspects of the design

Inherent safety features in balance-of-plant layout

International Nuclear Information System (INIS)

Wattelet, P.L.; Green, K.J.

1992-01-01

Future nuclear units must be more economical to construct and operate, and, at the same time, clearly incorporate advances in safety over the current generation of light water reactors. To achieve these goals, the root causes of safety issues must be addressed. In this way, global, cost-effective solutions can be implemented. With simple, direct design approaches, the licensing risk is minimized and configuration control is enhanced. With proper planning in the early stages of plant design, postulated accidents and events can often be mitigated by passive features inherent in the basic structure and layout, eliminating expensive added protective structures and components often found in current designs. Korea Electric Power Corporation's Yonggwang (YGN) Units 3 and 4, shown in an artist's rendering in Figure 1, are now under construction in Korea. Engineering is more than 85% complete, and Unit 3 construction is more than 50% complete. Significant steps toward design simplification and safety enhancement have been made by addressing safety concerns very early in the design effort. The tools used to achieve this were improved symmetry and separation, isolation of potential hazards, and an improved design process

Safety demonstration test (SR-1/S1C-1) plan of HTTR (Contract research)

Energy Technology Data Exchange (ETDEWEB)

Nakagawa, Shigeaki; Sakaba, Nariaki; Takada, Eiji; Tachibana, Yukio; Saito, Kenji; Furusawa, Takayuki; Sawa, Kazuhiro [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

2003-03-01

Safety demonstration tests in the HTTR (High Temperature Engineering Test Reactor) will be carried out in order to verify inherent safety features of the HTGR (High Temperature Gas-cooled Reactor). The first phase of the safety demonstration tests includes the reactivity insertion test by the control rod withdrawal and the coolant flow reduction test by the circulator trip. In the second phase, accident simulation tests will be conducted. By comparison of their experimental and analytical results, the prediction capability of the safety evaluation codes such as the core and the plant dynamics codes will be improved and verified, which will contribute to establish the safety design and the safety evaluation technologies of the HTGRs. The results obtained through its safety demonstration tests will be also utilised for the establishment of the safety design guideline, the safety evaluation guideline, etc. This paper describes the test program of the overall safety demonstration tests and the test method, the test conditions and the results of the pre-test analysis of the reactivity insertion test and the partial gas circulator trip test planned in March 2003. (author)

Transient behaviour and inherent safety research of LMFBR power plants

International Nuclear Information System (INIS)

Zhu Jizhou; Wang Ping; Yu Baoan

1995-06-01

Fast Breeder Reactor will be the next generation reactor for nuclear electricity production, the development of FBR will give the profits of efficient utilization of nuclear resources. The fast reactor safety analysis is the foundation and key of FBR research work. Therefore, a block-oriented mathematical model for the primary system of LMFBRs was constructed, and the dynamic simulating results which have been carried out on micro-computer are presented for various transients, i.e. TOP, LOFS, LOHS. The results agree well with the corresponding results of the code NATDEMO and experiment results of EBR-II. Based on previous analysis, various methods are discussed to confirm the inherent safety of LMFBR

Social learning of fear and safety is determined by the demonstrator's racial group.

Science.gov (United States)

Golkar, Armita; Castro, Vasco; Olsson, Andreas

2015-01-01

Social learning offers an efficient route through which humans and other animals learn about potential dangers in the environment. Such learning inherently relies on the transmission of social information and should imply selectivity in what to learn from whom. Here, we conducted two observational learning experiments to assess how humans learn about danger and safety from members ('demonstrators') of an other social group than their own. We show that both fear and safety learning from a racial in-group demonstrator was more potent than learning from a racial out-group demonstrator. © 2015 The Author(s) Published by the Royal Society. All rights reserved.

An inherently safe power reactor module

International Nuclear Information System (INIS)

Salerno, L.N.

1985-01-01

General Electric's long participation in liquid metal reactor technology has led to a Power Reactor Inherently Safe Module (PRISM) concept supported by DOE contract DE-AC06-85NE37937. The reactor module is sized to maximize inherent safety features. The small size allows factory fabrication, reducing field construction and field QA/QC labor, and allows safety to be demonstrated in full scale, to support a pre-licensed standard commercial product. The module is small enough to be placed underground, and can be combined with steam and electrical generating equipment to provide a complete electrical power producing plant in the range of 400-1200 MWe. Initial assessments are that the concept has the potential to be economically competitive with existing methods of power production used by the utility industry

Inherently safe characteristics of nuclear reactors

International Nuclear Information System (INIS)

1989-01-01

This report is based on a detailed study which was carried out by Colenco (a company of the Motor-Columbus Group) on behalf of the Commission of the European Communities (CEC). It presents a summary of this study and concentrates more on the generic issues involved in the subject of inherent safety in nuclear power plants. It is assumed that the reader is reasonably familiar with the design outline of the systems included in the report. The report examines the role of inherent design features in achieving the safety of nuclear power plants as an alternative to the practice, which is largely followed in current reactors, of achieving safety by the addition of engineered safety features. The report examines current reactor systems to identify the extent to which their characteristics are either already inherently safe or, on the other hand, have inherent characteristics that require protective action to be taken. It then considers the advantages of introducing design changes to improve their inherent safety characteristics. Next, it looks at some new reactor types for which claims of inherent safety are made to see to what extent these claims are justified. The general question is then considered whether adoption of the inherently safe reactors would give advantages (by reducing risk in real terms or by improving the public acceptability of nuclear power) which are sufficient to offset the expected high costs and the technical risks associated with any new technology

On the fundamentals of nuclear reactor safety assessment. Inherent threats and their implications

Energy Technology Data Exchange (ETDEWEB)

Hyvaerinen, J. [Finnish Centre for Radiation and Nuclear Safety, Helsinki (Finland). Nuclear Safety Dept.

1996-12-01

The thesis addresses some fundamental questions related to implementation and assessment of nuclear safety. The safety principles and assessment methods are described, followed by descriptions of selected novel technical challenges to nuclear safety. The novel challenges encompass a wide variety of technical issues, thus providing insights on the limitations of conventional safety assessment methods. Study of the limitations suggests means to improve nuclear reactor design criteria and safety assessment practices. The novel safety challenges discussed are (1) inherent boron dilution in PWRs, (2) metallic insulation performance with respect to total loss of emergency cooling systems in a loss-of-coolant accident, and (3) horizontal steam generator heat transfer performance at natural circulation conditions. (50 refs.).

On the fundamentals of nuclear reactor safety assessment. Inherent threats and their implications

International Nuclear Information System (INIS)

Hyvaerinen, J.

1996-12-01

The thesis addresses some fundamental questions related to implementation and assessment of nuclear safety. The safety principles and assessment methods are described, followed by descriptions of selected novel technical challenges to nuclear safety. The novel challenges encompass a wide variety of technical issues, thus providing insights on the limitations of conventional safety assessment methods. Study of the limitations suggests means to improve nuclear reactor design criteria and safety assessment practices. The novel safety challenges discussed are (1) inherent boron dilution in PWRs, (2) metallic insulation performance with respect to total loss of emergency cooling systems in a loss-of-coolant accident, and (3) horizontal steam generator heat transfer performance at natural circulation conditions. (50 refs.)

Project of a binary breeder reactor and its inherent safety

International Nuclear Information System (INIS)

Nascimento, J.A. do; Dias, A.F.; Ishiguro, Y.

1983-01-01

A core layout for the binary breeder reactor (BBR) is developed based on the results of preliminary burnup calculations. The apparent breeding ratio, in the U 233 /Th fueled inner core, is low due to the accumulation of Pa-233 in the first few months of operation. The loss of reactivity during this time is around 3%. The BBR requires more reactivity control than Pu/U-fueled LMFBRs and the core layout developed has 19 control rod assemblies in the inner core. Three aspects related to the inherent safety of the Binary Breeder Reactor have been studied: the radial distribution of the sodium-void reactivity zone-wise Doppler reactivity and the fractions of delayed neutrons. The results show excellent characteristics for the BRB safety. (Author) [pt

Project of a binary breeder reactor and its inherent safety

International Nuclear Information System (INIS)

Nascimento, J.A. do; Dias, A.F.; Ishiguro, Y.

1983-01-01

A core layout for the binary breeder reactor (BBR) is developed based on the results of preliminary burnup calculations. In the U 233 /TH fueled inner core, the apparent breeding ratio is low due to the accumulation of Pa-233 in the first few months of operation. The loss of reactivity during this time is approximatelly 3%. The BBR requires more reactivity control than Pu/U-fueled LMFBRs and the core layout developed has 19 control rod assemblies in the inner core. Three aspects related to the inherent safety of the BBR have been studied: radial distribution of the sodium-void reactivity, zone-wise Doppler reactivity, and the delayed neutron fractions. Results show excellent safety characteristics of the BBR. (Author) [pt

Multi-objective optimization of a cascade refrigeration system: Exergetic, economic, environmental, and inherent safety analysis

International Nuclear Information System (INIS)

Eini, Saeed; Shahhosseini, Hamidreza; Delgarm, Navid; Lee, Moonyong; Bahadori, Alireza

2016-01-01

Highlights: • A multi-objective optimization is performed for a cascade refrigeration cycle. • The optimization problem considers inherently safe design as well as 3E analysis. • As a measure of inherent safety level a quantitative risk analysis is utilized. • A CO 2 /NH 3 cascade refrigeration system is compared with a CO 2 /C 3 H 8 system. - Abstract: Inherently safer design is the new approach to maximize the overall safety of a process plant. This approach suggests some risk reduction strategies to be implemented in the early stages of design. In this paper a multi-objective optimization was performed considering economic, exergetic, and environmental aspects besides evaluation of the inherent safety level of a cascade refrigeration system. The capital costs, the processing costs, and the social cost due to CO 2 emission were considered to be included in the economic objective function. Exergetic efficiency of the plant was considered as the second objective function. As a measure of inherent safety level, Quantitative Risk Assessment (QRA) was performed to calculate total risk level of the cascade as the third objective function. Two cases (ammonia and propane) were considered to be compared as the refrigerant of the high temperature circuit. The achieved optimum solutions from the multi–objective optimization process were given as Pareto frontier. The ultimate optimal solution from available solutions on the Pareto optimal curve was selected using Decision-Makings approaches. NSGA-II algorithm was used to obtain Pareto optimal frontiers. Also, three decision-making approaches (TOPSIS, LINMAP, and Shannon’s entropy methods) were utilized to select the final optimum point. Considering continuous material release from the major equipment in the plant, flash and jet fire scenarios were considered for the CO 2 /C 3 H 8 cycle and toxic hazards were considered for the CO 2 /NH 3 cycle. The results showed no significant differences between CO 2 /NH 3 and

Inherent safety features of the HTTR revealed in the accident condition

International Nuclear Information System (INIS)

Kunitomi, K.; Shinozaki, M.; Baba, O.; Saito, S.

1992-01-01

The High Temperature Engineering Test Reactor (HTTR) being constructed by JAERI (Japan Atomic Energy Research Institute) is a graphite-moderated and helium-cooled reactor with an outlet gas temperature of 950degC. The inherent safety characteristics in the HTTR prevent temperature increase of reactor fuels and fission product release from the reactor core in postulated accident conditions. The reactor core can be cooled by a Vessel Cooling System (VCS) indirectly, even in the case that no forced cooling is expected during the accident such as primary pipe break. The VCS consists of independent water cooling loop and cooling panel around the reactor pressure vessel. The cooling panel whose temperature of 60-90degC cools the reactor pressure vessel by radiation and removes the decay heat from the core indirectly. Furthermore, even if failure of VCS is assumed during this accident as a severe accident, the reactor core is remained safe despite the temperature increase of biological concrete shield around the reactor pressure vessel. This paper describes the inherent safety features of the HTTR specially focused on the accident condition without forced cooling. The detailed analytical results of such an accident are described together with clarifying the role of the VCS. (author)

Concept of an inherently-safe high temperature gas-cooled reactor

International Nuclear Information System (INIS)

Ohashi, Hirofumi; Sato, Hiroyuki; Tachibana, Yukio; Kunitomi, Kazuhiko; Ogawa, Masuro

2012-01-01

As the challenge to ensure no harmful release of radioactive materials at the accidents by deterministic approach instead to satisfy acceptance criteria or safety goal for risk by probabilistic approach, new concept of advanced reactor, an inherently-safe high temperature gas-cooled reactor, is proposed based on the experience of the operation of the actual High Temperature Gas-cooled Reactor (HTGR) in Japan, High Temperature Engineering Test Reactor (HTTR), and the design of the commercial plant (GTHTR300), utilizing the inherent safety features of the HTGR (i.e., safety features based on physical phenomena). The safety design philosophy of the inherently-safe HTGR for the safety analysis of the radiological consequences is determined as the confinement of radioactive materials is assured by only inherent safety features without engineered safety features, AC power or prompt actions by plant personnel if the design extension conditions occur. Inherent safety features to prevent the loss or degradation of the confinement function are identified. It is proposed not to apply the probabilistic approach for the evaluation of the radiological consequences of the accidents in the safety analysis because no inherent safety features fail for the mitigation of the consequences of the accidents. Consequently, there are no event sequences to harmful release of radioactive materials if the design extension conditions occur in the inherently-safe HTGR concept. The concept and future R and D items for the inherently-safe HTGR are described in this paper.

Inherent and passive safety measures in accelerator driven systems: a safety strategy for ADS

International Nuclear Information System (INIS)

Maschek, W.; Rineiski, A.; Morita, K.; Flad, M.

2001-01-01

The efficiency of Accelerator Driven Systems (ADSs) for the transmutation and incineration of nuclear waste is strongly related to the utilization of so-called dedicated fuels. In the ideal case these fuels should consist of pure TRUs without fertile materials as 238 U or 232 Th to achieve highest incineration/transmutation rates. Dedicated fuels still have to be developed and programs are under way for their fabrication, irradiation and testing. These fertile-free fuels may suffer from deteriorated thermal or thermo-mechanical properties, as a reduced melting point, reduced thermal conductivity or even thermal instability. First analyses have shown that the use of dedicated fuels may lead to a strong deterioration of the safety parameters of the reactor core as e.g. the void worth, the Doppler or the kinetics quantities as neutron generation time and β eff . In addition, a dedicated core may contain multiple ''critical'' fuel masses, resulting in a considerable recriticality potential. Current knowledge on these dedicated fuels suggests that ''critical'' reactors may not be feasible, because of safety reasons. However, for ADSs, the salient hope has been promoted that due to the subcriticality of the system the poor safety features of such fuels could be coped with. Analyses are presented which show potential safety problems for such dedicated cores. Respecting the results of these analyses a safety strategy is proposed along the lines of defense approach in analogy with ideas formerly developed for fast reactors. Inherent and passive safety measures are integrated into the various defense lines. (author)

Efficiency of inherent protection mechanisms for an improved HTR safety concept

Energy Technology Data Exchange (ETDEWEB)

Petersen, K.; Buescher, R.; Gerwin, H.; Schenk, W. [Kernforschungsanlage Juelich G.m.b.H. (Germany, F.R.). Inst. fuer Reaktorentwicklung

1981-01-15

For a preliminary design of a 350 MWsub(th) annular core derived from AVR-reactor the efficiency of inherent protection mechanisms is discussed. After-heat removal and auto-shut down potential are demonstrated for intact and complete failure of core heat sinks.

The development of an inherent safety approach to the prevention of domino accidents.

Science.gov (United States)

Cozzani, Valerio; Tugnoli, Alessandro; Salzano, Ernesto

2009-11-01

The severity of industrial accidents in which a domino effect takes place is well known in the chemical and process industry. The application of an inherent safety approach for the prevention of escalation events leading to domino accidents was explored in the present study. Reference primary scenarios were analyzed and escalation vectors were defined. Inherent safety distances were defined and proposed as a metric to express the intensity of the escalation vectors. Simple rules of thumb were presented for a preliminary screening of these distances. Swift reference indices for layout screening with respect to escalation hazard were also defined. Two case studies derived from existing layouts of oil refineries were selected to understand the potentialities coming from the application in the methodology. The results evidenced that the approach allows a first comparative assessment of the actual domino hazard in a layout, and the identification of critical primary units with respect to escalation events. The methodology developed also represents a useful screening tool to identify were to dedicate major efforts in the design of add-on measures, optimizing conventional passive and active measures for the prevention of severe domino accidents.

Worldwide advanced nuclear power reactors with passive and inherent safety: What, why, how, and who

International Nuclear Information System (INIS)

Forsberg, C.W.; Reich, W.J.

1991-09-01

The political controversy over nuclear power, the accidents at Three Mile Island (TMI) and Chernobyl, international competition, concerns about the carbon dioxide greenhouse effect and technical breakthroughs have resulted in a segment of the nuclear industry examining power reactor concepts with PRIME safety characteristics. PRIME is an acronym for Passive safety, Resilience, Inherent safety, Malevolence resistance, and Extended time after initiation of an accident for external help. The basic ideal of PRIME is to develop power reactors in which operator error, internal sabotage, or external assault do not cause a significant release of radioactivity to the environment. Several PRIME reactor concepts are being considered. In each case, an existing, proven power reactor technology is combined with radical innovations in selected plant components and in the safety philosophy. The Process Inherent Ultimate Safety (PIUS) reactor is a modified pressurized-water reactor, the Modular High Temperature Gas-Cooled Reactor (MHTGR) is a modified gas-cooled reactor, and the Advanced CANDU Project is a modified heavy-water reactor. In addition to the reactor concepts, there is parallel work on super containments. The objective is the development of a passive ''box'' that can contain radioactivity in the event of any type of accident. This report briefly examines: why a segment of the nuclear power community is taking this new direction, how it differs from earlier directions, and what technical options are being considered. A more detailed description of which countries and reactor vendors have undertaken activities follows. 41 refs

Inherently safe high temperature gas-cooled reactor

International Nuclear Information System (INIS)

Yamada, Masao; Hayakawa, Hitoshi

1987-01-01

It is recognized in general that High Temperature Gas-cooled Reactors have remarkable characteristics in inherent safety and it is well known that credits of the time margin have been admitted for accident evaluation in the licensing of the currently operating prototype HTGRs (300 MWe class). Recently, more inherently safe HTGRs are being developed in various countries and drawing attention on their possibility for urban siting. The inherent safety characteristics of these HTRs differ each other depending on their design philosophy and on the features of the components/structures which constitute the plant. At first, the specific features/characteristics of the elemental components/structures of the HTRs are explained one by one and then the overall safety features/characteristics of these HTR plants are explained in connection with their design philosophy and combination of the elemental features. Taking the KWU/Interatom Modular Reactor System as an example, the particular design philosophy and safety characteristics of the inherently safe HTR are explained with a result of preliminary evaluation on the possibility of siting close to densely populated area. (author)

Inherently safe light water reactors

International Nuclear Information System (INIS)

Ise, Takeharu

1987-01-01

Today's large nuclear power reactors of world-wise use have been designed based on the philosophy. It seems that recent less electricity demand rates, higher capital cost and the TMI accident let us acknowledge relative small and simplified nuclear plants with safer features, and that Chernobyl accident in 1983 underlines the needs of intrinsic and passive safety characteristics. In such background, several inherently safe reactor concepts have been presented abroad and domestically. First describing 'Can inherently safe reactors be designed,' then I introduce representative reactor concepts of inherently safe LWRs advocated abroad so far. All of these innovative reactors employ intrinsic and passive features in their design, as follows: (1) PIUS, an acronym for Process Inherent Ultimate Safety, or an integral PWR with passive heat sink and passive shutdown mechanism, advocated by ASEA-ATOM of Sweden. (2) MAP(Minimum Attention Plant), or a self-pressurized, natural circulation integral PWR, promoted by CE Inc. of the U.S. (3) TPS(TRIGA Power System), or a compact PWR with passive heat sink and inherent fuel characteristics of large prompt temperature coefficient, prompted by GA Technologies Inc. of the U.S. (4) PIUS-BWR, or an inherently safe BWR employing passively actuated fluid valves, in competition with PIUS, prompted by ORNL of the U.S. Then, I will describe the domestic trends in Japan and the innovative inherently safe LWRs presented domestically so far. (author)

Spacecraft Fire Safety Demonstration

Data.gov (United States)

National Aeronautics and Space Administration — The objective of the Spacecraft Fire Safety Demonstration project is to develop and conduct large-scale fire safety experiments on an International Space Station...

Improved inherent safety in liquid fuel reactors

International Nuclear Information System (INIS)

Taube, M.

1982-01-01

The molten salt reactor system divided into core (thermal and fast) and breeding zone (fission breeder reactor, fusion hybrid system, accelerator-spallation system) has some unique inherent safety properties: a) reduced inventory of fission products during normal operation due to on-line chemical reprocessing and in-core gas purging; b) fast removal of freshly bred fissile nuclides and fission products from the breeding zone (the so called suppressed fission system); c) pressureless fuel and primary coolant system; d) elimination of the possibility of a violent exoenergetic chemical reaction with air, water or metals; e) elimination of the possibility of gaseous hydrogen production during an accident; f) provides a non-engineered feature of dumping of fuel from the core and heat exchanger to a safe drain tank; g) presence of a large heat sink in the form of an inactive diluent salt; h) possibility of natural convection heat removal during an accident and even normal operation (by means of gas lifting); i) dissipation of the remaining decayheat by spraying water on the containment from outside, which allows to manage the worst accident; i) Even in the case of the destruction in the war by conventional or nuclear weapon the contaminated land is significantly reduced. The world-wide present activity in the field of molten salt technology is reviewed. (orig.)

Demonstration of passive safety features in EBR-II

International Nuclear Information System (INIS)

Planchon, H.P. Jr.; Golden, G.H.; Sackett, J.I.

1987-01-01

Two tests of great importance to the design of future commercial nuclear power plants were carried out in the Experimental Breeder Reactor-II on April 3, 1986. These tests, (viewed by about 60 visitors, including 13 foreign LMR specialists) were a loss of flow without scram and a loss of heat sink without scram, both from 100% initial power. In these tests, inherent feedback shut the reactor down without damage to the fuel or other reactor components. This resulted primarily from advantageous characteristics of the metal driver fuel used in EBR-II. Work is currently underway at EBR-II to develop a control strategy that promotes inherent safety characteristics, including survivability of transient overpower accidents. In parallel, work is underway at EBR-II on the development of state-of-the-art plant diagnostic techniques

Cost-competitive, inherently safe LFMBR pool plant

International Nuclear Information System (INIS)

McDonald, J.S.; Brunings, J.E.; Chang, Y.I.; Hren, R.R.; Seidensticker, R.W.

1984-01-01

The Cost-Competitive, Inherently Safe LMFBR Pool Plant design was prepared in GFY 1983 under a DOE-sponsored program. This plant design was developed as a joint effort by Rockwell International and the Argonne National Laboratory with major contributions from the Bechtel Group, Inc.; Combustion engineering, Inc.; the Chicago Bridge and Iron Company; and the General Electric Company. Using current LMFBR technology, many innovative features were developed and incorporated into the design to meet the ultimate objectives of the Breeder Program, i.e., energy costs competitive with LWRs and inherent safety features to maintain the plant in a safe condition following assumed accidents without requiring operator action. This paper provides a description of the principal features that were incorporated into the design to achieve low cost and inherent safety

Prospects for inherently safe reactors

International Nuclear Information System (INIS)

Barkenbus, J.N.

1988-01-01

Public fears over nuclear safety have led some within the nuclear community to investigate the possibility of producing inherently safe nuclear reactors; that is, reactors that are transparently incapable of producing a core melt. While several promising designs of such reactors have been produced, support for large-scale research and development efforts has not been forthcoming. The prospects for commercialization of inherently safe reactors, therefore, are problematic; possible events such as further nuclear reactor accidents and superpower summits, could alter the present situation significantly. (author)

Inherently safe nuclear-driven internal combustion engines

International Nuclear Information System (INIS)

Alesso, P.; Chow, Tze-Show; Condit, R.; Heidrich, J.; Pettibone, J.; Streit, R.

1991-01-01

A family of nuclear driven engines is described in which nuclear energy released by fissioning of uranium or plutonium in a prompt critical assembly is used to heat a working gas. Engine performance is modeled using a code that calculates hydrodynamics, fission energy production, and neutron transport self-consistently. Results are given demonstrating a large negative temperature coefficient that produces self-shutoff of energy production. Reduced fission product inventory and the self-shutoff provide inherent nuclear safety. It is expected that nuclear engine reactor units could be scaled from 100 MW on up. 7 refs., 3 figs

Process Inherent Ultimate Safety (PIUS) reactor evaluation study: Final report

International Nuclear Information System (INIS)

1987-02-01

This report presents the results of an independent study by United Engineers and Constructors (UNITED) of the SECURE-P Process Inherent Ultimate Safety (PIUS) Reactor Concept which is presently under development by the Swedish light water reactor vendor ASEA-ATOM of Vasteras, Sweden. This study was performed to investigate whether there is any realistic basis for believing that the PIUS reactor could be a viable competitor in the US energy market in the future. Assessments were limited to the technical, economic and licensing aspects of PIUS. Socio-political issues, while certainly important in answering this question, are so broad and elusive that it was considered that addressing them with the limited perspective of one small group from one company would be of questionable value and likely be misleading. Socio-political issues aside, the key issue is economics. For this reason, the specific objectives of this study were to determine if the estimated PIUS plant cost will be competitive in the US market and to identify and evaluate the technical and licensing risks that might make PIUS uneconomical or otherwise unacceptable

MHTGR inherent heat transfer capability

International Nuclear Information System (INIS)

Berkoe, J.M.

1992-01-01

This paper reports on the Commercial Modular High Temperature Gas-Cooled Reactor (MHTGR) which achieves improved reactor safety performance and reliability by utilizing a completely passive natural convection cooling system called the RCCS to remove decay heat in the event that all active cooling systems fail to operate. For the highly improbable condition that the RCCS were to become non-functional following a reactor depressurization event, the plant would be forced to rely upon its inherent thermo-physical characteristics to reject decay heat to the surrounding earth and ambient environment. A computational heat transfer model was created to simulate such a scenario. Plant component temperature histories were computed over a period of 20 days into the event. The results clearly demonstrate the capability of the MHTGR to maintain core integrity and provide substantial lead time for taking corrective measures

Passengers' perception of the safety demonstration on board an aircraft

Science.gov (United States)

Ruenruoy, Ratchada

The cabin safety demonstration on board an aircraft is one of the methods to provide safety information for passengers before aircraft takeoff. However, passengers' enthusiasm toward safety demonstrations is normally low. Therefore, the study of passengers' perception toward safety briefings on board an aircraft is important in increasing the safety awareness for the travelling public on commercial aircraft. A survey was distributed to measure the perceptions of Middle Tennessee State University (MTSU) faculty and staff, Aerospace students, and international students who have traveled in the last year. It was generally found that watching the cabin safety demonstration before aircraft takeoff was believed to be important for passengers. However, the attention to the safety demonstration remained low because the safety briefings were not good enough in terms of clear communication, particularly in the recorded audio demonstration and the live safety demonstration methods of briefing.

Summary view on demonstration reactor safety

International Nuclear Information System (INIS)

Satoh, Kazuziro; Kotake, Shoji; Tsukui, Yutaka; Inagaki, Tatsutoshi; Miura, Masanori

1991-01-01

This work presents a summary view on safety design approaches for the demonstration fast breeder reactor (DFBR). The safety objective of DFBR is to be at lea as safe as a LWR. Major safety issues discussed in this paper are; reduction of sodium void reactivity worth, adoption of self-actuated mechanism in the backup shutdown system, use of the direct reactor auxiliary cooling system (DRACS), provision of the containment system. (author)

Toxic release consequence analysis tool (TORCAT) for inherently safer design plant

International Nuclear Information System (INIS)

Shariff, Azmi Mohd; Zaini, Dzulkarnain

2010-01-01

Many major accidents due to toxic release in the past have caused many fatalities such as the tragedy of MIC release in Bhopal, India (1984). One of the approaches is to use inherently safer design technique that utilizes inherent safety principle to eliminate or minimize accidents rather than to control the hazard. This technique is best implemented in preliminary design stage where the consequence of toxic release can be evaluated and necessary design improvements can be implemented to eliminate or minimize the accidents to as low as reasonably practicable (ALARP) without resorting to costly protective system. However, currently there is no commercial tool available that has such capability. This paper reports on the preliminary findings on the development of a prototype tool for consequence analysis and design improvement via inherent safety principle by utilizing an integrated process design simulator with toxic release consequence analysis model. The consequence analysis based on the worst-case scenarios during process flowsheeting stage were conducted as case studies. The preliminary finding shows that toxic release consequences analysis tool (TORCAT) has capability to eliminate or minimize the potential toxic release accidents by adopting the inherent safety principle early in preliminary design stage.

Improvement of inherent safety features in CSR (Coupled Spectrum Reactor) for treating MA

International Nuclear Information System (INIS)

Aziz, F.; Kitamoto, Asashi.

1996-01-01

Burning and/or transmutation (B/T) of MA is proposed here using a CSR (Coupled Spectrum Reactor) concept. CSR was based on a modified conventional 1150 MWe-PWR system, and consisted of two core regions for thermal and fast neutrons, respectively. The B/T fuel used was supposed such that MA discharged from 1 GWe-LWR were mixed homogeneously in LWR fuel. The geometry of B/T fuel in the outer region was left the same with that of PWR, while in the inner region the B/T fuel was arranged in a tight-lattice geometry that allowed a higher fuel to coolant volume ratio, (V m /V f ). In order to improve its inherent safety features, several cases of CSR were studied and compared, each case used different fuel type in the inner region. The result of the calculations showed that safety features can be improved by using composite fuel of ( 235 U-Pu- 238 U) in the inner region. The equilibrium of main isotopes in CSR can be achieved after about 5 recycle stages. This study also showed that the CSR can burn and transmute MA up to 808 kg/stage in a single reactor operated with a reactivity swing of 2.8 % Δk/kk'. (author)

An assessment of the low seismic risk of the inherently safe sodium advanced fast reactor (SAFR)

International Nuclear Information System (INIS)

Rutherford, P.D.

1988-01-01

A recent probabilistic risk assessment (PRA) of the sodium advanced fast reactor (SAFR) demonstrated the inherently low risk of advanced liquid-metal, pool-type fast reactors with inherent safety systems. As a result, it was recognized that external events, especially seismic events, may not only be a major contributor to risk (as shown in several LWR PRAs) but also may completely dominate the risk. Accordingly, a seismic risk assessment has been completed for SAFR, which resulted in a core damage frequency of 2 x 10 -7 /year and a large release frequency of 4 x 10 -9 /year. This paper reports that public health risk in terms of early fatality risk and latent fatality risk were also several orders of magnitude below the NRC safety goals and below recent LWR risks reported in NUREB/CR1150

Safety demonstration test (SR-3/S1C-3/S2C-3/SF-2) plan using the HTTR. Contract research

International Nuclear Information System (INIS)

Nakagawa, Shigeaki; Sakaba, Nariaki; Takamatsu, Kuniyoshi; Takada, Eiji; Tochio, Daisuke; Ohwada, Hiroyuki

2005-03-01

Safety demonstration tests using the HTTR are to be conducted from the FY2002 to verify the inherent safety features and to improve the safety design and evaluation technologies for HTGRs, as well as to contribute to not only the commercial HTGRs but also the research and development for the VHTR that is one of the Generation IV reactor candidates. This paper describes the reactivity insertion test (SR-3), the coolant flow reduction test by tripping of gas circulators (S1C-3, S2C-3), and the partial flow loss of coolant test (SF-2) planned in March 2005 with their detailed test method, procedure and results of pre-test analysis. From the analytical results, it was verified that the negative reactivity feedback effect of the core brings the reactor power safely to a stable level without a reactor scram. (author)

Implications of inherent safe nuclear power system

International Nuclear Information System (INIS)

Song, Yo-Taik

1987-01-01

The safety of present day nuclear power reactors and research reactors depends on a combination of design features of passive and active systems, and the alert judgement of their operators. A few inherently safe designs of nuclear reactors for power plants are currently under development. In these designs, the passive systems are emphasized, and the active systems are minimized. Also efforts are made to eliminate the potential for human failures that initiate the series of accidents. If a major system fails in these designs, the core is flooded automatically with coolants that flow by gravity, not by mechanical pumps or electromagnetic actuators. Depending on the choice of the coolants--water, liquid metal and helium gas--there are three principal types of inherently safe reactors. In this paper, these inherently safe reactor designs are reviewed and their implications are discussed. Further, future perspectives of their acceptance by nuclear industries are discussed. (author)

Key asset - inherent safety of LMFBR Pool Plant

International Nuclear Information System (INIS)

Marchaterre, J.F.; Sevy, R.H.; Lancet, R.T.; Mills, J.C.

1984-04-01

The safety approach used in the design of the Large Pool Plant emphasizes use of the intrinsic characteristics of Liquid Metal Fast Breeder Reactors to incorporate a high degree of safety in the design and reduce cost by providing simpler (more reliable) dedicated safety systems. Correspondingly, a goal was not to require the action of active systems to prevent significant core damage and/or provide large grace periods for all anticipated transients. The key safety features of the plant are presented and the analysis of representative flow and power transients are presented to show that the design goal has been satisfied

Key asset--Inherent safety of LFMBR pool plant

International Nuclear Information System (INIS)

Marchaterre, J.F.; Lancet, R.T.; Mills, J.C.; Sevy, R.H.

1984-01-01

The safety approach used in the design of the Large Pool Plant emphasizes use of the intrinsic characteristics of Liquid Metal Fast Breeder Reactors to incorporate a high degree of safety in the design and reduce cost by providing simpler (more reliable) dedicated safety systems. Correspondingly, a goal was not to require the action of active systems to prevent significant core damage and/or provide large grace periods for all anticipated transients. The key safety features of the plant are presented and the analysis of representative flow and power transients are presented to show that the design goal has been satisfied

Ordeals of Chernobyl and the rejustification of the inherently safe reactors

International Nuclear Information System (INIS)

Lu, Y.

1989-01-01

This paper presents the necessity of developing inherently safe economic reactors (ISERs). Two characteristics which define inherent safety are discussed on the basis of various applications of such a principle in practice. Different design concepts of ISERs are then evaluated and their possible role in the future nuclear program of PRC discussed. A three-stage development strategy of ISERs in PRC is proposed

Mitigation of Severe Accident Consequences Using Inherent Safety Principles

International Nuclear Information System (INIS)

Wigeland, R.A.; Cahalan, J.E.

2009-01-01

Sodium-cooled fast reactors are designed to have a high level of safety. Events of high probability of occurrence are typically handled without consequence through reliable engineering systems and good design practices. For accidents of lower probability, the initiating events are characterized by larger and more numerous challenges to the reactor system, such as failure of one or more major engineered systems and can also include a failure to scram the reactor in response. As the initiating conditions become more severe, they have the potential for creating serious consequences of potential safety significance, including fuel melting, fuel pin disruption and recriticality. If the progression of such accidents is not mitigated by design features of the reactor, energetic events and dispersal of radioactive materials may result. For severe accidents, there are several approaches that can be used to mitigate the consequences of such severe accident initiators, which typically include fuel pin failures and core disruption. One approach is to increase the reliability of the reactor protection system so that the probability of an ATWS event is reduced to less than 1 x 10-6 per reactor year, where larger accident consequences are allowed, meeting the U.S. NRC goal of relegating such accident consequences as core disruption to these extremely low probabilities. The main difficulty with this approach is to convincingly test and guarantee such increased reliability. Another approach is to increase the redundancy of the reactor scram system, which can also reduce the probability of an ATWS event to a frequency of less than 1 x 10-6 per reactor year or lower. The issues with this approach are more related to reactor core design, with the need for a greater number of control rod positions in the reactor core and the associated increase in complexity of the reactor protection system. A third approach is to use the inherent reactivity feedback that occurs in a fast reactor to

The inherently-safe power reactor DYONISOS (Dynamic Nuclear Inherently-Safe Reactor Operating with Spheres)

International Nuclear Information System (INIS)

Taube, M.; Lanfranchi, M.; Weissenfluh, Th. von; Ligou, J.; Yadigaroglu, G.; Taube, P.

1986-01-01

A philosophy of inherent safety is formulated and an inherently-safe thermal power reactor is presented. Solid fuel in the form of spheres a few centimetres in diameter is suspended under the hydro-dynamic pressure of molten lead coolant in vertical channels within the graphite moderator. Loss of main pump pressure, or a loss-of-coolant accident (LOCA), results in immediate removal of the fuel to rigid sieves below the core, with consequent subcriticality. Residual and decay heat are carried away by thermal conduction through the coolant or, in the case of a LOCA, by a combination of radiation and natural convection of cover gas or incoming air from the fuel to the reactor vessel and convection of air between the vessel and steel containment wall. All decay heat removal systems are passive, though actively initiated external spray cooling of the containment can be used to reduce wall temperature. This, however, is only necessary in the case of a LOCA and after a period of 24 h. (author)

Integral Inherently Safe Light Water Reactor (I2S-LWR)

International Nuclear Information System (INIS)

Petrovic, Bojan; Memmott, Matthew; Boy, Guy; Charit, Indrajit; Manera, Annalisa; Downar, Thomas; Lee, John; Muldrow, Lycurgus; Upadhyaya, Belle; Hines, Wesley; Haghighat, Alierza

2017-01-01

This final report summarizes results of the multi-year effort performed during the period 2/2013- 12/2016 under the DOE NEUP IRP Project ''Integral Inherently Safe Light Water Reactors (I 2 S-LWR)''. The goal of the project was to develop a concept of a 1 GWe PWR with integral configuration and inherent safety features, at the same time accounting for lessons learned from the Fukushima accident, and keeping in mind the economic viability of the new concept. Essentially (see Figure 1-1) the project aimed to implement attractive safety features, typically found only in SMRs, to a larger power (1 GWe) reactor, to address the preference of some utilities in the US power market for unit power level on the order of 1 GWe.

SVBR-75/100 multi-purpose modular inherent-safety fast reactor

International Nuclear Information System (INIS)

Dragunov, Yu.G.; Stepanov, V.S.; Klimov, N.N.; Dedul, A.V.; Zrodnokov, A.V.; Toshinsky, G.I.; Komlev, O.G.; Krushelnitsky, V.N.; Takh, S.M.

2006-01-01

In this century energy consumption, including electric power, will continue growing on a large scale especially in developing countries. Significant changes in electric power market needs are to be expected in the direction of decreasing and varying the capacity of power sources. To satisfy the expected growth of demand for electric power and to take a decision concerning the ways of further development of global power, including nuclear engineering, it is very important to continue the development of innovative concepts of nuclear power sources, which might successfully compete with alternative power technologies at the future power markets. The proposed nuclear power source (or in other words - reactor plant) of new generation is supposed: - to have small power capacity in the range of 10 - 100 MW (electric) and possibility of its multi-purpose application (independent nuclear power source for desalination installations and electricity supply, nuclear power plants (NPP) of various capacity and purpose; - to use modular principle of construction of NPP of various capacity on the basis of unified 'typical' reactor plants; - to have qualitatively new level of passive safety and possess properties of inherent safety, deterministically excluding any opportunity of severe accidents; - to have an opportunity to use different kinds of fuel and to work in various fuel cycles at various stages of development of nuclear power without change in the design. And also to have long (7-10 years, and in the long term 15-20 years) core life time and enrichment on U-235 not higher than 20 % (which is in compliance with recommendations of IAEA under non-proliferation condition); - to be completely factory-manufactured, and an opportunity of its safe transportation to and from the NPP site shall be provided. Unified multi-purpose reactor plant SVBR-75/100 (Lead-Bismuth Fast Reactor with equivalent electric power of 75 - 100 MW-e depending on the steam parameters) meets the set of the

The application of probabilistic risk assessment to inherently safe reactors

International Nuclear Information System (INIS)

Cave, L.; Kastenberg, W.E.

1987-01-01

In the development of safety goals and design criteria for 'inherently safe' reactors a question which arises is 'To what extent is PRA relevant.' To answer this question it is necessary to consider both the risk to the public and the investment risk to the utility. In this paper the factors which are likely to determine safety objectives and their allocation are presented. (orig.)

Inherently safe technologies-chemical and nuclear

International Nuclear Information System (INIS)

Weinberg, A.M.

1984-01-01

Probabilistic risk assessments show an inverse relationship between the likelihood and the consequences of nuclear and chemical plant accidents, but the Bhopal accident has change public complacency about the safety of chemical plants to such an extent that public confidence is now at the same low level as with nuclear plants. The nuclear industry's response was to strengthen its institutions and improve its technologies, but the public may not be convinced. One solution is to develop reactors which do not depend upon the active intervention of humans of electromechanical devices to deal with emergencies, but which have physical properties that limit the possible temperature and power of a reactor. The Process Inherent Ultimately Safe and the modular High-Temperature Gas-Cooled reactors are two possibilities. the chemical industry needs to develop its own inherently safe design precepts that incorporate smallness, safe processes, and hardening against sabotage. 5 references

PA activity by using nuclear power plant safety demonstration and analysis

International Nuclear Information System (INIS)

Tsuchiya, Mitsuo; Kamimae, Rie

1999-01-01

INS/NUPEC presents one of Public acceptance (PA) methods for nuclear power in Japan, 'PA activity by using Nuclear Power Plant Safety Demonstration and Analysis', by using one of videos which is explained and analyzed accident events (Loss of Coolant Accident). Safety regulations of The National Government are strictly implemented in licensing at each of basic design and detailed design. To support safety regulation activities conducted by the National Government, INS/NLTPEC continuously implement Safety demonstration and analysis. With safety demonstration and analysis, made by assuming some abnormal conditions, what impacts could be produced by the assumed conditions are forecast based on specific design data on a given nuclear power plants. When analysis results compared with relevant decision criteria, the safety of nuclear power plants is confirmed. The decision criteria are designed to help judge if or not safety design of nuclear power plants is properly made. The decision criteria are set in the safety examination guidelines by taking sufficient safety allowance based on the latest technical knowledge obtained from a wide range of tests and safety studies. Safety demonstration and analysis is made by taking the procedure which are summarized in this presentation. In Japan, various PA (Public Acceptance) pamphlets and videos on nuclear energy have been published. But many of them focused on such topics as necessity or importance of nuclear energy, basic principles of nuclear power generation, etc., and a few described safety evaluation particularly of abnormal and accident events in accordance with the regulatory requirements. In this background, INS/NUPEC has been making efforts to prepare PA pamphlets and videos to explain the safety of nuclear power plants, to be simple and concrete enough, using various analytical computations for abnormal and accident events. In results, PA activity of INS/NUPEC is evaluated highly by the people

Nuclear Criticality Safety Assessment for Tank 38H Salt Dissolution

International Nuclear Information System (INIS)

Davis, P.L.

1996-01-01

This assessment report of sample results of the accumulating insoluble solids from Tank 38H demonstrates that an inherent subcritical condition for nuclear criticality safety exists during saltcake dissolution. This report also defines criteria for future sampling of Tank 38H for continued verification of the inherent subcritical condition as saltcake dissolution proceeds

Vowel Inherent Spectral Change

CERN Document Server

Assmann, Peter

2013-01-01

It has been traditional in phonetic research to characterize monophthongs using a set of static formant frequencies, i.e., formant frequencies taken from a single time-point in the vowel or averaged over the time-course of the vowel. However, over the last twenty years a growing body of research has demonstrated that, at least for a number of dialects of North American English, vowels which are traditionally described as monophthongs often have substantial spectral change. Vowel Inherent Spectral Change has been observed in speakers’ productions, and has also been found to have a substantial effect on listeners’ perception. In terms of acoustics, the traditional categorical distinction between monophthongs and diphthongs can be replaced by a gradient description of dynamic spectral patterns. This book includes chapters addressing various aspects of vowel inherent spectral change (VISC), including theoretical and experimental studies of the perceptually relevant aspects of VISC, the relationship between ar...

Inherent safe design of advanced high temperature reactors - concepts for future nuclear power plants

International Nuclear Information System (INIS)

Hodzic, A.; Kugeler, K.

1997-01-01

This paper discusses the applicable solutions for a commercial size High Temperature Reactor (HTR) with inherent safety features. It describes the possible realization using an advanced concept which combines newly proposed design characteristics with some well known and proven HTR inherent safety features. The use of the HTR technology offers the conceivably best solution to meet the legal criteria, recently stated in Germany, for the future reactor generation. Both systems, block and pebble bed ,reactor, could be under certain design conditions self regulating in terms of core nuclear heat, mechanical stability and the environmental transfer. 23 refs., 7 figs

Food plant toxicants and safety - Risk assessment and regulation of inherent toxicants in plant foods

DEFF Research Database (Denmark)

Essers, A.J.A.; Alink, G.M.; Speijers, G.J.A.

1998-01-01

The ADI as a tool for risk management and regulation of food additives and pesticide residues is not readily applicable to inherent food plant toxicants: The margin between actual intake and potentially toxic levels is often small; application of the default uncertainty factors used to derive ADI...... values, particularly when extrapolating from animal data, would prohibit the utilisation of the food, which may have an overall beneficial health effect. Levels of inherent toxicants are difficult to control; their complete removal is not always wanted, due to their function for the plant or for human...... health. The health impact of the inherent toxicant is often modified by factors in the food, e.g. the bioavailability from the matrix and interaction with other inherent constituents. Risk-benefit analysis should be made for different consumption scenarios, without the use of uncertainty factors. Crucial...

Impacts of reactivity feedback uncertainties on inherent shutdown in innovative designs

International Nuclear Information System (INIS)

Mueller, C.J.

1986-01-01

The concept of inherent shutdown is emphasized in the approach to the design of innovative, small pool-type liquid-metal reactors (LMRs). This paper reports an evaluation of reactivity feedback uncertainties used in the analyses of anticipated transients without scram for innovative LMRs, and the associated impacts on safety margins and inherent shutdown success probabilities on unprotected loss-of-flow (LOF) events. It then assesses the ultimate importance of these uncertainties on LOF and transient overpower events in evolving metal and oxide innovative designs

Impacts of reactivity feedback uncertainties on inherent shutdown in innovative designs

International Nuclear Information System (INIS)

Mueller, C.J.

1986-01-01

The concept of ''inherent shutdown'' is emphasized in the approach to the design of innovative, small pool-type liquid metal reactors (LMRs). This paper reports an evaluation of reactivity feedback uncertainties used in the analyses of anticipated transients without scram (ATWS) for innovative LMRs, and the associated impacts on safety margins and inherent shutdown success probabilities on unprotected loss-of-flow (LOF) events. It then assesses the ultimate importance of these uncertainties on LOF and transient overpower (TOP) events in evolving metal and oxide innovative designs

Inherent Conservatism in Deterministic Quasi-Static Structural Analysis

Science.gov (United States)

Verderaime, V.

1997-01-01

The cause of the long-suspected excessive conservatism in the prevailing structural deterministic safety factor has been identified as an inherent violation of the error propagation laws when reducing statistical data to deterministic values and then combining them algebraically through successive structural computational processes. These errors are restricted to the applied stress computations, and because mean and variations of the tolerance limit format are added, the errors are positive, serially cumulative, and excessively conservative. Reliability methods circumvent these errors and provide more efficient and uniform safe structures. The document is a tutorial on the deficiencies and nature of the current safety factor and of its improvement and transition to absolute reliability.

HTGR safety philosophy

Energy Technology Data Exchange (ETDEWEB)

Joksimovic, V.; Fisher, C. R. [General Atomic Co., San Diego, CA (USA)

1981-01-15

The accident at the Three Mile Island has focused public attention on reactor safety. Many public figures advocate a safer method of generating nuclear electricity for the second nuclear era in the U.S. The paper discusses the safety philosophy of a concept deemed suitable for this second nuclear era. The HTGR, in the course of its evolution, included safety as a significant determinant in design philosophy. This is particularly evident in the design features which provide inherent safety. Inherent features cause releases from a wide spectrum of accident conditions to be low. Engineered features supplement inherent features. The significance of HTGR safety features is quantified and order-of-magnitude type of comparisons are made with alternative ways of generating electricity.

HTGR safety philosophy

International Nuclear Information System (INIS)

Joksimovic, V.; Fisher, C.R.

1981-01-01

The accident at the Three Mile Island has focused public attention on reactor safety. Many public figures advocate a safer method of generating nuclear electricity for the second nuclear era in the U.S. The paper discusses the safety philosophy of a concept deemed suitable for this second nuclear era. The HTGR, in the course of its evolution, included safety as a significant determinant in design philosophy. This is particularly evident in the design features which provide inherent safety. Inherent features cause releases from a wide spectrum of accident conditions to be low. Engineered features supplement inherent features. The significance of HTGR safety features is quantified and order-of-magnitude type of comparisons are made with alternative ways of generating electricity. (author)

HTGR safety philosophy

International Nuclear Information System (INIS)

Joskimovic, V.; Fisher, C.R.

1980-08-01

The accident at the Three Mile Island has focused public attention on reactor safety. Many public figures advocate a safer method of generating nuclear electricity for the second nuclear era in the US. The paper discusses the safety philosophy of a concept deemed suitable for this second nuclear era. The HTGR, in the course of its evolution, included safety as a significant determinant in design philosophy. This is particularly evident in the design features which provide inherent safety. Inherent features cause releases from a wide spectrum of accident conditions to be low. Engineered features supplement inherent features. The significance of HTGR safety features is quantified and order-of-magnitude type of comparisons are made with alternative ways of generating electricity

Progress on PRISM, an inherently safe, economic, and testable advanced liquid metal reactor

International Nuclear Information System (INIS)

Tippets, F.E.; Salerno, L.N.; Boardman, C.E.; Kwant, W.; Murata, R.E.; Snyder, C.R.

1987-01-01

This paper reports progress on the design of PRISM (Power Reactor Inherently Safe Module) under the DOE-sponsored innovative reactor program now in its third year at General Electric. The purpose of this program is to develop a design for an inherently safe, reliable, and marketable liquid metal fast reactor power plant. The PRISM design approach includes the following key elements: Compact sodium-cooled pool-type reactor modules that are sized to enable factory fabrication, economical shipment to inland as well as water-side sites, and economical full-scale prototype testing for design certification; Nuclear safety-related envelope limited to the reactor modules and their service systems; Inherent, passive shutdown heat removal for loss-of-cooling events; Inherent, passive reactivity shutdown for failure-to-scram events

Possibility of a pressurized water reactor concept with highly inherent heat removal following capability

International Nuclear Information System (INIS)

Araya, Fumimasa; Murao, Yoshio

1995-01-01

If the core power inherently follows change in heat removal rate from the primary coolant system within small thermal expansion of the coolant which can be absorbed in a practical size of pressurizer, reactor systems may have more safety and load following capability. In order to know possibility and necessary conditions of a concept on reactor core and primary coolant system of a pressurized water reactor (PWR) with such 'highly inherent heat removal following capability', transient analyses on an ordinary two-loop PWR have been performed for a transient due to 50% change in heat removal with the RETRAN-02 code. The possibility of a PWR concept with the highly inherent heat removal following capability has been demonstrated under the conditions of the absolute value of ratio of the coolant density reactivity coefficient to the Doppler reactivity coefficient more than 10x10 3 kg·cm 3 which is two to three times larger than that at beginning of cycle (BOC) in an ordinary PWR and realized by elimination of the chemical shim, the 12% lower average linear heat generation rate of 17.9 kW/m, and the 1.5 times larger pressurizer volume than those of the ordinary PWR. (author)

Integral Inherently Safe Light Water Reactor (I²S-LWR)

Energy Technology Data Exchange (ETDEWEB)

Petrovic, Bojan [Georgia Inst. of Technology, Atlanta, GA (United States); Memmott, Matthew [Brigham Young Univ., Provo, UT (United States); Boy, Guy [Florida Inst. of Technology, Melbourne, FL (United States); Charit, Indrajit [Univ. of Idaho, Moscow, ID (United States); Manera, Annalisa [Univ. of Michigan, Ann Arbor, MI (United States); Downar, Thomas [Univ. of Michigan, Ann Arbor, MI (United States); Lee, John [Univ. of Michigan, Ann Arbor, MI (United States); Muldrow, Lycurgus [Morehouse College, Atlanta, GA (United States); Upadhyaya, Belle [Univ. of Tennessee, Knoxville, TN (United States); Hines, Wesley [Univ. of Tennessee, Knoxville, TN (United States); Haghighat, Alierza [Virginia Polytechnic Inst. and State Univ. (Virginia Tech), Blacksburg, VA (United States)

2017-10-02

This final report summarizes results of the multi-year effort performed during the period 2/2013- 12/2016 under the DOE NEUP IRP Project “Integral Inherently Safe Light Water Reactors (I²S-LWR)”. The goal of the project was to develop a concept of a 1 GWe PWR with integral configuration and inherent safety features, at the same time accounting for lessons learned from the Fukushima accident, and keeping in mind the economic viability of the new concept. Essentially (see Figure 1-1) the project aimed to implement attractive safety features, typically found only in SMRs, to a larger power (1 GWe) reactor, to address the preference of some utilities in the US power market for unit power level on the order of 1 GWe.

Design strategy for control of inherently safe reactors

International Nuclear Information System (INIS)

Chisholm, G.H.

1984-01-01

Reactor power plant safety is assured through a combination of engineered barriers to radiation release (e.g., reactor containment) in combination with active reactor safety systems to shut the reactor down and remove decay heat. While not specifically identified as safety systems, the control systems responsible for continuous operation of plant subsystems are the first line of defense for mitigating radiation releases and for plant protection. Inherently safe reactors take advantage of passive system features for decay-heat removal and reactor shutdown functions normally ascribed to active reactor safety systems. The advent of these reactors may permit restructuring of the present control system design strategy. This restructuring is based on the fact that authority for protection against unlikely accidents is, as much as practical, placed upon the passive features of the system instead of the traditional placement upon the PPS. Consequently, reactor control may be simplified, allowing the reliability of control systems to be improved and more easily defended

Development status of PIUS/ISER - a inherently safe reactor for the international use

International Nuclear Information System (INIS)

Wakabayashi, Hiroaki

1987-01-01

It is just in early 1980s that LWR-based nuclear power has become a substantial power source. Though the safety level of nuclear power is always claimed to be sufficiently high by the industry, it rests on the idea of defense in depth, the calculation by probabilistic risk assessment (PRA) or probabilistic safety assessment (PSA). The TMI-2 and Chernobyl-4 accidents occurred in the industrially most advanced countries. In this paper, an alternative way to safe nuclear power is sought in so-called inherently safe reactors (ISR) including the LWR type PIUS/ISER. With proper consideration into the design of nuclear reactor plants, those can be made basically safe through the use of passive safe mechanism for their design. In short, an ISR is a nuclear power reactor which has passive and intrinsic core cooling capability and automatic shutdown capability. As the nuclear power reactors which are currently claimed to be inherently safe, there are the process inherent and ultimately safe reactor (PIUS) of ASEA-ATOM Sweden and the inherently safe and economical reactor (ISER) of the University of Tokyo, Japan, of LWR type. The current status of the development, the reliability, and some technical problems of ISER/PIUS and the attitude of various countries toward ISER/PIUS are described. (Kako, I.)

16 CFR 1211.13 - Inherent force activated secondary door sensors.

Science.gov (United States)

2010-01-01

... 16 Commercial Practices 2 2010-01-01 2010-01-01 false Inherent force activated secondary door sensors. 1211.13 Section 1211.13 Commercial Practices CONSUMER PRODUCT SAFETY COMMISSION CONSUMER PRODUCT... across the door so that the axis is perpendicular to the plane of the door. See Figure 6 of this part...

Probabilities of inherent shutdown of unprotected events in innovative liquid metal reactors

International Nuclear Information System (INIS)

Mueller, C.J.; Wade, D.C.

1988-01-01

The uncertainty in predicting the effectiveness of inherent shutdown in innovative liquid metal cooled reactors with metallic fuel results from three broad contributing areas of uncertainty: (1) the inability to exactly predict the frequency of ATWS events with potential to challenge the safety systems and require inherent shutdown; (2) the approximation of representing all such events by a selected set of ''generic scenarios''; and (3) the inability to exactly calculate the core response to the selected generic scenarios. This paper discusses the work being done to address each of these contributing areas, identifies the design and research approaches being used at Argonne National Laboratory to reducing the key contributions to uncertainties in inherent shutdown, and presents results. The conditional probabilities (given ATWS initiation) of achieving temperatures capable of defeating inherent shutdown are shown to range from /approximately/0.1% to negligible for current designs

EHS Open House: Learning Lab and Life Safety | Poster

Science.gov (United States)

Attendees of the Environment, Health, and Safety Program’s (EHS’) Open House had a chance to learn self-defense techniques, as well as visit with vendors demonstrating the latest trends in laboratory safety. “Working with sharps in labs is inherently dangerous, so EHS proactively focused on featuring equipment that would promote safer techniques,” said Siobhan Tierney, program

Preliminary safety analysis of unscrammed events for KLFR

International Nuclear Information System (INIS)

Kim, S.J.; Ha, G.S.

2005-01-01

The report presents the design features of KLFR; Safety Analysis Code; steady-state calculation results and analysis results of unscrammed events. The calculations of the steady-state and unscrammed events have been performed for the conceptual design of KLFR using SSC-K code. UTOP event results in no fuel damage and no centre-line melting. The inherent safety features are demonstrated through the analysis of ULOHS event. Although the analysis of ULOF has much uncertainties in the pump design, the analysis results show the inherent safety characteristics. 6% flow of rated flow of natural circulation is formed in the case of ULOF. In the metallic fuel rod, the cladding temperature is somewhat high due to the low heat transfer coefficient of lead. ULOHS event should be considered in design of RVACS for long-term cooling

Reactor safety research - visible demonstrations and credible computations

Energy Technology Data Exchange (ETDEWEB)

Loewenstein, W B; Divakaruni, S M

1985-11-01

EPRI has been conducting nuclear safety research for a number of years with the primary goal of assuring the safety and reliability of the nuclear plants. The visibility is emphasized by sponsoring or participating in large scale test demonstrations to credibly support the complex computations that are the basis for quantification of safety margins. Recognizing the success of the airline industry in receiving favorable public perception, the authors compare the design and operation practices of the airline industry with those of the nuclear industry practices to identify the elements contributing to public concerns and unfavorable perceptions. In this paper, authors emphasize the importance of proper communications of research results to the public in a manner that non-specialists understand. Further, EPRI supported research and results in the areas of source term, seismic and structural engineering research, analysis using probabilistic risk assessment (PRA), quantification of safety margins, digital technology development and implementation, and plant transient and performance evaluations are discussed in the paper. (orig./HP).

Reactor safety research - visible demonstrations and credible computations

International Nuclear Information System (INIS)

Loewenstein, W.B.; Divakaruni, S.M.

1985-01-01

EPRI has been conducting nuclear safety research for a number of years with the primary goal of assuring the safety and reliability of the nuclear plants. The visibility is emphasized by sponsoring or participating in large scale test demonstrations to credibly support the complex computations that are the basis for quantification of safety margins. Recognizing the success of the airline industry in receiving favorable public perception, the authors compare the design and operation practices of the airline industry with those of the nuclear industry practices to identify the elements contributing to public concerns and unfavorable perceptions. In this paper, authors emphasize the importance of proper communications of research results to the public in a manner that non-specialists understand. Further, EPRI supported research and results in the areas of source term, seismic and structural engineering research, analysis using probabilistic risk assessment (PRA), quantification of safety margins, digital technology development and implementation, and plant transient and performance evaluations are discussed in the paper. (orig./HP)

Spontaneous stabilization of HTGRs without reactor scram and core cooling—Safety demonstration tests using the HTTR: Loss of reactivity control and core cooling

Energy Technology Data Exchange (ETDEWEB)

Takamatsu, Kuniyoshi, E-mail: takamatsu.kuniyoshi@jaea.go.jp; Yan, Xing L.; Nakagawa, Shigeaki; Sakaba, Nariaki; Kunitomi, Kazuhiko

2014-05-01

It is well known that a High-Temperature Gas-cooled Reactor (HTGR) has superior safety characteristics; for example, an HTGR has a self-control system that uses only physical phenomena against various accidents. Moreover, the large heat capacity and low power density of the core result in very slow temperature transients. Therefore, an HTGR serves inherently safety features against loss of core cooling accidents such as the Tokyo Electric Power Co., Inc. (TEPCO)’s Fukushima Daiichi Nuclear Power Station (NPS) disaster. Herein we would like to demonstrate the inherent safety features using the High-Temperature Engineering Test Reactor (HTTR). The HTTR is the first HTGR in Japan with a thermal power of 30 MW and a maximum reactor outlet coolant temperature of 950 °C; it was built at the Oarai Research and Development Center of Japan Atomic Energy Agency (JAEA). In this study, an all-gas-circulator trip test was analyzed as a loss of forced cooling (LOFC) test with an initial reactor power of 9 MW to demonstrate LOFC accidents. The analytical results indicate that reactor power decreases from 9 MW to 0 MW owing to the negative reactivity feedback effect of the core, even if the reactor shutdown system is not activated. The total reactivity decreases for 2–3 h and then gradually increases in proportion to xenon reactivity; therefore, the HTTR achieves recritical after an elapsed time of 6–7 h, which is different from the elapsed time at reactor power peak occurrence. After the reactor power peak occurs, the total reactivity oscillates several times because of the negative reactivity feedback effect and gradually decreases to zero. Moreover, the new conclusions are as follows: the greater the amount of residual heat removed from the reactor core, the larger the stable reactor power after recriticality owing to the heat balance of the reactor system. The minimum reactor power and the reactor power peak occurrence are affected by the neutron source. The greater the

Proposed and existing passive and inherent safety-related structures, systems, and components (building blocks) for advanced light-water reactors

International Nuclear Information System (INIS)

Forsberg, C.W.; Moses, D.L.; Lewis, E.B.; Gibson, R.; Pearson, R.; Reich, W.J.; Murphy, G.A.; Staunton, R.H.; Kohn, W.E.

1989-10-01

A nuclear power plant is composed of many structures, systems, and components (SSCs). Examples include emergency core cooling systems, feedwater systems, and electrical systems. The design of a reactor consists of combining various SSCs (building blocks) into an integrated plant design. A new reactor design is the result of combining old SSCs in new ways or use of new SSCs. This report identifies, describes, and characterizes SSCs with passive and inherent features that can be used to assure safety in light-water reactors. Existing, proposed, and speculative technologies are described. The following approaches were used to identify the technologies: world technical literature searches, world patent searches, and discussions with universities, national laboratories and industrial vendors. 214 refs., 105 figs., 26 tabs

Proposed and existing passive and inherent safety-related structures, systems, and components (building blocks) for advanced light-water reactors

Energy Technology Data Exchange (ETDEWEB)

Forsberg, C.W.; Moses, D.L.; Lewis, E.B.; Gibson, R.; Pearson, R.; Reich, W.J.; Murphy, G.A.; Staunton, R.H.; Kohn, W.E.

1989-10-01

A nuclear power plant is composed of many structures, systems, and components (SSCs). Examples include emergency core cooling systems, feedwater systems, and electrical systems. The design of a reactor consists of combining various SSCs (building blocks) into an integrated plant design. A new reactor design is the result of combining old SSCs in new ways or use of new SSCs. This report identifies, describes, and characterizes SSCs with passive and inherent features that can be used to assure safety in light-water reactors. Existing, proposed, and speculative technologies are described. The following approaches were used to identify the technologies: world technical literature searches, world patent searches, and discussions with universities, national laboratories and industrial vendors. 214 refs., 105 figs., 26 tabs.

Gas-cooled fast reactor safety

International Nuclear Information System (INIS)

Rickard, C.L.; Simon, R.H.; Buttemer, D.R.

1977-01-01

Initial conceptual design work on the GCFR began in the USA in the early 1960s and since the later 1960s has proceeded with considerable international cooperation. A 300 MWe GCFR demonstration plant employing three main cooling loops is currently being developed at General Atomic. A major preapplication licensing review of this demonstration plant was initiated in 1971 leading in 1974 to publication of a Safety Evaluation Report by the USAEC Directorate of Licensing. The preapplication review is continuing by addressing areas of concern identified in this report such that a major part of the work necessary to support the actual licensing of a GCFR demonstration plant has been established. The safety performance of the GCFR demonstration plant is based upon its inherent safety characteristics among which are the single phase and chemically inert coolant which is not activated and has a low reactivity worth, the negative core power and temperature reactivity coefficients and the small and negative steam reactivity worth. Recent studies of larger core designs indicate that as the reactor size increases central fuel, clad and coolant reactivity worths decrease and the Doppler coefficient becomes more negative. These inherent safety characteristics are complemented by safety design features such as enclosing the entire primary coolant system within a prestressed concrete pressure vessel (PCRV), providing two independent and diverse shutdown systems and residual heat removal (RHR) systems, limiting the worth of control rods to less than $1, employing pressure-equalized fuel rods, a core supported rigidly at its upper end and otherwise unrestrained and coolant downflow within the core to enhance debris removal should local melting occur. The structurally redundant PCRV design allows the potential depressurization leak area to be controlled and, since the PCRV is located within a containment building, coolant is present even after a depressurization accident and each RHR

Safety standards for near surface disposal and the safety case and supporting safety assessment for demonstrating compliance with the standards

International Nuclear Information System (INIS)

Metcalf, P.

2003-01-01

The report presents the safety standards for near surface disposal (ICRP guidance and IAEA standards) and the safety case and supporting safety assessment for demonstrating compliance with the standards. Special attention is paid to the recommendations for disposal of long-lived solid radioactive waste. The requirements are based on the principle for the same level of protection of future individuals as for the current generation. Two types of exposure are considered: human intrusion and natural processes and protection measures are discussed. Safety requirements for near surface disposal are discussed including requirements for protection of human health and environment, requirements or safety assessments, waste acceptance and requirements etc

Development of physiotherapy inherent requirement statements - an Australian experience.

Science.gov (United States)

Bialocerkowski, Andrea; Johnson, Amanda; Allan, Trevor; Phillips, Kirrilee

2013-04-16

The United Nations Convention on the Rights of People with Disabilities promotes equal rights of people with a disability in all aspects of their life including their education. In Australia, Disability Discrimination legislation underpins this Convention. It mandates that higher education providers must demonstrate that no discrimination has occurred and all reasonable accommodations have been considered and implemented, to facilitate access and inclusion for a student with a disability. The first step to meeting legislative requirements is to provide students with information on the inherent requirements of a course. This paper describes the steps which were taken to develop inherent requirement statements for a 4-year entry-level physiotherapy program at one Australian university. Inherent requirement statements were developed using an existing framework, which was endorsed and mandated by the University. Items which described inherencies were extracted from Australian physiotherapy professional standards and statutory regulatory requirements, and units contained in the physiotherapy program. Data were integrated into the 8 prescribed domains: ethical behaviour, behavioural stability, legal, communication, cognition, sensory abilities, strength and mobility, and sustainable performance. Statements for each domain were developed using a 5-level framework (introductory statement, description of the inherent requirement, justification for inherency, characteristics of reasonable adjustments and exemplars) and reviewed by a University Review Panel. Refinement of statements continued until no further changes were required. Fifteen physiotherapy inherent requirement statements were developed. The eight domains identified in the existing framework, developed for Nursing, were relevant to the study of physiotherapy. The inherent requirement statements developed in this study provide a transparent, defensible position on the current requirements of physiotherapy study at

Development of inherent core technologies for advanced reactor

International Nuclear Information System (INIS)

Kim, Keung Koo; Noh, J.M.; Hwang, D.H.

1999-03-01

Recently, the developed countries made their effort on developing the advanced reactor which will result in significantly enhanced safety and economy. However, they will protect the advanced reactor and its design technology with patent and proprietary right. Therefore, it is very important to develop our own key core concepts and inherent core design technologies which can form a foundation of indigenous technologies for development of the domestic advanced reactor in order to keep the superiority in the nuclear plant building market among the developing countries. In order to provide the basic technology for the core design of advanced reactor, this project is for developing the inherent core design concepts with enhanced safety and economy, and associated methodologies and technologies for core analyses. The feasibility study of constructing domestic critical facilities are performed by surveying the status and utilization of foreign facilities and by investigating the demand for domestic facilities. The research results developed in this project, such as core analysis methodologies for hexagonal core, conceptual core design based on hexagonal fuel assemblies and soluble boron core design and control strategies, will provide a technical foundation in developing core design of domestic advanced reactor. Furthermore, they will strengthen the competitiveness of Korean nuclear technology. We also expect that some of the design concepts developed in this project to improve the reactor safety and economy can be applicable to the design of advanced reactor. This will significantly reduce the public anxiety on the nuclear power plant, and will contribute to the economy of construction and operation for the future domestic reactors. Even though the critical facility will not be constructed right now, the investigation of the status and utilization of foreign critical facility will contribute to the future critical facility construction. (author). 150 refs., 34 tabs., 103

Development of inherent core technologies for advanced reactor

Energy Technology Data Exchange (ETDEWEB)

Kim, Keung Koo; Noh, J.M.; Hwang, D.H. [and others

1999-03-01

Recently, the developed countries made their effort on developing the advanced reactor which will result in significantly enhanced safety and economy. However, they will protect the advanced reactor and its design technology with patent and proprietary right. Therefore, it is very important to develop our own key core concepts and inherent core design technologies which can form a foundation of indigenous technologies for development of the domestic advanced reactor in order to keep the superiority in the nuclear plant building market among the developing countries. In order to provide the basic technology for the core design of advanced reactor, this project is for developing the inherent core design concepts with enhanced safety and economy, and associated methodologies and technologies for core analyses. The feasibility study of constructing domestic critical facilities are performed by surveying the status and utilization of foreign facilities and by investigating the demand for domestic facilities. The research results developed in this project, such as core analysis methodologies for hexagonal core, conceptual core design based on hexagonal fuel assemblies and soluble boron core design and control strategies, will provide a technical foundation in developing core design of domestic advanced reactor. Furthermore, they will strengthen the competitiveness of Korean nuclear technology. We also expect that some of the design concepts developed in this project to improve the reactor safety and economy can be applicable to the design of advanced reactor. This will significantly reduce the public anxiety on the nuclear power plant, and will contribute to the economy of construction and operation for the future domestic reactors. Even though the critical facility will not be constructed right now, the investigation of the status and utilization of foreign critical facility will contribute to the future critical facility construction. (author). 150 refs., 34 tabs., 103

Structural safety assessment of a tokamak-type fusion facility for a through crack to cause cooling water leakage and plasma disruption

International Nuclear Information System (INIS)

Nakahira, Masataka

2004-01-01

A tokamak-type fusion machine has inherent safety associated with plasma shutdown. A small water leak can cause a plasma disruption although there is another possibility to terminate plasma without disruption. This plasma disruption will induce electromagnetic (EM) forces acting in the vacuum vessel (VV). From a radiological safety viewpoint, the VV is designed to form a physical barrier that encloses tritium and activated dust. If the VV can sustain an unstable fracture by EM forces from a through crack to cause the small leak, the structural safety will be assured and the inherent safety will be demonstrated. Therefore, a systematic approach to assure the structural safety is developed. A new analytical model to evaluate the through crack and leak rate of cooling water is proposed, with verification by experimental leak measurements. Based on the analysis, the critical crack length to terminate plasma is evaluated as about 2mm. On the other hand, the critical crack length for unstable fracture is obtained as about 400 mm. It is concluded that EM forces induced by the small leak to terminate plasma will not cause unstable fracture of the VV; thus the inherent safety is demonstrated. (author)

Integrating scientific results for a post-closure safety demonstration

International Nuclear Information System (INIS)

Taylor, E.C.; Ramspott, L.D.; Sinnock, S.; Sprecher, W.M.

1994-01-01

The U.S. Department of Energy (DOE) is developing a nuclear waste management system that will accept high-level radioactive waste, transport it, store it, and ultimately emplace it in a deep geologic repository. The key activity now is determining whether Yucca Mountain, Nevada is suitable as a site for the repository. If so, the crucial technological advance will be the demonstration that disposal of nuclear waste will be safe for thousands of years after closure. Recent regulatory, legal, and scientific developments imply that the safety demonstration must be simple. The scientific developments taken together support a simple set of hypotheses that constitute a post-closure safety argument for a repository at Yucca Mountain. If the understanding of Yucca Mountain hydrology presented in the Site Characterization Plan proves correct, then these hypotheses might be confirmed by combining results of Surface-Based Testing with early testing results in the Exploratory Studies Facility

Probabilities of inherent shutdown of unprotected events in innovative liquid metal reactors

International Nuclear Information System (INIS)

Mueller, C.J.

1987-01-01

The uncertainty in predicting the effectiveness of inherent shutdown (ISD) in innovative designs results from three broad contributing areas of uncertainty: (1) the inability to exactly predict the frequency of ATWS events with potential to challenge the safety systems and require ISD; (2) the approximation of representing all such ATWS events by a selected set of ''generic scenarios''; and (3) the inability to exactly calculate the core response to the selected generic scenarios. In this summary, the methodology and associated results of work used to establish probabilities of failure of inherent shutdown of innovative LMRs to the unprotected loss-of-flow (LOF) accident are discussed

Technical safety appraisal of the West Valley Demonstration Project

International Nuclear Information System (INIS)

1989-09-01

This report presents the results of one in a series of Technical Safety Appraisals (TSAs) being conducted of DOE nuclear operations by the Assistant Secretary for Environment, Safety, and Health Office of Safety Appraisals TSAs are one of the ititiatives announced by the Secretary of Energy on September 18, 1985, to enhance the DOE environment, safety and health program. This report presents the results of a TSA of the West Valley Demonstration Project (WVDP). The appraisal was conducted by a team of exerts assembled by the DOE Office of Safety Appraisal and was conducted during onsite visits of June 26-30 and July 10-21, 1989. West Valley, about 30 miles south of Buffalo, New York is the location of the only commercial nuclear fuel reprocessing facility operated in the United States. Nuclear Fuels Services, Inc. (NFS) operated the plant from 1966 to 1972 and processed about 640 metric tons of spent reactor fuel. The reprocessing operation generated about 560,000 gallons of high-level radioactive waste, which was transferred into underground tanks for storage. In 1972 NFS closed the plant and subsequently decided not to reopen it

MHTGR demonstration role in the NRC design certification process

International Nuclear Information System (INIS)

Kelley, A.P. Jr.; Jones, G.

1986-01-01

A modular high-temperature gas-cooled reactor (MHTGR) design is being developed by the US HTGR Program. Because of the small size of the individual modules that would make up a commercial facility, it appears feasible to design and construct a single-module demonstration plant within the funding constraints on the public and private-sector program participants. Furthermore, the safety margins that can be made inherent to the design permit full-scale testing that could supply a new basis for demonstrating investment protection and safety adequacy to the public, the US Nuclear Regulatory Commission (NRC), and potential users. With this in mind, a Project Definition Study was sponsored by Gas-Cooled Reactor Associates and the Tennessee Valley Authority to study the potential benefits of undertaking such a demonstration project. One of the areas investigated was the potential benefits of such a facility in supporting the NRC design certification process, which is envisioned as a necessary commercialization step for the MHTGR

Japan's ISER plant aims at inherent safety plus economy

International Nuclear Information System (INIS)

Kejv, L.

1987-01-01

Japan's ISER reactor concept, combining features of the sweden safety PIUS reactor with possibility of sufficient cost decrease, is briefly described. Comparative characteristics of two reactor concepts are presented

Demonstrating a correlation between the maturity of road safety practices and road safety incidents.

Science.gov (United States)

Amador, Luis; Willis, Christopher Joseph

2014-01-01

The objective of this study is to demonstrate a correlation between the maturity of a country's road safety practices and road safety incidents. Firstly, data on a number of road injuries and fatalities for 129 countries were extracted from the United Nations Global Status on Road Safety database. These data were subdivided according to road safety incident and accident causation factors and normalized based on vehicular fleet (per 1000 vehicles) and road network (per meter of paved road). Secondly, a road safety maturity model was developed based on an adaptation of the concept of process maturity modeling. The maturity of countries with respect to 10 road safety practices was determined through the identification of indicators recorded in the United Nations Global Status of Road Safety Database. Plots of normalized road safety performance of the 129 countries against their maturity scores for each road safety practice as well as an aggregation of the road safety practices were developed. An analysis of variance was done to determine the extent of the correlation between the road safety maturity of the countries and their performance. In addition, a full Bayesian analysis was done to confirm the correlation of each of the road safety practices with injuries and fatalities. Regression analysis for fatalities, injuries, and combined accidents identified maturity with respect to road safety practices associated with speed limits and use of alternative modes as being the most significant predictors of traffic fatalities. A full Bayesian regression confirms that there is a correlation between the maturity of road safety practices and road safety incidents. Road safety practices associated with enforcement of speed limits and promotion of alternative modes are the most significant road safety practices toward which mature countries have concentrated their efforts, resulting in a lower frequency of fatalities, injury rates, and property damage accidents. The authors

Occupational Safety and Health Program at the West Valley Demonstration Project

International Nuclear Information System (INIS)

L. M. Calderon

1999-01-01

The West Valley Nuclear Services Co. LLC (WVNS) is committed to provide a safe, clean, working environment for employees, and to implement U.S. Department of Energy (DOE) requirements affecting worker safety. The West Valley Demonstration Project (WVDP) Occupational Safety and Health Program is designed to protect the safety, health, and well-being of WVDP employees by identifying, evaluating, and controlling biological, chemical, and physical hazards in the work place. Hazards are controlled within the requirements set forth in the reference section at the end of this report. It is the intent of the WVDP Occupational Safety and Health Program to assure that each employee is provided with a safe and healthy work environment. This report shows the logical path toward ensuring employee safety in planning work at the WVDP. In general, planning work to be performed safely includes: combining requirements from specific programs such as occupational safety, industrial hygiene, radiological control, nuclear safety, fire safety, environmental protection, etc.; including WVDP employees in the safety decision-making processes; pre-planning using safety support re-sources; and integrating the safety processes into the work instructions. Safety management principles help to define the path forward for the WVDP Occupational Safety and Health Program. Roles, responsibilities, and authority of personnel stem from these ideals. WVNS and its subcontractors are guided by the following fundamental safety management principles: ''Protection of the environment, workers, and the public is the highest priority. The safety and well-being of our employees, the public, and the environment must never be compromised in the aggressive pursuit of results and accomplishment of work product. A graded approach to environment, safety, and health in design, construction, operation, maintenance, and deactivation is incorporated to ensure the protection of the workers, the public, and the environment

Preliminary risk assessment of the Integral Inherently-Safe Light Water Reactor

International Nuclear Information System (INIS)

McCarroll, Kellen R.; Lee, John C.; Manera, Annalisa; Memmott, Matthew J.; Ferroni, Paolo

2017-01-01

The Integral, Inherently Safe Light Water Reactor (I 2 S-LWR) concept seeks to significantly increase nuclear power plant safety. The project implements a safety-by-design philosophy, eliminating several initiating events and providing novel, passive safety systems at the conceptual phase. Pursuit of unparalleled safety employs an integrated development process linking design with deterministic and probabilistic safety analyses. Unique aspects of the I 2 S-LWR concept and design process present challenges to the probabilistic risk assessment (PRA), particularly regarding overall flexibility, auditability and resolution of results. Useful approaches to initiating events and conditional failures are presented. To exemplify the risk-informed design process using PRA, a trade-off study of two safety system configurations is presented. Although further optimization is required, preliminary results indicate that the I 2 S-LWR can achieve a core damage frequency (CDF) from internal events less than 1.01 × 10 −8 /ry, including reactor vessel ruptures. Containment bypass frequency due to primary heat exchanger rupture is found to be comparable to non-vessel rupture CDF.

Status of the safety concept and safety demonstration for an HLW repository in salt. Summary report

Energy Technology Data Exchange (ETDEWEB)

Bollingerfehr, W.; Buhmann, D.; Filbert, W.; and others

2013-12-15

safety demonstration are the integrity proofs for the geological and geotechnical barriers and analysis of backfill compaction. In addition, any possible radionuclide release from the repository to the environment has also to be assessed. The safety and demonstration concept developed in the course of the ISIBEL project was further evolved and applied in the course of the R and D project ''Vorlaeufige Sicherheitsanalyse Gorleben - VSG'' (preliminary safety analysis Gorleben) as an example for an HLW repository in a domal salt structure. The repository concepts also consider the requirement for retrievability of stored waste during the operational phase of the repository. The results of the R and D project VSG provide evidence that a safe HLW repository within a salt dome of a suitable geologic structure is feasible. The long-term safety can be ensured using state-of-the-art science and technology. In 2010, the Federal Ministry for the Environment, Nature Conservation and Nuclear Safety (BMU) issued new safety requirements for the disposal of heat-generating radioactive waste. These requirements have been included in the analysis. This study shows the depth of the geological and technical knowledge on final disposal of HLW in a salt dome with a suitable geologic structure and demonstrates that the tools required for safety evaluations are available and allow reliable safety assessments of HLW repositories in salt formations.

Operational-safety advantages of LMFBR's: the EBR-II experience and testing program

International Nuclear Information System (INIS)

Sackett, J.I.; Lindsay, R.W.; Golden, G.H.

1982-01-01

LMFBR's contain many inherent characteristics that simplify control and improve operating safety and reliability. The EBR-II design is such that good advantage was taken of these characteristics, resulting in a vary favorable operating history and allowing for a program of off-normal testing to further demonstrate the safe response of LMFBR's to upsets. The experience already gained, and that expected from the future testing program, will contribute to further development of design and safety criteria for LMFBR's. Inherently safe characteristics are emphasized and include natural convective flow for decay heat removal, minimal need for emergency power and a large negative reactivity feedback coefficient. These characteristics at EBR-II allow for ready application of computer diagnosis and control to demonstrate their effectiveness in response to simulated plant accidents. This latter testing objective is an important part in improvements in the man-machine interface

Safety design

International Nuclear Information System (INIS)

Kunitomi, Kazuhiko; Shiozawa, Shusaku

2004-01-01

JAERI established the safety design philosophy of the HTTR based on that of current reactors such as LWR in Japan, considering inherent safety features of the HTTR. The strategy of defense in depth was implemented so that the safety engineering functions such as control of reactivity, removal of residual heat and confinement of fission products shall be well performed to ensure safety. However, unlike the LWR, the inherent design features of the high-temperature gas-cooled reactor (HTGR) enables the HTTR meet stringent regulatory criteria without much dependence on active safety systems. On the other hand, the safety in an accident typical to the HTGR such as the depressurization accident initiated by a primary pipe rupture shall be ensured. The safety design philosophy of the HTTR considers these unique features appropriately and is expected to be the basis for future Japanese HTGRs. This paper describes the safety design philosophy and safety evaluation procedure of the HTTR especially focusing on unique considerations to the HTTR. Also, experiences obtained from an HTTR safety review and R and D needs for establishing the safety philosophy for the future HTGRs are reported

Regulatory and extra-regulatory testing to demonstrate radioactive material packaging safety

International Nuclear Information System (INIS)

Ammerman, D.J.

1997-01-01

Packages for the transportation of radioactive material must meet performance criteria to assure safety and environmental protection. The stringency of the performance criteria is based on the degree of hazard of the material being transported. Type B packages are used for transporting large quantities of radioisotopes (in terms of A 2 quantities). These packages have the most stringent performance criteria. Material with less than an A 2 quantity are transported in Type A packages. These packages have less stringent performance criteria. Transportation of LSA and SCO materials must be in open-quotes strong-tightclose quotes packages. The performance requirements for the latter packages are even less stringent. All of these package types provide a high level of safety for the material being transported. In this paper, regulatory tests that are used to demonstrate this safety will be described. The responses of various packages to these tests will be shown. In addition, the response of packages to extra-regulatory tests will be discussed. The results of these tests will be used to demonstrate the high level of safety provided to workers, the public, and the environment by packages used for the transportation of radioactive material

A compact, inherently safe liquid metal reactor plant concept for terrestrial defense power applications

International Nuclear Information System (INIS)

Magee, P.M.; Dubberley, A.E.; Lutz, D.E.; Palmer, R.S.

1987-01-01

A compact, inherently safe, liquid metal reactor concept based on the GE PRISM innovative LMR design has been developed for terrestrial defense power applications in the 2-50 MWe range. The concept uses a small, sodium-cooled, U-5%Zr metal fueled reactor contained within two redundant steel vessels. The core is designed to operate at a low power density and temperature (925 F) and can operate 30 years without refueling. One two primary coolant loops, depending upon the plant size, transport heat from the core to sodium-to-air, double-wall heat exchangers. Power is produced by a gas turbine operated in a closed ''bottoming'' cycle that employs intercoolers between the compressor stages and a recuperator. Inherent safety is provided by passive means only; operator action is not required to ensure plant safety even for events normally considered Beyond Design Basis Accidents. In addition to normal shutdown heat removal via the sodium-to-air heat exchangers, the design utilizes an inherently passive radiant vessel auxiliary cooling system similar to that designed for PRISM. The use of an air cycle gas turbine eliminates the cost and complexity of the sodium-water reactor pressure relief system required for a steam cycle sodium-cooled reactor

Safety Design Approach for the Development of Safety Requirements for Design of Commercial HTGR

International Nuclear Information System (INIS)

Ohashi, Hirofumi; Sato, Hiroyuki; Nakagawa, Shigeaki; Tachibana, Yukio; Nishihara, Tetsuo; Yan, Xing; Sakaba, Nariaki; Kunitomi, Kazuhiko

2014-01-01

The research committee on “Safety requirements for HTGR design” was established in 2013 under the Atomic Energy Society of Japan to develop the draft safety requirements for the design of commercial High Temperature Gas-cooled Reactors (HTGRs), which incorporate the HTGR safety features demonstrated using the High Temperature Engineering Test Reactor (HTTR), lessons learned from the accident of Fukushima Daiichi Nuclear Power Station and requirements for the integration of the hydrogen production plants. The safety design approach for the commercial HTGRs which is a basement of the safety requirements is determined prior to the development of the safety requirements. The safety design approaches for the commercial HTGRs are to confine the radioactive materials within the coated fuel particles not only during normal operation but also during accident conditions, and the integrity of the coated fuel particles and other requiring physical barriers are protected by the inherent and passive safety features. This paper describes the main topics of the research committee, the safety design approaches and the safety functions of the commercial HTGRs determined in the research committee. (author)

The inherent limits of predicting school violence.

Science.gov (United States)

Mulvey, E P; Cauffman, E

2001-10-01

The recent media hype over school shootings has led to demands for methods of identifying school shooters before they act. Despite the fact that schools remain one of the safest places for youths to be, schools are beginning to adopt identification systems to determine which students could be future killers. The methods used to accomplish this not only are unproven but are inherently limited in usefulness and often do more harm than good for both the children and the school setting. The authors' goals in the present article are to place school shootings in perspective relative to other risks of violence that children face and to provide a reasonable and scientifically defensible approach to improving the safety of schools.

Novel OSNR Monitoring Technique in Dense WDM Systems using Inherently Generated CW Monitoring Channels

DEFF Research Database (Denmark)

Petersen, Martin Nordal

2007-01-01

We present a simple, yet effective OSNR monitoring technique based on an inherent effect in the optical modulator. Highly accurate OSNR monitoring is demonstrated in a 40 Gb/s dense WDM system with 50 GHz channel spacing.......We present a simple, yet effective OSNR monitoring technique based on an inherent effect in the optical modulator. Highly accurate OSNR monitoring is demonstrated in a 40 Gb/s dense WDM system with 50 GHz channel spacing....

Radiation safety at the West Valley Demonstration Project

International Nuclear Information System (INIS)

Hoffman, R.L.

1997-01-01

This is a report on the Radiation Safety Program at the West Valley Demonstration Project (WVDP). This Program covers a number of activities that support high-level waste solidification, stabilization of facilities, and decontamination and decommissioning activities at the Project. The conduct of the Program provides confidence that all occupational radiation exposures received during operational tasks at the Project are within limits, standards, and program requirements, and are as low as reasonably achievable

Industrial Fuel Gas Demonstration Plant Program. Task III, Demonstration plant safety, industrial hygiene, and major disaster plan (Deliverable No. 35)

Energy Technology Data Exchange (ETDEWEB)

None

1980-03-01

This Health and Safety Plan has been adopted by the IFG Demonstration Plant managed by Memphis Light, Gas and Water at Memphis, Tennessee. The plan encompasses the following areas of concern: Safety Plan Administration, Industrial Health, Industrial Safety, First Aid, Fire Protection (including fire prevention and control), and Control of Safety Related Losses. The primary objective of this plan is to achieve adequate control of all potentially hazardous activities to assure the health and safety of all employees and eliminate lost work time to both the employees and the company. The second objective is to achieve compliance with all Federal, state and local laws, regulations and codes. Some thirty specific safe practice instruction items are included.

Development of physiotherapy inherent requirement statements – an Australian experience

Science.gov (United States)

2013-01-01

Background The United Nations Convention on the Rights of People with Disabilities promotes equal rights of people with a disability in all aspects of their life including their education. In Australia, Disability Discrimination legislation underpins this Convention. It mandates that higher education providers must demonstrate that no discrimination has occurred and all reasonable accommodations have been considered and implemented, to facilitate access and inclusion for a student with a disability. The first step to meeting legislative requirements is to provide students with information on the inherent requirements of a course. This paper describes the steps which were taken to develop inherent requirement statements for a 4-year entry-level physiotherapy program at one Australian university. Case presentation Inherent requirement statements were developed using an existing framework, which was endorsed and mandated by the University. Items which described inherencies were extracted from Australian physiotherapy professional standards and statutory regulatory requirements, and units contained in the physiotherapy program. Data were integrated into the 8 prescribed domains: ethical behaviour, behavioural stability, legal, communication, cognition, sensory abilities, strength and mobility, and sustainable performance. Statements for each domain were developed using a 5-level framework (introductory statement, description of the inherent requirement, justification for inherency, characteristics of reasonable adjustments and exemplars) and reviewed by a University Review Panel. Refinement of statements continued until no further changes were required. Fifteen physiotherapy inherent requirement statements were developed. The eight domains identified in the existing framework, developed for Nursing, were relevant to the study of physiotherapy. Conclusions The inherent requirement statements developed in this study provide a transparent, defensible position on the

Perception of parents as demonstrating the inherent merit of their values: relations with self-congruence and subjective well-being.

Science.gov (United States)

Yu, Shi; Assor, Avi; Liu, Xiangping

2015-02-01

This study focuses on the parenting practice of inherent value demonstration (IVD), involving parents' tendency to express their values in behaviours and appear satisfied and vital while doing so. Data from Chinese college students (n = 89) confirmed the hypothesis that offspring's perception of their parents as engaged in IVD predicts offspring's subjective well-being (SWB) through sense of self-congruence. Importantly, these relations emerged also when controlling for fundamental autonomy-supportive (FAS) parenting practices such as taking children's perspective, minimising control and allowing choice. These findings are consistent with the view that parents concerned with their children's sense of autonomy may do well to engage in IVD in addition to more fundamental autonomy-supportive practices. Future research may examine the role of IVD in promoting authentic values that serve as an internal compass that guides children to act in ways that feel self-congruent. © 2014 International Union of Psychological Science.

Demonstration of safety of decommissioning of facilities using radioactive material

International Nuclear Information System (INIS)

Batandjieva, Borislava; O'Donnell, Patricio

2008-01-01

Full text:The development of nuclear industry worldwide in the recent years has particular impact on the approach of operators, regulators and interested parties to the implementation of the final phases (decommissioning) of all facilities that use radioactive material (from nuclear power plants, fuel fabrication facilities, research reactors to small research or medical laboratories). Decommissioning is becoming an increasingly important activity for two main reasons - termination of the practice in a safe manner with the view to use the facility or the site for other purposes, or termination of the practice and reuse the facility or site for new built nuclear facilities. The latter is of special relevance to multi-facility sites where for example new nuclear power plants and envisaged. However, limited countries have the adequate legal and regulatory framework, and experience necessary for decommissioning. In order to respond to this challenge of the nuclear industry and assist Member States in the adequate planning, conduct and termination of decommissioning of wide range of facilities, over the last decade the IAEA has implemented and initiated several projects in this field. One of the main focuses of this assistance to operators, regulators and specialists involved in decommissioning is the evaluation and demonstration of safety of decommissioning. This importance of these Agency activities was also highlighted in the International Action Plan on Decommissioning, during the second Joint Convention meeting in 2006 and the International Conference on Lessons Learned from Decommissioning in Athens in 2006. The IAEA has been providing technical support to its Member States in this field through several mechanisms: (1) the establishment of a framework of safety standards on decommissioning and development of a supporting technical documents; (2) the establishment of an international peer review mechanism for decommissioning; (3) the technical cooperation projects

Are classical process safety concepts relevant to nanotechnology applications?

International Nuclear Information System (INIS)

Amyotte, Paul R

2011-01-01

The answer to the question posed by the title of this paper is yes - with adaptation to the specific hazards and challenges found in the field of nanotechnology. The validity of this affirmative response is demonstrated by relating key process safety concepts to various aspects of the nanotechnology industry in which these concepts are either already practised or could be further applied. This is accomplished by drawing on the current author's experience in process safety practice and education as well as a review of the relevant literature on the safety of nanomaterials and their production. The process safety concepts selected for analysis include: (i) risk management, (ii) inherently safer design, (iii) human error and human factors, (iv) safety management systems, and (v) safety culture.

Safety assessment for the 24 CANFLEX-NU bundle demonstration irradiation at Wolsong-1 generation

Energy Technology Data Exchange (ETDEWEB)

Suk, Ho Chun; Cho, M. S.; Jun, J. S. and others

2001-06-01

This document is a report on the safety assessment for the 24 CANFLEX-NU(CANDU Flexible fuelling - Natural Uranium) fuel bundle demonstration irradiation at Wolsong-1 Generating Station. The CANFLEX fuel bundle as a CANDU advanced fuel has been jointly developed by KAERI/AECL. This document describes the rationale for the demonstration irradiation and comments on the Korean government licensing issues such as the status of the CANFLEX fuel irradiations at NRU research reactor in AECL, status and plan of the CANFLEX fuel irradiations at a CANDU-6 power reactor, status of the water CHF(Critical Heat Flux) test at Stern Laboratories and the CHF correlation. This documents presents an assessment the consequences of postulated accidents with all safety system available during demonstration irradiation of 24 CANFLEX-NU fuel bundles at Wolsong-1 Generating Station. The assessment is made by two kinds of approaches. One approach is based on the document of the safety assessment for the 24 CANFLEX-NU fuel bundle demonstration irradiation at Point Lepreau Generating Station. The other approach is taken from the safety analyses using the analysis methods and assumptions used in the final safety reports on the 600 MWe CANDU-PHWR Wolsung-2, 3, and 4 Nuclear Power Plants for the Korea Electric Power Cooperation. The analyses are not comprehensive reviews of the postulated accidents, but examination of the expected difference in accident consequences because of the presence of 24 CANFLEX fuel bundles in two channels. The approach is to compare the difference to the safety margin for 37-element bundle cases.

Simulator platform for fast reactor operation and safety technology demonstration

International Nuclear Information System (INIS)

Vilim, R.B.; Park, Y.S.; Grandy, C.; Belch, H.; Dworzanski, P.; Misterka, J.

2012-01-01

A simulator platform for visualization and demonstration of innovative concepts in fast reactor technology is described. The objective is to make more accessible the workings of fast reactor technology innovations and to do so in a human factors environment that uses state-of-the art visualization technologies. In this work the computer codes in use at Argonne National Laboratory (ANL) for the design of fast reactor systems are being integrated to run on this platform. This includes linking reactor systems codes with mechanical structures codes and using advanced graphics to depict the thermo-hydraulic-structure interactions that give rise to an inherently safe response to upsets. It also includes visualization of mechanical systems operation including advanced concepts that make use of robotics for operations, in-service inspection, and maintenance.

Simulator platform for fast reactor operation and safety technology demonstration

Energy Technology Data Exchange (ETDEWEB)

Vilim, R. B.; Park, Y. S.; Grandy, C.; Belch, H.; Dworzanski, P.; Misterka, J. (Nuclear Engineering Division)

2012-07-30

A simulator platform for visualization and demonstration of innovative concepts in fast reactor technology is described. The objective is to make more accessible the workings of fast reactor technology innovations and to do so in a human factors environment that uses state-of-the art visualization technologies. In this work the computer codes in use at Argonne National Laboratory (ANL) for the design of fast reactor systems are being integrated to run on this platform. This includes linking reactor systems codes with mechanical structures codes and using advanced graphics to depict the thermo-hydraulic-structure interactions that give rise to an inherently safe response to upsets. It also includes visualization of mechanical systems operation including advanced concepts that make use of robotics for operations, in-service inspection, and maintenance.

A new assessment method for demonstrating the sufficiency of the safety assessment and the safety margins of the geological disposal system

International Nuclear Information System (INIS)

Ohi, Takao; Kawasaki, Daisuke; Chiba, Tamotsu; Takase, Toshio; Hane, Koji

2013-01-01

A new method for demonstrating the sufficiency of the safety assessment and safety margins of the geological disposal system has been developed. The method is based on an existing comprehensive sensitivity analysis method and can systematically identify the successful conditions, under which the dose rate does not exceed specified safety criteria, using analytical solutions for nuclide migration and the results of a statistical analysis. The successful conditions were identified using three major variables. Furthermore, the successful conditions at the level of factors or parameters were obtained using relational equations between the variables and the factors or parameters making up these variables. In this study, the method was applied to the safety assessment of the geological disposal of transuranic waste in Japan. Based on the system response characteristics obtained from analytical solutions and on the successful conditions, the classification of the analytical conditions, the sufficiency of the safety assessment and the safety margins of the disposal system were then demonstrated. A new assessment procedure incorporating this method into the existing safety assessment approach is proposed in this study. Using this procedure, it is possible to conduct a series of safety assessment activities in a logical manner. (author)

Experience on the demonstration of safety for older reactors

International Nuclear Information System (INIS)

Facer, R.

2001-01-01

The UK's oldest reactors are still operating. Built during the 1950's and commissioned between 1956 and 1960, eight reactors continue to provide electricity and process steam. It is still economically justified to keep them running. In addition to the economic considerations it is also necessary to justify that they can still continue to operate safely. This paper provides a brief review of how the Operator of these stations has justified the safety of operation to date and how they expect to continue to justify their operation for several more years. It is appropriate to consider why the Operator wishes to keep the plant operating. Among the most important reasons are that: The plant is built and paid for, Running costs are relatively low process steam is available for the adjacent sites It is a commercially viable electricity producer It is a reliable electricity source The operators have developed programmes for safety review of the plant and introduced a Continuing Operation Programme which had two main requirements which were, the demonstration of continuing acceptable safety the ensurance of commercial viability. (author)

Summary of advanced LMR [Liquid Metal Reactor] evaluations: PRISM [Power Reactor Inherently Safe Module] and SAFR [Sodium Advanced Fast Reactor

International Nuclear Information System (INIS)

Van Tuyle, G.J.; Slovik, G.C.; Chan, B.C.; Kennett, R.J.; Cheng, H.S.; Kroeger, P.G.

1989-10-01

In support of the US Nuclear Regulatory Commission (NRC), Brookhaven National Laboratory (BNL) has performed independent analyses of two advanced Liquid Metal Reactor (LMR) concepts. The designs, sponsored by the US Department of Energy (DOE), the Power Reactor Inherently Safe Module (PRISM) [Berglund, 1987] and the Sodium Advanced Fast Reactor (SAFR) [Baumeister, 1987], were developed primarily by General Electric (GE) and Rockwell International (RI), respectively. Technical support was provided to DOE, RI, and GE, by the Argonne National Laboratory (ANL), particularly with respect to the characteristics of the metal fuels. There are several examples in both PRISM and SAFR where inherent or passive systems provide for a safe response to off-normal conditions. This is in contrast to the engineered safety systems utilized on current US Light Water Reactor (LWR) designs. One important design inherency in the LMRs is the ''inherent shutdown'', which refers to the tendency of the reactor to transition to a much lower power level whenever temperatures rise significantly. This type of behavior was demonstrated in a series of unscrammed tests at EBR-II [NED, 1986]. The second key design feature is the passive air cooling of the vessel to remove decay heat. These systems, designated RVACS in PRISM and RACS in SAFR, always operate and are believed to be able to prevent core damage in the event that no other means of heat removal is available. 27 refs., 78 figs., 3 tabs

The Limits of Logic-Based Inherent Safety of Social Robots

DEFF Research Database (Denmark)

Bentzen, Martin Mose

2017-01-01

Social robots can reason and act while taking into accountsocial and cultural structures, for instance by complying withsocial or ethical norms or values. As social robots are likely to becomemore common and advanced and thus likely to interact withhuman beings in increasingly complex situations......-based safety for ethical robots is shown. Afterwards,an empirical study is used to show that there is a clash betweendeontic reasoning and most formal deontic logics. I give anexample as to how this clash can cause problems in human-robot interaction.I conclude that deontic logics closer to natural...... languagereasoning are needed and that logic only should play a limited partin the overall safety architecture of a social robot, which should alsobe based on other principles of safe design....

Compliance demonstration: What can be reasonably expected from safety assessment for geological repositories?

International Nuclear Information System (INIS)

Zuidema, P.; Smith, P.; Sumerling, T.

1999-01-01

When licensing a nuclear facility, it is important to demonstrate that it will comply with regulatory limits (e.g. individual dose limits) and also show that sufficient attention has been paid to optimisation of facility design and operation, such that any associated radiological impacts will be as low as reasonably achievable (ALARA). In general, in demonstrating compliance, experience can be drawn from the performance of existing and similar facilities, and monitoring plans can be specified that will confirm that actual radiological discharges during operations are within authorised limits for the facility. This is also true in respect of the operational period of a geological repository. For the post-closure phase of a repository, however, it is also necessary to show that possible releases will remain acceptably low even at long times in the future when, it is assumed, control of the facility has lapsed and there is no method of either monitoring releases or taking remedial action in the case of unexpected events or releases. In addition, within each country, a deep geological repository will be a first-of-a-kind development so that compliance arguments can be expected to be rigorously tested without any assistance from the precedent of licensing of similar facilities nationally. This puts heavy, and quite unusual, burdens on the long-term safety assessment for a geological repository to develop a case that is sufficiently strong to demonstrate compliance. This paper focuses on the problem of demonstrating compliance with long-term safety requirements for a geological repository, and explores: the overall aims and special difficulties of demonstrating compliance for a geological repository; the role of safety assessment in demonstrating compliance; the scope for optimisation of a geological repository and importance of robustness and lessons learnt from the application of safety assessment. In addition, some issues requiring further discussion and clarification

Passive and engineered safety features of the prototype fast reactor (PFR), Dounreay

International Nuclear Information System (INIS)

Gregory, C.V.

1991-01-01

Prototype fast reactor (PFR) combines passive and engineered safety features. Natural convection, a strong negative power coefficient, the decay heat removal system, and a fuel design able to operate beyond failure are all inherent and passive safety features of the PFR. The reliable shutdown system and the protection provided against SGU leaks are example of engineered protection. Experience at PFR demonstrates the worth and potential of a range of passive and engineered safeguards

Perspectives on Understanding and Verifying the Safety Terrain of Modular High Temperature Gas-Cooled Reactors

International Nuclear Information System (INIS)

Carlson, Donald E.

2014-01-01

The inherent safety characteristics of modular high temperature gas-cooled reactors (HTGRs) are conceptually well known and are largely supported by insights from past and ongoing research. This paper offers perspectives on selected issues in areas where further analysis and testing achievable within existing research and demonstration programs could help address residual uncertainties and better support the analysis of safety performance and the regulatory assessment of defense in depth. Areas considered include the evaluation of normal and anomalous core operating conditions and the analysis of accidents involving coolant depressurization, air ingress, moisture ingress, and reactivity insertion. In addition to discussing associated uncertainties and potential measures to address them, the paper also proposes supplemental “safety terrain” studies that would use realistic assessments of postulated extreme event sequences to establish a more comprehensive understanding of the inherent behaviors and ultimate safety capabilities of modular HTGRs. (author)

Regulatory requirements for demonstration of the achieved safety level at the Mochovce NPP before commissioning

International Nuclear Information System (INIS)

Lipar, M.

1997-01-01

A review of regulatory requirements for demonstration of the achieved safety level at the Mochovce NPP before commissioning is given. It contains licensing steps in Slovakia during commissioning; Status and methodology of Mochovce safety analysis report; Mochovce NPP safety enhancement program; Regulatory body policy towards Mochovce NPP safety enhancement; Recent development in Mochovce pre-operational safety enhancement program review and assessment process; Licensing steps in Slovakia during commissioning

Food plant toxicants and safety: risk assessment and regulation of inherent toxicants in plant foods.

NARCIS (Netherlands)

Essers, A.J.; Alink, G.M.; Speijers, G.J.A.; Alexander, J.; Bouwmeister, P.J.; Brandt, van den P.A.; Ciere, S.; Gry, J.; Herrman, J.; Kuiper, H.A.; Mortby, E.; Renwickn, A.G.

1998-01-01

The ADI as a tool for risk management and regulation of food additives and pesticide residues is not readily applicable to inherent food plant toxicants: The margin between actual intake and potentially toxic levels is often small; application of the default uncertainty factors used to derive ADI

Feasibility study of a dedicate nuclear desalination system: Low-pressure inherent heat sink nuclear desalination plant (LIND)

Energy Technology Data Exchange (ETDEWEB)

Kim, Ho Sik; No, Hee Cheon; Jo, Yu Gwan; Wivisono, Andhika Feri; Park, Byung Ha; Choi, Jin Young; Lee, Jeong Ik; Jeong, Yong Hoon; Cho, Nam Zin [Dept. of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

2015-04-15

In this paper, we suggest the conceptual design of a water-cooled reactor system for a low-pressure inherent heat sink nuclear desalination plant (LIND) that applies the safety-related design concepts of high temperature gas-cooled reactors to a water-cooled reactor for inherent and passive safety features. Through a scoping analysis, we found that the current LIND design satisfied several essential thermal-hydraulic and neutronic design requirements. In a thermal-hydraulic analysis using an analytical method based on the Wooton-Epstein correlation, we checked the possibility of safely removing decay heat through the steel containment even if all the active safety systems failed. In a neutronic analysis using the Monte Carlo N-particle transport code, we estimated a cycle length of approximately 6 years under 200 MW{sub th} and 4.5% enrichment. The very long cycle length and simple safety features minimize the burdens from the operation, maintenance, and spent-fuel management, with a positive impact on the economic feasibility. Finally, because a nuclear reactor should not be directly coupled to a desalination system to prevent the leakage of radioactive material into the desalinated water, three types of intermediate systems were studied: a steam producing system, a hot water system, and an organic Rankine cycle system.

Feasibility study of a dedicated nuclear desalination system: Low-pressure Inherent heat sink Nuclear Desalination plant (LIND

Directory of Open Access Journals (Sweden)

Ho Sik Kim

2015-04-01

Full Text Available In this paper, we suggest the conceptual design of a water-cooled reactor system for a low-pressure inherent heat sink nuclear desalination plant (LIND that applies the safety-related design concepts of high temperature gas-cooled reactors to a water-cooled reactor for inherent and passive safety features. Through a scoping analysis, we found that the current LIND design satisfied several essential thermal–hydraulic and neutronic design requirements. In a thermal–hydraulic analysis using an analytical method based on the Wooton–Epstein correlation, we checked the possibility of safely removing decay heat through the steel containment even if all the active safety systems failed. In a neutronic analysis using the Monte Carlo N-particle transport code, we estimated a cycle length of approximately 6 years under 200 MWth and 4.5% enrichment. The very long cycle length and simple safety features minimize the burdens from the operation, maintenance, and spent-fuel management, with a positive impact on the economic feasibility. Finally, because a nuclear reactor should not be directly coupled to a desalination system to prevent the leakage of radioactive material into the desalinated water, three types of intermediate systems were studied: a steam producing system, a hot water system, and an organic Rankine cycle system.

Safety of intrinsically safe and economical reactor (ISER)

International Nuclear Information System (INIS)

Asahi, Y.; Sugawara, I.; Yamanaka, K.

1988-01-01

Inherent safety of a reactor may be quantified by the grace period at various safety levels such as maintenance of fuel integrity, maintenance of fuel coolability and avoidance of core-melt. It is important to find out the grace period especially at the safety level of maintenance of fuel integrity. It has been conducted to design the ISER, which is characterized by the steel-made reactor pressure vessel. In addition to the passive nature of the safety design of the reactor itself, the ISER is equipped in the secondary system with a subsystem called the passive safety and shutdown system (PSSS), which will help to increase the grace period. It was found by the null transient analysis that check valves are needed at the top hot/cold interface. The analysis of the station blackout, which is one of the severest accident conceivable for the ISER, was made to examine inherent safety of the ISER with and without the PSSS. This paper reports that found out that the PSSS enhances inherent safety of the ISER

Innovation in the Safety of nuclear systems: fundamental aspects

International Nuclear Information System (INIS)

Herranz, L. E.

2009-01-01

Safety commercial nuclear reactors has been an indispensable condition for future enlargement of power generation based on nuclear technology. Its fundamental principle, defence in depth, far from being outdated, is still adopted as a key foundation in the advanced nuclear system (generations III and IV). Nevertheless, the cumulative experience gained in the operation and maintenance of nuclear reactors, the development of methodologies like the probabilistic safety analysis, the use of passive safety systems and, even, the inherent characteristics of some new design (which exclude accident scenarios), allow estimating safety figures of merit even more outstanding that those achieved in the second generation of nuclear reactors. This safety innovation of upcoming nuclear reactors has entailed a huge investigation program (generation III) that will be focused on optimizing and demonstrating the postulated safety of future nuclear systems (Generation IV). (Author)

Safety analysis report for packaging (onsite) transuranic performance demonstration program sample packaging

International Nuclear Information System (INIS)

Mccoy, J.C.

1997-01-01

The Transuranic Performance Demonstration Program (TPDP) sample packaging is used to transport highway route controlled quantities of weapons grade (WG) plutonium samples from the Plutonium Finishing Plant (PFP) to the Waste Receiving and Processing (WRAP) facility and back. The purpose of these shipments is to test the nondestructive assay equipment in the WRAP facility as part of the Nondestructive Waste Assay PDP. The PDP is part of the U. S. Department of Energy (DOE) National TRU Program managed by the U. S. Department of Energy, Carlsbad Area Office, Carlsbad, New Mexico. Details of this program are found in CAO-94-1045, Performance Demonstration Program Plan for Nondestructive Assay for the TRU Waste Characterization Program (CAO 1994); INEL-96/0129, Design of Benign Matrix Drums for the Non-Destructive Assay Performance Demonstration Program for the National TRU Program (INEL 1996a); and INEL-96/0245, Design of Phase 1 Radioactive Working Reference Materials for the Nondestructive Assay Performance Demonstration Program for the National TRU Program (INEL 1996b). Other program documentation is maintained by the national TRU program and each DOE site participating in the program. This safety analysis report for packaging (SARP) provides the analyses and evaluations necessary to demonstrate that the TRU PDP sample packaging meets the onsite transportation safety requirements of WHC-CM-2-14, Hazardous Material Packaging and Shipping, for an onsite Transportation Hazard Indicator (THI) 2 packaging. This SARP, however, does not include evaluation of any operations within the PFP or WRAP facilities, including handling, maintenance, storage, or operating requirements, except as they apply directly to transportation between the gate of PFP and the gate of the WRAP facility. All other activities are subject to the requirements of the facility safety analysis reports (FSAR) of the PFP or WRAP facility and requirements of the PDP

Safety characteristics of small heat producing reactors

International Nuclear Information System (INIS)

Seifritz, W.

1987-10-01

The primary objectives of protection in nuclear power plants are the possibility to shut the reactor down in case of emergency and keep it subcritical in the long run, the existence of a heat sink for post-decay heat removal in order to avoid overheating, let alone core meltdown, and the containment of radioactivity within the barriers designed for this purpose, thus preventing significant activity release. In principle, these objectives can be met in various ways, namely by active, passive or inherent technical safeguards systems. In practice, a mixture of these approaches is employed in almost all cases. What matters in the end is the assessment of the overall concept, not of some outstanding feature. Inherent characteristics are easier to achieve in small reactors. However, also in this case, inherent safety does not mean absolute safety. If inherent safety characteristics were all encompassing, they would have to include self-healing effects. However, inanimate matter is incapable of such self-organization. Consequently, inherent characteristics in nuclear technology by definition should include the increased use of dissipative processes in the thermal part of the plant. (author)

Accommodation of unprotected accidents by inherent safety design features in metallic and oxide-fueled LMFBRs

International Nuclear Information System (INIS)

Su, S.F.; Cahalan, J.E.; Sevy, R.H.

1985-01-01

This paper presents the results of a systematic study of the effectiveness of intrinsic design features to mitigate the consequences of unprotected accidents in metallic and oxide-fueled LMFBRs. The accidents analyzed belong to the class generally considered to lead to core disruption; unprotected loss-of-flow (LOF) and transient over-power (TOP). The results of the study demonstrate the potential for design features to meliorate accident consequences, and in some cases to render them benign. Emphasis is placed on the relative performance of metallic and oxide-fueled core designs, and safety margins are quantified in sensitivity studies. All analyses were carried out using the SASSYS LMFBR systems analysis code (1)

Experimental evidence for inherent Lévy search behaviour in foraging animals.

Science.gov (United States)

Kölzsch, Andrea; Alzate, Adriana; Bartumeus, Frederic; de Jager, Monique; Weerman, Ellen J; Hengeveld, Geerten M; Naguib, Marc; Nolet, Bart A; van de Koppel, Johan

2015-05-22

Recently, Lévy walks have been put forward as a new paradigm for animal search and many cases have been made for its presence in nature. However, it remains debated whether Lévy walks are an inherent behavioural strategy or emerge from the animal reacting to its habitat. Here, we demonstrate signatures of Lévy behaviour in the search movement of mud snails (Hydrobia ulvae) based on a novel, direct assessment of movement properties in an experimental set-up using different food distributions. Our experimental data uncovered clusters of small movement steps alternating with long moves independent of food encounter and landscape complexity. Moreover, size distributions of these clusters followed truncated power laws. These two findings are characteristic signatures of mechanisms underlying inherent Lévy-like movement. Thus, our study provides clear experimental evidence that such multi-scale movement is an inherent behaviour rather than resulting from the animal interacting with its environment. © 2015 The Author(s) Published by the Royal Society. All rights reserved.

40 CFR 88.312-93 - Inherently Low-Emission Vehicle labeling.

Science.gov (United States)

2010-07-01

... stroke width not less than 0.5 inches (1.3 centimeters). In addition, the words “INHERENTLY LOW-EMISSION..., the words “CLEAN AIR VEHICLE” must be present in lettering no smaller than 0.8 inches (2.0 centimeters... loan to someone who has not demonstrated eligibility for expanded TCMs available to ILEVs according to...

Nuclear safety as applied to space power reactor systems

International Nuclear Information System (INIS)

Cummings, G.E.

1987-01-01

To develop a strategy for incorporating and demonstrating safety, it is necessary to enumerate the unique aspects of space power reactor systems from a safety standpoint. These features must be differentiated from terrestrial nuclear power plants so that our experience can be applied properly. Some ideas can then be developed on how safe designs can be achieved so that they are safe and perceived to be safe by the public. These ideas include operating only after achieving a stable orbit, developing an inherently safe design, ''designing'' in safety from the start and managing the system development (design) so that it is perceived safe. These and other ideas are explored further in this paper

Design-related inherent safety characteristics in large LMFBR power plants

International Nuclear Information System (INIS)

Tzanos, C.P.; Barthold, W.P.; Bowers, C.H.; Ferguson, D.R.; Prohammer, F.G.; van Erp, J.B.

1976-01-01

Design-related safety-enhancing features such as (1) extended pump coastdown, (2) increased negative reactivity feedbacks, (3) reduced sodium void reactivity, and (4) self-actuated shutdown systems are evaluated. Primary emphasis is placed on preventing or limiting core damage. Attention is also given to features aimed at mitigation of the energetics potential of hypothetical core-disruptive accidents

Safety demonstration test on solvent fire in fuel reprocessing plant

International Nuclear Information System (INIS)

Nishio, Gunji; Hashimoto, Kazuichiro

1989-03-01

This report summarizes a fundamental of results obtained in the Reprocessing Plant Safety Demonstration Test Program which was performed under the contract between the Science and Technology Agency of Japan and the Japan Atomic Energy Research Institute. In this test program, a solvent fire was hypothesized, and such data were obtained as fire behavior, smoke behavior and integrity of exhaust filters in the ventilation system. Through the test results, it was confirmed that under the fire condition in hypothetical accident, the integrity of the cell and the cell ventilation system were maintained, and the safety function of the exhaust filters was maintained against the smoke loading. Analytical results by EVENT code agreed well with the present test data on the thermofluid flow in a cell ventilation system. (author)

Safety related experience in FFTF startup and operation

International Nuclear Information System (INIS)

Peterson, R.E.; Halverson, T.G.; Daughtry, J.W.

1982-06-01

The Fast Flux Test Facility (FFTF) is a 400 MW(t) sodium cooled fast reactor operating at the Hanford Engineering Development Laboratory, Richland, Washington, to conduct fuels and materials testing in support of the US LMFBR program. Startup and initial power ascension testing of the facility involved a comprehensive series of readiness reviews and acceptance tests, many of which relate to the inherent safety of the plant. Included are physics measurements, natural circulation, integrated containment leakage, shielding effectiveness, fuel failure detection, and plant protection system tests. Described are the measurements taken to confirm the design safety margins upon which the operating authorization of the plant was based. These measurements demonstrate that large margins of safety are available in the FFTF design

Methodology and applicability of a safety and demonstration concept for a HAW final repository on clays. Safety concept and verification strategy

International Nuclear Information System (INIS)

Ruebel, Andre; Meleshyn, Artur

2014-08-01

The report describes the site independent frame for a safety concept and verification strategy for a final repository for heat generating wastes in clay rock. In the safety concept planning specifications and technical measures are summarized that are supposed to allow a safe inclusion of radionuclides in the host rock. The verification strategy defines the systematic procedures for the development of fundamentals and scenarios as basis for the demonstration of the safety case and to allow the prognosis of appropriateness. The report includes the boundary conditions, the safety concept for the post-closure phase and the verification strategy for the post-closure phase.

Colonising Safety : creating risk through the enforcement of biomedical constructions of safety.

NARCIS (Netherlands)

Kadetz, P.

2013-01-01

In the normative health care discourse, safety is represented as a concept that is at once universal, irrefutable, and inherently beneficent. Yet, research at local levels in the Philippines challenges these assumptions embedded in the biomedical construction of safety. This article examines how the

Operation: Inherent Resolve

DEFF Research Database (Denmark)

Cramer-Larsen, Lars

2015-01-01

Kapitlet giver læseren indsigt i den internationale koalitions engagement mod IS igennem Operaton Inherent Resolve; herunder koalitionens strategi i forhold til IS strategi, ligesom det belyser kampagnens legalitet og folkeretlige grundlag, ligesom det giver et bud på overvejelser om kampagnens...

The role of structural integrity in liquid metal fast breeder reactor safety

International Nuclear Information System (INIS)

Holmes, J.A.G.

1982-01-01

Extensive studies have demonstrated the favourable safety characteristics of liquid metal fast breeder reactors, which are attributable to both their inherent features and the engineered safeguards which are included. This requires demonstration that there is no risk of sudden catastrophic failure of the core support system allowing the core to drop off the control rods to give a prompt critical reactivity excursion. An important part of our work in support of the safety case for the U.K. Commercial Demonstration Fast Reactor is to demonstrate that such a failure is virtually incredible. This covers design features, study of the fracture behaviour of stainless steel structures, and inspection and monitoring during fabrication and service. The paper gives a broad description of the relevant design features and supporting work programme

Demonstration tests for low level radioactive waste packaging safety

International Nuclear Information System (INIS)

Nagano, I.; Shimura, S.; Miki, T.; Tamamura, T.; Kunitomi, K.

1993-01-01

The transport packaging for low level radioactive waste (so-called the LLW packaging) has been developed to be utilized for transportation of LLW in 200 liter-drums from Japanese nuclear power stations to the LLW Disposal Center at Rokkashomura in Aomori Prefecture. Transportation is expected to start from December in 1992. We will explain the brief history of the development, technical features and specifications as well as two kinds of safety demonstration tests, namely one is '1.2 meter free drop test' and the other is 'ISO container standard test'. (J.P.N.)

System Guidelines for EMC Safety-Critical Circuits: Design, Selection, and Margin Demonstration

Science.gov (United States)

Lawton, R. M.

1996-01-01

Demonstration of safety margins for critical points (circuits) has traditionally been required since it first became a part of systems-level Electromagnetic Compatibility (EMC) requirements of MIL-E-6051C. The goal of this document is to present cost-effective guidelines for ensuring adequate Electromagnetic Effects (EME) safety margins on spacecraft critical circuits. It is for the use of NASA and other government agencies and their contractors to prevent loss of life, loss of spacecraft, or unacceptable degradation. This document provides practical definition and treatment guidance to contain costs within affordable limits.

Overall System Description and Safety Characteristics of Prototype Gen IV Sodium Cooled Fast Reactor in Korea

Directory of Open Access Journals (Sweden)

Jaewoon Yoo

2016-10-01

Full Text Available The Prototype Gen IV sodium cooled fast reactor (PGSFR has been developed for the last 4 years, fulfilling the technology demonstration of the burning capability of transuranic elements included in light water reactor spent nuclear fuel. The PGSFR design has been focused on the robustness of safety systems by enhancing inherent safety characteristics of metal fuel and strengthening passive safety features using natural circulation and thermal expansion. The preliminary safety information document as a major outcome of the first design phase of PGSFR development was issued at the end of 2015. The project entered the second design phase at the beginning of 2016. This paper summarizes the overall structures, systems, and components of nuclear steam supply system and safety characteristics of the PGSFR. The research and development activities to demonstrate the safety performance are also briefly introduced in the paper.

Overall system description and safety characteristics of Prototype Gen IV Sodium Cooled Fast Reactor in Korea

International Nuclear Information System (INIS)

Yoo, Jae Woon; Chang, Jin Wook; Lim, Jae Yong; Cheon, Jin Sik; Lee, Tae Ho; Kim, Sung Kyun; Lee, Kwi Lim; Joo, Hyung Kook

2016-01-01

The Prototype Gen IV sodium cooled fast reactor (PGSFR) has been developed for the last 4 years, fulfilling the technology demonstration of the burning capability of transuranic elements included in light water reactor spent nuclear fuel. The PGSFR design has been focused on the robustness of safety systems by enhancing inherent safety characteristics of metal fuel and strengthening passive safety features using natural circulation and thermal expansion. The preliminary safety information document as a major outcome of the first design phase of PGSFR development was issued at the end of 2015. The project entered the second design phase at the beginning of 2016. This paper summarizes the overall structures, systems, and components of nuclear steam supply system and safety characteristics of the PGSFR. The research and development activities to demonstrate the safety performance are also briefly introduced in the paper

A concept of passive safety pressurized water reactor system with inherent matching nature of core heat generation and heat removal

International Nuclear Information System (INIS)

Murao, Yoshio; Araya, Fumimasa; Iwamura, Takamichi; Okumura, Keisuke

1995-01-01

The reduction of manpower in operation and maintenance by simplification of the system are essential to improve the safety and the economy of future light water reactors. At the Japan Atomic Energy Research Institute (JAERI), a concept of a simplified passive safety reactor system JPSR was developed for this purpose and in the concept minimization of developing work and conservation of scale-up capability in design were considered. The inherent matching nature of core heat generation and heat removal rate is introduced by the core with high reactivity coefficient for moderator density and low reactivity coefficient for fuel temperature (Doppler effect) and once-through steam generators (SGs). This nature makes the nuclear steam supply system physically-slave for the steam and energy conversion system by controlling feed water mass flow rate. The nature can be obtained by eliminating chemical shim and adopting in-vessel control rod drive mechanism (CRDM) units and a low power density core. In order to simplify the system, a large pressurizer, canned pumps, passive residual heat removal systems with air coolers as a final heat sink and passive coolant injection system are adopted and the functions of volume and boron concentration control and seal water supply are eliminated from the chemical and volume control system (CVCS). The emergency diesel generators and auxiliary component cooling system of 'safety class' for transferring heat to sea water as a final heat sink in emergency are also eliminated. All of systems are built in the containment except for the air coolers of the passive residual heat removal system. The analysis of the system revealed that the primary coolant expansion in 100% load reduction in 60 s can be mitigated in the pressurizer without actuating the pressure relief valves and the pressure in 50% load change in 30 s does not exceed the maximum allowable pressure in accidental conditions in regardless of pressure regulation. (author)

The inherent catastrophic traps in retrograde CTO PCI.

Science.gov (United States)

Wu, Eugene B; Tsuchikane, Etsuo

2018-05-01

When we learn to drive, our driving instructor tells us how to check the side mirror and turn your head to check the blind spot before changing lanes. He tells us how to stop at stop signs, how to drive in slippery conditions, the safe stopping distances, and these all make our driving safe. Similarly, when we learn PCI, our mentors teach us to seat the guiding catheter co-axially, to wire the vessel safely, to deliver balloon and stents over the wire, to watch the pressure of the guiding, in order that we perform PCI safely and evade complications. In retrograde CTO PCI, there is no such published teaching. Also many individual mentors have not had the wide experience to see all the possible complications of retrograde CTO PCI and, therefore, may not be able to warn their apprentice. As the number of retrograde procedures increase worldwide, there is a corresponding increase in catastrophic complications, many of which, we as experts, can see are easily avoidable. To breach this gap in knowledge, this article describes 12 commonly met inherent traps in retrograde CTO PCI. They are inherent because by arranging our equipment in the manner to perform retrograde CTO PCI, these complications are either induced directly or happen easily. We hope this work will enhance safety of retrograde CTO PCI and avoid many catastrophic complications for our readers and operators. © 2017 Wiley Periodicals, Inc. © 2017 Wiley Periodicals, Inc.

Molten salt reactor as asymptotic safety nuclear system

International Nuclear Information System (INIS)

Novikov, V.M.; Ignatyev, V.V.

1989-01-01

Safety is becoming the main and priority problem of the nuclear power development. An increase of the active safety measures could hardly be considered as the proper way to achieve the asymptotically high level of nuclear safety. It seem that the more realistic way to achieve such a goal is to minimize risk factors and to maximize the use of inherent and passive safety properties. The passive inherent safety features of the liquid fuel molten salt reactor (MSR) technology are making it attractive for future energy generation. The achievement of the asymptotic safety in MSR is being connected with the minimization of such risk factors as a reactivity excess, radioactivity stored, decay heat, non nuclear energy stored in core. In this paper safety peculiarities of the different MSR concepts are discussed

Environment, Safety, Health, and Quality Plan for the Buried Waste Integrated Demonstration Program

International Nuclear Information System (INIS)

Walker, S.

1994-05-01

The Buried Waste Integrated Demonstration (BWID) is a program funded by the US Department of Energy Office of Technology Development. BWID supports the applied research, development, demonstration, testing, and evaluation of a suite of advanced technologies that together form a comprehensive remediation system for the effective and efficient remediation of buried waste. This document describes the Environment, Safety, Health, and Quality requirements for conducting BWID activities at the Idaho National Engineering Laboratory. Topics discussed in this report, as they apply to BWID operations, include Federal, State of Idaho, and Environmental Protection Agency regulations, Health and Safety Plans, Quality Program Plans, Data Quality Objectives, and training and job hazard analysis. Finally, a discussion is given on CERCLA criteria and System and Performance audits as they apply to the BWID Program

Aviation safety and ICAO

NARCIS (Netherlands)

Huang, Jiefang

2009-01-01

The thesis addresses the issue of aviation safety under the rule of law. Aviation safety is a global concern. While air transport is considered a safe mode of travel, it is susceptible to inherent risks of flight, the use of force, and terrorist acts. Consequently, within the framework of the

Chinese nuclear heating test reactor and demonstration plant

International Nuclear Information System (INIS)

Wang Dazhong; Ma Changwen; Dong Duo; Lin Jiagui

1992-01-01

In this report the importance of nuclear district heating is discussed. From the viewpoint of environmental protection, uses of energy resources and transport, the development of nuclear heating in China is necessary. The development program of district nuclear heating in China is given in the report. At the time being, commissioning of the 5 MW Test Heating Reactor is going on. A 200 MWt Demonstration Plant will be built. In this report, the main characteristics of these reactors are given. It shows this type of reactor has a high inherent safety. Further the report points out that for this type of reactor the stability is very important. Some experimental results of the driving facility are included in the report. (orig.)

Demonstrating safety: Lessons learnt by InSOTEC

International Nuclear Information System (INIS)

Kallenbach-Herbert, Beate; Brohmann, Bettina

2014-01-01

InSOTEC is a three-year collaborative social sciences research project funded under the European Atomic Energy Community's 7. Framework Programme FP7/2007-2011, under grant agreement no. 2699009.1 The project aims to generate a better understanding of the complex interplay between the technical and the social in radioactive waste management (RWM) and, in particular, in the context of the design and implementation of geological disposal. In doing so, InSOTEC wants to move beyond the social and technical division by treating RWM and geological disposal as 'socio-technical' challenges and in following the relationship and describing the context, one can identify the dependency as a socio-technical combination. InSOTEC focuses on situations and issues where the relationship between the technical and social components of geological disposal are still unstable, ambiguous or controversial, and where negotiations are taking place in terms of problem definitions and preferred solutions. Some concrete examples of socio-technical challenges are the question of siting and of introducing the notion of reversibility and retrievability or long-term repository monitoring into the concept of geological disposal. These examples show that the concept of geological disposal develops over time, not only because of evolutions in scientific knowledge, but also as a consequence of debates on how to implement this technology in the light of societal requirements. During the first year of the project, various research activities in the national context of InSOTEC partner countries as well as on the European and international levels contributed to the identification of the main socio-technical challenges in geological disposal. On this basis four topics were selected for in-depth analysis: - reversibility and retrievability; - demonstrating safety; - siting; - technology transfer; The aim of these analyses is to come to a better understanding of the relationships between social and technical

Transient safety performance of the PRISM innovative liquid metal reactor

International Nuclear Information System (INIS)

Magee, P.M.; Dubberley, A.E.; Rhow, S.K.; Wu, T.

1988-01-01

The PRISM sodium-cooled reactor concept utilizes passive safety characteristics and modularity to increase performance margins, improve licensability, reduce owner's risk and reduce costs. The relatively small size of each reactor module (471 MWt) facilitates the use of passive self-shutdown and shutdown heat removal features, which permit design simplification and reduction of safety-related systems. Key to the transient performance is the inherent negative reactivity feedback characteristics of the core design resulting from the use of metal (U-Pu-Zr) swing, and very low control rod runout worth. Selected beyond design basis events relying only on these core design features are analyzed and the design margins summarized to demonstrate the advancement in reactor safety achieved with the PRISM design concept

Environmental and safety issues of the fusion fuel cycle

International Nuclear Information System (INIS)

Crocker, J.G.

1980-01-01

This paper discusses the environmental and safety concerns inherent in the development of fusion energy, and the current Department of Energy programs seeking to: (1) develop safe and reliable techniques for tritium control; (2) reduce the quantity of activation products produced; and (3) provide designs to limit the potential for accidents that could result in release of radioactive materials. Because of the inherent safety features of fusion and the early start that has been made in safety problem recognition and solution, fusion should be among the lower risk technologies for generation of commercial power

Multi-dimensional Code Development for Safety Analysis of LMR

International Nuclear Information System (INIS)

Ha, K. S.; Jeong, H. Y.; Kwon, Y. M.; Lee, Y. B.

2006-08-01

A liquid metal reactor loaded a metallic fuel has the inherent safety mechanism due to the several negative reactivity feedback. Although this feature demonstrated through experiments in the EBR-II, any of the computer programs until now did not exactly analyze it because of the complexity of the reactivity feedback mechanism. A multi-dimensional detail program was developed through the International Nuclear Energy Research Initiative(INERI) from 2003 to 2005. This report includes the numerical coupling the multi-dimensional program and SSC-K code which is used to the safety analysis of liquid metal reactors in KAERI. The coupled code has been proved by comparing the analysis results using the code with the results using SAS-SASSYS code of ANL for the UTOP, ULOF, and ULOHS applied to the safety analysis for KALIMER-150

Nuclear desalination in the Arab world - Part II: Advanced inherent and passive safe nuclear reactors

International Nuclear Information System (INIS)

Karameldin, A.; Samer S. Mekhemar

2004-01-01

Rapid increases in population levels have led to greater demands for fresh water and electricity in the Arab World. Different types of energies are needed to contribute to bridging the gap between increased demand and production. Increased levels of safeguards in nuclear power plants have became reliable due to their large operational experience, which now exceeds 11,000 years of operation. Thus, the nuclear power industry should be attracting greater attention. World electricity production from nuclear power has risen from 1.7% in 1970 to 17%-20% today. This ratio had increased in June 2002 to reach more than 30%, 33% and 42% in Europe, Japan, and South Korea respectively. In the Arab World, both the public acceptance and economic viability of nuclear power as a major source of energy are greatly dependent on the achievement of a high level of safety and environmental protection. An assessment of the recent generation of advanced reactor safety criteria requirements has been carried out. The promising reactor designs adapted for the Arab world and other similar developing countries are those that profit from the enhanced and passive safety features of the new generation of reactors, with a stronger focus on the effective use of intrinsic characteristics, simplified plant design, and easy construction, operation and maintenance. In addition, selected advanced reactors with a full spectrum from small to large capacities, and from evolutionary to radical types, which have inherent and passive safety features, are discussed. The relevant economic assessment of these reactors adapted for water/electricity cogeneration have been carried out and compared with non-nuclear desalination methods. This assessment indicates that, water/electricity cogeneration by the nuclear method with advanced inherent and passive safe nuclear power plants, is viable and competitive. (author)

Conceptual design study for the demonstration reactor of JSFR. (3) Safety design and evaluation

International Nuclear Information System (INIS)

Tani, Akihiro; Shimakawa, Yoshio; Kubo, Shigenobu; Fujimura, Ken; Yamano, Hidemasa

2011-01-01

This paper describes the result of conceptual safety design and evaluation for the demonstration plant of Japan sodium-cooled fast reactor (JSFR), which was preliminarily conducted for providing information necessary to decide the plant specification for further design study. The plant major specifications except for output power and safety design concept are almost the same as those of the commercial JSFR. A set of safety evaluation for typical design basis events (DBEs) is mainly focused here, which was conducted for the 750 MWe design. Safety analyses for DBEs evaluation were performed on the basis of conservative assumptions using a one-dimensional flow network code with point kinetics. For representative DBEs, transient over power type events and loss of flow type events were analyzed. The long-term loss-of-offsite power event was also calculated to evaluate the natural circulation decay heat removal system. All analytical results showed to meet tentative safety criteria, thus it was confirmed that the safety design concept of JSFR is feasible against DBEs. (author)

Metal-fuel modeling for inherently safe reactor designs

International Nuclear Information System (INIS)

Miles, K.J. Jr.

1987-01-01

Current development of breeder reactor systems has led to the renewed interest in metal fuels. These fuels have properties that enhance the inherent safety of the system, such as high thermal conductivity, compatibility with liquid sodium, and low fuel/cladding mechanical interaction. While metal-fuel irradiation behavior is well understood, there are some areas where more information is needed to fully understand the various safety-related phenomena, such as fuel/cladding chemical interaction, eutectic melting and penetration, and axial relocation of molten fuel prior to cladding breach. Because many of these phenomena can cause changes in the reactivity state of the system, their effects on whole-core normal, anticipated, and hypothetical accident scenarios need to be studied. The metal-fuel behavior model DEFORM-5 is being developed to provide the necessary phenomenological basis for these studies. The first stage in the DEFORM-5 development has been completed. Presently, DEFORM-5 calculates the cladding strain, life fraction, and eutectic penetration thinning for Types D9, HT9, or 316 steels. This first stage of DEFORM-5 has been used to analyze the TREAT M2, M3, and M4 transients with irradiated Experimental Breeder Reactor-II driver fuel. The paper shows the DEFORM-5 and experimental results for failure times for the test pins. The results provide confidence and validation of the DEFORM-5 modeling of the cladding behavior

Conceptual design of inherently safe integral reactor

Energy Technology Data Exchange (ETDEWEB)

Kim, J. I.; Chang, M. H.; Lee, D. J. and others

1999-03-01

The design concept of a 300 MWt inherently safe integral reactor(ISIR) for the propulsion of extra large and superhigh speed container ship was developed in this report. The scope and contents of this report are as follows : 1. The state of the art of the technology for ship-mounted reactor 2. Design requirements for ISIR 3. Fuel and core design 4. Conceptual design of fluid system 5. Conceptual design of reactor vessel assembly and primary components 6. Performance analyses and safety analyses. Installation of two ISIRs with total thermal power of 600MWt and efficiency of 21% is capable of generating shaft power of 126,000kW which is sufficient to power a container ship of 8,000TEU with 30knot cruise speed. Larger and speedier ship can be considered by installing 4 ISIRs. Even though the ISIR was developed for ship propulsion, it can be used also for a multi-purpose nuclear power plant for electricity generation, local heating, or seawater desalination by mounting on a movable floating barge. (author)

A Holistic Approach to Protection and Safety

International Nuclear Information System (INIS)

Yankovich, L.T.

2017-01-01

Natural Ecosystems Undergo Inherent Fluctuations and Changes that are Related to Physical, Geological, Biological and other Natural Processes Associated with the Cycles of Life and the Earth. In other words, the effects of human activities must be evaluated in the context of the 'baseline' processes that are inherent to and imposed on natural ecosystems. The IAEA has established ahierarchy of Safety Standards, consisting of a set of ''Safety Fundamentals'' at the highest level, followed by ''Requirements'', then recommendations or ''guidance''. Human activities involving the use of radiation and radioactive substances can cause radiation exposure to people and the environment. This exposure should be regulated and monitored in accordance with international safety standards and national legislation

Engineered safeguards and passive safety features (safety analysis detailed report no. 6)

Energy Technology Data Exchange (ETDEWEB)

1988-01-15

The Safety-Analysis Summary lists the reactor's safety aspects for passive and active prevention of severe accidents and mitigation of accident consequences, i.e., intrinsic and passive protections of the plant; intrinsic and passive protections of the core; inherent decay-heat removal systems; rapid-shutdown systems; four physical containment barriers. This report goes into further details regarding some of this aspects.

Inherent wettability of different rock surfaces at nanoscale: a theoretical study

Science.gov (United States)

Chang, Xiao; Xue, Qingzhong; Li, Xiaofang; Zhang, Jianqiang; Zhu, Lei; He, Daliang; Zheng, Haixia; Lu, Shuangfang; Liu, Zilong

2018-03-01

Investigating the inherent wettability of rock surfaces at nanoscale is of great importance in ore floatation and oil recovery field. Using molecular dynamics simulations, we systematically study the wetting behavior of water on different rock surfaces (silica, calcite, gypsum, halite and graphite) at nanoscale. It is demonstrated that the inherent rock wettability follows the order of gypsum > calcite > halite > silica > graphite. Remarkably, we also manifest that the polarity of oil molecules can affect the water contact angles on silica surface. For example, the water contact angles on silica surface in hexane, dodecane, thiophene and toluene are 58 ± 2°, 63 ± 3°, 90 ± 1°, 118 ± 1°, respectively. Furthermore, we investigate the wetting behavior of water on heterogeneous rock surfaces and find that water molecules can move from hydrophobic surface to hydrophilic surface.

Inherent Anticipation in the Pharmaceutical and Biotechnology Industries.

Science.gov (United States)

Goldman, Michael; Evans, Georgia; Zappia, Andrew

2015-04-15

Pharmaceutical and biotech research often involves discovering new properties of, or new methods to use, existing compositions. The doctrine of inherent anticipation, however, prevents the issuance and/or validity of a patent for discoveries deemed to have been implicitly disclosed in the prior art. This can be a barrier to patent rights in these technologies. Inherent anticipation therefore creates uncertainty for patent protection in the pharmaceutical and biotech sciences. Despite this uncertainty, Federal Circuit jurisprudence provides guidance on the boundaries of the inherent anticipation doctrine. In view of the case law, certain strategies may be employed to protect inventions that may potentially be viewed as inherent in the prior art. Copyright © 2015 Cold Spring Harbor Laboratory Press; all rights reserved.

Can the inherence heuristic explain vitalistic reasoning?

Science.gov (United States)

Bastian, Brock

2014-10-01

Inherence is an important component of psychological essentialism. By drawing on vitalism as a way in which to explain this link, however, the authors appear to conflate causal explanations based on fixed features with those based on general causal forces. The disjuncture between these two types of explanatory principles highlights potential new avenues for the inherence heuristic.

Evaluation of methods and tools to develop safety concepts and to demonstrate safety for an HLW repository in salt. Final report

Energy Technology Data Exchange (ETDEWEB)

Bollingerfehr, W.; Buhmann, D.; Doerr, S.; and others

2017-03-15

safety demonstration are the integrity proofs for the geological and geotechnical barriers and analysis of backfill compaction. In addition, any possible radionuclide release from the repository to the environment has also to be assessed. The safety and demonstration concept developed in the course of the ISIBEL project was further evolved and applied in the course of the R and D project ''Vorlaeufige Sicherheitsanalyse Gorleben - VSG'' (preliminary safety analysis Gorleben) as an example for an HLW repository in a domal salt structure. The repository concepts also consider the requirement for retrievability of stored waste during the operational phase of the repository. The results of the R and D project VSG provide evidence that a safe HLW repository within a salt dome of a suitable geologic structure is feasible. The long-term safety can be ensured using state-of-the-art science and technology. In 2010, the Federal Ministry for the Environment, Nature Conservation and Nuclear Safety (BMU) issued new safety requirements for the disposal of heat-generating radioactive waste. These requirements have been included in the analysis. After completion of the VSG project in 2013 complementary work has been performed within the framework of the ISIBEL programme. In this context e.g. potential contributions of natural and antropogenic analogs to confidence building were addressed as well as the feasibility and limits of deriving a repository conc ept strictly from requirements. The report in hands provides a comprehensive summary of the results of R and D work regarding HLW disposal in domal salt formations that has been performed after launching the ISIBEL programme in 2005. This study shows the depth of the geological and technical knowledge on final disposal of HLW in a salt dome with a suitable geologic structure that had been gained up to now and demonstrates that the tools required for safety evaluations are available and allow reliable safety

Evaluation of methods and tools to develop safety concepts and to demonstrate safety for an HLW repository in salt. Final report

International Nuclear Information System (INIS)

Bollingerfehr, W.; Buhmann, D.; Doerr, S.

2017-03-01

safety demonstration are the integrity proofs for the geological and geotechnical barriers and analysis of backfill compaction. In addition, any possible radionuclide release from the repository to the environment has also to be assessed. The safety and demonstration concept developed in the course of the ISIBEL project was further evolved and applied in the course of the R and D project ''Vorlaeufige Sicherheitsanalyse Gorleben - VSG'' (preliminary safety analysis Gorleben) as an example for an HLW repository in a domal salt structure. The repository concepts also consider the requirement for retrievability of stored waste during the operational phase of the repository. The results of the R and D project VSG provide evidence that a safe HLW repository within a salt dome of a suitable geologic structure is feasible. The long-term safety can be ensured using state-of-the-art science and technology. In 2010, the Federal Ministry for the Environment, Nature Conservation and Nuclear Safety (BMU) issued new safety requirements for the disposal of heat-generating radioactive waste. These requirements have been included in the analysis. After completion of the VSG project in 2013 complementary work has been performed within the framework of the ISIBEL programme. In this context e.g. potential contributions of natural and antropogenic analogs to confidence building were addressed as well as the feasibility and limits of deriving a repository conc ept strictly from requirements. The report in hands provides a comprehensive summary of the results of R and D work regarding HLW disposal in domal salt formations that has been performed after launching the ISIBEL programme in 2005. This study shows the depth of the geological and technical knowledge on final disposal of HLW in a salt dome with a suitable geologic structure that had been gained up to now and demonstrates that the tools required for safety evaluations are available and allow reliable safety assessments of HLW

Safety performance of preliminary KALIMER conceptual design

Energy Technology Data Exchange (ETDEWEB)

Hahn Dohee; Kim Kyoungdoo; Kwon Youngmin; Chang Wonpyo; Suk Soodong [Korea atomic Energy Resarch Inst., Taejon (Korea)

1999-07-01

The Korea Atomic Energy Research Institute (KAERI) is developing KALIMER (Korea Advanced Liquid Metal Reactor), which is a sodium cooled, 150 MWe pool-type reactor. The safety design of KALIMER emphasizes accident prevention by using passive processes, which can be accomplished by the safety design objectives including the utilization of inherent safety features. In order to assess the effectiveness of the inherent safety features in achieving the safety design objectives, a preliminary evaluation of ATWS performance for the KALIMER design has been performed with SSC-K code, which is a modified version of SSC-L code. KAERI's modification of the code includes development of reactivity feedback models for the core and a pool model for KALIMER reactor vessel. This paper describes the models for control rod driveline expansion, gas expansion module and the thermal hydraulic model for reactor pool and the results of preliminary analyses for unprotected loss of flow and loss o heat sink. (author)

Safety performance of preliminary KALIMER conceptual design

International Nuclear Information System (INIS)

Hahn Dohee; Kim Kyoungdoo; Kwon Youngmin; Chang Wonpyo; Suk Soodong

1999-01-01

The Korea Atomic Energy Research Institute (KAERI) is developing KALIMER (Korea Advanced Liquid Metal Reactor), which is a sodium cooled, 150 MWe pool-type reactor. The safety design of KALIMER emphasizes accident prevention by using passive processes, which can be accomplished by the safety design objectives including the utilization of inherent safety features. In order to assess the effectiveness of the inherent safety features in achieving the safety design objectives, a preliminary evaluation of ATWS performance for the KALIMER design has been performed with SSC-K code, which is a modified version of SSC-L code. KAERI's modification of the code includes development of reactivity feedback models for the core and a pool model for KALIMER reactor vessel. This paper describes the models for control rod driveline expansion, gas expansion module and the thermal hydraulic model for reactor pool and the results of preliminary analyses for unprotected loss of flow and loss o heat sink. (author)

Demonstrating safety during license renewal should not be a large task

International Nuclear Information System (INIS)

Berto, D.S.

1993-01-01

The principal regulatory goal related to nuclear power plant operation is to ensure the health and safety of the public. The principal goal of extended plant operation via the license renewal process is also to ensure the health and safety of the public. The license renewal documentation issued by the Nuclear Regulatory Commission (NRC) provides guidance on what will be acceptable to the NRC in a license renewal application to demonstrate that this goal will be met. Application of this guidance is currently open to wide interpretation, with many of the current approaches proving to be extremely costly, complex, and uncertain of acceptability. This paper evaluates the requirements necessary to ensure the continued health and safety of the public during any license renewal term. This evaluation is based on the stated goals of the License Renewal Rule and on the published bases for the Rule. An approach to License Renewal is recommended that: (1) meets the stated goals of the NRC; (2) is consistent with current regulatory practices; and (3) will continue to ensure the health and safety of the public. This recommended approach is also much less costly than other current approaches, and can be easily agreed to by all participants. This approach will meet regulatory goals, while removing the cost and uncertainty obstacles currently being confronted by utilities. Providing a viable approach to license renewal will allow the renewal process to be pursued by utilities. Without such an approach, safe and reliable nuclear power plants will be permanently shut down at the arbitrary 40 year license limit

High temperature reactor safety and environment

International Nuclear Information System (INIS)

Brisbois, J.; Charles, J.

1975-01-01

High-temperature reactors are endowed with favorable safety and environmental factors resulting from inherent design, main-component safety margins, and conventional safety systems. The combination of such characteristics, along with high yields, prove in addition, that such reactors are plagued with few problems, can be installed near users, and broaden the recourse to specific power, therefore fitting well within a natural environment [fr

Initial Demonstration of the Real-Time Safety Monitoring Framework for the National Airspace System Using Flight Data

Science.gov (United States)

Roychoudhury, Indranil; Daigle, Matthew; Goebel, Kai; Spirkovska, Lilly; Sankararaman, Shankar; Ossenfort, John; Kulkarni, Chetan; McDermott, William; Poll, Scott

2016-01-01

As new operational paradigms and additional aircraft are being introduced into the National Airspace System (NAS), maintaining safety in such a rapidly growing environment becomes more challenging. It is therefore desirable to have an automated framework to provide an overview of the current safety of the airspace at different levels of granularity, as well an understanding of how the state of the safety will evolve into the future given the anticipated flight plans, weather forecast, predicted health of assets in the airspace, and so on. Towards this end, as part of our earlier work, we formulated the Real-Time Safety Monitoring (RTSM) framework for monitoring and predicting the state of safety and to predict unsafe events. In our previous work, the RTSM framework was demonstrated in simulation on three different constructed scenarios. In this paper, we further develop the framework and demonstrate it on real flight data from multiple data sources. Specifically, the flight data is obtained through the Shadow Mode Assessment using Realistic Technologies for the National Airspace System (SMART-NAS) Testbed that serves as a central point of collection, integration, and access of information from these different data sources. By testing and evaluating using real-world scenarios, we may accelerate the acceptance of the RTSM framework towards deployment. In this paper we demonstrate the framework's capability to not only estimate the state of safety in the NAS, but predict the time and location of unsafe events such as a loss of separation between two aircraft, or an aircraft encountering convective weather. The experimental results highlight the capability of the approach, and the kind of information that can be provided to operators to improve their situational awareness in the context of safety.

Inherent Safety Features and Passive Prevention Approaches for Pb/Bi-cooled Accelerator-Driven Systems

Energy Technology Data Exchange (ETDEWEB)

Carlsson, Johan

2003-03-01

This thesis is devoted to the investigation of passive safety and inherent features of subcritical nuclear transmutation systems - accelerator-driven systems. The general objective of this research has been to improve the safety performance and avoid elevated coolant temperatures in worst-case scenarios like unprotected loss-of-flow accidents, loss-of-heat-sink accidents, and a combination of both these accident initiators. The specific topics covered are emergency decay heat removal by reactor vessel auxiliary cooling systems, beam shut-off by a melt-rupture disc, safety aspects from locating heat-exchangers in the riser of a pool-type reactor system, and reduction of pressure resistance in the primary circuit by employing bypass routes. The initial part of the research was focused on reactor vessel auxiliary cooling systems. It was shown that an 80 MW{sub th} Pb/Bi-cooled accelerator-driven system of 8 m height and 6 m diameter vessel can be well cooled in the case of loss-of-flow accidents in which the accelerator proton beam is not switched off. After a loss-of-heat-sink accident the proton beam has to be interrupted within 40 minutes in order to avoid fast creep of the vessel. If a melt-rupture disc is included in the wall of the beam pipe, which breaks at 150 K above the normal core outlet temperature, the grace period until the beam has to be shut off is increased to 6 hours. For the same vessel geometry, but an operating power of 250 MW{sub th} the structural materials can still avoid fast creep in case the proton beam is shut off immediately. If beam shut-off is delayed, additional cooling methods are needed to increase the heat removal. Investigations were made on the filling of the gap between the guard and the reactor vessel with liquid metal coolant and using water spray cooling on the guard vessel surface. The second part of the thesis presents examinations regarding an accelerator-driven system also cooled with Pb/Bi but with heat-exchangers located

Inherent Safety Features and Passive Prevention Approaches for Pb/Bi-cooled Accelerator-Driven Systems

International Nuclear Information System (INIS)

Carlsson, Johan

2003-03-01

This thesis is devoted to the investigation of passive safety and inherent features of subcritical nuclear transmutation systems - accelerator-driven systems. The general objective of this research has been to improve the safety performance and avoid elevated coolant temperatures in worst-case scenarios like unprotected loss-of-flow accidents, loss-of-heat-sink accidents, and a combination of both these accident initiators. The specific topics covered are emergency decay heat removal by reactor vessel auxiliary cooling systems, beam shut-off by a melt-rupture disc, safety aspects from locating heat-exchangers in the riser of a pool-type reactor system, and reduction of pressure resistance in the primary circuit by employing bypass routes. The initial part of the research was focused on reactor vessel auxiliary cooling systems. It was shown that an 80 MW th Pb/Bi-cooled accelerator-driven system of 8 m height and 6 m diameter vessel can be well cooled in the case of loss-of-flow accidents in which the accelerator proton beam is not switched off. After a loss-of-heat-sink accident the proton beam has to be interrupted within 40 minutes in order to avoid fast creep of the vessel. If a melt-rupture disc is included in the wall of the beam pipe, which breaks at 150 K above the normal core outlet temperature, the grace period until the beam has to be shut off is increased to 6 hours. For the same vessel geometry, but an operating power of 250 MW th the structural materials can still avoid fast creep in case the proton beam is shut off immediately. If beam shut-off is delayed, additional cooling methods are needed to increase the heat removal. Investigations were made on the filling of the gap between the guard and the reactor vessel with liquid metal coolant and using water spray cooling on the guard vessel surface. The second part of the thesis presents examinations regarding an accelerator-driven system also cooled with Pb/Bi but with heat-exchangers located in the

Safety assessment of envisaged systems for automotive hydrogen supply and utilization

Energy Technology Data Exchange (ETDEWEB)

Landucci, Gabriele [Dipartimento di Ingegneria Chimica, Chimica Industriale e Scienza dei Materiali, Universita di Pisa, via Diotisalvi n.2, 56126 Pisa (Italy); Tugnoli, Alessandro; Cozzani, Valerio [Dipartimento di Ingegneria Chimica, Mineraria e delle Tecnologie Ambientali, Alma Mater Studiorum - Universita di Bologna, via Terracini n.28, 40131 Bologna (Italy)

2010-02-15

A novel consequence-based approach was applied to the inherent safety assessment of the envisaged hydrogen production, distribution and utilization systems, in the perspective of the widespread hydrogen utilization as a vehicle fuel. Alternative scenarios were assessed for the hydrogen system chain from large scale production to final utilization. Hydrogen transportation and delivery was included in the analysis. The inherent safety fingerprint of each system was quantified by a set of Key Performance Indicators (KPIs). Rules for KPIs aggregation were considered for the overall assessment of the system chains. The final utilization stage resulted by large the more important for the overall expected safety performance of the system. Thus, comparison was carried out with technologies proposed for the use of other low emission fuels, as LPG and natural gas. The hazards of compressed hydrogen-fueled vehicles resulted comparable, while reference innovative hydrogen technologies evidenced a potentially higher safety performance. Thus, switching to the inherently safer technologies currently under development may play an important role in the safety enhancement of hydrogen vehicles, resulting in a relevant improvement of the overall safety performance of the entire hydrogen system. (author)

Safety features and licensing of CNNC-ACP100

Energy Technology Data Exchange (ETDEWEB)

Zhong, F., E-mail: Zhongfj2000@163.com [Nuclear Power Inst. of China, National Key Lab. of Science and Technology on Reactor System Design Technology (China)

2014-07-01

ACP100 is an innovatory modular pressurized water reactor, the engineering safety systems fully adopt passive safety design technology. Its inherent safety and passive features/systems are verified via testing facilities and are highlighted at certain levels of defence in depth. The licensing of ACP 100 is within current LWR framework and meets up-to-date codes and requirements in nuclear safety. (author)

Development of the Digital Reactor Safety System

International Nuclear Information System (INIS)

Lee, Dong Young; Lee, C. K.; Hwang, I. K.

2008-04-01

Objectives of Project - Development of Digital Safety Grade PLC and Licensing - Development of Safety System(RPS) and Licensing - Development of Safety System(ESF-CCS) and Licensing Content and Result of Project - POSAFE-Q PLC : Development of PLC platform for Shin-UCN unit 1 and 2 ·Development Scope : Processor module, Power module, 3 kinds of Communication module, Bus extension module(Master and Slave), 16 kinds of Input and Output module ·PLC application software development tool(pSET) - IDiPS RPS and IDiPS ESF-CCS : Development of PPS for Sin-UCN 1 and 2 ·Development Scope - 4-channels RPS with the KNICS inherent architecture - A part of 1-channels ESF-CCS with the KNICS inherent architecture - Licensing ·optical Report Submitted and Expected to finish the licensing process until Aug. 2008

Thermosetting resins with high fractions of free volume and inherently low dielectric constants.

Science.gov (United States)

Lin, Liang-Kai; Hu, Chien-Chieh; Su, Wen-Chiung; Liu, Ying-Ling

2015-08-18

This work demonstrates a new class of thermosetting resins, based on Meldrum's acid (MA) derivatives, which have high fractions of free volume and inherently low k values of about 2.0 at 1 MHz. Thermal decomposition of the MA groups evolves CO2 and acetone to create air-trapped cavities so as to reduce the dielectric constants.

Relocation work of temporary thermocouples for measuring the vessel cooling system in the safety demonstration test

International Nuclear Information System (INIS)

Shimazaki, Yosuke; Shinohara, Masanori; Ono, Masato; Yanagi, Shunki; Tochio, Daisuke; Iigaki, Kazuhiko

2012-05-01

It is necessary to confirm that the temperature of water cooling panel of the vessel cooling system (VCS) is controlled under the allowable working temperature during the safety demonstration test because the water cooling panel temperature rises due to stop of cooling water circulation pumps. Therefore, several temporary thermocouples are relocated to the water cooling panel near the stabilizers of RPV and the side cooling panel outlet ring header of VCS in order to observe the temperature change of VCS. The relocated thermocouples can measure the temperature change with starting of the cooling water circulation pumps of VCS. So it is confirmed that the relocated thermocouples can observe the VCS temperature change in the safety demonstration test. (author)

Conceptual design of Inherently Safe Fast Reactor (ISFR)

International Nuclear Information System (INIS)

Asahi, Yoshiro

2003-01-01

ISFR is a boiling heavy water fast reactor of process inherent ultimate safety (PIUS) type. ISFR may breed fuel in the core. Owing to a positive void coefficient, the application of the PIUS concept to ISFR is not straightforward. Thus, the gap conductance is small so that the time constant τ α of the positive void feedback process is sufficiently large, while the initially-closed two-way check valves to be used as passive switches to the pumps are installed at the lower honeycombs. As a result, the passive shutdown mechanisms can come into effect sufficiently soon to suppress the positive feedback reactivity. Both large τ α and the passive switches also help stabilize the system so that ISFR can perform a constant power operation with a simple control logic for the main coolant pump speed. In a steam generator tube rupture, fuel temperature was found to smoothly decrease to the decay heat level with nucleate boiling. The feasibility of ISFR was proved only to some extent. (author)

Laser safety tools and training

CERN Document Server

Barat, Ken

2008-01-01

Lasers perform many unique functions in a plethora of applications, but there are many inherent risks with this continually burgeoning technology. Laser Safety: Tools and Training presents simple, effective ways for users in a variety of facilities to evaluate the hazards of any laser procedure and ensure they are following documented laser safety standards.Designed for use as either a stand-alone volume or a supplement to Laser Safety Management, this text includes fundamental laser and laser safety information and critical laser use information rarely found in a single source. The first lase

Licensing issues for inherently safe fast reactors

International Nuclear Information System (INIS)

Kastenberg, W.E.; Lee, S.; Okrent, D.

1986-01-01

There has been considerable interest recently in a new generation of liquid metal reactor (LMR) concepts in the US. Some significant changes in regulatory philosophy will be required if the anticipated cost advantages of inherently safe designs are to be achieved. The defense in depth philosophy will need to be significantly re-evaluated in the context of inherently safe reactors. It is the purpose of this paper to begin such a re-evaluation of this regulatory philosophy

Model development for the dynamic analysis of the OSU inherently safe reactor. Part 1

International Nuclear Information System (INIS)

Aybar, H.S.

1992-01-01

Faculty and students in the Nuclear Engineering Program at the Ohio State University (OSU) have proposed a conceptual design for an inherently safe 340 MWe power reactor. The design is based on the state-of-the-art technology of LWRs and the High Temperature Gas- cooled Reactors (HTGRs). The OSU Inherently Safe Reactor (OSU-ISR) concept uses shorter than standard BWR fuel elements in the reactor core. All the fluid on the primary side is contained within a Prestressed Concrete Reactor Vessel (PCRV). This important feature significantly reduces the probability of a LOCA. A new feature of the OSU-ISR is an operator independent steam driven Emergency Core Cooling System (ECCS) housed within the PCRV. In accident conditions where the steam generators are incapacitated, steam from the core drives a jet injector, which takes water from the suppression pool and pumps it into the core cavity to maintain core coverability. The preliminary analysis of the concept was performed as a design project in the Nuclear Engineering Program at the OSU during the Spring of 1985, and published in ''Nuclear Technology.'' The use of a PCRV for ducting and containment and the replacement of forced recirculation with natural circulation on the primary side significantly improve the inherent safety of the plant. Currently, work is in progress for the refinement of the OSU-ISR concept, partially supported by a grant from the U.S. Department of Energy

Alternate approaches to nuclear safety

International Nuclear Information System (INIS)

Crane, A.T.

1985-01-01

For the US nuclear power industry to expand, a greatly increased portion of the public must come to share the industry's confidence in reactor safety. Major obstacles to establishing this confidence are frequent incidents with potential safety implications and a lack of incontrovertible proof that the risk of a major accident is very low. The most important step toward overcoming these obstacles would be for each utility to operate, maintain, and evaluate its reactors according to far higher standards. With improvements in reliability and safety margins, existing plants would be a stimulus for building new ones rather than an impediment. If changes to the operation of existing plants and improvements to the design of future ones were inadequate, the only hope for a revival of the nuclear industry would be an alternative reactor so obviously safe that risk would no longer be an issue. Three possible concepts are the modular high-temperature gas reactor, the process inherent ultimate safety reactor, and the liquid-metal fast reactor. All three have inherent safety features that should make a meltdown essentially impossible. They cannot know just how great the advantage of these alternate reactors would be, but the benefits of developing one or more of the concepts appear great

Public education through safety culture demonstration

International Nuclear Information System (INIS)

Wanitsuksombut, Warapon

2005-01-01

The activities relating to nuclear energy have been world widely opposed against, because there have existed scars in the past; atomic bombs and a few accidents in nuclear facilities. It cannot be denied that the most effective education of public is through Medias such as news or documentary on newspaper and television. Once such cases appeared to public, it is difficult to erase the bad pictures from their memory. Since education for public is mainly depending on media, it is recommended putting harder effort on dissemination of information on regulation and regulatory function to public. The regulatory function of each country is the key of safe utilization of nuclear energy. Since prime responsibility of maintenance and operation are rested on the operators. To achieve the goal of safety, regulatory authority's task now is emphasized on encouraging operators of nuclear facilities to implement their safety culture. This will reduce the probability of unwanted events and therefore raising credit of nuclear energy. (author)

Modular High Temperature Gas-Cooled Reactor Safety Basis and Approach

Energy Technology Data Exchange (ETDEWEB)

David Petti; Jim Kinsey; Dave Alberstein

2014-01-01

Various international efforts are underway to assess the safety of advanced nuclear reactor designs. For example, the International Atomic Energy Agency has recently held its first Consultancy Meeting on a new cooperative research program on high temperature gas-cooled reactor (HTGR) safety. Furthermore, the Generation IV International Forum Reactor Safety Working Group has recently developed a methodology, called the Integrated Safety Assessment Methodology, for use in Generation IV advanced reactor technology development, design, and design review. A risk and safety assessment white paper is under development with respect to the Very High Temperature Reactor to pilot the Integrated Safety Assessment Methodology and to demonstrate its validity and feasibility. To support such efforts, this information paper on the modular HTGR safety basis and approach has been prepared. The paper provides a summary level introduction to HTGR history, public safety objectives, inherent and passive safety features, radionuclide release barriers, functional safety approach, and risk-informed safety approach. The information in this paper is intended to further the understanding of the modular HTGR safety approach. The paper gives those involved in the assessment of advanced reactor designs an opportunity to assess an advanced design that has already received extensive review by regulatory authorities and to judge the utility of recently proposed new methods for advanced reactor safety assessment such as the Integrated Safety Assessment Methodology.

Comparative analysis of operation and safety of subcritical nuclear systems and innovative critical reactors; Analyse comparative du fonctionnement et de la surete de systemes sous-critiques et de reacteurs critiques innovants

Energy Technology Data Exchange (ETDEWEB)

Bokov, P.M

2005-05-01

The main goal of this thesis work is to investigate the role of core subcriticality for safety enhancement of advanced nuclear systems, in particular, molten salt reactors, devoted to both energy production and waste incineration/transmutation. The inherent safety is considered as ultimate goal of this safety improvement. An attempt to apply a systematic approach for the analysis of the subcriticality contribution to inherent properties of hybrid system was performed. The results of this research prove that in many cases the subcriticality may improve radically the safety characteristics of nuclear reactors, and in some configurations it helps to reach the 'absolute' intrinsic safety. In any case, a proper choice of subcriticality level makes all analyzed transients considerably slower and monotonic. It was shown that the weakest point of the independent-source systems with respect to the intrinsic safety is thermohydraulic unprotected transients, while in the case of the coupled-source systems the excess reactivity/current insertion events remain a matter of concern. To overcome these inherent drawbacks a new principle of realization of a coupled sub-critical system (DENNY concept) is proposed. In addition, the ways to remedy some particular safety-related problems with the help of the core sub-criticality are demonstrated. A preliminary safety analysis of the fast-spectrum molten salt reactor (REBUS concept) is also carried out in this thesis work. Finally, the potential of the alternative (to spallation) neutron sources for application in hybrid systems is examined. (author)

Safety evaluation report of hot cell facilities for demonstration of advanced spent fuel conditioning process

International Nuclear Information System (INIS)

You, Gil Sung; Choung, W. M.; Ku, J. H.; Cho, I. J.; Kook, D. H.; Park, S. W.; Bek, S. Y.; Lee, E. P.

2004-10-01

The advanced spent fuel conditioning process(ACP) proposed to reduce the overall volume of the PWR spent fuel and improve safety and economy of the long-term storage of spent fuel. In the next phase(2004∼2006), the hot test will be carried out for verification of the ACP in a laboratory scale. For the hot test, the hot cell facilities of α- type and auxiliary facilities are required essentially for safe handling of high radioactive materials. As the hot cell facilities for demonstration of the ACP, a existing hot cell of β- type will be refurbished to minimize construction expenditures of hot cell facility. Up to now, the detail design of hot cell facilities and process were completed, and the safety analysis was performed to substantiate secure of conservative safety. The design data were submitted for licensing which was necessary for construction and operation of hot cell facilities. The safety investigation of KINS on hot cell facilities was completed, and the license for construction and operation of hot cell facilities was acquired already from MOST. In this report, the safety analysis report submitted to KINS was summarized. And also, the questionnaires issued from KINS and answers of KAERI in process of safety investigation were described in detail

Inherent-opening-controlled pattern formation in carbon nanotube arrays

International Nuclear Information System (INIS)

Huang Xiao; Zhou, Jijie J; Sansom, Elijah; Gharib, Morteza; Haur, Sow Chorng

2007-01-01

We have introduced inherent openings into densely packed carbon nanotube arrays to study self-organized pattern formation when the arrays undergo a wetting-dewetting treatment from nanotube tips. These inherent openings, made of circular or elongated hollows in nanotube mats, serve as dewetting centres, from where liquid recedes from. As the dewetting centres initiate dry zones and the dry zones expand, surrounding nanotubes are pulled away from the dewetting centres by liquid surface tension. Among short nanotubes, the self-organized patterns are consistent with the shape of the inherent openings, i.e. slender openings lead to elongated trench-like structures, and circular holes result in relatively round nest-like arrangements. Nanotubes in a relatively high mat are more connected, like in an elastic body, than those in a short mat. Small cracks often initialize themselves in a relatively high mat, along two or more adjacent round openings; each of the cracks evolves into a trench as liquid dries up. Self-organized pattern control with inherent openings needs to initiate the dewetting process above the nanotube tips. If there is no liquid on top, inherent openings barely enlarge themselves after the wetting-dewetting treatment

Measurement of inherent optical properties in the Arabian Sea

Digital Repository Service at National Institute of Oceanography (India)

Suresh, T.; Desa, E.; Kurian, J.; Mascarenhas, A.A.M.Q.

Inherent optical properties, absorption and began attenuation were measured in situ using a reflective tube absorption meter at nint wavelength, 412, 440, 488, 510, 555, 630, 650, 676 and 715 nm, in the Arabian Sea during March. Since inherent...

Development and implementation of setpoint tolerances for special safety systems

International Nuclear Information System (INIS)

Oliva, A.F.; Balog, G.; Parkinson, D.G.; Archinoff, G.H.

1991-01-01

The establishment of tolerances and impairment limits for special safety system setpoints is part of the process whereby the plant operator demonstrates to the regulatory authority that the plant operates safely and within the defined plant licensing envelope. The licensing envelope represents the set of limits and plant operating state and for which acceptably safe plant operation has been demonstrated by the safety analysis. By definition, operation beyond this envelope contributes to overall safety system unavailability. Definition of the licensing envelope is provided in a wide range of documents including the plant operating licence, the safety report, and the plant operating policies and principles documents. As part of the safety analysis, limits are derived for each special safety system initiating parameter such that the relevant safety design objectives are achieved for all design basis events. If initiation on a given parameter occurs at a level beyond its limit, there is a potential reduction in safety system effectiveness relative to the performance credited in the plant safety analysis. These safety system parameter limits, when corrected for random and systematic instrument errors and other errors inherent in the process of periodic testing or calibration, are then used to derive parameter impairment levels and setpoint tolerances. This paper describes the methodology that has evolved at Ontario Hydro for developing and implementing tolerances for special safety system parameters (i.e., the shutdown systems, emergency coolant injection system and containment system). Tolerances for special safety system initiation setpoints are addressed specifically, although many of the considerations discussed here will apply to performance limits for other safety system components. The first part of the paper deals with the approach that has been adopted for defining and establishing setpoint limits and tolerances. The remainder of the paper addresses operational

The Inherent Asymmetry of DNA Replication.

Science.gov (United States)

Snedeker, Jonathan; Wooten, Matthew; Chen, Xin

2017-10-06

Semiconservative DNA replication has provided an elegant solution to the fundamental problem of how life is able to proliferate in a way that allows cells, organisms, and populations to survive and replicate many times over. Somewhat lost, however, in our admiration for this mechanism is an appreciation for the asymmetries that occur in the process of DNA replication. As we discuss in this review, these asymmetries arise as a consequence of the structure of the DNA molecule and the enzymatic mechanism of DNA synthesis. Increasing evidence suggests that asymmetries in DNA replication are able to play a central role in the processes of adaptation and evolution by shaping the mutagenic landscape of cells. Additionally, in eukaryotes, recent work has demonstrated that the inherent asymmetries in DNA replication may play an important role in the process of chromatin replication. As chromatin plays an essential role in defining cell identity, asymmetries generated during the process of DNA replication may play critical roles in cell fate decisions related to patterning and development.

The representation of inherent properties.

Science.gov (United States)

Prasada, Sandeep

2014-10-01

Research on the representation of generic knowledge suggests that inherent properties can have either a principled or a causal connection to a kind. The type of connection determines whether the outcome of the storytelling process will include intuitions of inevitability and a normative dimension and whether it will ground causal explanations.

The development of a small inherently safe homogeneous reactor for the production of medical isotopes

Energy Technology Data Exchange (ETDEWEB)

Carlin, G.E.; Bonin, H.W., E-mail: george.carlin@rmc.ca [Royal Military College of Canada, Kingston, Ontario (Canada)

2013-07-01

The use of radioisotopes for various procedures in the health care industry has become one of the most important practices in medicine. New interest has been found in the use of liquid fueled nuclear reactors to produce these isotopes due to the ease of fuel processing and ability to efficiently use LEU as the fuel source. A version of this reactor is being developed at the Royal Military College of Canada to act as a successor to the SLOWPOKE-2 platform. The thermal hydraulic and transient characteristics of a 20 kWt version are being studied to verify inherent safety abilities. (author)

The safety case for a HLW repository in Opalinus clay: aims, methodology, first results

International Nuclear Information System (INIS)

Zuidema, Piet

2002-01-01

Piet Zuidema (Nagra, Switzerland) described the development of the safety case for a high level waste repository in Opalinus clay in which canisters would be placed in large vaults. The current phase of work was concerned with demonstrating the feasibility of the disposal concept. The Safety Case is taken to mean a set of arguments to support a statement that the proposed facility will meet relevant safety criteria and will include arguments giving the basis for confidence that those arguments are correct and properly taking account of uncertainties. The safety strategy was concerned both with the inherent robustness of the disposal concept and the adequacy of the assessment capability. As regards the former, the arguments being advanced were primarily qualitative. Key issues in terms of the documentation of the Safety Case were traceability and transparency of information, including how to ensure that key arguments did not become obscured because of the need to make available very large quantities of information

The safety of light water reactors

International Nuclear Information System (INIS)

Pershagen, B.

1986-04-01

The book describes the principles and practices of reactor safety as applied to the design, regulation and operation of both pressurized water reactors and boiling water reactors. The central part of the book is devoted to methods and results of safety analysis. Some significant events are described, notably the Three Mile Island accident. The book concludes with a chapter on the PIUS principle of inherent reactor safety as applied to the SECURE type of reactor developed in Sweden. (G.B.)

[Differences of inherent optical properties of inland lake water body in typical seasons].

Science.gov (United States)

Sun, De-Yong; Li, Yun-Mei; Wang, Qiao; Le, Cheng-Fen; Huang, Chang-Chun; Wang, Li-Zhen

2008-05-01

Inherent optical property is one of the important properties of water body, which lays the foundation for the establishment of water color analytical models. By using quantity filter technology (QFT) and BB9 backscattering meter, the absorption coefficients of chromophoric dissolved organic matter (CDOM) and total suspended matters (TSM) and the backscattering coefficient of TSM in the water body at Meiliang Bay of Taihu Lake were measured in summer and winter. Based on the spectral comparison of the absorption and backscattering coefficients, their differences between the two seasons were demonstrated, and the reasons that caused these differences were also explored in the context of their relations to the changes in water quality. Consequently, water environment condition could be revealed by using the inherent optical property. The relationship between the backscattering coefficient and the TSM concentration was established, which could provide supporting coefficients to the analytical models to be developed.

Safety demonstration tests of postulated solvent fire accidents in extraction process of a fuel reprocessing plant, (2)

International Nuclear Information System (INIS)

Tukamoto, Michio; Takada, Junichi; Koike, Tadao; Nishio, Gunji; Uno, Seiichiro; Kamoshida, Atsusi; Watanabe, Hironori; Hashimoto, Kazuichiro; Kitani, Susumu.

1992-03-01

Demonstration tests of hypothetical solvent fire in an extraction process of the reprocessing plant were carried out from 1984 to 1985 in JAERI, focusing on the confinement of radioactive materials during the fire by a large-scale fire facility (FFF) to evaluate the safety of air-ventilation system in the plant. Fire data from the demonstration test were obtained by focusing on fire behavior at cells and ducts in the ventilation system, smoke generation during the fire, transport and deposition of smoke containing simulated radioactive species in the ventilation system, confinement of radioactive materials, and integrity of HEPA filters by using the FFF simulating an air-ventilation system of the reference reprocessing plant in Japan. The present report is published in a series of the report Phase I (JAERI-M 91-145) of the demonstration test. Test results in the report will be used for the verification of a computer code FACE to evaluate the safety of postulated fire accidents in the reprocessing plant. (author)

A Mechanistic Reliability Assessment of RVACS and Metal Fuel Inherent Reactivity Feedbacks

Energy Technology Data Exchange (ETDEWEB)

Grabaskas, David; Brunett, Acacia J.; Passerini, Stefano; Grelle, Austin

2017-09-24

GE Hitachi Nuclear Energy (GEH) and Argonne National Laboratory (Argonne) participated in a two year collaboration to modernize and update the probabilistic risk assessment (PRA) for the PRISM sodium fast reactor. At a high level, the primary outcome of the project was the development of a next-generation PRA that is intended to enable risk-informed prioritization of safety- and reliability-focused research and development. A central Argonne task during this project was a reliability assessment of passive safety systems, which included the Reactor Vessel Auxiliary Cooling System (RVACS) and the inherent reactivity feedbacks of the metal fuel core. Both systems were examined utilizing a methodology derived from the Reliability Method for Passive Safety Functions (RMPS), with an emphasis on developing success criteria based on mechanistic system modeling while also maintaining consistency with the Fuel Damage Categories (FDCs) of the mechanistic source term assessment. This paper provides an overview of the reliability analyses of both systems, including highlights of the FMEAs, the construction of best-estimate models, uncertain parameter screening and propagation, and the quantification of system failure probability. In particular, special focus is given to the methodologies to perform the analysis of uncertainty propagation and the determination of the likelihood of violating FDC limits. Additionally, important lessons learned are also reviewed, such as optimal sampling methodologies for the discovery of low likelihood failure events and strategies for the combined treatment of aleatory and epistemic uncertainties.

Specific experiments carried out in Germany in order to demonstrate the safety of existing structures

International Nuclear Information System (INIS)

Krutzik, Norbert

2002-01-01

Specific experiments are carried out in Germany in order to demonstrate the safety of existing NPPs. HDR research program includes operational loads testing (pressure test, pressure and temperature test, thermal shock, fatigue); extreme loads (earthquake, aircraft crash, external explosion); internal emergency loads (blowdown, hydrogen combustion, fire, thermal shock, water hammer, condensation loads)

Technical and administrative approach for the West Valley Demonstration Project Safety Program

International Nuclear Information System (INIS)

Newsom, P.C.; Roberts, C.J.; Yuchien Yuan; Marchetti, S.

1987-06-01

The principal objective of the West Valley Demonstration Project (WVDP) is to vitrify the 2.2 million liters of high-level radioactive waste (HLW) stored at the Western New York Nuclear Service Center (WNYNSC). This simple statement of purpose, however, does not convey a sense of the complexity of the undertaking. The vitrification task is not only complex in and of itself, but requires a myriad of other activities to be accomplished on an intricate and fast paced schedule in order to support it. The West Valley Demonstration Project Act (P.L 96-368), U.S. Department of Energy Order DOE-5481.1A, Idaho Operations Office Order ID-5481.1 and standard nuclear industry practice all require that proposed systems and operations involving hazards not routinely encountered by the general public be analyzed to identify potential hazards and consequences, and to assure that reasonable measures are taken to eliminate, control, or mitigate these potential consequences. Virtually every substantive aspect of the WVDP involves hazards beyond those routinely encountered and accepted by the general public. In order to assure the safety of the public and the workers at the WVDP, a system of hazard identification, categorization, analysis and review has been established. In parallel with this system, a procedure for developing the minimum design specifications and quality assurance requirements has been developed for Project systems, components, and structures which play a role in the safety of a specific major facility or the overall Project. 29 refs., 3 figs., 6 tabs

Study on the KALIMER safety approach

International Nuclear Information System (INIS)

Kim, Eui Kwang; Han, Do Hee; Kim, Young Cheol.

1997-01-01

This study describes KALIMER's safety approach, how to establish the safety criteria and temperature limit, how to define safety evaluation events, and some safety research and development needs items. It is recommended that the KALIMER's approach to safety use seven levels of safety design and a defense-in-depth design approach with particular emphasis on inherent passive features. In order to establish as set DBEs for KALIMER safety evaluation, the procedure is explained how to define safety evaluation events. Final selection is to be determined later with the final establishment of design concepts. On the basis of preliminary studies and evaluation of the plant safety related areas, the KALIMER and PRISM have following three main difference that may require special research and development for KALIMER. (author). 7 refs., 6 tabs., 6 figs

Lessons learned in demonstration projects regarding operational safety during final disposal of vitrified waste and spent fuel

International Nuclear Information System (INIS)

Filbert, Wolfgang; Herold, Philipp

2015-01-01

The paper summarizes the lessons learned in demonstration projects regarding operational safety during the final disposal of vitrified waste and spent fuel. The three demonstration projects for the direct disposal of vitrified waste and spent fuel are described. The first two demonstration projects concern the shaft transport of heavy payloads of up to 85 t and the emplacement operations in the mine. The third demonstration project concerns the borehole emplacement operation. Finally, open issues for the next steps up to licensing of the emplacement and disposal systems are summarized.

Informing patients of risks inherent in treatment.

Science.gov (United States)

Griffith, Richard; Tengnah, Cassam

2009-11-01

Consent to treatment lies at the heart of autonomous decision making by patients who are entitled to make a free choice about whether to accept or refuse treatment. To help patients arrive at their decision district nurses must ensure that they give sufficient information about the nature and risks inherent in the treatment to allow an informed choice to be made. This article considers how much information regarding risks needs to be disclosed. It discusses how the law requires a different level of disclosure for patients who ask no questions about risks, those who make general enquiries about risks and those who ask specific questions about the risks inherent in treatment.

Passive Safety Features for Small Modular Reactors

International Nuclear Information System (INIS)

Ingersoll, Daniel T.

2010-01-01

The rapid growth in the size and complexity of commercial nuclear power plants in the 1970s spawned an interest in smaller, simpler designs that are inherently or intrinsically safe through the use of passive design features. Several designs were developed, but none were ever built, although some of their passive safety features were incorporated into large commercial plant designs that are being planned or built today. In recent years, several reactor vendors are actively redeveloping small modular reactor (SMR) designs with even greater use of passive features. Several designs incorporate the ultimate in passive safety they completely eliminate specific accident initiators from the design. Other design features help to reduce the likelihood of an accident or help to mitigate the accidents consequences, should one occur. While some passive safety features are common to most SMR designs, irrespective of the coolant technology, other features are specific to water, gas, or liquid-metal cooled SMR designs. The extensive use of passive safety features in SMRs promise to make these plants highly robust, protecting both the general public and the owner/investor. Once demonstrated, these plants should allow nuclear power to be used confidently for a broader range of customers and applications than will be possible with large plants alone.

Does the inherence heuristic take us to psychological essentialism?

Science.gov (United States)

Marmodoro, Anna; Murphy, Robin A; Baker, A G

2014-10-01

We argue that the claim that essence-based causal explanations emerge, hydra-like, from an inherence heuristic is incomplete. No plausible mechanism for the transition from concrete properties, or cues, to essences is provided. Moreover, the fundamental shotgun and storytelling mechanisms of the inherence heuristic are not clearly enough specified to distinguish them, developmentally, from associative or causal networks.

Demonstration of criticality safety for the modified TN-REG and TN-BRP transport/storage casks

International Nuclear Information System (INIS)

Parks, C.V.; Fox, P.B.

1989-01-01

An inability to model the structural performance of borated steel baskets under accident conditions forced the specially designed TN-BRP and TN-REG casks to be modified for half-loaded shipments. This paper discusses the approach used to demonstrate that the half-loaded casks would remain safely subcritical even if no credit were taken for the borated basket. Normal and accident configurations were analyzed with the KENO V.a code. The strategy conceived and the analyses performed to demonstrate an acceptable margin of safety are discussed. 5 refs., 3 figs., 2 tabs

PRISM [Power Reactor Inherently Safe Module] design concept enhances waste management

International Nuclear Information System (INIS)

Thompson, M.L.; Berglund, R.C.

1989-01-01

PRISM, a modular advanced liquid metal reactor (ALMR), has been designed conceptually by GE under the US Department of Energy sponsorship. The concept design and analyses have been primarily focused on passive safety and improved construction and operating costs. Significantly, the unique design of multiple modules and features of PRISM enhance waste management over conventional reactor systems. This paper provides an overview of PRISM of these enhancements. Inherent to the ALMR's, the sodium coolant precludes crud buildup on reactor surfaces and in components and waste for disposal. Preliminary evaluations indicate this fundamental feature results in factors of 2-4 less waste volume and 2-3 orders of magnitude less curies per megawatt-electric for ultimate disposal. For example, the tap designed for sodium cleanup is expected to be exchanged only once every thirty years. Also, inherent to ALMR's, burning waste actinides and selected fission products to preclude their accumulation and burial is very attractive. The hard neutron spectrum of ALMR burns the actinides efficiently and is not poisoned by the actinides and fission products. The modular design of PRISM components (and the fuel cycle equipment) permit replacement without expensive and potentially hazardous volume reduction. For example, the functional components of the reference electromagnetic pump and IHK can be removed intact for waste disposal. Although development of the reference metal fuel is not completed, it is estimated that (low-level) waste from recycle of the fuel will result in significantly less volume than would be generated by aqueous recycle of oxide fuel. 6 refs., 10 figs

Safety Analysis Of Actinide Recycled Fast Power Reactor

International Nuclear Information System (INIS)

Taufik, Mohammad

2001-01-01

Simulation for safety analysis of actinide recycled fast power reactor has been performed. The objective is to know reactor response about ULOF and ULOF and UTOP simultaneous accident. From parameter result such reactivity feedback, power, temperature, and cooled flow rate can conclusion that reactor have inherent safety system, which can back to new Equilibrium State

A real-life example of choosing an inherently safer process option

International Nuclear Information System (INIS)

Study, Karen

2007-01-01

While choosing an inherently safer alternative may seem straightforward, sometimes what seems to be the most obvious alternative may not provide the best risk reduction. The process designer must maintain a broad perspective to be able to recognize all potential hazards when evaluating design options. All aspects of operation such as start-up, shut-down, utility failure, as well as normal operation should be considered. Choosing the inherently safer option is best accomplished early in the option selection phase of a project; however, recycle back to the option selection phase may be needed if an option is not thoroughly evaluated early in the process. In this paper, a project to supply ammonia to a catalytic reactor will be reviewed. During the course of the project, an 'inherently safer' alternative was selected and later discarded due to issues uncovered during the detail design phase. The final option chosen will be compared to (1) the original design and (2) the initial 'inherently safer' alternative. The final option was inherently safer than both the original design and the initial 'inherently safer' alternative even though the design team initially believed that it would not be

Overview of fast reactor safety research and development in the USA

International Nuclear Information System (INIS)

Neuhold, R.J.; Avery, R.; Marchaterre, J.F.

1986-01-01

The liquid metal reactor (LMR) safety R and D program in the U.S. is presently focused on support of two modular innovative reactor concepts: PRISM - the General Electric Power Reactor Inherently Safe Module and SAFR - the Rockwell International Sodium Advanced Fast Reactor. These reactor plant concepts accommodate the use of either oxide fuel or the metal fuel which is under development in the Argonne National Laboratory (ANL) Integral Fast Reactor (IFR) program. Both concepts emphasize prevention of accidents through enhancement of inherent and passive safety characteristics. Enhancement of these characteristics is expected to be a major factor in establishing new and improved safety criteria and licensing arrangements with regulatory authorities for advanced reactors. Limited work is also continuing on the Large Scale Prototype Breeder (LSPB), a large pool plant design. Major elements of the current and restructured safety program are discussed. (author)

Inherent Risk or Risky Decision? Coach's Failure to Use Safety Device an Assumed Risk

Science.gov (United States)

Dodds, Mark A.; Bochicchio, Kristi Schoepfer

2013-01-01

The court examined whether a coach's failure to implement a safety device during pitching practice enhanced the risk to the athlete or resulted in a suboptimal playing condition, in the context of the assumption of risk doctrine.

Radioactive waste management in France: safety demonstration fundamentals.

Science.gov (United States)

Ouzounian, G; Voinis, S; Boissier, F

2012-01-01

The main challenge in development of the safety case for deep geological disposal is associated with the long periods of time over which high- and intermediate-level long-lived wastes remain hazardous. A wide range of events and processes may occur over hundreds of thousands of years. These events and processes are characterised by specific timescales. For example, the timescale for heat generation is much shorter than any geological timescale. Therefore, to reach a high level of reliability in the safety case, it is essential to have a thorough understanding of the sequence of events and processes likely to occur over the lifetime of the repository. It then becomes possible to assess the capability of the repository to fulfil its safety functions. However, due to the long periods of time and the complexity of the events and processes likely to occur, uncertainties related to all processes, data, and models need to be understood and addressed. Assessment is required over the lifetime of the radionuclides contained in the radioactive waste. Copyright © 2012. Published by Elsevier Ltd.

Regulatory review of safety cases and safety assessments - associated challenges

International Nuclear Information System (INIS)

Bennett, D.G.; Ben Belfadhel, M.; Metcalf, P.E.

2006-01-01

Regulatory reviews of safety cases and safety assessments are essential for credible decision making on the licensing or authorization of radioactive waste disposal facilities. Regulatory review also plays an important role in developing the safety case and in establishing stakeholders' confidence in the safety of the facility. Reviews of safety cases for radioactive waste disposal facilities need to be conducted by suitably qualified and experienced staff, following systematic and well planned review processes. Regulatory reviews should be sufficiently comprehensive in their coverage of issues potentially affecting the safety of the disposal system, and should assess the safety case against clearly established criteria. The conclusions drawn from a regulatory review, and the rationale for them should be reproducible and documented in a transparent and traceable way. Many challenges are faced when conducting regulatory reviews of safety cases. Some of these relate to issues of project and programme management, and resources, while others derive from the inherent difficulties of assessing the potential long term future behaviour of engineered and environmental systems. The paper describes approaches to the conduct of regulatory reviews and discusses some of the challenges faced. (author)

Perspectives on understanding and verifying the safety terrain of modular high temperature gas-cooled reactors

Energy Technology Data Exchange (ETDEWEB)

Carlson, Donald E., E-mail: donald@carlsonperin.net [11221 Empire Lane, Rockville, MD 20852 (United States); Ball, Sydney J., E-mail: beckysyd@comcast.net [100 Greywood Place, Oak Ridge, TN 37830 (United States)

2016-09-15

The passive safety characteristics of modular high temperature gas-cooled reactors (HTGRs) are conceptually well known and are largely supported by insights from past and ongoing research. This paper offers perspectives on selected issues in areas where further analysis and testing achievable within existing research and demonstration programs could help address residual uncertainties and better support the analysis of safety performance and the regulatory assessment of defense in depth. Areas considered include the evaluation of normal and anomalous core operating conditions and the analysis of accidents involving loss of forced cooling, coolant depressurization, air ingress, moisture ingress, and reactivity events. In addition to discussing associated uncertainties and potential measures to address them, this paper also proposes supplemental “safety terrain” studies that would use realistic assessments of postulated extreme event sequences to establish a more comprehensive understanding of the inherent behaviors and ultimate safety capabilities of modular HTGRs.

The safety feature of hydraulic driving system of control rod for 200 MW nuclear heating reactor

International Nuclear Information System (INIS)

Chi Zongbo; Wu Yuanqiang

1997-01-01

The hydraulic driving system of control rod is used as control rod drive mechanism in 200 MW nuclear heating reactor. Design of this system is based on passive system, integrating drive and guide of control rod. The author analyzes the inherent safety and the design safety of this system, with mechanism of control rod not ejecting when the pressure of pressure vessel is lost, and calculating result of core not exposing when the amount of coolant is drained by broken pipe. The results indicate that this system has good safety feature, and assures reactor safety under any accident conditions, providing important technology support for 200 MW nuclear heating reactor with inherent safety feature

Nuclear safety: operational aspects. 1. Demonstrating the Link Between Safety Culture and Competitiveness

International Nuclear Information System (INIS)

Chakoff, H. Elliot; Slider, James E.

2001-01-01

More than 20 years ago, we demonstrated a methodology for distinguishing the safety cultures of nuclear power plants. Using the content of licensee event reports, the methodology led to the identification of metrics that could be used to partition 12 pilot plants into better and poorer performers. The partitioning was validated by U.S. Nuclear Regulatory Commission (NRC) experts and shown to be statistically significant at the 95% level of confidence. We wanted to know if the passage of time had validated the differences in performance identified by the original methodology. Our follow-up confirmed the validity of the methodology and also revealed an order of magnitude difference in the long-term survival probability of the 12 pilot plants. The lessons learned from these studies could help plant owners improve safety culture and competitiveness in today's Darwinian marketplace. The original study sought to determine if it was possible to distinguish between better- and poorer-performing plants. The study found it was possible and developed a methodology for doing so. Key breakthroughs included the following: 1. recognizing that safety performance is a stochastic process; thus, performance data must be evaluated using appropriate methods; 2. developing a model for interpreting and transforming raw data into a root-cause domain; 3. maintaining a rigorous model design logic and selecting analytical tools and procedures consistent with that logic; 4. determining that the number of low significance events is an unreliable measure of performance; 5. recognizing that it is the relationship between events that is crucial to understanding performance and risk. Metrics were developed using a test population of 12 plants selected and grouped as 'good' or 'poor' performers by NRC's most senior inspectors. The test population included three plants that had significant events in a 2-yr period and nine that had none. The metrics validated differences in performance hypothesized

Safety and operating experience at EBR-II: lessons for the future

International Nuclear Information System (INIS)

Sackett, J.I.; Golden, G.H.

1981-01-01

EBR-II is a small LMFBR power plant that has performed safely and reliably for 16 years. Much has been learned from operating it to facilitate the design, licensing, and operation of large commercial LMFBR power plants in the US. EBR-II has been found relatively easy to keep in conformity with evolving safety requirements, largely because of inherent safety features of the plant. Such features reduce dependence on active safety systems to protect against accidents. EBR-II has experienced a number of plant-transient incidents, some planned, others inadvertent; none has resulted in any significant plant damage. The operating experience with EBR-II has led to the formulation of an Operational Reliability Test Program (ORTP), aimed at showing inherently safe performance of fuel and plant systems

Recommended safety procedures for the selection and use of demonstration-type gas discharge devices in schools

International Nuclear Information System (INIS)

1979-01-01

A 1972 survey of 30 Ottawa secondary schools revealed a total of 347 actual or potential X-ray sources available in these schools. More than half of these sources were gas discharge tubes. Some gas discharge tubes, in particular the cold cathode type, can emit X-rays at significantly high levels. Unless such tubes are used carefully, and with regard for good radiation safety practices, they can result in exposures to students that are in excess of the maximum levels recommended by the International Commission on Radiological Protection. Several cases of the recommended dose being exceeded were found in the classes surveyed. This document has been prepared to assist science teachers and others using demonstration-type gas discharge devices to select and use such devices so as to present negligible risk to themselves and students. Useful information on safety procedures to be followed when performing demonstrations or experiments is included. (J.T.A.)

Bavarian liquid hydrogen bus demonstration project - safety, licensing and acceptability aspects

Energy Technology Data Exchange (ETDEWEB)

Wurster, R.; Knorr, H.; Pruemm, W.

1999-07-01

A regular 12 m city bus of the MAN SL 202 type with an internal combustion engine adapted to hydrogen operation and auxiliary gasoline operation was demonstrated in the Bavarian cities of Erlangen and Munich between April 1996 and August 1998. Three bus operators, Erlanger Stadtwerke, Stadtwerke Muenchen and Autobus Oberbayern were testing the bus in three different operating schemes. In order to be able to perform this worldwide first public demonstration of a liquid hydrogen (LH{sub 2}) city bus in regular service, several requirements with respect to safety, licensing, training and acceptability had to be fulfilled. These activities were focusing mainly on the hydrogen specific issues such as (a) integration of onboard LH{sub 2} storage vessels, piping and instrumentation, (b) implementation of storage and refueling infrastructure in the operators' yards, (c) adaptation of the maintenance garages, (d) training of operating and maintenance personnel. During phase II of the demonstration activity a poll was performed on passengers traveling onboard the hydrogen-powered city bus in order to determined the level of acceptance among the users of the bus. The bus was designed and manufactured by MAN Nutzfahrzeuge Aktiengesellschaft. The cryogenic fuel storage and the refueling equipment were designed and manufactured by Linde AG. The realization of the hardware was financially supported by the European Commission (EC) within the Euro-Quebec Hydro-Hydrogen Pilot Project. The demonstration phase was financially supported by EC and the Bavarian State Government. Ludwig-Boelkow-Systemtechnik performed project monitoring for both funding organizations. The presentation will summarize the most important results of this demonstration phase and will address the measures undertaken in order to get the bus, the refueling infrastructure and the maintenance and operating procedures approved by the relevant authorities.

Safety demonstration tests of air-ventilation system for the postulated explosive burning in a cell of fuel-reprocessing plant

International Nuclear Information System (INIS)

Takada, Junichi; Suzuki, Motoe; Tukamoto, Michio; Koike, Tadao; Nishio, Gunji

1995-03-01

Safety demonstration tests of an explosive burning in a cell in the reprocessing plant has been carried out in JAERI under the auspices of the Science and Technology Agency, to evaluate the safety of an air-ventilation system during the hypothetical explosion. The postulated explosive burning of organic solvent mixed with nitric acid was simulated by solid explosives. The demonstration test was performed using an industrial scale experimental facility simulating to the ventilation system of the large scale reprocessing plant in JAPAN. Propagations of pressure, temperature, and gas velocity through cells and ducts in the ventilation system were measured during the explosive burning under deflagration. Experimental data in this report can be used to evaluate the transport phenomena of radioactive materials in the ventilation system during the explosion, and also to verify computer code CELVA for the safety analysis of ventilation system in the event of explosion accidents. (author)

ELFR: The European Lead Fast Reactor. Design, Safety Approach and Safety Characteristics

International Nuclear Information System (INIS)

Alemberti, Alessandro

2012-01-01

• In the framework of the LEADER project, the safety approach for a Lead cooled fast reactor has been defined and, in particular, all the possible challenges to the main safety functions and their mechanisms have been specified, in order to better define the needed provisions. • On the basis of the above and taking into account the results of the safety analyses performed during previous project (ELSY), a reference configuration of the ELFR plant has been consolidated, by improving and updating the plant design features. In particular, the emerged safety concerns have been analyzed in the LEADER project and a new set of design options and safety provisions have been proposed. • The combination of favourable Lead coolant inherent characteristics and plant design features, specifically developed to face identified challenges, resulted in a very robust and forgiving design, even in very extreme conditions, as a Fukushima-like scenario

[Exploration and demonstration study on drug combination from clinical real world].

Science.gov (United States)

Xie, Yan-ming; Wang, Lian-xin; Wang, Yong-yan

2014-09-01

Drug combination is extensive in the clinical real world,which is an important part and the inherent requirements of the post-marketing evaluation of traditional Chinese medicine (TCM). The key issues and technology include multi-domain and multi-disciplinary such as the rationality, efficacy and safety evaluation of combination drug starting from clinical real world, study on component in vivo and mechanism of combination drug, the risk/benefit assessment and cost-benefit evaluation of combination drug and so on. The topic has been studied as clinical demonstration on combination therapy of variety of diseases such as coronary heart disease, stroke, insomnia, depression, hepatitis, herpes zoster, psoriasis and ectopic pregnancy. Meanwhile, multi-disciplinary dynamic innovation alliance of clinical drug combination has been presented, which can promote the academic development and improving service ability and level of TCM.

Can a robot improve mine safety?

CSIR Research Space (South Africa)

Green, JJ

2010-09-01

Full Text Available Safety in mines is of paramount importance, especially in the labour intensive operations of South Africa, where upward of 300 000 people are employed on a daily basis in an environment that is inherently dangerous. On average approximately 50...

Thermally responsive polymer electrolytes for inherently safe electrochemical energy storage

Science.gov (United States)

Kelly, Jesse C.

Electrochemical double layer capacitors (EDLCs), supercapacitors and Li-ion batteries have emerged as premier candidates to meet the rising demands in energy storage; however, such systems are limited by thermal hazards, thermal runaway, fires and explosions, all of which become increasingly more dangerous in large-format devices. To prevent such scenarios, thermally-responsive polymer electrolytes (RPEs) that alter properties in electrochemical energy storage devices were designed and tested. These RPEs will be used to limit or halt device operation when temperatures increase beyond a predetermined threshold, therefore limiting further heating. The development of these responsive systems will offer an inherent safety mechanism in electrochemical energy storage devices, while preserving the performance, lifetimes, and versatility that large-format systems require. Initial work focused on the development of a model system that demonstrated the concept of RPEs in an electrochemical device. Aqueous electrolyte solutions of polymers exhibiting properties that change in response to temperature were developed for applications in EDLCs and supercapacitors. These "smart materials" provide a means to control electrochemical systems where polymer phase separation at high temperatures affects electrolyte properties and inhibits device performance. Aqueous RPEs were synthesized using N-isopropylacrylamide, which governs the thermal properties, and fractions of acrylic acid or vinyl sulfonic acids, which provide ions to the solution. The molecular properties of these aqueous RPEs, specifically the ionic composition, were shown to influence the temperature-dependent electrolyte properties and the extent to which these electrolytes control the energy storage characteristics of a supercapacitor device. Materials with high ionic content provided the highest room temperature conductivity and electrochemical activity; however, RPEs with low ionic content provided the highest "on

The inherent politics of quality in public park management

DEFF Research Database (Denmark)

Lindholst, Andrej Christian; Konijnendijk, Cecil Cornelis; Fors, Hanna

2012-01-01

In this paper, we highlight and illustrate the inherent politics embedded in “quality” as a concept for managing public parks. Reflecting more generic quality concepts, contemporary quality models in park management include concepts for both operational, strategic and stakeholder management as well...... managing the park organisation itself. However, quality concepts and their application through various management models include as well as exclude the access, values and worldviews of particular interests. In this way, any particular quality concept and model embeds its own politics by inherent...... allocations of ‘who gets what, when and how’. We illustrate the inherent politics by providing a case study of a widely adopted quality model for operational management that has been adopted and implemented in Denmark as part of new public management reforms. In perspective, other quality concepts and models...

Inherent work suit buoyancy distribution: effects on lifejacket self-righting performance.

Science.gov (United States)

Barwood, Martin J; Long, Geoffrey M; Lunt, Heather; Tipton, Michael J

2014-09-01

Accidental immersion in cold water is an occupational risk. Work suits and life jackets (LJ) should work effectively in combination to keep the airway clear of the water (freeboard) and enable self-righting. We hypothesized that inherent buoyancy, in the suit or LJ, would be beneficial for enabling freeboard, but its distribution may influence LJ self-righting. Six participants consented to complete nine immersions. Suits and LJ tested were: flotation suit (FLOAT; 85 N inherent buoyancy); oilskins 1 (OS-1) and 2 (OS-2), both with no inherent buoyancy; LJs (inherent buoyancy/buoyancy after inflation/total buoyancy), LJ-1 50/150/200 N, LJ-2 0/290/290 N, LJ-3 80/190/270 N. Once dressed, the subject entered an immersion pool where uninflated freeboard, self-righting performance, and inflated freeboard were measured. Data were compared using Friedman's test to the 0.05 alpha level. All suits and LJs enabled uninflated and inflated freeboard, but differences were seen between the suits and LJs. Self-righting was achieved on 43 of 54 occasions, irrespective of suit or LJ. On all occasions that self-righting was not achieved, this occurred in an LJ that included inherent buoyancy (11/54 occasions). Of these 11 failures, 8 occurred (73% of occasions) when the FLOAT suit was being worn. LJs that included inherent buoyancy, that are certified as effective on their own, worked less effectively from the perspective of self-righting in combination with a work suit that also included inherent buoyancy. Equipment that is approved for use in the workplace should be tested in combination to ensure adequate performance in an emergency scenario.

New perspectives on reactor safety

International Nuclear Information System (INIS)

Avery, R.

1986-01-01

Over the past few years a number of changes and new perspectives have come about in our approach to reactor safety. These changes have occurred over a period of time extending from as long ago as 1975, when WASH-1400 came out representing the first major application of probabilistic risk analysis (PRA) to US reactor plants. The period of change has extended from that time to the present, and includes new areas of focus such as safety goals, source term studies, and severe accident policy statement and approaches, including the IDCOR Program. It has also included a greatly increased interest in inherent safety. These areas are discussed in this paper

Solid Biomass Climate Change Interventions Examined in a Context of Inherent Safety, Media Shifting and Emerging Risks

DEFF Research Database (Denmark)

Hedlund, Frank Huess; Astad, John

2014-01-01

benefits over safety concerns seems to run deep and not confined to the realm of only solid biomass. Danish environmental ambitions are very high and the costs to society of introducing solid biomass fuels are breathtaking. In this setting, the general failure to address safety risk s appears particularly...

7T: Physics, safety, and potential clinical applications.

Science.gov (United States)

Kraff, Oliver; Quick, Harald H

2017-12-01

With more than 60 installed magnetic resonance imaging (MRI) systems worldwide operating at a magnetic field strength of 7T or higher, ultrahigh-field (UHF) MRI has been established as a platform for clinically oriented research in recent years. Profound technical and methodological developments have helped overcome the inherent physical challenges of UHF radiofrequency (RF) signal homogenization in the human body. The ongoing development of dedicated RF coil arrays was pivotal in realizing UHF body MRI, beyond mere brain imaging applications. Another precondition to clinical application of 7T MRI is the safety testing of implants and the establishment of safety concepts. Against this backdrop, 7T MRI and MR spectroscopy (MRS) recently have demonstrated capabilities and potentials for clinical diagnostics in a variety of studies. This article provides an overview of the immanent physical challenges of 7T UHF MRI and discusses recent technical solutions and safety concepts. Furthermore, recent clinically oriented studies are highlighted that span a broad application spectrum from 7T UHF brain to body MRI. 4 Technical Efficacy: Stage 1 J. Magn. Reson. Imaging 2017;46:1573-1589. © 2017 International Society for Magnetic Resonance in Medicine.

Safety physics inter-comparison of advanced concepts of critical reactors and ADS

International Nuclear Information System (INIS)

Slessarev, I.

2001-01-01

Enhanced safety based on the principle of the natural ''self-defence'' is one of the most desirable features of innovative nuclear systems (critical or sub-critical) regarding both TRU transmutation and ''clean'' energy producer concepts. For the evaluation of the ''self-defence'' domain, the method of the asymptotic reactivity balance has been generalised. The promising option of Hybrids systems (that use a symbiosis of fission and spallation in sub-critical cores) which could benefit the advantages of both Accelerated Driven Systems of the traditional type and regular critical systems, has been advocated. General features of Hybrid dynamics have been presented and analysed. It was demonstrated that an external neutron source of Hybrids can expand the inherent safety potential significantly. This analysis has been applied to assess the safety physics potential of innovative concepts for prospective nuclear power both for energy producers and for transmutation. It has been found, that safety enhancement goal defines a choice of sub-criticality of Hybrids. As for energy producers with Th-fuel cycle, a significant sub-criticality level is required due to a necessity of an improvement of neutronics together with safety enhancement task. (author)

Inherent Reward and Risk (Part I): Towards a Universal Paradigm for Investment Analysis

OpenAIRE

Liang Zou

2000-01-01

In this paper, a new paradigm is developed for analyzinginvestment strategies and pricing financial assets. This paradigmassumes that any investment strategy has its own “inherent reward”and “inherent risk” that can be judged with common sense. Ijustify axiomatically the existence and uniqueness (ratio scale)of inherent reward (U) and inherent risk (D) that could beregarded as universal measures of reward and risk for any giveninvestment strategy. Incorporating the notion of “inherentefficien...

Twenty years of improvements in LWR safety

International Nuclear Information System (INIS)

Franks, S. III; Mulkey, J.P.; Moonka, A.

1996-01-01

Substantial improvements have been made in the safety of light-water reactors in the US during the past two decades, making currently operating reactors safer than ever before. Safety improvements have resulted both from regulatory and operational changes and from new knowledge and technology. The US Nuclear Regulatory Commission, the US Department of Energy, and the American nuclear power industry have worked together and with the international community to enhance the safety of existing plants and to incorporate lessons learned from prior operation into designs for a new generation of advanced, inherently safer reactors

Analysis of solutions for passively activated safety shutdown devices for SFR

International Nuclear Information System (INIS)

Burgazzi, Luciano

2013-01-01

Highlights: • Innovative systems for emergency shut down of fast reactors are proposed. • The concepts of inherent and passive safety are put forward. • The relative analysis in terms of safety and reliability is presented. • A comparative assessment among the concepts is performed. • Path forward is tracked. -- Abstract: In order to enhance the inherent safety of fast reactors, innovative reactivity control systems have been proposed for intrinsic ultimate shut-down instead of conventional scram rods, to cope with the potential consequences of severe unprotected transient accidents, such as an energetic core disruptive accident, as in case of sodium fast reactors. The passive shut-down systems are designed to shut-down system only by inherent passive reactivity feedback mechanism, under unprotected accident conditions, implying failure of reactor protection system. They are conceived to be self-actuated without any signal elaboration, since the actuation of the system is triggered by the effects induced by the transient like material dilatation, in case of overheating of the coolant for instance, according to fast reactor design to meet the safety requirements. This article looks at different special shutdown systems specifically engineered for prevention of severe accidents, to be implemented on fast reactors, with main focus on the investigation of the performance of the self-actuated shutdown systems in sodium fast reactors

Finite-time and fixed-time leader-following consensus for multi-agent systems with discontinuous inherent dynamics

Science.gov (United States)

Ning, Boda; Jin, Jiong; Zheng, Jinchuan; Man, Zhihong

2018-06-01

This paper is concerned with finite-time and fixed-time consensus of multi-agent systems in a leader-following framework. Different from conventional leader-following tracking approaches where inherent dynamics satisfying the Lipschitz continuous condition is required, a more generalised case is investigated: discontinuous inherent dynamics. By nonsmooth techniques, a nonlinear protocol is first proposed to achieve the finite-time leader-following consensus. Then, based on fixed-time stability strategies, the fixed-time leader-following consensus problem is solved. An upper bound of settling time is obtained by using a new protocol, and such a bound is independent of initial states, thereby providing additional options for designers in practical scenarios where initial conditions are unavailable. Finally, numerical simulations are provided to demonstrate the effectiveness of the theoretical results.

International intercomparison and harmonization projects for demonstrating the safety of radioactive waste management, decommissioning and radioactive waste disposal

International Nuclear Information System (INIS)

Metcalf, Phil; O'Donnell, Patricio; Jova Sed, Luis; Batandjieva, Borislava; Rowat, John; Kinker, Monica

2008-01-01

Full text: The Joint Convention on the safety of spent fuel management and the safety of radioactive waste management and the international safety standards on radioactive waste management, decommissioning and radioactive waste disposal call for assessment and demonstration of the safety of facilities and activities; during siting, design and construction prior to operation, periodically during operation and at the end of lifetime or upon closure of a waste disposal facility. In addition, more recent revisions of the international safety standards require the development of a safety case for such facilities and activities, documentation presenting all the arguments supporting the safety of the facilities and activities covering site and engineering features, quantitative safety assessment and management systems. Guidance on meeting these safety requirements also indicates the need for a graded approach to safety assessment, with the extent and complexity of the assessment being proportional to the complexity of the activity or facility, and its propensity for radiation hazard. Safety assessment approaches and methodologies have evolved over several decades and international interest in these developments has been considerable as they can be complex and often subjective, which has led to international projects being established aimed at harmonization. The IAEA has sponsored a number of such initiatives, particularly in the area of disposal facility safety, but more recently in the areas of pre disposal waste management and decommissioning, including projects known as ISAM, ASAM, SADRWMS and DeSa. The projects have a number of common aspects including development of standardized methodological approaches, application on test cases and assessment review; they also have activity and facility specific elements. The paper presents an overview of the projects, the outcomes from the projects to date and their future direction aimed very much at practical application of

Risk management and safety

International Nuclear Information System (INIS)

Niehaus, F.; Novegno, A.

1985-01-01

Risk assessment, including probabilistic analyses, has made great progress over the past decade. In spite of the inherent uncertainties it has now become possible to utilize methods and results for decision making at various levels. This paper will, therefore, review risk management in industrial installations, risk management for energy safety policy and prospects of risk management in highly industrialized areas. (orig.) [de

Laboratory safety handbook

Science.gov (United States)

Skinner, E.L.; Watterson, C.A.; Chemerys, J.C.

1983-01-01

Safety, defined as 'freedom from danger, risk, or injury,' is difficult to achieve in a laboratory environment. Inherent dangers, associated with water analysis and research laboratories where hazardous samples, materials, and equipment are used, must be minimized to protect workers, buildings, and equipment. Managers, supervisors, analysts, and laboratory support personnel each have specific responsibilities to reduce hazards by maintaining a safe work environment. General rules of conduct and safety practices that involve personal protection, laboratory practices, chemical handling, compressed gases handling, use of equipment, and overall security must be practiced by everyone at all levels. Routine and extensive inspections of all laboratories must be made regularly by qualified people. Personnel should be trained thoroughly and repetitively. Special hazards that may involve exposure to carcinogens, cryogenics, or radiation must be given special attention, and specific rules and operational procedures must be established to deal with them. Safety data, reference materials, and texts must be kept available if prudent safety is to be practiced and accidents prevented or minimized.

Safety demonstration analyses at JAERI for severe accident during overland transport of fresh nuclear fuel

International Nuclear Information System (INIS)

Nomura, Yasushi; Kitao, Kohichi; Karasawa, Kiyonori; Yamada, Kenji; Takahashi, Satoshi; Watanabe, Kohji; Okuno, Hiroshi; Miyoshi, Yoshinori

2005-01-01

It is expected in the near future that more and more fresh nuclear fuel will be transported in a variety of transport packages to cope with increasing demand from nuclear fuel cycle facilities. Accordingly, safety demonstration analyses are planned and conducted at JAERI under contract with the Ministry of Economy, Trade and Industry of Japan. These analyses are conducted in a four year plan from 2001 to 2004 to verify integrity of packaging against leakage of radioactive material in the case of a severe accident postulated to occur during transportation, for the purpose of gaining acceptance of such nuclear fuel activities. In order to create the accident scenarios, actual transportation routes were surveyed, accident or incident records were tracked, international radioactive material transport regulations such as IAEA rules were investigated and thus, accident conditions leading to mechanical damages and thermal failure were determined to characterize the scenarios. As a result, the worst-case conditions of run-off-the-road accidents were set up to define the impact against a concrete or asphalt surface. For fire accident scenarios to be set up, collisions were assumed to occur with an oil tanker carrying lots of inflammable material in open air, or with a commonly used two-ton-truck inside a tunnel without ventilation. Then the cask models were determined for these safety demonstration analyses to represent those commonly used for fresh nuclear fuel transported throughout Japan. Following the postulated accident scenarios, the mechanical damages were analyzed by using the general-purpose finite element code LS-DYNA with three-dimensional elements. It was found that leak tightness of the package be maintained even in the severe impact scenario. Then the thermal safety was analyzed by using the general-purpose finite element code ABAOUS with three-dimensional elements to describe cask geometry. As a result of the thermal analyses, the integrity of the containment

Auxetic foam for snowsport safety devices

OpenAIRE

Allen, Tom; Duncan, Olly; Foster, Leon; Senior, Terry; Zampieri, Davide; Edeh, Victor; Alderson, Andrew

2017-01-01

Skiing and snowboarding are popular snow-sports with inherent risk of injury. There is potential to reduce the prevalence of injuries by improving and implementing snow-sport safety devices with the application of advanced materials. This paper investigates the application of auxetic foam to snow-sport safety devices. Composite pads - consisting of foam covered with a semi-rigid shell - were investigated as a simple model of body armour and a large 70 x 355 x 355 mm auxetic foam sample was fa...

Product Engineering Class in the Software Safety Risk Taxonomy for Building Safety-Critical Systems

Science.gov (United States)

Hill, Janice; Victor, Daniel

2008-01-01

When software safety requirements are imposed on legacy safety-critical systems, retrospective safety cases need to be formulated as part of recertifying the systems for further use and risks must be documented and managed to give confidence for reusing the systems. The SEJ Software Development Risk Taxonomy [4] focuses on general software development issues. It does not, however, cover all the safety risks. The Software Safety Risk Taxonomy [8] was developed which provides a construct for eliciting and categorizing software safety risks in a straightforward manner. In this paper, we present extended work on the taxonomy for safety that incorporates the additional issues inherent in the development and maintenance of safety-critical systems with software. An instrument called a Software Safety Risk Taxonomy Based Questionnaire (TBQ) is generated containing questions addressing each safety attribute in the Software Safety Risk Taxonomy. Software safety risks are surfaced using the new TBQ and then analyzed. In this paper we give the definitions for the specialized Product Engineering Class within the Software Safety Risk Taxonomy. At the end of the paper, we present the tool known as the 'Legacy Systems Risk Database Tool' that is used to collect and analyze the data required to show traceability to a particular safety standard

Cocaine Hydrolase Gene Transfer Demonstrates Cardiac Safety and Efficacy against Cocaine-Induced QT Prolongation in Mice

OpenAIRE

Murthy, Vishakantha; Reyes, Santiago; Geng, Liyi; Gao, Yang; Brimijoin, Stephen

2016-01-01

Cocaine addiction is associated with devastating medical consequences, including cardiotoxicity and risk-conferring prolongation of the QT interval. Viral gene transfer of cocaine hydrolase engineered from butyrylcholinesterase offers therapeutic promise for treatment-seeking drug users. Although previous preclinical studies have demonstrated benefits of this strategy without signs of toxicity, the specific cardiac safety and efficacy of engineered butyrylcholinesterase viral delivery remains...

78 FR 17140 - Upholstered Furniture Fire Safety Technology; Meeting and Request for Comments

Science.gov (United States)

2013-03-20

... retardant (FR) chemicals, specialty fibers/fabrics without FR chemicals, inherently fire resistant materials... Furniture Fire Safety Technology; Meeting and Request for Comments AGENCY: Consumer Product Safety... Commission (CPSC, Commission, or we) is announcing its intent to hold a meeting on upholstered furniture fire...

Environment, safety and health, management and organization compliance assessment, West Valley Demonstration Program, West Valley, New York

International Nuclear Information System (INIS)

1989-08-01

An Environment, Safety and Health ''Tiger Team'' Assessment was conducted at the West Valley Demonstration Project. The Tiger Team was chartered to conduct an onsite, independent assessment of WVDP's environment, safety and health (ES ampersand H) programs to assure compliance with applicable Federal and State laws, regulations, and standards, and Department of Energy Orders. The objective is to provide to the Secretary of Energy the following information: current ES ampersand H compliance status of each facility; specific noncompliance items; ''root causes'' for noncompliance items; evaluation of the adequacy of ES ampersand H organization and resources (DOE and contractor) and needed modifications; and where warranted, recommendations for addressing identified problem areas

IEEE standard for design qualification of safety systems equipment used in nuclear power generating stations

International Nuclear Information System (INIS)

Anon.

1980-01-01

This standard is written to serve as a general standard for qualification of all types of safety systems equipment, mechanical and instrumentation as well as electrical. It also establishes principles and procedures to be followed in preparing specific safety systems equipment standards. Guidance for qualifying specific safety systems equipment may be found in various specific equipment qualification standards that are now available or are being prepared. It is required that safety systems equipment in nuclear power generating stations meet or exceed its performance requirements throughout its installed life. This is accomplished by a disciplined program of design qualification and quality assurance of design, production, installation, maintenance and surveillance. This standard is for the design qualification section of the program only. Design qualification is intended to demonstrate the capability of the equipment design to perform its safety function(s) over the expected range of normal, abnormal, design basis event, post design basis event, and in-service test conditions. Inherent to design qualification is the requirement for demonstration, within limitations afforded by established technical state-of-the-art, that in-service aging throughout the qualified life established for the equipment will not degrade safety systems equipment from its original design condition to the point where it cannot perform its required safety function(s), upon demand. The above requirement reflects the primary role of design qualification to provide reasonable assurance that design- and age-related common failure modes will not occur during performance of safety function(s) under postulated service conditions

Response to 'Audiences, rationales and quantitative measure for demonstrations of nuclear safety and licensing by tests'

Energy Technology Data Exchange (ETDEWEB)

Taylor, J J [Electric Power Research Institute, Palo Alto, CA (United States)

1990-07-01

There are key overriding issues which are independent of the specific nature of the nuclear system itself which require concentrated attention to assure public safety and reliable, economic operation: - the need to keep the risk of external events to an acceptable level for all reactor systems; - the need to assure highly reliable operation of all elements of the system, many of which are the same regardless of what the nuclear system is composed of; - the importance of human proficiency in operating this total complex in a highly reliable manner. Nuclear system-specific demonstrations of public safety, although valuable, will not accomplish this and will not convince the public that there is zero risk. The very claim that a nuclear system or for that matter any big industrial complex, poses zero public risk raises a credibility gap with the public and is, therefore, counterproductive. So, we must take the dull, detailed technical steps to address the challenge: - define the minimal risk and accept that there is no zero risk; - demonstrate the achievement of that risk by detailed testing, conformance to standards and regulation, and trouble-free operation.

Response to 'Audiences, rationales and quantitative measure for demonstrations of nuclear safety and licensing by tests'

International Nuclear Information System (INIS)

Taylor, J.J.

1990-01-01

There are key overriding issues which are independent of the specific nature of the nuclear system itself which require concentrated attention to assure public safety and reliable, economic operation: - the need to keep the risk of external events to an acceptable level for all reactor systems; - the need to assure highly reliable operation of all elements of the system, many of which are the same regardless of what the nuclear system is composed of; - the importance of human proficiency in operating this total complex in a highly reliable manner. Nuclear system-specific demonstrations of public safety, although valuable, will not accomplish this and will not convince the public that there is zero risk. The very claim that a nuclear system or for that matter any big industrial complex, poses zero public risk raises a credibility gap with the public and is, therefore, counterproductive. So, we must take the dull, detailed technical steps to address the challenge: - define the minimal risk and accept that there is no zero risk; - demonstrate the achievement of that risk by detailed testing, conformance to standards and regulation, and trouble-free operation

Safety related terms for advanced nuclear plants

International Nuclear Information System (INIS)

1995-12-01

The terms considered in this document are in widespread current use without a universal consensus as to their meaning. Other safety related terms are already defined in national or international codes and standards as well as in IAEA's Nuclear Safety Standards Series. Most of the terms in those codes and standards have been defined and used for regulatory purposes, generally for application to present reactor designs. There is no intention to duplicate the description of such regulatory terms here, but only to clarify the terms used for advanced nuclear plants. The following terms are described in this paper: Inherent safety characteristics, passive component, active component, passive systems, active system, fail-safe, grace period, foolproof, fault-/error-tolerant, simplified safety system, transparent safety

Safety related terms for advanced nuclear plants

International Nuclear Information System (INIS)

1991-09-01

The terms considered in this document are in widespread current use without a universal consensus as to their meaning. Other safety related terms are already defined in national or international codes and standards as well as in IAEA's Nuclear Safety Standards Series. Most of the terms in those codes and standards have been defined and used for regulatory purposes, generally for application to present reactor designs. There is no intention to duplicate the description of such regulatory terms here, but only to clarify the terms used for advanced nuclear plants. The following terms are described in this paper: Inherent safety characteristics, passive component, active component, passive systems, active system, fail-safe, grace period, foolproof, fault-/error-tolerant, simplified safety system, transparent safety

Demonstration testing and evaluation of in situ soil heating. Health and safety plan (Revision 2)

Energy Technology Data Exchange (ETDEWEB)

Dev, H.

1994-12-28

This document is the Health and Safety Plan (HASP) for the demonstration of IITRI`s EM Treatment Technology. In this process, soil is heated in situ by means of electrical energy for the removal of hazardous organic contaminants. This process will be demonstrated on a small plot of contaminated soil located in the Pit Area of Classified Burial Ground K-1070-D, K-25 Site, Oak Ridge, TN. The purpose of the demonstration is to remove organic contaminants present in the soil by heating to a temperature range of 85{degrees} to 95{degrees}C. The soil will be heated in situ by applying 60-Hz AC power to an array of electrodes placed in boreholes drilled through the soil. In this section a brief description of the process is given along with a description of the site and a listing of the contaminants found in the area.

Demonstration testing and evaluation of in situ soil heating. Health and safety plan (Revision 2)

International Nuclear Information System (INIS)

Dev, H.

1994-01-01

This document is the Health and Safety Plan (HASP) for the demonstration of IITRI's EM Treatment Technology. In this process, soil is heated in situ by means of electrical energy for the removal of hazardous organic contaminants. This process will be demonstrated on a small plot of contaminated soil located in the Pit Area of Classified Burial Ground K-1070-D, K-25 Site, Oak Ridge, TN. The purpose of the demonstration is to remove organic contaminants present in the soil by heating to a temperature range of 85 degrees to 95 degrees C. The soil will be heated in situ by applying 60-Hz AC power to an array of electrodes placed in boreholes drilled through the soil. In this section a brief description of the process is given along with a description of the site and a listing of the contaminants found in the area

Public requirement to demonstrate safety

International Nuclear Information System (INIS)

Green, P.

1991-01-01

To many working within Government or industry, public concern over the disposal of radioactive waste is misplaced and has arisen out of an irrational and unscientific fear of technology, or even science in general. Members of the public, it is argued, are concerned because they do not understand the size of the risk in question. From the industry's point of view, the risk arising from the disposal of radioactive waste is ''negligible when compared to other everyday risks of life. Furthermore, any public exposure that may arise, either soon after closure of a facility or in the far future would comply with internationally accepted safety standards. In this context, the continuing concern over disposal of radioactive waste is viewed as evidence of the irrational and unscientific attitude of the public. The assessment and regulation of risk from waste disposal therefore is presented as a purely scientific question. Some of these issues are examined and public concern is shown not to be irrational but to be based upon legitimate questions over current waste management policy. An important question is not just ''how safe is safe, but who decides and how?''. (Author)

Functional safety of health information technology.

LENUS (Irish Health Repository)

Chadwick, Liam

2012-03-01

In an effort to improve patient safety and reduce adverse events, there has been a rapid growth in the utilisation of health information technology (HIT). However, little work has examined the safety of the HIT systems themselves, the methods used in their development or the potential errors they may introduce into existing systems. This article introduces the conventional safety-related systems development standard IEC 61508 to the medical domain. It is proposed that the techniques used in conventional safety-related systems development should be utilised by regulation bodies, healthcare organisations and HIT developers to provide an assurance of safety for HIT systems. In adopting the IEC 61508 methodology for HIT development and integration, inherent problems in the new systems can be identified and corrected during their development. Also, IEC 61508 should be used to develop a healthcare-specific standard to allow stakeholders to provide an assurance of a system\\'s safety.

Safety Justification and Safety Case for Safety-critical Software in Digital Reactor Protection System

International Nuclear Information System (INIS)

Kwon, Kee-Choon; Lee, Jang-Soo; Jee, Eunkyoung

2016-01-01

Nuclear safety-critical software is under strict regulatory requirements and these regulatory requirements are essential for ensuring the safety of nuclear power plants. The verification & validation (V and V) and hazard analysis of the safety-critical software are required to follow regulatory requirements through the entire software life cycle. In order to obtain a license from the regulatory body through the development and validation of safety-critical software, it is essential to meet the standards which are required by the regulatory body throughout the software development process. Generally, large amounts of documents, which demonstrate safety justification including standard compliance, V and V, hazard analysis, and vulnerability assessment activities, are submitted to the regulatory body during the licensing process. It is not easy to accurately read and evaluate the whole documentation for the development activities, implementation technology, and validation activities. The safety case methodology has been kwon a promising approach to evaluate the level and depth of the development and validation results. A safety case is a structured argument, supported by a body of evidence that provides a compelling, comprehensible, and valid case that a system is safe for a given application in a given operating environment. It is suggested to evaluate the level and depth of the results of development and validation by applying safety case methodology to achieve software safety demonstration. A lot of documents provided as evidence are connected to claim that corresponds to the topic for safety demonstration. We demonstrated a case study in which more systematic safety demonstration for the target system software is performed via safety case construction than simply listing the documents

Safety Justification and Safety Case for Safety-critical Software in Digital Reactor Protection System

Energy Technology Data Exchange (ETDEWEB)

Kwon, Kee-Choon; Lee, Jang-Soo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Jee, Eunkyoung [KAIST, Daejeon (Korea, Republic of)

2016-10-15

Nuclear safety-critical software is under strict regulatory requirements and these regulatory requirements are essential for ensuring the safety of nuclear power plants. The verification & validation (V and V) and hazard analysis of the safety-critical software are required to follow regulatory requirements through the entire software life cycle. In order to obtain a license from the regulatory body through the development and validation of safety-critical software, it is essential to meet the standards which are required by the regulatory body throughout the software development process. Generally, large amounts of documents, which demonstrate safety justification including standard compliance, V and V, hazard analysis, and vulnerability assessment activities, are submitted to the regulatory body during the licensing process. It is not easy to accurately read and evaluate the whole documentation for the development activities, implementation technology, and validation activities. The safety case methodology has been kwon a promising approach to evaluate the level and depth of the development and validation results. A safety case is a structured argument, supported by a body of evidence that provides a compelling, comprehensible, and valid case that a system is safe for a given application in a given operating environment. It is suggested to evaluate the level and depth of the results of development and validation by applying safety case methodology to achieve software safety demonstration. A lot of documents provided as evidence are connected to claim that corresponds to the topic for safety demonstration. We demonstrated a case study in which more systematic safety demonstration for the target system software is performed via safety case construction than simply listing the documents.

Students' Perception of Live Lectures' Inherent Disadvantages

Science.gov (United States)

Petrovic, Juraj; Pale, Predrag

2015-01-01

This paper aims to provide insight into various properties of live lectures from the perspective of sophomore engineering students. In an anonymous online survey conducted at the Faculty of Electrical Engineering and Computing, University of Zagreb, we investigated students' opinions regarding lecture attendance, inherent disadvantages of live…

Dynamic operator actions analysis for inherently safe fast reactors and light water reactors

International Nuclear Information System (INIS)

Ho, V.; Apostolakis, G.

1988-01-01

A comparative dynamic human actions analysis of inherently safe fast reactors (ISFRs) and light water reactors (LWRs) in terms of systems response and estimated human error rates is presented. Brief overviews of the ISFR and LWR systems are given to illustrate the design differences. Key operator actions required by the ISFR reactor shutdown and decay heat removal systems are identified and are compared with those of the LWR. It is observed that, because of the passive nature of the ISFR safety-related systems, a large time window is available for operator actions during transient events. Furthermore, these actions are fewer in number, are less complex, and have lower error rates and less severe consequences than those of the LWRs. We expect the ISFR operator errors' contribution to risk is smaller (at least in the context of the existing human reliability models) than that of the LWRs. (author)

Risk-informed approach for safety, safeguards, and security (3S) by design

International Nuclear Information System (INIS)

Suzuki, Mitsutoshi; Burr, Tom; Howell, John

2011-01-01

Over several decades the nuclear energy society worldwide has developed safety assessment methodology based on probabilistic risk analysis for incorporating its benefit into design and accident prevention for nuclear reactors. Although safeguards and security communities have different histories and technical aspects compared to safety, risk assessment as a supplement to their current requirements could be developed to promote synergism between Safety, Safeguards, and Security (3S) and to install effective countermeasures in the design of complex nuclear fuel cycle facilities. Since the 3S initiative was raised by G8 countries at Hokkaido Toyako-Summit in 2008, one approach to developing synergism in a 3S By Design (3SBD) process has been the application of risk-oriented assessment methodology. In the existing regulations of safeguards and security, a risk notion has already been considered for inherent threat and hazard recognition. To integrate existing metrics into a risk-oriented approach, several mathematical methods have already been surveyed, with attention to the scarcity of intentional acts in the case of safeguards and the sparseness of actual event data. A two-dimensional probability distribution composed of measurement error and incidence probabilities has been proposed to formalize inherent difficulties in the International Atomic Energy Agency (IAEA) safeguards criteria. In particular, the incidence probability that is difficult to estimate has been explained using a Markov model and game theory. In this work, a feasibility study of 3SBD is performed for an aqueous reprocessing process, and synergetic countermeasures are presented for preliminary demonstration of 3SBD. Although differences and conflicts between individual 'S' communities exist, the integrated approach would be valuable for optimization and balance between the 3S design features as well as for effective and efficient implementation under existing regulation frameworks. In addition

Probabilistic safety assessment based expert systems in support of dynamic risk assessment

International Nuclear Information System (INIS)

Varde, P.V.; Sharma, U.L.; Marik, S.K.; Raina, V.K.; Tikku, A.C.

2006-01-01

Probabilistic Safety Assessment (PSA) studies are being performed, world over as part of integrated risk assessment for Nuclear Power Plants and in many cases PSA insight is utilized in support of decision making. Though the modern plants are built with inherent safety provisions, particularly to reduce the supervisory requirements during initial period into the accident, it is always desired to develop an efficient user friendly real-time operator advisory system for handling of plant transients/emergencies which would be of immense benefit for the enhancement of operational safety of the plant. This paper discusses an integrated approach for the development of operator support system. In this approach, PSA methodology and the insight obtained from PSA has been utilized for development of knowledge based or rule based experts system. While Artificial Neural Network (ANN) approach has been employed for transient identification, rule-base expert system shell environment was used for the development of diagnostic module in this system. Attempt has been made to demonstrate that this approach offers an efficient framework for addressing requirements related to handling of real-time/dynamic scenario. (author)

Safety physics inter-comparison of advanced concepts of critical reactors and ADS

Energy Technology Data Exchange (ETDEWEB)

Slessarev, I. [CEA Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. d' Etudes des Reacteurs

2001-07-01

Enhanced safety based on the principle of the natural ''self-defence'' is one of the most desirable features of innovative nuclear systems (critical or sub-critical) regarding both TRU transmutation and ''clean'' energy producer concepts. For the evaluation of the ''self-defence'' domain, the method of the asymptotic reactivity balance has been generalised. The promising option of Hybrids systems (that use a symbiosis of fission and spallation in sub-critical cores) which could benefit the advantages of both Accelerated Driven Systems of the traditional type and regular critical systems, has been advocated. General features of Hybrid dynamics have been presented and analysed. It was demonstrated that an external neutron source of Hybrids can expand the inherent safety potential significantly. This analysis has been applied to assess the safety physics potential of innovative concepts for prospective nuclear power both for energy producers and for transmutation. It has been found, that safety enhancement goal defines a choice of sub-criticality of Hybrids. As for energy producers with Th-fuel cycle, a significant sub-criticality level is required due to a necessity of an improvement of neutronics together with safety enhancement task. (author)

Safety assessment document for spent fuel handling, packaging, and storage demonstrations at the E-MAD facility on the Nevada Test Site

International Nuclear Information System (INIS)

1985-04-01

The objectives for spent fuel handling and packaging demonstration are to develop the capability to satisfactorily encapsulate typical commercial nuclear reactor spent fuel assemblies and to establish the suitability of interim dry surface and near surface storage concepts. To accomplish these objectives, spent fuel assemblies from a pressurized water reactor have been received, encapsulated in steel canisters, and emplaced in on-site storage facilities and subjected to other tests. As an essential element of these demonstrations, a thorough safety assessment of the demonstration activities conducted at the E-MAD facility has been completed. This document describes the site location and characteristics, the existing E-MAD facility, and the facility modifications and equipment additions made specifically for the demonstrations. The document also summarizes the Quality Assurance Program utilized, and specifies the principal design criteria applicable to the facility modifications, equipment additions, and process operations. Evaluations have been made of the radiological impacts of normal operations, abnormal operations, and postulated accidents. Analyses have been performed to determine the affects on nuclear criticality safety of postulated accidents and credible natural phenomena. The consequences of postulated accidents resulting in fission product gas release have also been estimated. This document identifies the engineered safety features, procedures, and site characteristics that (1) prevent the occurrence of potential accidents or (2) assure that the consequences of postulated accidents are either insignificant or adequately mitigated

Psychological aspects of food safety risk perception

DEFF Research Database (Denmark)

Scholderer, Joachim

signals, motivating approach. Novelty, and the detection of certain olfactory and visual cues associated with spoilage or contamination, act as orientation or threat signals and motivate closer inspection or avoidance. Anticipatory affects are an inherent part of these behaviour regulation systems...... problematic food safety behaviours are likely to occur. The presentation will begin with an overview of the relevant psychological mechanisms that regulate approach and avoidance behaviour with respect to potentially hazardous foods. Learned representations of familiarity and reward value act as safety...

The inherence heuristic: an intuitive means of making sense of the world, and a potential precursor to psychological essentialism.

Science.gov (United States)

Cimpian, Andrei; Salomon, Erika

2014-10-01

We propose that human reasoning relies on an inherence heuristic, an implicit cognitive process that leads people to explain observed patterns (e.g., girls wear pink) predominantly in terms of the inherent features of their constituents (e.g., pink is a delicate color). We then demonstrate how this proposed heuristic can provide a unified account for a broad set of findings spanning areas of research that might at first appear unrelated (e.g., system justification, nominal realism, is-ought errors in moral reasoning). By revealing the deep commonalities among the diverse phenomena that fall under its scope, our account is able to generate new insights into these phenomena, as well as new empirical predictions. A second main goal of this article, aside from introducing the inherence heuristic, is to articulate the proposal that the heuristic serves as a foundation for the development of psychological essentialism. More specifically, we propose that essentialism - which is the common belief that natural and social categories are underlain by hidden, causally powerful essences - emerges over the first few years of life as an elaboration of the earlier, and more open-ended, intuitions supplied by the inherence heuristic. In the final part of the report, we distinguish our proposal from competing accounts (e.g., Strevens's K-laws) and clarify the relationship between the inherence heuristic and related cognitive tendencies (e.g., the correspondence bias). In sum, this article illuminates a basic cognitive process that emerges early in life and is likely to have profound effects on many aspects of human psychology.

Observation and analysis of water inherent optical properties

Science.gov (United States)

Sun, Deyong; Li, Yunmei; Le, Chengfeng; Huang, Changchun

2008-03-01

Inherent optical property is an important part of water optical properties, and is the foundation of water color analytical model establishment. Through quantity filter technology (QFT) and backscattering meter BB9 (WETlabs Inc), absorption coefficients of CDOM, total suspended minerals and backscattering coefficients of total suspended minerals had been observed in Meiliang Bay of Taihu lake at summer and winter respectively. After analyzing the spectral characteristics of absorption and backscattering coefficients, the differences between two seasons had been illustrated adequately, and the reasons for the phenomena, which are related to the changes of water quality coefficient, had also been explained. So water environment states can be reflected by inherent optical properties. In addition, the relationship models between backscattering coefficients and suspended particle concentrations had been established, which can support coefficients for analytical models.

Safety analysis for 'Fugen'

International Nuclear Information System (INIS)

1997-10-01

The improvement of safety in nuclear power stations is an important proposition. Therefore also as to the safety evaluation, it is important to comprehensively and systematically execute it by referring to the operational experience and the new knowledge which is important for the safety throughout the period of use as well as before the construction and the start of operation of nuclear power stations. In this report, the results when the safety analysis for ''Fugen'' was carried out by referring to the newest technical knowledge are described. As the result, it was able to be confirmed that the safety of ''Fugen'' has been secured by the inherent safety and the facilities which were designed for securing the safety. The basic way of thinking on the safety analysis including the guidelines to be conformed to is mentioned. As to the abnormal transient change in operation and accidents, their definition, the events to be evaluated and the standards for judgement are reported. The matters which were taken in consideration at the time of the analysis are shown. The computation programs used for the analysis were REACT, HEATUP, LAYMON, FATRAC, SENHOR, LOTRAC, FLOOD and CONPOL. The analyses of the abnormal transient change in operation and accidents are reported on the causes, countermeasures, protective functions and results. (K.I.)

Conceptual safety design analysis of Korea advanced liquid metal reactor

International Nuclear Information System (INIS)

Suk, S. D.; Park, C. K.

1999-01-01

The national long-term R and D program, updated in 1977, requires Korea Atomic Energy Research Institute (KAERI) to complete by the year 2006 the basic design of Korea Advanced Liquid Metal Reactor (KALIMER), along with supporting R and D work, with the capability of resolving the issue of spent fuel storage as well as with significantly enhanced safety. KALIMER is a 150 Mwe pool-type sodium cooled prototype reactor that uses metallic fuel. The conceptual design is currently under way to establish a self-consistent design meeting a set of major safety design requirements for accident prevention. Some of the current emphasis includes those for inherent and passive means of negative reactivity insertion and decay heat removal, high shutdown reliability, prevention of and protection from sodium chemical reaction, and high seismic margin, among others. All of these requirements affect the reactor design significantly and involve extensive supporting R and D programs. This paper summarizes some of the results of conceptual engineering and design analyses performed for the safety of KALIMER in the area of inherent safety, passive decay heat removal, sodium water reaction, and seismic isolation. (author)

Inherent reward & risk (Part I): Towards a universal paradigm for investment analysis

NARCIS (Netherlands)

Zou, L.

2000-01-01

In this paper, a new paradigm is developed for analyzinginvestment strategies and pricing financial assets. This paradigmassumes that any investment strategy has its own "inherent reward"and "inherent risk" that can be judged with common sense. Ijustify axiomatically the existence and uniqueness

Safety characteristics analysis of nuclear power plants with PHWR PT; Analiza sigurnosnh karakteristika nuklearnih elektrana sa reaktorima PHWR-PT tipa

Energy Technology Data Exchange (ETDEWEB)

Stosic, Z [Institute of nuclear sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

1983-07-01

The paper deals with analysis of basic safety characteristics of heavy water Candu reactor. Inherent safety characteristics, r/a material inventory, systematization of normal abnormal and transient conditions, safety systems and availability analysis are considered. (author)

The HTR safety concept demonstrated by selected examples

International Nuclear Information System (INIS)

Sommer, H.; Stoelzl, D.

1981-01-01

The licensing experience gained in the Federal Republic of Germany is based on the licensing procedures for the THTR-300 and the HTR-1160. In the course of the licensing procedures for these reactors a safety concept for an HTR has been developed. This experience constitutes the basis for the design of future HTR's. (author)

Adaptation of the ToxRTool to Assess the Reliability of Toxicology Studies Conducted with Genetically Modified Crops and Implications for Future Safety Testing.

Science.gov (United States)

Koch, Michael S; DeSesso, John M; Williams, Amy Lavin; Michalek, Suzanne; Hammond, Bruce

2016-01-01

To determine the reliability of food safety studies carried out in rodents with genetically modified (GM) crops, a Food Safety Study Reliability Tool (FSSRTool) was adapted from the European Centre for the Validation of Alternative Methods' (ECVAM) ToxRTool. Reliability was defined as the inherent quality of the study with regard to use of standardized testing methodology, full documentation of experimental procedures and results, and the plausibility of the findings. Codex guidelines for GM crop safety evaluations indicate toxicology studies are not needed when comparability of the GM crop to its conventional counterpart has been demonstrated. This guidance notwithstanding, animal feeding studies have routinely been conducted with GM crops, but their conclusions on safety are not always consistent. To accurately evaluate potential risks from GM crops, risk assessors need clearly interpretable results from reliable studies. The development of the FSSRTool, which provides the user with a means of assessing the reliability of a toxicology study to inform risk assessment, is discussed. Its application to the body of literature on GM crop food safety studies demonstrates that reliable studies report no toxicologically relevant differences between rodents fed GM crops or their non-GM comparators.

Demonstration of safety for geologic disposal

International Nuclear Information System (INIS)

Taylor, E.C.; Ramspott, L.D.; Sprecher, W.M.

1994-01-01

The US Department of Energy (DOE) is developing a nuclear waste management system that will accept high-level radioactive waste, transport it, store it, and ultimately emplace it in a deep geologic repository. The key activity now is determining whether Yucca Mountain, Nevada is suitable as a site for the repository. If so, the crucial technological advance will be the demonstration that disposal of nuclear waste will be safe for thousands of years after closure. This paper assesses the impact of regulatory developments, legal developments, and scientific developments on such a demonstration

Specific safety measures for emergency lanes and shoulders of motorways : a proposal for motorways' authorities in the framework of the European research project Safety Standards for Road Design and Redesign SAFESTAR, Workpackage 1.1.

NARCIS (Netherlands)

Braimaister, L.

1999-01-01

This workpackage is one of seven workpackages of the European SAFESTAR project, launched by DG VII. Directing on safety standards and recommendations for the Trans-European Roadway Network (TERN), the workpackage considered safety measures on emergency lanes (stopping strips), which are inherent

Charged-particle beam: a safety mandate

International Nuclear Information System (INIS)

Young, K.C.

1983-01-01

The Advanced Test Accelerator (ATA) is a recent development in the field of charged particle beam research at Lawrence Livermore National Laboratory. With this experimental apparatus, researchers will characterize intense pulses of electron beams propagated through air. Inherent with the ATA concept was the potential for exposure to hazards, such as high radiation levels and hostile breathing atmospheres. The need for a comprehensive safety program was mandated; a formal system safety program was implemented during the project's conceptual phase. A project staff position was created for a safety analyst who would act as a liaison between the project staff and the safety department. Additionally, the safety analyst would be responsible for compiling various hazards analyses reports, which formed the basis of th project's Safety Analysis Report. Recommendations for safety features from the hazards analysis reports were incorporated as necessary at appropriate phases in project development rather than adding features afterwards. The safety program established for the ATA project faciliated in controlling losses and in achieving a low-level of acceptable risk

Relevance of passive safety testing at the fast flux test facility to advanced liquid metal reactors - 5127

International Nuclear Information System (INIS)

Wootan, D.W.; Omberg, R.P.

2015-01-01

Significant cost and safety improvements can be realized in advanced liquid metal reactor (LMR) designs by emphasizing inherent or passive safety through crediting the beneficial reactivity feedbacks associated with core and structural movement. This passive safety approach was adopted for the Fast Flux Test Facility (FFTF), and an experimental program was conducted to characterize the structural reactivity feedback. Testing at the Rapsodie and EBR-II reactors had demonstrated the beneficial effect of reactivity feedback caused by changes in fuel temperature and core geometry mechanisms in a liquid metal fast reactor in a holistic sense. The FFTF passive safety testing program was developed to examine how specific design elements influenced dynamic reactivity feedback in response to a reactivity input and to demonstrate the scalability of reactivity feedback results from smaller cores like Rapsodie and EBR-II to reactor cores that were more prototypic in scale to reactors of current interest. The U.S. Department of Energy, Office of Nuclear Energy Advanced Reactor Technology program is in the process of preserving, protecting, securing, and placing in electronic format information and data from the FFTF, including the core configurations and data collected during the passive safety tests. Evaluation of these actual test data could provide insight to improve analytical methods which may be used to support future licensing applications for LMRs. (authors)

Long range stress correlations in the inherent structures of liquids at rest

Energy Technology Data Exchange (ETDEWEB)

Chowdhury, Sadrul; Abraham, Sneha; Hudson, Toby; Harrowell, Peter [School of Chemistry, University of Sydney, Sydney, NSW 2006 (Australia)

2016-03-28

Simulation studies of the atomic shear stress in the local potential energy minima (inherent structures) are reported for binary liquid mixtures in 2D and 3D. These inherent structure stresses are fundamental to slow stress relaxation and high viscosity in supercooled liquids. We find that the atomic shear stress in the inherent structures (IS’s) of both liquids at rest exhibits slowly decaying anisotropic correlations. We show that the stress correlations contribute significantly to the variance of the total shear stress of the IS configurations and consider the origins of the anisotropy and spatial extent of the stress correlations.

The deep geologic repository technology programme: toward a geoscience basis for understanding repository safety

International Nuclear Information System (INIS)

Jensen, M.R.

2007-01-01

Within the Deep Geologic Repository Technology Programme (DGRTP) several Geoscience activities are focused on advancing the understanding of groundwater flow system evolution and geochemical stability in a Canadian Shield setting as affected by long-term climate change. A key aspect is developing confidence in predictions of groundwater flow patterns and residence times as they relate to the safety of a deep geologic repository for used nuclear fuel waste. This is being achieved through a coordinated multi-disciplinary approach intent on: i) demonstrating coincidence between independent geo-scientific data; ii) improving the traceability of geo-scientific data and its interpretation within a conceptual descriptive model(s); iii) improving upon methods to assess and demonstrate robustness in flow domain prediction(s) given inherent flow domain uncertainties (i.e. spatial chemical/physical property distributions, boundary conditions) in time and space; and iv) improving awareness amongst geo-scientists as to the utility of various geo-scientific data in supporting a safety case for a deep geologic repository. This multi-disciplinary DGRTP approach is yielding an improved understanding of groundwater flow system evolution and stability in Canadian Shield settings that is further contributing to the geo-scientific basis for understanding and communicating aspects of DGR safety. (author)

Analysis of multiple failure accident scenarios for development of probabilistic safety assessment model for KALIMER-600

International Nuclear Information System (INIS)

Kim, T.W.; Suk, S.D.; Chang, W.P.; Kwon, Y.M.; Jeong, H.Y.; Lee, Y.B.; Ha, K.S.; Kim, S.J.

2009-01-01

A sodium-cooled fast reactor (SFR), KALIMER-600, is under development at KAERI. Its fuel is the metal fuel of U-TRU-Zr and it uses sodium as coolant. Its advantages are found in the aspects of an excellent uranium resource utilization, inherent safety features, and nonproliferation. The probabilistic safety assessment (PSA) will be one of the initiating subjects for designing it from the aspects of a risk informed design (RID) as well as a technology-neutral licensing (TNL). The core damage is defined as coolant voiding, fuel melting, or cladding damage. Accident scenarios which lead to the core damage should be identified for the development of a Level-1 PSA model. The SSC-K computer code is used to identify the conditions which lead to core damage. KALIMER-600 has passive safety features such as passive shutdown functions, passive pump coast-down features, and passive decay heat removal systems. It has inherent reactivity feedback effects such as Doppler, sodium void, core axial expansion, control rod axial expansion, core radial expansion, etc. The accidents which are analyzed are the multiple failure accidents such as an unprotected transient overpower, a loss of flow, and a loss of heat sink events with degraded safety systems or functions. The safety functions to be considered here are a reactor trip, inherent reactivity feedback features, the pump coast-down, and the passive decay heat removal. (author)

The near boiling reactor: Conceptual design of a small inherently safe nuclear reactor to extend the operational capability of the Victoria Class submarine

Science.gov (United States)

Cole, Christopher J. P.

Nuclear power has several unique advantages over other air independent energy sources for nuclear combat submarines. An inherently safe, small nuclear reactor, capable of supply the hotel load of the Victoria Class submarines, has been conceptually developed. The reactor is designed to complement the existing diesel electric power generation plant presently onboard the submarine. The reactor, rated at greater than 1 MW thermal, will supply electricity to the submarine's batteries through an organic Rankine cycle energy conversion plant at 200 kW. This load will increase the operational envelope of the submarine by providing up to 28 continuous days submerged, allowing for an enhanced indiscretion ratio (ratio of time spent on the surface versus time submerged) and a limited under ice capability. The power plant can be fitted into the existing submarine by inserting a 6 m hull plug. With its simplistic design and inherent safety features, the reactor plant will require a minimal addition to the crew. The reactor employs TRISO fuel particles for increased safety. The light water coolant remains at atmospheric pressure, exiting the core at 96°C. Burn-up control and limiting excess reactivity is achieved through movable reflector plates. Shut down and regulatory control is achieved through the thirteen hafnium control rods. Inherent safety is achieved through the negative prompt and delayed temperature coefficients, as well as the negative void coefficient. During a transient, the boiling of the moderator results in a sudden drop in reactivity, essentially shutting down the reactor. It is this characteristic after which the reactor has been named. The design of the reactor was achieved through modelling using computer codes such as MCNP5, WIMS-AECL, FEMLAB, and MicroShield5, in addition to specially written software for kinetics, heat transfer and fission product poisoning calculations. The work has covered a broad area of research and has highlighted additional areas

The near boiling reactor : conceptual design of a small inherently safe nuclear reactor to extend the operational capability of the Victoria Class submarine

International Nuclear Information System (INIS)

Cole, C.J.P.

2005-01-01

Nuclear power has several unique advantages over other air independent energy sources for nuclear combat submarines. An inherently safe, small nuclear reactor, capable of supply the hotel load of the 'Victoria' Class submarines, has been conceptually developed. The reactor is designed to complement the existing diesel electric power generation plant presently onboard the submarine. The reactor, rated at greater than 1 MW thermal, will supply electricity to the submarine's batteries through an organic Rankine cycle energy conversion plant at 200 kW. This load will increase the operational envelope of the submarine by providing up to 28 continuous days submerged, allowing for an enhanced indiscretion ratio (ratio of time spent on the surface versus time submerged) and a limited under ice capability. The power plant can be fitted into the existing submarine by inserting a 6 m hull plug. With its simplistic design and inherent safety features, the reactor plant will require a minimal addition to the crew. The reactor employs TRISO fuel particles for increased safety. The light water coolant remains at atmospheric pressure, exiting the core at 96 o C. Burn-up control and limiting excess reactivity is achieved through movable reflector plates. Shut down and regulatory control is achieved through the thirteen hafnium control rods. Inherent safety is achieved through the negative prompt and delayed temperature coefficients, as well as the negative void coefficient. During a transient, the boiling of the moderator results in a sudden drop in reactivity, essentially shutting down the reactor. It is this characteristic after which the reactor has been named. The design of the reactor was achieved through modelling using computer codes such as MCNP5, WIMS-AECL, FEMLAB, and MicroShield5, in addition to specially written software for kinetics, heat transfer and fission product poisoning calculations. The work has covered a broad area of research and has highlighted additional

Safety demonstration tests on pressure rise in ventilation system and blower integrity of a fuel-reprocessing plant

Energy Technology Data Exchange (ETDEWEB)

Takada, Junichi; Suzuki, Motoe; Tsukamoto, Michio; Koike, Tadao; Nishio, Gunji [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1996-12-01

In JAERI, the demonstration test was carried out as a part of safety researches of the fuel-reprocessing plant using a large-scale facility consist of cells, ducts, dumpers, HEPA filters and a blower, when an explosive burning due to a rapid reaction of thermal decomposition for solvent/nitric acid occurs in a cell of the reprocessing plant. In the demonstration test, pressure response propagating through the facility was measured under a blowing of air from a pressurized tank into the cell in the facility to elucidate an influence of pressure rise in the ventilation system. Consequently, effective pressure decrease in the facility was given by a configuration of cells and ducts in the facility. In the test, transient responses of HEPA filters and the blower by the blowing of air were also measured to confirm the integrity. So that, it is confirmed that HEPA filters and the blower under pressure loading were sufficient to maintain the integrity. The content described in this report will contribute to safety assessment of the ventilation system in the event of explosive burning in the reprocessing plant. (author)

Quantifying uncertainties in the estimation of safety parameters by using bootstrapped artificial neural networks

Energy Technology Data Exchange (ETDEWEB)

Secchi, Piercesare [MOX, Department of Mathematics, Polytechnic of Milan (Italy); Zio, Enrico [Department of Energy, Polytechnic of Milan, Via Ponzio 34/3, 20133 Milano (Italy)], E-mail: enrico.zio@polimi.it; Di Maio, Francesco [Department of Energy, Polytechnic of Milan, Via Ponzio 34/3, 20133 Milano (Italy)

2008-12-15

For licensing purposes, safety cases of Nuclear Power Plants (NPPs) must be presented at the Regulatory Authority with the necessary confidence on the models used to describe the plant safety behavior. In principle, this requires the repetition of a large number of model runs to account for the uncertainties inherent in the model description of the true plant behavior. The present paper propounds the use of bootstrapped Artificial Neural Networks (ANNs) for performing the numerous model output calculations needed for estimating safety margins with appropriate confidence intervals. Account is given both to the uncertainties inherent in the plant model and to those introduced by the ANN regression models used for performing the repeated safety parameter evaluations. The proposed framework of analysis is first illustrated with reference to a simple analytical model and then to the estimation of the safety margin on the maximum fuel cladding temperature reached during a complete group distribution header blockage scenario in a RBMK-1500 nuclear reactor. The results are compared with those obtained by a traditional parametric approach.

Demonstration of Risk Profiling for promoting safety in SME´s

DEFF Research Database (Denmark)

Jørgensen, Kirsten; Duijm, Nijs Jan; Troen, Hanne

2011-01-01

Purpose – The purpose of this paper is to identify and assess the risks and potential risks that may lead to accidents. It aims to look at how to improve risk assessment within SMEs for the benefit of all staff. Design/methodology/approach – The research included results from a Dutch project which...... identifies accident risks and safety barriers that are presented in a huge database and risk calculator. The method was first to develop a simple way of accessing this enormous amount of data, second, to develop a tool to observe risks and safety barriers in SMEs and to investigate the usefulness...... of the developed tools in real life, third, to collect data on risks and safety barriers in SMEs for two occupations by following 20 people for three days each and to create a risk profile for each occupations. Findings – The result is a simple way to go through all types of risks for accidents – a tool for risk...

Memory accessibility shapes explanation: Testing key claims of the inherence heuristic account.

Science.gov (United States)

Hussak, Larisa J; Cimpian, Andrei

2018-01-01

People understand the world by constructing explanations for what they observe. It is thus important to identify the cognitive processes underlying these judgments. According to a recent proposal, everyday explanations are often constructed heuristically: Because people need to generate explanations on a moment-by-moment basis, they cannot perform an exhaustive search through the space of possible reasons, but may instead use the information that is most easily accessible in memory (Cimpian & Salomon 2014a, b). In the present research, we tested two key claims of this proposal that have so far not been investigated. First, we tested whether-as previously hypothesized-the information about an entity that is most accessible in memory tends to consist of inherent or intrinsic facts about that entity, rather than extrinsic (contextual, historical, etc.) facts about it (Studies 1 and 2). Second, we tested the implications of this difference in the memory accessibility of inherent versus extrinsic facts for the process of generating explanations: Does the fact that inherent facts are more accessible than relevant extrinsic facts give rise to an inherence bias in the content of the explanations generated (Studies 3 and 4)? The findings supported the proposal that everyday explanations are generated in part via a heuristic process that relies on easily accessible-and often inherent-information from memory.

Is the inherent potential of maize roots efficient for soil phosphorus acquisition?

Directory of Open Access Journals (Sweden)

Yan Deng

Full Text Available Sustainable agriculture requires improved phosphorus (P management to reduce the overreliance on P fertilization. Despite intensive research of root adaptive mechanisms for improving P acquisition, the inherent potential of roots for efficient P acquisition remains unfulfilled, especially in intensive agriculture, while current P management generally focuses on agronomic and environmental concerns. Here, we investigated how levels of soil P affect the inherent potential of maize (Zea mays L. roots to obtain P from soil. Responses of root morphology, arbuscular mycorrhizal colonization, and phosphate transporters were characterized and related to agronomic traits in pot and field experiments with soil P supply from deficiency to excess. Critical soil Olsen-P level for maize growth approximated 3.2 mg kg(-1, and the threshold indicating a significant environmental risk was about 15 mg kg(-1, which represented the lower and upper levels of soil P recommended in current P management. However, most root adaptations involved with P acquisition were triggered when soil Olsen-P was below 10 mg kg(-1, indicating a threshold for maximum root inherent potential. Therefore, to maintain efficient inherent potential of roots for P acquisition, we suggest that the target upper level of soil P in intensive agriculture should be reduced from the environmental risk threshold to the point maximizing the inherent potential of roots.

Is the inherent potential of maize roots efficient for soil phosphorus acquisition?

Science.gov (United States)

Deng, Yan; Chen, Keru; Teng, Wan; Zhan, Ai; Tong, Yiping; Feng, Gu; Cui, Zhenling; Zhang, Fusuo; Chen, Xinping

2014-01-01

Sustainable agriculture requires improved phosphorus (P) management to reduce the overreliance on P fertilization. Despite intensive research of root adaptive mechanisms for improving P acquisition, the inherent potential of roots for efficient P acquisition remains unfulfilled, especially in intensive agriculture, while current P management generally focuses on agronomic and environmental concerns. Here, we investigated how levels of soil P affect the inherent potential of maize (Zea mays L.) roots to obtain P from soil. Responses of root morphology, arbuscular mycorrhizal colonization, and phosphate transporters were characterized and related to agronomic traits in pot and field experiments with soil P supply from deficiency to excess. Critical soil Olsen-P level for maize growth approximated 3.2 mg kg(-1), and the threshold indicating a significant environmental risk was about 15 mg kg(-1), which represented the lower and upper levels of soil P recommended in current P management. However, most root adaptations involved with P acquisition were triggered when soil Olsen-P was below 10 mg kg(-1), indicating a threshold for maximum root inherent potential. Therefore, to maintain efficient inherent potential of roots for P acquisition, we suggest that the target upper level of soil P in intensive agriculture should be reduced from the environmental risk threshold to the point maximizing the inherent potential of roots.

Dealing with uncertainty and pursuing superior technology options in risk management-The inherency risk analysis

International Nuclear Information System (INIS)

Helland, Aasgeir

2009-01-01

Current regulatory systems focus on the state of scientific evidence as the predominant factor for how to handle risks to human health and the environment. However, production and assessment of risk information are costly and time-consuming, and firms have an intrinsic disincentive to produce and distribute information about risks of their products as this could endanger their production opportunities and sales. An emphasis on more or better science may result in insufficient thought and attention going into the exploration of technology alternatives, and that risk management policies miss out on the possible achievement of a more favorable set of consequences. In this article, a method is proposed that combines risk assessment with the search for alternative technological options as a part of the risk management procedure. The method proposed is the inherency risk analysis where the first stage focuses on the original agent subject to investigation, the second stage focuses on identifying technological options whereas the third stage reviews the different alternatives, searching for the most attractive tradeoffs between costs and inherent safety. This is then used as a fundament for deciding which technology option to pursue. This method aims at providing a solution-focused, systematic technology-based approach for addressing and setting priorities for environmental problems. By combining risk assessment with a structured approach to identify superior technology options within a risk management system, the result could very well be a win-win situation for both company and the environment.

Is human fracture hematoma inherently angiogenic?

LENUS (Irish Health Repository)

Street, J

2012-02-03

This study attempts to explain the cellular events characterizing the changes seen in the medullary callus adjacent to the interfragmentary hematoma during the early stages of fracture healing. It also shows that human fracture hematoma contains the angiogenic cytokine vascular endothelial growth factor and has the inherent capability to induce angiogenesis and thus promote revascularization during bone repair. Patients undergoing emergency surgery for isolated bony injury were studied. Raised circulating levels of vascular endothelial growth factor were seen in all injured patients, whereas the fracture hematoma contained significantly higher levels of vascular endothelial growth factor than did plasma from these injured patients. However, incubation of endothelial cells in fracture hematoma supernatant significantly inhibited the in vitro angiogenic parameters of endothelial cell proliferation and microtubule formation. These phenomena are dependent on a local biochemical milieu that does not support cytokinesis. The hematoma potassium concentration is cytotoxic to endothelial cells and osteoblasts. Subcutaneous transplantation of the fracture hematoma into a murine wound model resulted in new blood vessel formation after hematoma resorption. This angiogenic effect is mediated by the significant concentrations of vascular endothelial growth factor found in the hematoma. This study identifies an angiogenic cytokine involved in human fracture healing and shows that fracture hematoma is inherently angiogenic. The differences between the in vitro and in vivo findings may explain the phenomenon of interfragmentary hematoma organization and resorption that precedes fracture revascularization.

Increasing the Fuel Economy and Safety of New Light-DutyVehicles

Energy Technology Data Exchange (ETDEWEB)

Wenzel, Tom; Ross, Marc

2006-09-18

One impediment to increasing the fuel economy standards forlight-duty vehicles is the long-standing argument that reducing vehiclemass to improve fuel economy will inherently make vehicles less safe.This technical paper summarizes and examines the research that is citedin support of this argument, and presents more recent research thatchallenges it. We conclude that the research claiming that lightervehicles are inherently less safe than heavier vehicles is flawed, andthat other aspects of vehicle design are more important to the on-roadsafety record of vehicles. This paper was prepared for a workshop onexperts in vehicle safety and fuel economy, organized by the William andFlora Hewlett Foundation, to discuss technologies and designs that can betaken to simultaneously improve vehicle safety and fuel economy; theworkshop was held in Washington DC on October 3, 2006.

Safety demonstration tests on thermal decomposition of nitrated solvent with nitric acid in nuclear fuel reprocessing plants. Contract research

International Nuclear Information System (INIS)

Tsukamoto, Michio; Takada, Junichi; Koike, Tadao; Watanabe, Koji; Uchiyama, Gunzou; Nishio, Gunji; Murata, Mikio

2001-03-01

The demonstration tests were conducted to investigate the safety of the ventilation system and integrity of the HEPA filters under the design basis accident (DBA) of the evaporator in the reprocessing plants. The tests were carried out by heating organic solvent (TBP/n- dodecane) mixed with nitric acid in a sealed vessel. It was possible to cause an explosive decomposition of TBP-complex formed by nitration of the solvent with nitric acid. The following was obtained by the analysis of the experimental results of the tests. From derivation by the experimental method, data on the maximum mass release rate and the maximum energy release rate in the explosion, as the solvent of 1 [kg] spouted out by the thermal decomposition, were obtained. They were 0.59 [kg/s] and 3240.3 [kJ/kg·s] respectively. The influence given on the cell ventilation system by this explosion was small and it was demonstrated that the safety of the HEPA filters could be secured. (author)

Inherent variation in multiple shoe-sole test impressions.

Science.gov (United States)

Shor, Yaron; Wiesner, Sarena; Tsach, Tsadok; Gurel, Ron; Yekutieli, Yoram

2018-04-01

Shoeprints left at crime scenes are seldom perfect. Many prints are distorted or contaminated by various materials. Noisy background often contributes to vagueness on the shoeprints as well. Test impressions made from the suspect's shoes in the laboratory are considered a genuine replication of the shoe-sole. This naïve attitude is far from being correct. Consecutive test impressions made in the laboratory under strict similar conditions revealed differences among the exemplars of the same sole. Some of them are minor, but some are major, and can mislead the less experienced practitioners during the comparison process. This article focuses on the inherent within source variability between controlled shoeprints made from the same shoe, as it appears on the RACs. To describe and analyze this variability, repeated test impressions were prepared, and datasets were created. Several RACs were marked on each test impression, using an expert assisting software tool (developed in the authors' lab). The variance in repeated test impressions is demonstrated and possible sources are discussed. This variance should be considered when trying to establish the degree of matching between individual characteristics. Copyright © 2017. Published by Elsevier B.V.

On the maximum entropy distributions of inherently positive nuclear data

Energy Technology Data Exchange (ETDEWEB)

Taavitsainen, A., E-mail: aapo.taavitsainen@gmail.com; Vanhanen, R.

2017-05-11

The multivariate log-normal distribution is used by many authors and statistical uncertainty propagation programs for inherently positive quantities. Sometimes it is claimed that the log-normal distribution results from the maximum entropy principle, if only means, covariances and inherent positiveness of quantities are known or assumed to be known. In this article we show that this is not true. Assuming a constant prior distribution, the maximum entropy distribution is in fact a truncated multivariate normal distribution – whenever it exists. However, its practical application to multidimensional cases is hindered by lack of a method to compute its location and scale parameters from means and covariances. Therefore, regardless of its theoretical disadvantage, use of other distributions seems to be a practical necessity. - Highlights: • Statistical uncertainty propagation requires a sampling distribution. • The objective distribution of inherently positive quantities is determined. • The objectivity is based on the maximum entropy principle. • The maximum entropy distribution is the truncated normal distribution. • Applicability of log-normal or normal distribution approximation is limited.

A general index of inherent risk

OpenAIRE

Schnytzer, Adi; Westreich, Sara

2009-01-01

We extend the pioneering work of Aumann and Serrano by presenting an index of inherent riskiness of a gamble having the desirable properties of their index, while being applicable to gambles with either positive or negative expectations. As such, our index provides a measure of riskiness which is of use for both risk lovers and risk aversive gamblers, and is defined for all discrete and a large class of continuous gambles. We analyze abstract properties of our index, and present in addition t...

Summary of FY 1997 work related to JAPC-U.S. DOE contract study on improvement of core safety - study on GEM (III)

International Nuclear Information System (INIS)

Burke, T.M.

1998-01-01

FFTF was originally designed/constructed/operated to develop LMFBR fuels and materials. Inherent safety became a major focus of the US nuclear industry in the mid 1980's. The inherent safety characteristics of LMFBRs were recognized but additional enhancement was desired. The presentation contents are: Fast Flux Test Facility history and status; Overview of contract activities; Summary of loss of flow without scram with GEMs testing; and Summary of pump start with GEMs testing

Demonstration of the LHC Safety Training Tunnel Mock-Up

CERN Multimedia

Brice, Maximilien

2014-01-01

Members of CERN's management visit the LHC tunnel mock-up at the Safety Training Centre on the Prévessin site. The facility is used to train personnel in emergency responses including the use of masks and safe evacuation.

Tactical supply chain planning models with inherent flexibility

DEFF Research Database (Denmark)

Esmaeilikia, Masoud; Fahimnia, Behnam; Sarkis, Joeseph

2016-01-01

Supply chains (SCs) can be managed at many levels. The use of tactical SC planning models with multiple flexibility options can help manage the usual operations efficiently and effectively, whilst improve the SC resiliency in response to inherent environmental uncertainties. This paper defines ta...

Criticality safety of spent fuel casks considering water inleakage

International Nuclear Information System (INIS)

Osgood, N.L.; Withee, C.J.; Easton, E.P.

2004-01-01

A fundamental safety design parameter for all fissile material packages is that a single package must be critically safe even if water leaks into the containment system. In addition, criticality safety must be assured for arrays of packages under normal conditions of transport (undamaged packages) and under hypothetical accident conditions (damaged packages). The U.S. Nuclear Regulatory Commission staff has revised the review protocol for demonstrating criticality safety for spent fuel casks. Previous review guidance specified that water inleakage be considered under accident conditions. This practice was based on the fact that the leak tightness of spent fuel casks is typically demonstrated by use of structural analysis and not by physical testing. In addition, since a single package was shown to be safe with water inleakage, it was concluded that this analysis was also applicable to an array of damaged packages, since the heavy shield walls in spent fuel casks neutronically isolate each cask in the array. Inherent in this conclusion is that the fuel assembly geometry does not change significantly, even under drop test conditions. Requests for shipping fuel with burnup exceeding 40 GWd/MTU, including very high burnups exceeding 60 GWD/MTU, caused a reassessment of this assumption. Fuel cladding structural strength and ductility were not clearly predictable for these higher burnups. Therefore the single package analysis for an undamaged package may not be applicable for the damaged package. NRC staff developed a new practice for review of spent fuel casks under accident conditions. The practice presents two methods for approval that would allow an assessment of potential reconfiguration of the fuel assembly under accident conditions, or, alternatively, a demonstration of the water-exclusion boundary through physical testing

Risk and safety perception on urban and rural roads: Effects of environmental features, driver age and risk sensitivity.

Science.gov (United States)

Cox, Jolene A; Beanland, Vanessa; Filtness, Ashleigh J

2017-10-03

The ability to detect changing visual information is a vital component of safe driving. In addition to detecting changing visual information, drivers must also interpret its relevance to safety. Environmental changes considered to have high safety relevance will likely demand greater attention and more timely responses than those considered to have lower safety relevance. The aim of this study was to explore factors that are likely to influence perceptions of risk and safety regarding changing visual information in the driving environment. Factors explored were the environment in which the change occurs (i.e., urban vs. rural), the type of object that changes, and the driver's age, experience, and risk sensitivity. Sixty-three licensed drivers aged 18-70 years completed a hazard rating task, which required them to rate the perceived hazardousness of changing specific elements within urban and rural driving environments. Three attributes of potential hazards were systematically manipulated: the environment (urban, rural); the type of object changed (road sign, car, motorcycle, pedestrian, traffic light, animal, tree); and its inherent safety risk (low risk, high risk). Inherent safety risk was manipulated by either varying the object's placement, on/near or away from the road, or altering an infrastructure element that would require a change to driver behavior. Participants also completed two driving-related risk perception tasks, rating their relative crash risk and perceived risk of aberrant driving behaviors. Driver age was not significantly associated with hazard ratings, but individual differences in perceived risk of aberrant driving behaviors predicted hazard ratings, suggesting that general driving-related risk sensitivity plays a strong role in safety perception. In both urban and rural scenes, there were significant associations between hazard ratings and inherent safety risk, with low-risk changes perceived as consistently less hazardous than high

Gas-cooled reactor safety and accident analysis

International Nuclear Information System (INIS)

1985-12-01

The Specialists' Meeting on Gas-Cooled Reactor Safety and Accident Analysis was convened by the International Atomic Energy Agency in Oak Ridge on the invitation of the Department of Energy in Washington, USA. The meeting was hosted by the Oak Ridge National Laboratory. The purpose of the meeting was to provide an opportunity to compare and discuss results of safety and accident analysis of gas-cooled reactors under development, construction or in operation, to review their lay-out, design, and their operational performance, and to identify areas in which additional research and development are needed. The meeting emphasized the high safety margins of gas-cooled reactors and gave particular attention to the inherent safety features of small reactor units. The meeting was subdivided into four technical sessions: Safety and Related Experience with Operating Gas-Cooled Reactors (4 papers); Risk and Safety Analysis (11 papers); Accident Analysis (9 papers); Miscellaneous Related Topics (5 papers). A separate abstract was prepared for each of these papers

Inherent Limitations in Mid-Wave and Long-Wave-IR Upconversion Detector

DEFF Research Database (Denmark)

Barh, Ajanta; Tseng, Yu-Pei; Pedersen, Christian

2017-01-01

Inherent limitations in terms of optical losses, selection of nonlinear crystal(s), detection efficiency and pumping conditions in mid-wave (3-5 µm) and long-wave (8-12 µm) infrared frequency upconversion modules are investigated in this paper.......Inherent limitations in terms of optical losses, selection of nonlinear crystal(s), detection efficiency and pumping conditions in mid-wave (3-5 µm) and long-wave (8-12 µm) infrared frequency upconversion modules are investigated in this paper....

Patient safety: Safety culture and patient safety ethics

DEFF Research Database (Denmark)

Madsen, Marlene Dyrløv

2006-01-01

,demonstrating significant, consistent and sometimes large differences in terms of safety culture factors across the units participating in the survey. Paper 5 is the results of a study of the relation between safety culture, occupational health andpatient safety using a safety culture questionnaire survey......Patient safety - the prevention of medical error and adverse events - and the initiative of developing safety cultures to assure patients from harm have become one of the central concerns in quality improvement in healthcare both nationally andinternationally. This subject raises numerous...... challenging issues of systemic, organisational, cultural and ethical relevance, which this dissertation seeks to address through the application of different disciplinary approaches. The main focus of researchis safety culture; through empirical and theoretical studies to comprehend the phenomenon, address...

New trends in safety approach for commercial LMFBRS after SPX1

International Nuclear Information System (INIS)

Bergeonneau, P.; Moreau, J.; Cowking, C.B.; Friedel, G.; Pezzxilli, M.

1988-01-01

The experience gained from SPX1 project safety studies shows the trends for the definition of the new safety approach for the next generation of commercial LMFBR's. New trends in safety criteria, as seen in Europe, are presented in the first part of this paper. It is shown that they greatly emphasize the prevention actions even for minor events which can, in certain cases, lead to severe accidents. In the second part, an attempt is made to compare these new trends in Europe with the ones developed in the USA that put forward the inherent safety approach

Radiation safety audit

International Nuclear Information System (INIS)

Kadadunna, K.P.I.K.; Mod Ali, Noriah

2008-01-01

Audit has been seen as one of the effective methods to ensure harmonization in radiation protection. A radiation safety audit is a formal safety performance examination of existing or future work activities by an independent team. Regular audit will assist the management in its mission to maintain the facilities environment that is inherently safe for its employees. The audits review the adequacy of facilities for the type of use, training, and competency of workers, supervision by authorized users, availability of survey instruments, security of radioactive materials, minimization of personnel exposure to radiation, safety equipment, and the required record keeping. All approved areas of use are included in these periodic audits. Any deficiency found in the audit shall be corrected as soon as possible after they are reported. Radiation safety audit is a proactive approach to improve radiation safety practices and identify and prevent any potential radiation accident. It is an excellent tool to identify potential problem to radiation users and to assure that safety measures to eliminate or reduce the problems are fully considered. Radiation safety audit will help to develop safety culture of the facility. It is intended to be the cornerstone of a safety program designed to aid the facility, staff and management in maintaining a safe environment in which activities are carried out. The initiative of this work is to evaluate the need of having a proper audit as one of the mechanism to manage the safety using ionizing radiation. This study is focused on the need of having a proper radiation safety audit to identify deviations and deficiencies of radiation protection programmes. It will be based on studies conducted on several institutes/radiation facilities in Malaysia in 2006. Steps will then be formulated towards strengthening radiation safety through proper audit. This will result in a better working situation and confidence in the radiation protection community

The safety characteristics of the HTR 500 reactor plant

International Nuclear Information System (INIS)

Wachholz, W.

1987-01-01

The HTR is a reactor having a passive safety. It is equipped with the usual active engineered safety systems in simplified form. Due to its inherent safety characteristics and the burst-safe prestressed concrete reactor vessel activity containment is ensured even without the effect of active safety systems. Even in the event of extremely hypothetical accidents the effect on the environment is low enough so that evacuation or relocation of the population is not required. Therefore large-scale damage of agricultural land and industrially used areas is safely ruled out. Thus the site selection for this type of reactor is not restricted i.e. an HTR can be constructed near industrial and urban center. (author)

Safety Analysis for Medium/Small Size Integral Reactor: Evaluation of Safety Characteristics for Small and Medium Integral Reactor

Energy Technology Data Exchange (ETDEWEB)

Kim, Hho jung; Seul, K W; Ahn, S K; Bang, Y S; Park, D G; Kim, B K; Kim, W S; Lee, J H; Kim, W K; Shim, T M; Choi, H S; Ahn, H J; Jung, D W; Kim, G I; Park, Y M; Lee, Y J [Korea Inst. of Nuclear Safety, Taejon (Korea, Republic of)

1997-07-01

The Small and medium integral reactor is developed to be utilized for non-electric areas such as district heating and steam production for desalination and other industrial purposes, and then these applications may typically imply a closeness between the reactor and the user. It requires the reactor to be designed with the adoption of special functional and inherent safety features to ensure and promote a high level of safety and reliability, in comparison with the existing nuclear power plants. The objective of the present study is to establish the bases for the development of regulatory requirements and technical guides to address the special safety characteristics of the small and medium integral reactor. In addition, the study aims to identify and to propose resolutions to the possible safety concerns in the design of the small and medium integral reactor. 34 refs., 20 tabs. (author)

Is it about "pink" or about "girls"? The inherence heuristic across social and nonsocial domains.

Science.gov (United States)

Kinzler, Katherine D; Sullivan, Kathleen R

2014-10-01

The inherence heuristic provides an intriguing and novel explanation for early thought in a variety of domains. Exploring similarities and differences in inherent reasoning across social and nonsocial domains can help us understand the role that inherent thinking plays in the development of human reasoning and the process by which more elaborate essentialist reasoning develops.

Bayesian estimation inherent in a Mexican-hat-type neural network

Science.gov (United States)

Takiyama, Ken

2016-05-01

Brain functions, such as perception, motor control and learning, and decision making, have been explained based on a Bayesian framework, i.e., to decrease the effects of noise inherent in the human nervous system or external environment, our brain integrates sensory and a priori information in a Bayesian optimal manner. However, it remains unclear how Bayesian computations are implemented in the brain. Herein, I address this issue by analyzing a Mexican-hat-type neural network, which was used as a model of the visual cortex, motor cortex, and prefrontal cortex. I analytically demonstrate that the dynamics of an order parameter in the model corresponds exactly to a variational inference of a linear Gaussian state-space model, a Bayesian estimation, when the strength of recurrent synaptic connectivity is appropriately stronger than that of an external stimulus, a plausible condition in the brain. This exact correspondence can reveal the relationship between the parameters in the Bayesian estimation and those in the neural network, providing insight for understanding brain functions.

Safety concept of high-temperature reactors based on the experience with AVR and THTR

International Nuclear Information System (INIS)

Wachholz, Winfried; Kroeger, Wolfgang

1990-01-01

In the Federal Republic of Germany a reactor is considered safe if verification has been furnished that the requirements contained in paragraph 7 of the Federal German Atomic Energy Act are met for this reactor: demonstration of sufficient precautions against damage required according to the actual state of the art, and especially compliance with the dose rate limits for normal operation and accidental conditions. These requirements result in a deterministic multi-stage safety concept with specified requirements for the engineered safety systems. In recent years, proposals for enhanced safety of nuclear power reactors or a radical change in safety philosophy have been made. This is characterised by 'inherently safe', 'super safe' and similar slogans. A quantitative definition of these requirements has not yet been established, but it is clear as a common objective that the event of beyond design basis accidents evacuation, relocation, and large scale contamination of ground should not occur. As a consequence of the Chernobyl accident the safety of all the NPPs in Germany has been reviewed. This analysis was completed for the THTR reactor in 1988. The same has been done for AVR reactor. The final evaluation of the HTR specific safety features have been fully confirmed. The HTR concepts under development are based on this experience. The HTR-Modul unit is currently being designed

Inherently safe in situ uranium recovery

International Nuclear Information System (INIS)

Krumhansl, James Lee; Beauheim, Richard Louis; Brady, Patrick Vane; Arnold, Bill Walter; Kanney, Joseph F.; McKenna, Sean Andrew

2009-01-01

Expansion of uranium mining in the United States is a concern to some environmental groups and sovereign Native American Nations. An approach which may alleviate some problems is to develop inherently safe in situ uranium recovery ('ISR') technologies. Current ISR technology relies on chemical extraction of trace levels of uranium from aquifers that, once mined, can still contain dissolved uranium and other trace metals that are a health concern. Existing ISR operations are few in number; however, high uranium prices are driving the industry to consider expanding operations nation-wide. Environmental concerns and enforcement of the new 30 ppb uranium drinking water standard may make opening new mining operations more difficult and costly. Here we propose a technological fix: the development of inherently safe in situ recovery (ISISR) methods. The four central features of an ISISR approach are: (1) New 'green' leachants that break down predictably in the subsurface, leaving uranium, and associated trace metals, in an immobile form; (2) Post-leachant uranium/metals-immobilizing washes that provide a backup decontamination process; (3) An optimized well-field design that increases uranium recovery efficiency and minimizes excursions of contaminated water; and (4) A combined hydrologic/geochemical protocol for designing low-cost post-extraction long-term monitoring. ISISR would bring larger amounts of uranium to the surface, leave fewer toxic metals in the aquifer, and cost less to monitor safely - thus providing a 'win-win-win' solution to all stakeholders.

Influence of Inherent Moisture Content on the Deformation ...

African Journals Online (AJOL)

Influence of Inherent Moisture Content on the Deformation. Properties of Coconut Tissues During Mechanical Oil. Expression. *J. J. Mpagalile1 and B. Clarke2. 1Department of ... The study confirmed that moisture content has an important role in the deformation of coconut ..... A micro penetration technique for mechanical.

Evaluation of design variants for improved inherent regulation of advanced small modular reactors - 15325

International Nuclear Information System (INIS)

Vilim, R.B.; Passerini, S.

2015-01-01

This paper examines design variants that can improve inherent regulation in Advanced Small Modular Reactors (ASMR). It looks at the nature of unprotected upsets and then develops appropriate design measures to ensure that no upset can override a capability for safe self-regulation. This work adopts a reference sodium fast reactor (SFR) design to serve as a baseline for operational and safety performance and for comparison with variants on this design. The effect of design measures on plant stability is then examined. It is found that compared to full-power operation, the stability margin is reduced under islanded-operation. Islanded-operation is more likely for an ASMR deployed in a small regional electric grid with high penetration of renewable energy sources. The stability of core power production is a function of the inlet temperature coefficient, coolant transport times, and temperature-front attenuation in heat exchangers. The interaction of these phenomena with the control system is described

Preliminary safety evaluation for CSR1000 with passive safety system

International Nuclear Information System (INIS)

Wu, Pan; Gou, Junli; Shan, Jianqiang; Zhang, Bo; Li, Xiang

2014-01-01

Highlights: • The basic information of a Chinese SCWR concept CSR1000 is introduced. • An innovative passive safety system is proposed for CSR1000. • 6 Transients and 3 accidents are analysed with system code SCTRAN. • The passive safety systems greatly mitigate the consequences of these incidents. • The inherent safety of CSR1000 is enhanced. - Abstract: This paper describes the preliminary safety analysis of the Chinese Supercritical water cooled Reactor (CSR1000), which is proposed by Nuclear Power Institute of China (NPIC). The two-pass core design applied to CSR1000 decreases the fuel cladding temperature and flattens the power distribution of the core at normal operation condition. Each fuel assembly is made up of four sub-assemblies with downward-flow water rods, which is favorable to the core cooling during abnormal conditions due to the large water inventory of the water rods. Additionally, a passive safety system is proposed for CSR1000 to increase the safety reliability at abnormal conditions. In this paper, accidents of “pump seizure”, “loss of coolant flow accidents (LOFA)”, “core depressurization”, as well as some typical transients are analysed with code SCTRAN, which is a one-dimensional safety analysis code for SCWRs. The results indicate that the maximum cladding surface temperatures (MCST), which is the most important safety criterion, of the both passes in the mentioned incidents are all below the safety criterion by a large margin. The sensitivity analyses of the delay time of RCPs trip in “loss of offsite power” and the delay time of RMT actuation in “loss of coolant flowrate” were also included in this paper. The analyses have shown that the core design of CSR1000 is feasible and the proposed passive safety system is capable of mitigating the consequences of the selected abnormalities

Standardized safety management of AP1000 nuclear power plant

International Nuclear Information System (INIS)

Li Xingwen; Cao Zhiqiang; Cong Jiuyuan

2011-01-01

In 2002, China published and implemented the Law of the People's Republic of China on Work Safety and promulgated a series of guidelines and policies, which strengthened the safety management supervision. Standardization of safety, as another important step on safety supervision, comes after safety assesment and safety production licensing system, is also a permanent solution. Standardization of safety is a strategic, long term and fundamental work, which is also the basic access to achieving scientific safety management and increasing the inherent safety of an enterprise. Haiyang AP1000 nuclear power plant, adopting the modularized, 'open-top' and parallel construction means, overturned the traditional construction theory of installation work comes after the civil work and greatly shorten the construction period. At the same time, the notable increase of oversize module transportation and lifting and parallel construction raises higher demands for safety management. This article combines the characteristics and difficulties of safety management for Haiyang AP1000 nuclear power plant, puts forward ideas and methods for standardized safety management, and could also serve as reference to the safety management for other AP1000 projects. (authors)

Program management plan for the conduct of a research, development, and demonstration program for improving the safety of nuclear powerplants

International Nuclear Information System (INIS)

1981-12-01

Congress passed Public Law 96-567, Nuclear Safety Research, Development, and Demonstration Act of 1980, (hereafter referred to as the Act) to provide for an accelerated and coordinated program of light water reactor safety research, development, and demonstration to be carried out by the Department of Energy. In order to assure that this program would be compatible with the needs of Nuclear Regulatory Commission (NRC) and industry, the Department of Energy (DOE) initiated its response to Section 4 of the Act by conducting individual information gathering meetings with NRC and a wide cross section of the nuclear industry. The Department received recommendations on needs of what type of activities would and would not be appropriate for the Department to assist in satisfying these needs. Based on the evaluation of these inputs, it is concluded that the Department's ongoing Light Water Reactor (LWR) safety program is responsive to the Act. Specifically, the Department's ongoing program includes tasks in the areas of regulatory assessment, risk assessment, fission product source term, and emergency preparedness as well as providing technical assistance to the Institute of Nuclear Power Operations (INPO) to improve training of nuclear power personnel. These were among the very high priority efforts that were identified as necessary and appropriate for support by the Department

Inherent reactor power controller for a metal-fueled ALMR

International Nuclear Information System (INIS)

Wood, R.T.; Wilson, T.L. Jr.

1990-01-01

Inherent power control for metal-fueled ALMR designs involves using reactivity thermal feedback effects to control reactor power. This paper describes how, using classical control design techniques, a control system for normal load following maneuvers was deigned for a pool-type ALMR. This design provides active control of power removal in the balance of plant, direct control of selected primary and intermediate loop temperatures, and passive control of reactor power. The inherent stability of the strong, fast reactivity feedback effects bring heat production in the core into balance with the heat removal system temperatures, which are controlled to meet power demand. A simulation of the control system successfully responded to a 10% step change in power demand by changing power at an acceptable rate without causing large temperature fluctuations or exceeding thermal limits

Iser: an international inherently safe reactor concept

International Nuclear Information System (INIS)

Wakabayashi, Hiroaki

1988-01-01

Iser is a modular standardised 200-300 MWe power reactor based on the PIUS principle. It differs from PIUS in being simpler, and making full use of existing steel-vessel-based LWR technology. Iser is an inherently safe reactor concept under development in Japan. It is a generic concept, not a patented commodity, and it is expected that an international association to develop the concept will be formed. (U.K.)

COLD-SAT feasibility study safety analysis

Science.gov (United States)

Mchenry, Steven T.; Yost, James M.

1991-01-01

The Cryogenic On-orbit Liquid Depot-Storage, Acquisition, and Transfer (COLD-SAT) satellite presents some unique safety issues. The feasibility study conducted at NASA-Lewis desired a systems safety program that would be involved from the initial design in order to eliminate and/or control the inherent hazards. Because of this, a hazards analysis method was needed that: (1) identified issues that needed to be addressed for a feasibility assessment; and (2) identified all potential hazards that would need to be controlled and/or eliminated during the detailed design phases. The developed analysis method is presented as well as the results generated for the COLD-SAT system.

Information about robustness, reliability and safety in early design phases

DEFF Research Database (Denmark)

Marini, Vinicius Kaster

methods, and an industrial case to assess how the use of information about robustness, reliability and safety as practised by current methods influences concept development. Current methods cannot be used in early design phases due to their dependence on detailed design information for the identification...... alternatives. This prompts designers to reuse working principles that are inherently flawed, as they are liable to disturbances, failures and hazards. To address this issue, an approach based upon individual records of early design issues consists of comparing failures and benefits from prior working...... principles, before making a decision, and improving the more suitable alternatives through this feedback. Workshops were conducted with design practitioners to evaluate the potential of the approach and to simulate decision-making and gain feedback on a proof-of-concept basis. The evaluation has demonstrated...

NASA's aviation safety research and technology program

Science.gov (United States)

Fichtl, G. H.

1977-01-01

Aviation safety is challenged by the practical necessity of compromising inherent factors of design, environment, and operation. If accidents are to be avoided these factors must be controlled to a degree not often required by other transport modes. The operational problems which challenge safety seem to occur most often in the interfaces within and between the design, the environment, and operations where mismatches occur due to ignorance or lack of sufficient understanding of these interactions. Under this report the following topics are summarized: (1) The nature of operating problems, (2) NASA aviation safety research, (3) clear air turbulence characterization and prediction, (4) CAT detection, (5) Measurement of Atmospheric Turbulence (MAT) Program, (6) Lightning, (7) Thunderstorm gust fronts, (8) Aircraft ground operating problems, (9) Aircraft fire technology, (10) Crashworthiness research, (11) Aircraft wake vortex hazard research, and (12) Aviation safety reporting system.

Quantification of Inherent Respirable Dust Generation Potential (IRDGP) of South African Coals

CSIR Research Space (South Africa)

Phillips, H

2003-08-01

Full Text Available Advisory Committee Project Summary Project Title: Inherent Respirable Dust Generation Potential (IRDGP) of South African Coals-SIM020604 Author(s): H.R.Phillips and B. K. Belle Agency: University of Witwatersrand Report Date: July2003... Related Projects: Health 607, Sim 02-06-03 Category: Occupational Health Applied Research Occupational Hygiene Summary Project SIM020604 was formulated to determine the Inherent Respirable Dust Generation Potential (IRDGP) of various South...

Safety design and evaluation policy for future FBRs in Japan

International Nuclear Information System (INIS)

Aizawa, Kiyoto

1991-01-01

The safety policy for fast breeder reactors (FBRs) has gradually matured in accordance with the development of FBRs. The safety assessment of the Japanese prototype FBR, Monju during the licensing process accelerated the maturity and the integration of knowledge and databases. Results are expected to be reflected in the establishment of the safety design and evaluation policy for FBRs. Although the methodologies and safety policies developed for LWRs are applicable in principle to future FBRs, it is neither rational nor realistic to treat safety only with these policies. It is recommended that one should develop the methodologies and safety policies starting from understanding of the inherent safety characteristics of FBR's through safety research, plant operating experience and design work. In the last few years, some technical committees were organized in Japan and have discussed key safety issues which are specific to FBRs in order to provide preparatory reports and to establish safety standards and guidelines for future commercial FBRs. (author)

Inherent hazards, poor reporting and limited learning in the solid biomass energy sector: A case study of a wheel loader igniting wood dust, leading to fatal explosion at wood pellet manufacturer

DEFF Research Database (Denmark)

Hedlund, Frank Huess; Astad, John; Nichols, Jeffrey

2014-01-01

are insufficiently developed and give ample room for potentially erroneous subjective individual judgment. It is a contributing factor that combustible dust, although with great hazard potential, is not classified as a dangerous substance. Accidents therefore fall outside the scope of systems designed to disseminate...... biomass, the accident investigation and any learning that subsequently took place. The paper argues that learning opportunities were missed repeatedly. Significant root causes were not identified; principles of inherent safety in design were ignored; the hazardous area classification was based on flawed...... lessons learned and prevent future accidents. More attention to safety is needed for the renewable energy and environmentally friendly biomass pellet industry also to become sustainable from a worker safety perspective....

Approaches to passive safety in advanced thermal reactors

International Nuclear Information System (INIS)

Moses, D.L.

1986-01-01

Since 1980, there has been a proliferation of thermal reactor designs which incorporate passive safety features. The evolution of this trend is briefly traced, and the nature of various passive safety features is discussed with regard to how they have been incorporated into evolving design concepts. The key aspects of the passive safety features include reduced core power density, enhanced passive heat sinks, inherent assured shutdown mechanisms, elimination/minimization of potential leak paths from the primary coolant systems, enhanced robustness of fuel elements and improved coolant chemistry and component materials. An increased reliance on purely passive safety features typically translates into larger reactor structures at reduced power ratings. Proponents of the most innovative concepts seek to offset the increased costs by simplifying licensing requirements and reducing construction time

Design and safety aspects of nuclear district heating reactors

International Nuclear Information System (INIS)

Brogli, R.; Mathews, D.; Pelloni, S.

1989-01-01

Extensive studies on the rationale, the potential and the technology of nuclear district heating have been performed in Switzerland. Beside economics the safety aspects were of primary importance. Due to the high costs to transport heat the heating reactor tend to be small and therefore, minimally staffed and located close to population centers. Stringed safety rules are therefore applying. Gas cooled reactors are well suited as district heating reactors since they have due to their characteristics several inherent features, significant safety margins and a remarkable radioactivity retention potential. Some ways to mitigate the effects of water ingress and graphite corrosion are under investigation. (author). 5 refs, 3 figs

GEOSAF Part II. Demonstration of the operational and long-term safety of geological disposal facilities for radioactive waste. IAEA international intercomparison and harmonization project

Energy Technology Data Exchange (ETDEWEB)

Kumano, Yumiko; Bruno, Gerard [International Atomic Energy Agency, Vienna (Austria). Vienna International Centre; Tichauer, Michael [IRSN, Institut de Radioprotection et de Surete Nucleaire, Fontenay-aux-Roses (France); Hedberg, Bengt [Swedish Radiation Safety Authority, Stockholm (Sweden)

2015-07-01

International intercomparison and harmonization projects are one of the mechanisms developed by the IAEA for examining the application and use of safety standards, with a view to ensuring their effectiveness and working towards harmonization of approaches to the safety of radioactive waste management. The IAEA has organized a number of international projects on the safety of radioactive waste management; in particular on the issues related to safety demonstration for radioactive waste management facilities. In 2008, GEOSAF, Demonstration of The Operational and Long-Term Safety of Geological Disposal Facilities for Radioactive Waste, project was initiated. This project was completed in 2011 by delivering a project report focusing on the safety case for geological disposal facilities, a concept that has gained in recent years considerable prominence in the waste management area and is addressed in several international safety standards. During the course of the project, it was recognized that little work was undertaken internationally to develop a common view on the safety approach related to the operational phase of a geological disposal although long-term safety of disposal facility has been discussed for several decades. Upon completion of the first part of the GEOSAF project, it was decided to commence a follow-up project aiming at harmonizing approaches on the safety of geological disposal facilities for radioactive waste through the development of an integrated safety case covering both operational and long-term safety. The new project was named as GEOSAF Part II, which was initiated in 2012 initially as 2-year project, involving regulators and operators. GEOSAF Part II provides a forum to exchange ideas and experience on the development and review of an integrated operational and post-closure safety case for geological disposal facilities. It also aims at providing a platform for knowledge transfer. The project is of particular interest to regulatory

Passive safety testing at the Fast Flux Test Facility

International Nuclear Information System (INIS)

Lucoff, D.M.

1989-01-01

During 1986, the Fast Flux Test Facility (FFTF) conducted several tests designed to improve the understanding of the passive safety characteristics of an oxide-fueled liquid-metal reactor (LMR). Static and dynamic tests were performed over a broad range of power, flow, and temperature conditions that extended beyond those for normal operation. Key results of these tests are presented. Stable operation at low power with natural circulation cooling was demonstrated. A passive safety enhancement feature, the gas expansion module (GEM) was developed specifically to offset the large amount of cooldown reactivity that needs to be controlled in an oxide-fueled LMR undergoing an unprotected loss-of-flow accident. Nine GEMs were built and successfully tested in FFTF. With the reactor at 50% power (200 MW (thermal)), the main coolant pumps were turned off and the normal control rod scram response was inhibited. The GEMs and inherent core reactivity feedback mechanisms took the core subcritical with a modest peak coolant temperature transient that reached 85 degrees C above the pretransient value and always maintained a >400 degrees C margin to the sodium boiling point (910 degrees C)

System safety program plan for the Isotope Brayton Ground Demonstration System (phase I)

International Nuclear Information System (INIS)

1976-01-01

The safety engineering effort to be undertaken in achieving an acceptable level of safety in the Brayton Isotope Power System (BIPS) development program is discussed. The safety organizational relationships, the methods to be used, the tasks to be completed, and the documentation to be published are described. The plan will be updated periodically as the need arises

[THE CYTOMETRIC TECHNIQUE OF BINDING OF EOSIN-5-MALEIMIDE IN DIAGNOSTIC OF INHERENT SPHEROCYTOSIS].

Science.gov (United States)

Kuzminova, J A; Plyasunova, S A; Jogov, V V; Smetanina, N S

2016-03-01

The laboratory diagnostic of inherent spherocytosis is based on detection of spherocytes in peripheral blood, decreasing of index of sphericity, decreasing of osmotic resistance of erythrocytes. The new test of diagnostic of hereditary spherocytosis build on molecular defect was developed on the basis of binding extracellular fragments of protein of band 3 with eosin-5-maleimide (EMA-test). The study was carried out to implement comparative analysis of sensitivity and specificity of techniques applied to diagnose inherent spherocytosis. The sampling of 94 patients with various forms of anemias was analyzed All patients were applied complex clinical laboratory examination including analysis of osmotic resistance of erythrocytes, erythrocytometry and EMA-test as specific techniques of diagnostic of inherent spherocytosis. In 51 out of 94 patients (54%) decreasing of values of EMA-test was detected and in 47 patients diagnosis of inherent spherocytosis was confirmed. The standard values of EMA-test were established in 43 patients (46%) and 12 patients out of them with established diagnosis of inherent spherocytosis. Therefore, sensitivity of EMA-test made up to 79% and specificity - 80%. The most sensitive techniques of diagnostic remain osmotic resistance of erythrocytes (91%) and index of sphericity (up to 96%). But the highest specificity in this respect has EMA-test (80%). Nowadays, none of implemented techniques of diagnostic of inherent spherocytosis can be applied as a universal one. The implementation of complex examination is needed for proper diagnostic of disease.

Failure and factors of safety in piping system design

International Nuclear Information System (INIS)

Antaki, G.A.

1993-01-01

An important body of test and performance data on the behavior of piping systems has led to an ongoing reassessment of the code stress allowables and their safety margin. The codes stress allowables, and their factors of safety, are developed from limits on the incipient yield (for ductile materials), or incipient rupture (for brittle materials), of a test specimen loaded in simple tension. In this paper, we examine the failure theories introduced in the B31 and ASME III codes for piping and their inherent approximations compared to textbook failure theories. We summarize the evolution of factors of safety in ASME and B31 and point out that, for piping systems, it is appropriate to reconsider the concept and definition of factors of safety

Safety philosophies in the history of the West German nuclear industry

International Nuclear Information System (INIS)

Radkau, J.

1989-01-01

The article discusses the term 'philosophy' within the framework of the safety debate, examines the philosophy of 'inherent safety' and that of 'power plant siting away from agglomerations', as well as other concepts and approaches in safety engineering, as e.g. the maximum credible accident, the MCA and probabilistic approach, the practice-oriented safety philosophy, and human factors. Participation of the public as a legal requirement is discussed as a means of balancing the interests of various groups of the society, taking into account the possibility of abandoning a technology altogether as an ultimate consequence of the principle of participation of the public. (HSCH) [de

Demonstration of the reliability of the safety pumps

International Nuclear Information System (INIS)

Durand, J.M.

1989-01-01

POMPES GUINARD is supplying about 60% of the Nuclear pumps for the French Program. To become the specialist of Safety Related Pumps POMPES GUINARD made a lot of efforts and investments to acquire knowledge and experience. This was possible mainly with test on special loops as it is the only way for a pump manufacturer to progress by controlling hydraulics, components, bearings, mechanical seals, inducer, mechanical and hydraulic behaviour of the units in process of time. We will describe hereafter some of the typical tests which were performed during the last fifteen years

Development of safety analysis technology for integral reactor; evaluation on safety concerns of integral reactor

Energy Technology Data Exchange (ETDEWEB)

Kim, Hee Chul; Kim, Woong Sik; Lee, J. H. [Korea Institute of Nuclear Safety, Taejeon (Korea)

2002-03-01

The Nuclear Desalination Plant (NDP) is being developed to produce electricity and fresh water, and is expected to locate near population zone. In the aspect of safety, it is required to protect the public and environment from the possible releases of fission products and to prevent the fresh water from the contamination of radioactivity. Thus, in this study, the safety characteristics of the integral reactor adopting passive and inherent safety features significantly different from existing nuclear power plants were investigated. Also, safety requirements applicable to the NDP were analyzed based on the regulatory requirements for current light water reactor and advanced reactor designs, and user requirements for small-medium size reactors. Based on these analyses, some safety concerns to be considered in the design stage have been identified and discussed. They include the use of proven technology for new safety features, systematic event classification and selection, strengthening containment function, and the safety impacts on desalination-related systems. The study presents the general safety requirements applicable to licensing of an integral reactor and suggests additional regulatory requirements, which need to be developed, based on the direction to resolution of the safety concerns. The efforts to identify and technically resolve the safety concerns in the design stage will provide the early confidence of SMART safety and the technical basis to evaluate the safety to designers and reviewers in the future. Suggestion on the development of additional regulatory requirements will contribute for the regulator to taking actions for licensing of an integral reactor. 66 refs., 5 figs., 24 tabs. (Author)

Experimental quantum control landscapes: Inherent monotonicity and artificial structure

International Nuclear Information System (INIS)

Roslund, Jonathan; Rabitz, Herschel

2009-01-01

Unconstrained searches over quantum control landscapes are theoretically predicted to generally exhibit trap-free monotonic behavior. This paper makes an explicit experimental demonstration of this intrinsic monotonicity for two controlled quantum systems: frequency unfiltered and filtered second-harmonic generation (SHG). For unfiltered SHG, the landscape is randomly sampled and interpolation of the data is found to be devoid of landscape traps up to the level of data noise. In the case of narrow-band-filtered SHG, trajectories are taken on the landscape to reveal a lack of traps. Although the filtered SHG landscape is trap free, it exhibits a rich local structure. A perturbation analysis around the top of these landscapes provides a basis to understand their topology. Despite the inherent trap-free nature of the landscapes, practical constraints placed on the controls can lead to the appearance of artificial structure arising from the resultant forced sampling of the landscape. This circumstance and the likely lack of knowledge about the detailed local landscape structure in most quantum control applications suggests that the a priori identification of globally successful (un)constrained curvilinear control variables may be a challenging task.

ARAMIS project: a more explicit demonstration of risk control through the use of bow-tie diagrams and the evaluation of safety barrier performance.

Science.gov (United States)

de Dianous, Valérie; Fiévez, Cécile

2006-03-31

Over the last two decades a growing interest for risk analysis has been noted in the industries. The ARAMIS project has defined a methodology for risk assessment. This methodology has been built to help the industrialist to demonstrate that they have a sufficient risk control on their site. Risk analysis consists first in the identification of all the major accidents, assuming that safety functions in place are inefficient. This step of identification of the major accidents uses bow-tie diagrams. Secondly, the safety barriers really implemented on the site are taken into account. The barriers are identified on the bow-ties. An evaluation of their performance (response time, efficiency, and level of confidence) is performed to validate that they are relevant for the expected safety function. At last, the evaluation of their probability of failure enables to assess the frequency of occurrence of the accident. The demonstration of the risk control based on a couple gravity/frequency of occurrence is also possible for all the accident scenarios. During the risk analysis, a practical tool called risk graph is used to assess if the number and the reliability of the safety functions for a given cause are sufficient to reach a good risk control.

Development of safety analysis technology for integral reactor

Energy Technology Data Exchange (ETDEWEB)

Sim, Suk K.; Song, J. H.; Chung, Y. J. and others

1999-03-01

Inherent safety features and safety system characteristics of the SMART integral reactor are investigated in this study. Performance and safety of the SMART conceptual design have been evaluated and confirmed through the performance and safety analyses using safety analysis system codes as well as a preliminary performance and safety analysis methodology. SMART design base events and their acceptance criteria are identified to develop a preliminary PIRT for the SMART integral reactor. Using the preliminary PIRT, a set of experimental program for the thermal hydraulic separate effect tests and the integral effect tests was developed for the thermal hydraulic model development and the system code validation. Safety characteristics as well as the safety issues of the integral reactor has been identified during the study, which will be used to resolve the safety issues and guide the regulatory criteria for the integral reactor. The results of the performance and safety analyses performed during the study were used to feedback for the SMART conceptual design. The performance and safety analysis code systems as well as the preliminary safety analysis methodology developed in this study will be validated as the SMART design evolves. The performance and safety analysis technology developed during the study will be utilized for the SMART basic design development. (author)

The project ANSICHT. Safety and demonstration methodology for a final repository in clay formations in Germany; Projekt ANSICHT. Sicherheits- und Nachweismethodik fuer ein Endlager im Tongestein in Deutschland. Synthesebericht

Energy Technology Data Exchange (ETDEWEB)

Jobmann, Michael; Bebiolka, Anke; Jahn, Steffen; and others

2017-03-30

Based on the status of science and technology and under consideration of international repository concepts the fundamental methodology for safety demonstration for a high-level radioactive waste final repository in clay formations Germany was developed. Basic elements of the safety concept are the geological site description and the geo-scientific long-term prognosis on future performance. Another important section is the closure and sealing concept for the mine shafts. In the frame of the project the fundamental elements were developed and documented for model regions in northern and southern Germany. Three independent safety proofs have to be performed: the demonstration of the geological barrier integrity (clay), the demonstration of the geo-technical barrier system integrity - i.e. closure constructions and backfilling of the shafts, and the radiological demonstration that the radionuclide release in the area is lower than the respective limiting value.

Radial core expansion reactivity feedback in advanced LMRs: uncertainties and their effects on inherent safety

International Nuclear Information System (INIS)

Wigeland, R.A.; Moran, T.J.

1988-01-01

An analytical model for calculating radial core expansion, based on the thermal and elastic bowing of a single subassembly at the core periphery, is used to quantify the effect of uncertainties on this reactivity feedback mechanism. This model has been verified and validated with experimental and numerical results. The impact of these uncertainties on the safety margins in unprotected transients is investigated with SASSYS/SAS4A, which includes this model for calculating the reactivity feedback from radial core expansion. The magnitudes of these uncertainties are not sufficient to preclude the use of radial core expansion reactivity feedback in transient analysis

National and international safety, safeguardability and security

International Nuclear Information System (INIS)

Wakabayashi, Hiroaki

1987-01-01

All nuclear power and fuel cycle facility development must comply with the predecided national regulation and security codes which each country's Atomic Energy Commission stipulates. Those codes will basically evolve as technologies and the social system will develop, change and shift. It is also to be noted that the IAEA's international guidelines have been adopted particularly by developing countries as a good reference for their proper establishment of their safety codes. The report first discusses the plant safety regulation of the inherently safe reactors in comparison to the existing code (or licensing guide) of the Japanese government. Then the new trend seen now in a regulatory body (the US NRC) is reviewed and a proposal of the smooth transition into the new philosophy is presented. In the second part of the paper, the fuel safeguarding and facility security (or physical protection) are discussed, because in the case of inherently safe reactors like ISER-PIUS, it seems that safety has much more to do with the safeguard and the security. In the third part, the international relevances to the security of the ISER-PIUS are discussed, because any ISER-PIUS will be meaningless unless they are used extensively and freely in any part of the world precluding the security concerns. In collaborative use of the state and international codes, regulatory guides and practices, it is evident that ISER-PIUS system can clear the requirements on all the aspects by ample margin. (Nogami, K.)

Safety Culture in Pre-operational Phases of Nuclear Power Plant Projects

Energy Technology Data Exchange (ETDEWEB)

NONE

2012-09-15

An abundance of information exists on safety culture related to the operational phases of nuclear power plants; however, pre-operational phases present unique challenges. This publication focuses on safety culture during pre-operational phases that span the interval from before a decision to launch a nuclear power programme to first fuel load. It provides safety culture insights and focuses on eight generic issues: safety culture understanding; multicultural aspects; leadership; competencies and resource competition; management systems; learning and feedback; cultural assessments; and communication. Each issue is discussed in terms of: specific challenges; desired state; approaches and methods; and examples and resources. This publication will be of interest to newcomers and experienced individuals faced with the opportunities and challenges inherent in safety culture programmes aimed at pre-operational activities.

Safety Culture in Pre-operational Phases of Nuclear Power Plant Projects

International Nuclear Information System (INIS)

2012-01-01

An abundance of information exists on safety culture related to the operational phases of nuclear power plants; however, pre-operational phases present unique challenges. This publication focuses on safety culture during pre-operational phases that span the interval from before a decision to launch a nuclear power programme to first fuel load. It provides safety culture insights and focuses on eight generic issues: safety culture understanding; multicultural aspects; leadership; competencies and resource competition; management systems; learning and feedback; cultural assessments; and communication. Each issue is discussed in terms of: specific challenges; desired state; approaches and methods; and examples and resources. This publication will be of interest to newcomers and experienced individuals faced with the opportunities and challenges inherent in safety culture programmes aimed at pre-operational activities.

On the rationale of resilience in the domain of safety: A literature review

International Nuclear Information System (INIS)

Bergström, Johan; Winsen, Roel van; Henriqson, Eder

2015-01-01

Resilience is becoming a prevalent agenda in safety research and organisational practice. In this study we examine how the peer-reviewed safety science literature (a) formulates the rationale behind the study of resilience; (b) constructs resilience as a scientific object; and (c) constructs and locates the resilient subject. The results suggest that resilience engineering scholars typically motivate the need for their studies by referring to the inherent complexities of modern socio-technical systems; complexities that make these systems inherently risky. The object of resilience then becomes the capacity to adapt to such emerging risks in order to guarantee the success of the inherently risky system. In the material reviewed, the subject of resilience is typically the individual, either at the sharp end or at higher managerial levels. The individual is called-upon to adapt in the face of risk to secure the continuous performance of the system. Based on the results from how resilience has been introduced in safety sciences we raise three ethical questions for the field to address: (1) should resilience be seen as people thriving despite of, or because of, risk?; (2) should resilience theory form a basis for moral judgement?; and finally (3) how much should resilience be approached as a trait of the individual? - Highlights: • The article reviews the object of resilience in the safety science literature. • The literature offers a clear link between the notions of complexity and danger. • Danger is managed through adaptive capacity (resilience), typically at the sharp end. • The ethical implications of accepting danger at the sharp end need to be debated

Indispensable role of biochar-inherent mineral constituents in its environmental applications: A review.

Science.gov (United States)

Xu, Xiaoyun; Zhao, Yinghao; Sima, Jingke; Zhao, Ling; Mašek, Ondřej; Cao, Xinde

2017-10-01

Biochar typically consists of both carbon and mineral fractions, and the carbon fraction has been generally considered to determine its properties and applications. Recently, an increasing body of research has demonstrated that mineral components inherent in biochar, such as alkali or alkaline earth metals in the form of carbonates, phosphates, or oxides, could also influence the properties and thus the applications. The review articles published thus far have mainly focused on multiple environmental and agronomic applications of biochar, including carbon sequestration, soil improvement, environmental remediation, etc. This review aims to highlight the indispensable role of the mineral fraction of biochar in these different applications, especially in environmental applications. Specifically, it provides a critical review of current research findings related to the mineral composition of biochar and the effect of the mineral fraction on the physicochemical properties, contaminant sorption, carbon retention and stability, and nutrient bioavailability of biochar. Furthermore, the role of minerals in the emerging applications of biochar, as a precursor for fuel cells, supercapacitors, and photoactive components, is also summarized. Overall, inherent minerals should be fully considered while determining the most appropriate application for any given biochar. A thorough understanding of the role of biochar-bound minerals in different applications will also allow the design or selection of the most suitable biochar for specific applications based on the consideration of feedstock composition, production parameters, and post-treatment. Copyright © 2017 Elsevier Ltd. All rights reserved.

Inherent veilige 80 km/uur-wegen : ontwikkeling van een strategie voor een duurzaam-veilige (her)inrichting van doorgaande 80 km/uur-wegen. Deel I: keuze van de relevante wegen en het opstellen van criteria en eisen.

NARCIS (Netherlands)

Minnen, J. van

1993-01-01

The former Dienst Verkeerskunde of Rijkswaterstaat commissioned SWOV Institute for Road Safety Research to develop a strategy for an "inherent safe" reconstruction of 80 km/h through roads in the Netherlands. The first part of this study is the conceptualization phase. It classifies 80 km/h through

SAFR: a marriage of safety and innovation in LMR design

International Nuclear Information System (INIS)

Lancet, R.T.; Mills, J.C.

1985-01-01

The Sodium Advanced Fast Reactor (SAFR) is a natural evolution of earlier designs, given the current economic and licensing environment. Stringent safety and economic goals have been established for the SAFR plant. This paper describes how these goals are being satisfied, with the primary emphasis being placed on safety. The top level safety goals are: (a) to provide inherently safe responses to all credible events (b) to minimize the potential for severe accidents, and (c) to eliminate the need for evacuation, (d) limited financial risk, (e) assured investment protection, (f) minimum development risk, (g) high capacity factor, (h) long plant life, and (i) low personnel radiation exposure

Liver-inherent immune system: its role in blood-stage malaria.

Science.gov (United States)

Wunderlich, Frank; Al-Quraishy, Saleh; Dkhil, Mohamed A

2014-01-01

The liver is well known as that organ which is obligately required for the intrahepatocyte development of the pre-erythrocytic stages of the malaria-causative agent Plasmodium. However, largely neglected is the fact that the liver is also a central player of the host defense against the morbidity- and mortality-causing blood stages of the malaria parasites. Indeed, the liver is equipped with a unique immune system that acts locally, however, with systemic impact. Its main "antipodal" functions are to recognize and to generate effective immunoreactivity against pathogens on the one hand, and to generate tolerance to avoid immunoreactivity with "self" and harmless substances as dietary compounds on the other hand. This review provides an introductory survey of the liver-inherent immune system: its pathogen recognition receptors including Toll-like receptors (TLRs) and its major cell constituents with their different facilities to fight and eliminate pathogens. Then, evidence is presented that the liver is also an essential organ to overcome blood-stage malaria. Finally, we discuss effector responses of the liver-inherent immune system directed against blood-stage malaria: activation of TLRs, acute phase response, phagocytic activity, cytokine-mediated pro- and anti-inflammatory responses, generation of "protective" autoimmunity by extrathymic T cells and B-1 cells, and T cell-mediated repair of liver injuries mainly produced by malaria-induced overreactions of the liver-inherent immune system.

DEM Simulation of Biaxial Compression Experiments of Inherently Anisotropic Granular Materials and the Boundary Effects

Directory of Open Access Journals (Sweden)

Zhao-Xia Tong

2013-01-01

Full Text Available The reliability of discrete element method (DEM numerical simulations is significantly dependent on the particle-scale parameters and boundary conditions. To verify the DEM models, two series of biaxial compression tests on ellipse-shaped steel rods are used. The comparisons on the stress-strain relationship, strength, and deformation pattern of experiments and simulations indicate that the DEM models are able to capture the key macro- and micromechanical behavior of inherently anisotropic granular materials with high fidelity. By using the validated DEM models, the boundary effects on the macrodeformation, strain localization, and nonuniformity of stress distribution inside the specimens are investigated using two rigid boundaries and one flexible boundary. The results demonstrate that the boundary condition plays a significant role on the stress-strain relationship and strength of granular materials with inherent fabric anisotropy if the stresses are calculated by the force applied on the wall. However, the responses of the particle assembly measured inside the specimens are almost the same with little influence from the boundary conditions. The peak friction angle obtained from the compression tests with flexible boundary represents the real friction angle of particle assembly. Due to the weak lateral constraints, the degree of stress nonuniformity under flexible boundary is higher than that under rigid boundary.

Safety of nuclear power in the decades ahead

International Nuclear Information System (INIS)

Niehaus, F.

1991-01-01

Technological advances and the implementation of a safety culture will achieve a safety level in future reactors of the present generation of a core melt probability of less than 10 -5 per year, and less than 10 -6 per year for large releases of radioactive materials. There are older reactors which do not comply with present safety thinking. The paper reviews findings of a recent design review of WWER 440/230 plants. Advanced evolutionary designs might be capable of reducing the probability of significant off-site releases to less than 10 -7 per year. For such reactors, there are inherent limitations to further increases in safety, due to the human element, complexity of design, and capability of the containment function. Therefore, revolutionary designs are being explored to eliminate the potential for off-site releases. In this context, it seems to be advisable to examine concepts in which the fuel is the ultimate safety barrier. (orig./HP) [de

Applying ethnography to the study of context in healthcare quality and safety.

Science.gov (United States)

Leslie, Myles; Paradis, Elise; Gropper, Michael A; Reeves, Scott; Kitto, Simon

2014-02-01

Translating and scaling healthcare quality improvement (QI) and patient safety interventions remains a significant challenge. Context has been identified as a major factor in this. QI and patient safety research have begun to focus on context, with ethnography seen as a promising methodology for understanding the professional, organisational and cultural aspects of context. While ethnography is used to investigate the context of a variety of QI and safety interventions, the challenges inherent in effectively importing a qualitative methodology and its social science practitioners into this work have been largely unexamined. We explain ethnography as a research practice grounded in theory and dependent on observations gathered and interpreted in particular ways. We then review the approach of health services literature to evaluating this sort of qualitative research. Although the study of context is an interest shared by both social scientists and healthcare QI and safety researchers, we identify three key points at which those 'exporting' ethnography as a methodology and those 'importing' it to deal with QI and safety challenges may diverge. We describe perspectival divergences on the methodology's mission, form and scale. At the level of mission we demonstrate how ethnography has been adapted to a 'describe and feed back' role in the service of QI. At the level of form, we show how the long-term embedded observation at the heart of ethnography can be adapted only so far to accommodate QI interests if both data quality and ethical standards are to be upheld. Finally, at the level of scale, we demonstrate one ethnographic study design that balances breadth of exposure with depth of experience in its observations and so generates a particular type of scalable findings. The effective export of ethnography into QI and safety research requires discussion and negotiation between social scientific and health services research perspectives, as well as creative approaches

Perspective channel-type reactor with enhanced safety

International Nuclear Information System (INIS)

Adamov, E.O.; Grozdov, I.I.; Kuznetsov, S.P.; Petrov, A.A.; Rozhdestvensky, M.I.; Cherkashov, Yu.M.

1994-01-01

Following the search for new design solutions to develop within the framework of channel trends the reactor with enhanced safety the Research and Development Institute of Power Engineering has developed the design of the multiloop boiling water reactor (MKER). The MKER enhanced safety is attained when involving the inherent safety features, passive safety systems as well as the accident consequences confinement devices. The design realizes several advantages which are typical of the channel-type reactors, namely: The design desintegration simplifying the manufacture, control, equipment delivery and decreasing, versus the pressure vessel reactors, the accident effect if it proceeds in an explosive manner; small operating reactivity margin and fuel burnup increased due to continuous refuelling; fuel cycle flexibility allowing comparatively easily to adopt the reactor to the conjuncture of the country fuel balance; multiloop circuit of the main coolant which reduces the degree and effect of the accidents connected with the equipment and pipings rupture; monitoring of the channels and fuel assemblies leak-tightness. (orig.)

Molten salt reactors - safety options galore

International Nuclear Information System (INIS)

Gat, U.; Dodds, H.L.

1997-01-01

Safety features and attributes of molten salt reactors (MSR) are described. The unique features of fluid fuel reactors of on-line continuous processing and the ability for so-called external cooling result in simple and safe designs with low excess reactivity, low fission product inventory, and small source term. These, in turn, make a criticality accident unlikely and reduce the severity of a loss of coolant to where they are no longer severe accidents. A melt down is not an accident for a reactor that uses molten fuel. The molten salts are stable, non-reactive and efficient heat transfer media that operate at high temperatures at low pressures and are highly compatible with selected structural materials. All these features reduce the accident plethora. Freeze valves can be used for added safety. An ultimate safe reactor (U.S.R) is described with safety features that are passive, inherent and non-tamperable (PINT)

Summary of LWR safety research in the USA

International Nuclear Information System (INIS)

Murley, T.E.; Tong, L.S.; Bennett, G.L.

1977-01-01

The U.S. Nuclear Regulatory Commission's water reactor safety research program is described and the basic results are presented. The USNRC water reactor safety research program consists of five basic research areas: integrity of vessel and piping, thermal-hydraulic test, fuel rod behaviour, code development and verification, and reactor operational safety. Results from the vessel and piping integrity research have demonstrated the high safety margins in scaled vessels and the analytical procedures for calculating vessel behaviour under pressure. Non-destructive examination techniques are being improved. Work is also proceeding to define the material constituents to reduce the susceptibility of irradiation embrittlement and stress corrosion cracking. The thermal-hydraulic tests have covered the various phases of a hypothetical loss of coolant accident (LOCA) and activation of the emergency core cooling system (ECCS). These tests have led to the development of engineering correlations to describe the phenomena to further quantify the safety margins in commercial nuclear power plants. Specifically, this paper presents selected experimental data and analytical predictions from the initial tests in LOFT and SEMISCALE. Comparisons and evaluations are made between the data and analytical predictions. Significant results and conclusions are presented regarding the behaviour of emergency core cooling systems in a LOCA environment: the ability to predict LOCA-type experiments over a scaling range of thirty and the thermal-hydraulic behaviour of components such as pumps in an integral system LOCA environment. The fuel behaviour research has provided valuable information on decay heat, cladding oxidation, fuel rod behaviour and fuel metling. Both the decay heat and the cladding oxidation have been shown to be lower than assumed in the licensing evaluations. The fuel behaviour and thermo-hydraulic research is being integrated into computer codes to be used to provide additional

The Fort McMurray Demonstration Project in social marketing: health- and safety-related behaviour among oil sands workers.

Science.gov (United States)

Guidotti, T L; Watson, L; Wheeler, M; Jhangri, G S

1996-08-01

This is the first round in a series of surveys conducted in Fort McMurray as part of the Fort McMurray Demonstration Project in social marketing. This component of the survey was intended to focus on the most prominent group of employed workers in the community and to compare their patterns of response with the community as a whole. Respondents to the survey were overwhelmingly male (96%), married (72.9%) and living in households of two to five persons (87.9%). They were predominantly aged 30-44 (55%) and graduates of high school (53.5%). Younger male workers (below age 30) were more likely to have a high school diploma (78.3%) or some additional technical or vocational training (21.7% compared to 12.5% overall) and to be unmarried or separated. Attitudes toward safety-related behaviours were stronger than for respondents from the community as a whole. Approximately 70-100% of all age groups and both sexes showed strong agreement with attitudes involving child car seats and the unacceptability of drinking and driving. These attitudes include strong advocacy of vigorous enforcement of occupational health and safety standards. However, they showed a variability similar to the community as a whole in behaviour at home compared to work, generally reporting more consistent use of personal protection on the job than in their own homes, particularly hearing protection. Even so, they were much less likely to perform stretching and warm-up exercises prior to exertion than community residents in general. The potential may exist to transfer the technology and attitudes from workplace health and safety to community safety. One possible strategy to accomplish this is to involve workers in this industry directly in community initiatives. This strategy may be generalizable to any community in which there are major employers who place a heavy emphasis on risk control and occupational health and safety.

Development of Probabilistic Safety Assessment with respect to the first demonstration nuclear power plant of high temperature gas cooled reactor in China

International Nuclear Information System (INIS)

Tong Jiejuan; Zhao Jun; Liu Tao; Xue Dazhi

2012-01-01

Due to the unique concept of HTR-PM (High Temperature Gas Cooled Reactor-Pebble Bed Module) design, Chinese nuclear authority has anticipated that HTR-PM will bring challenge to the present regulation. The pilot use of PSA (Probabilistic Safety Assessment) during HTR-PM design and safety review is deemed to be the necessary and efficient tool to tackle the problem, and is actively encouraged as indicated in the authority's specific policy statement on HTR-PM project. The paper summarizes the policy statement to set up the base of PSA development and application activities. The up-to-date status of HTR-PM PSA development and the risk-informed application activities are introduced in this paper as the follow-up response to the policy statement. For open discussion, the paper hereafter puts forward several technical issues which have been encountered during HTR-PM PSA development. Since HTR-PM PSA development experience has the general conclusion that many of the PSA elements can be and have been implemented successfully by the traditional PSA techniques, only the issues which extra innovative efforts may be needed are highlighted in this paper. They are safety goal and risk metrics, PSA modeling framework for the non-water reactors, passive system reliability evaluation, initiating events frequencies and component reliability data estimation techniques for the new reactors and so on. The paper presents the way in which the encountered technical issues were or will be solved, although the proposed way may not be the ultimate best solution. The paper intends to express the standpoint that although the PSA of new reactor has the inherent weakness due to the insufficient information and larger data uncertainty, the problem of component reliability data is much less severe than people have conceived. The unique design conception and functional features of the reactors can influence the results more significantly than the component reliability data. What we are benefited

An Innovative Hybrid Loop-Pool SFR Design and Safety Analysis Methods: Today and Tomorrow

International Nuclear Information System (INIS)

Hongbin Zhang; Haihua Zhao; Vincent Mousseau

2008-01-01

Investment in commercial sodium cooled fast reactor (SFR) power plants will become possible only if SFRs achieve economic competitiveness as compared to light water reactors and other Generation IV reactors. Toward that end, we have launched efforts to improve the economics and safety of SFRs from the thermal design and safety analyses perspectives at Idaho National Laboratory. From the thermal design perspective, an innovative hybrid loop-pool SFR design has been proposed. This design takes advantage of the inherent safety of a pool design and the compactness of a loop design to further improve economics and safety. From the safety analyses perspective, we have initiated an effort to develop a high fidelity reactor system safety code

Development of Basic Key Technologies for Gen IV SFR Safety Evaluation

International Nuclear Information System (INIS)

Jeong, Hae Yong; Kwon, Young Min; Kim, Tae Woon; Park, Soo Yong; Suk, Soo Dong; Lee, Kwi Lim; Lee, Yong Bum; Chang, Won Pyo; Ha, Kwi Seok; Hahn, Sang Hoon

2010-07-01

Safety issues and design requirements on control rod worth were identified through the evaluation of safety design characteristics and the preliminary safety evaluation. This results will be taken into account for the conceptual design studies of the demonstration reactor in the next stage. The Level-1 Pasa has been performed and a quantitative Cdf value was produced for the selected design from the several candidates. The inherent safety characteristics of the selected design were evaluated through the DBE and ATWS analyses. A surrogate material for Tru has been selected which is applicable to the study of liquidus/solidus temperature test for the metallic fuel containing Tru. A methodology for the regression analysis with surrogate material has been developed and valuable data on metal fuel liquidus/solidus temperature have been measured. A simple mechanistic model describing a bending of subassemblies has been formulated based on the foreign test data and existing models. Its applicability has been evaluated for the Phenix design. New criteria of the core damage for the SFR PSA were identified. The list of initiating events, system response event tree, and core response event tree, which constitute a PSA methodology for an SFR, have been introduced. By developing the SFR PIRT, phenomenological model features, which have to be satisfied in a safety code, were defined and the PIRT results were applied to the design of the PDRC test facility. Bases for a safety evaluation methodology for the SFR DBEs have been also prepared. A draft version of the topical report on the code for local fault analysis has been completed. Since 2007, the MARS-LMR code has been developed and assessments for model validation with the test data from EBR-II and Phenix reactor have been continued. The code has been applied to the evaluation of passive safety of a conceptual design of Gen IV SFR

Safety performance comparation of MOX, nitride and metallic fuel based 25-100 MWe Pb-Bi cooled long life fast reactors without on-site refuelling

International Nuclear Information System (INIS)

Su'ud, Zaki

2008-01-01

In this paper the safety performance of 25-100 MWe Pb-Bi cooled long life fast reactors based on three types of fuels: MOX, nitride and metal is compared and discussed. In the fourth generation NPP paradigm, especially for Pb-Bi cooled fast reactors, inherent safety capability is necessary against some standard accidents such as unprotected loss of flow (ULOF), unprotected rod run-out transient over power (UTOP), unprotected loss of heat sink (ULOHS). Selection of fuel type will have important impact on the overall system safety performance. The results of safety analysis of long life Pb-Bi cooled fast reactors without on-site fuelling using nitride, MOX and metal fuel have been performed. The reactors show the inherent safety pattern with enough safety margins during ULOF and UTOP accidents. For MOX fuelled reactors, ULOF accident is more severe than UTOP accident while for nitride fuelled cores UTOP accident may push power much higher than that comparable MOX fuelled cores. (author)

Safety related requirements on future nuclear power plants

International Nuclear Information System (INIS)

Niehaus, F.

1991-01-01

Nuclear power has the potential to significantly contribute to the future energy supply. However, this requires continuous improvements in nuclear safety. Technological advancements and implementation of safety culture will achieve a safety level for future reactors of the present generation of a probability of core-melt of less than 10 -5 per year, and less than 10 -6 per year for large releases of radioactive materials. There are older reactors which do not comply with present safety thinking. The paper reviews findings of a recent design review of WWER 440/230 plants. Advanced evolutionary designs might be capable of reducing the probability of significant off-site releases to less than 10 -7 per year. For such reactors there are inherent limitations to increase safety further due to the human element, complexity of design and capability of the containment function. Therefore, revolutionary designs are being explored with the aim of eliminating the potential for off-site releases. In this context it seems to be advisable to explore concepts where the ultimate safety barrier is the fuel itself. (orig.) [de

The Safety Case and Safety Assessment for the Disposal of Radioactive Waste

Energy Technology Data Exchange (ETDEWEB)

NONE

2012-09-15

This Safety Guide provides guidance and recommendations on meeting the safety requirements in respect of the safety case and supporting safety assessment for the disposal of radioactive waste. The safety case and supporting safety assessment provide the basis for demonstration of safety and for licensing of radioactive waste disposal facilities and assist and guide decisions on siting, design and operations. The safety case is also the main basis on which dialogue with interested parties is conducted and on which confidence in the safety of the disposal facility is developed. This Safety Guide is relevant for operating organizations preparing the safety case as well as for the regulatory body responsible for developing the regulations and regulatory guidance that determine the basis and scope of the safety case. Contents: 1. Introduction; 2. Demonstrating the safety of radioactive waste disposal; 3. Safety principles and safety requirements; 4. The safety case for disposal of radioactive waste; 5. Radiological impact assessment for the period after closure; 6. Specific issues; 7. Documentation and use of the safety case; 8. Regulatory review process.

Development of Safety Analysis Technology for Integral Reactor

Energy Technology Data Exchange (ETDEWEB)

Sim, S. K. [Korea Atomic Energy Research Institute, Taejeon (Korea); Seul, K. W.; Kim, W. S.; Kim, W. K.; Yun, Y. G.; Ahn, H. J.; Lee, J. S.; Sin, A. D. [Korea Institute of Nuclear Safety, Taejeon (Korea)

2000-03-01

The Nuclear Desalination Plant(NDP) is being developed to produce electricity and fresh water, and is expected to locate near population zone. In the aspect of safety, it is required to protect the public and environment from the possible releases of fission products and to prevent the fresh water from the contamination of radioactivity. Thus, in a present study, the safety characteristics of the integral reactor adopting passive and inherent safety features significantly different from existing nuclear power plants were investigated based on the design of foreign and domestic integral reactors. Also, safety requirements applicable to the NDP were analyzed based on the regulatory requirements for current and advanced reactor designs, and use requirements for small-medium size reactors. Based on these analyses, some safety concerns to be considered in the design stage have been identified. They includes the use of proven technology for new safety systems, the systematic classification and selection of design basis accidents, and the safety assurance of desalination-related systems. These efforts to identify and resolve the safety concerns in the design stage will provide the early confidence of SMART safety to designers, and the technical basis to evaluate the safety to reviewers in the future. 8 refs., 20 figs., 4 tabs. (Author)

The distribution of inherent phosphorus in fifteen water treatment ...

African Journals Online (AJOL)

The aim of this investigation was to characterise the chemical properties relevant to P-sorption/desorption processes of 15 South African WTR and to determine the inherent distribution of P within the WTR using a chemical fractionation procedure. The pH, exchangeable Ca and organic carbon content ranged from 4.77 to ...

Rational quantitative safety goals: a summary

International Nuclear Information System (INIS)

Unwin, S.D.; Hayns, M.R.

1984-08-01

We introduce the notion of a Rational Quantitative Safety Goal. Such a goal reflects the imprecision and vagueness inherent in any reasonable notion of adequate safety and permits such vagueness to be incorporated into the formal regulatory decision-making process. A quantitative goal of the form, the parameter x, characterizing the safety level of the nuclear plant, shall not exceed the value x 0 , for example, is of a non-rational nature in that it invokes a strict binary logic in which the parameter space underlying x is cut sharply into two portions: that containing those values of x that comply with the goal and that containing those that do not. Here, we utilize an alternative form of logic which, in accordance with any intuitively reasonable notion of safety, permits a smooth transition of a safety determining parameter between the adequately safe and inadequately safe domains. Fuzzy set theory provides a suitable mathematical basis for the formulation of rational quantitative safety goals. The decision-making process proposed here is compatible with current risk assessment techniques and produces results in a transparent and useful format. Our methodology is illustrated with reference to the NUS Corporation risk assessment of the Limerick Generating Station

A new small HTGR power plant concept with inherently safe features--An engineering and economic challenge

International Nuclear Information System (INIS)

McDonald, C.F.; Sonn, D.L.

1983-01-01

This paper outlines a small nuclear plant concept which is not meant to replace the large nuclear power plants that will continue to be needed by the industrialized nations, but rather recognizes the needs of the smaller energy user, both for special applications in the US and for the developing nations. The small High-Temperature Gas-Cooled Reactor (HTGR), whose introduction will be very dependent on market forces, represents only one approach to meet these needs. The design of a small power plant that could be inherently safer and that might have costs less than those indicated by the traditional reverse-economy-of-scale effect is discussed. Topics considered include power plant economics, the small steam cycle HTGR thermodynamic cycle, the reactor nuclear heat source layout, the reactor heat removal system (main loop cooling, a vessel cooling system with reactor pressurized, vessel cooling system with reactor depressurized), safety considerations, investment risk protection, the technology base, and applications for the small HTGR plant concept

Basis, evidences and consequences of the inherent stellar encocooning

CERN Document Server

Celis, L

2002-01-01

Based on 7093 observations with photoelectrical photometrical measurements of 191 Mira stars, the following equations (from the papers [1] to [18]) give the basis to establish the Inherent Stellar Encocooning with the spectro-photometric characteristics of the red giant variable stars, especially the Miras, which have large amplitudes (approx 50% of giant variables). The specific basis that justifies a progressive covering with ionized molecules, cold gases, dust and grains are: The relation of the visual amplitudes A sub v =A sub r +E sub A whose real luminosity separate the intrinsic pulsation and amplitude excess effects due to the presence of molecules [145] and an opaque envelope of cool gases; The relation of the visual absolute magnitudes M sub v =M sub v sub r (P)+M sub a (delta sub T sub i sub O V) which is affected by an inherent absorption and/or occultation, and; The relation that defines the probable absolute luminosity and depends on the period and the (Sa) spectral type at maximum M sub v =-2.2...

Impact of Passive Safety on FHR Instrumentation Systems Design and Classification

International Nuclear Information System (INIS)

Holcomb, David Eugene

2015-01-01

Fluoride salt-cooled high-temperature reactors (FHRs) will rely more extensively on passive safety than earlier reactor classes. 10CFR50 Appendix A, General Design Criteria for Nuclear Power Plants, establishes minimum design requirements to provide reasonable assurance of adequate safety. 10CFR50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors, provides guidance on how the safety significance of systems, structures, and components (SSCs) should be reflected in their regulatory treatment. The Nuclear Energy Institute (NEI) has provided 10 CFR 50.69 SSC Categorization Guideline (NEI-00-04) that factors in probabilistic risk assessment (PRA) model insights, as well as deterministic insights, through an integrated decision-making panel. Employing the PRA to inform deterministic requirements enables an appropriately balanced, technically sound categorization to be established. No FHR currently has an adequate PRA or set of design basis accidents to enable establishing the safety classification of its SSCs. While all SSCs used to comply with the general design criteria (GDCs) will be safety related, the intent is to limit the instrumentation risk significance through effective design and reliance on inherent passive safety characteristics. For example, FHRs have no safety-significant temperature threshold phenomena, thus enabling the primary and reserve reactivity control systems required by GDC 26 to be passively, thermally triggered at temperatures well below those for which core or primary coolant boundary damage would occur. Moreover, the passive thermal triggering of the primary and reserve shutdown systems may relegate the control rod drive motors to the control system, substantially decreasing the amount of safety-significant wiring needed. Similarly, FHR decay heat removal systems are intended to be running continuously to minimize the amount of safety-significant instrumentation needed to initiate

ASTRID: Advanced Sodium Technological Reactor for Industrial Demonstration

International Nuclear Information System (INIS)

Vasile, A.

2012-01-01

Conclusions: • R&D results [CEA-AREVA-EDF] obtained from 2007 to 2009 have contributed to ASTRID mid 2010 choice of options; • ASTRID has the objective to demonstrate at the industrial scale progress in the identified domains of SFR weakness (safety, operability, economy). and to perform transmutation demonstrations; • A lot of improvements are related to safety; • The first very important milestone is 2012 (June 2006 French Act on wastes management): – ASTRID pre-conceptual design studies: 2010-2012; – First investment cost evaluation; – First safety Authorities advice on the orientations for ASTRID safety; • With the ASTRID program funded by the French government, France has the opportunity to develop a GEN IV Sodium Fast Reactor

Compact All Solid State Oceanic Inherent Optical Property Sensor, Phase I

Data.gov (United States)

National Aeronautics and Space Administration — Light propagation in the sea and the consequent remote sensing signals seen by aircraft and spacecraft is fundamentally governed by the inherent optical properties...

Safety studies on Korean fusion DEMO plant using integrated safety assessment methodology

International Nuclear Information System (INIS)

Oh, Kyemin; Kang, Myoung-suk; Heo, Gyunyoung; Kim, Hyoung-chan

2014-01-01

Highlights: •The purpose of this paper is to suggest methodology that can investigate safety issues and provides a case study for Korean fusion DEMO plant. •The concepts of integrated safety assessment methodology (ISAM) that can be applied in addressing regulatory requirements and recognizing safety issues for K-DEMO were emphasized. •Phenomena identification and ranking table (PIRT) was proposed. It can recognize vulnerabilities of systems and identify the gaps in technical areas requiring additional researches. •This work is expected to contribute on the conceptual design of safety features for K-DEMO to design engineers and the guidance for regulatory requirements to licensers. -- Abstract: The purpose of this paper is to suggest methodology that can investigate safety issues and provides a case study for Korean fusion DEMO plant (K-DEMO) as a part of R and D program through the National Fusion Research Institute of Korea. Even though nuclear regulation and licensing framework is well setup due to the operating and design experience of Pressurized Water Reactors (PWRs) since 1970s, the regulatory authority of South Korea has concerns on the challenge of facing new nuclear facilities including K-DEMO due to the differences in systems, materials, and inherent safety feature from conventional PWRs. Even though the follow-up of the ITER license process facilitates to deal with significant safety issues of fusion facilities, a licensee as well as a licenser should identify the gaps between ITER and DEMO in terms of safety issues. First we reviewed the methods of conducting safety analysis for unprecedented nuclear facilities such as Generation IV reactors, particularly very high temperature reactor (VHTR), which is called as integrated safety assessment methodology (ISAM). Second, the analysis for the conceptual design of K-DEMO on the basis of ISAM was conducted. The ISAM consists of five analytical tools to develop the safety requirements from licensee

Safety studies on Korean fusion DEMO plant using integrated safety assessment methodology

Energy Technology Data Exchange (ETDEWEB)

Oh, Kyemin; Kang, Myoung-suk [Kyung Hee University, Youngin-si, Gyeonggi-do 446-701 (Korea, Republic of); Heo, Gyunyoung, E-mail: gheo@khu.ac.kr [Kyung Hee University, Youngin-si, Gyeonggi-do 446-701 (Korea, Republic of); Kim, Hyoung-chan [National Fusion Research Institute, Daejeon-si 305-333 (Korea, Republic of)

2014-10-15

Highlights: •The purpose of this paper is to suggest methodology that can investigate safety issues and provides a case study for Korean fusion DEMO plant. •The concepts of integrated safety assessment methodology (ISAM) that can be applied in addressing regulatory requirements and recognizing safety issues for K-DEMO were emphasized. •Phenomena identification and ranking table (PIRT) was proposed. It can recognize vulnerabilities of systems and identify the gaps in technical areas requiring additional researches. •This work is expected to contribute on the conceptual design of safety features for K-DEMO to design engineers and the guidance for regulatory requirements to licensers. -- Abstract: The purpose of this paper is to suggest methodology that can investigate safety issues and provides a case study for Korean fusion DEMO plant (K-DEMO) as a part of R and D program through the National Fusion Research Institute of Korea. Even though nuclear regulation and licensing framework is well setup due to the operating and design experience of Pressurized Water Reactors (PWRs) since 1970s, the regulatory authority of South Korea has concerns on the challenge of facing new nuclear facilities including K-DEMO due to the differences in systems, materials, and inherent safety feature from conventional PWRs. Even though the follow-up of the ITER license process facilitates to deal with significant safety issues of fusion facilities, a licensee as well as a licenser should identify the gaps between ITER and DEMO in terms of safety issues. First we reviewed the methods of conducting safety analysis for unprecedented nuclear facilities such as Generation IV reactors, particularly very high temperature reactor (VHTR), which is called as integrated safety assessment methodology (ISAM). Second, the analysis for the conceptual design of K-DEMO on the basis of ISAM was conducted. The ISAM consists of five analytical tools to develop the safety requirements from licensee

An analytic framework for developing inherently-manufacturable pop-up laminate devices

International Nuclear Information System (INIS)

Aukes, Daniel M; Goldberg, Benjamin; Wood, Robert J; Cutkosky, Mark R

2014-01-01

Spurred by advances in manufacturing technologies developed around layered manufacturing technologies such as PC-MEMS, SCM, and printable robotics, we propose a new analytic framework for capturing the geometry of folded composite laminate devices and the mechanical processes used to manufacture them. These processes can be represented by combining a small set of geometric operations which are general enough to encompass many different manufacturing paradigms. Furthermore, such a formulation permits one to construct a variety of geometric tools which can be used to analyze common manufacturability concepts, such as tool access, part removability, and device support. In order to increase the speed of development, reduce the occurrence of manufacturing problems inherent with current design methods, and reduce the level of expertise required to develop new devices, the framework has been implemented in a new design tool called popupCAD, which is suited for the design and development of complex folded laminate devices. We conclude with a demonstration of utility of the tools by creating a folded leg mechanism. (paper)

Late-onset Alzheimer's disease is associated with inherent changes in bioenergetics profiles.

Science.gov (United States)

Sonntag, Kai-C; Ryu, Woo-In; Amirault, Kristopher M; Healy, Ryan A; Siegel, Arthur J; McPhie, Donna L; Forester, Brent; Cohen, Bruce M

2017-10-25

Body-wide changes in bioenergetics, i.e., energy metabolism, occur in normal aging and disturbed bioenergetics may be an important contributing mechanism underlying late-onset Alzheimer's disease (LOAD). We investigated the bioenergetic profiles of fibroblasts from LOAD patients and healthy controls, as a function of age and disease. LOAD cells exhibited an impaired mitochondrial metabolic potential and an abnormal redox potential, associated with reduced nicotinamide adenine dinucleotide metabolism and altered citric acid cycle activity, but not with disease-specific changes in mitochondrial mass, production of reactive oxygen species, transmembrane instability, or DNA deletions. LOAD fibroblasts demonstrated a shift in energy production to glycolysis, despite an inability to increase glucose uptake in response to IGF-1. The increase of glycolysis and the abnormal mitochondrial metabolic potential in LOAD appeared to be inherent, as they were disease- and not age-specific. Our findings support the hypothesis that impairment in multiple interacting components of bioenergetic metabolism may be a key mechanism contributing to the risk and pathophysiology of LOAD.

Mathematical Safety Assessment Approaches for Thermal Power Plants

Directory of Open Access Journals (Sweden)

Zong-Xiao Yang

2014-01-01

Full Text Available How to use system analysis methods to identify the hazards in the industrialized process, working environment, and production management for complex industrial processes, such as thermal power plants, is one of the challenges in the systems engineering. A mathematical system safety assessment model is proposed for thermal power plants in this paper by integrating fuzzy analytical hierarchy process, set pair analysis, and system functionality analysis. In the basis of those, the key factors influencing the thermal power plant safety are analyzed. The influence factors are determined based on fuzzy analytical hierarchy process. The connection degree among the factors is obtained by set pair analysis. The system safety preponderant function is constructed through system functionality analysis for inherence properties and nonlinear influence. The decision analysis system is developed by using active server page technology, web resource integration, and cross-platform capabilities for applications to the industrialized process. The availability of proposed safety assessment approach is verified by using an actual thermal power plant, which has improved the enforceability and predictability in enterprise safety assessment.

Conceptual studies of construction and safety enhancement of ocean SMART mounted on GBS

International Nuclear Information System (INIS)

Kim, Min-Gil; Lee, Kang-Heon; Kim, Seong Gu; Woo, Il-Guk; Han, Jeong-Hoon; Lee, Phill-Seung; Lee, Jeong Ik

2014-01-01

Highlights: • We suggested the concept of coupling the SMART to the GBS, and we made suggested improvements. • We describe the design concepts and GA of SMART ONPP. • We analyzed seismic feature of SMART ONPP preliminarily. • We suggested the concept of coupling the IPSS to the SMART ONPP, and we made suggested improvements. - Abstract: From the Fukushima accident, protection of NPPs from any imaginable natural disasters became very important. In this study, the authors suggest a new concept of ocean nuclear power plant (ONPP) by using SMART as a reference reactor, which is the most recent Small Modular Reactor (SMR) developed by Korea, to demonstrate that the proposed concept can improve the safety of NPP from earthquake and tsunami. The proposed concept utilizes Gravity Based Structure (GBS), which is a widely spread construction technique of offshore plants. Because, floating type or submerged type NPPs can be easily affected by severe ocean environments such as tsunamis and storms, additional safety features have to be added to the existing land based plant. In contrast, the newly proposed GBS-type ONPP does not require going through significant design modifications due to inherent characteristics of the construction method. The authors have demonstrated this concept can be applied to the large nuclear power plant in the previous work and will expand this concept for SMRs in this paper. The authors discuss the new concept by presenting design parameters, design requirements, and the new total general arrangement. Furthermore, due to the unique configuration of ONPP SMART, innovative passive safety features can be added to the existing SMART design. The performance of proposed concept to resist earthquake as well as newly added passive safety feature will be discussed by presenting simplified analysis results

Conceptual studies of construction and safety enhancement of ocean SMART mounted on GBS

Energy Technology Data Exchange (ETDEWEB)

Kim, Min-Gil, E-mail: gggggtt@kaist.ac.kr [Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, 373-1 Guseong-dong Yuseong-gu, Daejeon 305-701 (Korea, Republic of); Lee, Kang-Heon, E-mail: welcome@kaist.ac.kr [Division of Ocean Systems Engineering, Korea Advanced Institute of Science and Technology, 373-1 Guseong-dong Yuseong-gu, Daejeon 305-701 (Korea, Republic of); Kim, Seong Gu, E-mail: skim07@kaist.ac.kr [Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, 373-1 Guseong-dong Yuseong-gu, Daejeon 305-701 (Korea, Republic of); Woo, Il-Guk, E-mail: igwoo@dsme.co.kr [Department of Energy System R and D (Plant R and D), Daewoo Shipbuilding and Marine Engineering Co., Ltd., 221-17, Nonhyun-Dong, Gangnam-Gu, Seoul 135-010 (Korea, Republic of); Han, Jeong-Hoon, E-mail: jhhan1@dsme.co.kr [Department of Energy System R and D (Plant R and D), Daewoo Shipbuilding and Marine Engineering Co., Ltd., 221-17, Nonhyun-Dong, Gangnam-Gu, Seoul 135-010 (Korea, Republic of); Lee, Phill-Seung, E-mail: philseung@kaist.edu [Division of Ocean Systems Engineering, Korea Advanced Institute of Science and Technology, 373-1 Guseong-dong Yuseong-gu, Daejeon 305-701 (Korea, Republic of); Lee, Jeong Ik, E-mail: jeongiklee@kaist.ac.kr [Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, 373-1 Guseong-dong Yuseong-gu, Daejeon 305-701 (Korea, Republic of)

2014-10-15

Highlights: • We suggested the concept of coupling the SMART to the GBS, and we made suggested improvements. • We describe the design concepts and GA of SMART ONPP. • We analyzed seismic feature of SMART ONPP preliminarily. • We suggested the concept of coupling the IPSS to the SMART ONPP, and we made suggested improvements. - Abstract: From the Fukushima accident, protection of NPPs from any imaginable natural disasters became very important. In this study, the authors suggest a new concept of ocean nuclear power plant (ONPP) by using SMART as a reference reactor, which is the most recent Small Modular Reactor (SMR) developed by Korea, to demonstrate that the proposed concept can improve the safety of NPP from earthquake and tsunami. The proposed concept utilizes Gravity Based Structure (GBS), which is a widely spread construction technique of offshore plants. Because, floating type or submerged type NPPs can be easily affected by severe ocean environments such as tsunamis and storms, additional safety features have to be added to the existing land based plant. In contrast, the newly proposed GBS-type ONPP does not require going through significant design modifications due to inherent characteristics of the construction method. The authors have demonstrated this concept can be applied to the large nuclear power plant in the previous work and will expand this concept for SMRs in this paper. The authors discuss the new concept by presenting design parameters, design requirements, and the new total general arrangement. Furthermore, due to the unique configuration of ONPP SMART, innovative passive safety features can be added to the existing SMART design. The performance of proposed concept to resist earthquake as well as newly added passive safety feature will be discussed by presenting simplified analysis results.

Confined Site Construction: A qualitative investigation of critical issues affecting management of Health and Safety

OpenAIRE

Spillane, John P.; Oyedele, Lukumon O.; Von Meding, Jason; Konanahalli, Ashwini; Jaiyeoba, Babatunde E.; Tijani, Iyabo K.

2011-01-01

The construction industry is inherently risky, with a significant number of accidents and disasters occurring, particularly on confined construction sites. This research investigates and identifies the various issues affecting successful management of health and safety in confined construction sites. The rationale is that identifying the issues would assist the management of health and safety particularly in inner city centres which are mostly confined sites. Using empiricism epistemology, th...

Safety

International Nuclear Information System (INIS)

Jones, P.M.S.

1987-01-01

Aspects of fission reactors are considered - control, heat removal and containment. Brief descriptions of the reactor accidents at the SL-1 reactor (1961), Windscale (1957), Browns Ferry (1975), Three Mile Island (1979) and Chernobyl (1986) are given. The idea of inherently safe reactor designs is discussed. Safety assessment is considered under the headings of preliminary hazard analysis, failure mode analysis, event trees, fault trees, common mode failure and probabalistic risk assessments. These latter can result in a series of risk distributions linked to specific groups of fault sequences and specific consequences. A frequency-consequence diagram is shown. Fatal accident incidence rates in different countries including the United Kingdom for various industries are quoted. The incidence of fatal cancers from occupational exposure to chemicals is tabulated. Human factors and the acceptability of risk are considered. (U.K.)

The Development and Evaluation of Inherent RPCS for the APR1400

Energy Technology Data Exchange (ETDEWEB)

Choi, Jae D.; Ryu, Seok H.; Jung, Dong C.; Kim, Joon S.; Baek, Byung C.; Sung, Song K.; You, Guk J. [KNF, Daejeon (Korea, Republic of); Kim, Han G. [KHNP, Daejeon (Korea, Republic of); Chi, Sung G. [KOPEC, Daejeon (Korea, Republic of)

2008-10-15

The APR1400 RPCS (Reactor Power Cutback System) is designed to rapidly reduce the core power to eliminate the need for a reactor trip following a large load rejection or a loss of two main feedwater pumps at high power. GDC (General Design Criteria) 25 says 'Protection system requirements for reactivity control malfunctions. The protection system shall be designed to assure that SAFDL (Specified Acceptable Fuel Design Limits) are not exceeded for any single malfunction of the reactivity control systems, such as accidental withdrawal (not ejection or dropout) of control rods.' In order to comply GDC 25, CPCS (Core Protection Calculator System) will apply a big power penalty({approx} 1.3) to determine the minimum DNBR (Departure from Nucleate Boiling Ratio), the maximum LPD(Local Power Density) and reactor may immediately trip by CPCS low DNBR and high LPD during high power operation for 12-finger CEA drop. The purpose of this study is to develop the inherent RPCS to avoid unwanted reactor trips due to a 12-finger CEA drop. In order to accomplish this purpose, the CPCS should be modified to send RPCS actuation signal and not to apply power penalty due to 12-finger CEA drop for a shot period. During this period which is determined by assessment of safety, the SAFDL will not be violated without any CPCS trip function. The system and CPCS performance is evaluated to verify that CPCS trip does not occurred during 12-finger CEA drop event.

The integral fast reactor (IFR) concept: Physics of operation and safety

International Nuclear Information System (INIS)

Wade, D.C.; Chang, Y.I.

1987-01-01

The IFR concept employs a pool layout, a U/Pu/Zr metal alloy fuel and a closed fuel cycle based on pyrometallurgical reprocessing and injection casting refabrication. The reactor physics issues of designing for inherent safety and for a closed fissile self-sufficient integral fuel cycle with uranium startup and potential actinide transmutation are discussed

The integral fast reactor (IFR) concept: physics of operation and safety

International Nuclear Information System (INIS)

Wade, D.C.; Chang, Y.I.

1987-01-01

The IFR concept employs a pool layout, a U/Pu/Zr metal alloy fuel and a closed fuel cycle based on pyrometallurgical reprocessing and injection casting refabrication. The reactor physics issues of designing for inherent safety and for a closed fissile self-sufficient integral fuel cycle with uranium startup and potential actinide transmutation are discussed

Safety and Mission Assurance: A NASA Perspective

Science.gov (United States)

Higginbotham, Scott A.

2016-01-01

Manned spaceflight is an incredibly complex and inherently risky human endeavor. As the result of the lessons learned through years of triumph and tragedy, the National Aeronautics and Space Administration (NASA) has embraced a comprehensive and integrated approach to the challenge of ensuring safety and mission success. This presentation will provide an overview of some of the techniques employed in this effort, with a focus on the processing operations performed at the Kennedy Space Center (KSC).

Chimera Type Behavior in Nonlocal Coupling System with Two Different Inherent Frequencies

Science.gov (United States)

Lin, Larry; Li, Ping-Cheng; Tseng, Hseng-Che

2014-03-01

From the research of Kuramoto and Strogatz, arrays of identical oscillators can display a remarkable pattern, named chimera state, in which phase-locked oscillators coexist with drifting ones in nonlocal coupling oscillator system. We consider further in this study, two groups of oscillators with different inherent frequencies and arrange them in a ring. When the difference of the inherent frequencies is within some specific parameter range, oscillators of nonlocal coupling system show two distinct chimera states. When the parameter value exceeds some threshold value, two chimera states disappear. They show different features. The statistical dynamic behavior of the system can be described by Kuramoto theory.

Safety methodology implementation in the conceptual design phase of a fusion reactor

International Nuclear Information System (INIS)

Rodriguez-Rodrigo, L.; Elbez-Uzan, J.

2007-01-01

The licensing of ITER in France represents the first process for licensing a fusion facility in the framework of an experimental device with a total Tritium inventory of 3 kg. The main ITER parameters are far from those expected in the future demonstration reactors where the fusion power will be at least 5 times higher and the additional heating power could also reach up to 5 times the one foreseen in ITER. Main safety requirements for these reactors are based, among other conditions, on their inherent features as low amount of fuel, very low impurity content of structural materials, minimum waste repository, no active systems for safe shut-down, and no need for evacuation of population after the most severe accident. The design of such reactors is at the stage of conceptual studies and is mainly dealing with plasma performances, tritium breeding, blanket/divertor designs and solution of engineering issues, as well as bounding accidents or classification of waste. The methodological approach for integrating safety analysis as a tool for optimizing the design of the overall fusion installation for future reactors in the conceptual design phase is sketched, including the machine itself and the different auxiliary nuclear buildings. (author)

Time-frames and the demonstration of safety for HLW disposal

International Nuclear Information System (INIS)

Watkins, B.; Kessler, J.

1999-01-01

An important principle which is often embodied in the criteria for the safe disposal of long-lived radioactive wastes is that a similar level of radiation protection should be provided to future generations as that provided for those alive today. This has resulted in the development of performance assessment methodologies to evaluate the potential long term impacts of HLW disposal on humans, usually in terms of individual dose or risk. However, the actual periods of time over which it is expected that there will be full control over high level waste disposals are extremely short in comparison with the times over which radionuclides in the wastes could potentially move from the deep repository and emerge into the surface environment. This leads to problems in setting quantitative dose or risk based standard appropriate for the short and long term, and in setting the time-frames for which calculations should be carried out. This is especially difficult in view of the uncertainty in predicting changes in human behaviour and changes in the biosphere and geosphere over the time-scales involved. Different assessment time-frames and approaches proposed by IAEA, Nordic countries, Britain and US guidance documents are briefly reviewed. Whilst accepting the basic radiation protection objective of protecting future generations, no international consensus bas been agreed on what time-frames should be used in performance assessments. It is recommended that different time-frames should be associated with different quantitative or qualitative performance measures. As a result, a range of indicators of safety may be appropriate in demonstrating compliance with regulatory performance criteria and the consequent overall assessment context. It is argued that what is required is a simple, robust yet defensible approach to time-frames and performance indicators which can be accepted by the public, regulators and the nuclear industry

Probabilistic safety criteria for improvement of Nuclear Power Plant design and operation

Energy Technology Data Exchange (ETDEWEB)

Cho, Nam Jin; Chung, Woo Sick; Park, Moon Kyu [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

1991-12-15

The procedure of this study is to : research on the status of IAEA(International Atomic Energy Agency) member states about the policy of safety goals, study figures of merit and demerit that inherently exist in the existing methodology for reliability allocation, develop an efficient methodology for allocating reliability from top-level safety goals to intermediate and low-level PSC, write a computer code on the basis of the methodology proposed in the study, and apply the methodology to Surry Unit 1 that is the type of PWR.

Inherently safe in situ uranium recovery

Science.gov (United States)

Krumhansl, James L; Brady, Patrick V

2014-04-29

An in situ recovery of uranium operation involves circulating reactive fluids through an underground uranium deposit. These fluids contain chemicals that dissolve the uranium ore. Uranium is recovered from the fluids after they are pumped back to the surface. Chemicals used to accomplish this include complexing agents that are organic, readily degradable, and/or have a predictable lifetime in an aquifer. Efficiency is increased through development of organic agents targeted to complexing tetravalent uranium rather than hexavalent uranium. The operation provides for in situ immobilization of some oxy-anion pollutants under oxidizing conditions as well as reducing conditions. The operation also artificially reestablishes reducing conditions on the aquifer after uranium recovery is completed. With the ability to have the impacted aquifer reliably remediated, the uranium recovery operation can be considered inherently safe.

Geometries inherent to N=1 supergravities

International Nuclear Information System (INIS)

Galperin, A.S.; Ogievetsky, V.I.; Sokatchev, E.S.

1981-01-01

The first part of the talk is devoted to a consideration of linearized N=1 supergravities. The second main part deals with complex geometries inherent to different N=1 supergravities. A special attention is paid to a new version with local symmetry. It is connected to the special nonminimal case (n=0) having a remarkable property of supervolume preservation in Csup(4.4) superspace. Therefore the superdeterminant of change of variables from left to right-handed Rsup(4.4) parametrization is a dimensionless scalar. This geometric invariant has to be constrained to obtain an action. Solving such a constraint on vector and spinor prepotentials in Wess-Zumino gauge one obtains the new supergravity with 12+12 fields and local symmetry. A possible relaxation of this constraint is briefly considered (16+16 fields version) [ru

A study for structural safety of ISER reactor building under impact load

International Nuclear Information System (INIS)

Takeuchi, Yoichiro; Hasegawa, Toshiyasu; Mutoh, Atsushi; Wakabayashi, Hiroaki.

1991-01-01

ISER (Inherently Safe and Economical Reactor) proposed in Japan by an academic circle and industries is expected to be used world-wide particularly in developing countries where an energy crunch is feared in the 21-st century. A certain level of hardened structures for plant safety seems to be effective and may be required by the regulatory body, since the ISER is claimed to be inherently safe even against a kind of external load. This paper concerns impact resistant design of ISER. A brief state-of-the-art review on related works, impact resistant design flow and results of some preliminary analysis of a proposed ISER model is also presented. (author)

A study on the methodology of probabilistic safety assessment for KALIMER

Energy Technology Data Exchange (ETDEWEB)

Jeong, Kwan Seong; Kwon, Young Min; Lee, Yong Bum; Jeong, Hae Yong; Yang, Joon Eon; Ha, Kyu Suk; Hahn, Do Hee [Korea Atomic Energy Research Institute, Taejeon (Korea)

2002-03-01

Existing Probabilistic Safety Assessment(PSA) is a method for Light Water Reactor or Pressurized Heavy Water Reactor. Because KALIMER is different from these reactor, the new methodology of PSA need to be developed. In this paper, the PSA of Power Reactor Inherently Safety Module(PRISM) is analyzed, and Initiating Event such as Experiential Assessment, Logical Assessment and Failure Mode Effect Analysis(FMEA) is reviewed. Also, Pipe Damage Frequency Method is suggested for KALIMER. And the Reliability Physical method of Passive System, which is a chief safety system of KALIMER, is reviewed and its applicability is investigated. Finally, for the Preliminary PSA of KALIMER, Intermediate Heat Transfer System is analyzed. 23 refs., 10 figs., 13 tabs. (Author)

Safety philosophy of gas turbine high temperature reactor (GTHTR300)

International Nuclear Information System (INIS)

Shoji Katanishi; Kazuhiko Kunitomi; Shusaku Shiozawa

2002-01-01

Japan Atomic Energy Research Institute (JAERI) has undertaken the study of an original design concept of gas turbine high temperature reactor, the GTHTR300. The general concept of this study is development of a greatly simplified design that leads to substantially reduced technical and cost requirements. Newly proposed design features enable the GTHTR300 to be an efficient and economically competitive reactor in 2010's. Also, the GTHTR300 fully takes advantage of its inherent safety characteristics. The safety philosophy of the GTHTR300 is developed based on the HTTR (High Temperature Engineering Test Reactor) of JAERI which is the first HTGR in Japan. Major features of the newly proposed safety philosophy for the GTHTR300 are described in this article. (authors)

Gaseous core nuclear-driven engines featuring a self-shutoff mechanism to provide nuclear safety

International Nuclear Information System (INIS)

Heidrich, J.; Pettibone, J.; Chow, Tze-Show; Condit, R.; Zimmerman, G.

1991-11-01

Nuclear driven engines are described that could be run in either pulsed or steady state modes. In the pulsed mode nuclear energy is released by fissioning of uranium or plutonium in a supercritical assembly of fuel and working gas. In a steady state mode a fuel-gas mixture is injected into a magnetic nozzle where it is compressed into a critical state and produces energy. Engine performance is modeled using a code that calculates hydrodynamics, fission energy production, and neutron transport self-consistently. Results are given demonstrating a large negative temperature coefficient that produces self-shutoff or control of energy production. Reduced fission product inventory and the self-shutoff provide inherent nuclear safety. It is expected that nuclear engine reactor units could be scaled up from about 100 MW e

Safety features of the MAPLE-X10 reactor design

International Nuclear Information System (INIS)

Lee, A.G.; Bishop, W.E.; Heeds, W.

1990-09-01

The MAPLE-X10 reactor is a D 2 0-reflected, H 2 0-cooled and -moderated pool-type reactor under construction at the Chalk River Nuclear Laboratories. This 10-MW reactor will produce key medical and industrial radio-isotopes such as 99 Mo, 125 I, and 192 Ir. As the prototype for the MAPLE research reactor concept, the reactor incorporates diverse safety features both inherent in the design and in the added engineered systems. The safety requirements are analogous to those of the Canadian CANDU power reactor since standards for the licensing of new research reactors have not been developed yet by the licensing authority in Canada

Safety features of the MAPLE-X10 reactor design

International Nuclear Information System (INIS)

Lee, A.G.; Bishop, W.E.; Heeds, W.

1990-01-01

This paper reports on the MAPLE-X10 reactor D 2 O-reflected, H 2 O-cooled and -moderated pool- type reactor, under construction at the Chalk River Nuclear Laboratories. This 10-MW will produce key medical and industrial radioisotopes such as 99 Mo, 125 I, and 192 Ir. The prototype for the MAPLE research reactor concept, the reactor incorporates diverse safety features both inherent in the design and in the added engineered systems. The safety requirements are analogous to those of the Canadian CANDU power reactor as standards for the licensing of new research reactors have not been developed by the licensing authority in Canada

Promoting and assessment of safety culture within regulatory body

International Nuclear Information System (INIS)

Awasthi, Sumit; Bhattacharya, D.; Koley, J.; Krishnamurthy, P.R.

2015-01-01

Regulators have an important role to play in assisting organizations under their jurisdiction to develop positive safety cultures. It is therefore essential for the regulator to have a robust safety culture as an inherent strategy and communication of this strategy to the organizations it supervises. Atomic Energy Regulatory Board (AERB) emphasizes every utility to institute a good safety culture during various stages of a NPP. The regulatory requirement for establishing organisational safety culture within utility at different stages are delineated in the various AERB safety codes which are presented in the paper. Although the review and assessment of the safety culture is a part of AERB’s continual safety supervision through existing review mechanism, AERB do not use any specific indicators for safety culture assessment. However, establishing and nurturing a good safety culture within AERB helps in encouraging the utility to institute the same. At the induction level AERB provides training to its staffs for regulatory orientation which include a specific course on safety culture. Subsequently, the junior staffs are mentored by seniors while involving them in various regulatory processes and putting them as observers during regulatory decision making process. Further, AERB established a formal procedure for assessing and improving safety culture within its staff as a management system process. The paper describes as a case study the above safety culture assessment process established within AERB

Tumor inherent interferons: Impact on immune reactivity and immunotherapy.

Science.gov (United States)

Brockwell, Natasha K; Parker, Belinda S

2018-04-19

Immunotherapy has revolutionized cancer treatment, with sustained responses to immune checkpoint inhibitors reported in a number of malignancies. Such therapeutics are now being trialed in aggressive or advanced cancers that are heavily reliant on untargeted therapies, such as triple negative breast cancer. However, responses have been underwhelming to date and are very difficult to predict, leading to an inability to accurately weigh up the benefit-to-risk ratio for their implementation. The tumor immune microenvironment has been closely linked to immunotherapeutic response, with superior responses observed in patients with T cell-inflamed or 'hot' tumors. One class of cytokines, the type I interferons, are a major dictator of tumor immune infiltration and activation. Tumor cell inherent interferon signaling dramatically influences the immune microenvironment and the expression of immune checkpoint proteins, hence regulators and targets of this pathway are candidate biomarkers of immunotherapeutic response. In support of a link between IFN signaling and immunotherapeutic response, the combination of type I interferon inducers with checkpoint immunotherapy has recently been demonstrated critical for a sustained anti-tumor response in aggressive breast cancer models. Here we review evidence that links type I interferons with a hot tumor immune microenvironment, response to checkpoint inhibitors and reduced risk of metastasis that supports their use as biomarkers and therapeutics in oncology. Copyright © 2018. Published by Elsevier Ltd.

Inherent safety that the reactivity effect of core bending in fast reactors brings about

International Nuclear Information System (INIS)

Nakagawa, Masatoshi; Yagawa, Genki.

1994-01-01

FBRs have the merit on safety by low operation pressure and the large heat capacity of coolant, in addition, due to the core temperature rise at the time of accidents and the thermal expansion of core structures, the negative feedback of reactivity can be expected. Recently, attention has been paid to the negative feedback of reactivity due to core bending. It can be expected also in the core of limited free bow type. Bending is caused by the difference of thermal expansion on six surfaces of hexagonal wrapper tubes. The bending changes core reactivity and exerts effects to fuel exchange force and operation, insertion of control rods and the structural soundness of fuel assemblies. for the purpose of limiting the effect that core bending exerts to core characteristics to allowable range, core constraint mechanism is installed. The behavior of core bending at the time of anticipated transient without scram is explained. The example of the analysis of PRISM reactor is shown. The experiment that confirmed the negative feedback of reactivity due to core bending under the condition of ULOF was that at the fast flux test facility. (K.I.)

Inherent Error in Asynchronous Digital Flight Controls.

Science.gov (United States)

1980-02-01

operation will be eliminated. If T* is close to T, the inherent error (eA) is a small value. Then the deficiency of the basic model, which is de... tK2 at m ~VCONTR0RKIPPIL I~+ R tKT 2 TKT 5 ~I G E 1 i V OTN TRIL 2-REDN HNE IjT e UT 3 ~49__I 4.) 4 -4 - 4.-4 U 4.) k-4E-- Iz E-4 P E-44)-4. 4.)l 1s...indicate the channel failure. To reduce this deficiency , the new model computes a tolerance value equal to the maximum steady-state sample covariance of the

Safety Teams: An Approach to Engage Students in Laboratory Safety

Science.gov (United States)

Alaimo, Peter J.; Langenhan, Joseph M.; Tanner, Martha J.; Ferrenberg, Scott M.

2010-01-01

We developed and implemented a yearlong safety program into our organic chemistry lab courses that aims to enhance student attitudes toward safety and to ensure students learn to recognize, demonstrate, and assess safe laboratory practices. This active, collaborative program involves the use of student "safety teams" and includes…

A literature review of safety culture.

Energy Technology Data Exchange (ETDEWEB)

Cole, Kerstan Suzanne; Stevens-Adams, Susan Marie; Wenner, Caren A.

2013-03-01

Workplace safety has been historically neglected by organizations in order to enhance profitability. Over the past 30 years, safety concerns and attention to safety have increased due to a series of disastrous events occurring across many different industries (e.g., Chernobyl, Upper Big-Branch Mine, Davis-Besse etc.). Many organizations have focused on promoting a healthy safety culture as a way to understand past incidents, and to prevent future disasters. There is an extensive academic literature devoted to safety culture, and the Department of Energy has also published a significant number of documents related to safety culture. The purpose of the current endeavor was to conduct a review of the safety culture literature in order to understand definitions, methodologies, models, and successful interventions for improving safety culture. After reviewing the literature, we observed four emerging themes. First, it was apparent that although safety culture is a valuable construct, it has some inherent weaknesses. For example, there is no common definition of safety culture and no standard way for assessing the construct. Second, it is apparent that researchers know how to measure particular components of safety culture, with specific focus on individual and organizational factors. Such existing methodologies can be leveraged for future assessments. Third, based on the published literature, the relationship between safety culture and performance is tenuous at best. There are few empirical studies that examine the relationship between safety culture and safety performance metrics. Further, most of these studies do not include a description of the implementation of interventions to improve safety culture, or do not measure the effect of these interventions on safety culture or performance. Fourth, safety culture is best viewed as a dynamic, multi-faceted overall system composed of individual, engineered and organizational models. By addressing all three components of

Study on the experimental VHTR safety with analysis for a hypothetical rapid depressurization accident

International Nuclear Information System (INIS)

Mitake, S.; Suzuki, K.; Ohno, T.; Okada, T.

1982-01-01

A hypothetical rapid depressurization accident of the experimental VHTR has been analyzed, including all phenomena in the accident, from its initiating depressurization of the coolant to consequential radiological hazard. Based on reliability analysis of the engineered safety features, all possible sequences, in which the safety systems are in success or in failure, have been investigated with event tree analysis. The result shows the inherent safety characteristics of the reactor and the effectiveness of the engineered safety features. And through the analysis, it has been indicated that further investigations on some phenomena in the accident, e.g., air ingress by natural circulation flow and fission product transport in the plant, will bring forth more reasonable and sufficient safety of the reactor

The French methodology for EBS confirmation and demonstration

International Nuclear Information System (INIS)

Plas, F.; Voinis, S.; Mayer, S.

2007-01-01

The December 30, 1991 French Waste Act entrusted ANDRA, the French national agency for radioactive waste management, with the task of assessing the feasibility of deep geological disposal of high- and medium-level long-lived waste (HLW and ILW, respectively C-waste and B-waste types in French) plus spent fuel (CU in French). In that context, the 'Dossier 2005 Argile' submitted by ANDRA presents the feasibility assessment - with regard to the technical capacity to accommodate all wastes, to reversibility, and to safety - of a radioactive waste disposal in a clay formation studied at the Meuse/Haute-Marne URL. This report was built upon an iterative approach between site characterisation, design, modelling, phenomenological analysis and safety analysis, in which two principles always guided the elaboration of the safety case: the principle of robustness - repository components must maintain their functionality given reasonable solicitations, taking into account uncertainties on the nature and level of these solicitations; and the principle of demonstrability - safety must be verified without requiring complex demonstrations, and based on multiple lines of evidence/argument (numerical simulation, qualitative arguments such as use of natural analogues, experiments and technological demonstrators). In that respect, the EBS definition, demonstration and confirmation of design is a part of the overall safety case. The 'Dossier 2005 Argile' was submitted to three independent peer reviews. The aim. of this article is to present the methodology that ANDRA implemented in the context of 'Dossier 2005 Argile' for defining, demonstrating and confirming the EBS design as well as the future programme with respect with the new Act of 28 June 2006. (author)

Disease resistance is related to inherent swimming performance in Atlantic salmon.

Science.gov (United States)

Castro, Vicente; Grisdale-Helland, Barbara; Jørgensen, Sven M; Helgerud, Jan; Claireaux, Guy; Farrell, Anthony P; Krasnov, Aleksei; Helland, Ståle J; Takle, Harald

2013-01-21

Like humans, fish can be classified according to their athletic performance. Sustained exercise training of fish can improve growth and physical capacity, and recent results have documented improved disease resistance in exercised Atlantic salmon. In this study we investigated the effects of inherent swimming performance and exercise training on disease resistance in Atlantic salmon.Atlantic salmon were first classified as either poor or good according to their swimming performance in a screening test and then exercise trained for 10 weeks using one of two constant-velocity or two interval-velocity training regimes for comparison against control trained fish (low speed continuously). Disease resistance was assessed by a viral disease challenge test (infectious pancreatic necrosis) and gene expression analyses of the host response in selected organs. An inherently good swimming performance was associated with improved disease resistance, as good swimmers showed significantly better survival compared to poor swimmers in the viral challenge test. Differences in mortalities between poor and good swimmers were correlated with cardiac mRNA expression of virus responsive genes reflecting the infection status. Although not significant, fish trained at constant-velocity showed a trend towards higher survival than fish trained at either short or long intervals. Finally, only constant training at high intensity had a significant positive effect on fish growth compared to control trained fish. This is the first evidence suggesting that inherent swimming performance is associated with disease resistance in fish.

Disease resistance is related to inherent swimming performance in Atlantic salmon

Directory of Open Access Journals (Sweden)

Castro Vicente

2013-01-01

Full Text Available Abstract Background Like humans, fish can be classified according to their athletic performance. Sustained exercise training of fish can improve growth and physical capacity, and recent results have documented improved disease resistance in exercised Atlantic salmon. In this study we investigated the effects of inherent swimming performance and exercise training on disease resistance in Atlantic salmon. Atlantic salmon were first classified as either poor or good according to their swimming performance in a screening test and then exercise trained for 10 weeks using one of two constant-velocity or two interval-velocity training regimes for comparison against control trained fish (low speed continuously. Disease resistance was assessed by a viral disease challenge test (infectious pancreatic necrosis and gene expression analyses of the host response in selected organs. Results An inherently good swimming performance was associated with improved disease resistance, as good swimmers showed significantly better survival compared to poor swimmers in the viral challenge test. Differences in mortalities between poor and good swimmers were correlated with cardiac mRNA expression of virus responsive genes reflecting the infection status. Although not significant, fish trained at constant-velocity showed a trend towards higher survival than fish trained at either short or long intervals. Finally, only constant training at high intensity had a significant positive effect on fish growth compared to control trained fish. Conclusions This is the first evidence suggesting that inherent swimming performance is associated with disease resistance in fish.

Software Safety Risk in Legacy Safety-Critical Computer Systems

Science.gov (United States)

Hill, Janice L.; Baggs, Rhoda

2007-01-01

Safety Standards contain technical and process-oriented safety requirements. Technical requirements are those such as "must work" and "must not work" functions in the system. Process-Oriented requirements are software engineering and safety management process requirements. Address the system perspective and some cover just software in the system > NASA-STD-8719.13B Software Safety Standard is the current standard of interest. NASA programs/projects will have their own set of safety requirements derived from the standard. Safety Cases: a) Documented demonstration that a system complies with the specified safety requirements. b) Evidence is gathered on the integrity of the system and put forward as an argued case. [Gardener (ed.)] c) Problems occur when trying to meet safety standards, and thus make retrospective safety cases, in legacy safety-critical computer systems.

Application of system safety engineering techniques for hazard prevention at the Superconducting Super Collider

International Nuclear Information System (INIS)

Hendrix, B.L.

1991-01-01

A primary goal of the Superconducting Super Collider Laboratory (SSCL) is to establish an exemplary safety program. Achieving this goal requires leadership, planning, coordination, and technical know-how. To ensure that safety is an inherent part of the design, the Environment, Safety and Health Office employs a systems engineering discipline and process known as System Safety. The goal of System Safety - hazard prevention - is accomplished by analyzing systems to identify hazards and to evaluate design and procedural options and countermeasures to prevent, eliminate, mitigate, or control hazards and risks. Establishment of safety and human factors design criteria at the outset of the project prevents unsafe designs and safety violations, reduces risks, and helps in avoiding costly design changes later. This process requires a considerable amount of coordination with a variety of technical disciplines and safety professionals to integrate methods of hazard prevention, mitigation, and risk reduction throughout the system life-cycle

Interdependency control : compensation strategies for the inherent vulnerability of critical infrastructure networks

International Nuclear Information System (INIS)

Mao, D.; Sotoodeh, M.; Monu, K.; Marti, J.R.; Srivastava, K.D.

2009-01-01

Today's increasingly interacting national critical infrastructures (NCIs) can tolerate most stochastic local disturbances. However, they are extremely fragile under global disturbances, as the latter may either push the whole system into a critical state or reveal many unexpected hidden interdependencies, inducing or triggering cascading failures among all possible layers. This robust yet fragile duality is an inherent vulnerability of modern infrastructures. It is therefore expected that weather-related disasters will be more frequent under a changing climate. This paper proposed an interdependency control strategy (ICS) that would maintain the survival of the most critical services, and compensate for this inherent vulnerability during emergency states. The paper also proposed a generalized adjacency matrix (GAM) to represent the physical interdependencies intra/inter of various infrastructure networks. The vulnerable section in the network can be identified, based on computed results of GAM, number of islands in the network, and influence domain(s) of each component. These features render ICS more effective and convincing. Last, the paper proposed a survivability index for isolated sub-networks and described relevant measures for improving this index during the four phases of emergency management. It was concluded that the proposed strategy is an effective means to reduce the inherent vulnerability and increase the resiliency of these critical infrastructures networks. 20 refs., 5 figs

Planificación e implementación de la Seguridad Sostenible en Holanda = Planning and implementation of Sustainable Safety in The Netherlands.

NARCIS (Netherlands)

Wegman, F.C.M. Aarts, L.T. & Bax, C.A.

2008-01-01

From the early 1990s Holland has been developing a concept of sustainable road safety to tackle the risks Inherent to mobility. Subsequent to introducing the sustainable safety concept in 1992, a series of steps were taken to apply measures in line with the Dutch concept. The most important step was

Safety demonstration analyses for severe accident of fresh nuclear fuel transport packages at JAERI

International Nuclear Information System (INIS)

Yamada, K.; Watanabe, K.; Nomura, Y.; Okuno, H.; Miyoshi, Y.

2004-01-01

It is expected in the near future that more and more fresh nuclear fuel will be transported in a variety of transport packages to cope with increasing demand from nuclear fuel cycle facilities. Accordingly, safety demonstration analyses of these methods are planned and conducted at JAERI under contract with the Ministry of Economy, Trade and Industry of Japan. These analyses are conducted part of a four year plan from 2001 to 2004 to verify integrity of packaging against leakage of radioactive material in the case of a severe accident envisioned to occur during transportation, for the purpose of gaining public acceptance of such nuclear fuel activities. In order to create the accident scenarios, actual transportation routes were surveyed, accident or incident records were tracked, international radioactive material transport regulations such as IAEA rules were investigated and, thus, accident conditions leading to mechanical damage and thermal failure were selected for inclusion in the scenario. As a result, the worst-case conditions of run-off-the-road accidents were incorporated, where there is impact against a concrete or asphalt surface. Fire accidents were assumed to occur after collision with a tank truck carrying lots of inflammable material or destruction by fire after collision inside a tunnel. The impact analyses were performed by using three-dimensional elements according to the general purpose impact analysis code LS-DYNA. Leak-tightness of the package was maintained even in the severe impact accident scenario. In addition, the thermal analyses were performed by using two-dimensional elements according to the general purpose finite element method computer code ABAQUS. As a result of these analyses, the integrity of the inside packaging component was found to be sufficient to maintain a leak-tight state, confirming its safety

PX–An Innovative Safety Concept for an Unmanned Reactor

Directory of Open Access Journals (Sweden)

Sung-Jae Yi

2016-02-01

Full Text Available An innovative safety concept for a light water reactor has been developed at the Korea Atomic Energy Research Institute. It is a unique concept that adopts both a fast heat transfer mechanism for a small containment and a changing mechanism of the cooling geometry to take advantage of the potential, thermal, and dynamic energies of the cold water in the containment. It can bring about rapid cooling of the containment and long-term cooling of the decay heat. By virtue of this innovative concept, nuclear fuel damage events can be prevented. The ultimate heat transfer mechanism contributes to minimization of the heat exchanger size and containment volume. A small containment can ensure the underground construction, which can use river or seawater as an ultimate heat sink. The changing mechanism of the cooling geometry simplifies several safety systems and unifies diverse functions. Simplicity of the present safety system does not require any operator actions during events or accidents. Therefore, the unique safety concept of PX can realize both economic competitiveness and inherent safety.

Safety margins in deterministic safety analysis

International Nuclear Information System (INIS)

Viktorov, A.

2011-01-01

The concept of safety margins has acquired certain prominence in the attempts to demonstrate quantitatively the level of the nuclear power plant safety by means of deterministic analysis, especially when considering impacts from plant ageing and discovery issues. A number of international or industry publications exist that discuss various applications and interpretations of safety margins. The objective of this presentation is to bring together and examine in some detail, from the regulatory point of view, the safety margins that relate to deterministic safety analysis. In this paper, definitions of various safety margins are presented and discussed along with the regulatory expectations for them. Interrelationships of analysis input and output parameters with corresponding limits are explored. It is shown that the overall safety margin is composed of several components each having different origins and potential uses; in particular, margins associated with analysis output parameters are contrasted with margins linked to the analysis input. While these are separate, it is possible to influence output margins through the analysis input, and analysis method. Preserving safety margins is tantamount to maintaining safety. At the same time, efficiency of operation requires optimization of safety margins taking into account various technical and regulatory considerations. For this, basic definitions and rules for safety margins must be first established. (author)

Geological boundary conditions for a safety demonstration and verification concept for a HLW repository in claystone in Germany. AnSichT

Energy Technology Data Exchange (ETDEWEB)

Stark, Lena; Bebiolka, Anke; Gerardi, Johannes [Federal Institute for Geosciences and Natural Resources (BGR), Hannover (Germany). Dept. of Underground Space for Storage and Economic Use; and others

2015-07-01

Within the framework of the R and D project ''AnSichT'', DBE TECHNOLOGY, BGR and GRS are developing a method to demonstrate the safety of a HLW repository in claystone in Germany. The methodological approach basing on a holistic concept, links the legal and geologic boundary conditions, the disposal and closure concept, the demonstration of barrier integrity, and the long-term analysis of the repository evolution as well. The geologic boundary conditions are specified by the description of the geological situation and generic models, the selection of representative parameters and geoscientific long-term predictions. They form a fundament for the system analysis.

Plants status monitor: Modelling techniques and inherent benefits

International Nuclear Information System (INIS)

Breeding, R.J.; Lainoff, S.M.; Rees, D.C.; Prather, W.A.; Fickiessen, K.O.E.

1987-01-01

The Plant Status Monitor (PSM) is designed to provide plant personnel with information on the operational status of the plant and compliance with the plant technical specifications. The PSM software evaluates system models using a 'distributed processing' technique in which detailed models of individual systems are processed rather than by evaluating a single, plant-level model. In addition, development of the system models for PSM provides inherent benefits to the plant by forcing detailed reviews of the technical specifications, system design and operating procedures, and plant documentation. (orig.)

Safer Systems: A NextGen Aviation Safety Strategic Goal

Science.gov (United States)

Darr, Stephen T.; Ricks, Wendell R.; Lemos, Katherine A.

2008-01-01

The Joint Planning and Development Office (JPDO), is charged by Congress with developing the concepts and plans for the Next Generation Air Transportation System (NextGen). The National Aviation Safety Strategic Plan (NASSP), developed by the Safety Working Group of the JPDO, focuses on establishing the goals, objectives, and strategies needed to realize the safety objectives of the NextGen Integrated Plan. The three goal areas of the NASSP are Safer Practices, Safer Systems, and Safer Worldwide. Safer Practices emphasizes an integrated, systematic approach to safety risk management through implementation of formalized Safety Management Systems (SMS) that incorporate safety data analysis processes, and the enhancement of methods for ensuring safety is an inherent characteristic of NextGen. Safer Systems emphasizes implementation of safety-enhancing technologies, which will improve safety for human-centered interfaces and enhance the safety of airborne and ground-based systems. Safer Worldwide encourages coordinating the adoption of the safer practices and safer systems technologies, policies and procedures worldwide, such that the maximum level of safety is achieved across air transportation system boundaries. This paper introduces the NASSP and its development, and focuses on the Safer Systems elements of the NASSP, which incorporates three objectives for NextGen systems: 1) provide risk reducing system interfaces, 2) provide safety enhancements for airborne systems, and 3) provide safety enhancements for ground-based systems. The goal of this paper is to expose avionics and air traffic management system developers to NASSP objectives and Safer Systems strategies.

Evaluation of common mode failure of safety functions for limiting fault events

International Nuclear Information System (INIS)

Rezendes, J.P.; Hyde, A.W.

2004-01-01

The draft U.S. Nuclear Regulatory Commission (NRC) policy on digital protection system software requires all Advanced Light Water Reactors (ALWRs) to be evaluated assuming a hypothetical common mode failure (CMF) which incapacitates the normal automatic initiation of safety functions. The System 80 + ALWR has been evaluated for such hypothetical conditions. The results show that the diverse automatic and manual protective systems in System 80 + provide ample safety performance margins relative to core coolability, offsite radiological releases. Reactor Coolant System (RCS) pressurization and containment integrity. This deterministic evaluation served to quantify the significant inherent safety margins in the System 80 + Standard Plant design even in the event of this extremely low probability scenario of a common mode failure. (author)

Inherent and antigen-induced airway hyperreactivity in NC mice

Directory of Open Access Journals (Sweden)

Tetsuto Kobayashi

1999-01-01

Full Text Available In order to clarify the airway physiology of NC mice, the following experiments were carried out. To investigate inherent airway reactivity, we compared tracheal reactivity to various chemical mediators in NC, BALB/c, C57BL/6 and A/J mice in vitro. NC mice showed significantly greater reactivity to acetylcholine than BALB/c and C57BL/6 mice and a reactivity comparable to that of A/J mice, which are known as high responders. Then, airway reactivity to acetylcholine was investigated in those strains in vivo. NC mice again showed comparable airway reactivity to that seen in A/J mice and a significantly greater reactivity than that seen in BALB/c and C57BL/6 mice. To investigate the effects of airway inflammation on airway reactivity to acetylcholine in vivo, NC and BALB/c mice were sensitized to and challenged with antigen. Sensitization to and challenge with antigen induced accumulation of inflammatory cells, especially eosinophils, in lung and increased airway reactivity in NC and BALB/c mice. These results indicate that NC mice exhibit inherent and antigen-induced airway hyperreactivity. Therefore, NC mice are a suitable strain to use in investigating the mechanisms underlying airway hyperreactivity and such studies will provide beneficial information for understanding the pathophysiology of asthma.

A concept of JAERI passive safety light water reactor system (JPSR)

Energy Technology Data Exchange (ETDEWEB)

Murao, Y.; Araya, F.; Iwamura, T. [Japan Atomic Energy Research Institute, Tokai-mura (Japan)

1995-09-01

The Japan Atomic Energy Research Institute (JAERI) proposed a passive safety reactor system concept, JPSR, which was developed for reducing manpower in operation and maintenance and influence of human errors on reactor safety. In the concept the system was extremely simplified. The inherent matching nature of core generation and heat removal rate within a small volume change of the primary coolant is introduced by eliminating chemical shim and adopting in-vessel control rod drive mechanism units, a low power density core and once-through steam generators. In order to simplify the system, a large pressurizer, canned pumps, passive engineered-safety-features-system (residual heat removal system and coolant injection system) are adopted and the total system can be significantly simplified. The residual heat removal system is completely passively actuated in non-LOCAs and is also used for depressurization of the primary coolant system to actuate accumulators in small break LOCAs and reactor shutdown cooling system in normal operation. All of systems for nuclear steam supply system are built in the containment except for the air coolers as a the final heat sink of the passive residual heat removal system. Accordingly the reliability of the safety system and the normal operation system is improved, since most of residual heat removal system is always working and a heat sink for normal operation system is {open_quotes}safety class{close_quotes}. In the passive coolant injection system, depressurization of the primary cooling system by residual heat removal system initiates injection from accumulators designed for the MS-600 in medium pressure and initiates injection from the gravity driven coolant injection pool at low pressure. Analysis with RETRAN-02/MOD3 code demonstrated the capability of passive load-following, self-power-controllability, cooling and depressurization.

Sensors and actuators inherent in biological species

Science.gov (United States)

Taya, Minoru; Stahlberg, Rainer; Li, Fanghong; Zhao, Ying Joyce

2007-04-01

This paper addresses examples of sensing and active mechanisms inherent in some biological species where both plants and animals cases are discussed: mechanosensors and actuators in Venus Fly Trap and cucumber tendrils, chemosensors in insects, two cases of interactions between different kingdoms, (i) cotton plant smart defense system and (ii) bird-of-paradise flower and hamming bird interaction. All these cases lead us to recognize how energy-efficient and flexible the biological sensors and actuators are. This review reveals the importance of integration of sensing and actuation functions into an autonomous system if we make biomimetic design of a set of new autonomous systems which can sense and actuate under a number of different stimuli and threats.

Spreading the word of the concept 'inherent safety' in a general industrial setting in the Dutch province of Zeeland

NARCIS (Netherlands)

Jongen, M.J.M.; Dijkman, A.; Zwanikken, S.; Zwetsloot, G.I.J.M.; Gort, J.

2007-01-01

Recent accidents in The Netherlands in different kinds of industries, like fire works storage, catering and energy industry, triggered the Dutch government to start a national program to enhance the enforcement of industrial safety at the regional and municipal level. Stimulated by this program the

"A Fiber Optic Ethernet With Inherent Migration Capability To FDDI"

Science.gov (United States)

Ferris, Kenneth D.; Chan, Tammy S.

1988-12-01

A Local Area Network (LAN) designed to a standard commercial interface, the Institute of Electrical and Electronics Engineers (IEEE) 802.3 or Ethernet, has been developed using fiber optics as the physical medium. The LAN, WhisperNet, operates in an active ring and thus has an inherent low cost migration path to a Fiber Distributed Data Interface (FDDI) implementation.

Bionic Ears: Their Development and Future Advances Using Neurotrophins and Inherently Conducting Polymers

Directory of Open Access Journals (Sweden)

Graeme M. Clark

2004-01-01

Full Text Available The development of the multiple-channel bionic ear for hearing and speech understanding in profoundly deaf people is the result of integrating biological and physical sciences with engineering. It is the first clinically successful restoration of sensory and brain function, and brings electronic technology into a direct functional relationship with human consciousness. It presently transmits essential place and coarse temporal information for the coding of frequency, but the fine temporal and place excitation of groups of nerve fibres is inadequate for high-fidelity sound. This is required for adequate musical appreciation and hearing in noise. Research has demonstrated that nerve growth factors preserve the peripheral processes of the auditory nerves so that an electrode array placed close to these fibres could produce this fine temporal and spatial coding. The nerve growth factors can be incorporated into inherently conducting polymers that are part of the array so the peripheral processes can be preserved at the same time as they are electrically stimulated.

Advancements in the design of safety-related systems and components of the MARS nuclear plant

International Nuclear Information System (INIS)

Caira, M.; Caruso, G.; Naviglio, A.; Sorabella, L.; Farello, C.E.

1992-01-01

In the paper, the advancements in the design of safety-related systems and components of the MARS nuclear plant, equipped with a 600 MW th PWR, are described. These advancements are due to the special safety features of this plant, which relies completely on inherent and passive safety. In particular, the new steps of the design of the innovative, completely passive, and with an unlimited autonomy Emergency core Cooling System are described, together with the characteristics of the last version of the steam generator, developed in a new design involving disconnecting components, for a fast erection and an easy maintenance. (author)

The US Advanced Liquid Metal Reactor and the Fast Flux Test Facility Phase IIA passive safety tests

International Nuclear Information System (INIS)

Shen, P.K.; Harris, R.A.; Campbell, L.R.; Dautel, W.A.; Dubberley, A.E.; Gluekler, E.L.

1992-07-01

This report discusses the safety approach of the Advanced Liquid Metal reactor program, sponsored by the US Department of Energy, which relies upon passive reactor responses to off-normal condition to limit power and temperature excursions to levels that allow safety margins. Gas expansion modules (GEM) have included in the design to provide negative reactivity to enhance these margins in the extremely unlikely event that pumping power is lost and the highly reliable scram system fails to operate. The feasibility and beneficial features of these devices were first demonstrated in the core of the Fast Flux Test Facility (FFTF) in 1986. Preapplication safety evaluations by the US Nuclear Regulatory Commission have identified areas that must be addressed if these devices are to be relied on. One of these areas is the response of the reactor when it is critical and the pumps are turned on, resulting in positive reactivity being added to the core. Tests to examine such transients have been performed as part of the continuing FFTF program to confirm the passive safety characteristics of liquid metal reactors (LMR). The primary tests consisted of starting the main coolant pumps, which forced sodium coolant into the GEMS, decreasing neutron leakage and adding positive reactivity. The resulting transients were shown to be benign and easily mitigated by the reactivity feedbacks inherent in the FFTF and all LMRs. Steady-state auxiliary tests of the GEM and feedback reactivity worths accurately predicted the transient results. The auxiliary GEM worth tests also demonstrated that the worth can be determined at a subcritical state, which allows for a verification of the GEM's availability prior to ascending to power

The secure heating reactor

International Nuclear Information System (INIS)

Pind, C.

1987-01-01

The SECURE heating reactor was designed by ASEA-ATOM as a realistic alternative for district heating in urban areas and for supplying heat to process industries. SECURE has unique safety characteristics, that are based on fundamental laws of physics. The safety does not depend on active components or operator intervention for shutdown and cooling of the reactor. The inherent safety characteristics of the plant cannot be affected by operator errors. Due to its very low environment impact, it can be sited close to heat consumers. The SECURE heating reactor has been shown to be competitive in comparison with other alternatives for heating Helsinki and Seoul. The SECURE heating reactor forms a basis for the power-producing SECURE-P reactor known as PIUS (Process Inherent Ultimate Safety), which is based on the same inherent safety principles. The thermohydraulic function and transient response have been demonstrated in a large electrically heated loop at the ASEA-ATOM laboratories

Risk-based approach to long-term safety assessment for near surface disposal of radioactive waste in Korea

International Nuclear Information System (INIS)

Jeong, C.W.; Kim, K.I.; Lee, J.I.

2000-01-01

This paper presents the Korean regulatory approach to safety assessment consistent with probabilistic, risk-based long-term safety requirements for near surface disposal facilities. The approach is based on: (1) From the standpoint of risk limitation, normal processes and probabilistic disruptive events should be integrated in a similar manner in terms of potential exposures; and (2) The uncertainties inherent in the safety assessment should be reduced using appropriate exposure scenarios. In addition, this paper emphasizes the necessity of international guidance for quantifying potential exposures and the corresponding risks from radioactive waste disposal. (author)

The PRISM concept for a safe, economic and testable liquid metal fast reactor plant

International Nuclear Information System (INIS)

Berglund, R.C.; Salerno, L.N.; Tippets, F.E.

1987-01-01

The PRISM project is underway at General Electric as part of an advanced reactor conceptual design program sponsored by the US Department of Energy. The PRISM concept emphasizes inherent safety, modular construction, and factory fabrication. These features are intended to improve the basis for public acceptance, reduce cost,improve licensability, and reduce the risk of schedule delays and cost increases during construction. A PRISM power plant comprises a number of reactor modules. The relatively small size of the reactor module facilitates the use of passive, inherent self-shutdown and shutdown heat removal features for safe accommodation of accidents. These inherent safety features permit simplification and reduction of conventional safety-related systems in the plant. Testing of a full-size prototype reactor module is planned in the late 1990's to demonstrate these inherent safety characteristics. It is intended that the results of the test be used to obtain certification of the design by the US Nuclear Regulatory Commission preparatory to use of reactor modules built to this standard design in licensed commercial plants

Early interferon-γ production in human lymphocyte subsets in response to nontyphoidal Salmonella demonstrates inherent capacity in innate cells.

Directory of Open Access Journals (Sweden)

Tonney S Nyirenda

2010-10-01

Full Text Available Nontyphoidal Salmonellae frequently cause life-threatening bacteremia in sub-Saharan Africa. Young children and HIV-infected adults are particularly susceptible. High case-fatality rates and increasing antibiotic resistance require new approaches to the management of this disease. Impaired cellular immunity caused by defects in the T helper 1 pathway lead to intracellular disease with Salmonella that can be countered by IFNγ administration. This report identifies the lymphocyte subsets that produce IFNγ early in Salmonella infection.Intracellular cytokine staining was used to identify IFNγ production in blood lymphocyte subsets of ten healthy adults with antibodies to Salmonella (as evidence of immunity to Salmonella, in response to stimulation with live and heat-killed preparations of the D23580 invasive African isolate of Salmonella Typhimurium. The absolute number of IFNγ-producing cells in innate, innate-like and adaptive lymphocyte subpopulations was determined.Early IFNγ production was found in the innate/innate-like lymphocyte subsets: γδ-T cells, NK cells and NK-like T cells. Significantly higher percentages of such cells produced IFNγ compared to adaptive αβ-T cells (Student's t test, P<0.001 and ≤0.02 for each innate subset compared, respectively, with CD4(+- and CD8(+-T cells. The absolute numbers of IFNγ-producing cells showed similar differences. The proportion of IFNγ-producing γδ-T cells, but not other lymphocytes, was significantly higher when stimulated with live compared with heat-killed bacteria (P<0.0001.Our findings indicate an inherent capacity of innate/innate-like lymphocyte subsets to produce IFNγ early in the response to Salmonella infection. This may serve to control intracellular infection and reduce the threat of extracellular spread of disease with bacteremia which becomes life-threatening in the absence of protective antibody. These innate cells may also help mitigate against the effect on IFN�

Labor unions and safety climate: perceived union safety values and retail employee safety outcomes.

Science.gov (United States)

Sinclair, Robert R; Martin, James E; Sears, Lindsay E

2010-09-01

Although trade unions have long been recognized as a critical advocate for employee safety and health, safety climate research has not paid much attention to the role unions play in workplace safety. We proposed a multiple constituency model of workplace safety which focused on three central safety stakeholders: top management, ones' immediate supervisor, and the labor union. Safety climate research focuses on management and supervisors as key stakeholders, but has not considered whether employee perceptions about the priority their union places on safety contributes contribute to safety outcomes. We addressed this gap in the literature by investigating unionized retail employee (N=535) perceptions about the extent to which their top management, immediate supervisors, and union valued safety. Confirmatory factor analyses demonstrated that perceived union safety values could be distinguished from measures of safety training, workplace hazards, top management safety values, and supervisor values. Structural equation analyses indicated that union safety values influenced safety outcomes through its association with higher safety motivation, showing a similar effect as that of supervisor safety values. These findings highlight the need for further attention to union-focused measures related to workplace safety as well as further study of retail employees in general. We discuss the practical implications of our findings and identify several directions for future safety research. 2009 Elsevier Ltd. All rights reserved.

Innovative grout/retrieval demonstration final report

International Nuclear Information System (INIS)

Loomis, G.G.; Thompson, D.N.

1995-01-01

This report presents the results of an evaluation of an innovative retrieval technique for buried transuranic waste. Application of this retrieval technique was originally designed for full pit retrieval; however, it applies equally to a hot spot retrieval technology. The technique involves grouting the buried soil waste matrix with a jet grouting procedure, applying an expansive demolition grout to the matrix, and retrieving the debris. The grouted matrix provides an agglomeration of fine soil particles and contaminants resulting in an inherent contamination control during the dusty retrieval process. A full-scale field demonstration of this retrieval technique was performed on a simulated waste pit at the Idaho National Engineering Laboratory. Details are reported on all phases of this proof-of-concept demonstration including pit construction, jet grouting activities, application of the demolition grout, and actual retrieval of the grouted pit. A quantitative evaluation of aerosolized soils and rare earth tracer spread is given for all phases of the demonstration, and these results are compared to a baseline retrieval activity using conventional retrieval means. 8 refs., 47 figs., 10 tabs

Nuclear safety guide: TID--7016, Revision 2

International Nuclear Information System (INIS)

Thomas, J.T.

1978-01-01

The Nuclear Safety Guide was first issued in 1956 as classified AEC report LA-2063 and was reprinted the next year, unclassified, as TID-7016. Revision 1, published in 1961, extended the scope and refined the guiding information. Revision 2 of the Guide differs significantly from its predecessor in that the latter was intentionally conservative in its recommendations. Firmly based on experimental evidence of criticality, the original Guide and the first revision were considered to be of most value to organizations whose activities with fissionable materials were not extensive and, secondarily, that it would serve as a point of departure for members of established nuclear safety teams experienced in the field. The advance of calculational capability has permitted validated calculations to extend and substitute for experimental data. The broadened data base has enabled better interpolation, extension, and understanding of available information, especially in areas previously addressed by undefined but adequate factors of safety. The content has been thereby enriched in qualitative guidance. The information inherently contains, and the user can recapture, the quantitative guidance characteristic of the former Guides by employing appropriate safety factors

The role of health and safety experts in the management of hazardous and toxic wastes in Indonesia

Science.gov (United States)

Supriyadi; Hadiyanto

2018-02-01

Occupational Safety and Health Experts in Indonesia have an important role in integrating environmental health and safety factors, including in this regard as human resources assigned to undertake hazardous waste management. Comprehensive knowledge and competence skills need to be carried out responsibly, as an inherent professional occupational safety and health profession. Management leaders should continue to provide training in external agencies responsible for science in the management of toxic waste to enable occupational safety and health experts to improve their performance in the hierarchy of control over the presence of hazardous materials. This paper provides an overview of what strategies and competencies the Occupational Safety and Health expert needs to have in embracing hazardous waste management practices.

Deliberate practice theory: perceived relevance, effort, and inherent enjoyment of music practice: study II.

Science.gov (United States)

Hyllegard, Randy; Bories, Tamara L

2009-10-01

This study, based on the theory of deliberate practice, examined the practice relevance, effort, and inherent enjoyment aspects of the theory. 25 college undergraduates practiced playing a melody on an electronic keyboard for three 20-min. practice sessions. Following each session, the perceived relevance of the practice for improving performance of the melody, the effort needed to learn the melody, and the inherent enjoyment of the practice were each rated on 10-point scales. Findings were consistent with theory and similar to previous studies also involving music practice and other tasks.

Manson Chicks and Microskirted Cuties : Pornification in Thomas Pynchon's Inherent Vice

NARCIS (Netherlands)

Cook, S.J.|info:eu-repo/dai/nl/411939432

2015-01-01

Many sexual encounters in Thomas Pynchon’s fiction have occurred beyond the mainstream, generating theatres of perversity which dramatise the death wish and enact power relations from wider arenas. However, in Inherent Vice they change in nature. With the exception of scenes which use Charles Manson

Applicability of LBB concept to tokamak-type fusion machine

International Nuclear Information System (INIS)

Nakahira, Masataka

2003-12-01

A tokamak-type fusion machine has been characterized as having inherent plasma shutdown safety. An extremely small leakage of impurities such as primary cooling water, i.e., less than 0.1 g/s, will cause a plasma disruption. This plasma disruption will induce electromagnetic forces (EM forces) acting in the Vacuum Vessel (VV) and plasma-facing components. The VV forms the physical barrier that encloses tritium and activated dust. If the VV has the possibility of sustaining an unstable fracture from a through crack caused by EM forces, the structural safety will be assured and the inherent safety will be demonstrated. This paper analytically assures the Leak-Before-Break (LBB) concept as applied to the VV and is based on experimental leak rate data of a through crack having a very small opening. Based on the analysis, the critical crack length to terminate plasma is evaluated as about 2 mm. On the other hand, the critical crack length for unstable fracture is obtained as about 400 mm. It is therefore concluded that EM forces induced by small leak to terminate plasma will not cause the unstable fracture of VV, and then the inherent safety is demonstrated. (author)

A simplified analysis of uncertainty propagation in inherently controlled ATWS events

International Nuclear Information System (INIS)

Wade, D.C.

1987-01-01

The quasi static approach can be used to provide useful insight concerning the propagation of uncertainties in the inherent response to ATWS events. At issue is how uncertainties in the reactivity coefficients and in the thermal-hydraulics and materials properties propagate to yield uncertainties in the asymptotic temperatures attained upon inherent shutdown. The basic notion to be quantified is that many of the same physical phenomena contribute to both the reactivity increase of power reduction and the reactivity decrease of core temperature rise. Since these reactivities cancel by definition, a good deal of uncertainty cancellation must also occur of necessity. For example, if the Doppler coefficient is overpredicted, too large a positive reactivity insertion is predicted upon power reduction and collapse of the ΔT across the fuel pin. However, too large a negative reactivity is also predicted upon the compensating increase in the isothermal core average temperature - which includes the fuel Doppler effect

Magnetic latch trigger for inherent shutdown assembly

International Nuclear Information System (INIS)