Safety evaluation for the inner canister closure station

International Nuclear Information System (INIS)

Glasscock, J.R.

1987-01-01

The Inner Canister Closure Station (ICCS), built by Remote Technology Corporation, will be operability tested. The ICCS is used to remotely leak test Inner Canister Closures (ICC's) and replace ICC's that are not water tight. After operability testing, the ICCS will be inspected and sent to the 717-F mock-up shop for remotability demonstration and dimensional checks, then installed in the Vitrification Building, 221-S. An analysis of potential safety hazards, equipment safety features, and procedural controls indicates that the ICCS can be operated without undue hazard to employees or to the public. A safety inspection and a new equipment inspection will be held before operation to verify that the ICCS meets Savannah River Site safety requirements. 4 refs., 6 figs

Cultivation of nuclear safety culture in Guangdong Nuclear Power Station (GNPS)

International Nuclear Information System (INIS)

Lu Wei; Tang Yanzhao

2004-01-01

Probed into the concept and developing phases of safety couture in the management of nuclear power station, especially analyzed the background and the road of cultivating nuclear safety culture in GNPS, highlighted the core concept of GNPS nuclear safety culture, presented GNPS safety culture indicators, summarized the major measures taken by GNPS, depicted the propagandizing process of transparency in GNPS, and systematically appraised the effect of GNPS in implementing nuclear safety culture. (authors)

Dam safety at Seven Sisters Generating Station

International Nuclear Information System (INIS)

Carson, R. W.; Gupta, R. C.

1996-01-01

A safety surveillance program for all hydraulic structures in Manitoba was first implemented in 1979, and updated in 1988. This contribution describes the current status of the program, and the nature of the issues that the program was designed to address. The Seven Sisters Station's dam on the Winnipeg River, about 90 km northeast of the City of Winnipeg, was used as an example. Extensive reviews of flood risks and downstream inundation potential at Seven Sisters' revealed a number of deficiencies; these findings will be incorporated into a corporate plan of overall remediation. Updating the program will also include efforts to ensure adherence to national dam safety guidelines. 5 figs

The Space Station Freedom - International cooperation and innovation in space safety

Science.gov (United States)

Rodney, George A.

1989-01-01

The Space Station Freedom (SSF) being developed by the United States, European Space Agency (ESA), Japan, and Canada poses novel safety challenges in design, operations, logistics, and program management. A brief overview discloses many features that make SSF a radical departure from earlier low earth orbit (LEO) space stations relative to safety management: size and power levels; multiphase manned assembly; 30-year planned lifetime, with embedded 'hooks and scars' forevolution; crew size and skill-mix variability; sustained logistical dependence; use of man, robotics and telepresence for on-orbit maintenance of station and free-flyer systems; closed-environment recycling; use of automation and expert systems; long-term operation of collocated life-sciences and materials-science experiments, requiring control and segregation of hazardous and chemically incompatible materials; and materials aging in space.

Dynamic safety assessment of natural gas stations using Bayesian network

Energy Technology Data Exchange (ETDEWEB)

Zarei, Esmaeil, E-mail: smlzarei65@gmail.com [Center of Excellence for Occupational Health Engineering, Research Center for Health Sciences, Faculty of Health, Hamadan University of Medical Sciences, Hamadan (Iran, Islamic Republic of); Azadeh, Ali [School of Industrial and Systems Engineering, Center of Excellence for Intelligent-Based Experimental Mechanic, College of Engineering, University of Tehran (Iran, Islamic Republic of); Khakzad, Nima [Safety and Security Science Section, Delft University of Technology, Delft (Netherlands); Aliabadi, Mostafa Mirzaei [Center of Excellence for Occupational Health Engineering, Research Center for Health Sciences, Faculty of Health, Hamadan University of Medical Sciences, Hamadan (Iran, Islamic Republic of); Mohammadfam, Iraj, E-mail: mohammadfam@umsha.ac.ir [Center of Excellence for Occupational Health Engineering, Research Center for Health Sciences, Faculty of Health, Hamadan University of Medical Sciences, Hamadan (Iran, Islamic Republic of)

2017-01-05

Graphical abstract: Dynamic cause-consequence analysis of the regulator system failure using BN. - Highlights: • A dynamic and comprehensive QRA (DCQRA) framework is proposed for safety assessment of CGSs. • Bow-tie diagram and Bayesian network are employed for accident scenario modeling. • Critical basic events and minimal cut sets are identified using probability updating. - Abstract: Pipelines are one of the most popular and effective ways of transporting hazardous materials, especially natural gas. However, the rapid development of gas pipelines and stations in urban areas has introduced a serious threat to public safety and assets. Although different methods have been developed for risk analysis of gas transportation systems, a comprehensive methodology for risk analysis is still lacking, especially in natural gas stations. The present work is aimed at developing a dynamic and comprehensive quantitative risk analysis (DCQRA) approach for accident scenario and risk modeling of natural gas stations. In this approach, a FMEA is used for hazard analysis while a Bow-tie diagram and Bayesian network are employed to model the worst-case accident scenario and to assess the risks. The results have indicated that the failure of the regulator system was the worst-case accident scenario with the human error as the most contributing factor. Thus, in risk management plan of natural gas stations, priority should be given to the most probable root events and main contribution factors, which have identified in the present study, in order to reduce the occurrence probability of the accident scenarios and thus alleviate the risks.

Dynamic safety assessment of natural gas stations using Bayesian network

International Nuclear Information System (INIS)

Zarei, Esmaeil; Azadeh, Ali; Khakzad, Nima; Aliabadi, Mostafa Mirzaei; Mohammadfam, Iraj

2017-01-01

Graphical abstract: Dynamic cause-consequence analysis of the regulator system failure using BN. - Highlights: • A dynamic and comprehensive QRA (DCQRA) framework is proposed for safety assessment of CGSs. • Bow-tie diagram and Bayesian network are employed for accident scenario modeling. • Critical basic events and minimal cut sets are identified using probability updating. - Abstract: Pipelines are one of the most popular and effective ways of transporting hazardous materials, especially natural gas. However, the rapid development of gas pipelines and stations in urban areas has introduced a serious threat to public safety and assets. Although different methods have been developed for risk analysis of gas transportation systems, a comprehensive methodology for risk analysis is still lacking, especially in natural gas stations. The present work is aimed at developing a dynamic and comprehensive quantitative risk analysis (DCQRA) approach for accident scenario and risk modeling of natural gas stations. In this approach, a FMEA is used for hazard analysis while a Bow-tie diagram and Bayesian network are employed to model the worst-case accident scenario and to assess the risks. The results have indicated that the failure of the regulator system was the worst-case accident scenario with the human error as the most contributing factor. Thus, in risk management plan of natural gas stations, priority should be given to the most probable root events and main contribution factors, which have identified in the present study, in order to reduce the occurrence probability of the accident scenarios and thus alleviate the risks.

Safety planning for nuclear power stations

International Nuclear Information System (INIS)

Tadmor, J.

1979-01-01

The article shows that compared to the many industries and other human activities, nuclear power stations are among the safest. A short description of the measures taken to prevent accidents and of the additional safety means entering into action if an accident does occur is presented. It is shown that in nuclear plants the death frequency following malfunctioning is 1 death in 100.000 years whereas deaths following other human activities is 1 in 2 to 100 years and following natural calamities like earthquakes and floods is 1 in 10 years. As an example it is shown that for a population of 15.000.000 living in a radius of 40 km around 100 power stations the average number of deaths will be of 2 per year as compared to 4200 from road accidents with the corresponding number of injuries of 20 and 375.000 respectively. (B.G.)

Preparedness of fire safety in underground train station: Comparison between train operators in Malaysia with other operators from the developed countries

Science.gov (United States)

Tajedi, Noor Aqilah A.; Sukor, Nur Sabahiah A.; Ismail, Mohd Ashraf M.; Shamsudin, Shahrul A.

2017-10-01

The purpose of this paper is to compare the fire evacuation plan and preparation at the underground train stations in the different countries. The methodology for this study was using the extended questionnaire survey to investigate the Rapid Rail Sdn Bhd, Malaysia's fire safety plan and preparation at the underground train stations. There were four sections in the questionnaire which included (i) background of the respondents, (ii) the details on the train stations, safety instruction and fire evacuation exercises (iii) technical systems, installation and equipment at the underground stations and (iv) procedures and technical changes related to fire safety that had been applied by the operators. Previously, the respondents from the different train operator services in the developed countries had completed the questionnaires. This paper extends the response from the Rapid Rail Sdn Bhd to compare the emergency procedures and preparation for fire event with the developed countries. As a result, this study found that the equipment and facilities that provided at the underground train stations that operated by Rapid Rail are relevant for fire safety procedures and needs. The main advantage for Rapid Rail is the underground stations were designed with two or more entrances/exits that may perform better evacuation compare to one main entrance/exit train stations in the other developed countries.

Efforts toward enhancing seismic safety at Kashiwazaki Kariwa nuclear power station

Energy Technology Data Exchange (ETDEWEB)

Yamashita, Kazuhiko

2010-09-15

Kashiwazaki-Kariwa Nuclear Power Station, 8212MW, was struck by M6.8 quakes in July 2007. TEPCO has steadily been conducting restoration and post-earthquake equipment integrity assessment, aiming to make it a disaster-resistant power station. 2 units among 7 resumed commercial operation by June 2010. This earthquake has provided a great deal of knowledge and information useful for nuclear safety improvement. It has also served as a valuable reference for the IAEA in developing earthquake-related guidelines. TEPCO would like to share the knowledge and information thereby contributing to improving the safety of nuclear power generation. We will introduce some of our activities.

Implications of Extension of Station Blackout Cooping Capability on Nuclear Power Plant Safety

International Nuclear Information System (INIS)

Volkanovski, Andrija

2015-01-01

The safety of the nuclear power plant depends on the availability of the continuous and reliable sources of electrical energy during all modes of operation of the plant. The station blackout corresponds to a total loss of all alternate current (AC) power as a result of complete failure of both offsite and on-site AC power sources. The electricity for the essential systems during station blackout is provided from the batteries installed in the nuclear power plant. The results of the probabilistic safety assessment show that station blackout is one of the main and frequently the dominant contributor to the core damage frequency. Results of the analysis of the implications of the strengthening of the SBO mitigation capability on safety of the NPP will be presented. The assessment is done with state-of-art deterministic and probabilistic methods and tolls with application on reference models of nuclear power plants. The safety analysis is done on reference model of the nuclear power plant. Obtained results show large decrease of the core damage frequency with strengthening of the station blackout mitigation capability. The time extension of blackout coping capability results in the delay of the core heat up for at least the extension time interval. Availability and operation of the steam driven auxiliary feedwater system maintains core integrity up to 72 h after the successful shutdown, even in the presence of the reactor coolant pumps seal leakage. The largest weighted decrease of the core damage frequency considering the costs for the modification is obtained for the modification resulting in extension of the station blackout coping capability. The importance of the common cause failures of the emergency diesel generators for the obtained decrease of the core damage frequency and overall safety of the plant is identified in the obtained results. (authors)

76 FR 52016 - NASA International Space Station Advisory Committee and the Aerospace Safety Advisory Panel; Meeting

Science.gov (United States)

2011-08-19

... NATIONAL AERONAUTICS AND SPACE ADMINISTRATION [Notice (11-074)] NASA International Space Station Advisory Committee and the Aerospace Safety Advisory Panel; Meeting AGENCY: National Aeronautics and Space... meeting of the NASA International Space Station Advisory Committee and the Aerospace Safety Advisory Panel...

Indicators for monitoring of safety operation and condition of nuclear power stations

International Nuclear Information System (INIS)

Manova, D.

2001-01-01

A common goal of all employees in the nuclear power field is safety operation of nuclear power stations. The evaluation and control of NPP safety operation are a part of the elements of safety management. The present report is related only to a part of the total assessment and control of the plant safety operation, namely - the indicator system for monitoring of Kozloduy NPP operation and condition. (author)

Safety aspects of station blackout at nuclear power plants

International Nuclear Information System (INIS)

1985-03-01

The principal focus of this report is on existing light water reactor nuclear power plants. However, many of the considerations discussed herein can be equally applied to new plants, i.e. those not yet in construction. This report is organized to provide a description of design and procedural factors which safety assessments and reviews of operating experience have shown to be important. These are divided into the off-site power system, the on-site AC power systems and alternate (or nearby) sources of power. The latter may be used in the unlikely event that both normal off-site and on-site sources fail. It must be emphasized that first priority should be placed on designing and maintaining high reliability of both the off-site and on-site AC power systems. This basic concept also applies to the capabilities for restoring power sources which failed and making use of all available alternative and nearby power sources during an emergency, to restore AC power in a prompt manner. Discussions on these aspects are provided in chapters 2 and 3 of this report. Because the expected event frequency and associated confidence in such estimations of station blackout are uncertain, preparations should be made to deal with a station blackout. The nature of those preparations, whether they be optimizing emergency procedures to use existing equipment, modifying this equipment to enhance capabilities, or adding new components or systems to cope with station blackout, must be made in light of plant-specific assessments and regulatory safety philosophies/requirements. Discussions on these matters are provided in chapter 4. General and specific conclusions and recommendations are provided in chapter 5. Appendix A provides a description of several case studies on station blackout and loss of off-site power. Abstracts of papers and presentations are provided in Appendix B with authors and affiliations identified to facilitate personal contact. The References and Bibliography contain a

Safety technology comments on Cattenom nuclear power station

International Nuclear Information System (INIS)

1986-01-01

The comments are based on the safety report for the identical Flamanville plant, on the contract documents for the DUP and on results of consultancy by the DFK Working Party and on replies from the French side in the context of a German list of questions. The comments refer to the state in spring 1982. According to the agreement made for the catalogue of questions Cattenom/KKP 2, the 1300 MW Philippsburg II nuclear power station is used as a German comparable plant in these comments. Reference is made to the German comparable plant in the description and assessment of important parts of the plant. The main points of the safety technology comments are in the assessment of site characteristics; the safety technology concept; the derivation of radio-active materials and the radiation exposure in the adjacent German area. (orig./HP) [de

Dynamic safety assessment of natural gas stations using Bayesian network

NARCIS (Netherlands)

Zarei, Esmaeil; Azadeh, Ali; Khakzad Rostami, N.; Mirzaei Aliabadi, Mostafa; Mohammadfam, Iraj

2017-01-01

Pipelines are one of the most popular and effective ways of transporting hazardous materials, especially natural gas. However, the rapid development of gas pipelines and stations in urban areas has introduced a serious threat to public safety and assets. Although different methods have been

The impact of cardinal rules on employee safety behaviour at power stations in Mpumalanga / Chauke, T.L.

OpenAIRE

Chauke, Tinyiko Lourence

2011-01-01

Occupational risk management can be a catalyst in generating superior returns for all stakeholders on a sustainable basis. A number of companies in South Africa have implemented Cardinal Rules of Safety adopted from international companies to ensure the safety of their employees. The purpose of this study was to measure the impact of the cardinal rules on employee safety behaviour implemented at power stations in Mpumalanga. The empirical study was done by using a questionnaire as measurin...

An approach for risk informed safety culture assessment for Canadian nuclear power stations

International Nuclear Information System (INIS)

Nelson, W.R.

2010-01-01

One of the most important components of effective safety and risk management for nuclear power stations is a healthy safety culture. DNV has developed an approach for risk informed safety culture assessment that combines two complementary paradigms for safety and risk management: loss prevention - for preventing and intervening in accidents; and critical function management - for achieving safety and performance goals. Combining these two paradigms makes it possible to provide more robust systems for safety management and to support a healthy safety culture. This approach is being applied to safety culture assessment in partnership with a Canadian nuclear utility. (author)

A study on activities of nuclear stations in the U.S. in the aim of enhancing safety culture

International Nuclear Information System (INIS)

Ichiki, Kuniyasu

2011-01-01

After the discovery of a Reactor Vessel Head Degradation at Davis Besse, the NRC, a regulatory body in the U.S., launched an evaluation scheme regarding nuclear stations' Safety Culture. In this evaluation, the NRC examines every inspection finding of conventional Reactor Oversight Program inspections to find out what was the most dominant causal factor with SC aspects. There are no similar analyses of this kind opened to public. This study has its goal of obtaining some useful clue for realizing how to project SC enhancement activities, by making analysis on every ROP inspection finding as well as collecting examples of good practices in improving organizational performance of stations in the U.S. As for the inspection results of ROP, there are no obvious change in the trend of number and types of ROP findings. On the other hand, because of the variety in average number of findings at stations, we could assume that actual condition of SC within a particular station quite differs from other stations. Looking at ROP reports minutely, you may realize that all U.S. stations have been making great effort to improve the efficiency of its Corrective Action Program which is regarded as the key ability to identify and fix undesirable incidents broke out at the station. It is a noteworthy aspect that CAP is not a simple program of quality assurance but also one of the items in power stations to be improved continuously. The concrete elements of this process are realizing current organizational condition by series of Performance Indicators, setting the next goals with the PIs, also evaluating the degree of the achievements with them, and modifying or re-defining new goal of the organization in accordance with new situation. This idea is introduced into broad range of organizational activities all over the U.S., which assure their improvement to be objective and effective. Simultaneously, such way of operation have its staff devoted in own responsibility, which eventually

Statement on safety requirements concerning the long-term operation of the Muehleberg nuclear power station

International Nuclear Information System (INIS)

2012-12-01

This report published by the Swiss Federal Nuclear Safety Inspectorate ENSI investigates the safety requirements with respect to the long-term operation of the Muehleberg nuclear power station in Switzerland. Relevant international requirements and Swiss legal stipulations concerning the long-term operation of the power station are stated. The management of aging processes is looked at. The regular verification of the integrity of various plant components such as containments, piping, steam generation system, etc. is looked at in detail. The state-of-the-art concerning deterministic accident analyses and refitting technology are discussed, as are automated safety systems. The applicable laws, decrees and guidelines are listed in appendices

Radiation safety assessment of mobile telephone base stations

International Nuclear Information System (INIS)

Mohd Yusof Mohd Ali; Mohd Anuar Majid; Mohd Amirul Nizam

2002-01-01

Mobile telephone is fast getting popular among users and in fact it has become one of the fastest selling electronic products in the world. More base stations are expected to be built to meet such high demands and this has caused great concerned among members of the public, especially those living close to the stations, about the potential harmful health effects of radiofrequency (RF) radiation produced by such facilities. A project was initiated by MINT in early 2000 with aims to assess the radiation levels present in the areas around the base stations and to establish baseline data on the pattern and trend of the radiation emission from each different set up of the facilities. This paper highlights some basics facts about mobile telephones and preliminary findings of the project. The assessment has been carried out at 16 base station sites and the results indicate that the radiation levels present around these sites are very low. Their broadband readings vary between below the detection limit of 0.3μWatts/cm 2 to 11 μWatts/cm 2 and they are comparable to normal background radiation present in places away from any base stations. The highest level observed was 1.5% of the exposure limit recommended for members of the public. However, locations at close distance in front of the the antenna can be very serious in term of radiation exposure since the radiation level here can easily exceed the permissible exposure limit for public. Safety precaution needs to be taken when entering these areas and they should be out of bound for members of the public. (Author)

Technology, safety and costs of decommissioning a Reference Boiling Water Reactor Power Station. Main report. Volume 1

Energy Technology Data Exchange (ETDEWEB)

Oak, H.D.; Holter, G.M.; Kennedy, W.E. Jr.; Konzek, G.J.

1980-06-01

Technology, safety and cost information is given for the conceptual decommissioning of a large (1100MWe) boiling water reactor (BWR) power station. Three approaches to decommissioning, immediate dismantlement, safe storage with deferred dismantlement and entombment, were studied to obtain comparisons between costs, occupational radiation doses, potential dose to the public and other safety impacts. It also shows the sensitivity of decommissioning safety and costs to the power rating of a BWR in the range of 200 to 1100 MWe.

The Spanish Nuclear Safety Council and nuclear power stations in operation in Spain

International Nuclear Information System (INIS)

Perello, M.

1984-01-01

On 20 April 1980 the Spanish Congress of Deputies passed an Act setting up the Nuclear Safety Council (CSN) as the sole organization responsible for nuclear safety and radiation protection. In this paper it is stated that that date marked the beginning of a new nuclear safety policy in Spain. As one of its objectives, this policy is aimed at the monitoring and testing of operating nuclear installations. A detailed description is given of the Operating Nuclear Installation Service (SINE), including its basic structure, its functions and the technical and manpower resources available to it. The maintenance of close relations with other organs of the CSN is considered of paramount importance in order for the tasks allotted to SINE to be fulfilled. International co-operation and outside contracting greatly assist importing countries which have limited manpower resources. A description is then given of the present state of the nuclear power stations in operation in Spain together with an account of the most important initiatives which have been taken so far. The year 1968 saw the beginning of commercial operation of the Jose Cabrera nuclear power station, which has the only single-loop PWR reactor in the world. At present, it is being subjected to the Systematic Evaluation Programme (SEP). The Santa Maria de Garona nuclear power station has been operating for over twelve years and is also being subjected to the SEP although design modifications derived from operating experience have already been introduced. The Vandellos I station was the last of the first generation and has also benefited from the operating experience of similar French plants. Unit 1 of the Almaraz power station opens the door to the second generation and the generic problem which has occurred with the steam generators is in process of being solved. Lastly, some general conclusions are presented about the organization of and experience acquired with operating nuclear power stations. (author)

The Safety Assessment of OPR-1000 for Station Blackout Applying Combined Deterministic and Probabilistic Procedure

Energy Technology Data Exchange (ETDEWEB)

Kang, Dong Gu; Ahn, Seung-Hoon; Cho, Dae-Hyung [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2015-05-15

This is termed station blackout (SBO). However, it does not generally include the loss of available AC power to safety buses fed by station batteries through inverters or by alternate AC sources. Historically, risk analysis results have indicated that SBO was a significant contributor to overall core damage frequency. In this study, the safety assessment of OPR-1000 nuclear power plant for SBO accident, which is a typical beyond design basis accident and important contributor to overall plant risk, is performed by applying the combined deterministic and probabilistic procedure (CDPP). In addition, discussions are made for reevaluation of SBO risk at OPR-1000 by eliminating excessive conservatism in existing PSA. The safety assessment of OPR-1000 for SBO accident, which is a typical BDBA and significant contributor to overall plant risk, was performed by applying the combined deterministic and probabilistic procedure. However, the reference analysis showed that the CDF and CCDP did not meet the acceptable risk, and it was confirmed that the SBO risk should be reevaluated. By estimating the offsite power restoration time appropriately, the SBO risk was reevaluated, and it was finally confirmed that current OPR-1000 system lies in the acceptable risk against the SBO. In addition, it was demonstrated that the proposed CDPP is applicable to safety assessment of BDBAs in nuclear power plants without significant erosion of the safety margin.

To dimension safety valves. Probabilist study

International Nuclear Information System (INIS)

Noel, Robert; Couvreur, Denis

1982-01-01

The gauge of safety valves of a steam pressure apparatus is usually determined according to an operating situation envelope which it is admitted covers all that can happen in reality. For the safety of the dryer-superheaters of turbines in nuclear power stations, Electricite de France and Alsthom-Atlantique made a reliability study; its method is exposed and the results are discussed. Such a study is heavy going and complex, but in return it permits a better quantitative understanding of the various dimension and operating parameters of an installation which condition its safety. It is therefore a source of progress [fr

International Cooperation in the Field of International Space Station (ISS) Payload Safety

Science.gov (United States)

Heimann, Timothy; Larsen, Axel M.; Rose, Summer; Sgobba, Tommaso

2005-01-01

In the frame of the International Space Station (ISS) Program cooperation, in 1998, the European Space Agency (ESA) approached the National Aeronautics and Space Administration (NASA) with the unique concept of a Payload Safety Review Panel (PSRP) "franchise" based at the European Space Technology Center (ESTEC), where the panel would be capable of autonomously reviewing flight hardware for safety. This paper will recount the course of an ambitious idea as it progressed into a fully functional reality. It will show how a panel initially conceived at NASA to serve a national programme has evolved into an international safety cooperation asset. The PSRP established at NASA began reviewing ISS payloads approximately in late 1994 or early 1995 as an expansion of the pre-existing Shuttle Program PSRP. This paper briefly describes the fundamental Shuttle safety process and the establishment of the safety requirements for payloads intending to use the Space Transportation System and International Space Station (ISS). The paper will also offer some historical statistics about the experiments that completed the payload safety process for Shuttle and ISS. The paper 1 then presents the background of ISS agreements and international treaties that had to be taken into account when establishing the ESA PSRP. The detailed franchising model will be expounded upon, followed by an outline of the cooperation charter approved by the NASA Associate Administrator, Office of Space Flight, and ESA Director of Manned Spaceflight and Microgravity. The resulting ESA PSRP implementation and its success statistics to date will then be addressed. Additionally the paper presents the ongoing developments with the Japan Aerospace Exploration Agency. The discussion will conclude with ideas for future developments, such to achieve a fully integrated international system of payload safety panels for ISS.

Integrated safety assessment report: Integrated Safety Assessment Program: Millstone Nuclear Power Station, Unit 1 (Docket No. 50-245): Draft report

International Nuclear Information System (INIS)

1987-04-01

The Integrated Safety Assessment Program (ISAP) was initiated in November 1984, by the US Nuclear Regulatory Commission to conduct integrated assessments for operating nuclear power reactors. The integrated assessment is conducted in a plant-specific basis to evaluate all licensing actions, licensee initiated plant improvements and selected unresolved generic/safety issues to establish implementation schedules for each item. In addition, procedures will be established to allow for a periodic updating of the schedules to account for licensing issues that arise in the future. This report documents the review of Millstone Nuclear Power Station, Unit No. 1, operated by Northeast Nuclear Energy Company (located in Waterford, Connecticut). Millstone Nuclear Power Station, Unit No. 1, is one of two plants being reviewed under the pilot program for ISAP. This report indicates how 85 topics selected for review were addressed. This report presents the staff's recommendations regarding the corrective actions to resolve the 85 topics and other actions to enhance plant safety. The report is being issued in draft form to obtain comments from the licensee, nuclear safety experts, and the Advisory Committee for Reactor Safeguards (ACRS). Once those comments have been resolved, the staff will present its positions, along with a long-term implementation schedule from the licensee, in the final version of this report

Preparations for the 1994 OSART [operational safety review] at Hunterston Power Station

International Nuclear Information System (INIS)

Moir, I.A.

1993-01-01

The International Atomic Energy Agency have accepted an invitation from the Scottish Office to carry out an Operational Safety Review at Hunterston Power Station in April 1994. This paper outlines the preparations being undertaken by Scottish Nuclear Limited to ensure the standards set by the International Atomic Energy Agency are met. (author)

Hinkley Point 'C' power station public inquiry: proof of evidence on safety criteria

International Nuclear Information System (INIS)

Taylor, R.H.

1988-09-01

A public inquiry has been set up to examine the planning application made by the Central Electricity Generating Board (CEGB) for the construction of a 1200 MW Pressurized Water Reactor power station at Hinkley Point (Hinkley Point ''C'') in the United Kingdom. The policy is to replicate the Sizewell ''B'' PWR design which was accepted as safe by an earlier enquiry. In this evidence to the Inquiry, subsequent developments are examined with a view to determining whether these would reverse the Sizewell decision. They are: the possible revision of radiation risk estimates upwards; whether cases of leukaemia occur with greater frequency around nuclear sites than elsewhere; publication of the Health and Safety Executive's consultative document ''The Tolerability of Risk from Nuclear Power Stations''. The overall conclusion is that these developments do not undermine the findings of the Sizewell ''B'' inquiry or the validity of the CEGB's safety criteria. (author)

Design of the system of maintenance operations occupational safety and health database application of nuclear power station

International Nuclear Information System (INIS)

Wang Xuehong; Li Xiangyang; Ye Yongjun

2011-01-01

Based on the KKS code of building equipment in nuclear power station, this paper introduces the method of establishing the system of maintenance operation occupational safety and health database application. Through the application system of maintenance occupational safety and health database, it can summarize systematically all kinds of maintenance operation dangerous factor of nuclear power station, and make a convenience for staff to learn the maintenance operation dangerous factors and the prevention measures, so that it can achieve the management concept of 'precaution crucial, continuous improvement' that advocated by OSHMS. (authors)

Regulatory analysis for the resolution of Unresolved Safety Issue A-44, Station Blackout. Draft report

International Nuclear Information System (INIS)

Rubin, A.M.

1986-01-01

''Station Blackout'' is the complete loss of alternating current (ac) electric power to the essential and nonessential buses in a nuclear power plant; it results when both offsite power and the onsite emergency ac power systems are unavailable. Because many safety systems required for reactor core decay heat removal and containment heat removal depend on ac power, the consequences of a station blackout could be severe. Because of the concern about the frequency of loss of offsite power, the number of failures of emergency diesel generators, and the potentially severe consequences of a loss of all ac power, ''Station Blackout'' was designated as Unresolved Safety Issue (USI) A-44. This report presents the regulatory analysis for USI A-44. It includes: (1) a summary of the issue, (2) the proposed technical resolution, (3) alternative resolutions considered by the Nuclear Regulatory Commission (NRC) staff, (4) an assessment of the benefits and costs of the recommended resolution, (5) the decision rationale, and (6) the relationship between USI A-44 and other NRC programs and requirements

The structural integrity safety case for Sizewell B power station

International Nuclear Information System (INIS)

Gerachty, J.E.

1993-01-01

This paper presents the safety case approach adopted for the components of the Sizewell 'B' Power Station for which a high degree of structural integrity is required. Such components include the Reactor Pressure Vessel, Steam Generator and Pressuriser for which Incredibility of Failure is claimed. The two parts of the case involve achievement and demonstration of integrity. This is achieved by extensive measures involving design, manufacture, materials and inspection. The demonstration has required a fracture mechanics approach. The specific role of inspection validation and its relation to critical defect size is described. (author)

Three-dimensional numerical analysis of stress and deformation of surrounding soil in safety construction of metro station

International Nuclear Information System (INIS)

Fung, I W H; Wang, J A; Au, A S K; Zhou, Y F

2008-01-01

Utilizing underground spaces synthetically has become an important task in urban planning and civil engineering in this century. Due to the complexity and uncertainty of technical issues in underground construction and soil engineering, construction safety arouses a great concern in underground construction industry. This paper focused on construction of an underground metro station, numerical analyses are carried out to estimate the possible stresses and deformations of surrounding soil and underground structures under the worse loading conditions to assess the structural design and construction safety. Influences of pit excavation on existing tunnel and surrounding spaces are discussed and the stability of main structure of metro station will be analyzed

Second periodic safety review of Angra Nuclear Power Station, unit 1

Energy Technology Data Exchange (ETDEWEB)

Martins, Carlos F.O.; Crepaldi, Roberto; Freire, Enio M., E-mail: ottoncf@tecnatom.com.br, E-mail: emfreire46@gmail.com, E-mail: robcrepaldi@hotmail.com [Tecnatom do Brasil Engenharia e Servicos Ltda, Rio de Janeiro, RJ (Brazil); Campello, Sergio A., E-mail: sacampe@eletronuclear.gov.br [Eletrobras Termonuclear S.A. (ELETRONUCLEAR), Rio de Janeiro, RJ (Brazil)

2015-07-01

This paper describes the second Periodic Safety Review (PSR2-A1) of Angra Nuclear Power Station, Unit 1, prepared by Eletrobras Eletronuclear S.A. and Tecnatom do Brasil Engenharia e Servicos Ltda., during Jul.2013-Aug.2014, covering the period of 2004-2013. The site, in Angra dos Reis-RJ, Brazil, comprises: Unit 1, (640 MWe, Westinghouse PWR, operating), Unit 2 (1300 MWe, KWU/Areva, operating) and Unit 3 (1405 MWe, KWU/Areva, construction). The PSR2-A1 attends the Standards 1.26-Safety in Operation of Nuclear Power Plants, Brazilian Nuclear Regulatory Commission (CNEN), and IAEA.SSG.25-Periodic Safety Review of Nuclear Power Plants. Within 18 months after each 10 years operation, the operating organization shall perform a plant safety review, to investigate the evolution consequences of safety code and standards, regarding: Plant design; structure, systems and components behavior; equipment qualification; plant ageing management; deterministic and probabilistic safety analysis; risk analysis; safety performance; operating experience; organization and administration; procedures; human factors; emergency planning; radiation protection and environmental radiological impacts. The Review included 6 Areas and 14 Safety Parameters, covered by 33 Evaluations.After document evaluations and discussions with plant staff, it was generated one General and 33 Specific Guide Procedures, 33 Specific and one Final Report, including: Description, Strengths, Deficiencies, Areas for Improvement and Conclusions. An Action Plan was prepared by Electronuclear for the recommendations. It was concluded that the Unit was operated within safety standards and will attend its designed operational lifetime, including possible life extensions. The Final Report was submitted to CNEN, as one requisite for renewal of the Unit Permanent Operation License. (author)

Second periodic safety review of Angra Nuclear Power Station, unit 1

International Nuclear Information System (INIS)

Martins, Carlos F.O.; Crepaldi, Roberto; Freire, Enio M.; Campello, Sergio A.

2015-01-01

This paper describes the second Periodic Safety Review (PSR2-A1) of Angra Nuclear Power Station, Unit 1, prepared by Eletrobras Eletronuclear S.A. and Tecnatom do Brasil Engenharia e Servicos Ltda., during Jul.2013-Aug.2014, covering the period of 2004-2013. The site, in Angra dos Reis-RJ, Brazil, comprises: Unit 1, (640 MWe, Westinghouse PWR, operating), Unit 2 (1300 MWe, KWU/Areva, operating) and Unit 3 (1405 MWe, KWU/Areva, construction). The PSR2-A1 attends the Standards 1.26-Safety in Operation of Nuclear Power Plants, Brazilian Nuclear Regulatory Commission (CNEN), and IAEA.SSG.25-Periodic Safety Review of Nuclear Power Plants. Within 18 months after each 10 years operation, the operating organization shall perform a plant safety review, to investigate the evolution consequences of safety code and standards, regarding: Plant design; structure, systems and components behavior; equipment qualification; plant ageing management; deterministic and probabilistic safety analysis; risk analysis; safety performance; operating experience; organization and administration; procedures; human factors; emergency planning; radiation protection and environmental radiological impacts. The Review included 6 Areas and 14 Safety Parameters, covered by 33 Evaluations.After document evaluations and discussions with plant staff, it was generated one General and 33 Specific Guide Procedures, 33 Specific and one Final Report, including: Description, Strengths, Deficiencies, Areas for Improvement and Conclusions. An Action Plan was prepared by Electronuclear for the recommendations. It was concluded that the Unit was operated within safety standards and will attend its designed operational lifetime, including possible life extensions. The Final Report was submitted to CNEN, as one requisite for renewal of the Unit Permanent Operation License. (author)

East/West cooperation on the safety of USSR-designed nuclear power stations

International Nuclear Information System (INIS)

Spencer, P.H.

1991-01-01

In the aftermath of the accident at the Chernobyl nuclear power station in the Soviet Union, nuclear power plant operators throughout the world came together in May 1989 to form the World Association of Nuclear Operators (WANO). When it became clear that the operators of plants of an early design supplied by the USSR needed assistance in the upgrading of the safety of these units, WANO was uniquely placed to assist and facilitate in this. In July 1990, WANO took the decision to form a special project to assist the operators of the VVER 440/230 plants in their efforts to increase the safety standards for these units. The work performed by this special project team is described

Qualitative Risk Assessment for an LNG Refueling Station and Review of Relevant Safety Issues

Energy Technology Data Exchange (ETDEWEB)

Siu, N.; Herring, J.S.; Cadwallader, L.; Reece, W.; Byers, J.

1998-02-01

This report is a qualitative assessment of the public and worker risk involved with the operation of a liquefied natural gas (LNG) vehicle refueling facility. This study includes facility maintenance and operations, tank truck deliveries, and end-use vehicle fueling; it does not treat the risks of LNG vehicles on roadways. Accident initiating events are identified by using a Master Logic Diagram, a Failure Modes and Effects Analysis, and historical operating experiences. The event trees were drawn to depict possible sequences of mitigating events following the initiating events. The phenomenology of LNG and other vehicle fuels is discussed to characterize the hazard posed by LNG usage. Based on the risk modeling and analysis, recommendations are given to improve the safety of LNG refueling stations in the areas of procedures and training, station design, and the dissemination of ``best practice`` information throughout the LNG community.

Central Station Design Options

DEFF Research Database (Denmark)

2011-01-01

. The work identifies the architecture, sizing and siting of prospective Central Stations in Denmark, which can be located at shopping centers, large car parking lots or gas stations. Central Stations are planned to be integrated in the Danish distribution grid. The Danish island of Bornholm, where a high...... overloading, more reference points might be necessary to represent various transformer loading levels. The subject of safety in Central Station is also addressed. A number of safety rules based on European standards apply to AC charging equipment up to 44 kW. The connection interlock and the automatic de......-energization are identified as fundamental requirements for safety in such a charging station. The connection interlock is a solution which ensures that no power is applied to the DC cable when the EV connector is not connected. The automatic de-energization device ensures that whenever a strain on the cable is detected, e...

Safety Evaluation Report related to the operation of River Bend Station (Docket No. 50-458)

International Nuclear Information System (INIS)

1984-10-01

Supplement No. 1 to the Safety Evaluation Report on the application filed by Gulf States Utilities Company as applicant and for itself and Cajun Electric Power Cooperative, as owners, for a license to operate River Bend Station has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. This supplement reports the status of certain items that had not been resolved at the time of publication of the Safety Evaluation Report

47 CFR 80.107 - Service of private coast stations and marine-utility stations.

Science.gov (United States)

2010-10-01

... 47 Telecommunication 5 2010-10-01 2010-10-01 false Service of private coast stations and marine...) SAFETY AND SPECIAL RADIO SERVICES STATIONS IN THE MARITIME SERVICES Operating Requirements and Procedures Operating Procedures-Land Stations § 80.107 Service of private coast stations and marine-utility stations. A...

Interim qualitative risk assessment for an LNG refueling station and review of relevant safety issues

Energy Technology Data Exchange (ETDEWEB)

Siu, N.; Herring, S.; Cadwallader, L.; Reece, W.; Byers, J.

1997-07-01

This report is a qualitative assessment of the public and worker risk involved with the operation of a liquefied natural (LNG) vehicle refueling facility. This study includes facility maintenance and operations, tanker truck delivers and end-use vehicle fueling; it does not treat the risks of LNG vehicles on roadways. Accident initiating events are identified by using a Master Logic Diagram, a Failure Modes and Effects analysis and historical operating experiences. The event trees were drawn to depict possible sequences of mitigating events following the initiating events. The phenomenology of LNG and other vehicle fuels is discussed to characterize the hazard posed by LNG usage. Based on the risk modeling and analysis, recommendations are given to improve the safety of LNG refueling stations in the areas of procedures and training, station design, and the dissemination of best practice information throughout the LNG community.

Ways of solving the problems of radiation safety and environmental protection in handling radioactive waste at atomic power stations in the USSR

International Nuclear Information System (INIS)

Gusev, D.I.; Belitskij, A.S.; Turkin, A.D.; Kozlov, V.M.

1977-01-01

Requirements of the State Sanitary Supervision on radiation safety of the personnel and population and on protection of the environment in handling radioactive wastes from nuclear power stations in the USSR are regulated by the Standards of Radiation Safety, the Main Sanitary Rules for Handling Radioactive Materials and by the Sanitary Rules for Designing Nuclear Power Stations. The regulations contained in these documents are obligatory for all the establishments at the stages of design, building and operation of nuclear power stations. The main requirement for handling radioactive wastes from nuclear power stations in the USSR is to dispose of them near the place of their production. In nuclear power station siting and designing the special territory is provided for liquid and solid radioactive wastes storage taking into account the whole period of nuclear power station operation. These storage sites are located within the controlled area. They are built as required, usually for five years. The report contains hygienic and hydrological requirements to the radiation waste burial sites and data on the accepted system of controlling leak-proof qualities of the disposal cavities and radioactivity of the ground water in this region. The results of long-term studies on radionuclide leaching from the bituminic blocks are given and it is shown that the bituminizing method used for solidification of intermediate activity wastes is very promising. In the USSR much attention is given to the problem of sanitary protection of the cooling ponds at nuclear power stations. No limits to the national-economic use of these ponds outside the nuclear power station site are established. Therefore in determining the requirements to the discharge of effluents into the cooling ponds of nuclear power stations the possibility of radionuclide transfer to the population through the aquaeous and terrestrial biological chains is taken into account. The possibility of human diet contamination

Evaluation of Station Blackout accidents at nuclear power plants. Technical findings related to Unresolved Safety Issue A-44. Draft report for comment

International Nuclear Information System (INIS)

Baranowsky, P.W.

1985-05-01

''Station Blackout,'' which is the complete loss of alternating current (ac) electrical power in a nuclear power plant, has been designated as Unresolved Safety Issue A-44. Because many safety systems required for reactor core decay heat removal and containment heat removal depend on ac power, the consequences of a station blackout could be severe. This report documents the findings of technical studies performed as part of the program to resolve this issue. The important factors analyzed include: the frequency of loss of offsite power; the probability that emergency or onsite ac power supplies would be unavailable; the capability and reliability of decay heat removal systems independent of ac power; and the likelihood that offsite power would be restored before systems that cannot operate for extended periods without ac power fail, thus resulting in core damage. This report also addresses effects of different designs, locations, and operational features on the estimated frequency of core damage resulting from station blackout events

Technology, safety and costs of decommissioning a reference boiling water reactor power station. Volume 1. Main report. Technical report, September 1977-October 1979

International Nuclear Information System (INIS)

Oak, H.D.; Holter, G.M.; Kennedy, W.E. Jr.; Konzek, G.J.

1980-06-01

Technology, safety and cost information is given for the conceptual decommissioning of a large (1100MWe) boiling water reactor (BWR) power station. Three approaches to decommissioning, immediate dismantlement, safe storage with deferred dismantlement and entombment, were studied to obtain comparisons between costs, occupational radiation doses, potential dose to the public and other safety impacts. It also shows the sensitivity of decommissioning safety and costs to the power rating of a BWR in the range of 200 to 1100 MWE

Technology, safety and costs of decommissioning a reference boiling water reactor power station. Volume 1. Main report. Technical report, September 1977-October 1979

Energy Technology Data Exchange (ETDEWEB)

Oak, H.D.; Holter, G.M.; Kennedy, W.E. Jr.; Konzek, G.J.

1980-06-01

Technology, safety and cost information is given for the conceptual decommissioning of a large (1100MWe) boiling water reactor (BWR) power station. Three approaches to decommissioning, immediate dismantlement, safe storage with deferred dismantlement and entombment, were studied to obtain comparisons between costs, occupational radiation doses, potential dose to the public and other safety impacts. It also shows the sensitivity of decommissioning safety and costs to the power rating of a BWR in the range of 200 to 1100 MWE.

Technology, safety and costs of decommissioning a reference boiling water reactor power station. Volume 2. Appendices. Technical report, September 1977-October 1979

International Nuclear Information System (INIS)

Oak, H.D.; Holter, G.M.; Kennedy, W.E. Jr.; Konzek, G.J.

1980-06-01

Technology, safety and cost information is given for the conceptual decommissioning of a large (1100MWe) boiling water reactor (BWR) power station. Three approaches to decommissioning, immediate dismantlement, safe storage with deferred dismantlement and entombment, were studied to obtain comparisons between costs, occupational radiation doses, potential dose to the public and other safety impacts. It also shows the sensitivity of decommissioning safety and costs to the power rating of a BWR in the range of 200 to 1100 MWE. This volume contains the appendices

Technology, safety and costs of decommissioning a reference boiling water reactor power station. Volume 2. Appendices. Technical report, September 1977-October 1979

Energy Technology Data Exchange (ETDEWEB)

Oak, H.D.; Holter, G.M.; Kennedy, W.E. Jr.; Konzek, G.J.

1980-06-01

Technology, safety and cost information is given for the conceptual decommissioning of a large (1100MWe) boiling water reactor (BWR) power station. Three approaches to decommissioning, immediate dismantlement, safe storage with deferred dismantlement and entombment, were studied to obtain comparisons between costs, occupational radiation doses, potential dose to the public and other safety impacts. It also shows the sensitivity of decommissioning safety and costs to the power rating of a BWR in the range of 200 to 1100 MWE. This volume contains the appendices.

Steam Pressure-Reducing Station Safety and Energy Efficiency Improvement Project

Energy Technology Data Exchange (ETDEWEB)

Lower, Mark D [ORNL; Christopher, Timothy W [ORNL; Oland, C Barry [ORNL

2011-06-01

The Facilities and Operations (F&O) Directorate is sponsoring a continuous process improvement (CPI) program. Its purpose is to stimulate, promote, and sustain a culture of improvement throughout all levels of the organization. The CPI program ensures that a scientific and repeatable process exists for improving the delivery of F&O products and services in support of Oak Ridge National Laboratory (ORNL) Management Systems. Strategic objectives of the CPI program include achieving excellence in laboratory operations in the areas of safety, health, and the environment. Identifying and promoting opportunities for achieving the following critical outcomes are important business goals of the CPI program: improved safety performance; process focused on consumer needs; modern and secure campus; flexibility to respond to changing laboratory needs; bench strength for the future; and elimination of legacy issues. The Steam Pressure-Reducing Station (SPRS) Safety and Energy Efficiency Improvement Project, which is under the CPI program, focuses on maintaining and upgrading SPRSs that are part of the ORNL steam distribution network. This steam pipe network transports steam produced at the ORNL steam plant to many buildings in the main campus site. The SPRS Safety and Energy Efficiency Improvement Project promotes excellence in laboratory operations by (1) improving personnel safety, (2) decreasing fuel consumption through improved steam system energy efficiency, and (3) achieving compliance with applicable worker health and safety requirements. The SPRS Safety and Energy Efficiency Improvement Project being performed by F&O is helping ORNL improve both energy efficiency and worker safety by modifying, maintaining, and repairing SPRSs. Since work began in 2006, numerous energy-wasting steam leaks have been eliminated, heat losses from uninsulated steam pipe surfaces have been reduced, and deficient pressure retaining components have been replaced. These improvements helped ORNL

47 CFR 80.1121 - Receipt and acknowledgement of distress alerts by ship stations and ship earth stations.

Science.gov (United States)

2010-10-01

... 47 Telecommunication 5 2010-10-01 2010-10-01 false Receipt and acknowledgement of distress alerts by ship stations and ship earth stations. 80.1121 Section 80.1121 Telecommunication FEDERAL COMMUNICATIONS COMMISSION (CONTINUED) SAFETY AND SPECIAL RADIO SERVICES STATIONS IN THE MARITIME SERVICES Global Maritime Distress and Safety System (GMDSS)...

Safety evaluation report related to the operation of River Bend Station (Docket No. 50-458)

International Nuclear Information System (INIS)

1984-05-01

The Safety Evaluation Report for the application filed by the Gulf States Utilities Company, as applicant and owner, for a license to operate the River Bend Station (Docket No. 50-458) has been prepared by the Office of Nuclear Reactor Regulation of US Nuclear Regulatory Commission. The facility is located near St. Francisville, Louisiana. Subject to favorable resolution of the items discussed in this report, the NRC staff concludes that the facility can be operated by the applicant without endangering the health and safety of the public

Safety technical investigation activities for shipment of damaged spent fuels from Fukushima Daiichi Nuclear Power Station

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-08-15

Japan Nuclear Energy Safety Organization(JNES) carries out the investigation for damaged fuel transportation from Fukushima Daiichi Nuclear Power Station(1F) under safety condition to support Nuclear Regulation Authority (NRA). In 2012 fiscal year, JNES carried out the investigation of spent fuel condition in unit 4 of 1F and actual result of leak fuel transport in domestic /other countries. From this result, Package containing damaged fuel from unit 4 in 1F were considered. (author)

Safety evaluation report related to the operation of River Bend Station (Docket No. 50-458)

International Nuclear Information System (INIS)

1985-08-01

Supplement No. 3 to the Safety Evaluation Report on the application filed by Gulf States Utilities Company as applicant and for itself and Cajun Electric Power cooperative, as owners, for a license to operate River Bend Station has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is located in West Feliciana Parish, near St. Francisville, Louisiana. This supplement reports the status of certain items that had not been resolved at the time of publication of the Safety Evaluation Report, Supplement No. 1, and Supplement No. 2

Safety procedures in operation of inspection and maintenance of pressure reduction and metering stations

International Nuclear Information System (INIS)

Villas Boas, Ademar Jose; Biesemeyer, Marco Aurelio R.

2000-01-01

Each local Natural Gas Distribution Company in Brazil has its own working procedures for operations of inspection and maintenance on equipment and accessories connected to the gas network. Some of these Companies developed a better elaborated and documented way of working routines, while others only work based on their operators experience. The objective of this work is to create a standard procedure for operations of inspection and maintenance of Pressure Reducing Stations and Metering Stations, mainly the ones concerned to safety aspects. This work has no intention of exhausting all aspects related to this subject but to become the first step to standardize these types of operations among Natural Gas Distribution Companies. (author)

Safety Evaluation Report related to the operation of Hope Creek Generation Station (Docket No. 50-354). Supplement No. 6

International Nuclear Information System (INIS)

1986-07-01

Supplement No. 6 to the Safety Evaluation Report on the application filed by Public Service Electric and Gas Company on its own behalf as co-owner and as agent for the other co-owner, the Atlantic City Electric Company, for a license to operate Hope Creek Generating Station has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is located in Lower Alloways Creek Township in Salem County, New Jersey. This supplement reports the status of certain items that has not been resolved at the time of the publication of the Safety Evaluation Report. This supplement supports the issuance of a full-power license to operate Hope Creek Generating Station

Technology, safety and costs of decommissioning a reference boiling water reactor power station: Comparison of two decommissioning cost estimates developed for the same commercial nuclear reactor power station

International Nuclear Information System (INIS)

Konzek, G.J.; Smith, R.I.

1990-12-01

This study presents the results of a comparison of a previous decommissioning cost study by Pacific Northwest Laboratory (PNL) and a recent decommissioning cost study of TLG Engineering, Inc., for the same commercial nuclear power reactor station. The purpose of this comparative analysis on the same plant is to determine the reasons why subsequent estimates for similar plants by others were significantly higher in cost and external occupational radiation exposure (ORE) than the PNL study. The primary purpose of the original study by PNL (NUREG/CR-0672) was to provide information on the available technology, the safety considerations, and the probable costs and ORE for the decommissioning of a large boiling water reactor (BWR) power station at the end of its operating life. This information was intended for use as background data and bases in the modification of existing regulations and in the development of new regulations pertaining to decommissioning activities. It was also intended for use by utilities in planning for the decommissioning of their nuclear power stations. The TLG study, initiated in 1987 and completed in 1989, was for the same plant, Washington Public Supply System's Unit 2 (WNP-2), that PNL used as its reference plant in its 1980 decommissioning study. Areas of agreement and disagreement are identified, and reasons for the areas of disagreement are discussed. 31 refs., 3 figs., 22 tabs

Thermohydraulic and safety analysis on China advanced research reactor under station blackout accident

International Nuclear Information System (INIS)

Tian Wenxi; Qiu Suizheng; Su Guanghui; Jia Dounan; Liu Xingmin; Zhang Jianwei

2007-01-01

A thermohydraulic and safety analysis code-TSACC has been developed using Fortran90 language to evaluate the transient thermohydraulic behavior of the China advanced research reactor (CARR) under station blackout accident (SBA). For the development of TSACC, a series of corresponding mathematical and physical models were applied. Point reactor neutron kinetics model was adopted for solving the reactor power. All possible flow and heat transfer conditions under station blackout accident were considered and the optional correlations were supplied. The usual finite difference method was abandoned and the integral technique was adopted to evaluate the temperature field of the plate type fuel elements. A new simple and convenient equation was proposed for the resolution of the transient behaviors of the main pump instead of the complicated four-quadrant model. Gear method and Adams method were adopted alternately for a better solution to the stiff differential equations describing the dynamic behavior of the CARR. The computational result of TSACC showed the adequacy of the safety margin of CARR under SBA. For the purpose of Verification and Validation (V and V), the simulated results of TSACC were compared with those of RELAP5/MOD3 and a good agreement was obtained. The adoption of modular programming techniques enables TASCC to be applied to other reactors by easily modifying the corresponding function modules

Safety Evaluation Report related to the operation of Hope Creek Generating Station (Docket No. 50-354)

International Nuclear Information System (INIS)

1984-10-01

The Safety Evaluation Report for the application filed by Public Service Electric and Gas Company, as applicant, for a license to operate the Hope Creek Generating Station (Docket No. 50-354), has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is located in Salem County, New Jersey. Subject to favorable resolution of the items discussed in this report, the NRC staff concludes that the facility can be operated by the applicant without endangering the health and safety of the public

Analysis of economics and safety to cope with station blackout in PWR

International Nuclear Information System (INIS)

Al Shehhi, Ahmed Saeed; Chang, Soon Heung; Kim, Sang Ho; Kang, Hyun Gook

2013-01-01

Highlights: • Proposed framework covers all aspects of very complicated decision making. • We addressed the various options against SBO. • Emergency water supply through the steam generator hookup was considered. • Optimal testing interval of EDG was determined in various design options. • Effect of risk aversion factor on decision making was quantitatively illustrated. - Abstract: Design and operation options that can reduce both the initiating event frequency and the accident mitigation probability were addressed in an integrated framework to cope with station blackout. The safety, engineering cost, water delivery cost and testing/maintenance cost of each option were quantitatively evaluated to calculate the cost variation and to find an optimal point in the reference reactor, OPR1000. Design variables that represent additional emergency water supply, diverse emergency diesel generator, and surveillance test period modification were investigated. Based on these design variables, we applied the developed formula to quantify cost items, which were presented as changes of the economics and the safety. A case study was provided to illustrate the change of the total cost. Different risk aversion factors that represent different attitudes of the public were also investigated. The result shows that the costs and benefits of various complicated options can be effectively addressed with the proposed risk-informed decision making framework

Power station instrumentation

International Nuclear Information System (INIS)

Jervis, M.W.

1993-01-01

Power stations are characterized by a wide variety of mechanical and electrical plant operating with structures, liquids and gases working at high pressures and temperatures and with large mass flows. The voltages and currents are also the highest that occur in most industries. In order to achieve maximum economy, the plant is operated with relatively small margins from conditions that can cause rapid plant damage, safety implications, and very high financial penalties. In common with other process industries, power stations depend heavily on control and instrumentation. These systems have become particularly significant, in the cost-conscious privatized environment, for providing the means to implement the automation implicit in maintaining safety standards, improving generation efficiency and reducing operating manpower costs. This book is for professional instrumentation engineers who need to known about their use in power stations and power station engineers requiring information about the principles and choice of instrumentation available. There are 8 chapters; chapter 4 on instrumentation for nuclear steam supply systems is indexed separately. (Author)

International Cooperation in the Field of International Space Station (ISS) Payload Safety

Science.gov (United States)

Grayson, C.; Sgobba, T.; Larsen, A.; Rose, S.; Heimann, T.; Ciancone, M.; Mulhern, V.

2005-12-01

In the frame of the International Space Station (ISS) Program cooperation, in 1998 the European Space Agency (ESA) approached the National Aeronautics and Space Administration (NASA) with the unique concept of a Payload Safety Review Panel (PSRP) "franchise" based at the European Space Technology Center (ESTEC), where the panel would be capable of autonomously reviewing flight hardware for safety. This paper will recount the course of an ambitious idea as it progressed into a fully functional reality. It will show how a panel initially conceived at NASA to serve a national programme has evolved into an international safety cooperation asset. The PSRP established at NASA began reviewing ISS payloads approximately in late 1994 or early 1995 as an expansion of the pre- existing Shuttle Program PSRP. This paper briefly describes the fundamental Shuttle safety process and the establishment of the safety requirements for payloads intending to use the Space Transportation System and ISS. The paper will also offer some historical statistics about the experiments that completed the payload safety process for Shuttle and ISS. The paper then presents the background of ISS agreements and international treaties that had to be considered when establishing the ESA PSRP. The paper will expound upon the detailed franchising model, followed by an outline of the cooperation charter approved by the NASA Associate Administrator, Office of Space Flight, and ESA Director of Manned Spaceflight and Microgravity. The paper will then address the resulting ESA PSRP implementation and its success statistics to date. Additionally, the paper presents ongoing developments with the Japan Aerospace Exploration Agency (JAXA). The discussion will conclude with ideas for future developments, such to achieve a fully integrated international system of payload safety panels for ISS.

Safety assessment of ammonia as a transport fuel

Energy Technology Data Exchange (ETDEWEB)

Duijm, N.J.; Markert, F.; Lundtang paulsen, Jette

2005-02-01

This report describes the safety study performed as part of the EU supported project 'Ammonia Cracking for Clean Electric Power Technology' The study addresses the following activities: safety of operation of the ammonia-powered vehicle under normal and accident (collision) conditions, safety of transport of ammonia to the refuelling stations and safety of the activities at the refuelling station (unloading and refuelling). Comparisons are made between the safety of using ammonia and the safety of other existing or alternative fuels. The conclusion is that the hazards in relation to ammonia need to be controlled by a combination of technical and regulatory measures. The most important requirements are: - Advanced safety systems in the vehicle - Additional technical measures and regulations are required to avoid releases in maintenance workshops and unauthorised maintenance on the fuel system - Road transport of ammonia to refuelling stations in refrigerated form - Sufficient safety zones between refuelling stations and residential or otherwise public areas. When these measures are applied, the use of ammonia as a transport fuel wouldnt cause more risks than currently used fuels (using current practice). (au)

Integrated Plant Safety Assessment, Systematic Evaluation Program: Yankee Nuclear Power Station (Docket No. 50-29)

International Nuclear Information System (INIS)

1987-10-01

The US Nuclear Regulatory Commission (NRC) has prepared Supplement 1 to the final Integrated Plant Safety Assessment Report (IPSAR) (NUREG-0825), under the scope of the Systematic Evaluation Program (SEP), for Yankee Atomic Electric Company's Yankee Nuclear Power Station located in Rowe, Massachusetts. The SEP was initiated by the NRC to review the design of older operating nuclear power plants to reconfirm and document their safety. This report documents the review completed under the SEP for those issues that required refined engineering evaluations or the continuation of ongoing evaluations after the Final IPSAR for the Yankee plant was issued. The review has provided for (1) an assessment of the significance of differences between current technical positions on selected safety issues and those that existed when Yankee was licensed, (2) a basis for deciding how these differences should be resolved in an integrated plant review, and (3) a documented evaluation of plant safety. 2 tabs

Safety of Ikata Nuclear Power Station from the accident of Three Mile Island

International Nuclear Information System (INIS)

Nonaka, Hiroshi

1979-01-01

The leak of radioactive substances occurred on March 28, 1979, in the No. 2 plant of Three Mile Island Nuclear Power Station, and this accident must be put to use to prevent similar accidents and to secure safety hereafter in the nuclear power stations being operated in Japan. In the TMI accident, too many problems concerning the operation management seemed to exist in a series of events. In this paper, a few matters related to the TMI accident among the aspects of the operation management in Ikata Nuclear Power Station are reported. As the problems of operation management, it is considered that the operation of the TMI plant was continued as the exit valve of auxiliary feed line was closed, that it took long time to close the root valve for a pressurizer relief valve manually, and that the ECCS was stopped manually. In TMI, the abnormal phenomenon of losing main feed water has occurred 6 times since the attainment of criticality in March, 1978, and the opening and sticking of pressurizer relief valves occurred at least twice in about 150 times of their actuation in the nuclear reactors designed by Babcock and Wilcox Co. In Ikata Nuclear Power Station, these problems are detected early and the suitable measures are taken immediately, therefore it never happens to continue the operation as the problems are left as they are. It is not conceivable that similar troubles occur many times. (Kako, I.)

Results of evaluation of periodic safety review for No. 1 plant in Mihama Power Station, Kansai Electric Power Co., Inc

International Nuclear Information System (INIS)

1994-01-01

No. 1 plant in Mihama Power Station started the commercial power generation in November, 1970, and has continued the operation for more than 23 years. During this period, the counter measures to troubles, periodic inspections and the maintenance by the electric power company have been carried out. These states of No. 1 plant in Mihama Power Station for more than 23 years are to be recollected from the view-points of the comprehensive evaluation of operation experiences and the reflection of latest technological knowledge, and the safety and reliability are to be further improved in the periodic safety review. Agency of Natural Resources and Energy evaluated the report of the periodic safety review for No. 1 plant in Mihama Power Station made by Kansai Electric Power Co., and summarized the results. The course of the evaluation of the report is shown. The facility utilization factor is 43.3% on the average of about 23 years, but in the last 10 years, it was improved to 69.4%. In the last five years, the rate of occurrence of unexpected shutoff was 0.6 times/year. These are the results of preventive maintenance and the improvement of the facilities and operation management. Operation management, maintenance management, fuel management, radiation control, and radioactive waste management have been carried out properly. The work plan for preventing disasters was established, and the experience of troubles and the latest technological knowledge were well reflected to improve the safety. (K.I.)

Safety-evaluation report related to operation of McGuire Nuclear Station, Units 1 and 2. Docket Nos. 50-369 and 50-370

International Nuclear Information System (INIS)

1983-05-01

This report supplements the Safety Evaluation Report Related to the Operation of McGuire Nuclear Station, Units 1 and 2 (SER (NUREG-0422)) issued in March 1978 by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission with respect to the application filed by Duke Power Company, as applicant and owner, for licenses to operate the McGuire Nuclear Station, Units 1 and 2 (Docket Nos. 50-369 and 50-370). The facility is located in Mecklenburg County, North Carolina, about 17 mi north-northwest of Charlotte, North Carolina. This supplement provides information related to issuance of a full-power authorization for Unit 2. The staff concludes that the McGuire Nuclear Station can be operated by the licensee without endangering the health and safety of the public

Contribution of NDT to the safety of pressurized components in power stations

International Nuclear Information System (INIS)

Mletzko, U.; Maier, H.J.

1994-01-01

In the eyes on the MPA Stuttgart, the nondestructive testing has a very high weight relating to the safety of pressure components in power stations (concept of basis safety). In this connection, the performance verification by NDT has a special significance. A qualification of NDT-techniques can be, indeed, executed in the initial stage at test bodies with artificial faults, known with respect to position, size and type. Even theoretical (modelling) considerations can be integrated. For a performance verification in a closer sense however, this is not sufficient. The performance verification should be effected for the overall system, composed of hardware, software and examination personnel at components having the scale of 1:1 (Full Scale) under realistic boundary conditions and given times. The components must have natural or quasi-natural faults in a certain quality. The informative value of performance verifications is considerably increased, when executed as authenitic dry runs, and when the fault state is subsequently verified by destructive (metallographical) methods. (orig.) [de

On the complex analysis of the reliability, safety, and economic efficiency of atomic electric power stations

International Nuclear Information System (INIS)

Emel'yanov, I.Ya.; Klemin, A.I.; Polyakov, E.F.

1977-01-01

The problem is posed of effectively increasing the engineering performance of nuclear electric power stations (APS). The principal components of the engineering performance of modern large APS are considered: economic efficiency, radiation safety, reliability, and their interrelationship. A nomenclature is proposed for the quantitative indices which most completely characterize the enumerated properties and are convenient for the analysis of the engineering performance. The urgent problem of developing a methodology for the complex analysis and optimization of the principal performance components is considered; this methodology is designed to increase the efficiency of the work on high-performance competitive APS. The principle of complex optimization of the reliability, safety, and economic-efficiency indices is formulated; specific recommendations are made for the practical realization of this principle. The structure of the complex quantiative analysis of the enumerated performance components is given. The urgency and promise of the complex approach to solving the problem of APS optimization is demonstrated, i.e., the solution of the problem of creating optimally reliable, fairly safe, and maximally economically efficient stations

Space station propulsion requirements study

Science.gov (United States)

Wilkinson, C. L.; Brennan, S. M.

1985-01-01

Propulsion system requirements to support Low Earth Orbit (LEO) manned space station development and evolution over a wide range of potential capabilities and for a variety of STS servicing and space station operating strategies are described. The term space station and the overall space station configuration refers, for the purpose of this report, to a group of potential LEO spacecraft that support the overall space station mission. The group consisted of the central space station at 28.5 deg or 90 deg inclinations, unmanned free-flying spacecraft that are both tethered and untethered, a short-range servicing vehicle, and a longer range servicing vehicle capable of GEO payload transfer. The time phasing for preferred propulsion technology approaches is also investigated, as well as the high-leverage, state-of-the-art advancements needed, and the qualitative and quantitative benefits of these advancements on STS/space station operations. The time frame of propulsion technologies applicable to this study is the early 1990's to approximately the year 2000.

Air and radiation monitoring stations

CERN Multimedia

AUTHOR|(SzGeCERN)582709

2015-01-01

CERN has around 100 monitoring stations on and around its sites. New radiation measuring stations, capable of detecting even lower levels of radiation, were installed in 2014. Two members of HE-SEE group (Safety Engineering and Environment group) in front of one of the new monitoring stations.

Refurbishment of Point Lepreau Generating Station

International Nuclear Information System (INIS)

Thompson, P.D.; Jaitly, R.; Ichiyen, N.; Petrilli, M.A.

2004-01-01

NB Power is planning to conduct an 18-month maintenance outage of the Point Lepreau Generating Station (PLGS) beginning in April 2008. The major activity would be the replacement of all 380 Fuel Channel and Calandria Tube Assemblies and the connecting feeder pipes. This activity is referred to as Retube. NB Power would also take advantage of this outage to conduct a number of repairs, replacements, inspections and upgrades (such as rewinding or replacing the generator, replacement of shutdown system trip computers, replacement of certain valves and expansion joints, inspection of systems not normally accessible, etc). These collective activities are referred to as Refurbishment. This would allow the station to operate for an additional 25 to 30 years. The scope of the project was determined from the outcome of a two-year study involving a detailed condition assessment of the station that examined issues relating to ageing and obsolescence. The majority of the plant components were found to be capable of supporting extended operation without needing replacement or changes. In addition to the condition assessment, a detailed review of Safety and Licensing issues associated with extended operation was performed. This included a review of known regulatory and safety issues, comparison of the station against current codes and standards, and comparison of the station against safety related modifications made to more recent CANDU 6 units. Benefit cost analyses (BCA) were performed to assist the utility in determining which changes were appropriate to include in the project scope. As a Probabilistic Safety Assessment (PSA) for PLGS did not exist at the time, a risk baseline for the station had to be determined for use in the BCA. Extensive dialogue with the Canadian Nuclear Safety Commission staff was also undertaken during this phase. A comprehensive Licensing Framework was produced upon which the CNSC provided feedback to NB Power. This feedback was important in terms of

Safety Evaluation Report related to the operation of Hope Creek Generating Station (Docket No. 50-354)

International Nuclear Information System (INIS)

1985-10-01

Supplement No. 3 to the Safety Evaluation Report on the application filed by Public Service Electric and Gas Company on its own behalf as co-owner and as agent for the other co-owner, the Atlantic City Electric Company, for a license to operate Hope Creek Generating Station has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is located in Lower Alloways Creek Township in Salem County, New Jersey. This supplement reports the status of certain items that had not been resolved at the time of publication of the Safety Evaluation Report. 6 tabs

Safety evaluation report related to the operation of Beaver Valley Power Station, Unit 2 (Docket No. 50-412)

International Nuclear Information System (INIS)

1986-05-01

This report, Supplement No. 1 to the Safety Evaluation Report for the application filed by the Duquesne Light Company et al. (the applicant) for a license to operate the Beaver valley Power Station, Unit 2 (Docket No. 50-412), has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. This supplement reports the status of certain items that had not been resolved at the time the Safety Evaluation Report was published

Safety-evaluation report related to the operation of Waterford Steam Electric Station, Unit No. 3. Docket No. 50-382

International Nuclear Information System (INIS)

1983-06-01

Supplement 5 to the Safety Evaluation Report for the application filed by Louisiana Power and Light Company for a license to operate the Waterford Steam Electric Station, Unit 3 (Docket No. 50-382), located in St. Charles Parish, Louisiana has been prepared by the Office of Nuclear Reactor Regulation of the Nuclear Regulatory Commission. The purpose of this supplement is to update the Safety Evaluation Report by providing the staff's evaluation of information submitted by the applicant since the Safety Evaluation Report and its four previous Supplements were issued

Safety Evaluation Report related to the operation of Waterford Steam Electric Station, Unit No. 3 (Docket No. 50-382)

International Nuclear Information System (INIS)

1985-03-01

Supplement 10 to the Safety Evaluation Report for the application filed by Louisiana Power and Light Company for a license to operate the Waterford Steam Electric Station, Unit 3 (Docket No. 50-382), located in St. Charles Parish, Louisiana, has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The purpose of this supplement is to update the Safety Evaluation Report by providing the staff's evaluation of information submitted by the licensee since the Safety Evaluation Report and its nine previous supplements were issued

Safety analysis report: A comparison of incidents from Safety Years 2006 through 2010, USDA Forest Service, Rocky Mountain Research Station Inventory and Monitoring Program

Science.gov (United States)

Devon Donahue

2012-01-01

This paper is an analysis of 5 years of accident data for the USDA Forest Service, Rocky Mountain Research Station (RMRS) Inventory and Monitoring (IM) Program that identifies past trends, allows for standardized self-comparison, and increases our understanding of the true costs of injuries and accidents. Measuring safety is a difficult task. While most agree that...

Organization of the operating quality in EDF nuclear power stations

International Nuclear Information System (INIS)

Stolz, J.

1976-01-01

The organization of operating quality in EDF nuclear stations cover a number of planned and systematic actions of technical and management order carried on at station level and Nuclear Safety Department level. Priority is given to safety quality which has to remain the same during the whole life of the stations; the safety of a station depending from its designing, realization and starting up quality on one hand and from its operating methods on the other [fr

47 CFR 80.1119 - Receipt and acknowledgement of distress alerts by coast stations and coast earth stations.

Science.gov (United States)

2010-10-01

... COMMUNICATIONS COMMISSION (CONTINUED) SAFETY AND SPECIAL RADIO SERVICES STATIONS IN THE MARITIME SERVICES Global Maritime Distress and Safety System (GMDSS) Operating Procedures for Distress and Safety Communications...

Safety Evaluation Report related to the operation of Beaver Valley Power Station, Unit 2 (Docket No. 50-412)

International Nuclear Information System (INIS)

1987-05-01

This report, Supplement No. 5 to the Safety Evaluation Report for the application filed by the Duquesne Light Company et al. (the applicant) for a license to operate the Beaver Valley Power Station, Unit 2 (Docket No. 50-412), has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. This supplement reports the status of certain items that had not been resolved when the Safety Evaluation Report and its Supplements 1, 2, 3, and 4 were published

Safety evaluation report related to the operation of Beaver Valley Power Station, Unit 2 (Docket No. 50-412)

International Nuclear Information System (INIS)

1987-03-01

This report, Supplement No. 4 to the Safety Evaluation Report for the application filed by the Duquesne Light Company et al. (the applicant) for a license to operate the Beaver Valley Power Station, Unit 2 (Docket No. 50-412), has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. This supplement reports the status of certain items that had not been resolved when the Safety Evaluation Report and its Supplements 1, 2, and 3 were published

76 FR 18395 - Safety Zone; Naval Air Station Corpus Christi Air Show, Oso Bay, Corpus Christi, TX

Science.gov (United States)

2011-04-04

...-AA00 Safety Zone; Naval Air Station Corpus Christi Air Show, Oso Bay, Corpus Christi, TX AGENCY: Coast... zone on the navigable waters of Oso Bay in Corpus Christi, Texas in support of the 2011 Naval Air... entities and very few recreational fisherman utilize this section of Oso Bay, the restriction of vessel...

Electricity supplies in a French nuclear power station

International Nuclear Information System (INIS)

2011-01-01

As the operation of a nuclear power station requires a power supply system enabling this operation as well as the installation safety, this document describes how such systems are designed in the different French nuclear power stations to meet the requirements during a normal operation (when the station produces electricity) or when it is stopped, but also to ensure power supply to equipment ensuring safety functions during an incident or an accident occurring on the installation. More precisely, these safety functions are provided by two independent systems in the French nuclear power stations. Their operation is briefly described. Two different types of nuclear reactors are addressed: pressurised water reactors (PWR) of second generation, EPR (or PWR of third generation)

Safety evaluation report related to the operation of Millstone Nuclear Power Station, Unit No. 3 (Docket No. 50-423)

International Nuclear Information System (INIS)

1985-09-01

The Safety Evaluation Report issued in August 1984 provided the results of the NRC staff review of Northeast Nuclear Energy Company's application for a license to operate the Millstone Nuclear Power Station, Unit No. 3. Supplement No. 1 to that report, issued in March 1985 updated the information contained in the Safety Evaluation Report and addressed the ACRS Report issued on September 10, 1984. The Report, Supplement No. 2 updates the information contained in the Safety Evaluation Report and Supplement No. 1 and addresses prior unresolved items. The facility is located in Waterford Township, New London, Connecticut. 11 refs., 9 tabs

Safety evaluation report related to the operation of River Bend Station (Docket No. 50-458). Supplement No. 2

International Nuclear Information System (INIS)

1985-08-01

Supplement No. 2 to the Safety Evaluation Report on the application filed by Gulf States Utilities Company as applicant and for itself and Cajun Electric Power Cooperative, as owners, for a license to operate River Bend Station has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is located in West Feliciana Parish, near St. Francisville, Louisiana. This supplement reports the status of certain items that had not been resolved at the time the Safety Evaluation Report was published

Safety evaluation report related to the operation of Beaver Valley Power Station, Unit 2 (Docket No. 50-412)

International Nuclear Information System (INIS)

1987-08-01

This report, Supplement No. 6 to the Safety Evaluation Report for the application filed by the Duquesne Light Company et al. (the licensee) for a license to operate the Beaver Valley Power Station, Unit 2 (Docket No. 50-412), has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. This supplement reports the status of certain items that had not been resolved when the Safety Evaluation Report and its Supplements 1, 2, 3, 4, and 5 were published

Hinkley Point 'C' power station public inquiry: proof of evidence on design and safety

International Nuclear Information System (INIS)

George, B.V.

1988-09-01

A public inquiry has been set up to examine the planning application made by the Central Electricity Generating Board (CEGB) for the construction of a 1200 MW Pressurized Water Reactor power station at Hinkley Point (Hinkley Point ''C'') in the United Kingdom. The policy is to replicate the Sizewell ''B'' PWR design. The Hinkley Point ''C'' design is described indicating where changes in the Sizewell ''B'' design have been made to accommodate site differences. These are associated with the civil engineering construction and some of the electrical systems and do not affect the safety case. External hazards differ from site to site and the effect on the safety case of those specific to Hinkley Point are examined. The Chernobyl accident and the assessment of the United Kingdom PWR which was carried out subsequently are reviewed. The assessment indicated that no changes in the Sizewell ''B'' design and safety case were called for as a result of this accident; accident management developments are also reviewed, however. The CEGB's approach to minimizing occupational radiation doses is described. (UK)

Lessons learned while implementing a safety parameter display system at the Comanche Peak steam electric station

International Nuclear Information System (INIS)

Hagar, B.

1987-01-01

With the completion of site Verification and Validation tests, the Safety Parameter Display System (SPDS) will be fully operational at the Comanche Peak Steam Electric Station. Implementation of the SPDS, which began in 1982, included: modifying generic Safety Assessment System Software; developing site-specific displays and features; installing and integrating system equipment into the plant; modifying station heating, ventilation, and air conditioning systems to provide necessary cooling; installing an additional uninterruptible power supply system to provide necessary power; and training station personnel in the operation and use of the system. Lessons learned during this project can be discussed in terms of an ideal SPDS implementation project. Such a project would design and implement an SPDS for a plant that is already under construction or operating, and would progress through a sequence of activities that includes: (1) developing and documenting the system design bases, and including all major design influences; (2) developing a database description and system functional specifications to clarify specific system requirements; (3) developing detailed system hardware and software design specifications to fully describe the system, and to enable identification of necessary site design changes early in the project; (4) implementing the system design; (5) configuring and extensively testing the system prior to routine system operation; and (6) tuning the system after the completion of system installation. The ideal project would include future system users in design development and system testing, and would use Verification and Validation techniques throughout the project to ensure that each sequential step is appropriate and correct

30 CFR 57.12085 - Transformer stations.

Science.gov (United States)

2010-07-01

... 30 Mineral Resources 1 2010-07-01 2010-07-01 false Transformer stations. 57.12085 Section 57.12085 Mineral Resources MINE SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT OF LABOR METAL AND NONMETAL MINE... Underground Only § 57.12085 Transformer stations. Transformer stations shall be enclosed to prevent persons...

Additional safety assessments, follow-up of stress tests of the French nuclear power stations. Action plan of the nuclear safety authority

International Nuclear Information System (INIS)

2012-01-01

This document presents the French national action plan defined by the Nuclear Safety Authority (ASN) in compliance with the recommendations made by the ENSREG (European nuclear safety regulators group). It refers to decisions taken at the national level after the Fukushima accident, recommendations after European stress tests, and recommendations after the extraordinary meeting of contracting parties at the Convention on Nuclear Safety of August 2012. For different topics, this document recalls the recommendations made by the peer review, indicates the ASN's opinion and progress or ASN's requirements for different power stations. The first part addresses the recommendations made by peers and based on the European review. Topics concern natural hazards (effects, detection, inspections and controls related to earthquakes, margin assessment with respect to flooding and natural hazards), the loss of safety systems (cooling systems, electricity supplies, actions, instruments, training, and so on), the management of a severe accident (reference levels, measures, guides, exercises, training, management of contaminated water, radiation protection, premises). The second part deals with topics addressed within the frame of the Convention: national organisations, organisations in an emergency and post-accidental situation, international cooperation. A third part addresses the follow-up of additional measures issued by the ASN

Safety evaluation report related to the operation of Hope Creek Generating Station (Docket No. 50-354). Supplement 2

International Nuclear Information System (INIS)

1985-08-01

Supplement No. 2 to the Safety Evaluation Report on the application filed by Public Service Electric and Gas Company as applicant for itself and Atlantic City Electric Company, as owners, for a license to operate Hope Creek Generating Station has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is located in Lower Alloways Creek Township in Salem County, New Jersey. This supplement reports the status of certain items that had not been resolved at the time of publication of the Safety Evaluation Report

IEEE standard requirements for reliability analysis in the design and operation of safety systems for nuclear power generating stations

International Nuclear Information System (INIS)

Anon.

1976-01-01

The purpose of this standard is to provide uniform, minimum acceptable requirements for the performance of reliability analyses for safety-related systems found in nuclear-power generating stations, but not to define the need for an analysis. The need for reliability analysis has been identified in other standards which expand the requirements of regulations (e.g., IEEE Std 379-1972 (ANSI N41.2-1972), ''Guide for the Application of the Single-Failure Criterion to Nuclear Power Generating Station Protection System,'' which describes the application of the single-failure criterion). IEEE Std 352-1975, ''Guide for General Principles of Reliability Analysis of Nuclear Power Generating Station Protection Systems,'' provides guidance in the application and use of reliability techniques referred to in this standard

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

Safety evaluation report related to the operation of Byron Station, Units 1 and 2 (Docket Nos. STN 50-454 and STN 50-455)

International Nuclear Information System (INIS)

1987-03-01

Supplement No. 8 to the Safety Evaluation Report related to Commonwealth Edison Company's application for licenses to operate the Byron Station, Units 1 and 2, located in Rockvale Township, Ogle County, Illinois, has been prepared by th Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. This supplement provides recent information regarding resolution of the license conditions identified in the SER. Because of the favorable resolution of the items discussed in this report, the staff concludes that the Byron Station, Unit 2 can be operated by the licensee at power levels greater than 5% without endangering the health and safety of the public

Alternative Fuels Data Center: Ethanol Fueling Stations

Science.gov (United States)

... More in this section... Ethanol Basics Benefits & Considerations Stations Locations Infrastructure fueling stations by location or along a route. Infrastructure Development Learn about ethanol fueling infrastructure; codes, standards, and safety; and ethanol equipment options. Maps & Data E85 Fueling Station

The next nuclear power station generation: Beyond-design accident concepts, methods, and action sequence

International Nuclear Information System (INIS)

Asmolov, V.G.; Khakh, O.Ya.; Shashkov, M.G.

1993-01-01

The problem of beyond-design accidents at nuclear stations will not be solved unless a safety culture becomes a basic characteristic of all lines of activity. Only then can the danger of accidents as an objective feature of nuclear stations be eliminated by purposive skilled and responsible activities of those implementing safety. Nuclear-station safety is provided by the following interacting and complementary lines of activity: (1) the design and construction of nuclear stations by properly qualified design and building organizations; (2) monitoring and supervision of safety by special state bodies; (3) control of the station by the exploiting organization; and (4) scientific examination of safety within the above framework and by independent organizations. The distribution of the responsibilities, powers, and right in these lines should be defined by a law on atomic energy, but there is not such law in Russian. The beyond-design accident problem is a key one in nuclear station safety, as it clear from the serious experience with accidents and numerous probabilistic studies. There are four features of the state of this topic in Russia that are of major significance for managing accidents: the lack of an atomic energy law, the inadequacy of the technical standards, the lack of a verified program package for nuclear-station designs in order to calculate the beyond-design accidents and analyze risks, and a lack of approach by designers to such accidents on the basis of international recommendations. This paper gives a brief description of three-forming points in the scientific activity: the general concept of nuclear-station safety, methods of analyzing and providing accident management, and the sequence of actions developed by specialists at this institute in recent years

Evolution of International Space Station Program Safety Review Processes and Tools

Science.gov (United States)

Ratterman, Christian D.; Green, Collin; Guibert, Matt R.; McCracken, Kristle I.; Sang, Anthony C.; Sharpe, Matthew D.; Tollinger, Irene V.

2013-01-01

The International Space Station Program at NASA is constantly seeking to improve the processes and systems that support safe space operations. To that end, the ISS Program decided to upgrade their Safety and Hazard data systems with 3 goals: make safety and hazard data more accessible; better support the interconnection of different types of safety data; and increase the efficiency (and compliance) of safety-related processes. These goals are accomplished by moving data into a web-based structured data system that includes strong process support and supports integration with other information systems. Along with the data systems, ISS is evolving its submission requirements and safety process requirements to support the improved model. In contrast to existing operations (where paper processes and electronic file repositories are used for safety data management) the web-based solution provides the program with dramatically faster access to records, the ability to search for and reference specific data within records, reduced workload for hazard updates and approval, and process support including digital signatures and controlled record workflow. In addition, integration with other key data systems provides assistance with assessments of flight readiness, more efficient review and approval of operational controls and better tracking of international safety certifications. This approach will also provide new opportunities to streamline the sharing of data with ISS international partners while maintaining compliance with applicable laws and respecting restrictions on proprietary data. One goal of this paper is to outline the approach taken by the ISS Progrm to determine requirements for the new system and to devise a practical and efficient implementation strategy. From conception through implementation, ISS and NASA partners utilized a user-centered software development approach focused on user research and iterative design methods. The user-centered approach used on

Design and implementation of an integrated safety management system for compressed natural gas stations using ubiquitous sensor network

International Nuclear Information System (INIS)

Yang, Jae Mo; Ko, Byung Seok; Park, Chulhwan; Ko, Jae Wook; Yoo, Byungtae; Shin, Dongil

2014-01-01

To increase awareness of safety in facilities where hazards may exist, operators, managers, and executive officers on the site should be able to monitor such facilities. However, most compressed natural gas (CNG) service stations in Korea use only local-mode monitoring, with only on-site operators to monitor the facility. To complement this local-mode monitoring, an online safety management system called Ubiquitous-gas safety management system (U-GSMS) was developed. The U-GSMS consists largely of software and hardware. The software consists of systems that can manage safety and operations, while the hardware consists of sensors installed in the gas facility and wireless communication systems using a ubiquitous sensor network (USN) technology that facilitates communication between sensors as well as between sensors and other devices. As these systems are web-based, on-site operators as well as managers and executive officers at the headquarters can more effectively and efficiently perform monitoring and safety management

Design and implementation of an integrated safety management system for compressed natural gas stations using ubiquitous sensor network

Energy Technology Data Exchange (ETDEWEB)

Yang, Jae Mo; Ko, Byung Seok; Park, Chulhwan; Ko, Jae Wook [Kwangwoon University, Seoul (Korea, Republic of); Yoo, Byungtae [National Disaster Management Institute, Seoul (Korea, Republic of); Shin, Dongil [Myongji University, Yongin (Korea, Republic of)

2014-03-15

To increase awareness of safety in facilities where hazards may exist, operators, managers, and executive officers on the site should be able to monitor such facilities. However, most compressed natural gas (CNG) service stations in Korea use only local-mode monitoring, with only on-site operators to monitor the facility. To complement this local-mode monitoring, an online safety management system called Ubiquitous-gas safety management system (U-GSMS) was developed. The U-GSMS consists largely of software and hardware. The software consists of systems that can manage safety and operations, while the hardware consists of sensors installed in the gas facility and wireless communication systems using a ubiquitous sensor network (USN) technology that facilitates communication between sensors as well as between sensors and other devices. As these systems are web-based, on-site operators as well as managers and executive officers at the headquarters can more effectively and efficiently perform monitoring and safety management.

Dungeness 'A' Nuclear Power Station. The findings of NII's assessment of Nuclear Electric's long term safety review

International Nuclear Information System (INIS)

1994-01-01

The assessment is reported of Nuclear Electrics' Long Term Safety Reviews (LTSR) of the Dungeness A magnox reactors. The assessment was undertaken by the Health and Safety Executive's Nuclear Installations Inspectorate (NII) which is responsible for regulating the safety of nuclear installations in the United Kingdom. This was one of a programme of LTSRs for all the UK magnox reactors. The LTSR for each plant was proceeded by a Generic Issues programme. The results of both the LTSR and the Generic Issues programme have been used by NII in forming the conclusions of this assessment. Overall the safety case for Dungeness A is satisfactory for continued operation. A programme of additional modifications and inspections has been put in place which further enhances the safety justification. Reactor operations will continue to be monitored and regulated in accordance with the inspections required under the licensing arrangements. Provided these requirements and the agreed further analysis, improvements and inspections give satisfactory results it is expected that the station will be able to operate safely till each reactor is at least 30 years old. Beyond this point a further Periodic Safety Review will be required. (UK)

Safety Evaluation Report related to the operation of Nine Mile Point Nuclear Station, Unit No. 2 (Docket No. 50-410)

International Nuclear Information System (INIS)

1987-07-01

This report supplements the Safety Evaluation Report (NUREG-1047, February 1985) for the application filed by Niagara Mohawk Power Corporation, as applicant and co-owner, for the license to operate Nine Mile Point Nuclear Station, Unit 2 (Docket No. 50-410). It has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is located near Oswego, New York. This report supports the issuance of the full-power license for Nine Mile Point Nuclear Station, Unit No. 2

Application of best estimate and uncertainty safety analysis methodology to loss of flow events at Ontario's Power Generation's Darlington Nuclear Generating Station

International Nuclear Information System (INIS)

Huget, R.G.; Lau, D.K.; Luxat, J.C.

2001-01-01

Ontario Power Generation (OPG) is currently developing a new safety analysis methodology based on best estimate and uncertainty (BEAU) analysis. The framework and elements of the new safety analysis methodology are defined. The evolution of safety analysis technology at OPG has been thoroughly documented. Over the years, the use of conservative limiting assumptions in OPG safety analyses has led to gradual erosion of predicted safety margins. The main purpose of the new methodology is to provide a more realistic quantification of safety margins within a probabilistic framework, using best estimate results, with an integrated accounting of the underlying uncertainties. Another objective of the new methodology is to provide a cost-effective means for on-going safety analysis support of OPG's nuclear generating stations. Discovery issues and plant aging effects require that the safety analyses be periodically revised and, in the past, the cost of reanalysis at OPG has been significant. As OPG enters the new competitive marketplace for electricity, there is a strong need to conduct safety analysis in a less cumbersome manner. This paper presents the results of the first licensing application of the new methodology in support of planned design modifications to the shutdown systems (SDSs) at Darlington Nuclear Generating Station (NGS). The design modifications restore dual trip parameter coverage over the full range of reactor power for certain postulated loss-of-flow (LOF) events. The application of BEAU analysis to the single heat transport pump trip event provides a realistic estimation of the safety margins for the primary and backup trip parameters. These margins are significantly larger than those predicted by conventional limit of the operating envelope (LOE) analysis techniques. (author)

Safety evaluation report related to the operation of Waterford Steam Electric Station, Unit No. 3 (Docket No. 50-382). Suppl.6

International Nuclear Information System (INIS)

1984-06-01

Supplement 6 to the Safety Evaluation Report for the application filed by Louisiana Power and Light Company for a license to operate the Waterford Steam Electric Station, Unit 3 (Docket No. 50-382), located in St. Charles Parish, Louisiana, has been prepared by the Office of Nuclear Reactor Regulation of the Nuclear Regulatory Commission. The purpose of this supplement is to update the Safety Evaluation Report by providing the staff's evaluation of information submitted by the applicant since the Safety Evaluation Report and its five previous supplements were issued

47 CFR 80.333 - Stations in the maritime mobile-satellite service.

Science.gov (United States)

2010-10-01

... SPECIAL RADIO SERVICES STATIONS IN THE MARITIME SERVICES Safety Watch Requirements and Procedures Distress, Alarm, Urgency and Safety Procedures § 80.333 Stations in the maritime mobile-satellite service. The... 47 Telecommunication 5 2010-10-01 2010-10-01 false Stations in the maritime mobile-satellite...

The safety assessment of OPR-1000 nuclear power plant for station blackout accident applying the combined deterministic and probabilistic procedure

Energy Technology Data Exchange (ETDEWEB)

Kang, Dong Gu, E-mail: littlewing@kins.re.kr [Korea Institute of Nuclear Safety, 62 Gwahak-ro, Yuseong-gu, Daejeon 305-338 (Korea, Republic of); Korea Advanced Institute of Science and Technology, 291 Daehak-ro, Yuseong-gu, Daejeon 305-701 (Korea, Republic of); Chang, Soon Heung [Korea Advanced Institute of Science and Technology, 291 Daehak-ro, Yuseong-gu, Daejeon 305-701 (Korea, Republic of)

2014-08-15

Highlights: • The combined deterministic and probabilistic procedure (CDPP) was proposed for safety assessment of the BDBAs. • The safety assessment of OPR-1000 nuclear power plant for SBO accident is performed by applying the CDPP. • By estimating the offsite power restoration time appropriately, the SBO risk is reevaluated. • It is concluded that the CDPP is applicable to safety assessment of BDBAs without significant erosion of the safety margin. - Abstract: Station blackout (SBO) is a typical beyond design basis accident (BDBA) and significant contributor to overall plant risk. The risk analysis of SBO could be important basis of rulemaking, accident mitigation strategy, etc. Recently, studies on the integrated approach of deterministic and probabilistic method for nuclear safety in nuclear power plants have been done, and among them, the combined deterministic and probabilistic procedure (CDPP) was proposed for safety assessment of the BDBAs. In the CDPP, the conditional exceedance probability obtained by the best estimate plus uncertainty method acts as go-between deterministic and probabilistic safety assessments, resulting in more reliable values of core damage frequency and conditional core damage probability. In this study, the safety assessment of OPR-1000 nuclear power plant for SBO accident was performed by applying the CDPP. It was confirmed that the SBO risk should be reevaluated by eliminating excessive conservatism in existing probabilistic safety assessment to meet the targeted core damage frequency and conditional core damage probability. By estimating the offsite power restoration time appropriately, the SBO risk was reevaluated, and it was finally confirmed that current OPR-1000 system lies in the acceptable risk against the SBO. In addition, it is concluded that the CDPP is applicable to safety assessment of BDBAs in nuclear power plants without significant erosion of the safety margin.

Integrated-plant-safety assessment Systematic Evaluation program. Millstone Nuclear Power Station, Unit 1, Northeast Nuclear Energy Company, Docket No. 50-245

International Nuclear Information System (INIS)

1982-11-01

The Systematic Evaluation Program was initiated in February 1977 to review the designs of older operating nuclear reactor plants to reconfirm and document their safety. The review provides: (1) an assessment of how these plants compare with current licensing safety requirements relating to selected issues, (2) a basis for deciding on how these differences should be resolved in an integrated plant review, and (3) a documented evaluation of plant safety. This report documents the review of the Millstone Nuclear Power Station, Unit 1, operated by Northeast Nuclear Energy Company (located in Waterford, Connecticut). Millstone Nuclear Power Station, Unit 1, is one of ten plants reviewed under Phase II of this program. This report indicates how 137 topics selected for review under Phase I of the program were addressed. Equipment and procedural changes have been identified as a result of the review. It is expected that this report will be one of the bases in considering the issuance of a full-term operating license in place of the existing provisional operating license

Safety Evaluation Report related to the operation of Clinton Power Station, Unit No. 1, Docket No. 50-461

International Nuclear Information System (INIS)

1983-05-01

Supplement No. 2 to the Safety Evaluation Report on the application filed by Illinois Power Company, Soyland Power Cooperative, Inc., and Western Illinois Power Cooperative, Inc., as applicants and owners, for a license to operate the Clinton Power Station, Unit No. 1, has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is located in Harp Township, DeWitt County, Illinois. This supplement reports the status of certain items that had not been resolved at the time of publication of the Safety Evaluation Report and Supplement No. 1

Safety Evaluation Report related to the operation of Beaver Valley Power Station, Unit 2 (Docket No. 50-412)

International Nuclear Information System (INIS)

1985-10-01

This Safety Evaluation Report on the application filed by Duquesne Light Company, as applicant and agent for the owners, for a license to operate the Beaver Valley Power Station Unit 2 (Docket No. 50-412) has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is located in Shippingport Borough, Beaver County, Pennsylvania, on the south bank of the Ohio River. Subject to the favorable resolution of the items discussed in this report, the staff concludes that the facility can be operated by the applicant without endangering the health and safety of the public

Safety Evaluation Report related to the operation of Hope Creek Generating Station (Docket No. 50-354). Supplement No. 1

International Nuclear Information System (INIS)

1985-03-01

Supplement No. 1 to the Safety Evaluation Report on the application filed by Public Service Electric and Gas Company as applicant for itself and Atlantic City Electric Company, as owners, for a license to operate Hope Creek Generating Station has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is located in Lower Alloways Creek Township in Salem County, New Jersey. This supplement reports the status of certain items that had not been resolved at the time of publication of the Safety Evaluation Report

Radioecological studies tied to the French nuclear power station programme: aims, nature and mode of execution

International Nuclear Information System (INIS)

Delile, G.

1980-01-01

Under present French practice, assessing the effects of radioactive discharges due to the various likely accident situations comes under 'L'Analyse de Surete Nucleaire' (Nuclear Safety Analysis). The assessment of radioactive discharges relating to normal working comes within the framework of radioecologic studies. Radioecology studies are undertkaen for every power station project. They consist in: - studying what is to be done with radioelements discharged as liquids or gases, - estimating their impact on the populations in the areas of influence of the power station and checking that this impact is permissible, - establishing a surveillance measuring programme for checking against the predictions made and, if required, determining the modifications to be made to the facilities or to their method of operation [fr

Studies on the aseismatic property of BWR type nuclear power stations

International Nuclear Information System (INIS)

Kasai, Hiroaki

1978-01-01

The social requirements for the safety of nuclear power stations are very severe, and in Japan where earthquakes occur frequently, consideration is given so that reactors are shut off safely in case of very large earthquakes which scarcely occur and keep that state. Nuclear power stations are the integrated system combining numerous equipments so as to demonstrate the prescribed functions, and there is large difference in the roles regarding safety of respective equipments. Therefore it is important to maintain the overall safety effectively. All buildings, equipments and pipings are classified according to the degree of importance, and the design is carried out so as to give the aseismatic property corresponding to the degree of importance. The tests and researches concerning the aseismatic property are roughly divided into those for establishing reasonable design conditions, those concerning the improvement of accuracy of the analysis of earthquake response, and those concerning the demonstration of the maintenance of functions. In this paper, the recent state of typical examples is described, such as making of artificial earthquake waves, analysis of building-ground system, vibration experiment on fuel assemblies and demonstration test on control rod insertion. In case of carrying out the design with large earthquake acceleration, it is necessary to evaluate earthquake response accurately by the analysis close to reality. (Kako, I.)

Nuclear Safety

International Nuclear Information System (INIS)

1978-09-01

In this short paper it has only been possible to deal in a rather general way with the standards of safety used in the UK nuclear industry. The record of the industry extending over at least twenty years is impressive and, indeed, unique. No other industry has been so painstaking in protection of its workers and in its avoidance of damage to the environment. Headings are: introduction; how a nuclear power station works; radiation and its effects (including reference to ICRP, the UK National Radiological Protection Board, and safety standards); typical radiation doses (natural radiation, therapy, nuclear power programme and other sources); safety of nuclear reactors - design; key questions (matters of concern which arise in the public mind); safety of operators; safety of people in the vicinity of a nuclear power station; safety of the general public; safety bodies. (U.K.)

IEEE Std 600: IEEE trial-use standard requirements for organizations that conduct qualification testing of safety systems equipment for use in nuclear power generating stations

International Nuclear Information System (INIS)

Anon.

1992-01-01

The purpose of this standard is to provide requirements for establishing a program for conducting qualification tests of safety systems equipment used in nuclear power generating stations. Compliance with the requirements of this standard does not assure the adequacy of the qualification tests performed. This standard applies to organizations that conduct qualification tests on equipment that has a definable safety function and is an identifiable part of a safety system for use in nuclear power generating stations. It requires a technical program, a quality assurance program, and a demonstrated ability to meet specified technical requirements. It does not apply to materials tests, production tests, normal performance testing, qualification by analysis, qualification by operating experience, or reliability tests such as diesel-generator multiple start tests. The intent of this standard is to achieve greater consistency, reliability, and reproducibility of test results and to provide adequate control of qualification testing of safety systems equipment

Safety Evaluation Report related to the operation of Waterford Steam Electric Station, Unit No. 3 (Docket No. 50-382). Supplement No. 8

International Nuclear Information System (INIS)

1984-12-01

Supplement 8 to the Safety Evaluation Report for the application filed by Louisiana Power and Light Company for a license to operate the Waterford Steam Electric Station, Unit 3 (Docket No. 50-382), located in St. Charles Parish, Louisiana, has been prepared by the Office of Nuclear Reactor Regulation of the Nuclear Regulatory Commission. The purpose of this supplement is to update the Safety Evaluation Report by providing the staff's evaluation of information submitted by the applicant since the Safety Evaluation Report and its seven previous supplements were issued

Operation feedback of hydrogen filling station

International Nuclear Information System (INIS)

Pregassame, S.; Barral, K.; Allidieres, L.; Charbonneau, T.; Lacombe, Y.

2004-01-01

One of the technical challenges of hydrogen technology is the development of hydrogen infrastructures which satisfy either safety requirements and reliability of filling processes. AIR LIQUIDE realized an hydrogen filling station in Sassenage (France) operational since September 2003. This station is able to fill 3 buses a day up to 350bar by equilibrium with high pressure buffers. In parallel with commercial stations, the group wanted to create a testing ground in real conditions running with several objectives: validate on a full scale bench a simulation tool able to predict the temperature of both gas and cylinder's materials during filling processes; define the best filling procedures in order to reach mass, temperature and filling time targets; analyse the temperature distribution and evolution inside the cylinder; get a general knowledge about hydrogen stations from safety and reliability point of view; operate the first full scale refuelling station in France. The station is also up-graded for 700bar filling from either a liquid hydrogen source or a gas booster, with cold filling possibility. This paper presents the results concerning 350bar filling : thermal effects, optimal filling procedures and influence of parameters such as climatic conditions are discussed. (author)

Safety Evaluation Report related to the operation of Beaver Valley Power Station, Unit 2 (Docket No. 50-412). Supplement No. 2

International Nuclear Information System (INIS)

1986-08-01

This report, Supplement No. 2 to the the Safety Evaluation Report for the application filed by the Duquesne Light Company, et al. (the applicant) for a license to operate the Beaver Valley Power Station Unit 2 (Docket No. 50-412), has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. This supplement reports the status of certain items that had not been resolved at the time the Safety Evaluation Report was published

Safety evaluation report related to the operation of Hope Creek Generation Station (Docket No. 50-354). Supplement No. 5

International Nuclear Information System (INIS)

1986-04-01

Supplement No. 5 to the Safety Evaluation Report on the application filed by Public Service Electric and Gas Company on its own behalf as co-owner and as agent for the other co-owner, the Atlantic City Electric Company, for a license to operate Hope Creek Generating Station has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is located in Lower Alloways Creek Township in Salem County, New Jersey. This supplement reports the status of certain items that had not been resolved at the time of the publication of the Safety Evaluation Report

Safety Evaluation Report related to the operation of Hope Creek Generating Station (Docket No. 50-354). Supplement No. 4

International Nuclear Information System (INIS)

1985-12-01

Supplement No. 4 to the Safety Evaluation Report on the application filed by Public Service Electric and Gas Company on its own behalf as co-owner and as agent for the other co-owner, the Atlantic City Electric Company, for a license to operate Hope Creek Generating Station has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is located in Lower Alloways Creek Township in Salem County, New Jersey. This supplement reports the status of certain items that had not been resolved at the time of publication of the Safety Evaluation Report

Space Station personal hygiene study

Science.gov (United States)

Prejean, Stephen E.; Booher, Cletis R.

1986-01-01

A personal hygiene system is currently under development for Space Station application that will provide capabilities equivalent to those found on earth. This paper addresses the study approach for specifying both primary and contingency personal hygiene systems and provisions for specified growth. Topics covered are system definition and subsystem descriptions. Subsystem interfaces are explored to determine which concurrent NASA study efforts must be monitored during future design phases to stay up-to-date on critical Space Station parameters. A design concept for a three (3) compartment personal hygiene facility is included as a baseline for planned test and verification activities.

Safety Evaluation Report related to the operation of Nine Mile Point Nuclear Station, Unit No. 2 (Docket No. 50-410)

International Nuclear Information System (INIS)

1985-02-01

The Safety Evaluation Report for the application filed by the Niagara Mohawk Power Corporation, as applicant and co-owner, for a license to operate the Nine Mile Point Nuclear Station, Unit 2 (Docket No. 50-410), has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is located near Oswego, New York. Subject to favorable resolution of the items discussed in this report, the NRC staff concludes that the facility can be operated by the applicant without endangering the health and safety of the public

Islands for nuclear power stations

International Nuclear Information System (INIS)

Usher, E.F.F.W.; Fraser, A.P.

1981-01-01

The safety principles, design criteria and types of artificial island for an offshore nuclear power station are discussed with particular reference to siting adjacent to an industrial island. The paper concludes that the engineering problems are soluble and that offshore nuclear power stations will eventually be built but that much fundamental work is still required. (author)

Safety Evaluation Report related to the full-term operating license for Dresden Nuclear Power Station, Unit 2 ( Docket No. 50-237)

International Nuclear Information System (INIS)

1990-10-01

The Safety Evaluation Report for the full-term operating license application filed by Commonwealth Edison Company for the Dresden Nuclear Power Station, Unit 2, has been prepared by the Office of Nuclear Regulation of the US Nuclear Regulatory Commission. The facility is located in Grundy County, Illinois. Subject to favorable resolution of the items discussed in this report, the staff concludes that the facility can continue to be operated without endangering the health and safety of the public. 72 refs

Fire safety in nuclear power stations

International Nuclear Information System (INIS)

Kench, R.L.

1988-01-01

This is the first of a three-part report on the fire hazards in nuclear power stations and some of the precautions necessary. This part lists the United Kingdom reactors, outlines how they work, the fuels used, the use of moderators and coolants and the control systems. Although the risk of fire is no higher than in fossil-fuel stations the consequences can be more serious. The radioactive materials used mean that there is biological shielding round the core, limitations on waste emissions allowed and limited access to some zones. Reliable shut-down systems are needed. Care in the use of water to fight fires must be exercised -it can act as a moderator and cause an otherwise safe core to go critical. The Wigner effect in graphite moderated reactors is explained. Fires in graphite can be extinguished by carbon dioxide. Argon, chlorine and sodium silicate can also be effective. In sodium cooled reactors fires can be allowed to burn themselves out, or TEC and argon could be used to extinguish the flame. (UK)

30 CFR 57.4261 - Shaft-station waterlines.

Science.gov (United States)

2010-07-01

....4261 Mineral Resources MINE SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT OF LABOR METAL AND NONMETAL MINE SAFETY AND HEALTH SAFETY AND HEALTH STANDARDS-UNDERGROUND METAL AND NONMETAL MINES Fire Prevention... located at underground shaft stations and are part of the mine's fire protection system shall have at...

Safety Evaluation Report, related to the operation of Byron Station, Units 1 and 2 (Docket Nos. STN 50-454 and STN 50-455)

International Nuclear Information System (INIS)

1983-11-01

Supplement No. 3 to the Safety Evaluation Report related to Commonwealth Edison Company's application for licenses to operate the Byron Station, Units 1 and 2, located in Rockvale Township, Ogle County, Illinois, has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. This supplement reports the status of certain items that had not been resolved at the time of publication of the Safety Evaluation Report

Safety Evaluation Report related to the operation of Byron Station, Units 1 and 2 (Docket Nos. STN 50-454 and STN 50-455)

International Nuclear Information System (INIS)

1984-05-01

Supplement No. 4 to the Safety Evaluation Report related to Commonwealth Edison Company's application for licenses to operate the Byron Station, Units 1 and 2, located in Rockvale Township, Ogle County, Illinois, has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. This supplement reports the status of certain items that had not been resolved at the time of publication of the Safety Evaluation Report

Safety evaluation report related to the operation of Byron Station, Units 1 and 2. Docket Nos. STN 50-454 and STN 50-455

International Nuclear Information System (INIS)

1983-01-01

Supplement No. 2 to the Safety Evaluation Report related to Commonwealth Edison Company's application for licenses to operate the Byron Station, Units 1 and 2, located in Rockvale Township, Ogle County, Illinois, has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. This supplement reports the status of certain items that had not been resolved at the time of publication of the Safety Evaluation Report

Plant Operation Station for HTR-PM Low Power and Shutdown operation Probabilistic safety analysis

International Nuclear Information System (INIS)

Liu Tao; Tong Jiejuan

2014-01-01

Full range Probabilistic safety analysis (PSA) is one of key conditions for nuclear power plant (NPP) licensing according to the requirement of nuclear safety regulatory authority. High Temperature Gas Cooled Reactor Pebble-bed Module (HTR-PM) has developed construction design and prepared for the charging license application. So after the normal power operation PSA submitted for review, the Low power and Shutdown operation Probabilistic safety analysis (LSPSA) also begin. The results of LSPSA will together with prior normal power PSA results to demonstrate the safety level of HTR-PM NPP Plant Operation Station (POS) is one of important terms in LSPSA. The definition of POS lays the foundation for LSPSA modeling. POS provides initial and boundary conditions for the following event tree and fault tree model development. The aim of this paper is to describe the state-of-the-art of POS definition for HTR-PM LSPSA. As for the first attempt to the high temperature gas cooled reactor module plant, the methodology and procedure of POS definition refers to the LWR LSPSA guidance, and adds to plant initial status analysis due to the HTR-PM characteristics. A specific set of POS grouping vectors is investigate and suggested for HTR-PM NPP, which reflects the characteristics of plant modularization and on-line refueling. As a result, seven POSs are given according to the grouping vectors at the end of the paper. They will be used to the LSPSA modelling and adjusted if necessary. The papers ’work may provide reference to the analogous NPP LSPSA. (author)

Technology, safety, and costs of decommissioning a reference pressurized water reactor power station

International Nuclear Information System (INIS)

Smith, R.I.; Konzek, G.J.; Kennedy, W.E. Jr.

1978-05-01

Safety and cost information was developed for the conceptual decommissioning of a large [1175 MW(e)] pressurized water reactor (PWR) power station. Two approaches to decommissioning, Immediate Dismantlement and Safe Storage with Deferred Dismantlement, were studied to obtain comparisons between costs, occupational radiation doses, potential radiation dose to the public, and other safety impacts. Immediate Dismantlement was estimated to require about six years to complete, including two years of planning and preparation prior to final reactor shutdown, at a cost of $42 million, and accumulated occupational radiation dose, excluding transport operations, of about 1200 man-rem. Preparations for Safe Storage were estimated to require about three years to complete, including 1 1 / 2 years for planning and preparation prior to final reactor shutdown, at a cost of $13 million and an accumulated occupational radiation dose of about 420 man-rem. The cost of continuing care during the Safe Storage period was estimated to be about $80 thousand annually. Accumulated occupational radiation dose during the Safe Storage period was estimated to range from about 10 man-rem for the first 10 years to about 14 man-rem after 30 years or more. The cost of decommissioning by Safe Storage with Deferred Dismantlement was estimated to be slightly higher than Immediate Dismantlement. Cost reductions resulting from reduced volumes of radioactive material for disposal, due to the decay of the radioactive containments during the deferment period, are offset by the accumulated costs of surveillance and maintenance during the Safe Storage period

Safety evaluation report related to the operation of Millstone Nuclear Power Station, Unit No. 3 (Docket No. 50-423). Supplement No. 5

International Nuclear Information System (INIS)

1986-01-01

This report supplements the Safety Evaluation Report (NUREG-1031) issued in July 1984, Supplement 1 issued in March 1985, Supplement 2 issued in September 1985, Supplement 3 issued in November 1985, and Supplement 4 issued in November 1985 by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission with respect to the application filed by Northeast Nuclear Energy Company (licensee and agent for the owners) for a license to operate Millstone Nuclear Power Station, Unit No. 3 (Docket 50-423). The supplement provides more recent information regarding resolution of license conditions identified in the SER. Because of the favorable resolution of the items discussed in this report, the staff concludes that Millstone Nuclear Power Station, Unit No. 3, can be operated by the licensee at power levels greater than 5% without endangering the health and safety of the public. 13 refs

Safety evaluation report related to the operation of Millstone Nuclear Power Station, Unit No. 3 (Docket No. 50-423)

International Nuclear Information System (INIS)

1984-07-01

The Safety Evaluation Report for the application filed by Northeast Nuclear Energy Company, as applicant and agent for the owners, for a license to operate the Millstone Nuclear Power Station Unit 3 (Docket No. 50-423), has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is located in the town of Waterford, New London County, Connecticut, on the north shore of Long Island Sound. Subject to favorable resolution of the items discussed in this report, the NRC staff concludes that the facility can be operated by the applicant without endangering the health and safety of the public

47 CFR 80.389 - Frequencies for maritime support stations.

Science.gov (United States)

2010-10-01

... 47 Telecommunication 5 2010-10-01 2010-10-01 false Frequencies for maritime support stations. 80.389 Section 80.389 Telecommunication FEDERAL COMMUNICATIONS COMMISSION (CONTINUED) SAFETY AND SPECIAL RADIO SERVICES STATIONS IN THE MARITIME SERVICES Frequencies Maritime Support Stations § 80.389...

47 CFR 80.99 - Radiotelegraph station identification.

Science.gov (United States)

2010-10-01

... 47 Telecommunication 5 2010-10-01 2010-10-01 false Radiotelegraph station identification. 80.99 Section 80.99 Telecommunication FEDERAL COMMUNICATIONS COMMISSION (CONTINUED) SAFETY AND SPECIAL RADIO...-General § 80.99 Radiotelegraph station identification. This section applies to coast, ship and survival...

Safety evaluation report related to the operation of Clinton Power Station, Unit No. 1 (Docket No. 50-461). Suppl. 3

International Nuclear Information System (INIS)

1984-05-01

Supplement No. 3 to the Safety Evaluation Report on the application filed by Illinois Power Company, Soyland Power Cooperative, Inc., and Western Illinois Power Cooperative, Inc., as applicants and owners, for a license to operate the Clinton Power Station, Unit No. 1, has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is located in Harp Township, DeWitt County, Illinois. This supplement reports the status of certain items that had not been resolved at the time of publication of the Safety Evaluation Report and Supplements No. 1 and 2

Safety-evaluation report related to the operation of Shoreham Nuclear Power Station, Unit No. 1 (Docket No. 50-322)

International Nuclear Information System (INIS)

1983-09-01

Supplement 4 (SSER 4) to the Safety Evaluation Report on Long Island Lighting Company's application for a license to operate the Shoreham Nuclear Power Station, Unit 1, located in Suffolk County, New York, has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. This supplement addresses several items that have been reviewed by the staff since the previous supplement was issued

Safety evaluation report related to the operation of Shoreham Nuclear Power Station, Unit No. 1. Docket No. 50-322

International Nuclear Information System (INIS)

1983-02-01

Supplement No. 3 to the Safety Evaluation Report of Long Island Lighting Company's application for a license to operate the Shoreham Nuclear Power Station, Unit 1, located in Suffolk County, New York, has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. This supplement addresses several items that have come to light since the previous supplement was issued

Safety Evaluation Report related to the operation of Shoreham Nuclear Power Station, Unit No. 1 (Docket No. 50-322)

International Nuclear Information System (INIS)

1989-04-01

Supplement 10 (SSER 10) to the Safety Evaluation Report on Long Island Lighting Company's application for a license to operate the Shoreham Nuclear Power Station, Unit 1, located in Suffolk County, New York, has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. This supplement addresses several items that have been reviewed by the staff since the previous supplement was issued

Safety evaluation report related to the operation of Braidwood Station, Units 1 and 2 (Docket Nos. 50-456 and 50-457)

International Nuclear Information System (INIS)

1983-11-01

The Safety Evaluation Report for the application filed by the Commonwealth Edison Company, as applicant and owner, for a license to operate Braidwood Station, Units 1 and 2 (Docket Nos. STN 50-456 and STN 50-457), has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is located in Reed Township, Will County, Illinois. Subject to favorable resolution of the items discussed in this report, the staff concludes that the facility can be operated by the applicant without endangering the health and safety of the public

Review of freeboard: Grand Rapids Generating Station

Energy Technology Data Exchange (ETDEWEB)

Groeneveld, J.L.; Harding, W.B.; Bonin, D.V.; Fuchs, D.M. [Acres Manitoba Ltd., Winnipeg, MB (Canada); Warner, B.J. [Manitoba Hydro, Winnipeg, MB (Canada)

2001-10-01

Constructed during the period 1960-1965, the Grand Rapids Generating Station is a 472 MW hydroelectric station located approximately 400 kilometres northwest of Winnipeg, Manitoba, on the Saskatchewan River. An intake structure, four penstocks, a four-unit plus house unit powerhouse, wing walls, extensive dyke structures and a four-bay spillway are the components of the generating station. A little over ten years ago, the Manitoba Hydro Dam Safety Program was initiated. The program included a detailed dam safety review of the Grand rapids Generating Station. A potential deficiency in the freeboard allowance for several of the earthen dykes was revealed by the review process. The dam safety guidelines sponsored by the Canadian Dam Association were not met. The occurrence of a 1:1000 year wind event from the critical direction when the reservoir was at or near its full supply level was compounded by the analysis. The establishment of a wind and wave monitoring program was included in the deficiency studies commissioned. The objective was to confirm the empirical estimates concerning wave height, the development and usage of a two dimensional numerical wave model, and additional freeboard analyses to refine estimates of the recurrence interval of the current level of protection. A statistical Monte Carlo analysis was performed in the case of the estimates of the recurrence interval to determine the joint probabilities of seasonal variations in wind direction, wind speed, and reservoir level. The estimate of the actual risk of overtopping was therefore refined, and the details pertaining to the methodology and the conclusions of the analysis are all included in the paper. 15 refs., 4 tabs., 9 figs.

A study of operation of station fire brigade with comparison for Japan and the U.S

International Nuclear Information System (INIS)

Ichiki, Kuniyasu

2011-01-01

With the occurrence of a fire at Kashiwazaki Kariwa No. 3 unit on a site transformer when Niigata Chuetsu Offshore Earthquake occurred in 2007, the movement of establishing station fire brigade broke out all over Japan. Fire brigade of U.S. power stations is an organization structured with the standards of clear requirements. Some of station fire event reports include examples of activation of station fire brigade. Preparing station fire brigade with good expertise is quite meaningful on the viewpoint of ruling out significant threats of fire against nuclear safety, such as Browns Ferry station fire occurred in 1975. The purpose of this study is to extract useful knowledge through making research on examples of operations of station fire brigade of both Japan and the U.S. As for fire occurrence, relatively frequent fires are seen at Japanese stations with miscellaneous components or some material for works. In the U.S., on the other hand, we can identify relatively frequent fires occurred with important equipment. This might make fire brigades of the U.S. well-prepared for fire fighting by establishing fire-coping strategies for every location in the station. Also, in some stations, fire brigade is continuously trained to enhance its firefighting and rescuing abilities. With the aspect of training of brigades, they are often assembled with devoted staff, which enable them acquiring frequent training and educations to improve their skills further. Fire fighting and rescuing are inevitably dangerous task. This is why frequent training is essential in order to prevent injuries and to keep up their motivation. Unfortunately, the circumstance for brigades in Japan is not helpful enough in experiencing special training needed for them. We have to figure out some way to get over such disadvantages. (author)

Human-factor operating concept for Borssele Nuclear Power Station

International Nuclear Information System (INIS)

Wieman, J.L.

1995-01-01

The safety level in the operation of a reactor is determined basically by human beings. The Borssele Nuclear Power Station has carried out measures for improving the man-machine interface through training and operating instructions for the shift personnel. The retrofitting of control technology relevant to safety engineering should avoid operating instructions which can cause potential failures. A safety study has shown that the remaining risk following all retrofitting measures remains dependent to the extent of 80% on human factors and that human factors as a whole have a positive effect on reactor safety. (orig.) [de

Safety evaluation report related to the operation of Nine Mile Point Nuclear Station, Unit No. 2 (Docket No. 50-410). Supplement No. 1

International Nuclear Information System (INIS)

1985-06-01

This report supplements the Safety Evaluation Report for the application filed by Niagara Mohawk Power Corporation, as applicant and co-owner, for a license to operate the Nine Mile Point Nuclear Station Unit 2. The facility is located near Oswego, New York. Subject to favorable resolution of the issues discussed in this report, the NRC staff concludes that the facility can be operated by the applicant without endangering the health and safety of the public. 1 fig., 3 tabs

AECB staff annual assessment of the Darlington Nuclear Generating Station for the year 1996

International Nuclear Information System (INIS)

1997-06-01

The Atomic Energy Control Board is the independent federal agency that controls all nuclear activities in Canada. A major use of nuclear energy in Canada is electricity production. The AECB assesses every station's performance against legal requirements, including the conditions in the operating licence. Each station is inspected and all aspects of the station's operation and management is reviewed. This report is the AECB staff assessment of safety at the Darlington Nuclear Generating Station for 1996. Ontario Hydro operated the station in a safe manner in 1996. All four special safety systems were fully available 100 percent of the time. There were more problems that affected the safety support systems in 1996 than in the previous year

Safety Evaluation Report related to the operation of Comanche Peak Steam Electric Station, Unit 2 (Docket No. 50-446)

International Nuclear Information System (INIS)

1992-09-01

This document supplement 25 to the Safety Evaluation Report related to the operation of the Comanche Peak Steam Electric Station (CPSES), Unit 2 (NUREG-0797), has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission (NRC). The facility is located in Somervell County, Texas, approximately 40 miles southwest of Fort Worth, Texas. This supplement reports the status of certain issues that had not been resolved when the Safety Evaluation Report and Supplements 1, 2, 3, 4, 6, 12, 21, 22, 23, and 24 to that report were published. This supplement deals primarily with Unit 2 issues; however, it also references evaluations for several Unit 1 licensing items resolved since Supplement 24 was issued

Safety evaluation report related to the operation of Nine Mile Point Nuclear Station, Unit No. 2 (Docket No. 50-410)

International Nuclear Information System (INIS)

1986-07-01

This report supplements the Safety Evaluation Report (NUREG-1047, February 1985) for the application filed by Niagara Mohawk Power Corporation, as applicant and co-owner, for a license to operate the Nine Mile Point Nuclear Station, Unit No. 2 (Docket No. 50-410). It has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is located near Oswego, New York. Supplement 1 to the Safety Evaluation Report was published in June 1985 and contained the report from the Advisory Committee on Reactor Safeguards as well as the resolution to a number of outstanding issues from the Safety Evaluation Report. Supplement 2 was published in November 1985 and contained the resolution to a number of outstanding and confirmatory issues. Subject to favorable resolution of the issues discussed in this report, the NRC staff concludes that the facility can be operated by the applicant without endangering the health and safety of the public

A study on safety climate at nuclear power plants

International Nuclear Information System (INIS)

Fukui, Hirokazu; Yoshida, Michio; Yoshiyama, Naohiro

2001-01-01

In the current study, we define safety climate as an organizational environment that induces members of the organization to give consideration to safety or take safety actions. It is of utmost importance that people holding managerial positions in an organization have a good understanding of the characteristics of the safety climate of the organization and implement safety promotion activities effectively. In the current research, we studied the rating scales and the characteristics of a safety climate. A survey was conducted, targeting technical engineers who belong to the three power stations of Kansai Electric Power Co., Inc. The questionnaire mainly consisted of questions concerning safety measures taken by individuals and questions concerning safety measures taken by the organization, to which the individuals belong. As a result of a factor analysis of the responses, we extracted five factors, namely, 'confidence in knowledge and skill', attitude of supervisors,' 'safety education in workplace', 'clarity of tasks' and 'safety confirmation/report'. In studying the rating scales of the safety climate, we selected five items from each of the above five factors, and used the total scores of the ratings of the five items as scores of each factor. Then, we examined the correlation between scores of personal factors and scores of organizational environment factors. We treated the scores of safety confirmation/report' and 'confidence in knowledge and skill', which are personal factors, as criterion variables, and the scores of 'attitude of supervisors', 'safety education in workplace' and 'clarity of tasks', which are organizational environment factors, as predictor variables. As a result, we found that levels of 'safety confirmation/report' and 'confidence in knowledge and skill' can be deduced from the scores of 'attitude of supervisors', 'safety education in workplace' and 'clarity of tasks.' Hence, we have decided to use these three organizational environment

Results of evaluation of periodic safety review for No. 1 plant in Fukushima No. 1 Nuclear Power Station, Tokyo Electric Power Co., Inc

International Nuclear Information System (INIS)

1994-01-01

No. 1 plant in Fukushima No. 1 Nuclear Power Station started the commercial power generation in March, 1971, and has continued the operation for more than 23 years. During this period, the countermeasures to troubles, periodic inspections, and the maintenance by the electric power company have been carried out. These states are to be recollected from the viewpoints of the comprehensive evaluation of the operation experiences and the reflection of the latest technological knowledge, and the safety and reliability are to be further improved in the periodic safety review. Agency of Natural Resources and Energy evaluated the report of the periodic safety review for No. 1 plant in Fukushima No. 1 Nuclear Power Station, and summarized the results. The course of the evaluation of the report is shown. The facility utilization factor was 50.1% on the average of about 23 years, but in the last 10 years, it was improved to 59.7%. In the last five years, the rate of occurrence of unexpected shutdown was 0.4 times/year. These are the results of preventive maintenance and the improvement of the facilities and operation management. Operation management, maintenance management, fuel management, radiation control, radioactive waste management and the reflection of the experience of troubles and the latest technological knowledge to the improvement of safety have been carried out properly. The work plan for disaster prevention was established. (K.I.)

AECB staff annual assessment of the Point Lepreau Nuclear Generating Station

International Nuclear Information System (INIS)

1997-06-01

The Atomic Energy Control Board is the independent federal agency that controls all nuclear activities in Canada. A major use of nuclear energy in Canada is electricity production. The AECB assesses every station's performance against legal requirements, including the conditions in the operating licence. Each station is inspected and all aspects of the station's operation and management is reviewed. This report is the AECB staff assessment of reactor safety at the Point Lepreau Generating Station in 1996. Point Lepreau operated safely but the worsening trends in NB Power's safety performance leads to the conclusion that urgent action is required. NB Power is required to report formally to the AECB on progress with measures to improve safety management every six months. Further licensing action will be taken on NB Power if it fails to make the improvements

47 CFR 80.1123 - Watch requirements for ship stations.

Science.gov (United States)

2010-10-01

... Section 80.1123 Telecommunication FEDERAL COMMUNICATIONS COMMISSION (CONTINUED) SAFETY AND SPECIAL RADIO SERVICES STATIONS IN THE MARITIME SERVICES Global Maritime Distress and Safety System (GMDSS) Operating... watches for broadcasts of maritime safety information on the appropriate frequency or frequencies on which...

New safety standards of nuclear power station with no requirements of site evaluation. No public dose limit published with possible inappropriateness of reactor site

International Nuclear Information System (INIS)

Takitani, Koichi

2013-01-01

Nuclear Regulation Authority was preparing new safety standards in order to aim at starting safety reviews of existing nuclear power station in July 2013. This article commented on issues of major accident, which was defined as severely damaged core event. Accumulated dose at the site boundary of the Fukushima Daiichi Nuclear Power Station totaled to about 234 mSv on March just after the accident with rare gas of 500 PBq, iodine 131 of 500 PBq, cesium 134 of 10 PBq and cesium 137 of 10 PBq released to the atmosphere, which was beyond 100 mSv. As measures for preventing containment vessel failure after core severely damaged, filtered venting system was required to be installed for low radiological risk to the public. However filter was not effective to rare gas. Accumulated doses at the site boundary of several nuclear power stations after filtered venting with 100% release of rare gas could be estimated to be 2-37 Sv mostly depending on the site condition, which might be surely greater than 100 mSv. Omitting site evaluation for major accident, which was beyond design basis accident, was great concern. (T. Tanaka)

Technical evaluation of the noise and isolation testing of the safety features actuation system at the Davis Besse Nuclear Power Station, Unit 1

International Nuclear Information System (INIS)

Selan, J.C.

1981-07-01

This report documents the technical evaluation of the noise and isolation testing of the safety features actuation system at the Davis Besse Nuclear Power Station, Unit 1. The tests were to verify that faults on the non-Class 1E circuits would not propagate to the Class 1E circuits and degrade them below acceptable levels. The tests conducted demonstrated that the safety features actuation system did not degrade below acceptable levels nor was the system's ability to perform its protective functions affected

Safety evaluation report related to the operation of Limerick Generating Station, Units 1 and 2 (Docket Nos. 50-352-50-353). Supplement 1

International Nuclear Information System (INIS)

1983-12-01

This report supplements the Safety Evaluation Report (NUREG-0991, August 1983) for the application filed by the Philadelphia Electric Company, as applicant and owner, for licenses to operate the Limerick Generating Station Units 1 and 2 (Docket Nos. 50-352 and 50-353). The facility is located near Pottstown, Pennsylvania. Subject to favorable resolution of the items discussed in this report, the NRC staff concludes that the facility can be operated by the applicant without endangering the health and safety of the public

Safety Evaluation Report related to the operation of LaSalle County Station, Units 1 and 2. Docket Nos. 50-373 and 50-374

International Nuclear Information System (INIS)

1984-03-01

This supplement to the Safety Evaluation Report of Commonwealth Edison Company's application for a license to operate its La Salle County Station, Unit 2, located in Brookfield Township, La Salle County, Illinois, has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. This supplement is to update evaluations on Unit 2 issues identified in the previous Safety Evaluation Report and Supplements that need resolution prior to issuance of the full power operating license for Unit 2

Safety-evaluation report related to the operation of Limerick Generating Station, Units 1 and 2 (Docket Nos. 50-352 and 50-353)

International Nuclear Information System (INIS)

1983-08-01

The Safety Evaluation Report for the application filed by the Philadelphia Electric Company, as applicant and owner, for licenses to operate the Limerick Generating Station Units 1 and 2 (Docket Nos. 50-352 and 50-353), has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is located near Pottstown, Pennsylvania. Subject to favorable resolution of the items discussed in this report, the NRC staff concludes that the facility can be operated by the applicant without endangering the health and safety of the public

Safety Evaluation Report related to the operation of Millstone Nuclear Power Station Unit No. 3 (Docket No. 50-4423). Supplement No. 1

International Nuclear Information System (INIS)

1985-03-01

In July 1984 staff of the Nuclear Regulatory Commission issued its Safety Evaluation Report regarding the application of Northeast Nuclear Energy Company (applicant and agent for the owners) for a license to operate Millstone Nuclear Power Station, Unit No. 3 (Docket 50-423). The facility is located in the town of Waterford, New London County, Connecticut, on the north shore of Long Island Sound. This first supplement to NUREG-1031 reports the status of certain items that remained unresolved at the time the Safety Evaluation Report was published

47 CFR 101.503 - Digital Electronic Message Service Nodal Stations.

Science.gov (United States)

2010-10-01

... Service § 101.503 Digital Electronic Message Service Nodal Stations. 10.6 GHz DEMS Nodal Stations may be... 47 Telecommunication 5 2010-10-01 2010-10-01 false Digital Electronic Message Service Nodal Stations. 101.503 Section 101.503 Telecommunication FEDERAL COMMUNICATIONS COMMISSION (CONTINUED) SAFETY...

Safety Evaluation Report related to the operation of Byron Station, Units 1 and 2 (Dockets Nos. STN 50-454 and STN 50-455)

International Nuclear Information System (INIS)

1984-10-01

Supplement No. 5 to the Safety Evaluation Report related to Commonwealth Edison Company's application for licenses to operate the Byron Station, Units 1 and 2, located in Rockvale Township, Ogle County, Illinois, has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. This supplement reports the status of certain items that had not been resolved at the time of publication of the Safety Evaluation Report. Because of the favorable resolution of the items discussed in this report, the staff concludes that there is reasonable assurance that the facility can be operated by the applicant without endangering the health and safety of the public

Hydrogen vehicle fueling station

Energy Technology Data Exchange (ETDEWEB)

Daney, D.E.; Edeskuty, F.J.; Daugherty, M.A. [Los Alamos National Lab., NM (United States)] [and others

1995-09-01

Hydrogen fueling stations are an essential element in the practical application of hydrogen as a vehicle fuel, and a number of issues such as safety, efficiency, design, and operating procedures can only be accurately addressed by a practical demonstration. Regardless of whether the vehicle is powered by an internal combustion engine or fuel cell, or whether the vehicle has a liquid or gaseous fuel tank, the fueling station is a critical technology which is the link between the local storage facility and the vehicle. Because most merchant hydrogen delivered in the US today (and in the near future) is in liquid form due to the overall economics of production and delivery, we believe a practical refueling station should be designed to receive liquid. Systems studies confirm this assumption for stations fueling up to about 300 vehicles. Our fueling station, aimed at refueling fleet vehicles, will receive hydrogen as a liquid and dispense it as either liquid, high pressure gas, or low pressure gas. Thus, it can refuel any of the three types of tanks proposed for hydrogen-powered vehicles -- liquid, gaseous, or hydride. The paper discusses the fueling station design. Results of a numerical model of liquid hydrogen vehicle tank filling, with emphasis on no vent filling, are presented to illustrate the usefulness of the model as a design tool. Results of our vehicle performance model illustrate our thesis that it is too early to judge what the preferred method of on-board vehicle fuel storage will be in practice -- thus our decision to accommodate all three methods.

NASA space station automation: AI-based technology review

Science.gov (United States)

Firschein, O.; Georgeff, M. P.; Park, W.; Neumann, P.; Kautz, W. H.; Levitt, K. N.; Rom, R. J.; Poggio, A. A.

1985-01-01

Research and Development projects in automation for the Space Station are discussed. Artificial Intelligence (AI) based automation technologies are planned to enhance crew safety through reduced need for EVA, increase crew productivity through the reduction of routine operations, increase space station autonomy, and augment space station capability through the use of teleoperation and robotics. AI technology will also be developed for the servicing of satellites at the Space Station, system monitoring and diagnosis, space manufacturing, and the assembly of large space structures.

Safety Evaluation Report related to the operation of Comanche Peak Steam Electric Station, Unit 2 (Docket No. 50-446)

International Nuclear Information System (INIS)

1993-02-01

Supplement 26 to the Safety Evaluation Report related to the operation of the Comanche Peak Steam Electric Station (CPSES), Unit 2, has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission (NRC). The facility is located in Somervell County, Texas, approximately 40 miles southwest of Fort Worth, Texas. This supplement reports the status of certain issues that had not been resolved when the Safety Evaluation Report and Supplements 1, 2, 3, 4, 6, 12, 21, 22, 23, 24, and 25 to that report were published. This supplement deals primarily with Unit 2 issues; however, it also references evaluations for several licensing issues that relate to Unit 1, which have been resolved since Supplement 25 was issued

Safety studies on oxide fuel elements for the fast neutron programme

International Nuclear Information System (INIS)

Ollier, H.; Truffert, J.

1976-01-01

Studies leading to the definition of a fuel element and instructions for its use overlap with safety studies to the extent that they must guarantee, for nominal running conditions, a failure rate consistent with the working rules of the Station. Given the present state of knowledge on the causes and kinetics of damage affecting the cladding, it is mainly dealt with the principles of this research. The advantages of applying statistical computing methods to the two factors which govern the lifetime of the pins: their working conditions and quality, are demonstrated [fr

Study on Evaluation Indicators System of Crowd Management for Transfer Stations Based on Pedestrian Simulation

Directory of Open Access Journals (Sweden)

Guanghou Zhang

2011-12-01

Full Text Available Improving safety and convenience of transfer is one of the most vital tasks in subway system planning, design and operation management. Because of complicated space layout and crowded pedestrian, crowd control is a big challenge for management of transfer stations. Thus, a quantitative evaluation should be done before improvement measures are carried out. Literature review showed that present evaluation indicators about crowd management in subway system were all based on fixed value or experience. Dynamic effect caused by pedestrian congestion and various facility combination cannot be represented based on these indicators. Thus, in this paper, based on the pedestrian simulation tool, dynamic evaluation indicators system of crowd management was established from the point of safety, cost-effectiveness and comfort. In order to aid decision makers to identify the most appropriate scenario to improve the effectiveness of crowd management, Matter-Element Analysis (MEA was used to rate different scenarios. A pedestrian simulation model of a designing intermodal transfer station was built and four different scenarios were tested to demonstrate how to use this indicators system. Simulation results were evaluated based on the dynamic indicators system and MEA. The application results show that the dynamic evaluation indicators system is operational and can reflect level of the crowd management in transfer station comprehensively and precisely.

Integrated plant safety assessment, Systematic Evaluation Program: Dresden Nuclear Power Station, Unit 2 (Docket No. 50-237)

International Nuclear Information System (INIS)

1989-10-01

The US Nuclear Regulatory Commission (NRC) has prepared Supplement 1 to the final Integrated Plant Safety Assessment Report (IPSAR) (NUREG-0823), under the scope of the Systematic Evaluation Program (SEP), for the Commonwealth Edison Company (CECo) Dresden Nuclear Power Station, Unit 2 located in Grundy County, Illinois. The NRC initiated the SEP to provide the framework for reviewing the design of older operating nuclear reactor plants to reconfirm and document their safety. This report documents the review completed by means of the SEP for those issues that required refined engineering evaluations or the continuation of ongoing evaluations subsequent to issuing the final IPSAR for Dresden Unit 2. The review was provided for (1) an assessment of the significance of differences between current technical positions on selected issues and those that existed when Dresden Unit 2 was licensed, (2) a basis for deciding on how these differences should be resolved in an integrated plant review, and (3) a documented evaluation of plant safety. The final IPSAR and this supplement forms part of the bases for considering the conversion of the existing provisional operating license to a full-term operating license. 83 refs., 9 tabs

Space station operating system study

Science.gov (United States)

Horn, Albert E.; Harwell, Morris C.

1988-01-01

The current phase of the Space Station Operating System study is based on the analysis, evaluation, and comparison of the operating systems implemented on the computer systems and workstations in the software development laboratory. Primary emphasis has been placed on the DEC MicroVMS operating system as implemented on the MicroVax II computer, with comparative analysis of the SUN UNIX system on the SUN 3/260 workstation computer, and to a limited extent, the IBM PC/AT microcomputer running PC-DOS. Some benchmark development and testing was also done for the Motorola MC68010 (VM03 system) before the system was taken from the laboratory. These systems were studied with the objective of determining their capability to support Space Station software development requirements, specifically for multi-tasking and real-time applications. The methodology utilized consisted of development, execution, and analysis of benchmark programs and test software, and the experimentation and analysis of specific features of the system or compilers in the study.

A study on safety climate at nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Fukui, Hirokazu [Institute of Nuclear Safety System Inc., Mihama, Fukui (Japan); Yoshida, Michio; Yoshiyama, Naohiro [Japan Institute for Group Dynamics, Fukuoka (Japan)

2001-09-01

In the current study, we define safety climate as an organizational environment that induces members of the organization to give consideration to safety or take safety actions. It is of utmost importance that people holding managerial positions in an organization have a good understanding of the characteristics of the safety climate of the organization and implement safety promotion activities effectively. In the current research, we studied the rating scales and the characteristics of a safety climate. A survey was conducted, targeting technical engineers who belong to the three power stations of Kansai Electric Power Co., Inc. The questionnaire mainly consisted of questions concerning safety measures taken by individuals and questions concerning safety measures taken by the organization, to which the individuals belong. As a result of a factor analysis of the responses, we extracted five factors, namely, 'confidence in knowledge and skill', attitude of supervisors,' 'safety education in workplace', 'clarity of tasks' and 'safety confirmation/report'. In studying the rating scales of the safety climate, we selected five items from each of the above five factors, and used the total scores of the ratings of the five items as scores of each factor. Then, we examined the correlation between scores of personal factors and scores of organizational environment factors. We treated the scores of safety confirmation/report' and 'confidence in knowledge and skill', which are personal factors, as criterion variables, and the scores of 'attitude of supervisors', 'safety education in workplace' and 'clarity of tasks', which are organizational environment factors, as predictor variables. As a result, we found that levels of 'safety confirmation/report' and 'confidence in knowledge and skill' can be deduced from the scores of 'attitude of supervisors', 'safety

On the safety assessment of human exposure in the proximity of cellular communications base-station antennas at 900, 1800 and 2170 MHz

International Nuclear Information System (INIS)

MartInez-Burdalo, M; MartIn, A; Anguiano, M; Villar, R

2005-01-01

In this work, the procedures for safety assessment in the close proximity of cellular communications base-station antennas at three different frequencies (900, 1800 and 2170 MHz) are analysed. For each operating frequency, we have obtained and compared the distances to the antenna from the exposure places where electromagnetic fields are below reference levels and the distances where the specific absorption rate (SAR) values in an exposed person are below the basic restrictions, according to the European safety guidelines. A high-resolution human body model has been located, in front of each base-station antenna as a worst case, at different distances, to compute whole body averaged SAR and maximum 10 g averaged SAR inside the exposed body. The finite-difference time-domain method has been used for both electromagnetic fields and SAR calculations. This paper shows that, for antenna-body distances in the near zone of the antenna, the fact that averaged field values be below the reference levels could, at certain frequencies, not guarantee guidelines compliance based on basic restrictions

On the safety assessment of human exposure in the proximity of cellular communications base-station antennas at 900, 1800 and 2170 MHz.

Science.gov (United States)

Martínez-Búrdalo, M; Martín, A; Anguiano, M; Villar, R

2005-09-07

In this work, the procedures for safety assessment in the close proximity of cellular communications base-station antennas at three different frequencies (900, 1800 and 2170 MHz) are analysed. For each operating frequency, we have obtained and compared the distances to the antenna from the exposure places where electromagnetic fields are below reference levels and the distances where the specific absorption rate (SAR) values in an exposed person are below the basic restrictions, according to the European safety guidelines. A high-resolution human body model has been located, in front of each base-station antenna as a worst case, at different distances, to compute whole body averaged SAR and maximum 10 g averaged SAR inside the exposed body. The finite-difference time-domain method has been used for both electromagnetic fields and SAR calculations. This paper shows that, for antenna-body distances in the near zone of the antenna, the fact that averaged field values be below the reference levels could, at certain frequencies, not guarantee guidelines compliance based on basic restrictions.

Nuclear safety review for qualification of class 1E motor inside containment for nuclear power stations

International Nuclear Information System (INIS)

Li Shixin; Wu Qi; Zhang Yunbo; Wu Caixia

2013-01-01

In nuclear power plants with pressurized water reactors, the review for class 1E motor inside containment qualification process and documents is an important aspect of nuclear safety equipment review, and the reviewers should make evaluations for the qualification test results in terms of the compliance with standard and regulation, and the consistency with inside containment environment. Firstly, this paper introduces the qualification test of class 1E motor inside containment for nuclear power generating stations, such as aging test and design-basis-event test. Second, there is a discussion about typical problems of review. At last, comparison of IEEE334 with RCC-E is conducted and explored. (authors)

Present status of nuclear power safety studies in JAERI

International Nuclear Information System (INIS)

1989-01-01

Japan Atomic Energy Research Institute carries out the safety research required for ensuring safety accompanying the development and utilization of atomic energy, dividing the roles with industrial circules, universities and other public research institutes. First, it carries out the research based on the yearly program of safety research, in which the subjects of safety research that the state should execute are decided. Next, it carries out a series of the safety verification test on atomic energy facilities entrusted by the state using the special account for power source development promotion countermeasures. Besides, it carries out the safety research on the specific subjects entrusted by outside organizations. The object and range of atomic energy safety at present are roughly divided into four fields, that is, nuclear power stations, nuclear fuel facilities, waste treatment and disposal, and environmental radioactivity. In order to advance the research effectively, in addition to the generalization department and the execution department, various committees are organized. In 1988, about 260 persons engaged in these four research fields, and the budget for fiscal year 1988 was about 17.4 billion yen. (K.I.)

Safety culture in nuclear installations. Management of safety and safety culture in Indian NPPs

International Nuclear Information System (INIS)

Rawal, S.C.

2002-01-01

Nuclear Power Corporation Of India Ltd. (NPCIL) is a company owned by Government of India and is responsible for Design, Construction, Commissioning, Operation and Decommissioning of Nuclear Power plants in India. Presently, a total of 13 Nuclear power Stations are in operation with an installed capacity of 2620 MWe and 2 VVR type PWR Units of 1000 MWe capacity each, 2 PHWR type units of 500 MWe capacity each and 4 PHWR type 220 MWe capacity each are under construction. NPPs generation capacity has been increased from 70% to 85% in the span Of last 7 years with high level of safety standards. This could be achieved through Management commitment towards building a strong Safety Culture. Safety culture is that assembly of characteristics and attitudes in organisation and individuals which establishes that as an overriding priority nuclear plant safety issues receives the attention warranted by their significance. This definition of safety culture brings out two major components in its manifestation. The framework within which individuals within the organisation works.The attitude and response of individual towards the safety issues over productivity and economics in the organisational work practices. The two attributes of safety culture are built in and upgraded in each individuals through special training at the time of entry in the organisation and later through in built procedures in the work practices, motivation and encouragement for free participation of each individuals. Individuals are encouraged to participate in Quality circle teams at the sectional level and review of safety proposal originated by individuals in Station operation Review Committee at Station level, in addition to this to continuously enhance the safety culture, refresher training courses are being organised at regular intervals. The safety related proposals are categorised in to two namely: Proposals from Operating Plants, and Proposals from projects and Design. The concept of safety

47 CFR 80.1073 - Radio operator requirements for ship stations.

Science.gov (United States)

2010-10-01

.... 80.1073 Section 80.1073 Telecommunication FEDERAL COMMUNICATIONS COMMISSION (CONTINUED) SAFETY AND SPECIAL RADIO SERVICES STATIONS IN THE MARITIME SERVICES Global Maritime Distress and Safety System (GMDSS... selecting HF DSC guard channels and receiving scheduled maritime safety information broadcasts; (4...

Analysis of tank safety with propane-butane on LPG distribution station

Directory of Open Access Journals (Sweden)

Krzysiak Zbigniew

2017-12-01

Full Text Available An analysis of the risk of failure in the safety valve – tank with propane-butane (LPG system has been conducted. An uncontrolled outflow of liquid LPG, caused by a failure of the above mentioned system has been considered as a threat. The main research goal of the study is the hazardous analysis of propane-butane gas outflow for the safety valve – LPG tank system. The additional goal is the development of an useful method to fast identify the hazard of a mismatched safety valve. The results of the research analysis have confirmed that safety valves are basic protection of the installation (tank against failures that can lead to loss of life, material damage and further undesired costs of their unreliability. That is why a new, professional computer program has been created that allows for the selection of safety valves or for the verification of a safety valve selection in installations where any technical or technological changes have been made.

Large Deployable Reflector (LDR) Requirements for Space Station Accommodations

Science.gov (United States)

Crowe, D. A.; Clayton, M. J.; Runge, F. C.

1985-01-01

Top level requirements for assembly and integration of the Large Deployable Reflector (LDR) Observatory at the Space Station are examined. Concepts are currently under study for LDR which will provide a sequel to the Infrared Astronomy Satellite and the Space Infrared Telescope Facility. LDR will provide a spectacular capability over a very broad spectral range. The Space Station will provide an essential facility for the initial assembly and check out of LDR, as well as a necessary base for refurbishment, repair and modification. By providing a manned platform, the Space Station will remove the time constraint on assembly associated with use of the Shuttle alone. Personnel safety during necessary EVA is enhanced by the presence of the manned facility.

Large Deployable Reflector (LDR) requirements for space station accommodations

Science.gov (United States)

Crowe, D. A.; Clayton, M. J.; Runge, F. C.

1985-04-01

Top level requirements for assembly and integration of the Large Deployable Reflector (LDR) Observatory at the Space Station are examined. Concepts are currently under study for LDR which will provide a sequel to the Infrared Astronomy Satellite and the Space Infrared Telescope Facility. LDR will provide a spectacular capability over a very broad spectral range. The Space Station will provide an essential facility for the initial assembly and check out of LDR, as well as a necessary base for refurbishment, repair and modification. By providing a manned platform, the Space Station will remove the time constraint on assembly associated with use of the Shuttle alone. Personnel safety during necessary EVA is enhanced by the presence of the manned facility.

Safety of nuclear installations

International Nuclear Information System (INIS)

1991-01-01

In accordance with the Nuclear Energy Act, a Licence may only be issued if the precautions required by the state of the art have been taken to prevent damage resulting from the construction and operation of the installation. The maximum admissible body doses in the area around the installation which must be observed in planning constructional and other technical protective measures to counter accidents in or at a nuclear power station (accident planning values, are established). According to the Radiological Protection Ordinance the Licensing Authority can consider these precautions to have been taken if, in designing the installation against accidents, the applicant has assumed the accidents which, according to the Safety Criteria and Guidelines for Nuclear Power Stations published in the Federal Register by the Federal Minister of the Interior after hearing the competent senior state authorities, must determine the design of a nuclear power station. On the basis of previous experience from safety analysis, assessment and operation of nuclear power stations, the accident guidelines published here define which accidents are determinative for the safety-related design of PWR power stations and what verification -particularly with regard to compliance with the accident planning values of the Radiological Protection Ordinance -must be provided by the applicant. (author)

Integrated Plant Safety Assessment: Systematic Evaluation Program. Yankee Nuclear Power Station, Yankee Atomic Electric Company, Docket No. 50-29. Final report

International Nuclear Information System (INIS)

1983-06-01

The Systematic Evaluation program was initiated in February 1977 by the US Nuclear Regulatory Commission to review the designs of older operating nuclear reactor plants to confirm and document their safety. The review provides: (1) an assessment of how these plants compare with current licensing safety requirements relating to selected issues, (2) a basis for deciding on how these differences should be resolved in an integrated plant review, and (3) a documented evaluation of plant safety. This report documents the review of Yankee Nuclear Power Station, operated by Yankee Atomic Electric Company. The Yankee plant is one of 10 plants reviewed under Phase II of this program. This report indicates how 137 topics selected for review under Phase I of the program were addressed. Equipment and procedural changes have been identified as a result of the review

Integrated Plant Safety Assessment, Systematic Evaluation Program. Yankee Nuclear Power Station, Yankee Atomic Electric Company, Docket No. 50-29. Draft report

International Nuclear Information System (INIS)

1983-02-01

The Systematic Evaluation Program was initiated in February 1977 by the US Nuclear Regulatory Commission to review the designs of older operating nuclear reactor plants to confirm and document their safety. The review provides (1) an assessment of how these plants compare with current licensing safety requirements relating to selected issues, (2) a basis for deciding on how these differences should be resolved in an integrated plant review, and (3) a documented evaluation of plant safety. This report documents the review of Yankee Nuclear Power Station, operated by Yankee Atomic Electric Company. The Yankee plant is one of 10 plants reviewed under Phase II of this program. This report indicates how 137 topics selected for review under Phase I of the program were addressed. Equipment and procedural changes have been identified as a result of the review

Safety evaluation report related to the operation of Clinton Power Station, Unit No. 1 (Docket No. 50-461)

International Nuclear Information System (INIS)

1987-03-01

Supplement No. 8 to the Safety Evaluation Report on the application filed by Illinois Power Company, Soyland Power Cooperative, Inc., and Western Illinois Power Cooperative, Inc., as applicants and owners, for a license to operate the Clinton Power Station, Unit No. 1, has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is located in Harp Township, DeWitt County, Illinois. This supplement reports the status of items that have been resolved by the staff since Supplement No. 7 was issued

Microwave radiation safety assessment around mobile telephone base station (MTBS) in Peninsular Malaysia

International Nuclear Information System (INIS)

Mohd Yusof Mohd Ali; Rozaimah Abd Rahim; Roha Tukimin; Mohd Anuar Abd Majid; Mohamad Amirul Nizam Mohamad Thari; Ahmad Fazli Ahmad Sanusi; Roslan Md Dan; Sahirudden Mohd Nor

2006-01-01

Mobile telephone is one of the fastest popular consumer product introduced in the market. Since more people are using mobile telephone, the number of mobile telephone base station (MTBS) in Malaysia had also increased in order to provide a better coverage services to consumer. The antennas that are required for the mobile (or cellular) telephone network are located at MTBS. This antenna emits radio frequency (RF) and microwave (MW) radiation. Due to the concerns that has been raised by the people that are living or working nearby to MTBS about the possibility of adverse health effects that might occur due to the exposure of this radiation, a project of microwave radiation safety assessment around MTBS by MINT was carried out (September 2003 - January 2006). It was involved with 128 MTBS from three biggest service providers in Malaysia. This assessment is required to establish a baseline data in term of pattern and trend of the radiation emission from the facilities as well as to develop a public confident. In this paper, it will describe the fact that radiation is critical to the MTBS system and without the radiation, the MTBS system is functionless. It will also highlight the result of the assessment's work that has been carried out by MINT around MTBS mounted on the rooftops and towers. The average reading varies between the detection limit of the instrument 2 ( 2 (7.204 V/m). The highest average reading corresponds to about 2.0% of the Suruhanjaya Komunikasi dan Multimedia Malaysia (MCMC) exposure limit for public. The finding of this measurement confirms that the presence of RF and MW radiation in public accessible area around the base station was very low and comparable to the radiation levels in other places away from MTBS. There is also no evidence, from any laboratory or epidemiology studies that the exposure to RF energy levels recommended limits has any health significance for humans. (Author)

NASA space station automation: AI-based technology review. Executive summary

Science.gov (United States)

Firschein, O.; Georgeff, M. P.; Park, W.; Cheeseman, P. C.; Goldberg, J.; Neumann, P.; Kautz, W. H.; Levitt, K. N.; Rom, R. J.; Poggio, A. A.

1985-01-01

Research and Development projects in automation technology for the Space Station are described. Artificial Intelligence (AI) based technologies are planned to enhance crew safety through reduced need for EVA, increase crew productivity through the reduction of routine operations, increase space station autonomy, and augment space station capability through the use of teleoperation and robotics.

AECB staff annual assessment of the Point Lepreau Nuclear Generating Station for the year 1994

International Nuclear Information System (INIS)

1995-06-01

This report is the Atomic Energy Control Board assessment of safety at Point Lepreau Generating Station for 1994. AECB on-site Project Officers, and Ottawa based specialists monitored the station throughout the year. The station operated safely during 1994. Performance of the special safety systems was very good, and NB Power made important progress improving the safety management of these systems. NB Power staff failed to comply with the conditions of our Operating Licence seven times during 1994. This is an improvement from last year, but the number of compliance problems is still too high. NB Power introduced safety culture training for its staff last year. AECB is optimistic that this will help to improve compliance. The station fully met the radiation safety requirements during the year. Releases of radioactive material to the environment are well below the limits set. Radiation exposure to workers are being well controlled, and AECB believe that NB Power is maintaining these exposures 'As Low As Reasonably Achievable' for this reactor type

Complementary safety evaluation of the Phenix power station (INB n 71) in the light of the Fukushima power station accident

International Nuclear Information System (INIS)

2011-01-01

This report proposes a complementary safety evaluation of the Phenix power station, one of the French basic nuclear installations (BNI, in French INB) in the light of the Fukushima accident. This evaluation takes the following risks into account: risks of flooding, earthquake, loss of power supply and loss of cooling, in addition to operational management of accident situations. It presents some characteristics of the Phenix installation (location, operator, industrial environment, installation characteristics), identifies the risks of cliff effect and the main structures and equipment, evaluates the seismic risk (installation sizing, installation conformity, margin evaluation), evaluates the flooding risk (installation sizing, installation conformity, margin evaluation), briefly examines other extreme natural phenomena (extreme meteorological conditions related to flooding, earthquake or flooding with a higher level than that for which the installation is designed). It analyzes the risk of a loss of power supply and of cooling (loss of external and internal electric sources, loss of the ultimate cooling system). It analyzes the management of severe accidents: crisis management organization, available intervention means, robustness of available means. It discusses the conditions of the use of subcontractors

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...

Safety Evaluation Report related to the operation of Braidwood Station, Units 1 and 2 (Docket Nos. 50-456 and 50-457). Supplement No. 1

International Nuclear Information System (INIS)

1986-09-01

In November 1983 the staff of the Nuclear Regulatory Commission issued its Safety Evaluation Report (NUREG-1002) regarding the application filed by the Commonwealth Edison Company, as applicant and owner, for a license to operate Braidwood Station, Units 1 and 2 (Docket Nos. STN 50-456 and STN 50-457). The facility is located in Reed Township, Will County, Illinois. This first supplement to NUREG-1002 reports the status of certain items that remained unresolved at the time the Safety Evaluation Report was published

Representations and types of organizational errors in nuclear power station

International Nuclear Information System (INIS)

Li Yongjuan; Wang Erping; Li Feng; Yu Guangtao

2003-01-01

The aim of this study is to explore the organizational errors that affect the safety of the nuclear power station,with the methods of interview and events re-analysis. Coding method has been used to change the qualitative data to quantitative data. After cluster analysis, five organizational factors in nuclear power field have been obtained, including technical management factors, non-technical management factors, information interface, organizational planning, and lack of holistic thinking. The study also finds that compared to the technical management factors, non-technical management factors have greater affects on the safety of the system. So the non-technical training is essential to insure the safety of the system

Shutdown Safety in NEK

International Nuclear Information System (INIS)

Gluhak, Mario; Senegovic, Marko

2014-01-01

Industry performance analysis since 2004 has revealed that 23% of the events reported to WANO occurred during outage periods. Given the fact that a plant is in the outage only 5 percent of the time, this emphasizes the importance of shutdown safety and measures station staffs undertake to maintain effective barriers to safety margins during the outage. Back in 1990s, the industry adopted guidance to meet safety requirements by focusing on safety functions. Both WANO and INPO released various documents, reports and guidelines to help accomplish those requirements. However, in the last decade inadequate 'defence in depth' has led to several events affecting shutdown safety and challenging one of the most important nuclear safety principles: 'The special characteristics of nuclear technology are taken into account in all decisions and actions. Reactivity control, continuity of core cooling, and integrity of fission product barriers are valued as essential, distinguishing attributes of nuclear station work environment'. NEK has recognized the importance of 'defence in depth'Industry performance analysis since 2004 has revealed that 23% of the events reported to WANO occurred during outage periods. Given the fact that a plant is in the outage only 5 percent of the time, this emphasizes the importance of shutdown safety and measures station staffs undertake to maintain effective barriers to safety margins during the outage. Back in 1990s, the industry adopted guidance to meet safety requirements by focusing on safety functions. Both WANO and INPO released various documents, reports and guidelines to help accomplish those requirements. However, in the last decade inadequate 'defence in depth' has led to several events affecting shutdown safety and challenging one of the most important nuclear safety principles: 'The special characteristics of nuclear technology are taken into account in all decisions and actions. Reactivity

Impact studies and nuclear power stations

International Nuclear Information System (INIS)

Chambolle, Thierry

1981-01-01

Impact studies form an essential part of environmental protection. The impact study discipline has enabled the EDF to have a better understanding of the effects of nuclear power stations on the environment and to remedy them at the project design stage [fr

Lunar base thermoelectric power station study

Science.gov (United States)

Determan, William; Frye, Patrick; Mondt, Jack; Fleurial, Jean-Pierre; Johnson, Ken; Stapfer, G.; Brooks, Michael D.; Heshmatpour, Ben

2006-01-01

Under NASA's Project Prometheus, the Nuclear Systems Program, the Jet Propulsion Laboratory, Pratt & Whitney Rocketdyne, and Teledyne Energy Systems have teamed with a number of universities, under the Segmented Thermoelectric Multicouple Converter (STMC) program, to develop the next generation of advanced thermoelectric converters for space reactor power systems. Work on the STMC converter assembly has progressed to the point where the lower temperature stage of the segmented multicouple converter assembly is ready for laboratory testing and the upper stage materials have been identified and their properties are being characterized. One aspect of the program involves mission application studies to help define the potential benefits from the use of these STMC technologies for designated NASA missions such as the lunar base power station where kilowatts of power are required to maintain a permanent manned presence on the surface of the moon. A modular 50 kWe thermoelectric power station concept was developed to address a specific set of requirements developed for this mission. Previous lunar lander concepts had proposed the use of lunar regolith as in-situ radiation shielding material for a reactor power station with a one kilometer exclusion zone radius to minimize astronaut radiation dose rate levels. In the present concept, we will examine the benefits and requirements for a hermetically-sealed reactor thermoelectric power station module suspended within a man-made lunar surface cavity. The concept appears to maximize the shielding capabilities of the lunar regolith while minimizing its handling requirements. Both thermal and nuclear radiation levels from operation of the station, at its 100-m exclusion zone radius, were evaluated and found to be acceptable. Site preparation activities are reviewed and well as transport issues for this concept. The goal of the study was to review the entire life cycle of the unit to assess its technical problems and technology

Safety Evaluation Report related to the operation of Seabrook Station, Units 1 and 2 (Docket Nos. 50-443 and 50-444)

International Nuclear Information System (INIS)

1990-03-01

This report is Supplement No. 9 to the Safety Evaluation Report (SER) (NUREG-0896, March 1983) for the application filed by the Public Service Company of New Hampshire, et al., for licenses to operate Seabrook Station, Units 1 and 2. It has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission and provides recent information on open items identified in the SER. The facility is located in Seabrook, New Hampshire. Subject to favorable resolution of the items discussed in this report, the staff concludes that the facility can be operated by the applicant without endangering the health and safety of the public. 70 refs., 1 fig., 1 tab

Safety evaluation report related to the operation of Seabrook Station, Units 1 and 2 (Docket Nos. 50-443 and 50-444)

International Nuclear Information System (INIS)

1986-05-01

This report is Supplement 4 to the Safety Evaluation Report (SER, NUREG-0896, March 1983) for the application filed by the Public Service Company of New Hampshire, et al., for licenses to operate Seabrook Station, Units 1 and 2 (Docket Nos. STN 50-443 and STN 50-444). It has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission and provides recent information on open items identified in the SER. The facility is located in Seabrook, New Hamphsire. Subject to favorable resolution of the items discussed in this report, the staff concludes that the facility can be operated by the applicant without endangering the health and safety of the public

Safety evaluation report related to the operation of Seabrook Station, Units 1 and 2 (Docket Nos. 50-443 and 50-444)

International Nuclear Information System (INIS)

1987-10-01

This report is Supplement No. 7 to the Safety Evaluation Report (NUREG-0896, March 1983) for the application filed by the Public Service Company of New Hampshire, et al. for licenses to operate Seabrook Station, Units 1 and 2 (Docket Nos. STN 50-443 and STN 50-444). It has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission and provides recent information on open items identified in the SER. The facility is located in Seabrook, New Hampshire. Subject to favorable resolution of the items discussed in this report, the staff concludes that the facility can be operated by the applicant without endangering the health and safety of the public

Safety-evaluation report related to the operation of Seabrook Station, Units 1 and 2. Docket Nos. 50-443 and 50-444

International Nuclear Information System (INIS)

1983-06-01

This report is Supplement 2 to the Safety Evaluation Report (NUREG-0896, March 1983) for the application filed by the Public Service Company of New Hampshire, et al., for licenses to operate Seabrook Station, Units 1 and 2 (Docket Nos. STN 50-443 and STN 50-444). It has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission and provides recent information on open items identified in the SER. The facility is located in Seabrook, New Hampshire. Subject to favorable resolution of the items discussed in this report, the staff concludes that the facility can be operated by the applicant without endangering the health and safety of the public

Hinkley Point 'C' power station public inquiry: statement of case

International Nuclear Information System (INIS)

1988-08-01

This Statement of Case contains full particulars of the case which the Central Electricity Generating Board (CEGB) proposes to put forward at the Hinkley Point ''C'' Inquiry. It relates to the planning application made by the CEGB for the construction of a 1200 MW Pressurized Water Reactor (PWR) power station at Hinkley Point in the United Kingdom, adjacent to an existing nuclear power station. The inquiry will consider economic, safety, environmental and planning matters relevant to the application and the implications for agriculture and local amenities of re-aligning two power transmission lines. The Statement contains submissions on the following matters: Topic 1 The Requirement for the Station; Topic 2 Safety and Design, including Radioactive Discharges; Topic 3 The On-Site Management of Radioactive Waste and Decommissioning of the Station; Topic 4 Emergency Arrangements; Topic 5 Local and Environmental Issues. (author)

Study of applicable methods on safety verification of disposal facilities and waste packages

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-08-15

Three subjects about safety verification on the disposal of low level radioactive waste were investigated in FY. 2012. For radioactive waste disposal facilities, specs and construction techniques of covering with soil to prevent possible destruction caused by natural events (e.g. earthquake) were studied to consider verification methods for those specs. For waste packages subject to near surface pit disposal, settings of scaling factor and average radioactivity concentration (hereafter referred to as ''SF'') on container-filled and solidified waste packages generated from Kashiwazaki Kariwa Nuclear Power Station Unit 1-5, setting of cesium residual ratio of molten solidified waste generated from Tokai and Tokai No.2 Power Stations, etc. were studied. Those results were finalized in consideration of the opinion from advisory panel, and publicly opened as JNES-EV reports. In FY 2012, five JNES reports were published and these have been used as standards of safety verification on waste packages. The verification method of radioactive wastes subject to near-surface trench disposal and intermediate depth disposal were also studied. For radioactive wastes which will be returned from overseas, determination methods of radioactive concentration, heat rate and hydrogen generation rate of CSD-C were established. Determination methods of radioactive concentration and heat rate of CSD-B were also established. These results will be referred to verification manuals. (author)

Performance assessment of Point Lepreau Generating Station

International Nuclear Information System (INIS)

Alikhan, S.

1991-01-01

The Point Lepreau Generating Station, a 680 MWe CANDU unit, is located about 40 km southwest of the city of Saint John, New Brunswick, Canada. It was declared in-service on 1 February, 1983 and, since then, has demonstrated an average cross capacity factor of over 93% up to the end of 1990. This paper compared the performance of the station with other sister CANDU units and the Light Water Reactors world-wide using the following ten performance indicators, as applicable: - gross capacity factor; - fuel burn-up; - heavy water upkeep; - unplanned reactor trips while critical; - forced outage rate; - fuel handling performance; - derived emission of radioactive effluents to environment; - personnel radiation dose; - industrial safety; - low-level solid radioactive wastes. The paper examines various areas of station activities including management and organization, operations and maintenance, technical support, fuel handling and health physics in order to highlight some of the 'good practices' which are believed to have made a significant contribution towards achieving the demonstrated performance of Point Lepreau G.S. In addition, several areas of potential improvement are discussed in order to maintain and enhance, where practicable, the safety, reliability and economic performance of the station. In this context, a careful review of the operating experiences, both in-house and at other stations, and a judicious application of lessons learned plays a significant role. (author)

Performance assessment of Point Lepreau Generating Station

Energy Technology Data Exchange (ETDEWEB)

Alikhan, S [Point Lepreau Generating Station, Lepreau, NB (Canada)

1991-04-01

The Point Lepreau Generating Station, a 680 MWe CANDU unit, is located about 40 km southwest of the city of Saint John, New Brunswick, Canada. It was declared in-service on 1 February, 1983 and, since then, has demonstrated an average cross capacity factor of over 93% up to the end of 1990. This paper compared the performance of the station with other sister CANDU units and the Light Water Reactors world-wide using the following ten performance indicators, as applicable: - gross capacity factor; - fuel burn-up; - heavy water upkeep; - unplanned reactor trips while critical; - forced outage rate; - fuel handling performance; - derived emission of radioactive effluents to environment; - personnel radiation dose; - industrial safety; - low-level solid radioactive wastes. The paper examines various areas of station activities including management and organization, operations and maintenance, technical support, fuel handling and health physics in order to highlight some of the 'good practices' which are believed to have made a significant contribution towards achieving the demonstrated performance of Point Lepreau G.S. In addition, several areas of potential improvement are discussed in order to maintain and enhance, where practicable, the safety, reliability and economic performance of the station. In this context, a careful review of the operating experiences, both in-house and at other stations, and a judicious application of lessons learned plays a significant role. (author)

Space station accommodations for lunar base elements: A study

Science.gov (United States)

Weidman, Deene J.; Cirillo, William; Llewellyn, Charles; Kaszubowski, Martin; Kienlen, E. Michael, Jr.

1987-01-01

The results of a study conducted at NASA-LaRC to assess the impact on the space station of accommodating a Manned Lunar Base are documented. Included in the study are assembly activities for all infrastructure components, resupply and operations support for lunar base elements, crew activity requirements, the effect of lunar activities on Cape Kennedy operations, and the effect on space station science missions. Technology needs to prepare for such missions are also defined. Results of the study indicate that the space station can support the manned lunar base missions with the addition of a Fuel Depot Facility and a heavy lift launch vehicle to support the large launch requirements.

Decommissioning of nuclear power stations

International Nuclear Information System (INIS)

Gregory, A.R.

1988-01-01

In the United Kingdom the Electricity Boards, the United Kingdom Atomic Energy Authority (UKAEA) and BNFL cooperate on all matters relating to the decommissioning of nuclear plant. The Central Electricity Generating Board's (CEGB) policy endorses the continuing need for nuclear power, the principle of reusing existing sites where possible and the building up of sufficient funds during the operating life of a nuclear power station to meet the cost of its complete clearance in the future. The safety of the plant is the responsibility of the licensee even in the decommissioning phase. The CEGB has carried out decommissioning studies on Magnox stations in general and Bradwell and Berkeley in particular. It has also been involved in the UKAEA Windscale AGR decommissioning programme. The options as to which stage to decommission to are considered. Methods, costs and waste management are also considered. (U.K.)

Summary of Great East Japan Earthquake response at Onagawa Nuclear Power Station and further safety improvement measures

International Nuclear Information System (INIS)

Sato, Toru

2013-01-01

A large earthquake occurred on March 11, 2011 and tsunami was generated following it. The East Japan suffered serious damage by the earthquake and tsunami. This is called the Great East Japan Earthquake. Onagawa Nuclear Power Station (NPS) is located closest to the epicenter of Great East Japan Earthquake. We experienced intense shake by the earthquake and some flooding from the tsunami, however, we have succeeded safely cold shutdown of the reactors. In this paper, we introduce summary of Great East Japan Earthquake response a Onagawa NPS and safety improvement measures which are based on both experience of Onagawa NPS and lesson from Fukushima Daiichi NPS accident. (author)

IAEA Leads Operational Safety Mission to Rajasthan Atomic Power Station 3 and 4

International Nuclear Information System (INIS)

2012-01-01

Full text: An international team of nuclear safety experts led by the International Atomic Energy Agency (IAEA) today completed a review of safety practices at Units 3 and 4 of the Rajasthan Atomic Power Station in Rawatbhata. The team noted a series of good practices and made recommendations and suggestions to reinforce safety practices. The IAEA assembled the Operational Safety Review Team (OSART) at the request of the Government of India. Led by the IAEA's Division of Nuclear Installation Safety, the team performed an in-depth operational safety review from 29 October to 14 November 2012. The team was comprised of experts from Canada, Belgium, Finland, Germany, Romania, Slovakia, Slovenia, Sweden and the IAEA. The team conducted an in-depth review of the aspects essential to the safe operation of the Power Plant. The conclusions of the review are based on the IAEA's Safety Standards and good international practices. The review covered the areas of Management, Organization and Administration; Training; Operations; Maintenance; Technical Support; Operating Experience; Radiation Protection; Chemistry; Emergency Planning and Preparedness; and Severe Accident Management. The OSART team identified a number of good practices of the plant. These will be shared in due course by the IAEA with the global nuclear industry for consideration. Examples include the following: - The Power Plant's safety culture cultivates a constructive work environment and a sense of accountability among the Power Plant personnel, and gives its staff the opportunity to expand skills and training; - The Power Plant's Public Awareness Programme provides educational opportunities to the local community about nuclear and radiation safety; - The Power Plant has a Management of Training and Authorization system for effective management of training activities; and - The Power Plant uses testing facilities and mockups to improve the quality of maintenance work and to reduce radiation doses. The OSART

Nuclear power and safety

International Nuclear Information System (INIS)

Saunders, P.; Tasker, A.

1991-01-01

Nuclear power currently provides about a fifth of both Britain's and the world's electricity. It is the largest single source of electricity in Western Europe; in France three quarters of electricity is generated by nuclear power stations. This booklet is about the safety of those plants. It approaches the subject by outlining the basic principles and approaches behind nuclear safety, describing the protective barriers and safety systems that are designed to prevent the escape of radioactive material, and summarising the regulations that govern the construction and operation of nuclear power stations. The aim is to provide a general understanding of the subject by explaining the general principles of the Advanced Gas Cooled Reactor and setting out the UKAEA strategy for nuclear safety, the objective being always to minimize risk. (author)

Sizewell 'B' power station public inquiry: CEGB proof of evidence

International Nuclear Information System (INIS)

Harrison, J.R.

1982-11-01

An overview is given of the Safety Case for the Sizewell B nuclear power station as presented in the Pre-Construction Safety Report. Information which has been made available since its publication is included. Safety is considered for normal operation of the reactor and under fault conditions. Faults considered are those due to external hazards such as earthquakes, extreme weather conditions and aircraft crashes, internal hazards such as fire and missiles and reactor faults. Reactor fault studies described include transient analyses of pressurised faults and of loss-of-coolant accidents and the evaluation of the radiological consequences of design basis faults. (U.K.)

Safety Evaluation Report related to the operation of Shoreham Nuclear Power Station, Unit No. 1 (Docket No. 50-322). Supplement No. 8

International Nuclear Information System (INIS)

1984-12-01

Supplement 8 (SSER 8) to the Safety Evaluation Report on Long Island Lighting Company's application for a license to operate the Shoreham Nuclear Power Station, Unit 1, located in Suffolk County, New York, has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. This supplement addresses several items that have been reviewed by the staff since the previous supplement was issued

Safety evaluation report related to the operation of Shoreham Nuclear Power Station, Unit No. 1 (Docket No. 50-322). Supplement No. 7

International Nuclear Information System (INIS)

1984-09-01

Supplement 7 (SSER 7) to the Safety Evaluation Report on Long Island Lighting Company's application for a license to operate the Shoreham Nuclear Power Station, Unit 1, located in Suffolk County, New York, has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. This supplement addresses several items that have been reviewed by the staff since the previous supplement was issued

Safety evaluation report related to the operation of Shoreham Nuclear Power Station, Unit No. 1 (Docket No. 50-322). Supplement No. 6

International Nuclear Information System (INIS)

1984-07-01

Supplement 6 (SSER 6) to the Safety Evaluation Report on Long Island Lighting Company's application for a license to operate the Shoreham Nuclear Power Station, Unit 1, located in Suffolk County, New York, has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. This supplement addresses several items that have been reviewed by the staff since the previous supplement was issued

Safety Evaluation Report related to the operation of Shoreham Nuclear Power Station, Unit No. 1 (Docket No. 50-322). Supplement No. 9

International Nuclear Information System (INIS)

1985-12-01

Supplement 9 (SSER 9) to the Safety Evaluation Report on Long Island Lighting Company's application for a license to operate the Shoreham Nuclear Power Station, Unit 1, located in Suffolk County, New York, has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. This supplement addresses several items that have been reviewed by the staff since the previous supplement was issued

Extended Station Blackout Analyses of an APR1400 with MARS-KS

International Nuclear Information System (INIS)

Kim, WoongBae; Jang, HyungWook; Oh, Seungjong; Lee, Sangyong

2016-01-01

The Fukushima Dai-ichi nuclear power plant accident shows that natural disasters such as earthquakes and the subsequent tsunamis can cause station blackout for several days. The electricity required for essential systems during a station blackout is provided from the emergency backup batteries installed at the nuclear power plant. In South Korea, in the event of an extended station blackout, the life of these emergency backup batteries has recently been extended from 8 hours to 24 hours at Shin-Kori 5, 6 and APR1400 for design certification. For a battery life of 24 hours, available safety means system, equipment and procedures are studied and analyzed in their ability to cope with an extended station blackout. A sensitivity study of reactor coolant pump seal leakage is performed to verify how different seal leakages could affect the system. For simulating of extended station blackout scenarios, the best estimate MARS-KS was used. In this paper, an APR1400 RELAP5 input deck was developed for station blackout scenario to analyze operation strategy by manually depressurizing the reactor coolant system through the steam generator's secondary side. Additionally, a sensitivity study was performed on reactor coolant pump seal leakage

Extended Station Blackout Analyses of an APR1400 with MARS-KS

Energy Technology Data Exchange (ETDEWEB)

Kim, WoongBae; Jang, HyungWook; Oh, Seungjong; Lee, Sangyong [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

2016-10-15

The Fukushima Dai-ichi nuclear power plant accident shows that natural disasters such as earthquakes and the subsequent tsunamis can cause station blackout for several days. The electricity required for essential systems during a station blackout is provided from the emergency backup batteries installed at the nuclear power plant. In South Korea, in the event of an extended station blackout, the life of these emergency backup batteries has recently been extended from 8 hours to 24 hours at Shin-Kori 5, 6 and APR1400 for design certification. For a battery life of 24 hours, available safety means system, equipment and procedures are studied and analyzed in their ability to cope with an extended station blackout. A sensitivity study of reactor coolant pump seal leakage is performed to verify how different seal leakages could affect the system. For simulating of extended station blackout scenarios, the best estimate MARS-KS was used. In this paper, an APR1400 RELAP5 input deck was developed for station blackout scenario to analyze operation strategy by manually depressurizing the reactor coolant system through the steam generator's secondary side. Additionally, a sensitivity study was performed on reactor coolant pump seal leakage.

Switching from motorcycle taxi to walking: A case study of transit station access in Bangkok, Thailand

Directory of Open Access Journals (Sweden)

Pornraht Pongprasert

2017-12-01

Full Text Available This study aims to find the factors affecting residents near transit stations within 1000 m, who are referred to as transit-oriented development (TOD residents, to reduce motorcycle taxi use and encourage walking to stations. These two modes of commuting are the most popular among over 85% of residents. However, motorcycle taxis are the main pedestrian barriers that hinder easy access and walkability in TODs of Bangkok, because they ride, stop, and provide services on sidewalks. From 2013 to 2015, these problems substantially increased the number of motorcycle taxis that are not willing and able to follow the rules. The increasing number of pedestrian accidents on sidewalks is related to the increase in the number of motorcycle taxis. According to a survey on pedestrian safety with 249 respondents, over 25% of walkers feel unsafe to walk, while 40% of motorcycle-taxi users riding to stations do not walk because they are afraid of accidents. In modal split, the share of walking reduces from 76% for areas < 500 m, to 25% for areas between 500 and 1000 m from transit stations, respectively. Hence, the number of motorcycle taxis in the 500–1000 m range is twice as high compared to that within the 500 m area. If motorcycle taxi users would accept a longer walking distance to station by 36 m or would be willing to walk to the station within 9.15 min, 54% of them may switch to walking to stations. Moreover, based on the estimation results of the logistic regression models, middle-adult aged residents, office employees, residents owning a car, and people living far from stations are less likely to walk. Average income households and commuters during non-peak hours tend to use motorcycle taxis more. On the other hand, residents living far from stations tend to use motorcycle taxis less, because most of the motorcycle taxi services are located near transit stations. Keywords: Transit accessibility, Pedestrian, Walkability

Hydrogen Filling Station

Energy Technology Data Exchange (ETDEWEB)

Boehm, Robert F; Sabacky, Bruce; Anderson II, Everett B; Haberman, David; Al-Hassin, Mowafak; He, Xiaoming; Morriseau, Brian

2010-02-24

future. Project partners also conducted a workshop on hydrogen safety and permitting. This provided an opportunity for the various permitting agencies and end users to gather to share experiences and knowledge. As a result of this workshop, the permitting process for the hydrogen filling station on the Las Vegas Valley Water District’s land was done more efficiently and those who would be responsible for the operation were better educated on the safety and reliability of hydrogen production and storage. The lessons learned in permitting the filling station and conducting this workshop provided a basis for future hydrogen projects in the region. Continuing efforts to increase the working pressure of electrolysis and efficiency have been pursued. Research was also performed on improving the cost, efficiency and durability of Proton Exchange Membrane (PEM) hydrogen technology. Research elements focused upon PEM membranes, electrodes/catalysts, membrane-electrode assemblies, seals, bipolar plates, utilization of renewable power, reliability issues, scale, and advanced conversion topics. Additionally, direct solar-to-hydrogen conversion research to demonstrate stable and efficient photoelectrochemistry (PEC) hydrogen production systems based on a number of optional concepts was performed. Candidate PEC concepts included technical obstacles such as inefficient photocatalysis, inadequate photocurrent due to non-optimal material band gap energies, rapid electron-hole recombination, reduced hole mobility and diminished operational lifetimes of surface materials exposed to electrolytes. Project Objective 1: Design, build, operate hydrogen filling station Project Objective 2: Perform research and development for utilizing solar technologies on the hydrogen filling station and convert two utility vehicles for use by the station operators Project Objective 3: Increase capacity of hydrogen filling station; add additional vehicle; conduct safety workshop; develop a roadmap for

Aseismatic design and safety of nuclear power generation facilities. On aseismatic capability of commercial nuclear power stations

International Nuclear Information System (INIS)

Kato, Muneaki

1995-01-01

In view of the great Hanshin earthquake, the aseismatic safety of the important facilities in nuclear power stations is ensured by the location direct on base rocks, the design with the earthquake force at least three times as large as that in the building standard, and the consideration of the earthquakes due to active faults as design earthquake. The basic policy of the aseismatic design of nuclear power stations is described. The determination of the earthquake motions due to strongest earthquake and utmost limit earthquake for design, the survey of the geological features and ground of the sites and so on are explained. In the aseismatic design of buildings and structures, structural planning, the modeling for the aseismatic analysis of buildings, the analysis of time historical response and so on are carried out. In the aseismatic design of equipment and piping systems, the planning of aseismatic support structures, the aseismatic design and the analysis of time historical response, the spectral modal analysis for other systems such as multiple material point system and so on are described. The tests and researches related to the aseismatic design are reported. (K.I.)

Study of the method of placing technical stations

Directory of Open Access Journals (Sweden)

Sattorov S.B.

2017-12-01

Full Text Available At present, the organization of transportation by rail is associated with the processing of trains and wagons along the route. Accommodation and equipping of technical stations play a major role in the organization of the transportation process. The uninterrupted operation of the technical stations ensures the stability of the railway transport as a whole. The distance between the technical stations determines the duration of operation of the locomotive crew, the efficient use of the locomotive, the required number of locomotive crews and locomotives for carrying out the transportation plan, and also improves the main indicators of the railway. However, often train locomotives do not fulfill their assigned tasks due to the lack of time for the locomotive crew in order to deliver the train to the nearest technical station, as well as the irrational use of the standard time of the locomotive crew. In connection with the expectation of delivery, the locomotive brigade replacing the locomotive, the idle time of the wagons at the intermediate stations increases. All this makes it difficult to control the movement of trains and reduces the speed of moving cars on the railway transport. The purpose of the study is to develop a method for placing technical stations on a railway test site in the course of an electric traction. To achieve this goal, it will be necessary to solve the following tasks: generalization of the main parameters of the location of technical stations; determination of half-races between technical stations in the operation of one locomotive brigade; search for the required number of change of the locomotive brigade; determination of the distance between technical stations. The methods of research are based on the generalization and analysis of existing results in finding the rational location of technical stations. Results: It is revealed that the first step for the rational placement of technical stations is to determine the

Safety Evaluation Report related to the operation of Seabrook Station, Units 1 and 2 (Docket Nos. 50-443 and 50-444)

International Nuclear Information System (INIS)

1989-05-01

This report is Supplement No. 8 to the Safety Evaluation Report (SER) (NUREG-0896, March 1983) for the application filed by the Public Service Company of New Hampshire, et al., for licenses to operate Seabrook Station, Units 1 and 2 (Docket Nos. STN 50-443 and STN 50-444). It has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission and provides recent information on open items identified in the SER. The facility is located in Seabrook, New Hampshire. Subject to favorable resolution of the items discussed in this report, the staff concludes that the facility can be operated by the applicant without endangering the health and safety of the public. 2 figs., 1 tab

Safety evaluation report related to the operation of LaSalle County Station, Units 1 and 2, (Docket Nos. 50-373 and 50-374). Supplement No. 7

International Nuclear Information System (INIS)

1983-12-01

Supplement No. 7 to the Safety Evaluation Report of Commonwealth Edison Company's application for a license to operate its La Salle County Station, Unit 2, located on Brookfield Township, La Salle County, Illinois, has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. This supplement is to update our evaluations on Unit 2 issues identified in the previous Safety Evaluation Report and Supplements that need resolution prior to issuance of the operating license for Unit 2

Examination policy concerning the additional installation of No. 3 and No. 4 reactors in Takahama Nuclear Power Station and No. 3 and No. 4 reactors in Fukushima No. 2 Nuclear Power Station

International Nuclear Information System (INIS)

1980-01-01

The Nuclear Safety Commission decided the annual examination policy on the modification of reactor installation in Takahama Nuclear Power Station to construct No. 3 and No. 4 reactors inquired under date of November 26, 1979, by the Minister of International Trade and Industry, so that the examination results of the accident in Three Mile Island nuclear power station are reflected to the examination for the purpose of improving reactor safety. The examination results of the accident in Three Mile Island power station are being investigated by the Committee on Examination of Reactor Safety, based on the policy shown in ''On the second report of the special committee examining the accident in a nuclear power station in the U.S.'' determined by the Nuclear Safety Commission under date of September 13, 1979. Though the Committee will further clarify the past guideline about the items concerning the criteria, design and operation management, the Committee decided the tentative policy to reflect it to safety examination. Further, a table is attached, in which 52 items to be reflected to the security measures are classified from the viewpoint of necessity to reflect them to the final examination. This table includes 13 items of criteria and examination, 7 items related to design, 10 items related to operation management, 10 antidisaster items, and 12 items related to safety research. (Wakatsuki, Y.)

Safety Evaluation Report related to the operation of Millstone Nuclear Power Station, Unit No. 3 (Docket No. 50-423). Supplement No. 3

International Nuclear Information System (INIS)

1985-11-01

This report supplements the Safety Evaluation Report (NUREG-1031) issued in July 1984, Supplement 1 issued in March 1985, and Supplement 2 issued in September 1985 by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission with respect to the application filed by Northeast Nuclear Energy Company (applicant and agent for the owners) for a license to operate Millstone Nuclear Power Station, Unit No. 3 (Docket 50-423). The facility is located in the Town of Waterford, New London County, Connecticut, on the north shore of Long Island Sound. This supplement provides more recent information regarding resolution or updating of some of the open and confirmatory items and license conditions identified in the Safety Evaluation Report

Safety Evaluation Report related to the operation of Wolf Creek Generating Station, Unit No. 1 (Docket No. STN 50-482). Supplement No. 6

International Nuclear Information System (INIS)

1985-06-01

This report supplements the Safety Evaluation Report (SER) for the application filed by the Kansas Gas and Electric Company, as applicant and agent for the owners, for a license to operate the Wolf Creek Generating Station, Unit 1 (Docket No. STN 50-482). The facility is located in Coffey County, Kansas. This supplement provides recent information regarding resolution of the license conditions identified in the SER. Because of the favorable resolution of the items discussed in this report, the staff concludes that the facility can be operated by the licensee at power levels greater than 5% without endangering the health and safety of the public

Integrated-plant-safety assessment Systematic Evaluation Program. Dresden Nuclear Power Station, Unit 2, Commonwealth Edison Company, Docket No. 50-237

International Nuclear Information System (INIS)

1982-10-01

The Systematic Evaluation Program was initiated in February 1977 by the US Nuclear Regulatory Commission to review the designs of older operating nuclear reactor plants to reconfirm and document their safety. The review provides: (1) an assessment of how these plants compare with current licensing safety requirements relating to selected issues; (2) a basis for deciding on how these differences should be resolved in an integrated plant review; and (3) a documented evaluation of plant safety. This report documents the review of Dresden Nuclear Generating Station, Unit 2 owned and operated by the Commonwealth Edison Company and located in Grundy County, Illinois. Dresden Unit 2 is one of ten plants reviewed under Phase II of this program, which indicates how 137 topics selected for review under Phase I of the program were addressed. Equipment and procedural changes have been identified as a result of the review. It is expected that this report will be one of the bases in considering the issuance of a full-term operating license in place of the existing provisional operating license

OPG waterways public safety program

Energy Technology Data Exchange (ETDEWEB)

Bennett, T [Ontario Power Generation Inc., Niagara Falls, ON (Canada)

2009-07-01

Ontario Power Generation (OPG) has 64 hydroelectric generating stations, 241 dams, and 109 dams in Ontario's registry with the International Commission on Large Dams (ICOLD). In 1986, it launched a formal dam safety program. This presentation addressed the importance of public safety around dams. The safety measures are timely because of increasing public interaction around dams; the public's unawareness of hazards; public interest in extreme sports; easier access by recreational vehicles; the perceived right of public to access sites; and the remote operation of hydroelectric stations. The presentation outlined the OPG managed system approach, with particular reference to governance; principles; standards and procedures; and aspects of implementation. Specific guidelines and governing documents for public safety around dams were identified, including guidelines for public safety of waterways; booms and buoys; audible warning devices and lights; public safety signage; fencing and barricades; and risk assessment for public safety around waterways. The presentation concluded with a discussion of audits and management reviews to determine if safety objectives and targets have been met. figs.

Socio-economic impacts of nuclear generating stations: Arkansas Nuclear One Station case study

International Nuclear Information System (INIS)

Pijawka, K.D.

1982-07-01

This report documents a case study of the socio-economic impacts of the construction and operation of the Arkansas Nuclear One nuclear power station. It is part of a major post-licensing study of the socio-economic impacts at twelve nuclear power stations. The case study covers the period beginning with the announcement of plans to construct the reactor and ending in the period 1980 to 1981. The case study deals with changes in the economy, population, settlement patterns and housing, local government and public services, social structure, and public response in the study area during the construction/operation of the reactor. A regional modeling approach is used to trace the impact of construction/operation on the local economy, labor market, and housing market. Emphasis in the study is on the attribution of socio-economic impacts to the reactor or other causal factors. As part of the study of local public response to the construction/operation of the reactor, the effects of the Three Mile Island accident are examined

Key techniques and risk management for the application of the Pile-Beam-Arch (PBA) excavation method: a case study of the Zhongjie subway station.

Science.gov (United States)

Guan, Yong-ping; Zhao, Wen; Li, Shen-gang; Zhang, Guo-bin

2014-01-01

The design and construction of shallow-buried tunnels in densely populated urban areas involve many challenges. The ground movements induced by tunneling effects pose potential risks to infrastructure such as surface buildings, pipelines, and roads. In this paper, a case study of the Zhongjie subway station located in Shenyang, China, is examined to investigate the key construction techniques and the influence of the Pile-Beam-Arch (PBA) excavation method on the surrounding environment. This case study discusses the primary risk factors affecting the environmental safety and summarizes the corresponding risk mitigation measures and key techniques for subway station construction using the PBA excavation method in a densely populated urban area.

Safety Evaluation Report related to the operation of Nine Mile Point Nuclear Station, Unit No. 2 (Docket No. 50-410). Supplement No. 5

International Nuclear Information System (INIS)

1986-10-01

This report supplements the Safety Evaluation Report (NUREG-1047, February 1985) for the application filed by Niagara Mohawk Power Corporation, as applicant and co-owner, for a license to operate Nine Mile Point Nuclear Station, Unit 2 (Docket No. 50-410). It has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is located near Oswego, New York. Supplement 1 to the Safety Evaluation Report was published in June 1985 and contained the report from the Advisory Committee on Reactor Safeguards as well as the resolution of a number of outstanding issues from the Safety Evaluation Report. Supplement 2 was published in November 1985 and contained the resolution of a number of outstanding and confirmatory issues. Supplement 3 was published in July 1986 and contained the resolution of a number of outstanding and confirmatory items, one new confirmatory item, the evaluation of the Engineering Assurance Program, and the evaluation of a number of exemption requests. Supplement 4 was published in September 1986 and contained the resolution of a number of outstanding and confirmatory issues and the evaluation of a number of exemption requests. This report contains the resolution of a number of issues that have been resolved since Supplement 4 was issued. It also contains the evaluation of a number of requests for exemption from the applicant. This report also supports the issuance of the low-power license for Nine Mile Point Nuclear Station, Unit 2

Risk investigation for safety assessment of nuclear power stations

International Nuclear Information System (INIS)

Heuser, F.W.; Hoertner, H.; Kersting, E.

1987-01-01

The most important results of the technical system investigations for Phase B of the German Risk Study and the insight gained from them are discussed in connection with the risk analyses in judging the safety-related design. The results of the German Precursor Study for the Biblis A and B plants are given. There is also a comparison of the German with more recent American risk analyses. (DG) [de

47 CFR 80.305 - Watch requirements of the Communications Act and the Safety Convention.

Science.gov (United States)

2010-10-01

... and the Safety Convention. 80.305 Section 80.305 Telecommunication FEDERAL COMMUNICATIONS COMMISSION (CONTINUED) SAFETY AND SPECIAL RADIO SERVICES STATIONS IN THE MARITIME SERVICES Safety Watch Requirements and Procedures Ship Station Safety Watches § 80.305 Watch requirements of the Communications Act and the Safety...

UK experience of safety requirements for thermal reactor stations

International Nuclear Information System (INIS)

Matthews, R.R.; Dale, G.C.; Tweedy, J.N.

1977-01-01

The paper summarises the development of safety requirements since the first of the Generating Boards' Magnox reactors commenced operation in 1962 and includes A.G.R. safety together with the preparation of S.G.H.W.R. design safety criteria. It outlines the basic principles originally adopted and shows how safety assessment is a continuing process throughout the life of a reactor. Some description is given of the continuous effort over the years to obtain increased safety margins for existing and new reactors, taking into account the construction and operating experience, experimental information, and more sophisticated computer-aided design techniques which have become available. The main safeguards against risks arising from the Generating Boards' reactors are the achievement of high standards of design, construction and operation, in conjunction with comprehensive fault analyses to ensure that adequate protective equipment is provided. The most important analyses refer to faults which can lead to excessive fuel element temperatures arising from an increase in power or a reduction in cooling capacity. They include the possibility of unintended control rod withdrawal at power or at start-up, coolant flow failure, pressure circuit failure, loss of boiler feed water, and failure of electric power. The paper reviews the protective equipment, and the policy for reactor safety assessments which include application of maximum credible accident philosophy and later the limited use of reliability and probability methods. Some of the Generating Boards' reactors are now more than half way through their planned working lives and during this time safety protective equipment has occasionally been brought into operation, often for spurious reasons. The general performance, of safety equipment is reviewed particularly for incidents such as main turbo-alternator trip, circulator failure, fuel element failures and other similar events, and some problems which have given rise to

Aviation Safety: FAA Oversight of Repair Stations Needs Improvement

Science.gov (United States)

1997-10-24

This report by the General Accounting Office examines the Federal Aviation : Administration's (FAA) oversight of the aviation repair station industry. : Specifically, this report addresses the following questions: (1) What is the : nature and scope o...

AECB staff annual assessment of the Pickering A and B Nuclear Generating Stations for the year 1995

International Nuclear Information System (INIS)

1996-06-01

This report is the Atomic Energy Control Board (AECB) staff assessment of safety at the Pickering Nuclear Generating Station (PNGS-A and PNGS-B) for 1995. Our on-site Project Officers and Ottawa-based specialists monitored the stations throughout the year. In 1995, compliance with the Transportation Packaging of Radioactive Materials Regulations and the Cost Recovery Fees Regulations was satisfactory. The performance of the special safety systems was good. Releases of radioactive materials from the station were low and well below the legal limits for public safety. 10 tabs., 7 figs

Safety study application guide

International Nuclear Information System (INIS)

1993-07-01

Martin Marietta Energy Systems, Inc., (Energy Systems) is committed to performing and documenting safety analyses for facilities it manages for the Department of Energy (DOE). Included are analyses of existing facilities done under the aegis of the Safety Analysis Report Upgrade Program, and analyses of new and modified facilities. A graded approach is used wherein the level of analysis and documentation for each facility is commensurate with the magnitude of the hazard(s), the complexity of the facility and the stage of the facility life cycle. Safety analysis reports (SARs) for hazard Category 1 and 2 facilities are usually detailed and extensive because these categories are associated with public health and safety risk. SARs for Category 3 are normally much less extensive because the risk to public health and safety is slight. At Energy Systems, safety studies are the name given to SARs for Category 3 (formerly open-quotes lowclose quotes) facilities. Safety studies are the appropriate instrument when on-site risks are limited to irreversible consequences to a few people, and off-site consequences are limited to reversible consequences to a few people. This application guide provides detailed instructions for performing safety studies that meet the requirements of DOE Orders 5480.22, open-quotes Technical Safety Requirements,close quotes and 5480.23, open-quotes Nuclear Safety Analysis Reports.close quotes A seven-chapter format has been adopted for safety studies. This format allows for discussion of all the items required by DOE Order 5480.23 and for the discussions to be readily traceable to the listing in the order. The chapter titles are: (1) Introduction and Summary, (2) Site, (3) Facility Description, (4) Safety Basis, (5) Hazardous Material Management, (6) Management, Organization, and Institutional Safety Provisions, and (7) Accident Analysis

AECB staff annual assessment of the Pickering A and B Nuclear Generating Stations for the year 1996

International Nuclear Information System (INIS)

1997-06-01

The Atomic Energy Control Board is the independent federal agency that controls all nuclear activities in Canada. A major use of nuclear energy in Canada is electricity production. The AECB assesses every station's performance against legal requirements, including the conditions in the operating licence. Each station is inspected and all aspects of the station's operation and management is reviewed. This report is the AECB staff assessment of reactor safety at the Pickering A and B Generating Stations for 1996. PNGS-A and PNGS-B operated safely during 1996. Although the risk to the workers and the public is low, major safety related changes are necessary at the stations and the sustainability of those changes needs to be demonstrated. Improvement is needed by Ontario Hydro in meeting the time limits for reporting reportable events. Ontario Hydro's follow up to events and causal factor analyses continue to need improvements. Improvements are needed to operational safety and reactor maintenance at both A and B. There are signs of improvement through Ontario Hydro's plan for recovery, and in station management changes. There also appears to be commitment to safety expressed at the highest level of the utility

Performance of Existing Hydrogen Stations

Energy Technology Data Exchange (ETDEWEB)

Sprik, Samuel [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Kurtz, Jennifer M [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Ainscough, Christopher D [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Saur, Genevieve [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Peters, Michael C [National Renewable Energy Laboratory (NREL), Golden, CO (United States)

2017-12-01

In this presentation, the National Renewable Energy Laboratory presented aggregated analysis results on the performance of existing hydrogen stations, including performance, operation, utilization, maintenance, safety, hydrogen quality, and cost. The U.S. Department of Energy funds technology validation work at NREL through its National Fuel Cell Technology Evaluation Center (NFCTEC).

Sizewell 'B' PWR pre-construction safety report

International Nuclear Information System (INIS)

1982-04-01

The Pre-Construction Safety Report (PCSR) for a PWR power station to be constructed as Sizewell 'B' is presented in 13 volumes containing 16 chapters. The PCSR has been submitted to the Nuclear Installations Inspectorate in support of the Central Electricity Generating Board's application for consent to the extension at Sizewell. It describes the design and provides the safety case for the proposed station, which comprises a 4-loop pressurized water reactor with associated generating plant and supporting auxiliary equipment. A general description of the station and its site is given. The strategy for ensuring nuclear safety is set out and the general design aspects of systems and plant outlined. The plant and systems, including their safety design bases and the fault analyses carried out for the design are described. Finally the way in which the plant will be decommissioned at the end of its useful life is outlined. (U.K.)

[Exploring nurse, usage effectiveness of mobile nursing station].

Science.gov (United States)

Chang, Fang-Mei; Lee, Ting-Ting

2013-04-01

A mobile nursing station is an innovative cart that integrates a wireless network, information technology devices, and online charts. In addition to improving clinical work and workflow efficiencies, data is integrated among different information systems and hardware devices to promote patient safety. This study investigated the effectiveness of mobile nursing cart use. We compared different distributions of nursing activity working samples to evaluate the nursing information systems in terms of interface usability and usage outcomes. There were two parts of this study. Part one used work sampling to collect nursing activity data necessary to compare a unit that used a mobile nursing cart (mobile group, n = 18) with another that did not (traditional group, n = 14). Part two applied a nursing information system interface usability questionnaire to survey the mobile unit with nurses who had used a mobile nursing station (including those who had worked in this unit as floating nurses) (n = 30) in order to explore interface usability and effectiveness. We found that using the mobile nursing station information system increased time spent on direct patient care and decreased time spent on indirect patient care and documentation. Results further indicated that participants rated interface usability as high and evaluated usage effectiveness positively. Comments made in the open-ended question section raised several points of concern, including problems / inadequacies related to hardware devices, Internet speed, and printing. This study indicates that using mobile nursing station can improve nursing activity distributions and that nurses hold generally positive attitudes toward mobile nursing station interface usability and usage effectiveness. The authors thus encourage the continued implementation of mobile nursing stations and related studies to further enhance clinical nursing care.

Extended station blackout analyses of an APR1400 with MARS-KS

Directory of Open Access Journals (Sweden)

Kim Woongbae

2016-01-01

Full Text Available The Fukushima Daiichi nuclear power plant accident shows that natural disasters such as earthquakes and the subsequent tsunamis can cause station blackout for several days. The electric energy required for essential systems during a station blackout is provided from emergency backup batteries installed at the nuclear power plant. In South Korea, in the event of an extended station blackout, the life of these emergency backup batteries has recently been extended from 8 hours to 24 hours at Shin-Kori 5, 6, and APR1400 for design certification. For a battery life of 24 hours, available safety means system, equipment and procedures are studied and analyzed in their ability to cope with an extended station blackout. A sensitivity study of reactor coolant pump seal leakage is performed to verify how different seal leakages could affect the system. For simulating extended station blackout scenarios, the best estimate MARS-KS computer code was used. In this paper, an APR1400 RELAP5 input deck was developed for station blackout scenario to analyze operation strategy by manually depressurizing the reactor coolant system through the steam generator's secondary side. Additionally, a sensitivity study on reactor coolant pump seal leakage was carried out.

The Heysham 2 and Torness nuclear power station projects

International Nuclear Information System (INIS)

Cameron, P.J.

1989-01-01

At the beginning of the design of the Heysham 2 and Torness Nuclear Power Stations in the UK, it was agreed that the Hinkley Point B/Hunterston B Advanced Gas Cooled reactor design would be repeated. However, the Hinkley Point B/Hunterston B designs dated from 1964, and inevitably safety requirements had escalated in the intervening years. Furthermore, operating experience gained from Hinkley Point B/Hunterston B showed that it was desirable to make changes to particular features. Design changes were therefore included where it was considered essential to improve performance, to improve safety and to improve engineering. The resulting station designs are described. (author)

Safety Evaluation Report related to the operation of Limerick Generating Station, Units 1 and 2 (Docket Nos. 50-352 and 50-353). Supplement 2

International Nuclear Information System (INIS)

1984-10-01

In August 1983 the staff of the Nuclear Regulatory Commission issued its Safety Evaluation Report (NUREG-0991) regarding the application of the Philadelphia Electric Company (the applicant) for licenses to operate the Limerick Generating Station, Units 1 and 2, located on a site in Montgomery and Chester Counties, Pennsylvania. This supplement addresses further issues that require resolution and closes them out

77 FR 76003 - Submission for OMB Review; Comment Request-Safety Standard for Omnidirectional Citizens Band Base...

Science.gov (United States)

2012-12-26

... Request--Safety Standard for Omnidirectional Citizens Band Base Station Antennas AGENCY: Consumer Product... information associated with the Commission's safety standard for omnidirectional citizens band base station... information required in the Safety Standard for Omnidirectional Citizens Band Base Station (16 CFR Part 1204...

The tolerability of risk from nuclear power stations: a discussion of the HSE's guidelines

International Nuclear Information System (INIS)

Ryder, E.A.; Woods, P.B.

1989-01-01

The Health and Safety Executive's discussion document, ''The Tolerability of risk from nuclear power stations'', published just a year ago considers the concept of risk and the broad principles of risk assessment and proposes guidelines on the tolerable levels of individual and societal risks from normal operation and from accidents at nuclear power stations. This paper discusses how these guidelines might be applied in safety assessments so as to ensure an acceptable level of safety and considers some of the problems inherent in the estimation of risk from nuclear power stations. It shows how the tolerable risk levels of the Health and Safety Executive's document are related to the standards used by HM Nuclear Installations Inspectorate when licensing nuclear installations. Some of the uncertainties in the estimation of acceptable risk are discussed as are the acceptance criteria used by the Inspectorate in its assessments of both normal and accident situations. (author)

Safety Evaluation Report related to the operation of Waterford Steam Electric Station, Unit No. 3 (Docket No. 50-382). Supplement 9

International Nuclear Information System (INIS)

1984-12-01

Supplement 9 to the Safety Evaluation Report for Louisiana Power and Light's application for a license to operate Waterford Steam Electric Station, Unit 3 (Docket No. 50-382), located in St. Charles Parish, Louisiana, has been jointly prepared by the Office of Nuclear Reactor Regulation and the Region IV Office of the US Nuclear Regulatory Commission. This supplement provides the results of the staff's completion of its evaluation of approximately 350 allegations and concerns of poor construction practices at the Waterford 3 facility

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.)

Safety of CANDU nuclear power stations

International Nuclear Information System (INIS)

Snell, V.G.

1978-11-01

A nuclear plant contains a large amount of radioactive material which could be a potential threat to public health. The plant is therefore designed, built and operated so that the risk to the public is low. Careful design of the normal reactor systems is the first line of defense. These systems are highly resistant to an accident happening in the first place, and can also be effective in stopping it if it does happen. Independent and redundant safety sytems minimize the effects of an accident, or stop it completely. They include shutdown systems, emergency core cooling systems, and containment systems. Massive impairment of any one safety system together with an accident can be tolerated. This 'defence in depth' approach recognizes that men and machines are imperfect and that the unexpected happens. The nuclear power plant need not be perfect to be safe. To allow meaningful judgements we must know how safe the plant is. The Atomic Energy Control Board guidelines give one such measure, but they may overestimate the true risk. We interpret these guidelines as an upper limit to the total risk, and trace their evolution. (author)

Regulatory experience in nuclear power station decommissioning

International Nuclear Information System (INIS)

Ross, W.M.; Waters, R.E.; Taylor, F.E.; Burrows, P.I.

1995-01-01

In the UK, decommissioning on a licensed nuclear site is regulated and controlled by HM Nuclear Installations Inspectorate on behalf of the Health and Safety Executive. The same legislative framework used for operating nuclear power stations is also applied to decommissioning activities and provides a continuous but flexible safety regime until there is no danger from ionising radiations. The regulatory strategy is discussed, taking into account Government policy and international guidance for decommissioning and the implications of the recent white paper reviewing radioactive waste management policy. Although each site is treated on a case by case basis as regulatory experience is gained from decommissioning commercial nuclear power stations in the UK, generic issues have been identified and current regulatory thinking on them is indicated. Overall it is concluded that decommissioning is an evolving process where dismantling and waste disposal should be carried out as soon as reasonably practicable. Waste stored on site should, where it is practical and cost effective, be in a state of passive safety. (Author)

Public safety assessment of electromagnetic radiation exposure from mobile base stations.

Science.gov (United States)

Alhekail, Z O; Hadi, M A; Alkanhal, M A

2012-09-01

Exposure of the general public to electromagnetic radiation originating from randomly selected GSM/WCDMA base stations in Riyadh, Kingdom of Saudi Arabia has been assessed in the context of the International Commission on Non-Ionising Radiation Protection (ICNIRP) guidelines. The purpose of the measurement was to record the maximum power density of signals to estimate possible worst case exposure at each measurement location. These power density measurements were carried out at 60 mobile base stations located in different regions of the city. For each of these sites, three sectors were operational, yielding a total of 180 sectors. Two positions were identified per site with the greatest power density values. Exposures from these base stations were generally found to be in the range of 0.313 to 0.00000149% of the ICNIRP general public reference level, and the greatest exposure near any of the base stations was 21.96 mW m(-2) for a wideband measurement in the 75-3000 MHz frequency range. Analysis of the measured data reveals several trends for different mobile bands with respect to maximum exposure in those locations. Additionally, a simplified calculation method for the electromagnetic fields was used to compare calculated and the measured data. It was determined, on the basis of both results of the measurements and calculations carried out for these selected base stations, that members of the public would not be exposed to in excess of a small fraction of the ICNIRP guidelines at any of those sites. These are first such measurements to be made in the Middle East and provide assurance that exposures in this region of the world do not seem to be any greater than elsewhere.

A molecular epidemiological study of methicillin-resistant Staphylococci environmental contamination in railway stations and coach stations in Guangzhou of China.

Science.gov (United States)

Lin, J L; Peng, Y; Ou, Q T; Lin, D X; Li, Y; Ye, X H; Zhou, J L; Yao, Z J

2017-02-01

Methicillin-resistant Staphylococcus aureus (MRSA) has caused a series of public health problems since it was first found in 1961. However, there are few research studies on the MRSA environmental contamination in railway stations and coach stations. Therefore, the aim of this study was to determine MRSA environmental contamination in public transport stations. Between December 2013 and January 2014, 380 surface samples from three railway stations (180) and four coach stations (200) in Guangzhou were collected to isolate and determine the prevalence and characteristics of Staphylococci strains. 39·21% of all samples were Staphylococci isolates, 1·58% of Staphylococci isolates were MRSA isolates, and 6·05% were methicillin-susceptible S. aureus. The proportion of multidrug resistant among 149 Staphylococci isolates was 75·84%. None of MRSA isolates was identified with the Panton-Valentine Leukocidin (PVL) genes, and one of them was identified with the qac gene. Four MRSA isolates were Staphylococcal Cassette Chromosome mec IVa, and the other two were nontypeable. Staphylococcus aureus isolates were classified into several sequence types (STs), and STs showed possible cross-transmissions of isolates from various sources. Methicillin-resistant Staphylococci contamination prevalence was high, and the environment of stations may be the vectors transmitting the Staphylococci to passengers. This is the first study to comprehensively report the prevalence, antibiotic resistance, and molecular characteristics of contamination of Staphylococci isolates in railway stations and coach stations of China. It will have great public health implications on infection control in community settings because of the serious hazard of Staphylococci, especially methicillin-resistant Staphylococci. Our findings have provided evidence for relevant departments to reduce the contamination of Staphylococci in environment of public transport stations. © 2016 The Society for Applied

Systems analysis of radiation safety during dismantling of power-plant equipment at a nuclear power station

International Nuclear Information System (INIS)

Bylkin, B.K.; Shpitser, V.Ya.

1993-01-01

A systems analysis of the radiation safety makes possible an ad hoc determination of the elements forming the system, as well as the establishment of the characteristics of their interaction with radiation-effect factors. Here the authors will present part of the hierarchical analysis procedure, consisting in general of four separate procedures. The purpose is to investigate and analyze the mean and stable (on the average) indices of radiation safety, within the framework of alternative mathematical models of dismantling the power-plant equipment of a nuclear power station. The following three of the four procedures are discussed: (1) simulated projection, of the processing of radioactive waste; (2) analysis of the redistribution of radionuclides during the industrial cycle of waste treatment; (3) planning the collective dose load during the dismantling operation. Within the framework of the first of these procedures, the solutions to the problem of simulating a waste-treatment operation of maximum efficiency are analyzed. This analysis is based on the use of a data base for the parameters of the installations, assemblies, and equipment, enabling the integration of these in a simulation of a complex automated facility. The results were visualized in an AUTOCAD-10 medium using a graphical data base containing an explanation of the rooms

Case study on the use of PSA methods: Station blackout risk at Millstone Unit 3

International Nuclear Information System (INIS)

1991-04-01

In Westinghouse pressurized water reactors, severe accidents sequences resulting from station blackout have been recognized to be significant contributors to risk of core damage and public consequences. To properly quantify the risk of station blackout it is necessary to consider all possible types of core damage scenarios. Having obtained an accurate representation of the types of core damage scenarios involved specific areas of vulnerability can be pinpointed for further improvement. In performing this analysis it was decided to use time dependent probabilistic safety assessment method to provide a more realistic treatment of time dependent failure and recovery. Overview of the analysis, calculation procedures and methods, interpretation of the results are discussed. Peer review process is described. 13 refs, 19 figs

Safety Evaluation Report related to the operation of Wolf Creek Generating Station, Unit No. 1 (Docket No. STN 50-482). Supplement No. 5

International Nuclear Information System (INIS)

1985-03-01

This report supplements the Safety Evaluation Report (SER) for the application filed by the Kansas Gas and Electric Company, as applicant and agent for the owners, for a license to operate the Wolf Creek Generating Station, Unit 1 (Docket No. STN 50-482). The facility is located in Coffey County, Kansas. This supplement has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission and provides recent information regarding resolution of the open items identified in the SER. Because of the favorable resolution of the items discussed in this report, the staff concludes that the facility can be operated by the applicant without endangering the health and safety of the public

Safety evaluation report related to the operation of Seabrook Station, Units 1 and 2. Docket Nos. 50-443 and 50-444. Suppl. 1

International Nuclear Information System (INIS)

1983-04-01

This report supplements the Safety Evaluation Report (NUREG-0896, March 1983) for the application filed by the Public Service Company of New Hampshire, et. al., for licenses to operate Seabrook Station, Units 1 and 2 (Docket Nos. STN 50-443 and STN 50-444). It has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission and provides recent information on open items identified in the SER. The facility is located in Seabrook, New Hampshire. Subject to favorable resolution of the items discussed in this report, the staff concludes that the facility can be operated by the applicant without endangering the health and safety of the public

Review on Selection and Suitability of Rail Transit Station Design Pertaining to Public Safety

Science.gov (United States)

Akabal, Farah Mohd; Masirin, Mohd Idrus Haji Mohd; Abidin Akasah, Zainal; Rohani, Munzilah Md

2017-08-01

Railway has emerged as a fast, convenient, safe, clean, and low-cost alternative to air and road transportation. Many countries have invested in rail transportation. In America, Europe and Asia, large investments are planned for rail transportation. This is because congestion problems can be reduced with the introduction of rail transportation. Rail transportation involves several components which are important to ensure the smooth and safe delivery of services such as locomotives, rail stations and railway tracks. Rail transit stations are places where trains stop to pick-up and drop-off passengers. Stations are vital for many to enable them to engage in work and social commitments. This paper focuses only on the rail transit station as it is one of the important components in rail transportation. It is also considered as a key public meeting place and space for interactions in a community. The role of rail transit station and the requirements of a good rail transit station are also described in this paper. Steps in selecting the location of rail transit station include the function and facilities in rail transit station are discussed with reference to best practices and handbooks. Selection of the appropriate rail transit station locations may help users indirectly. In addition, this paper will also elucidate on the design considerations for an efficient and effective rail transit station. Design selections for the rail transit station must be balanced between aesthetic value and functional efficiency. The right design selection may help conserve energy, assure and facilitate consumers even thought a rail transit station plays a smaller role in attracting consumers compared to a shopping complex or a residential building. This will contribute towards better and greener building for a green transportation facility. Thus, with this paper it is expected to assist the relevant authority to identify important elements in the selection and determination of suitable

KHNP special safety review

International Nuclear Information System (INIS)

Lee, Tae-Ho; Lee, Bang-Jin; Lee, Soung-Hee; Park, Goon-Cherl

2009-01-01

Commemorating the 30 year anniversary of commercial nuclear power plant operation in KOREA, Korea Hydro and Nuclear Power Co., Ltd. (KHNP) has conducted a Special Safety Review (SSR) of its 20 operating units to understand their safety performance and to identify any areas that need improvement. The SSR reviewed all 20 operating units for 2 weeks per site. Areas that were reviewed are Safety Margins, Plant Performance, Employee Safety, Employee Performance and Performance Improvement Process. Each review team consisted of international and domestic members. The international reviewers were from IAEA, WANO and INPO. The domestic reviewers consisted of professors, Engineering Company, Research Institute and KHNP experts. The review confirmed safe and reliable operations of the 20 nuclear units. The common understanding resulted from the SSR is as follows. Firstly, KHNP corporate and its plants confirmed and shared mutual understanding on recurring areas for improvements, especially in the areas of Organizational Effectiveness, Industrial Safety, Human Performance, Configuration Management, Operations, Equipment Performance and Material Condition. Secondly, KHNP understood that plant and department level performances are directly related to the leadership and competency of the management team including supervisors. Thirdly, the strengths of individual stations that consistently have produced good results need to be shared with the other KHNP stations. Finally, KHNP learned that strong corporate leadership and support are needed to resolve most of the areas for improvement since they are common to all KHNP stations. (author)

Reliability study of nuclear power stations

International Nuclear Information System (INIS)

Nemec, J.; Sedlacek, J.

1975-01-01

The paper considers the physical and mathematical bases underlying nuclear power station reliability from the standpoints of material fatigue, thermal yield and ageing. The risk of failure of nuclear power station components is determined by means of the Markov stochastic process

Station blackout core damage frequency in an advanced nuclear reactor

International Nuclear Information System (INIS)

Carvalho, Luiz Sergio de

2004-01-01

Even though nuclear reactors are provided with protection systems so that they can be automatically shut down in the event of a station blackout, the consequences of this event can be severe. This is because many safety systems that are needed for removing residual heat from the core and for maintaining containment integrity, in the majority of the nuclear power plants, are AC dependent. In order to minimize core damage frequency, advanced reactor concepts are being developed with safety systems that use natural forces. This work shows an improvement in the safety of a small nuclear power reactor provided by a passive core residual heat removal system. Station blackout core melt frequencies, with and without this system, are both calculated. The results are also compared with available data in the literature. (author)

Key Techniques and Risk Management for the Application of the Pile-Beam-Arch (PBA Excavation Method: A Case Study of the Zhongjie Subway Station

Directory of Open Access Journals (Sweden)

Yong-ping Guan

2014-01-01

Full Text Available The design and construction of shallow-buried tunnels in densely populated urban areas involve many challenges. The ground movements induced by tunneling effects pose potential risks to infrastructure such as surface buildings, pipelines, and roads. In this paper, a case study of the Zhongjie subway station located in Shenyang, China, is examined to investigate the key construction techniques and the influence of the Pile-Beam-Arch (PBA excavation method on the surrounding environment. This case study discusses the primary risk factors affecting the environmental safety and summarizes the corresponding risk mitigation measures and key techniques for subway station construction using the PBA excavation method in a densely populated urban area.

Aspects of safety and of functional construction and configuration in planning and designing nuclear heating stations

International Nuclear Information System (INIS)

Adam, E.; Mueller, R.; Boettger, M.; Kremtz, U.

1982-01-01

The present studies are based on the design of a technological project of a nuclear heating station with a unit power of 250 MW. Essentially, this nuclear heating station is a three-circuit plant, the primary coolant circuit being based on natural circulation through the reactor vessel with integrated heat exchangers. Starting from the social objective and the derived development structure of the territory, the siting problems in integrating the nuclear heating stations have to be solved. On the basis of the resulting dimensions of the containment the technical and economical specifications of different versions of containment design are evaluated. (author)

Studies and research relatives to the safety of the Super-Phenix project

International Nuclear Information System (INIS)

Anselin, F.; Penet, F.

1978-01-01

The analysis of safety reports concerning the Creys Malville power station (Superphenix) must be based on technical data supplied by the NERSA, responsible for the plant, and on results of research and development programme carried out in various establishments and at the CEA in particular. By virtue of the procedure laid down for the safety analysis, i.e. analysis by the barrier method, verification of reactor shut down rules at power, permanence of cooling, confinement of dangerous products in the event of hypothetical failure of the above two functions safety R and D programmes have a double aspect: accident prevention on the one hand and study of the development of accident, even the most hypothetical on the other. In the accident prevention field the studies deal with the resistance of barriers under normal and accidental working conditions, inspection systems and reactor safety functions allowing abnormal situations to be detected and the reactor shut down; whence the special emphasis placed on emergency shut-down and cooling systems. In the accident field the R and D activities cover a wide range of studies on phenomena liable to arise, independently of their probability of occurence during the lifetime of the reactor; heating in the mass or boiling of sodium, fuel, meeting, movements of fused materials, fuel-sodium thermal interaction, core deformation, resistance of confinement recovery of molten fuel, post-accident cooling, transfer of radioactivity and contamination outside the reactor, radiological consequences and means of confinement of dangerous products [fr

NRC/UBC fuelling station with intelligent compression

International Nuclear Information System (INIS)

Dada, A.; Boyd, B.; Law, L.; Semczyszyn, D.

2004-01-01

BOC Canada Ltd. will design, integrate and construct the second fueling station on the Hydrogen Highway. This station will be located at the National Research Council's Institute for Fuel Cell Innovation on the campus of the University of British Columbia. BOC's design will bring together an existing alkaline electrolyser, new compression, storage and dispensing. The station will be designed to serve fuel cell passenger vehicles using 350-bar storage. However, the flexible design concept will allow for many other user needs including the potential for servicing larger vehicles, as well as filling portable storage systems for use at satellite stations. The novel station design also offers the potential to fuel from multiple hydrogen sources. Together with NRC, this fueling station will be used to increase public, consumer and investor awareness of hydrogen technologies. Design and construction of this facility will assist in the development of industry codes and standards and familiarize authorities having jurisdiction with hydrogen fueling. The system concept offers the utmost attention to safety, novelty and flexibility. (author)

Reactor Safety Commission Code of Practice for Pressurized Water Reactors

International Nuclear Information System (INIS)

1990-01-01

The Reactor Safety Commission of the Federal German Republic has summarized in the form of Official Guidelines the safety requirements which, in the Commission's view, have to be met in the design, construction and operation of a nuclear power station equipped with a pressurized water reactor. The Third Edition of the RSK Guidelines for pressurized water reactors dated 14.10.81. is a revised and expanded version of the Second Edition dated 24.1.79. The Reactor Safety Commission will with effect from October 1981 use these Guidelines in consultations on the siting of and safety concept for the installation approval of future pressurized water reactors and will assess these nuclear power stations during their erection in the light of these Guidelines. They have not however been immediately conceived for the adaptation of existing nuclear power stations, whether under construction or in operation. The scope of application of these Guidelines to such nuclear power stations will have to be examined for each individual case. The main aim of the Guidelines is to simplify the consultation process within the reactor Safety Commission and to provide early advice on the safety requirements considered necessary by the Commission. (author)

Station blackout transient at the Browns Ferry Unit 1 Plant: a severe accident sequence analysis (SASA) program study

International Nuclear Information System (INIS)

Schultz, R.R.

1982-01-01

Operating plant transients are of great interest for many reasons, not the least of which is the potential for a mild transient to degenerate to a severe transient yielding core damage. Using the Browns Ferry (BF) Unit-1 plant as a basis of study, the station blackout sequence was investigated by the Severe Accident Sequence Analysis (SASA) Program in support of the Nuclear Regulatory Commission's Unresolved Safety Issue A-44: Station Blackout. A station blackout transient occurs when the plant's AC power from a comemrcial power grid is lost and cannot be restored by the diesel generators. Under normal operating conditions, f a loss of offsite power (LOSP) occurs [i.e., a complete severance of the BF plants from the Tennessee Valley Authority (TVA) power grid], the eight diesel generators at the three BF units would quickly start and power the emergency AC buses. Of the eight diesel generators, only six are needed to safely shut down all three units. Examination of BF-specific data show that LOSP frequency is low at Unit 1. The station blackout frequency is even lower (5.7 x 10 - 4 events per year) and hinges on whether the diesel generators start. The frequency of diesel generator failure is dictated in large measure by the emergency equipment cooling water (EECW) system that cools the diesel generators

Improving Safety on the International Space Station: Transitioning to Electronic Emergency Procedure Books on the International Space Station

Science.gov (United States)

Carter-Journet, Katrina; Clahoun, Jessica; Morrow, Jason; Duncan, Gary

2012-01-01

The National Aeronautics and Space Administration (NASA) originally designed the International Space Station (ISS) to operate until 2015, but have extended operations until at least 2020. As part of this very dynamic Program, there is an effort underway to simplify the certification of Commercial ]of ]the ]Shelf (COTS) hardware. This change in paradigm allows the ISS Program to take advantage of technologically savvy and commercially available hardware, such as the iPad. The iPad, a line of tablet computers designed and marketed by Apple Inc., was chosen to support this endeavor. The iPad is functional, portable, and could be easily accessed in an emergency situation. The iPad Electronic Flight Bag (EFB), currently approved for use in flight by the Federal Aviation Administration (FAA), is a fraction of the cost of a traditional Class 2 EFB. In addition, the iPad fs ability to use electronic aeronautical data in lieu of paper in route charts and approach plates can cut the annual cost of paper data in half for commercial airlines. ISS may be able to benefit from this type of trade since one of the most important factors considered is information management. Emergency procedures onboard the ISS are currently available to the crew in paper form. Updates to the emergency books can either be launched on an upcoming visiting vehicle such as a Russian Soyuz flight or printed using the onboard ISS printer. In both cases, it is costly to update hardcopy procedures. A new operations concept was proposed to allow for the use of a tablet system that would provide a flexible platform to support space station crew operations. The purpose of the system would be to provide the crew the ability to view and maintain operational data, such as emergency procedures while also allowing Mission Control Houston to update the procedures. The ISS Program is currently evaluating the safety risks associated with the use of iPads versus paper. Paper products can contribute to the flammability

Integrated plant safety assessment: Systematic Evaluation Program, San Onofre Nuclear Generating Station, Unit 1 (Docket No. 50-206): Final report

International Nuclear Information System (INIS)

1986-12-01

The Systematic Evaluation Program was initiated in February 1977 by the US Nuclear Regulatory Commission to review the designs of older operating nuclear reactor plants to reconfirm and document their safety. The review provides: (1) an assessment of how these plants compare with current licensing safety requirements relating to selected issues; (2) a basis for deciding on how these differences should be resolved in an integrated plant review; and (3) a documented evaluation of plant safety. This report documents the review of San Onofre Nuclear Generating Station, Unit 1, operated by Southern California Edison Company. The San Onofre plant is one of ten plants reviewed under Phase II of this program. This report indicates how 137 topics selected for review under Phase I of the program were addressed. Equipment and procedural changes have been identified as a result of the review. This report will be one of the bases in considering the issuance of a full-term operating license in place of the existing provisional operating license. This report also addresses the comments and recommendations made by the Advisory Committee on Reactor Safeguards in connection with its review of the draft report issued in April 1985

Commerical electric power cost studies. Capital cost addendum multi-unit coal and nuclear stations

International Nuclear Information System (INIS)

1977-09-01

This report is the culmination of a study performed to develop designs and associated capital cost estimates for multi-unit nuclear and coal commercial electric power stations, and to determine the distribution of these costs among the individual units. This report addresses six different types of 2400 MWe (nominal) multi-unit stations as follows: Two Unit PWR Station-1139 MWe Each, Two Unit BWR Station-1190 MWe Each, Two Unit High Sulfur Coal-Fired Station-1232 MWe Each, Two Unit Low Sulfur Coal-Fired Station-1243 MWe Each, Three Unit High Sulfur Coal-Fired Station-794 MWe Each, Three Unit Low Sulfur Coal-Fired Station-801 MWe Each. Recent capital cost studies performed for ERDA/NRC of single unit nuclear and coal stations are used as the basis for developing the designs and costs of the multi-unit stations. This report includes the major study groundrules, a summary of single and multi-unit stations total base cost estimates, details of cost estimates at the three digit account level and plot plan drawings for each multi-unit station identified

76 FR 50274 - Terrestrial Environmental Studies for Nuclear Power Stations

Science.gov (United States)

2011-08-12

... Draft Regulatory Guides in the ``Regulatory Guides'' collection of the NRC's Library at http://www.nrc... Stations AGENCY: Nuclear Regulatory Commission. ACTION: Draft regulatory guide; request for comment... draft regulatory guide (DG), DG-4016, ``Terrestrial Environmental Studies for Nuclear Power Stations...

Probabilistic assessment of the Juragua NPP response under Station Blackout conditions

International Nuclear Information System (INIS)

Valhuerdi Debesa, C.; Vilaragut Llanes, J.J.

1996-01-01

Assessment of the NPP response under SBO (station Blackout) conditions is a current safety issue of special interest, In the case of Juragua NPP, the safety assessment related to this topic is very important, taking into account the peculiarities of the Cuban Electro energetic System: small and long island, without possibilities of conexion beyond its borders and under the incidence of tropical phenomena In this papers a preliminary evaluation is presented of the potential incidence of Station Blackout conditions for Juragua NPP. the importance of this sort of events for the safety of the plant is evaluated, the factors which condition it are identified and measures for its prevention or recovering the normal situation if such an event takes place are proposed

Potentialities and type of integrating nuclear heating stations into district heating systems

International Nuclear Information System (INIS)

Munser, H.; Reetz, B.; Schmidt, G.

1978-01-01

Technical and economical potentialities of applying nuclear heating stations in district heating systems are discussed considering the conditions of the GDR. Special attention is paid to an optimum combination of nuclear heating stations with heat sources based on organic fuels. Optimum values of the contribution of nuclear heating stations to such combined systems and the economic power range of nuclear heating stations are estimated. Final considerations are concerned with the effect of siting and safety concepts of nuclear heating stations on the structure of the district heating system. (author)

Millstone 3 risk evaluation report. An overall review and evaluation of the Millstone Unit 3 probabilistic safety study

International Nuclear Information System (INIS)

Kelly, G.; Barrett, R.; Buslik, A.

1986-06-01

In 1981, the US Nuclear Regulatory Commission (NRC) requested Northeast Utilities to perform a design-specific probabilistic safety study (PSS) for Millstone Nuclear Power Station, Unit No. 3 (Millstone 3). In 1983, Northeast Utilities submitted the Millstone 3 Probabilistic Safety Study for review by the NRC staff. The NRC staff prepared the Millstone 3 Risk Evaluation Report, which discusses the findings regarding the PSS. The PSS estimates that the mean annual core damage frequency due to internal and external events is 5 x 10 -5 and 2 x 10 -5 , respectively. The NRC staff's Risk Evaluation Report estimates that the mean annual core damage frequency is about 2 x 10 -4 for internal events and lies between 1 x 10 -5 and 2 x 10 -4 for external events. The NRC staff estimates that station blackout dominates internal and external event core damage frequencies. The staff recommends that Northeast Utilities perform an engineering analysis on upgrading the diesel generator lube oil cooler anchorage system and on adding a manually operated, AC-independent containment spray system. The staff also recommends that Northeast Utilities prepare two emergency procedures (loss of room cooling and relay chatter due to an earthquake) to help reduce uncertainties. (Subsequent to the completion of this document, Northeast Utilities and the NRC staff have continued a dialogue regarding station blackout from events other than earthquakes. Both Northeast Utilities and the staff have performed additional evaluations, which have drawn their results closer together. Final requirements, if any, for the prevention or mitigation of station blackout from events other than earthquakes have not yet been determined.) 26 refs., 16 tabs

Safety assessment of RF and microwave radiation emitted by the mobile telephone base station (MTBS) in Malaysia: experience and challenge

International Nuclear Information System (INIS)

Roha Tukimin; Rozaimah Abd Rahim; Mohamad Amirul Nizam; Mohd Yusof Mohd Ali

2007-01-01

Non-ionising radiation (NIR) is known to be hazardous if the amount received is excessive. It is a fact that NIR, including extremely low frequency (ELF) electromagnetic fields, radiofrequency (RF) and microwave radiation can be found almost everywhere generated by both natural and man-made source. This is due to increase in demand for telecommunication and wireless technology which is become very important and as part of our lives. However, the widespread of the relevant technology contributed more NIR man-made sources exposure to the human. Due to public concern their potential of causing such health hazard, members of public and companies approached and request NIR Group of Nuclear Malaysia to carry out surveys and safety assessments of radiofrequency and microwave radiation emitted by the mobile telephone base station (MTBS) erected near the residential area or installed on the rooftop of the commercial building. Objective of the survey was to assess the presence of radiofrequency and microwave radiation and to identify radiation level which may lead to significant personnel exposure. Findings of the survey was compared to the standard guidelines issued by Malaysian Communication and Multimedia Commission (MCMC) and International Committee on Non-Ionising Radiation Protection (ICNIRP). This paper highlights the works that had been carried out by NIR Group of Nuclear Malaysia from 1997 to 2007. We will share the experience and challenge in carried out the NIR safety assessment at mobile telephone base station. Results of the assessment work will be used to develop non-ionising radiation database for future reference in Malaysia. (Author)

Advanced gas cooled reactors - Designing for safety

International Nuclear Information System (INIS)

Keen, Barry A.

1990-01-01

The Advanced Gas-Cooled Reactor Power Stations recently completed at Heysham in Lancashire, England, and Torness in East Lothian, Scotland represent the current stage of development of the commercial AGR. Each power station has two reactor turbo-generator units designed for a total station output of 2x660 MW(e) gross although powers in excess of this have been achieved and it is currently intended to uprate this as far as possible. The design of both stations has been based on the successful operating AGRs at Hinkley Point and Hunterston which have now been in-service for almost 15 years, although minor changes were made to meet new safety requirements and to make improvements suggested by operating experience. The construction of these new AGRs has been to programme and within budget. Full commercial load for the first reactor at Torness was achieved in August 1988 with the other three reactors following over the subsequent 15 months. This paper summarises the safety principles and guidelines for the design of the reactors and discusses how some of the main features of the safety case meet these safety requirements. The paper also summarises the design problems which arose during the construction period and explains how these problems were solved with the minimum delay to programme

Advanced gas cooled reactors - Designing for safety

Energy Technology Data Exchange (ETDEWEB)

Keen, Barry A [Engineering Development Unit, NNC Limited, Booths Hall, Knutsford, Cheshire (United Kingdom)

1990-07-01

The Advanced Gas-Cooled Reactor Power Stations recently completed at Heysham in Lancashire, England, and Torness in East Lothian, Scotland represent the current stage of development of the commercial AGR. Each power station has two reactor turbo-generator units designed for a total station output of 2x660 MW(e) gross although powers in excess of this have been achieved and it is currently intended to uprate this as far as possible. The design of both stations has been based on the successful operating AGRs at Hinkley Point and Hunterston which have now been in-service for almost 15 years, although minor changes were made to meet new safety requirements and to make improvements suggested by operating experience. The construction of these new AGRs has been to programme and within budget. Full commercial load for the first reactor at Torness was achieved in August 1988 with the other three reactors following over the subsequent 15 months. This paper summarises the safety principles and guidelines for the design of the reactors and discusses how some of the main features of the safety case meet these safety requirements. The paper also summarises the design problems which arose during the construction period and explains how these problems were solved with the minimum delay to programme.

AECB staff annual assessment of the Pickering A and B Nuclear Generating Stations for the year 1994

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-06-01

The Pickering Nuclear Generating Station (PNGS) is located on the north shore of Lake Ontario, about 32 km east of downtown Toronto. It consists of two stations, PNGS-A and PNGS-B. Each station contains four reactor units. PNGS-A consists of Units 1 to 4, while PNGS-B consists of Units 5 to 8. Each unit can generate about 540 megawatts of electricity. All eight units are located within a single enclosure. Ontario Hydro`s Pickering Nuclear Division has assigned one Station Director with authority over both stations, but each station has its own organization. AECB issue a separate operating licence for each station. This report presents the Atomic Energy Control Board staff assessment of the Pickering stations` safety performance in 1994 and other aspects that they consider to have significant impact on nuclear safety. AECB based their conclusions on their observations, audits, inspections and review of information that Ontario Hydro submits to them as required by the station Operating Licences. 11 tabs., 8 figs.

AECB staff annual assessment of the Pickering A and B Nuclear Generating Stations for the year 1994

International Nuclear Information System (INIS)

1995-06-01

The Pickering Nuclear Generating Station (PNGS) is located on the north shore of Lake Ontario, about 32 km east of downtown Toronto. It consists of two stations, PNGS-A and PNGS-B. Each station contains four reactor units. PNGS-A consists of Units 1 to 4, while PNGS-B consists of Units 5 to 8. Each unit can generate about 540 megawatts of electricity. All eight units are located within a single enclosure. Ontario Hydro's Pickering Nuclear Division has assigned one Station Director with authority over both stations, but each station has its own organization. AECB issue a separate operating licence for each station. This report presents the Atomic Energy Control Board staff assessment of the Pickering stations' safety performance in 1994 and other aspects that they consider to have significant impact on nuclear safety. AECB based their conclusions on their observations, audits, inspections and review of information that Ontario Hydro submits to them as required by the station Operating Licences. 11 tabs., 8 figs

Electromagnetic Radiation Exposure from Cellular Base Station: A ...

African Journals Online (AJOL)

Electromagnetic Radiation Exposure from Cellular Base Station: A Concern for Public ... as well as safety guidelines relating to exposure of non-ionizing radiation. Global System for Mobile Communication (GSM) operators claimed that their ...

47 CFR 101.1425 - RF safety.

Science.gov (United States)

2010-10-01

... 47 Telecommunication 5 2010-10-01 2010-10-01 false RF safety. 101.1425 Section 101.1425 Telecommunication FEDERAL COMMUNICATIONS COMMISSION (CONTINUED) SAFETY AND SPECIAL RADIO SERVICES FIXED MICROWAVE... safety. MVDDS stations in the 12.2-12.7 GHz frequency band do not operate with output powers that equal...

Case study: Massachusetts Department of Public Health's Southeastern Massachusetts health study on leukemia around Pilgrim Nuclear Power Station: Who won?

Energy Technology Data Exchange (ETDEWEB)

Flanagan, Anita [Boston Edison Company (United States)

1993-07-01

In October 1990, the Massachusetts Department of Public Health released their Southeastern Massachusetts health study. This is a study of leukemia incidence in 22 towns around Pilgrim NPP, for the period 1978 through 1986. Pilgrim Station had been returned to operation following a 3 year outage, from 1986-1989. During this period, some $300 million in capital outlay was invested by Boston Edison in upgrading safety systems and installing the so-Called Three Mile Island upgrades. A copy of the peer review panel report is attached to this paper. Because of the interest in the Southeastern Massachusetts Health Study in Europe. There are three major points the Panel made which can summarized: 1. No excess of leukemia was found around Pilgrim Station. 2. The Southeastern Massachusetts Health Study over-predicted by a factor of 90 the number of leukemia cases attributable to plant operation. 3. The Southeastern Massachusetts Health Study failed to account for exposure to natural background radiation, which represents far larger biological exposure than plant Operation. Given All Of This, One Might Ask Why Didn't Common Sense Prevail In The Beginning? One Answer Might Be The Energy Of The Media In Pursuing The Story And Playing It Up No Matter What. Another Answer Might Be That The Original Study Purported To Show 'What Everyone Knows'. No One Really Stopped To Question Whether The Study Was Politically Motivated, Given That The Division of Environmental Health's Budget Had Been Cut.

Case study: Massachusetts Department of Public Health's Southeastern Massachusetts health study on leukemia around Pilgrim Nuclear Power Station: Who won?

International Nuclear Information System (INIS)

Flanagan, Anita

1993-01-01

In October 1990, the Massachusetts Department of Public Health released their Southeastern Massachusetts health study. This is a study of leukemia incidence in 22 towns around Pilgrim NPP, for the period 1978 through 1986. Pilgrim Station had been returned to operation following a 3 year outage, from 1986-1989. During this period, some $300 million in capital outlay was invested by Boston Edison in upgrading safety systems and installing the so-Called Three Mile Island upgrades. A copy of the peer review panel report is attached to this paper. Because of the interest in the Southeastern Massachusetts Health Study in Europe. There are three major points the Panel made which can summarized: 1. No excess of leukemia was found around Pilgrim Station. 2. The Southeastern Massachusetts Health Study over-predicted by a factor of 90 the number of leukemia cases attributable to plant operation. 3. The Southeastern Massachusetts Health Study failed to account for exposure to natural background radiation, which represents far larger biological exposure than plant Operation. Given All Of This, One Might Ask Why Didn't Common Sense Prevail In The Beginning? One Answer Might Be The Energy Of The Media In Pursuing The Story And Playing It Up No Matter What. Another Answer Might Be That The Original Study Purported To Show 'What Everyone Knows'. No One Really Stopped To Question Whether The Study Was Politically Motivated, Given That The Division of Environmental Health's Budget Had Been Cut

Safety evaluation report related to the operation of Waterford Steam Electric Station, Unit No. 3 (Docket No. 50-382). Supplement No. 7

International Nuclear Information System (INIS)

1984-09-01

Supplement 7 to the Safety Evaluation Report for Louisiana Power and Light's application for a license to operate Waterford Steam Electric Station, Unit 3 (Docket No. 50-382), located in St. Charles Parish, Louisiana, has been jointly prepared by the Office of Nuclear Reactor Regulation and the Region IV Office of the US Nuclear Regulatory Commission. This supplement provides the results to date of the staff's evaluation of approximately 350 allegations and concerns of poor construction practices at the Waterford 3 facility

Health and safety annual report 1988

International Nuclear Information System (INIS)

1989-01-01

This report on health and safety provides a review of the impact of the Comapny's activities on its workforce, the public and the environment. New sections include safety auditing, emergency planning and health and safety research. BNFL operates five sites in north west England and southern Scotland. The head office and Engineering Design Centre is at Risley, near Warrington. Fuel is manufactured at Springfields near Preston, uranium is enriched for modern nuclear power stations at Capenhurst near Chester and spent fuel is reprocessed at Sellafield. BNFL also operate Calder Hall (Sellafield) and Chapelcross (Scotland) power stations and a disposal site for low-level radioactive wastes at Drigg near Sellafield. Radiation sources and BNFL's radioactive discharge are first explained generally and then specifically for each BNFL site. Industrial and radiological safety within BNFL are described. (UK)

Safety Evaluation Report related to the operation of Nine Mile Point Nuclear Station, Unit No. 2 (Docket No. 50-410). Supplement No. 4

International Nuclear Information System (INIS)

1986-09-01

This report supplements the Safety Evaluation Report (NUREG-1047, February 1985) for the application filed by Niagara Mohawk Power Corporation, as applicant and co-owner, for a license to operate Nine Mile Point Nuclear Station, Unit 2 (Docket No. 50-410). It has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is located near Oswego, New York. Supplement 1 to the Safety Evaluation Report was published in June 1985, and contained the report from the Advisory Committee on Reactor Safeguards as well as the resolution of a number of outstanding issues from the Safety Evaluation Report. Supplement 2 was published in November 1985, and contained the resolution of a number of outstanding and confirmatory issues. Supplement 3 was published in July 1986, and contained the resolution of a number of outstanding and confirmatory items, one new confirmatory item, the evaluation of the Engineering Assurance Program, and evaluation of a number of exemption requests

Safety Evaluation Report related to the operation of Clinton Power Station, Unit No. 1 (Docket No. 50-461). Supplement No. 7

International Nuclear Information System (INIS)

1986-09-01

Supplement No. 7 to the Safety Evaluation Report on the application filed by Illinois Power Company, Soyland Power Cooperative, Inc., and Western Illinois Power Cooperative, Inc., as applicants and owners, for a license to operate the Clinton Power Station, Unit No. 1, has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is located in Harp Township, DeWitt County, Illinois. This supplement reports the status of items that have been resolved by the staff since Supplement No. 6 was issued

Safety evaluation report related to the operation of Clinton Power Station, Unit No. 1 (Docket No. 50-461). Supplement No. 5

International Nuclear Information System (INIS)

1986-01-01

Supplement No. 5 to the Safety Evaluation Report on the application filed by Illinois Power Company, Soyland Power Cooperative, Inc., and Western Illinois Power Cooperative, Inc. as applicants and owners, for a license to operate the Clinton Power Station, Unit No. 1, has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is located in Harp Township, DeWitt County, Illinois. This supplement reports the status of items that have been resolved by the staff since supplement No. 4 was issued

Safety Evaluation Report related to the operation of Clinton Power Station, Unit No. 1 (Docket No. 50-461). Supplement No. 4

International Nuclear Information System (INIS)

1985-02-01

Supplement No. 4 to the Safety Evaluation Report on the application filed by Illinois Power Company, Soyland Power Cooperative, Inc., and Western Illinois Power Cooperative, Inc., as applicants and owners, for a license to operate the Clinton Power Station, Unit No. 1, has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is located in Harp Township, DeWitt County, Illinois. This supplement reports the status of items that have been resolved by the staff since Supplement No. 3 was issued

Safety Evaluation Report related to the operation of Clinton Power Station, Unit No. 1 (Docket No. 50-461). Supplement No. 6

International Nuclear Information System (INIS)

1986-07-01

Supplement No. 6 to the Safety Evaluation Report on the application filed by Illinois Power Company, Soyland Power Cooperative, Inc., and Western Illinois Power Cooperative, Inc., as applicants and owners, for a license to operate the Clinton Power Station, Unit No. 1, has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is located in Harp Township, DeWitt County, Illinois. This supplement reports the status of items that have been resolved by the staff since Supplement No. 5 was issued

Ensuring radiation safety during construction of the facility ''Ukrytie'' and restoration of unit 3 of the Chernobyl nuclear power station

International Nuclear Information System (INIS)

Belovodsky, L.F.; Panfilov, A.P.

1997-01-01

On April 26, 1986, an accident at the fourth power unit of the Chernobyl NPS (ChNPS) destroyed the reactor core and part of the power unit building, whereby sizeable amounts of radioactive materials, stored in reactor at operation, were released into the environment, and there were also highly active fragments of fuel elements and pieces of graphite from reactor spread on ChNPS site near to safety block. Information on the accident at ChNPS, including its cause and consequences, was considered at special meeting, conducted by IAEA on August 25-29, 1986, in Vienna. In final report of International Advisory Group for Nuclear Safety (IAGNS), prepared by results of meeting activities, the main stages of the accident effects elimination (AEE) immediately on the station site according to the data, received before August 1, 1986, were discussed. In 1987-1990 the published materials on the later period of AEE, completed by building ''Ukrytie'' installation at the fourth power unit of ChNPS

Report by USSR survey mission of Nuclear Safety Commission

International Nuclear Information System (INIS)

1990-01-01

The USSR survey mission of Nuclear Safety Commission drew up and presents the report as follows. In relation to the accident in Chernobyl Nuclear Power Station in USSR, in order to investigate into the present status of the countermeasures for nuclear power safety in USSR and to exchange opinion, the USSR survey mission inspected nuclear power station facilities and visited the government organs, research institutes and others in USSR. The survey mission comprised 13 members, and went to Moscow, Kiev and two nuclear power station sites, from October 22 to November 1, 1989, for 11 days. At present in USSR, 49 nuclear power plants of about 35 GWe are in operation, and by 2000, the operation of more nuclear power plants of about 30 GWe is needed, but due to the change of social situation in USSR, its attainment seems to be difficult. The plan of nuclear power generation in USSR, the ensuring of safety in general, the recent countermeasures for improving safety, the effect of the accident in Chenobyl Nuclear Power Station on health and so on are reported. The detailed record of the visit to Zaporozhe and Chernobyl Nuclear Power Stations and 7 other research institutes and government organs is given. (K.I.)

Safety evaluation report related to the operation of Catawba Nuclear Station, Units 1 and 2 (Docket Nos. 50-413 and 50-414). Supplement 2

International Nuclear Information System (INIS)

1984-06-01

This report supplements the Safety Evaluation Report (NUREG-0954) and Supplement 1 with respect to the application filed by Duke Power Company, North Carolina Municipal Power Agency Number 1, North Carolina Membership Corporation, and Saluda River Electric Cooperative, Inc., as applicants and owners, for licenses to operate the Catawba Nuclear Station, Units 1 and 2 (Docket Nos., 50-413 and 50-414, respectively). The facility is located in York County, South Carolina, approximately 9.6 km (6 mi) north of Rock Hill and adjacent to Lake Wylie. This supplement provides more recent information regarding resolution or updating of some of the open and confirmatory issues and license conditions identified in the Safety Evaluation Report

Safety Evaluation Report related to the operation of Seabrook Station, Units 1 and 2 (Docket Nos. 50-443 and 50-444). Supplement No. 5

International Nuclear Information System (INIS)

1986-07-01

This report is Supplement No. 5 to the Safety Evaluation Report (NUREG-0896, March 1983) for the application filed by the Public Service Company of New Hampshire, et al., for licenses to operate Seabrook Station, Units 1 and 2 Docket Nos. STN 50-443 and STN 50-444). It has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission and provides recent information on open items identified in the SER. The facility is located in Seabrook, New Hampshire. Subject to favorable resolution of the items discussed in this report, the staff concludes that the facility can be operated by the applicant without endangering the health and safety of the public

Non-safety piping operability review case study -- Today and tomorrow

International Nuclear Information System (INIS)

Flensburg, W.C.; Adams, T.M.

1995-01-01

During a 1993 Outage at the Perry Nuclear Power Station, a condition report was issued which identified potential intersystem loss of water between the Emergency Closed Cooling Water System and the Nuclear Closed Cooling Water System during a design basis event. The review of this condition report indicated that if a SSE (safe shutdown earthquake) event were to occur during a design basis event components important to plant safety could potentially be adversely affected if non-seismic/non-safety portions of the Nuclear Closed Cooling Water System could not maintain pressure boundary integrity as a result of the seismic loadings. Presented in this paper are steps, criteria, and methodology used to demonstrate the seismic acceptability of the affected portion of the Nuclear Closed Cooling Water System Piping. Also discussed are the potential benefits and applicability of a recently developed EPRI non-safety, non-seismic operability procedure. This discussion includes the potential cost savings which could have arisen from application of this recently developed procedure

Monitoring of the operation of a nuclear power station with design problems in an importing country: The Almaraz power station

International Nuclear Information System (INIS)

Reig, J.

1984-01-01

The purpose of this paper is to describe the regulatory activities carried out in Spain as a result of the design problem occurring in the steam generators during operation of Unit I of the Almaraz nuclear power station. First, a brief introduction is given to the operating history and characteristics of Unit I of Almaraz. Particular attention is paid to the specific licences issued subsequent to commercial operation which place limitations on the operation of the station and to the operational incidents of which the Nuclear Safety Council (CSN) has been notified. Next, a description is provided of the safety evaluation carried out by the CSN. Three aspects merit particular attention: methodology, evaluation and conclusions. The methodology applied by an importing country is normally based on that of the country of origin of the design, so that the overall evaluation by the NRC has been considered sufficiently representative of aspects specific to the Almaraz power station. In this regard the importance of international collaboration is clearly seen as a principal instrument for performing the evaluation. In the evaluation a distinction is made between general and specific aspects and between inspection programmes and quality assurance requirements. In addition, the conclusions leading to the requirement of the imposition of additional limitations on the operating licence are stated. Apart from the safety evaluation carried out by the CSN, other regulatory activities have been performed over this two-year period. These activities, which include site inspections, audits of the principal supplier company, other independent calculations and so on, are described. Lastly, the paper refers to the lessons learned from the operation of the above-mentioned unit, which are immediately applicable to other Spanish nuclear power stations. (author)

47 CFR 80.1115 - Transmission of a distress alert by a station not itself in distress.

Science.gov (United States)

2010-10-01

... (CONTINUED) SAFETY AND SPECIAL RADIO SERVICES STATIONS IN THE MARITIME SERVICES Global Maritime Distress and Safety System (GMDSS) Operating Procedures for Distress and Safety Communications § 80.1115 Transmission...

Technical limits on performance reserves and life expectancy in nuclear power stations with light water reactors

International Nuclear Information System (INIS)

Wanner, R.; Brosi, S.; Duijvestijn, G.

1990-01-01

The safety margin (i.e. the difference between the loads equipment can take and those actually imposed on components) in a reactor pressure vessel is a major factor in the life expectancy of a nuclear power station. This safety margin is reduced considerably by reductions in the toughness of equipment caused by neutron irradiation and growth of cracks. Once the minimum safety margin is infringed, the nuclear power station is at the end of its working life. 13 figs., 11 refs

Managing nuclear safety at Point Lepreau

Energy Technology Data Exchange (ETDEWEB)

Paciga, J [New Brunswick Power, Point Lepreau NGS, PQ (Canada)

1997-12-01

Managing nuclear safety at Point Lepreau nuclear power plant is described, including technical issues (station aging, definition of the safe operating envelope, design configuration management, code validation, safety analysis and engineering standards); regulatory issues (action items, probabilistic safety assessment, event investigation, periodic safety review, prioritization of regulatory issues, cost benefit assessment); human performance issues (goals and measures, expectations and accountability, supervisory training, safety culture, configuration management, quality of operations and maintenance).

Managing nuclear safety at Point Lepreau

International Nuclear Information System (INIS)

Paciga, J.

1997-01-01

Managing nuclear safety at Point Lepreau nuclear power plant is described, including technical issues (station aging, definition of the safe operating envelope, design configuration management, code validation, safety analysis and engineering standards); regulatory issues (action items, probabilistic safety assessment, event investigation, periodic safety review, prioritization of regulatory issues, cost benefit assessment); human performance issues (goals and measures, expectations and accountability, supervisory training, safety culture, configuration management, quality of operations and maintenance)

Radiation safety practice at nuclear power stations and estimation of dose burdens to the USSR general public in the context of the country's nuclear power development plans

International Nuclear Information System (INIS)

Vorob'ev, E.I.; Il'in, L.A.; Turovskij, V.D.; Buldakov, L.A.; Lusev, N.G.; Pavlovskij, O.A.; Parkhomenko, G.M.

1983-01-01

The paper sets forth the main features of the State system of health protection for staff and the general public, and likewise the essentials of environmental protection. The principles of standardizing radiation factors are given for power station personnel and for the general public, together with the main provisions of the health Standards and Rules for radiation protection at present valid in the USSR. Data are quoted on the radiation situation at nuclear power stations and on the size of releases of radioactive aerosols and liquid effluents to the environment. The paper pays particular attention to analyses of the radiation situation in districts where nuclear power stations are situated and also to the type and scope of monitoring of radioactive environmental contamination. An analysis of the coefficients achieved with Soviet pressurized water (WWER), high-power channel-type (RBMK) and fast (BN) reactors currently in large-scale use shows that in terms both of release levels of radioactive substances and of the dose burdens to staff and general public these reactors are comparable with the best foreign nuclear power installations. Values actually measured and values calculated for the basic parameters of the radiation situation in areas of the USSR where nuclear power stations are situated confirm the safety of these facilities as regards the health of the general public and the extremely low levels of their effects on the environment. In conclusion, the paper quotes estimates of the collective effective dose equivalent to the USSR population expected to result from implementation of the country's nuclear power programme up to the year 2000. Radiation safety problems associated with nuclear power production which still require solution are enumerated. (author)

Evaluation of River Bend Station Unit 1 Technical Specifications

International Nuclear Information System (INIS)

Baxter, D.E.; Bruske, S.J.

1985-08-01

This document was prepared for the Nuclear Regulatory Commission (NRC) to assist them in determining whether the River Bend Station Unit 1 Technical Specifications (T/S), which govern plant systems configurations and operations, are in conformance with the requirements of the Final Safety Analysis Report (FSAR) as amended, and the requirements of the Safety Evaluation Report (SER) as supplemented. A comparative audit of the FSAR as amended, and the SER as supplemented was performed with the River Bend T/S. Several discrepancies were identified and subsequently resolved through discussions with the cognizant NRC reviewer, NRC staff reviewers and/or utility representatives. The River Bend Station Unit 1 T/S, to the extent reviewed, are in conformance with the FSAR and SER

Present status of maintenance in nuclear power stations

International Nuclear Information System (INIS)

Fueki, Kensuke

1982-01-01

For nuclear power stations, the improvement of the rate of facility utilization is indispensable. As its means, the shortening of regular inspection period, the adoption of long term cycle operation, the shortening of plant shutdown period by the improvement of the reliability of installed equipment and so on are conceivable. In this paper, the maintenance techniques in nuclear power stations which constitute the basis of reliability improvement are explained. In nuclear power stations, the use of nuclear fuel, accordingly the existence of radioactivity are the remarkable features. At the time of an accident, the nuclear reaction in a reactor must be stopped, and the excessive heat must be removed. The radioactivity in nuclear fuel must not be released outside. The regular inspection of once a year is provided by the law, and routine tests are performed during normal operation. The check-up by operators also is a part of the safety measures. For the early detection of abnormality, the diagnosis system is developed, and the maintenance techniques during operation are examined for being taken into the safety test and standard. The improvement of reliability is attempted by the redundancy of systems. The activity of quality assurance, the organization for the maintenance of nuclear power stations and maintenance works, and the measures to raise the rate of operation are reported. (Kako, I.)

AECB staff annual assessment of the Bruce A Nuclear Generating Station for the year 1996

International Nuclear Information System (INIS)

1997-06-01

The Atomic Energy Control Board is the independent federal agency that controls all nuclear activities in Canada. A major use of nuclear energy in Canada is electricity production. The AECB assesses every station's performance against legal requirements, including the conditions in the operating licence. Each station is inspected and all aspects of the station's operation and management is reviewed. This report is the AECB staff assessment of reactor safety at the Bruce Nuclear Generating Station A for 1996. Ontario Hydro operated Bruce A safely in 1996, maintaining the risk to workers and the public at an acceptably low level. Special safety system performance at Bruce A was adequate. Availability targets were all met. Improvement is needed to reduce the number of operating licence non-compliances

Safety Evaluation Report related to the operation of Comanche Peak Steam Electric Station, Units 1 and 2 (Docket Nos. 50-445 and 50-446). Supplement No. 7

International Nuclear Information System (INIS)

1985-01-01

Supplement 7 to the Safety Evaluation Report for the Texas Utilities Electric Company application for a license to operate Comanche Peak Steam Electric Station, Units 1 and 2 (Docket Nos. 50-445, 50-446), located in Somervell County, Texas, has been jointly prepared by the Office of Nuclear Reactor Regulation and the Comanche Peak Technical Review of the US Nuclera Regulatory Commission. This supplement provides the results of the staff's evaluation and resolution of approximately 80 technical concerns and allegations in the areas of Electric/Instrumentation and Test Programs regarding construction and plant readiness testing practices at the Comanche Peak facility. Issues raised during Atomic Safety and Licensing Board hearings will be dealt with in future supplements to the Safety Evaluation Report

Safety Evaluation Report related to the operation of Comanche Peak Steam Electric Station, Units 1 and 2 (Docket Nos. 50-445 and 50-446). Supplement No. 8

International Nuclear Information System (INIS)

1985-02-01

Supplement 8 to the Safety Evaluation Report for the Texas Utilities Electric Company application for a license to operate Comanche Peak Steam Electric Station, Units 1 and 2 (Docket Nos. 50-445, 50-446), located in Somervell County, Texas, has been jointly prepared by the Office of Nuclear Reactor Regulation and the Comanche Peak Technical Review Team of the US Nuclear Regulatory Commission. This Supplement provides the results of the staff's evaluation and resolution of approximately 80 technical concerns and allegations relating to civil and structural and miscellaneous issues regarding construction and plant readiness testing practices at the Comanche Peak facility. Issues raised during recent Atomic Safety and Licensing Board hearings will be dealt with in future supplements to the Safety Evaluation Report

Finite element analysis of 2-Station hip himulator

Science.gov (United States)

Fazli, M. I. M.; Yahya, A.; Shahrom, A.; Nawawi, S. W.; Zainudin, M. R.; Nazarudin, M. S.

2017-10-01

This paper presented the analysis of materials and design architecture of 2-station hip simulator. Hip simulator is a machine used to conduct the joint and wear test of hip prosthetic. In earlier work, the hip simulator was modified and some improvement were made by using SolidWorks software. The simulator consists of 3DOF which controlled by separate stepper motor and a static load that set up by manual method in each station. In this work, finite element analysis (FEA) of hip simulator was implemented to analyse the structure of the design and selected materials used for simulator component. The analysis is completed based on two categories which are safety factor and stress tests. Both design drawing and FEA was done using SolidWorks software. The study of the two categories is performed by applying the peak load up to 4000N on the main frame that is embedded with metal-on-metal hip prosthesis. From FEA, the value of safety factor and degree of stress formation are successfully obtained. All the components exceed the value of 2 for safety factor analysis while the degree of stress formation shows higher value compare to the yield strength of the material. With this results, it provides information regarding part of simulator which are susceptible to destruct. Besides, the results could be used for design improvement and certify the stability of the hip simulator in real application.

Algological studies on the site of the Fessenheim nuclear power station

International Nuclear Information System (INIS)

Pierre, J.F.

1980-01-01

Systematic study of the algal flora at five stations situated on both sides of the nuclear power station at Fessenheim (department of Haut-Rhin, France). The analysis of the diatomaceae populations in 1977 and 1978, i.e. before and after the start of the reactors, does not indicate, in the composition and abundance of algae, any modifications susceptible to be directly connected to the implantation of the nuclear power station [fr

Safety Evaluation Report related to the full-term operating license for Millstone Nuclear Power Station, Unit No. 1 (Docket No. 50-245)

International Nuclear Information System (INIS)

1985-10-01

The Safety Evaluation Report for the full-term operating license application filed by the Connecticut Light and Power Company, the Hartford Electric Light Company, Western Massachusetts Electric Company and the Millstone Point Company [(now known as Connecticut Light and Power Company (CL and P) and Western Massachusetts Electric Company (WMECO) having authority to possess Millstone-1, 2, and 3, and the Northeast Nuclear Energy Company (NNECO) as the responsible entity for operation of the facilities)] for Millstone Nuclear Power Station Unit 1 has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is located in the town of Waterford, Connecticut. Subject to favorable resolution of the items discussed in this report, the staff concludes that the facility can continue to be operated without endangering the health and safety of the public

Licensing safety critical software

International Nuclear Information System (INIS)

Archinoff, G.H.; Brown, R.A.

1990-01-01

Licensing difficulties with the shutdown system software at the Darlington Nuclear Generating Station contributed to delays in starting up the station. Even though the station has now been given approval by the Atomic Energy Control Board (AECB) to operate, the software issue has not disappeared - Ontario Hydro has been instructed by the AECB to redesign the software. This article attempts to explain why software based shutdown systems were chosen for Darlington, why there was so much difficulty licensing them, and what the implications are for other safety related software based applications

Bradwell Nuclear Power Station

International Nuclear Information System (INIS)

1987-01-01

When built, the Magnox reactors were expected to have operating lifetimes of 20-25 years. In order to satisfy the licensing authorities of their continued safety, long term safety reviews (LTSRs) are being carried out as the reactors reach 20 years of operation. This is the Nuclear Installations Inspectorate's (NII) summary report on Bradwell nuclear power station. The objectives of the LTSR are stated. A description of the plant is followed by an explanation of the statutory position on licensing. The responsibilities of the Central Electricity Generating Board (CEGB) and the NII are defined. From the examination of the CEGB's LTSR it is concluded that this generally confirms the validity of the existing safety case for present operation. However, some recommendations are made as to work required for reactor operation up to 1992. A summary of the NII findings is presented. This includes the reactor pressure circuit integrity, effects of ageing and in-service wear and radiation doses. (U.K.)

16 CFR Appendix I to Part 1402 - Recommended Outline for Instruction Booklet on “How To Safely Install Your CB Base Station Antenna”

Science.gov (United States)

2010-01-01

... on âHow To Safely Install Your CB Base Station Antennaâ I Appendix I to Part 1402 Commercial Practices CONSUMER PRODUCT SAFETY COMMISSION CONSUMER PRODUCT SAFETY ACT REGULATIONS CB BASE STATION... Outline for Instruction Booklet on “How To Safely Install Your CB Base Station Antenna” I. Required...

The AP1000R nuclear power plant innovative features for extended station blackout mitigation

International Nuclear Information System (INIS)

Vereb, F.; Winters, J.; Schulz, T.; Cummins, E.; Oriani, L.

2012-01-01

Station Blackout (SBO) is defined as 'a condition wherein a nuclear power plant sustains a loss of all offsite electric power system concurrent with turbine trip and unavailability of all onsite emergency alternating current (AC) power system. Station blackout does not include the loss of available AC power to buses fed by station batteries through inverters or by alternate AC sources as defined in this section, nor does it assume a concurrent single failure or design basis accident...' in accordance with Reference 1. In this paper, the innovative features of the AP1000 plant design are described with their operation in the scenario of an extended station blackout event. General operation of the passive safety systems are described as well as the unique features which allow the AP1000 plant to cope for at least 7 days during station blackout. Points of emphasis will include: - Passive safety system operation during SBO - 'Fail-safe' nature of key passive safety system valves; automatically places the valve in a conservatively safe alignment even in case of multiple failures in all power supply systems, including normal AC and battery backup - Passive Spent Fuel Pool cooling and makeup water supply during SBO - Robustness of AP1000 plant due to the location of key systems, structures and components required for Safe Shutdown - Diverse means of supplying makeup water to the Passive Containment Cooling System (PCS) and the Spent Fuel Pool (SFP) through use of an engineered, safety-related piping interface and portable equipment, as well as with permanently installed onsite ancillary equipment. (authors)

Virginia power nuclear power station engineer training program

International Nuclear Information System (INIS)

Williams, T.M.; Haberstroh-Timpano, S.

1987-01-01

In response to the Institute of Nuclear Power Operations (INPO) accreditation requirements for technical staff and manager, Virginia Power developed the Nuclear Power Station Engineer Training Programs (NPSETP). The NPSETP is directed toward enhancing the specific knowledge and skills of company engineers, especially newly hired engineers. The specific goals of the program are to promote safe and reliable plant operation by providing engineers and appropriate engineering technicians with (1) station-specific basic skills; (2) station-specific specialized skills in the areas of surveillance and test, plant engineering, nuclear safety, and in-service inspection. The training is designed to develop, maintain, and document through demonstration the required knowledge and skills of the engineers in the identified groups at North Anna and Surry Power Stations. The program responds to American National Standards Institute, INPO, and US Nuclear Regulatory Commission standards

The safety of Ontario's nuclear power reactors. A scientific and technical review. Ontario Hydro Submission to the Ontario Nuclear Safety Review

International Nuclear Information System (INIS)

1987-01-01

Ontario Hydro is responsible for the safety of its nuclear stations: safety analysis, design and construction, training of operators, operating practices, and maintenance procedures. The utility must demonstrate to the regulatory body and the public that it is capable of operating nuclear stations safely. the dedicated attention of management and workers alike has been given to the achievement of an excellent safety record. Safety begins with well understood corporate goals, objectives and policies, and the clear assignment of responsibilities to well-trained, competent people who have the relevant experience and the right information and equipment. A prime cause of both the Chernobyl and the Three Mile Island accidents was a breakdown in operational procedures and human factors. On the contrary, the pressure tube failure at Pickering unit 2 in 1983 was understood almost immediately by the operators, who took the correct steps to shut down the reactor. This success is related to well-designed control room information systems and good understanding of fundamentals by the operators. Increasingly, in the design of nuclear plant control and instrumentation systems and in training in Ontario Hydro, the well-being, capabilities and limitations of humans are being taken into account. This report describes the series of barriers between the radioactive material in the fuel and the series of barriers between the radioactive material in the fuel and the environment, and the stringent quality control and technical measures taken to make the likelihood of malfunctions very small. Defence in depth protection for the public is a feature of all Ontario Hydro nuclear stations. As safety-related systems are updated in new stations, improvements are in some cases being backfitted to older stations

Technical evaluation report TMI action - NUREG-0737 (II.D.1) relief and safety valve testing for Clinton Power Station Unit 1. (Docket No. 50-461)

International Nuclear Information System (INIS)

Burr, T.K.; Magleby, H.L.

1985-05-01

Light water reactors operators have experienced a number of occurrences of improper performance by safety and relief valves installed in their primary coolant systems. Because of this, the authors of NUREG-0578 (TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations) recommended that programs be developed and completed which would reevaluate the performance capabilities of BWR safety and relief valves. This report has examined the response of the Licensee for the Clinton Power Station, Unit 1 to the requirements of NUREG-0578 and subsequently NUREG-0737 and finds that the Licensee has provided an acceptable response, reconfirming that the General Design Criteria 14, 15 and 30 of Appendix A to 10 CFR-50 have been met

(Case Study: Wastewater Pump Station of Khoramabad

Directory of Open Access Journals (Sweden)

Masoud Taheriyoun

2014-10-01

Full Text Available Evaluation of a Biofilter System for Removal of Hydrogen Sulfide Gas (Case Study: Wastewater Pump Station of Khorramabad Abstract The biofilter system is one of the methods commonly used for the removal of hydrogen sulfide as the main source of odors emitted from wastewater facilities. The system is based on using the contaminant material as bedding to feed microorganisms. To achieve the desirable removal efficiency, it is, therefore, essential to create the proper conditions for the bacteria to grow on the bedding. In this study, a pilot-scale biofilter made of compost and woodchip (with a compost/woodchip ratio of 5:1 was used as the bedding material at Khorrmabad wastewater pumping station to investigate the performance of the system under real conditions. The experiment was carried out over 75 days during which time the input and output H2S concentrations were measured on a regular basis. Moisture was adjusted between 40% and 60% throughout the experiment to provide optimal conditions for bacterial growth. The results showed that the concentration of H2S emitted from the pumping station during 24 hours varied greatly between 0 and 48 PPM. The maximum adsorption capacity of the biological bedding was recorded at 2.874 g/m3.hr and the mean efficiency of H2S removal including the startup time was 89%. The mean performance efficiency during the biological activity after the startup was recorded at 98%.

Hinkley Point 'C' power station public inquiry: outline statement of case

International Nuclear Information System (INIS)

1988-05-01

This outline statement relates to the public inquiry to be held into the planning application by the Central Electricity Generating Board (CEGB) to construct a 1200 MW Pressurized Water Reactor (PWR) power station at Hinkley Point (Hinkley Point ''C'') in the United Kingdom, adjacent to an existing nuclear power station. The inquiry will consider economic, safety, environmental and planning matters relevant to the application and the implications for agriculture and local amenities of the re-aligning of two 400 kV overhead transmission lines. The outline statement contains submissions on: policy contest and approach; the requirement for Hinkley Point ''C''; design and safety; local issues. (UK)

Fuel Receiving and Storage Station. Nuclear Regulatory Commission's safety evaluation report

International Nuclear Information System (INIS)

1976-01-01

The safety evaluation report covers design of structures, components, equipment, and systems; nuclear criticality safety; radiological safety; accident analysis; conduct of operations; quality assurance; common defense and security; financial qualifications; financial protection and indemnity requirements; and technical specifications

Safety Evaluation Report related to the operation of Braidwood Station, Units 1 and 2 (Docket Nos. 50-456 and 50-457)

International Nuclear Information System (INIS)

1987-05-01

In November 1983, the staff of the Nuclear Regulatory Commission issued its Safety Evaluation Report (NUREG-1002) regarding the application filed by the Commonwealth Edison Company, as applicant and owner, for a license to operate Braidwood Station, Units 1 and 2 (Docket Nos. 50-456 and 50-457). The first supplement to NUREG-1002 was issued in September 1986; the second supplement to NUREG-1002 was issued in October 1986. This third supplement to NUREG-1002 reports the status of certain items that remained unresolved at the time Supplement 2 was published. The facility is located in Reed Township, Will County, Illinois

Final environmental assessment for vegetation control at VHF stations, microwave stations, electrical substations, and pole yards

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-10-13

Southwestern Power Adm. operates very high frequency (VHF) and microwave radio stations, electrical substations, and pole yards for electric power transmission throughout AR, MO, and OK. Vegetation growth at the stations must be suppressed for safety of operation and personnel. Southwestern has been using a combination of mechanical/manual and herbicide control for this purpose; Federally- mandated reductions in staff and budgetary resources require Southwestern to evaluate all potentially efficient methods for vegetation control. Three alternatives were examined: no action, mechanical/manual control, and (proposed) a combination of mechanical/manual and herbicide control. Environmental impacts on air and water quality, wetlands, wildlife, endangered species, archaeological and other resources, farmland, human health, transportation, etc. were evaluated.

AECB staff annual assessment of the Point Lepreau Nuclear Generating Station for the year 1995

International Nuclear Information System (INIS)

1996-06-01

The Atomic Energy Control Board conducts a staff assessment of safety at Point Lepreau Generating Station for 1995. On-site Project Officers and Ottawa based specialists monitored the station throughout the year. Point Lepreau operated safely during 1995 which was an unusual year for the station. The station was shut down for a planned five month outage to allow NB Power to complete maintenance on the reactor's pressure tubes. NB Power failed to comply with the terms of the Operating Licence we issue on fourteen occasions in 1995. In addition, NB Power reported an unusually large number of events. None of the events themselves directly affected public safety, if this level of performance continues unchecked, it might result in increased risk from operation in the future. Human error was an important feature of these problems. NB Power had already introduced safety culture training for their staff, but they will need to undertake further work urgently to resolve the problems. Three of Point Lepreau's special safety systems failed to meet their availability targets during 1995. NB Power's safety analysis program progress slowed during 1995, due to the additional workload from the outage. The Point Lepreau's training, quality assurance and safeguards related activities continued to function satisfactorily. They have assigned additional resources to emergency preparedness planning, which should enable them to take a more pro-active approach in this area. 7 tabs., 5 figs

47 CFR 80.1135 - Transmission of maritime safety information.

Science.gov (United States)

2010-10-01

... 47 Telecommunication 5 2010-10-01 2010-10-01 false Transmission of maritime safety information. 80... RADIO SERVICES STATIONS IN THE MARITIME SERVICES Global Maritime Distress and Safety System (GMDSS) Operating Procedures for Distress and Safety Communications § 80.1135 Transmission of maritime safety...

Determination of safety distance limits for a human near a cellular base station antenna, adopting the IEEE standard or ICNIRP guidelines.

Science.gov (United States)

Cooper, Justin; Marx, Bernd; Buhl, Johannes; Hombach, Volker

2002-09-01

This paper investigates the minimum distance for a human body in the near field of a cellular telephone base station antenna for which there is compliance with the IEEE or ICNIRP threshold values for radio frequency electromagnetic energy absorption in the human body. First, local maximum specific absorption rates (SARs), measured and averaged over volumes equivalent to 1 and to 10 g tissue within the trunk region of a physical, liquid filled shell phantom facing and irradiated by a typical GSM 900 base station antenna, were compared to corresponding calculated SAR values. The calculation used a homogeneous Visible Human body model in front of a simulated base station antenna of the same type. Both real and simulated base station antennas operated at 935 MHz. Antenna-body distances were between 1 and 65 cm. The agreement between measurements and calculations was excellent. This gave confidence in the subsequent calculated SAR values for the heterogeneous Visible Human model, for which each tissue was assigned the currently accepted values for permittivity and conductivity at 935 MHz. Calculated SAR values within the trunk of the body were found to be about double those for the homogeneous case. When the IEEE standard and the ICNIRP guidelines are both to be complied with, the local SAR averaged over 1 g tissue was found to be the determining parameter. Emitted power values from the antenna that produced the maximum SAR value over 1 g specified in the IEEE standard at the base station are less than those needed to reach the ICNIRP threshold specified for the local SAR averaged over 10 g. For the GSM base station antenna investigated here operating at 935 MHz with 40 W emitted power, the model indicates that the human body should not be closer to the antenna than 18 cm for controlled environment exposure, or about 95 cm for uncontrolled environment exposure. These safe distance limits are for SARs averaged over 1 g tissue. The corresponding safety distance limits

Aseismic foundation system for nuclear power stations

International Nuclear Information System (INIS)

Jolivet, F.; Richli, M.

1977-01-01

The aseismic foundation system, as described in this paper, is a new development, which makes it possible to build standard nuclear power stations in areas exposed to strong earthquakes. By adopting proven engineering concepts in design and construction of components, great advantages are achieved in the following areas: safety and reliability; efficiency; design schedule; cost. The need for an aseismic foundation system will arise more and more, as a large part of nuclear power station sites are located in highly seismic zones or must meet high intensity earthquake criteria due to the lack of historic data. (Auth.)

Environmental qualification test of electrical penetration for nuclear power stations

International Nuclear Information System (INIS)

Kooziro, Tetsuya; Nakagawa, Akitoshi; Toyoda, Shigeru; Uno, Shunpei

1979-01-01

Environmental qualification test was conducted according to IEEE Std. 323-1974 in order to evaluate the safety and reliability of electrical penetration of PWR type nuclear power station. Electrical penetration is the assemblies of electric cables attached to the containment vessel and penetrate through the vessel. Since it is a part of the vessel, it is deemed to be one of the primary safety equipments that are important for the safety and reliability of nuclear power stations. Environmental tests were conducted continuously as to heat cycle, vibration and LOCA with the full size specimens of bushing type, pigtail type and triaxial cable type and at the same time thermal life and irradiation tests were conducted on the insulation materials used, in order to obtain the comprehensive evaluation of their electrical and mechanical characteristics. As the result, they all satisfied the requirements for the circuits for actual use during and after various environmental qualification tests according to IEEE Std. 323. (author)

The safety related aspects of pressure components in nuclear power plants

International Nuclear Information System (INIS)

Lindackers, K.H.

1979-01-01

Over the last two years the safety philosophy for nuclear power plants in the Federal Republic of Germany has changed considerably, as everyone working in the field perceives. The original and appropriate philosophy of risk minimalisation through graduated safety barriers has been more and more replaced by the utopian goal of total prevention of any damage. The reasons for this development are discussed briefly especially regarding pressure components. The very numerous pressure components of a nuclear power station are not all of equal importance with respect to safety. Although considerable efforts have been made, it has not been possible, to date, to achieve an agreement between operators, manufacturers, licensing authorities, independent experts, and other specialists about the safety related classification of the manifold pressure bearing parts in nuclear power stations. The background of this extremely regrettable situation is explained. In the last part of the paper the author suggests a simple and clear safety philosophy for pressure components in nuclear power stations. This philosophy is orientated both on Safety Regulations of the Radiation Protection Decree ('Strahlenschutzverordnung') of the 13th October 1976 and on the Safety Criteria for Nuclear Power Stations from 21st October 1977. Only a simple, clear framework can make a contribution to the further improvement of the already exceptional safety of nuclear facilities and to the removal of obstacles in the licensing procedure which, taken as a whole, tie up skilled personnel to a senseless degree, involve considerable financial expenditure, and have no relevance for the safety of nuclear power plants. (orig.) [de

Decommissioning situation and research and development for the decommissioning of the commercial nuclear power station in Japan

International Nuclear Information System (INIS)

Yamamoto, Tatsumi.

1996-01-01

There are 48 commercial nuclear power stations in operation in Japan as of January 1, 1995, which supplies about 28% (2.2 x 10 8 MWh) of total annual electricity generation in FY 1992. Accordingly, as the nuclear power contributes so much in electricity generation, there is a growing concern in the public toward the safety on decommissioning nuclear power station. It is gravely important to secure the safety throughout the decommissioning. This paper discusses: the decommissioning situation in Japan; the Japanese national policy for decommissioning of commercial nuclear power stations; R and D for decommissioning of commercial nuclear power stations in Japan; and the present conditions of low-level radioactive wastes disposal in Japan

Space Station Initial Operational Concept (IOC) operations and safety view - Automation and robotics for Space Station

Science.gov (United States)

Bates, William V., Jr.

1989-01-01

The automation and robotics requirements for the Space Station Initial Operational Concept (IOC) are discussed. The amount of tasks to be performed by an eight-person crew, the need for an automated or directed fault analysis capability, and ground support requirements are considered. Issues important in determining the role of automation for the IOC are listed.

Safety evaluation report: related to the operation of Seabrook Station, Units 1 and 2, Docket Nos. 50-443 and 50-444, Public Service Company of New Hampshire, et al

International Nuclear Information System (INIS)

1983-03-01

The Safety Evaluation Report for the application filed by the Public Service Company of New Hampshire, et al., for licenses to operate Seabrook Station, Units 1 and 2 (Docket Nos. STN 50-443 and STN 50-444), has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is located in Seabrook, New Hampshire. Subject to favorable resolution of the items discussed in this report, the staff concludes that the facility can be operated by the applicant without endangering the health and safety of the public

Fluid Studies on the International Space Station

Science.gov (United States)

Motil, Brian J.

2016-01-01

Will discuss the recent activities on the international space station, including the adiabatic two phase flow, capillary flow and interfacial phenomena, and boiling and condensation. Will also give a historic introduction to Microgravity Studies at Glenn Research Center. Talk will be given to students and faculty at University of Louisville.

Evaluation of Shoreham Nuclear Power Station, Unit 1 technical specifications

International Nuclear Information System (INIS)

Baxter, D.E.; Bruske, S.J.

1985-08-01

This document was prepared for the Nuclear Regulatory Commission (NRC) to assist them in determining whether the Shoreham Nuclear Power Station Unit 1 Technical Specifications (T/S), which govern plant systems configurations and operations, are in conformance with the assumptions of the Final Safety Analysis Report (FSAR) as amended, and the requirements of the Safety Evaluation Report (SER) as supplemented. A comparative audit of the FSAR as amended, and the SER as supplemented was performed with the Shoreham T/S. Several discrepancies were identified and subsequently resolved through discussions with the cognizant NRC reviewer, NRC staff reviewers and/or utility representatives. The Shoreham Nuclear Power Station Unit 1 T/S, to the extent reviewed, are in conformance with the FSAR and SER

Evaluation of Waterford Steam Electric Station Unit 3 technical specifications

International Nuclear Information System (INIS)

Baxter, D.E.; Bruske, S.J.

1985-09-01

This document was prepared for the Nuclear Regulatory Commission (NRC) to assist them in determining whether the Waterford Steam Electric Station Unit 3 Technical Specifications (T/S), which govern plant systems configurations and operations, are in conformance with the requirements of the Final Safety Analysis Report (FSAR) as amended, and the requirements of the Safety Evaluation Report (SER) as supplemented. A comparative audit of the FSAR as amended, and the SER as supplemented was performed with the Waterford T/S. Several discrepancies were identified and subsequently resolved by the cognizant NRC reviewer. Pending completion of the resolutions noted in Part 3 of this report, the Waterford Steam Electric Station Unit 3 T/S, to the extent reviewed, are in conformance with the FSAR and SER

Safety-barrier diagrams as a tool for modelling safety of hydrogen applications

DEFF Research Database (Denmark)

Duijm, Nijs Jan; Markert, Frank

2009-01-01

Safety-barrier diagrams have proven to be a useful tool in documenting the safety measures taken to prevent incidents and accidents in process industry. Especially during the introduction of new hydrogen technologies or applications, as e.g. hydrogen refuelling stations, safety-barrier diagrams...... are considered a valuable supplement to other traditional risk analysis tools to support the communication with authorities and other stakeholders during the permitting process. Another advantage of safety-barrier diagrams is that they highlight the importance of functional and reliable safety barriers in any...... system and here is a direct focus on those barriers that need to be subject to safety management in terms of design and installation, operational use, inspection and monitoring, and maintenance. Safety-barrier diagrams support both quantitative and qualitative approaches. The paper will describe...

Analysis of fuel handling system for fuel bundle safety during station blackout in 500 MWe PHWR unit of India

International Nuclear Information System (INIS)

Madhuresh, R.; Nagarajan, R.; Jit, I.; Sanatkumar, A.

1996-01-01

Situations of Station Blackout (SBO) i.e. postulated concurrent unavailability of Class Ill and Class IV power, could arise for a long period, while on-power refuelling or other fuel handling operations are in progress with the hot irradiated fuel bundles being anywhere in the system from the Reactor Building to the Spent Fuel Storage Bay. The cooling provisions for these fuel bundles are diverse and specific to the various stages of fuel handling operations and are either on Class Ill or on Class II power with particular requirements of instrument air. Therefore, during SBO, due to the limited availability of Class II power and instrument air, it becomes difficult to maintain cooling to these fuel bundles. However, some minimal cooling is essential, to ensure the safety of the bundles. As discussed in the paper, safety of these fuel bundles in the system and/or for those lying in the liner tube region of the reactor end fitting is ensured, during SBO, by resorting to passive means like 'stay-put', 'gravity- fill', 'D 2 0- steaming' etc. for cooling the bundles. The paper also describes various consequences emanating from these cooling schemes. (author). 6 refs., 2 tabs., 8 figs

Risk Perception and Occupational Accidents: A Study of Gas Station Workers in Southern Brazil

OpenAIRE

Cezar-Vaz, Marta Regina; Rocha, Laurelize Pereira; Bonow, Clarice Alves; da Silva, Mara Regina Santos; Vaz, Joana Cezar; Cardoso, Letícia Silveira

2012-01-01

The present study aimed to identify the perceptions of gas station workers about physical, chemical, biological and physiological risk factors to which they are exposed in their work environment; identify types of occupational accidents involving gas station workers and; report the development of a socioenvironmental intervention as a tool for risk communication to gas station workers. A quantitative study was performed with 221 gas station workers in southern Brazil between October and Decem...

Artificial intelligence for Space Station automation: Crew safety, productivity, autonomy, augmented capability

Science.gov (United States)

Firschein, O.; Georgeff, M. P.; Park, W.; Cheeseman, P. C.; Geldberg, J.

1986-01-01

Artificial intelligence (AI) R&D projects for the successful and efficient operation of the Space Station are described. The book explores the most advanced AI-based technologies, reviews the results of concept design studies to determine required AI capabilities, details demonstrations that would indicate the existence of these capabilities, and develops an R&D plan leading to such demonstrations. Particular attention is given to teleoperation and robotics, sensors, expert systems, computers, planning, and man-machine interface.

The AP1000{sup R} nuclear power plant innovative features for extended station blackout mitigation

Energy Technology Data Exchange (ETDEWEB)

Vereb, F.; Winters, J.; Schulz, T.; Cummins, E.; Oriani, L. [Westinghouse Electric Company LLC, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

2012-07-01

Station Blackout (SBO) is defined as 'a condition wherein a nuclear power plant sustains a loss of all offsite electric power system concurrent with turbine trip and unavailability of all onsite emergency alternating current (AC) power system. Station blackout does not include the loss of available AC power to buses fed by station batteries through inverters or by alternate AC sources as defined in this section, nor does it assume a concurrent single failure or design basis accident...' in accordance with Reference 1. In this paper, the innovative features of the AP1000 plant design are described with their operation in the scenario of an extended station blackout event. General operation of the passive safety systems are described as well as the unique features which allow the AP1000 plant to cope for at least 7 days during station blackout. Points of emphasis will include: - Passive safety system operation during SBO - 'Fail-safe' nature of key passive safety system valves; automatically places the valve in a conservatively safe alignment even in case of multiple failures in all power supply systems, including normal AC and battery backup - Passive Spent Fuel Pool cooling and makeup water supply during SBO - Robustness of AP1000 plant due to the location of key systems, structures and components required for Safe Shutdown - Diverse means of supplying makeup water to the Passive Containment Cooling System (PCS) and the Spent Fuel Pool (SFP) through use of an engineered, safety-related piping interface and portable equipment, as well as with permanently installed onsite ancillary equipment. (authors)

The selection of field component reliability data for use in nuclear safety studies

International Nuclear Information System (INIS)

Coxson, B.A.; Tabaie, Mansour

1990-01-01

The paper reviews the user requirements for field component failure data in nuclear safety studies, and the capability of various data sources to satisfy these requirements. Aspects such as estimating the population of items exposed to failure, incompleteness, and under-reporting problems are discussed. The paper takes as an example the selection of component reliability data for use in the Pre-Operational Safety Report (POSR) for Sizewell 'B' Power Station, where field data has in many cases been derived from equipment other than that to be procured and operated on site. The paper concludes that the main quality sought in the available data sources for such studies is the ability to examine failure narratives in component reliability data systems for equipment performing comparable duties to the intended plant application. The main benefit brought about in the last decade is the interactive access to data systems which are adequately structured with regard to the equipment covered, and also provide a text-searching capability of quality-controlled event narratives. (author)

Space station contamination control study: Internal combustion, phase 1

Science.gov (United States)

Ruggeri, Robert T.

1987-01-01

Contamination inside Space Station modules was studied to determine the best methods of controlling contamination. The work was conducted in five tasks that identified existing contamination control requirements, analyzed contamination levels, developed outgassing specification for materials, wrote a contamination control plan, and evaluated current materials of offgassing tests used by NASA. It is concluded that current contamination control methods can be made to function on the Space Station for up to 1000 days, but that current methods are deficient for periods longer than about 1000 days.

Technical evaluation of the susceptibility of safety-related systems to flooding caused by the failure of non-category 1 systems for the Yankee Rowe Nuclear Power Station

International Nuclear Information System (INIS)

Epps, R.C.

1980-11-01

This report documents the technical evaluation of the Maine Yankee Atomic Power Station. The purpose of this evaluation was to determine whether the failure of any non-Class I (seismic) equipment could result in a condition, such as flooding, that might adversely affect the performance of the safety-related equipment required for the safe shutdown of the facility, or to mitigate the consequences of an accident. Criteria developed by the US Nuclear Regulatory Commission were used to evaluate the acceptability of the existing protection system as well as measures taken by Maine Yankee Atomic Power Company (MYAPC) to minimize the danger of flooding and to protect safety-related equipment

Improving safety culture through the health and safety organization: a case study.

Science.gov (United States)

Nielsen, Kent J

2014-02-01

International research indicates that internal health and safety organizations (HSO) and health and safety committees (HSC) do not have the intended impact on companies' safety performance. The aim of this case study at an industrial plant was to test whether the HSO can improve company safety culture by creating more and better safety-related interactions both within the HSO and between HSO members and the shop-floor. A quasi-experimental single case study design based on action research with both quantitative and qualitative measures was used. Based on baseline mapping of safety culture and the efficiency of the HSO three developmental processes were started aimed at the HSC, the whole HSO, and the safety representatives, respectively. Results at follow-up indicated a marked improvement in HSO performance, interaction patterns concerning safety, safety culture indicators, and a changed trend in injury rates. These improvements are interpreted as cultural change because an organizational double-loop learning process leading to modification of the basic assumptions could be identified. The study provides evidence that the HSO can improve company safety culture by focusing on safety-related interactions. © 2013. Published by Elsevier Ltd and National Safety Council.

47 CFR 80.1133 - Transmission of safety communications.

Science.gov (United States)

2010-10-01

... SERVICES STATIONS IN THE MARITIME SERVICES Global Maritime Distress and Safety System (GMDSS) Operating... safety traffic frequencies specified in § 80.1077, or via the maritime mobile satellite service or on... 47 Telecommunication 5 2010-10-01 2010-10-01 false Transmission of safety communications. 80.1133...

AECB staff annual assessment of the Bruce B Nuclear Generating Station for the year 1996

International Nuclear Information System (INIS)

1997-06-01

The Atomic Energy Control Board is the independent federal agency that controls all nuclear activities in Canada. A major use of nuclear energy in Canada is electricity production. The AECB assesses every station's performance against legal requirements, including the conditions in the operating licence. Each station is inspected and all aspects of the station's operation and management is reviewed. This report is the AECB staff assessment of reactor safety at the Bruce Nuclear Generating Station B for 1996. It was concluded that Ontario Hydro operated Bruce B safely in 1996. Although the Bruce B plant is safe,it was noted that the number of outages and the number of secondary and tertiary equipment failures during reactor unit upsets increased. Ontario Hydro needs to pay special attention to prevent such a decrease in the safety performance at Bruce B

ENSI’s synopsis concerning the enhancement of safety margins; Zusammenfassung des ENSI zur Erhöhung der Sicherheitsmargen

Energy Technology Data Exchange (ETDEWEB)

NONE

2015-06-15

This press release issued by the Swiss Federal Nuclear Safety Inspectorate ENSI discusses the results of its probabilistic safety analysis on the safety margins for Swiss nuclear power stations that was made after the Fukushima nuclear accident. External incidents such as earthquakes and flooding and their impact on the power stations are analysed by the ERSIM project concerning the enhancement of safety margins. Three shut-down scenarios are discussed. The safety margins with respect to earthquakes and flooding for all four nuclear power station sites are discussed in detail.

Uncertainty and sensitivity analysis for the simulation of a station blackout scenario in the Jules Horowitz Reactor

International Nuclear Information System (INIS)

Ghione, Alberto; Noel, Brigitte; Vinai, Paolo; Demazière, Christophe

2017-01-01

Highlights: • A station blackout scenario in the Jules Horowitz Reactor is analyzed using CATHARE. • Input and model uncertainties relevant to the transient, are considered. • A statistical methodology for the propagation of the uncertainties is applied. • No safety criteria are exceeded and sufficiently large safety margins are estimated. • The most influential uncertainties are determined with a sensitivity analysis. - Abstract: An uncertainty and sensitivity analysis for the simulation of a station blackout scenario in the Jules Horowitz Reactor (JHR) is presented. The JHR is a new material testing reactor under construction at CEA on the Cadarache site, France. The thermal-hydraulic system code CATHARE is applied to investigate the response of the reactor system to the scenario. The uncertainty and sensitivity study was based on a statistical methodology for code uncertainty propagation, and the ‘Uncertainty and Sensitivity’ platform URANIE was used. Accordingly, the input uncertainties relevant to the transient, were identified, quantified, and propagated to the code output. The results show that the safety criteria are not exceeded and sufficiently large safety margins exist. In addition, the most influential input uncertainties on the safety parameters were found by making use of a sensitivity analysis.

Safety Evaluation Report related to the operation of Byron Station, Units 1 and 2 (Docket Nos. STN 50-454 and STN 50-455). Supplement No. 7

International Nuclear Information System (INIS)

1986-11-01

Supplement No. 7 to the Safety Evaluation Report related to Commonwealth Edison Company's application for licenses to operate the Byron Station, Units 1 and 2, located in Rockvale Township, Ogle County, Illinois, has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. This supplement provides additional information supporting the license for initial criticality and power ascension to full-power operation for Unit 2

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

Children Road Safety Training with Augmented Reality (AR) [Demo

OpenAIRE

Lugmayr, Artur; Tsang, Joyce; Williams, Toby; Lim, Casey X; Teo, Yeet Yung; Farmer, Matthew

2018-01-01

Children killed or seriously injured through road accidents can be avoided through an appropriate safety training. Through play and engagement children learn and understand hazards at i.e. railway stations, bus stops, crossings, school zones, train stations, footpaths, or while cycling. We developed a rapid prototype of an Augmented Reality (AR) safety training proof-of-concept demonstrator for a scaled real-world model of dangerous road hazards. Two scenarios have been picked to give childre...

Potable Water Supply Feasibility Study for Summit Station, Greenland

Science.gov (United States)

2011-03-01

station, Dronning Maud Land, Antarctica. Final comprehensive environmental evaluation report. British Antarctic Survey. 2007. Proposed construction...Troll in Dronning Maud Land, Antarctica to a permanent station. PolarPower.org. 2010. Examples: Princess Elisabeth Antarctica Station

Experimental and Theoretical Dynamic Study of the Aagesta Nuclear Power Station

Energy Technology Data Exchange (ETDEWEB)

Bliselius, P A; Vollmer, H; Aakerhielm, F

1969-12-15

The report presents a final review and summary of all dynamic investigations of the Aagesta nuclear power station. Special attention is paid to the final and unpublished experiments performed in 1965-66. These experiments are discussed and compared to the theoretical predictions. Transfer functions and step responses were measured by perturbations in reactivity and steam load. Three methods were used for transfer function measurements: step functions, trapeze waves and multifrequency functions based on the pseudo-random binary sequence (PRBS). From the frequency analysis we found that the different perturbation methods led to quite consistent results. For the Agesta application the PRBS method is demonstrated to be an accurate and practical method for obtaining experimental transfer functions. The step technique did not give satisfactory results for frequencies above approximately 0.01 Hz. From the static check of the model we may conclude that the experiments agree fairly well mutually and with theory. The measured reactivity coefficients tend to be smaller than the predicted ones. The predicted transients compare well with the measured ones. However, there is a tendency to more peaked power transients from the model. This is believed to be due to the assumption that the moderator could be regarded as one channel. The dynamic experiments carried out at the Agesta nuclear power station demonstrate both the inherent stability and the safety of this plant.

47 CFR 80.1109 - Distress, urgency, and safety communications.

Science.gov (United States)

2010-10-01

... 47 Telecommunication 5 2010-10-01 2010-10-01 false Distress, urgency, and safety communications. 80.1109 Section 80.1109 Telecommunication FEDERAL COMMUNICATIONS COMMISSION (CONTINUED) SAFETY AND SPECIAL RADIO SERVICES STATIONS IN THE MARITIME SERVICES Global Maritime Distress and Safety System (GMDSS...

Space Station Environmental Health System water quality monitoring

Science.gov (United States)

Vincze, Johanna E.; Sauer, Richard L.

1990-01-01

One of the unique aspects of the Space Station is that it will be a totally encapsulated environment and the air and water supplies will be reclaimed for reuse. The Environmental Health System, a subsystem of CHeCS (Crew Health Care System), must monitor the air and water on board the Space Station Freedom to verify that the quality is adequate for crew safety. Specifically, the Water Quality Subsystem will analyze the potable and hygiene water supplies regularly for organic, inorganic, particulate, and microbial contamination. The equipment selected to perform these analyses will be commercially available instruments which will be converted for use on board the Space Station Freedom. Therefore, the commercial hardware will be analyzed to identify the gravity dependent functions and modified to eliminate them. The selection, analysis, and conversion of the off-the-shelf equipment for monitoring the Space Station reclaimed water creates a challenging project for the Water Quality engineers and scientists.

The approach associated with the continued operation of the Calder Hall and Chapelcross nuclear power stations to 50 years

International Nuclear Information System (INIS)

Ayres, G.

1996-01-01

Calder Hall was the world's first commercial nuclear power station, commencing operation in 1956, and with its sister station at Chapelcross has operated successfully, with consistently high load factors, for approximately 40 years. The first part of this paper reviews the operating history of the stations. Secondly, the paper will briefly describe both the work carried out under the Long Term Safety Review which has supported operation to 40 years and the work being carried out as part of a Periodic Safety Review to support continued operation of both stations to 50 years. The commercial improvements, some of which, of course, do have some nuclear safety significance, will be briefly described in the context of operating within what is increasingly becoming a demanding privatized electricity market in the United Kingdom. Finally, potential life limiting features will be identified and the monitoring programmes described leading to the conclusion that there is no reason why the stations should not continue to operate to at least 50 years. (author). 4 refs

Case study: beverage temperature at aid stations in ironman triathlon.

Science.gov (United States)

Burdon, Catriona A; Johnson, Nathan A; Chapman, Phillip G; Munir Che Muhamed, Ahmad; O'Connor, Helen T

2013-08-01

The aim of this study was to measure the effect of environmental conditions and aid-station beverage- cooling practices on the temperature of competitor beverages. Environmental and beverage temperatures were measured at three cycling and two run course aid stations at the 2010 Langkawi, Malaysia (MA), and Port Macquarie, Australia (AU), Ironman triathlon events. To measure the specific effect of radiant temperature, additional fluid-filled (600 ml) drink bottles (n = 12) were cooled overnight (C) and then placed in direct sun (n = 6) or shade (n = 6) near to a cycle aid station at AU. During both events, beverage temperature increased over time (p beverage temperature ranged between 14-26°C and during both events was above the palatable range (15-22°C) for extended periods. At AU, bottles placed in direct sunlight heated faster (6.9 ± 2.3 °C·h-1) than those in the shade (4.8 ±1.1°C·h-1, p = .05). Simple changes to Ironman aid-station practices, including shade and chilling beverages with ice, result in the provision of cooler beverages. Future studies should investigate whether provision of cool beverages at prolonged endurance events influences heat-illness incidence, beverage-consumption patterns, and competitor performance.

Tokai earthquakes and Hamaoka Nuclear Power Station

International Nuclear Information System (INIS)

Komura, Hiroo

1981-01-01

Kanto district and Shizuoka Prefecture are designated as ''Observation strengthening districts'', where the possibility of earthquake occurrence is high. Hamaoka Nuclear Power Station, Chubu Electric Power Co., Inc., is at the center of this district. Nuclear power stations are vulnerable to earthquakes, and if damages are caused by earthquakes in nuclear power plants, the most dreadful accidents may occur. The Chubu Electric Power Co. underestimates the possibility and scale of earthquakes and the estimate of damages, and has kept on talking that the rock bed of the power station site is strong, and there is not the fear of accidents. However the actual situation is totally different from this. The description about earthquakes and the rock bed in the application of the installation of No.3 plant was totally rewritten after two years safety examination, and the Ministry of International Trade and Industry approved the application in less than two weeks thereafter. The rock bed is geologically evaluated in this paper, and many doubtful points in the application are pointed out. In addition, there are eight active faults near the power station site. The aseismatic design of the Hamaoka Nuclear Power Station assumes the acceleration up to 400 gal, but it may not be enough. The Hamaoka Nuclear Power Station is intentionally neglected in the estimate of damages in Shizuoka Prefecture. (Kako, I.)

Improvement and optimization for in-service inspection of M310 nuclear power station

International Nuclear Information System (INIS)

Wang Chen; Sun Haitao; Gao Chen; Deng Dong

2015-01-01

In-service inspection (ISI) is an important method to ensure the safety of the mechanical equipment in nuclear power stations. According to the in-service inspection experience feedback from the domestic nuclear power stations, the reasonableness of some provisions in the RSE-M code are discussed and the applications of risk-informed in-service inspection (RI-ISI) are introduced, and the advices for the optimization of the ISI of the domestic M310 nuclear power stations are proposed. (authors)

Full scale reactor safety experiments performed in the Marviken Power Station Sweden

International Nuclear Information System (INIS)

Thoren, H.G.; Ericson, L.

1977-01-01

Since 1972 experiments oriented towards increasing the understanding of reactor safety processes have been performed at the Marviken Power Station. This was originally built as a direct cycle BHWR but was never taken into nuclear operation. In addition to Sweden, the countries represented in these experiments are Denmark, the Federal Republic of Germany, Finland, Norway, the United States, the Netherlands, France and Japan. The first series of sixteen experiments included studies of the response of the PS-containment to simulated ruptures in the pipe systems that are connected to the pressure vessel. These tests were completed in 1973 and also included experimental studies of iodine transport, containment leakage, the behaviour of auxiliary components under accident conditions and pressure fluctuations in the wetwell water pool. One of the more essential findings of the tests was that the containment performance was in accordance with the pre-test calculations. A second series of eight blowdown tests was begun in February 1976. The main purpose of these tests is to provide additional information as to the characteristics of the pressure oscillations inside the containment and primarily in the wetwell water pool under different conditions. These oscillations were observed in the first series of blowdowns but only low frequencies could then be detected due to limitations in the measurement system. The measurement system was therefore substantially extended for this second series of experiments. A summary of the results from these two sets of blowdown tests are given in the paper. In 1976 preparations for a new test program were initiated. The objective of these tests is to improve the understanding of critical flow in the low quality and subcooled flow regions through short length, large diameter pipes. Extensive modifications of the test facility will be necessary in order to allow a discharge flow through openings which are up to 500 mm in diameter. Advanced plans

Implementing an effective self-assessment program at Millstone Station

International Nuclear Information System (INIS)

Venable, J.M.

1996-01-01

The nuclear industry is becoming ever more reliant on self-assessments to ensure operational safety and to meet our increasingly competitive business challenges. This trend includes utility assessments modeled after major U.S. Nuclear Regulatory Commission (NRC) inspections such as safety system functional inspections (SSFI). Utility conducted SSFIs leveraged the limited resources of the NRC, making possible many evaluations that simply would not have been conducted otherwise. This report describes a self-assessment program at the Millstone Station plant

Noise test system of rotating machinery in nuclear power station based on microphone array

Science.gov (United States)

Chang, Xincai; Guan, Jishi; Qi, Liangcai

2017-12-01

Rotating machinery plays an important role in all walks of life. Once the equipment fails, equipment maintenance and shutdown will cause great social harm and economic losses. Equipment safety operations at nuclear power stations have always been of top priority. It is prone to noise when the equipment is out of order or aging. Failure to find or develop equipment at the initial stage of equipment failure or ageing will pose a serious threat to the safety of the plant’s equipment. In this paper, sound imaging diagnosis technology is applied as a supplementary method to the condition monitoring and diagnosis system of rotating machinery in nuclear power stations. It provides a powerful guarantee for the condition monitoring and fault diagnosis of rotating machinery in nuclear power stations.

AECB staff annual assessment of the Bruce A Nuclear Generating Station for the year 1994

International Nuclear Information System (INIS)

1995-06-01

AECB believe that Ontario Hydro operated Bruce A in a safe manner during 1994, and that the risk to workers and the public has been maintained at an acceptably low level. Radiation doses to workers and releases to the environment were well below regulatory limits. All special safety systems met availability targets. We noted improvements in operation and maintenance but some further improvements are still required. This is particularly true of the station's compliance with the Operating Licence. AECB believe that the station continues to be well managed, with a high priority placed on safety. However, there is a need for increased capability in the area of safety analysis and assessment. 4 tabs., 4 figs

Alternative off-site power supply improves nuclear power plant safety

International Nuclear Information System (INIS)

Gjorgiev, Blaže; Volkanovski, Andrija; Kančev, Duško; Čepin, Marko

2014-01-01

Highlights: • Additional power supply for mitigation of the station blackout event in NPP is used. • A hydro power plant is considered as an off-site alternative power supply. • An upgrade of the probabilistic safety assessment from its traditional use is made. • The obtained results show improvement of nuclear power plant safety. - Abstract: A reliable power system is important for safe operation of the nuclear power plants. The station blackout event is of great importance for nuclear power plant safety. This event is caused by the loss of all alternating current power supply to the safety and non-safety buses of the nuclear power plant. In this study an independent electrical connection between a pumped-storage hydro power plant and a nuclear power plant is assumed as a standpoint for safety and reliability analysis. The pumped-storage hydro power plant is considered as an alternative power supply. The connection with conventional accumulation type of hydro power plant is analysed in addition. The objective of this paper is to investigate the improvement of nuclear power plant safety resulting from the consideration of the alternative power supplies. The safety of the nuclear power plant is analysed through the core damage frequency, a risk measure assess by the probabilistic safety assessment. The presented method upgrades the probabilistic safety assessment from its common traditional use in sense that it considers non-plant sited systems. The obtained results show significant decrease of the core damage frequency, indicating improvement of nuclear safety if hydro power plant is introduced as an alternative off-site power source

Construction work management for nuclear power stations

International Nuclear Information System (INIS)

Yoshikawa, Yuichiro

1982-01-01

Nuclear power generation is positioned as the nucleus of petroleum substitution. In the Kansai Electric Power Co., efforts have been made constantly to operate its nuclear power plants in high stability and safety. At present, Kansai Electric Power Co. is constructing Units 3 and 4 in the Takahama Nuclear Power Station in Fukui Prefecture. Under the application of the management of construction works described here, both the nuclear power plants will start operation in 1985. The activities of Kansai Electric Power Co. in the area of this management are described: an outline of the construction works for nuclear power stations, the management of the construction works in nuclear power stations (the stages of design, manufacturing, installation and test operation, respectively), quality assurance activities for the construction works of nuclear power plants, important points in the construction work management (including the aspects of quality control). (J.P.N.)

Safety Evaluation Report related to the operation of Comanche Peak Steam Electric Station, Units 1 and 2. Docket Nos. 50-445 and 50-446

International Nuclear Information System (INIS)

1983-03-01

Supplement No. 3 to the Safety Evaluation Report (SER) related to the operation of the Comanche Peak Steam electric Station, Units 1 and 2, has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. the facility is located in Somervell County, Texas. Subject to favorable resolution of the items identified in this supplement, the staff concludes that the facility can be operated by the applicatn without endangering the health and safety of the public. This document provides the NRC staff's evaluation of the outstanding and confirmatory issues that have been resolved since Supplement No. 2 was issued in January 1982, and addresses changes to the SER and its earlier supplements which have resulted from the receipt of additonal information from the applicant during the period of January throught October 1982

Space Station flight telerobotic servicer functional requirements development

Science.gov (United States)

Oberright, John; Mccain, Harry; Whitman, Ruth I.

1987-01-01

The Space Station flight telerobotic servicer (FTS), a flight robotic system for use on the first Space Station launch, is described. The objectives of the FTS program include: (1) the provision of an alternative crew EVA by supporting the crew in assembly, maintenance, and servicing activities, and (2) the improvement of crew safety by performing hazardous tasks such as spacecraft refueling or thermal and power system maintenance. The NASA/NBS Standard Reference Model provides the generic, hierarchical, structured functional control definition for the system. It is capable of accommodating additional degrees of machine intelligence in the future.

Reactor safety

International Nuclear Information System (INIS)

Meneley, D.A.

The people of Ontario have begun to receive the benefits of a low cost, assured supply of electrical energy from CANDU nuclear stations. This indigenous energy source also has excellent safety characteristics. Safety has been one of the central themes of the CANDU development program from its very beginning. A great deal of work has been done to establish that public risks are small. However, safety design criteria are now undergoing extensive review, with a real prospect of more stringent requirements being applied in the future. Considering the newness of the technology it is not surprising that a consensus does not yet exist; this makes it imperative to discuss the issues. It is time to examine the policies and practice of reactor safety management in Canada to decide whether or not further restrictions are justified in the light of current knowledge

Computer simulation for risk management: Hydrogen refueling stations and water supply of a large region

DEFF Research Database (Denmark)

Markert, Frank; Kozine, Igor

2012-01-01

in applying DES models to the analysis of large infrastructures for refueling stations and water supply. Two case studies are described which are concerned with the inherently safer supply and storage of hydrogen at refueling stations and an established drinking water supply system of a large metropolitan...... area, respectively. For both, the simulation aims at identifying points of potential improvement from the reliability point of view. This allows setting up a list of activities and safety measures to reduce risk and possible losses and mitigate the consequences of accidents. Based on the cases...

Public opinion poll on safety and regulations of nuclear energy

Energy Technology Data Exchange (ETDEWEB)

Park, M. I.; Park, B. I.; Lee, S. M. [Gallup Korea, Seoul (Korea, Republic of)

2004-02-15

The purpose of this poll is not only to research understanding on safety and regulations of nuclear energy and to compare the result by time series followed 2003 to 2002 years, also to establish the public relations strategies and to offer information for developing long-term policies. The contents of the study are on the general perception, safety, management of nuclear power station, regulations and surroundings about nuclear energy.

Nuclear safety: an operational constraint or necessity

International Nuclear Information System (INIS)

Gauvenet, A.

1983-01-01

Different aspects of the nuclear safety in the operation of power stations are analysed. There is always a danger that safety is considered as a constraint at operator level, but it is essential that human factors and working conditions be taken into consideration [fr

TEPCO plans to construct Higashidori Nuclear Power Station

International Nuclear Information System (INIS)

Tsuruta, Atsushi

2008-01-01

In 2006, TEPCO submitted to the government plans for the construction of Higashidori Nuclear Power Station. The application was filed 41 years after the project approved by the Higashidori Village Assembly. This nuclear power station will be the first new nuclear power plant constructed by TEPCO since the construction of Units No.6 and 7 at the Kashiwazaki Kariwa Nuclear Power Station 18 years ago. Higashidori Nuclear Power Station is to be constructed at a completely new site, which will become the fourth TEPCO nuclear power station. Higashidori Nuclear Power Station Unit No.1 will be TEPCO's 18th nuclear reactor. Unit No.1 will be an advanced boiling water reactor (ABWR), a reactor-type with a proven track record. It will be TEPCO's third ABWR. Alongside incorporating the latest technology, in Higashidori Nuclear Power Station Unit No.1, the most important requirement is for TEPCO to reflect in the new unit information and experience acquired from the operation of other reactors (information and experience acquired through the experience of operating TEPCO's 17 units at Fukushima Daiichi Nuclear Power Station, Fukushima Daini Nuclear Power Station and Kashiwazaki Kashiwa Nuclear Power Station in addition to information on non-conformities at nuclear power stations in Japan and around the world). Higashidori Nuclear Power Station is located in Higashidori-Village (Aomori Prefecture) and the selected site includes a rich natural environment. From an environmental perspective, we will implement the construction with due consideration for the land and sea environment, aiming to ensure that the plant can co-exist with its natural surroundings. The construction plans are currently being reviewed by the Nuclear and Industrial Safety Agency. We are committed to making progress in the project for the start of construction and subsequent commercial operation. (author)

Operating experience and performance at Narora Atomic Power Station

International Nuclear Information System (INIS)

Mittal, Subhash; Gupta, J.P.

1998-01-01

Narora Atomic Power Station consists of two units of 220 MWe capacity each. These are Pressurized Heavy Water Reactors, fuelled by natural uranium, moderated and cooled by heavy water. The Station is owned by Nuclear Power Corporation of India Ltd., which is responsible for design, construction, commissioning, and operation of all nuclear power stations in the country. NAPS was the first opportunity to apply operating experiences in design, keeping in view the evolving safety and seismicity requirements, ease of maintenance, inservice inspection needs, improved construction ability and standardization. Both the units of NAPS are having improved safety standards of current international levels. All the equipment are indigenous with improved quality and reliability. The first unit of the station went critical in March 1989 and synchronized to the grid in July 1989. The second units followed with its criticality in October 1991 and synchronization in January 1992. Considering the initial stabilizing period, the performance of both units of NAPS has progressively improved over the years. The annual capacity factor for NAPS - 1 was 90.01% and for NAPS - 2 was 89.01% for the financial year 1997-1998. This paper presents an analysis of the performance during the last three years and measures taken to improve it. The stated enhanced performance could be achieved by improvement in human performance by training/re-training, scrupulous monitoring and review of equipment/systems, institution of adequate procedure and ensuring their adherence. (authors)

Reliability and safety of nuclear power stations

International Nuclear Information System (INIS)

Stepanek, S.

1979-01-01

The main problems are briefly discussed associated with the assessment of the safety and reliability of reactor pressure vessels. Two approaches are being applied to the assessment: one is based on the crack arrest temperature, the other on the determination of conditions corresponding to brittle fracture formation and on the determination of the critical defect size. The importance is stressed of continuous in-service inspection which may increase the factor of reliability by up to 10 4 times. (Z.M.)

Refurbishment and replacement efforts to mitigate ageing at Tarapur Atomic Power Station - an overview

International Nuclear Information System (INIS)

Katiyar, S.C.; Thattey, V.; Das, P.K.

2006-01-01

Tarapur Atomic Power Station (TAPS) - a twin Boiling Water Reactor unit and India's first Atomic Power Station was commissioned in April 1969, and was declared commercial in November 1969. Since then the light water moderated, low enriched uranium BWR with its demonstrated reliability and favourable economics is playing a vital role as a reliable source of power for the states of Maharashtra and Gujarat. The Power Station played a key role as a technology demonstrator validating the nuclear energy as safe and environmentally benign and economically viable alternate source of power generation in India. Built in the late sixties with state-of-the-art safety features prevailing then, TAPS has further evolved to be a safe plant with renovation and refurbishment efforts. Ageing Management Programme is in place at TAPS. Identification of systems, structures and components (SSCs) important to safety and availability, assessment of ageing degradation of these SSCs and mitigation through repair, replacement and refurbishment based on the investigations have enhanced the plant safety and reliability. The station's operating experience and feedback from BWRs operating abroad have also given inputs to Ageing Management Programme. A good number of major equipment have been replaced to mitigate ageing. Primary system piping, process heat exchangers, feed water heaters, turbine extraction system piping, turbine blades, emergency condenser tube bundles, various pumps, station batteries, electrical cables, circuit breakers etc. are some of them. Obsolescence is another aspect of ageing of a plant. Replacement of obsolete equipment and components particularly in C and I is another area where much headway has been made. Replacement and refurbishment of equipment have been done after detailed study and analysis so that current standards are met. Retrofitting the indigenously developed and fabricated equipment in a compact plant like TAPS was a difficult task and required lot of

Smoke flow temperature beneath tunnel ceiling for train fire at subway station: Reduced-scale experiments and correlations

International Nuclear Information System (INIS)

Meng, Na; Wang, Qiang; Liu, Zhaoxia; Li, Xiao; Yang, He

2017-01-01

Highlights: • Reduced-scale experiments on train fire at subway station. • Smoke flow temperature beneath tunnel ceiling measured and correlated. • Effect of platform-tunnel conjunction door type on smoke temperature is clarified. - Abstract: This paper is to investigate the smoke flow temperature beneath tunnel ceiling for a train on fire stopping besides a subway station. Experiments were carried out in a reduced-scale (1:10) subway station model to study the maximum smoke temperature and the longitudinal temperature distribution beneath the tunnel ceiling by considering platform-tunnel conjunction doors of two types: the full-seal platform screen door (PSD) and the full-height safety door. For the maximum temperature beneath the tunnel ceiling, it is found to be well correlated non-dimensionally with heat release rate by a 3.65 and a 2.92 power law function for the full-seal platform screen door and the full-height safety door, respectively. For the longitudinal temperature distribution along the tunnel ceiling, it can be well correlated by an exponential function for both types of platform-tunnel conjunction doors. Concerning the effect of the door type, the maximum temperature is lower and the longitudinal temperature decays faster for full-height safety door than that for full-seal PSD. This is due to that with the full-height safety door, the effective width of the tunnel ceiling is widened, which results in more heat losses from the smoke flow to the ceiling.

Darlington Station outage - a maintenance perspective

International Nuclear Information System (INIS)

Plourde, J.; Marczak, J.; Stone, M.; Myers, R.; Sutton, K.

1997-01-01

Ontario Hydro's Darlington Nuclear Generating Station (4x881MW(e)net) has carried out its first station outage since full commercial operation. The outage presented challenges to the organization in terms of outage planning, support, management, and safe execution within the constraints of schedule, budget and resources. This paper will focus on the success of the outage maintenance program, identifying the major work programs - a vacuum structure and containment outage, an emergency service water system outage, an emergency coolant injection system outage, intake channel inspections, low pressure service water inspections, and significant outage maintenance work on each of the four reactor units. Planning for the outage was initiated early in anticipation of this important milestone in the station's life. Detailed safety reviews - nuclear, radiation, and conventional - were conducted in support of the planned maintenance program. System lineup and work protection were provided by the Station Operator work group. Work protection permitry was initiated well in advance of the outage. Station maintenance staff resources were bolstered in support of the outage to ensure program execution could be maintained within the schedule. Training programs were in place to ensure that expectations were clear and that high standards would be maintained. Materials management issues in support of maintenance activities were given high priority to ensure no delays to the planned work. Station management review and monitoring in preparation for and during the outage ensured that staff priorities remained focused. Lessons learned from the outage execution are being formalized in maintenance procedures and outage management procedures, and shared with the nuclear community. (author)

Integrated plant safety assessment: systematic evaluation program. Oyster Creek nuclear generating station. GPU Nuclear Corporation and Jersey Central Power and Light Company. Docket No. 50-219

International Nuclear Information System (INIS)

1982-09-01

The Systematic Evaluation Program was initiated in February 1978 by the US Nuclear Regulatory Commission to review the designs of older operating nuclear reactor plants to reconfirm and document their safety. The review provides (1) an assessment of how these plants compare with current licensing safety requirements relating to selected issues, (2) a basis for deciding on how these differences should be resolved in an integrated plant review, and (3) a documented evaluation of plant safety. This report documents the review of the Oyster Creek Nuclear Generating Station (located in Ocean County, New Jersey), one of ten plants reviewed under Phase II of this program, and indicates how 137 topics selected for review under Phase I of the program were addressed. Equipment and procedural changes have been identified as a result of the review. It is expected that this report will be one of the bases in considering the issuance of a full-term operating license in place of the existing provisional operating license

Safety evaluation report related to the operation of Susquehanna Steam Electric Station, Units 1 and 2 (Docket Nos. 50-387 and 50-388). Suppl.6

International Nuclear Information System (INIS)

1984-03-01

In April 1981, the staff of the Nuclear Regulatory Commission issued its Safety Evaluation Report (NUREG-0776) regarding the application of the Pennsylvania Power and Light Company (the applicant and/or licensee) and the Allegheny Electric Cooperative, Inc. (co-applicant) for licenses to operate the Susquehanna Steam Electric Station, Units 1 and 2, located on a site in Luzerne County, Pennsylvania. This supplement to NUREG-0776 addresses the remaining issues that required resolution before licensing operation of Unit 2 and closes them out

Integral Safety Assessment of Underground Storage of CO2 in Barendrecht, the Netherlands

International Nuclear Information System (INIS)

Vijgen, L.; Nitert, M.; Buijtendijk, B.; Van Dalen, A.

2009-10-01

The DCMR Environmental Protection Agency Rijnmond in the Netherlands conducted an Integral Safety Assessment of Underground Storage of CO2 in Barendrecht, the Netherlands, in cooperation with the involved safety and supervision authorities. The following aspects of the entire storage project and its safety issues have been examined: the compressor station in Pernis; the underground pipes between the compressor station and the injection locations; and the injection locations Barendrecht-Ziedewij and Barendrecht. [nl

Transient Simulations in Hydropower Stations Based on a Novel Turbine Boundary

Directory of Open Access Journals (Sweden)

Yanna Liu

2016-01-01

Full Text Available Most accidents in hydropower stations happened during transient processes; thus, simulation of these processes is important for station design and safety operation. This study establishes a mathematical model of the transient process in hydropower stations and presents a new method to calculate the hydraulic turbine boundary based on an error function of the rotational speed. The mathematical derivation shows that the error function along the equal-opening characteristic curve is monotonic and has opposite signs at the two sides, which means that a unique solution exists to make the error function null. Thus, iteration of the transient simulation is unique and monotonous, which avoids iterative convergence or false solution and improves the solution efficiency compared with traditional methods. Simulation of an engineering case illustrates that the results obtained by the error function are reasonable. Then, the accuracy and feasibility of the mathematical model using the proposed solution are verified by comparison with model and field tests.

Space station automation and robotics study. Operator-systems interface

Science.gov (United States)

1984-01-01

This is the final report of a Space Station Automation and Robotics Planning Study, which was a joint project of the Boeing Aerospace Company, Boeing Commercial Airplane Company, and Boeing Computer Services Company. The study is in support of the Advanced Technology Advisory Committee established by NASA in accordance with a mandate by the U.S. Congress. Boeing support complements that provided to the NASA Contractor study team by four aerospace contractors, the Stanford Research Institute (SRI), and the California Space Institute. This study identifies automation and robotics (A&R) technologies that can be advanced by requirements levied by the Space Station Program. The methodology used in the study is to establish functional requirements for the operator system interface (OSI), establish the technologies needed to meet these requirements, and to forecast the availability of these technologies. The OSI would perform path planning, tracking and control, object recognition, fault detection and correction, and plan modifications in connection with extravehicular (EV) robot operations.

Development and testing of an objective structured clinical exam (OSCE) to assess socio-cultural dimensions of patient safety competency

Science.gov (United States)

Ginsburg, Liane R; Tregunno, Deborah; Norton, Peter G; Smee, Sydney; de Vries, Ingrid; Sebok, Stefanie S; VanDenKerkhof, Elizabeth G; Luctkar-Flude, Marian; Medves, Jennifer

2015-01-01

Background Patient safety (PS) receives limited attention in health professional curricula. We developed and pilot tested four Objective Structured Clinical Examination (OSCE) stations intended to reflect socio-cultural dimensions in the Canadian Patient Safety Institute's Safety Competency Framework. Setting and participants 18 third year undergraduate medical and nursing students at a Canadian University. Methods OSCE cases were developed by faculty with clinical and PS expertise with assistance from expert facilitators from the Medical Council of Canada. Stations reflect domains in the Safety Competency Framework (ie, managing safety risks, culture of safety, communication). Stations were assessed by two clinical faculty members. Inter-rater reliability was examined using weighted κ values. Additional aspects of reliability and OSCE performance are reported. Results Assessors exhibited excellent agreement (weighted κ scores ranged from 0.74 to 0.82 for the four OSCE stations). Learners’ scores varied across the four stations. Nursing students scored significantly lower (p<0.05) than medical students on three stations (nursing student mean scores=1.9, 1.9 and 2.7; medical student mean scores=2.8, 2.9 and 3.5 for stations 1, 2 and 3, respectively where 1=borderline unsatisfactory, 2=borderline satisfactory and 3=competence demonstrated). 7/18 students (39%) scored below ‘borderline satisfactory’ on one or more stations. Conclusions Results show (1) four OSCE stations evaluating socio-cultural dimensions of PS achieved variation in scores and (2) performance on this OSCE can be evaluated with high reliability, suggesting a single assessor per station would be sufficient. Differences between nursing and medical student performance are interesting; however, it is unclear what factors explain these differences. PMID:25398630

A feasibility assessment of nuclear reactor power system concepts for the NASA Growth Space Station

Science.gov (United States)

Bloomfield, H. S.; Heller, J. A.

1986-01-01

A preliminary feasibility assessment of the integration of reactor power system concepts with a projected growth Space Station architecture was conducted to address a variety of installation, operational, disposition and safety issues. A previous NASA sponsored study, which showed the advantages of Space Station - attached concepts, served as the basis for this study. A study methodology was defined and implemented to assess compatible combinations of reactor power installation concepts, disposal destinations, and propulsion methods. Three installation concepts that met a set of integration criteria were characterized from a configuration and operational viewpoint, with end-of-life disposal mass identified. Disposal destinations that met current aerospace nuclear safety criteria were identified and characterized from an operational and energy requirements viewpoint, with delta-V energy requirement as a key parameter. Chemical propulsion methods that met current and near-term application criteria were identified and payload mass and delta-V capabilities were characterized. These capabilities were matched against concept disposal mass and destination delta-V requirements to provide a feasibility of each combination.

A feasibility assessment of nuclear reactor power system concepts for the NASA growth Space Station

International Nuclear Information System (INIS)

Bloomfield, H.S.; Heller, J.A.

1986-01-01

A preliminary feasibility assessment of the integration of reactor power system concepts with a projected growth space station architecture was conducted to address a variety of installation, operational, disposition and safety issues. A previous NASA sponsored study, which showed the advantages of space station related concepts, served as the basis for this study. A study methodology was defined and implemented to assess compatible combinations of reactor power installation concepts, disposal destinations, and propulsion methods. Three installation concepts that met a set of integration criteria were characterized from a configuration and operational viewpoint, with end-of-life disposal mass identified. Disposal destinations that met current aerospace nuclear safety criteria were identified and characterized from an operational and energy requirements viewpoint, with delta-V energy requirement as a key parameter. Chemical propulsion methods that met current and near-term application criteria were identified and payload mass and delta-V capabilities were characterized. These capabilities were matched against concept disposal mass and destination delta-V requirements to provide a feasibility of each combination

Optimization of station battery replacement

International Nuclear Information System (INIS)

Jancauskas, J.R.; Shook, D.A.

1994-01-01

During a loss of ac power at a nuclear generating station (including diesel generators), batteries provide the source of power which is required to operate safety-related components. Because traditional lead-acid batteries have a qualified life of 20 years, the batteries must be replaced a minimum of once during a station's lifetime, twice if license extension is pursued, and more often depending on actual in-service dates and the results of surveillance tests. Replacement of batteries often occurs prior to 20 years as a result of systems changes caused by factors such as Station Blackout Regulations, control system upgrades, incremental load growth, and changes in the operating times of existing equipment. Many of these replacement decisions are based on the predictive capabilities of manual design basis calculations. The inherent conservatism of manual calculations may result in battery replacements occurring before actually required. Computerized analysis of batteries can aid in optimizing the timing of replacements as well as in interpreting service test data. Computerized analysis also provides large benefits in maintaining the as-configured load profile and corresponding design margins, while also providing the capability of quickly analyze proposed modifications and response to internal and external audits

Persuasion to use personal protective equipment in constructing subway stations: application of social marketing.

Science.gov (United States)

Shamsi, Mahmoud; Pariani, Abbas; Shams, Mohsen; Soleymani-nejad, Marzieh

2016-04-01

To study the effects of an intervention based on social marketing to persuade workers to use personal protective equipment (PPE) in constructing subway stations in Isfahan, Iran. This was a quasi-experimental study. Two stations were selected as intervention and control groups. Intervention was designed based on results of a formative research. A free package containing a safety helmet with a tailored message affixed to it, mask and gloves and an educational pamphlet was delivered to the intervention group. After 6 weeks, behaviours in the intervention and control stations were measured using an observational checklist. After the intervention, the percentage of workers who used PPE at the intervention station increased significantly. OR for helmet and mask usage was 7.009 and 2.235, respectively, in the intervention group. Social marketing can be used to persuade workers to use PPE in the workplace. Published by the BMJ Publishing Group Limited. For permission to use (where not already granted under a licence) please go to http://www.bmj.com/company/products-services/rights-and-licensing/

Analysis of fuel handling system for fuel bundle safety during station blackout in 500 MWe PHWR unit of India

Energy Technology Data Exchange (ETDEWEB)

Madhuresh, R; Nagarajan, R; Jit, I; Sanatkumar, A [Nuclear Power Corporation of India Ltd., Mumbai (India)

1997-12-31

Situations of Station Blackout (SBO) i.e. postulated concurrent unavailability of Class Ill and Class IV power, could arise for a long period, while on-power refuelling or other fuel handling operations are in progress with the hot irradiated fuel bundles being anywhere in the system from the Reactor Building to the Spent Fuel Storage Bay. The cooling provisions for these fuel bundles are diverse and specific to the various stages of fuel handling operations and are either on Class Ill or on Class II power with particular requirements of instrument air. Therefore, during SBO, due to the limited availability of Class II power and instrument air, it becomes difficult to maintain cooling to these fuel bundles. However, some minimal cooling is essential, to ensure the safety of the bundles. As discussed in the paper, safety of these fuel bundles in the system and/or for those lying in the liner tube region of the reactor end fitting is ensured, during SBO, by resorting to passive means like `stay-put`, `gravity- fill`, `D{sub 2}0- steaming` etc. for cooling the bundles. The paper also describes various consequences emanating from these cooling schemes. (author). 6 refs., 2 tabs., 8 figs.

AECB staff annual assessment of the Darlington Nuclear Generating Station for the year 1995

International Nuclear Information System (INIS)

1996-06-01

This report is the Atomic ENergy Control Board staff assessment of safety at the Darlington Nuclear Generating Station for 1995. The report is based on observations made by our staff, and on information submitted to us by Ontario Hydro. Performance was satisfactory for all four special safety systems. In 1995, Ontario Hydro complied with the regulations made under the Atomic Energy Control Act, except for two instances of non-compliance with the Transport Packaging of Radioactive Materials Regulations. Radiation doses received by Ontario Hydro station staff were below the regulatory limits. In general Ontario Hydro's maintenance program was found satisfactory. 9 tabs

Safety Evaluation Report related to the operation of Braidwood Station, Units 1 and 2 (Docket Nos. 50-456 and 50-457). Supplement No. 2

International Nuclear Information System (INIS)

1986-10-01

In November 1983, the staff of the Nuclear Regulatory Commission issued its Safety Evaluation Report (NUREG-1002) regarding the application filed by the Commonwealth Edison Company, as applicant and owner, for a license to operate Braidwood Station, Units 1 and 2 (Docket Nos. 50-456 and 50-457). The first supplement to NUREG-1002 was issued in September 1986. This second supplement to NUREG-1002 reports the status of certain items that remained unresolved at the time Supplement 1 was published. The facility is located in Reed Township, Will County, Illinois

Technical Specifications, Seabrook Station, Unit 1 (Docket No. 50-443). Appendix ''A'' to License No. NPF-56

International Nuclear Information System (INIS)

1986-10-01

This report provides specifications for the Seabrook Station Unit 1 reactor concerning: safety limits and limiting safety settings; limiting conditions for operation and surveillance requirements; design features; and administrative controls

Socio-economic impacts of nuclear generating stations: summary report on the NRC post-licensing studies

International Nuclear Information System (INIS)

Chalmers, J.; Pijawka, D.; Branch, K.; Bergmann, P.; Flynn, J.; Flynn, C.

1982-07-01

Information is presented concerning the conceptual framework for the assessment of socioeconomic impacts; methodology for the post-licensing case studies; socioeconomic changes due to the construction and operation of nuclear generating stations; public response to the construction and operation of nuclear generating stations; socioeconomic consequences of the accident at Three Mile Island; the significance of socioeconomic change due to the construction and operation of nuclear generating stations; findings of the post-licensing studies relative to the nuclear station impact literature; and implications of the findings for projective assessments and planning studies

Enhancement of safety for reprocessing facilities

International Nuclear Information System (INIS)

2012-06-01

The adequacy of the safety measures for utility loss accidents in nuclear fuel reprocessing facilities which have been formulated by the nuclear enterprises is investigated in JNES which organizes an advanced committee to specifically study this problem. The results are reviewed in the present report including the case of such severe accidents as in Fukushima Daiichi Nuclear Power Plant. The report also represents a tentative proposal for examination standards of such unimaginable severe accidents as 'station blackout,' urgent safety measures necessary for reoperation of nuclear power plants and requested by nuclear and industrial safety agency, and pointing out and clarification of the potential weakness from the safety point of view, and collective and composite evaluation of safety of the relevant facilities. Furthermore, the definition of accident management is given as of controlled condition and the authorized way of thinking for the cases of plural events happening at the same time and the cases when risks exist radioactivity emits with explosion. (S. Ohno)

Esfas: An information system on worldwide nuclear power stations

International Nuclear Information System (INIS)

Melis, M.

1990-01-01

While performing the analysis and transcoding of about 30.000 abnormal events happened in nuclear power stations, in the frame of the AORS - Abnormal Occurrences Reporting System project (CEC-Joint Research Centre, Ispra), it was clear to the transcoders (12 nuclear engineers) that, for a good understanding of the true sequence and safety relevance of events, it was necessary to identify the plant layout and the characteristics of safety and auxiliary systems. This exigence, together with the systematic collection of publicly available information (safety reports, utilities descriptions of plants, etc.) was the starting point for the development of ESFAS - Engineered Safety Features and Auxiliary Systems data base, conceived as a tool for the various phases of nuclear plant/system design, up to operational data analysis

Safety-Evaluation Report related to the operation of Grand Gulf Nuclear Station, Units 1 and 2. Docket Nos. 50-416 and 50-417

International Nuclear Information System (INIS)

1983-05-01

Supplement 4 to the Safety Evaluation Report for Mississippi Power and Light Company, et. al., joint application for licenses to operate the Grand Gulf Nuclear Station, Units 1 and 2, located on the east bank of the Mississippi River near Port Gibson in Claiborne County, Mississippi, has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. This supplement reports the status on the resolution of those issues that required further evaluation before authorizing operation of Unit 1 above 5% rated power and other issues that were to be evaluated during the first cycle of power operation

Preventing violence : service station employer handbook

Energy Technology Data Exchange (ETDEWEB)

NONE

2007-07-01

According to part 3 of British Columbia's Workers Compensation Act, employers must ensure the health and safety of their employees and any other workers present at their worksite. Workers are also responsible for following established safe work procedures and protecting their own health and safety. This handbook was designed for service station employers who do not already have adequate violence-prevention procedures. In addition to providing guidelines, it describes employment standards for workplace health and safety. It describes general duties of supervisors, owners and suppliers and includes the forms needed to fill out, notably an inspection list; an action plan; a violent incident report for workers to fill out in the event of a violent incident; a suspect and vehicle identification form; an employer incident investigation report; and a safety and security feedback report for workers. Regulations that relate to young and new worker orientation and training were also provided along with regulations for working alone or in isolation, violence in the workplace, and high-visibility apparel.

Pickering seismic safety margin

International Nuclear Information System (INIS)

Ghobarah, A.; Heidebrecht, A.C.; Tso, W.K.

1992-06-01

A study was conducted to recommend a methodology for the seismic safety margin review of existing Canadian CANDU nuclear generating stations such as Pickering A. The purpose of the seismic safety margin review is to determine whether the nuclear plant has sufficient seismic safety margin over its design basis to assure plant safety. In this review process, it is possible to identify the weak links which might limit the seismic performance of critical structures, systems and components. The proposed methodology is a modification the EPRI (Electric Power Research Institute) approach. The methodology includes: the characterization of the site margin earthquake, the definition of the performance criteria for the elements of a success path, and the determination of the seismic withstand capacity. It is proposed that the margin earthquake be established on the basis of using historical records and the regional seismo-tectonic and site specific evaluations. The ability of the components and systems to withstand the margin earthquake is determined by database comparisons, inspection, analysis or testing. An implementation plan for the application of the methodology to the Pickering A NGS is prepared

Vacuum switchgear for power station auxiliary switchboards

International Nuclear Information System (INIS)

Coombs, P.E.

1992-01-01

Sizewell B is the first UK power station in which vacuum switchgear is used for the auxiliary switchboards. Previously the 3.3kV, 6.6kV or 11kV switchgear has used air-break circuit breakers and fused air-break contactors, known as motor starting devices or fused switching devices (FSD). The use of vacuum interrupters is therefore a new technology in this application, although it has been established in the UK distribution network and in industrial installations from the mid 1970s. Vacuum switchgear was already in use in the USA for power station auxiliary switchgear at the time that it was proposed for Sizewell B. The Sizewell B high voltage auxiliary switchgear comprises eight Unit and Station Auxiliary Switchboards at 3.3kV and 11kV, and four 3.3kV Essential Switchboards for the essential safety related circuits, making a total of 65 circuit breakers plus FSD panels. (Author)

Conditions of radiation protection in medical stations

International Nuclear Information System (INIS)

Sa, L.R.B.S.; Tomaz Neto, A.; Pires, A.; Azevedo, H.; Boasquevisque, E.M.

1988-01-01

The purpose of this study is to clear up what safety procedures are normally' observed for occupational and environmental radiology. 30 Public Medical station in Rio de Janeiro were investigated. A questionaire of 13 questions was prepared to be filled up by the professionals directly involved with the radiologic work, intending to evaluate, the personal and environmental aspect of radioprotection, the individual responsability of each worker and of the whole institution. It was also verified that knowledge of safety norms is doubtful and precarious in the despite of the fact that a great number of the people in question declare to have specific graduation for the activity. Only 45% from the total really make use of the dosimeters, the periodical medical examinations are not frequent (65%), and fewer employes make use of this lead apron (23%). We come to the conclusion that there is a remarkable bewilderment as for the personal observences about the work conditions in controlled areas. (author) [pt

Plant computer system in nuclear power station

International Nuclear Information System (INIS)

Kato, Shinji; Fukuchi, Hiroshi

1991-01-01

In nuclear power stations, centrally concentrated monitoring system has been adopted, and in central control rooms, large quantity of information and operational equipments concentrate, therefore, those become the important place of communication between plants and operators. Further recently, due to the increase of the unit capacity, the strengthening of safety, the problems of man-machine interface and so on, it has become important to concentrate information, to automate machinery and equipment and to simplify them for improving the operational environment, reliability and so on. On the relation of nuclear power stations and computer system, to which attention has been paid recently as the man-machine interface, the example in Tsuruga Power Station, Japan Atomic Power Co. is shown. No.2 plant in the Tsuruga Power Station is a PWR plant with 1160 MWe output, which is a home built standardized plant, accordingly the computer system adopted here is explained. The fundamental concept of the central control board, the process computer system, the design policy, basic system configuration, reliability and maintenance, CRT display, and the computer system for No.1 BWR 357 MW plant are reported. (K.I.)

Safety evaluation report related to the operation of Catawba Nuclear Station, Units 1 and 2. Docket Nos. 50-413 and 50-414. Suppl. 1

International Nuclear Information System (INIS)

1983-04-01

This reort supplements the Safety Evaluation Report (NUREG-0954) issued in February 1983 by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission with respect to the application filed by Duke Power Company, North Carolina Municipal Power Agency Number 1, North Carolina Membership Corporation, and Saluda River Electric Cooperative, Inc. as applicants and owners, for licenses to operate the Catawba Nuclear Station, Units 1 and 2 (Docket Nos. 50-413 and 50-414, respectively). The facility is located in York County, South Carolina, approximately 9.6 km (6 mi) north of Rock Hill and adjacent to Lake Wylie. This supplement provides more recent information regarding resolution or updating of some of the open and confirmatory items and license conditions identified in the Safety Evaluation Report, and discusses the recommendations of the Advisory Committee on Reactor Safeguards in its report dated March 15, 1983

Lifetime management of Magnox power stations

International Nuclear Information System (INIS)

Smitton, C.

1998-01-01

Magnox Electric, which is, a subsidiary of BNFL, operates six nuclear power plants that have an average age of about 33 years. The procedures developed to maintain the plants and ensure nuclear safety in longer-term operation are reviewed. The technical limit on station lifetimes is expected to be determined by the effect of ageing on major reactor structures where replacement is impractical. Examination of the effect of ageing confirms that the stations are capable of operating to a life of at least 40 years. The economic factors affecting operation are reviewed, recognising the need to sell electricity in a competitive market. Recently Magnox Electric and BNFL have merged and all plant supporting Magnox operations are now within a single integrated company that will provide further opportunities for improved efficiency. (author)

Nordic studies in reactor safety

International Nuclear Information System (INIS)

Pershagen, N.

1993-01-01

The Nordic Nuclear Safety Research Programme SIK programme in reactor safety is part of a major joint Nordic research effort in nuclear safety. The report summarizes the achievements of the SIK programme, which was carried out during 1990-1993 in collaboration between Nordic nuclear utilities, safety authorities, and research institutes. Three main projects were successfully completed dealing with: 1) development and application of a living PSA concept for monitoring the risk of core damage, and of safety indicators for early warning of possible safety problems; 2) review and intercomparison of severe accident codes, case studies of potential core melt accidents in nordic reactors, development of chemical models for the MAAP code, and outline of a system for computerized accident management support; 3) compilation of information about design and safety features of neighbouring reactors in Germany, Lithuania and Russia, and for naval reactors and nuclear submarines. The report reviews the state-of-the-art in each subject matter as an introduction to the individual project summaries. The main findings of each project are highlighted. The report also contains an overview of reactor safety research in the Nordic countries and a summary of fundamental reactor safety principles. (au) (69 refs.)

Training of power station staff

International Nuclear Information System (INIS)

Dusserre, J.

1993-01-01

ELECTRICITE DE FRANCE currently operates 51 generating stations with 900 and 1300 MW Pressurized Water Reactors while, only 15 years ago, France possessed only a very small number of such stations. It was therefore vital to set up a major training organization to produce staff capable of starting, controlling and maintaining these facilities with a constant eye to improving quality and safety. Operator and maintenance staff training is based on highly-structured training plans designed to match both the post to be filled and the qualifications possessed by the person who is to fill it. It was essential to set up suitable high-performance training resources to handle this fast growth in staff. These resources are constantly being developed and allow EDF to make steady progress in a large number of areas, varying from the effects of human factors to the procedures to be followed during an accident

The opinion of the IRSN on the safety and radiation protection of the French electronuclear fleet in 2011

International Nuclear Information System (INIS)

2013-01-01

In its first part, this annual report proposes a global assessment of safety and radiation protection in the exploited electronuclear fleet (trends noticed in 2011 in the field of safety and radiation protection). The second part discusses the various events, incidents and anomalies which occurred in 2011: anomalies in studies for safety demonstration, rate unbalance between safety injection lines in 900 MWe reactors, defects in a penetration at the bottom of a vessel of reactor nr 1 of the Gravelines power station, anomalies concerning pipe supports, incident of the 4 May 2011 on the reactor nr 1 of the Tricastin power station, human and organisational failures in reactor control. The last part comments significant evolutions: EDF approach for a continuous safety improvement, control of reactor ageing effects, high room temperature for safety injection pumps, hybrid cores, boil-over risk at the vicinity of Gravelines

Development of safety performance indicators in Japan

International Nuclear Information System (INIS)

Ohashi, H.; Tamao, S.; Tanaka, J.; Sawayama, T.

2001-01-01

For the purpose of safety regulations of operating nuclear power stations in Japan, the regulatory authorities utilize two types of regulations. One is the direct regulation, such as periodical inspection to inspect the function and performance of equipment important to safety, and the other is the audit type regulation such as preservation inspection to audit the compliance with the safety preservation rules. As performance indicators are expected to be an effective tool to evaluate the activities by audit type regulations, NUPEC is studying a comprehensive set of operational performance indicators to meet the effective evaluation method for the safety preservation activities in the audit type regulations under the frame of current safety regulation system. The study includes the establishment of comprehensive operational performance indicators applicable in Japan, the effective application of performance indicators to the current Japanese regulation, the clarification of the applicable scope of utilization, the possibility of applying the performance indicators. This report describes the present status of our performance indicator studies. After the completion of these studies the regulatory authorities will evaluate if and how the new set of comprehensive performance indicators could be introduced to Japanese regulatory scheme. (author)

Safety evaluation report related to the operation of Catawba Nuclear Station, Units 1 and 2. Docket Nos. 50-413 and 50-414, Duke Power Company, et al

International Nuclear Information System (INIS)

1983-02-01

The Safety Evaluation Report for the application filed by Duke Power Company, North Carolina Municipal Power Agency Number 1, North Carolina Electric Membership Corporation, and Saluda River Electric Cooperative, Inc. as applicants and owners, for licenses to operate the Catawba Nuclear Station, Units 1 and 2 (Docket Nos. 50-413 and 50-414), has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is located in York County, South Carolina, approximately 9.6 km (6 mi) north of Rock Hill and adjacent to Lake Wylie. Subject to favorable resolution of the items discussed in this report, the staff concludes that the facility can be operated by the applicant without endangering the health and safety of the public

Safety evaluation report related to the operation of Comanche Peak Steam Electric Station, Units 1 and 2 (Docket Nos. 50-445 and 50-446): Supplement No. 21

International Nuclear Information System (INIS)

1989-04-01

Supplement 21 to the Safety Evaluation Report related to the operation of the Comanche Peak Steam Electric Station (CPSES), Units 1 and 2 (NUREG-0797), has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission (NRC). The facility is located in Somervell County, Texas, approximately 40 miles southwest of Fort Worth, Texas. This supplement reports the status of certain issues that had not been resolved when the Safety Evaluation Report and Supplements 1, 2, 3, 4, 6, and 12 to that report were published. This supplement also lists the new issues that have been identified since Supplement 12 was issued and includes the evaluations for licensing items resolved in this interim period. 21 refs

Reactor safety training for decision making

International Nuclear Information System (INIS)

Scott, C.K.

2003-01-01

The purpose of this paper is to describe an approach to reactor safety training for technical staff working at an operating station. The concept being developed is that, when the engineer becomes a registered professional engineer, they have sufficient reactor safety knowledge to perform independent technical work without compromising the safety of the plant. This goal would be achieved with a focused training program while working as an engineer-in-training (four years in NB). (author)

Case Study of Multi-Unit Risk: Multi-Unit Station Black-Out

Energy Technology Data Exchange (ETDEWEB)

Oh, Kyemin; Jang, Seung-cheol [KAERI, Daejeon (Korea, Republic of); Heo, Gyunyoung [Kyung Hee University, Yongin (Korea, Republic of)

2015-05-15

After Fukushima Daiichi Accident, importance and public concern for Multi-Unit Risk (MUR) or Probabilistic Safety Assessment (PSA) have been increased. Most of nuclear power plant sites in the world have more than two units. These sites have been facing the problems of MUR or accident such as Fukushima. In case of South Korea, there are generally more than four units on the same site and even more than ten units are also expected. In other words, sites in South Korea also have been facing same problems. Considering number of units on the same site, potential of these problems may be larger than other countries. The purpose of this paper is to perform case study based on another paper submitted in the conference. MUR is depended on various site features such as design, shared systems/structures, layout, environmental condition, and so on. Considering various dependencies, we assessed Multi-Unit Station Black-out (MSBO) accident based on Hanul Unit 3 and 4 model. In this paper, case study for multi-unit risk or PSA had been performed. Our result was incomplete to assess total multi-unit risk because of two challenging issues. First, economic impact had not been evaluated to estimate multi-unit risk. Second, large uncertainties were included in our result because of various assumptions. These issues must be resolved in the future.

Case Study of Multi-Unit Risk: Multi-Unit Station Black-Out

International Nuclear Information System (INIS)

Oh, Kyemin; Jang, Seung-cheol; Heo, Gyunyoung

2015-01-01

After Fukushima Daiichi Accident, importance and public concern for Multi-Unit Risk (MUR) or Probabilistic Safety Assessment (PSA) have been increased. Most of nuclear power plant sites in the world have more than two units. These sites have been facing the problems of MUR or accident such as Fukushima. In case of South Korea, there are generally more than four units on the same site and even more than ten units are also expected. In other words, sites in South Korea also have been facing same problems. Considering number of units on the same site, potential of these problems may be larger than other countries. The purpose of this paper is to perform case study based on another paper submitted in the conference. MUR is depended on various site features such as design, shared systems/structures, layout, environmental condition, and so on. Considering various dependencies, we assessed Multi-Unit Station Black-out (MSBO) accident based on Hanul Unit 3 and 4 model. In this paper, case study for multi-unit risk or PSA had been performed. Our result was incomplete to assess total multi-unit risk because of two challenging issues. First, economic impact had not been evaluated to estimate multi-unit risk. Second, large uncertainties were included in our result because of various assumptions. These issues must be resolved in the future

Research at United States Antarctic stations during the International Magnetosphere Study

International Nuclear Information System (INIS)

Rosenberg, T.J.

1982-01-01

During the International Magnetospheric Study (IMS) the U.S. operated programs at McMurdo, Siple, South Pole, and Palmer stations and at the Soviet Vostok station. Details concerning measurement locations are considered, and program summaries are provided. The programs are related to the study of geomagnetic variations, magnetic pulsations in the polar cap, cosmic noise absorption, VLF radio waves, auroral photometry, the morphology and dynamics of visible auroral forms, cosmic ray intensity variations, and auroral infrasonic waves. One program is based on the utilization of VHF Doppler auroral radar

Nuclear safety chains

International Nuclear Information System (INIS)

Robbins, M.C.; Eames, G.F.; Mayell, J.R.

1981-01-01

An original scheme has been developed for expressing the complex interrelationships associated with the engineered safeguards provided for a nuclear power station. This management tool, based upon network diagrams called Nuclear Safety Chains, looks at the function required of a particular item of safety plant, defines all of the vital supplies and support features necessary for successful operation, and expresses them in visual form, to facilitate analysis and optimisation for operations and maintenance staff. The safety chains are confined to manual schemes at present, although they are designed to be compatible with modern computer techniques. Their usefulness with any routine maintenance planning application on high technology plant is already being appreciated. (author)

On the troubles happened in nuclear power stations, 1995

International Nuclear Information System (INIS)

1996-01-01

The troubles which happened at the nuclear power stations of Japan in the fiscal year of 1995 are described in this report. The number of troubles in those power stations reported from the corporations of electric industry to Nuclear Safety Commission according to The Law for Regulation of Nuclear Fertile Material, Nuclear Fuel Material and Reactors and Utility Industry Law were 14 in the year and so, the number per reactor was 0.3. The details of the trouble cases were as follows; one and nine cases for automatic and manual shutdowns in operation, respectively and 4 cases found during a down-time of the reactor. But, there was no influence on the environment surrounding those nuclear power stations by the radioactive materials in either of the cases. (M.N.)

Electric motorcycle charging station powered by solar energy

Science.gov (United States)

Siriwattanapong, Akarawat; Chantharasenawong, Chawin

2018-01-01

This research proposes a design and verification of an off-grid photovoltaic system (PVS) for electric motorcycle charging station to be located in King’s Mongkut’s University of Technology Thonburi, Bangkok, Thailand. The system is designed to work independently (off-grid) and it must be able to fully charge the batteries of a typical passenger electric motorcycle every evening. A 1,000W Toyotron electric motorcycle is chosen for this study. It carries five units of 12.8V 20Ah batteries in series; hence its maximum energy requirement per day is 1,200Wh. An assessment of solar irradiation data and the Generation Factor in Bangkok, Thailand suggests that the charging system consists of one 500W PV panel, an MPPT charge controller, 48V 150Ah battery, a 1,000W DC to AC inverter and other safety devices such as fuses and breakers. An experiment is conducted to verify the viability of the off-grid PVS charging station by collecting the total daily energy generation data in the raining season and winter. The data suggests that the designed off-grid solar power charging station for electric motorcycle is able to supply sufficient energy for daily charging requirements.

Measuring Safety Culture on Ships Using Safety Climate: A Study among Indian Officers

Directory of Open Access Journals (Sweden)

Yogendra Bhattacharya

2015-12-01

Full Text Available Workplace safety continues to be an area of concern in the maritime industry due to the international nature of the operations. The effectiveness of extensive legislation to manage shipboard safety remains in doubt. The focus must therefore shift towards the human element - seafarers and their perceptions of safety. The study aims to understand the alignment that exists between safety culture and safety climate on board ships as perceived by seafarers. The underlying factors of safety climate were identified using factor analysis which isolated seven factors - Support on Safety, Organizational Support, Resource Availability, Work Environment, Job Demands, ‘Just’ Culture, and Safety Compliance. The perception of safety level of seafarers was found to be low indicating the existence of misalignments between safety culture values and the actual safety climate. The study also reveals that the safety perceptions of officers employed directly by ship owners and those by managers do not differ significantly, nor do they differ between senior and junior officers. A shift in perspective towards how seafarers themselves feel towards safety might provide more effective solutions – instead of relying on regulations - and indeed aid in reducing incidents on board. This paper details practical suggestions on how to identify the factors that contribute towards a better safety climate on board ships.

Safety Evaluation Report related to the operation of Grand Gulf Nuclear Station, Units 1 and 2 (Docket Nos. 50-416 and 50-417). Supplement 6

International Nuclear Information System (INIS)

1984-08-01

Supplement 6 to the Safety Evaluation Report for Mississippi Power and Light Company et al. joint application for licenses to operate the Grand Gulf Nuclear Station, Units 1 and 2, located on the east bank of the Mississippi River near Port Gibson in Claiborne County, Mississippi, has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. This supplement reports the NRC staff's evaluation of open items from previous supplements and Technical Specification changes required before authorizing operation of Unit 1 above 5% of rated power

Risk perception and occupational accidents: a study of gas station workers in southern Brazil.

Science.gov (United States)

Cezar-Vaz, Marta Regina; Rocha, Laurelize Pereira; Bonow, Clarice Alves; Silva, Mara Regina Santos da; Vaz, Joana Cezar; Cardoso, Letícia Silveira

2012-07-01

The present study aimed to identify the perceptions of gas station workers about physical, chemical, biological and physiological risk factors to which they are exposed in their work environment; identify types of occupational accidents involving gas station workers and; report the development of a socioenvironmental intervention as a tool for risk communication to gas station workers. A quantitative study was performed with 221 gas station workers in southern Brazil between October and December 2010. Data collection was performed between October to December 2010 via structured interviews. The data were analyzed using SPSS 19.0. The participants identified the following risk types: chemical (93.7%), physical (88.2%), physiological (64.3%) and biological (62.4%). In this sample, 94.1% of gas station workers reported occupational accidents, and 74.2% reported fuel contact with the eyes (p accidents as an indicator of the dangerous nature of their work environment.

Qualification of safety-related valve actuators

International Nuclear Information System (INIS)

Anon.

1981-01-01

This Standard describes the qualification of all types of power-driven valve actuators, including damper actuators, for safety-related functions in nuclear power generating stations. It may also be used to separately qualify actuator components. This Standard establishes the minimum requirements for, and guidance regarding, the methods and procedures for qualification of all safety-related functions of power-driven valve actuators

[Risk-oriented model of the control of the level of electric magnetic fields of base stations of cellular communications].

Science.gov (United States)

Lutsenko, L A; Tulakin, A V; Egorova, A M; Mikhailova, O M; Gvozdeva, L L; Chigryay, E K

The purpose of this study was to give the description of harmful effects of the impact of electromagnetic radiations from base stations of cellular communication as the most common sources of radio frequencies of electromagnetic fields in the environment. The highest values of the energy flux density were measured on the roofs of houses where antennas are installed - more than 10 pW/cm. The lowest values were recorded in inside premises with expositions of 0.1-1 pW/cm. In the close location of the railway station to the base stations of the cellular communication there was seen a cumulative effect. There are proposed both new safe hygienic approaches to the control for the safety of the work of base station and protective measures.

Space Station data system analysis/architecture study. Task 1: Functional requirements definition, DR-5

Science.gov (United States)

1985-01-01

The initial task in the Space Station Data System (SSDS) Analysis/Architecture Study is the definition of the functional and key performance requirements for the SSDS. The SSDS is the set of hardware and software, both on the ground and in space, that provides the basic data management services for Space Station customers and systems. The primary purpose of the requirements development activity was to provide a coordinated, documented requirements set as a basis for the system definition of the SSDS and for other subsequent study activities. These requirements should also prove useful to other Space Station activities in that they provide an indication of the scope of the information services and systems that will be needed in the Space Station program. The major results of the requirements development task are as follows: (1) identification of a conceptual topology and architecture for the end-to-end Space Station Information Systems (SSIS); (2) development of a complete set of functional requirements and design drivers for the SSIS; (3) development of functional requirements and key performance requirements for the Space Station Data System (SSDS); and (4) definition of an operating concept for the SSIS. The operating concept was developed both from a Space Station payload customer and operator perspective in order to allow a requirements practicality assessment.

Planning and preparedness for radiological emergencies at nuclear power stations

International Nuclear Information System (INIS)

Thomson, R.; Muzzarelli, J.

1996-01-01

The Radiological Emergency Preparedness (REP) Program was created after the March 1979 accident at the Three Mile Island nuclear power station. The Federal Emergency Management Agency (FEMA) assists state and local governments in reviewing and evaluating state and local REP plans and preparedness for accidents at nuclear power plants, in partnership with the US Nuclear Regulatory Commission (NRC), which evaluates safety and emergency preparedness at the power stations themselves. Argonne National Laboratory provides support and technical assistance to FEMA in evaluating nuclear power plant emergency response exercises, radiological emergency plans, and preparedness

Systematic evaluation program review of NRC safety topic VII-2 associated with the electrical, instrumentation and control portions of the ESF system control logic and design for the Dresden Station, Unit II nuclear power plant

International Nuclear Information System (INIS)

St Leger-Barter, G.

1980-11-01

This report documents the technical evaluation and review of NRC Safety Topic VII-2, associated with the electrical, instrumentation, and control portions of the ESF system control logic and design for the Dresden Station Unit II nuclear power plant, using current licensing criteria

Nuclear safety. Improvement programme

International Nuclear Information System (INIS)

2000-01-01

In this brochure the improvement programme of nuclear safety of the Mochovce NPP is presented in detail. In 1996, a 'Mochovce NPP Nuclear Safety Improvement Programme' was developed in the frame of unit 1 and 2 completion project. The programme has been compiled as a continuous one, with the aim to reach the highest possible safety level at the time of commissioning and to establish good preconditions for permanent safety improvement in future. Such an approach is in compliance with the world's trends of safety improvement, life-time extension, modernisation and nuclear station power increase. The basic document for development of the 'Programme' is the one titled 'Safety Issues and their Ranking for WWER 440/213 NPP' developed by a group of IAEA experts. The following organisations were selected for solution of the safety measures: EUCOM (Consortium of FRAMATOME, France, and SIEMENS, Germany); SKODA Prague, a.s.; ENERGOPROJEKT Prague, a.s. (EGP); Russian organisations associated in ATOMENERGOEXPORT; VUJE Trnava, a.s

Reactor safety in Eastern Europe. Proceedings

International Nuclear Information System (INIS)

1995-02-01

The papers presented to the GRS colloquium refer to the cooperative activities for reactor accident analysis and modification of the GRS computer codes for their application to reactors of the Russian design types of WWER or RBMK. Another topic is the safety of RBMK reactors in particular, and the current status of investigations and studies addressing the containment of unit 4 of the Chernobyl reactor station. (HP) [de

Experience gained in the operation of the Beznau nuclear power station

International Nuclear Information System (INIS)

Kueffer, K. von.

1976-01-01

The 24th December 1969, when the Beznau 1 nuclear power station was placed in commercial operation, marked the beginning of electricity production from nuclear energy in Switzerland. Beznau 2 followed on the 15th March 1972. Together with the Muehleberg nuclear power station, nuclear energy accounts for approximately 20 percent of the total electricity production in Switzerland. Until the end of 1975, Switzerland's three nuclear power stations had a mean energy utilisation factor of 71.3 percent which, as compared with a mean energy utilization factor of 60.5 percent for all the nuclear power stations in the West, suggests fairly good operational results. Problems that arose during operation are discussed in detail. By way of summary it is stated that the operation of the Beznau nuclear power station has so far proved a success. The production of electrical energy has always remained within the limits imposed by the law and by the safety aspects. (Auth.)

Safety Evaluation Report related to the operation of Palo Verde Nuclear Generating Station, Units 1, 2, and 3 (Docket Nos. STN 50-528, STN 50-529, and STN 50-530)

International Nuclear Information System (INIS)

1984-10-01

Supplement No. 6 to the Safety Evaluation Report for the application filed by Arizona Public Service Company, et al., for licenses to operate the Palo Verde Nuclear Generating Station, Units 1, 2, and 3 (Docket Nos. STN 50-528/529/530), located in Maricopa County, Arizona, has been prepared by the Office of Nuclear Reactor Regulation of the Nuclear Regulatory Commission. The purpose of this supplement is to update the Safety Evaluation Report by providing an evaluation of (1) additional information submitted by the applicant since Supplement No. 5 was issued and (2) matters that the staff had under review when Supplement No. 5 was issued

Seismological studies carried out by the CEA in connection with the safety of nuclear sites

International Nuclear Information System (INIS)

Barbreau, A.; Ferrieux, H.; Mohammadioun, B.

1975-01-01

In order to evaluate the seismic risk at nuclear sites, the Department of Nuclear Safety of the French Atomic Energy Commission (CEA) has been conducting a programme of seismological studies for several years past. This programme is aimed at acquiring a better knowledge of seismic phenomena, in particular the spectral distribution of the energy of earthquakes, considered to be the only correct approach to the problem of earthquake protection, as well as a better knowledge of the seismic activity of the areas surrounding nuclear sites. The authors propose defining the design spectrum of the site on the basis of the probable energy at the source, the distance from the epicentre and the transfer function of the geological formations. The need - for the purpose of defining this spectrum - to acquire data on the characteristics of French earthquakes and on regional seismicity led the Department of Nuclear Safety to set up a network of seismic stations. It now has an observatory at the Cadarache Nuclear Research Centre and mobile stations with automatic magnetic recording for studying aftershock sequences and the activity of faults in the vicinity of nuclear sites, and for making the measurements necessary to calculate the transfer functions. With this equipment it was possible to record six aftershocks of the Oleron earthquake on 7 September 1972 close to the epicentre, and to calculate the spectra therefrom. The latter contained a lot of high frequencies, which is in agreement with the data obtained from other sources for earthquakes of low energy. The synthetic spectra calculated on the basis of one magnitude and one distance are in good agreement with the spectra obtained experimentally

Safety evaluation report related to the operation of Catawba Nuclear Station, Units 1 and 2 (Docket Nos. 50-413 and 50-414)

International Nuclear Information System (INIS)

1984-07-01

The report supplements the Safety Evaluation Report (NUREG-0954) issued in February 1983 by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission with respect to the application filed by Duke Power Company, North Carolina Municipal Power Agency Number 1, North Carolina Membership Corporation, and Saluda River Electric Cooperative, Inc. as applicants and owners, for licenses to operate the Catawba Nuclear Station, Units 1 and 2 (Docket Nos. 50-413 and 50-414, respectively). The facility is located in York County, South Carolina, approximately 9.6 km (6 mi) north of Rock Hill and adjacent to Lake Wylie. This supplement provides additional information supporting the license for fuel loading and precriticality testing for Unit 1

Reactor safety in Eastern Europe

International Nuclear Information System (INIS)

1995-02-01

The papers presented to the GRS colloquium refer to the cooperative activities for reactor accident analysis and modification of the GRS computer codes for their application to reactors of the Russian design types of WWER or RBMK. Another topic is the safety of RBMK reactors in particular, and the current status of investigations and studies addressing the containment of unit 4 of the Chernobyl reactor station. All papers are indexed separately in report GRS--117. (HP)

Wolf Creek Generating Station containment model

International Nuclear Information System (INIS)

Nguyen, D.H.; Neises, G.J.; Howard, M.L.

1995-01-01

This paper presents a CONTEMPT-LT/28 containment model that has been developed by Wolf Creek Nuclear Operating Corporation (WCNOC) to predict containment pressure and temperature behavior during the postulated events at Wolf Creek Generating Station (WCGS). The model has been validated using data provided in the WCGS Updated Safety Analysis Report (USAR). CONTEMPT-LT/28 model has been used extensively at WCGS to support plant operations, and recently, to support its 4.5% thermal power uprate project

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

Electrical system design and reliability at Ontario Hydro nuclear generating stations

Energy Technology Data Exchange (ETDEWEB)

Royce, C. J. [Ontario Hydro, 700 University Avenue, Toronto, Ontario M5G 1X6 (Canada)

1986-02-15

This paper provides an overview of design practice and the predicted and actual reliability of electrical station service Systems at Ontario Nuclear Generating Stations. Operational experience and licensing changes have indicated the desirability of improving reliability in certain instances. For example, the requirement to start large emergency coolant injection pumps resulted in the turbine generator units in a multi-unit station being used as a back-up power supply. Results of reliability analyses are discussed. To mitigate the effects of common mode events Ontario Hydro adopted a 'two group' approach to the design of safety related Systems. This 'two group' approach is reviewed and a single fully environmentally qualified standby power supply is proposed for future use. (author)

NB Power's public safety perspective

Energy Technology Data Exchange (ETDEWEB)

Sisk, P [New Brunswick Power, Fredericton, NB (Canada)

2009-07-01

New Brunswick Power Generation (NB Power) - Genco operates and maintains one of North America's most diverse generating systems. It consists of 15 hydro, coal, oil and diesel-powered generating stations and supplies approximately 75 per cent of the in-province load. It also exports energy to neighbouring New England, Quebec, Prince Edward Island, and Nova Scotia. This presentation provided some history regarding public safety issues at NB Power. The Mactaquac generating system was discussed with particular reference to its activities, challenges and control measures such as signage, audible alarms, visuals, security fencing, and coast guard navigation buoys. Several recommendations were presented, such as developing a tool to conduct risk assessment at all hydro stations; developing a public safety campaign; installing booms where required; standardizing signs; evaluating security risks; and conducting pre-spill inspections to remove the public from dangerous areas. figs.