Development of Mitigation Strategy for Beyond Design Basis External Events for NRC Design Certification

International Nuclear Information System (INIS)

Kim, Dong Hak; Lee, Jae Jong; Kim, Myung Ki

2013-01-01

In this study, how to develop FLEX strategy for beyond-design-basis external events for U. S. NRC design certification is examined. The development method of FLEX strategy for U. S. NRC design certification is examined. The applicants should make unit-specific FLEX strategy and establish the minimum coping capabilities consistent with unit-specific evaluation of the potential impacts and responses to BDBEEs. NEI 12-06 outlines the process to define and deploy the diverse and flexible mitigation strategies(FLEX strategy) that will increase defense-in-depth for beyond-design-basis scenarios to address the extended loss of alternating current (ac) power (ELAP) and loss of normal access to the ultimate heat sink (LUHS) occurring simultaneously at all units on a site. The order (EA-12-049) is issued to all reactor licensees, including holders of active, Construction Permit (CP) holders, and Combined License (COL) holders. Applicants for the new reactor design certification should prepare and submit FLEX strategy for NRC staff's review. Site-specific data related with the new reactor can't be determined during the new reactor design certification applications so that the unit-specific FLEX strategy should be developed

PHWR safety: design, siting and construction

International Nuclear Information System (INIS)

Sharma, V.K.

2002-01-01

In all activities associated with NPPs viz. siting, design, construction, commissioning and operation, safety is given overriding importance. The safety design principles of PHWRs are based on defence-in-depth approach, physical and functional separation between process and safety systems and also among various safety systems, redundancy to meet single failure criteria and postulation of a number of design basis events for which the plant must be designed. Apart from engineered safety systems, PHWRs have inherent characteristics which contribute to safety. In siting of a NPP, it is required to ensure that the given site does not pose undue radiological hazard to public and the environment both during normal operation as well as during and following an accident condition. For this purpose, all site related external events, both natural and man induced, are assessed for their effect on the plant and are considered as part of the design basis. Possible radiological impact of the NPP on environment and surrounding population is assessed and ensured to be within acceptable limits. During construction phase, it is essential that the NPP be built in accordance with design intent and with required quality of workmanship to ensure that the NPP will remain safe during all states of operation. This is achieved through careful execution and QA activities encompassing all aspects of component fabrication at manufacturer works, civil construction, site erection, assembly, and commissioning. Future trends in nuclear safety will continue to be based on existing principles which have proved to be sound. These will be further strengthened by features such as increasing use of passive means of performing safety functions and a more explicit treatment of severe accidents. (author)

BWR NSSS design basis documentation

International Nuclear Information System (INIS)

Vij, R.S.; Bates, R.E.

2004-01-01

In 1985 an incident at Toledo Edison's Davis Besse plant caused the U.S. Nuclear Regulatory Commission (NRC) to re-evaluate the technical information that the utilities had readily available to support the design of their plants. The Design Basis programs, currently on going in most U.S. utilities, have been the nuclear industry's response to the needs identified by this re-evaluation. In order to understand the Design Basis programs which have been implemented by the U.S. nuclear utilities, it is necessary to understand the problem as it was perceived by the nuclear industry (the utilities, the original NSSS designers and the regulators) after the Davis-Besse incident, the subsequent programs undertaken by the industry under the leadership of INPO and NUMARC, the NRC's actions, and the overall evolution of the industry's vision in relation to this problem. This paper presents the history of the design basis efforts from the first recognition of the problem by the NRC after the Davis-Besse incident, describes the actions taken by the NRC, INPO, NUMARC, the U.S. utilities and the NSSS designers, and brings the problem statement up-to-date in relation to the vision presently held by the U.S. nuclear industry. It then presents a technical discussion to develop a detailed definition of design basis information to support the problem statement. The information originally supplied by the NSSS designers during the plant design and construction is discussed as well as its relationship to the previously defined design basis information. This section of the paper concludes by defining the additional information needed by nuclear utilities to satisfy the requirements developed from the problem statement. Having developed a definition of the additional information (i.e., information not originally supplied during design and construction) required to solve the design basis problem as it is presently perceived by the U.S. nuclear industry, the paper then discusses design basis

Understanding and capturing NSSS design basis

International Nuclear Information System (INIS)

Palo, W.J.; Miller, B.

1993-01-01

Changes to, and technical evaluations of nuclear generating station designs are often warranted. Comprehensive documentation and understanding of the NSSS Design Basis are essential to support these activities. Effective configuration management tools are also needed to maintain the plant within design basis limits. Efficient design basis reconstitution can be realized via: In-depth understanding of the design process; Utilization of effective data collection methodology; State of the art data basing tools. A database can be created to generate a Design Basis Manual (DBM). This database can communicate electronically with other plant databases. A living document vice a static snapshot of the plant design is the goal. A design basis database can serve as the cornerstone for a global electronic information control system

Design basis 2

Energy Technology Data Exchange (ETDEWEB)

Larsen, G.; Soerensen, P. [Risoe National Lab., Roskilde (Denmark)

1996-09-01

Design Basis Program 2 (DBP2) is comprehensive fully coupled code which has the capability to operate in the time domain as well as in the frequency domain. The code was developed during the period 1991-93 and succeed Design Basis 1, which is a one-blade model presuming stiff tower, transmission system and hub. The package is designed for use on a personal computer and offers a user-friendly environment based on menu-driven editing and control facilities, and with graphics used extensively for the data presentation. Moreover in-data as well as results are dumped on files in Ascii-format. The input data is organized in a in-data base with a structure that easily allows for arbitrary combinations of defined structural components and load cases. (au)

Design basis ground motion (Ss) required on new regulatory guide

International Nuclear Information System (INIS)

Kamae, Katsuhiro

2013-01-01

New regulatory guide is enforced on July 8. Here, it is introduced how the design basis ground motion (Ss) for seismic design of nuclear power reactor facilities was revised on the new guide. Ss is formulated as two types of earthquake ground motions, earthquake ground motions with site specific earthquake source and with no such specific source locations. The latter is going to be revised based on the recent observed near source ground motions. (author)

Guidelines for determining design basis ground motions

International Nuclear Information System (INIS)

1993-11-01

This report develops and applies a method for estimating strong earthquake ground motion. The emphasis of this study is on ground motion estimation in Eastern North America (east of the Rocky Mountains), with particular emphasis on the Eastern United States and southeastern Canada. Specifically considered are ground motions resulting from earthquakes with magnitudes from 5 to 8, fault distances from 0 to 500 km, and frequencies from 1 to 35 Hz. The two main objectives were: (1) to develop generic relations for estimating ground motion appropriate for site screening; and (2) to develop a guideline for conducting a thorough site investigation needed to define the seismic design basis. For the first objective, an engineering model was developed to predict the expected ground motion on rock sites, with an additional set of amplification factors to account for the response of the soil column over rock at soil sites. The results incorporate best estimates of ground motion as well as the randomness and uncertainty associated with those estimates. For the second objective, guidelines were developed for gathering geotechnical information at a site and using this information in calculating site response. As a part of this development, an extensive set of geotechnical and seismic investigations was conducted at three reference sites. Together, the engineering model and guidelines provide the means to select and assess the seismic suitability of a site

Advanced Test Reactor Safety Basis Upgrade Lessons Learned Relative to Design Basis Verification and Safety Basis Management

International Nuclear Information System (INIS)

G. L. Sharp; R. T. McCracken

2004-01-01

The Advanced Test Reactor (ATR) is a pressurized light-water reactor with a design thermal power of 250 MW. The principal function of the ATR is to provide a high neutron flux for testing reactor fuels and other materials. The reactor also provides other irradiation services such as radioisotope production. The ATR and its support facilities are located at the Test Reactor Area of the Idaho National Engineering and Environmental Laboratory (INEEL). An audit conducted by the Department of Energy's Office of Independent Oversight and Performance Assurance (DOE OA) raised concerns that design conditions at the ATR were not adequately analyzed in the safety analysis and that legacy design basis management practices had the potential to further impact safe operation of the facility.1 The concerns identified by the audit team, and issues raised during additional reviews performed by ATR safety analysts, were evaluated through the unreviewed safety question process resulting in shutdown of the ATR for more than three months while these concerns were resolved. Past management of the ATR safety basis, relative to facility design basis management and change control, led to concerns that discrepancies in the safety basis may have developed. Although not required by DOE orders or regulations, not performing design basis verification in conjunction with development of the 10 CFR 830 Subpart B upgraded safety basis allowed these potential weaknesses to be carried forward. Configuration management and a clear definition of the existing facility design basis have a direct relation to developing and maintaining a high quality safety basis which properly identifies and mitigates all hazards and postulated accident conditions. These relations and the impact of past safety basis management practices have been reviewed in order to identify lessons learned from the safety basis upgrade process and appropriate actions to resolve possible concerns with respect to the current ATR safety

Design characteristics of EU-APR1400 on-site power system

International Nuclear Information System (INIS)

Kim, D.H.; Kim, Y.S.; Kim, Y.S.

2014-01-01

In the global nuclear market, US and European design requirements have been largely used to develop the design of nuclear power plants(NPPs). The APR1400 design was developed on the basis of US regulatory guide and EPRI utility requirements document(URD). In order to enlarge the export market of APR1400, KHNP (Korea Hydro & Nuclear Power Co., Ltd) has developed the EU-APR1400 design which complies with the European nuclear design requirements. In this paper, the design characteristics of EU-APR1400 on-site power system developed according to the European design requirements of electrical power system are described. The European main design requirements of electrical power system involve 50 Hz rated frequency, 400/110 kV grid voltage, the application of the diversity and the redundancy, and so on. The EU-APR1400 on-site power system has been developed on the basis of these requirements. The representative designs include the redundancy, diversity, independence design, the emergency power supply design, the design for providing electrical power to the dedicated severe accident systems, and the design for European grid requirements. (author)

Early Site Permit Demonstration Program: Guidelines for determining design basis ground motions

International Nuclear Information System (INIS)

1993-01-01

This report develops and applies a methodology for estimating strong earthquake ground motion. The motivation was to develop a much needed tool for use in developing the seismic requirements for structural designs. An earthquake's ground motion is a function of the earthquake's magnitude, and the physical properties of the earth through which the seismic waves travel from the earthquake fault to the site of interest. The emphasis of this study is on ground motion estimation in Eastern North America (east of the Rocky Mountains), with particular emphasis on the Eastern United States and southeastern Canada. Eastern North America is a stable continental region, having sparse earthquake activity with rare occurrences of large earthquakes. While large earthquakes are of interest for assessing seismic hazard, little data exists from the region to empirically quantify their effects. The focus of the report is on the attributes of ground motion in Eastern North America that are of interest for the design of facilities such as nuclear power plants. This document, Volume II, contains Appendices 2, 3, 5, 6, and 7 covering the following topics: Eastern North American Empirical Ground Motion Data; Examination of Variance of Seismographic Network Data; Soil Amplification and Vertical-to-Horizontal Ratios from Analysis of Strong Motion Data From Active Tectonic Regions; Revision and Calibration of Ou and Herrmann Method; Generalized Ray Procedure for Modeling Ground Motion Attenuation; Crustal Models for Velocity Regionalization; Depth Distribution Models; Development of Generic Site Effects Model; Validation and Comparison of One-Dimensional Site Response Methodologies; Plots of Amplification Factors; Assessment of Coupling Between Vertical ampersand Horizontal Motions in Nonlinear Site Response Analysis; and Modeling of Dynamic Soil Properties

Evolution of Safety Basis Documentation for the Fernald Site

International Nuclear Information System (INIS)

Brown, T.; Kohler, S.; Fisk, P.; Krach, F.; Klein, B.

2004-01-01

The objective of the Department of Energy's (DOE) Fernald Closure Project (FCP), in suburban Cincinnati, Ohio, is to safely complete the environmental restoration of the Fernald site by 2006. Over 200 out of 220 total structures, at this DOE plant site which processed uranium ore concentrates into high-purity uranium metal products, have been safely demolished, including eight of the nine major production plants. Documented Safety Analyses (DSAs) for these facilities have gone through a process of simplification, from individual operating Safety Analysis Reports (SARs) to a single site-wide Authorization Basis containing nuclear facility Bases for Interim Operations (BIOs) to individual project Auditable Safety Records (ASRs). The final stage in DSA simplification consists of project-specific Integrated Health and Safety Plans (I-HASPs) and Nuclear Health and Safety Plans (N-HASPs) that address all aspects of safety, from the worker in the field to the safety basis requirements preserving the facility/activity hazard categorization. This paper addresses the evolution of Safety Basis Documentation (SBD), as DSAs, from production through site closure

Designation of facility usage categories for Hanford Site facilities

International Nuclear Information System (INIS)

Wodrich, D.; Ellingson, D.; Scott, M.; Schade, A.

1991-01-01

This report summarizes the Hanford Site methodology used to ensure facility compliance with the natural phenomena design criteria set forth in the US Department of Energy orders and guidance. In particular, the Hanford Site approach to designating a suitable facility open-quotes Usage Category,close quotes is presented. The current Hanford Site methodology for Usage Category designation is based on an engineered feature's safety function and on the feature's assigned Safety Class. At the Hanford Site, Safety Class assignments are deterministic in nature and are based on the consequences of failure, without regard to the likelihood of occurrence. The report also proposes a risk-based approach to Usage Category designation, which is being considered for future application at the Hanford Site. To establish a proper Usage Category designation, the safety analysis and engineering design processes must be coupled. This union produces a common understanding of the safety function(s) to be accomplished by the design feature(s) and a sound basis for the assignment of Usage Categories to the appropriate systems, structures, and components

Designation of facility usage categories for Hanford Site facilities

International Nuclear Information System (INIS)

Woodrich, D.D.; Ellingson, D.R.; Scott, M.A.; Schade, A.R.

1991-10-01

This report summarizes the Hanford Site methodology used to ensure facility compliance with the natural phenomena design criteria set forth in the US Department of Energy Orders and guidance. The current Hanford Site methodology for Usage Category designation is based on an engineered feature's safety function and on the feature's assigned Safety Class. At the Hanford Site, Safety Class assignments are deterministic in nature and are based on teh consequences of failure, without regard to the likelihood of occurrence. The report also proposes a risk-based approach to Usage Category designation, which is being considered for future application at the Hanford Site. To establish a proper Usage Category designation, the safety analysis and engineering design processes must be coupled. This union produces a common understanding of the safety function(s) to be accomplished by the design feature(s) and a sound basis for the assignment of Usage Categories to the appropriate systems, structures, and components. 4 refs., 9 figs., 1 tab

A basis for standardized seismic design (SSD) for nuclear power plants/critical facilities

International Nuclear Information System (INIS)

O'Hara, T.F.; Jacobson, J.P.; Bellini, F.X.

1991-01-01

US Nuclear Power Plants (NPP's) are designed, engineered and constructed to stringent standards. Their seismic adequacy is assured by compliance with regulatory standards and demonstrated by both probabilistic risk assessments (PRAs) and seismic margin studies. However, present seismic siting criteria requires improvement. Proposed changes to siting criteria discussed here will provide a predictable licensing process and a stable regulatory environment. Two recent state-of-the-art studies evaluate the seismic design for all eastern US (EUS) NPP'S: a Lawrence Livermore National Labs study (LLNL, 1989) funded by the NRC and similar research by the Electric Power Research Institute (EPRI, 1989) supported by the utilities. Both confirm that Appendix A 10CFR Part 100 has not provided consistent seismic design levels for all sites. Standardized Seismic Design (SSD) uses a probabilistic framework to accommodate alternative deterministic interpretations. It uses seismic hazard input from EPRI or LLNL to produce consistent bases for future seismic design. SSD combines deterministic and probabilistic insights to provide a comprehensive approach for determining a future site's acceptable seismic design basis

Development and comparision of techniques for estimating design basis flood flows for nuclear power plants

International Nuclear Information System (INIS)

1980-05-01

Estimation of the design basis flood for Nuclear Power Plants can be carried out using either deterministic or stochastic techniques. Stochastic techniques, while widely used for the solution of a variety of hydrological and other problems, have not been used to date (1980) in connection with the estimation of design basis flood for NPP siting. This study compares the two techniques against one specific river site (Galt on the Grand River, Ontario). The study concludes that both techniques lead to comparable results , but that stochastic techniques have the advantage of extracting maximum information from available data and presenting the results (flood flow) as a continuous function of probability together with estimation of confidence limits. (author)

Safety Basis Report

International Nuclear Information System (INIS)

R.J. Garrett

2002-01-01

As part of the internal Integrated Safety Management Assessment verification process, it was determined that there was a lack of documentation that summarizes the safety basis of the current Yucca Mountain Project (YMP) site characterization activities. It was noted that a safety basis would make it possible to establish a technically justifiable graded approach to the implementation of the requirements identified in the Standards/Requirements Identification Document. The Standards/Requirements Identification Documents commit a facility to compliance with specific requirements and, together with the hazard baseline documentation, provide a technical basis for ensuring that the public and workers are protected. This Safety Basis Report has been developed to establish and document the safety basis of the current site characterization activities, establish and document the hazard baseline, and provide the technical basis for identifying structures, systems, and components (SSCs) that perform functions necessary to protect the public, the worker, and the environment from hazards unique to the YMP site characterization activities. This technical basis for identifying SSCs serves as a grading process for the implementation of programs such as Conduct of Operations (DOE Order 5480.19) and the Suspect/Counterfeit Items Program. In addition, this report provides a consolidated summary of the hazards analyses processes developed to support the design, construction, and operation of the YMP site characterization facilities and, therefore, provides a tool for evaluating the safety impacts of changes to the design and operation of the YMP site characterization activities

Safety Basis Report

Energy Technology Data Exchange (ETDEWEB)

R.J. Garrett

2002-01-14

As part of the internal Integrated Safety Management Assessment verification process, it was determined that there was a lack of documentation that summarizes the safety basis of the current Yucca Mountain Project (YMP) site characterization activities. It was noted that a safety basis would make it possible to establish a technically justifiable graded approach to the implementation of the requirements identified in the Standards/Requirements Identification Document. The Standards/Requirements Identification Documents commit a facility to compliance with specific requirements and, together with the hazard baseline documentation, provide a technical basis for ensuring that the public and workers are protected. This Safety Basis Report has been developed to establish and document the safety basis of the current site characterization activities, establish and document the hazard baseline, and provide the technical basis for identifying structures, systems, and components (SSCs) that perform functions necessary to protect the public, the worker, and the environment from hazards unique to the YMP site characterization activities. This technical basis for identifying SSCs serves as a grading process for the implementation of programs such as Conduct of Operations (DOE Order 5480.19) and the Suspect/Counterfeit Items Program. In addition, this report provides a consolidated summary of the hazards analyses processes developed to support the design, construction, and operation of the YMP site characterization facilities and, therefore, provides a tool for evaluating the safety impacts of changes to the design and operation of the YMP site characterization activities.

System Design and the Safety Basis

International Nuclear Information System (INIS)

Ellingson, Darrel

2008-01-01

The objective of this paper is to present the Bechtel Jacobs Company, LLC (BJC) Lessons Learned for system design as it relates to safety basis documentation. BJC has had to reconcile incomplete or outdated system description information with current facility safety basis for a number of situations in recent months. This paper has relevance in multiple topical areas including documented safety analysis, decontamination and decommissioning (D and D), safety basis (SB) implementation, safety and design integration, potential inadequacy of the safety analysis (PISA), technical safety requirements (TSR), and unreviewed safety questions. BJC learned that nuclear safety compliance relies on adequate and well documented system design information. A number of PIS As and TSR violations occurred due to inadequate or erroneous system design information. As a corrective action, BJC assessed the occurrences caused by systems design-safety basis interface problems. Safety systems reviewed included the Molten Salt Reactor Experiment (MSRE) Fluorination System, K-1065 fire alarm system, and the K-25 Radiation Criticality Accident Alarm System. The conclusion was that an inadequate knowledge of system design could result in continuous non-compliance issues relating to nuclear safety. This was especially true with older facilities that lacked current as-built drawings coupled with the loss of 'historical knowledge' as personnel retired or moved on in their careers. Walkdown of systems and the updating of drawings are imperative for nuclear safety compliance. System design integration with safety basis has relevance in the Department of Energy (DOE) complex. This paper presents the BJC Lessons Learned in this area. It will be of benefit to DOE contractors that manage and operate an aging population of nuclear facilities

Development of a design basis tornado and structural design criteria for Lawrence Livermore Laboratory's Site 300

International Nuclear Information System (INIS)

McDonald, J.R.; Minor, J.E.; Mehta, K.C.

1975-11-01

Criteria are prescribed and guidance is provided for professional personnel who are involved with the evaluation of existing buildings and facilities at Site 300 near Livermore, California to resist the possible effects of extreme winds and tornadoes. The development of parameters for the effects of tornadoes and extreme winds and guidelines for evaluation and design of structures are presented. The investigations conducted are summarized and the techniques used for arriving at the combined tornado and extreme wind risk model are discussed. The guidelines for structural design methods for calculating pressure distributions on walls and roofs of structures and methods for accommodating impact loads from missiles are also presented

PRELIMINARY SELECTION OF MGR DESIGN BASIS EVENTS

International Nuclear Information System (INIS)

Kappes, J.A.

1999-01-01

The purpose of this analysis is to identify the preliminary design basis events (DBEs) for consideration in the design of the Monitored Geologic Repository (MGR). For external events and natural phenomena (e.g., earthquake), the objective is to identify those initiating events that the MGR will be designed to withstand. Design criteria will ensure that radiological release scenarios resulting from these initiating events are beyond design basis (i.e., have a scenario frequency less than once per million years). For internal (i.e., human-induced and random equipment failures) events, the objective is to identify credible event sequences that result in bounding radiological releases. These sequences will be used to establish the design basis criteria for MGR structures, systems, and components (SSCs) design basis criteria in order to prevent or mitigate radiological releases. The safety strategy presented in this analysis for preventing or mitigating DBEs is based on the preclosure safety strategy outlined in ''Strategy to Mitigate Preclosure Offsite Exposure'' (CRWMS M andO 1998f). DBE analysis is necessary to provide feedback and requirements to the design process, and also to demonstrate compliance with proposed 10 CFR 63 (Dyer 1999b) requirements. DBE analysis is also required to identify and classify the SSCs that are important to safety (ITS)

Development of a design basis tornado and structural design criteria for the Nevada Test Site, Nevada. Final report

International Nuclear Information System (INIS)

McDonald, J.R.; Minor, J.E.; Mehta, K.C.

1975-06-01

In order to evaluate the ability of critical facilities at the Nevada Test Site to withstand the possible damaging effects of extreme winds and tornadoes, parameters for the effects of tornadoes and extreme winds and structural design criteria for the design and evaluation of structures were developed. The meteorological investigations conducted are summarized, and techniques used for developing the combined tornado and extreme wind risk model are discussed. The guidelines for structural design include methods for calculating pressure distributions on walls and roofs of structures and methods for accommodating impact loads from wind-driven missiles. Calculations for determining the design loads for an example structure are included

Exploratory Shaft Facility design basis study report

International Nuclear Information System (INIS)

Langstaff, A.L.

1987-01-01

The Design Basis Study is a scoping/sizing study that evaluated the items concerning the Exploratory Shaft Facility Design including design basis values for water and methane inflow; flexibility of the design to support potential changes in program direction; cost and schedule impacts that could result if the design were changed to comply with gassy mine regulations; and cost, schedule, advantages and disadvantages of a larger second shaft. Recommendations are proposed concerning water and methane inflow values, facility layout, second shaft size, ventilation, and gassy mine requirements. 75 refs., 3 figs., 7 tabs

Early Site Permit Demonstration Program: Guidelines for determining design basis ground motions

International Nuclear Information System (INIS)

1993-01-01

This report develops and applies a methodology for estimating strong earthquake ground motion. The motivation was to develop a much needed tool for use in developing the seismic requirements for structural designs. An earthquake's ground motion is a function of the earthquake's magnitude, and the physical properties of the earth through which the seismic waves travel from the earthquake fault to the site of interest. The emphasis of this study is on ground motion estimation in Eastern North America (east of the Rocky Mountains), with particular emphasis on the Eastern United States and southeastern Canada. Eastern North America is a stable continental region, having sparse earthquake activity with rare occurrences of large earthquakes. While large earthquakes are of interest for assessing seismic hazard, little data exists from the region to empirically quantify their effects. Therefore, empirically based approaches that are used for other regions, such as Western North America, are not appropriate for Eastern North America. Moreover, recent advances in science and technology have now made it possible to combine theoretical and empirical methods to develop new procedures and models for estimating ground motion. The focus of the report is on the attributes of ground motion in Eastern North America that are of interest for the design of facilities such as nuclear power plants. Specifically considered are magnitudes M from 5 to 8, distances from 0 to 500 km, and frequencies from 1 to 35 Hz. This document, Volume IV, provides Appendix 8.B, Laboratory Investigations of Dynamic Properties of Reference Sites

System requirements and design description for the document basis database interface (DocBasis)

International Nuclear Information System (INIS)

Lehman, W.J.

1997-01-01

This document describes system requirements and the design description for the Document Basis Database Interface (DocBasis). The DocBasis application is used to manage procedures used within the tank farms. The application maintains information in a small database to track the document basis for a procedure, as well as the current version/modification level and the basis for the procedure. The basis for each procedure is substantiated by Administrative, Technical, Procedural, and Regulatory requirements. The DocBasis user interface was developed by Science Applications International Corporation (SAIC)

Archaeological data as a basis for repository marker design

International Nuclear Information System (INIS)

Kaplan, M.F.

1982-10-01

This report concerns the development of a marking system for a nuclear waste repository which is very likely to survive for 10,000 years. In order to provide a background on the subject, and for the preliminary design presented in this report, a discussion is presented about the issues involved in human interference with the repository system and the communication of information. A separate chapter summarizes six ancient man-made monuments including: materials, effects of associated textual information on our understanding of the monument, and other features of the ancient monument relevant to marking a repository site. The information presented in the two chapters is used to provide the basis and rationale for a preliminary marker system design presented in a final chapter. 86 refs., 22 figs., 1 tab

Archaeological data as a basis for repository marker design

Energy Technology Data Exchange (ETDEWEB)

Kaplan, M.F.

1982-10-01

This report concerns the development of a marking system for a nuclear waste repository which is very likely to survive for 10,000 years. In order to provide a background on the subject, and for the preliminary design presented in this report, a discussion is presented about the issues involved in human interference with the repository system and the communication of information. A separate chapter summarizes six ancient man-made monuments including: materials, effects of associated textual information on our understanding of the monument, and other features of the ancient monument relevant to marking a repository site. The information presented in the two chapters is used to provide the basis and rationale for a preliminary marker system design presented in a final chapter. 86 refs., 22 figs., 1 tab.

A risk-informed framework for establishing a beyond design basis safety basis for external hazards

Energy Technology Data Exchange (ETDEWEB)

Amico, P. [Hughes Associates, Inc, Baltimore, MD (United States); Anoba, R. [Hughes Associates, Inc, Raleigh, NC (United States); Najafi, B. [Hughes Associates, Inc., Los Gatos, CA (United States)

2014-07-01

The events at Fukushima Daiichi taught us that meeting a deterministic design basis requirement for external hazards does not assure that the risk is low. As observed at the plant, the two primary reasons for this are failure cliffs above the design basis event and that combined hazard effects are not considered in design. Because the possible combinations of design basis exceedences and external hazard combinations are very large and complex, an approach focusing only on the most important ones is needed. For this reason, a risk informed approach is the most effective approach, which is discussed in this paper. (author)

Basic repository environmental assessment design basis, Richton Dome site

International Nuclear Information System (INIS)

1988-01-01

This study examines the engineering factors and costs associated with the construction, operation, and decommissioning of a high-level nuclear waste repository in salt in the Gulf Interior Region at Richton Dome in Perry County, Mississippi. The study assumes a repository capacity of 36,000 metric tons of heavy metal (MTHM) of unreprocessed spent fuel and 36,000 MTHM of commercial high-level reprocessing waste, along with 7,020 canisters of defense high-level reprocessing waste and associated quantities of remote-and contact-handled transuranic waste (TRU). With the exception of TRU, all the waste forms are placed in 300- to 1,000-year-life carbon-steel waste packages in a collocated waste handling and packaging facility (WHPF), which is also described. The construction, operation, and decommissioning of the proposed repository is estimated to cost approximately $4.49 billion. Costs include those for the WHPF, engineering, and contingency, but exclude waste form assembly and shipment to the site and waste package fabrication and shipment to the site. These costs reflect the relative average wage rates of the region and the relatively easy access to the site. Construction would require an estimated 6.25 years. Engineering factors and costs are not strongly influenced by environmental considerations. 52 refs., 24 figs., 20 tabs

Basic repository environmental assessment design basis, Davis Canyon site

International Nuclear Information System (INIS)

1984-01-01

This study examines the engineering factors and costs associated with the construction, operation, and decommissioning of a high-level nuclear waste repository in salt in the Paradox Basin in Davis Canyon, Utah. The study assumes a repository capacity of 36,000 metric tons of heavy metal (MTHM) of unreprocessed spent fuel and 36,000 MTHM of commercial high-level reprocessing waste, along with 7,020 canisters of defense high-level reprocessing waste and associated quantities of remote- and contact-handled transuranic waste (TRU). With the exception of TRU, all the waste forms are placed in 300- to 1,000-year-life carbon-steel waste packages in a collected waste handling and packaging facility (WHPF), which is also described. The construction, operation, and decommissioning of the proposed repository is estimated to cost approximately $5.49 billion. Costs include those for the collocated WHPF, engineering, and contingency, but exclude waste form assembly and shipment to the site and waste package fabrication and shipment to the site. These costs reflect the relative average wage rates of the region and the relatively sound nature of the salt at this site. Construction would require an estimated 7.75 years. Engineering factors and costs are not strongly influenced by environmental considerations. 50 refs., 24 figs., 20 tabs

Basic repository environmental assessment design basis, Lavender Canyon site

International Nuclear Information System (INIS)

1988-01-01

This study examines the engineering factors and costs associated with the construction, operation, and decommissioning of a high-level nuclear waste repository in salt in the Paradox Basin in Lavender Canyon, Utah. The study assumes a repository capacity of 36,000 metric tons of heavy metal (MTHM) of unreprocessed spent fuel and 36,000 MTHM of commercial high-level reprocessing waste, along with 7020 canisters of defense high-level reprocessing waste and associated quantities of remote- and contact-handled transuranic waste (TRU). With the exception of TRU, all the waste forms are placed in 300- to 1000-year-life carbon-steel waste packages in a collocated waste handling and packaging facility (WHPF), which is also described. The construction, operation, and decommissioning of the proposed repository is estimated to cost approximately $5.51 billion. Costs include those for the collocated WHPP, engineering, and contingency, but exclude waste form assembly and shipment to the site and waste package fabrication and shipment to the site. These costs reflect the relative average wage rates of the region and the relatively sound nature of the salt at this site. Construction would require an estimated 7.75 years. Engineering factors and costs are not strongly influenced by environmental considerations. 51 refs., 24 figs., 20 tabs

Design basis programs and improvements in plant operation

International Nuclear Information System (INIS)

Metcalf, M.F.

1991-01-01

Public Service Electric and Gas (PSE and G) Company operates three commercial nuclear power plants in southern New Jersey. The three plants are of different designs and vintages (two pressurized water reactors licensed in 1976 and 1980 and one boiling water reactor licensed in 1986). As the industry recognized the need to develop design basis programs, PSE and G also realized the need after a voluntary 52-day shutdown of one unit because of electrical design basis problems. In its drive to be a premier electric utility, PSE and G has been aggressively active in developing design basis documents (DBDs) with supporting projects and refined uses to obtain the expected value and see the return on investment. Progress on Salem is nearly 75% complete, while Hope Creek is 20% complete. To data, PSE and G has experienced success in the use of DBDs in areas such as development of plant modifications, development of the reliability-centered maintenance program, procedure upgrades, improved document retrieval, resolution of regulatory issues, and training. The paper examines the design basis development process, supporting projects, and expected improvements in plant operations as a result of these efforts

Development of Probabilistic Design Basis Earthquake (DBE) Parameters for Moderate and High Hazard Facilities at INEEL

International Nuclear Information System (INIS)

Payne, S. M.; Gorman, V. W.; Jensen, S. A.; Nitzel, M. E.; Russell, M. J.; Smith, R. P.

2000-01-01

Design Basis Earthquake (DBE) horizontal and vertical response spectra are developed for moderate and high hazard facilities or Performance Categories (PC) 3 and 4, respectively, at the Idaho National Engineering and Environmental Laboratory (INEEL). The probabilistic DBE response spectra will replace the deterministic DBE response spectra currently in the U.S. Department of Energy Idaho Operations Office (DOE-ID) Architectural Engineering Standards that govern seismic design criteria for several facility areas at the INEEL. Probabilistic DBE response spectra are recommended to DOE Naval Reactors for use at the Naval Reactor Facility at INEEL. The site-specific Uniform Hazard Spectra (UHS) developed by URS Greiner Woodward Clyde Federal Services are used as the basis for developing the DBE response spectra. In 1999, the UHS for all INEEL facility areas were recomputed using more appropriate attenuation relationships for the Basin and Range province. The revised UHS have lower ground motions than those produced in the 1996 INEEL site-wide probabilistic ground motion study. The DBE response spectra were developed by incorporating smoothed broadened regions of the peak accelerations, velocities, and displacements defined by the site-specific UHS. Portions of the DBE response spectra were adjusted to ensure conservatism for the structural design process

Early Site Permit Demonstration Program: Guidelines for determining design basis ground motions. Volume 2, Appendices

Energy Technology Data Exchange (ETDEWEB)

1993-03-18

This report develops and applies a methodology for estimating strong earthquake ground motion. The motivation was to develop a much needed tool for use in developing the seismic requirements for structural designs. An earthquake`s ground motion is a function of the earthquake`s magnitude, and the physical properties of the earth through which the seismic waves travel from the earthquake fault to the site of interest. The emphasis of this study is on ground motion estimation in Eastern North America (east of the Rocky Mountains), with particular emphasis on the Eastern United States and southeastern Canada. Eastern North America is a stable continental region, having sparse earthquake activity with rare occurrences of large earthquakes. While large earthquakes are of interest for assessing seismic hazard, little data exists from the region to empirically quantify their effects. The focus of the report is on the attributes of ground motion in Eastern North America that are of interest for the design of facilities such as nuclear power plants. This document, Volume II, contains Appendices 2, 3, 5, 6, and 7 covering the following topics: Eastern North American Empirical Ground Motion Data; Examination of Variance of Seismographic Network Data; Soil Amplification and Vertical-to-Horizontal Ratios from Analysis of Strong Motion Data From Active Tectonic Regions; Revision and Calibration of Ou and Herrmann Method; Generalized Ray Procedure for Modeling Ground Motion Attenuation; Crustal Models for Velocity Regionalization; Depth Distribution Models; Development of Generic Site Effects Model; Validation and Comparison of One-Dimensional Site Response Methodologies; Plots of Amplification Factors; Assessment of Coupling Between Vertical & Horizontal Motions in Nonlinear Site Response Analysis; and Modeling of Dynamic Soil Properties.

Basic repository environmental assessment design basis: Deaf Smith County site

International Nuclear Information System (INIS)

1988-03-01

This study examines the engineering factors and costs associated with the construction, operation, and decommissioning of a high-level nuclear waste repository in salt in the Palo Duro Basin in Deaf Smith County, Texas. The study assumes a repository capacity of 36,000 metric tons of heavy metal (MTHM) of unreprocessed spent fuel and 36,000 MTHM of commercial high-level reprocessing waste, along with 7,020 canisters of defense high-level reprocessing waste and associated quantities or remote- and contact-handled transuranic waste (TRU). With the exception of TRU, all the waste forms are placed in 300- to 1000-year-life carbon-steel waste packages in a collocated waste handling and packaging facility (WHPF), which is also described. The construction, operation, and decommissioning of the proposed repository is estimated to cost approximately $4.64 billion. Costs include those for the collocate WHPF, engineering, and contingency, but exclude waste form assembly and shipment to the site and waste package fabrication and shipment to the site. These costs reflect the relative average wage rates of the region, the relatively easy access to the site, and the relatively weak nature of the salt at this site. Construction would require an estimated 7 to 7.5 years. Engineering factors and costs are not strongly influenced by environmental considerations. 62 refs., 24 figs., 20 tabs

Study of seismic design bases and site conditions for nuclear power plants

International Nuclear Information System (INIS)

1980-04-01

This report presents the results of an investigation of four topics pertinent to the seismic design of nuclear power plants: Design accelerations by regions of the continental United States; review and compilation of design-basis seismic levels and soil conditions for existing nuclear power plants; regional distribution of shear wave velocity of foundation materials at nuclear power plant sites; and technical review of surface-founded seismic analysis versus embedded approaches

Study of seismic design bases and site conditions for nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

1980-04-01

This report presents the results of an investigation of four topics pertinent to the seismic design of nuclear power plants: Design accelerations by regions of the continental United States; review and compilation of design-basis seismic levels and soil conditions for existing nuclear power plants; regional distribution of shear wave velocity of foundation materials at nuclear power plant sites; and technical review of surface-founded seismic analysis versus embedded approaches.

Basis for NGNP Reactor Design Down-Selection

Energy Technology Data Exchange (ETDEWEB)

L.E. Demick

2010-08-01

The purpose of this paper is to identify the extent of technology development, design and licensing maturity anticipated to be required to credibly identify differences that could make a technical choice practical between the prismatic and pebble bed reactor designs. This paper does not address a business decision based on the economics, business model and resulting business case since these will vary based on the reactor application. The selection of the type of reactor, the module ratings, the number of modules, the configuration of the balance of plant and other design selections will be made on the basis of optimizing the Business Case for the application. These are not decisions that can be made on a generic basis.

Remedial action plan and site design for stabilization of the inactive uranium mill tailings site at Lowman, Idaho: Attachment 2, Geology report

International Nuclear Information System (INIS)

1991-09-01

Detailed investigations of geologic, geomorphic, and seismic conditions at the Lowman site in central Idaho were conducted by the Technical Assistance Contractor. The purpose of these investigations was basic site characterization and the identification of potential geologic hazards that could affect long-term site stability. Subsequent engineering studies (e.g., analyses of the hydrologic regime and liquefaction potential) use this data . The geomorphic analysis is employed in the design of effective erosion protection. Studies of the regional and local seismotectonic setting, which included a detailed search for possible capable faults within a 65-km (40-mile) radius of the site, provided the basis for estimating seismic design parameters

Design basis document open-item resolution and reportability

International Nuclear Information System (INIS)

Gambhir, S.K.; Livingston, B.R.; Purcell, J.J.; Erickson, E.A.

1989-01-01

In the process of reconstituting the design bases for older nuclear power plants, information or references may not be available to fully define the design requirements or to document and verify the adequacy of the design. Also, information that is in conflict with other data is identified. The missing and conflicting information must be reconstituted in order to adequately document the design bases of the plant. For these operating facilities, the identification, tracking, and resolution of missing or conflicting information is very important when the reporting requirements stipulated by 10CFR21, 10CFR50.72, and 10CFR50.73 are considered. Additionally, controlled documentation (calculations, drawings, etc.) used to develop the design basis documents may contain conflicting data. In some cases, conflicts between the as-built design and licensing or design basis requirements established in specific commitments to the U.S. Nuclear Regulatory Commission may be identified. Furthermore, concerns regarding the adequacy of safety-related systems or components to perform their required function may be identified that would warrant prompt action by the licensee. The approach discussed in this paper was used by Omaha Public Power District for the ongoing design basis reconstitution effort at the Fort Calhoun nuclear plant

Ground motion following selection of SRS design basis earthquake and associated deterministic approach

International Nuclear Information System (INIS)

1991-03-01

This report summarizes the results of a deterministic assessment of earthquake ground motions at the Savannah River Site (SRS). The purpose of this study is to assist the Environmental Sciences Section of the Savannah River Laboratory in reevaluating the design basis earthquake (DBE) ground motion at SRS during approaches defined in Appendix A to 10 CFR Part 100. This work is in support of the Seismic Engineering Section's Seismic Qualification Program for reactor restart

Basic repository environment assessment design basis, Cypress Creek Dome Site

International Nuclear Information System (INIS)

1988-03-01

This study examines the engineering factors and costs associated with the construction, operation, and decommissioning of a high-level nuclear waste repository in salt in the Gulf Interior Region at Cypress Creek Cone, Mississippi. The study assumes a repository capacity of 36,000 metric tons of heavy metal (MTHM) of unreprocessed spent fuel and 36,000 MTHM of commercial high-level reprocessing waste, along with 7020 canisters of defense high-level reprocessing waste and associated quantities of remote- and contact-handled transuranic waste (TRU). With the exception of TRU, all the waste forms are placed in 300- to 1000-year-life carbon-steel waste packages in a collocated waste handling and packaging facility (WHPF), which is also described. The construction, operation, and decommissioning of the proposed repository is estimated to cost approximately $4.66 billion. Costs include those for the collocated WHPF, engineering, and contingency, but exclude waste from assembly and shipment to the site and waste package fabrication and shipment to the site. These costs reflect the relatively easy access to the site. Construction would require an estimated 7 years. Engineering factors and costs are not strongly influenced by environmental considerations. 53 refs., 24 figs., 10 tabs

Integral Monitored Retrievable Storage (MRS) Facility conceptual basis for design

International Nuclear Information System (INIS)

1985-10-01

The purpose of the Conceptual Basis for Design is to provide a control document that establishes the basis for executing the conceptual design of the Integral Monitored Retrievable Storage (MRS) Facility. This conceptual design shall provide the basis for preparation of a proposal to Congress by the Department of Energy (DOE) for construction of one or more MRS Facilities for storage of spent nuclear fuel, high-level radioactive waste, and transuranic (TRU) waste. 4 figs., 25 tabs

Establishing 'design basis threat' in Norway

International Nuclear Information System (INIS)

Maerli, M.B.; Naadland, E.; Reistad, O.

2002-01-01

Full text: INFCIRC 225 (Rev. 4) assumes that a state's physical protection system should be based on the state's evaluation of the threat, and that this should be reflected in the relevant legislation. Other factors should also be considered, including the state's emergency response capabilities and the existing and relevant measures of the state's system of accounting for and control of nuclear material. A design basis threat developed from an evaluation by the state of the threat of unauthorized removal of nuclear material and of sabotage of nuclear material and nuclear facilities is an essential element of a state's system of physical protection. The state should continuously review the threat, and evaluate the implications of any changes in that threat for the required levels and the methods of physical protection. As part of a national design basis threat assessment, this paper evaluates the risk of nuclear or radiological terrorism and sabotage in Norway. Possible scenarios are presented and plausible consequences are discussed with a view to characterize the risks. The need for more stringent regulatory requirements will be discussed, together with the (positive) impact of improved systems and procedures of physical protection on nuclear emergency planning. Special emphasis is placed on discussing the design basis threat for different scenarios in order to systemize regulatory efforts to update the current legislation, requirement for operators' contingency planning, response efforts and the need for emergency exercises. (author)

Simulant Basis for the Standard High Solids Vessel Design

Energy Technology Data Exchange (ETDEWEB)

Peterson, Reid A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Fiskum, Sandra K. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Suffield, Sarah R. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Daniel, Richard C. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Gauglitz, Phillip A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Wells, Beric E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

2017-09-30

The Waste Treatment and Immobilization Plant (WTP) is working to develop a Standard High Solids Vessel Design (SHSVD) process vessel. To support testing of this new design, WTP engineering staff requested that a Newtonian simulant and a non-Newtonian simulant be developed that would represent the Most Adverse Design Conditions (in development) with respect to mixing performance as specified by WTP. The majority of the simulant requirements are specified in 24590-PTF-RPT-PE-16-001, Rev. 0. The first step in this process is to develop the basis for these simulants. This document describes the basis for the properties of these two simulant types. The simulant recipes that meet this basis will be provided in a subsequent document.

Selection of design basis event for modular high temperature gas-cooled reactor

International Nuclear Information System (INIS)

Sato, Hiroyuki; Nakagawa, Shigeaki; Ohashi, Hirofumi

2016-06-01

Japan Atomic Energy Agency (JAEA) has been investigating safety requirements and basic approach of safety guidelines for modular High Temperature Gas-cooled Reactor (HTGR) aiming to increase internarial contribution for nuclear safety by developing an international HTGR safety standard under International Atomic Energy Agency. In this study, we investigate a deterministic approach to select design basis events utilizing information obtained from probabilistic approach. In addition, selections of design basis events are conducted for commercial HTGR designed by JAEA. As a result, an approach for selecting design basis event considering multiple failures of safety systems is established which has not been considered as design basis in the safety guideline for existing nuclear facility. Furthermore, selection of design basis events for commercial HTGR has completed. This report provides an approach and procedure for selecting design basis events of modular HTGR as well as selected events for the commercial HTGR, GTHTR300. (author)

ITER technical basis

Energy Technology Data Exchange (ETDEWEB)

NONE

2002-01-01

Following on from the Final Report of the EDA(DS/21), and the summary of the ITER Final Design report(DS/22), the technical basis gives further details of the design of ITER. It is in two parts. The first, the Plant Design specification, summarises the main constraints on the plant design and operation from the viewpoint of engineering and physics assumptions, compliance with safety regulations, and siting requirements and assumptions. The second, the Plant Description Document, describes the physics performance and engineering characteristics of the plant design, illustrates the potential operational consequences foe the locality of a generic site, gives the construction, commissioning, exploitation and decommissioning schedule, and reports the estimated lifetime costing based on data from the industry of the EDA parties.

ITER technical basis

International Nuclear Information System (INIS)

2002-01-01

Following on from the Final Report of the EDA(DS/21), and the summary of the ITER Final Design report(DS/22), the technical basis gives further details of the design of ITER. It is in two parts. The first, the Plant Design specification, summarises the main constraints on the plant design and operation from the viewpoint of engineering and physics assumptions, compliance with safety regulations, and siting requirements and assumptions. The second, the Plant Description Document, describes the physics performance and engineering characteristics of the plant design, illustrates the potential operational consequences foe the locality of a generic site, gives the construction, commissioning, exploitation and decommissioning schedule, and reports the estimated lifetime costing based on data from the industry of the EDA parties

Early Site Permit Demonstration Program: Guidelines for determining design basis ground motions

International Nuclear Information System (INIS)

1993-01-01

This report develops and applies a methodology for estimating strong earthquake ground motion. The motivation was to develop a much needed tool for use in developing the seismic requirements for structural designs. An earthquake's ground motion is a function of the earthquake's magnitude, and the physical properties of the earth through which the seismic waves travel from the earthquake fault to the site of interest. The emphasis of this study is on ground motion estimation in Eastern North America (east of the Rocky Mountains), with particular emphasis on the Eastern United States and southeastern Canada. Eastern North America is a stable continental region, having sparse earthquake activity with rare occurrences of large earthquakes. While large earthquakes are of interest for assessing seismic hazard, little data exists from the region to empirically quantify their effects. Therefore, empirically based approaches that are used for other regions, such as Western North America, are not appropriate for Eastern North America. Moreover, recent advances in science and technology have now made it possible to combine theoretical and empirical methods to develop new procedures and models for estimating ground motion. The focus of the report is on the attributes of ground motion in Eastern North America that are of interest for the design of facilities such as nuclear power plants. Specifically considered are magnitudes M from 5 to 8, distances from 0 to 500 km, and frequencies from 1 to 35 Hz

Seismic methodology in determining basis earthquake for nuclear installation

International Nuclear Information System (INIS)

Ameli Zamani, Sh.

2008-01-01

Design basis earthquake ground motions for nuclear installations should be determined to assure the design purpose of reactor safety: that reactors should be built and operated to pose no undue risk to public health and safety from earthquake and other hazards. Regarding the influence of seismic hazard to a site, large numbers of earthquake ground motions can be predicted considering possible variability among the source, path, and site parameters. However, seismic safety design using all predicted ground motions is practically impossible. In the determination of design basis earthquake ground motions it is therefore important to represent the influences of the large numbers of earthquake ground motions derived from the seismic ground motion prediction methods for the surrounding seismic sources. Viewing the relations between current design basis earthquake ground motion determination and modem earthquake ground motion estimation, a development of risk-informed design basis earthquake ground motion methodology is discussed for insight into the on going modernization of the Examination Guide for Seismic Design on NPP

Linkage between site characterization program and performance and design data needs

International Nuclear Information System (INIS)

Blanchard, M.B.; Dobson, D.C.; Younker, J.L.; Voegele, M.D.

1990-01-01

The basis for identifying site-specific data needs for the Yucca Mountain site was a thorough review and evaluation of the regulatory requirements for a geologic repository. The US Nuclear Regulatory Commission's (NRC) regulation governing disposal of high-level waste in a geologic repository utilizes a multiple barrier approach that includes reliance on both engineered and natural elements. Given an understanding of the regulatory requirements, the next step was to identify the natural features of the site and the components of the engineered barriers considered most likely to make significant contributions to isolation of waste at Yucca Mountain. The intent was to develop a site testing program that focused on understanding those natural site characteristics and processes most likely to contribute to waste isolation; and providing site data necessary to design and assess the performance of engineered barriers. 3 refs., 3 figs

Design basis tropical cyclone for nuclear power plants

International Nuclear Information System (INIS)

1984-01-01

The general characteristics of tropical cyclones are discussed in this Safety Guide, with particular emphasis on their pressure and wind structures in the light of available data. General methods are given for the evaluation of the relevant parameters of a Probable Maximum Tropical Cyclone (PMTC), which can be used as the Design Basis Tropical Cyclone (DBTC); these parameters then serve as inputs for the derivation of a design basis surge and a design basis wind. A possible method is also given for the evaluation of the PMTC pressure and wind field based on an approach valid primarily for a particular region. This method depends on the results of a theoretical study on the tropical cyclone structure and makes use of a large amount of data, including aircraft reconnaissance observations for 170 most intense tropical cyclones near the coast of Japan, Taiwan and the Philippines for the period 1960-1974, as well as detailed analyses of all the extreme storms along the Gulf of Mexico and the east coast of the USA during 1900-1978, for the determination of the necessary parameters

Remedial Action Plan and site design for stabilization of the inactive uranium mill tailings site at Durango, Colorado: Attachment 2, Geology report

International Nuclear Information System (INIS)

1991-12-01

Detailed investigations of geologic, geomorphic, and seismic conditions at the Bodo Canyon disposal site were conducted. The purpose of these investigations was basic site characterization and identification of potential geologic hazards that could affect long-term site stability. Subsequent engineering studies, such as analyses of hydrologic and liquefaction hazards, used the data developed in these studies. The geomorphic analysis was employed in the design of effective erosion protection. Studies of the regional and local seismotectonic setting, which included a detailed search for possible capable faults within a 65 kilometer radius of the site, provided the basis for seismic design parameters. The scope of work performed included the following: Compilation and analysis of previous published and unpublished geologic literature and maps; Review of historical and instrumental earthquake data; Review of site-specific subsurface geologic data, including lithologic and geophysical logs of exploratory boreholes advanced in the site area; Photogeologic interpretations of existing conventional aerial photographs; and, Ground reconnaissance and mapping of the site region

Assessment of WWER fuel condition in design basis accident

International Nuclear Information System (INIS)

Bibilashvili, Yu.; Sokolov, N.; Andreeva-Andrievskaya, L.; Vlasov, Yu.; Nechaeva, O.; Salatov, A.

1994-01-01

The fuel behaviour in design basis accidents is assessed by means of the verified code RAPTA-5. The code uses a set of high temperature physico-chemical properties of the fuel components as determined for commercially produced materials, fuel rod simulators and fuel rod bundles. The WWER fuel criteria available in Russia for design basis accidents do not generally differ from the similar criteria adopted for PWR's. 12 figs., 11 refs

46 CFR 177.310 - Satisfactory service as a design basis.

Science.gov (United States)

2010-10-01

... 46 Shipping 7 2010-10-01 2010-10-01 false Satisfactory service as a design basis. 177.310 Section... (UNDER 100 GROSS TONS) CONSTRUCTION AND ARRANGEMENT Hull Structure § 177.310 Satisfactory service as a design basis. When scantlings for the hull, deckhouse, and frames of the vessel differ from those...

Preliminary Evaluation Methodology of ECCS Performance for Design Basis LOCA Redefinition

International Nuclear Information System (INIS)

Kang, Dong Gu; Ahn, Seung Hoon; Seul, Kwang Won

2010-01-01

To improve their existing regulations, the USNRC has made efforts to develop the risk-informed and performance-based regulation (RIPBR) approaches. As a part of these efforts, the rule revision of 10CFR50.46 (ECCS Acceptance Criteria) is underway, considering some options for 4 categories of spectrum of break sizes, ECCS functional reliability, ECCS evaluation model, and ECCS acceptance criteria. Since the potential for safety benefits and unnecessary burden reduction from design basis LOCA redefinition is high relative to other options, the USNRC is proceeding with the rulemaking for design basis LOCA redefinition. An instantaneous break with a flow rate equivalent to a double ended guillotine break (DEGB) of the largest primary piping system in the plant is widely recognized as an extremely unlikely event, while redefinition of design basis LOCA can affect the existing regulatory practices and approaches. In this study, the status of the design basis LOCA redefinition and OECD/NEA SMAP (Safety Margin Action Plan) methodology are introduced. Preliminary evaluation methodology of ECCS performance for LOCA is developed and discussed for design basis LOCA redefinition

Assessment of WWER fuel condition in design basis accident

Energy Technology Data Exchange (ETDEWEB)

Bibilashvili, Yu; Sokolov, N; Andreeva-Andrievskaya, L; Vlasov, Yu; Nechaeva, O; Salatov, A [Vsesoyuznyj Nauchno-Issledovatel` skij Inst. Neorganicheskikh Materialov, Moscow (Russian Federation)

1994-12-31

The fuel behaviour in design basis accidents is assessed by means of the verified code RAPTA-5. The code uses a set of high temperature physico-chemical properties of the fuel components as determined for commercially produced materials, fuel rod simulators and fuel rod bundles. The WWER fuel criteria available in Russia for design basis accidents do not generally differ from the similar criteria adopted for PWR`s. 12 figs., 11 refs.

Characterisation of Liquefaction Effects for Beyond-Design Basis Safety Assessment of Nuclear Power Plants

Science.gov (United States)

Bán, Zoltán; Győri, Erzsébet; János Katona, Tamás; Tóth, László

2015-04-01

Preparedness of nuclear power plants to beyond design base external effects became high importance after 11th of March 2011 Great Tohoku Earthquakes. In case of some nuclear power plants constructed at the soft soil sites, liquefaction should be considered as a beyond design basis hazard. The consequences of liquefaction have to be analysed with the aim of definition of post-event plant condition, identification of plant vulnerabilities and planning the necessary measures for accident management. In the paper, the methodology of the analysis of liquefaction effects for nuclear power plants is outlined. The case of Nuclear Power Plant at Paks, Hungary is used as an example for demonstration of practical importance of the presented results and considerations. Contrary to the design, conservatism of the methodology for the evaluation of beyond design basis liquefaction effects for an operating plant has to be limited to a reasonable level. Consequently, applicability of all existing methods has to be considered for the best estimation. The adequacy and conclusiveness of the results is mainly limited by the epistemic uncertainty of the methods used for liquefaction hazard definition and definition of engineering parameters characterizing the consequences of liquefaction. The methods have to comply with controversial requirements. They have to be consistent and widely accepted and used in the practice. They have to be based on the comprehensive database. They have to provide basis for the evaluation of dominating engineering parameters that control the post-liquefaction response of the plant structures. Experience of Kashiwazaki-Kariwa plant hit by Niigata-ken Chuetsu-oki earthquake of 16 July 2007 and analysis of site conditions and plant layout at Paks plant have shown that the differential settlement is found to be the dominating effect in case considered. They have to be based on the probabilistic seismic hazard assessment and allow the integration into logic

Japan Catastrophic Earthquake and Tsunami in Fukushima Daiichi NPP; Is it Beyond Design Basis Accident or a Design Deficiency and Operator Unawareness?

International Nuclear Information System (INIS)

Gaafar, M.A.; Refeat, R.M.; EL-Kady, A.A.

2012-01-01

On March 11, 2011 a catastrophic earthquake and tsunami struck the north east coast of Japan. This catastrophe damaged fully or partially the six units of the Fukushima Daiichi Nuclear Power Plant.Questions were raised following the aftermath, whether it is beyond design basis accident caused by severe natural event or a failure by the Japanese authorities to plan to deal with such accident. There are many indications that the Utility of Fukushima Daiichi NPP, Tokyo Electric Power Company (TEPCO), did not pay enough attention to numerous facts about the incompatibility of the site and several design defects in the plant units. In fact there are three other NPP sites nearby Fukushima Daiichi Plant (about 30 to 60 Km far from Fukushima Daiichi NPP), with different site characteristics, which survived the same catastrophic earthquake and tsunami, but they were automatically turned into a safe shutdown state. These plants sites are Fukushima Daini Plant (4 units), Onagawa Plant (3 units) and Tokai Daini (II) Plant (one unit). In this paper, the aftermath Fukushima Daiichi plant integrity is pointed out. Some facts about the site and design concerns which could have implications on the accident are discussed. The response of Japan Authority is outlined and some remarks about their actions are underlined. The impacts of this disaster on the Nuclear Power Program worldwide are also discussed.

Technical basis for the ITER-FEAT outline design

Energy Technology Data Exchange (ETDEWEB)

NONE

2000-11-01

This ITER EDA Documentation Series issue summarizes the results of the ITER Engineering Design Activities on the technical basis for the ITER-FEAT outline design. This issue also comprises some physical analysis activities as well as structure and goals of the Physics Expert Group activities.

Technical basis for the ITER-FEAT outline design

International Nuclear Information System (INIS)

2000-01-01

This ITER EDA Documentation Series issue summarizes the results of the ITER Engineering Design Activities on the technical basis for the ITER-FEAT outline design. This issue also comprises some physical analysis activities as well as structure and goals of the Physics Expert Group activities

The Hifrensa Site of Vandellos I NPP, Basis for a New Future

International Nuclear Information System (INIS)

Albet, V.

2000-01-01

This article presents the HIFRENSA Site in Vandellos as the basis for a new future and explains the process developed to achieve this new reality. This includes a brief description of the site, an explanation of the analysis that has been carried out on installation reuse, and a review of the potential uses. Finally, it reflects on the strategic value of the Site that gives meaning to all the business efforts and actions undertaken by HIFRENSA to manage the future value of this asset both for the company itself and for society in general. (Author)

The power of simplification: Operator interface with the AP1000R during design-basis and beyond design-basis events

International Nuclear Information System (INIS)

Williams, M. G.; Mouser, M. R.; Simon, J. B.

2012-01-01

The AP1000 R plant is an 1100-MWe pressurized water reactor with passive safety features and extensive plant simplifications that enhance construction, operation, maintenance, safety and cost. The passive safety features are designed to function without safety-grade support systems such as component cooling water, service water, compressed air or HVAC. The AP1000 passive safety features achieve and maintain safe shutdown in case of a design-basis accident for 72 hours without need for operator action, meeting the expectations provided in the European Utility Requirements and the Utility Requirement Document for passive plants. Limited operator actions may be required to maintain safe conditions in the spent fuel pool (SFP) via passive means. This safety approach therefore minimizes the reliance on operator action for accident mitigation, and this paper examines the operator interaction with the Human-System Interface (HSI) as the severity of an accident increases from an anticipated transient to a design basis accident and finally, to a beyond-design-basis event. The AP1000 Control Room design provides an extremely effective environment for addressing the first 72 hours of design-basis events and transients, providing ease of information dissemination and minimal reliance upon operator actions. Symptom-based procedures including Emergency Operating Procedures (EOPs), Abnormal Operating Procedures (AOPs) and Alarm Response Procedures (ARPs) are used to mitigate design basis transients and accidents. Use of the Computerized Procedure System (CPS) aids the operators during mitigation of the event. The CPS provides cues and direction to the operators as the event progresses. If the event becomes progressively worse or lasts longer than 72 hours, and depending upon the nature of failures that may have occurred, minimal operator actions may be required outside of the control room in areas that have been designed to be accessible using components that have been designed

Designing a responsive web site

OpenAIRE

Fejzić , Diana

2016-01-01

Due to the increasing prevalence of smartphones and tablet computers design became a crucial part of web design. For a user, responsive web design enables the best user experience, regardless of whether a user is visiting the site via a mobile phone, a tablet or a computer. This thesis covers the process of planning, designing and responsive web site development, for a fictitious company named “Creative Design d.o.o.”, with the help of web technologies. In the initial part of the thesis, w...

Information management needs for Fort Calhoun's design basis reconstitution project

International Nuclear Information System (INIS)

Beach, D.R.; Erickson, E.A.; Gambhir, S.K.; Parsons, R.D.

1989-01-01

While the need for information management is not new to the nuclear industry or Omaha Public Power District (OPPD), the interrelationship among design information, multiple systems, and design basis issues has necessitated the management of this information in new ways. The project team involved in the reconstitution of the design basis for OPPD's Fort Calhoun nuclear station has experienced the need for the developed effective methods for managing the vast amount of interrelated information associated with this effort. This management of information has been necessary to ensure that design basis documents (DBDs) adequately reflect the interrelated nature of component, system, and plant design; are complete and accurate; and are produced and maintained in a cost-effective manner. Fort Calhoun's aggressive design basis reconstitution project began in early 1987. The present scope of the project includes the production of 52 system and plant level DBDs; currently the project is ∼50% complete with DBDs in various stages of completion, from pilot DBDs through DBDs with approved formats, which have been issued for use. The experience in producing these documents has lead to a growing understanding of the special need for information management in each stage of the project. The development of the information tracking and management processes for the various stages of DBD development has proven to be cost-effective and gives a level of assurance that information has been included in the DBDs consistently and accurately

Site-specific design optimization of wind turbines

DEFF Research Database (Denmark)

Fuglsang, P.; Bak, C.; Schepers, J.G.

2002-01-01

This article reports results from a European project, where site characteristics were incorporated into the design process of wind turbines, to enable site-specific design. Two wind turbines of different concept were investigated at six different sites comprising normal flat terrain, offshore...... and complex terrain wind farms. Design tools based on numerical optimization and aeroelastic calculations were combined with a cost model to allow optimization for minimum cost of energy. Different scenarios were optimized ranging from modifications of selected individual components to the complete design...... of a new wind turbine. Both annual energy yield and design-determining loads depended on site characteristics, and this represented a potential for site-specific design. The maximum variation in annual energy yield was 37% and the maximum variation in blade root fatigue loads was 62%. Optimized site...

The power of simplification: Operator interface with the AP1000{sup R} during design-basis and beyond design-basis events

Energy Technology Data Exchange (ETDEWEB)

Williams, M. G.; Mouser, M. R.; Simon, J. B. [Westinghouse Electric Company, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

2012-07-01

The AP1000{sup R} plant is an 1100-MWe pressurized water reactor with passive safety features and extensive plant simplifications that enhance construction, operation, maintenance, safety and cost. The passive safety features are designed to function without safety-grade support systems such as component cooling water, service water, compressed air or HVAC. The AP1000 passive safety features achieve and maintain safe shutdown in case of a design-basis accident for 72 hours without need for operator action, meeting the expectations provided in the European Utility Requirements and the Utility Requirement Document for passive plants. Limited operator actions may be required to maintain safe conditions in the spent fuel pool (SFP) via passive means. This safety approach therefore minimizes the reliance on operator action for accident mitigation, and this paper examines the operator interaction with the Human-System Interface (HSI) as the severity of an accident increases from an anticipated transient to a design basis accident and finally, to a beyond-design-basis event. The AP1000 Control Room design provides an extremely effective environment for addressing the first 72 hours of design-basis events and transients, providing ease of information dissemination and minimal reliance upon operator actions. Symptom-based procedures including Emergency Operating Procedures (EOPs), Abnormal Operating Procedures (AOPs) and Alarm Response Procedures (ARPs) are used to mitigate design basis transients and accidents. Use of the Computerized Procedure System (CPS) aids the operators during mitigation of the event. The CPS provides cues and direction to the operators as the event progresses. If the event becomes progressively worse or lasts longer than 72 hours, and depending upon the nature of failures that may have occurred, minimal operator actions may be required outside of the control room in areas that have been designed to be accessible using components that have been

Emergency procedures beyond design basis ''Feed and Bleed''

International Nuclear Information System (INIS)

Dominguez Bautista, M.T.; Campuzano Pena, F.

1994-01-01

The incorporation of Beyond-Design-Basis Emergency Procedures, also called the Emergency Manual or Severe Accident Manual, has been an important step forward in nuclear power plant safety. These procedures cover situations in which the deterministic criteria used in plant design have been contravened. In such situations new accident scenarios, unforeseen system actions or a combination of both, need to be considered. Establishing these procedures is actually the last in a sequence of activities the sequence includes definition of scenarios, study of their phenomena, analysis of optional system actions, verification of their effectiveness and finally, implementation of the procedure. The systematization of these new strategies is supported by the results of the probabilistic analyses which serve in this case to pinpoint the objectives of these strategies. This paper describes the application of this methodology in the definition of a procedure for heat sink recovery on the secondary side (feed and bleed) if this has been totally or partially lost in a beyond-design-basis event. (Author)

Design-Load Basis for LANL Structures, Systems, and Components

Energy Technology Data Exchange (ETDEWEB)

I. Cuesta

2004-09-01

This document supports the recommendations in the Los Alamos National Laboratory (LANL) Engineering Standard Manual (ESM), Chapter 5--Structural providing the basis for the loads, analysis procedures, and codes to be used in the ESM. It also provides the justification for eliminating the loads to be considered in design, and evidence that the design basis loads are appropriate and consistent with the graded approach required by the Department of Energy (DOE) Code of Federal Regulation Nuclear Safety Management, 10, Part 830. This document focuses on (1) the primary and secondary natural phenomena hazards listed in DOE-G-420.1-2, Appendix C, (2) additional loads not related to natural phenomena hazards, and (3) the design loads on structures during construction.

Environmental conditions using thermal-hydraulics computer code GOTHIC for beyond design basis external events

International Nuclear Information System (INIS)

Pleskunas, R.J.

2015-01-01

In response to the Fukushima Dai-ichi beyond design basis accident in March 2011, the Nuclear Regulatory Commission (NRC) issued Order EA-12-049, 'Issuance of Order to Modify Licenses with Regard to Requirements for Mitigation Strategies Beyond-Design-Basis-External-Events'. To outline the process to be used by individual licensees to define and implement site-specific diverse and flexible mitigation strategies (FLEX) that reduce the risks associated with beyond design basis conditions, Nuclear Energy Institute document NEI 12-06, 'Diverse and Flexible Coping Strategies (FLEX) Implementation Guide', was issued. A beyond design basis external event (BDBEE) is postulated to cause an Extended Loss of AC Power (ELAP), which will result in a loss of ventilation which has the potential to impact room habitability and equipment operability. During the ELAP, portable FLEX equipment will be used to achieve and maintain safe shutdown, and only a minimal set of instruments and controls will be available. Given these circumstances, analysis is required to determine the environmental conditions in several vital areas of the Nuclear Power Plant. The BDBEE mitigating strategies require certain room environments to be maintained such that they can support the occupancy of personnel and the functionality of equipment located therein, which is required to support the strategies associated with compliance to NRC Order EA-12-049. Three thermal-hydraulic analyses of vital areas during an extended loss of AC power using the GOTHIC computer code will be presented: 1) Safety-related pump and instrument room transient analysis; 2) Control Room transient analysis; and 3) Auxiliary/Control Building transient analysis. GOTHIC (Generation of Thermal-Hydraulic Information for Containment) is a general purpose thermal-hydraulics software package for the analysis of nuclear power plant containments, confinement buildings, and system components. It is a volume/path/heat sink

Remedial Action Plan and site design for stabilization of the inactive uranium mill tailings site at Durango, Colorado: Attachment 2, Geology report. Revised final report

Energy Technology Data Exchange (ETDEWEB)

1991-12-01

Detailed investigations of geologic, geomorphic, and seismic conditions at the Bodo Canyon disposal site were conducted. The purpose of these investigations was basic site characterization and identification of potential geologic hazards that could affect long-term site stability. Subsequent engineering studies, such as analyses of hydrologic and liquefaction hazards, used the data developed in these studies. The geomorphic analysis was employed in the design of effective erosion protection. Studies of the regional and local seismotectonic setting, which included a detailed search for possible capable faults within a 65 kilometer radius of the site, provided the basis for seismic design parameters. The scope of work performed included the following: Compilation and analysis of previous published and unpublished geologic literature and maps; Review of historical and instrumental earthquake data; Review of site-specific subsurface geologic data, including lithologic and geophysical logs of exploratory boreholes advanced in the site area; Photogeologic interpretations of existing conventional aerial photographs; and, Ground reconnaissance and mapping of the site region.

Design basis of off-site emergency response plans for fuel cycle installations

International Nuclear Information System (INIS)

Rzepka, J.P.; Dubiau, Ph.; Jouve, A.C.; Charles, T.; Mercier, J.P.

1995-01-01

In France, the term 'off-site emergency response plan' refers to all the arrangements which should be made by the government authorities to protect the population in the event of an accident affecting the installations of the site considered. The outline of the method of defining typical accidents, evaluation of 'source-terms' and health consequences is presented. Two applications to installations from the front-end and from the back-end of the fuel cycle are discussed. (K.A.). 1 tab

Configuration management after design basis reconstitution

International Nuclear Information System (INIS)

Purcell, J.J.; Livingston, B.R.

1991-01-01

Over the last few years, Fort Calhoun station (FCS) has implemented a number of programs to enhance plant operability and readiness. The design basis document (DBD) reconstitution project was the cornerstone of this effort. Vendor manual upgrade, operating procedures upgrade, plant equipment data-base verification, equipment labeling, and warehousing improvements were also implemented as part of this improvement program. With the completion of these programs, plant documentation was current to the baselines established by each program, and a configuration management program (CMP) was established to maintain this level of accuracy throughout the remaining life of FCS. Change control throughout the organization has been reviewed and upgraded to ensure that all changes are evaluated for impact to the design bases

Design basis flood for nuclear power plants on coastal sites

International Nuclear Information System (INIS)

1983-01-01

This Guide discusses the phenomena causing coastal floods (storm surge, seiche, tsunami and wind-wave) and gives a general description of the methods used and the critical factors involved in the evaluation of such floods and of their associated effects. In addition, some treatment is presented of the possible combinations of two or more of these phenomena to produce a DBF. Methods are also provided for evaluating the reference water levels, taking into account the effect of tides, sea level anomalies and changes in lake level and river flow. Sites vulnerable to coastal flooding are located on open coastal regions, semi-enclosed bodies of water and enclosed bodies of water. Open coastal regions are those portions of land directly exposed to and having a shore on a major body of water. Semi-enclosed bodies of water are lagoons, river estuaries, gulfs, fjords and rias. Enclosed bodies of water are lakes and reservoirs. The phenomena of the lowering of the water level at coastal sites caused by offshore winds, low tides, wave effects or of drawdown caused by tsunamis are discussed. The static and dynamic effects of floods resulting from the various combinations (independent and interdependent) of surface waves of varying frequency are also discussed. Consideration is also given to shoreline instabilities and to the effects of erosion. Estimated flood levels and related effects on the nuclear power plant, which will vary according to the method of analysis and the type of flooding considered, shall be compared with available historical data where this is relevant, to check the conservativeness of the evaluated results

Data base pertinent to earthquake design basis

International Nuclear Information System (INIS)

Sharma, R.D.

1988-01-01

Mitigation of earthquake risk from impending strong earthquakes is possible provided the hazard can be assessed, and translated into appropriate design inputs. This requires defining the seismic risk problem, isolating the risk factors and quantifying risk in terms of physical parameters, which are suitable for application in design. Like all other geological phenomena, past earthquakes hold the key to the understanding of future ones. Quantificatio n of seismic risk at a site calls for investigating the earthquake aspects of the site region and building a data base. The scope of such investigations is il lustrated in Figure 1 and 2. A more detailed definition of the earthquake problem in engineering design is given elsewhere (Sharma, 1987). The present document discusses the earthquake data base, which is required to support a seismic risk evaluation programme in the context of the existing state of the art. (author). 8 tables, 10 figs., 54 refs

Instrumental basis utilizing a sodium iodide detector for radioactive soil evaluations for site remediation

International Nuclear Information System (INIS)

Davis, A.I.

1996-07-01

This document provides the design rational for using a 5.08 by 5.08 cm (2 by 2 in.) sodium iodide detector to aid in boundary mapping of the radioactive contaminated soils at the Hanford Site. The detector can provide adequate response to a homogeneously contaminated soil volume of at least 1.22 to 1.52 m in diameter and up to 45.7 cm deep. The gamma activity basis is described by a soil column in a right circular cylinder that is uniformly isotopic. The detection techniques and characteristics are used to define contaminated areas requiring excavation by locating regions where gamma activities exceed cleanup standards. To determine the areas and boundary of contamination, the boundary and adjacent surfaces will be mapped using the detector operated in a windowed gross-gamma mode

Exploratory Shaft Seismic Design Basis Working Group report

International Nuclear Information System (INIS)

Subramanian, C.V.; King, J.L.; Perkins, D.M.; Mudd, R.W.; Richardson, A.M.; Calovini, J.C.; Van Eeckhout, E.; Emerson, D.O.

1990-08-01

This report was prepared for the Yucca Mountain Project (YMP), which is managed by the US Department of Energy. The participants in the YMP are investigating the suitability of a site at Yucca Mountain, Nevada, for construction of a repository for high-level radioactive waste. An exploratory shaft facility (ESF) will be constructed to permit site characterization. The major components of the ESF are two shafts that will be used to provide access to the underground test areas for men, utilities, and ventilation. If a repository is constructed at the site, the exploratory shafts will be converted for use as intake ventilation shafts. In the context of both underground nuclear explosions (conducted at the nearby Nevada Test Site) and earthquakes, the report contains discussions of faulting potential at the site, control motions at depth, material properties of the different rock layers relevant to seismic design, the strain tensor for each of the waveforms along the shaft liners, and the method for combining the different strain components along the shaft liners. The report also describes analytic methods, assumptions used to ensure conservatism, and uncertainties in the data. The analyses show that none of the shafts' structures, systems, or components are important to public radiological safety; therefore, the shafts need only be designed to ensure worker safety, and the report recommends seismic design parameters appropriate for this purpose. 31 refs., 5 figs., 6 tabs

Guidance on the Implementation of Modifications to Mitigate Beyond Design Basis Accidents

Energy Technology Data Exchange (ETDEWEB)

Dermarkar, F.; Marczak, J.; O’Neill, M., E-mail: fred.dermarkar@opg.com [Ontario Power Generation, Pickering, Ontario (Canada)

2014-10-15

Following the events at Fukushima, Canadian Nuclear Power Plants (NPP) procured equipment and initiated modifications to improve response capability for Beyond Design Basis Accidents (BDBA). These changes were not typical of other design modifications to the nuclear power plants and reinforced the need for additional guidance for modifications to address BDBA. This paper describes the guidance that was developed to guide the design, procurement, installation, operation, and maintenance of equipment and modifications to mitigate BDBAs. The guidance developed prescribes a graded approach based on a categorization of the nature of the modification. Four categories of modifications are introduced, with the distinction being the degree of interface with existing design basis systems, structures and components (SSCs). This has resulted in a cost-effective means of implementing additional capability to mitigate BDBA conditions, and yet ensure the design basis capability of SSCs is maintained. (author)

The earthquake problem in engineering design: generating earthquake design basis information

International Nuclear Information System (INIS)

Sharma, R.D.

1987-01-01

Designing earthquake resistant structures requires certain design inputs specific to the seismotectonic status of the region, in which a critical facility is to be located. Generating these inputs requires collection of earthquake related information using present day techniques in seismology and geology, and processing the collected information to integrate it to arrive at a consolidated picture of the seismotectonics of the region. The earthquake problem in engineering design has been outlined in the context of a seismic design of nuclear power plants vis a vis current state of the art techniques. The extent to which the accepted procedures of assessing seismic risk in the region and generating the design inputs have been adherred to determine to a great extent the safety of the structures against future earthquakes. The document is a step towards developing an aproach for generating these inputs, which form the earthquake design basis. (author)

FRG conceptual design and design basis

International Nuclear Information System (INIS)

Roethemeyer, H.

1979-01-01

For the site-independent conceptual design the following requirements have been laid down: (1) for safety reasons retrievability is not considered; (2) standard mining techniques and experience gained at Asse should be used; (3) two shafts should be sufficient; (4) different waste forms and containers shall be disposed of in different storage areas; (5) ventilated sections must allow the shutting off of each storage area from the rest of the mine; (6) the mining method of retreat working should be applied; (7) the mine works shall have a lateral safety distance to the caprock of 200 m and a vertical safety zone beneath salt level of 300 m; (8) all disposal areas shall be on one level; (9) salt and waste shall be transported in different drifts, mainly in a one way system

New Reactor Siting in Finland, Hanhikivi Site in Pyhaejoki - STUK preliminary safety assessment

International Nuclear Information System (INIS)

Nevalainen, Janne

2013-01-01

STUK has performed a preliminary assessment of the Decision-in-Principle on the Fennovoima application. A variety of factors must be considered in the selection of a site, including effects of the site on the plant design and the effects of the plant on the site environment. These include external hazards, both natural and human-induced. Since this is a new site, an extensive siting process is followed, that can include an EIA. A site survey is performed to identify candidate sites, after investigating a large region and rejecting unsuitable sites. The remaining sites are then screened and compared on the basis of safety and other considerations to select one or more preferred sites. Natural hazards include geology, seismology, hydrology and meteorology. Offshore ice will be a particular hazard for this plant, since the site is on average only 1.5 m above sea level. The design basis earthquake corresponds to a return frequency of 100,000 years, with 50 % confidence. The existing sites in southern Finland used a design peak ground acceleration of 0.1 g with the ground response spectrum maximum at 10 Hz. The candidate sites in northern Finland will require a peak ground acceleration of 0.2 g with the ground response spectrum maximum at 25 Hz

Design Load Basis for Offshore Wind turbines

DEFF Research Database (Denmark)

Natarajan, Anand; Hansen, Morten Hartvig; Wang, Shaofeng

2016-01-01

DTU Wind Energy is not designing and manufacturing wind turbines and does therefore not need a Design Load Basis (DLB) that is accepted by a certification body. However, to assess the load consequences of innovative features and devices added to existing offshore turbine concepts or new offshore...... turbine concept developed in our research, it is useful to have a full DLB that follows the current design standard and is representative of a general DLB used by the industry. It will set a standard for the offshore wind turbine design load evaluations performed at DTU Wind Energy, which is aligned...... with the challenges faced by the industry and therefore ensures that our research continues to have a strong foundation in this interaction. Furthermore, the use of a full DLB that follows the current standard can improve and increase the feedback from the research at DTU Wind Energy to the international...

Guidance on the implementation of modifications to mitigate beyond design basis accidents

Energy Technology Data Exchange (ETDEWEB)

Harris, S.; Marczak, J.; O' Neill, M. [Ontario Power Generation, Pickering, ON (Canada)

2014-07-01

Following the events at Fukushima, Canadian Nuclear Power Plants (NPP) procured equipment and initiated modifications to improve response capability for Beyond Design Basis Accidents (BDBA). These changes were not typical of other design modifications to the nuclear power plants and reinforced the need for additional guidance for modifications to address BDBA. This paper describes the guidance that was developed to guide the design, procurement, installation, operation, and maintenance of equipment and modifications to mitigate BDBAs. The guidance developed prescribes a graded approach based on a categorization of the nature of the modification. Four categories of modifications are introduced, with the distinction being the degree of interface with existing design basis systems, structures and components (SSCs). This has resulted in a cost-effective means of implementing additional capability to mitigate BDBA conditions, and yet ensure the design basis capability of SSCs is maintained. Operating experience with use of the guidance is also discussed. (author)

Guidance on the implementation of modifications to mitigate beyond design basis accidents

International Nuclear Information System (INIS)

Harris, S.; Marczak, J.; O'Neill, M.

2014-01-01

Following the events at Fukushima, Canadian Nuclear Power Plants (NPP) procured equipment and initiated modifications to improve response capability for Beyond Design Basis Accidents (BDBA). These changes were not typical of other design modifications to the nuclear power plants and reinforced the need for additional guidance for modifications to address BDBA. This paper describes the guidance that was developed to guide the design, procurement, installation, operation, and maintenance of equipment and modifications to mitigate BDBAs. The guidance developed prescribes a graded approach based on a categorization of the nature of the modification. Four categories of modifications are introduced, with the distinction being the degree of interface with existing design basis systems, structures and components (SSCs). This has resulted in a cost-effective means of implementing additional capability to mitigate BDBA conditions, and yet ensure the design basis capability of SSCs is maintained. Operating experience with use of the guidance is also discussed. (author)

DART - for design basis justification and safety related information management

International Nuclear Information System (INIS)

Billington, A.; Blondiaux, P.; Boucau, J.; Cantineau, B.; Doumont, C.; Mared, A.

2000-01-01

DART is the acronym for Design Analysis Re-engineering Tool. It embodies a systematic and integrated approach to NPP safety re-assessment and configuration management, that makes use of Reverse Failure Mode and Effect Analysis in conjunction with a state-of-the-art relational database and a standardized data format, to permit long-term management of plant safety related information. The plant design is reviewed in a step-by-step logical fashion by constructing fault trees that identify the link between undesired consequences and their causes. Each failure cause identified in a fault tree is addressed by defining functional requirements, which are in turn addressed by documenting the specific manner in which the plant complies with the requirement. The database can be used to generate up-to-date plant safety related documents, including: SAR, Systems Descriptions, Technical Specifications and plant procedures. The approach is open-minded by nature and therefore is not regulatory driven, however the plant licensing basis will also be reviewed and documented within the same database such that a Regulatory Conformance Program may be integrated with the other safety documentation. This methodology can thus reconstitute the plant design bases in a comprehensive and systematic way, while allowing to uncover weaknesses in design. The original feature of the DART methodology is that it links all the safety related documents together, facilitating the evaluation of the safety impact resulting from any plant modification. Due to its capability to retrieve the basic justifications of the plant design, it is also a useful tool for training the young generation of plant personnel. The DART methodology has been developed for application to units 2, 3 and 4 at Vattenfall's Ringhals site in Sweden. It may be applied to any nuclear power plant or industrial facility where public safety is a concern. (author)

DART - for design basis justification and safety related information management

International Nuclear Information System (INIS)

Billington, A.; Blondiaux, B.; Boucau, J.; Cantineau, B.; Mared, A.

2001-01-01

DART is the acronym for Design Analysis Re-Engineering Tool. It embodies a systematic and integrated approach to NPP safety re-assessment and configuration management, that makes use of Reverse Failure Mode and Effect Analysis in conjunction with a state-of-the-art relational database and a standardized data format, to permit long-term management of plant safety related information. The plant design is reviewed in a step-by-step logical fashion by constructing fault trees that identify the link between undesired consequences and their causes. Each failure cause identified in a fault tree is addressed by defining functional requirements, which are in turn addressed by documenting the specific manner in which the plant complies with the requirement. The database can then be used to generate up-to-date plant safety related documents, including: SAR, Systems Descriptions, Technical Specifications and plant procedures. The approach is open-minded by nature and therefore is not regulatory driven, however the plant licensing basis will also be reviewed and documented within the same database such that a Regulatory Conformance Program may be integrated with the other safety documentation. This methodology can thus reconstitute the plant design bases in a comprehensive and systematic way, while allowing to uncover weaknesses in design. The original feature of the DART methodology is that it links all the safety related documents together, facilitating the evaluation of the safety impact resulting from any plant modification. Due to its capability to retrieve the basic justifications of the plant design, it is also a useful tool for training the young generation of plant personnel. The DART methodology has been developed for application to units 2, 3 and 4 at Vattenfall's Ringhals site in Sweden. It may be applied to any nuclear power plant or industrial facility where public safety is a concern. (author)

Design Status and Applications of Small reactors without On-site Refuelling

International Nuclear Information System (INIS)

Kuznetsov, V.

2006-01-01

Small reactors without on-site refuelling are the reactors that can operate without reloading and shuffling of fuel for a reasonably long period, consistent with plant economy and considerations of energy security, with no fresh or spent fuel being stored at a site during reactor operation. Such reactors could simplify the implementation of safeguards and provide certain guarantees of sovereignty to those countries that would prefer to lease fuel from a foreign vendor or, perhaps, an international fuel cycle centre. About 30 concepts of such reactors are being analyzed or developed in 6 IAEA Member States. They cover all principle reactor lines: water cooled, fast gas cooled, sodium cooled, lead or lead bismuth cooled and molten salt cooled reactors. An increased refuelling interval could be achieved with reduced core power density, burnable absorbers, or high conversion ratio. The design goals for small reactors without on-site refuelling, inter alia, include: difficult unauthorized access to fuel; design provisions to facilitate the implementation of safeguards; capability to survive all postulated accident scenarios without requiring emergency response in the public domain; economic competitiveness for anticipated market conditions and applications; the capability to achieve higher manufacturing quality through factory mass production, design standardization and common basis for design certification; and a flexibility in siting and applications. Such reactors are often considered in conjunction with fuel or NPP leasing Small reactors without on-site refuelling have many common technology development issues related to the provision of lifetime core operation, economic competitiveness, high level of safety and proliferation resistance. Reestablishment of a practice of licensing by test and establishment of legal provisions and the insurance scheme for a transit of fuel loads or factory fabricated reactors through the territory of a third country are mentioned as

Report of the Siting Policy Task Force

International Nuclear Information System (INIS)

1979-08-01

In August 1978, the Nuclear Regulatory Commission directed the staff to develop a general policy statement on nuclear power reactor siting. A Task Force was formed for that purpose and has prepared a statement of current NRC policy and practice and has recommended a number of changes to current policy. Recommendations were made to accomplish the following goals: (1) To strengthen siting as a factor in defense in depth by establishing requirements for site approval that are independent of plant design consideration. The present policy of permitting plant design features to compensate for unfavorable site characteristics has resulted in improved designs but has tended to deemphasize site isolation. (2) To take into consideration in siting the risk associated with accidents beyond the design basis (Class 9) by establishing population density and distribution criteria. Plant design improvements have reduced the probability and consequences of design basis accidents, but there remains the residual risk from accidents not considered in the design basis. Although this risk cannot be completely reduced to zero, it can be significantly reduced by selective siting. (3) To require that sites selected will minimize the risk from energy generation. The selected sites should be among the best available in the region where new generating capacity is needed. Siting requirements should be stringent enough to limit the residual risk of reactor operation but not so stringent as to eliminate the nuclear option from large regions of the country. This is because energy generation from any source has its associated risk, with risks from some energy sources being greater than that of the nuclear option

Quantum mechanical design of enzyme active sites.

Science.gov (United States)

Zhang, Xiyun; DeChancie, Jason; Gunaydin, Hakan; Chowdry, Arnab B; Clemente, Fernando R; Smith, Adam J T; Handel, T M; Houk, K N

2008-02-01

The design of active sites has been carried out using quantum mechanical calculations to predict the rate-determining transition state of a desired reaction in presence of the optimal arrangement of catalytic functional groups (theozyme). Eleven versatile reaction targets were chosen, including hydrolysis, dehydration, isomerization, aldol, and Diels-Alder reactions. For each of the targets, the predicted mechanism and the rate-determining transition state (TS) of the uncatalyzed reaction in water is presented. For the rate-determining TS, a catalytic site was designed using naturalistic catalytic units followed by an estimation of the rate acceleration provided by a reoptimization of the catalytic site. Finally, the geometries of the sites were compared to the X-ray structures of related natural enzymes. Recent advances in computational algorithms and power, coupled with successes in computational protein design, have provided a powerful context for undertaking such an endeavor. We propose that theozymes are excellent candidates to serve as the active site models for design processes.

Early Site Permit Demonstration Program: Station design alternatives report

International Nuclear Information System (INIS)

1993-03-01

This report provides the results of investigating the basis for including Station Design Alternatives (SDAs) in the regulatory guidance given for nuclear plant environmental reports (ERs), explains approaches or processes for evaluating SDAs at the early site permit (ESP) stage, and applies one of the processes to each of the ten systems or subsystems considered as SDAS. The key objective o this report s to demonstrate an adequate examination of alternatives can be performed without the extensive development f design data. The report discusses the Composite Suitability Approach and the Established Cutoff Approach in evaluating station design alternatives and selects one of these approaches to evaluate alternatives for each of the plant or station that were considered. Four types of ALWRs have been considered due to the availability of extensive plant data: System 80+, AP600, Advanced Boiling Reactor (ABWR), and Simplified Boiling Water Reactor (SBWR). This report demonstrates the feasibility of evaluating station design alternatives when reactor design detail has not been determined, quantitatively compares the potential ental impacts of alternatives, and focuses the ultimate selection of a alternative on cost and applicant-specific factors. The range of alternatives system is deliberately limited to a reasonable number to demonstrate the or to the three most commonly used at operating plants

Design basis II: Design for events

International Nuclear Information System (INIS)

Frisch, W.

1982-01-01

In a lecture of this title, it could be expected that all events which are a basis for system and component design are described. According to the title of the Course 'Instrumentation and Control of Nuclear Power Plants' emphasis is put on events originating within the plant (no consideration of external events such as air plane crash or earth-quake). The lecture is divided into the two parts 'Transients' and 'Loss of coolant accidents (LOCAs)'. Due to the complex interaction between systems and components during transients, the first part is the main part of the lecture, while the second part (LOCAs) is only a very brief description of emergency core cooling system functions and the typical course of a large and small LOCA event. The first part on anticipated transients with intact primary coolant system boundary (non-LOCA-transients) covers several aspects of the analysis, such as classification, brief system description, transient description, analysis of anticipated transients without scram (ATWS) and analytical methods. Due to the time restriction necessary within the course, only a small section of the entire area can be presented in this paper. (orig.)

Canister storage building design basis accident analysis documentation

International Nuclear Information System (INIS)

KOPELIC, S.D.

1999-01-01

This document provides the detailed accident analysis to support HNF-3553, Spent Nuclear Fuel Project Final Safety Analysis Report, Annex A, ''Canister Storage Building Final Safety Analysis Report.'' All assumptions, parameters, and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the Canister Storage Building Final Safety Analysis Report

Development of in-structure design spectra for dome mounted equipment on underground waste storage tanks at the Hanford Site

International Nuclear Information System (INIS)

Julyk, L.J.

1995-09-01

In-structure response spectra for dome mounted equipment on underground waste storage tanks at the Hanford Site are developed on the basis of recent soil-structure-interaction analyses. Recommended design spectra are provided for various locations on the tank dome

Accident beyond the design basis management with the coolant loss at the NPP with WWER

International Nuclear Information System (INIS)

Skalozubov, V.I.; Klyuchnikov, A.A.; Kolykhanov, V.N.

2010-01-01

The analysis of status and experience of development on modelling and accident beyond the design basis management, including the severe accidents, at the nuclear power plants is carried out. The methodical providing of manuals on the accident beyond the design basis management with the coolant loss on the basis of simulated critical system configurations providing the necessary safety function performance on reactor unit is proposed. The project of symptom-oriented manuals on accident beyond the design basis management with the coolant loss on the serial power unit with WWER-1000 on the basis of developed methodical providing and well known results of deepened safety analysis is presented.

Site characterization plan: Yucca Mountain site, Nevada research and development area, Nevada: Consultation draft, Nuclear Waste Policy Act

Energy Technology Data Exchange (ETDEWEB)

NONE

1988-01-01

Chapter six describes the basis for facility design, the completed facility conceptual design, the completed analytical work relating to the resolution of design issues, and future design-related work. The basis for design and the conceptual design information presented in this chapter meet the requirements of the Nuclear Waste Policy Act of 1982, for a conceptual repository design that takes into account site-specific requirements. This information is presented to permit a critical evaluation of planned site characterization activities. Chapter seven describes waste package components, emplacement environment, design, and status of research and development that support the Nevada Nuclear Waste Storage Investigation (NNWSI) Project. The site characterization plan (SCP) discussion of waste package components is contained entirely within this chapter. The discussion of emplacement environment in this chapter is limited to considerations of the environment that influence, or which may influence, if perturbed, the waste packages and their performance (particularly hydrogeology, geochemistry, and borehole stability). The basis for conceptual waste package design as well as a description of the design is included in this chapter. The complete design will be reported in the advanced conceptual design (ACD) report and is not duplicated in the SCP. 367 refs., 173 figs., 68 tabs.

Site characterization plan: Yucca Mountain site, Nevada research and development area, Nevada: Consultation draft, Nuclear Waste Policy Act

International Nuclear Information System (INIS)

1988-01-01

Chapter six describes the basis for facility design, the completed facility conceptual design, the completed analytical work relating to the resolution of design issues, and future design-related work. The basis for design and the conceptual design information presented in this chapter meet the requiremenrs of the Nuclear Waste Policy Act of 1982, for a conceptual repository design that takes into account site-specific requirements. This information is presented to permit a critical evaluation of planned site characterization activities. Chapter seven describes waste package components, emplacement environment, design, and status of research and development that support the Nevada Nuclear Waste Storage Investigation (NNWSI) Project. The site characterization plan (SCP) discussion of waste package components is contained entirely within this chapter. The discussion of emplacement environment in this chapter is limited to considerations of the environment that influence, or which may influence, if perturbed, the waste packages and their performance (particularly hydrogeology, geochemistry, and borehole stability). The basis for conceptual waste package design as well as a description of the design is included in this chapter. The complete design will be reported in the advanced conceptual design (ACD) report and is not duplicated in the SCP. 367 refs., 173 figs., 68 tabs

Technical basis for classification of low-activity waste fraction from Hanford site tanks

Energy Technology Data Exchange (ETDEWEB)

Petersen, C.A., Westinghouse Hanford

1996-07-17

The overall objective of this report is to provide a technical basis to support a U.S. Nuclear Regulatory Commission determination to classify the low-activity waste from the Hanford Site single-shell and double-shell tanks as `incidental` wastes after removal of additional radionuclides and immobilization.The proposed processing method, in addition to the previous radionuclide removal efforts, will remove the largest practical amount of total site radioactivity, attributable to high-level wastes, for disposal in a deep geologic repository. The remainder of the waste would be considered `incidental` waste and could be disposed onsite.

Technical basis for classification of low-activity waste fraction from Hanford site tanks

Energy Technology Data Exchange (ETDEWEB)

Petersen, C.A.

1996-09-20

The overall objective of this report is to provide a technical basis to support a U.S. Nuclear Regulatory Commission determination to classify the low-activity waste from the Hanford Site single-shell and double-shell tanks as `incidental` wastes after removal of additional radionuclides and immobilization.The proposed processing method, in addition to the previous radionuclide removal efforts, will remove the largest practical amount of total site radioactivity, attributable to high-level waste, for disposal is a deep geologic repository. The remainder of the waste would be considered `incidental` waste and could be disposed onsite.

Canister storage building design basis accident analysis documentation

Energy Technology Data Exchange (ETDEWEB)

KOPELIC, S.D.

1999-02-25

This document provides the detailed accident analysis to support HNF-3553, Spent Nuclear Fuel Project Final Safety Analysis Report, Annex A, ''Canister Storage Building Final Safety Analysis Report.'' All assumptions, parameters, and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the Canister Storage Building Final Safety Analysis Report.

Determination of a Basis for Design of a Yam (Dioscorea Spp ...

African Journals Online (AJOL)

Manual separation is both tedious and expensive, so the work reported here was done to determine a suitable basis for the design of a mechanical minisett sorter. Results from this study showed that the minisetts cut from the regions of the parent tuber can be separated on the basis of characteristic dimensions of arc length ...

A User-centered Model for Web Site Design

Science.gov (United States)

Kinzie, Mable B.; Cohn, Wendy F.; Julian, Marti F.; Knaus, William A.

2002-01-01

As the Internet continues to grow as a delivery medium for health information, the design of effective Web sites becomes increasingly important. In this paper, the authors provide an overview of one effective model for Web site design, a user-centered process that includes techniques for needs assessment, goal/task analysis, user interface design, and rapid prototyping. They detail how this approach was employed to design a family health history Web site, Health Heritage . This Web site helps patients record and maintain their family health histories in a secure, confidential manner. It also supports primary care physicians through analysis of health histories, identification of potential risks, and provision of health care recommendations. Visual examples of the design process are provided to show how the use of this model resulted in an easy-to-use Web site that is likely to meet user needs. The model is effective across diverse content arenas and is appropriate for applications in varied media. PMID:12087113

An Improved Setpoint Determination Methodology for the Plant Protection System Considering Beyond Design Basis Events

International Nuclear Information System (INIS)

Lee, C.J.; Baik, K.I.; Baek, S.M.; Park, K.-M.; Lee, S.J.

2013-06-01

According to the nuclear regulations and industry standards, the trip setpoint and allowable value for the plant protection system have been determined by considering design basis events. In order to improve the safety of a nuclear power plant, an attempt has been made to develop an improved setpoint determination methodology for the plant protection system trip parameter considering not only a design basis event but also a beyond design basis event. The results of a quantitative evaluation performed for the Advanced Power Reactor 1400 nuclear power plant in Korea are presented herein. The results confirmed that the proposed methodology is able to improve the nuclear power plant's safety by determining more reasonable setpoints that can cover beyond design basis events. (authors)

Assessment Of Source Term And Radiological Consequences For Design Basis Accident And Beyond Design Basis Accident Of The Dalat Nuclear Research Reactor

International Nuclear Information System (INIS)

Luong Ba Vien; Le Vinh Vinh; Huynh Ton Nghiem; Nguyen Kien Cuong; Tran Tri Vien

2011-01-01

The paper presents results of the assessment of source terms and radiological consequences for the Design Basis Accident (DBA) and Beyond Design Basis Accident (BDBA) of the Dalat Nuclear Research Reactor. The dropping of one fuel assembly during fuel handling operation leading to the failure of fuel cladding and the release of fission products into the environment was selected as a DBA for the analysis. For the BDBA, the introduction of a step positive reactivity due to the falling of a heavy block from the rotating bridge crane in the reactor hall onto a part of the platform where are disposed the control rod drives is postulated. The result of the radiological consequence analyses shows that doses to members of the public are below annual dose limit for both DBA and BDBA events. However, doses from exposure to operating staff and experimenters working inside the reactor hall are predicted to be very high in case of BDBA and therefore the protective actions should be taken when the accident occurs. (author)

77 FR 64564 - Implementation of Regulatory Guide 1.221 on Design-Basis Hurricane and Hurricane Missiles

Science.gov (United States)

2012-10-22

...-Basis Hurricane and Hurricane Missiles AGENCY: Nuclear Regulatory Commission. ACTION: Proposed interim...-ISG-024, ``Implementation of Regulatory Guide 1.221 on Design-Basis Hurricane and Hurricane Missiles....221, ``Design-Basis Hurricane and Hurricane Missiles for Nuclear Power Plants.'' DATES: Submit...

Survey report of site selection process on the basis of local partnerships. Slovenia and Belgium

International Nuclear Information System (INIS)

Yoshioka, Tatsuji; Nakata, Hisakazu; Kanzaki, Noriko; Osawa, Hideaki; Hayakawa, Tsuyoshi

2014-01-01

Local partnership approach to achieve stakeholder consensus on low-level waste disposal, with particular focus on the siting of repositories are adopted in Europe. Especially in Belgium and Slovenia, final repository site had been decided in each country. The authors investigate on these two cases on the basis of interviews with implementing organizations for final disposal, representatives of local partnerships and citizens concerned so that the problem of introducing a local partnership in Japan can be examined. Its results are used as a reference for studying measures to determine the radioactive final repository site in Japan. (author)

SLAC site design aesthetics

International Nuclear Information System (INIS)

Hall, F.F.

1985-10-01

Stanford Linear Accelerator Center (SLAC) is a single mission laboratory dedicated to basic research in high energy particle physics. SLAC site also houses Stanford Synchrotron Radiation Laboratory (SSRL) which is a multi-mission laboratory for research using beams of ultraviolet light and low energy photons as emitted tangentially from SLAC colliding beam facilities. This paper discusses various aspects of SLAC site design aesthetics under the following headings: (1) imposed footprint of SLAC, (2) description of selected site, (3) use of earth cover for radiation and sight screens, (4) use of landscaping for cosmetic purposes, (5) use of exterior paint colors to soften SLAC impact on neighbors, (6) relocation of SLAC main entrance, (7) relocation of SLAC collider arcs and experimental hall, (8) parking lots and storage yards, and (9) land use zoning at SLAC

WSDM : A user-centred design method for web sites

NARCIS (Netherlands)

de Troyer, O.M.F.; Leune, C.J.

1998-01-01

WSDM is a user-centered method for the design of kiosk Web Sites. By explicitly starting from the requirements of the users or visitors, WSDM solves Web site problems that are primarily caused by that fact that a site has no underlying design at all, or that the design is mostly data-driven.

Geohydrologic aspects for siting and design of low-level radioactive-waste disposal

Science.gov (United States)

Bedinger, M.S.

1989-01-01

The objective for siting and design of low-level radioactive-waste repository sites is to isolate the waste from the biosphere until the waste no longer poses an unacceptable hazard as a result of radioactive decay. Low-level radioactive waste commonly is isolated at shallow depths with various engineered features to stabilize the waste and to reduce its dissolution and transport by ground water. The unsaturated zone generally is preferred for isolating the waste. Low-level radioactive waste may need to be isolated for 300 to 500 years. Maintenance and monitoring of the repository site are required by Federal regulations for only the first 100 years. Therefore, geohydrology of the repository site needs to provide natural isolation of the waste for the hazardous period following maintenance of the site. Engineering design of the repository needs to be compatible with the natural geohydrologic conditions at the site. Studies at existing commercial and Federal waste-disposal sites provide information on the problems encountered and the basis for establishing siting guidelines for improved isolation of radioactive waste, engineering design of repository structures, and surveillance needs to assess the effectiveness of the repositories and to provide early warning of problems that may require remedial action.Climate directly affects the hydrology of a site and probably is the most important single factor that affects the suitability of a site for shallow-land burial of low-level radioactive waste. Humid and subhumid regions are not well suited for shallow isolation of low-level radioactive waste in the unsaturated zone; arid regions with zero to small infiltration from precipitation, great depths to the water table, and long flow paths to natural discharge areas are naturally well suited to isolation of the waste. The unsaturated zone is preferred for isolation of low-level radioactive waste. The guiding rationale is to minimize contact of water with the waste and to

Structural Basis for Linezolid Binding Site Rearrangement in the Staphylococcus aureus Ribosome.

Science.gov (United States)

Belousoff, Matthew J; Eyal, Zohar; Radjainia, Mazdak; Ahmed, Tofayel; Bamert, Rebecca S; Matzov, Donna; Bashan, Anat; Zimmerman, Ella; Mishra, Satabdi; Cameron, David; Elmlund, Hans; Peleg, Anton Y; Bhushan, Shashi; Lithgow, Trevor; Yonath, Ada

2017-05-09

An unorthodox, surprising mechanism of resistance to the antibiotic linezolid was revealed by cryo-electron microscopy (cryo-EM) in the 70S ribosomes from a clinical isolate of Staphylococcus aureus This high-resolution structural information demonstrated that a single amino acid deletion in ribosomal protein uL3 confers linezolid resistance despite being located 24 Å away from the linezolid binding pocket in the peptidyl-transferase center. The mutation induces a cascade of allosteric structural rearrangements of the rRNA that ultimately results in the alteration of the antibiotic binding site. IMPORTANCE The growing burden on human health caused by various antibiotic resistance mutations now includes prevalent Staphylococcus aureus resistance to last-line antimicrobial drugs such as linezolid and daptomycin. Structure-informed drug modification represents a frontier with respect to designing advanced clinical therapies, but success in this strategy requires rapid, facile means to shed light on the structural basis for drug resistance (D. Brown, Nat Rev Drug Discov 14:821-832, 2015, https://doi.org/10.1038/nrd4675). Here, detailed structural information demonstrates that a common mechanism is at play in linezolid resistance and provides a step toward the redesign of oxazolidinone antibiotics, a strategy that could thwart known mechanisms of linezolid resistance. Copyright © 2017 Belousoff et al.

Cold Vacuum Drying Facility Design Basis Accident Analysis Documentation

International Nuclear Information System (INIS)

PIEPHO, M.G.

1999-01-01

This document provides the detailed accident analysis to support HNF-3553, Annex B, Spent Nuclear Fuel Project Final Safety Analysis Report, ''Cold Vacuum Drying Facility Final Safety Analysis Report (FSAR).'' All assumptions, parameters and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the FSAR

Cold Vacuum Drying facility design basis accident analysis documentation

International Nuclear Information System (INIS)

CROWE, R.D.

2000-01-01

This document provides the detailed accident analysis to support HNF-3553, Annex B, Spent Nuclear Fuel Project Final Safety Analysis Report (FSAR), ''Cold Vacuum Drying Facility Final Safety Analysis Report.'' All assumptions, parameters, and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the FSAR. The calculations in this document address the design basis accidents (DBAs) selected for analysis in HNF-3553, ''Spent Nuclear Fuel Project Final Safety Analysis Report'', Annex B, ''Cold Vacuum Drying Facility Final Safety Analysis Report.'' The objective is to determine the quantity of radioactive particulate available for release at any point during processing at the Cold Vacuum Drying Facility (CVDF) and to use that quantity to determine the amount of radioactive material released during the DBAs. The radioactive material released is used to determine dose consequences to receptors at four locations, and the dose consequences are compared with the appropriate evaluation guidelines and release limits to ascertain the need for preventive and mitigative controls

Cold Vacuum Drying facility design basis accident analysis documentation

Energy Technology Data Exchange (ETDEWEB)

CROWE, R.D.

2000-08-08

This document provides the detailed accident analysis to support HNF-3553, Annex B, Spent Nuclear Fuel Project Final Safety Analysis Report (FSAR), ''Cold Vacuum Drying Facility Final Safety Analysis Report.'' All assumptions, parameters, and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the FSAR. The calculations in this document address the design basis accidents (DBAs) selected for analysis in HNF-3553, ''Spent Nuclear Fuel Project Final Safety Analysis Report'', Annex B, ''Cold Vacuum Drying Facility Final Safety Analysis Report.'' The objective is to determine the quantity of radioactive particulate available for release at any point during processing at the Cold Vacuum Drying Facility (CVDF) and to use that quantity to determine the amount of radioactive material released during the DBAs. The radioactive material released is used to determine dose consequences to receptors at four locations, and the dose consequences are compared with the appropriate evaluation guidelines and release limits to ascertain the need for preventive and mitigative controls.

Beyond Trust: Web Site Design Preferences Across Cultures

OpenAIRE

Dianne Cyr; Carole Bonanni; John Bowes; Joe Ilsever

2005-01-01

The growth of Internet shopping motivates a better understanding of how e-loyalty is built online between businesses and consumers. In this study, Web site design and culture are advanced as important to Web site trust, Web site satisfaction, and e-loyalty in online business relationships. Based on data collected in Canada, the U.S., Germany, and Japan, the research considers (1) examining within culture preferences for design elements of a local vs. a foreign Web site and subsequent particip...

Beyond-design-basis accident management in the RF regulation documents

International Nuclear Information System (INIS)

Bukrinskij, A.M.

2010-01-01

The article observes the issues of the management of beyond-design-basis accidents (BDBA) in the existing regulations in Russia. The ideology of the approach to the definition of the BDBA list to formulate the management guidelines has been proposed [ru

Canister Storage Building (CSB) Design Basis Accident Analysis Documentation

International Nuclear Information System (INIS)

CROWE, R.D.; PIEPHO, M.G.

2000-01-01

This document provided the detailed accident analysis to support HNF-3553, Spent Nuclear Fuel Project Final Safety Analysis Report, Annex A, ''Canister Storage Building Final Safety Analysis Report''. All assumptions, parameters, and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the Canister Storage Building Final Safety Analysis Report

Canister Storage Building (CSB) Design Basis Accident Analysis Documentation

International Nuclear Information System (INIS)

CROWE, R.D.

1999-01-01

This document provides the detailed accident analysis to support ''HNF-3553, Spent Nuclear Fuel Project Final Safety, Analysis Report, Annex A,'' ''Canister Storage Building Final Safety Analysis Report.'' All assumptions, parameters, and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the Canister Storage Building Final Safety Analysis Report

[Basis for designing a medical course curriculum].

Science.gov (United States)

Villarreal, R; Bojalil, L F; Mercer, H

1977-01-01

This article sets forth the reasons for the structure given to the Division of Biology and Health on the Xochimilco campus of Metropolitan Autonomous University in Mexico: to adjust the university to the process of social change going forward in the country and gear the university to the problems of the present by avoiding the rigidity of its structure. The basic aspects of curriculum design are cited against a background of an historical analysis of the socioeconomic structure of education and health. The principles underlying the curriculum and the course work are then described on the basis of that analysis.

Response spectra by blind faults for design purpose of stiff structures on rock site

International Nuclear Information System (INIS)

Hiroyuki Mizutani; Kenichi Kato; Masayuki Takemura; Kazuhiko Yashiro; Kazuo Dan

2005-01-01

The goal of this paper is to propose the response spectra by blind faults for seismic design of nuclear power facilities. It is impossible to evaluate earthquake ground motions from blind faults, because the size and the location of blind fault cannot be identified even if the detailed geological surveys are conducted. From the viewpoint of seismic design, it is crucial to investigate the upper level of earthquake ground motions due to blind faults. In this paper, 41 earthquakes that occurred in the upper crust in Japan and California are selected and classified into the active and the blind fault types. On the basis of near-source strong motion records observed on rock sites, upper level of response spectra by blind faults is examined. The estimated upper level is as follows: the peak ground acceleration is 450 cm/s 2 , the flat level of the acceleration response spectra is 1200 cm/s 2 , and the flat level of the velocity response spectra is 100 cm/s on rock sites with shear wave velocity Vs of about 700 m/s. The upper level can envelop the observed response spectra in near-source region on rock sites. (authors)

Reduced design load basis for ultimate blade loads estimation in multidisciplinary design optimization frameworks

DEFF Research Database (Denmark)

Pavese, Christian; Tibaldi, Carlo; Larsen, Torben J.

2016-01-01

The aim is to provide a fast and reliable approach to estimate ultimate blade loads for a multidisciplinary design optimization (MDO) framework. For blade design purposes, the standards require a large amount of computationally expensive simulations, which cannot be efficiently run each cost...... function evaluation of an MDO process. This work describes a method that allows integrating the calculation of the blade load envelopes inside an MDO loop. Ultimate blade load envelopes are calculated for a baseline design and a design obtained after an iteration of an MDO. These envelopes are computed...... for a full standard design load basis (DLB) and a deterministic reduced DLB. Ultimate loads extracted from the two DLBs with the two blade designs each are compared and analyzed. Although the reduced DLB supplies ultimate loads of different magnitude, the shape of the estimated envelopes are similar...

Design basis for resistance to shock and vibration

International Nuclear Information System (INIS)

Glass, R.E.; Gwinn, K.W.

1989-01-01

Sandia National Laboratories, in conjunction with its participation in the American National Standards Institute (ANSI) writing groups, has undertaken to provide an experimental and analytical basis for the design of components of radioactive materials packages to resist normal transport shock and vibration loads. Previous efforts have resulted in an overly conservative shock spectra description of the loads in the tie-downs and cask attachment points anticipated during normal shipment. The present effort is aimed at predicting the actual loads so that the design basis can be accurately determined. This goal is being accomplished with road simulator and over-the-road tests and the development of an analytical model. This model is used to parametrically evaluate and envelop the transportation systems' responses. The parameters to be varied include damping, stiffness, geometry, and cargo mass. The over-the-road tests provide operational data that are used to validate the selection of environments for the road simulator tests. The road simulator tests provide verification for the model. This verification is accomplished since the road simulator tests provide not only the system response which can be measured in over-the-road tests but also the system input. Finally, when the model has been verified, it can be used to vary parameters to envelop a wide range of normal transport conditions

Design basis for resistance to shock and vibration

International Nuclear Information System (INIS)

Glass, R.E.; Gwinn, K.W.

1989-01-01

Sandia National Laboratories, in conjunction with its participation in the American National Standards Institute (ANSI) writing groups, has undertaken to provide an experimental and analytical basis for the design of components of radioactive materials packages to resist normal transport shock and vibration loads. Previous efforts have resulted in an overly conservative shock spectra description of the loads in the tie-downs and cask attachment points anticipated during normal shipment. The present effort is aimed at predicting the actual loads so that the design basis can be accurately determined. This goal is being accomplished with road simulator and over-the-road tests and the development of an analytical model. This model is used to parametrically evaluate and envelop the transportation systems responses. The parameters to be varied include damping, stiffness, geometry, and cargo mass. The over-the-road tests provide operational data that are used to validate the selection of environments for the road simulator tests. The road simulator tests provide verification for the model. This verification is accomplished since the road simulator tests provide not only the system response which can be measured in over-the-road tests but also the system input. Finally, when the model has been verified, it can be used to vary parameters to envelope a wide range of normal transport conditions

Report of the working group 'Regulatory requirements on AM - Concept of nuclear and radiation safety during beyond-design-basis accidents'

International Nuclear Information System (INIS)

Bobaly, P.

2001-01-01

The developed working group report contains the following main paragraphs: legal basis and basis for regulatory requirements for on-site and off-site Accident Management (AM), regulatory requirements or recommendations for on-site AM and for emergency preparedness, background information concerning the implementation and review of an AM program as a basis for an AM guideline. Overview about AM/SAM implementation in member countries of the SAMINE project; measure and candidates for high level actions based upon US SAMG; interactions of severe accident research and the regulatory positions, relationship between different components of an accident management programme are also given

Canister Storage Building (CSB) Design Basis Accident Analysis Documentation

Energy Technology Data Exchange (ETDEWEB)

CROWE, R.D.

1999-09-09

This document provides the detailed accident analysis to support ''HNF-3553, Spent Nuclear Fuel Project Final Safety, Analysis Report, Annex A,'' ''Canister Storage Building Final Safety Analysis Report.'' All assumptions, parameters, and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the Canister Storage Building Final Safety Analysis Report.

Spent Nuclear Fuel (SNF) Project Design Basis Capacity Study

International Nuclear Information System (INIS)

CLEVELAND, K.J.

2000-01-01

This study of the design basis capacity of process systems was prepared by Fluor Federal Services for the Spent Nuclear Fuel Project. The evaluation uses a summary level model of major process sub-systems to determine the impact of sub-system interactions on the overall time to complete fuel removal operations. The process system model configuration and time cycle estimates developed in the original version of this report have been updated as operating scenario assumptions evolve. The initial document released in Fiscal Year (FY) 1996 varied the number of parallel systems and transport systems over a wide range, estimating a conservative design basis for completing fuel processing in a two year time period. Configurations modeling planned operations were updated in FY 1998 and FY 1999. The FY 1998 Base Case continued to indicate that fuel removal activities at the basins could be completed in slightly over 2 years. Evaluations completed in FY 1999 were based on schedule modifications that delayed the start of KE Basin fuel removal, with respect to the start of KW Basin fuel removal activities, by 12 months. This delay resulted in extending the time to complete all fuel removal activities by 12 months. However, the results indicated that the number of Cold Vacuum Drying (CVD) stations could be reduced from four to three without impacting the projected time to complete fuel removal activities. This update of the design basis capacity evaluation, performed for FY 2000, evaluates a fuel removal scenario that delays the start of KE Basin activities such that staffing peaks are minimized. The number of CVD stations included in all cases for the FY 2000 evaluation is reduced from three to two, since the scenario schedule results in minimal time periods of simultaneous fuel removal from both basins. The FY 2000 evaluation also considers removal of Shippingport fuel from T Plant storage and transfer to the Canister Storage Building for storage

Experimental site and design

Energy Technology Data Exchange (ETDEWEB)

Guenette, C. C. [SINTEF Applied Cemistry, Trondheim (Norway)

1999-08-01

Design and site selection criteria for the Svalbard oil spill experiments are described. All three experimental sites have coarse and mixed sediment beaches of sand and pebble; within each site wave exposure is very similar; along-shore and across-shore sediment characteristics are also relatively homogeneous. Tidal range is in the order of 0.6 m at neaps, and 1.8 m at springs. All three sites are open to wave action and are ice-free during the experimental period of mid-July to mid-October. Study plots at each site were selected for different treatments from within the continuous stretch of oiled shoreline, with oiled buffer zones between plots and at either end of the oiled zone. Treatments included mixing (tilling), sediment relocation (surf washing) and bioremediation (nutrient enrichment). Measurements and observations were carried out during the summers of 1997 and 1998. The characteristics measured were: wave and wind conditions; beach topography and elevation; sediment grain size distribution; mineral fines size distribution and mineral composition; background hydrocarbons; concentration of oil within experimental plots and the rate of oil loss over time; depth of oil penetration and thickness of the oiled sediment layer; oil concentration and toxicity of near-shore benthic sediments; mineral composition of suspended particulate material captured in sub-tidal sediment traps; and oil-fines interaction in near-shore water samples. 1 fig.

Experimental site and design

Energy Technology Data Exchange (ETDEWEB)

Guenette, C. C. [SINTEF Applied Cemistry, Trondheim (Norway)

1999-07-01

Design and site selection criteria for the Svalbard oil spill experiments are described. All three experimental sites have coarse and mixed sediment beaches of sand and pebble; within each site waveexposure is very similar; along-shore and across-shore sediment characteristics are also relatively homogeneous. Tidal range is in the order of 0.6 m at neaps, and 1.8 m at springs. All three sites are open to wave action and are ice-free during the experimental period of mid-July to mid-October. Study plots at each site were selected for different treatments from within the continuous stretch of oiled shoreline, with oiled buffer zones between plots and at either end of the oiled zone. Treatments included mixing (tilling), sediment relocation (surf washing) and bioremediation (nutrient enrichment). Measurements and observations were carried out during the summers of 1997 and 1998. The characteristics measured were: wave and wind conditions; beach topography and elevation; sediment grain size distribution; mineral fines size distribution and mineral composition; background hydrocarbons; concentration of oil within experimental plots and the rate of oil loss over time; depth of oil penetration and thickness of the oiled sediment layer; oil concentration and toxicity of near-shore benthic sediments; mineral composition of suspended particulate material captured in sub-tidal sediment traps; and oil-fines interaction in near-shore water samples. 1 fig.

Design Basis Provisions for New and Existing Nuclear Power Plants and Nuclear Fuel Cycle Facilities in India

International Nuclear Information System (INIS)

Soni, R.S.

2013-01-01

India has 3-Stage Nuclear Power Program. • Various facilities under design, construction or operation. • Design Basis Knowledge Management (DBKM) is an important and challenging task. • Design Basis Knowledge contributes towards: - Safe operation of running plants; - Design and construction of new facilities; - Addresses issues related to future decommissioning activities

NPP Design Basis Handover and Knowledge Preservation from Subcontractors, Vendors and EPC

International Nuclear Information System (INIS)

Freeland, Kent

2013-01-01

Using PLM-based Workflow for Configuration Management (CM) in the Nuclear Power Industry Advantages – some work to do! • NPP’s must adapt to using PLM-based solutions to support CM and to synchronize design changes to asset or product changes, and reduce “slipstreaming”. In the NPP world, this often appears as events that circumvent CM – for example, non-approved parts substitutions and “temporary” plant modifications that are never removed. • PLM serves as the method for unifying the application of requirements to design changes, processes and workflow. In NPP’s, requirements are generally considered only relevant to designs – not process and workflow. • PLM supports Configuration Management and Design Basis in Regulator Action Tracking for NPP’s, and application of PLM-based CM to regulator action and compliance systems. This is a poorly-understood application of CM in NPP’s, yet these elements control large parts of the NPP design basis. • Suppliers, EPC’s and Technology Vendors must also understand the role of CM, SE and PLM in construction of new standards-driven NPP designs (like EPR and Westinghouse AP-1000 NPP designs), as well as understanding the role and handling of Knowledge Systems

Acceptable risk as a basis for design

International Nuclear Information System (INIS)

Vrijling, J.K.; Hengel, W. van; Houben, R.J.

1998-01-01

Historically, human civilisations have striven to protect themselves against natural and man-made hazards. The degree of protection is a matter of political choice. Today this choice should be expressed in terms of risk and acceptable probability of failure to form the basis of the probabilistic design of the protection. It is additionally argued that the choice for a certain technology and the connected risk is made in a cost-benefit framework. The benefits and the costs including risk are weighed in the decision process. A set of rules for the evaluation of risk is proposed and tested in cases. The set of rules leads to technical advice in a question that has to be decided politically

Design Basis Threat (DBT) Approach for the First NPP Security System in Indonesia

International Nuclear Information System (INIS)

Ign Djoko Irianto

2004-01-01

Design Basis Threat (DBT) is one of the main factors to be taken into account in the design of physical protection system of nuclear facility. In accordance with IAEA's recommendations outlined in INFCIRC/225/Rev.4 (Corrected), DBT is defined as: attributes and characteristics of potential insider and/or external adversaries, who might attempt unauthorized removal of nuclear material or sabotage against the nuclear facilities. There are three types of adversary that must be considered in DBT, such as adversary who comes from the outside (external adversary), adversary who comes from the inside (internal adversary), and adversary who comes from outside and colludes with insiders. Current situation in Indonesia, where many bomb attacks occurred, requires serious attention on DBT in the physical protection design of NPP which is to be built in Indonesia. This paper is intended to describe the methodology on how to create and implement a Design Basis Threat in the design process of NPP physical protection in Indonesia. (author)

Design basis for the safe disposal of radioactive waste

International Nuclear Information System (INIS)

Lewi, J.; Kaluzny, Y.

1990-01-01

All radioactive waste disposal sites, regardless of disposal concept, are designed to isolate the radioactive substances contained in such waste for a period at least equal to the time it may remain potentially harmful. Isolation is achieved through the use of containment barriers. This paper summarises the function and limits of different types of barrier used in various disposal systems. For each type of barrier, the paper describes and comments on the site selection criteria and waste packaging requirements applicable in various countries. 13 refs., 1 fig [fr

Adaptation of the ITER facility design to a Canadian site

International Nuclear Information System (INIS)

Smith, S.

2001-01-01

This paper presents the status of Canadian efforts to adapt the newly revised ITER facility design to suit the specific characteristics of the proposed Canadian site located in Clarington, west of Toronto, Ontario. ITER Canada formed a site-specific design team in 1999, comprising participants from three Canadian consulting companies to undertake this work. The technical aspects of this design activity includes: construction planning, geotechnical investigations, plant layout, heat sink design, electrical system interface, site-specific modifications and tie-ins, seismic design, and radwaste management. These areas are each addressed in this paper. (author)

SRS BEDROCK PROBABILISTIC SEISMIC HAZARD ANALYSIS (PSHA) DESIGN BASIS JUSTIFICATION (U)

Energy Technology Data Exchange (ETDEWEB)

(NOEMAIL), R

2005-12-14

This represents an assessment of the available Savannah River Site (SRS) hard-rock probabilistic seismic hazard assessments (PSHAs), including PSHAs recently completed, for incorporation in the SRS seismic hazard update. The prior assessment of the SRS seismic design basis (WSRC, 1997) incorporated the results from two PSHAs that were published in 1988 and 1993. Because of the vintage of these studies, an assessment is necessary to establish the value of these PSHAs considering more recently collected data affecting seismic hazards and the availability of more recent PSHAs. This task is consistent with the Department of Energy (DOE) order, DOE O 420.1B and DOE guidance document DOE G 420.1-2. Following DOE guidance, the National Map Hazard was reviewed and incorporated in this assessment. In addition to the National Map hazard, alternative ground motion attenuation models (GMAMs) are used with the National Map source model to produce alternate hazard assessments for the SRS. These hazard assessments are the basis for the updated hard-rock hazard recommendation made in this report. The development and comparison of hazard based on the National Map models and PSHAs completed using alternate GMAMs provides increased confidence in this hazard recommendation. The alternate GMAMs are the EPRI (2004), USGS (2002) and a regional specific model (Silva et al., 2004). Weights of 0.6, 0.3 and 0.1 are recommended for EPRI (2004), USGS (2002) and Silva et al. (2004) respectively. This weighting gives cluster weights of .39, .29, .15, .17 for the 1-corner, 2-corner, hybrid, and Greens-function models, respectively. This assessment is judged to be conservative as compared to WSRC (1997) and incorporates the range of prevailing expert opinion pertinent to the development of seismic hazard at the SRS. The corresponding SRS hard-rock uniform hazard spectra are greater than the design spectra developed in WSRC (1997) that were based on the LLNL (1993) and EPRI (1988) PSHAs. The

A design method for an intuitive web site

Energy Technology Data Exchange (ETDEWEB)

Quinniey, M.L.; Diegert, K.V.; Baca, B.G.; Forsythe, J.C.; Grose, E.

1999-11-03

The paper describes a methodology for designing a web site for human factor engineers that is applicable for designing a web site for a group of people. Many web pages on the World Wide Web are not organized in a format that allows a user to efficiently find information. Often the information and hypertext links on web pages are not organized into intuitive groups. Intuition implies that a person is able to use their knowledge of a paradigm to solve a problem. Intuitive groups are categories that allow web page users to find information by using their intuition or mental models of categories. In order to improve the human factors engineers efficiency for finding information on the World Wide Web, research was performed to develop a web site that serves as a tool for finding information effectively. The paper describes a methodology for designing a web site for a group of people who perform similar task in an organization.

THE IMPORTANCE OF WEB DESIGN: VISUAL DESIGN EVALUATION OF DESTINATION WEB SITES

OpenAIRE

Fırlar, Belma; Okat Özdem, Özen

2013-01-01

As in the literature, the researchs about web site efficiency are mostly about site context. The analysis about function are mostly superficial. Whereas, controlling every little part of a web site respective is a necessity to show its efficiency. Here in this context in the study of perception and response event web sites that play an important role in visual design criteria are below the lens as featured and the web sites evaulated by heuristic evaluation method.The research focus of this s...

Development of methodology for the analysis of fuel behavior in light water reactor in design basis accidents

International Nuclear Information System (INIS)

Salatov, A. A.; Goncharov, A. A.; Eremenko, A. S.; Kuznetsov, V. I.; Bolnov, V. A.; Gusev, A. S.; Dolgov, A. B.; Ugryumov, A. V.

2013-01-01

The report attempts to analyze the current experience of the safety fuel for light-water reactors (LWRs) under design-basis accident conditions in terms of its compliance with international requirements for licensing nuclear power plants. The components of fuel behavior analysis methodology in design basis accidents in LWRs were considered, such as classification of design basis accidents, phenomenology of fuel behavior in design basis accidents, system of fuel safety criteria and their experimental support, applicability of used computer codes and input data for computational analysis of the fuel behavior in accidents, way of accounting for the uncertainty of calculation models and the input data. A brief history of the development of probabilistic safety analysis methodology for nuclear power plants abroad is considered. The examples of a conservative approach to safety analysis of VVER fuel and probabilistic approach to safety analysis of fuel TVS-K are performed. Actual problems in development of the methodology of analyzing the behavior of VVER fuel at the design basis accident conditions consist, according to the authors opinion, in following: 1) Development of a common methodology for analyzing the behavior of VVER fuel in the design basis accidents, implementing a realistic approach to the analysis of uncertainty - in the future it is necessary for the licensing of operating VVER fuel abroad; 2) Experimental and analytical support to the methodology: experimental studies to identify and study the characteristics of the key uncertainties of computational models of fuel and the cladding, development of computational models of key events in codes, validation code on the basis of integral experiments

Cold Vacuum Drying (CVD) Facility Design Basis Accident Analysis Documentation

Energy Technology Data Exchange (ETDEWEB)

PIEPHO, M.G.

1999-10-20

This document provides the detailed accident analysis to support HNF-3553, Annex B, Spent Nuclear Fuel Project Final Safety Analysis Report, ''Cold Vacuum Drying Facility Final Safety Analysis Report (FSAR).'' All assumptions, parameters and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the FSAR.

Impact of Site Elevation on Mg Smelter Design

Science.gov (United States)

Baker, Phillip W.

Site elevation has many surprising and significant impacts on the engineering design of metallurgical plant of all types. Electrolytic magnesium smelters maybe built at high elevation for a variety of reasons including availability of raw material, energy or electric power. Because of the unit processes they typically involve, Mg smelters can be extensively impacted by site elevation. In this paper, generic examples of the design changes required to adapt a smelter originally designed for sea level to operate at 2700 m are presented. While the examples are drawn from a magnesium plant design case, these changes are generically applicable to all industrial plants utilizing similar unit processes irrespective of product.

Potentials for site-specific design of MW sized wind turbines

DEFF Research Database (Denmark)

Thomsen, K.; Fuglsang, P.; Schepers, G.

2001-01-01

The potential for site specific design of MW sized wind turbines is quantified by comparing design loads for wind turbines installed at a range of different sites. The sites comprise on-shore normal flat terrain stand-alone conditions and wind farm conditions together with offshore and mountainous...

Site characterization design and techniques used at the Southern Shipbuilding Corporation site

International Nuclear Information System (INIS)

Mueller, J.P.; Geraghty, C.A.; Moore, G.W.; Mullins, J.R.

1995-01-01

The Southern Shipbuilding Corporation (SSC) site is an inactive barge/ship manufacturing and repair facility situated on approximately 54 acres in Slidell, St. Tammany Parish, Louisiana. Two unlined surface impoundments (North and South impoundments) are situated on the northwest portion of the site and are surrounded on three sides by Bayou Bonfouca. These impoundments are the sources of carcinogenic polynuclear aromatic hydrocarbon (CPAH) contamination at the site. Inadequate containment has resulted in the release of impoundment wastes into the bayou. To evaluate potential response alternatives for the site, an Engineering Evaluation/Cost Analysis (EE/CA) field investigation was conducted from July through October 1994. A two phase sampling approach was used in combination with innovative and traditional sampling techniques, field screening technologies, and exploitation of the visual characteristics of the waste to determine the extent of waste migration with limited off-site laboratory confirmation. A skid-mounted mobile drilling unit, secured to a specialized sampling platform designed for multiple applications, was used for collection of sediment cores from the bayou as well as tarry sludge cores from the impoundments. Field screening of core samples was accomplished on site using an organic vapor analyzer and a total petroleum hydrocarbon (TPH) field analyzer. Pollutants of concern include metals, cyanide, dioxin, and organic compounds. This paper presents details on the sampling design and characterization techniques used to accomplish the EE/CA field investigation

Comparison of deterministic and stochastic techniques for estimation of design basis floods for nuclear power plants

International Nuclear Information System (INIS)

Solomon, S.I.; Harvey, K.D.

1982-12-01

The IAEA Safety Guide 50-SG-S10A recommends that design basis floods be estimated by deterministic techniques using probable maximum precipitation and a rainfall runoff model to evaluate the corresponding flood. The Guide indicates that stochastic techniques are also acceptable in which case floods of very low probability have to be estimated. The paper compares the results of applying the two techniques in two river basins at a number of locations and concludes that the uncertainty of the results of both techniques is of the same order of magnitude. However, the use of the unit hydrograph as the rainfall runoff model may lead in some cases to nonconservative estimates. A distributed non-linear rainfall runoff model leads to estimates of probable maximum flood flows which are very close to values of flows having a 10 6 - 10 7 years return interval estimated using a conservative and relatively simple stochastic technique. Recommendations on the practical application of Safety Guide 50-SG-10A are made and the extension of the stochastic technique to ungauged sites and other design parameters is discussed

Analysis of regulatory requirement for beyond design basis events of SMART

International Nuclear Information System (INIS)

Kim, W. S.; Seol, K. W.

2000-01-01

To enhance the safety of SMART reactor, safety and regulatory requirements associated with beyond design basis events (beyond BDE), which were developed and applied to advanced light water reactor designs, were analyzed along with a design status of passive reactor. And, based on these requirements, their applicability on the SMART design was evaluated. In the design aspect, severe accident prevention and mitigation features, containment performance, and accident management were analyzed. The evaluation results show that the requirement related to beyond DBE such as ATWS, loss of residual heat removal during shutdown operation, station blackout, fire, inter-system LOCA, and well-known events from severe accident phenomena is applicable to the SMART design. However, comprehensive approach against beyond DBE is not yet provided in the SMART design, and then it is required to designate and analyze the beyond DBE-related features. This study is expected to contribute to efforts to improve plant safety and to establish regulatory requirements for safety review

Design basis event consequence analyses for the Yucca Mountain project

International Nuclear Information System (INIS)

Orvis, D.D.; Haas, M.N.; Martin, J.H.

1997-01-01

Design basis event (DBE) definition and analysis is an ongoing and integrated activity among the design and analysis groups of the Yucca Mountain Project (YMP). DBE's are those that potentially lead to breach of the waste package and waste form (e.g., spent fuel rods) with consequent release of radionuclides to the environment. A Preliminary Hazards Analysis (PHA) provided a systematic screening of external and internal events that were candidate DBE's that will be subjected to analyses for radiological consequences. As preparation, pilot consequence analyses for the repository subsurface and surface facilities have been performed to define the methodology, data requirements, and applicable regulatory limits

NDARC-NASA Design and Analysis of Rotorcraft Theoretical Basis and Architecture

Science.gov (United States)

Johnson, Wayne

2010-01-01

The theoretical basis and architecture of the conceptual design tool NDARC (NASA Design and Analysis of Rotorcraft) are described. The principal tasks of NDARC are to design (or size) a rotorcraft to satisfy specified design conditions and missions, and then analyze the performance of the aircraft for a set of off-design missions and point operating conditions. The aircraft consists of a set of components, including fuselage, rotors, wings, tails, and propulsion. For each component, attributes such as performance, drag, and weight can be calculated. The aircraft attributes are obtained from the sum of the component attributes. NDARC provides a capability to model general rotorcraft configurations, and estimate the performance and attributes of advanced rotor concepts. The software has been implemented with low-fidelity models, typical of the conceptual design environment. Incorporation of higher-fidelity models will be possible, as the architecture of the code accommodates configuration flexibility, a hierarchy of models, and ultimately multidisciplinary design, analysis and optimization.

Structural analysis of the CAREM-25 nuclear power plant subjected to the design basis accident and seismic loads

International Nuclear Information System (INIS)

Ambrosini, Daniel; Codina, Ramón H.; Curadelli, Oscar; Martínez, Carlos A.

2017-01-01

Highlights: • Structural analysis of CAREM-25 NPP is presented. • Full 3D numerical model was developed. • Transient thermal and static structural analyses were performed. • Modeling guidelines for numerical structural analysis of NPP are recommended. • Envelope condition of DBA dominates the structural behavior. - Abstract: In this paper, a numerical study about the structural response of the Argentine nuclear power plant CAREM-25 subjected to the design basis accident (DBA) and seismic loads is presented. Taking into account the hardware capabilities available, a full 3D finite element model was adopted. A significant part of the building was modeled using more than 2 M solid elements. In order to take into account the foundation flexibility, linear springs were used. The springs and the model were calibrated against a greater model used to study the soil-structure interaction. The structure was subjected to the DBA and seismic loads as combinations defined by ASME international code. First, a transient thermal analysis was performed with the conditions defined by DBA and evaluating the time history of the temperature of the model, each 1 h until 36 h. The final results of this stage were considered as initial conditions of a static structural analysis including the pressure defined by DBA. Finally, an equivalent static analysis was performed to analyze the seismic response considering the design basis spectra for the site. The different loads were combined and the abnormal/extreme environmental combination was the most unfavorable for the structure, defining the design.

Definitive design status of the SP-100 Ground Engineering System Test Site

International Nuclear Information System (INIS)

Renkey, E.J. Jr.; Bazinet, G.D.; Bitten, E.J.; Brackenbury, P.J.; Carlson, W.F.; Irwin, J.J.; Edwards, P.A.; Shen, E.J.; Titzler, P.A.

1989-05-01

The SP-100 reactor will be ground tested at the SP-100 Ground Engineering System (GES) Test Site on the US Department of Energy (DOE) Hanford Site near Richland, Washington. Project direction and the flight system design evolution have resulted in a smaller reactor size and the consequential revision to Test Site features to accommodate the design changes and reduce Test Site costs. The significant design events since the completion of the Conceptual Design are discussed in this paper

Definitive design status of the SP-100 Ground Engineering System Test Site

Energy Technology Data Exchange (ETDEWEB)

Renkey, E.J. Jr.; Bazinet, G.D.; Bitten, E.J.; Brackenbury, P.J.; Carlson, W.F.; Irwin, J.J.; Edwards, P.A.; Shen, E.J.; Titzler, P.A.

1989-05-01

The SP-100 reactor will be ground tested at the SP-100 Ground Engineering System (GES) Test Site on the US Department of Energy (DOE) Hanford Site near Richland, Washington. Project direction and the flight system design evolution have resulted in a smaller reactor size and the consequential revision to Test Site features to accommodate the design changes and reduce Test Site costs. The significant design events since the completion of the Conceptual Design are discussed in this paper.

Study on effective prestressing effects on concrete containment under the design-basis pressure condition

International Nuclear Information System (INIS)

Sun Feng; Pan Rong; Wang Lu; Mao Huan; Yang Yu

2013-01-01

Prestressing technology is widely used in nuclear power plant containment building, and the durability of containment structure is affected directly by the distribution and loss of prestressing value under design-basis pressure. Containment structure and the distribution of prestressing system are introduced briefly. Furthermore, the calculating process of horizontal prestressing bunch loss near the equipment hatch hole is put forward in details, and the containment structure prestressing loss when 5-year pressure test is obtained. Based above analysis, the finite element model of the prestressed concrete containment structure is built by using ANSYS code, the prestressing effect on concrete containment is analysed. The results show that most of the design pressure is bore by the prestressing system under the design-basis pressure, so the containment structure is safe. These conclusions are consistent with prestressing containment system design concepts, which can provide reference to the engineering staff. (authors)

Transient and accident analyses topical design basis documents

International Nuclear Information System (INIS)

Chi, Larry; Eckert, Eugene; Grim, Brit

2004-01-01

The designers and operators of nuclear power plants have extensively documented system functions, licensing performance, and operating procedures for all conditions. This paper presents a complementary, systematic approach for the documentation of all requirements that are based on the analysis of operational transients, abnormal transients, accidents, and other events which are included in the design and licensing basis for the plant. Up to now, application of the approach has focused on required mitigation actions (automatic or manual). All mitigation actions are directly identified with all applicable reactor events, as well as the plant-unique systems that work together to perform each function. The approach is also applicable to all operational functions. The approach makes extensive use of data base methods, thereby providing effective ways to interrogate the information for the varied users of this information. Examples of use include: evaluations of system design changes and equipment modifications, safety evaluations of any plant change (e.g., USNRC 10CFR50.59 review), plant operations (e.g., manual actions during unplanned events), system interactions, classification of safety-related equipment, environmental qualification of equipment, and mitigation requirements for different reactor operating states. This approach has been applied in customized ways to several boiling water reactor (BWR) units, based on the desires and needs of the specific utility. (author)

Advances in the physics basis for the European DEMO design

Science.gov (United States)

Wenninger, R.; Arbeiter, F.; Aubert, J.; Aho-Mantila, L.; Albanese, R.; Ambrosino, R.; Angioni, C.; Artaud, J.-F.; Bernert, M.; Fable, E.; Fasoli, A.; Federici, G.; Garcia, J.; Giruzzi, G.; Jenko, F.; Maget, P.; Mattei, M.; Maviglia, F.; Poli, E.; Ramogida, G.; Reux, C.; Schneider, M.; Sieglin, B.; Villone, F.; Wischmeier, M.; Zohm, H.

2015-06-01

In the European fusion roadmap, ITER is followed by a demonstration fusion power reactor (DEMO), for which a conceptual design is under development. This paper reports the first results of a coherent effort to develop the relevant physics knowledge for that (DEMO Physics Basis), carried out by European experts. The program currently includes investigations in the areas of scenario modeling, transport, MHD, heating & current drive, fast particles, plasma wall interaction and disruptions.

The RCC-MR design code for LMFBR components. A useful basis for fusion reactor design tools development

International Nuclear Information System (INIS)

Acker, D.; Chevereau, G.

1986-01-01

LMFBR and fusion reactors exhibit common features with regard to structural materials, temperature service level, loading types. So, design and construction rules used in France for LMFBR, that is to say RCC-MR Code, can constitute a good basis for fusion reactors design. Some original aspects of RCC-MR design rules are described, relating to unsignificant creep, ratchetting effect, fatigue and creep damage limits, creep damage evaluation, fatigue damage evaluation, buckling. The main originality of RCC-MR consists to propose comprehensive simplified rules based on elastic calculations and extended from classical cold temperatures to the elevated temperature domain. (author)

Design surface covers: an approach to long-term waste site stabilization

International Nuclear Information System (INIS)

Beedlow, P.A.; Cadwell, L.L.; McShane, M.C.

1983-02-01

The wide range of existing environmental conditions, potential contaminants and available cover materials at waste disposal sites necessitates site-specific designing of surface covers for effective long-term erosion resistance. This paper presents a systematic approach to designing surface covers for hazardous waste repositories that can be tailored to conditions at any site. The approach consists of three phases: (1) an assessment, during which the degree of required surface protection (erosion potential) is determined; (2) a preliminary design that integrates surface cover design with the need to minimize transport of contaminants; and (3) a final design, where the cost and effectiveness of the surface cover are determined. 1 figure

Overview of Mobile Equipment Used in Case of Beyond Design Basis Accident at NPP Krsko

International Nuclear Information System (INIS)

Lukacevic, H.; Kopinc, D.; Ivanjko, M.

2016-01-01

Terrorist attack in USA in the September 11, 2001 and accident at the Fukushima - Daiichi Nuclear Power Station in the March 11, 2011 highlight the importance of mitigating strategies in responding to Beyond Design Basis Accident (BDBA), while ensuring cooling of reactor core, containment and spent fuel pool. Nuclear Power Plant Krsko (NEK) has acquired additional mobile equipment and made necessary modifications on existing systems for the connection of this equipment (fast couplers). Usage of mobile equipment is not only limited to design basis accident (DBA), but, also to prevent and mitigate the consequences in case of BDBA, when other plant systems are not available. NEK also decided to take steps for upgrade of safety measures and prepared Safety Upgrade Program (SUP), which is consistent with the nuclear industry response to the Fukushima accident and is implementing main projects and modifications related to SUP. NEK mobile equipment is not required to operate under normal reactor plant operation except for periodic surveillance testing and is incorporated into the normal training process. Equipment is dislocated from the reactor building and most of the equipment is located in the new building, able to withstand extreme natural events, including earthquakes and tornadoes. The usage of all mobile equipment is prescribed as an additional option in NEK operating procedures in following cases and enables following options: filling various tanks, filling the steam generators, filling the containment, additional compressed air source, spent fuel pool refilling and spraying, alternative power supply. This document provides an overview of NEK mobile equipment, which consists of various mobile fire protection pumps, air compressors, protective equipment, fire trucks, diesel generators. Sufficient fuel supply for the equipment is provided on site for a minimum three days of operation. (author).

Beyond design basis external flooding. Generic design assessment and lessons learned from the Fukushima Dai-ichi accident

International Nuclear Information System (INIS)

MacLeod, Tanya; Smith, Leslie; Allmark, Tim; Ford, Peter

2017-01-01

New reactors intended for construction in GB undergo the Office for Nuclear Regulation's (ONR's) Generic Design Assessment (GDA). GDA is a pre-licensing process that provides requesting parties with the opportunity to demonstrate at an early stage that the design is capable of meeting the legal requirements of Great Britain. During GDA, the intended reactor site may not yet be known. Therefore, requesting parties usually define a 'Generic Site' with characteristics typical for Great Britain. These characteristics should, as far as possible, bound the characteristics of known potential sites so that reactors of the proposed type could potentially be built at various suitable locations. This paper critically reviews ONR's approach to ensuring that external flooding is appropriately addressed at the GDA stage and covers: An overview of ONR's approach to post-Fukushima assessment. Changes to ONR's SAPs (Safety Assessment Principles) related to External Flooding. Two examples of post-Fukushima GDA approaches to External Flooding. Uncertainty and the provision of adequate safety measures. The paper concludes that the identification of potential vulnerabilities in the design to external flooding combined with a consideration of post-Fukushima resilience enhancements has led to increased regulatory confidence in the robustness of new reactor designs in GB against external flooding. (author)

Remedial action plan and site design for stabilization of the inactive uranium mill tailings sites at Rifle, Colorado

Energy Technology Data Exchange (ETDEWEB)

1992-02-01

This appendix assesses the present conditions and data gathered about the two inactive uranium mill tailings sites near Rifle, Colorado, and the designated disposal site six miles north of Rifle in the area of Estes Gulch. It consolidates available engineering, radiological, geotechnical, hydrological, meteorological, and other information pertinent to the design of the Remedial Action Plan (RAP). The data characterize conditions at the mill, tailings, and disposal site so that the Remedial Action Contractor (RAC) may complete final designs for the remedial actions.

Grid fault and design-basis for wind turbines - Final report

DEFF Research Database (Denmark)

Hansen, Anca Daniela; Cutululis, Nicolaos Antonio; Markou, Helen

, have been performed and compared for two cases, i.e. one when the turbine is immediately disconnected from the grid when a grid fault occurs and one when the turbine is equipped with a fault ride-through controller and therefore it is able to remain connected to the grid during the grid fault......This is the final report of a Danish research project “Grid fault and design-basis for wind turbines”. The objective of this project has been to assess and analyze the consequences of the new grid connection requirements for the fatigue and ultimate structural loads of wind turbines....... The fulfillment of the grid connection requirements poses challenges for the design of both the electrical system and the mechanical structure of wind turbines. The development of wind turbine models and novel control strategies to fulfill the TSO’s requirements are of vital importance in this design. Dynamic...

Comparison of deterministic and stochastic techniques for estimation of design basis floods for nuclear power plants

International Nuclear Information System (INIS)

Solomon, S.I.; Harvey, K.D.; Asmis, G.J.K.

1983-01-01

The IAEA Safety Guide 50-SG-S10A recommends that design basis floods be estimated by deterministic techniques using probable maximum precipitation and a rainfall runoff model to evaluate the corresponding flood. The Guide indicates that stochastic techniques are also acceptable in which case floods of very low probability have to be estimated. The paper compares the results of applying the two techniques in two river basins at a number of locations and concludes that the uncertainty of the results of both techniques is of the same order of magnitude. However, the use of the unit hydrograph as the rain fall runoff model may lead in some cases to non-conservative estimates. A distributed non-linear rainfall runoff model leads to estimates of probable maximum flood flows which are very close to values of flows having a 10 6 to 10 7 years return interval estimated using a conservative and relatively simple stochastic technique. Recommendations on the practical application of Safety Guide 50-SG-10A are made and the extension of the stochastic technique to ungauged sites and other design parameters is discussed

Site characterization plan conceptual design report for a high-level nuclear waste repository in salt, vertical emplacement mode: Volume 1

International Nuclear Information System (INIS)

1987-12-01

This Conceptual Design Report describes the conceptual design of a high-level nuclear waste repository in salt at a proposed site in Deaf Smith County, Texas. Waste receipt, processing, packing, and other surface facility operations are described. Operations in the shafts underground are described, including waste hoisting, transfer, and vertical emplacement. This report specifically addresses the vertical emplacement mode, the reference design for the repository. Waste retrieval capability is described. The report includes a description of the layout of the surface, shafts, and underground. Major equipment items are identified. The report includes plans for decommissioning and sealing of the facility. The report discusses how the repository will satisfy performance objectives. Chapters are included on basis for design, design analyses, and data requirements for completion of future design efforts. 105 figs., 52 tabs

Site characterization plan conceptual design report for a high-level nuclear waste repository in salt, vertical emplacement mode: Volume 1

Energy Technology Data Exchange (ETDEWEB)

1987-12-01

This Conceptual Design Report describes the conceptual design of a high-level nuclear waste repository in salt at a proposed site in Deaf Smith County, Texas. Waste receipt, processing, packing, and other surface facility operations are described. Operations in the shafts underground are described, including waste hoisting, transfer, and vertical emplacement. This report specifically addresses the vertical emplacement mode, the reference design for the repository. Waste retrieval capability is described. The report includes a description of the layout of the surface, shafts, and underground. Major equipment items are identified. The report includes plans for decommissioning and sealing of the facility. The report discusses how the repository will satisfy performance objectives. Chapters are included on basis for design, design analyses, and data requirements for completion of future design efforts. 105 figs., 52 tabs.

Site characterization plan conceptual design report for a high-level nuclear waste repository in salt, horizontal emplacment mode: Volume 1

International Nuclear Information System (INIS)

1987-12-01

This Conceptual Design Report describes the conceptual design of a high-level nuclear waste repository in salt at a proposed site in Deaf Smith County, Texas. Waste receipt, processing, packaging, and other surface facility operations are described. Operations in the shafts and underground are described, including waste hoisting, transfer, and horizontal emplacement. This report specifically addresses the horizontal emplacement mode, the passive alternate design for the repository. Waste retrieval capability is described. The report includes a description of the layout of the surface, shafts, and underground. Major equipment items are identified. The report includes plans for decommissioning and sealing of the facility. The report discusses how the repository will satisfy performance objectives. Chapters are included on basis for design, design analyses, and data requirements for completion of future design efforts. 105 figs., 52 tabs

Designing Raster Cells as the Basis for Developing Personal Graphic Language

Directory of Open Access Journals (Sweden)

Jana Z. Vujić

2011-05-01

Full Text Available Continuous work in creating new designer solutions points towards the need to create personal routines as personalcommunication in the relation comprising design, algorithms, and original computer graphics. This paper showsprocedures for developing a control language for creating graphic designs with individual raster elements (screeningelement obtaint by halftoning. Personal commands should set routines in a language understood by the printer andthe designer. The PostScript basis is used because we mix vector and pixel graphics in the same program stream, aswell as different colour systems, and our own raster forms. The printing raster is set with the target of special designmulti-use, and this includes the field of security graphics and art computer reproduction. Each raster form assumesmodifications, creating their raster family. The raster cell content is transformed with PostScript, allowing the settingof basic values, angle and liniature for each pixel separately. Raster cells are mixed in multi-colour graphics to thelevel of individual designs with variable values of parameters determining them.

Research on high level radioactive waste repository seismic design criteria

International Nuclear Information System (INIS)

Jing Xu

2012-01-01

Review seismic hazard analysis principle and method in site suitable assessment process of Yucca Mountain Project, and seismic design criteria and seismic design basis in primary design process. Demonstrated spatial character of seismic hazard by calculated regional seismic hazard map. Contrasted different level seismic design basis to show their differences and relation. Discussed seismic design criteria for preclosure phrase of high level waste repository and preference goal under beyond design basis ground motion. (author)

Supercollider: Design changes

International Nuclear Information System (INIS)

Anon.

1990-01-01

A major task faced by the new US Superconducting Supercollider (SSC) Laboratory as it began life early last year was to prepare a baseline design, cost and schedule for the site in Ellis County, Texas. The conceptual design developed by the SSC Central Design Group at Berkeley in 1986 was for a generic site, and served as the basis of the cost estimate submitted to Congress for fiscal year 1990. Under the terms of the contract between Universities Research Association (URA - the SSC parent organization) and the US Department of Energy to establish the SSC Laboratory, signed in January 1989, a report on the SiteSpecific Conceptual Design Report was complete by December

Supercollider: Design changes

Energy Technology Data Exchange (ETDEWEB)

Anon.

1990-04-15

A major task faced by the new US Superconducting Supercollider (SSC) Laboratory as it began life early last year was to prepare a baseline design, cost and schedule for the site in Ellis County, Texas. The conceptual design developed by the SSC Central Design Group at Berkeley in 1986 was for a generic site, and served as the basis of the cost estimate submitted to Congress for fiscal year 1990. Under the terms of the contract between Universities Research Association (URA - the SSC parent organization) and the US Department of Energy to establish the SSC Laboratory, signed in January 1989, a report on the SiteSpecific Conceptual Design Report was complete by December.

Calculation of particulate dispersion in a design-basis tornadic storm from the Battelle Memorial Institute, Columbus, Ohio

Energy Technology Data Exchange (ETDEWEB)

Pepper, D.W.

1980-10-01

A three-dimensional numerical model is used to calculate ground-level air concentration and deposition (due to precipitation scavenging) after a hypothetical tornado strike at the Battelle Memorial Institute at Columbus, Ohio. Plutonium particles less than 20 ..mu..m in diameter are assumed to be lifted into the tornadic storm cell by the vortex. The rotational characteristics of the tornadic storm are embedded within the larger mesoscale flow of the storm system. The design-basis translational wind value is based on probabilities associated with existing records of tornado strikes in the vicinity of the plant site. Turbulence exchange coefficients are based on empirical values deduced from experimental data in severe storms and from theoretical assumptions obtained from the literature. The method of moments is used to incorporate subgrid-scale resolution of the concentration within a grid cell volume.

Calculation of particulate dispersion in a design-basis tornadic storm from the Battelle Memorial Institute, Columbus, Ohio

International Nuclear Information System (INIS)

Pepper, D.W.

1980-10-01

A three-dimensional numerical model is used to calculate ground-level air concentration and deposition (due to precipitation scavenging) after a hypothetical tornado strike at the Battelle Memorial Institute at Columbus, Ohio. Plutonium particles less than 20 μm in diameter are assumed to be lifted into the tornadic storm cell by the vortex. The rotational characteristics of the tornadic storm are embedded within the larger mesoscale flow of the storm system. The design-basis translational wind value is based on probabilities associated with existing records of tornado strikes in the vicinity of the plant site. Turbulence exchange coefficients are based on empirical values deduced from experimental data in severe storms and from theoretical assumptions obtained from the literature. The method of moments is used to incorporate subgrid-scale resolution of the concentration within a grid cell volume

Ames expedited site characterization demonstration at the former manufactured gas plant site, Marshalltown, Iowa

International Nuclear Information System (INIS)

Bevolo, A.J.; Kjartanson, B.H.; Wonder, J.D.

1996-03-01

The goal of the Ames Expedited Site Characterization (ESC) project is to evaluate and promote both innovative technologies (IT) and state-of-the-practice technologies (SOPT) for site characterization and monitoring. In April and May 1994, the ESC project conducted site characterization, technology comparison, and stakeholder demonstration activities at a former manufactured gas plant (FMGP) owned by Iowa Electric Services (IES) Utilities, Inc., in Marshalltown, Iowa. Three areas of technology were fielded at the Marshalltown FMGP site: geophysical, analytical and data integration. The geophysical technologies are designed to assess the subsurface geological conditions so that the location, fate and transport of the target contaminants may be assessed and forecasted. The analytical technologies/methods are designed to detect and quantify the target contaminants. The data integration technology area consists of hardware and software systems designed to integrate all the site information compiled and collected into a conceptual site model on a daily basis at the site; this conceptual model then becomes the decision-support tool. Simultaneous fielding of different methods within each of the three areas of technology provided data for direct comparison of the technologies fielded, both SOPT and IT. This document reports the results of the site characterization, technology comparison, and ESC demonstration activities associated with the Marshalltown FMGP site. 124 figs., 27 tabs

Some conditions affecting the definition of design basis accidents relating to sodium/water reactions

International Nuclear Information System (INIS)

Bolt, P.R.

1984-01-01

The possible damaging effects of large sodium/water reactions on the steam generator, IHX and secondary circuit are considered. The conditions to be considered in defining the design basis accidents for these components are discussed, together with some of the assumptions that may be associated with design assessments of the scale of the accidents. (author)

Cold shock to aquatic organisms: guidance for power-plant siting, design, and operation

International Nuclear Information System (INIS)

Coutant, C.C.

1977-01-01

Problems of cold-shock damages to aquatic organisms have arisen at some condenser cooling-water discharges of thermal power stations when the warm-water releases have suddenly terminated. The basis for such damage lies in the exposure of resident organisms to a rapid decrease in temperature and a sustained exposure to low temperature that induces abnormal behavioral or physiological performance and often leads to death. Although some spectacular fish kills from cold shock have occurred, the present knowledge of the hydraulic and biological processes involved can provide guidance for the siting, design, and operation of power-plant cooling systems to minimize the likelihood of significant cold-shock effects. Preventing cold-shock damages is one consideration in minimizing overall environmental impacts of power-plant cooling and in balancing plant costs with environmental benefits

Comparative analysis of safety related site characteristics

International Nuclear Information System (INIS)

Andersson, Johan

2010-12-01

This document presents a comparative analysis of site characteristics related to long-term safety for the two candidate sites for a final repository for spent nuclear fuel in Forsmark (municipality of Oesthammar) and in Laxemar (municipality of Oskarshamn) from the point of view of site selection. The analyses are based on the updated site descriptions of Forsmark /SKB 2008a/ and Laxemar /SKB 2009a/, together with associated updated repository layouts and designs /SKB 2008b and SKB 2009b/. The basis for the comparison is thus two equally and thoroughly assessed sites. However, the analyses presented here are focussed on differences between the sites rather than evaluating them in absolute terms. The document serves as a basis for the site selection, from the perspective of long-term safety, in SKB's application for a final repository. A full evaluation of safety is made for a repository at the selected site in the safety assessment SR-Site /SKB 2011/, referred to as SR-Site main report in the following

Comparative analysis of safety related site characteristics

Energy Technology Data Exchange (ETDEWEB)

Andersson, Johan (ed.)

2010-12-15

This document presents a comparative analysis of site characteristics related to long-term safety for the two candidate sites for a final repository for spent nuclear fuel in Forsmark (municipality of Oesthammar) and in Laxemar (municipality of Oskarshamn) from the point of view of site selection. The analyses are based on the updated site descriptions of Forsmark /SKB 2008a/ and Laxemar /SKB 2009a/, together with associated updated repository layouts and designs /SKB 2008b and SKB 2009b/. The basis for the comparison is thus two equally and thoroughly assessed sites. However, the analyses presented here are focussed on differences between the sites rather than evaluating them in absolute terms. The document serves as a basis for the site selection, from the perspective of long-term safety, in SKB's application for a final repository. A full evaluation of safety is made for a repository at the selected site in the safety assessment SR-Site /SKB 2011/, referred to as SR-Site main report in the following

Interior design conceptual basis

CERN Document Server

Sully, Anthony

2015-01-01

Maximizing reader insights into interior design as a conceptual way of thinking, which is about ideas and how they are formulated. The major themes of this book are the seven concepts of planning, circulation, 3D, construction, materials, colour and lighting, which covers the entire spectrum of a designer’s activity. Analysing design concepts from the view of the range of possibilities that the designer can examine and eventually decide by choice and conclusive belief the appropriate course of action to take in forming that particular concept, the formation and implementation of these concepts is taken in this book to aid the designer in his/her professional task of completing a design proposal to the client. The purpose of this book is to prepare designers to focus on each concept independently as much as possible, whilst acknowledging relative connections without unwarranted influences unfairly dictating a conceptual bias, and is about that part of the design process called conceptual analysis. It is assu...

Unites States position paper on sodium fires. Design basis and testing

International Nuclear Information System (INIS)

Lancet, R.T.; Johnson, R.P.; Matlin, E.; Vaughan, E.U.; Fields, D.E.; Glueckler, E.; McCormack, J.D.; Miller, C.W.; Pedersen, D.R.

1989-01-01

This paper focuses on designs, analyses, and tests performed since the last Sodium Fires Meeting of the IAEA International Working Group on Fast Reactors in May 1982. Since the U.S. Liquid Metal Reactor (LMR) program is focused on the two advanced LMRs, SAFR and PRISM, the paper relates this work to these designs. First, the design philosophy and approach taken by these advanced pool reactors are described. This includes methods of leak detection, the design basis leaks, and passive accommodation of sodium fires. Then the small- and large-scale sodium fire tests performed in support of the Clinch River Breeder Reactor Plant (CRBRP) program, including post-accident cleanup, are presented and related to the advanced LMR designs. Next, the assessment and behavior of the aerosols generated are discussed including generation rate, behavior within structures, release and dispersal, and deposition on safety-grade equipment. Finally, the impact of these aerosols on the performance of safety-grade decay heat removal heat exchange surfaces is discussed including some test results as well as planned tests. (author)

Direct-to-consumer advertising via the Internet: the role of Web site design.

Science.gov (United States)

Sewak, Saurabh S; Wilkin, Noel E; Bentley, John P; Smith, Mickey C

2005-06-01

Recent attempts to propose criteria for judging the quality of pharmaceutical and healthcare Web sites do not distinguish between attributes of Web site design related to content and other attributes not related to the content. The Elaboration Likelihood Model from persuasion literature is used as a framework for investigating the effects of Web site design on consequents like attitude and knowledge acquisition. A between-subjects, 2 (high or low involvement)x2 (Web site designed with high or low aspects of visual appeal) factorial design was used in this research. College students were randomly assigned to these treatment groups yielding a balanced design with 29 observations per treatment cell. Analysis of variance results for the effects of involvement and Web site design on attitude and knowledge indicated that the interaction between the independent variables was not significant in both analyses. Examination of main effects revealed that participants who viewed the Web site with higher visual appeal actually had slightly lower knowledge scores (6.32) than those who viewed the Web site with lower visual appeal (7.03, F(1,112)=3.827, P=.053). Results of this research seem to indicate that aspects of Web site design (namely aspects of visual appeal and quality) may not play a role in attaining desired promotional objectives, which can include development of favorable attitudes toward the product and facilitating knowledge acquisition.

Strategic Basis for License Application Planning for a Potential Yucca Mountain Repository

International Nuclear Information System (INIS)

Newberry, C. M.; Brocoum, S. J.; Gamble, R. P.; Murray, R. C.; Cline, M.

2002-01-01

If Yucca Mountain, Nevada is designated as the site for development of a geologic repository for disposal of spent nuclear fuel and high-level radioactive waste, the Department of Energy (DOE) must obtain Nuclear Regulatory Commission (NRC) approval first for repository construction, then for an operating license, and, eventually, for repository closure and decommissioning. The licensing criteria defined in Code of Federal Regulations, Title 10, Part 63 (10 CFR Part 63) establish the basis for these NRC decisions. Submittal of a license application (LA) to the NRC for authorization to construct a repository at the Yucca Mountain site is, at this point, only a potential future action by the DOE. The policy process defined in the Nuclear Waste Policy Act (NWPA), as amended, for recommendation and designation of Yucca Mountain as a repository site makes it difficult to predict whether or when the site might be designated. The DOE may only submit a LA to the NRC if the site designation takes effect. In spite of this uncertainty, the DOE must take prudent and appropriate action now, and over the next several years, to prepare for development and timely submittal of a LA. This is particularly true given the need for the DOE to develop, load, and certify the operation of its electronic information system to provide access to its relevant records as part of the licensing support network (LSN) in compliance with NRC requirements six months prior to LA submittal. The DOE must also develop a LA, which is a substantially different document from those developed to support a Site Recommendation (SR) decision. The LA must satisfy NRC licensing criteria and content requirements, and address the acceptance criteria defined by the NRC in its forthcoming Yucca Mountain Review Plan (YMRP). The content of the LA must be adequate to facilitate NRC acceptance and docketing for review, and the LA and its supporting documents must provide the documented basis for the NR C findings required

Design criteria and design basis for the 100-HR-3 and 100-KR-4 pump-and-treat projects

International Nuclear Information System (INIS)

McKinley, W.S.; Winters, J.N.

1996-06-01

The 100-HR-3 and 100-KR-4 Operable Units are located in the 100 Area at the Hanford Site in Richland, Washington. The document describes the project objectives and design criteria to be used for the 100-HR-3 and 100-KR-4 groundwater pump-and-treat design activities

Status of small reactor designs without on-site refuelling

International Nuclear Information System (INIS)

2007-01-01

There is an ongoing interest in member states in the development and application of small and medium sized reactors (SMRs). In the near term, most new NPPs are likely to be evolutionary designs building on proven systems while incorporating technological advances and often the economics of scale, resulting from the reactor outputs of up to 1600 MW(e). For the longer term, the focus is on innovative designs aiming to provide increased benefits in the areas of safety and security, non-proliferation, waste management, resource utilization and economy, as well as to offer a variety of energy products and flexibility in design, siting and fuel cycle options. Many innovative designs are reactors within the small-to-medium size range, having an equivalent electric power less than 700 MW(e) or even less than 300 MW(e). A distinct trend in design and technology development, accounting for about half of the SMR concepts developed worldwide, is represented by small reactors without on-site refuelling. Such reactors, also known as battery-type reactors, could operate without reloading and shuffling of fuel in the core over long periods, from 5 to 25 years and beyond. Upon the advice and with the support of IAEA member states, within its Programme 1 'Nuclear Power, Fuel Cycle, and Nuclear Science', the IAEA provides a forum for the exchange of information by experts and policy makers from industrialized and developing countries on the technical, economic, environmental, and social aspects of SMRs development and implementation in the 21st century, and makes this information available to all interested Member States by producing status reports and other publications dedicated to advances in SMR technology. The objective of this report is to provide Member States, including those just considering the initiation of nuclear power programmes and those already having practical experience in nuclear power, with a balanced and objective information on important development trends and

Ground motion for the design basis earthquake at the Savannah River Site, South Carolina based on a deterministic approach

International Nuclear Information System (INIS)

Youngs, R.R.; Coppersmith, K.J.; Silva, W.J.; Stephenson, D.E.

1991-01-01

Ground motion assessments are presented for evaluation of the seismic safety of K-Reactor at the Savannah River Site. Two earthquake sources were identified as the most significant to seismic hazard at the site, a M 7.5 earthquake occurring at Charleston, South Carolina, and a M 5 event occurring in the site vicinity. These events control the low frequency and high frequency portions of the spectrum, respectively. Three major issues were identified in the assessment of ground motions for the Savannah River site; specification of the appropriate stress drop for the Charleston source earthquake, specification of the appropriate levels of soil damping at large depths for site response analyses, and the appropriateness of western US recordings for specification of ground motions in the eastern US

Calculation of particulate dispersion in a design-basis tornadic storm from Westinghouse PFDL, Cheswick, Pennsylvania

International Nuclear Information System (INIS)

Pepper, D.W.

1978-07-01

A three-dimensional numerical model is used to calculate ground-level air concentration and deposition (due to precipitation scavenging) after a hypothetical tornado strike at the Westinghouse Plutonium Fuel Development Laboratory (PFDL) at Cheswick, Pennsylvania. Plutonium particles less than 20 μm in diameter are assumed to be lifted into the tornadic storm cell by the vortex. The rotational characteristics of the tornadic storm are embedded within the larger mesoscale flow of the storm system. The design-basis translational wind values are based on probabilities associated with existing records of tornado strikes in the vicinity of the plant site. Turbulence exchange coefficients are based on empirical values deduced from experimental data in severe storms and from theoretical assumptions obtained from the literature. The method of moments is used to incorporate subgrid-scale resolution of the concentration within a grid cell volume. This method is a quasi-Lagrangian scheme which minimizes numerical error associated with advection. In all case studies, the effects of updrafts and downdrafts, coupled with scavenging of the particulates by precipitation, account for most of the material being deposited within 20-45 km downwind of the plant site. Ground-level isopleths in the x-y plane show that most of the material is deposited behind and slightly to the left of the centerline trajectory of the storm. Approximately 5% of the material is dispersed into the stratosphere and anvil section of the storm

Brownfields Recommendations for Sustainable Site Design — Green Landscape Plan

Science.gov (United States)

The assessment of conditions contained in this report focuses on site-specific environmental and soil conditions that might affect recommendations related to sustainable landscaping and site design, stormwater management, and stormwater reuse.

Calculation of particulate dispersion in a design-basis tornadic storm from the Babcock and Wilcox Plant, Leechburg, Pennsylvania

Energy Technology Data Exchange (ETDEWEB)

Pepper, D.W.

1978-03-01

A three-dimensional numerical model is used to calculate ground-level air concentration and deposition (due to precipitation scavenging) after a hypothetical tornado strike at the Babcock and Wilcox Plutonium Fabrication Facility at Leechburg, Pennsylvania. Plutonium particles lss than 20 ..mu..m in diameter are assumed to be lifted into the tornadic storm cell by the vortex. The rotational characteristics of the tornadic storm are embedded within the larger mesoscale flow of the storm system. The design-basis translational wind values are based on probabilities associated with existing records of tornado strikes in the vicinity of the plant site. Turbulence exchange coefficients are based on empirical values deduced from experimental data in severe storms and from theoretical assumptions obtained from the literature. The quasi-Lagrangian method of moments is used to model the transport of concentration within a grid cell volume.

Calculation of particulate dispersion in a design-basis tornadic storm from the Babcock and Wilcox Plant, Leechburg, Pennsylvania

International Nuclear Information System (INIS)

Pepper, D.W.

1978-03-01

A three-dimensional numerical model is used to calculate ground-level air concentration and deposition (due to precipitation scavenging) after a hypothetical tornado strike at the Babcock and Wilcox Plutonium Fabrication Facility at Leechburg, Pennsylvania. Plutonium particles lss than 20 μm in diameter are assumed to be lifted into the tornadic storm cell by the vortex. The rotational characteristics of the tornadic storm are embedded within the larger mesoscale flow of the storm system. The design-basis translational wind values are based on probabilities associated with existing records of tornado strikes in the vicinity of the plant site. Turbulence exchange coefficients are based on empirical values deduced from experimental data in severe storms and from theoretical assumptions obtained from the literature. The quasi-Lagrangian method of moments is used to model the transport of concentration within a grid cell volume

Building arrangement and site layout design guides for on site low level radioactive waste storage facilities

International Nuclear Information System (INIS)

McMullen, J.W.; Feehan, M.J.

1986-01-01

Many papers have been written by AE's and utilities describing their onsite storage facilities, why they are needed, NRC regulations, and disposal site requirements. This paper discusses a typical storage facility and address the design considerations and operational aspects that are generally overlooked when designing and siting a low level radioactive waste storage facility. Some topics to be addressed are: 1. Container flexibility; 2. Modular expansion capabilities; 3. DOT regulations; 4. Meterological requirements; 5. OSHA; 6. Fire protection; 7. Floods; 8. ALARA

Preliminary site design for the SP-100 ground engineering test

International Nuclear Information System (INIS)

Cox, C.M.; Miller, W.C.; Mahaffey, M.K.

1986-04-01

In November, 1985, Hanford was selected by the Department of Energy (DOE) as the preferred site for a full-scale test of the integrated nuclear subsystem for SP-100. The Hanford Engineering Development Laboratory, operated by Westinghouse Hanford Company, was assigned as the lead contractor for the Test Site. The nuclear subsystem, which includes the reactor and its primary heat transport system, will be provided by the System Developer, another contractor to be selected by DOE in late FY-1986. In addition to reactor operations, test site responsibilities include preparation of the facility plus design, procurement and installation of a vacuum chamber to house the reactor, a secondary heat transport system to dispose of the reactor heat, a facility control system, and postirradiation examination. At the conclusion of the test program, waste disposal and facility decommissioning are required. The test site must also prepare appropriate environmental and safety evaluations. This paper summarizes the preliminary design requirements, the status of design, and plans to achieve full power operation of the test reactor in September, 1990

Computational design of trimeric influenza-neutralizing proteins targeting the hemagglutinin receptor binding site

Energy Technology Data Exchange (ETDEWEB)

Strauch, Eva-Maria; Bernard, Steffen M.; La, David; Bohn, Alan J.; Lee, Peter S.; Anderson, Caitlin E.; Nieusma, Travis; Holstein, Carly A.; Garcia, Natalie K.; Hooper, Kathryn A.; Ravichandran, Rashmi; Nelson, Jorgen W.; Sheffler, William; Bloom, Jesse D.; Lee, Kelly K.; Ward, Andrew B.; Yager, Paul; Fuller, Deborah H.; Wilson, Ian A.; Baker , David (UWASH); (Scripps); (FHCRC)

2017-06-12

Many viral surface glycoproteins and cell surface receptors are homo-oligomers1, 2, 3, 4, and thus can potentially be targeted by geometrically matched homo-oligomers that engage all subunits simultaneously to attain high avidity and/or lock subunits together. The adaptive immune system cannot generally employ this strategy since the individual antibody binding sites are not arranged with appropriate geometry to simultaneously engage multiple sites in a single target homo-oligomer. We describe a general strategy for the computational design of homo-oligomeric protein assemblies with binding functionality precisely matched to homo-oligomeric target sites5, 6, 7, 8. In the first step, a small protein is designed that binds a single site on the target. In the second step, the designed protein is assembled into a homo-oligomer such that the designed binding sites are aligned with the target sites. We use this approach to design high-avidity trimeric proteins that bind influenza A hemagglutinin (HA) at its conserved receptor binding site. The designed trimers can both capture and detect HA in a paper-based diagnostic format, neutralizes influenza in cell culture, and completely protects mice when given as a single dose 24 h before or after challenge with influenza.

Implementation of an Industrial-Based Case Study as the Basis for a Design Project in an Introduction to Mechanical Design Course

Science.gov (United States)

Lackey, Ellen

2011-01-01

The purpose of this paper is to discuss the implementation of an industrial-based case study as the basis for a design project for the Spring 2009 Introduction to Mechanical Design Course at the University of Mississippi. Course surveys documented the lack of student exposure in classes to the types of projects typically experienced by engineers…

Progress on building design and site layout, and preparation for procurement

International Nuclear Information System (INIS)

Haange, R.

2004-01-01

The ITER site layout and buildings retain many of the major features worked out for the original 1998 design of ITER. By mid-2001, the design had been adjusted to take account of the reductions in size of the tokamak and in the performance and capability of its supporting plant systems, with a view to providing a building construction cost reduction near the target of 50%. The present status of the site layout and buildings is that, with the exception of the hot cell building, designs are available for the ITER buildings and site layout, in different degrees of detail, ready for final review as soon as the drawings and 3D models have been updated. The ITER site layout is determined by the functions of the various plant systems provided in support of the ITER tokamak. These systems are located in different buildings on the ITER site. Changes in the plant layout inside buildings can, therefore, have an important impact on the overall site layout. The general layout of the site assumes electrical power comes in from the west, heat rejection is on the east, access into the main building for assembly is from the south, and removal of waste is from the north. These orientations are purely notional, determining only the relative positions. Both the European and Japanese potential ITER sites will most likely use seismic isolation of the main tokamak complex. As a result, the detailed design of these buildings and in particular the piping connections to them developed for the 'generic' ITER site will require redesign. Also, the footprint of the main tokamak complex will increase, and hence the layout of the utility tunnel network will have to be updated. A draft Site Facilities Procurement Specification has been written by the International Team. This will require considerable specific input when the site has been selected. Furthermore, the present site layout will have to be reviewed in detail. The generic site has been kept compact. Once a site is available, it may be

External human induced events in site evaluation for nuclear power plants. Safety guide

International Nuclear Information System (INIS)

2004-01-01

The purpose of the present Safety Guide is to provide recommendations and guidance for the examination of the region considered for site evaluation for a plant in order to identity hazardous phenomena associated with human induced events initiated by sources external to the plant. In some cases it also presents preliminary guidance for deriving values of relevant parameters for the design basis. This Safety Guide is also applicable for periodic site evaluation and site evaluation following a major human induced event, and for the design and operation of the site's environmental monitoring system. Site evaluation includes site characterization. Consideration of external events that could lead to a degradation of the safety features of the plant and cause a release of radioactive material from the plant and/or affect the dispersion of such material in the environment. And consideration of population issues and access issues significant to safety (such as the feasibility of evacuation, the population distribution and the location of resources). The process of site evaluation continues throughout the lifetime of the facility, from siting to design, construction, operation and decommissioning. The external human induced events considered in this Safety Guide are all of accidental origin. Considerations relating to the physical protection of the plant against wilful actions by third parties are outside its scope. However, the methods described herein may also have some application for the purposes of such physical protection. The present Safety Guide may also be used for events that may originate within the boundaries of the site, but from sources which are not directly involved in the operational states of the nuclear power plant units, such as fuel depots or areas for the storage of hazardous materials for the construction of other facilities at the same site. Special consideration should be given to the hazardous material handled during the construction, operation and

Key issues on safety design basis selection and safety assessment

International Nuclear Information System (INIS)

An, S.; Togo, Y.

1976-01-01

In current fast reactor design in Japan, four design accident conditions and four design seismic conditions are adopted as the design base classifications. These are classified by the considerations on both likelihood of occurrence and the severeness of the consequences. There are several major problem areas in safety design consideration such as core accident problems which include fuel sodium interaction, fuel failure propagation and residual decay heat removal, and decay heat removal systems problems which is more or less the problem of selection of appropriate system and of assurance of high reliability of the system. In view of licensing, two kinds of accidents are postulated in evaluating the adequacy of a reactor site. The one is the ''major accident'' which is the accident to give most severe radiation hazard to the public from technical point of view. The other is the ''hypothetical accident'', induced public accident of which is severer than that of major accident. While the concept of the former is rather unique to Japanese licensing, the latter is almost equivalent to design base hypothetical accident of the US practice. In this paper, design bases selections, key safety issues and some of the licensing considerations in Japan are described

Repository surface design site layout analysis

International Nuclear Information System (INIS)

Montalvo, H.R.

1998-01-01

The purpose of this analysis is to establish the arrangement of the Yucca Mountain Repository surface facilities and features near the North Portal. The analysis updates and expands the North Portal area site layout concept presented in the ACD, including changes to reflect the resizing of the Waste Handling Building (WHB), Waste Treatment Building (WTB), Carrier Preparation Building (CPB), and site parking areas; the addition of the Carrier Washdown Buildings (CWBs); the elimination of the Cask Maintenance Facility (CMF); and the development of a concept for site grading and flood control. The analysis also establishes the layout of the surface features (e.g., roads and utilities) that connect all the repository surface areas (North Portal Operations Area, South Portal Development Operations Area, Emplacement Shaft Surface Operations Area, and Development Shaft Surface Operations Area) and locates an area for a potential lag storage facility. Details of South Portal and shaft layouts will be covered in separate design analyses. The objective of this analysis is to provide a suitable level of design for the Viability Assessment (VA). The analysis was revised to incorporate additional material developed since the issuance of Revision 01. This material includes safeguards and security input, utility system input (size and location of fire water tanks and pump houses, potable water and sanitary sewage rates, size of wastewater evaporation pond, size and location of the utility building, size of the bulk fuel storage tank, and size and location of other exterior process equipment), main electrical substation information, redundancy of water supply and storage for the fire support system, and additional information on the storm water retention pond

On site PWR fuel inspection measurements for operational and design verification

International Nuclear Information System (INIS)

1996-01-01

The on-site inspection of irradiated Pressurized Water Reactor (PWR) fuel and Non-Fuel Bearing Components (NFBC) is typically limited to visual inspections during refuelings using underwater TV cameras and is intended primarily to confirm whether the components will continue in operation. These inspections do not normally provide data for design verification nor information to benefit future fuel designs. Japanese PWR utilities and Nuclear Fuel Industries Ltd. designed, built, and performed demonstration tests of on-site inspection equipment that confirms operational readiness of PWR fuel and NFBC and also gathers data for design verification of these components. 4 figs, 3 tabs

Brownfields Green Avenue Sites: Technical Memorandum - Conceptual Design for Sustainable Redevelopment

Science.gov (United States)

This technical memorandum briefly describes the site and proposed conceptual site plan, indicates conceptual design considerations, specifies recommended green and sustainable features, and offers other recommendations

Maximal design basis accident of fusion neutron source DEMO-TIN

Energy Technology Data Exchange (ETDEWEB)

Kolbasov, B. N., E-mail: Kolbasov-BN@nrcki.ru [National Research Center Kurchatov Institute (Russian Federation)

2015-12-15

When analyzing the safety of nuclear (including fusion) facilities, the maximal design basis accident at which the largest release of activity is expected must certainly be considered. Such an accident is usually the failure of cooling systems of the most thermally stressed components of a reactor (for a fusion facility, it is the divertor or the first wall). The analysis of safety of the ITER reactor and fusion power facilities (including hybrid fission–fusion facilities) shows that the initial event of such a design basis accident is a large-scale break of a pipe in the cooling system of divertor or the first wall outside the vacuum vessel of the facility. The greatest concern is caused by the possibility of hydrogen formation and the inrush of air into the vacuum chamber (VC) with the formation of a detonating mixture and a subsequent detonation explosion. To prevent such an explosion, the emergency forced termination of the fusion reaction, the mounting of shutoff valves in the cooling systems of the divertor and the first wall or blanket for reducing to a minimum the amount of water and air rushing into the VC, the injection of nitrogen or inert gas into the VC for decreasing the hydrogen and oxygen concentration, and other measures are recommended. Owing to a continuous feed-out of the molten-salt fuel mixture from the DEMO-TIN blanket with the removal period of 10 days, the radioactivity release at the accident will mainly be determined by tritium (up to 360 PBq). The activity of fission products in the facility will be up to 50 PBq.

Deep repository. Underground design premises. Edition D1/1

International Nuclear Information System (INIS)

2004-09-01

This report proposes design premises and methodology for application in the preliminary design of underground excavations within the framework of SKB's site investigations. The design premises and supporting instructions for rock mechanical and hydrogeological analyses are a result of planning work carried out between the autumn of 2002 and the spring of 2003. Other overall planning has also been done at the same time for the Deep Repository Project. The first draft underwent revision following a seminar on 30 June 2003 primarily with respect to structure and coordination aspects within the project, as well as certain factual questions. That version served as a basis for an application exercise based on site descriptive model version 1.1 Forsmark. The purpose of the application exercise was to: Test relevant portions of the proposed methodology for rock engineering for the initial site investigations (ISI); Evaluate the application of preliminary design requirements; Identify and develop proposals for collaboration between design and site and site modelling units for the best possible feedback in future work, especially iterations for meeting the needs of design; Identify and develop proposals for collaboration between safety assessment and design for realistic proposals for site adaptation. The design application exercise was evaluated at the end of 2003 and its results serve as an important basis for revisions. The update was commenced with a workshop in February 2004, where the goals were established

10 CFR 435.6 - Sustainable principles for siting, design and construction. [Reserved

Science.gov (United States)

2010-01-01

... 10 Energy 3 2010-01-01 2010-01-01 false Sustainable principles for siting, design and construction. [Reserved] 435.6 Section 435.6 Energy DEPARTMENT OF ENERGY ENERGY CONSERVATION ENERGY EFFICIENCY STANDARDS...-Rise Residential Buildings. § 435.6 Sustainable principles for siting, design and construction...

10 CFR 433.6 - Sustainable principles for siting, design and construction. [Reserved

Science.gov (United States)

2010-01-01

... 10 Energy 3 2010-01-01 2010-01-01 false Sustainable principles for siting, design and construction. [Reserved] 433.6 Section 433.6 Energy DEPARTMENT OF ENERGY ENERGY CONSERVATION ENERGY EFFICIENCY STANDARDS... BUILDINGS § 433.6 Sustainable principles for siting, design and construction. [Reserved] ...

OSU TOMF Program Site Selection and Preliminary Concept Design Report

Energy Technology Data Exchange (ETDEWEB)

Spadling, Steve [Oklahoma State Univ., Stillwater, OK (United States)

2012-05-10

The purpose of this report is to confirm the programmatic requirements for the new facilities, identify the most appropriate project site, and develop preliminary site and building concepts that successfully address the overall project goals and site issues. These new facilities will be designed to accommodate the staff, drivers and maintenance requirements for the future mixed fleet of passenger vehicles, Transit Style Buses and School Buses.

Design/licensing of on-site package for core component

International Nuclear Information System (INIS)

Ogasawara, K.; Chohzuka, T.; Shimura, T.; Kikuchi, T.; Fujiwara, R.; Karigome, S.; Takani, M.

1993-01-01

For storage of used core components which are produced from reactors, Tohoku EPCO decided to construct a site bunker at Onagawa site. It was also decided to develop and fabricate one packaging to transport core components from the reactor buildings to the site bunker. The packaging will be used within the power station; therefore, it shall comply with 'The Law for the Business of Electric Power' and relevant Notification. The main requirements of the packaging are as follows: 1) The number of contents, such as channel boxes and control rods, shall be as large as possible. 2) The weight and the outer dimensions of the packaging shall be within the limitation of the reactor building and the site bunker. 3) Materials shall be selected from those which have been already applied for existing packagings and utilized without any problems. 4) It shall be considered during design of trunnions that handling equipment, such as lifting beam, can be used for not only this packaging but also for existing spent fuel packagings. The design of the packaging is completed and has been licensed. The packaging is scheduled to be utilized from November, 1993. (J.P.N.)

Beyond Information Architecture: A Systems Integration Approach to Web-site Design

Directory of Open Access Journals (Sweden)

Krisellen Maloney

2017-09-01

Full Text Available Users' needs and expectations regarding access to information have fundamentally changed, creating a disconnect between how users expect to use a library Web site and how the site was designed. At the same time, library technical infrastructures include legacy systems that were not designedf or the Web environment. The authors propose a framework that combines elements of information architecture with approaches to incremental system design and implementation. The framework allows for the development of a Web site that is responsive to changing user needs, while recognizing the need for libraries to adopt a cost-effective approach to implementation and maintenance.

The content and design of Web sites : an empirical study

NARCIS (Netherlands)

Huizingh, EKRE

2000-01-01

To support the emergence of a solid knowledge base for analyzing Web activity, we have developed a framework to analyze and categorize the capabilities of Web sites. This distinguishes content from design. Content refers to the information, features, or services that are offered in the Web site,

DESIGNING ALGORITHMS FOR SOLVING PHYSICS PROBLEMS ON THE BASIS OF MIVAR APPROACH

Directory of Open Access Journals (Sweden)

Dmitry Alekseevich Chuvikov

2017-05-01

Full Text Available The paper considers the process of designing algorithms for solving physics problems on the basis of mivar approach. The work also describes general principles of mivar theory. The concepts of parameter, relation and class in mivar space are considered. There are descriptions of properties which every object in Wi!Mi model should have. An experiment in testing capabilities of the Wi!Mi software has been carried out, thus the model has been designed which solves physics problems from year 8 school course in Russia. To conduct the experiment a new version of Wi!Mi 2.1 software has been used. The physics model deals with the following areas: thermal phenomena, electric and electromagnetic phenomena, optical phenomena.

Reconfigurable Flight Control Design using a Robust Servo LQR and Radial Basis Function Neural Networks

Science.gov (United States)

Burken, John J.

2005-01-01

This viewgraph presentation reviews the use of a Robust Servo Linear Quadratic Regulator (LQR) and a Radial Basis Function (RBF) Neural Network in reconfigurable flight control designs in adaptation to a aircraft part failure. The method uses a robust LQR servomechanism design with model Reference adaptive control, and RBF neural networks. During the failure the LQR servomechanism behaved well, and using the neural networks improved the tracking.

Probabilistic site dependent design spectra for a NPP

International Nuclear Information System (INIS)

Chavez, M.; Arroyo, M.; Romo, M.P.

1985-01-01

A methodology is proposed to compute the design spectra for a NPP site. Near field earthquakes are included by using an appropriately scaled sample of response spectra. Site effects are considered through a probabilistic site response analysis in the frequency domain which considers nonlinear behaviour of soils. The uncertainties of the soil shear modulus, G, are introduced by using Rosenblueth's point estimates. Strong motion duration is treated by using sensitivity analysis. The procedure is applied to a NPP site and the results are: a) the USNCR R.G.1.60 underestimate the spectral amplitudes for frequencies of interest; b) the omission of the uncertainties on the G leads to under or over-estimate the spectral amplitudes at certain frequency bands; c) the effect of considering the actual strong motion duration instead of an average value is to reduce the peak spectral amplitudes by a ten per cent. (orig.)

KIT multi-physics tools for the analysis of design and beyond design basis accidents of light water reactors

International Nuclear Information System (INIS)

Sanchez, Victor Hugo; Miassoedov, Alexei; Steinbrueck, M.; Tromm, W.

2016-01-01

This paper describes the KIT numerical simulation tools under extension and validation for the analysis of design and beyond design basis accidents (DBA) of Light Water Reactors (LWR). The description of the complex thermal hydraulic, neutron kinetics and chemo-physical phenomena going on during off-normal conditions requires the development of multi-physics and multi-scale simulations tools which are fostered by the rapid increase in computer power nowadays. The KIT numerical tools for DBA and beyond DBA are validated using experimental data of KIT or from abroad. The developments, extensions, coupling approaches and validation work performed at KIT are shortly outlined and discussed in this paper.

KIT multi-physics tools for the analysis of design and beyond design basis accidents of light water reactors

Energy Technology Data Exchange (ETDEWEB)

Sanchez, Victor Hugo; Miassoedov, Alexei; Steinbrueck, M.; Tromm, W. [Karlsruhe Institute of Technology (KIT), Eggenstein-Leopoldshafen (Germany)

2016-05-15

This paper describes the KIT numerical simulation tools under extension and validation for the analysis of design and beyond design basis accidents (DBA) of Light Water Reactors (LWR). The description of the complex thermal hydraulic, neutron kinetics and chemo-physical phenomena going on during off-normal conditions requires the development of multi-physics and multi-scale simulations tools which are fostered by the rapid increase in computer power nowadays. The KIT numerical tools for DBA and beyond DBA are validated using experimental data of KIT or from abroad. The developments, extensions, coupling approaches and validation work performed at KIT are shortly outlined and discussed in this paper.

Design considerations for on-site spent-fuel transfer systems

International Nuclear Information System (INIS)

Jones, R.H.; Jones, C.R.

1989-06-01

Studies on spent fuel shipping logistics and operation make it clear that the use of large casks, i.e., 100--125 tons, is superior to smaller casks of similar construction. This superiority manifests itself in both transportation and/or shipping economics and safety as well as reduced personnel exposure in the processing of the casks. An on-site system for the transfer of spent fuel from the storage pool to a large shipping or storage cask, as well as the transfer of spent fuel directly from a storage cask to a shipping cask, could bring the large cask benefits to those restricted reactors. Sensing the need to look more closely at this opportunity, EPRI contracted with S. Levy, Incorporated of Campbell, CA to develop a set of design considerations for such transfer systems. Rather then embark on another design study, EPRI decided to first identify the system considerations that must be factored into any design. The format for this effort presents both the Consideration and the Rationale for the consideration. The resulting work identified thirty-six General Considerations and two Special Considerations. The Considerations are in the form of mandatory requirements and desirable but nonmandatory requirements. Additionally, a brief economic study was performed to get a feel for the cost considerations of on-site transfers. The study results suggest a relatively narrow set of scenarios where on-site transfers are economically superior to alternatives. These scenarios generally involve the use of concrete casks as on-site storage devices

Probabilistic risk assessment for the Los Alamos Meson Physics Facility worst-case design-basis accident

International Nuclear Information System (INIS)

Sharirli, M.; Butner, J.M.; Rand, J.L.; Macek, R.J.; McKinney, S.J.; Roush, M.L.

1992-01-01

This paper presents results from a Los Alamos National Laboratory Engineering and Safety Analysis Group assessment of the worse-case design-basis accident associated with the Clinton P. Anderson Meson Physics Facility (LAMPF)/Weapons Neutron Research (WNR) Facility. The primary goal of the analysis was to quantify the accident sequences that result in personnel radiation exposure in the WNR Experimental Hall following the worst-case design-basis accident, a complete spill of the LAMPF accelerator 1L beam. This study also provides information regarding the roles of hardware systems and operators in these sequences, and insights regarding the areas where improvements can increase facility-operation safety. Results also include confidence ranges to incorporate combined effects of uncertainties in probability estimates and importance measures to determine how variations in individual events affect the frequencies in accident sequences

New "persona" concept helps site designers cater to target user segments' needs.

Science.gov (United States)

2004-09-01

Using the relatively new "persona" design concept, Web strategists create a set of archetypical user characters, each one representing one of their site's primary audiences. Then, as their site is constructed or upgraded, they champion the personas, arguing on their behalf and forcing the design team to take each audience's needs and wants into account.

Deep repository. Underground design premises. Edition D1/1

Energy Technology Data Exchange (ETDEWEB)

NONE

2004-09-01

This report proposes design premises and methodology for application in the preliminary design of underground excavations within the framework of SKB's site investigations. The design premises and supporting instructions for rock mechanical and hydrogeological analyses are a result of planning work carried out between the autumn of 2002 and the spring of 2003. Other overall planning has also been done at the same time for the Deep Repository Project. The first draft underwent revision following a seminar on 30 June 2003 primarily with respect to structure and coordination aspects within the project, as well as certain factual questions. That version served as a basis for an application exercise based on site descriptive model version 1.1 Forsmark. The purpose of the application exercise was to: Test relevant portions of the proposed methodology for rock engineering for the initial site investigations (ISI); Evaluate the application of preliminary design requirements; Identify and develop proposals for collaboration between design and site and site modelling units for the best possible feedback in future work, especially iterations for meeting the needs of design; Identify and develop proposals for collaboration between safety assessment and design for realistic proposals for site adaptation. The design application exercise was evaluated at the end of 2003 and its results serve as an important basis for revisions. The update was commenced with a workshop in February 2004, where the goals were established.

Environmental assessment, Richton Dome site, Mississippi

International Nuclear Information System (INIS)

1986-05-01

The Nuclear Waste Policy Act of 1982 (42 USC Sections 10101-10226) requires the environmental assessment of a potential site to include a statement of the basis for the nomination of a site as suitable for characterization. Volume 2 of this environmental assessment provides a detailed evaluation of the Richton Dome Site and its suitability as the site for a radioactive waste disposal facility under DOE siting guidelines, as well as a comparison of the Richton Dome site with other proposed sites. Evaluation of the Richton Dome site is based on the reference repository design, but the evaluation will not change if based on the Mission Plan repository concept. The comparative evaluation of proposed sites is required under DOE guidelines, but is not intended to directly support the subsequent recommendation of three sites for characterization as candidate sites. 428 refs., 24 figs., 62 tabs

Conceptual design of a cover system for the degmay uranium tailings site

Energy Technology Data Exchange (ETDEWEB)

Vatsidin, Saidov; David, S. Kessel; Kim, Chang Lak [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

2016-06-15

The Republic of Tajikistan has ten former uranium mining sites. The total volume of all tailings is approximately 55 million tonnes, and the covered area is more than 200 hectares. The safe management of legacy uranium mining and tailing sites has become an issue of concern. Depending on the performance requirements and site-specific conditions (location in an arid, semiarid or humid region), a cover system for uranium tailings sites could be constructed using several material layers using both natural and man-made materials. The purpose of this study is to find a feasible cost-effective cover system design for the Degmay uranium tailings site which could provide a long period (100 years) of protection. The HELP computer code was used in the evaluation of potential Degmay cover system designs. As a result of this study, a cover system with 70 cm thick percolation layer, 30 cm thick drainage layer, geomembrane liner and 60 cm thick barrier soil layer is recommended because it minimizes cover thickness and would be the most cost-effective design.

NGA-West 2 GMPE average site coefficients for use in earthquake-resistant design

Science.gov (United States)

Borcherdt, Roger D.

2015-01-01

Site coefficients corresponding to those in tables 11.4–1 and 11.4–2 of Minimum Design Loads for Buildings and Other Structures published by the American Society of Civil Engineers (Standard ASCE/SEI 7-10) are derived from four of the Next Generation Attenuation West2 (NGA-W2) Ground-Motion Prediction Equations (GMPEs). The resulting coefficients are compared with those derived by other researchers and those derived from the NGA-West1 database. The derivation of the NGA-W2 average site coefficients provides a simple procedure to update site coefficients with each update in the Maximum Considered Earthquake Response MCER maps. The simple procedure yields average site coefficients consistent with those derived for site-specific design purposes. The NGA-W2 GMPEs provide simple scale factors to reduce conservatism in current simplified design procedures.

Extreme load alleviation using industrial implementation of active trailing edge flaps in a full design load basis

DEFF Research Database (Denmark)

Barlas, Athanasios; Pettas, Vasilis; Gertz, Drew Patrick

2016-01-01

The application of active trailing edge flaps in an industrial oriented implementation is evaluated in terms of capability of alleviating design extreme loads. A flap system with basic control functionality is implemented and tested in a realistic full Design Load Basis (DLB) for the DTU 10MW...

A cliff edge evaluation for CANDU-6 beyond design basis accidents

Energy Technology Data Exchange (ETDEWEB)

Kim, S.M.; Kho, D.W., E-mail: wolsong@khnp.co.kr [Korea Hydro and Nuclear Power Co., Ltd., Daejeon (Korea, Republic of); Yi, S.D.; Kang, S.H.; Kim, S.R. [Nuclear Engineering Service and Solution Co., Ltd., Daejeon (Korea, Republic of)

2015-07-01

The condition of nuclear power plant in the event of station black out (SBO) accompanying large-scale natural disaster exceeding design basis accident (DBA) was evaluated. Additional scenarios were added to the evaluation to review capability of the plant to endure different conditions with different actions. The analysis resulted that the key action required from the operator was to ensure the opening of main steam safety valves (MSSVs) in the secondary side and of motor-operated valves for high pressure injection of Emergency Core Cooling System (HPECCS) to mitigate accidents or extend the cliff edge. (author)

Ambient air monitoring to support HLW repository site characterization

International Nuclear Information System (INIS)

Fransioli, P.M.; Dixon, W.R.

1993-01-01

Site characterization at the Yucca Mountain site includes an ambient air quality and meteorological monitoring program to provide information for environmental and site characterization issues. The program is designed to provide data for four basic purposes: Atmospheric dispersion calculations to estimate impacts of possible airborne releases of radiological material; Engineering design and extreme weather event characterization; Local climate studies for environmental impact analyses and climate characterization; and, Air quality permits required for site characterization work. The program is compiling a database that will provide the basis for analyses and reporting related to the purposes of the program. Except for reporting particulate matter and limited meteorological data to the State of Nevada for an air quality permit condition, the data have yet to be formally analyzed and reported

Design and Modeling of RF Power Amplifiers with Radial Basis Function Artificial Neural Networks

OpenAIRE

Ali Reza Zirak; Sobhan Roshani

2016-01-01

A radial basis function (RBF) artificial neural network model for a designed high efficiency radio frequency class-F power amplifier (PA) is presented in this paper. The presented amplifier is designed at 1.8 GHz operating frequency with 12 dB of gain and 36 dBm of 1dB output compression point. The obtained power added efficiency (PAE) for the presented PA is 76% under 26 dBm input power. The proposed RBF model uses input and DC power of the PA as inputs variables and considers output power a...

Calculation of particulate dispersion in a design-basis tornadic storm from the Exxon Nuclear Company, Richland, Washington

Energy Technology Data Exchange (ETDEWEB)

Pepper, D.W.

1978-07-01

A three-dimensional numerical model is used to calculate ground-level air concentration and deposition (due to precipitation scavenging) after a hypothetical tornado strike at the Exxon Nuclear Company at Richland, Washington. Plutonium particles less than 20 ..mu..m in diameter are assumed to be lifted into the tornadic storm cell by the vortex. The rotational characteristics of the tornadic storm are embedded within the larger mesoscale flow of the storm system. The design-basis translational wind values are based on probabilities associated with existing records of tornado strikes in the vicinity of the plant site. Turbulence exchange coefficients are based on empirical values deduced from experimental data in severe storms and from theoretical assumptions obtained from the literature. The method of moments is used to incorporate subgrid-scale resolution of the concentration within a grid cell volume. This method is a quasi-Lagrangian scheme which minimizes numerical error associated with advection. In all case studies, the effects of updrafts and downdrafts, coupled with scavenging of the particulates by precipitation, account for most of the material being deposited within 20 to 45 km downwind of the plant site. Ground-level isopleths in the x-y plane show that most of the material is deposited behind and slightly to the left of the centerline trajectory of the storm. Approximately 5% of the material is dispersed into the stratosphere and anvil section of the storm.

Calculation of particulate dispersion in a design-basis tornadic storm from the Exxon Nuclear Company, Richland, Washington

International Nuclear Information System (INIS)

Pepper, D.W.

1978-07-01

A three-dimensional numerical model is used to calculate ground-level air concentration and deposition (due to precipitation scavenging) after a hypothetical tornado strike at the Exxon Nuclear Company at Richland, Washington. Plutonium particles less than 20 μm in diameter are assumed to be lifted into the tornadic storm cell by the vortex. The rotational characteristics of the tornadic storm are embedded within the larger mesoscale flow of the storm system. The design-basis translational wind values are based on probabilities associated with existing records of tornado strikes in the vicinity of the plant site. Turbulence exchange coefficients are based on empirical values deduced from experimental data in severe storms and from theoretical assumptions obtained from the literature. The method of moments is used to incorporate subgrid-scale resolution of the concentration within a grid cell volume. This method is a quasi-Lagrangian scheme which minimizes numerical error associated with advection. In all case studies, the effects of updrafts and downdrafts, coupled with scavenging of the particulates by precipitation, account for most of the material being deposited within 20 to 45 km downwind of the plant site. Ground-level isopleths in the x-y plane show that most of the material is deposited behind and slightly to the left of the centerline trajectory of the storm. Approximately 5% of the material is dispersed into the stratosphere and anvil section of the storm

Chemical data for the calculation of fission product releases in design basis faults in PWRs

International Nuclear Information System (INIS)

Ali, S.M.; Bawden, R.J.; Garbett, K.; Deane, A.M.; Large, N.R.

1982-04-01

This review considers the chemistry of caesium and iodine and their volatility under the conditions which would exist during a number of design-basis faults. It recommends values which should be used for the distribution of these elements between liquid and gas phases. (author)

Data Requirements and the Basis for Designing Health Information Kiosks.

Science.gov (United States)

Afzali, Mina; Ahmadi, Maryam; Mahmoudvand, Zahra

2017-09-01

Health kiosks are an innovative and cost-effective solution that organizations can easily implement to help educate people. To determine the data requirements and basis for designing health information kiosks as a new technology to maintain the health of society. By reviewing the literature, a list of information requirements was provided in 4 sections (demographic information, general information, diagnostic information and medical history), and questions related to the objectives, data elements, stakeholders, requirements, infrastructures and the applications of health information kiosks were provided. In order to determine the content validity of the designed set, the opinions of 2 physicians and 2 specialists in medical informatics were obtained. The test-retest method was used to measure its reliability. Data were analyzed using SPSS software. In the proposed model for Iran, 170 data elements in 6 sections were presented for experts' opinion, which ultimately, on 106 elements, a collective agreement was reached. To provide a model of health information kiosk, creating a standard data set is a critical point. According to a survey conducted on the various literature review studies related to the health information kiosk, the most important components of a health information kiosk include six categories; information needs, data elements, applications, stakeholders, requirements and infrastructure of health information kiosks that need to be considered when designing a health information kiosk.

Site characterization plan: Conceptual design report, Volume 1: Chapters 1-3

International Nuclear Information System (INIS)

MacDougall, H.R.; Scully, L.W.; Tillerson, J.R.

1987-09-01

The site for the prospective repository is located at Yucca Mountain in southwestern Nevada, and the waste emplacement area will be constructed in the underlying volcanic tuffs. The target horizon for waste emplacement is a sloping bed of densely welded tuff more than 650 ft below the surface and typically more than 600 ft above the water table. The conceptual design described in this report is unique among repository designs in that it uses ramps in addition to shafts to gain access to the underground facility, the emplacement horizon is located above the water table, and it is possible that 300- to 400-ft-long horizontal waste emplacement boreholes will be used. This report summarizes the design bases (site and properties of the waste package), design and performance criteria, and the design analyses performed. The current status of meeting the preclosure performance objectives for licensing and of resolving the repository design and preclosure issues is presented. The repository design presented in this report will be expanded and refined during the advanced conceptual design, the license application design, and the final procurement and construction design phases

Geo-metadata design for the GIS of the pre-selected site for China's high-level radioactive waste repository

International Nuclear Information System (INIS)

Zhong Xia; Wang Ju; Huang Shutao; Wang Shuhong; Gao Min

2008-01-01

The information system for the geological disposal of high-level radioactive waste aims at the integrated management and full application of multi-sourceful information in the research for geological disposal of high-level radioactive waste. And the establishment and operation of the system need geo-metadata's support of multi-sourceful information. In the paper, on the basis of geo-data analysis for pre-selected site of disposal of high-level radioactive waste, we can apply the existing metadata standards. Also we can research and design the content information, management pattern and application for geo-metadata of the multi-sourceful information. (authors)

Screening and analysis of beyond design basis accident of 49-2 SPR

International Nuclear Information System (INIS)

Zhang Yadong; Guo Yue; Wu Yuanyuan; Zou Yao

2015-01-01

The beyond design basis accident was analyzed to ensure safe operation of 49-2 Swimming Pool Reactor (SPR) after design life. Because it's difficult to use PSA method, the unconditional assumed severe accidents were adopted to obtain a conservative result. The main conclusions were obtained by analyzing anticipated transients without scram in station blackout (SBO ATWS), horizontal channel rupture, core uncovering after shutdown and emergency response capacity. The results show that the core is safe in SBO ATWS, and the fuel elements will not melt as long as the core are not exposed in 2.5 h in loss of coolant accident caused by horizontal channel rupture and other factors. The passive siphon breaker function and various ways of emergency core makeup can ensure that the core is not exposed. (authors)

Functional mimicry of a discontinuous antigenic site by a designed synthetic peptide

NARCIS (Netherlands)

Villen, J.; Borras, E.; Schaaper, W.M.M.; Meloen, R.H.; Davila, M.; Domingo, E.; Giralt, E.; Andreu, D.

2002-01-01

Functional reproduction of the discontinuous antigenic site D of foot-and-mouth disease virus (FMDV) has been achieved by means of synthetic peptide constructions that integrate each of the three protein loops that define the antigenic site into a single molecule. The site D mimics were designed on

ITER interim design report package and relevant documents

International Nuclear Information System (INIS)

1996-01-01

This publication documents the technical basis which underlay the Interim Design Report, Cost Review and Safety Analysis submitted to the ITER Councils (IC-8 and IC-9) Records of decisions and the ''ITER Interim Design Report Package''. This publication contains ITER Site Requirements and ITER Site Design Assumptions, TAC-8 Report, SRG Report, CP's Report on Tentative Sequence of Events and Parties' Views on the IDR Package and Parties' Technical Comments on the IDR Package. Figs, tabs

National Ignition Facility subsystem design requirements NIF site improvements SSDR 1.2.1

International Nuclear Information System (INIS)

Kempel, P.; Hands, J.

1996-01-01

This Subsystem Design Requirements (SSDR) document establishes the performance, design, and verification requirements associated with the NIF Project Site at Lawrence Livermore National Laboratory (LLNL) at Livermore, California. It identifies generic design conditions for all NIF Project facilities, including siting requirements associated with natural phenomena, and contains specific requirements for furnishing site-related infrastructure utilities and services to the NIF Project conventional facilities and experimental hardware systems. Three candidate sites were identified as potential locations for the NIF Project. However, LLNL has been identified by DOE as the preferred site because of closely related laser experimentation underway at LLNL, the ability to use existing interrelated infrastructure, and other reasons. Selection of a site other than LLNL will entail the acquisition of site improvements and infrastructure additional to those described in this document. This SSDR addresses only the improvements associated with the NIF Project site located at LLNL, including new work and relocation or demolition of existing facilities that interfere with the construction of new facilities. If the Record of Decision for the PEIS on Stockpile Stewardship and Management were to select another site, this SSDR would be revised to reflect the characteristics of the selected site. Other facilities and infrastructure needed to support operation of the NIF, such as those listed below, are existing and available at the LLNL site, and are not included in this SSDR. Office Building. Target Receiving and Inspection. General Assembly Building. Electro- Mechanical Shop. Warehousing and General Storage. Shipping and Receiving. General Stores. Medical Facilities. Cafeteria services. Service Station and Garage. Fire Station. Security and Badging Services

Probabilistic evaluation of concrete containment capacity for beyond design basis internal pressures

International Nuclear Information System (INIS)

Tang, H.T.; Dameron, R.A.; Rashid, Y.R.

1995-01-01

For beyond design basis internal pressure loading, experimental studies have demonstrated that the most probable failure mode governing the ultimate functional capacity of concrete containments is leak rather than break. Based on leak rates measured in experiments, a prediction formula for leak rate as functions of containment liner size and internal pressure has been postulated. The determination of liner tear is cast in a probabilistic framework. In calculating leakage, particular attention is paid to the evaluation of leakage versus rupture and the loading rates that may be required to leapfrog over a leakage mode. (orig.)

Environmental assessment, Richton Dome site, Mississippi (US)

Energy Technology Data Exchange (ETDEWEB)

none,

1986-05-01

The Nuclear Waste Policy Act of 1982 (42 USC Sections 10101-10226) requires the environmental assessment of a potential site to include a statement of the basis for the nomination of a site as suitable for characterization. Volume 2 of this environmental assessment provides a detailed evaluation of the Richton Dome Site and its suitability as the site for a radioactive waste disposal facility under DOE siting guidelines, as well as a comparison of the Richton Dome site with other proposed sites. Evaluation of the Richton Dome site is based on the reference repository design, but the evaluation will not change if based on the Mission Plan repository concept. The comparative evaluation of proposed sites is required under DOE guidelines, but is not intended to directly support the subsequent recommendation of three sites for characterization as candidate sites. 428 refs., 24 figs., 62 tabs. (MHB)

Site characterization requirements for nuclear-cratering design

International Nuclear Information System (INIS)

Terhune, R.W.; Carlson, R.C.

1977-01-01

A material properties measurement program for the design of large engineering nuclear-excavation projects by computer calculation is presented. Material properties of the site and their relative effect on crater size are analyzed and ordered in relation to their importance in determining the overall cratering efficiency. The measurement program includes both in situ logging and laboratory measurement of core samples, together with the reason for each measurement and its use in the calculations

Study on risk factors of PWR accidents beyond design basis

International Nuclear Information System (INIS)

Ahn, Seung Hoon; Nah, W. J.; Bang, Y. S.; Oh, D. Y.; Oh, S. H.

2005-01-01

Development of the regulatory guidelines for Beyond Design Basis Accidents (BDBA) with high risk requires a detailed investigation of major factors contributing to the event risk. In this study, each event was classified by the level of risk, based on the probabilistic safety assessment results, so that BDBA with high risk could be selected, with consideration of foreign and domestic regulations, and operating experiences. The regulatory requirements and technical backgrounds for the selected accidents were investigated, and effective regulatory approaches for risk reduction of the accidents. The following conclusions were drawn from this study: - Selected high risk BDBA is station blackout, anticipated without scram, total loss of feedwater. - Major contributors to the risk of selected events were investigated, and appropriate assessment of them was recommended for development of the regulatory guidelines

10 CFR 960.3-1-5 - Basis for site evaluations.

Science.gov (United States)

2010-01-01

... comparative evaluations of sites in terms of the capabilities of the natural barriers for waste isolation and.... Comparative site evaluations shall place primary importance on the natural barriers of the site. In such... the sensitivity of the natural barriers to such realistic engineered barriers. For a better...

Systematic approach to designing surface covers for uranium-mill tailings

International Nuclear Information System (INIS)

Beedlow, P.A.; Cadwell, L.L.; McShane, M.C.

1982-01-01

The wide range of environmental conditions present at uranium mill tailings sites precludes the use of a single type of surface cover. Surface covers must be designed on a site-specific basis. To facilitate site specific designs the UMTRA program is developing guidelines for designing surface covers. This paper presents a systematic approach to designing surface covers for tailings that can be applied under any site condition. The approach consists of three phases: (1) An assessment during which the degree of surface protection is determined. (2) A preliminary design that facilitates interaction with those designing other containment system elements. (3) A final design where the cost and effectiveness of the surface cover are determined. The types of information required to apply this approach are discussed

Site locality identification study: Hanford Site. Volume I. Methodology, guidelines, and screening

International Nuclear Information System (INIS)

1980-07-01

Presented in this report are the results of the site locality identification study for the Hanford Site using a screening process. To enable evaluation of the entire Hanford Site, the screening process was applied to a somewhat larger area; i.e., the Pasco Basin. The study consisted of a series of screening steps that progressively focused on smaller areas which are within the Hanford Site and which had a higher potential for containing suitable repository sites for nuclear waste than the areas not included for further study. Five site localities, designated H-1, H-2, H-3, H-4, H-5 (Figure A), varying in size from approximately 10 to 50 square miles, were identified on the Hanford Site. It is anticipated that each site locality may contain one or more candidate sites suitable for a nuclear waste repository. The site locality identification study began with definition of objectives and the development of guidelines for screening. Three objectives were defined: (1) maximize public health and safety; (2) minimize adverse environmental and socioeconomic impacts; and (3) minimize system costs. The screening guidelines have numerical values that provided the basis for the successive reduction of the area under study and to focus on smaller areas that had a higher likelihood of containing suitable sites

Remedial action plan and site conceptual design for stabilization of the inactive uranium mill tailings sites at Rifle, Colorado. Appendix D, Final report

International Nuclear Information System (INIS)

1992-02-01

This appendix assesses the present conditions and data gathered about the two designated inactive uranium mill tailings sites near Rifle, Colorado, and the proposed disposal site six miles north of Rifle in the area of Estes Gulch. It consolidates available engineering, radiological, geotechnical, hydrological, meteorological, and other information pertinent to the design of the Remedial Action Plan (RAP). The data characterize conditions at the mill, tailings, and disposal site so that the Remedial Action Contractor (RAC) may complete final designs for the remedial actions

Remedial action plan and site conceptual design for stabilization of the inactive uranium mill tailings sites at Rifle, Colorado. Appendix D, Final report

Energy Technology Data Exchange (ETDEWEB)

NONE

1992-02-01

This appendix assesses the present conditions and data gathered about the two designated inactive uranium mill tailings sites near Rifle, Colorado, and the proposed disposal site six miles north of Rifle in the area of Estes Gulch. It consolidates available engineering, radiological, geotechnical, hydrological, meteorological, and other information pertinent to the design of the Remedial Action Plan (RAP). The data characterize conditions at the mill, tailings, and disposal site so that the Remedial Action Contractor (RAC) may complete final designs for the remedial actions.

Considerations on Fail Safe Design for Design Basis Accident (DBA) vs. Design Extension Condition (DEC): Lesson Learnt from the Fukushima Accident

International Nuclear Information System (INIS)

Ha, Jun Su; Kim, Sungyeop

2014-01-01

The fail safety design is referred to as an inherently safe design concept where the failure of an SSC (System, Structure or Component) leads directly to a safe condition. Usually the fail safe design has been devised based on the design basis accident (DBAs), because the nuclear safety has been assured by securing the capability to safely cope with DBAs. Currently regards have been paid to the DEC (Design Extension Condition) as an extended design consideration. Hence additional attention should be paid to the concept of the fail safe design in order to consider the DEC, accordingly. In this study, a case chosen from the Fukushima accident is studied to discuss the issue associated with the fail safe design in terms of DBA and DEC standpoints. For the fail safe design to be based both on the DBA and the DEC, a Mode Changeable Fail Safe Design (MCFSD) is proposed in this study. Additional discussions on what is needed for the MCFSD to be applied in the nuclear safety are addressed as well. One of the lessons learnt from the Fukushima accident should include considerations on the fail-safe design in a changing regulatory framework. Currently the design extension condition (DEC) including severe accidents should be considered during designing and licensing NPPs. Hence concepts on the fail safe design need to be changed to be based on not only the DBA but also the DEC. In this study, a case on a fail-safe design chosen from the Fukushima accident is studied to discuss the issue associated with the fail safe design in terms of DBA and DEC conditions. For the fail safe design to be based both on the DBA and the DEC, a Mode Changeable Fail Safe Design (MCFSD) is proposed in this study. Additional discussions on what is needed for the MCFSD to be applied in the nuclear safety are addressed as well

Radial basis function (RBF) neural network control for mechanical systems design, analysis and Matlab simulation

CERN Document Server

Liu, Jinkun

2013-01-01

Radial Basis Function (RBF) Neural Network Control for Mechanical Systems is motivated by the need for systematic design approaches to stable adaptive control system design using neural network approximation-based techniques. The main objectives of the book are to introduce the concrete design methods and MATLAB simulation of stable adaptive RBF neural control strategies. In this book, a broad range of implementable neural network control design methods for mechanical systems are presented, such as robot manipulators, inverted pendulums, single link flexible joint robots, motors, etc. Advanced neural network controller design methods and their stability analysis are explored. The book provides readers with the fundamentals of neural network control system design.   This book is intended for the researchers in the fields of neural adaptive control, mechanical systems, Matlab simulation, engineering design, robotics and automation. Jinkun Liu is a professor at Beijing University of Aeronautics and Astronauti...

A comparison of the consequences of the design basis accident of the Greek Research Reactor with those of a serious realistic accident

International Nuclear Information System (INIS)

Kollas, J.G.; Anoussis, J.N.

1985-12-01

An analysis of the radiological consequences of the design basis and the coolant flow blockage accidents of the Greek Research Reactor is presented. The results indicate that the consequences of the coolant flow blockage accident are practically trivial being 1-2 orders of magnitude lower than the corresponding consequences of the design basis accident. (author)

Remedial design of the Fultz Landfill Site, Byesville, Ohio

International Nuclear Information System (INIS)

Rajaram, V.; Riesing, R.; Bloom, T.

1994-01-01

The Fultz Landfill Superfund (Fultz) site is a 30-acre hazardous waste landfill located near Byesville, Ohio. The site is approximately 75 miles east of Columbus and 3 miles southwest of Cambridge, the largest city in Guernsey County, Ohio. The landfill is situated on the north slope of a ridge that overlies abandoned coal mines in the Upper Freeport Coal seam. The north half of the landfill lies in an unreclaimed strip mine in the Upper Freeport Coal seam, where saturated portions of surface mine spoils and natural soils form the ''shallow aquifer''. The south half of the landfill lies 40 to 50 feet (ft.) above an abandoned, flooded deep mine in the same coal seam. The flooded deep mine forms an aquifer referred to as the ''coal mine aquifer''. This paper presents the results of design studies completed by PRC Environmental Management, Inc. (PRC), during 1993, and the remedial design (RD) of the components specified by the US Environmental Protection Agency (EPA) Record of Decision (ROD) for the Fultz site (EPA 1991). The remedy specified in the ROD includes a multilayer landfill cap that is compliant with Resource Conservation and Recovery Act (RCRA) Subtitle C guidelines, a leachate collection and groundwater extraction and treatment system, and stabilizing mine voids underlying the southern portion of the site. Vinyl chloride is the only contaminant exceeding a maximum contaminant limit (MCL) in the coal mine aquifer

Approach to developing a ground-motion design basis for facilities important to safety at Yucca Mountain

International Nuclear Information System (INIS)

King, J.L.

1990-01-01

This paper discusses a methodology for developing a ground-motion design basis for prospective facilities at Yucca Mountain that are important to safety. The methodology utilizes a guasi-deterministic construct called the 10,000-year cumulative-slip earthquake that is designed to provide a conservative, robust, and reproducible estimate of ground motion that has a one-in-ten chance of occurring during the preclosure period. This estimate is intended to define a ground-motion level for which the seismic design would ensure minimal disruption to operations engineering analyses to ensure safe performance are included

Establishing design basis threats for the physical protection of nuclear materials and facilities

International Nuclear Information System (INIS)

Chetvergov, S.

2001-01-01

In the area of nuclear energy utilization, the Republic of Kazakhstan follows the standards of international legislation and is a participant of the Nuclear Weapons Non-proliferation Treaty as a country that does not have nuclear weapons. In the framework of this treaty, Kazakhstan provides for the measures to ensure the regime of nonproliferation. The Republic signed the Agreement with the IAEA on the guarantee that was ratified by the Presidential Decree in 1995. Now the Government of the RK is considering the Convention on Physical Protection of Nuclear Materials. Kazakhstan legislation in the area of nuclear energy utilization is represented by a set of laws: the main of them is the Law of the Republic of Kazakhstan 'On the utilization of atomic energy', dated April 14, 1997. According to the Law, the issues of physical protection are regulated by interdepartmental guideline documents. Nuclear science and industry of RK include: Enterprises on uranium mining and processing; Ulba metallurgical plant, manufacturing fuel pellets of uranium dioxide for heat release assemblies of RBMK and WWR reactor types, with the enrichment on U235 1.6-4.4%; Power plant in Aktau for heat and power supply and water desalination, based on fast breeder reactor BN-350; Research reactors of National Nuclear Center: WWR-K - water-water reactor, with 10 MW power, uses highly enriched uranium (up to 36% of U-235); IVG.1M - water-water heterogeneous reactor of vessel type on thermal neutrons, maximum power is 35 MW; IGR - impulse homogeneous graphite reactor on thermal neutrons, with graphite reflector; RA - high temperature gas cooled reactor on thermal neutrons, 0.5 MW power. The establishment of design basis threats for nuclear objects in the Republic of Kazakhstan is an urgent problem because of the developing military-political situation in the region. It is necessary to specify important elements affecting the specific features of the design basis threat: military operations of

Investigation of layered covers designed to limit infiltration at waste disposal sites

International Nuclear Information System (INIS)

Johnson, T.M.; Herzog, B.L.; Cartwright, K.; Larson, T.H.

1983-01-01

Layered soils of highly contrasting texture have been shown to act as a barrier to infiltration because of differences in unsaturated hydraulic properties. Water infiltrating into a fine-grained soil overlying an unsaturated gravel will not enter the gravel until the overlying material is nearly saturated. As a result of what is commonly called the wick effect, moisture will flow laterally through the fine-grained soil above the gravel layer. Layered covers of compacted soil materials utilizing this phenomenon have been proposed to limit infiltration at waste disposal sites. A study funded by the US Nuclear Regulatory Commission is underway to evaluate the potential use of layered trench covers to minimize infiltration at low-level radioactive waste disposal sites. Moisture movement through layered covers consisting of compacted fine-grained soils and sand or gravel was modeled, using a one-dimensional finite-difference model to compare with laboratory column experiments and a two-dimensional finite-element model for field-scale simulations. Soil-moisture characteristic data were obtained from laboratory experiments and the hydraulic conductivity was calculated as a function of capillary pressure. The numerical simulations indicate that layered covers using various combinations of compacted fine-grained soils and coarse-grained material are able to substantially reduce infiltration. Laboratory column experiments using a dual-gamma ray attenuation system to measure moisture content and density simultaneously have been used to observe the behavior of layered soils and to select materials for field testing. Currently underway are field-scale studies of alternative cover designs selected on the basis of two-dimensional modeling

Final disposal of spent nuclear fuel - basis for site selection

International Nuclear Information System (INIS)

Anttila, P.

1995-05-01

International organizations, e.g. IAEA, have published several recommendations and guides for the safe disposal of radioactive waste. There are three major groups of issues affecting the site selection process, i.e. geological, environmental and socioeconomic. The first step of the site selection process is an inventory of potential host rock formations. After that, potential study areas are screened to identify sites for detailed investigations, prior to geological conditions and overall suitability for the safe disposal. This kind of stepwise site selection procedure has been used in Finland and in Sweden. A similar approach has been proposed in Canada, too. In accordance with the amendment to the Nuclear Energy Act, that entered into force in the beginning of 1995, Imatran Voima Oy has to make preparations for the final disposal of spent fuel in the Finnish bedrock. Relating to the possible site selection, the following geological factors, as internationally recommended and used in the Nordic countries, should be taken into account: topography, stability of bedrock, brokenness and fracturing of bedrock, size of bedrock block, rock type, predictability and natural resources. The bedrock of the Loviisa NPP site is a part of the Vyborg rapakivi massif. As a whole the rapakivi granite area forms a potential target area, although other rock types or areas cannot be excluded from possible site selection studies. (25 refs., 7 figs.)

Design Characteristics as Basis for Design Languages

DEFF Research Database (Denmark)

Mortensen, Niels Henrik

1997-01-01

The application of modern feature based CAD systems has in many companies lead to significant rationalisation of design, particulary the "down stream" acticities such as NC code generation, FEM analysis, mould flow simulation and documentation. The subject of this paper is the "up stream" activit......The application of modern feature based CAD systems has in many companies lead to significant rationalisation of design, particulary the "down stream" acticities such as NC code generation, FEM analysis, mould flow simulation and documentation. The subject of this paper is the "up stream...

Molecular Basis for Drug Resistance in HIV-1 Protease

Directory of Open Access Journals (Sweden)

Celia A. Schiffer

2010-11-01

Full Text Available HIV-1 protease is one of the major antiviral targets in the treatment of patients infected with HIV-1. The nine FDA approved HIV-1 protease inhibitors were developed with extensive use of structure-based drug design, thus the atomic details of how the inhibitors bind are well characterized. From this structural understanding the molecular basis for drug resistance in HIV-1 protease can be elucidated. Selected mutations in response to therapy and diversity between clades in HIV-1 protease have altered the shape of the active site, potentially altered the dynamics and even altered the sequence of the cleavage sites in the Gag polyprotein. All of these interdependent changes act in synergy to confer drug resistance while simultaneously maintaining the fitness of the virus. New strategies, such as incorporation of the substrate envelope constraint to design robust inhibitors that incorporate details of HIV-1 protease’s function and decrease the probability of drug resistance, are necessary to continue to effectively target this key protein in HIV-1 life cycle.

Natural phenomena hazards, Hanford Site, south central Washington

International Nuclear Information System (INIS)

Tallman, A.M.

1996-01-01

This document presents the natural phenomena hazard (NPH) loads for use in implementing DOE Order 5480.28, Natural Phenomena Hazards Mitigation, at the Hanford Site in south-central Washington State. The purpose of this document is twofold: (1) summarize the NPH that are important to the design and evaluation of structures, systems, and components at the Hanford Site; (2) develop the appropriate natural phenomena loads for use in the implementation of DOE Order 5480.28. The supporting standards, DOE-STD-1020-94, Natural Phenomena Hazards Design and Evaluation Criteria for Department of Energy Facilities (DOE 1994a); DOE-STD-1022-94, Natural Phenomena Hazards Site Characteristics Criteria (DOE 1994b); and DOE-STD-1023-95, Natural Phenomena Hazards Assessment Criteria (DOE 1995) are the basis for developing the NPH loads

Environmental assessment, Deaf Smith County site, Texas

Energy Technology Data Exchange (ETDEWEB)

1986-05-01

The Nuclear Waste Policy Act of 1982 (42 USC sections 10101-10226) requires the environmental assessment of a proposed site to include a statement of the basis for nominating a site as suitable for characterization. Volume 2 provides a detailed statement evaluating the site suitability of the Deaf Smith County Site under DOE siting guidelines, as well as a comparison of the Deaf Smith County Site to the other sites under consideration. The evaluation of the Deaf Smith County Site is based on the impacts associated with the reference repository design, but the evaluation will not change if based on the Mission Plan repository concept. The second part of this document compares the Deaf Smith County Site to Davis Canyon, Hanford, Richton Dome and Yucca Mountain. This comparison is required under DOE guidelines and is not intended to directly support subsequent recommendation of three sites for characterization as candidate sites. 259 refs., 29 figs., 66 refs. (MHB)

Environmental assessment, Deaf Smith County site, Texas

International Nuclear Information System (INIS)

1986-05-01

The Nuclear Waste Policy Act of 1982 (42 USC sections 10101-10226) requires the environmental assessment of a proposed site to include a statement of the basis for nominating a site as suitable for characterization. Volume 2 provides a detailed statement evaluating the site suitability of the Deaf Smith County Site under DOE siting guidelines, as well as a comparison of the Deaf Smith County Site to the other sites under consideration. The evaluation of the Deaf Smith County Site is based on the impacts associated with the reference repository design, but the evaluation will not change if based on the Mission Plan repository concept. The second part of this document compares the Deaf Smith County Site to Davis Canyon, Hanford, Richton Dome and Yucca Mountain. This comparison is required under DOE guidelines and is not intended to directly support subsequent recommendation of three sites for characterization as candidate sites. 259 refs., 29 figs., 66 refs

Remedial action plan and site design for stabilization of the inactive uranium mill tailings sites at Slick Rock, Colorado

International Nuclear Information System (INIS)

1993-07-01

The Slick Rock uranium mill tailings sites are located near the small town of Slick Rock, in San Miguel County, Colorado. There are two designated UMTRA sites at Slick Rock, the Union Carbide (UC) site and the North Continent (NC) site. Both sites are adjacent to the Dolores River. The UC site is approximately 1 mile (mi) [2 kilometers (km)] downstream of the NC site. Contaminated materials cover an estimated 55 acres (ac) [22 hectares (ha)] at the UC site and 12 ac (4.9 ha) at the NC site. The sites contain former mill building concrete foundations, tailings piles, demolition debris, and areas contaminated by windblown and waterborne radioactive materials. The total estimated volume of contaminated materials is approximately 620, 000 cubic yards (yd 3 ) [470,000 cubic meters (m 3 )]. In addition to the contamination at the two processing site areas, four vicinity properties were contaminated. Contamination associated with the UC and NC sites has leached into groundwater

Position paper: Seismic design criteria

International Nuclear Information System (INIS)

Farnworth, S.K.

1995-01-01

The purpose of this paper is to document the seismic design criteria to be used on the Title 11 design of the underground double-shell waste storage tanks and appurtenant facilities of the Multi-Function Waste Tank Facility (MWTF) project, and to provide the history and methodologies for determining the recommended Design Basis Earthquake (DBE) Peak Ground Acceleration (PGA) anchors for site-specific seismic response spectra curves. Response spectra curves for use in design are provided in Appendix A

Safety in nuclear power plant siting. A code of practice

International Nuclear Information System (INIS)

1978-01-01

This publication is brought out within the framework of establishing Codes of Practice and Safety Guides for nuclear power plants: NUSS programme. The scope of the document encompasses site and site-plant interaction factors related to operational states and accident conditions. The purpose of the Code is to give criteria and procedures to be applied as appropriate to operational states and accident conditions, including those which could lead to emergency situations. This Code is mainly concerned with severe events of low probability which relate to the siting of nuclear power plants and have to be considered in designing a particular nuclear power plant. Annex: Examples of natural and man-made events relevant for design basis evaluation

Geological-Hydrological Site Evaluation for NPP Planning

Energy Technology Data Exchange (ETDEWEB)

Faust, Brigitte; Mini, Paolo [Nordostsschweizerische Kraftwerke AG NOK, Parkstrasse 23, 5401 Baden (Switzerland)

2008-07-01

NOK is investigating the potential replacement of the current NPP in Beznau. In order to meet the requirements with respect to a general licence application, geological, seismological, and geotechnical engineering, but also hydrological boundary conditions have been defined. These conditions define the nature of necessary investigations to obtain the geological, seismic, geotechnical and hydrological data themselves forming the basis to determine the site suitability. Viability has to be provided that a NPP can be built and operated at the proposed site without compromising public health, safety and environment. The collected data are also the basis for the design of all safety relevant structures, systems, and components. For example, the latter have to withstand the effects of natural phenomena such as earthquakes and human induced impact such as airplane crash without loosing their capability to perform the assigned safety functions. (authors)

Geological-Hydrological Site Evaluation for NPP Planning

International Nuclear Information System (INIS)

Faust, Brigitte; Mini, Paolo

2008-01-01

NOK is investigating the potential replacement of the current NPP in Beznau. In order to meet the requirements with respect to a general licence application, geological, seismological, and geotechnical engineering, but also hydrological boundary conditions have been defined. These conditions define the nature of necessary investigations to obtain the geological, seismic, geotechnical and hydrological data themselves forming the basis to determine the site suitability. Viability has to be provided that a NPP can be built and operated at the proposed site without compromising public health, safety and environment. The collected data are also the basis for the design of all safety relevant structures, systems, and components. For example, the latter have to withstand the effects of natural phenomena such as earthquakes and human induced impact such as airplane crash without loosing their capability to perform the assigned safety functions. (authors)

Site study plan for EDBH [Engineering Design Boreholes] seismic surveys, Deaf Smith County site, Texas: Revision 1

International Nuclear Information System (INIS)

Hume, H.

1987-12-01

This site study plan describes seismic reflection surveys to run north-south and east-west across the Deaf Smith County site, and intersecting near the Engineering Design Boreholes (EDBH). Both conventional and shallow high-resolution surveys will be run. The field program has been designed to acquire subsurface geologic and stratigraphic data to address information/data needs resulting from Federal and State regulations and Repository program requirements. The data acquired by the conventional surveys will be common-depth- point, seismic reflection data optimized for reflection events that indicate geologic structure near the repository horizon. The data will also resolve the basement structure and shallow reflection events up to about the top of the evaporite sequence. Field acquisition includes a testing phase to check/select parameters and a production phase. The field data will be subjected immediately to conventional data processing and interpretation to determine if there are any anamolous structural for stratigraphic conditions that could affect the choice of the EDBH sites. After the EDBH's have been drilled and logged, including vertical seismic profiling, the data will be reprocessed and reinterpreted for detailed structural and stratigraphic information to guide shaft development. The shallow high-resulition seismic reflection lines will be run along the same alignments, but the lines will be shorter and limited to immediate vicinity of the EDBH sites. These lines are planned to detect faults or thick channel sands that may be present at the EDBH sites. 23 refs. , 7 figs., 5 tabs

Replacement cross-site transfer system project W-058 safety class upgrade summary report

International Nuclear Information System (INIS)

Schlosser, R.L.

1998-01-01

This report evaluates the design of the replacement cross-site transfer system structures, systems, and components for safety related applications as defined in the Tank Waste Remediation Systems Basis for Interim Operations

Remedial design services for Montclair/West Orange and Glen Ridge Superfund sites

International Nuclear Information System (INIS)

Urbaniak, T.F.; Tomiczek, P.W. Jr.

1994-01-01

The Montclair/West Orange and Glen Ridge Superfund Sites are located 12 miles west of New York City in Essex County, New Jersey. The sites are contaminated with waste materials from radium-processing facilities which operated in the area during the early 1900's. The waste materials, containing radium and other radioactive isotopes were placed in three separate landfill sites. Major public health risks are indoor radon gas build-up and indoor/ outdoor gamma radiation. In 1989, the EPA issued a Record of Decision (ROD) which chose excavation and off-site disposal of material as the preferred alternative. The purpose of this presentation is to highlight key elements of the design process for the remedial action at Montclair. Those key elements are as follows: meeting community relations challenges; measuring radioactive contamination; developing plans and specifications; packaging of remedial action contacts; and continually improving both the process and the designs

Safety-related site characteristics - a relative comparison of the Forsmark reference areas

International Nuclear Information System (INIS)

Winberg, Anders

2010-12-01

SKB has over the years from 2002 to 2008 conducted site investigations in Forsmark and Laxemar, with associated site modeling, design and safety analysis. In mid-2009 Forsmark was selected on the basis of analysis made as site for a future repository for spent nuclear fuel. Based on defined safety-related geoscientific location factors data from Forsmark are compared in relative terms with data from a number of locations in Sweden, previously studied by SKB. The factors compared include: the rock's composition and structures, future climate evolution, rock mechanical conditions, earthquakes, groundwater flow, groundwater composition, delay of solutes, and the ability to characterize and describe the location. Past comparisons of these properties for the selected sites show that none of these sites collectively show any significant benefit over Forsmark site for a repository. This does not preclude that there may be places on the basis of an overall assessment of geoscientific location factors could be equivalent to Forsmark

How to open & operate a financially successful web site design business

CERN Document Server

Evans, Charlotte

2009-01-01

The Pricing & Ethical Guidelines Handbook published by the Graphic Arts Guild reports that the average cost of designing a Web site for a small corporation can range from 7,750 to 15,000. It is incredibly easy to see the enormous profit potential. Web design businesses can be run part- or full-time and can easily be started in your own home. As such, they are one of the fastest growing segments of the Internet economy. Here is the manual you need to cash in on this highly profitable segment of the industry. This book is a comprehensive and detailed study of the business side of Web site des

Hydrologic and geologic aspects of low-level radioactive-waste site management

International Nuclear Information System (INIS)

Cutshall, N.H.; Vaughan, N.D.; Haase, C.S.; Olsen, C.R.; Huff, D.D.

1982-01-01

Hydrologic and geologic site characterization is a critical phase in development of shallow land-burial sites for low-level radioactive-waste disposal, especially in humid environments. Structural features such as folds, faults, and bedding and textural features such as formation permeability, porosity, and mineralogy all affect the water balance and water movement and, in turn, radionuclide migration. Where these features vary over short distance scales, detailed mapping is required in order to enable accurate model predictions of site performance and to provide the basis for proper design and planning of site-disposal operations

Engineered surface barriers for waste disposal sites: lysimeter facility design and construction

International Nuclear Information System (INIS)

Phillips, S.J.; Ruben, M.S.; Kirkham, R.R.

1988-01-01

A facility to evaluate performance of engineered surface carriers for confinement of buried wastes has been designed, constructed, and operations initiated. The Field Lysimeter Test Facility is located at the US Department of Energy's Hanford Site in Richland, Washington. The facility consists of 18 one-dimensional drainage and weighing lysimeters used to evaluate 7 replicated barrier treatments. Distinct layers of natural earth materials were used to construct layered soil and rock barriers in each lysimeter. These barrier designs are capable in principal of significantly reducing or precluding infiltration of meteoric water through barriers into underlying contaminated zones. This paper summarizes salient facility design and construction features used in testing of the Hanford Site's engineered surface barriers

Why there is a need to revise the design basis threat concept

International Nuclear Information System (INIS)

Kondratov, S.; Steinhaeusler, F.

2005-01-01

Full text: The coordinated terrorist attacks in the United States on September 11, 2001, necessitated the review of the proven concept of the Design Basis Threat (DBT) for nuclear installations. It is safe to assume that revised and upgraded DBT will result in costly technical solutions. Since infrastructure deficits and financial limitations in many countries have already limited the practical application of the DBT in many instances, the revised threat assessment is likely to worsen the current dissatisfactory situation further. Therefore, a new realism in the use of the DBT concept is proposed, based on a three-level approach. This will enable countries to tailor the design of their physical protection systems in accordance with their means by implementing either a minimum required security level protecting only against the most probable threat, or aiming for an intermediate protection level reflecting the newly introduced AHARA - as high as reasonably achievable - principle, or providing the optimum protection level based on an externally reviewed, fully comprehensive DBT. (author)

Why there is a need to revise the Design Basis Threat concept

International Nuclear Information System (INIS)

Kondratov, S.; Steinhausler, F.

2006-01-01

The terrorist attacks in the USA on 11 September 2001 necessitated a review of the proven concept of the Design Basis Threat (DBT) for nuclear installations. It can be assumed that revised and upgraded DBT will result in costly technical solutions. Since infrastructure deficits and financial limitations in many countries have already limited the practical application of the DBT, the revised threat assessment is likely to worsen the current unsatisfactory situation. Therefore, a new realism in the use of the DBT concept is proposed based on a three-level approach. This will enable countries to tailor the design of their physical protection systems in accordance with their means by implementing either a minimum required security level protecting only against the most probable threat, or an intermediate protection level reflecting the newly introduced AHARA (As High As Reasonably Achievable) principle, or the optimum protection level based on an externally reviewed, fully comprehensive DBT. (author)

The concept of risk in the design basis threat

International Nuclear Information System (INIS)

Reynolds, J.M.

2001-01-01

Full text: Mathematically defined, risk is a product of one or more probability factors and one or more consequences. Actuarial analysis of risk requires the creation of a numeric algorithm that reflects the interaction of different probability factors, where probability data usually draws on direct measurements of incidence. For physical protection purposes, the algorithms take the general form: Risk = Probability of successful attack x Consequence where the overall probability of a successful attack will be determined by the product of, amongst other things, the probability of there being sufficient intent, the probability of there being available hostile resources, the probability of deterrence, and the probability that a hostile act will be detected and prevented. Deliberate, malevolent acts against nuclear facilities are rare. In so far as it is possible to make an actuarial type of judgement, the probability of malevolent activity against a nuclear facility is almost zero. This creates a problem for a numerical assessment of risk for nuclear facilities where the value (consequence) term could be almost infinite. As can be seen from the general equation above, a numerical algorithm of risk of malevolent activity affecting nuclear facilities could only yield a zero or infinite result. In such circumstances, intelligence-based threat assessments are sometimes thought of as a substitute for historic data in the determination of probability. However, if the paucity of historic data reflects the actual threat - which by and large it should - no amount of intelligence is likely to yield a substantially different conclusion. This mathematical approach to analysing risk appears to lead us either to no risk and no protection or to an infinite risk demanding every conceivable protective measure. The Design Basis Threat (DBT) approach offers a way out of the dilemma. Firstly, it allows us to eliminate from further consideration all zero or near zero probabilities

75 FR 54497 - Ocean Dumping; Guam Ocean Dredged Material Disposal Site Designation

Science.gov (United States)

2010-09-08

.... SUMMARY: The EPA is designating the Guam Deep Ocean Disposal Site (G- DODS) as a permanent ocean dredged... administration of ocean disposal permits; (2) development and maintenance of a site monitoring program; (3... include: (1) Regulating quantities and types of material to be disposed, including the time, rates, and...

Siting of research reactors

International Nuclear Information System (INIS)

1987-01-01

The purpose of this document is to develop criteria for siting and the site-related design basis for research reactors. The concepts presented in this document are intended as recommendations for new reactors and are not suggested for backfitting purposes for facilities already in existence. In siting research reactors serious consideration is given to minimizing the effects of the site on the reactor and the reactor on the site and the potential impact of the reactor on the environment. In this document guidance is first provided on the evaluation of the radiological impact of the installation under normal reactor operation and accident conditions. A classification of research reactors in groups is then proposed, together with a different approach for each group, to take into account the relevant safety problems associated with facilities of different characteristics. Guidance is also provided for both extreme natural events and for man-induced external events which could affect the safe operation of the reactor. Extreme natural events include earthquakes, flooding for river or coastal sites and extreme meteorological phenomena. The feasibility of emergency planning is finally considered for each group of reactors

Nevada test site low-level and mixed waste repository design in the unsaturated zone

International Nuclear Information System (INIS)

Kawamura, T.A.; Warren, D.M.

1989-01-01

The Area 5 Radioactive Waste Management Site (RWMS) at the Nevada Test Site (NTS) is used for shallow land disposal of Low-Level Radioactive (LLW) and for retrievable disposal of Mixed Wastes (MW) from various Department of Energy (DOE) facilities. The site is situated in southern Nevada, one of the most arid regions of the United States. Design considerations include vadose zone monitoring in lieu of groundwater monitoring, stringent waste acceptance and packaging criteria, a waste examination and real-time radiography facility, and trench design. 4 refs

Three types of children’s informational web sites: an inventory of design conventions

NARCIS (Netherlands)

Jochmann-Mannak, Hanna; Lentz, Leo; Huibers, Theo W.C.; Sanders, Ted

2012-01-01

"Purpose: Research on Web design conventions has an almost exclusive focus on Web design for adults. There is far less knowledge about Web design for children. For the first time, an overview is presented of the current design conventions for children's informational Web sites. Method: In this study

Regulatory standpoints on the design-basis capability of safety-related motor-operated valves(MOVs) and power-operated gate valves(POGVs)

International Nuclear Information System (INIS)

Kim, W. T.; Kum, O. H.

1999-01-01

The weakness in the design-basis capability of Motor-Operated Valves(MOVs) and the susceptibility to Pressure Locking and Thermal Binding phenomena of Power-Operated Gate Valves(POGVs) have been major concerns to be resolved in the nuclear society in and abroad since Three Mile Island accident occurred in the USA in 1979. Through detailed analysis of operating experience and regulatory activities, some MOVs and POGVs have been found to be unreliable in performing their safety functions when they are required to do so under certain conditions, especially under design-basis accident conditions. Further, it is well understood that these safety problems may not be identified by the typical valve in-service testing(IST). USNRC has published three Generic Letters, GL 89-10, GL 95-07, and GL 96-05, requiring nuclear plant licensees to take appropriate actions to resolve the problems mentioned above. Korean nuclear regulatory body has made public an administration measure called 'Regulatory recommendation to verify safety functions of the safety-related MOVs and POGVs' on June 13, 1997, and in this administration measure Korean utility is asked to submit written documents to show how it assure design-basis capability of these valves. The following are among the major concerns being considered from a regulation standpoint. Program scope and implementation priority, dynamic tests under differential pressure conditions, accuracy of diagnostic equipment, torque switch setting and torque bypass percentage, weak link analysis, motor actuator sizing, corrective actions taken to resolve pressure locking and thermal binding susceptibility, and a periodic verification program for the valves once design-basis capability has been verified

Conceptual design considerations for providing hook-up type schemes for tracking beyond design basis events (BDBE) for 700 MWe PHWR project

International Nuclear Information System (INIS)

Vhora, S.F.; Inder Jit; Bhardwaj, S.A.

2005-01-01

A broad review of major nuclear accidents such as Chernobyl reveals that provision of access to the reactor core for cooling purpose had to be made from outside the reactor building by tunneling. Also the NAPS fire incident could be mitigated once the fire water injection to the steam generators could be ensured. In this case the boiler room which was outside the primary containment was accessible relatively easily for mitigation after the initial period. Both of the above had accident scenarios which can be termed Beyond Design Basis (BDBE) since the accident initiation/scenario did not fit into the events under postulated initiating events (PIES) or Design Basis Events (DBEs). These accidents or events reveal that some sort of access to the core or the components inside the Reactor building becomes necessary. It is also to be noted that manual intervention beyond the initial period of half an hour or earlier in the Emergency operating procedure (EOP) is inevitably called for as a recovery action in order to mitigate the severity and minimize long term consequences. This paper attempts to discuss the type of concepts which can give access to the core or associated systems which can then provide continued heat sink. The discussions would include the criteria for design of such concepts and give examples of such concepts already implemented and proposes schemes to be implemented in the 700 MWe Project. (author)

Calculation of particulate dispersion in a design-basis tornadic storm from the General Electric Vallecitos Nuclear Center, Vallecitos, California

International Nuclear Information System (INIS)

Pepper, D.W.

1979-11-01

A three-dimensional numerical model is used to calculate ground-level air concentration and deposition (due to precipitation scavenging) after a hypothetical tornado strike at the General Electric Vallecitos Nuclear Center at Vallecitos, California. Plutonium particles less than 20 μm in diameter are assumed to be lifted into the tornadic storm cell by the vortex. The rotational characteristics of the tornadic storm are embedded within the larger mesoscale flow of the storm system. The design-basis translational wind values are based on probabilities associated with existing records of tornado strikes in the vicinity of the plant site. Turbulence exchange coefficients are based on empirical values deduced from experimental data in severe storms and from theoretical assumptions obtained from the literature. The method of moments is used to incorporate subgrid-scale resolution of the concentration within a grid cell volume. This method is a quasi-Lagrangian scheme which minimizes numerical error associated with advection. In all case studies, the effects of updrafts and downdrafts, coupled with scavenging of the particulates by precipitation, account for most of the material being deposited within 20 to 45 km downwind of the plant site. Ground-level isopleths in the x-y plane show that most of the material is deposited behind and slightly to the left of the centerline trajectory of the storm. Approximately 5% of the material is dispersed into the stratosphere and anvil section of the storm

Calculation of particulate dispersion in a design-basis tornadic storm from the General Electric Vallecitos Nuclear Center, Vallecitos, California

Energy Technology Data Exchange (ETDEWEB)

Pepper, D.W.

1979-11-01

A three-dimensional numerical model is used to calculate ground-level air concentration and deposition (due to precipitation scavenging) after a hypothetical tornado strike at the General Electric Vallecitos Nuclear Center at Vallecitos, California. Plutonium particles less than 20 ..mu..m in diameter are assumed to be lifted into the tornadic storm cell by the vortex. The rotational characteristics of the tornadic storm are embedded within the larger mesoscale flow of the storm system. The design-basis translational wind values are based on probabilities associated with existing records of tornado strikes in the vicinity of the plant site. Turbulence exchange coefficients are based on empirical values deduced from experimental data in severe storms and from theoretical assumptions obtained from the literature. The method of moments is used to incorporate subgrid-scale resolution of the concentration within a grid cell volume. This method is a quasi-Lagrangian scheme which minimizes numerical error associated with advection. In all case studies, the effects of updrafts and downdrafts, coupled with scavenging of the particulates by precipitation, account for most of the material being deposited within 20 to 45 km downwind of the plant site. Ground-level isopleths in the x-y plane show that most of the material is deposited behind and slightly to the left of the centerline trajectory of the storm. Approximately 5% of the material is dispersed into the stratosphere and anvil section of the storm.

Constructing sites at a large scale - towards new design (education) methods

DEFF Research Database (Denmark)

Braae, Ellen Marie; Tietjen, Anne

2010-01-01

of the design disciplines within the development of our urban landscapes. At the same time, urban and landscape designers are confronted with new methodological problems. Within a strategic transformation perspective the formulation of the design problem or brief becomes an integrated part of the design process......Since the 1990s the regional scale has regained importance in urban and landscape design. In parallel, the focus in design tasks has shifted from master plans for urban extension to strategic urban transformation projects. The current paradigm of planning by projects reinforces the role....... This paper discusses new design (education) methods based on a relational concept of urban sites and design processes using the actor-network-theory as theoretical frame....

The EPR-a comprehensive design concept against external events

International Nuclear Information System (INIS)

Stoll, U.; Waas, U.

2006-01-01

The main objective of design provisions against external hazards is to ensure that the safety functions required to bring the plant to safe shutdown are not inadmissibly affected by any external hazards that might be postulated for the intended site of the plant. In the design of the European Pressurized Water Reactor (EPR) particular attention was paid to external hazards such as earthquake, airplane crash, and explosion pressure wave. The standard EPR covers a large range of possible site conditions, the design earthquake enveloping safe shutdown earthquakes (SSE) to be expected for potential sites. The loads for the design basis airplane crash and - if required - for the design extension airplane crash as well as for external Explosion Pressure Wave are defined depending on site specific requirements. Protection against other external load cases such as extreme winds and external flooding is also included in the standard design

Design of an environmental site assessment template for open radioactive site contamination : a radioecological risk approach and case study

International Nuclear Information System (INIS)

Nguyen, T.

2004-01-01

To reduce redundancy, cost, and time, while at the same time ultimately increasing the effectiveness of the radioactive risk management process, a logical framework incorporating risk assessments (human cancer and environmental risks) into the environmental site assessment process was designed for radioactive open site contamination. Risk-based corrective action is becoming an increasingly more acceptable approach for the remediation of contaminated sites. In the past, cleanup goals were usually established without any regard to the risk involved, by mandating remediation goals based solely on maximum contamination levels. Now, a multi-stage environmental site assessment template has been developed on a radioecological approach. The template gives a framework for making environmentally sound decisions based on relevant regulations and guidelines. The first stage involves the comparison of the background screening activity level to the regulated activity level, the second stage involves the use of site-specific information to determine the risk involved with the contamination, and the third stage provides a remediation decision matrix based on results from the first two stages. This environmental site assessment template is unique because it incorporates the modified Canadian National Classification System for radioactive contaminated sites and two different types of risk assessments (human cancer risks and the newly designed ecological risk) into the decision making process. The template was used to assess a radiologically contaminated site at the Canadian Forces Base at Suffield (Alberta) as a case study, and it reaffirms the Department of National Defence's action as appropriate. This particular site is a Class 3, has an overall insignificant human cancer risk ( -6 ) and a low environmental risk, and conforms to all regulated guidelines. Currently, it is restricted and should be left as is, provided that the subsurface is not disturbed. (author)

iROCS: Integrated accident management framework for coping with beyond-design-basis external events

International Nuclear Information System (INIS)

Kim, Jaewhan; Park, Soo-Yong; Ahn, Kwang-Il; Yang, Joon-Eon

2016-01-01

Highlights: • An integrated mitigating strategy to cope with extreme external events, iROCS, is proposed. • The strategy aims to preserve the integrity of the reactor vessel as well as core cooling. • A case study for an extreme damage state is performed to assess the effectiveness and feasibility of candidate mitigation strategies under an extreme event. - Abstract: The Fukushima Daiichi accident induced by the Great East Japan earthquake and tsunami on March 11, 2011, poses a new challenge to the nuclear society, especially from an accident management viewpoint. This paper presents a new accident management framework called an integrated, RObust Coping Strategy (iROCS) to cope with beyond-design-basis external events (BDBEEs). The iROCS approach is characterized by classification of various plant damage conditions (PDCs) that might be impacted by BDBEEs and corresponding integrated coping strategies for each of PDCs, aiming to maintain and restore core cooling (i.e., to prevent core damage) and to maintain the integrity of the reactor pressure vessel if it is judged that core damage may not be preventable in view of plant conditions. From a case study for an extreme damage condition, it showed that candidate accident management strategies should be evaluated from the viewpoint of effectiveness and feasibility against accident scenarios and extreme damage conditions of the site, especially when employing mobile or portable equipment under BDBEEs within the limited time available to achieve desired goals such as prevention of core damage as well as a reactor vessel failure.

Providing a Scientific and Technical Basis for Repository Decisions

International Nuclear Information System (INIS)

2001-01-01

The Nuclear Waste Policy Act (NWPA) of 1982 directed the U. S. Department of Energy (DOE) to research sites and design a deep geologic repository for the disposal of our nation's spent nuclear fuel and high-level radioactive waste. In 1987, Congress amended the NWPA and directed the DOE to focus only on Yucca Mountain, Nevada, to determine whether it is a suitable site for a repository. For more than 15 years, the DOE has been studying Yucca Mountain and has accumulated an enormous amount of scientific and technical information about the mountain and the area surrounding it. The secretary of energy will decide whether to recommend Yucca Mountain to the president as a suitable site for a repository. This decision will be based on the scientific and technical information resulting from the Department of Energy's studies of Yucca Mountain and on the views and comments submitted by other governmental groups and the public. One required basis for the secretary's decision will be a scientific analysis called a total system performance assessment

Radiotracers for per studies of neurotransmitter binding sites: Design considerations

International Nuclear Information System (INIS)

Kilbourn, M.R.

1991-01-01

Neurotransmitter binding sites, such as receptors, neuronal uptake systems, and vesicular uptake systems, are important targets for new radiopharmaceutical design. Selection of potential radioligands can be guided by in vitro laboratory data including such characteristics as selectivity and affinity for specific binding sites. However, development of PET radiotracers for use in vivo must include considerations of in vivo pharmacokinetics and metabolism. Introduction of potential radioligands is further narrowed by the demands of the radiochemical synthesis, which must produce radioligands of high chemical and radiochemical purity and of high specific activity. This paper will review examples of previous and current attempts by radiopharmaceutical chemists to meet these demands for new positron emitter-labeled radioligands for PET studies of a wide array of neurotransmitter binding sites

Evaluation of dynamic properties, local site effects and design ground motions: recent advances

International Nuclear Information System (INIS)

Sitharam, T.G.; Vipin, K.S.; James, Naveen

2011-01-01

Evidences from past earthquakes clearly shows that the damages due to an earthquake and its severity at a site are controlled mainly by three factors i.e., earthquake source and path characteristics, local geological and geotechnical characteristics, structural design and quality of the construction. Seismic ground response at a site is strongly influenced by local geological and soil conditions. The exact information of the geological, geomorphological and geotechnical data along with seismotectonic details are necessary to evaluate the ground response. The geometry of the subsoil structure, the soil type, the lateral discontinuities and the surface topography will also influence the site response at a particular location. In the case of a nuclear power plant, the details obtained from the site investigation will have multiple objectives: (i) for the effective design of the foundation (ii) assessment of site amplification (iii) for liquefaction potential evaluation. Since the seismic effects on the structure depend fully on the site conditions and assessment of site amplification. The first input required in evaluation of geotechnical aspect of seismic hazard is the rock level peak horizontal acceleration (PHA) values. The surface level acceleration values need to be calculated based on the site conditions and site amplification values. This paper discusses various methods for evaluating the site amplification values, dynamic soil properties, different field and laboratory tests required and various site classification schemes. In addition to these aspects, the evaluation of liquefaction potential of the site is also presented. The paper highlights on the latest testing methods to evaluate dynamic properties (shear modulus and damping ratio) of soils and techniques for estimating local site effects. (author)

41 CFR 102-76.20 - What issues must Federal agencies consider in providing site planning and landscape design services?

Science.gov (United States)

2010-07-01

... agencies consider in providing site planning and landscape design services? 102-76.20 Section 102-76.20... What issues must Federal agencies consider in providing site planning and landscape design services? In providing site planning and design services, Federal agencies must— (a) Make the site planning and landscape...

Design Basis Knowledge Management for New Build Projects & Ageing Plants - A Perspective

International Nuclear Information System (INIS)

Weightman, Mike

2013-01-01

Summary: • KM for Design Basis of New and Ageing nuclear facilities is at a crossroads; • Needs leadership, vision, cultural change and resources; • Outcome of this workshop is vital; • Information is not knowledge; • Knowledge includes the WHAT, the HOW, the WHY, the Environment and, importantly, Application; • In general, Industry and Regulators are behind the curve; • Develop and apply the principles rigorously; • Keep it simple - focus first on Leadership, values (e.g. questioning attitude), culture, and prioritise – risk informed; • KM is a complex organic creature and needs to be nurtured, fed, learn, grow, evolve in response to a changing environment, and discharge what is not needed to prosper

Remedial action plan and site design for stabilization of the inactive uranium mill tailings sites at Rifle, Colorado. Volume 2, Appendices D and E: Final report

Energy Technology Data Exchange (ETDEWEB)

1992-02-01

This appendix assesses the present conditions and data gathered about the two inactive uranium mill tailings sites near Rifle, Colorado, and the designated disposal site six miles north of Rifle in the area of Estes Gulch. It consolidates available engineering, radiological, geotechnical, hydrological, meteorological, and other information pertinent to the design of the Remedial Action Plan (RAP). The data characterize conditions at the mill, tailings, and disposal site so that the Remedial Action Contractor (RAC) may complete final designs for the remedial actions.

Extreme load alleviation using industrial implementation of active trailing edge flaps in a full design load basis

OpenAIRE

Barlas, Athanasios; Pettas, Vasilis; Gertz, Drew Patrick; Aagaard Madsen , Helge

2016-01-01

The application of active trailing edge flaps in an industrial oriented implementation is evaluated in terms of capability of alleviating design extreme loads. A flap system with basic control functionality is implemented and tested in a realistic full Design Load Basis (DLB) for the DTU 10MW Reference Wind Turbine (RWT) model and for an upscaled rotor version in DTU's aeroelastic code HAWC2. The flap system implementation shows considerable potential in reducing extreme loads in components o...

Conceptual Design Report: Nevada Test Site Mixed Waste Disposal Facility Project

International Nuclear Information System (INIS)

2009-01-01

Environmental cleanup of contaminated nuclear weapons manufacturing and test sites generates radioactive waste that must be disposed. Site cleanup activities throughout the U.S. Department of Energy (DOE) complex are projected to continue through 2050. Some of this waste is mixed waste (MW), containing both hazardous and radioactive components. In addition, there is a need for MW disposal from other mission activities. The Waste Management Programmatic Environmental Impact Statement Record of Decision designates the Nevada Test Site (NTS) as a regional MW disposal site. The NTS has a facility that is permitted to dispose of onsite- and offsite-generated MW until November 30, 2010. There is not a DOE waste management facility that is currently permitted to dispose of offsite-generated MW after 2010, jeopardizing the DOE environmental cleanup mission and other MW-generating mission-related activities. A mission needs document (CD-0) has been prepared for a newly permitted MW disposal facility at the NTS that would provide the needed capability to support DOE's environmental cleanup mission and other MW-generating mission-related activities. This report presents a conceptual engineering design for a MW facility that is fully compliant with Resource Conservation and Recovery Act (RCRA) and DOE O 435.1, 'Radioactive Waste Management'. The facility, which will be located within the Area 5 Radioactive Waste Management Site (RWMS) at the NTS, will provide an approximately 20,000-cubic yard waste disposal capacity. The facility will be licensed by the Nevada Division of Environmental Protection (NDEP)

International Peer Reviews of Design Basis

International Nuclear Information System (INIS)

Hughes, Peter

2013-01-01

International peer reviews: Design and safety assessment review service: - Review of design requirements; - Review in support of licensing; - Review in support of severe accident management; - Review in support of modifications; - Review in relation to periodic safety, or life extension; - Reviews take place at any time in NPP lifecycle from concept, through design and operations

Geo-scientific basis for the investigation site Oberbauenstock. Volume 1

International Nuclear Information System (INIS)

Schneider, T.R.; Kappeler, S.W.

1984-12-01

The geological and hydrogeological conditions prevailing in the region of the investigation site Oberbauen Stock have been chosen as model data-set, in the frame of project Gewaehr 1985, for a type B (low and intermediate level) waste repository. The report, in its capacity as a reference report for project Gewaehr presents all available, relevant site data applicable to modeling of the site. The sources of the data are presented. Geological studies and investigations, boreholes and geophysical investigations prior to the construction of the Seelisberg tunnel (as well as all underground facilities pertaining to the said tunnel) are described. In the section devoted to geology, the stratigraphy and the tectonics of the helvetic nappes in the area of the site are described. Further, the host rock, the neighbouring formations and the local tectonics are presented. In conclusion the geological boundary conditions of the area are summarized. The section devoted to hydrogeology deals with the observations and measurements carried out at the surface and in the underground facilities. As well, the present state of knowledge concerning the water chemistry is described. The section also includes a description of the regional hydrogeology. Concerning rock mechanics a summary and discussion of all available parameters and excavation classes is presented. A description of all the experiences acquired during construction follows as a second part. The section ends with a presentation of the observations and measurements of the gas content of the host rock and other formations. A section concerning long-term behaviour concludes the report. Possible future changes of the geology and the climate are investigated. The consequences for the model site are discussed. It is concluded that a negative impact to a repository on the site is improbable on a timescale of 10 5 years. (author)

Remedial action plan and site design for stabilization of the inactive uranium processing site at Naturita, Colorado

International Nuclear Information System (INIS)

1994-03-01

The uranium processing site near Naturita, Colorado, is one of 24 inactive uranium mill sites designated to be cleaned up by the US Department of Energy (DOE) under the Uranium Mill Tailings Radiation Control Act of 1978 (UMTRCA), 42 USC section 7901 et seq. Part of the UMTRCA requires that the US Nuclear Regulatory Commission (NRC) concur with the DOE's remedial action plan (RAP) and certify that the remedial action conducted at the site complies with the standards promulgated by the US Environmental Protection Agency (EPA). Included in the RAP is this Remedial Action Selection Report (RAS), which describes the proposed remedial action for the Naturita site. An extensive amount of data and supporting information has been generated and evaluated for this remedial action. These data and supporting information are not incorporated into this single document but are included or referenced in the supporting documents. The RAP consists of this RAS and four supporting documents or attachments. This Attachment 2, Geology Report describes the details of geologic, geomorphic, and seismic conditions at the Dry Flats disposal site

Hanford Site waste tank farm facilities design reconstitution program plan

International Nuclear Information System (INIS)

Vollert, F.R.

1994-01-01

Throughout the commercial nuclear industry the lack of design reconstitution programs prior to the mid 1980's has resulted in inadequate documentation to support operating facilities configuration changes or safety evaluations. As a result, many utilities have completed or have ongoing design reconstitution programs and have discovered that without sufficient pre-planning their program can be potentially very expensive and may result in end-products inconsistent with the facility needs or expectations. A design reconstitution program plan is developed here for the Hanford waste tank farms facility as a consequence of the DOE Standard on operational configuration management. This design reconstitution plan provides for the recovery or regeneration of design requirements and basis, the compilation of Design Information Summaries, and a methodology to disposition items open for regeneration that were discovered during the development of Design Information Summaries. Implementation of this plan will culminate in an end-product of about 30 Design Information Summary documents. These documents will be developed to identify tank farms facility design requirements and design bases and thereby capture the technical baselines of the facility. This plan identifies the methodology necessary to systematically recover documents that are sources of design input information, and to evaluate and disposition open items or regeneration items discovered during the development of the Design Information Summaries or during the verification and validation processes. These development activities will be governed and implemented by three procedures and a guide that are to be developed as an outgrowth of this plan

Design report for the interim waste containment facility at the Niagara Falls Storage Site

International Nuclear Information System (INIS)

1986-05-01

Low-level radioactive residues from pitchblende processing and thorium- and radium-contaminated sand, soil, and building rubble are presently stored at the Niagara Falls Storage Site (NFSS) in Lewiston, New York. These residues and wastes derive from past NFSS operations and from similar operations at other sites in the United States conducted during the 1940s by the Manhattan Engineer District (MED) and subsequently by the Atomic Energy Commission (AEC). The US Department of Energy (DOE), successor to MED/AEC, is conducting remedial action at the NFSS under two programs: on-site work under the Surplus Facilities Managemnt Program and off-site cleanup of vicinity properties under the Formerly Utilized Sites Remedial Action Program. On-site remedial action consists of consolidating the residues and wastes within a designated waste containment area and constructing a waste containment facility to prevent contaminant migration. The service life of the system is 25 to 50 years. Near-term remedial action construction activities will not jeopardize or preclude implementation of any other remedial action alternative at a later date. Should DOE decide to extend the service life of the system, the waste containment area would be upgraded to provide a minimum service life of 200 years. This report describes the design for the containment system. Pertinent information on site geology and hydrology and on regional seismicity and meteorology is also provided. Engineering calculations and validated computer modeling studies based on site-specific and conservative parameters confirm the adequacy of the design for its intended purposes of waste containment and environmental protection

Methodology to evaluate the site standard seismic motion to a nuclear facility

International Nuclear Information System (INIS)

Soares, W.A.

1983-01-01

For the seismic design of nuclear facilities, the input motion is normally defined by the predicted maximum ground horizontal acceleration and the free field ground response spectrum. This spectrum is computed on the basis of records of strong motion earthquakes. The pair maximum acceleration-response spectrum is called the site standard seismic motion. An overall view of the subjects involved in the determination of the site standard seismic motion to a nuclear facility is presented. The main topics discussed are: basic principles of seismic instrumentation; dynamic and spectral concepts; design earthquakes definitions; fundamentals of seismology; empirical curves developed from prior seismic data; available methodologies and recommended procedures to evaluate the site standard seismic motion. (Author) [pt

Toxic industrial chemicals (TICs) as asymmetric weapons: the design basis threat

International Nuclear Information System (INIS)

Skinner, L.

2009-01-01

Asymmetric warfare concepts relate well to the use of improvised chemical weapons against urban targets. Sources of information on toxic industrial chemicals (TICs) and lists of high threat chemicals are available that point to likely choices for an attack. Accident investigations can be used as a template for attacks, and to judge the possible effectiveness of an attack using TICs. The results of a chlorine rail car accident in South Carolina, USA and the Russian military assault on a Moscow theater provide many illustrative points for similar incidents that mighty be carried out deliberately. Computer modeling of outdoor releases shows how an attack might take into consideration issues of stand-off distance and dilution. Finally, the preceding may be used to estimate with some accuracy the design basis threat posed by the used of TICs as weapons.(author)

Molecular dynamics explorations of active site structure in designed and evolved enzymes.

Science.gov (United States)

Osuna, Sílvia; Jiménez-Osés, Gonzalo; Noey, Elizabeth L; Houk, K N

2015-04-21

This Account describes the use of molecular dynamics (MD) simulations to reveal how mutations alter the structure and organization of enzyme active sites. As proposed by Pauling about 70 years ago and elaborated by many others since then, biocatalysis is efficient when functional groups in the active site of an enzyme are in optimal positions for transition state stabilization. Changes in mechanism and covalent interactions are often critical parts of enzyme catalysis. We describe our explorations of the dynamical preorganization of active sites using MD, studying the fluctuations between active and inactive conformations normally concealed to static crystallography. MD shows how the various arrangements of active site residues influence the free energy of the transition state and relates the populations of the catalytic conformational ensemble to the enzyme activity. This Account is organized around three case studies from our laboratory. We first describe the importance of dynamics in evaluating a series of computationally designed and experimentally evolved enzymes for the Kemp elimination, a popular subject in the enzyme design field. We find that the dynamics of the active site is influenced not only by the original sequence design and subsequent mutations but also by the nature of the ligand present in the active site. In the second example, we show how microsecond MD has been used to uncover the role of remote mutations in the active site dynamics and catalysis of a transesterase, LovD. This enzyme was evolved by Tang at UCLA and Codexis, Inc., and is a useful commercial catalyst for the production of the drug simvastatin. X-ray analysis of inactive and active mutants did not reveal differences in the active sites, but relatively long time scale MD in solution showed that the active site of the wild-type enzyme preorganizes only upon binding of the acyl carrier protein (ACP) that delivers the natural acyl group to the active site. In the absence of bound ACP

The Role of Tectonic and Seismicity in Siting of Nuclear Power Plant

International Nuclear Information System (INIS)

Abdel Aziz, M.A.H.

2008-01-01

The site selection for the a nuclear power plant (NPP) is controlled by many criteria. One of the most important criterion is the tectonic and seismicity of the site and its surroundings. Since, it is preferable the site in concern is characterized by low tectonic and low seismicity to avoid the damage effects associated with the occurrence of destructive earthquakes. The investigation of the tectonic and seismicity maps of egypt has been carried out to candidate potential areas or sites for nuclear power plant installation from seismicity point of view. Also, the design basis ground motion in terms of peak ground acceleration and response spectra of some of the potential sites are defined through the conduct of probabilistic seismic hazard analysis The study revealed that although there is no criterion to exclude areas of high tectonic and high seismicity as potential sites for nuclear power plant installation but, it is preferable avoiding such areas. This is attributed to the critical seismic curve that characterizes such areas and is required high seismic design levels to resist the destructive vibratory ground motion associated with the expected earthquake. Consequently, the required high seismic design levels will have a negative impact on the economic cost of the facility compared with that built in low and moderate seismic areas. Hence, areas like the gulf of suez, the northern part of the Red Sea and the southern part of Sinai Peninsula should be avoided as potential sites for NPP from the tectonic and seismicity point of view. On the other hand, areas like Nile delta and its valley, the Northern and Southern parts of Western desert and the central and southern parts of the Eastern Desert should be candidate as potential sites on condition, the other criteria meet the IAEA's regulations. Also, the seismic hazard curve of the Northwest littoral zone reflects low design basis ground motion values compared with the Nile delta region

The discussion of nuclear power plant's cooling chain design for freezing site

International Nuclear Information System (INIS)

Hu Jian; Yang Ting; Jiang Xulun

2014-01-01

The Component cooling water system (RRI) and Essential service water system (SEC) are composed of Nuclear Power Plant's (NPP) cooling chain, which has its special requirement for freezing site from system design and safety point of view. The feature and difficulty of cooling chain design at freezing condition (when the intake water temperature is below O ℃) are represented. At present, several NPPs are in operation or under construction at freezing site in the world, including Pressurized Water Reactor (PWR) and Canadian Deuterium Uranium reactor (CANDU). By analyzing the thoughts and applicability of different kinds of cooling chain design at freezing site, one solution called 'SEC thermal discharge reflux' is proposed to remove the residual heat from Nuclear Island (NI) into heat sink safely in winter. The solution has been approved by National Nuclear Safety Administration (NNSA) in China and applied in one of CPR NPP in the north of China, which is able to solve several problems compared with the traditional solutions, such as 'Reactor low power operation', 'Reactor start-up for the first time', and 'Changeover of RRI/SEC trains in winter'. The solution is also able to prevent RRI/SEC heat exchanger from icing and avoid low flowrate in SEC pipes. Besides, considering of the economical efficiency, simple operation and control strategy is designed. (authors)

Understanding the evolution of the repository and the olkiluoto site

International Nuclear Information System (INIS)

Koskinen, K.; Pastina, B.

2008-01-01

Posiva Safety Case is organised in a portfolio including ten main reports: Site, Spent Fuel Characteristics and Inventories, Canister Design, Repository Design, Process, Evolution of the Repository and the Site, Biosphere Assessment, Radionuclide Transport, Complementary Evaluations of Safety, and Summary. This portfolio constitutes the basis of the Preliminary Safety Assessment Report, which will be presented to the authorities in 2012 as part of the repository construction license application. The Evolution report [1], which is the focus of this paper, is the main advance in the Safety Case portfolio since the implementation of the Safety Case plan [2] in 2005. The report provides the status of current knowledge with respect to the evolution of the site and the engineered barrier system and highlights areas where better understanding is needed. (authors)

Perceptions of Web Site Design Characteristics: A Malaysian/Australian Comparison.

Science.gov (United States)

Fink, Dieter; Laupase, Ricky

2000-01-01

Compares the perceptions of Malaysians and Australians for four Web site design characteristics--atmospherics, news stories, signs, and products and services--as part of the integrated Internet marketing model. Hypothesizes that the predominant culture is not generalized to another culture, discusses validity and reliability, and suggest further…

Development, use and maintenance of the design basis threat. Implementing guide

International Nuclear Information System (INIS)

2009-01-01

threat to those assets. As described in this publication, an understanding of the threat can lead to a detailed description of potential adversaries (the design basis threat), which, in turn, is the basis of an appropriately designed physical protection system. This direct link gives confidence that protection would be effective against an adversary attack. International experience in using a design basis threat to protect assets of high consequence is largely based on the protection of nuclear material and facilities. Furthermore, the nuclear security documents defining and recommending that physical protection be based upon the threat - The Physical Protection Objectives and Fundamental Principles (GOV/2001/41/ Attachment), the Recommendations on the Physical Protection of Nuclear Facilities and Nuclear Material (INFCIRC/225/Rev. 4 (corrected)), and the Convention on Physical Protection of Nuclear Facilities and Nuclear Material as Amended (INFCIRC/274) (adopted on 8 July 2005; (GOV/2005/57)) - do so exclusively for the protection of nuclear material and facilities. Given the historical background, and its continuing contemporary relevance, it has been necessary to draw on that nuclear protection experience in developing this publication. However, the general approach can also be applied to protecting other assets that require a high degree of confidence in the effectiveness of their protection, such as high-activity radioactive material. Specialists from France, Germany, Japan, the Russian Federation, Spain, the United Kingdom, and the United States of America assisted the IAEA in preparing this publication. A draft was presented to an open-ended technical meeting in December 2006, and subsequently circulated for comment to all Member States. This publication is consistent with The Physical Protection Objectives and Fundamental Principles; the Convention on Physical Protection of Nuclear Facilities and Nuclear Material as Amended; and the Recommendations on the

Conceptual design analyses for Hanford Site deployable remote spectroscopy systems

International Nuclear Information System (INIS)

Philipp, B.L.; Reich, F.R.

1994-09-01

This document identifies potential remote, NIR spectroscopic waste surface moisture monitoring system design alternatives to be operated inside one of the Hanford Site, high level, nuclear waste storage tanks. Potential tank waste moisture data impacts from the remote NIR signal transfer through high humidity vapor space is evaluated

Volcanoes and associated topics in relation to nuclear power plant siting

International Nuclear Information System (INIS)

1997-01-01

The main purpose of this report is to provide draft guidance on the criteria and methods for the evaluation of a site for a nuclear power plant with respect to the potential effects of volcanic activity which may jeopardize its safety and to elicit feedback from Member States. Different types of phenomena associated with volcanism are discussed in terms of their influence on site acceptability and on derivation of design basis parameters. This report was developed for application to new nuclear power plant sites. It does not address the issue of the re-evaluation of existing nuclear power sites to the potential effects of volcanic activity, although it contains general information and criteria useful for this purpose. The guidelines and procedures discussed in this report can appropriately be used as the basis for the safe siting and design of nuclear power plants in different volcanic environments. In this report, the description of the phenomena associated with volcanism and the collection of required data and information are separated from the criteria for hazard assessment. Thus Section 2 gives the non-specialist a general description of the different types of volcanic phenomena and Section 4 provides indication on the acquisition of the database. Section 3 outlines the general requirements to be fulfilled during site selection and evaluation. Sections 5, 6 and 7 provide guidance to perform the hazard assessment and to derive the design basis parameters. Finally, Section 8 deals with monitoring systems. As general information for the non-specialist, Annex I provides the major divisions of geological time. With the same spirit, and recognizing that a complete consensus has not been reached in the scientific community on the use and meaning of some terms, a glossary of volcanological definitions is given in Annex II, applicable only to the use of this report. Finally, Annex III provides an example of a classification of volcanoes that may be used for capability

Central repository for low- and intermediate-level waste (ALMA) conceptual design, siting and safety study

International Nuclear Information System (INIS)

Kjellbert, N.; Haeggblom, H.; Cederstroem, M.; Lundgren, T.

1980-07-01

A generic design, siting and safety study of a proposed repository for low- and intermediate-level waste has been made. Special emphasis has been placed on safety characterostics. The conceptual design and the generic site, on which the study is based, are realistically chosen in accordance with present construction techniques and the existing geohydrological conditions in Sweden. (Auth.)

Environment as a basis for the design of advertising structures by forming

Science.gov (United States)

Khmelevsky, Y. P.; Seryakov, V. A.; Mamontov, G. Y.; Tsarenko, D. T.

2017-01-01

A few different neighbouring styles of architectural forms are quite frequent in the cities of great historical past. As a result, a designer or architect has to solve the complex problem while designing the objects within such environment, i.e. one has to fit them naturally into the existing site development. Often, form making is found to be hard, due to the fact that the existing architectural forms of totally different stylistic execution coexist in the visual proximity. Presently, placement of the advertising bills in urban environment is both an urgent and debatable issue. On the one hand, advertising providers are keen to present their product bigger and brighter, on the other hand, the overall and eye-catching exhibition stands can be disharmonious with the surrounding architectural ensemble of the city. This situation is relevant for every cultural city.

ASME N510 test results for Savannah River Site AACS filter compartments

Energy Technology Data Exchange (ETDEWEB)

Paul, J.D.; Punch, T.M. [Westinghouse Savannah River Company, Aiken, SC (United States)

1995-02-01

The K-Reactor at the Savannah River Site recently implemented design improvements for the Airborne Activity Confinement System (AACS) by procuring, installing, and testing new Air Cleaning Units, or filter compartments, to ASME AG-11, N509, and N510 requirements. Specifically, these new units provide documentable seismic resistance to a Design Basis Accident earthquake, provide 2 inch adsorber beds with 0.25 second residence time, and meet all AG-1, N509, and N510 requirements for testability and maintainability. This paper presents the results of the Site acceptance testing and discusses an issue associated with sample manifold qualification testing.

Tests of qualification of national components of nuclear power plants under design basis accident

International Nuclear Information System (INIS)

Mesquita, A.Z.

1990-01-01

With the purpose of qualifying national components of nuclear power plants, whose working must be maintained during and after an accident, the Thermohydraulic Division of CDTN have done tests to check the equipment stability, under Design Basis Accident conditions. Until this moment, the following components were tested: electrical junction boxes (connectors); coating systems for wall, inside cover and steel containment; hydraulics components of personnel and equipment airlock. This work describes the test instalation, the tests performed and its results. The components tested, in a general way, fulfil the specified requirements. (author) [pt

Reducing Production Basis Risk through Rainfall Intensity Frequency (RIF) Indexes: Global Sensitivity Analysis' Implication on Policy Design

Science.gov (United States)

Muneepeerakul, Chitsomanus; Huffaker, Ray; Munoz-Carpena, Rafael

2016-04-01

The weather index insurance promises financial resilience to farmers struck by harsh weather conditions with swift compensation at affordable premium thanks to its minimal adverse selection and moral hazard. Despite these advantages, the very nature of indexing causes the presence of "production basis risk" that the selected weather indexes and their thresholds do not correspond to actual damages. To reduce basis risk without additional data collection cost, we propose the use of rain intensity and frequency as indexes as it could offer better protection at the lower premium by avoiding basis risk-strike trade-off inherent in the total rainfall index. We present empirical evidences and modeling results that even under the similar cumulative rainfall and temperature environment, yield can significantly differ especially for drought sensitive crops. We further show that deriving the trigger level and payoff function from regression between historical yield and total rainfall data may pose significant basis risk owing to their non-unique relationship in the insured range of rainfall. Lastly, we discuss the design of index insurance in terms of contract specifications based on the results from global sensitivity analysis.

Grid fault and design-basis for wind turbines. Final report

Energy Technology Data Exchange (ETDEWEB)

Hansen, A.D.; Cutululis, N.A.; Markou, H.; Soerensen, Poul; Iov, F.

2010-01-15

This is the final report of a Danish research project 'Grid fault and design-basis for wind turbines'. The objective of this project has been to assess and analyze the consequences of the new grid connection requirements for the fatigue and ultimate structural loads of wind turbines. The fulfillment of the grid connection requirements poses challenges for the design of both the electrical system and the mechanical structure of wind turbines. The development of wind turbine models and novel control strategies to fulfill the TSO's requirements are of vital importance in this design. Dynamic models and different fault ride-through control strategies have been developed and assessed in this project for three different wind turbine concepts (active stall wind turbine, variable speed doublyfed induction generator wind turbine, variable speed multipole permanent magnet wind turbine). A computer approach for the quantification of the wind turbines structural loads caused by the fault ride-through grid requirement, has been proposed and exemplified for the case of an active stall wind turbine. This approach relies on the combination of knowledge from complimentary simulation tools, which have expertise in different specialized design areas for wind turbines. In order to quantify the impact of the grid faults and grid requirements fulfillment on wind turbines structural loads and thus on their lifetime, a rainflow and a statistical analysis for fatigue and ultimate structural loads, respectively, have been performed and compared for two cases, i.e. one when the turbine is immediately disconnected from the grid when a grid fault occurs and one when the turbine is equipped with a fault ride-through controller and therefore it is able to remain connected to the grid during the grid fault. Different storm control strategies, that enable variable speed wind turbines to produce power at wind speeds higher than 25m/s and up to 50m/s without substantially increasing

Licensing topical report: application of probabilistic risk assessment in the selection of design basis accidents

International Nuclear Information System (INIS)

Houghton, W.J.

1980-06-01

A probabilistic risk assessment (PRA) approach is proposed to be used to scrutinize selection of accident sequences. A technique is described in this Licensing Topical Report to identify candidates for Design Basis Accidents (DBAs) utilizing the risk assessment results. As a part of this technique, it is proposed that events with frequencies below a specified limit would not be candidates. The use of the methodology described is supplementary to the traditional, deterministic approach and may result, in some cases, in the selection of multiple failure sequences as DBAs; it may also provide a basis for not considering some traditionally postulated events as being DBAs. A process is then described for selecting a list of DBAs based on the candidates from PRA as supplementary to knowledge and judgments from past licensing practice. These DBAs would be the events considered in Chapter 15 of Safety Analysis Reports of high-temperature gas-cooled reactors

Design and implementation of an identification system in construction site safety for proactive accident prevention.

Science.gov (United States)

Yang, Huanjia; Chew, David A S; Wu, Weiwei; Zhou, Zhipeng; Li, Qiming

2012-09-01

Identifying accident precursors using real-time identity information has great potential to improve safety performance in construction industry, which is still suffering from day to day records of accident fatality and injury. Based on the requirements analysis for identifying precursor and the discussion of enabling technology solutions for acquiring and sharing real-time automatic identification information on construction site, this paper proposes an identification system design for proactive accident prevention to improve construction site safety. Firstly, a case study is conducted to analyze the automatic identification requirements for identifying accident precursors in construction site. Results show that it mainly consists of three aspects, namely access control, training and inspection information and operation authority. The system is then designed to fulfill these requirements based on ZigBee enabled wireless sensor network (WSN), radio frequency identification (RFID) technology and an integrated ZigBee RFID sensor network structure. At the same time, an information database is also designed and implemented, which includes 15 tables, 54 queries and several reports and forms. In the end, a demonstration system based on the proposed system design is developed as a proof of concept prototype. The contributions of this study include the requirement analysis and technical design of a real-time identity information tracking solution for proactive accident prevention on construction sites. The technical solution proposed in this paper has a significant importance in improving safety performance on construction sites. Moreover, this study can serve as a reference design for future system integrations where more functions, such as environment monitoring and location tracking, can be added. Copyright © 2011 Elsevier Ltd. All rights reserved.

Interregional Knowledge Management Workshop on Life Cycle Management of Design Basis Information. Issues, Challenges, Approaches

International Nuclear Information System (INIS)

Šula, Radek

2013-01-01

Introduction and objectives: • It is evident that the design basis area is from the point of view of knowledge sharing extremely complicated. • Time is changing and puts on us ever greater demands. • We have to analyze the near and remote surroundings and have to simplified the problem of knowledge sharing in that area. • I believe that it is graspable task for knowledge management and I will try to outline some possible context and approaches

The Swedish Utilities joint approach to form common basis for design requirements for the future

International Nuclear Information System (INIS)

Hansson, B.

1998-01-01

The Owners of the Swedish Nuclear Power Plants have decided to form a document that should state the design principals and requirement for cost-effective and continuous development of the reactor safety in the future. The development of this document will be a part of the modernization and development of the Swedish Nuclear Power Plants. The basis for this document is an evaluation of Swedish and International standards and regulations as IAEA/INSAG, US-regulations, EUR etc. (author)

Design and control strategy for a hybrid green energy system for mobile telecommunication sites

Science.gov (United States)

Okundamiya, Michael S.; Emagbetere, Joy O.; Ogujor, Emmanuel A.

2014-07-01

The rising energy costs and carbon footprint of operating mobile telecommunication sites in the emerging world have increased research interests in green technology. The intermittent nature of most green energy sources creates the problem of designing the optimum configuration for a given location. This study presents the design analysis and control strategy for a cost effective and reliable operation of the hybrid green energy system (HGES) for GSM base transceiver station (BTS) sites in isolated regions. The design constrains the generation and distribution of power to reliably satisfy the energy demand while ensuring safe operation of the system. The overall process control applies the genetic algorithm-based technique for optimal techno-economic sizing of system's components. The process simulation utilized meteorological data for 3 locations (Abuja, Benin City and Sokoto) with varying climatic conditions in Nigeria. Simulation results presented for green GSM BTS sites are discussed and compared with existing approaches.

Heritage planning and rethinking the meaning and values of designating heritage sites in a post-disaster context: The case of Aceh, Indonesia

Science.gov (United States)

Meutia, Z. D.; Akbar, R.; Zulkaidi, D.

2018-05-01

importance of understanding how local values create heritage, that heritage is not only about material and physical aspects but it is also a cultural process. The methodology used is ethnography with in-depth interviews to uncover meaning and values in this context. This paper uses the examples of how the sites in Aceh that were affected by the tsunami disaster in 2004 such as the PLTD Apung ship and the Mosque of Ulee Lheue are sites that a mutual and role of historical city relating to the potential as signs of civilization, historical values, commemorative values, religious values, and values of consensus within the community as outstanding local values, as a basis of consideration in rethinking the designations of heritage sites of less than 50 years old.

Development of Pupils Picture Aesthetic Competences on the Basis of IT-didactic Designs of Digital Picture Production

DEFF Research Database (Denmark)

Rasmussen, Helle

: The research method refers to Design Based Research, since the project is based on a design theoretical view of learning. (Cobb et. All 2003, Van den Akker 2006, Collins 2004). Learning is here to be understood as “a sign producing activity in a specific situation within an institutional framing”, which makes...... Education” (English Title), The Danish University of Education Cobb, P. et al. (2003): “Design Experiments in Educational Research” in “Educational Researcher”, vol. 32, no. 1. Collins, Allan et. al. (2004): “Design Research: Theoretical and Metodological Issuses” in “Journal of the Learning Sciences”, Vol...... Competences on the Basis of IT-didactic Designs of Digital Picture Production Proposal information: The topic for this presentation is an ongoing investigation of the connection between the learning outcome of digital picture production and IT-didactic designs, and it refers to a Ph.D.-project in progress...

Design and development of self-powered sensors on wireless sensor network for standalone plant critical data management during SBO and beyond design basis events

International Nuclear Information System (INIS)

Aparna, J.; Dulera, I.V.; Rama Rao, A.; Vijayan, P.K.

2015-01-01

Advanced reactors are designed with an aim of maximum safety, optimized fuel utilization and effective system design. Safety aspects in reactor designs are being viewed for all possible vulnerabilities, and as a result, robust self-regulating passive safety features have been favored in Gen IV and advanced reactor designs. In addition to passive systems, the accidents scenarios at Fukushima indicate the dire need of reliable and stand-alone self-powered sensors, for monitoring plant critical parameters for effective damage control actions. There is a strong need for plant critical data management and situation awareness during the unavailability of all conventional power sources in a nuclear power plant, during extended station blackout (SBO) conditions. These self-powered sensors would assist the operators in managing events like SBO and help in containing any Beyond Design Basis Events (BDBE) conditions, well away from the public domain

HTP-OligoDesigner: An Online Primer Design Tool for High-Throughput Gene Cloning and Site-Directed Mutagenesis.

Science.gov (United States)

Camilo, Cesar M; Lima, Gustavo M A; Maluf, Fernando V; Guido, Rafael V C; Polikarpov, Igor

2016-01-01

Following burgeoning genomic and transcriptomic sequencing data, biochemical and molecular biology groups worldwide are implementing high-throughput cloning and mutagenesis facilities in order to obtain a large number of soluble proteins for structural and functional characterization. Since manual primer design can be a time-consuming and error-generating step, particularly when working with hundreds of targets, the automation of primer design process becomes highly desirable. HTP-OligoDesigner was created to provide the scientific community with a simple and intuitive online primer design tool for both laboratory-scale and high-throughput projects of sequence-independent gene cloning and site-directed mutagenesis and a Tm calculator for quick queries.

Conceptual design report for regional low-level waste interim storage site

International Nuclear Information System (INIS)

Bird, M.V.; Thompson, J.D.

1981-08-01

An interim storage site design concept was developed for receiving 100,000 ft 3 low-level waste per year, in the form of solidified wastes in 55-gallon drums with a dose rate of < 200 mrem per hour at contact

Remedial Action Plan and site design for stabilization of the inactive uranium mill tailings site at Rifle, Colorado

International Nuclear Information System (INIS)

1990-02-01

This document has been structured to provide a comprehensive understanding of the remedial action proposed for the Rifle sites. That remedial action consists of removing approximately 4,185,000 cubic yards (cy) of tailings and contaminated materials from their current locations, transporting, and stabilizing the tailings material at the Estes Gulch disposal site, approximately six miles north of Rifle. The tailings and contaminated materials are comprised of approximately 597,000 cy from Old Rifle, 3,232,000 cy from New Rifle, and 322,000 cy from vicinity properties and about 34,000 cy from demolition. The remedial action plan includes specific design requirements for the detailed design and construction of the remedial action. An extensive amount of data and supporting information have been generated for this remedial action and cannot all be incorporated into this document. Pertinent information and data are included with reference given to the supporting documents

Remedial action plan and site design for stabilization of the inactive uranium processing site at Naturita, Colorado

International Nuclear Information System (INIS)

1993-08-01

The US Environmental Protection Agency (EPA) has established health and environmental protection regulations to correct and prevent groundwater contamination resulting from processing activities at inactive uranium milling sites (40 CFR 192). The Uranium Mill Tailings Radiation Control Act (UMTRCA) of 1978 designated responsibility to the US Department of Energy (DOE) for assessing the inactive uranium milling sites. The DOE has determined that each assessment shall include information on site characterization, a description of the proposed action, and a summary of the water resources protection strategy that describes how the proposed action will comply with the EPA groundwater protection standards. To achieve compliance with the proposed US Environmental Protection Agency (EPA) groundwater protection standards, the US Department of Energy (DOE) proposes that supplemental standards be applied at the Dry Flats disposal site because of Class III (limited use) groundwater in the uppermost aquifer (the basal sandstone of the Cretaceous Burro Canyon Formation) based on low yield. The proposed remedial action will ensure protection of human health and the environment

Remedial action plan and site design for stabilization of the inactive uranium processing site at Naturita, Colorado

International Nuclear Information System (INIS)

1993-08-01

The uranium processing site near Naturita, Colorado, is one of 24 inactive uranium mill sites designated to be cleaned up by the US Department of Energy (DOE) under the Uranium Mill Tailings Radiation Control Act of 1978 (UMTRCA), Public Law 95-604. Part of the UMTRCA requires that the US Nuclear Regulatory Commission (NRC) concur with the DOE's remedial action plan (RAP) and certify that the remedial action conducted at the site complies with the standards promulgated by the US Environmental Protection Agency (EPA). Included in the RAP is this Remedial Action Selection Report (RAS), which serves two purposes. First, it describes the activities that are proposed by the DOE to accomplish remediation and long-term stabilization and control of the radioactive materials at the inactive uranium processing site near Naturita, Colorado. Second, this document and the rest of the RAP, upon concurrence and execution by the DOE, the state of Colorado, and the NRC, become Appendix B of the cooperative agreement between the DOE and the State of Colorado

The regulatory basis for site selection for radioactive buildings

International Nuclear Information System (INIS)

Sirag, N.M.

2012-01-01

The architectural bases and the design standard for radiation research buildings differ than other research building, that the nature of the used radioactive materials in the nature of the carried an work in the building will decide the planning in the design of the building, and we shall discus in this research what already had been reached of results about the architectural bases in the used standard in the used safe design of research building and establishing design standard, that archaistic rely upon when designing these project which have specific and special characters

Comment and response document for the final remedial action plan and site design for stabilization of the inactive uranium mill tailings sites at Slick Rock, Colorado. Revision 2

International Nuclear Information System (INIS)

1996-05-01

This document for the final remedial action plan and site design has been prepared for US Department of Energy Environmental Restoration Division as part of the Uranium Mill Tailings Remedial Action plan. Comments and responses are included for the site design for stabilization of the inactive uranium mill tailings sites at Slick Rock, Colorado

Retrofitting a spent fuel pool spray system for alternative cooling as a strategy for beyond design basis events

Energy Technology Data Exchange (ETDEWEB)

Hartmann, Christoph; Vujic, Zoran [Westinghouse Electric Germany GmbH, Mannheim (Germany)

2017-06-15

Due to requirements for nuclear power plants to withstand beyond design basis accidents, including events such as happened in 2011 in the Fukushima Daiichi Nuclear Power Plant in Japan, alternative cooling of spent fuel is needed. Alternative spent fuel cooling can be provided by a retrofitted spent fuel pool spray system based on the AP1000 plant design. As part of Krsko Nuclear Power Plant's Safety Upgrade Program, Krsko Nuclear Power Plant decided on, and Westinghouse successfully designed a retrofit of the AP1000 {sup registered} plant spent fuel pool spray system to provide alternative spent fuel cooling.

Guide to radiological accident considerations for siting and design of DOE nonreactor nuclear facilities

International Nuclear Information System (INIS)

Elder, J.; Graf, J.M.

1984-01-01

DOE Office of Nuclear Safety has sponsored preparation of a guidance document to aid field offices and contractors in their analyses of consequences of postulated major accidents. The guide addresses the requirements of DOE Orders 5480.1A, Chapter V, and 6430.1, including the general requirement that DOE nuclear facilities be sited, designed, and operated in accordance with standards, codes, and guides consistent with those applied to comparable licensed nuclear facilities. The guide includes both philosophical and technical information in the areas of: siting guidelines doses applied to an offsite reference person; consideration also given to an onsite reference person; physical parameters, models, and assumptions to be applied when calculating doses for comparison to siting criteria; and potential accident consequences other than radiological dose to a reference person which might affect siting and major design features of the facility, such as environmental contamination, population dose, and associated public health effects. Recommendations and/or clarifications are provided where this could be done without adding new requirements. In this regard, the guide is considered a valuable aid to the safety analyst, especially where requirements have been subject to inconsistent interpretation or where analysis methods are in transition, such as use of dose model (ICRP 2 or ICRP 30) or use of probabilistic methods of risk analysis in the siting and design of nuclear facilities

SIR - small is safe [in reactor design

International Nuclear Information System (INIS)

Hayns, M.

1989-01-01

A joint USA-UK venture has been initiated to design a small nuclear reactor which offers low capital cost, greater flexibility and a potentially lower environmental impact. Called Safe Integral Reactor (SIR), the lead unit could be built in the United Kingdom Atomic Energy Authority's (UKAEA's) Winfrith site if the design is accepted by the UK Nuclear Installations Inspectorate (NII). This article describes the 320 MWe reactor unit that is the basis of the design being developed. (author)

Are Biophilic-Designed Site Office Buildings Linked to Health Benefits and High Performing Occupants?

Directory of Open Access Journals (Sweden)

Tonia Gray

2014-11-01

Full Text Available This paper discusses the first phase of a longitudinal study underway in Australia to ascertain the broad health benefits of specific types of biophilic design for workers in a building site office. A bespoke site design was formulated to include open plan workspace, natural lighting, ventilation, significant plants, prospect and views, recycled materials and use of non-synthetic materials. Initial data in the first three months was gathered from a series of demographic questions and from interviews and observations of site workers. Preliminary data indicates a strong positive effect from incorporating aspects of biophilic design to boost productivity, ameliorate stress, enhance well-being, foster a collaborative work environment and promote workplace satisfaction, thus contributing towards a high performance workspace. The longitudinal study spanning over two years will track human-plant interactions in a biophilic influenced space, whilst also assessing the concomitant cognitive, social, psychological and physical health benefits for workers.

Estimating open population site occupancy from presence-absence data lacking the robust design.

Science.gov (United States)

Dail, D; Madsen, L

2013-03-01

Many animal monitoring studies seek to estimate the proportion of a study area occupied by a target population. The study area is divided into spatially distinct sites where the detected presence or absence of the population is recorded, and this is repeated in time for multiple seasons. However, when occupied sites are detected with probability p Ecology 84, 2200-2207) developed a multiseason model for estimating seasonal site occupancy (ψt ) while accounting for unknown p. Their model performs well when observations are collected according to the robust design, where multiple sampling occasions occur during each season; the repeated sampling aids in the estimation p. However, their model does not perform as well when the robust design is lacking. In this paper, we propose an alternative likelihood model that yields improved seasonal estimates of p and Ψt in the absence of the robust design. We construct the marginal likelihood of the observed data by conditioning on, and summing out, the latent number of occupied sites during each season. A simulation study shows that in cases without the robust design, the proposed model estimates p with less bias than the MacKenzie et al. model and hence improves the estimates of Ψt . We apply both models to a data set consisting of repeated presence-absence observations of American robins (Turdus migratorius) with yearly survey periods. The two models are compared to a third estimator available when the repeated counts (from the same study) are considered, with the proposed model yielding estimates of Ψt closest to estimates from the point count model. Copyright © 2013, The International Biometric Society.

Calculation of particulate dispersion in a design-basis tornadic storm from the Atomics International Nuclear Material Development Facility, Santa Susana, California

Energy Technology Data Exchange (ETDEWEB)

Pepper, D.W.

1980-07-01

A three-dimensional numerical model is used to calculate ground-level air concentration and deposition (due to precipitation scavenging) after a hypothetical tornado strike at the Atomics International Nuclear Material Development Facility at Santa Susana, California. Plutonium particles less than 20 ..mu..m in diameter are assumed to be lifted into the tornadic storm cell by the vortex. The rotational characteristics of the tornadic storm are embedded within the larger mesoscale flow of the storm system. The design-basis translational wind values are based on probabilities associated with existing records of tornado strikes in the vicinity of the plant site. Turbulence exchange coefficients are based on empirical values deduced from experimental data in severe storms and from theoretical assumptions obtained from the literature. The method of moments is used to incorporate subgrid-scale resolution of the concentration within a grid cell volume. This method is a quasi-Lagrangian scheme which minimizes numerical error associated with advection. In all case studies, the effects of updrafts and downdrafts, coupled with scavenging of the particulates by precipitation, account for most of the material being deposited within 50 km downwind of the plant site. Ground-level isopleths in the x-y plane show that most of the material is deposited behind and slightly to the left of the centerline trajectory of the storm. Approximately 5% of the material is dispersed into the stratosphere and anvil section of the storm.

Calculation of particulate dispersion in a design-basis tornadic storm from the Atomics International Nuclear Material Development Facility, Santa Susana, California

International Nuclear Information System (INIS)

Pepper, D.W.

1980-07-01

A three-dimensional numerical model is used to calculate ground-level air concentration and deposition (due to precipitation scavenging) after a hypothetical tornado strike at the Atomics International Nuclear Material Development Facility at Santa Susana, California. Plutonium particles less than 20 μm in diameter are assumed to be lifted into the tornadic storm cell by the vortex. The rotational characteristics of the tornadic storm are embedded within the larger mesoscale flow of the storm system. The design-basis translational wind values are based on probabilities associated with existing records of tornado strikes in the vicinity of the plant site. Turbulence exchange coefficients are based on empirical values deduced from experimental data in severe storms and from theoretical assumptions obtained from the literature. The method of moments is used to incorporate subgrid-scale resolution of the concentration within a grid cell volume. This method is a quasi-Lagrangian scheme which minimizes numerical error associated with advection. In all case studies, the effects of updrafts and downdrafts, coupled with scavenging of the particulates by precipitation, account for most of the material being deposited within 50 km downwind of the plant site. Ground-level isopleths in the x-y plane show that most of the material is deposited behind and slightly to the left of the centerline trajectory of the storm. Approximately 5% of the material is dispersed into the stratosphere and anvil section of the storm

Design methodology for the physical protection upgrade rule requirements for fixed sites. Technical report

International Nuclear Information System (INIS)

Evans, L.J. Jr.; Allen, T.

1980-06-01

This Design Methodology document aids the licensee in understanding how the fixed site requirements of the Physical Protection Upgrade Rule affect the design of physical protection systems for fuel processing plants, fuel manufacturing plants, or other fixed site special nuclear material operations involving possession or use of formula quantities of strategic special nuclear material. The document consists of three major elements: Logic Trees, Safeguards Jobs and Component Matrices, and Effectiveness Test Questionnaires. The work is based upon a previous study conducted by Sandia Laboratories for the Nuclear Regulatory Commission

Earthquakes and associated topics in relation to nuclear power plant siting

International Nuclear Information System (INIS)

1979-01-01

The main emphasis of the Guide is on the determination of the design basis ground motions for the nuclear power plant and on the determination of the potential for surface faulting at the site. Additionally, the Guide treats initiation of seismically induced flooding and the ground failure phenomena of subsidence and collapse. Volcanic activity is not dealt with except in connection with tsunamis. The methods and procedures discussed in this Guide should be taken as the basis for the design of seismically safe nuclear power plants and are primarily suited to areas of high and medium seismicity; for areas of low seismicity the Guide may not be applicable in its entirety and may need to be supplemented by other methods

Comments and responses on the Remedial Action Plan and site design for stabilization of the Inactive Uranium Mill Tailings Site, Grand Junction, Colorado. Revision 1

International Nuclear Information System (INIS)

1994-01-01

This report contains information concerning public comments and responses on the remedial action plan and site design for stabilization of the inactive uranium mill tailings site in Grand Junction, Colorado

An overview of the UK regulatory expectation for design basis accident analysis

International Nuclear Information System (INIS)

Trimble, Andy

2013-01-01

The UK Health and Safety Executive published its most recent regulatory expectations in the 2006 version of it's safety assessment principles (SAPs). This built on experience regulating the full range of facilities for which it is responsible. Thus the principles underpinning all regulatory safety case assessment are the same but the implementation differs depending on the application. This paper will describe the published design basis accident analysis (DBAA) logic in context with other technical aspects of the regulatory expectation for safety cases. It will further illustrate the DBAA methodology with practical examples from actual experience on reprocessing plant gained over the last 15 years or so. Among the examples will be the relevance of conventional safety fault initiators to nuclear safety assessment. It will further demonstrate the derivation of facility limits and conditions necessary for nuclear safety. (authors)

Review and comment on the US Department of Energy Site Characterization Plan Conceptual Design report

International Nuclear Information System (INIS)

1988-10-01

The ''Site Characterization Plan Conceptual Design Report,'' as prepared by the United States Department of Energy and its contractors, depicts the conceptual designs being proposed for the Yucca Mountain, Nevada High-Level Waste Repository. Much of the report is comprised of designs based on large assumptions that lead to inconclusiveness in terms of technical integrity and reasonableness. As a result, the likely success of the designs proposed within the report is highly questionable from a technical and even constructibility standpoint. Grave deficiencies have been found throughout the report ranging from the over-simplification of the geologic and geohydrologic setting to the incorrect determination of the repository horizon's host rock strength. All of these deficiencies found within the report relate directly to the probable outcome of the site characterization of Yucca Mountain and the possible design and construction of a high-level waste repository at the site. It does not appear that a concerted effort has been put forth to develop many of the concepts within the report to a reasonable level of engineering and scientific integrity that would be suitable to meet the needs of and provide the results to the Nation's High-Level Waste Program

Confidence assessment. Site-descriptive modelling SDM-Site Laxemar

International Nuclear Information System (INIS)

2009-06-01

The objective of this report is to assess the confidence that can be placed in the Laxemar site descriptive model, based on the information available at the conclusion of the surface-based investigations (SDM-Site Laxemar). In this exploration, an overriding question is whether remaining uncertainties are significant for repository engineering design or long-term safety assessment and could successfully be further reduced by more surface-based investigations or more usefully by explorations underground made during construction of the repository. Procedures for this assessment have been progressively refined during the course of the site descriptive modelling, and applied to all previous versions of the Forsmark and Laxemar site descriptive models. They include assessment of whether all relevant data have been considered and understood, identification of the main uncertainties and their causes, possible alternative models and their handling, and consistency between disciplines. The assessment then forms the basis for an overall confidence statement. The confidence in the Laxemar site descriptive model, based on the data available at the conclusion of the surface based site investigations, has been assessed by exploring: - Confidence in the site characterization data base, - remaining issues and their handling, - handling of alternatives, - consistency between disciplines and - main reasons for confidence and lack of confidence in the model. Generally, the site investigation database is of high quality, as assured by the quality procedures applied. It is judged that the Laxemar site descriptive model has an overall high level of confidence. Because of the relatively robust geological model that describes the site, the overall confidence in the Laxemar Site Descriptive model is judged to be high, even though details of the spatial variability remain unknown. The overall reason for this confidence is the wide spatial distribution of the data and the consistency between

Confidence assessment. Site-descriptive modelling SDM-Site Laxemar

Energy Technology Data Exchange (ETDEWEB)

2008-12-15

The objective of this report is to assess the confidence that can be placed in the Laxemar site descriptive model, based on the information available at the conclusion of the surface-based investigations (SDM-Site Laxemar). In this exploration, an overriding question is whether remaining uncertainties are significant for repository engineering design or long-term safety assessment and could successfully be further reduced by more surface-based investigations or more usefully by explorations underground made during construction of the repository. Procedures for this assessment have been progressively refined during the course of the site descriptive modelling, and applied to all previous versions of the Forsmark and Laxemar site descriptive models. They include assessment of whether all relevant data have been considered and understood, identification of the main uncertainties and their causes, possible alternative models and their handling, and consistency between disciplines. The assessment then forms the basis for an overall confidence statement. The confidence in the Laxemar site descriptive model, based on the data available at the conclusion of the surface based site investigations, has been assessed by exploring: - Confidence in the site characterization data base, - remaining issues and their handling, - handling of alternatives, - consistency between disciplines and - main reasons for confidence and lack of confidence in the model. Generally, the site investigation database is of high quality, as assured by the quality procedures applied. It is judged that the Laxemar site descriptive model has an overall high level of confidence. Because of the relatively robust geological model that describes the site, the overall confidence in the Laxemar Site Descriptive model is judged to be high, even though details of the spatial variability remain unknown. The overall reason for this confidence is the wide spatial distribution of the data and the consistency between

KBS-3H design description 2005

Energy Technology Data Exchange (ETDEWEB)

Autio, Jorma (Saanio and Riekkola Oy, Helsinki (Finland))

2008-05-15

SKB and Posiva are performing an RandD program over the period of 2002-2007 with the overall aim to develop the KBS-3H to an alternative to the KBS-3V concept for disposal of spent nuclear fuel. A feasibility study of the KBS-3H concept was carried out in 2002, followed by the setting up of basic design in 2003. Several problems related to the behavior of the design and scope of future research and development work were addressed. Therefore the design basis was developed further and two candidate designs were developed: 1) previous Basic Design (BD) was developed more robust and tolerable to inflows. Parallel to that a novel modified 2) DAWE design with Drainage, Air evacuation and Watering and was developed to function robustly at various inflow situations. The candidate designs presented in this report include several novel components, such as fixing rings and steel plugs which have been designed without support of applicable design guidelines, regulations or standards available. The design basis and performance of these components include uncertainties, which should be studied and verified. It is possible that a feasible site specific design can be based on using both alternatives

KBS-3H design description 2005

International Nuclear Information System (INIS)

Autio, Jorma

2008-05-01

SKB and Posiva are performing an RandD program over the period of 2002-2007 with the overall aim to develop the KBS-3H to an alternative to the KBS-3V concept for disposal of spent nuclear fuel. A feasibility study of the KBS-3H concept was carried out in 2002, followed by the setting up of basic design in 2003. Several problems related to the behavior of the design and scope of future research and development work were addressed. Therefore the design basis was developed further and two candidate designs were developed: 1) previous Basic Design (BD) was developed more robust and tolerable to inflows. Parallel to that a novel modified 2) DAWE design with Drainage, Air evacuation and Watering and was developed to function robustly at various inflow situations. The candidate designs presented in this report include several novel components, such as fixing rings and steel plugs which have been designed without support of applicable design guidelines, regulations or standards available. The design basis and performance of these components include uncertainties, which should be studied and verified. It is possible that a feasible site specific design can be based on using both alternatives

Soil structural analysis tools and properties for Hanford site waste tank evaluation

International Nuclear Information System (INIS)

Moore, C.J.; Holtz, R.D.; Wagenblast, G.R.; Weiner, E.D.; Marlow, R.S.

1995-09-01

As Hanford Site contractors address future structural demands on nuclear waste tanks, built as early as 1943, it is necessary to address their current safety margins and ensure safe margins are maintained. Although the current civil engineering practice guidelines for soil modeling are suitable as preliminary design tools, future demands potentially result in loads and modifications to the tanks that are outside the original design basis and current code based structural capabilities. For example, waste removal may include cutting a large hole in a tank. This report addresses both spring modeling of site soils and finite-element modeling of soils. Additionally seismic dynamic modeling of Hanford Site soils is also included. Of new and special interest is Section 2.2 that Professor Robert D. Holtz of the University of Washington wrote on plane strain soil testing versus triaxial testing with Hanford Site application to large buried waste tanks

Soil structural analysis tools and properties for Hanford site waste tank evaluation

Energy Technology Data Exchange (ETDEWEB)

Moore, C.J.; Holtz, R.D.; Wagenblast, G.R.; Weiner, E.D.; Marlow, R.S.

1995-09-01

As Hanford Site contractors address future structural demands on nuclear waste tanks, built as early as 1943, it is necessary to address their current safety margins and ensure safe margins are maintained. Although the current civil engineering practice guidelines for soil modeling are suitable as preliminary design tools, future demands potentially result in loads and modifications to the tanks that are outside the original design basis and current code based structural capabilities. For example, waste removal may include cutting a large hole in a tank. This report addresses both spring modeling of site soils and finite-element modeling of soils. Additionally seismic dynamic modeling of Hanford Site soils is also included. Of new and special interest is Section 2.2 that Professor Robert D. Holtz of the University of Washington wrote on plane strain soil testing versus triaxial testing with Hanford Site application to large buried waste tanks.

The Mixed Waste Management Facility. Design basis integrated operations plan (Title I design)

International Nuclear Information System (INIS)

1994-12-01

The Mixed Waste Management Facility (MWMF) will be a fully integrated, pilotscale facility for the demonstration of low-level, organic-matrix mixed waste treatment technologies. It will provide the bridge from bench-scale demonstrated technologies to the deployment and operation of full-scale treatment facilities. The MWMF is a key element in reducing the risk in deployment of effective and environmentally acceptable treatment processes for organic mixed-waste streams. The MWMF will provide the engineering test data, formal evaluation, and operating experience that will be required for these demonstration systems to become accepted by EPA and deployable in waste treatment facilities. The deployment will also demonstrate how to approach the permitting process with the regulatory agencies and how to operate and maintain the processes in a safe manner. This document describes, at a high level, how the facility will be designed and operated to achieve this mission. It frequently refers the reader to additional documentation that provides more detail in specific areas. Effective evaluation of a technology consists of a variety of informal and formal demonstrations involving individual technology systems or subsystems, integrated technology system combinations, or complete integrated treatment trains. Informal demonstrations will typically be used to gather general operating information and to establish a basis for development of formal demonstration plans. Formal demonstrations consist of a specific series of tests that are used to rigorously demonstrate the operation or performance of a specific system configuration

Methodology applied in Cuba for siting, designing, and building a radioactive waste repository under safety conditions

International Nuclear Information System (INIS)

Orbera, L.; Peralta, J.L.; Franklin, R.; Gil, R.; Chales, G.; Rodriguez, A.

1993-01-01

The work presents the methodology used in Cuba for siting, designing, and building a radioactive waste repository safely. This methodology covers both the technical and socio-economic factors, as well as those of design and construction so as to have a safe siting for this kind of repository under Cuba especial condition. Applying this methodology will results in a safe repository

Site characterization plan: Conceptual design report, Volume 2: Chapters 4-9: Nevada Nuclear Waste Storage Investigations Project

International Nuclear Information System (INIS)

MacDougall, H.R.; Scully, L.W.; Tillerson, J.R.

1987-09-01

This document presents a description of a prospective geologic repository for high-level radioactive waste to support the development of the Site Characterization Plan for the Yucca Mountain site. The target horizon for waste emplacement is a sloping bed of densely welded tuff more than 650 ft below the surface and typically more than 600 ft above the water table. The conceptual design described in this report is unique among repository designs in that it uses ramps in addition to shafts to gain access to the underground facility, the emplacement horizon is located above the water table, and it is possible that 300- to 400-ft-long horizontal waste emplacement boreholes will be used. This report summarizes the design bases (site and properties of the waste package), design and performance criteria, and the design analyses performed. The current status of meeting the preclosure performance objectives for licensing and of resolving the repository design and preclosure issues is presented. The repository design presented in this report will be expanded and refined during the advanced conceptual design, the license application design, and the final procurement and construction design phases. 147 refs., 145 figs., 83 tabs

Two crystal structures of dihydrofolate reductase-thymidylate synthase from Cryptosporidium hominis reveal protein–ligand interactions including a structural basis for observed antifolate resistance

Energy Technology Data Exchange (ETDEWEB)

Anderson, Amy C., E-mail: aca@dartmouth.edu [Dartmouth College, Department of Chemistry, Burke Laboratories, Hanover, NH 03755 (United States)

2005-03-01

An analysis of the protein–ligand interactions in two crystal structures of DHFR-TS from C. hominis reveals a possible structural basis for observed antifolate resistance in C. hominis DHFR. A comparison with the structure of human DHFR reveals residue substitutions that may be exploited for the design of species-selective inhibitors. Cryptosporidium hominis is a protozoan parasite that causes acute gastrointestinal illness. There are no effective therapies for cryptosporidiosis, highlighting the need for new drug-lead discovery. An analysis of the protein–ligand interactions in two crystal structures of dihydrofolate reductase-thymidylate synthase (DHFR-TS) from C. hominis, determined at 2.8 and 2.87 Å resolution, reveals that the interactions of residues Ile29, Thr58 and Cys113 in the active site of C. hominis DHFR provide a possible structural basis for the observed antifolate resistance. A comparison with the structure of human DHFR reveals active-site differences that may be exploited for the design of species-selective inhibitors.

Seismic site evaluation practice and seismic design guide for NPP in Continent of China

Energy Technology Data Exchange (ETDEWEB)

Yuxian, Hu [State Seismological Bureau, Beijing, BJ (China). Inst. of Geophysics

1997-03-01

Energy resources, seismicity, NPP and related regulations of the Continent of China are briefly introduced in the beginning and two codes related to the seismic design of NPP, one on siting and another on design, are discussed in some detail. The one on siting is an official code of the State Seismological Bureau, which specifies the seismic safety evaluation requirements of various kinds of structures, from the most critic and important structures such as NPP to ordinary buildings, and including also engineering works in big cities. The one on seismic design of NPP is a draft subjected to publication now, which will be an official national code. The first one is somewhat unique but the second one is quite similar to those in the world. (author)

Seismic site evaluation practice and seismic design guide for NPP in Continent of China

International Nuclear Information System (INIS)

Hu Yuxian

1997-01-01

Energy resources, seismicity, NPP and related regulations of the Continent of China are briefly introduced in the beginning and two codes related to the seismic design of NPP, one on siting and another on design, are discussed in some detail. The one on siting is an official code of the State Seismological Bureau, which specifies the seismic safety evaluation requirements of various kinds of structures, from the most critic and important structures such as NPP to ordinary buildings, and including also engineering works in big cities. The one on seismic design of NPP is a draft subjected to publication now, which will be an official national code. The first one is somewhat unique but the second one is quite similar to those in the world. (author)

Aerosol measurements at the Southern Great Plains Site: Design and surface installation

Energy Technology Data Exchange (ETDEWEB)

Leifer, R.; Knuth, R.H.; Guggenheim, S.F.; Albert, B. [Department of Energy, New York, NY (United States)

1996-04-01

To impropve the predictive capabilities of the Atmospheric Radiation Measurements (ARM) program radiation models, measurements of awserosol size distributions, condensation particle concentrations, aerosol scattering coefficients at a number of wavelenghts, and the aerosol absorption coefficients are needed at the Southern Great Plains (SGP) site. Alos, continuous measurements of ozone concnetrations are needed for model validation. The environmental Measuremenr Laboratory (EMK) has the responsibility to establish the surface aerosol measurements program at the SGP site. EML has designed a special sampling manifold.

Steric Hindrance as a Basis for Structure-Based Design of Selective Inhibitors of Protein-Tyrosine Phosphatases

DEFF Research Database (Denmark)

Iversen, L. F.; Andersen, H. S.; Møller, K. B.

2001-01-01

Utilizing structure-based design, we have previously demonstrated that it is possible to obtain selective inhibitors of protein-tyrosine phosphatase 1B (PTP1B). A basic nitrogen was introduced into a general PTP inhibitor to form a salt bridge to Asp48 in PTP1B and simultaneously cause repulsion...... in PTPs containing an asparagine in the equivalent position [Iversen, L. F., et al. (2000) J. Biol. Chem. 275, 10300−10307]. Further, we have recently demonstrated that Gly259 in PTP1B forms the bottom of a gateway that allows easy access to the active site for a broad range of substrates, while bulky...... in accessibility to the active site among various PTPs. We show that a general, low-molecular weight PTP inhibitor can be developed into a highly selective inhibitor for PTP1B and TC-PTP by introducing a substituent, which is designed to address the region around residues 258 and 259. Detailed enzyme kinetic...

Creep-Fatigue Life Design with Various Stress and Temperature Conditions on the Basis of Lethargy Coefficient

International Nuclear Information System (INIS)

Park, Jung Eun; Yang, Sung Mo; Han, Jae Hee; Yu, Hyo Sun

2011-01-01

High temperature and stress are encounted in power plants and vehicle engines. Therefore, determination of the creep-fatigue life of a material is necessary prior to fabricating equipment. In this study, life design was determined on the basis of the lethargy coefficient for different temperatures, stress and rupture times. SP-Creep test data was compared with computed data. The SP-Creep test was performed to obtain the rupture time for X20CrMoV121 steel. The integration life equation was considered for three cases with various load, temperature and load-temperature. First, the lethargy coefficient was calculated by using the obtained rupture stress and the rupture time that were determined by carrying out the SP-Creep test. Next, life was predicted on the basis of the temperature condition. Finally, it was observed that life decreases considerably due to the coupling effect that results when fatigue and creep occur simultaneously

Review of the ANSTO submission on the site geological investigations for the RRR at Lucas Heights

International Nuclear Information System (INIS)

2002-01-01

Preliminary Safety Analysis Report (PSAR) and ANSTO's Consolidated Seismic Report were reviewed by ARPANSA's Regulatory Branch, and by an IAEA consultant to ARPANSA. It included a recommendation by the IAEA Peer Review Team (June 2001) that an additional study be undertaken for the Lucas Heights site in accordance with IAEA Safety Guide 50-SG-S1. In particular this related to regional and local investigations to determine whether there exists a potential for surface faulting at the site area, to formulate the investigations that should be used to determine whether or not any faults in the area should be considered capable. The CEO of ARPANSA requested ANSTO to undertake such a task. The reviews that have been undertaken by experts from Geoscience Australia and the IAEA expert contracted by ARPANSA give confidence that the seismic design of the reactor structures, systems, and components important for safety should not be affected by the fault through the excavations for the foundations of the reactor building. The basis is the conclusions reached by these experts that the fault discovered is very old and not capable under either the USNRC Criteria or the IAEA Criteria. It is not now active, has not been active in the geologically recent past and is unlikely to re-activate within the foreseeable lifespan of the proposed facility. Consequently, this pre-existing fault presents no greater risk than the surrounding unfaulted material. Thus, the replacement reactor need not be designed for such displacement. The existence of this 'incapable' fault does not impact on the probabilistic seismic hazard assessment from which the seismic design basis for the reactor was derived. Therefore, the existing design response spectrum accepted in the Regulatory Assessment Report as the basis for seismic design of structures, systems and components, and approved in the Construction Licence, has not changed and remains adequately conservative with regard to vibratory motion at the site

Breckinridge Project, initial effort. Report XI, Volume V. Critical review of the design basis. [Critical review

Energy Technology Data Exchange (ETDEWEB)

None

1982-01-01

Report XI, Technical Audit, is a compendium of research material used during the Initial Effort in making engineering comparisons and decisions. Volumes 4 and 5 of Report XI present those studies which provide a Critical Review of the Design Basis. The Critical Review Report, prepared by Intercontinental Econergy Associates, Inc., summarizes findings from an extensive review of the data base for the H-Coal process design. Volume 4 presents this review and assessment, and includes supporting material; specifically, Design Data Tabulation (Appendix A), Process Flow Sheets (Appendix B), and References (Appendix C). Volume 5 is a continuation of the references of Appendix C. Studies of a proprietary nature are noted and referenced, but are not included in these volumes. They are included in the Limited Access versions of these reports and may be reviewed by properly cleared personnel in the offices of Ashland Synthetic Fuels, Inc.

Study of site layout in the Rokkasho site

International Nuclear Information System (INIS)

Sato, Kazuyoshi; Tamura, Kousaku; Yagenji, Akira; Sekiya, Shigeki; Takahashi, Hideo; Neyatani, Yuzuru; Uehara, Masaharu; Motohashi, Keiichi; Hashimoto, Masayoshi; Ogino, Shunji; Nagamatsu, Nobuhide

2006-03-01

The Final Design Report (FDR) of the International Thermonuclear Experimental Reactor (ITER) was published on July 2001 as a summary of the Engineering Design Activity (EDA). After the EDA, site dependent design has been investigated for the invitation of ITER toward Rokkasho Site (Iyasakadai area) in Aomori prefecture. This report describes the results of site layout of major buildings and structures of ITER in the Rokkasho-Site. The data of the ground near the site and the results of site dependent design in Japan were applied to this study. Through this study, the most appropriate site layout has been constructed with satisfaction of following conditions. (1) Bedrock level at the tokamak complex building is relatively high and it can be reduced the cost of excavation and foundation work. (2) Total amount of excavation soil for site preparation is minimized and the flexibility of the layout is ensured with flat ground level. (3) Accessibility of human and equipments, reduction of noise and vibration to the environment can be obtained. Total length of ducts and piping between buildings in site is minimized. (author)

Remedial Action Plan and site design for stabilization of the inactive uranium mill tailings site at Falls City, Texas

International Nuclear Information System (INIS)

Chernoff, A.R.; Lacker, D.K.

1992-09-01

The uranium processing site near Falls City, Texas, was one of 24 inactive uranium mill sites designated to be remediated by the US Department of Energy (DOE) under Title I of the Uranium Mill Tailings Radiation Control Act of 1978 (UMTRCA). The UMTRCA requires that the US Nuclear Regulatory Commission (NRC) concur with the DOE's remedial action plan (RAP) and certify that the remedial action conducted at the site complies with the standards promulgated by the US Environmental Protection Agency (EPA). The RAP, which includes this summary remedial action selection report (RAS), serves a two-fold purpose. First, it describes the activities proposed by the DOE to accomplish long-term stabilization and control of the residual radioactive materials at the inactive uranium processing site near Falls City, Texas. Second, this document and the remainder of the RAP, upon concurrence and execution by the DOE, the State of Texas, and the NRC, becomes Appendix B of the Cooperative Agreement between the DOE and the State of Texas

Seismic design of circular-section concrete-lined underground openings: Preclosure performance considerations for the Yucca Mountain Site

International Nuclear Information System (INIS)

Richardson, A.M.; Blejwas, T.E.

1992-01-01

Yucca Mountain, the potential site of a repository for high-level radioactive waste, is situated in a region of natural and man-made seismicity. Underground openings excavated at this site must be designed for worker safety in the seismic environment anticipated for the preclosure period. This includes accesses developed for site characterization regardless of the ultimate outcome of the repository siting process. Experience with both civil and mining structures has shown that underground openings are much more resistant to seismic effects than surface structures, and that even severe dynamic strains can usually be accommodated with proper design. This paper discusses the design and performance of lined openings in the seismic environment of the potential site. The types and ranges of possible ground motions (seismic loads) are briefly discussed. Relevant historical records of underground opening performance during seismic loading are reviewed. Simple analytical methods of predicting liner performance under combined in situ, thermal, and seismic loading are presented, and results of calculations are discussed in the context of realistic performance requirements for concrete-lined openings for the preclosure period. Design features that will enhance liner stability and mitigate the impact of the potential seismic load are reviewed. The paper is limited to preclosure performance concerns involving worker safety because present decommissioning plans specify maintaining the option for liner removal at seal locations, thus decoupling liner design from repository postclosure performance issues

Natural phenomena analyses, Hanford Site, Washington

International Nuclear Information System (INIS)

Tallman, A.M.

1989-01-01

Probabilistic seismic hazard studies completed for the Washington Public Power Supply System's Nuclear Plant 2 and for the US Department of Energy's N Reactor sites, both on the Hanford Site, suggested that the Lawrence Livermore National Laboratory seismic exposure estimates were lower than appropriate, especially for sites near potential seismic sources. A probabilistic seismic hazard assessment was completed for those areas that contain process and/or waste management facilities. the lower bound magnitude of 5.0 is used in the hazard analysis and the characteristics of small-magnitude earthquakes relatively common to the Hanford Site are addressed. The recommended ground motion for high-hazard facilities is somewhat higher than the Lawrence Livermore National Laboratory model and the ground motion from small-magnitude earthquakes is addressed separately from the moderate- to large-magnitude earthquake ground motion. The severe wind and tornado hazards determined for the Hanford Siste are in agreement with work completed independently using 43 years of site data. The low-probability, high-hazard, design-basis flood at the Hanford Site is dominated by dam failure on the Columbia River. Further evaluation of the mechanisms and probabilities of such flooding is in progress. The Hanford Site is downwind from several active Cascade volcanoes. Geologic and historical data are used to estimate the ashfall hazard

Site-specific probabilistic seismic hazard analyses for the Idaho National Engineering Laboratory. Volume 1: Final report

International Nuclear Information System (INIS)

1996-05-01

This report describes and summarizes a probabilistic evaluation of ground motions for the Idaho National Engineering Laboratory (INEL). The purpose of this evaluation is to provide a basis for updating the seismic design criteria for the INEL. In this study, site-specific seismic hazard curves were developed for seven facility sites as prescribed by DOE Standards 1022-93 and 1023-96. These sites include the: Advanced Test Reactor (ATR); Argonne National Laboratory West (ANL); Idaho Chemical Processing Plant (ICPP or CPP); Power Burst Facility (PBF); Radioactive Waste Management Complex (RWMC); Naval Reactor Facility (NRF); and Test Area North (TAN). The results, probabilistic peak ground accelerations and uniform hazard spectra, contained in this report are not to be used for purposes of seismic design at INEL. A subsequent study will be performed to translate the results of this probabilistic seismic hazard analysis to site-specific seismic design values for the INEL as per the requirements of DOE Standard 1020-94. These site-specific seismic design values will be incorporated into the INEL Architectural and Engineering Standards

Design basis for the NRC Operations Center

Energy Technology Data Exchange (ETDEWEB)

Lindell, M.K.; Wise, J.A.; Griffin, B.N.; Desrosiers, A.E.; Meitzler, W.D.

1983-05-01

This report documents the development of a design for a new NRC Operations Center (NRCOC). The project was conducted in two phases: organizational analysis and facility design. In order to control the amount of traffic, congestion and noise within the facility, it is recommended that information flow in the new NRCOC be accomplished by means of an electronic Status Information Management System. Functional requirements and a conceptual design for this system are described. An idealized architectural design and a detailed design program are presented that provide the appropriate amount of space for operations, equipment and circulation within team areas. The overall layout provides controlled access to the facility and, through the use of a zoning concept, provides each team within the NRCOC the appropriate balance of ready access and privacy determined from the organizational analyses conducted during the initial phase of the project.

Design basis for the NRC Operations Center

International Nuclear Information System (INIS)

Lindell, M.K.; Wise, J.A.; Griffin, B.N.; Desrosiers, A.E.; Meitzler, W.D.

1983-05-01

This report documents the development of a design for a new NRC Operations Center (NRCOC). The project was conducted in two phases: organizational analysis and facility design. In order to control the amount of traffic, congestion and noise within the facility, it is recommended that information flow in the new NRCOC be accomplished by means of an electronic Status Information Management System. Functional requirements and a conceptual design for this system are described. An idealized architectural design and a detailed design program are presented that provide the appropriate amount of space for operations, equipment and circulation within team areas. The overall layout provides controlled access to the facility and, through the use of a zoning concept, provides each team within the NRCOC the appropriate balance of ready access and privacy determined from the organizational analyses conducted during the initial phase of the project

Basis of valve operator selection for SMART

International Nuclear Information System (INIS)

Kang, H. S.; Lee, D. J.; See, J. K.; Park, C. K.; Choi, B. S.

2000-05-01

SMART, an integral reactor with enhanced safety and operability, is under development for use of the nuclear energy. The valve operator of SMART system were selected through the data survey and technical review of potential valve fabrication vendors, and it will provide the establishment and optimization of the basic system design of SMART. In order to establish and optimize the basic system design of SMART, the basis of selection for the valve operator type were provided based on the basic design requirements. The basis of valve operator selection for SMART will be used as a basic technical data for the SMART basic and detail design and a fundamental material for the new reactor development in the future

Basis of valve operator selection for SMART

Energy Technology Data Exchange (ETDEWEB)

Kang, H. S.; Lee, D. J.; See, J. K.; Park, C. K.; Choi, B. S

2000-05-01

SMART, an integral reactor with enhanced safety and operability, is under development for use of the nuclear energy. The valve operator of SMART system were selected through the data survey and technical review of potential valve fabrication vendors, and it will provide the establishment and optimization of the basic system design of SMART. In order to establish and optimize the basic system design of SMART, the basis of selection for the valve operator type were provided based on the basic design requirements. The basis of valve operator selection for SMART will be used as a basic technical data for the SMART basic and detail design and a fundamental material for the new reactor development in the future.

GEOTECHNICAL DESIGN OF SOLID WASTE LANDFILL SITES

Directory of Open Access Journals (Sweden)

Suat AKBULUT

2003-02-01

Full Text Available Solid waste landfills are important engineering structures for protection of wastes, decrease of environmental pollution, and especially prevention of soil and water pollution. Solid wastes should conveniently be maintained in landfill areas to control environmental pollution caused by waste disposals. Until the middle of this century clay liners were used for maintenance of waste disposal, but it was observed that these liner systems were insufficient. Today thinner and less permeable liner systems are constructed by using synthetic materials. In this study, by evaluating the waste landfills, site assessment of landfills and construction of natural and synthetic liner systems were summarized respectively, and especially the design properties of these systems were examined intensively. Also, leachate collection and removal facilities, landfill gas collection unites, and final cover unites were evaluated in a detailed way.

Risk assessment data bank design at the Savannah River Site

International Nuclear Information System (INIS)

Townsend, C.S.; Johnson, K.B.

1992-01-01

The Savannah River Site has designed and implemented a database system containing a series of compilations of incidents used primarily for risk assessment. Four databases have been designed and implemented using advanced database management system computer software. These databases exist for reprocessing, fuel fabrication, waste management, and the Savannah River Technology Center. They are combined into one system caged the Risk Assessment Methodology (RAM) Fault Tree Data Banks. This paper will discuss the logical design of the data, the menus, and the operating platform. Built-in updating features, such as batch and on-line data entry; data validation methods; automatic update features; and expert system programs, will also be discussed. User functions, such as on-line search/view/report and statistical functions, will be presented. Security features and backup and recovery methods will also be covered

Basis of the tubesheet heat exchanger design rules used in the French pressure vessel code

International Nuclear Information System (INIS)

Osweiller, F.

1990-01-01

For about 40 years most tubesheet heat exchangers have been designed according to the standards of TEMA. Partly due to their simplicity, these rules do not assure a safe heat-exchangers design in all cases. This is the main reason why new tubesheet design rules were developed in 1981 in France for the French pressure vessel code CODAP. For fixed tubesheet heat exchangers the new rules account for the elastic rotational restraint of the shell and channel at the outer edge of the tubesheet. For floating-head and U- tube exchangers an approach was selected with some modifications. In both cases the tubesheet is replaced by an equivalent solid plate with adequate effective elastic constants, and the tube bundle is simulated by an elastic foundation. The elastic restraint at the edge of the tubesheet due the shell and channel is accounted for in different ways in the two types of heat exchangers. The purpose of the paper is to present the main basis of these rules and to compare them to TEMA rules

Alternate technologies for MRS design

International Nuclear Information System (INIS)

Smith, R.I.; Triplett, M.B.; Ashton, W.B.; Kelly, W.S.

1984-01-01

This paper describes the process conducted by the Pacific Northwest Laboratory (PNL) to evaluate candidate MRS concepts and to recommend the two most preferred concepts. The eight concepts studied are: metal cask (stationary and transportable); concrete cask (silo); concrete cask-in-trench; field drywell; tunnel drywell; open cycle vault; closed cycle vault; and tunnel rack vault. To achieve a more equitable comparison of the concepts, conceptual design analyses were performed for the candidate concepts using a common set of specified design requirements selected with consideration of the MRS mission. Using this normalized conceptual design information, the MRS concepts were evaluated and compared on the basis of their relative performance on seven criteria: flexibility, concept maturity, cost, environmental impacts, safety and licensing, socioeconomic impacts, and siting requirements. These seven criteria were judged to form a reasonable and complete basis for the evaluation of MRS concepts' ability to satisfy the MRS mission requirements. 5 references, 8 figures, 1 table

Design of turning hydraulic engines for manipulators of mobile machines on the basis of multicriterial optimization

Directory of Open Access Journals (Sweden)

Lagerev I.A.

2016-12-01

Full Text Available In this paper the mathematical models of the main types of turning hydraulic engines, which at the present time widely used in the construction of handling systems of domestic and foreign mobile transport-technological machines wide functionality. They allow to take into consideration the most significant from the viewpoint of ensuring high technical-economic indicators of hydraulic efficiency criteria – minimum mass (weight, their volume and losses of power. On the basis of these mathematical models the problem of multicriterial constrained optimization of the constructive sizes of turning hydraulic engines are subject to complex constructive, strength and deformation limits. It allows you to de-velop the hydraulic engines in an optimized design which is required for the purpose of designing a comprehensive measure takes into account efficiency criteria. The multicriterial optimization problem is universal in nature, so when designing a turning hydraulic engines allows for one-, two - and three-criteria optimization without making any changes in the solution algorithm. This is a significant advantage for the development of universal software for the automation of design of mobile transport-technological machines.

Development of uniform hazard response spectra for Tarapur, Trombay and Kakrapar sites

International Nuclear Information System (INIS)

Ghosh, A.K.; Rao, K.S.; Kushwaha, H.S.

2003-08-01

Traditionally, the seismic design basis ground motion has been specified by normalised response spectral shapes and peak ground acceleration (PGA). The mean recurrence interval (MRI) used to be computed for PGA only. The present work develops uniform hazard response spectra i.e. spectra having the same MRI at all frequencies for Tarapur, Trombay and Kakrapar sites. These results determine the seismic hazard at the given site and the associated uncertainties. Typical examples are given to show the sensitivity of the results to changes in various parameters. (author)

Application of the MELCOR code to design basis PWR large dry containment analysis.

Energy Technology Data Exchange (ETDEWEB)

Phillips, Jesse; Notafrancesco, Allen (USNRC, Office of Nuclear Regulatory Research, Rockville, MD); Tills, Jack Lee (Jack Tills & Associates, Inc., Sandia Park, NM)

2009-05-01

The MELCOR computer code has been developed by Sandia National Laboratories under USNRC sponsorship to provide capability for independently auditing analyses submitted by reactor manufactures and utilities. MELCOR is a fully integrated code (encompassing the reactor coolant system and the containment building) that models the progression of postulated accidents in light water reactor power plants. To assess the adequacy of containment thermal-hydraulic modeling incorporated in the MELCOR code for application to PWR large dry containments, several selected demonstration designs were analyzed. This report documents MELCOR code demonstration calculations performed for postulated design basis accident (DBA) analysis (LOCA and MSLB) inside containment, which are compared to other code results. The key processes when analyzing the containment loads inside PWR large dry containments are (1) expansion and transport of high mass/energy releases, (2) heat and mass transfer to structural passive heat sinks, and (3) containment pressure reduction due to engineered safety features. A code-to-code benchmarking for DBA events showed that MELCOR predictions of maximum containment loads were equivalent to similar predictions using a qualified containment code known as CONTAIN. This equivalency was found to apply for both single- and multi-cell containment models.

Volcanic Hazards in Site Evaluation for Nuclear Installations

Energy Technology Data Exchange (ETDEWEB)

NONE

2012-10-15

This publication provides comprehensive and updated guidance for site evaluation in relation to volcanic hazards. It includes recommendations on assessing the volcanic hazards at a nuclear installation site, in order to identify and characterize, in a comprehensive manner, all potentially hazardous phenomena that may be associated with future volcanic events. It describes how some of these volcanic phenomena may affect the acceptability of the selected site, resulting in exclusion of a site or determining the corresponding design basis parameters for the installation. This Safety Guide is applicable to both existing and new sites, and a graded approach is recommended to cater for all types of nuclear installations. Contents: 1. Introduction; 2. Overview of volcanic hazard assessment; 3. General recommendations; 4. Necessary information and investigations (database); 5. Screening of volcanic hazards; 6. Site specific volcanic hazard assessment; 7. Nuclear installations other than nuclear power plants; 8. Monitoring and preparation for response; 9. Management system for volcanic hazard assessment; Annex I: Volcanic hazard scenarios; Annex II: Worldwide sources of information.

Catalysis by Design: Well-Defined Single-Site Heterogeneous Catalysts

KAUST Repository

Pelletier, Jeremie

2016-03-09

ConspectusHeterogeneous catalysis, a field important industrially and scientifically, is increasingly seeking and refining strategies to render itself more predictable. The main issue is due to the nature and the population of catalytically active sites. Their number is generally low to very low, their "acid strengths" or " redox properties" are not homogeneous, and the material may display related yet inactive sites on the same material. In many heterogeneous catalysts, the discovery of a structure-activity reationship is at best challenging. One possible solution is to generate single-site catalysts in which most, if not all, of the sites are structurally identical. Within this context and using the right tools, the catalyst structure can be designed and well-defined, to reach a molecular understanding. It is then feasible to understand the structure-activity relationship and to develop predictable heterogeneous catalysis. Single-site well-defined heterogeneous catalysts can be prepared using concepts and tools of surface organometallic chemistry (SOMC). This approach operates by reacting organometallic compounds with surfaces of highly divided oxides (or of metal nanoparticles). This strategy has a solid track record to reveal structure-activity relationship to the extent that it is becoming now quite predictable. Almost all elements of the periodical table have been grafted on surfaces of oxides (from simple oxides such as silica or alumina to more sophisticated materials regarding composition or porosity).Considering catalytic hydrocarbon transformations, heterogeneous catalysis outcome may now be predicted based on existing mechanistic proposals and the rules of molecular chemistry (organometallic, organic) associated with some concepts of surface sciences. A thorough characterization of the grafted metal centers must be carried out using tools spanning from molecular organometallic or surface chemistry. By selection of the metal, its ligand set, and the

Evaluation of E-Learning Web Sites Using Fuzzy Axiomatic Design Based Approach

Directory of Open Access Journals (Sweden)

2010-04-01

Full Text Available High quality web site has been generally recognized as a critical enabler to conduct online business. Numerous studies exist in the literature to measure the business performance in relation to web site quality. In this paper, an axiomatic design based approach for fuzzy group decision making is adopted to evaluate the quality of e-learning web sites. Another multi-criteria decision making technique, namely fuzzy TOPSIS, is applied in order to validate the outcome. The methodology proposed in this paper has the advantage of incorporating requirements and enabling reductions in the problem size, as compared to fuzzy TOPSIS. A case study focusing on Turkish e-learning websites is presented, and based on the empirical findings, managerial implications and recommendations for future research are offered.

Impact of liability and site closure and long-term care issues on future siting efforts

International Nuclear Information System (INIS)

Carlin, E.M.; Hana, S.L.A.

1988-01-01

Washington's research in the area of financial responsibility for liability and cleanup for radioactive materials licensees and low-level radioactive waste permittees is offered to assist unsited states and regions in their planning and development of new low-level waste disposal capacity. The state considered the need for third party bodily injury and property damage financial responsibility and determined that the USDOT requirements comprehensively cover transport of wastes. In regard to licensees' facilities, it is the state's opinion that an adequate technical basis for third party requirements has not yet been developed. Also considered was the need for financial assurance for cleanup, which is covered for transportation, but generally not available for facilities. Three options are examined to provide such coverage, and the economic impact on licensees assessed. Finally, the current low-level waste disposal site operator's insurance coverage is analyzed and deficiencies are identified. Washington is also conducting research into site closure and perpetual care and maintenance requirements for the commercial low-level radioactive waste disposal facility located on the Hanford reservation near Richland, Washington. This research includes a site assessment and identification and formulation of site-specific design elements for closure and long-term care

Beyond Design Basis Severe Accident Management as an Element of DiD Concept Strengthening

Energy Technology Data Exchange (ETDEWEB)

Kuznetsov, M., E-mail: kuznetsov_mv@vosafety.ru [FSUE VO “Safety”, Moscow (Russian Federation)

2014-10-15

The 4{sup th} Level of DiD is ensured by management of beyond design basis accidents which is achieved by implementation of the Beyond Design Basis Accidents Management Guidance (BDBAMG) and, if necessary, by additional technical devices and organizational measures at NPP Unit. BDBAMG is located between Levels 3 and 5 in DiD and is related to them. It is connected with Level 3 by means of conditions generated at this Level and according to which BDBAM should be initiated (Level 4). It is associated with Level 5 by conditions which necessitate implementation of Emergency planning. Both types of conditions should be identified in BDBAMG. BDBAs including the phase of severe damage of fuel and protective barriers (severe accidents) in accordance with Russian regulatory framework are a subset of all BDBAs set. In this connection, such accident scenarios meet the representativeness criterion for further analysis and development of Guidance for their management. BDBAMG availability, as it provides robustness of DiD as a whole, is an obligatory condition for obtaining a NPP operational license. In the process of BDBAMG development and implementation a feedback with technical and organizational measures, comprising Level 1 and, to a less extent, Level 2, comes up. BDBAMG verification is an important final stage of its development. Addressing severe accidents, it is a challenging issue for a full scope simulator and may require its software modernization to make it responsive to severe accident phenomena. The existing BDBAMGs should be updated due to NPP Unit modernizations and in conjunction with the latest knowledge on severe accident phenomenology and lessons learnt from known events (e.g. NPP Fukushima). Thus, improvements incorporated in BDBAMG, enhance the strength of DiD. (author)

Feasibility studies for pump and treat technology at leaking underground storage tank sites in Michigan

International Nuclear Information System (INIS)

O'Brien, J.M.; Pekas, B.S.

1993-01-01

Releases from underground storage tanks have resulted in impacts to groundwater at thousands of sites across the US. Investigations of these sites were initiated on a national basis with the implementation of federal laws that became effective December 22, 1989 (40 CFR 280). Completion of these investigations has led to a wave of design and installation of pump and treat aquifer restoration systems where impacts to groundwater have been confirmed. The purpose of this paper is to provide managers with a demonstration of some of the techniques that can be used by the consulting industry in evaluating the feasibility of pump and treat systems. With knowledge of these tools, managers can better evaluate proposals for system design and their cost effectiveness. To evaluate the effectiveness of typical pump and treat systems for leaking underground storage tank (LUST) sites in Michigan, ten sites where remedial design had been completed were randomly chosen for review. From these ten, two sites were selected that represented the greatest contrast in the types of site conditions encountered. A release of gasoline at Site 1 resulted in contamination of groundwater and soil with benzene, toluene, ethylbenzene, and xylenes

Design and implementation of a computer based site operations log for the ARM Program

International Nuclear Information System (INIS)

Tichler, J.L.; Bernstein, H.J.; Bobrowski, S.F.; Melton, R.B.; Campbell, A.P.; Edwards, D.M.; Kanciruk, P.; Singley, P.T.

1992-01-01

The Atmospheric Radiation Measurement (ARM) Program is a Department of Energy (DOE) research effort to reduce the uncertainties found in general circulation and other models due to the effects of clouds and solar radiation. ARM will provide an experimental testbed for the study of important atmospheric effects, particularly cloud and radiative processes, and testing of parameterizations of the processes for use in atmospheric models. The design of the testbed known as the Clouds and Radiation Testbed (CART), calls for five, long-term field data collection sites. The first site, located in the Southern Great Plains (SGP) in Lamont, OK began operation in the spring of 1992. The CART Data Environment (CDE) is the element of the testbed which acquires the basic observations from the instruments and processes them to meet the ARM requirements. A formal design was used to develop a description of the logical requirements for the CDE. This paper discusses the design and prototype implementation of a part of the CDE known as the site operations log, which records metadata defining the environment within which the data produced by the instruments is collected

Recharge at the Hanford Site: Status report

International Nuclear Information System (INIS)

Gee, G.W.

1987-11-01

A variety of field programs designed to evaluate recharge and other water balance components including precipitation, infiltration, evaporation, and water storage changes, have been carried out at the Hanford Site since 1970. Data from these programs have indicated that a wide range of recharge rates can occur depending upon specific site conditions. Present evidence suggests that minimum recharge occurs where soils are fine-textured and surfaces are vegetated with deep-rooted plants. Maximum recharge occurs where coarse soils or gravels exist at the surface and soils are kept bare. Recharge can occur in areas where shallow-rooted plants dominate the surface, particularly where soils are coarse-textured. Recharge estimates have been made for the site using simulation models. A US Geological Survey model that attempts to account for climate variability, soil storage parameters, and plant factors has calculated recharge values ranging from near zero to an average of about 1 cm/yr for the Hanford Site. UNSAT-H, a deterministic model developed for the site, appears to be the best code available for estimating recharge on a site-specific basis. Appendix I contains precipitation data from January 1979 to June 1987. 42 refs., 11 figs., 11 tabs

The EPR-a comprehensive design concept against external events

International Nuclear Information System (INIS)

Waas, U.; Stoll, U.

2005-01-01

The main objective of design provisions against external hazards is to ensure that the safety functions required to bring the plant to safe shutdown are not inadmissible affected. In the design of the EPR particular attention was paid to the following external hazards: Earthquake, Airplane crash, Explosion pressure wave. The design concept for these events is discussed below. The standard EPR covers a large range of possible site conditions, the design earthquake enveloping safe shutdown earthquakes (SSE) to be expected for potential sites. The Basic Design was developed for the seismic loads given in the European Utility Requirements with a horizontal free field Peak Ground Acceleration (PGA) for all site conditions of 0.25 g. The seismic protection is based on a deterministic design approach, with the intention of ensuring the safety functions in case of SSE. The loads for the design basis airplane crash and - if required - for the design extension airplane crash are defined depending on site specific requirements. For the design basis airplane crash as defined in Finland the safety goals are fulfilled for postulated single failure and preventive maintenance as well as for specific unlikely scenarios with local impacts where one redundant train is assumed to be lost. For the design extension airplane crash no single failure and preventive maintenance are assumed. Reactor building (RB), fuel building (FB), safeguard building (SB) 2 and 3 are protected by design against airplane crash. The common base mat of the RB, FB and SBs ensures global stability. To avoid penetration the wall thickness of the outer building structures of RB, FB and SB2/3 is set at 1.80 m (result of an optimization process). To rule out major induced vibrations due to airplane crash the inner building structures are decoupled from the outer walls. The SB 1 and SB4, the main steam and feedwater valve compartments, the diesel buildings and the service water pump buildings are protected against

Flood risk analysis of Cold Creek near the Hanford Site

International Nuclear Information System (INIS)

Skaggs, R.L.; Walters, W.H.

1981-01-01

The Pacific Northwest Laboratory has analyzed the flood potential at the reference repository location located on the Hanford Site near Richland, Washington. It is emphasized that this work is not intended as a basis for engineering design, but rather as an initial, regional appraisal of whether detailed engineering design analysis will be required. In order to achieve the detail required for engineering design specifications, the study results should be refined using more detailed channel geometry data, and the topography of the western portion of the reference repository location should be mapped using a contour interval of not less than 2 ft. 19 refs., 15 figs., 5 tabs

The design on high slope stabilization in waste rock sites of uranium mines

International Nuclear Information System (INIS)

Liu Taoan; Zhou Xinghuo; Liu Jia

2005-01-01

Design methods, reinforcement measures, and flood control measures concerning high slope stabilization in harnessing waste rock site are described in brief according to some examples of two uranium mines in Hunan province. (authors)

Design basis and design features of WWER-440 model 213 nuclear power plants. Reference plant: Bohunice V2 (Slovakia)

International Nuclear Information System (INIS)

1994-05-01

The prime objective of the IAEA Technical Co-operation Project on Evaluation of Safety Aspects of WWER-440 model 213 NPPs is to co-ordinate and to integrate assistance to national organizations in studying selected aspects of safety for the same type of reactors. Consequently, the study integrated the results generated by national activities carried out in the Czech Republic, Hungary, Slovakia and Ukraine and co-ordinated through the IAEA. Valuable assistance in carrying out the tasks was also provided by Bulgaria and Poland. A set of publications is being prepared to present the results of the project. The publications are intended to facilitate the review and utilization of the results of the project. They are also providing assistance in further refinement and/or extension of plant specific safety evaluation of model 213 NPPs. This Technical Document addressing the design basis and safety related design features of WWER-440 model 213 plants is the first of the series to be published. It is hoped that this document will be useful to anyone working in the field of WWER safety, and in particular to experts planning, executing or reviewing studies related to the subject. Refs, 36 figs, tabs

Study of integrated optimization design of wind farm in complex terrain

DEFF Research Database (Denmark)

Xu, Chang; Chen, Dandan; Han, Xingxing

2017-01-01

wind farm design in complex terrain and setting up integrated optimization mathematical model for micro-site selection, power lines and road maintenance design etc.. Based on the existing 1-year wind measurement data in the wind farm area, the genetic algorithm was used to optimize the micro......-site selection. On the basis of location optimization of wind turbine, the optimization algorithms such as single-source shortest path algorithm and minimum spanning tree algorithm were used to optimize electric lines and maintenance roads. The practice shows that the research results can provide important...

KBS-3H design description 2005

International Nuclear Information System (INIS)

Autio, J.

2007-03-01

SKB and Posiva are performing an R and D program over the period of 2002-2007 with the overall aim to develop the KBS-3H to an alternative to the KBS-3V concept for disposal of spent nuclear fuel. A feasibility study of the KBS-3H concept was carried out in 2002, followed by the setting up of basic design in 2003. Several problems related to the behavior of the design and scope of future research and development work were addressed. Therefore the design basis was developed further and two candidate designs were developed: (1) previous Basic Design (BD) was developed more robust and tolerable to inflows. Parallel to that a novel modified (2) DAWE design with Drainage, Air evacuation and Watering and was developed to function robustly at various inflow situations. The candidate designs presented in this report include several novel components, such as fixing rings and steel plugs which have been designed without support of applicable design guidelines, regulations or standards available. The design basis and performance of these components include uncertainties, which should be studied and verified. It is possible that a feasible site specific design can be based on using both alternatives. (orig.)

Custom-Designed Molecular Scissors for Site-Specific Manipulation of the Plant and Mammalian Genomes

Science.gov (United States)

Kandavelou, Karthikeyan; Chandrasegaran, Srinivasan

Zinc finger nucleases (ZFNs) are custom-designed molecular scissors, engineered to cut at specific DNA sequences. ZFNs combine the zinc finger proteins (ZFPs) with the nonspecific cleavage domain of the FokI restriction enzyme. The DNA-binding specificity of ZFNs can be easily altered experimentally. This easy manipulation of the ZFN recognition specificity enables one to deliver a targeted double-strand break (DSB) to a genome. The targeted DSB stimulates local gene targeting by several orders of magnitude at that specific cut site via homologous recombination (HR). Thus, ZFNs have become an important experimental tool to make site-specific and permanent alterations to genomes of not only plants and mammals but also of many other organisms. Engineering of custom ZFNs involves many steps. The first step is to identify a ZFN site at or near the chosen chromosomal target within the genome to which ZFNs will bind and cut. The second step is to design and/or select various ZFP combinations that will bind to the chosen target site with high specificity and affinity. The DNA coding sequence for the designed ZFPs are then assembled by polymerase chain reaction (PCR) using oligonucleotides. The third step is to fuse the ZFP constructs to the FokI cleavage domain. The ZFNs are then expressed as proteins by using the rabbit reticulocyte in vitro transcription/translation system and the protein products assayed for their DNA cleavage specificity.

Nuclear installations sites safety

International Nuclear Information System (INIS)

Barber, P.; Candes, P.; Duclos, P.; Doumenc, A.; Faure, J.; Hugon, J.; Mohammadioun, B.

1988-11-01

This report is divided into ten parts bearing: 1 Safety analysis procedures for Basis Nuclear Installations sites (BNI) in France 2 Site safety for BNI in France 3 Industrial and transport activities risks for BNI in France 4 Demographic characteristics near BNI sites in France 5 Meteorologic characteristics of BNI sites in France 6 Geological aspects near the BNI sites in France 7 Seismic studies for BNI sites in France 8 Hydrogeological aspects near BNI sites in France 9 Hydrological aspects near BNI sites in France 10 Ecological and radioecological studies of BNI sites in France [fr

Report of JLC site study group

CERN Document Server

Hasegawa, T; Yamashita, S

2003-01-01

This study group selected some good sites for construction of JLC (Electron-Positron Linear Collider) on the basis of investigation of data and field survey. The aims, activity, use of underground of private land, conditions of site, selection of site at present and future, summary and proposal are reported. 9 sites (Hidaka, Kitakami, Murayama, Abukuma, Kitaibaraki, Aichi and Gifu, Takamatsu, Hiroshima and Seburi range) are selected for the construction on the basis of firm ground and 4 sites (Okinawa, Harima, Tsukuba and Mutsuogawara) for development and researches. 9 sites area consists of plutonic rock or old strata of Paleozoic era. Many problems in each site are reported. There are three following proposals; 1) the self-governing communities of the sites have to understand JLC and start to construct it by information, 2) a site evaluation committee consists of specialist of civil engineering, building, social and natural environment and disaster prevention and 3) the vibration test should be carried out ...

Consideration of the band-gap tunability of BaSi2 by alloying with Ca or Sr based on the electronic structure calculations

International Nuclear Information System (INIS)

Imai, Yoji; Watanabe, Akio

2007-01-01

The electronic structures and total energies of BaSi 2 -SrSi 2 and BaSi 2 -CaSi 2 systems have been calculated using the first-principle pseudopotential method to clarify the band gap tunability of BaSi 2 by alloying with Sr or Ca. From an energetic consideration of the compounds where all the Ba I sites or all the Ba II sites of the BaSi 2 lattice are preferentially replaced by Sr or Ca, it is expected that the Ba I site will be preferentially replaced by Sr rather than the Ba II sites. Compounds where all the Ba II sites are replaced by Sr or all the Ba II or all the Ba I sites are replaced by Ca are energetically unfavorable compared to the undissolved system of BaSi 2 and SrSi 2 or CaSi 2 . The effect of the addition of Sr or Ca into the BaSi 2 lattice on the gap value is different depending on the replaced sites of Ba. The replacement of Ba I site by Sr will broaden the band gap of BaSi 2 , which is consistent with the observed results

On elastic limit margins for earthquake design

International Nuclear Information System (INIS)

Buchhardt, F.; Matthees, W.; Magiera, G.

1987-01-01

In the Federal Republic of Germany KTA rule 2201 being the basis for the design of nuclear power plants against seismic events is now under discussion for revisions. One of the main demands to modify KTA rule 2201 consists in cancelling the existing design philosophy, i.e. design against an operating basis earthquake (AEB) as well as against a safe shutdown earthquake (SEB). When using the present rule the 'lower' earthquake (AEB) can become design-predominant, since for AEB and SEB different types of load cases are to be superimposed with different safety factors. The scope of this study is to quantify by parametric analyses so-called 'elastic bearing capacity limit margins' for seismic events; hereby different seismic input criteria - conventional as well as recently proposed are taken into account to investigate the influence of eventual modifications in seismic design philosophy. This way a relation between AEB and SEB has to be defined so that SEB is just still predominant for the design while AEB still will yield to elastic behaviour. The study covers all German site conditions

Subsurface moisture regimes and tracer movement under two types of trench-cap designs for shallow land burial sites

International Nuclear Information System (INIS)

Perkins, B.A.; Cokal, E.J.

1986-03-01

The Los Alamos work has focused on proper design of shallow land burial (SLB) sites in arid and semiarid regions and on applying corrective measures to existing sites. One of the most important design features affecting the probability of movement of radionuclides in SLB sites is the type of trench cap placed over the waste. The cap influences such interdependent parameters as erosion, water infiltration and percolation, and biointrusion. To obtain experimental data for arid and semiarid sites, two different designs of trench caps, one with topsoil underlain with a cobble/gravel biobarrier and one with topsoil underlain with crushed tuff, were compared with respect to (1) seasonal changes in volumetric soil water content, and (2) downward migration of tracers emplaced directly below each type of trench cap. The causes for the large differences in concentrations found in this experiment need to be investigated further. Problems in environmental modeling and monitoring of arid and semiarid SLB sites because of heterogeneities in the soil profiles and their implications for SLB waste management need to be better understood. More work in trench-cap design and its influence on the many pathways available for mobilization is needed

Preliminar calculation of tornado risk in the site of Ipero

Energy Technology Data Exchange (ETDEWEB)

Pacheco, Rafael R.; Costa, Saulo Barros, E-mail: rafael.rade@ctmsp.mar.mil.br, E-mail: saulo.costa@ctmsp.mar.mil.br [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), Sao Paulo, SP (Brazil); Andrade, Delvonei A., E-mail: delvonei@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2015-07-01

General Design Criterion (GDC) 2 to 10 CFR 50 requires that 'structures, systems, and components that are important to safety shall be designed to withstand the effects of natural phenomena, such as tornadoes, without loss of capability to perform their safety functions'. According to Regulatory Guide 1.76, the design-basis intensity of a tornado for a nuclear power plant shall not exceed the intensity of the strongest tornado that occurs with the frequency of 10-7/years. Reinforcing the plant to achieve this goal represents a high increase in the costs of the project, and correspondently increase in the time required to have it commissioned. This way, the right definition of tornado risk in a site would represent savings in money for the project and in time for the licensing of a nuclear power plants. This works aims to establish a preliminary calculation of the tornado risk in the site of Ipero, where will work LABGENE from Brazilian Navy, and RMB from CNEN. (author)

AHTR Mechanical, Structural, And Neutronic Preconceptual Design

Energy Technology Data Exchange (ETDEWEB)

Varma, Venugopal Koikal [ORNL; Holcomb, David Eugene [ORNL; Peretz, Fred J [ORNL; Bradley, Eric Craig [ORNL; Ilas, Dan [ORNL; Qualls, A L [ORNL; Zaharia, Nathaniel M [ORNL

2012-10-01

This report provides an overview of the mechanical, structural, and neutronic aspects of the Advanced High Temperature Reactor (AHTR) design concept. The AHTR is a design concept for a large output Fluoride salt cooled High-temperature Reactor (FHR) that is being developed to enable evaluation of the technology hurdles remaining to be overcome prior to FHRs becoming a commercial reactor class. This report documents the incremental AHTR design maturation performed over the past year and is focused on advancing the design concept to a level of a functional, self-consistent system. The AHTR employs plate type coated particle fuel assemblies with rapid, off-line refueling. Neutronic analysis of the core has confirmed the viability of a 6-month 2-batch cycle with 9 weight-percent enriched uranium fuel. Refueling is intended to be performed automatically under visual guidance using dedicated robotic manipulators. The present design intent is for used fuel to be stored inside of containment for at least 6 months and then transferred to local dry wells for intermediate term, on-site storage. The mechanical and structural concept development effort has included an emphasis on transportation and constructability to minimize construction costs and schedule. The design intent is that all components be factory fabricated into rail transportable modules that are assembled into subsystems at an on-site workshop prior to being lifted into position using a heavy-lift crane in an open-top style construction. While detailed accident identification and response sequence analysis has yet to be performed, the design concept incorporates multiple levels of radioactive material containment including fully passive responses to all identified design basis or non-very-low frequency beyond design basis accidents. Key building design elements include: 1) below grade siting to minimize vulnerability to aircraft impact, 2) multiple natural circulation decay heat rejection chimneys, 3) seismic

Reliability-Based Optimal Design of Experiment Plans for Offshore Structures

DEFF Research Database (Denmark)

Sørensen, John Dalsgaard; Faber, Michael Havbro; Kroon, I. B.

1993-01-01

Design of cost optimal experiment plans on the basis of a preposterior analysis is discussed. In particular the planning of on-site response measurements on offshore structures in order to update probabilistic models for fatigue life estimation is addressed. Special emphasis is given to modelling...

REVIEW REPORT: BUILDING C-400 THERMAL TREATMENT 90 PERCENT REMEDIAL DESIGN REPORT AND SITE INVESTIGATION, PGDP, PADUCAH, KENTUCKY

International Nuclear Information System (INIS)

Looney, B; Jed Costanza, J; Eva Davis, E; Joe Rossabi, J; Lloyd Stewart, L; Hans Stroo, H

2007-01-01

On 9 April 2007, the U.S. Department of Energy (DOE) Headquarters, Office of Soil and Groundwater Remediation (EM-22) initiated an Independent Technical Review (ITR) of the 90% Remedial Design Report (RDR) and Site Investigation (RDSI) for thermal treatment of trichloroethylene (TCE) in the soil and groundwater in the vicinity of Building C-400 at the Paducah Gaseous Diffusion Plant (PGDP). The general ITR goals were to assess the technical adequacy of the 90% RDSI and provide recommendations sufficient for DOE to determine if modifications are warranted pertaining to the design, schedule, or cost of implementing the proposed design. The ultimate goal of the effort was to assist the DOE Paducah/Portsmouth Project Office (PPPO) and their contractor team in ''removing'' the TCE source zone located near the C-400 Building. This report provides the ITR findings and recommendations and supporting evaluations as needed to facilitate use of the recommendations. The ITR team supports the remedial action objective (RAO) at C-400 to reduce the TCE source area via subsurface Electrical Resistance Heating (ERH). Further, the ITR team commends PPPO, their contractor team, regulators, and stakeholders for the significant efforts taken in preparing the 90% RDR. To maximize TCE removal at the target source area, several themes emerge from the review which the ITR team believes should be considered and addressed before implementing the thermal treatment. These themes include the need for: (1) Accurate and site-specific models as the basis to verify the ERH design for full-scale implementation for this challenging hydrogeologic setting; (2) Flexible project implementation and operation to allow the project team to respond to observations and data collected during construction and operation; (3) Defensible performance metrics and monitoring, appropriate for ERH, to ensure sufficient and efficient clean-up; and (4) Comprehensive (creative and diverse) contingencies to address the

Environmental assessment overview: Richton Dome site, Mississippi

International Nuclear Information System (INIS)

1986-05-01

In February 1983, the US Department of Energy (DOE) identified the Richton Dome site in Mississippi as one of the nine potentially acceptable sites for a mined geologic repository for spent nuclear fuel and high-level radioactive waste. To determine their suitability, the Richton Dome site and the eight other potentially acceptable sites have been evaluated in accordance with the DOE's General Guidelines for the Recommendation of Sites for the Nuclear Waste Repositories. The site is in the Gulf interior region, which is one of five distinct geohydrologic settings considered for the first repository. On the basis of the evaluations reported in this EA, the DOE has found that the Richton Dome site is not disqualified under the guidelines. On the basis of these findings, the DOE is nominating the Richton Dome site as one of five sites suitable for characterization. 3 figs

Assessment of value-impact associated with the elimination of postulated pipe ruptures from the design basis for nuclear power plants

International Nuclear Information System (INIS)

Holman, G.S.; Chou, C.K.

1985-01-01

The US Nuclear Regulatory Commission is proposing to amend the regulations that currently require that the design basis for nuclear power plants include the postulation of dynamic effects from loss of coolant accidents up to and including the double-ended rupture of the largest pipe in the reactor coolant system. Proposed modifications would allow analyses to serve as a sufficient basis for excluding dynamic effects, including but not necessarily limited to pipe whip and jet impingement, associated with specific pipe ruptures. Only dynamic effects would be impacted; current design requirements for containment sizing and discharge capacity of emergency core cooling systems would remain unchanged. This report presents a detailed analysis of value-impact associated with the proposed amendment for PWR reactor coolant loop piping and for BWR recirculation loop piping. The effect of extending application of the proposed rule change to other piping systems is also assessed in a less quantitative manner

Transuranic waste storage and assay facility (TRUSAF) interim safety basis

International Nuclear Information System (INIS)

Gibson, K.D.

1995-09-01

The TRUSAF ISB is based upon current facility configuration and procedures. The purpose of the document is to provide the basis for interim operation or restrictions on interim operations and the authorization basis for the TRUSAF at the Hanford Site. The previous safety analysis document TRUSAF hazards Identification and Evaluation (WHC 1977) is superseded by this document

Modifications to the remedial action plan and site design for stabilization of the inactive Uranium Mill Tailings Site at Green River, Utah

International Nuclear Information System (INIS)

1994-09-01

Modifications to the water resources protection strategy detailed in the remedial action plan for the Green River, Utah, disposal site are presented. The modifications are based on new information, including ground water quality data collected after remedial action was completed and on a revised assessment of disposal cell design features, surface conditions, and site hydrogeology. The modifications will result in compliance with the U.S. EPA proposed ground water standards (52 FR 36000 (1987))

Parameters for characterizing sites for disposal of low-level radioactive waste

International Nuclear Information System (INIS)

Lutton, R.J.; Malone, P.G.; Meade, R.B.; Patrick, D.M.

1982-05-01

Sixty-seven site parameters and parameter groups are identified as important characteristics of sites for disposal of low-level radioactive waste and require detailed evaluation. Several of the most important parameters are needed for hydrological analysis while others are needed for facility design, construction, and operation. Still others are needed for baseline and detection stages of monitoring. It is recommended that all parameters be evaluated by technically qualified personnel. Appropriate tests and documentation methods are discussed in a second report, which will follow. However, site-specific testing or elaborate field measurement will not always be necessary, i.e., where indicated to be unnecessary on a technical basis. Much of this report, Appendices A through G, is directed to explaining the importances of parameters and to establishing site-specific limitations

Earthquakes and associated topics in relation to nuclear power plant siting