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Sample records for plant pfp safety

  1. Plutonium Finishing Plant (PFP) Final Safety Analysis Report (FSAR) [SEC 1 THRU 11

    Energy Technology Data Exchange (ETDEWEB)

    ULLAH, M K

    2001-02-26

    The Plutonium Finishing Plant (PFP) is located on the US Department of Energy (DOE) Hanford Site in south central Washington State. The DOE Richland Operations (DOE-RL) Project Hanford Management Contract (PHMC) is with Fluor Hanford Inc. (FH). Westinghouse Safety Management Systems (WSMS) provides management support to the PFP facility. Since 1991, the mission of the PFP has changed from plutonium material processing to preparation for decontamination and decommissioning (D and D). The PFP is in transition between its previous mission and the proposed D and D mission. The objective of the transition is to place the facility into a stable state for long-term storage of plutonium materials before final disposition of the facility. Accordingly, this update of the Final Safety Analysis Report (FSAR) reflects the current status of the buildings, equipment, and operations during this transition. The primary product of the PFP was plutonium metal in the form of 2.2-kg, cylindrical ingots called buttoms. Plutonium nitrate was one of several chemical compounds containing plutonium that were produced as an intermediate processing product. Plutonium recovery was performed at the Plutonium Reclamation Facility (PRF) and plutonium conversion (from a nitrate form to a metal form) was performed at the Remote Mechanical C (RMC) Line as the primary processes. Plutonium oxide was also produced at the Remote Mechanical A (RMA) Line. Plutonium processed at the PFP contained both weapons-grade and fuels-grade plutonium materials. The capability existed to process both weapons-grade and fuels-grade material through the PRF and only weapons-grade material through the RMC Line although fuels-grade material was processed through the line before 1984. Amounts of these materials exist in storage throughout the facility in various residual forms left from previous years of operations.

  2. Plutonium Finishing Plant (PFP) Standards/Requirements Identification Document (S/RID)

    Energy Technology Data Exchange (ETDEWEB)

    Maddox, B.S.

    1996-01-01

    This Standards/Requirements Identification Document (S/RID) sets forth the Environmental Safety and Health (ESH) standards/requirements for the Plutonium Finishing Plant (PFP). This S/RID is applicable to the appropriate life cycle phases of design, construction, operation, and preparation for decommissioning. These standards/requirements are adequate to ensure the protection of the health and safety of workers, the public, and the environment.

  3. Plutonium Finishing Plant (PFP) Standards/Requirements Identification Document (S/RID)

    International Nuclear Information System (INIS)

    Maddox, B.S.

    1996-01-01

    This Standards/Requirements Identification Document (S/RID) sets forth the Environmental Safety and Health (ESH) standards/requirements for the Plutonium Finishing Plant (PFP). This S/RID is applicable to the appropriate life cycle phases of design, construction, operation, and preparation for decommissioning. These standards/requirements are adequate to ensure the protection of the health and safety of workers, the public, and the environment

  4. Plutonium Finishing Plant safety evaluation report

    International Nuclear Information System (INIS)

    1995-01-01

    The Plutonium Finishing Plant (PFP) previously known as the Plutonium Process and Storage Facility, or Z-Plant, was built and put into operation in 1949. Since 1949 PFP has been used for various processing missions, including plutonium purification, oxide production, metal production, parts fabrication, plutonium recovery, and the recovery of americium (Am-241). The PFP has also been used for receipt and large scale storage of plutonium scrap and product materials. The PFP Final Safety Analysis Report (FSAR) was prepared by WHC to document the hazards associated with the facility, present safety analyses of potential accident scenarios, and demonstrate the adequacy of safety class structures, systems, and components (SSCs) and operational safety requirements (OSRs) necessary to eliminate, control, or mitigate the identified hazards. Documented in this Safety Evaluation Report (SER) is DOE's independent review and evaluation of the PFP FSAR and the basis for approval of the PFP FSAR. The evaluation is presented in a format that parallels the format of the PFP FSAR. As an aid to the reactor, a list of acronyms has been included at the beginning of this report. The DOE review concluded that the risks associated with conducting plutonium handling, processing, and storage operations within PFP facilities, as described in the PFP FSAR, are acceptable, since the accident safety analyses associated with these activities meet the WHC risk acceptance guidelines and DOE safety goals in SEN-35-91

  5. Plutonium Finishing Plant (PFP) hazards assessment

    International Nuclear Information System (INIS)

    Campbell, L.R.

    1998-01-01

    This report documents the hazards assessment for the Plutonium Finishing Plant (PFP) located on the US Department of Energy (DOE) Hanford Site. This hazards assessment was conducted to provide the emergency planning technical basis for the PFP. DOE Orders require an emergency planning hazards assessment for each facility that has the potential to reach or exceed the lowest level emergency classification

  6. Plutonium Finishing Plant (PFP) Waste Composition and High Efficiency Particulate Air Filter Loading

    Energy Technology Data Exchange (ETDEWEB)

    ZIMMERMAN, B.D.

    2000-12-11

    This analysis evaluates the effect of the Plutonium Finishing Plant (PFP) waste isotopic composition on Tank Farms Final Safety Analysis Report (FSAR) accidents involving high-efficiency particulate air (HEPA) filter failure in Double-Contained Receiver Tanks (DCRTs). The HEPA Filter Failure--Exposure to High Temperature or Pressure, and Steam Intrusion From Interfacing Systems accidents are considered. The analysis concludes that dose consequences based on the PFP waste isotopic composition are bounded by previous FSAR analyses. This supports USQD TF-00-0768.

  7. System Design Description PFP Thermal Stabilization

    International Nuclear Information System (INIS)

    RISENMAY, H.R.

    2000-01-01

    The purpose of this document is to provide a system design description (SDD) and design basis for the Plutonium Finishing Plant (PFP) Thermal Stabilization project. The chief objective of the SDD is to document the Structures, Systems, and Components (SSCs) that establish and maintain the facility Safety Envelope necessary for normal safe operation of the facility; as identified in the FSAR, the OSRs, and Safety Assessment Documents (SADs). This safety equipment documentation should satisfy guidelines for the SDD given in WHC-SD-CP-TI-18 1, Criteria for Identification and Control of Equipment Necessary for Preservation of the Safety Envelope and Safe Operation of PFP. The basis for operational, alarm response, maintenance, and surveillance procedures are also identified and justified in this document. This document and its appendices address the following elements of the PFP Thermal Stabilization project: Functional and design requirements; Design description; Safety Envelope Analysis; Safety Equipment Class; and Operational, maintenance and surveillance procedures

  8. Plutonium Finishing Plant (PFP) Treatment and Storage Unit Waste Analysis Plan

    International Nuclear Information System (INIS)

    PRIGNANO, A.L.

    2000-01-01

    The purpose of this waste analysis plan (WAP) is to document waste analysis activities associated with the Plutonium Finishing Plant Treatment and Storage Unit (PFP Treatment and Storage Unit) to comply with Washington Administrative Code (WAC) 173-303-300(1), (2), (4)(a) and (5). The PFP Treatment and Storage Unit is an interim status container management unit for plutonium bearing mixed waste radiologically managed as transuranic (TRU) waste. TRU mixed (TRUM) waste managed at the PFP Treatment and Storage Unit is destined for the Waste Isolation Pilot Plant (WIPP) and therefore is not subject to land disposal restrictions [WAC 173-303-140 and 40 CFR 268]. The PFP Treatment and Storage Unit is located in the 200 West Area of the Hanford Facility, Richland Washington (Figure 1). Because dangerous waste does not include source, special nuclear, and by-product material components of mixed waste, radionuclides are not within the scope of this documentation. The information on radionuclides is provided only for general knowledge

  9. Definition and means of maintaining the process vacuum liquid detection interlock systems portion of the PFP safety envelope

    International Nuclear Information System (INIS)

    LINTHO, J.E.

    2003-01-01

    The purpose of this document is to record the technical evaluation of the Technical Safety Requirements described in the Plutonium Finishing Plant (PFP) Safety Technical Requirements, HNF-SD-CP-OSR-010/Rev.1, Section 3.1.1, ''Criticality Prevention System.'' This document also defines the Safety Envelope (SE) for the liquid detection interlock system in the Process Vacuum System. The SE is derived FR-om information in the Plutonium Finishing Plant Final Safety Analysis Report (PFP FSAR), HNF-SD-CP-SAR-021, Rev 4, and the Criticality Safety Analysis Report (CSAR) for the 26-inch Hg Vacuum System, WHC-SD-SQA-CSA-20159, Rev 0-A. This document, with its appendices, provides the following: (1) The system functional requirements for determining system operability (Section 3). (2) Evaluations of equipment to determine the safety envelope boundary for the system (Section 4 list of SE boundary drawings). (3) A list of the safety envelope equipment (Appendix B). (4) Functional requirements for the individual safety envelope equipment, including appropriate set points and process parameters (Section 4). (5) A list of the operational and surveillance procedures necessary to operate and maintain the system equipment within the safety envelope (Sections 5 and 6 and Appendix A)

  10. Walkdown procedure: Seismic adequacy review of safety class 3 ampersand 4 commodities in 2736-Z ampersand ZB buildings at PFP facility

    International Nuclear Information System (INIS)

    Ocoma, E.C.

    1995-01-01

    Seismic evaluation of existing safety class (SC) 3 and non-SC 4 commodities at the Plutonium Finishing Plant (PFP) is integrated into an area walkdown program. Field walkdowns of potential PFP seismic deficiencies associated with structural failure and falling will be performed using the DOE SQUG/EPRI methodology. Potential proximity interactions are also addressed. Objective of the walkdown is to qualify as much of the equipment as practical and to identify candidates for further evaluation

  11. HANFORD PLUTONIUM FINISHG PLAN (PFP) COMPLETES PLUTONIUM STABILIZATION KEY SAFETY ISSUES CLOSED

    International Nuclear Information System (INIS)

    GERBER, M.S.

    2004-01-01

    A long and intense effort to stabilize and repackage nearly 18 metric tons (MT) of plutonium-bearing leftovers from defense production and nuclear experiments concluded successfully in February, bringing universal congratulations to the Department of Energy's Hanford Site in southeast Washington State. The victorious stabilization and packaging endeavor at the Plutonium Finishing Plant (PFP), managed and operated by prime contractor Fluor Hanford, Inc., finished ahead of all milestones in Hanford's cleanup agreement with regulators, and before deadlines set by the Defense Nuclear Facilities Safety Board (DNFSB), a part of the federal Executive Branch that oversees special nuclear materials. The PFP stabilization and packaging project also completed under budget for its four-year tenure, and has been nominated for a DOE Secretarial Award. It won the Project of the Year Award in the local chapter competition of the Project Management Institute, and is being considered for awards at the regional and national level

  12. Definition and means of maintaining the criticality detectors and alarms portion of the PFP safety envelope

    International Nuclear Information System (INIS)

    White, W.F.

    1997-01-01

    The Criticality Alarm System (CAS) provides continuous detection for high radiation (criticality) events and automatically initiates an evacuation signal to affected personnel. The Safety Envelope (SE) for PFP includes the necessary equipment and the required procedures to ensure the CAS is capable of performing its intended function. This document provides the definition and means of maintaining the SE for PFP related to the CAS. This document also identifies and provides a justification for those portions of the CAS excluded from the PFP Safety Envelope

  13. The Integrated Safety Management System Verification Enhancement Review of the Plutonium Finishing Plant (PFP)

    International Nuclear Information System (INIS)

    BRIGGS, C.R.

    2000-01-01

    The primary purpose of the verification enhancement review was for the DOE Richland Operations Office (RL) to verify contractor readiness for the independent DOE Integrated Safety Management System Verification (ISMSV) on the Plutonium Finishing Plant (PFP). Secondary objectives included: (1) to reinforce the engagement of management and to gauge management commitment and accountability; (2) to evaluate the ''value added'' benefit of direct public involvement; (3) to evaluate the ''value added'' benefit of direct worker involvement; (4) to evaluate the ''value added'' benefit of the panel-to-panel review approach; and, (5) to evaluate the utility of the review's methodology/adaptability to periodic assessments of ISM status. The review was conducted on December 6-8, 1999, and involved the conduct of two-hour interviews with five separate panels of individuals with various management and operations responsibilities related to PFP. A semi-structured interview process was employed by a team of five ''reviewers'' who directed open-ended questions to the panels which focused on: (1) evidence of management commitment, accountability, and involvement; and, (2) consideration and demonstration of stakeholder (including worker) information and involvement opportunities. The purpose of a panel-to-panel dialogue approach was to better spotlight: (1) areas of mutual reinforcement and alignment that could serve as good examples of the management commitment and accountability aspects of ISMS implementation, and, (2) areas of potential discrepancy that could provide opportunities for improvement. In summary, the Review Team found major strengths to include: (1) the use of multi-disciplinary project work teams to plan and do work; (2) the availability and broad usage of multiple tools to help with planning and integrating work; (3) senior management presence and accessibility; (4) the institutionalization of worker involvement; (5) encouragement of self-reporting and self

  14. Plutonium Finishing Plant (PFP) Criticality Alarm System Commercial Grade Item (CGI) Critical Characteristics

    International Nuclear Information System (INIS)

    WHITE, W.F.

    1999-01-01

    This document specifies the critical characteristics for Commercial Grade Items (CGI) procured for PFP's criticality alarm system as required by HNF-PRO-268 and HNF-PRO-1819. These are the minimum specifications that the equipment must meet in order to properly perform its safety function. There may be several manufacturers or models that meet the critical characteristics for any one item. PFP's Criticality Alarm System includes the nine criticality alarm system panels and their associated hardware. This includes all parts up to the first breaker in the electrical distribution system. Specific system boundaries and justifications are contained in HNF-SD-CP-SDD-003, ''Definition and Means of Maintaining the Criticality Detectors and Alarms Portion of the PFP Safety Envelope.'' The procurement requirements associated with the system necessitates procurement of some system equipment as Commercial Grade Items in accordance with HNF-PRO-268, ''Control of Purchased Items and Services.''

  15. Project plan remove special nuclear material from PFP project plutonium finishing plant

    International Nuclear Information System (INIS)

    BARTLETT, W.D.

    1999-01-01

    This plan presents the overall objectives, description, justification and planning for the Plutonium Finishing Plant (PFP) Remove Special Nuclear Material (SNM) Materials. The intent of this plan is to describe how this project will be managed and integrated with other facility stabilization and deactivation activities. This plan supplements the overall integrated plan presented in the Plutonium Finishing Plant Integrated Project Management Plan (IPMP), HNF-3617, Rev. 0. This project plan is the top-level definitive project management document for PFP Remove SNM Materials project. It specifies the technical, schedule, requirements and the cost baselines to manage the execution of the Remove SNM Materials project. Any deviations to the document must be authorized through the appropriate change control process

  16. CSER 99-001: PFP LAB Dentirating calciner

    International Nuclear Information System (INIS)

    MILLER, E.M.; DOBBIN, K.D.

    1999-01-01

    A criticality safety evaluation report was prepared for the Plutonium Finishing Plant (PFP) laboratory denigrating calciner, located in Glovebox 188-1, that converts Pu(NO 3 ) 4 solutions to the high fired stable oxide PuO 2 . Fissile mass limits and volume limits are set for the glovebox for testing operations and training operators using only nitric acid feed to a plutonium oxide bed in the calciner

  17. PFP requirements development planning guide

    International Nuclear Information System (INIS)

    SINCLAIR, J.C.

    1999-01-01

    The PFP Requirements Development Planning Guide presents the strategy and process used for the identification, allocation, and maintenance of requirements within the Plutonium Finishing Plant (PFP) integrated project baseline. Future revisions to this document will be included as attachments (e.g., results of the PFP Requirements Analysis attributable to this approach). This document is intended be a Project-owned management tool. As such, this document will periodically require revisions resulting from improvements of the information, processes, and techniques as now described. Future updates may be made to this document by PFP management and final approval of the content will be accomplished in a Baseline Change Request as it impacts the Multi-Year Work Plan, or baseline information managed in the Hanford Site Systems Engineering Baseline

  18. Project Plan For Remove Special Nuclear Material (SNM) from Plutonium Finishing Plant (PFP) Project

    International Nuclear Information System (INIS)

    BARTLETT, W.D.

    1999-01-01

    This plan presents the overall objectives, description, justification and planning for the Plutonium Finishing Plant (PFP) Remove SNM Materials. The intent of this plan is to describe how this project will be managed and integrated with other facility stabilization and deactivation activities. This plan supplements the overall integrated plan presented in the Plutonium Finishing Plant Integrated Project Management Plan (IPMP), HNF-3617. This project plan is the top-level definitive project management document for the PFP Remove SNM Materials project. It specifies the technical, schedule, requirements and the cost baseline to manage the execution of the Remove SNM Materials project. Any deviation to the document must be authorized through the appropriate change control process. The Remove SNM Materials project provides the necessary support and controls required for DOE-HQ, DOE-RL, BWHC, and other DOE Complex Contractors the path forward to negotiate shipped/receiver agreements, schedule shipments, and transfer material out of PFP to enable final deactivation

  19. Project plan remove special nuclear material from PFP project plutonium finishing plant; TOPICAL

    International Nuclear Information System (INIS)

    BARTLETT, W.D.

    1999-01-01

    This plan presents the overall objectives, description, justification and planning for the Plutonium Finishing Plant (PFP) Remove Special Nuclear Material (SNM) Materials. The intent of this plan is to describe how this project will be managed and integrated with other facility stabilization and deactivation activities. This plan supplements the overall integrated plan presented in the Plutonium Finishing Plant Integrated Project Management Plan (IPMP), HNF-3617,Rev. 0. This project plan is the top-level definitive project management document for PFP Remove SNM Materials project. It specifies the technical, schedule, requirements and the cost baselines to manage the execution of the Remove SNM Materials project. Any deviations to the document must be authorized through the appropriate change control process

  20. PFP vault operations containers for Plutonium Handling and Storage Critical Characteristics

    International Nuclear Information System (INIS)

    BONADIE, E.P.

    2000-01-01

    This document specifies the critical characteristics for containers procured for Plutonium Finishing Plant's (PFP's) Vault Operations system as required by HNF-PRO-268 and HNF-PRO-1819. These are the minimum specifications that the equipment must meet in order to perform its safety function

  1. PFP functional development planning guide

    International Nuclear Information System (INIS)

    SINCLAIR, J.C.

    1999-01-01

    The PFP Functional Development Planning Guide presents the strategy and process used for the identification, development, and analysis of functions (activities) necessary to satisfy the requirements within the Plutonium Finishing Plant (PFP) integrated project baseline. The functional analysis will provide the basis for the development of a function driven work breakdown structure. Future revisions to this document will include as attachments the results of the PFP Functional Analysis resulting from this approach. This document is intended be a Project-owned management tool. As such, the guide will periodically require revisions resulting from improvements of the information, processes, and techniques as now described

  2. Definition and Means of Maintaining the Criticality Prevention Design Features Portion of the PFP Safety Envelope

    International Nuclear Information System (INIS)

    RAMBLE, A.L.

    2000-01-01

    The purpose of this document is to record the technical evaluation of the Operational Safety Requirements described in the Plutonium Finishing Plant Final (PFP) Operational Safety Requirements, WHC-SD-CP-OSR-010. Rev. 0-N , Section 3.1.1, ''Criticality Prevention System.'' This document, with its appendices, provides the following: (1) The results of a review of Criticality Safety Analysis Reports (CSAR), later called Criticality Safety Evaluation Reports (CSER), and Criticality Prevention Specifications (CPS) to determine which equipment or components analyzed in the CSER or CPS are considered as one of the two unlikely, independent, and concurrent changes before a criticality accident is possible. (2) Evaluations of equipment or components to determine the safety boundary for the system (Section 4). (3) A list of essential drawings that show the safety system or component (Appendix A). (4) A list of the safety envelope (SE) equipment (Appendix B). (5) Functional requirements for the individual safety envelope equipment (Sections 3 and 4). (6) A list of the operational and surveillance procedures necessary to maintain the system equipment within the safety envelope (Section 5)

  3. Definition and means of maintaining the criticality detectors and alarms portion of the PFP safety envelope

    Energy Technology Data Exchange (ETDEWEB)

    White, W.F.

    1997-05-13

    The purpose of this document is to provide the definition and means of maintaining the Safety Envelope (SE) related to the Criticality Alarm System (CAS). This document provides amplification of the Limiting Condition for Operation (LCO) described in the Plutonium Finishing Plant (PFP) Operational Safety Requirements (OSR), WHC-SD-CP-OSR-010, Rev. 0, 1994, Section 3.1.2, Criticality Detectors and Alarms. This document, with its appendices, provides the following: (1) System functional requirements for determining system operability (Section 3); (2) A list of annotated system block diagrams which indicate the safety envelope boundaries (Appendix C); (3) A list of the Safety Class 1 and 2 Safety Envelope (SC-1/2 SE) equipment for input into the Master Component Index (Appendix B); (4) Functional requirements for individual SC-1/2 SE components, including appropriate setpoints and process parameters (Section 6 and Appendix A); (5) A list of the operational, maintenance and surveillance procedures necessary to operate and maintain the SC-1/2 SE components as required by the LCO (Section 6 and Appendix A).

  4. Definition and means of maintaining the criticality detectors and alarms portion of the PFP safety envelope

    International Nuclear Information System (INIS)

    White, W.F.

    1997-01-01

    The purpose of this document is to provide the definition and means of maintaining the Safety Envelope (SE) related to the Criticality Alarm System (CAS). This document provides amplification of the Limiting Condition for Operation (LCO) described in the Plutonium Finishing Plant (PFP) Operational Safety Requirements (OSR), WHC-SD-CP-OSR-010, Rev. 0, 1994, Section 3.1.2, Criticality Detectors and Alarms. This document, with its appendices, provides the following: (1) System functional requirements for determining system operability (Section 3); (2) A list of annotated system block diagrams which indicate the safety envelope boundaries (Appendix C); (3) A list of the Safety Class 1 and 2 Safety Envelope (SC-1/2 SE) equipment for input into the Master Component Index (Appendix B); (4) Functional requirements for individual SC-1/2 SE components, including appropriate setpoints and process parameters (Section 6 and Appendix A); (5) A list of the operational, maintenance and surveillance procedures necessary to operate and maintain the SC-1/2 SE components as required by the LCO (Section 6 and Appendix A)

  5. PFP supply fan motor starters

    International Nuclear Information System (INIS)

    Keck, R.D.

    1995-01-01

    The Plutonium Finishing Plant (PFP) is currently stabilizing about 25 kg of Pu sludge; upon completion of this task, PFP will be maintained in a safe standby condition to await decision from the PFP NEPA review. It can take about 10 years to initiate and complete terminal cleanout after this; the facility will then be decommissioned and decontaminated. The 234-5Z ventilation system must continue to operate until terminal cleanout. Part of the ventilation system is the seismic fan shutdown system which shuts down the ventilation supply fans in case of strong earthquake. This document presents criteria for installing solid state, reduced voltage motor starters and isolation contactors for the 8 main ventilation supply fans. The isolation contactors will shutdown the supply fans in event of earthquake

  6. PFP dangerous waste training plan

    International Nuclear Information System (INIS)

    Khojandi, J.

    1996-01-01

    This document establishes the minimum training requirements for the Plutonium Finishing Plant (PFP) personnel who are responsible for management of dangerous waste. The training plan outlines training requirements for handling of solid dangerous waste during generator accumulation and liquid dangerous waste during treatment and storage operations. The implementation of this training plan will ensure the PFP facility compliance with the training plan requirements of Dangerous Waste Regulation. Chapter 173-303-330. Washington Administrative Code (WAC). The requirements for such compliance is described in Section 11.0 of WHC-CM-7-5 Environmental Compliance Manual

  7. PFP deactivation project management plan

    International Nuclear Information System (INIS)

    Bogen, D.M.

    1997-01-01

    This document identifies the overall approach for deactivation of the Plutonium Finishing Plant (PFP) Complex, excluding the vaults, and includes a draft set of End Point Criteria for all buildings being deactivated

  8. PFP Emergency Lighting Study

    International Nuclear Information System (INIS)

    BUSCH, M.S.

    2000-01-01

    NFPA 101, section 5-9 mandates that, where required by building classification, all designated emergency egress routes be provided with adequate emergency lighting in the event of a normal lighting outage. Emergency lighting is to be arranged so that egress routes are illuminated to an average of 1.0 footcandle with a minimum at any point of 0.1 footcandle, as measured at floor level. These levels are permitted to drop to 60% of their original value over the required 90 minute emergency lighting duration after a power outage. The Plutonium Finishing Plant (PFP) has two designations for battery powered egress lights ''Emergency Lights'' are those battery powered lights required by NFPA 101 to provide lighting along officially designated egress routes in those buildings meeting the correct occupancy requirements. Emergency Lights are maintained on a monthly basis by procedure ZSR-12N-001. ''Backup Lights'' are battery powered lights not required by NFPA, but installed in areas where additional light may be needed. The Backup Light locations were identified by PFP Safety and Engineering based on several factors. (1) General occupancy and type of work in the area. Areas occupied briefly during a shiftly surveillance do not require backup lighting while a room occupied fairly frequently or for significant lengths of time will need one or two Backup lights to provide general illumination of the egress points. (2) Complexity of the egress routes. Office spaces with a standard hallway/room configuration will not require Backup Lights while a large room with several subdivisions or irregularly placed rooms, doors, and equipment will require Backup Lights to make egress safer. (3) Reasonable balance between the safety benefits of additional lighting and the man-hours/exposure required for periodic light maintenance. In some plant areas such as building 236-Z, the additional maintenance time and risk of contamination do not warrant having Backup Lights installed in all rooms

  9. PFP total operating efficiency calculation and basis of estimate

    International Nuclear Information System (INIS)

    SINCLAIR, J.C.

    1999-01-01

    The purpose of the Plutonium Finishing Plant (PFP) Total Operating Efficiency Calculation and Basis of Estimate document is to provide the calculated value and basis of estimate for the Total Operating Efficiency (TOE) for the material stabilization operations to be conducted in 234-52 Building. This information will be used to support both the planning and execution of the Plutonium Finishing Plant (PFP) Stabilization and Deactivation Project's (hereafter called the Project) resource-loaded, integrated schedule

  10. The PBMR fuel plant: Proven technology in an advanced safety environment

    International Nuclear Information System (INIS)

    Braehler, G.; Froschauer, K.; Welbers, P.; Boyes, D.

    2008-01-01

    The PBMR Fuel Plant (PFP), to be constructed at the Pelindaba site near Johannesburg will fuel the first South African Pebble Bed Modular Reactor. The qualification of the PBMR fuel shall be based on past experience with fuel which was produced in the German NUKEM/HOBEG plant and irradiated in the German AVR reactor. Accordingly, the PFP must produce the same fuel as the German plant did, and consequently, the design of the PFP has in essence to be a copy of the NUKEM/HOBEG plant. As a reminder this plant had been operated in accordance with the German regulatory rules which were defined in the years 1970/80. Since then, the requirements with regard to radiological protection, criticality safety and emission control have been significantly tightened, and of course the PFP must be designed in accordance with the most advanced international norms and standards. The implications which follow from these two potentially conflicting requirements, as defined above, are highlighted, and technical solutions are presented. Hence, the change from administrative criticality safety control to technical control, i.e. the application of safe geometry as far as possible. and the introduction of technical solutions for the remaining safe mass regime will be described. A lot of equipment in the Kernel area and in the recycling areas needed to be redesigned in safe geometry. The sensitive processes for Kernel Calcining, for the Coating and the Over-coating remain under safe mass regime, but the safety against criticality is completely independent from staff activities and based on technical measures. A new concept for safe storage of large volumes of Uranium-containing liquids has been developed. Also, the change from relatively open handling of Uranium to the application of containment enclosures wherever release of radioactivity into the room atmosphere is possible, will be addressed. This change required redesign of all process steps requiring the handling of dry Uranium oxides

  11. PFP issues/assumptions development and management planning guide

    International Nuclear Information System (INIS)

    SINCLAIR, J.C.

    1999-01-01

    The PFP Issues/Assumptions Development and Management Planning Guide presents the strategy and process used for the identification, allocation, and maintenance of an Issues/Assumptions Management List for the Plutonium Finishing Plant (PFP) integrated project baseline. Revisions to this document will include, as attachments, the most recent version of the Issues/Assumptions Management List, both open and current issues/assumptions (Appendix A), and closed or historical issues/assumptions (Appendix B). This document is intended be a Project-owned management tool. As such, this document will periodically require revisions resulting from improvements of the information, processes, and techniques as now described. Revisions that suggest improved processes will only require PFP management approval

  12. PFP solution stabilization

    International Nuclear Information System (INIS)

    Aftanas, B.L.

    1996-01-01

    This Functional Design Criteria (FDC) addresses remediation of the plutonium-bearing solutions currently in inventory at the Plutonium Finishing Plant (PFP). The recommendation from the Environmental Impact Statement (EIS) is that the solutions be treated thermally and stabilized as a solid for long term storage. For solutions which are not discardable, the baseline plan is to utilize a denitration process to stabilize the solutions prior to packaging for storage

  13. Technical Basis Document for PFP Area Monitoring Dosimetry Program

    CERN Document Server

    Cooper, J R

    2000-01-01

    This document describes the phantom dosimetry used for the PFP Area Monitoring program and establishes the basis for the Plutonium Finishing Plant's (PFP) area monitoring dosimetry program in accordance with the following requirements: Title 10, Code of Federal Regulations (CFR), part 835, ''Occupational Radiation Protection'' Part 835.403; Hanford Site Radiological Control Manual (HSRCM-1), Part 514; HNF-PRO-382, Area Dosimetry Program; and PNL-MA-842, Hanford External Dosimetry Technical Basis Manual.

  14. Technical Basis Document for PFP Area Monitoring Dosimetry Program

    International Nuclear Information System (INIS)

    COOPER, J.R.

    2000-01-01

    This document describes the phantom dosimetry used for the PFP Area Monitoring program and establishes the basis for the Plutonium Finishing Plant's (PFP) area monitoring dosimetry program in accordance with the following requirements: Title 10, Code of Federal Regulations (CFR), part 835, ''Occupational Radiation Protection'' Part 835.403; Hanford Site Radiological Control Manual (HSRCM-1), Part 514; HNF-PRO-382, Area Dosimetry Program; and PNL-MA-842, Hanford External Dosimetry Technical Basis Manual

  15. Technical Basis Document for PFP Area Monitoring Dosimetry Program

    Energy Technology Data Exchange (ETDEWEB)

    COOPER, J.R.

    2000-04-17

    This document describes the phantom dosimetry used for the PFP Area Monitoring program and establishes the basis for the Plutonium Finishing Plant's (PFP) area monitoring dosimetry program in accordance with the following requirements: Title 10, Code of Federal Regulations (CFR), part 835, ''Occupational Radiation Protection'' Part 835.403; Hanford Site Radiological Control Manual (HSRCM-1), Part 514; HNF-PRO-382, Area Dosimetry Program; and PNL-MA-842, Hanford External Dosimetry Technical Basis Manual.

  16. PFP total process throughput calculation and basis of estimate

    International Nuclear Information System (INIS)

    SINCLAIR, J.C.

    1999-01-01

    The PFP Process Throughput Calculation and Basis of Estimate document provides the calculated value and basis of estimate for process throughput associated with material stabilization operations conducted in 234-52 Building. The process throughput data provided reflects the best estimates of material processing rates consistent with experience at the Plutonium Finishing Plant (PFP) and other U.S. Department of Energy (DOE) sites. The rates shown reflect demonstrated capacity during ''full'' operation. They do not reflect impacts of building down time. Therefore, these throughput rates need to have a Total Operating Efficiency (TOE) factor applied

  17. Anticipated Radiological Dose to Worker for Plutonium Stabilization and Handling at PFP - Project W-460

    International Nuclear Information System (INIS)

    WEISS, E.V.

    2000-01-01

    This report provides estimates of the expected whole body and extremity radiological dose, expressed as dose equivalent (DE), to workers conducting planned plutonium (Pu) stabilization processes at the Hanford Site Plutonium Finishing Plant (PFP). The report is based on a time and motion dose study commissioned for Project W-460, Plutonium Stabilization and Handling, to provide personnel exposure estimates for construction work in the PFP storage vault area plus operation of stabilization and packaging equipment at PFP

  18. Anticipated Radiological Dose to Worker for Plutonium Stabilization and Handling at PFP - Project W-460

    CERN Document Server

    Weiss, E V

    2000-01-01

    This report provides estimates of the expected whole body and extremity radiological dose, expressed as dose equivalent (DE), to workers conducting planned plutonium (Pu) stabilization processes at the Hanford Site Plutonium Finishing Plant (PFP). The report is based on a time and motion dose study commissioned for Project W-460, Plutonium Stabilization and Handling, to provide personnel exposure estimates for construction work in the PFP storage vault area plus operation of stabilization and packaging equipment at PFP.

  19. PFP Wastewater Sampling Facility

    International Nuclear Information System (INIS)

    Hirzel, D.R.

    1995-01-01

    This test report documents the results obtained while conducting operational testing of the sampling equipment in the 225-WC building, the PFP Wastewater Sampling Facility. The Wastewater Sampling Facility houses equipment to sample and monitor the PFP's liquid effluents before discharging the stream to the 200 Area Treated Effluent Disposal Facility (TEDF). The majority of the streams are not radioactive and discharges from the PFP Heating, Ventilation, and Air Conditioning (HVAC). The streams that might be contaminated are processed through the Low Level Waste Treatment Facility (LLWTF) before discharging to TEDF. The sampling equipment consists of two flow-proportional composite samplers, an ultrasonic flowmeter, pH and conductivity monitors, chart recorder, and associated relays and current isolators to interconnect the equipment to allow proper operation. Data signals from the monitors are received in the 234-5Z Shift Office which contains a chart recorder and alarm annunciator panel. The data signals are also duplicated and sent to the TEDF control room through the Local Control Unit (LCU). Performing the OTP has verified the operability of the PFP wastewater sampling system. This Operability Test Report documents the acceptance of the sampling system for use

  20. Plutonium Finishing Plant (PFP) Safety Class and Safety Significant Commercial Grade Items (CGI) Critical Characteristic

    International Nuclear Information System (INIS)

    THOMAS, R.J.

    2000-01-01

    This document specifies the critical characteristics for Commercial Grade Items (CGI) procured for use in the Plutonium Finishing Plant as required by HNF-PRO-268 and HNF-PRO-1819. These are the minimum specifications that the equipment must meet in order to properly perform its safety function. There may be several manufacturers or models that meet the critical characteristics of any one item

  1. PLUTONIUM FINISHING PLANT (PFP) 241-Z LIQUID WASTE TREATMENT FACILITY DEACTIVATION AND DEMOLITION

    International Nuclear Information System (INIS)

    JOHNSTON GA

    2008-01-01

    Fluor Hanford, Inc. (FH) is proud to submit the Plutonium Finishing Plant (PFP) 241-Z liquid Waste Treatment Facility Deactivation and Demolition (D and D) Project for consideration by the Project Management Institute as Project of the Year for 2008. The decommissioning of the 241-Z Facility presented numerous challenges, many of which were unique with in the Department of Energy (DOE) Complex. The majority of the project budget and schedule was allocated for cleaning out five below-grade tank vaults. These highly contaminated, confined spaces also presented significant industrial safety hazards that presented some of the most hazardous work environments on the Hanford Site. The 241-Z D and D Project encompassed diverse tasks: cleaning out and stabilizing five below-grade tank vaults (also called cells), manually size-reducing and removing over three tons of process piping from the vaults, permanently isolating service utilities, removing a large contaminated chemical supply tank, stabilizing and removing plutonium-contaminated ventilation ducts, demolishing three structures to grade, and installing an environmental barrier on the demolition site . All of this work was performed safely, on schedule, and under budget. During the deactivation phase of the project between November 2005 and February 2007, workers entered the highly contaminated confined-space tank vaults 428 times. Each entry (or 'dive') involved an average of three workers, thus equaling approximately 1,300 individual confined -space entries. Over the course of the entire deactivation and demolition period, there were no recordable injuries and only one minor reportable skin contamination. The 241-Z D and D Project was decommissioned under the provisions of the 'Hanford Federal Facility Agreement and Consent Order' (the Tri-Party Agreement or TPA), the 'Resource Conservation and Recovery Act of 1976' (RCRA), and the 'Comprehensive Environmental Response, Compensation, and Liability Act of 1980

  2. System design description PFP thermal stabilization

    International Nuclear Information System (INIS)

    RISENMAY, H.R.

    1998-01-01

    The purpose of this document is to provide a system design description and design basis for the Plutonium Finishing P1ant (PFP) Thermal Stabilization project. The sources of material for this project are residues scraped from glovebox floors and materials already stored in vault storage that need further stabilizing to meet the 3013 storage requirements. Stabilizing this material will promote long term storage and reduced worker exposure. This document addresses: function design, equipment, and safety requirements for thermal stabilization of plutonium residues and oxides

  3. Water bath and air bath calorimeter qualification for measuring 3013 containers of plutonium oxide at the Hanford Plutonium Finishing Plant (PFP)

    International Nuclear Information System (INIS)

    WELSH, T.L.

    2003-01-01

    The purpose of this paper is to present qualification data generated from water and air-bath calorimeters measuring radioactive decay heat from plutonium oxide in DOE STD-3013-2000 (3013) containers at the Hanford Plutonium Finishing Plant (PFP). Published data concerning air and water bath calorimeters and especially 3013-qualified calorimeters is minimal at best. This paper will address the data from the measurement/qualification test plan, the heat standards used, and the calorimeter precision and accuracy results. The 3013 package is physically larger than earlier plutonium oxide storage containers, thereby necessitating a larger measurement chamber. To accommodate the measurements of the 3013 containers at PFP, Los Alamos National Laboratory (LANL) supplied a water bath dual-chambered unit and the Savannah River Technology Center (SRTC) provided two air-bath calorimeters. Both types of Calorimeters were installed in the analytical laboratory at PFP. The larger 3013 containers presented a new set of potential measurement problems: longer counting times, heat conductivity through a much larger container mass and wall thickness, and larger amounts of copper shot to assist sample thermal conductivity. These potential problems were addressed and included in the measurement/qualification test plan

  4. Engineering report (conceptual design) PFP solution stabilization

    Energy Technology Data Exchange (ETDEWEB)

    Witt, J.B.

    1997-07-17

    This Engineering Report (Conceptual Design) addresses remediation of the plutonium-bearing solutions currently in inventory at the Plutonium Finishing Plant (PFP). The recommendation from the Environmental Impact Statement (EIS) is that the solutions be treated thermally and stabilized as a solid for long term storage. For solutions which are not discardable, the baseline plan is to utilize a denitration process to stabilize the solutions prior to packaging for storage.

  5. Engineering report (conceptual design) PFP solution stabilization

    International Nuclear Information System (INIS)

    Witt, J.B.

    1997-01-01

    This Engineering Report (Conceptual Design) addresses remediation of the plutonium-bearing solutions currently in inventory at the Plutonium Finishing Plant (PFP). The recommendation from the Environmental Impact Statement (EIS) is that the solutions be treated thermally and stabilized as a solid for long term storage. For solutions which are not discardable, the baseline plan is to utilize a denitration process to stabilize the solutions prior to packaging for storage

  6. CSER 00-003: Criticality Safety Evaluation report for PFP Magnesium Hydroxide Precipitation Process for Plutonium Stabilization Glovebox 3

    International Nuclear Information System (INIS)

    LAN, J.S.

    2000-01-01

    This Criticality Safety Evaluation Report analyzes the stabilization of plutonium/uranium solutions in Glovebox 3 using the magnesium hydroxide precipitation process at PFP. The process covered are the receipt of diluted plutonium solutions into three precipitation tanks, the precipitation of plutonium from the solution, the filtering of the plutonium precipitate from the solution, the scraping of the precipitate from the filter into boats, and the initial drying of the precipitated slurry on a hot plate. A batch (up to 2.5 kg) is brought into the glovebox as plutonium nitrate, processed, and is then removed in boats for further processing. This CSER establishes limits for the magnesium hydroxide precipitation process in Glovebox 3 to maintain criticality safety while handling fissionable material

  7. PLUTONIUM FINISHING PLANT (PFP) SUB-GRADE EE/CA EVALUATION OF ALTERNATIVES: A NEW MODEL

    International Nuclear Information System (INIS)

    HOPKINS, A.M.

    2007-01-01

    An engineering evaluation/cost analysis (EE/CA) was performed at the Hanford Site's Plutonium Finishing Plant (PFP). The purpose of the EVCA was to identify the sub-grade items to be evaluated; determine the Comprehensive Environmental Response, Compensation, and Liability Act of 1980 (CERCLA) hazardous substances through process history and available data; evaluate these hazards; and as necessary, identify the available alternatives to reduce the risk associated with the contaminants. The sub-grade EWCA considered four alternatives for an interim removal action: (1) No Action; (2) Surveillance and Maintenance (S and M); (3) Stabilize and Leave in Place (Stabilization); and (4) Remove, Treat and Dispose (RTD). Each alternative was evaluated against the CERCLA criteria for effectiveness, implementability, and cost

  8. PLUTONIUM FINISHING PLANT (PFP) 241-Z LIQUID WASTE TREATMENT FACILITY DEACTIVATION AND DEMOLITION

    Energy Technology Data Exchange (ETDEWEB)

    JOHNSTON GA

    2008-01-15

    Fluor Hanford, Inc. (FH) is proud to submit the Plutonium Finishing Plant (PFP) 241-Z liquid Waste Treatment Facility Deactivation and Demolition (D&D) Project for consideration by the Project Management Institute as Project of the Year for 2008. The decommissioning of the 241-Z Facility presented numerous challenges, many of which were unique with in the Department of Energy (DOE) Complex. The majority of the project budget and schedule was allocated for cleaning out five below-grade tank vaults. These highly contaminated, confined spaces also presented significant industrial safety hazards that presented some of the most hazardous work environments on the Hanford Site. The 241-Z D&D Project encompassed diverse tasks: cleaning out and stabilizing five below-grade tank vaults (also called cells), manually size-reducing and removing over three tons of process piping from the vaults, permanently isolating service utilities, removing a large contaminated chemical supply tank, stabilizing and removing plutonium-contaminated ventilation ducts, demolishing three structures to grade, and installing an environmental barrier on the demolition site . All of this work was performed safely, on schedule, and under budget. During the deactivation phase of the project between November 2005 and February 2007, workers entered the highly contaminated confined-space tank vaults 428 times. Each entry (or 'dive') involved an average of three workers, thus equaling approximately 1,300 individual confined -space entries. Over the course of the entire deactivation and demolition period, there were no recordable injuries and only one minor reportable skin contamination. The 241-Z D&D Project was decommissioned under the provisions of the 'Hanford Federal Facility Agreement and Consent Order' (the Tri-Party Agreement or TPA), the 'Resource Conservation and Recovery Act of 1976' (RCRA), and the 'Comprehensive Environmental Response, Compensation, and

  9. Definition and means of maintaining the emergency notification and evacuation system portion of the plutonium finishing plant safety envelope

    International Nuclear Information System (INIS)

    WHITE, W.F.

    1999-01-01

    The Emergency Evacuation and Notification System provides information to the Plutonium Finishing Plant (PFP) Building Emergency Director to assist in determining appropriate emergency response, notifies personnel of the required response, and assists in their response. The report identifies the equipment in the Safety Envelope (SE) for this System and the Administrative, Maintenance, and Surveillance Procedures used to maintain the SE Equipment

  10. Definition and means of maintaining the emergency notification and evacuation system portion of the plutonium finishing plant safety envelope; TOPICAL

    International Nuclear Information System (INIS)

    WHITE, W.F.

    1999-01-01

    The Emergency Evacuation and Notification System provides information to the Plutonium Finishing Plant (PFP) Building Emergency Director to assist in determining appropriate emergency response, notifies personnel of the required response, and assists in their response. The report identifies the equipment in the Safety Envelope (SE) for this System and the Administrative, Maintenance, and Surveillance Procedures used to maintain the SE Equipment

  11. History and stabilization of the Plutonium Finishing Plant (PFP) complex, Hanford Site

    Energy Technology Data Exchange (ETDEWEB)

    Gerber, M.S., Fluor Daniel Hanford

    1997-02-18

    The 231-Z Isolation Building or Plutonium Metallurgy Building is located in the Hanford Site`s 200 West Area, approximately 300 yards north of the Plutonium Finishing Plant (PFP) (234-5 Building). When the Hanford Engineer Works (HEW) built it in 1944 to contain the final step for processing plutonium, it was called the Isolation Building. At that time, HEW used a bismuth phosphate radiochemical separations process to make `AT solution,` which was then dried and shipped to Los Alamos, New Mexico. (AT solution is a code name used during World War II for the final HEW product.) The process was carried out first in T Plant and the 224-T Bulk Reduction Building and B Plant and the 224-B Bulk Reduction Building. The 224-T and -B processes produced a concentrated plutonium nitrate stream, which then was sent in 8-gallon batches to the 231-Z Building for final purification. In the 231-Z Building, the plutonium nitrate solution underwent peroxide `strikes` (additions of hydrogen peroxide to further separate the plutonium from its carrier solutions), to form the AT solution. The AT solution was dried and shipped to the Los Alamos Site, where it was made into metallic plutonium and then into weapons hemispheres.` The 231-Z Building began `hot` operations (operations using radioactive materials) with regular runs of plutonium nitrate on January 16, 1945.

  12. PFP Commercial Grade Food Pack Cans for Plutonium Handling and Storage Critical Characteristics

    International Nuclear Information System (INIS)

    BONADIE, E.P.

    1999-01-01

    This document specifies the critical characteristics for Commercial Grade Items (CGI) procured for PFP's Vault Operations system as required by HNF-PRO-268 and HNF-PRO-1819. These are the minimum specifications that the equipment must meet in order to perform its safety function

  13. Project Management Plan to Maintain Safe and Compliant Conditions at the Plutonium Finishing Plant

    International Nuclear Information System (INIS)

    COX, G.J.

    1999-01-01

    This Project Management Plan presents the overall plan, description, mission, and workscope for the Plutonium Finishing Plant (PFP) maintain safe and compliant conditions project at PFP. This plan presents the overall description, mission, work scope, and planning for the Plutonium Finishing Plant (PFP) Maintain Safe and Compliant Conditions Project at PFP. This project includes all tasks required to maintain the safety boundary for the PFP Complex, except for the 2736-2 Vault Complex and the 234-52 vaults and vault-type rooms. The intent of this plan is to describe how this project will be managed and integrated with the stabilization, and deactivation activities. This plan supplements the overall integrated plan presented in the Plutonium Finishing Plant Integrated Project Management Plan (IPMP), HNF-3617, Rev. 0. This is the top-level definitive project management document that specifies the technical (work scope), schedule, and cost baselines that will manage the execution of this project. It describes the organizational approach and roles/responsibilities implemented to execute the project. This plan is under configuration management and any deviations must be authorized by appropriate change control action

  14. Plan for the Startup of HA-21I Furnace Operations at the Plutonium Finishing Plant (PFP)

    International Nuclear Information System (INIS)

    WILLIS, H.T.

    2000-01-01

    Achievement of Thermal Stabilization mission elements require the installation and startup of three additional muffle furnaces for the thermal stabilization of plutonium and plutonium bearing materials at the Plutonium Finishing Plant (PFP). The release to operate these additional furnaces will require an Activity Based Startup Review. The conduct of the Activity Based Startup Review (ABSR) was approved by Fluor Daniel Hanford on October 15, 1999. This plan has been developed with the objective of identifying those activities needed to guide the controlled startup of five furnaces from authorization to unrestricted operations by adding the HA-211 furnaces in an orderly and safe manner after the approval to Startup has been given. The Startup Plan provides a phased approach that bridges the activities between the completion of the Activity Based Startup Review authorizing the use of the three additional furnaces and the unrestricted operation of the five thermal stabilization muffle furnaces. The four phases are: (1) the initiation of five furnace operations using three empty (simulated full) boat charges from HA-211 and two full charges from HC-21C; (2) three furnace operations (one full charge from HA-211 and two full charges from HC-21C); (3) four furnace operations (two full charges from HA-211 and two full charges from HC-21C); and (4) integrated five furnace operations and unrestricted operations. Phase 1 of the Plan will be considered as the cold runs. This Plan also provides management oversight and administrative controls that are to be implemented until unrestricted operations are authorized. It also provides a formal review process for ensuring that all preparations needed for full five furnace operations are completed and formally reviewed prior to proceeding to the increased activity levels associated with five furnace operations. Specific objectives include: (1) To ensure that activities are conducted in a safe manner. (2) To provide supplemental

  15. Operability test procedure for PFP wastewater sampling facility

    International Nuclear Information System (INIS)

    Hirzel, D.R.

    1995-01-01

    Document provides instructions for performing the Operability Test of the 225-WC Wastewater Sampling Station which monitors the discharge to the Treated Effluent Disposal Facility from the Plutonium Finishing Plant. This Operability Test Procedure (OTP) has been prepared to verify correct configuration and performance of the PFP Wastewater sampling system installed in Building 225-WC located outside the perimeter fence southeast of the Plutonium Finishing Plant (PFP). The objective of this test is to ensure the equipment in the sampling facility operates in a safe and reliable manner. The sampler consists of two Manning Model S-5000 units which are rate controlled by the Milltronics Ultrasonic flowmeter at manhole No.C4 and from a pH measuring system with the sensor in the stream adjacent to the sample point. The intent of the dual sampling system is to utilize one unit to sample continuously at a rate proportional to the wastewater flow rate so that the aggregate tests are related to the overall flow and thereby eliminate isolated analyses. The second unit will only operate during a high or low pH excursion of the stream (hence the need for a pH control). The major items in this OTP include testing of the Manning Sampler System and associated equipment including the pH measuring and control system, the conductivity monitor, and the flow meter

  16. Plutonium Finishing Plan (PFP) Treatment and Storage Unit Interim Status Closure Plan

    International Nuclear Information System (INIS)

    PRIGNANO, A.L.

    2000-01-01

    This document describes the planned activities and performance standards for closing the Plutonium Finishing Plant (PFP) Treatment and Storage Unit. The PFP Treatment and Storage Unit is located within the 234-52 Building in the 200 West Area of the Hanford Facility. Although this document is prepared based upon Title 40 Code of Federal Regulations (CFR), Part 265, Subpart G requirements, closure of the unit will comply with Washington Administrative Code (WAC) 173-303-610 regulations pursuant to Section 5.3 of the Hanford Federal Facility Agreement and Consent Order (Tri-Party Agreement) Action Plan (Ecology et al. 1996). Because the PFP Treatment and Storage Unit manages transuranic mixed (TRUM) waste, there are many controls placed on management of the waste. Based on the many controls placed on management of TRUM waste, releases of TRUM waste are not anticipated to occur in the PFP Treatment and Storage Unit. Because the intention is to clean close the PFP Treatment and Storage Unit, postclosure activities are not applicable to this closure plan. To clean close the unit, it will be demonstrated that dangerous waste has not been left onsite at levels above the closure performance standard for removal and decontamination. If it is determined that clean closure is not possible or is environmentally impractical, the closure plan will be modified to address required postclosure activities. The PFP Treatment and Storage Unit will be operated to immobilize and/or repackage plutonium-bearing waste in a glovebox process. The waste to be processed is in a solid physical state (chunks and coarse powder) and will be sealed into and out of the glovebox in closed containers. The containers of immobilized waste will be stored in the glovebox and in additional permitted storage locations at PFP. The waste will be managed to minimize the potential for spills outside the glovebox, and to preclude spills from reaching soil. Containment surfaces will be maintained to ensure

  17. Plutonium vulnerability issues at Hanford's Plutonium Finishing Plant

    International Nuclear Information System (INIS)

    Feldt, E.; Templeton, D.W.; Tholen, E.

    1995-01-01

    The Plutonium Finishing Plant (PFP) at the Hanford, Washington Site was operated to produce plutonium (Pu) metal and oxide for national defense purposes. Due to the production requirements and methods utilized to meet national needs and the abrupt shutdown of the plant in the late 1980s, the plant was left in a condition that poses a risk of radiation exposure to plant workers, of accidental radioactive material release to the environment, and of radiation exposure to the public. In early 1994, an Environmental Impact Statement (EIS) to determine the best methods for cleaning out and stabilizing Pu materials in the PFP was started. While the EIS is being prepared, a number of immediate actions have been completed or are underway to significantly reduce the greatest hazards in the PFP. Recently, increased attention his been paid to Pu risks at Department of Energy (DOE) facilities resulting in the Department-wide Plutonium Vulnerability Assessment and a recommendation by the Defense Nuclear Facilities Safety Board (DNFSB) for DOE to develop integrated plans for managing its nuclear materials

  18. CSER 00-006 Storage of Plutonium Residue Containers in 55 Gallon Drums at the PFP

    Energy Technology Data Exchange (ETDEWEB)

    DOBBIN, K.D.

    2000-05-24

    This criticality safety evaluation report (CSER) provides the required limit set and controls for safe transit and storage of these drums in the 234-5Z Building at the PFP. A mass limit of 200 g of plutonium or fissile equivalent per drum is acceptable

  19. Nuclear criticality safety: general. 6. Application of Fixed Neutron Absorbers in the New Hanford PFP Horizontal Rack Design

    International Nuclear Information System (INIS)

    Lan, J.S.; Miller, E.M.; Toffer, H.; Mo, B.S.

    2001-01-01

    The Hanford Plutonium Finishing Plant (PFP) is currently in a waste cleanup and plutonium stabilization mode. Plutonium-bearing materials are processed through thermal treatment, creating forms of oxides suitable for long-term storage. Stabilized materials at PFP are stored in a variety of cans such as the bag-less transfer cans (BTCs), which are ultimately contained in the U.S. Department of Energy (DOE) 3013 can; both cans are larger than previously used plutonium storage containers and hold more plutonium. To compensate for the increased plutonium loadings, added engineered safety features were considered in the storage facilities. The vaults in PFP, subdivided into concrete-walled cubicles, will contain both new and older cans. The DOE 3013 and BTC cans may be loaded with up to 4.4 kg of plutonium as a compound (mostly oxide). New racks that store cans horizontally are being constructed to hold both new and older containers. The loading objective is to accommodate 70 kg of plutonium per cubicle. Two design analysis approaches for the new racks were considered. The first approach incorporated neutron absorption provided by the structural materials of the rack and the cans in determining a safe configuration. A rack loading arrangement was determined as shown in Fig. 1 and specified in Table I. This approach provides compliance with criticality control requirements; however, added administrative controls were needed to accommodate a sufficient number of cans in specific locations to achieve 70 kg of plutonium per cubicle. The 4.4-kg plutonium container can be placed only in predetermined locations. The second approach evaluated the addition of a fixed neutron absorber plate along the back wall of the cubicle (Fig. 1). The location of the special plate facilitates installation of the racks and provides additional criticality safety margin beyond the first approach. Its presence permits loading of racks with up to 4.4-kg plutonium cans in any storage locations

  20. Total Measurement Uncertainty for the Plutonium Finishing Plant (PFP) Segmented Gamma Scan Assay System

    CERN Document Server

    Fazzari, D M

    2001-01-01

    This report presents the results of an evaluation of the Total Measurement Uncertainty (TMU) for the Canberra manufactured Segmented Gamma Scanner Assay System (SGSAS) as employed at the Hanford Plutonium Finishing Plant (PFP). In this document, TMU embodies the combined uncertainties due to all of the individual random and systematic sources of measurement uncertainty. It includes uncertainties arising from corrections and factors applied to the analysis of transuranic waste to compensate for inhomogeneities and interferences from the waste matrix and radioactive components. These include uncertainty components for any assumptions contained in the calibration of the system or computation of the data. Uncertainties are propagated at 1 sigma. The final total measurement uncertainty value is reported at the 95% confidence level. The SGSAS is a gamma assay system that is used to assay plutonium and uranium waste. The SGSAS system can be used in a stand-alone mode to perform the NDA characterization of a containe...

  1. Preliminary safety evaluation for the plutonium stabilization and packaging system

    International Nuclear Information System (INIS)

    Shapley, J.E.

    1997-01-01

    This Preliminary Safety Evaluation (PSE) describes and analyzes the installation and operation of the Plutonium Stabilization and Packaging System (SPS) at the Plutonium Finishing Plant (PFP). The SPS is a combination of components required to expedite the safe and timely storage of Plutonium (Pu) oxide. The SPS program will receive site Pu packages, process the Pu for storage, package the Pu into metallic containers, and safely store the containers in a specially modified storage vault. The location of the SPS will be in the 2736- ZB building and the storage vaults will be in the 2736-Z building of the PFP, as shown in Figure 1-1. The SPS will produce storage canisters that are larger than those currently used for Pu storage at the PFP. Therefore, the existing storage areas within the PFP secure vaults will require modification. Other modifications will be performed on the 2736-ZB building complex to facilitate the installation and operation of the SPS

  2. Collaborative Negotiations: A Successful Approach for Negotiation Compliance Milestones for the transition of the PFP Hanford Nuclear Reservation

    International Nuclear Information System (INIS)

    HOPKINS, A.M.

    2003-01-01

    The new approach to negotiations was termed collaborative (win-win) rather than positional (win-lose). Collaborative negotiations were conducted to establish milestones for the decommissioning of the Plutonium Finishing Plant, PFP

  3. High knee abduction moments are common risk factors for patellofemoral pain (PFP) and anterior cruciate ligament (ACL) injury in girls: is PFP itself a predictor for subsequent ACL injury?

    Science.gov (United States)

    Myer, Gregory D; Ford, Kevin R; Di Stasi, Stephanie L; Foss, Kim D Barber; Micheli, Lyle J; Hewett, Timothy E

    2015-01-01

    Identifying risk factors for knee pain and anterior cruciate ligament (ACL) injury can be an important step in the injury prevention cycle. We evaluated two unique prospective cohorts with similar populations and methodologies to compare the incidence rates and risk factors associated with patellofemoral pain (PFP) and ACL injury. The 'PFP cohort' consisted of 240 middle and high school female athletes. They were evaluated by a physician and underwent anthropometric assessment, strength testing and three-dimensional landing biomechanical analyses prior to their basketball season. 145 of these athletes met inclusion for surveillance of incident (new) PFP by certified athletic trainers during their competitive season. The 'ACL cohort' included 205 high school female volleyball, soccer and basketball athletes who underwent the same anthropometric, strength and biomechanical assessment prior to their competitive season and were subsequently followed up for incidence of ACL injury. A one-way analysis of variance was used to evaluate potential group (incident PFP vs ACL injured) differences in anthropometrics, strength and landing biomechanics. Knee abduction moment (KAM) cut-scores that provided the maximal sensitivity and specificity for prediction of PFP or ACL injury risk were also compared between the cohorts. KAM during landing above 15.4โ€…Nm was associated with a 6.8% risk to develop PFP compared to a 2.9% risk if below the PFP risk threshold in our sample. Likewise, a KAM above 25.3โ€…Nm was associated with a 6.8% risk for subsequent ACL injury compared to a 0.4% risk if below the established ACL risk threshold. The ACL-injured athletes initiated landing with a greater knee abduction angle and a reduced hamstrings-to-quadriceps strength ratio relative to the incident PFP group. Also, when comparing across cohorts, the athletes who suffered ACL injury also had lower hamstring/quadriceps ratio than the players in the PFP sample (p15โ€…Nm of knee abduction load

  4. PFP Interface identification and management planning guide

    International Nuclear Information System (INIS)

    SINCLAIR, J.C.

    1999-01-01

    The purpose of-this planning guide is to present the process used to identify, document, and control PFP Stabilization and Deactivation Project interfaces. Revisions to this document will include, as attachments, the most recent version of the Project Interface Management List. A preliminary Interface Management List is included in Appendix A. This document is intended be a Project owned management tool. As such, this document will periodically require revisions resulting from improvements of the information, processes, and techniques as now described. For most revisions that suggest improved processes, PFP management approval is all that will be required

  5. PFP Commercial Grade Food Pack Cans for Plutonium Handling and Storage Critical Characteristics

    International Nuclear Information System (INIS)

    BONADIE, E.P.

    1999-01-01

    This document specifies the critical characteristics for Commercial Grade Items (CGI) procured for PFP's Vault Operations system as required by HNF-PRO-268 and HNF-PRO-1819. These are the minimum specifications that the equipment must meet in order to perform its safety function. The changes in these specifications have no detrimental effect on the descriptions and parameters related to handling plutonium solids in the authorization basis. Because no parameters or sequences exceed the limits described in the authorization bases, no accident or abnormal conditions are affected. The specifications prescribed in this critical characteristics document do not represent an unreviewed safety question

  6. Plutonium Finishing Plant

    Data.gov (United States)

    Federal Laboratory Consortium โ€” The Plutonium Finishing Plant, also known as PFP, represented the end of the line (the final procedure) associated with plutonium production at Hanford.PFP was also...

  7. PFP: Automated prediction of gene ontology functional annotations with confidence scores using protein sequence data.

    Science.gov (United States)

    Hawkins, Troy; Chitale, Meghana; Luban, Stanislav; Kihara, Daisuke

    2009-02-15

    Protein function prediction is a central problem in bioinformatics, increasing in importance recently due to the rapid accumulation of biological data awaiting interpretation. Sequence data represents the bulk of this new stock and is the obvious target for consideration as input, as newly sequenced organisms often lack any other type of biological characterization. We have previously introduced PFP (Protein Function Prediction) as our sequence-based predictor of Gene Ontology (GO) functional terms. PFP interprets the results of a PSI-BLAST search by extracting and scoring individual functional attributes, searching a wide range of E-value sequence matches, and utilizing conventional data mining techniques to fill in missing information. We have shown it to be effective in predicting both specific and low-resolution functional attributes when sufficient data is unavailable. Here we describe (1) significant improvements to the PFP infrastructure, including the addition of prediction significance and confidence scores, (2) a thorough benchmark of performance and comparisons to other related prediction methods, and (3) applications of PFP predictions to genome-scale data. We applied PFP predictions to uncharacterized protein sequences from 15 organisms. Among these sequences, 60-90% could be annotated with a GO molecular function term at high confidence (>or=80%). We also applied our predictions to the protein-protein interaction network of the Malaria plasmodium (Plasmodium falciparum). High confidence GO biological process predictions (>or=90%) from PFP increased the number of fully enriched interactions in this dataset from 23% of interactions to 94%. Our benchmark comparison shows significant performance improvement of PFP relative to GOtcha, InterProScan, and PSI-BLAST predictions. This is consistent with the performance of PFP as the overall best predictor in both the AFP-SIG '05 and CASP7 function (FN) assessments. PFP is available as a web service at http://dragon.bio.purdue.edu/pfp

  8. PFP Interface identification and management planning guide; TOPICAL

    International Nuclear Information System (INIS)

    SINCLAIR, J.C.

    1999-01-01

    The purpose of-this planning guide is to present the process used to identify, document, and control PFP Stabilization and Deactivation Project interfaces. Revisions to this document will include, as attachments, the most recent version of the Project Interface Management List. A preliminary Interface Management List is included in Appendix A. This document is intended be a Project owned management tool. As such, this document will periodically require revisions resulting from improvements of the information, processes, and techniques as now described. For most revisions that suggest improved processes, PFP management approval is all that will be required

  9. Fire hazard analysis for Plutonium Finishing Plant complex

    International Nuclear Information System (INIS)

    MCKINNIS, D.L.

    1999-01-01

    A fire hazards analysis (FHA) was performed for the Plutonium Finishing Plant (PFP) Complex at the Department of Energy (DOE) Hanford site. The scope of the FHA focuses on the nuclear facilities/structures in the Complex. The analysis was conducted in accordance with RLID 5480.7, [DOE Directive RLID 5480.7, 1/17/94] and DOE Order 5480.7A, ''Fire Protection'' [DOE Order 5480.7A, 2/17/93] and addresses each of the sixteen principle elements outlined in paragraph 9.a(3) of the Order. The elements are addressed in terms of the fire protection objectives stated in paragraph 4 of DOE 5480.7A. In addition, the FHA also complies with WHC-CM-4-41, Fire Protection Program Manual, Section 3.4 [1994] and WHC-SD-GN-FHA-30001, Rev. 0 [WHC, 1994]. Objectives of the FHA are to determine: (1) the fire hazards that expose the PFP facilities, or that are inherent in the building operations, (2) the adequacy of the fire safety features currently located in the PFP Complex, and (3) the degree of compliance of the facility with specific fire safety provisions in DOE orders, related engineering codes, and standards

  10. Fire hazard analysis for Plutonium Finishing Plant complex

    Energy Technology Data Exchange (ETDEWEB)

    MCKINNIS, D.L.

    1999-02-23

    A fire hazards analysis (FHA) was performed for the Plutonium Finishing Plant (PFP) Complex at the Department of Energy (DOE) Hanford site. The scope of the FHA focuses on the nuclear facilities/structures in the Complex. The analysis was conducted in accordance with RLID 5480.7, [DOE Directive RLID 5480.7, 1/17/94] and DOE Order 5480.7A, ''Fire Protection'' [DOE Order 5480.7A, 2/17/93] and addresses each of the sixteen principle elements outlined in paragraph 9.a(3) of the Order. The elements are addressed in terms of the fire protection objectives stated in paragraph 4 of DOE 5480.7A. In addition, the FHA also complies with WHC-CM-4-41, Fire Protection Program Manual, Section 3.4 [1994] and WHC-SD-GN-FHA-30001, Rev. 0 [WHC, 1994]. Objectives of the FHA are to determine: (1) the fire hazards that expose the PFP facilities, or that are inherent in the building operations, (2) the adequacy of the fire safety features currently located in the PFP Complex, and (3) the degree of compliance of the facility with specific fire safety provisions in DOE orders, related engineering codes, and standards.

  11. CSER 94-014: Storage of metal-fuel loaded EBR-II casks in concrete vault on PFP grounds

    International Nuclear Information System (INIS)

    Hess, A.L.

    1994-01-01

    A criticality safety evaluation is presented to permit EBR-2 spent fuel casks loaded with metallic fuel rods to be stored in an 8-ft diameter, cylindrical concrete vault inside the PFP security perimeter. The specific transfer of three casks with Pu alloy fuel from the Los Alamos Molten Plutonium Reactor Experiment from the burial grounds to the vault is thus covered. Up to seven casks may be emplaced in the casing with 30 inches center to center spacing. Criticality safety is assured by definitive packaging rules which keep the fissile medium dry and at a low effective volumetric density

  12. Genotypic and phenotypic diversity in populations of plant-probiotic Pseudomonas spp. colonizing roots

    Science.gov (United States)

    Picard, Christine; Bosco, Marco

    2008-01-01

    Several soil microorganisms colonizing roots are known to naturally promote the health of plants by controlling a range of plant pathogens, including bacteria, fungi, and nematodes. The use of theses antagonistic microorganisms, recently named plant-probiotics, to control plant-pathogenic fungi is receiving increasing attention, as they may represent a sustainable alternative to chemical pesticides. Many years of research on plant-probiotic microorganisms (PPM) have indicated that fluorescent pseudomonads producing antimicrobial compounds are largely involved in the suppression of the most widespread soilborne pathogens. Phenotype and genotype analysis of plant-probiotic fluorescent pseudomonads (PFP) have shown considerable genetic variation among these types of strains. Such variability plays an important role in the rhizosphere competence and the biocontrol ability of PFP strains. Understanding the mechanisms by which genotypic and phenotypic diversity occurs in natural populations of PFP could be exploited to choose those agricultural practices which best exploit the indigenous PFP populations, or to isolate new plant-probiotic strains for using them as inoculants. A number of different methods have been used to study diversity within PFP populations. Because different resolutions of the existing microbial diversity can be revealed depending on the approach used, this review first describes the most important methods used for the assessment of fluorescent Pseudomonas diversity. Then, we focus on recent data relating how differences in genotypic and phenotypic diversity within PFP communities can be attributed to geographic location, climate, soil type, soil management regime, and interactions with other soil microorganisms and host plants. It becomes evident that plant-related parameters exert the strongest influence on the genotypic and phenotypic variations in PFP populations.

  13. Interface Control Document Between the Double Shell Tanks (DST) System and the Plutonium Finishing Plan (PFP)

    International Nuclear Information System (INIS)

    MAY, T.H.

    1999-01-01

    This document identifies the requirements and responsibilities for all parties to support waste transfer from the Plutonium Finishing Plant (PFP) facility to the Double-Shell Tank (DST) System of the River Protection Project (RPP). This Interface Control Document (ICD) will not attempt to control the physical portion of this interface because the physical equipment making up this interface, and any associated interface requirements, are already in place, operational and governed by existing operating specifications and other documentation. The PFP and DST Systems have a direct physical interface (the waste transfer pipeline) that travels between the 241-2 Building (TK-D5) and DST SY-102 via 244-TX double-contained receiver tank (DCRT). The purpose of the ICD process is to formalize working agreements between the RPP DST System and organization/companies internal and external to RPP. This ICD has been developed as part of the requirements basis for design of the DST System to support the Phase I Privatization effort

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

    International Nuclear Information System (INIS)

    Mccoy, J.C.

    1997-01-01

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

  15. Evaluation of Need and Location for a Thermogravimetric Analyzer in the Plutonium Finishing Plant (PFP) Materials Stabilization

    International Nuclear Information System (INIS)

    WILLIS, H.T.

    2000-01-01

    This plan provides an analysis for locating a TGA to support PFP Thermal Stabilization processes. The scope of this document is to evaluate the need for, and location for, installation of a TGA system as a supplement to the SFE equipment for moisture measurement in pure oxides. A location assessment for the SFE equipment was previously performed (HNF 1999). Based on that assessment, co-location of the TGA system with the SFE system is the preferred option. This would enable thermally stabilized material to be analyzed for residual moisture by either the TGA system or SFE system or both This evaluation considers glovebox locations in the PFP 234-52 Building Analytical Laboratory or operating areas for the installation of the TGA system and it's supporting equipment. This evaluation considers using existing gloveboxes along with an alternative of adding a new glovebox to existing process lines. The location evaluation criteria focuses mainly on glovebox size, with qualitative consideration of relative cost and schedule impacts associated with system implementation, radiological control, and interaction with other laboratory operations and processes. In addition, the possible co-location of a TGA furnace system with the SFE system was considered

  16. Time and Temperature Test Results for PFP Thermal Stabilization Furnaces

    International Nuclear Information System (INIS)

    COMPTON, J.A.

    2000-01-01

    The national standard for plutonium storage acceptability (standard DOE-STD-3013-99, generally known as ''the 3013 standard'') has been revised to clarify the requirement for processes that will produce acceptable storage materials. The 3013 standard (Reference 1) now states that ''Oxides shall be stabilized by heating the material in an oxidizing atmosphere to a Material Temperature of at least 950 C (1742 F) for not less than 2 hours.'' The process currently in use for producing stable oxides for storage at the Plutonium Finishing Plant (PFP) heats a furnace atmosphere to 1000 C and holds it there for 2 hours. The temperature of the material being stabilized is not measured directly during this process. The Plutonium Process Support Laboratories (PPSL) were requested to demonstrate that the process currently in use at PFP is an acceptable method of producing stable plutonium dioxide consistently. A spare furnace identical to the production furnaces was set up and tested under varying conditions with non-radioactive surrogate materials. Reference 2 was issued to guide the testing program. The process currently in use at the PFP for stabilizing plutonium-bearing powders was shown to heat all the material in the furnace to at least 950 C for at least 2 hours. The current process will work for (1) relatively pure plutonium dioxide, (2) dioxide powders mixed with up to 20 weight percent magnesium oxide, and (3) dioxide powders with up to 11 weight percent magnesium oxide and 20 weight percent magnesium nitrate hexahydrate. Time and temperature data were also consistent with a successful demonstration for a mixture containing 10 weight percent each of sodium and potassium chloride; however, the molten chloride salts destroyed the thermocouples in the powder and temperature data were unavailable for part of that run. These results assume that the current operating limits of no more than 2500 grams per furnace charge and a powder height of no more than 1.5 inches remain

  17. In-depth performance evaluation of PFP and ESG sequence-based function prediction methods in CAFA 2011 experiment

    Directory of Open Access Journals (Sweden)

    Chitale Meghana

    2013-02-01

    Full Text Available Abstract Background Many Automatic Function Prediction (AFP methods were developed to cope with an increasing growth of the number of gene sequences that are available from high throughput sequencing experiments. To support the development of AFP methods, it is essential to have community wide experiments for evaluating performance of existing AFP methods. Critical Assessment of Function Annotation (CAFA is one such community experiment. The meeting of CAFA was held as a Special Interest Group (SIG meeting at the Intelligent Systems in Molecular Biology (ISMB conference in 2011. Here, we perform a detailed analysis of two sequence-based function prediction methods, PFP and ESG, which were developed in our lab, using the predictions submitted to CAFA. Results We evaluate PFP and ESG using four different measures in comparison with BLAST, Prior, and GOtcha. In addition to the predictions submitted to CAFA, we further investigate performance of a different scoring function to rank order predictions by PFP as well as PFP/ESG predictions enriched with Priors that simply adds frequently occurring Gene Ontology terms as a part of predictions. Prediction accuracies of each method were also evaluated separately for different functional categories. Successful and unsuccessful predictions by PFP and ESG are also discussed in comparison with BLAST. Conclusion The in-depth analysis discussed here will complement the overall assessment by the CAFA organizers. Since PFP and ESG are based on sequence database search results, our analyses are not only useful for PFP and ESG users but will also shed light on the relationship of the sequence similarity space and functions that can be inferred from the sequences.

  18. Plutonium Finishing Plant Treatment and Storage Unit Dangerous Waste Training Plan

    International Nuclear Information System (INIS)

    ENTROP, G.E.

    2000-01-01

    The training program for personnel performing waste management duties pertaining to the Plutonium Finishing Plant (PFP) Treatment and Storage Unit is governed by the general requirements established in the Plutonium Finishing Plant Dangerous Waste Training Plan (PFP DWTP). The PFP Treatment and Storage Unit DWTP presented below incorporates all of the components of the PFP DWTP by reference. The discussion presented in this document identifies aspects of the training program specific to the PFP Treatment and Storage Unit. The training program includes specifications for personnel instruction through both classroom and on-the-job training. Training is developed specific to waste management duties. Hanford Facility personnel directly involved with the PFP Treatment and Storage Unit will receive training to container management practices, spill response, and emergency response. These will include, for example, training in the cementation process and training pertaining to applicable elements of WAC 173-303-330(1)(d). Applicable elements from WAC 173-303-330(1)(d) for the PFP Treatment and Storage Unit include: procedures for inspecting, repairing, and replacing facility emergency and monitoring equipment; communications and alarm systems; response to fires or explosions; and shutdown of operations

  19. Maintenance implementation plan for the Plutonium Finishing Plant. Revision 3

    International Nuclear Information System (INIS)

    Meldrom, C.A.

    1996-03-01

    This document outlines the Maintenance Implementation Plan (MIP) for the Plutonium Finishing Plant (PFP) located at the Hanford site at Richland, Washington. This MIP describes the PFP maintenance program relative to DOE order 4330.4B. The MIP defines the key actions needed to meet the guidelines of the Order to produce a cost-effective and efficient maintenance program. A previous report identified the presence of significant quantities of Pu-bearing materials within PFP that pose risks to workers. PFP's current mission is to develop, install and operate processes which will mitigate these risks. The PFP Maintenance strategy is to equip the facility with systems and equipment able to sustain scheduled PFP operations. The current operating run is scheduled to last seven years. Activities following the stabilization operation will involve an Environmental Impact Statement (EIS) to determine future plant activities. This strategy includes long-term maintenance of the facility for safe occupancy and material storage. The PFP maintenance staff used the graded approach to dictate the priorities of the improvement and upgrade actions identified in Chapter 2 of this document. The MIP documents PFP compliance to the DOE 4330.4B Order. Chapter 2 of the MIP follows the format of the Order in addressing the eighteen elements. As this revision is a total rewrite, no sidebars are included to highlight changes

  20. Engineering work plan for PFP criticality alarm panel first unit re-build

    International Nuclear Information System (INIS)

    Clem, W.E.

    1994-01-01

    This document describes the first step in increasing the quality, reliability, and ease of maintenance of the nine Criticality Alarm Panels (CAP) at PFP. Development control practices and guidelines of WHC-CM-6-1, EP-2.4 and WHC-IP-1026, EPG-2.4 are applied to develop a prototype of a replacement Criticality Alarm Panel (CAP) with facility-use potential. During the development of the prototype CAP, the design requirements of all of PFP's nine CAPs are considered to develop standardized hardware and detailed design drawings that are tailored to PFP maintenance needs. Increased quality and reliability is achieved through quality hardware, proven technology and design techniques, and the use of the Class 1E workmanship standards of WHC-CM-8-1. The end result of the work described by this work plan is a verified/read-to-install replacement for CAP Z4 and verified/released H-2 drawings that are formatted such that they can easily be replicated when producing design drawings for the other eight CAPs

  1. Assessment of the Plutonium Finishing Plant Criticality Alarm System U.S. Department of Energy Richland Operations Office

    International Nuclear Information System (INIS)

    NIRIDER, L.T.

    2002-01-01

    At the request of the Assistant Manager for Safety and Engineering, the U.S. Department of Energy Richland Operations Office (RL) Engineering Support Division, performed an oversight review of the Plutonium Finishing Plant (PFP) nuclear Criticality Alarm System (CAS). The review was conducted to satisfy requirements and agreements associated with Defense Nuclear Facility Safety Board (DNFSB) Recommendation 2000-2, ''Vital Safety Systems.'' The PFP is managed by Fluor Hanford, Inc. for RL. The field assessment and staff interviews were conducted August 12 through August 19,2002. This was a limited scope assessment that consisted of a review of the nuclear CAS operations, maintenance, and compliance with National Consensus Standards Requirements. The main purpose of the assessment was to determine the adequacy of the existing alarm system and its associated infrastructure to support the PFP facility mission through the remaining facility lifetime. The Review Plan was modeled upon Criteria and Review Approach Documents (CRAD) developed for DNFSB Recommendation 2000-2 reviews conducted across the Hanford Site. Concerns regarding component degradation and failure, increasing numbers of occurrence reports associated with the alarm system, and reliability issues were addressed. Additionally, RL performed a review of the engineering aspects of the CAS including the functions of design authorities and aspects of systems engineering. However, the focus of the assessment was on operations, maintenance, and reliability of the CAS, associated procurement practices, adequacy of safety and engineering policies and procedures, safety documentation, and fundamental engineering practices including training, qualification, and systems engineering. This assessment revealed that the PFP CAS and its associated infrastructure, administrative procedures, and conduct of operations are generally effective. There are no imminent criticality safety issues associated with the operation of the

  2. A New Property of Conjugated Polymer PFP: Catalytic Degradation of Methylene Blue Aqueous Solution

    Institute of Scientific and Technical Information of China (English)

    ๆ— 

    2005-01-01

    A new property of conjugated polymer poly(furancarbinol-co-phenol)(PFP) was studied.The target copolymer was used as a catalyst after proper heating treatment. And dye methylene blue (MB) could be fully degraded and largely mineralized on PFP, under natural light or even in dark, in a few minutes. Furthermore, the catalytic activity could be preserved after several runs and the catalyst was readily separated. The effect of calcination temperature was also observed.

  3. Definition and means of maintaining the emergency notification and evacuation system portion of the Plutonium Finishing Plant safety envelope

    International Nuclear Information System (INIS)

    White, W.F.

    1997-01-01

    The Emergency Evacuation and Notification System provides information to the PFP Building Emergency Director to assist in determining appropriate emergency response, notifies personnel of the required response, and assists in their response. The report identifies the equipment in the Safety Envelope (SE) for this System and the Administrative, Maintenance, and Surveillance Procedures used to maintain the SE Equipment

  4. THE DEACTIVATION, DECONTAMINATION AND DECOMMISSIONING OF THE PLUTONIUM FINISHING PLANT, A FORMER PLUTONIUM PROCESSING FACILITY AT DOE'S HANFORD SITE

    International Nuclear Information System (INIS)

    CHARBONEAU, S.L.

    2006-01-01

    The Plutonium Finishing Plant (PFP) was constructed as part of the Manhattan Project during World War II. The Manhattan Project was developed to usher in the use of nuclear weapons to end the war. The primary mission of the PFP was to provide plutonium used as special nuclear material (SNM) for fabrication of nuclear devices for the war effort. Subsequent to the end of World War II, the PFP's mission expanded to support the Cold War effort through plutonium production during the nuclear arms race and later the processing of fuel grade mixed plutonium-uranium oxide to support DOE's breeder reactor program. In October 1990, at the close of the production mission for PFP, a shutdown order was prepared by the Department of Energy (DOE) in Washington,; DC--and issued to the Richland DOE field office. Subsequent to the shutdown order, a team from the Defense Nuclear Facilities Safety Board (DNFSB) analyzed the hazards at PFP associated with the continued storage of certain forms of plutonium solutions and solids. The assessment identified many discrete actions that were required to stabilize the different plutonium forms into stable form and repackage the material in high integrity containers. These actions were technically complicated and completed as part of the PFP nuclear material stabilization project between 1995 and early 2005. The completion of the stabilization project was a necessary first step in deactivating PFP. During stabilization, DOE entered into negotiations with the U.S. Environmental Protection Agency (EPA) and the State of Washington and established milestones for the Deactivation and Decommissioning (DandD) of the PFP. The DOE and its contractor, Fluor Hanford (Fluor), have made great progress in deactivating, decontaminating and decommissioning the PFP at the Hanford Site as detailed in this paper. Background information covering the PFP DandD effort includes descriptions of negotiations with the State of Washington concerning consent

  5. Test plan for N2 HEPA filters assembly shop stock used on PFP E4 exhaust system

    International Nuclear Information System (INIS)

    DICK, J.D.

    1999-01-01

    At Plutonium Finishing Plant (PFP) and Plutonium Reclamation Facility (PRF) Self-contained HEPA filters, encased in wooden frames and boxes, are installed in the E4 Exhaust Ventilation System to provide confinement of radioactive releases to the environment and confinement of radioactive contamination within designated zones inside the facility. Recently during the routine testing in-leakage was discovered downstream of the Self-contained HEPA filters boxes. This Test Plan describes the approach to conduct investigation of the root causes for the in-leakage of HEPA filters

  6. Safety of nuclear power plants: Design. Safety requirements

    International Nuclear Information System (INIS)

    2000-01-01

    The present publication supersedes the Code on the Safety of Nuclear Power Plants: Design (Safety Series No. 50-C-D (Rev. 1), issued in 1988). It takes account of developments relating to the safety of nuclear power plants since the Code on Design was last revised. These developments include the issuing of the Safety Fundamentals publication, The Safety of Nuclear Installations, and the present revision of various safety standards and other publications relating to safety. Requirements for nuclear safety are intended to ensure adequate protection of site personnel, the public and the environment from the effects of ionizing radiation arising from nuclear power plants. It is recognized that technology and scientific knowledge advance, and nuclear safety and what is considered adequate protection are not static entities. Safety requirements change with these developments and this publication reflects the present consensus. This Safety Requirements publication takes account of the developments in safety requirements by, for example, including the consideration of severe accidents in the design process. Other topics that have been given more detailed attention include management of safety, design management, plant ageing and wearing out effects, computer based safety systems, external and internal hazards, human factors, feedback of operational experience, and safety assessment and verification. This publication establishes safety requirements that define the elements necessary to ensure nuclear safety. These requirements are applicable to safety functions and the associated structures, systems and components, as well as to procedures important to safety in nuclear power plants. It is expected that this publication will be used primarily for land based stationary nuclear power plants with water cooled reactors designed for electricity generation or for other heat production applications (such as district heating or desalination). It is recognized that in the case of

  7. Safety of nuclear power plants: Design. Safety requirements

    International Nuclear Information System (INIS)

    2004-01-01

    The present publication supersedes the Code on the Safety of Nuclear Power Plants: Design (Safety Series No. 50-C-D (Rev. 1), issued in 1988). It takes account of developments relating to the safety of nuclear power plants since the Code on Design was last revised. These developments include the issuing of the Safety Fundamentals publication, The Safety of Nuclear Installations, and the present revision of various safety standards and other publications relating to safety. Requirements for nuclear safety are intended to ensure adequate protection of site personnel, the public and the environment from the effects of ionizing radiation arising from nuclear power plants. It is recognized that technology and scientific knowledge advance, and nuclear safety and what is considered adequate protection are not static entities. Safety requirements change with these developments and this publication reflects the present consensus. This Safety Requirements publication takes account of the developments in safety requirements by, for example, including the consideration of severe accidents in the design process. Other topics that have been given more detailed attention include management of safety, design management, plant ageing and wearing out effects, computer based safety systems, external and internal hazards, human factors, feedback of operational experience, and safety assessment and verification. This publication establishes safety requirements that define the elements necessary to ensure nuclear safety. These requirements are applicable to safety functions and the associated structures, systems and components, as well as to procedures important to safety in nuclear power plants. It is expected that this publication will be used primarily for land based stationary nuclear power plants with water cooled reactors designed for electricity generation or for other heat production applications (such as district heating or desalination). It is recognized that in the case of

  8. Integrated Safety Management System Phase I Verification for the Plutonium Finishing Plant (PFP) [VOL 1 & 2

    Energy Technology Data Exchange (ETDEWEB)

    SETH, S.S.

    2000-01-10

    U.S. Department of Energy (DOE) Policy 450.4, Safety Management System Policy commits to institutionalizing an Integrated Safety Management System (ISMS) throughout the DOE complex as a means of accomplishing its missions safely. DOE Acquisition Regulation 970.5204-2 requires that contractors manage and perform work in accordance with a documented safety management system.

  9. Safety of Nuclear Power Plants: Design. Specific Safety Requirements

    International Nuclear Information System (INIS)

    2012-01-01

    On the basis of the principles included in the Fundamental Safety Principles, IAEA Safety Standards Series No. SF-1, this Safety Requirements publication establishes requirements applicable to the design of nuclear power plants. It covers the design phase and provides input for the safe operation of the power plant. It elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. Contents: 1. Introduction; 2. Applying the safety principles and concepts; 3. Management of safety in design; 4. Principal technical requirements; 5. General plant design; 6. Design of specific plant systems.

  10. Safety of nuclear power plants: Operation. Safety requirements

    International Nuclear Information System (INIS)

    2004-01-01

    The safety of a nuclear power plant is ensured by means of its proper siting, design, construction and commissioning, followed by the proper management and operation of the plant. In a later phase, proper decommissioning is required. This Safety Requirements publication supersedes the Code on the Safety of Nuclear Power Plants: Operation, which was issued in 1988 as Safety Series No. 50-C-O (Rev. 1). The purpose of this revision was: to restructure Safety Series No. 50-C-O (Rev. 1) in the light of the basic objectives, concepts and principles in the Safety Fundamentals publication The Safety of Nuclear Installations. To be consistent with the requirements of the International Basic Safety Standards for Protection against Ionizing Radiation and for the Safety of Radiation Sources. And to reflect current practice and new concepts and technical developments. Guidance on fulfillment of these Safety Requirements may be found in the appropriate Safety Guides relating to plant operation. The objective of this publication is to establish the requirements which, in the light of experience and the present state of technology, must be satisfied to ensure the safe operation of nuclear power plants. These requirements are governed by the basic objectives, concepts and principles that are presented in the Safety Fundamentals publication The Safety of Nuclear Installations. This publication deals with matters specific to the safe operation of land based stationary thermal neutron nuclear power plants, and also covers their commissioning and subsequent decommissioning

  11. Safety of nuclear power plants: Operation. Safety requirements

    International Nuclear Information System (INIS)

    2003-01-01

    The safety of a nuclear power plant is ensured by means of its proper siting, design, construction and commissioning, followed by the proper management and operation of the plant. In a later phase, proper decommissioning is required. This Safety Requirements publication supersedes the Code on the Safety of Nuclear Power Plants: Operation, which was issued in 1988 as Safety Series No. 50-C-O (Rev. 1). The purpose of this revision was: to restructure Safety Series No. 50-C-O (Rev. 1) in the light of the basic objectives, concepts and principles in the Safety Fundamentals publication The Safety of Nuclear Installations. To be consistent with the requirements of the International Basic Safety Standards for Protection against Ionizing Radiation and for the Safety of Radiation Sources. And to reflect current practice and new concepts and technical developments. Guidance on fulfillment of these Safety Requirements may be found in the appropriate Safety Guides relating to plant operation. The objective of this publication is to establish the requirements which, in the light of experience and the present state of technology, must be satisfied to ensure the safe operation of nuclear power plants. These requirements are governed by the basic objectives, concepts and principles that are presented in the Safety Fundamentals publication The Safety of Nuclear Installations. This publication deals with matters specific to the safe operation of land based stationary thermal neutron nuclear power plants, and also covers their commissioning and subsequent decommissioning

  12. Safety of nuclear power plants: Operation. Safety requirements

    International Nuclear Information System (INIS)

    2000-01-01

    The safety of a nuclear power plant is ensured by means of its proper siting, design, construction and commissioning, followed by the proper management and operation of the plant. In a later phase, proper decommissioning is required. This Safety Requirements publication supersedes the Code on the Safety of Nuclear Power Plants: Operation, which was issued in 1988 as Safety Series No. 50-C-O (Rev. 1). The purpose of this revision was: to restructure Safety Series No. 50-C-O (Rev. 1) in the light of the basic objectives, concepts and principles in the Safety Fundamentals publication The Safety of Nuclear Installations; to be consistent with the requirements of the International Basic Safety Standards for Protection against Ionizing Radiation and for the Safety of Radiation Sources; and to reflect current practice and new concepts and technical developments. Guidance on fulfillment of these Safety Requirements may be found in the appropriate Safety Guides relating to plant operation. The objective of this publication is to establish the requirements which, in the light of experience and the present state of technology, must be satisfied to ensure the safe operation of nuclear power plants. These requirements are governed by the basic objectives, concepts and principles that are presented in the Safety Fundamentals publication The Safety of Nuclear Installations. This publication deals with matters specific to the safe operation of land based stationary thermal neutron nuclear power plants, and also covers their commissioning and subsequent decommissioning

  13. Design of plant safety model in plant enterprise engineering environment

    International Nuclear Information System (INIS)

    Gabbar, Hossam A.; Suzuki, Kazuhiko; Shimada, Yukiyasu

    2001-01-01

    Plant enterprise engineering environment (PEEE) is an approach aiming to manage the plant through its lifecycle. In such environment, safety is considered as the common objective for all activities throughout the plant lifecycle. One approach to achieve plant safety is to embed safety aspects within each function and activity within such environment. One ideal way to enable safety aspects within each automated function is through modeling. This paper proposes a theoretical approach to design plant safety model as integrated with the plant lifecycle model within such environment. Object-oriented modeling approach is used to construct the plant safety model using OO CASE tool on the basis of unified modeling language (UML). Multiple views are defined for plant objects to express static, dynamic, and functional semantics of these objects. Process safety aspects are mapped to each model element and inherited from design to operation stage, as it is naturally embedded within plant's objects. By developing and realizing the plant safety model, safer plant operation can be achieved and plant safety can be assured

  14. 1999 Annual Cathodic Protection Survey Report for PFP

    International Nuclear Information System (INIS)

    BOWMAN, T.J.

    2000-01-01

    This cathodic protection (CP) report documents the results of the 1999 annual CP survey of the underground piping within PFP property. An annual survey of CP systems is required by Washington Administrative Code (WAC). A spreadsheet to document the 1999 annual survey polarization data is included in this report. Graphs are included to trend the cathodic voltages and the polarization voltages at each test station on PFP property. The trending spans from 1994 to 1999. Graphs are also included to trend voltage and amperage outputs of each rectifier during the annual surveys. During the annual survey, resistance testing between the underground piping was conducted at each test station. The testing showed that all piping (with test leads into the test stations) was continuous with every pipe represented in the test stations. The resistance data is not documented in this report but can be accessed in work package 22-99-01003. During the annual survey, the wiring configurations of anode junction boxes AJB(R45-1) and AJB(45-1) were documented. The sketches can be accessed from the JCS work record of work package 22-99-01003. Analysis, conclusions, and recommendations of the 1999 annual CP survey results are included in this report

  15. IMPLEMENTING CHANGES TO AN APPROVED AND IN-USE DOCUMENTED SAFETY ANALYSIS

    International Nuclear Information System (INIS)

    KING JP

    2008-01-01

    The Plutonium Finishing Plant (PFP) has refined a process to ensure a comprehensive and complete DSA/TSR change implementation. Successful Nuclear Facility Safety Basis implementation is essential to avoid creating a Potential Inadequacy in Safety Analysis (PISA) situation, or implementing a facility into a non-compliance that can result in a TSR violation. Once past initial implementation, additional changes to Documented Safety Analysis (DSA) and Technical Safety Requirements (TSRs) are often needed due to needed requirement clarifications, operating experience indicating that Conditions/Required Actions/Surveillance Requirements could be improved, changes in facility conditions, or changes in facility mission etc. An effective change implementation process is essential to ensuring compliance with 10 CFR 830.202(a), 'The contractor responsible for a hazard category 1,2, or 3 DOE nuclear facility must establish and maintain the safety basis for the facility'

  16. Disposal of TRU Waste from the PFP in pipe overpack containers to WIPP Including New Security Requirements

    International Nuclear Information System (INIS)

    HOPKINS, A.M.

    2003-01-01

    The Department of Energy is responsible for the safe management and cleanup of the DOE complex. As part of the cleanup and closure of the Plutonium Finishing Plant (PFP) located on the Hanford site, the nuclear material inventory was reviewed to determine the appropriate disposition path. Based on the nuclear material characteristics, the material was designated for stabilization and packaging for long term storage and transfer to the Savannah River Site, or a decision for discard was made. The discarded material was designated as waste material and slated for disposal to the Waste Isolation Pilot Plant (WIPP). Prior to preparing any residue wastes for disposal at the WIPP, several major activities need to be completed. As detailed a processing history as possible of the material including origin of the waste must be researched and documented. A technical basis for termination of safeguards on the material must be prepared and approved. Utilizing process knowledge and processing history, the material must be characterized, sampling requirements determined, acceptable knowledge package and waste designation completed prior to disposal. All of these activities involve several organizations including the contractor, DOE, state representatives and other regulators such as EPA. At PFP, a process has been developed for meeting the many, varied requirements and successfully used to prepare several residue waste streams including Rocky Flats incinerator ash, hanford incinerator ash and Sand, Slag and Crucible (SS and C) material for disposal. These waste residues are packed into Pipe Overpack Containers for shipment to the WIPP

  17. Characterization of past and present solid waste streams from the plutonium finishing plant

    Energy Technology Data Exchange (ETDEWEB)

    Duncan, D.R.; Mayancsik, B.A. [Westinghouse Hanford Co., Richland, WA (United States); Pottmeyer, J.A.; Vejvoda, E.J.; Reddick, J.A.; Sheldon, K.M.; Weyns, M.I. [Los Alamos Technical Associates, Kennewick, WA (United States)

    1993-02-01

    During the next two decades the transuranic (TRU) wastes now stored in the burial trenches and storage facilities at the Hanford Site are to be retrieved, processed at the Waste Receiving and Processing (WRAP) Facility, and shipped to the Waste Isolation Pilot Plant (WIPP) near Carlsbad, New Mexico for final disposal. Over 50% of the TRU waste to be retrieved for shipment to the WIPP has been generated at the Plutonium Finishing Plant (PFP), also known as the Plutonium Processing and Storage Facility and Z Plant. The purpose of this report is to characterize the radioactive solid wastes generated by the PFP since its construction in 1947 using process knowledge, existing records, and history-obtained from interviews. The PFP is currently operated by Westinghouse Hanford Company (WHC) for the US Department of Energy (DOE).

  18. Characterization of past and present solid waste streams from the plutonium finishing plant

    International Nuclear Information System (INIS)

    Duncan, D.R.; Mayancsik, B.A.; Pottmeyer, J.A.; Vejvoda, E.J.; Reddick, J.A.; Sheldon, K.M.; Weyns, M.I.

    1993-02-01

    During the next two decades the transuranic (TRU) wastes now stored in the burial trenches and storage facilities at the Hanford Site are to be retrieved, processed at the Waste Receiving and Processing (WRAP) Facility, and shipped to the Waste Isolation Pilot Plant (WIPP) near Carlsbad, New Mexico for final disposal. Over 50% of the TRU waste to be retrieved for shipment to the WIPP has been generated at the Plutonium Finishing Plant (PFP), also known as the Plutonium Processing and Storage Facility and Z Plant. The purpose of this report is to characterize the radioactive solid wastes generated by the PFP since its construction in 1947 using process knowledge, existing records, and history-obtained from interviews. The PFP is currently operated by Westinghouse Hanford Company (WHC) for the US Department of Energy (DOE)

  19. Plant air systems safety study: Portsmouth Gaseous Diffusion Plant

    International Nuclear Information System (INIS)

    1982-05-01

    The Portsmouth Gaseous Diffusion Plant Air System facilities and operations are reviewed for potential safety problems not covered by standard industrial safety procedures. Information is presented under the following section headings: facility and process description (general); air plant equipment; air distribution system; safety systems; accident analysis; plant air system safety overview; and conclusion

  20. Integrated Safety Management System Phase I Verification for the Plutonium Finishing Plant (PFP) [VOL 1 and 2

    International Nuclear Information System (INIS)

    SETH, S.S.

    2000-01-01

    U.S. Department of Energy (DOE) Policy 450.4, Safety Management System Policy commits to institutionalizing an Integrated Safety Management System (ISMS) throughout the DOE complex as a means of accomplishing its missions safely. DOE Acquisition Regulation 970.5204-2 requires that contractors manage and perform work in accordance with a documented safety management system

  1. Evaluation of the Magnesium Hydroxide Treatment Process for Stabilizing PFP Plutonium/Nitric Acid Solutions

    Energy Technology Data Exchange (ETDEWEB)

    Gerber, Mark A.; Schmidt, Andrew J.; Delegard, Calvin H.; Silvers, Kurt L.; Baker, Aaron B.; Gano, Susan R.; Thornton, Brenda M.

    2000-09-28

    This document summarizes an evaluation of the magnesium hydroxide [Mg(OH)2] process to be used at the Hanford Plutonium Finishing Plant (PFP) for stabilizing plutonium/nitric acid solutions to meet the goal of stabilizing the plutonium in an oxide form suitable for storage under DOE-STD-3013-99. During the treatment process, nitric acid solutions bearing plutonium nitrate are neutralized with Mg(OH)2 in an air sparge reactor. The resulting slurry, containing plutonium hydroxide, is filtered and calcined. The process evaluation included a literature review and extensive laboratory- and bench-scale testing. The testing was conducted using cerium as a surrogate for plutonium to identify and quantify the effects of key processing variables on processing time (primarily neutralization and filtration time) and calcined product properties.

  2. Safety assessment and verification for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2001-01-01

    This publication supports the Safety Requirements on the Safety of Nuclear Power Plants: Design. This Safety Guide was prepared on the basis of a systematic review of all the relevant publications including the Safety Fundamentals, Safety of Nuclear Power Plants: Design, current and ongoing revisions of other Safety Guides, INSAG reports and other publications that have addressed the safety of nuclear power plants. This Safety Guide also provides guidance for Contracting Parties to the Convention on Nuclear Safety in meeting their obligations under Article 14 on Assessment and Verification of Safety. The Safety Requirements publication entitled Safety of Nuclear Power Plants: Design states that a comprehensive safety assessment and an independent verification of the safety assessment shall be carried out before the design is submitted to the regulatory body. This publication provides guidance on how this requirement should be met. This Safety Guide provides recommendations to designers for carrying out a safety assessment during the initial design process and design modifications, as well as to the operating organization in carrying out independent verification of the safety assessment of new nuclear power plants with a new or already existing design. The recommendations for performing a safety assessment are suitable also as guidance for the safety review of an existing plant. The objective of reviewing existing plants against current standards and practices is to determine whether there are any deviations which would have an impact on plant safety. The methods and the recommendations of this Safety Guide can also be used by regulatory bodies for the conduct of the regulatory review and assessment. Although most recommendations of this Safety Guide are general and applicable to all types of nuclear reactors, some specific recommendations and examples apply mostly to water cooled reactors. Terms such as 'safety assessment', 'safety analysis' and 'independent

  3. Remote-Reading Safety and Safeguards Surveillance System for 3013 Containers

    International Nuclear Information System (INIS)

    Lechelt, W. M.; Skorpik, J. R.; Silvers, K. L.; Szempruch, R. W.; Douglas, D. G.; Fein, K. O.

    2002-01-01

    At Hanford's Plutonium Finishing Plant (PFP), plutonium oxide is being loaded into stainless steel containers for long-term storage on the Hanford Site. These containers consist of two weld-sealed stainless steel cylinders nested one within the other. A third container holds the plutonium within the inner cylinder. This design meets the U.S. Department of Energy (DOE) storage standard, DOE-STD- 3013-2000, which anticipates a 50-year storage lifetime. The 3013 standard also requires a container surveillance program to continuously monitor pressure and to assure safeguards are adequate. However, the configuration of the container system makes using conventional measurement and monitoring methods difficult. To better meet the 3013 monitoring requirements, a team from Fluor Hanford (who manages the PFP), Pacific Northwest National Laboratory (PNNL), and Vista Engineering Technologies, LLC, developed a safer, cost-efficient, remote PFP 3013 container surveillance system. This new surveillance system is a combination of two successfully deployed technologies: (1) a magnetically coupled pressure gauge developed by Vista Engineering and (2) a radio frequency (RF) tagging device developed by PNNL. This system provides continuous, 100% monitoring of critical parameters with the containers in place, as well as inventory controls. The 3013 container surveillance system consists of three main elements: (1) an internal magnetic pressure sensor package, (2) an instrument pod (external electronics package), and (3) a data acquisition storage and display computer. The surveillance system described in this paper has many benefits for PFP and DOE in terms of cost savings and reduced personnel exposure. In addition, continuous safety monitoring (i.e., internal container pressure and temperature) of every container is responsible nuclear material stewardship and fully meets and exceeds DOE's Integrated Surveillance Program requirements

  4. Safety of Nuclear Power Plants: Design. Specific Safety Requirements (Chinese Ed.)

    International Nuclear Information System (INIS)

    2012-01-01

    On the basis of the principles included in the Fundamental Safety Principles, IAEA Safety Standards Series No. SF-1, this Safety Requirements publication establishes requirements applicable to the design of nuclear power plants. It covers the design phase and provides input for the safe operation of the power plant. It elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. Contents: 1. Introduction; 2. Applying the safety principles and concepts; 3. Management of safety in design; 4. Principal technical requirements; 5. General plant design; 6. Design of specific plant systems.

  5. Safety of Nuclear Power Plants: Design. Specific Safety Requirements (French Ed.)

    International Nuclear Information System (INIS)

    2012-01-01

    On the basis of the principles included in the Fundamental Safety Principles, IAEA Safety Standards Series No. SF-1, this Safety Requirements publication establishes requirements applicable to the design of nuclear power plants. It covers the design phase and provides input for the safe operation of the power plant. It elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. Contents: 1. Introduction; 2. Applying the safety principles and concepts; 3. Management of safety in design; 4. Principal technical requirements; 5. General plant design; 6. Design of specific plant systems.

  6. Safety of Nuclear Power Plants: Design. Specific Safety Requirements (Arabic Ed.)

    International Nuclear Information System (INIS)

    2012-01-01

    On the basis of the principles included in the Fundamental Safety Principles, IAEA Safety Standards Series No. SF-1, this Safety Requirements publication establishes requirements applicable to the design of nuclear power plants. It covers the design phase and provides input for the safe operation of the power plant. It elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. Contents: 1. Introduction; 2. Applying the safety principles and concepts; 3. Management of safety in design; 4. Principal technical requirements; 5. General plant design; 6. Design of specific plant systems.

  7. Safety of Nuclear Power Plants: Design. Specific Safety Requirements (Spanish Edition)

    International Nuclear Information System (INIS)

    2012-01-01

    This publication is a revision of Safety Requirements No. NS-R-1, Safety of Nuclear Power Plants: Design. It establishes requirements applicable to the design of nuclear power plants and elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. It will be useful for organizations involved in the design, manufacture, construction, modification, maintenance, operation and decommissioning of nuclear power plants, as well as for regulatory bodies. Contents: 1. Introduction; 2. Applying the safety principles and concepts; 3. Management of safety in design; 4. Principal technical requirements; 5. General plant design; 6. Design of specific plant systems.

  8. Safety of Nuclear Power Plants: Design. Specific Safety Requirements (Russian Edition)

    International Nuclear Information System (INIS)

    2012-01-01

    This publication is a revision of Safety Requirements No. NS-R-1, Safety of Nuclear Power Plants: Design. It establishes requirements applicable to the design of nuclear power plants and elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. It will be useful for organizations involved in the design, manufacture, construction, modification, maintenance, operation and decommissioning of nuclear power plants, as well as for regulatory bodies. Contents: 1. Introduction; 2. Applying the safety principles and concepts; 3. Management of safety in design; 4. Principal technical requirements; 5. General plant design; 6. Design of specific plant systems.

  9. COLLABORATIVE NEGOTIATIONS A SUCCESSFUL APPROACH FOR NEGOTIATING COMPLIANCE MILESTONES FOR THE TRANSITION OF THE PLUTONIUM FINISHING PLANT (PFP), HANFORD NUCLEAR RESERVATION, AND HANFORD, WASHINGTON

    International Nuclear Information System (INIS)

    Hebdon, J.; Yerxa, J.; Romine, L.; Hopkins, AM; Piippo, R.; Cusack, L.; Bond, R.; Wang, Oliver; Willis, D.

    2003-01-01

    The Hanford Nuclear Reservation is a former U. S. Department of Energy Defense Production Site. The site is currently listed on the National Priorities List of the Comprehensive Environmental Response Compensation and Liability Act of 1980 (CERCLA) and is undergoing cleanup and environmental restoration. The PFP is a former Plutonium metal production facility. The operating mission of the PFP ended with a DOE Headquarters shutdown letter in October of 1996. Generally, the receipt of a shutdown letter initiates the start of Transition (as the first step of Decommissioning) of a facility. The Hanford site is subject to the Hanford Federal Facilities Compliance Act and Consent Order (HFFCCO), an order on consent signed by the DOE, the U. S. Environmental Protection Agency, (EPA) and the Washington Department of Ecology (WDOE). Under the HFFCCO, negotiations for transition milestones begin within six months after the issuance of a shutdown order. In the case of the PFP, the Nuclear Materials disposition and stabilization activities, a DOE responsibility, were necessary as precursor activities to Transition. This situation precipitated a crisis in the negotiations between the agencies, and formal negotiations initiated in 1997 ended in failure. The negotiations reached impasse on several key regulatory and operational issues. The 1997 negotiation was characterized by a strongly positional style. DOE and the regulatory personnel took hard lines early in the negotiations and were unable to move to resolution of key issues after a year and a half. This resulted in unhappy stakeholders, poor publicity and work delays as well as wounded relationships between DOE and the regulatory community. In the 2000-2001 PFP negotiations, a completely different approach was suggested and eventually initiated: Collaborative Negotiations. The collaborative negotiation style resulted in agreement between the agencies on all key issues within 6 months of initiation. All parties were very

  10. COLLABORATIVE NEGOTIATIONS A SUCCESSFUL APPROACH FOR NEGOTIATING COMPLIANCE MILESTONES FOR THE TRANSITION OF THE PLUTONIUM FINISHING PLANT (PFP), HANFORD NUCLEAR RESERVATION, AND HANFORD, WASHINGTON

    Energy Technology Data Exchange (ETDEWEB)

    Hebdon, J.; Yerxa, J.; Romine, L.; Hopkins, AM; Piippo, R.; Cusack, L.; Bond, R.; Wang, Oliver; Willis, D.

    2003-02-27

    The Hanford Nuclear Reservation is a former U. S. Department of Energy Defense Production Site. The site is currently listed on the National Priorities List of the Comprehensive Environmental Response Compensation and Liability Act of 1980 (CERCLA) and is undergoing cleanup and environmental restoration. The PFP is a former Plutonium metal production facility. The operating mission of the PFP ended with a DOE Headquarters shutdown letter in October of 1996. Generally, the receipt of a shutdown letter initiates the start of Transition (as the first step of Decommissioning) of a facility. The Hanford site is subject to the Hanford Federal Facilities Compliance Act and Consent Order (HFFCCO), an order on consent signed by the DOE, the U. S. Environmental Protection Agency, (EPA) and the Washington Department of Ecology (WDOE). Under the HFFCCO, negotiations for transition milestones begin within six months after the issuance of a shutdown order. In the case of the PFP, the Nuclear Materials disposition and stabilization activities, a DOE responsibility, were necessary as precursor activities to Transition. This situation precipitated a crisis in the negotiations between the agencies, and formal negotiations initiated in 1997 ended in failure. The negotiations reached impasse on several key regulatory and operational issues. The 1997 negotiation was characterized by a strongly positional style. DOE and the regulatory personnel took hard lines early in the negotiations and were unable to move to resolution of key issues after a year and a half. This resulted in unhappy stakeholders, poor publicity and work delays as well as wounded relationships between DOE and the regulatory community. In the 2000-2001 PFP negotiations, a completely different approach was suggested and eventually initiated: Collaborative Negotiations. The collaborative negotiation style resulted in agreement between the agencies on all key issues within 6 months of initiation. All parties were very

  11. Industrial safety in power plants

    International Nuclear Information System (INIS)

    1987-01-01

    The proceedings of the VGB conference 'Industrial safety in power plants' held in the Gruga-Halle, Essen on January 21 and 22, 1987, contain the papers reporting on: Management responsibility for and legal consequences of industrial safety; VBG 2.0 Industrial Accident Prevention Regulation and the power plant operator; Operational experience gained with wet-type flue gas desulphurization systems; Flue gas desulphurization systems: Industrial-safety-related requirements to be met in planning and operation; the effects of the Hazardous Substances Ordinance on power plant operation; Occupational health aspects of heat-exposed jobs in power plants; Regulations of the Industrial Accident Insurance Associations concerning heat-exposed jobs and industrial medical practice; The new VBG 30 Accident Prevention Regulation 'Nuclear power plants'; Industrial safety in nuclear power plants; safe working on and within containers and confined spaces; Application of respiratory protection equipment in power plants. (HAG) [de

  12. Safety assessment and verification for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2004-01-01

    This Safety Guide was prepared under the IAEA programme for safety standards for nuclear power plants. The present publication is a revision of the IAEA Safety Guide on Management of Nuclear Power Plants for Safe Operation issued in 1984. It supplements Section 2 of the Safety Requirements publication on Safety of Nuclear Power Plants: Operation. Nuclear power technology is different from the customary technology of power generation from fossil fuel and by hydroelectric means. One major difference between the management of nuclear power plants and that of conventional generating plants is the emphasis that should be placed on nuclear safety, quality assurance, the management of radioactive waste and radiological protection, and the accompanying national regulatory requirements. This Safety Guide highlights the important elements of effective management in relation to these aspects of safety. The attention to be paid to safety requires that the management recognize that personnel involved in the nuclear power programme should understand, respond effectively to, and continuously search for ways to enhance safety in the light of any additional requirements socially and legally demanded of nuclear energy. This will help to ensure that safety policies that result in the safe operation of nuclear power plants are implemented and that margins of safety are always maintained. The structure of the organization, management standards and administrative controls should be such that there is a high degree of assurance that safety policies and decisions are implemented, safety is continuously enhanced and a strong safety culture is promoted and supported. The objective of this publication is to guide Member States in setting up an operating organization which facilitates the safe operation of nuclear power plants to a high level internationally. The second objective is to provide guidance on the most important organizational elements in order to contribute to a strong safety

  13. Safety assessment and verification for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2005-01-01

    This Safety Guide was prepared under the IAEA programme for safety standards for nuclear power plants. The present publication is a revision of the IAEA Safety Guide on Management of Nuclear Power Plants for Safe Operation issued in 1984. It supplements Section 2 of the Safety Requirements publication on Safety of Nuclear Power Plants: Operation. Nuclear power technology is different from the customary technology of power generation from fossil fuel and by hydroelectric means. One major difference between the management of nuclear power plants and that of conventional generating plants is the emphasis that should be placed on nuclear safety, quality assurance, the management of radioactive waste and radiological protection, and the accompanying national regulatory requirements. This Safety Guide highlights the important elements of effective management in relation to these aspects of safety. The attention to be paid to safety requires that the management recognize that personnel involved in the nuclear power programme should understand, respond effectively to, and continuously search for ways to enhance safety in the light of any additional requirements socially and legally demanded of nuclear energy. This will help to ensure that safety policies that result in the safe operation of nuclear power plants are implemented and that margins of safety are always maintained. The structure of the organization, management standards and administrative controls should be such that there is a high degree of assurance that safety policies and decisions are implemented, safety is continuously enhanced and a strong safety culture is promoted and supported. The objective of this publication is to guide Member States in setting up an operating organization which facilitates the safe operation of nuclear power plants to a high level internationally. The second objective is to provide guidance on the most important organizational elements in order to contribute to a strong safety

  14. ASSESSING CHEMICAL HAZARDS AT THE PLUTONIUM FINISHING PLANT FOR PLANNING FUTURE DECONTAMINATION AND DECOMMISSIONING

    International Nuclear Information System (INIS)

    HOPKINS, A.M.; KLOS, D.B.; MINETT, M.J.

    2007-01-01

    This paper documents the fiscal year (FY) 2006 assessment to evaluate potential chemical and radiological hazards associated with vessels and piping in the former plutonium process areas at Hanford's Plutonium Finishing Plant (PFP). Evaluations by PFP engineers as design authorities for specific systems and other subject-matter experts were conducted to identify the chemical hazards associated with transitioning the process areas for the long-term layup of PFP before its eventual final decontamination and decommissioning (D and D). D and D activities in the main process facilities were suspended in September 2005 for a period of between 5 and 10 years. A previous assessment conducted in FY 2003 found that certain activities to mitigate chemical hazards could be deferred safely until the D and D of PFP, which had been scheduled to result in a slab-on-grade condition by 2009. As a result of necessary planning changes, however, D and D activities at PFP will be delayed until after the 2009 time frame. Given the extended project and plant life, it was determined that a review of the plant chemical hazards should be conducted. This review to determine the extended life impact of chemicals is called the ''Plutonium Finishing Plant Chemical Hazards Assessment, FY 2006''. This FY 2006 assessment addresses potential chemical and radiological hazard areas identified by facility personnel and subject-matter experts who reevaluated all the chemical systems (items) from the FY 2003 assessment. This paper provides the results of the FY 2006 chemical hazards assessment and describes the methodology used to assign a hazard ranking to the items reviewed

  15. Leveraging the Partnership for Patients' Initiative to Improve Patient Safety and Quality Within the Military Health System.

    Science.gov (United States)

    King, Heidi B; Kesling, Kimberly; Birk, Carmen; Walker, Theodore; Taylor, Heather; Datena, Michael; Burgess, Brittany; Bower, Lyndsay

    2017-03-01

    Partnership for Patients (PfP) was a national initiative sponsored by the Department of Health and Human Services, Centers for Medicare and Medicaid Services, to reduce preventable hospital acquired conditions (HACs) by 40% and readmissions (within 30 days) by 20%, by the end of 2013 (as compared to the baseline of CY2010). Along with partners across the nation, the Assistant Secretary of Defense for Health Affairs, Dr. Jonathan Woodson, pledged to support PfP in June 2011. Participation of the Military Health System (MHS) in PfP marked the implementation of the first enterprise-wide patient safety initiative. Three phases of the MHS initiative were developed to meet the aims of the national PfP initiative: (1) Planning and Design, (2) Implementation, and (3) Monitoring and Sustainment. The Planning and Design phase focused on the identification of evidence-based practices (Table III); the development of implementation guides; the implementation of various communication, education, and improvement strategies; and the development of methods by which to track progress and share successes. The implementation phase focused on identifying roles and responsibilities across all levels of care; creating, disseminating, and implementing evidence-based practices at participating military treatment facilities; and establishing a structured learning action network. Finally, during the monitoring and sustainment phase, per the guidance of the Agency for Healthcare Research and Quality, an overall HAC rate was developed for quarterly analysis. The HAC rate per 1,000 dispositions (i.e., discharges) was an aggregate of all PfP HACs. Using the HAC rate, the improvement rate was calculated by comparing the current quarter's HAC rate to the baseline (CY2010). This allowed the MHS to track the overall progress across the enterprise. The MHS achieved a number of accomplishments, including a 15.8% cumulative reduction in HACs by the end of 2013, an 11.1% reduction in readmissions

  16. Update on the Department of Energy's 1994 plutonium vulnerability assessment for the plutonium finishing plant

    International Nuclear Information System (INIS)

    HERZOG, K.R.

    1999-01-01

    A review of the environmental, safety, and health vulnerabilities associated with the continued storage of PFP's inventory of plutonium bearing materials and other SNM. This report re-evaluates the five vulnerabilities identified in 1994 at the PFP that are associated with SNM storage. This new evaluation took a more detailed look and applied a risk ranking process to help focus remediation efforts

  17. Nuclear power plant safety

    International Nuclear Information System (INIS)

    Otway, H.J.

    1974-01-01

    Action at the international level will assume greater importance as the number of nuclear power plants increases, especially in the more densely populated parts of the world. Predictions of growth made prior to October 1973 [9] indicated that, by 1980, 14% of the electricity would be supplied by nuclear plants and by the year 2000 this figure would be about 50%. This will make the topic of international co-operation and standards of even greater importance. The IAEA has long been active in providing assistance to Member States in the siting design and operation of nuclear reactors. These activities have been pursued through advisory missions, the publication of codes of practice, guide books, technical reports and in arranging meetings to promote information exchange. During the early development of nuclear power, there was no well-established body of experience which would allow formulation of internationally acceptable safety criteria, except in a few special cases. Hence, nuclear power plant safety and reliability matters often received an ad hoc approach which necessarily entailed a lack of consistency in the criteria used and in the levels of safety required. It is clear that the continuation of an ad hoc approach to safety will prove inadequate in the context of a world-wide nuclear power industry, and the international trade which this implies. As in several other fields, the establishment of internationally acceptable safety standards and appropriate guides for use by regulatory bodies, utilities, designers and constructors, is becoming a necessity. The IAEA is presently planning the development of a comprehensive set of basic requirements for nuclear power plant safety, and the associated reliability requirements, which would be internationally acceptable, and could serve as a standard frame of reference for nuclear plant safety and reliability analyses

  18. Power plants and safety 1982

    International Nuclear Information System (INIS)

    1982-01-01

    The papers of this volume deal with the whole range of safety issues from planning and construction to the operation of power plants, and discuss also issues like availability and safety of power plants, protective clothes and their incommodating effect, alternatives for rendering hot-water generators safe and the safety philosophy in steam turbine engineering. (HAG) [de

  19. ALARA Design Review for the Resumption of the Plutonium Finishing Plant (PFP) Cementation Process Project Activities

    CERN Document Server

    Dayley, L

    2000-01-01

    The requirements for the performance of radiological design reviews are codified in 10CFR835, Occupational Radiation Protection. The basic requirements for the performance of ALARA design reviews are presented in the Hanford Site Radiological Control Manual (HSRCM). The HSRCM has established trigger levels requiring radiological reviews of non-routine or complex work activities. These requirements are implemented in site procedures HNF-PRO-1622 and 1623. HNF-PRO-1622 Radiological Design Review Process requires that ''radiological design reviews [be performed] of new facilities and equipment and modifications of existing facilities and equipment''. In addition, HNF-PRO-1623 Radiological Work Planning Process requires a formal ALARA Review for planned activities that are estimated to exceed 1 person-rem total Dose Equivalent (DE). The purpose of this review is to validate that the original design for the PFP Cementation Process ensures that the principles of ALARA (As Low As Reasonably Achievable) were included...

  20. Plan for the Initiation of HA-211 Furnace Operations at the Plutonium Finishing Plan (PFP)

    International Nuclear Information System (INIS)

    WILLIS, H.T.

    2000-01-01

    This plan provides a phased approach authorizing the use of three additional muffle furnaces for thermal stabilization. Achievement of Thermal Stabilization mission elements require the installation and startup of three additional muffle furnaces for the thermal stabilization of plutonium and plutonium bearing materials at the Plutonium Finishing Plant (PFP). The release to operate these additional furnaces will require an Activity Based Startup Review. The conduct of the Activity Based Startup Review (ABSR) was approved by Fluor Daniel Hanford on October 15, 1999. This plan has been developed with the objective of identifying those activities needed to guide the controlled startup of five furnaces from authorization to unrestricted operations by adding the HA-211 furnaces in an orderly and safe manner after the approval to Startup has been given

  1. Safety of Nuclear Power Plants: Commissioning and Operation. Specific Safety Requirements

    International Nuclear Information System (INIS)

    2016-01-01

    This publication describes the requirements to be met to ensure the safe operation of nuclear power plants. It takes into account developments in areas such as long term operation of nuclear power plants, plant ageing, periodic safety review, probabilistic safety analysis and risk informed decision making processes. In addition, the requirements are governed by, and must apply, the safety objective and safety principles that are established in the IAEA Safety Standards Series No. SF-1, Fundamental Safety Principles. A review of Safety Requirements publications was commenced in 2011 following the accident in the Fukushima Daiichi nuclear power plant in Japan. The review revealed no significant areas of weakness and resulted in just a small set of amendments to strengthen the requirements and facilitate their implementation, which are contained in the present publication

  2. Definition and means of maintaining the supply ventilation system seismic shutdown portion of the PFP safety envelope. Revision 2

    International Nuclear Information System (INIS)

    Keck, R.D.

    1995-01-01

    This report describes the modifications to the ventilation system for the Plutonium Finishing Plant. Topics discussed in this report include; system functional requirements, evaluations of equipment, a list of drawings showing the safety envelope boundaries; list of safety envelope equipment, functional requirements for individual safety envelope equipment, and a list of the operational, maintenance and surveillance procedures necessary to operate and maintain the system equipment

  3. Plant designer's view of the operator's role in nuclear plant safety

    International Nuclear Information System (INIS)

    Corcoran, W.R.; Church, J.F.; Cross, M.T.; Porter, N.J.

    1981-01-01

    The nuclear plant operator's role supports the design assumptions and equipment with four functional tasks. He must set up th plant for predictable response to disturbances, operate the plant so as to minimize the likelihood and severity of event initiators, assist in accomplishing the safety functions, and feed back operating experiences to reinforce or redefine the safety analyses' assumptions. The latter role enhances the operator effectiveness in the former three roles. The Safety Level Concept offers a different perspective that enables the operator to view his roles in nuclear plant safety. This paper outlines the operator's role in nuclear safety and classifies his tasks using the Safety Level Concept

  4. Periodic safety reviews of nuclear power plants

    International Nuclear Information System (INIS)

    Toth, Csilla

    2009-01-01

    Operational nuclear power plants (NPPs) are generally subject to routine reviews of plant operation and special safety reviews following operational events. In addition, many Member States of the International Atomic Energy Agency (IAEA) have initiated systematic safety reassessment, termed periodic safety review (PSR), to assess the cumulative effects of plant ageing and plant modifications, operating experience, technical developments, site specific, organizational and human aspects. These reviews include assessments of plant design and operation against current safety standards and practices. PSRs are considered an effective way of obtaining an overall view of actual plant safety, to determine reasonable and practical modifications that should be made in order to maintain a high level of safety throughout the plant's operating lifetime. PSRs can be used as a means to identify time limiting features of the plant. The trend is to use PSR as a condition for deciding whether to continue operation of the plant beyond the originally established design lifetime and for assessing the status of the plant for long term operation. To assist Member States in the implementation of PSR, the IAEA develops safety standards, technical documents and provides different services: training courses, workshops, technical meetings and safety review missions for the independent assessment of the PSR at NPPs, including the requirements for PSR, the review process and the PSR final reports. This paper describes the PSR's objectives, scopes, methods and the relationship of PSR with other plant safety related activities and recent experiences of Member States in implementation of PSRs at NPPs. (author)

  5. Safety of Nuclear Power Plants: Commissioning and Operation. Specific Safety Requirements

    International Nuclear Information System (INIS)

    2017-01-01

    This publication is a revision of IAEA Safety Standards Series No. NS-R-2, Safety of Nuclear Power Plants: Operation, and has been extended to cover the commissioning stage. It describes the requirements to be met to ensure the safe commissioning, operation, and transition from operation to decommissioning of nuclear power plants. Over recent years there have been developments in areas such as long term operation of nuclear power plants, plant ageing, periodic safety review, probabilistic safety analysis review and risk informed decision making processes. It became necessary to revise the IAEAโ€™s Safety Requirements in these areas and to correct and/or improve the publication on the basis of feedback from its application by both the IAEA and its Member States. In addition, the requirements are governed by, and must apply, the safety objective and safety principles that are established in the IAEA Safety Standards Series No. SF-1, Fundamental Safety Principles. A review of Safety Requirements publications, initiated in 2011 following the accident in the Fukushima Daiichi nuclear power plant in Japan, revealed no significant areas of weakness but resulted in a small set of amendments to strengthen the requirements and facilitate their implementation. These are contained in the present publication.

  6. Seismic safety of nuclear power plants

    International Nuclear Information System (INIS)

    Guerpinar, A.; Godoy, A.

    2001-01-01

    This paper summarizes the work performed by the International Atomic Energy Agency in the areas of safety reviews and applied research in support of programmes for the assessment and enhancement of seismic safety in Eastern Europe and in particular WWER type nuclear power plants during the past seven years. Three major topics are discussed; engineering safety review services in relation to external events, technical guidelines for the assessment and upgrading of WWER type nuclear power plants, and the Coordinated Research Programme on 'Benchmark study for the seismic analysis and testing of WWER type nuclear power plants'. These topics are summarized in a way to provide an overview of the past and present safety situation in selected WWER type plants which are all located in Eastern European countries. Main conclusion of the paper is that although there is now a thorough understanding of the seismic safety issues in these operating nuclear power plants, the implementation of seismic upgrades to structures, systems and components are lagging behind, particularly for those cases in which the re-evaluation indicated the necessity to strengthen the safety related structures or install new safety systems. (author)

  7. Modifications to nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2004-01-01

    This Safety Guide was prepared under the IAEA's programme for safety standards for nuclear power plants. It supplements Section 7 of the Safety Requirements publication on Safety of Nuclear Power Plants: Operation, which establishes the safety requirements for the modification of nuclear power plants. Reasons for carrying out modifications to nuclear power plants may include: (1) maintaining or strengthening existing safety provisions and thus maintaining consistency with or improving on the current design. (2) recovering from plant faults. (3) improving the thermal performance or increasing the power rating of the plant. (4) increasing the maintainability of the plant, reducing the radiation exposure of personnel or reducing the costs of plant maintenance. And (5) extending the design life of the plant. Most modifications, made on the basis of operating experience, are intended to improve on the design or to improve operational performance and flexibility. Some are rendered necessary by new regulatory requirements, ageing of the plant or obsolescence of equipment. However, the benefits of regularly updating the plant design can be jeopardized if modifications are not kept under rigorous control throughout the lifetime of the plant. The need to reduce costs and improve efficiency, in combination with changes to the structure of the electricity generation sector of the economy in many countries, has led many companies to make changes in the structure of the operating organization for nuclear power plants. Whatever the reason for such organizational changes, consideration should be given to the effects of those changes with the aim of ensuring that they would have no impacts that would compromise the safety of the plant. The objective of this Safety Guide is to provide guidance and recommendations on controlling activities relating to modifications at nuclear power plants in order to reduce risk and to ensure that the configuration of the plant is at all times under

  8. Modifications to nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2007-01-01

    This Safety Guide was prepared under the IAEA's programme for safety standards for nuclear power plants. It supplements Section 7 of the Safety Requirements publication on Safety of Nuclear Power Plants: Operation, which establishes the safety requirements for the modification of nuclear power plants. Reasons for carrying out modifications to nuclear power plants may include: (1) maintaining or strengthening existing safety provisions and thus maintaining consistency with or improving on the current design. (2) recovering from plant faults. (3) improving the thermal performance or increasing the power rating of the plant. (4) increasing the maintainability of the plant, reducing the radiation exposure of personnel or reducing the costs of plant maintenance. And (5) extending the design life of the plant. Most modifications, made on the basis of operating experience, are intended to improve on the design or to improve operational performance and flexibility. Some are rendered necessary by new regulatory requirements, ageing of the plant or obsolescence of equipment. However, the benefits of regularly updating the plant design can be jeopardized if modifications are not kept under rigorous control throughout the lifetime of the plant. The need to reduce costs and improve efficiency, in combination with changes to the structure of the electricity generation sector of the economy in many countries, has led many companies to make changes in the structure of the operating organization for nuclear power plants. Whatever the reason for such organizational changes, consideration should be given to the effects of those changes with the aim of ensuring that they would have no impacts that would compromise the safety of the plant. The objective of this Safety Guide is to provide guidance and recommendations on controlling activities relating to modifications at nuclear power plants in order to reduce risk and to ensure that the configuration of the plant is at all times under

  9. Safety assessment principles for nuclear plants

    International Nuclear Information System (INIS)

    1992-01-01

    The present Safety Assessment Principles result from the revision of those which were drawn up following a recommendation arising from the Sizewell-B enquiry. The principles presented here relate only to nuclear safety; there is a section on risks from normal operation and accident conditions and the standards against which those risks are assessed. A major part of the document deals with the principles that cover the design of nuclear plants. The revised Safety assessment principles are aimed primarily at the safety assessment of new nuclear plants but they will also be used in assessing existing plants. (UK)

  10. Safety culture in nuclear power plants

    International Nuclear Information System (INIS)

    Weihe, G. von; Pamme, H.

    2003-01-01

    Experience shows that German nuclear power plants have always been operated reliably and safely. Over the years, the safety level in these plants has been raised considerably so that they can stand any comparison with other countries. This is confirmed by the two reports published by the Federal Ministry for the Environment on the nuclear safety convention. Behind this, there must obviously stand countless appropriate 'good practices' and a safety management system in nuclear power plants. (orig.) [de

  11. Plutonium Finishing Plant (PFP) HVAC System Component Index; FINAL

    International Nuclear Information System (INIS)

    DICK, J.D.

    1999-01-01

    This document identities the components, design media, procedures and defines the critical characteristics of Commercial Grade Items necessary to ensure the HVAC system provides these functions. This document lists safety class (SC) and safety significant (SS) components for the Heating Ventilation Air Conditioning (HVAC) and specifies the critical characteristics for Commercial Grade Items (CGI), as required by HNF-PRO-268 and HNF-PRO-1819. These are the minimum specifications that the equipment must meet in order to properly perform its safety function. There may be several manufacturers or models that meet the critical characteristics for any one item

  12. The critical safety functions and plant operation

    International Nuclear Information System (INIS)

    Corcoran, W.R.; Church, J.F.; Cross, M.T.; Guinn, W.M.; Porter, N.J.

    1981-01-01

    The operator's role in nuclear safety is outlined and the concept of ''safety functions'' introduced. Safety functions are a group of actions that prevent core melt or minimize radiation releases to the general public. They can be used to provide a hierarchy of practical plant protection that an operator should use. The plant safety evaluation uses four inputs in predicting the results of an event: the event initiator, the plant design, the initial plant conditions and setup, and the operator actions. If any of these inputs are not as assumed in the evaluation, confidence that the consequences will be as predicted is reduced. Based on the safety evaluation, the operator has three roles in assuring that the consequences of an event will be no worse than the predicted acceptable results: Maintain plant setup in readiness to properly respond. Operate the plant in a manner such that fewer, milder events minimize the frequency and the severity of adverse events. Monitor the plant to verify that the safety functions are accomplished. The operator needs a systematic approach to mitigating the consequences of an event. The concept of safety functions introduces this systematic approach and presents a hierarchy of protection. If the operator has difficulty identifying an event for any reason, the systematic safety function approach allows accomplishing the overall path of mitigating consequences. Ten functions designed to protect against core melt, preserve containment integrity, prevent indirect release of radioactivity, and maintain vital auxiliaries needed to support the other safety functions are identified

  13. Safety principles for nuclear power plants

    International Nuclear Information System (INIS)

    Vuorinen, A.

    1993-01-01

    The role and purpose of safety principles for nuclear power plants are discussed. A brief information is presented on safety objectives as given in the INSAG documents. The possible linkage is discussed between the two mentioned elements of nuclear safety and safety culture. Safety culture is a rather new concept and there is more than one interpretation of the definition given by INSAG. The defence in depth is defined by INSAG as a fundamental principle of safety technology of nuclear power. Discussed is the overall strategy for safety measures, and features of nuclear power plants provided by the defence-in-depth concept. (Z.S.) 7 refs

  14. Safety of Nuclear Power Plants: Design. Specific Safety Requirements (French Edition)

    International Nuclear Information System (INIS)

    2017-01-01

    This publication establishes requirements applicable to the design of nuclear power plants and elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. It will be useful for organizations involved in design, manufacture, construction, modification, maintenance, operation and decommissioning of nuclear power plants, as well as for regulatory bodies. A review of Safety Requirements publications was commenced in 2011 following the accident in the Fukushima Daiichi nuclear power plant in Japan. The review revealed no significant areas of weakness and resulted in just a small set of amendments to strengthen the requirements and facilitate their implementation, which are contained in the present publication.

  15. Safety of Nuclear Power Plants: Design. Specific Safety Requirements (Russian Edition)

    International Nuclear Information System (INIS)

    2016-01-01

    This publication establishes requirements applicable to the design of nuclear power plants and elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. It will be useful for organizations involved in design, manufacture, construction, modification, maintenance, operation and decommissioning of nuclear power plants, as well as for regulatory bodies. A review of Safety Requirements publications was commenced in 2011 following the accident in the Fukushima Daiichi nuclear power plant in Japan. The review revealed no significant areas of weakness and resulted in just a small set of amendments to strengthen the requirements and facilitate their implementation, which are contained in the present publication.

  16. Safety of Nuclear Power Plants: Design. Specific Safety Requirements (Arabic Edition)

    International Nuclear Information System (INIS)

    2017-01-01

    This publication establishes requirements applicable to the design of nuclear power plants and elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. It will be useful for organizations involved in design, manufacture, construction, modification, maintenance, operation and decommissioning of nuclear power plants, as well as for regulatory bodies. A review of Safety Requirements publications was commenced in 2011 following the accident in the Fukushima Daiichi nuclear power plant in Japan. The review revealed no significant areas of weakness and resulted in just a small set of amendments to strengthen the requirements and facilitate their implementation, which are contained in the present publication.

  17. Safety design of Qinshan Nuclear Power Plant

    International Nuclear Information System (INIS)

    Ouyang Yu; Zhang Lian; Du Shenghua; Zhao Jiayu

    1984-01-01

    Safety issues have been greatly emphasized through the design of the Qinshan Nuclear Power Plant. Reasonable safety margine has been taken into account in the plant design parameters, the design incorporated various safeguard systems, such as engineering safety feature systems, safety protection systems and the features to resist natural catastrophes, e. g. earthquake, hurricanes, tide and so on. Preliminary safety analysis and environmental effect assessment have been done and anti-accident provisions and emergency policy were carefully considered. Qinshan Nuclear Power Plant safety related systems are designed in accordance with the common international standards established in the late 70's, as well as the existing engineering standard of China

  18. The safety of nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    1988-01-01

    Do nuclear power plants present an unjustifiable risk Can there be confidence in their safety The Uranium Institute invited a group of senior safety experts from eight different Western countries operating different types of reactors to provide an authoritative explanation for non-specialists of the basic principles of reactor safety, their application and their implications. The report presents the group's opinion on the level of safety achieved in the Western nuclear power plants with which the authors are directly familiar. Although many of the points made may well also be true for non-Western reactors, the report does not cover them except where specifically stated. It does describe and discuss the causes of the Chernobyl disaster. It does not compare nuclear power with other fuels, nor does it deal with its benefits, since however great the benefits from the peaceful use of nuclear power, and its own advantages over other fuels, they could not compensate for lack of safety. The conclusion reached is that the risk associated with electricity production at nuclear power plants can be kept very low. Proper use of the extensive knowledge available today can guarantee operation of nuclear power plants at very high safety levels, carrying very low risks, both to health and of contamination of the environment: risks that are continually lowered by upgrading existing plants and their operation, and by the design of future power plants. (author).

  19. Safety of Nuclear Power Plants: Commissioning and Operation

    International Nuclear Information System (INIS)

    2011-01-01

    This publication is a revision of Safety Requirements No. NS-R-2, Safety of Nuclear Power Plants: Operation, and has been extended to cover the commissioning stage. It describes the requirements to be met to ensure the safe operation of nuclear power plants. Over recent years there have been developments in areas such as long term operation, plant ageing, periodic safety review, probabilistic safety analysis and risk informed decision making processes. It became necessary to revise the IAEA's safety requirements in these areas and to correct and/or improve the publication on the basis of feedback from its application by both the IAEA and its Member States. In addition, the requirements are governed by, and must apply, the safety objective and safety principles that are established in the Fundamental Safety Principles. Contents: 1. Introduction; 2. Safety objectives and principles; 3. The management and organizational structure of the operating organization; 4. Management of operational safety; 5. Operational safety programmes; 6. Plant commissioning; 7. Plant operations; 8. Maintenance, testing, surveillance and inspection; 9. Preparation for decommissioning.

  20. Definition and means of maintaining the ventilation system confinement portion of the PFP safety envelope

    Energy Technology Data Exchange (ETDEWEB)

    Dick, J.D.; Grover, G.A.; O`Brien, P.M., Fluor Daniel Hanford

    1997-03-05

    The Plutonium Finishing Plant Heating Ventilation and Cooling system provides for the confinement of radioactive releases to the environment and provides for the confinement of radioactive contamination within designated zones inside the facility. This document identifies the components and procedures necessary to ensure the HVAC system provides these functions. Appendices E through J provide a snapshot of non-safety class HVAC equipment and need not be updated when the remainder of the document and Appendices A through D are updated.

  1. The role of probabilistic safety assessment and probabilistic safety criteria in nuclear power plant safety

    International Nuclear Information System (INIS)

    1992-01-01

    The purpose of this Safety Report is to provide guidelines on the role of probabilistic safety assessment (PSA) and a range of associated reference points, collectively referred to as probabilistic safety criteria (PSC), in nuclear safety. The application of this Safety Report and the supporting Safety Practice publication should help to ensure that PSA methodology is used appropriately to assess and enhance the safety of nuclear power plants. The guidelines are intended for use by nuclear power plant designers, operators and regulators. While these guidelines have been prepared with nuclear power plants in mind, the principles involved have wide application to other nuclear and non-nuclear facilities. In Section 2 of this Safety Report guidelines are established on the role PSA can play as part of an overall safety assurance programme. Section 3 summarizes guidelines for the conduct of PSAs, and in Section 4 a PSC framework is recommended and guidance is provided for the establishment of PSC values

  2. Plutonium finishing plant dangerous waste training plan

    International Nuclear Information System (INIS)

    ENTROP, G.E.

    1999-01-01

    This training plan describes general requirements, worker categories, and provides course descriptions for operation of the Plutonium Finish Plant (PFP) waste generation facilities, permitted treatment, storage and disposal (TSD) units, and the 90-Day Accumulation Areas

  3. Barsebaeck power plant - safety and emergency measures

    International Nuclear Information System (INIS)

    Anon.

    1981-01-01

    A Swedish-Danish Committee on safety at the Swedish nuclear power plant Barsebaeck was established in 1979 in order to evaluate the nuclear safety at Barsebaeck with a view to the reactor accident at the Three-Mile-Island nuclear power plant March 28, 1979. According to the committees mandate the investigations of the Kemeny Commission, the Rogouin investigation, investigations of the American Nuclear Regulatory Commission, and the Swedish report ''Safe nuclear power'' have been taken into consideration by the Committee. Furthermore, it has formed the basis for the Committees work that the authority responsibility for the safety at Barsebaeck lies with the Swedish authorities, and that these authorities have evaluated the safety aspects before the permissions for operation of the Barsebaeck power plant were given and hereafter currently in connection with the inspection of the power plant. The report prepared by the Commission treats aspects as: a) Nuclear safety at the Barsebaeck power plant, b) reactor safety and emergency provisions, c) common elements in the emergency provision situation in Sweden and Denmark, d) ongoing investigations on course of events during accidents and release limiting safety systems. (BP)

  4. Safety of Nuclear Power Plants: Commissioning and Operation. Specific Safety Requirements (French Edition)

    International Nuclear Information System (INIS)

    2016-01-01

    This publication describes the requirements to be met to ensure the safe operation of nuclear power plants. It takes into account developments in areas such as long term operation of nuclear power plants, plant ageing, periodic safety review, probabilistic safety analysis and risk informed decision making processes. In addition, the requirements are governed by, and must apply, the safety objective and safety principles that are established in the IAEA Safety Standards Series No. SF-1, Fundamental Safety Principles. A review of Safety Requirements publications was commenced in 2011 following the accident in the Fukushima Daiichi nuclear power plant in Japan. The review revealed no significant areas of weakness and resulted in just a small set of amendments to strengthen the requirements and facilitate their implementation, which are contained in the present publication.

  5. Safety in nuclear power plants

    International Nuclear Information System (INIS)

    Koeberlein, K.

    1987-01-01

    In nuclear power plants large amounts of radioactive fission products ensue from the fission of uranium. In order to protect the environment, the radioactive material is confined in multiple 'activity barriers' (crystal matrix of the fuel, fuel cladding, coolant boundary, safety containment, reactor building). These barriers are protected by applying a defense-in-depth concept (high quality requirements, protection systems which recognize and terminate operational incidents, safety systems to cope with accidents). In spite of a favorable safety record of German nuclear power plants it is obvious - and became most evident by the Chernobyl accident - that absolute safety is not achievable. At Chernobyl, however, design disadvantages of that reactor type (like positive reactivity feedback of coolant voiding, missing safety containment) played an important role in accident initiation and progression. Such features of the Russian 'graphite-moderated pressure tube boiling water reactor' are different from those of light water reactors operating in western countries. The essential steps of the waste management of the nuclear fuel cycle ('Entsorgung') are the interim storage, the shipment, and the reprocessing of the spent fuel and the final repository of radioactive waste. Reprocessing means the separation of fossil material (uranium, plutonium) from radioactive waste. Legal requirements for radiological protection of the environment, which are identical for nuclear power plants and reprocessing plant, are complied with by means of comprehensive filter systems. Safety problems of a reprocessing plant are eased considerably by the fact that system pressures, process temperatures and energy densities are low. In order to confine the radioactive waste from the biosphere for a very long period of time, it is to be discarded after appropriate treatment into the deep geological underground of salt domes. (orig./HP) [de

  6. Basic safety principles for nuclear power plant

    International Nuclear Information System (INIS)

    Zhang Shiguan

    1989-01-01

    To ensure the safety operation of nuclear power plant, one should strictly adhere to the implelmentation of safety codes and the establishment of nuclear safety code system, as well as the applicable basic safety principles of nuclear power plants. This article briefly introduce the importance of nuclear codes and its economic benefits and the implementation of basic safety principles to be accumulated in practice for many years by various countries

  7. Safety of Nuclear Power Plants: Commissioning and Operation

    International Nuclear Information System (INIS)

    2011-01-01

    The safety of a nuclear power plant is ensured by means of proper site selection, design, construction and commissioning, and the evaluation of these, followed by proper management, operation and maintenance of the plant. In a later phase, a proper transition to decommissioning is required. The organization and management of plant operations ensures that a high level of safety is achieved through the effective management and control of operational activities. This publication is a revision of the Safety Requirements publication Safety of Nuclear Power Plants: Operation, which was issued in 2000 as IAEA Safety Standards Series No. NS-R-2. The purpose of this revision was to restructure Safety Standards Series No. NS-R-2 in the light of new operating experience and new trends in the nuclear industry; to introduce new requirements that were not included in Safety Standards Series No. NS-R-2 on the operation of nuclear power plants; and to reflect current practices, new concepts and technical developments. This update also reflects feedback on the use of the standards, both from Member States and from the IAEA's safety related activities. The publication is presented in the new format for Safety Requirements publications. The present publication reflects the safety principles of the Fundamental Safety Principles. It has been harmonized with IAEA Safety Standards Series No. GS-R-3 on The Management System for Facilities and Activities. Guidance on the fulfilment of the safety requirements is provided in supporting Safety Guides. The terminology used in this publication is defined and explained in the IAEA Safety Glossary. The objective of this publication is to establish the requirements which, in the light of experience and the present state of technology, must be satisfied to ensure the safe operation of nuclear power plants. These requirements are governed by the safety objective and safety principles that are established in the Fundamental Safety Principles. This

  8. Radioactive Air Emissions Notice of Construction for the Magnesium Hydroxide Precipitation Process at the Plutonium Finishing Plant (PFP)

    International Nuclear Information System (INIS)

    JANSKY, M.T.

    1999-01-01

    The following description and any attachments and references are provided to the Washington State Department of Health (WDOH), Division of Radiation Protection, Air Emissions and Defense Waste (WAC) 246-247, Radiation Protection-Air Emissions. The WAC 246-247-060, ''Applications, registration, and licensing'', states ''This section describes the information requirements for approval to construct, modify, and operate an emission unit. Any NOC requires the submittal of information listed in Appendix A.'' Appendix A (WAC 246-247-1 10) lists the requirements that must be addressed. Additionally, the following description, attachments and references are provided to the US Environmental Protection Agency (EPA) as an NOC, in accordance with Title 40, Code of Federal Regulations (CFR), Part 61, ''National Emission Standards for Hazardous Air Pollutants.'' The information required for submittal to the EPA is specified in 40 CFR 61.07. The potential emissions from this activity are estimated to provide greater than 0.1 millirem per year total effective dose equivalent (TEDE) to the hypothetical offsite maximally exposed individual (MEI), and commencement is needed within a short time. Therefore, this application also is intended to provide notification of the anticipated date of initial startup in accordance with the requirement listed in 40 CFR 61.09(a)(1), and it is requested that approval of this application also will constitute EPA acceptance of this initial startup notification. Written notification of the actual date of initial startup, in accordance with the requirement listed in 40 CFR 61.09(a)(2) will be provided at a later date. This NOC covers the activities associated with the Construction and operation activities involving the magnesium hydroxide precipitation process of plutonium solutions within the Plutonium Finishing Plant (PFP)

  9. Safety of Nuclear Power Plants: Commissioning and Operation. Specific Safety Requirements (Arabic Edition)

    International Nuclear Information System (INIS)

    2017-01-01

    This publication is a revision of IAEA Safety Standards Series No. NS-R-2, Safety of Nuclear Power Plants: Operation, and has been extended to cover the commissioning stage. It describes the requirements to be met to ensure the safe commissioning, operation, and transition from operation to decommissioning of nuclear power plants. Over recent years there have been developments in areas such as long term operation of nuclear power plants, plant ageing, periodic safety review, probabilistic safety analysis review and risk informed decision making processes. It became necessary to revise the IAEAโ€™s Safety Requirements in these areas and to correct and/or improve the publication on the basis of feedback from its application by both the IAEA and its Member States. In addition, the requirements are governed by, and must apply, the safety objective and safety principles that are established in the IAEA Safety Standards Series No. SF-1, Fundamental Safety Principles. A review of Safety Requirements publications, initiated in 2011 following the accident in the Fukushima Daiichi nuclear power plant in Japan, revealed no significant areas of weakness but resulted in a small set of amendments to strengthen the requirements and facilitate their implementation. These are contained in the present publication.

  10. Safety standards and safety record of nuclear power plants

    International Nuclear Information System (INIS)

    Davis, A.B.

    1984-01-01

    This paper focuses on the use of standards and the measurement and enforcement of these standards to achieve safe operation of nuclear power plants. Since a discussion of the safety standards that the Nuclear Regulatory Commission (NRC) uses to regulate the nuclear power industry can be a rather tedious subject, this discussion will provide you with not only a description of what safety standards are, but some examples of their application, and various indicators that provide an overall perspective on safety. These remarks are confined to the safety standards adopted by the NRC. There are other agencies such as the Environmental Protection Agency, the Occupational Safety and Health Administration, and the state regulatory agencies which impact on a nuclear power plant. The NRC has regulatory authority for the commercial use of the nuclear materials and facilities which are defined in the Atomic Energy Act of 1954 to assure that the public health and safety and national security are protected

  11. Sludge stabilization at the Plutonium Finishing Plant, Hanford Site, Richland, Washington

    International Nuclear Information System (INIS)

    1994-10-01

    This Environmental Assessment evaluates the proposed action to operate two laboratory-size muffle furnaces in glovebox HC-21C, located in the Plutonium Finishing Plant (PFP), Hanford Site, Richland, Washington. The muffle furnaces would be used to stabilize chemically reactive sludges that contain approximately 25 kilograms (55 pounds) of plutonium by heating to approximately 500 to 1000 degrees C (900 to 1800 degrees F). The resulting stable powder, mostly plutonium oxide with impurities, would be stored in the PFP vaults. The presence of chemically reactive plutonium-bearing sludges in the process gloveboxes poses a risk to workers from radiation exposure and limits the availability of storage space for future plant cleanup. Therefore, there is a need to stabilize the material into a form suitable for long-term storage. This proposed action would be an interim action, which would take place prior to completion of an Environmental Impact Statement for the PFP which would evaluate stabilization of all plutonium-bearing materials and cleanout of the facility. However, only 10 percent of the total quantity of plutonium in reactive materials is in the sludges, so this action will not limit the choice of reasonable alternatives or prejudice the Record of Decision of the Plutonium Finishing Plant Environmental Impact Statement

  12. Instrumentation and control systems important to safety in nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2005-01-01

    This Safety Guide was prepared under the IAEA programme for establishing safety standards for nuclear power plants. It supplements Safety Standards Series No. NS-R-1: Safety of Nuclear Power Plants: Design (the Requirements for Design), which establishes the design requirements for ensuring the safety of nuclear power plants. This Safety Guide describes how the requirements should be met for instrumentation and control (I and C) systems important to safety. This publication is a revision and combination of two previous Safety Guides: Safety Series Nos 50-SG-D3 and 50-SG-D8, which are superseded by this new Safety Guide. The revision takes account of developments in I and C systems important to safety since the earlier Safety Guides were published in 1980 and 1984, respectively. The objective of this Safety Guide is to provide guidance on the design of I and C systems important to safety in nuclear power plants, including all I and C components, from the sensors allocated to the mechanical systems to the actuated equipment, operator interfaces and auxiliary equipment. This Safety Guide deals mainly with design requirements for those I and C systems that are important to safety. It expands on paragraphs of Ref in the area of I and C systems important to safety. This publication is intended for use primarily by designers of nuclear power plants and also by owners and/or operators and regulators of nuclear power plants. This Safety Guide provides general guidance on I and C systems important to safety which is broadly applicable to many nuclear power plants. More detailed requirements and limitations for safe operation specific to a particular plant type should be established as part of the design process. The present guidance is focused on the design principles for systems important to safety that warrant particular attention, and should be applied to both the design of new I and C systems and the modernization of existing systems. Guidance is provided on how design

  13. Nuclear power plant's safety and risk

    International Nuclear Information System (INIS)

    Franzen, L.F.

    1975-01-01

    Starting with a comprehensive safety strategy as evolved over the past years and the present legal provisions for the construction and operation of nuclear power plants, the risk of the intended operation, of accidents and unforeseen events is discussed. Owing to the excellent safety record of nuclear power plants, main emphasis in discussing accidents is given to the precautionary analysis within the framework of the licensing procedure. In this context, hypothetical accidents are mentioned only as having been utilized for general risk comparisons. The development of a comprehensive risk concept for a completely objective safety assessment of nuclear power plants remains as a final goal. (orig.) [de

  14. Empirical analysis of selected nuclear power plant maintenance factors and plant safety

    International Nuclear Information System (INIS)

    Olson, J.; Osborn, R.N.; Thurber, J.A.; Sommers, P.E.; Jackson, D.H.

    1985-07-01

    This report contains a statistical analysis of the relationship between selected aspects of nuclear power plant maintenance programs and safety related performance. The report identifies a large number of maintenance resources which can be expected to influence maintenance performance and subsequent plant safety performance. The resources for which data were readily available were related statistically to two sets of performance indicators: maintenance intermediate safety indicators and final safety performance indicators. The results show that the administrative structure of the plant maintenance program is a significant predictor of performance on both sets of indicators

  15. Safety assessment, safety performance indicators at the Paks Nuclear Power Plant

    International Nuclear Information System (INIS)

    Baji, C.; Vamos, G.; Toth, J.

    2001-01-01

    The Paks Nuclear Power Plant has been using different methods of safety assessment (event analysis, self-assessment, probabilistic safety analysis), including performance indicators characterizing both operational and safety performance since the early years of operation of the plant. Regarding the safety performance, the indicators include safety system performance, number of scrams, release of radioactive materials, number of safety significant events, industrial safety indicator, etc. The Paks NPP also reports a set of ten indicators to WANO Performance Indicator Programme which, among others, include safety related indicators as well. However, a more systematic approach to structuring and trending safety indicators is needed so that they can contribute to the enhancement of the operational safety. A more comprehensive set of indicators and a systematic evaluation process was introduced in 1996. The performance indicators framework proposed by the IAEA was adapted to Paks in this year to further improve the process. Safety culture assessment and characterizing safety culture is part of the assessment process. (author)

  16. Addendum 1 to CSER 96-025: PFP storage of 9.25/9.5 inch tall, 4.4 kg Pu cans on existing Vault 4 pedestals

    International Nuclear Information System (INIS)

    Hillesland, K.E.

    1997-01-01

    A nuclear criticality safety analysis has been performed to increase the approved plutonium mass limit for cans stored in Vault number-sign 4 cubicles at PFP. The original CSER 96-025 accommodated the storage of 4.4 kg of plutonium in PuO, (5.0 kg PuO,) in Vault number-sign 4 by requiring that half the cubicles be left vacant. This addendum allows for all the cubicles to be used, but with a fissile plutonium mass limit of 58 kg per cubicle. A mass limit for each cubical allows for storage of a larger number of cans if some have less than the 4.4 kg Pu limit per can. The highest k., calculated is 0.932 + 0.003 when an overbatched can is present in every fourth cubicle. This is below the criticality safety limit of kff 0.935, and consequently, an increase of plutonium mass to 4.4 kg per can is within acceptable safety limits for the given mass limit

  17. Safety culture in nuclear power plants. Proceedings

    International Nuclear Information System (INIS)

    1994-12-01

    As a consequence of the INSAG-4 report on 'safety culture', published by the IAEA in 1991, the Federal Commission for the Safety of Nuclear Power Plants (KSA) decided to hold a one-day seminar as a first step in this field. The KSA is an advisory body of the Federal Government and the Federal Department of Transport and Energy (EVED). It comments on applications for licenses, observes the operation of nuclear power plants, assists with the preparation of regulations, monitors the progress of research in the field of nuclear safety, and makes proposals for research tasks. The objective of this seminar was to familiarise the participants with the principles of 'safety culture', with the experiences made in Switzerland and abroad with existing concepts, as well as to eliminate existing prejudices. The main points dealt with at this seminar were: - safety culture from the point of view of operators, - safety culture from the point of view of the authorities, - safety culture: collaboration between power plants, the authorities and research organisations, - trends and developments in the field of safety culture. Invitations to attend this seminar were extended to the management boards of companies operating Swiss nuclear power plants, and to representatives of the Swiss authorities responsible for the safety of nuclear power plants. All these organisations were represented by a large number of executive and specialist staff. We would like to express our sincerest thanks to the Head of the Federal Department of Transport and Energy for his kind patronage of this seminar. (author) figs., tabs., refs

  18. Safety provisions of nuclear power plants

    International Nuclear Information System (INIS)

    Niehaus, F.

    1994-01-01

    Safety of nuclear power plants is determined by a deterministic approach complemented by probabilistic considerations. Much use has been made of the wealth of information from more than 6000 years of reactor operation. Design, construction and operation is governed by national and international safety standards and practices. The IAEA has prepared a set of Nuclear Safety Standards as recommendations to its Member States, covering the areas of siting, design, operations, quality assurance, and governmental organisations. In 1988 the IAEA published a report by the International Nuclear Safety Advisory Group on Basic Safety Principles for Nuclear Power Plants, summarizing the underlying objectives and principles of excellence in nuclear safety and the way in which its aspects are interrelated. The paper will summarize some of the key safety principles and provisions, and results and uses of Probabilistic Safety Assessments. Some comments will be made on the safety of WWER 440/230 and WWER-1000 reactors which are operated on Bulgaria. 8 figs

  19. Alternative off-site power supply improves nuclear power plant safety

    International Nuclear Information System (INIS)

    Gjorgiev, Blaลพe; Volkanovski, Andrija; Kanฤev, Duลกko; ฤŒepin, Marko

    2014-01-01

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

  20. Artificial intelligence and distance learning philosophy in support of PfP mandate

    OpenAIRE

    Antoliลก, Krunoslav

    2003-01-01

    Computers have long been utilised in the legal environment. The main use of computers however, has merely been to automate office tasks. More exciting is the prospect of using artificial intelligence (AI) technology to create computers that can emulate the substantive legal jobs performed by lawyers, to create computers that can autonomously reason with the law to determine legal solutions, for example: structuring and support of Partnership for Peace (PfP) mandate. Such attempts have not bee...

  1. Nuclear power plant's safety and risk (requirements of safety and reliability)

    International Nuclear Information System (INIS)

    Franzen, L.F.

    1977-01-01

    Starting out from the given safety objectives as they have evolved during the past few years and from the present legal and regulatory provisions for the construction and operation of nuclear power plants, the hazards involved in regular operation, accidents and emergency situations are discussed. In compliance with the positive safety balance of nuclear power plants in the FRG, special attention is focused on the preventive safety analysis within the frame of the nuclear licensing procedure. Reference is made to the beginnings of a comprehensive hazard concept for an unbiased plant assessment. Emergency situations are discussed from the point of view of general hazard comparisons. (orig.) [de

  2. Systematic safety evaluation of old nuclear power plants

    International Nuclear Information System (INIS)

    Dredemis, G.; Fourest, B.

    1984-01-01

    The French safety authorities have undertaken a systematic evaluation of the safety of old nuclear power plants. Apart from a complete revision of safety documents (safety analysis report, general operating rules, incident and accident procedures, internal emergency plan, quality organisation manual), this examination consisted of analysing the operating experience of systems frequently challenged and a systematic examination of the safety-related systems. This paper is based on an exercise at the Ardennes Nuclear Power Plant which has been in operation for 15 years. This paper also summarizes the main surveys and modifications relating to this power plant. (orig.)

  3. Reprocessing plants safety

    International Nuclear Information System (INIS)

    Davies, A.G.; Leighton, C.; Millington, D.

    1989-01-01

    The reprocessing of irradiated nuclear fuel at British Nuclear Fuels (BNFL) Sellafield site consists of a number of relatively self-contained activities carried out in separate plants across the site. The physical conditions and time scales applied in reprocessing and storage make it relatively benign. The potential for minor releases of radioactivity under fault conditioning is minimised by plant design definition of control procedures, training and supervision. The risks to both the general public and workforce are shown to be low with all the safety criteria being met. Normal operating conditions also have the potential for some occupational radiation exposure and the plant and workers are monitored continuously. Exposure levels have been reduced steadily and will continue to fall with plant improvements. (U.K.)

  4. Deterministic Safety Analysis for Nuclear Power Plants. Specific Safety Guide (Russian Edition)

    International Nuclear Information System (INIS)

    2014-01-01

    The objective of this Safety Guide is to provide harmonized guidance to designers, operators, regulators and providers of technical support on deterministic safety analysis for nuclear power plants. It provides information on the utilization of the results of such analysis for safety and reliability improvements. The Safety Guide addresses conservative, best estimate and uncertainty evaluation approaches to deterministic safety analysis and is applicable to current and future designs. Contents: 1. Introduction; 2. Grouping of initiating events and associated transients relating to plant states; 3. Deterministic safety analysis and acceptance criteria; 4. Conservative deterministic safety analysis; 5. Best estimate plus uncertainty analysis; 6. Verification and validation of computer codes; 7. Relation of deterministic safety analysis to engineering aspects of safety and probabilistic safety analysis; 8. Application of deterministic safety analysis; 9. Source term evaluation for operational states and accident conditions; References

  5. Selection and verification of safety parameters in safety parameter display system for nuclear power plants

    International Nuclear Information System (INIS)

    Zhang Yuangfang

    1992-02-01

    The method and results for safety parameter selection and its verification in safety parameter display system of nuclear power plants are introduced. According to safety analysis, the overall safety is divided into six critical safety functions, and a certain amount of safety parameters which can represent the integrity degree of each function and the causes of change are strictly selected. The verification of safety parameter selection is carried out from the view of applying the plant emergency procedures and in the accident man oeuvres on a full scale nuclear power plant simulator

  6. DEACTIVATION AND DECOMMISSIONING ENVIRONMENTAL STRATEGY FOR THE PLUTONIUM FINISHING PLANT COMPLEX, HANFORD NUCLEAR RESERVATION

    International Nuclear Information System (INIS)

    Hopkins, A.M.; Heineman, R.; Norton, S.; Miller, M.; Oates, L.

    2003-01-01

    Maintaining compliance with environmental regulatory requirements is a significant priority in successful completion of the Plutonium Finishing Plant (PFP) Nuclear Material Stabilization (NMS) Project. To ensure regulatory compliance throughout the deactivation and decommissioning of the PFP complex, an environmental regulatory strategy was developed. The overall goal of this strategy is to comply with all applicable environmental laws and regulations and/or compliance agreements during PFP stabilization, deactivation, and eventual dismantlement. Significant environmental drivers for the PFP Nuclear Material Stabilization Project include the Tri-Party Agreement; the Resource Conservation and Recovery Act of 1976 (RCRA); the Comprehensive Environmental Response, Compensation and Liability Act of 1980 (CERCLA); the National Environmental Policy Act of 1969 (NEPA); the National Historic Preservation Act (NHPA); the Clean Air Act (CAA), and the Clean Water Act (CWA). Recent TPA negotiation s with Ecology and EPA have resulted in milestones that support the use of CERCLA as the primary statutory framework for decommissioning PFP. Milestones have been negotiated to support the preparation of Engineering Evaluations/Cost Analyses for decommissioning major PFP buildings. Specifically, CERCLA EE/CA(s) are anticipated for the following scopes of work: Settling Tank 241-Z-361, the 232-Z Incinerator, , the process facilities (eg, 234-5Z, 242, 236) and the process facility support buildings. These CERCLA EE/CA(s) are for the purpose of analyzing the appropriateness of the slab-on-grade endpoint Additionally, agreement was reached on performing an evaluation of actions necessary to address below-grade structures or other structures remaining after completion of the decommissioning of PFP. Remaining CERCLA actions will be integrated with other Central Plateau activities at the Hanford site

  7. The operating organization for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2001-01-01

    This Safety Guide was prepared under the IAEA programme for safety standards for nuclear power plants. The present publication is a revision of the IAEA Safety Guide on Management of Nuclear Power Plants for Safe Operation issued in 1984. It supplements Section 2 of the Safety Requirements publication on Safety of Nuclear Power Plants: Operation. Nuclear power technology is different from the customary technology of power generation from fossil fuel and by hydroelectric means. One major difference between the management of nuclear power plants and that of conventional generating plants is the emphasis that should be placed on nuclear safety, quality assurance, the management of radioactive waste and radiological protection, and the accompanying national regulatory requirements. This Safety Guide highlights the important elements of effective management in relation to these aspects of safety. The attention to be paid to safety requires that the management recognize that personnel involved in the nuclear power programme should understand, respond effectively to, and continuously search for ways to enhance safety in the light of any additional requirements socially and legally demanded of nuclear energy. This will help to ensure that safety policies that result in the safe operation of nuclear power plants are implemented and that margins of safety are always maintained. The structure of the organization, management standards and administrative controls should be such that there is a high degree of assurance that safety policies and decisions are implemented, safety is continuously enhanced and a strong safety culture is promoted and supported. The objective of this publication is to guide Member States in setting up an operating organization which facilitates the safe operation of nuclear power plants to a high level internationally. The second objective is to provide guidance on the most important organizational elements in order to contribute to a strong safety

  8. The operating organization for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2005-01-01

    This Safety Guide was prepared under the IAEA programme for safety standards for nuclear power plants. The present publication is a revision of the IAEA Safety Guide on Management of Nuclear Power Plants for Safe Operation issued in 1984. It supplements Section 2 of the Safety Requirements publication on Safety of Nuclear Power Plants: Operation. Nuclear power technology is different from the customary technology of power generation from fossil fuel and by hydroelectric means. One major difference between the management of nuclear power plants and that of conventional generating plants is the emphasis that should be placed on nuclear safety, quality assurance, the management of radioactive waste and radiological protection, and the accompanying national regulatory requirements. This Safety Guide highlights the important elements of effective management in relation to these aspects of safety. The attention to be paid to safety requires that the management recognize that personnel involved in the nuclear power programme should understand, respond effectively to, and continuously search for ways to enhance safety in the light of any additional requirements socially and legally demanded of nuclear energy. This will help to ensure that safety policies that result in the safe operation of nuclear power plants are implemented and that margins of safety are always maintained. The structure of the organization, management standards and administrative controls should be such that there is a high degree of assurance that safety policies and decisions are implemented, safety is continuously enhanced and a strong safety culture is promoted and supported. The objective of this publication is to guide Member States in setting up an operating organization which facilitates the safe operation of nuclear power plants to a high level internationally. The second objective is to provide guidance on the most important organizational elements in order to contribute to a strong safety

  9. The operating organization for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2004-01-01

    This Safety Guide was prepared under the IAEA programme for safety standards for nuclear power plants. The present publication is a revision of the IAEA Safety Guide on Management of Nuclear Power Plants for Safe Operation issued in 1984. It supplements Section 2 of the Safety Requirements publication on Safety of Nuclear Power Plants: Operation. Nuclear power technology is different from the customary technology of power generation from fossil fuel and by hydroelectric means. One major difference between the management of nuclear power plants and that of conventional generating plants is the emphasis that should be placed on nuclear safety, quality assurance, the management of radioactive waste and radiological protection, and the accompanying national regulatory requirements. This Safety Guide highlights the important elements of effective management in relation to these aspects of safety. The attention to be paid to safety requires that the management recognize that personnel involved in the nuclear power programme should understand, respond effectively to, and continuously search for ways to enhance safety in the light of any additional requirements socially and legally demanded of nuclear energy. This will help to ensure that safety policies that result in the safe operation of nuclear power plants are implemented and that margins of safety are always maintained. The structure of the organization, management standards and administrative controls should be such that there is a high degree of assurance that safety policies and decisions are implemented, safety is continuously enhanced and a strong safety culture is promoted and supported. The objective of this publication is to guide Member States in setting up an operating organization which facilitates the safe operation of nuclear power plants to a high level internationally. The second objective is to provide guidance on the most important organizational elements in order to contribute to a strong safety

  10. Nuclear power plants: a unique challenge to fire safety

    International Nuclear Information System (INIS)

    Nowlen, S.P.

    1992-01-01

    The evaluation of fire safety in a nuclear power plant must include the consideration of the impact of a fire on the operability of plant safety equipment and systems. This issue is not typical of the life safety and property protection issues which dominate traditional fire safety concerns. This paper provides a general discussion of the issue of nuclear power plant fire safety as it currently exists in the USA. Included is a discussion of the past history of nuclear power plant fire events, the development of nuclear industry specific fire safety guidelines, the adverse experience associated with the inadvertent operation of fire suppression systems, and the anticipated direction of fire safety requirements for future reactor designs in the USA. (Author)

  11. Effort on Nuclear Power Plants safety

    International Nuclear Information System (INIS)

    Prayoto.

    1979-01-01

    Prospects of nuclear power plant on designing, building and operation covering natural safety, technical safety, and emergency safety are discussed. Several problems and their solutions and nuclear energy operation in developing countries especially control and permission are also discussed. (author tr.)

  12. Assessment of IAEA safety series no. 75-INSAG-3 - ''basic safety principles for nuclear power plants''

    International Nuclear Information System (INIS)

    1989-01-01

    The International Atomic Energy Agency Safety Series No. 75-INSAG--3, 'Basic Safety Principles for Nuclear Power Plants' is reviewed in the light of the Advisory Committee on Nuclear Safety reports ACNS--2, 'Safety Objectives for Nuclear Activities in Canada', and ACNS--4, 'Recommended General Safety Requirements for Nuclear Power Plants'. The INSAG safety objectives are consistent with the safety objectives stated in ACNS--2 but are less general, applying only to nuclear power plants. The INSAG safety principles are, in general, consistent with the requirements stated in ACNS--4 but put more emphasis on 'safety culture'. They give little attention to reactor plant effluents, waste management, or decommissioning. (fig., 5 refs.)

  13. Safety culture in the maintenance of nuclear power plants

    International Nuclear Information System (INIS)

    2005-01-01

    Safety culture is the complexity of beliefs, shared values and behaviour reflected in making decisions and performing work in a nuclear power plant or nuclear facility. The definition of safety culture and the related concepts presented in the IAEA literature are widely known to experts. Since the publication of Safety Culture, issued by the IAEA as INSAG-4 in 1991, the IAEA has produced a number of publications on strengthening the safety culture in organizations that operate nuclear power plants and nuclear facilities. However, until now the focus has been primarily on the area of operations. Apart from operations, maintenance in plants and nuclear facilities is an aspect that deserves special attention, as maintenance activities can have both a direct and an indirect effect on equipment reliability. Adverse safety effects can arise, depending upon the level of skill of the personnel involved, safety awareness and the complexity of the work process. Any delayed effects resulting from challenges to maintenance can cause interruptions in operation, and hence affect the safety of a plant or facility. Building upon earlier IAEA publications on this topic, this Safety Report reviews how challenges to the maintenance of nuclear power plants can affect safety culture. It also highlights indications of a weakening safety culture. The challenges described are in areas such as maintenance management; human resources management; plant condition assessment and the business environment. The steps that some Member States have taken to address safety culture aspects are detailed and singled out as good practices, with a view to disseminating and exchanging experiences and lessons learned. Although this report is primarily directed at plant maintenance organizations, the subject matter is applicable to a wider audience, including plant contracting organizations and regulatory authorities

  14. Properties of the malarial proteins Pf2, Pf9 and PfP0, which support ...

    Indian Academy of Sciences (India)

    Properties of the malarial proteins Pf2, Pf9 and PfP0, which support their roles as immune targets. Antibodies raised to each of these proteins (or purified from immune adults) inhibit the growth of Plasmodium falciparum at the red cell invasion step. The proteins are localized on the parasite cell surface. Each protein isย ...

  15. Safety classification of items in Tianwan Nuclear Power Plant

    International Nuclear Information System (INIS)

    Sun Yongbin

    2005-01-01

    The principle of integrality, moderation and equilibrium should be considered in the safety classification of items in nuclear power plant. The basic ways for safety classification of items is to classify the safety function based on the effect of the outside enclosure damage of the items (parts) on the safety. Tianwan Nuclear Power Plant adopts Russian VVER-1000/428 type reactor, it safety classification mainly refers to Russian Guidelines and standards. The safety classification of the electric equipment refers to IEEE-308(80) standard, including 1E and Non 1E classification. The safety classification of the instrumentation and control equipment refers to GB/T 15474-1995 standard, including safety 1E, safety-related SR and NC non-safety classification. The safety classification of Tianwan Nuclear Power Plant has to be approved by NNSA and satisfy Chinese Nuclear Safety Guidelines. (authors)

  16. Criticality safety evaluation in Tokai Reprocessing Plant

    International Nuclear Information System (INIS)

    Shirai, Nobutoshi; Nakajima, Masayoshi; Takaya, Akikazu; Ohnuma, Hideyuki; Shirouzu, Hidetomo; Hayashi, Shinichiro; Yoshikawa, Koji; Suto, Toshiyuki

    2000-04-01

    Criticality limits for equipments in Tokai Reprocessing Plant which handle fissile material solution and are under shape and dimension control were reevaluated based on the guideline No.10 'Criticality safety of single unit' in the regulatory guide for reprocessing plant safety. This report presents criticality safety evaluation of each equipment as single unit. Criticality safety of multiple units in a cell or a room was also evaluated. The evaluated equipments were ones in dissolution, separation, purification, denitration, Pu product storage, and Pu conversion processes. As a result, it was reconfirmed that the equipments were safe enough from a view point of criticality safety of single unit and multiple units. (author)

  17. Safety in waste management plants: An Indian perspective

    International Nuclear Information System (INIS)

    Shekhar, P.; Ozarde, P.D.; Gandhi, P.M.

    2000-01-01

    Assurance of safety of public and plant workers and protection of the environment are prime objectives in the design and construction of Waste Management Plants. In India, waste management principles and strategies have been evolved in accordance with national and international regulations and standards for radiation protection. The regulations governing radiation protection have a far-reaching impact on the management of the radioactive waste. The wastes arise at each stages of the fuel cycle with varying chemical nature, generation rate and specific activity levels depending upon the type of the facility. Segregation of waste based on its chemical nature and specific activity levels is an essential feature, as its aids in selection of treatment and conditioning process. Selection of the process, equipment and materials in the plant, are governed by safety consideration alongside factors like efficiency and simplicity. The plant design considerations like physical separation, general arrangement, ventilation zoning, access control, remote handling, process piping routing, decontamination etc. have major role in realizing waste safety. Stringent quality control measures during all stages of construction have helped in achieving the design intended safety. These aspects together with operating experience gained form basis for the improved safety features in the design and construction of waste management plants. The comprehensive safety is derived from adoption of waste management strategies and appropriate plant design considerations. The paper briefly brings safety in waste management programme in India, in its current perspective. (author)

  18. Safety upgrading at PAKS Nuclear Power Plant

    International Nuclear Information System (INIS)

    Bajsz, J.; Elter, J.

    2000-01-01

    The operation of Paks NPP has reached its half time. Until this time the plant fulfilled expectations raised before its construction: the four units have produced safely and reliably more than 200 TWh electricity. The production of the plant has been at the stable level since its construction and has provided 43-38 % of electricity consumed in Hungary. The annual production is around 14 TWh, which means a load factor higher than 85 %. Safety upgrading activities [1] at Paks had started in the late eighties, when the commissioning work of units 3 and 4 were carried out. That time the main emphases were put to lessons learned of the TMI and Chernobyl accidents. The international reviews hosted by our plant widened our review's scope. To systematize our approach a complete safety review, the AGNES (Advanced General Safety and New Evaluation of Safety) project was started in 1991. The goal of the project was to evaluate to what extent Paks NPP satisfied the current international safety expectations and to help in determining the priorities for safety enhancement and upgrading measures. The project completed in 1994 ranked our safety upgrading measures by safety significance, which became a basis for technical design work and financial scheduling. The other important outcome of the AGNES project was the introduction the Periodical Safety Review regime by our nuclear authority. These periodical reviews held after 10 years of operation offer the possibility - and obligation for the licensee - to perform a comprehensive assessment of the safety of the plant, to evaluate the integral effects of changes of circumstances happened during the review period. The goal of these reviews is to deal with cumulative effects of NPP ageing, modifications, operating experience and technical developments aimed at ensuring a high level of safety throughout plant service life. The execution of our safety-upgrading program is well advancing. For the whole program from 1996 to 2002 250

  19. Plutonium finishing plant safety systems and equipment list

    International Nuclear Information System (INIS)

    Bergquist, G.G.

    1995-01-01

    The Safety Equipment List (SEL) supports Analysis Report (FSAR), WHC-SD-CP-SAR-021 and the Plutonium Finishing Plant Operational Safety Requirements (OSRs), WHC-SD-CP-OSR-010. The SEL is a breakdown and classification of all Safety Class 1, 2, and 3 equipment, components, or system at the Plutonium Finishing Plant complex

  20. Periodic Safety Review of Nuclear Power Plants: Experience of Member States

    International Nuclear Information System (INIS)

    2010-04-01

    Routine reviews of nuclear power plant operation (including modifications to hardware and procedures, operating experience, plant management and personnel competence) and special reviews following major events of safety significance are the primary means of safety verification. In addition, many Member States of the IAEA have initiated systematic safety reassessments, termed periodic safety reviews, of nuclear power plants, to assess the cumulative effects of plant ageing and plant modifications, operating experience, technical developments and siting aspects. The reviews include an assessment of plant design and operation against current safety standards and practices, and they have the objective of ensuring a high level of safety throughout the plant's operating lifetime. They are complementary to the routine and special safety reviews and do not replace them. Periodic safety reviews of nuclear power plants are considered an effective way to obtain an overall view of actual plant safety, and to determine reasonable and practical modifications that should be made in order to maintain a high level of safety. They can be used as a means of identifying time limiting features of the plant in order to determine nuclear power plant operation beyond the designed lifetime. The periodic safety review process can be used to support the decision making process for long term operation or licence renewal. Since 1994, the use of periodic safety reviews by Member States has substantially broadened and confirmed its benefits. Periodic safety review results have, for example, been used by some Member States to help provide a basis for continued operation beyond the current licence term, to communicate more effectively with stakeholders regarding nuclear power plant safety, and to help identify changes to plant operation that enhance safety. This IAEA-TECDOC is intended to assist Member States in the implementation of a periodic safety review. This publication complements the

  1. Improving plant state information for better operational safety

    International Nuclear Information System (INIS)

    Girard, C.; Olivier, E.; Grimaldi, X.

    1994-01-01

    Nuclear Power Plant (NPP) safety is strongly dependent on components' reliability and particularly on plant state information reliability. This information, used by the plant operators in order to produce appropriate actions, have to be of a high degree of confidence, especially in accidental conditions where safety is threatened. In this perspective, FRAMATOME, EDF and CEA have started a joint research program to prospect different solutions aiming at a better reliability for critical information needed to safety operate the plant. This paper gives the main results of this program and describes the developments that have been made in order to assess reliability of different information systems used in a Nuclear Power Plant. (Author)

  2. Safety analysis of nuclear power plants

    International Nuclear Information System (INIS)

    Selvatici, E.

    1981-01-01

    A study about the safety analysis of nuclear power plant, giving emphasis to how and why to do is presented. The utilization of the safety analysis aiming to perform the licensing requirements is discussed, and an example of the Angra 2 and 3 safety analysis is shown. Some presented tendency of the safety analysis are presented and examples are shown.(E.G.) [pt

  3. Operating plant safety analysis needs

    International Nuclear Information System (INIS)

    Young, M.Y.; Love, D.S.

    1992-01-01

    The primary objective for nuclear power station owners is to operate and manage their plants safely. However, there is also a need to provide economical electric power, which requires that the unit be operated as efficiently as possible, consistent with the safety requirements. The objectives cited above can be achieved through the identification and use of available margins inherent in the plant design. As a result of conservative licensing and analytical approaches taken in the past, many of these margins may be found in the safety analysis limits within which plants currently operate. Improvements in the accuracy of the safety analysis, and a more realistic treatment of plant initial and boundary conditions, can make this margin available for a variety of uses which enhance plant performance, help to reduce O and M costs, and may help to extend licensed operation. Opportunities for improvement exist in several areas in the accident analysis normally performed for Chapter 15 of the FSAR. For example, recent modifications to the ECCS rule, 10CFR50.46 and Appendix K, allow use of margins previously unavailable in the analysis of the Loss of Coolant Accident (LOCA). To take advantage of this regulatory change, new methods are being developed to analyze both the large and small break loss of coolant accident (LOCA). As this margin is used, enhancements in the analysis of other transients will become necessary. The paper discusses accident analysis methods, future development needs, and analysis margin utilization in specific accident scenarios

  4. Dukovany nuclear power plant safety

    International Nuclear Information System (INIS)

    1999-01-01

    Presentation covers recommended safety issues for the Dukovany NPP which have been solved with satisfactory conclusions. Safety issues concerned include: radiation safety; nuclear safety; security; emergency preparedness; health protection at work; fire protection; environmental protection; chemical safety; technical safety. Quality assurance programs at all stages on NPP life time is described. Report includes description of NPP staff training provision, training simulator, emergency operating procedures, emergency preparedness, Year 2000 problem, inspections and life time management. Description of Dukovany Plant Safety Analysis Projects including integrity of the equipment, modernisation, equipment innovation and safety upgrading program show that this approach corresponds to the actual practice applied in EU countries, and fulfilment of current IAEA requirements for safety enhancement of the WWER 440/213 units in the course of MORAWA Equipment Upgrading program

  5. Safety aspects of nuclear power plant ageing

    International Nuclear Information System (INIS)

    1990-01-01

    The nuclear community is facing new challenges as commercial nuclear power plants (NPPs) of the first generation get older. At present, some of the plants are approaching or have even exceeded the end of their nominal design life. Experience with fossil fired power plants and in other industries shows that reliability of NPP components, and consequently general plant safety and reliability, may decline in the middle and later years of plant life. Thus, the task of maintaining operational safety and reliability during the entire plant life and especially, in its later years, is of growing importance. Recognizing the potential impact of ageing on plant safety, the IAEA convened a Working Group in 1985 to draft a report to stimulate relevant activities in the Member States. This report provided the basis for the preparation of the present document, which included a review in 1986 by a Technical Committee and the incorporation of relevant results presented at the 1987 IAEA Symposium on the Safety Aspects of the Ageing and Maintenance of NPPs and in available literature. The purpose of the present document is to increase awareness and understanding of the potential impact of ageing on plant safety; of ageing processes; and of the approach and actions needed to manage the ageing of NPP components effectively. Despite the continuing growth in knowledge on the subject during the preparation of this report it nevertheless contains much that will be of interest to a wide technical and managerial audience. Furthermore, more specific technical publications on the evaluation and management of NPP ageing and service life are being developed under the Agency's programme, which is based on the recommendations of its 1988 Advisory Group on NPP ageing. Refs, figs and tabs

  6. Safety of Nuclear Power Plants: Commissioning and Operation (Spanish Edition)

    International Nuclear Information System (INIS)

    2012-01-01

    This publication is a revision of Safety Requirements No. NS-R-2, Safety of Nuclear Power Plants: Operation, and has been extended to cover the commissioning stage. It describes the requirements to be met to ensure the safe operation of nuclear power plants. Over recent years there have been developments in areas such as long term operation, plant ageing, periodic safety review, probabilistic safety analysis and risk informed decision making processes. It became necessary to revise the IAEA's safety requirements in these areas and to correct and/or improve the publication on the basis of feedback from its application by both the IAEA and its Member States. In addition, the requirements are governed by, and must apply, the safety objective and safety principles that are established in the Fundamental Safety Principles. Contents: 1. Introduction; 2. Safety objectives and principles; 3. The management and organizational structure of the operating organization; 4. Management of operational safety; 5. Operational safety programmes; 6. Plant commissioning; 7. Plant operations; 8. Maintenance, testing, surveillance and inspection; 9. Preparation for decommissioning.

  7. Safety of Nuclear Power Plants: Commissioning and Operation (French Edition)

    International Nuclear Information System (INIS)

    2012-01-01

    This publication is a revision of Safety Requirements No. NS-R-2, Safety of Nuclear Power Plants: Operation, and has been extended to cover the commissioning stage. It describes the requirements to be met to ensure the safe operation of nuclear power plants. Over recent years there have been developments in areas such as long term operation, plant ageing, periodic safety review, probabilistic safety analysis and risk informed decision making processes. It became necessary to revise the IAEA's safety requirements in these areas and to correct and/or improve the publication on the basis of feedback from its application by both the IAEA and its Member States. In addition, the requirements are governed by, and must apply, the safety objective and safety principles that are established in the Fundamental Safety Principles. Contents: 1. Introduction; 2. Safety objectives and principles; 3. The management and organizational structure of the operating organization; 4. Management of operational safety; 5. Operational safety programmes; 6. Plant commissioning; 7. Plant operations; 8. Maintenance, testing, surveillance and inspection; 9. Preparation for decommissioning.

  8. Safety of Nuclear Power Plants: Commissioning and Operation. Arabic Edition

    International Nuclear Information System (INIS)

    2011-01-01

    This publication is a revision of Safety Requirements No. NS-R-2, Safety of Nuclear Power Plants: Operation, and has been extended to cover the commissioning stage. It describes the requirements to be met to ensure the safe operation of nuclear power plants. Over recent years there have been developments in areas such as long term operation, plant ageing, periodic safety review, probabilistic safety analysis and risk informed decision making processes. It became necessary to revise the IAEA's safety requirements in these areas and to correct and/or improve the publication on the basis of feedback from its application by both the IAEA and its Member States. In addition, the requirements are governed by, and must apply, the safety objective and safety principles that are established in the Fundamental Safety Principles. Contents: 1. Introduction; 2. Safety objectives and principles; 3. The management and organizational structure of the operating organization; 4. Management of operational safety; 5. Operational safety programmes; 6. Plant commissioning; 7. Plant operations; 8. Maintenance, testing, surveillance and inspection; 9. Preparation for decommissioning.

  9. The critical safety functions and plant operation

    International Nuclear Information System (INIS)

    Corcoran, W.R.; Church, J.F.; Porter, N.J.; Cross, M.T.; Guinn, W.M.

    1981-01-01

    The paper outlines the operator's role in nuclear safety and introduces the concept of ''safety functions''. Safety functions are a group of actions that prevent core melt or minimize radiation releases to the general public. They can be used to provide a hierarchy of practical plant protection that an operator should use. ''An accident identical to that at Three Mile Island is not going to happen again'', said the Rogovin investigators. The concepts put forward in this paper are intended to help the operator avoid serious consequence from the next unexpected threat. On the basis of the safety evaluation, the operator has three roles in assuring that the consequences of an event will be no worse than the predicted acceptable results. These three operator roles are: first, maintain plant setup in readiness to properly respond; second, operate the plant in a manner such that fewer, milder events minimize the frequency and the severity of adverse events; third, the operator needs to monitor the plant to verify that the safety functions are accomplished. The operator needs a systematic approach to mitigating the consequences of an event. The concept of ''safety function'' introduces that systematic approach and prevents a hierarchy of protection. If the operator has difficulty in identifying an event for any reason, the systematic safety function approach allows ones to accomplish the overall path of mitigating consequences. There are ten identified functions designed to protect against core melt, preserve containment integrity, prevent indirect release of radioactivity, and maintain vital auxiliaries needed to support the other safety functions. The paper describes in detail the operator's role and the safety functions, and provides many examples of the use of alternative success paths to accomplish the safety function

  10. Safety goals for commercial nuclear power plants

    International Nuclear Information System (INIS)

    Roe, J.W.

    1988-01-01

    In its official policy statement on safety goals for the operation of nuclear power plants, the Nuclear Regulatory Commission (NRC) set two qualitative goals, supported by two quantitative objectives. These goals are that (1) individual members of the public should be provided a level of protection from the consequences of nuclear power plant operation such that individuals bear no significant additional risk to life and health; and (2) societal risks to life and health from nuclear power plant operation should be comparable to or less than the risks of generating electricity by viable competing technologies and should not be a significant addition to other societal risks. As an alternative, this study proposes four quantitative safety goals for nuclear power plants. It begins with an analysis of the NRC's safety-goal development process, a key portion of which was devoted to delineating criteria for evaluating goal-development methods. Based on this analysis, recommendations for revision of the NRC's basic benchmarks for goal development are proposed. Using the revised criteria, NRC safety goals are evaluated, and the alternative safety goals are proposed. To further support these recommendations, both the NRC's goals and the proposed goals are compared with the results of three major probabilistic risk assessment studies. Finally, the potential impact of these recommendations on nuclear safety is described

  11. Safety of Nuclear Power Plants: Design. Specific Safety Requirements (Russian Edition); Bezopasnost' atomnykh ehlektrostantsij: proektirovanie. Konkretnye trebovaniya bezopasnosti

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2012-04-15

    This publication is a revision of Safety Requirements No. NS-R-1, Safety of Nuclear Power Plants: Design. It establishes requirements applicable to the design of nuclear power plants and elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. It will be useful for organizations involved in the design, manufacture, construction, modification, maintenance, operation and decommissioning of nuclear power plants, as well as for regulatory bodies. Contents: 1. Introduction; 2. Applying the safety principles and concepts; 3. Management of safety in design; 4. Principal technical requirements; 5. General plant design; 6. Design of specific plant systems.

  12. Problems of nuclear power plant safety evaluation

    International Nuclear Information System (INIS)

    Suchomel, J.

    1977-01-01

    Nuclear power plant safety is discussed with regard to external effects on the containment and to the human factor. As for external effects, attention is focused on shock waves which may be due to explosions or accidents in flammable material transport and storage, to missiles, and to earthquake effects. The criteria for evaluating nuclear power plant safety in different countries are shown. Factors are discussed affecting the reliability of man with regard to his behaviour in a loss-of-coolant accident in the power plant. Different types of PWR containments and their functions are analyzed, mainly in case of accident. Views are discussed on the role of destructive accidents in the overall evaluation of fast reactor safety. Experiences are summed up gained with the operation of WWER reactors with respect to the environmental impact of the nuclear power plants. (Z.M.)

  13. Safety Assessment - Swedish Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Kjellstroem, B. [Luleaa Univ. of Technology (Sweden)

    1996-12-31

    After the reactor accident at Three Mile Island, the Swedish nuclear power plants were equipped with filtered venting of the containment. Several types of accidents can be identified where the filtered venting has no effect on the radioactive release. The probability for such accidents is hopefully very small. It is not possible however to estimate the probability accurately. Experiences gained in the last years, which have been documented in official reports from the Nuclear Power Inspectorate indicate that the probability for core melt accidents in Swedish reactors can be significantly larger than estimated earlier. A probability up to one in a thousand operating years can not be excluded. There are so far no indications that aging of the plants has contributed to an increased accident risk. Maintaining the safety level with aging nuclear power plants can however be expected to be increasingly difficult. It is concluded that the 12 Swedish plants remain a major threat for severe radioactive pollution of the Swedish environment despite measures taken since 1980 to improve their safety. Closing of the nuclear power plants is the only possibility to eliminate this threat. It is recommended that until this is done, quantitative safety goals, same for all Swedish plants, shall be defined and strictly enforced. It is also recommended that utilities distributing misleading information about nuclear power risks shall have their operating license withdrawn. 37 refs.

  14. Safety Assessment - Swedish Nuclear Power Plants

    International Nuclear Information System (INIS)

    Kjellstroem, B.

    1996-01-01

    After the reactor accident at Three Mile Island, the Swedish nuclear power plants were equipped with filtered venting of the containment. Several types of accidents can be identified where the filtered venting has no effect on the radioactive release. The probability for such accidents is hopefully very small. It is not possible however to estimate the probability accurately. Experiences gained in the last years, which have been documented in official reports from the Nuclear Power Inspectorate indicate that the probability for core melt accidents in Swedish reactors can be significantly larger than estimated earlier. A probability up to one in a thousand operating years can not be excluded. There are so far no indications that aging of the plants has contributed to an increased accident risk. Maintaining the safety level with aging nuclear power plants can however be expected to be increasingly difficult. It is concluded that the 12 Swedish plants remain a major threat for severe radioactive pollution of the Swedish environment despite measures taken since 1980 to improve their safety. Closing of the nuclear power plants is the only possibility to eliminate this threat. It is recommended that until this is done, quantitative safety goals, same for all Swedish plants, shall be defined and strictly enforced. It is also recommended that utilities distributing misleading information about nuclear power risks shall have their operating license withdrawn. 37 refs

  15. Safety indicators as a tool for operational safety evaluation of nuclear power plants

    International Nuclear Information System (INIS)

    Araujo, Jefferson Borges; Melo, Paulo Fernando Ferreira Frutuoso e; Schirru, Roberto

    2009-01-01

    Performance indicators have found a wide use in the conventional and nuclear industries. For the conventional industry, the goal is to optimize production, reducing loss of time with accidents, human error and equipment downtimes. In the nuclear industry, nuclear safety is an additional goal. This paper presents a general methodology to the establishment, selection and use of safety indicators for a two loop PWR plant, as Angra 1. The use of performance indicators is not new. The NRC has its own methodology and the IAEA presents methodology suggestions, but there is no detailed documentation about indicators selection, criteria and bases used. Additionally, only the NRC methodology performs a limited integrated evaluation. The study performed identifies areas considered critical for the plant operational safety. For each of these areas, strategic sub-areas are defined. For each strategic sub-area, specific safety indicators are defined. These proposed Safety Indicators are based on the contribution to risk considering a quantitative risk analysis. For each safety indicator, a goal, a bounded interval and proper bases are developed, to allow for a clear and comprehensive individual behavior evaluation. On the establishment of the intervals and boundaries, a probabilistic safety study, operational experience, international and national standards and technical specifications were used. Additionally, an integrated evaluation of the indicators, using expert systems, was done to obtain an overview of the plant general safety. This evaluation uses well-defined and clear rules and weights for each indicator to be considered. These rules were implemented by means of a computational language, on a friendly interface, so that it is possible to obtain a quick response about operational safety. This methodology can be used to identify situations where the plant safety is challenged, by giving a general overview of the plant operational condition. Additionally, this study can

  16. Seismic safety of nuclear power plants in Eastern Europe

    International Nuclear Information System (INIS)

    Gurpinar, A.; Godoy, A.

    1995-01-01

    This paper summarizes the work performed by the International Atomic Energy Agency in the areas of safety reviews and applied research in support of programmes for the assessment and enhancement of seismic safety in WWER type nuclear power plants during the past five years. Three major topics are discussed; engineering safety review services in relation to external events, technical guidelines for the assessment and upgrading of WWER type nuclear power plants, and the Coordinated Research Programme on B enchmark study for the seismic analysis and testing of WWER type nuclear power plants . These topics are summarized in a way to provide an overview of the past and present safety situation in selected WWER type plants which are all located in Eastern European countries. Main conclusion of the paper is that although there is now a thorough understanding of the seismic safety issues in these operating nuclear power plants, the implementation of seismic upgrades to structures, systems and components are lagging behind, particularly for those cases in which the re-evaluation indicated the necessity to strengthen the safety related structures or install new safety systems. (author)

  17. Nitrogen-system safety study: Portsmouth Gaseous Diffusion Plant

    International Nuclear Information System (INIS)

    1982-07-01

    The Department of Energy has primary responsibility for the safety of operations at DOE-owned nuclear facilities. The guidelines for the analysis of credible accidents are outlined in DOE Order 5481.1. DOE has requested that existing plant facilities and operations be reviewed for potential safety problems not covered by standard industrial safety procedures. This review is being conducted by investigating individual facilities and documenting the results in Safety Study Reports which will be compiled to form the Existing Plant Final Safety Analysis Report which is scheduled for completion in September, 1984. This Safety Study documents the review of the Plant Nitrogen System facilities and operations and consists of Section 4.0, Facility and Process Description, and Section 5.0, Accident Analysis, of the Final Safety Analysis Report format. The existing nitrogen system consists of a Superior Air Products Company Type D Nitrogen Plant, nitrogen storage facilities, vaporization facilities and a distribution system. The system is designed to generate and distribute nitrogen gas used in the cascade for seal feed, buffer systems, and for servicing equipment when exceptionally low dew points are required. Gaseous nitrogen is also distributed to various process auxiliary buildings. The average usage is approximately 130,000 standard cubic feet per day

  18. Different aspects of safety in Nuclear Fuel Plant at Pitesti, Romania

    International Nuclear Information System (INIS)

    Ivana, T.; Epure, Gh.

    2009-01-01

    Nuclear Fuel Plant (FCN) is a facility that produces fuel bundles of CANDU-6 type for the CANDU nuclear power plant. Only natural and depleted uranium in bulk and itemized form are present as nuclear materials in this facility. Uranium and wastes from the plant are handled, processed, treated and stored throughout the entire facility. The nuclear materials with natural and depleted uranium are entirely under nuclear safeguards. The amount of uranium present in the plant in different forms and activities together with zircaloy, beryllium and other hazardous substances, wastes, explosive materials at high temperatures, etc. lead to special measures undertaken by Nuclear Safety Department (DNS) to ensure nuclear safety. Different aspects of safety are continuously monitored in the plant: operational safety, industrial safety, radiological safety, labour safety, informational safety. The emergency preparedness and response, physical protection and the security of the plant and of the transportation of radioactive materials are contributing to cover the multitude of safety aspects. The safety culture of workers built directly on the safety components completes this activity in the plant. In addition the aspects of safety, security and safeguards are in permanent synergy, parts of the three components being included in each other. In the future the policy of FCN will be focused so that any improvement of one of the safety components will be reflected in improving the other safety aspects. (authors)

  19. Safety aspects of a fuel reprocessing plant

    International Nuclear Information System (INIS)

    Donoghue, J.K.; Charlesworth, F.R.; Fairbairn, A.

    1977-01-01

    The establishment of the basic process must include the determination of the sensitivity of the process to operational errors or plant failures. The probability, and consequences of escapes of activity must be evaluated and emergency procedures set up to deal with accidents which might lead to such escapes. The administrative arrangements for safety should include a safety evaluation and advisory service independent of line management. A quality assurance strategy for the construction and commissioning stages is important. The design and construction of the plant must include: (i) Attention to plant reliability. Maintenance and inspection procedures to maintain reliability must be adopted and the design should include measures to facilitate in-service inspection of highly-active plant. (ii) Suitable and sufficient means of detection and prevention of malfunction, including criticality, bearing in mind both the timescale of development of the fault and its consequences. (iii) Measures for containment of activity. Penetrations from active into operating areas should be eliminated or minimised and maintenance should be separated from operational areas. Secondary containment beyond that provided for operations of a significant magnitude. A ventilation system with appropriate gas clean-up, monitoring and discharge facilities is required. (iv) Adequate shielding, with particular attention paid to multiple activities in a single operational area which might lead to an operator being exposed to radiation from operations which are beyond his control. (v) Means of accounting for active materials and for their recovery, transfer and disposal in the event of a forced shut down. (vi) Suitable methods for segregation and control of wastes within the plant and for their discharge. Solid or liquid wastes should be subject to delay and monitoring procedures before release. Facilities for storage of waste must be subject to the same safety principles as the plant itself. (vii) Final

  20. Addendum 2 to CSER 79-002: Extension of the 150 gram fissile limit used in room 187 of PFP

    International Nuclear Information System (INIS)

    Friar, D.E.

    1994-01-01

    The PFP operating organization requests that the limit set permitting 150 grams fissile be extended to the Hoods 4 and 5 of Room 187. The request for the limit change is explained in the attached request for analysis

  1. Code on the safety of nuclear power plants: Design

    International Nuclear Information System (INIS)

    1988-01-01

    This Code is a compilation of nuclear safety principles aimed at defining the essential requirements necessary to ensure nuclear safety. These requirements are applicable to structures, systems and components, and procedures important to safety in nuclear power plants embodying thermal neutron reactors, with emphasis on what safety requirements shall be met rather than on specifying how these requirements can be met. It forms part of the Agency's programme for establishing Codes and Safety Guides relating to land based stationary thermal neutron power plants. The document should be used by organizations designing, manufacturing, constructing and operating nuclear power plants as well as by regulatory bodies

  2. Mathematical Safety Assessment Approaches for Thermal Power Plants

    Directory of Open Access Journals (Sweden)

    Zong-Xiao Yang

    2014-01-01

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

  3. Reviewing industrial safety in nuclear power plants

    International Nuclear Information System (INIS)

    1990-02-01

    This document contains guidance and reference materials for Operational Safety Review Team (OSART) experts, in addition to the OSART Guidelines (TECDOC-449), for use in the review of industrial safety activities at nuclear power plants. It sets out objectives for an excellent industrial safety programme, and suggests investigations which should be made in evaluating industrial safety programmes. The attributes of an excellent industrial safety programme are listed as examples for comparison. Practical hints for reviewing industrial safety are discussed, so that the necessary information can be obtained effectively through a review of documents and records, discussions with counterparts, and field observations. There are several annexes. These deal with major features of industrial safety programmes such as safety committees, reporting and investigation systems and first aid and medical facilities. They include some examples which are considered commendable. The document should be taken into account not only when reviewing management, organization and administration but also in the review of related areas, such as maintenance and operations, so that all aspects of industrial safety in an operating nuclear power plant are covered

  4. 46. The goals of safety engineering department of the plant

    International Nuclear Information System (INIS)

    Ivanov, A.V.

    1993-01-01

    The goals of safety engineering department of the plant, including elaboration of instructions on safety engineering on all specialities, safety engineering training of all labours working on the plant and control for abidance by the instructions on safety engineering were discussed.

  5. Software for computer based systems important to safety in nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2004-01-01

    Computer based systems are of increasing importance to safety in nuclear power plants as their use in both new and older plants is rapidly increasing. They are used both in safety related applications, such as some functions of the process control and monitoring systems, as well as in safety critical applications, such as reactor protection or actuation of safety features. The dependability of computer based systems important to safety is therefore of prime interest and should be ensured. With current technology, it is possible in principle to develop computer based instrumentation and control systems for systems important to safety that have the potential for improving the level of safety and reliability with sufficient dependability. However, their dependability can be predicted and demonstrated only if a systematic, fully documented and reviewable engineering process is followed. Although a number of national and international standards dealing with quality assurance for computer based systems important to safety have been or are being prepared, internationally agreed criteria for demonstrating the safety of such systems are not generally available. It is recognized that there may be other ways of providing the necessary safety demonstration than those recommended here. The basic requirements for the design of safety systems for nuclear power plants are provided in the Requirements for Design issued in the IAEA Safety Standards Series.The IAEA has issued a Technical Report to assist Member States in ensuring that computer based systems important to safety in nuclear power plants are safe and properly licensed. The report provides information on current software engineering practices and, together with relevant standards, forms a technical basis for this Safety Guide. The objective of this Safety Guide is to provide guidance on the collection of evidence and preparation of documentation to be used in the safety demonstration for the software for computer based

  6. Software for computer based systems important to safety in nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2005-01-01

    Computer based systems are of increasing importance to safety in nuclear power plants as their use in both new and older plants is rapidly increasing. They are used both in safety related applications, such as some functions of the process control and monitoring systems, as well as in safety critical applications, such as reactor protection or actuation of safety features. The dependability of computer based systems important to safety is therefore of prime interest and should be ensured. With current technology, it is possible in principle to develop computer based instrumentation and control systems for systems important to safety that have the potential for improving the level of safety and reliability with sufficient dependability. However, their dependability can be predicted and demonstrated only if a systematic, fully documented and reviewable engineering process is followed. Although a number of national and international standards dealing with quality assurance for computer based systems important to safety have been or are being prepared, internationally agreed criteria for demonstrating the safety of such systems are not generally available. It is recognized that there may be other ways of providing the necessary safety demonstration than those recommended here. The basic requirements for the design of safety systems for nuclear power plants are provided in the Requirements for Design issued in the IAEA Safety Standards Series.The IAEA has issued a Technical Report to assist Member States in ensuring that computer based systems important to safety in nuclear power plants are safe and properly licensed. The report provides information on current software engineering practices and, together with relevant standards, forms a technical basis for this Safety Guide. The objective of this Safety Guide is to provide guidance on the collection of evidence and preparation of documentation to be used in the safety demonstration for the software for computer based

  7. Software for computer based systems important to safety in nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2000-01-01

    Computer based systems are of increasing importance to safety in nuclear power plants as their use in both new and older plants is rapidly increasing. They are used both in safety related applications, such as some functions of the process control and monitoring systems, as well as in safety critical applications, such as reactor protection or actuation of safety features. The dependability of computer based systems important to safety is therefore of prime interest and should be ensured. With current technology, it is possible in principle to develop computer based instrumentation and control systems for systems important to safety that have the potential for improving the level of safety and reliability with sufficient dependability. However, their dependability can be predicted and demonstrated only if a systematic, fully documented and reviewable engineering process is followed. Although a number of national and international standards dealing with quality assurance for computer based systems important to safety have been or are being prepared, internationally agreed criteria for demonstrating the safety of such systems are not generally available. It is recognized that there may be other ways of providing the necessary safety demonstration than those recommended here. The basic requirements for the design of safety systems for nuclear power plants are provided in the Requirements for Design issued in the IAEA Safety Standards Series.The IAEA has issued a Technical Report to assist Member States in ensuring that computer based systems important to safety in nuclear power plants are safe and properly licensed. The report provides information on current software engineering practices and, together with relevant standards, forms a technical basis for this Safety Guide. The objective of this Safety Guide is to provide guidance on the collection of evidence and preparation of documentation to be used in the safety demonstration for the software for computer based

  8. Planning and evaluation of plant under safety aspects

    International Nuclear Information System (INIS)

    Strnad, H.

    1985-01-01

    Plant denotes a technical product characterized as being structured, complex, comprising the use of energy, and that of measuring, automatic control and monitoring systems to keep track of present, control and monitor processes. Particular attention is paid to methods of developing plant concepts, measures to exclude or detect risks, integration of safety engineering into the course of planning, safety concept and ergonomics in plant design. (DG) [de

  9. Plant safety review from mass criticality accident

    International Nuclear Information System (INIS)

    Susanto, B.G.

    2000-01-01

    The review has been done to understand the resent status of the plant in facing postulated mass criticality accident. From the design concept of the plant all the components in the system including functional groups have been designed based on favorable mass/geometry safety principle. The criticality safety for each component is guaranteed because all the dimensions relevant to criticality of the components are smaller than dimensions of 'favorable mass/geometry'. The procedures covering all aspects affecting quality including the safety related are developed and adhered to at all times. Staff are indoctrinated periodically in short training session to warn the important of the safety in process of production. The plant is fully equipped with 6 (six) criticality detectors in strategic places to alert employees whenever the postulated mass criticality accident occur. In the event of Nuclear Emergency Preparedness, PT BATAN TEKNOLOGI has also proposed the organization structure how promptly to report the crisis to Nuclear Energy Control Board (BAPETEN) Indonesia. (author)

  10. Safety and security aspects in design of digital safety I and C in nuclear power plants

    International Nuclear Information System (INIS)

    Ding, Yongjian; Waedt, Karl

    2016-01-01

    The paper describes a safety objective oriented systematic design approach of digital (computerized) safety I and C in modern nuclear power plants which considers the plant safety requirements as well as cybersecurity needs. The defence in depth philosophy is applied by using different defence lines in the I and C architecture and protection zones in the plant IT environment.

  11. Safety and security aspects in design of digital safety I and C in nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Ding, Yongjian [University of Applied Sciences Magdeburg-Stendal, Magdeburg (Germany). Inst. of Electrical Engineering; Waedt, Karl [Areva GmbH, Erlangen (Germany). PEAS-G

    2016-05-15

    The paper describes a safety objective oriented systematic design approach of digital (computerized) safety I and C in modern nuclear power plants which considers the plant safety requirements as well as cybersecurity needs. The defence in depth philosophy is applied by using different defence lines in the I and C architecture and protection zones in the plant IT environment.

  12. Method of safety evaluation in nuclear power plants

    International Nuclear Information System (INIS)

    Kuraszkiewicz, P.; Zahn, P.

    1988-01-01

    A novel quantitative technique for evaluating safety of subsystems of nuclear power plants based on expert estimations is presented. It includes methods of mathematical psychology recognizing the effect of subjective factors in the expert estimates and, consequently, contributes to further objectification of evaluation. It may be applied to complementing probabilistic safety assessment. As a result of such evaluations a characteristic 'safety of nuclear power plants' is obtained. (author)

  13. Tools for plant safety engineer

    International Nuclear Information System (INIS)

    Fabic, S.

    1996-01-01

    This paper contains: - review of tools for monitoring plant safety equipment reliability and readiness, before and accident (performance indicators for monitoring the risk and reliability performance and for determining when degraded performance alert levels are achieved) - brief reviews of tools for use during an accident: Emergency Operating Procedures (EOPs), Emergency Response Data System (ERDS), Reactor Safety Assessment System (RSAS), Computerized Accident Management Support

  14. Safety prediction technique for nuclear power plants

    International Nuclear Information System (INIS)

    Henry, C.D. III; Anderson, R.T.

    1985-01-01

    This paper presents a safety prediction technique (SPT) developed by Reliability Technology Associates (RTA) for nuclear power plants. It is based on a technique applied by RTA to assess the flight safety of US Air Force aircraft. The purpose of SPT is to provide a computerized technique for objective measurement of the effect on nuclear plant safety of component failure or procedural, software, or human error. A quantification is determined, called criticality, which is proportional to the probability that a given component or procedural-human action will cause the plant to operate in a hazardous mode. A hazardous mode is characterized by the fact that there has been a failure/error and the plant, its operating crew, and the public are exposed to danger. Whether the event results in an accident, an incident, or merely the exposure to danger is dependent on the skill and reaction of the operating crew as well as external influences. There are three major uses of SPT: (a) to predict unsafe situations so that corrective action can be taken before accidents occur, (b) to quantify the impact of equipment malfunction or procedural, software, or human error on safety and thereby establish priorities for proposed modifications, and (c) to provide a means of evaluating proposed changes for their impact on safety prior to implementation and to provide a method of tracking implemented changes

  15. Licensee responsibility for nuclear power plant safety

    International Nuclear Information System (INIS)

    Schneider, Horst

    2010-01-01

    Simple sentences easy to grasp are desirable in regulations and bans. However, in a legal system, their meaning must be unambiguous. Article 6, Paragraph 1 of the EURATOM Directive on a community framework for the nuclear safety of nuclear facilities of June 2009 states that 'responsibility for the nuclear safety of a nuclear facility is incumbent primarily on the licensee.' The draft 'Safety Criteria for Nuclear Power Plants, Revision D, April 2009' of the German Federal Ministry for the Environment, Nature Conservation, and Nuclear Safety (BMU) (A Module 1, 'Safety Criteria for Nuclear Power Plants: Basic Safety Criteria' / '0 Principles' Paragraph 2) reads: 'Responsibility for ensuring safety rests with the licensee. He shall give priority to compliance with the safety goal over the achievement of other operational objectives.' In addition, the existing rules and regulations, whose rank is equivalent to that of international regulations, assign priority to the safety goal to be pursued by the licensee over all other objectives of the company. The operator's responsibility for nuclear safety can be required and achieved only on the basis of permits granted, which must meet legal requirements. The operator's proximity to plant operation is the reason for his 'primary responsibility.' Consequently, verbatim incorporation of Article 6, Paragraph 1 of the EURATOM Directive would only be a superscript added to existing obligations of the operator - inclusive of a safety culture designed as an incentive to further 'the spirit of safety-related actions' - without any new legal contents and consequences. In the reasons of the regulation, this would have to be clarified in addition to the cryptic wording of 'responsibility.. primarily,' at the same time expressing that operators and authorities work together in a spirit of openness and trust. (orig.)

  16. Safety criteria for design of nuclear power plants

    International Nuclear Information System (INIS)

    1997-01-01

    In Finland the general safety requirements for nuclear power plants are presented in the Council of State Decision (395/91). In this guide, safety principles which supplement the Council of State Decision and which are to be used in the design of nuclear power plants are defined

  17. Preliminary safety analysis report for project 89-GEB-610 Plutonium Finishing Plant instrumentation upgrade. Revision 1

    International Nuclear Information System (INIS)

    Huber, T.E.

    1995-01-01

    This document consists of an analysis of the MICON system upgrade. This project shall install a Micon Co. distributed process monitor and control system with Sparc Sun workstation operator interfaces. The Sparc workstations are housed in consoles custom designed to human factors specifications. The distributed control system (DCS) shall have the installed capacity to monitor and control all related instruments and equipment presently connected to the panels in the PFP Power Control Room 321A as listed in the input/output list. This also includes all devices monitored and controlled by the 2736-ZB Allen Bradley programmable logic controller. The system has since assumed the control and monitoring responsibilities for Projects B- 680H ''Low Level Waste Treatment Facility'' and C-031H ''PFP Liquid Effluent Facilities''. Part of the new en's change area in Building 234-5ZA, Room 712, has been remodeled to house two consoles and one supervisor console. Local control units containing the microprocontrollers and the input/output interface circuit boards shall be wired to the instrumentation and controlled equipment. These units communicate with the Sparc workstations via a redundant data communications highway and shall be strategic, throughout the PFP facility. The DCS has already been purchased from Micon Co., located in Houston Texas, presently on site

  18. Plant safety and performance indicators for regulatory use

    International Nuclear Information System (INIS)

    Ferjancic, M.; Nemec, T.; Cimesa, S.

    2004-01-01

    Slovenian Nuclear Safety Administration (SNSA) supervises nuclear and radiological safety of Krsko NPP. This SNSA supervision is performed through inspections, safety evaluations of plant modifications and event analyses as well as with the safety and performance indicators (SPI) which are a valuable data source for plant safety monitoring. In the past SNSA relied on the SPI provided by Krsko NPP and did not have a set of SPI which would be more appropriate for regulatory use. In 2003 SNSA started with preparation of a new set of SPI which would be more suitable for performing the regulatory oversight of the plant. New internal SNSA procedure which is under preparation will define use and evaluation of SPI and will include definitions for the proposed set of SPI. According to the evaluation of SPI values in comparison with the limiting values and/or trending, the procedure will define SNSA response and actions. (author)

  19. Nuclear power plant systems, structures and components and their safety classification

    International Nuclear Information System (INIS)

    2000-01-01

    The assurance of a nuclear power plant's safety is based on the reliable functioning of the plant as well as on its appropriate maintenance and operation. To ensure the reliability of operation, special attention shall be paid to the design, manufacturing, commissioning and operation of the plant and its components. To control these functions the nuclear power plant is divided into structural and functional entities, i.e. systems. A systems safety class is determined by its safety significance. Safety class specifies the procedures to be employed in plant design, construction, monitoring and operation. The classification document contains all documentation related to the classification of the nuclear power plant. The principles of safety classification and the procedures pertaining to the classification document are presented in this guide. In the Appendix of the guide, examples of systems most typical of each safety class are given to clarify the safety classification principles

  20. IAEA Completes Safety Review at Czech Nuclear Power Plant

    International Nuclear Information System (INIS)

    2012-01-01

    Full text: An international team of nuclear safety experts, led by the International Atomic Energy Agency (IAEA), today completed a review of safety practices at Temelin Nuclear Power Station in the Czech Republic. The team highlighted the Power Plant's good practices and also recommended improvements to some safety measures. At the request of the Government of the Czech Republic, the IAEA assembled a team of nuclear installation safety experts to send an Operational Safety Review Team (OSART) to the Power Plant, and the mission was conducted from 5 to 22 November 2012. The team was comprised of experts from Brazil, Hungary, Slovakia, South Africa, Sweden, Ukraine and the United Kingdom. An OSART mission is designed as a review of programmes and activities essential to operational safety. It is not a regulatory inspection, nor is it a design review or a substitute for an exhaustive assessment of the Plant's overall safety status. The team at Temelin conducted an in-depth review of the functions essential to the safe operation of the Power Plant, which are under the responsibility of the site's management. The review covered the areas of management, organization and administration; operations; maintenance; technical support; operating experience; radiation protection; chemistry; and severe accident management. The conclusions of the review are based on the IAEA's Safety Standards and proven good international practices. The OSART team has identified good plant practices, which will be shared with the rest of the nuclear industry for consideration of potential application elsewhere. Examples include the following: - The Power Plant has adopted effective computer software to improve the efficiency of the plant to prepare and isolate equipment for maintenance; - The Power Plant undertakes measures to control precisely the chemical parameters that limit corrosion in the reactor's coolant system, which in turn reduce radiation exposure to the workforce; and - The Temelin

  1. Safety culture of nuclear power plant

    International Nuclear Information System (INIS)

    Zheng Beixin

    2008-01-01

    This paper is a summary on the basis of DNMC safety culture training material for managerial personnel. It intends to explain the basic contents of safety, design, management, enterprise culture, safety culture of nuclear power plant and the relationship among them. It explains especially the constituent elements of safety culture system, the basic requirements for the three levels of commitments: policy level, management level and employee level. It also makes some analyses and judgments for some typical safety culture cases, for example, transparent culture and habitual violation of procedure. (authors)

  2. Westinghouse Advances in Passive Plant Safety

    International Nuclear Information System (INIS)

    Bruschi, H. J.; Manager, General; Gerstenhaber, E.

    1993-01-01

    On June 26, 1992, Westinghouse submitted the Ap600 Standard Safety Analysis Report and comprehensive PIRA results to the U. S. NRC for review as part of the Ap600 design certification program. This major milestone was met on time on a schedule set more than 3 years before submittal and is the result of the cooperative efforts of the U. S. Department of Energy (DOE), the Electric Power Requirements Program, and the Westinghouse Ap600 design team. These efforts were initiated in 1985 to develop a 600 MW advanced light water reactor plant design based on specific technical requirements established to provide the safety, simplicity, reliability, and economics necessary for the next generation of nuclear power plants. The Ap600 design achieves the ALRR safety requirements through ample design margins, simplified safety systems based on natural driving forces, and on a human-engineered man-machine interface system. Extensive Probabilistic Risk evolution, have recently shown that even if none of the active defense-in-depth safety systems are available, the passive systems alone meet safety goals. Furthermore, many tests in an extensive test program have begun or have been completed. Early tests show that passive safety perform well and meet design expectations

  3. Plant functional modelling as a basis for assessing the impact of management on plant safety

    International Nuclear Information System (INIS)

    Rasmussen, Birgitte; Petersen, Kurt E.

    1999-01-01

    A major objective of the present work is to provide means for representing a chemical process plant as a socio-technical system, so as to allow hazard identification at a high level in order to identify major targets for safety development. The main phases of the methodology are: (1) preparation of a plant functional model where a set of plant functions describes coherently hardware, software, operations, work organization and other safety related aspects. The basic principle is that any aspect of the plant can be represented by an object based upon an Intent and associated with each Intent are Methods, by which the Intent is realized, and Constraints, which limit the Intent. (2) Plant level hazard identification based on keywords/checklists and the functional model. (3) Development of incident scenarios and selection of hazardous situation with different safety characteristics. (4) Evaluation of the impact of management on plant safety through interviews. (5) Identification of safety critical ways of action in the management system, i.e. identification of possible error- and violation-producing conditions

  4. Strengthening of nuclear power plant construction safety management

    International Nuclear Information System (INIS)

    Yu Jun

    2012-01-01

    The article describes the warning of the Fukushima nuclear accident, and analyzes the major nuclear safety issues in nuclear power development in China, problems in nuclear power plants under construction, and how to strengthen supervision and management in nuclear power construction. It also points out that the development of nuclear power must attach great importance to the safety, and nuclear power plant construction should strictly implement the principle of 'safety first and quality first'. (author)

  5. Probabilistic safety assessment in nuclear power plant management

    International Nuclear Information System (INIS)

    Holloway, N.J.

    1989-06-01

    Probabilistic Safety Assessment (PSA) techniques have been widely used over the past few years to assist in understanding how engineered systems respond to abnormal conditions, particularly during a severe accident. The use of PSAs in the design and operation of such systems thus contributes to the safety of nuclear power plants. Probabilistic safety assessments can be maintained to provide a continuous up-to-date assessment (Living PSA), supporting the management of plant operations and modifications

  6. Safety of industrial irradiation plants

    International Nuclear Information System (INIS)

    1992-01-01

    Radiation is nowadays used in many applications in industry and medicine; accidental exposure, however, can have grave consequences as large doses of radiation occur in the 600 accelerator or gamma source plants in use around the world. This film explains the operation of irradiation plants and the safety procedures that must be followed to prevent accidents and to ensure safe use

  7. Views on safety culture at Swedish and Finnish nuclear power plants

    International Nuclear Information System (INIS)

    Hammar, L.; Wahlstroem, B.; Kettunen, J.

    2000-02-01

    The report presents the results of interviews about safety culture at Swedish and Finnish nuclear power plants. The aim is to promote the safety work and increase the debate about safety in nuclear power plants, by showing that the safety culture is an important safety factor. The interviews point out different threats, which may become real. It is therefor necessary that the safety aspects get support from of the society and the power plant owners. (EHS)

  8. The safety of future nuclear power plants in France

    International Nuclear Information System (INIS)

    Queniart, D.

    1988-10-01

    The present paper concerns certain personal thoughts on the safety of future French power plants, which will come into operation at the beginning of the next century. These reflections, which are made on the author's own behalf and, under no circumstances, implicate at this stage the official views of the French safety authorities, are aimed at defining some directions for the improvement of safety in these future plants as compared with that of plants presently in operation or under construction

  9. Aspects of nuclear safety at power plants and fuel cycle plants in the USSR

    International Nuclear Information System (INIS)

    Kozlov, N.I.; Efimov, E.; Dubovskij, B.G.; Dikarev, V.; Lyubchenko, V.; Kruglov, A.K.

    1977-01-01

    The paper discusses the problems of organizing inspection monitoring of power plants including the development of some regulations and norms and the interaction between the USSR State Nuclear Safety Organization, scientific and designing organizations and power plants. The principles of computer use to work out advice for operational staff and warning signals and commands for the reactor control and protection system are discussed. Some attention is turned to the importance of using high-speed computers to calculate prompt reactivity values and to determine impurity concentrations in the coolant and margins to permissible operational limits. In particular, reactimeters are considered as signal generators in monitor and protection systems. Some problems of nuclear safety inspection, the issue and inculcation of some regulations and operational documents on nuclear safety, and instrumentation of plants reprocessing or processing fuel elements are presented. Methods of determining the critical parameters of technological units are described, together with the fundamental principles of fuel cycle plant nuclear safety, providing margin coefficients, accounting for deviations from the normal operational process and other problems, as well as methods of keeping the restrictions on nuclear safety requirements at fuel cycle plants. (author)

  10. Key asset - inherent safety of LMFBR Pool Plant

    International Nuclear Information System (INIS)

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

    1984-04-01

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

  11. Key asset--Inherent safety of LFMBR pool plant

    International Nuclear Information System (INIS)

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

    1984-01-01

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

  12. Safety guide on fire protection in nuclear power plants

    International Nuclear Information System (INIS)

    1976-01-01

    The purpose of the Safety Guide is to give specific design and operational guidance for protection from fire and explosion in nuclear power plants, based on the general guidance given in the relevant sections of the 'Safety Code of Practice - Design' and the 'Safety Code of Practice - Operation' of the International Atomic Energy Agency. The guide will confine itself to fire protection of safety systems and items important to safety, leaving the non-safety matters of fire protection in nuclear power plants to be decided upon the basis of the various available national and international practices and regulations. (HP) [de

  13. Radioactive waste management plan for the PBMR (Pty) Ltd fuel plant

    International Nuclear Information System (INIS)

    Makgae, Mosidi E.

    2009-01-01

    The Pebble Bed Modular Reactor (Pty) Ltd Fuel Plant (PFP) radioactive waste management plan caters for waste from generation, processing through storage and possible disposal. Generally, the amount of waste that will be generated from the PFP is Low and Intermediate Level Waste. The waste management plan outlines all waste streams and the management options for each stream. It also discusses how the Plant has been designed to ensure radioactive waste minimisation through recycling, recovery, reuse, treatment before considering disposal. Compliance to the proposed plan will ensure compliance with national legislative requirements and international good practice. The national and the overall waste management objective is to ensure that all PFP wastes are managed appropriately by utilising processes that minimize, reduce, recover and recycle without exposing employees, the public and the environment to unacceptable impacts. Both International Atomic Energy Agency (IAEA) and Department of Minerals and Energy (DME) principles act as a guide in the development of the strategy in order to ensure international best practice, legal compliance and ensuring that the impact of waste on employees, environment and the public is as low as reasonably achievable. The radioactive waste classification system stipulated in the Radioactive Waste Management Policy and Strategy 2005 will play an important role in classifying radioactive waste and ensuring that effective management is implemented for all waste streams, for example gaseous, liquid or solid wastes.

  14. Peak pressures from hydrogen deflagrations in the PFP thermal stabilization glovebox

    International Nuclear Information System (INIS)

    Van Keuren, J.C.

    1998-01-01

    This document describes the calculations of the peak pressures due to hydrogen deflagrations in the glovebox used for thermal stabilization (glovebox HC-21A) in PFP. Two calculations were performed. The first considered the burning of hydrogen released from a 7 inch Pu can in the Inert Atmosphere Confinement (IAC) section of the glovebox. The peak pressure increase was 12400 Pa (1.8 psi). The second calculation considered burning of the hydrogen from 25 g of plutonium hydride in the airlock leading to the main portion of the glovebox. Since the glovebox door exposes most of the airlock when open, the deflagration was assumed to pressurize the entire glovebox. The peak pressure increase was 3860 Pa (0.56 psi)

  15. Results of dilution studies with waste from tank 241-AN-104

    International Nuclear Information System (INIS)

    HERTING, D.L.

    1999-01-01

    This report documents the completion of the B and W Hanford Company functional and Safety Review Board reviews and the Fluor Daniel Hanford review for the Plutonium Finishing Plant (PFP) Final Safety Analysis Report (FSAR), HNF-SD-CP-SAR-021, Revision 1. The reviews for the FSAR were conducted during the period from December 9, 1998 to January 14, 1999

  16. Special safety requirements applied to Brazilian nuclear power plant

    International Nuclear Information System (INIS)

    Lepecki, W.P.S.; Hamel, H.J.E.; Koenig, N.; Vieira, P.C.R.; Fritzsche, J.C.

    1981-01-01

    Some safety aspects of the Angra 2 and 3 nuclear power plants are presented. An analysis of the civil and mechanical project of these nuclear power plant having in view a safety analysis is done. (E.G.) [pt

  17. Code on the safety of nuclear power plants: Siting

    International Nuclear Information System (INIS)

    1988-01-01

    This Code provides criteria and procedures that are recommended for safety in nuclear power plant siting. It forms part of the Agency's programme for establishing Codes and Safety Guides relating to land based stationary thermal neutron power plants

  18. Functional enrichment analyses and construction of functional similarity networks with high confidence function prediction by PFP

    Directory of Open Access Journals (Sweden)

    Kihara Daisuke

    2010-05-01

    Full Text Available Abstract Background A new paradigm of biological investigation takes advantage of technologies that produce large high throughput datasets, including genome sequences, interactions of proteins, and gene expression. The ability of biologists to analyze and interpret such data relies on functional annotation of the included proteins, but even in highly characterized organisms many proteins can lack the functional evidence necessary to infer their biological relevance. Results Here we have applied high confidence function predictions from our automated prediction system, PFP, to three genome sequences, Escherichia coli, Saccharomyces cerevisiae, and Plasmodium falciparum (malaria. The number of annotated genes is increased by PFP to over 90% for all of the genomes. Using the large coverage of the function annotation, we introduced the functional similarity networks which represent the functional space of the proteomes. Four different functional similarity networks are constructed for each proteome, one each by considering similarity in a single Gene Ontology (GO category, i.e. Biological Process, Cellular Component, and Molecular Function, and another one by considering overall similarity with the funSim score. The functional similarity networks are shown to have higher modularity than the protein-protein interaction network. Moreover, the funSim score network is distinct from the single GO-score networks by showing a higher clustering degree exponent value and thus has a higher tendency to be hierarchical. In addition, examining function assignments to the protein-protein interaction network and local regions of genomes has identified numerous cases where subnetworks or local regions have functionally coherent proteins. These results will help interpreting interactions of proteins and gene orders in a genome. Several examples of both analyses are highlighted. Conclusion The analyses demonstrate that applying high confidence predictions from PFP

  19. Safety/security interface assessments at commercial nuclear power plants

    International Nuclear Information System (INIS)

    Byers, K.R.; Brown, P.J.; Norderhaug, L.R.

    1985-01-01

    The findings of the Haynes Task Force Committee (NUREG-0992) are used as the basis for defining safety/security assessment team activities at commercial nuclear power plants in NRC Region V. A safety/security interface assessment outline and the approach used for making the assessments are presented along with the composition of team members. As a result of observing simulated plant emergency conditions during scheduled emergency preparedness exercises, examining security and operational response procedures, and interviewing plant personnel, the team has identified instances where safety/security conflicts can occur

  20. Safety/security interface assessments at commercial nuclear power plants

    International Nuclear Information System (INIS)

    Byers, K.R.; Brown, P.J.; Norderhaug, L.R.

    1985-07-01

    The findings of the Haynes Task Force Committee (NUREG-0992) are used as the basis for defining safety/security assessment team activities at commercial nuclear power plants in NRC Region V. A safety/security interface assessment outline and the approach used for making the assessments are presented along with the composition of team members. As a result of observing simulated plant emergency conditions during scheduled emergency preparedness exercises, examining security and operational response procedures, and interviewing plant personnel, the team has identified instances where safety/security conflicts can occur. 2 refs

  1. Safety targets for nuclear power plants

    International Nuclear Information System (INIS)

    Herttrich, P.M.

    1985-01-01

    By taking as an example the safety targets of the American nuclear energy authority US-NRC, this paper explains what is meant by global, quantitative safety targets for nuclear power plants and what expectations are associated with the selecton of such safety targets. It is shown how probabilistic methods can be an appropriate completion of proven deterministic methods and what are the sectors where their application may become important in future. (orig./HP) [de

  2. Integrated Plant Safety Assessment, Systematic Evaluation Program, Palisades Plant (Docket No. 50-255)

    International Nuclear Information System (INIS)

    1983-11-01

    This report documents the review completed under the SEP for those issues that required refined engineering evaluations or the continuation of ongoing evaluations after the Final IPSAR for the Palisades Plant was issued. The review has provided for (1) an assessment of the significance of differences between current technical positions on selected safety issues and those that existed when the Palisades Plant was licensed, (2) a basis for deciding on how these differences should be resolved in an integrated plant review, and (3) a documented evaluation of plant safety when the supplement to the Final IPSAR and the Safety Evaluation Report for converting the license from a provisional to a full-term license have been issued. The Final IPSAR and its supplement will form part of the bases for considering the conversion of the provisional operating license to a full-term operating license

  3. Safety aspects and operating experience of LWR plants in Japan

    International Nuclear Information System (INIS)

    Aoki, S.; Yoshioka, T.; Toyota, M.; Hinoki, M.

    1977-01-01

    To develop nuclear power generation for the future, it is necessary to put further emphasis on safety assurance and to endeavour to devise measures to improve plant availability, based on the careful analysis of causes that reduce plant availability. The paper discusses the results of studies on the following items from such viewpoints: (1) Safety and operating experience of LWR nuclear power plants in Japan: operating experience with LWRs; improvements in LWR design during the past ten years; analysis of the factors affecting plant availability; (2) Assurance of safety and measures to increase availability: measures for safety and environmental protection; measures to reduce radiation exposure of employees; appropriateness of maintenance and inspection work; measures to increase plant availability; measures to improve reliability of equipment and components; (3) Future technical problems. (author)

  4. The safety of the new reprocessing plants of La Hague

    International Nuclear Information System (INIS)

    Devillers, C.; Dubois, G.

    1987-09-01

    In this document the authors show the main guiding lines on which is based the safety of the new reprocessing plant of La Hague. They are: - the objectives: to limit the impacts on workers and environment - the methods: safety analysis based on the checking and evaluation of significant risks. - the means: to make a safety plant by the use of quality assurance in the conception and in the plant construction [fr

  5. IAEA Concludes Safety Review at Gravelines Nuclear Power Plant, France

    International Nuclear Information System (INIS)

    2012-01-01

    Full text: An IAEA-led international team of nuclear safety experts noted a series of good practices and made recommendations to reinforce some safety measures during a review of operational safety at France's Gravelines Nuclear Power Plant (NPP) that concluded today. The Operational Safety Review Team (OSART) was assembled at the French Government's request. The in-depth review, which began 12 November 2012, focused on aspects essential to the safe operation of the NPP. The team was composed of experts from Bulgaria, China, Germany, Hungary, Japan, Romania, Slovakia, South Africa, Spain, Ukraine and the IAEA. The review covered the areas of management, organization and administration; training and qualification; operations; maintenance; technical support; operating experience; radiation protection; chemistry; emergency planning and preparedness; and severe accident management. The conclusions of the review are based on the IAEA's Safety Standards. The OSART team has identified good plant practices, which will be shared with the rest of the nuclear industry for consideration of their possible use elsewhere. Examples include the following: - The Power Plant uses a staff-skills mapping process that significantly enhances knowledge of the facility's collective and individual skills and provides proactive management to address the loss of such skills; - As a measure to reduce the risk of workers' radiation exposure, the Power Plant uses a system to ensure that dose rate measurements are carried out at a precise distance from the source of radiation; and - Flood protection of the Power Plant is supported by special technical guidance documents and associated arrangements. The team identified a number of proposals for improvements to operational safety at Gravelines NPP. Examples include the following: - The Power Plant should reinforce its measures to prevent foreign objects from entering plant systems; - The Power Plant should ensure the 24-hour presence of an operator

  6. Meteorological events in site evaluation for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2005-01-01

    This Safety Guide provides recommendations and guidance on conducting hazard assessments of extreme and rare meteorological phenomena. It is of interest to safety assessors and regulators involved in the licensing process as well as to designers of nuclear power plants. This Safety Guide was prepared under the IAEA programme for safety standards for nuclear power plants. It supplements the IAEA Safety Requirements publication on Site Evaluation for Nuclear Facilities which is to supersede the Code on the Safety of Nuclear Power Plants: Siting, Safety Series No. 50-C-S (Rev. 1), IAEA, Vienna (1988). The present Safety Guide supersedes two earlier Safety Guides: Safety Series No. 50-SG-S11A (1981) on Extreme Meteorological Events in Nuclear Power Plant Siting, Excluding Tropical Cyclones and Safety Series No. 50-SG-S11B (1984) on Design Basis Tropical Cyclone for Nuclear Power Plants. The purpose of this Safety Guide is to provide recommendations and guidance on conducting hazard assessments of extreme and rare meteorological phenomena. This Safety Guide provides interpretation of the Safety Requirements publication on Site Evaluation for Nuclear Facilities and guidance on how to fulfil these requirements. It is aimed at safety assessors or regulators involved in the licensing process as well as designers of nuclear power plants, and provides them with guidance on the methods and procedures for analyses that support the assessment of the hazards associated with extreme and rare meteorological events. This Safety Guide discusses the extreme values of meteorological variables and rare meteorological phenomena, as well as their rates of occurrence, according to the following definitions: (a) Extreme values of meteorological variables such as air temperature and wind speed characterize the meteorological or climatological environment. And (b) Rare meteorological phenomena

  7. Design of reactor containment systems for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2008-01-01

    This Safety Guide was prepared under the IAEA programme for safety standards for nuclear power plants. It is a revision of the Safety Guide on Design of the Reactor Containment Systems in Nuclear Power Plants (Safety Series No. 50-Sg-D1) issued in 1985 and supplements the Safety Requirements publication on Safety of Nuclear Power Plants: Design. The present Safety Guide was prepared on the basis of a systematic review of the relevant publications, including the Safety of Nuclear Power Plants: Design, the Safety fundamentals publication on The Safety of Nuclear Installations, Safety Guides, INSAG Reports, a Technical Report and other publications covering the safety of nuclear power plants. 1.2. The confinement of radioactive material in a nuclear plant, including the control of discharges and the minimization of releases, is a fundamental safety function to be ensured in normal operational modes, for anticipated operational occurrences, in design basis accidents and, to the extent practicable, in selected beyond design basis accidents. In accordance with the concept of defence in depth, this fundamental safety function is achieved by means of several barriers and levels of defence. In most designs, the third and fourth levels of defence are achieved mainly by means of a strong structure enveloping the nuclear reactor. This structure is called the 'containment structure' or simply the 'containment'. This definition also applies to double wall containments. 1.3. The containment structure also protects the reactor against external events and provides radiation shielding in operational states and accident conditions. The containment structure and its associated systems with the functions of isolation, energy management, and control of radionuclides and combustible gases are referred to as the containment systems

  8. Design of reactor containment systems for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2004-01-01

    This Safety Guide was prepared under the IAEA programme for safety standards for nuclear power plants. It is a revision of the Safety Guide on Design of the Reactor Containment Systems in Nuclear Power Plants (Safety Series No. 50-Sg-D1) issued in 1985 and supplements the Safety Requirements publication on Safety of Nuclear Power Plants: Design. The present Safety Guide was prepared on the basis of a systematic review of the relevant publications, including the Safety of Nuclear Power Plants: Design, the Safety fundamentals publication on The Safety of Nuclear Installations, Safety Guides, INSAG Reports, a Technical Report and other publications covering the safety of nuclear power plants. 1.2. The confinement of radioactive material in a nuclear plant, including the control of discharges and the minimization of releases, is a fundamental safety function to be ensured in normal operational modes, for anticipated operational occurrences, in design basis accidents and, to the extent practicable, in selected beyond design basis accidents. In accordance with the concept of defence in depth, this fundamental safety function is achieved by means of several barriers and levels of defence. In most designs, the third and fourth levels of defence are achieved mainly by means of a strong structure enveloping the nuclear reactor. This structure is called the 'containment structure' or simply the 'containment'. This definition also applies to double wall containments. 1.3. The containment structure also protects the reactor against external events and provides radiation shielding in operational states and accident conditions. The containment structure and its associated systems with the functions of isolation, energy management, and control of radionuclides and combustible gases are referred to as the containment systems

  9. IAEA Concludes Safety Review at Chooz Nuclear Power Plant in France

    International Nuclear Information System (INIS)

    2013-01-01

    Full text: An IAEA-led international team of nuclear safety experts noted good practices and made recommendations to reinforce safety measures during a review of operational safety at France's Chooz Nuclear Power Plant (NPP) that concluded today. The Operational Safety Review Team (OSART) was assembled at the French Government's request. The in-depth review, which began 17 June, focused on aspects essential to the safe operation of the NPP. The team comprised experts from Switzerland, Belgium, Germany, China, India, United Kingdom, Czech Republic, Canada, Hungary and the IAEA. The review covered the areas of management, organization and administration; training and qualification of personnel; operations; maintenance; technical support; operating experience; radiation protection; chemistry; emergency planning and preparedness; and severe accident management. The conclusions of the review are based on the IAEA's Safety Standards. The OSART team identified good plant practices that will be shared with the rest of the nuclear industry for consideration. Examples include: The plant has a professional development programme as part of a joint employment effort shared by the plant and its contractors. This enables trainees to develop professional capability, understand practices and gain experience from other nuclear power plants in terms of work planning and coordination; The plant has built a strong relationship between the on-shift response team of the plant and the local fire brigade to improve firefighting and rescue operations; Self-assessment groups discuss and resolve specific issues within operations, empowering operations personnel to take ownership of improvement programmes; and The plant has improved warnings at entrances to all o range zones , areas of elevated dose rates to which only authorized staff have access. The team identified a number of improvements to operational safety at Chooz NPP. Examples include: The plant should review its process for the

  10. Safety classification of nuclear power plant systems, structures and components

    International Nuclear Information System (INIS)

    1992-01-01

    The Safety Classification principles used for the systems, structures and components of a nuclear power plant are detailed in the guide. For classification, the nuclear power plant is divided into structural and operational units called systems. Every structure and component under control is included into some system. The Safety Classes are 1, 2 and 3 and the Class EYT (non-nuclear). Instructions how to assign each system, structure and component to an appropriate safety class are given in the guide. The guide applies to new nuclear power plants and to the safety classification of systems, structures and components designed for the refitting of old nuclear power plants. The classification principles and procedures applying to the classification document are also given

  11. Operational safety of nuclear power plants

    International Nuclear Information System (INIS)

    Tanguy, P.

    1987-01-01

    The operational safety of nuclear power plants has become an important safety issue since the Chernobyl accident. A description is given of the various aspects of operational safety, including the importance of human factors, responsibility, the role and training of the operator, the operator-machine interface, commissioning and operating procedures, experience feedback, and maintenance. The lessons to be learnt from Chernobyl are considered with respect to operator errors and the management of severe accidents. Training of personnel, operating experience feedback, actions to be taken in case of severe accidents, and international cooperation in the field of operational safety, are also discussed. (U.K.)

  12. Safety of Nuclear Power Plants: Design. Specific Safety Requirements (Spanish Edition); Seguridad de las centrales nucleares: Diseno. Requisitos de seguridad especificos

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2012-04-15

    This publication is a revision of Safety Requirements No. NS-R-1, Safety of Nuclear Power Plants: Design. It establishes requirements applicable to the design of nuclear power plants and elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. It will be useful for organizations involved in the design, manufacture, construction, modification, maintenance, operation and decommissioning of nuclear power plants, as well as for regulatory bodies. Contents: 1. Introduction; 2. Applying the safety principles and concepts; 3. Management of safety in design; 4. Principal technical requirements; 5. General plant design; 6. Design of specific plant systems.

  13. Safety performance indicators used by the Russian Safety Regulatory Authority in its practical activities on nuclear power plant safety regulation

    International Nuclear Information System (INIS)

    Khazanov, A.L.

    2005-01-01

    The Sixth Department of the Nuclear, Industrial and Environmental Regulatory Authority of Russia, Scientific and Engineering Centre for Nuclear and Radiation Safety process, analyse and use the information on nuclear power plants (NPPs) operational experience or NPPs safety improvement. Safety performance indicators (SPIs), derived from processing of information on operational violations and analysis of annual NPP Safety Reports, are used as tools to determination of trends towards changing of characteristics of operational safety, to assess the effectiveness of corrective measures, to monitor and evaluate the current operational safety level of NPPs, to regulate NPP safety. This report includes a list of the basic SPIs, those used by the Russian safety regulatory authority in regulatory activity. Some of them are absent in list of IAEA-TECDOC-1141 ('Operational safety performance indicators for nuclear power plants'). (author)

  14. An engineer-constructor's view of nuclear power plant safety

    International Nuclear Information System (INIS)

    Landis, J.W.; Jacobs, S.B.

    1984-01-01

    At SWEC we have been involved in the development of safety features of nuclear power plants ever since we served as the engineer-constructur for the first commerical nuclear power station at Shippingport, Pennsylvania, in the 1950s. Our personnel have pioneered a number of safety innovations and improvements. Among these innovations is the subatmospheric containment for pressurized water reactor (PWR) power plants. This type of containment is designed so that leakage will terminate within 1 to 2 hours of the worst postulated loss of coolant accident. Other notable contributions include first use of reinforced-concrete atmospheric containments for PWR power plants and of reinforced-concrete, vapor-suppression containments for boiling water reactor (BWR) power plants. Both concepts meet rigorous U.S. safety requirements. SWEC has performed a substantial amount of work on developing standardized plant designs and has developed standardized engineering and construction techniques and procedures. Standardization concepts are being developed in Canada, France, USSR, and Germany, as well as in the United States. The West German convoy concept, which involves developing a number of standardized plants in a common effort, has been quite successful. We believe standardization contributes to safety in a number of ways. Use of standardized designs, procedures, techniques, equipment, and methods increases efficiency and results in higher quality. Standardization also reduces the design variations with which plant operators, emergency teams, and regulatory personnel must be familiar, thus increasing operator capability, and permits specialized talents to be focused on important safety considerations. (orig./RW)

  15. Surveillance of items important to safety in nuclear power plants

    International Nuclear Information System (INIS)

    1990-01-01

    The Guide was prepared as part of the IAEA's programme, referred to as the NUSS Programme, for establishing Codes and Safety Guides relating to nuclear power plants. THe Guide supplements the Code on the Safety of Nuclear Power Plants: Operation, IAEA Safety Series No. 50-C-O(Rev.1). The operating organization has overall responsibility for the safe operation of the nuclear power plant. Therefore, it shall ensure that adequate surveillance activities are carried out in order to verify that the plant is operated within the prescribed operational limits and conditions, and to detect in time any deterioration of structures, systems and components as well as any adverse trend that could lead to an unsafe condition. These activities can be classified as: Monitoring plant parameters and system status; Checking and calibrating instrumentation; Testing and inspecting structures, systems and components. This Safety Guide provides guidance and recommendations on surveillance activities to ensure that structures, systems and components important to safety are available to perform their functions in accordance with design intent and assumptions

  16. Safety aspects and operating experience of LWR plants in Japan

    International Nuclear Information System (INIS)

    Aoki, S.; Hinoki, M.

    1977-01-01

    From the outset of nuclear power development in Japan, major emphasis has been placed on the safety of the nuclear power plants. There are now twelve nuclear power plants in operation with a total output of 6600 MWe. Their operating records were generally satisfactory, but in the 1974 to 1975 period, they experienced somewhat declined availability due to the repair work under the specific circumstances. After investigation of causes of troubles and the countermeasures thereof were made to ensure safety, they are now keeping good performance. In Japan, nuclear power plants are strictly subject to sufficient and careful inspection in compliance with the safety regulation, and are placed under stringent radiation control of employees. Under the various circumstances, however, the period of annual inspection tends to be prolonged more than originally planned, and this consequently is considered to be one of the causes of reduced availability. In order to develop nuclear power generation for the future, it is necessary to put further emphasis on the assurance of safety and to endeavor to devise measures to improve availability of the plants, based on the careful analysis of causes which reduce plant availability. This paper discusses the results of studies made for the following items from such viewpoints: (1) Safety and Operating Experience of LWR Nuclear Power Plants in Japan; a) Operating experience with light water reactors b) Improvements in design of light water reactors during the past ten years c) Analysis of the factors which affect plant availability; 2) Assurance of Safety and Measures to Increase Availability a) Measures for safety and environmental protection b) Measures to reduce radiation exposure of employees c) Appropriateness of maintenance and inspection work d) Measures to increase plant availability e) Measures to improve reliability of equipments and components; and 3) Future Technical Problems

  17. EDITORIAL: Safety aspects of fusion power plants

    Science.gov (United States)

    Kolbasov, B. N.

    2007-07-01

    This special issue of Nuclear Fusion contains 13 informative papers that were initially presented at the 8th IAEA Technical Meeting on Fusion Power Plant Safety held in Vienna, Austria, 10-13 July 2006. Following recommendation from the International Fusion Research Council, the IAEA organizes Technical Meetings on Fusion Safety with the aim to bring together experts to discuss the ongoing work, share new ideas and outline general guidance and recommendations on different issues related to safety and environmental (S&E) aspects of fusion research and power facilities. Previous meetings in this series were held in Vienna, Austria (1980), Ispra, Italy (1983), Culham, UK (1986), Jackson Hole, USA (1989), Toronto, Canada (1993), Naka, Japan (1996) and Cannes, France (2000). The recognized progress in fusion research and technology over the last quarter of a century has boosted the awareness of the potential of fusion to be a practically inexhaustible and clean source of energy. The decision to construct the International Thermonuclear Experimental Reactor (ITER) represents a landmark in the path to fusion power engineering. Ongoing activities to license ITER in France look for an adequate balance between technological and scientific deliverables and complying with safety requirements. Actually, this is the first instance of licensing a representative fusion machine, and it will very likely shape the way in which a more common basis for establishing safety standards and policies for licensing future fusion power plants will be developed. Now that ITER licensing activities are underway, it is becoming clear that the international fusion community should strengthen its efforts in the area of designing the next generations of fusion power plantsโ€”demonstrational and commercial. Therefore, the 8th IAEA Technical Meeting on Fusion Safety focused on the safety aspects of power facilities. Some ITER-related safety issues were reported and discussed owing to their potential

  18. Bohunice Nuclear Power Plant Safety Upgrading Program

    International Nuclear Information System (INIS)

    Toth, A.; Fagula, L.

    1996-01-01

    Bohunice nuclear Power Plant generation represents almost 50% of the Slovak republic electric power production. Due to such high level of commitment to nuclear power in the power generation system, a special attention is given to safe and reliable operation of NPPs. Safety upgrading and operational reliability improvement of Bohunice V-1 NPP was carried out by the Bohunice staff continuously since the plant commissioning. In the 1990 - 1993 period extensive projects were realised. As a result of 'Small Reconstruction of the Bohunice V-1 NPP', the standards of both the nuclear safety and operational reliability have been significantly improved. The implementation of another modifications that will take place gradually during extended refuelling outages and overhauls in the course of 1996 through 1999, is referred to as the Gradual Reconstruction of the Bohunice V-1 Plant. The general goal of the V-1 NPP safety upgrading is the achievement of internationally acceptable level of nuclear safety. Extensive and financially demanding modification process of Bohunice V-2 NPP is likely to be implemented after a completion of the Gradual Reconstruction of the Bohunice V-1 NPP, since the year 1999. With this in mind, a first draft of the strategy of the Bohunice V-2 NPP upgrading program based on Probabilistic Safety assessment consideration was developed. A number of actions with a general effect on Bohunice site safety is evident. All these activities are aimed at reaching the essential objective of Bohunice NPP Management - to ensure a safe, reliable and effective electric energy and heat generation at the Bohunice site. (author)

  19. Standardized safety management of AP1000 nuclear power plant

    International Nuclear Information System (INIS)

    Li Xingwen; Cao Zhiqiang; Cong Jiuyuan

    2011-01-01

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

  20. Safety and licensing of nuclear heating plants

    International Nuclear Information System (INIS)

    Snell, V.G.; Hilborn, J.W.; Lynch, G.F.; McAuley, S.J.

    1989-09-01

    World attention continues to focus on nuclear district heating, a low-cost energy from a non-polluting fuel. It offers long-term security for countries currently dependent on fossil fuels, and can reduce the burden of fossil fuel transportation on railways and roads. Current initiatives encompass large, centralized heating plants and small plants supplying individual institutions. The former are variants of their power reactor cousins but with enhanced safety features. The latter face the safety and licensing challenges of urban siting and remotely monitored operation, through use of intrinsic safety features such as passive decay heat removal, low stored energy and limited reactivity speed and depth in the control systems. Small heating reactor designs are compared, and the features of the SLOWPOKE Energy System, in the forefront of these designs, are summarized. The challenge of public perception must be met by clearly presenting the characteristics of small heating reactors in terms of scale and transparent safety in design and operation, and by explaining the local benefits

  1. Upgrading the safety assessment of exported nuclear power plants

    International Nuclear Information System (INIS)

    Rosen, M.

    1978-01-01

    An examination of the safety aspects of exported nuclear power plants demonstrates that additional and somewhat special considerations exist for these plants, and thus that some new approaches may be required to insure their safety. In view of the generally small regulatory staffs of importing countries, suggestions are given for measures which should be taken by the various organizations involved in the export and import of nuclear power facilities to raise the level of the very essential safety assessment. These include the upgrading of the 'export edition' of the traditionally supplied safety documentation by use of a Supplementary Information Report, written specifically for the needs of a smaller and/or less technically qualified staff, which highlights the differences that exist between the facility to be constructed and the supposedly similar reference plant of the supplier country; by improvement of supporting safety documentation to allow for adequate understanding of significant safety parameters; and by attention to the needs of smaller countries in the critical Operating Regulations (Technical Specifications for Operation). Consideration is also given to upgrading the regulatory effort and to the obligations of principal organizations involved with exported nuclear plants, including national and international, for insuring the importing countries' technical readiness and the adequacy of the regulatory effort. Special attention is directed towards the project contract as a means of implementing programmes to achieve these goals. (author)

  2. A PIP chart for nuclear plant safety

    International Nuclear Information System (INIS)

    Suzuki, Tatsujiro; Yamaoka, Taiji

    1992-01-01

    While it is known that social and political aspects of nuclear safety issues are important, little study has been done on identifying the breadth of stakeholders whose policies have important influences over nuclear plant safety in a comprehensive way. The objectives of this study are to develop a chart that visually identifies important stakeholders and their policies and illustrates these influences in a hierarchical representation so that the relationship between stakeholders and nuclear safety will be better understood. This study is based on a series of extensive interviews with major stakeholders, such as nuclear plant managers, corporate planning vice presidents, state regulators, news media, and public interest groups, and focuses on one US nuclear power plant. Based on the interview results, the authors developed a conceptual policy influence paths (PIP) chart. The PIP chart illustrates the hierarchy of influence among stakeholders. The PIP chart is also useful in identifying possible stakeholders who can be easily overlooked without the PIP chart. In addition, it shows that influence flow is circular rather than linear in one direction

  3. Safety criteria for siting a nuclear power plant

    International Nuclear Information System (INIS)

    2001-01-01

    The guide sets forth requirements for safety of the population and the environment in nuclear power plant siting. It also sets out the general basis for procedures employed by other competent authorities when they issue regulations or grant licences. On request STUK (Radiation and Nuclear Safety Authority of Finland) issues case-specific statements about matters relating to planning and about other matters relating to land use in the environment of nuclear power plants

  4. IAEA Leads Operational Safety Mission to Armenian Nuclear Power Plant

    International Nuclear Information System (INIS)

    2011-01-01

    Full text: An international team of nuclear installation safety experts, led by the International Atomic Energy Agency (IAEA), has reviewed the Armenian Nuclear Power Plant (ANPP) near Metsamor for its safety practices and has noted a series of good practices, as well as recommendations to reinforce them. The IAEA assembled an international team of experts at the request of the Government of the Republic of Armenia to conduct an Operational Safety Review (OSART) of the NPP. Under the leadership of the IAEA's Division of Nuclear Installation Safety, the OSART team performed an in-depth operational safety review from 16 May to 2 June 2011. The team was made up of experts from Finland, France, Lithuania, Hungary, Netherlands, Slovakia, UK, USA, EC and the IAEA. An OSART mission is designed as a review of programmes and activities essential to operational safety. It is not a regulatory inspection, nor is it a design review or a substitute for an exhaustive assessment of the plant's overall safety status. Experts participating in the IAEA's June 2010 International Conference on Operational Safety of Nuclear Power Plants (NPP) reviewed the experience of the OSART programme and concluded: In OSART missions NPPs are assessed against IAEA safety standards which reflect the current international consensus on what constitutes a high level of safety; and OSART recommendations and suggestions are of utmost importance for operational safety improvement of NPPs. Armenia is commended for openness to the international nuclear community and for actively inviting IAEA safety review missions to submit their activities to international scrutiny. Examples of IAEA safety reviews include: Design Safety Review in 2003; Review of Probabilistic Safety Assessment in 2007; and Assessment of Seismic Safety Re-Evaluation in 2009. The team at ANPP conducted an in-depth review of the aspects essential to the safe operation of the plant, which is largely under the control of the site management

  5. Study on 'Safety qualification of process computers used in safety systems of nuclear power plants'

    International Nuclear Information System (INIS)

    Bertsche, K.; Hoermann, E.

    1991-01-01

    The study aims at developing safety standards for hardware and software of computer systems which are increasingly used also for important safety systems in nuclear power plants. The survey of the present state-of-the-art of safety requirements and specifications for safety-relevant systems and, additionally, for process computer systems has been compiled from national and foreign rules. In the Federal Republic of Germany the KTA safety guides and the BMI/BMU safety criteria have to be observed. For the design of future computer-aided systems in nuclear power plants it will be necessary to apply the guidelines in [DIN-880] and [DKE-714] together with [DIN-192]. With the aid of a risk graph the various functions of a system, or of a subsystem, can be evaluated with regard to their significance for safety engineering. (orig./HP) [de

  6. The safety of nuclear power plants in Eastern Europe

    International Nuclear Information System (INIS)

    Hoehn, J.; Niehaus, F.

    1997-01-01

    Nuclear power plant operators and nuclear organizations from the West and from the East cooperate at many levels. The G7 and G24 nations have taken it upon themselves to improve the safety of Eastern nuclear power plants. The European Union has launched support programs, i.e. Technical Assistance to the Commonwealth of Independent States (Tacis) and Pologne-Hangrie: Aide a la Reconstruction Economique (Phare), and founded the European Bank for Reconstruction and Development. The countries of Central and Eastern Europe operate nuclear power plants equipped with VVER-type pressurized water reactors and those equipped with RBMK-type reactors. The safety of these two types of plants is judged very differently. Among the VVER plants, a distinction is made between the older and the more recent 440 MWe lines and the 1000 MWe line. Especially the RBMK plants (Chernobyl-type plants) differ greatly as a function of location and year of construction. Even though they do not meet Western safety standards and at best can be backfitted up to a certain level, it must yet be assumed that they will remain in operation to the end of their projected service lives for economic reasons. (orig.) [de

  7. Modernisation for maintaining and improving safety at Nordic nuclear power plants

    International Nuclear Information System (INIS)

    Hammer, L.; Wahlstroem, B.; Simola, K.

    1998-02-01

    The safety practices in Finland and Sweden are described and compared in regard of effecting modernisation for safety of the nuclear plants in the two countries, considering new technology and advancing safety requirements as proposed for new reactors. Particular attention is given to strategies for applying new safety requirements to reactors built to earlier standards, and to the interplay between the nuclear utilities and the safety authorities. Overviews are given of past and current modernisation of the nuclear power plants in Finland and Sweden. The management procedures in controlling the implementation of modifications to the nuclear power plants are described and discussed in regard of prevailing differences between Finnish and Swedish practices. A formal modelling technique (SADT) was applied for capture of the essential contents of the relevant documented procedures. Two examples of recent plant modifications in the Finnish nuclear plants in Olkiluoto and Loviisa are described and discussed in greater detail. Recommendations are given. (au)

  8. A proposed approach for enhancing design safety assurance of future plants

    International Nuclear Information System (INIS)

    Oh, Kyu Myeng; Ahn, Sang Kyu; Lee, Chang Ju; Kim, Inn Seock

    2010-01-01

    This paper provides various insights from a detailed review of deterministic approaches typically applied to ensure design safety of nuclear power plants (NPPs) and risk-informed approaches proposed to evaluate safety of advanced reactors such as Generation IV reactors. Also considered herein are the risk-informed safety analysis (RISA) methodology suggested by Westinghouse as a means to improve the conventional accident analysis, together with the Technology Neutral Framework recently suggested by the U.S. NRC for safety evaluation of future plants. These insights from the comparative review of deterministic and risk-informed approaches could be used in further enhancing the methodology for design safety assurance of future plants

  9. Organization and conduct of IAEA fire safety reviews at nuclear power plants

    International Nuclear Information System (INIS)

    1998-01-01

    The importance of fire safety in the safe and productive operation of nuclear power plants is recognized worldwide. Lessons learned from experience in nuclear power plants indicate that fire poses a real threat to nuclear safety and that its significance extends far beyond the scope of a conventional fire hazard. With a growing understanding of the close correlation between the fire hazard in nuclear power plants and nuclear safety, backfitting for fire safety has become necessary for a number of operating plants. However, it has been recognized that the expertise necessary for a systematic independent assessment of fire safety of a NPP may not always be available to a number of Member States. In order to assist in enhancing fire safety, the IAEA has already started to offer various services to Member States in the area of fire safety. At the request of a Member State, the IAEA may provide a team of experts to conduct fire safety reviews of varying scope to evaluate the adequacy of fire safety at a specific nuclear power plant during various phases such as construction, operation and decommissioning. The IAEA nuclear safety publications related to fire protection and fire safety form a common basis for these reviews. This report provides guidance for the experts involved in the organization and conduct of fire safety review services to ensure consistency and comprehensiveness of the reviews

  10. INERT Atmosphere confinement operability test procedure

    International Nuclear Information System (INIS)

    RISENMAY, H.R.

    1999-01-01

    This Operability Test Procedure (OTP) provides instructions for testing operability of the Inert Atmosphere Confinement (IAC). The Inert Atmosphere Confinement was designed and built for opening cans of metal items that might have hydrided surfaces. Unreviewed Safety Question (USQ) PFP-97-005 addresses the discovery of suspected plutonium hydride forming on plutonium metal currently stored in the Plutonium Finishing Plant vaults. Plutonium hydride reacts quickly with air, liberating energy. The Inert Atmosphere Confinement was designed to prevent this sudden liberation of energy by opening the material in an inert argon atmosphere instead of the normal glovebox atmosphere. The IAC is located in glovebox HC-21A, room 230B of the 234-5Z Building at the Plutonium Finishing Plant (PFP) in the 200-West Area of the Hanford Site

  11. Results of research into nuclear power plant safety

    International Nuclear Information System (INIS)

    Polak, V.; Hladky, E.; Moravek, J.; Suchomel, J.; Stehlik, J.

    1976-01-01

    A survey is given of computer programmes for the safety analysis of nuclear power plants with WWER type reactors and with fast breeder reactors. The programmes solve accidents in the core, the primary circuit and the containment. A comparison is made of Czechoslovak and foreign computer programmes and their agreement proved. Also studied is the problem of radiation safety of nuclear power plants with regard to the leakage of radioactive isotopes and their detection. (J.B.)

  12. Safety-related incidents at the Finnish nuclear power plants

    International Nuclear Information System (INIS)

    Lehtinen, P.

    1986-03-01

    This report contains detailed descriptions of operating incidents and other safety-related matters at the Finnish nuclear power plants regarded as significant by the regulatory authority, the Finnish Centre for Radiation and Nuclear Safety. In this connection, an account is given of the practical actions caused by the incidents, and their significance to reactor safety is evaluated. The main features of the incidents are also described in the general Quartely Reports, Operation of Finnish Nuclear Power Plants, which are supplemented by this report intended for experts. (author)

  13. Safety-related incidents at the Finnish nuclear power plants

    International Nuclear Information System (INIS)

    Lehtinen, P.

    1985-01-01

    This report contains detailed descriptions of operating incidents and other safety-related matters at the Finnish nuclear power plants regarded as significant by the regulatory authority, the Finnish Centre for Radiation and Nuclear Safety. In this connection, an account is given of the practical actions caused by the incidents, and their significance to reactor safety is evaluated. The main features of the incidents are also described in the general Quartely Reports, Operation of Finnish Nuclear Power Plants, which are supplemented by this report intended for experts. (author)

  14. Integrated plant safety assessment. Systematic evaluation program, Big Rock Point Plant (Docket No. 50-155). Final report

    International Nuclear Information System (INIS)

    1984-05-01

    The Systematic Evaluation Program was initiated in February 1977 by the U.S. Nuclear Regulatory Commission to review the designs of older operating nuclear reactor plants to reconfirm and document their safety. The review provides (1) an assessment of how these plants compare with current licensing safety requirements relating to selected issues, (2) a basis for deciding how these differences should be resolved in an integrated plant review, and (3) a documented evaluation of plant safety when the supplement to the Final Integrated Plant Safety Assessment Report has been issued. This report documents the review of the Big Rock Point Plant, which is one of ten plants reviewed under Phase II of this program. This report indicates how 137 topics selected for review under Phase I of the program were addressed. It also addresses a majority of the pending licensing actions for Big Rock Point, which include TMI Action Plan requirements and implementation criteria for resolved generic issues. Equipment and procedural changes have been identified as a result of the review

  15. IAEA Operational Safety Team Reviews Cattenom Nuclear Power Plant

    International Nuclear Information System (INIS)

    2011-01-01

    Full text: An international team of nuclear installation safety experts led by the International Atomic Energy Agency (IAEA) has reviewed operational safety at France's Cattenom Nuclear Power Plant (NPP) noting a series of good practices as well as recommendations and suggestions to reinforce them. The IAEA assembled an international team of experts at the request of the Government of France to conduct an Operational Safety Review (OSART) of Cattenom NPP. Under the leadership of the IAEA's Division of Nuclear Installation Safety in Vienna, the OSART team performed an in-depth operational safety review of the plant from 14 November to 1 December 2011. The team was made up of experts from Belgium, the Czech Republic, Finland, Germany, Hungary, Japan, Russia, Slovakia, South Africa, Sweden, Ukraine, the United Kingdom and the IAEA. The team at Cattenom conducted an in-depth review of the aspects essential to the safe operation of the NPP, which is largely under the control of the site management. The conclusions of the review are based on the IAEA's Safety Standards. The review covered the areas of Management, Organization and Administration; Training and Qualification; Operations; Maintenance; Technical Support; Operating Experience; Radiation Protection; Chemistry; Emergency Planning and Preparedness; and Severe Accident Management. Cattenom is the first plant in Europe to voluntarily undertake a Severe Accident Management review during an OSART review. The OSART team has identified good plant practices, which will be shared with the rest of the nuclear industry for consideration of their application. Examples include: Sheets are displayed in storage areas where combustible material is present - these sheets are updated readily and accurately by the area owner to ensure that the fire limits are complied with; A simple container is attached to the neutron source handling device to ensure ease and safety of operations and reduce possible radiation exposure during use

  16. PA activity by using nuclear power plant safety demonstration and analysis

    International Nuclear Information System (INIS)

    Tsuchiya, Mitsuo; Kamimae, Rie

    1999-01-01

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

  17. Operational safety performance indicators for nuclear power plants

    International Nuclear Information System (INIS)

    2000-05-01

    Since the late 1980s, the IAEA has been actively sponsoring work in the area of indicators to monitor nuclear power plant (NPP) operational safety performance. The early activities were mainly focused on exchanging ideas and good practices in the development and use of these indicators at nuclear power plants. Since 1995 efforts have been directed towards the elaboration of a framework for the establishment of an operational safety performance indicator programme. The result of this work, compiled in this publication, is intended to assist NPPs in developing and implementing a monitoring programme, without overlooking the critical aspects related to operational safety performance. The framework proposed in this report was presented at two IAEA workshops on operational safety performance indicators held in Ljubljana, Slovenia, in September 1998 and at the Daya Bay NPP, Szenzhen, China, in December 1998. During these two workshops, the participants discussed and brainstormed on the indicator framework presented. These working sessions provided very useful insights and ideas which where used for the enhancement of the framework proposed. The IAEA is acknowledging the support and contribution of all the participants in these two activities. The programme development was enhanced by pilot plant studies. Four plants from different countries with different designs participated in this study with the objective of testing the applicability, usefulness and viability of this approach

  18. Assessment of the overall fire safety arrangements at nuclear power plants

    International Nuclear Information System (INIS)

    1996-01-01

    The present publication has been developed with the help of experts from regulatory, operating and engineering organizations, all with practical experience in the field of fire safety of nuclear power plants. The publication comprises a detailed checklist of the specific elements to be addressed when assessing the adequacy and effectiveness of the overall fire safety arrangements of operating nuclear power plants. The publication will be useful not only to regulators and safety assessors but also to operators and designers. The book addresses a specialized topic outlined in Safety Guide No. 50-SG-D2 (Rev.1), Fire Protection in Nuclear Power Plants, and it is recommended that it be used in conjunction with this Safety Series publication

  19. High temperature reactor module power plant. Plant and safety concept June 1986 - 38.07126.2

    International Nuclear Information System (INIS)

    1986-06-01

    The modular HTR power plant is a universally applicable energy source for the co-generation of electricity, process steam or district heating. The modular HTR concept is characterized by the fact that standardized reactor units with power ratings of 200 MJ/s (so-called modules) can be combined to form power plants with a higher power rating. Consequently the special safety features of small high-temperature reactors (HTR) are also available at higher power plant ratings. The safety features, the technical design and the mode of operation are briefly described in the following, taking a power plant with two HTR-Modules for the co-generation of electricity and process steam as an example. Due to its universal applicability and excellent safety features, the modular HTR power plant is suitable for erection on any site, but particularly on sites near other industrial plants or in densely populated areas. The co-generation of electricity and process steam or district heating with a modular HTR power plant as described here is primarily tailored to the requirements of industrial and communal consumers. The site for such a plant is a typical industrial one. The anticipated features of such sites were taken into consideration in the design of the modular HTR power plant

  20. High temperature reactor module power plant. Plant and safety concept June 1986 - 38.07126.2

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1986-06-15

    The modular HTR power plant is a universally applicable energy source for the co-generation of electricity, process steam or district heating. The modular HTR concept is characterized by the fact that standardized reactor units with power ratings of 200 MJ/s (so-called modules) can be combined to form power plants with a higher power rating. Consequently the special safety features of small high-temperature reactors (HTR) are also available at higher power plant ratings. The safety features, the technical design and the mode of operation are briefly described in the following, taking a power plant with two HTR-Modules for the co-generation of electricity and process steam as an example. Due to its universal applicability and excellent safety features, the modular HTR power plant is suitable for erection on any site, but particularly on sites near other industrial plants or in densely populated areas. The co-generation of electricity and process steam or district heating with a modular HTR power plant as described here is primarily tailored to the requirements of industrial and communal consumers. The site for such a plant is a typical industrial one. The anticipated features of such sites were taken into consideration in the design of the modular HTR power plant.

  1. Status of safety issues at licensed power plants

    International Nuclear Information System (INIS)

    1991-05-01

    As part of ongoing US Nuclear Regulatory Commission (NRC) efforts to ensure the quality and accountability of safety issue information, a program has been established whereby an annual NUREG report will be published on the status of licensee implementation and NRC verification of safety issues in major NRC requirement areas. This report, the second volume of a three-volume series, addresses the status of unresolved safety issues (USIs) at licensed plants. The data contained in these NUREG reports are a product of the NRC's Safety Issues Management System (SIMS) database, which is maintained by the Project Management Staff in the Office of Nuclear Reactor Regulation and by NRC regional personnel. The purpose of this report is to provide a comprehensive description of the status of implementation and verification of the 27 safety issues designated as USIs and to make this information available to other interested parties, including the public. A corollary purpose of this NUREG report is to serve as a follow-on to NUREG-0933, ''A Prioritization of Generic Safety Issues,'' which tracks safety issues up until requirements are approved for imposition at licensed plants. 3 figs., 4 tabs

  2. Adapting a reactor safety assessment system for specific plants

    International Nuclear Information System (INIS)

    Ballard, T.L.; Cordes, G.A.

    1991-01-01

    The Reactor Safety Assessment System (RSAS) is an expert system being developed by the Idaho National Engineering Laboratory, the University of Maryland (UofM) and US Nuclear Regulatory Commission (NRC) for use in the NRC Operations center. RSAS is designed to help the Reactor Safety Team monitor and project core status during an emergency at a licensed nuclear power plant. Analysis uses a hierarchical plant model based on equipment availability and automatically input parametric plant information. There are 3 families of designs of pressurized water reactors and 75 plants using modified versions of the basic design. In order to make an RSAS model for each power plant, a generic model for a given plant type is used with differences being specified by plant specific files. Graphical displays of this knowledge are flexible enough to handle any plant configuration. A variety of tools have been implemented to make it easy to modify a design to fit a given plant while minimizing chance for error. 3 refs., 4 figs

  3. Safety-related occurrences at the Finnish nuclear power plants

    International Nuclear Information System (INIS)

    Reponen, H.; Viitasaari, O.

    1985-04-01

    This report contains detailed descriptions of operating incidents and other safety-related matters at the Finnish nuclear power plants regarded as significant by the regulatory authority, the Finnish Centre for Radiation and Nuclear Safety. In this connection, an account is given of the practical actions caused by the incidents, and their significance to reactor safety is evaluated. The main features of the incidents are also described in the general Quartely Report for this period, Operation of Finnish Nuclear Power Plants (STUK-B-YTO 7), which is supplemented by this report intended for experts. (author)

  4. Operational characteristics of nuclear power plants - modelling of operational safety

    International Nuclear Information System (INIS)

    Studovic, M.

    1984-01-01

    By operational experience of nuclear power plants and realize dlevel of availability of plant, systems and componenst reliabiliuty, operational safety and public protection, as a source on nature of distrurbances in power plant systems and lessons drawn by the TMI-2, in th epaper are discussed: examination of design safety for ultimate ensuring of safe operational conditions of the nuclear power plant; significance of the adequate action for keeping proess parameters in prescribed limits and reactor cooling rquirements; developed systems for measurements detection and monitoring all critical parameters in the nuclear steam supply system; contents of theoretical investigation and mathematical modeling of the physical phenomena and process in nuclear power plant system and components as software, supporting for ensuring of operational safety and new access in staff education process; program and progress of the investigation of some physical phenomena and mathematical modeling of nuclear plant transients, prepared at faculty of mechanical Engineering in Belgrade. (author)

  5. Domestic Regulation for Periodic Safety Review of Nuclear Power Plants

    International Nuclear Information System (INIS)

    Kim, Daesik; Ahn, Seunghoon; Auh, Geunsun; Lee, Jonghyeok

    2015-01-01

    The so-called Periodic Safety Review (PSR) has been carried out such safety assessment throughout its life, on a periodic basis. In January 2001, the Atomic Energy Act and related regulations were amended to adopt the PSR institutional scheme from IAEA Nuclear Safety Guide 50-SG-O12. At that time the safety assessment was made to review the plant safety on 10 safety factors, such as aging management and emergency planning, where the safety factor indicates the important aspects of safety of an operating NPP to be addressed in the PSR. According to this legislation, the domestic utility, the KHNP has conducted the PSR for the operating NPP of 10 years coming up from operating license date, starting since May 2000. Some revisions in the PSR rule were made to include the additional safety factors last year. This paper introduces the current status of the PSR review and regulation, in particular new safety factors and updated technical regulation. Comprehensive safety assessment for Korea Nuclear Power Plants have performed a reflecting design and procedure changes and considering the latest technology every 10 years. This paper also examined the PSR system changes in Korea. As of July 2015, reviews for PSR of 18 units have been completed, with 229 nuclear safety improvement items. And implementation have been completed for 165 of them. PSR system has been confirmed that it has contributed to improvement of plant safety. In addition, this paper examined the PSR system change in Korea

  6. Nuclear power plant safety in Brazil

    International Nuclear Information System (INIS)

    Lederman, L.

    1980-01-01

    The Code of Practice for the Safe Operation of Nuclear Power Plants states that: 'In discharging its responsibility for public health and safety, the government should ensure that the operational safety of a nuclear reactor is subject to surveillance by a regulatory body independent of the operating organization'. In Brazil this task is being carried out by the Comissao Nacional de Energia Nuclear in accordance with the best international practice. (orig./RW)

  7. Code on the safety of nuclear power plants: Governmental organization

    International Nuclear Information System (INIS)

    1988-01-01

    This Code recommends requirements for a regulatory body responsible for regulating the siting, design, construction, commissioning, operation and decommissioning of nuclear power plants for safety. It forms part of the Agency's programme for establishing Codes and Safety Guides relating to land based stationary thermal neutron power plants

  8. Intergrated plant safety assessment. Systematic evaluation program. Palisades plant, Consumers Power Company, Docket No. 50-255. Final report

    International Nuclear Information System (INIS)

    1982-10-01

    The Nuclear Regulatory Commission (NRC) has published its Final Integrated Plant Safety Assessment Report (IPSAR) (NUREG-0820), under the scope of the Systematic Evaluation Program (SEP), for Consumers Power Company's Palisades Plant located in Covert, Van Buren County, Michigan. The SEP was initiated by the NRC to review the design of older operating nuclear reactor plants to reconfirm and document their safety. This report documents the review completed under the SEP for the Palisades Plant. The review has provided for (1) as assessment of the significance of differences between current technical positions on selected safety issues and those that existed when the Palisades Plant was licensed, (2) a basis for deciding on how these differences should be resolved in an integrated plant review, and (3) a documented evaluation of plant safety when all supplements to the Final IPSAR and the Safety Evaluation Report for converting the license from a provisional to a full-term license have been issued. The report also addresses the comments and recommendations made by the Advisory Committee on Reactor Safeguards in connection with its review of the Draft Report, issued in April 1982

  9. Facility Effluent Monitoring Plan for the Plutonium Finishing Plant (PFP); FINAL

    International Nuclear Information System (INIS)

    FRAZIER, T.P.

    1999-01-01

    A facility effluent monitoring plan is required by the U. S. Department of Energy in DOE Order 5400.1 for any operations that involve hazardous materials and radioactive substances that could impact employee or public safety or the environment. This facility effluent monitoring plan assesses effluent monitoring systems and evaluates whether these systems are adequate to ensure the public health and safety as specified in applicable federal, state, and local requirements. To ensure the long-range integrity of the effluent monitoring systems, an update to this facility effluent monitoring plan is required whenever a new process or operation introduces new hazardous materials or significant radioactive materials. This document is reviewed annually even if there are no operational changes, and is updated, at a minimum, every 3 years

  10. General design safety principles for nuclear power plants

    International Nuclear Information System (INIS)

    1986-01-01

    This Safety Guide provides the safety principles and the approach that have been used to implement the Code in the Safety Guides. These safety principles and the approach are tied closely to the safety analyses needed to assist the design process, and are used to verify the adequacy of nuclear power plant designs. This Guide also provides a framework for the use of other design Safety Guides. However, although it explains the principles on which the other Safety Guides are based, the requirements for specific applications of these principles are mostly found in the other Guides

  11. Research on fuzzy comprehensive assessment method of nuclear power plant safety culture

    International Nuclear Information System (INIS)

    Xiang Yuanyuan; Chen Xukun; Xu Rongbin

    2012-01-01

    Considering the traits of safety culture in nuclear plant, 38 safety culture assessment indexes are established from 4 aspects such as safety values, safety institution, safety behavior and safety sub- stances. Based on it, a comprehensive assessment method for nuclear power plant safety culture is constructed by using AHP (Analytic Hierarchy Process) approach and fuzzy mathematics. The comprehensive assessment method has the quality of high precision and high operability, which can support the decision making of safety culture development. (authors)

  12. Improving the safety of LWR power plants. Final report

    International Nuclear Information System (INIS)

    1980-04-01

    This report documents the results of the Study to identify current, potential research issues and efforts for improving the safety of Light Water Reactor (LWR) power plants. This final report describes the work accomplished, the results obtained, the problem areas, and the recommended solutions. Specifically, for each of the issues identified in this report for improving the safety of LWR power plants, a description is provided in detail of the safety significance, the current status (including information sources, status of technical knowledge, problem solution and current activities), and the suggestions for further research and development. Further, the issues are ranked for action into high, medium, and low priority with respect to primarily (a) improved safety (e.g. potential reduction in public risk and occupational exposure), and secondly (b) reduction in safety-related costs

  13. Nuclear power plant safety, the merits of separation

    International Nuclear Information System (INIS)

    Helander, L.I.; Tiren, L.I.

    1977-01-01

    The United States AEC General Design Criteria for Nuclear Power Plants are used worldwide as a basis for the assessment of nuclear plant safety. Several of these criteria require redundancy of safety systems, separation of protection and control systems, consideration of natural phenomena, etc. All these criteria point in one particular direction: the necessity for physically separating the various safety-related systems of a nuclear power plant, particularly with regard to single occurrences that may yield a multiple failure. Requirements in this regard have been amplified by the United States NRC Regulatory Guides and by IEEE Standards. The single occurrence that yields a multiple failure may be, for example, fire, pipe whip, missiles, flooding, hurricanes, or lightning. The paper discusses protection, against the quoted events and others, obtained through applying criteria regarding redundancy and separation of safety-related structures, systems and components. Such criteria affect nuclear plant design in many areas, such as building lay-out, arrangements for fire protection and ventilation, separation of mechanical systems and components, in particular emergency cooling systems, and separation of electric equipment and cables. Implementation of the ensuing design criteria for a BWR power plant is described. This design involves the separation of Emergency Cooling Systems into four 50% Capacity Systems which are independent and separated, including the distribution network for electric power from on-site standby diesel generators and the circuitry for the reactor protection system. The plant is subdivided into a number of fire zones each with its own independent ventilation system. The fire zones are further subdivided into a multitude of fire cells such that redundant subsystems are housed in separate cells. These design precautions with regard to fire are complemented by extensive fire fighting systems

  14. Uranium isotope separation by gaseous diffusion and plant safety

    International Nuclear Information System (INIS)

    Simeon, Claude; Dumas, Maurice.

    1980-07-01

    This report constitutes a safety guide for operators of uranium isotope separation plants, and includes both aspects of safety and protection. Taking into account the complexity of safety problems raised at design and during operation of plants which require specialized guides, this report mainly considers both the protection of man, the environment and goods, and the principles of occupational safety. It does not claim to be comprehensive, but intends to state the general principles, the particular points related to the characteristics of the basic materials and processes, and to set forth a number of typical solutions suitable for various human and technical environments. It is based on the French experience gained during the last fifteen years [fr

  15. Safety issues and their ranking for 'small series' WWER-1000 nuclear power plants. A publication of the extrabudgetary programme on the safety of WWER and RBMK nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-09-01

    This report presents the safety issues in 'small series' WWER-1000 nuclear power plants (NPPs). Safety issues are deviations from current recognized safety practices in design and operation judged to be safety significant by their impact on the plants' defence in depth. This report is intended to serve as reference for the development of plant specific safety improvement programmes and for the evaluation of measures proposed and/or implemented. The identification of safety issues is based on safety studies conducted by the operators of 'small series' WWER-1000 units and by organizations dealing with these reactors, on findings of IAEA safety missions to 'small series' WWER-1000 plants in South Ukraine, at Novovoronezh and Kalinin, and on information obtained from specialists from various countries during an IAEA consultants meeting, 8-12 September 1997 in Vienna, within the framework of the Extra budgetary Programme on the Safety of WWER and RBMK NPPs. Safety issues are first presented according to their impact on the main safety functions and are then described individually. The safety issues are characterized by issue title and specified by issue clarification. Safety issues connected with plant design are followed by the ranking of the issue and ranking justification. Altogether 85 safety issues have been identified, 12 of which are in Category III (defence in depth is insufficient, immediate corrective action is necessary), 38 in Category 11 (defence in depth is degraded, action is needed to resolve the issue) and 22 in Category I (departure from international practices, to be addressed as part of actions to resolve higher priority issues). In the case of operational safety issues (13 safety issues) no ranking is provided as the available material was considered insufficient. For each safety issue, comments and recommendations are made by the IAEA; the status of corresponding measures to improve safety implemented or planned at each site are presented in the

  16. Safety issues and their ranking for 'small series' WWER-1000 nuclear power plants. A publication of the extrabudgetary programme on the safety of WWER and RBMK nuclear power plants

    International Nuclear Information System (INIS)

    2000-09-01

    This report presents the safety issues in 'small series' WWER-1000 nuclear power plants (NPPs). Safety issues are deviations from current recognized safety practices in design and operation judged to be safety significant by their impact on the plants' defence in depth. This report is intended to serve as reference for the development of plant specific safety improvement programmes and for the evaluation of measures proposed and/or implemented. The identification of safety issues is based on safety studies conducted by the operators of 'small series' WWER-1000 units and by organizations dealing with these reactors, on findings of IAEA safety missions to 'small series' WWER-1000 plants in South Ukraine, at Novovoronezh and Kalinin, and on information obtained from specialists from various countries during an IAEA consultants meeting, 8-12 September 1997 in Vienna, within the framework of the Extra budgetary Programme on the Safety of WWER and RBMK NPPs. Safety issues are first presented according to their impact on the main safety functions and are then described individually. The safety issues are characterized by issue title and specified by issue clarification. Safety issues connected with plant design are followed by the ranking of the issue and ranking justification. Altogether 85 safety issues have been identified, 12 of which are in Category III (defence in depth is insufficient, immediate corrective action is necessary), 38 in Category 11 (defence in depth is degraded, action is needed to resolve the issue) and 22 in Category I (departure from international practices, to be addressed as part of actions to resolve higher priority issues). In the case of operational safety issues (13 safety issues) no ranking is provided as the available material was considered insufficient. For each safety issue, comments and recommendations are made by the IAEA; the status of corresponding measures to improve safety implemented or planned at each site are presented in the

  17. Safety strategy and safety analysis of nuclear power plants

    International Nuclear Information System (INIS)

    Franzen, L.F.

    1976-01-01

    The safety strategy for nuclear power plants is characterized by the fact that the high level of safety was attained not as a result of experience, but on the basis of preventive accident analyses and the finding derived from such analyses. Although, in these accident analyses, the deterministic approach is predominant, it is supplemented by reliability analyses. The accidents analyzed in nuclear licensing procedures cover a wide spectrum from minor incidents to the design basis accidents which determine the design of the safety devices. The initial and boundary conditions, which are essentail for accident analyses, and the determination of the loads occurring in various states during regular operation and in accidents flow into the design of the individual systems and components. The inevitable residual risk and its origins are discussed. (orig.) [de

  18. Evaluation of seismic hazards for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2002-01-01

    The main objective of this Safety Guide is to provide recommendations on how to determine the ground motion hazards for a plant at a particular site and the potential for surface faulting, which could affect the feasibility of construction and safe operation of a plant at that site. The guidelines and procedures presented in this Safety Guide can appropriately be used in evaluations of site suitability and seismic hazards for nuclear power plants in any seismotectonic environment. The probabilistic seismic hazard analysis recommended in this Safety Guide also addresses the needs for seismic hazard analysis of external event PSAs conducted for nuclear power plants. Many of the methods and processes described may also be applicable to nuclear facilities other than power plants. Other phenomena of permanent ground displacement (liquefaction, slope instability, subsidence and collapse) as well as the topic of seismically induced flooding are treated in Safety Guides relating to foundation safety and coastal flooding. Recommendations of a general nature are given in Section 2. Section 3 discusses the acquisition of a database containing the information needed to evaluate and address all hazards associated with earthquakes. Section 4 covers the use of this database for construction of a seismotectonic model. Sections 5 and 6 review ground motion hazards and evaluations of the potential for surface faulting, respectively. Section 7 addresses quality assurance in the evaluation of seismic hazards for nuclear power plants

  19. Safety culture in nuclear power plants. Proceedings; Sicherheitskultur im Kernkraftwerk. Seminarbericht

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1994-12-01

    As a consequence of the INSAG-4 report on `safety culture`, published by the IAEA in 1991, the Federal Commission for the Safety of Nuclear Power Plants (KSA) decided to hold a one-day seminar as a first step in this field. The KSA is an advisory body of the Federal Government and the Federal Department of Transport and Energy (EVED). It comments on applications for licenses, observes the operation of nuclear power plants, assists with the preparation of regulations, monitors the progress of research in the field of nuclear safety, and makes proposals for research tasks. The objective of this seminar was to familiarise the participants with the principles of `safety culture`, with the experiences made in Switzerland and abroad with existing concepts, as well as to eliminate existing prejudices. The main points dealt with at this seminar were: - safety culture from the point of view of operators, - safety culture from the point of view of the authorities, - safety culture: collaboration between power plants, the authorities and research organisations, - trends and developments in the field of safety culture. Invitations to attend this seminar were extended to the management boards of companies operating Swiss nuclear power plants, and to representatives of the Swiss authorities responsible for the safety of nuclear power plants. All these organisations were represented by a large number of executive and specialist staff. We would like to express our sincerest thanks to the Head of the Federal Department of Transport and Energy for his kind patronage of this seminar. (author) figs., tabs., refs.

  20. Safety-related occurrences at the Finnish nuclear power plants

    International Nuclear Information System (INIS)

    Viitasaari, O.; Rantavaara, A.

    1984-03-01

    This report contains detailed descriptions of operating incidents and other safety-related matters at the Finnish nuclear power plants regarded as significant by the regulatory authority, the Finnish Centre for Radiation and Nuclear Safety. In this connection, an account is given of the practical actions caused by the incidents, and their significance to reactor safety is evaluated. The main features of the incidents are also described in the general Quartely Report for this period, Operation of Finnish Nuclear Power Plants (STL-B-RTO-83/7), which is supplemented by this report intended principally for experts. (author)

  1. Evaluation of the safety of the operating nuclear power plants built to earlier standards

    International Nuclear Information System (INIS)

    Menteseoglu, S.

    2001-01-01

    The objective of this paper is to provide practical assistance on judging the safety of a nuclear power plant, on the basis of a comparison with current safety standards and operational practices. For nuclear power plants built to earlier standards for which there are questions about the adequacy of the maintenance of the plant design and operational practices, a safety review against current standards and practices can be considered a high priority. The objective of reviewing nuclear power plants built to earlier standards against current standards and practices is to determine whether there are any deviations which would have an impact on plant safety. The safety significance of the issues identified should be judged according to their implications for plant design and operation in terms of basic safety concepts such as defence in depth and safety culture. In addition, this paper provides assistance on the prioritization of corrective measures and their implementation so as to approach an acceptable level of safety

  2. IAEA Leads Operational Safety Mission to Muehleberg Nuclear Power Plant

    International Nuclear Information System (INIS)

    2012-01-01

    Full text: An international team of nuclear safety experts led by the International Atomic Energy Agency today concluded a review of the safety practices at the Muehleberg Nuclear Power Plant (NPP) near Bern in Switzerland. The team noted a series of good practices and made recommendations and suggestions to reinforce them. The IAEA assembled the Operational Safety Review Team at the request of the Swiss government. The team, led by the IAEA's Division of Nuclear Installation Safety, performed an in-depth operational safety review from 8 to 25 October 2012. The team comprised experts from Belgium, the Czech Republic, Finland, Germany, Hungary, Slovakia, Sweden, the United Kingdom and the United States as well as experts from the IAEA. The team conducted an in-depth review of the aspects essential to the safe operation of the Muehleberg NPP. The conclusions of the review are based on the IAEA's Safety Standards and proven good international practices. The review covered the areas of Management, Organization and Administration; Training; Operations; Maintenance; Technical Support; Operating Experience; Radiation Protection; Chemistry, Emergency Planning and Preparedness, Severe Accident Management and Long-Term Operation. The OSART team made 10 recommendations and 11 suggestions related to areas where operations of Muehleberg NPP could be further improved, for example: - Plant management could improve the operating experience program and methods throughout the plant to ensure corrective actions are taken in a timely manner; - In the area of Long-Term Operation, the ageing management review for some systems and components is not complete and the environmental qualification of originally installed safety cables has not yet been revalidated for long-term operation; and - The plant provisions for the protection of persons on the site during an emergency with radioactive release can be improved to minimize health risks to plant personnel. The team also identified 10 good

  3. Safety criteria of uranium enrichment plants

    International Nuclear Information System (INIS)

    Nardocci, A.C.; Oliveira Neto, J.M. de

    1994-01-01

    The applicability of nuclear reactor safety criteria applied to uranium enrichment plants is discussed, and a new criterion based on the soluble uranium compounds and hexafluoride chemical toxicities is presented. (L.C.J.A.). 21 refs, 4 tabs

  4. Recommended general safety requirements for nuclear power plants

    International Nuclear Information System (INIS)

    1983-06-01

    This report presents recommendations for a set of general safety requirements that could form the basis for the licensing of nuclear power plants by the Atomic Energy Control Board. In addition to a number of recommended deterministic requirements the report includes criteria for the acceptability of the design of such plants based upon the calculated probability and consequence (in terms of predicted radiation dose to members of the public) of potential fault sequences. The report also contains a historical review of nuclear safety principles and practices in Canada

  5. Innovative Modelling Approach of Safety Culture Assessment in Nuclear Power Plant

    International Nuclear Information System (INIS)

    Ahn, N.

    2016-01-01

    A culture is commonly defined as the shared set of norms and values that govern appropriate individual behavior. Safety culture is the subset of organizational culture that reflects the general attitude and approaches to safety and risk management. While safety is sometimes narrowly defined in terms of human death and injury, we use a more inclusive definition that also considers mission loss as a safety problem and is thus applicable to nuclear power plants and missions. The recent accident reports and investigations of the nuclear power plant mission failures (i.e., TMI, Chernobyl, and Fukushima) point to safety cultural problems in nuclear power plants. Many assessment approaches have been developed by organizations such as IAEA and INPO based on the assessment of parameters at separate levels โ€” individuals, groups, and organizations.

  6. Safety-related decision making at a nuclear power plant

    International Nuclear Information System (INIS)

    Vaurio, J.K.

    1998-01-01

    The decision making environment of an operating nuclear power plant is presented. The organizations involved, their roles and interactions as well as the main influencing factors and decision criteria are described. The focus is on safety-related decisions, and the framework is based on the situation at Loviisa power station. The role of probabilistic safety assessment (PSA) is illustrated with decisions concerning plant modifications, optimization, acceptance of temporary configurations and extended repair times. Suggestions are made for rational and flexible risk-based control of allowed times to operate the plant with some components out of service. (orig.)

  7. Analysis of effect of safety classification on DCS design in nuclear power plants

    International Nuclear Information System (INIS)

    Gou Guokai; Li Guomin; Wang Qunfeng

    2011-01-01

    By analyzing the safety classification for the systems and functions of nuclear power plants based on the general design requirements for nuclear power plants, especially the requirement of availability and reliability of I and C systems, the characteristics of modem DCS technology and I and C products currently applied in nuclear power field are interpreted. According to the requirements on the safety operation of nuclear power plants and the regulations for safety audit, the effect of different safety classifications on DCS design in nuclear power plants is analyzed, by considering the actual design process of different DCS solutions in the nuclear power plants under construction. (authors)

  8. Safety philosophy for nuclear power plants in egypt

    International Nuclear Information System (INIS)

    Mervat, S.A.; Hammad, F.H.

    1988-01-01

    This work establishes the basic principles of a safety philosophy for nuclear power plants in egypt. A number of deterministic requirements stemming the multiple barriers and the defense-in-depth concept are emphasised. other requirements in the areas of siting, operational safety, safety analysis, special issues, and experience feedback are also identified. The role of international cooperation in nuclear safety technology-transfer and nuclear emergencies is highlighted. In addition probabilistic ally based guidelines are set for acceptable risk and dose limits

  9. Safety issues and their ranking for WWER-1000 model 320 nuclear power plants. A publication of the extrabudgetary programme on the safety of WWER and RBMK nuclear power plants

    International Nuclear Information System (INIS)

    1996-03-01

    The objective of this report is to present a consolidated list of safety deficiencies, called safety issues, ranked according to their safety significance and the corrective measures to improve overall safety. It is intended for use as a reference to facilitate the development of plant specific safety improvement programmes and to serve as a basis for reviewing their implementation. To the extent that information was made available to the IAEA, the country/plant specific status with respect to each safety issue is described. Section 2 provides an overview of the impact of the relevant issues on the main safety functions in different operational conditions and other aspects important to overall plant safety. A summary of the safety issues and their respective ranking is given in Tables 1 and 2 at the end of Section 2. Section 3 deals with individual safety issues identified in the design which are presented according to the structure below. Section 4 presents the safety issues related to operational safety according to a similar structure but without the ranking. 73 refs, 3 tabs

  10. Safety issues and their ranking for WWER-1000 model 320 nuclear power plants. A publication of the extrabudgetary programme on the safety of WWER and RBMK nuclear power plants

    International Nuclear Information System (INIS)

    1997-04-01

    The objective of this report is to present a consolidated list of safety deficiencies, called safety issues, ranked according to their safety significance and the corrective measures to improve overall safety. It is intended for use as a reference to facilitate the development of plant specific safety improvement programmes and to serve as a basis for reviewing their implementation. To the extent that information was made available to the IAEA, the country/plant specific status with respect to each safety issue is described. Section 2 provides an overview of the impact of the relevant issues on the main safety functions in different operational conditions and other aspects important to overall plant safety. A summary of the safety issues and their respective ranking is given in Tables 1 and 2 at the end of Section 2. Section 3 deals with individual safety issues identified in the design which are presented according to the structure below. Section 4 presents the safety issues related to operational safety according to a similar structure but without the ranking

  11. Study on a quantitative evaluation method of equipment maintenance level and plant safety level for giant complex plant system

    International Nuclear Information System (INIS)

    Aoki, Takayuki

    2010-01-01

    In this study, a quantitative method on maintenance level which is determined by the two factors, maintenance plan and field work implementation ability by maintenance crew is discussed. And also a quantitative evaluation method on safety level for giant complex plant system is discussed. As a result of consideration, the following results were obtained. (1) It was considered that equipment condition after maintenance work was determined by the two factors, maintenance plan and field work implementation ability possessed by maintenance crew. The equipment condition determined by the two factors was named as 'equipment maintenance level' and its quantitative evaluation method was clarified. (2) It was considered that CDF in a nuclear power plant, evaluated by using a failure rate counting the above maintenance level was quite different from CDF evaluated by using existing failure rates including a safety margin. Then, the former CDF was named as 'plant safety level' of plant system and its quantitative evaluation method was clarified. (3) Enhancing equipment maintenance level means an improvement of maintenance quality. That results in the enhancement of plant safety level. Therefore, plant safety level should be always watched as a plant performance indicator. (author)

  12. Safety related terms for advanced nuclear plants

    International Nuclear Information System (INIS)

    1995-12-01

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

  13. Safety related terms for advanced nuclear plants

    International Nuclear Information System (INIS)

    1991-09-01

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

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

  16. Guidelines for the Layout and Contents of Safety Reports for Stationary Nuclear Power Plants

    International Nuclear Information System (INIS)

    1970-01-01

    The purpose of the present document is to suggest guidelines for the organization and contents of the Safety Reports which support the request for authorization to construct and operate a nuclear power plant incorporating one or more reactors. Safety Reports represent the principal communication between the applicant and the Regulatory Body, as outlined in the Code of Practice for the Safe Operation of Nuclear Power Plants. It should be understood that these Safety Reports will be a valuable document for the applicant. They should contain, therefore, precise information on the plant and its operating conditions. The writing of Safety Reports should be considered an opportunity to enhance the safety of the plant and its operating conditions. Their main purpose is to provide information to permit the assessment of the nuclear safety implications which may arise from the establishment of the plant at the chosen site with due consideration to the health and safety of the general public and the operating personnel. Safety Reports should include information such as design bases, site and plant characteristics, limits and conditions, conduct of operation and safety analyses, in such way that the Regulatory Body may be able to evaluate the safety of the plant. The applicant should consider the present guidelines as a series of recommendations to be interpreted according to each specific case.

  17. Design of the reactor coolant system and associated systems in nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2008-01-01

    This Safety Guide was prepared under the IAEA programme for establishing safety standards for nuclear power plants. The basic requirements for the design of safety systems for nuclear power plants are established in the Safety Requirements publication, Safety Standards Series No. NS-R-1 on Safety of Nuclear Power Plants: Design, which it supplements. This Safety Guide describes how the requirements for the design of the reactor coolant system (RCS) and associated systems in nuclear power plants should be met. 1.2. This publication is a revision and combination of two previous Safety Guides, Safety Series No. 50-SG-D6 on Ultimate Heat Sink and Directly Associated Heat Transport Systems for Nuclear Power Plants (1981), and Safety Series No. 50-SG-D13 on Reactor Coolant and Associated Systems in Nuclear Power Plants (1986), which are superseded by this new Safety Guide. 1.3. The revision takes account of developments in the design of the RCS and associated systems in nuclear power plants since the earlier Safety Guides were published in 1981 and 1986, respectively. The other objectives of the revision are to ensure consistency with Ref., issued in 2000, and to update the technical content. In addition, an appendix on pressurized heavy water reactors (PHWRs) has been included

  18. Applications of computer based safety systems in Korea nuclear power plants

    International Nuclear Information System (INIS)

    Won Young Yun

    1998-01-01

    With the progress of computer technology, the applications of computer based safety systems in Korea nuclear power plants have increased rapidly in recent decades. The main purpose of this movement is to take advantage of modern computer technology so as to improve the operability and maintainability of the plants. However, in fact there have been a lot of controversies on computer based systems' safety between the regulatory body and nuclear utility in Korea. The Korea Institute of Nuclear Safety (KINS), technical support organization for nuclear plant licensing, is currently confronted with the pressure to set up well defined domestic regulatory requirements from this aspect. This paper presents the current status and the regulatory activities related to the applications of computer based safety systems in Korea. (author)

  19. Integrated safety assessment report, Haddam Neck Plant (Docket No. 50-213): Integrated Safety Assessment Program: Draft report

    International Nuclear Information System (INIS)

    1987-07-01

    The integrated assessment is conducted on a plant-specific basis to evaluate all licensing actions, licensee initiated plant improvements and selected unresolved generic/safety issues to establish implementation schedules for each item. Procedures allow for a periodic updating of the schedules to account for licensing issues that arise in the future. The Haddam Neck Plant is one of two plants being reviewed under the pilot program. This report indicates how 82 topics selected for review were addressed, and presents the staff's recommendations regarding the corrective actions to resolve the 82 topics and other actions to enhance plant safety. 135 refs., 4 figs., 5 tabs

  20. [Safety assessment of foods derived from genetically modified plants].

    Science.gov (United States)

    Pรถting, A; Schauzu, M

    2010-06-01

    The placing of genetically modified plants and derived food on the market falls under Regulation (EC) No. 1829/2003. According to this regulation, applicants need to perform a safety assessment according to the Guidance Document of the Scientific Panel on Genetically Modified Organisms of the European Food Safety Authority (EFSA), which is based on internationally agreed recommendations. This article gives an overview of the underlying legislation as well as the strategy and scientific criteria for the safety assessment, which should generally be based on the concept of substantial equivalence and carried out in relation to an unmodified conventional counterpart. Besides the intended genetic modification, potential unintended changes also have to be assessed with regard to potential adverse effects for the consumer. All genetically modified plants and derived food products, which have been evaluated by EFSA so far, were considered to be as safe as products derived from the respective conventional plants.

  1. Safety and regulatory requirements of nuclear power plants

    International Nuclear Information System (INIS)

    Kumar, S.V.; Bhardwaj, S.A.

    2000-01-01

    A pre-requisite for a nuclear power program in any country is well established national safety and regulatory requirements. These have evolved for nuclear power plants in India with participation of the regulatory body, utility, research and development (R and D) organizations and educational institutions. Prevailing international practices provided a useful base to develop those applicable to specific system designs for nuclear power plants in India. Their effectiveness has been demonstrated in planned activities of building up the nuclear power program as well as with unplanned activities, like those due to safety related incidents etc. (author)

  2. Safety and operation of the Stade nuclear power plant

    International Nuclear Information System (INIS)

    Salcher, H.

    1991-01-01

    The concept of PreussenElektra is to continuously increase the existing safety standard of the Stade nuclear power station using experience gained from faults and operation in nuclear power stations and the progressive state of the art. Modifications to achieve the most gentle operation of the plant have been completed and other are on-going. To do so instruments were attached to those components which are susceptible to fatigue to record the transients and extensive calculatory records were kept. Although the plant has almost 20 years successful operation behind it, it can still stand up well to comparisons with more recent plants as far as safety aspects are concerned. 6 figs

  3. Rock siting of nuclear power plants from a reactor safety standpoint

    International Nuclear Information System (INIS)

    1975-11-01

    The study has aimed at surveying the advantages and disadvantages of a rock sited nuclear power plant from a reactor safety standpoint. The studies performed are almost entirely concentrated on the BWR alternative. The design of a nuclear power plant in rock judged most appropriate has been studied in greater detail, and a relatively extensive safety analysis has been made. It is found that the presented technical design of the rock sited alternative is sufficiently advanced to form a basis for further projecting treatment. The chosen technical design of the reactor plant demands a cavern with a 45-50 metre span. Caverns without strengthening efforts with such spans are used in mines, but have no previously been used for industrial plants. Studies of the stability of such caverns show that a safety level is attainable corresponding to the safety required for the other parts of the nuclear power plant. The conditions are that the rock is of high quality, that necessary strengthening measures are taken and that careful studies of the rock are made before and during the blasting, and also during operation of the plant. When locating a rock sited nuclear power plant, the same criteria must be considered as for an above ground plant, with additional stronger demands for rock quality. The presented rock sited nuclear power plant has been assessed to cost 20 % more in total construction costs than a corresponding above ground plant. The motivations for rock siting also depend on whether a condensing plant for only electricity production, or a plant for combined power production and district heating, is considered. The latter would under certain circumstances make rock siting look more attractive. (author)

  4. Waste Isolation Pilot Plant Safety Analysis Report

    International Nuclear Information System (INIS)

    1995-11-01

    The following provides a summary of the specific issues addressed in this FY-95 Annual Update as they relate to the CH TRU safety bases: Executive Summary; Site Characteristics; Principal Design and Safety Criteria; Facility Design and Operation; Hazards and Accident Analysis; Derivation of Technical Safety Requirements; Radiological and Hazardous Material Protection; Institutional Programs; Quality Assurance; and Decontamination and Decommissioning. The System Design Descriptions'' (SDDS) for the WIPP were reviewed and incorporated into Chapter 3, Principal Design and Safety Criteria and Chapter 4, Facility Design and Operation. This provides the most currently available final engineering design information on waste emplacement operations throughout the disposal phase up to the point of permanent closure. Also, the criteria which define the TRU waste to be accepted for disposal at the WIPP facility were summarized in Chapter 3 based on the WAC for the Waste Isolation Pilot Plant.'' This Safety Analysis Report (SAR) documents the safety analyses that develop and evaluate the adequacy of the Waste Isolation Pilot Plant Contact-Handled Transuranic Wastes (WIPP CH TRU) safety bases necessary to ensure the safety of workers, the public and the environment from the hazards posed by WIPP waste handling and emplacement operations during the disposal phase and hazards associated with the decommissioning and decontamination phase. The analyses of the hazards associated with the long-term (10,000 year) disposal of TRU and TRU mixed waste, and demonstration of compliance with the requirements of 40 CFR 191, Subpart B and 40 CFR 268.6 will be addressed in detail in the WIPP Final Certification Application scheduled for submittal in October 1996 (40 CFR 191) and the No-Migration Variance Petition (40 CFR 268.6) scheduled for submittal in June 1996. Section 5.4, Long-Term Waste Isolation Assessment summarizes the current status of the assessment

  5. Waste Isolation Pilot Plant Safety Analysis Report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-11-01

    The following provides a summary of the specific issues addressed in this FY-95 Annual Update as they relate to the CH TRU safety bases: Executive Summary; Site Characteristics; Principal Design and Safety Criteria; Facility Design and Operation; Hazards and Accident Analysis; Derivation of Technical Safety Requirements; Radiological and Hazardous Material Protection; Institutional Programs; Quality Assurance; and Decontamination and Decommissioning. The System Design Descriptions`` (SDDS) for the WIPP were reviewed and incorporated into Chapter 3, Principal Design and Safety Criteria and Chapter 4, Facility Design and Operation. This provides the most currently available final engineering design information on waste emplacement operations throughout the disposal phase up to the point of permanent closure. Also, the criteria which define the TRU waste to be accepted for disposal at the WIPP facility were summarized in Chapter 3 based on the WAC for the Waste Isolation Pilot Plant.`` This Safety Analysis Report (SAR) documents the safety analyses that develop and evaluate the adequacy of the Waste Isolation Pilot Plant Contact-Handled Transuranic Wastes (WIPP CH TRU) safety bases necessary to ensure the safety of workers, the public and the environment from the hazards posed by WIPP waste handling and emplacement operations during the disposal phase and hazards associated with the decommissioning and decontamination phase. The analyses of the hazards associated with the long-term (10,000 year) disposal of TRU and TRU mixed waste, and demonstration of compliance with the requirements of 40 CFR 191, Subpart B and 40 CFR 268.6 will be addressed in detail in the WIPP Final Certification Application scheduled for submittal in October 1996 (40 CFR 191) and the No-Migration Variance Petition (40 CFR 268.6) scheduled for submittal in June 1996. Section 5.4, Long-Term Waste Isolation Assessment summarizes the current status of the assessment.

  6. Use of operational experience in fire safety assessment of nuclear power plants

    International Nuclear Information System (INIS)

    2000-01-01

    Fire hazard has been identified as a major contributor to a plant's operational risk and the international nuclear power industry has been studying and developing tools for defending against this hazard. Considerable progress in design and regulatory requirements for fire safety, in fire protection technology and in related analytical techniques has been made in the past two decades. Substantial efforts have been undertaken worldwide to implement these advances in the interest of improving fire safety both at new and existing nuclear power plants. To assist in these efforts, the IAEA initiated a programme on fire safety that was intended to provide assistance to Member States in improving fire safety in nuclear power plants. In order to achieve this general objective, the IAEA programme aimed at the development of guidelines and good practices, the promotion of advanced fire safety assessment techniques, the exchange of state of the art information between practitioners and the provision of engineering safety advisory services and training in the implementation of internationally accepted practices. During the period 1993-1994, the IAEA activities related to fire safety concentrated on the development of guidelines and good practice documents related to fire safety and fire protection of operating plants. One of the first tasks was the development of a Safety Guide that formulates specific requirements with regard to the fire safety of operating nuclear power plants. Several documents, which provide advice on fire safety inspection, were developed to assist in its implementation. In the period 1995-1996, the programme focused on the preparation of guidelines for the systematic analysis of fire safety at nuclear power plants (NPPs). The IAEA programme on fire safety for 1997-1998 includes tasks aimed at promoting systematic assessment of fire safety related occurrences and dissemination of essential insights from this assessment. One of the topics addressed is the

  7. Safety criteria for nuclear chemical plants

    International Nuclear Information System (INIS)

    Ball, P.W.; Curtis, L.M.

    1983-01-01

    Safety measures have always been required to limit the hazards due to accidental release of radioactive substances from nuclear power plants and chemical plants. The risk associated with the discharge of radioactive substances during normal operation has also to be kept acceptably low. BNFL (British Nuclear Fuels Ltd.) are developing risk criteria as targets for safe plant design and operation. The numerical values derived are compared with these criteria to see if plants are 'acceptably safe'. However, the criteria are not mandatory and may be exceeded if this can be justified. The risk assessments are subject to independent review and audit. The Nuclear Installations Inspectorate also has to pass the plants as safe. The assessment principles it uses are stated. The development of risk criteria for a multiplant site (nuclear chemical plants tend to be sited with many others which are related functionally) is discussed. This covers individual members of the general public, societal risks, risks to the workforce and external hazards. (U.K.)

  8. The effect of management and organizational structure on nuclear power plant safety

    International Nuclear Information System (INIS)

    Thurber, J.A.

    1986-01-01

    Many informed observers have proposed that utility management is a key element underlying the safe operation of nuclear power plants (NPP). One way that management likely influences plant safety performance is through the organizational structures it consciously creates or allows to exist. This paper describes an empirical analysis of the relationships between some important dimensions of plant organizational structure and measures of plant safety performance

  9. A new approach to preparing safety cases for existing nuclear plant (COSR)

    International Nuclear Information System (INIS)

    Rice, S.A.; Buchan, A.B.

    2000-01-01

    BNFL is committed to achieving world class safety performance, through a process of continuously reviewing and improving its safety practices. In the mid 1990s, as part of this process, the company began to develop a new type of safety case, for existing non-reactor nuclear plants, called the continued operation safety report (COSR). Following a significant amount of development work from experts within BNFL and important contributions from its regulators, the first approved COSR was recently completed and submitted to the Nuclear Installations Inspectorate. The COSR aims to provide a visibly integrated safety and engineering case for the adequacy of continued operation of a nuclear facility. It achieves this by identifying the main plant structures, systems and components that have a safety function and provides the appropriate supporting engineering substantiation. The COSR aims to explore plant safety and identify worthwhile improvements. The document also aims to be reader-friendly by focusing on the main safety issues. It is therefore a slim safety summary which provides operators, safety specialists and regulators with an overview and introduction into the broader, more detailed safety case. This paper provides an overview of the COSR and its production process, describing the safety case improvements that have been made by comparing it to its predecessor, the fully developed safety case. The paper also illustrates the benefits of the COSR by providing current examples of its application on existing BNFL plant. Finally, the paper describes ongoing development work aimed at further improving the COSR and its production process. (author)

  10. The working of RVNRL pilot plant of Rubber Board and it's safety devices

    International Nuclear Information System (INIS)

    Britto, I.J.; Thomas, E.V.

    1996-01-01

    A pilot plant for producing radiation vulcanized natural rubber latex (RVNRL) was established at Rubber Board, India in 1992. Irradiation is done by a batch process in the plant. The plant has a versatile safety system for safety of operators and people working in and around the plant

  11. Operational safety review programmes for nuclear power plants. Guidelines for assessment

    International Nuclear Information System (INIS)

    2002-01-01

    The IAEA has been offering the Operational Safety Review Team (OSART) programme to provide advice and assistance to Member States in enhancing the operational safety of nuclear power plants (NPPs). Simultaneously, the IAEA has encouraged self-assessment and review by Member States of their own nuclear power plants to continuously improve nuclear safety. Currently, some utilities have been implementing safety review programmes to independently review their own plants. Corporate or national operational safety review programmes may be compliance or performance based. Successful utilities have found that both techniques are necessary to provide assurance that (i) as a minimum the NPP meets specific corporate and legal requirements and (ii) management at the NPP is encouraged to pursue continuous improvement principles. These programmes can bring nuclear safety benefits to the plants and utilities. The IAEA has conducted two pilot missions to assess the effectiveness of the operational review programme. Based on these missions and on the experience gained during OSART missions, this document has been developed to provide guidance on and broaden national/corporate safety review programmes in Member States, and to assist in maximizing their benefits. These guidelines are intended primarily for the IAEA team to conduct assessment of a national/corporate safety review programme. However, this report may also be used by a country or utility to establish its own national/corporate safety review programme. The guidelines may likewise be used for self-assessment or for establishing a baseline when benchmarking other safety review programmes. This report consists of four parts. Section 2 addresses the planning and preparation of an IAEA assessment mission and Sections 3 and 4 deal with specific guidelines for conducting the assessment mission itself

  12. Overview of IAEA guidelines for fire safety inspection and operation in nuclear power plants

    International Nuclear Information System (INIS)

    Mowrer, D.S.

    1998-01-01

    In 1992, the International Atomic Energy Agency began an ambitious project on fire safety in nuclear power plants. The purpose of this ongoing project is to provide specific guidance on compliance with the requirements set forth through the IAEA Nuclear Safety Standards program established in 1974. The scope of the Fire Safety project encompasses several tasks, including the development of new standards and guidelines to assist Member States in assessing the level of fire safety in existing plants. Five new Safety Practices, one new Safety Guide and a Technical Document have been developed for use by the fire safety community. The primary intent of these new documents is to provide detailed guidance and a consistent format for the assessment of the overall level of fire safety being provided in existing nuclear power plants around the world and especially in developing countries. Sufficient detail is provided in the Safety Guide and Safety Practices to allow technically knowledgeable plant personnel, outside consultants or other technical experts to assess the adequacy of fire safety within the plant facilities. This paper describes topics addressed by each of the IAEA Fire Safety documents and discussed the relationship of each document to others in the series. (author)

  13. Periodic safety review of operational nuclear power plants. A publication within the NUSS programme

    International Nuclear Information System (INIS)

    1994-01-01

    This Safety Guide which supplements the IAEA Safety Fundamentals: The Safety of Nuclear Installations and the Code on the Safety of Nuclear Power Plants: Operation, forms part of the Agency's programme, referred to as the NUSS programme, for establishing Codes and Guides relating to nuclear power plants. A list of NUSS publications is given at the end of this book. This Guide was drafted on the basis of a systematic review approach that was endorsed by the IAEA Conference on the Safety of Nuclear Power: Strategy for the Future. The purpose of this Safety Guide is to provide guidance on the conduct of Periodic Safety Reviews (PSRs) for an operational nuclear power plant. The Guide is directed at both owners/operators and regulators. This Safety Guide deals with the PSR of an operational nuclear power plant. A PSR is a comprehensive safety review addressing all important aspects of safety, carried out at regular intervals. 22 refs, 4 figs

  14. Trends in safety objectives for nuclear district heating plants

    Energy Technology Data Exchange (ETDEWEB)

    Brogli, R [Paul Scherrer Inst., Villigen (Switzerland)

    1997-09-01

    Safety objectives for dedicated nuclear heating plants are strongly influenced on the one hand by what is accepted for electricity nuclear stations, and on the other hand by the requirement that for economical reasons heating reactors have to be located close to population centers. The paper discusses the related trends and comes to the conclusion that on account of the specific technical characteristics of nuclear heating plants adequate safety can be provided even for highly populated sites. (author). 8 refs.

  15. Female PFP patients present alterations in eccentric muscle activity but not the temporal order of activation of the vastus lateralis muscle during the single leg triple hop test.

    Science.gov (United States)

    Kalytczak, Marcelo Martins; Lucareli, Paulo Roberto Garcia; Dos Reis, Amir Curcio; Bley, Andrรฉ Serra; Biasotto-Gonzalez, Daniela Aparecida; Correa, Joรฃo Carlos Ferrari; Politti, Fabiano

    2018-04-07

    This study aimed to compare the concentric and eccentric activity and the temporal order of peak activity of the hip and knee muscles between women with patellofemoral pain (PFP) and healthy women during the single leg triple hop test (SLTHT). Electromyographic (EMG) and Kinematic data were collected from 14 healthy women (CG) and 14 women diagnosed with PFP (PFG) during a single session of the single leg triple hop test. Integral surface electromyography (iEMG) data of the hip and knee muscles in eccentric and concentric phases and the length of time that each muscle needed to reach the maximal peak of muscle activity were calculated. The iEMG in the eccentric phase was significantly higher (pโ€ฏ<โ€ฏ0.05) than the concentric phase, for the gluteus maximus and gluteus medius muscles (CG and PFG) and for the vastus lateralis muscle (PFG). The vastus lateralis muscle was the first muscle to reach the highest peak of activity in the PFG, and the third to reach this peak in the CG. In the present study, the activity of the vastus lateralis muscle during the eccentric phase of the jump was greater than concentric phase, as a temporal anticipation of its peak in activity among women with PFP. Copyright ยฉ 2018 Elsevier B.V. All rights reserved.

  16. Studies on environment safety and application of advanced reactor for inland nuclear power plants

    International Nuclear Information System (INIS)

    Wei, L.; Jie, L.

    2014-01-01

    To study environment safety assessment of inland nuclear power plants (NPPs), the impact of environment safety under the normal operation was researched and the environment risk of serious accidents was analyzed. Moreover, the requirements and relevant provisions of site selection between international nuclear power plant and China's are comparatively studied. The conclusion was that the environment safety assessment of inland and coastal nuclear power plant have no essential difference; the advanced reactor can meet with high criteria of environment safety of inland nuclear power plants. In this way, China is safe and feasible to develop inland nuclear power plant. China's inland nuclear power plants will be as big market for advanced reactor. (author)

  17. Operational limits and conditions and operating procedures for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2005-01-01

    This Safety Guide was prepared as part of the Agency's programme for establishing safety standards relating to nuclear power plants. The present Safety Guide supersedes the IAEA Safety Guide on Operational Limits and Conditions for Nuclear Power Plants which was issued in 1979 as Safety Series No. 50-SG-O3. For a nuclear power plant to be operated in a safe manner, the provisions made in the final design and subsequent modifications shall be reflected in limitations on plant operating parameters and in the requirements on plant equipment and personnel. Under the responsibility of the operating organization, these shall be developed during the design safety evaluation as a set of operational limits and conditions (OLCs). A major contribution to compliance with the OLCs is made by the development and utilization of operating procedures (OPs) that are consistent with and fully implement the OLCs. The requirements for the OLCs and OPs are established in Section 5 of the IAEA Safety Requirements publication Safety of Nuclear Power Plants: Operation, which this Safety Guide supplements. The purpose of this Safety Guide is to provide guidance on the development, content and implementation of OLCs and OPs. The Safety Guide is directed at both regulators and owners/operators. This Safety Guide covers the concept of OLCs, their content as applicable to land based stationary power plants with thermal neutron reactors, and the responsibilities of the operating organization regarding their establishment, modification, compliance and documentation. The OPs to support the implementation of the OLCs and to ensure their observance are also within the scope of this Safety Guide. The particular aspects of the procedures for maintenance, surveillance, in-service inspection and other safety related activities in connection with the safe operation of nuclear power plants are outside the scope of this Safety Guide but can be found in other IAEA Safety Guides. Section 2 indicates the

  18. Operational limits and conditions and operating procedures for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2000-01-01

    This Safety Guide was prepared as part of the Agency's programme for establishing safety standards relating to nuclear power plants. The present Safety Guide supersedes the IAEA Safety Guide on Operational Limits and Conditions for Nuclear Power Plants which was issued in 1979 as Safety Series No. 50-SG-O3. For a nuclear power plant to be operated in a safe manner, the provisions made in the final design and subsequent modifications shall be reflected in limitations on plant operating parameters and in the requirements on plant equipment and personnel. Under the responsibility of the operating organization, these shall be developed during the design safety evaluation as a set of operational limits and conditions (OLCs). A major contribution to compliance with the OLCs is made by the development and utilization of operating procedures (OPs) that are consistent with and fully implement the OLCs. The requirements for the OLCs and OPs are established in Section 5 of the IAEA Safety Requirements publication Safety of Nuclear Power Plants: Operation, which this Safety Guide supplements. The purpose of this Safety Guide is to provide guidance on the development, content and implementation of OLCs and OPs. The Safety Guide is directed at both regulators and owners/operators. This Safety Guide covers the concept of OLCs, their content as applicable to land based stationary power plants with thermal neutron reactors, and the responsibilities of the operating organization regarding their establishment, modification, compliance and documentation. The OPs to support the implementation of the OLCs and to ensure their observance are also within the scope of this Safety Guide. The particular aspects of the procedures for maintenance, surveillance, in-service inspection and other safety related activities in connection with the safe operation of nuclear power plants are outside the scope of this Safety Guide but can be found in other IAEA Safety Guides. Section 2 indicates the

  19. Current trends in codal requirements for safety in operation of nuclear power plants

    International Nuclear Information System (INIS)

    Srivasista, K.; Shah, Y.K.; Gupta, S.K.

    2006-01-01

    The Code of practice on safety in nuclear power plant operation states the requirements to be met during operation of a nuclear power plant for assuring safety. Among various stages of authorization, regulatory body issues authorization for operation of a nuclear power plant, monitors and enforces regulatory requirements. The responsible organization shall have overall responsibility and the plant management shall have the primary responsibility for ensuring safe and efficient operation of its nuclear power plants. A set of codal requirements covering technical and administrative aspects are mandatory for the plant management to implement to ensure that the nuclear power plant is operated in accordance with the design intent. Requirements on operating procedures and instructions establish operation and maintenance, inspection and testing of the plant in a planned and systematic way. The requirements on emergency preparedness programme establish with a reasonable assurance that, in the event of an emergency situation, appropriate measures can be taken to mitigate the consequences. Commissioning requirements verify performance criteria during commissioning to ensure that the design intent and QA requirements are met. Several modifications in systems important to safety required during operation of a nuclear power plant are regulated. However new operational codal requirements arising out of periodic safety review, operational experience feedback, life management, probabilistic safety assessment, physical security, safety convention and obligations and decommissioning are not covered in the present code of practice for safety in nuclear power plant operation. Codal provisions on 'Review by operating organization on aspects of design having implications on operability' are also required to be addressed. The merits in developing such a methodology include acceptance of the design by operating organization, ensuring maintainability, proper layout etc. in the new designs

  20. Safety assessment of emergency power systems for nuclear power plants

    International Nuclear Information System (INIS)

    1992-01-01

    This publication is intended to assist the safety assessor within a regulatory body, or one working as a consultant, in assessing the safety of a given design of the emergency power systems (EPS) for a nuclear power plant. The present publication refers closely to the NUSS Safety Guide 50-SG-D7 (Rev. 1), Emergency Power Systems at Nuclear Power Plants. It covers therefore exactly the same technical subject as that Safety Guide. In view of its objective, however, it attempts to help in the evaluation of possible technical solutions which are intended to fulfill the safety requirements. Section 2 clarifies the scope further by giving an outline of the assessment steps in the licensing process. After a general outline of the assessment process in relation to the licensing of a nuclear power plant, the publication is divided into two parts. First, all safety issues are presented in the form of questions that have to be answered in order for the assessor to be confident of a safe design. The second part presents the same topics in tabulated form, listing the required documentation which the assessor has to consult and those international and national technical standards pertinent to the topics. An extensive reference list provides information on standards. 1 tab

  1. 78 FR 25488 - Qualification Tests for Safety-Related Actuators in Nuclear Power Plants

    Science.gov (United States)

    2013-05-01

    ... Nuclear Power Plants AGENCY: Nuclear Regulatory Commission. ACTION: Draft regulatory guide; request for... regulatory guide (DG), DG-1235, ``Qualification Tests for Safety-Related Actuators in Nuclear Power Plants... entitled ``Qualification Tests for Safety-Related Actuators in Nuclear Power Plants'' is temporarily...

  2. Application of probabilistic safety goals to regulation of nuclear power plants in Canada

    Energy Technology Data Exchange (ETDEWEB)

    Rzentkowski, G.; Akl, Y.; Yalaoui, S. [Canadian Nuclear Safety Commission, Ottawa, Ontario (Canada)

    2013-07-01

    In the Canadian nuclear regulatory framework, Safety Goals are formulated in addition to the deterministic design requirements and the dose acceptance criteria so that risk to the public that originates from accidents outside the design basis is considered. In principle, application of the Safety Goals ensures that the likelihood of accidents with serious radiological consequences is extremely low, and the potential radiological consequences from severe accidents are limited as far as practicable. Effectively, the Safety Goals extend the plant design envelope to include not only the capabilities of the plant to successfully cope with various plant states, but also practical measures to halt the progression of severe accidents. This paper describes the general approach to the development of the Safety Goals and their application to the existing nuclear power plants in Canada. This general approach is consistent with the currently accepted international practice and Canadian regulatory experience. The results of probabilistic safety assessments indicate that the Safety Goals meet or exceed international safety objectives due to effective implementation of the defence-in-depth principle in the reactor design and plant operation. At the same time, the application of the Safety Goals reveal that practicable measures exist to further enhance the overall level of reactor safety by focusing on severe accident prevention and mitigation. These measures are being currently implemented through refurbishment projects and feedback on operating experience. (author)

  3. Integrated plant safety assessment: systematic evaluation program. Haddam Neck Plant, Connecticut Yankee Atomic Power Company. Docket No. 50-213

    International Nuclear Information System (INIS)

    1983-03-01

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

  4. Recruitment, qualification and training of personnel for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2005-01-01

    The objective of this Safety Guide is to outline the various factors that should to be considered in order to ensure that the operating organization has a sufficient number of qualified personnel for safe operation of a nuclear power plant. In particular, the objective of this publication is to provide general recommendations on the recruitment and selection of plant personnel and on the training and qualification practices that have been adopted in the nuclear industry since the predecessor Safety Guide was published in 1991. In addition, this Safety Guide seeks to establish a framework for ensuring that all managers and staff employed at a nuclear power plant demonstrate their commitment to the management of safety to high professional standards. This Safety Guide deals specifically with those aspects of qualification and training that are important to the safe operation of nuclear power plants. It provides recommendations on the recruitment, selection, qualification, training and authorization of plant personnel. That is, of all personnel in all safety related functions and at all levels of the plant. Some parts or all of this Safety Guide may also be used, with due adaptation, as a guide to the recruitment, selection, training and qualification of staff for other nuclear installations (such as research reactors or nuclear fuel cycle facilities). Section 2 gives guidance on the recruitment and selection of suitable personnel for a nuclear power plant. Section 3 gives guidance on the establishment of personnel qualification, explains the relationship between qualification and competence, and identifies how competence may be developed through education, experience and training. Section 4 deals with general aspects of the training policy for nuclear power plant personnel: the systematic approach, training settings and methods, initial and continuing training, and the keeping of training records. Section 5 provides guidance on the main aspects of training programmes

  5. Recruitment, qualification and training of personnel for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2002-01-01

    The objective of this Safety Guide is to outline the various factors that should to be considered in order to ensure that the operating organization has a sufficient number of qualified personnel for safe operation of a nuclear power plant. In particular, the objective of this publication is to provide general recommendations on the recruitment and selection of plant personnel and on the training and qualification practices that have been adopted in the nuclear industry since the predecessor Safety Guide was published in 1991. In addition, this Safety Guide seeks to establish a framework for ensuring that all managers and staff employed at a nuclear power plant demonstrate their commitment to the management of safety to high professional standards. This Safety Guide deals specifically with those aspects of qualification and training that are important to the safe operation of nuclear power plants. It provides recommendations on the recruitment, selection, qualification, training and authorization of plant personnel; that is, of all personnel in all safety related functions and at all levels of the plant. Some parts or all of this Safety Guide may also be used, with due adaptation, as a guide to the recruitment, selection, training and qualification of staff for other nuclear installations (such as research reactors or nuclear fuel cycle facilities). Section 2 gives guidance on the recruitment and selection of suitable personnel for a nuclear power plant. Section 3 gives guidance on the establishment of personnel qualification, explains the relationship between qualification and competence, and identifies how competence may be developed through education, experience and training. Section 4 deals with general aspects of the training policy for nuclear power plant personnel: the systematic approach, training settings and methods, initial and continuing training, and the keeping of training records. Section 5 provides guidance on the main aspects of training programmes

  6. IAEA Leads Operational Safety Mission to Smolensk Nuclear Power Plant

    International Nuclear Information System (INIS)

    2011-01-01

    Full text: An international team of nuclear safety experts led by the International Atomic Energy Agency (IAEA) has reviewed the Smolensk Nuclear Power Plant (NPP) near Desnogorsk, in Russia's Smolensk region, for its safety practices and has noted a series of good practices as well as recommendations and suggestions to reinforce them. The IAEA assembled the team at the request of the Government of the Russian Federation to conduct an Operational Safety Review (OSART) of the NPP. Under the leadership of the IAEA's Division of Nuclear Installation Safety, the OSART team performed an in-depth operational safety review from 5 to 22 September 2011. The team was made up of experts from China, India, Lithuania, Slovakia, South Africa, Sweden, UK, USA, the World Association of Nuclear Operators and the IAEA. The team conducted an in-depth review of the aspects essential to the safe operation of the Smolensk NPP. The conclusions of the review are based on the IAEA's Safety Standards and proven good international practices. The review covered the areas of Management, Organization and Administration; Training; Operations; Maintenance; Technical Support; Operating Experience; Radiation Protection; and Chemistry. Throughout the review, the exchange of information between the OSART experts and plant personnel was very open, professional and productive. The plant's staff were found to be motivated, well trained, knowledgeable and experienced. The OSART team has identified good plant practices which will be shared with the rest of the nuclear industry for consideration of their application. Examples include the following: Illuminated hot-spot wire to identify higher radiation levels is used in the radiation-controlled area to reduce exposures when working in the controlled area; Modern and state-of-the-art training infrastructure and facilities are available at the plant. These include: maintenance training centre; multimedia simulator for the refueling machine; and safety

  7. Safety and radiation protection at the Swedish nuclear power plants 2000

    International Nuclear Information System (INIS)

    2001-04-01

    During 2000 no events occurred, or discoveries were made, that seriously affected the reactor safety at the Swedish nuclear plants. The basic safety strategy is designed so that hidden faults and deficiencies shall not lead to any serious consequences for the plants. It is of outmost importance that the safety work at the plants is performed with the best effort and quality in order to realize this strategy. Especially in the new economic situation of the utilities after deregulation of the electricity market. The total radiation dose to the personnel and contracted workers at the plants was the lowest ever recorded with all NPPs running (8.1 man Sv). Corrosion damages led to a stand-still of two reactors during a long period, and thorough analyses were performed before the Inspectorate allowed a restart

  8. The Creative Application of Science, Technology and Work Force Innovations to the Decontamination and Decommissioning of the Plutonium Finishing Plant at the Hanford Nuclear Reservation

    International Nuclear Information System (INIS)

    Charboneau, S.; Klos, B.; Heineman, R.; Skeels, B.; Hopkins, A.

    2006-01-01

    The Plutonium Finishing Plant (PFP) consists of a number of process and support buildings for handling plutonium. Building construction began in the late 1940's to meet national priorities and became operational in 1950 producing refined plutonium salts and metal for the United States nuclear weapons program The primary mission of the PFP was to provide plutonium used as special nuclear material for fabrication into a nuclear device for the war effort. Subsequent to the end of World War II, the PFP's mission expanded to support the Cold War effort through plutonium production during the nuclear arms race. PFP has now completed its mission and is fully engaged in deactivation, decontamination and decommissioning (D and D). At this time the PFP buildings are planned to be reduced to ground level (slab-on-grade) and the site remediated to satisfy national, Department of Energy (DOE) and Washington state requirements. The D and D of a highly contaminated plutonium processing facility presents a plethora of challenges. PFP personnel approached the D and D mission with a can-do attitude. They went into D and D knowing they were facing a lot of challenges and unknowns. There were concerns about the configuration control associated with drawings of these old process facilities. There were unknowns regarding the location of electrical lines and the condition and contents of process piping containing chemical residues such as strong acids and caustics. The gloveboxes were highly contaminated with plutonium and chemical residues. Most of the glovebox windows were opaque with splashed process chemicals that coated the windows or etched them, reducing visibility to near zero. Visibility into the glovebox was a serious worker concern. Additionally, all the gloves in the gloveboxes were degraded and unusable. Replacing gloves in gloveboxes was necessary to even begin glovebox clean-out. The sheer volume of breathing air needed was also an issue. These and other challenges and PFP

  9. Qualification of safety-critical software for digital reactor safety system in nuclear power plants

    International Nuclear Information System (INIS)

    Kwon, Kee-Choon; Park, Gee-Yong; Kim, Jang-Yeol; Lee, Jang-Soo

    2013-01-01

    This paper describes the software qualification activities for the safety-critical software of the digital reactor safety system in nuclear power plants. The main activities of the software qualification processes are the preparation of software planning documentations, verification and validation (V and V) of the software requirements specifications (SRS), software design specifications (SDS) and codes, and the testing of the integrated software and integrated system. Moreover, the software safety analysis and software configuration management are involved in the software qualification processes. The V and V procedure for SRS and SDS contains a technical evaluation, licensing suitability evaluation, inspection and traceability analysis, formal verification, software safety analysis, and an evaluation of the software configuration management. The V and V processes for the code are a traceability analysis, source code inspection, test case and test procedure generation. Testing is the major V and V activity of the software integration and system integration phases. The software safety analysis employs a hazard operability method and software fault tree analysis. The software configuration management in each software life cycle is performed by the use of a nuclear software configuration management tool. Through these activities, we can achieve the functionality, performance, reliability, and safety that are the major V and V objectives of the safety-critical software in nuclear power plants. (author)

  10. South Ukraine NPP: Safety improvements through Plant Computer upgrade

    International Nuclear Information System (INIS)

    Brenman, O.; Chernyshov, M. A.; Denning, R. S.; Kolesov, S. A.; Balakan, H. H.; Bilyk, B. I.; Kuznetsov, V. I.; Trosman, G.

    2006-01-01

    This paper summarizes some results of the Plant Computer upgrade at the Units 2 and 3 of South Ukraine Nuclear Power Plant (NPP). A Plant Computer, which is also called the Computer Information System (CIS), is one of the key safety-related systems at VVER-1000 nuclear plants. The main function of the CIS is information support for the plant operators during normal and emergency operational modes. Before this upgrade, South Ukraine NPP operated out-of-date and obsolete systems. This upgrade project wax founded by the U.S. DOE in the framework of the International Nuclear Safety Program (INSP). The most efficient way to improve the quality and reliability of information provided to the plant operator is to upgrade the Human-System Interface (HSI), which is the Upper Level (UL) CIS. The upgrade of the CIS data-acquisition system (DAS), which is the Lower Level (LL) CIS, would have less effect on the unit safety. Generally speaking, the lifetime of the LL CIS is much higher than one of the UL CIS. Unlike Plant Computers at the Western-designed plants, the functionality of the WER-1000 CISs includes a control function (Centralized Protection Testing) and a number of the plant equipment monitoring functions, for example, Protection and Interlock Monitoring and Turbo-Generator Temperature Monitoring. The new system is consistent with a historical migration of the format by which information is presented to the operator away from the traditional graphic displays, for example, Piping and Instrument Diagrams (P and ID's), toward Integral Data displays. The cognitive approach to information presentation is currently limited by some licensing issues, but is adapted to a greater degree with each new system. The paper provides some lessons learned on the management of the international team. (authors)

  11. Psychology in nuclear power plants: an integrative approach to safety - general statement

    International Nuclear Information System (INIS)

    Shikiar, R.

    1983-08-01

    Since the accident at the Three Mile Island nuclear power plant on March 28, 1979, the commercial nuclear industry in the United States has paid increasing attention to the role of humans in overall plant safety. As the regulatory body with primary responsibility for ensuring public health and safety involving nuclear operations, the United States Nuclear Regulatory Commission (NRC) has also become increasingly involved with the ''human'' side of nuclear operations. The purpose of this symposium is to describe a major program of research and technical assistance that the Pacific Northwest Laboratory is performing for the NRC that deals with the issues of safety at nuclear power plants (NPPs). This program addresses safety from several different levels of analysis, which are all important within the context of an integrative approach to system safety

  12. A proposal of safety indicators aggregation to assess the safety management effectiveness of nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Carvalho, Jose Antonio B.; Saldanha, Pedro L.C. [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil). Coordenacao-Geral de Reatores e Ciclo Combustivel], e-mail: jantonio@cnen.gov.br, e-mail: saldanha@cnen.gov.br; Melo, Paulo F.F. Frutuoso e [Coordenacao dos Programas de Pos-graduacao de Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear], e-mail: frutuoso@con.ufrj.br

    2009-07-01

    Safety management has changed with the evolution of management methods, named Quality Systems, moving from Quality Control, where the focus was the product, passing through Quality Assurance, which takes care of the whole manufacturing process and reaching the Total Quality Management, where policies and goals are established. Nowadays, there is a trend towards Management Systems, which integrate all different aspects related to the management of an organization (safety, environment, security, quality, costs and, etc), but it is necessary to have features to establish and assure that safety overrides the remaining aspects. The most usual way to reach this goal is to establish a policy where safety is a priority, but its implementation and the assessment of its effectiveness are no so simple. Nuclear power plants usually have over a hundred safety indicators in many processes dedicated to prevent and detect problems, although a lot of them do not evaluate these indicators in an integrated manner or point out degradation trends of organizational aspects, which can affect the plant safety. This work develops an aggregation of proactive and reactive safety indicators in order to evaluate the effectiveness of nuclear power plant safety management and to detect, at early stages, signs of process degradation or activities used to establish, maintain and assure safety conditions. The aggregation integrates indicators of the usual processes and is based on the manner the management activities have been developed in the last decades, that is: Planning, Doing, Checking and Acting - known as PDCA cycle - plus a fifth element related to the capability of those who perform safety activities. The proposed aggregation is in accordance to Brazilian standards and international recommendations and constitutes a friendly link between the top management level and the daily aspects of the organization. (author)

  13. A proposal of safety indicators aggregation to assess the safety management effectiveness of nuclear power plants

    International Nuclear Information System (INIS)

    Carvalho, Jose Antonio B.; Saldanha, Pedro L.C.; Melo, Paulo F.F. Frutuoso e

    2009-01-01

    Safety management has changed with the evolution of management methods, named Quality Systems, moving from Quality Control, where the focus was the product, passing through Quality Assurance, which takes care of the whole manufacturing process and reaching the Total Quality Management, where policies and goals are established. Nowadays, there is a trend towards Management Systems, which integrate all different aspects related to the management of an organization (safety, environment, security, quality, costs and, etc), but it is necessary to have features to establish and assure that safety overrides the remaining aspects. The most usual way to reach this goal is to establish a policy where safety is a priority, but its implementation and the assessment of its effectiveness are no so simple. Nuclear power plants usually have over a hundred safety indicators in many processes dedicated to prevent and detect problems, although a lot of them do not evaluate these indicators in an integrated manner or point out degradation trends of organizational aspects, which can affect the plant safety. This work develops an aggregation of proactive and reactive safety indicators in order to evaluate the effectiveness of nuclear power plant safety management and to detect, at early stages, signs of process degradation or activities used to establish, maintain and assure safety conditions. The aggregation integrates indicators of the usual processes and is based on the manner the management activities have been developed in the last decades, that is: Planning, Doing, Checking and Acting - known as PDCA cycle - plus a fifth element related to the capability of those who perform safety activities. The proposed aggregation is in accordance to Brazilian standards and international recommendations and constitutes a friendly link between the top management level and the daily aspects of the organization. (author)

  14. Improving Chemical Plant Safety Training Using Virtual Reality

    OpenAIRE

    Nasios, Konstantinos

    2002-01-01

    The chemical engineering industry often requires people to work in hazardous environments and to operate complicated equipment which often limits the type of training that be carried out on site. The daily job of chemical plant operators is becoming more demanding due to the increasing plant complexity together with increasing requirements on plant safety, production capacity, product quality and cost effectiveness. The importance of designing systems and environments that are as safe as poss...

  15. Improved Management of Part Safety Classification System for Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jin Young; Park, Youn Won; Park, Heung Gyu; Park, Hyo Chan [BEES Inc., Daejeon (Korea, Republic of)

    2016-10-15

    As, in recent years, many quality assurance (QA) related incidents, such as falsely-certified parts and forged documentation, etc., were reported in association with the supply of structures, systems, components and parts to nuclear power plants, a need for a better management of safety classification system was addressed so that it would be based more on the level of parts . Presently, the Korean nuclear power plants do not develop and apply relevant procedures for safety classifications, but rather the safety classes of parts are determined solely based on the experience of equipment designers. So proposed in this paper is a better management plan for safety equipment classification system with an aim to strengthen the quality management for parts. The plan was developed through the analysis of newly introduced technical criteria to be applied to parts of nuclear power plant.

  16. Draft report on compilation of generic safety issues for light water reactor nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-07-01

    A generally accepted approach to characterizing the safety concerns in nuclear power plants is to express them as safety issues which need to be resolved. When such safety issues are applicable to a generation of plants of a particular design or to a family of plants of similar design, they are termed generic safety issues. Examples of generic safety issues are those related to reactor vessel embrittlement, control rod insertion reliability or strainer clogging. The safety issues compiled in this document are based on broad international experience. This compilation is one element in the framework of IAEA activities to assist Member States in reassessing the safety of operating nuclear power plants. Refs.

  17. Draft report on compilation of generic safety issues for light water reactor nuclear power plants

    International Nuclear Information System (INIS)

    1997-07-01

    A generally accepted approach to characterizing the safety concerns in nuclear power plants is to express them as safety issues which need to be resolved. When such safety issues are applicable to a generation of plants of a particular design or to a family of plants of similar design, they are termed generic safety issues. Examples of generic safety issues are those related to reactor vessel embrittlement, control rod insertion reliability or strainer clogging. The safety issues compiled in this document are based on broad international experience. This compilation is one element in the framework of IAEA activities to assist Member States in reassessing the safety of operating nuclear power plants. Refs

  18. Some aspects of nuclear power plant safety under war conditions

    International Nuclear Information System (INIS)

    Stritar, A.; Mavko, B.; Susnik, J.; Sarler, B.

    1993-01-01

    In the summer of 1991, the Krsko nuclear power plant in Slovenia found itself in an area of military operations. This was probably the first commercial nuclear power plant to have been threatened by an attack by fighter jets. A number of never-before-asked questions had to be answered by the operating staff and supporting organizations. Some aspects of nuclear power plant safety under war conditions are described, such as the selection of the best plant operating state before the attack and the determination of plant system vulnerability and dose releases from the potentially damaged spent fuel in the spent-fuel pit. The best operating mode to which the plant should be brought before the attack is cold shutdown, and radiological consequences to the environment after the spent fuel is damaged and the water in the pit is lost are not very high. The problem of nuclear power plant safety under war conditions should be addressed in more detail in the future

  19. Probabilistic safety assessment of nuclear power plants: a monograph

    International Nuclear Information System (INIS)

    Solanki, R.B.; Prasad, Mahendra

    2007-11-01

    This monograph on probabilistic safety assessment (PSA) is addressed to the wide community of professionals engaged in the nuclear industry and concerned with the safety issues of nuclear power plants (NPPs). While the monograph describes PSA of NPPs, the principles described in this monograph can be extended to other facilities like spent fuel storage, fuel reprocessing plants and non-nuclear facilities like chemical plants, refineries etc. as applicable. The methodology for risk assessment in chemical plants or refineries is generally known as quantitative risk analysis (QRA). The fundamental difference between NPP and chemical plant is that in NPPs the hazardous material (fuel and fission products) are contained at a single location (i.e. inside containment), whereas in a chemical plant and reprocessing plants, the hazardous material is present simultaneously at many places, like pipelines, reaction towers, storage tanks, etc. Also unlike PSA, QRA does not deal with levels; it uses an integrated approach combining all the levels. The monograph covers the areas of broad interest in the field of PSA such as historical perspective, fundamentals of PSA, strengths and weaknesses of PSA, applications of PSA, role of PSA in the regulatory decision making and issues for advancement of PSA

  20. Safety principles and design management of Chashma Nuclear Power Plant

    International Nuclear Information System (INIS)

    Geng Qirui; Cheng Pingdong

    1997-01-01

    The basic safety consideration and detailed design principles in the design of Chashma Nuclear Power Plant is elaborated. The management within the frame setting up by 'safety culture' and 'quality culture'

  1. Quality assurance for safety in nuclear power plants. A code of practice

    International Nuclear Information System (INIS)

    1978-01-01

    The Code of Practice is a part of the International Atomic Energy Agency's programme, referred to as the NUSS programme (Nuclear Safety Standards), for establishing Codes of Practice and Safety Guides relating to land-based stationary thermal neutron power plants. The documents are based on documentation and experience from various national systems and practices. The present document provides the recommended principles and objectives for the establishment and implementation of a quality assurance programme during design, manufacture, construction, commissioning and operation of structures, system and components important to safety. They are applicable by all those responsible for the power plant, by plant designers, suppliers, architect-engineers, plant constructors, plant operators and other organizations participating in activities affecting quality. The Lists of relevant definition and the Provisional List of NUSS Programme Titles are given

  2. Safety analysis of an expert reactor protection system in nuclear power plants

    International Nuclear Information System (INIS)

    El-Kafas, A.A.

    1997-01-01

    The purpose of the dissertation is to develop real time expert reactor protection system (ERPS) for operational safety of pressurized water reactor nuclear power plant. The system is developed to diagnose plant failures and for identification plant transients (with and without scram). For this erps, probabilistic safety analysis techniques are used to check the availability and priority of the recommended safety system in case of plant accidents. The real - time information during transients and accidents can be obtained to assess the operator in his decision - making. Also, the ERPS is able to give advice for the reactor operator to take the appropriate corrective action during abnormal situations. 5-15 figs., 42 refs

  3. Plasma, a plant safety monitoring and assessment system for VVER-440 reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hornaes, A.; Hulsund, J. E. [Institutt for energiteknikk (IFE), OECD Halden Reactor Project, Halden (Norway); Lipcsei, S.; Major, Cs.; Racz, A.; Vegh, J. [KFKI, Atomic Energy Research Institute, Budapest (Hungary); Eiler, J. [Paks, Nuclear Power Plant Ltd, Paks (Hungary)

    1999-05-15

    The objective with the Plant Safety Monitoring and Assessment System (PLASMA) is to develop an operator support system to support the execution of new symptom-based Emergency Operating Procedures for application in VVER reactors, with the Paks NPP in Hungary as the target plant. Many of the VVER reactors are rewriting their EOPs to comply more with Western standards of symptom-based EOPs. In this connection it is desirable to improve the data validation, information integration and presentation for operators when executing the EOPs. The entry-point to a symptom-oriented procedure is defined by the occurrence of a well-defined reactor operation status, with all its symptoms. However, the application of the EOF benefits from an operator support system, which performs plant status and symptom identification reliably and accurately. The development of the PLASMA system is a joint venture between Institutt for energiteknikk (IFE) and KFKI with the NPP Paks as the target plant. The project has been initiated and partly funded by the Science and Technology Agency (STA), Japan through the OECD NEA assistance program. In Hungary, considerable effort has concentrated on the safety reassessment of the Paks NPP and new EOPs are being written, but no comprehensive Operator Support System (OSS) for plant safety assessment is installed. Some safety parameter display functions are incorporated into diverse operator support systems, but an online 'plant safety monitoring and assessment system' is still missing. The present project comprises designing, constructing, testing and installing such an OSS, which to a great extent could support plant operators in their safety assessment work (author) (ml)

  4. A proposal for safety design philosophy of HTGR for coupling hydrogen production plant

    International Nuclear Information System (INIS)

    Sato, Hiroyuki; Ohashi, Hirofumi; Tazawa, Yujiro; Imai, Yoshiyuki; Nakagawa, Shigeaki; Tachibana, Yukio; Kunitomi, Kazuhiko

    2013-06-01

    Japan Atomic Energy Agency (JAEA) has been conducting research and development for hydrogen production utilizing heat from High Temperature Gas-cooled Reactors (HTGRs). Towards the realization of nuclear hydrogen production, coupled hydrogen production plants should not be treated as an extension of a nuclear plant in order to open the door for the entry of non-nuclear industries as well as assuring reactor safety against postulated abnormal events initiated in the hydrogen production plants. Since hydrogen production plant utilizing nuclear heat has never been built in the world, little attention has been given to the establishment of a safety design for such system including the High Temperature engineering Test Reactor (HTTR). In the present study, requirements in order to design, construct and operate hydrogen production plants under conventional chemical plant standards are identified. In addition, design considerations for safety design of nuclear facility are suggested. Furthermore, feasibility of proposed safety design and design considerations are evaluated. (author)

  5. Nuclear power plant with a safety enclosure

    International Nuclear Information System (INIS)

    Keller, W.; Krueger, J.; Ropers, J.; Schabert, H.P.

    1976-01-01

    A nuclear power plant has a safety enclosure for a nuclear reactor. A fuel element storage basin is also located in this safety enclosure and a fuel element lock extends through the enclosure, with a cross-sectional size proportioned for the endwise passage of fuel elements, the lock including internal and external valves so that a fuel element may be locked endwise safely through the lock. The lock, including its valves, being of small size, does not materially affect the pressure resistance of the safety enclosure, and it is more easily operated than a lock large enough to pass people and fuel element transport vessels

  6. Holistic safety analysis for advanced nuclear power plants

    International Nuclear Information System (INIS)

    Alvarenga, M.A.B.; Guimaraes, A.C.F.

    1992-01-01

    This paper reviews the basic methodology of safety analysis used in the ANGRA-I and ANGRA-II nuclear power plants, its weaknesses, the problems with public acceptance of the risks, the future of the nuclear energy in Brazil, as well as recommends a new methodology, HOLISTIC SAFETY ANALYSIS, to be used both in the design and licensing phases, for advanced reactors. (author)

  7. Criteria Document for B-plant's Surveillance and Maintenance Phase Safety Basis Document

    International Nuclear Information System (INIS)

    SCHWEHR, B.A.

    1999-01-01

    This document is required by the Project Hanford Managing Contractor (PHMC) procedure, HNF-PRO-705, Safety Basis Planning, Documentation, Review, and Approval. This document specifies the criteria that shall be in the B Plant surveillance and maintenance phase safety basis in order to obtain approval of the DOE-RL. This CD describes the criteria to be addressed in the S and M Phase safety basis for the deactivated Waste Fractionization Facility (B Plant) on the Hanford Site in Washington state. This criteria document describes: the document type and format that will be used for the S and M Phase safety basis, the requirements documents that will be invoked for the document development, the deactivated condition of the B Plant facility, and the scope of issues to be addressed in the S and M Phase safety basis document

  8. General view about reactor safety nuclear power plants in Brazil

    International Nuclear Information System (INIS)

    Gasparian, A.E.; Silva, D.E.; Salvatore, J.E.L.; Lima, J.M. de

    1991-01-01

    In this paper the authors describe the principles and goals that have guided, as well as the methods that have been used by the National Commission of Nuclear Energy (CNEN) to set forth measures aiming at providing safety to the Brazilian nuclear power plants. The status of the licensing process of these power plants is shown. The performance and the results obtained so far in relation to the nuclear safety are also described. (author)

  9. Radiation protection aspects of design for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2005-01-01

    . The IAEA takes seriously the enduring challenge for users and regulators everywhere: that of ensuring a high level of safety in the use of nuclear materials and radiation sources around the world. Their continuing utilization for the benefit of humankind must be managed in a safe manner, and the IAEA safety standards are designed to facilitate the achievement of that goal. This Safety Guide has been prepared as a part of the IAEA programme on safety standards for nuclear power plants. It includes recommendations on how to satisfy the requirements established in the Safety Requirements publication on the Safety of Nuclear Power Plants: Design. It addresses the provisions that should be made in the design of nuclear power plants in order to protect site personnel, the public and the environment against radiological hazards for operational states, decommissioning and accident conditions. The recommendations on radiation protection provided in this Safety Guide are consistent with the International Basic Safety Standards for Protection against Ionizing Radiation and for the Safety of Radiation Sources (BSS), which were jointly sponsored by the Food and Agriculture Organization of the United Nations (FAO), the IAEA, the International Labour Organisation (ILO), the OECD Nuclear Energy Agency (OECD/NEA), the Pan American Health Organization (PAHO) and the World Health Organization (WHO). This Safety Guide supersedes Safety Series No. 50-SG-D9, Design Aspects of Radiation Protection for Nuclear Power Plants, published in 1985. Effective radiation protection is a combination of good design, high quality construction and proper operation. Procedures that address the radiation protection aspects of operation are covered in the Safety Guide on Radiation Protection and Radioactive Waste Management in the operation of Nuclear Power Plants

  10. ESRS guidelines for software safety reviews. Reference document for the organization and conduct of Engineering Safety Review Services (ESRS) on software important to safety in nuclear power plants

    International Nuclear Information System (INIS)

    2000-01-01

    The IAEA provides safety review services to assist Member States in the application of safety standards and, in particular, to evaluate and facilitate improvements in nuclear power plant safety performance. Complementary to the Operational Safety Review Team (OSART) and the International Regulatory Review Team (IRRT) services are the Engineering Safety Review Services (ESRS), which include reviews of siting, external events and structural safety, design safety, fire safety, ageing management and software safety. Software is of increasing importance to safety in nuclear power plants as the use of computer based equipment and systems, controlled by software, is increasing in new and older plants. Computer based devices are used in both safety related applications (such as process control and monitoring) and safety critical applications (such as reactor protection). Their dependability can only be ensured if a systematic, fully documented and reviewable engineering process is used. The ESRS on software safety are designed to assist a nuclear power plant or a regulatory body of a Member State in the review of documentation relating to the development, application and safety assessment of software embedded in computer based systems important to safety in nuclear power plants. The software safety reviews can be tailored to the specific needs of the requesting organization. Examples of such reviews are: project planning reviews, reviews of specific issues and reviews prior final acceptance. This report gives information on the possible scope of ESRS software safety reviews and guidance on the organization and conduct of the reviews. It is aimed at Member States considering these reviews and IAEA staff and external experts performing the reviews. The ESRS software safety reviews evaluate the degree to which software documents show that the development process and the final product conform to international standards, guidelines and current practices. Recommendations are

  11. Organizational factors and nuclear power plant safety

    International Nuclear Information System (INIS)

    Haber, S.B.

    1995-01-01

    There are many organizations in our society that depend on human performance to avoid incidents involving significant adverse consequences. As our culture and technology have become more sophisticated, the management of risk on a broad basis has become more and more critical. The safe operation of military facilities, chemical plants, airlines, and mass transit, to name a few, are substantially dependent on the performance of the organizations that operate those facilities. The nuclear power industry has, within the past 15 years, increased the attention given to the influence of human performance in the safe operation of nuclear power plants (NPP). While NPPs have been designed through engineering disciplines to intercept and mitigate events that could cause adverse consequences, it has been clear from various safety-related incidents that human performance also plays a dominant role in preventing accidents. Initial efforts following the 1979 Three Mile Island incident focused primarily on ergonomic factors (e.g., the best design of control rooms for maximum performance). Greater attention was subsequently directed towards cognitive processes involved in the use of NPP decision support systems and decision making in general, personnel functions such as selection systems, and the influence of work scheduling and planning on employees' performance. Although each of these approaches has contributed to increasing the safety of NPPS, during the last few years, there has been a growing awareness that particular attention must be paid to how organizational processes affect NPP personnel performance, and thus, plant safety. The direct importance of organizational factors on safety performance in the NPP has been well-documented in the reports on the Three Mile Island and Chernobyl accidents as well as numerous other events, especially as evaluated by the U.S. Nuclear Regulatory Commission (NRC)

  12. Safety and environmental aspects of deuterium--tritium fusion power plants: work shop summary

    International Nuclear Information System (INIS)

    1978-05-01

    In September of 1977 a workshop was held on the safety and environmental aspects of fusion power plants to consider potential safety and environmental problems of fusion power plants and to reveal solutions or methods of solving those problems. The objective was to promote incorporation of safety and environmental protection into reactor design, thereby reducing the expense and delay of backfitting safety systems after reactor designs are complete. A dialogue was established between fusion reactor designers and safety and environmental researchers. Four topics, each with several subdivisions, were selected for discussion: radiation exposure, accidents, environmental effects, and plant safety. For each topic, discussion focused on the significance of the problem, and adequacy of current technology to solve the problem, design solutions available and research needed to solve the problem

  13. Survey of numerical safety targets for nuclear power plants

    International Nuclear Information System (INIS)

    Kelley, A.P. Jr.; Buttemer, D.R.

    1981-04-01

    The construction of a nuclear power plant implies, as does the construction of any major public work, the acceptance of a finite degree of risk. This risk can be reduced by an increased investment in engineered safeguards. However, at some level of risk, overinvestment in safety can render the project uneconomical. Because of the desirability of fixing safety standards on an absolute basis, there has long been an interest in establishing numerical risk criteria for the design, construction, and operation of nuclear power plants. Interest in the subject of numerical safety goals has recently been intensified by the Three Mile Island Action Plan. The USNRC has been directed by Congress to develop a national safety goal for reactor regulation. This report summarizes actions which have been historically, and are currently, taking place toward establishing national numerical risk targets for reactor regulation. Emphasis is placed upon actions taken, or currently being taken, by federal regulatory agencies and directly associated advisory bodies

  14. Safety provision for nuclear power plants during remaining running time

    International Nuclear Information System (INIS)

    Rossnagel, Alexander; Hentschel, Anja

    2012-01-01

    With the phasing-out of the industrial use of nuclear energy for the power generation, the risk of the nuclear power plants has not been eliminated in principle, but only for a limited period of time. Therefore, the remaining nine nuclear power plants must also be used for the remaining ten years according to the state of science and technology. Regulatory authorities must substantiate the safety requirements for each nuclear power plant and enforce these requirements by means of various regulatory measures. The consequences of Fukushima must be included in the assessment of the safety level of nuclear power plants in Germany. In this respect, the regulatory authorities have the important tasks to investigate and assess the security risks as well as to develop instructions and orders.

  15. A study on safety assessment methodology for a vitrification plant

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Y. C.; Lee, G. S.; Choi, Y. C.; Kim, G. H. [Yonsei Univ., Seoul (Korea, Republic of)

    2002-03-15

    In this study, the technical and regulatory status of radioactive waste vitrification technologies in foreign and domestic plants is investigated and analyzed, and then significant factors are suggested which must be contained in the final technical guideline or standard for the safety assessment of vitrification plants. Also, the methods to estimate the stability of vitrified waste forms are suggested with property analysis of them. The contents and scope of the study are summarized as follows : survey of the status on radioactive waste vitrification technologies in foreign and domestic plants, survey of the characterization methodology for radioactive waste form, analysis of stability for vitrified waste forms, survey and analysis of technical standards and regulations concerned with them in foreign and domestic plants, suggestion of significant factors for the safety assessment of vitrification plants, submission of regulated technical standard on radioactive waste vitrification plats.

  16. Safety review on unit testing of safety system software of nuclear power plant

    International Nuclear Information System (INIS)

    Liu Le; Zhang Qi

    2013-01-01

    Software unit testing has an important place in the testing of safety system software of nuclear power plants, and in the wider scope of the verification and validation. It is a comprehensive, systematic process, and its documentation shall meet the related requirements. When reviewing software unit testing, attention should be paid to the coverage of software safety requirements, the coverage of software internal structure, and the independence of the work. (authors)

  17. Safety analysis of an expert reactor protection system in nuclear power plants

    International Nuclear Information System (INIS)

    EL-Kafas, A.E.A.E.

    1996-01-01

    the purpose of the dissertation is to develop a real time expert reactor protection system (ERPS) for operational safety of pressurized water reactor nuclear power plant. The system is developed to diagnose plant failures and for identification of plant transients (with and without scram). for this ERPS. probabilistic safety analysis techniques are used to check the availability and priority of the recommended safety system in case of plant accidents . the real- time information during transients and accidents can be obtained to asses the operator in his decision - making . Also, the ERPS is able to give advice for the reactor operator to take the appropriate corrective action during abnormal situations. The system model consists of the dynamic differential equations for reactor core, pressurizer, steam generator, turbine and generator, piping and plenums. The system of equations can be solved by appropriate codes also displayed directly from sensors of the plant. All scenarios of transients, accidents and fault tress for plant systems are learned to ERPS

  18. A generalized framework for assessment of safety margins in nuclear power plants

    International Nuclear Information System (INIS)

    Gavrilas, M.; Youngblood, B.; Prelewicz, D.; Meyer, Jim

    2004-01-01

    The protection of public health and safety, and the environment from inadvertent releases of radioactive materials from nuclear power plants relies on the implementation of the defense-in-depth strategy. The term defense-in-depth evolved historically, and thus its application has not always been uniform. The use of the term in the context of the U.S. Nuclear Regulatory Commission (NRC) safety philosophy entails the reliance of a nuclear facility on successive compensatory measures in preventing accidents or mitigating damage caused by malfunctions, accidents, or naturally occurring events. The introduction of probabilistic risk analyses with NUREG-74/014 and subsequent evolution in risk assessment techniques, are leading to the implementation of risk informed regulation to ensure the safety of the public and the environment. Risk informed regulation minimizes the likelihood of overlooking potentially significant accident sequences while limiting unnecessary burdens imposed on licensees. The proposed framework merges fundamental elements of safety regulation: defense-in depth, safety margins and probabilistic risk. It formalizes the relationship between probabilistic risk assessment (PRA) methods and data, and deterministic analyses in a manner consistent with NRC's defense-in-depth philosophy. Succinctly put, the likelihood and consequences of accident scenarios are considered simultaneously and quantified by a plant safety metric. The integration of these fundamental elements into a practically applicable safety framework is consistent with the NRC policy statement on use of probabilistic risk assessment methods and the November 2002 Regulatory Guide on risk informed decisions on plant-specific changes to the licensing basis. Safety information resulting from the application of the framework supersedes traditional safety figures of merit. Safety quantifiers, referred herein as safety indices, expand on the qualifier outcomes that currently accompany fault tree

  19. Planning and architectural safety considerations in designing nuclear power plants

    International Nuclear Information System (INIS)

    Konsowa, Ahmed A.

    2009-01-01

    To achieve optimum safety and to avoid possible hazards in nuclear power plants, considering architectural design fundamentals and all operating precautions is mandatory. There are some planning and architectural precautions should be considered to achieve a high quality design and construction of nuclear power plant with optimum safety. This paper highlights predicted hazards like fire, terrorism, aircraft crash attacks, adversaries, intruders, and earthquakes, proposing protective actions against these hazards that vary from preventing danger to evacuating and sheltering people in-place. For instance; using safeguards program to protect against sabotage, theft, and diversion. Also, site and building well design focusing on escape pathways, emergency exits, and evacuation zones, and the safety procedures such as; evacuation exercises and sheltering processes according to different emergency classifications. In addition, this paper mentions some important codes and regulations that control nuclear power plants design, and assessment methods that evaluate probable risks. (author)

  20. External Events Excluding Earthquakes in the Design of Nuclear Power Plants. Safety Guide

    International Nuclear Information System (INIS)

    2008-01-01

    This Safety Guide provides recommendations and guidance on design for the protection of nuclear power plants from the effects of external events (excluding earthquakes), i.e. events that originate either off the site or within the boundaries of the site but from sources that are not directly involved in the operational states of the nuclear power plant units. In addition, it provides recommendations on engineering related matters in order to comply with the safety objectives and requirements established in the IAEA Safety Requirements publication, Safety of Nuclear Power Plants: Design. It is also applicable to the design and safety assessment of items important to the safety of land based stationary nuclear power plants with water cooled reactors. Contents: 1. Introduction; 2. Application of safety criteria to the design; 3. Design basis for external events; 4. Aircraft crash; 5. External fire; 6. Explosions; 7. Asphyxiant and toxic gases; 8. Corrosive and radioactive gases and liquids; 9. Electromagnetic interference; 10. Floods; 11. Extreme winds; 12. Extreme meteorological conditions; 13. Biological phenomena; 14. Volcanism; 15. Collisions of floating bodies with water intakes and UHS components; Annex I: Aircraft crashes; Annex II: Detonation and deflagration; Annex III: Toxicity limits.

  1. Safety of WWER type nuclear power plants - viewing from Hungary

    International Nuclear Information System (INIS)

    Voeroess, L.

    1991-01-01

    An evaluation of WWER type nuclear power plants operating in Hungary is given, relative to the safety requirements accepted internationally; how safe can they be regarded and what can be done to assure a high level of safety in all case. After an overview of general safety criteria, an overall description of WWER-440 type nuclear reactors is presented. Design safety, operational safety issues are treated in detail. Safety inspection and safety-related research and development is discussed. Regarding the future, five different issues associated with nuclear reactor safety should be considered. (R.P.) 20 refs.; 12 figs.; 3 tabs

  2. Management and organization in nuclear power plant safety

    International Nuclear Information System (INIS)

    Osborn, R.N.

    1983-08-01

    In the immediate aftermath of the Three Mile Island accident, the Nuclear Regulatory Commission-sponsored investigations of the relation between human issues and safety tended to focus on individual and, at most, group level phenomena. This initial bottom up view of organizational safety has continued to be investigated by the Nuclear Regulatory Commission, as evidence by the four previous papers. Recently, however, work has begun which adopts a top down management/organization approach to nuclear power plant safety. This paper reports on the research, to date, on this focus

  3. Definitions of engineered safety features and related features for nuclear power plants

    International Nuclear Information System (INIS)

    1986-01-01

    In light water moderated, light water cooled nuclear power plants, definitions are given of engineered safety features which are designed to suppress or prevent dispersion of radioactive materials due to damage etc. of fuel at the times of power plant failures, and of related features which are designed to actuate or operate the engineered safety features. Contents are the following: scope of engineered safety features and of related features; classification of engineered safety features (direct systems and indirect systems) and of related features (auxiliaries, emergency power supply, and protective means). (Mori, K.)

  4. A probabilistic method for optimization of fire safety in nuclear power plants

    International Nuclear Information System (INIS)

    Hosser, D.; Sprey, W.

    1986-01-01

    As part of a comprehensive fire safety study for German Nuclear Power Plants a probabilistic method for the analysis and optimization of fire safety has been developed. It follows the general line of the American fire hazard analysis, with more or less important modifications in detail. At first, fire event trees in selected critical plant areas are established taking into account active and passive fire protection measures and safety systems endangered by the fire. Failure models for fire protection measures and safety systems are formulated depending on common parameters like time after ignition and fire effects. These dependences are properly taken into account in the analysis of the fire event trees with the help of first-order system reliability theory. In addition to frequencies of fire-induced safety system failures relative weights of event paths, fire protection measures within these paths and parameters of the failure models are calculated as functions of time. Based on these information optimization of fire safety is achieved by modifying primarily event paths, fire protection measures and parameters with the greatest relative weights. This procedure is illustrated using as an example a German 1300 MW PWR reference plant. It is shown that the recommended modifications also reduce the risk to plant personnel and fire damage

  5. Generic safety issues for nuclear power plants with pressurized heavy water reactors and measures for their resolution

    International Nuclear Information System (INIS)

    2007-06-01

    The IAEA Conference on The Safety of Nuclear Power: Strategy for the Future in 1991 was a milestone in nuclear safety. The objective of this conference was to review nuclear power safety issues for which achieving international consensus would be desirable, to address concerns on nuclear safety and to formulate recommendations for future actions by national and international authorities to advance nuclear safety to the highest level. Two of the important items addressed by this conference were ensuring and enhancing safety of operating plants and treatment of nuclear power plants built to earlier safety standards. Publications related to these two items, that have been issued subsequent to this conference, include: A Common Basis for Judging the Safety of Nuclear Power Plants Built to Earlier Standards, INSAG-8 (1995), the IAEA Safety Guide 50-SG-O12, Periodic Safety Review of Operational Nuclear Power Plants (1994) and an IAEA publication on the Safety Evaluation of Operating Nuclear Power Plants Built to Earlier Standards - A Common Basis for Judgement (1997). Some of the findings of the 1991 conference have not yet been fully addressed. An IAEA Symposium on Reviewing the Safety of Existing Nuclear Power Plants in 1996 showed that there is an urgent need for operating organizations and national authorities to review operating nuclear power plants which do not meet the high safety levels of the vast majority of plants and to undertake improvements, with assistance from the international community if required. Safety reviews of operating nuclear power plants take on added importance in the context of the Convention on Nuclear Safety and its implementation. To perform safety reviews and to reassess the safety of operating nuclear power plants in a uniform manner, it is imperative to have an internationally accepted reference. Existing guidance needs to be complemented by a list of safety issues which have been encountered and resolved in other plants and which can

  6. Protection against internal fires and explosions in the design of nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2004-01-01

    Experience of the past two decades in the operation of nuclear power plants and modern analysis techniques confirm that fire may be a real threat to nuclear safety and should receive adequate attention from the beginning of the design process throughout the life of the plant. Within the framework of the NUSS programme, a Safety Guide on fire protection had therefore been developed to enlarge on the general requirements given in the Code. Since its first publication in 1979, there has been considerable development in protection technology and analysis methods and after the Chernobyl accident it was decided to revise the existing Guide. This Safety Guide supplements the requirements established in Safety of Nuclear Power Plants: Design. It supersedes Safety Series No. 50-SG-D2 (Rev. 1), Fire Protection in Nuclear Power Plants: A Safety Guide, issued in 1992.The present Safety Guide is intended to advise designers, safety assessors and regulators on the concept of fire protection in the design of nuclear power plants and on recommended ways of implementing the concept in some detail in practice

  7. SAFETY STUDIES TO MEASURE EXOTHERMIC REACTIONS OF SPENT PLUTONIUM DECONTAMINATION CHEMICALS USING WET AND DRY DECONTAMINATION METHODS

    International Nuclear Information System (INIS)

    HOPKINS, A.M.; JACKSON, G.W.; MINETTE, M.; EWALT, J.; COOPER, T.; SCOTT, P.; JONES, S.; SCHEELEY, R.

    2005-01-01

    The Plutonium Finishing Plant (PFP) at the Hanford site in Eastern Washington is currently being decommissioned by Fluor Hanford. Chemicals being considered for dccontamination of gloveboxes in PFP include cerium (IV) nitrate in a nitric acid solution, and proprietary commercial solutions that include acids and sequestering agents. Aggressive chemicals are commonly used to remove transuranic contaminants from process equipment to allow disposal of the equipment as low level waste. Fluor's decontamination procedure involves application of chemical solutions as a spray on the contaminated surfaces, followed by a wipe-down with rags. Alternatively, a process of applying oxidizing Ce IV ions contained in a gel matrix and vacuuming a dry gel material is being evaluated. These processes effectively transfer the transuranic materials to rags or a gel matrix which is then packaged as TRU waste and disposed

  8. Fire safety study of Dodewaard and Borssele nuclear power plants

    International Nuclear Information System (INIS)

    1988-03-01

    From the nuclear and conventional fire safety audits of both Dutch nuclear power plants performed under supervision of the Nuclear Safety Inspectorate and the Inspectorate for the Fire Services it turns out that the fire safety is sufficient however amenable for improvement. Besides a detailed description of the method of examination, the study 'Fire Safety' contains the results of the audit and 14 respectively 15 recommendations for improvement of the fire safety in Dodewaard and Borssele. The suggested recommendations which are applicable to both power plants are the following: fire fighting training for operators and guards, a complete emergency ventilation system of the control room, increase in the number of detectors and alarms, an increase in the quantity of water available for high-pressure fire fighting, improvement of fire barriers between several redundancies of nuclear safety systems, an investigation and possible improvement of the heat and radiation protection offered by presently used fire fighting suits. For Dodewaard a closed emergency staircase in the reactor building (secondary containment) is deemed necessary for personnel emergency escape routes and continued fire fighting if required

  9. Safety issues and their ranking for WWER-440 model 213 nuclear power plants. A publication of the extrabudgetary programme on the safety of WWER and RBMK nuclear power plants

    International Nuclear Information System (INIS)

    1996-04-01

    The objective of this report is to present a consolidated list of generic safety concerns, called safety issues, ranked according to their safety significance and the corrective measures to improve safety. It is intended for use as a reference to facilitate the development of plant specific safety improvement programmes and to serve as a basis for reviewing their implementation. Section 2 provides and overview of the impact of all relevant issues on the main safety functions and other aspects important to overall plant safety. Section 3 presents safety issues identified in design according to the structure described below. Section 4 presents the safety issues in the area of operation, according to the same structure except that no ranking is given. At the end of Section 2, Tables 1 and 2 present a summary of all safety issues in a tabular form. 138 refs, tabs

  10. Quality and safety of nuclear plants: the role of the administrative authorities

    International Nuclear Information System (INIS)

    Queniart, D.

    1977-10-01

    After specifying the notions of 'safety' and 'quality', the terms and conditions governing the intervention of the public authorities in the matter of safety of nuclear plants are described: individual permits, the establishing and application of technical rules of a general character, surveillance of the plants. The criteria and regulations guiding the evaluation of safety and quality and, in conclusion, insisting on the necessity for permanent discussions among the various organizations concerned are presented

  11. Safety goals for nuclear power plant operation

    International Nuclear Information System (INIS)

    1983-05-01

    This report presents and discusses the Nuclear Regulatory Commission's, Policy Statement on Safety Goals for the Operation of Nuclear Power Plants. The safety goals have been formulated in terms of qualitative goals and quantitative design objectives. The qualitative goals state that the risk to any individual member of the public from nuclear power plant operation should not be a significant contributor to that individual's risk of accidental death or injury and that the societal risks should be comparable to or less than those of viable competing technologies. The quantitative design objectives state that the average risks to individual and the societal risks of nuclear power plant operation should not exceed 0.1% of certain other risks to which members of the US population are exposed. A subsidiary quantitative design objective is established for the frequency of large-scale core melt. The significance of the goals and objectives, their bases and rationale, and the plan to evaluate the goals are provided. In addition, public comments on the 1982 proposed policy statement and responses to a series of questions that accompanied the 1982 statement are summarized

  12. Good safety culture maintenance at Leningrad nuclear power plant

    International Nuclear Information System (INIS)

    Ardanov, A.

    1996-01-01

    The evidence in favour of the Leningrad NPP commitment to safety tasks, as the case is in the international practice, is The Safety Policy Statement document where safety is declared to be more significant than the power generation related issues, with the entire responsibility for the safety provision taken over by the operating utility. To avoid the situation when the stated safety tasks and policy remain only a declaration, the organizational structure of the operating utility was expanded to include The Safety Control Department and The Quality Control Department whose tasks encompass the control of the achieved safety level, development of recommendations, measures and actions aimed at the safety culture improvement, assessment and revision of the criteria and requirements to the personnel and management. Each individual at LNPP whose activity affects the plant safety has been familiarized with The Safety Policy Statement document

  13. Good safety culture maintenance at Leningrad nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Ardanov, A [Safety Control Dept., Leningrad Nuclear Power Plant, Leningrad (Russian Federation)

    1997-12-31

    The evidence in favour of the Leningrad NPP commitment to safety tasks, as the case is in the international practice, is The Safety Policy Statement document where safety is declared to be more significant than the power generation related issues, with the entire responsibility for the safety provision taken over by the operating utility. To avoid the situation when the stated safety tasks and policy remain only a declaration, the organizational structure of the operating utility was expanded to include The Safety Control Department and The Quality Control Department whose tasks encompass the control of the achieved safety level, development of recommendations, measures and actions aimed at the safety culture improvement, assessment and revision of the criteria and requirements to the personnel and management. Each individual at LNPP whose activity affects the plant safety has been familiarized with The Safety Policy Statement document.

  14. Generic safety issues for nuclear power plants with light water reactors and measures taken for their resolution

    International Nuclear Information System (INIS)

    1998-09-01

    The IAEA Conference on 'The Safety of Nuclear Power: Strategy for the Future' in 1991 was a milestone in nuclear safety. Two of the important items addressed by this conference were ensuring and enhancing safety of operating plants and treatment of nuclear power plants built to earlier safety standards. A number of publications related to these two items issued subsequent to this conference were: A Common Basis for Judging the Safety of Nuclear Power Plants Built to Earlier Standards, INSAG-9 (1995), the IAEA Safety Guide 50-SG-O12, periodic Safety Review of Operational Nuclear Power Plants (1994) and an IAEA publication on the Safety Evaluation of Operating Nuclear Power Plants Built to Earlier Standards - A Common Basis for Judgement (1997). Some of the findings of the 1991 Conference have not yet been fully addressed. An IAEA Symposium on reviewing the Safety of Existing Nuclear Power Plants in 1996 showed that there is an urgent need for operating organizations and national authorities to review operating nuclear power plants which do not meet the high safety levels of the vast majority of plants and to undertake improvements with assistance from the international community if required. Safety reviews of operating nuclear power plants take on added importance in the context of the Convention on Nuclear safety and its implementation. The purpose of this TECDOC compilation based on broad international experience, is to assist the Member States in the reassessment of operating plants by providing a list of generic safety issues identified in nuclear power plants together with measures taken to resolve these issues. These safety issues are generic in nature with regard to light water reactors and the measures for their resolution are for use as a reference for the safety reassessment of operating plants. The TECDOC covers issues thought to be significant to Member States based on consensus process. It provides an introduction to the use of generic safety issues for

  15. Upgrading of fire safety in nuclear power plants. Proceedings of an International Symposium

    International Nuclear Information System (INIS)

    1998-04-01

    The document includes 40 papers presented at the International Symposium on Upgrading of Fire Safety in Nuclear Power Plants held in Vienna between 18-21 November 1997. The symposium presentations were grouped in 6 sessions: Fire safety reviews (5 papers), Fire safety analysis - Methodology (6 papers), Fire safety analysis - Applications (3 papers), Panel 1 - Identification of deficiencies in fire safety in nuclear power plants - Operational experience and data (7 papers), Panel 2 - Experience based data in fire safety assessment - Fire safety regulations and licensing (7 papers), Upgrading programmes (10 papers), and a closing session (2 papers). A separate abstract was prepared for each paper

  16. Upgrading of fire safety in nuclear power plants. Proceedings of an International Symposium

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-04-01

    The document includes 40 papers presented at the International Symposium on Upgrading of Fire Safety in Nuclear Power Plants held in Vienna between 18-21 November 1997. The symposium presentations were grouped in 6 sessions: Fire safety reviews (5 papers), Fire safety analysis - Methodology (6 papers), Fire safety analysis - Applications (3 papers), Panel 1 - Identification of deficiencies in fire safety in nuclear power plants - Operational experience and data (7 papers), Panel 2 - Experience based data in fire safety assessment - Fire safety regulations and licensing (7 papers), Upgrading programmes (10 papers), and a closing session (2 papers). A separate abstract was prepared for each paper Refs, figs, tabs

  17. Defense-in-depth for common cause failure of nuclear power plant safety system software

    International Nuclear Information System (INIS)

    Tian Lu

    2012-01-01

    This paper briefly describes the development of digital I and C system in nuclear power plant, and analyses the viewpoints of NRC and other nuclear safety authorities on Software Common Cause Failure (SWCCF). In view of the SWCCF issue introduced by the digitized platform adopted in nuclear power plant safety system, this paper illustrated a diversified defence strategy for computer software and hardware. A diversified defence-in-depth solution is provided for digital safety system of nuclear power plant. Meanwhile, analysis on problems may be faced during application of nuclear safety license are analyzed, and direction of future nuclear safety I and C system development are put forward. (author)

  18. Safety implications of computerized process control in nuclear power plants

    International Nuclear Information System (INIS)

    1991-02-01

    Modern nuclear power plants are making increasing use of computerized process control because of the number of potential benefits that accrue. This practice not only applies to new plants but also to those in operation. Here, the replacement of both conventional process control systems and outdated computerized systems is seen to be of benefit. Whilst this contribution is obviously of great importance to the viability of nuclear electricity generation, it must be recognized that there are major safety concerns in taking this route. However, there is the potential for enhancing the safety of nuclear power plants if the full power of microcomputers and the associated electronics is applied correctly through well designed, engineered, installed and maintained systems. It is essential that areas where safety can be improved be identified and that the pitfalls are clearly marked so that they can be avoided. The deliberations of this Technical Committee Meeting are a step on the road to this goal of improved safety through computerized process control. This report also contains the papers presented at the technical committee meeting by participants. A separate abstract was prepared for each of these 15 presentations. Refs, figs and tabs

  19. Safety-related instrumentation and control systems for nuclear power plants

    International Nuclear Information System (INIS)

    1984-01-01

    This Safety Guide deals mainly with design requirements for those I and C systems that are important to safety but are not safety systems. The Guide is intended to expand paragraphs 3.1, 3.2 and 3.3 of the Code of Practice on Design for Safety of Nuclear Power Plants (IAEA Safety Series No.50-C-D) in the area of I and C systems important to safety and refers to them as safety-related I and C systems. It also gives guidance and enumerates requirements for multiplexing and the use of the digital computers employed in this area

  20. Insights from the U.S. department of Energy plant safety evaluation program of VVER and RBMK reactors

    International Nuclear Information System (INIS)

    Petri, M.C.; Binder, J.L.; Pasedag, W.F.

    2001-01-01

    Throughout the years 1990 the U.S. Department of Energy has worked build capability in countries of the former Soviet Union to assess the safety of their VVER and RBMK reactors. Through this Plant Safety Evaluation Program, deterministic and probabilistic analyses have been used to provide a documented plant risk profile to support safe plant operation and to set priorities for safety upgrades. Work has been sponsored at thirteen nuclear power plant sites in eight countries. The Plant Safety Evaluation Program has resulted in immediate and long-term safety benefits for the Soviet-designed nuclear plants. (author)

  1. Core management and fuel handling for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2004-01-01

    This Safety Guide supplements and elaborates upon the safety requirements for core management and fuel handling that are presented in Section 5 of the Safety Requirements publication on the operation of nuclear power plants. The present publication supersedes the IAEA Safety Guide on Safety Aspects of Core Management and Fuel Handling, issued in 1985 as Safety Series No. 50-SG-010. It is also related to the Safety Guide on the Operating Organization for Nuclear Power Plants, which identifies fuel management as one of the various functions to be performed by the operating organization. The purpose of this Safety Guide is to provide recommendations for core management and fuel handling at nuclear power plants on the basis of current international good practice. The present Safety Guide addresses those aspects of fuel management activities that are necessary in order to allow optimum reactor core operation without compromising the limits imposed by the design safety considerations relating to the nuclear fuel and the plant as a whole. In this publication, 'core management' refers to those activities that are associated with fuel management in the core and reactivity control, and 'fuel handling' refers to the movement, storage and control of fresh and irradiated fuel. Fuel management comprises both core management and fuel handling. This Safety Guide deals with fuel management for all types of land based stationary thermal neutron power plants. It describes the safety objectives of core management, the tasks that have to be accomplished to meet these objectives and the activities undertaken to perform those tasks. It also deals with the receipt of fresh fuel, storage and handling of fuel and other core components, the loading and unloading of fuel and core components, and the insertion and removal of other reactor materials. In addition, it deals with loading a transport container with irradiated fuel and its preparation for transport off the site. Transport

  2. Core management and fuel handling for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2002-01-01

    This Safety Guide supplements and elaborates upon the safety requirements for core management and fuel handling that are presented in Section 5 of the Safety Requirements publication on the operation of nuclear power plants. The present publication supersedes the IAEA Safety Guide on Safety Aspects of Core Management and Fuel Handling, issued in 1985 as Safety Series No. 50-SG-010. It is also related to the Safety Guide on the Operating Organization for Nuclear Power Plants, which identifies fuel management as one of the various functions to be performed by the operating organization. The purpose of this Safety Guide is to provide recommendations for core management and fuel handling at nuclear power plants on the basis of current international good practice. The present Safety Guide addresses those aspects of fuel management activities that are necessary in order to allow optimum reactor core operation without compromising the limits imposed by the design safety considerations relating to the nuclear fuel and the plant as a whole. In this publication, 'core management' refers to those activities that are associated with fuel management in the core and reactivity control, and 'fuel handling' refers to the movement, storage and control of fresh and irradiated fuel. Fuel management comprises both core management and fuel handling. This Safety Guide deals with fuel management for all types of land based stationary thermal neutron power plants. It describes the safety objectives of core management, the tasks that have to be accomplished to meet these objectives and the activities undertaken to perform those tasks. It also deals with the receipt of fresh fuel, storage and handling of fuel and other core components, the loading and unloading of fuel and core components, and the insertion and removal of other reactor materials. In addition, it deals with loading a transport container with irradiated fuel and its preparation for transport off the site. Transport

  3. Geotechnical aspects of site evaluation and foundations for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2003-01-01

    This publication is a revision of the former safety standards of IAEA Safety Series No. 50-SG-S8. The scope has been extended to cover not only foundations but also design questions related to geotechnical science and engineering, such as the bearing capacity of foundations, design of earth structures and design of buried structures. Seismic aspects also play an important role in this field, and consequently the Safety Guide on Evaluation of Seismic Hazards for Nuclear Power Plants, Safety Standards Series No. NS-G-3.3, which discusses the determination of seismic input motion, is referenced on several occasions. The present Safety Guide provides an interpretation of the Safety Requirements on Site Evaluation for Nuclear Installations and guidance on how to implement them. It is intended for the use of safety assessors or regulators involved in the licensing process as well as the designers of nuclear power plants, and it provides them with guidance on the methods and procedures for analyses to support the assessment of the geotechnical aspects of the safety of nuclear power plants

  4. Geotechnical aspects of site evaluation and foundations for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2006-01-01

    This publication is a revision of the former safety standards of IAEA Safety Series No. 50-SG-S8. The scope has been extended to cover not only foundations but also design questions related to geotechnical science and engineering, such as the bearing capacity of foundations, design of earth structures and design of buried structures Seismic aspects also play an important role in this field, and consequently the Safety Guide on Evaluation of Seismic Hazards for Nuclear Power Plants, Safety Standards Series No. NS-G-3.3, which discusses the determination of seismic input motion, is referenced on several occasions. The present Safety Guide provides an interpretation of the Safety Requirements on Site Evaluation for Nuclear Installations and guidance on how to implement them. It is intended for the use of safety assessors or regulators involved in the licensing process as well as the designers of nuclear power plants, and it provides them with guidance on the methods and procedures for analyses to support the assessment of the geotechnical aspects of the safety of nuclear power plants

  5. Design of the reactor coolant system and associated systems in nuclear power plants. Safety guide (Spanish Edition)

    International Nuclear Information System (INIS)

    2010-01-01

    This Safety Guide was prepared under the IAEA programme for establishing safety standards for nuclear power plants. The basic requirements for the design of safety systems for nuclear power plants are established in the Safety Requirements publication, Safety Standards Series No. NS-R-1 on Safety of Nuclear Power Plants: Design, which it supplements. This Safety Guide describes how the requirements for the design of the reactor coolant system (RCS) and associated systems in nuclear power plants should be met. This publication is a revision and combination of two previous Safety Guides, Safety Series No. 50-SG-D6 on Ultimate Heat Sink and Directly Associated Heat Transport Systems for Nuclear Power Plants (1982), and Safety Series No. 50-SG-D13 on Reactor Coolant and Associated Systems in Nuclear Power Plants (1987), which are superseded by this new Safety Guide. The revision takes account of developments in the design of the RCS and associated systems in nuclear power plants since the earlier Safety Guides were published in 1982 and 1987, respectively. The other objectives of the revision are to ensure consistency with Ref., issued in 2004, and to update the technical content. In addition, an appendix on pressurized heavy water reactors (PHWRs) has been included.

  6. The German nuclear power plant safety study

    International Nuclear Information System (INIS)

    1979-01-01

    With this study a new approach has been chosen, taking nuclear power plants as an example to assess and to describe the risks arising from the use of modern technology, including those hazards emanating from the rather hypothetical possibility of occurrence of very serious accidents. Following the definition of basic concepts and methods to be applied in risk assessment studied, as well as a brief account of the design and operating mode of nuclear power plants with PWRs', accidents and failures to be considered in a safety study are described. Using the course-of-event and fault tree analysis, the probability of fission product release as a consequence of failures in safety systems or of core meltdown is evaluated. Subsequently, the theoretical model for assessment of reactor accident consequences is presented, discussing such aspects as the dispersion of radioactivity in the atmosphere, the radiation dose model, safety and countermeasures, the model for the evaluation of health hazards as well as methods and calculations for estimating the reliability of risk assessments together with the remaining uncertainties. In an appendix to this study, the analyses presented in the study are discussed in the light of the TMI-2 event. This safety study showing the possibilities of detecting, keeping in check and minimizing harmful effects, can be regarded as a contribution to a better understanding of our modern, highly industrialised society, and eventually to an improvement of the quality of life. (GL) 891 GL/GL 892 MB [de

  7. Social contention about safety of nuclear power plant

    International Nuclear Information System (INIS)

    Nemoto, Kazuyasu

    1978-01-01

    In Japan, the contentions and arguments on the safety of nuclear power generation have been active since its first introduction, and these are greatly influenced by the nation's experiences of atomic bombs in Hiroshima, Nagasaki, and Bikini. As the result, the attitude of peoples toward the acceptance of nuclear power plants is significantly different from that in other countries. The situation in Japan of social contentions about nuclear power safety is explained in two aspects: acceptance of the safety, by peoples and Japanese pattern of safety contentions. In both upstream and downstream of nuclear power generation, not only the safety but also the right or wrong for nuclear power generation itself is discussed. The problem of nuclear power safety has gone into the region beyond the technological viewpoint. The pattern of safety contentions in Japan is the entanglement of three sectors; i.e. local people, labor unions and political parties, enterprises and administration, and intellectuals. (Mori, K.)

  8. IAEA Sees Safety Commitment at Spainโ€™s Almaraz Nuclear Power Plant, Areas for Enhancement

    International Nuclear Information System (INIS)

    2018-01-01

    An International Atomic Energy Agency (IAEA) team of experts said the operator of Spainโ€™s Almaraz Nuclear Power Plant demonstrated a commitment to the long-term safety of the plant and noted several good practices to share with the nuclear industry globally. The team also identified areas for further enhancement. The Operational Safety Review Team (OSART) today concluded an 18-day mission to Almaraz, whose two 1,050-MWe pressurized-water reactors started commercial operation in 1983 and 1984, respectively. Centrales Nucleares Almaraz-Trillo (CNAT) operates the plant, located about 200 km southwest of Madrid. OSART missions aim to improve operational safety by objectively assessing safety performance using the IAEAโ€™s safety standards and proposing recommendations for improvement where appropriate. Nuclear power generates more than 21 per cent of electricity in Spain, whose seven operating power reactors all began operation in the 1980s.The mission made a number of recommendations to improve operational safety, including: โ€ข The plant should implement further actions related to management, staff and contractors to enforce standards and expectations related to industrial safety. โ€ข The plant should take measures to reinforce and implement standards to enhance the performance of reactivity manipulations in a deliberate and carefully-controlled manner. โ€ข The plant should improve the support, training and documented guidance for Severe Accident Management Guideline users in order to mitigate complex severe accident scenarios. The team provided a draft report of the mission to the plantโ€™s management. The plant management and the Nuclear Safety Council (CSN), which is responsible for nuclear safety oversight in Spain, will have the opportunity to make factual comments on the draft. These will be reviewed by the IAEA and the final report will be submitted to the Government of Spain within three months. The plant management said it would address the areas

  9. Regulatory practices and safety standards for nuclear power plants

    International Nuclear Information System (INIS)

    1989-01-01

    The International Symposium on Regulatory Practices and Safety Standards for Nuclear Power Plants was jointly organized by the International Atomic Energy Agency (IAEA), for Nuclear Energy Agency of the OECD and the Government of the Federal Republic of Germany with the objective of providing an international forum for the exchange of information on regulatory practices and safety standards for nuclear power plants. The Symposium was held in Munich, Federal Republic of Germany, from 7 to 10 November 1988. It was attended by 201 experts from some 32 Member States and 4 international organizations. Fifty-one papers from 19 Member States and 2 international organizations were presented and discussed in 5 technical sessions covering the following subjects: National Regulatory Practices and Safety Standards (14 papers); Implementation of Regulatory Practices - Technical Issues (8 papers); Implementation of Regulatory Practices - Operational Aspects (8 papers); Developments and Trends in Safety Standards and Practices (11 papers); International Aspects (10 papers). A separate abstract was prepared for each of these papers. Refs, figs and tabs

  10. Design characteristics of safety parameter display system for nuclear power plants

    International Nuclear Information System (INIS)

    Zhang Yuangfang

    1992-02-01

    The design features of safety parameter display system (SPDS) developed by Tsinghua University is introduced. Some new features have been added into the system functions and they are: (1) hierarchical display structure; (2) human factor in the display format design; (3)automatic diagnosis of safety status of nuclear power plant; (4) extension of SPDS use scope; (5) flexible hardware structure. The new approaches in the design are: (1)adopting the international design standards; (2) selecting safety parameters strictly; (3) developing software under multitask operating system; (4) using a nuclear power plant simulator to verify the SPDS design

  11. Basic Safety Considerations for Nuclear Power Plant Dealing with External Human Induced Events

    Energy Technology Data Exchange (ETDEWEB)

    Salem, W., E-mail: wafaasalem21@yahoo.com [Nuclear and Radiological Regulatory Authority (Egypt)

    2014-10-15

    Facilities and human activities in the region in which a nuclear power plant is located may under some conditions affect its safety. The potential sources of human induced events external to the plant should be identified and the severity of the possible resulting hazard phenomena should be evaluated to derive the appropriate design bases for the plant. They should also be monitored and periodically assessed over the lifetime of the plant to ensure that consistency with the design assumptions is maintained. External human induced events that could affect safety should be investigated in the site evaluation stage for every nuclear power plant site. The region is required to be examined for facilities and human activities that have the potential, under certain conditions, to endanger the nuclear power plant over its entire lifetime. Each relevant potential source is required to be identified and assessed to determine the potential interactions with personnel and plant items important to safety. (author)

  12. Probabilistic safety assessment technology for commercial nuclear power plant security evaluation

    International Nuclear Information System (INIS)

    Liming, J.K.; Johnson, D.H.; Dykes, A.A.

    2004-01-01

    Commercial nuclear power plant physical security has received much more intensive treatment and regulatory attention since September 11, 2001. In light of advancements made by the nuclear power industry in the field of probabilistic safety assessment (PSA) for its power plants over that last 30 years, and given the many examples of successful applications of risk-informed regulation at U. S. nuclear power plants during recent years, it may well be advisable to apply a 'risk-informed' approach to security management at nuclear power plants from now into the future. In fact, plant PSAs developed in response to NRC Generic Letter 88-20 and related requirements are used to help define target sets of critical plant safety equipment in our current security exercises for the industry. With reasonable refinements, plant PSAs can be used to identify, analyze, and evaluate reasonable and prudent approaches to address security issues and associated defensive strategies at nuclear power plants. PSA is the ultimate scenario-based approach to risk assessment, and thus provides a most powerful tool in identifying and evaluating potential risk management decisions. This paper provides a summary of observations of factors that are influencing or could influence cost-effective or 'cost-reasonable' security management decision-making in the current political environment, and provides recommendations for the application of PSA tools and techniques to the nuclear power plant operational safety response exercise process. The paper presents a proposed framework for nuclear power plant probabilistic terrorist risk assessment that applies these tools and techniques. (authors)

  13. Conduct of Operations at Nuclear Power Plants. Safety Guide (Spanish Edition)

    International Nuclear Information System (INIS)

    2012-01-01

    This Safety Guide identifies the main responsibilities and practices of nuclear power plant (NPP) operations departments in relation to their responsibility for the safe functioning of the plant. The guide presents the factors to be considered in structuring the operations department of an NPP; setting high standards of performance; making safety related decisions in an effective manner; conducting control room and field activities in a thorough and professional manner; and maintaining an NPP within established operational limits and conditions. Contents: 1. Introduction; 2. Management and organization of plant operations; 3. Shift complement and functions; 4. Shift routines and operating practices; 5. Control of equipment and plant status; 6. Operations equipment and operator aids; 7. Work control and authorization.

  14. Nuclear power plants. Electrical equipment of the safety system. Qualification

    International Nuclear Information System (INIS)

    2001-01-01

    This International Standard applies to electrical parts of safety systems employed at nuclear power plants, including components and equipment of any interface whose failure could affect unfavourably properties of the safety system. The standard also applies to non-electrical safety-related interfaces. Furthermore, the standard describes the generic process of qualification certification procedures and methods of qualification testing and related documentation. (P.A.)

  15. Upgrading safety documentation for exported nuclear power plants

    International Nuclear Information System (INIS)

    Rosen, M.

    1978-01-01

    In view of the generally small regulatory staffs of importing countries, suggestions are given for upgrading the ''export edition'' of the traditionally supplied safety documentation by use of a Supplementary Information Report, written specifically for the needs of a smaller and/or less technically qualified staff, which would highlight the differences that exist between the facility to be constructed and the supposedly similar reference plant of the supplier country; by improvement of supporting safety documentation to allow for adequate understanding of significant safety parameters; and by attention to the needs of smaller countries in the critical operating regulations (Technical Specifications for Operation). (author)

  16. Developments in safety and operations culture in BNFL's thorp reprocessing plant, Sellafield, Cumbria

    International Nuclear Information System (INIS)

    Kett, P.J.

    2000-01-01

    One of the best descriptions of Culture is 'how we do things around here'. In a stable organisation it is extremely difficult to change any type of culture, whether it is an operations, customer service or safety culture. To change culture one of two elements are essential. There must be either a significant external pressure felt by all in the organisation or a change in senior management, with authority to set a new direction for the organisation. BNFL had a unique opportunity through the commissioning and operation of the Thorp Reprocessing Plant at Sellafield to shape a new Safety and Operations Culture. Both the key elements for change were present. Thorp was a high profile flagship plant that had attracted multinational investment. It incorporated new technology. The workforce had volunteered to operate the plant. A strong senior management team was specially selected. The plant was being commissioned in an environment where there was significant opposition by 'anti nuclear' groups. It was essential to both BNFL and the wider international nuclear community that Thorp was commissioned and operated safely. A strong operating culture was developed with safety as the corner stone. The culture comprises three key components. Rigorous plant safety case and risk assessments before work commences and modifications to the plant occur; A high level of involvement by all levels of the workforce in both operations and safety matters; Strong supportive leadership which does not allow safety standards to be compromised and encourages open debate on how to improve. During commissioning and early operation of Thorp the robustness of the Safety and Operations Culture was demonstrated. On several occasions, despite intense commercial pressure, operations were halted until the situation was resolved both technically and procedurally. This paper describes how the Safety and Operations Culture was developed. The key factors for success include recruitment, team selection

  17. Integrated Plant Safety Assessment: Systematic Evaluation Program. Haddam Neck Plant, Connecticut Yankee Atomic Power Company, Docket No. 50-213. Final report

    International Nuclear Information System (INIS)

    1983-01-01

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

  18. Personnel Safety for Future Magnetic Fusion Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Lee Cadwallader

    2009-07-01

    The safety of personnel at existing fusion experiments is an important concern that requires diligence. Looking to the future, fusion experiments will continue to increase in power and operating time until steady state power plants are achieved; this causes increased concern for personnel safety. This paper addresses four important aspects of personnel safety in the present and extrapolates these aspects to future power plants. The four aspects are personnel exposure to ionizing radiation, chemicals, magnetic fields, and radiofrequency (RF) energy. Ionizing radiation safety is treated well for present and near-term experiments by the use of proven techniques from other nuclear endeavors. There is documentation that suggests decreasing the annual ionizing radiation exposure limits that have remained constant for several decades. Many chemicals are used in fusion research, for parts cleaning, as use as coolants, cooling water cleanliness control, lubrication, and other needs. In present fusion experiments, a typical chemical laboratory safety program, such as those instituted in most industrialized countries, is effective in protecting personnel from chemical exposures. As fusion facilities grow in complexity, the chemical safety program must transition from a laboratory scale to an industrial scale program that addresses chemical use in larger quantity. It is also noted that allowable chemical exposure concentrations for workers have decreased over time and, in some cases, now pose more stringent exposure limits than those for ionizing radiation. Allowable chemical exposure concentrations have been the fastest changing occupational exposure values in the last thirty years. The trend of more restrictive chemical exposure regulations is expected to continue into the future. Other issues of safety importance are magnetic field exposure and RF energy exposure. Magnetic field exposure limits are consensus values adopted as best practices for worker safety; a typical

  19. Personnel Safety for Future Magnetic Fusion Power Plants

    International Nuclear Information System (INIS)

    Cadwallader, Lee

    2009-01-01

    The safety of personnel at existing fusion experiments is an important concern that requires diligence. Looking to the future, fusion experiments will continue to increase in power and operating time until steady state power plants are achieved; this causes increased concern for personnel safety. This paper addresses four important aspects of personnel safety in the present and extrapolates these aspects to future power plants. The four aspects are personnel exposure to ionizing radiation, chemicals, magnetic fields, and radiofrequency (RF) energy. Ionizing radiation safety is treated well for present and near-term experiments by the use of proven techniques from other nuclear endeavors. There is documentation that suggests decreasing the annual ionizing radiation exposure limits that have remained constant for several decades. Many chemicals are used in fusion research, for parts cleaning, as use as coolants, cooling water cleanliness control, lubrication, and other needs. In present fusion experiments, a typical chemical laboratory safety program, such as those instituted in most industrialized countries, is effective in protecting personnel from chemical exposures. As fusion facilities grow in complexity, the chemical safety program must transition from a laboratory scale to an industrial scale program that addresses chemical use in larger quantity. It is also noted that allowable chemical exposure concentrations for workers have decreased over time and, in some cases, now pose more stringent exposure limits than those for ionizing radiation. Allowable chemical exposure concentrations have been the fastest changing occupational exposure values in the last thirty years. The trend of more restrictive chemical exposure regulations is expected to continue into the future. Other issues of safety importance are magnetic field exposure and RF energy exposure. Magnetic field exposure limits are consensus values adopted as best practices for worker safety; a typical

  20. Nuclear accidents and safety measures of domestic nuclear power plants

    International Nuclear Information System (INIS)

    Song Zurong; Che Shuwei; Pan Xiang

    2012-01-01

    Based on the design standards for the safety of nuclear and radiation in nuclear power plants, the three accidents in the history of nuclear power are analyzed. And the main factors for these accidents are found out, that is, human factors and unpredicted natural calamity. By combining the design and operation parameters of domestic nuclear plants, the same accidents are studied and some necessary preventive schemes are put forward. In the security operation technology of domestic nuclear power plants nowadays, accidents caused by human factors can by prevented completely. But the safety standards have to be reconsidered for the unpredicted neutral disasters. How to reduce the hazard of nuclear radiation and leakage to the level that can be accepted by the government and public when accidents occur under extreme conditions during construction and operation of nuclear power plants must be considered adequately. (authors)

  1. Safety assessment of plant food supplements (PFS)

    NARCIS (Netherlands)

    Berg, van den S.J.P.L.; Serra-Majem, L.; Coppens, P.; Rietjens, I.

    2011-01-01

    Botanicals and botanical preparations, including plant food supplements (PFS), are widely used in Western diets. The growing use of PFS is accompanied by an increasing concern because the safety of these PFS is not generally assessed before they enter the market. Regulatory bodies have become more

  2. Comparative safety assessment of plant-derived foods

    NARCIS (Netherlands)

    Kok, E.J.; Keijer, J.; Kleter, G.A.; Kuiper, H.A.

    2008-01-01

    The second generation of genetically modified (GM) plants that are moving towards the market are characterized by modifications that may be more complex and traits that more often are to the benefit of the consumer. These developments will have implications for the safety assessment of the resulting

  3. Selecting safety standards for nuclear power plants

    International Nuclear Information System (INIS)

    1981-01-01

    Today, many thousands of documents are available describing the requirements, guidelines, and industrial standards which can be used as bases for a nuclear power plant programme. Many of these documents relate to nuclear safety which is currently the focus of world-wide attention. The multitude of documents available on the subject, and their varying status and emphasis, make the processes of selection and implementation very important. Because nuclear power plants are technically intricate and advanced, particularly in relation to the technological status of many developing countries, these processes are also complicated. These matters were the subject of a seminar held at the Agency's headquarters in Vienna last December. The IAEA Nuclear Safety Standards (NUSS) programme was outlined and explained at the Seminar. The five areas of the NUSS programme for nuclear power plants cover, governmental organization, siting, design; operation; quality assurance. In each area the Agency has issued Codes of Practice and is developing Safety Guides. These provide regulatory agencies with a framework for safety. The Seminar recognized that the NUSS programme should enable developing countries to identify priorities in their work, particularly the implementation of safety standards. The ISO activities in the nuclear field are carried out in the framework of its Technical Committee 85 (ISO/TC85). The work is distributed in sub-committees. Seminar on selection and implementation of safety standards for nuclear power plants, jointly organized by the IAEA and the International Organization for Standardization (ISO), and held in Vienna from 15 to 18 December 1980 concerned with: terminology, definitions, units and symbols (SC-1), radiation protection (SC-2), power reactor technology (SC-3), nuclear fuel technology (SC-5). There was general agreement that the ISO standards are complementary to the NUSS codes and guides. ISO has had close relations with the IAEA for several years

  4. Using game technologies to improve the safety of construction plant operations.

    Science.gov (United States)

    Guo, Hongling; Li, Heng; Chan, Greg; Skitmore, Martin

    2012-09-01

    Many accidents occur world-wide in the use of construction plant and equipment, and safety training is considered by many to be one of the best approaches to their prevention. However, current safety training methods/tools are unable to provide trainees with the hands-on practice needed. Game technology-based safety training platforms have the potential to overcome this problem in a virtual environment. One such platform is described in this paper - its characteristics are analysed and its possible contribution to safety training identified. This is developed and tested by means of a case study involving three major pieces of construction plant, which successfully demonstrates that the platform can improve the process and performance of the safety training involved in their operation. This research not only presents a new and useful solution to the safety training of construction operations, but illustrates the potential use of advanced technologies in solving construction industry problems in general. Copyright ยฉ 2011 Elsevier Ltd. All rights reserved.

  5. Complex nuclear safety evaluation of the Bohunice V-1 nuclear power plant

    International Nuclear Information System (INIS)

    Kriz, Z.

    1991-01-01

    The safety concept of V-230 type reactor units dates back to the late 1960s. The units fail to be sufficiently dimensioned for emergency cooling of the reactor core and are fitted with no containment. So far, operating experience is good. The availability factor is 71.5% for unit 1 and 77.8% for unit 2. There occur 1 to 3 unscheduled shutdowns annually. The quality of steam generator tubes is very good. A complex safety assessment of the plant was accomplished in 1990. It concerned the concept and criteria of safety assessment, the earthquake situation, the condition of the primary coolant circuit equipment, the control system, the effect of the human factor, and preparedness of emergency plans. OSART and ASSET missions were accomplished at the plant. Based on the results of the missions as well as of inspections by the State Surveillance over Nuclear Safety, the decision has been adopted to operate the plant not longer than till 1995; the further fate of the plant will be decided on according to a future technical and economic analysis. (M.D.)

  6. Self-assessment of operational safety for nuclear power plants

    International Nuclear Information System (INIS)

    1999-12-01

    Self-assessment processes have been continuously developed by nuclear organizations, including nuclear power plants. Currently, the nuclear industry and governmental organizations are showing an increasing interest in the implementation of this process as an effective way for improving safety performance. Self-assessment involves the use of different types of tools and mechanisms to assist the organizations in assessing their own safety performance against given standards. This helps to enhance the understanding of the need for improvements, the feeling of ownership in achieving them and the safety culture as a whole. Although the primary beneficiaries of the self-assessment process are the plant and operating organization, the results of the self-assessments are also used, for example, to increase the confidence of the regulator in the safe operation of an installation, and could be used to assist in meeting obligations under the Convention on Nuclear Safety. Such considerations influence the form of assessment, as well as the type and detail of the results. The concepts developed in this report present the basic approach to self-assessment, taking into consideration experience gained during Operational Safety Review Team (OSART) missions, from organizations and utilities which have successfully implemented parts of a self-assessment programme and from meetings organized to discuss the subject. This report will be used in IAEA sponsored workshops and seminars on operational safety that include the topic of self-assessment

  7. Kozloduy nuclear power plant. Units 1-4. Status of safety assessment activities. Rev. 2

    International Nuclear Information System (INIS)

    1999-01-01

    This paper presents the results of the status of safety assessment activities carried out by the Kozloduy Nuclear Power Plant (KNPP) in order to evaluate the current status of the safety of its reactor units 1-4. The steam supply system of this units is based of the reactor WWER-440/ B-230, which is a PWR of Russian design developed according to the safety standards in force in USSR in late 60-s. Now a days 10 reactor units of this type are in operation in four NPPs. Despite of efforts of the different plants to implement safety improvements measures during first 10-15 years of operation of this type of reactor its major safety problems were not eliminated and were a subject of international concern. The systematic evaluation of the deficiencies of the original design of this type of reactors have been initiated by IAEA in the beginning of 1990 and brought to developing a comprehensive list of safety problems which required urgent implementation of safety measures in all plants. To solve this problems in 1991 KNPP initiated implementation of so called 'short term' safety improvement program, developed with the help of WANO under agreement with Bulgarian Nuclear Safety Authority (BNSA) and consortium RISKAUDIT. The program was based on a stage approach and was foreseen to be implemented by tree stages in very tight time schedule in order to achieve significant and rapid improvements of the level of safety in operation of the units. The Short Tenn Program was implemented between the years 1991 and 1997 thanks of the strong safety commitment of NEK and KNPP staff and the broad international cooperation and financial support. Important part of resources were supplied under PHARE program of CEC, EBRD grant agreement and EDF support. The plant current safety level analysis has been performed using IAEA analytical methodology according to 50-SG-O12 standard 'Periodic safety review of operational nuclear power plants'. The approach and criteria for acceptable safety level

  8. New requirements on safety of nuclear power plants according to the IAEA safety standards

    International Nuclear Information System (INIS)

    Misak, J.

    2005-01-01

    In this presentation author presents new requirements on safety of nuclear power plants according to the IAEA safety standards. It is concluded that: - New set of IAEA Safety Standards is close to completion: around 40 standards for NPPs; - Different interpretation of IAEA Safety Standards at present: best world practices instead of previous 'minimum common denominator'; - A number of safety improvements required for NPPs; - Requirements related to BDBAs and severe accidents are the most demanding due to degradation of barriers: hardware modifications and accident management; - Large variety between countries in implementation of accident management programmes: from minimum to major hardware modifications; -Distinction between existing and new NPPs is essential from the point of view of the requirements; WWER 440 reactors have potential to reflect IAEA Safety Standards for existing NPPs; relatively low reactor power offers broader possibilities

  9. The reevaluation of seismic safety of existing nuclear power plant

    International Nuclear Information System (INIS)

    Kitagawa, Hiroshi; Tominaga, Shohei; Kumagai, Chiyoshi; Koshiba, Koremutsu; Kono, Tomonori; Agawa, Kazuyoshi; Kuwata, Kenichiro

    2003-01-01

    We have carried out additional geological surveys in order to enrich our database on geological faults in the vicinity of Shimane Nuclear Power Plant (NPP). Prior to additional geological surveys, given the social importance of nuclear power plants, we hypothetically assumed that almost the whole length of an area covered by surveys would be an active fault that must be considered in seismic design, and tried to reevaluate the seismic safety of the NPP by applying an input earthquake ground motion larger than the level at the design stage. As a result, we have confirmed that seismic safety of the NPP can be maintained. This paper describes the method that we employed to reevaluate the seismic safety of Shimane NPP. (author)

  10. Development of a dynamical systems model of plant programmatic performance on nuclear power plant safety risk

    International Nuclear Information System (INIS)

    Hess, Stephen M.; Albano, Alfonso M.; Gaertner, John P.

    2005-01-01

    Application of probabilistic risk assessment (PRA) techniques to model nuclear power plant accident sequences has provided a significant contribution to understanding the potential initiating events, equipment failures and operator errors that can lead to core damage accidents. Application of the lessons learned from these analyses has resulted in significant improvements in plant operation and safety. However, this approach has not been nearly as successful in addressing the impact of plant processes and management effectiveness on the risks of plant operation. The research described in this paper presents an alternative approach to addressing this issue. In this paper we propose a dynamical systems model that describes the interaction of important plant processes on nuclear safety risk. We discuss development of the mathematical model including the identification and interpretation of significant inter-process interactions. Next, we review the techniques applicable to analysis of nonlinear dynamical systems that are utilized in the characterization of the model. This is followed by a preliminary analysis of the model that demonstrates that its dynamical evolution displays features that have been observed at commercially operating plants. From this analysis, several significant insights are presented with respect to the effective control of nuclear safety risk. As an important example, analysis of the model dynamics indicates that significant benefits in effectively managing risk are obtained by integrating the plant operation and work management processes such that decisions are made utilizing a multidisciplinary and collaborative approach. We note that although the model was developed specifically to be applicable to nuclear power plants, many of the insights and conclusions obtained are likely applicable to other process industries

  11. IR-360 nuclear power plant safety functions and component classification

    International Nuclear Information System (INIS)

    Yousefpour, F.; Shokri, F.; Soltani, H.

    2010-01-01

    The IR-360 nuclear power plant as a 2-loop PWR of 360 MWe power generation capacity is under design in MASNA Company. For design of the IR-360 structures, systems and components (SSCs), the codes and standards and their design requirements must be determined. It is a prerequisite to classify the IR-360 safety functions and safety grade of structures, systems and components correctly for selecting and adopting the suitable design codes and standards. This paper refers to the IAEA nuclear safety codes and standards as well as USNRC standard system to determine the IR-360 safety functions and to formulate the principles of the IR-360 component classification in accordance with the safety philosophy and feature of the IR-360. By implementation of defined classification procedures for the IR-360 SSCs, the appropriate design codes and standards are specified. The requirements of specific codes and standards are used in design process of IR-360 SSCs by design engineers of MASNA Company. In this paper, individual determination of the IR-360 safety functions and definition of the classification procedures and roles are presented. Implementation of this work which is described with example ensures the safety and reliability of the IR-360 nuclear power plant.

  12. IR-360 nuclear power plant safety functions and component classification

    Energy Technology Data Exchange (ETDEWEB)

    Yousefpour, F., E-mail: fyousefpour@snira.co [Management of Nuclear Power Plant Construction Company (MASNA) (Iran, Islamic Republic of); Shokri, F.; Soltani, H. [Management of Nuclear Power Plant Construction Company (MASNA) (Iran, Islamic Republic of)

    2010-10-15

    The IR-360 nuclear power plant as a 2-loop PWR of 360 MWe power generation capacity is under design in MASNA Company. For design of the IR-360 structures, systems and components (SSCs), the codes and standards and their design requirements must be determined. It is a prerequisite to classify the IR-360 safety functions and safety grade of structures, systems and components correctly for selecting and adopting the suitable design codes and standards. This paper refers to the IAEA nuclear safety codes and standards as well as USNRC standard system to determine the IR-360 safety functions and to formulate the principles of the IR-360 component classification in accordance with the safety philosophy and feature of the IR-360. By implementation of defined classification procedures for the IR-360 SSCs, the appropriate design codes and standards are specified. The requirements of specific codes and standards are used in design process of IR-360 SSCs by design engineers of MASNA Company. In this paper, individual determination of the IR-360 safety functions and definition of the classification procedures and roles are presented. Implementation of this work which is described with example ensures the safety and reliability of the IR-360 nuclear power plant.

  13. Radiation protection aspects in the design of nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2008-01-01

    . The IAEA takes seriously the enduring challenge for users and regulators everywhere: that of ensuring a high level of safety in the use of nuclear materials and radiation sources around the world. Their continuing utilization for the benefit of humankind must be managed in a safe manner, and the IAEA safety standards are designed to facilitate the achievement of that goal. This Safety Guide has been prepared as a part of the IAEA programme on safety standards for nuclear power plants. It includes recommendations on how to satisfy the requirements established in the Safety Requirements publication on the Safety of Nuclear Power Plants: Design. It addresses the provisions that should be made in the design of nuclear power plants in order to protect site personnel, the public and the environment against radiological hazards for operational states, decommissioning and accident conditions. The recommendations on radiation protection provided in this Safety Guide are consistent with the International Basic Safety Standards for Protection against Ionizing Radiation and for the Safety of Radiation Sources (BSS), which were jointly sponsored by the Food and Agriculture Organization of the United Nations (FAO), the IAEA, the International Labour Organisation (ILO), the OECD Nuclear Energy Agency (OECD/NEA), the Pan American Health Organization (PAHO) and the World Health Organization (WHO). This Safety Guide supersedes Safety Series No. 50-SG-D9, Design Aspects of Radiation Protection for Nuclear Power Plants, published in 1985. Effective radiation protection is a combination of good design, high quality construction and proper operation. Procedures that address the radiation protection aspects of operation are covered in the Safety Guide on Radiation Protection and Radioactive Waste Management in the operation of Nuclear Power Plants

  14. Safety aspects of nuclear plant licensing in Canada

    International Nuclear Information System (INIS)

    Jennekens, J.H.F.

    1975-01-01

    The legislative authority is laid down in the Atomic Energy Control Act, 1946, declaring atomic energy a matter of national interest and establishing the Atomic Energy Control Board (AECB) as the competent body for regulating all aspects of atomic energy. The Act also vests a Minister designated by the Government with research and exploitation functions; thus, by Ministerial order, Atomic Energy of Canada Limited was established in 1952 as a State-owned company. The Nuclear Liability Act, 1970, channels all liability for nuclear damage to the operator of a nuclear installation and requires him to obtain insurance in the amount of $75 million, part of which may be re-insured by the Government. The licensing requirements comprise the issuance of a site approval, a construction licence and an operating licence. The AECB is assisted in its licensing functions by its Nuclear Plant Licensing Directorate and by the Reactor Safety Advisory Committee co-operating with each other in making extensive safety assessments of a licence application. A site evaluation report, a preliminary safety report and a final safety report are required in relation to the siting, construction and operation of a nuclear power plant. The Canadian reactor safety philosophy is based on the concept of defence in depth, implemented through a multi-step approach, which includes avoidance of malfunctions, provision of special safety systems, periodic inspection and testing, and avoidance of human errors. Specific criteria and principles have evolved in applying this basic safety philosophy and radiation protection standards are derived from international recommendations. Stringent control is exercised over the management of radioactive waste and management facilities must meet the engineering and procedural requirements of AECB before they can be placed in operation. (author)

  15. Discussion of important safety requirements for new nuclear power plants

    International Nuclear Information System (INIS)

    Zhang Lin; Jia Xiang; Yan Tianwen; Li Wenhong; Li Chun

    2014-01-01

    This paper presents the analysis of several important safety requirements and improvement direction. Technical view of security goals on site safety evaluation, internal and external events fortification, serious accident prevention and mitigation, as well as the core, containment system and instrument control system design and engineering optimization, and etc are indicated. It will be useful for new plant design, construction and safety improvement. (authors)

  16. European passive plant program preliminary safety analyses to support system design

    International Nuclear Information System (INIS)

    Saiu, Gianfranco; Barucca, Luciana; King, K.J.

    1999-01-01

    In 1994, a group of European Utilities, together with Westinghouse and its Industrial Partner GENESI (an Italian consortium including ANSALDO and FIAT), initiated a program designated EPP (European Passive Plant) to evaluate Westinghouse Passive Nuclear Plant Technology for application in Europe. In the Phase 1 of the European Passive Plant Program which was completed in 1996, a 1000 MWe passive plant reference design (EP1000) was established which conforms to the European Utility Requirements (EUR) and is expected to meet the European Safety Authorities requirements. Phase 2 of the program was initiated in 1997 with the objective of developing the Nuclear Island design details and performing supporting analyses to start development of Safety Case Report (SCR) for submittal to European Licensing Authorities. The first part of Phase 2, 'Design Definition' phase (Phase 2A) was completed at the end of 1998, the main efforts being design definition of key systems and structures, development of the Nuclear Island layout, and performing preliminary safety analyses to support design efforts. Incorporation of the EUR has been a key design requirement for the EP1000 form the beginning of the program. Detailed design solutions to meet the EUR have been defined and the safety approach has also been developed based on the EUR guidelines. The present paper describes the EP1000 approach to safety analysis and, in particular, to the Design Extension Conditions that, according to the EUR, represent the preferred method for giving consideration to the Complex Sequences and Severe Accidents at the design stage without including them in the design bases conditions. Preliminary results of some DEC analyses and an overview of the probabilistic safety assessment (PSA) are also presented. (author)

  17. Experience gained in enhancing operational safety at ComEd's nuclear power plants

    International Nuclear Information System (INIS)

    Elias, D.

    1997-01-01

    The following aspects of experience gained in enhancing operational safety at Comed's nuclear power plants are discussed: nuclear safety policy; centralization/decentralization; typical nuclear operating organization; safety review boards; human performance enhancement; elements of effective nuclear oversight

  18. The safety approach in the operation of EDF power plants

    International Nuclear Information System (INIS)

    Bertron, L.; Mira, J.J.

    1988-01-01

    To get a view on what is involved in maintaining a high level of safety in the operation of EdF nuclear power plants, it may be recalled that in 1987, 76 % of the EdF production was nuclear. The nuclear plants include thirty-four standard PWR 900 plants, fourteen PWR 1300 plants, the 305 MW SENA PWR, the four 500 MW GCR: CHINON A3 plant, St-LAURENT A1 (390 MW), A2 (450 MW) and BUGEY 1 (540 MW), the 233 MW PHENIX fast breeder reactor and the CREYS-MALVILLE 1200 MW fast breeder reactor, now being prepared for a new startup after the 1987 incident. So the importance of a safe operation of this investment is considerable for EdF, which is the designer, owner, industrial architect and operator. According to the French regulations, EdF is responsible for the safe operation of its power plants. A considerable human component is also at stake, as the safe operation of plants implies all the personnel to varying degrees. There are 15,000 such employees, all of whom have to be trained, competent and motivated. The operation of this system for 340 reactor-years has to-date resulted in no incident of any significant impact on the environment. Right from the start, safety in operation has always been an essential and clearly stated priority. Among other lessons the Three-Mile Island and Chernobyl accidents have reinforced the conviction that the human factors, the man-machine interface, and the safety culture were determining elements. With forty-eigh PWR plants in service, the problem is to maintain safe operation of a system now running at cruising speed, but also including some units (particularly the GCRs) that must be prepared for decommissioning. In addition EDF has to demonstrate the safe operations of CREYS MALVILLE, fast breeder reactor

  19. Radiation safety and protection on the nuclear power plants

    International Nuclear Information System (INIS)

    Nosovskij, A.V.; Bogorad, V.I.; Vasil'chenko, V.N.; Klyuchnikov, A.A.; Litvinskaya, T.V.; Slepchenko, A.Yu.

    2008-01-01

    The main issues of the radiation safety and protection provision on the nuclear power plants are considered in this monograph. The description of the basic sources of the radiation danger on NPPs, the principles, the methods and the means of the safety and radiation monitoring provision are shown. The special attention is paid to the issues of the ionizing radiation regulation

  20. Organizational analysis and safety for utilities with nuclear power plants: an organizational overview. Volume 1

    International Nuclear Information System (INIS)

    Osborn, R.N.; Olson, J.; Sommers, P.E.; McLaughlin, S.D.; Jackson, M.S.; Scott, W.G.; Connor, P.E.

    1983-08-01

    This two-volume report presents the results of initial research on the feasibility of applying organizational factors in nuclear power plant (NPP) safety assessment. A model is introduced for the purposes of organizing the literature review and showing key relationships among identified organizational factors and nuclear power plant safety. Volume I of this report contains an overview of the literature, a discussion of available safety indicators, and a series of recommendations for more systematically incorporating organizational analysis into investigations of nuclear power plant safety

  1. IAEA Leads Operational Safety Mission To Gravelines Nuclear Power Plant, France

    International Nuclear Information System (INIS)

    2012-01-01

    Full text: An IAEA-led international team of experts today began an in-depth operational safety review of the Gravelines Nuclear Power Plant in France. The review, conducted at the invitation of the French government, focuses on programmes and activities essential to the safe operation of the nuclear power plant. The three-week review will cover the areas of Management, Organization and Administration; Training and Qualification; Operations; Maintenance; Technical Support; Operating Experience; Radiation Protection; Chemistry; Emergency Planning and Preparedness; and Severe Accident Management. The conclusions of the review will be based on the IAEA Safety Standards and on well-established international good practices. The mission is not a regulatory inspection, a design review or a substitute for an exhaustive assessment of the plant's overall safety status. The team, led by the IAEA's Division of Nuclear Installation Safety, comprises experts from Bulgaria, China, Germany, Hungary, Japan, Romania, Slovakia, South Africa, Spain and Ukraine. The Gravelines mission is the 173rd conducted as part of the IAEA's Operational Safety Review Team programme, which began in 1982. France participates actively in the programme and the Gravelines mission is the 24th hosted by the country. General information about OSART missions can be found on the IAEA Website: OSART Missions. (IAEA)

  2. Software important to safety in nuclear power plants

    International Nuclear Information System (INIS)

    1994-01-01

    The report provides guidance on current practices, documenting their strengths and weaknesses for dealing with the important issues of software engineering that nuclear power plant system designers, software producers and regulators are facing. The focus of the report is on safety critical applications of general purpose processors controlled by custom developed software; however, it should also have application in safety related applications and for other types of computers. In addition to system designers, software producers and regulators, the intended readership of this report includes users of software based systems, who should be aware of the relevant issues in specifying and obtaining software for systems important to safety. Refs, 1 fig., tabs

  3. Safety in Swiss nuclear power plants

    International Nuclear Information System (INIS)

    Cederqvist, H.

    1992-01-01

    Safety-related facilities and equipment are continuously backfitted in Swiss nuclear power plants. In the Beznau-1 and -2 nuclear generating units, the measures taken under the heading of 'Backfitting of Emergency Systems' included provisions to enhance the protection against earthquakes, airplane crash, and fire; in addition, the emergency power system was upgraded. In Muehleberg, the stack exhaust air monitoring system was optimized. The containment pressure suppression system of the plant has been designed to withstand a hypothetical accident exceeding the design basis. The BKM-Crud computer simulation model simulates steps taken to reduce radiation exposure. The power of Swiss nuclear power stations will be raised by 4% to 15% within the 'Energy 2000' action program. (orig.) [de

  4. ENSI's technical view on the periodic safety review 2008 of the nuclear power plant Goesgen

    International Nuclear Information System (INIS)

    2012-08-01

    The owner of a license for a nuclear power plant operation in Switzerland has to undergo every 10 years a comprehensive safety check called 'periodic safety review' (PSR). The regulatory authority, the Swiss Federal Nuclear Safety Inspectorate (ENSI), reviews the documents supplied by the licensee. The Goesgen power plant (KKG) obtained its operation license and started operation in 1978. A first PSR was performed in the years 1996 to 1998 (PSR 1998) and reported. KKG delivered an analysis of the safety status, an evaluation of subsystems as well as test reports. The new PSR covers the period 1998 to 2007. The basis of the evaluation by ENSI is the new nuclear energy law in force since 1 February 2005. In comparison to PSR 1998, new aspects have to be considered like the description of the safety concept, including the technical safety classification of buildings, systems and components, or consideration of the protection objective 'limitation of the radiation exposure'. The PSR 2008 is focussed on the estimate of the nuclear safety of KKG. Basically, for the operation of a nuclear power plant, a sufficient protection has to be guaranteed against the release of radioactive materials to the environment as well as the irradiation of persons, during normal operation as well as in the case of accidents. The licensee of a nuclear power plant in operation must retrofit his plant according to the experience already gained and the state-of-the-art. The purpose of the PSR is to check the quality of the plant in the domain of safety. A probabilistic safety assessment (PSA) study must prove that the probability of damages to the reactor core is smaller than 10 -5 /year. In Switzerland the life time of a nuclear power plant is not limited by a fixed maximum time of operation. On the contrary, the limitation proceeds from safety criteria. Insufficiencies in the plant design are often recognized only through the evolution of the technique or some unexpected events. Ageing

  5. Integration of safety culture in transient analyses for nuclear power plants

    International Nuclear Information System (INIS)

    Stosic, Zoran V.; Stoll, Uwe

    2009-01-01

    In the nuclear field Safety Culture is the arrangement of attitudes and characteristics in individuals and organisations which determines first and foremost that nuclear power plant safety issues receive adequate attention due to their outstanding significance. It differs from general Corporate Culture via its concept of core hazards and the potentially large effects associated with the release of radioactivity. One can talk about positive and negative Safety Cultures. A positive Safety Culture assumes that the whole is more than the sum of the parts. The different parts interact to increase the overall effectiveness. In a negative Safety Culture the opposite is the case, with the action of some individuals restricted by the cynicism of others. Some examples of issues that contribute to a negative safety culture are: non-adherence to the established instructions and procedures, unclear definition of responsibilities, disinterest and inattentiveness, overestimation of own capabilities and arrogance, unclear rules, and mistrust between involved organisations. In addition to differentiation and importance of Safety Culture, necessary commitment levels, safety management framework, the paper discusses integration of Safety Culture in transient analyses of nuclear power plants. In this course the commitment to Safety Culture is defined as: a good Safety Culture depends on the continuous commitment and fulfilment of all involved organizations, persons and processes without any exception. (author)

  6. Organizational processes and nuclear power plant safety

    International Nuclear Information System (INIS)

    Landy, F.J.; Jacobs, R.R.; Mathieu, J.

    1991-01-01

    The paper describes the effects organizational factors have on the risk associated with the operation of nuclear power plants. The described research project addresses three methods for identifying the organizational factors that impact safety. The first method consists of an elaborate theory-based protocol dealing with decision making procedures, interdepartmental coordination of activities, and communications. The second, known as goals/means/measures protocol, deals with identifying safey related goals. The third method is known as behaviorally anchored rating scale development. The paper discusses the importance of the convergence of these three methods to identify organizational factors essential to reactor safety

  7. Manual on quality assurance for computer software related to the safety of nuclear power plants

    International Nuclear Information System (INIS)

    1988-01-01

    The objective of the Manual is to provide guidance in the assurance of quality of specification, design, maintenance and use of computer software related to items and activities important to safety (hereinafter referred to as safety related) in nuclear power plants. This guidance is consistent with, and supplements, the requirements and recommendations of Quality Assurance for Safety in Nuclear Power Plants: A Code of Practice, 50-C-QA, and related Safety Guides on quality assurance for nuclear power plants. Annex A identifies the IAEA documents referenced in the Manual. The Manual is intended to be of use to all those who, in any way, are involved with software for safety related applications for nuclear power plants, including auditors who may be called upon to audit management systems and product software. Figs

  8. A cross-cultural study of organizational factors on safety: Japanese vs. Taiwanese oil refinery plants.

    Science.gov (United States)

    Hsu, Shang Hwa; Lee, Chun-Chia; Wu, Muh-Cherng; Takano, Kenichi

    2008-01-01

    This study attempts to identify idiosyncrasies of organizational factors on safety and their influence mechanisms in Taiwan and Japan. Data were collected from employees of Taiwanese and Japanese oil refinery plants. Results show that organizational factors on safety differ in the two countries. Organizational characteristics in Taiwanese plants are highlighted as: higher level of management commitment to safety, harmonious interpersonal relationship, more emphasis on safety activities, higher devotion to supervision, and higher safety self-efficacy, as well as high quality of safety performance. Organizational characteristics in Japanese plants are highlighted as: higher level of employee empowerment and attitude towards continuous improvement, more emphasis on systematic safety management approach, efficient reporting system and teamwork, and high quality of safety performance. The casual relationships between organizational factors and workers' safety performance were investigated using structural equation modeling (SEM). Results indicate that the influence mechanisms of organizational factors in Taiwan and Japan are different. These findings provide insights into areas of safety improvement in emerging countries and developed countries respectively.

  9. Kozloduy nuclear power plant. Units 1-4. Status of safety improvements. Rev. 2

    International Nuclear Information System (INIS)

    1999-01-01

    This paper presents the results of the safety improvements activities carried out by the Kozloduy Nuclear Power Plant (KNPP) within the period 1990-1998. The steam supply system of this units is based of the reactor WWER-440/ B-230, which is a PWR of russian design developed according to the safety standards in force in USSR in late sixties. Up to now 10 reactor units of this type are in operation in four NPPs. Despite of efforts of the different plants to implement safety improvements measures during first 10-15 years of operation of this type of reactor its major safety problems were not eliminated and were a subject of international concern. The systematic evaluation of the deficiencies of the original design of this type of reactors have been initiated by IAEA in the beginning of 1990 and brought to developing a comprehensive list of safety problems which required urgent implementation of safety measures in all plants. To solve this problems in 1991 KNPP initiated implementation of so called 'short term' safety improvement program, developed with the help of WANO under agreement with Bulgarian Nuclear Safety Authority (BNSA) and consortium RISKAUDIT. The program was based on a stage approach and was foreseen to be implemented by tree stages in very tight time schedule in order to achieve significant and rapid improvements of the level of safety in operation of the units. The Short term program was implemented between from 1991 to 1997 owing to strong safety commitment of NEC and KNPP staff as well as broad international cooperation and financial support. Important part of resources were supplied under PHARE program of CEC, EBRD grant agreement and EDF support. In parallel a special assessment process started in 1995 in order to evaluate the level of safety, achieved by Short Term Program, according to current safety standards and to define the measures, which should be implemented by the Utility to complete the process of improving the safety in future

  10. Integrated approach to knowledge acquisition and safety management of complex plants with emphasis on human factors

    International Nuclear Information System (INIS)

    Kosmowski, K.T.

    1998-01-01

    In this paper an integrated approach to the knowledge acquisition and safety management of complex industrial plants is proposed and outlined. The plant is considered within a man-technology-environment (MTE) system. The knowledge acquisition is aimed at the consequent reliability evaluation of human factor and probabilistic modeling of the plant. Properly structured initial knowledge is updated in life-time of the plant. The data and knowledge concerning the topology of safety related systems and their functions are created in a graphical CAD system and are object oriented. Safety oriented monitoring of the plant includes abnormal situations due to external and internal disturbances, failures of hard/software components and failures of human factor. The operation and safety related evidence is accumulated in special data bases. Data/knowledge bases are designed in such a way to support effectively the reliability and safety management of the plant. (author)

  11. Safety aspects of LWR fuel reprocessing and mixed oxide fuel fabrication plants

    International Nuclear Information System (INIS)

    Fischer, M.; Leichsenring, C.H.; Herrmann, G.W.; Schueller, W.; Hagenberg, W.; Stoll, W.

    1977-01-01

    The paper is focused on the safety and the control of the consequences of credible accidents in LWR fuel reprocessing plants and in mixed oxide fuel fabrication plants. Each of these plants serve for many power reactor (about 50.000 Mwel) thus the contribution to the overall risk of nuclear energy is correspondingly low. Because of basic functional differences between reprocessing plants, fuel fabrication plants and nuclear power reactors, the structure and safety systems of these plants are different in many respects. The most important differences that influence safety systems are: (1) Both fuel reprocessing and fabrication plants do not have the high system pressure that is associated with power reactors. (2) A considerable amount of the radioactivity of the fuel, which is in the form of short-lived radionuclides has decayed. Therefore, fuel reprocessing plants and mixed oxide fuel fabrication plants are designed with multiple confinement barriers for control of radioactive materials, but do not require the high-pressure containment systems that are used in LWR plants. The consequences of accidents which may lead to the dispersion of radioactive materials such as chemical explosions, nuclear excursions, fires and failure of cooling systems are considered. A reasonable high reliability of the multiple confinement approach can be assured by design. In fuel reprocessing plants, forced cooling is necessary only in systems where fission products are accumulated. However, the control of radioactive materials can be maintained during normal operation and during the above mentioned accidents, if the dissolver off-gas and vessel off-gas treatment systems provide for effective removal of radioactive iodine, radioactive particulates, nitrogen oxides, tritium and krypton 85. In addition, the following incidents in the dissolver off-gas system itself must be controlled: failures of iodine filters, hydrogen explosion in O 2 - and NOsub(x)-reduction component, decomposition of

  12. The reliability of nuclear power plant safety systems

    International Nuclear Information System (INIS)

    Susnik, J.

    1978-01-01

    A criterion was established concerning the protection that nuclear power plant (NPP) safety systems should afford. An estimate of the necessary or adequate reliability of the total complex of safety systems was derived. The acceptable unreliability of auxiliary safety systems is given, provided the reliability built into the specific NPP safety systems (ECCS, Containment) is to be fully utilized. A criterion for the acceptable unreliability of safety (sub)systems which occur in minimum cut sets having three or more components of the analysed fault tree was proposed. A set of input MTBF or MTTF values which fulfil all the set criteria and attain the appropriate overall reliability was derived. The sensitivity of results to input reliability data values was estimated. Numerical reliability evaluations were evaluated by the programs POTI, KOMBI and particularly URSULA, the last being based on Vesely's kinetic fault tree theory. (author)

  13. Transportable nuclear power plant TEC-M with two reactor plants of improved safety

    International Nuclear Information System (INIS)

    Ogloblin, B.G.; Sazonov, A.G.; Svishchev, A.M.; Gromov, B.F.; Zelensky, V.N.; Komkova, O.I.; Sidorov, V.I.; Tolstopyatov, V.P.; Toshinsky, G.I.

    1993-01-01

    Liquid metals are the best to meet the requirements of inherently safety nuclear power plants among the coolants used. A great experience has been gained in lead coolant power plant development and operation as applied to transportable power set-ups. Low chemical activity of this coolant with respect to air-water interaction is a determining factor for this coolant. The transportable nuclear power plant is described. It is intended to generate electric power for populated areas placed a long distance from the main electric power supply sources where it is difficult or not economical to deliver the conventional types of fuel. There are several remote areas in Siberia, Kamchatka in need of this type of power plant

  14. Safety evaluation of the nuclear power plant at Cattenom

    International Nuclear Information System (INIS)

    Anon.

    1987-01-01

    This is a systematic compilation of the material which was dealt with at the level of the German-French Commission (on questions of the safety of nuclear installations) in this discussions about the nuclear power plant at Cattenom. As a supplement to the report published already in 1982, the Commission has officially released its deliberation results on the subjects constructive safety measures, radiological effects, and precautions in case of an emergency. The allegations according to which the installation is wanting in safety are countered by the joint statement of the chairmen of GPR (Permanent Group on Reactors) and RSK (German Commission on Reactor Safety) of August 29, 1986. (HSCH) [de

  15. The principal approaches to the problem of nuclear power plant safety in the USSR

    International Nuclear Information System (INIS)

    Sidorenko, V.A.; Kovalevich, O.M.; Kramerov, A.Ya.; Bagdasarov, Yu.E.

    1977-01-01

    The paper sets forth methods of ensuring the safety of nuclear power plants in the USSR on the basis of the scientific and engineering experience gained during the design, construction and operation of such plants, and describes the complex of technical and organizational problems whose solution determines the actual safety of nuclear power plants in the USSR. High-quality nuclear power plant equipment and components and their constant checking during the whole life of the plant are the prerequisites for preventing failures and accidents. The pattern of protective measures is discussed on the basis of possible failures and 'safe limits' for failures. The potentialities of the quantitative probabilistic method are analysed together with the need for a deterministic approach. The relationship of the maximum design accident with the protection and localization systems is considered in the case of nuclear power plants of different generations. The authors deal with the questions of State regulation of power plant safety on the basis of the adopted organizational structure and the system of standards. In conclusion, they briefly consider the application of the safety approach here described to power plants using water-water reactors, high-power boiling-water reactors and fast reactors in accordance with their place and role in the nuclear power development programme of the USSR. (author)

  16. Summary report on safety objectives in nuclear power plants

    International Nuclear Information System (INIS)

    1989-01-01

    The special Task Force on Safety Objectives of the Commission of the European Communities (CEC) Working Group on the Safety of Light Water Reactors reported in May 1983 on its review of existing overall safety objectives in nuclear power plants. Since then much relevant worlwide activity has taken place. This report reviews those activities that have taken place since 1983 in European Community Member States, including more recent Members, as well as in Sweden and Finland. The report confines itself to issues related to probabilistic safety objectives, and concludes that significant progress has been made in many areas. Mutual understanding of safety objectives is leading to a convergence of views and approaches, but it is noted that much work remains to be completed

  17. Safety goals for nuclear power plants: a discussion paper

    International Nuclear Information System (INIS)

    1982-02-01

    This report includes a proposed policy statement on safety goals for nuclear power plants published by the Commission for public comment and a supporting discussion paper. Proposed qualitative goals and associated numerical guidelines for nuclear power-plant accident risks are presented. The significance of the goals and guidelines, their bases and rationale, and their proposed mode of implementation are discussed

  18. Basic safety principles for nuclear power plants

    International Nuclear Information System (INIS)

    1988-01-01

    Nuclear power plant safety requires a continuing quest for excellence. All individuals concerned should constantly be alert to opportunities to reduce risks to the lowest practicable level. The quest, however, is most likely to be fruitful if it is based on an understanding of the underlying objectives and principles of nuclear safety, and the way in which its aspects are interrelated. This report is an attempt to provide a logical framework for such an understanding. The proposed objectives and principles of nuclear safety are interconnected and must be taken as a whole; they do not constitute a menu from which selection can be made. The report takes account of current issues and developments. It includes the concept of safety objectives and the use of probabilistic safety assessment. Reliability targets for safety systems are discussed. The concept of a 'safety culture' is crucial. Attention has been paid to the need for planning for accident management. The report contains objectives and principles. The objectives state what is to be achieved; the principles state how to achieve it. In each case, the basic principle is stated as briefly as possible. The accompanying discussion comments on the reasons for the principle and its importance, as well as exceptions, the extent of coverage and any necessary clarification. The discussion is as important as the principle it augments. 4 figs

  19. IAEA Operational Safety Team Review Bohunice Nuclear Power Plant, Slovak Republic

    International Nuclear Information System (INIS)

    2010-01-01

    Full text: An international team of nuclear installation safety experts, led by the International Atomic Energy Agency (IAEA), has reviewed Slovakia's Bohunice Nuclear Power Plant (BNPP) for its safety practices and has noted a series of good practices as well as recommendations to reinforce them. The IAEA assembled an international team of experts at the request of the Government of Slovak Republic to conduct an Operational Safety Review (OSART) of Bohunice NPP. Under the leadership of the IAEA's Division of Nuclear Installation Safety, the OSART team performed an in-depth operational safety review from 1 to 18 November 2010. The team was made up of experts from Belgium, Canada, China, the Czech Republic, France, Sweden, the United Kingdom and the IAEA. An OSART mission is designed as a review of programmes and activities essential to operational safety. It is not a regulatory inspection, nor is it a design review or a substitute for an exhaustive assessment of the plant's overall safety status. The team at BNPP conducted an in-depth review of the aspects essential to the safe operation of the NPP, which largely is under the control of the site management. The conclusions of the review are based on the IAEA's Safety Standards and proven good international practices. The review covered the areas of Management, Organization and Administration; Operations; Maintenance; Technical Support; Operating Experience; Radiation Protection; Chemistry and Emergency Planning and Preparedness. Long Term Operation assessment has been requested by the plant in addition to the standard OSART program. The OSART team has identified good plant practices which will be shared with the rest of the nuclear industry for consideration of their application. Examples include: BNPP has implemented a comprehensive set of technical and organizational measures which have significantly reduced the production of liquid radioactive waste; BNPP has developed an automatic transfer of dosimetry data

  20. Safety Second: the NRC and America's nuclear power plants

    International Nuclear Information System (INIS)

    Adato, M.; MacKenzie, J.; Pollard, R.; Weiss, E.

    1987-01-01

    In 1975, Congress created the Nuclear Regulatory Commission (NRC). Its primary responsibility was to be the regulation of the nuclear power industry in order to maintain public health and safety. On March 28, 1979, in the worst commercial nuclear accident in US history, the plant at Three Mile Island began to leak radioactive material. How was Three Mile Island possible? Where was the NRC? This analysis by the Union of Concerned Scientists (UCS) of the NRC's first decade, points specifically to the factors that contributed to the accident at Three Mile Island. The NRC, created as a watchdog of the nuclear power industry, suffers from problems of mindset, says the UCS. The commission's problems are political, not technical; it repeatedly ranks special interests above the interest of public safety. This book critiques the NRC's performance in four specific areas. It charges that the agency has avoided tackling the most pervasive safety issues; has limited public participation in decision making and power plant licensing; has failed to enforce safety standards or conduct adequate regulation investigations; and, finally, has maintained a fraternal relationship with the industry it was created to regulate, serving as its advocate rather than it adversary. The final chapter offers recommendations for agency improvement that must be met if the NRC is to fulfill its responsibility for safety first

  1. Assessing nuclear power plant safety and recovery from earthquakes using a system-of-systems approach

    International Nuclear Information System (INIS)

    Ferrario, E.; Zio, E.

    2014-01-01

    We adopt a โ€˜system-of-systemsโ€™ framework of analysis, previously presented by the authors, to include the interdependent infrastructures which support a critical plant in the study of its safety with respect to the occurrence of an earthquake. We extend the framework to consider the recovery of the system of systems in which the plant is embedded. As a test system, we consider the impacts produced on a nuclear power plant (the critical plant) embedded in the connected power and water distribution, and transportation networks which support its operation. The Seismic Probabilistic Risk Assessment of such system of systems is carried out by Hierarchical modeling and Monte Carlo simulation. First, we perform a top-down analysis through a hierarchical model to identify the elements that at each level have most influence in restoring safety, adopting the criticality importance measure as a quantitative indicator. Then, we evaluate by Monte Carlo simulation the probability that the nuclear power plant enters in an unsafe state and the time needed to recover its safety. The results obtained allow the identification of those elements most critical for the safety and recovery of the nuclear power plant; this is relevant for determining improvements of their structural/functional responses and supporting the decision-making process on safety critical-issues. On the test system considered, under the given assumptions, the components of the external and internal water systems (i.e., pumps and pool) turn out to be the most critical for the safety and recovery of the plant. - Highlights: โ€ข We adopt a system-of-system framework to analyze the safety of a critical plant exposed to risk from external events, considering also the interdependent infrastructures that support the plant. โ€ข We develop a hierarchical modeling framework to represent the system of systems, accounting also for its recovery. โ€ข Monte Carlo simulation is used for the quantitative evaluation of the

  2. Safety and Radiation Protection at Swedish Nuclear Power Plants 2007

    Energy Technology Data Exchange (ETDEWEB)

    2008-07-01

    The safety level of the plants is maintained at an acceptable level. SKI has in its regulatory supervision not found any known deficiencies in the barriers which could result in release of radioactive substances in excess of the permitted levels. SKI considers that improvements have been implemented during the year in the management, control and following up of safety work at the plants. In some cases, SKI has imposed requirements that improvements be made. Extensive measures are under way at the nuclear power plants to comply with the safety requirements in SKI's regulations, SKIFS 2004:2 concerning the design and construction of nuclear power reactors, and the stricter requirements regarding physical protection. Concurrently preparations are underway at eight of the ten units for thermal power increases. At the Forsmark plant considerable efforts have been during the year to correct the deficiencies in the safety culture and quality assurance system that became apparent in 2006. A programme to improve the execution of activities has been established in accordance with SKI's decision. SKI considers that the plant has developed in a positive direction but that there are further possibilities for improvement with regard to internal control. This is amongst other things concerns the areas internal auditing, independent safety review function, and working methods. SKI has had special supervision of the plant since 28 September, 2006. At the Oskarshamn plant work has been carried out to improve the organisation and routines in several areas. The plant has established routines which provide the basis to ensure that decisions are taken in a stringent manner. The quality assurance system has a clearer structure and there is a better defined division of work. Some measures remain to be dealt with in 2008. The Ringhals plant has also worked with attitudes to routines and internal control. SKI considers that the measures have good prerequisites to provide a

  3. Final report of the programme on the safety of WWER and RBMK nuclear power plants. A publication of the extrabudgetary programme on the safety of WWER and RBMK nuclear power plants

    International Nuclear Information System (INIS)

    1999-05-01

    The review of the extrabudgetary programme on the safety of WWER and RBMK nuclear power plants focuses on the wide scope of the activities aimed at identifying safety deficiencies, ranking their importance on the results of safety improvement programmes and on areas where future work is necessary. The information in the report reflects to a large extent, the situation as it stood when individual IAEA tasks actually took place. It deals with the IAEA activities and it discusses selected safety issues and safety review results as they apply to each reactor type. The results, recommendations and conclusions resulting from the IAEA Programme are intended to assist national decision makers who have the sole responsibilities for the regulation and safety operation of their nuclear power plants

  4. Safety Evaluation Approach with Security Controls for Safety I and C Systems on Nuclear Power Plants

    International Nuclear Information System (INIS)

    Kim, D. H.; Jeong, S. Y.; Kim, Y. M.; Park, H. S.; Lee, M. S.; Kim, T. H.

    2016-01-01

    This paper addresses concepts of safety and security and relations between them for assessing effects of security features in safety systems. Also, evaluation approach for avoiding confliction with safety requirements and cyber security features which may be adopted in safety-related digital I and C system will be described. In this paper, safety-security life cycle model based confliction avoidance method was proposed to evaluate the effects when the cyber security control features are implemented in the safety I and C system. Also, safety effect evaluation results using the proposed evaluation method were described. In case of technical security controls, many of them are expected to conflict with safety requirements, otherwise operational and managerial controls are not relatively. Safety measures and cyber security measures for nuclear power plants should be implemented not to conflict with one another. Where safety function and security features are both required within the systems, and also where security features are implemented within safety systems, they should be justified

  5. Safety Evaluation Approach with Security Controls for Safety I and C Systems on Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Kim, D. H.; Jeong, S. Y.; Kim, Y. M.; Park, H. S. [KINS, Daejeon (Korea, Republic of); Lee, M. S.; Kim, T. H. [Formal Works Inc., Seoul (Korea, Republic of)

    2016-05-15

    This paper addresses concepts of safety and security and relations between them for assessing effects of security features in safety systems. Also, evaluation approach for avoiding confliction with safety requirements and cyber security features which may be adopted in safety-related digital I and C system will be described. In this paper, safety-security life cycle model based confliction avoidance method was proposed to evaluate the effects when the cyber security control features are implemented in the safety I and C system. Also, safety effect evaluation results using the proposed evaluation method were described. In case of technical security controls, many of them are expected to conflict with safety requirements, otherwise operational and managerial controls are not relatively. Safety measures and cyber security measures for nuclear power plants should be implemented not to conflict with one another. Where safety function and security features are both required within the systems, and also where security features are implemented within safety systems, they should be justified.

  6. Safety assessment in plant layout design using indexing approach: Implementing inherent safety perspective

    International Nuclear Information System (INIS)

    Tugnoli, Alessandro; Khan, Faisal; Amyotte, Paul; Cozzani, Valerio

    2008-01-01

    Layout planning plays a key role in the inherent safety performance of process plants since this design feature controls the possibility of accidental chain-events and the magnitude of possible consequences. A lack of suitable methods to promote the effective implementation of inherent safety in layout design calls for the development of new techniques and methods. In the present paper, a safety assessment approach suitable for layout design in the critical early phase is proposed. The concept of inherent safety is implemented within this safety assessment; the approach is based on an integrated assessment of inherent safety guideword applicability within the constraints typically present in layout design. Application of these guidewords is evaluated along with unit hazards and control devices to quantitatively map the safety performance of different layout options. Moreover, the economic aspects related to safety and inherent safety are evaluated by the method. Specific sub-indices are developed within the integrated safety assessment system to analyze and quantify the hazard related to domino effects. The proposed approach is quick in application, auditable and shares a common framework applicable in other phases of the design lifecycle (e.g. process design). The present work is divided in two parts: Part 1 (current paper) presents the application of inherent safety guidelines in layout design and the index method for safety assessment; Part 2 (accompanying paper) describes the domino hazard sub-index and demonstrates the proposed approach with a case study, thus evidencing the introduction of inherent safety features in layout design

  7. 77 FR 50720 - Test Documentation for Digital Computer Software Used in Safety Systems of Nuclear Power Plants

    Science.gov (United States)

    2012-08-22

    ... Used in Safety Systems of Nuclear Power Plants AGENCY: Nuclear Regulatory Commission. ACTION: Draft... Computer Software used in Safety Systems of Nuclear Power Plants.'' The DG-1207 is proposed Revision 1 of... for Digital Computer Software Used in Safety Systems of Nuclear Power Plants'' is temporarily...

  8. Radiological safety of nuclear power plants in India

    International Nuclear Information System (INIS)

    Sathish, A.V.

    2015-01-01

    Safety in nuclear power plants (NPPs) is often less understood and more talked about, thus the author wanted to share the facts to clear the myths. Safety is accorded overriding priority in all the activities. All nuclear facilities are sited, designed, constructed, commissioned and operated in accordance with strict quality and safety standards. Principles of defence in depth, redundancy and diversity are followed in the design of all nuclear facilities and their systems/components. PPs in India are not only safe but are also well regulated, have proper radiological protection of workers and the public, regular surveillance, approved standard operating and maintenance procedures, a well-defined waste management methodology, periodically rehearsed emergency preparedness and disaster management plans. The regulatory framework in the country is robust, with the independent Atomic Energy Regulatory Board (AERB) having powers to frame the policies, laying down safety standards, monitoring and enforcing all the safety provisions. As a result, India's safety record has been excellent in over 400 reactor years of operation of power reactors

  9. Safety related requirements on future nuclear power plants

    International Nuclear Information System (INIS)

    Niehaus, F.

    1991-01-01

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

  10. Forum for fire protection and safety in power plants[Norway

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2005-07-01

    The conference contains 16 presentations on topics in the fields of fire protection and safety in plants in Western Norway, reorganization and reconstruction of power systems and plants in Norway, various aspects of risk and vulnerability analysis, technological aspects of plant management and construction and problems and risks with particularly transformers. Some views on challenges of the fire departments and the new Norwegian regulations for electrical power supply systems are included. One presentation deals with challenges for Icelandic power production plants.

  11. Aging of safety class 1E transformers in safety systems of nuclear power plants

    International Nuclear Information System (INIS)

    Roberts, E.W.; Edson, J.L.; Udy, A.C.

    1996-02-01

    This report discusses aging effects on safety-related power transformers in nuclear power plants. It also evaluates maintenance, testing, and monitoring practices with respect to their effectiveness in detecting and mitigating the effects of aging. The study follows the US Nuclear Regulatory Commission's (NRC's) Nuclear Plant-Aging Research approach. It investigates the materials used in transformer construction, identifies stressors and aging mechanisms, presents operating and testing experience with aging effects, analyzes transformer failure events reported in various databases, and evaluates maintenance practices. Databases maintained by the nuclear industry were analyzed to evaluate the effects of aging on the operation of nuclear power plants

  12. Specific safety aspects of the water-steam cycle important to nuclear power plant project

    International Nuclear Information System (INIS)

    Lobo, C.G.

    1986-01-01

    The water-steam cycle in a nuclear power plant is similar to that used in conventional power plants. Some systems and components are required for the safe nuclear power plant operation and therefore are designed according to the safety criteria, rules and regulations applied in nuclear installations. The aim of this report is to present the safety characteristics of the water-steam cycle of a nuclear power plant with pressurized water reactor, as applied for the design of the nuclear power plants Angra 2 and Angra 3. (Author) [pt

  13. Definitive closure of nuclear power plants. Aspects concerning physical safety

    International Nuclear Information System (INIS)

    Rodriguez, C.; Puntarulo, L.; Canibano, J.

    1988-01-01

    This paper analyzes the various safety requirements that must be fulfilled by nuclear power plants for their operation without restrictions, such as safeguards, nuclear safety and physical protection. Physical protection, the subject most extensively dealt by the authors, is defined as safety measures aimed at providing protection against deliberate hostile deeds, such as robberies or non-authorized transport of radioactive materials or sabotage in nuclear facilities, performed either by individuals or by groups of individuals. (Author)

  14. Technical and institutional safety features of nuclear power plants in Brazil

    International Nuclear Information System (INIS)

    Rosa, L.P.

    1986-01-01

    This work reports technical, political and institutional safety features of nuclear power plants in Brazil. It is mainly concerned with reactor accidents and personnel safety. The three mile Island and Chernobyl accidents are also discussed and taken as examples. (A.C.A.S.)

  15. Relation of management, supervision, and personnel practices to nuclear power plant safety

    International Nuclear Information System (INIS)

    Layton, W.L.; Turnage, J.J.

    1980-01-01

    The knowledge base of industrial/organization psychology suggests three major areas of research with important implications for nuclear power plant safety. These areas are: Management and Supervision: Personnel Selection, Training and Placement; and Organizational Climate. Evidence drawn from several Three Mile Island investigations confirms that organizational structure of plants and supervisory practices, the selection and training of personnel, and organizational climate are important factors. Difficulties in decision making and coordination of personnel are pinpointed. Deficiencies in training are highlighted and the climate of working atmosphere is discussed. These matters are related to nuclear power plant safety. Future research directions are presented

  16. Management of operational safety in nuclear power plants. INSAG-13. A report by the International Nuclear Safety Advisory Group

    International Nuclear Information System (INIS)

    1999-01-01

    The International Atomic Energy Agency's activities relating to nuclear safety are based upon a number of premises. First and foremost, each Member State bears full responsibility for the safety of its nuclear facilities. States can be advised, but they cannot be relieved of this responsibility. Secondly, much can be gained by exchanging experience; lessons learned can prevent accidents. Finally, the image of nuclear safety is international; a serious accident anywhere affects the public's view of nuclear power everywhere. With the intention of strengthening its contribution to ensuring the safety of nuclear power plants, the IAEA established the International Nuclear Safety Advisory Group (INSAG), whose duties include serving as a forum for the exchange of information on nuclear safety issues of international significance and formulating, where possible, commonly shared safety principles. Engineering issues have received close attention from the nuclear community over many years. However, it is only in the last decade or so that organizational and cultural issues have been identified as vital to achieving safe operation. INSAG's publication No. 4 has been widely recognized as a milestone in advancing thinking about safety culture in the nuclear community and more widely. The present report deals with the framework for safety management that is necessary in organizations in order to promote safety culture. It deals with the general principles underlying the management of operational safety in a systematic way and provides guidance on good practices. It also draws on the results of audits and reviews to highlight how shortfalls in safety management have led to incidents at nuclear power plants. In addition, several specific issues are raised which are particularly topical in view of organizational changes that are taking place in the nuclear industry in various countries. Advice is given on how safety can be managed during organizational change, how safety

  17. The site of a nuclear power plant and environmental safety; Ydinvoimalaitoksen sijaintipaikka ja ympaeristoen turvallisuus

    Energy Technology Data Exchange (ETDEWEB)

    Rossi, J. [VTT Energy, Espoo (Finland)

    2001-11-01

    The purpose of this report is to give the reader a general view of the things associated with the site of a nuclear power plant. In this context the effect of a nuclear power plant and site on environmental safety is considered. Planning, construction and operating a nuclear power plant require several judgements and licenses based on different laws. The location of the planned nuclear facility project and environmental conditions contribute in great detail to the compliance arguments of permits. At first the environmental impacts of the siting project and its alternatives shall be investigated in the Environmental Impact Assessment procedure. Then the decision in principle according to the Nuclear Energy Act can be applied from the Council of State, the decision shall further be confirmed by Parliament. When the decision in principle is considered the overall good of society shall be assessed by means of considering i.a. site alternatives and safety. The safety related basic principle is that operation of a nuclear power plant may not cause danger to the environment, public or property. After the affirmative principle approval the construction license and later on the operation license can be applied from the Council of State, these licenses need to be supported i.a. by building and environmental licenses of separate authorities. Also some international contracts concern realisation of a nuclear power plant siting. The nuclear power plant site shall be suitable for the needs of the electricity production and the transmission system and it shall be technically appropriate for building and operation of a power plant. The site shall be safe enough on the other hand from the view of external events threatening the power plant - although one can be partly prepared for these things in the design of the plant - and on the other hand from the point of public safety. Requirements for the safety of the site are directed in the decision of the Council of State's general

  18. Methodology for safety classification of PWR type nuclear power plants items

    International Nuclear Information System (INIS)

    Oliveira, Patricia Pagetti de

    1995-01-01

    This paper contains the criteria and methodology which define a classification system of structures, systems and components in safety classes according to their importance to nuclear safety. The use of this classification system will provide a set of basic safety requirements associated with each safety class specified. These requirements, when available and applicable, shall be utilized in the design, fabrication and installation of structures, systems and components of Pressurized Water Reactor Nuclear Power Plants. (author). 13 refs, 1 tab

  19. Application of meteorology to safety at nuclear plants

    International Nuclear Information System (INIS)

    1968-01-01

    This report was prepared on behalf of the International Atomic Energy Agency by an international panel of experts who met at the Agency's headquarters from 10 to 14 April 1967. The application of meteorology to safety at nuclear plants is discussed in connection with site selection, design and construction, operation, and emergency planning and action. The final chapter considers the training to be given to operators and health and safety personnel on meteorology problems. The appendix gives a simple method for computing air concentration values at ground level. An extensive bibliography is also included.

  20. The importance of the reliability study for the safety operation of chemical plants. Application in heavy water plants

    International Nuclear Information System (INIS)

    Dumitrescu, Maria; Lazar, Roxana Elena; Preda, Irina Aida; Stefanescu, Ioan

    1999-01-01

    Heavy water production in Romania is based on H 2 O-H 2 S isotopic exchange process followed by vacuum isotopic distillation. The heavy water plant are complex chemical systems, characterized by an ensemble of static and dynamic equipment, AMC components, enclosures. Such equipment must have a high degree of reliability, a maximum safety in technological operation and a high availability index. Safety, reliable and economical operation heavy water plants need to maintain the systems and the components at adequate levels of reliability. The paper is a synthesis of the qualitative and quantitative assessment reliability studies for heavy water plants. The operation analysis on subsystems, each subsystems being a well-defined unit, is required by the plant complexity. For each component the reliability indicators were estimated by parametric and non-parametric methods based on the plant operation data. Also, the reliability qualitative and quantitative assessment was done using the fault tree technique. For the dual temperature isotopic exchange plants the results indicate an increase of the MTBF after the first years of operation, illustrating both the operation experience increasing and maintenance improvement. Also a high degree of availability was illustrated by the reliability studies of the vacuum distillation plant. The establishment of the reliability characteristics for heavy water plant represents an important step, a guide for highlighting the elements and process liable to failure being at the same time a planning modality to correlate the control times with the maintenance operations. This is the way to minimise maintenance, control and costs. The main purpose of the reliability study was the safety increase of the plant operation and the support for decision making. (authors)

  1. Research on integrated managing system based on CIMS for nuclear power plant safety

    International Nuclear Information System (INIS)

    Zhou Gang

    2006-01-01

    In order to improve safety, economy and reliability of operation for nuclear power plant (NPP), a novel integrated managing method was proposed based on the ideas of computer and contemporary integrated manufacturing system (CIMS). The application of CIMS to nuclear power plant safety management was researched. In order to design an integrated managing system to meet the needs of NPP safety management, all work related to nuclear safety is divided into different category according to its characters. On basis of this work, general integrated managing system was designed at first. Then subsystems were designed and every subsystem implements a category of nuclear safety management work. All subsystems are independent relatively on the one hand and are interrelated on other hand by global information system. (authors)

  2. 33 CFR 165.115 - Safety and Security Zones; Pilgrim Nuclear Power Plant, Plymouth, Massachusetts.

    Science.gov (United States)

    2010-07-01

    ... 33 Navigation and Navigable Waters 2 2010-07-01 2010-07-01 false Safety and Security Zones; Pilgrim Nuclear Power Plant, Plymouth, Massachusetts. 165.115 Section 165.115 Navigation and Navigable... Coast Guard District ยง 165.115 Safety and Security Zones; Pilgrim Nuclear Power Plant, Plymouth...

  3. Improved safety of the system 80+TM standard plants design through increased diversity and redundancy of safety systems

    International Nuclear Information System (INIS)

    Matzie, Regis A.; Carpentino, Frederick L.; Robertson, James E.

    1996-01-01

    Safely systems in the System 80+ TM Standard Plant are designed with more redundancy, diversity and simplicity than earlier nuclear power plant designs. These gains were accomplished by an evolutionary process that preserved the desirable and proven features in currently operating nuclear plants, while improving reliability and defense-in-depth. The System 80+ safety systems are the primary contributors to a core damage frequency that is more than 100 times lower than 1980's vintage U. S. designs, including the predecessor System 80 R standard nuclear steam supply system (NSSS) design. The System 80+ design includes significant improvements to the safety injection system, emergency feedwater system, shutdown cooling system, containment spray system, reactor coolant gas vent system, and to their vital support systems. These improvements enhance performance for traditional design basis events and significantly reduce the probability of a severe accident. The System 80+ design also incorporates safety systems to mitigate a severe accident. The added systems include the rapid depressurization system, the in-containment refueling water storage tank, the cavity flooding system. These systems fully address the U. S. Nuclear Regulatory Commission's (US NRC) severe accident policy. The System 80+ safety systems are integrated with the System 80+ Nuclear Island (NI) design. The NI general arrangement provides quadrant separation of the safety systems for protection from fire and flooding, and large equipment pull spaces and lay down areas for maintenance. This paper will describe the System 80+ safety systems advanced design features, the improved accident prevention and mitigation capabilities, and startup, operating and maintenance benefits

  4. IAEA-led Operational Safety Team Reviews Dukovany Nuclear Power Plant, Czech Republic

    International Nuclear Information System (INIS)

    2011-01-01

    Full text: An international team of nuclear installation safety experts, led by the International Atomic Energy Agency (IAEA), has commended the Dukovany Nuclear Power Plant (NPP) in the Czech Republic for its safety practices and has also made a series of recommendations to reinforce them. The IAEA assembled an international team of experts at the request of the Government of the Czech Republic to conduct an Operational Safety Review (OSART) of Dukovany NPP. Under the leadership of the IAEA's Division of Nuclear Installation Safety in Vienna, the OSART team performed an in-depth operational safety review of the plant from 6 to 23 June 2011. The team was made up of experts from Armenia, Germany, Hungary, Romania, Slovenia, Sweden, the UK and the USA. An OSART mission is designed as a review of programmes and activities essential to operational safety. It is not a regulatory inspection, nor is it a design review or a substitute for an exhaustive assessment of the plant's overall safety status. The team at Dukovany conducted an in-depth review of the aspects essential to the safe operation of the NPP, which is largely under the control of the site management. The conclusions of the review are based on the IAEA's Safety Standards and proven good international practices. The review covered the areas of Management, Organization and Administration; Training and Qualification; Operations; Maintenance; Technical Support; Operating Experience; Radiation Protection; Chemistry; and Emergency Planning and Preparedness. The OSART team has identified good plant practices, which will be shared with the rest of the nuclear industry for consideration of their application. Examples include: The plant uses an integrated approach to recruit, select, psychologically assess and train new employees. This approach has resulted in consistently high success rates for licensed operator examinations and the identification of potential candidates for various plant departments; The performance

  5. Safety and Radiation Protection at Swedish Nuclear Power Plants 2007

    International Nuclear Information System (INIS)

    2008-01-01

    The safety level of the plants is maintained at an acceptable level. SKI has in its regulatory supervision not found any known deficiencies in the barriers which could result in release of radioactive substances in excess of the permitted levels. SKI considers that improvements have been implemented during the year in the management, control and following up of safety work at the plants. In some cases, however, SKI has imposed requirements that improvements be made. Extensive measures are under way at the nuclear power plants to comply with the safety requirements in SKI's regulations, SKIFS 2004:2 concerning the design and construction of nuclear power reactors, and the stricter requirements regarding physical protection. Concurrently preparations are underway at eight of the ten units for thermal power increases. At the Forsmark plant considerable efforts have been during the year to correct the deficiencies in the safety culture and quality assurance system that became apparent in 2006. A programme to improve the execution of activities has been established in accordance with SKI's decision. SKI considers that the plant has developed in a positive direction but that there are further possibilities for improvement with regard to internal control. This is amongst other things concerns the areas internal auditing, independent safety review function, and working methods. SKI has had special supervision of the plant since 28 September, 2006. At the Oskarshamn plant work has been carried out to improve the organisation and routines in several areas. The plant has established routines which provide the basis to ensure that decisions are taken in a stringent manner. The quality assurance system has a clearer structure and there is a better defined division of work. Some measures remain however to be dealt with in 2008. The Ringhals plant has also worked with attitudes to routines and internal control. SKI considers that the measures have good prerequisites to provide a

  6. Experience gained in enhancing operational safety at ComEd`s nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Elias, D [Commonwealth Edison Co. (United States)

    1997-09-01

    The following aspects of experience gained in enhancing operational safety at Comed`s nuclear power plants are discussed: nuclear safety policy; centralization/decentralization; typical nuclear operating organization; safety review boards; human performance enhancement; elements of effective nuclear oversight.

  7. The use of probabilistic safety assessment based maintenance indicators to increase the availability of safety related systems in nuclear power plants

    International Nuclear Information System (INIS)

    Kirchsteiger, C.

    1991-04-01

    This work describes the theoretical development of a Probabilistic Safety Assessment (PSA) based Performance Indicator (PI) model for a comprehensive Maintenance Efficiency Analysis (MEA) and its practical application to past operational history data of a certain Nuclear Power Plant. Plant specific equipment history and maintenance work order data have been collected and analysed using various advanced statistical procedures (nonparametric methods, multivariate analysis) in order to be able to estimate safety system related equipment and maintenance process trends. The main results of such a MEA case study are the trends in the (in)effectiveness of the performance of a selected safety system and its dominant maintenance related causes of its bad (good) equipment performance. Finally, the therefrom gained results are used to propose a new set of safety system based and maintenance related Performance Indicators, including suggestions for a corresponding plant specific maintenance data collection system. (author)

  8. Risk-based safety performance indicators for nuclear power plants

    International Nuclear Information System (INIS)

    Chakraborty, S.; Prohaska, G.; Flodin, Y.; Grint, G.; Habermacher, H.; Hallman, A.; Isasia, R.; Melendez, E.; Verduras, E.; Karsa, Z.; Khatib-Rahbar, M.; Koeberlein, K.; Schwaeger, C.; Matahri, N.; Moravcik, I.; Tkac, M.; Preston, J.

    2003-01-01

    In a Concerted Action (CA), sponsored by the European Commission within its 5th Framework Program, a consortium of eleven partners from eight countries has reviewed and evaluated the application of Safety Performance Indicators (SPIs), which - in combination with other tools - can be used to monitor and improve the safety of nuclear power plants. The project was aimed at identification of methods that can be used in a risk-informed regulatory system and environment, and to exploit PSA techniques for the development and use of meaningful additional/alternative SPIs. The CA included the review of existing indicator systems, and the collection of information on the experience from indicator systems by means of a specific questionnaire. One of the most important and challenging issues for nuclear plant owners and/or regulators is to recognize early signs of deterioration in safety performance, caused by influences from management, organization and safety culture (MOSC), before actual events and/or mishaps take place. Most of the existing SPIs as proposed by various organizations are considered as 'lagging' indicators, that is, they are expected to show an impact only when a downward trend has already started. Furthermore, most of the available indicators are at a relatively high level, such that they will not provide useful information on fundamental weaknesses causing the problem in the first place. Regulators' and utilities' views on the use of a Safety Performance Indicator System have also been a part of the development of the CA. (author)

  9. Basic safety principles of KLT-40C reactor plants

    International Nuclear Information System (INIS)

    Beliaev, V.; Polunichev, V.

    2000-01-01

    The KLT-40 NSSS has been developed for a floating power block of a nuclear heat and power station on the basis of ice-breaker-type NSSS (Nuclear Steam Supply System) with application of shipbuilding technologies. Basic reactor plant components are pressurised water reactor, once-through coil-type steam generator, primary coolant pump, emergency protection rod drive mechanisms of compensate group-electromechanical type. Basic RP components are incorporated in a compact steam generating block which is arranged within metal-water shielding tank's caissons. Domestic regulatory documents on safety were used for the NSSS design. IAEA recommendations were also taken into account. Implementation of basic safety principles adopted presently for nuclear power allowed application of the KLT-40C plant for a floating power unit of a nuclear co-generation station. (author)

  10. Safety improvement plant modifications at Forsmark 3, 1986-1995

    Energy Technology Data Exchange (ETDEWEB)

    Kjellander, M. [Kaernkraftsaekerhet och utbildning, Nykoeping (Sweden)

    1998-10-01

    All important plant modifications implemented in safety-related equipment or software at Forsmark 3 are compiled in this report. The report covers the period from the start of commercial operation in 1985 up to and including 1995. The plant modifications, which were carried out by different suppliers during the guarantee period, are not included in the report since they have not been administered by the Forsmark organisation. The report contains references to relevant modification notices and to files and file divider numbers. These data refer to the Safety Department central archives. The report is based on Forsmark 3 Technical Specifications (STF) which means that Chapter 3 is divided into the same sections as in the STF. Modifications, which cannot be directly attributed to any specific STF chapter, and major modifications are described separately

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

    International Nuclear Information System (INIS)

    1987-10-01

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

  12. IAEA effort on the evaluation and management of safety aspects of nuclear power plant ageing

    International Nuclear Information System (INIS)

    Pachner, J.; Yaremy, E.M.

    1991-01-01

    The questions attached to nuclear power plant (NPP) ageing and the need for monitoring and assessment of plant condition will grow in importance as more NPPs approach and pass the end of their nominal design lives. Ageing in nuclear plants must be effectively managed to ensure plant safety during their entire service life. This paper provides an overview of the IAEA programme and its results on the evaluation and management of safety aspects of NPP ageing. Under this programme, three generic guidance documents have been prepared on: data collection and record keeping; ageing management methodology; and the use of probabilistic safety assessment in plant life extension considerations. A two pronged strategy involving both technical and regulatory aspects has been adopted for the current and future work. (author)

  13. Application of the Safety Classification of Structures, Systems and Components in Nuclear Power Plants

    International Nuclear Information System (INIS)

    2016-04-01

    This publication describes how to complete tasks associated with every step of the classification methodology set out in IAEA Safety Standards Series No. SSG-30, Safety Classification of Structures, Systems and Components in Nuclear Power Plants. In particular, how to capture all the structures, systems and components (SSCs) of a nuclear power plant to be safety classified. Emphasis is placed on the SSCs that are necessary to limit radiological releases to the public and occupational doses to workers in operational conditions This publication provides information for organizations establishing a comprehensive safety classification of SSCs compliant with IAEA recommendations, and to support regulators in reviewing safety classification submitted by licensees

  14. Safety assessment for the passive system of the nuclear power plants (NPPs) using safety margin estimation

    International Nuclear Information System (INIS)

    Woo, Tae-Ho; Lee, Un-Chul

    2010-01-01

    The probabilistic safety assessment (PSA) for gas-cooled nuclear power plants has been investigated where the operational data are deficient, because there is not any commercial gas-cooled nuclear power plant. Therefore, it is necessary to use the statistical data for the basic event constructions. Several estimations for the safety margin are introduced for the quantification of the failure frequency in the basic event, which is made by the concept of the impact and affordability. Trend of probability of failure (TPF) and fuzzy converter (FC) are introduced using the safety margin, which shows the simplified and easy configurations for the event characteristics. The mass flow rate in the natural circulation is studied for the modeling. The potential energy in the gravity, the temperature and pressure in the heat conduction, and the heat transfer rate in the internal stored energy are also investigated. The values in the probability set are compared with those of the fuzzy set modeling. Non-linearity of the safety margin is expressed by the fuzziness of the membership function. This artificial intelligence analysis of the fuzzy set could enhance the reliability of the system comparing to the probabilistic analysis.

  15. Study on safety subsidiary objective of nuclear power plant in USA

    International Nuclear Information System (INIS)

    Chen Yan; Zhang Chunming; Fu Zhiwei; Song Wei; Li Chaojun; Wang Zhe; Zuo Jiaxu

    2013-01-01

    This paper reviewed the development of the quantitative safety objective and subsidiary objective in USA. The expressions of CDF and LERF were obtained according to NUREG-1150. The relationship between the subsidiary objective and the quantitative safety objective was derived. The method was compared with that used in NUREG-1860. The requirements of safety objective for the future nuclear power plant and the development of probabilistic safety analysis (PSA) technology in USA were studied and can be used as reference in China. (authors)

  16. The quality and safety of nuclear plants: the part played by the administrative authorities

    International Nuclear Information System (INIS)

    Queniart, Daniel

    1976-01-01

    After specifying the notions of 'safety' and 'quality', the terms and conditions governing the intervention of the public authorities in the matter of safety of nuclear plants are described: individual permits, the establishing and application of technical rules of a general character, surveillance of the plants. The criteria and regulations guiding the evaluation of safety and quality and, in conclusion, insisting on the necessity for permanent discussions among the various organizations concerned are presented [fr

  17. Regulatory supervision of safety indicators; experience with radiation safety indicators in Dukovany nuclear power plant performance

    International Nuclear Information System (INIS)

    Urbancik, L.; Kulich, V.

    2004-01-01

    The State Office for Nuclear Safety uses three sets of indicators describing the following aspects of a favourable nuclear power plant operation: smooth operation in normal circumstances, low risk to the population, and operation with a positive safety attitude. These are three safety-related areas for assessment. Each area has its own set of indicators. Overall operational safety performance indicators were identified for each attribute. From this point, a level of strategic indicators was developed, and finally, a set of specific indicators was set up. While neither the overall indicators nor the strategic indicators are directly measurable, the specific indicators are directly measurable and are targeted during inspection. (author)

  18. Study of system safety evaluation on LTO of national project. NISA safety research project on system safety of nuclear power plants

    International Nuclear Information System (INIS)

    Takizawa, Masayuki; Sekimura, Naoto; Miyano, Hiroshi; Aoyama, Katsunobu

    2012-01-01

    Japanese safety regulatory body, that is, Nuclear and Industrial Safety Agency (NISA) started a 5-year national safety research project as 'the first stage' from 2006 FY to 2010 FY whose objective is 'Improve the technical information basis in order to utilize knowledge as well as information related to ageing management and maintenance of NPPs. Fukushima disaster happened in March 2011, and the priority of research needs for ageing management dramatically changed in Japan. The second-stage national project started in October 2011 with the concept of 'system safety' of NNPs where not only ageing management on degradation phenomena of important components but also safety management on total plant systems are paid attention to. The second-stage project is so called 'Japanese Ageing Management Program for System Safety (JAMPSS)'. (author)

  19. Aging of safety class 1E transformers in safety systems of nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Roberts, E.W.; Edson, J.L.; Udy, A.C. [Lockheed Idaho Technologies Co., Idaho Falls, ID (United States)

    1996-02-01

    This report discusses aging effects on safety-related power transformers in nuclear power plants. It also evaluates maintenance, testing, and monitoring practices with respect to their effectiveness in detecting and mitigating the effects of aging. The study follows the US Nuclear Regulatory Commission`s (NRC`s) Nuclear Plant-Aging Research approach. It investigates the materials used in transformer construction, identifies stressors and aging mechanisms, presents operating and testing experience with aging effects, analyzes transformer failure events reported in various databases, and evaluates maintenance practices. Databases maintained by the nuclear industry were analyzed to evaluate the effects of aging on the operation of nuclear power plants.

  20. OSART mission highlights 1989-1990: Operational safety practices in nuclear power plants

    International Nuclear Information System (INIS)

    1992-12-01

    The IAEA Operational Safety Review Team (OSART) programme provides advice and assistance to Member States in enhancing the operational safety of nuclear power plants. OSART reviews are available to all countries with nuclear power plants in operation or approaching operation. Most of these countries have participated in the programme, by hosting one or more OSART missions or by making experts available to participate in missions. Careful design and high quality of construction are prerequisites for a safe nuclear power plant. However, a plant's safety depends ultimately on the ability and conscientiousness of the operating personnel and on their tools and work methods. OSART missions assess a facility's operational practices in comparison with those used successfully in other countries, and exchange, at the working level, ideas for promoting safety. Both the plants reviewed and the organizations providing experts have benefited from the programme. The observations of the OSART members are documented in technical notes which are then used as source material for the official OSART Report submitted to the government of the host country. The technical notes contain recommendations for improvements and descriptions of recommendable good practices. The same notes have been used to compile the present summary report which is intended for wide distribution to all organizations constructing, operating or regulating nuclear power plants. This report is the fourth in a series following IAEA-TECDOC-458, IAEA-TECDOC-497 and IAEA-TECDOC-570 and covers the period June 1989 to December 1990. Reference is also made to a summary report of Pre-OSART missions, which is in preparation. In addition, a report presenting OSART Good Practices has been published (IAEA-TECDOC-605)